Text

License Amendment Request to Reconcile 10 CFR 50 and 10 CFR 72 Criticality Requirements for Loading and Unloading Dry Spent Fuel Storage Canisters in the Spent Fuel Pool, License Amendment Request No. 2006-009
	ML060720030
Person / Time
Site:	Oconee
Issue date:	03/01/2006
From:	Brandi Hamilton; Duke Power Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	Download: ML060720030 (76)
	v • d • e

MkDuke iBRUCE H HAMILTON I OPowere Oconee Nuclear Station Duke Power ONO0 VP / 7800 Rocherster Highway Seneca, SC 29672 864 885 3487 864 885 4208 fax March 1, 2006 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555-0001

Subject:

Oconee Nuclear Site, Units 1, 2, and 3 Docket Numbers 50-269, 50-270, and 50-287 License Amendment Request to Reconcile 10 CFR 50 and 10 CFR 72 Criticality Requirements for Loading and Unloading Dry Spent Fuel Storage Canisters in the Spent Fuel Pool License Amendment Request No. 2005-009

Reference:

NRC Regulatory Issue Summary 2005-05, "Regulatory Issues Regarding Criticality Analyses for Spent Fuel Pools and Independent Spent Fuel Storage Installations,"

dated March 23, 2005.

In accordance with 10 CFR 50.90, Duke Energy Corporation (Duke) proposes to amend Renewed Facility Operating Licenses Nos. DPR-38, DPR-47, and DPR-55. If granted, this amendment request will allow spent fuel loading, unloading, and handling operations in the Oconee Nuclear Site (Oconee) Spent Fuel Pools (SFP) that support spent fuel transfer to an Independent Spent Fuel Storage Installation (ISFSI) licensed under 10 CFR 72.

As the Nuclear Regulatory Commission (NRC) noted in the Reference, there are differences in the criticality requirements of 10 CFR 50 for SFPs and 10 CFR 72 for ISFSIs. Duke has completed a review of the Oconee 10 CFR 72 licensing basis for the NUHOMS.-24PHB and the NUHOMS,-24P storage systems and concluded they do not meet the requirements of 10 CFR 50.68(b)(1) during loading and unloading operations in the Oconee SFP.

To demonstrate acceptable subcriticality margins for cask loading and unloading operations in accordance with 10 CFR 50 requirements, Duke proposes to revise applicable sections of the Oconee Technical Specifications (TS), TS Bases, and Updated Final Safety Analysis Report (UFSAR) to incorporate changes based on the results of a new criticality analysis. The new criticality limits for spent fuel dry storage casks are derived from a methodology that is very similar to that approved by the NRC for the McGuire Nuclear Site spent fuel storage racks.

Results of this new analysis show that application of the new and revised TS and TS Bases will assure that there is acceptable subcriticality margin for cask loading and unloading operations in both Oconee SFP. The changes proposed in this amendment request will not affect the 10 CFR Aowc www. dukepower. comn 0 fn

Nuclear Regulatory Commission License Amendment Request No. 2005-009 March 1, 2006 Page 2 72 license, but will reconcile the differences in the 10 CFR 50 and 10 CFR 72 licenses and are hereby proposed for NRC approval.

At the meeting held between Duke and the NRC staff on November 1, 2005, the NRC indicated that Duke should address 10 CFR 50.68(b) in this submittal. Accordingly, Enclosure 4 describes how Oconee complies with 10 CFR 50.68(b). Duke will also update applicable sections of the Oconee UFSAR and submit these changes per 10 CFR 50.71(e).

In accordance with Duke administrative procedures and the Quality Assurance Program Topical Report, these proposed changes have been reviewed and approved by the Plant Operations Review Committee and Nuclear Safety Review Board. Additionally, a copy of this license amendment request is being sent to the State of South Carolina in accordance with 10 CFR 50.91 requirements. In order to support required ISFSI transfers in Summer 2006, Duke respectfully requests that the amendment be issued by June 1, 2006, with a 90-day implementation period from the date of issuance.

Inquiries on this proposed amendment request should be directed to Reene' Gambrell of the Oconee Regulatory Compliance Group at (864) 885-3364.

Sincerely, B. H. Hamilton, Vice President Oconee Nuclear Site

Enclosures:

1. Notarized Affidavit

2. Evaluation of Proposed Change

3. Oconee Nuclear Site - NUHOMSO-24P/24PHB DSC Criticality Analysis

4. Compliance with 10 CFR 50.68(b)

Attachments:

1. Technical Specifications - Mark Ups

2. Technical Specifications - Reprinted Pages

3. List of Regulatory Commitments

Nuclear Regulatory Commission License Amendment Request No. 2005-009 March 1, 2006 Page 3 bc w/enclosures and attachments:

Mr. W. D. Travers, Regional Administrator U. S. Nuclear Regulatory Commission - Region II Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, Georgia 30303 Mr. L. N. Olshan, Project Manager Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop 0-14 H25 Washington, D. C. 20555 Mr. M. C. Shannon Senior Resident Inspector Oconee Nuclear Site Mr. Henry Porter, Director Division of Radioactive Waste Management Bureau of Land and Waste Management Department of Health & Environmental Control 2600 Bull Street Columbia, SC 29201

ENCLOSURE 1 NOTARIZED AFFIDAVIT

- Notarized Affidavit License Amendment Request No. 2005-009 March 1, 2006 Page 1 AFFIDAVIT B. H. Hamilton, being duly sworn, states that he is Vice President, Oconee Nuclear Site, Duke Energy Corporation, that he is authorized on the part of said Company to sign and file with the U. S. Nuclear Regulatory Commission this revision to the Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55; and that all statements and matters set forth herein are true and correct to the best of his knowledge.

B. H. Hamilton, Vice President Oconee Nuclear Site Subscribed and sworn to before me this / X day of 2006 Notary Public My Commission Expires:

-Date SEAL

ENCLOSURE 2 EVALUATION OF PROPOSED CHANGE

- Evaluation of Proposed Change License Amendment Request No. 2005-009 March 1, 2006 Page 1

Subject:

License Amendment Request to Reconcile 10 CFR 50 and 10 CFR 72 Criticality Requirements for Loading and Unloading Dry Spent Fuel Storage Canisters in the Spent Fuel Pool

1. DESCRIPTION

2. PROPOSED CHANGE

3. BACKGROUND

4. TECHNICAL ANALYSIS

5. REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration 5.2 Applicable Regulatory Requirements/Criteria

6. ENVIRONMENTAL CONSIDERATION

- Evaluation of Proposed Change License Amendment Request No. 2005-009 March 1, 2006 Page 2

1.0 DESCRIPTION

Oconee Nuclear Site (Oconee) currently stores spent fuel assemblies in its Spent Fuel Pools (SFPs) and at its Independent Spent Fuel Storage Installation (ISFSI). These storage locations have been previously licensed utilizing criteria from different sections of the code of federal regulations (CFR). In the spent fuel assembly transfer process (from underwater pool to dry cask storage), there are instances where regulatory requirements of both the facility operating licenses (FOLs) and ISFSI licenses apply; however, as described in Regulatory Issue Summary (RIS) 2005-05, there are notable differences in the criticality requirements of 10 CFR 50 (for SFPs) and 10 CFR 72 (for ISFSIs). Duke Energy Corporation (Duke) has completed a review of the 10 CFR 72 licensing bases for the NUHOMSD-24PHB and the NUHOMS,&-24P storage systems and concluded they do not meet the requirements of 10 CFR 50.68(b)(1) during cask loading and unloading operations in the Oconee SFP.

2.0 PROPOSED CHANGE

Duke proposes to revise applicable sections of the Oconee Units 1, 2, and 3 Technical Specifications (TS) and TS Bases to establish boron concentration limits for spent fuel cask loading and unloading operations and to restrict the burnup of spent fuel assemblies that can be loaded into a spent fuel storage cask while in a SFP. Specifically, TS 3.7.12, "Spent Fuel Boron Concentration," will be revised to add cask loading and unloading operations to the current SFP application and a new TS 3.7.18, "Dry Spent Fuel Storage Cask Loading and Unloading" will be added to address burnup limits for fuel assemblies loaded into casks while in a SFP. In addition, TS Section 4.0, "Design Features," will be revised to include spent fuel storage cask loading and unloading operations, and associated TS Bases will be revised or added as necessary. These changes are needed to ensure the requirements of 10 CFR 50.68(b)(1) are met when loading and unloading the NUHOMSv dry storage canisters (DSCs) at Oconee.

3.0 BACKGROUND

Oconee uses the NUHOMSx dry spent fuel storage system at its ISFSI. Forty NUHOMS-24P DSCs have been loaded under Duke's specific license (SNM-2503).

Another forty-four NUHOMS&-24P DSCs have been loaded under Duke's general license. Certificate of Compliance (CoG) 72-1004 is applicable to the DSCs loaded under the general license. For future loadings, Oconee will use the NUHOMS-24PHB which has been approved by the NRC as Amendment 6 to CoC 72-1004.

The minimum dissolved boron concentration for the SFP at Oconee is provided in TS 3.7.12 for the Renewed Facility Operating License. TS Surveillance Requirement (SR)

- Evaluation of Proposed Change License Amendment Request No. 2005-009 March 1, 2006 Page 3 3.7.12.1 requires verification that the pool boron concentration is within the limits of the Core Operations Limits Report (COLR) and greater than or equal to 2,220 ppm.

4.0 TECHNICAL ANALYSIS

The criticality analysis' of the NUHOMS&-24P/24PHB DSC, for loading and unloading operations in the Oconee SFPs, has been performed in accordance with the requirements of 10 CFR 50.68(b). This evaluation takes partial credit for soluble boron in the SFPs.

Minimum bumup requirements were developed for fuel to be placed without location restrictions in the NUHOMSo-24P/24PHB DSC. These bumup requirements, applicable for eligible fuel assemblies with a minimum 5 years post-irradiation cooling time, are a function of initial U-235 enrichment.

In the DSC criticality analysis, the maximum 95/95 keff with no boron in the DSC submerged in the Oconee SFP was calculated to be 0.9980. The criticality evaluation also confirmed that with 430 ppm of partial soluble boron credit, the maximum 95/95 keff of 0.9264 remains well below the regulatory requirement that the maximum 95/95 kff be less than 0.95 for all normal conditions. Additionally, the criticality analysis demonstrated that the current minimum boron concentration required in the Oconee SFPs (2220 ppm) is adequate to maintain the maximum 95/95 kff below 0.95 for all credible accident scenarios associated with loading and unloading fuel assemblies into the NUHOMS&-24P/24PHB DSCs.

5.0 REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration Pursuant to 10 CFR 50.91, Duke has made the determination that this amendment request involves a No Significant Hazards Consideration by applying the standards established by the NRC regulations in 10 CFR 50.92. This ensures that operation of the facility in accordance with the proposed amendment would not:

1) Involve a significant increase in the probability or consequences of an accident previously evaluated.

The applicable accidents are the dropped fuel assembly and drop of the 100 ton spent fuel cask into the SFP. This amendment request does not change the fuel assemblies or any of the Part 50 structures, systems, or components 1 Reference Enclosure 3

- Evaluation of Proposed Change License Amendment Request No. 2005-009 March 1, 2006 Page 4 involved in fuel assembly or cask handling or any of the operations involved.

Therefore, this amendment request does not affect the probability of an accident previously evaluated.

The proposed change does not increase the consequences of an accident previously evaluated for the following reasons: there is no increase in radiological source terms for the fuel; there is no change to the SFP water level; subcriticality is maintained for normal and accident conditions for the spent fuel storage racks and for cask loading and unloading; and the same boron concentrations that were previously credited for the spent fuel storage racks are assumed in the criticality analysis for cask loading and unloading.

2) Create the possibility of a new or different kind of accident from any accident previously evaluated.

Handling of fuel assemblies and the NUHOMS spent fuel cask have been previously evaluated for Oconee. These activities lead to evaluation of the fuel handling accident (dropped fuel assembly) and drop of the 100 ton spent fuel cask onto spent fuel stored in the Oconee SFP. These elements of the license amendment request (LAR) are not new, and thus do not create the potential for new or different kinds of accidents.

The new element of this LAR is the application of additional criticality controls (i.e., minimum burnup requirements for the fuel assemblies) beyond the 10 CFR 72 controls already in place for the NUHOMS,> spent fuel cask.

However, application of such criticality controls is not a new activity for Oconee, since similar criticality controls are currently applied to the spent fuel storage racks. Fuel assembly misloading is not a new accident; as discussed in, Section 6.5, fuel assembly misloading has been considered previously for the NUHOMS spent fuel cask and for the Oconee spent fuel pool racks. Furthermore, the criticality analysis for cask loading and unloading evaluates the same boron concentrations, moderator temperatures, and misloading scenario as previously evaluated for the spent fuel storage racks. The analysis demonstrates that a criticality accident does not occur under these conditions. It is concluded that the possibility of a criticality accident is not created since application of criticality controls is not new and the analysis demonstrates that criticality does not occur. More generally, this supports the conclusion that the potential for new or different kinds of accidents is not created.

- Evaluation of Proposed Change License Amendment Request No. 2005-009 March 1, 2006 Page 5

3) Involve a significant reduction in a margin of safety.

This LAR involves the application of additional criticality controls (minimum burnup requirements) to the 10 CFR 72 controls already in place for the NIJHOMS spent fuel cask. The criticality analysis demonstrates subcriticality margins are maintained for normal and accident conditions consistent with 10 CFR 50.68(b) and other NRC guidance. Margins previously established for Oconee's spent fuel storage racks are not altered.

Therefore, this LAR does not result in a reduction in a margin of safety.

5.2 Applicable Regulatory Reguirements/Criteria 5.2.1 Spent Fuel Pool Storage (10 CFR 50):

As described in UFSAR Section 9.1.2.3.2, "Criticality Analysis,"

criticality of fuel assemblies outside the reactor is precluded by adequate design of fuel transfer, shipping and storage facilities and by administrative control procedures. The two principal methods of preventing criticality are limiting the fuel assembly array size and limiting assembly interaction by fixing the minimum separation between assemblies and/or inserting neutron poisons between assemblies.

The design basis for preventing criticality outside the reactor is that, considering possible variations, there is a 95 percent probability at a 95 percent confidence level that the effective multiplication factor (keff) of the fuel assembly array will be less than or equal to 0.95, with partial credit for soluble boron.

6.0 ENVIRONMENTAL CONSIDERATION

Duke has evaluated this license amendment request against the criteria for identification of licensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.21. Duke has determined that this license amendment request meets the criteria for a categorical exclusion set forth in 10 CFR 51.22(c)(9). This determination is based on the fact that this change is being proposed as an amendment to a license issued pursuant to 10 CFR 50 that changes a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or that changes an inspection or a surveillance requirement, and the amendment meets the following specific criteria.

- Evaluation of Proposed Change License Amendment Request No. 2005-009 March 1, 2006 Page 6 (i)

The amendment involves no significant hazards consideration.

As demonstrated in Section 5.1, this proposed amendment does not involve significant hazards consideration.

(ii) There is no significant change in the types or significant increase in the amounts of any effluent that may be released offsite.

Additional criticality safety requirements do not impact effluents. Therefore, there will be no significant change in the types or significant increase in the amounts of any effluents released offsite.

(iii) There is no significant increase in individual or cumulative occupational radiation exposure.

Additional criticality safety requirements do not impact individual or cumulative occupational radiation exposure. Therefore, there will be no significant increase in individual or cumulative occupational radiation exposure resulting from this change.

ENCLOSURE 3 OCONEE NUCLEAR SITE NUHOMS-24P/24PHB DSC CRITICALITY ANALYSIS

- NUHOMS-24P/24PHB DSC Criticality Analysis License Amendment Request No. 2005-09 March 1, 2006 Page 1 1 Introduction This analysis examines the criticality aspects of fuel storage in the NUHOMS-24P and NUHOMS-24PHB dry storage canisters (DSCs), during loading and unloading operations in the spent fuel pools (SFPs) at Oconee Nuclear Site (Oconee). The analysis is intended to address the concerns documented by the Nuclear Regulatory Commission in Regulatory Issue Summary 2005-05 (Reference 1).

The objective of the DSC criticality evaluation is to demonstrate that eligible fuel assemblies enriched up to 5.0 wt % U-235 may be placed without location restrictions in the DSCs during loading/unloading operations in the Oconee SFPs, if specific requirements for minimum burnup, fuel assembly design, and cooling time are met.

The DSC criticality analysis examines the NUHOMS systems in use at Oconee, to show that the placement of irradiated fuel assemblies in these DSCs complies with the requirements of 10 CFR 50.68(b). In accordance with that regulation, the DSC criticality evaluation must show subcriticality in unborated water, but may take partial credit for soluble boron in the SFP water to achieve a keff less than 0.95. The current licensing basis for fuel assembly storage in the Oconee SFPs (Reference 2) allows 430 ppm soluble boron credit for normal conditions.

2 NUHOMS-24P/24PHB DSC Description Three variations of the NUHOMS-24P DSC design have been or are planned to be employed at Oconee. These include the NUHOMS&-24P (site-specific license), the NUHOMS-24P (general license), and the NUHOMS-24PBB (general license). Table 1 provides the nominal design data for the NUHOMS-24P/24PHB DSC components.

Note that the only significant difference among these DSC designs, from a criticality modeling perspective, is with the guide sleeves, which are full-axial length stainless steel storage cells in the DSCs. The NUHOMSv-24P (site-specific license) design has a reduced thickness for twelve of its outer guide sleeves as compared with the other two DSC designs. This design difference is evaluated in Section 6.3.

Figure 1 depicts a cross-sectional slice of the NUHOMSOD-24P/24PBB DSC, through one of the 2-inch-thick spacer disks that hold the guide sleeves in place. During loading/unloading operations, the DSC sits inside of a transfer cask comprising several inches of stainless steel, lead, carbon steel, and a cementatious neutron poison material.

The transfer cask effectively isolates the fuel assemblies placed in the DSC from any neighboring fuel stored in the Oconee SFP racks. DSC loading/unloading takes place in the cask pit area of the SFP. The cask pit is adjacent to the spent fuel storage racks in each of the Oconee SFPs, and is open to the rest of the SFP at all times.

- NUHOMS-24P/24PHB DSC Criticality Analysis License Amendment Request No. 2005-09 March 1, 2006 Page 2 Ligament Thickness A = 1.25 in Ligament Thickness B = 1.00 in Ligament Thickness C = 0.75 in Spacer Disk Hole Width D = 9.28 in (square)

Support Rod Diameter E = 3.25 in Figure 1. Oconee NUHOMS-24P/24PHB DSC Radial Geometry (Spacer Disk Detail)

- NUHOMS,&-24P/24PHB DSC Criticality Analysis License Amendment Request No. 2005-09 March 1, 2006 Page 3 Table 1. Design Data for NUHOMS-24P and -24PHB DSCs and Transfer Cask Parameter Nominal Dimension (inches)

Stainless Steel (SS) Guide Sleeve ID 8.90 Guide Sleeve Thickness 0.105*

SS Support Rod OD 3.25 Support Rod (x,y) coordinates 23.95, 13.92 Ligament (Spacing) Thicknesses 1.25, 1.00, 0.75 Spacer Disk Hole ID 9.28 (square)

Distances to Guide Sleeve Hole Centers from DSC radial origin (x 5.265, 15.545, 25.575 or y coordinates)

Spacer Disk OD 65.50 Axial distance between 21.1 Spacer Disks SS Spacer Disk Thickness 2

DSC SS Shell Thickness 0.625 DSC Shell OD 67.19 Transfer Cask SS inner shell ID 68.00 Transfer Cask Pb shield ID 69.00 Transfer Cask Carbon Steel Support 76.00 Shell ID Transfer Cask Neutron Shield ID 79.00 Transfer Cask SS outer shell ID 85.00 Transfer Cask SS outer shell OD 85.75

-- note that the NUHOMS,-24P site-specific design originally used at Oconee had a reduced guide sleeve thickness (0.06 inches) for twelve (12) of its outer guide sleeves

- NUHOMS-24P/24PHB DSC Criticality Analysis License Amendment Request No. 2005-09 March 1, 2006 Page 4 3 Fuel Assembly Designs Considered The following Babcock & Wilcox (B&W) 15x15 fuel types that have been employed at Oconee are eligible to be loaded in the NUHOMS&-24P or NUHOMS<-24PHB DSCs: MkB2, MkB3, MkB4, MkB4Z, MkB5, MkB5Z, MkB6, MkB7, MkB8, MkB9, MkB1OD, MkB1OE, MkBlOF, MkBlOG, MkBIOL. Though other fuel assembly designs have been irradiated in the Oconee reactors, they are not allowed to be loaded in the NUHOMS-24P or NUHOMSe-24PHB DSCs - in accordance with Reference 3 - and thus are not considered in this analysis.

Each of the eligible fuel assembly types can be classified as one of three bounding (in terms of criticality parameters) generic designs with the following shorthand names: mbz, mbf, and mbl. The important design data for these generic fuel types are listed in Table 2.

To maximize fuel assembly kff, it is assumed that irradiated fuel assemblies to be placed in the NUHOMS-24P/24PHB DSCs contained discrete A1203-B4C burnable poison rod assemblies (BPRAs) with the highest feasible B4C loading during operation in the Oconee reactors. The Reference 4 analysis shows that higher BPRA boron concentrations yield greater fuel assembly khffs once the BPRAs are removed from the assemblies after reactor irradiation.

The design data for the bounding Oconee BPRAs are provided in Table 3.

Table 2. Design Data for Generic Fuel Assembly "Types" Storage in NUHOMS,-24P/24PHB DSCs Eligible for Parameter mbz mbf mbl Average U0 2 fuel density (g/cc) 10.28 10.21 10.53 Fuel Pellet OD (inches) 0.3686 0.3700 0.3735 Cladding ID (inches) 0.377 0.377 0.380 Cladding OD (inches) 0.430 0.430 0.430 Cladding Material Zircaloy Zircaloy Zircaloy Fuel Pin Pitch (inches) 0.568 0.568 0.568 Fuel Pin Array Size 15x15 15x15 15x15 Guide Tube ID (inches) 0.498 0.498 0.498 Guide Tube OD (inches) 0.530 0.530 0.530 MkB4Z, MkB5, MkB2, MkB3, MkB 0l, Specific 15x15 MkB5Z, MB6:

MvkB4, MkB9, MkB lOG, Fuel Designs Represented MB7, MkB8

MkB1OD, MkB1OL MkBIOE

- NUHOMS-24P/24PHB DSC Criticality Analysis License Amendment Request No. 2005-09 March 1, 2006 Page 5 Table 3. Design Data for Bounding BPRAs Inserted into Fuel Assemblies During Oconee Reactor Irradiation Parameter Value Poison Pellet Density (g/cc) 3.38 Poison Pellet Diameter (inches) 0.340 Blo concentration (wt %)

0.5726 BII concentration (wt %)

2.5588 C concentration (wt %)

0.8686 Al concentration (wt %)

50.807 o concentration (wt %)

45.193 Number of rodlets (fingers) in BPRA 16

- NUHOMSz-24P/24PHB DSC Criticality Analysis License Amendment Request No. 2005-09 March 1, 2006 Page 6 4 Criticality Computer Code Validation The main neutronics codes employed in the NUHOMS&-24P/24PHB DSC criticality analysis are SCALE 4.4/KENO V.a and CASMO-3/SLMIULATE-3. These codes are well-suited to wet fuel storage criticality applications, and have been extensively benchmarked to critical experiments and reactor operational data. KENO V.a is a 3-D Monte Carlo criticality module in the SCALE 4.4 (Reference 5) package. CASMO-3 (Reference 6) is a 2-D transport code that performs fuel criticality and depletion calculations, using a 70-group cross-section library that is based on ENDF/B-IV.

CASMO-3 also produces nodal macro-group cross-sections that can be used by SIMULATE-3 (Reference 7), its counterpart 3-D nodal diffusion code, for applications involving arrays of fuel assemblies with varying enrichments or burnups.

SCALE 4.4/KENO V.a is used for explicit evaluation of the 3-D NUHOMS-24P/24PHB DSC described in Section 2. This analysis involves unirradiated fuel assemblies loaded in the DSC. The SCALE 4.4/KENO V.a computations are performed to confirm the conservatism of a simplified uniform-array DSC model, and are described and documented in Section 6.3.

CASMO-3/SIMULATE-3 is used for all DSC irradiated-fuel cases because this is the only code system qualified by Duke Power to perform criticality analyses using burnup credit. Note that KENO V.a is capable of doing calculations for burned fuel, using isotopic data produced via the SAS2H module of SCALE 4.4. However, because SAS2H (which was not originally intended for fuel criticality applications) is a 1-D transport code, it is preferable to use a more explicit 2-D transport code such as CASMO-3 for irradiated fuel evaluations. 2-D calculations should more accurately model fuel assemblies that are not radially uniform, such as the fuel types described in Section 3 that contain BPRAs during initial reactor irradiation.

The following subsections discuss the benchmarking validation that has been performed for both SCALE 4.4/KENO V.a and CASMO-3/SIMULATE-3. Note that the same code benchmarking results were employed in the McGuire SFP amendment request approved per Reference 16. Given the types of critical experiments with which these code systems have been validated (low-enriched uranium fuel rod lattices with configurations similar to those of fuel assemblies in SFP storage), the use of these code packages is appropriate for the NUHOMS-24P/24PHB DSC criticality evaluations.

- NUHOMSO-24P/24PHB DSC Criticality Analysis License Amendment Request No. 2005-09 March 1, 2006 Page 7 4.1 Validation of Benchmark Critical Experiments for SCALE 4.4/KENO V.a Duke Power has performed a SCALE 4.4/KENO V.a benchmark analysis of a large number of critical experiments to determine calculational biases and uncertainties for both the 44-group and 238-group cross-section libraries included with the SCALE 4.4 package.

For NUHOMS-24P/24PHB DSC criticality applications, the SCALE 4.4/KENO V.a benchmark biases and uncertainties are based on 58 critical experiments carried out by Pacific Northwest Laboratories (see References 8 to 10). The critical experiments evaluated cover a wide range of enrichment (2.35 and 4.31 wt % U-235), and include both over-and under-moderated lattices.

The results from the benchmark analyses indicate that the 238-group cross-section library yields the more consistent results (i.e., smaller variations in reactivity bias) across the ranges of moderation and enrichment considered. Therefore, the 238-group cross-section library is used for all the SCALE 4.4/KENO V.a computations performed in the DSC criticality analysis.

The 58 experiments used in the benchmarking resulted in a calculational bias of +0.0064 Ak and an uncertainty of +/-0.0066 Ak. These biases and uncertainties will be used in determining the total bounding 95/95 system kffs for each DSC configuration modeled with SCALE 4.4/KENO V.a.

4.2 Validation of Benchmark Critical Experiments for CASMO-3/SIMULATE-3 For all of the irradiated-fuel criticality evaluations for the NUHOMSD-24P/24PHB DSC, the CASMO-3/SIMULATE-3 code set is used. All CASMO-3 calculations will be carried out with the fine-energy-group (70-group) neutron cross-section library available with that code. Duke Power has performed a benchmark analysis of 10 B&W critical experiments with CASMO-3 and SIMULATE-3. These B&W critical experiments (Reference 11) were specifically designed for reactivity benchmarking purposes. Results from the 10 B&W critical benchmark cases yielded a calculational bias of -.0015 Ak (average over-prediction of keff) and an uncertainty of +/-0.0121 Ak. Even though CASMO-3/SIMULATE-3 tends to over-predict kff, the negative bias will be conservatively ignored. The uncertainty, however, will still be used in computing the overall 95/95 kffs for the DSC irradiated-fuel storage cases described in Section 6.5.

- NUHOMSO-24P/24PHB DSC Criticality Analysis License Amendment Request No. 2005-09 March 1, 2006 Page 8 5 Computation of the Maximum 95/95 kff For each fuel assembly design, enrichment, and burnup combination that is considered in the scope of the NUHOMS-24P/24PHB DSC criticality analyses, a nominal kff is calculated. This keff is only the base value, however. A total keff is determined by adding several pertinent reactivity biases and uncertainties, to provide an overall 95 percent probability, at a 95 percent confidence level (95/95), that the true system keff does not exceed the 95/95 keff for that particular storage condition.

The total 95/95 keff equation has the following form:

keff = knomilnal +

B

+

ks 2 where:

knon-dnal is the keff computed for the nominal case being considered.

B,,

is a pertinent bias, as indicated in Table 4.

ksx is the pertinent 95/95 independent uncertainty on knoninal, as indicated in Table 4.

Table 4 lists the various biases and uncertainties that are considered in the NUHOMSO-24P/24PHB DSC criticality analyses. Each of these biases and uncertainties is discussed in more detail below:

Benchmark Method Bias This bias is determined from the benchmarking of the code system used (SCALE 4.4/KENO V.a or CASMO-3/SIMULATE-3), and represents how much the code system is expected to overpredict (negative bias) or underpredict (positive bias) the "true keff of the physical system being modeled. The critical experiment benchmarks for these codes are discussed in Sections 4.1 and 4.2. The bias for SCALE 4.4/KENO V.a with its 238-group cross-section library is +0.0064 Ak. The bias for CASMO-3/

SIMULATE-3 with its 70-group cross-section library is -0.0015 Ak. Note that negative biases are conservatively ignored in this calculation, as recommended in Reference 12.

- NUTHOMSe-24P/24PHB DSC Criticality Analysis License Amendment Request No. 2005-09 March 1, 2006 Page 9

Axial Burnup Bias Section 6.4 discusses the method for determining the reactivity bias associated with the difference between the system keff calculated using an average 2-D fuel assembly bumup, and the k~ff using the 3-D axial burnup distribution for that assembly. Section 6.4 analyzes these keff differences for a bounding set of fuel assemblies and calculates an axial burnup bias as a linear function of assembly-average burnup. This bias is of the form:

Ak = 0.00105912*BU - 0.02189 where BU is assembly-average burnup, in gigawatt-days/tonne uranium (GWD/MTU). Any calculated biases that are negative are conservatively ignored.

Benchmark Method Uncertainty This uncertainty is determined from the benchmarking of the code system used (SCALE 4.4/KENO V.a or CASMO-3/SIMULATE-3), and is a measure of the expected variance (95/95 one-sided uncertainty) of predicted reactivity from the "true keff" of the physical system being modeled. The critical experiment benchmarks for these codes are discussed in Sections 4.1 and 4.2. The method uncertainty for SCALE 4.4/KENO V.a with its 238-group cross-section library is +/-0.0066 Ak. The uncertainty for CASMO-3/SIMULATE-3, with its 70-group cross-section library, is

+/-0.0121 Ak.

Monte Carlo Computational Uncertainty For the SCALE 4.4/KENO V.a computations performed in this analysis to determine 95/95 kffs, the Monte Carlo computational uncertainty is equal to 1.727* no,jnai. The anominal factor is the calculated standard deviation of knominai (the nominal keff for that particular case). The 1.727 multiplier is the one-sided 95/95 tolerance factor for 1000 neutron generations (Reference 18). Each of the SCALE 4.4/KENO V.a cases modeled in the DSC criticality analysis counted 1000 neutron generations.

Mechanical Uncertainties The "mechanical uncertainties" represent the total reactivity contributions of various independent DSC-related and fuel manufacturing-related uncertainties. These include reactivity effects for possible variations in fuel enrichment (+0.05 wt % U-235), fuel pellet diameter, fuel density, cladding dimensions, DSC guide sleeve thickness, DSC cell center-to-center spacing, and fuel assembly positioning within the DSC guide sleeve. The following bounding total mechanical uncertainties have been determined for the NUHOMS-24P/24PHB DSC criticality analyses:

+/-0.0280 Ak (no boron in SFP water)

+/-0.0304 Ak (430 ppm boron in SFP water)

- NUHOMS-24P/24PHB DSC Criticality Analysis License Amendment Request No. 2005-09 March 1, 2006 Page 10

Burnup Computational Uncertainty This burnup-related uncertainty represents the ability of the CASMO-3/SIMULATE-3 codes to accurately determine the isotopic content, and hence kff, of irradiated fuel assemblies.

As a conservative alternative to determining this uncertainty directly, Reference 12 notes that "... In the absence of any other determination of the depletion uncertainty, an uncertainty equal to 5 percent of the reactivity decrement to the burnup of interest is an acceptable assumption." This approach is used for the burnup credit cases evaluated in Section 6.5, and the largest uncertainty calculated is applied to the total 95/95 kefffor all assemblies loaded into the DSCs that have minimum burnup requirements. The maximum burnup computational uncertainty determined in Section 6.5 is *0.0151 Ak.

Burnup Measurement Uncertainty This uncertainty represents the reactivity penalty associated with the difference between measured burnup and actual burnup. Measured bumups, which are used for Technical Specification verification, have various sources of instrumentation error that can contribute to overall measurement inaccuracies. Comparison of measured burnup data to core follow predicted burnups for a large sample of discharged Oconee fuel assemblies shows that a four (4) percent burnup measurement uncertainty is conservative. Note that Reference 17 suggests that a two (2) percent measurement uncertainty may be sufficient for pressurized-water reactor (PWR) fuel.

The largest Ak penalty associated with a 4 percent uncertainty among the burnup credit cases considered in Section 6.5 is applied to the total 95/95 keff computations for fuel requiring any amount of bumup. The bounding burnup measurement uncertainty calculated in this manner is :t0.0104 Ak.

- NUHOMSoD-24P/24PHB DSC Criticality Analysis License Amendment Request No. 2005-09 March 1, 2006 Page I 1 Table 4. Pertinent 95/95 Biases and Uncertainties to be Considered in the NUHOMS-24P/24P1B DSC Criticality Analysis Include for Include for SCALE 4.4/

CASMO-3/

Biases KENO V.a SIMULATE-3 Calculations?

Calculations?

Benchmark Method Bias Axial Burnup Bias Uncertainties Benchmark Method Uncertainty Monte Carlo Computational Uncertainty Mechanical Uncertainties V

V Burnup Computational Uncertainty Burnup Measurement Uncertainty V

- NUHOMSo-24P/24PHB DSC Criticality Analysis License Amendment Request No. 2005-09 March 1, 2006 Page 12 6 NUHOMSe-24P/24PHB Dry Storage Canister Criticality Analysis 6.1 General Analysis Requirements In order to address the concerns documented in RIS 2005-05 (Reference 1) for the placement of fuel assemblies in the NUHOMS-24P/24PHB DSCs, the following requirements of 10 CFR 50.68(b) are satisfied:

"... If credit is taken for soluble boron, the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with borated water, and the k-effective must remain below 1.0 (subcritical), at a 95 percent probability, 95 percent confidence level, if flooded with unborated water."

In addition, for evaluations of irradiated fuel, Reference 12 provides the following general criteria:

"A reactivity uncertainty due to uncertainty in the fuel depletion calculations should be developed and combined with other calculational uncertainties."

"A correction for the effect of the axial distribution in burnup should be determined and, if positive, added to the reactivity calculated for uniform axial burnup distribution."

6.2 DSC Criticality Analysis Assumptions / Bases The following assumptions and bases are employed for the NUHOMS-24P/24PHB DSC criticality evaluations:

1)

The main DSC criticality calculations with irradiated fuel are performed in two dimensions, using the CASMO-3 transport code. The conservative infinite-array DSC model described in Section 6.3 is used. Separate three-dimensional computations are performed with the SIMULATE-3 nodal code to determine appropriate axial burnup biases to apply to the two-dimensional CASMO-3 computational results - see Section 6.4. The 3-D SIMULATE-3 model includes 23 axial fuel zones, along with top and bottom axial reflectors containing a mix of water, steel, and Zircaloy. Reference 13 supports the assumption that using 23 axial fuel segments is sufficient to accurately capture the reactivity effects associated with axial variations in fuel burnup. Extensive historic core-follow axial burnup predictions are employed to determine a conservative axial bumup bias for use in the total 95/95 DSC keff calculations.

- NUHOMSO-24P/24PHB DSC Criticality Analysis License Amendment Request No. 2005-09 March 1, 2006 Page 13

2)

No credit is taken for any short-lived Xe-135 poisons in the fuel assemblies loaded in the DSCs, consistent with Reference 12.

3)

No credit is taken for fuel assembly spacer grids.

4)

No credit is taken for any BPRAs in the fuel assemblies loaded in the DSCs.

As with spacer grids, even depleted BPRAs act as a modest poison in unborated or low-borated SFP water.

5)

In order to ensure the most conservative isotopic content, and hence, kff, for irradiated fuel assemblies to be loaded into the NUHOMS-24P/24PHB DSCs, conservative depletion parameters are used for the Oconee reactor fuel burnup computations performed by CASMO-3. These parameters include a high average fuel temperature (1054 'F), high outlet moderator temperature (630 'F), high cycle-average soluble boron concentration (700 ppm), and one-cycle maximum BPRA exposure (25 GWD/MTU).

6)

Credit for the reactivity reduction associated with fuel burnup and 5 years of post-irradiation cooling time is employed for the DSC criticality analysis.

Reactivity reduction with cooling time is primarily attributable to Pu-241 decay (-14.3 yr half-life), and Gd-155 buildup (via Eu-155 decay with - 4.7 yr half-life).

7)

Partial soluble boron credit of 430 ppm in the Oconee SFP is taken in order to achieve a system kdff < 0.95. This is in accordance with the regulatory subcriticality criteria defined in 10 CFR 50.68(b), as well as the guidance provided in Reference 12. The 430 ppm boron credit is the amount allowed for normal conditions in the current Oconee SFP criticality licensing basis (Reference 2).

8)

For accident conditions in the DSC, the minimum Oconee SFP boron concentration, as specified in the Core Operating Limits Report and Technical Specification 3.7.12 (2220 ppm), is available. Per the double contingency principle (see Reference 12), it is allowable to assume that the minimum boron concentration is present in the event of an accident condition - such as a misloaded fuel assembly - in the DSC.

The assumptions and bases listed above indicate that the criticality analysis of the NUHOMSo9-24P/24PHB DSC is extremely conservative. The major sources of the large quantity of Ak margin include the simplified infinite-array DSC model, the mechanical and burnup-related uncertainties discussed in Section 5, the in-reactor depletion parameters, and the axial burnup bias. The combination of these conservatisms amounts to more than 0.05 Ak.

- NUHOMS-24P/24PHB DSC Criticality Analysis License Amendment Request No. 2005-09 March 1, 2006 Page 14 6.3 DSC Model Simplification for Burnup Credit Computations Note that the NUHOMS-24P/24PBB DSC, as depicted in Figure 1, has a variable radial spacing between fuel assembly storage positions. While this variable spacing is straightforward to model with SCALE 4.4/KENO V.a, it is not really feasible using the CASMO-3/SIMULATE-3 code system.

To allow CASMO-3/SIMULATE-3 to be used for burnup credit criticality analyses of the NUHOMSO-24P/24PHB DSC, a conservative model simplification is employed, with the following features:

Infinite array of fuel assemblies (perfect radial reflection)

Uniform storage cell pitch (10.28 inches, corresponding to intermediate spacing "B" in Figure 1)

Guide sleeves included in model (0.105-inch thickness per Table 1)

Elimination of all other DSC structural material, including spacer disks, support rods, DSC shell, and surrounding transfer cask Figure 2 illustrates the proposed infinite-array DSC model. To demonstrate the conservatism of this model for subsequent criticality evaluations, keff calculations are performed for both the full-detail radial DSC model in Figure 1 (using SCALE 4.4/KENO V.a) and the simplified DSC layout in Figure 2 (using both SCALE 4.4/KENO V.a and CASMO-3). These calculations are all performed using perfect axial reflection. The Figure 1 calculations are performed for both the "site-specific" NUHOMS-24P design as well as the general license NUHOMSO-24P/24PBB DSC.

The difference between these designs is discussed in Section 2.

Table 5 presents the results of the criticality evaluations of the full-detail and simplified DSC models, which were performed using unirradiated "mbl" fuel assemblies, as described in Table 2, in the DSC storage cells. Calculations were carried out for two different enrichments (2.0 and 5.0 wt % U-235) and SFP water temperatures (68 and 150 OF).

As the results in Table 5 demonstrate, the infinite-array Figure 2 DSC model is extremely conservative, with calculated keffs more than 0.04 Ak higher than those associated with the full-detail Figure 1 DSC. Such a large amount of conservatism provides assurance that the Figure 2 DSC model is appropriate for the CASMO-3/SIMULATE-3 burnup-credit calculations described in Sections 6.4 and 6.5.

The 2-D infinite-array keff values in Table 5 show good agreement between SCALE 4.4/KENO V.a and CASMO-3. Note that some of the relative conservatism in the CASMO-3 model is attributable to ignoring its negative method bias (see Section 5).

- NUHOMS-24P/24PHB DSC Criticality Analysis License Amendment Request No. 2005-09 March 1, 2006 Page 15 Assembly center-to-center spacing - 10.28 inches

' 1000 assemblies) was used for the evaluation. The assembly bumup distributions (in 23 axial nodes) were available from SIMULATE-3 Oconee core follow predictions. End-of-cycle axial bumup data from Oconee 1 Cycles 16-20, and Oconee 3 Cycles 15-18, were chosen for this analysis, as these cycles provided sufficient non-blanketed and blanketed fuel assembly burnup information for assessment.

CASMO-3 core depletion cases were carried out with the fuel assembly types listed in Table 2, for several different enrichments that bounded the core follow operational data. The conservative depletion parameters identified in Section 6.2 were used, and the fuel assemblies were cooled for 5 years following reactor irradiation.
A SIMULATE-3 nodal model of the NUHOMS-24P/24PHB DSC was constructed, using the nominal infinite-array CASMO-3 DSC data described in Section 6.3. Axial reflectors containing a mixture of water, Zircaloy, and steel were specified.
SIMULATE-3 cases were performed to calculate kffs for the core follow fuel assemblies compiled in the first step of this procedure. For a given fuel type considered, two SIMULATE calculations were carried out for each core follow fuel assembly: 1) using the real (predicted) axial bumup profile; and 2) using a flat axial profile, with each of the 23 nodes at the equivalent assembly-average bumup. These SIMULATE-3 cases showed that the "mbz" fuel type from Table 2 yielded the highest (most positive) axial bumup biases among the three fuel types considered.
The individual "mbz" fuel assembly keff differences between real and flat axial profile cases were computed, and these Akffs were then plotted as a function of burmup.
Figure 3 shows all of the Akeffs that were calculated in the last step of the above procedure. A 95/95 axial bumup bias as a linear function of enrichment is determined by drawing a line through the data points circled on Figure 3.
- NUHOMSO-24P/24PHB DSC Criticality Analysis License Amendment Request No. 2005-09 March 1, 2006 Page 18 The resulting linear equation for the bias is:
Ak = 0.00105912*BU - 0.02189 where BU is assembly-average bumup, in GWD/MTU.
Note that this 95/95 "bounding" line does not include a few of the data points in Figure 3.
Table A-31 in Reference 14 confirms that the "bounding" axial bias line in Figure 3 provides 95% probability, at a 95% confidence level, that the true axial burnup bias for a particular assembly does not exceed the value calculated per the above equation.
0.040 0.0250- --- -- -- ---- --- -- -- -- - -r- -- -- -- -- ------ -- -- -- - - --- ---
I I
I I
4 I
0.025- -- ---------
-- -r i earaxiala 0.010 -- __
-- -- -- -- ----- I-les-qae (0
.0 0
5 - -_
- -+
- -t_
l 0.005 (0.0 10 ) - -__ P^ __ __ _
(0.015) ___
(0.020)
Avg Assembly Burnup (GWD/MTU)
Figure 3. Individual Fuel Assembly Axial End Effect Biases
{from Oconee Core Follow Data}
- NUHOMSO-24P/24PHB DSC Criticality Analysis License Amendment Request No. 2005-09 March 1, 2006 Page 19 6.5 DSC Criticality Analysis Results Using the infinite-array DSC model validated in Section 6.3, the main criticality calculations were performed with CASMO-3, using the three generic fuel assembly designs described in Section 3. The normal-condition criticality calculations for the NUHOMS-24P/24PHB DSC were performed with no boron in the SFP water [to satisfy the 95/95 keff < 1.0 criterion of 10 CFR 50.68(b)I, and with 430 ppm of soluble boron credit (to satisfy the 95/95 kff < 0.95 criterion of the same regulation).
Table 6 documents the maximum kff results, at various enrichments and corresponding burnups, for the no-boron normal-condition CASMO-3 DSC criticality calculations.
These results are provided for the fuel assembly type (mbl) and SFP water temperature (150 'F) that yielded the highest kffs. The biases and uncertainties in this table are taken from Section 5. Note that the maximum computed 95/95 keff is 0.9980, meeting the subcriticality criterion identified in the previous paragraph.
Using the fuel assembly burnup requirements shown in Table 6, DSC calculations with 430 ppm boron yield the highest nominal kff (0.8645) at 5.0 wt % U-235 (as expected per the Reference 15 report). Applying the mechanical uncertainty documented in Section 5 for 430 ppm conditions (+/-0.0304 Ak), and assuming the remaining biases and uncertainties are the same as the no-boron values listed in Table 6, the maximum 95/95 keff for the NUHOMS-24P/24PHB DSC in 430 ppm water is 0.9264.
The minimum fuel assembly burnup requirements shown in Table 6 are plotted as a function of enrichment in Figure 4. Because the data points in this figure show a high degree of linearity (the coefficient of determination, or R2, is 0.9994), it is appropriate to perform a linear interpolation between neighboring data points in Figure 4, when determining the minimum burnup requirement for a fuel enrichment not specifically evaluated in Table 6.
Among the Reference 12 accident conditions that need to be considered, (abnormal water temperatures, water voiding, fuel assembly drop, misload, and placement immediately adjacent to the DSC), the fuel assembly misload is the worst-case event from a criticality perspective. The most severe type of misload is the placement of an unirradiated 5.0 wt
% U-235 fuel assembly in the DSC. Among all the fuel types that have been used at Oconee (including those not eligible to be stored in the NUHOMS-24P/24PHB DSCs),
conservative criticality computations - using the SCALE 4.4/KENO V.a code with the simplified infinite-array model described in Section 6.3 - show that a misloaded MkB 11 fuel assembly requires the most soluble boron, 630 ppm, to maintain DSC system keff below 0.95. This is still much less than the amount of soluble boron available for accident conditions (2220 ppm, as discussed in Section 6.2).
- NUHOMS0-24P/24PHB DSC Criticality Analysis License Amendment Request No. 2005-09 March 1, 2006 Page 20 Note that a fuel assembly misleading event in the SFP, whether it occurs in the fuel storage racks or a NUHOMSo-24P/24PHB DSC, is not a new type of accident.
Reference 2 mentions the misloading of a fresh 5.0 wt % U-235 assembly in the Oconee SFP storage racks as the most severe of the criticality accident scenarios. Reference 3 discusses the misloading of an unqualified high enrichment fuel assembly in the NUHOMS9-24P DSC as a postulated "off-normal" condition.
Table 6. Maximum 95/95 kffs for NUHOMS-24P/24PHB DSC (Infinite-Array CASMO-3/SIMULATE-3 DSC Model)
- Normal (non-accident) Conditions -
{bounding "mbl" fuel, unborated SFP water at 150 'F)
Enrichment (wt % U-235) 1.60 2.00 2.50 3.00 3.50 4.00 4.50 5.00 Burnup (GWD/MTU) 0 8.93 15.34 21.02 27.12 32.78 38.33 43.77 Nominal CASMO-3 kff 0.9673 0.9624 0.9625 0.9620 0.9556 0.9496 0.9437 0.9380 Benchmark Method Bias (0.0015)* (0.0015)* (0.0015)* (0.0015)* (O.0015)* (0.0015)* (0.0015)* (0.0015)*
Axial Burnup Bias 0
(0.0124)* (0.0056)*
0.0004 0.0068 0.0128 0.0187 0.0245 Benchmark Method Uncert 0.0121 0.0121 0.0121 0.0121 0.0121 0.0121 0.0121 0.0121 Mechanical Uncerts 0.0280 0.0280 0.0280 0.0280 0.0280 0.0280 0.0280 0.0280 Burnup Comp Uncert 0
0.0151 0.0151 0.0151 0.0151 0.0151 0.0151 0.0151 Burup Meas Uncert 0
0.0104 0.0104 0.0104 0.0104 0.0104 0.0104 0.0104 Total 95/95 krr 0.9978 0.9980 0.9980 1 0.9980 0.9980 0.9980 0.9980
-- negative bias conservatively ignored
- NUHIOMSM-24P/24PHB DSC Criticality Analysis License Amendment Request No. 2005-09 March 1, 2006 Page 21 9-0.
a)
CD CD L.
50-45-40-35 30 25 20 15 10 5-0-
A -
- L- -
-4
-4 L- -
I L -
-I-L----
I
-I
)O 2.50 3.00 3.50 4.00 Enrichment (wt % U-235) 4.50 5.00 Figure 4. NUHOMS-24P/24PHB DSC Minimum Burnup Requirements to meet 10 CFR 50.68(b) Criteria (Minimum 5 Years Post-Irradiation Cooling Time)
{MkB2-B8, MkB9, MkB1O Fuel Types)
- NUHOMS&-24P/24PHB DSC Criticality Analysis License Amendment Request No. 2005-09 March 1, 2006 Page 22 7 Conclusions The criticality analysis of the NUHOMS-24P/24PHB DSC, for loading and unloading operations in the Oconee SFPs, has been performed in accordance with the requirements of 10 CFR 50.68(b). This evaluation takes partial credit for soluble boron in the SFPs.
Minimum burnup requirements were developed for fuel to be placed without location restrictions in the NUHOMS-24P/24PHB DSC. These burnup requirements, applicable for eligible fuel assemblies with a minimum 5 years post-irradiation cooling time, are a function of initial U-235 enrichment.
In the DSC criticality analysis, the maximum 95/95 kff with no boron in the Oconee SFP was calculated to be 0.9980. This meets the no-boron 95/95 keff < 1.0 criterion in 10 CFR 50.68(b). The criticality evaluation also confirmed that with 430 ppm of partial soluble boron credit, the maximum 95/95 kff of 0.9264 remains well below the regulatory requirement that the maximum 95/95 keff be less than 0.95 for all normal conditions.
Finally, the criticality analysis demonstrated that the current minimum boron concentration required in the Oconee SFPs (2220 ppm) is adequate to maintain the maximum 95/95 keff below 0.95 for all credible accident scenarios associated with loading fuel assemblies into the NUHOMS,-24P/24PHB DSCs.
- NUHOMSe-24P/24PBB DSC Criticality Analysis License Amendment Request No. 2005-09 March 1, 2006 Page 23 8 References
1. "NRC Regulatory Issue Summary 2005-05, Regulatory Issues Regarding Criticality Analyses for Spent Fuel Pools and Independent Spent Fuel Storage Installations," U.S. NRC, March 23, 2005.
2.
"Oconee Nuclear Station Units 1, 2, and 3 Re: Issuance of Amendments (TAC NOs MB0894, MB0895, and MB0896)," Letter from L. Olshan (U.S.
NRC) to W. McCollum (Duke), April 22, 2002.
3.
Final Safety Analysis Report for the Standardized NUIHOMSo Horizontal Modular Storage System for Irradiated Nuclear Fuel, NUH-003, Revision 8, Transnuclear Inc., June 2004.
4.
Parametric Study of the Effect of Burnable Poison Rods for PWR Burnup Credit, NUREG/CR-6761, Oak Ridge National Laboratory, March 2002.
5.
SCALE 4.4 - A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation, NUREG/CR-0200 (Rev. 5),
CCC-545, Oak Ridge National Laboratory, March 1997.
6.
CASMO A Fuel Assembly Burnup Program, STUDSVIK/NFA-89/3, June 1993.
7.
SIMULATE Advanced Three-Dimensional Two-Group Reactor Analysis Code, STUDSVIKISOA-95/15, October 1995.
8. Criticality Experiments with Subcritical Clusters of 2.35 and 4.31 wt %
U-235 Enriched U02 Rods in Water at a Water to Fuel Volume Ratio of 1.6, PNL-3314, July 1980.
9.
Critical Separation Between Subcritical Clusters of 2.35 wt % U-235 Enriched U02 Rods in Water with Fixed Neutron Poisons, PNL-2438, October 1977.
10. Criticality Experiments to Provide Benchmark Data on Neutron Flux Traps, PNL-6205, June 1988.
11. Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel, B&W-1484-7, July 1979.
- NUHOMS-24P/24PHB DSC Criticality Analysis License Amendment Request No. 2005-09 March 1, 2006 Page 24
12. "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants," Memorandum from L. Kopp (NRC) to T. Collins (NRC), U.S. Nuclear Regulatory Commission, August 19, 1998.
13. Review of Axial Burnup Distribution Considerations for Burnup Credit Calculations, ORNUJTM-1999/246, Oak Ridge National Laboratory, March 2000.
14. Experimental Statistics, Handbook 91, Mary Gibbons Natrella, National Bureau of Standards, October 1966.
15. NUREGICR-6683, A Critical Review of the Practice of Equating the Reactivity of Spent Fuel to Fresh Fuel in Burnup Credit Criticality Safety Analyses for PWR Spent Fuel Pool Storage, J. Wagner and C.
Parks, Oak Ridge National Laboratory, prepared for the U.S. Nuclear Regulatory Commission, September 2000.
16. 'McGuire Nuclear Station, Units 1 and 2 Re: Issuance of Amendments (TAC NOs MC0945 and MC0946)," Letter from J. Shea (U.S. NRC) to G.
Peterson (Duke), March 17, 2005.
17. Determination of the Accuracy of Utility Spent-Fuel Burnup Records, EPRI TR-1 12054, Electric Power Research Institute, July 1999.
18. Factors for One-Sided Tolerance Limits and for Variables Sampling Plans, SCR-607, Sandia Corporation, March 1963.
ENCLOSURE 4 COMPLIANCE WITH 10 CFR 50.68(b)
- Compliance with 10 CFR 50.68(b)
License Amendment Request No. 2005-009 March 1, 2006 Page 1 The eight criteria in 10 CFR 50.68(b) are listed below, along with a discussion of how Oconee complies with each criterion. This information is being provided for information only purposes and may be revised in the future consistent with 10 CFR 50.59. Therefore, this information does not represent specific commitments:
(1)
Plant procedures shall prohibit the handling and storage at any one time of more fuel assemblies than have been determined to be safely subcritical under the most adverse moderation conditions feasible by unborated water.
All storage locations have been evaluated to ensure that fuel can be stored safely at all times, under the most adverse moderator conditions as specified by appropriate regulations. Controls and procedures are in place to ensure that fuel assemblies are only placed into an acceptable storage configuration as allowed by appropriate regulatory criteria. Procedures and other administrative controls for the handling of fuel assemblies ensure that the movement of a fuel assembly is performed safely and that the fuel assembly being moved remains subcritical even under the most adverse moderation conditions.
(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.
Oconee has no fresh fuel storage racks, so this criterion is not applicable.
(3)
If optimum moderation of fresh fuel in the fresh fuel storage racks occurs when the racks are assumed to be loaded with fuel of the maximum fuel assembly reactivity and filled with low-density hydrogenous fluid, the k-effective corresponding to this optimum moderation must not exceed 0.98, at a 95 percent probability, 95 percent confidence level.
This evaluation need not be performed if administrative controls and/or design features prevent such moderation or if fresh fuel storage racks are not used.
Oconee has no fresh fuel storage racks, so this criterion is not applicable.
(4)
If no credit for soluble boron is taken, the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with unborated water. If credit is taken for soluble boron, the k-effective of the spent fuel storage racks loaded
- Compliance with 10 CFR 50.68(b)
License Amendment Request No. 2005-009 March 1, 2006 Page 2 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.
Duke is currently taking credit for 430 ppm soluble boron in the Oconee SFPs, per Technical Specification (TS) 4.3.1.c and proposed TS 4.4.1.c.
For fuel assemblies in the Oconee spent fuel storage racks, the NRC Safety Evaluation for the current licensing basis ["Oconee Nuclear Station Units 1, 2, and 3 Re: Issuance of Amendments (TAC NOs MB0894, MB0895, and MB0896)," Letter from L. Olshan (U.S.
NRC) to W. McCollum (Duke), April 22,2002] noted the 10 CFR 50.68(b)(4) requirement, and approved the methodology Duke employed to meet the dual subcriticality criteria in this regulation.
For fuel assemblies to be loaded into the NUHOMSO-24P and NUHOMS-24PHB dry storage canisters (DSCs), in accordance with the proposed new TS 4.4 and TS 3.7.18 requirements, the Enclosure 3 DSC criticality analysis shows that the maximum 95195 kff with no boron in the Oconee SFP is 0.9980. The DSC criticality analysis also confirms that with 430 ppm soluble boron credit, the maximum 95/95 keff is 0.9264, well below the 0.95 criterion.
(5)
The quantity of SNM, other than nuclear fuel stored onsite, is less than the quantity necessary for a critical mass.
Excluding the nuclear fuel, there is a limited amount of SNM stored at various locations onsite at Oconee. The total quantity of non-fuel SNM (< 200 grams of fissile material) is below the amount for a critical mass defined in 10 CFR 70.4.
(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.
Fuel assemblies are stored and handled in areas of the plant discussed below. Radiation monitoring is provided for these areas to detect excessive radiation levels and will provide an alarm to alert personnel if a potential radiation hazard is present.
1.
Unit 1 and 2 Fuel Building; includes the fuel receiving area.
2.
Unit 1 and 2 Spent Fuel Pool; includes the cask loading pit, the decon area, the new fuel elevator, the fuel transfer tube area and the spent fuel storage area/racks.
3.
Unit 1 Reactor Building; includes the fuel transfer tube area, the reactor core and the refueling canal.
- Compliance with 10 CFR 50.68(b)
License Amendment Request No. 2005-009 March 1, 2006 Page 3
4.
Unit 2 Reactor Building; includes the fuel transfer tube area, the reactor core and the refueling canal.
5.
Unit 3 Fuel Building; includes the fuel receiving area.
6.
Unit 3 Spent Fuel Pool; includes the cask loading pit, the decon area, the new fuel elevator, the fuel transfer tube area and the spent fuel storage area/racks.
7.
Unit 3 Reactor Building; includes the fuel transfer tube area, the reactor core and the refueling canal.
Another area in which fuel assemblies are stored at Oconee is the ISFSI, which has been licensed in accordance with 10 CFR Part 72. As such, this area of Oconee is not addressed by this response.
(7)
The maximum nominal U-235 enrichment of the fresh fuel assemblies is limited to five (5.0) percent by weight.
This limit is provided in TS 3.7.13 and 4.3.
(8)
The FSAR is amended no later than the next update which § 50.7 1(e) of this part requires, indicating that the licensee has chosen to comply with § 50.68(b).
The Oconee UFSAR is updated on an annual basis and a submittal made to the NRC at the end of June. Following approval of this amendment request, applicable sections of the Oconee Nuclear Site UFSAR will be updated to fully credit 10 CFR 50.68(b) criteria no later than June 30, 2007.
ATTACHMENT 1 TECHNICAL SPECIFICATIONS - MARK UPS
TABLE OF CONTENTS 3.7.14 Secondary Specific Acivfty."_.I.............................
xj 7.14-1 3.7.16 Decay Trne for Fuel Assemblies in Spent Fuel Pool _
3.7.16 S
onlrd lRoom Area Cooling Systems (CRAGS)............................. 7.16 -1 mw~zv a
I
-. V-l 3.8.1A Sources -Opertig.-----
3.8.2 AC Sources -
3.3 DC Sources - Oporaing.
3.ZA DC Sources-Stown......................
U.8 Battery CeN Pararneters........
3.8.6 Vital Inverters - Operating _
8.87 Vital Iverters - Shutdown
_3.8.7 1
3.88 DiStrbutiOn Sstemrs -
3.9 DistrbutIon Systems -
Wown.............
39 ag.1 3.92 8.9A 3.9.5 3.0.7 32.7 3.10 8.10.1 1102 4.0 4.1 42 4.B 5.0 5.1 REFRJNG OPERATIONS......
BoronC._
Nuclear Instrurnenlalion Contawnment Penetations.
Decay Heat Removal (DHR) and Crt
- High Water Le1 Decay Heat Removal (DHR) and Chvulon - Low Water Le%
Fuel Transfer Canal Water Level Unborated Water Source Islatic STANDBY SHUTDOWN FACIUTY..
Standby Shfdown Flity (SSF, Stant S~hdown Feck S Ceo Pare_..ters DESIGN FEATURES Site Lwaon............
ReactorCore Fuel Stage AnAAIMITA1 VF vrna
+/-9.1-1
+/-82.1-1
.3.92-1 Mn Vahes
.3.9.6-1
+/-."8-1 I.39.7-1
+/-10.1-1 Q In Q.4
_P.
sv.
g-l dftnsn-I
__~~

. _. __ _...iJl

. I
-A.0-1
..A.0fi fLn-t
.u-
~Iwlldw%.
I a
.o.5.0-1
&nr tunAR L....V......
I OCONEE UNITS 1, 9, 8, 3 i\\,
Anendment Nos.
& s& I
Spent Fuel Pool Boron Concentration 3.7.12 3.7 PLANT SYSTEMS 3.7.12 Spent Fuel Pool Boron Concentration LCO 3.7.12 The spent fuel pool boron concentration Emit shall be within lmits.
APPUCIASILTY:
A.
Spent fuel pool boron oonmentaraon not wthin Imit.
_Nn NOTE-LCO 3S.0. Is not applicable.
A.1 Suspend Movement of fuel assembles In the spent fuel pool.
Irnnimdately AD A.2 Inlliate action to
=
store hueatel spenttuel pool boron I concerarion to Within I Knint.
OCOIYEE UNITS 1, 2, & 3 3.7.12-1 Amendment Nos.
W2 & 241
\\ 2 -/
Dry Spent Fuel Storage Cask Loading and Unloading I 3.7.18 1 3.7 PLANTSYSTEMS 3.7.18 Dry Spent Fuel Storage Cask Loading and Unloading LCO 3.7.18 The combination of initial enrichment, burnup and post-irradiation cooling time of each fuel assembly in a dry spent fuel storage cask shall meet the criteria of Table 3.7.18-1.
Whenever any fuel assembly is in a dry spent fuel storage cask located in the spent fuel pool.
APPLICABILITY:
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Requirements of the A.1
--- NOTE-----------
LCO not met.
LCO 3.0.3 is not applicable.
Initiate action to move Immediately the noncomplying fuel assembly to an acceptable storage location.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.18.1 Verify by administrative means the Prior to placing the fuel assembly initial enrichment, burnup, and post-into a dry spent fuel storage cask irradiation cooling time of the fuel for loading assembly is in accordance with Table 3.7.18-1.
AND Prior to placing a dry spent fuel storage cask into the spent fuel pool for unloading.
OCONEE UNITS 1, 2, & 3 3.7.1 8-1 Amnendment Nos. 3XX, 3XX, & 3XX I
Dry Spent Fuel Storage Cask Loading and Unloading l 3.7.18 1 Table 3.7.18-1 (page 1 of 1)
Minimum Qualifying Burnup versus Design Maximum Enrichment for Dry Spent Fuel Storage Cask Loading and Unloading Initial Design Minimum Maximum Assembly Enrichment Bumup (Weeioht% U-235)
(GWD/MTU) 1.60 (or less) 0 2.00 8.93 2.50 15.34 3.00 21.02 3.50 27.12 4.00 32.78 4.50 38.33 5.00 43.77 50 -
40 i
20 10 o
I ACCEPTABLE l
/ I UNACCEPTABLE I
2.00 2.50 3.00 3.50 4.00 INITIAL DESIGN MA)UMUM ENRICHMENT, %U-235 4.50 5.00 NOTES:
The Design Maximum enrichment indicated above is the nominal maximum enrichment of any fuel pin in the fuel assembly being considered. The as-built enrichment of a fuel assembly may exceed its specified Design Maximum by up to 0.05 wt % U-235 and still be loaded in accordance with the above burnup limits for that Design Maximum enrichment. The minimum burnup requirements indicated above are based on a minimum post-irradiation cooling time of 5 years.
Fuel which differs from those designs used to determine the requirements of Table 3.7.18-1 may be qualified by means of an analysis using NRC approved methodology to assure that ke is less than 1.0 with no boron and less than or equal to 0.95 with credit for soluble boron.
OCONEE UNITS 1, 2, & 3 3.7.1 8-2 Amendment Nos. 3XX, 3XX, & 3XX I
Design Features 4.0 4.0 DE8IGN FEATURES 4.3 Fuel Storage (continued)
b.
kjc < 1.01 fully 1ooded with unborated water, which Includes an allowance for uncertainties a5 described In Secdon 9.1 of he UFSAR;
c.
lcmdS 0.5If fullflooded with water borated to 430 ppm. which Includes an allowance for uncertainties as described In Sectlon 91 of the UFSAR. Mantalning the normal Vent fuel pool boron concentralion wMin the TS limits assures kd S 0.95 for any accident condition;
d.
A nominal 10.65 Inch center to center distance between fuel assemblies placed In spent fuel storage racks servIng Units I and P4
e.
A nomdnal 10.60 Ih center to center distance between fuel 4
assenmles placed In spent fuel strage raft servirg Unit 3;
f.
A norninal 25.75 Inch center to center spacing between fuel assembis placed In te fuel transfer canal.
4.3.2 5:apap The spent fuel storage pool Is designed and shall be maintained with a storage capacity Dmrted to no more than 1812 fuel assemblies In th spent fuel storage racks serving Units I and 2 and 825 f assemblies I the spent fuel storage racks serving Unit a In alddition, up to 4 assembees endwor faled fuel container may be stored In each fuel transfer canal when,tb canal Is at refuermg lovel. Spent fuel may also be stored In the Oconee Nuclear Statlon Idependent Spent Fuel Storage Instaflation.
OCONEE UNITS 1, 2, & 3 40-2 Arnendmtent Nos. 318
& :4-j
Design Features 4.0 4.0 DESIGN FEATURES 4.4 Dry Spent Fuel Storage Cask Loading and Unloading 4.4.1 Criticalitv Dry spent fuel storage cask loading or unloading in the spent fuel pool shall be maintained with:
a.
Fuel assemblies having a maximum nominal U-235 enrichment of 5.0 weight percent;
b.
ke < 1.0 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.1 of the UFSAR;
c.
kef < 0.95 if fully flooded with water borated to 430 ppm, which includes an allowance for uncertainties as described in Section 9.1 of the UFSAR. Maintaining the normal spent fuel pool boron concentration within the TS limits assures kff < 0.95 for any accident condition;
d.
Dry spent fuel storage cask designs limited to NUHOMS-24P or NUHOMS&-24PHB.
OCONEE UNITS 1, 2, & 3 4.0-3 Amendment Nos. 3XX, 3XX, & 3XX I
TABLE OF CONTENTS B 3.7 B 3.7.9 B 3.7.10 B 8.7.11 83.7.12 S 37.13 6 3.7.14 B 3.7.15 B3.7.16 Bmt B 3.8.1 B3.8A 83.8.8 B 3.8.6
-B&U7 B.8.8 B 3.8.9 PLANT SYSTEMS (coniinued)
Ontol Room Ventilation System (CRVS) Booeter Fans....
9
.,...._.B 37.9 Not....
1.........._.B7.0-1 zr-Spent Fu Pod WarLeve...
1...1 Spent Fuel Pool Boron o B 8.7.12-1 FuAssembly Storage...................
e._
8.7.13-1 Secondary Speofot ACoty....------............... ___.
,B 7.14-1 Deay Time for Fuel Assernbies In Spent Fuel Pool (SFP)........-.-.-.....
__.B 3.7.15-1 Control Room Area Coon Sytems CRACS).........
87.164
-arl 9b 1W
_s7__
AC Sources -Shutdoan.............
00 Sourcs - Operatw
._g.......w.................w__
DC Sourcs - Shutdow................
Batery Cell Parameters 9.8.1-1 S.8.3-1 8.8.6-11 947-11 M1_
3.8.8-1 t d.
vfl W
ierr Vital
_w.
niet f
&4Aw et.^,6
-'U
-B 6
a, vwzuubuuual "AYMMU10 L
%AN lstrbution Systems - Shutdown.
1.,_.O.-.C.C._4 63.9 B 3.9.1 B 3.92 B S.O.S B 3.9A 8 39Q.
B 3.9.6 83.9.7 B 3.10 B3.10.1 B S.10.2 REFUELING OPERATIONS_
Boron Conoerlon_......
NYtW M
uU.RWHnnL
..B 3.9.1-1
.B 3.9.1-1
.J 8.e2-1
..B 8.9.3-1 dl A a Containment Penetrations_..
Decay Heat Removal (
) an arcldation - High Water Le Decay Heat Renmoval W P"an CLadadon -Lo Water to Fuel Transfer Canal Water Levi Unborae Water Source 11o at STANDBY SHUTDOWN FACILITY.
Stanby Shutdown Fadility (S8i Stant Shutdown Facility (SSI Cell Paramrs.___..
zbwF_**as*ll~e*X.
~~~~~D.M..
n.Va..es....._
..B 3.9.6-1
.8.9.8-1
,.B 8.9.7-1
. 8.10.1 -1
.,B 3.10.1-1 B 3.10.2-1
-p OCONEE UNItTS 1, Z2 k 3 Iv Amendmrent Nos.
fflX sC 9
39.
Aw & VTI
Spent Fuel Pool Boron Concentration B 3.7.12 B 3.7 PLANT SYSTEMS B 3.7.12 Spent Fuel Pool Boron Concentration BASES BACKG ROUND 7v
1 -
Tecone? seant elstage ra. scon in Bor flex ne tron-abs rbing anels at su ound ch sto age ce on all ur si s (e ept forenip eral sid s). T e func on of t se Bo aflex p els is/o en ure t t th reactivi of t store fuel as embli is ma tainedithi equir$d limits Boraex, as anufa tured, s a silic n rubb r ma ial that tains powd r of bar n car e (B ) neutr n abso ing aterial.
Th BorafI pan s are eclose n a fo ed st less s eel wr pper s et tha is spo weld to the torag tube. T e wrap er sh et is b t t each nd to ompiet the e closur of the raflex anel.
he Borafl pane is con med i the p1 num ar betwe n the orage be and e wra er p1 a. Si e the rapper ate end sure i not se ed, spe t fuel eol war is f e to ci ulate Iough the plenu
. It ha been a erved,nat a r Bar flex re ives a gh gain dose rom th store adiate fuel 1010 ds) it n begi a degrad and c1 solve i the w t enviro ent.
us he B4 oison aterial ca be re oved, t ereby reduc g thefois worth f the B aflex shea.
Thiephano enon i docy aente in C Ge nc Lett 96-04, "B raflex egrad ion in S ntFu Pa Stora Racks/
o ad ess his de dation,he Oconee pent f I stora racks ave been na zad ta g credi or soluble ron as Ilowed i Refer ce 1.
Th met odolog ensure that the spe t fuel r ok multip cation f ctor, ke, is I ss tha or equa a 0.95 as re amie ed in A hANS 7.2-1 83 ef. 2) nd NR guidance (R. 3). T e spent at star e ra s re nalyza to allow torage of fu asse lies with nnichm nts u to a Imum ominal richment of.00 we ht perce (wt %)
rani 2j5 whil maintai ng k.< 0.95,ncludi uncerta' ties, tole anc iases, nd cred' for soluble b on. N te that th criticality anal sis accouts for a aximum as-bitt enrin ment tol rance of.05
% U-235. or exa hle, for a sp fied inimum de ign enric en of 5.00
% -235, an s-built enric ent uqto 5.05 w ight parc t is ocepta e.
oluble bo n credit is u d to off at uncert inties, tol anc Si and ff-ormal c ditions and t provide ubcriticaJmargin s h thathe sp nt fuel poo k0f < 0.95.
e solubi boron co centratia requlad to/
maint n kf. 0.95 u rden nor I conditi s is 430 m. I additin, su critic lity of the pa (kef < 1.0 is assur on a 95/ 5 bas with the pre ence of the s ubla boro in the I (excludi g cartin bur up-re ted uncertai ies descri d in the riticality avalysis. The riticalty OCONEE UNITS 1, 2, & 3 B 3.7.12-1 Amendment Nos. 3;f, 3,
& 3K X X
)9 C
Insert 1.
Each Oconee spent fuel pool (SFP) contains racks for fuel assembly storage and a cask pit area for loading assemblies into a NUHOMSO-24P/24PHB dry storage canister (DSC). Criticality analyses have been performed for both SFP rack storage and DSC loading/unloading operations, in accordance with the regulation (Ref. 1) and the guidance in References 2 and 3. The SFP and DSC criticality analyses each take credit for 430 ppm soluble boron during normal conditions, in order to achieve system keff S 0.95. This partial soluble boron credit is included in Specifications 4.3.1 c. (SFP storage racks) and 4.4.1 c. (DSC).
The SFP storage rack criticality analysis yields fuel assembly storage configuration requirements and associated minimum bumup values (as a function of initial U-235 enrichment), which are specified in LCO 3.7.13. The DSC criticality evaluation establishes minimum burnup requirements for the loading of fuel assemblies into a NUHOMSe-24P/24PHB DSC without location restrictions. The DSC burnup requirements are provided in LCO 3.7.18.
The minimum SFP boron concentration of 2220 ppm (per SR 3.7.12.1) allows sufficient time to detect and mitigate all credible boron dilution scenarios, well before the SFP boron concentration drops to 430 ppm. The minimum 2220 ppm boron is available for all accident conditions evaluated in the SFP rack and DSC criticality analyses, per the double contingency principle (Ref. 4).
(ii;411uel Pool Boron Concentration Sll H
1I~
t peA+
B 3.7.12 BiG D/
/nalys/
prgmed how itat~th acdeptanb critea fo 9{tcal tre gonued)
/
met rr Ithe ragy of fu~f as rr li e
whr/redits take,4 for re oivity
/ /
~einde to uel b fupin fri or e Bo
~lex n utron v~Orb
/
/ /
,aesndsoage c hnig
!aon, and erihn mim specie by APPLICABLE ntn eei v
SAFETY ANALYSES SPe~i fdlpog1 EafIpe g: th ccidt coptons/re tthe/rop of f
mf sfrbf on to o Kck 'throp~d a fuf ass ~rbly be ven r rSd)s(~kdesn pr cudes tis icoyition) ad,fedropp a fe asseily fetwe d
rc/ndleand the poc wll/owev o,,
for/
ac~dntfan b po~latd wwhcldres~ti increa,e in r~itvt i/te s qnt flstrePool, 1hfirs is/ dropo placqfent qfafue ssmfl int tegsk load!r, ara The/Geo>di a los of ngfmal/
coolir t Jesp fue po~ wa r whicW caus aninqfaeiitepo wat tneratf h t(r i ml^h aitdinio a fuf sefl n,
logdtoni wigh the resllton lotion, nrichmyn andun/r r~d safied g~e fourt is a Zrop of d eay~ load oft the Setfb
_ l1 lFoia irrencep lhd osty ted ccden t te do dbe cor tngey fcij discuss i n eNI N-i.-1 7 and tt~ April,98NF lett 3 e
,ef./4 can be plid.Tis Kates hat one't not r ~ired t:/sst e
/wnlikely, ireeet churn evento ensut proefo jainst a ¢rtclt pie% hsVor t,se pst~at acient' cldtir's the
,peseneoadinl sJbe orn in t spent fel pool>aeabv/
he 430 pf eie ofain
~in kf < p
.95 ur er norra stag
/
cnditi 5d) eafb assdmd1 a reati initial onditio, sinc o
assufng its ~ene wo be a stndunfi y eve".
Cs§clatior w re~eFor~pe to d~erminee ao ht of s lu le b l~rn rqurdpof the hi hest re ctnIt iCeas ca-ed bftee
/psuai
cjets, maintnH Q5 It wXs founv that a/pent fepoo o
conc nrationfo 222 pmwsuff iciet tomelti l"S.5trthe wrtcas tpst ulw criticaptyrelat dacent (te/
he ofrope ent). Sp ciicatn 3..2nsrfh sht fuel 50 citisdequa<< disso,>e boro t comn nsate f~rthe i reased/
eattcuedby theyi postuyie aco' ents.
/h inimu ybron /ocent ion limJ/esures/h SFP/ron/
cycetrtiv is aduate treet t sub-criti~lt rec ren)-o fuel foe nteSP^rthe rstlimi g accidef in the MF:Ack drop yoto fuelnteSP OCONEE UNITS 1, 2, & 3 B 3.7.12-2 Amendment Nos. 3,25,3g-d & 3?4 KKC X1 KX
Insert 2.
Reference 3 discusses several criticality accident conditions that should be considered in SFP storage rack criticality analyses. Applicable accidents for the Oconee SFP storage racks include: 1) drop of a fuel assembly on top of the SFP storage rack; 2) drop of a fuel assembly outside of the storage rack modules; 3) abnormal SFP water temperatures outside the normal temperature range; 4) the misloading of a fuel assembly in a storage cell for which restrictions on location, enrichment or burnup are not satisfied; and 5) the drop of a heavy load (transfer cask) onto the SFP storage racks (NUREG-0612). Of these SFP storage rack accidents, the heavy load drop event requires the largest amount of soluble boron (almost 2200 ppm) to maintain SFP ke
0.95.
The accident scenarios (Ref. 3) that are valid for the loading/unloading of a NUHOMSe-24P/24PHB DSC include: 1) drop of a fuel assembly on top of the DSC storage cells; 2) drop of a fuel assembly immediately outside of the transfer cask containing the DSC; 3) abnormal SFP water temperatures beyond the normal temperature range; and 4) the misloading of a fresh 5.0 wt % U-235 fuel assembly in one of the DSC storage cells. Of these DSC accidents, the misload event requires the largest amount of soluble boron (630 ppm) to achieve a system keff s 0.95.
Note that it is plausible to consider a loss of normal SFP cooling accident occurring in conjunction with a boron dilution event in the Oconee SFPs. In this unlikely scenario, with SFP water temperatures up to 212 QF, the largest concentration of soluble boron required to maintain system keff s 0.95 is 500 ppm (for the SFP storage racks). This amount of soluble boron is still much less than that remaining after the worst-case credible dilution event (825 ppm).
Therefore, maintaining the SFP boron concentration 2 2220 ppm per SR 3.7.12.1 ensures that keff s 0.95 for any accident conditions in the SFP storage rack or NUHOMSe-24P/24PHB DSC. This minimum boron concentration limit includes allowance for analytical, mechanical, and instrument measurement uncertainties.
The concentration of dissolved boron in the SFP satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii) (Ref. 5).
G;~uel Pool Boron Concentration 5f*
B 3.7.12 BASES LCO iis Fe~rn6ogtrai/
lifs frj spe tWfelol prervei t~ e as~sdPt'~~ usdin t anys/
th,Oteni Iacc--en seari
/$dscibd Foe ~iskoc~
t Of diSolv boo is theziilr nfetl rare gncev~ati~fr Ifile aseb ~rag and yove,>etyihr APPLICABILITY lTsLOalie hevul sentise oidesetfyl, ACTIONS A.1 and A.2 The Required Actions are modified by a Note indicating that LCO 3.0.3 does not apply.
OCONEE UNITS 1, 2, & 3 B 3.7.12-3 Amendment Nos. 3$, 3P6, & 3;4 Wk xx Y
Insert 3.
The minimum concentration of dissolved boron in the SFP (2220 ppm) preserves the assumptions used in the analyses of the potential accident scenarios described above. This minimum boron concentration ensures that the system kel for the SFP storage rack or the NUHOMSO-24P/24PHB DSC will remain below 0.95 for all credible criticality accident scenarios and boron dilution events.
Insert 4.
This LCO applies whenever fuel assemblies are stored in the SFP storage racks, or whenever fuel assemblies are being loaded into a NUHOMSO-24P/24PHB DSC in the SFP.
6R~Fuel Pool Boron Concentration B 3.7.12 BASES ACTIONS A.1 and A.2 (continued)
If moving irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action. If moving Irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operation. Therefore, inability to suspend movement of fuel assemblies is not a sufficient reason to require a reactor shutdown.
When the concentration of boron in theie ig is less than required, immediate action must be taken to the occurrence of an accident or to mitigate the consequences of an accident in progress.
This is achieved by immediately suspending the movement of the fuel assemblies. This does not preclude movement of a fuel assembly to a safe position. Immediate action is also required to initiate action to restore the SFP boron concentration to within limits.
SURVEILLANCE SR 3.7.12.1 SFP REQUIREMENTS This SR verifies that the concentration of boron in th e
is within the required limit. As long as this SR is met, the analyzed incidents are fully addressed. The 7 day Frequency is appropriate because no major replenishment of pool water Is expected to take place over a short period of time. The COLR revision process assures that the minimum boron concentration specified in the COLR bounds the limit specified by this SR.
REFERENCES
1.
J-1§416-P-e
2.
American Nuclear Society, "American National Standard Design Requirements for Light Water Reactor Fuel Storage Facilities at Nuclear Power Plants," ANSI/ANS-57.2-1983, October 7, 1983.
3.
Nuclear Regulatory Commission, Memorandum to Timothy Collins from Laurence Kopp, 'Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Ught Water Reactor Power Plants," August 19, 1998.
OCONEE UNITS 1, 2, & 3 B 3.7.12-4 Amendment Nos. 3P; 3?X' & 3;<
S KYC Xk
6;i)Fuel Pool Boron Concentration 45+
B83.7.12 BASES REFERENCES
4.
Double contingency principle of ANSI N1 6.1-1975, as (continued) specified in the April 14, 1978 NRC letter (Section 1.2) and implied in the proposed revision to Regulatory Guide 1.13 (Section 1.4, Appendix A).
5.
10 CFR 50.36.
OCONEE UNITS 1, 2, & 3 B 3.7.12-5 Amendment Nos. 3?(326, & 37f fiX XV4
p Dry Spent Fuel Cask Loading and Unloading B 3.7.18 3.7 PLANT SYSTEMS B 3.7.18 Dry Spent Fuel Storage Cask Loading and Unloading BASES BACKGROUND Fuel loading and unloading operations for the NUHOMSD-24P and NUHOMS-24PHB dry storage canisters (DSCs) take place in the cask pit area of the spent fuel pool. The cask pit is adjacent to the spent fuel storage racks in each of the Oconee spent fuel pools, and is open to the rest of the spent fuel pool at all times. The NUHOMS-24P and NUHOMS-24PHB DSCs contain storage cells for 24 fuel assemblies.
Eligible B&W 15x1 5 fuel assemblies (MkB2-B8, MkB9, and MkB1 0) with initial enrichments s 5.0 wt % U-235 may be stored in the NUHOMS-24P or NUHOMS-24PHB DSC, as long as the fuel assemblies meet the minimum burnup and cooling time requirements specified in Table 3.7.18-1.
For normal conditions in the spent fuel pool, the NUHOMS&-24P and NUHOMS-24PHB DSCs have been analyzed using credit for soluble boron as allowed in Reference 1. This ensures that the system multiplication factor, ke, is < 0.95 as recommended in ANSI/ANS-57.2-1983 (Ref. 2) and NRC guidance (Ref. 3). The DSC is analyzed to allow loading/unloading of eligible fuel assemblies while maintaining keff < 0.95, including uncertainties, tolerances, biases, and credit for 430 ppm soluble boron. Note that the criticality analysis accounts for a maximum as-built enrichment tolerance of 0.05 wt % U-235. For example, for a specified maximum design enrichment of 5.00 wt % U-235, an as-built enrichment up to 5.05 weight percent is acceptable. The 430 ppm soluble boron credit must provide sufficient subcritical margin to maintain the DSC k.e <
0.95. In addition, sub-criticality of the DSC (ke < 1.0) must be assured on a 95/95 basis, without the presence of any soluble boron in the spent fuel pool.
The dual k" criteria identified in the above paragraph are satisfied for fuel assemblies meeting the minimum burnup and post-irradiation cooling time requirements specified in Table 3.7.18-1. Reactivity reduction with cooling time is primarily attributable to Pu-241 decay and Gd-1 55 buildup (via Eu-1 55 decay).
Specification 4.4.1 c. requires that the DSC ke be < 0.95 when flooded with water borated to 430 ppm. A spent fuel pool boron dilution analysis has been performed that confirms that sufficient time is available to detect and mitigate a dilution of the spent fuel pool before the 0.95 kt OCONEE UNITS 1, 2, & 3 B 3.7.18-1 Amendment Nos. 3XX, 3XX, & 3XX I
Dry Spent Fuel Cask Loading and Unloading B 3.7.18 BASES BACKGROUND design basis is exceeded. The spent fuel pool boron dilution analysis (continued) concluded that an unplanned or inadvertent event which could result in the dilution of the spent fuel pool boron concentration to 430 ppm is not a credible event.
APPLICABLE Several accident conditions (Ref. 3) are considered that could result SAFETY ANALYSES in an increase in system klfl for a DSC being loaded or unloaded in the spent fuel pool. These accident conditions include the drop of a fuel assembly on top of the DSC storage cells, the drop of a fuel assembly just outside the transfer cask containing the DSC, a higher than normal spent fuel pool water temperature, and the misleading of a fresh 5.0 wt %
U-235 assembly in one of the DSC storage cells.
For an occurrence of these postulated accidents, the double contingency principle discussed in ANSI N-16.1-1975 and the April 1978 NRC letter (Ref. 4) can be applied. This double contingency principle does not require assuming two unlikely, independent, concurrent events to ensure protection against a criticality accident. Thus, for these postulated accident conditions, the presence of additional soluble boron in the spent fuel pool water (above the 430 ppm required to maintain kf < 0.95 under normal DSC loading/unloading conditions) can be assumed as a realistic initial condition since not assuming its presence would be a second unlikely event.
Calculations were performed to determine the amount of soluble boron required to offset the highest reactivity increase associated with these postulated accidents, in order to maintain kA
< 0.95. It was found that a spent fuel pool boron concentration of 630 ppm was sufficient to maintain kff < 0.95 for the worst-case postulated criticality-related accident (the fresh fuel assembly misloaded in a DSC storage cell). Specification 3.7.12 ensures the spent fuel pool contains adequate dissolved boron to compensate for the increased reactivity caused by these postulated accidents.
For normal storage conditions, Specification 4.3.1 c. requires that the spent fuel rack kff be < 0.95 when flooded with water borated to 430 ppm. A spent fuel pool boron dilution analysis was performed which confirmed that sufficient time is available to detect and mitigate a dilution of the spent fuel pool before the 0.95 kef design basis is exceeded. The spent fuel pool boron dilution analysis concluded that an unplanned or inadvertent event which could result in the dilution of the spent fuel pool boron concentration to 430 ppm is not a credible event.
OCONEE UNITS 1,2, &3 B 3.7.18-2 Amendment Nos. 3XX, 3XX, & 3XX I
ADD) fr I Dry Spent Fuel Cask Loading and Unloading B 3.7.18 BASES APPLICABLE SAFETY ANALYSIS (continued)
The configuration of fuel assemblies in the DSC and the concentration of dissolved boron in the spent fuel pool satisfy Criterion 2 of 10 CFR 50.36 (Ref. 5)
LCO The keff of the dry spent fuel storage cask (NUHOMS-24P or NUHOMSe-24PHB DSC), during loading and unloading operations in the spent fuel pool, will always remain < 0.95, assuming the spent fuel pool is flooded with water borated to at least 430 ppm, and that each loaded fuel assembly meets the initial enrichment, burnup, and post-irradiation cooling time of Table 3.7.18-1.
APPLICABILITY This LCO applies whenever any fuel assembly is in a dry spent fuel storage cask located in the spent fuel pool.
ACTIONS A.1 Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply.
If moving fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action. If moving fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operation. Therefore, in either case, inability to move fuel assemblies is not sufficient reason to require a reactor shutdown.
When the configuration of fuel assemblies loaded in the NUHOMS-24P or NUHOMS-24PHB DSC is not in accordance with the LCO, immediate action must be taken to make the necessary fuel assembly movement(s) to bring the configuration into compliance with the LCO.
SURVEILLANCE REQUIREMENTS SR 3.7.18.1 This SR verifies by administrative means that the initial enrichment, burnup, and post-irradiation cooling time of the fuel assembly to be loaded into or removed from the NUHOMSv-24P or NUHOMS-24PHB DSC is in accordance with Table 3.7.18-1.
OCONEE UNITS 1, 2, & 3 B 3.7.1 8-3 Amendment Nos. 3XX, 3XX, & 3XX I
BASES Dry Spent Fuel Cask Loading and Unloading B 3.7.18 l REFERENCES
1.
10 CFR 50.68(b)(4)
2.
American Nuclear Society, "American National Standard Design Requirements for Light Water Reactor Fuel Storage Facilities at Nuclear Power Plants," ANSI/ANS-57.2-1983, October 7, 1983.
3.
Nuclear Regulatory Commission, Memorandum to Timothy Collins from Laurence Kopp, "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light Water Reactor Power Plants," August 19,1998.
4.
Double contingency principle of ANSI N16.1-1975, as specified in the April 14, 1978 NRC letter (Section 1.2) and implied in the proposed revision to Regulatory Guide 1.13 (Section 1.4, Appendix A).
5.
10 CFR 50.36 OCONEE UNITS 1, 2, & 3 B 3.7.18-4 Amendment Nos. 3XX, 3XX, & 3XX I
ATTACHMENT 2 TECHNICAL SPECIFICATIONS-REPRINTED PAGES
TABLE OF CONTENTS 3.7.14 Secondary Specific Activity..............................................
3.7.14-1 3.7.15 Decay Time for Fuel Assemblies in Spent Fuel Pool (SFP)..........................................
3.7.15-1 3.7.16 Control Room Area Cooling Systems (CRACS)................................. 3.7.16-1 3.7.17 Spent Fuel Pool Ventilation System (SFPVS).................................... 3.7.17-1 3.7.18 Dry Spent Fuel Storage Cask Loading and Unloading....................... 3.7.18-1 3.8 ELECTRICAL POWER SYSTEMS..........................................
3.8.1-1 3.8.1 AC Sources - Operating..........................................
3.8.1-1 3.8.2 AC Sources - Shutdown..........................................
3.8.2-1 3.8.3 DC Sources - Operating..........................................
3.8.3-1 3.8.4 DC Sources - Shutdown..........................................
3.8.4-1 3.8.5 Battery Cell Parameters..........................................
3.8.5-1 3.8.6 Vital Inverters - Operating..........................................
3.8.6-1 3.8.7 Vital Inverters - Shutdown..........................................
3.8.7-1 3.8.8 Distribution Systems - Operating..........................................
3.8.8-1 3.8.9 Distribution Systems - Shutdown..........................................
3.8.9-1 3.9 REFUELING OPERATIONS..........................................
3.9.1-1 3.9.1 Boron Concentration..........................................
3.9.1-1 3.9.2 Nuclear Instrumentation..........................................
3.9.2-1 3.9.3 Containment Penetrations..........................................
3.9.3-1 3.9.4 Decay Heat Removal (DHR) and Coolant Circulation - High Water Level...
3.9.4-1 3.9.5 Decay Heat Removal (DHR) and Coolant Circulation - Low Water Level.................................
3.9.5-1 3.9.6 Fuel Transfer Canal Water Level.................................
3.9.6-1 3.9.7 Unborated Water Source Isolation Valves.................................
3.9.7-1 3.10 STANDBY SHUTDOWN FACILITY.................................
3.10.1-1 3.10.1 Standby Shutdown Facility (SSF).................................
3.10.1-1 3.10.2 Standby Shutdown Facility (SSF) Battery Cell Parameters..........................................
3.10.2-1 4.0 DESIGN FEATURES..........................................
4.0-1 4.1 Site Location..........................................
4.0-1 4.2 Reactor Core..........................................
4.0-1 4.3 Fuel Storage..........................................
4.0-1 4.4 Dry Spent Fuel Storage Cask Loading and Unloading............................... 4.0-1 5.0 ADMINISTRATIVE CONTROLS..........................................
5.0-1 5.1 Responsibility..........................................
5.0-1 OCONEE UNITS 1, 2, & 3 iv Amendment Nos. 3XX, 3XX, & 3XX I
Spent Fuel Pool Boron Concentration 3.7.12 3.7 PLANT SYSTEMS 3.7.12 Spent Fuel Pool Boron Concentration LCO 3.7.12 The spent fuel pool boron concentration limit shall be within limits.
APPLICABILITY:
When fuel assemblies are stored in the spent fuel pool and when fuel assemblies are in a dry spent fuel storage cask located in the spent fuel pool.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Spent fuel pool boron

NOTE-----------------

concentration not within LCO 3.0.3 is not applicable.
limit.
A.1 Suspend movement of Immediately fuel assemblies in the spent fuel pool.
AND A.2 Initiate action to restore Immediately spent fuel pool boron concentration to within limit.
OCONEE UNITS 1, 2, & 3 3.7.12-1 Amendment Nos. 3XX, 3XX, & 3XX I
Dry Spent Fuel Storage Cask Loading and Unloading l 3.7.18 1 3.7 PLANT SYSTEMS 3.7.18 Dry Spent Fuel Storage Cask Loading and Unloading LCO 3.7.18 The combination of initial enrichment, burnup and post-irradiation cooling time of each fuel assembly in a dry spent fuel storage cask shall meet the criteria of Table 3.7.18-1.
APPLICABILITY:
Whenever any fuel assembly is in a dry spent fuel storage cask located in the spent fuel pool.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Requirements of the A.1 NOTE----------
LCO not met.
LCO 3.0.3 is not applicable.
Initiate action to move Immediately the noncomplying fuel assembly to an acceptable storage location.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.18.1 Verify by administrative means the Prior to placing the fuel assembly initial enrichment, burnup, and post-into a dry spent fuel storage cask irradiation cooling time of the fuel for loading assembly is in accordance with Table 3.7.18-1.
AND Prior to placing a dry spent fuel storage cask into the spent fuel pool for unloading.
OCONEE UNITS 1, 2, & 3 3.7.1 8-1 Amendment Nos. 3XX, 3XX, & 3XX I
Dry Spent Fuel Storage Cask Loading and Unloading 3.7.18 Table 3.7.18-1 (page 1 of 1)
Minimum Qualifying Burnup versus Design Maximum Enrichment for Dry Spent Fuel Storage Cask Loading and Unloading Initial Design Minimum Maximum Assembly Enrichment Burnup (Weight% U-235)
(GWD/MTU) 1.60 (or less) 0 2.00 8.93 2.50 15.34 3.00 21.02 3.50 27.12 4.00 32.78 4.50 38.33 5.00 43.77 40 I30 j20 10 I ACCEPTABLE I T
UNACCEPTBLE 2.00 2.50 3.00 3.50 4.00 INITIAL DESIGN MAXIMUM ENRICHMENT, %U-235 4.50 5.00 NOTES:
The Design Maximum enrichment indicated above is the nominal maximum enrichment of any fuel pin in the fuel assembly being considered. The as-built enrichment of a fuel assembly may exceed its specified Design Maximum by up to 0.05 wt % U-235 and still be loaded in accordance with the above burnup limits for that Design Maximum enrichment. The minimum burnup requirements indicated above are based on a minimum post-irradiation cooling time of 5 years.
Fuel which differs from those designs used to determine the requirements of Table 3.7.18-1 may be qualified by means of an analysis using NRC approved methodology to assure that kO is less than 1.0 with no boron and less than or equal to 0.95 with credit for soluble boron.
OCONEE UNITS 1, 2, & 3 3.7. 18-2 Amendment Nos. 3XX, 3XX, & 3XX I
Design Features 4.0 4.0 DESIGN FEATURES 4.4 Dry Spent Fuel Storage Cask Loading and Unloading 4.4.1 Criticality Dry spent fuel storage cask loading or unloading in the spent fuel pool shall be maintained with:
a.
Fuel assemblies having a maximum nominal U-235 enrichment of 5.0 weight percent;
b.
krff < 1.0 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.1 of the UFSAR;
c.
kff < 0.95 if fully flooded with water borated to 430 ppm, which includes an allowance for uncertainties as described in Section 9.1 of the UFSAR. Maintaining the normal spent fuel pool boron concentration within the TS limits assures keff < 0.95 for any accident condition;
d.
Dry spent fuel storage cask designs limited to NUHOMS-24P or NUHOMSe-24PHB.
OCONEE UNITS 1, 2, & 3 4.0-3 Amendment Nos. 3XX, 3XX, & 3XX I
Spent Fuel Pool Boron Concentration I B 3.7.12 B 3.7 PLANT SYSTEMS B 3.7.12 Spent Fuel Pool Boron Concentration BASES BACKGROUND Each Oconee spent fuel pool (SFP) contains racks for fuel assembly storage and a cask pit area for loading assemblies into a NUHOMSO -
24P/24PHB dry storage canister (DSC). Criticality analyses have been performed for both SFP rack storage and DSC loading/unloading operations, in accordance with the regulation (Ref. 1) and the guidance in References 2 and 3. The SFP and DSC criticality analyses each take credit for 430 ppm soluble boron during normal conditions, in order to achieve system kff < 0.95. This partial soluble boron credit is included in TS 4.3.1 c. (SFP storage racks) and 4.4.1 c. (DSC).
The SFP storage rack criticality analysis yields fuel assembly storage configuration requirements and associated minimum burnup values (as a function of initial U-235 enrichment), which are specified in LCO 3.7.13.
The DSC criticality evaluation establishes minimum burnup requirements for the loading of fuel assemblies into a NUHOMSO -24P/24PHB DSC without location restrictions. The DSC burnup requirements are provided in LCO 3.7.18.
The minimum SFP boron concentration of 2220 ppm (per SR 3.7.12.1) allows sufficient time to detect and mitigate all credible boron dilution scenarios, well before the SFP boron concentration drops to 430 ppm.
The minimum 2220 ppm boron is available for all accident conditions evaluated in the SFP rack and DSC criticality analyses, per the double contingency principle (Ref. 4).
APPLICABLE Reference 3 discusses several criticality accident conditions that should SAFETY ANALYSES be considered in SFP storage rack criticality analyses. Applicable accidents for the Oconee SFP storage racks include: 1) drop of a fuel assembly on top of the SFP storage rack; 2) drop of a fuel assembly outside of the storage rack modules; 3) abnormal SFP water temperatures outside the normal temperature range; 4) the misloading of a fuel assembly in a storage cell for which restrictions on location, enrichment, burnup, or post-irradiation cooling time are not satisfied; and
5) the drop of a heavy load (transfer cask) onto the SFP storage racks (NUREG-0612). Of these SFP storage rack accidents, the heavy load drop event requires the largest amount of soluble boron (almost 2200 ppm) to maintain SFP kff < 0.95.
OCONEE UNITS 1, 2, & 3 B 3.7.12-1 Amendment Nos. 3XX, 3XX, & 3XX I
Spent Fuel Pool Boron Concentration B 3.7.12 BASES ACTIONS A.1 and A.2 (continued)
If moving irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operation. Therefore, inability to suspend movement of fuel assemblies is not a sufficient reason to require a reactor shutdown.
When the concentration of boron in the SFP is less than required, immediate action must be taken to preclude the occurrence of an accident or to mitigate the consequences of an accident in progress.
This is achieved by immediately suspending the movement of the fuel assemblies. This does not preclude movement of a fuel assembly to a safe position. Immediate action is also required to initiate action to restore the SFP boron concentration to within limits.
SURVEILLANCE SR 3.7.12.1 REQUIREMENTS This SR verifies that the concentration of boron in the SFP is within the required limit. As long as this SR is met, the analyzed incidents are fully addressed. The 7 day Frequency is appropriate because no major replenishment of pool water is expected to take place over a short period of time. The COLR revision process assures that the minimum boron concentration specified in the COLR bounds the limit specified by this SR.
REFERENCES
1.
10 CFR 50.68(b).
2.
American Nuclear Society, "American National Standard Design Requirements for Light Water Reactor Fuel Storage Facilities at Nuclear Power Plants," ANSI/ANS-57.2-1983, October 7, 1983.
3.
Nuclear Regulatory Commission, Memorandum to Timothy Collins from Laurence Kopp, "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light Water Reactor Power Plants," August 19,1998.
4.
Double contingency principle of ANSI N16.1-1975, as specified in the April 14,1978 NRC letter (Section 1.2) and implied in the proposed revision to Regulatory Guide 1.13 (Section 1.4, Appendix A).
OCONEE UNITS 1, 2, & 3 B 3.7.12-3 Amendment Nos. 3XX, 3XX, & 3XX
Spent Fuel Pool Boron Concentration B 3.7.12 BASES REFERENCES
5.
10 CFR 50.36.
(continued)
Amendment Nos. 3XX, 3XX, & 3XX I OCONEE UNITS 1, 2, & 3 B 3.7.12-4
Dry Spent Fuel Cask Loading and Unloading B 3.7.18 l B 3.7 PLANT SYSTEMS B 3.7.18 Dry Spent Fuel Storage Cask Loading and Unloading BASES BACKGROUND Fuel loading and unloading operations for the NUHOMS-24P and NUHOMS&-24PHB dry storage canisters (DSCs) take place in the cask pit area of the spent fuel pool. The cask pit is adjacent to the spent fuel storage racks in each of the Oconee spent fuel pools, and is open to the rest of the spent fuel pool at all times. The NUHOMSqD-24P and NUHOMS-24PHB DSCs contain storage cells for 24 fuel assemblies.
Eligible B&W 15x1 5 fuel assemblies (MkB2-B8, MkB9, and MkB1 0) with initial enrichments 5 5.0 wt % U-235 may be stored in the NUHOMS-24P or NUHOMS-24PHB DSC, as long as the fuel assemblies meet the minimum burnup and cooling time requirements specified in Table 3.7.18-1.
For normal conditions in the spent fuel pool, the NUHOMS-24P and NUHOMS-24PHB DSCs have been analyzed using credit for soluble boron as allowed in Reference 1. This ensures that the system multiplication factor, keff, is < 0.95 as recommended in ANSIIANS-57.2-1983 (Ref. 2) and NRC guidance (Ref. 3). The DSC is analyzed to allow loading/unloading of eligible fuel assemblies while maintaining kef< 0.95, including uncertainties, tolerances, biases, and credit for 430 ppm soluble boron. Note that the criticality analysis accounts for a maximum as-built enrichment tolerance of 0.05 wt % U-235. For example, for a specified maximum design enrichment of 5.00 wt % U-235, an as-built enrichment up to 5.05 weight percent is acceptable. The 430 ppm soluble boron credit must provide sufficient subcritical margin to maintain the DSC kff <
0.95. In addition, sub-criticality of the DSC (kft < 1.0) must be assured on a 95/95 basis, without the presence of any soluble boron in the spent fuel pool.
The dual keff criteria identified in the above paragraph are satisfied for fuel assemblies meeting the minimum burnup and post-irradiation cooling time requirements specified in Table 3.7.18-1. Reactivity reduction with cooling time is primarily attributable to Pu-241 decay and Gd-1 55 buildup (via Eu-1 55 decay).
Specification 4.4.1 c. requires that the DSC kff be < 0.95 when flooded with water borated to 430 ppm. A spent fuel pool boron dilution analysis has been performed that confirms that sufficient time is available to detect and mitigate a dilution of the spent fuel pool before the 0.95 kff OCONEE UNITS 1, 2, & 3 B 3.7.1 8-1 Amendment Nos. 3XX, 3XX, & 3XX I
Dry Spent Fuel Cask Loading and Unloading l B 3.7.18 l BASES BACKGROUND (continued) design basis is exceeded. The spent fuel pool boron dilution analysis concluded that an unplanned or inadvertent event which could result in the dilution of the spent fuel pool boron concentration to 430 ppm is not a credible event.
APPLICABLE Several accident conditions (Ref. 3) are considered that could result SAFETY ANALYSES in an increase in system kee for a DSC being loaded or unloaded in the spent fuel pool. These accident conditions include the drop of a fuel assembly on top of the DSC storage cells, the drop of a fuel assembly just outside the transfer cask containing the DSC, a higher than normal spent fuel pool water temperature, and the misleading of a fresh 5.0 wt %
U-235 assembly in one of the DSC storage cells.
For an occurrence of these postulated accidents, the double contingency principle discussed in ANSI N-1 6.1-1975 and the April 1978 NRC letter (Ref. 4) can be applied. This double contingency principle does not require assuming two unlikely, independent, concurrent events to ensure protection against a criticality accident. Thus, for these postulated accident conditions, the presence of additional soluble boron in the spent fuel pool water (above the 430 ppm required to maintain kff < 0.95 under normal DSC loading/unloading conditions) can be assumed as a realistic initial condition since not assuming its presence would be a second unlikely event.
Calculations were performed to determine the amount of soluble boron required to offset the highest reactivity increase associated with these postulated accidents, in order to maintain ke < 0.95. It was found that a spent fuel pool boron concentration of 630 ppm was sufficient to maintain keff < 0.95 for the worst-case postulated criticality-related accident (the fresh fuel assembly misloaded in a DSC storage cell). Specification 3.7.12 ensures the spent fuel pool contains adequate dissolved boron to compensate for the increased reactivity caused by these postulated accidents.
For normal storage conditions, Specification 4.3.1 c. requires that the spent fuel rack kern be < 0.95 when flooded with water borated to 430 ppm. A spent fuel pool boron dilution analysis was performed which confirmed that sufficient time is available to detect and mitigate a dilution of the spent fuel pool before the 0.95 kf design basis is exceeded. The spent fuel pool boron dilution analysis concluded that an unplanned or inadvertent event which could result in the dilution of the spent fuel pool boron concentration to 430 ppm is not a credible event.
OCONEE UNITS 1, 2, & 3 B 3.7.1 8-2 Amendment Nos. 3XX, 3XX, & 3XX I
Dry Spent Fuel Cask Loading and Unloading l B 3.7.18 BASES APPLICABLE The configuration of fuel assemblies in the DSC and the concentration of SAFETY ANALYSIS dissolved boron in the spent fuel pool satisfy Criterion 2 of 10 CFR 50.36 (continued)
(Ref. 5)
LCO The k" of the dry spent fuel storage cask (NUHOMS-24P or NUHOMS-24PHB DSC), during loading and unloading operations in the spent fuel pool, will always remain < 0.95, assuming the spent fuel pool is flooded with water borated to at least 430 ppm, and that each loaded fuel assembly meets the initial enrichment, burnup, and post-irradiation cooling time of Table 3.7.18-1.
APPLICABILITY This LCO applies whenever any fuel assembly is in a dry spent fuel storage cask located in the spent fuel pool.
ACTIONS A.1 Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply.
If moving fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action. If moving fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operation. Therefore, in either case, inability to move fuel assemblies is not sufficient reason to require a reactor shutdown.
When the configuration of fuel assemblies loaded in the NUHOMSO-24P or NUHOMS-24PHB DSC is not in accordance with the LCO, immediate action must be taken to make the necessary fuel assembly movement(s) to bring the configuration into compliance with the LCO.
SURVEILLANCE SR 3.7.18.1 REQUIREMENTS This SR verifies by administrative means that the initial enrichment, burnup, and post-irradiation cooling time of the fuel assembly to be loaded into or removed from the NUHOMS-24P or NUHOMS-24PHB DSC is in accordance with Table 3.7.18-1.
OCONEE UNITS 1, 2, & 3 B 3.7.18-3 Amendment Nos. 3XX, 3XX, & 3XX I
Dry Spent Fuel Cask Loading and Unloading B 3.7.18 BASES REFERENCES
1.
10 CFR 50.68(b)(4)
2.
American Nuclear Society, "American National Standard Design Requirements for Light Water Reactor Fuel Storage Facilities at Nuclear Power Plants," ANSIIANS-57.2-1983, October 7, 1983.
3.
Nuclear Regulatory Commission, Memorandum to Timothy Collins from Laurence Kopp, "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light Water Reactor Power Plants," August 19, 1998.
4.
Double contingency principle of ANSI N16.1-1975, as specified in the April 14,1978 NRC letter (Section 1.2) and implied in the proposed revision to Regulatory Guide 1.13 (Section 1.4, Appendix A).
5.
10 CFR 50.36 OCONEE UNITS 1,2, &3 B 3.7.18-4 Amendment Nos. 3XX, 3XX, & 3XX I
ATTACHMENT 3 LIST OF REGULATORY COMMITMENTS
- List of Regulatory Commitments License Amendment Request No. 2005-009 March 1, 2006 Page 1 The following commitment table identifies those actions committed to by Duke Energy Corporation (Duke) in this submittal. Other actions discussed in the submittal represent intended or planned actions by Duke. They are described to the Nuclear Regulatory Commission (NRC) for the NRC's information and are not regulatory commitments.
Commitment Implementation Date Upon NRC approval of this license amendment request, applicable Prior to June 30, sections of the Oconee Nuclear Site Updated Final Safety Analysis 2007 Report (UFSAR) will be updated and submitted in the annual report.

v t e Letter Sequence Request
TAC:MC0238, (Approved, Closed)
Initiation Request Acceptance... Administration Questions Answers 26 April 2006 Results Approval, Approval Other: ML042040460, ML042050060
MONTHYEAR2006-03-01 [Table View]

ML060720030

Text

Subject:

Reference:

Enclosures:

Subject:

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

6.0 ENVIRONMENTAL CONSIDERATION

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