Text

License Amendment Request to Support the Use of Metamic@ Polson Insert Assembles in the Spent Fuel Pool
	ML062220440
Person / Time
Site:	Arkansas Nuclear
Issue date:	07/27/2006
From:	Forbes J; Entergy Operations
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	1CAN070603
	Download: ML062220440 (111)
	v • d • e

Entergy Operations, Inc.

i1448S R 333

~Entergy Russellville, AR 72802 Tei 479-858-4888 Jeffery S. Forbes Vice Pres~dent Operations AN4O ICAN070603 July 27, 2006 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

Ucense Amendment Request to Support the Use of MetamicO Poison Insert Assemblies in the Spent Fuel Pool Arkansas Nuclear One, Unit I (ANO-1)

Docket No. 50-313 Ucense No. DPR-51

REFERENCES:

1. Letter to the NRC dated March 18, 200A, 'Degradation of Boraflex in ANO-1 Spent Fuel Poor' (ICAA1030203)

2. Letter to the NRC dated August 8, 2002, "Use of Metamic In Fuel Pool Applications' (OCAN0802Q1)

3. Letter to the NRC dated Apdl 2, 2003, 'License Amendment Request to Modify the Fuel Assembly Enrichment, the Spent Fuel Pool (SFP) Boron Concentration Technical Specification (TS) 3.7.14, the Loading Restrictions in the SFP in TS 3.7.15, and to Modify the Fuel Storage Design Features in TS 4.3' (1CAN040302)

4. Letter from the NRC dated June 17, 2003, "Arkansas Nuclear One, Units 1.and 2 - Safety Evaluation for Holtec Report HI-2022871 Regarding Use of Metamic in Fuel Pool Applications' (TAC NOS. MB5862 and MB5863)

5. Letter to the NRC dated June 24, 2004, Withdrawal of Ucense Amendment Request" (1CAN060404)

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Entergy Operations, Inc. (Entergy) hereby requests an amendment to the Arkansas Nuclear One, Unit 1 (ANO-1) Technical Specification (TS) 3.7.14, Spent Fuel Pool Concentration, TS 3.7.15, Spent Fuel Pool Storage, and TS 4.3, Fuel Storage.

1CAN070603 Page 2 of 3 The proposed TS changes support a planned modification to the ANO-1 spent fuel pool (SFP) that will utilize Metamic poison insert assemblies (PIAs). Entergy submitted by letter dated August 8, 2002, a topical report (Reference 2) to allow the use of Metamic@ in SFP applications.

The report, prepared by Holtec International, describes the physical and chemical properties of Metamic and includes the test results for the use of Metamic in fuel pool applications. The topical report was approved by the NRC on June 17, 2003 (Reference 4).

The ANO-1 SFP racks are currently divided into two regions. The Region 1 racks contain Boraflex. By letter dated March 18, 2002 (Reference 1), Entergy described its plans to administratively control spent fuel loading based on the degradation of Boraflex. Continued degradation of Boraflex is projected and therefore the proposed change will take no credit for Boraflex@ in Region I and will include new loading restrictions for Region 1.

The Region 2 racks do not contain any fixed poison panel assemblies. The proposed change will modify a portion of the SFP storage racks in Region 2 by the insertion of Metamic PIAs.

The area with the Metamic PIAs will be defined as Region 3. Loading restrictions will be applied to the remaining Region 2 racks and will also apply to the newly defined Region 3.

In addition to the above proposed change and plant modification, Entergy proposes to increase the SFP boron concentration and credit boron to assure a five percent subcriticality margin is maintained during normal and accident conditions. The allowance to credit boron is consistent with 10 CFR 50.68, Criticalityaccidentrequirements.

To accommodate future reload plans, Entergy proposes to increase the allowable initial fuel assembly U-235 enrichment from 4.1 weight percent (wt%) to a maximum U-235 enrichment of 4.95 wt%. Criticality analyses were performed using the higher enrichment for the SFP racks as well as the new fuel storage racks. New loading patterns are proposed for the new fuel storage racks.

In accordance with the conditions of the topical report (Reference 4), the proposed change will include a coupon sampling program to monitor the potential changes in the characteristics of Metamic).

The proposed change has been evaluated in accordance with 10 CFR 50.91 (a)(1) using criteria in 10 CFR 50.92(c) and it has been determined that this change involves no significant hazards considerations. The bases for these determinations are included in the enclosed Attachments.

The proposed change includes new commitments that are listed in Attachment 7.

In April 2003 (Reference 3) Entergy submitted a change to support the use of Metamic PIAs.

The submittal was in final NRC technical reviews when a manufacturing issue was identified, which resulted in the withdrawal of the submittal (Reference 5). During the NRC's review process, additional information was requested to assist in the review of the April 2003 submittal.

Where appropriate, Entergy's responses to the NRC's request for additional information associated with the April 2003 submittal have been included in the enclosed attachments.

Entergy requests approval of the proposed amendment by February 1, 2007, in order to support insertion of the Metamic PIAs prior to the spring 2007 refueling outage. Once approved, the amendment shall be implemented within 60 days. Although this request is neither exigent nor emergency, your prompt review is requested.

I CAN070603 Page 3 of 3 Ifyou have any questions or require additional information, please contact Dana Millar at 601-368-5445.

I declare under penalty of perjury that the foregoing is true and correct. Executed on July 27, 2006.

Sincerely, JSF/dm Attachments:

1. Analysis of Proposed Technical Specification Change

2. Proposed Technical Specification Changes (mark-up)

3. Proposed Technical Specification Changes (revised)

4. Changes to Technical Specification Bases Pages (For Information Only)

5. Spent Fuel Pool Racks Modifications with Poison Material Inserts in ANO Unit 1

6. Evaluation of Spent Fuel Pool Structural Integrity for Increased Loads from Spent Fuel Racks

7. List of Regulatory Commitments cc: Dr. Bruce S. Mallett Arkansas Department of Health & Human Regional Administrator Services U. S. Nuclear Regulatory Commission Division of Health Region IV P.O. Box 1437 611 Ryan Plaza Drive, Suite 400 Slot H-30 Arlington, TX 76011-8064 Little Rock, AR 72203-1437 NRC Senior Resident Inspector U. S. Nuclear Regulatory Commission Arkansas Nuclear One Attn: Mr. Drew G. Holland P. O. Box 310 MS O-7D1 London, AR 72847 Washington, DC 20555-0001

Attachment I ICAN070603 Analysis of Proposed Technical Specification Change to 1CAN070603 Page 1 of 23

1.0 DESCRIPTION

This letter is a request to amend Operating License DPR-51 for Arkansas Nuclear One, Unit I (ANO-1).

The proposed changes will revise the ANO-1 Technical Specifications (TSs) to:

" Allow insertion of Metamic@ poison insert assemblies (PIAs) into the flux traps of a newly defined Region 3 of the ANO-1 spent fuel pool (SFP).

Redefine the loading pattern in the current Region I racks, taking no credit for Boraflex.

Redefine the loading pattern in the remaining Region 2 racks.

Define the loading pattern for the newly created Region 3 racks.

" Modify the SFP boron concentration from Ž: 1600 parts per million (ppm) to > 2000 ppm.

" Modify the applicability of TS 3.7.14 to specify that the TS is applicable any time fuel assemblies are stored in the SFP regardless of whether a SFP verification has been performed or not.

Allow an increase in the maximum fuel assembly uranium-235 (U-235) enrichment from the current U-235 enrichment of 4.1 weight percent (wt%) to a maximum of 4.95 wt%/o.

Redefine storage patterns in the new fuel storage racks.

" Add a Metamic coupon sampling program (Surveillance Requirement (SR) 3.7.15.2 and TS 5.5.17 will be added).

Changes are proposed to the following ANO-1 TSs:

TS 3.7.14, Spent Fuel Pool Boron Concentration

TS 3.7.15 and the associated Figure 3.7.15-1

TS 4.3, Fuel Storage and the associated Figure 4.3.1.2-1

SR 3.7.15.2 will be added to direct the performance of the coupon sampling program, which will be reflected in proposed TS 5.5.17 Appropriate changes will also be made to the associated TS Bases. A markup of the TS Bases is included for information only in Attachment 4.

The changes are desired in order to address the degradation of Boraflex and to support the creation of Region 3 in which Metamic PIAs will be installed. In addition, the increase in maximum fuel assembly U-235 enrichment is desired to support future reactor fuel loading capabilities. Approval of the proposed changes is desired by February 1, 2007 to support installation of the PIAs in the SFP prior to the spring 2007 refueling outage. NRC's approval of this amendment request will allow ANO-1 to maintain full core offload capability during and following the spring outage.

to 1CAN070603 Page 2 of 23

2.0 PROPOSED CHANGE

Technical Specification 3.7.14, Spent Fuel Pool Boron Concentration ANO-1 TS 3.7.14 and SR 3.7.14.1 define the minimum required SFP boron concentration. The proposed change increases the requirement for the minimum boron concentration to greater than 2000 ppm. The proposed increase in boron concentration provides sufficient margin that assures the maximum neutron multiplication factor (keff) will remain below 0.95 in the unlikely event of a criticality accident. The upper limit on SFP boron concentration is 3500 ppm per ANO-1 Safety Analysis Report (SAR) Section 9.6.2.4.3.4.

In addition, TS 3.7.14 is currently applicable whenever fuel assemblies are stored in the SFP and when SFP verification has not been performed since the last movement of fuel assemblies in the SFP. The proposed change will modify the applicability to require the designated boron concentration any time fuel assemblies are stored in the SFP regardless of whether SFP verification has been performed. Deletion of this portion of the applicability requires the deletion of Action A.2.2. These changes are needed to support the new criticality analysis that credits boron to assure the required subcritical margin is maintained. Crediting soluble boron to maintain a five percent subcritical margin is allowed by 10 CFR 50.68, Criticalityaccident requirements.

A minor format change will be made to the numbering of the Actions.

Technical Specification 3.7.15, Spent Fuel Pool Storage ANO-1 TS 3.7.15 and Figure 3.7.15-1 define loading restrictions for fuel assemblies that are stored in Region 2 of the ANO-1 SFP. Currently, no loading restrictions are required by TS in Region 1 of the SFP.

Region 1 contains Boraflex poison panels. Calculations indicate that the Boraflex content in Region 1 will degrade below the Boraflex content assumed in the criticality analysis. This condition is currently being tracked and responded to through the ANO 10 CFR 50 Appendix B corrective action program. Based on the continued projected degradation of Boraflex, the proposed change will not credit Boraflex in Region 1 and will include loading restrictions for Region 1. This will result in changes to TS 3.7.15, the deletion of Figure 3.7.15-1 and the creation of a new Table 3.7.15-1 based on new SFP criticality analysis.

The TS will be changed to apply the loading restrictions to any fuel assembly that is stored in the SFP and will be applicable whenever a fuel assembly is stored in the SFP. The Action will be modified to require the non-conforming fuel assembly to be placed in an acceptable storage location in accordance with the appropriate loading restrictions.

Surveillance requirement 3.7.15.1 will be modified to reflect its applicability to the parameters defined in the proposed Table 3.7.15-1. The frequency will be changed to delete reference to Region 2. The parameters associated with the fuel assembly must be satisfied prior to storing a fuel assembly in the SFP, as loading restrictions now apply throughout the pool.

Attachment I to 1CAN070603 Page 3 of 23 Region 1 loading restrictions will be reflected in Table 3.7.15-1 as follows:

Region 1 - Minimum Burnup Requirements at Varying Initial U-235 Enrichment and Cooling Time (Notes 1 & 2)

Enrichment 2.0 2.5 3.0 3.5 4.0 4.5 5.0 Cooling Time Minimum Bumup (GWD/MTU)

(Years) _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

0 2.1 8.6 15.0 21.8 27.2 32.7 38.4 5 2.0 8.2 14.1 20.7 26.2 30.8 36.4 10 1.9 7.9 13.7 19.5 25.5 29.5 34.6 15 1.8 7.7 13.2 18.8 24.7 28.8 33.8 20 1.8 7.5 13.0 18.4 24.2 28.2 33.0 The criticality analysis also results in changes to the current loading restrictions imposed on the Region 2 racks and reflected in current Figure 3.7-15-1. The current figure will be deleted and new loading restrictions will be reflected in Table 3.7.15-1 for Region 2, as follows:

Region 2 - Minimum Bumup Requirements at Varying Initial U-235 Enrichment and Cooling Time (Notes 1 & 2)

Enrichment 2.0 2.5 3.0 3.5 4.0 4.5 5.0 Cooling Time Minimum Bumup (GWD/MTU)

(Years) _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

0 4.0 11.2 18.2 25.0 30.1 36.6 42.7 5 3.8 10.6 17.1 23.9 28.8 34.4 40.3 10 3.6 10.1 16.2 22.7 27.8 32.7 38.3 15 3.5 9.8 15.8 21.9 27.1 31.7 37.1 20 3.3 9.6 15.2 21.4 26.6 30.7 36.2 Note 1 associated with the two tables will be added stating:

"Linear interpolation between bumups for a given cooling time is allowed. However, linear interpolation between cooling times is not allowed, therefore the cooling time of a given assembly must be rounded down to the nearest cooling time."

Note 2, which is also associated with Regions I and 2, will be added to state the following:

"When it is necessary to store fuel assemblies in Region 1 or Region 2 that do not meet the burnup versus U-235 enrichment restrictions, fuel assemblies, including fresh or irradiated fuel assemblies with a maximum U-235 enrichment of 4.95 wt%, may be stored in a 2 x 2 checkerboard (i.e., 2 assemblies and 2 empty cells) arrangement."

A portion of the SFP racks in Region 2 will be modified by the installation of Metamic@ PIAs.

This will result in the creation of a new Region 3 that will include loading restrictions which will be reflected in Table 3.7.15-1 as follows:

Attachment I to 1CAN070603 Page 4 of 23 Region 3 Loading Restrictions Unrestricted storage is allowed for fuel assemblies with an initial U-235 enrichment less than or equal to 4.35 wt%

For fuel assemblies with an initial U-235 enrichment greater than 4.35 wt%, the bumup of at least one fuel assembly in each 2 x 2 section of storage cells is at least 20.1 GWD/MTU.

Rack interface requirements have been evaluated and will be included in Table 3.7.15-1 as follows:

Rack Interface Requirements In addition to the above requirements for each individual rack, the following requirements must be met on the interfaces between and within racks:

a. In the Region I and Region 2 racks, a fresh fuel checkerboard and uniform spent fuel loading may be placed in the same rack.

b. In Region I and Region 2 racks, if adjacent racks contain a checkerboard of fresh fuel assemblies, the checkerboard must be maintained across the gap, i.e., fresh fuel assemblies may not face each other across a gap.

c. In Region 3, uniform loading of fresh fuel with a maximum U-235 enrichment of 4.35 wt%

may be combined with 3 of 4 loading in the same rack as long as a row of fresh and spent fuel in the 3 of 4 loading pattern faces the uniform loading of all fresh fuel with a maximum U-235 enrichment of 4.35 wt%.

d. If adjacent Region 3 racks contain different loading patterns (one rack contains all fresh fuel with a maximum U-235 enrichment of 4.35 wt% and the other rack contains a 3 of 4 loading pattern), both fresh and spent fuel must be in the outer row of the rack containing the 3 of 4 pattern.

e. If adjacent Region 3 racks both contain 3 of 4 loading patterns, both racks may not have fresh fuel facing the other rack. A loading pattern with both Region 3 racks containing 3 of 4 patterns with all fresh fuel in the outer row of one rack and fresh and spent fuel in the outer row of the second rack is allowed.

f. All interfaces between dissimilar racks (Region 1-Region 3 and Region 2-Region 3) are permitted.

A new SR will be added as SR 3.7.15.2 to direct performance of the coupon sampling program.

The coupon sampling program will be added to the Administrative section of the TSs as TS 5.5.17. The SR will be included as follows:

to I CAN070603 Page 5 of 23 SURVEILLANCE FREQUENCY SR 3.7.15.2 Verify Metamic properties are in accordance with, and In accordance with are maintained within the limits of, the Metamic the Metamic Coupon Sampling Program. Coupon Sampling Program.

Technical Specifications 4.3.1. Criticality ANO-1 TS 4.3.1.1 a defines a maximum U-235 enrichment of 4.1 wt%. The proposed change will allow the maximum U-235 enrichment to be 4.95 wt%.

ANO-1 TS 4.3.1.1 b defines that k.1f (the effective neutron multiplication factor) will be maintained less than or equal to 0.95 if the spent fuel pool racks are fully flooded with unborated water. The criticality analysis, as allowed by 10 CFR 50.68, will credit boron (457 parts per million (ppm)) to assist in maintaining ke*-- 0.95 during normal operating conditions.

TS 4.3.1.1 c will be added, which will describe that a ke* less than 1.0 will be maintained when the pool is flooded with unborated water. The addition of TS 4.3.1.1 c will result in the currently designated TSs 4.3.1.1 c, d, and e to be re-indexed as TSs 4.3.1.1 d, e, and f, respectively.

Current ANO-1 TS 4.3.1.1 d and TS 4.3.1.1 e define loading restrictions for Regions 1 and 2.

These will be modified in accordance with the changes proposed to TS 3.7.15 and as follows (current TS 4.3.1.1 d is reflected as me" and current TS 4.3.1.1 e is reflected as "f below):

e. New or partially spent fuel assemblies stored in accordance with Table 3.7.15-1 in the spent fuel pool storage racks;

f. New or partially spent fuel assemblies with cooling times, U-235 enrichment or discharge bumup in the unacceptable range of Table 3.7.15-1 for fuel stored in either Region 1 or Region 2 may be stored in a 2 x 2 checkerboard configuration (i.e., 2 assemblies and 2 empty cells); and To describe the design features of Region 3, TS 4.3.1.1 g will be added and will state:

g. Neutron absorber (Metamic) installed between fuel assemblies in the Region 3 racks.

ANO-1 TS 4.3.1.2 a defines a maximum U-235 enrichment of 4.1 wt%. The proposed change will allow the maximum U-235 enrichment to be 4.95 wt%.

ANO-1 TS 4.3.1.2 e references Figure 4.3.1.2-1 which depicts locations in the fresh fuel storage racks in which fuel loading is prohibited. Based on the increase in fuel assembly U-235 enrichment from 4.1 wt% to a maximum U-235 enrichment of 4.95 wt%, two loading pattern configurations will be proposed, which will result in the addition of Figure 4.3.1.2-2. Figure 4.3.1.2-1 will depict the loading pattern associated with fuel assemblies with a maximum U-235 enrichment up to 4.95 wt% and Figure 4.3.1.2-2 will illustrate the loading configuration for fuel

Attachment I to 1CAN070603 Page 6 of 23 assemblies with a maximum U-235 enrichment up to 4.2 wt%. A change is also proposed to TS 4.3.1.2 e to state that the fuel storage racks shall be maintained in accordance with one of the two proposed figures, based on U-235 fuel enrichment.

Technical Specification 5.5.17, Metamic Coupon Samplingq Program The following program will be added to address the coupon sampling program requirements:

5.5.17 Metamic Coupon Sampling Program A coupon surveillance program will be implemented to maintain surveillance of the Metamic absorber material under the radiation, chemical, and thermal environment of the SFP. The purpose of the program is to establish the following:

Coupons will be examined on a two year basis for the first three intervals with the first coupon retrieved for inspection being on or before February 2009 and thereafter at increasing intervals over the service life of the inserts.

" Measurements to be performed at each inspection will be as follows:

A) Physical observations of the surface appearance to detect pitting, swelling or other degradation, B) Length, width, and thickness measurements to monitor for bulging and swelling C) Weight and density to monitor for material loss, and D) Neutron attenuation to confirm the B-1 0 concentration or destructive chemical testing to determine the boron content.

" The provisions of SR 3.0.2 are applicable to the Metamic Coupon Sampling Program.

" The provisions of SR 3.0.3 are not applicable to the Metamic Coupon Sampling Program.

A minor format change was made to page 5.0-25 by deleting the double end line on the page since it is no longer the last program in the Programs and Manuals, section 5.5.

In summary, the proposed change will modify the ANO-1 TSs to support the insertion of Metamic@ PIAs. Changes are proposed to TS 3.7.14 to address the applicability and boron concentration. Technical Specification 3.7.15 will be revised to apply the appropriate loading restrictions in the SFP. TS 4.3.1 will be changed to support higher fuel assembly U-235 enrichment and the new loading restrictions in the SFP and in the new fuel storage vault. A new coupon sampling program will be added as TS 5.5.17.

to 1CAN070603 Page 7 of 23

3.0 BACKGROUND

3.1 Spent Fuel Pool Racks The ANO-1 SFP provides 968 storage locations for new and spent fuel assemblies or other items. Spent fuel pool storage is currently divided into two regions. Region 1 utilizes Borafiex as a neutron absorbing material. Region 2 does not utilize neutron absorbing materials.

Region 1 is designed to accommodate non-irradiated fully enriched fuel. Region 2 is designed to accommodate irradiated fuel that has sustained approximately 80 percent of the design bumup. Placement of fuel in Region 2 is determined by burnup calculations and controlled administratively. Fuel, which does not meet this criterion, may be placed in Region 2 in a checkerboard fashion. In these cases, currently vacant spaces adjacent to the faces of any fuel assembly that does not meet the Region 2 burnup criteria (non-restricted) are physically blocked before any such fuel assembly may be placed in Region 2. The racks meet the requirements of the NRC Position for Review and Acceptance of Spent Fuel Storage and Handling Applications, dated April 14, 1978, and modified January 18, 1979, with the exception that, for Region 2 storage, credit is taken for fuel bumup based on the proposed Revision 2 of Regulatory Guide 1.13.

The Region I storage racks are composed of individual storage cells made of type 304 stainless steel and conform to the requirements of ASME B&PV Code,Section III, Subsection NF. These racks utilize a neutron absorbing material, Boraflex, which is attached to each cell. The cells within a module are interconnected by grid assemblies to form an integral structure. Each rack module is provided with leveling pads which contact the SFP floor and are remotely adjustable from above through the cells at installation. The modules are neither anchored to the floor nor braced to the pool walls.

The Region 2 storage racks consist of type 304 stainless steel cells assembled in a checkerboard pattern, producing a honeycomb type structure. Each cell has attached to its outer wall a stainless steel wrapper plate creating a pocket opened at the top and bottom. This is referred to as a "spacer pocket" or "flux trap" design. The flux traps were designed to accept poison panel inserts if future need arises. This design is also provided with leveling pads which contact the SFP floor and are remotely adjustable from above through the cells at installation.

The modules are neither anchored to the floor nor braced to the pool walls.

The proposed change will modify two of the Region 2 storage racks with the insertion of Metamic@ PIAs in the existing flux traps. The racks where Metamico PIAs are inserted will be designated as Region 3 storage racks.

3.2 New Fuel Storagqe Area The new fuel storage area is a separate and protected area for the dry storage of new fuel assemblies. The new fuel storage module consists of a nine by eight array with a 21-inch pitch in both directions. Currently, ten interior storage cells are precluded from use prior to storage of any new fuel assemblies in the new fuel rack.

The new fuel storage racks are structural frames consisting of beams, columns and diagonal bracing acting as a unit to provide both vertical support at the bottom of the fuel element and lateral support at the top and bottom of the element. The racks are constructed of aluminum and are designed for gravity loads from the racks and fuel elements as well as the Design Basis Earthquake (DBE) in accordance with ASCE Paper 3341.

Attachment I to 1CAN070603 Page 8 of 23 The proposed change will allow storage of new fuel with a maximum U-235 enrichment of 4.95 wt% or a maximum U-235 enrichment of 4.2 wt% in storage patterns. If any of the fuel assemblies stored in the new fuel vault have U-235 enrichments greater than 4.2 wt%, then the assemblies will be stored in accordance with the loading restrictions for 4.95 wt%. Only if all assemblies stored contain fuel of 4.2 wt% U-235 enrichment or less, would the 4.2 wt%

configuration be applicable.

Section 9.6 of ANO-1 SAR contains description of the ANO-1 SFP and new fuel storage area.

3.3 Spent Fuel Cooling System The Spent Fuel Cooling (SFC) System is designed to maintain the water quality and clarity and to remove the decay heat from the stored fuel in the SFP. It is designed to maintain the SFP water at less than or equal to 150°F while removing the total decay heat load from the combination of stored fuel assemblies. This was determined from analysis performed using the guidelines of ASB 9-2, Residual Decay Energy for Light Water Reactors for Long Term Cooling.

In meeting the foregoing design bases, the system has the capability of maintaining the SFP water at approximately 120°F with a heat load based on the decay heat generated from approximately one-third of the core fuel assemblies discharged at the end of a given cycle.

In addition to its primary function, the system provides for purification of the SFP water, the fuel transfer canal water, and the contents of the Borated Water Storage Tank (BWST) in order to remove fission and corrosion products and to maintain water clarity for fuel handling operations.

The system also provides for filling the fuel transfer canal, the incore instrumentation tank, and the cask loading area from the BWST.

The SFP coolers are designed to maintain the temperature of the SFP as noted. The SFP coolers reject heat to the nuclear intermediate cooling water system which subsequently rejects its heat to the service water system.

The two SFP circulating pumps take suction from the SFP and recirculate the fluid back to the pool after passing through the coolers. At least one pump is normally in service. Cold water is discharged into the pool through two nozzles simultaneously. One of these nozzles is located near the water surface while the other one is near the bottom of the pool. The suction nozzle for the SFP circulating pumps is located on the opposite end of the pool from the discharge nozzles and is near the water surface. This arrangement provides thermal mixing and insures uniform water temperature. Significant convection currents, which contribute to the mixing in the pool, are also created from heat being transferred out of the stored spent fuel assemblies.

3.4 Spent Fuel Pool Makeup When makeup is required to replenish the SFP level, the following methods are available:

" From the BWST using the borated water recirculation pump discharging through the spent fuel purification loop.

" Boric acid from the boric acid addition tank (T-6) followed by demineralized (DI) water via the SFP circulating pump.

" From DI water only, if addition is required only for making up for evaporative losses.

" From service water via the service connection. Makeup from service water to the SFP is intended for emergency use only.

Attachment I to 1CAN070603 Page 9 of 23 Leakage, although unexpected, would be indicated by receipt of the control room SFP low level annunciator. As directed by procedure, the control room staff would visually verify actual level and commence makeup to the SFP as necessary. If unborated water sources are used for makeup, which requires manual operation, then chemistry samples are taken to ensure the SFP boron concentration remains within the TS required limit.

Section 9.4 of the ANO-1 SAR contains a detailed description of the SFP systems. No modifications are proposed to these systems in order to support the proposed change.

3.5 Previously Approved Amendments TS 3.7.15, Figure 3.7.15-1, TS 4.3.1.1.d, and TS 4.3.1.1.e currently define loading restrictions for Region 2 based on the combination of initial U-235 enrichment and bumup of each fuel assembly. The loading restrictions were defined by the approval of TS Amendment 76 (NRC Safety Evaluation Report (SER) dated April 15, 1983) which allowed modification of the ANO-1 SFP storage capabilities from 589 spaces to 968 spaces. The expansion was accomplished by installing new racks. The region designated as Region I of the new racks contained Boraflex while the remainder of the racks did not include poison inserts.

TS 4.3.1.1.a and TS 4.3.1.2.a currently define the maximum U-235 enrichment for fuel assemblies that will be stored in the SFP and the new fuel storage racks, respectively.

TS 4.3.1.2 and Figure 4.3.1.2-1 currently define the new fuel storage rack loading restrictions based on the new fuel maximum U-235 enrichment of 4.1 wt%. The loading restrictions were defined by the approval of TS Amendment 76 (NRC SER dated April 15, 1983) and TS Amendment 166 (NRC SER dated June 28, 1993).

3.6 Loading Pattern / Storage Procedural Controls The controls used in determining the storage location for new and irradiated fuel in the new or spent fuel storage racks are governed by procedure. The procedure currently contains guidelines pertaining to restricted and unrestricted fuel storage as reflected by the TS. The new loading pattern restrictions and the addition of Region 3 will continue to be governed by procedure. Checkerboard storage configurations and vacant spaces will be administratively controlled by procedure.

Prior to moving a fuel assembly, the fuel assembly is required to be classified based on initial U-235 fuel enrichment, decay time, and burnup to determine where it can be placed in the SFP.

ANO administratively controls the proper placement of fuel assemblies based upon their fuel assembly classification. A qualified engineer using the procedure for controlling special nuclear material checks the fuel assembly classification for each assembly. An independent review by a qualified engineer assures the proper fuel assembly classification.

The above defined classification process is used before placement of fuel assemblies into either Region I or Region 2. The process for placement of fuel assemblies within the pool is controlled by site procedures as described below.

Attachment I to I CAN070603 Page 10 of 23 If it is desired to move a fuel assembly to Region 1, then the classification of that fuel assembly for that region is checked. Ifthe fuel assembly is unrestricted for Region 1, then that assembly can be stored unrestricted within Region 1. Ifthe fuel assembly is restricted for Region 1, then the assembly will be stored in a checkerboard arrangement. The same logic is used to determine placement of an assembly in Region 2. The qualified engineer then develops a special nuclear material transfer report. The engineer checks the most recent inventory map, any restricted storage areas, and checks any outstanding transfer reports that have been performed after the last update to the map. The qualified engineer verifies that the cell where the fuel assembly will be stored is empty and the cell is located in the correct region. If a checkerboard loading pattern is required, then the qualified engineer verifies that the surrounding cells where the restricted fuel assembly will be stored meet the checkerboard loading pattern requirements. The area that is designated to require a checkerboard pattern is classified as a checkerboard area and the empty spaces are procedurally controlled. If required, the qualified engineer ensures assemblies are moved out of cells that are required to be empty of special nuclear material prior to placing a restricted fuel assembly in the checkerboard arrangement. The qualified engineer will complete the transfer report.

The entire process for controlling the movement of fuel assemblies is independently reviewed by another qualified engineer. The Reactor Engineering Superintendent approves the report and ensures that the report has been independently reviewed. The transfer report is then sent to operations for review.

Movement of fuel assemblies is performed by qualified fuel handling personnel. Prior to grappling and un-grappling a fuel assembly the qualified fuel handler verifies the fuel handling grapple or fuel assembly is over that correct location designated by the transfer report. These locations are independently verified prior to grappling and un-grappling the assembly. Upon completion of the special nuclear transfer report two qualified engineers then independently perform a survey of the storage locations to verify that the special nuclear material has been moved to the proper location. After the special nuclear material location verification is complete, the inventory maps are updated along with the special nuclear material location record.

As part of the standard implementation process for an approved TS change, the process and procedures described above will be modified to reflect the proposed loading restrictions for the different regions in the ANO-1 SFP.

3.7 Technical Specification Surveillance Requirements and Response ANO-1 SR 3.7.13.1 requires monitoring of the SFP level. To satisfy this SR, the ANO-1 Operations staff monitors SFP level daily using a control room indication and weekly using local indication at the SFP. The Operations staff is trained to monitor for unexpected deviations in their recorded readings and take appropriate actions to determine and correct the cause of the deviation as needed. In the unlikely event a decreasing trend is not appropriately observed through daily monitoring, a SFP low level annunciator, which alarms at 6 inches below normal pool level, is provided in the ANO-1 control room. Upon receipt of the alarm, an Operator is dispatched to the SFP to verify actual level and if level is dropping, the SFP cooling pumps are secured, if necessary, and SFP level is adjusted to within limits. Ifthe SFP level alarm malfunctioned at this magnitude of inventory loss, it would take a long period of time for the SFP level to significantly drop (the SFP contains over 600 gallons of borated water per inch).

Eventually, there would be an increase in radiation levels that would be detected by the SFP area radiation monitor, which is monitored by Operations personnel every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and alarms in the ANO-1 control room.

Attachment I to 1CAN070603 Page 11 of 23 Additionally, the SFP boron concentration is sampled at least weekly as required by TS SR 3.7.14.1. If there were an unexpected downward trend in boron concentration, Operations would be notified to investigate the cause and take appropriate corrective actions.

3.8 SFP Structural Integrity Spent fuel pool structural integrity is monitored under the Maintenance Rule structural monitoring program established in ANO Engineering Standard CES-19. This standard requires periodic condition monitoring of key structures at ANO including the spent fuel storage structure.

The scope of the inspection includes, but is not limited to, visual examination for any evidence of cracks wider than 1/16", spalling, scaling, stratification, water infiltration, rust bleeding, exposed or corroded rebar, discoloration, concrete erosion or chemical attack. The initial baseline inspection of the ANO-1 SFP structure identified no deficiencies.

3.9 Fuel Type No fuel types other than B&W 15 x 15 fuel assemblies are currently stored in the ANO-1 SFP. If different fuel types are used in the future, changes to the fuel assembly design, key fuel assembly mechanical features, and the changes in operating strategy will be evaluated under 10 CFR 50.59, Changes, tests and experiments.

4.0 TOPICAL REPORT AND COUPON SAMPLING PROGRAM Entergy submitted to the Nuclear Regulatory Commission (NRC) by letter dated August 8, 2002, a topical report that supports the use of Metamic@ in SFP applications. The NRC approved Entergy's submittal by letter dated June 17, 2003. The topical report describes the manufacturing process, the material composition, the corrosion testing results, and the resistance of Metamic to radiation damage. The report also describes various coupon sampling programs that have been established at test facilities to monitor the physical and chemical property changes over time.

Although the conditions and limitations listed in the NRC Safety Evaluation (SE) are not explicitly referenced as conditions 1, 2, 3 and 4, for the sake of communication the conditions are broken out into separate line items below. Entergy's response to the conditions and limitations is as follows:

Condition 1 The staff conditions the use of the material [Metamic] upon a coupon sampling program to ensure consistent performance with that described in the Holtec report.

Entergy's Response To ensure the physical and chemical properties of Metamic behave in a similar manner as that found at the test facilities, Entergy will establish a coupon sampling program which will be reflected in TS 5.5.17, Metamic Coupon Sampling Program. A new TS Surveillance Requirement (SR) 3.7.15.2 will be created to direct the performance of the sampling program.

to 1CAN070603 Page 12 of 23 Condition 2 In addition, the staff requests that any application of this SE include a discussion of the following:

" Size and types of coupons to be used (i.e., similar in fabrication and layout as the proposed insert including welds and proximity to stainless steel);

Technique for measuring the initial B4C content of the coupons;

Simulation of scratches on the coupons;

Frequency of coupon sampling and its justification; and

Tests to be performed on coupons (e.g., weight measurements, measurement of dimensions (length, width and thickness), and B4C content);

these tests should also address, as a minimum, any bubbling, blistering, cracking, flaking, or areal density changes of the coupons, any dose changes to the coupons, or the effects of any fluid movement and temperature fluctuations of the pool water.

Entergy's Response Entergy will establish a coupon sampling program. The coupon measurement program is intended to monitor for the physical properties of the MetamicOD absorber material and thus provide a method of verifying that the assumptions used in the SFP criticality analyses remain valid.

The surveillance coupons will be approximately 7" x 5" and 0.100" thick, identical in composition and manufacturing process as the Metamic in the inserts (i.e., created from the same manufacturing lot used to manufacture the Metamic@ PIAs). The coupons will be mounted in stainless steel jackets simulating the actual insert design and held by a stainless steel coupon tree. The coupons are made from production runs and under go the same cleaning process as the poison insert assemblies.

The coupon tree will have ten or more coupons. To better simulate the conditions of the Metamic panels, the coupon tree has been designed to be installed within a flux trap. The coupon tree will be located in the Region 2 racks due to the availability of full length uncovered flux traps that do not and will not contain poison material (i.e., Region 1 contains Boraflex and the newly established Region 3 will contain Metamic PIAs).

The coupons will be staggered and placed adjacent to the active fuel region where, based on the burnup profile, the localized bumup is greater than the assembly average burnup. This design will maximize the dose received by the coupons as compared to the Metamic PIAs and accurately simulate the flow characteristics, pool chemistry, and differential metal interfaces that the Metamic PIAs will experience. No welding will be used on the Metamic as per the PIA design.

The initial B4C content of the surveillance coupons will be determined by the same chemical analysis technique used to establish the B4C content of the Metamico in the PIA. The B4 C used in the production of the Metamic was tested to determine the concentration of boron within each batch. The B4 C and aluminum matrix were accurately weighed and combined in a ratio to ensure the panels have a B4C content higher than 24.5 weight percent. Samples of the B4C and aluminum mixture were then chemically tested to determine the B4C concentration.

Additional samples were taken of the final product, which were also chemically tested to confirm

Attachment I to 1CAN070603 Page 13 of 23 the concentration of B4C in Metamic. Representative samples of the batches used will undergo neutron attenuation or chemical testing to validate the B10 loading in the Metamic panels and coupons. The testing of the initial B4C content in the Metamic PIAs and the Metamic coupons is performed under a Quality Assurance program using written procedures.

Scratches will be simulated by the mechanical etching or scribing the surface of the coupons.

The scratches will be formed using hardened materials made out of carbon steel, stainless steel, and Metamic. The scratches will not be cleaned after being applied to ensure an evaluation will be performed of the corrosion affects of leaving the trace material in a scratch.

The surveillance coupons will be removed and examined on a regular schedule to be established by TS 5.5.17. Coupons will be examined on a two year basis for the first three intervals and thereafter on a 4 to 5 year interval over the service life of the inserts. During the first six years, freshly discharged fuel assemblies will be placed on two sides of the coupon tree to ensure that the dose to the coupons is maximized. The initial two year period is based on providing sufficient exposure time to the environment such that any degradation would be observable.

The following table summarizes the sampling period for the Metamic coupon program. The number of coupons was based on the testing the Metamic for at least 40 years from 2007, which bounds the current operating license for ANO-1. Even though there are a sufficient number of coupons to last the existing operating life of the plant, the coupons may be placed back on the coupon tree after evaluations have been completed at the discretion of the reviewing engineer. However, if the integrity of the coupon is compromised or if the contamination levels are extremely high, the coupon will not be returned to the coupon tree.

In general the Metamic coupon testing program is very similar to testing programs used on Boral surveillance coupon testing programs.

Sample Coupon Measurement Schedule Coupon # Duration in SFP Years Sampling Period Years 1 2 2 2 4 2 3 6 2 4 10 4 5 15 5 6 20 5 7 25 5 8 30 5 9 35 5 10 40 5 11* Spare At Any Time 12* Spare At Any Time

The last two coupons are not required. These coupons will be pulled only if additional testing is desired.

to 1CAN070603 Page 14 of 23 Measurements to be performed at each inspection will be as follows:

" Physical observations of the surface appearance to detect pitting, swelling or other degradation,

Length, width, and thickness measurements to monitor for bulging and swelling (Measurements will be taken in five procedurally defined locations prior to placing the coupons in the ANO-1 SFP. When the coupon is removed, measurements will be taken in the same locations as the original measurements.)

" Weight and density to monitor for material loss, and

Neutron attenuation to confirm the B10 concentration or destructive chemical testing to determine the boron content.

Condition 3 In addition, applications for the use of Metamic should include a description of the anodizing process if anodized Metamic is used, and should include the cleaning technique to ensure sufficient removal of surface contaminants prior to installation.

Entergy's Response Anodized Metamic will not be used.

The only instance in which the Metamic is mechanically cleaned using air blasting is after the billet is extruded into a "bloom" stock. After pressing and sintering a specified mixture of boron carbide and type 6061 aluminum powders, the sintered billet is extruded through a specially designed die. This extrusion process is the primary source of surface contamination as the very hard boron carbide powder gouges minute particles of steel (iron) from the extrusion die. These minute particles of iron must be removed from the bloom stock in order to avoid a possible source of corrosion (pitting). A small amount of the surface of the Metamic and surface impurities are removed by blasting with air containing finely-divided aluminum oxide powder or small glass beads. Air blasting is performed until a visually uniform luster (appearance) is observed over the entire panel (both sides) and the panel is free from the presence of inclusions or foreign material embedded in the surface. In the EPRI tests some pitting corrosion was observed in samples that had not been adequately cleaned of these foreign particles. Cleaning the bloom stock prior to rolling the final plates is expected to be adequate to reduce the surface contamination in the final panels to reasonable and acceptable levels.

After shearing and dimensional verification and immediately prior to packaging for shipment, the Metamic panel is chemically cleaned by wiping the panel with a clean lint-free cloth wetted with dilute nitric acid solution. This is followed with wiping the panel using a clean lint-free cloth wetted with denatured alcohol or acetone.

Local areas may be cleaned chemically as described previously or mechanically using Scotch-Briteor similar material. Both steps are followed with wiping the panel using a clean lint-free cloth wetted with denatured alcohol or acetone.

Condition 4 The staff also limits the use of this SE to a B4 C content of 31 wt.% as evaluated and discussed in the Holtec report.

Entergy's Response The Metamic PIAs are designed with a B4C content of 25 wt%.

to 1CAN070603 Page 15 of 23

5.0 TECHNICAL ANALYSIS

The proposed change will result in defining a new region, designated as Region 3, in the SFP in which MetamicO poison panels will be inserted. New loading patterns are proposed in the existing Region 1, in which no credit for Boraflex will be taken, and in Region 2. In addition, Entergy is proposing to increase the fuel assembly maximum U-235 enrichment. Attachments 5 and 6 to this letter provide detailed technical analyses in support of the proposed change. A brief summary of the content of the analysis contained in these attachments is included in sections 5.2 through 5.7.

5.1 Dose Conseauences Associated with Increased U-235 Fuel Enrichment The current license basis for the potential offsite and control room dose radiological consequences of a Fuel Handling Accident (FHA) was performed considering the maximum U-235 enrichment of 4.95 wt%.

A FHA is a postulated accident involving damage to an irradiated fuel assembly during refueling.

Two possibilities exist for this accident: mechanical damage resulting in a release of activity and/or a criticality accident. Section 4 of Attachment 5 prepared by Holtec International addresses the criticality aspects of the FHA. The dose consequences associated with a FHA, which were calculated assuming a maximum bumup of 60,000 megawatt-days/ton of uranium (see ANO-1 SAR Table 14-24 for additional parameter assumptions), resulted in the following:

With Filtration (1) Without Filtration 25% of (rem) (rem) 10 CFR 100 Limits (rem)

Exclusion Area Boundary Whole 0.3 0.3 6 Body Dose Exclusion Area Boundary 10.4 69.1 75 Thyroid Dose (1) Filtration is through the charcoal filters in the normal SFP ventilation system.

From the information provided above, it can be seen that the dose consequences result in only a small fraction of the 25% of 10 CFR 100 limits.

5.2 Material Considerations It is proposed that MetamicD will be inserted in the newly defined Region 3. The physical and chemical properties of MetamicOD were included in the topical report that supports the use of MetamicV in SFP applications.

to I CAN070603 Page 16 of 23 5.3 Criticality Considerations A criticality safety evaluation was performed for storage of fresh and spent fuel in the ANO-1 SFP. The evaluation considered three regions that are designated as Region 1, Region 2, and Region 3. The criticality analysis currently in place for Region 1 assumes the presence of Borafiex. In the new analyses, no credit was taken for the Boraflex in Region 1. The new analysis also assumes Metamic PIAs are installed in Region 3. It was concluded that in order to assure keff remains less than 0.95 for the allowable storage configuration for both spent fuel and fresh fuel assemblies, a minimum soluble boron concentration is required. The proposed change to TS 3.7.14 will require a minimum boron concentration in the SFP of greater than 2000 ppm. The boron concentrations for each region determined by the analyses to assure Kff remains below 0.95 are bounded by the TS value. The fuel loading patterns are defined by the criticality safety evaluation included in the proposed changes and will be governed, as they are now, by procedure.

The criticality analyses also include consideration for storage of fuel assemblies with a maximum U-235 enrichment of 4.95 wt%. Analyses were performed for the new fuel storage racks and the SFP racks considering storage of the higher U-235 enrichment. Two new storage configurations are included in the proposed change for the new fuel storage racks based on different initial U-235 fuel enrichments.

5.4 Thermal Hydraulic Considerations A thermal hydraulic analysis conservatively demonstrated that natural circulation of the pool water for the proposed configuration provides adequate cooling of all fuel assemblies in the event of a loss of external cooling. Additionally, corrective actions can be taken prior to SFP boiling. The analysis also demonstrated that fuel cladding will not be subjected to departure from nucleate boiling under the postulated accident scenario of the loss of all SFP cooling and that cladding integrity would be maintained. None of the temperature limits or corrective actions for the SFP cooling system will be changed.

5.5 Mechanical Accident In line with the current approved philosophies, the postulated fuel assembly drop events for Region 3 of the SFP racks were evaluated. The criticality safety evaluation conservatively analyzed the Region 3 racks under the postulated assumption that all poison inserts are damaged. It should be noted that no postulated single drop accident event can damage all the Metamic@) panels in the spent fuel racks. For this non-credible and extremely conservative postulated scenario, the racks were determined to remain subcritical, when credit was taken for 1600 ppm soluble boron in the SFP.

5.6 Structural/Seismic Analysis A structural analysis of the spent fuel rack with the new poison panel inserts was considered for all loading configurations. The analysis evaluated normal, seismic, and accident conditions.

The evaluations demonstrate margins of safety in all storage racks.

The structural integrity of the new PIAs under normal and seismic conditions is essential to maintaining the assumptions of the criticality analysis. The PIAs design has been evaluated for normal and seismic conditions and all safety factors are greater than 1.0.

Attachment I to 1CAN070603 Page 17 of 23 5.7 SFP Structural Integrity for Increased Loads from SFP Racks An evaluation of the SFP structural integrity for the effects of the increased loads from the SFP racks was performed. The evaluation demonstrated that the structural integrity of the pool structure is maintained.

6.0 REGULATORY ANALYSIS

6.1 Applicable Regulatory Requirements/Criteria The proposed changes have been evaluated to determine whether applicable regulations and requirements continue to be met. The applicable regulations and requirements used to support the proposed changes and reflection of their continued compliance are included in subsequent attachments to this letter.

Arkansas Nuclear One, Unit 1 (ANO-1) is currently exempt from the requirements of 10 CFR 70.24, Criticalityaccident requirements. The exemption was granted on October 6, 1998 (TAC NOS. MA1278 and MA1279). Upon approval of the proposed change, the exemption to 10 CFR 70.24 will no longer be required. ANO-1 will fully comply with 10 CFR 50.68 paragraph (b) as follows:

50.68(b)(1) - Plant procedures shall prohibit the handling and storage at any one time of more fuel assemblies than have been determined to be safely subcritical under the most adverse moderation conditions feasible by unborated water.

It has been determined that movement of only one fuel assembly at a time assures that subcdticality is maintained under the most adverse moderation conditions feasible by unborated water. ANO fuel handling procedures for the spent fuel pool (SFP) and reactor refueling bridges will exclusively prohibit the movement of more than one fuel assembly over the SFP or the refueling canal. Movement of a fuel assembly using the upender frame is allowed while the fuel handling bridges are moving fuel assemblies because it has been determined that for the worst case geometry the effective neutron multiplication factor (ke*) is less than 0.95 at a 95% probability with a 95% confidence level. The fuel receipt procedure only allows one new fuel assembly to be moved at a time. Only one fuel assembly at a time is procedurally allowed to be moved into a dry fuel storage cask.

Storage of fuel assemblies is procedurally controlled to assure k* remains below 1.0, at a 95% probability, 95% confidence level, when flooded with unborated water. The storage patterns assure subcriticality under the most adverse moderation conditions by unborated water. The storage patterns will also insure reactivity will not exceed 0.95 at a 95%

probability with a 95% confidence level when credit is taken for 457 ppm boron during normal conditions. Ifa fuel assembly were to be misloaded, reactivity will not exceed 0.95 at a 95% probability with a 95% confidence level when credit is taken 889 ppm boron. For the overly conservative fuel drop accident that assumes the loss of all MetamicQ, a boron concentration of 1600 ppm ensures reactivity will not exceed 0.95 at a 95% probability with a 95% confidence level.

50.68(b)(2) - The estimated ratio of neutron production to neutron absorption and leakage (k-effective) of the fresh fuel in the fresh fuel storage racks shall be calculated assuming

Attachment I to ICAN070603 Page 18 of 23 the racks are loaded with fuel of the maximum fuel assembly reactivity and flooded with unborated water and must not exceed 0.95, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such flooding or if fresh fuel storage racks are not used.

Criticality calculations have been performed on the new fuel vault fully loaded with B&W 15 x 15 fresh fuel assemblies and filled with the most reactive unborated water. The results of these calculations showed that reactivity did not exceed 0.95, at a 95% probability, 95% confidence level.

50.68(b)(3) - If optimum moderation of fresh fuel in the fresh fuel storage racks occurs when the racks are assumed to be loaded with fuel of the maximum fuel assembly reactivity and filled with low density hydrogenous fluid, the k-effective corresponding to this optimum moderation must not exceed 0.98, at a 95 percent probability, 95 percent confidence level.

This evaluation need not be performed if administrative controls and/or design features prevent such moderation or if fresh fuel storage racks are not used.

Criticality calculations were performed on the new fuel vault fully loaded with B&W 15 x 15 fresh fuel assemblies and filled with the most reactive low density hydrogenous fluid. The results of these calculations showed that reactivity does not exceed 0.98, at a 95% probability, 95% confidence level. Hydrogenous fluid are not used in the new fuel vault area and they would only be used in the most extreme cases where the use of fire water was not able to contain a fire in the new fuel vault area.

50.68(b)(4) - If no credit for soluble boron is taken, the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with unborated water. If credit is taken for soluble boron, the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with borated water, and the k-effective must remain below 1.0 (subcritical), at a 95 percent probability, 95 percent confidence level, if flooded with unborated water.

Soluble boron credit will be taken in the SFP storage racks. The criticality calculations included in the proposed change show that kg remains below 1.0, at a 95% probability, 95% confidence level, when flooded with unborated water. Reactivity (Kff ) will not exceed 0.95 at a 95% probability with a 95% confidence level when credit is taken for 457 ppm boron during normal operations. Ifa fuel assembly were to be misloaded, reactivity will not exceed 0.95 at a 95% probability with a 95% confidence level when credit is taken 889 ppm boron. For the overly conservative fuel drop accident that assumes the loss of all Metamic, a boron concentration of 1600 ppm ensures reactivity will not exceed 0.95 at a 95% probability with a 95% confidence level.

50.68(b)(5) - The quantity of SNM, other than nuclear fuel stored onsite, is less than the quantity necessary for a critical mass.

Any quantity of SNM (special nuclear material), other than nuclear fuel, that is received on site is tracked to ensure that the total quantity remains less than that needed to form a critical mass.

Attachment I to ICAN070603 Page 19 of 23 50.68(b)(6) - Radiation monitors are provided in storage and associated handling areas when fuel is present to detect excessive radiation levels and to initiate appropriate safety actions.

A radiation monitor is located in the ANO-1 SFP area and alarms within the control room.

During fuel movement activities additional radiation monitors may be located directly on the fuel handling bridges to provide an additional audible indication of excessive radiation levels. The fuel handling procedures require at least one radiation monitor to be in place when fuel movement is in progress.

50.68(b)(7) - The maximum nominal U-235 enrichment of the fresh fuel assemblies is limited to five (5.0) percent by weight.

The proposed change to the ANO-1 Technical Specifications (TS) 4.3.1.1.a and 4.3.1.2.a will limit the maximum U-235 fuel enrichment to 4.95 weight percent including enrichment uncertainties.

50.68(b)(8) - The FSAR [Final Safety Analysis Report] is amended no later than the next update which 50.71(e) of this part requires, indicating that the licensee has chosen to comply with 50.68(b).

The ANO-1 FSAR will be amended no later than the next required update after the proposed TS change is approved and implemented. This FSAR update will indicate that ANO-1 has chosen to comply with 50.68(b).

Entergy has determined that the proposed changes do not require any exemptions or relief from regulatory requirements, other than the TS, and do not affect conformance with General Design Criteria (GDC) 61, Fuel storage and handling and radioactivitycontrol, or GDC 62, Prevention of criticalityin fuel storage and handing,differently than described in the FSAR.

6.2 No Significant Hazards Consideration The proposed changes will modify the Arkansas Nuclear One, Unit I (ANO-1) Technical Specifications (TSs) related to fuel storage and uranium-235 (U-235) fuel assembly enrichment.

The ANO-1 spent fuel pool (SFP) is currently divided into two regions (designated as Region 1 and Region 2) in which specific loading restrictions are imposed based on assembly average bumup and initial assembly average U-235 enrichment (up to 4.1 weight percent (wt%) U-235).

The SFP racks in Region 1 contain Boraflex@ as a neutron absorber while the SFP racks in Region 2 contain no neutron absorbers. Based on calculations, which indicate that the neutron absorption characteristics of Boraflex are degrading, Entergy has determined that the reactivity worth of Boraflex@ should no longer be credited in the reactivity analysis and thus more stringent loading restrictions should be imposed in Region 1. Therefore, the proposed changes include modifications to the loading restrictions in Region 1. Changes to the loading restrictions in Region 2 are also proposed. In addition, a portion of the current Region 2 will be designated as a new Region 3. The new region will contain Metamic@ poison panel insert assemblies (PIAs) which will provide the neutron absorption capability required to allow storage of various combinations of fuel burnup and uranium-235 (U-235) enrichment.

to 1CAN070603 Page 20 of 23 In addition to the above proposed change and plant modification, Entergy proposes to increase the SFP boron concentration and credit boron to assure a five percent subcriticality margin is maintained during normal and accident conditions. The allowance to credit boron is consistent with 10 CFR 50.68, Criticalityaccidentrequirements.

In order to accommodate future reactor core loading flexibility, Entergy is also proposing to increase the allowable U-235 fuel assembly enrichment to a maximum of 4.95 weight percent (wt%.)

The above proposed changes will result in modifications to TSs covering the SFP boron concentration, the SFP storage, and the design features related to fuel storage. In addition, a coupon sampling program will be added.

Entergy Operations, Inc. has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

Fuel Handling Accidents The current licensing bases for the dose consequences associate with a fuel handling accident (FHA), which was performed considering a maximum U-235 enrichment of 4.95 wt% and a maximum bumup of 60,000 megawatt-days/ton of uranium, does not exceed 25% of 10 CFR 100 limits. The proposed change does not impact the current analysis and therefore, there is no increase in the dose consequences associated with a FHA.

The probability of having a FHA has not increased. Although it could be postulated that a Metamic panel could be dropped during installation, the approximate 50 pound weight of the panel falling on the racks is bounded by the current fuel assembly drop analysis.

Criticality Accidents associated with a Dropped Fuel Assembly The three fuel assembly drop accidents described below can be postulated to increase reactivity. However, for these accident conditions, the double contingency principle of ANS N16.1-1975 is applied. This states that it is unnecessary to assume two unlikely, independent, concurrent events to ensure protection against a criticality accident. Thus, for accident conditions, the presence of soluble boron in the storage pool water can be assumed as a realistic initial condition since its absence would be a second unlikely event.

Three types of drop accidents have been considered: a vertical drop accident, a horizontal drop accident, and an inadvertent drop of an assembly between the outside periphery of the rack and the pool wall. The structural damage to the pool liner, the racks, and fuel assembly resulting from a dropped fuel assembly striking the rack, the

Attachment I to 1CAN070603 Page 21 of 23 pool floor, or another assembly located in the racks is primarily dependent on the mass of the falling object and drop height. Since these two parameters are not changed by the proposed modification, the postulated structural damage to these items remains unchanged. In all cases the proposed TS limit for boron concentration ensures that a five percent subcriticality margin is met for the postulated accidents.

Criticality Accidents associated with a Misplaced Fuel Assembly The fuel assembly misplacement accident was considered for all storage configurations.

An assembly with high reactivity is assumed to be placed in a storage location which requires restricted storage based on initial U-235 loading, cooling time, and bumup. The presence of boron in the pool water assumed in the analysis has been shown to offset the worst case reactivity effect of a misplaced fuel assembly for any configuration. This boron requirement is less than the boron concentration required by the ANO-1 TS.

Thus, a five percent subcriticality margin is met for postulated accidents, since any reactivity increase will be much less than the negative worth of the dissolved boron.

Optimum Moderation Accident For fuel storage applications in the SFP, water is usually present. An "optimum moderation" accident is not a concern in SFP storage racks because the rack design prevents the preferential reduction of water density between the cells of a rack (e.g.,

boiling between cells). In addition, the criticality analysis has demonstrated that Kff will remain less than 1.0 when the SFP is fully flooded with unborated water.

An "optimum moderation" accident in the new fuel vault was evaluated and the conclusions of that evaluation confirmed that the reactivity effect is less than the regulatory limit of 0.98 for kff.

Loss of SFP Cooling The proposed changes to the ANO-1 SFP racks do not result in changes to the SFP cooling system and therefore the probability of a loss of SFP cooling is not increased.

The consequences of a loss of spent fuel pool cooling were evaluated and found to not involve a significant increase as a result of the proposed changes. A thermal-hydraulic evaluation for the loss of SFP cooling was performed. The analysis determined that the minimum time to boil is more than three hours following a complete loss of forced cooling. This provides sufficient time for the operators to restore cooling or establish an alternate means of cooling before the water shielding above the top of the racks falls below 10 feet. Therefore, the proposed change represents no increase in the consequences of loss of pool cooling.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

to lCAN070603 Page 22 of 23

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The presence of soluble boron in the pool water assumed in the criticality analysis is less than the boron concentration required by the ANO-1 TSs. Thus, a five percent subcriticality margin is met for postulated accidents, since any reactivity increase will be much less than the negative worth of the dissolved boron.

No new or different types of fuel assembly drop scenarios are created by the proposed change. During the installation of the Metamic@ panels, the possible drop of a panel is bounded by the current fuel assembly drop analysis. No new or different fuel assembly misplacement accidents will be created. Administrative controls currently exist to assist in assuring fuel misplacement does not occur.

No changes are proposed to the spent fuel pool cooling system or makeup systems and therefore no new accidents are considered related to the loss of cooling or makeup capability.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

With the presence of a nominal boron concentration, the SFP storage racks will be designed to assure a subcritical array with a five percent subcritical margin (95%

probability at the 95 % confidence level). This has been verified by criticality analyses.

Credit for soluble boron in the SFP water is permitted under accident conditions. The proposed modification that will allow insertion of Metamic poison panels does not result in the potential of any new misplacement scenarios. Criticality analyses have been performed to determine the required boron concentration that would ensure the maximum Ke, does not exceed 0.95. The ANO-1 TS for the minimum SFP boron concentration is greater than that required to ensure Kff does not exceed 0.95.

Therefore, the margin of safety defined by taking credit for soluble boron will be maintained.

The structural analysis of the spent fuel racks along with the evaluation of the SFP structure indicated that the integrity of these structures will be maintained with the addition of the PIAs. The structural requirements were shown to be satisfied, thus the safety margins were maintained.

In addition the proposed change includes a coupon sampling program that will monitor the physical properties of the Metamic absorber material. The monitoring program provides a method of verifying that the assumptions used in the SFP criticality analyses remain valid.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

to 1CAN070603 Page 23 of 23 Based on the above, Entergy concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

6.3 Environmental Considerations The proposed amendment does not involve (i) a significant hazards consideration, (ii)a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

Attachment 2 ICAN070603 Proposed Technical Specification Changes (mark-up)

Spent Fuel Pool Boron Concentration 3.7.14 3.7 PLANT SYSTEMS 3.7.14 Spent Fuel Pool Boron Concentration LCO 3.7.14 The spent fuel pool boron concentration shall be >_4600-2000 ppm. I APPLICABILITY: When fuel assemblies are stored in the spent fuel pool spent fuel

, -a,-

Pool erificn-ialI il h-aInot boon codIrfmotId giItnc thp I f I r .....

I mo mont of fuel ascsmmcs in thc cpcnt ufui pool.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Spent fuel pool boron ---- NOTE-----

concentration not within LCO 3.0.3 is not applicable.

limit.

A.1 Suspend movement of fuel Immediately assemblies in the spent fuel pool.

AND A.24 Initiate action to restore Immediately I spent fuel pool boron concentration to within limit.

OR A-2.2 initiate actien to perform a 1Immediately 6pent fuel pool '.erification.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.14.1 Verify the spent fuel pool boron concentration is

>_4600-2000 ppm.

7 days I

ANO-1 3.7.14-1 Amendment No. 245,

Spent Fuel Pool Storage 3.7.15 3.7 PLANT SYSTEMS 3.7.15 Spent Fuel Pool Storage

~YL. ~ !A!...3 LA ~..J L.

LCO 3.7.15 "H9 cOm, ina o . n ...f-initial .nt n.. aind

-. C -

.uri.. u.

-- *. buu...

I- -- -- L irr.. .- I assembly assemblies shall be stored in Region 2 shall be the spent fuel pool within the acceptable Fange imits of F--i*guTable 3.7.15-1 or in accordance with Specification 4.3.1.1.

APPLICABILITY: Whenever any fuel assembly is stored in Regio)-2-ef-the spent fuel pool. I ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the LCO A.1 ----NOTE----

not met. LCO 3.0.3 is not applicable.

Initiate action to move the Immediately non-complying fuel assembly f49R Re.ien.2to an acceptable storage location in accordance with Table 3.7.15-1.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.15.1 Verify by administrative means the initial nrichmont Prior to storing the and bU.*up parameters associated with ef the fuel fuel assembly in assembly is-are in accordance with Aguwe-Table Re! the spent 3.7.15-1 or Specification 4.3.1.1. fuel pool.

SR 3.7.15.2 Verify Metamic properties are in accordance with, In accordance with and are maintained within the limits of, the Metamic the Metamic Coupon Coupon Sampling Program. Sampling Program.

ANO-1 3.7.15-1 Amendment No. 245,

Spent Fuel Pool Storage 3.7.15 iu Table 3.7.15-1 Bun*up 'Y"e"u Enr"chMRnt Cur..Loadina Restrictions for Spent Fuel Storage Racks MINIMUMIRkllRBURNUP

! i~~r IR VS. INITIAL idII Ilr' AI ENRICHMENT M II.RnM n wi

n i*wl*m i h WI Wm*IW*

(36,4.1) 36 36 32 32 2a8 Non-restricted - I (acceptable range) - --

24 0

0. 20 E

CD 0 16 E

12 Al f /II

.0 8

4

- /

- - Restricted to Checkerboard Spacing I

(unacceptable range) 1.0 (0,1.4) 2.0 3.0 4.0 (0,4.1)

Initial Assembly Average Enrichment (w/o U-235)

ANO-1 3.7.15-2 Amendment No. 24-5,

Spent Fuel Pool Storage 3.7.15 Region I - Minimum Bumup Requirements at Varying Initial U-235 Enrichment and Cooling Time (Notes I & 2)

Enrichment 2_.0 L2.5 3.0 3.5 4.0 . 4.5 1 5.0 Cooling Time Minimum Burnup (GWD/MTU)

(Years) 0 2.1 8.6 15.0 21.8 . 27.2 32.7 38.4 5 2.0 8.2 14.1 20.7 26.2 30.8 36.4 10 1.9 7.9 13.7 19.5 25.5 29.5 34.6 15 1.8 7.7 13.2 18.8 24.7 28.8 33.8 20 1.8 71..5 13.0 18.4 24.2 28.2 33.0 Region 2 - Minimum Burnup Requirements at Varying Initial U-235 Enrichment and Cooling Time (Notes I & 2)

Enrichment 2.0 2.5 1 3.0 1 3.5 4.0 4.5 5.0 Cooling Time Minimum Burnup (GWD/MTU)

(Years) 0 4.0 11.2 18.2 25.0 30.1 36.6 42.7 5 3.8 10.6 17.1 . 23.9 28.8 34.4 40.3 10 3.6 10.1 16.2 22.7 27.8 32.7 38.3 15 3.5 9.8 15.8 21.9 27.1 31.7 37.1 20 3._3 9.6 15.2 21.4 26.6 30.7 36.2 Region 3 Loadinq Restrictions Unrestricted storaae is allowed for fuel assemblies with an initial U-235 enrichment less than or equal to 4.35 wt%.

For fuel assemblies with an initial U-235 enrichment greater than 4.35 wt%, the bumup of at least one fuel assembly in each 2 x 2 section of storage cells is at least 20.1 GWD/MTU.

Note 1:Linear interpolation between burnups for a given cooling time is allowed. However, linear interpolation between cooling times is not allowed, therefore the coolina time of a aiven assembly must be rounded down to the nearest coolinq time.

Note 2:When it is necessary to store fuel assemblies in Region I or Region 2 that do not meet the burnup versus U-235 enrichment restrictions, fuel assemblies, including fresh or irradiated fuel assemblies with a maximum U-235 enrichment of 4.95 wt%, may be stored in a 2 x 2 checkerboard (i.e., 2 assemblies and 2 empty cells) arrangement.

ANO-1 3.7.15-2 Amendment No. 24-,

Spent Fuel Pool Storage 3.7.15 Table 3.7.15-1 (continued)

Loading Restrictions for the Spent Fuel Storage Racks Rack Interface Requirements In addition to the above requirements for each individual rack, the following requirements must be met on the interfaces between and within racks:

a. In the Region 1 and Region 2 racks, a fresh fuel checkerboard and uniform spent fuel loading may be placed in the same rack.

b. In Region 1 and Region 2 racks, if adiacent racks contain a checkerboard of fresh fuel assemblies, the checkerboard must be maintained across the gap, i.e., fresh fuel assemblies may not face each other across a gap.

c. In Region 3. uniform loading of fresh fuel with a maximum U-235 enrichment of 4.35 wt%

may be combined with 3 of 4 loading in the same rack as long as a row of fresh and spent fuel in the 3 of 4 loading pattem faces the uniform loading of all fresh fuel with a maximum U-235 enrichment of 4.35 wt%.

d. If adiacent Region 3 racks contain different loading pattems (one rack contains all fresh fuel with a maximum U-235 enrichment of 4.35 wt% and the other rack contains a 3 of 4 loading pattem), both fresh and spent fuel must be in the outer row of the rack containing the 3 of 4 pattern.

e. If adiacent Region 3 racks both contain 3 of 4 loading patterns, both racks may not have fresh fuel facing the other rack. A loading pattem with both Region 3 racks containing 3 of 4 pattems with all fresh fuel in the outer row of one rack and fresh and spent fuel in the outer row of the second rack is allowed.

f. All interfaces between dissimilar racks (Region 1-Region 3 and Region 2-Region 3) are permitted.

ANO-1 3.7.15-3 Amendment No. I

Fuel Storage 4.3 4.0 DESIGN FEATURES 4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:

a. Fuel assemblies having a maximum U-235 enrichment of 44-.4.95 weight percent;

b. kff _<0.95 iffully flooded with 457 ppm of o~borated water, which includes an allowance for uncertainties as described in Section 9.6.2.4.3 of the SAR;

c. kff < 1.0 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.6.2.4.3 of the SAR; Gd. A nominal 10.65 inch center to center distance between fuel assemblies placed in the storage racks; de. New or partially spent fuel assemblies with a di--har.gc in

'bunUP the "'cceptable rnngc" of stored in accordance with PguTe-Iable 3.7.15-1 allowed un.e.t..ted s.e*ago in eithe -the spent fuel storage racks Region . or Region 2, and ef. New or partially spent fuel assemblies with cooling times, U-235 enrichment or a-discharge burnup in the !unacceptable range! of Figue-Table 3.7.15-1 for fuel assemblies stored in Region 1 or Region 2 may be stored in a 2 x 2,-e--iA~ checkerboard configuration (i.e., 2 assemblies and 2 empty cells); and-ig Reglen

a. Neutron absorber (Metamic) installed between fuel assemblies in the Region 3 racks.

4.3.1.2 The new fuel storage racks are designed and shall be maintained with:

a. Fuel assemblies having a maximum U-235 enrichment of 444.95 weight percent;

b. kff *5 0.95 under normal conditions, which includes an allowance for uncertainties as described in Section 9.6.2.4.3 of the SAR;

c. krf

0.98 with optimum moderation, which includes an allowance for uncertainties as described in Section 9.6.2.4.3 of the SAR;

d. A nominal 21 Inch center to center distance between fuel assemblies placed in the storage racks; and

e. Ten interiFo eoage cl.. , -"Fuel assembly loading prohibited in interior storage cells as shown in Figures 4.3.1.2-1 or 4.3.1.2-2.

based on U-235 fuel enrichment., ,precde4 4fGr-su during fu,,

24.

ANO-14.0-3AmendenteN.

ANO-1 4.0-3 Amendment No. 24.5,

Design Features 4.0 Figure 4.3.1.2-1 Fresh Fuel Storage Rack Loading Pattern for a Maximum Enrichment of 4.95 wt% U-235 I NO NO NO NO NO NO NO NO NO NO NO NO NO NO "NO" Indicates a location in which fuel loading is prohibited.

ANO-1 4.0-5 Amendment No. 245,

Desiqn Features 4.0 Figure 4.3.1.2-2 Fresh Fuel Storage Rack Loading Pattern for a Maximum Enrichment of 4.2 wt% U-235 NO NO NO NO NO NO NO NO "NO" Indicates a location in which fuel loading is prohibited.

ANO-1 4.0-6 Amendment No. I

Programs and Manuals 5.5 5.0 ADMINSTRATIVE CONTROLS 5.5 Programs and Manuals 5.5.16 Reactor Building Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the reactor building as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995, except that the next Type A test performed after the April 16, 1992 Type A test shall be performed no later than April 15, 2007.

In addition, the reactor building purge supply and exhaust isolation valves shall be leakage rate tested once prior to entering MODE 4 from MODE 5 if not performed within the previous 92 days.

The peak calculated reactor building internal pressure for the design basis loss of coolant accident, P., is 54 psig.

The maximum allowable reactor building leakage rate, La, shall be 0.20% of containment air weight per day at P,.

Reactor Building leakage rate acceptance criteria is < 1.OLa. During the first unit startup following each test performed in accordance with this program, the leakage rate acceptance criteria are < 0.60L" for the Type B and Type C tests and

< 0.75L, for Type A tests.

The provisions of SR 3.0.2 do not apply to the test frequencies specified in the Reactor Building Leakage Rate Testing Program.

The provisions of SR 3.0.3 are applicable to the Reactor Building Leakage Rate Testing Program.

ANO-1 5.0-2 Amendment No. 2-14,249,

Programs and Manuals 5.5 5.0 ADMINSTRATIVE CONTROLS 5.5 Programs and Manuals 5.5.17 Metamic Coupon Sampling Program A coupon surveillance program will be implemented to maintain surveillance of the Metamic absorber material under the radiation, chemical, and thermal environment of the SFP. The purpose of the program is to establish the following:

Coupons will be examined on a two year basis for the first three intervals with the first coupon retrieved for inspection being on or before February 2009 and thereafter at increasing intervals over the service life of the inserts.

" Measurements to be performed at each inspection will be as follows:

A) Physical observations of the surface appearance to detect pitting, swelling or other degradation, B) Length, width, and thickness measurements to monitor for bulging and swelling C) Weight and density to monitor for material loss, and D) Neutron attenuation to confirm the B-10 concentration or destructive chemical testing to determine the boron content.

The provisions of SR 3.0.2 are applicable to the Metamic Coupon Sampling Program.

" The provisions of SR 3.0.3 are not applicable to the Metamic Coupon Sampling Program.

ANO-1 5.0-25a Amendment No. I

Attachment 3 ICAN070603 Proposed Technical Specification Changes (revised)

Spent Fuel Pool Boron Concentration 3.7.14 3.7 PLANT SYSTEMS 3.7.14 Spent Fuel Pool Boron Concentration LCO 3.7.14 The spent fuel pool boron concentration shall be > 2000 ppm. I APPLICABILITY: When fuel assemblies are stored in the spent fuel pool. I ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Spent fuel pool boron ---- NOTE------

concentration not within LCO 3.0.3 is not applicable.

limit.

A.1 Suspend movement of fuel assemblies in the spent fuel Immediately pool.

AND A.2 Initiate action to restore I spent fuel pool boron Immediately concentration to within limit.

I SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.14.1 Verify the spent fuel pool boron concentration is 7 days

> 2000 ppm. I ANO-1 3.7.14-1 Amendment No. 24-5,

Spent Fuel Pool Storage 3.7.15 3.7 PLANT SYSTEMS 3.7.15 Spent Fuel Pool Storage LCO 3.7.15 Fuel assemblies shall be stored in the spent fuel pool within the acceptable limits of Table 3.7.15-1 or in accordance with Specification 4.3.1.1.

APPLICABILITY: Whenever any fuel assembly is stored in the spent fuel pool. I ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the LCO A.1 -NOTE---

not met. LCO 3.0.3 is not applicable.

Initiate action to move the Immediately non-complying fuel assembly to an acceptable storage location in accordance with Table 3.7.15-1.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.15.1 Verify by administrative means the parameters Prior to storing the associated with the fuel assembly are in accordance fuel assembly in the with Table 3.7.15-1 or Specification 4.3.1.1. spent fuel pool.

SR 3.7.15.2 Verify Metamic properties are in accordance with, In accordance with and are maintained within the limits of, the Metamic the Metamic Coupon Coupon Sampling Program. Sampling Program.

ANO-1 3.7.15-1 Amendment No. 24-,

Spent Fuel Pool Storage 3.7.15 Table 3.7.15-1 Loading Restrictions for Spent Fuel Storage Racks Region I - Minimum Burnup Requirements at Varying Initial U-235 Enrichment and Cooling Time (Notes I & 2)

Enrichment 2.0 1 2.5 3.0 3.5 4.0 4.5 5.0 Cooling Time Minimum Bumup (GWD/MTU)

(Years) 0 2.1 8.6 15.0 21.8 27.2 32.7 38.4 5 2.0 8.2 14.1 20.7 26.2 30.8 36.4 10 1.9 7.9 13.7 19.5 25.5 29.5 34.6 15 1.8 7.7 13.2 18.8 24.7 28.8 33.8 20 1.8 7.5 13.0 18.4 24.2 28.2 33.0 Region 2 - Minimum Burnup Requirements at Varying Initial U-235 Enrichment and Cooling Time (Notes I & 2)

Enrichment 2.0 1 2.5 3.0 3.5 4.0 4.5 5.0 Cooling Time Minimum Burnup (GWD/MTU)

(Years) 0 4.0 11.2 18.2 25.0 30.1 36.6 42.7 5 3.8 10.6 17.1 23.9 28.8 34.4 40.3 10 3.6 10.1 16.2 22.7 27.8 32.7 38.3 15 3.5 9.8 15.8 21.9 27.1 31.7 37.1 20 3.3 9.6 15.2 21.4 26.6 30.7 36.2 Region 3 Loading Restrictions Unrestricted storage is allowed for fuel assemblies with an initial U-235 enrichment less than or equal to 4.35 wt%.

For fuel assemblies with an initial U-235 enrichment greater than 4.35 wt%/o, the bumup of at least one fuel assembly in each 2 x 2 section of storage cells is at least 20.1 GWD/MTU.

Note 1:Linear interpolation between bumups for a given cooling time is allowed. However, linear interpolation between cooling times is not allowed, therefore the cooling time of a given assembly must be rounded down to the nearest cooling time.

Note 2:When it is necessary to store fuel assemblies in Region 1 or Region 2 that do not meet the bumup versus U-235 enrichment restrictions, fuel assemblies, including fresh or irradiated fuel assemblies with a maximum U-235 enrichment of 4.95 wt%, may be stored in a 2 x 2 checkerboard (i.e., 2 assemblies and 2 empty cells) arrangement.

ANO-11 3.7.15-2 Amendment No. 24-,

Spent Fuel Pool Storage 3.7.15 Table 3.7.15-1 (continued)

Loading Restrictions for the Spent Fuel Storage Racks Rack Interface Requirements In addition to the above requirements for each individual rack, the following requirements must be met on the interfaces between and within racks:

a. In the Region 1 and Region 2 racks, a fresh fuel checkerboard and uniform spent fuel loading may be placed in the same rack.

b. In Region 1 and Region 2 racks, if adjacent racks contain a checkerboard of fresh fuel assemblies, the checkerboard must be maintained across the gap, i.e., fresh fuel assemblies may not face each other across a gap.

c. In Region 3, uniform loading of fresh fuel with a maximum U-235 enrichment of 4.35wt%

may be combined with 3 of 4 loading in the same rack as long as a row of fresh and spent fuel in the 3 of 4 loading pattern faces the uniform loading of all fresh fuel with a maximum U-235 enrichment of 4.35 wt%.

d. If adjacent Region 3 racks contain different loading patterns (one rack contains all fresh fuel with a maximum U-235 enrichment of 4.35 wt% and the other rack contains a 3 of 4 loading pattern), both fresh and spent fuel must be in the outer row of the rack containing the 3 of 4 pattern.

e. If adjacent Region 3 racks both contain 3 of 4 loading patterns, both racks may not have fresh fuel facing the other rack. A loading pattern with both Region 3 racks containing 3 of 4 patterns with all fresh fuel in the outer row of one rack and fresh and spent fuel in the outer row of the second rack is allowed.

f. All interfaces between dissimilar racks (Region 1-Region 3 and Region 2-Region 3) are permitted.

ANO-1 3.7.15-3 Amendment No. I

Fuel Storage 4.3 4.0 DESIGN FEATURES 4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:

a. Fuel assemblies having a maximum U-235 enrichment of 4.95 weight percent;

b. kff -<0.95 if fully flooded with 457 ppm of borated water, which includes an allowance for uncertainties as described in Section 9.6.2.4.3 of the SAR;

c. kff < 1.0 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.6.2.4.3 of the SAR;

d. A nominal 10.65 inch center to center distance between fuel assemblies placed in the storage racks;

e. New or partially spent fuel assemblies stored in accordance with Table 3.7.15-1 in the spent fuel storage racks;

f. New or partially spent fuel assemblies with cooling times, U-235 enrichment or discharge bumup in the unacceptable range of Table 3.7.15-1 for fuel assemblies stored in Region I or Region 2 may be stored in a 2 x 2 checkerboard configuration (i.e.,

2 assemblies and 2 empty cells); and

g. Neutron absorber (Metamic) installed between fuel assemblies in the Region 3 racks.

4.3.1.2 The new fuel storage racks are designed and shall be maintained with:

a. Fuel assemblies having a maximum U-235 enrichment of 4.95 weight percent;

b. ke -*0.95 under normal conditions, which includes an allowance for uncertainties as described in Section 9.6.2.4.3 of the SAR;

c. keff 0.98 with optimum moderation, which includes an allowance for uncertainties as described in Section 9.6.2.4.3 of the SAR;

d. A nominal 21 inch center to center distance between fuel assemblies placed in the storage racks; and

e. Fuel assembly loading prohibited in interior storage cells as shown in Figures 4.3.1.2-1 or 4.3.1.2-2, based on U-235 fuel enrichment.

ANO-1 4.0-3 Amendment No. 245,

Design Features 4.0 Figure 4.3.1.2-1 Fresh Fuel Storage Rack Loading Pattern for a Maximum Enrichment of 4.95 wt% U-235 I NO NO NO NO NO NO NO NO NO NO I I .~ .t I "NO" Indicates a location in which fuel loading is prohibited.

ANO-1 4.0-5 Amendment No. 246,

Design Features 4.0 Figure 4.3.1.2-2 Fresh Fuel Storage Rack Loading Pattern for a Maximum Enrichment of 4.2 wt% U-235 NO NO NO NO NO NO NO NO

4. 4 4 4 6 I

4. 4. 4 4 I I 4.

"NO" Indicates a location in which fuel loading is prohibited.

ANO-1 4.0-6 Amendment No. I

Programs and Manuals 5.5 5.0 ADMINSTRATIVE CONTROLS 5.5 Programs and Manuals 5.5.16 Reactor Buildinq Leakage Rate Testinq Program A program shall be established to implement the leakage rate testing of the reactor building as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995, except that the next Type A test performed after the April 16, 1992 Type A test shall be performed no later than April 15, 2007.

In addition, the reactor building purge supply and exhaust isolation valves shall be leakage rate tested once prior to entering MODE 4 from MODE 5 if not performed within the previous 92 days.

The peak calculated reactor building internal pressure for the design basis loss of coolant accident, P,, Is 54 psig.

The maximum allowable reactor building leakage rate, L8, shall be 0.20% of containment air weight per day at P,.

Reactor Building leakage rate acceptance criteria is _ 1.0L.. During the first unit startup following each test performed in accordance with this program, the leakage rate acceptance criteria are < 0.60L, for the Type B and Type C tests and

< 0.75L, for Type A tests.

The provisions of SR 3.0.2 do not apply to the test frequencies specified in the Reactor Building Leakage Rate Testing Program.

The provisions of SR 3.0.3 are applicable to the Reactor Building Leakage Rate Testing Program.

ANO-1 5.0-25 Amendment No. 241,2449,

Programs and Manuals 5.5 5.0 ADMINSTRATIVE CONTROLS 5.5 Programs and Manuals 5.5.17 Metamic Coupon Samplincq Pro-gram A coupon surveillance program will be implemented to maintain surveillance of the Metamic absorber material under the radiation, chemical, and thermal environment of the SFP. The purpose of the program is to establish the following:

" Coupons will be examined on a two year basis for the first three intervals with the first coupon retrieved for inspection being on or before February 2009 and thereafter at increasing intervals over the service life of the inserts.

" Measurements to be performed at each inspection will be as follows:

A) Physical observations of the surface appearance to detect pitting, swelling or other degradation, B) Length, width, and thickness measurements to monitor for bulging and swelling C) Weight and density to monitor for material loss, and D) Neutron attenuation to confirm the B-10 concentration or destructive chemical testing to determine the boron content.

The provisions of SR 3.0.2 are applicable to the Metamic Coupon Sampling Program.

The provisions of SR 3.0.3 are not applicable to the Metamic Coupon Sampling Program.

ANO-1 5.0-25a Amendment No. I

Attachment 4 ICAN070603 Changes to Technical Specification Bases Pages For Information Only

Spent Fuel Pool Boron Concentration B 3.7.14 B 3.7 PLANT SYSTEMS B 3.7.14 Spent Fuel Pool Boron Concentration BASES BACKGROUND As described in the Bases for LCO 3.7.15, "Spent Fuel Pool Storage," fuel assemblies are stored in the spent fuel pool racks in accordance with criteria based on initial U-235 enrichment, cooling time, and discharge bumup. Although the water in the spent fuel pool is normally borated to !>4600-2000_ppm, the criteria that limit the storage of a fuel assembly to specific rack locations are conservatively developed withut taking credit for a boron concentration of 457 ppm in the spent fuel pool water.

The spent fuel storage pool is divided into twe-three separate and distinct regions as shown in SAR Figure 9-53 which, for the purpose of criticality considerations, are considered as separate peelginfinite arrays._ Regon I ajs deoignod to ac..ommodat new fuel with a mvaximum ,ichmnent of 4.105.0 m'% U-235, or spe nt.(Oadiatd) fuel regardless of thee disc.harge fuel bumup. Region I 'nd-AoThespent fuel pool racks 2 Is are designed to accommodate fuel of various initial U-235 enrichments whieh ha've accumudated- bumup6,within

.m..inimum

. the acceptable domain according to FigUre-Table 3.7.15-1. Fuel assemblies not meeting the criteria of Figu e-Table 3.7.15-1 shall be stored in accordance with Specification 4.3.1.1.ef. The criticality considerat.on for the cask are the same as rcgUired forF Region I of1th st fuel pool 6torage locationS.

The water in the spent fuel storage pool normally contains soluble boron, which results in large subcriticality margins under actual operating conditions. However, the NRC guidelines (10 CFR 50.68) specify that the limiting kIf of 0-*51.0 be evaluated in the absence of soluble boron. The NRC -guidelinesalso require that the limiting k,ff of 0.95 may be evaluated considering soluble boron or the absence of soluble boron. Hence, the design of beth-the three regions is based on the use of uriborated water, which maintains each region in a subcritical condition during normal operation with the regions fully loaded. The double contingency principle discussed in ANSI N-16.1-1975 and the April 1978, NRC letter (Ref. 1) allows credit for soluble boron under other abnormal or accident conditions, since only a single accident need be considered at one time. Thus, for accident conditions, the presence of soluble boron in the spent fuel pool water can be assumed as a realistic condition. For example, accident scenarios are postulated which could potentially increase the reactivity and reduce the margin to criticality. To mitigate these postulated criticality related accidents, boron is dissolved in the pool water. Safe operation of the high density storage racks with no movement of assemblies may therefore be achieved by controlling the location of each assembly in accordance with LCO 3.7.15, "Spent Fuel Pool Storage." Prior to movement of an assembly, it is necessary to perform SR 3.7.15.1.

ANO-1 B 3.7.14-1 Amendment No. 24-6, Rev.

Spent Fuel Pool Boron Concentration B 3.7.14 APPLICABLE SAFETY ANALYSES MoGt acciadent conditions YAIl not result inan incroase in K.4+of the rack. Examplesar the less of cooling systems (reactivity decreasce with Iccrasin water density)-and droppin;g a fuol assembly,on top of the rack (the rack truc,tur-e pert1nent for criticality is net d-eformcd and the assembly has more that eight inches of water separating it from the active fuel in the rest of the Frck which precludes, interaction). Ho~eweor,-Aaccidents can be postulated which would increase reactivity such as inadvertent drop of an assembly between the outside periphery of the rack and the pool wall. Thus, for accident conditions, the presence of soluble boron in the storage pool water is assumed as a realistic initial condition.

The presence of greater than 1600 ppm boron in the pool water will decrease reactiVit by, conservatively assures (based on the loss of all Metamic) the fuel assemblies will be maintained in a subcritical array with a..pprimately 30% AK. Thus Keff 9 0.95 eaR-be easily met for postulatdin the event of a postulated drop accidentes., -snc* a..Feacti"*y incease will be mziuch loss;r- than the negative worth of the dissolved boron.

Analysis has shown that, during a postulated misplacement accident with the fuel stored within the limits of this specification, that a Kf, _<0.95 will be maintained when the boron concentration is at or above 889 ppm.

The concentration of dissolved boron in the fuel storage pool satisfies Criterion 2 of 10 CFR 50.36 (Ref. 4).

LCO The specified concentration-s> 4600-2000 ppm of dissolved boron in the spent fuel pool conservatively preserves the assumption used in the analyses of the potential accident scenarios. This concentration of dissolved boron is the minimum required concentration for fuel assembly storage and movement within the spent fuel pool.

APPLICABILITY This LCO applies whenever fuel assemblies are stored in the spent fuel pool, or cask loading pit (when the gate is open.), until a complete spent fuel pool verification has been peofermed fellow~n th las mo ivement of fuel assemblies in the spent fuel pool.-

This LCGO does net apply following the verification since thever~inficaio Would confrm thatthrea:re no misloadod fuel assemblies. With nofurther fuel assembly movemenQt i progress, there is no potential for a misloaded fuel assembly Or a dropped fuel ACTIONS A.1- and A.2.4,--and-A.2.2 The Required Actions are modified by a Note indicating that LCO 3.0.3 does not apply.

If moving irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operation. Therefore, inability to suspend movement of fuel assemblies is not a sufficient reason to require a reactor shutdown.

ANO-1 B 3.7.14-2 Amendment No. 245, Rev. 4-3,

Spent Fuel Pool Boron Concentration B 3.7.14 ACTIONS (continued)

A.1T and A.2.,Iad .A..222 (continued)

When the concentration of boron in the spent fuel pool is less than required, immediate action must be taken to preclude the occurrence of an accident or to mitigate the consequences of an accident in progress. This is most efficiently achieved by immediately suspending the movement of the fuel assemblies. This does not preclude movement of a fuel assembly to a safe position. In addition, action must be immediately initiated to restore the spent fuel pool boron concentration to within its limit. An acceptable altcrnatiy.e is to immediately initiate porfor~a ncc of a Spent fuel pool Yrification to ensure proper locations of the fuel since the last moeomcnt of fuel assemblios in the fuol pool. HoPe r-,Pprior

.pent p to resuming movement of fuel assemblies, the concentration of boron must be restored. Either of these actions are acceptable, and once initiated must be con-rtinued until the action is completed-.The immediate Completion Time for initiation of these actions reflects the importance of maintaining a controlled environment for irradiated fuel.

SURVEILLANCE REQUIREMENTS SR 3.7.14.1 This SR verifies that the concentration of boron in the spent fuel pool and cask loading pit is within the required limit. As long as this SR is met, the analyzed incidents are fully addressed. The 7 day Frequency is appropriate because no major replenishment of pool water is expected to take place over a short period of time.

REFERENCES

1. Double contingency principle of ANSI N16.1-1975, as specified in the April 14, 1978, NRC letter (Section 1.2) and implied in the proposed revision to Regulatory Guide 1.13 (Section 1.4, Appendix A).

2. SAR, Section 14.2.2.3.

3. Safety Evaluation Report, Section 2.1.3, License Amendment No. 76, April 15, 1983.

4. 10 CFR 50.36.

5. 10 CFR 50.68.

ANO-1 B 3.7.14-3 A^"i*ondment No. 215 Rev.43,

Spent Fuel Pool Storage B 3.7.15 B.3.7 PLANT SYSTEMS B 3.7.15 Spent Fuel Pool Storage BASES BACKGROUND The spent fuel assembly storage facility is designed to store either new (nonirradiated) nuclear fuel assemblies, or burned (irradiated) fuel assemblies in a vertical configuration underwater. The spent fuel pool is sized to store 968 fuel assemblies and is connected to a pit for loading shipping or dry fuel storage casks. The spent fuel storage cells are installed in parallel rows with center to center spacing of 10.65 inches in each direction.

The cask configuration is in accordance with the cask vendors ve..de.s-Certificate of Compliance.

The spent fuel storage pool is divided into two-three separate and distinct regions as shown in SAR Figure 9-53 which, for the purpose of criticality considerations, are considered as separate pools. Region 1 is designed to acc,,mmodate now fuel with a maximu enrichmcnt of 1.10 ':.'% U 235, or spent (irradiated) fuel regardless of the dicharge fuol bumup. -nd Region 2 is-are designed to accommodate fuel of various initial U-235 enrichments which have accumulated minimum burnups within the acceptable domain according to rigu Table 3.7.15-1. Fuel assemblies not meeting the GRtea-acceptable range of Figure-Table 3.7.15-1 shall be stored in accordance with paragraph 4.3.1.1 .e-f in SAR-TSSection 4.3, Fuel Storage. Region 3 is designed to accommodate new fuel with loading restrictions with a maximum initial fuel assembly enrichment of 4.35 or 4.95 wt% U-235 in accordance with Table 3.7.15-1. The supporting analysis included U-235 fuel enrichments of up to 5.0wt.% plus 0.05wt.% for uncertainties (Ref.5). Te criticality considerationS for thA. cask are the same as ro.uired for Rcgion I of the spcnt fucl pool storage locations.

APPLICABLE SAFETY ANALYSES Criticality of fuel assemblies in the spent fuel storage rack and casks is prevented by the design of the rack or cask, which limits fuel assembly interaction. This is done by fixing the minimum separation between assembhlies and inserting neutFro poison betw:een assemblies in Region I .Region 1 and- Region 2 controls fuel assembly interaction by fixing the minimum separation between assemblies and by setting U-235 enrichment.,

aRd burnup, and cooling time criterion to limit fissile materials. Region 3 controls fuel assembly interaction similar to Region 2 and utilizes Metamic poison panels. ,Iis-ie Sufcient to maintain Aa kff of:9 0.95 for spent fuel of original U-235 enrichment of up to 4-.04.95 wt% is accomplished by taking credit for boron (457 rrmn"nd itions. rquire n" -"'; - s),. A k6 of < 1.0 is accomplished when no credit for boron is taken.

However, fuel assemblies to be stored inthe spent ful:6 pool Region 2 whic~h doe not meet kbrOf 0.95 or less. n;order to prevetinderet fuel assembly insertion-r into9 two adjacn storaeloations, vacant spacoes adjacent to the faces of any fuel asebl which does. not meet the Region 2 bDurnup critera (unrestricted) are physically bloced before any such fuel assembly is placed in Region 2 (Ref-. 1). in addition, the area designated for checkerboard arrangemnent is,divide from the normal storage in Region 2 by a row of vacant storage spaces (Ref. 2.Ž ANO-1 B 3.7.15-1 Amendment hno 245, Rev. 43,

Spent Fuel Pool Storage B 3.7.15 Required Soluble Boron Concentrations for Accident Conditions TS 3.7.14 includes the requirement for greater than 46002000 ppm boron concentration to assure the fuel assemblies will be maintained in a subcritical array with K _0.95 in the event of a postulated drop accident. Analysis has shown that, during a postulated misplacement accident with the fuel stored within the limits of this specification, that a Kff *<0.95 will be maintained when the boron concentration is at or above 889 Dpm.

The spent fuel pool storage satisfies Criterion 2 of 10 CFR 50.36 (Ref. 3).

ANO-1 B 3.7.15-1 .A.mondmot No. 245, Rev. 4-3,

Spent Fuel Pool Storage B 3.7.15 LCO The restrictions on the placement of fuel assemblies within the fuel pool, according to Figure-Table 3.7.15-1 or equivalent cask criticality analysis, ensure that the k,, of the spent fuel pool will always remain _ 0.95 assuming the pool to be flooded with uaborated water. The restrictions are consistent with the criticality safety analysis performed for the spent fuel pool. Fuel assemblies not meeting the U-235 enrichment and bumup criteria shall be stored in accordance with Specification 4.3.1.1.

The kf of the spent fuel pool will remain below 1.0 when the pool is flooded with unborated water.

In the event a checkerboard storage configuration is deemed necessary fe.F a pe*tieAef Region 2, spaces, void of fissionable material, adjacent to the faces of any fuel assembly which does not meet the RegieR-2-burnup criteria (non-restricted) shall be physiGall, bkwAkedconfirmed before any such fuel assembly may be placed in44Re-21the spent fuel pool. This will prevent inadvertent fuel assembly insertion into two adjacent storage locations.

APPLICABILITY This LCO applies whenever any fuel assembly is stored in Region-2-2f-the spent fuel pool.

ACTIONS A._1 Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply. If moving irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operation. Therefore, in either case, inability to move fuel assemblies is not sufficient reason to require a reactor shutdown.

When the configuration of fuel assemblies stored in the spent fuel pool is not in accordance with Fi4gue-Table 3.7.15-1, immediate action must be taken to make the necessary fuel assembly movement(s) to bring the configuration into compliance with Fiqu e-Table 3.7.15-1 or Specification 4.3.1.1.

SURVEILLANCE REQUIREMENTS SR 3.7.15.1 This SR verifies by administrative means that the initial U-235 enrichment, cooling time, and burnup of the fuel assembly is-are in accordance with Figu e-Table 3.7.15-1.in the accompanying LCO or Specification 4.3.1.1. For fuel assemblies in the unacceptable range of Figure-Table 3.7.15-1, performance of the SR will ensure compliance with Specification 4.3.1.1.

ANO-1 B 3.7.15-3 A..mdo...et N. 215 Rev.

Spent Fuel Pool Storage B 3.7.15 SURVEILLANCE REQUIREMENTS (continued) I SR 3.1.15.2 This SR verifies the properties of Metamic@ are maintained in accordance with the Metamic Coupon Sampling Program (TS 5.5.17).

REFERENCES

1. SAR, Section 9.6.2.

2. SER for ANO-1 License Amendment No. 76, Section 2.1 (0CNA048314), dated April 15, 1983.

3. 10 CFR 50.36.

4. 10 CFR 50.68.

5. CALC-ANO-ER-02-016, Rev 0. Criticality Safety Evaluation of the ANO-1 Spent Fuel Pool Storage Racks ANO-1 B 3.7.15-3 Amendment No. 2215 Rev.

Attachment 6 ICAN070603 Evaluation of Spent Fuel Pool Structural Integrity for Increased Loads from Spent Fuel Racks

STRUCTURAL/SEISMIC CONSIDERATIONS FOR ADDITION OF METAMIC PANELS TO THE FLUX TRAPS OF TWO SPENT FUEL RACKS AT ANO-1

Stevenson & Associates Report 06Q3571.01-01 1.0 Introduction The overall design objectives of the spent fuel storage pool at Arkansas Nuclear One (ANO)

Unit 1 are governed by various Regulatory Guides, the Standard Review Plan, and industry standards. The addition of Metamic poison panels to the flux traps of two of the spent fuel storage racks adds additional deadweight to these racks and consequently to the pool.

Additionally, this added weight (mass) changes the seismic load effects on the affected racks and on the pool structure. The structural adequacy of the Spent Fuel Pool (SFP) maximum density spent fuel racks at ANO Unit 1 and the SFP structure, with the poison inserts included, was evaluated using the appropriate regulatory and design standards. Postulated loadings for normal, seismic, and accident conditions at the ANO Unit 1 site are considered in this analysis and evaluation.

The design adequacy of the racks and the poison inserts, are confirmed with analyses that are performed in compliance with the USNRC Standard Review Plan [1], the USNRC Office of Technology Position Paper [2], Lawrence Livermore Report UCRL52342 [3] and ANO Specification APL-C-502 [4]. This report is a summary of the Ref. [5] detailed calculation performed to assess the design adequacy of the racks with the poison inserts and the Reference

[26] calculation performed to assess the effect of the change in loads to the spent fuel pool structure. This report includes a description of the rack layout in Section 2, the methodology used to analyze the rack structures in Section 3, the development of the SOLVIA [6 and 7]

structural dynamic models in Section 4, the applicable load combinations in Section 5, a summary of all the analyses for the racks performed in Section 6, the acceptance criteria in Section 7, the analysis results in Section 8, Description of the Pool Structure Review in Section 9, Conclusions in Section 10, and References in Section 11.

2.0 Rack Layout and Description 2.1 Rack Layout Description The ANO Unit 1 Spent Fuel Pool contains eight independent rack structures designed to hold the spent fuel assemblies and rod cluster control assemblies in storage for long term decay. There are three regions of racks. The Region 1 racks employ Boroflex as the poison material. Region 2 racks do not have any poison material. Region 3 racks are Region 2 racks that will be modified by inserting Metamic poison material strips into the flux traps of the cells. The pool layout is illustrated in Figure 2.1, including the rack modules and the X and Y coordinate axes used in the model development.

The racks are free standing on fourteen feet that rest on the bottom of the pool. The eight racks, originally designed by Westinghouse, are self-supporting and are not connected to each other or to the SFP walls. There are two Region 1 racks, four Region 2 racks and two proposed Region 3 racks (that are modified Region 2 racks).

Page 2 of 50

Stevenson & Associates Report 06Q3571.01-01 m-24c 11x1l Rack II-IHI M-244D 11xl1 Rack 121 Cells 121 Cells I I I I I I I I I I I I I Region 2 Rack, w h Flux Traps nal*dyssRadc #5 6000 Sedes Nodes t

Noirth

.- 2,4II I4A M-244B 11WxlRack 11xl1 Rack 121 Cells 121 Cealls Regionm2 Racl, wi h Flux Traps Analys~s Rack#4 Anayss Rack #3 30Seies Irr' II

%Nodes 2 edes Nodes

--- M-243A M-243B LL 11x12 Rack 11x12 Rack 132 Cells 132 Cells 241 Flux Traps _ 241 Flux Traps Proposed Location of Reon 3 Rai ks with the Inserted Poison II Analysis Rack #2 Analyss Rack #1 1000 Series Nodes Base Sedes Nodes

- - I IK I I M-242 M-241 100l Rack

- 10xl Rack 110 Cells 110 Cells Region I IRa.ks witg i I rail ,x Analysis Rac8 Analss Rack#7 7000 Series Nodes 6 es Nodes Figure 2.1 - ANO Unit 1 Spent Fuel Pool Layout Page 3 of 50

Stevenson & Associates Report 06Q3571.01-01 2.2 Material Properties of Rack, Fuel and Poison Inserts (Design Inputs)

The high density storage rack weights from Reference [8] are listed in Table 2.1 below. The weights used in the analysis were within about 3% of the values below as discussed in Section 4.1.

Table 2.1 RACK WEIGHT DATA Rack # Empty Rack Dry Weight Per Figure 2.1 Per Ref. 8 Cells/Module Array Size (lbs) 1 (M-243B-Region 3) 3 132 1lx12 19,150 2 (M-243A-Region 3) 3 132 1lx12 19,150 3 (M-244B--Region 2) 4 121 llxl1 17,650 4 (M-244A-Region 2) 4 121 1lxl1 17,650 5 (M-244D--Region 2) 4 121 1lxl1 17,650 6 (M-244C-Region 2) 4 121 llxl1 17,650 7 (M-241-Region 1) 1 110 1Oxil 27,650 8 (M-242-Region 1) 2 110 1Oxi 1 27,650 The Cartesian coordinate system used within the rack dynamic model has the following nomenclature:

X = Horizontal axis along plant North Y = Horizontal axis along plant West Z = Vertical axis upward from the rack base The material properties for the rack and support material are summarized in Table 2.2 below.

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Stevenson & Associates Report 06Q3571.01-01 Table 2.2 RACK MATERIAL DATA (ASME - Section 11, Part D)

MATERIAL DATA (To = 150-F)

Stainless Steel Young's Modulus Yield Strength Tensile Strength Material E Sy S (psi) (psi) (psi)

SA240, Type 304 27.7 x 106 27,500 73,000 SA479, Type 304 27.7 x 106 27,500 73,000 MATERIAL DATA (Ta = 250-F)

SA240, Type 304 27.3 x 106 23,750 68,500 SA479, Type 3041 27.3 x 106 23,750 68,500 3.0 Overview of Rack Structural Analysis Methodology The response of a free-standing rack module to seismic loadings is nonlinear and involves a complex combination of motions (rocking, twisting, turning and sliding). This could potentially cause impacts within the structure (fuel assemblies to the cell walls) and between modules, or between modules and the pool walls. Rack dynamic behavior includes a large portion of the total structural mass in a confined rattling motion. The rack pedestals are restricted from lateral motion only by friction at the base. In addition, there are large fluid coupling effects due to water around the assemblies and the independent adjacent structures.

Linear dynamic analysis methods cannot reasonably simulate the structural response of these highly nonlinear structures when subjected to earthquake loadings. An appropriate simulation can only be obtained by direct integration of the nonlinear equations of motion with three directional pool slab time-histories applied as forcing functions acting on the structures simultaneously.

Whole Pool Multi-Rack (WPMR) analysis is used to obtain final analysis results in order to simulate the dynamic behavior of the storage rack structures. This section describes the methodology used in the analysis.

3.1 Analysis Methodology Background Reliable assessment of stresses within the rack components and stored fuel behavior within the rack modules requires a dynamic model that incorporates the appropriate attributes of the actual Page 5 of 50

Stevenson & Associates Report 06Q3571.01-01 structure. The model must feature the ability to simultaneously simulate concurrent motions compatible with the rack and fuel storage installation.

The model has the capability to affect interactions, which occur due to rattling of fuel assemblies inside storage cells, and lift-off of the support pedestals on the pool floor. The contribution of the water mass in the spaces around the rack modules and within the storage cells is modeled in an accurate manner as described below.

The friction coefficient at the pedestal-to-pool liner (or bearing pad) interface may lie in a rather wide range and a conservative value of friction cannot be prescribed without performing bounding simulations. Different friction coefficients provide the governing results for different analysis parameters. For example, the lower bound friction results in the largest overall rack displacement, which may seem obvious, however other parameters such as the impact force between the rack and fuel assembly being largest with the upper bound friction is a result not immediately predictable.

The approach used in this evaluation was to develop single rack models for the Region 3 type rack structures, since these are the racks being modified relative to the current seismic qualification analysis (Reference 9). The three-dimensional single rack dynamic model addresses the parameters discussed above. Single rack simulations may not by themselves be sufficient in determining the maximum dynamic response. This is due to the participation of water around the racks, with hydraulic interaction that may either increase or decrease rack motion. The results of this evaluation confirm that the dynamics of one rack affects the motion of the others in the pool. Therefore, the dynamic simulation of one rack, while providing a great deal of insight into this behavior, may not adequately predict the motion or structural response (applied forces and internal stresses) of rack modules.

For this reason, the hydraulic and dynamic interaction of closely spaced racks is simulated by including all modules in one comprehensive simulation using a WPMR model. All rack modules are modeled simultaneously and the coupling effect due to multi-body motion is included in the analysis. Region 2 rack models for the whole pool model were developed from the Region 3 rack model with corrections for one less row of cells. The Region 2 and Region 3 racks are identical in construction except for the number of rows of cells. Similarly, the Region 1 rack models for the whole pool model were developed from the Region 3 rack model, with appropriate changes to section properties due to the differences in cell cross-section and additional framing members present in the Region 1 racks.

The models developed as described below, consist of beam elements to model the rack, fuel elements, and Metamic inserts. Spring/gap elements and contact surfaces were used to account for the racks being unanchored and for possible impacts between the racks and pool walls, rack to rack, and fuel element or Metamic insert to rack cell wall. Mass elements were used to include the added mass to account for hydrodynamic effects.

Modeling of the racks, fuel assemblies, and inserts in this manner for analysis using SOLVIA was shown to be adequate by performing a fluid-structure analysis study using fluid elements and shell elements to model box type fuel rack structures. From this fluid-structure analysis study, good agreement on the behavior of the beam/stick model was demonstrated.

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Stevenson & Associates Report 06Q3571.01-01 Also from the fluid-structure study, it was determined that the presence and movement of the racks in the pool during seismic loading does not significantly effect the hydrodynamic pressure on the pool walls when compared to the response of the water only, with no submerged structures included.

3.2 Equation of Motion Program SOLVIA was used for the dynamic non-linear time history analysis of the single rack and WPMR model of the structures. Using the direct time integration method, the equations of motion are solved at each time step for acceleration time histories in each of the three degrees of freedom. The basic equations that SOLVIA is operating on are:

Mfti(t) + Cti(t) = R(t) - F(t) where:

M constant mass matrix, C = constant damping matrix, R(t) = external load vector applied at time t, F(t) = nodal point force vector equivalent to the element stresses at time t, A superimposed dot denotes time derivative, e.g.,

Q(t) = nodal point velocity vector at time t.

ii(t) = nodal point acceleration vector at time t.

An implicit time integration method is employed for this structural vibration problem.

There are several non-linear attributes and unique hydrodynamic properties of this structure that are modeled. The model has been built by modeling each attribute and checking their effects one at a time. Each single rack model is developed by appropriately combining these attributes. The WPMR is modeled by combining the eight modules and including the appropriate off diagonal stiffness matrix and mass matrix terms that include the interactions between the modules.

3.3 Friction Coefficient Between Rack Supports and Pool Floor It is not possible to determine an accurate coefficient of friction (g) between the pedestal supports and the pool floor. Data on austenitic stainless steel plates submerged in water show a mean value of p to be 0.503 [Ref. 13] with a standard deviation of 0.125. Upper and lower bounds (based on twice the standard deviation) are 0.753 and 0.253, respectively. Therefore, coefficient of friction values of 0.2 (lower limit) and 0.8 (upper limit) as well as a best estimate value of 0.5 provide reasonable limits and provide a reasonable envelope for calculating the upper bound module response for each design parameter.

The friction interface between rack support pedestal and liner in the fuel rack simulations is simulated by linear contact (friction) elements. These elements function only when the pedestal is physically in contact with the pool floor. Friction elements are also included at the base of the fuel rod to rack base interface to reasonably model the behavior of the rod at this juncture. The Page 7 of 50

Stevenson & Associates Report 06Q3571.01-01 coefficient of friction modeled at this interface was consistent with that used for the pedestal/pool bottom interface for a given analysis.

3.4 Rack Beam Behavior The structural model using an equivalent beam stiffness developed for the full cell structure, was modeled using linear beam members to represent the elastic bending and twisting action.

The equivalent moment of inertia for the beam was estimated using a shell element model of a row of cells with the appropriate number of cells included for each horizontal direction. The axial area was estimated using a single cell model. The overall combined section properties for each type of rack module were then estimated from results of analysis of these models for applied unit displacements.

3.5 Impact Behavior To include the impact behavior, compression-only gap elements are used to provide for opening and closing of interfaces such as the pedestal-to-pool floor interface and the fuel assembly-to-cell wall interface. These interface gaps are modeled using nonlinear spring elements (Gapped Truss elements in SOLVIA). The nonlinear spring is the mathematical representation of the condition where a restoring force is zero until the gap is closed and then is linearly proportional to displacement.

3.6 Fuel Loading to Cell Wall Behavior The fuel assemblies are conservatively assumed to rattle in unison, which provides an upper bound for the contribution of impact against the cell wall. This is modeled with a single spent fuel assembly, which is a combination of all the assemblies contained in the rack. This single assembly is allowed to rattle against the cell walls of the rack modeled as an equivalent beam element. This results in the impact load being a combination of all 132 fuel assemblies hitting the wall at the same time.

From Reference 3, it is noted that impact damping is a significant source of damping for multiple impacting members. The same effective damping due to fuel to cell impact as a function of mass and stiffness presented in Reference [9] was used. From Reference [9], the damping coefficient was calculated as:

C=2xdampingx*i5ý*Af where C = effective damping coefficient K= impact stiffness m = mass Af = area damping = 2%

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Stevenson & Associates Report 06Q3571.01-01 3.7 Fluid - Rack Coupling The WPMR model used for this analysis handles simultaneous simulation of all racks in the pool as a WPMR three dimensional analysis. The WPMR analysis is appropriate for predicting maximum structural stresses with reasonable predictions of rack dynamic response.

During an earthquake, all racks in the pool are subject to the input excitation simultaneously.

While the possibility of inter-rack impact is not a common occurrence and depends on rack spacing, the effect of water (the fluid coupling effect) is a factor. It is, therefore, essential that the contribution of the fluid forces be included in a comprehensive manner. This is possible when all racks in the pool are included in a three dimensional simulation using a mathematical model that includes all modules moving simultaneously. The fluid coupling effect encompasses interaction between every set of racks in the pool. The motion of one rack effects the fluid forces on all other racks and on the pool walls. Therefore, both near-field and far-field fluid coupling effects are included in the analysis.

3.8 Poison Insert Analysis Methodology The poison inserts are nominally smaller in cross-section than the inside dimensions of the flux traps. As such, it is likely that most of the inserts will rattle inside the flux traps and impact the sides similar to the behavior of the fuel bundles inside the cells. For this reason, the poison inserts were analyzed together with the rack and spent fuel assemblies by including them as additional beam elements in the structural model. Two beam elements were used for the poison inserts due to 121 inserts oriented 90 degrees to the other 120 inserts.

Properties for the beam elements representing the poison inserts were calculated by modeling a single insert, and obtaining equivalent beam properties in a similar manner as was done for the rack. Forces and moments on the inserts were then obtained directly from the single rack and whole pool model. Stresses on individual inserts are calculated using these forces and moments from overall rack module models.

3.9 Whole Pool Multi-Rack (WPMR) Methodology The WPMR analysis must deal with both stress displacement and impact criteria. The model development and analysis steps that are undertaken are summarized in the following steps.

a. The section and mass properties of a single cell are developed.

b. Using the single cell section and mass properties, equivalent properties for each rack module are developed.

c. Similarly, single element properties are calculated for the fuel assembly, poison inserts and the base pedestals. These are also used to develop equivalent properties for the rack module.

d. Individual stiffness used in the gap elements are calculated for each of the interfaces included in the model. These include the pedestal base to pool floor, rack to rack and rack to wall stiffness and fuel assembly to rack wall interface.

For the new Region 3 racks, the interfaces between the poison inserts and the flux Page 9 of 50

Stevenson & Associates Report 06Q3571.01-01 traps were also included. These are also appropriately combined to get equivalent module properties.

e. Calculate the appropriate hydrodynamic properties for the spent fuel assemblies and rack. This includes the hydrodynamic mass and the off-diagonal hydrodynamic mass matrix terms.

f. Develop the individual or single rack models in the pool.

g. Combine the single rack models into one three-dimensional dynamic model suitable for a time-history analysis of the racks. These models include the assemblage of all rack modules in the pool. Include all fluid coupling interactions and mechanical coupling appropriate to performing an accurate non-linear simulation.

h. Perform the three-dimensional dynamic analyses on various physical conditions (such as coefficient of friction and extent of cells containing fuel assemblies).

Archive the appropriate displacement and load outputs from the dynamic model for post-processing.

i. Using the force and moment outputs from the dynamic analyses, perform stress analysis of high stress areas for the limiting cases. Use simple modeling techniques to evaluate the local regions of the structure that need to be evaluated.

Demonstrate compliance with ASME Code Section III, Subsection NF limits on stress and displacement.

4.0 Rack Model Development 4.1 Single Rack Module Development The Region 3 rack includes 132 (11 by 12) cells. The weight of each component from Reference 8 is as follows:

Rack weight = 19,150 lb Fuel weight = 1,700 lb* 132 Assemblies = 224,400 lb Inserted Poison Panels = 100 lb

241 = 24,100 lb Total weight (dry) = 267,650 lb In order to verify the Region 3 rack weight the weight was estimated using the actual mass densities and using the structural drawings. The weight of the various structure components were calculated as follows:

Base plate Wbp = 128.8 x 118.5 x 0.5 x 0.29 = 2.213x10 3 lb Rack walls W,. = 373.256 x 162 x 0.29 = 17,536 lb Total WR = 17,536 + 2,210 = 19,746 lb (use 19,750 lb)

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Stevenson & Associates Report 06Q3571.01-01 Although the total weight is close to the 19,150 lb weight from Ref. [8] (within about 3%) a weight of 19,750 lb calculated above was used in the analysis for the Region 3 racks. Using the same method 18,100 lb was calculated and used in the analysis for the Region 2 racks. The weight of the Region 1 racks was not verified, however, based on the verification above, the 27,650 lb weight from Ref. [8] was justified and used.

The material properties for the stainless steel racks used in the analysis are as follows:

Type 304, SA240 18CR-8N (Ref. 18):

Modulus of Elasticity: E= 27.7 x 106 psi Poisson's Ratio: ,= 0.3 (Steel)

Density (Stainless Steel--weight units): = 0.29 lb/in3 6(,w)

The calculated material properties used for the pool concrete from Ref. 22:

Compressive Strength, f, = 4000 psi Modulus of Elasticity, E= 5700047. = 5700044000 psi = 3.60E6 psi Poisson's Ratio /= 0.16 (Concrete)

Density (Concrete--weight units) 5(w) -0.0868 lb/in 3 Fuel weight Wf = 1700 lb (assume the weight is uniformly distributed)

The Single Rack combined structural section properties [Moment of Inertias (I and Iy) and Area, A] for the Region 2 and 3 modules are as follows:

Region 3: 1, = 256,670 in 4 Region 2: 1 = 215,383 in 4 I, = 234,960 in 4 I = 215,383 in 4 A = 373.256 in 2 A = 348.256 in 2 The stiffness for the Gap compression only element at the base is as follows:

Kped= 1.16x 107 lb/in.

4.2 Single Rack to Multi-Rack Model Development The single rack models are combined into the WPMR model and the inter-rack gap stiffness springs are attached. When the gaps are closed the following stiffness in Table 4.1 will be in effect between these interfaces:

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Stevenson & Associates Report 06Q3571.01-01 Table 4.1 Interface Springs Between Interfaces in Multi-Rack Model Impact Spring Type Spring Constant [lb/in]

Rack to Rack (top) 2.82 x10' Rack to Rack (bottom) 1.01 x10 6 Rack to Pool Wall (top) 2.75 xl05 Rack to Pool Wall (bottom) 9.25 xl055 Fuel to Rack (Note 1) 6.2x10 Metamic Insert to Rack (Note 2) 2.0x 105 (weak direction) 3.0x10 5 (strong direction)

Note: 1 - The fuel to rack stiffness is a summation of all the fuel to rack stiffnesses of each fuel assembly in each individual cell.

Note 2 - The Insert to rack stiffnesses are a summation of 120/121 of the insert to rack stiffnesses of an insert assembly in each individual flux trap. The inserts are modeled as two sticks, with the properties of the two sticks modeled effectively 900 to each other.

4.3 Model Details and Description The rack structure dynamic model was prepared by considering nonlinearities and parametric variations. Particulars of modeling details and assumptions for the WPMR analysis of racks are given in the following subsections.

4.4 Modeling Details and Assumptions

a. The model for the rack is supported at the base level, on four (comer) pedestals, modeled using non-linear compression-only gap spring elements and eight linear friction spring elements. These elements are located with respect to the centerline of the rack beam to allow for arbitrary rocking and sliding motions.

b. The fuel rack structure motion is simulated by modeling the rack using 6 degrees-of-freedom at each mass point of the model. This includes the displacements and rotations at each of these points. The response of the module relative to the base is simulated in the dynamic analyses using suitable springs to couple the rack degrees-of-freedom and simulate rack stiffness.

c. Fluid coupling between the rack and fuel assemblies and between the rack and wall is simulated by appropriately modeling the off diagonal mass matrix terms.

Inclusion of these effects uses rack/assembly coupling and rack-to-rack coupling as described in subsection 4.6.

d. Fluid damping and velocity drag due to water particle velocity are not modeled.

These effects are considered implicitly in the fluid coupling and fluid assumption mass modeling described in c. and i.

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Stevenson & Associates Report 06Q3571.01-01

e. Rattling fuel assemblies within the rack are modeled by five lumped masses located at H, 0.75H, 0.5H, 0.25H, and at the rack base (H is the rack height measured above the base-plate). Each lumped fuel mass has two horizontal displacement degrees-of-freedom. Vertical motion of the fuel assembly mass is assumed equal to rack vertical motion at the base-plate level.

f. Seismic motion of a fuel rack is characterized assuming that fuel assemblies in their individual storage location move together in phase. This is the worst case computed dynamic loading on the rack structure for this phenomenon.

g. Potential impacts between the cell walls of the racks and the contained fuel assemblies are accounted for by appropriate compression-only gap elements between the masses involved. The possible incidence of rack-to-wall or rack-to-rack impact is simulated by gap elements at the top and bottom of the rack in two horizontal directions. Bottom gap elements are located at the base-plate elevation.

The initial gaps reflect the presence of base-plate extensions, and the rack stiffnesses are chosen to simulate the local structural detail.

h. Pedestals are modeled using gap elements in the vertical direction and as "rigid links" for transferring horizontal forces. Each pedestal support is linked to the pool liner (or bearing pad) by two friction springs. The spring rate for the friction springs includes any lateral elasticity of the stub pedestals. Local pedestal vertical spring stiffness accounts for floor elasticity and for local rack elasticity just above the pedestal.

i. Rattling of fuel assemblies inside the storage locations causes the gap between fuel assemblies and cell wall to change from a maximum of twice the nominal gap to a theoretical zero gap. Fluid coupling coefficients are based on the nominal gap in order to provide a measure of fluid resistance to gap closure.

j. Sloshing is found to be negligible at the top of the rack and is, therefore, neglected in the analysis of the rack.

4.5 Element Details The dynamic model of a single rack is shown in Figure 4.1. The figure shows many of the characteristics of the model including the fuel to rack gap springs, the rack and fuel bundle elements and the gapped and friction springs at the base that are linked with rigid members.

Page 13 of 50

Stevenson & Associates Report 06Q3571.01-01 Metamlc Insert to Cell Truss/Gap Element VP21: Modeled +/- each horlzontal direction Fuel to Rack Cell Gap/Truss W(Elements (tp)

Note: Modeled +/-each horizontal direction (typ) 9 Fuel Bundles Metamic Insert Elastic Beam Elements Mass Point (typ)

(typ) I Rack Maýs JW* Fuel (typ) Mass (typ) I Metamic Base Insert P P Wl or Baseplate Elastic Beam Gap/Truss Elements

((typ)

I Pedestal--GapfTruss Elements (typical)

/*/ Rack to Pool Bottom Figure 4.1 - Schematic of the Single Rack Dynamic Model Note that the Metamic Inserts were modeled as two sticks, since about half are oriented 90 degrees to the other half.

4.6 Hydrodynamic Coupling Modeling (Single and Multi-Body Coupling)

The hydrodynamic coupling between any two masses is described as "adding" force due to relative motion of the two masses in the X direction. The formulation for this added force is given in Ref. [11] and is summarized using the following mass matrix formulation:

F,I= M,+M 2 +M,, - (MI + MH~X F.-2 -(M, + MH) MR I 'k21 where, adding force acted on Mass 1 Page 14 of 50

Stevenson & Associates Report 06Q3571.01-01 F, - adding force acted on Mass 2 (Mass 2 is assumed contained inside Mass 1)

A - water mass enclosed by Mass 1 M2 - displaced water mass by Mass 2 MH - hydrodynamic mass ji' - absolute acceleration of Mass 1

,'2 - absolute acceleration of Mass 2 Therefore, the mass matrix for adding the hydrodynamic coupling force between any two masses is included in the solution process by adding the water masses MI, M2, and the hydrodynamic mass MH in each direction to the SOLVIA structural model.

As shown in the above formulation, the motion of one body affects the force field on another.

This force field is a function of inter-body gap and can be large when the gaps are small. The lateral motion of a fuel assembly inside a storage location encounters this effect. The rack analysis contains inertial fluid coupling terms, which model the effect of fluid in the gaps between adjacent racks.

Rack-to-rack gap elements have initial gaps set to the entire physical gap between the racks or between outermost racks and the adjacent pool walls.

Using the above formulation, the motion of one body affects the force field on another. This force field is a function of inter-body gap and can be large when the gaps are small. The lateral motion of a fuel assembly inside a storage location encounters this effect. The rack analysis contains inertial fluid coupling terms, which model the effect of fluid in the gaps between adjacent racks.

Rack-to-rack gap elements have initial gaps set to the entire physical gap between the racks or between outermost racks and the adjacent pool walls. Masses including the hydrodynamic mass were calculated by setting the Kinetic energy of hydrodynamic mass to be equal to kinetic energy of the fluid flow and maintaining continuity between the body and fluid flow area and combining the mass for all of the cells.

4.7 Stiffness Element There are three element types used in the SOLVIA rack module models. The first element type is linear elastic beam elements used to represent the beam-like behavior of the integrated rack cell matrix. The second element type is the linear friction springs used to develop the forces between the rack pedestals and the supporting floor. The third element type is non-linear gap elements, which model gap closures and impact loadings between fuel assemblies and the storage cell inner walls and racks.

The gap elements modeling impacts between fuel assemblies and racks have local stiffness Ki.

Support pedestal spring rates K, are modeled by gap elements. Local behavior of the pedestal on the concrete floor is included in K,. The type 2 friction elements are included as Kf. The beam elements for the rack and fuel model the combined stiffness of these components for the racks.

Page 15 of 50

Stevenson & Associates Report 06Q3571.01-01 Friction at the support to pool floor interface is modeled by the linear friction springs with stiffness Kf up to the limiting lateral load pN, where N is the current compression load at the interface between support and liner. At every time-step during time history analysis, the current value of N (either zero, if the pedestal has lifted off the floor, or a restraining force) is computed.

The modeling of the effective compression stiffness with the gap element of stiffness K, includes the pedestal stiffness and local stiffness of the underlying pool slab.

4.8 Poison Insert Modeling The metamic poison inserts consist of two nominal 7" wide x 155" long by 0.1" thick metamic panels per flux trap. Each metamic panel is enclosed in a formed 26 gage sheet metal channel.

The channels are nominally 7.12" deep with .43" flanges x 155" long. The metamic panels are held in the channels by a 4" long x 26 gage sheet metal channel-shaped band at the top and bottom, and four 6" long x 26 gage sheet metal channel-shaped bands between the 4" bands, spaced at about 30" to 32" center to center. These bands are spot welded to the outside channel "flanges" with two spot welds on each side. The inserts are detailed on References 14 and 15.

The design of the metamic inserts has evolved over time to the present configuration where the sheet metal channels with the metamic panels are now held (front to front) at a nominal 1.20" apart, measured out to out between the backs of the channel sections. They are held by ten to fourteen pairs of 0.075" thick x 0.95" wide x 4" long plates, spaced 7" to 20" welded to the channel flanges along the length, plus a pair of 0.075" thick x 0.90" wide x 4" long plates welded to the channel flanges at each end.

There are four 22 Gage sheet metal formed bands described as horizontal supports, between the two insert halves along with four pairs of corresponding 22 Gage sheet metal formed bands described as vertical supports. The original purpose of these "supports" was to hold the opposing channel/panel sections together during shipping and insertion into the flux traps. Now that the design is such that the panel/channel sections are fixed by the plates on the flanges, these "supports" no longer have any function. They are classified as non-safety related and provide no structural support, and assumed no interference. The total mass/weight of these supports is very small (z0. 1 lb each). In the rack models, the inserts are modeled with more than double their actual weight. Hence, the horizontal and vertical supports are not considered explicitly.

There are also three sets of wedge blocks made from Y2" plate, with one set at the top, one at the bottom, and the other located about 2/5 the distance from the top, between the other two. These wedge blocks were originally designed to hold the insert halves apart and against the inside walls of the flux traps after insertion into the flux traps. Since the panels with their wrapper channels are now fixed by the added plates at a nominal 1.20" outside width and the opposing panel assemblies are now offset vertically 113/16", the wedge blocks likely are not in contact with each other, and hence cannot be counted on to provide any structural support. They are included for mass/weight effects only in the insert model, with mass distribution varied proportionately to account for their locations in the rack models.

To determine equivalent structural properties for the metamic insert panels for inclusion in the rack models, a model of a typical metamic panel was developed using shell elements.

Page 16 of 50

Stevenson & Associates Report 06Q3571.01-01 4.9 Friction Modeling Between Rack Supports and Pool Floor As discussed in 3.3 simulations are performed with friction coefficients of 0.2, 0.5 and 0.8 in order to bound the range of realistic results for the earthquake event.

5.0 Load Combinations and Load Development 5.1 Loads and Load Combinations The applicable loads and load combinations to be considered in the seismic analysis of rack modules are taken from the OT Position [2] and are included in Table 5.1 below: The acceptance criteria is defined in Subsection NF of the ASME Code [16].

[ Table 5.1 - Load Combinations for the SFP Rack Analysis Loading Combination(t) Acceptance Limit D +L Normal Limits ofNF3231.1a(2), Ref. [16]

D+L+E D+L + To D + L + To + E Lesser of 2 Sy or Su - Stress Range(')

D +L + Ta+E D + L + Ta + E' Faulted Condition Limits of NF 323 1.1c(3 ), Ref. [16]

Notes:

1) The thermal loadings have been addressed in detail in the Ref. [9] calculation and shown to not control the structural evaluations of the racks. Since the normal acceptance limits for the load condition D + L + E are less than the stress range limits, the loading combinations with the stress range limits including thermal loads generally are not controlling. Thermal loads on the racks were therefore not reanalyzed. For completeness, for this analysis, the thermal stresses obtained from Ref. [9] are used for the rack components. For the inserts, a thermal analysis of the insert was performed in this analysis.

2) The design basis is ASME Subsection NF, 1980 through the Winter 1981 Addendum. The rack is evaluated to these code requirements. The new inserts are evaluated to ASME 1998 NF requirements. It is noted that there is no significant difference between these code versions for this application.

3) Faulted conditions in the ASME code are defined as Service Level D condition

[16]. NF3231.1c ultimately references Appendix F for this evaluation.

Where:

D = Dead weight-induced loads (including fuel assembly and poison insert weights)

Page 17 of 50

Stevenson & Associates Report 06Q3571.01-01 L = Live Load (not applicable for the fuel rack, since there are no moving objects in the rack load path)

E = Operating Basis Earthquake (OBE), including the effects of impacts occurring during the earthquake event.

Ef = Safe Shutdown Earthquake (DBE), including the effects of impacts occurring during the earthquake event.

To = Differential temperature induced loads (normal operating or shutdown condition based on the most critical transient or steady state condition)

Ta = Differential temperature induced loads (the highest temperature associated with the postulated abnormal design conditions).

As discussed in Note 1 above, temperature loadings have been evaluated previously and were not explicitly evaluated, except for the Metamic inserts.

The basic generally governing load combination evaluated is as follows:

D+L+ E (Acceptance Limit Normal Limits of NF3231.1a, Ref. [16])

For ease of analysis, the elastic modulus at 1500 F was used for both the OBE and DBE dynamic analyses, which results in the best estimate global forces and displacements. The allowable stresses calculated in Sections 7.2 and 7.3 use the yield and ultimate strength properties at 1500 F and 2500 F respectively for the OBE and DBE. As discussed in the OT Position [2], "for impact loading the ductility ratios utilized to absorb kinetic energy in the tensile, flexural, compressive, and shearing modes should be quantified." Maximum impact loads and therefore maximum ductility ratios will be derived from the DBE event, also ductility ratios are applicable only for faulted condition limits. Therefore, impact loading was only evaluated for the DBE load case.

In addition the impact acceptance criteria includes a provision that insures that the consequent impact loads on the fuel assembly does not lead to damage of the fuel in accordance with the OT Position [2].

5.2 Synthetic Earthquake Time Histories OBE and DBE The synthetic time-histories in three orthogonal directions (N-S, E-W, and vertical) are generated in accordance with the provisions of SRP [1], Section 3.7.1. In order to prepare an acceptable set of acceleration time-histories, Stevenson and Associates' commercial code THSPEC [17] is used. It is noted that program THSPEC is a derivative of Program SIMQKE, developed at MIT.

The response spectrum and the power spectral density (PSD) corresponding to the generated acceleration time-history is to envelope their target (design basis) spectrum and PSD with only finite enveloping infractions. The target floor response spectra were developed by interpolating the 2% damped horizontal OBE spectra between 354' and 372' to obtain a spectrum at 362'. The vertical OBE spectra at all elevations in the building were used for the vertical response spectra target. It is noted that time history acceleration is independent of damping level, however, due to smoothing and enveloping when developing design spectra, the time history may not envelop all Page 18 of 50

Stevenson & Associates Report 06Q3571.01-01 response spectra at a given location developed with different damping coefficients. It is reasonable to use a 2% damped target since this is the damping used in the analysis of welded steel structures. The DBE horizontal design time histories were developed by simply multiplying the OBE horizontal time histories by 1.8 and the OBE vertical time history by 2.0 in accordance with APL-C-502 [4]. The time-histories used for the rack analyses were generated to satisfy the enveloping criterion for the synthetic time-histories in Section 3.7.1 of the SRP [1].

The seismic files also satisfy the requirements of statistical independence required by SRP 3.7.1

[1]. The absolute value of correlation function of the three time-histories relative to one another were calculated to be 0.176, 0.167 and 0.142 respectively, which are less than 0.30 (the statistical independence criterion) indicating that the three data sets are statistically independent.

Plots showing the comparison of the Response Spectra generated by each of the artificial time histories to the target floor response spectra are provided as follows:

ComparisonBetween Generatedand Target RS Elev 362' MrkzonaI - I Damping rmfo.O.O2 rIO -I 7

C S

4 I S 0.1 1 to00 R'squcyjPWz ThSpe 1.,2 Sltvsnam A Assocites, Romanb Figure 5.1 - Comparison of Generated Response Spectrum (solid line) to Target Floor Response Spectrum (Dashed Line), Horizontal, Direction 1 Page 19 of 50

Stevenson & Associates Report 06Q3571.01-01 ComparisonBetween Generatedand Target RS IontaI 2 D2wnpng Rauo-0.02 x0 I 2

/.0 0

0.1 to too ft.qu.ncyMit) e 4* At..n.,.

A..ne/ab. R.,a.,L*

Figure 5.2 - Comparison of Generated Response Spectrum (solid line) to Target Floor Response Spectrum (Dashed Line), Horizontal, Direction 2 Comparison Between GenerateG and Target RS Vewil, DampingRNVi..2 x 16.2it 12 to I

4 10 j

100 Frqu-crJ YhSný 1.2 Bwmmon AAssodaces, RomanA Figure 5.3 - Comparison of Generated Response Spectrum (solid line) to Target Floor Response Spectrum (Dashed Line), Vertical Direction Page 20 of 50

Stevenson & Associates Report 06Q3571.01-01 TIME-HISTORY STATISTICAL CORRELATION RESULTS OBE and DBE X direction to Y direction .176 X direction to Z direction .167 Y direction to Z direction .142 5.3 Impact Load Consideration and Combination with other Loads The impact loading effect on the global rack assemblies is implicitly included by the modeling and dynamic simulations. As described in the modeling, impacts are considered as the gap elements open and close during the analysis.

6.0 Summary of Analyses Performed 6.1 Single Rack Analysis As previously discussed in Section 4.1, single rack models were developed for each module type in order to use them as building blocks for the WPMR analysis. In addition the single rack models are employed to study the effect of top loading the rack with miscellaneous equipment.

The top loaded rack simulation is performed using the 0.8 coefficient of friction, DBE load case to produce the maximum overturning moment. A 2,000 lbf mass, with three translational degrees-of-freedom, is rigidly attached to the rack 24" above the top of the cell structure. The analysis results, with and without the weight, are studied. It is noted that the results indicate that the additional mass has an insignificant effect on the rack module analysis results.

6.2 Whole Pool Multi-Rack (WPMR) Analysis The multiple rack models use the fluid coupling effects for all racks in the pool. The eight racks are modeled with proper interface fluid gaps and a coefficient of friction at the support interface locations as described in Section 4.3. The response to both DBE and OBE seismic excitation is determined.

6.3 Parametric Simulations 6.3.1 Friction Coefficient Variation The WPMR simulations listed in Table 6.1 have been performed to investigate the structural integrity of the racks, including the new poison inserts.

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Stevenson & Associates Report 06Q3571.01-01 Table 6.1 LIST OF WPMR AND SINGLE RACK SIMULATIONS Case Model Load Case COF Event 1 WPMR All racks fully loaded 0.5 OBE 2 WPMR All racks fully loaded 0.2 OBE 3 WPMR All racks fully loaded 0.8 OBE 4 WPMR All racks fully loaded 0.5 DBE 5 WPMR All racks fully loaded 0.2 DBE 6 WPMR All racks fully loaded 0.8 DBE 7 WPMR Racks 50% Full(') 0.5 DBE COF = Coefficient of Friction Note 1: The 50% full simulation was performed to determine whether there was a possibility that the racks could exhibit greater displacement when all the cells within the rack are not in use.

Note 2: No numerical convergence or instability problems were encountered in any of the analyses.

7.0 Acceptance Criteria Development 7.1 Displacement and Rocking Acceptance Criteria According to Section 3.8.5 of Ref. [1], the minimum required safety margins against overturning under the OBE and DBE events are 1.5 and 1.1 respectively. The maximum rotations of the rack (about the two principal axes) are obtained from a post processing of the rack time-history response output. The margin of safety against overturning is given by the ratio of the rotation required to produce incipient tipping in either principal plane to the actual maximum rotation in that plane predicted by the time-history solution.

0 requiredfor overturning Marginof SafetyQ*

0 predicted All ratios for the OBE and DBE events should be greater than 1.5 and 1.1 respectively, to satisfy the regulatory acceptance criteria.

The 9 required for overturning is calculated as follows:

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Stevenson & Associates Report 06Q3571.01-01 The height of the rack is 165" with the center of gravity of the rack to be about the center. The width of the rack between the outside feet is 85.2". The center of the rack has to therefore rock over half the distance between the feet. This angle is defined as:

e = tan -' 42.6" (half the distance between feet)/82.5" (distance to the Center of Gravity) e = 27.30 7.2 Stress Evaluations - OBE Load Case The stress limits presented apply to the rack structure and are derived from the ASME Code,Section III, Subsection NF [16]. Parameters and terminology are in accordance with the ASME Code. Material properties are obtained from the ASME Code Appendices and are listed in Table 2.2.

7.2.1 Tension Allowable Stress - OBE Allowable stress in tension on a net section is:

F, = 0.6Sy F, = 0.6

27,500 psi = 16,500 psi Where Sy = yield stress at temperature, and F, is equivalent to primary membrane stress.

7.2.2 Compression Allowable Stress - OBE Allowable stress in compression on a net section is:

F. = S{.47 4-r F = 27,500(. 47- 44rj where kl/r for the main rack body is based on the full height and cross section of the honeycomb region and does not exceed 120 for all sections.

I = unsupported length of component k = length coefficient which gives influence of boundary conditions.

r = radius of gyration of component This is applicable to the rack as a whole.

For local buckling considerations of the cell walls, the critical buckling stress is given by:

Fcr E Page 23 of 50

Stevenson & Associates Report 06Q3571.01-01 where F,, = critical buckling stress k = buckling stress coefficient (= 4.0 for simply supported unloaded edges)

E = initial modulus of elasticity

p. = poisson's ratio t = plate thickness b = effective width For the metamic inserts, due to their construction and inherent eccentricities, the effective buckling load or axial load capacity is determined by a linearized buckling analysis.

7.2.3 Shear Allowable Stress - OBE Allowable stress in shear on a net section is:

F, = 0.4 Sy F, = 0.4

27,500 psi = 11,000psi 7.2.4 Bending Allowable Stress - OBE Maximum allowable bending stress at the outermost fiber of a net section, due to flexure about one plane of symmetry is:

Fb = 0.60 Sy (equivalent to primary bending)

Fb = 0.6 * (27,500 psi) = 16,500 psi 7.2.5 Combined Bending and Tension or Compression Allowable Stress - OBE Combined bending and compression on a net section satisfies:

f. + .A <1 F. DbFxC DFfby where:

f Direct compressive stress in the section b.,x = Maximum bending stress along x-axis fby = Maximum bending stress along y-axis C,,= = 0.85 Cmy = 0.85 D. = I(faF'e.)

Dy 1- ,F'ey)

I=-

F',y 2 (R2 E)/(2.15 (kl/r) ,Y)

E = Young's Modulus Page 24 of 50

Stevenson & Associates Report 06Q3571.01-01 and subscripts x and y reflect the particular bending plane.

Combined flexure and compression (or tension) on a net section:

f.a f. xLby

+j <j1.

0.6S b. Fby The above requirements are to be met for both direct tension and compression.

7.2.6 Bearing Allowable Stress - OBE Allowable Bearing Stress from Section NF-3226.1 of the ASME Code [16]:

Fb = Sy = 27,500 psi 7.2.7 Weld Allowable Stress or Force (By Analysis and Test) - OBE Allowable maximum shear stress on the net section of a weld is given by:

F, = 0.3 S, (on the weld material) or F, = 0.4 Sy (on the base metal material in shear)

& = 0.6 Sy (on the base metal material in tension) where S, is the weld material ultimate strength at temperature and Sy is the base metal yield strength at temperature. Per Ref. [9] the weld material used is an E80 electrode with an Su = 80 ksi. For fillet weld legs in contact with base metal, the shear stress on the gross section is limited to 0.4Sy, where Sy is the base material yield strength at temperature.

Therefore the allowable weld stress is:

& = 0.3 S, = .3

80 ksi = 24 ksi (on the weld material)

F, = 0.4 Sy = 0.4

27,500 psi = 11,000 psi (on the base metal material in shear)

F, = 0.6 Sy = 0.6

27,500 psi = 16,500 psi (on the base metal material in tension)

The spot weld allowables were determined by test and were taken as:

F = TL. * (S/ S)

Where: TL. is the mean ultimate capacity test results. From Ref. [19] the mean of the 15 test samples for the spot weld = 680 lb (rounded to the nearest 10 lb)

S = ASME Code Allowable Stress S = 17.2 ksi from Ref. [18] (Note that failure of the test was a base metal failure, therefore, S, is the base metal allowable stress.)

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Stevenson & Associates Report 06Q3571.01-01 S. = Ultimate Strength from the ASME Code = 68.5 ksi (@2500 F) from Ref.

[16]. (Note that failure of the test was a base metal failure, therefore, S., is the base metal ultimate stress.)

Therefore the allowable spot weld load for the OBE case is:

F = 680 lb * (17.2 / 68.5) = 170 lb 7.3 Stress Evaluations - DBE Load Case Section F-1334 (ASME Section III, Appendix F [16]) states that limits for the Level D condition (the stress limits that are applicable to faulted conditions) are the smaller of 2 or 1.167S./Sy times the corresponding limits for the Level A condition. Examination of material properties for Type 304 stainless steel demonstrates that two times the Level A allowable stress from Service Limit A controls.

7.3.1 Tension Allowable Stress - DBE Allowable stress in tension on a net section is:

Ft = 2.0

0.6

23,750 psi = 28,500 psi 7.3.2 Compression Allowable Stress - DBE Axial Compression Loads are limited to 2/3 of the calculated buckling load or no greater than the allowable tension load:

F, = 0.667

F, < 28,500 psi Where: Fe is the Euler Buckling Load For local buckling considerations of the cell walls and buckling of the metamic inserts, acceptance criteria is taken as the same for OBE above.

7.3.3 Shear Allowable Stress - DBE Stresses in shear shall not exceed the lesser of 0.72S. or 0.42S.. In the case of the Austenitic Stainless material used here, 0.72Sy governs.

Allowable stress in shear on a net section is:

F, = 0.72

23,750 psi = 17,100psi 7.3.4 Bending Allowable Stress - DBE Maximum allowable bending stress at the outermost fiber of a net section due to flexure about one plane of symmetry is:

Fb = 2.0

0.6 * (23,750 psi) = 28,500 psi Page 26 of 50

Stevenson & Associates Report 06Q3571.01-01 7.3.5 Combined Bending and Tension or Compression Allowable Stress - DBE Combined bending and compression on a net section satisfies:

f. + C + C.fby <1 4fu 0.667*Fe DxF, DYFbY Where all of the terms have been defined in Subsections 7.2.5 and 7.3.2.

Combined flexure and compression (or tension) on a net section:

f. + y <1.0 0.667*Fe FbX Fby Where 0.667

Fe is limited to the tension allowable of 28,500 psi. The above requirements are to be met for both direct tension and compression.

7.3.6 Bearing Allowable Stress - DBE Per Section F-1334.10 [16], Bearing Stress need not be evaluated for loads when Limit D Service Limits are specified.

7.3.7 Weld Allowable Stress and Force (By Test) - DBE For welds, the allowable maximum weld stress is not specified in Appendix F of the ASME Code. An appropriate limit for weld throat stress is conservatively set here as:

F, = 0.3 S. x factor (on the weld material)

& = 0.4 Sy x factor (on the base metal in shear)

& = 0.6 Sy x factor (on the base metal in tension) where: factor = (Level D shear stress limit) / (Level A shear stress limit) =

17,100/11,000 = 1.55 and S,, and Sy were defined in Section 7.2.7 Therefore the allowable weld stress is:

F, = 0.3 S,, x factor - 1.55*0.3

80 ksi = 37.2 ksi (weld material)

& = 0.4 Sy x factor = 1.55

0.4

23,750 psi = 14,725 psi (base metal in shear)

& = 0.6 Sy x factor = 1.55 *0.6

23,750 psi = 22,088 psi (base metal in tension)

The spot weld allowables were determined by test from F-1332.7 of Appendix F [16] and were taken as:

F&=0.7 *TL. * (S,,/ S,*)

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Stevenson & Associates Report 06Q3571.01-01 Where: T.L. and S, were defined previously in Section 7.2.7 and S.* is the ultimate strength at the testing temperature. Therefore, the ratio S. /S,* is essentially 1.0. Using the Ref. [19] mean test results of 680 lbs the allowable spot weld load for the DBE case is:

F, = 680 lbs

0.7 = 475 lbs 7.3.8 Impact Acceptance Criteria - DBE Impact allowable stress will be calculated in accordance with Appendix F of the ASME Code

[16], Section F1341.2 for Plastic Analysis.

In accordance with Section F-1341.2 the general Primary Membrane Stress; Pm <0. 7S,, = 0.7

68,500 psi = 47,950 psi the maximum Primary Stress (including bending from the impact);

Pm < 0.9S. = 0.9

68,500 psi = 61,650 psi 8.0 Analysis Results and Comparison to Acceptance Criteria 8.1 Time-History Simulation Results The results from the analyses are contained in the raw data output files. However, due to the huge quantity of output data, a post-processor is used to scan for worst case conditions. Further reduction in this bulk of information is provided in this section by extracting the worst case values from the parameters of interest; namely displacements, support pedestal forces, impact loads, and stress factors. Table 8.1 and 8.2 below summarize the overall global response of the various Single Rack and WPMR DBE and OBE Analyses respectively.

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Stevenson & Associates Report 06Q3571.01-01 Table 8.1 Result Summary - Load Case = D + L + DBE Rack Global Single Rack Analysis Full Pool Analysis Forces F.Coeff=0.2 F.Coeff=0.5 F.Coeff=0.8 F.Coeff=O.2 F.Coeff=0.5 F.Coeff=0.8 Base Sh* (lb) 1.36x10 5 1.42x10 5 1.89x10 5 6.682x10 4 1.681x10s 1.90x105 Base Shy (ib) 6.86x10 4 1.29x105 1.97x10s 7.393x10 4 1.633x105 1.98x105 Axial F (lb) -1.75x104 -1.74x10 4 -1.74x10 4 -1.75x10 4 -1.74x104 -1.74x10 4 Base M,, 8.27x10 6 1.532x10 7 2.48x10 7 8.01x10 6 1.654x10 7 2.28x10 7 (lb-in) 1.15x10 7 1.738x10 7 2.41x10 7 6.75x10 6 1.615x10 7 2.27x10 7 Base M,,

(lb-in)

Base D. (in.) 2.87 1.303 0.522 1.537 0.258 0.0986 Base Dy (in.) 4.66 1.411 0.524 3.305 0.733 0.199 Top D. (in.) 2.89 1.328 0.552 1.541 0.271 0.126 Top Dy (in.) 4.67 1.422 0.556 3.301 0.752 0.237 Base Acc1 175.18 177.50 196.99 254.58 486.6 206.16 (in./S2)

Base Accy 129.33 149.56 177.42 170.52 314.6 418.59 (in./S2)

Top Acc1 178.83 237.95 292.68 354.86 466.95 543.98 (in ./S2)

_ 268.65 155.81 291.9 315.34 Top Accy 141.50 181.05 2

(in./s )

Max. Fuel 8.81x10 4 9.23x10 4 1.001x10 5 1.072x10 5 1.63x105 2.167x10I Impact (lb)

Max. Insert 7.25x10 3 8.27x10 3 1.736x10 4 5.23x10 3 9.8x10 3 1.718x10 4 Impact Foot-I (lb) -9.57x10 4 -1.456x10 5 -2.019xi0s -1.0x10s -1.42x10 5 -1.92x10 5 Foot-2 (lb) -9.21x10 4 -1.493x10 5 -1.988x10s -9.345x10 4 -1.36x10 5 -1.76x10s Foot-3 (lb) -1.02x10s -1.61 1x10s -1.999x10 5

-9.258x10 4

-1.34x10 5

-1.70x10 5 Foot-4 (lb) -1.12x10 5 -1.53x10 5 -2.043x10 5 -9.691x10 4 -1.40xlO 5 -1.85x10 5 Notes: (1) Results are for the 12xI 1 rack (group 3)

(2) Foot-1, Foot-2, Foot-3, and Foot-4 are the pedestal forces (Note, models include 4 feet per rack, actual racks have 14 pedestals.) and include deadweight, vertical seismic, and impact effects.

(3) Maximum Fuel Impact and Maximum Insert Impact are the maximum impact force between the fuel assemblies and the cell walls and between the metamic inserts and cell walls respectively.

Page 29 of 50

Stevenson & Associates Report 06Q3571.01-01 Table 8.2 Result Summary - Load Case = D + L + OBE Rack Global Single Rack Analysis Full Pool Analysis Forces F.Coeff=0.2 F.Coeff=0.5 F.Coeff=0.8 F.Coeff=0.2 F.Coeff=0.5 F.Coeff=0.8 Base Sh, (lb) 6.07x10 4 1.21x10 5 1.733x10 5 6.08x10 4 1.04x105 1.30x10 5 Base Shy (lb) 5.75x10 4 1.15x10 5 1.742x10 5 7.60x10 4 1.20x10 5 1.77x10 5 Axial F (lb) -1.638x104 -1.637x10 4

-1.637x10 4

-1.637x10 4 -1.637x10 4 -1.637x10 4 Base M 7.029x10W 1.44x10 7 2.07x10 7 6.86x10 6 1.42x10 7 2.01x10 7 (lb-in)

Base M)Y 7.20x10 6 1.55x10 7 2.00x10 7 6.72x 106 1.27x10 7 1.45x10 7 (lb-in)

Base D, (in.) 1.162 0.175 0.074 0.278 0.0205 0.0026 Base Dy (in.) 1.511 0.199 0.049 0.774 0.0771 0.0172 Top D. (in.) 1.166 0.185 0.102 0.281 0.0376 0.0323 Top Dy (in.) 1.518 0.222 0.079 0.780 0.1029 0.0565 Base Acc, 119.66 108.83 137.60 182.89 86.87 28.14 2

(in.s )

Base Accy 109.01 97.13 130.96 154.61 87.37 90.47 (in./s 2 )

Top Accx 88.71 183.89 249.80 181.65 320.78 262.25 (inJs2)

Top Acc, 85.18 148.23 181.42 118.12 246.75 211.05 (in.Is 2 )

Max. Fuel 7.52x10 4 1.231x10 5 1.287x10 3 9.41x10 4 1.205x10 5 1.145x10I Impact (ib)

Max. Insert 3.68x10 3 7.15x10 3 1.109x10 4 5.199x10 3 6.43x103 8.35x10 3 Impact Foot-I (lb) -9.17x10 4 -1.43x10 5 -1.96x10 5 -8.71x10 4 -1.47x10 5 -1.54x10 5 Foot-2 (lb) -9.44x10 4 -1.45x10 5 -1.83x10 5 -9.21x10 4 -1.36x10 5 -1.58x10 5 Foot-3 (lb) -9.09x10 4 -1.31x10 5 -2.05x10 5 -8.99x10 4 -1.27x10 5 -1.67x10 5 Foot-4 (lb) -8.68x10 4 -1.41x10 5 -1.91x10 5 -8.85x10 4 -1.28x10 5 -1.45x10 5 Notes: (1) Results are for the 12xl I rack (group 3)

(2) Foot-i, Foot-2, Foot-3, and Foot-4 are the pedestal forces (Note, models include 4 feet per rack, actual racks have 14 pedestals.) and include deadweight, vertical seismic, and impact effects.

(3) Maximum Fuel Impact and Maximum Insert Impact are the maximum impact force between the fuel assemblies and the cell walls and between the metamic inserts and cell walls respectively.

Output results from SOLVIA for the controlling case for maximum fuel-cell impact are shown in Figures 8.1 and 8.2 for the single rack model and Figure 8.3 and 8.4 for the WPMR model.

The Subsections that follow summarize additional analyses performed to develop and evaluate structural member stresses, which are not determined by the post processor.

Page 30 of 50

Single Rack Model (11x12 cells) DBE - Friction coefF=0.8 ORIGINAL F- - 100. Z ORIGINAL -- -- 1 100.

MAX DISPL. F-1 0.87714 SY MAX DISPL. i- 0.44153 TIME 3 X TIME 3.2 CD ZONE ZI ZONE ZI CD 01 0 C

C)

It CD ra CD ha C

0 ORIGINAL -- -4 100. Z ORIGINAL H-- -- 100. 0%

MAX DISPL. i-i 0.70794 SY MAX DISPL. --- 0.41859 ~0 CD TIME 3.4 X TIME 3.6 -J Co ZONE Zi ZONE Zi -A

-A 0

-A 0*

CD II SOLVIA-POST 03 STEVENSON & ASSOCIATES, CLEVELAND, OHIO SOLVIA-POST 03 STEVENSON & ASSOCIATES, CLEVELAND, OHIO

Single Rack Model (11x12 cells) DBE - Friction coefF=0.8 LU-0 90 tLL 0

0- 0 -. 4 a

N CL LU 0 0 Ln 0

0 1w 03 I-0 0 5 10 Is 0 TIME TIME CD ON I

Lu 0* 0 0 UU ILL I-A

.q-i co 0 0

LU C)

(D II LUT LUJ 0 5 10 15 00 TIME TIME SOLVIA-POST 03 STEVENSON & ASSOCIATES, CLEVELAND, OHIO

Full Pool Model (8 racks) DBE - Friction coeff=0.8 ORIGINAL -- -- 200. ORIGINAL -- -- 200. Z MAX DISPL. ---4.2413 MAX DISPL. F-4 5.7561 SY TIME 3.8 TIME 4 X 00 ZONE ZI ZONE ZI rA 0

~m.

U)

U) 0 0

U2 IV U) 0 CD ORIGINAL -- -- 200. ORIGINAL -- -1 200. Z C MAX DISPL. i-- 6.0762 MAX DISPL. -i 4.8586 SY TIME 4.2 TIME 4.4 X ZONE ZI ZONE Zi -J C

C 0

0.

0 II P

00 SOLVIA-POST 03 STEVENSON & ASSOCIATES, CLEVELAND, OHIO SOLVIA-POST 03 STEVENSON & ASSOCIATES. CLEVELAND. OHIO

Full Pool Model (8 racks) DBE - Friction coeff=0.8 (Y)

CD

-. 4 LU CD CL (A LI) C LU_

LUJ V_ ;~b.

(A (A

(D Ld C CD 0 5 10 Is 0 5 10 15 CD (A

CD TIME TIME CD 0

C 0

0F 7IV' , LU ~0 CA LU 0T 0

L. 0 (z

LU_

a-

-.9 0 LU 0

N 0 5 10 15 0 5 10 is TIME TIME SOLVIA-POST 03 STEVENSON & ASSOCIATES, CLEVELAND, OHIO

Single Rack Model (11x12 cells) DBE - Friction coeff=0.2 ORIGINAL i-- -1 100. Z ORIGINAL -- -i 100.

MAX DISPL. i-ý 4.0923 Y MAX DISPL. -- 4 4.7676 90 LA TIME 3 X ZONE 3.2 TIME ZI ZONE ZI CD CD CD CD 0A w ORIGINAL i-- --A 100. Z ORIGINAL -- -- 100.

MAX DISPL. i- 4.7966 Y MAX DISPL. i-- 4.7076 X TIME TIME 3.4 ZONE 3.6 ZI C6 ZONE Zi 0

SOLVIA-POST 03 STEVENSON & ASSOCIATES, CLEVELAND, OHIO SOLVIA-POST 03 STEVENSON & ASSOCIATES, CLEVELAND. OHIO

Full Pool Model (8 racks) DBE - Friction coeff=0.2

'~21 ORIGINAL i-- -- 200. Z ORIGINAL i-- ---1 200.

MAX DISPL. F--- 2.9057 SY MAX DISPL. i-- 3.4558 TIME 3 X TIME 3.2 00 ZONE ZI ZONE Zi w

pip rJ2 a

a C CN 0

ORIGINAL H-- -- 200. z ORIGINAL -- --- 200.

MAX DISPL. i-- 3.3819 Y MAX DISPL. - 3.4614 TIME 3.4 X TIME ZONE 3.6 ZI 0" ZONE Zi 0

a II 0

SOLVIA-POST 03 STEVENSON & ASSOCIATES, CLEVELAND, OHIO SOLVIA-POST 03 STEVENSON & ASSOCIATES, CLEVELAND, OHIO

Stevenson & Associates Report 06Q3571.01-01 8.2 Maximum Rack Displacements and Rocking The maximum rack displacements are obtained from the time-histories of the motion of the upper and lower four comers of each rack in each of the simulations. The maximum displacements in either direction reported from the WPMR analyses are 3.3" for the DBE event and 0.77" for the OBE event. The maximum displacements in either direction reported from the single rack analysis are 4.7". The rack height is 165" from the floor to the top of the rack.

Making the conservative assumption that the displacement of the rack at the base is 0" (somewhat unrealistically conservative since the majority of this displacement is from sliding) results in the following:

The rocking angle of the rack for the OBE displacement is:

e (predicted)= sin-' (0.78" / 165") = 0.270 Marginof Safety= 0 requiredfor overturning = 27.30 (calculated in Subsection 0 predicted 6.1)/0.270 = 116.3 >> 1.5 (from Subsection 6.1) OK The rocking angle of the rack for the DBE small displacement is:

e (predicted)= sin-' 4.7"/ 165" = 1.630 Marginof Safety = 0 requiredfor overturning = 27.30 (calculated in Subsection 0 predicted 6.1)/ 1.600 = 19.05 >> 1.1 (from Subsection 7.1) OK 8.3 Pedestal Evaluation 8.3.1 Maximum Pedestal Vertical Forces The maximum individual vertical pedestal force obtained in the Single Rack or WPMR simulations was 41,000 lb for the OBE Condition. Total combined downward load on all pedestals for OBE at the time of this maximum is 300,637 lb.

The maximum vertical pedestal force obtained in the Single Rack or WPMR simulations was 40,860 lb for the DBE Condition. Total combined downward load on all pedestals for DBE at the time of this maximum is 297,705 lb.

8.3.2 Maximum Pedestal Horizontal Forces (From Friction)

The maximum interface shear force value bounding all pedestals in the WPMR simulations for the OBE in the X direction is 12,379 lb and in the Y direction 12,643 lb. The maximum interface shear force value bounding all pedestals in the WPMR simulations for the DBE in the X direction is 13,571 lb and in the Y direction 14,143 lb.

Page 37 of 50

Stevenson & Associates Report 06Q3571.01-01 8.3.3 Pedestal and Pedestal Connection Structural Evaluation The time-history results from the analyses provide the pedestal normal and lateral interface forces, which may be converted to the limiting bending moment and shear force at the bottom baseplate-pedestal interface. Maximum values are determined for every pedestal in the array of racks. With this information available, the structural integrity of the pedestal was assessed. The net section maximum bending moments and shear forces can also be determined at the bottom baseplate-rack cellular structure interface for each spent fuel rack in the pool. Using these forces and moments, the maximum stress on the worst case pedestal was calculated.

For the OBE condition the maximum stress for axial and bending loads calculated for the pedestal supports was:

9,323 psi < 16,500 psi allowable (Section 7.2.1, 7.2.4 and 7.2.5) OK Note that for this interaction, the interaction equation f- + f-' + AL < 1.0 controls the 0.6Sf Fb, FbY combined bending and axial load interaction and the 0.6 Sy = Fb = 16,500 psi For the DBE condition the maximum stress for axial and bending loads calculated for the pedestal supports was:

9,847 psi < 28,500 psi allowable (Section 7.3.1, 7.3.4 and 7.3.5) OK Note that for this interaction, the interaction equation f. +A + 1.0 controls since 0.667*F, Fb Fb

0. 667

Fe is limited to 28,500 psi = 2.0

Fb.

For the OBE condition the maximum bearing stress calculated for the pedestal supports was:

8,274 psi < 27,500 psi allowable (Section 7.2.6) OK The DBE condition does not have to be evaluated for bearing as discussed in Section 7.3.

For the OBE condition the maximum shear stress in the pedestal support threads was:

8129 psi < 11,000 psi allowable (Section 7.2.3) OK For the DBE condition the maximum shear stress in the pedestal support threads was:

8550 psi < 17,100 psi allowable (Section 7.3.3) OK 8.4 Rack Structural Evaluation 8.4.1 Rack Member Evaluations The time-history results from the analyses provide the maximum internal section forces and moments, which may be converted to the limiting stresses within the rack. The limiting maximum combined rack stress interaction coefficient for axial and bending stresses for the OBE Page 38 of 50

Stevenson & Associates Report 06Q3571.01-01

= 0.62 < 1.0 allowable and for the DBE = 0.41 < 1.0 allowable. These evaluations include the worst case rack members in the rack.

8.4.2 Rack Connection Evaluations Weld locations subjected to significant seismic loading are at the bottom of the rack at the baseplate-to-cell connection, at the top of the pedestal support at the baseplate connection, and at cell-to-cell connections. Bounding values of resultant loads are used to qualify the connections.

a. Baseplate-to-Rack Cell Welds Weld stresses are produced through the analysis of the rack cell welds for the maximum loads on the sections. In the case of the baseplate to the rack cell the base metal section controlled the evaluation.

The highest predicted cell to baseplate base metal stress in tension for the OBE is calculated as:

12,790 psi < 16,500 psi allowable (Section 7.2.7) OK The highest predicted cell to baseplate base metal stress in tension for the DBE is calculated as:

15,375 psi < 22,088 psi allowable (Section 7.3.7) OK

b. Wrapper Plate Welds The fuel cells are composed of two L-shaped formed sections, placed to form a box section and welded at the opposite corners. The wrapper plates form the flux traps on the outside of each side of the cell box. The wrapper plates are welded to each cell side by 0.03 inch fillet welds, 1.5" long at 8" centers. Alternatively, spot welds could be used with three options on size and spacing. The maximum force in these welds occurs near the base. The limiting shear on these welds was calculated based on comparison to the Ref. [9] results. Note that for these welds the base metal material in shear governs.

Limiting Shear for OBE 8,847 psi < 11,000 psi allowable (Section 7.2.7) OK Limiting Shear for DBE 10,661 psi<14,725 psi allowable (Section 7.3.7) OK

c. Cell Seam Welds The cell seam welds are at opposite corners of each cell and are nominally 0.035" fillet welds that are 4" long at the bottom of the cell and on 8" centers at the top.

An optional butt weld of 0.055" is adequate if the 0.035" fillet welds is shown to be adequate. The maximum force on these welds occurs near the base.

The limiting shear on these welds is also calculated based on comparison to the Ref. [9] results. Note that for these welds the weld material governs.

Page 39 of 50

Stevenson & Associates Report 06Q3571.01-01 Max. Shear for OBE is 18,786 psi < 24,000 psi allowable (Section 7.2.7) OK Max. Shear for DBE is 22,637 psi < 37,200 psi allowable (Section 7.3.7) OK

d. Cell-to-Cell Welds Cell-to-cell connections are by a series of connecting welds along the cell height.

The weld stress is calculated based on the cell wall base metal stresses resulting from the maximum shear flow developed between two adjacent cells under OBE and DBE conditions. Note that the contributions from impact loads are included in the maximum shear flow.

The limiting shear on these welds is also calculated based on comparison to the Ref. [9] results. Note that for these welds the weld material governs.

Max. Shear for OBE is 19,681 psi < 24,000 psi allowable (Section 7.2.7) OK Max. Shear for DBE is 23,716 psi < 37,200 psi allowable (Section 7.3.7) OK 8.5 Support Plate Evaluation The total shear stress on the support plate was calculated for the loads transmitted from the pedestals. The maximum shear stress for OBE loads was calculated to be:

f, = 2544 psi < 11,000 psi allowable (Section 7.2.3) OK The maximum shear stress for DBE loads was calculated to be:

fv = 2629 psi < 17,100 psi allowable (Section 7.3.3) OK 8.6 Poison Insert Evaluations The poison insert assemblies were evaluated using the time-history results from the analyses for the maximum internal section forces and moments. The limiting maximum combined stress interaction coefficient for the inserts is for axial and bending stresses. For the OBE this stress interaction coefficient is 0.26 < 1.0 allowable and for the DBE = 0.28 < 1.0 allowable. These evaluations consider the worst case forces and moments for the insert members.

The shear force loads on the spot welds were for seismic loads were not calculated, but due to the very low stresses in the assemblies, are much less than the 170 lb capacity for the welds (from Section 7.2.7).

Thermal stresses in the Metamic panel assemblies were analyzed separately using the model developed to obtain equivalent section properties. Results of the thermal analysis show the stresses to be well within the allowable stress. Results are included in the summaries in Section 9.0.

Page 40 of 50

Stevenson & Associates Report 06Q3571.01-01 The Metamic panel assemblies were also analyzed for buckling capacity, using the model developed to obtain equivalent section properties and a linearized buckling analysis using SOLVIA. Local buckling was found to control panel capacity interaction. The linearized buckling analysis was verified by a large displacement incremental load analysis of the model with initial perturbed conditions based on the first buckling mode shape obtained from the linearized buckling analysis. Results are included in the summaries in Section 9.0.

8.7 Impact Evaluation 8.7.1 Local Stress Evaluations Due to Impact Between the Fuel Assembly and Cell Wall Local cell wall integrity is conservatively estimated from peak impact loads. Plastic analysis is used to obtain the limiting impact load that could lead to gross permanent deformation. As shown in Table 8.3, the maximum impact force is between the fuel and the rack, and occurs for the DBE loading for the single rack analysis with a coefficient of friction of 0.8.

The impact force applicable for a single cell is:

2.167x10 5 / 132 = 1642 lb This load was applied to a two cell finite element model of the cell wall to determine the local stress on the wall. A plastic analysis of the two cell model was analyzed for this load with and without the wedge blocks and considering the wedge blocks located at the point of impact. The maximum primary stress for this case occurs for the case without the wedge blocks with a value of:

Pm = 12,017 psi < 47,950 psi allowable (Section 7.3.8) OK The plastic analysis performed ensures that the primary membrane plus bending loads are limited to about the yield strength of the material and therefore, the primary membrane plus bending stress limitation is implicitly enforced.

8.7.2 Evaluation of the Fuel Assembly The permissible lateral load on an irradiated spent fuel assembly has been studied by the Lawrence Livermore National Laboratory (LLNL). The LLNL report [20] states that "...for the most vulnerable fuel assembly, axial buckling varies from 82g's at initial storage to 95g's after 20 years' storage. In a side drop, no yielding is expected below 63g's at initial storage to 74g's after 20 years' [dry] storage."

The maximum fuel-to-storage cell rattling force from the WPMR runs is 1,642 lb calculated above. The weight of a fuel assembly is 1700 lb. By inspection, the impact force from a side drop at 63 g's of the 1700 lb assembly is much greater than the 1642 lb impact load from the analysis and therefore, the fuel assembly is acceptable.

Page 41 of 50

Stevenson & Associates Report 06Q3571.01-01 8.7.3 Rack to Wall or Rack to Rack Impact Loads The storage racks do not impact the pool walls or adjacent racks under any simulation. The rack to rack or rack to wall gap elements did not close during the analytical simulations.

8.8 Consideration of Miscellaneous Equipment Loads For the Region 2 racks, an additional load of 2000 lbs can temporarily be set on top of these racks. The center of gravity for this equipment can be 24" above the top of the racks, and the equipment occupies a planar area of about 3' x 3'. Per ANO personnel, fuel bundles in the cells below where this equipment is placed would be removed. As analyzed, for a 3' x 3' area, this would include about 9 cells. Fuel weight for these 9 cells as analyzed is 9(1700 lb) =15,300 lb.

This mass, although distributed top to bottom for the rack, is more than 7.5 times the 2000 lb equipment mass. It is concluded from this plus the consideration of the dominance of the hydrodynamic mass for the dynamic response of the racks, that the presence of this equipment does not impact the seismic qualification of the racks. Similarly, a 200 lb weight, considered to represent testing equipment which may be placed anywhere on the top of any rack has an inconsequential effect on the seismic adequacy of the racks.

8.9 Analysis considering Half Full Fuel Racks An additional analysis was performed for a condition where the Region 3 racks were considered half full. This was run for the DBE Full Pool Analysis for a 0.8 coefficient of friction since this was the controlling case for all full pool analyses with the racks full. When compared to the results for the full rack conditions, this case was found not to control for the racks.

8.10 Comparison of Analysis Results to Westinghouse Ref. [9] Results The Westinghouse analysis in Ref. [9] evaluated the racks without the Metamic inserts. The methodology was similar to that used in the Stevenson & Associates (S&A) evaluation of Ref.

[5]. Table 8.3 below presents a comparison of component stresses that were independently calculated. As shown in the table, the results were very similar. This comparison is a further validation of the S&A Ref. [5] evaluation and that the use of the Westinghouse results for the Wrapper welds, cell seam weld and cell-to-cell weld is justified. Note that the S&A results are generally slightly less than those by Westinghouse. Potential reasons for this include the likelihood of less conservatism in the time history functions used by S&A and more accurate modeling of the fuel assemblies.

Page 42 of 50

Stevenson & Associates Report 06Q3571.01-01 Table 8.3 - Comparison of the Stress Results for OBE Load Case between Westinghouse 19] and Stevenson & Associates [5]

Component Stress D + L+ E Type Stress from S&A Stress from % Difference Analysis Westinghouse (S&A results as (psi) Analysis base)

(psi)

Cell Axial+Bending 10,209 11,826 15.8%

Cell to Base Plate Shear 12790 14,369 12.3%

Welds Support Pad Axial 9323 9,654 3.5%

Shear 2,116 2,544 20.2%

Bearing 8,274 8,492 2.6%

Threads Shear 8,129 8,275 1.8%

Support Plate Shear 2,544 2,951 16.0%

Note, the load case of D+L+E (E=OBE) in general controls the qualification of the racks (i.e. for all applicable load combinations, ratios of applied stress or limits to allowable values are the highest for this load combination.)

9.0 Evaluation of Spent Fuel Pool Structural Integrity for Increased Loads from Spent Fuel Racks The ANO-1 spent fuel pool consists of 6'-0" thick reinforced concrete walls and a 5'-6" thick floor slab. The pool is supported below by thick foundation walls. Concrete compressive strength for structural analysis for the ANO-I spent fuel pool is 5000 psi and reinforcement used was Grade 40.

The spent fuel pool was originally designed by Bechtel Corp. in accordance with the ACI 318-63 reinforced concrete building code, for loadings including deadweight of the structure, water, and spent fuel racks, hydrodynamic pressure from the water, operating thermal, accident thermal, seismic, tornado and flood loads. Rack loads were treated as a uniform load spread across the pool floor slab.

In 1981-1982, a reanalysis of the spent fuel pool structure including the foundation walls, refueling canal, and cask storage area was performed by Structural Dynamics Inc. in support of the re-rack project for ANO-1. Finite element methodology was used for this analysis. The same loads as described above in the Bechtel design were included in the analysis. The loads from the spent fuel racks included their deadweight (treated as live load on the pool floor slab) and vertical and horizontal seismic load effects. Rack loads were provided by Westinghouse Corp. This analysis, again used the acceptance criteria in the ACI 318-63 code, but supplemented the strength design methodology using provisions from the ACI 349-80 Nuclear Structure Reinforced Concrete Code. The load combinations used were in accordance with Standard Review Plan, Section 3.8.

Page 43 of 50

Stevenson & Associates Report 06Q3571.01-01 The dominant load effects were due to thermal expansion from the accident thermal loading for both analyses of the ANO-1 Spent Fuel Pool.

The recent evaluation of the spent fuel racks by Holtec International was performed to evaluate the effects on two of the racks for the additional weight of proposed poison inserts to two of the racks. This evaluation included analyses of the racks as described in Sections 3 through 8, and resulted in revised loads imparted from the racks to the pool floor slab. ANO engineering also conservatively redefined the total deadweight of all the racks, and included 5000 lb contingency loads between each rack and the pool walls around the periphery of the pool, (60,000 lb total), to account for miscellaneous items stored in this area of the pool. Additionally, Holtec specified a conservative hydrodynamic pressure resulting from the seismic displacement of the racks, which loads the pool walls for the height of the racks.

A review of the pool structure was performed using the 1981-1982 analysis by Structural Dynamics with the applied loads including the rack load effects. These effects were amplified using conservatively determined factors to account for the increased loads from the racks.

Specifically, the deadweight loading of the racks was factored up by the ratio of the maximum increase for any of the racks. The seismic load contribution (which consisted of combined seismic effects for the pool structure, the water, plus the rack seismic loads) was in general recalculated by factoring the seismic rack loads, by the maximum ratio calculated for the worst case rack in either the horizontal or vertical directions.

The 1981-1982 analysis checked 21 points for section moment, transverse shear, and in-plane shear. These 21 points were the highest stressed points for the various elements of the pool structure (e.g. the highest stressed points for each direction for the pool floor slab, the highest stressed point in each of the pool walls, etc.). Of these locations, two were for the pool floor slab and three were for the pool foundation walls. The spent fuel racks are supported only by the pool floor slab, which transmits load effects from the racks to the foundation walls, to the ground. Above the pool slab level, the rack loads have little impact on the pool structural elements.

Hence, the five critical locations for the pool floor slab and foundation walls were reviewed in detail, with the rack deadweight (live load case) and the seismic loading combinations factored as described above. The following table summarizes the results of the review of these five locations.

Page 44 of 50

Stevenson & Associates Report 06Q3571.01-01 Table 9.1 - Summary of Section Strength Review of Selected Locations Location Section Strength Previous Analysis Ratio Conservative Estimate of Ratio Parameter to Code Allowable to Code Allowable for Increased

_______________Rack Loads Pool Floor Slab Moment 0.76 0.76 East-West Section Transverse Shear 0.24 0.337 In-Plane Shear 0.16 0.174 Pool Floor Slab Moment 0.59 0.642 North-South Section -Transverse Shear 0.29 0.35 1

____________In-Plane Shear 0.42 0.47 1 Pool Foundation Moment 0.39 0.409 South Wall Transverse Shear 0.40 0.869

____________In-Plane Shear 0.55 0.859 Pool Foundation Moment 0.42 0.486 East Wall Transverse Shear 0.48 0.657

____________In-Plane Shear 0.96 0.998 Pool Foundation Moment 0.49 0.527 West Wall LTransverse Shea 0.58 0.660 In-Plane 1_________ Shear 10.80 10.951 The greatest increase for any location reviewed was for the South Pool Foundation Wall, for in-plane shear. The previous ratio for indicated shear to the allowable was 0.55. The conservatively estimated increased ratio is 0.859 or about a 49% increase. The highest stressed point was for the East Pool Foundation Wall, with a previous ratio of indicated in-plane shear to allowable of 0.96. The increased ratio is 0.998, which is about a 4% increase.

It should be noted that the added poison inserts increase the deadweight for only two of the racks by a total of about 11,500 pounds for each of the two racks. This is about a 4.7% increase in deadweight for these two racks. As analyzed, the inserts were considered with more than twice their actual weight, and compared to the previous analysis of the pool structure, represent an 11.45% increase in deadweight for these racks. For the other racks, the increase in deadweight over that previously analyzed was 3.3% for two Region 2 racks, 5.5% for the other two Region 2 racks, and 21% for the Region I racks. The average deadweight increase for racks and other items in the pool over the previous analysis was about 14.4%. For the review performed however, the weights for all racks were effectively increased about 29.5%. Hence, the increased load effects as applied for this review were conservatively overestimated. Seismic loads from the racks were similarly conservatively considered.

The results of this review demonstrate that the indicated increased loads from the racks have minimal effects on the pool structural elements, and that the structural integrity of the pool structure is maintained.

Page 45 of 50

Stevenson & Associates Report 06Q3571.01-01 10.0 Conclusions The overall design objectives of the spent fuel storage pool at ANO Unit 1 have been shown to meet the various Regulatory Guides, the Standard Review Plan, and industry standards. The structural adequacy of the SFP maximum density spent fuel racks at ANO Unit 1 with the new poison inserts have been evaluated using the appropriate regulatory and design standards.

Postulated loadings for normal, seismic, and accident conditions at the ANO Unit 1 site were considered in this analysis and evaluation. The design adequacy of the racks and the poison inserts has been confirmed with analyses that were performed in compliance with the USNRC Standard Review Plan [1], the USNRC Office of Technology Position Paper [2], Lawrence Livermore Report UCRL52342 [3] and ANO Specification APL-C-502 [4]. All applicable displacement and stress acceptance criteria have been met for the racks and the new poison inserts, as summarized for the OBE and DBE in Tables 10.1, 10.2 and 10.3 below. Results for the Pool Structure Analysis are summarized in Table 9.1 above.

Table 10.1 - Summary of Stre.~s Results fnr flRE Load Case Component Stress D +L+ E Type Applied Stress Allowable Stress Stress (psi) (psi) Interaction Cell Axial+Bending 10209.4 16500 0.619 Buckling 5485.4 7612 0.721 Wrapper Welds Shear 8847 11000 0.804 Cell Seam Welds Shear 18786 24000 0.783 Cell to Cell Welds at Top Shear 12259 24000 0.511 at Bottom Shear 19681 24000 0.820 Cell to Base Plate Tension 12790 16500 0.775 Welds Support Pad Axial 9322.9 16500 0.565 Shear 2115.9 11000 0.192 Bearing 8274.2 27500 0.301 Shear 8128.5 11000 0.739 Threads Support Plate Axial 330.1 16500 0.020 Shear 2543.9 11000 0.231 Metamic Insert Enclosure Channel Axial + Bending 44.3 151/16500 0.261 Buckling Axial Compression 94.2 lb 360 lb 0.262 Spot Welds Shear Force - 170 1 -

Note: 1) Shear force on the spot welds are insignificant from seismic loading.

Page 46 of 50

Stevenson & Associates Report 06Q3571.01-01 Table 10.2 - Summary of Stress Results for OBE Load Case With Thermal Stresses Component Stress D + L + E + T, ")

Type Applied Stress Allowable Stress Stress (psi) (psi) Interaction Cell Axial+Bending 12857 55000 0.234 Buckling 6055 7612 0.796 Wrapper Welds Shear 8847 31735 0.279 Cell Seam Welds Shear 18786 31735 0.592 Cell to Cell Welds at Top Shear 19111 31735 0.602 at Bottom Shear 25544 31735 0.805 Cell to Base Plate Welds Shear 25544 31735 0.805 Support Pad Axial 7660 55000 0.139 Shear 1887 31735 0.060 Bearing 6769 55000 0.123 Threads Shear 6635 31735 0.209 Support Plate Axial 330.1 55000 0.006 Shear 2273 31735 0.072 Metamic Insert Enclosure Channel Axial + Bending 3628.3 55000 0.066 Spot Welds Shear Force 44 lb 170 lb 0.26 M -,t M"lJ t St-ae l. . T Th*.. -

trt00 a'- ttt-..# A f--

.-. 1fO 1 T.M.

IA "la. II.S.IaA- I I*-M-1 - *A*RMA. n*r inserts were analyzed and obtained in this calculation. Thermal stresses are limited by normal force and friction for the 0

Support Pad, Threads and Support Plate, and consider a maximum temperature of 212 F within the cells for Cell and Cell welds.

Page 47 of 50

Stevenson & Associates Report 06Q3571.01-01 Table 10.3 - Summary of Stress Results for DBE Load Case Component Stress D + L + E' Type Applied Stress Allowable Stress Stress Interaction (psi) (psi)

Cell Axial+Bending 12263.5 28500 0.430 Wrapper Welds Shear 10661 14725 0.724 Cell Seam Welds Shear 22637 37200 0.609 Cell to Cell Welds at Top Shear 14772 37200 0.397 at Bottom Shear 23715.6 37200 0.638 Cell to Base Plate Welds Tension 15374.5 22088 0.696 Support Pad Axial 9847.3 28500 0.346 Shear 2350.8 17100 0.138 Bearing 8716.4 49500 N/A Threads Shear 8549.6 17100 0.500 Support Plate Axial 360 28500 0.013 Shear 2629.1 17100 0.154 Impact Loads on Cells Membrane Tension 12017 47950 0.251 Metamic Insert Enclosure Channel Axial + Bending 46.9 151/16500 (OBE) 0.311 Buckling Axial Compression 100 lb 360 lb 0.278 Spot Welds Shear Force -- 475 lb Note: 1) Shear force on the spot welds are insignificant from seismic loading.

It is noted that the addition of the inserts to the existing racks has only a slight impact on the rack dynamic response. The added stiffhess of the inserts is small and the mass is extremely small in comparison to the hydrodynamic mass, which dominates the response. The region 3 spent fuel racks have been shown to be seismically adequate with the addition of the Metamic poison inserts. Additionally, the poison inserts have been shown adequate for seismic induced loads.

Page 48 of 50

Stevenson & Associates Report 06Q3571.01-01 11.0 References

[1] USNRC NUREG-0800, Standard Review Plan, June 1987.

[2] (USNRC Office of Technology) "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," dated April 14,1978, and January 18, 1979 amendment thereto.

[3] UCRL52342, "Effective Mass and Damping of Submerged Structures," Lawrence Livermore National Laboratory, April 1, 1978.

[4] ANO Technical Specification APL-C-502, "Technical Specifications for Earthquake Resistance Design of Structures and/or Components Located in the Auxiliary Building for the Arkansas Nuclear One Unit 1 Power Plant," Rev. 2, 4-22-87.

[5] Stevenson & Associates Calculation ANO-ER-02-051, "Seismic Re-qualification of the Arkansas Nuclear One Unit 1 Spent Fuel Racks," Rev. 1, June 28, 2006.

[6] SOLVIA, Finite Element System, Version 99, Solvia Engineering, AB, Sweden, 1987-2001.

[7] SOLVIA, Finite Element System, Version 03, Solvia Engineering, AB, Sweden, 1987-2006.

[8] Westinghouse Drawing 1-W62A-017(1)-0

[9] Calculation 91E-0079-01, Revision 1, "Design Report for Spent Fuel Storage Racks for AP&L Co.," 9-20-96 (analysis performed in 1982).

[10] Levy, S. and Wilkinson, J.P.D., "The Component Element Method in Dynamics with Application to Earthquake and Vehicle Engineering," McGraw Hill, 1976.

[12] Stevenson & Associates Report, "Independent Evaluation of Seismic Response of Spent Fuel Storage Racks Currently Being Procured for Salem Nuclear Power Plant," Dec. 1, 1993.

[13] NUREG/CR-5912, BNL-NUREG-52335, "Review of the Technical Basis and Verification of Current Analysis Methods Used to Predict Seismic Response of Spent Fuel Storage Racks," Brookhaven National Laboratory, October 1992.

[14] Holtec International Drawing 4127, "Fabrication Drawing of Metamic Inserts for ANO-1 Spent Fuel Racks," Revision 12, 2/23/06.

[15] Holtec International Drawing 3917, "Metamic Inserts and Fuel Leadins for Spent Fuel Racks," Revision 10, 2/13/06.

Page 49 of 50

Stevenson & Associates Report 06Q3571.01-01

[16] ASME Boiler & Pressure Vessel Code,Section III, Subsection NF, 1980, through Winter 1981 Addendum.

[17] Stevenson and Associates, Program THSPEC - Verification and User's Manual for Computer Code THSPEC.

[18] ASME Boiler & Pressure Vessel Code,Section III, Appendices, 1980, through Winter 1981 Addendum.

[19] Stork Herron Testing Labs, Specimen Load Tests, Dec. 18, 2003.

[20] Chun, R., Witte, M. and Schwartz, M., "Dynamic Impact Effects on Spent Fuel Assemblies," UCID-21246, Lawrence Livermore National Laboratory, October 1987.

[21] ACI 349-85, Code Requirements for Nuclear Safety Related Concrete Structures, American Concrete Institute, Detroit, Michigan, 1985.

[22] ACI 318-95, Building Code Requirements for Structural Concrete, American Concrete Institute, Detroit, Michigan, 1995.

[23] Entergy letter to the NRC dated April 2, 2003. "License Amendment Request to Modify the Fuel Assembly Enrichment, the Spent Fuel Pool (SFP) Boron Concentration Technical Specification (TS) 3.7.14, the Loading Restrictions in the SFP in TS 3.7.15, and to Modify the Fuel Storage Design Features in TS 4.3."

[24] Entergy letter to Stevenson & Associates - ANO-2003-00144, dated Dec. 8, 2003,

Subject:

"Spent fuel Pool Inputs Required for the Unit 1 Rack Seismic Analysis."

[25] ANO-1 SFP Rack Design Inputs, Design Input Record, Document Number ANO-2005-00252, 11/16/2005.

[26] Stevenson & Associates Calculation ANO-ER-02-010, "Review of Structural Analysis of the Arkansas Nuclear One Unit 1 Spent Fuel Pool For Revised Fuel Rack Loads," Rev. 0, June 28, 2006.

[27] Chajes, A., "Principles of Structural Stability Theory," Prentice-Hall, New Jersey, 1974.

[28] ANO-1 SFP Rack Design Inputs, Design Input Record, Document Number ANO-2005-00252, 11/16/2005.

Page 50 of 50

Attachment 7 ICAN070603 List of Regulatory Commitments ICAN070603 Page 1 of 2 List of Regulatory Commitments The following table identifies those actions committed to by Entergy in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.

TYPE (Check one) SCHEDULED ONE- CONTINUING COMPLETION COMMITMENT TIME CONTINCE DATE (If ACTION COMPLIANCE Required)

The surveillance coupons will be approximately 7" x 5" and 0.100" thick, identical in composition and manufacturing process as the Metamic@ in the x inserts (i.e., created from the same manufacturing lot used to manufacture the Metamic PIAs).

The coupons will be mounted in stainless steel jackets simulating the actual insert design. x The coupon tree will have ten or more coupons. x The coupon tree will be installed within a flux trap in Region 2. x The coupons will be staggered and placed adjacent to the active fuel region where, based on the bumup profile, the localized burnup is greater than the assembly average bumup.

No welding will be used on the Metamic@ as per the PIA design. x Scratches will be simulated by the mechanical etching or scribing the surface of the coupons. The scratches will be formed using hardened materials made out of carbon steel, stainless steel, and Metamic. The scratches will not be cleaned after x being applied to ensure an evaluation will be performed of the corrosion affects of leaving the trace material in a scratch.

Coupons will be examined on a two year basis for the first three intervals and thereafter on a 4 to 5 x year interval over the service life of the inserts.

During the first six years, freshly discharged fuel assemblies will be placed on two sides of the coupon tree to ensure that the dose to the coupons is maximized.

1CAN070603 Page 2 of 2 TYPE (Check one) SCHEDULED ONE- CONTINUING COMPLETION COMMITMENT TIME CONTINCE DATE (If ACTION COMPLIANCE Required)

Measurements to be performed at each inspection will be as follows:

Physical observations of the surface appearance to detect pitting, swelling or other degradation,

" Length, width, and thickness measurements to monitor for bulging and swelling (Measurements will be taken in five procedurally defined locations prior to x placing the coupons in the ANO-1 SFP.

When the coupon is removed, measurements will be taken in the same locations as the original measurements.)

Weight and density to monitor for material loss, and

Neutron attenuation to confirm the B10 concentration or destructive chemical testing to determine the boron content.

The ANO-1 SAR will be amended no later than the next required update after the proposed TS change is approved and implemented. This SAR update will x indicate that ANO-1 has chosen to comply with 50.68(b).

Attachment 5 1CAN070603 Spent Fuel Pool Racks Modifications with Poison Material Inserts in ANO Unit I

v t e Letter Sequence Request
TAC:MD2674, (Approved, Closed)
Initiation Request, Request, Request Acceptance... Supplement, Supplement, Supplement Administration Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance Questions Answers Results Approval, Approval Other: 0CAN071601, Response to Generic Letter 2016-01, Monitoring of Neutron-Absorbing Materials in Spent Fuel Pools, ML063470192, ML063470197, ML070240605, ML070440038, ML070440042, ML072630086
MONTHYEAR2007-11-05 [Table View]

1CAN070603, License Amendment Request to Support the Use of Metamic@ Polson Insert Assembles in the Spent Fuel Pool

Text

SUBJECT:

REFERENCES:

Dear Sir or Madam:

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

5.0 TECHNICAL ANALYSIS

6.0 REGULATORY ANALYSIS

Subject:

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1CAN070603, License Amendment Request to Support the Use of Metamic@ Polson Insert Assembles in the Spent Fuel Pool

Text

SUBJECT:

REFERENCES:

Dear Sir or Madam:

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

5.0 TECHNICAL ANALYSIS

6.0 REGULATORY ANALYSIS

Subject:

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