ML082820400

License Amendment Request to Modify Allowable Storage Patterns in Spent Fuel Storage Racks

ML082820400
Person / Time
Site:	Cook
Issue date:	09/25/2008
From:	Weber L Indiana Michigan Power Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
AEP:NRC:8431
Download: ML082820400 (59)

Text

Indiana Michigan Power Company INDIANA Nuclear Generation Group IOne Cook Place MICHIGAN Bridgman, MI 49106 PCRER aep.com September 25, 2008 AEP:NRC:8431 10 CFR 50.90 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop O-P1-17 Washington, DC 20555-0001

SUBJECT:

Donald C. Cook Nuclear Plant Units 1 and 2 Docket Nos. 50-315 and 50-316 License Amendment Request to Modify Allowable Storage Patterns in Spent Fuel Storage Racks

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Indiana Michigan Power Company (I&M),

the licensee for Donald C. Cook Nuclear Plant (CNP) Units 1 and 2, proposes to amend the Appendix A Technical Specifications (TS) to Facility Operating Licenses DPR-58 and DPR-74. I&M proposes to modify two TS figures showing allowable locations for nuclear fuel in the spent fuel pool storage racks.

These figures show two different allowable storage patterns for four of the storage rack modules.

I&M proposes to modify these two figures such that fuel may be located in any of these four individual modules in accordance with either figure. The proposed amendment will allow continued placement of new and intermediate burnup fuel in the spent fuel pool as the storage racks approach a full condition. to this letter provides an affirmation statement pertaining to the information contained herein. Enclosure 2 provides I&M's evaluation of the proposed change. Enclosure 3 provides a technical evaluation by Holtec International regarding the proposed change. Attachments 1 and 2 to this letter provide TS pages marked to show changes for Unit 1 and Unit 2, respectively.

Associated TS Bases changes will be made in accordance with the CNP Bases Control Program.

I&M requests approval of the proposed change by August 31, 2009, to support fuel movements for the fall 2009 Unit 1 refueling outage. I&M will inform the Nuclear Regulatory Commission (NRC)

Licensing Project Manager if changes to the Unit 1 operating schedule affect the date by which approval is requested. The proposed change will be implemented within 45 days of NRC approval.

Copies of this letter and its attachments are being transmitted to the Michigan Public Service Commission and Michigan Department of Environmental Quality, in accordance with the requirements of 10 CFR 50.91.

bD\\"

U. S. Nuclear Regulatory Commission AEP:NRC:8431 Page 2 There are no new regulatory commitments made in this letter. Should you have any questions, please contact Mr. John A. Zwolinski, Regulatory Affairs Manager, at (269) 466-2478.

Sincerely, Lawrence J. Weber Site Vice President JRW/rdw

Enclosures:

1.

Affirmation

2.

Proposed License Amendment to Modify Allowable Storage Patterns in Four Spent Fuel Storage Modules

3.

Holtec Technical Evaluation Attachments:

1.

Donald C. Cook Nuclear Plant Unit 1 Technical Specification Pages Marked To Show Changes

2.

Donald C. Cook Nuclear Plant Unit 2 Technical Specification Pages Marked To Show Changes c:

T. A. Beltz, NRC Washington, DC J. L. Caldwell, NRC Region III K. D. Curry, Ft. Wayne AEP, w/o enclosures/attachments J. T. King, MPSC MDEQ - WHMD/RPS NRC Resident Inspector to AEP:NRC:8431 AFFIRMATION I, Lawrence J. Weber, being duly sworn, state that I am Site Vice President of Indiana Michigan Power Company (I&M), that I am authorized to sign and file this request with the Nuclear Regulatory Commission on behalf of I&M, and that the statements made and the matters set forth herein pertaining to I&M are true and correct to the best of my knowledge, information, and belief.

Indiana Michigan Power Company Lawrence J. Weber Site Vice President SWORN TO AND SUBSCRIBED BEFORE ME THIS o/5, DAY OF-eD-ewbe-2008 MCms Notary Public

(___

My Commission Expires IW0wawfbi.*tQ *,ft to AEP:NRC:8431 Proposed License Amendment to Modify Allowable Storage Patterns in Four Spent Fuel Storage Modules Documents referenced in this enclosure are identified in Section 8.0.

1.0 DESCRIPTION

Pursuant to 10 CFR 50.90, Indiana Michigan Power Company (!&M), the licensee for Donald C. Cook Nuclear Plant (CNP) Units 1 and 2, proposes to amend'the Appendix A Technical Specifications (TS) to Facility Operating Licenses DPR-58 and DPR-74.

I&M proposes to modify two TS figures showing allowable locations for nuclear fuel in the spent fuel pool storage racks. These figures-show two different allowable storage patterns for four of the storage rack modules. I&M proposes to modify these two figures such that fuel may be located in any of these four individual modules in accordance with either figure.

The proposed amendment will allow continued placement of new and intermediate burnup fuel in the spent fuel pool as the storage racks approach a full condition.

2.0 PROPOSED CHANGE

I&M proposes to modify Figure 4.3-1 and Figure 4.3-2 in the Unit 1 and Unit 2 TS as follows:

Add the alpha-numeric designations Al, C1, El, and A5, to the four fuel modules that comprise the vertical row on the left side of the figure.

Add a note indicating that the storage pattern for any of these individual modules, Al, C1, El, or A5, may be in accordance with Figure 4.3-1 or Figure 4.3-2.

Delete the value for the total number of cells in each of the three storage regions (Region 1, Region 2, and Region 3), and on Figure 4.3-2 only, delete the value for the

.total number of cells that are to be left empty.

Attachments 1 and 2 to this letter provide TS pages marked to show proposed changes for Unit 1 and Unit 2, respectively.

New text on these pages is indicated by an underline, and deleted text is indicated by a strikethrough.

New clean Unit 1 and Unit 2 TS pages with proposed changes incorporated will be provided to the Nuclear Regulatory Commission (NRC)

Licensing Project Manager when requested. Associated TS Bases changes will be made in accordance with the CNP Bases Control Program.

3.0 BACKGROUND

3.1 Spent Fuel Storage Rack Design and Operation The current CNP spent fuel storage racks were installed in the early 1990s. The design and operation of the storage racks were detailed in the 1991 license amendment request transmitted by Reference 1.

The associated amendment was approved by the NRC in 1993 as documented by Reference 2. A summary of the storage rack design and operation is provided below. Additional details are provided in Reference 1.

to AEP:NRC:8431 Page 2 Irradiated fuel from Unit 1 and Unit 2 is stored in spent fue[ storage racks located in a common spent fuel pool. New fuel is also stored in the spent fuel storage racks prior to being placed in the reactor vessel during refueling operations. The storage racks are comprised of twenty-three free-standing, self-supporting storage rack modules that rest on the floor of the spent fuel pool.

Each module is divided into cells which hold individual fuel elements. The twenty-three modules contain a total of 3613 storage cells. There are several sizes of modules with varying numbers of cells. Within each module, the cells are grouped in parallel rows with a nominal distance between centers of 8.97 inches in both directions. The modules are constructed primarily of stainless steel. The only other material used in the modules is a boron carbide and aluminum neutron absorber with the product name "Boral." Each cell wall includes a Boral panel. The module baseplates extend outwards such that a two inch nominal gap is established between the adjacent walls of neighboring modules. Consequently, there are two neutron Boral absorber panels and a nominal two inch water flux trap between fuel assemblies located in peripheral cells of adjacent modules.

In establishing allowable fuel loading patterns for the modules, the cells are assigned to one of three regions or assigned as empty. Region 1 cells are designated for new fuel or irradiated fuel of any burnup. Region 2 cells are designated for high burnup fuel.

Region 3 cells are designated for intermediate or high burnup fuel. In the configuration shown in TS Figure 4.3-1, the Region 1 cells are located in alternate cells along the outside edges of the outside modules (where neutron leakage reduces reactivity), and alternate cells along the edges between two modules (where the water gap provides a flux-trap which reduces reactivity). High burnup fuel in Region 2 provides a low-reactivity barrier between new fuel assemblies and fuel of intermediate burnup in Region 3. The configuration shown in TS Figure 4.3-1 is designated as the "Normal Storage Pattern."

As shown in TS Figure 4.3-2, another storage pattern, termed the "Interim Storage Pattern," is also permitted.

This pattern allows the central cells in four modules to be loaded in a checkerboard pattern of Region 1 cells and empty cells.

The modules in which the checkerboard pattern is allowed are identified as modules Al, C1, El, and A5 in Reference 1 and Figure 9.7-2 of the Updated Final Safety Analysis Report (UFSAR).

The checkerboard pattern provides more Region 1 cells than the pattern shown in TS Figure 4.3-1.

The checkerboard pattern was intended primarily to facilitate a temporary full core off-load and storage of new fuel when needed for refueling outages prior to the time that the central cells in modules Al, C1, El, or A5 are beginning to fill with intermediate or high burnup fuel.

3.2 Current Licensing Basis Criticality Analyses The amendment request transmitted by Reference 1 addressed the applicable design considerations for the spent fuel storage modules. The principal design consideration of interest with respect to the amendment proposed by this letter is criticality control. A summary of the criticality analyses detailed in Reference I is provided below.

Conditions Analyzed, Criticality analyses of two conditions were performed. Analyses were performed for normal operating conditions in which fuel is located as shown in TS Figure 4.3-1 or Figure 4.3-2.

to AEP:NRC:8431 Page 3 Analyses were also performed for abnormal or accident conditions in which a fuel assembly is placed in a location not in accordance with TS Figure 4.3-1.

Analysis Assumptions Except as noted, the following conservative assumptions were applied to both the normal and the abnormal or accident condition analyses to assure that the actual reactivity remains lower than calculated reactivity:

A room temperature (20 degrees centigrade) moderator was assumed because it yields the highest reactivity within the operating temperature range.

No poison (soluble boron) in the spent fuel pool water was assumed for the normal criticality analysis.

Neutron leakage was only assumed for module boundary storage cells.

Neutron absorption in minor structural members was disregarded.

Additional margins of conservatism resulted from the time that irradiated fuel is in the spent fuel pool which leads to a decrease in reactivity, and the negative moderator coefficient of reactivity inherent in the pool water. These margins insure the actual reactivity of the TS Figure 4.3-1 and Figure 4.3-2 configurations will remain less than the maximum calculated reactivity (keff) of 0.94, which in turn satisfies the TS requirement that keff be maintained < 0.95.

The analyses were performed by modeling the standard Westinghouse 15 x 15 fuel assemblies containing U0 2 with a maximum enrichment of 4.95 +/- 0.05 weight percent U-235' used in Unit 1.

The fuel assemblies were assumed to be uniform, having the high reactivity characteristic to the central enriched zone. This model yields higher reactivity values than those for standard and optimized 17 x 17 Westinghouse and Advanced Nuclear Fuel (ANF) designs used in Unit 2.

Methodology The KENO-5a Monte Carlo computer package was used in conjunction with the 27-group SCALE cross-section library and the NITAWL subroutine for U-238 resonance shielding effects.

A total of 1,250,000 histories were run to guarantee calculation convergence, thus, minimizing statistical variances.

The two-dimensional transport theory code CASMO-3 was used.to determine equivalent enrichments, as burnup capability was not available.

Normal Operating Condition Analyses The criticality calculations for TS Figure 4.3-1 or Figure 4.3-2 configurations yielded a maximum reactivity of 0.940. The maximum calculated reactivity includes a margin for uncertainty in reactivity calculations including mechanical tolerances.

All uncertainties are statistically combined, such that the final keff will be equal to or less than 0.95 with a 95% probability at a 95% confidence level. This satisfies the TS 4.3.1.1.b requirement that keff be maintained < 0.95.

to AEP:NRC:8431 Page 4 Abnormal or Accident Condition Analyses The assumed abnormal or accident condition for the TS Figure 4.3-1 storage pattern consisted of placement of a new fuel assembly in a Region 2 position where the remainder of the rack is filled with fuel of the highest permissible reactivity.

This condition could result in a reactivity increase of +0.065 5k. The calculation determined that a spent fuel pool boron concentration of 450 parts per million (ppm) would be sufficient to provide a keff of 0.95. A boron concentration well in excess of this value is assured by TS 3.7.15 which requires that the spent fuel pool boron concentration be _Ž2400 ppm when fuel assemblies are stored in the pool and a storage pool verification has not been performed since the last movement of fuel assemblies in the pool.

3.3 Reason for Requesting Amendment The cells in the center of modules Al, Cl, El, and A5 are normally kept empty so they can be loaded in the interim checkerboard pattern shown in TS Figure 4.3-2 to support refueling outages. However, the cells in modules other than Al, Cl, El, and A5 are approaching a full condition. It will then be necessary to load irradiated fuel into modules Al, Cl, El, and A5 in accordance with the normal pattern shown in TS Figure 4.3-1. Once the normal pattern is used in one of the four modules (Al, Cl, El, and A5), the current TS Figure 4.3-1 and Figure 4.3-2 would preclude use of the interim checkerboard pattern in the other three modules.

Without the allowance to use the interim pattern in the other three modules, it will be necessary to rearrange fuel in modules other than Al, Cl, El, and A5 to accomplish upcoming refueling outages while maintaining compliance with requirements identified in Section B.5.b of NRC Order EA-02-026.

This would significantly increase the number and distance of fuel moves needed to support the outages. Eventually, a condition will be reached in which there will not be enough Region 1 cells available to accommodate the new fuel and low burnup fuel for a core reload. Therefore, the allowance to use the interim checkerboard pattern in some of the four modules and the normal pattern in others will minimize the number and distance of fuel moves for upcoming outages, and will ultimately allow continued plant operation until fuel is relocated to dry cask storage facilities (expected to begin in mid-2011).

Accordingly, I&M is requesting a TS change that would allow fuel to be located in any of these four individual modules (Al, Cl, El, or A5) in accordance with either the normal pattern of Figure 4.3-1 or the interim checkerboard pattern of TS Figure 4.3-2.

4.0 TECHNICAL ANALYSIS

4.1 Design Considerations To support the TS change proposed by this amendment request, I&M contracted Holtec International to evaluate the effect of the proposed changes on the design considerations addressed in Reference 1. The Holtec technical evaluation is provided as Enclosure 3 to this letter.

The Holtec technical evaluation determined that compliance with the design criteria addressed in Reference 1 will not be impacted by the proposed change. I&M has verified that there have been no changes to the CNP design and licensing basis subsequent to NRC approval of Reference 1 that would alter the conclusions documented in the technical to AEP:NRC:8431 Page 5 evaluation. Note that the Holtec technical evaluation uses the term "fresh fuel," while the TS use the term "new fuel." Both terms refer to fuel that has not been subjected to a neutron flux in an operating reactor.

In addition to evaluating the effect of the proposed change on the design considerations addressed in Reference 1, Holtec performed an analysis of an abnormal or accident case in which a fuel assembly is placed in a location not in accordance with the interim checkerboard pattern shown in TS Figure 4.3-2.

As described in Section 4.1 of Enclosure 3, the Holtec analysis used similar assumptions to the analysis documented in Reference 1, but used a newer computer code, MCNP4a, rather than the KENO-5a code. Use of the MCNP4a code has been previously approved by the NRC as described in Section 7.0 of this enclosure. Section 4.1 of Enclosure 3 documents that the maximum keff resulting from placement of a fuel assembly in a location not in accordance with the interim checkerboard pattern shown in TS Figure 4.3-2 would be significantly less than 0.95 with the boron concentration required during fuel movement by the TS.

4.2 Specific TS Changes The reasons for the specific changes to TS Figure 4.3-1 and TS Figure 4.3-2 are described below.

Addition of alpha-numeric designations Al, C1, El, and A5 The alpha-numeric designations were added to differentiate the modules involved in the proposed change (those on the left side of the figure) from the remainder of the modules. The alpha-numeric designation for each of the four modules corresponds to that shown in Figure 2.1.1 of Reference 1 and Figure 9.7-2 of the UFSAR.

Addition of notes indicating that the storage pattern for any of the individual modules (Al, C1, El, or A5) may be in accordance with Figure 4.3-1 or Figure 4.3-2.

The notes were added to accomplish the fundamental objective of the proposed amendment, i.e., to allow any of the specified modules to be loaded in accordance with either the normal pattern (Figure 4.3-1) or the interim checkerboard pattern (Figure 4.3-2).

Deletion of the value for the total number of cells in Region 1, Region 2, Region 3, and for Figure 4.3-2 only, deletion of the value for the total number of empty cells.

The values for the total number of cells in each category were deleted because the values can change depending on the loading pattern used for each of the individual modules Al, C1, El, and A5. The analyses discussed in the Holtec technical evaluation provided as Enclosure 3 to this letter demonstrate that compliance with the storage patterns specified in TS Figure 4.3-1 and Figure 4.3-2 in accordance with this proposed amendment will assure safety. Therefore, the specific values for the total number of cells in each category are not important to safety, and the inclusion of these values in TS is not required by 10 CFR 50.36(d)(4).

to AEP:NRC:8431 Page 6 5.0 REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration Pursuant to 10 CFR 50.90, Indiana Michigan Power Company (I&M), the licensee for Donald C. Cook Nuclear Plant (CNP) Units 1 and 2, proposes to amend the Appendix A Technical Specifications (TS) to Facility Operating Licenses DPR-58 and DPR-74.

I&M proposes to modify two TS figures showing allowable locations for nuclear fuel in the spent fuel pool storage racks. These figures show two different allowable storage patterns for four of the storage rack modules. I&M proposes to modify these two figures such that fuel may be located in any of the four individual modules in accordance with either figure.

I&M has evaluated whether a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

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

Response: No The accidents and events of concern involving fuel located in the spent fuel pool storage racks are a criticality accident, a fuel handling accident, and inadequate decay heat removal.

The proposed change will not increase the probability of a criticality accident because analyses demonstrate that sub-criticality will be maintained for the fuel storage configurations allowed by the change.

The proposed change will not increase the probability of a fuel handling accident because it does not affect the manner in which fuel is moved or handled.. The proposed change will decrease the number of fuel moves needed for upcoming refueling outages. The proposed change will not increase the probability of inadequate decay heat removal because thermal-hydraulic analyses demonstrating adequate heat removal will remain valid for the storage configurations allowed by the change. Therefore, the probability of occurrence of a previously evaluated accident will not be significantly increased.

The proposed change does not adversely affect the ability to perform the intended safety functions of any system, structure, or component (SSC) credited for mitigating a criticality accident, a fuel handling accident, or inadequate decay heat removal.

Therefore, the consequences of a previously evaluated accident will not be significantly increased.

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

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

Response: No to AEP:NRC:8431 Page 7 The proposed change does not alter the design function or operation of any SSC. The proposed change does not affect the capability of the SSCs involved with the storage of fuel in the spent fuel pool to perform their function. As a result, no new failure mechanisms, malfunctions, or accident initiators are created. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No The margins of safety involved with the storage of fuel in the spent fuel pool are the margins associated with the prevention of criticality, mitigation of a fuel handling accident, and assurance of adequate decay heat removal. The proposed amendment involves no change in the capability of any SSC that maintains these margins. Therefore, there is no significant reduction in a margin of safety as a result of the proposed amendment.

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

Based on the above, I&M concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

5.2 Applicable Regulatory Requirements/Criteria As described below, compliance with the applicable regulatory requirements and criteria is not affected by the proposed amendment.

Code of Federal Regulations 10 CFR 50.36 requires that the TS for a facility include a description of design features, such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety, and are not covered by Safety Limits, Limiting Safety System

Settings, Limiting Control Settings, Limiting Conditions for Operation, or Surveillance Requirements. TS Figures 4.3-1 and Figure 4.3-2 will continue to satisfy this criterion in that the fuel storage patterns allowed by the figures will assure safety with respect to the licensing basis design considerations.

10 CFR 70.24 requires that licensees authorized to possess specified quantities and forms of special nuclear material maintain provisions for monitoring criticality.and mitigating the consequences of criticality accidents.

By Reference 3, the Nuclear Regulatory, Commission (NRC) granted I&M an exemption from these requirements for CNP.

The basis for the exemption was that inadvertent or accidental criticality will be precluded through compliance with the TS, the geometric spacing of fuel assemblies in the new fuel storage facility and spent fuel storage pool, and administrative controls imposed on fuel handling procedures. The basis for this exemption remains valid in that geometric spacing of fuel assemblies will continue to maintain keff less than the TS limit of 0.95. Additionally, appropriate administrative controls on fuel handling will be maintained..

to AEP:NRC:8431 Page.8 Plant Specific Desiqn Criteria (PSDC)

The CNP PSDC are described in Section 1.4 of the UFSAR. These criteria differ from the criteria stated in Appendix A of 10 CFR 50 which were published in 1971 after the CNP construction permits were issued.

CNP PSDC Criterion 66, "Prevention of Fuel Storage Criticality," requires that criticality in the spent fuel pool be prevented by physical systems or processes. Compliance with this criterion will be unaffected by the proposed change in that, as described in Enclosure 3, criticality will be prevented by the allowed storage configurations.

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

6.0 ENVIRONMENTAL CONSIDERATION

S A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement.

However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

7.0 PRECEDENT The proposed change is specific to the configuration of the CNP spent fuel pool, the CNP storage rack arrangement, and the CNP allowable fuel storage patterns. Consequently, there are no apparent precedent license amendments that allow a change identical to that proposed for CNP. However, a license amendment for Crystal River Nuclear Plant was approved by the NRC (Reference 4) based on use of the MCNP4a three-dimensional Monte Carlo code. The MCNP4a code was used for the analysis of a postulated misloading accident in the interim storage configuration at CNP as described in Enclosure 3.

8.0 REFERENCES

1. Letter from E. E. Fitzpatrick, I&M, to NRC Document Control Desk, "Spent Fuel Pool Reracking Technical Specification Changes," dated July 26, 1991.

2.

Letter from W. M. Dean, NRC, to E. E. Fitzpatrick, I&M, "Donald C. Cook Nuclear Plant, Units 1 and 2 - Amendment Nos. 169 And 152 to Facility Operating License Nos. DPR-58 and DPR-74 (TAC Nos. M80615 and M80616)," dated January 14, 1993 (ML021060153).

to AEP:NRC:8431 Page 9

3. Letter from J. B. Hickman, NRCI to E. E. Fitzpatrick, I&M, "Exemption from 10 CFR 70.24 Criticality Monitoring Requirements - Donald C. Cook Nuclear Plant, Units 1 and Unit 2 (TAC Nos. M95197 and M95198)," dated October 28, 1996 (ML021070041).

4.

Letter from S. N. Bailey, NRC, to D. E. Young, Crystal River Nuclear Plant, "Crystal River, Unit 3 - Issuance of Amendment Regarding Fuel Storage Patterns in the Spent Fuel Pool (TAC No. MD3308)," dated October 25, 2007 (ML072910317).

to AEP:NRC:8431 HOLTEC TECHNICAL EVALUATION

Technical Evaluation of Additional Configurations in the D.C. Cook Spent Fuel Pool Revision 0 July 25, 2008 1.0 Introduction This position paper documents the acceptability of a proposed configuration change for the D.C.

Cook spent fuel pool. Each of the major technical disciplines as documented in Holtec Report HI-90488, "Licensing Report for Storage of Densification of D.C. Cook Spent Fuel Pool" [1] are addressed in the following subsections.

2.0

Background

The D.C. Cook spent fuel pool was re-racked in the early 1990's with high density spent fuel storage racks to achieve the maximum spent fuel capacity possible as shown in Figure 1. All racks installed were non-flux-trap style racks, with a single neutron absorber panel (Boral) between each fuel assembly. The racks were approved by the USNRC to accommodate a mixed zone three region (MZTR) storage scheme where fresh fuel is interspersed with high burnup fuel on the edge of the rack, with moderate burnup fuel stored in the center of each rack. Figure 2 (Figure 4-1 from [1]) shows the normal storage pattern for the spent fuel pool with the following fuel stored in each of the 3 regions:

• Region 1 is designed to accommodate new fuel with a maximum enrichment of 4.95 +

0.05 wt% 2 35U, or spent fuel regardless of the discharge fuel burnup.

" Region 2 is designed to accommodate fuel of 4.95 wt% initial enrichment burned to at least 50,000 MWD/MTU (assembly average), or fuel of other enrichments with a burnup yielding an equivalent reactivity.

" Region 3 is designed to accommodate fuel of 4.95 wt% initial enrichment burned to at least 38,000 MWD/MTU (assembly average), or fuel of other enrichments with a buinup yielding an equivalent reactivity.

Additionally, an interim storage pattern was licensed to accommodate a full-core offload of low-burned fuel where the center region of four racks (A1, Cl, El and A5) in the pool are replaced with a checkerboard of fresh fuel and empty storage cells as shown in Figure 3 (Figure 4-2 from

[1]).

The following subsections provide technical justification for allowing any combination of the four rack modules to contain either fuel assemblies meeting the Region 3 criteria or a checkerboard of fresh fuel and empty storage cells in the central region of these modules. The numbering of the following subsections matches the section numbering in [1].

H a L T E c This Technical Evaluation is a QA Validated Document Page 1 of 14 INTERNATIONAL

Technical Evaluation of Additional Configurations in the D.C. Cook Spent Fuel Pool Revision 0 July 25, 2008 3.0 Construction of Rack Module No changes are proposed to the construction of the rack modules, only changes to the acceptable loading configurations in the spent fuel pool. Therefore, the discussion in [1] on construction of the rack modules is not affected.

4.0 Criticality Safety Analysis The criticality analysis presented in [1] addresses both the normal storage configuration and the interim storage configuration separately. The MZTR concept is based on neutronically isolating fresh fuel assemblies by surrounding them with high burned fuel, while placing moderately burned fuel in the center of the rack. The storage cells in each rack module are of the non-flux trap type, with a single sheet of neutron absorber between adjacent assemblies.

Each rack module is separated from adjacent rack modules by a nominal spacing of 2 inches with neutron absorber on the outside surfaces of the rack module, effectively creating a "flux-trap1"' between rack modules. Additionally, fuel assemblies in the center region (Region 3) of adjacent racks are separated by four rows of fuel and the flux-trap between rack modules, ensuring that the central regions of adjacent racks are neutronically decoupled.

KENO5a models were created which represent a single rack containing the MZTR loading pattern.

These models represent an 1 lxlI storage rack, as shown in Figure 4. Reflecting boundary conditions were used on all sides of the model, effectively creating an infinite array of 1 lxi I storage racks. Additional models for the interim storage configuration were created with the same geometry, except the Region 3 fuel assemblies were replaced with a checkerboard of Region 1 fuel assemblies (fresh fuel) and empty storage locations. The results show that the maximum klff of the interim and normal storage configurations are statistically equivalent (within 2c) and therefore either storage configuration is acceptable.

Given that the analysis demonstrates 'that either of the two configurations is acceptable for a single rack module, any combination of the four storage rack modules with the center region containing either Region 3 fuel assemblies or a checkerboard of fresh fuel and empty storage cells is acceptable.

The previous analysis made use of reactivity equivalencing, which was the subject of an NRC issue summary 2001-12 [2].

The reactivity equivalencing method used in the analysis was reviewed and determined that the equivalent enrichments were calculated in the storage cell configuration and therefore are acceptable. For accident and abnormal conditions, 450 ppm of soluble boron in the spent fuel pool was credited to ensure that the maximum keff would remain less than or equal to 0.95. Given that the spent fuel pool has a Technical Specification 3.7.15

'A flux-trap construction implies that there is a water gap between adjacent storage cells such that the neutrons emanating from a fuel assembly are thermalized before reaching an adjacent assembly (i.e., in the water gap between assemblies).

H O L T E c This Technical Evaluation is a QA Validated Document Page 2 of 14 INTERNATIONAL

Technical Evaluation of Additional Configurations in the D.C. Cook Spent Fuel Pool Revision 0 July 25, 2008 requirement that the fuel storage pool concentration be > 2400ppm during fuel movement, the reactivity in the event of an accident or abnormal condition will be much lower than 0.95.

Based on the discussion above and the fact that the previous analysis was sufficient to bound the proposed configuration, no new analysis was performed to justify the proposed configuration.

However, a single calculation was performed to address an accident condition not analyzed when the racks were first licensed.

The analysis for this accident conditions is presented in the following subsections.

4.1 Criticality Analysis - Misloading Accident in Interim Storage Configuration During the review of the licensing report [1] to justify the proposed configuration, the existing analysis analyzed the misloading accident in the normal storage configuration, and showed that a soluble boron amount of 450 ppm was necessary to ensure that the maximum keff remained below 0.95. However, the misloading accident in the interim storage configuration where a fresh, unburned assembly is placed into a storage location intended to remain empty, was not performed. The criticality safety analysis is performed to determine the necessary soluble boron amount to ensure that the maximum k1ff remained below 0.95 for the misloaded assembly in the interim storage configuration, where the misloaded assembly is face adjacent to four fresh fuel assemblies.

4.1.1 Calculational Approach The basic approach for the additional calculation was to model a misplaced assembly in the center of the storage rack in the interim storage configuration, where 4 fresh fuel assemblies are face adjacent to the misloaded assembly. This location was chosen to maximize keff as the misplaced assembly is adjacent to high reactivity, fresh fuel. The calculational model, as shown in Figure 5, is a 5x5 array of storage cells, with a checkerboard of fresh fuel assemblies and empty storage cells and the center storage cell of the model including the misloaded fresh fuel assembly (in a location intended to be empty).

This model captures the reactivity effect of the misloaded fresh fuel assembly in the interim storage configuration without the use of reactivity equivalencing that was used for the spent fuel in the original analysis [1]. Additionally, reflecting boundary conditions are used in the radial direction, effectively creating an infinite model in the radial direction, where a misloaded assembly is present in every 5x5 array of cells. Finally, the uncertainties and bias from the previous analysis [1] were included in the calculation of the maximuml kff in a conservative manner where applicable, as described in the relevant subsections.

4.1.2 Methodology The principal method for the criticality analysis is the three-dimensional Monte Carlo code MCNP4a [4]. MCNP4a is a continuous energy three-dimensional Monte Carlo code developed at the Los Alamos National Laboratory. MCNP4a was selected because it has been used previously H O L T E C This Technical Evaluation is a QA Validated Document Page 3 of 14 INTERNATIONAL

Technical Evaluation of Additional Configurations in the D.C. Cook Spent Fuel Pool Revision 0 July 25, 2008 and verified for criticality analyses and has all of the necessary features for this analysis.

Additionally, this code has been used in all recent spent fuel storage rack licensing efforts by Holtec International that have been reviewed and approved by the USNRC.

MCNP4a calculations used continuous energy cross-section data based on ENDF/B-V and ENDF/B-VI supplied with the code.

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

+ 0.0011 for MCNP4a, evaluated with a 95% probability at the 95% confidence level [3]. The calculations for this analysis utilize the same computer platform and cross-section libraries used for the benchmark calculations discussed in Appendix A.

The convergence of a Monte Carlo criticality problem is sensitive to the following parameters:

(1) number of histories per cycle, (2) the number of cycles skipped before averaging, (3) the total number of cycles and (4) the initial source distribution. The MCNP4a criticality output contains a great deal of useful information that may be used to determine the acceptability of the problem convergence. This information has been used in parametric studies to develop appropriate values for the aforementioned criticality parameters to be used in storage rack criticality calculations.

Based on these studies, a minimum of 10,000 histories were simulated per cycle, a minimum of 20 cycles were skipped before averaging, a minimum of 100 cycles were accumulated, and the initial source was specified as uniform over the fueled regions (assemblies). Further, the output was reviewed to ensure that each calculation achieved acceptable convergence. These parameters represent an acceptable compromise between calculational precision and computational time.

The maximum kff is determined from the MCNP4a calculated kfr, the calculational bias and the applicable uncertainties and tolerances (bias uncertainty, calculational uncertainty, rack tolerances, fuel tolerances) using the following formula:

Max klff = Calculated kff + biases + [Zl (Uncertaintyi)2] 1/2 In the geometric models used for the calculations, each fuel rod and its cladding were described explicitly and reflecting boundary conditions were used in the radial direction which has the effect of creating an infinite radial array of storage cells.

4.1.3 Input data All rack and fuel input data provided in [1] is sufficient to perform the criticality analysis for the misloaded assembly in the interim storage configuration. Fuel data for the design basis assembly was provided in Table 4.4 of [1] and rack data was provided in Figure 4-4 of [112: The design 2 Table 4.4 and Figure 4-4 of [1] have different dimensions for the guide tube (thimble) i.d. and o.d., but the same thickness. The dimensions in Table 4.4 are used in the analysis. The discrepancy between the dimensions for the guide tube has a negligible effect on reactivity.

H M L T E c This Technical Evaluation is a QA Validated Document Page 4 of 14 INTERNATIONAL

Technical Evaluation of Additional Configurations in the D.C. Cook Spent Fuel Pool Revision 0 July 25, 2008 basis assembly is the Westinghouse 15x15 assembly, with an enrichment of 5.0 wt%, which was found in [1] to have the highest reactivity of all assembly types present in the D.C. Cook spent fuel pool.

4.1.4 Assumptions/Conservatisms To assure the true reactivity will always be less than the calculated reactivity, the following conservative design criteria and assumptions were employed:

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

2) The effective multiplication factor of an infinite radial array of fuel assemblies was used in the analysis.

3) An enrichment of 5.0 wt% is conservatively used for all fresh fuel assemblies in the analysis.

The analysis in [1] used a maximum enrichment of 4.95 wt% with a 0.05 wt% enrichment tolerance. The reactivity effect of the enrichment tolerance is conservatively included in the calculation of the maximum klf as shown in Table 1.

4) The boron loading in the neutron absorber panels is assumed to be at the minimum of 0.0300 g '°B/cm 2 (nominal of 0.0345 g l°B/cm2). The reactivity effect for the neutron absorber loading tolerance from [1] is conservatively included in the rack manufacturing tolerances as shown in Table 1.

5) The presence of burnable absorbers (B4C, Gadolinium, Erbium, IFBA) in fresh fuel is neglected. This is conservative as burnable absorbers would reduce the reactivity of the fresh fuel assembly.

6) The calculation consists of a 5x5 array with a misloaded fuel assembly in the center storage cell.

Reflecting boundary conditions conservatively create an infinite array of a single misloaded assembly in every 5x5 array.

7) The biases and uncertainties are applied in a conservative manner to maximize the maximimnum keff. The axial burnup effect and depletion uncertainty from [1] is applied although this model contains only fresh fuel, to account for the high burned fuel in the interim storage configuration.

4.1.5 Computer Codes The following computer codes were used during this analysis.

H kL"T E c This Technical Evaluation is a QA Validated Document Page 5 of 14 INTERNATIONAL

Technical Evaluation of Additional Configurations in the D.C. Cook Spent Fuel Pool Revision 0 July 25, 2008 0 MCNP4a [4] is a three-dimensional continuous energy Monte Carlo code developed at Los Alamos National Laboratory.

This code offers the capability of performing full three-dimensional calculations for the loaded storage racks.

MCNP4a was run on the PCs at Holtec.

4.1.6 Temperature Effects Pool water temperature effects on reactivity -in the spent fuel storage racks were previously calculated and the results were presented in Table 4.6 of [1]. The results show that the spent fuel pool temperature coefficient of reactivity is negative, i.e. a lower temperature results in a higher reactivity.

In MCNP4a, the Doppler treatment and cross-sections are valid only at 300K (27 °C). Therefore, a Ak is determined from 27 'C to 4 'C from Table 4.6 of [1], and is included in the final keff calculation as a bias. The bias shown in Table 1 for 27 'C (80.33 'F) to 4 'C (39.2 'F).

4.1.7 Uncertainties Due to Manufacturing Tolerances In the calculation of the final kff, the effect of manufacturing tolerances on reactivity must be included. CASMO-3 was used to perform these calculations in the previous analysis [1]. As prescribed in [5], the methodology employed to calculate the tolerance effects combine both the worst-case bounding value and sensitivity study approaches. The evaluations include tolerances of the rack dimensions and tolerances of the fuel dimensions. As for the bounding assembly, calculations are performed for a nominal initial enrichment of 4.95 wt% 235U. The reference condition is the condition with nominal dimensions and properties. To determine the Ak associated with a specific manufacturing tolerance, the kinf calculated for the reference condition is compared to the kinf from a calculation with the tolerance included. Note that for the individual parameters associated with a tolerance, no statistical approach is utilized. Instead, the full tolerance value was utilized to determine the maximum reactivity effect. All of the Ak values from the various tolerances are statistically combined (square root of the sum of the squares) to determine the final reactivity allowance for manufacturing tolerances. Only the Ak values in the positive direction (increasing reactivity) were used in the statistical combination. The fuel and rack tolerances included in this analysis are shown in Table 1.

4.1.8 Criticality Analysis The calculational models for the misloaded assembly situation consist of 5x5 checkerboard of assemblies with reflective boundary conditions on all four sides. Figure 5 shows the MCNP4a calculational model of the spent fuel storage cells, as drawn by the two-dimensional plotter in MCNP4a for the analyzed accident condition.

This model is a checkerboard of fresh fuel Eu...

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Technical Evaluation of Additional Configurations in the D.C. Cook Spent Fuel Pool Revision 0 July 25, 2008 assemblies and empty storage locations, with the center location filled with a fresh fuel assembly rather than remaining empty. Reactivity effects of manufacturing tolerances and uncertainties are taken from [1] and applied to the accident condition analyzed here. A summary of the calculation of the maximum keff with a maximum enrichment of 5.0 wt% 235U is shown in Table

1. The result of the analysis shows that the maximum keff under this accident scenario is significantly less than 0.95 with 800 ppm of soluble boron credited. The 800 ppm is much less than the 2400 ppm required during fuel movement by the Technical Specifications.

5.0 Thermal-Hydraulic Considerations The thermal hydraulic evaluation provided in [1] considers several core-offload scenarios such as normal discharge and back-to-back core offloads. The discharge scenarios are not dependant upon the loading configurations illustrated in Figure 2 and Figure 3. Therefore, the thermal analysis bounds any loading combination allowed by the proposed configuration change.

6.0 Rack Structural Considerations The rack structural/seismic analysis provided in [1] considers all racks to be fully loaded with spent fuel. This analysis is bounding for both the normal storage configuration and the interim storage configuration, and therefore is also bounding for the proposed configuration.

7.0 Accident Analysis and Miscellaneous Structural Evaluations The accident analysis and miscellaneous structural evaluations presented in [1] considers such events as a dropped fuel assembly, local cell wall buckling, analysis of welded joints in an isolated hot cell and crane uplift loads. In all cases these events are not affected by the proposed configuration.

8.0 Static and Dynamic Analyses of Fuel Pool Structure The spent fuel pool structural analysis provided in [1] assumes all racks to be fully loaded with spent fuel. This analysis is bounding for both the normal storage configuration and the interim storage configuration, and therefore is also bounding for the proposed configuration.

9.0 Radiolo2ical Evaluation The radiological analysis of a fuel handling accident presented in [1] is based upon a single assembly at the maximum bumup and shortest cooling time. The normal operational dose rates around the spent fuel pool due to the proposed configuration would be no higher than the dose ml..'

H O L T E c This Technical Evaluation is a QA Validated Document Page 7 of 14 INTERNATIONAL

Technical Evaluation of Additional Configurations in the D.C. Cook Spent Fuel Pool Revision 0 July 25, 2008 rates with the configuration shown in Figure 2.

Therefore, the accident and the normal operational radiological analyses in [1] are not affected by the proposed configuration.

10.0 In-Service Surveillance Program The in-service surveillance of the Boral neutron absorber coupons is not affected by the proposed change as no hardware changes are being implemented.

11.0 References

[1]

"Licensing Report for Storage Densification of D.C. Cook Spent Fuel Pool," Holtec Report HI-90488, Revision 6, transmitted as Attachment 4 to letter from E.E. Fitzpatrick, I&M, to NRC Document Control Desk, "Spent Fuel Pool Reracking Technical Specification Changes,".dated July 26, 1991.

[2]

"Nonconservatism in Pressurized Water Reactor Spent Fuel Storage Pool Reactivity Equivalencing Calculations," NRC Regulatory Issue Summary 2001-12, May 18, 2001

[3]

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

[4]

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

[5]

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

Collins, August 19, 1998.

Technical Author /Technical Reviewer Kristopher Cummi Ogsj Bret Brickner Emmmm HOLTEC IN I F,

f RN A A

I This Technical Evaluation is a QA Validated Document Page 8 of 14

Technical Evaluation of Additional Configurations in the D.C. Cook Spent Fuel Pool Revision 0 July 25, 2008 58'-3 1/8" 2 111 t 2

1121 10811icy 10 1~t I

IR

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-11 11d/m

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Typical Rack-to-Rack Gap: 2 lA51-) I Total Storage: 3616 cells (Include 3 triangular corner cells)

Figure 1: D.C. Cook Spent Fuel Pool Layout I.E..

HOLTEC IN rERNATrIONAL This Technical Evaluation is a QA Validated Document Page 9 of 14

Technical Evaluation of Additional Configurations in the D.C. Cook Spent Fuel Pool Revision 0 July 25, 2008 i504 Region 1 Cells 01439 Region 2 Cells *1670 Region 3 Cells Figure 2: Normal Storage Pattern nunnu HOLTEC INTERNATIONAL This Technical Evaluation is a QA Validated Document Page 10 of 14

Technical Evaluation of Additional Configurations in the D.C. Cook Spent Fuel Pool Revision 0 July 25, 2008 ir.4 a

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H O LT E C This Technical Evaluation is a QA Validated I INTERNATIONAL 2 Cells f1355 Region 3 Cells

)ocument Page 11 of 14

Technical Evaluation of Additional Configurations in the D.C. Cook Spent Fuel Pool Revision 0 July 25, 2008 Figure 4: Criticality Analysis KENO model for the Normal Storage Configuration at D.C. Cook 3 3 The KENO model for the interim storage configuration is identical except that the Region 3 fuel assemblies are replaced by a checkerboard of Region 1 fuel assemblies (fresh fuel) and empty storage locations.

I.E..

HOLTEC INTERNATIONAL This Technical Evaluation is a QA Validated Document Page 12 of 14

Technical Evaluation of Additional Configurations in the D.C. Cook Spent Fuel Pool Revision 0 July 25, 2008 Figure 5:

A Two-Dimensional Representation of the Misloaded Assembly Calculation Model used for the Interim Storage Configuration. This Figure was Drawn (To Scale) with the Two-Dimensional Plotter in MCNP4a.

MENEM HOLTEC INTERNATIONAL This Technical Evaluation is a QA Validated Document Page 13 of 14

Technical Evaluation of Additional Configurations inthe D.C. Cook Spent Fuel Pool Revision 0 July 25, 2008 Table 1 Summary of the Criticality Safety Analyses for a Misloaded Fresh Fuel Assembly in the Interim Storage Configuration with Soluble Boron Design Basis Burnup at 5.0 wt% 235U 0 GWD/MTU Soluble Boron 800 ppm MCNP Filename d5r2b4 Uncertainties Bias Uncertainty (95%/95%)

+/- 0.0011 Calculational Statistics (95%/95%, 2.Oxa)

+/- 0.0014 Fuel Eccentricity

+ 0.0019 Rack Tolerances 5

+/- 0.0078 Fuel Tolerances 6

+ 0.0049 Depletion Uncertainty 7

+/- 0.0047 Statistical Combination of Uncertainties

+/- 0.0107 Reference keff (MCNP4a) 0.9108 Total Uncertainty (above) 0.0107 Axial Bumup Effect8 0.0037 Temperature Bias9 0.0034 Calculational Bias (see Appendix A) 0.0009 Maximum kerr 0.9295 Regulatory Limiting keff 0.9500 4 All input files for the calculations are stored in the directory \\Projects\\1784\\Technical Evalautions\\1784001 and its subdirectories on the Holtec server.

5 This includes manufacturing tolerances of the'0B loading in the Boral panel (0.0061), Boral panel width (0.0009),

storage cell ID (0.00 15), storage cell wall thickness (0.0009) and water gap between adjacent racks (0.0045).

6 This includes manufacturing tolerances of the fuel density (0.0035) and fuel enrichment (0.0034) 7 Depletion uncertainty from [ 1 ] for 4.95 wt% fuel burned to 50 GWD/MTU.

8 From Table 4.2 of [1]. Conservatively included to account for burned fuel in the periphery of the rack. This is the axial burnup effect for the normal storage configuration where the rack is filled with spent and fresh fuel as shown in Figure 4.

9 Calculated as the difference in reactivity for a temperature of 80.33 'F (300K) to 39.2°F from Table 4.6 of [1] for fresh fuel. The temperature bias is calculated by linear interpolation between the points provided in Table 4.6 of [1]

for Region 1.

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

Note: because this appendix was taken from a previously NRC approved report, the next page is labeled "Appendix 4A, Page 1".

Eu.l.

HOLTEC INTERNATIONAL This Technical Evalaution is a QA Validated Document

APPENDIX 4A: BENCHMARK CALCULATIONS 4A. 1 INTRODUCTION AND

SUMMARY

Benchmark calculations have been made on selected critical experiments, chosen, in so far as possible, to bound the range of variables in the rack designs. Two independent methods of analysis were used, differing in cross section libraries and in the treatment of the cross sections. MCNP4a [4A. 1] is a continuous energy Monte Carlo code and KENO5a [4A.2]

uses group-dependent cross sections. For the KENO5a analyses reported here, the 238-group library was chosen, processed through the NITAWL-II [4A.2] program to create a working library and to account for resonance self-shielding in uranium-238 (Nordheimn integral treatment). The 238 group library was chosen to avoid or minimize the errorst (trends) that have been reported (e.g., [4A.3 through 4A.5]) for calculations with collapsed cross section sets.

In rack designs, the three most significant parameters affecting criticality are (1) the fuel enrichment, (2) the '°B loading in the neutron absorber, and (3) the lattice spacing (or water-gap thickness if a flux-trap design is used). Other parameters, within the normal range of rack and fuel designs, have a smaller effect, but are also included in the analyses.

Table 4A. 1 summarizes results of the benchmark calculations for all cases selected and analyzed, as referenced in the table. The effect of the major variables are discussed in subsequent sections below. It is important to note that there is obviously considerable overlap in parameters since it is not possible to vary a single parameter and maintain criticality; some other parameter or parameters must be concurrently varied to maintain criticality.

One possible way of representing the data is through a spectrum index that incorporates all of the variations in parameters. KENO5a computes and prints the "energy of the average lethargy causing fission" (EALF). In MCNP4a, by utilizing the tally option with the identical 238-group energy structure as in KENO5a, the number of fissions in each group may be collected and the EALF determined (post-processing).

Small but observable trends (errors) have been reported for calculations with the 27-group and 44-group collapsed libraries. These errors are probably due to the use of a single collapsing spectrum when the spectrum should be different for the various cases analyzed, as evidenced by the spectrum indices.

Appendix 4A, Page 1

Figures 4A. 1 and 4A.2 show the calculated klf for the benchmark critical experiments as a function of the EALF for MCNP4a and KENO5a, respectively (U0 2 fuel only). The scatter in the data (even for comparatively minor variation in critical parameters) represents experimental error' in performing the critical experiments within each laboratory, as well as between the various testing laboratories. The B&W critical experiments show a larger experimental error than the PNL criticals. This would be expected since the B&W criticals encompass a greater range of critical parameters than the PNL criticals.

Linear regression analysis of the data in Figures 4A. 1 and 4A.2 show that there are no trends, as evidenced by very low values of the correlation coefficient (0.13 for MCNP4a and 0.21 for KENO5a). The total bias (systematic error, or mean of the deviation from a k~ff of exactly 1.000) for the two methods of analysis are shown in the table below.

Calculational Bias of MCNP4a and KENO5a MCNP4a 0.0009 _+0.0011 KENO5a 0.0030+/-0.0012 The bias and standard error of the bias were derived directly from the calculated kff values in Table 4A. 1 using the following equations', with the standard error multiplied by the one-sided K-factor for 95 % probability at the 95 % confidence level from NBS Handbook 91 [4A. 18] (for the number of cases analyzed, the K-factor is -2.05 or slightly more than 2).

k=1Z k

(4A.1) ni t

A classical example of experimental error is the corrected enrichment in the PNL experiments, first as an addendum to the initial report and, secondly, by revised values in subsequent reports for the same fuel rods.

tt These equations may be found in any standard text on statistics, for example, reference

[4A.6] (or the MCNP4a manual) and is the same methodology used in MCNP4a and in KENO5a.

Appendix 4A, Page 2

n G 2 k)2 (4A.2) k n (n-1)

Bias =(I-k )

K a-(4A.3) where k, are the calculated reactivities of n critical experiments; a, is the unbiased estimator of the standard deviation of the mean (also called the standard error of the bias (mean)); K is the one-sided multiplier for 95 % probability at the 95 % confidence level (NBS Handbook 91 [4A. 18]).

Formula 4.A.3 is based on the methodology of the National Bureau of Standards (now NIST) and is used to calculate the values presented on page 4.A-2. The first portion of the equation, ( 1-k ), is the actual bias which is added to the MCNP4a and KENO5a results.

The second term, Kai, is the uncertainty or standard error associated with the bias. The K values used were obtained from the National Bureau of Standards Handbook 91 and are for one-sided statistical tolerance limits for 95 % probability at the 95 % confidence level. The actual K values for the 56 critical experiments evaluated with MCNP4a and the 53 critical experiments evaluated with KENO5a are 2.04 and 2.05, respectively.

The bias values are used to evaluate ihe maximum k~ff values for the rack designs.

KENO5a has a slightly larger systematic error than MCNP4a, but both result in greater precision than published data [4A.3 through 4A.5] would indicate for collapsed cross section sets in KENO5a (SCALE) calculations.

4A.2 Effect of Enrichment The benchmark critical experiments include those with enrichments ranging from 2.46 w/o to 5.74 w/o and therefore span the enrichment range for rack designs. Figures 4A.3 and 4A.4 show the calculated klT values (Table 4A. 1) as a function of the fuel enrichment reported for the critical experiments. Linear regression analyses for these data confirms that there are no trends, as indicated by low values of the correlation coefficients (0.03 for MCNP4a and 0.38 for KENO5a). Thus, there are no corrections to the bias for the various enrichments.

Appendix 4A, Page 3

As further confirmation of the absence of any trends with enrichment, a typical configuration was calculated with both MCNP4a and KENO5a for various enrichments.

The cross-comparison of calculations with codes of comparable sophistication is suggested in Reg. Guide 3.41. Results of this comparison, shown in Table 4A.2 and Figure 4A.5, confirm no significant difference in the calculated values of kfr for the two independent codes as evidenced by the 45' slope of the curve. Since it is very unlikely that two independent methods of analysis would be subject to the same error, this comparison is considered confirmation of the absence of an enrichment effect (trend) in the bias.

4A.3 Effect of '0B Loading Several laboratories have performed critical experiments with a variety of thin absorber panels similar to the Boral panels in the rack designs. Of these critical experiments, those performed by B&W are the most representative of the rack designs. PNL has also made some measurements with absorber plates, but, with one exception (a flux-trap experiment),

the reactivity worth of the absorbers in the PNL tests is very loW and any significant errors that might exist in the treatment of strong thin absorbers could not be revealed.

Table 4A.3 lists the subset of experiments using thin neutron absorbers (from Table 4A. 1) and shows the reactivity worth (Ak) of the absorbernt No trends with reactivity worth of the absorber are evident, although based on the calculations shown in Table 4A.3, some of the B&W critical experiments seem to have unusually large experimental errors. B&W made an effort to report some of their experimental errors. Other laboratories did not evaluate their experimental errors.

To further 'confirmr the absence of a significant trend with 10B concentration in the absorber, a cross-comparison was made with MCNP4a and KENO5a (as suggested in Reg.

Guide 3.41). Results are shown in Figure 4A.6 and Table 4A.4 for a typical geometry.

These data substantiate the absence of any error (trend) in either of the two codes for the conditions analyzed (data points fall on a 450, line, within an expected 95 % probability limit).

The reactivity worth of the absorber panels was determined by repeating the calculation with the absorber analytically removed and calculating the incremental (Ak) change in reactivity due to the absorber.

Appendix 4A, Page 4

4A.4 Miscellaneous and Minor Parameters 4A.4.1 Reflector Material and Spacings PNL has performed a number of critical experiments with thick steel and lead reflectors~t Analysis of these critical experiments are listed in Table 4A.5 (subset of data in Table 4A. 1). There appears to be a small tendency toward overprediction of kff at the lower spacing, although there are an insufficient number of data points in each series to allow a quantitative determination of any trends. The tendency toward overprediction at close spacing means that the rack calculations may be slightly more conservative than otherwise.

4A.4.2 Fuel Pellet Diameter and Lattice Pitch The critical experiments selected for analysis cover a range of fuel pellet diameters from 0.311 to 0.444 inches, and lattice spacings from 0.476 to 1.00 inches. In the rack designs, the fuel pellet diameters range from 0.303 to 0.3805 inches O.D. (0.496 to 0.580 inch lattice spacing) for PWR fuel and from 0.3224 to 0.494 inches O.D. (0.488 to 0.740 inch lattice spacing) for BWR fuel. Thus, the critical experiments analyzed provide a reasonable representation of power reactor fuel. Based on the data in Table 4A. 1, there does not appear to be any observable trend with either fuel pellet diameter or lattice pitch, at least over the range of the critical experiments applicable to rack designs.

4A.4.3 Soluble Boron Concentration Effects Various soluble boron concentrations were used in the B&W series of critical experiments and in one PNL experiment, with boron concentrations ranging up to 2550 ppm. Results of MCNP4a (and one KENO5a) calculations are shown in Table 4A.6. Analyses of the very high boron concentration experiments (> 1300 ppm) show a tendency to slightly overpredict reactivity for the three experiments exceeding 1300 ppm. In turn, this would suggest that the evaluation of the racks with higher soluble boron concentrations could be slightly conservative.

Parallel experiments with a depleted uranium reflector were also performed but not included in the present analysis since they are not pertinent to the Holtec rack design.

Appendix 4A, Page 5

4A.5 MOX Fuel The number of critical experiments with PuO2 bearing fuel (MOX) is more limited than for U0 2 fuel. However, a number of MOX critical experiments have been analyzed and the results are shown in Table 4A.7. Results of these analyses are generally above a K,, of 1.00, indicating that when Pu is present, both MCNP4a and KENO5a overpredict the reactivity. This may indicate that calculation for MOX fuel will be expected to be conservative, especially with MCNP4a. It may be noted that for the larger lattice spacings, the KENO5a calculated reactivities are below 1.00, suggesting that a small trend may exist with KENO5a. It is also possible that the overprediction in kff for both codes may be due to a small inadequacy in the determination of the Pu-241 decay and Am-241 growth. This possibility is supported by the consistency in calculated lKff over a wide range of the spectral index (energy of the average lethargy causing fission).

Appendix 4A, Page 6

4A.6 References

[4A. 1]

J.F. Briesmeister, Ed., "MCNP4a - A General Monte Carlo N-Particle Transport Code, Version 4A; Los Alamos National Laboratory, LA-12625-M (1993).

[4A.2]

SCALE 4.3, "A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation", NUREG-0200 (ORNL-NUREG-CSD-2/U2/R5, Revision 5, Oak Ridge National Laboratory, September 1995.

[4A.3] M.D. DeHart and S.M. Bowman, "Validation of the SCALE Broad Structure 44-G Group ENDF/B-Y Cross-Section Library for Use in Criticality Safety Analyses", NUREG/CR-6102 (ORNL/TM-12460)

Oak Ridge National Laboratory, September 1994.

[4A.4]

W.C. Jordan et al., "Validation of KENOV.a", CSD/TM-238, Martin Marietta. Energy Systems, Inc., Oak Ridge National Laboratory, December 1986.

[4A.5]

O.W. Hermann et al., "Validation of the Scale System for PWR Spent Fuel Isotopic Composition Analysis", ORNL-TM-12667, Oak Ridge National Laboratory, undated.

[4A.6]

R.J. Larsen and M.L. Marx, An Introduction to Mathematical Statistics and its Applications, Prentice-Hall, 1986.

[4A.7]

M.N. Baldwin et al., Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel, BAW-1484-7, Babcock and Wilcox Company, July 1979.

[4A.8]

G.S. Hoovier et al., Critical Experiments Supporting Underwater Storage of Tightly Packed Configurations of Spent Fuel Pins, BAW-1645-4, Babcock & Wilcox Company, November 1991.

[4A.9]

L.W. Newman et al., Urania Gadolinia: Nuclear Model Development and Critical Experiment Benchmark, BAW-18 10, Babcock and Wilcox Company, April 1984.

Appendix 4A, Page 7

[4A. 10] J.C. Manaranche et al., "Dissolution and Storage Experimental Program with 4.75 w/o Enriched Uranium-Oxide Rods,t" Trans.

Am. Nucl. Soc. 33: 362-364 (1979).

[4A. 11] S.R. Bierman and E.D. Clayton, Criticality Experiments with Subcritical Clusters of 2.35 w/o and 4.31 w/o 235U Enriched U0 2 Rods in Water with Steel Reflecting Walls, PNL-3602, Battelle Pacific Northwest Laboratory, April 1981.

[4A. 12] S.R. Bierman et al., Criticality Experiments with Subcritical Clusters of 2.35 w/o and 4.31 w/o 235U Enriched U0 2 Rods in Water with Uranium or Lead Reflecting Walls, PNL-3926, Battelle Pacific Northwest Laboratory, December, 1981.

[4A. 13] S.R. Bierman et al., Critical Separation Between Subcritical Clusters of 4.31 w/o '3'U Enriched U0 2 Rods in Water with Fixed Neutron Poisons, PNL-2615, Battelle Pacific Northwest Laboratory, October 1977.

[4A. 14] S.R. Bierman, Criticality Experiments with Neutron Flux Traps Containing Voids, PNL-7167, Battelle Pacific Northwest Laboratory, April 1990.

[4A. 151 B.M. Durst et al., Critical Experiments with 4.31 wt % 235U Enriched U0 2 Rods in Highly Borated Water Lattices, PNL-4267, Battelle Pacific Northwest Laboratory, August 1982.

'[4A. 16] S.R. Bierman, Criticality Experiments with Fast Test Reactor Fuel Pins in Organic Moderator, PNL-5803, Battelle Pacific Northwest Laboratory, December 1981.

[4A. 17] E.G. Taylor et al., Saxton Plutonium Program Critical Experiments for the Saxton Partial Plutonium Core, WCAP-3385-54, Westinghouse Electric Corp., Atomic Power Division, December 1965.

[4A. 18] M.G. Natrella, Experimental Statistics, National Bureau of Standards, Handbook 91, August 1963.

Appendix 4A, Page 8

Table 4A.1 Summary of Criticality Benchmark Calculations Calculated kr EALF ' (eV)

MCNP4a KENO5a Reference Identification Enrich.

MCNP4a KENO5a 1

B&W-1484 (4A.7)

Core I 2.46 0.9964 +/- 0.0010 0.9898+/- 0.0006 0.1759 0.1753 2

B&W-1484 (4A.7)

Core H1 2.46 1.0008 +/- 0.0011 1.0015 +/- 0.0005 0.2553 0.2446 3

B&W-1484 (4A.7)

Core mI 2.46 1.0010 +/- 0.0012 1.0005 +/- 0.0005 0.1999 0.1939 4

B&W-1484 (4A.7)

Core IX 2.46 0.9956 +/- 0.0012 0.9901 +/- 0.0006 0.1422 0.1426 5

B&W-1484 (4A.7)

Core X 2.46 0.9980 + 0.0014 0.9922 +/- 0.0006 0.1513 0.1499 6

B&W-1484 (4A.7)

Core XN 2.46 0.9978 + 0.0012 1.0005 +/- 0.0005 0.2031 0.1947 7

B&W-1484 (4A.7)

Core XII 2.46 0.9988 +/- 0.0011 0.9978 +/- 0.0006 0.1718 0.1662 8

B&W-1484 (4A.7)

Core XII 2.46 1.0020 +/- 0.0010 0.9952 +/- 0.0006 0.1988 0.1965 9

B&W-1484 (4A.7)

Core XIV 2.46 0.9953 +/- 0.0011 0.9928 +/-10.0006 0.2022 0.1986 10 B&W-1484 (4A.7)

Core XV 2.46 0.9910'+/- 0.0011 0.9909 +/- 0.0006 0.2092 0.2014 11 B&W-1484 (4A.7)

Core XVI 2.46 0.9935 +/- 0.0010 0.9889 +/- 0.0006 0.1757 0.1713 12 B&W-1484 (4A.7)

Core XVII 2.46 0.9962 +/- 0.0012 0.9942 - 0.0005 0.2083 0.2021 13 B&W-1484 (4A.7)

Core XVIII 2.46 1.0036 +/- 0.0012 0.9931 +/- 0.0006 0.1705 0.1708 Appendix 4A, Page 9

Table 4A.1 Summary of Criticality Benchmark Calculations Calculated k.,

Reference Identification Enrich.

MCNP4a KENO5a EALFt (eV)

MCNP4a KENO5a 14 B&W-1484 (4A.7)

Core XIX 2.46 0.9961 +/- 0.0012 0.9971 +/- 0.0005 0.2103 0.2011 15 B&W-1484 (4A.7)

Core XX 2.46 1.0008 +/- 0.0011 0.9932 +/- 0.0006 0.1724 0.1701 16 B&W-1484 (4A.7)

Core XXI 2.46 0.9994 +/- 0.0010 0.9918 +/- 0.0006 0.1544 0.1536 17 B&W-1645 (4A.8)

S-type Fuel, w/886 ppm B 2.46 0.9970 +/- 0.0010 0.9924 +/- 0.0006 1.4475 1.4680 18 B&W-1645 (4A.8)

S-type Fuel, w/746 ppm B 2.46 0.9990 +/- 0.0010 0.9913 +/- 0.0006 1.5463 1.5660 19 B&W-1645 (4A.8)

SO-type Fuel, w/1156 ppm B 2.46 0.9972 +/- 0.0009 0.9949 +/- 0.0005 0.4241 0.4331 20 B&W-1810 (4A.9)

Case 1 1337 ppm B 2.46 1.0023 +/- 0.0010 NC 0.1531 NC 21 B&W-1810 (4A.9)

Case 12 1899 ppm B 2.46/4.02 1.0060 +/- 0.0009 NC 0.4493 NC 22-French (4A.10)

Water Moderator 0 gap 4.75 0.9966 +/- 0.0013 NC 0.2172 NC 23 French (4A.10)

Water Moderator 2.5 cm gap 4.75 0.9952 +/- 0.0012 NC 0.1778 NC 24 French (4A.10)

Water Moderator 5 cm gap 4.75 0.9943 +/- 0.0010 NC

0. 1677 NC 25 French (4A.10)

Water Moderator 10 cm gap 4.75 0.9979 +/- 0.0010 NC 0.1736 NC 26 PNL-3602 (4A.11)

Steel Reflector, 0 separation 2.35 NC 1.0004 + 0.0006 NC 0.1018 Appendix 4A, Page 10

Table 4A.1 Summary of Criticality Benchmark Calculations Calculated k__

EAL ' (eb~

MCNP4a KENO5a Reference Identification Enrich.

MCNP4a KENO5a 22 27 PNL-3602 (4A.11)

Steel Reflector, 1.321 cm sepn.

2.35 0.9980 +/- 0.0009 0.9992 +/- 0.0006 0.1000 0.0909 28 PNL-3602 (4A.11)

Steel Reflector, 2.616 cm sepn 2.35 0.9968 +/- 0.0009 0.9964 +/- 0.0006 0.0981 0.0975 29 PNL-3602 (4A.11)

Steel Reflector, 3.912 cm sepn.

2.35 0.9974 +/- 0.0010 0.9939 +/- 0.0006 0.0976 0.0970 30 PNL-3602 (4A.11)

Steel Reflector, infinite sepn.

2.35 0.9962 +/- 0.0008 0.9939 +/-- 0.0006 0.0973 0.0968 31 PNL-3602 (4A.11)

Steel Reflector, 0 cm sepn.

4.306 NC 1.0003 +/- 0.0007 NC 0.3282 32 PNL-3602 (4A.11)

Steel Reflector, 1.321 cm sepn.

4.306 0.9997 +/- 0.0010 1.0012 +/- 0.0007 0.3016 0.3039 33 PNL-3602 (4A.11)

Steel Reflector, 2.616 cm sepp.

4.306 0.9994 +/- 0.0012 0.9974 +/- 0.0007 0.2911 0.2927 34 PNL-3602 (4A.11)

Steel Reflector, 5.405 cm sepn.

4.306 0.9969 +/- 0.0011 0.9951 +/- 0.0007 0.2828 0.2860 35 PNL-3602 (4A.11)

Steel Reflector, Infinite sepn.

4.306 0.9910 +/- 0.0020 0.9947 +/- 0.0007 0.2851 0.2864 36 PNL-3602 (4A.11)

Steel Reflector, with Boral Sheets 4.306 0.9941 +/- 0.0011 0.9970 +/- 0.0007 0.3135 0.3150 37 PNL-3926 (4A.12)

Lead Reflector, 0 cm sepn.

4.306 NC 1.0003 +/- 0.0007 NC 0.3159 38 PNL-3926 (4A.12)

Lead Reflector, 0.55 cm sepn.

4.306 1.0025 +/- 0.0011 0.9997 +/- 0.0007 0.3030 0.3044 39 PNL-3926 (4A.12)

Lead Reflector, 1.956 cm sepn.

4.306 1.0000 +/- 0.0012 0.9985 +/- 0.0007 0.2883 0.2930 Appendix 4A, Page 11

Table 4A. 1 Summary of Criticality Benchmark Calculations

. Calculated k..

EALF I (eV)

MCNP4a KENOSa Reference Identification Enrich.

MCNP4a KENO5a 40 PNL-3926 (4A.12)

Lead Reflector, 5.405 cm sepn.

4.306 0.9971 +/- 0.0012 0.9946 +/- 0.0007 0.2831 0.2854 41 PNL-2615 (4A.13)

Experiment 004/032 - no absorber 4.306 0.9925 +/- 0.0012 0.9950 +/- 0.0007 0.1155 0.1159 42 PNL-2615 (4A.13)

Experiment 030

- Zr plates 4.306 NC 0.9971 +/- 0.0007 NC 0.1154 43 PNL-2615 (4A.13)

Experiment 013

- Steel plates 4.306 NC 0.9965 +/- 0.0007 NC 0.1164 44 PNL-2615 (4A.13)

Experiment 014

- Steel plates 4.306 NC 0.9972 +/- 0.0007 NC 0.1164 45 PNL-2615 (4A.13)

Exp. 009 1.05% Boron-Steel plates 4.306 0.9982 +/- 0.0010 0.9981 +/- 0.0007 0.1172 0.1162 46 PNL-2615 (4A.13)

Exp. 012 1.62% Boron-Steel plates 4.306 0.9996 +/- 0.0012 0.9982 +/- 0.0007 0.1161 0.1173 47 PNL-2615 (4A.13)

Exp. 031 - Boral plates 4.306 0.9994 +/- 0.0012 0.9969 +/- 0.0007 0.1165 0.1171 48 PNL-7167 (4A.14)

Experiment 214R - with flux trap 4.306 0.9991 +/- 0.0011 0.9956 +/- 0.0007 0.3722 0.3812 49 PNL-7167 (4A.14)

Experiment 214V3 - with flux trap 4.306 0.9969 +/- 0.0011 0.9963 +/- 0.0007 0.3742 0.3826 50 PNL-4267 (4A.15)

Case 173 - 0 ppm B 4.306 0.9974 +/- 0.0012 NC 0.2893 NC 51 PNL-4267 (4A.15)

Case 177 - 2550 ppm B 4.306 1.0057 +/- 0.0010 NC 0.5509 NC 52 PNL-5803 (4A.16)

MOX Fuel - Type 3.2 Exp. 21 20% Pu 1.0041 +/- 0.0011 1.0046 +/- 0.0006 0.9171 0.8868 Appendix 4A, Page 12

Table 4A.1 Summary of Criticality Benchmark Calculations Calculated k.

EALF (eV)

Reference Identification Enrich.

MCNP4a KENO5a MCNP4a KENO5a 53 PNL-5803 (4A.16)

MOX Fuel - Type 3.2 Exp. 43 20% Pu 1.0058 +/- 0.0012 1.0036 +/- 0.0006 0.2968 0.2944 54 PNL-5803 (4A.16)

MOX Fuel - Type 3.2 Exp. 13 20% Pu 1.0083 +/- 0.0011 0.9989 +/- 0.0006 0.1665 0.1706 55 PNL-5803 (4A.16)

MOX Fuel - Type 3.2 Exp. 32 20% Pu 1.0079 +/- 0.0011 0.9966 +/- 0.0006 0.1139 0.1165 56 WCAP-3385 (4A.17)

Saxton Case 52 PuO2 0.52" pitch 6.6% Pu 0.9996 +/- 0.0011 1.0005 +/- 0.0006 0.8665 0.8417 57 WCAP-3385 (4A.17)

Saxton Case 52 U 0.52" pitch 5.74 1.0000 +/- 0.0010 0.9956 +/- 0.0007 0.4476 0.4580 58 WCAP-3385 (4A.17)

Saxton Case 56 PuO2 0.56" pitch 6.6% Pu 1.0036 +/- 0.0011 1.0047 + 0.0006 0.5289 0.5197 59 WCAP-3385 (4A.17)

Saxton Case 56 borated PuO2 6.6% Pu 1.0008 +/- 0.0010 NC 0.6389 NC 60 WCAP-3385 (4A.17)

Saxton Case 56 U 0.56" pitch 5.74 0.9994 +/- 0.0011 0.9967 +/- 0.0007 0.2923 0.2954 61 WCAP-3385 (4A.17)

Saxton Case 79 PuO2 0.79" pitch 6.6% Pu 1.0063 +/- 0.0011 1.0133 +/- 0.0006 0.1520 0.1555 62 WCAP-3385 (4A.17)

Saxton Case 79 U 0.79" pitch 5.74 1.0039 +/- 0.0011 1.0008 +/- 0.0006 0.1036 0.1047 Notes: NC stands for not calculated.

t EALF is the energy of the average lethargy causing fission.

tt These experimental results appear to be statistical outliers (> 3o) suggesting the possibility of unusually large experimental error. Although they could justifiably be excluded, for conservatism, they were retained in determining the calculational basis.

Appendix 4A, Page 13

Table 4A.2 COMPARISON OF MCNP4a AND KENO5a CALCULATED REACTIVITIESt FOR VARIOUS ENRICHMENTS Calculated k,,, + la Enrichment MCNP4a KENO5a 3.0 0.8465 +/- 0.0011 0.8478 +/- 0.0004 3.5 0.8820 +/- 0.0011 0.8841 +/- 0.0004 3.75 0.9019 + 0.0011 0.8987 +/- 0.0004 4.0 0.9132 +/-:0.0010 0.9140 +/- 0.0004 4.2 0.9276 +/-:0.0011 0.9237 +/- 0.0004 4.5 0.9400 + 0.0011 0.9388 +/- 0.0004 Based on the GE 8x8R fuel assembly.

Appendix 4A, Page 14

Table 4A.3 MCNP4a CALCULATED REACTIVITIES FOR CRITICAL EXPERIMENTS WITH NEUTRON ABSORBERS Ak MCNP4a Worth of Calculated EALF t Ref.

Experiment Absorber k1 (eV) 4A. 13 PNL-2615 Boral Sheet 0.0139 0.9994+/-0.0012 0.1165 4A.7 B&W-1484 Core XX-0.0165 1.0008+/-0.0011 0.1724 4A. 13 PNL-2615 1.62% Boron-steel 0.0165 0.9996+/-0.0012 0.1161 4A.7 B&W-1484 Core XIX 0.0202 0.9961+/-0.0012 0.2103 4A.7 B&W-1484 Core XXI 0.0243 0.9994+/-0.0010 0.1544 4A.7 B&W-1484 Core XVII 0.0519 0.9962+/-0.0012 0.2083 4A.11 PNL-3602 Boral Sheet 0.0708 0.9941+/-0.0011 0.3135 4A.7 B&W-1484 Core XV 0.0786 0.9910+/-0.0011 0.2092 4A.7 B&W-1484 Core XVI 0.0845 0.9935+/-0.0010 0.1757 4A.7 B&W-1484 Core XIV 0.1575 0.9953+/-0.0011 0.2022 4A.7 B&W-1484 Core XIII 0.1738 1.0020+/-0.0011 0.1988 4A.14 PNL-7167 Expt 214Rflux trap 0.1931 0.9991+/-0.0011 0.3722 tEALF is the energy of the average lethargy causing fission.

Appendix 4A, Page 15

Table 4A.4 COMPARISON OF MCNP4a AND KENO5a CALCULATED REACTIVITIESt FOR VARIOUS 10B LOADINGS Calculated kf + lo

'OB, glcm2 MCNP4a KENO5a 0.005 1.0381 +/- 0.0012 1.0340 +/- 0.0004 0.010 0.9960 +/- 0.0010 0.9941 +/- 0.0004 0.015 0.9727 + 0.0009 0.9713 +/- 0.0004 0.020 0.9541 +/- 0.0012 0.9560 + 0.0004 0.025 0.9433 +/- 0.0011 0.9428 +/- 0.0004 0.03 0.9325 -I0.0011 0.9338 +/- 0.0004 0.035 0.9234 +/- 0.0011 0.9251 +/- 0.0004 0.04 0.9173 +/- 0.0011 0.9179 +/- 0.0004 Based on a 4.5% enriched GE 8x8R fuel assembly.

Appendix 4A, Page 16

Table 4A.5 CALCULATIONS FOR CRITICAL EXPERIMENTS WITH THICK LEAD AND STEEL REFLECTORSt Separation, Ref.

Case E, wt%

cm MCNP4a k,,

KENO5a k,,

4A. 11 Steel 2.35 1.321 0.9980+/-0.0009 0.9992+/-0.0006 Reflector 2.35 2.616 0.9968+/-0.0009 0.9964+/-0.0006 2.35 3.912 0.9974+/-0.0010 0.9980+/-0.0006 2.35 00 0.9962-+/-0.0008 0.9939+/-0.0006 4A. 11 Steel 4.306 1.321 0.9997+/-0.0010 1.0012+/-0.0007 Reflector 4.306 2.616 0.9994+/-0.0012 0.9974+/-0.0007 4.306 3.405 0.9969+/-0.0011 0.9951+/-0.0007 4.306 c

0.9910+/-0.0020 0.9947+/-0.0007 4A. 12 Lead 4.306 0.55 1.0025+/-0.0011 0.9997+/-0.0007 Reflector 4.306 1.956 1.0000+/-0.0012 0.9985+/-0.0007 4.306 5.405 0.9971+/-0.0012 0.9946+/-0.0007 t

Arranged in order of increasing reflector-fuel spacing.

Appendix 4A, Page 17

Table 4A. 6 CALCULATIONS FOR CRITICAL EXPERIMENTS WITH VARIOUS SOLUBLE BORON CONCENTRATIONS Calculated keg Boron Concentration, Reference Experiment ppm MCNP4a KENO5a 4A. 15 PNL-4267 0

0.9974 +/- 0.0012 4A.8.

B&W-1645 886 0.9970 +/- 0.0010 0.9924 +/- 0.0006 4A.9 B&W-1810 1337 1.0023 +/- 0.0010 4A.9 B&W-1810 1899 1.0060 +/--0.0009 4A. 15 PNL-4267 2550 1.0057 +/- 0.0010 Appendix 4A, Page 18

Table 4A.7 CALCULATIONS FOR CRITICAL EXPERIMENTS WITH MOX FUEL MCNP4a KENO5a Reference Caset kff EALF" kdr EALF" PNL-5803 MOX Fuel - Exp. No. 21 1.0041+/-0.0011 0.9171 1.0046+/-0.0006 0.8868

[4A. 16]

MOX Fuel - Exp. No. 43 1.0058+/-0.0012 0.2968 1.0036+/-0.0006 0.2944 MOX Fuel - Exp. No. 13 1.0083 +/-0.0011 0.1665 0.9989+/-0.0006 0.1706 MOX Fuel - Exp. No. 32 1.0079+/-0.0011 0.1139 0.9966+/-0.0006 0.1165 WCAP-Saxton @ 0.52" pitch 0.9996+/-0.0011 0.8665 1.0005+/-0.0006 0.8417 3385-54 14A. 17]

Saxton @ 0.56" pitch 1.0036+/-0.0011 0.5289 1.0047+/-0.0006 0.5197.

Saxton @ 0.56" pitch borated 1.0008+/-0.0010 0.6389 NC NC Saxton @ 0.79" pitch 1.0063+/-0.0011 0.1520 1.0133+/-0.0006 0.1555 Note, NC stands for not calculated t

Arranged in order of increasing lattice spacing.

tt EALF is the energy of the average lethargy causing fission.

Appendix 4A, Page 19

-Linear Regression with Correlation Coefficient of 0.13 1.010 1.005 a)

.4-C.)

a) a)

Va)

.4-a

3 C.)

00 1.000 0.995 0.990 0.1 1

Energy of Average Lethargy Causing Fission (Log Scale)

FIGURE 4A.1 MCNP CALCULATED k-eff VARIOUS VALUES OF THE VALUES for SPECTRAL INDEX

Linear Regression with Correlation Coefficient of 0.21 1.010 1.005 ai 1.000 4--

-,0.995 0

0.990 0.985 0.1 Energy of Average Lethargy Causing Fission (Log Scale) 1 FIGURE 4A.2 KENO5a VARIOUS CALCULATED VALUES OF k-eff VALUES FOR THE SPECTRAL INDEX

- - -Linear Regression with Correlation Coefficient of 0.03 1.010 1.005 U

1.000 a) 4-0 U

0.995 0.990 Enrichment,.w/o U-235 FIGURE 4A.3 MCNP CALCULATED k.eff VALUES AT VARIOUS U-235 ENRICHMENTS

Linear Regression with Correlation Coefficient of 0.38 1.010 1.005 4-1, 1.000 4--

"0 4-

.2 0.995 0.990 0.985

___Bias 0

li I

I I If I I III III.

I II T 7 T

77 I

1111111I 11II1II 1II1II1 I I

I-

1. '1 6.0 Z.°U L.Z.

1). U

.5.

4.0 4.5 5.0 5.5 6.0 Enrichment, w/o U-235 FIGURE 4A.4 KENO CALCULATED k-eff VALUES AT VARIOUS U-235 ENRICHMENTS

0.94 0.92 E

CD 0

z 0.90 0.88 0.86 0.84 MCNP k-eff Calculations FIGURE 4A.5 COMPARISON OF MCNP AND KENO5A CALCULATIONS FOR VARIOUS FUEL ENRICHMENTS

1.04-1.03-

,i, 1.92-z 2J*

1.0-3~0.99-S 0.96-

,5 0.1.96-~

0A 0.020 g/-q

.g 0.5 0.025

//cmq 0.94-

• 0.93

.030 g/m 0.035 g/am5q 0. 9 2 0./4 0.020.04or/a q

0.900 0.920 0.940 0.960 0.980 1.000 1

Reactivity Calculated with KENO5a FIGURE 4A.6 COMPARISON OF MCNP AND KEN05o CALCULATIONS FOR VARIOUS BORON-I1 AREAL DENSITIES to AEP:NRC:8431 DONALD C. COOK NUCLEAR PLANT UNIT 1 TECHNICAL SPECIFICATION PAGES MARKED TO SHOW CHANGES 4.0-4 4.0-5

Design Features 4.0 MODULE Al MODULE cl

• MODULE El MODULE A5 I

/

0 -504REGION I CELLS ED 4439 REGION 2 CELLS

]

1678 REGION 3 CELLS The storage pattern for any of these individual modules may be as shown in this figure or Figure 4.3-2.

Figure 4.3-1 (page 1 of 1)

Normal Storage Pattern (Mixed Three Zone)

Cook Nuclear Plant Unit 1 4.0-4 Amendment No. 287

Design Features 4.0 MODULE Al MODULE C1 MODULE El MODULE A5 ME M 0 M M M WOMMOM11111 mioeoo Q4 0 11111 I r-4 M V-4 M Q4 *4

• 4 Q4 Q M Q

r-4 M

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[I

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-664-REGION I CELLS El 1439 REGION 2 CELLS

[

4366. REGION 3 CELLS The storage pattern for any of these individual modules may be as shown in this fi-gure or Figure 4.3-1.

Figure 4.3-2 (page 1 of 1)

Interim Storage Pattern (Checkerboard)

Cook Nuclear Plant Unit 1 4.0-5 Amendment No. 287 to AEP:NRC:8431 DONALD C. COOK NUCLEAR PLANT UNIT 2 TECHNICAL SPECIFICATION PAGES MARKED TO SHOW CHANGES 4.0-4 4.0-5

Design Features 4.0 MODULE Al MODULE A5 0 --- 4REGION I CELLS E

41439 REGION 2 CELLS

]

1678 REGION 3 CELLS The storage pattern for any of these individual modules may be as shown in this figure or Fi-gure 4.3-2.

Figure 4.3-1 (page 1 of 1)

Normal Storage Pattern (Mixed Three Zone)

Cook Nuclear Plant Unit 2 4.0-4 Amendment No. 269

Design Features 4.0 MODULE Al MODULE C1 MODULE El MODULE A5 so 0 0 a 0 a o e

e

ý0

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1-55"REGION 3 CELLS The storage pattern for any of these individual modules may be as shown in this figure or Figure 4.3-1.

Figure 4.3-2 (page 1 of 1)

Interim Storage Pattern (Checkerboard)

Cook Nuclear Plant Unit 2 4.0-5 Amendment No. 269