Text

Request for Exemption from Nuhoms Certificate of Compliance No. 1004, Amendment No. 8
	ML061650157
Person / Time
Site:	Fort Calhoun, 07201004
Issue date:	06/09/2006
From:	Gambhir S; Omaha Public Power District
To:	; Document Control Desk, Office of Nuclear Material Safety and Safeguards
References
	LIC-06-0056, NUH-003
	Download: ML061650157 (127)
	v • d • e

Omaha Public Power District 444 South 16th Street Mall Omaha NE 68102-2247 June 9, 2006 LIC-06-0056 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk, Director, Spent Fuel Project Office, Office of Nuclear Material Safety and Safeguards, Washington, D.C. 20555

References:

1. Docket Nos. 50-285 and 72-054

2. "Standardized NUHOMSO Horizontal Modular Storage System for Irradiated Nuclear Fuel, Updated Final Safety Analysis Report," NUH-003

3. Certificate of Compliance No. 1004 for the Standardized NUHOMSO System, Amendment 8, effective December 5, 2005

SUBJECT:

Fort Calhoun Station-Request for Exemption from NUTIOMS Certificate of Compliance No. 1004, Amendment No. 8 Pursuant to the provisions of 10 CFR 72.7, "Specific exemptions," Omaha Public Power District (OPPD) requests an exemption from requirements ,specified in 10 CFR 72.2 12, "Conditions of general license issued under §72.2 10." The specificexemption would be from the requirements of 10 CFR 72.212(a)(2), 10 CFR 72.212(b)(2)(i)(A), 10 CFR 72,212(b)(7), and 10 CFR 72.214, all of which require the licensee to comply with the tenns and conditions of the NRC issued certificate of compliance (CoC). In connection with these requirements, OPPD also requests an exemption -- to the extent necessary -- from 10 CFR 72.48(c)(1)(B), which requires that design changes do not involve changes to the terms, conditions, or specifications of the CoC. This includes exemption from Technical Specifications 1.2.1, 1.2.11, and 1.2.1 7a associated with the CoC No. 1004, Amendment No. 8 for the standardized NUHOMSO-32PT storage system and the 0S197L Transfer Cask. Finally, OPPD requests an exemption from 10 CFR 72.48(c)(2)(viii) to use a method of thermal analysis (for the transfer trailer) that the NRC has determiined during inspection activities to be a departure from methodology described in the NUHOMS6 UFSAR.

The transfer trailer thermal analysis has been re-performed using spent fuel attributes specific to Fort Calhoun Station instead of generic design basis attributes in order to demonstrate ample margin from the peak clad temperature limit. Details of the requested exemption are contained in Attachment 1 of this letter.

There is insufficient time for the items included in this exemption request to be addressed through the Certificate of Compliance amendment process. The requested exemption is limited to a single loading campaign of four 32PT canisters and will support the Fort Calhoun Station refueling outage, which starts September 9, 2006. The outage includes the replacement of major components of the reactor coolant system, i.e., two steam generators, the reactor vessel head, and the pressurizer. Due to the scope of these activities and the associated risks of inadvertent Employment with Equal Opportunity

U.S. Nuclear Regulatory Commission LIC-06-0056 Page 2 introduction of foreign material into the system, OPPD needs to preserve full core offload capability following the 2006 refueling outage in order to safely and efficiently address potential problems.

Approval of this exemption will allow Fort Calhoun Station to maintain full core offload capability after the 2006 refueling outage, will allow receipt and storage of new fuel, and will allow better management of decay heat loads within the Spent Fuel Pool (including minimization of fuel handling activities) until the spring 2008 refueling outage.Section VI of Attachment I describes the need for loading four canisters to provide these capabilities.

Expedited approval of this exemption request is needed in order to complete the canister loadings and thereby minimize outage scheduling and radiological impacts on required preparations for the 2006 refueling outage. These activities include preparations for the major component replacements that are taking place in the area exterior to the Auxiliary Building, immediately adjacent to the spent fuiel pool and spent fuiel transfer trailer travel route. Other outage preparation activities requiring use of the same area include receipt, inspection, and storage for 44 new fuel bundles and 49 new control element assemblies (control rods) before their placement into the spent fuel pool. All of these activities enhance nuclear, radiological, and industrial safety.

Please note that the only elements of the spent fuel loading operations impacted by this exemption are related to the spent fuiel transfer activities. Specifically, the storage mode at the Independent Spent Fuel Storage Installation is not affected.

To support the loading of four 32 PT canisters prior to the fall 2006 Outage, OPPD requests approval of the exemption before July 7, 2006.

If you require additional information, please contact Thomas C. Matthews at (402) 533-6938.

Si:ely, 4ambhir Division Manager - Nuclear Projects Fort Calhoun Station SKG/rlj

U.S. Nuclear Regulatory Commission LIC-06-0056 Page 3 Attachments:

1. Omaha Public Power District - Fort Calhoun Station Request for Exemption from NUHOMS8 Certification of Compliance No. 1004, Amendment No. 8

2. FSAR Change Notice - FCN 721004-321, Rev. 1

3. TN Calculation 1121-0504, Revision 1, 0S197L 75 Ton Transfer Cask Dose Rates Calculation to be used with OPPD Exemption Request

4. TN Calculation 1121-0400, Revision 1, Calculation of 0S197L Cask Shell Temperature with 11.0 and 18.4 kW Heat Loads

5. TN Calculation 1121-0401, Revision 1, 0S197L 75 Ton Transfer Cask Thermal Analysis to be used with OPPD Exemption Request (18.4 kWIDSC & 11.0 kW/DSC) cc: Director of Consumer Health Services, Department of Regulation and Licensure, Nebraska Health and Human Services, State of Nebraska

LIC-06-056 Page 1 Omaha Public Power District Fort Calhoun Station Request for Exemption from NUHOMS8 Certification of Compliance No. 1004, Amendment No. 8 I. Exemption Request II. General Background III. Technical Specification Exemption Requests A. Technical Specification 1.2.1 (Bases)

1. Exemption Request - TS 1.2.1

2. Background -TS 1.2.1

3. Technical Justification - TS 1.2.1 B. Technical Specification 1.2.11

1. Exemption Request - TS 1.2.11

2. Background -TS 1.2.11

3. Technical Justification - TS 1.2.11 C. Technical Specification 1.2.1 7a

1. Exemption Request - TS 1.2.17a

2. Background -TS 1.2.17a

3. Technical Justification - TS 1.2.17a IV. Technical Justification - Remaining Sections of the UFSAR Associated with TS 1.2.1, 1.2.11, 1.2.17a Exemptions V. 72.48(c)(2)(viii) - Methods of Analysis Exemption Request

1. Exemption Request

2. Background - 72.48(c)(2)(viii) - Methods of Analysis Exemption

3. Technical Justification - 72.48(c)(2)(viii) - Methods of Analysis Exemption VI. Environmental Assessment - TS 1.2.1, 1.2.11, 1.2.17a VII. References

LIC-06-056 Page 2 Omaha Public Power District Fort Calhoun Station Request for Exemption from NUHOMSO Certificate of Compliance No. 1004, Amendment 8

1. Exemption Request Pursuant to 10 CFR 72.7, "Specific exemptions," Omaha Public Power District (OPPD) requests an exemption from the requirements of 10 CFR 72.2 12(a)(2),

10 CFR 72.212(b)(2)(i)(A), 10 CFR 72.212(b)(7), and 10 CFR 72.214, all of which require the licensee to comply with the terms and conditions of the NRC issued certificate of compliance. In connection with these exemptions, OPPD also requests an exemption -- to the extent necessary -- from 10 CFR 72.48(c)(1)(B), which requires a verification that design changes do not involve changes to the terms, conditions, or specifications of the CoG. Specifically, OPPD is requesting an exemption from three of the Technical Specifications that are part of Certificate of Compliance (CoC) No. 1004, Amendment No. 8 for the standardized NIJHOMS8-32PT storage system [Reference 1] to be used at Fort Calhoun Station (FCS).

The affected CoC 1004 Technical Specifications (TS) are 1.2.1, 1.2.11, and 1.2.17a.

These specifications address transfer cask dose rates and the canister vacuum drying time limits.

In addition, to address NRC concerns, OPPD requests exemption from 10 CFR 72.48(c)(2)(viii) to use a method of thermal analysis (for the transfer trailer) that the NRC has determined during inspection activities to be a departure from that described in the NUHOMS0 UFSAR [Reference 2]. The transfer trailer thermal analysis has been re-performed using spent fuel attributes specific to Fort Calhoun Station instead of generic design basis attributes in order to demonstrate ample margin from the peak clad temperature limit.

OPPD, will comply with all other Conditions of Use and Technical Specifications of CoC No. 1004, Amendment No. 8 during fuel loading at ECS.

OPPD requests that an exemption be issued before July 7, 2006, because there is insufficient time for the items included in this exemption request to be addressed through the Certificate of Compliance amendment process. Expedited approval of this exemption' request is needed in order to complete canister loadings and thereby minimize outage scheduling and radiological impacts on activities whose completion are necessary preparations for the 2006 refueling outage. These activities include preparations for the major component replacements that are taking place in the area exterior to the Auxiliary Building, immediately adjacent to the spent fuel pool and spent fuel transfer trailer travel route. Other outage preparation activities requiring use of the same area include receipt, inspection, and storage for 44 new fuel bundles and 49 new control element assemblies (control rods) before their placement in the spent fuel pool. All of these activities enhance nuclear, radiological, and industrial safety.

LIC-06-056 Page 3 The requested exemption term is limited to loading of four 32PT canisters with spent fufel and will apply only during loading activities (there are no exemptions necessary with respect to storage of the canisters). The requested exemption is limited to loading of four 32PT canisters with fuel having a maximum total canister heat load of 11 kW.

11. General Background Updated ESAR (UFSAR) [Reference 2] contains two alternate versions of a Transfer Cask (TC). They are referred to as the Standardized Cask (with a solid neutron shield) and OS 197/0S1I97H/OSL197FC cask (with a liquid neutron shield). This TC is used for loading/unloading fuel into/from a canister and moving a loaded Dry Shielded Canister (DSC) from the spent fuel pool to the Horizontal Storage Module (HSM). In general the use of this cask requires a nominal 100 ton crane lift capacity.

In an effort to expand the capability of the NUHOMSO system to plants with reduced crane capacity, a light weight configuration of the OS 197 TC was developed, referred to as the 05197L. The design intent of this cask is to allow for the loading/unloading and transfer of the licensed DSCs (24P, 52B, 6lBT, 24PT2, 32PT and 24PHB) and maintain the bounding crane load to less than 75 tons.

TN, the certificate holder for the NUHOMSO system, utilized the 72.48 process to add the 0S197L TC to the NUHOMSO system and issued an UFSAR Change Notice (FCN) that added the description, analysis description, and analysis results for the 0S197L system to the CoC 1004 IJFSAR. In the following discussions the UFSAR, Revision 9, including FCN 321, Revision 1, is referred to as the UFSAR.

OPPD plans to use the 0S197L TC to load four 32PT canisters using the 75-ton capacity Auxiliary Building crane. This will allow Fort Calhoun Station to maintain full core offload capability after the 2006 refueling outage, will allow receipt and storage of new fuel, and will allow better management of decay heat loads within the Spent Fuel Pool (including minimization of fuel handling activities) until the spring 2008 refuieling outage.Section VI describes the need for loading of four canisters to provide these capabilities.

As part of the inspection process at FCS prior to initiation of fuel loading, NRC identified and verbally communicated several issues/concerns, which were responded to by OPPD. Continued communications among NRC, TN, and OPPD management has determined that submittal of an exemption request is the optimum path for use of the OS5197L TC at FCS.

For conservatism, the exemption request is based on loading of specific FCS fuel that is less reactive, and with significantly lower decay heat and source term, than the Design Basis fuel assembly in CoC 1004 for the 32PT DSC. These decay heat reductions are applied on both the total canister level and the specific Fuel Assembly (FA) level. This is addressed in Table 1 and subsequent discussion.

LIC-06-056 Page 4 OPPD has determined that it is acceptable and conservative to load the FCS FAs into the 32PT DSC utilizing the 0S197L TC. The FCS FAs are bounded by a substantial margin by the design basis fuel assemblies that are evaluated for the current license in the UFSAR. As such, the requested exemption will not endanger life or property or the common defense and security.

The exemption will be in the public interest in that it will allow for the safe and efficient storage of spent nuclear fuel at FCS. NRC approval of the exemption will preserve full core off-load capability, allow receipt and storage of new fuel, and will allow better management of decay heat loads within the Spent Fuel Pool (including minimization of fuiel handling activities) until the spring 2008 refueling outage.

Please note that the only elements of the spent fuel loading operations impacted by this exemption are related to the spent fuel transfer activities. Specifically, the storage mode at the Independent Spent Fuel Storage Installation is not affected.

Table 1 Summary of Key Parameters for FCS CE 14 x 14 Fuel Assemblies and NUIIOMS'-32PT DSC Design Basis Fuel Assemblies FCS CE 14 x 14 NUHOMSý-32PT Fuel Assembly Design Basis Fuel

___________________Parameters Parameters Maximum Total Decay Heat load per NUHOMS*-32PT 11.0 24.0 DSC (kW)

Maximum Total Decay Heat Ranges from Load per Fuel Assembly 0.50 maximum 0.60 to 1.20 (kW) ___________

Maximum Assembly Average Burnup 42,049 45,000 (MWD/MTU)_ _ _ _ _

Maximum Initial Bundle Average Enrichment 4.5 5.0 (wt% U235)

Maximum Initial Uranium 0.377 0.475 Content(MTU/Assembly) ___________ __________

The proposed FCS fuel assembly zoning configuration was redefined to incorporate the lower fuel assembly decay heat load (0.50 kW maximum) zoning limits and therefore results in reduced peak fuel clad temperatures. (See Figure 1 and Reference 6)

LIC-06-05 6 Page 5 FIGURE 1 - Proposed FCS Fuel Loading Configuration (I11 kWIDSC) 3 3

- mm 2 2

r. 3 2 I 2 3 ml -

1 0-0-- 1 3 '2- 1 I 2 3 2 2 3

Heat.Zone 1 2 3 4 J

ofFuel Assemblies 48 8 1 4 Max Heat Load / Assembly (IcV) 0.16 j0.35 0.40 j0.50 0.50 Max Heat Load/I Zone (kW%) 0.64 j 2.80 3.20 ,4.00 , 2.00 Max Heat Load / DSC (LW) 11.0 Note: This is a bounding fuel load for the fuel assemblies (CE 14x14) to be loaded at FCS in the 32 PT DSC. Conservatively, the total heat generation used in the 3D DSC thermal model based on the above fuel loading configuration is 12.64 kW per DSC. The total heat load to be loaded at FCS in the Phase I campaign will be less than 11 kW per DSC.

LIC-06-056 Page 6 An evaluation is presented below for the structural, thermal, and shielding NIJHOMSO UFSAR sections as they are affected by the reduced parameters of the proposed FCS FAs.

Structural Evaluation The structural evaluation of the NUHOMSe-32PT DSC is documented in the UFSAR. The design parameters for the design basis fuel assembly used in Chapter M.3 of the UFSAR (e.g., total fuel assembly weight and total decay heat load) are unchanged and bound the FCS fuel assembly. This exemption has no adverse impact on the structural evaluation.

Thermal Evaluation The thermal evaluation of the NUHOMSO-32PT DSC is documented in Chapter MA4 of the UFSAR. The method used for the thermal evaluation of the FCS fuel: 1)

Evaluated the effective fuel properties (thermal conductivity) of the FCS fuel assembly and compared it with the NUHOMS'ý-32PT DSC design basis fuel

[Reference 2]; and 2).Determined the margin of conservatism between the FCS FA and the Design Basis FA to demonstrate that the corresponding thermal analysis results for the NUH{OMSO-32PT DSC with the FCS fuel assembly can be conservatively used for thermal evaluation of the NUHOMSO-32PT DSC design basis fuel in the UPSAR.

As shown in Table 1, the maximum decay heat per DSC and maximum assembly average burnup for the FCS FAs are all bounded by the design basis values used for the NUIHOMSw-32PT thermal evaluation. (See Section V)

The FCS fuel assembly effective thermal conductivities are addressed by a review of Figures M.4-19 through M.4-21 of the UFSAR. These figures demonstrate the significantly higher thermal conductivity of the ECS CE 14 x 14 FA over that of the Design Basis WE 14 x 14. As shown in the figures, the CE 14 x 14 FA has thermal conductivity values nominally 20% or more higher than the Design Basis WE 14 x

14. This higher conductivity will reduce the AT across and along the FA, and result in lower fuel clad temperatures.

The maximum decay heat load for the FCS fuel assemblies proposed to be loaded in the NU}IOMSP-32PT DSC is 11.0 kW per DSC. This is approximately 55% less than the design basis heat load for NUHOMSO-32PT DSC, [Reference 2] and a 77%

reduction in individual FA heat load for the four center FA's, compared to the design basis configuration (Configuration 3). The FCS fuel thermal properties as described above will result in lower AT, across and along the FA, and therefore the thermal evaluation in the UFSAR for the design basis fuel in the NUHOMSý-32PT DSC during transfer for normal, off-normal and accident conditions remains bounding for the FCS fuel assemblies with decay heat loads less than or equal to 11.0 kWIDSC and individual FA decay heat loads limited per the proposed FCS zoning configuration.

LIC-06-056 Page 7 Shielding Evaluation Chapter M.5 of the UFSAR documents the shielding evaluation for the NUHOMSO-32PT DSC. Chapter M.5 of the FSAR states:

"The B&W 15x15 assembly is the bounding fuel assembly design for shielding purposes because it has the highest initial heavy metal loading as compared to the 14x14, other 15xl5, and 17xl7 fuel assemblies which are also authorized contents of the NUHIOMSO-32PT DSC. In addition, the maximum Co59 content of the hardware regions for each assembly type is less than that of the B&W 15x15 Mark B fuel assembly."

The maximum initial heavy metal content of the FCS fuiel assembly is 0.377 MTU per assembly while the shielding design basis fuel assembly is 0.475 MTU. In addition, the total canister decay heat load (nominally proportional to source term) is reduced from the Design Basis 24 kW to 11.0 kW. Therefore, the design basis radiation source terms for all burnup, initial enrichment and cooling time combinations allowed to be stored in the NUIHOMSO-32PT DSC remain bounding for the FCS fuel assembly. As a result, all of the storage dose rates reported in the tables in the UFSAR and storage dose rate limits reported in the Technical Specifications remain bounding for FCS FAs. The exemption to TS 1.2.1 and 1.2.11 addresses OS 197L TC dose rates.

11I. Technical Specification Exemption Requests A. Technical Specification 1.2.1 (Bases)

I1. Exemption Request - TS 1.2.1 Certificate of Compliance No. 1004 (CoC), Amendment No. 8, includes Technical Specification 1.2.1 which controls fuel specifications. An exemption is requested in connection with the statements in the Bases section that describe the transfer cask surface dose rates for the 24P and the 52B canisters. The exemption would allow disregarding the wording on the transfer cask surface dose rates in the Technical Specification Bases. All other elements of the Limit/Specifications and Applicability remain in force.

2. Background - TS 1.2.1 OPPD acknowledges that the description of the transfer cask surface dose rates, described in the Bases section of Technical Specification 1.2.1 for the 24P and 52B canisters, cannot be met with the use of the bare 0S197L TC, and therefore an exemption is requested. It is noted that OPPD will only load the 32PT canister, and not the 24P or the 52B DSCs.

3. Technical Justification - TS 1.2.1 10 CFR 72.7 specifies that "... the Commission may, upon application by any interested person or upon its own initiative, grant such exemptions from the

LIC-06-056 Page 8 requirements of the regulations in this part as it determines are authorized by law and will not endanger life or property or the common defense and security and are otherwise in the public interest."

The safety analysis of the NUHOMSO-32PT system is described in the current license in Appendix M of the Updated Final Safety Analysis Report (UFSAR) for the standardized NUHOMS system [Reference 2]. The current Technical Specifications (TS) for CoC No. 1004 [Reference 1] that are issued to TN for the standardized NUJ{OMS8 system contain the following requirements regarding the authorized content of the DSC:

Technical Specification 1.2.1, "Fuel Specifications, Functional and Operating Limits," states that "The characteristics of the spent fuel which is allowed to be stored in the standardized NUHOMS system are limited by those included in Tables i-ia, 1-1b, i-ic, i-id, i-ie, i-if, i-ig, i-li, 1-1j, i-il, and I-im."

The fuel assemblies (CE i4 x i4) that OPPD proposes to load in the 32PT canister are fully compliant with the relevant Tables (I - I e, 1-If, and I-I g) of the Technical Specification.

In addition, the Bases section of this TS contains the following wording regarding the radiological criterion:

"The radiological design criterion is that fuel stored in the NUIIOMSO system must not increase the average calculated HSM or transfer cask surface dose rates beyond those calculated for the 24P, 24PHB, 52B, 6iBT, or 32PT canister full of design basis fuel assemblies with or without BPRAs. The design value average HSM and cask surface dose rates for the 24P and 52B canisters were calculated to be 48.6 mremlhr and 591.8 mremlhr respectively based on storing twenty four (24) Babcock and Wilcox 15x15 PWR assemblies (without BPRAs) with 4.0 wt. % U-235 initial enrichment, irradiated to 40,000 MWCI!MTU, and having a post irradiation time of five years.

To account for BPRAs, the fuel assembly cooling required times are increased to maintain the above dose rate limits."

OPPD proposes to load only the 32PT canisters at FCS under this exemption.

Specific dose rate limits for the 32PT are not included in this Bases section.

Specific 0S197L TC dose rate measurements will be taken as noted in connection with the proposed exemption for TS 1.2.i11.

An evaluation of the required NUHOMS8 UFSAR sections has been completed. The results of that evaluation for the structural, thermal, shielding, and criticality disciplines are summarized below for each of the three Technical Specifications to be exempted. An evaluation summarizing the impact on other UIFSAR sections is provided for all three TSs at the end of this section.

LIC-06-056 Page 9 Structural Evaluation The structural evaluation of the NUHOMSO-32PT DSC is documented in the UFSAR. The design parameters for the design basis fuel assembly used in Chapter M.3 of the UFSAR (e.g., total fuel assembly weight and total decay heat load) are unchanged and bound the FCS fuel assembly. This TS, and the exemption, have no adverse impact on the structural evaluation.

Thermal Evaluation The thermal evaluation of the NUHOMSO-32PT DSC is documented in Chapter M.4 of the UFSAR. The design parameters for the design basis fuel assembly used in Chapter M.4 of the UFSAR (e.g., total fuel assembly burnup, cooling time, and decay heat) bound the proposed FCS FAs (CE 14 x

14) limited to 11.0 kW and the proposed zoning configuration. As discussed in the General Background above, significant thermal margin is present. As a result, the thermal evaluation results reported in Chapter MA4 of the UFSAR remain bounding.

Shielding Evaluation The shielding evaluation of the NUHOMS'ý-32PT DSC is documented in Chapter M.5 of the UFSAR. The design parameters for the design basis fuel assembly used in Chapter M.5 of the UFSAR (e.g., total fuel assembly burnup, cooling time, and source term) bound the proposed FCS FAs (CE 14 x 14) limited to 11.0 kW As a result, the shielding evaluation results reported in Chapter M.5 of the UFSAR remain bounding and ALARA principles will be met.

Criticality Analysis The criticality evaluation of the NUHOMSO-32PT DSC is documented in Chapter M.6 of the UFSAR. The design parameters for the design basis fuel assembly used in Chapter M.6 of the UFSAR (e.g., fuel assembly configuration, enrichment, and pool boron loading) are unchanged and the DSC geometry and materials are unchanged. As a result, the criticality evaluation results reported in Chapter M.6 of the UFSAR remain bounding.

LIC-06-056 Page 10 B. Technical Specification 1.2.11

1. Exemption Request - TS 1.2.11 Certificate of Compliance No. 1004 (CoG), Amendment No. 8, includes Technical Specification 1.2.11 which provides dose rate limits for the transfer cask. An exemption is requested in connection with all elements of the TS, including the Limit/Specification and Applicability. The exemption is justified because equivalent dose rate limits can be met through use of the shielding elements of the 0S197L system at the FCS site using the 32PT DSC.

The exemption would be only for the 32PT DSC to be loaded with FCS FAs, and as such no exemption to TS 1.2.11 a, 1.2.1 lb, or 1.2.11 c is necessary.

2. Background -TS 1.2.11 OPPD acknowledges that the description of the transfer cask surface dose rates, described in the Technical Specification 1.2.11, does not explicitly address the supplemental shielding, and that the limits cannot be met with the use of the bare OS1I97L TC alone.

3. Technical Justification - TS 1.2.11 10 CFR 72.7 specifies that "... the Commission may, upon application by any interested person or upon its own initiative, grant such exemptions from the requirements of the regulations in this part as it determines are authorized by law and will not endanger life or property or the common defense and security and are otherwise in the public interest."

The safety analysis of the NUHOMSO-32PT system is described in the current license in Appendix M of the Ugdated Final Safety Analysis Report (UIFSAR) for the standardized NUHOMS system [Reference 2]. The intent of the TS, as stated in the Objective, is to identify a FA misload, and to support ALARA operations. These objectives are captured through use of the following dose rate measurements.

Dose rate measurements of the 0S197L containing a loaded 32PT DSC will be taken at FCS at selected steps in the loading sequence. The specific dose rate parameters for the OS I97L TC are contained in Table 2 and Table 3:

LIC-06-056 Page 11 Table 2 Axial Dose Rate Measurement Configuration

32PT DSC inside 051 97L inside decon sleeve/bell

Water drained from DSC

TC/DSC annulus full (within approximately 1 foot of top)

TC neutron shield full

Top Shield Plug in place and included in axial shielding

Inner Top Cover Plate in place and included in axial shielding

" Automated Welding System (AWS) with integral shield in place and included in axial shielding

" Measurement taken at vertical centerline of DSC, 3 feet from AWS shield Calculated Dose Rate [Reference 4] ....... 128 nirem/hr Proposed Dose Rate Limit.............. 170 mrem/hr

LIC-06-056 Page 12 Table 3 Radial Dose Rate Measurement Configuration

32PT DSC inside OS 197L inside decon sleeve/bell

Water drained from DSC 0 TC/DSC annulus full (within approximately 1 foot of top) 0 TC neutron shield full

6 inch nominal thickness carbon steel decon sleeve/bell in place and included in radial shielding

Measurement taken at outside surface (contact) of decon sleeve/bell Calculated Dose Rate [Reference 4] ......... 68 mrem/hr Proposed Dose Rate Limit............... 110 mrem/hr In addition to the above dose rate measurements to be taken as a replacement to TS 1.2.11, OPPD will take additional dose rate measurements during the dry fuel loading process as part of the normal OPPD Part 50 Radiation Protection (RP) and ALARA programs. To assist in ALARA planning, Transnuclear has calculated the bare cask surface dose rates in calculation 1121-0505 Rev.0 [Reference 7]. The 0S197L TC average surface dose rate for the Fort Calhoun specific fuel being loaded under this exemption (11.0 kW DSC heat load) was calculated to be 13 Rem/H~r. This is significantly lower than -the 53 Rem/Hr dose rate calculated with design basis fuel. This calculation is available onsite for NRC review.

In the axial direction, the shielded configuration of the 32PT DSC inside the 0S197L is unchanged from the 32PT inside the OS 197. The limits are based on a calculation of the configuration [Reference 4] and the exemption explicitly states the configuration for which the dose rates were calculated and for which the measurements should be taken.

The axial measurement is the most favorable for identification of a fuel assembly misload condition. An analysis of the axial and radial dose rates for a potential fuel assembly misload of a 32PT DSC have been evaluated in TN calculation NUHO6L-0502 [Reference 3)1. The result of this calculation shows that a design basis assembly with design basis burnup and 1 year cooling time located in the center of the DSC cavity, would be easily detected by axial dose

LIC-06-056 Page 13 rate measurements from either the OS 197 TC or the 0S197L TC, since the calculated doses for this misloaded configuration far exceed TS 1.2.11 Limit/Specification "a" dose limits. This is expected, since the OS 197 and 0S1 97L TCs do not differ in axial shielding configuration and measurement in the axial direction eliminates the potential of other FAs to self-shield the misloaded FA.

The radial measurement serves as an ALARA check, and misload detection through radial measurements are not reliable. The misload analysis of Reference 3 also demonstrates that a radial dose rate measurement for both the OS 197 and 0S197L TCs would not detect a misloading via TS 1.2.11 Limit/Specification "b" for a misloaded assembly (1 year cooled) placed at the center of the DSC. However, a misloaded assembly placed other than at the center of the DSC may be detected via the dose rate measurements of TS 1.2.11 Limit/Specification "b," depending on the FA location within the 32PT basket, for both the OS 197 and 0S197L TC configurations. Therefore the only reliable dose rate measurement to detect a misload is the axial measurement. The proposed radial dose rate limits are based on a calculation of the configuration [Reference 4] and the exemption explicitly states the configuration for which the dose rates were calculated and for which the measurements should be taken.

The equivalent ALARA goal is accomplished with respect to TS 1.2.11 Limit.

The requirement of ALARA is reduction of exposure and the proposed dose rate limit will still assure that the 32PT used with the 0S197L TC at FCS will provide equivalent or better shielding than the OS 197 TC.

The exemption alternative actions will also utilize the same steps and actions of the TS Action Statement:

"If specified dose rates are exceeded, place temporary shielding around affected areas of transfer cask and review the plant records of fuel assemblies which have been placed in DSC to. ensure they conform to the fuel specifications of Section 1.2.1. Submit a letter report to the NRC within 30 days summarizing the action taken and results of the surveillance, investigation and findings. The report must be submitted using instructions in 10 CFR 72.4 with a copy sent to the administrator of the appropriate NRC regional office."

Note that the major movements utilizing the bare 0S197L TC will be

.performed remotely to minimize dose. Remote placement of the TC inside the decon shielding or onto the TT is no more complex or difficult than placement of the TC within the fuel pool. The remote operation has already been demonstrated. In addition, should all these features fail, proper positioning of the OS197L TC when being handled by a crane can be

LIC-06-056 Page 14 achieved by using a combination of remote and manual alignment practices.

In both cases the result is a safely landed TC.

Additionally, the unlikely event of a hung load due to crane malfunction has been addressed through contingency planning. Reliability of the crane to be used for dry fuel storage operations has been demonstrated through extensive usage during internal and NRC demonstrations. Crane operators have been trained to manually operate the crane in order to minimize radiological exposure in the event of a crane malfunction during movement of the TC.

Projected personnel doses which would be received during this operation have been determined to be ALARA and within the Fort Calhoun Station administrative limits.

Therefore, the use of the OS1 97L TC with the dose rate measurements noted in this exemption will allow OPPD to meet the dual objectives of detecting fuel misload and maintaining dose rates ALARA.

An evaluation of the required NUHOMSO UFSAR sections was completed.

The results of that evaluation for the structural, thermal, shielding, and criticality disciplines are summarized below for each of the three Technical Specifications to be exempted, followed by a combined summary of the remaining UFSAR sections.

Structural Evaluation The structural evaluation of the NUHOMSO-32PT DSC is documented in the UFSAR. The design parameters for the design basis fuel assembly used in Chapter M.3 of the UFSAR (e.g., total fuel assembly weight and total decay heat load) are unchanged and bound the FCS fuel assembly. This TS, and the exemption, have no adverse impact on the structural evaluation.

Thermal Evaluation The thermal evaluation of the NUHOMS'ý-32PT DSC is documented in Chapter MA4 of the UFSAR. This TS and the exemption have no adverse impact on the thermal evaluation. As a result, the thermal evaluation results reported in Chapter M.4 of the UFSAR remain bounding.

Shielding Evaluation The shielding evaluation of the NUHOMS(ý-32PT DSC is documented in Chapter M.5 of the UFSAR. As stated above, the axial configuration of the OS 197L TC was evaluated in [Reference 4]. The exemption explicitly defines the configuration to be measured and the dose rate limit to be used.

The radial shielding of the OS1 97L TC, including the decon area sleeve/bell, is documented in FCN 32 1, Rev. 1 [Reference 2], showing that the OS 197L

LIC-06-056 Page 15 TC system configuration has increased shielding. As the radial measurement is not a reliable misload detection method, and serves only as an ALARA check, the increased shielding supports this ALARA function. The exemption explicitly defines the configuration to be measured and the dose rate limit to be used [Reference 4].

Criticality Analysis The criticality evaluation of the NUIIOMSý-32PT DSC is documented in Chapter M.6 of the UFSAR. This TS, and the exemption, have no adverse impact on the criticality evaluation. As a result, the criticality evaluation results reported in Chapter M.6 of the UFSAR remain bounding.

C. Technical Specification 1.2.1 7a

1. Exemption Request - TS 1.2.17a Certificate of Compliance No. 1004 Amendment No. 8, includes Technical Specification 1.2.17a which provides time limits for vacuum drying for the 32PT. An exemption is requested in connection with the Limit/Specification wording on the start of vacuum drying. Instead, OPPD will conservatively start the time limit for vacuum drying earlier in the loading sequence and will use helium as the backfill gas. No change to the time duration or the elements of the Action Statement is requested.

The exemption would be only for the 32PT DSC and as such no exemption to TS 1.2.17, 1.2.17b, or 1.2.17c is necessary.

Justification of the specific exemption is that the start of the time limit for vacuum drying, irrespective of canister heat load, will occur at the time that the initial 750 gallons draindown from the canister is achieved. Additional water will continue to be drained to meet crane weight limits. This will occur at FCS as the 0S197L TC with 32PT is lifted to the fuel pool surface and the drain down of water from the canister cavity is performed.

2. Background - TS 1.2.17a OPPD proposes to start the. vacuum drying clock (time limit) at the time that the initial. 750 gallon draindown from the canister is achieved. This will ensure that the initial conditions assumed in the vacuum drying analysis are met and therefore that the. fuel clad temperature limits, which are the subject of this TS, are maintained below the values listed in the UFSAR.

LIC-06-056 Page 16

3. Technical Justification - TS 1.2.17a 10 CFR 72.7 specifies that "... the Commission may, upon application by any interested person or upon its own initiative, grant such exemptions from the requirements of the regulations in this part as it determines are authorized by law and will not endanger life or property or the common defense and security and are otherwise in the public interest."

The safety analysis of the NUHOMSO-32PT system is described in the current license in Appendix M of the 14,dated Final Safety Analysis Report (UFSAR) for the standardized NUHOMS system [Reference 2]. This analysis includes in M.4, calculation of fuel cladding temperatures for the vacuum drying condition.

These analyses conservatively assume that the initial fuiel clad temperature at the start of vacuum drying (and therefore the start of the vacuum drying duration time limit) is 215*F. OPPD proposes to start the clock at the time that the initial 750 gallons draindown from the canister is achieved, which is prior to movement of the cask/canister to the decon area. The FA and fuel cladding at this time will have just been removed from the fuel pool water within the canister (draining 750 gallons is anticipated to take approximately 1-2 hours) and the FA will be approximately half submerged. Therefore the 215'F initial fuel clad temperature assumption is bounded.

All of the operations subsequent to the start of the vacuum drying clock, which include continued draining, movement of the cask/canister to the decon area, any refilling of the canister with fuel pool water, placement of the inner top cover plate and Automated Welding System (AWS), inner top cover plate welding, and start of actual vacuum drying, do not adversely impact the analysis assumptions of MA4 The cask/canister annulus will be maintained full to assure the boundary DSC exterior temperature of 215'F maximum.

Therefore the vacuum drying thermal analysis documented in M.4 bounds this sequence.

Once the cask/canister is placed in the decon area, the remaining steps for canister closure (Inner Top Cover Plate placement and welding) will be performed.

It is anticipated, based upon a review of the loading sequence, that vacuum drying can be completed within the 31 hour3.587963e-4 days 0.00861 hours 5.125661e-5 weeks 1.17955e-5 months limit of TS 1.2.1 7a. Should the end of vacuum drying not be accomplished within the TS 1.2.17a time limits, the existing TS action statement shall be entered, which requires that a helium atmosphere greater than 0.1 atm be established within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . Once accomplished, the vacuum drying clock can be reset and the vacuum drying restarted.

LIC-06-056 Page 17 An evaluation of the required NUHOMSO UFSAR sections has been completed. The results of that evaluation for the structural, thermal, shielding, and criticality disciplines are summarized below for the Technical Specification to be exempted, followed by a combined summary of the remaining UFSAR sections.

Structural Evaluation The structural evaluation of the NUHOMSý-32PT DSC is documented in the UFSAR. The design parameters for the design basis fuel assembly used in Chapter M.3 of the UFSAR (e.g., total fuiel assembly weight and total decay heat load) are unchanged and bound the FCS fuel assembly. This TS, and the exemption, have no adverse impact on the structural evaluation.

Thermal Evaluation The thermal evaluation of the NUHOMSO-32PT DSC is documented in Chapter MA4 of the UFSAR. This TS, and the exemption, have no adverse impact on the thermal evaluation. As a result, the thermal evaluation results reported in the UFSAR remain bounding. Fuel cladding temperatures during operations from the fuel pool to final vacuum drying will be maintained at or below the values listed in the UIFSAR.

Shielding Evaluation The shielding evaluation of the NUHOMSO-32PT DSC is documented in Chapter M.5 of the UFSAR. This TS, and the exemption, have no adverse impact on the shielding evaluation. As a result, the shielding evaluation results reported in the UFSAR remain bounding.

Criticality Analysis The criticality evaluation of the NUHOMSO-32PT DSC is documented in Chapter M.6 of the UFSAR. This TS, and the exemption, have no adverse impact on the criticality evaluation. As a result, the criticality evaluation results reported in the UFSAR remain bounding.

IV. Technical Justification - Remaining Sections of the UFSAR Associated with TS 1.2.1, 1.2.11, 1.2.17a Exemptions Confinement Evaluation The confinement evaluation of the system is documented in Chapter M.7 of the UFSAR. This section of the UFSAR is not adversely affected by the transfer cask dose rates or the start time/time limits for vacuum drying and therefore remains

LIC-06-056 Page 18 applicable when the revised cask dose rate measurements and vacuum drying start times are implemented.

Operating Systems The operating procedures for the system are documented in Chapter M.8 of the UFSAR. This section of the UFSAR is not adversely affected by the specifics of transfer cask dose rate measurements. The transfer cask dose rates will be measured in the decon area, as listed in the discussion for the TS 1.2.11 exemption. The sequence of canister loading is unchanged by the earlier start of the vacuum drying clock, and the major sequence steps are not altered.

Test and Maintenance Program The Test and Maintenance program for the system is documented in Chapter M.9 of the UFSAR. This section of the UFSAR is not adversely affected by the transfer cask dose rates or the start time/time limits for vacuum drying and therefore remains applicable when the revised cask dose rate measurements and vacuum drying start times are implemented.

Radiation Protection Occupational, Exposure and Off-site dose evaluations for the system are presented in Chapter M. 10 of the UFSAR. As addressed in the Shielding Evaluation discussion above, the Occupational Exposure and Off-site dose evaluations presented in Chapter M. 10 of the UFSAR remain bounding.

Accident Analysis Accident analyses for the system are presented in Chapter 11 of the UFSAR (includes FCN 321). As addressed in the disc'ussion for the Structural Evaluation, Thermal Evaluation, Shielding Evaluation, and Criticality Evaluation above, this section of the UFSAR is not adversely affected by the transfer cask dose rates or the start time/time limits for vacuum drying. Therefore, the accident analysis results presented in the UFSAR remain bounding.

Conditions for Cask Use - Onerating Controls and Limits or Technical Specification Conditions for cask use operating controls and limits or technical specifications for the system are presented in Chapter 12 of the UFSAR which refers to the Technical Specifications for the CoC No. 1004. Except for the exemptions described above for TS 1.2.1, 1.2.11, and 1.2.17a, all other TSs remain limiting.

Ouality Assurance The Quality Assurance program to be applied to the system is described in Chapter M. 13 of the UFSAR. This section of the UFSAR is not adversely

LIC-06-056 Page 19 affected by the transfer cask dose rates or the start time/time limits for vacuum drying and therefore remains applicable when the revised cask dose rate measurements and vacuum drying start times are implemented.

Decommissioning The decommissioning evaluation for the system is described in Chapter 14 of the UFSAR. This section of the UFSAR is not adversely affected by the transfer cask dose rates or the start time/time limits for vacuum drying and therefore remains applicable when the revised cask dose rate measurements and vacuum drying start times are implemented.

V. 72.48(c)(2)(viii) - Methods of Analysis Exemption Request

1. Exemption Request Pursuant to the provisions of 10 CFR 72.7, "Specific exemptions," OPPD, Fort Calhoun Station (FCS) requests an exemption from a requirement specified in 10 CFR 72.212, "Conditions of general license issued under §72.210. "

72.2 12 requires compliance with the requirements of 72.48 for all changes implemented using this provision. An exemption is requested from 72.48(c)(2)(viii).

2. Background - 72.48(c)(2)(viii) - Methods of Analysis Exemption During inspection activities, the NRC determined that the thermal analysis of the OS 197L TC on the transfer trailer with additional shielding utilizes a change in method of analysis, as defined in 72.48(c)(2)(viii). To address this issue, OPPD is submitting the thermal analysis for this configuration as part of the exemption

[References 5 & 6].

3. Technical Justification - 72.48(c)(2)(viii) - Methods of Analysis Exemption 10 CFR 72.7 specifies that "... the Commission may, upon application by any interested person or upon its own initiative, grant such exemptions from the requirements of the regulations in this part as it determines are authorized by law and will. not endanger life or property or the common defense and security and are otherwise in the public interest."

The thermal analysis submitted [Reference 5] with this exemption utilized a CED analysis Code (FLUENT) to calculate TC surface temperatures during transfer in the transfer trailer. These TC surface temperatures were then used in [Reference 6] to calculate DSC shell and peak fuel clad temperatures within the 32PT DSC.

The calculation was performed using the reduced heat load of 11.0 kW, and the reduced fuel assembly heat loads, proposed earlier. In addition, the analysis

LIC-06-056 Page 20 utilized the thermal conductivity values of the FCS FA (CE 14 x 14). The results of the calculation are presented below.

Table 4 Calculation of Peak FA Clad Temperature [Reference 6]

DSC Heat Maximum Peak Fuel Allowable Fuel Clad Load Clad Temperature Temperature (IF)

(kW) (OF) 11.0 472 752 The results demonstrate that significant thermal margin (>275 IF) is present.

VI. Environmental Assessment - TS 1.2.1, 1.2.11, 1.2.17a The following information is provided in support of an environmental assessment and finding of no significant impact for the proposed exemption:

Identification of the Proposed Action Pursuant to the provisions of 10 CFR 72.7, "Specific exemptions," OPPD requests an exemption from requirements specified in 10 CFR 72.2 12(a)(2), 10 CFR 72.212(b)(2)(i)(A), 10 CFR 72.212(b)(7), 10 CFR 72.214, and 10 CFR 72.48(c)(1)(B), all of which require the licensee to comply with the terms and conditions of the COC. The exemption would be from conditions in Amendment 8 to CoC No. 1004 for the NUHOMSO-32PT storage system. Specifically, OPPD is requesting an exemption from TS 1.2. 1(Bases), 1.2.11, and 1.2.1 7a.

In addition OPPD is requesting exemption from 72.48(c)(2)(viii), specifically to use a method of analysis that the NRC determined to be a departure from that described in the NUHOMSO UFSAR. These exemptions would allow OPPD to store fuel assemblies using the 32PT DSC and the OS 197L TC.

The Need for the Proposed Action Approval of this exemption will allow Fort Calhoun Station to maintain full core offload capability after the 2006 refueling outage, will allow receipt and storage of new fuel, and will allow better management of decay heat loads within the Spent Fuel Pool (including minimization of fuel handling activities) until the

LIC-06-056 Page 21 spring 2008 refueling outage.Section VI of Attachment I describes the need for loading of four canisters to provide these capabilities.

Expedited approval of this exemption request is needed in order to complete canister loadings and thereby minimize outage scheduling and radiological impacts on required preparations for the 2006 refueling outage. These activities include preparations for the major component replacements, that are taking place in the area exterior to the Auxiliary Building, immediately adjacent to the spent fuel pool and spent fuiel transfer trailer travel route. Other outage preparation activities requiring use of the same area include receipt, inspection, and storage for 44 new fuel bundles and 49 new control element assemblies (control rods) before their placement in the spent fulel pool. All of these activities enhance nuclear, radiological, and industrial safety.

There is insufficient time for the items included in this exemption request to be addressed through the Certificate of Compliance amendment process.

Part 10 CFR 72.7 specifies that the NRC may grant, exemptions from the requirements of 10 CFR Part 72 when the exemptions are authorized by law and will not endanger life or property or the common defense and security, and are otherwise in the public interest. OPPD has concluded that the conditions for granting an exemption are met and has provided the justification in this submittal.

The Need for Loading of Four Canisters Fort Calhoun Station's (FCS) Spent Fuel Pool (SFP) is designed with a capacity of 1083 cells. Prior to the 2006 Refuieling Outage (RFO) in September, FCS has 142 empty cells. Due to several operational constraints, FCS has detenmined that a minimum of four (4) dry shielded canisters need to be loaded and stored via this exemption request to prudently manage adverse impacts to nuclear, radiological and industrial safety. Five operational constraints are summarized below.

Operational Constraints Requiring The Loading of Four Dry Shielded Canisters Full core offload capability during and following the 2006 RFO

Management of decay heat loads if a full core offload is required

Receipt and inspection of new fuel assemblies control rods for the 2006 RFO

Compliance with NRC Regulatory Issue Summary 2005-05:

Regulatory Issues Regarding Criticality Analyses for Spent Fuel Pools and Independent Spent Fuel Storage Installations

Temporary storage of ultrasonic fuel cleaning filters and an incore thimble The first operational constraint is the need for FCS to offload the 133 assembly core during and following the 2006 RFO. The large amount of Reactor Coolant System (RCS) components being replaced during the outage raises the likelihood

LIC-06-056 Page 22 that foreign material could be introduced and potentially deposited under the core support. plate in the reactor vessel. This scenario would require the core to be offloaded to the SEP and the reactor core barrel to be removed to allow removal of the foreign material.

The second constraint is ensuring 50 additional cells are identified and empty following the 2006 REQ in order to manage decay heat loads if a fulll core offload is required. Technical Specifications dictate that fuel assemblies must be stored in one of two regions of the SEP based on each assembly's bum-up and enrichment. Additionally, FCS must distribute specific bundles throughout Region 2 of the SFP to adhere to NRC decay heat separation criteria and SFP concrete gamma heating operational constraints. If a post-2006 core offload is required, most of the newer fuel assemblies (Batch BB and CC) must be stored in Region 1 of the SEP and cannot be stored in a decay heat dispersion pattern.

However, approximately 50 fuel assemblies (20 to 80 assemblies depending on bumn-up at the time of the offload) can and should be stored within a heat dispersion pattern in Region 2 which will significantly lower the decay heat loading within Region 1. Removing 50 fuel assemblies now enables FCS to eliminate at least 50% of the fuel movements later, depending on core bum-up at the time of the offload, thereby substantially reducing the number of required fuel movements and the risk associated with these movements.

The third constraint results from the need to receive and inspect 44 new fuel assemblies and 49 new control rods for the 2006 REQ. Once inspections are complete, the assemblies and control rods are systematically transferred from the New Fuel Storage Rack into the SEP. This fuel handling operation requires more resources, presents more radiological challenges, and is more complicated than normal intra-SEP fuel movements. Consequently, it is ECS practice to perform these operations prior to a refueling outage.

The fourth constraint results from the requirements of two issues. The first requirement pertains to compliance with NRC Regulatory Issue Summary (RIS) 2005-05, "Regulatory Issues Regarding Criticality Analyses for Spent Fuel Pools and.Independent Spent Fuel Storage Installations." Complying with RIS 2005-05 requires the station to ensure the first row of SEP cells adjacent to the cask pit area (CPA) is empty. Keeping these 23 cells empty addresses criticality issues related to neutronic coupling of fuel in the dry shielded canister (DSC) during dry fuel storage activities. The second requirement is to keep the same 23 cells vacant of fuel assemblies while the ultrasonic fuel cleaning (UFC) chamber is stored in the CPA. The UEC vendor recommends this constraint to minimize long-term radiation exposure to the transducer equipment. FCS plans on installing the UFC after core offload and then leaving the UFC stored in the CPA for radiological reasons.

The fifth constraint is based on the need to temporarily store ultrasonic fuel cleaning (UFC) filters and an incore thimble storage can mn five (5) SEP cells.

LIC-06-056 Page 23 Dose rates from filters will exceed lOORem/hr and therefore must be stored in the SEP until sufficient decay time has elapsed. FCS currently has over twenty filters stored in all available locations on the walls of the SFP. The filters must be moved to four (4) SEP Cells in order to allow additional 2006 RFO filters to be stored in the SEP. In addition, a new storage can is being placed within a SFP cell to accommodate the incore thimbles that will be replaced during the 2006 RFO.

In summary, the operational constraints above will require the availability of 211 empty cells in the spent fuel pool at the end of the 2006 REG. Loading only 3 canisters will provide 194 empty cells in the spent fuel pool at the end of the 2006 RFO and thus will negatively impact the management of decay heat loads.

Loading 4 canisters will provide 226 empty cells in the spent fuiel pool at the end of the 2006 RFO, allowing OPPD to better manage nuclear, radiological, and industrial safety by fully addressing the operational constraints.

Environmental I~mpacts of the Proposed Action The NRC completed an Environmental Assessment of COG No. 1004, Amendment No. 8, in March 2005 and reached the following conclusions:

"Considering the specific design requirements for each accident condition, the design of the cask would prevent loss of containment, shielding, and criticality control. Without the loss of containment, shielding, or criticality control, the risk to public health and safety is not compromised.

The staff reviewed. the proposed changes and confirmed that the changes provide reasonable assurance that the spent fuel can be stored safely and that the changes meet the acceptance criteria specified in 10 CFR Part 72. The staff documented its findings in a Safety Evaluation Report. The occupational exposure is not significantly increased, and offsite dose rates remain well within the 10 CFR Part 20 limits. Therefore, the proposed action now under consideration would not change the potential environmental effects assessed in the initial rulemaking.

Therefore, the NRC staff has determined that an acceptable safety margin is maintained and that no significant environmental impacts occur as a result of the amendment.

Because the proposed changes will not change the environmental requirements for the storage of spent fuel, no change in environmental impact is anticipated."

OPPD concludes that the conclusions reached by the NRC in the Environmental Assessment for Amendment No. 8 remain valid with the implementation of more explicit transfer cask dose rate measurements, the earlier start time/time limits for vacuum drying, and the use of the submitted thermal analysis of the TC on the transfer trailer.

LIC-06-056 Page 24 The fuel assemblies which OPPD plans to load into the NUHOLMS'-32PT DSC are bounded by the design basis fuel assemblies for the 32PT DSC as evaluated in the UFSAR. The procedures that OPPD has used for selecting, loading and storing its spent fuel will also meet all the CoC requirements. The exemption will not significantly increase the probability or consequences of accidents. The use of remote handling techniques for the 051 97L TC during loading operations will be consistent with ALARA principles and will not adversely increase occupational or public radiation exposures. There are no changes being made in the types or amounts of effluents that may be released offsite, and there is no significant increase in occupational or public radiation exposure as a result of the proposed activities. Therefore, there are no significant radiological environmental impacts associated with the proposed exemption.

With regard to potential non-radiological environmental impacts, OPPD has determined that the proposed exemption has no potential to affect any historic sites. It does not affect non-radiological plant effluents and has no other environmental impact. Therefore, there are no significant non- radiological environmental impacts associated with the requested exemption.

Environmental Impacts of the Alternatives to the Proosed Action As an alternative to the requested exemption, the NRC could consider denial (i.e.,

the "no-action" alternative). Denial of the exemption would result in no change to the current environmental impacts. OPPD considers the "no-action" alternative to potentially impact OPPD's ability to provide safe, affordable, competitive, and reliable electrical power generation.

VI1. References I1. Certificate of Compliance No. 1004 for the Standardized NUHOMS8 System, Amendment 8, effective December 5, 2005.

2. "Standardized NUHOMSO Horizontal Modular Storage System for Irradiated Nuclear Fuel, Final Safety Analysis Report," NUH-003, Revision 9, dated January 2006, including all issues FSAR Change Notices f{Includes FCN 321, Rev. 1 (Attached)}.

3. TN Calculation NUHO6L-0502, Revision 0, 05197/05197L Transfer Cask Shielding Evaluation for Single Assembly Misloading.

4. TN Calculation 1121-0504, Revision 1, 0S197L 75 Ton Transfer Cask Dose Rates Calculation to be used with OPPD Exemption Request. (Attached)

5. TN Calculation 1121-0400, Revision 1, Calculation of 0S197L Cask Shell Temperature with 11.0 and 18.4 kW Heat Loads. (Attached)

6. TN Calculation 1121-0401, Revision 1, 0S197L 75 Ton Transfer Cask Thermal Analysis to be used with OPPD Exemption Request (18.4 kWIDSC

& 11.0 kW/DSC). (Attached)

7. TN Calculation 1121-0505, Revision 0, 0S197L 75 Ton Bare Transfer Cask Dose Rates for as Loaded Configuration to be Used With OPPD Exemption Request (18.4 kW/DSC & 11.0 kW/DSC).

LIC-06-.056 Page 1 FSAR Change Notice - FCN 721004-321, Rev. 1

FCN 721004-321 Rev. 1' Page 1 of 99 A

TRANSNUCLEAR

~ Change Notice

~~~~~FSAR (FCN)

CNo:

72.48 Review F 71032 LR 721004-321, AN AREACOMPAN No.: Revision 1 Preparer/Date Verifier/Date-Miguel Manrigue / James Axiine Usama Farradj 31`3 (Ac G Approved: U.B. Chopra ae 4xfs /3 1r 5"k--

Jayant Bondre Date:

A marked up copy of the applicable FSAR pages and drawings clearly showing the change implemented by the 72.48 Review must be attached to this form:

" List changed UFSAR pages (if applicable) attached to this FSAR Change Notice:

Page 1.1-2a Page 1.1-2b Page 1.3-3 Page 1.3-3a Page 3.1-4 Page 3.2-7 Page 4.2-9 Page 5.1-1 Page 7.1 -1 Page 8.1 -1 Page E-2 Add new Appendix W (NUHOMSO OSI 97L Transfer Cask)

" List changed FSAR Drawings (if applicable) attached to this FSAR Change Notice:

None Form 3.5-4 Revislon 0 Page I of 1

FCN 721004-321 Rev. 1 Page 2 of 99 The NUHOMSR-24PTH system adds a new canister with three alternate configurations (designated as DSC Type 24PTH-S, -24PTH-L, or -24PTH-S-LC), a new module designated as HSM-H, and a modified version of 0S197/0S197H transfer cask designated as OS 197FC/0S 197H FC.

A detailed description of the 24PTH system, including drawings, authorized payload contents and supporting safety analyses for this system are provided in Appendix P of this UPSAR.

Amendment 8 to CoC also authorized storage of low enrichment and reconstituted fuel in the 32PT DSC. In addition, the authorized contents of the 24PHB DSC were revised to include additional fuel types. A detailed description of the changes implemented to the 32PT and 24PHB DSCs are provided in Appendices M and N, respectively.

.TN has added two alternate HSMs, designated as HSM Model 152 and HSM Model 202, to the standardized NUHOMSO system. These alternate HSM designs provide enhanced shielding features while meeting the heat rejection requirements. A detailed description of the HSM Model 152 and HSM Model 202 and supporting analyses are provided in Appendices R and V, respectively.

Chapters 1 though 8 and Appendices A through H of this ESAR provide the supporting licensing basis for the Standardized NUHOMSO-24P and -52B systems only.

Appendix W has been added to the UPSAR to incorporatea light weight (75 ton) version of the 05197 onsite transfercask A complete description of the new systems addressed by the above listed amendments, including supporting safety analysis, is located within self-contained Appendices to this FSAR as sunmmarized in the following table:

NUJH-003 Revision 9 Page 1.1-2a January 2006

FCN 721004-321 Rev. 1 Page 3 of 99 Amendment DsrpinLocation of Supporting No. D cipinLicensing Basis 3 Addition of the NUHOMS*-GlBT DSC to the contents of Appendix K the Standardized NUROMS system __________

N/A Addition of the NUHOMS"-24PT2 DSC to the contents Appendix L of the Standardized NUHOMSO system__________

4 Addition of low burnup fuel to the contents of the Chapter 3 NUHOMSP-24P DSC 5 Addition of the NUHOMS'-32PT DSC to the Apni Standardized NUH-OMSO system Apni 6 Addition of the NUHOMSO-24PHB DSC to the Appendix N Standardized NUHOMS4 system __________

7 Addition of damaged fuel to the contents of the Appendix K NUliIOMS0-6IBT DSC (a) Addition of the NUHOMS 24PTH system to the Appendix P Standardized NUHOMS'P system 8 (b) Revision of the authorized contents of the 32PT DSC Appendix M, to include low enrichment and reconstituted fuel (c) Revision of the authorized contents of the 24PHIB Appendix N DSC to include additional fuel types___________

Addition of an alternate version of the HISM, designated N/A as H-SM Model 152, to the Standardized NUHQMSO Appendix R system Addition of an alternate version of the HSM, designated N/A as HSM Model 202, to the Standardized NUHOMS4 Appendix V

___________system Addition of an alternateversion of the 0S197 Transfer N/A Cask designated as OS)97L, to the Standardized Appendix W NUHOMrS system _ _ _ _ _ _ _

NUII-003 Revision 9 Page 1.1-2b Page .1-2bJanuary 2006 1

FCN 721004-321 Rev. 1 Page 4 of 99 controlled access. The necessary civil work required to prepare the ISFSI site is the same as that for an ISFSI utilizing vertical storage casks.

Two alternate designs of the standardized HSM are available for licensees' use: the original HSM, now designated as HSM Model 80 and HSM Model 102. USM Model 102 design is similar to USM Model 80 design except for the following two features:

The steel encased composite door of HSM Model 80 design is replaced by a two foot thick reinforced concrete door with a steel liner on its inside surface. The steel liner mitigates DSC damage from spalled concrete due to tornado generated missile impact.

The inlet and outlet vents, which are formed in concrete for HSM Model 80, are lined with IPY" steel plates.

The above features included with HSM Model 102 are improvements to the original HSM Model 80 design that increase the shielding capabilities of the HSM. The heat transfer capability (decay heat rejection from the DSC to the HSM and heat removal from the HSM by natural convection) of both HSM Model 80 and HSM Model 102 designs are equivalent. Appendix E drawings show both models. Each model can store a DSC with maximum weight up to 102 kips which includes 24P, 52B, 24PT2 and 61BT DSCs.

1.3.2 Transfer Systems Descriptions 1.3.2.1 On-Site TC The transfer cask used in the NUHOMSt system provides shielding and protection from potential hazards during the DSC closure operations and transfer to the HSM. Fouralternate configurations of the transfer cask are available for the licensees' use. The basic configuration, where the cask is provided with a solid neutron shield, is described herein as the "Standardized Cask." An alternate configuration, where a liquid neutron shield is provided instead, is described in this SAR as the "OS 197, OS 197H or 0S197L Cask."

The configuration of the OS 197 is a slightly modified version of the NRC approved cask (with a liquid neutron shield) as described in the NUHOMS-24P Topical Report (1.10). The standardized transfer cask documented in this SAR has a gross weight of less than 90.7 Te (100 tons) and is limited to on-site use under 10CFR72. The OS 197 and 05197H1 transfer casks, which are also limited to on-site use under I00FR72, have a maxhimum gross weight of 94.6 Te (104.25 tons) and 113.4 Te (125 Tons), respectively. In addition, the licensee may also elect to utilize a future transfer cask having a gross weight of about 113.4 Te (125 tons) which can be used on-site under 10CFR72, but is also suitable for future off-site shipment of intact NUHOMSO canisters under 10CFR7J.

Where applicable, any other NRC licensed NUHOMSD transfer or transportation cask is acceptable for use with the standardized NUHOMSO system subject to an application specific safety evaluation.

The third configuration of the transfer cask, designated as OS197FC/0S197H FC, is a modified version of 0S197/0S197H equipped with a modified lid to allow air circulation through the TC/DSC annulus, and is described in Appendix P.

NUII-003 Revision 9 Page 1.3-3 January2006

FCN 721004-321 Rev. 1 Page 5 of 99 A fourth configuration of the transfer cask; designated as 0S197.L TC, is a lighter version of the 0S197L 7)Z It is designedfor use byfacilities with a crane capacit of 75 tons and is described in Appendix W.

NUJH-003 Revision 9 Page 1.3-3a Januaty 2006

FCN 721004-321 Rev. 1 Page 6 of 99 Once inside the HSM, the DSC and its payload of SFAs is in passive dry storage. Safe storage in the HSM is assured by a natural convection heat removal system, and massive concrete walls and slabs which act as biological radiation shields. The storage operation of the HSMs and DSCs is totally passive. No active systems are required.

3.1.2.1 -Handling and Transfer Equipment The handling and transfer equipment required to implement the NUHOMSo system includes a cask handling crane at the reactor fuel pool, a cask lifting yoke, a transfer cask, a cask support skid and positioning system, a low profile heavy haul transport trailer and a hydraulic ram system. This equipment is designed and tested to applicable governmental and industrial standards and is maintained and operated according to the manufacturer's specifications. Performance criteria for this equipment, excluding the fuel/reactor building cask handling crane, is given in the following sections. The criteria are summarized in Table 3.1-7.

On-Site Transfer Cask: The on-site transfer cask used for the NUHOMSO system has certain basic features. The DSC is transferred from the plant's fuel pool to the HSM inside the transfer cask. The cask provides neutron and gamma shielding adequate for biological protection at the outer surface of the cask. The cask is capable of rotation, from the vertical to the horizontal position on the support skid. The cask has a top cover plate which is fitted with a lifting eye allowing removal when the cask is oriented horizontally. The cask is capable of rejecting the design basis decay heat load to the atmosphere assuming the most severe ambient conditions postulated to occur during normal, off-niormal and accident conditions. For the NUHOMSE)-24P, 24PHB DSC or the NUHOMSO-24PT2 DSC, the standardized transfer cask has a cylindrical cavity of 1.73m (68 inches) diameter and 4.75m (186.75 inches) in length and a maximum dry payload capacity of 42,321 Kg (93,300 pounds). For the NUIHOMSP-52B3 or NUHOMS-ý6IBT, the standardized transfer cask is fitted with an extension collar to accommodate the longer BWR DSC and fuel. Alternatively, the OS 197 and 0S197H transfer casks with a full length cavity of 5.Om (196.75 inches) may be used for the NUHOMSO-24P, 24PHB (with cask spacer), NUHQMSO-52B, NUHOMSO-61BT DSCs, NUHOMS0-24PT2 DSC (with cask spacer) or NUHOMSP-32PT DSC (with cask spacer). The 0S197 and 0S197H casks can carry a maximum dry payload of 44,100 kg (97,250 lb) and 52,600 kg (116,000 Ib), respectively. These payload capacities are based on a transfer cask weight of 111,250 pounds. The cask and the associated lifting yoke are designed and operated such that the consequences of a postulated drop satisfy the current IOCFR50 licensing bases for the vast majority of plants. Appendix Wprovldes a detailed description of the 0S197L transfer cask The NUHOMSO transfer cask is designed to meet the requirements of 10CFR72 (3.6) for normal, off-normal and accident conditions. The NUHOMSO transfer cask is designed for the following conditions:

NUHI-003 Revision 9 Paize 3.1-4 Januar2006

FCN 721004-321 Rev. 1 Page 7 of 99 summarizes the stress criteria for DSC non-pressure boundary components (except for support rods). The spacer discs are designed using the component stress criteria from ASME Code Subsection NB (for Service Levels A, B, C) and ASME Code Appendix F (Service Level D, Elastic and Elastic/Plastic analysis). The support rods are designed using the criteria of ASME Code Subsection NF for linear type component supports (for Service levels A, B, C) and ASME Code Appendix F (for Service Level D stress or stability criteria). For Service Level A the limits of NF-3 322 are used. For Service Levels B and C the factors of Table NF-3523(b)-l are used. For Service Level D, the criteria from Appendix F is used. The 24P guide sleeves and oversleeves are designed using the stress criteria of ASME Code Subsection NB and ASME Code Appendix F, and the stability criteria of Subsection NF and Appendix F, as applicable. All non-pressure boundary partial penetration and fillet welds are designed using the stress criteria of ASME Code Subsection NE and ASME Code Appendix F.

Other components of the DSC include the support ring, the lifting lugs, the shield plugs, the grapple ring and grapple ring support plate, and all welds associated with these components. The support ring is designed using the ASME Code Subsection NB criteria.

The associated weld to the DSC shell is a partial penetration weld evaluated to the ASME Code Subsection NF and Appendix F requirements, as applicable. The lifting lugs and associated welds are designed using Subsection NE allowables. The grapple ring, grapple ring support plates and associated welds are designed using the ASME Code Subsection NB design criteria. The shield plugs are non-pressure boundary components and need only to maintain their structural integrity. The shield plugs are evaluated using Subsection NB primary stress limits. The shield plugs stiffener welds in the long cavity basket are full penetration welds designed to Subsection NE.

3.2.5.3 On-site Transfer Cask The on-site transfer cask is a non-pressure retaining component which conservatively is designed by analysis to meet the stress allowables of the ASME Code (3.14) Subsection NC for Class 2 components. The cask is conservatively designed by utilizing linear elastic analysis methods. The load combinations considered for the transfer cask normal, off-normal, and postulated accident loadings are shown in Table 3.2-7. Service Levels A and B allowables are used for all normal operating and off-normal loadings. Service Levels C and D allowables are used for load combinations which include postulated accident loadings. Allowable stress limits for the upper lifting trunnions and upper trunnion sleeves are conservatively developed to meet the requirements of ANSI N 14.6-1993 (3.37) for a non-redundant lifting device for all cask movements within the fuel/reactor building. The maximum shear stress theory is used to calculate principal stresses in the cask structural shell. The appropriate dead load and thermal stresses are combined with the calculated drop accident scenario stresses to determine the worst case design stresses. The transfer cask structural design criteria are summarized in Table 3.2-11 and Table 3.2-12. The transfer cask accident analyses are presented in Section 8.2. The effects of fatigue on the transfer cask due to thermal cycling are addressed in Section 8.2.10. Appendices K, L, M and N address the effects of handling the NUHOMSP-61BT, -20P12, 32PT and 24PHB DSC in the transfer cask, respectively.

The effects of handling the licensed (-<4. 0 kW) DSCs (24P, 52B, 6JBT, 24PT2, 32PT and 24PHB) in the 0S197L TC are addressed in Appendix W.

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FCN 721004-321 Rev. 1 Page 8 of 99 OR

The coarse and fine aggregates to be one or a mix of the following: limestone, dolomite, marble, basalt, granite, rhyolite, gabbro. Determination of the aggregate constituents shall be done in accordance with the same methods described above.

For all PWR and BWR HSM components the above aggregate requirements can be waived if the criteria established by Appendix D for strength reduction is further validated by strength tests performed on the actual concrete mix to be used for construction subjected to elevated temperatures established by the design. Alternatively the minimum compressive strength requirements for the concrete may be increased to account for an appropriate reduction in concrete strength. This approach removes the need to reevaluate the HSM design analyses.

4.2.3.3 On-Site Transfer Cas The on-site transfer cask is a nonpressure-retaining cylindrical vessel with a welded bottom assembly and bolted top cover plate. The transfer cask is designed for on-site transport of the DSC to and from the plant's spent fuel pool and the ISFSI as shown in Figure 4.2-10 and Figure 4.2-11. The transfer cask provides the principal biological shielding and heat rejection mechanism for the DSC, and SFAs during handling in the fuel/reactor building, DSC closure operations, transport to the ISFSI, and transfer to the HSM. The transfer cask also provides primary protection for the loaded DSC during off-normal and drop accident events postulated to occur during the transport operations. The NUHOMSO transfer cask is illustrated in Figure 1.3-6. Drawings of the transfer cask are contained in Appendix E.

The transfer cask may be fitted with a shielded collar to extend the cask cavity length to accommodate the longer NUJHOMS'O-52B DSC as shown in Figure 4.2-12. The collar is a heavy forged steel ring with a bolt circle to match that of the transfer cask top flange and cover plate. Alternatively, a NUHOMS" transfer cask with a longer cavity length may be used for DSCs with PWR (with cask spacer) or BWR fuel.

The transfer cask to be used by a utility may be any one of the designs documented in Appendix B, including the standardized cask, OS 197, 0S197H or 0S197L. The licensee may also use any other previously NRC reviewed and approved design such as the transfer cask designs documented in the NUHOMS4-24P Topical Report [4.13], the Oconee Nuclear Station ISFSI Safety Analysis Report [4.16], and the Calvert Cliffs ISFSI Safety Analysis Report [4.17], provided it is demonstrated prior to use that the limiting conditions of use as described in CoC 1004 can be met.

The transfer cask is constructed from three concentric cylindrical shells to form an inner and outer annulus. These are filled with lead and a neutron absorbing material. The two inner shells are welded to heavy forged ring assemblies at the top, and bottom ends of the NUH-003 Revision 9 Page 4.2-9 January 2006

FCN 721004-321 Rev. 1 Page 9 of 99

5. OPERATION SYSTEMS This Chapter presents the operating procedures for the, standardized NUHOMS' system described in previous chapters and shown on the drawings in Appendix E for the 24P and 52B systems. The operating procedures for the 6lBT, 24PT2, 32PT, 24PHB and 24PTH systems are described in Appendices K. L, M, N and P, respectively. The procedures include preparation of the DSC and fuel loading, closure of the DSC, transport to the ISFSI, DSC transfer into the HSM Model 80 and Model 102, monitoring operations, and DSC retrieval from. the HSM Model 80 and Model 102. The operating procedures involving the HSM-H, HSM Model 152, and HSM Model 202 are described in Appendices P, R, and V, respectively. The standardized NUHOMSO transfer equipment, and the existing plant systems and equipment are, used to accomplish these operations.

Procedures are delineated here to describe how these operations are to be performed and are not intended to be limiting. Standard fuel and cask handling operations performed under the plant's IOCFR50 operating license are described in less detail. Existing operational procedures may be revised by the licensee and new ones may be developed according to the requirements of the plant, provided that the limiting conditions of operation specified in Technical Specifications, Functional and Operating Limits of the NUHOMSO' CoC (5.6) are not exceeded.

Appendix W.8 provides a description of the changes in operationalsequences when each of the licensed DSCs *524.0 kW (241P, 52B, 6JBT, 241'T2, 32PT and 24PHB) are transferredin an OS)197L TC.

5.1 Operation Description The following sections outline the typical operating procedures for the standardized NUHOMSO system. These generic NUHOMSo procedures have been developed to minimize, the amount of time required to complete the subject operations, to minimize personnel exposure, and to assure that all operations required for DSC loading, closure, transfer, and storage are performed safely. Plant specific ISFSI procedures are to be developed by each licensee in accordance with the requirements of IlOCFR72.24 (h) and the guidance of Regulatory Guide 3.61 (5.7). The generic procedures presented here are provided as a guide for the preparation of plant specific procedures and serve to point out how the NUHOMS4 system operations are to be accomplished. They are not intended to be limiting, in that the licensee may judge that alternate acceptable means are available to accomplish the same operational objective.

The generic operating procedures presented herein also do not address the use of auxiliary equipment which is optional or represents a level of detail which a licensee may choose to implement based on licensee preference. Examples of such auxiliary items are the Neutron Shield Overflow Tank (used with OS 197 or 0S197H Cask only), TC/DSC Annulus Pressurization Tank, and the Shield Plug Restraints.

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7. RADIATION PROTECTION The analysis presented in this Chapter is specifically applicable to the storage of the NUHOMSO-24P and -52B DSCs in the HSM Model 80 and Model 102 and transfer in the standardized cask; 0S197 or 0S197H TOs. Appendices J, K and L pro~vide similar evaluations for the NUHOMS0-24P long cavity, -61BT and -24PT2 systems, respectively. Appendices R and V provide an evaluation of these various DSCs stored in the HSM Model 152 and HSM Model 202, respectively.

Shielding analysis of the licensed DSCs (-<24. 0 k99 (24P, 52B, 6IBT, 24PT2, 32PT and 24PHB) when transferredin 0S197L TC areprovided in.Appendix W.

7.1 Ensuring That Q "cuational Radiation Exposures Are As-Low-As-Reasonby Achievable (ALARM) 7.1.1 Policy Considerations The licensee's existing radiation safety and ALARA policies for the plant should be applied to the ISFSI. The ALARA program should follow the general guidelines of Regulatory Guides 1.8, 8.8, 8.10 and IOCFR2O. ISFSI personnel should be trained and updated on ALARA practices and dose reduction techniques. Implementation of ISFSI systems and equipment procedures should be reviewed by the licensee to ensure ALARA exposure during all phases of operations, maintenance and surveillance.

7.1.2 Design Considerations The design of the NUHOMSO DSC and HSM comply with IOCFR72 ALARA requirements. Features of the NUHOMSO system design that are directed toward ensuring ALARA are:

A. Thick concrete walls and roof on the HSM to minimize the on-site and off-site dose contribution from the ISFSI.

B. A thick shield plug on each end of the DSC to reduce the dose to plant workers performing drying and sealing operations, and during transfer and storage of the DSC in the HSM.

C. Use of a heavy shielded transfer cask for DSC handling and transfer operations to ensure that the dose to plant and ISFSI workers is minimized.

D. Fuel loading procedures which follow accepted practice and build on existing experience.

E. A recess in the HSM access opening to dock and secure the transfer cask during DSC transfer so as to reduce direct and scattered radiation exposure.

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8. ANALYSIS OF DESIGN EVENTS In previous chapters of this SAR, the features of the standardized NUJHOMS system which are important to safety have been identified and discussed. The purpose of this chapter is to present the engineering analyses for normal and off-normal operating conditions, and to establish and qualify the system for a range of credible and hypothetical accidents. As stated in Chapter 1, the analyses presented in-this section are applicable to the standard length 24P and 52B canisters.

An evaluation of the long cavity 24P canister, for the same design criteria, is provided in Appendix H and 1. Appendices K, L, M, N and P provide the evaluation for the NUHOMS"-

61BT 24PT2, 32PT, 24PHB and 24PTH DSC, respectively. Also, as noted in Chapter 1, the structural, thermal, and shielding evaluations for the HSM-H, HSM Model 152, and HSM Model 202 are provided in Appendices P, R, and V, respectively. Evaluations for other canisters and modules may be included as additional appendices at a later time.

The structuraland thermal analysis of the licensed DSCs (_Q4.O0 kff9 (24P, 52B, 6JBT, 24PT2, 32PT and 24PHB) when transferred in 0S197L TC are provided in Appendix W.3 and W.4, ,

respectively.

In accordance with NRC Regulatory Guide 3.48 (8.1), the design events identified by ANSI/ANS 57.9-1984, (8.2) form the basis for the accident analyses, performed for the standardized NUHOMSO system. Four categories of design events are defined. Design event Types I and II cover normal and off-normal events and are addressed in Section 8.1. Design event Types III and IV cover a range of postulated accident events and are addressed in Section 8.2. These events provide a means of establishing that the NUHOMS6 system design satisfies the applicable operational and safety acceptance criteria as delineated herein.

It is important to note that, given the generic nature of this SAR,ý the majority of the analyses presented throughout this chapter are based on bounding conservative assumptions and methodologies, with the objective of establishing upper bound values for the responses of the primary components and structures of the standardized NIJHOMSO system for the design basis events. Because of the conservative approach adopted herein, the reported temperatures and stresses in this chapter envelope the actual temperatures or states of stress for the various operating and postulated accident conditions. More rigorous and detailed analyses and/or more realistic assumptions and loading conditions would result in temperatures and states of stress which are significantly lower than the reported values.

8.1 Normal and Off-Normal Operations Normal operating design conditions consist of a set of events that occur regularly, or frequently, in the course of normal operation of the NUHOMSO system. These normal operating conditions are addressed in Section 8.1.1. Off-normal operating design conditions are events that could occur with moderate frequency, possibly once during any calendar year of operation. These off-NLTH-003 Revision 9 Pam 81-1 .Januar. 2006lI

FCN 721004-321 Rev. 1 Page 12 of 99 This appendix contains the following items:

EA Drawings for NUHOMS'0 Dry Shielded Canisters(')

E. 1.1 Standardized NUHOMS4-24P DSC Drawings E.1.2 Standardized NUHOMS'-52B DSC Drawings E. 1.3 Standardized NUHOMS 6 -24P Long Cavity DSC Drawings E.2 Drawings for NUJHOMSe Horizontal Storage Module(2) (HSM Model 80 and Model 102 only)

E.3 Drawings for NUHOMSO On-Site Transfer Cask(3)

I

"~The drawings for the NUHOMSO-6IBT, 24PT2 and 32PT DSCs are contained in Appendices K, L and M, respectively. The drawings for the NUHOMS0-24PHB DSCs are contained in Appendices E and N. The drawings for the NUHOMSe-24PTH system (24PTH DSC, HSM-H and OS 197FC transfer cask) are contained in Appendix P.

(2 The drawings for the NUHOMS" ESM Model 152 are contained in Section R.1.5 of Appendix P.

(3 The drawingsfor the NUHOMLSO 0S1 97L transfer cask are contained in Section W.I15 of Appendix W.

NUTH-003 (PROPRIETARY INFORMATION)

Revision 9 E-2 January2006

FCN 721004-321 Rev. 1 Page 13 of 99 APPENDIX W NUHOMSO OS1I97L TRANSFER CASK NLTH-003 Revision 9 Page i Page January 2006

FCN 721004-321 Rev. 1 Page 14 of 99 TABLE OF CONTENTS Page W.1 General Description.......................................................................... W. 1-1 W.1.1 Introduction ......................................................................... W.1-1 W. 1.2 General Description of the NUHOMSO OS 197L TC ........................... W. 1-1 W.1.3 Identification of Agents and Contractors ......................................... W. 1-3 W. 1.4 Generic Cask Arrays ............................................................... W. 1-3 W.1.5 Supplemental Data ................................................................. W.1-4 W.1.6 References........................................................................... W. 1-4 W.2 Principal Design Criteria .................................................................... W.2-1 W.2.1 Spent Fuel To Be Stored ........................................................... W.2-1 W.2.2 Design Criteria for Environmental Conditions and Natural Phenomena......W.2-1 W.2.3 Safety Protection Systems ......................................................... W.2-2 W.2.4 Decommissioning Considerations................................................. W.2-2 W.2.5 Summary of NUHOMSO OS197L TC Design Criteria.......................... W.2-2 W.3 Structural Evaluation......................................................................... W.3-1 W.3.1 0S197L TC Description ........................................................... W.3-1 W.3.2 Design Criteria...................................................................... W3-2 W.3.3 OS197L TC Weight ................................................................ W.3-2 W.3.4 Mechanical Properties of Materials ............................................... W.3-2 W.3.5 General Standards for Casks....................................................... W.3-2 W.3.6 Normal/Off-Normal Structural Evaluation ....................................... W.3-3 W.3.7 Applicability of 05 197 TC Accident Drop Evaluations to the OS 197L TC... W.3-3 W.3.8 Effect of Increased OS5197L Temperatures on DSC Shell and Basket Components......................................................................... W.3-4 W.3.9 References........................................................................... W.3-4 WA4 Thermal Evaluation .......................................................................... W.4-1 WA4. Discussion........................................................................... W.4-1 W.4.2 Summary of Thermal Properties of Materials.................................... W.4-1 W.4.3 Specifications for Components .................................................... W.4-2 WA4A Thermal Analysis of an OS5197L TC Containing a 32PT DSC with 24 kW Heat Load ........................................................................... W.4-2 W.5 Shielding Evaluation......................................................................... W.5-1 W.5.1 Methodology........................................................................ W.5-1 W.5.2 Model Specification ................................................................ W.5-1 W.5.3 Shielding Evaluation ............................................................... W.5-1 W.5.4 References........................................................................... W.5-3 W.6 Criticality ..................................................................................... W.6- 1 W.7 Confinement.................................................................................. W.7-1 W.8 Operating Procedures........................................................................ W.8-1 NUH-003 Revision 9 Page ii January 2006

FCN 721004-321 Rev. 1 Page15 of 99 W.8.1 Operational Differences between 0S197L and 0S197 TCs ................... W.8-1 1 W.9 Acceptance Criteria and Maintenance Program............................................ W.9-1 W. 10 Radiation Protection........................................................................ W. 10-1 W. 11 Accident Analyses .......................................................................... W.11-1 W. 11.1 Postulated Accidents.............................................................. W. 11-1 W.12 Operating Controls and Limits............................................................. W. 12-1 W. 13 Quality Assurance .......................................................................... W. 13-1 W.14 Decommissioning........................................................................... W.14-1 NUH-003 Revision 9 Page iii January 2006

FCN 721004-321 Rev. 1 Page 16 of 99 LIST OF TABLES Page Table W.1-1 Comparison of Key Parameters of NUHOMSO OS 197 Versus OS 197L TCs.................................................................................. W.1-5 Table W. 1-2 0S197L TC UFSAR Sections Affected ......................................... W.1-6 Table W.2-1 0S197L TC System Components and Safety Classification ................... W.2-3 Table W.3-1 Summary of OS197L TC Weights................................................ W.3-5 Table W.3-2 Summary of Maximum Stress Ratios for Critical Lifts ......................... W.3-6 Table W.3-3 Summary of Maximum Stress Ratios for Level A Load Combinations ......W.3-6 Table W.3-4 Summary of Maximum Stress Ratios for Level C Load Combinations ......W.3-6 Table W.4-1 32PT DSC Shell and OS197L TC Component Maximum Temperatures without Supplemental Shielding................................. W.4-6 Table W.4-2 32PT DSC Shell Maximum Temperatures for 0S197L (without Supplemental Skid Shielding) and OS 197 TCs.................................. W.4-7 Table W.4-3 32PT DSC Shell and OS5197L TC Component Maximum Temperatures (Supplemental Skid Shielding Effect Included)................. W.4-8 Table W.4-4 Summary of Maximum Fuel Cladding and DSC Component Temperature Increase during Transfer in 0S197L TC.......................... W.4-9 Table W.4-5 Maximum DSC Internal Pressure during Transfer in OS 197L TC ........... W.4-10 Table W.4-6 Maximum Component Temperatures for OS 197L TC during Fire Accident........................................................................... W.4- 11 Table W.5-1 OS 197L TC Normal Condition Dose Rates ..................................... W.5-4 Table W.5-2 Dose Rate Results for Two Trunnion Designs (niremlhr) ...................... W.5-4 NUH-003 Revision 9 Page iv Page ivJanuary 2006

FCN 721004-321 Rev. 1 Page 17 of 99 LIST OF FIGURES Page Figure W.1-1 0S197L TC Configuration........................................................ W.1-8 Figure W.1-2 NUHOMSa' 05197L TC System Decontamination Area Shielding........... W.1-9 Figure W. 1-3 0S51 97L Transfer Equipment Schematic ....................................... W. 1-10 Figure W. 1-4 OS I97L TC System on Transfer Trailer with Shielding ...................... W. 1-11 Figure W.3-1 0S197L TC ANSYS Stress Analysis Model .................................... W.3-7 Figure W.3-2 OS I97L TC ANSYS Analysis Model - Upper and Lower Trunnions Detail................................................................................ W.3-8 Figure W.3-3 ANSYS Model Stress Analysis Results - Upper Trunnion Region............ W.3-9 Figure W.4-1 0S197L TC ANSYS Model..................................................... W.4-12 Figure W.4-2 Details of OS197L TC ANSYS Model......................................... W.4-13 Figure W.4-3 Temperature Plot for 32PT DSC (24 kW) in OS 197L TC without Supplemental Shielding, T~b=1000 F, Insolation .............................. W.4-14 Figure W.5-1 OS5197L TC and Decontamination Area Shielding Model Geometry ......... W.5-5 NUH-003 Revision 9 Page v 2006 Page vJanuary

FCN 721004-321 Rev. 1 Page 18 of 99 W. 1 General Description Appendix W to the NUHOMSO Updated Final Safety Analysis Report (UFSAR) addresses the Important to Safety aspects of adding the 0S197L TC to the Standardized NUHOMS~ system described in the UFSAR. The OS 197L TC is added to the UFSAR as an alternative to the OS 197 and OS 19711 TCs. The primary reason for adding the OS 197L TC design is to include a transfer cask that can be used by facilities with a crane capacity of 75 tons.

The 0S197L TC accommodates both PWR (187") and BWR (197") length Dry Shielded Canisters (DSCs), including the 24P, 5213, 24PT2, 6 1BT, 32PT and 24PH{B DSCs.

The format of this Appendix follows the guidance provided in NRC Regulatory Guide 3.61 [1.1].

The analyses presented in this Appendix demonstrate that the OS 197L TC system meets all the requirements of 10 CFR 72 [1.2].

Several sections of this Appendix have been identified as "No Change." For these sections, the description or analysis presented in the corresponding sections of the UFSAR for the Standardized NUIOMSO system is also applicable to the OS197L TC. In addition, tables and figures presented in the UFSAR which remain unchanged due to the addition of the OS 197L TC to the Standardized NUHOMSO system are not repeated in this Appendix. Table W.1-2 provides a summary of the sections of the main body of the UFSAR applicable to the OS 197 TC and addresses the impact of the 0S197L TC on these sections.

Note: References to sections or chapters within this Appendix are identified with a prefix W (e.g., Section W.2.3 or Chapter W.2). References to sections or chapters of the UFSAR outside of this Appendix (i.e., main body of the UFSAR) are identified with the applicable UFSAR section, chapter number or Appendix number (e.g., Section 2.3, Chapter 2 or Appendix K). The references used in this Appendix are identified as [X.X] (e.g., [1.1] is reference 1.1 at the end of Chapter W. 1).

OS 197 and OS197H TCs in the remainder of this Appendix will be referred to as OS 197 TC.

W. 1.1 Introduction As stated in Section 1.2. 1, the body of this UFSAR is dedicated to three on-site transfer cask types: the Standard cask, NUHOMS"-OS 197 and NUHOMSPkOS 197H TCs. The purpose of this Appendix is to provide the safety analysis of the design of a fourth type of on-site transfer cask, designated as the NUHOMSO OS 197L TC, for use with the standardized NUHOMSO system.

W.1.2 General Description of the NUHOMSO 0S197L TC The 68 metric ton (Te) (75 tons) OS I97L TC on-site transfer cask is designed to accommodate plants whose crane capacity can not accommodate the use of the 94.6 Te (104.25 tons) OS 197 TC or the 113.4 Te (125 Tons) 0S197H TC cask for fuel transfer. The major differences between the OS 197L TC and the OS 197/0S 197H1 casks are:

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FCN 721004-321 Rev. 1

" Reduced cask weight Page 19 of 99

- No lead shielding (one 2.68" nominal thickness steel shell instead of a combination of a 0.5" nominal thickness steel inner liner, 3.5" nominal thickness lead shield and 1.5" nominal thickness steel structural shell)

One piece solid trunnion configuration for the upper and lower cask trunnions

" Two piece neutron shield (inner and outer shell of 1/4" nominal thickness versus an outer shell of 3/16" nominal thickness)

The OS 197L TC key design parameters are compared to the OS 197 TC in Table W.1I- 1.

The OS I97L TC used in the NUJHOMSO system provides shielding and protection from potential hazards during the DSC fuel loading/unloading operations and transfer to the HSM. The design and configuration of the OS 197L TC is a modified version of the NRC approved OS 197 and OS I97H TCs described in Section 1.3.2.1 of the UFSAR and is limited to on-site use under I1OCFR72. The OS I97L TC can be configured to meet a gross weight limit of 68 Te (75 tons).

Figure W.1I-1I provides an overview of the OS I97L TC. The OS 197L TC configuration also requires the use of additional shielding in the decontamination area (see Figure W. 1-2) and on the skid/trailer (see Figure W. 1-3).

W. 1.2.1.1 Transfer Equipmen Transport Trailer: The NUHOMS" 0S197L TC transport trailer consists of a heavy industrial trailer with a payload capacity of 136 Te (150 tons), including the skid and loaded cask. The 0S197L TC transport trailer is the same as the one shown in Figure 1.3-7 of the UFSAR.

Cask Support Skid: The 0S197L TC support skid differs from the 0S197 TC support skid shown in UFSAR Figure 1.3-8 as described below:

1. The OS 197L TC support skid has permanently mounted 2.5" thick side shielding and accommodates an additional 3" thick side shielding bolted to the permanent shielding when transferring the OS 197L TC.

2. The OS197L TC also has a 2.5" shielding inner top cover and an additional 3" shielding outer top cover to shield the upper sections of the cask.

The OS197L TC support skid utilized for the standardized NUHOMS0 system is illustrated in Figure W.1-3.

Hydraulic Ram: The high capacity hydraulic ram system is similar to the hydraulic ram system described in the UFSAR. The capacity of this ram is increased in order to increase the ram capacity margin (and to accommodate other future DSC designs). There is no change to the maximum ram forces allowed (80 kips) during system operation.

A picture of the OS 197L TC system is provided in Figure W. 1-4.

NUH-003 Revision 9 Page W. 1-2 January 2006

FCN 721004-321 Rev. 1 Page 20 of 99 W. 1.2.2 Operational Features The primary operations with the OS I97L TC (in sequence of occurrence) for the NUHOMSO system are the same as the systems operation described in Section 1.3.3 of the UFSAR except as noted below for operations 8 and 13 (of Section 1.3.3):

Lifting Cask from Pool: The loaded OS 197 TC is lifted out of the pool and placed (in the vertical position) in a decontamination area shield on the drying pad in the decon pit. Prior to the lift, the DSC water is pumped out and a helium or nitrogen gas blanket is provided for the fuel assemblies.

The OS 197 TC neutron shield and the TC/DSC annulus is maintained full.

Placement of Cask on Transport Trailer Skid: The OS 197 TC is then lifted onto the cask support skid. The neutron shield may be drained during this operation if water is maintained in the DSC/cask annulus with an interim cover. The plant's crane is used to downend the cask from a vertical to a horizontal position. Inner top shielding is added to the skid and the cask is also covered with an additional outer top shielding. The outer top additional shielding is to be installed inside the fuel handling building if the floor loads can accommodate it (if floor loading is a concern, the additional shielding may be placed on the skid outside the fuel handling building). The neutron shield is filled, if previously drained, prior to draining of the annulus and replacement of the interim cover with the standard cask cover. The cask is then secured to the skid and readied for the subsequent transport operations.

W. 1.3 Identification of Agents and Contractors Transnuclear, Inc. (TN) provides the design, analysis, licensing support and quality assurance for the NUHOMSO OS197L TC. Fabrication of the NUHOMSO OS 197L TC is done by one or more qualified fabricators under TN's quality assurance program described in Chapter W. 13.

This program is written to satisfy the requirements of Subpart G of I1OCFR72, [1.2] and covers control of design, procurement, fabrication, inspection, testing, operations and corrective action.

TN provides specialized services for the nuclear fuel cycle that support transportation, storage and handling of spent nuclear fuel, radioactive waste and other radioactive materials. TN is the holder of NUHOMSO CoC 1004 [1.3].

W.1.4 Generic Cask Arrays No change.

NUH-003 Revision 9 Pane W. 1-3 January 2006

FCN 721004-321 Rev. 1 Page 21 of 99 W.1.5 Supplemental Data The following TN drawings are enclosed:

1. NUHOMSO OS 197L Onsite Transfer Cask, Cask Body Assembly, Drawing NUH-03-8008-SAR.

2. NUHOMSO OS 197L Onsite Transfer Cask, Light Neutron Shield Assembly, Drawing NUH-03-8009-SAR.

3. NUHOMSO OS 197L Onsite Transfer Cask, OS5197L Main Assembly, Drawing NUH-03-80 10-SAR.

W. 1.6 References

[1.1] U.S. Nuclear Regulatory Commission, Regulatory Guide 3.61, Standard Format and Content for a Topical Safety Analysis Report for a Spent Fuel Dry Storage Cask, February, 1989.

[1.2] 10CFR72, Rules and Regulations, Title 10, Chapter 1, Code of Federal Regulations -

Energy, U.S. Nuclear Regulatory Commission, Washington, D.C., "Licensing Requirements for the Independent Storage of Spent Nuclear Fuel and High-Level Radioactive Waste."

[1.3] NUHOMSO Certificate of Compliance for Dry Spent Fuel Storage Casks, Certificate Number 1004, Amendment No. 8, December 2005 (Docket 72-1004).

NUH-003 Revision 9 Paue W.1-4 January 2006

Figure Withheld Under 10 CFR 2.390 PqCN 721004-321 ' ~

A TRANSNUCLEAR AN AREVA COWARY

s ~NUHO4SO 0S197L A ONSITE TRANSFER CASK CASK BODY ASSEMBLY 2 I

Figure Withheld Under 10 CFR 2.390 ROM DRAWING7 MADE NU23.8O0S8.jSVJ

Figure Withheld Under 10 CFR 2.390 MADE FROM DRAWING I I NUH-23-8008' REV10 2 1

Figure Withheld Under 10 CFR 2.390 MADE FROM DRAWING IA

/NUH-03-8o8, REV 0 2 I

Figure Withheld Under 10 CFR 2.390 MAEFROMDRAWINGI

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22H-3-8OO8, REV 0 F-- 3O-WS-511". I -hiK'Ibj 2 I 1 '

Figure Withheld Under 10 CFR 2.390 IMDEROM DRAWING L NU-03-008,REV 0 V7ZZ~smIT-I

Figure Withheld Under 10 CFR 2.390 MAD FOM DRAWING NU-380O8, REV.2j 2 ý 0 r~

-A I 187FII

Figure Withheld Under 10 CFR 2.390

'R fCN 721004-321 A

TRANSNUCLEAR APOARLVA COMPANY NUHOMS'0S197L ONSITE TRANSFER CASK

-* UGHT NEUTRON SHIELD ASSEMBLY

"-" -*N...S.. I "'N& 17'-0 2 I 1

Figure Withheld Under 10 CFR 2.390 MADErROMDRAI;NG RV0 2 1

Figure Withheld Under 10 CFR 2.390 F-MDE ROM DR.AWNG LŽ.2 3-009,REV 02

Figure Withheld Under 10 CFR 2.390 NMUDEROM ORARWEING 80 REv 0 2 I 1

Figure Withheld Under 10 CFR 2.390 MADE FROM DRAWING NUH-03-8009, REV 0 2

I 11 T,

i-3-8"-SA* I

Figure Withheld Under 10 CFR 2.390 MAEFROMDRAWtNG7 LUH-3O809. RV 0 2 I

Figure Withheld Under 10 CFR 2.390

"' TRANSNUCLEAR AN AREVA ComPAw o,* NUHOMS8 0S197L ONSITE TRANSFER CASK OSI97L MAIN ASSEMBLY 2I1

Figure Withheld Under 10 CFR 2.390 SECTIONLP-_B rRMA~ RMORAW7NF1-A NUH-03-80 10. REV

FCN 721004-321 Rev. 1 Page 38 of 99 Table W.1-1 Comnparison of Key Parameters of NUHOMS 0S197 Versus 0S197L TOs____

Same?

Characteristic 0S197 TC OS197L TC (Yes/No)

Note No.

Physical Data Outside Diameter 85.50" 80.36" No (1)

Outside Length 207.22" 207.22" Yes Cavity Diameter 68" 68" Yes Cavity Length 197.75" 197.75" Yes Ram Access2222Ye Penetration Diameter2"22Ye 106,670 lbs 57,400 lbs Weight, Empty (includes cask top cover plate assembly (includes cask top cover plate assembly No (2) and neutron shield without water) and neutron shield without water)

Cask Materials ~ <>>> ;K7>$ k ~

Outer Jacket 3/16" thick plate, ASTM A240, Type 304 1/4' thick plate, ASTM A240, Type 304 No (3)

Neutron Shielding 3" of Water inannulus 3"of Water in annulus Yes Structural Shell 1-1/2" thick Type SA-240 plate,304 ASME 2.68" thick plate, SA-240 ASME Type 304 No (4)

Gamma Shielding 3.56" thick, ASTM 829 Chemical Copper No lead shielding No (4)

________________Lead Inne Linr1t"thik plte, SME A-24 Typ 304No separate inner liner N 4 InnerLine 1/2 thik plte, SME A-24 Typ 304 (consists of structural shell) N 4 Consists of 3"thick ASME

' thck SMESA-20, ypeSA-240, Type 304 structural plate awith Consstsof a thick of 3" at A1E/A4' Consistsutua h thick thin 1/4" thick shell encapsulating solid Top Cover Assembly 304l stnctualpslate itha thlin 1/4uthick Neutron Absorbing Material (NS-3) Yes shlnApsulatingMatesoli Neutro During downending inside the fuel Absobin Maeria (N-3) building, an interim aluminum cover may be used to reduce crane loading Top Flange ASME SA-1 82, Type F304N ASME SA-1 82, Type F304N Yes Uppr Lftig Tunnon ASME SA-564, Grade 630 steel trunnion Solid monolithic Trunnion made of ASME N ifingTrnnonwith sleeve encapsulating a solid Neutron Uppr SA-1 82, Type FXM-19 N Absorbing Material (NS-3)

Lower Support ASME SA-240, Type F304 steel trunnion Solid monolithic Trunnion made of ASME Trninwith sleeve encapsulating a solid Neutron SA-1 82, N TrunionAbsorbing Material (NS-3) Type F304 Canister Rails ASTM A240 Nitronic 60 ASTM A240 Nitronic 60 Yes Bottom End Plate 2" thick,Type ASME Yes 304SA-240, 2" thick,Type ASME 304SA-240, Bottom Support Ring, ASME SA-182, Type F304N ASME SA-182, Type F304N Yes PeeRatio Rings ASME SA-1 82, Type F304N ASME SA-1 82, Type F304N Yes Cask Payload ~477 DSC.Type 24P, 528, 61 BT, 24PHB, 24PT2, 32PT J24P, 52B, 6IBT, 24PHB, 24PT2, 32PT Yes Heat Load 24 kW 124 kW J Yes Notes:

1. The diameter of the OS1 97L TC is smaller, reflecting the reduced radial shielding. The 2.68" thick SS structural shell replaces the combined thickness of %4" of inner liner, 3.50" of lead, and 1.50" of structural shell, a reduction of approximately 5.5" diametrical.

2. The reduced weight of the 0S197L TC reflects the reduced radial shielding. Utilizing the yoke (approximate weight 5,000 lbs.) results in an estimated maximum hook weight of 148-149 kips.

3. The outer panel of the neutron shield is increased in thickness to stiffen the assembly.

4. The reduced shielding is a result of the lead shielding that is eliminated and the combined inner liner and structural shell.

NUH-003 Revision 9 Page W. 1-5 January 2( )06

FCN 721004-321 Rev. 1 Page 39 of 99 Table W.1-2 0S197L TC UIFSAR Sections Affected Seq Section/Page Description 0S197L 1 1.1(3)11.1-3 Description of TC for transport of DSC No Change 2 Figures 1.1-2/1.1-6 NUHOMSO System Components including TC See Section W.1 3 Figures 1.1-3/1.1-7 NUHOMS System Components including TC See Section WA1 4 1..3/12-3Description of Operating and Handling Systems Changes addressed in Section 4 12./1.-3including TC W.1.

5 Table 1.2-2/1.2-8 Key Design Parameters for NUHOMS System See Section WA1 6 Table 1.2-3/1.2-9 NUHOMSO System Operations Overview See Section W.8 7 Section 1.3.2.1/1.3-3 Description of On-Site TC See Section WA1 8 Section 1.3.2.2/1.3-4 Description of Transfer Equipment (Trailer and See Section WA Skid) 9 Table 1.3-1/1 .3-10 Components, Structures and Equipment for the See Section W.1 Standardized NUHOMSO System ____________

10 Figure 1.3-6/1.3-18 NUHOMS' On-Site TC See Section W. 1 _____

11 Figure 1.3-8/1.3-20 Cask Support Skid for NUHOMS System See Section W. 1 ______

12 Figure 1.3-10/1.3-22 NUHOMS6 System Operational Overview See Section W. 1 _____

13 2.0 Site Characteristics No Change No change to loading conditions 14 3.1.2.1/3.1-4 On-Site Transfer Cask See Section W.1I for OS I97L

________________________________description.

15 Table 3.1-7/3.1-13 NUHOMSO Transfer Equipment Criteria No Change On-Site Transfer Cask Load Combinations and No Change to load combinations 16 3.2.5.3/3.2-7 StutrlDsg rtraor criteria. See Section W.3 for Strutura CiteraDesgn 01 97L structural results 17 Table 3.2-1/3.2-11 Summary of NUHOMS Component Design No Change

_______________Loadings_______________

18 Table 3.2-7/3.2-20 On-Site Transfer Cask Load Combinations and No Change

______________Service Levels 19 Table 3.2-11/3.2-25 Structural Design Criteria for On-Site Transfer No Change Cask ______________

20 Table 3.2-12/3.2-26 Structural Design Criteria for Bolts No Change 21 3.3.5.2/3.3-31 Radiological Protection-Shielding See Section W.5 22 Table 3.3-1/3.3-36 NUHOMS System Components Important To See Table W.2-1.

______Safety 23 3.4.4.1/3.4-2 Classification of Structures,

______Systems- Transfer Cask andComponents, Yoke and No Change 24 3.4.4.2/3.4-2 Classification of Structures, Components, and No Change

______Systems- Other Transfer Equipment 25 Table 3.4-1/3.4-4 NUH-OMS Major Components and Safety See Table W.2-1

______Classification 26 4.2.1/4.2-1 Storage Structures - Structural Specifications No Change 27 4.2.3.3/4.2-9 Individual Unit Description - On-Site Transfer See Sections WA1 and W.3 for Cask trunnion load test.

28 Figure 4.2-10/4.2-21 Composite View of NUHOMSD Transfer Cask- See Section W.1

_____________24P 29 Figure 4.2-11/4.2-22 Composite View of NUHOMSe Transfer Cask- See Section W.1~

___________ I 52B NUH-003 9 ~Page Revision W. 1-6 Jnay20 January 2006

FCN 721004-321 Rev. I

_______________________Page 40 of 99 Seq Section/Page Description 0S1 97L 30 Figure 4.2-12/4.2-23 NUHOMSO On-Site Transfer Cask with BWR No Change

_________________________Collar 31 Figure 4.2-15a/4.2- NUHOMS 75 Ton Transfer Cask Lifting Yoke 26a I__________________ I No Change Table W.1-2 0S197L TC UFSAR Sections Affected Seq Section/Page Description 0S1 97L 32 4.5/4.5-1 Transfer Maintenance Cask and Lifting Hardware Repair and No Change 33 4.7.3.2/4.7-5 Individual Unit Descriptions - Transfer Cask See Section W.1 34 4.7.3.8/4.7-10 Individual Unit Descriptions - Cask Support See Section WA1 Skid 35 4/491ASME4.9/.9-1Cask Code Exceptions List for the Transfer See Section W.3 36 Table 4.9-1/4.9-3 ASME Code Exceptions List for the Transfer See Section W.3 Cask _____________

37 5.0/5.1-1 Operation Systems See Section W.8 38 6.0 Waste Confinement and Management No Change 39 7.1/7.1-1 Radiation Protection-design Considerations See Section W.5 40 7.3.2.2.F/7.3-6 Transfer Cask Surface Dose Rates See Section W.5 Tables 7.3-2 through 41 7.3-5/7.3-9 through Shielding Analysis Results See Section W.5 7.3-14 __________ __

42 7.4.1/7.4-1 Operational Dose Assessment See Section W.5 43 Tbe74l743 NUHOMSO System Operations - Occupational See Section W.5 Table7.4-11.4-3 Dose Calculations ______________

See:

44 80 Anlysi ofDesin EvntsSection W.3 - Structural 44 80 Anlysi ofDesin EvntsSection WA4 - Thermal Section W. 11 - Accident 45 9.0 Conduct of Operations No Change 46 10.0 Operating Controls and Limits No Change 47 11.0 Quality Assurance No Change 48 Appendix A Details of Shielding Models of the NUHOMS See Section W.5

______ ~~~System _____________

49 ApedxBDetails Appendix BNUHOMS of Heat Transfer Analysis of the System No Change 50 Appendix C.A Deleted No Change 51 Appendix C.2 Transfer Cask Drop Analysis See Section W.3 52 Appendix C.3 Transfer Cask Side Drop Analysis See Section W.3 53 Appendix CAA. DSC Fatigue Evaluation No Change 54 Appendix C.4.2 Transfer Cask Fatigue Evaluation No Change 55 Appendix C.*5 Transfer Cask Structural Analysis NRC See Section W.3 for DBT events

______Question Resolutions 56 Appendix C.6 References No Change 57 Appendix D Review of Concrete Behavior under Sustained No Change

______elevated Temperature 58 Appendix E Drawings See Section W. 1.5 59 ApedxFNUHOMS 24P Topical Report - NRC No Change Appenix FQuestions NUH-003 Page W. 1-7 Reviion92006 Pae W1-7January

FCN 721004-321 Rev. 1

___________________Page 41 of 99 Seq Section/Page Description 0S197L 60 Appendix G Deleted No Change 61 Appendix H NUHOMS 24P - Long Cavity DSC Evaluation No Change for Storing PWR fuel Without BPRA's 62 Appendix I Deleted No Change 63 Appendix J NUHOMS 24P - Long Cavity DSC Evaluation No Change

_______________for Storing PWR fuel With BPRA's ______________

NUH-003 Revision 9 Page W. 1-8 Page

.1-8January 2006

FCN 721004-321 Rev. 1 Page 42 of 99 BOTTOM COVER ASSEMBLY r- RAM ACCESS PENETRATION LIGHT NEUTRON SHIELD ASSEMBLY TOP COVER ASSEMBLY LOWER (SUPPORT)

TRUNN ION CASK DRAIN PORT (NOT VISIBLE, AT BOTTOM IOF CAVITY)

STRUCTURAL SHELL UPPER (LIFTING) TRUNNION INTERIM. TOP COVER (OPTIONAL)

Figure W.1-1 0S197L TC Configuration NUH-003 Revision 9 Page W. 1-9 Page

.1-9January 2006

FCN 721004-321 Rev. 1 Page 43 of 99 ti UPPER CASK SHIELD (SHIELDING BELL) 6" THICK 0S197L CASK

~-LOWER CASK SHIELD 6" THICK F-41 Figure W.1-2 NUHOMSO 0S197L TC System Decontamination Area Shielding NUH-003 Revision 9 Page W.1-10 aury20 January 2006

FCN 721004-321 Rev. 1 Page 44 of 99 5.5". REAR SADDLE SHIELD 5.5" REAR TRUNNION SHIELDING 3-STAGE HYDRAULIC RAM Figure W.1-3 0S197L Transfer Equipment Schematic NUJH-003 Revision 9 Page W.1-1 I Page

.1-11January 2006

FCN 721004-321 Rev. 1 Page 45 of 99 Figure W.1-4 0S197L TC System on Transfer Trailer with Shielding NUH-003 Revision 9 Page W.1-12 Page

.1-12January 2006

FCN 721004-321 Rev. 1 Page 46 of 99 W.2 Princip~al Designi Criteria This section provides the principal design criteria for the NUHOMSO OS197L TC System. The principal design criteria for the NUIIOMSO OS197L TC are the same as the NUHOMSO OS 197 TC as described in Chapter 3. Section W.2.1 presents a general description of the spent fuel to be stored. Section W.2.2 provides the design criteria for environmental conditions and natural phenomena. Section W.2.3 provides a description of the systems which have been designated as important to safety. Section W.2.4 discusses decommissioning considerations. Section W.2.5 summarizes the NUHOMSO OS197L TC design criteria.

W.2.1 Spent Fuel To Be Stored The NUHOMS'O DSCs are designed to handle a total of 24 or 32 PWR fuel assemblies and 52 or 61 BWR fuel assemblies with the same characteristics as those described in Chapter 3 (24P and 52B DSCs) and Appendices K.2 (6113T DSC), L.2 (24PT2 DSC), M.2 (32PT DSC), and N.2 (24PHB DSC).

W.2. 1.1 General Operating Functions No change.

W.2.2 Desigg Criteria for Environmental Conditions and Natural Phenomena The NUHOMS4D OS 197L TC is handled and utilized in the same manner as the existing NUHOMS4D OS 197 TC System. The environmental conditions, natural phenomena and design criteria are the same as described for the NUHOMS(O OS 197 TC in Chapter 3. Design criteria for the NIJHOMSO DSC and HSM remain unchanged.

W.2.2.1 Tornado Wind and Tornado Missiles No change.

W.2.2.2 Water Level (Elood) Design No change.

W.2.2.3 Seismic Design No change.

W.2.2.4 Snow and Ice Loading No change.

W.2.2.5 Combined Load Criteria No change.

NUH--003 Revision 9 Paue W.2-1 January 2006

FCN 721004-321 Rev. 1 Page 47 of 99 W.2.3 Safet Protection Systems W.2.3. 1 General Table W.2-1 provides the safety classification of the 0S197L TC system components.

W.2.3.2 Protection By Multiple Confinement Barriers and Systems No change.

W.2.3.3 Protection By Equipment and Instrumentation Selection No change.

W.2.3.4 Nuclear Criticality Safet W.2.3.4.1 Control Methods for Prevention of Criticality No change.

W.2.3.4.2 Error Contingency Criteria No change.

W.2.3 .4.3 Verification Analysis-Benchmarking No change.

W.2.3.5 Radiological Protection No change.

W.2.3.6 Fire and Explosion Protection No change.

W.2.4 Decommissioning Considerations No change.

W.2.5 Summarya of NUIIOMSe 0S197L TC Design Criteria The principal design criteria for the NUHOMSO OS 197L TC are the same as those presented for the NUHOMSO OS 197 TC in Chapter 3. The NUHOMSO OS197L TC is designed to handle a DSC loaded with PWR or BWR fuel assemblies identical to those stored in a NUHOMSO OS 197 TC as described in Chapter 3 and Appendices Chapters K.2, L.2, M.2 and N.2.

NUH-003 Revision 9 Page W.2-2 January 2006

FCN 721004-321 Rev. 1 Page 48 of 99 Table W.2-1 0S197L TC System Components and Safety Classification 0S197L TC System Components Safety Classification Onsite Transfer Cask

- Structural Shell and Cover Plates Important to Safety(')

- Upper and Lower Trunnions Important to Safety(')

- Decontamination Area Shield Not Important to Safety(')

- Trailer Shielding Important to Safety~l)

Transfer Equipment 2

- Cask Lifting Yoke Safety Related ()

- Transport Trailer/Skid Not Important to Safety(')

-Ram Assembly Not Important to Safetyý')

- Dry Film Lubricant Not Important to Safety(T)

Notes:

(1) Structures, systems and components 'important to safety" are defined in 10CFR 72.3 as those features of the ISFSI whose function is (1)to maintain the conditions required to store spent fuel safety, (2) to prevent damage to the spent fuel container during handling and storage, or (3) to provide reasonable assurance that spent fuel can be received, handled, packaged, stored, and retrieved without undue risk to the health and safety of the public.

(2) Yoke and rigid or sling lifting members are classified as "Safety Related" inaccordance with 1OCIFR50.

NUH-003 Revision 9 Page W.2-3 Page 2006

.2-3January

FCN 721004-321 Rev. I Page 49 of 99 W.3 Structural Evaluation This section describes the structural evaluation of the NUHOMSO OS197L Transfer Cask (TC).

The OS 197L TC is a modified version of the OS1I97/0S 197H TCs (henceforth referred as the OS 197 TC) designed to enable "under-the-hook" lift weights of 75 tons. The 051 97L TC may be used for transfer of loaded DSCs currently licensed under CoC 1004 (24P, 52B, 6113T, 24PT2, 32PT and 24PHB) [3.11]. The structural evaluation for the OS 197L TC is based on the OS 197 TC evaluations documented in Chapter 8, and additional evaluations as described in Appendices K, L, M and N for payloads associated with the 6 1BT, 24PT2, 32PT and 24PHB DSCs, respectively. The additional evaluations provided in this section address specific design differences between the OS1I97L TC and the OS 197 TC.

W.3.1 OS197L TC Description The specific design differences in the OS 197L TC relative to OS 197 TC are summarized below:

The 1.5" thick structural shell and the 0.5" thick inner liner (both SA-240 stainless steel) are replaced with a single thicker 2.68" thick shell of the same material. This represents an increase in the TC shell structural capacity relative to the OS 197 TC.

The encapsulated 3.56" thick lead thickness in the OS 197 TC is eliminated to achieve the desired weight reduction.

A neutron shield assembly is provided with the inner and outer shells made from '/"

thick plate material instead of a neutron shield assembly that is integral to the structural shell on the inside and a 3/16" thick outer shell. The neutron shield materials (type 304),

total annulus water thickness of 3"and the configuration of the internal stiffening elements remain unchanged.

The two-piece upper trunnions assemblies made from SA-564 Type 630 steel trunnion and welded into a forged Type 304 steel trunnion sleeve with encapsulated NS-3 for the OS 197 TC are replaced with one solid trunnion design made from SA- 182 Type FXM- 19 stainless steel. This modified trunnion design results in a stronger trunnion as it eliminates the 5A564, Type 630 to SA 240, Type 304 weld.

The two-piece lower trunnions made from Type 304 stainless with encapsulated NS-3 are replaced with solid Type 304 forgings.

Specific evaluations are performed to address the modified OS1I97L TC trunnion configuration.

The evaluations also address the effect on local shell stresses. Thermal stresses of the cask are also evaluated. All other structural analyses for the OS 197 TC bound the OS197L TC because the cask structural shell capacity of the OS I97L TC is higher than that provided by the OS 197 and the top and bottom forging assemblies are unchanged.

NUJH-003 Revision 9 Page W.3-1 January 2006

FCN 721004-321 Rev. 1 Page 50 of 99 W.3.2 Design Criteria The structural design criteria for the 05 197L TC are the same as that applicable to the OS 197 TC as summarized in Chapter 3. Similar to the OS 197 TC, the OS 197L TC is designed to meet the stress allowables of the ASME Code [3.2] Subsection NC for Class 2 components. The OS 197 TC criteria summarized in Table 3.2-1 (component design loadings, as applicable), Table 3.2-7 (load combinations), Table 3.2-11 (stress criteria) and Table 3.2-12 (bolts design criteria) are applicable to the OS I97L TC. The OS 197 TC ASME Code exceptions described in Table 4.9-1 is also applicable to the OS 197L TC.

The test load criteria for the upper trunnions of the 05 197L TC are the same as described in Section 4.2.3.3, except that the test load is conservatively equal to 300% of the design load (instead of 150% for the OS 197 TC).

W.3.3 OS 197L TC Weigh The dry weight of the 0S197L TC is presented in Table W.3-1. The total weight of the cask, including neutron shield water, is approximately 62,000 lbs. This compares with the corresponding weight of 111,250 lbs for the OS 197 TC. To provide flexibility during transfer from the decontamination area to the trailer, a 1"thick aluminum cask top lid that weights approximately 500 lbs may be used in lieu of the stainless cask top lid.

The OS5I97L TC weights as described in Table W.3-1 are to be used in conjunction with the payload weights for the various DSCs as described in the applicable sections in Chapter 8 (Tables 8.1-4 and 8.1-5 for the 24P and 52B DSCs), and Appendices K.3, L.3, M.3 and N.3.

Each specific user is to evaluate the total under-the-hook lift weights against plant specific crane capacity limits in accordance with the requirements of 10CFR7 1.212.

W.3 .4 Mechanical Properties of Materials The materials properties for the OS I97L TC are specified in Section 8. 1, Table 8.1-3.

W.3.5 General Standards for Casks The OS1 97L is fabricated using the same materials as the OS 197 TC. Thus, there are no changes to the documentation in Chapter 4 and Appendices K.3, L.3, M.3 and N.3 relative to chemical and galvanic reactions.

The evaluation of the OS 197L TC is based on critical lift weights of 250,000 lbs.

The thermal analysis of the OS 197L along with a summary of the effect on pressures and temperatures is described in Section WA.4 NUH-003 Revision 9 Pane W.3-2 January 2006

FCN 721004-321 Rev. 1 Page 51 of 99 W.3 .6 Normal/Off-Normal Structural Evaluation W.3.6.1 Evaluation of the One-Piece OS1I97L Trunnions As discussed above, the OS 197L TC upper trunnions consist of one piece solid trunnion forgings made from SA- 182 Type FXM- 19 stainless steel which results in a stronger trunnion design as it eliminates the SA564 (trunnion) to Type 304 (trunnion sleeve) weld and associated inconel weld.

Loads considered in the stress evaluation include lifting, transfer handling, HSM loading/unloading and seismic. The trunnions are evaluated for a maximum TC loaded weight during lifting and handling of 125 tons. For critical lifts, the maximum TC loaded weight is increased by a factor of 1.15. This results in critical lift load of 144 kips/trunnion. The trunnion evaluations are performed using hand calculations and applying the ANSI N14.6 [3.3] design criteria, including load testing. In addition, an ANSYS model of the cask, including the upper and lower trunnions, is developed to determine cask shell stresses at the trunnion-shell interface as well as within 3" to 4" away from the trunnions. These stresses are evaluated against the ASME stress criteria in Table 3.2-11. The ANSYS model of OSl197L TC is shown in Figure W.3-1 and Figure W.3-2.

In addition, a thermal stress analysis of the OS 197L TC with the trunnions is performed. The stresses obtained from the thermal stress analysis are combined with the mechanical stresses to determine total stresses at and near (within 3"to 4") of the trunnion-structural shell interface.

The stress distribution in the region of the upper trunnion is shown in Figure W.3-3. The structural analyses results of the 0S197L TC trunnions are summarized in Table W.3-2 through Table W.3-4. The maximum stress ratio is 0.74, therefore the 0S197L TC trunnion configuration has significant margin with respect to ASME/ANSI N14.6 code allowables.

W.3.6.2 Thermal Stress Analysis of OS197L A conservative thermal profile was developed for the purpose of calculating bounding thermal stresses in the OS 197L TC components. The thermal stress analysis is performed using the same three-dimensional ANSYS model shown in Figure W.3-1. The calculated maximum thermal stress intensity in the OS 197L is 17.4 ksi and occurs in the structural shell away from the trunnion region. Conservatively combining this thermal stress intensity with the maximum mechanical stresses which occur in the trunnion region results in an enveloping (structural shell top and bottom forgings) combined stress intensity for normal and off-normal conditions of 37.5 ksi, which is well below the allowable stress intensity for primary plus secondary stress of 56.1 ksi at temperature.

W.3.7 Applicability of OS 197 TC Accident Drop Evaluations to the 0S197L TC The fully loaded weight of the OS197L TC is bounded by the OS 197 TC loaded weight.

Therefore, an evaluation was performed to determine if the bounding accelerations used for the postulated accident drop evaluations of the OS 197 TC remain applicable to the 0S197L TC.

NUH-003 Revision 9 Page W.3-3 January 2006

FCN 721004-321 Rev. 1 Page 52 of 99 As reported in Section 8.2.5.1C, the g loads for the OS 197 TC were determined to be 59 g for the end drop, 49 g for the side drop and 25 g for a comner drop. Based on these accelerations, bounding accelerations of 75g for the horizontal (side) and vertical drops and 25g for the corner drop were used for the OS 197 TC drop evaluations. The OS 197 TC evaluations are documented in Chapter 8. Using the same methodology as that described in Section 8.2.5. 1C for the OS 197 TC, the equivalent loads for the 0S197L TC are 75 g for an end drop, 61 g for a side drop and 25 g for a corner drop. Therefore, the 75g accident drop evaluation results for the side and end drops and the 25g evaluations for the comner drop performed for the OS 197 TC and reported in Section 8.2 remain bounding and are applicable to the OS1I97L TC. These g-loads are conservative with respect to shell stresses since the thicker OS 197L TC shell has a higher load capacity than the OS 197 TC shell configuration. Hence, all the cask accident drop results reported in Section 8.2, and Appendices K.3, L.3, M.3 and N.3 remain bounding and, thus, are not affected.

W.3.8 Effect of Increased OS1I97L Temperatures on DSC Shell and Basket Compvonents Table W.4-4 shows maximum temperature increases for the various DSCs allowed for transfer in the 0S197L TC (24P, 52B, 61BT, 24PT2, 32PT and 24PHB). The maximum temperature increase applicable to the DSC shell and basket internals is approximately 13'F for normal conditions, 220 F7 for off-normal and less than 18*F for accident conditions. This magnitude increase will not appreciably affect the material properties or the allowables used for the evaluation of these DSCs as documented in Chapter 8 and Appendices K.3, L.3, M.3 and N.3.

Therefore, the documented results of these DSCs are not affected due to use of the 0S197L TC.

W.3.9 References

[3.1] NUHOMSO Certificate of Compliance for Dry Spent Fuel Storage Casks, Amendment No. 8, December 2005, Docket No. 72-1004.

[3.2] American Society of Mechanical Engineers, ASME Boiler and Pressure Vessel Code,Section III, Division 1, 1983 Edition with Winter 1985 Addenda.

[3.3] American National Standard, "For Radioactive Materials - Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds (4500 Kg) or More," ANSI N14.6-1986, American National Standards Institute, Inc. New York, New York (1993)

NUH-003 Revision 9 Paae W.3-4 January 2006

FCN 721004-321 Rev. 1 Page 53 of 99 Table W.3-1 Summary of 0S197L TC Weights Item Weight (Ibs) 0S197 Maximum Dry Weight including Neutron 5,0 Shield Assembly and Top 5,0 Cask Lid Neutron Shield Water 4,600 Top Cask Lid 5,150 NUH-003 Revision 9 Page W.3-5 Page

.3-5January 2006

FCN 721004-321 Rev. 1 Page 54 of 99 Table W.3-2 Summary of Maximum Stress Ratios for Critical Lifts Enveloping Stress Ratios Critical Lilt Load Combinations T-PPm ICalculated IAllowable IRatio

+ Pb Calculated Allowable Ratio ff Pm + Pb +0Q Calculated Allowable Ratio Notes Cask Shell, ANSYS Evaluations I ASME Criteria near Trunnion(s) t! I 4.21 ksl I18.7 ksi I0.23 16.6 ksi 28.1 ksi 0.59 28.1 ksi 56.10 0.50 400OF Level A Trunnion Evaluations, Hand Calculations (ASME-Lower; ANSI N14.6-Upper)

(A1/A2/A3) JPM Pm + Pb Pm+ Pb + Q ASMVE Criteria L ower Trunnion 2.55 ksi 120.3 ksi 10.13 14.81 ksi 130.5 ksiT 0.16 n/a nela n/a Type F304N Upe TrnnonShear

______ Stress Normal Stress N14.6 Criteria

[-U___ er Trunnion__ 2.86 ksi 4.07 ksi 10.70 15.01 ksi 6.78 ksi 10.74 n/a n/a In/a Type FXM-19 Table W.3-3 Summary of Maximum Stress Ratios for Level A Load Combinations Enveloping Stress Ratios - Leve A (non-Critical Lilt) Combinations____________ _____

Enveloping Stress Ratios _ __ Pm __ __ Pm + Pb Pm + Pb + Qa __

Non-Critical Level A Comb.- Calculated Allowable Ratio calculated Allowable Ratio Calculated Allowable Ratio Notes Shell at Upper Trunnion/Lower Trunnion 6.05 ksi 18.7 ksl 0.32 I

I-20.1 ksi 28.1 ksi 0.72 35.8 ksi 56.1 ksi 0.84 ANSYS Level A (A4)A5)

Shell near Upper Trunnion/Lower Trunnion") 4.44 ksi 118.7 kal 10.24 1.___________________

114.7 ksi 28.1 ksi 0.56 il27.4 ksi 56.1 ksi 10.49 Analysis Upper Trunnion _____ii ____ I ____ ____ ___

n/a Hn 2.96 ksi 45.3 ksi 0.07 Wea n/a r (FXM-1 9) 2.51 ksi 30.2 ksi 0.08 LWer Trunnion (OType F304N) 2.41 kal 20.3 ksi 0.12 4.55 ksi 30.5 ksi 0.15 Wea n/a n/a Cluain Max: 0.32 Max: 0.72 Max: 0.64 Table W.3-4 Summary of Maximum Stress Ratios for Level C Load Combinations Enveloping Stress Ratios ____ Pm ____Pm + Pb _ __Pm + Pb + 0 For Level C Combinations calculated Allowable Ratio calculated Allowablel Ratio CalculatedT Alowable IRatio Notes Shell at Upper Trunnion/Lower Trunnion 5.12 kal 18.7 ksi 0.27 17.3 kal 28.1 kal 0.62 not re uired for Level C ANSYS Shell near Upper Level C Trunnion/Lower (Cl/C2) Trunnion1l)

Upper Trunnion 3.23 ksil

____J_______

18.7 ksi 10.17 f________I l 111.9 kal 28.1 kal 0.43 n/a

____I_______

n/a n/a n/a IAnalysis Hand 2.33 ksi 54.4 ksi 0.04 n/a n/a (FXM-19) 1.44 ksi 36.2 ksl 0.04 (Type F304N) 1.19 ksi 24.4 ksi 0.05 12.25 ksi 136.5 ksi 0.06 n/a n/a n/a [Calculationsa Max: 1 0.27 Max: 1 0.62 Note: (1) 4" from upper trunnion/shell interface and 3"from lower trunnion/shell interface.

NUH-003 Revision 9 Page W.3-6 January 2006

FCN 721004-321 Rev. 1 Page 55 of 99 ANSY 8.1

'EB 23 2006 411,34:\50

)V =.58512 TV =.18969 ZY -- .78845:

tb$ST:85 tXF~ -21.e6 tIZr =99.515

ZtBUFFER d2)

Note:

Material properties were assigned as follows:

Purple = SA-1 82 Type F304N (forgings)

Gray = SA-240 Type 304 Blue= SA-1 82 Type FXM-1 9 (Type XMV-11 9 Forging)

Figure W.3-1 0S197L TC ANSYS Stress Analysis Model NUH-003 Page W.3-7 2006 Reviion9 Pae W3-7January

FCN 721004-321 Rev. 1 Page 56 of 99 Upper Trunnion Lower Trunnion Figure W.3-2 0S197L TC ANSYS Analysis Model - Upper and Lower Trunnions Detail NUH-003 Revision 9 Page W.3-8 Jan iuaty 2006

FCN 721004-321 Rev. 1 Page 57 of 99 Figure W.3-3 ANSYS Model Stress Analysis Results - Upper Trunnion Region NUH-003 Revision 9 Page W.3-9 Page

.3-9January 2006

FCN 721004-321 Rev. 1 Page 58 of 99 W.4 Thermal Evaluation W.4. 1 Discussion This chapter documents the thermal evaluation of the OS 197L TC for the loading and transfer of the DSCs currently licensed under CoC 1004 (52B3, 24P, 6 1BT, 24PT2, 32PT and 24PHB).

The OS1I97L TC is a modified version of the OS1I97/OS I97H TCs (henceforth referred to as OS 197 TC) designed to allow use with crane load limit of 75 tons. From a thermal analyses perspective, the following relevant modifications are implemented in the OS I97L TC relative to the OS 197 TCs:

" The 1.5" thick structural shell, the encapsulated 3.56" thick lead and the 0.5" thick inner liner (both SA-240, Type 304 stainless steel) in the OS 197 are replaced with a single 2.68" thick shell made of SA-240, Type 304 stainless steel material.

" The neutron assembly that is integral to the structural shell on the inside and includes a 3/16" thick outer shell in the OS 197 TC is replaced with a neutron shield assembly consisting of inner and outer shells made from 1/4/" thick plate material in the OS5197L TC.

The neutron shield materials, total water annulus thickness of 3" and the configuration of the internal stiffening elements remain unchanged.

" Supplemental shielding is used around the OS 197L TC as part of the OS1I97L TC system, when the TC is in the vertical orientation in the decontamination area and when the TC is in the horizontal orientation on the transfer trailer/skid.

The OS I97L TC shielding system in the decontamination area consists of a two-part, 6" thick cylindrical shaped upper and lower shields made from A-36 steel with rectangular openings at the top and at the bottom that allow free convection boundary layer development along the DSC shell.

The OS 197L TC shielding system on the transfer skid consists of a series of plates that are attached to the sides and ends of the transfer skid. Two upper sections fit like a clamshell over the cask and skid after the cask is placed on the transfer skid. Clearances provided at the support legs of the skid and other openings at the ends of the skid permit cooling airflow to enter the enclosure and pass around the enclosed cask and exit via a long slot opening at the top of the upper sections of the shielding.

The OS 197 TC and the bare OS197L TC have comparable radial thermal resistances (the thermal resistance provided by the lead to shell gaps in the OS 197 TC is compensated by the gap between the neutron shield assembly and the structural shell in the OS197L TC).

W.4.2 Summar of Thermal Properties of Materials The thermal properties of the materials used in the thermal evaluation are the same as those specified in Appendix M, Section M.4.2. Effective thermal conductivity of water and air-filled neutron shield with axial stiffeners are listed in Appendix M, Section M.4.9.

NUIH-003 Revision 9 Paue W.4-1 January 2006

FCN 721004-321 Rev. 1 Page 59 of 99 W.4.3 Specifications for Components Mechanical properties of the materials are the same as those described in Section 8. 1, Table 8.1-3.

W.4.4 Thermal AnaUlsis of an 0S197L TC Containing a 32PT DSC with 24 kW Heat Load A two-dimensional model of the 0S197L TC and 32PT DSC shell with 24 kW heat load was developed using ANSYS to provide temperature distributions for the TC and DSC shell. The 2D) model considers the hottest cross-section of the fuel and conservatively neglects heat transfer in the axial direction. The TC thermal model and analysis methodology are consistent with the methodology described in Appendix M.4.4 for the OS 197 TC with a 32PT DSC payload with changes implemented to account for the configuration changes in the OS 197L TC relative to the OS 197 TC. The OS I97L TC model is shown in Figure W.4-1 and Figure W.4-2.

The heat flux applied to the OS 197 TC model is identical to that described in Appendix M.4.4.1.6 for OS 197 TC with a 32PT DSC.

The following table summarizes the OS 197L TC dimensions used in the thermal analyses.

Dimensions used in 0S197L TC ANSYS Thermal Analysis Model DSCICask Component 0S197L DSC Shell Outside Diameter, in. 67.17 DSC Shell Thickness, in. 0.5 Cask Inner Radius, in. 34.0 Structural Shell Thickness, in. 2.68 Structural Shell Outside Radius, in. 36.68 Structural Shell-Neutron Shield Inner Panel Gap, in. 0.03 Neutron Shield Inside Radius, in. 36.71 Neutron Shield Inner Panel Thickness, in. 0.25 Neutron Shield Inner Radius, in. 36.96 Neutron Shield Thickness, in. 3.0 Neutron Shield Outside Radius/Neutron Shield 39.96 Outer Panel Inner Radius, in.

Neutron Shield Outer Panel Thickness, in. 0.25 Neutron Shield Outer Panel Outside Radius, in. 40.21 The results of the thermal analysis of the 0S197L TC with a 32PT DSC payload but without the supplemental skid shielding are summarized in Table W.4- 1. A plot of the OS 197L TC model temperature distribution for the 100*F ambient condition with insolation is shown in Figure W.4-3. The temperatures incorporating the effect of the skid shielding are discussed in Section W.4.4.3.

NUH-003 Revision 9 PNee W.4-2 January 200i 6

FCN 721004-321 Rev. 1 Page 60 of 99 The thermal performance of the OS 197L TC (without supplemental skid shielding) is very similar to the OS 197/OS 197H as demonstrated by a comparison of the maximum 32PT DSC shell temperatures in Table W.4-2.

W.4.4.1 Effect of the Decontamination Shield on OS I97L TC Thermal Performance An evaluation is performed using manual calculations to confirm that the radial gap between the OS1I97L TC and the inner diameter of the decontamination shield is sufficiently large, and that the size of the top and bottom cut out openings are of sufficient size as to not adversely affect the thermal performance of the OS197L TC. The evaluation is based on analysis of the free convection turbulent boundary layer development along the outer 0S197L TC surface during vacuum drying and helium backfilling operations. The results of the evaluation confirm that the DSC shell-decontamination shield gap and the area of the inlet and outlet openings are adequate and, thus, the decontamination area shield does not adversely impact the cask boundary conditions assumed in the thermal analysis.

W.4.4.2 Effect of Draining Neutron Shield during Transfer from Decon Area to Trailer To reduce cask weight during transfer from the decon area to the trailer, the neutron shield may be drained. To maintain DSC shell temperatures within previously analyzed conditions for the DSC vacuum drying, backfilling, and welding operations, the DSC/cask annulus is maintained full during this transfer. The DSC/annulus is maintained at atmospheric pressure by venting.

W.4.4.3 Effect of the Supplemental Skid Shielding on OS197L TC Thermal Performance As discussed above, supplemental shielding is installed on the OS 197L skid to compensate for the reduced shielding capability of the bare OS 197L TC. While the skid shielding prevents direct insolation heating of the cask surface, it also affects the convection and radiation heat transfer from the cask. To address the effect of the supplemental skid shielding on the thermal performance of the OS 197L TC with a 32PT DSC payload, a computational fluid dynamics (CFD) analysis is performed for the OS 197L TC with the supplemental shielding.

The FLUENT Code CFD methodology is used to define the boundary condition temperatures for the OS197L TC on the transfer trailer. The use of FLUENT is appropriate to model the natural convection driven air flow inside the trailer shielding as it flows around the cask.

Since the OS I97L TC is provided with side and top shielding, the cask boundary sees a temperature which exceeds the ambient temperature. To define an appropriate boundary condition for the cask, the CFD analysis evaluated a prototypic segment of the cask and skid geometry, including the air flow paths into the cask/skid shielding enclosure. The shielding enclosure is provided with openings between the skid beams and the trailer deck to allow air to enter the enclosure. Air exits the enclosure through an opening at the top of the enclosure.

The analysis of the OS 197L TC employed the following:

1) The analysis involved the passive cooling of a cylindrical, heat dissipating body housed within an enclosure wherein the cooling airflow enters at the bottom of the enclosure and exits at the top.

NUH-003 Revision 9 Paae W.4-3 January 2006

FCN 721004-321 Rev. 1 Page 61 of 99

2) The analysis used the FLUENT CFD Code Version 6.2 [4.1].

3) The analysis employed second order discretization solution schemes for energy, momentum, and turbulence.

Table W.4-3 summarizes the thermal analysis results of the 0S197L TC with a 32PT DSC payload, including the effect of the skid shielding.

W.4.4. Evaluation of DSC (24P. 52B. 6lBT. 24PT2 and 24PHB DSCs) Shell Temperatures inside OS 197L TC The 32PT DSC shell temperatures when transferred inside the OS 197L TC are determined as discussed in Section W.4.4.2. These results are used as discussed below to estimate the DSC shell temperatures for all licensed DSCs with *ý24.0 kW heat load inside the 0S197L TC. This approach is conservative since the analyzed 32PT DSC has the shortest DSC cavity and thus the highest heat flux for a given heat load.

The DSC shell temperatures determined in Sections 8.1, 8.2, Appendices K.4, LA4 and NA4 for normal, off-normal and accident conditions are applicable for transfer inside OS 197 TC. These average DSC temperatures are increased by the same ratio as the 32PT DSC for transfer inside the OS 197L TC with allowance for specific canister heat loads and ambient temperatures as applicable. The maximum temperature rise of the DSC components is conservatively assumed equal to the corresponding average DSC shell temperature increase.

These results are summarized in Table W.4-4.

W.4.4.5 Evaluation of Maximum Fuel Cladding Temperatures inside OS 197L TC To calculate the maximum fuel cladding temperature for each DSC when transferred inside the OS 197L TC, a methodology identical to that described above in Section W.4.4.3 above is used.

A summary of the maximum fuel cladding and DSC component temperatures evaluated for the various DSC types allowed for transfer in the OS1I97L TC is shown in Table W.4-4. In all cases, calculated fuiel cladding temperatures are less than the regulatory limit.

W.4.4.6 Evaluation of DSC Internal Pressures The maximum DSC internal pressures during transfer in the OS197L are calculated based on results for the DSC in OS 197 TC. The increase is based on the change in the helium backfill temperature due to the increased temperature of the cask, due to the skid supplemental shielding.

This temperature increase is conservatively assumed equal to DSC component temperature increase shown in Table W.4-4.

The internal pressure for each DSC type is calculated as PDSC in0S197L = PDSC in05197TTDSC He av inOS 197L/TDSC He ,vin0S197, NUH-003 Revision 9 Panye W.4-4 January 2006

FCN 721004-321 Rev. 1 Page 62 of 99 where THe avin05197, OR - average helium temperature in DSC during transfer in 0S197 TC, THe avin05197L, OR - average helium temperature in DSC during transfer in 0S197L TC, THeav in OS I97L = THe v inOS 197 + ATHeav.

Table W.4-5 lists maximum DSC internal pressure for each DSC authorized for transfer within the OS 197L TC in comparison with the design pressures.

As seen from Table W.4-5, the DSC internal pressure during transfer in OS 197L TC is within design limits for all the DSCs.

Therefore, all DSCs considered in Table W.4-5 are qualified for transfer in OS 197L TCs.

W.4.4.7 Thermal Performance of OSlI97L TC during Fire Accident Conditions The fire transient analysis is performed for the OS 197L TC on the transfer trailer, using the same assumptions as for the OS 197 TC fire accident analysis as described in Appendices K.4, L.4, MA4 and NA4 The calculated maximum temperatures are summarized in Table W.4-6. No credit was taken for the trailer shielding in this analysis. This is conservative because the shielding on the trailer will not, in reality, be impacted by the fire condition.

W.4.4.8 References

[4.1] FLUEN1~m CFD Code Version 6.2, Fluent Inc., 10 Cavendish Court, Lebanon, NH 03766.

NUH-003 Revision 9 Paue W.4-5 January 2006

FCN 721004-321 Rev. 1 Page 63 of 99 Table W.4-1 32PT DSC Shell and 0S197L TC Component Maximum Temperatures without Supplemental Shielding Operating Conditions OFmb OF OFxpOFN5pmx OF Thm 100 309 280 239 431 Normal, transfer 100, insolation~l) 311 281 258 448 Off-normal, transfer 117 314 285 244 435 Accident, transfer (Loss of supplemental 117, 573 548 265 607 skid shielding and liquid insolation neutron shield) I_ _ __ _ _ I _ _ __ _ __ _ I__ _ _ I__

where:

Tstrah max - cask structural shell maximum temperature, TinNS p - cask inner NS panel maximum temperature, Tout NS p max cask outer NS panel maximum temperature, Tah max - DSC shell maximum temperature.

Notes:

1. Temperature values with insolation are provided for comparison with OS1 97 TC cases.

2. Values shown are for 32PT DSC shell. For increases inmaximum shell temperature for all DSCs, see Table W.4-4 NUH-003 Revision 9 Page W.4-6 January 200 6

FCN 721004-321 Rev. 1 Page 64 of 99 Table W.4-2 32PT DSC Shell Maximum Temperatures for 0S197L (without Supplemental Skid Shielding) and 0S197 TCs Operating Conditions T.~b, OS197L 0S197 OF Tah maxy Tah max, Normal, transfer inslaio 448 445 Off-normal, transfer 117 435 433 Accident, transfer 17 (Loss of sun shade and 117,ton 67 0

,liquid neutron shield) Inslto 60I0 where:

T~h ,ax - DSC shell maximum temperature.

NUH-003 Revision 9 Page W.4-7 Page

.4-7January 2006

FCN 721004-321 Rev. 1 Page 65 of 99 Table W.4-3 32PT DSC Shell and 0S197L TC Component Maximum Temperatures (Supplemental Skid Shielding Effect Included)

Operating Conditions Tamb, T.s,. ah max, Tinn NS pmax, T0~t NS p maxax Normal, transfer 100 353 324 283 444 Off-normal, transfer 117 358 329 288 454 Accident, transfer (Loss of supplemental 117, 53548 265 607 skidd shielding and liquid insolation neutron shield)_______ ______

where:

Tatrsa, ma - cask structural shell maximum temperature, Tinn NS p, - cask inner NS panel maximum temperature, TOUR NS p max- cask outer NS panel maximum temperature, Tahmax - DSC shell maximum temperature.

Notes:

1. Values shown are for 32PT DSC shell. For the maximum shell temperature of all DSCs allowed for transfer in OS1I97L, see Table W.4-4.

NUH-003 Revision 9 Paae W.4-8 January 20i06 06

FCN 721004-321 Rev. 1 Page 66 of 99 Table W.4-4 Summary of Maximum Fuel Cladding and DSC Component Temperature Increase during Transfer in 0S197L TC DSC Type I24P I52B I24PT2SIL 161 BT I24PHB I32PT Heat load, kW 24 1124 24 W 24 24 Tamb 100 100 100 100 100 100 4

~ATosc compmax(') 13 10(4) 13 10( ) 13 13

&Tfuq ~(2) 8 7 7 10 4 9 TWOetmax1 743 1 769 1 735 648 1 726 1 729 Ttueiiw 1058 1058 1058 1058 752 752 Tamb 125 125 125 125 117 117

1) 22(5) 18(4.6) 22(5)

S ATDSC compna,( 22(s) 19(4,S) 19 S ATtuel max( 2) 14 12 12 15 7 15 TWOeimax 749 774 740 655 729 730 Tfe mt 1058 1058 1058 1058 1058 752 Twnb 125 125 125 125 117 117 ATosc comipmax(l) 18(4) 16(4) 18(4) 16(4) 18(4) 10

~, ATIuemax( 2) 11 10 e118 15 6 10 Ot Thw max 1 882(6) 915(6) 798 824(6) 744 873 I___________1 1058 1058 1058 1058 1058 1058 Notes:

1. Increase in DSC shell temperature in0S197L TC relative to the 0S197 TC.

Conservatively, it isassumed that the temperature for the DSC internal basket components increases by the same delta. The delta (increase) isevaluated as: AToSc ax = ATshav= T,,OS197L - TalVOS1g7.

2. ATtuew max = Trusi max Dsc in0S197L - Twte max Dsc in0S197

3. Not used.

4. AT~u m-. account for Tomb and heat load difference

5. Adjustment to 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months average ambient temperature difference applied.

6. Accident storage case isconservatively used for evaluation.

NUH-003 Revision 9 Page W.4-9 Page

.4-9January 2006

FCN 721004-321 Rev. 1 Page 67 of 99 Table W.4-5 Maximum DSC Internal Pressure during Transfer in 0S197L TC DSC Type 7 24P I 52B I24PT2S/L I61BT I24PHB I 32PT Heat load, kW 24 19.24 24 18.3 24 24 PDSC,PSig 7.4 1 6.5 8.2 9.3 1 6.6 4.5 S PDSCI!mtt,PSig 10.0 j 10.0 10.0 10.0 15.0 15.0 S Posc,pSig 10.9(1) 8.5 10.6(1) 12.1 11.8 15.5

Po~sc,psig :7.2 31.7 58.4 46.9 64.4 105.0 IK PoSCiiMjtpsi 10.0 100 10.0 20.0 20.0 205.0 Note:

1. The maximum calculated pressure exceeds the design pressure. This isacceptable based on paragraph NB-3223(a) of ASME Section 111, Subsection NB. See Tables 8.1-6 and J.4-2 for the 24P DSC and Section L.4.4.4.8 for the 24PT2 DSC.

NUH-003 Revision 9 Page W.4-10 Page.4-1OJanuary 2006

FCN 721004-321 Rev. 1 Page 68 of 99 Table W.4-6 Maximum Component Temperatures for 0S197L TC during Fire Accident Maximum Temperature at Allowable Temperature, Component 2000 Min,O DSC Shell 571*

Cask Structural Shell f506*

The components perform their intended safety function within the operating range.

Bounded by the accident transfer (loss of sunshade and liquid neutron shield) case.

NUH-003 Revision 9 Page W.4-11 January 2006

FCN 721004-321 Rev. 1 Page 69 of 99 QS197L ITC Figure W.4-1 0S197L TC ANSYS Model NUH-003 9 ~Page Revision W.4-12 Jnay20 January 2006

FCN 721004-321 Rev. 1 Page 70 of 99 Figure W.4-2 Details of 0S197L TC ANSYS Model NUH-003 Revision 9 Page W.4-13 Jnay20 January 2006

FCN 721004-321 Rev. 1 Page 71 of 99 Figure W.4-3 Temperature Plot for 32PT DSC (24 kW) in 0S197L TC without Supplemental Shielding, Tamb=100 0 F, Insolation NUJH-003 Revision 9 Page W.4-14 Jnay20 January 2006

FCN 721004-321 Rev. 1 Page 72 of 99 W.5 Shielding Evaluation This Appendix presents the shielding evaluation of the OS 197L TC when used for fuel loading and transfer of the DSCs currently licensed under CoC 1004 (52B, 24P, 6 1BT, 24PT2, 32PT and 24PHB).

The shielding analysis is performed for the 32PT DSC design basis source terms. The results for normal operations demonstrate that exposures for OS 197L TC activities with operational personnel present are bounded by OS 197 TC exposures (remote crane operation is used and no personnel are present while the cask is on the crane hook).

W.5.1 Methodology Two radiation transport codes are used in the shielding analysis performed in the OS 197L TC calculation: ANISN [5. 1] and MCNP 5 [5.2]. ANISN is primarily used for the scoping analysis to determine (a) spectral distribution on the side of the transfer cask, (b) bum-up, enrichment, cooling time combination(s) which would result in the highest dose rate, and (c) optimum layout of shielding materials to meet certain restrictions on dose rates, etc. After desired optimal parameters are established with ANISN calculations these parameters are incorporated in a more rigorous, 3 dimensional models used in MCNP runs to calculate the final results.

The MCNP analysis is performed for the 32PT-DSC as a typical payload in order to quantify the effect on dose rates of the use of the 0S197L TC. The resulting analysis demonstrates that the 0S1 97L TC dose rates with the decontamination area shielding and the trailer shielding are similar and less than the dose rates for the OS 197 TC.

W.5.2 Model Specification See Figure 5.1 for a description of the 3-D OS 197 TC model which used a 3" steel shell which corresponds to the cask shell and the neutron shield inner shell. The OS5197L TC neutron shield outer shell is also 1/16" thicker than that used in the model. The 3-D MCNP analysis model of a 32PT DSC inside an OS 197L TC is similar to that used in Appendix P.5 for the OS 197 TC with a 24PTH DSC. In this model, the 32PT DSC design basis source terms are used as a baseline analysis for the OS 197L TC. The data obtained is compared against a 2-D DORT model for the OS 197 TC with the 32PT DSC source terms as described in Appendix M.5. This comparison is used to document that the additional decontamination area and trailer shielding, in conjunction with the 0S197L TC, provide an equivalent level of shielding as the OS 197 TC. The increased surface dose rates for the OS 197L TC while on the crane hook will not impact operational dose rates since crane operations will be performed by remote crane control and using cameras and laser alignment systems.

W.5.3 Shielding Evaluation The use of the 0S197L (75 ton) TC is not expected to have any significant adverse impact on personnel dose rates during normal operation since crane operations will be performed remotely.

The maximum dose rates on the side of the cask for normal conditions (neutron shield in place NUH-003 Revision 9 Pane W.5-1 January 2006

FCN 721004-321 Rev. 1 Page 73 of 99 and filled with water) are shown in Table W.5-1. For the transfer from the decon area to the trailer with an empty neutron shield and a filled DSC/cask annulus, the dose rates are conservatively estimated using the dose rates for the accident condition (no neutron shield),

shown in the first table in Section W. 11. 1.4 for accident dose rates.

The dose rates associated with the OS 197L TC during the short time duration lifts from the pool to the decontamination area (54 rem/hr surface dose) and from the decontamination area to the trailer (13 8 rem/hr surface dose rate, second table in Section W. 11. 1.4, OS 197L) are significantly higher compared to OS 197 TC operational doses (346 mrem/hr surface dose). All operations associated with these two cask movements will be performed using remote crane operation using a laser/optical targeting system and cameras for confirmation of the cask location without the need for personnel in the vicinity of the cask. Should a failure of the crane occur during these operations, procedures will be in place to manually position the load in a safe, shielded location. Therefore, the dose received by operations personnel resulting from this high dose operation will be minimal as these operations are short duration and are performed remotely with no personnel in the vicinity.

The dose rates associated with the cask in the decontamination area and on the trailer (122 mrem/hr surface dose), are approximately one-third of the dose rates for the current configuration and relative to the precision of shielding analysis, can be considered to have similar shielding (346 mreni/hr surface dose, identified as OS 197 TC in Table W.5-1). The data provided for the UFSAR configuration above, is the data using the specific model used in the UFSAR. The data provided for the OS 197 TC configuration credits some additional shielding that was not credited in the IJFSAR analysis. The above data is for a 32PT-DSC payload but is provided for evaluation of relative doses. The relative effect of the 0S197L TC configuration and the decontamination area/trailer shielding configurations with respect to the OS 197 TC configurations shown above is representative of the relative effect for all CoC 1004 licensed DSC payloads for the OS5197L TC.

The loss of neutron shield accident dose rates are addressed in Appendix W. 11. These dose rates bound the doses from accident fire condition because the shielding on the trailer is not affected by the fire condition.

W.5.3.1 Solid One Piece Trunnion Dose Rate Evaluation Analyses are performed to compare the effect of the solid steel trunnion design to the original trunnion design (multiple pieces) which used NS-3 neutron absorber to reduce neutron dose.

The result of this analysis indicates that this change does result in an increase in neutron dose, however, since the majority of the dose contribution is gamma; the overall dose is reduced in the solid steel trunnion configuration. A comparison of the dose rates is provided in Table W.5-2.

In summary, the use of a one-piece trunnion reduces the total calculated dose rate by a factor greater than ten, thus providing a beneficial impact on occupational dose rates.

NUH-003 Revision 9 Page W.5-2 January 2006

FCN 721004-321 Rev. 1 Page 74 of 99 W.5.3.2 Removable Two Piece Neutron Shield Dose Rate Evaluation The two piece neutron shield provides the same level of shielding as the OS 197 TC neutron shield. The water cavity thickness is unchanged. The outer shell of the OS 197L TC neutron shield is slightly thicker than that used in the OS 197 TC (.25" versus. .18"). The addition of the seam between the two halves would reduce gamma dose in the vicinity of the seam but would increase neutron dose due to less water in the vicinity. As discussed for the trunnion modification above, since the total dose is primarily gamma, the increase in steel will result in a net decrease in total dose in the vicinity of the seams.

W.5.4 References

[5.1] One-Dimensional Discrete Ordinates Transport Code System with Anisotropic Scattering," CCC-254, Oak Ridge National Laboratory, RSICC Computer Code Collection, April 1991.

[5.2] A General Monte Carlo N-Particle Transport Code, Version 5, Volume 11: User's Guide, LA-CP-03-0245, 2003.

NUJH-003 Revision 9 Page W.5-3 January 2006

FCN 721004-321 Rev. 1 Page 75 of 99 Table W.5-1 0S197L TC Normal Condition Dose Rates Dose Rates at Different Distances from Side Surface -

Transfer Normal Condition Neutron Shield Filled Cniuats o CompoenRte On Side Surface 4.57 meters (15') 100 meters 609.9 meters (2000')

n Dose Rate, Dose Rate, Dose Rate, Dose Rate, mremlhr mremlhr mremlhr mremlhr UFSAR Neutron 261 Not Caic. Not Caic. Not Caic.

(Table M.5-5 Gamma 784 Not Calc. Not Caic. Not Caic and Section M. 11.2.5.3) Total 950 Not Caic Not Calc 0.01 Neutron 102 7.20 0.006 7.09e-6 0S197 TC Gamma 248 20.3 0.03 5.29e-5 Total 346 25.9 0.03 5.67e-5 Neutron 247 18.2 0.018 2.19e-5 0S197L TC Gamma 53,031 3906 4.52 9.70e-3 Bare Cask________

Total 53,249 3922 4.53 9.70e-3 OS197L TC Neutron 28 2 0.002 1.31e-6 with Decon Area or Gamma 94 11 0.02 2.44e-5 Trailer Additional Total 122 13 0.02 2.57e-5 Shielding __ _ I__ I_______

________ I________

Table W.5-2 Dose Rate Results for Two Trunnion Designs (mremlhr)

Trunnion Neutron Gamma Total Type Dose Rate Dose Rate Dose Rate Original Upper 0.2 621 621.2 Solid Steel 51.1 .14 61.24 Upper_______ _______

Original Lower 1.0 1702 1703 Solid Steel 79.5 1.3 80.8 Lower NUH-003 Revision 9 Page W.5-4 Page 2006

.5-4January

FCN 721004-321 Rev. I Page 76 of 99 BOTTOM TOP END END I----------- - - -11, 1 K F6260 DETAIL 1 S ..L 3.00" WATER JACKET 12.00" AIR CAP 5.50" STEEL DETAIL 1 Figure W.5-1 0S197L TC and Decontamination Area Shielding Model Geometry NUH-003 Revision 9 Page W.5-5 Page

.5-5January 2006

FCN 721004-321 Rev. I Page 77 of 99 W.6 Criticality The modifications associated with the OS 197L TC will not have a significant adverse impact on the criticality analyses performed for the OS 197 TC. The changes are in an area of relatively insignificant importance to criticality - no change in the fuel geometry / poison loading / or borated water concentration. The changes only affect the outer surface of the cask. The UFSAR shows that a reflective boundary, simulating an infinite cask array, was employed that further reduces the sensitivity of the analysis to TC design changes. In addition the outside diameter of the OS 197 TC and OS197L TC are basically the same. Therefore, these changes will have a negligible impact on the criticality analyses.

NUH-003 Revision 9 Page W.6- 1 January 2006

FCN 721004-321 Rev. 1 Page 78 of 99 W.7 Confinement There are no confinement features associated with the OS i197L TC on-site transfer cask since the cask is designed as a non-pressure retaining system. The DSC is the confinement system.

NUH-003 Revision 9 Paize W.7-1 January 2006

FCN 721004-321 Rev. 1 Page 79 of 99 W.8 Operating Procedures The following is a description of the operational sequences for use of the 0DS1 97L TC. In general, the steps are similar to those for the 0DS197 TC, described in detail in Chapter 5 of the UFSAR, and Chapter 8 of the canister-specific appendices (e.g., M.8 for the 32PT DSC). This chapter highlights the differences in operational steps when using ODS 197L TC relative to the 05S197 TC. Figures are provided to illustrate these steps.

Note: The applicable Technical Specification requirements for loading/unloading operations as listed in UPSAR Chapter 5 or Chapter 8 of the canister specific appendix are also applicable for this chapter when using OS 197L TC.

Placement of the DSC into the IDS197L TC and preparations for placement of the TC into the fuel pool are the same as for the OS 197 TC. The DSC/TC annulus is filled with clean water and sealed with the annulus seal. The TC neutron shield is also filled with clean water. As there is no fuel in the DSC at this time, the 75 ton limit is not approached, and the DSC may be filled with fuel pool water prior to lowering into the pool. This may be done either prior to the lift to the fuel pooi, or the IDS197L TC lowered to within a few feet of submergence and the DSC filled at that time. The IDS197L TC with DSC is then lowered to the fuel pool bottom and landed, and the yoke removed. Sequence 1 below shows the cask as it enters the pool.

OPTIONAL.

CASK COVER

~FUEL POOL-.,

fCASK ARA m O 0OS197L IS BROUGHT TO SURFACE NUH-003 Revision 9 Paize W.8-1 January 2006

FCN 721004-321 Rev. 1 Page 80 of 99 ALARA practices implemented during the TC movement are intended to reduce operational exposure, including temporary contamination barriers. In this case an optional flexible plastic/fabric cask protective cover may be used to keep the cask surface from being contaminated in the fuel pool. The protective cover is intended to prevent fuel pool water from coming in contact with the exterior surface of the cask, and would be removed as the cask is lifted from the fuel pool. As an alternate to the cask protective cover, the trailer bed could be lined with plastic/fabric to provide a barrier for any cask surface contamination. This barrier, in combination with the other parts of the transfer trailer, would prevent dispersal of any loose (smearable) material on the surface of the cask during transfer from the fuel handling/reactor building to the ISFSI and back. The use of the optional cask protective cover is shown in Sequence 5, where if used, it would be removed as the cask is lifted from the pool.

The use of the flexible plastic/fabric cask protective cover has no impact on the structural, shielding, or thermal analysis since it is only used in the pool and has minimal weight. During this time the DSC/cask annulus is filled with water and this defines the DSC shell temperature.

The use of a protective cover with the OS 197L TC will not adversely impact the criticality analyses performed for the OS 197 TC. Specifically the possible use of unborated water between the cover and the cask exterior has a negligible effect on the criticality analysis as the unborated water is outside the cask neutron shield. The use of a protective cover with the OS 197L TC will not adversely impact the mechanical interfaces. The cover will be used only in the fuel pool, and thus will have no interface with the trailer or HSM. It is anticipated that the cover will have a thickness less than 1/16" and will be compatible with fuel pool chemistry.

Selected Fuel Assemblies (FA's) are then placed into the DSC. Following fuel verification, the top shield plug is lowered into place and set. The yoke is then lowered and connected to the OS197L TC. The cask is then lifted until the cask top just breaks the surface of the fuel pool. At this time the water weight in the DSC and cask is offset by the buoyancy of the OS197L TC and allows for the hook weight to remain below 75 tons. However, further raising of the DSC and cask would exceed the 75 ton limit. This is shown as Sequences 2 through 4.

NUH-003 Revision 9 Pane W.8-2 JniTn1rv 2006

FCN 721004-321 Rev. I Page 81 of 99 ARELOADEDINTO 0S197L G)FUEL ASSEMBLIES 0 TOP SHIELD PLUG IS LOWEREDINTO DSC (7)S197L IS BROUGHT TO SURFACE. AND WATER WITHIN THE DSC IS PUMPED OUT NUHi-003 Revision 9 Page W.8-3 Page

.8-3January 2006

FCN 721004-321 Rev. 1 Page 82 of 99 Connections are then made to the DSC siphon and vent ports and the water within the DSC removed (p~umped out). During this water removal, a nitrogen or helium gas blanket will be supplied through the vent port as the water is drained. The neutron shield will not be drained during this step and the DSC/cask annulus will be maintained full. This is shown as Sequence 4.

After water has been pumped out from the DSC (approximately 13,600 lbs.), the OS 197L TC will be lifted from the fuel pool to the decontamination area. The 75 ton cask itself has significantly reduced shielding and employs draining of the water in the DSC to achieve the 75 ton limit. However, the OS1 97L TC operations utilize additional shielding and measures to achieve shielding capacity similar to the OS 197 TC. The OS 197L TC system consists of the bare cask and the upper and lower cask shielding utilized in the decontamination area, and the additional shielding provided on the cask support skid. The bare cask is in this reduced shielding configuration ONLY during the movement of the cask from the fuel pool to the decontamination area and from the decontamination area to the transfer trailer. Both of these operations are of short time duration (i.e. minutes). During bare cask movement from the fuel pool to the decontamination area and from the decontamination area to the trailer, remote crane operation and/or an optical targeting system with remote camera monitoring will be used to minimize personnel exposure to the reduced shielding configuration. This remote operation is shown in Sequence 5.

T FUELPOOLTODECONAREA T CASKIS MOVED (REMOTE FROM OPERATION)

OPrnONALCASKCOVER IS REMOVED

-AS CASKIS LIFTED OFROM FUEL POOL TO DECONTAMINATION AREA.

OSIM97L IS PLACED IN SHIELDING SLEEVE (PART OF 0S197L)

NUH-003 Revision 9 Paize W.8-4 January 2006

FCN 721004-321 Rev. 1 Page 83 of 99 In the decontamination area, the bare cask is placed in a shielding sleeve (lower cask shield) which provides shielding below the trunnions. An upper cask shield (shielding bell) is then placed on top of the shielding sleeve to shield the upper section of the cask. The shielding sleeve and shield bell are nominally 6" thick carbon steel. Placement of the cask in the shielding sleeve and placement of the shielding bell on the cask is performed using remote crane operation and/or an optical targeting system with remote camera monitoring. The OS 197L TIC system configuration of the cask and shielding sleeve and bell is shown as Sequences 5 and 6.

SIMILAR SHIELDING TO OS197 OTOP SHIELDING BELL COMPONENT OF 0S197L IS PLACED (REMOTE OPERATION)

The combination of the bare 0S197L TC and these shielding structures provide a similar level of shielding as the OS 197 TC in the radial direction. As stated in the Technical Specification 1.2.11 basis, which defines transfer cask dose rate limits, the determination of the cask dose rate limits is based on the shielding analysis documented in the UFSAR. This UFSAR analysis modeled both the axial and radial shielding of the OS 197 TC, and therefore, similar shielding levels shall be used when measuring dose rates for comparison to the TS. The configuration of the OS197L TC system in the decontamination area, within the shielding sleeve and bell, is one configuration for which the TS requirements for dose rates apply.

While in the shielding sleeve and bell, the canister is completely drained, vacuum dried, and helium backfilled, and all top covers welded in place. The OS 197L TC neutron shield will NUH-003 Reviqion 9 Paize W.8-5 January 2006

FCN 721004-321 Rev. 1 Page 84 of 99 remain filled and vented, similar to OS 197 TC operations, during these steps. During these operations, the cask and the shielding sleeve and bell provide occupational radiation shielding for personnel necessary to perform the canister closure operations. These operations are essentially unchanged from those listed in the UFSARý, Section 5.0 and the canister specific Appendices, such as M.8 for the 32PT. The shielding sleeve and the bell are designed to not interfere with the NUHOMSO AWS system or other equipment of the canister sealing operations. This is shown in Sequence 7.

OCANISTER IS PREPARED FOR CLOSURE OPERATIONS Once the DSC is sealed, the DSC/cask annulus will be drained and the cask top cover installed prior to downending onto the transfer trailer. In the event that the neutron shield is to be drained to reduce weight during the transfer from the decon area to the trailer, the DSC annulus will remain filled and the interim cover will be installed using a gasket to prevent annulus water from leaking during downending operations. The annulus will remain vented to the atmosphere through the annulus fill port in the cask side and/or through fittings in the interim cask cover.

Again during the downending process, the bare OS197L TC movement is of short time duration and is performed using remote crane operation and/or an optical targeting system with remote camera monitoring. This remote operation is shown in Sequence 8.

Note: See UFSAR Section W.8.1.5 regarding the use of a reduced weight interim cask cover.

NUH-003 Revision 9 Paize W.8-6 January 2006

FCN 721004-321 Rev. 1 Page 85 of 99

()CASK IS MOVEDFROM DECONAREATO TRANSFERTRAILER (REMOTEOPERATION)

Once on the transfer trailer, the skid provides 5.5" of carbon steel shielding to the sides of the cask up to the trunnions. A 2.5" thick carbon steel shield will be placed over the cask/skid inside the fuel building, after which a 3" thick carbon steel shield will be placed over the 2.5" thick shield providing a total of 5.5" of shielding on the skid. These shields may be placed on the skid inside the fuel handling building, or if load limits exist within the building, the 3" outer shield may be placed on the skid once the trailer exits the building. Placement of the inner shield on the skid inside the fuel handling building will be performed in accordance with the plants heavy loads procedures, and is evaluated within the plant 72.212 (50.59) for the dry fuel loading process. Sequence 9 shows this remote operation. If the neutron shield was drained during transfer from the decon area to the trailer (with the annulus filled), the neutron shield will be refilled and the annulus drained. The interim cover plate will then be replaced with the standard cask cover plate prior to exiting the fuel handling building.

NUH-003 Revision 9 Paze W.8-7 January 2006

FCN 721004-321 Rev. 1 Page 86 of 99 I a I I SUPPORT SKID INNER TOP SHIELDING (D INSTALLATION OF If fuel assembly weights are of a magnitude that would exceed the 75 ton limit, the standard cask top cover may be replaced with a reduced weight interim cover during transfer from the decontamination area to the trailer. Following placement of the cask on the trailer, and placement of the inner top shield on the transfer trailer, the interim cask top cover would be removed and the standard top cask cover installed prior to exiting the spent fuel/reactor building.

This is shown in Sequence 10.

NUH-003 Revision 9 Page W.8-8 January 2006

FCN 721004-321 Rev. 1 Page 87 of 99 es OINTERIM CASK TOP COVER IS REPLACED WITH STANDARD TOP COVER Following placement of the standard cask top cover, the trailer with the OS 197L TC may be moved out of the fuel building and the outer top trailer shielding installed outside, if the fuel building weight limits preclude placement of the outer top trailer shielding inside the fuel building. This is shown in Sequence 11. The OS 197L TC system shielding (6" of shielding) provided in the decontamination area and the 5.5" provided on the trailer, along with the shielding provided by the bare OS 197L TC, provides a level of shielding equivalent to that provided by the standard OS 197 TC (with lead shielding) and is the bounding condition of the two from a dose perspective (decon area and transfer trailer). Therefore the Technical Specification 1.2.11 limits for cask dose rates are to be measured in the trailer configuration.

NUH-003 Revision 9 Panye W.8-9 JanurWv 20 06 06

FCN 721004-321 Rev. 1 Page 88 of 99 FUEL BUILDING (PART 50 AND 72)

Si' INSTALLATION OF PART 72 SUPPORT SKID OUTER TOP SHIELDING The transfer trailer, with loaded OS 197L TC including the supplemental shielding, is then moved to the ISFSI and the Cask docked with the HSM. The DSC is then inserted into the HSM using the same methods as the OS 197 TC. This is shown in Sequence 12.

12TRANSFER TRAILER IS DOCKED TO HSM AND CANISTER IS TRANSFERRED NUH-003 Revision 9 Page W.8-10 .8-1OJanuary Page 2006

FCN 721004-321 Rev. 1 Page 89 of 99 W.8.1 Operational Differences between OS I97L and OS 197 TCs Listed below are each of the IJFSAR sections for preparation and loading of fuel assemblies into the HSM using the NUHOMSO OS1I97L TC system. In each section, which mirrors that of Section 5 of the UFSAR and Section 8 of the canister-specific appendices (e.g., M.8 for the 32PT DSC) the differences specific to the 0S197L TC are listed.

W.8.1.1 UFSAR Section 5.1.1.1 - Preparation of the Transfer Cask and DSC

The transfer cask may be filled in the fuel pooi or prior to placement in the fuel pool.

A decontamination cover may be provided to limit the potential for contamination of the exterior of the cask. Decontamination of the cask external surface as the cask is removed from the pool, if necessary, should be performed ALARA, due to the high cask dose rate for the bare cask in the 75 ton configuration on the crane.

W.8.1.2 UFSAR Section 5.1.1.2 - DSC Fuel Loading

A preliminary measurement of Technical Specification 1.2.11 or 1.2.11La limits for the dose rates at 3 feet from the top of the cask with the shield plug installed and water in the DSC cavity is performed.

Water is pumped out of the DSC when the cask breaks the surface of the fuel pool to reduce cask weight.

A helium or nitrogen back fill will be provided during initial draining to eliminate exposure of the fuel to air. The shield plug restraints will be installed to prevent shield plug movement.

The decontamination cover, if provided, is removed from the cask while the cask is removed from the pool. This may be performed by connecting the decontamination cover to a fixed point above the pool to strip the cover without the need for personnel in the area.

" Personnel are evacuated from the area, as specified by plant's ALARA practices, due to the high cask dose rates. Crane operations will be performed remotely using cameras and laser/target positioning.

The cask is placed into the decontamination area shielding sleeve once removed from the fuel pool. A decontamination area shielding bell is then placed over the side of the cask above the upper trunnions. The shielding sleeve and bell provide the additional shielding to produce similar shielding as the OS 197 TC.

W.8.1.3 UFSAR Section 5.1.1.3 - DSC Drying and Backfillinig 0 Technical Specification 1.2.11 or 1.2.11La limits for DSC dose rates at 3 feet from the top of the cask with the Automated Welding System (AWS) installed on the inner top cover plate are to be verified after the cask is placed in the decontamination area shields. This NUH-003 Revision 9 Page W.8-11 January 2006

FCN 721004-321 Rev. 1 Page 90 of 99 configuration offers shielding similar to that of the OS 197 TC. Verification of TS dose limits is performed to determine if a fuel misload has not occurred.

W.8. 1.4 UFSAR Section 5.1.1.4 - DSC Sealing Operations

" No change

If the neutron shield is to be drained to reduce weight during the transfer from the decon area to the trailer, the DSC/cask annulus will be maintained full and the interim cask cover will be installed with a gasket to prevent annulus water leakage during downending. The annulus will be vented to atmosphere through a cask fitting or a fitting on the interim cask cover.

CAUTION: During the DSC closure operations, the opening at the top and bottom of the decontamination area shielding shall be monitored (visual inspection) to assure no significant blockage of openings. Although blockage is improbable as all 16 openings would require sealing, personnel shall perform visual inspection of shielding sleeve and bell openings during the operations when DSC is in the sleeve.

W.8.1.5 UFSAR Section 5.1.1.5 - Transfer Cask Downending and Transport to ISFSI

Crane operations for removal of the decontamination area shielding bell, engagement of the yoke to the cask trunnions, movement of the cask to the trailer, lowering of the cask onto the trailer and placement of the trailer shielding on the cask will be performed remotely using cameras and laser/target positioning, due to bare cask dose rates. The additional trailer shielding may be placed on the trailer at this time or the outer top shield may be placed on the trailer once the trailer exits the auxiliary building, as applicable based on site specific weight restrictions.

If fuel assembly weights are of a magnitude that would exceed the 75 ton limit, the standard cask top cover may be replaced with a reduced weight interim cover during transfer from the decontamination area to the trailer. If water from the neutron shield was drained for transfer from the decon area to the trailer, the neutron shield will be refilled after the cask is placed on the trailer and after placement of the inner shield cover on the trailer. The DSC/cask annulus will then be drained. Following placement of the cask on the trailer, and placement of the inner top shield on the transfer trailer, the interim cask top cover would be removed and the standard top cask cover installed prior to exiting the spent fuel/reactor building.

" The 0S197L TC system shielding (6" of shielding) provided in the decontamination area and the 5.5" provided on the trailer, along with the shielding provided by the bare OS 197L TC, provides a level of shielding similar to that provided by the standard OS 197 TC (with lead shielding) and is the bounding condition of the two (decon area and transfer trailer). Therefore the Technical Specification 1.2.11 limits for cask dose rates are to be measured in the trailer configuration.

" The interim top cask cover, is an aluminum plate (nominal I" thick and 78.62" diameter) that interfaces with the cask top bolting, similar to the standard top cask cover.

Following placement of the cask on the trailer, and placement of the inner top shield on NUH-003 Revision 9 Page W.8-12 January 2006

FCN 721004-321 Rev. 1 Page 91 of 99 the transfer trailer, the interim, cask top cover would be removed and the standard top cask cover installed prior to exiting the spent fuel/reactor building. The aluminum cover plate is approximately 4, 000 lbs. lighter than the standard cover. The function of the interim cask top cover is to provide some additional shielding in the axial direction, but more importantly to provide assurance that the DSC will remain within the TC under events that are beyond design basis. The interim cask top cover will also prevent any shifting of the DSC within the TC prior to placement of the standard cask cover. The interim cask top cover will not be used outside the fuel building. The interim top cover will be placed on the cask with a gasket if the DSC/cask annulus is to be maintained fuill.

The 1" aluminum cover will see minimal stress due to the hydraulic head of the annulus water level.

The interim cover will itself have lifting points that meet ANSI N14.6 and is anticipated to weigh less than 500 lbs.

The effect on personnel doses will be minimal since the timeframe for use of this cover is short and significant shielding is provided at the top of the canisters.

The use of this cover will not impact the criticality analysis and will provide a slight improvement in thermal performance (heat rejection from the DSC).

W.8.1.6 UFSAR Section 5.1.1.6 -DSC Transfer to the HSM

" Following placement of the standard cask top cover, the trailer with the OS 197L TC may be moved out of the fuel building and the outer top trailer shielding installed outside, if the fuel building weight limits preclude placement of the outer top trailer shielding inside the fuel building.

Install the cask top centerline alignment target, through the trailer shielding.

CAUTION: During the actual movement of the Transfer Cask on the transfer trailer to the 1SF SI, the gap between the transfer deck and bottom of the skid shall be monitored (visual inspection) to assure no significant blockage of airflow. Although blockage is improbable as over 60 feet of gap would require sealing, personnel shall maintain a visual scan of the trailer.

The operational differences specified above for loading operations will also apply for unloading operations.

NUH-003 Revision 9 Page W.8-13 January 2006

FCN 721004-321 Rev. 1 Page 92 of 99 W.9 Acceptance Criteria and Maintenance Program All acceptance criteria and maintenance requirements for the OS 197L TC are identical to those of the OS 197 and OS 197H TCs described throughout the body of this UFSAR.

NUH-003 Revision 9 Paize W.9-1 January 2006

FCN 721004-321 Rev. I Page 93 of 99 W. 10 Radiation Protection As discussed in Section W.5, use of the OS1 97L TC does not significantly affect personnel dose rates (during closure operations, handling, or storage) or site boundary dose rates. The 051 97L TC is used only for loading/unloading and transfer operations, and the storage conditions are unchanged. Therefore, the personnel doses, occupational exposures and site bounding dose rates documented for each DSC/IISM storage configuration in Section 7.4 and Appendices K.l10, L. 10, M. 10 and N. 10 remain unchanged and are applicable to operations using the OS1I97L TC.

The use of the OS 197L TC is not expected to have any significant impact on personnel dose rates during normal operation since the operations for placement and removal of bare 051 97L TC from the fuel pool into the decontamination area shielding sleeve, placement and removal of the decontamination area shielding bell, engagement of the yoke to the cask trunnions, movement of the cask to the trailer, lowering of the cask onto the trailer and placement of the trailer shielding on the cask will be performed remotely using cameras and laser/target positioning.

NUH-003 Revision 9 Page W.10-1 January 2006

FCN 721004-321 Rev. 1 Page 94 of 99 W. 11I Accident Analyses This section describes the postulated accident events that could occur during fuel loading, draining, drying, welding and transfer of the DSC using a NUHOMSO OS 197L TC. Sections which do not affect the evaluation presented in Chapter 8 or Appendices K. 11, L. 11, M. 11 and N. 11I for various DSC designs are identified as "No change." Detailed analysis of the events are provided in other sections and are referenced herein.

W. 11. 1 Postulated Accidents Only those accidents affecting the OS 197L TC are addressed in this section. There is no change to accident evaluations affecting other NUHOMSO components.

WIL1L1. OS1 97L TC Missile Impact Analysis This event is described in Section 8.2.2.4. The 0S197L TC uses a 2.68" steel shell in lieu of a 1.5" steel shell with a nominal 3.5" lead annulus and a 0.5" inner liner for OS 197 TC. The missile impact analyses for the OS 197 TC are therefore bounding for the OS 197L TC.

W.1 1.1.2 Earthqguake This event is described in Section 8.2.3.D. The OS 197L TC configuration (cg location, cask length, trunnion location and bottom forging configuration) does not significantly differ from that of the OS 197 TC. The 0S197L TC remains stable when subjected to the design basis earthquake.

W. 11.1.2.1 OS1 97L TC in a Vertical Configuration during Vacuum Drying and Welding Operations The bottom forging on which the cask is resting during vertical cask operations, is the same size and configuration as the OS 197 TC. The 0S197L TC cg location is not significantly altered by the change in the cask shell configuration. The addition of the decontamination area shield will provide a larger diameter, more stable shell, outside the cask envelope, thereby potentially enhancing the OS I97L TC seismic capacity.

W.1 1.1.2.2 0S197L TC in a Horizontal Configuration during Transfer Operations The cask seismic stresses for the OS 197L TC are bounded by the OS 197 TC stresses due to the similar configurations of the cask ends (top and bottom forgings and covers) and larger thickness structural shell.

The trailer with the OS I97L TC, with the additional shielding, remains stable for the design basis seismic accelerations.

W.1 1.1.3 0S197L TC Accidental Cask Drop This event is described in Sections 8.2.5.2.13, D and E.

NUH-003 Revision 9 Panye W. 11-1 January 2006

FCN 721004-321 Rev. 1 Page 95 of 99 See Section W.3.1 .3 for a discussion of the 0S197L TC drop accident. This drop accident is bounded by the results for the 0S197 TC drop accident discussed in Section 8.2.

W. 11.1.4 Loss of Neutron Shield This event is described in Section 8.2.5.3.

For the accident condition (the unlikely cask drop scenario) a complete loss of the CISI 97L TC neutron shield is postulated similar to the OS 197 TC evaluation described in Section 8.2.5.3. In addition, since the trailer shield is not important to safety, the analysis conservatively assumes that all the trailer shielding is lost. However, the trailer shield is fabricated using two sets of plate shields (the inside shield is 2.5" thick, the outside shield is 3" thick) which may be damaged in a drop but are unlikely to separate completely from the skid and cask.

Assuming the non-mechanistic drop scenario occurs and the trailer shields and the cask are dislodged completely from the trailer and skid, recovery actions are required to manipulate the shields or providing supplemental shielding to reduce dose rates to a reasonable value until a long term recovery plan is in place.

0S197L TC ACCIDENT CONDITION DOSE RATES Dose Rates at Different Distances from Side Surface -

Transfer Cask Dose Rate _________ Accident Condition No Neutron Shield Configuration Component On Side Surface 4.57 meters (15') 100 meters 609.9 meters (2000')

Dose Rate, mremlhr Dose Rate, mremlhr Dose Rate, mremlhr Dose Rate, mrmlihr Neutron 3,780 Not Caic. Not Caic. Not Calc.

UFSAR (Table Gamma 1,070 Not Caic. Not Caic. Not Caic M.1 1-2)

_______ Total 4,640 Not Caic Not Caic 0.01 Neutron 1,282 66 0.067 1.87e-5 0S1 97 TC Gamma 291 30 0.04 5.14e-5

_______ Total 1573 84 0.10 6.48e-5 7 C OS Neutron 3,691 187 0.20 1.06e-4 (Bare Cask) Gamma 134,328 11,576 12.7 3.19e-2 Total 138,019 11,763 12.9 3.20e-2 The dose rates provided for the UFSAR configuration above, are based on 32PT DSC with design basis source terms inside OS 197 TC. The shielding analysis for the CIS197L TC configuration presented in W.5 credits some additional shielding such as the 32PT DSC basket aluminum rails and other basket structures that were not included in the CIS 197 TC evaluation (see Appendix M.5.4) due to limitation of the previous analysis methods. The above data for the 0S197L TC bare cask is for a 32PT DSC payload but is provided for evaluation of relative doses. The relative effect of the OS 197L TC configuration and the decontamination area/trailer shielding configurations with respect to the OIS 197 TC configurations shown above is representative of the relative effect for all CoC 1004 licensed DSC payloads for the 0S197L TC.

The dose rates on the ends of the OS 197L TC will be the same as the OS 197 TC since the top and bottom forging and cover plate configurations have not been modified.

NUH-003 Revision 9 Page W. 11-2 January 2006

FCN 721004-321 Rev. 1 Page 96 of 99 As shown in the table below, the dose rates at the site boundary, assuming a 100 meter site boundary, would be approximately 13 mrem/hr during the timefirame that the cask trailer shield is dislodged from the cask and until the trailer shield is repositioned.

The 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months of recovery period assumed is appropriate because the repositioning of the trailer shields will be performed using lifting hardware pre-positioned prior to transfer operations. This will facilitate quick positioning using a crane to minimize the need for personnel to approach the cask.

A comparison of the OS 197 TC and 0S197L TC accident dose analyses using the 32PT DSC as a representative payload is provided below:

Recovery Total Person- Total Person-Contact Dose at Dose at 100 Dose at Period Dose at 100 Dose at 2000 Dose 15 feet meters 2000 feet Assumed meters feet Cask (mrem/hr) (mremlhr) (mremlhr) (mremlhr) (hours) (mrem) (mrem)

UFSAR (Section 4,640 700 5.25 .011 8 N/A 0.09 M. 11.2.5.3) __ _ 1__ _ _ 1__ _ __ _ 1__

_ __ 1__

__ _ _ __ _ _ __ _ _ _ __ __1 _ _ _

0S197 TO 1,573 84 0.10 6.48e-5 8 0.8 5.1 84e-4 OS197LTC 138,019 111,763 12.9 0.032 8 103.2 0.25 The increase in dose rates at the site boundary (100 meters) is significant (approximately 130 times) between the OS 197 TC and OS 197L TC values. However, the total dose at 100 meter site boundary still remains very low (103 mrem) and below the regulatory limit of 5,000 mrem.

A review of the UPSAR shows that the TC payload that produces the highest 100 meter dose rate is the 24PHB DSC (Appendix N. 11). This is a 7 mrem/hr dose rate (Section N. 11.2.5.3). Using the ratio of UFSAR dose rate and OS197L TC dose rate from the table above results in a factor of 12.9/5.25=2.45. Applying this factor to the 7 mremlhr dose rate for the 24PHB DSC results in a 100 meter dose rate for a 24PHB DSC within the 0S197L TC of 2.45x7=18 mrem/hr. This dose rate, applied over the 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period, results in a total person-dose of 18x18= 144 mrem. The 144 mrem is approximately 3% of the 5000 mrem limit for offsite exposure.

NUH-003 Revision 9 Page W. 11-3 January 2006

FCN 721004-321 Rev. 1 Page 97 of 99 W.12 Operating Controls and Limits The addition of OS197L TC to the standardized NUHOMSO system does not result in any change to the Technical Specifications, Functional and Operating Limits described in NUHOMSO CoG 1004 Amendment 8.

NUJH-003 Revision 9 Paae W. 12-1 January 2006

FCN 721004-321 Rev. 1 Page 98 of 99 W. 13 Quality Assurance Chapter 11I provides a description of the Quality Assurance Program to be applied to the safety-related and important-to-safety activities associated with the standardized NUHOMSQ system.

The addition of OSl197L TC to the NUHOMS" system does not require any changes to the quality assurance requirements stipulated in Chapter 11.

NUJH-003 Revision 9 Pane W. 13-1 January 200d5 6

FCN 721004-321 Rev. 1 Page 99 of 99 W. 14 Decommissioning No change to the decommissioning evaluation presented in Section 9.6 due to the addition of the OS1I97L TC to the NUHOMSO system.

NUH-003 Revision 9 Pagye W.14-1 Janiirn 2006

LIC-06-0056, Request for Exemption from Nuhoms Certificate of Compliance No. 1004, Amendment No. 8

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LIC-06-0056, Request for Exemption from Nuhoms Certificate of Compliance No. 1004, Amendment No. 8

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