W3F1-2008-0052, Transmittal of Waterford Steam Electric Station, Unit 3, License Amendment Request to Modify Technical Specification Section 5.6, Fuel Storage and Add New Technical Specification 3/4 9.12, Spent Fuel Pool Boron Concentration

Transmittal of Waterford Steam Electric Station, Unit 3, License Amendment Request to Modify Technical Specification Section 5.6, Fuel Storage and Add New Technical Specification 3/4 9.12, Spent Fuel Pool Boron Concentration

ML082660649
Person / Time
Site:	Waterford
Issue date:	09/17/2008
From:	Walsh K Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NPF-38-277, W3F1-2008-0052
Download: ML082660649 (106)

Text

Entergy Nuclear South Entergy Operations, Inc.

SEn tergy 17265 River Road Killona, LA 70057 Tel 504 739 6660 Fax 504 739 6678 kwalshl @entergy.com Kevin T. Walsh Vice President, Operations Attachment 3 contains Proprietary Information Waterford 3 W3F1-2008-0052 September 17, 2008 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

License Amendment Request NPF-38-277 License Amendment Request to Modify Technical Specification Section 5.6, Fuel Storage and Add New Technical Specification 3/4 9.12, Spent Fuel Pool Boron Concentration Waterford Steam Electric Station, Unit 3 Docket No. 50-382 License No. NPF-38

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Entergy Operations, Inc. (Entergy) hereby requests the following amendment to Waterford Steam Electric Station, Unit 3 (Waterford 3) Technical Specifications (TS). The Waterford 3 TS are being revised to take credit for soluble boron in Region 1 (cask storage pit) and Region 2 (spent fuel pool and refueling canal) fuel storage racks for the storage of both Standard and Next Generation Fuel (NGF) assemblies. In accordance with 10CFR 50.68, the limits for keff of the spent fuel storage racks are appropriately revised based on analysis to maintain keff less than 1.0 when flooded with unborated water, and less than, or equal to, 0.95 when flooded with water having a minimum boron concentration of 447 ppm, during normal conditions. A new Technical Specification is added which includes a surveillance that ensures the required boron concentration is maintained in the spent fuel storage racks. This added Technical Specification Surveillance conforms to the guidance of NUREG-1432. The change is evaluated for both normal operation and accident conditions.

This change will provide more flexibility in storing the more reactive NGF assemblies in the spent fuel storage racks.

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

There are no new regulatory commitments contained in this submittal.

Entergy requests NRC approval of the proposed amendment by September 10, 2009, in order to support the Fall 2009 planned refueling outage. Once approved, the amendment will be implemented within 60 days of receipt.

A O0/

W3F1-2008-0052 Page 2 contains an analysis of the proposed TS change. Attachment 2 contains the proposed TS changes. Attachment 3 documents the criticality safety evaluation for the storage of Standard and Next Generation Fuel assemblies in Holtec-designed Region 1 and Region 2 style high-density spent fuel storage racks (SFSRs) at Waterford Unit 3, and the evaluation provides the analytical basis for the Technical Specification changes. Attachment 3 contains proprietary information. The proprietary information was provided to Entergy in a Holtec International transmittal that is referenced by an affidavit. Holtec requests the enclosed proprietary information identified in Attachment 3 be withheld from public disclosure in accordance with the provisions of 10CFR 2.390 and 10CFR 9.17. Attachment 4 contains the affidavit for withholding the proprietary information contained in Attachment 3. Attachment 5 contains the Non-Proprietary Licensing Report documenting the criticality safety evaluation for the storage of Standard and Next Generation Fuel assemblies'in Holtec-designed Region 1 and Region 2 style high-density spent fuel storage racks (SFSRs) at Waterford Unit 3.

Please contact Robert J. Murillo, Manager, Licensing at (504) 739-6715 if there are any questions concerning this matter.

I declare under penalty of perjury that the foregoing is true and correct. Executed on September 17, 2008.

Sincerel K /GCS/ssf Attachments:

1. Analysis of Proposed Technical Specification Change

2. Proposed Technical Specification Changes (mark-up) 3 Holtec Report: Licensing Report for Waterford 3 NGF Criticality Analysis (Proprietary Information) 4 Affidavit for withholding Proprietary Information

5. Holtec Report: Licensing Report for Waterford 3 NGF Criticality Analysis (Non Proprietary Information)

W3F1 -2008-0052 Page 3 cc: Mr. Elmo E. Collins Regional Administrator U. S. Nuclear Regulatory Commission Region IV 612 E. Lamar Blvd., Suite 400 Arlington, TX 76011-8064 NRC Senior Resident Inspector Waterford 3 P. 0. Box 822 Killona, LA 70066-0751 U. S. Nuclear Regulatory Commission Attn: Mr. N. Kalyanam MS 0-07 D1 Washington, DC 20555-0001 American Nuclear Insurers Attn: Library 95 Glastonbury Blvd.

Suite 300 Glastonbury, CT 06033-4443 Wise, Carter, Child & Caraway Attn: J. Smith P.O. Box 651 Jackson, MS 39205 Louisiana Department of Environmental Quality Office of Environmental Compliance Surveillance Division P. Oz Box 4312 Baton Rouge, LA 70821-4312 Winston & Strawn ATTN: N.S. Reynolds 1700 K Street, NW Washington, DC 20006-3817 Morgan, Lewis & Bockius LLP ATTN: T.C. Poindexter 1111 Pennsylvania Avenue, NW Washington, DC 20004

Attachment 1 to W3F1-2008-0052 Analysis of Proposed Technical Specification Change to W3F17-2008-0052 Page 1 of 10

1.0 DESCRIPTION

This letter is a request to amend Operating License NPF-38 for Waterford Steam Electric Station, Unit 3 (Waterford 3). The Waterford 3 TS are being revised to take credit for soluble boron in the Region 1 (cask storage pit) and Region 2 (spent fuel pool and refueling canal) fuel storage racks for the storage of both Standard and Next Generation Fuel (NGF) assemblies. In accordance with 10 CFR50.68, the limits for keff of the spent fuel storage racks are appropriately revised based on analysis to maintain keff less than 1.00 when flooded with unborated water, and less than, or equal to, 0.95 when flooded with water having a minimum boron concentration of 447 ppm during normal conditions. A new Technical Specification is added which includes a surveillance that ensures the required boron concentration is maintained in the spent fuel storage racks. The boron concentration will be maintained at a minimum of 1900 ppm per the new Technical Specification, significantly exceeding the required concentration levels to maintain keff within regulatory requirements. The change is evaluated for both normal operation and accident conditions. This change will provide more flexibility in storing the more reactive NGF assemblies in the spent fuel storage racks. This change does not involve a Significant Hazards Consideration, and the change is in conformance with regulatory requirements.

2.0 PROPOSED CHANGE

The proposed change will modify TS section 5.6 as follows:

a. TS 5.6.1a wording will be changed to: "For Region 1 (cask storage pit) and Region 2, (spent fuel pool and refueling canal) racks, a maximum kLIrr of less than 1.00 when flooded with unborated water, and less than, or equal to, 0.95 when flooded with water having a boron concentration of 447 ppm."

b. TS 5.6.1d will be revised to replace the words "New or partially spent" with "Fresh and irradiated" and will read as follows: "Fresh and irradiated fuel assemblies may be allowed unrestricted storage in Region 1 racks."

c. TS 5. 6. le will be revised to replace the word "New" with "Fresh" and will read as follows: "Fresh fuel assemblies may be stored in the Region 2 racks provided that they are stored in a "checkerboard pattern" as illustrated in Figure 5.6-1."

d. TS 5.6.1f will be revised to replace the words "Partially spent" with "Irradiated" and delete the word "discharge" and will read as follows: "Irradiated fuel assemblies with a burnup in'the "acceptable range" of Figure 5.6-2 may be allowed unrestricted storage in the Region 2 racks."

e. TS 5.6.1g will be revised to replace the words "Partially spent" with "Irradiated" and delete the word "discharge" and replace the word "spent" with "irradiated" and will read as follows: "Irradiated fuel assemblies with a burnup in the "unacceptable range" of Figure 5.6-2 may be stored in the Region 2 racks provided that they are stored in a "checkerboard pattern," as illustrated in Figure 5.6-1 with fuel in the "acceptable range" of Figure 5.6-3."

to W3F1-2008-0052 Page 2 of 10

f. TS 5.6.2 will be revised to replace the word "new" with "fresh" and will read as follows:

"The keff for fresh fuel stored in the new fuel storage racks shall be less than or equal to 0.95 when flooded with unborated water and shall not exceed 0.98 when aqueous foam moderation is assumed."

g. TS Figure 5.6-1 will be replaced by a new Figure 5.6-1 to show the new Alternative Checkerboard Storage Arrangements.

h. TS Figure 5.6-2 will be replaced by a new Figure 5.6-2 to show the new Acceptable Burnup Domain for Unrestricted Storage of Irradiated Fuel in Region 2 of the Spent Fuel Pool.

i. TS Figure 5.6-3 will be replaced by a new Figure 5.6-3 to show the Acceptable Burnup Domain for Irradiated Fuel in a Checkerboard Arrangement with Fuel of 5 wt %

Enrichment, or less, at 27 GWd/MTU Burnup, or higher, in Region 2 of the Spent Fuel Pool.

j. A new TS Figure, TS Figure 5.6-4, will be added to show Examples of Contiguous Fresh and Irradiated Fuel Checkerboards Which Meet Interface Requirements.

The proposed change will add TS 3/4.9.12 as follows:

1. TS 3/4.9.12 will be entitled "Spent Fuel Pool (SFP) Boron Concentration."

2. A Limiting Condition for Operation will read as follows: "3.9.12 The spent fuel pool boron concentration shall be > 1900 ppm."

3. The Applicability will read as follows: "When fuel assemblies are stored in the SFP."

4. The Action statement will read as follows: "With the spent fuel boron concentration not within limits immediately suspend movement of fuel in the SFP and immediately initiate actions to restore boron concentration to within limits."

5. The Surveillance Requirements will read as follows: "4.9.12 Verify the spent fuel pool concentration is within limits once per 7 days."

3.0 BACKGROUND

License Amendment 214, approved on May 9, 2008, allowed the use of Next Generation Fuel (NGF) for Waterford 3. The new NGF fuel assemblies have a higher fuel pellet density, smaller rod diameter and thinner fuel rod cladding which results in the NGF fuel assembly being more reactive than the current Standard fuel assemblies. The acceptable storage patterns of the NGF assemblies in the spent fuel storage racks are currently limited due to the higher reactivity of these assemblies. The proposed TS changes will provide more flexibility in the storage pattern for NGF stored, in the spent fuel storage racks by taking credit for soluble boron to ensure that keff remains within regulatory limits. 'Criticality analysis has demonstrated that taking credit for soluble boron in the spent fuel storage racks will ensure that keff remains within regulatory limits.

to VV3F1-2008-0052 Page 3 of 10 The purpose of the spent fuel storage racks is to maintain the fresh and irradiated assemblies in a safe storage condition. The current licensing basis as defined by the existing Technical Specification Requirements, Section 5.6, and federal code requirements, 10 CFR 50.68, specify the normal and accident parameters associated with maintaining the fresh and irradiated assemblies in a safe storage condition. Per the existing Technical Specification, the keff of the spent fuel storage racks are designed to be maintained less than or equal to 0.95 when flooded with unborated water. 10 CFR 50.68 defines the criticality accident requirements associated with the spent fuel racks and states the following: "If no credit for soluble boron is taken, 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% probability, 95% confidence level, if flooded with unborated water. 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% probability, 95% confidence level, if flooded with borated water, and the keff must remain below 1.0, subcritical, at a 95% probability, 95% confidence level, if flooded with unborated water."

Waterford 3's current Technical Specification does not take credit for soluble boron to maintain keff < 0.95. Accordingly, Waterford 3 is in compliance with 10 CFR 50.68 which states "If no credit for soluble boron is taken, 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% probability, 95% confidence level, if flooded with unborated water." The analysis shows that a minimum soluble boron concentration of 447 ppm is required to maintain keff within the regulatory requirement of _ 0.95. Based on the proposed amendment, which will credit boron to maintain regulatory conformance, the following excerpt from 10 CFR50.68 applies: "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% probability, 95%

confidence level, if flooded with borated water, and the keff must remain below 1.00, subcritical, at a 95% probability, 95% confidence level, if flooded with unborated water." The proposed applicable Technical Specification change is in compliance with the above statement and reads as follows: "For Region 1 and Region 2 racks, a maximum keff of less than 1.0 when flooded with unborated water, and less than, or equal to 0.95 when flooded with water having a boron concentration of 447 ppm."

An analysis, provided in Attachment 3 as a Holtec International Report, demonstrated that the effective neutron multiplication factor, keff, is less than 1.00 with the racks fully loaded with fuel of the highest anticipated reactivity, and flooded with unborated water at a temperature corresponding to the highest reactivity. In addition, the analysis demonstrated that keff is less than or equal to 0.95 with the racks fully loaded with fuel of the highest anticipated reactivity, and flooded with borated water at a temperatire corresponding to the highest reactivity. The maximum calculated reactivity included a margin for uncertainty in reactivity calculations including manufacturing tolerances and is shown to be less than 0.95 with a 95% probability at a 95% confidence level. Reactivity effects of abnormal and'accident conditions were also evaluated to assure that under all credible abnormal and accident conditions, the keff will not exceed the regulatory limit of 0.95 under borated conditions or 1.0 with unborated water.

to W3F1-2008-0052 Page 4 of 10

4.0 TECHNICAL ANALYSIS

Holtec Report No. HI-2084014, entitled "Licensing Report for Waterford Unit 3 Spent Fuel Pool Criticality Analysis" (Attachment 3) provides the technical analysis for the proposed change to store Standard and Next Generation Fuel (NGF) assemblies in Holtec-designed Region 1 and Region 2 style high-density spent fuel storage racks at Waterford 3. The report analyzed the impact of the change on Region 1 and Region 2 of the spent fuel storage racks and the resultant interfaces within and between the racks in each region. Also, the report performed and evaluated various calculations related to the Fuel Transfer Carriage Criticality, Upender Criticality, New Fuel Elevator Criticality, Boron Dilution Accident Evaluation, Low Flow Rate Dilution, High Flow Dilution, Temporary Storage Racks (in the refueling pool.inside containment), Fuel Pin Storage Container, and New Fuel Storage Vault.

The results of the analysis determined that the high-density spent fuel storage racks for Waterford 3 were designed using applicable codes and standards. The analysis showed that the effective neutron multiplication factor, keff, is less than 1.00 with the racks fully loaded with the fuel of the highest anticipated reactivity, and flooded with unborated water at a temperature corresponding to the highest reactivity. The report demonstrated that keff is less than or equal to "0.95 with the racks fully loaded with fuel of the highest anticipated reactivity, and flooded with borated water at a temperature corresponding to the highest reactivity. Also, reactivity effects of abnormal and accident conditions will not result in keff exceeding the regulatory limit of 0.95 under borated conditions.

5.0 REGULATORY ANALYSIS

5.1 Applicable Regulatory Requirements/Criteria The proposed changes have been evaluated to determine whether applicable regulations and requirements continue to be met.

By letter dated February 9, 1983, the Nuclear Regulatory Commission issued a Material License (no. SNM-1913) to Waterford 3 which authorizes the receipt, possession, inspection, and storage of uranium enriched in the U-235 isotope, contained in fuel assemblies, and the receipt, possession, and use of two Pu-Be neutron sources. In the letter, the NRC granted Waterford 3 an exemption (related to criticality alarm systems) from the requirements of 10 CFR 70.24, "CriticalityAccident Requirements." With the approval of the proposed change this exemption is no longer required. Waterford 3 currently complies with the requirements of 10 CFR 50.68, "CriticalityAccident Requirements."

Waterford 3 proposed change will comply with the requirements of 10 CFR 50.68, "Criticality Accident Requirements."

There are eight criteria that must be satisfied in the regulation. Waterford 3 complies with these as follows:

1) 10 CFR 50.68, (b) (1) - Plant procedures shall prohibit the handling and storage at any one time of more fuel assembliesthan have been determined to be safely subcritical under the most adverse moderation conditions fea'sible by unborated water.

to W3F1-2008-0052 Page 5 of 10 W3's fuel handling procedures ensure that subcriticality is maintained in the reactor vessel and the spent fuel storage racks under the most adverse moderation conditions feasible by unborated water. Storage of fuel assemblies is procedurally controlled to assure keff remains below 1.0, at a 95% probability, 95% confidence level, when flooded with unborated water.

2) 10 CFR 50.68, (b) (2) - The estimated ratio of neutron production to neutron absorption and leakage (k-effective) of the fresh fuel in the fresh fuel storage racks shall be calculated assuming the racks are loaded with fuel of the maximum fuel assembly reactivity and flooded with unborated water and must not exceed 0.95, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such flooding or if fresh fuel storage racks are not used.

Criticality calculations of the new fuel vault fully loaded with Standard and NGF fresh fuel assemblies and filled with the most reactive unborated water showed that reactivity did not exceed 0.95, at a 95% probability, 95% confidence level.

3) 10 CFR 50.68, (b) (3) - If optimum moderation of fresh fuel in the fresh fuel storage racks occurs when the racks are assumed to be loaded with fuel of the maximum fuel assembly reactivity and filled with low density hydrogenous fluid, the k-effective corresponding to this optimum moderation must not exceed 0.98, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent suchmoderation or if fresh fuel storage racks are not used.

Criticality calculations were performed on the new fuel vault fully loaded with Standard and NGF fresh fuel assemblies and filled with the most reactive low density hydrogenous fluid. The results of these calculations showed that reactivity did not exceed 0.98, at a 95% probability,. 95% confidence level.

4) 10 CFR'50.68, (b) (4) - If no credit for soluble boron is taken, the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with unborated water. If credit is taken for soluble boron, the 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, and the k-effective must remain below 1.0 (subcritical), at a 95 percent probability, 95 percent confidence level, if flooded with unborated water.

Soluble boron credit will be taken in the spent fuel storage racks. Reactivity will not exceed 0.95 at a 95% probability with a 95% confidence level with at least 447 ppm boron. The criticality calculations included in the proposed change show that keff remains below 1.0, at a 95% probability, 95% confidence level, when flooded with unborated water.

5) 10 CFR 50.68, (b) (5) - The quantity of SNM, other than nuclear fuel stored onsite, is less than the quantity necessary for a critical mass.

to W3F1-2008-0052 Page 6 of 10 W3 does not currently have a quantity of SNM, other than the nuclear fuel, stored on site to establish a critical mass.

6) 10 CFR 50.68, (b) (6) - Radiation monitors are provided in storage and associated handling areas when fuel is present to detect excessive radiation levels and to initiate appropriate safety actions.

Radiation monitors are located in the spent fuel storage area which alarm in the control room. When fuel movement is in progress additional radiation monitors are placed directly on the fuel handling bridges to provide an additional audible indication of excessive radiation levels.

7) 10 CFR 50.68, (b) (7) - The maximum nominal U-235 enrichment of the fresh fuel assemblies is limited to five (5.0) percent by weight.

Per W3 TS 5.6.1 h, the maximum U-235 fuel enrichment limit is 5.0 weight percent.

8) 10 CFR 50.68, (b) (8) - The FSAR is amended no later than the next update which 50.71(e) of this part requires, indicating that the licensee has chosen to comply with 50.68(b).

The W3 FSAR will be amended no later than the next required update after the proposed TS change is approved and implemented.

Entergy has determined that the proposed changes do not require any exemptions or relief from regulatory requirements, other than the TS, and do not affect conformance with any General Design Criterion (GDC) differently than described in the Updated Final Safety Analysis Report (UFSAR).

5.2 No Significant Hazards Consideration

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

Response: No, the proposed change to take credit for soluble boron and revise the loading patterns in the Region 1 (cask storage pit) and Region 2 (main storage pool) of the spent fuel storage racks for the storage of Standard and Next Generation Fuel (NGF) assemblies will not involve a significant increase in the probability or consequences of an accident previously evaluated.

The following potential accident scenarios have been evaluated:

• Dropped assembly - horizontal

• Dropped assembly - vertical

• Misloaded fresh assembly

• Mislocated fresh fuel assembly to W3F1-2008-0052 Page 7 of 10 The proposed change in criticality limits will not increase the probability of any of these potential accidents.

For the situation in which a fuel assembly is assumed to be dropped above a spent fuel storage rack and come to rest horizontally on top of the rack, the minimum separation distance between the dropped assembly and the top ofthe active fuel region of assemblies in the racks would be more than 12 inches, which would be neutronically an infinite separation, thereby precluding a significant increase in reactivity.

A vertical drop of an assembly onto another assembly in a spent fuel storage rack has been shown to cause no significant damage to either fuel assembly. A vertical drop into an empty storage cell could result in a small, localized deformation in the rack baseplate which could produce a misalignment between the active fuel region of the dropped assembly and the neutron absorbing Boral of the rack. The corresponding reactivity increase would be small, and would be bounded by the reactivity increase resulting from the misloading-of a fresh fuel assembly in the Region 2 racks.

,The Region 1 racks have been shown analytically to be qualified for the storage of fresh fuel assemblies with a maximum enrichment of 5.0 wt% U-235. Therefore, the misloading of a fuel assembly within the Region 1 racks is not a concern.,

The inadvertent misloading of a fresh fuel assembly into a Region 2 storage cell which was intended for the storage of an irradiated fuel assembly would not result in exceeding the regulatory kITf limit of 0.95 if a soluble boron level of 838 ppm or more were present. The concentration of boric acid in the water during fuel movement is maintained > 1900 ppm in accordance with Technical Specification 3/4.9.12.

The mislocation of a fresh fuel assembly with a maximum enrichment of 5.0 wt% U-235 outside of a Region 1 or Region 2 rack and adjacent to other fuel assemblies would not result in exceeding the regulatory k,, limit of 0.95 if a soluble boron level of 534 ppm or more were present.

Therefore, it is concluded that the proposed changes do not significantly increase the probability or consequences of any accident previously evaluated.

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

Response: No The proposed change to take credit for soluble boron and revise the-loading restrictions in the Region 1 (cask storage pit) and Region 2 (main storage pool) of the spent fuel storage racks for the storage f&-Standard and Next Generation Fuel (NGF) assemblies will not create the possibility of a new or different kind 'accident from any accident previously evaluated.

to W3F1 -2008-0052 Page 8 of 10 Soluble boron has been maintained in the Region 1 and Region 2 water and is currently required by procedures. Therefore, crediting soluble boron in the spent fuel storage rack criticality analysis will have no effect on normal pool operation and maintenance. Crediting soluble boron will only result in increased sampling to verify the boron concentration in accordance with new TS 3/4/9.12. This increased sampling ensures that a new kind of accident not previously evaluated, boron dilution in the spent, is not created. A dilution of the spent fuel storage rack soluble boron has always been a possibility. However, the boron dilution event previously had no consequences, since boron was not previously credited in the accident analysis. The initiating events that were considered for having the potential to cause dilution of the boron in the spentfuel storage pool to a level below that credited in the criticality analyses fall into three categories: dilution by flooding, dilution by loss-of-coolant induced makeup, and dilution by loss-of-cooling system induced makeup. The addition of large amounts of unborated water would be necessary to. reduce the boron concentration from the normal level of >1900 ppm to either 838 ppm or 447 ppm. This amount of water would be detected by the Operator and secured prior to reaching these boron concentrations.

A small dilution flow might result from a leak from the cooling system into the spent fuel pool. A dilution flow of 2 gpm from in-'leakage might not be immediately detected, but would require more than 1.35 days to reduce the boron concentration in the spent fuel storage racks to the minimum required 447 ppm under normal conditions, and more than 72 days to reach the 838 ppm which would be required to accommodate the most limiting fuel misloading accident. These time periods are based on a ,

conservative starting point of 1720 ppm boron. It is expected that routine surveillance measurements of the soluble boron concentration conducted every 7 days per new TS 3/4.9.12 would readily detect the reduction in concentration and provide sufficient time for corrective action prior to exceeding the regulatory limits.

The continuous operation of the Condensate Storage Pool makeup pump could add a large amount of unborated water to the spent fuel pool. Conservatively assuming instantaneous mixing of unborated water with the pool water, it would take approximately 648 minutes to reduce the soluble boron, concentration to 447 ppm which is the minimum concentration required to maintain krf below 0.95 under normal operating conditions. During this dilution accident, 389,000 gallons of water would be released into the spent fuel storage racks. For this high flow rate scenario, 346 minutes would elapse before reaching the 838 ppm concentration which is the level needed to address the most limiting fuel misloading accident. These time periods are also based on a conservative starting point of 1720 ppm boron.

A high flow rate dilution accident would result in multiple alarms alerting the control room to the situation, including the fuel pool high-level alarm, Fuel Handling Building sump high level alarm, and the Liquid Waste Management Trouble alarm. It is not considered to be credible that multiple alarms would fail or be ignored by Operators in the control room. Spilling of large volumes of water from the spent fuel storage pool would be observed by plant personnel during these calculated time periods and result in corrective actions prior to exceeding regulatory limits.

to W3F1`-2008-0052 Page 9 of 10 In the unlikely event that soluble boron in the spent fuel storage racks is completely diluted, the fuel in the spent fuel storage racks will remain subcritical by a design margin of at least 0.005 Ak, and so the kefr of the fuel in the spent fuel storage racks will remain below 1.00.

The proposed change will not result in any other change in the plant configuration or equipment design.

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

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

Response: No The proposed change to take credit for soluble boron and revise the loading restrictions in the Region 1 (cask storage pit) and Region 2 (main storage pool) of the spent fuel storage racks for the storage of Standard and NGF assemblies does not involve a significant reduction in a margin of safety.

Detailed analysis with approved and benchmarked methods has shown with a 95%

probability at a 95% confidence level that the neutron multiplication factor, kefr, of the Region 1 and Region 2 high-density spent fuel storage racks loaded with either Standard or NGF assemblies, and including biases, tolerances, and uncertainties, is less than 1.00 with unborated water, and less than 0.95 with 447 ppm of soluble boron credited. In addition, the effects of abnormal and accident conditions have been evaluated to demonstrate that under credible conditions the k,1r will not exceed 0.95 with soluble boron credited. To ensure that the margin of safety for subcriticality is maintained and that k~rr will be below 0.95, a new TS requires a soluble boron level of

> 1900 ppm in the spent fuel pool. This is much greater than the required soluble concentration of 447 ppm under normal conditions, and 838 ppm for all credible accident conditions.

Therefore, it is concluded that the proposed changes do not involve a significant reduction in the margin of safety.

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

to W3F1-2008-0052 Page 10 of 10 6.0 PRECEDENCE The NRC has approved similar submittals for ANO-2 and Calvert Cliffs in NRC SERs dated 9/30/2003 and 6/3/2004, respectively.

7.0 REFERENCES

a. Entergy License Amendment Request to Support Next Generation Fuel, Letter W3F1-2007-0037, dated August 2, 2007.

b. NRC letter dated May 9, 2008, Correction to Amendment No. 214 Re: Request to Support Next Generation Fuel, Review and Approval of Revised Emergency Core Cooling System (ECCS) Performance Analysis, and Review and Approval of Supplement to the ECCS Performance Analysis.

c. NRC letter to Louisiana Power and Light Company dated February 9, 1983 granting Waterford 3 exemption to 10 CFR 70.24 requirements.

Attachment 2 W3F1-2008-0052 Proposed Technical Specification Changes (mark-up) to W3F1-2008-0052 Page 1 of 10 DESIGN FEATURES 5.6 FUEL STORAGE CRITICALITY 5.6.1 The spent fuel storage racks are designed and shall be maintained with;

a. Xnormal k*, of le al to 0.95 whenoded unborated y , eU14C.,i{L wh include conservative wance f ncertainties.

b. A nominal 10.185 inch center-to-center distance between fuel assemblies placed in Region 1 (cask storage pit) spent fuel storage racks.

c. A nominal 8,692 inch center-to-center distance between fuel assemblies in the Region 2 (spent fuel pool and refuelling canal) racks, except for the four southern-most racks in the spent fuel pool which have an increased N-S center-to-center nominal distance of 8.892 inches.

F~reshuX lrradi,-,,'Ae*

d, eJerA4- e fuel assemblies may be allowed unrestricted storage in Region 1 racks.

e. rrle fuel assemblies may be stored in the Region 2 racks provided that they are stored in a "checkerboard pattern" as illustrated in Figure 5.6-1.

f. Pallielty sppemfuel assemblies with aj 4wag*e burnup in the "accceptable range' of Figure 5.6-2 may be allowed unrestricted storage in the Region 2 racks,

g. Peialy tpentfuel assemblies w a e burnup in the "unacceptable range" of Figure 5.6-2 may be storedlirtheRec ion 2 racks provided that they are stored in a "checkerboard pattern", as illu rate in Figuie 5.6-1, with*,ýp fuel in the "acceptable range" of Figure 5.6-3. irr Clo fe h, Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent.

56.2" The k,, forp j fuel stored in the new fuel storage racks shall be less than or equal to 0,95 when flooded with unborated water and shall not exceed 0.98 when aqueous foam moderation is assumed.

DRAINAGE 5.6.3 The spent fuel pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation +40,0 MSL. When fuel is being stored in the cask storage pit and/or the refueling canal, these areas will also be maintained at +40.0 MSL, CAPACITY 5.6.4 The spent fuel pool is designed and shall be maintained with a storage capacity limited to no more than 1849 fuel assemblies in the main pool, 255 fuel assemblies in the cask storage pit and after permanent plant shutdown 294 fuel assemblies in the refueling canal, The heat load from spent fuel stored in the refueling canal racks shall not exceed 1,72x10E6 BTU/Hr. Fuel shall not be stored in the spent fuel racks in the cask storage pit or the refueling canal unless all of the racks are installed in each respective area per the design.

5.7 NOT USED AMENDMENT NO. +0-t, 8 4 4,4186--

WATERFORD - UNIT 3 5-6 4199. 200u, to W3F1-2008-0052 Page 2 of 10 Item A For Region 1 (cask storage pit) and Region 2 (spent fuel pool and refueling canal) racks, a maximum krIT of less than 1.00 when flooded with unborated water, and less than, or equal to, 0.95 when flooded with water having a boron concentration of 447 ppm.

to W3F1-2008-0052 Page 3 of 10 4 ..~. A

,, -.I 4... 4

'U o" M to 5%

,,, // ,

h rend Cf/

Note: Either of these Cho orbackd Arrangements may be used in areas con guous to each other or to areas of unrestrictl storage in Region 2,

• 27Ulm wflhFmauup iMW/ICU d Ce0wlh d A 0ofb Wfled

• 41W.rwd ofSp@fd AWnn",wU-b Si

./

Ftgiie 5.6-1 AfMtswlv Checkerboard Arffragsm0*f WATERFORD-UNIT 3 5-6a 6aAENENNO AMENDMENT NO. 144 14 to W3F1-2008-0052 page 4 of 10

~: ~<.

jotedili fre~hfuel of lew, than.

Checkerboard Arrangement of Fresh Fuel A.~tcanblies and Empty Storage Colls

• Cells loaded with fiel of 27 GWNIMTU rhumup, or higher

((. Cels loaded with fuel having (le enrichment-bunuip

' \. ncbinafijus speeificd in Figure5.6-3 i__

Checkerboard of Fuel Askwnhlies with Btrnups of 27 GWd.MTU. or higher, and Fuel Assemblies of Specified Fnrichnient-Burnup Combinations Note: Either of these checkerboard arrangements may be used in areas contiguous to areas of unrestricted storage in Region 2. For interfaces between a fresh fuel checkerboard and an irradiated fuel checkcrb(ard, each high-reactivity irTradiated assembly ( e~g, 27 GWdIMTU ) may he face-adjacent to ii more than one fresh fuel assembly: each fresh fuel assembly may be tace-a djaccnt with up to I t,(ytw!

high-reactivity irradiated fuel assemblies: See Figure 5.6-4 tbr examples of contiguous fresh fuel checkerboards and irradiated fuel checkerboards which meet these requircments.

Figure 5.6-1 Alternative Checkerboard Storage Arrangements WAIFRFOIIDIJUfl 3  ;-6a ~ FD1N

.- MkNINDINfE NT N O 0.

to W3FI-2008-0052 Page 5 of 10

-V.

.4 N

C L.

Lm~-

,Initial. Fuel Enrichment, wt. z U-235 Fiqv*

56-2Ac~ptbl Fuai MnfOr SpentDomain rnup Unmtv~sfktd Re[on 2 5.6-2of F~gUVW 5-6b WAIENDMENT NO. 144 WATERFORD-UNIT 3 to ,

W3F1-2008-0052 Page 6 of 10

--J ..

35 ...........

33.4 Note: it is acceptable to llnearluyinterpolate between data points 30 ,*..

25 Acceptable Surnup Domain

• 20

' 15 Unla~zceptable Burnup Domain its8 10 5.,

• 2.0 2.5 3,0 3.5 4.0 4.5 5,0 Initial Fuel Enrichment ( wt% U-235)

Figure 5.6-2 Acceptable Burnup Domain for Unrestricted Storage of Irradiated Fuel in Region 2 of the Spent Fuel Pool WATRFI FORI)-UNIT 3,56 5-01) AFI~lN AN\I EMNI) ENT NO. O to W3F1-2008-0052 Page 7 of 10

,, ý( , - -ý I ,

C A N

70 60 Acepfabi Surr up Don so- -

\

00 CM 340

,,3 "1 30 "

L6D

/ Initial Fuel Enrichment, wt. x U-235 1

/.--

/

/

InAcceptable Bumu Checkerboard rOm-ent Spent W"*In focwith FuelFusi of

/ I t'/

5 9 Enrkhrnent (or low) at 27 MWID/KqU

'5-6c AMENDMENT N 0. 144 WATERFORD-UNIT 3 to W3F1-2008-0052 Page 8 of 10

-. x.

45 Note: it is acceptable to lineadylinterpolate between data points .2 40 35 30 Acceptable Burnup Domain S25 S20 Uacceptabie Burnup Dbmain U. 15 10 5

.4 2,0 2.5 30 1,5 4,0 4.5 5.0 lnitial Fuel Enrichment (wt% U-235)

Figure 5.6-3 Acceptable Burnup Domain for Irradiated Fuel in a Checkerboard Arrangement with Fuel of 5 wt% Enrichment, or Less, at 27 GWdAiMTU Burnup, or Iligher, in Region 2 of the Spent Fuel Pool NVIAERFORI)-UNrr 3 5-6c

-cAE)MNN. AMENRMENT NO.

to W3F1-2008-0052 Page 9 of 10 A AA Irradiated fuel at, or above, checkerboard curve B < 5 wt% U-235 fresh fuel 7C Empty

< 5 wt%storage U-235,cell

> 27 GWd/MTU C "B C A*:AC B C MB A *0 C 0 A- A CB CB Irradiated fuel at, or above, checkerboard curve B< 5 wt% U-235 fresh fuel W

7_< Empty 5 wt%storage U-235, cell

> 27 GWd/MTU Figure 5.6-4 Examples of Contiguous Fresh Fuel and Irradiated Fuel Checkerboards Which Meet Interface Requirements WATERFORD-UNIT 3 .5-6d AMENDMENTNO.

to W3F1-2008-0052 Page 10 of 10 ADD TS 3/ 4 9.12 3/4.9.12 SPENT FUEL POOL (SFP) BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.9.12 The spent fuel pool boron concentration shall be > 1900 ppm.

APPLICABILITY: When fuel assemblies are stored in the SFP ACTION:

a. With the spent fuel pool boron concentration not within limits immediately suspend movement of fuel in the SFP and immediately initiate actions to restore boron concentration to within limits.

b. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.12 Verify the spent fuel pool concentration is within limits once per 7 days.

3/4 9 -12

Attachment 4 W3F1-2008-0052 Affidavit for Proprietary Information

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk AFFIDAVIT PURSUANT TO 10 CFR 2.390 I, Evan Rosenbaum, being duly sworn, depose and state as follows:

(1) I am the Holtec International Wet Storage Technical Lead for the Waterford Steam Electric Station (WSES) Unit 3 Criticality Analysis Project and 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 Revision 0 of Holtec Report HI-2084014 containing 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)(1) 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 of FOIA 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).

1 of 5

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk 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. Infonmation 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 paragraphs 4.a and 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 2 of 5

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk AFFIDAVIT PURSUANT TO 10 CFR 2.390 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 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 3 of 5

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk AFFIDAVIT PURSUANT TO 10 CFR 2.390 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.

(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.

4 of 5

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk AFFIDAVIT PURSUANT TO 10 CFR 2.390 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.

STATE OF NEW JERSEY s

) ss "

COUNTY OF BURLINGTON)

Mr. Evan Rosenbaum, 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 Marlton, New Jersey, this 2 0th day of August, 2008..

Evan Rosenbaum, P.E.

Holtec International Subscribed and sworn before me this 0 day of 312008.

MARIA C MASS1 fNOTAY PUBLIC OF NEW.JERSEY Aprl 25,2010 My Commission Expires

Attachment 5 W3F1-2008-0052 Licensing Report for W3 NGF Criticality Analysis (Non Proprietary Information)

Eu...

H OLTEC Holtec Center, 555 Lincoln Drive West, Marlton, NJ 08053 Telephone (856) 797- 0900 Fax (856) 797 - 0909 INTERNATIONAL U

LICENSING REPORT FOR WA TERFORD UNIT 3 SPENT FUEL POOL CRITICALITY ANAL YSIS FOR ENTERGY Holtec Report No: HI-2084014 Holtec Project No: 1712 Report Class : SAFETY RELATED

-. , OMPANY.,PRIVATE:,, .

Th *is doc9menýjs C .. the.pr'operty ofHoltecinternati'onal and its-Client. It is to be usedonly in oifinectionvith'the performnce'of c Internati oItnal or its deosigntedsUcontractors.

Reproduction publication or repr ier pupose.by any party other than th Cient is expressly forbidden.y-~*,

HOLTEC INTERNATIONAL DOCUMENT ISSUANCE AND REVISION STATUS' DOCUM..ENT N... AME. I:CEMTNSING REPORT F

.':i REVATEIRFOD UNIT'3 SPANTU1i:(i: "IL POOL.

DOCU MENTNO.M:E: CRITICALITY ANALYSIS JI\1I-2O84Ol4. CATEGORY": U GEERIC PROJ ECTNO.: PROJ). ECTPCII p

1:712 Rev. Date, ALuthor"S No.7 A\rovedL Initials V IR T'17/ i K,,rickncr 8892 Q DOCUMENT CATEGORIZATION In.accordance. itlh thce 1oltecQtil i A\ssuran0e Manual and assdciated Iiloltec! Quality4 PI-ocedi-1eS

(,I*IQ)S). this d0iiCi1t11011** -i,&

it. i- ed aZIN  :

5 CalCulation Package 3 (Per I1HQP 3.2) .jTcchnlieal Rep~ort (PIer IHQP -3.2)

(Su-'Lch ISa Ijiesn eot 5 Design Criterion Documen10t (Per.HQI 1'3A4) 5 Desfign "Specification1 ('P&HQP1 1 3.4)

l -other (Spc.iiýy,):

.... DOCUMENT FORMATTING The irntt of ilic coiiicnt's ofithis document is in ~acc:ordOnce \with:thC iIIstruLCtikonS of I1I)[' 1.2 or-33.4 except as noted belowý:

DECLARATION OF PROPRIETARY STATUS I~Nopopita-v 5 Hotc roretr 5 '-rivIegeCd In~teleCC E]

This document contains extremely valu'able intellectual property of Holtec International. Holtec's nmettiods, models, and p-,recepts described in this documnent are prOtected againlSt unlauthorized any other party underth~e U.S.. and ilnternational intellectual property laws. U~nauthorized dissem do'cumenrt by the reicipient will tie' deem 1ed. to constitute aIwillfufl breach of contract governing thi,,

this dAocume.n.t bears sole responsibilit~y to honor. Holtec's unabridgeýd ownership rights of this do an tolmt s to the purpose for which it was delivered to the r~ecipipnt. Portionw confdentalit, subiect to.cdbnvriah. rtetio -,a .inst unau.tIhorized re.oroduction by a third oartv..

Summary of Revisions:

Revision 0: Original Issue Project No. 1,712 Report No. HI-2084014 Pagei Shaded Areas Indicate Where Proprietary Information Has Been Removed

Table of Contents

1. INTRODUCTION ................................................................................. 3

2. METHODOLOGY ................................................................................ 4

2. CRITICALITY ANALYSIS.................................................................................. 4 2.2 BORON'DILUTION ACCIDENT.......................................................................... 5

3. ACCEPTANCE CRITERIA ..................................................................... 6

4. ASSUMPTIONS ...................... e.............................................................. 7

5. INPUT DATA ...................................................................................... 8 5.1 FUEL ASSEMBLY SPECIFICATION......................................................................8 5.2 CORE OPERATING PARAMETERS .......................  :............................................... 8 5.3 AxIAL BURNUP DISTRIBUTION ........................................................................ 8 5.4 BURNABLE ABSORBERS ................................................................................ 8 5.5 STORAGE RACK SPECIFICATION ........... .......................................................... 8 5.5.1 Region I Style Storage Racks ................................................................ 9 5.5.2 Region 2 Style Storage Racks ................................................................ 9 5.5.3 Rack Interfaces................................................................................. 9 5.6 ADDITIONAL CALCULATIONS..........................................................................9 5.6.1 Fuel Transfer Carriage Criticality............................................................ 9 5.6.2 Upender Criticality ........................ .................................................. 10 5.6.3 New Fuel Elevator Criticality ............................................................. *...

10 5.6.4 Boron Dilution Accident Evaluation ....................................................... 10 5.6.5 Temporary Storage Racks .. ................................................................. 10 5.6.6 Fuel Pin Storage Container.................................................................. 10 5.6.7 New Fuel Storage Vault......................................................................1I

6. COMPUTER CODES ........................................................................... 11

7. ANALYSIS .................................I.............................. o......................... 12 7.1 REGION I....................................................................................... ........ 12 7.1 .1 Identification of Reference Fuel Assembly .............................................. 12 7.1.2 Eccentric Fuel Assembly Positioning..................................................... 13 7.1.3 Uncertainties Due to Manufacturing Tolerances ........................................ 13 7.1.4 Temperature and Water Density Effects ........................................... I........14 7.1.5 Calculation of Maximum kff.............................................................. 14 7.1 .6 Abnormal and Accident Conditions ...................................................... 15 7.2 REGION 2............................................................................... 16 7.2.1 Identification of Reference Fuel Assembly .............................................. 16 7.2.2 Reactivity Effect of Burnable Absorbers During Depletion............................ 17 7.2.3 Reactivity Effect of Axial Burnup Distribution.......................................... 17 7.2.4 Isotopic Compositions...................................................................... 17 7.2.5 Uncertainty in Depletion Calculations.................................................... 18 Project No. 1712 Report No. HI-2084014 Page I Shaded Areas Indicate Where Proprietary Information Has Been Removed

7.2.6 Eccentric Fuel Assembly Positioning ................................................................... 18 7.2.7 Uncertainties Due to Manufacturing Tolerances .................................................. 18 7.2.8 Temperature and Water Density Effects .............................................................. 19 7.2.9 Calculation of Maximum keff ............................................................................... 20 7.2.10 Abnormal and Accident Conditions ...................................................................... 20 7.3 INTERFACES WITHIN AND BETWEEN RACKS ...................................................................... 21 7.3.1 Gaps Between Region I Racks ............................................................................ 22 7.3.2 Gaps Between Region 2 Racks ..................................... 22 7.3.3 Gaps Between Region I and Region 2 Racks ....................................................... 22 7.3.4 Patterns Within Region 2 Racks .......................................................................... 22 7.4 ADDITIONAL CALCULATIONS ....................................................................................... 23 7.4.1 Fuel Transfer Carriage Criticality ....................................................................... .23 7.4.2 U pender C riticality .............................................................................................. 23 7.4.3 New Fuel Elevator Criticality ............................................................................. 23 7.4.4 Boron Dilution Accident Evaluation ................................................................... .23 7.4.4.1 Low Flow Rate Dilution ........................................................................................ 24 7.4.4.2 High Flow Rate Dilution ..................................................................................... 24 7.4.5 Temporary Storage Racks ......................................... 25 7.4.6 Fuel Pin Storage Container ............................................................................... 25 7.4.7 New Fuel Storage Vault ....................................................................................... 25 REFERENCES ............................................................................................................................. 26 APPENDIX A: Benchmark Calculations ............................................................ A-1 Project No. 1712 Report No. HI-2084014 Page 2 Shaded Areas Indicate Where Proprietary Information Has Been Removed

1. INTRODUCTION This report documents the criticality safety evaluation for the storage of Standard and Next Generation Fuel (NGF) assemblies in Holtec Region 1 & 2 style high-density spent fuel storage racks (SFSRs) at the Waterford Unit 3 nuclear power plant operated by Entergy Nuclear. The purpose of the present analysis is to re-perform the original criticality analysis, taking credit for soluble boron, in order to qualify the racks, etc. for the storage and handling of fuel assemblies having new fuel parameters.

Additional calculations are also documented such as the criticality analysis for storing fuel with an initial enrichment of up to 5.0 wt% 235U in the Reactor Building Temporary Storage Rack (TSR) and storing fuel rods with an initial enrichment of up to 5.0 wt% 235U in the Fuel Pin Storage Container (FPSC) in the spent fuel pool, a boron dilution analysis of the spent fuel pool, a criticality analysis of additional spent fuel pool equipment and also the New Fuel Storage Vault (NFV) (See Section 5.6).

The results of the Region 1 calculations are summarized in Table 7.1 through Table 7.6. The calculations demonstrate that maximum keff is less than 1.0 without credit for soluble boron and less than or equal to 0.95 with 61 ppm soluble boron. Furthermore, all reactivity effects of abnormal and accident conditions have also been evaluated to assure that under all credible abnormal and accident conditions, the reactivity will not exceed the regulatory limit of 0.95 with 170 ppm soluble boron present.

The results of the Region 2 calculations are summarized in Table 7.7 through Table 7.22, and Table 7.26 through Table 7.27. Under normal conditions, a soluble boron concentration of 447 ppm is required in the spent fuel pool. Under credible accident conditions, a soluble boron concentration of 838 ppm is required (see Table 7.21).

Three loading patterns have been qualified for the Region 2 racks (See Tables 7.16 through Table 7.20):

" a uniform loading of spent fuel meeting the burnup versus enrichment requirements of Table 7.26,

• a checkerboard of high and low reactivity fuel (i.e., spent fuel checkerboard). The high reactivity fuel assembly must have an enrichment no greater than 5.0 wt% 235 U and a bumup greater than 27 GWD/MTU and the low reactivity fuel must meet the bumup versus enrichment requirements of Table 7.27,

• a checkerboard of fresh fuel up to 5.0 wt% 235U and empty cell locations (i.e., fresh fuel checkerboard).

Within Region 2 racks, several interfaces are possible with the three loading patterns qualified for storage. The permissible interface conditions are summarized as follows:

• No restrictions are necessary between the uniform loading pattern and either of the checkerboard loading patterns (fresh or spent).

Project No. 1712 Report No. HI-2084014 Page 3 Shaded Areas Indicate Where Proprietary Information Has Been Removed

• For interfaces between a fresh fuel checkerboard and spent fuel checkerboard, the high reactivity spent fuel assembly (5.0 wt% 23 .U, 27 GWD/MTU) may be face adjacent to no, more than one fresh fuel assembly. The fresh fuel assembly may be face adjacent with up to 2 high reactivity spent fuel assemblies. Figure 7.4 shows one example of an acceptable 3x3 fresh fuel checkerboard within the center of a spent fuel checkerboard that meets these requirements.

2. METHODOLOGY 2.1 Criticality Analysis The principal method for the criticality analysis of the high-density storage racks is the use of the three-dimensional Monte Carlo code MCNP4a [2]. MCNP4a is a continuous energy three-dimensional Monte Carlo code developed at the Los Alamos National Laboratory. MCNP4a was selected because it has been used previously and verified for criticality analyses and has all of the necessary features for this analysis. MCNP4a calculations used continuous energy cross-section data predominantly based on ENDF/B-V and ENDF/B-VI. Exceptions are two lumped fission products calculated by the CASMO-4 depletion code, which do not have corresponding cross sections in MCNP4a. For these isotopes, the CASMO-4 cross sections are used in MCNP4a. This approach has been validated in [3] by showing that the cross sections result in the same reactivity effect in both CASMO-4 and MCNP4a.

Benchmark calculations, presented in Appendix A, indicate a bias of 0.0009 with an uncertainty of+/-

0.0011 for MCNP4a, evaluated with a 95% probability at the 95% confidence level [1]. The calculations for this analysis utilize the same computer platform and cross-section libraries used for the benchmark calculations discussed in Appendix A.

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. The MCNP4a criticality output contains a great .deal of useful information that may be used to determine the acceptability of the problem convergence. This information has been used in parametric studies to develop appropriate values for the aforementioned criticality parameters to be used in storage rack criticality calculations. Based on these studies, a minimum of 10,000 histories were simulated per cycle, a minimum of 50 cycles *were skipped before averaging, a minimum of 100 cycles were accumulated, and the initial source was usually specified as uniform over the fueled regions (assemblies). Further, the output was reviewed to ensure that, each calculation achieved acceptable convergence. These parameters represent an acceptable compromise between calculational precision and computational time.

Fuel depletion analyses during core operation were performed with CASMO-4 (using the 70-group cross-section library), a two-dimensional multigroup transport theory code based on the Method of Characteristics [4-6]. Detailed neutron energy spectra for each rod type are obtained in collision probability micro-group calculations for use in the condensation of the cross sections. CASMO-4 is used to determine the isotopic composition of the spent fuel. In addition, the CASMO-4 Project No. 1712 Report No. HI-2084014 Page 4 Shaded Areas Indicate Where Proprietary Information Has Been Removed

calculations are restarted in the storage rack geometry, yielding the two-dimensional infinite multiplication factor (kinf) for the storage rack to determine the reactivity effect of fuel and rack tolerances, temperature variation, and to perform various studies. For all calculations in the spent fuel pool racks, the Xe-135 concentration in the fuel is conservatively set to zero.

The maximum kefr is determined from the MCNP4a calculated k'ff, the calculational bias, the temperature bias, and the applicable uncertainties and tolerances (bias uncertainty, calculational uncertainty, rack tolerances, fuel tolerances, depletion uncertainty) using the following formula:

2 2 Max kff = Calculated kerr + biases + [Yi (Uncertainty) ]"1 In the geometric models used for the calculations, each fuel rod and its cladding were described explicitly, and reflecting or periodic boundary conditions were used in the radial direction which has the effect of creating an infinite radial array of storage cells, except for the assessment of certain accident conditions.

2.2 Boron Dilution Accident The methodology related to the Boron Dilution accident follows the general equation for boron dilution which is, F

C, = Coe v where Ct = boron concentration at time t, CO = initial boron concentration, V = volume of water in the pool, and F = flow rate of un-borated water into the pool This equation conservatively assumes the un-borated water flowing into the pool mixes instantaneously with the water in the pool.

For convenience, the above equation may be re-arranged to permit calculating the time required to dilute the soluble boron from its initial concentration to a specified minimum concentration, which is given below.

V t=--In(C /C,)

F If V is expressed in gallons and F in gallons per minute (gpm), the time, t, will be in minutes.

Project No. 1712 Report No. HI-2084014 Page 5 Shaded Areas Indicate Where Proprietary Information Has Been Removed

3. ACCEPTANCE CRITERIA The high-density spent fuel PWR storage racks for Waterford Unit 3 are designed in accordance with the applicable codes and standards listed below. The objective of this evaluation is to show that the effective neutron multiplication factor, keff, is less than 1.0 with the racks fully loaded with fuel of the highest anticipated reactivity, and flooded with un-borated water at a temperature corresponding to the highest reactivity. In addition, it is to be demonstrated that kff is less than or. equal to 0.95 with the racks fully loaded with fuel of the highest anticipated reactivity, and flooded with borated water at a temperature corresponding to the highest reactivi-ty. The maximum calculated reactivity includes a margin for uncertainty in reactivity calculations including manufacturing tolerances and is shown to be less than 0.95 with a 95%

probability at a 95% confidence level [1]. Reactivity effects of abnormal and accident conditions have also been evaluated to assure that under all credible abnormal and accident conditions, the reactivity will not exceed the regulatory limit of 0.95 under borated conditions.

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

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

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

• USNRC letter of April 14, 1978, to all Power Reactor Licensees - OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications (GL-78-01 1),

including modification letter dated January 18, 1979 (GL-79-004).

• 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.

" USNRC Regulatory Guide 1.13, Spent Fuel Storage Facility Design Basis, Rev. 2, March 2007.

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

• Code of Federal Regulations, Title 10, Part 50, Section 68, "Criticality Accident Requirements."

The New Fuel Storage Vault is intended for the receipt and storage of fresh fuel under normally dry conditions where the reactivity is very low. To assure criticality safety under accident conditions and to conform to the requirements of 10 CFR 50.68, these two accident condition criteria must be met:

Project No. 1712 Report No. HI-2084014 Page 6 Shaded Areas Indicate Where Proprietary Information Has Been Removed

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

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

These criteria preclude a secondary accident per ANSI 8.1 or accidents under dry conditions.

4. ASSUMPTIONS To assure the true reactivity will always be less than the calculated reactivity, the following conservative design criteria and assumptions were employed:

1) Moderator is borated or un-borated water at a temperature in the operating range that results in the highest reactivity, as determined by the analysis.

2) Neutron absorption in minor structural members is neglected, i.e., spacer grids are replaced by water.

3) The effective multiplication factor of an infinite radial array of fuel assemblies was used in the analyses, except for the assessment of certain abnormal/accident conditions and conditions where leakage is inherent.

4) The neutron absorber length is modeled to be the same length as the active region of the fuel.

5) No cooling time is credited in the rack calculations.

6) The presence of burnable absorbers in fresh fuel is neglected. This is conservative.as burnable absorbers would reduce the reactivity of the fresh fuel assembly.

7) The presence of annular pellets is neglected. This is conservative as it is bounded by the solid fuel.

8) All structural materials of the new fuel storage racks are conservatively neglected and replaced with water at the appropriate density.

9) The concrete wall of the transfer canal is conservatively modeled as 100 cm thick.

10) The FPSC tubes holes were not modeled; however, the other steel structures of the FPSC were modeled as water. Therefore, the neglecting of the tube holes is conservative.

11) The concrete walls of the vault are 'Conservatively modeled as 100 cm thick.

Project No. 1712 Report No. HI-2084014 . Page 7 Shaded Areas Indicate Where Proprietary Information Has Been Removed

12) The two inch redwood planks in the NFV are assumed to be 1.5 inches thick.

13) In MCNP4a, the Doppler treatment and cross-sections are valid at 300K (80.33 'F);

however, in the NFV calculations no temperature bias is applied to the results to account for the actual temperature of the water.

14) In the NFV the eccentric fuel positioning condition is covered by the fuel cell spacing tolerance.

5. INPUT DATA 5.1 Fuel Assembly Specification The spent fuel storage racks are designed to accommodate various 16x16 fuel assemblies used at the Waterford Unit 3 facility. The design specifications for these fuel assemblies, which were used for this analysis, are given in Table 5.1.

5.2 Core Operating Parameters Core operating parameters are rlecessary for fuel depletion calculations performed with CASMO-4. The core parameters used for the depletion calculations are presented in Table 5.2.

Temperature and soluble boron values are taken as the upper bound (most conservative) of the core operating parameters of Waterford Unit 3. The neutron spectrum is hardened by each of these parameters, leading to a greater production of plutonium during depletion, which results in conservative reactivity values.

5.3 Axial Bumup Distribution Generic axial burnup profiles provided by the client are specified at node centers for 24 equally-spaced axial sections for burnups of less than 25 GWD/MTU and greater than 25 GWD/MTU.

The resulting profiles are presented in Table 5.3.

5.4 Burnable Absorbers At the Waterford Unit 3 facility there is the potential for either B4 C, erbia or IFBA burnable absorbers to be located in the fuel assembly as integral absorbers. In [10] it is clearly seen that the reactivity of the fuel assembly with IFBA bound those with B 4 C or erbiaand therefore only the IFBA is considered in this analysis. The design specifications for the IFBA rods are given in Table 5.1 and are further discussed in Section 7.2.2.

5.5 Storage Rack Specification Project No. 1712 Report No. HI-2084014 Page 8 Shaded Areas Indicate Where Proprietary Information Has Been Removed

The storage cell characteristics are summarized in Table 5.4.

5.5.1 Region I Style Storage Racks The Region I storage cells are composed of stainless steel boxes separated by a water gap, with fixed neutron absorber panels centered on each side. The steel walls define the storage cells, and stainless steel sheathing supports the neutron absorber panel and defines the boundary of the flux-trap water-gap used to augment reactivity control. Stainless steel channels connect the storage cells in a rigid structure and define the flux-trap between the neutron absorber panels.

Neutron absorber panels are installed on all exterior walls facing other racks.

The calculational models consist of a single cell with reflective boundary conditions through-the centerline of the water gaps, thus simulating an infinite array of Region I storage cells. Figure 5.1 shows the actual calculational model containing the reference 16x16 assembly, as drawn by the two-dimensional plotter in MCNP4a. The calculations are described in Section 7.1.

5.5.2 Region 2 Style Storage Racks The Region 2 storage cells are composed of stainless steel boxes with a single fixed neutron absorber panel, (attached by stainless steel sheathing) centered on each side. The stainless steel boxes are arranged in an alternating pattern such that the connection of the box corners form storage cells between those of the stainless steel boxes.

The calculational models consist of a group of four identical cells surrounded by reflective boundary conditions through the centerline of the composite of materials between the cells, thus simulating an infinite array of Region 2 storage cells. Figure 5.2 shows the actual calculational model containing the 16xI 6 assembly as drawn by the two-dimensional plotter in MCNP4a.

5.5.3 Rack Interfaces Based on the layout of the spent fuel pool, there are no Region I to Region 2 interfaces. The gap between adjacent Region 2 racks is conservatively neglected. The Region 2 to Region 2 rack loading pattern interfaces are analyzed in Section 7.3.

5.6 Additional Calculations 5.6.1 Fuel Transfer Carriage Criticality The fuel transfer carriage conveys the fuel assemblies through the fuel transfer tube and is capable of accommodating two fuel assemblies at a time, carried in stainless steel boxes. The results of this calculation can be found in Section 7.4.1.

Project No. 1712 Report No. HI-2084014 Page 9 Shaded Areas Indicate Where Proprietary Information Has Been Removed

5.6.2 Upender Criticality The fuel upender is a machine located at each end of the transfer tube. The criticality of this component is bounded by the fuel transfer carriage. No input required. See Section 7.4.2.

5.6.3 New Fuel Elevator Criticality The new fuel elevator has a capacity of a single fuel assembly and is utilized to lower new fuel from the operating level of the fuel handling building to the bottom of the spent fuel pool. See Section 7.4.3.

5.6.4 Boron Dilution Accident Evaluation The spent fuel pool at Waterford Unit 3 has a minimum soluble boron concentration of 1720 ppm. The spent fuel pool volume is considered to be 38,600 ft3 . Under certain abnormal conditions, un-borated water may dilute this concentration below the requirements determined in Section 7.

Makeup to the spent fuel storage pool is from the Refueling Water Storage Pool and/or the Condensate Storage Pool. Makeup from the Refueling Water Storage Pool is provided by the refueling water pool purification pump which has a capacity of 150 gpm. The Refueling Water Storage Pool has a minimum boron concentration of 2050 ppm. The component cooling water makeup pumps provide makeup from the Condensate Storage Pool and have a capacity of 600 gpm. For the accident case a high flow rate of 600 gpm is therefore assumed. The results of these calculations are shown in Section 7.4.4.

5.6.5 Temporary Storage Racks The TSR storage cell locations are arranged in a row of 5 cells with the geometric dimensions in Table 5.5. The design basis calculational model places 5 fresh fuel assemblies enriched to 5.0 wt% 235 U in the storage rack. No steel structural material is included. For simplification, the following tolerances are included in the design basis model: fuel density, lattice pitch and enrichment.

5.6.6 Fuel Pin Storage Container The FPSC is a square stainless steel container that fits in a fuel assembly storage rack in the spent fuel pool. It has 81 stainless steel tubes that may contain fuel rods of up to 5.0 wt% 235U (See Table 5.5). The FPSC was modeled as 81 solid steel tubes of equal diameter, each containing 1 fresh fuel rod with the maximum enrichment. All other steel components of the container were neglected. The model includes 100 cm of water surrounding the FPSC or fuel assembly.

The criticality analysis of the FRSC is performed by comparing the reactivity of the FRSC loaded with the maximum number of fresh fuel pins to the reactivity of various fuel assemblies Project No. 1712 Report No. HI-2084014 Page 10 Shaded Areas Indicate Where Proprietary Information Has Been Removed

and determine which cases bound the FRSC. These calculations are performed with the fuel assembly surrounded by 100 cm of water, meaning no storage racks, poison material or structural materials are considered (the steel tubes of the FRSC are modeled). No tolerances are included. Reflective boundary conditions are applied on all sides to maximize reactivity.

5.6.7 New Fuel Storage Vault The NGF assembly is the only fuel assembly type to be stored in the NFV. The design input data is tabulated in Table 5.1 and Table 5.6. The storage locations are arranged in 8 modules providing a total of 16 rows of 5 cells each for a total of 80 storage locations. The cells are located on a 21 inch pitch within each module, and on a 49 inch cell center to cell center spacing between modules in the east-west direction and a 58 inch cell center to cell center spacing between modules in the north-south direction. Normally, fuel is stored in the dry condition with very low reactivity. Graphic representations of the analytical model are shown in Figure 7.5 and 7.6. These figures were drawn (to scale) with a two-dimensional plotter.

The reactivity uncertainties associated with various manufacturing tolerances for the NFV were calculated by the difference between two MCNP4a calculations, one with the nominal value and a second independent calculation with the tolerance parameter changed. Based on the nominal condition results, it was determined that the 100% moderator condition, i.e. 1.0 g/cc, represented the maximum reactivity condition and therefore the tolerance calculations were performed with 100% moderator density. These tolerance effects each include the combination of statistical errors in the MCNP4a calculations due to the random nature of Monte Carlo calculations, at the 95% confidence level (Ak+(.2)*2*a). In evaluating the uncertainties due to tolerances, the following tolerances were used:

• Enrichment Tolerance of+ 0.05 3 wt% 235U

• Density of +/-0.165 g U0 2/cm

• Fuel Storage Cell Spacing of +/-0.8125 in.

The fuel storage cell spacing tolerance was only used in the 21 inch assembly pitch. In determining the maximum kn', the effects of these manufacturing tolerances were statistically combined (square root of the sum of the squares) with the MCNP4a bias uncertainty from the benchmarking results and the MCNP4a calculational statistics (2*a) to determine the total uncertainty.

6. COMPUTER CODES The following computer codes were used during this analysis.

MCNP4a [2] is a three-dimensional continuous energy Monte Carlo code developed at Los Alamos National Laboratory. This code offers the capability of performing full three-dimensional calculations for the loaded storage racks. MCNP4a was run on the PCs at Holtec.

Project No. 1712 Report No. HI-2084014 Page 11 Shaded Areas Indicate Where Proprietary Information Has Been Removed 2

• CASMO-4, Version 2.05.14 [4-6] is a two-dimensional multigroup transport theory code developed by Studsvik Scandpower, Inc. CASMO-4 performs cell criticality calculations and burnup. CASMO-4 has the capability of analytically'restarting burned fuel assemblies in the rack configuration. This code was used to determine the reactivity effects of tolerances and fuel depletion.

7. ANALYSIS This section describes the calculations that were used to determine the acceptable storage criteria

• for the Region I and Region 2 style racks. In addition, this section discusses the possible abnormal and accident conditions.

Unless otherwise stated, all calculations assumed nominal characteristics for the fuel and the fuel storage cells. The effect of the manufacturing tolerances is accounted for with a reactivity adjustment as discussed below.

As discussed in Section 2, MCNP4a was the primary code used in the PWR calculations.

CASMO-4 was used to determine the reactivity effect of- tolerances and for depletion calculations. MCNP4a'was used for reference cases and to perform calculations which are not possible with CASMO-4 (e.g., eccentric fuel positioning, axial burnup distributions, and fuel misloading).

Figures 5.1 and 5.2 are pictures of the basic calculational models used in MCNP4a. These pictures were created with the two-dimensional plotter in MCNP4a and clearly indicate the explicit modeling of fuel rods in each fuel assembly. In CASMO-4, a single cell is modeled, and since CASMO-4 is a two-dimensional code, the fuel assembly hardware above and below the active fuel length is not represented. The three-dimensional MCNP4a models that included axial leakage assumed approximately 30 cm of water above and below the active fuel length.

Additional models with more storage cells were generated with MCNP4a to investigate the effect of abnormal and normal conditions. These models are discussed in the appropriate section.

7.1 Region I The goal of the criticality calculations for the Region 1 style racks is to qualify the racks for storage of fuel assemblies with design specifications as shown in Table 5.1 and a maximum nominal initial enrichment of 5.0 wt% 2 35 U.

7.1.1 Identification of Reference Fuel Assembly CASMO-4 calculations were performed to determine which of the two assembly types in Table 5.1 is bounding in the Region 1 racks. The presence of burnable absorbers in the fuel assembly (IFBA) was neglected for determination of the reference fuel assembly. The results in Table 7.1 Project No. 1712 Report No. HI-2084014 Page 12 Shaded Areas Indicate Where Proprietary Information Has Been Removed

shows that the NGF assembly has the highest reactivity and this assembly type is therefore used in all subsequent calculations.

7.1.2 Eccentric Fuel Assembly Positioning The fuel assemblies are assumed to be normally located in the center of the storage rack cell. To investigate the potential reactivity effect of eccentric positioning of assemblies in the cells, MCNP4a calculations were performed with the fuel assemblies assumed to be in the corner of the storage rack cell (four-assembly cluster at closest approach). The highest reactivity, therefore, corresponds to the reference design with the fuel assemblies positioned in the center of the storage cells. The results of this calculation is shown in Table 7.6.

7.1.3 Uncertainties Due to Manufacturing Tolerances In the calculation of the final kif, the effect of manufacturing tolerances on reactivity must be included. CASMO-4 was 'used to perform these calculations. As allowed in [7], the methodology employed to calculate the tolerance effects combine both the worst-case bounding value and sensitivity study approaches. The evaluations include tolerances of the rack and fuel dimensions. As for the bounding assembly, calculations are performed at an enrichment of 5.0 wt% 235U. The reference condition is the condition with nominal dimensions and properties. To determine the Ak associated with a specific manufacturing tolerance, the kinf calculated for the reference condition is compared to the kinf from a calculation with the tolerance included. Note that for the individual, parameters associated with a tolerance, no statisticalapproach is utilized.

Instead, the full tolerance value is utilized to determine the maximum reactivity effect. All of the Ak values from the various tolerances are statistically combined (square root of the sum of the squares) to determine the final reactivity allowance for manufacturing tolerances. The fuel and rack tolerances included in this analysis are described below; the fuel density and enrichment tolerances are typical values:

Fuel Tolerances 3

• Increased Fuel Density: +0.165 g/cm

• Increased Fuel Enrichment: 0.05 wt% 235U

• Fuel Rod Pitch: +0.01 in.

• Fuel Rod Cladding Outside Diameter: +/- 0.00 15 in.

• Fuel Rod Cladding Thickness min: 0.021 in.

• Fuel Pellet Outside Diameter: +/- 0.0005 in.

• Guide Tube Outside Diameter: +/- 0.003 in.

• Guide Tube Thickness min: 0.036 in.

Rack Tolerances

• Cell Inner Dimension:

• Box Wall Thickness: Lii2J

" Cell Pitch: E7 7 1..

r Boral Width:2 R Project No. 1712 Report No. HI-2084014 Page 13 Shaded Areas Indicate Where Proprietary Information Has Been Removed

- Poison Gap min:

• Poison Loading mi: t.

Regarding the tolerance calculations, the following needs to be noted:

In some cases it is not obvious whether an increase or decrease of the parameter will lead to an increase in reactivity. In these cases, the reactivity effect of both increase and decrease of the parameter are calculated, and the positive reactivity effect is used when calculating the statistical combination.

The tolerance in the flux trap is conservatively captured in'the tolerances of the cell ID and cell pitch, since variations of the cell ID are evaluated for a constant cell pitch and vice versa.

• Tolerance calculations were performed for pure water only since the presence of soluble boron in the pool lowers reactivity and reactivity effects of tolerances, and therefore the pure water'case bounds the soluble boron case.

The results of the calculations of the manufacturing tolerances are presented in Table 7.2.

7.1.4 Temperature and Water Density Effects Pool water temperature effects on reactivity in the Region 1 racks have been calculated with CASMO-4 for various enrichments with a maximum value of 5.0 wt% 235 U and the results are presented in Table 7.3. The results show that the Region 1 spent fuel pool temperature coefficient of reactivity is negative, i.e., a lower temperature results in a higher reactivity.

Consequently, the design basis calculations are evaluated at 0 'C (32 'F) for normal conditions.

In MCNP4a, the Doppler treatment and cross-sections are valid only at 300K (80.33 'F).

Therefore, a Ak is determined in CASMO-4 from 32 'F to 80.33 'F, and is included in the final kerr calculation as a bias. Table 7.3 shows the calculation of the bias. The temperature bias is calculated with pure water.

7.1.5 Calculation of Maximum keff Using the calculational model shown in Figure 5.1 and the reference 16x16 NGF fuel assemblies, the keff in the Region I storage racks has been calculated with MCNP4a. The calculations of the maximum ker values, based on the formula in Section 2, are shown in Table 7.4 and Table 7.5. In summary, the results show that the maximum keff of the Region I racks is less than 1.0 at a 95% probability at a 95% confidence level with no credit for soluble boron, and by linear interpolation, less than or equal to 0.95 with 61 ppm soluble boron.

Project No. 1712 Report No. HI-2084014 Page 14 Shaded Areas Indicate Where Proprietary Information Has Been Removed

7.1.6 Abnormal and Accident Conditions The effects on reactivity of credible abnormal and accident conditions are examined in this section. This section identifies which of the credible abnormal or accident conditions will result in exceeding the limiting reactivity (kerr -<0.95). For those accident or abnormal conditions that result in exceeding the limiting reactivity, a minimum soluble boron concentration is determined to ensure that keff < 0.95. The double contingency principal of ANS-8.I/N16.1-1975 [8] (and the USNRC letter of April 1978; see Section 3.0) specifies that it shall require at least two unlikely, independent and concurrent events to produce a criticality accident. This principle precludes the necessity of considering the simultaneous occurrence of multiple accident conditions.

7.1.6.1 Abnormal Temperature All calculations for Region 1 are performed at a pool temperature of 32°F. As -shown in Section 7.1.4 above, the temperature coefficient of reactivity is negative, therefore any increase in temperature above 32'F would cause a reduction in the reactivity. Therefore, no further evaluations of abnormal temperatures are performed.

7.1.6.2 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 too of the rack with a minimum separation distance from the active fuel region of more than 12 inches, 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, the soluble boron in the spent fuel pool water assures that the true reactivity is always less than the limiting value for this dropped fuel accident.

7.1.6.3 Dropped Assembly - Vertical Into Fuel Cell It is also possible to vertically drop an assembly into a location that might be occupied by another assembly or that might be empty. Such a vertical impact onto another assembly has previously been shown to cause no damage to either fuel assembly. A vertical drop into an empty storage cell could result in a small deformation of the baseplate. The resultant effect would be the lowering of a single fuel assembly by the amount of the deformation. This could potentially result in further misalignment between the active fuel region and the Boral. However, the amount of deformation for this drop would be small and restricted to a localized area of the rack around the storage cell where the drop occurs. Furthermore, the soluble boron in the spent fuel pool water assures that the true reactivity is always less than the limiting value for this dropped fuel accident.

Project No. 1712 Report No. HI-2084014 Page 15 Shaded Areas Indicate Where Proprietary Information Has Been Removed

7.1.6.4 Abnormal Location of a Fuel Assembly 7.1.6.4.1 Misloaded Fresh Fuel Assembly The Region I racks are qualified for the storage of fresh, unburned fuel assemblies with the maximum permissible enrichment (5.0 wt% 235U). Therefore, the abnormal location of a fuel assembly within normal Region 1 cells is of no concern.

7.1.6.4.2 Mislocated Fresh Fuel Assembly The mislocation of a fresh unburned fuel assembly could, in the absence of soluble poison, result in exceeding the regulatory limit (keff of 0.95). This could possibly occur if a fresh fuel assembly of the highest permissible enrichment (5.0 wt% 235U) were to be accidentally mislocated outside of a storage rack adjacent to other fuel assemblies. The results of the analysis are 9hown in Table 7.6 and show by linear interpolation that a soluble boron level of 170 ppm is sufficient to ensure that the maximum kff value for this condition remains at or below 0.95 7.2 Region 2 The goal of the criticality calculations for the Region 2 style racks is to qualify the racks for storage of fuel assemblies with design specifications as shown in Table 5.1 and a maximum nominal initial enrichment of 5.0 wt% 235U. Specifically, the purpose of the criticality calculations is to determine the initial enrichment and burnup combinations required for the storage of spent fuel assemblies with nominal initial enrichments up to 5.0 wt% 235 U. Three loading configurations were analyzed to create bumup versus enrichment curves:

• a uniform loading of spent fuel meeting the burnup versus enrichment requirements of Table 7.26,

• a checkerboard loading pattern of high and low reactivity fuel with the high reactivity fuel at an enrichment of 5.0 wt% 235U and a burnup of 27 GWDiMTU and the low reactivity fuel must meet the burnup versus enrichment requirements of Table 7.27;

" a checkerboard of fresh fuel up to 5.0 wt% 235U and empty cell locations (i.e., fresh fuel checkerboard).

7.2.1 Identification of Reference Fuel Assembly CASMO-4 calculations were performed to determine which of the two assembly types are bounding in the Region 2 racks. In the calculations, the fuel assembly is burned in the cores configuration and restarted in the rack configuration. For all assemblies, the presence of burnable absorbers in the fuel assembly (BPRA, IFBA) was neglected for determination of the reference fuel assembly (see Section 7.2.2 for a discussion the effect of burnable poison). The results are shown in Table 7.7 (selected enrichments and burnups) and show that the NGF assembly has the highest reactivity for all enrichments and burnups relative to the final burnup versus enrichment curve.

Project No. 1712 Report No. HI-2084014 Page 16 Shaded Areas Indicate Where Proprietary Information Has Been Removed

7.2.2 Reactivity Effect of Burnable Absorbers During Depletion The Waterford Unit 3 fuel makes use of burnable absorbers of either B4 C, erbia or integral fuel burnable absorber (IFBA) rods with a thin coating of ZrB2 on the U0 2 pellet.

Generic studies [10] have investigated the effect that integral burnable absorbers (IBAs) have on the reactivity of spent fuel assemblies. These studies have concluded that there is a small positive reactivity effect associated with the presence of IFBA rods, which therefore bounds the negative effects of the B 4C and erbia. Therefore, only the IFBA is considered in this analysis.

To determine the reactivity effect for the Waterford Unit 3 spent fuel racks, depletion calculations were performed for selected configurations of IFBA rods provided by Entergy. The reactivity of the fuel assembly with IFBA rods is compared to the reactivity of the respective fuel assembly without IFBA rods. The results are presented in Table 7.8 and an IFBA bias of 0.0070 is conservatively applied to the final kr to bound all IFBA configurations.

7.2.3 Reactivity Effect of Axial Burnup Distribution Initially, fuel loaded into the reactor will burn with a slightly skewed cosine power distribution.

As burnup progresses, the burnup distribution will tend to flatten, becoming more highly burned in the central regions than in the upper and lower ends. At high burnup, the more reactive fuel near the ends, of the fuel assembly (less than average burnup) occurs in regions of lower reactivity worth due to neutron leakage. Consequently, it would be expected that over most of the burnup history, distributed burnup fuel assemblies would exhibit a slightly lower reactivity than that calculated for the average burnup. As burnup progresses, the distribution, to some extent, tends to be self-regulating as controlled by the axial power distribution, precluding the existence of large regions of significantly reduced burnup.

Generic analytic results of the axial burnup effect for assemblies without axial blankets have been provided by Turner [9] based upon calculated and measured axial burnup distributions.

These analyses confirm the minor and generally negative reactivity effect of the axially distributed burnup compared to a flat distribution, becoming positive at burnups greater than about 30 GWD/MTU. The trends observed in [9] suggest the possibility of a small positive reactivity effect above 30 GWD/MTU, increasing to slightly over 1% Ak at 40 GWD/MTU. The required burnup for the maximum enrichment is higher than 30 GWD/MTU. Therefore, a positive reactivity effect of the axially distributed burnup is possible. Calculations are conservatively performed with the axial burnup distribution shown in Table 5.3 (see Section 5.3) and with an axially constant burnup, and the higher. reactivity is used in the analyses.

7.2.4 Isotopic Compositions To perform the criticality evaluation for spent fuel in MCNP4a, the isotopic composition of the fuel is calculated with the depletion code CASMO-4 and then specified as input data for Project No. 1712 Report No. HI-2084014 Page 17 Shaded Areas Indicate Where Proprietary Information Has Been Removed

MCNP4a. The CASMO-4 calculations performed to,, obtain the isotopic compositions for MCNP4a were performed generically, with one calculation for each enrichment, and burnups in increments of 2.5 GWD/MTU or less. The isotopic composition for any given burnup is then determined by linear interpolation.

7.2.5 Uncertainty in Depletion Calculations Since critical experiment data with spent fuel is not available for determining the uncertainty in burnup-dependent reactivity calculations, an allowance for uncertainty in reactivity was assigned based upon other considerations. Based on the recommendation in [7], a burnup dependent uncertainty in reactivity for burnup calculations of 5% of the reactivity decrement is used. This allowance is statistically combined with the other reactivity allowances in the determination of the maximum keff for normal conditions where assembly bumup is credited.

7.2.6 Eccentric Fuel Assembly Positioning The fuel assembly is assumed to be normally located in the center of the storage rack cell. In the absence of a fixed neutron absorber, the eccentric location of fuel assemblies in the storage cells may produce a positive reactivity effect. Therefore, the eccentric positioning is performed in a very conservative manner in MCNP4a, assuming 4 assemblies in the corners of the storage cell (four-assembly cluster at closest approach), and that these clusters of four assemblies are repeated throughout the rack. The results of these calculations are shown in Table 7.9 and indicate that eccentric fuel positioning results in a decrease in reactivity.

7.2.7 Uncertainties Due to Manufacturing Tolerances I6 the calculation of the final keff, the effect of manufacturing tolerances on reactivity must be included. CASMO-4 was used to perform these calculations. As allowed in [7], the methodology employed to calculate the tolerance effects combine both the worst-case bounding value and sensitivity study approaches. The evaluations include tolerances of the rack and fuel dimensions. As for the bounding assembly, calculations are performed for different enrichments and burnups with a maximum value of 5.0 wt% 235U. The reference condition is the condition with nominal dimensions and properties. To determine the Ak associated with a specific manufacturing tolerance, the kinf calculated for the reference condition is compared to the kif from a calculation with the tolerance included. Note that for the individual parameters associated with a tolerance, no statistical approach is utilized. Instead, the full tolerance value is' utilized to determine the maximum reactivity effect. All of the Ak values from the various tolerances are statistically combined (square root of the sum of the squares) to determine the final reactivity allowance for manufacturing tolerances. Only the Ak values in the positive direction (increasing reactivity) were used in the statistical combination. The fuel and rack tolerances included in this analysis are described below; the fuel density and enrichment tolerances are typical values:

Project No. 1712 Report No. HI-2084014 Page 18 Shaded Areas Indicate Where Proprietary Information Has Been Removed'

Fuel Tolerances 3

• Increased Fuel Density: +0.165 g/cm

• Increased Fuel Enrichment: 0.05 wt% 235U

. Fuel Rod Pitch: +/-0.0I in.

• Fuel Rod Cladding Outside Diameter: +/- 0.0015 in.

• Fuel Rod Cladding Thickness min: 0.021 in.

• Fuel Pellet Outside Diameter: +/- 0.0005 in.

• Guide Tube Outside Diameter: +/- 0.003 in.

• Guide Tube Thickness min: 0.036 in.

Rack Tolerances Cell Inner Dimension ,

• Box Wall Thickness: :"

• Poison Width:

• Poison Gap minimum:

• Boral B-10 Loading min Regarding the tolerance calculations, the following needs to be noted:

• In some cases it is not obvious whether an increase or decrease of the parameter will lead to an increase in reactivity. In these cases, the reactivity effect of both increase and decrease of the parameter are calculated, and the positive reactivity effect is used when calculating the statistical combination.

• In the CASMO-4 model used, the tolerance calculation for the Cell ID resulted in a negative reactivity for both increases and decreases in Cell ID. Conservatively, the least negative value was used as a positive reactivity effect.

• Tolerance calculations were performed for pure water only since the presence of soluble boron in the pool lowers reactivity and reactivity effects of tolerances, and therefore the pure water case bounds the soluble boron case.

7.2.8 Temperature and Water Density Effects Pool water temperature effects on reactivity in the Region 2 racks have been calculated with CASMO-4 for various enrichments with a maximum value of 5.0 wt% 235U and the results are presented in Table 7.12. The results show that the Region 2 spent fuel pool temperature coefficient of reactivity is negative, i.e., a higher temperature results in a lower reactivity.

Consequently, all CASMO-4 calculations are evaluated at 32 OF.

In MCNP4a, the Doppler treatment and cross-sections are valid only at 300K (80.33 OF).

Therefore, a Ak is determined in CASMO-4 from 32 OF to 80.33 OF, and is included in the final kcff calculation as a bias.

Project No. 1712 Report No. HI-2084014 Page 19 Shaded Areas Indicate Where Proprietary Information Has Been Removed

7.2.9 Calculation of Maximum kff Using the calculational model shown in Figure 5.2 and the reference 16x16 NGF fuel assembly, the keff in the Region 2 storage racks has been calculated with MCNP4a for the cases discussed in Section 7.2. The determination of the maximum k~ff values, based on the formula in Section 2, is shown in, for initial enrichments between 2.0 wt% 235U and 5.0 wt% 235U, Table 7.13.for the uniform loading case, Table 7.14 for the spent fuel checkerboard loading case, and Table 7.15 for the fresh fuel checkerboard case. A summary of the calculations for non-accident conditions of the maximum kef for spent fuel of maximum nominal enrichment of 5.0 wt% 235U is shown in Table 7.16 for the uniform loading of spent fuel without soluble boron and Table 7.17 with soluble boron, Table 7.18 for the spent fuel checkerboard without soluble boron and Table 7.19 with soluble boron, and Table 7.20 for the fresh fuel checkerboard fuel. Table 7.26 and Figure 7.1 present the burnup versus enrichment requirements for the uniform loading of spent fuel and Table 7.27 and Figure 7.2 present the burnup versus enrichment requirements for the low reactivity fuel assemblies in the spent fuel checkerboard. The results show that the maximum keff of the Region 2 racks is less than 1.0 at a 95% probability and at a 95% confidence level for the three loading patterns and less than 0.95 at a 95% probability and at a 95% confidence level with 447 ppm soluble boron.

7.2.10 Abnormal and Accident Conditions The effects on reactivity of credible abnormal and accident conditions are examined in this section. This section identifies which of the credible abnormal or accident conditions will result in exceeding the limiting reactivity (keff < 0.95). For those accident or abnormal conditions that, result in exceeding the limiting reactivity, a minimum soluble boron concentration is determined to ensure that k~ff < 0.95. The double contingency principal of ANS-8.1/N 16.1-1975 [8] (and the USNRC letter of April 1978; see Section 3.0) specifies that it shall require at least two unlikely, independent and concurrent events to produce a criticality accident. This principle precludes the necessity of considering the simultaneous occurrence of multiple accident conditions.

7.2.10.1 Abnormal Temperature All calculations for Region 2 are performed at a pool temperature of 32 'F. As shown in Section 7.2.8 above, the temperature coefficient of reactivity is negative, therefore no additional calculations are required.

7.2.10.2 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, 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, the soluble boron in the spent fuel pool Project No. 1712 Report No. H1-2084014 Page 20 Shaded Areas Indicate Where Proprietary Information Has Been Removed

water assures that the true reactivity is always less than the limiting value for this dropped fuel accident.

7.2.10.3 Dropped Assembly - Vertical It is also 'possible to vertically drop an assembly into a location that might be occupied by another assembly or that might be empty. Such a vertical impact onto another assembly has preyiously been shown to cause no damage to either fuel assembly. A vertical drop into an empty storage cell could result in a small deformation of the baseplate. The resultant effect would be the lowering of a single fuel assembly by the amount of the deformation. This could potentially result in further misalignment between the active fuel, region and the Boral. However, the amount of deformation for this drop would be small and restricted to a localized area of the rack around the storage cell where the drop occurs. Furthermore, the reactivity increase would be small compared to the reactivity increase created by the misloading of a fresh assembly discussed in the following section. The vertical drop is therefore bounded by this misloading accident and no separate calculation is performed for the drop accident.

7.2.10.4 Abnormal Location of a Fuel Assembly 7.2.10.4.1 Misloaded Fresh Fuel Assembly The misloading of a'fresh unburned fuel assembly could, in the absence of soluble poison, result in exceeding the regulatory limit (k~ff of 0.95). This could possibly occur if a fresh fuel assembly of the highest permissible enrichment (5.0 wt% 235U) were to be inadvertently misloaded into a storage cell intended to be used for spent fuel. The results of this accident are shown in Table 7.21.

7.2.10.4.2 Mislocated Fresh Fuel Assembly The mislocation of a fresh unburned fuel assembly could, in the absence of soluble poison, result in exceeding the regulatory limit (keff of 0.95). This could possibly occur if a fresh fuel assembly of the highest permissible enrichment (5.0 wt% 235U) were to be accidentally mislocated outside of a Region 2 storage rack adjacent to other fuel assemblies The MCNP4a model consists of an array of Region 2 fuel storage cells with a single fresh, unburned assembly placed adjacent to the rack as close to the rack faces as possible to maximize the possible reactivity effect. The results of the analysis are shown in Table 7.21.

7.3 Interfaces Within and Between Racks The calculations in Sections 7.1 and 7.2 assume laterally infinite arrangements of rack cells. This section evaluates the potential effect of the interfaces between and within rack modules.

Project No. 1712 Report No. HI-2084014 Page 21 Shaded Areas Indicate Where Proprietary Information Has Been Removed

7.3.1 Gaps Between Region 1 Racks Region 1 racks have poison panels on all peripheral walls facing other racks. Furthermore, the assembly distance across the gaps between Region 1 racks is larger than the assembly distance within the racks. Under abnormal conditions, in the event of lateral rack movement, the baseplate extensions will maintain a minimum rack to rack gap that is bounded by the infinite array calculations, and no further evaluations are necessary.

7.3.2 Gaps Between Region 2 Racks Under normal conditions, the assembly distance across the gaps between Region 2 racks is larger than the assembly distance within these racks. Since there is at least one Boral panel between adjacent assemblies for these rack to rack interfaces, the condition in the gap is therefore bounded by the infinite array calculations, and no further evaluations are necessary.

7.3.3 Gaps Between Region 1 and Region 2 Racks According to the data provided by Entergy, Region I and Region 2 are separated by distances that exceed the gaps between racks within either region, and therefore the condition is bounded by the infinite array calculations and no further evaluations are necessary.

7.3.4 Patterns Within Region 2 Racks.

The Region 2 racks are qualified for three types of fuel loading pattern: a uniform loading of spent fuel, a spent fuel checkerboard loading pattern, and a fresh fuel checkerboard loading pattern with empty cells. Within the Region 2 racks, various interfaces between these patterns are qualified. To show that the selected interfaces are acceptable, the following conditions are analyzed:

• An interface between the spent fuel uniform loading pattern and the spent fuel checkerboard. The configuration was chosen so that the high reactivity assembly in the spent fuel checkerboard pattern (5.0 wt%/27 GWD/MTU) is face adjacent to three low reactivity assemblies from the spent fuel checkerboard pattern (see Table 7.22), and face adjacent to 1 assembly meeting the uniform spent fuel requirement (see Table 7.22).

" Two interfaces are evaluated between checkerboards of spent fuel and fresh fuel/empty cells. The bounding case is the case where the fresh fuel assemblies face the high reactivity assembly in the spent fuel checkerboard pattern (5.0 wt%/27 GWD/MTU) on two sides, and has an empty cell on the other two sides. This condition bounds other interfaces between fresh and spent fuel, since the spent fuel with the highest permissible reactivity is used.

Project No. 1712 Report No. HI-2084014 Page 22 Shaded Areas Indicate Where Proprietary Information Has Been Removed

The interface configuration is acceptable, when the resulting keff is equivalent to, or less than the maximum keff of the individual pattern. The results are shown in Table 7.22 and show that this requirement is fulfilled for all analyzed cases and therefore:

• No restrictions are necessary between the uniform loading pattern and either of the checkerboard loading patterns (fresh or spent).

" For interfaces between the fresh fuel checkerboard and spent fuel checkerboard, the high reactivity spent fuel assembly (5.0 wt% 235 U, 27 GWD/MTU) may be face adjacent to no more than one fresh fuel assembly. The fresh fuel assembly may be face adjacent with up to 2 high reactivity spent fuel assemblies. Figure 7.5 shows one example of an acceptable 3x3 fresh fuel checkerboard within the center of a spent fuel checkerboard that meets these requirements.

7.4 Additional Calculations 7.4.1 Fuel Transfer Carriage Criticality The transfer carriage is capable of accommodating two fuel assemblies at a time, carried in stainless steel boxes. The fuel transfer carriage is conservatively modeled as two fuel assemblies at 5.0 wt% 235U and zero burnup separated by 5.06 inches of water only. The calculation of the criticality of the fuel transfer carriage accounts for both the carriage and the transfer tube. The results of the MCNP4a calculations are shown in Table 7.23.

Based on the design of the fuel transfer carriage, a fuel assembly could be mislocated outside the carriage. Two additional calculations were performed with a fresh fuel assembly mislocated directly adjacent to one of the two, fuel assemblies in the carriage. The results of the MCNP4a calculations are shown in Table 7.23.

7.4.2 Upender Criticality The criticality of the Upender is bounded by the calculation of the fuel transfer carriage in Section 7.4.1.

7.4.3 New Fuel Elevator Criticality The criticality of the New Fuel Elevator is bounded by the calculation of the fuel transfer carriage in Section 7.4.1.

7.4.4 Boron Dilution Accident Evaluation Project No. 1712 Report No. HI-2084014 Page 23 Shaded Areas Indicate Where Proprietary Information Has Been Removed

The soluble boron in the spent fuel pool water is conservatively assumed to contain a minimum of 1720 ppm under- operating conditions. Significant loss or dilution of the soluble boron concentration is extremely unlikely, if not incredible. Nonetheless, an evaluation was performed based on the data provided by Entergy.

The required minimum soluble boron concentration is 447 ppm under normal conditions and 838 ppm for the most serious credible accident scenario (see Table 7.17 and Table 7.21). The volume of water in the pool is approximately 288,748 gallons. Large amounts of un-borated water would be necessary to reduce the boron concentration from 1720 ppm to.-838 ppm or to 447 ppm.

Abnormal or accident conditions are discussed below for either low dilution rates (abnormal conditions) or high dilution rates (accident conditions).

7.4.4.1 Low Flow Rate Dilution Small dilution flow around pump seals and valve stems or mis-aligned valves could possibly occur in the normal soluble boron control system or related systems. Such failures might not be immediately detected. These flow rates would be of the order of 2 gpm maximum and the increased frequency of makeup flow might not be observed. However, an assumed loss flow-rate of 2 gpm dilution flow rate would require approximately 135 days to reduce the boron concentration to the minimum required 447 ppm under normal conditions or 72 days to reach the 838 ppm required for the most severe fuel handling accident. Routine surveillance measurements of the soluble, boron concentration would readily detect the reduction in soluble boron concentration with ample time for corrective action.

Administrative controls require a measurement of the soluble boron concentration in the pool water at least weekly. Thus, the longest time period that a potential boron dilution might exist without a direct measurement of the boron concentration is 7 days. In this time period, an undetected dilution flow rate of 38.6 gpm would be required to reduce the boron concentration to 447 ppm. No known dilution flow rate of this magnitude has been identified. Further, a total of more than 389,000 gallons of un-borated water would be associated with the dilution event and such a large flow of un-borated water would be readily evident by high-level alarms and by visual inspection on daily walk-downs of the storage pool area.

7.4.4.2 High Flow Rate Dilution Under certain accident conditions, it is conceivable that a high flow rate of un-borated water cOuld flow into the spent fuel pool. As discussed in Section 5.6.4, the component cooling water makeup pumps provide makeup fromithe Condensate Storage Pool and have a capacity of 600 gpm. Such an accident scenario could result from the continuous operation of the Condensate Storage Pool pump and a flow rate of up to 600 gpm which could possibly contribute large amounts of un-borated water into the spent fuel.

Conservatively assuming that all the un-borated water from the pump poured into the pool and further assuming instantaneous mixing of the un-borated water with the pool water, it would take approximately 648 minutes to dilute the soluble boron concentration to 447 ppm, which is the Project No. 1712 Report No. HI-2084014 Page 24 Shaded Areas Indicate Where Proprietary Information Has Been Removed

minimum required concentration to maintain keff below 0.95 under normally operating conditions. In this dilution accident, some 389,000 gallons of water would be released into the spent fuel pool and multiple alarms would have alerted the control room of the accident consequences (including the fuel pool high-level alarm and the Fuel Handling Building sump high level alarm and Liquid Waste Management Trouble, alarm). For this high flow rate condition, 346 minutes would be required to reach the 838 ppm required for the most severe fuel handling accident.

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

7.4.5 Temporary Storage Racks The results of the TSR are summarized in Table 7.24. These results show that the TSR is qualified for loading fuel assemblies with an. initial enrichment' of up to 5.0 wt% 235U. Based on information provided by Entergy, a fuel assembly may be mislocated on the exterior of the TSR.

The mislocated fresh fuel assembly was modeled at the closest approach (See Table 5.5). For simplification, the following tolerances are included in the design basis model: fuel density, lattice pitch and enrichment (See Table 5.5). The results of the mislocated case and the necessary soluble boron amount are present in Table 7.24.

7.4.6 Fuel Pin Storage Container The FPSC calculation involved comparing the reactivity of the FPSC to three cases of NGF fuel assemblies under equivalent modeling conditions: a fresh fuel assembly, a burnup of 27 GWD/MTU and a bumup of 33.4 GWD/MTU, all at 5.0 wt% 235U. These three cases match the most reactive fuel assembly for the three loading patterns analyzed in the main body of the report. The results of these comparisons can be seen in Table 7.25. Therefore the FPSC can be placed in any location intended for fresh or spent fuel.

7.4.7 New Fuel Storage Vault The maximum calculated reactivity of the NFV is listed in Table 7.28. The calculated reactivity as a function of water density is also shown in Figure 7.7. The results show that the optimum moderator density occurs at 100% water density and this maximum k'fr is below the regulatory limit.

Project No. 1712 Report No. HI-2084014 Page 25 Shaded Areas Indicate Where Proprietary Information Has Been Removed

REFERENCES'

1. M.G. Natrella, Experimental Statistics, National Bureau of Standards, Handbook 91, August 1963.

2. J.F. Briesmeister, Editor, "MCNP - A General Monte Carlo N-Particle Transport Code, Version 4A," LA-12625, Los Alamos National Laboratory (1993).

3. "Lumped Fission Product and Pm148m Cross Sections for MCNP," Holtec Report HI-2033031, Rev 0, September 2003.

4. M. Edenius, K. Ekberg, B.H. Forssdn, and D. Knott, "CASMO-4 A Fuel Assembly Burnup Program User's Manual," Studsvik/SOA-95/1, Studsvik of America, Inc. and Studsvik Core Analysis AB (proprietary).

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

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

(proprietary).

7. L.I. Kopp, "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants," NRC Memorandum from L. Kopp to T.

Collins, August 19, 1998.

8. ANS-8.]/N16.1-1975, "American National Standard for Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors," April 14, 1975.

9. S.E. Turner, "Uncertainty Analysis - Burnup Distributions," presented at the DOE/SANDIA Technical Meeting on Fuel Burnup Credit, Special Session, ANS/ENS Conference, Washington, D.C., November 2, 1988.

10. "Study of the Effect of Integral Burnable Absorbers for PWR Burnup Credit," NUREG/CR-6760, ORNL/TM-2000-321, March 2002.

Note: The revision status of Holtec documents cited above is subject to updates as the project progresses. This document will be revised if a revision to any of the above-referenced Holtec work materially affects the instructions, results, conclusions or analyses contained in this document. Otherwise, a revision to this document will not be made and the latest revision of the referenced Holtec documents shall be assumed to supercede the revision numbers cited above. The Holtec Project Manager bears the undivided responsibility to ensure that there is no intra-document conflict with respect to the information contained in all Holtec generated documents on a safety-significant project. The latest revision number of all documents produced by Holtec International in a safety significant project is readily available from the company's electronic network.

Project No. 1712 Report No. HI-2084014 Page 26 Shaded Areas Indicate Where Proprietary Information Has Been Removed

Table 5.1 Fuel Assembly Specification Assembly Type 16x16 Standard 16x16 NGF Stack Density, g/cm 3 10. 412 10.522 Fuel Rod Pitch, in 0.506 0.506 Number of Fuel Rods 236 236 Number of Guide Tubes 5 5 Fuel Rod Clad OD, in 0.382 0.374 Fuel Rod Clad ID, in 0.332 0.329 Active Length, in 149.61-150.0 150.0 Fuel Pellet Diameter, in 0.325 0.3225 Guide Tube OD, in 0.98 0.98 Guide Tube ID, in 0.9 0.9 ZrB2 Rod Coating 3.14 Loading (mgm 1°B/inch) 3.14 ZrB2 Rod Coating 0.0004167 0.000417 Thickness (inches)

ZrB 2 Rod Coating 136 138 Length (inches)

Fuel Assembly Width n/a 8.125 (min), in.

Bottom of Active Fuel to Bottom of Fuel n/a 5.402 Assembly, in.

Project No. 1712 Report No. HI-2084014 Page 27 Shaded Areas Indicate Where Proprietary Information Has Been Removed

Table 5.2 Core Operating Parameter for Depletion Analyses Parameter Value Soluble Boron Concentration (bounding cycle 1000 average), ppm Reactor Specific Power, MW/MTU 40.5 Core Average Fuel Temperature, 'F 1041.0 Core Average Moderator Temperature at the 614.0 Top of the Active Region, 'F In-Core Assembly Pitch, Inches 8.18 Project No. 1712 Report No. HI-2084014 Page 28 Shaded Areas Indicate Where Proprietary Information Has Been Removed

Table 5.3 Axial Burnup Profiles Node Center Relative Burnup Relative Burnup (cm) < 25 GWD/MT > 25 GWD/MT 7.62 0.54 0.593 22.86 0.773 0.819 38.1 0.921 0.961 53.34 1.013 1.028 68.58 1.055 1.051 83.82 1.065 1.057 99.06 1.064 1.058 114.3 1.061 1.058 129.54 1.058 1.057 144.78 1.056 1.056 160.02 1.054 1.055 175.26 1.053 1.054 190.5 1.052 1.054 205.74 1.051 1.053 220.98 1.05 1.051 236.22 1.047 1.049 251.46 1.046 1.048 266.7 1.044 1.046 281.94 1.04 1.043 297.18 1.031 1L036 312.42 0.994 1.021 327.66 0.92 0.966 342.9 0.81 0.873 358.14 0.655 0.725 373.38 0.441 0.508 Project No. 1712 Report No. HI-2084014 Page 29 Shaded Areas Indicate Where Proprietary Information Has Been Removed

Table 5.4 Storage Rack and Spent Fuel Pool Parameter Specification Region 1 Parameter Value Tolerance Cell ID, in 8.5 Cell Wall thickness, in 0.075 2Cell Pitch, in 10.185 Boundary Sheathing Thickness, in 0.075 Inner Sheathing Thickness, in 0.0235 3Poison Thickness, in 0.089 Poison Width, in 7.25 Poison Gap, (nominal) in 0.096 Flux Trap (nominal) in 1.3 B-10 Loading, (nom)rg/cm2 0.028 ,

Region 2 Parameter Value Tolerance Cell ID, in .8.5 Cell Wall thickness, in 0.075 " .

Cell Pitch, in 8.692 Boundary Sheathing Thickness, in 0.075 Inner Sheathing Thickness, in 0.035 Poison Thickness, in 0.075 !4 Poison Width, in 7.25 Poison Gap, in (nominal) 0.082 B-10 Loading, (nom) g/cm2 0.0216 Additional Spent Fuel Pool Information Parameter Value Tolerance Soluble Boron Concentration, ppm 1720 n/a Spent Fuel Pool Volume, cf 38,600 n/a Fuel Transfer Carriage Gap, in 5.06 n/a Refueling Water Storage Pool (min), ppm .2050 n/a Refueling Water Pool Purification Pump, gpm 150 n/a Component Cooling Water Makeup Pumps, gpm 600 n/a 2 Note that [4] indicates a larger cell-cell pitch for the North-South direction. The value used is bounding.

3 Note that the actual model used 0.075 inches for the poison thickness for conservatism.

Project No. 1712 Report No. HI-2084014 Page 30 Shaded Areas Indicate Where Proprietary Information Has Been Removed

Table 5.5 Reactor Building Temporary Storage Rack Parameter Value Number of Storage Cells 5 Pitch, in. 18 +/- 0.02 Rack Opening, in. 8.62 +/-0.06 Canal Wall to Cell Center, in. 8.06 Distance from Outside Edge of Cell Wall to Outside Edge of Structural 2.25 Material of Cell, in.

Enrichment Tolerance, wt% 235U + 0.05 Fuel DensityTolerance, g U0 2/cm 3 +/-0.165 Rack Pitch Spacing 4 Tolerance, in. +/-0.555 Fuel Pin Storage Container Parameter Value Steel Tube Outer Diameter 5, in. 0.625 Steel Tube Thickness, in. 0.035 Steel Tube Pitch, in. 0.917 4 The rack pitch spacing is used to account for the possible gaps between the fuel assembly and rack inner wall. This value is used in the place of the much smaller pitch tolerance listed.

5 Note: 4 tubes have a larger outer diameter; the smaller diameter is used to conservatively model less steel.

Project No. 1712 Report No. HI-2084014 Page 31 Shaded Areas Indicate Where Proprietary Information Has Been Removed

Table 5.6 New Fuel Vault Parameters Parameter Value Vault North-South width, ft. 27.5 Vault East-West width, ft. 29.25 Rack Cell Opening, in. 8.9375 Thickness of Redwood Planks, in. 1.5 Rack Cell Pitch, in. 21 East-West Rack Module Center-to- 49 Center Cell Separation, in.

North-South Rack Module Center-to-Center Cell Separation, in.

Distance from Fuel Assembly Center 12.25 to North Wall, in.

Distance from Fuel Assembly Center 60 to East and West Wall, in.

Distance from Fuel Assembly Center 91.75 to South Wall, in.

Depth of Rack Cell, in. 190 Project No. 1712 Report No. HI-2084014 Page 32 Shaded Areas Indicate Where Proprietary Information Has Been Removed

Table 7.1 Results of the Region 1 Reference Fuel Assembly Calculations 235 Assembly Type at 5.0 wt% U Calculated kaf Standard 0.9164 NGF 0.9268 Project No. 1712 Report No. HI-2084014 Page 33 Shaded Areas Indicate Where Proprietary Information Has Been Removed

Table 7.2 Region I Manufacturing Tolerances and Uncertainty Calculations Calculated Parameter Cluae Delta-k Reference Case CASMO 0.9268 n/a Storage Cell ID Increase 0.9370 0.0102 Storage Cell ID Decrease 0.9205 -0.0063 Storage Cell Pitch Increase 0.9184 -0.0084 Storage Cell Pitch Decrease 0.9350 0.0082 Storage Cell Poison Width Increase 0.9250 -0.0018 Storage Cell Poison Width Decrease 0.9289 0.0021 Storage Cell Poison Gap Minimum 0.9263 -0.0005 Storage Cell Box Wall Decrease 0.9242 -0.0026 Storage Cell Box Wall Increase 0.9285 0.0017 Storage Cell Poison B- 10 Loading Minimum 0.9291 0.0023 Fuel Rod Pitch Increase 0.9277 0.0009 Fuel Rod Pitch Decrease 0.9259 -0.0009 Fuel Rod Clad OD Increase 0.9248 -0.0020 Fuel Rod Clad OD Decrease 0.9288 0.0020 Fuel Rod Clad Thickness Minimum 0.9267 -0.0001 Fuel Pellet OD Increase 0.9271 0.0003 Fuel Pellet OD Decrease 0.9265 -0.0003 Guide Tube OD Increase 0.9268 0.0000 Guide Tube OD Decrease 0.9268 0.0000 Guide Tube Thickness Minimum 0.9272 0.0004 Fuel Pellet Enrichment Increase 0.9284 0.0016 Fuel Pellet Density Increase 0.9285 0.0017 Statistical Combination of Positive Reactivity Uncertainties: 0.0140 Project No. 1712 Report No. HI-2084014 Page 34 Shaded Areas Indicate Where Proprietary Information Has Been Removed

Table 7.3 Region 1 Temperature and Water Density Effects Results Case Calculated Cllt Delta-k kef Reference Temperature 32 F 0.9268 0.0000 39.2 F 0.9266 -0.0002 68 F 0.9253 -0.0015 80.33 F 0.9244 -0.0024 140 F 0.9188 -0.0080 255 F 0% voids 0.9028. -0.0240 255 F 10% voids 0.8681 -0.0587 255 F 20% voids 0.8295 -0.0973 Bias to 80.33 F: 0.0024 Project No. 1712 Report No. HI-2084014 Page 35 Shaded Areas Indicate Where Proprietary Information Has Been Removed

Table 7.4 Summary of the Criticality Safety Analysis for Region I Without Soluble Boron Uncertainties:

Bias Uncertainty (95%/95%) + 0.0011 Calculation Statistics (95%/95%,2.Oxa) + 0.0014 Fuel Eccentricity Negative Manufacturing Tolerances +/- 0.0140 Statistical Combination of Uncertainties - 0.0141 Reference Calculated keff (MCNP4a) 0.9354 Total Uncertainty (above) 0.0141 Bias to 80.33 OF 0.0024 Calculation Bias (see Appendix A) 0.0009 Maximum kff 0.9527 Regulatory Limit ker 1.0000 Project No. 1712 Report No. HI-2084014 Page 36 Shaded Areas Indicate Where Proprietary Information Has Been Removed

Table 7.5 Summary of the Criticality Safety Analysis for Region I with Soluble Boron Requirement Soluble Boron ppm 60.7 Uncertainties:

Bias Uncertainty (95%/95%) +/- 0.0011 Calculation Statistics (95%/95%,2.0xa) +/- 0.0014 Fuel Eccentricity Negative Manufacturing Tolerances + 0.0140 Statistical Combination of Uncertainties + 0.0141 Reference Calculated ktff (MCNP4a) 0.9277 Total Uncertainty (above) 0.0141 Bias to 80.33 'F 0.0024 Calculation Bias (see Appendix A) 0.0009 Maximum keff 0.9450 Regulatory Limit keff 0.9500 f

Project No. 1712 Report No. HI-2084014 Page 37 Shaded Areas Indicate Where Proprietary Information Has Been Removed

Table 7.6 Results of Associated Region I Reactivity Calculations Eccentric Positioning Case Case ker Reference 0.9354 Eccentric 0.9332 Delta-k -0.0022 Soluble Boron Case ppm Boron keff 0 0.9354 200 0.9099 Target keff 0.9277 Calculated ppm 61 Mislocated FA Case ppm Boron keff 0 0.9510 400 0.8962 Target klff 0.9277 Calculated ppm 170 Project No. 1712 Report No. HI-2084014 Page 38 Shaded Areas Indicate Where Proprietary Information Has Been Removed

Table 7.7 Region 2 Calculations for the Reference Fuel Assembly Enrichment 2.0 wt% 235 U Burnup (GWD/MTU) Standard NGF Ak 0.0 0.9568 0.9631 0.0063 0.1 0.9537 0.9600 0.0063 2.0 0.9391 0.9448 010057 4.0 0.9231 0.9283 0.0052 Enrichment 3.5 wt% 235U Burnup (GWD/MTU) Standard NGF Ak 0.0 1.1113 1.1179 0.0067 0.1 1.1089 1.1156 0.0067 2.0 1.0887 1.0952 0.0064 4.0 1.0719 1.0782 0.0062 6.0 1.0547 1.0607 0.0061 8.0 1.0377 1.0435 0.0058 10.0 1.0211 1.0267 0.0055 11.0 1.0130 1.0184 0.0054 12.5 1.0012 1.0063 0.0052 15.0 0.9819 0.9867 0.0048 17.5 0.9631 0.9674 0.0043 20.0 0.9446 0.9484 0.0038 22.5 0.9265 0.9298 0.0033 25.0 0.9088 0.9115 0.0027 Project No. 1712 Report No. HI-2084014 Page 39 Shaded Areas Indicate Where Proprietary Information Has Been Removed

11 Table 7.7 Continued Burnup (GWD/MTU) Standard NGF Ak Enrichment 5.0 wt% 5U 23 0.0 1.1932 1.1998' 0.0066 0.1 1.1914 1.1980 0.0066 2.0 1.1708 1.1773 0.0065 4.0 1.1558 1.1623 0.0064 6.0 1.1406 1.1470 0.0064 8.0 1.1254 1.1317 0.0063 10.0 1.1106 1.1168 0.0062 11.0 A1.1034 1.1095 0.0062 12.5 1.0927 1.0987 0.0060 15.0 1.0753 1.0812 0.0059 17.5 1.0584 1.0640 0.0056 20.0 1.0417 1.0471 0.0054 22.5 1.0254 1.0305 0.0051 25.0 1.0093 1.0141 0.0048 27.5 0.9934 0.9979 0.0044 30.0 0.9776 0.9817 0.0041 32.5 0.9620 0.9656 0.0037 35.0 0.9464 0.9497 0.0032 37.5 0.9310 0.9338 0.0028 40.0 0.9157 0.9180 0.0023 42.5 0.9005 0.9023 0.0018 Project No. 1712 Report No. HI-2084014 Page 40 Shaded Areas Indicate Where Proprietary Information Has Been Removed

Table 7.8 Region 2 Calculations for NGF Fuel IFBA Rods Reactivity Effect wt% U235 3.5 5.0 Number of IFBA Rods 0 148 Delta k 0 148 Delta k Burnup GWD/MTU 0.0 1.1179 0.8007 -0.3172 1.1998 0.9152 -0.2846 0.1 1.1156 0.8026 -0.3130 1.1980, 0.9162 -0.2818 2.0 1.0952 0.8564 -0.2388 1.1773 0.9476 -0.2297 4.0 1.0782 0.9013 -0.1769 1.1623 0.9774 -0.1848 6.0 1.0607 0.9330 -0.1278 1.1470 1.0000 -0.1469 8.0 1.0435 0.9537 -0.0898 1.1317 1.0165 -0.1153 10.0 1.0267 0.9655 -0.0611 1.1168 1.0276 -0.0892 11.0 1.0184 0.9686 -0.0498 1.1095 1.0315 -0.0780 12.5 1.0063 0.9704 -0.0359 1.0987 1.0353 -0.0635 15.0 0.9867 0.9673 -0.0194 1.0812 1.0371 -0.0441 17.5 0.9674 0.9585 -0.0089 1.0640 1.0343 -0.0297 20.0 0.9484 0.9461 -0.0024 1.0471 1.0279 -0.0192 22.5 0.9298 0.9315 0.0017 1.0305 1.0188 -0.0117 25.0 0.9115 0.9156 0.0041 1.0141 1.0076 -0.0065 27.5 0.8935 0.8990 0.0055 0.9979 0.9951 -0.0028 30.0 0.8758 0.8821 0.0063 0.9817 0.9815 -0.0002 32.5 0.8585 0.8653 0.0067 0.9656 0.9673 0.0016 35.0 0.8417 0.8486 0.0069 0.9497 0.9525 0.0029 37.5 0.8253 0.8323 0.0070 0.9338 0.9375 0.0037 40.0 0.8095 0.8165 0.0070 0.9180 0.9223 0.0043 42.5 0.7942 0.8011 0.0069 0.9023 0.9070 0.0047 45.0 0.7796 0.7864 0.0068 0.8868 0.8918 0.0050 47.5 n/a 0.8714 0.8766 0.0052 50.0 n/a 0.8562 0.8616 0.0053 N

52.5 n/a 0.8413 0.8468 0.0054 55.0 n/a 0.8267 0.8322 0.0055 57.5 n/a 0.8125 0.8180 0.0055 60.0 n/a 0.7986 0.8041 0.0055 Project No. 1712 Report No. HI-2084014 Page 41 Shaded Areas Indicate Where Proprietary Information Has Been Removed

Table 7.9 Region 2 Calculations for Eccentric Fuel Positioning Case Calculated Delta k keff Reference Uniform Loading 0.9570 Spent Fuel Uniform Loading 0.9517 -0.0053 Eccentric Positioning Reference Spent Fuel Checkerboard 0.9719 Loading -0.0044 Spent Fuel Checkerboard Loading 0.9675 Eccentric Positioning Reference Fresh Checkerboard 0.8256 Fresh Fuel Checkerboard Eccentric 0.8224-0.0032 0.8224 Positioning Project No. 1712 Report No. HI-2084014 Page 42 Shaded Areas Indicate Where Proprietary Information Has Been Removed

Table 7.10 Region 2 Calculations for Manufacturing Tolerance Uncertainties for Fuel Storage Cell B-10 Sttsia BWurnup Enrichment Ref ID + ID- Poison Poison Poison Box Box Loading Statistical GWD/MTU E h Case Width Width Gap Wall + Wall - oin Combo

+ Min M 0.0 2 0.9631 -0.0023 -0.0013 -0.0020 0.0026 0.0001- 0.0001 -0.0001 0.0034 0.0045 2.0 2 0.9448 -0.0024 -0.0012 -0.0020 0.0025 0.0001 0.0001 -0.0001 0.0034 0.0043 4.0 2.5 0.9897 -0.0029 -0.0009 -0.0021 0.0026 0.0001 0.0001 -0.0001 0.0035 0.0045 8.0 2.5 0.9534 -0.0028 -0.0008 -0.0020 0.0025 0.0001 0.0001 -0.0001 0.0034 0.0043 11.0 3 0.9769 -0.0030 -0.0006 -0.0021 0.0025 0.0001 0.0000 -0.0001 0.0035 0.0043 15.0 3 0.9443 -0.0029 -0.0006 -0.0020 0.0024 0.0001 0.0001 -0.0001 0.0034 0.0042 15.0 3.5 0.9867 -0.0032 -0.0004 -0.0021 0.0026 0.0001 0.0001 -0.0001 0.0035 0.0044 22.5 3.5 0.9298 -0.0029 -0.0005 -0.0020 0.0024 0.0001 0.0000 -0.0001 0.0033 0.0041 22.5 4 0.9679 -0.0032 -0.0003 -0.0020 0.0025 0.0001 0.0000 -0.0001 0.0034 0.0043 27.5 4 0.9326 -0.0030 -0.0004 -0.0020 0.0024 0.0001 0.0000 -0.0001 0.0033 0.0041 27.5 4.5 0.9673 -0.0032 -0.0002 -0.0020 0.0025 0.0001 0.0000 -0.0001 0.0034 0.0042 32.5 4.5 0.9338 -0.0031 -0.0003 -0.0020 0.0024 0.0001 0.0000 -0.0001 0.0033 0.0041 32.5 5 0.9656 -0.0033 -0.0002 -0.0020 0.0025 0.0001 0.0000 -0.0001 0.0034 0.0042 40.0 5 0.9180 -0.0030 -0.0002 -0.0019 0.0024 0.0001 0.0000 -0.0001 0.0032 0.0040 Project No. 1712 Report No. HI-2084014 Page 43 Shaded Areas Indicate Where Proprietary Information Has Been Removed

Table 7.11 Region 2 Calculations for Fuel Tolerance Uncertainties GuideFuel Bumup Clad Fuel Fuel Guide Guide Guide Fuel Pellet Sttistical GWD/ Enr Ref Pitch + Pitch - Clad Clad Thickness Pellet Pellet Tube Tube Tube Pellet MTU Case OD+ OD - Min OD + OD- OD + OD- Thickness Enr + Densit Combo.

Min +

0.0 2.0 0.9631 0.0007 -0.0007 -0.0009 0.0009 0.0005 0.0004 -0.0004 0.0000 0.0000 0.0002 0.0074 0.0022 0.0079 2.0 2.0 0.9448 0.0007 -0.0007 -0.0008 0.0008 0.0005 0.0004 -0.0004 0.0000 0.0000 0.0002 0.0070 0.0022 0.0075 4.0 2.5 0.9897 0.0008 -0.0008 -0.0008 0.0008 0.0005 0.0004 -0.0004 0.0000 0.0000 0.0002 0.0054 0.0019 0.0059 8.0 2.5 0.9534 0.0008 -0.0008 -0.0007 0.0007 0.0005 0.0004 -0.0004 0.0000 0.0000 0.0002 0.0054 0.0020 0.0059 11.0 3.0 0.9769 0.0009 -0.0009 -0.0008 0.0007 0.0005 0.0003 -0.0004 0.0000 0.0000 0.0002 0.0045 0.0018 0.0050 15.0 3.0 0.9443 0.0008 -0.0008 -0.0006 0.0006 0.0004 0.0004 -0.0004 0.0000 0.0000 0.0002 0.0046 0.0020 0.0051 15.0 3.5 0.9867 0.0009 -0.0009 -0.0007 0.0007 0.0004 0.0003 -0.0003 0.0000 0.0000 0.0002 0.0039 0.0017 0.0044 22.5 3.5 0.9298 0.0009 -0.0008 -0.0005 0.0005 0.0004 0.0004 -0.0004 0.0000 0.0000 0.0001 0.0041 0.0020 0.0047 22.5 4.0 0.9679 0.0009 -0.0009 -0.0007 0.0006 0.0004 0.0003 -0.0004 0.0000 0.0000 0.0002 0.0035 0.0017 0.0041 27.5 4.0 0.9326 0.0009 -0.0009 -0.0005 0.0005 0.0004 0.0004 -0.0004 0.0000 0.0000 0.0001 0.0037 0.0019 0.0043 27.5 4.5 0.9673 0.0010 -0.0009 -0.0006 0.0006 0.0004 0.0003 -0.0003 0.0000 0.0000 0.0002 0.0032 0.0016 0.0038 32.5 4.5 0.9338 0.0009 -0.0009 '-0.0005 0.0005 0.0004 0.0004 -0.0004 0.0000 0.0000 0.0001 0.0033 0.0018 0.0040 32.5 5.0 0.9656 0.0010 -0.0010 -0.0006 0.0006 0.0004 0.0003 -0.0003 0.0000 0.0000 0.0002 0.0030 0.0015 0.0036 40.0 5.0 0.9180 0.0009 -0.0009 -0.0004 0.0004 0.0004 0.0004 -0.0004 0.0000 0.0000 0.0001 0.0031 0.0019 0.0039 Project No. 1712 Report No. HI-2084014 Page 44 Shaded Areas Indicate Where Proprietary Information Has Been Removed

N Table 7.12 Region 2 Calculations for Pool Temperature Tolerance Uncertainties Burnup Enr Ref Case T=39.2F T=80.33F T=255F, T=255F, T=255F, GWDIMTU T =32 F 0% Voids 10% Voids 20% Voids 0.0 2.0 0.9631 -0.0008 -0.0056 -0.0318 -0.0495 -0.0714 2.0 2.0 0.9448 -0.0007 -0.0051 -0.0291 -0.0462. -0.0675 4.0 2.5 0.9897 -0.0006 -0.0046 -0.0273 -0.0458 -0.0684 8.0 2.5 0.9534 -0.0005 -0.0041 -0.0248 -0.0431 -0.0655 11.0 3.0 0.9769 -0.0005 -0.0038 -0.0242 -0.0435 -0.0667 15.0 3.0 0.9443 -0.0004 -0.0035 -0.0225 -0.0414 -0.0643 15.0 3.5 0.9867 -0.0004 -0.0036 -0.0234 -0.0433 -0.0671 22.5 3.5 0.9298 -0.0004 -0.0031 -0.0208 -0.0400 -0.0631 22.5 4.0 0.9679 -0.0004 -0.0032 -0.0219 -0.0419 -0.0658 27.5 4.0 0.9326 -0.0003 -0.0029 -0.0203 -0.0399 -0.0633 27.5 4.5 0.9673 -0.0003 -0.0030 -0.0213 -0.0416 -0.0658 32.5 4.5 0.9338 -0.0003 -0.0028 -0.0199 -0.0398 -0.0635 32.5 5.0 0.9656 -0.0003 -0.0029 -0.0208 -0.0414 -0.0657 40.0 5.0 0.9180 -0.0003 -0.0025 -0.0189 -0.0388 -0.0625 Project No. 1712 Report No. HI-2084014 Page 45 Shaded Areas Indicate Where Proprietary Information Has Been Removed

Table 7.13 Region 2 Results for the Spent Fuel Uniform Loading Enrichment (wt% 235U) 2.0 2.5 3.0 3.5 4.0 4.5 5.0 Burnup (GWD/MTU) 0 5.89 11.77 17.50 23.55 28.15 33.39 CASMO Burnup for 0.0 4.0 11.0 15.0 22.5 27.5 32.5 Tolerances Depletion Uncertainty 0.0000 0.0019 0.0051 0.0066 0.0091 0.0105 0.0117 Manufacturing Uncertainty 0.0045 0.0045 0.0043 0.0044 0.0043 0.0042 0.0042 Fuel Uncertainty . 0.0079 0.0059 0.0050 0.0044 0.0041 0.0038 0.0036 Calculational Uncertainty 0.0012 0.0012 0.0014 0.0014 0.0012 0.0012 0.0014 Code Uncertainty 0.0011 0.0011 0.0011 0.0011 0.0011 0.0011 0.0011 TotalUncertainty 0.0092 0.0078 0.0085 0.0092 0.0110 0.0121 0.0131 Code Bias 0.0009 0.0009 0.0009 0.0009 0.0009 0.0009 0.0009 Temperature Bias 0.0056 0.0046 0.0038 0.0036 0.0032 0.0030 0.0029 IFBA Bias 0.0070 0.0070 0.0070 0.0070 0.0070 0.0070 0.0070 MCNP kff 0 ppm Boron 0.9613 0.9747 0.9747 0.9743 0.9729 0.9720 0.9712 MCNP kff 600 ppm Boron 0.8560 n/a n/a 0.8948 n/a n/a 0.9040 Max keff 0 ppm Boron 0.9950 0.9950 0.9950 0.9950 0.9950 0.9950 0.9950 Max kffwith 600 ppm 0.8787 n/a n/a 0.9175 n/a n/a 0.9267 Boron Project No. 1712 Report No. HI-2084014 Page 46 Shaded Areas Indicate Where Proprietary Information Has Been Removed

Table 7.14 Region 2 Results for the Spent Fuel Checkerboard Loading Enrichment (wt% 235U) 2.0 2.5 3.0 3.5 4.0 4.5 5.0 Burnup (GWD/MTU) 2.41 9.42 16.21 23.55 29.53 34.64 41.23 CASMO Burnup for 2.0 8.0 15.0 22.5 27.5 32.5 40.0 Tolerances Depletion Uncertainty 0.0009 0.0038 0.0067 0.0094 0.0109 0.0122 0.0141 Manufacturing Uncertainty 0.0043 0.0043 0.0042 0.0041 0.0041 0.0041 0.0040 Fuel Uncertainty 0.0075 0.0059 0.0051 0.0047 0.0043 0.0040 0.0039 Calculational Uncertainty 0.0012 0.0012 0.0012 0.0012 0.0012 0.0012 0.0012 Code Uncertainty 0.0011 0.0011 0.0011 0.0011 0.0011 0.0011 0.0011 Total Uncertainty 0.0088 0.0083 0.0096 0.0114 0.0125 0.0135 0.0152 Code Bias 0.0009 0.0009 0.0009 0.0009 0.0009 0.0009 0.0009 Temperature Bias 0.0051 0.0041 0.0035 0.0031 0.0029 0.0028 0.0025 IFBA Bias 0.0070 0.0070 0.0070 0.0070 0.0070 0.0070 0.0070 MCNP kff 0 ppm Boron 0.9731 0.9747 0.9740 0.9726 0.9717 0.9708 0.9693 MCNP kff 600 ppm Boron 0.8893 n/a n/a 0.8969 n/a n/a 0.9008 Max keff 0 ppm Boron 0.9950 0.9950 0.9950 0.9950 0.9950 0.9950 0.9950 Max ker- with 600 ppm 0.9112 n/a n/a 0.9193 n/a n/a 0.9265 Boron Project No. 1712 Report No. HI-2084014 Page 47 Shaded Areas Indicate Where Proprietary Information Has Been Removed

Table 7.15 Region 2 Results for the Fresh Checkerboard Loading, 5.0 wt% 235U Enrichment (wt% 235U) 5.0 Burnup (GWD/MTU) 0 Manufacturing Uncertainty 0.0053 Fuel Uncertainty 0.0029 Calculational Uncertainty 0.0014 Code Uncertainty 0.0011 Total Uncertainty 0.0063 Code Bias 0.0009 Temperature Bias 0.0034 IFBA Bias 0.0070 MCNP keff 0 ppm Boron 0.8256 Max keff without Boron 0.8432

'I Project No. 1712 Report No. HI-2084014 Page 48 Shaded Areas Indicate Where Proprietary Information Has Been Removed

Table 7.16 Summary of the Criticality Safety Analysis for Region 2, Spent Fuel Uniform Loading, 0 ppm Soluble Boron Enrichment (wt% 235U) 5.0 Bumup (GWD/MTU) 33.4 Soluble Boron ppm 0.0 Fuel EccentricityT negative Statistical Combination of Uncertainties 0.0131 Calculated keff (MCNP4a) 0.9712 IFBA Bias 0.0070 Bias to 80.33 'F 0.0029 Calculation Bias (see Appendix A) 0.0009 Maximum keff 0.9950 Regulatory Limit k1t 1.0000 Project No. 171,2 Report No. HI-2084014 Page 49 Shaded Areas Indicate Where Proprietary Information Has Been Removed

Table 7.17 Summary of the Criticality Safety Analysis for Region'2, Spent Fuel Uniform Loading, 447 ppm Soluble Boron Enrichment (wt% 231U) 5.0 Burnup (GWD/MTU) 33.4 Soluble Boron (ppm) 447 Statistical Combination of Uncertainties + 0.0131 Calculated keff (MCNP4a) 0.9212 lFBA Bias 0.0070 Bias to 80.33 'F 0.0029 Calculation Bias (see Appendix A) 0.0009 Maximum keff 0.9450 Regulatory Limit keff 0.9500 Project No. 1712 Report No. HI-2084014 Page 50 Shaded Areas Indicate Where Proprietary Information Has Been Removed

Table 7.18 Summary of the Criticality Safety Analysis for Region 2, Spent Fuel Checkerboard Loading, 0 ppm Soluble Boron Enrichment (wt% 235U) 5.0 Burnup (GWD/MTU) 41.2 Soluble Boron (ppm) 0.0 Fuel Eccentricity negative Statistical Combination of Uncertainties +/- 0.0152 Calculated kff (MCNP4a) 0.9693 IFBA Bias 0.0070 Bias to 80.33 'F 0.0025 Calculation Bias (see Appendix A) 0.0009 Maximum keff 0.9950 Regulatory Limit keff 1.0000 Project No. 1712 Report No. HI-2084014 Page 51 Shaded Areas Indicate Where Proprietary Information Has Been Removed

Table 7.19 Summary of the Criticality Safety Analysis for Region 2, Spent Fuel Checkerboard Loading, 438 ppm Soluble Boron 235U)

Enrichment (wt% 5.0 Burnup (GWD/MTU) 41.2 Soluble Boron (ppm) 438 Statistical Combination of Uncertainties + 0.0152 Calculated kff (MCNP4a) 0.9193 IFBA Bias 0.0070 Bias to 80.33 'F 0.0025 Calculation Bias (see Appendix A) 0.0009 Maximum klff 0.9450 Regulatory Limit keff 0.9500 2

Project No. 1712 Report No. HI-2084014 Page 52 Shaded Areas Indicate Where Proprietary Information Has Been Removed

Table 7.20 Summary of the Criticality Safety Analysis for Region 2, Fresh' Fuel Checkerboard Loading, 0 ppm Soluble Boron Enrichment (wt% 235U) 5.0 Burnup (GWD/MTU) 0.0 Soluble Boron (ppm) 0.0 Fuel Eccentricity negative Statistical Combination of Uncertainties + 0.0063 Calculated keff (MCNP4a) 0.8256 IFBA Bias 0.0070 Bias to 80.33 'F 0.0034 Calculation Bias (see Appendix A) 0.0009 Maximum keff 0.8432 Regulatory Limit keff 1.0000 Project No. 1712 Report No. HI-2084014 Page 53 Shaded Areas Indicate Where Proprietary Information Has Been Removed

Table 7.21 Summary of Region 2 Accident Cases Case Result Dropped Fuel Assembly - Horizontal On Negligible Top of Cells Dropped Fuel Assembly - Vertical into Negligible Storage Cell Misloaded Fuel Assembly, Spent Fuel Checkerboard Loading, 5.0 wt% 235U 8386 (ppm Soluble Boron)

Mislocated Fuel Assembly, Spent Fuel 235U 5347 Checkerboard Loading, 5.0 wt%

(ppm Soluble Boron) 6 This case was the maximum for the misloaded assembly in the spent fuel uniform loading, spent fuel checkerboard loading, or fresh fuel checkerboard.

7 This case was the maximum for the mislocated assembly in the spent fuel uniform loading, spent fuel Checkerboard loading, or fresh fuel checkerboard.

Project No. 1712 Report No. HI-2084014 Page 54 Shaded Areas Indicate Where Proprietary Information Has Been Removed

Table 7.22 Region 2 Calculation Results for the Interface Cases Axial Burnup Ref kf Description Profile Enr (GWD/MTU) keff (at curve)

Interface Segmented 2.0 2.41 0.9598 0.9731 between rackeof fhalf a Spent fuel chekfuerbr Segmented 3.5 23.55 0.9513 0.9726 rack of fresh checkerboard fuel loading, fresh Segmented 5.0 41.23 0.9464 0.9693 checkerboard FA adjacent 27 Uniform 2.0 2.41 0M959 0.9731 and half a rack GWD/MTU, 5.0 Uniform 3.5 23.55 0.9539 0.9726 of spent fuel wt% 135U FA checkerboard Uniform 5.0 41.23 0.9506 0.9693 Interface Segmented 2.0 2.41 0.9734 0.9731 between a 3x3 set of fresh Spent fuel Segmented 3.5 23.55 0.9659 0.9726 checkerboard checkerboard Segmented 5.0 41.23 0.9611 0.9693 (fresh in loading, fresh center) FA adjacent 27 Uniform 2.0 2.41 0.9716 0.9731 surrounded by GWD/MTU, 5.0 Uniform 3.5 23.55 0.9679 0.9726 a rack of spent wt% 235U FA fuel Uniform 5.0 41.23 0.9642 0.9693 checkerboard Segmented 2.0 2.41 0.9697 0.9731 Segmented 3.5 23.55 0.9694 0.9726 Interface between a set of spent Segmented 5.0 41.23 0.9667 0.9693 fuel checkerboard loading fuel and spent uniform loading fuel. Uniform 2.0 2.41 0.97 0.9731 Uniform 3.5 23.55 0.9706 0.9726 Uniform 5.0 41.23 0.9653 0.9693 C

Project No. 1712 Report No. HI-2084014 Page 55 Shaded Areas Indicate Where Proprietary Information Has Been Removed

Table 7.23 Results of the Calculation of the Fuel Transfer Carriage Descriptio Calculated keff n

Reference 0.9436 Case Mislocated 1.0612 Case 800 ppm 0.9209 Boron Case f

Project No. 1712 Report No. HI-2084014 Page 56 Shaded Areas Indicate Where Proprietary Information Has Been Removed

Table 7.24 Results of the Criticality Analysis for the TSR Description Calculated kefr TSR Design Basis Model 0.9297 TSR Mislocated Fuel Assembly 1.0204 Model TSR Mislocated Fuel Assembly 0.8525 Model with 800 ppm Soluble Boron Extrapolated TSR Soluble Boron Requirement for Mislocated 359 Accident, ppm_

Project No. 1712 Report No. HI-2084014 Page 57 Shaded Areas Indicate Where Proprietary Information Has Been Removed

Table 7.25 Results of the Criticality Analysis for the FPSC Description Calculated keff FPSC Design Basis Model 0.6715 5.0 wt% 235U Fuel Assembly at 0.7521 33.4 GWD/MTU 5.0 wt% 235U Fuel Assembly at 27 0.7784 GWD/MTU Fresh NGF Fuel Assembly 0.9226 Project No. 1712 Report No. HI-2084014 Page 58 Shaded Areas Indicate Where Proprietary Information Has Been Removed

Table 7.26 Region 2 Burnup Versus Enrichment Curve for Spent Fuel Uniform Loading Enrichment (wt% 235U) Burnup (GWD/MTU) 2.0 0.0 2.5 5.9 3.0 11.8 3.5 17.5 4.0 23.5 4.5 28.1 5.0 33.4 Project No. 1712 Report No. HI-2084014 Page 59 Shaded Areas Indicate Where Proprietary Information Has Been Removed

Table 7.27 Region 2 Burnup Versus Enrichment Curve for Spent Fuel Checkerboard Loading 235 Enrichment (wt% U) Burnup (GWD/MTU) 2.0 2.4 2.5 9.4 3.0 16.2 3.5 23.6 4.0 29.5 4.5 34.6 5.0 41.2 Project No. 1712 Report No. HI-2084014 Page 60 Shaded Areas Indicate Where Proprietary Information Has Been Removed

Table 7.28 Summary of the Criticality Safety Analysis for New Fuel Vault, 100% Moderator Density Tolerances:

Enrichment kerr 0.9195 0.0008 Enrichment Uncertainty

• 0.0034 Pellet Density krff 0.9192 0.0008 Pellet Density Uncertainty

• 0.0031 Storage Rack Pitch kefr 0.9187 +/- 0.0007 Storage Rack Pitch Uncertainty 0.0023

+/--

Bias Uncertainty (95%/95%) 0.0011 Calculation Statistics (95%/95%,2xa) 0.0014 Statistical Combination of

- 0.0054 Uncertainties Calculated keff (MCNP4a) 0.9184 Calculation Bias (see Appendix A) 0.0009 Maximum keff 0.9247 Regulatory Limit keff 0.9500 Project No. 1712 Report No. HI-2084014 Page 61 Shaded Areas Indicate Where Proprietary Information Has Been Removed

Figure 5.1 Region I Model Project No. 1712 Report No. HI-2084014 Page 62 Shaded Areas Indicate Where Proprietary Information Has Been Removed

Figure 5.2 Region 2 Model Project No. 1712 Report No. HI-2084014 Page 63 Shaded Areas Indicate Where Proprietary Information Has Been Removed

Figure 7.1 Region 2 Spent Fuel Uniform Loading Burnup versus Enrichment Curve 40 35 30 525 o 2020 a-I m 15 10 5

0 2.0 2.5 3.0 3.5 4.0 4.5 5.0 Enrichment wt% U-235 Project No. 1712 Report No. HI-2084014 Page 64 Shaded Areas Indicate-Where Proprietary Information Has Been Removed

Figure 7.2 Region 2 Spent Fuel Checkerboard Loading Bumup versus Enrichment Curve 45 41.2 40 34.6 35 29.5 30 2 25 23.6 S20o CL E ,16.

15 10 .9.

0 2.0 2.5 3.0 3.5 4.0 4.5 5.0 Enrichment wt% U-235 Project No. 1712 Report No. HI-2084014 Page 65 Shaded Areas Indicate Where Proprietary Information Has Been Removed

Figure 7.3 Region 2 intra-rack interface between half a rack of Fresh fuel checkerboard and half a rack of spent fuel checkerboard A A A-,A-'

. 5wt% 1 3 5U27 GWD/MTU BSpent Fuel At Spent Fuel Checkerboard Curve 5 wt% 2 3 5 U Fresh Fuel Empty Cell Project No. 1712 Report No. HI-2084014 Page 66 Shaded Areas Indicate Where Proprietary Information Has Been Removed

Figure 7.4 Region 2 intra-rack interface between a 3x3 set of Fresh fuel checkerboard (fresh in center) surrounded by a rack of spent fuel checkerboard B Spent Fuel At Checkerboard Curve 5 wt% 235U Fresh Fuel Empty Cell Project No. 1712 Report No. HI-2084014 Page 67 Shaded Areas Indicate Where Proprietary Information Has Been Removed

Figure 7.5 Two-Dimensional Representation of the Actual Calculations Model used for the New Fuel Vault as seen from above.

Project No. 1712 Report No. HI-2084014 Page 68 Shaded Areas Indicate Where Proprietary Information Has Been Removed

Figure 7.6 Two-Dimensional Representation of the Actual Calculations Model used for the New Fuel Vault as seen from the side.

Project No. 1712 Report No. HI-2084014 Page 69 Shaded Areas Indicate Where Proprietary Information Has Been Removed

Figure 7.7 Results of the Waterford Unit 3 New Fuel Vault Criticality Analysis As a Function of Water Density 0.95 0.9 0.85 0.8

• 0.75 0.7 0.65 0.6 0.55 0.5 10 100

% Moderator Density Project No. 1712 Report No. HI-2084014 Page 70 Shaded Areas Indicate Where Proprietary Information Has Been Removed

Appendix A Benchmark Calculations (total number of pages: 26 including this page)

HOLTEC PROPRIETARY APPENDIX HAS BEEN REMOVED IN IT'S ENTIRETY Project No. 1712 Report No. HI-2084014 Page A- I