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License Amendment Request Regarding Spent Fuel Storage Pool Criticality Analyses
	ML17193B165
Person / Time
Site:	Harris
Issue date:	06/28/2017
From:	Hamilton T; Duke Energy Progress
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
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ML17193B163	List: HNP-17-008, License Amendment Request Regarding Spent Fuel Storage Pool Criticality Analyses;
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	HNP-17-008
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Tanya M. Hamilton Vice President Harris Nuclear Plant 5413 Shearon Harris Road New Hill, NC 27562-9300 PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 4, THIS LETTER IS DECONTROLLED 10 CFR 50.90 June 28, 2017 Serial: HNP-17-008 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Shearon Harris Nuclear Power Plant, Unit 1 Docket No. 50-400/Renewed License No. NPF-63

Subject:

License Amendment Request Regarding Spent Fuel Storage Pool Criticality Analyses Ladies and Gentlemen:

Pursuant to 10 CFR 50.90, Duke Energy Progress, LLC (Duke Energy), hereby requests a revision to the Technical Specifications (TS) for the Shearon Harris Nuclear Power Plant, Unit 1 (HNP). The proposed license amendment modifies the TS for fuel storage criticality to account for the use of Metamic neutron absorbing spent fuel pool rack inserts and soluble boron for the purpose of criticality control in the Boiling Water Reactor (BWR) storage racks that currently credit Boraflex. Specifically, TS 5.6.1.3, BWR Storage Racks in Pools A and B, will be modified to reflect the respective design features of the two BWR rack types utilized in these pools, including adjusted requirements for the racks that credit the use of neutron absorbing Metamic inserts and soluble boron for criticality control in place of the existing credited Boraflex.

This license amendment request is required to resolve a current operable but degraded condition and is not a voluntary request from a licensee to change its licensing basis. Therefore, it is not subject to forward fit considerations as described in the letter from S. Burns (NRC) to E. Ginsburg (NEI), dated July 14, 2010 (ADAMS Accession No. ML01960180).

The proposed changes have been evaluated in accordance with 10 CFR 50.91(a)(1) using criteria in 10 CFR 50.92(c), and it has been concluded that the proposed changes involve no significant hazards consideration. Attachment 1 of this license amendment request provides Duke Energys evaluation of the proposed changes. Attachment 2 provides a copy of the proposed TS changes. Attachment 3 is the affidavit supporting the request for withholding the proprietary information in Attachment 4 from public disclosure. Attachments 4 and 5 provide the proprietary and nonproprietary versions of Holtec International Report No. HI-2177590, Licensing Report for Use of DREAM Neutron Absorber Inserts in the Spent Fuel Pools A and B at Shearon Harris NPP. Attachment 6 provides a copy of the TS Bases markup based on the proposed changes (for information only).

U.S. Nuclear Regulatory Commission Page2 HNP-17-008 PROPRIETARY INFORMATION -WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 4, THIS LETTER IS DECONTROLLED Approval of the proposed license amendment is requested within 18 months of acceptance. The amendment shall be implemented within 60 days following approval.

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated North Carolina State Official.

This document contains no new Regulatory Commitments.

Please refer any questions regarding this submittal to Jeffery Robertson, HNP Regulatory Affairs Manager, at (919) 362-3137.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on June .:i_g , 2017.

Sincerely, Tanya M. Hamilton Attachments:

1. Evaluation of the Proposed Change

2. Proposed Technical Specification Changes

3. Affidavit for Withholding of Proprietary Information

4. Holtec International Report No. Hl-2177590, "Licensing Report for Use of DREAM Neutron Absorber Inserts in the Spent Fuel Pools 'A' and 'B' at Shearon Harris NPP," Revision 1 (Proprietary)

5. Holtec International Report No. Hl-2177590, "Licensing Report for Use of DREAM Neutron Absorber Inserts in the Spent Fuel Pools 'A' and 'B' at Shearon Harris NPP," Revision 1 (Nonproprietary)

6. Proposed Technical Specification Bases Change (For Information Only) cc: J. Zeiler, NRC Sr. Resident Inspector, HNP W. L Cox, Ill, Section Chief N.C. DHSR M. Barillas, NRC Project Manager, HNP C. Haney, NRC Regional Administrator, Region II

U.S. Nuclear Regulatory Commission Page 2 HNP-17-008 PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 4, THIS LETTER IS DECONTROLLED Approval of the proposed license amendment is requested within 18 months of acceptance. The amendment shall be implemented within 60 days following approval.

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated North Carolina State Official.

This document contains no new Regulatory Commitments.

Please refer any questions regarding this submittal to Jeffery Robertson, HNP Regulatory Affairs Manager, at (919) 362-3137.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on June , 2017.

Sincerely, Tanya M. Hamilton Attachments:

1. Evaluation of the Proposed Change

2. Proposed Technical Specification Changes

3. Affidavit for Withholding of Proprietary Information

4. Holtec International Report No. HI-2177590, Licensing Report for Use of DREAM Neutron Absorber Inserts in the Spent Fuel Pools A and B at Shearon Harris NPP, Revision 1 (Proprietary)

5. Holtec International Report No. HI-2177590, Licensing Report for Use of DREAM Neutron Absorber Inserts in the Spent Fuel Pools A and B at Shearon Harris NPP, Revision 1 (Nonproprietary)

6. Proposed Technical Specification Bases Change (For Information Only) cc: J. Zeiler, NRC Sr. Resident Inspector, HNP W. L. Cox, III, Section Chief N.C. DHSR M. Barillas, NRC Project Manager, HNP C. Haney, NRC Regional Administrator, Region II

U.S. Nuclear Regulatory Commission Serial HNP-17-008 HNP-17-008 ATTACHMENT 1 EVALUATION OF THE PROPOSED CHANGE SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400 RENEWED LICENSE NUMBER NPF-063 13 PAGES PLUS COVER

U.S. Nuclear Regulatory Commission Page 1 of 13 Serial HNP-17-008 Evaluation of the Proposed Change

Subject:

License Amendment Request Regarding Spent Fuel Storage Pool Criticality Analyses 1.0

SUMMARY

DESCRIPTION In accordance with the provisions of 10 CFR 50.90, Duke Energy Progress, LLC (Duke Energy),

hereby requests a revision to the Technical Specifications (TS) for the Shearon Harris Nuclear Power Plant, Unit 1 (HNP). The proposed license amendment modifies the TS for fuel storage criticality to account for the use of Metamic neutron absorbing spent fuel pool (SFP) rack inserts and soluble boron for the purpose of criticality control in the Boiling Water Reactor (BWR) storage racks that currently credit Boraflex. Specifically, TS 5.6.1.3, BWR Storage Racks in Pools A and B, will be modified to reflect the respective design features of the two BWR rack types utilized in these pools, including adjusted requirements for the racks that credit the use of neutron absorbing Metamic inserts and soluble boron for criticality control in place of the existing credited Boraflex.

2.0 DETAILED DESCRIPTION 2.1 System Design and Operation The design of HNP incorporates the use of three spent fuel pools, one new fuel pool, and a cask loading pool, all of which are connected by a fuel transfer canal system. Spent fuel storage is provided by the New Fuel Storage Pool (Pool A) and the three SFPs commonly referred to as Pools B, C, and D. The four pools are licensed to include 3404 Pressurized Water Reactor (PWR) storage cells and 4628 BWR storage cells for a total storage capacity of 8032 fuel assemblies. The BWR racks are used for storage of spent fuel assemblies from the Brunswick Steam Electric Plant (BSEP), as authorized by HNPs Operating License.

The SFPs at HNP contain both PWR fuel racks and BWR fuel racks. Both the PWR and BWR racks use a neutron absorber in the rack design for reactivity control. Two types of neutron absorbers, Boral and Boraflex, are used in the BWR fuel racks. This proposed license amendment addresses the Westinghouse-designed rack for spent BWR fuel in use in HNP Pools A and B, which utilizes Boraflex as the neutron absorber.

The fuel storage racks consist of individual vertical cells fastened together through top and bottom supporting grid structures, or intermittent welds along the corners of adjacent fabricated cell boxes, to form an integral array of storage cells. The bottom of the storage cell array is connected to a baseplate, which provides additional rigidity and also serves to support the fuel assemblies in storage. The rack modules are free-standing and self-supporting. Either Boraflex or Boral neutron absorbing material is encapsulated into the stainless steel walls of the BWR racks in SFPs A and B.

2.2 Current Technical Specification Requirements The current HNP Design Feature TS 5.6.1.3 for fuel storage criticality in the BWR storage racks in SFPs A and B encompasses both BWR rack types (Boraflex and Boral) and does not credit the soluble boron present in the SFPs. The design for BWR racks in Pools A and B under the

U.S. Nuclear Regulatory Commission Page 2 of 13 Serial HNP-17-008 current licensing basis is such that the keff for the racks will not exceed 0.95 with the SFP flooded with unborated water. In its current state, it reads:

3. BWR Storage Racks in Pools "A" and "B"

a. keff less than or equal to 0.95 when flooded with unborated water.

b. The reactivity margin is assured for BWR racks in pools "A" and "B" by maintaining a nominal 6.25 inch center-to-center distance in the BWR storage racks.

Per TS 3.7.14, Fuel Storage Pool Boron Concentration, the boron concentration of SFPs shall be greater than or equal to 2000 parts per million (ppm) at all times for pools that contain nuclear fuel. In the event that the spent fuel pool boron concentration is not within this limit, immediate action is taken to suspend the movement of fuel assemblies and initiate restoration of the concentration to within the limit. However, the soluble boron requirement is not credited in the criticality analysis for the BWR racks.

2.3 Reason for the Proposed Change This license amendment request is required to resolve a current operable but degraded condition and is not a voluntary request from a licensee to change its licensing basis. Therefore, it is not subject to forward fit considerations as described in the letter from S. Burns (NRC) to E. Ginsburg (NEI), dated July 14, 2010 (ADAMS Accession No. ML01960180).

Significant industry experience has indicated that Boraflex degrades at a higher rate than originally anticipated. This degradation may lead to a reduction in the Boraflexs ability to hold down reactivity sufficiently to perform its design function. The Boraflex is not required for structural integrity of the racks and is not required for maintaining the required minimum spacing between assemblies. NRC Generic Letter (GL) 96-04 was issued in June 1996 informing power plants of concerns with the use of Boraflex material in spent fuel racks. The result of the degradation was acknowledged in the HNP supplemental response to GL 96-04 on April 25, 2005 (ADAMS Accession No. ML051230320) to the NRC, which details the actions taken and the coupon monitoring program and silica monitoring of the fuel pools. HNP previously removed credit for Boraflex from the PWR racks (which utilized it as the neutron absorber) by taking credit for both SFP dissolved boron and burnup credit, as approved by the NRC in the letter dated March 10, 2006 (ADAMS Accession No. ML060600349).

The degradation of the Boraflex material in these racks has resulted in Duke Energy planning to install Metamic rack inserts into the three Westinghouse BWR Boraflex racks in SFP A and the five in SFP B for reactivity control. Over the course of the next year, HNP will be installing a maximum of 968 BWR rack inserts in the identified racks.

The proposed rack enhancement program does not require any physical modifications to the existing storage rack arrays, other than insertion of the rack inserts into the storage cells. The inserts, known as DREAM inserts, are designed and supplied by Holtec International. Crediting Metamic as the neutron absorber, they provide a robust means of ensuring that the current inventory of stored irradiated fuel can be accommodated without losing spent fuel storage locations while maintaining an acceptable neutron multiplication factor. The Boraflex panels will remain in place providing additional, albeit not credited, neutron absorption.

U.S. Nuclear Regulatory Commission Page 3 of 13 Serial HNP-17-008 2.4 Description of the Proposed Change The proposed change requests NRC approval for use of an alternate mechanism other than Boraflex for criticality control. This application requests approval to credit the Metamic neutron absorbing rack inserts that will be installed in the BWR Boraflex storage rack cells in SFPs A and B, in combination with the soluble boron present in the pools, as a replacement for the neutron absorbing properties of the Boraflex panels.

Duke Energy proposes changes to the HNP TS as follows:

Design Feature TS 5.6.1.3 is revised to read:

3. BWR Storage Racks in Pools A and B

a. Racks with Metamic neutron absorber inserts

1. keff less than or equal to 0.95 when flooded with water borated to 1000 ppm.

2. keff less than 1.0 when flooded with unborated water.

3. The reactivity margin is assured for BWR racks in pools A and B by maintaining a nominal 6.25 inch center-to-center distance in the BWR storage racks.

4. The following restrictions are also imposed through administrative controls:

a. Storage of BWR fuel designs limited to GE3, GE4, GE5, GE6, and GE7 fuel designs.

b. Rack insert orientation is limited to that shown in Figure 5.6-3 and Figure 5.6-4.

c. No fuel shall be stored in Storage Location A11 of Rack C1 in Spent Fuel Pool A.

b. Racks with Boral neutron absorber

1. keff less than or equal to 0.95 when flooded with unborated water.

2. The reactivity margin is assured for BWR racks in pools A and B by maintaining a nominal 6.25 inch center-to-center distance in the BWR storage racks.

Add TS Figure 5.6-3 and Figure 5.6-4 to identify the required orientation of the rack inserts in the applicable BWR racks in SFPs A and B, respectively.

In addition to the proposed changes to the TS, the TS 3.7.14 Bases will be revised to account for the changes to the fuel storage criticality design feature requirements for BWR racks in SFPs A and B to remove reliance on Boraflex as a neutron absorber and credit Metamic rack inserts and soluble boron for criticality control. Changes to the TS Bases will be made in accordance with the TS Bases Control Program following approval of the requested amendment and are provided in Attachment 6 for information only.

U.S. Nuclear Regulatory Commission Page 4 of 13 Serial HNP-17-008

3.0 TECHNICAL EVALUATION

The proposed changes are intended to resolve the issue of Boraflex degradation in the HNP SFP A and B Westinghouse BWR racks. This evaluation relies significantly on Holtec Report HI-2177590, Licensing Report for Use of DREAM Neutron Absorber Inserts in the Spent Fuel Pools A and B at Shearon Harris NPP (Attachment 4 of this submittal, hereafter referenced as Attachment 4). The following text will provide a roadmap of all the requisite evaluations with substantial reference to Attachment 4.

3.1 Insert Mechanical Design The design objective of the rack insert is to provide a neutron absorber, with material composition and dimensions suitable for co-residence with fuel in a storage cell, that can be easily inserted and relocated within the storage racks, of which the neutron absorption properties must be sufficient to eliminate reliance on the existing Boraflex. Designed for insertion in any Westinghouse BWR rack storage cell containing a fuel assembly, the insert consists of a Metamic panel attached to an aluminum upper block used to provide a robust attachment for handling purposes. Installed, the insert blankets two of the four walls of the host storage cell. The inserts upper aluminum block is equipped with interfaces for lifting and handling by an appropriate custom-designed tool. The position of the aluminum block at the top of the fuel assembly will be visually evident and it provides confirmation of the orientation of the insert within each cell.

The design of the insert ensures that, when seated, the active fuel region is shadowed by Metamic. The dimensional differences between a fuel assembly (5.787 inches square) and the inside dimension of a storage rack cell (nominally 6.05 inches square) provide a sufficient gap (0.263 inches) for the insert to be inserted. The bottom edge of the Metamic panel on each insert is skew cut and beveled to ensure that the insert will readily slide into this gap and not snag on any fuel assembly components.

In order to move the fuel assembly from its cell, the insert must first be removed. Different tools are used to manipulate fuel assemblies and rack inserts. As such, in addition to removing the insert, a tool change is required to move a fuel assembly from a cell that has an insert.

3.2 Material Considerations Manufactured by the Orrvilon division of Holtec International in Ohio, the Metamic neutron absorber material is the principal material for manufacturing the insert. Comprised of aluminum alloy 6061 and boron carbide (B4C), Metamic has been subjected to rigorous testing and been approved by the NRC for use in wet storage (i.e., fuel pool) applications, including PWR SFP racks installed in HNP SFPs C and D (NRC letter dated January 29, 2009, ADAMS Accession No. ML090270022). All other components of the insert are manufactured from aluminum alloys, which are chemically compatible with Metamic.

Section 3 of Attachment 4 summarizes the considerations that provide assurance that the inserts that are being installed in HNP Pool A and B Westinghouse BWR racks will perform their intended function for the design life of the fuel racks.

U.S. Nuclear Regulatory Commission Page 5 of 13 Serial HNP-17-008 3.3 Criticality Section 4 of Holtec Report HI-2177590, Revision 1, submitted as Attachment 4 of this LAR, documents the SFP criticality calculations performed for Pools A and B. These calculations were performed for the BWR Boraflex storage racks without credit for the degraded Boraflex and with credit for Metamic inserts. The objective of the analysis was to demonstrate that the effective neutron multiplication factor (keff) is less than 1.0 with the pool flooded with unborated water and that the keff is less than or equal to 0.95 with the pool flooded with borated water. The maximum calculated keff includes a margin for uncertainty in reactivity calculations including manufacturing tolerances and is shown to be less than the regulatory limit with a 95%

probability at a 95% confidence level as presented in Table 4.7.1 of Attachment 4. Reactivity effects of abnormal and accident conditions have also been evaluated to assure that under all credible abnormal and accident conditions, the keff will not exceed the regulatory limit of 0.95 with credit for soluble boron.

Criticality control in the BWR Boraflex storage racks relies upon the following:

Administrative Restrictions:

1. The BWR fuel designs allowed in the Harris Pool A and Pool B BWR Boraflex storage racks are limited to the GE3, GE4, GE5, GE6, and GE7 fuel designs.

2. The orientation of the Metamic inserts in Harris Pool A and Pool B BWR Boraflex storage racks are limited to the orientation as shown in Figures 4.1.1 and 4.1.2 of Attachment 4 (also provided in the Attachment 2 markup of the HNP TS for Figure 5.6-3 and Figure 5.6-4).

3. No fuel shall be stored in the storage cell at the northeast corner of the BWR Boraflex storage racks nearest the interface with the PWR Boraflex storage racks on the north and east side in Pool A.

Soluble boron for normal and accident conditions in accordance with 10 CFR 50.68(b)(4).

Holtec Metamic inserts.

Criticality control in the BWR Boraflex storage racks does not rely on residual Boraflex, burnup, or gadolinium.

Section 4.1.1 of Attachment 4 outlines the special considerations taken into account during the development of the criticality analysis, while Section 4.2 outlines the methodology utilized to perform the analysis, ensuring that the results are below the regulatory limit with a 95%

probability at a 95% confidence level. Section 4.3 provides a listing of the codes, standards, and regulations, or pertinent sections thereof, which are applicable to the analyses. The analysis itself is provided in Section 4.7.

As provided in TS 3.7.14, the Limiting Condition of Operation (LCO) limit for boron concentration is 2000 ppm and applies at all times to any pool that stores nuclear fuel. The most limiting boron requirement of any of the pools as required by the criticality analysis is 1000 ppm. The difference between 2000 ppm and 1000 ppm provides margin for boron measurement uncertainties and the detection and mitigation of an accidental boron dilution event. The

U.S. Nuclear Regulatory Commission Page 6 of 13 Serial HNP-17-008 required minimum boron concentration for normal conditions and for the limiting credible accident scenario is consistent with the PWR Boraflex racks in SFPs A and B, which credit soluble boron as per approved HNP Operating License Amendment No. 121 (ADAMS Accession No. ML060600356).

Operation in accordance with the proposed amendment will not change the probability of a boron dilution event because the systems and events that could affect spent fuel soluble boron are unchanged. The consequences of a boron dilution event are unchanged because the proposed amendment utilizes a soluble boron value in the analysis below the SFP TS requirement and the maximum possible water volume displaced by the inserts is an insignificant fraction of the total spent fuel pool water volume. In addition, the criticality analysis concludes that keff remains less than 1.0, even with unborated water in the SFPs. As such, the boron dilution analysis for the PWR Boraflex racks remains bounding and is applicable for the BWR racks with Metamic inserts.

3.4 Thermal-Hydraulic Evaluation With respect to thermal-hydraulic performance of the spent fuel pool cooling function, the Westinghouse BWR racks with DREAM inserts were evaluated to the following acceptance criteria:

1. Following the planned offload fuel assemblies from the Harris reactor and with forced cooling available, the bulk SFP temperatures shall be limited to 150 °F.

2. Under a complete failure of active cooling during the limiting fuel offload scenario, the water surface is allowed to reach saturation. Sufficient time must be available before the onset of bulk boiling to implement corrective measures.

3. Local water and fuel cladding temperatures for the fuel assemblies within the Westinghouse BWR racks shall not exceed the local saturation temperature of water. , Section 5 provides a summary of the methods, models, analyses, and numerical results that demonstrate the Westinghouse BWR racks in HNP SFPs A and B will continue to meet the thermal-hydraulic requirements for safe storage of spent fuel following the installation of Holtec-supplied DREAM inserts. With respect to bulk thermal effects, it demonstrates that the existing licensing bases for equilibrium SFP temperature and time-to-boil remain bounding. With respect to local thermal effects, it demonstrates that the local water temperature and fuel rod cladding temperature remain within the acceptance criteria.

3.5 Rack Structural Evaluation/Seismic Considerations Section 6 of Attachment 4 describes the structural adequacy of the HNP SFP racks after DREAM inserts have been added to the existing Westinghouse BWR racks located in Pools A and B. Loadings considered were postulated to occur during normal, seismic, and accident conditions. The evaluation of the structural design bases for HNP SFPs A and B considered the following specific areas:

Seismic Qualification of Existing BWR Boraflex Racks Fuel Pool Structural Qualification Pool Liner Qualification Mechanical Accident Evaluation

U.S. Nuclear Regulatory Commission Page 7 of 13 Serial HNP-17-008 The structural design bases for the existing Boraflex racks in HNP SFPs A and B, as well as the structural qualification of the pool structures, were determined to be not adversely affected by the planned installation of neutron absorber inserts as the weight of the neutron absorber inserts are negligibly small in comparison to the overall dead weight of these structures and their contents. Additionally, the structural integrity of the DREAM inserts under normal (i.e.,

installation and handling) and accident condition (i.e., seismic events) loads was found to be adequate to perform their intended design function.

3.6 Metamic Surveillance Program In conjunction with the installation of the Metamic rack inserts into the eight Westinghouse-fabricated BWR Boraflex spent fuel racks installed in SFPs A and B, HNP will implement a surveillance coupon monitoring program to verify that the Metamic rack inserts continue to provide the criticality control relied upon in the criticality analysis. The purpose of such a surveillance program is to characterize certain properties of Metamic in order to provide the data needed to evaluate the ability of Metamic to perform its intended function. The monitoring program will be capable of identifying whether changes to the Metamic are occurring, and if those changes are occurring, that the anticipated characteristics of change can be verified.

The Metamic rack insert monitoring program will rely primarily upon the results of a coupon testing program. Other activities supporting the monitoring program will include: SFP water chemistry monitoring, maintaining an awareness of ongoing research and development, participation in industry groups that share operating experience amongst plants, and evaluation of the relevance of outside data on the in-service material. Acceptance criteria provide the basis for the comparison of results in order to determine whether material performance is acceptable or actions are necessary to address performance issues.

3.6.1 Coupon Testing Program Overview The coupon testing program consists of a population of small sections of the same neutron absorber material lots used to fabricate the rack inserts. The coupons are suspended on a mounting called a "tree", placed in a designated cell, and surrounded by spent fuel. The designated cell location and surrounding fuel will be chosen to ensure the coupons see conditions which will bound that of the installed inserts. The coupon "tree" contains 14 Metamic test coupons which will provide adequate samples to support the monitoring program beyond the end of HNPs current operating license.

The coupons are fully exposed to the spent fuel pool environment except where they are attached to the coupon "tree". The coupons are attached to the "tree" with hardware fabricated from the same materials of the spent fuel rack structural components which will provide some indication of performance at the interface of different material types. Pre-characterization of each coupon will be done prior to insertion to obtain "before" reference initial values for comparison with "after" irradiation measurements. The coupons are removed from the array on a prescribed schedule for testing. The stability and integrity of the Metamic in the storage cells may be inferred from the measurement of certain physical properties.

3.6.2 Coupon Testing Program Details The specific criteria of HNP's surveillance program include:

U.S. Nuclear Regulatory Commission Page 8 of 13 Serial HNP-17-008

Metamic coupon dimensions: 8" x 4" x 0.072" nominal thickness;

Coupons are from the lots in which the inserts were fabricated with no welds; proximity to stainless steel (rack wall) is accomplished via bolt/stainless steel washer combination where the coupon is attached to the tree;

Initial B4C content of the coupons was determined by measuring constituents during material manufacture;

B4C measured content as identified in Table 2.2 of Attachment 4;

Boron in B4C shall contain a minimum of 18.3 weight percent of B-10;

Coupon scratches and other surface anomalies will not be added to individual coupons since coupons are of the same lot/sections as the production Metamic and will contain same surface defects;

Nominal B-10 areal density as found in Table 4.5.4 of Attachment 4;

The coupon tree will be placed in a rack location surrounded by bounding fuel assemblies. The HNP BWR fuel population is static (no additional BWR assemblies to be received) therefore the fuel assemblies will not require periodic changeout.

For neutron attenuation testing of the coupons, acceptance criterion will be established such that the B-10 areal density of the test coupon is the same as its original B-10 areal density within the uncertainty of the measurement.

Any bubbling, blistering, cracking or flaking of the coupons, any dose changes to the coupons, the effects of any fluid movement and temperature fluctuations of the pool water will all be addressed through visual observation, dimensional measurement, weight measurement and specific gravity measurements; any areal density changes of the coupons will be monitored through neutron attenuation measurements;

The initial testing interval is set at 5 years and will expand to 10 years provided acceptable performance is observed. This schedule is based upon Metamics excellent performance to date in spent fuel pool applications while recognizing its limited experience in the industry.

HNP will utilize the following methods to monitor the physical properties of the Metamic coupons in the surveillance program for changes:

Visual observation and photography

Neutron attenuation testing

Dimensional measurements (length, width, thickness)

Weight and specific gravity measurements 3.6.3 Evaluating Coupon Monitoring Program Results The results from the Metamic coupon monitoring program fall within the broad categories of: 1) confirmation that no material changes are occurring; 2) confirmation that anticipated changes are occurring; and/or 3) determination if unanticipated changes are occurring. The coupon testing program procedure will be used to evaluate results of the monitoring program. If no changes are observed, or if anticipated changes are occurring that have already been accounted for, then the material condition continues to be adequately represented in the criticality analysis.

If unanticipated changes are identified (either new mechanisms or anticipated mechanisms at rates or levels beyond those anticipated), the issue will be entered into the Corrective Action Program for further evaluation. This evaluation will address the following items as applicable:

U.S. Nuclear Regulatory Commission Page 9 of 13 Serial HNP-17-008 Determine if unanticipated changes could result in a loss of B-10 areal density.

Evaluation of the effects of B-10 areal density on the criticality analysis will be performed and addressed through appropriate licensee processes. Additionally, monitoring or test results that indicate potential degradation are evaluated and trended, even if it does not challenge the criticality analysis.

Determine if unanticipated changes not resulting in loss of B-10 areal density have an impact on the criticality analyses. Dimensional or non-neutron absorbing material changes (e.g. formation of gaps, localized displacement of moderator, or superficial scratches) may have no or little impact on the criticality analyses. However, the potential effects of these changes on the criticality analysis will be evaluated and addressed through appropriate licensee processes.

3.7 Conclusion The proposed change is necessary to resolve the issue of Boraflex degradation in the HNP SFP A and B Westinghouse BWR racks. The proposed change to take credit for Holtec DREAM rack inserts in the spent fuel pool storage racks has been evaluated and shown to be a safe and effective manner in which to resolve the Boraflex degradation issue for the remaining time that BSEP BWR spent fuel needs to be stored in the HNP spent fuel pool storage racks.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 10 CFR 50.68, Criticality Accident Requirements, paragraph (b)(4) states that if credit is taken for soluble boron, the keff of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with borated water. The keff must also remain below 1.0, at a 95 percent probability, 95 percent confidence level, if flooded with unborated water. The rack insert criticality analysis, as discussed in Attachment 4 to this submittal, demonstrates that this requirement is met.

10 CFR 50 Appendix A General Design Criterion (GDC) 61, Fuel Storage and Handling and Radioactivity Control, requires that the fuel storage and handling systems be designed to assure adequate safety under normal and postulated accident conditions. As previously discussed in this request, the proposed use of the Holtec DREAM rack inserts to resolve the Boraflex degradation issue does not impact HNPs compliance with this criterion.

10 CFR 50 Appendix A GDC 62, Prevention of Criticality in Fuel Storage and Handling, specifies that criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations. The changes proposed in this request do not impact the capability of the other storage racks in SFPs A and B to comply with this criterion. Additionally, the criticality analysis crediting the Metamic rack inserts and soluble boron has been performed to demonstrate that keff will remain less than or equal to 0.95 with no credit taken for the Boraflex neutron poison material present in the spent fuel storage racks. It also demonstrates that the keff will remain below 1.0 in the event of being flooded with unborated water.

U.S. Nuclear Regulatory Commission Page 10 of 13 Serial HNP-17-008 4.2 Precedent The NRC previously approved the use of Metamic in SFPs C and D at HNP as a neutron absorber in the form of fuel storage racks via letter dated January 29, 2009 (ADAMS Accession No. ML090270022). While not directly related to crediting the use of Metamic rack inserts in place of degraded Boraflex in BWR storage racks, it does establish that the Metamic neutron absorber is compatible with the environment of the SFP. It also established that the proposed surveillance program for the Metamic racks was capable of detecting potential degradation that could impair its neutron absorption capability. The surveillance program outlined in Section 3.6 of this submittal is similarly structured for the Metamic rack inserts.

Additionally, the NRC has approved the use of Metamic rack inserts to replace the credit taken for Boraflex in the criticality analysis for Turkey Point, Units 2 and 3, to address remedying the degradation of Boraflex neutron absorbing material (ADAMS Accession No. ML071800198).

While this example doesnt provide exact precedence due to HNP seeking to credit Metamic rack inserts in BWR Boraflex racks versus PWR Boraflex racks, it does provide for the crediting of Metamic rack inserts to replace the reliance on Boraflex as a neutron absorber in a PWR SFP environment with credited soluble boron. This would be the first application of the DREAM rack insert in BWR racks.

The NRC also has approved the crediting of soluble boron and burnup as a means of counterbalancing the removal of the Boraflex rack poison credit in the PWR Boraflex racks in HNP SFPs A and B via letter dated March 10, 2006 (ADAMS Accession No. ML060600349).

While this example doesnt provide exact precedence, as HNP is not requesting fuel burnup credit be assessed for the BWR Boraflex storage racks, it does provide for the limited crediting of soluble boron as a means to offset the reactivity increase due to Boraflex removal.

4.3 No Significant Hazards Consideration Determination Pursuant to 10 CFR 50.90, Duke Energy Progress, LLC (Duke Energy), proposes a license amendment request (LAR) for the Shearon Harris Nuclear Power Plant, Unit 1 (HNP) Technical Specifications (TS). The proposed license amendment modifies the TS for fuel storage criticality to account for the use of neutron absorbing spent fuel pool (SFP) rack inserts and soluble boron for the purpose of criticality control in the Boiling Water Reactor (BWR) storage racks that currently credit Boraflex. Specifically, TS 5.6.1.3, BWR Storage Racks in Pools A and B, will be modified to reflect the respective design features of the two BWR rack types utilized in these pools, including adjusted requirements for the racks that credit the use of neutron absorbing Metamic inserts and soluble boron for criticality control in place of the existing credited Boraflex.

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

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

The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. As this amendment request pertains only to the spent fuel pool, only those accidents that are related to movement and storage of fuel assemblies in the SFP could potentially be affected by the proposed

U.S. Nuclear Regulatory Commission Page 11 of 13 Serial HNP-17-008 change. The proposed change modifies HNP TS 5.6.1.3 to reflect the respective design features of the two BWR rack types utilized in these pools, including adjusted requirements for the Boraflex racks that account for use of neutron absorbing inserts.

The change is necessary to ensure that, with continued Boraflex degradation over time, the effective neutron multiplication factor, keff, is less than or equal to 0.95 if the spent fuel pool is flooded with borated water, and that it is less than 1.0 if flooded with unborated water, as required by 10 CFR 50.68(b)(4).

The installation of DREAM rack inserts and credit for soluble boron does not result in a significant increase in the probability of an accident previously analyzed because there are no changes in the manner in which spent fuel is handled, moved, or stored in the rack cells. The probability that a fuel assembly would be dropped or misloaded is unchanged by the installation of the DREAM rack inserts and use of additional administrative controls on BWR fuel storage in these racks. These events involve failures of administrative controls, human performance, and equipment failures that are unaffected by the presence or absence of Boraflex and the rack inserts. The probability of a SFP dilution event is also unchanged. The soluble boron is already present in SFPs A and B and no changes are proposed regarding the manner in which soluble boron is managed. The current controls in place remain applicable.

The installation of the DREAM rack inserts and crediting of soluble boron does not result in a significant increase in the consequences of an accident previously analyzed because there is no change in the fuel assemblies that provide the source terms used in calculating the radiological consequences of a fuel handling accident. In addition, consistent with the current design, only one fuel assembly will be moved at a time. Thus, the consequences of dropping a fuel assembly onto any other fuel assembly or other structure remain bounded by the previously analyzed fuel handling accident. The proposed change does not affect the effectiveness of the other engineered design features, such as filtration systems, that limit the offsite dose consequences of a fuel handling accident.

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 previously evaluated?

The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. The rack inserts are passive devices that, when inside a spent fuel storage rack cell, perform the same function as the previously licensed Boraflex neutron absorber panels in that cell. These devices do not add any limiting structural loads or affect the removal of decay heat from the assemblies.

No change in total heat load in the spent fuel pool is being made. The inserts will maintain their design function over the life of the spent fuel pool. The existing fuel handling accident, which assumes the drop of a fuel assembly, bounds the drop of a rack insert and/or rack insert installation tool. This proposed change does not create the possibility of misloading an assembly into a spent fuel storage rack cell.

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

U.S. Nuclear Regulatory Commission Page 12 of 13 Serial HNP-17-008

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

The proposed change does not involve a significant reduction in a margin of safety. The DREAM rack inserts are being installed to restore the spent fuel pool criticality margin, compensating for the degraded Boraflex neutron absorber. The DREAM rack inserts, once approved for crediting, will replace the existing Boraflex as the credited neutron absorber for controlling spent fuel pool reactivity, even though the Boraflex absorber will remain in place.

The proposed HNP TS 5.6.1.3.a.1 requires that the BWR spent fuel storage racks with Metamic rack inserts maintain the effective neutron multiplication factor, keff, less than or equal to 0.95 when flooded with water borated to 1000 ppm. This includes an allowance for uncertainties in such that the TS limit for boron concentration in the SFPs, HNP TS 3.7.14, shall be greater than or equal to 2000 ppm at all times for pools that contain nuclear fuel. Therefore, for criticality, the required safety margin is 5% including a conservative margin to account for engineering and manufacturing uncertainties.

The proposed change provides a method to ensure that keff continues to be less than or equal to 0.95, thus preserving the required safety margin of 5%. The criticality analyses demonstrate that the required margin to 5%, including a conservative margin to account for engineering and manufacturing uncertainties, is maintained. In addition, the radiological consequences of a dropped fuel assembly onto a spent fuel storage rack cell containing a fuel assembly with a rack insert is bounded by the radiological consequences of a dropped fuel assembly without a rack insert. The proposed change also maintains the capacity of the HNP spent fuel pools.

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

Based on the above, Duke Energy concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92, and, accordingly, a finding of "no significant hazards consideration" is justified.

4.4 Conclusions Based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

Duke Energy has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR Part 20, and would change an inspection or surveillance requirement. However, the proposed changes do not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

U.S. Nuclear Regulatory Commission Page 13 of 13 Serial HNP-17-008 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.

U.S. Nuclear Regulatory Commission Serial HNP-17-008 HNP-17-008 ATTACHMENT 2 PROPOSED TECHNICAL SPECIFICATION CHANGES SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400 RENEWED LICENSE NUMBER NPF-063 8 PAGES PLUS COVER

DESIGN FEATURES No changes to this page 5.6 FUEL STORAGE CRITICALITY 5.6.1 The fuel storage racks are designed and shall be maintained with:

L PWR storage racks in pools "A" and "B"

a. kett 1ess than or equal to 0. 95 if fully flooded with water borated to 2000ppm.

b. ke11 less than 1. 0 if flooded with unborated water.

c. A nominal 10 .5 inch center-to-center distance between fuel assemblies.

d. Assemblies must be within the "acceptable range" of the burnup

re~triLlrnns ~ho~-.:;, inf ;gure 5.6-2 prior to storage in unrestricted storage.

e. Assemblies that do not meet the requirements of 5.6.1.1.d shall be stored in a 2-of-4 checkerboard within and across rack module boundaries. Less dense storage patterns (e.g. 1-of-4 or l-of-5) are acceptable in place of 2-of-4.

f. The empty spaces (water holes) in the 5.6.1.1.e checkerboard may be occupied by non-fuel items (e.g .. containment specimen and trash baskets. mock fuel assemblies etc.) up to a li mit of one per every 6 storage spaces.

g. If fuel that meets the requirement of 5.6.1.1.d and fuel that does not meet 5.6.1.1.d are stored in the same rack module.an interface region must exist between the two regions. The interface region shall either be an empty row/column or a row/column of fuel that meets the requirements of 5.6.1 .1.d in a checkPrhonrrl pattPrn with the restricted

( 5. 6 .1.1. e) r*eg ion'.

2. Dry New Fuel PWR Storage Racks

a. Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent.

b. ke 11 ~.95 if fully flooded with unborated water without credit for Boraflex in the rack module .

c. ke11 < 0.98 i11 *an op1.iF:..:*:1 moderation event.

d. A nominal 10.5 inch center to center distance between storage cells with alternating rows and columns blocked such that fuel is stored in a l-of-4 pattern.

SHEARON HARRIS - UNIT 1 5-7 Amendment No. 121

~--

DESIGN FEATURES unborated

4. PWR and BWR racks in pools "C" and "D"

a. kett less than or equal to 0.95 when flooded with unborated water.

b. The reactivity margin i s assured for poo1s "C" and "D" by maintaining a nominal 9.017 inch center-to-center distance uetween fuel assemblies placed in the non-flux trap style PWR storage racks and 6.25 inch center-to-center distance in the BWR storage racks.

c. The following restrictions are also imposed through administrative controls:

1. PWR assemblies must be within the "acceptable range" of the burnup restrictions shown in Figure 5.6-1 prior to storage in pools "C" and "D".

2. BWR assemblies are acceptable for storage in pool "C" provided the maximum planar average enrichments are less than 4.6 wt .% U235 and K~t is less than or equal to 1.32 for the standard cold core geometry CSCCG).

5. In each case. kett includes allowances for uncertainties as described in Section 4.3.2.6 of the FSAR .

ORA1NAGt 5.6.2 The pools "A". "B". "C" and "D" are designed and shall be maintained to prevent inadvertent draining of the pools below elevation 277.

CAPACITY 5.6.3.a Pool "A" contains six (6 x 10 cell) flux trap type PWR racks and three (11 x 11 cell) BWR racks for a total storage capacity of 723 assemblies.

Pool "B" contains six (7 x 10 cell). five (6 x 10 cell). and one (6 x 8 cell) flu~ trap sLyle PWR racks and seventeen Cll x 11 cell) BWR racks and is licensed for one additional Cll x 11 cell) BWR rack that will be installed as needed. The combined pool "A" and "B" licensed storage capacity is 3669 assemblies.

SHEARON HARRIS - UNIT 1 5-7a Amendment No . ~

3. BWR Storage Racks in Pools A and B

a. Racks with Metamic neutron absorber inserts

1. keff less than or equal to 0.95 when flooded with water borated to 1000 ppm.

2. keff less than 1.0 when flooded with unborated water.

3. The reactivity margin is assured for BWR racks in pools A and B by maintaining a nominal 6.25 inch centertocenter distance in the BWR storage racks.

4. The following restrictions are also imposed through administrative controls:

a. Storage of BWR fuel designs limited to GE3, GE4, GE5, GE6, and GE7 fuel designs.

b. Rack insert orientation is limited to that shown in Figure 5.63 and Figure 5.64.

c. No fuel shall be stored in Storage Location A11 of Rack C1 in Spent Fuel Pool A.

b. Racks with Boral neutron absorber

1. keff less than or equal to 0.95 when flooded with unborated water.

2. The reactivity margin is assured for BWR racks in pools A and B by maintaining a nominal 6.25 inch centertocenter distance in the BWR storage racks.

DESIGN FEATURES No changes to this page 5.6 .3.b Pool "C" is designed to contain a combination of PWR and BWR assemblies Pool "C" can contain two 01 x 9 cell) and thirteen (9 X 9 cell)

PWR racks for storage of 1251 PWR assembl ies. Pool "C" can contain two (8 x 13 cell). two C8 x 11 cell). six (13 x 11 cell), and five (13 x 13 cell) BWR racks for storage of 2087 BWR assemblies. The ( 9 x 9 ce 11) PWR racks and the (13 x 13 cell) BWR racks are dimensioned to allow intercllangeab1lity between PWR or BWR si.orage rack styles as required . The racks in pool C will be installed as needed .

5 6.3 c Pool "D" contains a variable number of PWR storage spaces. These racks will be installed as needed . Pool "D" is designed for a maximum storage capacity of 1025 PWR assemblies 5.6.3.d The heat load from fuel stored in Pools "C" and "D" shall not exceed

7. 0 MBtu/hr.

5.7 COMPONENT CYCLIC OR TRANSIENT LJMIT 5 7 l The components idenL1fied in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of Table 5 7- 1 SHEARON HARRIS - UNIT 1 5*7b Amendment No 129

No changes to this DESIGN FEATURES page 40000 Acceptable Bumup Domain 35000 - - - - - + - - - - - - + - - - - -- - - - - - + - - - - - A - - - - - - - 1 Bumup = 12114 x Enrichment-19123 I-

~ 25000 - - - - - + - - - - - - - + - - - ---+----

c

~

- 20000

E a.

~ -----+-----+-~-+------+----

...::J m

Unacceptable Bumup Domain 5000 - - - - + - - - - - - - --------- ---~----l--- ----t I

2.0 2.5 3.0 3.5 4.0 4.5 5.0 Enrichment (wt% U-235)

FIGURE 5 .6-1 POOLS "C" and "D" BURNUP VERSUS ENRICHMENT FOR PWR FUEL SHEARON HARRIS - UNIT 1 5-7c Amendment No. 121

No changes to this DESIGN FEATURES page 45000 ~--~--~--

_ _ Acceptable Burnup Domain 40000 35000 Burnup = 12230 x Enrichment - 18770 30000 -+----+- -+--+---+-----+--_,,.---I------<

I-

~ 25000 -+-----+---

c 3:

1!

a.

...:l~ 20000 -+------+---

co 15000 -+-----+-----+-

Unacceptable Burnup Domain Note: The Failed Rod Storage Canister is exempt from the requirements of this curve for 5000 -+---~F+------tstorage in the unrestricted region.

1.5 2 2.5 3 3.5 4 4.5 5 Enrichment (wto/o U-235)

FIGURE 5. 6-2 POOLS "A" and "B" BURNUP VERSUS ENRICHMENT FOR PWR FUEL SHEARON HARRIS - UNIT 1 5-7d Amendment No. 121

INSERT: New Page/Figure DESIGN FEATURES

PWR Racks BWR Rack BWR Rack BWR Rack PWR Racks With Inserts With Inserts With Inserts Plant North Insert Orientation

FIGURE 5.6-3 POOL A METAMIC RACK INSERT ORIENTATION SHEARON HARRIS - UNIT 1 5-7e Amendment No.

INSERT: New Page/Figure DESIGN FEATURES

PWR Racks BWR Racks BWR Rack BWR Rack BWR Rack With With Inserts With Inserts With Inserts BWR Racks Boral With Boral BWR Rack BWR Rack BWR Rack With Inserts With Inserts With Boral Plant North Insert Orientation

FIGURE 5.6-4 POOL B METAMIC RACK INSERT ORIENTATION

SHEARON HARRIS - UNIT 1 5-7f Amendment No.

U.S. Nuclear Regulatory Commission Serial HNP-17-008 HNP-17-008 ATTACHMENT 3 AFFIDAVIT FOR WITHHOLDING OF PROPRIETARY INFORMATION SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400 RENEWED LICENSE NUMBER NPF-063 5 PAGES PLUS COVER

HOLTEC INTERNATIONAL Holtec Center, 555 Lincoln Drive West, Marlton, NJ 08053 Telephone (856) 797-0900 Fax (856) 797-0909 Holtec International Document ID 2635001--AFFIDAVIT-01 AFFIDAVIT PURSUANT TO 10 CFR 2.390 I, Debabrata (Debu) Mitra Majumdar, being duly sworn, depose and state as follows:

( 1) I have reviewed the information described in paragraph (2) which is sought to be withheld, and am authorized to apply for its withholding.

(2) The information sought to be withheld is information in the following report.

a. HI-2177590, "Licensing Report for Use of DREAM Neutron Absorber Inserts in the Spent Fuel Pools "A" and "B" at Shearon Harris Nuclear Power Plant", Revision l" This report contains Holtec Proprietary Information.

(3) In making this application for withholding of proprietary information of which it is the owner, Holtec International relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4) and the Trade Secrets Act, 18 USC Sec. 1905, and NRC regulations 10CFR Part 9.17(a)(4), 2.390(a)(4), and 2.390(b)(l) for "trade secrets and commercial or financial information obtained from a person and privileged or confidential" (Exemption 4 ). The material for which exemption from disclosure is here sought is all "confidential commercial information", and some portions also qualify under the narrower definition of "trade secret", within the meanings assigned to those terms for purposes ofFOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission, 975F2d871 (DC Cir. 1992),

and Public Citizen Health Research Group v. FDA, 704F2d1280 (DC Cir.

1983).

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Holtec International Document ID 2635001-AFFIDAVIT-Ol AFFIDAVIT PURSUANT TO 10 CFR 2.390 (4) Some examples of categories of information which fit into the definition of proprietary information are:

a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by Holtec's competitors without license from Holtec International constitutes a competitive economic advantage over other companies;

b. Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product.

c. Information which reveals cost or price information, production, capacities, budget levels, or commercial strategies of Holtec International, its customers, or its suppliers;

d. Information which reveals aspects of past, present, or future Holtec International customer-funded development plans and programs of potential commercial value to Holtec International;

e. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraph 4.b, above.

(5) The information sought to be withheld is being submitted to the NRC in confidence. The information (including that compiled from many sources) is of a sort customarily held in confidence by Holtec International, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by Holtec International. No public disclosure has been made, and it is not available in public sources. All disclosures to third parties, including any required transmittals to the NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary 2 of5

Holtec International Document ID 2635001-AFFIDAVIT-Ol AFFIDAVIT PURSUANT TO 10 CFR 2.390 agreements which provide for maintenance of the information in confidence. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following.

(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge.

Access to such documents within Holtec International is limited on a "need to know" basis.

(7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or his designee), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation.

Disclosures outside Holtec International are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements.

(8) The information classified as proprietary was developed and compiled by Holtec International at a significant cost to Holtec International. This information is classified as proprietary because it contains detailed descriptions of analytical approaches and methodologies not available elsewhere. This information would provide other parties, including competitors, with information from Holtec International's technical database and the results of evaluations performed by Holtec International. A substantial effort has been expended by Holtec International to develop this information. Release of this information would improve a competitor's position because it would enable Holtec' s competitor to copy our technology and offer it for sale in competition with our company, causing us financial injury.

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Holtec International Document ID 2635001-AFFIDAVIT-01 AFFIDAVIT PURSUANT TO 10 CFR 2.390 (9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to Holtec International's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of Holtec International's comprehensive spent fuel storage technology base, and its commercial value extends beyond the original development cost. The value of the technology base goes beyond the extensive physical database and analytical methodology, and includes development of the expertise to determine and apply the appropriate evaluation process.

The research, development, engineering, and analytical costs comprise a substantial investment of time and money by Holtec International.

The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial.

Holtec International's competitive advantage will be lost if its competitors are able to use the results of the Holtec International experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

The value of this information to Holtec International would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive Holtec International of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing these very valuable analytical tools.

4 of5

Holtec International Document ID 2635001-AFFIDAVIT-Ol AFFIDAVIT PURSUANT TO 10 CFR 2.390 STATE OF NEW JERSEY )

) ss:

COUNTY OF CAMDEN)

Mr. Debabrata (Debu) Mitra Majumdar, being duly sworn, deposes and says:

That he has read the foregoing affidavit and the matters stated therein are true and correct to the best of his knowledge, information, and belief.

Executed at Camden, New Jersey, this 5th day of June, 2017 Debabrata (Debu) Mitra Majumdar, Ph.D.

Corporate Director - Engineering Analysis Holtec International Subscribed and sworn before me this __5__ day of _ _ _Ju_n_e___, 2017.

5 of5

U.S. Nuclear Regulatory Commission Serial HNP-17-008 HNP-17-008 ATTACHMENT 5 HOLTEC INTERNATIONAL REPORT NO. HI-2177590, LICENSING REPORT FOR USE OF DREAM NEUTRON ABSORBER INSERTS IN THE SPENT FUEL POOLS A AND B AT SHEARON HARRIS NPP, REVISION 1 (NONPROPRIETARY)

SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400 RENEWED LICENSE NUMBER NPF-063 120 PAGES PLUS COVER

*

________________________________________________ H<Jltei_~_ Tei~~r:i<:>l<:>gyc:;eir:i!ei~. _ _Qriei . H<Jl!lc;f?<J_[Jll\f?f9,C::?_~glr:i.!'J~Q?1Q4 Telephone (856) 797-0900 H0 LTEC INTERNATIONAL Fax (856) 797-0909 LICENSING REPORT FOR USE OF DREAM NEUTRON ABSORBER INSERTS IN THE SPENT FUEL POOLS "A" AND "B" AT SHEARON HARRIS NPP FOR DUKE ENERGY NON-PROPRIETARY VERSION Holtec Report No: Hl-2177590 Holtec Project No: 2635 Sponsoring Holtec Division: NPD Report Class : SAFETY RELATED

SUMMARY

OF REVISIONS Revision 0 - Original Issue Revision 1 - Editorial/verbiage changes were made to Chapter 4. All changes are marked by revision bars.

Holtec Report HI-2177590 Holtec Project 2635

TABLE OF CONTENTS Section Section Title I.O INTRODUCTION ..................................................................................................... I-I I.I References .................................................................................................................. I-3 2.0 PROPOSED MODIFICATION, PRINCIPAL DESIGN CRITERIA &

REFERENCES .......................................................................................................... 2- I

2. I Introduction ................................................................................................................ 2-I 2.2 Summary of Principal Design Criteria ....................................................................... 2-I 2.3 Applicable Codes and Standards ............................................................................... 2-3 2.4 Quality Assurance Program ....................................................................................... 2-4 2.5 DREAM' Insert Mechanical Design ....................................................................... 2-5 3.0 MATERIAL CONSIDERATIONS ........................................................................... 3-I

3. I Introduction ................................................................................................................ 3- I 3.2 Materials Used in the DREAM' Insert .................................................................... 3-I 3.3 Neutron Absorbing Material ...................................................................................... 3- I 3.4 Compatibility with Environment ............................................................................... 3-4 3.5 Potential for Abrasion ................................................................................................ 3-4 3.6 References .................................................................................................................. 3-6 4.0 CRITICALITY SAFETY ANALYSES .................................................................... 4-I

4. I Introduction and Summary ........................................................................................ 4-I 4.2 Methodology .............................................................................................................. 4-4 4.3 Acceptance Criteria .................................................................................................. 4-26 4.4 Assumptions ............................................................................................................. 4-27 4.5 Input Data................................................................................................................. 4-27 4.6 Computer Codes ....................................................................................................... 4-29 4.7 Analysis .................................................................................................................... 4-29 4.8 Conclusion ............................................................................................................... 4-33 4.9 References ................................................................................................................ 4-35 Appendix 4A - Analysis Results Holtec Report HI-2I 77590 ii Holtec Project 2635

TABLE OF CONTENTS (continued)

Section Section Title 5.0 THERMAL-HYDRAULIC EVALUATION ............................................................ 5-1 5.1 Introduction ................................................................................................................ 5-1 5.2 Acceptance Criteria .................................................................................................... 5-2 5.3 Assumption and Design Data..................................................................................... 5-3 5.4 Bulk SFP Temperatures ............................................................................................. 5-6 5.5 Local Water and Fuel Cladding Temperatures .......................................................... 5-8 5.6 References ................................................................................................................ 5-14 6.0 EXISTING RACK STRUCTURAL/SEISMIC CONSIDERATIONS ..................... 6-1 6.1 Introd ucti on ................................................................................................................ 6-1 6.2 Seismic Qualification of Existing BWR Boraflex Racks .......................................... 6-1 6.3 Fuel Pool Structural Qualification ............................................................................. 6-2 6.4 Pool Liner Qualification ............................................................................................ 6-3 6.5 Mechanical Accident Evaluation ............................................................................... 6-3 6.6 Conclusion ................................................................................................................. 6-3 Holtec Report HI-2177590 lll Holtec Project 2635

1.0 INTRODUCTION

The Shearon Harris Nuclear Power Plant (HNP), owned and operated by Duke Energy Progress, is located in the extreme southwest comer of Wake County, North Carolina, and the southeast comer of Chatham County, North Carolina. The design of HNP incorporates the use of three spent fuel pools and one new fuel pool, as well as a cask loading pool. All of these pools are connected by a fuel transfer canal system. Spent fuel storage is provided by the New Fuel Storage Pool (Pool A) and the three spent fuel pools commonly referred to as Pool B, C, and D.

The four pools are licensed to include 3404 PWR storage cells and 4628 BWR storage cells for a total storage capacity of 8032 fuel assemblies.

There is one Westinghouse design for spent BWR fuel storage. The limiting design in Pools A and B is such that the keff for the racks will not exceed 0.95 with the spent fuel pool flooded with unborated water. With this limit on assembly reactivity, all fuel assemblies located in Brunswick Steam Electric Plant (BSEP or BNP) Unit 1 through reload 5 and all fuel assemblies located in BSEP Unit 2 through reload 6 are conservatively bounded and may be stored at HNP. The BWR design for all four pools do not require or contain flux traps, since subcriticality of all fuel is ensured by considering storage of fuel with the highest reactivity. BWR storage locations do not have a lead-in, since the fuel nozzle design facilitates insertion into the storage cell.

Since the installation of the BWR Boraflex spent fuel racks in Pools A and B, significant industry experience has indicated that the Boraflex degrades at a higher rate than originally anticipated. This degradation leads to a reduction of the ability of Boraflex to hold down reactivity sufficiently to ensure safe storage of fuel. NRC Generic Letter (GL 96-04) was issued in June 1996 informing power plants of concerns with the use of Boraflex material in spent fuel racks. The result of the degradation was acknowledged in the HNP supplemental response to GL 96-04 on April 25, 2005 to the NRC. The response details the actions taken and the coupon monitoring program and silica monitoring of the fuel pools. The Boraflex degradation phenomenon does not impact other BWR spent fuel storage racks at HNP that use Boral for reactivity suppression.

Holtec Report HI-2177590 1-1 Holtec Project 2635

Due to the degradation of the neutron absorbing materials in these racks that is credited in the HNP Technical Specifications, and the inability to relocate this fuel because of insufficient storage capacity, Duke Energy Progress, Inc., is seeking to install Rack Inserts into three (3)

Westinghouse BWR Boraflex Racks in Spent Fuel Pool (SFP) A and five (5) in SFP B. Pools A and B are located at the south end of the Fuel Handling Building and provide storage for new PWR and spent PWR and BWR fuel assemblies using a combination of various rack modules sizes.

The fuel rack enhancement program proposed by this application intends to rely on Metamic' inserts, designed and supplied by Holtec International, for reactivity control. Crediting Metamic' as the neutron absorber provides a robust means of ensuring that the current inventory of stored irradiated fuel can be accommodated without losing spent fuel storage locations and while maintaining an acceptable neutron multiplication factor. The Boraflex panels will remain in place providing additional (not credited) neutron absorption.

The Holtec inserts are known as DREAM' inserts. DREAM' is an acronym for Device for Reactivity Mitigation. This is a fully developed product and Holtec has the capability to mass-produce them efficiently. The use of similar inserts, supplied by Holtec to Florida Power and Light, has been reviewed and approved by the USNRC [1.7]. The method for storage cell criticality control enhancement proposed in this license amendment request is disclosed in several U.S. Patents [1.1-1.5].

The proposed rack enhancement program does not require any physical modifications to the existing storage rack arrays, other than insertion of the DREAM' inserts into the storage cells.

This report documents the design and analyses performed to demonstrate that the DREAM' inserts in the existing racks will meet all governing requirements of the applicable codes and standards; in particular the "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications" [1.6]. The aspects of the proposed rack enhancement program that are to Holtec Report HI-2177590 1-2 Holtec Project 2635

be implemented via this request to amend the station's operating license are described in the following paragraphs, and in the balance of this report.

Sections 2 and 3 of this report provide a brief abstract of the design and material information for the existing racks and a detailed description of the new DREAM' inserts. Section 4 provides a summary of the methods and results of criticality evaluations performed for the existing racks with the DREAM' inserts. Section 5 and 6 provide summary of the methodology and acceptance criteria used for the thermal and structural qualifications.

All computer programs utilized to perform the analyses documented in this report are benchmarked and verified. Holtec International has utilized these programs in numerous license applications over the past decade.

The analyses presented herein demonstrate that when the existing fuel storage racks at HNP Pools A and B are equipped with DREAM' inserts, the resulting modules possess wide margins of safety with respect to all the nuclear subcriticality considerations specified in the OT Position Paper [1.6].

1.1 References

[1.1] Metamic' U.S. Patent # 5,965,829 entitled "Radiation Absorbing Refractory Composition and Method of Manufacture" Dr. Kevin Anderson, Thomas G. Haynes III,

& Edward Oschmann, issued Oct. 12, 1999

[1.2] Metamic' U.S. Patent # 6,042,779 entitled "Extrusion Fabrication Process for Discontinuous Carbide Particulate Metal and Super Hypereutectic Al/Si Alloys" Thomas G. Haynes III and Edward Oschmann, issued March 28, 2000.

[1.3] Metamic' U.S. Patent # 6,332,906 entitled "Aluminum - Silicon Alloy Formed by Powder" Thomas G. Haynes III and Dr. Kevin Anderson, issued Dec. 25, 2001.

[1.4] Metamic' U.S. Patent Application 09/433773 entitled "High Surface Area Metal Matrix Composite Radiation Absorbing Product" Thomas G. Haynes III and Goldie Oliver, filed May 1, 2002.

Holtec Report HI-2177590 1-3 Holtec Project 2635

[1.5] U.S. Patent # 8,681,924B2 entitled "Singe-Plate Neutron Absorbing Apparatus and Method of Manufacturing the Same", Evan Rosenbaum, Thomas G. Haynes and Krishna P. Singh, March 25, 2014.

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

[1.7] Turkey Point Plant, Units 3 and 4 - Issuance of Amendments Regarding Spend Fuel Pool Boraflex Remedy," USNRC Letter from B. Mozafari to J.A. Stall, dated 17 July 2007.

Holtec Report HI-2177590 1-4 Holtec Project 2635

2.0 PROPOSED MODIFICATION, PRINCIPAL DESIGN CRITERIA & REFERENCES 2.1 Introduction As noted in Section 1, Duke Energy Progress, Inc. is seeking to install Rack Inserts into three (3)

BWR Westinghouse Boraflex Racks in Spent Fuel Pool (SFP) A and five (5) in SFP B. Pools A and B are located at the south end of the Fuel Handling Building and provide storage for new PWR and spent PWR and BWR fuel assemblies using a combination of various rack modules.

Each rack is a freestanding module, made primarily of austenitic stainless steel, and containing an array of interconnected storage cells. Nominal data for the Westinghouse BWR rack modules is presented in Table 2.1.

The rack enhancement proposed by this license amendment seeks to equip certain storage cells with Holtec DREAM inserts made of the neutron absorber Metamic'. This rack enhancement is designed to compensate for the ongoing loss in neutron attenuation capability of the originally installed Boraflex material. Details of the DREAM' insert design are provided in Section 2.5.

2.2 Summary of Principal Design Criteria The key design criteria for spent fuel storage racks are set forth in the USNRC memorandum entitled "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications" [2.7]. The "OT Position Paper" remains applicable for re-evaluation of the existing racks to consider the addition of the DREAM' inserts. The individual sections of this report address the specific design bases derived from the above-mentioned "OT Position Paper".

The design bases for the racks with DREAM' inserts are summarized in the following:

a. Rack Module Configuration: The rack modules remain freestanding during a seismic event.

Holtec Report HI-2177590 2-1 Holtec Project 2635

b. Kinematic Stability: Each freestanding module must be kinematically stable (resist tipping or overturning) under the plant's design basis seismic events.

c. Structural Compliance: All primary stresses in the rack modules must satisfy the limits in Section III Subsection NF of the ASME B&PV Code.

The DREAM' inserts to be installed in the rack modules are non-structural components.

Nevertheless, to ensure that they will continue to perform their intended function under all service conditions, the following requirement is imposed:

The allowable load for buckling is limited to 2/3 of the critical buckling load for conservatism and to ensure embedded margin in the design.

The above structural integrity requirement on the DREAM inserts is derived from Holtec Intemational's fuel rack design practice; it is not a prescribed requirement in the NRC or Code documents applicable to this project. Additionally, it is noted that this acceptance criterion is conservative in some respects, since failure of certain insert welds does not necessarily compromise the design function of the insert.

d. Criticality Compliance:

The objective of the criticality analysis is to ensure that the effective neutron multiplication factor (keff) is less than or equal to 0.95 with the pool flooded with borated water, and that it is less than 1.0 under the assumed accident of the loss of soluble boron in the pool water, i.e.

assuming unborated water in the spent fuel pool, all for 95% probability at a 95% confidence level.

e. Thermal Compliance: The objective of the thermal analysis is to demonstrate that the Westinghouse-supplied BWR spent fuel storage racks (SFSRs) in Shearon Harris spent fuel Holtec Report HI-2177590 2-2 Holtec Project 2635

pools (SFPs) A & B will continue to meet the thermal-hydraulic requirements for safe storage of spent fuel following installation of Holtec-supplied DREAM inserts.

The foregoing design bases are further articulated in Sections 4, 5 and 6 of this licensing report.

2.3 Applicable Codes and Standards The following codes, standards and practices are used as applicable for the design, construction, and assembly of DREAM' inserts. Because DREAM' inserts do not perform a structural function, only the criticality safety related codes and standards cited hereunder are germane to the evaluations and analyses presented in this report. Additional specific references related to detailed analyses are also provided, as appropriate.

[2.1] ASTM C750 - Standard Specification for Nuclear-Grade Boron Carbide Powder.

[2.2] ASTM C992 - Standard Specification for Boron-Based Neutron Absorbing Material Systems for Use in Nuclear Spent Fuel Storage Racks.

[2.3] ANSI N45.2.l - Cleaning of Fluid Systems and Associated Components during Construction Phase of Nuclear Power Plants - 1973 (R.G. 1.37).

[2.4] ANSI N45.2.2 - Packaging, Shipping, Receiving, Storage and Handling of Items for Nuclear Power Plants - 1972 (R.G. 1.38).

[2.5] ASME NQA Quality Assurance Program Requirements for Nuclear Facilities.

[2.6] ASME NQA Quality Assurance Requirements for Nuclear Power Plants.

[2. 7] "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.

[2.8] ANSI/ANS 8.1 - Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors.

[2.9] ANSI/ANS 8.17 - Criticality Safety Criteria for the Handling, Storage, and Transportation ofLWR Fuel Outside Reactors.

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[2.10] 10CFR21 - Reporting of Defects and Non-compliance.

[2.11] 10CFR50 Appendix B - Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants.

[2.12] 10CFR50.68 "Criticality Accident Requirements".

2.4 Quality Assurance Program The governing quality assurance requirements for design and fabrication of the spent fuel storage equipment are stated in 10CFR50 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, such as the DREAM' inserts, in accordance with the applicable codes, specifications, and regulatory requirements.

In recognition of the central role of the neutron absorber in maintaining subcriticality, Holtec International utilizes appropriately rigorous technical and quality assurance criteria and acceptance protocols to ensure satisfactory neutron absorber performance over the service life of the inserts.

Holtec Intemational's Q.A. program ensures that the neutron absorber material will be manufactured under the control and surveillance of a Quality Assurance/Quality Control Program that conforms to the requirements of 10CFR50 Appendix B, "Quality Assurance Criteria for Nuclear Power Plants". Consistent with its role in reactivity control, all neutron absorbing material in the Holtec products is categorized as Safety Related (SR). SR manufactured items, as required by Holtec's NRC-approved Quality Assurance program, must be produced to essentially preclude the potential of an error in the procurement of constituent materials and the manufacturing processes. Accordingly, material and manufacturing control processes must be established to eliminate the incidence of errors, and inspection steps are implemented to serve as an independent set of barriers to ensure that all critical characteristics defined for the material by Holtec's design team are met in the manufactured product.

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All major steps in the manufacture of Metamic TM are governed by formalized procedures. Raw materials (Al-6061 and B4C) used to make Metamic are obtained from qualified suppliers and overcheck analyses are performed to confirm the claims of the materials vendors. Separate mass spectroscopic determination of the fraction of the boron-10 nuclide in the boron is performed for each lot of B4C. Each batch mixture of B4C and Al-6061 is chemically analyzed to assure a composition that conforms to the design specification for the weight percentage of B4C.

Permanent records of these analyses with unique identification numbers are maintained in the Holtec QA files. Each completed Metamic' panel has a unique identification number that permits traceability to the material lot numbers of the constituent powders. Once the powders are thoroughly mixed, there is no known mechanism that might cause re-segregation of the powders.

After the isostatic pressing and sintering, the ingots are extruded and cleaned by glass-beading.

At this point, visual inspection confirms the removal of foreign particles from the surface of the extrusion piece. The extrusion piece is then rolled to a specified thickness and dimensions are confirmed with a precision jig. Random samples from the rolled panels are measured by neutron attenuation to confirm the proper B10 areal density and to qualify the homogeneity achieved in the fabrication process. As a qualified process, further neutron attenuation testing of the finished product is not required.

The Quality Assurance system enforced on the manufacturer's shop floor shall provide for all controls necessary to fulfill all quality assurance requirements. The final inspection and acceptance criteria of the manufactured DREAM' inserts focus on the insert's dimensions, bow, twist, profile, cleanliness, and identifying markings.

2.5 DREAM' Insert Mechanical Design The design objective for DREAM' inserts is to provide a neutron absorber having material composition and dimensions suitable for co-residence with fuel in a storage cell, and that can be easily inserted and relocated within the storage racks. Further, neutron absorption properties of the insert must be sufficient to eliminate reliance on the existing Boraflex. A major goal of the Holtec Report HI-2177590 2-5 Holtec Project 2635

program to eliminate reliance on Boraflex is to allow the use of every storage cell within the Westinghouse BWR Racks.

DREAM' inserts are designed for insertion in any Westinghouse BWR Rack storage cell subsequent to placing a fuel assembly in that cell. Each insert consists of a Metamic' panel attached to an aluminum upper "block" used to provide a robust attachment for handling purposes. Table 2.2 provides some basic data on the DREAM' insert design, which is shown in Figure 2.1.

An installed DREAM insert blankets two of the four walls of the host storage cell. The insert's upper aluminum block is equipped with interfaces for lifting and handling by an appropriate custom-designed tool. Different tools are used to manipulate fuel assemblies and DREAM' inserts. The position of the aluminum block at the top of the fuel assembly will be visually evident and it provides confirmation of the orientation of the DREAM' insert within each cell.

The design of the insert ensures that, when seated, the active fuel region is shadowed by Metamic'. The dimensional differences between a fuel assembly (5.787 inches square) and the inside dimension of a storage rack cell (nominally 6.05 inches square) provide a sufficient gap (0.263 inches) for the DREAM' insert to be inserted. The bottom edge of the Metamic' panel on each insert is skew cut and beveled to ensure that the insert will readily slide into this gap and not snag on any fuel assembly components.

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Table 2.1 MODULE DATA FOR EXISTING WESTINGHOUSE BWR RACKS 1 Storage cell inside nominal dimension 6.05 in Cell wall thickness 0.075 in Cell pitch 6.25 in Storage cell height 169 in Poison material Boraflex 1 All dimensions are nominal values.

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Table 2.2 DREAM' INSERT PHYSICAL PARAMETERS Maximum Width of DREAM' Insert Minimum Length of DREAM' Insert Nominal Thickness ofMetamic' Panel Metamic' Panel Minimum B4C Loading Approximate Weight of DREAM' Insert Holtec Report HI-2177590 2-8 Holtec Project 2635

Figure 2.1: DREAM' INSERT Holtec Report Hl-2177590 2-9 Holtec Project 2635

3.0 MATERIAL CONSIDERATIONS 3.1 Introduction A primary consideration in design of the DREAM' insert proposed in this amendment request is that materials introduced into the pool water be of proven durability and compatible with the fuel pool environment. This section summarizes the considerations that provide assurance that the DREAM' inserts installed in HNP Pool A and B Westinghouse BWR Racks will perform their intended function for the design life of the fuel racks.

3.2 Materials Used in the DREAM' Insert The Metamic' neutron absorber material is the principal material for manufacturing the insert.

Metamic' itself is comprised of aluminum alloy 6061 and boron carbide (B4C). All other components of the insert are manufactured from aluminum alloys, which are chemically compatible with Metamic.

3.3 Neutron Absorbing Material The Metamic' neutron absorber material is manufactured by the Orrvilon division of Holtec International in Ohio. As discussed below, Metamic' has been subjected to rigorous tests by various organizations, including Holtec International, and has been approved by the USNRC for use in wet storage (i.e., fuel pool) applications.

Holtec Report HI-2177590 3-1 Holtec Project 2635

Holtec Report HI-2177590 3-2 Holtec Project 2635 Holtec Report HI-2177590 3-3 Holtec Project 2635 3.4 Compatibility with Environment Typically, DREAM' inserts installed at HNP pools will experience an environment of heated demineralized water. Usual bulk water temperatures will be between 80°F and 150°F. Because both constituents of Metamic' (the Al 6061 alloy and boron carbide) are known to maintain physical and chemical stability, and Metamic' has no internal porosity (i.e., panels are fabricated at essentially 100% of the theoretical density) there is no known mechanism for Metamic's degradation in the HNP spent fuel pools. Further, over many years, it has been shown that galvanic corrosion of aluminum and aluminum alloys in contact with other metals (i.e.,

zircaloy or stainless steel) does not occur in water [3.1, 3.2, 3.4].

3.5 Potential for Abrasion

Holtec Report HI-2177590 3-4 Holtec Project 2635

Holtec Report HI-2177590 3-5 Holtec Project 2635 3.6 References

[3.1] "Qualification of METAMIC for Spent-Fuel Storage Applications", Report 1003137, EPRI, Palo Alto, CA, October 2001.

[3.2] "METAMIC Neutron Shielding", by K. Anderson, T. Haynes, and R. Kazmier, EPRI Boraflex Conference, November 19-20 (1998).

[3.3] "METAMIC 6061 + 40% boron Carbide Metal Matrix Composite Test Program for NAC International, Inc.", California Consolidated Technology, Inc. (2001).

[3.4] "METAMIC" Qualification Program for Nuclear Fuel Storage Applications, Final Test Results", Report NET 152-03, Prepared for Reynolds Metal Company, Inc. by Northeast Technology Corporation.

[3.5] "Use of METAMIC in Fuel Pool Applications," Holtec Information Report No. HI-2022871, Revision 1 (2002).

[3.6] "Sourcebook for Metamic' Performance Assessment" by Dr. Stanley Turner, Holtec Report No. HI-2043215 (2004).

[3.7] "Qualification of Metamic' for Use as a Neutron Poison Material, Holtec Report No.

HI-2033129 (2004).

Holtec Report HI-2177590 3-6 Holtec Project 2635

4. CRITICALITY SAFETY ANALYSIS 4.1 Introduction and Summary This report documents the spent fuel pool (SFP) criticality calculations performed for Duke Energy for the Harris site Pool A and B. The Harris SFP is designed for storage of Pressurized Water Reactor (PWR) fuel, but also contains storage racks designed for Boiling Water Reactors (BWR) fuel. The Harris SFP therefore also contains permanently discharged fuel previously shipped from the Brunswick Unit I and 2 BWR's. The purpose of this analysis is to qualify the BWR BORAFLEX' storage racks designed by Westinghouse in both Pool A and B using Holtec Metamic inserts.

Criticality control in the BWR BORAFLEX' storage racks DOES rely on the following:

Administrative Restrictions:

o Restriction I The BWR fuel designs allowed in the Harris Pool A and Pool B BWRBORAFLEX' storage racks are limited to the GE3, GE4, GE5, GE6 and GE7 fuel designs.

o Restriction 2 The orientation of the Metamic inserts in Harris Pool A and Pool B BWR BORAFLEX' storage racks are limited to the orientation shown in Figure 4.1.1 and Figure 4.1.2.

o Restriction 3 No fuel shall be stored in the storage cell at the northeast comer of the BWR BORAFLEX' storage racks nearest the interface with the PWR BORAFLEX' storage racks on the north and east side in Pool A.

Soluble boron for normal and accident conditions in accordance with 10CFR50.68(b)(4).

Holtec Metamic inserts.

Criticality control in the BWR BORAFLEX' storage racks DOES NOT rely on:

Residual amount ofBORAFLEX'.

Bumup or residual Gadolinium (Gd).

Holtec Report HI-2177590 Page 4-1 Holtec Project 2635

4.1.1 Special Considerations The criticality analysis presented in this report is unique in several ways from other BWR SFP criticality analyses. Additionally, the Metamic inserts used for criticality control have not been used for BWR fuel previously. Thus, the following special considerations exist:

Two SFP's are being considered, Pool A and Pool B.

SFP A has two rack designs, one for PWR fuel and one for BWR fuel.

SFP B has three rack designs, one for PWR fuel and two for BWR fuel.

The PWR racks in both Pool A and Pool Bare of the flux trap design and were originally Boraflex racks, however, no credit is taken for the Boraflex in the PWR racks.

The BWR racks in Pool B are both Region 2 style high density racks. One type is the Boraflex design (the same design as Pool A and thus intended to be controlled with Metamic inserts and therefore are bounded by the calculations in this analysis) and the other type is the Holtec Baral design. The Holtec Baral racks were designed to be dimensionally equivalent to the Boraflex design with the exception of using Bora! instead of Boraflex. Thus, the Baral racks have a much lower reactivity than the Boraflex racks (the case for both the original Boraflex credit and now with credit for Metamic inserts).

Since the SFP's contain PWR fuel (Harris is a PWR site), soluble boron credit is taken and established by the PWR analysis ofrecord [4.11].

The BWR fuel is from another site and was transferred to Harris and therefore the population of fuel is old and static. Additionally, BWR fuel movement is unlikely and not a normal part of plant operations.

Most BWR fuel is channeled, though some BWR fuel has been de-channeled.

All BWR fuel has significant bumup and extended cooling time.

The Holtec Metamic insert design, while new for use with BWR fuel, is not an entirely new design.

The Metamic insert rests in the storage rack and is not mechanically fixed to the rack cell.

The BWR fuel does not need to be removed prior to Metamic insert installation and the Metamic inserts must be removed prior to BWR fuel movement.

This analysis is specific only to the BWR Boraflex racks with Boraflex credit removal and Metamic insert credit established.

Three Administrative Restrictions are put in place via this criticality analysis to control various SFP operation parameters.

The special considerations listed above, as well as the various design parameters of the SFP and various storage racks, result in several unique approaches with respect to the criticality analysis.

Unique approaches for normal conditions:

A missing insert is required for the design basis model because the Metamic inserts must be removed prior to BWR fuel movement.

The normal condition soluble boron requirement, if needed, is already established by the PWR fuel analysis. Thus, any normal condition soluble boron calculations are performed with the Holtec Report HI-2177590 Page 4-2 Holtec Project 2635

limit amount. No interpolation is required as long as the limit requirement demonstrates regulatory compliance.

The Metamic insert orientation requirement was selected to ensure additional safety margin will exist. This margin is not credited in the analysis but the analysis considers that the Metamic insert panel was NOT always positioned between the BWR Boraflex racks and the PWR racks. Thus, the analysis already considers the bounding Metamic insert orientation for the interface of the BWR Boraflex racks and the PWR racks while the as-installed configuration will always be bounded by the analysis condition.

Rack to rack interfaces are evaluated in a conservative manner by considering minimum rack to rack gaps over both Pool A and Pool B. Thus, whichever rack to rack gaps from both pools are bounding (smallest gap) and are applied to the interface calculations so that both pools are bounded by the interface evaluations.

The interface evaluation is only applicable to the BWR Boraflex racks with Metamic inserts.

The other rack designs are not qualified by this analysis.

The analysis considers only fresh BWR fuel. An Administrative Restriction is applied via this analysis to ensure that the fuel in Pool A and Pool B is bounded by the analysis. No fuel movement is required to meet this Administrative Restriction. The purpose of the Administrative Restriction is therefore to ensure consistency with an already established requirement and to ensure that no changes are necessary for the soluble boron requirements established by another analysis.

Unique approaches for accident conditions:

The Metamic inserts have a specific orientation controlled by an Administrative Restriction.

The purpose of the orientation requirement is to maintain the Metamic inserts in a conservative configuration (the analysis considers the worst case) that is always bounded by this analysis. As a result ofthe orientation requirement a new accident condition is created if the Metamic inserts are not in the controlled orientation. While this is unlikely since the orientation of the insert will be administratively controlled through various plant operating controls, and the Metamic insert orientation is easily observable from above, the possibility remains that the Metamic insert, or more than one Metamic insert, could be orientated incorrectly. Thus, a new accident condition must be considered for multiple mis-oriented Metamic inserts. This accident is only important relative to the situation where an insert is already removed for the purpose of fuel movement. Thus, in that case, if the inserts around that location are mis-oriented, then that configuration would be an accident configuration.

The Metamic insert orientation requirement was also selected for conservatism with respect to the mislocated fuel assembly accidents. The orientation thus ensures that NO Metamic insert panel is between the mislocated fuel assembly and the BWR fuel.

For the mislocated fuel assembly accident, the PWR 17xl 7 fuel design is used. The PWR fuel is used because of the following conservatisms:

o The BWR fuel population is static and does not move.

o The BWR fuel is all spent and has a low reactivity.

o This PWR fuel may be fresh and thus have a high reactivity.

o The PWR fuel considered in the mislocated fuel assembly accident is treated as 5.0 wt% U-235, fresh and with no absorbers. Thus, while this is not possible since fresh 17xl 7 PWR fuel with a maximum enrichment will always have burnable absorbers, the Holtec Report HI-2177590 Page 4-3 Holtec Project 2635

mislocated accident calculations are very conservative.

o There are locations where the BWR fuel, if considered in place of the PWR fuel, could fit in locations where PWR fuel cannot fit. However, the accident cases considered bound all those locations. In one such case, the PWR fuel is made to fit where it normally could not by making small adjustments to the rack to rack gaps.

In general, the accident calculations considered are performed using Pool A configurations.

The Pool A configurations bound the Pool B configurations for the following reasons:

o The rack to rack gaps bound both pools.

o The BWR Boral racks, which are only in Pool B, are a much lower reactivity rack and thus they do not impact the reactivity of the BWR Boraflex racks.

o *The BWR Boral racks in Pool B might reduce the reactivity effect of the various mislocated accidents and thus it is conservative to neglect them.

o The Metamic insert orientation is controlled such that the analysis considers the worst case for the accident scenarios. Thus, the analysis calculations bound both pools.

o Therefore, every accident considered is applicable to both Pool A and Pool B, or, covers every possibility for Pool A and Pool B.

The accident condition soluble boron requirement is already established by the PWR fuel analysis. Thus, all accident condition calculations are performed with both pure water and the soluble boron limiting requirement. No interpolation is required as long as the limit requirement demonstrates regulatory compliance. Furthermore, in some cases accident scenarios may be more reactive at 0 ppm but another variation may be more reactive at the soluble boron requirement. Thus, the most reactive variation should be selected from the highest reactivity with soluble boron.

The analysis considers only fresh BWR fuel. An Administrative Restriction is applied via this analysis to ensure that the fuel in Pool A and Pool B is bounded by the analysis. No fuel movement is required to meet this Administrative Restriction. However, the accident analysis therefore requires special consideration for the unrealistic possibility that the most reactive BWR fuel design, the GE13, is accidently placed in Pool A or Pool B BWR Boraflex racks.

This accident scenario is technically a misload accident. The misload accident is performed in a very conservative manner to demonstrate the significant level of conservatisms which exists in the analysis. The GE13 fuel assembly is modeled as 5.0 wt% U-235 fresh while the GE13 actually is spent fuel that originally had a maximum enrichment of 4.2% U-235. The misload calculations also consider the fresh GE13 in every location, and also include the missing insert required for normal conditions. The results of those calculations show that even with that level of conservatism the results could have been used for normal conditions but are not to avoid potential changes to the TS requirements for soluble boron.

The specific details of the special considerations discussed above are provided in greater detail the sections below.

4.2 Methodology 4.2.1 General Approach In general, the analysis approach is to be as conservative as possible and whenever possible.

Additionally, the analysis is performed in a manner such that the results are below the regulatory Holtec Report HI-2177590 Page 4-4 Holtec Project 2635

limit with a 95% probability at a 95% confidence level. The calculations are performed using either the worst case bounding approach or the statistical analysis approach with respect to the various calculation parameters. The approach considered for each parameter is discussed below.

4.2.2 Computer Codes and Cross Section Libraries 4.2.2.1 MCNP5-l.51 MCNP5-l .51 [4.1] is used forthe criticality analyses. MCNP5-l .51 is a three-dimensional Monte Carlo code developed at the Los Alamos National Laboratory. MCNP was selected because it has a long history of successful use in fuel storage criticality analyses and has all of the necessary features for the analysis to be performed for Harris Pool A and B. MCNP5-1.51 calculations use continuous energy cross-section data predominantly based on ENDF/B-VII [4.2]. The default ENDF/B-VII cross sections are adjusted for temperature dependence using the appropriate continuous energy cross-section data processed with NJOY 99.396 code using ENDF/B-VII library [4.3, 4.4].

The convergence of a Monte Carlo criticality problem is sensitive to the following parameters: (1) number of histories per cycle, (2) the number of cycles skipped before averaging, (3) the total number of cycles and (4) the initial source distribution. All MCNP5-l.51 calculations are performed with a minimum of 12,000 histories per cycle, a minimum of 400 skipped cycles before averaging, and a minimum of 800 cycles that are accumulated. The initial source is specified the fueled regions (assemblies) and confirmed to converge. It is a well-known fact [4.5] that keff (eigenvalue), which is an integral quantity, converges much faster than the fission source spatial distribution (eigenfunction). However, a convergence of the spatial source distribution is important for estimating local quantities, such as pin power. To assist users in assessing the convergence of the fission source spatial distribution, MCNP5 computes a quantity called the Shannon entropy of the fission source distribution, Hsrc [4.1]. The Shannon entropy [4.5] is a well- known concept from information theory that has been shown to be an effective diagnostic measure for characterizing convergence and provides a single number for each cycle to help characterize convergence of the fission source distribution. It has been found that the Shannon entropy converges to a single steady-state value as the source distribution approaches stationarity. Therefore, the convergence of the power iteration process is ensured using the Shannon entropy, as implemented in MCNP5 [4.1].

Since the eigenvalue (keff) converges faster than the fission source distribution, the convergence of the keffiS assured by the convergence of the source distribution. The Shannon entropy convergence has been checked for each calculation.

4.2.2.2 MCNP5-l.51 Validation Benchmarking ofMCNP5-l .51 for criticality calculations is documented in [4.6]. The benchmarking is based on the guidance in [4.7], and includes calculations for a total of 562 critical experiments with fresh U02 fuel, fresh MOX fuel, and fuel with simulated actinide composition of spent fuel (HTC experiments [4.6]). The benchmarking area of applicability is presented in Table 4.2.1. The results of the benchmarking calculations for the full set of all 562 experiments are presented in Table 4.2.2 along with trending analysis. The statistical treatment used to determine those values considered the variance of the population about the mean and used appropriate confidence factors and trend Holtec Report HI-2177590 Page 4-5 Holtec Project 2635

analysis.

Trend analyses are also performed in [4.6], and the significant trends determined for various subsets and parameters are presented in Table 4.2.2. In order to determine the maximum bias that is applicable to the calculations in this report, the trend equations from [4.6] are evaluated for the specific parameters of the current analyses in Table 4.2.3. The results presented in Table 4.2.3 show the maximum bias and bias uncertainty associated with the benchmark subsets. This maximum bias and bias uncertainty is applied to all analysis calculations to determine keff.

4.2.3 Analysis Methods The overall analysis method considers a bounding analysis approach for the fuel, Metamic inserts and various storage rack models. The analysis models consider 12 inches of water above and below the active length of the fuel, thus the rack baseplate and other materials are modeled as water. This approach is acceptable because for fresh fuel the maximum reactivity is in the center of the fuel active length, not at the top or bottom. These bounding approaches and storage rack models are summarized below:

Bounding Fuel Designs and Fuel Assembly Parameters:

Each design basis analysis calculation considers fresh fuel with a uniform enrichment equal to the initial maximum planar average enrichment (IMPAE) plus the enrichment tolerance. The same bounding enrichment is considered along the entire active length for each fuel pin [4.8].

No Gd is included. Lower enriched blankets are neglected. Therefore, there is no axial or radial variation in fuel along the entire active length. This bounding approach provides analysis simplicity and margin since the fuel has significant bumup and cooling time and thus inherent negative reactivity. See Section 4.2.3 .1.1.

The bounding fuel assembly parameters are considered. This bounding approach provides analysis simplicity and margin since the fuel assembly parameter tolerance uncertainties are treated as a bias, not an uncertainty. See Section 4.2.3 .1.2.

Bounding Storage Rack Parameters:

The bounding rack design parameters are considered. This bounding approach provides analysis simplicity and margin since the rack design parameter tolerance uncertainties are treated as a bias, not an uncertainty. See Section 4.2.3.2.

The BORAFLEX' is replaced by water. This bounding approach provides analysis simplicity and margin since the remaining BORAFLEX' has negative reactivity that is not credited.

Each Metamic insert within a particular rack is administratively controlled to be installed with the same orientation (see Figure 4.1.1 and Figure 4.1.2), however, the storage cell in the center of the array is modeled with a missing Metamic insert. The center location is left with no Metamic insert because the Metamic insert design requires removal of the Metamic insert in order move a fuel assembly. Thus, after all locations have Metamic inserts installed, if a fuel assembly is to be moved, the Metamic insert must first be removed. Therefore the analysis considers a Metamic insert removed in the design basis model for both normal conditions and accident conditions.

For the misload accident, the missing insert location is also used for the misloaded fuel Holtec Report HI-2177590 Page 4-6 Holtec Project 2635

assembly location. This bounding approach provides analysis simplicity and margin since the worst case location for a missing insert is considered for both normal and accident conditions.

Bounding Melamie Insert Parameters:

The bounding thickness, width and B-10 loading for the Metamic inserts is considered.

Additionally, a slot along the entire length in the bend region is included in the model. This bounding approach provides analysis simplicity and margin since the reactivity effect of the insert design parameters are treated as an analysis bias rather than an uncertainty. Furthermore, the Metamic in the insert bend region that is not slotted (i.e. that remains after slotting) is ignored and modeled as water. See Section 42.3.3.

Bounding SFP Moderator Temperature:

The bounding SFP moderator temperature and density are used for all design basis calculations.

The calculations include NJOY corrected cross sections and S(a,p) cards. This bounding approach provides analysis simplicity and margin since the reactivity effect of the SFP reduction in temperature accident is treated for all normal and accident conditions, rather than as an additional accident condition. See Section 4.2.3.4.

Bounding Radial Positioning of Fuel assembly and Insert Location

The radial position of the fuel assembly and Metamic insert in the storage cell, as well as the presence or lack of the fuel assembly channel, is considered in a bounding fashion. This bounding approach provides analysis simplicity and margin since the reactivity effect of the radial location of the Metamic insert, fuel assembly along with the presence or lack of a channel is treated as an analysis bias rather than an uncertainty. See Section 42.3 .5.

Various fuel, rack and Metamic insert models are used in the analysis. The base model used for the analysis calculations in Section 4.2.3 consider the following MCNP5-1.51 rack model (with additional variations for the various evaluations described in each subsection):

1lxl1 array of storage cells with the BORAFLEX' replaced by water.

Cell geometry along the exterior considers a simple cell wall.

Adjacent to the exterior cell wall is a water gap with a thickness equal to half the BWR BORAFLEX' rack to BWR BORAFLEX' rack gap.

Periodic boundary conditions are considered along the water gap, thus creating a laterally infinite array.

All Metamic inserts have the orientations shown in Figure 4.1.1 and 4.1.2.

The center location of the 1lxl1 array does not include any Metamic insert.

The Metamic inserts are located between the fuel assembly and cell comer and have a bounding (assumed minimum) loading, width and thickness.

The storage rack design parameters are nominal values.

The SFP moderator is at 39.2 °F.

All materials in the model have the same temperature as the SFP moderator.

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Additional details and analysis methodology discussions are provided below.

4.2.3.l Bounding Fuel Design and Fuel Assembly Parameters The Harris SFP contains various GE BWR fuel designs (i.e. GE3, GE4, GE5, GE6, GE7, GE8, GE9, GEIO, and GE13), of which a subset (i.e. GE3, GE4, GE5, GE6 and GE7) is selected for potential storage in the Pool A and B BWR BORAFLEX' racks. The most reactive of that subset of fuel designs and that fuel designs bounding design parameters are determined below.

4.2.3.1.1 Bounding Fuel Design For normal conditions, the GE7 fuel design is expected to be the design basis assembly. The GE7 fuel design is expected to be more reactive than the GE3, GE4, GE5 and GE6 fuel designs and is therefore used to bound those lower reactivity fuel designs and qualify the GE3 through GE7 fuel designs for storage in the Pool A and B BWR BORAFLEX' racks. Note that the GE5 and GE6 are essentially the same and therefore the GE6 is used to evaluate both.

Since there is a restriction on which fuel designs can be stored in the Pool A and B BWR BORAFLEX' racks, an administrative requirement will control which fuel designs are allowed in the Pool A and B BWR BORAFLEX' racks. Therefore, consideration is taken for the potential misload accident where a fuel design not qualified for storage in the Pool A and B BWR BORAFLEX' racks is accidently loaded in a Pool A or B BWR BORAFLEX' storage cell location. Note that this is highly unlikely since the movement ofBWR fuel is uncommon and all non-qualified fuel designs have been removed from Pool A and B. However, for the purpose of this analysis, the accident condition is considered. Therefore, since the GE 13 is expected to be the most reactive of all fuel designs, the GE13 is used as the misload accident fuel design [4.8].

The bounding fuel design evaluation calculates the maximum reactivity of each fuel design using the 1 lxl 1 rack model discussed in Section 4.2.3 (with the center cell Metamic insert missing) and the following generic fuel design parameters unless otherwise noted:

IMPAE including the enrichment tolerance, unless otherwise stated.

Dimension parameters are nominal values.

The fuel assembly is cell centered.

The fuel channel present.

The following cases are evaluated:

Case 4.2.3.1.1.1: GE3 with maximum IMPAE +enrichment tolerance.

Case 4.2.3.1.1.2: GE4 with maximum IMPAE +enrichment tolerance.

Case 4.2.3 .1.1.3: GE6 with maximum IMPAE + enrichment tolerance.

Case 4.2.3.1.1.4: GE7 with maximum IMPAE + enrichment tolerance.

Case 4.2.3.1.1.5: GE 13 with maximum IMPAE + enrichment tolerance.

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The bounding fuel design parameters are determined using the same 1lxl1 rack model as discussed in Section 4.2.3 (with the center cell Metamic insert missing) with the exception for variations in fuel assembly parameters. For the fuel assembly parameter evaluations, each fuel assembly in the model has the tolerance applied at the same time.

Note that the fuel density increase and fuel pellet diameter increase cases are not considered since it is a well-known effect that maximum parameter values will increase reactivity (considering that the pellet to clad gap is not flooded in the model). Therefore, the reference case below (and all tolerance cases) includes the maximum fuel pellet density and maximum fuel pellet outer diameter.

The following BWR fuel tolerances [4.8] are considered for the bounding fuel design for nominal conditions (expected to be the GE7):

Case 4.2.3.1.2.1: Reference case. All tolerance parameters nominal (except fuel pellet diameter and density which are maximum).

Case 4.2.3.1.2.2: Minimum cladding thickness.

Case 4.2.3.1.2.3: Maximum cladding thickness.

Case 4.2.3.1.2.4: Minimum fuel rod pitch.

Case 4.2.3.1.2.5: Maximum fuel rod pitch.

Case 4.2.3.1.2.6: Minimum fuel channel thickness.

Case 4.2.3.1.2.7: Maximum fuel channel thickness.

Case 4.2.3.1.2.8: Minimum water rod thickness.

Case 4.2.3.1.2.9: Maximum water rod thickness.

Case 4.2.3.1.2.10: Pin specific enrichment for the IMPAE considered in Case 4.2.3.1.1.4.

Note that the fuel rod pitch tolerance calculations were increased and decreased by applying a 2 cr approach using the following equation:

Tolerance= pitch+/- 2*(tolerance)/sqrt(number of fuel rods)

The reactivity effect of each tolerance is determined from:

delta-kcalc = (kcalc2- kcalcl) +/- 2 * >I (m 2+ cri) where+/- 2 * -'1 (cr1 2+ cr22) is called the 95/95 uncertainty.

The results are then used to determine the bounding fuel assembly parameters for use in all design basis calculations.

Case 4.2.3.1.2.10 is evaluated to show that using the IMPAE is acceptable.

4.2.3.2 Bounding BWR BORAFLEX' Rack Parameters The bounding BWR BORAFLEX' rack parameters are determined using the same 1lxl1 rack model as discussed in Section 4.2.3 (with the center cell Metamic insert missing) with GE7 fuel with nominal fuel design parameters and with the exception for variations in BWR BORAFLEX' rack Holtec Report HI-2177590 Page 4-9 Holtec Project 2635

parameters.

The following tolerances are considered:

Case 4.2.3.2.1: Reference case. All storage rack parameters nominal.

Case 4.2.3.2.2: Minimum storage cell inner diameter.

Case 4.2.3.2.3: Maximum storage cell inner diameter.

Case 4.2.3.2.4: Minimum storage cell wall thickness.

Case 4.2.3.2.5: Maximum storage cell wall thickness.

Case 4.2.3.2.6:* Minimum storage cell pitch.

The reactivity effect of each tolerance is determined from:

delta-kcalc = (kcalc2- kcaicl) +/- 2 * '1 (cn 2+ cr22) where+/- 2 * '1 (cr1 2+ cri2) is called the 95/95 uncertainty.

Note that there is no maximum storage cell pitch tolerance.

The results are then used to determine the bounding BWR BORAFLEX' rack parameters for use in all design basis calculations.

4.2.3.3 Bounding Metamic Insert Parameters The bounding Metamic insert design parameters are determined using the same 1lxl1 rack model as discussed in Section 4.2.3 (with the center cell Metamic insert missing) with GE7 fuel with nominal fuel design parameters and with the exception for variations in Metamic insert design parameters.

Additionally, since it is expected that the position of the insert relative to the fuel will have an impact on reactivity (see Section 4.2.3.5), the Metamic insert position in the calculations in this section is adjacent to the storage cell comer. Therefore, the results of the calculations with various Metamic insert design parameter changes will resolve the reactivity impact of the Metamic insert design parameters and not the slight changes in relative distance to the fuel assembly.

The following tolerances are considered:

Case 4.2.3.3.1: Reference case. Minimum B4C wt%, nominal insert thickness and width (nominal loading at minimum B4C wt%). Metamic insert position is adjacent to the storage cell comer.

Case 4.2.3.3.2: Same as Case 4.2.3.3.1, but with alternative missing insert location along rack edge, in center location. Minimum B4C wt%, nominal insert thickness and width (nominal loading at minimum B4C wt%).

Case 4.2.3.3.3: Same as Case 4.2.3.3.1, but with alternative missing insert location along rack edge, in comer location. Minimum B4C wt%, nominal insert thickness and width (nominal loading at minimum B4C wt%).

Case 4.2.3.3.4: Same as Case 4.2.3.3.1, but with minimum B4C wt%, maximum thickness and width (maximum loading at minimum B4C wt%).

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Case 4.2.3.3.5: Same as Case 4.2.3.3.1, but with minimum B4C wt%, minimum thickness and width (minimum loading at minimum B4C wt%).

Case 4.2.3.3.6: Coupon testing measurement uncertainty. Same as Case 4.2.3.3.1, but with minimum B4C wt% - 5%, minimum thickness and width (minimum loading at minimum B4C wt% -

5%).

The reactivity effect of each tolerance is determined from:

delta-kca1c= (kca1c2-kcalcl) +/- 2 *.,,; (cr1 2+ cri2) where+/- 2 * -'1 (cr1 2+ cr22) is called the 95/95 uncertainty.

Case 4.2.3.3.2 and Case 4.2.3.3.3 are included to show that the missing Metamic insert in the center location is the bounding configuration.

Case 4.2.3.3.6 is provided for the coupon measurement uncertainty and the results of those calculations are not included in the design basis model. Rather, the results for Case 4.2.3.3.6 is statistically combined with the other analysis uncertainties to determine keff.

The results for Case 4.2.3.3.4 and Case 4.2.3.3.5 are evaluated to determine the bounding Metamic insert design parameters for use in all design basis calculations.

4.2.3.4 Reactivity Effect of Spent Fuel Pool Water Temperature The criticality analysis should be performed at the most reactive temperature and density [4.9].

Additionally there may be temperature-dependent cross section effects in MCNP5- l .51 that need to be considered. In general, both density and cross section effects are not necessarily the same for all storage rack scenarios, since configurations with strong neutron absorbers typically show a higher reactivity at lower water temperature, while configurations without such neutron absorbers typically show a higher reactivity at a higher water temperature. For the Harris Pool A and B BWR BORAFLEX' storage racks with Metamic inserts, the maximum reactivity condition therefore is expected to be the minimum SFP water temperature and maximum density.

Additionally, the standard cross section temperature in MCNP5-l.51 is 300 K. Cross sections are also available at other temperatures, however not usually at the desired temperature for SFP criticality analysis. MCNP5- l .51 has the ability to automatically adjust the cross sections to the specified temperature when using the TMP card. Additionally, MCNP5-1.51 has the ability to make a molecular energy adjustment for select materials (such as water) by using the S(a,~) card.

The S(a,p) card is provided for certain fixed temperatures which are not always applicable to SFP criticality analysis. Rather, there are limited temperature options, i.e. 300 K and 350 K, etc.

Additionally, MCNP5-1.51 does not have the ability to adjust the S(a,p) card for temperatures as it does for the TMP card discussed above. Therefore, the cross sections and S(a,p) card are adjusted using NJOY [4.3, 4.4], and these adjusted cross sections use a temperature that is reasonably close to the SFP specific values. This approach is acceptable because the system is poisoned and the reactivity trend is well known.

The Harris SFP have a normal water temperature operating range of 85 to 105 °F [4.8].

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Temperatures above and below this range are considered accidents. However, for conservatism, the design basis model will consider the bounding temperature.

Studies are performed to demonstrate the reactivity effect of the moderator temperature and density over the range 39.2 °F through 212 °F using temperature adjusted cross sections and S(a,~)cards.

The bounding temperature is determined using the same 1lxl1 rack model as discussed in Section 4.2.3 (with the center cell Metamic insert missing) with GE7 fuel (GE7 with IMPAE at 4.0 wt% U-235 and nominal fuel design parameters) and with the exception for variations temperature. The following studies are performed:

Case 4.2.3.4.1: Reference case. Temperature of39.2 °F (NJOY cross sections are 39.2 °F).

Same model as discussed in Section 4.2.3 .1. l.

Case 4.2.3.4.2. Minimum nominal temperature 85 °F (NJOY cross sections are 69 °F).

Case 4.2.3.4.3. Maximum nominal temperature 105 °F (NJOY cross sections are 149°F).

Case 4.2.3.4.4. Maximum possible temperature 212 °F (NJOY cross sections are 212 °F).

The reactivity effect of each tolerance is determined from:

delta-kcalc = (kcalc2- kcaicl) +/- 2 * -'1 (cn 2+ cri) where+/- 2 * -'1 (cr1 2+ cr22) is called the 95/95 uncertainty.

All design basis calculations consider the bounding moderator temperature and density using temperature adjusted cross sections and S(a,~) cards. Note that void formation is not considered credible [4.8].

4.2.3.5 Reactivity Effect of Fuel Assembly Channel and Radial Positioning of Fuel Assembly and Metamic Insert The reactivity effect off the fuel assembly with and without the channel should be considered coincidently with the radial position of the fuel and Metamic insert. The fuel assembly can be in the following main radial locations: cell centered, centered between the insert and opposite comer, and eccentrically positioned towards all four comers. Additionally, the two main positions for the fuel in the 1lxl1 array are all fuel towards the center location and all fuel away from the center location. The Metamic insert can be in the following main positions: adjacent to the cell comer and as far from the cell comer as possible while adjacent to the fuel assembly. Note that the incorrect orientation of an insert is considered an accident condition and is therefore discussed in Section 4.2.5.8. The presence or lack of a channel impacts the physical distance for the various radial locations for Metamic inserts and fuel assembly. The dominant reactivity effect from these parameters is expected to be the radial position of the fuel assembly. The bounding radial location for fuel and insert, along with presence or absence of the fuel channel, is evaluated using the same 1lxl1 rack model as discussed in Section 4.2.3 (with the center cell Metamic insert missing) with GE7 fuel (GE7 with IMPAE at 4.0 wt% U-235 and nominal fuel design parameters) and with the exception for the variations described below:

Fuel Assembly Cell Centered, Fuel Channel Present:

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Case 4.2.3.5.1: Reference case. Metamic inserts are between fuel assembly and cell corner.

Case 4.2.3.5.2: Metamic inserts are adjacent to the cell corner.

Case 4.2.3.5.3: Metamic inserts are adjacent to the fuel assembly.

Fuel Assembly Cell Centered, Fuel Channel Missing:

Case 4.2.3.5.4: Metamic inserts are centered between fuel assembly and cell corner.

Case 4.2.3.5.5: Metamic inserts are adjacent to the cell corner.

Case 4.2.3.5.6: Metamic inserts are adjacent to the fuel assembly.

Fuel Assembly Positioned Towards Rack Center, Fuel Channel Present:

Case 4.2.3.5.7: Metamic inserts are adjacent to the cell corner. All fuel assemblies are positioned towards the rack center while center location is centered in rack cell.

Case 4.2.3.5.8: Metamic inserts are adjacent to the cell corner. All fuel assemblies are positioned towards the rack center while center location is eccentric in rack cell.

Fuel Assembly Positioned Towards Rack Center, Fuel Channel Missing:

Case 4.2.3.5.9: Metamic inserts are adjacent to the cell corner. All fuel assemblies are positioned towards the rack center while center location is centered in rack cell.

Case 4.2.3.5.10: Metamic inserts are adjacent to the cell corner. All fuel assemblies are positioned towards the rack center while center location is centered in rack cell.

Fuel Assembly Positioned Away From Rack Center, Fuel Channel Present:

Case 4.2.3.5.11: Metamic inserts are adjacent to the cell comer. All fuel assemblies are positioned away from the rack center while center location is centered in rack cell.

Case 4.2.3.5.12: Metamic inserts are adjacent to the cell corner. All fuel assemblies are positioned away from the rack center while center location is eccentric in rack cell.

Fuel Assembly Positioned Away From Rack Center, Fuel Channel Missing:

Case 4.2.3.5.13: Metamic inserts are adjacent to the cell corner. All fuel assemblies are positioned away from the rack center while center location is centered in rack cell.

Case 4.2.3.5.14: Metamic inserts are adjacent to the cell corner. All fuel assemblies are positioned away from the rack center while center location is eccentric in rack cell.

The reactivity effect of each tolerance is determined from:

delta-kcalc = (kcaic2- kca!cl) +/- 2 * '1 (cr1 2+ cri) where+/- 2 * '1 (cr1 2+ cr22) is called the 95/95 uncertainty.

The maximum reactivity case is used for all design basis calculations.

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4.2.3.6 Design Basis Model The design basis model contains various bounding parameters that are determined from the results of the studies discussed in:

Section 4.2.3.1 for the bounding fuel design (expected to be GE7 for normal conditions and GE13 for the misload accident)

Section 4.2.3.2 for the bounding BWR BORAFLEX' storage rack parameters

Section 4.2.3.3 for the bounding Metamic insert parameters (expected to be minimum)

Section 4.2.3.4 for the bounding SFP moderator temperature and density (expected to be 39.2 OF)

Section 4.2.3.5 for the bounding radial position of the fuel assembly in the storage rack cell, the Metamic insert and the presence or lack of a fuel assembly channel.

The results of those studies are all considered in the design basis model for simplicity and margin.

Additionally, as discussed in Section 4.2.3 the design basis model considers:

1lxl1 array of storage cells with the BORAFLEX' replaced by water.

The 1lxl1 storage rack model replaces the cell geometry along the exterior with a simple cell wall.

Adjacent to the exterior cell wall is water with a thickness equal to half the rack to rack gap.

Periodic boundary conditions are considered along the water gap, thus creating a laterally infinite array.

The center location of the 1lxl1 array does not include any insert.

All materials in the model have the same temperature as the SFP moderator.

The design basis model described above is used for the following calculations:

Case 4.2.3.6.1: Design basis model maximum kca1cwith GE7 fuel with design IMPAE 3.282 wt% U-235.

Case 4.2.3 .6.2: Design basis model except with nominal fuel rod cladding thickness.

Case 4.2.3.6.3: Design basis model except with nominal fuel rod pitch.

Case 4.2.3.6.4: Design basis model except with nominal storage rack cell pitch.

Case 4.2.3.6.5: Design basis model except with maximum Metamic thickness and width.

Case 4.2.3.6.6: Design basis model with 500 ppm soluble boron.

The design basis model is presented in Figure 4.2.1.

4.2.3.6.1 Model Simplifications While the fuel and rack models used in the analyses are very detailed, they still contain a number of modeling simplifications. The following is a list of those simplifications:

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The 1lxl1 storage rack model replaces the formed cells along the exterior with the poison pocket and sheathing with a simple cell wall. Adjacent to the exterior cell wall is water with a thickness equal to half the rack to rack gap. Periodic boundary conditions are considered along the water gap, thus creating a laterally infinite array. This is acceptable for cells in the center of the racks, where the effect of any lateral neutron leakage would be negligible. This is conservative for cells on the periphery of the racks, specifically on the periphery of the pool, since it conservatively neglects the lateral leakage in those areas which will reduce reactivity.

Dishing and chamfering of the fuel pellets is neglected, i.e. the fuel is always modeled as solid cylinder inside the cladding. This is acceptable since the amount of fuel is maintained, and the water-to-fuel ratio and the principal location of the fuel remain unchanged.

Minor parts of the fuel and rack construction are neglected and replaced by water. Those include grid straps and minor structural rack components.

The residual B-10 amount in the BORAFLEX' in the racks is neglected in the analyses. The BORAFLEX' is replaced by water in the MCNP5-1.51 model. This is appropriate because as the absorber material degrades, the polymer and silica substrate which hold the B4C are removed together with the B4C and replaced by the water in the SPF.

All fuel cladding material is modeled as pure zirconium, while the actual fuel cladding consists of one of several zirconium alloys. This is acceptable since the model neglects the trace elements in the alloy which may provide additional neutron absorption.

4.2.3.7 Storage Rack Interfaces The Harris site Pool A and B BWR BORAFLEX' storage racks have the following rack to rack interfaces:

Pool A

BWR BORAFLEX' to PWRBORAFLEX'

BWR BORAFLEX'to BWR BORAFLEX' PoolB

BWR BORAFLEX'to PWRBORAFLEX'

BWRBORAFLEX'toBWRBORAL

BWR BORAFLEX'to BWR BORAFLEX' For both Pool A and B, the design basis model described in Section 4.2.3.6 considers the minimum BWR BORAFLEX' storage rack to rack gap and therefore bounds all BWR BORAFLEX' storage rack to storage rack interfaces for Pool A and Pool B.

For the non BWR BORAFLEX' storage rack to storage rack interfaces, the interface conditions are evaluated using the interface assumption: the interface does not result in a more reactive condition than the infinite array calculations. This is because the geometry and material conditions (of the racks) is assumed to create a physical configuration that does not allow for neutron coupling between racks. Since the more reactive condition (infinite arrays) is considered for the Holtec Report HI-2177590 Page 4-15 Holtec Project 2635

determination of keff, regulatory compliance is determined by the validation of the infinite array calculations and not the interface calculations. Rather, the interface qualification is determined by demonstration that the basic assumption is valid.

The interface assumption validation for the rack to rack interfaces is complicated by the fact that for any given reactivity calculations, the reactivity determined (i.e. kcaic) is necessarily the maximum reactivity at any single point in the calculation model, no matter how large or small the model. For example, if there are two racks in a spent fuel pool where Rack A is a high reactivity rack with a keff of 0.99 and Rack B is a low reactivity rack with a keff of 0.97 (where kerris from the infinite array calculations at the 95/95 level to meet regulatory requirements), and an interface calculation is performed with Rack A very far from Rack B (far enough to preclude any neutron coupling), the resulting calculated reactivity (kcalc) will necessarily be due to Rack A. This example calculation does not provide any useful information whatsoever about the reactivity of Rack B or about the interface between Rack A and Rack B and therefore cannot be used to validate the interface assumption. For the interface calculation, if the calculation is performed by placing Rack A very close to Rack B and the result of the calculation is that the interface kcalc is less than the infinite array kcalc for the loading pattern in Rack A, then the interface calculation has determined that the interface assumption for Rack A only is validated. If the kcalc of the interface calculation is higher than the infinite array calculations for the loading pattern in Rack A then no determination can be made about which rack has the higher reactivity and the interface assumption has not been validated. Therefore, because of this complication, various approaches have been developed to validate the interface assumption for such cases. The approach selected for the Harris Pool A and B BWR BORAFLEX' storage rack interfaces is discussed below.

Note that the interface evaluation presented in this analysis is used to qualify the BWR BORAFLEX' storage racks with Metamic inserts only. The BWR BORAL storage racks and PWR BORAFLEX' storage racks are qualified under separate analysis [4.11, 4.12], respectively.

Additionally, note that:

In [4.1 O] depletion calculations were performed to generate the 17x 17 PWR spent fuel isotopic compositions for use in the interface and accident evaluations. Those same isotopic compositions are used in this analysis. The spent PWR fuel isotopic compositions are not recalculated, therefore the methodology for their determination is referred to in [4.1 O].

The PWR BORAFLEX' storage racks do not credit residual BORAFLEX' and may contain either a checkerboard of fresh and empty cells or spent fuel in a uniform loading pattern [4.11].

The BWR BORAL racks credit BORAL and consider fresh BWR fuel only [4.12]. The BWR BORAL racks have the same BWR fuel design restrictions that are imposed on the BWR BORAFLEX'racks (i.e. the GE7 is the design basis fuelassembly).

The BWR BORAL storage rack contains 90% of the minimum B-10 and has the minimum thickness (see Table 4.5.2).

For the interface calculations, the BWR BORAL storage racks and the BWR BORAFLEX' storage rack are fully loaded with fresh GE7 fuel at the design IMPAE and consider that same fuel design parameters as determined for the BWR BORAFLEX' storage rack with Metamic inserts bounding design basis model (see Case 4.2.3.6.1).

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The 17x 17 fuel design which is fully characterized in Table 4.5. l .a is considered bounding of the other PWR fuel designs at Harris [4.11]. Note that the PWR BORAFLEX' storage rack design parameters, as well as the fuel design parameters, are treated at nominal values.

The interface assumption for the BWR BORAFLEX' storage rack with Metamic inserts is validated using the following approach:

A single storage rack model with reflective boundary conditions along the half rack to rack water gap is developed for the three storage rack designs.

The reactivity of the infinite array for each rack design is evaluated to determine if additional calculations are needed.

o The bounding BWR BORAFLEX' storage rack model, i.e. Case 4.2.3.6.1, is evaluated since this is the model which shows compliance with the regulatory requirements.

o The evaluation of the BWR BORAL storage racks with GE7 fresh fuel and PWR BORAFLEX'storage racks with both fresh and spent 17xl 7 fuel will determine ifthe reactivity of those configurations is significantly lower than the BWR BORAFLEX' storage racks with Metamic inserts.

For the PWR BORAFLEX' storage rack infinite array model with spent fuel, it is necessary to perform additional full pool interface calculations since the bumup requirements were determined in another analysis [4.11]. The full pool interface calculations require that the PWR fuel reactivity that is essentially the same as the reactivity of the BWR BORAFLEX' storage rack with Metamic inserts. Therefore, the evaluation of the PWR BORAFLEX' storage rack infinite array model is evaluated at a bumup which yields the appropriate infinite array reactivity. The same bumup will then be used in the full pool interface calculations. See Section 4.5.1 for a description of the PWR spent fuel isotopics.

Full pool models are developed to create interface models that are used to calculate the reactivity of the various interfaces.

The interface assumption is then validated by evaluating the reactivity of the full pool interface model(s). The requirement is that the reactivity of the full pool interface model must be no greater than the reactivity of the BWR BORAFLEX'storage rack with Metamic inserts.

The following interface assumption criteria cases are evaluated:

Case 4.2.3.7.l: BWR BORAFLEX' storage rack infinite array. Same as design basis model (Case 4.2.3.6.1 ).

Case 4.2.3.7.2: BWR BORAL storage rack infinite array. Fuel is centered in storage rack. See Figure 4.2.2.

Case 4.2.3. 7 .3: PWR BORAFLEX' storage rack infinite array with a uniform loading of spent PWR with an initial enrichment of 5.0 wt% U-235 fuel at a bumup which yields essentially the same reactivity as Case 4.2.3.7.1. Fuel is centered in storage rack. See Figure 4.2.3.

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Case 4.2.3.7.4: PWR BORAFLEX' storage rack infinite array with a checkerboard of empty and fresh PWR with an enrichment of 5.0 wt% U-235. Fuel is centered in storage rack. See Figure 4.2.4.

The interface evaluations are performed, if needed, for the various rack to rack interfaces using full pool models for Harris Pool A. Note that the BWR BORAL racks are essentially identical to the BWR BORAFLEX TM racks with the exception of the neutron absorber. Therefore, it is expected that the reactivity of the BWR BORAL racks will be bounded (significantly lower reactivity) by the reactivity of the BWR BORAFLEX TM racks. Since the rack designs are essentially identical, and the same fuel storage restrictions apply to both storage rack designs, it is expected that additional interface calculations for the BWR BORAFLEX TM storage rack to BWR BORAL storage rack interface will not be necessary because the design basis BWR BORAFLEX

'model (Case 4.2.3.6.1) bounds both rack designs. Furthermore, since Pool A and B contain PWR BORAFLEX TM storage racks, but Pool A does not contain BWR BORAL storage racks, only Pool A full pool interface calculations will be necessary.

The full pool model (Pool A only) interface cases are:

Case 4.2.3.7.5: BWR BORAFLEX' storage racks are the same as Case 4.2.3.6.l (bounding design basis configuration). The PWR BORAFLEX' storage rack contain a uniform loading of spent PWR with an initial enrichment of 5.0 wt% U-235 fuel and the same bumup as Case 2.3.7.3. The PWR fuel is positioned eccentrically towards the BWR BORAFLEX' storage racks. See Figure 4.2.5.

Case 4.2.3.7.6: Same as Case 4.2.3.7.5 except that the PWR fuel is cell centered.

Case 4.2.3.7.7: The model is the same as Case 4.2.3.7.5 except that the PWR BORAFLEX' storage rack contain a checkerboard of empty and fresh PWR with an enrichment of 5.0 wt% U-235.

Case 4.2.3.7.8: Same as Case 4.2.3.7.7 except that the PWR fuel is cell centered.

Case 4.2.3.7.9: Same as Case 4.2.3.7.5 except the BWR fuel is eccentrically positioned towards the PWR fuel.

Case 4.2.3.7.10: Same as Case 4.2.3.7.9 except the PWR fuel is cell centered.

Case 4.2.3.7.11: Same as Case 4.2.3.7.7 except the BWR fuel is eccentrically positioned towards the PWR fuel.

Case 4.2.3.7.12: Same as Case 4.2.3.7.11 except the PWR fuel is cell centered.

Case 4.2.3.7.13: Same as Case 4.2.3.7.5 except there is no PWR fuel in the PWRracks.

Case 4.2.3.7.14: Same as Case 4.2.3.7.13 except the BWR fuel is eccentrically positioned towards the PWR racks.

Note that the rack to rack gaps are considered in a conservative manner. The absolute minimum gaps allowed by the rack baseplate extensions are considered [4.8], with this minimum distance being measured from exterior rack wall to exterior rack wall (i.e. not from exterior sheathing to exterior sheathing). Additionally, there are minor differences in the gaps between racks for the full pool model (Case 4.2.3.7.5 through Case 4.2.3.7.14) when compared to the infinite array cases models (Case 4.2.3.7.l through Case 4.2.3.7.4). For the infinite array models (Case 4.2.3.7.1 through Case 4.2.3.7.4), the periodic boundary conditions along the half water gap defines the rack Holtec Report HI-2177590 Page 4-18 Holtec Project 2635

to rack gaps. The full pool model is more complex and therefore the model ensures that the gaps remain conservative (i.e. as small as possible) by making them smaller in some cases. Furthermore, the large gap in Pool A between the rows of PWR BORAFLEX' storage racks is modeled as a large gap, as installed and described in [4.8].

4.2.4 Fuel Movement, Reconstitution etc The Harris SFP is essentially a repository for spent BWR fuel. Therefore, typical SFP activities such as BWR fuel movement, reconstitution, channeling/dechanning efforts are not normal activities for the BWR fuel. In the case where BWR fuel is moved, the various dropped accident and misload activities are covered by the accident conditions in Section 4.2.5. Fuel activities related to the PWR fuel are also addressed considering the accident conditions (see Section 4.2.5).

4.2.5 Accident Conditions There are various potential accident conditions that must be evaluated. The credible accidents to be evaluated are:

The effect of SFP temperature exceeding the normal range.

A dropped fuel assembly:

A misloaded fuel assembly (a fuel assembly in the wrong location within the storage rack)

A mislocated fuel assembly (a fuel assembly in the wrong location outside the storage rack).

Rack movement due to seismic activity.

Mis-orientation ofMetamic inserts.

As discussed previously in Section 4.2.3.7, the 17xl 7 5.0 wt% fuel design is considered for the interface and accident analysis. This fuel design is considered bounding of the other 15x15 PWR fuel designs at Harris [4.11] and therefore no separate calculations are performed. The l 7xl 7 fuel design is fully characterized in Table 4.5.1.

For each accident condition considered, an additional calculation is performed with a soluble boron concentration of 1000 ppm. From the full set of accident conditions, the most reactive case is determined from the highest reactivity at 1000 ppm soluble boron.

4.2.5.l Temperature and Water Density Effects The SFP water temperature accident conditions for consideration are the decrease and increase in SFP water temperature below and above the nominal SFP temperature range of 85 to 105 °F. The decrease and increase in temperature is evaluated in Section 4.2.3.4. The increase in temperature accident that leads to boiling in the Harris SFP A and B is riot considered credible (see Section 5.0).

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4.2.5.2 Dropped Assembly-Horizontal For the case in which a fuel assembly is assumed to be dropped 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 of more than 12 inches [4.8], which is sufficient to preclude neutron coupling (i.e.,

an effectively infinite separation). Consequently, the horizontal fuel assembly drop accident will not result in a significant increase in reactivity. Furthermore, any reactivity increase would be small compared to the reactivity increase created by the misloading of a fresh assembly discussed in one of the following sections. The horizontal drop is therefore bounded by this misloading accident and no separate calculation is performed for this drop accident.

4.2.5.3 Dropped Assembly- Vertical into a Storage Cell It is also possible to vertically drop an assembly into a location that might be occupied by another assembly with a Metamic insert or that might be empty. Such a vertical impact would at most cause a small compression of the stored assembly, if present, or result in a small deformation of the baseplate for an empty cell. The damage to the Metamic insert would be minimal due to the necessary presence of the fuel assembly which would absorb majority of the drop accident deformation. These deformations could potentially increase reactivity. However, the reactivity increase would be small compared to the reactivity increase created by the misload of a fresh assembly discussed in one of the following sections. The vertical drop is therefore bounded by this misload accident and no separate calculation is performed for this drop accident.

4.2.5.4 Misloaded Fresh Fuel Assembly As discussed in Section 4.1, an administrative requirement to restrict the fuel assembly designs allowed in Pool A and B will preclude the presence of the higher reactivity designs GE8, GE9, GElO and GE13. However, it is possible that one of those restricted fuel assembly designs could be accidently placed in the BWR BORAFLEX' racks in Pool A and B. Therefore, the follow multiple misload accident condition is evaluated:

Case 4.2.5.4.1: Same model as Case 4.2.3.6.1 except the GE7 fuel assembly in every location is replaced by a fresh 5.0 wt% U-235 GE13 fuel assembly. The GE13 fuel in the model considers nominal design parameters and is channeled and include the following assumptions apply: no part length rods (considered the more reactive lattice [4.11]) or axial blankets and the axial length increased to 150 inches. Note that the enrichment of the fresh GE13 fuel assembly is increased to 5.0 wt% U-235 whereas the actual IMPAE is less than 4.2 wt% U-235. Note that this configuration is very conservative and bounds all possible multiple misload configurations.

Thus, studies are not performed with the GE13 to determine the bounding fuel design parameters. The SFP moderator has 0 ppm soluble boron.

Case 4.2.5.4.2: Same model as Case 4.2.5.4.l exceptthatthe SFP moderator has 1000 ppm soluble boron.

The misload accident calculations are performed with models that also cover the unlikely scenario of multiple misloaded GE13 fuel assemblies.

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4.2.5.5 Mislocated Fresh Fuel Assembly The possibility exists that a BWR or PWR fuel assembly could be accidently mislocated outside of the BWR BORAFLEX' racks in either Pool A or B. The three bounding cases are the fresh GE 13 (similar to the mislead accident discussed in Section 4.2.5.4), the fresh PWR fuel assembly and the spent PWR fuel assembly. Of those three cases the fresh 17xl 7 PWR assembly with no burnable absorbers is the bounding mislocated accident scenario due to the larger size and actual physical presence of the fresh PWR fuel in the SFP (the GE13 fuel is all spent and administratively restricted from Pool A and B).

The Harris SFP A and B layout was examined to determine possible worst case locations for a mislocated fresh PWR fuel assembly adjacent to the BWR BORAFLEX' storage racks. The worst possible locations would be in locations where the mislocated fuel assembly would be face adjacent to two or more fuel assemblies in storage racks with no poison between them. The worst case locations are shown in Figure 4.2.6. This location, although not physically possible, bounds all other possible configurations because the mislocated fuel may be face adjacent, with no neutron absorber between them, to up to three fuel assemblies (one BWR and up to two PWR assemblies in the racks).

Therefore, this scenario, although not physically possible, is evaluated and no other mislocated cases are required.

The base case model for the mislocated fuel assembly considers the following:

The model is a full pool model (see Figure 4.2.6) with minimum rack to rack gaps .considered along with a water reflector on the exterior of the rack. The water reflector is acceptable because the reactivity in the model is dominated either in the center of the BWR BORAFLEX'rack or in the location of the mislocated PWR fuel assembly. Thus, the geometry at the exterior of the racks along the pool wall is of no concern.

The fresh GE7 BWR fuel is eccentrically positioned toward the center BWR BORAFLEX'rack center (i.e. the bounding configuration for normal conditions). All BWR fuel and BORAFLEX' rack parameters are considered in the bounding manner, as concluded in Section 4.2.3.6, including that the central location does not include a Metamic insert. All Metamic inserts have the orientation shown in Figure 4.1.1.

The PWR fuel is a uniform loading of spent fuel and the PWR fuel is cell centered.

The gap between the PWR racks where the PWR fuel assembly is mislocated is widened slightly (by allowing the surface which defines the rack exterior to cut off most of the external sheathing).

This is done to allow the mislocated fuel to just fit in the gap.

The mislocated fresh 5.0 wt% U-235 17xl 7 PWR fuel assembly is face adjacent and at its closest approach to the BWR fuel.

The SFP moderator has 0 ppm soluble boron.

The following cases are therefore considered with variations to the above base case as noted:

Case 4.2.5.5.l: Same as base case.

Case 4.2.5.5.2: Same as Case 4.2.5.5.1 except the SFP moderator has 1000 ppm soluble boron.

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Case 4.2.5.5.3: Same as Case 4.2.5.5.1 except the PWR fuel is a checkerboard of fresh fuel and empty cells.

Case 4.2.5.5.4: Same as Case 4.2.5.5.3 except the SFP moderator has 1000 ppm soluble boron.

Case 4.2.5.5.5: Same as Case 4.2.5.5.1 except the BWR fuel is eccentrically positioned toward the mislocated PWR fuel.

Case 4.2.5.5.6: Same as Case 4.2.5.5.5 except the SFP moderator has 1000 ppm soluble boron.

Case 4.2.5.5.7: Same as Case 4.2.5.5.5 except the PWR fuel is a checkerboard of fresh fuel and empty cells.

Case 4.2.5.5.8: Same as Case 4.2.5.5.7 except the SFP moderator has 1000 ppm soluble boron.

Case 4.2.5.5.9: Same as Case 4.2.5.5.1 except the mislocated PWR fuel face adjacent to the PWR fuel in the PWR BORAFLEX' racks.

Case 4.2.5.5.10: Same as Case 4.2.5.5.9 except the SFP moderator has 1000 ppm soluble boron.

Case 4.2.5.5.11: Same as Case 4.2.5.5.3 except the mislocated PWR fuel face adjacent to the PWR fuel in the PWR BORAFLEX'racks.

Case 4.2.5.5.12: Same as Case 4.2.5.5.11 except the SFP moderator has 1000 ppm soluble boron.

Case 4.2.5.5.13: Same as Case 4.2.5.5.5 except the mislocated PWR fuel face adjacent to the PWR fuel in the PWR BORAFLEX' racks.

Case 4.2.5.5.14: Same as Case 4.2.5.5.13 except the SFP moderator has 1000 ppm soluble boron.

Case 4.2.5.5.15: Same as Case 4.2.5.5.7 except the mislocated PWR fuel face adjacent to the PWR fuel in the PWR BORAFLEX' racks.

Case 4.2.5.5.16: Same as Case 4.2.5.5.15 except the SFP moderator has 1000 ppm soluble boron.

Administrative Restriction 3 Variations As discussed in Section 4.1, Administrative Restriction 3 requires that:

No fuel shall be stored in the storage cell at the northeast comer of the BWR BORAFLEX' storage racks nearest the interface with the PWR BORAFLEX' storage racks on the north and east side in Pool A.

Therefore, to determine the reactivity effect of Restriction 3, the more bounding mislocated accident conditions discussed in Cases 4.2.5.5.1through4.2.5.5.16 above are modified to remove the BWR fuel assembly in the model. The following cases are evaluated:

Case 4.2.5.5.17: Same as Case 4.2.5.5.1 except the comer BWR fuel assembly is removed.

Case 4.2.5.5.18: Same as Case 4.2.5.5.17 except the SFP moderator has 1000 ppm soluble boron.

Case 4.2.5.5.19: Same as Case 4.2.5.5.17 except the PWR fuel is a checkerboard of fresh fuel and empty cells.

Case 4.2.5.5.20: Same as Case 4.2.5.5.19 except the SFP moderator has 1000 ppm soluble boron.

Case 4.2.5.5.21: Same as Case 4.2.5.5.17 except the BWR fuel is eccentrically positioned toward the mislocated PWR fuel.

Case 4.2.5.5.22: Same as Case 4.2.5.5.21 except the SFP moderator has 1000 ppm soluble boron.

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Case 4.2.5.5.23: Same as Case 4.2.5.5.21 except the PWR fuel is a checkerboard of fresh fuel and empty cells.

Case 4.2.5.5.24: Same as Case 4.2.5.5.23 except the SFP moderator has 1000 ppm soluble boron.

Case 4.2.5.5.25: Same as Case 4.2.5.5.17 except the mislocated PWR fuel face adjacent to the PWR fuel in the PWR BORAFLEX' racks.

Case 4.2.5.5.26: Same as Case 4.2.5.5.25 except the SFP moderator has 1000 ppm soluble boron.

Case 4.2.5.5.27: Same as Case 4.2.5.5.19 except the mislocated PWR fuel face adjacent to the PWR fuel in the PWR BORAFLEX' racks.

Case 4.2.5.5.28: Same as Case 4.2.5.5.27 except the SFP moderator has 1000 ppm soluble boron.

Case 4.2.5.5.29: Same as Case 4.2.5.5.21 except the mislocated PWR fuel face adjacent to the PWR fuel in the PWR BORAFLEX' racks.

Case 4.2.5.5.30: Same as Case 4.2.5.5.29 except the SFP moderator has 1000 ppm soluble boron.

Case 4.2.5.5.31: Same as Case 4.2.5.5.23 except the mislocated PWR fuel face adjacent to the PWR fuel in the PWR BORAFLEX' racks.

Case 4.2.5.5.32: Same as Case 4.2.5.5.31 except the SFP moderator has 1000 ppm soluble boron.

Additional Mislocated Accident Cases The two sets of mislocated accident cases above, i.e. Case 4.2.5.5.l through Case 4.2.5.5.32, considered a water reflector on the outside of the storage racks in the full pool model. This assumption is acceptable because the mislocated fuel assembly is in the center of the storage racks.

Additional calculations are performed forthe same configurations as Case 4.2.5.5.1 through Case 4.2.5.5.32 except with a concrete wall surrounding the SFP. Additionally, an alternative mislocated accident fuel assembly location, i.e. between the SFP wall and BWR rack is also provided to demonstrate that the bounding configuration is provided in Case 4.2.5.5.1 through Case 4.2.5.5.32.

The concrete wall is assumed to have the bounding material composition described in [4.15] and a thickness of40 cm (an infinite reflector for concrete). The pool liner is neglected. The descriptions of these cases is presented in Appendix 4.B.

4.2.5.6 Rack Movement During seismic activity the storage racks may move. Since Pool A only credits the Metamic inserts (i.e. Pool B has BWR BORAL racks) only Pool A is considered for this accident. The following evaluations are performed:

The base case model for the rack movement accident considers the following:

The model is a full pool model (similar to Figure 4.2.6) which considers nearly closed rack to rack gaps along with a water reflector on the exterior of the rack. The water reflector is acceptable because the reactivity in the model is dominated either in the center of the BWR BORAFLEX' rack or in the location in the center of the model where the PWR racks are adjacent to two sides of the BWR racks. Thus, the geometry at the exterior of the racks along the pool wall is of no concern.

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The fresh GE7 BWR fuel is eccentrically positioned toward the center BWR BORAFLEX'rack center (i.e. the bounding configuration for normal conditions). All BWR fuel and BORAFLEX' rack parameters are considered in the bounding manner, as concluded in Section 4.2.3.6, including that the central location does not include a Metamic insert. All Metamic inserts have the orientation shown in Figure 4.1.1.

The PWR fuel is a uniform loading of spent fuel and the PWR fuel is cell centered.

All gaps are reduced so that the racks are closer than possible due to the baseplate extensions.

The SFP moderator has 0 ppm soluble boron.

The following cases are therefore considered with variations to the above base case as noted:

Case 4.2.5 .6.1: Same as base case.

Case 4.2.5.6.2: Same as Case 4.2.5.6.1 except the SFP moderator has 1000 ppm soluble boron.

Case 4.2.5.6.3: Same as Case 4.2.5.6.1 except the PWR fuel is a checkerboard of fresh fuel and empty cells.

Case 4.2.5.6.4: Same as Case 4.2.5.6.3 except the SFP moderator has 1000 ppm soluble boron.

Case 4.2.5.6.5: Same as Case 4.2.5.6.1 except the BWR fuel is eccentrically positioned toward the location where two PWR racks are adjacent to the BWRrack.

Case 4.2.5.6.6: Same as Case 4.2.5.6.5 except the SFP moderator has 1000 ppm soluble boron.

Case 4.2.5.6.7: Same as Case 4.2.5.6.5 except the PWR fuel is a checkerboard of fresh fuel and empty cells.

Case 4.2.5.6.8: Same as Case 4.2.5.6.7 except the SFP moderator has 1000 ppm soluble boron.

4.2.5.7 Missing Metamic Inserts As discussed in Section 4.2.3.6, the design basis models consider the missing Metamic insert already, in conjunction with the misload accident (see Section 4.2.5.4). Therefore, no further evaluations are required.

4.2.5.8 Mis-orientation of Metamic Inserts As discussed in Section 4.1, the orientation of the Metamic inserts is controlled by an administrative requirement. Therefore, the accident condition of mis-oriented inserts should be evaluated. For the mis-orientation accident, the Metamic inserts are present in all storage rack locations, however, one or more are not oriented as shown in Figures 4.1.1 and 4.1.2. In general, since the Metamic inserts are not missing, but mis-oriented, the worst configuration is where there is a 2x2 array with four fuel assembly to fuel assembly gaps with no Metamic panel; however, this configuration still has the four inserts albeit in a different configuration. For simplicity and conservatism, the mis-orientation accident is covered by the more reactive configuration with four missing inserts adjacent to the missing insert in the central location of the design basis model (Case 4.2.3.6.1). In this configuration, the fuel assembly in the center of the model (which has no Metamic insert) is face adjacent to four other locations with no Metamic insert. Each of those four other locations which are missing Metamic inserts also therefore have a fuel assembly face adjacent to another Holtec Report HI-2177590 Page 4-24 Holtec Project 2635

fuel assembly with no Metamic insert between them. See Figure 4.2.7. The following cases are evaluated:

Case 4.2.5 .8.1: The model is the same as Case 4.2.3.6.1 except that the center location, and the four locations which are face adjacent to the center location have no Metamic insert. The SFP has 0 ppm soluble boron.

Case 4.2.5.8.2: The model is the same as Case 4.2.5.8.l except that the SFP has 1000 ppm soluble boron.

4.2.5.9 Calculations to Determine keffValues The calculation of the maximum keff for both normal and accident conditions for the BWR BORAFLEX' storage racks in the Harris Pool A and B includes the following conservative biases within the design basis model:

Bounding fuel assembly parameters

Bounding storage rack parameters

Bounding Metamic insert design parameters

Bounding radial positioning of the Metamic insert and fuel assembly Therefore, the calculated reactivity is conservatively biased. The maximum kerds determined using the following equation:

keff= kcalc +uncertainty+ bias where kcalc includes:

Maximum reactivity normal case kcaic or,

Maximum reactivity accident case.

where uncertainty includes:

Coupon Measurement uncertainty

MCNP5-l .5 l bias uncertainty (95% probability at a 95% confidence level)

MCNP5-1.5 l calculations statistics (95% probability at a 95% confidence level, 2cr) and the bias includes

MCNP5- l.51 bias

1% NRC Administrative Margin Note that each uncertainty is statistically combined with other uncertainties, while biases are added together in order to determine keff. The approach used here takes credit for soluble boron under normal conditions (see Section 4.3). Under this approach, the limiting condition is the non-borated Holtec Report HI-2177590 Page 4-25 Holtec Project 2635

condition, which needs to be shown to result in a maximum keffOfless than 1.0 at the 95/95 level.

Note that for the cases with credit for soluble boron, where the regulatory limit is 0.95, no specific target keffis defined, rather a soluble boron concentration is selected and the kerds shown to meet the regulatory limit.

4.2.5.10 Margin Evaluation The criticality analysis methodology conservatively includes biases in the design basis model for simplicity and margin. The following is a summary of the approach:

Bounding fuel assembly parameters treated as a bias rather than as an uncertainty.

Bounding storage rack parameters treated as a bias rather than as an uncertainty.

Bounding Metamic insert design parameters treated as a bias.

Bounding radial positioning of the Metamic insert and fuel assembly treated as a bias.

Planar average enrichment versus pin specific enrichments (Section 4.2.3.5.3) treated as a bias (see Case 4.2.3.1.2.10).

Additionally a 1% bias is applied to bias the results of the analysis for the NRC to use as administrative margin.

4.3 Acceptance Criteria Codes, standard, and regulations or pertinent sections thereof that are applicable to these analyses 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, Title 10, Part 50, Section 68, "Criticality Accident Requirements"

USNRC Standard Review Plan, NUREG-0800, Section 9.1.1, Criticality Safety of Fresh and Spent Fuel Storage and Handling, Rev. 3 -March 2007.

L. 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.

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

USNRC, NUREG/CR-6698, Guide for Validation of Nuclear Criticality Safety Calculational Methodology, January 2001.

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DSS-ISG-2010-01, Revision 0, Staff Guidance Regarding the Nuclear Criticality Safety Analysis for Spent Fuel Pools.

Consistent with the requirements in 10CFR50.68(b)(4), the objective of this analysis is to ensure that the effective neutron multiplication factor (keff) is less than or equal to 0.95 with the pool flooded with borated water, and that it is less than 1.0 under the assumed accident of the loss of soluble boron in the pool water, i.e. assuming unborated water in the spent fuel pool, all for 95%

probability at a 95% confidence level.

4.4 Assumptions The analyses apply a number of assumptions, either for conservatism or to simplify the calculation approach. Each assumption is appropriately discussed and justified in the text. Important aspects of applying those assumptions are as follows:

Bounding or sufficiently conservative inputs and assumptions are used essentially throughout the entire analyses, and in most cases, studies are presented to show that the selected inputs and parameters are in fact conservative or bounding.

An evaluation is performed to estimate the overall margins of the analyses. This evaluation includes considerations for potential non-conservatisms throughout the analyses, to ensure those are covered by the margin.

4.5 Input Data 4.5.1 Fuel Assembly Designs The BWR and PWR fuel assembly data used in the analysis is presented in Table 4.5.la and Table 4.5.lb. The GE7 pin map is provided in [4.14].

All BWR fuel is modeled as fresh fuel with no Gd, while the PWR fuel is both fresh (no reactivity control devices) and spent.

The GE13 is modeled with no axial blankets, no part length rods, and the axial length increased to 150 inches.

For the PWR fuel, only the 17xl 7 design is considered since this design bounds the 15x15 design

[4.11]. All 17xl 7 spent fuel isotopic compositions are taken directly from [4.1 O], with the exception of the lumped fission products [4.1 O] which were removed.

For all fuel designs, the fuel assembly is explicitly modeled in terms of fuel pin, cladding, water holes and channels (BWR fuel) and guide tubes (PWR fuel). The water holes, channels and guide tubes are all considered to be the same length as the fuel.

4.5.2 Storage Rack Designs The storage rack designs data used in the analysis is presented in Table 4.5.2.

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The MCNP5-1.51 storage rack models consider the axial height of the rack beginning at the top of the rack baseplate. The baseplate itself is not considered, and everything below the bottom of the fuel is replaced by water. The storage rack height however extends above the active length of the fuel to its full height. This is done because the various fuel designs considered in this analysis have various axial lengths and therefore for those cases where various rack designs exist in the same model the axial variation is considered. This approach is not expected to have any impact on the results of the analysis since only fresh fuel is considered with no axial variations and therefore the maximum reactivity will be in the center of the fuel assembly. Note that the BWR racks and PWR racks baseplate tops are at different heights and this difference is also considered.

4.5.2.1 BWR BORAFLEX' Spent Fuel Storage Racks The BWR BORAFLEX' spent fuel storage racks are located in both Pool A (three 1lxl1 racks) and Pool B (five 1lxl1 racks) and were designed by Westinghouse. These racks contain 968 storage spaces that are the focus of this analysis. The BORAFLEX' used in the design has known degradation issues and it is assumed to be completely gone for this analysis and is replaced by water in the MCNP5-1.51 models. The storage racks are composed of an 1lxl1 array of stainless steel boxes, joined at the comers in an egg-crate structure such that there are two cell types, one that is fabricated and one that is formed. The original BORAFLEX' material was fixed to the fabricated cell wall exterior with sheathing.

As discussed in Section 4.2, various BWR BORAFLEX' rack models are considered for the various evaluations.

4.5.2.2 PWR BORAFLEX' Storage Racks The PWR BORAFLEX' storage racks were designed by Westinghouse and are located in both Pool A (six 6x10 racks) and Pool B (five 6x10, six 7x10 and one 6x8 racks). These racks are qualified (with soluble boron credit) to contain both fresh (in a checkerboard of fresh and empty cells) and spent fuel (in a uniform loading pattern) [4.11]. As discussed in Section 4.5.2.l the BORAFLEX'used in the design has known degradation issues and it is assumed to be completely gone for this analysis and is replaced by water in the MCNP5-1.51 models. The various storage racks are low density flux trap style racks. This geometry is explicitly modeled.

4.5.2.3 BWR BORAL Storage Racks The BWR BORAL racks were designed by Holtec International and are located only in Pool B (twelve 1lxl1 racks). These storage racks credit the BORAL, do not take credit for soluble boron and are qualified to store the same GE fuel designs considered in this analysis for the BWR BORAFLEX'storage racks [4.12]. The analysis in [4.12] considered the GE7 as the design basis analysis and therefore the current analysis considers the BWR BORAL racks fully loaded with GE7 fuel.

The various storage racks are composed of arrays of stainless steel boxes, joined at the comers in an egg-crate structure such that there are two cell types, one that is fabricated and one that is Holtec Report HI-2177590 Page 4-28 Holtec Project 2635

formed. This geometry is explicitly modeled. Each storage rack has a Boral panel fixed along the exterior [4.12].

4.5.2.4Rack to Rack Gaps In Pool A there are both BWR BORAFLEX' and PWR BORAFLEX' storage racks and in Pool B there are BWR BORAFLEX', PWR BORAFLEX' and BWR BORAL storage racks. The gaps between these racks vary from pool to pool and within each pool for the various rack types.

Therefore, minimum rack spacing values were determined for use in the accident and interface calculations. These values are presented in Table 4.5.3 and are applicable to both pools.

4.5.3 Metamic Insert Design The Metamic insert design parameters are presented in Table 4.5.4. The Metamic insert is fabricated using a single extruded sheet of Metamic which is rolled to the desired thickness. The Metamic sheet is then trimmed to the desired dimensions and the slots are cut for bending using a waterjet.

The insert is heated prior to each bend, including bending of top of the insert for mounting of the head piece and a full length continuous bend. Following the final bend, the insert is placed in a flattening fixture, and allowed to cool. It may be necessary to perform multiple flattening iterations to meet the desired flatness requirements. The final step of fabrication is to machine and install the insert head piece. All Metamic insert design parameters are then verified using the appropriate QA processes.

4.5.4 Material Composition The material compositions for the various MCNP models are presented in Table 4.5.5.

4.6 Computer Codes See Section 4.2.2.

4.7 Analysis As discussed in Section 4.2, the analysis is performed using a combination of bounding analysis parameters and statistical uncertainties. The use of bounding analysis parameters allows for simplicity and inclusion of analysis margin in the results. The following analysis parameters are treated in a bounding manner:

Fuel assembly manufacturing parameters.

Holtec Report HI-2177590 Page 4-29 Holtec Project 2635

Storage rack manufacturing parameters.

Metamic insert manufacturing parameters.

SFP moderator temperature.

Radial positioning of fuel assembly, Metamic insert location, and presence of fuel assembly channel.

1lxl1 array used in design basis model includes a missing insert in center location.

Additional discussions are provided below.

4.7.l Bounding Fuel Design and Fuel Assembly Parameters As discussed in Section 4.2.3.1.1 and 4.2.3.1.2, the analysis performs evaluations to determine the most reactive fuel design and fuel design parameters for use as the design basis assembly.

4.7.1.1 Bounding Fuel Design As discussed in Section 4.1, the following BWR GE fuel designs are permitted for storage in the Harris Pool A and B BWR BORAFLEX' storage racks: GE3, GE4, GE5, GE6 and GE7.

Additionally, for accident conditions, all administratively non-permitted designs, i.e. the GE8, GE9, GEIO and GE13, are evaluated to determine the bounding fuel assembly for the accident analysis. For simplicity and analysis margin, the GE13 is selected as the bounding accident fuel design with a conservative IMPAE of 5.0 wt% U-235. Thus, further evaluations of the administratively non-permitted fuel designs are not required since the GE13, while already the fuel design with the highest IMPAE, is evaluated with the conservative IMPAE of 5.0 wt% U-235.

As discussed in Section 4.2.3.1.1, evaluations are performed and the results are presented in Appendix 4.A, Table 4.A. l. The GE fuel designs evaluated are fully characterized in Table 4.5.1.a and Table 4.5.1.b.

4.7.1.2 Bounding Fuel Design Parameters As discussed in Section 4.2.3 .1.2, evaluations are performed for the bounding fuel design, the GE7, and the accident case fuel design, the GE13. The results of the evaluations for the bounding fuel design parameters is presented in Appendix 4.A, Table 4.A.2. The results show that the bounding GE7 fuel design parameters are:

Maximum fuel rod density and pellet OD.

Minimum fuel rod clad thickness.

Maximum fuel rod pitch.

The water rod and channel thickness have minimal impact. Therefore, for calculations with those parameters nominal dimensions are used.

Holtec Report HI-2177590 Page 4-30 Holtec Project 2635

Therefore, for all design basis calculations the bounding set of fuel assembly parameters is used.

This is conservative and provides analysis margin because the reactivity effect of the fuel assembly parameters is treated as a bias rather than an analysis uncertainty.

Note that the results for the pin specific enrichment cases, Case 4.2.3.1.2.10, show that it is acceptable to use the IMPAE rather than pin specific enrichments.

4.7.2 Bounding BWR BORAFLEX' Storage Rack Parameters As discussed in Section 4.2.3.2, the BWR BORAFLEX' storage rack parameters were evaluated to determine the bounding set for use in the design basis model. The results of the evaluations are presented in Appendix 4.A, Table 4.A.3. The results presented in Appendix 4.A, Table 4.A.3 show that the bounding BWR BORAFLEX' storage rack parameters are:

Minimum storage cell pitch.

Note that the storage cell inner diameter and cell wall thickness had minimal impact. Therefore those parameters are treated as nominal.

Therefore, for all design basis calculations the bounding set ofBWR BORAFLEX' storage rack parameters is used. This is conservative and provides analysis margin because the reactivity effect of the BWR BORAFLEX' storage rack parameters is treated as a bias rather than an analysis uncertainty.

4.7.3 Bounding Metamic Insert Parameters As discussed in Section 4.2.3.3, the Metamic insert parameters were evaluated to determine the bounding set for use in the design basis model. The results of the evaluations are presented in Appendix 4.A, Table 4.A.4. The results presented in Appendix 4.A, Table 4.A.4 show that the bounding Metamic insert parameters are:

Missing insert in the center location.

Minimum Metamic loading, thickness and width.

Therefore, for all design basis calculations the bounding set ofMetamic insert parameters is used.

This is conservative and provides analysis margin because the reactivity effect of the Metamic insert parameters is treated as a bias rather than an analysis uncertainty.

Note that the results presented in Appendix 4.A, Table 4.A.4 also provide the statistical uncertainty calculations for the coupon measurement uncertainty (Case 4.2.3.3.6).

4.7.4 Reactivity Effect of SFP Water Temperature As discussed in Section 4.2.3.4, the reactivity effect of SFP water temperature and density is evaluated so that the bounding values can be used in the design basis mode. The results of the evaluations are presented in Appendix 4.A, Table 4.A.5. The results presented in Appendix 4.A, Holtec Report HI-2177590 Page 4-31 Holtec Project 2635

Table 4.A.5 show that the bounding temperature (and corresponding density) are the minimum temperature and maximum density. Therefore, these values are used in the design basis model.

4.7.5 Reactivity Effect of Fuel Assembly Channel and Radial Positioning of Fuel Assembly and Metamic Insert As discussed in Section 4.2.3.5, the reactivity effect of the fuel and Metamic insert radial location, in conjunction with the presence or not of the fuel channel is evaluated to determine the bounding configuration for use in the design basis model. The results of the evaluations are presented in Appendix 4.A, Table 4.A.6. The results presented in Appendix 4.A, Table 4.A.6 show that the bounding radial configurations are:

No fuel assembly channel present

Metamic insert adjacent to the storage cell comer.

All fuel eccentrically positioned towards the center of the storage rack.

Fuel assembly in the central storage rack location positioned in the center of the storage cell.

Therefore, all design basis calculations for normal conditions include the bounding configuration.

This is conservative and provides analysis margin because the reactivity effect of the radial position of the fuel and Metamic insert is treated as a bias rather than an analysis uncertainty.

4.7.6 Design Basis Model As discussed in Section 4.2.3 .6, various evaluations have been performed to determine the bounding set of parameters for the design basis modal. The results of these evaluations have been discussed in the previous sections. Based on the results of those evaluations, the design basis calculations for normal conditions have been performed. The results of the design basis calculations for normal conditions are presented in Appendix 4.A, Table 4.A.7. The design basis models consist of the following parameters:

Bounding fuel assembly parameters

Bounding storage rack parameters

Bounding Metamic insert design parameters

Bounding radial positioning of the Metamic insert and fuel assembly

Missing Metamic insert in most reactive location (center ofrackarray).

The results of the design basis model calculations are used to determine keff(see Section 4.7.9).

4.7.7 Storage Rack Interfaces As discussed in Section 4.2.3.7, the Harris Pool A and B storage rack interfaces are evaluated to determine if the design basis model is bounding by validating the interface assumption. The results of the interface calculations are presented in Appendix 4A, Table 4A.8. The results of the calculations show that the interface assumption is validated.

Holtec Report HI-2177590 Page 4-32 Holtec Project 2635

4.7.8 Accident Conditions As discussed in Section 4.2.5, the following accident conditions have been evaluated:

The effect of SFP temperature exceeding the normal range.

A dropped fuel assembly:

A misloaded fuel assembly (a fuel assembly in the wrong location within the storage rack)

A mislocated fuel assembly (a fuel assembly in the wrong location outside the storage rack).

Rack movement due to seismic activity.

Mis-orientation ofMetamic inserts.

Soluble boron dilution accident.

The results of the accident condition evaluations are presented in Appendix 4.A, Table 4.A.9.

The maximum reactivity case is used to determine kerrfor accident conditions (see Section 4.7.9).

The results ofthe additional mislocated fuel assembly accident cases, as discussed in Section 4.2.5.5, are presented in Appendix 4.B, Tables 4.B.1through4.B.3. A summary of the maximum reactivity cases from all of the mislocated fuel assembly cases is presented in Appendix 4.B, Table 4.B.4.

The results presented in Appendix 4.B show that the bounding configuration, i.e. the mislocated fuel assembly in the center of the SFP, was selected. The results also show that using a water reflector for those cases is acceptable.

4.7.9 Calculation of the Maximum keff As discussed in Section 4.2.5.9, the maximum keff for both normal and accident conditions is determined for both pure water and borated water. The results of the keffcalculations are presented in Table 4.7.1.

4.7.10 Margin Evaluation The criticality analysis methodology that is used in this report is described in detail in Section 4.2.

The methodology allows for the use of both nominal and bounding parameters for the design basis calculations, and also performs various studies that quantify potential conservatisms and non-conservatisms. The reactivity effect of various analysis methods is presented in Table 4.7.2. As it can be seen from Table 4.7.2, the primary conservatisms are the use of planar average enrichment, bounding Metamic insert parameters and the eccentric positioning of the fuel in the storage cell (with no fuel channel present). Smaller impacts are associated with the fuel design and storage rack manufacturing parameters. In summary, the analysis methodology contains significantly conservative approaches which yield substantial margin.

4.8 Conclusion The criticality calculations for the Harris SFP Pool A and B have been performed for the BWR BORAFLEX' storage racks without credit for the degraded BORAFLEX' and with credit for Holtec Report HI-2177590 Page 4-33 Holtec Project 2635

Metamic inserts. The objective of the analysis is to demonstrate that the effective neutron multiplication factor (kerr) is less than 1.0 with the pool flooded with un-borated water and that keff is less than or equal to 0.95 with the pool flooded with borated water. The maximum keff includes a margin for uncertainty in reactivity calculations including manufacturing tolerances and is shown to be less than the regulatory limit with a 95% probability at a 95% confidence level as presented in Table 4. 7 .1. Reactivity effects of abnormal and accident conditions have also been evaluated to assure that under all credible abnormal and accident conditions, the maximum keffwill not exceed the regulatory limit of 0.95 with credit for soluble boron.

Holtec Report HI-2177590 Page 4-34 Holtec Project 2635

4.9 References

[4.1] "MCNP - A General Monte Carlo N-Particle Transport Code, Version 5," Los Alamos National Laboratory, LA-UR-03-1987 (2003, Revised 2/1/2008).

[4.2] ENDF/B-VII.O Evaluated Nuclear Data Library, Release December 15, 2006.

[4.3] NJOY99.0, Code System for Producing Pointwise and Multigroup Neutron and Photon Cross Section from ENDF/B Data, PSR-480, March 2000, ORNL.

[4.4] HI-2156450, "MCNP5 Temperature Dependent Library Construction with NJOY",

latest revision, Holtec International.

[4.5] F. Brown, A Review of Monte Carlo Criticality Calculations - Convergence, Bias, Statistics, 2009 International Conference on Mathematics, Computational Methods and Reactor Physics, Saratoga Springs, NY, 2009.

[4.6] "Nuclear Group Computer Code Benchmark Calculations", Holtec Report HI-2104790 Rev. 3.

[4.7] Guide for Validation of Nuclear Criticality Safety Calculational Methodology, NUREG/CR- 6698, January 2001.

[4.8] Brunswick Nuclear Plant Calculation OB21-0203, "Supplemental Information to Support Criticality Analysis Removing Credit for Boraflex in BWR Spent Fuel Storage Racks at HNP," Latest Revision.

[4.9] DSG-ISG-2010-01, Staff Guidance Regarding the Nuclear Criticality Safety Analysis for Spent Fuel Pools, Revision 0.

[4.10] HI-2104763, current revision, "Criticality Analysis for Harris Spent Fuel Pool A/B BWR Boraflex Racks Without Credit for Boraflex".

[4.11] Framatome ANP, Inc. Document 77-5069740-NP-OO, Dated August 2005, "Shearon Harris Criticality Evaluation".

[4.12] HI-2115044, current revision, "Criticality Analysis for Harris Spent Fuel Pool B BWR Boral Racks".

[4.13] Deleted.

[4.14] Email from Patrick Washington (Duke Energy) to Patrick Washington, dated August 2, 2016, subject "HNP BWR Rack Inserts- GE 7x7 Fuel Pin Map".

[4.15] EPRI Report 3002003073, "Sensitivity Analysis for Spent Fuel Pool Criticality," 2014.

Holtec Report HI-2177590 Page 4-35 Holtec Project 2635

Table 4.2.1 Summary of the Area of Applicability of the MCNP5-1.51 Benchmark Design Parameter Benchmarks Validated Application Fissionable Material 23su, 239Pu, 241pu 23su, 239Pu, 241Pu 23su, 239Pu, 241pu Isotopic Composition 23su;u < 5.0wt% 1.57-10% < 10wt%

Pu/(U+Pu) n/a 1.104 - 20 % <20wt%

Physical Form U02 U02,MOX U02,MOX Fuel Density (g/cm3) 10.31 - 10.63 6.1 -10.4 6.1 - 10.7 1 Moderator Material (coolant) H H H Physical Form H20 H20 HiO Density (g/cm3) around 1.0 g/cm3 around 1.0 g/cm3 around 1.0 g/cm3 Reflector Material H H H Physical Form HiO H20 H20 Density (g/cm3) around 1.0 g/cm 3 around 1.0 g/cm3 around 1.0 g/cm3 Interstitial Reflector Material Plate Steel Steel or Lead Steel or Lead Absorber Material None, Boron (15 - None, Boron (0 -

None, Boron (0- 2550 ppm) or 2550 ppm) or Soluble 1000 ppm) Gadolinium (48 - Gadolinium (48 to 197 ppm) 197 ppm)

B4C, Pyrex, Rods n/a Vicor, Steel or B- Boron Al Separating Material Water, B-SS, Baral, Water, B-SS, Baral, Plate Water, Baral Boroflex, Zircaloy or Boroflex, Zircaloy or Cadmium Cadmium Geometry Lattice type Square Square, Triangle Square, Triangle 1.44- 1.87 Lattice Pitch (cm) (BWR) 0.7to 4.318 0.7 to 4.318 1.26 (PWR)

Neutron Energy Thermal spectrum Thermal spectrum Thermal spectrum 1

See Table 2.3 in [4.7].

Holtec Report HI-2177590 Page 4-36 Holtec Project 2635

Table 4.2.2 MCNP5-l.51 Benchmarking Bias and Bias Uncertainty and Trending Analysis

[4.6]

No. Normality Residuals Experiment Bias of Bias x2 Linear Correlation Normality, Description Uncertainty exp. (Pct(x2;d)) (Pct(x2;d))

All 0.0072 165.89 experiments 562 0.0001 (0.0111) (0.0%)

None -

All except those with 35.50 Gadolinium, 389 0.0004 0.0061 (19%)

None -

Cadmium and Lead2 All with 33.43 311 0.0008 0.0062 None 0.19%

Fresh Water (26.90%)

All with Density 4.54 33%

Borated 78 -0.0004 0.0054 k(x) = 1.0654 + (-6.428E-(21%) (Significant)

Water 03)*x Fresh U02 0.0054 18.24 EALF 44%

Fuel with 178 0.0011 (0.0042) (1.09%) k(x) = 1.0001 + (5.349E-03)*x (Significant)

Fresh Water Fresh U02 0.0046 9.73 Fuel with 53 -0.0007 (0.0078) (2.22%)

None -

Borated EALF 25%

HTC+

k(x) = 0.9981 + (1.024E-02)*x (Significant)

MOX Fuel 3.78 133 -0.0003 0.0083 with Fresh (80%)

Pu Enrichment 12%

Water k(x) = 0.9986 + (5.552E-04)*x (Significant)

Rod-OD 31%

HTC+

k(x) = 0.9904 + (1.125E-02)*x (Significant)

MOX Fuel 1.19 25 0.0016 0.0092 with Borated (75%) Density 21%

Water k(x) = 1.0754 + (-7.198E-(Significant) 03)*x 2

Note: Critical experiments with Gadolinium, Cadmium and Lead were excluded from all subsequent subsets.

Holtec Report HI-2177590 Page 4-37 Holtec Project 2635

Table 4.2.3 Summary ofMCNP5-1.51 Benchmarking Bias and Bias Uncertainty Significant Trending Analysis Analysis Analysis Analysis Experiment Linear Parameter [4.6]

Parameter Parameter Trend Description Correlation Trend Bias Trend Value3 Bias Uncertainty Table Density All with k(x) = 1.0654 + 10.6312 g/cm3 Borated -0.0029 0.0087 D.3-14

(-6.428E-03)*x (max)

Water (g/cm3l EALF Fresh U02 k(x) = 1.0001 + 0.22 eV Fuel with 0.0013 0.0053 (5.349E-03)*x (min)

Fresh Water (eV)

EALF 0.22 eV k(x) = 0.9981 + 0.0000 0.0081 HTC+ (min)

(1.024E-02)*x MOX Fuel Pu Enrichment with Fresh k(x) = 0.9986 + 0%

Water -0.0014 0.0081 (5.552E-04)*x (min) D.3-16 (wt%)

Rod-OD k(x) = 0.9904 + 0.95 cm 0.0011 0.0108 HTC+ (l.125E-02)*x (min)

MOX Fuel (cm) with Density Borated k(x) = 1.0754 + 10.6312 g/cm3 Water -0.0011 0.0124

(-7.198E-03)*x (max)

(g/cm 3)

Holtec Report HI-2177590 Page 4-38 Holtec Project 2635

Table 4.5.1.a Specification of the Fuel Assembly Parameters [4.8]

GE5 andGE6 Fuel Assembly Design GE3 GE4 GE7 GE 13 17xl7 Fuel Assembly Overall Length, 176.16 176.16 176.16 176.16 n/a n/a Inches Distance from Bottom of Fuel 7.585 7.385 7.385 7.385 n/a n/a Assembly to Beginning of Active Length, Inches 0.563 +/- 0.493 +/- 0.483 +/- 0.483 +/- 0.44+/-

Clad Outer Diameter, Inches 0.374 0.0025 0.0025 0.0025 0.0025 0.0025 0.489 +/- 0.425 +/- 0.419 +/- 0.419+/- 0.384+/-

Clad Inner Diameter, Inches 0.329 0.0025 0.0025 0.0025 0.0025 0.0025 Clad Material Zr-2 Zr-2 Zr-2 Zr-2 Zr-2 Zr-4 0.477 +/- 0.416 +/- 0.411 +/- 0.411 +/- 0.376 +/-

Pellet Diameter, Inches 0.3225 0.0005 0.0005 0.0005 0.0005 0.0005 10.31+/-

Stack Density glee 10.32 +/- 0.17 10.32 +/- 0.17 10.44 +/- 0.17 10.45 +/- 0.17 10.6312 0.17 Full Length Rods Active 144 146 150 150 146 144 Part Length Rods Active n/a n/a n/a n/a 108 n/a Number of Full Length Rods 49 63 62 62 66 264 Number of Part Length Rods n/a n/a n/a n/a 8 n/a Fuel Rod Array 7x7 8x8 8x8 8x8 9x9 17xl7 6 top, 6 6 top, 6 8 top, 6 Blanket Length, Inches n/a n/a n/a bottom bottom bottom Blanket Enrichment n/a n/a natural natural natural n/a 0.738 +/- 0.64+/- 0.64+/- 0.64+/- 0.566 +/-

Fuel Rod Pitch, Inches 0.496 0.0055 0.0055 0.0055 0.0055 0.0055 Number of Water Rods 0 l 2 2 2 25 0.493 +/- 0.591 +/- 0.591 +/- 0.98+/-

Water Rod Outer Diameter, n/a 0.474 0.0025 0.0025 0.0025 0.0025 Inches 0.425 +/- 0.531 +/- 0.531 +/- 0.92+/-

Water Rod Inner Diameter, n/a 0.45 0.0025 0.0025 0.0025 0.0025 Inches Water Rod Pitch, Inches n/a n/a n/a n/a 1.024 n/a 5.278 +/-

Channel Inner Diameter, Inches 5.278 +/- 0.01 5.278 +/- 0.01 5.278 +/- 0.01 5.278 +/- 0.01 n/a 0.01 0.08+/-

Channel Thickness, Inches 0.08 +/- 0.004 0.08 +/- 0.004 0.08 +/- 0.004 0.07 +/- 0.004 n/a 0.004 Channel Comer Radius, Inches 0.38 0.38 0.38 0.38 0.45 n/a Channel Height Above Active 20.175 18.175 14.175 n/a n/a n/a Fuel, Inches Holtec Report HI-2177590 Page 4-39 Holtec Project 2635

Table 4.5.1.b Additional Fuel Assembly Parameters

[4.10]

Maximum Fuel Minimum Fuel Assembly Type Assembly Type Fuel Assembly IMPAE4 in SFP Average Bumup Design (wt% U-235) (MWD/MTU)

GE3 2.301 4902.7 GE4 2.737 11970.8 GE5 and GE6 3.192 20729.2 GE7 3.192 24723.6 GE8 3.688 31280.1 GE9-1 3.768 13506.7 GE9-2 3.778 34211.5 GEIO 3.968 27430.7 GE13 4.178 34675.4 Note: there is only one GE9-l assembly in the Harris SFP.

4 Note: The IMPAE in this table does not include the enrichmenttolerance (except for GE13).

Holtec Report HI-2177590 Page 4-40 Holtec Project 2635

Table 4.5.2 Specification of the Pool A and B Fuel Storage Racks [4.8]

Parameter Value BWR PWR Rack Type BWRBORAL BORAFLEX' BORAFLEX' 3 Pool A, 5 Number of Racks 12 Pool B 6 Pool A, 12 Pool B Poo!B Storage Rack Material 304 SS 304 SS 304 SS Rack Height (top of baseplate to top of rack),

169 169 168.56 Inches Distance from Pool Liner to Rack Baseplate, 8.92 8.92 6 Inches Distance Baseplate Extension Protrudes from 0.4 0.25 0.25 Rack Envelope, Inches Distance from Rack Baseplate to Bottom of 5.28 5.28 4.5 Neutron Absorber, Inches Length of Neutron Absorber, Inches 150 150 144.25 Neutron Absorber Thickness, Inches 0.045 0.075 0.075 Neutron Absorber Width, Inches 5.1 5.0 7.46 Storage Cell Inner Diameter, Inches 6.05 +/- 0.02 6.06 8.75 Storage Cell Pitch, Inches 6.25 +/- 0.03 6.25 10.5 Storage Cell Box Wall Thickness, Inches 0.075 +/- 0.004 0.075 0.0747 Storage Cell Neutron Absorber Pocket 0.07 +/- 0.01 0.08 0.1 Thickness, Inches Storage Cell Neutron Absorber Sheathing 0.035 +/- 0.002 0.035 0.035 Thickness, Inches Storage Cell Neutron Absorber Sheathing Inside 5.15 +/- 0.0625 5.0 7.46 Width, Inches Degraded Degraded BORAFLEX' BORAFLEX' Neutron Absorber Type BORAL (modeled as water) (modeled as water)

Neutron Absorber areal density (g/sqcm) 5 n/a 0.015 min n/a Neutron Absorber Density, g/cc n/a 2.665 n/a 5

Note: For the Bora! racks, 90% of the value in the table is used per [4.12].

Holtec Report HI-2177590 Page 4-41 Holtec Project 2635

Table 4.5.3 Rack to Rack Gaps [4.8]

Description Value PWR to PWR rack (inches) 1.625 PWR to PWR rack N-S Pool A (inches) 8.5 PWR to B WR rack (inches) 1.5 BWR to BWR rack (inches) 1.875 Holtec Report HI-2177590 Page 4-42 Holtec Project 2635

Table 4.5.4 Metamic Insert Specification Description Value Thickness (inches)

Width6 (inches)

Length (inches)

Areal density (g/cm2)

\

Holtec Report HI-2177590 Page 4-43 Holtec Project 2635

Table 4.5.5 (1 of 5)

Material Compositions MCNPZAID7 Element Weight Fraction

[4.l],[4.4]

Steel (density 7.92 glee) [4.10]

24050.21c 0.00790496 24052.21c 0.15852660 Cr 24053.21c 0.01832179 24054.21c 0.00464667 Mn 25055.21c 0.02001000 26054.21c 0.03898260 26056.21c 0.63458000 Fe 26057.2lc 0.01491736 26058.21c 0.00202002 28058.21c 0.06719768 28060.2lc 0.02677598 Ni 28061.21c 0.00118336 28062.21c 0.00383482 28064.21c 0.00100815 Zr (density= 6.55 glee) [4.10]

40090.21c 0.50706120 40091.21c 0.11180900 Zr 40092.21c 0.17278100 40094.21c 0.17891100 40096.21c 0.02943790 Holtec Report HI-2177590 Page 4-44 Holtec Project 2635

Table 4.5.5 (2 of 5)

Material Compositions MCNPZAID 8 Element Weight Fraction

[4.1],[4.4]

B 5010.21c 5011.21c ---

Metamic (density = 2.6 glee) c Al B

6000.21c 13027.21c BORAL (density= 2.67 glee) [4.10]

5010.21c 5011.21c 0.04322397 0.19129520 c 6000.21c 0.06513745 Al 13027.21c 0.70034320 Pure water (density = 1.0 glee) 1001.21c 0.11188600 H

1002.21c 0.00002572 8016.21c 0.88579510 0

8017.21c 0.00229319 500 ppm Borated Water (density= 1.0 glee) 1001.21c 0.11183010 H

1002.21c 0.00002570 8016.21c 0.88535220 0

8017.21c 0.00229204 5010.21c 0.00009215 B

5011.21c 0.00040784 1000 ppm Borated Water (density= 1.0 glee) 1001.21c 0.11177410 H

1002.21c 0.00002569 8016.21c 0.88490930 0

8017.21c 0.00229090 5010.21c 0.00018430 B

5011.21c 0.00081569 78 The MCNP ZAID used is customized for ENDF/B-VII NJOY adjusted cross Holtec Report HI-2177590 Page 4-45 Holtec Project 2635

Table 4.5.5 (3 of 5)

Material Compositions MCNPZAID9 Element Weight Fraction

[4.1],[4.4]

GE7 Fresh 3.282 wt% U-235 Fuel 10 92235.21c 0.02893000 u

92238.21c 0.85257000 0 8016.21c 0.11850000 BORAL (density= 2.67 glee) [4.10]

5010.21c 0.04322397 B

5011.21c 0.19129520 c 6000.21c 0.06513745 Al 13027.21c 0.70034320 Pure water (density= 1.0 glee) 1001.21c 0.11188600 H

1002.21c 0.00002572 8016.21c 0.88579510 0

8017.21c 0.00229319 500 ppm Borated Water (density= 1.0 glee) 1001.21c 0.11183010 H

1002.21c 0.00002570 8016.21c 0.88535220 0

8017.21c 0.00229204 5010.2lc 0.00009215 B

5011.21c 0.00040784 1000 ppm Borated Water (density= 1.0 glee) 1001.21c 0.11177410 H

1002.21c 0.00002569 8016.21c 0.88490930 0

8017.21c 0.00229090 5010.21c 0.00018430 B

501 l.21c 0.00081569 9

The MCNP ZAID used is customized for ENDF/B-VII NJOY adjusted cross sections.

10 Other fresh fuel compositions are used, the design basis case is provided as an example.

Holtec Report HI-2177590 Page 4-46 Holtec Project 2635

Table 4.5.5 (4of5)

Material Compositions MCNPZAID 11 PWR Spent Fuel Isotopic Composition [4.1 O] (weight fractions vary depending on axial node and bumup considered) 92234.21c 45105.2lc 92235.21c 47109.21c 92236.21c 53135.21c 92238.21c 5413 l.21c 92239.21c 55133.21c 93237.21c 55134.21c 94239.21c 55135.21c 94238.21c 55137.21c 94239.21c 60143.21c 94240.21c 60145.21c 94241.21c 61147.21c 94242.21c 61148.21c 95241.21c 61149.21c 95242.21c 62147.21c 95243.21c 62149.21c 96242.21c 62150.21c 96243.21c 62151.21c 96244.21c 62152.21c 96245.21c 63153.21c 96246.21c 63154.21c 36083.21c 63155.21c 45103.21c 64155.21c 11 The MCNP ZAID used is customized for ENDF/B-VII NJOY adjusted cross sections.

Holtec Project 2635 Page 4-47 Holtec Report HI-2177590

Table 4.5.5 (5 of 5)

Material Compositions MCNPZAID 12 Weight Fraction Element

[4.l],[4.4] r4.I 51 Bounding Concrete Composition [4.15]

8016.21c 0.424102100 0

8017.21c 0.001097938 c 6000.21c 0.140000000 Na 11023.21c 0.023200000 12024.21c 0.058696310 Mg 12025.21c 0.007740914 12026.21c 0.008862759 Al 13027.21c 0.027200000 14028.21c 0.247507000 Si 14029.21c 0.013016850 14030.21c 0.008876131 20040.21c 0.034024950 20042.21c 0.000238430 20043.21c 0.000050900 Ca 20044.21c 0.000805311 20046.21c 0.000001610 20048.21c 0.000078800 26054.21c 0.000254050 26056.21c 0.004135569 Fe 26057.21c 0.000097200 26058.2lc 0.000013200 12 The MCNP ZAID used is customized for ENDF/B-VII NJOY adjusted cross sections.

Holtec Project 2635 Page 4-48 Holtec Report HI-2177590

Table 4.7.l Summary of the Analysis Results Analysis Criticality Results Summary For Normal Conditions Normal conditions maximum kcalc, 0 ppm (Case 4.2.3.6.1) 0.9455 Normal conditions maximum kcalc, 500 ppm (Case 4.2.3.6.6) 0.8714 Analysis Criticality Results Summary For Accident Conditions Accident conditions maximum kcalc, 0 ppm (Case 4.2.5.5.29)- With 1.0303 Administrative Requirement 3 Accident conditions maximum kcalc, 1000 ppm (Case 4.2.5.5.22)-

0.8880 With Administrative Requirement 3 Analysis Uncertainties Summary MCNP5-1.51 bias uncertainty (95/95) (Table 4.2.3) 0.0124 MCNP5-1.51 calculation statistics (2 sigma) 0.0006 Metamic Coupon Measurement uncertainty (Case 4.2.3.3.6) 0.0149 Analysis Bias Summary NRC Administrative Margin 0.0100 MCNP5-l.51 bias (Table 4.2.3) 0.0011 Determination of kerrfor Normal Conditions Normal Conditions maximum keff, 0 ppm 0.9660 Normal Conditions margin to regulatory limit for 0 ppm 0.0240 Normal Conditions maximum keff, 500 ppm 0.8919 Normal Conditions margin to regulatory limit for 500 ppm 0.0481 Determination of kerrfor Accident Conditions 13 Accident Conditions maximum keff, 0 ppm 1.0495 Accident Conditions maximum keff, 1000 ppm 0.9085 Accident Conditions margin to regulatory limit for 1000 ppm 0.0315 13 The maximum reactivity case is selected from the most reactive of the 1000 ppm calculations from among the Administrative Requirement 3 cases. The calculations for NO application of Administrative Requirement 3 cases are presented in Appendix 4.A for information only.

Holtec Project 2635 Page 4-49 Holtec Report HI-2177590

Table 4.7.2 Reactivity Effect of Various Analysis Methods Bounding Nominal Parameter Delta Case kcalc Case kcalc 2.3.6.2 0.9439 0.0016 Fuel Assembly Parameters 2.3.6.1 0.9455 2.3.6.3 0.9445 0.0010 Storage Rack Parameters 2.3.6.1 0.9455 2.3.6.4 0.9457 -0.0002 Planar Average enrichment 2.3.1.2. l 0.9044 2.3.1.2.10 0.8870 0.0174 versus pin specific enrichments Metamic Insert Parameters 2.3.6.1 0.9455 2.3.6.5 0.9259 0.0196 Eccentric Positioning and Fuel 2.3.5.9 0.9395 2.3.5.l 0.9015 0.0380 Channel Holtec Project 2635 Page 4-50 Holtec Report HI-2177590

Figure 4.1.1 Pool A Metamic Rack Insert Orientation PWR BORAFLEX' Racks BWR BWR BWR BORAFLEX' BORAFLEX' BORAFLEX' PWR BORAFLEX' Racks Rack 1 Rack2 Rack3 Plant North 1

Insert Orientation I Holtec Project 2635 Page 4-51 Holtec Report HI-2177590

Figure 4.1.2 Pool B Metamic Rack Insert Orientation PWR BORAFLEX' Racks BWR BWR BWR BWR BORAL BORAFLEX' BORAFLEX' BORAFLEX' Racks Rack4 Rack5 Rack6 BWR BORAL Racks BWR BWR BWRBORAL BORAFLEX' BORAFLEX' Rack Rack7 Rack8 Plant North Insert Orientation J Holtec Project 2635 Page 4-52 Holtec Report HI-2177590

Figure 4.2.1 MCNP5-l .51 BWR BORAFLEX' Design Basis Model This is a partial 2D representation of the design basis model showing the centerlocation with a missing insert. The GE7 fuel has no channel (the outline of the channel location is visible) and is eccentrically positioned towards the center location. All Metamic inserts are adjacent to the upper left corner of the storage cell.

Holtec Project 2635 Page 4-53 Holtec Report HI-2177590

Figure 4.2.2 MCNP5-1.51 BWR BORAL Interface Assumption Model GE7 with no 1~~~~¥1':;;~111-+ofl!!---. channel This is a partial 2D representation of the BWR BORAL model. The GE7 fuel is cell centered and has no channel (the outline of the channel location is visible).

Holtec Project 2635 Page 4-54 Holtec Report HI-2177590

Figure 4.2.3 MCNP5-1.51 PWR BORAFLEX' Interface Assumption Model with Spent PWR 17xl 7 Fuel

-- PWR 17x17 fuel This is a partial 2D representation of the PWR model. The PWR spent fuel is cell centered, and there is no BORAFLEX' present.

Holtec Project 2635 Page 4-55 Holtec Report HI-2177590

Figure 4.2.4 MCNP5-l.51 PWR BORAFLEX' Interface Assumption Model with Fresh PWR 17xl 7 Fuel n

This is a partial 2D representation of the PWR model. The PWR fresh fuel is cell centered, and there is no BORAFLEX' present.

Holtec Project 2635 Page 4-56 Holtec Report Hl-2177590

Figure 4.2.5 MCNP5-1.5 l Pool A Interface Calculation Model PWR fuel GE7 fuel This is a partial 2D representation of the full pool model showing the center of the pool where the two rows of PWR racks meet with the row ofBWR racks. The PWR fuel is eccentrically positioned towards the BWR fuel and the BWR fuel is in the same configuration as the design basis model (see Case 4.2.3.6.1).

\

Holtec Project 2635 Page 4-57 Holtec Report HI-2177590

Figure 4.2.6 Mislocated Accident Location Mislocated PWR fuel Administrative Restriction 3 location.

This is a partial 2D representation of the full pool model showing the center of the pool wherethe two rows of PWR racks meet with the row of BWR racks. The spent PWR fuel in the racks is eccentrically positioned towards the BWR fuel and the BWR fuel is in the same configuration as the design basis model (see Case 4.2.3.6.1). The mislocated fresh PWR fuel assembly is positioned in a location that is slightly smaller than its actual size, and as close to the BWR rack as possible.

There is no neutron absorber between the mislocated PWR fuel and any other fuel in the three adjacent racks. This is Case 4.2.5.5.1. Note that for Administrative Restriction 3 the BWR fuel assembly directly adjacent to the mislocated PWR fuel assembly is removed (the storage location is assumed to be blocked).

Holtec Project 2635 Page 4-58 Holtec Report HI-2177590

Figure 4.2.7 Mis-oriented Metamic Insert Accident Configuration Missing Metamic insert locations This is a partial 2D representation of the 1lxl1 BWR BORAFLEX' rack model for the mis-orientation of Metamic inserts accident case showing the center of the rack. The B WR fuel is in the same configuration as the design basis model (see Case 4.2.3.6.1. The location of the missing four additional missing Metamic inserts is shown, along with the design basis center location missing Metamic insert.

Holtec Project 2635 Page 4-59 Holtec Report HI-2177590

Appendix 4.A Analysis Results Report No. HI-2177590 Page 4.A-1 Holtec Project 2635

Table 4.A.1 MCNP5-l.51 Results for Fuel Design Reactivity Case kcalc sigma Description GE3 with maximum IMPAE +

4.2.3.1.1.1 0.8203 0.0003 enrichment tolerance.

GE4 with maximum IMPAE +

4.2.3.1.1.2 0.8595 0.0003 enrichment tolerance.

GE6 with maximum IMP AE +

4.2.3.1.1.3 0.8991 0.0003 enrichment tolerance.

GE7 with maximum IMPAE +

4.2.3.1.1.4 0.9015 0.0003 enrichment tolerance.

GE13 with maximum IMPAE +

4.2.3.1.1.5 0.9546 0.0003 enrichment tolerance.

Report No. HI-2177590 Page 4.A-2 Holtec Project 2635

Table 4.A.2 MCNP5-1.51 Results for Bounding Fuel Design Parameters Delta kcalc Case kcalc sigma Description (95/95)

GE7 with maximum IMPAE +enrichment tolerance, 4.2.1.3 .1.4 0.9015 0.0003 ref nominal fuel dimensions.

All tolerance parameters nominal (except fuel pellet 4.2.3 .1.2.1 0.9044 0.0003 0.0035 diameter and density which are maximum).

4.2.3 .1.2.2 0.9053 0.0003 0.0015 Minimum cladding thickness.

4.2.3.1.2.3 0.9017 0.0003 -0.0021 Maximum cladding thickness.

4.2.3 .1.2.4 0.9030 0.0003 -0.0008 Minimum fuel rod pitch.

4.2.3 .1.2.5 0.9057 0.0003 0.0019 Maximum fuel rod pitch.

4.2.3.1.2.6 0.9040 0.0003 0.0003 Minimum fuel channel thickness.

4.2.3 .1.2. 7 0.9040 0.0003 0.0002 Maximum fuel channel thickness.

4.2.3.1.2.8 0.9035 0.0003 -0.0003 Minimum water rod thickness.

4.2.3.1.2.9 0.9033 0.0003 -0.0004 Maximum water rod thickness.

Pin specific enrichment for the IMP AE considered in 4.2.3.1.2.10 0.8870 0.0003 -0.0144 Case 4.2.3.1.1.4.

Holtec Report HI-2177590 Page 4.A-3 Holtec Project 2635

Table 4.A.3 MCNP5-l .51 Results for Bounding BWR BORAFLEX Storage Rack Parameters Delta kcalc Case kc ale sigma (95/95) Description 4.2.3.2.1 0.9015 0.0003 ref Reference case. All storage rack parameters nominal.

4.2.3.2.2 0.9005 0.0003 -0.0004 Minimum storage cell inner diameter.

4.2.3.2.3 0.9023 0.0003 0.0015 Maximum storage cell inner diameter.

4.2.3.2.4 0.9017 0.0003 0.0008 Minimum storage cell wall thickness.

4.2.3.2.5 0.9014 0.0003 0.0006 Maximum storage cell wall thickness.

4.2.3.2.6 0.9058 0.0003 0.0049 Minimum storage cell pitch.

Holtec Report HI-2177590 Page 4.A-4 Holtec Project 2635

Table 4.A.4 MCNP5-1.51 Results for Bounding Metamic Insert Parameters Delta kcalc Case kc ale sigma Description (95195)

Reference case. Minimum B4C wt%, nominal insert thickness and 4.2.3.3.1 0.9100 0.0003 ref width (nominal loading at minimum B4C wt%).

Alternative missing insert location along rack edge, in center location.

4.2.3.3.2 0.9017 0.0003 -0.0076 Minimum B4C wt%, nominal insert thickness and width (nominal loading at minimum B4C wt%).

Alternative missing insert location along rack edge, in corner location.

4.2.3.3.3 0.8984 0.0003 -0.0110 Minimum B4C wt%, nominal insert thickness and width (nominal loading at minimum B4C wt%).

Minimum B4C wt%, maximum thickness and width (maximum 4.2.3.3.4 0.9025 0.0003 -0.0069 loading at minimum B4C wt%).

Minimum B4C wt%, minimum thickness and width (minimum 4.2.3.3.5 0.9206 0.0003 0.0113 loading at minimum B4C wt%).

Coupon testing measurement uncertainty. Minimum B4C wt% - 5%,

4.2.3.3.6 0.9243 0.0003 0.0149 minimum thickness and width (minimum loading at minimum B4C wt% -5%).

Holtec Report HI-2177590 Page 4.A-5 Holtec Project 2635

Table 4.A.5 MCNP5-l.51 Results for Bounding SPP Water Temperature Delta kcalc Case kcalc sigma Description (95/95) 4.2.3.4.1 0.9015 0.0003 ref Reference case. Same model as discussed in Section 4.2.3 .1.1.

4.2.3.4.2 0.9007 0.0003 -0.0002 Minimum nominal temperature 85 °P.

4.2.3.4.3 0.8919 0.0003 -0.0089 Maximum nominal temperature 105 °P.

4.2.3.4.4 0.8856 0.0003 -0.0152 Maximum possible temperature 212 °P.

Holtec Report HI-2177590 Page 4.A-6 Holtec Project 2635

Table 4.A.6 MCNP5-1.51 Results for Fuel Assembly Channel and Radial Positioning of Fuel Assembly and Metamic Insert Delta kcalc Description Case kc ale sigma (95/95)

Fuel Assembly Cell Centered, Fuel Channel Present 4.2.3.5.I 0.9015 0.0003 ref Reference case. Metamic inserts are between fuel assembly and cell comer.

4.2.3.5.2 0.9219 0.0003 0.0210 Metamic inserts are adjacent to the cell comer.

4.2.3.5.3 0.8934 0.0003 -0.0074 Metamic inserts are adjacent to the fuel assembly.

Fuel Assembly Cell Centered, Fuel Channel Missing 4.2.3.5.4 0.8980 0.0003 -0.0029 Metamic inserts are centered between fuel assembly and cell comer.

4.2.3.5.5 0.9195 0.0003 0.0222 Metamic inserts are adjacent to the cell comer.

4.2.3.5.6 0.8589 0.0003 -0.0420 Metamic inserts are adjacent to the fuel assembly.

Fuel Assembly Positioned Towards Rack Center, Fuel Channel Present Metamic inserts are adjacent to the cell comer. All fuel assemblies are positioned towards 4.2.3.5.7 0.9368 0.0003 0.0359 the rack center while center location is centered in rack cell.

Metamic inserts are adjacent to the cell comer. All fuel assemblies are positioned towards 4.2.3.5.8 0.9348 0.0003 0.0340 the rack center while center location is eccentric in rack cell.

Fuel Assembly Positioned Towards Rack Center, Fuel Channel Missing Metamic inserts are adjacent to the cell comer. All fuel assemblies are positioned towards 4.2.3.5.9 0.9395 0.0003 0.0386 the rack center while center location is centered in rack cell.

Metamic inserts are adjacent to the cell comer. All fuel assemblies are positioned towards the 4.2.3.5.10 0.9329 0.0003 0.0321 rack center while center location is centered in rack cell.

Fuel Assembly Positioned A way From Rack Center, Fuel Channel Present Metamic inserts are adjacent to the cell comer. All fuel assemblies are positioned away from 4.2.3.5.11 0.9001 0.0003 -0.0008 the rack center while center location is centered in rack cell.

Metamic inserts are adjacent to the cell comer. All fuel assemblies are positioned away from 4.2.3.5.12 0.8996 0.0003 -0.0013 the rack center while center location is eccentric in rack cell.

Fuel Assembly Positioned Away From Rack Center, Fuel Channel Missing Metamic inserts are adjacent to the cell comer. All fuel assemblies are positioned away from 4.2.3.5.13 0.8740 0.0003 -0.0268 the rack center while center location is centered in rack cell.

Metamic inserts are adjacent to the cell comer. All fuel assemblies are positioned away from 4.2.3.5.14 0.8735 0.0003 -0.0274 the rack center while center location is eccentric in rack cell.

Holtec Report HI-2177590 Page 4.A-7 Holtec Project 2635

Table 4.A.7 MCNP5-1.51 Results for the Design Basis Model Delta kcalc Case kcalc sigma Description (95/95)

Design basis model maximum kcalc with GE7 fuel with design 4.2.3.6.l 0.9455 0.0003 ref IMPAE 3.282 wt% U-235.

4.2.3.6.2 0.9439 0.0003 -0.0010 Design basis model except with nominal fuel rod cladding thickness.

4.2.3.6.3 0.9445 0.0003 -0.0004 Design basis model except with nominal fuel rod pitch.

4.2.3.6.4 0.9457 0.0003 0.0008 Design basis model except with nominal storage rack cell pitch.

Design basis model except with maximum Metamic thickness and 4.2.3.6.5 0.9259 0.0003 -0.0190 width.

4.2.3.6.6 0.8714 0.0003 -0.0736 Design basis model with 500 ppm soluble boron.

Holtec Report HI-2177590 Page 4.A-8 Holtec Project 2635

Table 4.A.8 MCNP5-1.51 Results for the Interface Calculations Case kcaic sigma Description 4.2.3.7.1 0.9455 0.0003 BWR BORAFLEX' storage rack infinite array. Same as design basis model (Case 4.2.3.7.2 0.9128 0.0003 BWR BORAL storage rack infinite array.

PWR BORAFLEX' storage rack infinite array with a uniform loading of spent PWR fuel at 4.2.3.7.3 0.9455 0.0003 about 45 GWD/MTU.

PWR BORAFLEX' storage rack infinite array with a checkerboard of empty and fresh 4.2.3.7.4 0.9289 0.0003 PWR with an enrichment of 5.0 wt% U-235.

Pool A, bounding BWR BORAFLEX configuration, spent PWR fuel eccentric towards the 4.2.3.7.5 0.9455 0.0003 BWR racks.

4.2.3.7.6 0.9456 0.0003 Pool A, bounding BWR BORAFLEX configuration, spent PWR fuel cell centered.

Pool A, bounding BWR BORAFLEX configuration, fresh PWR fuel eccentric towards the 4.2.3.7.7 0.9459 0.0003 BWRracks.

4.2.3.7.8 0.9463 0.0003 Pool A, bounding BWR BORAFLEX configuration, fresh PWR fuel cell centered.

Pool A, BWR fuel eccentric toward PWR fuel, spent PWR fuel eccentric towards the BWR 4.2.3.7.9 0.9464 0.0003 racks.

4.2.3.7.10 0.9366 0.0003 Pool A, BWR fuel eccentric toward PWR fuel, spent PWR fuel cell centered.

Pool A, BWR fuel eccentric toward PWR fuel, fresh PWR fuel eccentric towards the BWR 4.2.3.7.11 0.8972 0.0003 racks.

4.2.3.7.12 0.8965 0.0003 Pool A, BWR fuel eccentric toward PWR fuel, fresh PWR fuel cell centered.

4.2.3.7.13 0.9460 0.0003 Pool A, bounding BWR BORAFLEX configuration, no PWR fuel.

4.2.3.7.14 0.8974 0.0003 Pool A, BWR fuel eccentric toward PWR racks, no PWR fuel.

Ho ltec Report HI-2177 590 Page 4.A-9 Holtec Project 2635

Table 4.A.9 MCNP5-1.51 Results for the Accident Calculations Case kcalc sigma Description 4.2.5.4.l 1.0335 0.0003 Misloaded fresh 5.0 wt% U-235 GE13 fuel assembly, 0 ppm soluble boron.

4.2.5.4.2 0.9011 0.0003 Misloaded fresh 5.0 wt% U-235 GE13 fuel assembly, 1000 nnm soluble boron Mislocated fresh 5.0 wt% U-235 l 7xl 7 PWR fuel assembly, PWR racks with 4.2.5.5.l 1.0499 0.0003 spent fuel, 0 ppm soluble boron.

4.2.5.5.2 0.9062 0.0003 Same as Case 4.2.5.5.l except the SFP moderator has 1000 ppm soluble boron.

Same as Case 4.2.5.5.l except the PWR fuel is a checkerboard of fresh fuel and 4.2.5.5.3 0.9925 0.0003 empty cells.

4.2.5.5.4 0.8552 0.0003 Same as Case 4.2.5.5.3 except the SFP moderator has 1000 ppm soluble boron.

Same as Case 4.2.5.5. l except the BWR fuel is eccentrically positioned toward 4.2.5.5.5 1.0652 0.0003 the mislocated PWR fuel.

4.2.5.5.6 0.9266 0.0003 Same as Case 4.2.5.5.5 except the SFP moderator has 1000 ppm soluble boron.

Same as Case 4.2.5.5.5 except the PWR fuel is a checkerboard of fresh fuel and 4.2.5.5.7 1.0017 0.0003 empty cells.

4.2.5.5.8 0.8692 0.0003 Same as Case 4.2.5.5.7 except the SFP moderator has 1000 ppm soluble boron.

Same as Case 4.2.5.5.l except the mislocated PWR fuel face adjacent to the 4.2.5.5.9 1.0363 0.0003 PWR fuel in the PWR BORAFLEX TM racks.

4.2.5.5.10 0.8889 0.0003 Same as Case 4.2.5.5.9 except the SFP moderator has 1000 ppm soluble boron.

Same as Case 4.2.5.5.3 except the mislocated PWR fuel face adjacent to the 4.2.5.5.11 0.9797 0.0003 PWR fuel in the PWR BORAFLEX TM racks.

4.2.5.5.12 0.8390 0.0003 Same as Case 4.2.5.5.11 except the SFP moderator has 1000 ppm soluble boron.

Same as Case 4.2.5.5.5 except the mislocated PWR fuel face adjacent to the 4.2.5.5.13 1.0636 0.0003 PWR fuel in the PWR BORAFLEX TM racks.

4.2.5.5.14 0.9075 0.0003 Same as Case 4.2.5 .5 .13 except the SFP moderator has 1000 ppm soluble boron.

Same as Case 4.2.5.5.7 except the mislocated PWR fuel face adjacent to the 4.2.5.5.15 1.0016 0.0003 PWR fuel in the PWR BORAFLEX TM racks.

4.2.5.5.16 0.8542 0.0003 Same as Case 4.2.5.5.15 except the SFP moderator has 1000 ppm soluble boron.

4.2.5.6.1 0.9798 0.0003 Rack movement accident, PWR racks with spent fuel, 0 ppm soluble boron.

4.2.5.6.2 0.8378 0.0003 Same as Case 4.2.5.6.1 except the SFP moderator has 1000 ppm soluble boron.

Same as Case 4.2.5.6.l except the PWR fuel is a checkerboard of fresh fuel and 4.2.5.6.3 0.9460 0.0003 empty cells.

4.2.5.6.4 0.8102 0.0003 Same as Case 4.2.5 .6.3 except the SFP moderator has 1000 ppm soluble boron.

Same as Case 4.2.5.6.1 except the BWR fuel is eccentrically positioned toward 4.2.5.6.5 0.9802 0.0003 the central BWR rack location where two PWR racks are adjacent to the BWR 4.2.5.6.6 0.8376 0.0003 Same as Case 4.2.5.6.5 except the SFP moderator has 1000 ppm soluble boron.

Same as Case 4.2.5.6.5 except the PWR fuel is a checkerboard of fresh fuel and 4.2.5.6.7 0.8944 0.0003 empty cells.

4.2.5.6.7 0.8944 0.0003 Same as Case 4.2.5.6.7 except the SFP moderator has 1000 ppm soluble boron.

4.2.5.8.l 1.0590 0.0003 Mis-orientation of Metamic inserts.

4.2.5.8.2 0.8888 0.0003 Same as Case 4.2.5.8.l except the SFP moderator has 1000 ppm soluble boron.

Holtec Project 2635 Page 4.A-10 Holtec Report HI-2177590

Table 4.A. l 0 MCNP5-1.51 Results for the Mislocated Fuel Assembly Accident Calculations with Administrative Requirement 3 Case kcalc sigma Description 4.2.5.5.17 1.0185 0.0003 Same as Case 4.2.5.5.l except administrative requirement 3 is applied.

Same as Case 4.2.5.5.17 except the SFP moderator has 1000 4.2.5.5.18 0.8783 0.0003 ppm soluble boron.

Same as Case 4.2.5.5.17 except the PWR fuel is a checkerboard 4.2.5.5.19 0.9597 0.0003 of fresh fuel and empty cells.

Same as Case 4.2.5.5.19 except the SFP moderator has 1000 4.2.5.5.20 0.8292 0.0003 nnm soluble boron.

Same as Case 4.2.5.5.17 except the BWR fuel is 4.2.5.5.21 1.0290 0.0003 eccentrically positioned toward the mislocated PWR fuel.

Same as Case 4.2.5.5.21 except the SFP moderator has 1000 4.2.5.5.22 0.8880 0.0003 ppm soluble boron.

Same as Case 4.2.5.5.21 except the PWR fuel is a checkerboard 4.2.5.5.23 0.9670 0.0005 of fresh fuel and emPtv cells.

Same as Case 4.2.5.5.23 except the SFP moderator has 1000 4.2.5.5.24 0.8330 0.0003 ppm soluble boron.

Same as Case 4.2.5.5.17 except the mislocated PWR fuel face 4.2.5.5.25 1.0231 0.0003 adjacent to the PWR fuel in the PWR BORAFLEX TM racks.

Same as Case 4.2.5.5.25 except the SFP moderator has 1000 4.2.5.5.26 0.8812 0.0003 nnm soluble boron.

Same as Case 4.2.5.5.19 except the mislocated PWR fuel face 4.2.5.5.27 0.9670 0.0003 adjacent to the PWR fuel in the PWR BORAFLEX TM racks.

Same as Case 4.2.5.5.27 except the SFP moderator has 1000 4.2.5.5.28 0.8332 0.0003 ppm soluble boron.

Same as Case 4.2.5.5.21 except the mislocated PWR fuel face 4.2.5.5.29 1.0303 0.0003 adjacent to the PWR fuel in the PWR BORAFLEX TM racks.

Same as Case 4.2.5.5.29 except the SFP moderator has 1000 4.2.5.5.30 0.8853 0.0003 nnm soluble boron.

Same as Case 4.2.5.5.23 except the mislocated PWR fuel face 4.2.5.5.31 0.9704 0.0003 adjacent to the PWR fuel in the PWR BORAFLEX TM racks.

Same as Case 4.2.5.5.31 except the SFP moderator has 1000 4.2.5.5.32 0.8348 0.0003 ppm soluble boron.

Holtec Report HI-2177590 Page 4.A-11 Holtec Project 2635

Appendix 4.B Additional Mislocated Accident Results (Total of 9 pages)

Holtec Report HI-2177590 Page 4.B-1 Holtec Project 2635

Table 4.B.1 Additional MCNP5-1.51 Results for the Mislocated Fuel Assembly Accident Calculations with Administrative Requirement 3 Case kcalc sigma Description 4.2.5.5.33 1.0194 0.0003 Same as Case 4.2.5.5.17 except the SFP concrete wall is present.

4.2.5.5.34 0.8793 0.0003 Same as Case 4.2.5.5.33 except the SFP moderator has 1000 ppm soluble boron.

Same as Case 4.2.5.5.33 except the PWR fuel is a checkerboard of fresh fuel 4.2.5.5.35 0.9608 0.0003 and emotv cells.

4.2.5.5.36 0.8296 0.0003 Same as Case 4.2.5.5.35 except the SFP moderator has 1000 ppm soluble boron.

Same as Case 4.2.5.5.33 except the BWR fuel is eccentrically positioned 4.2.5.5.37 1.0284 0.0003 toward the mislocated PWR fuel.

4.2.5.5.38 0.8869 0.0003 Same as Case 4.2.5.5.37 except the SFP moderator has 1000 ppm soluble boron.

Same as Case 4.2.5.5.37 except the PWR fuel is a checkerboard of fresh fuel 4.2.5.5.39 0.9653 0.0003 and empty cells.

4.2.5.5.40 0.8329 0.0003 Same as Case 4.2.5.5.39 except the SFP moderator has 1000 ppm soluble boron.

Same as Case 4.2.5.5.33 except the mislocated PWR fuel face adjacent to 4.2.5.5.41 1.0224 0.0003 the PWR fuel in the PWR BORAFLEX TM racks.

4.2.5.5.42 0.8803 0.0003 Same as Case 4.2.5.5.41 except the SFP moderator has 1000 ppm soluble boron.

Same as Case 4.2.5.5.35 except the mislocated PWR fuel face adjacent to the 4.2.5.5.43 0.9675 0.0003 PWR fuel in the PWR BORAFLEX TM racks.

4.2.5.5.44 0.8329 0.0003 Same as Case 4.2.5.5.43 except the SFP moderator has 1000 ppm soluble boron.

Same as Case 4.2.5.5.37 except the mislocated PWR fuel face adjacent to 4.2.5.5.45 1.0309 0.0003 the PWR fuel in the PWR BORAFLEX TM racks.

4.2.5.5.46 0.8869 0.0003 Same as Case 4.2.5.5.45 except the SFP moderator has 1000 ppm soluble boron.

Same as Case 4.2.5.5.39 except the mislocated PWR fuel face adjacent to 4.2.5.5.47 0.9701 0.0003 the PWR fuel in the PWR BORAFLEX TM racks.

4.2.5.5.48 0.8348 0.0003 Same as Case 4.2.5.5.47 except the SFP moderator has 1000 ppm soluble boron.

Holtec Report HI-2177590 Page 4.B-2 Holtec Project 2635

Table 4.B.2 Additional MCNP5-l .51 Results for the Mislocated Fuel Assembly Accident Calculations with NO Administrative Requirement 3 Case kcalc sigma Description 4.2.5.5.49 1.0505 0.0003 Same as Case 4.2.5.5.1 except with concrete wall present.

4.2.5.5.50 0.9063 0.0003 Same as Case 4.2.5.5.49 except the SFP moderator has 1000 ppm soluble boron.

Same as Case 4.2.5.5.49 except the PWR fuel is a checkerboard of fresh fuel 4.2.5.5.51 0.9921 0.0003 and empty cells.

4.2.5.5.52 0.8547 0.0003 Same as Case 4.2.5.5.51 except the SFP moderator has 1000 ppm soluble boron.

Same as Case 4.2.5.5.49 except the BWR fuel is eccentrically positioned toward 4.2.5.5.53 1.0651 0.0003 the mislocated PWR fuel.

4.2.5.5.54 0.9261 0.0003 Same as Case 4.2.5.5.53 except the SFP moderator has 1000 ppm soluble boron.

Same as Case 4.2.5.5.53 except the PWR fuel is a checkerboard of fresh fuel 4.2.5.5.55 1.0018 0.0003 and empty cells.

4.2.5.5.56 0.8703 0.0003 Same as Case 4.2.5.5.55 except the SFP moderator has 1000 ppm soluble boron.

Same as Case 4.2.5.5.49 except the mislocated PWR fuel face adjacent to the PWR 4.2.5.5.57 1.0369 0.0003 fuel in the PWR BORAFLEX TM racks.

4.2.5.5.58 0.8887 0.0003 Same as Case 4.2.5.5.57 except the SFP moderator has 1000 ppm soluble boron.

Same as Case 4.2.5.5.51 except the mislocated PWR fuel face adjacent to the PWR 4.2.5.5.59 0.9799 0.0003 fuel in the PWR BORAFLEX TM racks.

4.2.5.5.60 0.8392 0.0003 Same as Case 4.2.5.5.59 except the SFP moderator has 1000 ppm soluble boron.

Same as Case 4.2.5.5.53 except the mislocated PWR fuel face adjacent to the PWR 4.2.5.5.61 1.0627 0.0003 fuel in the PWR BORAFLEX TM racks.

4.2.5.5.62 0.9065 0.0003 Same as Case 4.2.5.5.61 except the SFP moderator has 1000 ppm soluble boron.

Same as Case 4.2.5.5.55 except the mislocated PWR fuel face adjacent to the PWR 4.2.5.5.63 1.0018 0.0003 fuel in the PWR BORAFLEX TM racks.

4.2.5.5.64 0.8542 0.0003 Same as Case 4.2.5.5.63 except the SFP moderator has 1000 ppm soluble boron.

Holtec Report HI-2177590 Page 4.B-3 Holtec Project 2635

Table 4.B.3 MCNP5-l .51 Results for the Mislocated Fuel Assembly Accident Outside the BWR BORAFLEX' Rack Case kcalc sigma Description Same as Case 4.2.5 .5 .1 except the mislocated PWR fuel is between 4.2.5.5.65 0.9973 0.0003 the BWR rack and the concrete wall and adjacent the BWR rack. See Figure 4.B. l.

Same as Case 4.2.5.5.65 except the SFP moderator has 1000 4.2.5.5.66 0.8384 0.0003 ppm soluble boron.

Same as Case 4.2.5.5.3 except the mislocated PWR fuel is between 4.2.5.5.67 0.9978 0.0003 the BWR rack and the concrete wall and adjacent to the BWR rack.

Same as Case 4.2.5.5.67 except the SFP moderator has 1000 4.2.5.5.68 0.8385 0.0003 ppm soluble boron.

Same as Case 4.2.5.5.65 except the BWR fuel is eccentric towards 4.2.5.5.69 1.0155 0.0003 the mislocated PWR fuel. See Figure 4.B.2.

Same as Case 4.2.5.5.69 except the SFP moderator has 1000 4.2.5.5.70 0.8643 0.0003 ppm soluble boron.

Same as Case 4.2.5.5.67 except the BWR fuel is eccentric towards 4.2.5.5.71 1.0162 0.0003 the mislocated PWR fuel.

Same as Case 4.2.5.5.71 except the SFP moderator has 1000 4.2.5.5.72 0.8643 0.0003 ppm soluble boron.

Same as Case 4.2.5.5.65 except the mislocated PWR fuel face 4.2.5.5.73 1.0005 0.0003 adjacent to the concrete wall. See Figure 4.B.3.

Same as Case 4.2.5.5.73 except the SFP moderator has 1000 4.2.5.5.74 0.8390 0.0003 ppm soluble boron.

Same as Case 4.2.5.5.67 except the mislocated PWR fuel is adjacent 4.2.5.5.75 1.0000 0.0003 to the concrete wall.

Same as Case 4.2.5.5.75 except the SFP moderator has 1000 4.2.5.5.76 0.8397 0.0004 ppm soluble boron.

Holtec Report HI-2177590 Page 4.B-4 Holtec Project 2635

Table 4.B.4 Summary of the Mislocated Accident Results Result Bounding kcalc 0 kcalc Description Table Case ppm 1000 ppm Mislocated in center of pool, NO Administrative 4.A.9 4.2.5.5.5 1.0652 0.9266 Requirement 3, water reflector around SFP.

Mislocated in center of pool, Administrative 4.A.10 4.2.5.5.22 1 1.0290 0.8880 Requirement 3, water reflector around SFP.

Mislocated in center of pool, Administrative 4.B.l 4.2.5.5.45 1.0309 0.8869 Requirement 3, concrete reflector around SFP.

Mislocated in center of pool, NO Administrative 4.B.2 4.2.5.5.53 1.0651 0.9261 Requirement 3, concrete reflector around SFP.

Mislocated outside rack, NO Administrative 4.B.3 4.2.5.5.71 1.0162 0.8643 Requirement 3, concrete reflector around SFP.

1 The maximum reactivity case is selected from the most reactive of the 1000 ppm calculations.

Holtec Report HI-2177590 Page 4.B-5 Holtec Project 2635

Figure 4.B.1 (1 of2)

Additional Mislocated Fuel Assembly Accident Location Concrete wall mislocated fuel assembly A low resolution 2D image of the mislocated fuel assembly accident outside the BWR BORAFLEX' rack.

Holtec Project 2635 Page 4.B-6 Holtec Report HI-2177590

Figure 4.B. 1 (2of2)

A partial 2D representation of the MCNP model for Case 4.2.5.5.65 with BWR fuel in bounding eccentric position.

Holtec Report HI-2177590 Page 4.B-7 Holtec Project 2635

Figure 4.B.2 Additional Mislocated Fuel Assembly Accident Configuration A partial 2D representation of the MCNP model for Case 4.2.5.5.69 with BWR fuel eccentric towards the mislocated PWR fuel assembly.

Holtec Report HI-2177590 Page 4.B-8 Holtec Project 2635

Figure 4.B.3 Additional Mislocated Fuel Assembly Accident Configuration Adjacent to the SFP Concrete Wall A partial 2D representation of the MCNP model for Case 4.2.5.5.73 with BWR fuel eccentric towards the mislocated PWR fuel assembly.

Holtec Report HI-2177590 Page4.B-9 Holtec Project 2635

CHAPTER 5: THERMAL-HYDRAULIC EVALUATION 5.1 Introduction This chapter provides a summary of the methods, models, analyses and numerical results to demonstrate that the Westinghouse-supplied BWR spent fuel storage racks (SFSRs) in Shearon Harris spent fuel pools (SFPs) A & B will continue to meet the thermal-hydraulic requirements for safe storage of spent fuel following installation of Holtec-supplied DREAM inserts. Similar thermal-hydraulic analyses have been used for spent fuel storage licensing applications at Harris in the past.

The following specific thermal-hydraulic analyses are performed:

1. Assessment of the plant's current SFP bulk thermal evaluation to determine ifit will continue to apply following installation of the DREAM inserts.

2. Assessment of the plant's current SFP time-to-boil evaluation to determine if it will continue to apply following installation of the DREAM inserts.

3. A rigorous Computational Fluid Dynamics (CFD) based study to conservatively quantify the peak local water temperatures in the Westinghouse-supplied BWR SFSRs following installation of the DREAM inserts.

4. Determination of a bounding maximum fuel cladding temperature in the Westinghouse-supplied BWR SFSRs following installation of the DREAM inserts.

These analyses are described in detail in Sections 5.3 through 5.5. A single scenario is postulated and analyzed, with all Westinghouse BWR SFSRs loaded with fuel assemblies having the maximum decay heat per assembly permitted in the Vectra IF-300 shipping cask used to transport them to Shearon Harris and the SFP bulk temperature set to the bulk temperature limit of 150°F.

In the sections that follow, analysis methods are described, a single scenario is evaluated and results are presented.

Holtec Report HI-2177590 5-1 Holtec Project 2635

5.2 Acceptance Criteria Applicable codes, standards and regulations include the following:

a. NUREG-0800, Standard Review Plan, Section 9.1.3 [5.2.1].

b. USNRC OT Position Paper for Review and Acceptance of Spent Fuel Storage and Handling Application, 4/78 [5.2.2].

The design of the DREAM inserts must ensure that all fuel assemblies in the Westinghouse BWR SFSRs will continue to be adequately cooled by circulation of water for the design-basis scenario.

The Westinghouse BWR SFSRs with DREAM inserts are evaluated to the following criteria:

1. Following the planned offload fuel assemblies from the Harris reactor and with forced cooling available, the bulk SFP temperatures shall be limited to 150°F. This criterion can be met by demonstrating that the existing licensing basis remains applicable.

2. Under a complete failure of active cooling during the limiting fuel offload scenario, the water surface is allowed to reach saturation. Sufficient time must be available before the onset of bulk boiling to implement corrective measures. This criterion can be met by demonstrating that the existing licensing basis remains applicable.

3. Local water and fuel cladding temperatures for the fuel assemblies within the Westinghouse BWR SFSRs shall not exceed the local saturation temperature of water.

Holtec Report HI-2177590 5-2 Holtec Project 2635

5.3 Assumptions and Design Data 5 .3 .1 Assumptions The following assumptions are applied to render a conservative portrayal of thermal-hydraulic conditions in Shearon Harris SFPs A and B.

1. Heat loss by natural convection, mass diffusion and thermal radiation from the surface of the SFP water is neglected, as is conduction heat transfer through the SFP structure. Thus, all decay heat loads are considered to be removed by the SFP cooling system alone, maximizing computed temperatures.

2. No downcomer flow is assumed to exist between the Westinghouse BWR SFSR modules in the SFPs, minimizing the ability of cooled water to enter the bottom of the rack cells.

3. All SFSR cells are assumed to have the inlet flow holes geometry of the pedestal cells. This conservatively reduces the water flow area into the storage cells, thereby increasing the hydraulic resistance.

4. An additional heat transfer resistance of 0.01 hrxft2 x°F/Btu is conservatively imposed on the outside of the fuel rods, to account for any crud layer, thereby increasing the calculated fuel cladding superheat.

5. The maximum local water temperature (at the SFSR cell exits) and the peak heat flux (typically near the mid-height of the active fuel region) are considered to occur co-incidentally. The superposition of these two maximum values ensures that the calculated peak fuel cladding temperature bounds the fuel cladding temperature anywhere along the length of the fuel assembly.

5 .3 .2 Design Data The principal design data employed to determine if the current licensing-basis time-to-boil remains applicable are summarized in Table 5.3.1. The principal design data employed for the local thermal-hydraulic analyses are presented in Table 5.3.2.

Holtec Report HI-2177590 5-3 Holtec Project 2635

Table 5.3.1

SUMMARY

OF INPUTS FOR TIME-TO-BOIL EVALUATION INPUT DATA VALUE Westinghouse BWR SFSR Weight 10890 lb.

Holtec BWR SFSR Weight 18200 lb.

Number of Westinghouse BWR SFSRs 8 DREAM Insert Weight 20 lb.

Maximum Number of DREAM Inserts 967 Stainless Steel Density 501 lb./ft. 3 Metamic Density 2.65 gm./cm.3 Holtec Report HI-2177590 5-4 Holtec Project 2635

Table 5.3.2

SUMMARY

OF INPUTS FOR LOCAL TEMPERATURE ANALYSES INPUT DATA VALUE SFP Water Elevations Floor 246 ft.

Water Surface 284 ft.

SFP Rack-to-Wall Gaps SFP A- South Wall 10 3/4 in.

SFP B- West Wall 12 3/4 in.

SFP B - South Wall 7 3/4 in.

SFP B - East Wall 2 3/4 in.

SFSR Plan Width 5 ft. 9 1/2 in.

Racks-to-Floor Plenum Height 6 1/2 in.

Rack Cell Length 169 in.

Active Fuel Length 144 in.

Fuel Assembly Array Size 9x9 Rack Cell Pitch 6 1/4 in.

Fuel Assembly Channel ID 5.278 in.

Number of Flow Holes per Pedestal 4 Rack Pedestal Flow Holes Diameter 2 in.

Maximum Nominal Fuel Assembly Heat 2353 Btu/hr.

Assembly Axial Peaking Factor 1.25 Assembly Radial Peaking Factor 1.6 Holtec Report HI-2177590 5-5 Holtec Project 2635

5.4 Bulk SFP Temperatures 5.4.l Equilibrium SFP Temperature Evaluation The effect of the addition of the inserts on the equilibrium SFP temperatures is evaluated by reviewing the current licensing basis evaluation to identify whether the SFP water volume or thermal inertia is credited in performing the calculation. This review indicates that steady-state heat balances are used to determine the equilibrium bulk temperatures. No credit is taken for heat energy storage by the SFP water, so neither SFP water volume nor thermal inertia is credited.

Because no water volume is credited the displacement of water by the DREAM inserts has no impact and the existing licensing basis evaluation remains applicable following addition of the inserts. This satisfies Acceptance Criterion 1.

5.4.2 Time-To-Boil Evaluation The addition of the DREAM inserts displaces a quantity of SFP water, which slightly reduces the thermal inertia of the SFP. The current licensing basis time-to-boil calculation is a transient evaluation, so the SFP thermal inertia is credited. The effect of the addition of the inserts on the time-to-boil is evaluated by reviewing the current licensing basis evaluation to determine if the credited water volume contains sufficient margin to bound the addition of the DREAM inserts.

The results of this comparison are presented in Table 5.4.1. The credited water volume in the current licensing basis evaluation is low enough to bound the addition of the inserts. This satisfies Acceptance Criterion 2.

Holtec Report HI-2177590 5-6 Holtec Project 2635

Table 5.4.1

SUMMARY

OF TIME-TO-BOIL EVALUATION RESULTS Conservatism in Credited Water Volume 117.9 ft. 3 Total Dream Insert Displaced Volume 116.9 ft. 3 Holtec Report HI-2177590 5-7 Holtec Project 2635

5.5 Local Water and Fuel Cladding Temperatures The objective of the local temperature analyses is to demonstrate that the principal thermal-hydraulic criterion of ensuring local subcooled conditions in the Westinghouse BWR SFSRs is met. Adequate cooling is demonstrated by performing a rigorous evaluation of the coupled velocity and temperature fields in the Westinghouse BWR SFSRs in each SFP.

For determining the maximum local water temperature, three-dimensional Computational Fluid Dynamics (CFD) analyses are implemented. There are several significant geometric and thermal-hydraulic features of the Westinghouse BWR SFSRs in each 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 bulk region outside the Westinghouse BWR SFSRs, where the classical Navier-Stokes equations are solved with turbulence effects included. The other region is the heat-generating zone of Westinghouse BWR SFSRs loaded with fuel assemblies, where water flow is directed vertically upwards by the buoyancy forces through relatively small flow channels formed by the fuel assembly rod arrays in each rack cell. The Westinghouse BWR SFSRs are modeled as porous medium regions in which Darcy's Law [5.5.1] governs fluid flow.

The CFD analyses are performed using version 6.3.26 of the Fluent [5.5.2] fluid flow and heat transfer modeling program. The Fluent code enables buoyancy flow and turbulence effects to be included in the CFD analysis. Turbulence effects are modeled by relating time-varying "Reynolds' Stresses" to the mean bulk flow quantities by the standard k-s turbulence model.

The peak fuel rod cladding temperature is computed by following a series of calculation steps as outlined below:

Step 1: Compute the maximum local water temperatures as just described above.

Step 2: Compute the maximum cladding to local water temperature difference (~Tc).

Step 3: Compute a bounding maximum fuel rod cladding temperatures by adding~ Tc to the maximum local water temperatures.

Holtec Report HI-2177590 5-8 Holtec Project 2635

The procedure to perform Step 2 is presented next.

The maximum specific decay power of a single fuel assembly is denoted by QA. The most emissive fuel rod can produce fr times the average heat emission rate, where fr is the radial peaking factor.

A fuel rod can also produce fz times the average heat emission rate over a small length, where fz is the axial peaking factor. The axial heat distribution in a fuel rod is highest in the central region, and tapers off at its two extremities. Thus, peak cladding heat flux per unit heat transfer area of fuel rod is given by the equation:

QA xf, xf_

qpeak = A -

rods where Arocts is the total external heat transfer area of the cladding in the active fuel region of a single fuel assembly.

Within each fuel assembly rod sub-channel, water is continuously heated by the cladding as it moves axially upwards from bottom to top under laminar flow conditions. Rohsenow and Hartnett

[5.5.3] report a Nusselt number, Nu, for heat transfer in a laminar flow situation through a heated channel as:

Nu=~xDh=4.364 kwater where:

kwater is the water thermal conductivity, Btu/(hr.-ft.-°F) he is the laminar flow convective heat transfer coefficient, Btu/(hr.-ft. 2-°F)

Dh is the sub-channel hydraulic diameter, ft.

In order to introduce some additional conservatism in the analysis, it is assumed that the fuel cladding has a crud deposit thermal resistance, Rcruct, which covers the entire surface. Therefore, Holtec Report HI-2177590 5-9 Holtec Project 2635

the overall heat transfer coefficient U, considering a crud deposit resistance Rcruct, can be defined by the following:

The temperature drop, ~Tc, between the outer surface of the fuel cladding and the water flowing up through the assembly at the peak cladding flux location is computed by the following:

~T = qpeak c u Finally, the maximum fuel rod temperature (Step 3 above) is defined by the following:

where:

Troct is the maximum fuel clad temperature T1oca1 is the maximum local water temperature A solution of each CFD model is performed to obtain the coupled flow and temperature fields, the maximum local water temperature is extracted from each temperature field, and then the maximum fuel cladding temperatures are computed as described. The maximum local water temperatures, fuel cladding superheat and bounding fuel cladding temperatures are summarized in Table 5.5.1.

Temperature contours in a vertical plane through the center of the Westinghouse BWR SFSRs in each SFP are shown in Figures 5.5.l and 5.5.2. At the top of the active fuel length, the local saturation temperature is approximately 240°F. From the local water and fuel cladding temperature results, it is concluded that local water and fuel cladding temperatures remain below saturation, satisfying Acceptance Criterion 3.

Holtec Report HI-2177590 5-10 Holtec Project 2635

Table 5.5.1

SUMMARY

OF LOCAL TEMPERATURE RESULTS PARAMETER CALCULATED VALUE Maximum Water Temperature SFPA 152.6°F SFPB 152.7°F Fuel Cladding Superheat l.4°F Bounding Fuel Cladding Temperature SFPA 154.0°F SFPB 154.l °F Holtec Report HI-2177590 5-11 Holtec Project 2635

152.56 PLANT 152.43 NORTH 152.30 152.17 152.04 iJ 151.92 151.79 151.66 151.53 151.41 151.28 151.15 151.02 150.89 150.77 150.64 150.51 150.38 150.26 ~

150.13 't---X ,_,,___.----*-*-*-***"""_________ .

150.00 Contours of Static Temperature (f) Mar 03, 2017 FLUENT 6.3 (3d, dp, pbns, ske)

FIGURE 5.5.1: CONTOURS OF STATIC TEMPERATURE IN A PLANE THROUGH THE CENTER OF THE WESTINGHOUSE BWR SFSRs IN SFP A Holtec Report HI-2177590 5-12 Holtec Project 2635

152.68 152.55 152.41 PLANT 152.28 NORTH 152.14 152.01 151.88 151.74 151.61 151.47 151.34 151.21 151.07 150.94 150.80 150.67 150.54 150.40 150.27 150.13 \_ __x 150.00 Contours of Static Temperature (f) Mar 03, 2017 FLUENT 6.3 (3d, dp, pbns, ske)

FIGURE 5.5.2: CONTOURS OF STATIC TEMPERATURE IN A PLANE THROUGH THE CENTER OF THE WESTINGHOUSE BWR SFSRs IN SFP B Holtec Report HI-2177590 5-13 Holtec Project 2635

5.6 References

[5 .2.1] "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants - L WR Edition," NUREG-0800, Revision 2, June 1987.

[5.2.2] "OT Position Paper for Review and Acceptance of Spent Fuel Storage and Handling Applications," April 14, 1978.

[5.5.1] "Flow of Fluids Through Valves, Fittings, and Pipe," Crane Technical Paper No. 410, Crane Valve Company, Twenty-Second Printing, 1985.

[5.5.2] Fluent Computational Fluid Dynamics Software, Ansys Inc.

[5.5.3] Rohsenow, W.M. and J.P. Hartnett, "Handbook of Heat Transfer," McGraw Hill Book Company, NY, 1973.

Holtec Report HI-2177590 5-14 Holtec Project 2635

6.0 EXISTING RACK STRUCTURAL/SEISMIC CONSIDERATIONS 6.1 Introduction This section examines the structural adequacy of the Shearon Harris Spent Fuel Pool (SFP) racks, after Dream' inserts have been added to the existing BWR Boraflex racks located in Pools A and B. Loadings postulated to occur during normal, seismic, and accident conditions have been considered. The effects of the Dream' inserts on the structural design bases are evaluated by reviewing the existing analysis reports listed as References [6.7.l] through [6.7.3].

The evaluation of the structural design bases for Harris Pools A and B considered the following specific areas:

Seismic Qualification of Existing BWR Boraflex Racks

Fuel Pool Structural Qualification

Pool Liner Qualification

Mechanical Accident Evaluation The findings and conclusions with respect to each of these structural areas are presented below.

6.2 Seismic Qualification of Existing BWR Boraflex Racks There are three (3) existing 1 lxll BWR Boraflex racks in Pool A and five (5) existing 1lxl1 BWR Boraflex racks in Pool B that will receive Dream' inserts. Per Table 2.2 of Chapter 2, a single Dream' insert weighs only which is a small fraction of the total weight of a loaded rack as shown below.

Quantity x Weight Item Quantity Dry Weight (lb) Source (lb) l lxl 1 BWR Boraflex Rack 10,890 10,890 Reference [6.7.1]

BWR Fuel Assembly 121 681 (max.)

82,401 95,711 Table C-5 ofRef. [6.7.2]

Holtec Report HI-2177590 6-1 Project 2635

In percentage terms, the Dream' inserts account for only~ of the total rack weight. Furthermore, the existing seismic qualification for the spent fuel racks in Pool B (Ref. [6.7.3]) conservatively uses an input value of 13,100 lb for the dry weight of an empty llxl 1 BWR Boraflex Rack (as opposed to actual weight of 10,890 lb). This additional weight (13,100 lb - 10,890 lb = 2,210 lb) is almost equal to the total weight of the Dream' inserts installed in the rack - Thus, the conservatism in the existing analysis for Pool B offsets the added weight of the Dream' such that the net difference in the loaded weight of a 1lxl1 BWR Boraflex Rack is very small(< 1%). Given that the Dream'inserts have a negligible effect on the loaded rack weight, it can also be concluded that their planned use in Pools A and B will likewise have a negligible impact on the existing seismic qualification of the 1lxl1 BWR Boraflex Racks.

6.3 Fuel Pool Structural Qualification The Dream' inserts have even a smaller impact on the fuel pool structural qualifications for Pool A and Pool B. This is because the total weight of the installed Dream' inserts is extremely small in comparison to the other load contributors (e.g., pool water, rack weight, fuel weight). For example, the table below lists the static loads on the pool floor for Pool B, which has the largest number of Dream' inserts.

Quantity x Weight Item Quantity Weight (lb) Source (lb) l lxl 1 BWR Boraflex Rack 5 10,890 54,450 Reference [6.7.1]

l lxl 1 BWR Holtec Rack 13 13,100 170,300 Reference [6.7.3]

6x10 PWR Boraflex Rack 5 18,000 90,000 Reference r6.7.3l 7x 10 PWR Boraflex Rack 6 21,000 126,000 Reference [6.7.31 \

6x8 PWR Boraflex Rack 1 14,400 14,400 Reference [6.7.3]

BWR Fuel Assembly 2,178 681 (max.) 1,483,218 Table C-5 of Ref. [6.7.21 PWR Fuel Assembly Neutron Absorber Insert (5 BWR racks)

SFP Water (50' L x 27' W x 35' min.

height)

Total Weight 768 605 1

1,440 (max.)

2,948,400 1,105,920 2,943,675

> 6,000,000 Table B-2 of Ref. [6.7.21 Assumed height; water density = 62.3 pcf Holtec Report HI-2177590 6-2 Project 2635

From the above table, it is obvious that the Dream inserts have an infinitesimal impact on the TM fuel pool structural qualification for Pool B, as they represent only . . of the total load on the pool floor. Although it is smaller in size and capacity, the conclusion is the same for Pool A.

6.4 Pool Liner Qualification During a seismic event, the Spent Fuel Racks transmit vertical and horizontal forces to the pool liner through the support pedestals at the base of the racks. The loads acting on the liner are proportional to the loaded weight of the Spent Fuel Racks. As discussed above, the Dream TM inserts represent a very small portion of the total loaded weight of a 1lxl1 BWR Boraflex rack, and therefore the inserts will have minimal effect on the pool liner qualification.

6.5 Mechanical Accident Evaluation The addition of the Dream TM inserts can only improve the result of the mechanical accident analysis as the insert would provide some reinforcement to the cell wall and absorb some of the impact energy during an accidental fuel assembly drop. The accidental drop of a neutron absorber insert plus its handling tool onto the top of a Spent Fuel Rack is also bounded by the existing drop analysis as the weight of the insert plus handling tool (< 200 lb) is much less than a BWR fuel assembly.

6.6 Conclusion In conclusion, the structural design bases for the existing 1lxl1 BWR Boraflex racks in Harris Pools A and B, as well as the structural qualification of the pool structures, are not adversely affected by the planned installation of neutron absorber inserts in three (3) 1lxl1 BWR Boraflex racks in Pool A (363 locations) and in five (5) llxl 1 BWR Boraflex racks in Pool B (605 locations). The weight of the neutron absorber inserts are negligibly small in comparison to the overall dead weight of these structures and their contents.

Holtec Report HI-2177590 6-3 Project 2635

Furthermore, the structural integrity of the Dream' inserts under normal (i.e. installation and handling) and accident condition (i.e seismic events) loads has been evaluated in [6.7.4]. Per the analysis in [6.7.4], Dream' inserts are found to be structurally adequate to perform their intended design function under both normal and seismic conditions.

6. 7 References

[6.7.1] CP&L Cale ID SF-0038, "Spent Fuel Pool Heat Up Rate I Time to Boil Calculation", Revision 2.

[6.7.2] CP&L Calculation No. HNP-F/NFSA-0076, "Spent Fuel Shipping and Storage Parameters", Revision 2.

[6.7.3] Holtec Report No. HI-90526, "Sourcebook for New BWR Racks and Whole Pool Module Layout for the Shearon Harris Nuclear Power Plant",

Revision 0.

[6.7.4] Holtec Report No. HI-2167295, "Structural Evaluation of Harris Dream Insert", Revision 1.

Holtec Report HI-2177590 6-4 Project 2635

U.S. Nuclear Regulatory Commission Serial HNP-17-008 HNP-17-008 ATTACHMENT 6 PROPOSED TECHNICAL SPECIFICATION BASES CHANGE (FOR INFORMATION ONLY)

SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400 RENEWED LICENSE NUMBER NPF-063 1 PAGE PLUS COVER

and the BWR racks that utilize rack inserts racks' or that contained in either the rack inserts

v t e Letter Sequence Request
CAC:MF9996, (Approved, Closed)
Initiation Request Acceptance... Supplement, Supplement, Supplement Administration Withholding Request Acceptance Questions 8 January 2018 Answers 18 January 2018 Results Approval Other:
MONTHYEAR2018-01-08 [Table View]

HNP-17-008, License Amendment Request Regarding Spent Fuel Storage Pool Criticality Analyses

Text

Subject:

Subject:

SUMMARY

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

SUMMARY

1.0 INTRODUCTION

SUMMARY

SUMMARY

SUMMARY

SUMMARY

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