ML18088A180

to Application for Amendment 15 to Standardized Nuhoms Certificate of Compliance No. 1004 for Spent Fuel Storage Casks, Supplemental Response to First Request for Additional Information

ML18088A180
Person / Time
Site:	07201004
Issue date:	03/22/2018
From:	Bondre J Orano USA, TN Americas LLC
To:	Document Control Desk, Office of Nuclear Material Safety and Safeguards
References
CAC 001028, E-51144, EPID: L-2017-LLA-0012
Download: ML18088A180 (41)

Text

0 orano C olumbi a Office 71 3 5 Min st r e l Way C olumbi a, MD 2 1045 Tel: (4 1 0) 91 0-69 00 @Or a n o_U SA U.S. Nuclear Regulatory Commission Attn: Document Control Desk One White Flint North 11555 Rockville Pike Rockville , MD 20852 March 22 , 2018 E-51144

Subject:

Revision 3 to Application for Amendment 15 to Standardized NU HOMS Certificate of Compliance No. 1004 for Spent Fuel Storage Casks , Supplemental Response to First Request for Additional Information (CAC No. 001028, Docket No. 72-1004, EPID: L-2017-LLA-OO 12)

Reference:

(1) U.S. NRC Conversation Record by Christian Jacobs (NRC), " Clarification Call on RAI #1 response for CoC 1004 Amendment 15 application

," dated March 12, 2018 , ADAMS Accession No. ML 18071A059 (2) Application for Amendment 15 to Standardized NUHOMS Certificate of Compliance No. 1004 for Spent Fuel Storage Casks, Revision 2 , Response to First Request for Additional Information , TN Americas LLC submittal E-50037 transmitted to the NRC on December 14 , 2017 This subm i ttal provides revised responses to three questions (RAI 7-1 , RAI 7-3, and RAI 8-3) o ri ginally provided in Reference 2 , plus other information requested during a clarificat i on call with NRC staff on March 8 , 2018 described in Reference

1. Enclosure 2 of this submittal provides a proprietary version of the revised response to each of the identified request for additional information (RAI) items , with the public version provided in Enclosure

3. These revised RAI responses supersede the prior responses to these three RAls provided in Reference 2 in their entirety. There we r e no changes to the response to RAI 8-3; however , the updated final safety analys i s report (UFSAR) changed pages have been updated to reflect the information provided in the response. Enclosure 4 provides the proprietary version of the UFSAR changed pages , with the public version of the UFSAR change pages provided in Enclosure

5. Enclosure 6 provides the proposed Technical Specificat i ons changed pages. An updated description of the Amendment 15 changes is include d as Enclosure

7. There were no changes to th e scope of Amendment 15 (as described in Section 1.0 of Enclosure

7) from that provided in the original Amendment 15 (Rev i sion 0) submittal.

However , a consolidated description of the changes to the T e chnical Spec i fications associated with the RSI Document Control Desk March 22 , 2018 E-51144 Page 2 of 2 and RAI responses is included in Section 2.0 of Enclosure

7. Enclosure 8 (computer disk) provides sample computer input files , as described in the response to RAI 7-3. For the UFSAR , replacement and new Amendment 15, Revision 3 pages and drawings are provided. New or changed pages are annotated as Revision 3 , with changes indicated by italicized text and revision bars. Changes made in response to RAls are outlined in red and identified by the number of the RAI that necessitated the change. For the TS, Amendment 15 changes on Revision 3 pages are indicated by italicized text , revision bars and outlined in red to distinguish them from the changes proposed in Revisions 0 , 1 and 2 of the Amendment 15 application. In accordance with 10 CFR 2.390 , TN Americas LLC is providing an affidavit (Enclosure 1), specifically requesting that proprietary information be withheld from public disclosure. With the exception of Enclosure 8 , public versions of portions containing proprietary and/or security-related information are provided. Should NRC staff have any questions or require additional information to support the review of this application , please contact Mr. Dennis Williford by telephone at (704) 805-2223 , or by email at Dennis.Will i ford@orano

.group. Sincerely , -4,o~ Jayant Bondre Chief Technical Officer cc: Chris Jacobs (NRC SFM) as follows , all provided in a separate mailing:

• One paper copy of this cover letter

• Six computer disks , each containing this cover letter and Enclosures 1 through 8

• One computer disk of the complete CoC 1004 Proposed Amendment 15 , Revision 3 Technical Specifications in MS-WORD format

Enclosures:

1. Affidavit Pursuant to 10 CFR 2.390 2. Revised Responses to Request for Additional Information (Proprietary Version) 3. Revised Responses to Request for Additional Information (Public Version) 4. CoC 1004 Amendment 15 , Revision 3 UFSAR Changed Pages (Proprietary)

5. CoC 1004 Amendment 15 , Revision 3 UFSAR Changed Pages (Public Version) 6. Coe 1004 Amendment 15 , Revision 3 Technical Specifications Changed Pages 7. Description of Amendment 15 Changes 8. Sample Computer Input Files Computer Disk (Proprietary)

Enclosure 3 to E-51144 Revised Responses to Request for Additional Information (Public Version)

Responses to Request for Additional Information (public version) 7 -Criticality Evaluation Enclosure 3 to E-51144 1. Revise the Safety Analysis Report (SAR) to clarify if Figure M.1-2 is applicable to both 84C Poison Rod Assemblies (PRAs) a n d silver-indium-cadmium (AIC) PRAs, or 84C PRAs only. If Figure M.1-2 is applicable only to B 4 C PRAs, provide a separate figure showing the arrangement of AIC PRAs. In either case , provide a specification for the minimum AIC absorbe r material rod diameter.

Figure M.1-2 shows the layout and dimensions for PRAs to be used in the 32PT Dry Shielded Canister (DSC). However, it is not clear if this figure applies to B 4 C PRAs only, or t o both B 4 C and AIC PRAs. The Figure should be revised to indicate which PRA types it is applicable to. If it is not intended to b e applicable to the AIC PRA, a separate figure should be provided. In either case , the minimum AIC absorber material rod diameter should be specified. For B 4 C PRAs , the minimum absorber material rod diameter is defined by the rod outer diameter and clad thickness dimensions , as the absorber material is B 4 C powder which fills the entire clad interior volume. The AIC absorber alloy is a solid material , however , which will necessarily have a smaller diameter than that defined by the clad dimensions. The minimum diameter of AIC absorber material should be limited to the dimension shown to maintain the cask subcritical in the criticality analysis. Th i s information is needed to ensure that the Standardized NU HOMS 32PT DSC will continue to meet the criticality safety requirements of 10 CFR 72.124. RESPONSE TO RAI 7-1 Figure M .1-2 of the Final Safety Analysis Report has been revised to clarify its applicability to both B 4 C and AIC PRAs and to incorporate the AIC PRA absorber dimensions. Application Impact: FSAR Figure M.1-2 has been revised as described in the response. Page 1 of 4 Responses to Request for Additional Information (public version) Enclosure 3 to E-51144 3. Clarify the amount of credit taken in the 32PT DSC criticality analysis for the silver content in the AIC PRAs, in relation to the amount of silver required per Note 3 of Table 1-1 h of the Technical Specifications.

Section M.6 of the SAR states that 40% of the actual silver content is credited as being present , with 75% of that credited in the criticality analysis. However , Note 3 of Table 1-1 h of the Technical Specifications specifies 2.46 grams of silver per centimeter of absorber rod in the AIC PRA. The footnote to Table M.2-4a of the SAR states that 75% of t hat value is credited in the criticality analysis. It is not clear how the minimum 2.46 grams per centimeter of absorber rod is determined , since it appears that , based on the dimensions in Figure M.1-2, this value represents roughly 44% of the density of elemental silver (at 10.49 g/cm 3), or roughly 56% of the density of silver in typical 80% silver, 15% indium , 5% cadmium AIC alloys. Revise the SAR to state the expected density of absorber material in the AIC PRA , as well as the minimum diameter of absorber material in the rods of the AIC PRA. Note that conservatism in the modeling of AIC material is used to account for the lack of representative benchmark experiments with AIC material in the computer code validat i on analysis. This information i s needed to ensure that the Standardized NU HOMS 32PT DSC will cont i nue to meet the criticality safety requirements of 10 CFR 72.124. RESPONSE TO RAI 7-3 The critical requirement for the silver-indium

-cadmium (AIC) poison rod assembly (PRA) is the minimum linear mass of the silver con t ent , Ag , per rod as shown in Table 1-1 h of the Technical Specifications (TS) or Final Safety Analysis Report (FSAR) Table M.2-4a. The initial AIC composition is 80 wt.% Ag , 15 wt.% In , and 5 wt.% Cd and its density is 10.17 g/cm 3 (IAEA-TECDOC-949 , June 1997). [ ] This translates into a linear density of 2.60 g/cm [ ] The AIC PRA description in FSAR Table M.2-4a has been revised to provide the updated poison length , pellet diameter and minimum required Ag content per rod and the silver content basis. The results presented in FSAR Appendix M.6 (M.6 , M.6.3.2 , M.6.4.1.3) have been revised to clarify and provide a consistent bas i s for the required silver content in the AIC PRA. In addition , FSAR Table M.6-5 has bee n revised to include a footnote for the AIC PRA that the silver composition is credited in the KENO base model. Page 2 of 4 Responses to Request fo r Additional Information (public version) Enclosure 3 to E-51144 While the base model utilizes an AIC pellet radius of 0.4902 cm , the AIC pellet radius for a WE17x17 fuel assembly may be smaller , as shown in Table 4 of NUREG/CR-6759. [ ] FSAR Section M.6.4.4 has been revised to incorporate the sensitivity analyses information and to state the PRA modeling assumptions and the sensitivity analyses performed to demonstrate the applicability of the linear density requ i rement. TS Table 1-1 h has also been revised to indicate that the minimum silver content per rod is 2.60 g/cm. In addition , sample input files used in the KENO base case and sensitivity analyses are provided in Enclosure 8 of this submittal.

Application Impact: FSAR Sections M.6 , M.6.3.2 , M.6.4.1.3 and M.6.4.4 have been revised as described in the response. In addition , FSAR Tables M.2-4a and M.6-5 have been revised as described in the response. Technical Specifications Table 1-1h has been revised as described in the response.

Page 3 of 4 Responses to Request for Additional Information (public version) 8 -Materials Evaluation Enclosure 3 to E-51144 3. Update the Technical Specifications Table 1-1 e to include the following information in the definition of Fuel Damage: The extent of damage in the fuel assembly, including non-cladding damage , is to be limited such that a fuel assembly is able to be handled by normal means. The extent of damage in the fuel rods is to be limited such that a fuel pellet is not able to pass through the damaged cladding during handling and retrievability is ensured following normal and off-normal conditions.

The applicant provided this description of damaged fuel for the 32PT DSC in the UFSAR change pages in section M2.1 Spent fuel to be stored. This information is necessary to assure compliance with 10 CFR 72.236(9).

RESPONSE TO RAI 8-3 As requested , Technical Specification Table 1-1e has been revised to clarify the definition of Fuel Damage by adding the following informat i on on fuel rod damage: " The extent of damage in the fuel rods is to be limited such that a fuel pellet is not able to pass through the damaged cladding during handling and retrievability is ensured follow i ng normal and off-normal conditions

." Similar changes have been made to the defin i tion of fuel damage in TS Tables 1-1 i , 1-1j , 1-11 , 1-1t, 1-1aa , 1-1 gg and 1-111. In addition , the definition of fuel damage has been clarified for other DSCs in the FSAR by add i ng the sentence above to FSAR Sections N.2.1 , T.1.1 , U.1.1 , U.2.1 , Y.2.1 and Z.2.1. Application Impact: FSAR Sections N.2.1 , T.1.1 , U.1.1 , U.2.1 , Y.2.1 and Z.2.1 have been r evised as descr i bed in the response.

TS Tables 1-1e , 1-1 i , 1-1j , 1-11 , 1-1t , 1-1 aa , 1-1gg and 1-111 have been revised as desc ri bed in the response. Page 4 of 4 Enclosure 5 to E-51144 Coe 1004 Amendment 15 Application, Revision 3 UFSAR Changed Pages (Public Version)

March 2018 R ev ision 3 All indicated changes are in response to RAI 7-1 PLATE .25" TH K "A" "B" DIMENSION ABSORBER ROD OD NOMINAL (IN) ABSORBER ROD DIMENSION

" A"(IN) ABSORBER STACK HE IGHT, " B" (IN) CLAD TH ICK NESS NOMINAL (I N) No. OF RODS CLAD MATERIAL THE SPECIFIC F UEL ASSEMBLY GRAPPLE r REMOVABLE FIXTURE TO MATE WITH ,--1

• THESE DIMENSIONS ARE FOR USE WITH WESTINGHOUSE 17xl 7 FUEL ASSEMBLIES. DIMENSIONS WILL VARY AS REQUIRED BY FUEL ASSEMBLY TYPE. ABSORBER TYPE B.c FUEL ASSEMBLY TYPE WE 17x17 8&:W 15x15 WE 15 x 15 CE 14x14 .362 .438 .450 .975 156 160 156 143 151 151 150 129 .018 .022 .023 .049 24 16 20 5 304 SST 304 SST 304 SST 304 SST 4.00" +/-.21 (SHORT DSC) ABSORBER ROD WE 14 x14 .432 156 150 .022 16 304 SST 12.5 0" +/-.2-! (LONG DSC) AIC WE 17x 17 .381 156 150 .018 24 304 SST ABSORBER PELLET NOMINAL .341 OUTER DIAMETER (IN) -----Figure M.1-2 Poison Rod Assemblies (B 4 C-PRAs or AIC-PRAs) 72-1004Am e ndm e nt No. 15 PageM.1-121 Assembl y Class W E 1 7 xl7 All indicated changes are in response to RAI 7-3 Table M.2-4 Poison Rod Assembly (PRA) Description Asse mbl y Minimum N umber Mo deled B 4 C Co ntent MinimumB 4 C per Rod (g/cm) Content per Rod C la ss WE 17x17 B&W 15 x 15 WE 15 x 15 CE 14 xl4 WE 14 x l4 N umb e r of R od s/A IC PRA 24 ofRods/PRA (75% C redit) 24 0.59 16 0.72 20 0.72 5 3.1 4 16 0.72 Table M.2-4a AIC PRA Description

!No minal!P oison !No minal!P e ll e l L e ng/It (ill.) diam eter (in.) I 1 50 j 0 I 0.341 I (g/cm) 0.79 0.96 0.96 4.19 0.96 Minimum R e quire d Ag Content p e r Rod (2 1c m) (IJ 1 2.6 (or 6.6 f!/in.) I (1) The Az conte nt svecified i s ba sed on a [ I March 2018 R e vi s ion 3 1 1 Table M.2-4b B-10 Specification for the NUHOMS-32PT Poison Plates NUHOM~-32PT Number of Minimum B-10 Areal DSC Ba ske t Type B 4 C PRA s D e n s i ty, gmlc m 2 A 0 0.00 7 0 A l, A l-3 2 0 0.0150 A2 , A 2-32 0 0.0200 B 4 0.00 7 0 Bl 4 0.0150 B 2 4 0.0200 C 8 0.0070 C l 8 0.0 1 50 D 1 6 0.00 7 0 NUH O M~-32PT N umb e r of Minimum B-10 A r ea l DSC Ba sket Type A IC PRA s D ens i (v, gm!cm 2 Bl-r 4 0.0 1 50 B2-r 4 0.0200 C l-r 8 0.0150 C2-r 8 0.0200 7 2-1 00 4A m e ndm e ntNo.15 Page M .2-l 6a All indicated changes are in response to RAI 7-3 M.6 Criticality Evaluation The design criteria for the NUHOMS-32PT DSC requires that the fuel loaded in the DSC remain subcritical under normal, and accident conditions as defined in 1 OCFR Part 72. The NUHOMS-32PT DSC system's criticality safety is ensured by fixed neutron absorbers, soluble boron in the pool and favorable geometry.

Burnup credit is not taken in this criticality evaluation.

The fixed neutron absorbers are present in the form of borated metallic plates and non-irradiated poison rod assemblies (PRAs) with B 4 C absorber or Ag-In-Cd absorber , which are inserted in the guide tubes of certain assemblies in the basket. The PRA with Ag-In-Cd absorber is designated as AIC PRA while the PRAs with B 4 C absorber is designated as B 4 C PRA or PRA, as applicable.

These materials are ideal for long-term use in the radiation and thermal environments of a DSC. A basket may contain 0, 4, 8, or 16 PRAs and is designated a Type "A," Type "B," Type "C" or Type "D" basket , resrectively. Each of these basket types has the same minimum Boron-IO content of 0.007 gm/cm , (90% credit taken in the criticality analysis or 0.0063 gm/cm\ Two additional basket types, designated as Type Al and A2 , have a higher minimum Boron-IO content of 0.015 gm/cm 2 and 0.020 gm/cm 2 , respectively , (90% credit taken in the criticality analysis or 0.0135 gm/cm 2 and 0.018 gm/cm 2 , respectively).

Type A 1 and A2 basket configurations are qualified for the 24 poison plate basket design only with O PRAs. Metal Matrix Composites (MMCs) at a minimum areal density of 0.0070g/cm 2 have been qualified for use as a neutron absorber with 90% credit as justified in Section M.9.1.7.2.

Similarly, Section M.9.1.7.1 provides the justification for the use of 90% credit for borated aluminum. In addition to the fixed neutron poison in the basket , PRAs may be required for the center four, eight or sixteen assemblies depending on fuel assembly design and initial enrichment.

The minimum required B 4 C content of the PRAs is 40% Theoretical Density (TD) with 75% credit taken in the criticality analysis or 30% TD. The minimum required Ag content of the AIC PRAs includes a [ 1----------------------'

Based on Basket Type Al and A2, two more basket types are added and designated as Type Al-32 and A2-32. These two new basket types are identical to Type A l and A2 except that they have 32 poison plates. Based on Basket Type B , four more basket types are added and designated as Type Bl , B2 , Bl-r, and B2-r. Type Bl and B2 are identical to Type B except that they have higher minimum Boron-JO content of0.015 gm/c m 2 and 0.020 gm/cm 2. Type BJ-rand B2-r are identical to Type BI and B2 except that they contain 4 AIC P RAs instead of 4 B 4 C P RAs. Based on Basket Type C, three more basket types are added and designated as Type CJ, C J-r , and C2-r. Type CJ is id entical to Type C except that it has higher minimum B-10 content of 0.015 gm/cm 2. Type Cl-r i s id entica l to Type CJ excep t that it contains 8 AIC PRAs instead of 8 B4C PRAs. Typ e C2-r is identi cal to Type CJ-r except that it ha s higher minimum B-10 content of 0. 020 gm/cm 2. Basket designations as a function of number of PRAs and required B-10 loadin g for metallic plates are as specified in Table M2-4b: March 20 1 8 R e vision 3 7 2-1004 A m e ndm e nt N o. 15 Page M.6-1 I All indicated changes are in response to RAI 7-3 M.6.3.2 Package Regional Densities The Oak Ridge National Laboratory (ORNL) SCALE code package [6-1] contains a standard materia l data library for common elements , compounds, and mixtures. All the materials used for the TC and canister analysis are available in this data library. Table M.6-5 provides a complete list of all the relevant materials used for the criticality evaluation.

The material density for the B-10 in the poison plates includes a 10% reduction, a 25% reduction for the PRAs , and a [ ,.__ ____________________________

__. March2018 R ev ision 3 7 2-1004 A m e ndm e nt N o. 15 1 Pa ge M.6-5 I All i ndicated changes are in response to RAI 7-3 The TC is modeled with KENO V.a using the available geometry input. This option allows a model to be constructed that uses regular geometric shapes to define the material boundaries. The following conservative assumptions are also incorporated into the criticality calculations

1. No burnable poison s are accounted for in the fuel , 2. Water density is at optimum moderator density , 3. Unirradiated fuel -no credit taken for fissile depletion due to burnup or fission product poison mg , 4. The maximum lattice average fuel enrichment is modeled as uniform everywhere throughout the assembly. Natural uranium blankets and axial or radial enrichment zones are modeled as enriched uranium. It is assumed that the fuel assemblies are of uniform enrichment everywhere.

5. All fuel rods are filled with 100% moderator in the pellet/cladding gap , 6. Only the active fuel length of each assembly type is explicitly modeled with reflective boundary conditions on the ends; therefore the model is effectively infinitely long , 7. The neutron shield and stainless steel skin of the TC are stripped away and the infinite array of TCs are pushed close together with moderator in the interstitial spaces , 8. The least material condition (LMC) i s assumed for the fuel compartment and fuel structure assemblies are pushed together toward the center maximizing reactivity; the reduction in steel thickness also reduces neutron absorption in the steel in the basket , 9. The transition rails between the basket and the canister shell are modeled as 100% aluminum. Steel and open space in the transition rails reduces reactivity because the s e materials have much higher absorption cross sections as compared to the aluminum , 10. Only 90% credit is taken for the B-10 in the poison plates and 75% credit for the B-10 in PRAs are credited in the KENO models , 11. Temperature is 20°C (293K), 12. Ninety six percent theoretical density for fuel which conservatively increases the total fuel content in the model, and 13. All stainless steel, including XM-19 , is modeled as SS304; the small differences in the composition of the various stainless steels have no effect on results of the calculation.

14. The most reactive geometry remains unaffected for the 16 poison plate , 2 4 and 32 poison plate configurations. 15.[ Mar c h 2018 R e vi s ion 3 72-1004 A m e ndm e nt N o. 1 5 D 1 Pa ge M.6-7 I All indicated changes are in response to RAI 7-3 plate/tube thickness , minimum fuel compartment width, minimum assembly-to-assembly pitch and uniform maximum planar enrichment.

The following analysis uses this configuration to determine the maximum allowed initial enrichment as a function of ba sket type and soluble boron loading for each assembly class. All four poison plate configurations (20 poison plate configuration, 16 poison plate configuration, 24 poison plate configuration, and 32 poison plate configuration) are evaluated in these calculations.

Calculations at 2500 ppm soluble boron are performed for all assembly classes. For the CE 14xl4, WE 14x14, and CE 15x15 assembly classes, calculations are also performed for a range of boron soluble loadin gs (1800-2500 ppm) for the 16 poison plate and the 24 poison plate configurations.

Ca lculation s are performed for the CE I 5xl 5 and WE 1 7x l 7 loaded in the 32 poison plate basket design. The most reactive assembly type for each assembly class is used for each evaluation. In addition , for each case the internal moderator density is varied to determine the peak reactivity for the specific configuration.

The maximum initial enrichment for each assembly class and PRA configuration are provided in Table M.6-1. The canister/TC model for this evaluation differs from the actual design in the following ways:

• the B-10 content in the poison plates is I 0% lower than the minimum required,

• the B-10 content in the PRAs is 25% lower than the minimum required,

• the Ag content in the AIC-PRA is 25% lower than minimum required,

• the stainless steel/aluminum transition rails that provide support to the fuel compartment grid are modeled as various so lid materials to determine the most reactive condition,

• BPRAs , when modeled, are modeled as solid 11 B 4 C in the guide tubes and instrument tubes ,

• the neutron shield and the skin of the TC are conservatively replaced with water between the TCs , and

• the worst-case material conditions , as determined in Section M.6.4.2 above, are modeled. The input file for the case with the highest calculated reactivity is included in Section M.6.6.4. March 2018 R e vision 3 72-1004Am e ndm e nt No. 15 Pag e M.6-JOb I All indicated changes are in response to RAI 7-3 WE 17xl 7 Class Assemblies The most reacti ve WE 17x17 class assembly is the WE 17x17 LOPARJstandard assembly as demonstrated in Table M.6-6. The results for the WE 17xl 7 class assembly calculations for the 20 poison plate configuration are listed in Table M.6-13 and Table M.6-14 for cases without and with BPRA s, respectively.

The results for the WE l 7x 17 class assembly calculations for the 16 poison plate configuration are listed in Table M.6-27 and Table M.6-28 for cases without and wit h BPRAs, respectively.

The results for the WE I 7x 17 class assembly calculations for the 24 poison plate Type A/B/C/D basket configuration are listed in Table M.6-29 and Table M.6-30 for cases without and with BPRAs, respectively.

The 24 poison plate Type AJ and A2 basket configuration results are list ed in Tables M.6-51 and M.6-57, respectively.

Additional configu,rations with the 24 poison plate desi n include B 4 C PRAs 04 and 08 in number) and AIC PRAs (04 and 08 in number). 'he AIC PRA configu,rations are evaluated at 2800 ppm soluble boron concentration, and the maximum a owa e enric ments are determined.

The results for the!AIC P RA !c oefigu,rations are given in Table M 6-63 .... ! (-P-ar_t_s_4_a_n_d__,! cm For the AIC PRA co,iflgu,rations, the poison rod radius is modeled as 0.4902 cm (0.386" diameter) in the KENO base model. Considering the initial AIC composition is 80 wt. % Ag, 15 wt. % In and 5 wt. % Cd with a density of 10.1 7 glcm 3 , the initial density of Ag is 8.14 g/c m 3 (6.14 glcm) per rod of AIC. [ J Table M6-1 (Part 3 of 3) shows the maximum initial enrichments for each co,iflgu,ration with AIC P RA. [ 1 A sensitivity evaluation is performed to demonstrate the applicability of the linear density requirement , [ } for a smaller AIC diameter. The sensitivity analysis utilizes the following model for Type Bl-r case with 4 AIC PRAs presented in Table M6-1 (Part 3 of 3):

• WE 1 7x l 7 fael assembly without CC, 4. 60 wt%, in presence of 2800 ppm boron. The maximum k eJJiS 0.93 7 0 (80% water density-2800 ppm boron), Table M6-63 (Part 4 o/5). The sensitivity analysis employs a smaller poison radius of0.4330 7 cm (0.341-in. diameter or 0.86614 cm diameter).

See AIC diameter Control Rod specification for a WE 1 7x 1 7 fuel assembly in Table 4 of NUREG I CR-6 7 59. In order to obtain a statistically equivalent k eJJ tO 0.93 7 0 , a linear Ag density of [ J is required.

Based on the results of the base and sensitivity analyses, a minimum Ag lin ear density of 2.60 g/cm is specified for AIC P RAs, which corresponds to a [ } For a nominal absorber diameter of 0.341 inches , as indicated in Figu,re Ml-2 and Table M2-4a, [ March 20 1 8 R evis ion 3 1 72-1004 Amendment No. 15 Pa ge M.6-11 I All indicated changes are in response to RAI 7-3 Criticality analysis for the 32 poison plate configu,ration is performed with and wit hout CCs. This configu,ration is evaluated at 2500 ppm and 2800 ppm soluble boron concentrations and the maximum allowable enrichments are determined.

The results for this configuration are given in Table M6-63 (Parts 2 and 3 of 5). B& W J 5xl 5 Class Assemblies The most reactive B& W l 5xl 5 class assembly is the B& W l 5xl 5 Mark B assembly as demonstrated in Table M.6-6. The results for the B& W l 5x 15 class assembly calculations for the 20 poison plate configuration are listed in Table M.6-15 and Table M.6-16 for cases without and with BPRAs , respectively.

The results for the B&W 15x15 class assembly calculations for the 16 poison plate configuration are listed in Table M.6-31 and Table M.6-32 for cases without and with BPRAs, respectively.

The results for the B&W l 5xl 5 class assembly calculations for the 24 poison plate Type A/B/C/D basket configuration are listed in Table M.6-33 and Table M.6-34 for cases without and with BPRAs , respectively.

The 24 poison plate Type Al and A2 basket configuration results are listed in Tables M.6-49 and M.6-55, respectively.

CE l 5xl 5 Class Assemblies The most reactive CE l 5x 15 class assembly is the CE l 5x l 5 Palisades assembly as demonstrated in Table M.6-6. The results for the CE l 5xl 5 class assembly calculations for the 20 poison plate configuration are listed in Table M.6-17 for cases without BPRAs. BPRAs are not authorized to be stored with CE l 5xl 5 class assemblies.

The addition of plugging cluster assemblies, i.e., steel rods, into each of the eight guide tubes of a CE 15x15 class assembly reduces the maximum reactivity of the payload. The introduction of the steel rods displaces both moderator and soluble boron within the assemblies.

The plugging clusters are assumed to extend approximately 1 inch into the top of the assembly's active fuel region, and the resulting change in the maximum reactivity is less than the statistical uncertainties of the calculations.

To demonstrate the effect of displacing the borated water on system reactivity, CE l 5xl 5 Palisades cases with the highest fuel enrichments and highest soluble boron loadings (2300, 2400, and 2500 ppm boron) were reevaluated with steel in the g uide tubes. Two scenarios were evaluated:

full length steel rods and l inch long steel rods. The March 20 1 8 Revision 3 72-1004AmendmentNo.

15 Page M6-Jl a I All indicated changes are in response to RAI 7-3 Table M.6-5 Material Property Data Material Density Element Weight% g/cm 3 U-235 3.00 U02 10.52 U-238 85.15 (Enrichment

-3.4 wt%) 0 11.85 U-235 4.41 U02 10.52 U-238 83.73 (E nrichm ent -5.0 wt%) 0 11.86 Zr 98.23 Sn 1.45 Zircaloy-4 6.56 Fe 0.21 Cr 0.10 Hf O.ol H 0.11 Borated Water 0 0.8 9 (2500 ppm Boron) 1.000 4.602E-04 B-10 B-11 2.038E-03 H 11.1 Water 0.998 88.9 0 C 0.080 Si 1.000 p 0.045 Stain l ess Steel (SS3 04) 7.94 Cr 19.000 Mn 2.000 Fe 68.375 Ni 9.500 Aluminum 2.70 Al 100.0 Lead 11.34 Pb 100.0 Aluminum -Boron Poison Plate B-10 1.34 (0.0063 g/cm 2 B-10 Type A/B/C/D) 2.465(I) Al 98.66 Aluminum-Boron Poison Plate B-10 2.88 2.465(I) (0.0135 g/cm 2 B-10 Type Al) Al 98.60 Aluminum-Boron Poison Plate B-10 3.84 (0.0200 g/cm 2 B-10 Type A2) 2.465(I) Al 95.74 B-10 14.42 B 4 C in PRA 0.756 B-11 63.83 C 21.75 11 B 4 C in BPRA B-11 78.56 2.555 C 21.44 AIC PR Af::1 Ag-10 7 N I A N I A Ag-109 N I A Note: Atom Density (atoms/b-cm) 8.0797E-04 2.2666E-02 4.6948E-02 1.1882E-03 2.2290E-02 4.6956E-02 4.2541E-02 4.8254E-04 1.4856 E-04 7.5978E-05 2.2133E-06 6.6769E-02 3.3385E-02 2.7713 E-05 1.l 155 E-04 6.6769E-02 3.3385E-02 3.1877 E-04 l.7025E-03 6.9468E-05 l.7473E-02 l.7407E-03 5.8545E-02 7.7402E-03 6.0307E-02 3.2969E-02 1.9847 E-03 5.4276E-02 4.2606E-03 5.3 203E-02 5.6809£-03 5.2618E-02 6.5599E-03 2.6405£-02 8.241 lE-03 1.0988£-01 2.7470£-02 [ J [ 1 (1) Note in some models the a l uminum-boron poi so n i s modeled w ith a density of2.693 g/c m 3 (see the s ample input in Section M.6.6.5 , al thou h the number den s i of B-10 i s equiva l e nt. (2) Ag co mp osition c r ed ited in KE NO base mod e l. March 2018 Revision 3 72-1004 Amendment No. 15 Page M.6-34 I All indicated changes are in response to RAI 8-3 N.2.1 Spent Fuel to be Stored There are two design configurations for the NUHOMS-24PHB DSC: the 24PHBS and 24PHBL, which are nearly identical to the standard and long cavity 24P DSCs, respectively.

Each of the DSC configurations is designed to store 24 intact PWR fuel assemblies, including reconstituted assemblies or up to 4 damaged and balance intact PWR fuel assemblies (including reconstituted) with characteristics described in Table N .2-1. The 24PHB DSC is designed to store intact or damaged B&W I 5x15 , intact WE 17xl 7, intact WE I 5xl 5 , intact CE 14xl4 , and intact WE 14xl4 Class PWR fuel assemblies as specified in Table N.2.1. Control Components (CCs) and damaged fuel assemblies are allowed only in the B&W 15x15 class fuel assembly.

Replacement assemblies by other manufacturers are also allowed provided they meet limiting features listed in Table N.2-1. Damaged PWRfuel assemblies are assemblies containing missing or partial fuel rods,fuel rods with known or suspected cladding defects greater than hairline cracks or pinhole leaks. The extent of damage in the fuel assembly , including non-cladding damage , is to be limited such that a fuel assembly is able to be handled b normal means and retrievabili is ensured ollowin normal and off-normal conditions.

The extent of damage in the fuel rods is to be limited such that a fuel pellet is not able to pass through the damaged cladding during handling and retrievability is ensured following normal and off-normal conditions.

The DSC bas et cells that store damaged fuel assemblies are provided with top and bottom end caps to ensure retrievability. The NUHOMS-24PHB DSC may store PWR fuel assemblies arranged in one of two alternate Heat Load Zoning Configurations with a maximum decay heat of 1.3 kW per assembly and a maximum heat load of 24 kW per DSC. The Heat Load Zoning Configurations are shown in Figure N.2-1 and Figure N.2-2. The NUHOMS-24PHB DSC is vacuum dried and backfilled with helium at the time of loading. The maximum (bounding) fuel assembly weight of 1682 lbs with a CC is identical to the NUHOMS-24P DSC design. The maximum fuel cladding temperature limit of 400 °C (752 °F) is applicable to normal conditions of storage and all short term operations from spent fuel pool to ISFST pad including vacuum drying and helium backfilling of the 24PHB DSC per the guidance provided in NUREG-1536 [2.1). In addition , NUREG-1536 does not permit thermal cycling of the fuel cladding with temperature differences greater than 65 °C (117 °F) during DSC drying , backfilling and transfer operations.

The maximum fuel cladding temperature limit of 570 °C (1058 °F) is applicable to accidents or off-normal thermal transients

[2.1]. The information provided in Table N .2-1 is based on the design basis B& W l 5x 15 fuel which is the bounding fuel assembly.

The types of spent fuel considered in Appendix N include the following:

• B&W 15x15 MarkB2 , B3, B4, B4Z , BZ , B5 , B5Z, B6, B7, B8, B9, BIO , BIOD , Bl OE , B 1 OF , B 1 OG , B 1 OL , B 11 , and B 11 A fuel assemblies , with or without CC s.

• B&W 15xl5 reconstituted fuel assemblies with a maximum of IO stainless steel rods per assembly or unlimited number of lower enrichment U0 2 rods instead of zirconium-alloy clad enriched U0 2 rods. The stainless steel rods are assumed to have two thirds the irradiation time as the zirconium-alloy rods of the assembly.

The reconstituted U0 2 rods are assumed to have the same irradiation history as the entire fuel assembly.

The reconstituted rods can be at M arch2018 R e vi s ion 3 7 2-1004Am e ndm e nt N o. 15 Pa g e N.2-2 I All indicated changes are in response to RAI 8-3 T.1.1 Introduction The NUHOMS-61BTH S y stem is designed to store up to 61 intact (including reconstituted) or up to 16 damaged with up to 4failedfuel cans (FFCs) loaded with failed fuel with the remainder intact BWR fuel assemblies with or without fuel channels.

Alternatively, 61 damaged fuels can be stored in the NUHOMS-61BTH DSC as shown in Figure T.2-9. The fuel to be stored is limited to a maximum initial lattice average initial enrichment of 5.0 wt.%, a maximum assembly average burnup of 62 GWd/MTU , and a minimum cooling time of 3.0 years. The design characteristics , including physical and radiological parameters of the payload, are described in Appendix T.2. Reconstituted assemblies containing up to 10 replacement stainless steel rods per assembly or up to 61 lower enrichment U0 2 rods instead of Zircaloy clad enriched U0 2 rods are acceptable for storage in 61BTH DSC as intact fuel assemblies with a slightly longer cooling time than that required for a standard assembly.

The maximum number ofreconstituted fuel assemblies per DSC is four with stainless steel rods or 61 with U0 2 rods. Provisions have been made for storage of up to 61 damaged fuel assemblies in lieu of an equal number of intact assemblies in cells located at the outer edge of the 61BTH basket. Damaged BWR fuel assemblies are assemblies containing missing or partial fuel rods , fuel rods with known or suspected cladding defects greater than hairline cracks or pinhole leaks. The extent of damage in the fuel assembly, including non-cladding damage, i s to be limited su c h that a fu e l assembly is able to be handled by normal means and th e retrievability is ensur e d following the normal and off-normal conditions.

The extent of damage in th e fuel rod s is to be limited such that a e l p e llet i s not a e to pass through the damaged cladding during handling and retrievability is ensured following normal and off-normal conditions. The DSC basket cells that store damaged fuel assemblies are provided with top and bottom end caps to ensure retrievability.

Provisions have also been made for storage of up to four failed fuel assemblies in the corner cells , along with up to 12 damaged fuel assemblies in the cells located at the outer periphery of the 61BTH basket and balance intact as described in Appendix T.2. The NUHOMS-61BTH System consists of the following new or modified components:

• A 61 BTH DSC, with two alternate configurations , designated as Type 1 61BTH DSC or Type 2 61BTH , is described in detail in Section T.1.2. It provides confinement , an inert environment , structural support, heat rejection, and criticality control for the 61 BWR fuel assemblies ,

• A modified HSM-H module, as described in Section T.1.2 , or HSM Model 80/102/152/202 , with no modifications to the configuration as described in UFSAR Chapter 1 , is provided for environmental protection , shielding and heat rejection during storage ,

• An OS 197 or OS 197H Transfer Cask (TC) with no modifications to the configuration as described in UFSAR Chapter 1, or a modified version of the 0Sl97FC TC , designated as OS 197FC-B , described in Section T.1.2 , is provided for on site transfer of the 61 BTH DSCs ,

• An upgraded version of the HSM-H , designated as HSM-HS, is provided to allow storage of the NUHOMS-61BTH DSC in locations where higher seismic levels exist. The HSM-HS design configuration , described in Appendix U.1 , is modified to accommodate the smaller diameter of the NUHOMS-61BTH DSC , and March 20 1 8 R ev i s ion 3 7 2-1004 A m e ndm e nt No. 15 P age T.1-3 I All indicated changes are in response to RAI 8-3 U.1.1 Introduction The NUHOMS-32PTH1 System is designed to store up to 32 (including reconstituted)

B&W 15xl5 , WE 17xl 7 , CE 15xl5 , WE 15xl5 , CE 14x14 , and WE 14x14 class PWR fuel assemblies. The fuel to be stored is limited to a maximum assembly average initial enrichment of 5.0 wt. % U-235, a maximum assembly average burnup of 62 GWd/MTU , and a minimum cooling time of 2.0 years. Each of the 32PTH1 DSC types is designed to store up to 32 Control Components (CCs) which include burnable poison rod assemblies (BPRAs), thimble plug assemblies (TPAs), control rod assemblies (CRAs), rod cluster control assemb l ies (RCCAs), axial power shaping rod assemblies (APSRAs), orifice rod assemblies (ORAs), vibration suppression inserts (VSis), and neutron source assemblies (NSAs). The design characteristics , including physical and radiological parameters of the payload , are described in Chapter U.2. Reconstituted assemblies containing up to 10 replacement irradiated stainless steel rods per assembly or an unlimited number of lower enrichment U0 2 rods instead of Zircaloy clad enriched U0 2 rods or Zr rods or Zr pellets or unirradiated stainless steel rods are acceptable for storage in 32PTH1 DSC as intact fuel assemblies.

The maximum number of reconstituted fuel assemblies with irradiated stainless steel rods per DSC is four. Provisions have been made for storage of up to 16 damaged fuel assemblies in lieu of an equal number of intact assemblies in the cells located at the center of the 32PTH1 basket. Damaged PWR fuel assemblies are assemblies containing missing or partial fuel rods , fuel rods with known or suspected cladding defects greater than hairline cracks or pinhole leaks. The extent of damage in the fuel assembl y, including non-cladding damage , is to be limit e d such that a fuel assembly is able to be handled by normal means and the retrievability is ensured following the normal and off-normal conditions.

The extent of damage in the fuel rods is to be limited such that a fuel pellet is not a e to pass through the damaged cladding during handling and retrievability is ensured following normal and off-normal conditions. The DSC basket cells that store amaged fuel assemblies are provided with top and bottom end caps to assure retrievability. Provisions have also been made for storage of up to four failed fuel cans (FFCs) in cells located at the corners of the interior 4x4 compartment cells of the 32PTH1 basket or up to 16 FFC s in a checkerboard pattern, as described in Chapter U.2. The NUHOMS-32PTH1 System consists of the following new or modified components

• A 32PTH1 DSC, with three alternate configurations, described in detail in Section U.1.2, provides confinement, an inert environment , structural support , and criticality control for the 32 PWR fuel assemblies ,

• A modified HSM-H module , described in Section U.1.2 , is provided for environmental protection , shielding and heat rejection during storage , and

• OS200 or OS200FC TC for onsite transfer of the 32PTH1 DSCs. The NUHOMS-32PTH1 System requires the use of non-safety related auxiliary transfer equipment similar to those described in Section 1.3.2.2 (for OS200 TC) and Appendix P (for OS200FC TC) of the UFSAR. There i s no change to any of the design features of the auxiliary M ar c h 2 01 8 R e vi s ion 3 72-1004 Am e ndm e nt N o. 1 5 P ag e U.1-3 I All indicated changes are in response to RAI 8-3 Fuel assemblies that contain fixed integral non-fuel rods are also considered as intact fuel assemblies.

These fuel assemblies are different than reconstituted assemblies because fuel rods are not "replaced" by non-fuel rods, rather the non-fuel rods are pat1 of the initial fuel design. The non-fuel rods displace the same amount of moderator, with zirconium-alloy ( or aluminum) cladding and typically contain burnable absorber (or other non-fuel) material.

The radiation and thermal source terms for the non-fuel rods are significantly lower than those of the fuel rods since there is no significant radioactive decay source. The internal pressure of the non-fuel rods after irradiation is lower than those of the fuel rods since there is no fission gas generation.

The reactivity of the fuel rods (from a criticality standpoint) is significantly higher than that of fuel rods. In summary, the mechanical , thermal, shielding, and criticality evaluations for these rods are bounded by those of the regular fuel rods. Therefore, no further evaluations are required for the qualification of these fuel assemblies.

Reconstituted assemblies containing up to IO replacement irradiated stainless steel rods per assembly or an unlimited number of lower enrichment U0 2 rods instead of Zircaloy clad enriched U0 2 rods, or Zr rods or Zr pellets, or unirradiated stainless steel rods are acceptable for storage in 32PTH1 DSC as intact fuel assemblies. The stainless steel rods are assumed to have two-thirds the irradiation time as the remaining fuel rods of the assembly.

The reconstituted U0 2 rods are assumed to have the same irradiation history as the entire fuel assembly.

The reconstituted rods can be at any location in the fuel assemblies.

The maximum number of reconstituted fuel assemblies per DSC is four with irradiated stainless steel replacement rods or 32 with U0 2 replacement rods. The NUHOMS-32PTH1 DSCs can also accommodate up to a maximum of 16 damaged fuel assemblies placed in the center cells of the DSC as shown in Figure U.2-1 through Figure U.2-3. Damaged PWR fuel assemblies are assemblies containing missing or partial fuel rods , fuel rods with known or suspected cladding defects greater than hairline cracks , or pinhole leaks, including non-cladding damage. The extent of damage in the fuel assembly is to be limited such that a fuel assembly is being able to be handled b normal means and retrievabili is assured following normal and off-normal conditions.

The extent of damage in the fuel rods is to be limited such that a fuel pellet is not able to pass through the damaged claddin during handling and retrievability is ensured following normal and o -normal conditions.

he DSC basket cells w 1c store amage ue assemblies are provided with top and bottom end caps to assure retrievability.

The NUHOMS-32PTH1F DSC, an alternate version of the NUHOMS-32PTH1 DSC, is designed to accommodate failed fuel in up to a maximum o f four failed fuel cans (FFCs) placed in the corner cells of the interior 4x4 compartment cells of the basket, as shown in Figure U.2-5 or up to 16 FFCs in a checkerboard loading as shown in Figure U.2-3. Failed fuel is defined as fuel rods that have been removed from a fuel assembly and placed in a secondary container , such as a rod storage basket. Failed fuel may contain breached rods, grossly breached rods , and other defects such as missing or partial rods , missing grid spacers, or damaged spacers to the extent that the assembly cannot be handled by normal means. Individual fuel rods that are not failed can be stored directly in the FFC without a secondary container such as an RSB. The maximum number of fuel rods that may be stored in a failed fuel can is 100, with a total uranium loading limited to 250 kg initial uranium. The total weight of the failed fuel can plus all its contents shall be less than 1715 lbs, 1625 lbs, and 1665 lbs , for the 32PTH1-L , 32PTH1-M , and 32PTH1-S DSCs, respectively. M ar c h 2018 R e vi s ion 3 72-1004 Am e ndm e n t N o. 1 5 Pa ge U.2-2a I All indicated changes are in response to RAI 8-3 Y.2.1 Spent Fuel to be Stored Y.2.1.1 Intact or Damaged Fuel As described in Appendix Y.l, the NUHOMS-69BTH DSC is designed to store intact (including reconstituted) and/or damaged (boiling water reactor) BWR fuel assemblies as specified in Table Y.2-1 and Table Y.2-2. The fuel to be stored is limited to a maximum lattice average initial enrichment of 5.0 wt. % U-235. The maximum allowable fuel assembly average burnup is limited to 62 GWd/MTU. The minimum required cooling time for fuel to be stored with 170 kgU/F A and J 98 kgU/F A is explicitly specified as a function of burnup and enrichment in Tables Y.2-5 through Y.2-17b. For fuel with a kgU/FA loading between these two values , the minimum required cooling time for fuel to be stored as a function of burn up and enrichment is determined by using the instructions provided in the the notes and examp l es following Table Y.2-16. The NUHOMS-69BTH DSC is also authorized to store fuel assemblies containing blended low enriched uranium (BLEU) fuel material.

Fuel pellets containing BLEU fuel material are no different than U02 fuel pellets except for the presence of a higher quantity of cobalt impurity.

The consideration of cobalt impurity affects only the gamma source terms for fuel assemblie s located in the DSC periphery.

This does not affect any criticality , thermal or structural analysis inputs for evaluation of fuel assemblies with BLEU material.

The qualification of fuel assemblies containing BLEU fuel pellets will require an additional coo lin g time of three years to ensure that the source terms calculated with U0 2 material are bounding.

Reconstituted fuel assemblies containing up to IO replacement irradiated stainless steel rods per assembly or 69 lower enrichment U0 2 rods instead of zircaloy clad enriched U0 2 rods are acceptable for storage in 69BTH DSCs as intact fuel assemb li es. The stainless steel rods are assumed to have two-thirds the irradiation time as the remaining fuel rods of the assembly.

The reconstituted U0 2 rods are assumed to have the same irradiation history as the entire fuel assembly. The reconstituted rods can be at any location in the fuel assemblies. The maximum number of reconstituted fuel assemblies per DSC is four with irradiated stainless steel rods or 69 with U0 2 rods or Zr rods or Zr pellets or unirradiated stainless steel rods. The NUHOMS-69BTH DSCs can also accommodate up to a maximum of24 damaged fuel assemblies placed in the four outer "six compartment" arrays located at the outer edge of the DSC as shown in Figure Y.2-7. Damaged BWR fuel assemblies are assemblies containing missing or partial fuel rods , fuel rods with known or suspected cladding defects greater than hairline cracks or pinhole leaks. The extent of damage in the fuel assembly is to be limited such that the fuel assembly , including non-cladding damag e , is able to be handled b y normal means and the retrievabili is as s ured ollowin the normal and o -normal conditions.

Missin fuel rods are allowed. The e x tent of damage in the fuel rods is to be limited such that a fuel pellet is not able to pass through th e damaged cladding during hand/in and retrievabili is ensured following normal and off-normal conditions. The DSC basket cells which store damaged fuel assem 1es are prov1 e wit top an ottom end caps. A 69BTH DSC containing less than 69 fuel assemblies may contain dummy fuel assemblies in the empty slots. The dummy assemblies are unirradiated , aluminum blocks that approximate the weight and center of gravity of a fuel assembly.

M arch 2018 R e vision 3 7 2-1004 A m e ndm e nt N o. 15 Pa ge Y.2-2 I All indicated changes are in response to RAI 8-3 The stainless steel rods are assumed to have two-thirds the irradiation time as the remaining fuel rods of the assembly.

The reconstituted U0 2 rods are assumed to have the same irradiation history as the entire fuel assembly.

The reconstituted rods can be at any location in the fuel assemblies.

The maximum number of reconstituted fuel assemblies per DSC is four with irradiated stainless steel replacement rods or 37 with U0 2 replacement rods. The NUHOMS-37PTH DSCs can also accommodate up to a maximum of four damaged fuel assemblies placed in the outer cells of the DSC as shown in Figure Z.2-2 and Figure Z.2-3. Damaged PWR fuel assemblies are assemblies containing missing or partial fuel rods, fuel rods with known or suspected cladding defects greater than hairline cracks , or pinhole leaks. The extent of damage in the fuel assembly , including non-cladding damage, is to be limited such that a fuel assembly will still be able to be handled by normal means and the retrievaµ.LL* LL. :JL....L:.

• ~==.1.., following the normal and off-normal conditions.

Missing fuel rods are allowed. The extent of damage in the fuel rods is to be limited such that a fuel pellet is not able to pas s through the damaged cladding durin hand/in and retrievabili is ensured ollowin normal and o -normal conditions. The DSC basket cells which store damaged fuel assemblies are provided with top and bottom end caps. A 37PTH DSC containing less than 37 fuel assemblies may contain dummy fuel assemblies in the empty slots. The dummy assemblies are unirradiated, stainless steel encased structures that approximate the weight and center of gravity of a fuel assembly.

The 37PTH DSC basket is designed with solid aluminum transition rails for support and to facilitate heat removal , since the solid aluminum rails allow a more direct heat conduction path from the basket edge to the DSC shell. The NUHOMS-37PTH DSCs may store up to 37 PWR fuel assemblies arranged in any of the two alternate heat load zoning configurations (HLZC) as shown in Figure Z.2-2 and Figure Z.2-3. The maximum decay heat per fuel assembly and the maximum canister heat load allowed is as specified in Figure Z.2-2 and Figure Z.2-3. The maximum allowed heat load for the various 37PTH system configurations are presented in Table Z.2-17. The NUHOMS-37PTH DSC configuration is analyzed for three alternate DSC basket designs for critica l ity control. In Option 1, the poison plate configuration consists of a 0.05in. nominal thickness aluminum plate in combination with a 0.075in. nominal thickness neutron absorber plate. Opt i on 2 and Option 3 consist of a single neutron absorber plate of 0. 125 in. and 0.105 in. nominal thickness , respectively. Option 1 results in the most reactive configuration.

In addition , the NUHOMS-37PTH DSC basket is provided with three alternate neutron absorber plate materials (poison material) for criticality control: borated aluminum alloy , boron carbide/aluminum metal matrix composite (MMC) and Baral. For criticality analysis , 90% of B 10 content present in the borated aluminum and MMC poison plates is credited , while only 75% is credited for Baral. The minimum B-10 poison loadings allowed are presented in Table Z.2-16. The use of Baral is restricted to Option 1 DSC basket designs only. The selection of the poison material does not have any impact on the thermal analysis , since it is based on the limiting thermal conductivity of Baral as discussed in Appendix Z.4 , Section Z.4.3. Mar c h2018 R e vision 3 72-100 4 A m e ndm e nt N o. 15 Pa ge Z.2-3 I Enclosure 6 to E-51144 Coe 1004 Amendment 15, Revision 3 Technical Specifications Changed Pages All indicated changes are in response to RAI 8-3 Tables Table 1-1e PWR Fuel Specifications for Fuel to be Stored in the NUHOMS-32PT DSC PHYSICAL PARAMETERS:

Fuel Assembly Class Intact (including reconstituted) or damaged or failed B&W 15x15, WE 17x17 , CE 15x15 , WE 15x15, CE 14x14 and WE 14x14 class PWR assemblies that are enveloped by the fuel assembly design characteristics listed in Table 1-1 f. Damaged and/or failed fuel assemblies beyond the definitions contained below are not authorized for storage. 32 assemblies per DSC with up to 56 irradiated Reconstituted Fuel Assembl ies stainless steel rods per assembly or unlimited number of lower enrichment U0 2 rods per assembly.

Damaged PWR fuel assemblies are assemblies containing missing or partial fuel rods , fuel rods with known or suspected cladding defects greater than hairline cracks or pinhole leaks. The extent of damage in the fuel assembly , including non-cladding damage , is to be limited in such a way that a fuel assembly is able to be handled by Fuel Damage normal means. Missina fuel rods are allowed.rFt;;;" extent of damage in the fuel rods is to be limited such that a fuel pellet is not able to pass through the damaged cladding during handling and retrievability is ensured following normal and off-normal conditions.

1 Damaged fuel assemblies shall also contain top and bottom end fittings or nozzles or tie plates depending on the fuel type. Failed fuel is defined as ruptured fuel rods , severed fuel rods , loose fuel pellets , or fuel assemblies that cannot be handled by normal means. Fuel assemblies may contain breached rods , grossly breached rods , and other defects such as missing or partial rods , missing grid spacers , or damaged spacers to the extent that the assembly cannot be handled by normal means. Fuel debris and fuel rods that have been removed from a fuel assembly and placed in a rod storage Failed Fuel basket are also considered as failed fuel. Loose fuel debris , not contained in a rod storage basket must be placed in a failed fuel can for storage , provided the size of the debris is larger than the failed fuel can screen mesh opening and it is located at a position of at least 4" above the top of the bottom shield plug of the DSC. Fuel debris may be associated with any type of U0 2 fuel provided that the maximum uranium content and initial enrichment limits are met. The total weight of each failed fuel can plus all its contents shall be Jess than 1682 lb. Standardized NUHOMS System Technical Specifications (continued)

T-5 CoC 1004 Am endme nt 15 Application , Revision 3 All indicated changes are in response to RAI 7-3 Tables Table 1-1h Specification for the NUHOMS-32PT Poison Plates and PRAs NUHOMS-32PT Number of Minimum B-10 Areal DSC Basket Type B 4 C PRAs r 1 H 2 J Density, gmlcm 2 A 0 0.0070 A1 , A 1-32 0 0.0150 A2 , A2-32 0 0.0200 8 4 0.0070 81 4 0.0150 82 4 0.0200 C 8 0.0070 C1 8 0.0150 D 16 0.0070 NUHOMS-32PT Number of Minimum B-10 Areal DSC Basket Type AIC PRAs r 1 H 3 J Density, gm!cm 2 81-r 4 0.0150 82-r 4 0.0200 C1-r 8 0.0150 C2-r 8 0.0200 Notes: (1) Figure 1-5 , Figure 1-6 and Figure 1-7 provide the required PRA configurations (2) PRAs with Boron Carbide absorber are specified as B4C PRAs. Minimum B4C content per absorber rod is 0. 79 grams/cm (3) PRAs with Silver-Indium-Cadmium absorber are s ecified as AIC PRAs. Minimum Silver content per absorber rod is 2. 60 g/cm Standardized NUHOMS System Technical Specifications CoC 1004 Am e ndment 15 Application , Revision 3 T-13 All indicated changes are in response to RAI 8-3 Tables Table 1-1i PWR Fuel Specifications for Fuel to be Stored in the Standardized NUHOMS-24PHB DSC PHYSICAL PARAMETERS Fuel Class Intact or damaged, unconsolidated B&W 15x15 (with or without CCs), intact WE 17x17 , intact WE 15x15, intact CE 14x14 and intact WE 14x14 Class PWR fuel assemblies (all without CCs) or equivalent reload fuel manufactured by other vendor , with the following requirements

Damaged fuel assemblies beyond the definition contained below are not authorized for storage. Maximum Number of Irradiated Stainless 40 Steel Rods in Reconstituted Assemblies per DSC Maximum Number of Irradiated Stainless 10 Steel Rods per Reconstituted Assembly Maximum Number of Reconstituted 24 Assemblies per DSC with Low Enriched Uranium Oxide Rods Fuel Damage Damaged PWR fuel assemblies are assemblies containing missing or partial fuel rods , fuel rods with known or suspected cladding defects g r eater than hairline cracks or pinhole leaks. The extent of damage in the fuel assembly , including non-cladding damage , is to be limite d such that a fuel assembly is able to be handled by normal means. Missing fuel rods are allowed. I The extent of dam age in the fuel rods is to be limited such that a fuel pellet is not able to pass through the damaged cladding during handling and retrievability is ensured following normal and off-normal conditions

.I Damaged tuel assemblies shall also contain top and bottom end fittings or nozzles or tie plates depending on the fuel type. Control Components

• Up to 24 CCs are authorized for storage in 24PHBL DSCs only.

• Authorized CCs include Burnable Poison Rod Assemblies (BPRAs), Thimble Plug Assemblies (TPAs), Control Rod Assembl ies (CRAs), Rod Cluster Control Assemblies (RCCAs), Axial Power Shaping Rod Assemblies (APSRAs), Orifice Rod Assemblies (ORAs), V i bration Suppression Inserts (VSls), Neutron Source Assemblies (NSAs), and Neutron Sources. Non-f uel hardware that are positioned within the fuel assembly after the fuel assembly is discharged from the core such as Guide Tube or Instrument Tube Tie Rods or Anchors , Guide Tube Inserts , BPRA Spacer Plates or devices that are positioned and operated within the fuel assembly during react or opera tion such as those listed above are also considered as CCs.

• Design basis thermal and radiolo~ical characteristics for the CCs are listed in Table 1-1n (1. Physical Parameters (without CCs) Maximum Assembly Length (unirradiated, 165.785 in (Standard Cavity) intact assembly with Maximum Burnup :5 55 171.23 in (Long Cavity) GWd/M T U) Maximum Assembly Length (unirradi ated , 165.785 in (Stan dard Cav ity) damaged assembly with Maximum Burnup :5 171.23 in (Long Cavity) 45 GWd/MTU) (continued)

Standardized NU HOMS System Technical Specifications T-14 CoC 1004 Amendment 15 Application , Revisi on 3 All indicated changes are in response to RAI 8-3 Tables Table 1-1j BWR Fuel Specification of Damaged Fuel to be Stored in the Standardized NUHOMS-6181 DSC PHYSICA L PARAMETERS:

7x7 , 8x8 BWR damaged fuel assemblies that are enveloped by the Fuel assembly des i gn characteristics Fuel Design listed in Table 1-1d for the 7x7 and 8x8 designs only. Damaged fuel assemblies beyond the definition contained below are not authorized for storage. Claddina Material Zircaloy Damaged BWR fuel assemblies are fuel assemblies containing fuel rods with known or suspected claddin~ defects qreater than hairline cracks or pinhole leaks. The extent of damage in the fuel assembly , including non-cladding damage , is to be limited such that the fuel assembly is able to be handled by normal means. The Fuel Damage extent of damage in the fuel rods is to be limited such that a fuel pellet is not able to pass through the damaged cladding during handling and retrievability is ensured following normal and off-normal conditions.

l Damaged fuel shall be stored with Top and Bottom Caps. Damaged fuel may only be stored in the 2x2 compartments of the " Type C" NUHOMS-61BT Canister described in Table 1-1k. Channels Fuel may be stored with or without fuel channels. Maximum Assemblv Lenath (un i rradiated) 176.2 in Nominal Assembly Width (excludinq channels) 5.44 in Maximum Assembly Weiqht 7051bs RADIOLOGICAL PARAMETERS:'~'

No interpolation of Radiological Parameters is permitted between qroups. Group 1 Maximum Burnuo 27,000 MWd/MTU Minimum Coolinq Time 5-years Maximum Initial Lattice Averaqe Enrichment 4.0 wt. % U-235 Maximum Pellet Enrichment 4.4 wt. % U-235 Minimum Initial Assembly Averaqe Enrichment 2.0 wt. % U-235 Maximum Initial Uranium Content 198 kq/assembly Maximum Decay Heat 300 W/assembly Group 2 Maximum Burnup 35 000 MWd/MTU Minimum Coolinq Time 8-years Maximum Initial Lattice Average Enrichment 4.0 wt.% U-235 Maximum Pellet Enrichment 4.4 wt. % U-235 Minimum Initial Assembly Averaae Enrichment 2.65 wt. % U-235 Maximum Init i al Uranium Content 198 kq/assembly Maximum Decay Heat 300 W/assembly Group 3 Maximum Burnuo 37 , 200 MWd/MTU Minimum Coolinq Time 6.5-years 1 1 1 Maximum Initial Lattice Averaqe Enrichment 4.0 wt. % U-235 Maximum Pellet Enrichment 4.4 wt. % U-235 Minimum Initial Assembly Averaqe Enrichment 3.38 wt. % U-235 Maximum Initial Uranium Content 198 kq/assembly Maximum Decay Heat 300 W/a ssembly ( continued)

I Standardized NUHOMS System Technical Specifications T-16 CoC 1004 Amendment 15 Appl icati on , Revision 3 All indicated changes are in response to RAI 8-3 Tables Table 1-11 PWR Fuel Specification for the Fuel to be Stored in the NUHOMS-24PTH DSC PHYSICAL PARAMETERS:

Fuel Class Intact or damaged or failed unconsolidated B&W 15x15 , WE 17x17 , CE 15x15 , WE 15x15 , CE 14x14 and WE 14x14 class PWR assemblies (with or without control components) that are enveloped by the fuel assembly design characteristics listed in Table 1-1 m. Damaged and/or failed fuel assemblies beyond the definitions contained below are not authorized for storaoe. Fuel Damage Damaged PWR fuel assemblies are assemblies conta i ning m i ssing or partial fuel rods , fuel rods with known or suspected cladding defects greater than hairline cracks or pinhole leaks. The extent of damage in the fuel assembly , including non-cladding damage , is to be limited such that the fuel assembly is able to be handled bv normal means. Missina fuel rods ar e allo w ed. I The e x tent of damage in t he fuel rods is to be limited su c h that a fuel p ellet is not able to pass through the dam a ge d c ladding durin g handling and r e t rievab i l it v i s ensured follo w ing normal a n d off-normal conditions. !Damaged fuel assemblies shall also contain top and bott o m end fittings or no zz les or t ie plates depend i ng on the fuel t voe. Failed F uel Failed fuel is defined as ruptured fuel rods , severed fuel rods , loose fuel pellets , or fuel assembl i es that cannot be handled by normal means. Fuel assembl i es may contain breached rods, grossly breached rods , and other defects such as m i ss i ng or part i al rods , missing grid spacers , or damaged spacers to the extent that the assembly cannot be handled by normal means. Fuel debris and fuel rods that have been removed from a fuel assembly and placed in a rod storage basket are also considered as failed fuel. L oose fuel debr i s , not contained in a rod storage basket must be placed i n a failed fuel can for storage , provided the s iz e of the deb ri s is larger than the failed fuel can screen mesh opening and i t is located a t a position o f at least 1 O" above the top of the bottom shield plug of the DSC. Fuel deb r is may be associated with any type of U0 2 fuel provided that the maximum uranium content and initial enrichment limits are met. The total we i ght of each failed fuel can plus all its contents shall be less than 1682 lb. Partial Length Shield Assemblies (PLSAs) WE 15x15 c l ass PLSAs which have only ever been irradiated in peripheral core locations with following characte ri s ti cs are autho riz ed:

• Maxi m um burnup , 40 GWd/MTU

• Min i mum cooling time , 6.5 years

• Maximum decav heat , 900 watts (continued)

Standardized NUHOMS System Technical Specifications T-19 C o C 100 4 A mend m e n t 15 App l ica tion , R e v i si o n 3 All indicated changes are in response to RAI 8-3 Tables Table 1-1t BWR Fuel Specification for the Fuel to be Stored in the NUHOMS-618TH DSC PHYSICAL PARAMETERS:

Fuel Class Fuel Damage Failed Fuel RECONSTITUTED FUEL ASSEMBLIES:

• Maximum Number of Irradiated Stainless Steel Rods in Reconstituted Assemblies per DSC

• Maximum Number of Irradiated Stainless Steel Rods per Reconstituted Fuel Assembly

• Maximum Number of Reconstituted Assembl i es per DSC with unlimited number of low enriched U0 2 rods or Zr Rods or Zr Pellets or Unirradiated Stainless Steel Rods Number of Intact Assemblies Intact or damaged or failed 7x7, 8x8 , 9x9 , 10x10 or 11 x 11 BWR assemblies that are enveloped by the fuel assembly design characteristics listed in Table 1-1 u. Damaged and/or failed fuel assemblies beyond the definitions contained below are not authorized for storaqe. Damaged BWR fuel assemblies are assemblies containing fuel rods with known or suspected cladding defects greater than hairline cracks or pinhole leaks. The extent of cladding damage in the fuel assembly , including e/adding damage , is to be limited such that a fuel assembly needs to be handled by normal means. Missing fuel rods are allowed. I The extent of damage in the fuel rods is to be limited such that a fuel pellet is not able to pass through the damaged cladding during handling and ret ri evabilitv is ensured fof/owina normal and off-normal conditions

.I Damaged fuel assemblies shall also contain top and bottom end fittings or nozzles or tie plates depending on the fuel type. Failed fuel is defined as ruptured fuel rods , severed fuel rods , loose fuel pellets , or fuel assemblies that cannot be handled by normal means. Failed fuel assemblies may contain breached rods, grossly breached rods, and other defects such as missing or partial rods , missing grid spacers , or damaged spacers to the extent that the assembly cannot be handled by normal means. Fuel debris and fuel rods that have been removed from a fuel assembly and placed in a rod storage basket are also considered as failed fuel. Loose fuel debris , not conta i ned in a rod storage basket must be placed in a failed fuel can for storage , provided the size of the debris is larger than the failed fuel can screen mesh opening and it is located at a position of at least 1 O" above the top of the bottom shield plug of the DSC. Fuel debris may be associated with any type of U0 2 fuel provided that the maximum uranium content and initial enrichment limits are met. The total weight of each failed fuel can plus all its content shall be less than 705 lb. 40 10 61 < 61 (continued)

Standardized NUHOMS System Technical Specifications T-32 CoC 1004 Am e ndment 15 Appli ca tion , Revision 3 All indicated changes are in response to RAI 8-3 Tables Table 1-1aa PWR Fuel Specification for the Fuel to be Stored in the NUHOMS-32PTH1 DSC PHYSICAL PARAMETERS:

Fuel Class Fuel Damage Failed Fuel Reconstituted Fuel Assemblies:

Intact or damaged or failed unconsolidated B&W 15x15 , WE 17x17 , CE 15x15 , WE 15x15 , CE 14 x 14 , WE 14x14 and CE 16x 16 class PWR assemblies (with or without CCs) that are envelo p ed by the fuel assembly des i gn characteristics listed in Table 1-1bb. Damaged and/or failed fuel assemblies beyond the defin i t i ons contained below are not authorized for storage. Damaged PWR fuel assemblies are assemblies contain i ng missing or partial fuel rods , fuel rods with known or suspected cladding defects greater than hairline cracks or pinhole leaks. The e x tent of damage in the fuel assembly , including none/adding damage , is to be limited such that the fuel assembly will st i ll be able to be handled by norma l means. Missina fuel rods are allowed. I The e x tent of damage in the fuel rods is to l be ltmtted such that a fuel pellet is not a ble to pass through the damaged c ladding during handling and retrievabilitv is ensured following normal and off-normal conditions

.\Damaged fuel assemblies shall also contain top and bottom end fittings or nozzles o r t i e plates depending on the f u el tvoe. Failed fuel is defined as fuel rods that have been removed from a fuel assembly , breached rods , grossly breached rods , and other defective rods. Fuel rods that have been removed from a fuel assembly may be placed in a secondary container , such as a rod storage basket. Individual fuels rods that are not failed can be stored in a failed fuel can i ster (FFC) without a secondary container such as a rod storage basket. The maximum number of fuel rods that may be stored in the FFC is 100 with a total uranium loading limited to 2.50 kg initial uranium per rod.

• Maximum Number of Irradiated Stainless Steel 40 Rods in Reconstituted Assembl i es per DSC

• Maximum Number of Irradiated Stainless Steel 10 Rods per Reconstituted Fuel Assembly

• Maximum Number of Reconstituted Assemblies 32 per DSC with unlimited number of low enriched U0 2 rods , or Zr Rods or Zr Pellets or Unirradiated Stainless Steel Rods Control Components (CCs)

• Up to 32 CCs are authorized for storage i n 32PTH1-S , 32PTH1-M and 32PTH1-L DSCs.

• Authorized CCs include burnable poison rod assembl i es (BPRAs), thimble plug assemblies (TPAs), control rod assemblies (CRAs), rod cluster contro l assemblies (RCCAs), axial power shaping rod assemblies (APSRAs), or i fice r od assemblies (ORAs), vibrat i on suppression i nserts (VSls), neutron source assemblies (NSAs) and neutron sources. Non-fuel hardware that are positioned within the fuel assembly after the fuel assembly is discharged from the core such as guide tub e or instrument tube tie rods or anchors , guide tube inserts , BPRA spacer plates or devices that are positioned and operated with i n the fuel assembly during r eactor operat i o n such as those l i sted above are also cons i de r ed as CCs.

• Design basis thermal and radiologica l characteristics for the CCs are listed in Table 1-1ee. ( continued)

Standardized NUHOMS System Technical Specifications T-42 CoC 1004 Amendment 15 Application , R e vis i on 3 All indicated changes are in response to RAI 8-3 Tables Table 1-1gg BWR Fuel Specification for the Fuel to be Stored in the NUHOMS-69BTH DSC PHYSICAL PARAMETERS

Fuel class Intact or damaged 7x7 , 8x8 , 9x9 or 1Ox10 BWR assemblies that are enveloped by the fuel assembly design characteristics listed in Table 1-1 ii. Damaged fuel assemblies beyond the definition contained below are not authorized for storage. Fuel damage Damaged BWR fue l assemblies are assemblies containing fuel rods with known , suspected cladding defects greater than hairline cracks or pinhole leaks. The extent of damage in the fuel assembly , including non-cladding damage , is to be limited such that the fuel assembly will still be able to be handled by nnrm;:il mP.;:in!':

Mic:c;inn fuel rods are ::illnwPrl I The extent of damage in the fuel rods is to be limited such that a fuel pellet is not able to pass through the damaged cladding during handling and retrievability is ensured following normal and off-normal conditions. Damaged fuel assemblies shall also contain top and bottom end fittings or nozzles or tie plates depending on the fuel type. RECONSTITUTED FUEL ASSEMBLIES

• Maximum Number of Irradiated Stainless Steel 40 Rods in Reconstituted Assemblies per DSC

• Maximum Number of Irradiated Stainless Steel 10 Rods per Reconstituted Fuel Assembly

• Maximum Number of Reconstituted Assemblies 69 per DSC with unlimited number of low enriched U0 2 rods or Zr rods or Zr pellets or Unirradiated Stainless Steel Rods Number of intact assemblies 69 Number and location of damaged assemblies Up to 24 damaged fuel assemblies, with balance intact or dummy assemblies, are authorized for storage in 698TH DSC. Damaged fuel assemblies may only be stored in the locations shown in Figure 1-37. The DSC basket cells which store damaged fuel assemblies are provided with top and bottom end caps. Channels Fuel may be stored with or without channels, channel fasteners or finger springs. Fissile Material U0 2 U0 2 198 kq/assembly Maximum assembly weiqht includinq channels 7051b (continued)

Standardized NUHOMSSystem Technical Specifications T-53 CoC 1004 Amendment 15 Application , Revision 3 All indicated changes are in response to RAI 8-3 Tab l es T abl e 1-111 PW R Fuel Specification for the Fue l to be Stored in the NUHOMS-37PTH DSC PH Y SICAL PARAMETERS:

Intact or damaged unconsolidated WE 17x17 , CE 16X16, CE 15x15 , WE 15x15 , CE 14x14 , and WE 14x14 c l ass PWR assembl i es (with or without control components) that are enveloped by the fue l assembly Fuel C l ass design characteristics listed in Table 1-1nn. Damaged fuel assemblies beyond the defin iti on contained below are not authorized for st o raqe. Damaged P WR fuel assemblies are assemblies contain in g missing or part i a l fuel rods , fuel rods with known or s u spected cladding defects greater than hairline cracks or pinhole leaks. The extent of damage in the fuel assembly , including non-cladding damage, is to be limited such that a fuel assembly is able to be handled by normal means. Missing fuel rods are Fuel Damage """'.,"'rl 1 , ne excem or aamage m me rue! rocJs ,s to be l imited such that a fuel pe ll et is not able to pass through the damaged c l adding during handling and retrievabi l ity is ensured fo ll owing normal and off-normal conditions. Damaged fue l assemblies shall also contain top and bottom end fitt i ngs or nozz l es or tie plates depending on the fuel tvoe. Reconst i t u ted Fuel Assemblies:

• Maximum Number of Irradiated Stain l ess Steel Rods 40 in Reconstituted Assemblies per DSC

• Maxim u m Number of Irradiated Stain l ess Steel Rods 10 per Reconstituted Fue l Assembly

• Maximum Number of Reconstituted Assemblies per 37 DSC w i th Unlimited Number of Low Enriched U0 2 Rods , or Zr Rods or Zr Pellets or U ni rrad i ated Sta i n l ess Steel Rods

• Up to 37 CCs are authorized for storage in 37PTH-S, and 37PTH-M DSCs.

• Authorized CCs inc l ude burnable poison rod assemb l ies (BPRAs), thimble plug assemblies (TPAs), control rod assemb l ies (CRAs), rod cluster control assemblies (RCCAs), axial power shaping rod assemb l ies (APSRAs), orifice rod assemblies (ORAs), neutron source assemblies (NSAs), vibration suppression inserts (VSls) and neutron Control Components (CCs) sources. Non-fuel hardware that are positioned within the fuel assemb l y after the fuel assembly is discharged from the core such as guide tube or instrument tube tie rods or anchors, guide tube inserts, BPRA spacer plates or devices that are positioned and operated within the fuel assembly during reactor operation such as those li sted above are also considered as CCs.

• Design basis thermal and radiological characteristics for the CCs are listed in Table 1-1aa. (continued)

Standardized NU HOMS System Technical Specifications T-59 CoC 1004 Amendment 15 Appli cat ion , Revision 3 Enclosure 7 to E-51144 Description of Amendment 15 Changes Enclosure 7 to E-51144 DESCRIPTION OF AMENDMENT 15 CHANGES

1.0 INTRODUCTION

The scope of Amendment 15 to CoC 1004 includes the changes described below. Change No. 1: Unify and standardize the fuel qualification tables for four PWR systems (32PT, 24PTH, 32PTH1 and 37PTH) in order to simplify the Technical Specifications (TS). The standardized fuel qualification tables (FQTs) provide for minimum required cooling times , as low as two years , as a funct i on of enrichment and burnup (BU) for all the heat loads described in the various heat load zoning configurations (HLZCs) for these four PWR systems. Further, the FQTs are generated for three MTU loadings per fuel assembly (FA) and allow for interpolation between MTU loadings and to establish cooling times for FAs that fall into the unanalyzed regions of the FQTs. For this purpose , the source term , dose rate , occupational exposure and site dose analyses have been revised for the four PWR systems described above. The TS and Updated Final Safety Analysis (UFSAR) Appendices M , P , U and Z have been revised accordingly.

Change No. 2: For the 32PT System , add a new HLZC #4 to allow for the loading of FAs with decay heat up to 2.2 kW corresponding to a 2-year cooled fuel. The TS and UFSAR Appendix M have been revised to incorporate this new HLZC. Change No. 3: For the 32PT System , increase the maximum assembly average BU from 55 GWd/MTU to 62 GWd/MTU. The TS and UFSAR Appendix M have been revised to incorporate this change. Change No. 4: For the 32PT System , allow for the loading of damaged fuel assemblies confined within top and bottom end caps and failed fuel assemblies loaded within individual failed fuel canisters. Provide for a basket option to increase the number of poison plates from 24 to 32 resulting in an increase in the allowable enrichment of the authorized contents. Expand the definition of the PRAs to include RCCA materials, specifically silver neutron absorber. This change also caused us to clarify the definition for damaged fuel for all DSCs in the UFSAR sections and TS tables. Additionally the TS now have a separate definition for intact fuel. The TS and UFSAR Appendix M have been revised to incorporate this change. Change No. 5: For the 32PT System , include other zirconium alloy cladding materials such as ZIRLO and M5. The TS and UFSAR Appendix M have been revised to incorporate this change. Change No. 6: For the 24PTH System , add a new HLZC #6 to allow for the loading of FAs with decay heat up to 2.5 kW corresponding to a 2-year cooled fuel , and a total heat load of 35 kW per basket. The TS and UFSAR Appendix P have been revised to incorporate this new HLZC and editorial changes are made to the TS for the descriptions of basket types. Page 1 of 5 Enclosure 7 to E-51144 Change No. 7: For the 24PTH System, the OS197 is added as an author i zed transfer cask (TC) for the transfer of the 24PTH-S-LC DSC in addition to the standardized TC. UFSAR Chapters P.1 , P.2 and P.4 have been revised to incorporate this change. Change No. 8: For the 61 BTH System , revise the exist i ng HLZC #10 to allow loading FAs with decay heat up to 1.2 kW corresponding to a 2-year cooling time. GNF-2 and ATR I UM-11 FA designs are also added as authorized contents. Additionally , the FQTs with minimum cooling times of two years are generated for MTU loadings of 0.180 and 0.198 per fuel assembly at a decay heat of 1.2 kW and to establ i sh coo l ing times for FAs that fall into the unanalyzed regions of all the FQ T s. The TS and UFSAR Appendix T have been revised to incorporate these changes. Change No. 9: For the 32PTH1 System , add new HLZC #5 to a ll ow for the l oad i ng of FAs with decay heat up to 1.1 kW for a total heat load of 35.2 kW per basket and HLZC #6 to allow for loading of FAs with decay heat up to 1.3 kW for a total heat load of 37.6 kW per basket. This is applicable for Type 1 DSCs using solid aluminum rails only. The TS and UFSAR Appendix U have been revised to incorporate these changes. Change No. 10: Provide a description in the UFSAR for the so l ar shield currently described in the TS for the TC during transfer operations. UFSAR Chapter 10 has been revised to incorporate this change. Change No. 11: Technical Specification 4.3.3 Item 11 is changed to add flexibility to general licensees in verifying compliance regard i ng the storage pad location and the soil-structure interaction , which may affect the response of loaded HSMs.

2.0 DESCRIPTION

OF THE CHANGES TO THE STANDARDIZED NU HOMS CoC 1004 TECHNICAL SPECIF I CATIONS The table below provides technical specification pages which changed , a brief description of the subject and/or change , and a reference to the scope item from Section 1.0 and applicable RSI / RAI response which relates to the change or changes. Scope RS I or TS page TS Number Description Item RAI Cover Page N/A Disclaimer regarding Amendment 14 none N/A chanQes beinQ incorporated TOC/LOT/LOF N/A Table of Contents , etc. automated none N/A updates Definition for intact , damaged and failed 1-1 1.1 fuel broken into INTACT, then 4 N/A DAMAGED/FAILED 3-7 3.1.3 HLZC #6 added for 24PTH and HLZC #5 6 , 9 N/A and #6 added for 32PTH1 P age 2 o f 5 Enclosure 7 to E-51144 Scope RSI or TS page TS Number Description I tem RAI 3-11 3.2.1 Added boron concentration tables 4 N I A references for 32PT Added flexibility to general licensees in 4-39 4.3.3 verifying compliance regarding the 11 NIA storage pad location and the soil-structure interaction 5-7 5.2.4.e Updated TC dose rates for 32PT, 24PTH , 1 NIA 32PTH1, and 37PTH Updated HSM dose rates for 32PT, 5-14 5.4.2 24PTH, 61 BTH , 32PTH1, 698TH, and 1 NIA 37PTH Removed the Table 1-1e row for " Fuel Cladding Material" in order to remove specificity to " Zircaloy" and therefore allow other c l adding materials , consistent T-5 Table 1-1e with the approach in other TS tables such 5 RSI 8-1 as TS Tab l e 1-1 i for the 24PH B system , Table 1-11 for the 24PTH System, Tab l e 1-1aa for the 32PTH1 System, and Table 1-111 for the 37PTH System, etc. Changes made to the fuel specification T-5 and T-6 Table 1-1e table for the 32PT -including intact fuel 1, 4 RAI 8-3 description and fuel damage definition clarification T-7 Table 1-1f Changes made to the FA design 4 NIA characteristics for the 32PT Changes made to the 32PT table for certain T-8 Table 1-1g fuel assemb l y parameters (intact fuel) and 4 RAI 7-5 to clarify the control component (CC) configurations Changes made to the 32PT table for T-9 and T-10 Table 1-1g1 certain basket parameters (intact fuel) 4 RAI 7-5 and to clarify the contro l component (CC) configurations New 32PT table for certain basket T-11 Table 1-1g2 parameters (damaged fuel) and to clarify 4 RAls 7-2 the control component (CC) and 7-5 configurations New 32PT table for certain basket RAls 7-2, T-12 Table 1-1g3 parameters (damaged or failed fuel) and 4 7-5 and to clarify the control component (CC) 7-6 configurations Changes made to the B 1 O specification T-13 Table 1-1h table for the 32PT , and update to the 4 RAI 7-3 minimum silver content per AIC absorber rod Page 3 of 5 Enclosure 7 to E-51144 TS page TS Number Scope RS I or Descript i on Item RA I Changes made to the fuel specificat i on T-14 Table 1-1i table for the 24PHB , inc l uding clarification 4 RAI 8-3 of the fuel damage definition Changes made to the fue l specification T-16 Table 1-1j table for the 61 BT , including clarification 4 RAI 8-3 of the fuel damage definition Changes made to the fuel specification T-19 and T-21 Table 1-11 table for the 24PTH , including clarification 1 , 4 RA I 8-3 of the fuel damage definition and fuel class description T-30 Table 1-1r Changes made to the B 10 specification 6 NIA table for the 24P T H Changes made to the fuel specification T-32 and T-34 Table 1-1t for t he 61 BTH , including clarification of 4 , 8 RAI 8-3 the fuel damage definition and fue l class desc r iption T-35 Table 1-1 u Changes made to the FA design 8 NIA characteristics table for the 61 BTH T-36 Table 1-1v Changes made to the 61 BTH for certain 8 NIA basket parameters (intact fuel) T-37 Table 1-1w Changes made to 61 BTH for certain 8 NIA basket parameters (damaqed fuel) Changes made to 61 BTH for certain T-38 Table 1-1w1 basket paramete r s (fai l ed and damaged 8 NIA fuel) Changes made to the fuel specification T-42 and T-43 Table 1-1aa table for the 32PTH1 , including 1 , 4 RAI 8-3 clarification of the fuel damage definition and fuel c l ass description T-44 Table 1-1 bb Changes made to the FA design 1 NIA characteristics for the 32PTH1 Changes made to the fuel specification T-53 Table 1-1gg for the 698TH , including clarification of 4 RAI 8-3 the fuel damaqe definition Changes made to the fuel specificat i on T-59 and T-60 Table 1-111 for the 37PTH , including clarification of 1 RAI 8-3 the fuel damage definition T-69 Table 1-2c Revision bar signi fi es deletion of Tables 1 NIA paqe 1-2d to 1-2m T-70 none Tables 1-2d through 1-2m are deleted 1 NIA T-75 to T-122 Tables 1-3a These tables are the new , generic PWR 1 NIA to 1-3-p FQTs T-123 to T-1 2 5 These are the notes for the new , generic 1 NIA PWR FQTs T-126 to T-13 7 Tabl e s 1-4a Changes made t o 61 BTH FQT 8 NIA Page 4 o f 5 Enclosure 7 to E-51144 TS page Scope RS I or TS Number Description Item RA I to 1-4f T-138 Table 1-4g Changes made to 61 BTH FQT 8 N/A T-139 to T-142 Tables 1-4h Changes made to 61 BTH FQT 8 N/A and 1-4i T-143 to T-1 44 none Changes made to the notes for 61 BTH 8 N/A FQT T-145 none Tables 1-5a through 1-5g are deleted 1 NIA T-179 none Revision bar signifies de l etion of Tab l es 1 N/A 1-8a to 1-8f F-2 Figure 1-2 Changes made to HLZC #1 for 32PT for 4 N/A damaqed FAs F-3 F i gure 1-3 Changes made to HLZC #2 for 32PT for 4 N/A damaged and failed FAs F-4 Figure 1-4 Changes made to HLZC #3 for 32PT for 4 N/A damage FAs F-5 Figure 1-4a New HLZC# 4 for the 32PT 2 N/A F-6 Figure 1-4b New figure for location of damaged and 4 N/A failed fue l in the 32PT F-14 Figure 1-11 Changes made to HLZC #1 for 24PTH 1 N/A F-15 Figure 1-12 Changes made to HLZC #2 for 24PTH 1 N/A F-16 Figure 1-13 Changes made to H LZC #3 for 24PTH 1 N/A F-17 Figure 1-14 Changes made to HLZC #4 for 24PTH 1 N/A F-18 Figure 1-15 Changes made to HLZC #5 for 24PTH 1 N/A F-19 Figure 1-15a New HLZC #6 for the 24PTH 6 N/A F-31 Figure 1-25b Changes made to HLZC #1 O for 61 BTH 8 N/A F-36 Figure 1-28b New HLZC #5 for 32PTH1 9 N/A F-37 Figure 1-28c New HLZC #6 for 32PTH1 9 N/A Page 5 of 5