Text

License Amendment Request to Modify the Fuel Assembly Enrichment, the Spent Fuel Pool (SFP) Boron Concentration TS 3. 7.14, the Loading Restrictions in the SFP in TS 3,7,15m and to Modify the Fuel Storage Design Features in TS 4.3
	ML030970336
Person / Time
Site:	Arkansas Nuclear
Issue date:	04/02/2003
From:	Anderson C; Entergy Operations
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	1CAN040302
	Download: ML030970336 (185)
	v • d • e

Entergy Operations, Inc.

1448 SR 333

- Entergy Russeliville, AR 72801 Tel 501-858-4888 Craig Anderson Vice President Operations ANO 1CAN040302 April 2, 2003 U.S. Nuclear Regulatory Commission Attn Document Control Desk Washington, DC 20555 SUBJECT Arkansas Nuclear One, Unit 1 (ANO-1)

Docket No. 50-313 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 REFERENCES

1. Letter to the NRC dated August 8, 2002, "Use of Metamic In Fuel Pool Applications" (OCAN080201)

2. Letter to the NRC dated March 18, 2002, "Degradation of Boraflex( in ANO-1 Spent Fuel Pool" (1CAN030203)

3. Letter to the NRC dated September 6, 2000, "License Renewal Application RAls" (1CAN090002) 4 Letter to the NRC dated January 29, 2003, "License Amendment Request to Change the Spent Fuel Pool (SFP)

Loading Restrictions (2CAN010304)

Dear Sir or Madam.

Pursuant to 10CFR50.90, 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 The ANO-1 spent fuel pool (SFP) storage racks are divided into two Regions. The Region 1 racks contain Boraflex The Region 2 racks do not contain any poison panel inserts. A portion of the SFP storage racks in Region 2 will be modified by the insertion of Metamic poison panels. The area with the Metamic poison panels will be defined as Region 3. No loading restrictions will apply to the newly defined Region 3 (i.e., new and spent fuel can be stored in Region 3 without restrictions). By letter dated March 18, 2002 (Reference 2), Entergy described its plans to administratively control spent fuel loading based on the degradation of BoraflexD. Continued degradation of Boraflex)is projected and therefore the proposed change will take no credit for Boraflex in Region 1 and will include loading restrictions for Region 1. In addition to the above P} b IN

1CAN040302 Page 2 of 3 modification, Entergy proposes to increase the allowable initial fuel assembly uranium-235 (U-235) enrichment from 4.1 weight percent (w/o) to a maximum enrichment of 5.0 w/o. Analyses have been performed which include the effects of the U-235 enrichment, the insertion of the Metamic poison panels, and not crediting BoraflexO in the Region 1 racks. In addition, a criticality analysis has been performed to demonstrate that the new fuel storage racks can accommodate storage of the higher enriched fuel assemblies. The proposed change will modify the minimum SFP boron concentration requirements in TS 3.7.14, the spent fuel loading restrictions defined by TS 3.7.15, and the initial fuel enrichment and effective multiplication factor (Kff) and the associated Figure in TS 4.3.

Entergy submitted by letter dated August 8, 2002, a topical report (Reference 1) prepared by Holtec Intemational that describes the physical and chemical properties of Metamic. The report also includes the test results for the use of Metamic in fuel pool applications. Approval of the topical report is required to support the requested plant specific TS changes addressed in the attachments to this letter.

The proposed change has been evaluated in accordance with 10CFR50.91(a)(1) using criteria in 10CFR50.92(c) and it has been determined that this change involves no significant hazards considerations. The bases for these determinations are included in the attached submittal.

The proposed change includes new commitments that are listed in Attachment 6. The Nuclear Regulatory Commission has not approved TS changes for other facilities that allow the use of Metamic) poison panels in wet storage applications. However, by letter dated January 29, 2003 (Reference 4) Entergy submitted a similar TS change for ANO, Unit 2 (ANO-2). The proposed change to the ANO-1 TS is similar to the ANO-2 change with the following exceptions:

1) ANO-1 is requesting to increase the initial fuel assembly U-235 enrichment and change the loading pattern in the new fuel storage racks; 2) the ANO-2 SFP boron concentration is being increased to allow for various loading configurations; and 3) the concentration of Boron Carbide content in the Metamic panel differs between the two units Entergy requests approval of the proposed amendment by November 15, 2003, in order to support fuel re-arrangement in the ANO-1 SFP in late 2003 or early 2004. Once approved, the amendment shall be implemented within 90 days. Although this request is neither exigent nor emergency, your prompt review is requested.

If you have any questions or require additional information, please contact Dana Millar at 601-368-5445.

1CAN040302 Page 3 of 3 I declare under penalty of perjury that the foregoing is true and correct. Executed on April 2, 2003.

Sincerely, CGA/dm Attachments:

1. Analysis of Proposed Technical Specification Change

2. Proposed Technical Specification Changes (mark-up)

3. Changes to TS Bases pages (for information only)

4. Holtec License Report

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

6. Commitments cc: Mr. Ellis W. Merschoff Regional Administrator U. S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064 NRC Senior Resident Inspector Arkansas Nuclear One P. O. Box 310 London, AR 72847 U. S. Nuclear Regulatory Commission Attn: Mr. John L. Minns MS 0-7 D1 Washington, DC 20555-0001 U. S. Nuclear Regulatory Commission Attn: Mr. Thomas W. Alexion MS 0-7 D1 Washington, DC 20555-0001 Mr. Bernard R. Bevill Director Division of Radiation Control and Emergency Management Arkansas Department of Health 4815 West Markham Street Little Rock, AR 72205

Attachment I I CAN040302 Analysis of Proposed Technical Specification Change to 1CAN040302 Page 1 of 10

1.0 DESCRIPTION

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

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

Allow insertion of MetamicO poison panels in a newly defined Region 3 of the ANO-1 spent fuel pool (SFP).

Redefine the loading pattern in the current Region 1,taking no credit for Boraflex.

Redefine the loading pattern in the current Region 2.

Modify the SFP boron concentration from 21600 ppm to >1600 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 enrichment of Uranium-235 (U-235) from the current enrichment of 4.1 weight percent (w/o) to a maximum of 5.0 w/o.

Redefine storage patterns in the new fuel storage racks.

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.

Appropriate changes will also be made to the associated TS Bases. These are included for information only as Attachment 3.

The changes are desired in order to address the degradation of Boraflex and to support the creation of a new Region 3 in which MetamicO poison panel inserts will be installed. In addition the increase in maximum fuel assembly enrichment is desired to support reactor fuel loading capabilities. Approval of the proposed changes is desired by November 15, 2003 to support fuel shuffling in late 2003 or early 2004.

2.0 PROPOSED CHANGE

ANO-1 TS 3.7.14 and Surveillance Requirement 3.7.14.1 define the minimum required SFP boron concentration. The proposed change will result in changing the "greater than or equal to" sign (2) to a "greater than" sign (>). The proposed change, although minor, results in a more accurate representation of the analysis assumption of greater than 1600 parts per million (ppm) for the boron concentration. In addition the TS 3.7.14 currently applies whenever fuel assemblies are stored in the SFP and a 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 a SFP verification has been performed or not. This change is needed to support the new criticality analysis that takes credit for boron to assure the required subcritical margin is maintained.

to 1CAN040302 Page 2 of 10 ANO-1 TS 3.7.15 and Figure 3.7.15-1 define loading restrictions for any fuel assembly that is stored in Region 2 of the ANO-1 SFP. Currently, no loading restrictions are required by TS in SFP Region 1, which allows storage of new and spent fuel. Region 1 contains Boraflex poison panels. Calculations indicate that the Boraflex content in Region 1 will degrade below the original Boraflex content assumed in the criticality analysis (Reference 2). Based on continued projected degradation of Boraflex, the proposed change will take no credit for BoraflexD in Region 1 and will include loading restrictions for Region 1. This will result in changes to TS 3.7.15 and Figure 3.7.15-1, and the creation of a new Figure 3.7.15-2 based on new criticality analysis for Regions I and 2. In addition, a portion of the SFP racks in Region 2 will be modified by the installation of Metamic poison panel inserts. This will result in the creation of a new Region 3 that will allow unrestricted storage of new and spent fuel. Entergy has submitted a topical report (Reference 1) that describes the physical and chemical properties of Metamic. Approval of the topical report is required to allow the use of Metamic in SFP applications.

ANO-1 TS 4.3.1.1a and 4.3.1.2a define a maximum U-235 enrichment of 4.1 w/o. The proposed change will allow the maximum enrichment to be 5.0 w/o.

ANO-1 Figure 4.3.1.2-1 depicts locations in the fresh fuel storage racks in which fuel loading is prohibited. Based on the increase in fuel assembly enrichment from 4.1 w/o to a maximum enrichment of 5.0 w/o, two loading pattern configurations will be proposed, which will result in the addition of a new Figure 4.3.1.2-2. Figure 4.3.1.2-2 will depict the loading pattern associated with fuel assemblies whose enrichment is up to 4.2 w/o and Figure 4.3.1.2-1 will illustrate the loading configuration for fuel assemblies whose enrichment is up to 5.0 w/o. A change is also proposed to TS 4.3.1.2.d and e to state that the fuel storage racks shall be maintained in accordance with the two proposed figures which are based on fuel enrichment.

TS 4.3.1.1 b states that an effective multiplication factor (kff) of less than or equal to 0.95 is achieved when the SFP racks are fully flooded with unborated water. This subcriticality margin will be maintained by taking credit for boron, which results in the proposed change that states that a kff of less than or equal to 0.95 is maintained when the pool is flooded with 400 ppm borated water.

A new TS 4.3.1.1 c will be added, which will describe that a kff 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.

In summary, the proposed change will modify TS 3.7.15 to define Region 3 in which Metamic poison panels will be inserted and impose new loading restrictions in Region 1 and Region 2.

TS 4.3.1 will be changed to support higher fuel assembly enrichment.

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. SFP storage is currently divided into two regions. Region 1 utilizes Boraflex 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 to 1CAN040302 Page 3 of 10 accommodate irradiated fuel that has sustained approximately 80 percent of the design burnup.

Placement of fuel in Region 2 is determined by bumup 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 which 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 burnup based on the proposed Revision 2 of Regulatory Guide 1.13.

The Region 1 storage racks are composed of individual storage cells made of type 304L stainless steel and conform to the requirements of ASME B&PV Code,Section III, Subsection IV. 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 spent fuel pool 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 304L stainless steel cells assembled in a checkerboard pattem, 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 are designed to accept poison inserts if future need arises. This design is also provided with leveling pads which contact the spent fuel pool 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.

3.2 New Fuel Storage Area The new fuel storage area is a separate and protected area for the dry storage of new fuel assemblies in the fuel storage and handling area. 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 any storage 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.

Section 9.6 of ANO-1 Safety Analysis Report (SAR) contains a detailed description of the ANO-1 SFP and new fuel storage area.

3.3 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 enrichment and bumup of each spent 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 to 1CAN040302 Page 4 of 10 replacing the existing racks. The region designated as Region 1 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 w/o. 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).

3.4 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 TS Figure 3.7.15-1. 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.

3.5 Spent Fuel Pool 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 spent fuel pool. It is designed to maintain the SFP water at less than or equal to approximately 1500 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 spent fuel pool water at approximately 1200F with a heat load based on the decay heat generated from approximately one-third of the core fuel assemblies discharged at the end of any 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 spent fuel 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 spent fuel pool circulating pumps take suction from the SFP and recirculate the fluid back to the pool after passing through the coolers. 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. The system provides purification of the SFP water through the SFP demineralizer and SFP filters in order to remove fission and corrosion products and to maintain water clarity for fuel handling operations.

to 1CAN040302 Page 5 of 10 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.

4.0 TOPICAL REPORT AND COUPON SAMPLING PROGRAM Entergy submitted to the Nuclear Regulatory Commission by letter dated August 8, 2002 (Reference 1), a topical report that will support the use of MetamicO poison panel inserts in SFP applications. 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. 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. Coupons suspended on a mounting tree will be inserted into either an empty fuel cell or an empty flux trap in an ANO-1 SFP rack that is surrounded by spent fuel assemblies. Ten coupons will be created from the same manufacturing lot that will be used to manufacture the Metamic poison panels and inserted into the SFP. The coupon measurement program is intended to monitor for changes of the following physical properties of the MetamicO absorber material: visual observation and photography, neutron attenuation, dimensional measurements (length, width and thickness),

weight and specific gravity. The physical changes observed upon evaluation will reflect the probable changes that are occurring in the Metamic poison panel inserts and thus provide a method of verifying that the assumptions used in the SFP criticality analyses remain valid.

5.0 TECHNICAL ANALYSIS

The proposed change will result in defining a new region, designated as Region 3, in the SFP in which Metamic poison panels will be inserted. New loading patterns are proposed in the existing Region 1, in which no credit for BoraflextD will be taken, and in Region 2. In addition, Entergy is proposing to increase the fuel assembly maximum U-235 enrichment. Attachments 4 and 5 to this letter provide detailed technical analyses in support of the proposed change. A brief summary of the content in these attachments is included in sections 5.2 through 5.7.

5.1 Dose Consequences Associated with Increased Fuel Enrichment An evaluation of the potential offsite and control room dose radiological consequences of a Fuel Handling Accident (FHA) was performed considering the change in radiological consequences associated with the increase in maximum U-235 enrichment from 4.1 w/o to 5.0 w/o. 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 activates and/or a criticality accident. The enclosed attachment prepared by Holtec International addresses the criticality aspects of the FHA. The following dose consequences for the increase in fuel enrichment were calculated for the FHA.

to 1CAN040302 Page 6 of 10 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 change in dose consequences results in only a small fraction of the 25% 10 CFR 100 limits.

5.2 Material Considerations It is proposed that Metamic will be inserted in the newly defined Region 3. The physical and chemical properties of Metamic have been submitted by letter to the NRC (Reference 1).

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 designed as Region 1, Region 2, and Region 3. The criticality analysis currently in place for Region 1 assumes the presence of Boraflex. In the new analyses, no credit was taken for the Boraflex in Region 1. The new analyses also assume Metamic poison panel inserts are installed in Region 3. It was concluded that in order to assure Kff remains less than 0.95 in the various storage configurations that are allowed considering the storage of both spent 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 1600 parts per million (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 by procedure.

The criticality analyses also include consideration for storage of fuel assemblies with a maximum U-235 enrichment of 5.0 w/o. Analyses were performed for the new fuel storage racks and the SFP racks considering storage of the higher enrichment. A new storage configuration is included in the proposed change for the new fuel storage racks.

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 to 1CAN040302 Page 7 of 10 that cladding integrity would be maintained. None of the temperature limits or corrective actions for the SFP cooling system change.

5.5 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 large margins of safety in all storage modules.

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

However, it is expected that minor changes to the design will occur during product development and testing. All changes will be reflected in the finalized evaluation of the poison insert structural analysis.

5.6 Mechanical Accident In line with the current approved philosophies, the postulated fuel assembly drop events for Region 3 of the SFP racks were conservatively evaluated and conclude that the poison inserts, as well as the cell wall of the impacted rack cell could be significantly damaged. Conservatively assuming that all poison inserts in Region 3 were damaged, the evaluation concludes that the racks will remain subcritical when credit is taken for the proposed TS limit of greater than 1600 ppm soluble boron in the pool.

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, Criticality accident requirements. The exemption was granted on October 6, 1998 (TAC NOS. MA1278 and MA1279). Upon approval of the proposed change, ANO-1 will fully comply with 10 CFR 50.68 paragraph (b), and the exemption to 10 CFR 70.24 will no longer be required.

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 any GDC differently than described in the SAR.

to 1CAN040302 Page 8 of 10 6.2 No Significant Hazards Consideration The proposed changes will modify the Arkansas Nuclear One, Unit 1 (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 burnup and initial assembly average U-235 enrichment (up to 4.1 weight percent (wlo) 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 inserts which will provide the neutron absorption capability required to allow storage of various combinations of fuel burnup and enrichment without loading restrictions. 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 5.0 w/o. The criticality analyses associated with these changes will require credit for boron in the SFP to assure the SFP remains subcritical with an effective multiplication factor (kef) less than 0.95. The above proposed changes will result in modifications to TSs covering the SFP boron concentration and the SFP storage and design features related to fuel storage. The proposed change to the SFP boron concentration is a minor editorial change.

Entergy Operations, Inc. has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) 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.

The three fuel handling 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.

A vertical drop directly upon a cell will cause damage to the racks in the active fuel region. The current 1600 ppm soluble boron concentration TS limit will ensure that Keff does not exceed 0.95.

to 1CAN040302 Page 9 of 10

A fuel assembly dropped on top of the rack horizontally will not deform the rack structure such that criticality assumptions are invalidated. The rack structure is such that an assembly positioned horizontally on top of the rack results in a minimum separation distance from the upper end of the active fuel region of the stored assemblies. This distance is sufficient to preclude interaction between the dropped assembly and the stored fuel.

An inadvertent drop of an assembly between the outside periphery of the rack and the pool wall is bounded by the worst case fuel misplacement accident condition.

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 1600 ppm currently required by the AN0-1 TS. Thus, a five percent subcriticality margin can be easily met for postulated accidents, since any reactivity increase will be much less than the negative worth of the dissolved boron.

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

boiling between cells). An "optimum moderation" accident in the new fuel pit was previously evaluated and the conclusions of that evaluation have not changed as a result of the fuel enrichment.

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

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

Response: No.

The proposed changes will define a portion of the current Region 2 as Region 3. The new region will contain Metamic poison panel inserts and will allow unrestricted storage of fuel assemblies with various enrichments and burnup. To support the proposed change new criticality analyses have been performed. The analyses resulted in new loading restrictions in Region 1 and Region 2. The presence of boron in the pool water assumed in the analysis is less than the 1600 ppm currently required by the ANO-1 TSs. Thus, a five percent subcriticality margin can be easily 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.

to I CAN040302 Page 10 of 10 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 that fuel assemblies of less than or equal to five weight percent U-235 enrichment when loaded in accordance with the proposed loading restrictions will be maintained within 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 Kff 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 currently 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 showed that the integrity of these structures will be maintained with the addition of the poison inserts. The structural requirements were shown to be satisfied, so the safety margins were maintained.

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

Based on the above, Entergy concludes that the proposed amendment(s) present 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 1CAN040302 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 > >1600 ppm. I APPLICABILITY: When fuel assemblies are stored in the spent fuel pool-and-a-spent-uel pool-verification has-not-been performed -sinre-the last-movement of-fuel assemblies4nrhe-spent-fuepwool.

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.2.1 Initiate action to restore Immediately spent fuel pool boron concentration to within limit.

OR A424nitiate-action-o-perfor-m-a Immediately spent -fuel pool-verification-SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.14.1 Verify the spent fuel pool boron concentration is 7 days 2> 1600 ppm.

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 LCO 3.7.15 The combination of initial enrichment cooling time, and burnup of each spent fuel assembly stored in Rqe~iojn, anid Region 2 shall be within the acceptable range of Figures 3.7.15-1 and 3.7.15-2 respvely or in accordance with Specification 4.3.1.1.

APPLICABILITY: Whenever any fuel assembly is stored in Region IoRegion 2 of the spent fuel pool.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME I. Requirements of the A.1 -------- NOTE LCO not met. LCO 3.0.3 is not applicable.

Initiate action to move the Immediately noncomplying fuel assembly from Region 2the affected Region.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.15.1 Verfy by administrative means the initial enrichment. Prior to storing the coo !rig time and bumup of the fuel assembly is in fuel assembly in accordance with Figure 3.7.15-1Fjre 3.7.15-2 or Region 1 or 2.

Specification 4.3.1.1.

ANO-1 3.7.1 5-1 Amendment No. 21-5,

Spent Fuel Pool Storage 3.7.15 Fiure -3.-7,1-11 D. @,. . .. L r. . C--

Spent-Fuel Storage Plaoks MINIMUM BURNUP VS. INITIAL ENRICHMENT FOR REGION 2 STORAGE A 36 I II I II I I I [A I;_= _

J l T I 1 I10- A 32 I T 1I T I T I T I I IX I I I_ I , I I.

23

_ Non-restrictod -- ___s_

__ (acceptable range)__ _ _.__ _

24

=I I I I = =I =;

0

_~~~1- _- I Lr.

^u a E

m a

CR 16 W

E 12 4

_X_ __ft___

a

- Restricted to

- Checkerboard Spacing 4

(unacc~eptable range) 0 1.0 (0.1.4) Z2 3.0 4.0 (0. .11 Initial Assembly Average Enrichment (w/o U-235)

ANO-1 3.7.1 Amendment AN No. 24-5,

Spent Fuel Pool Storage 3.7.15 Figure 3.7.15-1 Burnup versus Enrichment Curve for SFP Storage Racks Region I

'''1'''''''''''''''''''1'''''''''''''''''1'''''''''''''''''1''''.......

35 30 R 25 Decay Time M

0.

CD 0 Years

- 20 -- -- 5 Years L.

m -- IO Years a,

- -- -- 15Years

- - -- - 20 Years t.0 15 E01 1,

10 5

0 2 2.5 3 3.5 4 4.5 5 Initial Assembly Average Enrichment (w/o U-235)

ANO-1 3.7.15-2 Amendment No. 2415,

Spent Fuel Pool Storage 3.7.15 Figure 3.7.15-2 Burnup versus Enrichment Curve For SFP Storage Racks Region 2.

40

/ I 35 Non-restricted / / , -

30 (acceptable range)

Decay Time

<925 0 Years

- - - -5 Years

.- 10 Years MI20

15 Years 20 Years Inu 2 U,

0.15 ci to Checkrboard

.Restriced

!2o'>'Spacing

,.(unacceptable range) 10

/,4 :=.

5 0

2 2.5 3 3.5 4 4.5 5 Initial Assembly Average Enrichment (wlo U-235)

ANO-1 3.7.1 S3 Amendment No. 215,

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 4115.0 weight percent;

b. kuf, < 0.95 if fully flooded with 400.ppm.lofunborated water, which includes an allowance for uncertainties as described in Section 9.6.2.4.3 of the SAR;

c. rkf <1.0 if fully flooded with unborated water, which includes an

~~~~~~~~~~~~~~~. ....... ..... .. ......... ........ ... .. ......... I.. ........... ......... .. ............................

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 coqloig tmes.ppD ,a discharge bumup in the "acceptable range" of Figure 3.7.15-1 or Figure3.7.15-2 allowed unrestricted storage in either fuel storage rack Region 1,or Region 2; or Region 3. and ef. New or partially spent fuel assemblies with cooling times and a discharge burnup in the "unacceptable range" of Figure 3.7.15-1 o.Figupe3.7.157-,2stored in Region 413, or in checkerboard configuration in either Region -1or Region 2.

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 1415.0 weight percent;

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

c. kff < 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 show, .Fuel loa,,ding derneas proh bited in interior storage cells as depicted in Figures 4.3.1.2-1 and 4.3.1,-,2, based on fuel enrichment ,ed-from-use-during fuel storage 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 5.0 w/o U-235

NORTH 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. 21-5,

Design Features 4.0 Figure 4.3.1.2-2 Fresh Fuel Storaqe Rack Loading Pattern for a Maximum Enrichment of 4.2 w/o U-235

4 NORTH I I I- -

NO NO NO NO NO NO NO .NO

__ __ __ I__ __ .1 1 .1 "NO" Indicates a location in which fuel Goadin~g isprohibited.

ANO-1 4.0-6 Amendment No. 245,

Attachment 3 1CAN040302 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 enrichment coolintimre, and discharge bumup. Although the water in the spent fuel pool is normally borated to t> 1600 ppm, the criteria that limit the storage of a fuel assembly to specific rack locations are conservatively developed without taking credit for boron in the spent fuel pool water.

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 poolsinfinite arrays. Region .13 is designed to accommodate new fuel with a maximum enrichment of 44050Q wt% U-235, or spent (irradiated) fuel regardless of the discharge fuel bumup.

Region 1 and 2 s-are designed to accommodate fuel of various initial enrichments which have accumulated minimum cooling times and burnups within the acceptable domain according to Figure 3.7.15-1land 3.7.15-2. Fuel assemblies not meeting the criteria of Figure 3.7.15-1 or 3.7.152 shall be stored in accordance with Specification 4.3.1.1.ef.

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 specify that the limiting keff of ".51:.0 be evaluated in the absence of soluble boron. Hence, the design of both the three regions is based on the use of unborated 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-1 6.1-1975 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. 245,

Spent Fuel Pool Boron Concentration B 3.7.14 APPLICABLE SAFETY ANALYSES Most-accident-conditions-will-not -result-in an inrease in KI.Sof the rack:

E-xamnples-are4he4oss-Gf-cooling systems reactivity-decreases-with-deoFeasing water density) and dropping a fuel assembly on-top-of -the racw Sthe rack struoAureert tor-Grnticality-s-nt-defofmed-andthe-asserbly-has-mre that-eight- inches -of-water-separating-it-from the active fuel in the-rest-of-the raGk-whh-precudesInteraGtlon-4--w-ever,-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 wil-dec-rease reactivity- byis -toassure the fuel assemblies will be maintained in a subcritical arrayNwith -approximate %-AK. T-hu5-Keff < 0.95 Ganrbe-eaay-me postulatedin the event of a postulated drop accidents. since any--reactivity inebe- man he-negati -worth ofth-dssolved boron.Analysis has shown that, during a postulated misplacement accident with the fuel stored within the limits of this specification,,a Ko < 0.95 will be maintained when the boron concentration is at or above 800 ppm.

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

LCO The specified concentrationŽ> 1600 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.j until a -complete spent fuel pool verification-has been -performed following-the last-movement-of-fue-assemblies4n-the-spent-uelt-pool:This-L-O-does-not apply -following the-verification since the verification-would-confirTn-that there are- no misloaded-fuel -assermblies.---With-no-furt~herfuel-asssenbly moverment-in progfess-4her4is potentiat4or-a-misod la ssembly-Gr-a-dropped-fuel assembly.

ACTIONS A 1,and A2.1;-and-A 2.2 ANO-1 B 3.7.14-2 Amendment No. 24-5,

Spent Fuel Pool Boron Concentration B 3.7.14 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,

Spent Fuel Pool Boron Concentration B 3.7.14 ACTIONS (continued)

A.1 and 2.1, and A.2.2 (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 afuel 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 alternative -isto immediateytnae-peormanceofaspent-uerp GbenfiatonAo-ensufe proper-locations of-the-fuel since-the-last-movement-of-fuel assemblies-in the spen4ueloow -leowevepprio-r-dor to resuming movement of fuel assemblies, the concentration of boron must be restored. Either-of these aotiGnsare-aGeeptable,-and-once-nia must ted. be-Gontinued--Rn4he-auetIk-s completed--The immediate Completion Time for initiation of these-this 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 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. 1QCFR 50,68 ANO-1 B 3.7.14-3 Amendment No. 245,

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. The spent fuel storage cells are installed in parallel rows with center to center spacing of 10.65 inches in each direction.

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-accommodate--new-fuel with-a maxim-um-e.srichment-of-4A-.AO-).vt%-U-935,-or-spent-(irradiated)-fuel-regardless of-the discharge-fuel burnup. and Region 2 is are designed to accommodate fuel of various initial enrichments which have accumulated minimum burnups within the acceptable domain according to Figures 3.7.15-1 and 3.7.15-2. Regiq 3ned to accommodate

,6'e ' ...A new fuel with a maximum enrichment of 5.Owt% U-235 or spent (irradiated) fuel regardless of the discharge fuel burnup. Fuel assemblies not meeting the criteria acceptable range of Figures 3.7.15-1 or 3.7.15-2 shall be stored in accordance with paragraph 4.3.1.1.e-f in SAR-TS Section 4.3, Fuel Storage.

APPLICABLE SAFETY ANALYSES Criticality of fuel assemblies in the spent fuel storage rack is prevented by the design of the rack which limits fuel assembly interaction. Thswis-done-byfixing-the-minimum separation between assemblies -and inserting neutron-poison between-assemblies -in Region-1.RegiLon 1 and- Region 2 controls fuel assembly interaction by fixing the minimum separation between assemblies and by setting enrichment and burnup criterion to limit fissile materials. Reqion 3 controls fuel assebly intera wth the insertion of Metamic poison panels This is sufficient-to -maintain -Aa keff of

0.95 for spent fuel of original enrichment of up to 4:1:05.Owt is t g credit for boron (nominal conditions require 400 ppm for all regions). Akn of < 10Qis accomplished when no credit for boron is taken. However uel-assembles4o--be--stred4n4he-spentfuel--ooRegion-2 which do not-meet enrichment -and-burnup criterion must be stored in -a checkeboard pattem4o-maintainra-k ,-95-Gr-essnworder-to-pr-eventinadve-tenuel-assembly 1

insertion into two-adjacent storage locations, -vacant-spaces adjacent-to the-f aces-of-any fueassemblywh hfot-t~ egiun-umrup-criteria-unrestriGted}-a-e physically bIl ed-before-any-such-fuel assembly is. placed in Region-2.(Ref.- 4).-4n adio,4e-a eskjnaterekboardar-angemet4sfn4aI storage in Region-2-by-a-row of--vacant-storage spaces,--Ref: .2:)

AN0-1 B 3.7.15-1 Amendment No. 245,

Spent Fuel Pool Storage B 3 7.15 Regaion 1 and Region Fresh or spent fuel with enrichments up to 5.0 wt% U-235 may be stored in Region 1 and/or Region 2 perT 3..5Fgrs371- Rgo )ad3.7.15-2(Rgo2)

The fiqgures.use initialaU-235 enrichment and bumup to determine loading restrictions. Th ivecuyrvye~s"o..n each figure reresent decay time in years. Decay time is based on the time at which an assembly

................................ .................................... ............ J.........

is permanently placed in the pool

.. ...... ' ~ ' h (i.e., if an assembly is placed in the pool one cycle and then returned to the core, the initial time in the pool is not included in the decav time once the assemb lyis

,m,,,ove.d from...

p..e.rm..,m..a.n.e..n,.t.lyrem.... .... thecore........ ..) Thef....ol.l,.,o.wing

.. . equa.tio.. ns .....

we.... .... s..d..

re.u..

generate the decay cur and may be used toinntemolate fuelplcmnin esn restricted or non-restricted re9 ions:

.R.egion.. ecayTime Curves:

Decay Time., Bumup, GWDIMTU Years 0 (O0.1661*E4)+ (2.791*E 3) -(17.013*E2) +(55.795*E) -62.57 5 (-0.3014*E 4) + (4.6636*E 3) - (26.447*E 2 ) + (75.84*E) - 77.88 10 (-0.443*E 4) + (6.6239*E 3) - (36.324*E2) + (96.837*E) - 93.94 15 (-0.5782*E 4 ) + (84965*E23) - (45.758*E 2 ) + (116.88*E) - 109.27 20 + (10.457*E 3)_-(55.635*E2.) +(. - 125.3 Re~~on.?.pecay.Time,.Cur.ves:

Decay Time, Bumup, GWD/MTU Years

.5 ~~~~~~~~(6.Q.8934*7)(,5....580,,.,2)3*,E, + (,36. 1.2,,93, * ,E,) 46Q.0- 3 5 (q44*q W) ~ .(.?3..) - 40.3 10 (0.3634*_3) -- (4.1511*22) + (26.244*E) -34.7 15 (0.231 1*E3)_ ~(2.776*E 2) + (21.186*E) -29.0 20 (0.1011* E3 ) - (1.4265*E 2) + (16.19*E) - 23.4 ANO-1 B 3.7.15-1 Amendment No. 21-5,

Spent Fuel Pool Storage B 3.7.15 Region 1 and Reqion 2 Restricted Loading Pattern If restricted loadingq in either Region I or Region 2 storage cells is required by Figures 3.7.15-1 and/o'rj3.7.5-2, then a check..e..rb~oa"r'd'.. l-oa~d'in'..g...a"r'ra..ngqem'e'..n't' is required (alternate storage cells filled only with water or non-fuel bearing materials checkerboarde with cells containing ffu aseblies)..

Region I is comprised of two separate modules and Region 2 has four separate moueswthawter wit, ......s.. ..w....t, r.... qap p...esa,~ t.n paratiqqeac q...a..c h.. module . ..ut.f........ fromo . .........the..........

e.others. If it is. desired o h. .............. ........d........to l.oa .pt f.M due .ithecheckerboard loadin Dpattern and partof.the module without loading water or non-fuel bearing materials) must be maintained as a barrier between the two loadingt s. Th row of tclls serves as the water ap between the modulesto .prevent neutron interaction between the two. loadng patterns and thereby ensures proper criticality control.

If a checkerboard loading arrangement is not desirable, then the fuel assembly may

. ,tpr.e.d.. ..... . gi...n...

.3.

Repgion.3 Fr~es~h fuel assemblies with enrichments up to a maximum of 5 wt% U-235 orsWet fu .uel. b s b hst .e itnn . t elnR , goon3....w.h.t..r.s..tr a tu..o. t...

Required Soluble Boron Concentrations for Accident Conditions TS 3.7.14 includes the requirement for great than 1600 ppm boron concentration to assure the fuel assemblies will be maintained in a subcritical array with Kef, *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 Thespeificinat s ea a K.. ... ..0.95 ..... .will ......be ... maintained

.............. when ... ......the.. boron

...... ...concentration

.. .. ...... is at orabove .800 Dom.

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

ANO-1 B 3.7.15-1 Amendment No. 21-5,

Spent Fuel Pool Storage B 3.7.15 LCO The restrictions on the placement of fuel assemblies within the fuel pool, according to Figures 3.7.15-l.and.3.7.15-2, ensure that the kff of the spentfuel pool will always remain

<*0.95 assuming the pool to be flooded with 400 pm unborated water. The restrictions are consistent with the criticality safety analysis performed for the spent fuel pool. Fuel assemblies not meeting the enrichment and burnup criteria shall be stored in accordance with Specification 4.3.1.1.

The k~oT~be...8~f..o~f..t.,h.

h e,,,s~pe,.n~t,.f.u~e.

setfe pool .will remain. below 1.0 when the P.. pooi is flooded with

,p.o. ....i .!l...... .. .i.n ..b...o,.,........1 . w,.............. h~..... !..~

.. ..f..

..... . d..e..... w...i..

unborated water.

In the event a checkerboard storage configuration is deemed necessary for a portion of Region 1 or 2, spaces, void of fissionable material, adjacent to the faces of any fuel assembly which does not meet the Region 1 or 2 burnup criteria (non-restricted) shall be physially blockedconfirmed before any such fuel assembly may be placed in Region I or

2. This will prevent inadvertent fuel assembly insertion into two adjacent storage locations.

APPLICABILITY This LCO applies whenever any fuel assembly is stored in Region 1.or 2 of 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 Figure 3.7.15-1 or 3.7.15-2, immediate action must be taken to make the necessary fuel assembly movement(s) to bring the configuration into compliance with Figure 3.7.15-1, Figure.7.1.5-.2? or Specification 4.3.1.1.

SURVEILLANCE REQUIREMENTS SR 3.7.15.1 This SR verifies by administrative means that the initial enrichment, cooling time, and burnup of the fuel assembly is in accordance with Figure 3.7.15-1.lor.Figure 3.7.15-2 in the accompanying LCO or Specification 4.3.1.1. For fuel assemblies in the unacceptable range of Figure 3.7.15-1 .or.3.7.15-2, performance of the SR will ensure compliance with l Specification 4.3.1.1.

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

Spent Fuel Pool Storage B 3.7.15 REFERENCES

1. SAR, Section 9.6.2.

2. SER forANO-1 License Amendment No. 76, Section 2.1 (OCNA048314), dated April 15, 1983.

3. 10 CFR 50.36.

AN0-1 B 3.7.15-3 Amendment No 245,

Attachment 4 1CAN040302 Holtec License Report

i mm 0 m E ....... .. Holtc Center, 5565 Lncoln Drive West, Marlton, NJ 09053 furTelephone H I Y r (8566787-0900 H 0 L I a C Fax (856) 797-0900 INTERNATIONAL SPENT FUEL POOL RACKS MODIFICATIONS WITH POISON MATERIAL INSERTS IN ANO UNIT-1 FOR ENTERGY Holtec Report No: HI-2022867 Holtec Project No: 1196 Report Class: SAFETY RELATED

-!M

TABLE OF CONTENTS

1.0 INTRODUCTION

.1-1 1.1 References ...... , , 1-3 2.0 SPENT FUEL RACK FLUX TRAP GAP POISON INSERT DESIGN .................. 2-1 2.1 Introduction ........................... 2-1 2.2 Summary of Principal Design Criteria ........................... 2-1 2.3 Applicable Codes and Standards ....................... 2-2 2.4 Quality Assurance Program ....................... 2-9 2.5 Mechanical Design ....................... 2-9 2.6 Fabrication ....................... 2-11 2.7 Installation ....................... , 2-12 3.0 MATERIAL CONSIDERATIONS ....................... 3-1 31 Introduction ....................... 3-1 3.2 Structural Materials ................... 3-1 3.3 Neutron Absorbing Material .......................... 3-1 3.3.1 METAMICe Material Characteristics .......................... , 3-2 3.4 Compatibility with Environment ...................... 3-3 3.5 Heavy Load Considerations ...................... 3-3 3.6 References ...... 3-4 4.0 CRITICALITY SAFETY EVALUATION ........................... 4-1 4.1 Design Bases ...................... 4-1 4.2 Summary of Criticality Analyses ...................... 4-5 4.2.1 Normal Operating Conditions .................... , 4-5 4.2.1.1 Region I .................... 4-5 4.2.1.2 Region 2 ..... . .. ................................. 4-6 4.2.1.3 Region 3 .............................. ................................ 4........................-.......

-7 4.3 Reference Design Input Data ...................... 4-7 4.3.1 Reference Fuel Assembly .4-7 4.3.2 Region I Fuel Storage Cells ...................... 4-7 4.3.3 Region 2 Fuel Storage Cells .4-7 4.3.4 Region 3 Fuel Storage Cells .4-8 4.4 Analytical Methodology .4-8 4.4 1 Reference Design Calculations ................................ 4-8 4.4.2 Fuel Burnup Calculations and Uncertainties .4-9 4.4.3 Effect of Axial Burnup Distribution..................................................................... 4-9 4.4 4 BPRAs and APSRs ........................ 4-10 4.4.5 MCNP4a Temperature Correction ........................ , 4-10 4.4.6 Long-Term Changes in Reactivity ........................ 4-10 4.5 Region 1 Criticality Analyses and Tolerances . 4-11 4.5.1 Nominal Design Case ............... 4-11 4.5.2 Uncertainties Due to Tolerances ....................... 4-11 4.5.3 Eccentric Fuel Positioning. 4-11 4.6 Region 2 Criticality Analyses and Tolerances . 4-12 4.6.1 Nominal Design Case ....................... 4-12 4.6.2 Uncertainties Due to Tolerances .4-12 4.6.3 Eccentric Fuel Positioning ....................... 4-12 4.7 Region 3 Criticality Analyses and Tolerances...................................................4-12 4.7.1 Nominal Design Case . 4-12 4.7.2 Uncertainties Due to Tolerances......................................................................4-13 4.7.3 Eccentric Fuel Positioning . 4-13 TTl r1olux x -- Tconfl-

-_nReporL - -LZb file UzLbu / 1 1196

TABLE OF CONTENTS 4.8 Abnormal and Accident Conditions in the Spent Fuel Pool Racks .................... 4-13 4.8.1 Temperature and Water Density Effects ................................................ 4-13 4.8.2 Lateral Rack Movement .............................. 4-13 4.8.3 Abnormal Location of a Fuel Assembly .............................. 4-14 4.8.4 Dropped Fuel Assembly .............................. 4-14 4.9 Soluble Boron Dilution Evaluation.. ............................... ,.,.4-15 4.10 New Fuel Storage Racks Criticality Analysis .............................. 4-18 4.11 Fuel Handling Equipment ..................... 4-19 4.12 References ....... 4-20 5.0 THERMAL-HYDRAULIC CONSIDERATIONS ............ ..................... 5-1 5.1 Introduction .................... 5-1 5.2 Cooling Systems Description .................... 5-2 5.3 Spent Fuel Pool Decay Heat Loads ..................................... 5-3 5.4 Minimum Time-to-Boil and Maximum Boiloff Rate ..................................... 5-4 5.5 Maximum SFP Local Water Temperature ..................................... 5-5 5.6 Fuel Rod Cladding Temperature ....................................... 5-8 5.7 Results ..................................... 5-9 5.7.1 Decay Heat ..................................... 5-9 5.7.2 Minimum Time-to-Boil and Maximum Boiloff Rate ...................................... ,,.5-9 5.7.3 Local Water and Fuel Cladding Temperatures ..................................... 5-10 5.8 References ................................. 5-11 6.0 STRUCTURAIUSEISMIC CONSIDERATIONS ..................................... 6-1 6.1 Introduction ..................................... 6-1 6.2 Overview of Rack Structural Analysis Methodology ................... .................. 6-1 6 2.1 Background of Analysis Methodology ....................................... 6-2 6.3 Description of Racks ................. 6-5 6.4 Synthetic Time-Histories ................. 6-6 6.5 WPMR Methodology ......................... 6-6 6.5.1 Model Details for Spent Fuel Racks . 6-7 6.5.1.1 Assumptions ......................... 6-7 6.5.1.2 Element Details .......... 6-9 6.5.2 Fluid Coupling Effect ............................ 6-10 6.5.2.1 Multi-Body Fluid Coupling Phenomena ............................. ,6-11 6.5.3 Stiffness Element Details..........6......................................-................................11 6.5 4 Coefficients of Friction ...................... 6-12 6.5.5 Governing Equations of Motion.......................................................................6 13 6.6 Structural Evaluation of Spent Fuel Rack Design ......................................... 6-14 6.6.1 Kinematic and Stress Acceptance Criteria .............................. 6-14 6.6.2 Stress Limit Evaluations .............................. 6-15 6.6.3 Dimensionless Stress Factors .............................. 6-15 6.6.4 Loads and Loading Combinations for Spent Fuel Racks .............................. 6-19 6.7 Parametric Simulations................................................................................6-20 6.8 Time History Simulation Results ....................... 6-21 6.8.1 Rack Displacements ....................... 6-21 6.8.2 Pedestal Vertical Forces ... 6.....................,

6-22 6.8.3 Pedestal Friction Forces ................. , 6-22 6.8.4 Rack Impact Loads .................. 6-22 6.8.4.1 Rack to Rack Impacts ............... 6-22 6.8.4.2 Rack to Wall Impacts ...................... 6-23 6.8.4.3 Fuel to Cell Wall Impact Loads ...................... 6-23 6.9 Rack Structural Evaluation... 6-24 rvalue

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TABLE OF CONTENTS 6 9.1 Rack Stress Factors .................................. 6-24 6.9.2 Pedestal Thread Shear Stress .................................. 6-24 6.9.3 Local Stresses Due to Impacts .................................. 6-25 6.9.4 Weld Stresses .................................. 6-25 6.10 Poison Insert Structural Evaluation .................................. 6-28 6.11 Level A Evaluation .................................. 6-28 6.12 Hydrodynamic Loads on Pool Walls.................................................................6-28 6.13 Local Stress Considerations .................................. 6-28 6.13.1 Cell Wall Buckling............................................................................................6-29 6.13.2 Thermal loads .................................. 6-29 6.14 Evaluation of Postulated Stuck Fuel Assembly .................................. 6-29 6.15 Conclusion .................................. 6-29 6.16 References .......................... 6-30 7.0 MECHANICAL ACCIDENTS .......................... 7-1 7.1 Introduction .......................... 7-1 7.2 Description of Mechanical Accidents................................................................. 7-1 7.3 Evaluation of Mechanical Accidents .......................... 7-1 7.4 Conclusion .......................... 7-2 7.5 References .......................... 7-2 TJT-.i.. f b fY n %nTre *

"01tec report HI-ZUZZ416 I 111 1196

TABLE OF CONTENTS Tables 3.3.1 Chemical Composition and Physical Properties of Aluminum (6061 Alloy) .............. ..... 3-5 3.3.2 Chemical Composition and Physical Properties of Boron Carbide ............ .................... 3-6 4.1.1 Fuel Assembly Specifications .................................................................. 4-22 4.2.1 Summary of Criticality Safety Analyses for Storage of Spent Fuel Assemblies in Region I Racks .......... 4-23 4.2.2 Minimum Bumup Required for Storage of Spent Fuel Assemblies in the Region 1 Racks ... 2...............................

424 4.2.3 Bounding Polynomial Fits to Determine Minimum Acceptable Burnup for Storage of Spent Fuel Assemblies Storage in Region I Racks as a Function of Initial Enrichment ...................................................... 4-25 4.2.4 Summary of the Criticality Safety Analyses for 2 of 4 Checkerboard Storage of Fresh Fuel Assemblies and Empty Cells in Region I Racks .................................................. 4-26 4.2.5 Summary of the Criticality Safety Analyses for Storage of Spent Fuel Assemblies in Region 2 Racks ......... 4-27 4.2.6 Minimum Burnup Required for Storage of Spent Fuel Assemblies in the Region 2 Racks ......... ...... 428 4.2.7 Bounding Polynomial Fits to Determine Minimum Acceptable Bumup for Storage of Spent Fuel Assemblies Storage in Region 2 Racks as a Function of Initial Enrichment ...... 4-29 4.2.8 Summary of the Criticality Safety Analyses for 2-of-4 Checkerboard Storage of Fresh Fuel Assemblies and Empty Cells in Region 2 Racks. , 4-30 4.2.9 Summary of the Criticality Analyses for Storage of Fresh Fuel Assemblies in ANO Unit I Region 3 Racks ...................................... 4-31 4.9.1 Required Soluble Boron Concentration in the SFP Water ............................................ 4-32 4.10.1 Summary of New Fuel Vault Criticality Safety Analysis ................................................ 4-33 5.3.1 Key Input Data for Decay Heat Computations..............................................................5-12 5 3.2 Offload Schedule ................................................................. 5-13 5.4.1 Key Input Data for T ime-To-Boil Evaluation ................................................... ............5-14 5.5.1 Key Input Data for Local Temperature Evaluation ........................................................ 5-15 5.7.1 Result of SFP Decay Heat Calculations ................................................................. 5-16 5.7.2 Results of Loss-of-Forced Cooling Evaluations ........................................................... 5-17 5.7.3 Results of Maximum Local Water and Fuel Cladding Temperature Evaluations ........ 5-18 6.2.1 Partial Listing of Fuel Rack Applications Using Dynarack ........................ 6-33 through 6-35 6.3.1 Rack Material Data (150 TF) (ASME - Section II, Part D).............................................. 6-36 6.4.1 Time-History Statistical Correlation Results ...............................................6-37 6.5.1 Degrees-of-Freedom ............................................... 6-38 6.9.1 Comparison of Bounding Calculated Stresses vs. Code Allowables at Impact Locations and at Welds ................ 6-39 noitec--

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TABLE OF CONTENTS Finures 1.1 Location of the Different Rack Types in the Spent Fuel Pool 2.5.1 Schematic of the Poison Insert Mechanism 2.5.2 Lead-in Device 4.1.1 3-Dimensional Plot of Minimum Fuel Bumups for Fuel in Unit 1 Region 1 for Enrichments and/or Cooling Times 4.1.2 3-Dimensional Plot of Minimum Fuel Bumups for Fuel in Unit 1 Region 2 for Enrichments and/or Cooling Times 4.3.1 A Cross-Sectional View of the Calculational Model Used for the Region I Rack Analysis 4.3.2 A Cross-Sectional View of the Calculational Model Used for the Region II Rack Analysis 4.10.1 Reactivity of the New Fuel Vault as a Function of Moderator Density (4.95% Enrichment) 4.10.2 Acceptable New Fuel Storage Vault Configuration for 4.95% Enrichment Fresh Fuel 4.10.3 Acceptable New Fuel Storage Vault Configuration for 4.20% Enrichment Fresh Fuel 5.5.1 Schematic of the CFD Model of the ANO-1 SFP 5.7.1 Partial Core Offload Bounding Spent Fuel Heat Load 5.7.2 Full Core Offload Bounding Spent Fuel Heat Load 5.7.3 Contours of Static Temperature in a Vertical Plane Through the Center of the SFP 5.7.4 Velocity Vector Plot in a Vertical Plane Through the Center of the SFP 6.1.1 ANO-I Spent Fuel Pool Layout 6.4.1 ANO Unit 1, DBE, 2% Damping, 362' Elevation E-W Time Acceleration 6.4.2 ANO Unit 1, DBE, 2% Damping, 362' Elevation N-S Time Acceleration 6.4.3 ANO Unit 1, DBE, 2% Damping, 362' Elevation Vert. Time Acceleration 6.4.4 ANO Unit 1, OBE, 2% Damping, 362' Elevation E-W Time Acceleration 6.4.5 ANO Unit 1, OBE, 2% Damping, 362' Elevation N-S Time Acceleration 6,4.6 ANO Unit 1, OBE, 2% Damping, 362' Elevation Vert. Time Acceleration 6.5.1 Schematic of the Single Rack Dynamic Model Used in Dynarack 6.5.2 Fuel-to-Rack Gap/Impact Elements at Level of Rattling Mass 6.5.3 Two Dimensional View of the Spring-Mass Simulation 6.5.4 Shear and Bending Springs Representing Rack Elasticity in X-Z and Y-Z Planes 6.5.5 Rack Periphery Gap/impact Elements 6.11.1 Maximum Instantaneous Hydrodynamic Pressures Holtec Report HI-2022867 V 1196

1.0 INTRODUCTION

Arkansas Nuclear One, Unit 1 (ANO-1), operated by Entergy Operations, is located approximately 70 miles northwest of Little Rock, Arkansas and about five miles west of Russeliville. ANO-1 is a Babcock

& Wilcox pressurized water reactor and has been in commercial operation since 1974. The ANO-1 reactor is licensed for a thermal power level of 2568 megawatts The reactor core contains 177 fuel assemblies and the spent fuel pool (SFP) is licensed for the storage of 968 assemblies.

The racks in the SFP of ANO-1 are free-standing and self supporting racks of Westinghouse design.

The principal fabrication materials are ASTM A-240, Type 304 stainless steel for the structural members and shapes. "Boraflex', a product of BISCO (a division of Brand, Inc.) was originally used to augment reactivity control.

The ANO-1 SFP was designed to hold spent fuel assemblies (or rod cluster control assemblies) in underwater storage for long-term decay after their removal from the reactor core. The structures are seismic Category 1, heavy walled, reinforced concrete pool, located on grade outside the containment structure. The interior of the pool is lined with stainless steel plate (Type 304L).

The ANO-1 spent fuel racks consist of individual cells with a square pitch of 10.65 inches, each of which accommodates a single B&W 15x15 fuel assembly or equivalent. The ANO-1 SFP is divided into two regions, designated Region 1 and Region 2. Region 1 racks employ Boraflex as the poison material and are presently qualified to store fresh fuel assemblies with enrichments up to 4.1 weight percent (wt%)

235U. Region 2 racks are designed with flux-traps and are currently used to store spent fuel assemblies with various initial enrichments that have accumulated certain minimum burn-ups. These racks do not have any poison material. Some of the Region 2 racks will be modified by the insertion of Metamic absorber panels into the flux trap region to create a Region 3. These different regions are depicted in Figure 1-1. These poison inserts will have two borated Aluminum (Metamice) panels as neutron absorbers. Each poison insert panel will be held in the flux trap along the cell wall by a spring mechanism. The insertion of the Metamic poison panels into the new region, as shown by analyses later in this report, will enable storage of fresh fuel with a maximum enrichment up to 5.0 weight percent (wt%) in the ANO-1 SFP. The Region 3 flux traps will be fitted with lead-ins on the top of the flux traps, which will act to prevent any possible uplifting of the poison panel insert. The lead-in devices will also help guide the fuel assemblies into the storage cells.

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The existing Region 1 racks have been reanalyzed to establish new fuel storage requirements without crediting the presence of Boraflex. The Region 2 racks, which will not be converted to Region 3 racks, were re-analyzed to establish more flexible fuel storage requirements. These racks have been re-analyzed to establish their capability for fresh fuel storage in a 2-of-4 checkerboard arrangement or to store spent fuel assemblies of specified enrichment-burnup limits. The New Fuel Vault and fuel handling equipment is also analyzed to confirm acceptability of fuel with initial enrichment up to 5.0 wt% 235U.

Sections 2.0 and 3.0 of this report provide an abstract of the design and material information on the poison inserts and the racks.

Section 4.0 provides a summary of the methods and results of the criticality evaluations performed for the spent fuel storage racks. Credit for soluble boron in the pool has been taken, in accordance with 10CFR50.68, to assure the criticality safety of the spent fuel storage racks. The analyses show that the neutron multiplication factor (keff) for the stored fuel array are subcritical under assumed condition of the loss of all soluble boron in the pool water. Additional analyses are required to determine the soluble boron requirements to maintain kff below 0.95 for both normal storage and accident conditions. The criticality safety analysis sets the requirements on the Metamice poison insert panel length and the amount of B4 C (i.e., loading density) of the Metamice inserts for the Region 3 SFP racks.

Thermal-hydraulic considerations require that fuel cladding will not fail due to excessive thermal stress.

The thermal-hydraulic analyses carried out in support of the modification of some of the existing Region 2 racks are described in Section 5.0.

Rack module structural criteria require that the primary stresses in the rack module structure will remain below the ASME B&PV Code (Subsection NF) [1] allowables. Demonstrations of seismic and structural adequacy are presented in Section 6.0. The structural qualification also requires that the subcriticality of the stored fuel be maintained under all postulated mechanical accident scenarios. The structural consequences of these postulated accidents are evaluated and presented in Section 7.0 of this report.

Results of the analyses presented in this report establish acceptable restrictions on combinations of initial enrichment and discharge bumups for Region 1 and Region 2, as well as showing that the insertion of poison inserts into the newly defined Region 3 racks will permit storage of fresh fuel assemblies in these racks. The storage racks meet all requirements of the applicable USNRC guidelines and regulations, and applicable ANSI/ANS standards (References 2 - 6). The analysis Holtec Report HI-2022867 1-2 1196

methodologies employed are a direct evolution of previous license applications reviewed and approved by the USNRC, including nuclear subcriticality, thermal-hydraulic safety, seismic and structural adequacy, and mechanical integrity.

All computer programs utilized to perform the analyses documented in section 1.0 through 7.0 are benchmarked and verified. These programs have been utilized by Holtec International in numerous license applications over the past decade. The analyses presented herein clearly demonstrate that the rack module arrays with the addition of the poison inserts and the lead-ins possess wide margins of safety in respect to all considerations of safety specified in the OT Position Paper [3], namely, nuclear subcriticality, thermal-hydraulic safety, seismic and structural adequacy, and mechanical integrity.

1.1 REFERENCES

[11 American Society of Mechanical Engineers (ASME), Boiler & Pressure Vessel Code, Section 1I1, 1989 Edition, Subsection NF, and Appendices.

[21 General Design Criterion 62, Prevention of Criticality in Fuel Storage and Handling.

[3] USNRC, "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications, April 14, 1978, and Addendum dated January 18, 1979.

[4] Code of Federal Regulations 1DCFR50.68, Criticality Accident Requirements

[5] ANSI-8.17-1984, Criticality Safety Criteria for the Handling, Storage and Transportation of LWR Fuel Outside Reactors.

[6] L. I. Kopp, 'Guidance on the Regulatory Requirements for Criticality Analysis Of Fuel Storage At Light-Water Reactor Plants", USNRC Internal Memorandum L. Kopp to Timothy Collins, August 19 1998.

Holtec Report HI-2022867 1-3 1196

N Figure 1.1: Location of the Different Rack Types in the Spent Fuel Pool Wnotnp WIJafVVl-)mIn7 -

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2.0 SPENT FUEL RACK FLUX TRAP GAP POISON INSERT DESIGN

2.1 INTRODUCTION

The ANO-1 Spent Fuel Pool contains eight fuel racks with a total storage capacity of 968 fuel assemblies. As described in Section 1.0, there are currently two regions in the ANO-1 SFP that are designated as Region 1, which contains Boraflex, and Region 2, which does not contain poison inserts. A portion of Region 2 will be designated as Region 3 in which panels of Metamic containing a high loading of the B4 C (up to 25% by weight of B4C) will be inserted into the flux traps to provide appropriate neutron attenuation between adjacent storage cells. With the insert of MetamicO, it is proposed to allow the initial enrichment limit for fresh fuel storage in Region 3 racks to be 5.0 wt% (4.95 +/-0.05 wt%).

In addition to the poison inserts, the Region 3 racks will also be fitted with independent lead-in devices to help guide the fuel assemblies into the storage cells and prevent debris from entering the flux trap.

2.2

SUMMARY

OF PRINCIPAL DESIGN CRITERIA The key design criteria for the spent fuel racks are set forth in the USNRC memorandum entitled "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications,"

dated April 14, 1978 as modified by amendment dated January 18, 1979. The individual sections of this report expound on the specific design bases derived from the above-mentioned "OT Position Paper." The design bases for the spent fuel racks with the poison inserts in them are summarized below:

a. Kinematic Stability: All freestanding modules must be kinematically stable (against tipping or overturning) if a seismic event is imposed on any module.

b. Structural Compliance: All primary stresses in the rack modules must satisfy the limits postulated in Section III subsection NE of the ASME B & PV Code.

c. Thermal-Hvdraulic Compliance: The spatial average bulk pool temperature is required to remain below 150 OF. No localized boiling is permitted.

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d. Criticality Compliance: The New Fuel Storage Racks (NFSR) and Spent Fuel Storage Racks (SFSR) must be able to store Zircaloy dad fuel of 5.0 weight percent maximum enrichment while maintaining the reactivity (kff) less than 0.95.

e. Accident Events- In the event of postulated drop events (uncontrolled lowering of a fuel assembly, for instance), it is necessary to demonstrate that the stored fuel remains subcritical.

The foregoing design bases are further articulated in Sections 4.0 through 7.0 of this report.

2.3 APPLICABLE CODES AND STANDARDS The following codes, standards and practices are used as applicable for the design, construction, and assembly of the poison inserts. Additional specific references related to detailed analyses are given in each section.

a. Desicjn Codes (1) ASME B & PV Code Section 1I1,1998 Edition.

(2) American Society for Nondestructive Testing SNT-TC-1A, June 1980, Recommended Practice for Personnel Qualifications and Certification in Non-destructive Testing.

(3) ASME Y14.5M, Dimensioning and Tolerancing (4) ASME B & PV Code, Section lI-Part D, 1998 Edition.

b. Standards of American Society for Testing and Materials (ASTM)

(1) ASTM A240 - Standard Specification for Heat-Resisting Chromium and Chromium-Nickel Stainless Steel Plate, Sheet and Strip for Pressure Vessels.

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(2) ASTM A262 - Standard Practices for Detecting Susceptibility to Intergranular Attack in Austenitic Stainless Steel.

(3) ASTM C750 - Standard Specification for Nuclear-Grade Boron Carbide Powder.

(4) ASTM A380 - Standard Practice for Cleaning, Descaling, and Passivation of Stainless Steel Parts, Equipment and Systems.

(5) ASTM C992 - Standard Specification for Boron-Based Neutron Absorbing Material Systems for Use in Nuclear Spent Fuel Storage Racks.

(6) ASTM E3 - Standard Practice for Preparation of Metallographic Specimens.

(7) ASTM E190 - Standard Test Method for Guided Bend Test for Ductility of Welds.

c. Welding Code:

(1) ASME B & PV Code,Section IX - Welding and Brazing Qualifications, latest applicable edition and addenda.

d. Quality Assurance, Cleanliness, Packaging, Shipping. Receiving Storage.

and Handlinq (1) ANSI N45.2.1 - Cleaning of Fluid Systems and Associated Components during Construction Phase of Nuclear Power Plants - 1980 (R.G. 1.37)

(2) ANSI N45.2.2 - Packaging, Shipping, Receiving, Storage and Handling of Items for Nuclear Power Plants - 1978 (R.G. 1.38).

(3) ANSI N45.2.6 - Qualifications of Inspection, Examination, and Testing 1oiec Xeo'

_I2226 2-3 holtec Report HI-2022867 2-3 1196

Personnel for the Construction Phase of Nuclear Power Plants - 1973.

(R.G. 1.58).

(4) ANSI N45.2.8 - Supplementary Quality Assurance Requirements for Installation, Inspection and Testing of Mechanical Equipment and Systems for the Construction Phase of Nuclear Plants - 1975 (R.G. 1.116).

(5) ANSI N45.2.11 - Quality Assurance Requirements for the Design of Nuclear Power Plants - 1978 (R.G. 1.64).

(6) ANSI N45.2.12 - Requirements for Auditing of Quality Assurance Programs for Nuclear Power Plants - 1977 (R.G 1.144).

(7) ANSI N45.2.13 - Quality Assurance Requirements for Control of Procurement of Items and Services for Nuclear Power Plants - 1976 (R.G.

1.123).

(8) ANSI N45.2.23 - Qualification of Quality Assurance Program Audit Personnel for Nuclear Power Plants - 1978 (R.G. 1.146).

(9) ASME B & PV Code,Section V, Nondestructive Examination, 1983 Edition.

(10) ANSI N16.9 Validation of Calculation Methods for Nuclear Criticality Safety.

(11) ASME NQA-1 - Quality Assurance Program Requirements for Nuclear Facilities.

(12) ASME NQA-2 - Quality Assurance Requirements for Nuclear Power Plants.

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e. USNRC Documents (1) 'OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," dated April 14, 1978, and the modifications to this document of January 18, 1979.

(2) NUREG 0612, "Control of Heavy Loads at Nuclear Power Plants", USNRC, Washington, D.C., July, 1980.

(3) NUREG 0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants", USNRC, Washington, D.C., July, 1981.

f. Other ANSI Standards (not listed in the preceding)

(1) ANSI/ANS 8.1 - Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors, 1983.

(2) ANSI N45.2.9 - Requirements for Collection, Storage and Maintenance of Quality Assurance Records for Nuclear Power Plants - 1974.

(3) ANSI N45.2.10 - Quality Assurance Terms and Definitions - 1973 (4) ANSI/ASME N626 Qualification and Duties of Specialized Professional Engineers, 1977.

9. Code-of-Federal Regulations (CFR)

(1) 10 CFR 20 - Standards for Protection Against Radiation.

(2) 10 CFR 21 - Reporting of Defects and Non-compliance.

(3) 10 CFR 50 Appendix A - General Design Criteria for Nuclear Power Plants.

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(4) 10 CFR 50 Appendix B - Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants.

(5) 10 CFR 100 - Reactor Site Criteria

h. Reaulatorv Guides (RG)

(1) RG 1.13 - Spent Fuel Storage Facility Design Basis (Revision 2 Proposed).

(2) RG 1.25 - Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors, Rev. 0 -

March, 1972.

(3) RG 1.28 - Quality Assurance Program Requirements - Design and Construction, Rev. 2 - February, 1979 (endorses ANSI N45.2).

(4) RG 1.33 - Quality Assurance Program Requirements.

(5) RG 1.29 - Seismic Design Classification, Rev. 2 - February, 1976.

(6) RG 1.31 - Control of Ferrito Content in Stainless Steel Weld Metal, Rev. 3.

(7) RG 1.38 - Quality Assurance Requirements for Packaging, Shipping, Receiving, Storage and Handling of Items for Water-Cooled Nuclear Power Plants, Rev. 2 - May, 1977 (endorses ANSI N45.2 2).

(8) RG 1.44 - Control of the Use of Sensitized Stainless Steel.

(9) RG 1.58 - Qualification of Nuclear Power Plant Inspection, Examination, and Testing Personnel, Rev. I - September 1980 (endorses ANSI N45.2.6)

Holtec Report HI-2022867 2-6 1196

(10) RG 1.60 - Design Response Spectra for Seismic Design of Nuclear Power Plants (11) RG 1.61 - Damping Values for Seismic Design of Nuclear Power Plants, Rev. 0, 1973.

(12) RG 1.64 - Quality Assurance Requirements for the Design of Nuclear Power Plants, Rev. 2 - June, 1976 (endorses ANSI N45.2.11).

(13) RG 1.71 - Welder Qualifications for Areas of Limited Accessibility.

(14) RG 1.74 - Quality Assurance Terms and Definitions, Rev. 2 - February, 1974 (endorses ANSI N45.2.10).

(15) RG 1.85 - Materials Code Case Acceptability - ASME Section III, Division 1.

(16) RG 1.88 - Collection, Storage and Maintenance of Nuclear Power Plant Quality Assurance Records, Rev. 2 - October, 1976 (endorses ANSI N45.2.9).

(17) RG 1.92 - Combining Modal Responses and Spatial Components in Seismic Response Analysis, Rev. 1 - February, 1976.

(18) RG 1.116 - Quality Assurance Requirements for Installation, Inspection and Testing of Mechanical Equipment and Systems, Rev. 0-R - May,1977 (endorses ANSI N45.2.8-1975)

(19) RG 1.123 - Quality Assurance Requirements for Control of Procurement of Items and Services for Nuclear Power Plants, Rev. 1 - July, 1977 (endorses ANSI N45.2.13).

(20) RG 1.124 - Service Limits and Loading Combinations for Class 1 Linear-Type Component Supports, Rev. 1 - January,1978.

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(21) RG 1.144 - Auditing of Quality Assurance Programs for Nuclear Power Plants, Rev.1 - September, 1980 (endorses ANSI N45.2.12-1977)

(22) RG 8.8 - Information Relative to Ensuring that Occupational Radiation Exposures at Nuclear Power Stations will be as Low as Reasonably Achievable (ALARA).

(23) IE Information Notice 83 Fuel Binding Caused by Fuel Rack Deformation.

(24) RG 8.38 - Control of Access to High and Very High Radiation Areas in Nuclear Power Plants, June, 1993.

Branch Technical Position (1) CPB 9.1 Criticality In Fuel Storage Facilities.

(2) ASB 9 Residual Decay Energy for Light-Water Reactors for Long-Term Cooling - November, 1975.

j American Welding Societv (AWS) Standards (1) AWS D1.1 - Structural Welding Code - Steel.

(2) AWS D1.3 - Structure Welding Code - Sheet Steel.

(3) AWS D9.1 - Sheet Metal Welding Code.

(4) AWS A2.4 - Standard Symbols for Welding, Brazing and Nondestructive Examination.

(5) AWS A3.0 - Standard Welding Terms and Definitions.

Holtec Report HJ-2022867 2-8 1196

(6) AWS A5.12 - Specification for Tungsten and Tungsten Alloy Electrodes for Arc-Welding and Cutting (7) AWS CC1 - Standard for AWS Certification of Welding Inspectors.

(8) AWS 5.4 - Specification for Stainless Steel Electrodes for Shielded Metal Arc Welding.

(9) AWS 5 9 - Specification for Bare Stainless Steel Welding Electrodes and Rods.

k. Other References (1) ANO Unit 1 & 2 Operating Licenses and Technical Specifications, License No. DPR-51 & NPF-6.

(2) ANO Unit 1 & 2 Updated Final Safety Analysis Report (UFSAR).

2.4 QUALITY ASSURANCE PROGRAM The governing quality assurance requirements for design and fabrication of the poison inserts are stated in 10 CFR 50 Appendix B. Holtec's Nuclear Quality Assurance program complies with this regulation and Is designed to provide a system for the design, analysis, and licensing of customized components in accordance with various codes, specifications, and regulatory requirements.

The Quality Assurance System that will be used by Entergy Operations to install the poison inserts is controlled by the AND-1 Quality Assurance Program.

2.5 MECHANICAL DESIGN The mechanical design of the poison insert consists of two poison panels separated by a mechanism to maintain the water gap specified by criticality considerations. The poison panels r-1UJLVU MUPUR Jrjj-ZUZ4tjjj 2-9 1196

are independent flat panels sized to cover the active fuel region. The poison panels will extend all the way to the SFP rack base plate. The poison panel will be nominally 0.10 inch thick 6061 aluminum plus boron carbide metal matrix manufactured by Metamic.

The poison panels will be held together with a frame that is fabricated from SA240-304 stainless steel. A schematic of the arrangement is shown in Figure 2.5.1. The poison panels and the mechanical framework form the poison insert. The poison insert will be able to collapse smaller than the flux trap opening prior to installation. Figure 4.3.2 depicts the current Region 2 cell with four flux traps. The insert is designed to expand in the flux trap once fully inserted.

Figure 2.5.1 Schematic of the Poison Insert Mechanism The lead-in device, which is depicted in Figure 2.5.2, is fabricated from SA240-304L stainless steel. The device is designed to rest on top of the flux trap, and it is secured in place by two slotted plates, which straddle the cell wall at the corners external to the flux trap. The size and shape of the lead-in is such that it will not interfere with the square opening of the cell. Each lead-in device weighs less than 3 lb. The lead-in contains flow holes in the mounting plate to provide an uninterrupted flow path for the water entering at the bottom of the flux trap and exiting at the top of the flux trap.

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Figure 2.5.2 Lead-in Device 2.6 FABRICATION The object of this section is to provide a brief description of the poison insert construction activities, which enable an independent appraisal of the adequacy of design. The pertinent methods used in manufacturing the poison inserts may be stated as follows:

1. The poison panels are extruded and rolled from a powder metallurgy billet then cut to the specified rectangular size.

2. The fabrication process involves operational sequences that permit immediate accessibility for verification by the inspection staff.

3. The poison inserts are fabricated per the manufacturer's Appendix B Quality Assurance program, which ensures, and documents that the fabricated poison inserts meet all of the requirements of the design and fabrication documents.

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2.7 INSTALLATION The poison insert is placed in an upending cradle on the fuel bridge. The poison insert is then upended and connected to the poison insert handling tool. All installation activities will be performed remotely, from the fuel bridge, using long handled installation tools. Subsequent to the upending process, the poison insert is lowered into the spent fuel pool and guided into the appropriate flux trap with installation tools. Then, the lead-ins are installed onto a flux trap that received a poison insert.

Holtec Report HI-2022867 2-12 1196

3.0 MATERIAL CONSIDERATIONS

3.1 INTRODUCTION

Safe storage of nuclear fuel in the pool requires that the materials utilized in the poison inserts be of proven durability and compatible with the pool water environment. This section provides a synopsis of the considerations with regard to long-term design service life of 60 years.

3.2 STRUCTURAL MATERIALS The only structural material utilized in the fabrication of the poison inserts is SA240 Type 304 stainless steel.

3.3 NEUTRON ABSORBING MATERIAL In addition to the structural materials, the poison inserts employ Metamice, a patented product of Metamice, Inc., as the neutron absorber material A brief description of Metamice follows.

Metamice Is a neutron absorber material developed by the Reynolds Aluminum Company in the mid-1990s for spent fuel reactivity control in dry and wet storage applications [3.3.1]. Metallurgically, Metamice is a metal matrix composite (MMC) consisting of a high purity 6061 aluminum matrix reinforced with Type 1 ASTM C750-89, isotopically graded boron carbide (B4 C). MetamicO is characterized by an extremely fine aluminum spherical powder (325 mesh or better) and boron carbide powder (average particle size under 10 microns). The high performance reliability of Metamice derives from the particle size distribution of its constituents, namely, high purity aluminum 6061 alloy powder and isotopically graded B4C particulate, rendered into an isotropic metal matrix composite state by the powder metallurgy process which yields excellent homogeneity, and which prevents B4C from clustering in the final product.

The powders are carefully blended together without binders, chelating agents, or other additives that could potentially become retained in the final product and deleteriously influence performance. The maximum percentage of B4 C that will be dispersed in the aluminum alloy 6061 matrix is 25% by weight.

The pure blend of powders is cold isostatically compacted into a green billet and vacuum sintered to a U.,I14-- C3 -- LJO f'~~

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high theoretical density'. An extrusion process is used to bring the matrix into final density. Billets can vary in diameter, size and weight depending on a number of variables including loading and final panel dimensions.

Metamice has been subjected to an extensive array of tests sponsored by the Electric Power Research Institute (EPRI) that evaluated the functional performance of the material at elevated temperatures (up to 900 OF) and radiation levels (1E+11 rads gamma). The results of the tests documented in an EPRI report (3.3.2] indicate that Metamico maintains its physical and neutron absorption properties with little variation in its properties from the unirradiated state. The main conclusions provided in the above-referenced EPRI report are summarized below:

The isotropic metal matrix configuration produced by the powder metallurgy process with a complete absence of interconnected internal porosity in Metamice ensures that its density is essentially equal to the maximum theoretical density.

Measurements of boron carbide particle distnbution show extremely small particle-to-particle distance2 and near-perfect homogeneity.

The physical and neutronic properties of Metamice are essentially unaltered under exposure to elevated temperatures (750 'F - 900 'F).

I No detectable change in the neutron attenuation characteristics under accelerated test conditions has been observed.

Holtec International's Q.A. program ensures that Metamico is manufactured under the control and surveillance of a Quality Assurance/Quality Control Program that conforms to the requirements of 10 CFR 50 Appendix B, "Quality Assurance Criteria for Nuclear Power Plants".

3.3 1 METAMIC" Material Characteristics Aluminum: Aluminum is a silvery-white, ductile metallic element. The 6061 alloy aluminum is used 1 The theoretical density of Metamico, before being hot worked, is 82% to 98% depending on the B C content.

4 2 Medium measured neighbor-to-neighbor distance is 10.08 microns according to the article, 'METAMIC Neutron Shielding [3.3.31.

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extensively in heat exchangers, pressure and storage tanks, chemical equipment, reflectors, and sheet metal work.

It has high resistance to corrosion in industrial and marine atmospheres. Aluminum has atomic number of 13, atomic weight of 26.98, specific gravity of 2.69 and valence of 3. The physical, mechanical and chemical properties of the 6061 alloy aluminum are listed in Table 3.3.1.

The excellent corrosion resistance of the 6061 alloy aluminum is provided by the protective oxide film that quickly develops on its surface from exposure to the atmosphere or water. This film prevents the loss of metal from general corrosion or pitting corrosion.

Boron Carbide: The boron carbide contained in Metamice is a fine granulated powder that conforms to ASTM C750-89 nuclear grade Type 1. The material conforms to the chemical composition and properties listed in Table 3.3.2.

References [3.3.1] and [3.3.2] provide rurther discussion as to the suitability of these materials for use in spent fuel storage applications.

3.4 COMPATIBILITY WITH ENVIRONMENT All materials used in the construction of the poison inserts have been determined to be compatible with the ANO Spent Fuel Pools, and have an established history of in-pool usage. Austenitic stainless steel (e.g., Type 304) is a widely used stainless alloy in nuclear power plants, and it has an established history of in-pool usage. Metamic is likewise an excellent material for spent fuel applications based on its high resistance to corrosion and its functional performance at elevated temperatures and radiation levels.

3.5 HEAVY LOAD CONSIDERATIONS There are no heavy loads involved in the proposed installation of poison inserts. The estimated weight of a single poison insert is less than 40 pounds LlzAl- Da Id - -- ^ -n -

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3.6 REFERENCES

[3.3.1] "Use of METAMIC in Fuel Pool Applications," Holtec International, HI-2022871, Revision 1, August 2002.

[3.3.2] 'Qualification of METAMIC for Spent Fuel Storage Application," EPRI, 1003137, Final Report, October 2001.

13.3.3] K Anderson et al., WMETAMIC Neutron Shielding,' EPRI Boraflex Conference, November 19-20, 1998.

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Table 3.3.1 Chemical Composition and Physical Properties of Aluminum (6061 Alloy)

Chemical Composition 0.8-1.2% Magnesium 0.40-0.8% Silicone 0.15-0.40% Copper 0.15% max. Iron 0.25% max. Zinc 0.15% max. Titanium 50 ppm max. Nickel 10 ppm max. Cobalt 10 ppm max. Manganese 10 ppm max. Chromium 0.15% max. Other Remainder Aluminum Physical Properties Density 0.098 lblin3 2.71 g/cm 3 Melting Range 1080 'F - 1205 OF 582 'C- 652 0C Thermal Conductivity (77 OF) 1250 BTU/hr-fl 2 -OF/in

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Table 3.3.2 Chemical Composition and Physical Properties of Boron Carbide _ .

Chemical Composition (Weight Percent)

Total boron 76.5 min.

B10 isotope 19.9 +/- 0.30 alo HN 3 soluble

. boron 0.5 max.

Water soluble boron 0.2 max.

Fluoride 25 pglg max.

Chloride 75 pglg max.

Calcium 0.3 max.

Iron 1.0 max.

Total boron plus total carbon 98.0 min.

Physical Properties Chemical formula 04C Boron content (weight percent) 78.28%

Carbon content (weight percent) 21.72%

Crystal structure rhombohedral Density 0.0907 lb/in3

_ 2.51 gfcm3 Melting Point 4442 'F 2450 0 C Boiling Point 6332 OF 3500 DC l -- - - 3-6PA_ --- 11RE6 nVILlCL meporl rIl-zuzzoo/ 3-6 1196

4.0 CRITICALITY SAFETY EVALUATION 4.1 DESIGN BASES This section of the report documents the criticality safety evaluation for the storage of fresh and spent nuclear fuel assemblies in the ANO-1 high-density spent fuel storage racks. ANO-1 spent fuel pool currently has two regions of storage, which are currently licensed to store fuel assemblies with a maximum enrichment of 4.1 wt%.

1. Region I racks: These racks were originally designed with Boraflex as the poison material in a flux-trap configuration.

2. Region 2 racks: These racks are designed to store spent fuel assemblies of a specified combination of initial enrichment and discharge bumup. These racks do not currently have any poison material within the water gap between cells.

Due to the Boraflex degradation in the Region 1 racks, future credit for Boraflex is not feasible in these racks. The proposed resolution is to re-evaluate the criticality safety of the racks without credit for Boraflex and to insert poison material strips into the flux trap region of some of the Region 2 type racks.

These modified Region 2 racks are identified as Region 3 racks. The new Region 3 racks will enable unrestricted fresh fuel (maximum enrichment of 5.0 wt% or nominally 4.95 +/-0.05 wt%) storage capability in that region. The calculations are performed with no credit for Boraflex in the Region 1 racks and with poison inserts in the Region 3 racks All racks, including the remaining Region 2 racks were re-evaluated under the provisions of 10 CFR 50.68.

Specifically, the following evaluations were performed for ANO-1:

The Region I racks were evaluated for storage of spent fuel assemblies with specific bumup requirements for the spent fuel assemblies, as a function of initial enrichments and decay times (up to 20 years). Results are summarized In Figure 4.1.1 and tabulated in Table 4.2.1, 4.2.2, and 4.2.3.

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Fresh fuel storage in Region I has been assessed based in a "2-of-4" checkerboard loading with empty storage cells. (i.e., filled only with water or non-fuel bearing materials). Results are shown in Table 4.2.4 for fuel of 4.95 _0.05 wt% 2 35U enrichment.

Region 2 racks were evaluated for storage of spent fuel assemblies with burnup requirements for the spent fuel assemblies, determined as a function of Initial enrichments and decay times (up to 20 years). Results are summarized in Figure 4.1.2 and Tables 4.2.5, 4.2.6, and 4.2.7.

Fresh fuel storage in Region 2 has been assessed based on a "2- of -4" checkerboard loading with empty storage cells (i.e., filled only with water or non-fuel bearing materials). Results are shown in Table 4.2.8 for fuel assemblies of 4.95 +/-0.05 wt% 2 35U enrichment.

The new Region 3 racks were evaluated with Metamice panels inserted in the water gap, for storage of fresh fuel assemblies, with 235U enrichments up to 4.95 +/-0.05 wt%. Results are shown in Table 4.2.9. Region 3 may also accommodate spent fuel of any bumup for fuel assemblies up to 4.95 +/-0.05 wt% 235U enrichment.

The racks have been evaluated for Babcox & Wilcox (B&W) 15x15 spent and fresh fuel assemblies with an initial average uniform enrichment up to 5.0 wt% 235U (4.95

+/-0.05 wt%). Credit is taken for poison inserts, fuel burnup, cooling time, and soluble boron in pool water as applicable per 10 CFR 50.68 and Reference 4.1.2.

The objective of this analysis is to ensure, per 10 CFR 50.68, that the racks shall remain subcritical under normal conditions with no credit for soluble boron and less than or equal to 0.95 when partial credit is taken for soluble boron in pool water, including calculation uncertainties and effects of mechanical tolerances. Reactivity effects of abnormal and accident conditions have also been evaluated to determine the required soluble boron concentration in the pool to assure that under all credible abnormal and accident conditions, the maximum reactivity will not exceed the regulatory limit of 0.95. The required soluble boron concentrations are summarized in Table 4.9.1. In this context "abnormal" refers to conditions, which may reasonably be expected to occur during the lifetime of the plant and "accident" refers to conditions, which are not expected to occur but nevertheless must be protected against. The double contingency principle [4.1.2] allows full credit for soluble boron under other abnormal or accident conditions, since only a single independent accident need be considered at one time.

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Applicable codes, standards, and regulations or pertinent sections thereof, include the following

Code of Federal Regulations, Title 10, Part 50, Appendix A, General Design Criterion 62, "Prevention of Criticality in Fuel Storage and Handling."

Code of Federal Regulations 10 CFR 50.68, "Criticality Accident Requirements'

USNRC Standard Review Plan, NUREG-0800, Section 9.1.2, 'Spent Fuel Storage," Rev. 1 - July 1981.

USNRC letter of April 14, 1978, to all Power Reactor Licensees - "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," including modification letter dated January 18, 1979.

L. 1. Kopp, 'Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants," NRC Memorandum from L. Kopp to T. Collins, August 19, 1998.

USNRC Regulatory Guide 1.13, "Spent Fuel Storage Facility Design Basis," Rev. 2 (proposed),

December 1981.

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

To assure the true reactivity will be less than the calculated reactivity, the following conservative design criteria and assumptions have been employed:

Criticality safety analyses were based upon an infinite radial array of cells; i.e., no credit was taken for radial neutron leakage, except for evaluating accident conditions along the rack outer boundary where neutron leakage is inherent.

Minor structural materials were neglected; i.e., spacer grids were conservatively assumed to be replaced by water.

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Because the temperature coefficient of reactivity is positive in the absence of neutron absorber panels, the analyses for Region 1 (no credit is taken in the current evaluations for the existing Boraflex poison material in these racks) and Region 2 (no poison material currently exists in these racks) type racks assumed a temperature of 150 cF. This is the design basis maximum pool water temperature. Higher temperatures would be an accident condition for which full soluble boron credit is permitted and the reactivity effects would be mitigated by the presence of the large amounts of soluble boron in the pool water.

For Region 3 type racks, the moderator is assumed to be un-borated water at a temperature within the operating range (4 0C or 39 'F) that results in highest reactivity. Criticality calculations were performed at 20 'C (68 °F) and temperatures below 20 'C (68 'F) were assumed to be abnormal events and the reactivity effects combined additively with other uncertainties.

Manufacturing tolerances of the Metamice neutron absorber (width, thickness and B4C loading) is included in the criticality safety evaluation. Steel components associated with the inserts are replaced with water in the analysis.

The analyses used a B&W 15x15 fuel assembly, with a maximum 235U enrichment of 4.95 +/-0.05 wt%.

No axial blankets were assumed to be present in the fuel rods. The entire active fuel length was assumed to be of uniform enrichment.

In-core depletion calculations have assumed conservative operating conditions, conservative fuel and moderator temperature, and an allowance for the average soluble boron concentrations during in-core operations.

The spent fuel storage racks are designed to accommodate the fuel assembly type listed in Table 4.1.1 with a maximum nominal initial enrichment of 4.95 +/-0.05 wt% 235U.

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4.2

SUMMARY

OF CRITICALITY ANALYSES 4.2.1 Normal Operating Conditions The criticality analyses for each of the three separate regions of the spent fuel storage pool for the design basis storage conditions are summarized in Tables 4.2.1 through 4.2.9. For the fuel acceptance criteria defined in the previous section, the maximum effective neutron multiplication factor (keff ) values are shown to be less than 1.0 (95% probability at the 95% confidence level) in each of the regions when no credit is taken for the presence of soluble boron in the pool. Credit for soluble boron is required to ensure koo is maintained less than 0.95. The required soluble boron concentrations are summarized in Table 4.9.1.

4.2.1.1 Region 1 The maximum keff values for storage of spent fuel were determined assuming an infinite radial array of storage cells with a finite axial length, water reflected. For each spent fuel cooling time, minimum bumup values were determined that assure the maximum ket, Including calculational and manufacturing uncertainties, remains subcritical under the assumed accident condition of the loss of all soluble boron.

Table 4.2.1 summarizes the results of these analyses at zero cooling time for spent fuel assemblies with an initial enrichment of 4.95 +/-0.05 wt% 235U. Figure 4.1.1 and Table 4.2.2 show the minimum acceptable bumup for storage of fuel assemblies of various initial enrichments and cooling times in the spent fuel. The calculated maximum reactivity includes the reactivity effect of the axial distribution in bumup and provides an additional margin for the uncertainty in the depletion calculations. The minimum soluble boron concentration required to maintain keff below 0.95, including all manufacturing and calculational tolerances, for the storage of spent fuel in the Region 1 racks, is 225 ppm.

For convenience, the minimum (limiting) burnup data shown in Table 4.2.2 may be described as a function of the nominal initial enrichment, E, in wt% 235U by a bounding polynomial expression as shown in Table 4.2.3 Fuel assemblies with enrichments less than 2.0 wt% 23sU will conservatively be required to meet the bumup requirements of 2.0 wt% 235U assemblies. Since the data is nearly linear, linear interpolation between the points listed in Table 4.2.2 is acceptable.

The maximum keff, for storage of fresh fuel assemblies of 4.95 +/-0.05 wt% initial enrichment in the Region 1 type racks in a 2-of-4 checkerboard pattern with the alternate cells remaining empty of fuel, is 0.9443. Table 4.2.4 summarizes the results of this analysis. Based on this result, this arrangement is HI-t., 0--4 LJI nnA'R7 X I I...L." 1%VU&I IIL-rUovU A- 1196

acceptable for storage of fresh fuel with no credit for soluble boron or for spent fuel regardless of burnup.

4.2.1.2 Region 2 The maximum keff values for storage of spent fuel were determined assuming an infinite radial array of storage cells with a finite axial length, water reflected. For each spent fuel cooling time, minimum burnup values were determined that assure the maximum keff, including calculational and manufacturing uncertainties, remains subcritical under the assumed accident condition of the loss of all soluble boron.

Table 4.2.5 summarizes the results of these analyses at zero cooling time for spent fuel assemblies with an initial enrichment of 4.95 +/-0.05 wt%. Figure 4.1.2 and Table 4.2.6 show the minimum acceptable bumup for storage of fuel assemblies of various initial enrichments and cooling times in the SFP. The calculated maximum reactivity includes the reactivity effect of the axial distribution in burnup and provides an additional margin for the uncertainty in the depletion calculations. The minimum soluble boron concentration required to maintain keff below 0.95, including all manufacturing and calculational tolerances, for the storage of spent fuel allowed in the Region 2 racks is 180 ppm.

For convenience, the minimum (limiting) bumup data shown in Table 4.2.6 may be described as a function of the nominal initial enrichment, E, in wt% 235U by a bounding polynomial expression as shown in Table 4.2.7. Fuel assemblies with enrichments less than 2.0 wt% 236U will conservatively be required to meet the bumup requirements of 2.0 wt% 235U assemblies. Since the data is nearly linear, linear interpolation between the points listed in Table 4.2.6 is acceptable.

The maximum keff, for storage of fresh unburned fuel assemblies of enrichment 4.95 +/-0.05 wt% in the Region 2 in a 2-of-4 checkerboard pattern with the alternate cells remaining empty of fuel, is 0.9621.

Table 4.2.8 summarizes the results of this analysis and confirms that this arrangement is acceptable for storage of fresh fuel or spent fuel regardless of burnup, without requiring any credit for soluble boron.

The minimum soluble boron concentration required to maintain keff below the 0.95 limit, including all manufacturing and calculational tolerances, for the storage of fresh fuel assemblies in this configuration in the Region 2 racks is 180 ppm.

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4.2.1.3 Region 3 The Region 3 racks were analyzed for the storage of 4.95 +/-0.05 wt% fresh fuel assemblies. The maximum keff, for storage of fresh fuel assemblies in the Region 3 racks is 0 9939. Table 4.2.9 summarizes the results of this analysis, and confirms that this arrangement is acceptable for storage of fresh unburned fuel or spent fuel regardless of burnup. The minimum soluble boron concentration required in Region 3 to maintain keff below 0.95, including all manufacturing and calculational tolerances, is 400 ppm.

4.3 REFERENCE DESIGN INPUT DATA 4.3.1 Reference Fuel Assembly The spent fuel storage racks are designed to accommodate B&W 15x15 fuel assemblies. The design specifications for the B&W fuel assembly designs, as used for this analysis, are given in Table 4.1.1.

4.3.2 Region 1 Fuel Storage Cells Figure 4.3.1 shows the calculational model of the nominal Region 1 spent fuel storage cell. The Region 1 storage cells are composed of stainless steel boxes separated by a gap. The 0.075 +/-0.004 thick steel walls define the storage cells, which have an 8.970 -0.05/+0.025 inch inside dimension. A 0.020 inch stainless steel sheath is around the gap and defines the boundary of the flux-trap water-gap used to augment reactivity control. The cells are located on a lattice spacing of 10.65 inches in both directions. Stainless steel channels connect the storage cells in a rigid structure and define the flux-trap of 1.31 inches, between the sheathing of adjacent cells.

4.3.3 Region 2 Fuel Storage Cells Figure 4.3.2 shows the calculational model of the nominal Region 2 spent fuel storage cell. The Region 2 storage cells area also have a flux trap between adjacent cells and are composed of stainless steel boxes separated by a gap. The straight portion of the flux trap Is 7.5 inches. The measured flux trap water gap of 1.456 +0.12/-0.05 inches was used in the analysis. The 0.062 +/-0.004 thick steel walls define the storage cells, which have an 8.97 +0.050/-0.025 inch nominal inside dimension. The measured value of the flux trap water gap corresponded to a Box inside dimension (ID) of 9.07 inches because of the bow in the cell walls. This value of the Box ID was usei in the analysis. The cells are L-a.t,..- Mfl.... I it , _

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located on a lattice spacing of 10.65 inches in both directions. No additional water gaps exist between adjacent Region 2 cells in a rack.

4.3.4 Region 3 Fuel Storage Cells The Region 3 storage cells are identical to Region 2 storage cells except that the Metamice Poison panels will be inserted into the flux trap gaps. The poison panels are designed to be 7.0 +/-0.0625 inches wide with a with a minimum Boron Carbide (84C) content of 25 weight percent. The criticality safety analysis conservatively assumes the poison panels are 6.7 inches wide. These Metamice panels are held by appropriate mechanisms to remain close to the straight portion of the flux trap walls.

4.4 ANALYTICAL METHODOLOGY 4.4.1 Reference Design Calculations The principal methods for the criticality analyses of the storage racks include the following codes:

(1) MCNP4a [4.4.1], (2) KENO Va [4.4.2], and (3) CASMO-4 [4.4.5-4.4.7]. MCNP4a is a continuous energy three-dimensional Monte Carlo code developed at the Los Alamos National Laboratory. KENO Va is a three-dimensional multi-group Monte Carlo code developed at the Oak Ridge National Laboratory as part of the SCALE 4.3 package [4.4.3]. The KENO Va calculations used the 238-group SCALE cross-section library and NITAWL [4.4.4] for 238U resonance shielding effects (Nordheim integral treatment). Benchmark calculations, presented in Appendix 4A, indicate a bias of 0.0009 with an uncertainty of +/-0.0011 for MCNP4a and 0.0030 +/-0.0012 for KENO Va, both evaluated with the 95%

probability at the 95% confidence level [4.1.1].

Fuel depletion analyses during core operation were performed with CASMO-4, two-dimensional multi-group transport theory codes based on transmission probabilities [4.4.5 - 4.4.71. Restarting the CASMO-4 calculations In the storage rack geometry yields the two-dimensional infinite multiplication factor (koo) for the storage rack. CASMO-4 was also used to determine the reactivity uncertainties (differential calculations) of manufacturing tolerances and the reactivity effects of variations in the water temperature and density.

In the geometric models used for the calculations, each fuel rod and its cladding were described explicitly and reflecting boundary conditions were used in the radial direction, which has the effect of wnitar- D~n.-4 wLnn,"aa,7

, n-vw Ius I~~l 4-0 1196

creating an infinite radial array of storage cells. Monte Carlo calculations inherently include a statistical uncertainty due to the random nature of neutron tracking. To minimize the statistical uncertainty of the MCNP4a and KENO Va calculated reactivity and to assure convergence, a minimum of 1 million neutron histories were accumulated in each calculation. Three-dimensional MCNP calculations were necessary to describe the geometry of the checkerboard cases. However, MCNP cannot perform depletion calculations, and depletion calculations were performed with CASMO-4. Explicit description of the fission product nuclide concentrations in the spent fuel was determined from the CASMO-4 calculations and used in the MCNP calculations. To compensate for those few fission product nuclides that are not in the MCNP library, an equivalent boron-10 concentration in the fuel was determined which produced the same reactivity in MCNP as the CASMO-4 result. This methodology explicitly incorporates approximately 40 of the most important fission products, accounting for all but about 1% in reactivity.

The remaining -1 % in reactivity is included by the equivalent B-10 concentration in the fuel.

4.4.2 Fuel Bumup Calculations and Uncertainties CASMO-4 was used for burnup calculations in the hot operating condition. Conservatively bounding moderator and fuel temperatures and the average operating soluble boron concentrations (900 ppm) were used to assure the high plutonium production and hence conservatively high values of reactivity.

Since critical experiment data with spent fuel is not available for determining the uncertainty in depletion calculations, an allowance for uncertainty in reactivity was assigned based upon other considerations 14.1.2]. Assuming the uncertainty in depletion calculations is less than 5% of the total reactivity decrement, a bumup dependent uncertainty in reactivity for bumup calculations was assigned. Thus, the burnup uncertainty varies (increases) with bumup. This allowance for burnup uncertainty was included in determination of the acceptable bumup versus enrichment combinations.

4.4.3 Effect of Axial Bumup Distribution Initially, fuel loaded into the reactor will bum with a slightly skewed cosine power distribution. As bumup progresses, the burnup distribution will tend to flatten, becoming more highly burned in the central regions than in the upper and lower regions. At high bumup, the more reactive fuel near the ends of the fuel assembly (less than average bumup) occurs in regions of high neutron leakage. Consequently, it is expected that over most of the bumup history, fuel assemblies with distributed burnups will exhibit a slightly lower reactivity than that calculated for the uniform average bumup. As bumup progresses, the distribution, to some extent, tends to be self-regulating as controlled by the axial power distribution, precluding the existence of large regions of significantly reduced burnup.

Holtec Report 2022867 4-9 1196

Among others, Turner [4.4.8] has provided generic analytic results of the axial bumup effect based upon calculated and measured axial burnup distributions. These analyses confirm the minor and generally negative reactivity effect of the axially distributed bumups at values less than about 30 GWD/MTU with small positive reactivity effects at higher bumup values. Calculations were performed based upon a bumup distribution provided by ANO. These calculations were performed in MCNP4a with 10 zone axial calculations, using specific (CASMO-4) concentrations of actinides and fission product nuclides in each zone. Results of these calculations, therefore, inherently include the effect of the axial distribution in burnup.

4.4.4 BPRAs and APSRs The fuel assemblies used at the ANO-1 Nuclear Power Plant contain poison rods such as the Burnable Poison Rods Assemblies (BPRAs) and Axial Power Shaping Rods (APSRs). These rods are inserted into the guide tubes during the in-core depletion process and are stored in fuel assemblies in the rack cells. Analyses show that the fuel assemblies, which contained the BPRAs during depletion in the core, are more reactive in the racks than those that contained APSRs or no burnable absorber in the bumup range of interest in this analysis. CASMO-4 calculations performed to obtain the isotopic compositions of the spent fuel assemblies, which were subsequently used In the MCNP calculations, were performed with BPRAs in the fuel assemblies during core operation.

4.4.5 MCNP4a Temperature Correction The reactivity for non-poisoned racks in the spent fuel pool increases with pool water temperature. The maximum bulk pool water temperature is 150 OF. However, since the Doppler treatment and cross-sections in MCNP4a are valid only at 20 IC (68 0F), the delta-k determined in CASMO-4 from 20 DC (68 0 F) to the limiting temperature described in section 4.8.1 is included in the final keff calculation.

4.4.6 Long-Term Changes in Reactivitv At reactor shutdown, the reactivity of the fuel initially decreases due to the growth of Xe-135.

Subsequently, the Xenon decays and the reactivity increase to a maximum at about a hundred hours when the Xenon is gone. Therefore, for conservatism, the Xe is set to zero in the calculations to assure maximum reactivity. During the next 50 years, the reactivity continuously decreases due primarily to LJ^d4+-- 10. UII nnl'V1O&'

..... t nluL IXVVU "A-cUd-49 4-1 1196

241 Pu 241Am decay and growth. Credit for this decay and for changes in fission product concentrations is included in calculations of the decrease in reactivity in long-term storage (up to 20 years). The CASMO-4 code includes the capability of tracking the decay of the actinides and the most significant product nuclides during long-term storage.

4.5 REGION 1 CRITICALITY ANALYSES AND TOLERANCES 4.5.1 Nominal Design Case For the nominal storage cell design in Region 1, the criticality safety analyses for the two different storage patterns are summarized in Tables 4.2.1, 4.2.2, and 4.2.4. This data confirms that the maximum reactivity in Region 1 remains subcritical (less than the regulatory limit keff S 1.0) under the assumed condition of the loss of all soluble boron in the pool water. Figure 4.1.1 shows the limiting burnup values for fuel of other enrichments and cooling times (see also Table 4.2.2).

4.5.2 Uncertainties Due to Tolerances The reactivity effects of manufacturing tolerances are summarized in Tables 4.2.1 and 4.2.4. All of the individual reactivity allowances were separately calculated for the reference fuel assembly and a statistical combination of uncertainties was used. The tolerances include the fuel storage area LD., rack material thickness, water gap, fuel enrichment and density.

4 5.3 Eccentric Fuel Positioning The fuel assembly is assumed to be normally located in the center of the storage rack cell. However, calculations were also made with the fuel assemblies assumed to be in the comer of the storage rack cell (four-assembly cluster at closest approach). These calculations indicated that the reactivity effect is slightly positive. Therefore, the uncertainty for eccentricity is included in the calculations for the final keff In Tables 4.2.1 and 4.2.4.

A I1U-14 - n_-L UlInO I uILV%- FNUPrl nl-dtV~foV I 4-11 1196

4.6 REGION 2 CRITICALITY ANALYSES AND TOLERANCES 4.6.1 Nominal Design Case For the nominal storage cell design in Region 2, the criticality safety analyses are summarized in Tables 4.2.5, 4.2.6, and 4.2.8. This data confirms that the maximum reactivity in Region 2 remains subcritical (less than the regulatory limit keff < 1.0) under the assumed condition of the loss of all soluble boron in the pool water. Figure 4.1.2 (and table 4 2.6) summarizes the limiting fuel burnups for fuel assemblies of other enrichments and cooling times.

4.6.2 Uncertainties Due to Tolerances The reactivity effects of manufacturing tolerances are summarized in Tables 4.2.5 and 4.2.8. All of the individual reactivity allowances were separately calculated for the reference fuel assembly and a statistical combination of uncertainties was used. The tolerances include the fuel storage area I.D, rack material thickness, water gap, fuel enrichment and density.

4.6.3 Eccentric Fuel Positioning The fuel assembly is assumed to be normally located in the center of the storage rack cell. However, calculations were also made with the fuel assemblies assumed to be in the comer of the storage rack cell (four-assembly cluster at closest approach). These calculations indicate that the reactivity effect is slightly positive. Therefore, the uncertainty for eccentricity is included in the calculations for the final keff in Tables 4.2.5 and 4.2.8.

4.7 REGION 3 CRITICALITY ANALYSES AND TOLERANCES 4.7.1 Nominal Design Case For the nominal storage cell design in Region 3, the criticality safety analyses are summarized in Table 4.2.9. This data confirms that the maximum reactivity in Region 3 remains subcritical (less than the regulatory limit keff 5 1.0) under the assumed condition of the loss of all soluble boron in the pool water.

-E LIE fle. .

nuIVLtnc rsUpul t nl-uo 1-U40 4-12 1196

4.7.2 Uncertainties Due to Tolerances The reactivity effects of manufacturing tolerances are tabulated in Table 4.2,9. All of the individual reactivity allowances were separately calculated for the reference fuel assembly and a statistical combination of uncertainties was used. The tolerances include the fuel storage area l.D., rack material thickness, water gap, poison panel width, poison panel thickness, boron loading in the poison panel, fuel enrichment, and density.

4.7.3 Eccentric Fuel Positioning The fuel assembly is assumed to be normally located in the center of the storage rack cell. However, calculations were also made with the fuel assemblies assumed to be in the comer of the storage rack cell (four-assembly cluster at closest approach). These calculations indicate that the reactivity effect is slightly negative, so it is conservatively not included.

4.8 ABNORMAL AND ACCIDENT CONDITIONS IN THE SPENT FUEL POOL RACKS 4 8.1 Temperature and Water Density Effects The moderator temperature coefficient of reactivity in both Region 1 and Region 2 is positive. A moderator temperature of 20 'C (68 'F) was assumed for the reference MCNP4a calculations and the increase in reactivity to the maximum permissible bulk pool water temperature of 150 0 F is included (CASMO-4 calculation) as a bias in the calculation of the maximum keff. This assures that the true reactivity will always be lower over the expected range of water temperatures. The reactivity effects of the pool water temperature effects on reactivity have been evaluated using CASMO-4.

The moderator temperature coefficient of reactivity for the Region 3 racks is negative and, therefore, the reference MCNP4a calculation provides the maximum reactivity. The effect of a reduction in temperature to 4 OC (39 0F) is included as an uncertainty in Table 4.2.9.

4.8.2 Lateral Rack Movement Lateral motion of the storage racks under postulated seismic conditions could potentially alter the spacing between racks. Under these conditions, credit for the soluble boron (permitted under accident Holtec Report HI-2022867 4-13 1196

conditions) would maintain the keff at a value well below the maximum allowable. The double contingency principle requires the consideration of only one accident at one time. Nevertheless, the separation (water-gap) between rack modules is sufficiently large that even for the maximum movement expected under seismic excitation, the water gap remains larger than the water gap within the Region 1 modules. In the Region 2 and the Region 3 racks, the keff Is independent of the inter-module water gap and is not sensitive to any potential seismic induced movement of the modules. The water-gap structure in each cell is included in the analysis and precludes any closer proximity between modules.

4.8.3 Abnormal Location of a Fuel Assembly The misplacement of a fresh unburned fuel assembly of the highest permissible reactivity, would, in the absence of soluble poison, result in exceeding the regulatory limit (keff

1.0). This could occur if a fresh fuel assembly of the highest permissible initial enrichment (4.95 +/-0.05 wt% 2 35U) were to be inadvertently loaded into a Region I or Region 2 storage cell, which are Intended to store spent fuel assemblies or remain empty. Calculations confirmed that the highest reactivity, including uncertainties, for the worst case postulated fuel mis-loading accident condition (fresh fuel assembly in Region 2 cell intended to remain empty) would exceed the limit on reactivity in the absence of soluble boron. Soluble boron in the spent fuel pool water, for which credit is permitted under these accident conditions, would assure that the reactivity is maintained substantially less than the design limitation. Calculations indicate that a soluble boron concentration of 800 ppm is adequate to assure that the maximum keff does not exceed 0.95.

In addition, the mislocation of a fresh unburned fuel assembly could occur if a fresh fuel assembly of the highest permissible initial enrichment (4.95 +/-0.05 wt%) were to be accidentally mislocated outside of a Region 1 or Region 2 storage rack, with the rack fully loaded. However, this is an area of high neutron leakage and the reactivity effect would be bounded by that of a fuel assembly accidentally misloaded internal to Region 1 or Region 2 storage module. The gaps between the racks are sufficiently small to preclude the accidental mislocation of a fuel assembly in between storage rack modules.

4.8.4 Dropped Fuel Assembly For the case in which a fuel assembly is assumed to be dropped horizontally on top of a rack, the fuel assembly will come to rest horizontally on top of the rack with a minimum separation distance from the active fuel region sufficient to preclude neutron coupling (with fuel in the storage rack). Consequently, nI wc- epr- - -4 .-

rioneu Mepon IUit 4-14 1196

the horizontal fuel assembly drop accident will not result in a significant increase in reactivity.

Furthermore, the soluble boron in the spent fuel pool water assures that the true reactivity is always less than the limiting value for this dropped fuel accident.

Analyses were performed to evaluate the potential loss of Metamicr poison panels in the Region 3 racks by means of a dropped fuel assembly. A very conservative bounding accident condition was analyzed postulating the loss of all Metamice absorber material throughout the entire Region 3 storage racks. For this accident analysis, the Tech Spec limit of 1600 ppm soluble boron was assumed. The results of this postulated accident condition show that the maximum k,,g is well below 0.95, including bias and tolerance uncertainties. This is a very conservative evaluation since an actual dropped assembly at most, would be expected to only damage a maximum of eight Metamice panels (4 inserts).

It is also possible to vertically drop an assembly into a location occupied by another assembly. Such a vertical impact, would, at most cause a small compression of the stored assembly, reducing the water-to-fuel ratio and thereby reducing reactivity. In addition the distance between the active fuel regions of both assemblies will be more than sufficient to ensure no neutron interaction between the two assemblies.

Dropping of an assembly into an unoccupied cell could result in a localized deformation of the baseplate of the rack. The immediate eight surrounding fuel cells could also be affected. However, the amount of deformation for these cells would be considerably less. The resultant effect would be the lowering of a few fuel assemblies in the area near the deformation. Since the Metamic inserts in Region 3 are designed to sit on the baseplate, the localized deformation of the baseplate was considered. The Metamice poison panel inserts in the Region 3 type racks are designed to sit on the baseplate and could potentially move downward and uncover a portion of the active fuel. The resulting geometry is bound by the previously discussed configuration in which a complete loss of all Metamice inserts was assumed with no change in the active fuel region alignment. Therefore, the presence of the boron concentration assures the maximum ken is well below the 0.95 acceptance criteria.

4.9 SOLUBLE BORON DILUTION EVALUATION The soluble boron in the spent fuel pool water is normally a minimum of 1600 ppm under operating conditions. Significant loss or dilution of the soluble boron concentration is extremely unlikely, if not incredible. Nonetheless, an evaluation was performed based on the ANO spent fuel pool data. The minimum required soluble boron concentration in the spent fuel pool water for various conditions is F-1UIL- i'IU LIU1Onn 4-1 n uketnepo l-uzo 4-15b 1196

summarized in Table 4.9.1.

The required minimum soluble boron concentration is 400 ppm under normal conditions and 800 ppm for the most serious credible accident scenario. The volume of water in the pool is 256,860 gallons. Large amounts of unborated water would be necessary to reduce the boron concentration from 1600 ppm to 800 ppm or 400 ppm. Abnormal or accident conditions are discussed below for either low dilution rates (abnormal conditions) or high dilution rates (accident conditions). It should be noted that routine surveillances to measure the soluble boron concentrations in the pool water are required by Technical Specifications.

Small failures or mis-aligned values could possibly occur In the normal soluble boron control system or related systems. Such failures might not be immediately detected. These flow rates would be of the order of 2 gpm (comparable to normal evaporative loss) and the Increased frequency of makeup flow might not be observed. However, an assumed loss flow-rate of 2 gpm dilutions flow rate would require some 123 days to reduce the boron concentration to the minimum required 400 ppm required under normal conditions or 62 days to reach the B00 ppm required for the most severe fuel handling accident.

Routine surveillance measurements of the soluble boron concentration would readily detect the reduction in soluble boron concentration with ample time for corrective action.

Under certain accident conditions, it is conceivable that a high flow rate of unborated water could flow onto the top of the pool. Such an accident scenario could result from rupture of a demineralized water supply line or possibly the rupture of a fire protection system header, both events potentially allowing unborated water to spray onto the pool. A flow rate of up to 2500 gpm could possibly flow onto the spent fuel pool as a result of a rupture of the fire protection line. This would be the most serious condition and bounds all other accident scenarios. Conservatively assuming that all the unborated water from the break poured onto the top of the pool and further assuming instantaneous mixing of the unborated water with the pool water, it would take approximately 142 minutes to dilute the soluble boron concentration to 400 ppm, which is the minimum required concentration to maintain keff below 0.95 under normally operating conditions. In this dilution accident, some 355,000 gallons of water would spill on the auxiliary building floor. Well before the spilling of such a large volume of water, multiple alarms would have alerted the control room of the accident consequences (including the fuel pool high-level alarm, the fire protection system pump operation alarm, and the floor drain receiving tank high level alarm). For this high flow rate condition, 71 minutes would be required to reach the 800 ppm required for the most severe fuel handling accident.

lt a " I n v .

"Ulwec rnepon r l-ewzzoo 4-16 1196

Instantaneous mixing of pool water with the water from the rupture of the demineralized water supply line is an extremely conservative assumption. Water falling on to the pool surface would mix with the top layers of pool water and the portions of the mixed volumes would continuously spill out of the pool. The density difference between water at 1500 F (maximum permissible pool bulk water temperature) and at the temperature of the demineralizer water supply is small. This density difference will not cause the water falling on to the pool surface to instantaneously sink down into the racks overcoming the principal driving force for the flow in the pool, which is the buoyancy force generated in the spent fuel pool racks region due to the heat generation from the spent fuel in the racks. This would further enhance the mixing process between the pool water and spilled water above the racks For the fire water system line break, upon the initial break, the fire protection system header pressure would drop to the auto start setpoint of the fire protection pumps. The start is accompanied with an alarm in the main control room. The annunciator response is to dispatch an operator to find the source of the pump start. Approximately 5 minutes into the event, a Spent Fuel Pool high level alarm would be received in the main control room, assuming that the Spent Fuel Pool level started at the low alarm. The annunciator response for high Spent Fuel Pool level is to investigate the cause. The coincidence of the 2 alarms would quickly lead to the discovery of the failure of the fire protection system and sufficient time to isolate the failure.

The maximum flow rate for a failure of the demineralized water header would provide approximately 900 gpm into the Spent Fuel Pool. Failure of the demineralized water header is not accompanied with an alarm; however, the time to dilute the Spent Fuel Pool from 1600 to 400 ppm is greater than the bounding case described above. In this scenario, there is sufficient time to isolate the failure and to prevent the spilling of some 355,000 gallons of water.

The analysis assume that for a double-ended break in the tire water system piping, the stream of water will arch through the air some 40 feet falling on top of the pool. This is virtually an incredible event.

Should the stream of water fall upon the pool deck, a 3 inch high curb would channel some of the water to the pool drain and prevent all of the water from reaching the pool. Furthermore, the evaluation also assumes at least 3 independent and concurrent accidents occur simultaneously:

Large amount of water flowing from the double-ended pipe break would remain un-detected and is ignored.

Pool water high level alarms either fail or are ignored.

4f-- (--4 19167n77

, ,-Lzw I=ul XU n tl-AW44001 4-17 1196

Alarms indicating large amounts of water flowing into the floor drain have failed or are ignored.

Considering all related facts, a significant dilution of the pool soluble boron concentration in a short period of time without corrective action is not considered a credible event.

It is not considered credible that multiple alarms would fail or be ignored or that the spilling of large volumes of water would not be observed. Therefore, such a major failure would be detected in sufficient time for corrective action to avoid violation of an administrative guideline and to assure that the health and safety of the public is protected.

4.10 NEW FUEL STORAGE RACKS CRITICALITY ANALYSIS The New Fuel Storage Vault is intended for the receipt and storage of fresh fuel under normally dry conditions where the reactivity is very low. To assure criticality safety under accident conditions and to conform to the requirements of General Design Criterion 62, "Prevention of Criticality in Fuel Storage and Handling," two separate criteria must be satisfied as defined in NUREG-0800, Standard Review Plan 9.1.1, "New Fuel Storage." These criteria are as follows:

When fully loaded with fuel of the highest anticipated reactivity and flooded with clean unborated water, the maximum reactivity, including uncertainties, shall not exceed a keff of 0.95.

With fuel of the highest anticipated reactivity in place and assuming the optimum hypothetical low density moderation, (i.e., fog or foam), the maximum reactivity shall not exceed a keff of 0.98.

The New Fuel Storage Vault provides two 4 x 9 storage rack modules with cell array of storage locations arranged on a 21.00 inch lattice spacing. Calculations were made with the 238-group NITAWL KENO Va code package (SCALE 4.3), a three-dimensional Monte Carlo analytical technique, with fresh fuel assemblies with 4.95 wt% initial enrichment. These calculations were made for various moderator densities and the results shown in Figure 4.10 1, shows that the peak reactivity (optimum moderation) occurs at 9% moderator density. The calculations for the configuration illustrated in Figure 4.10.2 confirm that five locations in each of the storage racks are required to remain empty in order to meet the regulatory limits. Results of the criticality safety analyses are summarized in Table 4.10.1 for the two accident conditions for fuel assemblies of 4.95 +/-0.05 wt% initial enrichment. The maximum reactivity at LJ1N-t.r a L11 fln'nln7a I lulms. ItNUUIL n7-4U4ogi 4-18U 1196

9% moderator density is 0.9708, Including uncertainties, which Is within the regulatory limit of 0.98, thus confirming the acceptability of the New Fuel Vault for 4.95 +/-0.05 wt% fuel.

Additional calculations at 9% moderator density, performed for the storage pattern depicted in Figure 4.10.3, show that this storage configuration is acceptable for storage of fresh fuel assemblies of up to 4.20 wt% enrichment with four locations in each rack array required to remain empty.

In the flooded condition (clean unborated water), the storage locations are essentially isolated from each other (neutronically). Under these conditions and with fuel of 4.95 +/-0.05 wt% enrichment, the maximum reactivity, including all known uncertainties is less than the regulatory limit of 0.95 for keff, thus confirming the acceptability of the NFV for 4.95 wt% fuel in the fully flooded accident condition. At 4.2 wt% enrichment in the flooded condition, the reactivity will be substantially lower than that for 4.95 +/- 0.05 wt% enrichment and would therefore be acceptable for storage. (Note: in the fully flooded conditions the fuel assemblies are neutronically isolated from each other and the reactivity is independent of the presence or number of blanked-off cells.)

4.11 FUEL HANDLING EQUIPMENT Criticality safety evaluations were also performed for handling of fresh fuel assemblies during transfer from the new fuel vault to the reactor core, including the new-fuel elevator, (he up-ender and fuel carriage, and the temporary storage rack within the transfer canal. The new fuel elevator is located on the south wall of the pool facing the Region I spent fuel storage racks This device can position a fresh fuel assembly 16 inches (assembly center line) from the wall. The distance from the wall to the edge of the rack is 24.5 inches A distance of 7.845 inches exists between the centerline of the assembly in the elevator and the edge of the closest fuel storage cell in the rack. For normal operation, the maximum reactivity with fuel in the New Fuel Elevator (with Region 1 loaded) is 0.9512 in unborated water. This is reduced to 0.9271 with credit for 100 ppm soluble boron. Additional calculations were performed to evaluate the effect of accidentally dropping or misplacing an assembly adjacent to the New Fuel Elevator while it is loaded. A most reactive location for the dropped assembly was determined. A credit of 700 ppm boron will ensure k-effective remains below 0.95 should such an event occur. The New Fuel Elevator therefore meets the criticality acceptance criteria defined in 10 CFR 50.68. The Upenderl Fuel Carriage device handles a single assembly. The maximum reactivity of a single fresh assembly containing 4.95 wt% +0.05 enriched fuel in water is 0.9425. Furthermore for a postulated accident in which a second fresh assembly was positioned near the Upender/Fuel Carriage, the presence of soluble fUI1t1- r .pr i1I-LIe f- _ 19 r1VILUL; Mt1PrJrJ f'JJ-ZUd,400J 4-7H 1196

boron (1600 ppm minimum) excludes the possibility of any criticality concern.

The transfer canal incorporates a 7-cell temporary storage rack on a linear array at a 21-1/8 inch spacing (6 locations for fuel assemblies and I location for a damaged fuel). The maximum k-effective for normal operation of this rack was determined to be 0.9413. Evaluations of a potential mis-placement of a fresh fuel assembly at a position of closest approach to another assembly in the spent fuel rack, separated only by the structure of the temporary rack, shows that the maximum keff (in the absence of any soluble boron) would be 0.9698. The presence of 200 ppm soluble boron would be sufficient to maintain the maximum k-effective below 0.95. However, the transfer canal, during operations, would always contain the minimum 1600 ppm boron (or usually more), significantly reducing reactivity and further eliminating any criticality concern.

4.12 REFERENCES

[4.1.1] M. G. Natrella, Experimental Statistics, National Bureau of Standards Handbook 91, August 1963.

[4.1.2] L. I. Kopp, "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants," NRC Memorandum from L. Kopp to T. Collins, August 19, 1998.

[4.4.1] J. F. Briesmeister, Editor, UMCNP - A General Monte Carlo N-Particle Transport Code, Version 4A," LA-1 2625, Los Alamos National Laboratory (1993).

[4.4.2] L. M. Petrie and N.F. Landers, "KENO Va - An Improved Monte Carlo Criticality Program with Supergrouping,' Volume 2, Section F11 from "SCALE: A Modular System for Performing Standardized Computer Analysis for Licensing Evaluation" NUREGICR-0200, Rev. 4, January 1990.

[4 4.3] 'SCALE 4.3: A Modular System for Performing Standardized Computer Analysis for Licensing Evaluation For Workstations and Personal Computers, Volume 0," CCC-545, ORNL-RSICC, Oak Ridge National Laboratory (1995).

I I-I r Apoi -11 as uI n.. _4 A rivituc mepan ril-'ue'dut 4-20 1196

[4.4.4] N. M. Greene, L. M. Petrie and R. M. Westfall, MNITAWL-Il: Scale System Module for Performing Shielding and Working Ubrary Production," Volume 1, Section F1 from "SCALE: A Modular System for Performing Standardized Computer Analysis for Licensing Evaluation' NUREGICR-0200, Rev. 4, January 1990.

[4.4.5] M. Edenius, K. Ekberg, B. H. Forssen, and D. Knott, "CASMO-4 A Fuel Assembly Burnup Program User's Manual," StudsviklSOA-9511, Studsvik of America, Inc. and Studsvik Core Analysis AB (proprietary).

[4.4.6] D. Knott, 'CASMO-4 Benchmark Against Critical Experiments", SOA-94/13, Studsvik of America, Inc., (proprietary).

[4.4.7] D. Knott, "CASMO-4 Benchmark Against MCNP," SOA-94/12, Studsvik of America, Inc.,

(proprietary).

[4.4.8] S E. Turner, "Uncertainty Analysis - Burnup Distributions", presented at the DOEISANDIA Technical Meeting on Fuel Bumup Credit, Special Session, ANS/ENS Conference, Washington, D.C., November 2, 1988.

rIc-jiiecI-- o- -'( 4--21 riollec Keporl mineuz-2tsul 4-21 1196

Table 4.1.1 Fuel Assembly Specifications Assembly Data .-

Rod Arra Size 15x15 Rod Pitch (inches) 0.5680 Active Fuel Length (inches) 144 Fuel Rod Data Total Number of Fueled Rods 208 Fuel Rod Outer Diameter (inches) 0.430 Fuel Rod Inner Diameter (inches) 0.377 Cladding Thickness (inches) 0.0265 Cladding Material Zircaloy Pellet Diameter (inches) 0.370 U02 Stack Density, gms/cc 10.412 +/-0.20 Guide Tube Data Number of Tubes 16 Outer Diameter (inches) 0.530 Wall Thickness (inches) 0.016 Material Zircaloy Instrument Tube Data Number of Tubes I Outer Diameter (inches) 0,493 Wall Thickness (inches) 0.026 Material Zircaloy Holtec Report HI-2022867 4-22 1196

Table 4.2.1 Summary of the Criticality Safety Analyses for Storage of Spent Fuel Assemblies in Region 1 Racks Reference keff 0.9710 Bumup, MWDIKgU 35.4 MCNP4a Bias +0.0009 Temperature Bias 0.0100 MCNP4a Bias Uncertainty +/-0.0011 MCNP4a Statistical (95195) Uncertainty +/-0.0007 Manufacturing Tolerance Uncertainty +/-0.0041 Enrichment Tolerance Uncertainty +/-0.0027 Depletion Uncertainty +/-0.01 19 Fuel Eccentric Positioning Uncertainty +/-0.0020 Statistical Combination of Uncertainties +/-0.0131 Maximum keff 0.9950 Regulatory Limiting keff 1.0000 LJ-I#-- 0---.4 UI A IUILOL UXpIJUI t "I-euV4oorUf 4-Z3 1196

Table 4.2.2 Minimum Bumup Required for Storage of Spent Fuel Assemblies in the Region 1 Racks BURNUP, MWDJKgU Average 0 Years 5 Years 10 Years 15 Years 20 Years Enrichment, Cooling Cooling Cooling Cooling Cooling 235 wt% U Time Time Time Time Time 2 0.50 0.375 0.25 0.125 0.00 2.5 7.50 7.35 7.20 7.05 6.90 3.0 13.60 13.10 12.60 12.10 11.60 3.5 18.70 18.03 17.35 16.68 16.00 4.0 24.27 23.45 22.64 21.82 21.00 4.5 30.00 29.05 28.10 27.15 26.20 4.95 35.40 34.08 32.75 31.43 30.10 IHnl+.,

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14-/4 1196

Table 4.2.3 Bounding Polynomial Fits to Determine Minimum Acceptable Bumup for Storage of Spent Fuel Assemblies Storage in Region 1 Racks as a Function of Initial Average Enrichment Decay Time, Years Burnup, MWDIKgU 0 - 0.1661*E 4 +2.791*E 3 - 17.013*E2 + 55.795*E - 62.57 5 - 0.3014*E 4+4.6636*E 3 - 26.447*E 2 +75.84*E - 77.88 10 - 0.443*E 4+6.6239*E 3 - 36.324*E2 + 96.837*E - 93.94 15 - 0.5782*E 4+8.4965*E 3 - 45.758*E 2 + 116.88*E - 109.27 20 -0.7198*E 4 +10.457*E 3 - 55.635*E 2 + 137.88*E- 125.30 Note: E = Initial average enrichment in wt% 23 5U

-5 11_6 Holiec Report HI-Q-022867 4-25 1136

Table 4.2.4 Summary of the Criticality Safety Analyses for a 2-of-4 Checkerboard Storage of Fresh Fuel Assemblies and Empty Cells in Region I Racks Reference keff 0.9264 MCNP4a Bias +0.0009 Temperature Effect +0.0117 Axial Bumup Distribution Penalty Not Applicable MCNP4a Bias Uncertainty +0.0011 MCNP4a Statistics (95195) Uncertainty +/-o 0007 Manufacturing Tolerance Uncertainty +/-0.0044 Enrichment Tolerance Uncertainty +/-0.0019 Depletion Uncertainty Not Applicable Fuel Eccentric Positioning Uncertainty +/-O.0017 Statistical Combination of Uncertainties +0.0053 Maximum keff 0.9443 Regulatory Limiting keff 1.0000 4-26_l 11__9__.

6 HOWC K6POrt hi-ZO22667 4-26 1196

Table 4.2.5 Summary of the Criticality Safety Analyses for Storage of Spent Fuel Assemblies in Region 2 Racks Initial Enrichment, wt% 235U 4.95+/-0.05 Bumup, MWD/KgU 41.00 Cooling Time, years 0 Reference keff 0 9619 MCNP4a Bias +0.0009 Temperature Effect 0.0112 MCNP4a Bias Uncertainty +/-O.0011 MCNP4a Statistics (95195) Uncertainty +/-0.0007 Manufacturing Tolerance Uncertainty +/--0.0094 Enrichment Tolerance Uncertainty +/-0.0028 Depletion Uncertainty +/-0.0133 Fuel Eccentric Positioning Uncertainty +/-0.0088 Statistical Combination of Uncertainties iO.0188 Maximum keff 0.9928 Regulatory Limiting keff 1.0000 o t - JN~jJ1 - ( 1-198 rlult-u n-e-U4o4 II 14-/Z7 1196

Table 4.2.6 Minimum Burnup Required for Storage of Spent Fuel Assemblies in the Region 2 Racks BURNUP, MWDIKgU Average 0 Years 5 Years 10 Years 15 Years 20 Years Enrichment, Cooling Cooling Cooling Cooling Cooling Wt% 236U Time Time Time Time Time 2.00 4.00 4.00 4.00 4.00 4.00 2.50 11.50 11.00 10.50 10.00 9.50 3.00 17.80 17.10 16.40 15.70 15.00 3.50 23.50 22.63 21.75 20.88 20.00 4.00 29.20 28.15 27.10 26.05 25.00 4.50 35.00 33.63 32.25 30.88 29.50 4.95 41.00 39.25 37.50 35.75 34.00 Anrolte PRnoir4 Wi-In9')R7f AO -

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I PI-LULLC4v 4;-ZO 1196

Table 4.2.7 Bounding Polynomial Fits to Determine Minimum Acceptable Bumup for Storage of Spent Fuel Assemblies Storage in Region 2 Racks as a Function of Initial Enrichment Decay Time, Years Burnup, MWDIKgU 0 0.6257E3 -6.8758E + 36.293E - 46.0

_ _ __ _ 0.4934*E 3 - 5.5023*E 2 +31.23*E - 40.3 I __ 0.3634*E 3 4.1511*E +26.244*E - 34.7 15 _ 0.2311 *E3 2.776*E + 21.186*E-29.0 20 0.1011*E 3 _1.4265-E2 + 16.19*E-23.4 Note: E = Initial average enrichment in wt% 235U W -D~IuluXstVwl

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Ui

. n& =u D 4-ZU 1196

Table 4.2.8 Summary of the Criticality Safety Analyses for a 2-of-4 Checkerboard Storage of Fresh Fuel Assemblies and Empty Cells in Region 2 Racks Reference keff 0.9359 MCNP4a Bias +/-0.0009 Temperature Effect 0.0129 MCNP4a Bias Uncertainty +0.0011 MCNP4a Statistics (95195) Uncertainty +/-0.0007 Manufacturing Tolerance Uncertainty +/-0.0119 Enrichment Tolerance Uncertainty +0.0020 Depletion Uncertainty Not Applicable Fuel Eccentricity Positioning Uncertainty +0.0026 Statistical Combination of Uncertainties +/-0.0124 Maximum keff 0.9621 Regulatory Limiting keff 1.0000 H1------ Re orHI-2022867 Holec Report 19 Holtec 4-30 1196

Table 4.2.9 Summary of the Criticality Safety Analyses for Storage of Fresh Fuel Assemblies in ANO Unit 1 Region 3 Racks Reference keff 0.9776 MCNP4a Bias +/-0.0009 MCNP4a Bias Uncertainty _ +0.0011 MCNP4a Statistics (95/95) Uncertainty +/-0.0007 Manufacturing Tolerance Uncertainty +/-0.0152 Enrichment Tolerance Uncertainty +/-0.0016 Depletion Uncertainty Not Applicable Fuel Eccentric Positioning Uncertainty Negative Pool Water Temperature Uncertainty +0.0012 Statistical Combination of Uncertainties +/-0.0154 Maximum keff 0.9939 Regulatory Limiting keff 1.0000 ontec t(eport rioiiec Keport hI-U27 Hl-Z2o2atj7 4-31 4-31 1196

Table 4.9.1 Required Soluble Boron Concentrations in the SFP Water.

Solule Bron Soluble Boron Condition oRequire fork<1 Required for k<o.95 Region 1: All Spent Fuel Assemblies 0 225 Region 1: Accident condition of 1 fresh fuel assembly mis-placed into a cell intended to store - 475 spent fuel .-

Region 1: Accident condition of 1 fresh fuel assembly mis-placed into a cell intended to remain empty 675 Region 2: All Spent Fuel Assemblies 0 180 Region 2: Accident condition of I fresh fuel assembly mis-placed into a cell intended to store spent fuel 450 Region 2: Checkerboard Arrangement 0 100 Region 2: Accident condition of I fresh fuel assembly mis-placed into a cell intended to remain 800 empty 0 Region 3: All Fresh Fuel Assemblies

. 400 Region 3: Dropped fuel assembly with all fresh fuel ___

assemblies with no poison inserts. 1600

-. 0.2.

Re or HI-2022867

.. ol e Report

_I .7_4_3 2 119 Holtec 4-32 1196

Table 4.1 0.1 Summary of New Fuel Vault Criticality Safety Analysis Optimum Flooded Moderation Optimum*

Moderation Moderation (Figure 4.10.1) (Figure 4.10.1) (Figure 4.10.2)

Initial Enrichment, wt% 4.95+/- 0.05 4.95+/- 0.05 4.20 +/- 0.05 Temperature for 20 0 C (68 0 F) 20 -C (68 -F) 20 -C (68 0 F) analysis Reference kgr 0.9639 0.9350 0.9650 Calculational bias, Ak 0.0030 0.0030 0.0030 Uncertainties_

KENO Bias +/-0.0012 +/-0.0012 +/-0.0012 KENO Statistics +/-0.0004 +/-0.0004 +/-0.0004 Lattice Spacing +/-0.0019 +/-0.0016 +/-0.0018 Fuel Density +/-0.0025 +/-0.0026 +/-0.0022 Fuel Enrichment +/-0.0019 +/-0.0030 +0.0028 Statistical Combination +/-0.0039 +/-0.0045 +/-0.0042 Total kff 0.9669 +/- 0.0039 0.9380 +/- 0.0045 0.9680 +/- 0.0042 Maximum k-eff 0.9708 0.9425 0.9722 Regulatory Limit 0.98 0.95 0.98

At 4.2 wt% enrichment, the flooded condition reactivity is much lower than the 4.95 wt% case.

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10 650 Pitch Reference d .

0.075 +0.004 0.020 +/-0.004 0 090-0.010

. 970 +0.025 j 0 _ C, 1.31 Water Gap Figure 4.3.1: A Cross-Sectional View of the Calculational Model Used for the Region 1 Rack Analysis (NOT TO SCALE).

4-36 1196 Holtec Report 1-11-2022867 Report Hl-2022857 4-36 1196

10 650 Pich Figure 4.3.2: A Cross-Sectional View of the Calculational Model Used for the Region 2 Rack Analysis (NOT TO SCALE).

Note: The straight portion of the flux trap is modeled as 7.2 inches. In order to preserve the pitch due to a conservative reduction of the flux trap gap width from a design reference value of 1.556 inches to 1.456 inches (based on measurements which accounts for the bow in the cell walls), the cell ID was modeled as 9.07 +0.050/-0.025 inches.

HI-202267 4-37 1196 Report MI-20221367 Holtec Report 4-37 1196

1.00 PIi U

La 0.83 I LI C

S CS

.3 z1 0.70 0 55 Pmnr~ tJcder~r D sftty Figure 4.10.1 Reactivity of the New Fuel Vault as a Function of Moderator Density (4.95% +/-0.05 Enrichment).

Holtec Reporlt 1-2022867 4-38 1196

N 8 = = = == = l 7 _______

5 X X X- X X 4 _ X X X X X _

A B C D E F G H K Figure 4.10.2: Acceptable New Fuel Storage Vault Configuration for up to 4.95 wt% Enrichment Fresh Fuel Note: X's show the locations where fuel assemblies will not be stored.

5 X X X-_

4 _

X

_X

- x __

A B C D E F G H K Figure 4.10.3: Acceptable New Fuel Storage Vault Configuration for up to 4.2 wt% Enrichment Fresh Fuel.

Note: X's show the locations where fuel assemblies will not be stored.

Holtec Report HI-2022867 4-39 1196

5.0 THERMAL-HYDRAULIC CONSIDERATIONS

5.1 INTRODUCTION

This document requests an operating license amendment to modify the spent fuel storage capacity at the ANO-1 nuclear power plant to preserve the capability of storing fresh fuel assemblies in the SFP racks. As discussed in Section 1.0, this will be achieved by placing poison inserts into the flux trap area of some of the existing Region 2 spent fuel storage racks (SFSRs). This section provides a summary of the analyses performed to demonstrate the compliance of the SFP and its attendant cooling system with the provisions of USNRC Standard Review Plan (SRP) 9.1.3 (Spent Fuel Pool Cooling and Cleanup System, Rev. 1, July 1981) and Section III of the USNRC MOT Position Paper for Review and Acceptance of Spent Fuel Storage and Handling Applications," (April 14, 1978). Similar methods of thermal-hydraulic analysis have been used in the licensing evaluations for other SFP capacity expansion projects.

The thermal-hydraulic qualification analyses for the modified rack array may be broken down into the following categories:

i. Evaluation of bounding maximum decay heat versus time profiles, used as input to subsequent analyses.

Ii. Evaluation of loss-of-forced cooling scenarios, to establish minimum times to perform corrective actions and the associated makeup water requirements.

iii. Determination of the maximum local water temperature, at the instant when the pool decay heat reaches its maximum value, to establish that localized boiling in the SFSRs is not possible while forced cooling is operating. The bulk pool temperature is postulated to be at the maximum limit.

iv. Evaluation of the maximum fuel rod cladding temperature, at the instant when the pool decay heat reaches its maximum value, to establish that nucleate boiling is not possible while forced cooling is operating. The bulk pool temperature is postulated to be at the maximum limit.

Holtec Report I-2022867 5-1 1196

The following sections present plant system descriptions, analysis methodologies and assumptions, a synopsis of the input data employed, and summaries of the calculated results.

5.2 COOLING SYSTEMS DESCRIPTION The Spent Fuel Cooling (SFC) System is designed to maintain the water quality and clanty and to remove the decay heat from the stored fuel in the spent fuel pool. It is designed to maintain the spent fuel pool water at less than or equal to approximately 150 "F. This is accomplished by recirculating spent fuel coolant water from the spent fuel pool through the pumps and coolers and back to the pool. The spent fuel pool coolers reject heat to the nuclear intermediate cooling water system, which subsequently rejects its heat to the service water system.

The spent fuel coolant pumps take suction from the spent fuel pool and deliver the water through the tube side of two coolers arranged in parallel back to the pool. In addition to its primary function, the system provides for purification of the spent fuel pool 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. A bypass purification loop is provided to maintain the purity of the water in the spent fuel pool. This loop is also utilized to purify the water in the BWST following refueling and to maintain clarity in the fuel transfer canal during refueling. Water from the BWST or fuel transfer canal can be purified by using the borated water recirculation pump. The system also provides for filling the fuel transfer canal, the incore instrumentation tank, and the cask loading area from the BWST.

The spent fuel storage pool is provided with a makeup system design which meets the intent of Safety Guide 13. That is, the Borated Water Storage Tank (BWST) is a Seismic Category 1 vessel; all connecting piping is located in a Seismic Category 1 structure; a backup system for supplying water to the pool is provided through a temporary connection to the Seismic Category 1 service water system; and, there is sufficient time to rig a temporary makeup water supply to the pool in the event of failure of the normal source. The service water system can be supplied from either the Dardanelle Reservoir or the emergency cooling pond.

Holtec Report HI-2022867 5-2 1196

5.3 SPENT FUEL POOL DECAY HEAT LOADS The decay heat in the SFP is generated in the spent fuel assemblies stored therein.

In order to conservatively simplify the decay heat calculations, the total decay heat is considered as coming from two different groups of assemblies:

i. Fuel assemblies from previous offloads already stored in the SFP ii. Fuel assemblies that are being offloaded from the reactor to the SFP The fuel assemblies in the first group are referred to as previously offloaded fuel. Over the relatively short transient evaluation periods of this report the heat generation rate of these assemblies reduces very slowly with time, due to the exponential nature of radioactive decay and their relatively long decay periods. The decay heat contribution of these assemblies can therefore be conservatively treated as constant, neglecting any reduction in their decay heat contribution during the evaluation period. The fuel assemblies in the second group are referred to as recently offloaded fuel. The heat generation rate of these assemblies reduces rapidly with time, so the decay heat contribution of these assemblies is treated as time varying. The following equation defines the total decay heat generation in the SFP.

QcG (T)= Qp +F(r)XQR (r) (-

where:

QGEN(S) is the total time-varying decay heat generation rate in SFP, Btulhr Qp is the decay heat contribution of the previously offloaded fuel, Btu/hr F(l) is the fraction of the recently offloaded fuel transferred to the SFP QR(T) Is the decay heat contribution of the recently offloaded fuel, Btu/hr xris the fuel decay time after reactor shutdown, hrs Prior to the start of fuel transfer from the reactor to the SFP, F(T) is equal to zero and the total decay heat in the SFP will be equal to the invariant portion Qp. During the fuel transfer, F(c) will increase linearly from zero to one, and the total decay heat in the SFP will increase to QP+QR(.r).

Following the completion of fuel transfer, the total decay heat in the SFP will decrease as QR(T) decreases.

Holtec Report HI-2022867 5-3 1196

The decay heat contributions of both the previously and recently offloaded fuel are determined using the Holtec QA validated computer program DECOR [5.3.1]. This computer program incorporates the Oak Ridge National Laboratory (ORNL) ORIGEN2 computer code [5.3.2] for performing decay heat calculations. The ORIGEN2 code is not modified for its incorporation into the Holtec program and should give the same results as DECOR.

Based on the input data provided in Tables 5.3.1 and 5.3.2, the fuel decay heat is determined for the following two offload scenarios:

1. Partial Core Offload - A refueling batch of 76 assemblies is offloaded from the plant's reactor into the SFP, completely filling all available storage locations. The total SFP inventory prior to the offload is 912 fuel assemblies, for a final post-offload inventory of 988 fuel assemblies. This slightly exceeds the storage capacity of the ANO-1 SFP (and the ANO-1 TS 4.3.3 limit of 968 assemblies) and is used for calculation of decay heat loads, which is conservative.

2. Full Core Offload - The full core of 177 assemblies is offloaded from the plant's reactor into the SFP, completely filling all available storage locations. The total SFP inventory prior to the offload is 836 fuel assemblies, for a final post-offload inventory of 1013 fuel assemblies. This slightly exceeds the storage capacity of the ANO-1 SFP (and the ANO-1 TS 4.3 3 limit of 968 assemblies) and is used for calculation of decay heat loads, which is conservative.

5.4 MINIMUM TIME-TO-BOIL AND MAXIMUM BOILOFF RATE In this section, we present the methodology for calculating the minimum time-to-boil and corresponding maximum boiloff rate.

The following conservatisms and assumptions are applied in the time-to-boil and boiloff rate calculations:

The initial SFP bulk temperature is assumed to be equal to the bulk temperature limit of 150 OF.

Holtec Report HI-2022867 5-4 1196

The thermal inertia (thermal capacity) of the SFP is based on the net water volume only.

This conservatively neglects the considerable thermal inertia of the fuel assemblies, stainless steel racks and stainless steel SFP liners.

During the loss of forced cooling evaluations, it is assumed that makeup water is not available. This minimizes the thermal capacity of the SFP as water is boiled off, thus increasing the water level drop rate.

The loss of forced cooling is assumed to occur coincident with the peak SFP bulk temperature and the maximum pool decay heat. Maximizing the initial temperature and the pool decay heat conservatively minimizes the calculated time-to-boil.

The governing enthalpy balance equation for this condition, subject to these conservative assumptions, can be written as:

dl' C~T+r)rO)(5-2) d = Q 8¢"

where:

C(Xr) = Time-varying SFP thermal capacity (BTUfF)

T = Time after cooling is lost (hr) nO = Loss of cooling time after shutdown (hr)

T=Pool water temperature, ('F)

Equation 5-2 is solved to obtain the bulk pool temperature as a function of time, the time-to-boil, boil-off rate and water depth versus time. The major input values for these analyses are summarized in Table 5 4.1.

5.5 MAXIMUM SFP LOCAL WATER TEMPERATURE In this section, a summary of the methodology for evaluating the maximum SFP local water temperature is presented. The results of these evaluations are maximum local water temperatures.

In order to determine an upper bound on the maximum local water temperature, a series of conservative assumptions are made. The most important of these assumptions are:

Holtec Report H--2022867 5-5 1196

The walls and floor of the SFP are all modeled as adiabatic surfaces, thereby neglecting conduction heat loss through these items.

Heat losses by thermal radiation and natural convection from the hot SFP surface to the environment are neglected.

a No downcomer flow is assumed to exist between the rack modules.

The hydraulic resistance parameters for the rack cells, permeability and inertial resistance, are conservatively adjusted by 10%.

The bottom plenum heights used in the model are less than the actual heights.

The hydraulic resistance of every SFSR cell is determined based on the most restrictive water inlet geometry of the cells over rack support pedestals (i.e., all baseplate holes are completely blocked). These cells have a reduced water entrance area, caused by the pedestal blocking the baseplate hole, and a correspondingly increased hydraulic resistance.

The hydraulic resistance of every SFSR cell includes the effects of blockage due to an assumed dropped fuel assembly lying horizontally on top of the SFSRs.

The objective of this study is to demonstrate that the thermal-hydraulic criterion of ensuring local subcooled conditions in the SFP is met for all postulated fuel offload scenarios. The local thermal-hydraulic analysis is performed such that slight fuel assembly variations are bounded.

An outline of the Computational Fluid Dynamics (CFD) approach is described in the following.

There are several significant geometric and thermal-hydraulic features of the ANO-1 SFP that need to be considered for a rigorous CFD analysis. From a fluid flow modeling standpoint, there are two regions to be considered. One region is the SFP bulk region where the classical Navier-Stokes equations [5.5.1] are solved, with turbulence effects included. The other region is the SFSRs containing heat generating fuel assemblies, located near the bottom of the SFP.

in this region, water flow is directed vertically upwards due to buoyancy forces through relatively small nI e- i- p ... I- U22_ 5-6 nollec Kepon HI-ZUZ2667 5-6 1196

flow channels formed by the B&W 15x15 fuel assemblies in each SFSR cell. This situation is modeled as a porous region with pressure drop in the flowing fluid governed by Darcy's Law as:

.=-- V- Cp IVIY!.(-3 ax, K(i) - 2 where MP/aX; is the pressure gradient, K(i), Vi and C are the corresponding permeability, velocity and inertial resistance parameters and ti is the fluid viscosity. These terms are added as sink terms to the classic Navier-Stokes equations. The permeability and inertial resistance parameters for the rack cells loaded with B&W 15x15 fuel assemblies are determined based on friction factor correlations for the laminar flow conditions that would exist due to the low buoyancy induced velocities and the small size of the flow channels.

The ANO-1 SFP geometry requires an adequate portrayal of both large scale and small scale features, spatially distributed heat sources In the SFSRs and water inlet/outlet piping. Relatively cooler bulk water normally flows down between the perimeter of the fuel rack array and wall liner, a clearance known as the downcomer. Near the bottom of the racks the flow turns from a vertical to horizontal direction into the bottom plenum, supplying cooling water to the rack cells.

Heated water issuing out of the top of the racks mixes with the bulk water. An adequate modeling of these features on the CFD program involves meshing the large scale bulk SFP region and small scale downcomer and bottom plenum regions with sufficient number of computational cells to capture both the global and local features of the flow field.

The distributed heat sources in the SFP racks are modeled by identifying distinct heat generation zones considering recently offloaded fuel, bounding peaking effects, and the presence of background decay heat from previous offloads. Two heat generating zones are modeled. The first consists of background fuel from previous offloads. The second zone consists of fuel from recently offloaded fuel assemblies. This is a conservative model, since all of the hot fuel assemblies from the recent offload are placed in a contiguous area. A uniformly distributed heat generation rate was applied throughout each distinct zone (i.e., there were no variations in heat generation rate within a single zone).

The CFD analysis was performed on the commercially available FLUENT [5.5.2] Computational Fluid Dynamics program, which has been benchmarked under Holtec's QA program. The 5-7 1196 Repon MI-U2b(

oirec tepan Holiec HI-M-2867 57 1196

FLUENT code enables buoyancy flow and turbulence effects to be included in the CFD analysis.

Buoyancy forces are included by specifying a temperature-dependent density for water and applying an appropriate gravity vector. Turbulence effects are modeled by relating time-varying Reynolds' Stresses to the mean bulk flow quantities with the standard k-e turbulence model.

Some of the major input values for this analysis are summarized in Table 5.5.1. An isometric view of the assembled CFD model is presented in Figure 5.5.1.

5.6 FUEL ROD CLADDING TEMPERATURE In this section, the method to calculate the temperature of the fuel rod cladding is presented.

The maximum fuel rod cladding temperature is determined to establish that nucleate boiling is not possible while forced cooling is operating. This requires demonstrating that the highest fuel rod cladding temperatures are less than the local saturation temperature of the adjacent SFP water. The maximum fuel cladding superheat above the local water temperature is calculated for two different peak fuel rod heat emission rates.

A fuel rod can produce F, times the average heat emission rate over a small length, where F, is the axial peaking factor. The axial heat distribution in a rod is generally a maximum in the central region, and tapers off at its two extremities. Thus, peak cladding heat flux over an infinitesimal rod section is given by the equation:

=Q (5-4) where Q is the rod average heat emission and A, is the total cladding external heat transfer area in the active fuel length region. The axial peaking factor is given in Table 5.5.1.

As described previously, the maximum local water temperature was computed. Within each fuel assembly sub-channel, water is continuously heated by the cladding as it moves axially upwards under laminar flow conditions. Rohsenow and Hartnett [5.6.1] report a Nusselt-number for laminar flow heat transfer in a heated channel. The film temperature driving force (ATf) at the peak cladding flux location is calculated as follows:

5-8 1196..

rioltec Report Hl-2022B67 5-8 1196

A Tf = q-hf (5-5) hf =NuKw Dh where hf is the waterside film heat transfer coefficient, Dh is sub-channel hydraulic diameter, Kw is water thermal conductivity and Nu is the Nusselt number for laminar flow heat transfer.

In order to introduce some additional conservatism in the analysis, we assume that the fuel cladding has a crud deposit resistance R, (equal to 0.0005 ft2 -hr-°F1/tu) which covers the entire surface. Thus, including the temperature drop across the crud resistance, the cladding to water local temperature difference (AT,) is given by the equation AT, = ATr + Rc x qr.

5.7 RESULTS This section contains results from the analyses performed for the postulated offload scenarios.

5.7.1 Decay Heat For the offload scenarios described in Section 5.3, the calculated SFP decay heat loads are summarized in Table 5.7.1. Given the conservatisms incorporated Into the calculations, actual decay heat loads will be lower than these calculated values. Figures 5.7.1 and 5.7.2 each present profiles of net decay heat load versus time for the evaluated transient scenarios.

5.7.2 Minimum Time-to-Boil and Maximum Boiloff Rate For the offload/cooling described in Section 5.3, the calculated times-to-boil and maximum boil-off rates are summarized in Table 5.7.2. These results show that, in the extremely unlikely event of a failure of forced cooling to the SFP, there would be at least 3.23 hours2.662037e-4 days 0.00639 hours 3.80291e-5 weeks 8.7515e-6 months available for corrective actions prior to SFP boiling. Given the conservatisms incorporated into the calculations, actual times-to-boil will be higher than these calculated values. It is noted that a complete failure of forced cooling Is extremely unlikely. The maximum water boiloff rate is less than 85 gpm.

I-nlf-,, DF2.,,,.4

- LJI e n

- 91vjJWJ nH-CUC.&Q Iv~pl 0-to 1196

5.7.3 Local Water and Fuel Cladding Temperatures Consistent with our approach to make conservative assessments of temperature, the local water temperature calculations are performed for a SFP with a total decay heat generation equal to the calculated decay heat load coincident with the maximum SFP bulk temperature. Thus, the local water temperature evaluation is a calculation of the temperature increment over the theoretical spatially uniform value due to local hot spots (due to the presence of highly heat emissive fuel assemblies). As described in Subsection 5.6, the peak fuel clad superheats (i e., the maximum clad-to-local water temperature difference) are determined. The resultant bounding superheat values were used to calculate bounding maximum fuel clad temperatures.

The numeric results of the maximum local water temperature and the bounding fuel cladding temperature evaluations are presented in Table 5.7.3. Figure 5.7.3 presents converged temperature contours in a vertical slice through the hot fuel region. Figure 5.7.4 presents converged velocity vectors in a vertical slice through the hot fuel region.

Both the maximum local water temperatures and the bounding fuel cladding temperatures are substantially lower than the 240 0F local boiling temperature at the top of the SFSRs. These results demonstrate that boiling, including nucleate boiling on clad surfaces, cannot occur anywhere within the ANO-1 SFP.

Under a postulated accident scenario of the loss of all cooling, the water temperature will rise.

Assuming a temperature of 212 0 F at the inlet to the rack cells, and conservatively using the bounding bulk-to-local and local-to-clad temperature differences from Table 5.7.3, the maximum possible cladding temperature will be 285.5 'F, which is greater than the saturation temperature at the top of the active fuel length. Due to the low maximum assembly heat flux (approximately 7000 W/rn2 ) and the critical heat flux required for departure from nucleate boiling (on the order of 106 W/m 2 ), it can be concluded that the fuel cladding will not be subjected to departure from nucleate boiling even under the postulated accident scenario of the loss of all SFP cooling and the cladding integrity would be maintained.

5-10 1196 ronec epon i-ionec Report MI-UL(

Hk-Z202267 5-10 1196

5.8 REFERENCES

[5.3.1] `QA Documentation for DECOR," Holtec Report HI-971734, Revision 0.

[5.3.2] A. G. Croff, "ORIGEN2 - A Revised and Updated Version of the Oak Ridge Isotope Generation and Depletion Code," ORNL-5621, Oak Ridge National Laboratory, 1980.

[5.5.1] Batchelor, G. K, "An Introduction to Fluid Dynamics", Cambridge University Press, 1967.

[5.5.2] 'Validation of FLUENT Version 5.1", Holtec Report HI-992276, Revision 0.

[5.6.1] Rohsenow, N. M., and Hartnett, J. P., "Handbook of Heat Transfer", McGraw Hill Book Company, New York, 1973.

'-ioite -e o I----------1-11 Honec Report Hl-ZUZZ667 5-11 1196

Table 5.3.1 Key Input Data for Decay Heat Computations Input Data Parameter Value Reactor Thermal Power (MWt) 2800 Number of Assemblies in Reactor Core 177 Maximum Number of Storage Cells in SFP 968 Bounding Discharge Schedule Table 5.3.2 Minimum In-Core Hold Time (hr) 100 Fuel Discharge Rate 5 per hour Holtec Report HI-2022867 5-12 1196

Table 5.3.2 Offload Schedule Cycle Offload Date Number of Average Initial 235U Assembly U Number (mmlddlyyyy) Assemblies Bumup Enrichment Weight (kgU)

(MWdIMTU) (Wt.%)(4) 1 01101/2002(1) 76 75,000 4.00 463.8 2 01/0112004 76 75,000 4.00 463.8 3 01/01/2006 76 75,000 4.00 463.8 4 01101/2008 76 75,000 4.00 463.8 5 01/0112010 76 75,000 4.00 463.8 6 01/01/2012 76 75,000 4.00 463.8 7 01/01/2014 76 75,000 4.00 463.8 8 01/01/2016 76 75,000 4.00 463.8 9 01/01/2018 76 75,000 4.00 463.8 10 01/01/2020 76 75,000 4.00 463.8 11 01/01/2022 76 75,000 4.00 463.8 12 01/01/2024 76 or 177(2 75,000 4.00 463.8 13 01/01/2026 76 or 017) 75,000 4.00 463.8 Table Notes:

(1) Dates are arbitrarily set to yield two years between offloads. While historic (ca. 2002) offloads were on 18-month cycles, the use of the longer24-month cycle will have a negligible impacton the total SFP heat load. This Is due to the use of bounding burnups and initial enrichments for the historic offloads, as well as the extremely long cooling times for these fuel assemblies at the point in time where the SFP becomes filled.

(2) 76 assemblies for partial core offload (previously discharged fuel), 177 assemblies for full core offload (recently discharged fuel).

(3) 76 assemblies for partial core offload (recently discharged fuel), 0 assemblies for full core offload.

(4) Initial enrichments may be as high as 5.0 wt.%. For a given bumup, a lower enrichment will yield a higher calculated decay heat. Thus, the use of 4.0 wt.% is conservative for the purposes of the thermal analysis.

-ii-LJ22b7 5-13 Honec eporz i-ioirac Report HI-2022867 5-13 1196

Table 5.4.1 Key Input Data for Time-To-Boil Evaluation SFP Surface Area 1012 ft2 Minimum Pool Water Depth 37.0 feet SFP Net Water Volume 34,340 ft3 Note: The net water volume is the gross water volume (i.e., area tmes depth) minus the volume displaced by the fuel racks and stored fuel assemblies.

5-14 119 multut; report ril-4uzzaa( 5-14 1196

Table 5.5.1 Key Input Data for Local Temperature Evaluation Axial Peaking Factor 1.65 Number of Fuel Assemblies 968 Cooled SFP Water Flow Rate 1000 gpm*

through SFPCS Heat Exchanger Fuel Assembly Type B&W 15x15 Fuel Rod Outer Diameter 0.430 inches Active Fuel Length 140.6 inches Number of Rods per Assembly 208 rods Rack Cell Inner Dimension 8.97 inches Rack Cell Length 162 inches Modeled Bottom Plenum Height 3 inches

Conservatively, only one pump flow is credited In the analysis.

- Conservatively, the lowerbound value for the active fuel length for ANO-1 fuel assemblies is used in the analysis.

Holtec Report HI-2022867 5-15 1196

Table 5.7.1 Result of SFP Decay Heat Calculations Heat Load Component Partial Core Offload Value Full Core Offload Value (Btu/hr) (Btulhr)

Previously Discharged Fuel 5.739x106 5.462x1006 Recently Discharged Fuel at 15.598x10 6 34.014x106 End of Transfer Total Bounding Decay Heat 21.337x106 39

.47 6x106 SFP Pump Heat (2 operating) 0.2040106 O.204x1V Total Bounding SFP Heat 21.541x106 39.680x106 Holtec Report HI-2022867 5-16 1196

Table 5.7.2 Results of Loss-of-Forced Cooling Evaluations Calculate Result Parameter Partial Core Offload Value Full Core Offload Value Minimum Time-to-Boil 8.87 hours0.00101 days 0.0242 hours 1.438492e-4 weeks 3.31035e-5 months 3.23 hours2.662037e-4 days 0.00639 hours 3.80291e-5 weeks 8.7515e-6 months Maximum Boiloff Rate 45.84 gallons per minute 84.80 gallons per minute Minimum Time for Water to 63.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months 34.3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months Drop to Top of Racks Moi ec Ke or- -I--- - - s 17--

Holtec Report Hl-Z202287 5-17 1196

Table 5.7.3 Results of Maximum Local Water and Fuel Cladding Temperature Evaluations Parameter Value Peak Local Water Temperature 1 192 OF Peak Cladding Superheat 31.50 F Peak Local Fuel Cladding Temperature 223.50 F Holtec Report Hl-2022867 5-18 1196

Grid I Feb 15, 2062' FLUENT 5.1 (3d, segregated; ke)

Figure 5.5.1 Schematic of the CFD Model of the ANO-1 SFP. segregated; ke)

Holtec Report H1-2022867 5-19 1196

Figure 5.7.1 - Partial Core Offload Bounding Spent Fuel Pool Heat Load (including previouslyand recently discharged fuel and 2 SFP pumps) 0 40 80 120 160 200 240 280 320 360 400 440 480 Time After Reactor Shutdown (hr) 1196nrffi HAlt-r tlnssnM:u I..IplUv "I-d;pu em Ixt~ 5-20 1196

Figure 5.7.2 - Full Core Offload Bounding Spent Fuel Pool Heat Load (including previously and recently discharged fuel and 2 SFP pumps) 4.0 E+07 In._ _ And --

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6.0 STRUCTURALUSEISMIC CONSIDERATIONS

6.1 INTRODUCTION

This section considers the structural adequacy of the existing spent fuel racks with new poison inserts under all loadings postulated for normal and seismic conditions at the Arkansas Nuclear One (ANO) Unit 1 reactor. The module layout is illustrated in Figure 6.1.1, along with the X and Y coordinate axes used to identify displacement orientation.

The analyses undertaken to confirm the structural integrity of the racks and the poison inserts, are performed in compliance with the OT Position Paper [6.1.2], ANO Specification No. AP&L-C-502 [6.1.3], and the ANO Unit I SAR [6.1.4]. An abstract of the methodology, modeling assumptions, key results, and summary of the parametric evaluation is presented. Delineation of the relevant criteria is discussed in the text associated with each analysis.

6.2 OVERVIEW OF RACK STRUCTURAL ANALYSIS METHODOLOGY The response of a freestanding rack module to seismic Inputs is highly nonlinear and involves a complex combination of motions (sliding, rocking, twisting, and turning), resulting in potential impacts and friction effects. Some of the unique attributes of the rack dynamic behavior include a large fraction of the total structural mass in a confined rattling motion, friction support of rack pedestals against lateral motion, and large fluid coupling effects due to deep submergence and independent motion of closely spaced adjacent structures.

Linear methods, such as modal analysis and response spectrum techniques, cannot accurately simulate the structural response of such a highly nonlinear structure to seismic excitation. An accurate simulation is obtained only by direct integration of the nonlinear equations of motion with the three pool slab acceleration time-histories applied as the forcing functions acting simultaneously.

Whole Pool Multi-Rack (WPMR) analysis is the vehicle utilized in this project to simulate the dynamic behavior of the complex storage rack structures. The following sections provide the basis for this selection and discussion on the development of the methodology.

Holtec Report HI-2022867 6-1 1196

6.2.1 Backaround of Analysis Methodoloqy Reliable assessment of the stress field and kinematic behavior of the rack modules calls for a conservative dynamic model incorporating all key attributes of the actual structure. This means that the model must feature the ability to execute the concurrent motion forms compatible with the freestanding installation of the modules.

The model must possess the capability to effect momentum transfers which occur due to rattling of fuel assemblies inside storage cells and the capability to simulate lift-off and subsequent impact of support pedestals with the pool liner (or bearing pad). The contribution of the water mass in the interstitial spaces around the rack modules and within the storage cells must be modeled in an accurate manner, since erring in quantification of fluid coupling on either side of the actual value is no guarantee of conservatism.

The Coulomb 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 analyzing this effect. In fact, a review of the results of rack dynamic analyses in numerous dockets (Table 6.2.1) indicates that an upper bound value of the coefficient of friction often maximizes the computed rack displacements as well as the equivalent elastostatic stresses.

In short, there are a large number of parameters with potential influence on the rack kinematics.

The comprehensive structural evaluation must deal with all of these without sacrificing conservatism.

The three-dimensional single rack dynamic model introduced by Holtec International in the Enrico Fermi Unit 2 rack project (ca. 1980) and used in some 50 rerack projects since that time (Table 6.2.1) addresses most of the above mentioned array of parameters. The details of this methodology are also published in the permanent literature [6.2.1]. Despite the versatility of the 3-D seismic model, the accuracy of the single rack simulations has been suspect due to one key element; namely, hydrodynamic participation of water around the racks. During dynamic rack motion, hydraulic energy is either drawn from or added to the moving rack, modifying its submerged motion in a significant manner. Therefore, the dynamics of one rack affects the motion of all others in the pool.

i-iaiiec i-ceport H11-2022u867 HI-UZb7 5-2 1196 iontec Report 6-2 1196

A dynamic simulation, which treats only one rack, or a small grouping of racks, is intrinsically inadequate to predict the motion of rack modules with any quantifiable level of accuracy. Three-dimensional Whole Pool Multi-Rack analyses carried out on several previous plants demonstrate that single rack simulations may under predict rack displacement during seismic responses

[6.2.2].

Briefly, the 3-D rack model dynamic simulation, involving one or more spent fuel racks, handles the array of variables as follows:

Interface Coefficient of Friction: Parametric runs are made with upper bound and lower bound values of the coefficient of friction. The limiting values are based on experimental data which has been found to be bounded by the values 0.2 and 0.8. Simulations are also performed with the array of pedestals having randomly chosen coefficients of friction in a Gaussian distribution with a mean of 0.5 and lower and upper limits of 0.2 and 0.8, respectively. In the fuel rack simulations, the Coulomb friction interface between rack support pedestal and liner is simulated by piecewise linear (friction) elements. These elements function only when the pedestal is physically in contact with the pool liner or bearing pad.

Rack Beam Behavior: Rack elasticity, relative to the rack base, is included In the model by introducing linear springs to represent the elastic bending action, twisting, and extensions.

Impact Phenomena: Compression-only gap elements are used to provide for opening and closing of interfaces such as the pedestal-to-bearing pad interface, and the fuel assembly-to-cell wall interface. These Interface gaps are modeled using nonlinear spring elements. The term nonlinear spring" is a generic term used to denote the mathematical representation of the condition where a restoring force is not linearly proportional to displacement.

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Fuel Loading Scenarios: The fuel assemblies are conservatively assumed to rattle in unison which exaggerates the contribution of impact against the cell wall.

Fluid Coupling: Holtec International extended Fritz's classical two-body fluid coupling model to multiple bodies and utilized it to perform the first two-dimensional multi-rack analysis (Diablo Canyon, ca. 1987). Subsequently, laboratory experiments were conducted to validate the multi-rack fluid coupling theory. This technology was incorporated in the computer code DYNARACK

[6.2.4] which handles simultaneous simulation of all racks in the pool as a Whole Pool Multi-Rack 3-D analysis. This development was first utilized in Chinshan, Oyster Creek, and Shearon Harris plants [6.2.1, 6.2.3] and, subsequently, in numerous other rerack projects. The WPMR analyses have corroborated the accuracy of the single rack 3-D solutions in predicting the maximum structural stresses, and also serve to improve predictions of rack kinematics.

For closely spaced racks, demonstration of kinematic compliance is verified by including all modules in one comprehensive simulation using a WPMR model. Additional more conservative single rack analyses are performed to confirm kinematic stability under the most adverse conditions such as fuel loading eccentricities and interim reracking configurations. In WPMR analysis, all rack modules are modeled simultaneously and the coupling effect due to this multi-body motion is included in the analysis. Due to the superiority of this technique in predicting the dynamic behavior of closely spaced submerged storage racks, the Whole Pool Multi-Rack analysis methodology is used for this project.

Holtec Report HI-2022867 6-4 1196

6.3 DESCRIPTION

OF RACKS The storage racks are analyzed as follows:

RACK WEIGHT DATA Dry Weight Total Weight

of Poison of Rack wt% of Rack plus Rack # Array Size # of Cells Inserts Poison Poison Inserts (lb) Inserts" (lb) 1 1i x 1 121 0 17,650 17,650 2 11 x I1 121 0 17,650 17,650 3 11 x 12 132 241 19,150 43,250 4 i 0 xI 110 0 27,650t 27,650 5 11 x I1 121 0 17,650 17,650 6 11 x 1 121 0 17,650 17,650 7 11 x 12 132 241 19,150 43,250 8 1lxil lID 0 27x6510 27,650 I Contains Boraflex.

ttPoison inserts are conservatively assumed to weigh 100 lb each. The actual weight of a poison insert plus a lead-in device is approximately 43 lb total.

For the purpose of modeling, the racks are numbered I through 5 as shown in Figure 6.1.1.

The Cartesian coordinate system utilized within the rack dynamic model has the following nomenclature:

x = Horizontal axis along plant South y = Horizontal axis along plant East z = Vertical axis upward from the rack base MOJtC Hi-UZ2tii xeport Hl-Z0ZZ1867 5-5 1196 Holtec Report 6-5 1196

6.4 SYNTHETIC TIME-HISTORIES The synthetic time-histories in three orthogonal directions (N-S, E-W, and vertical) are generated in accordance with the provisions of SRP [6.1.2], Section 3.7.1. In order to prepare an acceptable set of acceleration time-histories, Holtec International's proprietary code GENEQ

[6.4.1] is utilized.

A preferred criterion for the synthetic time-histories in SRP 3.7.1 calls for both the response spectrum and the power spectral density corresponding to the generated acceleration time-history to envelope their target (design basis) counterparts with only finite enveloping infractions.

The time-histories for the pools have been generated to satisfy this preferred criterion. The seismic files also satisfy the requirements of statistical independence mandated by SRP 3.7.1.

Figures 6.4.1 through 6.4.3 provide plots of the time-history accelerograms which were generated over a 20 second duration for the DBE event. Figures 6.4.4 through 6.4.6 provide plots of the time-history accelerograms which were generated over a 20 second duration for the OBE event. These artificial time-histories are used in all non-linear dynamic simulations of the racks.

Results of the correlation function of the three time-histories are given in Table 6.4.1. Absolute values of the correlation coefficients are shown to be less than 0.15, indicating that the desired statistical independence of the three data sets has been met.

65 WPMR METHODOLOGY Recognizing that the analytical work effort must deal with both stress and displacement criteria, the sequence of model development and analysis steps that are undertaken are summarized in the following a Prepare 3-D dynamic models suitable for a time-history analysis of the maximum density racks. These models include the assemblage of all rack modules in each pool. Include all fluid coupling interactions and mechanical coupling appropriate to performing an accurate non-linear simulation. This 3-D simulation is referred to as a Whole Pool Mulli-Rack model.

L_lt4A_ Da.a LolOn.X:

nobleut mepoltrlm-Lu~zzoo F64 1196

b. Perform 3-D dynamic analyses on various physical conditions (such as coefficient of friction and extent of cells containing fuel assemblies). Archive appropriate displacement and load outputs from the dynamic model for post-processing.

c. Perform stress analysis of high stress areas for the limiting case of all the rack dynamic analyses. Demonstrate compliance with ANO Specification No. AP&L-C-502 [6.1.3] and ANO Unit I SAR [6.1.4] limits on stress and displacement.

6.5.1 Model Details for Spent Fuel Racks The dynamic modeling of the rack structure is prepared with special consideration of all nonlinearities and parametric variations. Particulars of modeling details and assumptions for the Whole Pool Multi-Rack analysis of racks are given in the following:

6.5.1.1 Assumptions

a. The fuel rack structure motion is captured by modeling the rack as a 12 degree-of-freedom structure. Movement of the rack cross-section at any height is described by six degrees-of-freedom of the rack base and six degrees-of-freedom at the rack top. In this manner, the response of the module, relative to the base-plate, is captured in the dynamic analyses once suitable springs are introduced to couple the rack degrees-of-freedom and simulate rack stiffness.

b. Rattling fuel assemblies within the rack are modeled by five lumped masses located at H, .75H, .5H, .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. The centroid of each fuel assembly mass can be located off-center, relative to the rack structure centroid at that level, to simulate a partially loaded rack.

c. Seismic motion of a fuel rack is characterized by random rattling of fuel assemblies In their individual storage locations. All fuel assemblies are assumed to move in-phase within a rack. This exaggerates computed dynamic loading on the rack structure and, therefore, yields conservative results.

ille cpr I22876719 Holtec Report HI-2022867 6-7 1196

d. Fluid coupling between the rack and fuel assemblies, and between the rack and wall, is simulated by appropriate inertial coupling in the system kinetic energy.

Inclusion of these effects uses the methods of [6.5.2, 6.5.3] for rack/assembly coupling and for rack-to-rack coupling.

e. Sloshing is negligible 20 feet below the surface of the pool, where the racks reside, and it is, therefore, neglected in the analysis of the rack.

f. Potential impacts between the cell walls of the racks and the contained fuel assemblies are accounted for by appropriate compression-only gap elements between 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 baseplate extensions, and the rack stiffnesses are chosen to simulate local structural detail,

9. Pedestals are modeled by gap elements In the vertical direction and as 'rigid links" for transferring horizontal stress. 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.

h. 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 conservative measure of fluid resistance to gap closure.

i. The model for the rack is considered supported, at the base level, on four pedestals modeled as non-linear compression only gap spring elements and eight piecewise linear friction spring elements. These elements are properly located with respect to the centerline of the rack beam, and allow for arbitrary rocking and sliding motions. In reality each rack has 14 pedestals. This assumption increases the load transmitted to a single pedestal, thereby resulting in higher pedestal stresses.

6-8 1196 rionec eport r-ioitec Report MJ-UZ1 Hl-Y20ZZ557 6-8 1196

j. The mass of the poison inserts, which is conservatively assumed as 100 lb per insert, is lumped with the mass of the rack. Although the model does not allow the poison insert to move inside the flux traps, the potential impact force between the poison inserts and the flux trap walls has a negligible effect on the dynamic behavior of the rack. Moreover, the potential impacts involving the poison inserts are bounded by the conservative accounting of the poison insert weight.

6.5.1.2 Element Details Figure 6.5.1 shows a schematic of the dynamic model of a single rack The schematic depicts many of the characteristics of the model including all of the degrees-of-freedom and some of the spring restraint elements.

Table 6.5.1 provides a complete listing of each of the 22 degrees-of-freedom for a rack model.

Six translational and six rotational degrees-of-freedom (three of each type on each end) describe the motion of the rack structure, Rattling fuel mass motions (shown at nodes 1', 2', 3', 4, and 5 in Figure 6.5.1) are described by ten horizontal translational degrees-of-freedom (two at each of the five fuel masses). The vertical fuel mass motion is assumed (and modeled) to be the same as that of the rack baseplate.

Figure 6.5.2 depicts the fuel to rack impact springs (used to develop potential impact loads between the fuel assembly mass and rack cell innerwalls) in a schematic isometric. Only one of the five fuel masses is shown in this figure. Four compression only springs, acting in the horizontal direction, are provided at each fuel mass.

Figure 6.5.3 provides a 2-D schematic elevation of the storage rack model, discussed in more detail in Section 6.5.3. This view shows the vertical location of the five storage masses and some of the support pedestal spring members.

Figure 6.5 4 shows the modeling technique and degrees-of-freedom associated with rack elasticity. In each bending plane a shear and bending spring simulate elastic effects [6.5.4].

Linear elastic springs coupling rack vertical and torsional degrees-of-freedom are also included in the model.

Figure 6.5.5 depicts the inter-rack impact springs (used to develop potential impact loads between racks or between rack and wall).

..Nonec OI.. Keor M-----------i 9 Report HI-20228637 6-9 1196

6.5.2 Fluid Coupling Effect In its simplest form, the so-called "fluid coupling effect" [6.5.2, 6.5.3] can be explained by considering the proximate motion of two bodies under water. If one body (mass mi) vibrates adjacent to a second body (mass m2), and both bodies are submerged in frictionless fluid, then Newton's equations of motion for the two bodies are:

(mi + Mlr) Al+ M12 A2= applied forces on mass m1 + 0 (X1 2)

M2 i Ai + (M2 + M22)A2 = applied forces on mass M2 + 0 (X22)

Al and A2 denote absolute accelerations of masses ml and in , respectively, and the notation 2

O(X2 ) denotes nonlinear terms.

M1 , M12, M21, and M22 are fluid coupling coefficients which depend on body shape, relative disposition, etc. Fritz [6.5.3] gives data for M, for various body shapes and arrangements. The fluid adds mass to the body (M11 to mass m1 ), and an inertial force proportional to acceleration of the adjacent body (mass M2). Thus, acceleration of one body affects the force field on another.

This force field is a function of inter-body gap, reaching large values for small gaps. Lateral motion of a fuel assembly inside a storage location encounters this effect. For example, fluid coupling behavior will be experienced between nodes 2 and 2* in Figure 6.5.1. The rack analysis also contains inertial fluid coupling terms, which model the effect of fluid in the gaps between adjacent racks Terms modeling the effects of fluid flowing between adjacent racks in a single rack analysis suffer from the inaccuracies described earlier. These terms are usually computed assuming that all racks adjacent to the rack being analyzed are vibrating in-phase or 1800 out of phase.

The WPMR analyses do not require any assumptions with regard to phase.

Rack-to-rack gap elements have initial gaps set to 100% of the physical gap between the racks or between outermost racks and the adjacent pool walls Holtec Report H1-2022867 6-10 1196

6.5.2.1 Multi-Body Fluid Coupling Phenomena During the seismic event, all racks in the pool are subject to the input excitation simultaneously.

The motion of each free-standing module would be autonomous and independent of others as long as they did not impact each other and no water were present in the pool. While the scenario of inter-rack impact is not a common occurrence and depends on rack spacing, the effect of water (the so-called fluid coupling effect) is a universal factor. As noted In Ref. [6.5.2, 6.5.4], the fluid forces can reach rather large values in closely spaced rack geometries. It is, therefore, essential that the contribution of the fluid forces be included in a comprehensive manner. This is possible only if all racks in the pool are allowed to execute 3-D motion in the mathematical model. For this reason, single rack or even multi-rack models involving only a portion of the racks in the pool, are inherently inaccurate. The Whole Pool Multi-Rack model removes this intrinsic limitation of the rack dynamic models by simulating the 3-D motion of all modules simultaneously. The fluid coupling effect, therefore, encompasses interaction between every set of racks in the pool, i.e., the motion of one rack produces fluid forces on all other racks and on the pool wails. Stated more formally, both near-field and far-field fluid coupling effects are included in the analysis.

The derivation of the fluid coupling matrix [6.5.5] relies on the classical inviscid fluid mechanics principles, namely the principle of continuity and Kelvin's recirculation theorem. While the derivation of the fluid coupling matrix is based on no artificial construct, it has been nevertheless verified by an extensive set of shake table experiments [6.5.5].

6.5.3 Stiffness Element Details Three element types are used in the rack models. Type 1 elements are linear elastic elements used to represent the beam-like behavior of the integrated rack cell matrix. Type 2 elements are the piece-wise linear friction springs used to develop the appropriate forces between the rack pedestals and the supporting bearing pads. Type 3 elements are non-linear gap elements, which model gap closures and subsequent impact loadings i.e, between fuel assemblies and the storage cell inner walls, and rack outer periphery spaces.

If the simulation model is restricted to two dimensions (one horizontal motion plus one vertical motion, for example), for the purposes of model clarification only, then Figure 6.5.3 describes the Ulnit,... M tA

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auli rwprlt ril-4UZZ001 6-11 1196

configuration. This simpler model is used to elaborate on the various stiffness modeling elements.

Type 3 gap elements modeling impacts between fuel assemblies and racks have local stiffness K, in Figure 6.5.3. Support pedestal spring rates Ks are modeled by type 3 gap elements. Local compliance of the concrete floor is included in Ks The type 2 friction elements are shown in Figure 6.5.3 as K,. The spring elements depicted in Figure 6.5.4 represent type 1 elements.

Friction at support/liner interface is modeled by the piecewise linear friction springs with suitably large stiffness K, 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 transient analysis, the current value of N (either zero if the pedestal has lifted off the liner, or a compressive finite value) is computed The gap element Ks, modeling the effective compression stiffness of the structure in the vicinity of the support, includes stiffness of the pedestal, local stiffness of the underlying pool slab, and local stiffness of the rack cellular structure above the pedestal.

The previous discussion is limited to a 2-D model solely for simplicity. Actual analyses incorporate 3-D motions.

6.5.4 Coefficients of Friction To eliminate the last significant element of uncertainty in rack dynamic analyses, multiple simulations are performed to adjust the friction coefficient ascribed to the support pedestal/pool bearing pad interface. These friction coefficients are chosen consistent with the two bounding extremes from Rabinowicz's data [6.5.1]. Simulations are also performed by imposing friction coefficients developed by a random number generator with Gaussian normal distribution characteristics. The assigned values are then held constant during the entire simulation in order to obtain reproducible results.t Thus, in this manner, the WPMR analysis results are brought closer to the realistic structural conditions.

t It is noted that DYNARACK has the capability to change the coefficient of friction at any pedestal at each instant of contact based on a random reading of the computer clock cycle. However, exercising this option would yield results that could not be reproduced.

Therefore, the random choice of coefficients is made only once per run.

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The coefficient of friction (p) between the pedestal supports and the pool floor is indeterminate.

According to Rabinowicz 16.5.1], results of 199 tests performed on austenitic stainless steel plates submerged in water show a mean value of p to be 0.503 with standard deviation of 0.125.

Upper and lower bounds (based on twice standard deviation) are 0.753 and 0.253, respectively.

Analyses are therefore performed for coefficient of friction values of 0.2 (lower limit) and 0.8 (upper limit), as well as for random friction values clustered about a mean of 0.5. The bounding values of p = 0.2 and 0.8 have been found to envelope the upper limit of module response in previous rerack projects 6.5.5 Governing Equations of Molion Using the structural model discussed in the foregoing, equations of motion corresponding to each degree-of-freedom are obtained using Lagrange's Formulation [6.5.4]. The system kinetic energy includes contnbutions from solid structures and from trapped and surrounding fluid. The final system of equations obtained have the matrix forn:

/Ml[t21 = [ QJ +r[GI

[d4]

where:

[M] - total mass matrix (including structural and fluid mass contributions). The size of this matrix will be 22n x 22n for a WPMR analysis (n = number of racks in the model).

q - the nodal displacement vector relative to the pool slab displacement (the term with q indicates the second derivative with respect to time, i.e., acceleration)

[G] - a vector dependent on the given ground acceleration

[Q] - a vector dependent on the spring forces (linear and nonlinear) and the coupling between degrees-of-freedom The above column vectors have length 22n. The equations can be rewritten as follows:

Holtec Report H--2022867 6-13 1196

[dt

[2 d m]

21 = [Ml[ -/[Q]

Q + XM M~ I-[G]

This equation set is mass uncoupled, displacement coupled at each instant in time. The numerical solution uses a central difference scheme built into the proprietary computer program DYNARACK [6.2.4]

6.6 STRUCTURAL EVALUATION OF SPENT FUEL RACK DESIGN 6.6.1 Kinematic and Stress Acceptance Criteria There are two sets of criteria to be satisfied by the rack modules:

a. Kinematic Criteria According to Section 3.8.5 of Ref. [6.1.1], 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 required for overturning Margin of Safety =

e 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. However, to be conservative, the OBE safety factor of 1.5 will be applied to the worst case displacements from DBE. This is conservative, since the displacement for the DBE simulations exceed those for the OBE simulation.

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, -.- - -1cO LinlA.C--uLCOV -14 1196

b. Stress Limit Criteria Stress limits must not be exceeded under the postulated load combinations provided herein.

6.6.2 Stress Limit Evaluations The stress limits presented below apply to the rack structure and are derived from the ASME Code,Section III, Article XVII [6.6.1]. Parameters and terminology are in accordance with the ASME Code. Material properties are obtained from the ASME Code Appendices [6.6.2], and are listed in Table 6.3.1. The yield and ultimate strengths are taken at 150 OF, which is the maximum allowable temperature for ANO Spent Fuel Pools. Per the "double contingency principle" [6.6.3],

two simultaneous accident events, such as the loss of Spent Fuel Pool cooling together with an earthquake, need not be evaluated. Therefore, 150 'F is an appropriate temperature for the determination of material properties.

(i) Normal & Upset Conditions (Level A & B)

a. Allowable stress in tension on a net section is:

F, = 0 6 Sy Where, Sy = yield stress at temperature, and Ft is equivalent to primary membrane stress.

b Allowable stress in shear on a net section is.

F, = .4 Sy

c. Allowable stress in compression on a net section is:

Fa=Sy kl) 444 ra) where kllr for the main rack body is based on the full height and cross section of LI-1.-- DA Ad- .u 4^N0S nosesU MUePOo rj1--ZvzZjC 6-15 1196

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. The following values are appropriate for the described end conditions:

1 (simple support both ends) 2 (cantilever beam)

% (clamped at both ends) r = radius of gyration of component

d. Maximum allowable bending stress at the outermost fiber of a net section, due to flexure about one plane of symmetry is:

Fb = 0.60 S. (equivalent to primary bending)

e. Combined bending and compression on a net section satisfies:

f.a+ C. fbr + CMV fby<

Fa D. Fb. Dy Fby where:

f. = Direct compressive stress in the section fbx = Maximum bending stress along x-axis fby = Maximum bending stress along y-axis Cm, = 0.85 CMY = 0.85 Dx = I (fa/F',O Dy = 1 - (f./F'y)

F',,,y = (n2 E)1(2.15 (klIr) X,)

E = Young's Modulus and subscripts x,y reflect the particular bending plane.

Unitor- "IAAaloona rlultu. nUpUIL fli-4us6oQi 6-16b 1196

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

fa + fhr + fby <J.O 0.6Sy F& Fby The above requirements are to be met for both direct tension and compression.

g. Welds Allowable maximum shear stress on the net section of a weld is given by:

F= 0.3 S, where S, is the weld material ultimate strength at temperature. For fillet weld legs in contact with base metal, the shear stress on the gross section is limited to

0. 4 Sy, where Sy is the base material yield strength at temperature (ii) Faulted Conditions (Level D)

Per ANO Specification No. AP&L-C-502 [6.1.3], the stress limits for DBE conditions are the stress limits for Level A and B, as defined above, multiplied by 1.5.

Exceptions to the above general multiplier are the following:

a) Stresses in shearshall not exceed 0.5Sf.

b) The maximum allowable stress In bending and tension is 0.9 SY..

c) For this licensing application, AISC N-690 is conservatively used for welds and the limit on weld shear stress is set as:

Weld Allowable = (0.3 x S,) x 1.4 Holtec Report HI-2022867 6-17 1196

6.6.3 Dimensionless Stress Factors For convenience, the stress results are in dimensionless form. Dimensionless stress factors are defined as the ratio of the actual developed stress to the specified limiting stress value.

Based on the stress limits of ASME Code Article XVII for Level A and Level D conditions, the limiting value of each stress factor is 1.0. For this evaluation, however, the stress limits defined in ANO Specification No. AP&L-C-502 [6.1.3] for Level D conditions are more restrictive than the ASME Code. As a result, the limiting values for the stress factors are less than 1.0 under DBE conditions. The following table provides the list of dimensionless stress factors along with the limiting values for OBE and DBE.

Dimensionless Stress Factors Limiting Value OBE DBE R= Ratio of direct tensile or compressive stress on a net section to its 1.0 0.75 allowable value (note pedestals only resist compression)

R2 = Ratio of gross shear on a net section in the x-direction to its 1.0 0.694 allowable value R3= Ratio of maximum bending stress due to bending about the x-axis 1.0 0.75 to its allowable value for the section R4 = Ratio of maximum bending stress due to bending about the y-axis to its allowable value for the section 1.0 0.75 Rs = Combined flexure and compression factor (as defined in Section 3.4.1.5 above) 1.0 0. 75 R6 = Combined flexure and tension (or compression) factor (as defined in Section 3.4.1.6 above) 1.0 10.75 R7 = Ratio of gross shear on a net section in the y-direction to its allowable value 1.0 10.694 Holtec Report HI-2022867 6-18 1196

6.6.4 Loads and Loadina Combinations for SDent Fuel Racks The applicable loads and their combinations, which are considered in the seismic analysis of rack modules, are excerpted from the OT Position [6.1.2], ANO Specification No. AP&L-C-502

[6.1.3], the Unit 1 SAR, and Section 3.8.4 of the USNRC Standard Review Plan (SRP)

[6.1.1].

The SRP load combinations are not required to be met by ANO. However, for thoroughness the load equations outlined in SRP 3.8.4 are used. The load combinations considered are identified below:

Loading Combination Service Levelt Section Reference D+L Level A 6.11 D + L + T. 6.11, 6.13.2 D + L + T, + E 6.8, 6.9, 6.13.2 D + L + T, + E Level B 6.13.2 D + L + T. + Pr 614 D + L+T,+E' LevelD 6.8,6.9 D + L + T. + Fd The functional capability of the 7.0 fuel racks must be demonstrated.

The allowable stress limits for Service Levels A, B, and D are given in Section 6 6.2.

Where:

D = Dead weight-induced loads L = Live Load (including stored fuel assemblies, poison inserts, and miscellaneous equipment loads)

Pr = Upward force on the racks caused by postulated stuck fuel assembly Fd = Impact force from accidental drop of the heaviest load from the maximum possible height.

E Operating Basis Earthquake (OBE)

E' = Design Basis Earthquake (DBE)

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

T, = Differential temperature induced loads (the highest temperature associated with the postulated abnormal design conditions) Ta and T, produce local thermal stresses. The worst thermal stress field in a fuel Holtec Report HI-2022867 6-19 1196

rack is obtained when an isolated storage location has a fuel assembly generating heat at maximum postulated rate and surrounding storage locations contain no fuel. Heated water makes unobstructed contact with the inside of the storage walls, thereby producing maximum possible temperature difference between adjacent cells. Secondary stresses produced are limited to the body of the rack; that is, support pedestals do not experience secondary (thermal) stresses.

6.7 PARAMETRIC SIMULATIONS The multiple rack models employ the fluid coupling effects for all racks in the pool, as discussed above, and these simulations are referred to as WPMR evaluations. In addition, single rack models are also developed for additional study of the effect of various parameters on rack displacement. The models are described as follows:

(I) Whole Pool Multi Rack Model An array of eight racks is modeled with proper interface fluid gaps and a coefficient of friction at the support interface locations with the bearing pad generated by a Gaussian distribution random number generator with 0.5 as the mean and 0.15 standard deviation. The response to both DBE and OBE seismic excitation is determined.

(11) Single Rack Model A single rack model is employed to study the effect of top loading the rack with miscellaneous equipment, which represents a future storage possibility. Rack number 3 is chosen for this simulation because it is one of only two racks in the pool, both of identical size, that have poison inserts. The top loaded rack simulation (Case

7) is performed using the 0.8 coefficient of friction and the DBE event since this combination generally produces the largest rack top displacements. A fictitious 2,000 Ibf mass, with three translational degrees of freedom, is rigidly attached to the rack 24" above the top of the cell structure. The displacements calculated from the single rack run are used as a further check for kinematic stability.

The Whole Pool and Single Rack simulations listed in the following table have been performed to Holtec Report HI-2022867 6-20 1196

investigate the structural integrity of the rack array, including the new poison inserts.

LIST OF WPMR AND SINGLE RACK SIMULATIONS Case Model Load Case COF Event 1 WPMR All racks fully loaded Random OBE 2 WPMR All racks fully loaded 0.2 OBE 3 WPMR All racks fully loaded 0.8 OBE 4 WVPMR All racks fully loaded Random DBE 5 WPMR All racks fully loaded 0.2 DBE 6 WPMR All racks fully loaded 0.8 DBE 7 iFully loaded rack 7 Single w 2,000 lb load overhead 0.8 DBE where Random = Gaussian distribution with a mean coeff. of friction of 0.5.

(upper and lower limits of 0.8 and 0.2, respectively)

COF = Coefficient of Friction 6.8 TIME HISTORY SIMULATION RESULTS The results from the DYNARACK runs may be seen 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 and develop the stress factors discussed in subsection 6.6.3. 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. This section also summarizes additional analyses performed to develop and evaluate structural member stresses which are not determined by the post processor.

6.8.1 Rack Displacements The maximum rack displacements are obtained from the time histories of the motion of the upper and lower four corners of each rack In each of the simulations. The maximum absolute value of displacement in the two horizontal directions, relative to the pool slab, is determined by the post-Ml.. .- I It -j - F n

-h r-IUILM; MUPWI rjJ-ZtjgZ00( 6-21 1196

processor for each rack, at the top and bottom comers. The maximum displacements in either direction reported from the WPMR analyses are 0.726" at the top of Rack 4 during the DBE events and 0.384" at the top of Rack 4 during the OBE events. The maximum displacement in either direction reported from the single rack analysis is 0.236", which was performed for Rack 3.

It is obvious from these small displacements at the top of the racks that the safety factors for tipping are met.

6.8.2 Pedestal Vertical Forces The maximum vertical pedestal force obtained in the WPMR simulations is 203,000 lbf for Rack 4 under DBE conditions. The maximum vertical pedestal force obtained in the OBE simulation is 138,000 Ibf for Rack 8.

6.8.3 Pedestal Friction Forces The maximum interface shear force value in any direction bounding all pedestals in the WPMR simulations is 89,700 Ibf for Rack 3 in Case 6.

6.8.4 Rack Impact Loads A freestanding rack, by definition, is a structure subject to potential impacts during a seismic event. Impacts arise from rattling of the fuel assemblies in the storage rack locations and, in some instances, from localized impacts between the racks, or between a peripheral rack and the pool wall. The following sections discuss the bounding values of these impact loads.

6.8 4.1 Rack to Rack Impacts Gap elements track the potential for impacts between adjacent racks. The results for each simulation have been scanned for non-zero impact forces. The simulation results show that no gap element between any two racks closes. Thus, there are no rack-to-rack impacts Holtec Report Hl-2022867 6-22 1196

6.8.4.2 Rack to Wall Impacts The storage racks do not impact the pool walls under any simulation.

6.8.4.3 Fuel to Cell Wall Impact Loads A review of all simulations performed allows determination of the maximum instantaneous impact load between fuel assembly and fuel cell wall at any modeled impact site. The maximum fuel/cell wall impact loads are 1263 lbf in Rack 4 in the DBE case of the WPMR analyses and 853 lbf for the OBE case in Rack 4. The cell wall integrity under this instantaneous impact load has been evaluated and shown to remain intact with no permanent damage.

The permissible lateral load on an irradiated spent fuel assembly has been studied by the Lawrence Livermore National Laboratory. The LLNL report [6.8.1] 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 most significant load on the fuel assembly arises from rattling during the seismic event. For the five lumped mass model, the limiting lateral load, therefore, is equal to Fe, where Fe= (w x a)14 where:

w= weight of one fuel assembly (upper bound value = 1700 Ibs) a= permissible lateral acceleration in g's (a = 63)

Therefore, Fe = 26,775 lbs The maximum fuel-to-storage cell rattling force from the WPMR runs is 1263 lbs. Therefore, the nominal factor of safety against fuel failure is roughly equal to 21.

Holtec Report HJ-2022867 6-23 1196

6.9 RACK STRUCTURAL EVALUATION 6.9.1 Rack Stress Factors The time history results from the DYNARACK solver 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. In particular, maximum values for the previously defined stress factors are determined for every pedestal in the array of racks. With this information available, the structural integrity of the pedestal can be assessed and reported. The net section maximum (in time) 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 in the limiting rack cell (box) can be evaluated.

The stress factor results for male and female pedestals, and for the entire spent fuel rack cellular cross section just above the baseplate have been determined. These factors are reported for every rack in each simulation, and for each pedestal in every rack. These locations are the most heavily loaded net sections in the structure so that satisfaction of the stress factor criteria at these locations ensures that the overall structural criteria set forth in Section 6.6 are met.

The maximum pedestal stress factor for OBE is 0.221, which occurs in Rack 7 during Case 3, and for DBE is 0.216, which occurs in Rack 4 during Case 6. The maximum cell wall stress factor is computed to be 0.344 for OBE in Rack 7 during Case 3 and 0.245 for DBE in Rack 2 during Case 6. An evaluation of the stress factors, for all of the simulations performed, leads to the conclusion that all stress factors are less than the limits specified in Section 6.6.3.

Therefore, the requirements of Section 6.6.2 are indeed satisfied for, the load levels considered for every limiting location in the rack.

6.9.2 Pedestal Thread ShearStress From the WPMR simulations, the maximum thread engagement stresses under seismic conditions for every pedestal for every rack In the pool are 12,843 psi under DBE conditions and 8,730 psi under OBE conditions. The yield stress for the pedestal material is 27,500 psi. The allowable shear stress for Level B conditions is 0.4 times the yield stress, which gives 11,000 psi. Since 8,730 psi (occurs during OBE run) is less than the OBE allowable of 11,000 psi, the U14-ia F U1r wlo~s;:

IulJUL6 NDIAML "I-eu adsQ b-24 1196

female pedestal threads are shown to be acceptable. The maximum shear stress in the pedestal thread under DBE condition is 12,843 psi, which is less than 0.5Sv (or 13,750 psi).

6.9.3 Local Stresses Due to Impacts Impact loads at the pedestal base (discussed in subsection 6.8.4.1) produce stresses in the pedestal for which explicit stress limits are prescribed in the Code. However, impact loads on the cellular region of the racks, as discussed in subsection 6.8.4.3 above, produce stresses which attenuate rapidly away from the loaded region. This behavior is characteristic of secondary stresses.

Even though limits on secondary stresses are not prescribed in the Code for class 3 NF structures, evaluations are made to ensure that the localized impacts do not lead to plastic deformations in the storage cells which affect the sub-criticality of the stored fuel array.

a. Impact Loadinn Between 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 which would lead to gross permanent deformation The limiting impact load of 2,436 lbf (including a safety factor of 2.0) is much greater than the highest calculated impact load value of 1263 lbf (see subsection 6.8.4.3) obtained from any of the rack analyses. Therefore, fuel impacts do not represent a significant concern with respect to fuel rack cell deformation.

b. Impacts Between Adiacent Racks As may be seen from subsection 6.8.4.1, no impacts are predicted between adjacent racks.

6.9.4 Weld Stresses 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.

6-25I-- 1-1----

"nudms rwpvior Ill-ZUZdu~t 6-25 1196

a. Baseplate-to-Rack Cell Welds For Level A or B conditions, Ref. [6.6.1] permits an allowable weld stress of - = .3 S, =

21903 psi. As stated in subsection 6.6.2, the allowable may be increased for Level D by an amplification factor which is equal to 1.4.

Weld dimensionless stress factors are produced through the use of a simple conversion (ratio) factor applied to the corresponding stress factor in the adjacent rack material. The ratio 0.819 is developed from the differences In material thickness and length versus weld throat dimension and length:

RATIO = (Cell wall thickness

Avg. Cell Wall Length)J(Effective Weld Size

Weld Length)

= (0.062 ' 8.97)/ (0.12

0.7071

8.0)

The highest predicted cell to baseplate weld stress is calculated based on the highest R6 value for the rack cell region tension stress factor and R2 and R7 values for the rack cell region shear stress factors. Refer to subsection 6.6.3 for definition of these factors.

These cell wall stress factors may be converted into weld stress values as follows:

[R6 * (1.2) + R2 * (0.72) + R7 * (0.72)]

S

Ratio =

[0.245 * (1.2) + 0.027 * (0.72) + 0.029 (0.72)]

27,500 - 0.819 = 7,530 psi This calculation is conservative for the following reasons:

1) The directional stresses associated with the normal stress -r and the two shear stresses and T.should be combined using SRSS instead of direct summation.

'y

2) The maximum stress factors used above do not all occur at the same time instant, in the same storage rack, or during the same simulation.

The DBE condition governs since the ratio of the DBE weld stress to the OBE weld stress fl.1.--.n

- - L11fflme rVpU1L fI-ULZZoI "U1i0L, O-Zb 1196

is greater than 1.4 (see paragraph 6.6.2(ii)(c)). The maximum weld stress is less than the DBE allowable for base metal shear stress, which is 13,750 psi. Therefore, the welds are acceptable, with a safety factor of (I 3,750/7,530) = 1.83

b. Baseplate-to-Pedestal Welds The weld between baseplate and support pedestal is checked using finite element analysis to determine the maximum weld stresses. The maximum weld stresses under OBE and DBE conditions are 12,150 and 26,920 psi, respectively. These calculated stress values are below the corresponding OBE and DBE weld allowables of 21,900 psi and 30,660 psi, respectively. Therefore, these welds have been determined to be acceptable The maximum shear stresses in the metal adjacent to the weld are 1,736 psi and 856 psi, respectively, under DBE and OBE conditions. Both of these calculated stress values are well below OBE metal allowable of 11,000 psi.

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

Stresses in storage cell to cell welds develop due to fuel assembly impacts with the cell wall, These weld stresses are conservatively calculated by assuming that fuel assemblies in adjacent cells are moving out of phase with one another so that impact loads in two adjacent cells are in opposite directions; this tends to separate the two cells from each other at the weld.

Table 6.9.1 gives the computed results for the maximum allowable load that can be transferred by these welds based on the available weld area. The upper bound on the applied load transferred is also given in Table 6.9.1. This upper bound value is very conservatively obtained by applying the bounding rack-to-fuel impact load from any simulation in two orthogonal directions simultaneously, and multiplying the result by 2 to account for the simultaneous impact of two assemblies In adjacent cells moving in opposing directions. An equilibrium analysis at the connection then yields the upper bound load to be transferred. As shown in Table 6.9.1, the calculated shear stress in the base metal adjacent to the weld is 7,363 psi under DBE condition and 4,997 psi under U1l2I n - 196 MURKo;e MUpOIE trl-,Ued-'U1 6-27 1196

OBE conditions. These values are less than the allowable ORE and DBE stress limits of 11,000 psi and 13,750 psi, respectively.

6.10 POISON INSERT STRUCTURAL EVALUATION A structural evaluation will be performed to demonstrate that the stresses in the poison insert, under normal and accident conditions, meet the appropriate stress limits from ANO Specification No. AP&L-C-502 [6.1.3] and the ANO Unit I SAR [6.1.3]. The design details of the poison insert are provided in Section 2.5. The minimum calculated safety factor for the poison insert for all loading conditions will be greater than 1 0.

6.11 LEVEL A EVALUATION The stress allowables are the same for Level A and Level B conditions. The Upset (ORE) condition controls over normal (Gravity) condition. Therefore, no further evaluation is required for these load cases.

6.12 HYDRODYNAMIC LOADS ON POOL WALLS The hydrodynamic pressures that develop between adjacent racks and the pool walls can be developed from the archived results produced by the WPMR analysis. Of the racks next to the SFP walls, the one that resulted in the maximum displacement generates the maximum hydrodynamic load on its adjacent wall. The maximum hydrodynamic pressure is considered as an individual load in the structure qualification of the spent fuel pool. The pressure plots on the four walls of the SFP at the time of maximum (in absolute value) instantaneous hydrodynamic pressure for the OBE and DBE events are shown in Figure 6.1 1.1.

6.13 LOCAL STRESS CONSIDERATIONS This section presents the results of evaluations for the possibility of cell wall buckling and the secondary stresses produced by temperature effects.

LJ-t*AP MA -- l . Q71-nULug.; reupJUi n l-2uLLOr( 6-28 1196

6.13.1 Cell Wall Buckling The allowable local buckling stresses in the fuel cell walls are obtained by using classical plate buckling analysis. The evaluation for cell wall buckling is based on the applied stress being uniform along the entire length of the cell wall In the actual fuel rack, the compressive stress comes from consideration of overall bending of the rack structures during a seismic event, and as such Is negligible at the rack top, and maximum at the rack bottom.

The critical buckling stress is determined to be 4,663 psi. The average compressive stress in the cell wall, based on the R6 stress factor, is 2,838 psi. Therefore, there is a 64.3% margin of safety against local cell wall buckling.

6.13.2 Thermal Loads The rack Is freestanding; thus, there is minimal or no restraint against free thermal expansion of the rack. The high thermal conductivity of the stainless steel material, the buoyancy driven flow of the SFP water, and the thin sheet metal construction tend to diminish the thermal gradients In the spent fuel racks. Thermal loads applied to the rack are, therefore, not included in the stress combinations.

6.14 EVALUATION OF POSTULATED STUCK FUEL ASSEMBLY The ability of the spent fuel racks to withstand the uplift forces due to a postulated stuck fuel assembly is also evaluated. Strength of materials formulas are used to determine the effects of a 5,000 lb force, which bounds the fuel handling crane's cut-off limit, applied at various locations on the storage cell. For a load applied vertically anywhere along a cell wall, the tensile stress is 8,991 psi, which is below the yield stress of the material. When the load is applied at a 45 degree angle to the top of a cell wall, the damaged region extends downward, from the top of the rack, no more than 2.59 inches, which is well short of the poison insert.

6.15 CONCLUSION Seven discrete freestanding dynamic simulations of maximum density spent fuel storage racks have been performed to establish the structural margins of safety. Of the seven parametric I-lr1#__ I~nat 1l 1 - -A7^_r "uLILLA, rtpJIL ni-4u440ooI 6s-29 1196

analyses, six simulations consisted of modeling all 8 fuel racks in the pool in one comprehensive Whole Pool Multi Rack (WPMR) model The remaining run was carried out with the classical single rack 3-D model. The parameters varied in the different runs consisted of the rack/pool liner interface coefficient of friction and the type of seismic input (DBE or OBE). Maximum (maximum in time and space) values of pedestal vertical, shear forces, displacements and stress factors have been post-processed from the array of runs and summarized in tables in this chapter. The results show that:

(i) All stresses are well below the specified allowable limits.

(ii) There is no rack-to-rack or rack-to-wall impact anywhere in the cellular region of the rack modules (iii) The factor of safety against overturning of a rack is in excess of 60.

In conclusion, all evaluations of structural safety, mandated by the OT Position Paper [6.1.2] and the contemporary fuel rack structural analysis practice have been carried out. They demonstrate consistently large margins of safety in all storage modules 6.16 REFERENCES (6.1.11 USNRC NUREG-0800, Standard Review Plan, June 1987.

[6.1.2] (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.

[6.1.3] Technical Specification for Earthquake Resistance Design of Structures and/or Components Located in the Auxiliary Building for the Arkansas Nuclear One Unit 1 Power Plant, Arkansas Power and Light Company, Little Rock, Arkansas, Specification Number AP&L-C-502, Revision 2.

[6 1.4] Arkansas Nuclear One Unit 1 Safety Analysis Report (SAR).

W-nfto,~ D- LII 13V1Q'

- NFGruI 11-oudoCof fa-Zi 1196

[6.2.1] Soler, A. I. and Singh, K. P., "Seismic Responses of Free Standing Fuel Rack Constructions to 3-D Motions", Nuclear Engineering and Design, Vol. 80, pp.

315-329 (1984).

[6.2.2] Soler, A. I. and Singh, K. P., "Some Results from Simultaneous Seismic Simulations of All Racks in a Fuel Pool", INNM Spent Fuel Management SeminarX, January, 1993.

[6.2.3] Singh, K. P. and Soler, A. I., "Seismic Qualification of Free Standing Nuclear Fuel Storage Racks - the Chin Shan Experience, Nuclear Engineering International, UK (March 1991).

[6.2.4] Holtec Proprietary Report HI-961465 - WPMR Analysis User Manual for Pre &

Post Processors & Solver, August, 1997.

[6.4.1] Holtec Proprietary Report HI-89364 - Verification and Users Manual for Computer Code GENEQ, January, 1990.

[6.5.1) Rabinowicz, E., "Friction Coefficients of Water Lubricated Stainless Steels for a Spent Fuel Rack Facility," MIT, a report for Boston Edison Company, 1976.

[6.5 2] Singh, K P. and Soler, A. I., "Dynamic Coupling in a Closely Spaced Two-Body System Vibrating in Liquid Medium, The Case of Fuel Racks," 3rd International Conference on Nuclear Power Safety, Keswick, England, May 1982.

[6.5.3] Fritz, R. J., "The Effects of Liquids on the Dynamic Motions of Immersed Solids," Journal of Engineering for Industry, Trans. of the ASME, February 1972, pp 167-172.

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

wnlrtne- WL-l2n-"QMa7 -

G AVISik l\G8UI L ale-- uL<oU 01i 1196

[6.5.5] Paul, B., "Fluid Coupling in Fuel Racks: Correlation of Theory and Experiment', (Proprietary), NUSCOIHoltec Report HI-88243

[6.6.1] ASME Boiler & Pressure Vessel Code, Section l1l, Article XVII, 1980 Edition with Winter 1981 Addenda.

[6.6.2] ASME Boiler & Pressure Vessel Code,Section III, Appendix I, Table 1-3, 1980 Edition with Winter 1981 Addenda.

[6.6.3] Kopp, L., "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light Water Reactor Power Plants," USNRC Internal Memorandum, 1998.

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

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

[6.9.2] ACI 318-95, Building Code requirements for Structural Concrete," American Concrete Institute, Detroit, Michigan, 1995.

[6.9.3] Holtec report H--89330, Rev. 1, 'Seismic Analysis of High Density Fuel Racks; Part IlIl: Structural Design Calculations - Theory."

[6.10.1] ASME Boiler& Pressure Vessel Code, Section 1II,Subsection NF, 1998 Edition.

Mote Keor _1-U2~ 6-3 1196__ @l monec Report F11-20228r37 6-32 1196

E;.X PLANT DOCKET NUMBER(s) YEAR Enrico Fermi Unit 2 USNRC 50-341 1980 Quad Cities I & 2 USNRC 50-254, 50-265 1981 Rancho Seco USNRC 50-312 1982 Grand Gulf Unit 1 USNRC 50-416 1984 Oyster Creek USNRC 50-219 1984 Pilgrim USNRC 50-293 1985 V.C. Summer USNRC 50-395 1984 Diablo Canyon Units 1 & 2 USNRC 50-275, 50-323 1986 Byron Units I & 2 USNRC 50-454, 50-455 1987 Braidwood Units 1 & 2 USNRC 50-456, 50-457 1987 Vogtle Unit 2 USNRC 50-425 1988 St. Lucie Unit 1 USNRC 50-335 1987 Millstone Point Unit 1 USNRC 50-245 1989 Chinshan Taiwan Power 1988 D.C Cook Units 1 & 2 USNRC 50-315, 50-316 1992 Indian Point Unit 2 USNRC 50-247 1990 Three Mile Island Unit 1 USNRC 50-289 1991 James A FitzPatrick USNRC50-333 1990 Shearon Harris Unit 2 USNRC 50-401 1991 Hope Creek USNRC 50-354 1990 Kuosheng Units 1 & 2 Taiwan Power Company 1990 Ulchin Unit 2 Korea Electric Power Co. 1990 Laguna Verde Units I & 2 Comision Federal de 1991

________ .Electricidad __. -

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v. r s v ......

>~ ~~ ~ ~ ~ ~ps8wi ~ ~~M vBl7t^.tR^

S:, - x,.--eze9 .^ :.A.- ?-- '*ro _x ov- ........ < s>O:.e.-8;t~~o'se':9 '{2 , .............

PLANT DOCKET NUMBER(s) YEAR Zion Station Units 1 & 2 USNRC 50-295, 50-304 1992 Sequoyah USNRC 50-327, 50-328 1992 LaSalle Unit I USNRC 50-373 1992 Duane Amold Energy Center USNRC 50-331 1992 Fort Calhoun USNRC 50-285 1992 Nine Mile Point Unit 1 USNRC 50-220 1993 Beaver Valley Unit I USNRC 50-334 1992 Salem Units 1 & 2 USNRC 50-272, 50-311 1993 Limerick USNRC 50-352, 50-353 1994 Ulchin Unit 1 KINS 1995 Yonggwang Units I & 2 KINS 1996 Kori-4 KINS 1996 Connecticut Yankee USNRC 50-213 1996 Angra Unit 1 Brazil 1996 Sizewell B United Kingdom 1996 Waterford 3 USNRC 50-382 1997 JA. Fitzpatrick USNRC 50-333 1998 Callaway USNRC 50-483 1998 Nine Mile Unit 1 USNRC 50-220 1998 Chin Shan Taiwan Power Company 1998 Vermont Yankee USNRC 50-271 1998 Millstone 3 USNRC 50-423 1998 ByronlBraidwood USNRC 50-454, 50-455, 1999 50-567, 50-457

- r - - 3JI a It - -

rIVILUL r-,pUJIL fnl-ZU4L01I 6-34 1196

U. ~, -.. 3.

A. 'b,. A -C.N-. MPM e..C ".33:33.

.N:pW l ick.A5 Iica. oi_________________

PLANT DOCKET NUMBER(s) YEAR Wolf Creek USNRC 50-482 1999 Plant Hatch Units 1 & 2 USNRC 50-321, 50-366 1999 Harris Pools C and D USNRC 50-401 1999 Davis-Besse USNRC 50-346 1999 Enrico Fermi Unit 2 USNRC 50-341 2000 Kewaunee USNRC 50-305 2001 I U Ilf L. rM-J- uI nizu epUorVtUI "IJ ~ L O r-11-4V=ZZ-0 11 IV 6-35 1196

Stainless Steel Young's Modulus Yield Strength Ultimate Strength Material E SY Su (psi) (psi) (psi)

SA240, Type 304 (cell boxes) 27.9 x 10 27,500 73,000 SA240, Type 304 (upper part 27.9 x 106 27,500 73,000 of support feet)

SA479, Type 304 (lower part 27.9 x 106 27,500 73,000 of support feet) tA-I&--

n'.anw r¶upurI I nilI-euL.oaI

1. ---

b-36 1196

OBE Datal to Data2 0.025 Datal to Data3 0.100 Data2 to Data3 0.032 DBE Datal to Data2 0.020 Datal to Data3 0.103 Data2 to Data3 0.034 Datal corresponds to the time-history acceleration values along the X axis (South)

Data2 corresponds to the time-history acceleration values along the Y axis (East)

Data3 corresponds to the time-history acceleration values along the Z axis (Vertical) 6-I1- 9 nUILOG MePDFE r1I-,eUZ;eUOf 6-37 1196

.1......

LOCATION _Node) DISPLACEMENT ROTATION Ur Uy uz Ox 1 PP2 P3 q4 q5 q6 2 P17 P1i Pig q20 _q21 q 22 Node 1 is assumed to be attached to the rack at the bottom most point.

Node 2 is assumed to be attached to the rack at the top most point.

Refer to Figure 6.5.1 for node Identification.

2 PT Ps 3 P9 Pic 4 Pi P12 5 P13 P14 1 Pis P16 where the relative displacement variables p, are defined as:

P. q.(t) + U.(t) i = 1.7,9,11,13,15,17 q,(t) + Uy(t) i = 2,8,10,12,14,16,18

= qi(t) + U(t) i = 3,19 q,(t) i = 4,5,6.20,21,22 p,denotes absolute displacement with respect to inertial space q,denotes relative rotation with respect to the floor slab 0(t) are the three known earthquake displacements

denotes fuel mass nodes

.ec

..Noltec e or H---------6--

Report HI-20221367 6-38 1196

A

........ ........ .1..1. '* .: 21. ' ';c '. - I .

V W : . ' .Xl-

.:Louisiana

'X' .'ASEAN,., .1 . Si 3XJ I J1 xz_ "

M.Par,IS#.Wofso. Ka" ._nf. - .. W.

. .. UN&;...

Al

.Kg RO., S.

DBEt Item/Location Calculated Allowable Rackibaseplate weld, psi 7,530 13,750 Female pedestal/baseplate weld, psi 26,920 30,660 Cellcell welds, psi tt 7,363 13,750 t DBE controls over OBE.

Wnlf=r~ D -V UJ. a

--- c -F-~ . IUxOp DtJ3d 1196

T I I . . I Ch-;ll I -1t:0, Cs.." -I . .

eJ-ltisA

+ *

@8 *" 4'

.,C5w26X _ . [T.1744.

117.373

.1 I* ia.a 4, I

+ OI1-4.J

___llit___I +t

'0'.

.OO I= it so J O M .

.4 ______

- -Glxo_. oa+ +

4,. Y

... I O. A cz1-wn go7

I. .

- s . .

Lx Figure 6.1.1 - ANO-1 Spent Fuel Pool Layout I-Ilfn &n 1&.1.4 LJO nn-I*n_-7

, . L~V il'JFJVIL nl-UzVf4Do 6-40 1196

-ANO 1.

UNIT 1, DBE, 2% DAMP.V[ING,¢ 362 ' ELEVATION 0 . 40 e -l AWUL.tA I ION 0.40

-J20 U

U

0 I j . I I I I

'TItlE So)

Figure 6.4.1 LUIWL F¶._ F flI-L- LL-- -4(

nonut; meporl rlFzuzzUt> 6-41 1196

ANO UNIT 1. DDE. 2% DAMPINIG, 262@ ELtEVATION

.N-S T IME ACCELERATION

<0 .

U.:

. .040- ;

TIME ( sn0 Figure 6.4.2 6-214-- 11"96 I ji nuderk; rweport rl-zuxzoct 6-42 1196

ANO UNIT 1, DBE, 2% DAMPING. 362' ELEVATI ON

0. 10 VENT TTME ACCELERATION

.- 4 ..

0

. li-e -,

fi -0 Il.-

W-..

Figure 6.4.3 I-N c .4 E3--- LISILnnnno7 S I IuIvLz; ixUurIL r-eU4400I 6-43 1196

ANO UiT 1, Q9Ei. 2z<DAMPING, 362"' ELEVAT ION E- W TINE ACCELER~ATION 0.0S0 . 0........... 1

e. 0. ...

IT-M

-0.20- I -t L l, t,4 tItt '1r[

Figure 6.4.4

_I.20 2 6 Re or HI-2022867

..te Report 6.44. 1196__

Holtec 6-44 1196

ANO UNIT 1, OBE,. .2% DAMPING 362' ELEVATION N-S TIME ACCELERATION.

Lr..

U.

ffi Edi a

-0.

20.00 Tl~t~ ( onc)

Figure 6.4.5 Holtec Report F... ova uu b=-4!5 1196 b4L

ANO UNlT 1, .OBE-,2% DAIMPING, 362' ELEVATION VERT.. TIME ACCELERATION..

0.04 5_ .

z .

D 0.00 w ,

I-l U

. - 0. 02

-. 4

-0. 08Y.. -

. - a.

TIMg (8qcv)

Figure 6.4.6 w.

Ppnnrtf WLWA9na.7 Hto!tpR

. . . . -- lV u)-4sv 1196

r FRICTIOM PEDEXIAL/POJNA TION EL 8EM~N - -AK,/ SAP ELEMENT-Figure 6.5.1 -Schematic of Single Rack Dynamic Model Used in Dynarack Holtec Report HI-2022867 6-47 1196

FUEL ASSEMBLY/CELL GAP/ELEMENT IMPACT Figure 6.5.2 - Fuel-to-Rack Gap/lmpact Elements at Level of Raffling Mass 6-48 1196nN,,Nt,

IUILtU. fXCUIL rni-ciZjo4O 6-48 11V9 6

FUEL ASS;/CELL:

GAP/ILPAVT ELEMT, Ki TYP~rfAy. ,FUEL ASS O RACK PERIPHERY GAP/IKPACT

.LE1n s. g-1/4 F 1 F771~TL.h'N:

GAPWIMPACT FRICTION - ELEMENT ,KS ITERFACE

- ELE#ENT.Kf '777 Figure 6.5.3 - Two Dimensional View of Spring-Mass Simulation 1- e as__ - I---e---- ._

ron~ec Report Hl-1Z0Z2867 5649 1196

H/2 1 .410 KSY Kux

-RACK DEGREES-OF-FRBEDO :FOR -Y-Z-PLANE BENDING WI-TH SHEAR AND BENDING SPRING i q7 R/2 I I:

. . Ksx

%%I . .

B/2

. q5 RACK DEGREES-OF-FREEDOM FOR X-Z PLANE BENDING WITH

- --SHEAR AND BENDING SPRING Figure 6.5.4 - Shear and Bending Springs Representing Rack Elasticity In X-Z and Y-Z Planes l

LUIIA An5 nI-..

riultm; meport rMl-Zu~zoo I &so0 1196

TYPICAL TOP IMPACT ELEMENT RACK -STRUCTURE TYPICAL BOTTOM \

IMPACT ELEMENTM Figure 6.5.5 - Rack Periphery Gap/lmpact Elements 6-51 riolteC eport rioitec K~epor MI-ZUat$bf Hl-eUZbt67 6-51 1196

. .PrasEUres by Pmp. CHAPLSIO at time a 12.S9

- - (ni Denates Channel n, '1' Marks Channel ends t-S9 at vx 500 - - 1000 Distance Along Wail (Inchoe)

GBIL 0 2 Coafficicat of oition D

Pressures by Prog. CHAPLSIO at time - 8,78

{-n)Denotes Channel n; '*1 Marki Channel Ends 500 . 1000 Dlstance Along Wall (Inahns)

DB!Cncyaeictof Fjctgu Figure 6.11.1 - Maximum Instantaneous Hydrodynamic Pressures t~tI,*. -c U II ^n-nn- -- -

noULVLec-UV[

vspr la-Zuzz00 5-52 1196

7.0 MECHANICAL ACCIDENTS

7.1 INTRODUCTION

The USNRC OT position paper [7.1.1] specifies that the design of the rack must ensure the functional integrity of the spent fuel racks under all credible drop events.

The postulated fuel drop events on the ANO-1 spent fuel pool Region 3 racks, which will be inserted with Metamic material in the rack flux traps with lead-ins installed on the top of flux traps, are evaluated.

The proposed change to the rack does not impact conclusions in the current licensing basis on the potential fuel damage due to mechanical accidents.

7.2 DESCRIPTION

OF MECHANICAL ACCIDENTS The postulated drop accidents assume that a fuel assembly, along with the portion of handling tool will drop vertically and hit the top of the rack at one of two enveloping locations: the cell wall edge, or the cell wall comer intersection. The weight and drop height in the postulated fuel drop accident are at most 2,580 lbs and 29.25 inches, respectively.

7.3 EVALUATION OF MECHANICAL ACCIDENTS To obtain conservative results, the postulated mechanical drop accidents were evaluated based on the maximum impact energy, a thinner rack wall thickness, weakest weld size and configuration, and worst case fabrication tolerance for the ANO-1 Region 3 racks. The evaluation of the postulated drop events demonstrated that, with the previously described conservative considerations, the postulated mechanical drop accidents would result in significant damage to the impacted cell wall down into the active fuel region of the racks, leading to the failure of Metamice inserts inside the flux trap.

fl-11L- I"pn r111- -~L-Of noeutvc AUP01 ril-.Uw~a /

7-I 1196

7.4 CONCLUSION

The fuel assembly drop events postulated for the ANO-1 spent fuel pool Region 3 racks were conservatively evaluated and found that the poison inserts, as well as the cell wall, of the impacted rack cell could be significantly damaged under the postulated accidental events. To ensure the functional integrity of the rack, the criticality safety evaluation (reported in Section 4.0) conservatively analyzed the Region 3 racks under the postulated assumption that all poison inserts in the impacted cell are damaged. The racks were determined to remain subcritical even under this extremely conservative postulate, when credit was taken for the technical specification limit of 1600 ppm soluble boron in the pool.

7.5 REFERENCES

[7.1.1] "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," dated April 14, 1978, and addendum dated 1979.

T 1.. -- UT rTnT - A_

Kpr11i-ZUVZLL?

nFJLLeCL;M:LF 7 -2 1 196

Attachment 5 1CAN040302 Evaluation of Spent Fuel Pool Structural Integrity for Increased Loads from Spent Fuel Racks to 1CAN040302 Page 1 of 4 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 the ANO-1 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.

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 6 of , 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.

to 1CAN040302 Page 2 of 4 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 factored in their entirety, by the maximum ratio calculated for the worst case rack in either the horizontal or vertical directions. This also conservatively accounted for the added hydrodynamic pressure on the pool walls.

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.

to 1CAN040302 Page 3 of 4 Summary of Section Strength Review of Selected Locations Location Section Strength Previous Analysis Conservative Estimate of Parameter Ratio to Code Ratio to Code Allowable Allowable for Increased Rack Loads Pool Floor Slab Moment 0.76 0.76 East-West Transverse Shear 0.24 0.213 Section In-Plane Shear 0.16 0.177 Pool Floor Slab Moment 0.59 0.636 North-South Transverse Shear 0.29 0.265 Section In-Plane Shear 0.42 0.477 Pool Moment 0.39 0.402 Foundation Transverse Shear 0.40 0.443 South Wall In-Plane Shear 0.55 0.821 Pool Moment 0.42 0.456 Foundation Transverse Shear 0.48 0.431 East Wall In-Plane Shear 0.96 0.982 Pool Moment 0.49 0.496 Foundation Transverse Shear 0.58 0.588 West Wall In-Plane Shear 0.80 0.906 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 0.82 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 0.982, which is about a 2% increase. The allowable of 1.0 is not exceeded.

It should be noted that the added poison inserts increase the deadweight for only two of the racks by a total of about 16870 pounds, which is about a 3.5% increase in deadweight for these two racks. For the review performed however, the weights for all racks were effectively increased about 29.5%. Additionally, the factored seismic loads included factoring the pool to 1CAN040302 Page 4 of 4 structural seismic loads as well, which have not changed. Hence in general, it can be seen that the increased load effects as applied for this review resulted in only slight increases, and if only the added poison inserts were considered, the changes would be very small.

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.

Attachment 6 1CAN040302 List of Regulatory Commitments 1CAN040302 Page 1 of 1 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 COMPLIANCE DATE (If ACTION Required)

Loading pattern restrictions including the vacant x space required by checkerboard storage configurations will be procedurally controlled Entergy will establish a coupon sampling program to x Within 90 ensure that the physical and chemical properties of days of TS Metamic behave in a similar manner to that found approval at the test facilities.

Entergy will complete the analysis of the structural x Within 90 integrity of the poison panel inserts for normal and days of TS seismic conditions considering the finalized design approval modification to ensure all safety factors are greater than 1.0.

A SAR change will be submitted to Licensing x Within 90 reflecting the changes made by this amendment days of TS approval

1CAN040302, License Amendment Request to Modify the Fuel Assembly Enrichment, the Spent Fuel Pool (SFP) Boron Concentration TS 3. 7.14, the Loading Restrictions in the SFP in TS 3,7,15m and to Modify the Fuel Storage Design Features in TS 4.3

Contents

Text

4.3 REFERENCES

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

5.0 TECHNICAL ANALYSIS

6.0 REGULATORY ANALYSIS

=I I I I = =I =;

1.0 INTRODUCTION

1.0 INTRODUCTION

1.1 REFERENCES

2.1 INTRODUCTION

SUMMARY

3.1 INTRODUCTION

3.6 REFERENCES

SUMMARY

5.1 INTRODUCTION

5.8 REFERENCES

6.1 INTRODUCTION

6.3 DESCRIPTION

7.1 INTRODUCTION

7.2 DESCRIPTION

7.4 CONCLUSION

7.5 REFERENCES

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1CAN040302, License Amendment Request to Modify the Fuel Assembly Enrichment, the Spent Fuel Pool (SFP) Boron Concentration TS 3. 7.14, the Loading Restrictions in the SFP in TS 3,7,15m and to Modify the Fuel Storage Design Features in TS 4.3

Text

4.3 REFERENCES

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

5.0 TECHNICAL ANALYSIS

6.0 REGULATORY ANALYSIS

=I I I I = =I =;

1.0 INTRODUCTION

1.0 INTRODUCTION

1.1 REFERENCES

2.1 INTRODUCTION

SUMMARY

3.1 INTRODUCTION

3.6 REFERENCES

SUMMARY

5.1 INTRODUCTION

5.8 REFERENCES

6.1 INTRODUCTION

6.3 DESCRIPTION

7.1 INTRODUCTION

7.2 DESCRIPTION

7.4 CONCLUSION

7.5 REFERENCES

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