ML061240463

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 - Non Proprietary Response to NRC RAI

ML061240463
Person / Time
Site:	Oconee
Issue date:	04/26/2006
From:	Brandi Hamilton Duke Energy Corp, Duke Power Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LAR 2005-009
Download: ML061240463 (26)

Text

BRUCE H HAMILTON

_Duke Vice President vEn1er'gye Oconee Nuclear Station Duke Energy Corporation ON0IVP / 7800 Rochester Highway Seneca, SC 29672 864 885 3487 864 885 4208 fax bhhamilton@duke-ene,'gy.corn April 26, 2006 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555-0001

Subject:

Duke Power Company LLC d/b/a Duke Energy Carolinas, LLC 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 - Duke Response to NRC Request For Additional Information License Amendment Request (LAR) 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 Power Company LLC d/b/a Duke Energy Carolinas, LLC (Duke) submitted an amendment to Renewed Facility Operating Licenses Nos. DPR-38, DPR-47, and DPR-55 on March 1, 2006. If granted, this amendment request will allow spent fuel loading, unloading, and handling operations in the Oconee Nuclear Site (Oconee) Spent Fuel Pools (5FP) that support spent fuel transfer to an Independent Spent Fuel Storage Installation (ISFSI) licensed under 10 CFR 72.

In a meeting with the Nuclear Regulatory Commission (NRC) on March 22, 2006, Duke provided an overview of the amendment request and discussed NRC Staff's initial concerns.

Duke also restated its commitment to respond to those concerns expeditiously in order to facilitate approval of the amendment request by June 1, 2006. On April 3, 2006, a request for additional information (RAI) was discussed in a conference call between the Staff and Duke.

Duke received the RAI on April 5, 2006 and this document is in response to the Staffs request.

There ale no new commitments being made as a result of this document.

www. duke-energy. corn

Nuclear Regulatory Commission LAR No. 2005-009 - Duke Response to NRC Request for Additional Information April 26, 2006 Page 2 Enclosures 3 and 4 contain the RAI responses. Enclosure 4 contains information proprietary to Transnuclear, Inc. and Areva NP. The RAI responses in Enclosure 4 have been reproduced in their entirety for ease of review. Affidavits from Transnuclear, Inc. and Areva NP are included in Enclosure 2. The affidavits set forth the basis on which the information may be withheld from public disclosure by the NRC pursuant to 10 CFR 2.790.

Inquiries on this 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. Affidavits for Transnuclear, Inc. and Areva NP

3. Duke Response to NRC Request for Additional Information - Non Proprietary

4. Duke Response to NRC Request for Additional Information - Proprietary

Nuclear Regulatory Commission LAR No. 2005-009 - Duke Response to NRC Request for Additional Information April 26, 2006 Paoe 3 bc w/enzlosures 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

Nuclear Regulatory Commission LAR No. 2005-009 - Duke Response to NRC Request for Additional Information April 26, 2006 Page 4 bcc w/enclosures and attachments:

B. Ci. Davenport R. M. Glover S. ]'. Nesbit G. R. Walden J. P. Coletta W. J. Murphy C. D. Fago S. J1.Perrero S. C. Newman R. V. Gambrell L. F. Vaughn S. ]). Capps T. 1P. Gillespie R.:L. Gill - NRI&IA R. D. Hart - CNS C. .1.Thomas - MNS NSRB, EC05N ELL, ECO50 File - T.S. Working ON'S Document Management

ENCLOSURE 1 NOTARIZED AFFIDAVIT

Enclosure I - Notarized Affidavit LAR No. 2005-009 - Duke Response to NRC Request for Additional Information April 26,2006 Page 1 AFFIDAVIT B. H. Hamilton, being duly sworn, states that he is Vice President, Oconee Nuclear Site, Duke Energy Carolinas, LLC 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 conrect to the best of his knowledge.

/44c~a~*~

B. H. Hamilton, Vice President Oconee Nuclear Site Subscribed and sworn to before me this 2*. day of . 2006 Notary ]?ublic My Commission Expires:

/ a -te

-I- Date SEAL

ENCLOSURE 2 AFFIDAVITS FOR TRANSNUCLEAR, INC. AND AREVA NP

E-23512 Page 1 of 3 April 4, 2006 AFFIDAVIT STATE OF MARYLAND }

3 COUNTY OF HOWARD }

Before me, the undersigned authority, personally appeared Tara J. Nelder who, being by me duly sworn according to law, deposes and says that she is

-authorized to execute this Affidavit on behalf of Transnuclear, Inc. and that the averments of fact set forth in this Affidavit are true and correct to the best of her knowledge, information, and belief:

-TARAJ. NEIDER Sworn to and subscribed before me this 4 day of ,2006, Commislo piotary P / 200 My Commission Expirest 208

E-23512 Page 2 of 3 April 4, 2006 (1) I am President and Chief Operating Officer of Transnuclear, Inc. and my responsibilities include reviewing the proprietary information sought to be withheld from public disclosure in connection with the licensing of spent fuel transport cask systems or spent fuel storage cask systems. I am authorized to apply for its withholding on behalf of Transnuclear, Inc.

(2) I am making this Affidavit in conformance with the provisions of 10 CFR Section 2.390 of the commission's regulations and in conjunction with the Transnuclear application for withholding accompanying this Affidavit.

(3) I have personal knowledge of the criteria and procedures utilized by Transnuclear in designating information as a trade secret, privileged or as confidential commercial or financial information.

(4) The following information is furnished pursuant to the provisions of paragraph 10 CFR 2.390(b)(4) to determine whether the information sought to be withheld from public disclosure should be withheld.

(i) The information sought to be withheld from public disclosure is owned and has been held In confidence by Transnuclear.

(ii) The information Is of a type customarily held in confidence by Transnuclear, is not customarily disclosed to the public and is transmitted to the commission in confidence.

(iii) The information sought to be protected is not now available in public sources to the best of our knowledge and belief and the release of such information might result in a loss of competitive advantage as follows:

(a) It reveals the distinguishing aspects of a storage system where prevention of its use by any of Transnuclear's competitors without license from Transnuclear constitutes a competitive economic advantage over other companies.

(b) It consists of supporting data, including analytical models, relative to a component or material, the application of which secures a competitive economic or technical advantage.

(c) Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.

E-2 3512 Page 3 of3.

April 4, 2006 (5) The information is being transmitted to the commission in confidence and, under the provision of 10 CFR Section 2.390, it is to be received in confidence by the Commission.

(6) The information sought to be protected is not available in public sources to the best of our knowledge and belief.

(7) The proprietary information, as shown, sought to be withheld is information contained in Amendment 6 of the TN NUHOMS-24P CoC 72-1004, as referenced in IOCFR72 Section 72.214.

(8) This information should be held in confidence because it provides details of analytical methods that were developed at significant expense. This information has substantial commercial value to Transnuclear in connecting with competition with other vendors for contracts.

The subject information could only be duplicated by competitors if they were to invest time and effort equivalent to that invested by Transnuclear provided they have the requisite talent and experience.

Public disclosure of this information is likely to cause substantial harm to the competitive position of Transnuclear, because it would simplify design and evaluation tasks without requiring a commensurate investment of time and effort.

4

AFFIDAVIT COMMONWEALTH OF VIRGINIA )

) ss.

CITY OF LYNCHBURG )

1. My name is Gayle F. Elliott. I am Manager, Product Licensing in Regulatory Affairs, forAREVA NP, and as such I am authorized to execute this Affidavit.

2. 1am familiar with the criteria applied by AREVA NP to determine whether certain AREVA NP Information is proprietary. I am familiar with the policies established by AREVA NP to ensure the proper application of these criteria.

3. 1am familiar with the attributes listed in Attachment A and referred to herein as "Document." Information contained in this Document has been classified by AREVA NP as proprietary in accordance with the policies established by AREVA NP for the control and protection of proprietary and confidential information.

4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by AREVA NP and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained In this Document as proprietary and confidential.

5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the Information contained in this Document be withheld from public disclosure.

6. The following criteria are customarily applied by AREVA NP to determine W-ether information should be classified as proprietary:

(a) The information reveals details of AREVA NP's research and development plans and programs or their results.

(b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.

(c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for AREVA NP.

(d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for AREVA NP in product optimization or marketability.

(e) The information is vital to a competitive advantage held by AREVA NP, wouN be helpful to competitors to AREVA NP, and would likely cause substantial harm to the competitive position of AREVA NP.

7. In accordance with AREVA NP's policies governing the protection and control of information, proprietary information contained in this Document has been made available, on a limited basis, to others outside AREVA NP only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8. AREVA NP policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.

9. The foregoing statements are true and correct to the best of my knowledge, information, and belief.

SUBSCRIBED before me this ______

day of, (LA u

, . P.

I K . .

. 2006.

_

(3."O 0" apsy Brenda C. Maddox NOTARY PUBLIC, COMMONWEALTH OF VIRGINIA MY COMMISSION EXPIRES: 7/31107

ENCLOSURE 3 DUKE RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION - NON PROPRIETARY

- Duke Response to NRC Request for Additional Information - Non-proprietary License Amendment Request No. 2005-09 April 26, 2006 Page 1 Enclosure 3 Duke Response to NRC Request for Additional Information - Non-proprietary Ouestion I Provide a description of the benchmark analysis and results used to determine the SCALE 4.4/KENO V.a bias and uncertainty.

Question 2 Provide a description of the benchmark analysis and results used to determine the CASMO-3/SIMULATE-3 bias and uncertainty.

Response to Questions 1 and 2 Table 1 lists the 58 specific critical experiments (from References I to 3) that were analyzed for benchmarking purposes with SCALE 4.4/KENO V.a. The calculated keff values for the SCALE 4.4/KENO V.a models of these experiments are also provided in Table 1. To determine the SCALE 4.4IKENO V.a method bias and uncertainty to be applied to the NUJHOMS-24P/24PHB DSC analysis, the following equations from Reference 4 are used:

N k

~ i2 Average ker k =N I N

E (o x NG,)

Average Variance VARavg N ENG, N

Method Bias BiasMethod = (Y)XK-j Method Uncertainty UnCAlfthd = f95,95

.Enclosure 3 - Duke Response to NRC Request for Additional Information - Non-proprietary License Amendment Request No. 2005-09 April 26, 2006 Page '

Where:

ki; = KENO V.a calculated krf for critical experiment i a1 = KENO V.a standard deviation for critical experiment i NG, = number of neutron generations used in KENO V.a analysis for critical experiment i (400for all experiments mnodeled in Table I)

N = number of KENO V.a critical experiments (58)

K; = measured value of kff for critical experiment i (1.000for each of 58 experiments) f95N5 = 95/95 one-sided tolerance factor (2.03for 58 experiments per Reference 5)

Table 2 lists the 10 benchmark critical experiments from Reference 8 that were evaluated with CASMO-3/SIMULATE-3. The CASMO-3/SIMULATE-3 calculated kenf values and experimentally measured keff values are included in Table 2. Because CASMO-3/SIMULATE-3 calculations yield deterministic solutions, the method bias and uncertainty calculations simplify to the following:

N Method Bias BiashChOd = (-) X E (K,-k,)

lNN l(K, - k,- BiasMe,,d )2 Method Uncertainty UnCAnethod =95/95X (N- 1)

Where:

ki = CASMO-3/SIMULATE-3 calculated kff for critical experiment i N = number of CASMO-3/SIMULATE-3 critical experiments (10)

K; = measured value of klff for critical experiment i (see Table 2) fg5,95 = 95/95 one-sided tolerance factor (2.911 for 10 experiments per Reference 5)

Note that the Reference 7 submittal for the Oconee spent fuel pool storage racks employed the same critical experiments for its criticality code benchmarking. Because the SCALE and SIMULATE versions used in Reference 7 have since been updated, the method biases and uncertainties resulting from analysis of these critical experiments are slightly different. Note also that the Reference 10 submittal employed many of the critical experiments in Tables 1 and 2 for its code benchmarking.

The fuel design parameters, storage cell spacing, and SFP conditions associated with loading fuel assemblies into the NUHOMS-24P/24PHB DSCs are quite similar to those associated with storage of assemblies in the Oconee SFP racks. The applicability of the set of critical experiments used in Reference 7 to the conditions in the Oconee SFPs thus extends to the loading of the NUHOMS-24P/24PHB DSCs.

Table 3 lists a set of important criticality parameters and the range of values for these parameters in the evaluated SCALE 4.4/KENO V.a and CASMO-3/SIMULATE-3 critical experiments.

Included for comparison are the values of these parameters for the NUHOMS-24P/24PHB DSCs (simplified infinite-array model described in Reference 6). This table shows that the selected benchmark critical experiments are appropriate for application to the NUHOMS-24P/24F'HB DSC model.

.Enclosure 3 - Duke Response to NRC Request for Additional Information - Non-proprietary License Amendment Request No. 2005-09 April 26, 2006 Page 3 Table 1. Calculated keff Values for the SCALE 4.4/KENO V.a Benchmark Critical Experiments Experiment Calculated Experiment Calculated Report Number a¢¢ cy (kff) Report Number ker C a (ker)

PNL-3314 043 0.99799 0.00198 PNL-3314 085 0.99430 0.00203 PNL-3314 045 0.99613 0.00181 PNL-3314 094 0.99733 0.00199 PNJL-3314 046 0.99185 0.00161 PNL-3314 095 0.99723 0.00198 PNL-3314 047 0.99937 0.00204 PNL-3314 096 0.99669 0.00198 PN4L-3314 048 0.99728 0.00193 PNL-3314 097 0.99767 0.00194 PNL-3314 04c 0.99604 0.00178 PNL-3314 098 0.99657 0.00204 PNL-3314 051 0.98920 0.00200 PNL-3314 100 0.99292 0.00198 PNL-3314 053 0.98302 0.00225 PNL-3314 101 0.99493 0.00213 PNL-3314 055 0.99403 0.00186 PNL-3314 105 0.99548 0.00195 PNL-3314 056 0.99130 0.00218 PNL-3314 106 0.99325 0.00206 PNL-3314 057 0.98979 0.00201 PNL-3314 107 0.99696 0.00214 PNL-3314 058 0.99355 0.00188 PNL-3314 131 0.99050 0.00170 PNL-3314 059 0.99184 0.00185 PNL-3314 996 0.98675 0.00173 PNL-3314 060 0.99099 0.00179 PNL-3314 997 0.98970 0.00187 PNL-3314 061 0.99213 0.00202 PNL-2438 005 0.99298 0.00151 PNL-3314 062 0.99537 0.00206 PNL-2438 014 0.99163 0.00174 PNL-3314 064 0.99351 0.00226 PNL-2438 015 0.99359 0.00174 PNL-3314 065 0.99185 0.00195 PNL-2438 021 0.99123 0.00182 PNL-3314 066 0.99018 0.00225 PNL-2438 026 0.99216 0.00164 PNL-3314 067 0.98951 0.00207 PNL-2438 027 0.98934 0.00155 PNL-3314 068 0.99025 0.00199 PNL-2438 028 0.99260 0.00148 PNL-3314 069 0.99716 0.00193 PNL-2438 029 0.99524 0.00175 PNL-3314 06d 1.00418 0.00161 PNL-2438 034 0.99118 0.00194 PNL-3314 070 0.98758 0.00184 PNL-2438 035 0.98978 0.00173 PNL-3314 071 0.99521 0.00181 PNL-6205 214 0.99190 0.00241 PNL-3314 072 0.99304 0.00181 PNL-6205 223 1.00122 0.00192 PNL-3314 073 0.98938 0.00176 PNL-6205 224 0.99256 0.00219 PNL-3314 083 0.99749 0.00178 PNL-6205 229 0.99829 0.00170 PNL-3314 084 0.99680 0.00269 PNL-6205 230 0.99744 0.00193 Average Calculated krf = 0.9936 SCALE 4.4/KENO V.a Method Bias = +0.0064 Ak (average)

SCALE 4.4/KENO V.a Method Uncertainty = +/-0.0066 Ak

.Enclosure 3 - Duke Response to NRC Request for Additional Information - Non-proprietary License Amendment Request No. 2005-09 April 26, 2006 Page 4 Table 2. Calculated kerf Values for the CASMO-3/SIMULATE-3 Benchmark Critical Experiments BAW-1484-7 Critical Experimentally SIMULATE-3 Ak Experiment Measured Calculated kf (Measured kff minus Core Number k,ff Calculated kff) 2 1.0001 1.00274 (0.00264) 3B 1.0000 1.00320 (0.00320) 9 1.0030 0.99905 0.00395 10 1.0001 0.99793 0.00217 11 1.0000 1.00497 (0.00497) 13B 1.0000 1.00926 (0.00926) 14 1.0001 1.00461 (0.00451) 15 0.9988 0.99611 0.00269 17 1.0000 0.99891 0.00109 19 1.0002 1.00003 0.00017 Average - (0.00145)

Deviation - { 0.00416 1ASMO-3/SIMULATE-3 Method Bias = -0.0015 Ak (average)

CASMO-3/SIMULATE-3 Method Uncertainty = 2.911*0.00416 = +/-0.0121 Ak Table 3. Important NUHOMS-24P/24PHB DSC Criticality Analysis Parameters and their Values for Selected Benchmark Critical Experiments Range of Values in Range of Values in Range of Values in Reference 6 Table 1 SCALE Table 2 CASMO-3' Parameter simplified infinite- 4.4/KENO V.a Critical SIMULATE-3 array DSC model Experiments Critical Experiments Lattice water-to-fuel volume ratio 1.62- 1.66 1.60 (48) and 2.92 (10) 1.84 U-235 Enrichment (wt % U-235) 1.60 - 5.00 2.35 - 4.31 2.46 Separation between Rod Arrays (cm) 4.47 0- 19.81 0 - 6.54 Solubl Boron Concentration (ppm) 0 - 630 0 0 - 1037

.Enclosure 3 - Duke Response to NRC Request for Additional Information - Non-proprietary License Amendment Request No. 2005-09 April 26, 2006 Page 5 Questioin 3 This LAR lists two mechanical uncertainties, one with unborated water, the other with water borated to 430 ppm of boron. Each mechanical uncertainty encompasses all manufacturing tolerances and uncertainties for the fuel and the cask.

a. Provide a list of the mechanical uncertainties associated with the fuel assemblies used to determine the two uncertainties listed in the LAR. Explain why these uncertainties are appropriate and how they bound all of the fuel designs listed in Table 2 of the LAR.

b. Provide a list of the mechanical uncertainties associated with the DSC used to determine the two uncertainties listed in the LAR.

Explain why these uncertainties are appropriate.

c. Explain the difference between the two mechanical uncertainties, one with unborated water, the other with water borated to 430 ppm of boron, used in the LAR.

d. Explain the method used to combine all of the above uncertainties into the two mechanical uncertainties listed in the LAR.

Response to Ouestion 3a Table 4 lists the parameters associated with fuel assemblies whose tolerances were observed to acquire mechanical uncertainty factors at both 0 and 430 ppm soluble boron. In addition, their individual contributions to the final results are provided.

With the exception of the guide tube parameters, each of these fuel assembly parameters were considered in the Reference 7 analysis of the Oconee spent fuel pools. Furthermore, the parameters considered here are consistent with those outlined, in Reference 9.

In deter-nining the mechanical uncertainty factors at 0 and 430 ppm soluble boron, the reactivity effects of the parameters in Table 4 were observed for each of the fuel designs listed in Table 2 of Enclosure 3 of Reference 6. The maximum resulting total mechanical uncertainty at 0 and 430 ppm soluble boron was chosen to conservatively bound all three of the fuel designs.

Enclosure 3 - Duke Response to NRC Request for Additional Information - Non-proprietary License Amendment Request No. 2005-09 April 26, 2006 Page 6 Table 4. Mechanical Uncertainty Factors at 0 and 430 ppm Soluble Boron, Fuel Assembly-Related Parameters Reactivity Effect Parameter Tolerance 0 ppm 430 ppm boron boron Fuel Enrichment [ ] 0.00194Ak 0.0022Ak Fuel Pellet Dish Volume Varies by type. 0.00048Ak 0.00052Ak Fuel Theoretical Density [ ] 0.0009Ak 0.00173Ak Fuel Pellet Outer Diameter 0.00023Ak 0.00047Ak Fuel Clad Outer Diameter [ ] 0.00177Ak 0.00108Ak Guide Tube Inner Diameter [ ] 0.00017Ak 0.00012Ak Guide Tube Outer Diameter [ ] 0.00018Ak 0.00013Ak Fuel Eccentricity (location in + 0.192" 0.0077Ak 0.00571Ak cell) (x and y coord.)

Response to Question 3b Table 5 lists the parameters associated with the DSC whose tolerances were observed to acquire mechanical uncertainty factors at both 0 and 430 ppm soluble boron. In addition, their individual contributions to the final results are provided.

These parameters in Table 5 were observed in the Reference 7 analysis, but for spentfuel pool storage cells (as opposed to DSC storage cells). The methodology is unchanged and remains valid as DSC loading conditions are similar to storage conditions in the Oconee spent fuel storage racks. In addition, the parameters in Table 5 are the only structural characteristics of the DSC that are considered in the homogeneous DSC model (Section 6.3 of Enclosure 3 of Reference 6).

Table 5. Mechanical Uncertainty Factors at 0 and 430 ppm Soluble Boron, DSC-Related Parameters Reactivity Effect Parameter Tolerance 0 ppm 430 ppm boron boron

[DSC] Cell Inside Dimension [ ] 0.01004Ak 0.0095Ak

[DSC] Cell Wall Thickness [ ] 0.00298Ak 0.00153Ak DSC] Cell Center-to-Center Spacing [ ] 0.02457Ak 0.02806Ak

Enclosure 3 - Duke Response to NRC Request for Additional Information - Non-proprietary License Amendment Request No. 2005-09 April 26, 2006 Page 7 Response to Question 3c The two mechanical uncertainties used in the Reference 6 analysis, 0.0280Ak and 0.0304Ak, were calculated at 0 and 430 ppm soluble boron, respectively. The same mechanical parameters in Tables 4 and 5 were observed in both cases; however, the presence (or lack) of soluble boron changes the reactivity impact of a given mechanical tolerance. Thus, the statistically-combined overall mechanical uncertainty factor will vary with soluble boron concentration based on the variations of its individual contributors.

Of the mechanical parameters listed in Tables 4 and 5, the DSC cell center-to-center spacing provided the largest reactivity effects at both 0 and 430 ppm soluble boron. This parameter in particular was analyzed in a very conservative fashion, with the pitch of the infinite homogeneous array model being increased/decreased by the [ ] tolerance from Table 5. The total statistically-combined mechanical uncertainty factors at 0 and 430 ppm, respectively, fromn of Reference 6 are 0.0280Ak and 0.0304Ak. This disparity in mechanical uncertainty values stems primarily from the increased reactivity worth of the borated water, which is displaced by the reduction in the DSC cell center-to-center spacing. Consulting the reactivity effect values in Tables 4 and 5 confirms that, indeed, the center-to-center spacing parameter is impacted the most by the addition of 430 ppm soluble boron.

Response to Question 3d Reference 9 states that the reactivity effects of tolerance variations may be combined statistically if they are independent. The following equations were used to statistically combine the independent tolerance variations into an overall mechanical uncertainty, which was then applied to the calculated nominal multiplication factor:

AkktechUnr = IX + (f25 2k2 anom)

Vi for KENO Va.

Aki = 4(ki -knom) +k(f9 5 *a,)

or, AkAtechUnr = Z(Ak,)2 Vi for CASMO-3 Ak5 =k1 - kno

.Enclosure 3 - Duke Response to NRC Request for Additional Information - Non-proprietary License Amendment Request No. 2005-09 April 26, 2006 Page 8 where,

!k; = uncertainty due to tolerance (i),

knom = CASMO-3 or KENO V.a Calculated keff for nominal mechanical parameters, Uinom = KENO V.a one-sigma uncertainty for nominal mechanical parameters, ki = CASMO-3 or KENO V.a Calculated krff for mechanical tolerance (i),

y = KENO V.a one-sigma uncertainty for mechanical tolerance (i),

£F95 = 95th percentile one-sided tolerance factor (1.727 for 1000 generations)

Both the upper and lower tolerances of each mechanical parameter were evaluated to determine which had a greater effect on the multiplication factor. Only the direction of the tolerance (upper or lower) which produced the largest positive difference of {k1 - knom) was used to determine the uncertainty in keff due to that particular parameter.

Questioin 4 Please explain the effect of placing burnable poison rod assemblies (BPRA) into fuel assemblies loaded into the DSC while in the SFP. Include any limitations on quantity or location of BPRAs.

Response to Question 4 In order to determine whether the presence of BPRA components produces a net increase in reactivity with the displacement of borated water, a CASMO-3 job was executed to analyze the infinite homogeneous DSC model with fresh (unirradiated) 5.00 wt% U-235 mbl fuel at 630 ppm soluble boron. Fully depleted (i.e. 100% A12 03 ) BPRA components were conservatively placed in each assembly of the infinite model. The maximum 95/95 keff from this case was then compared with the same model executed with no BPRA components present. As this comparison was performed with unirradiated, maximum enrichment (5.00 wt% U-235) fuel assemblies with fully depleted BPRA components infinitely modeled and with the maximum soluble boron concentration credited in Reference 6, this analysis fully bounds all DSC loading conditions.

The maximum 95/95 keff for unirradiated 5.00 wt% U-235 mbl fuel in the infinite homogeneous DSC model with 630 ppm soluble boron and depleted BPRA components is 1.15765, while the maximum 95/95 keff for the same model with no BPRA components modeled is 1.16650. The presence of depleted BPRAs led to a net decrease in reactivity of 0.00885Ak, indicating that, at the maximum (630 ppm) soluble boron concentration credited in Reference 6, it remains conservative and bounding to assume that no BPRA components are present.

Enclosure 3 - Duke Response to NRC Request for Additional Information - Non-proprietary License Amendment Request No. 2005-09 April 26), 2006 Page 9 Questioin 5 The LAR briefly describes the accident analysis associated with the misloading of a single unirradiated 5.0 w/o U-235 fuel assembly in the DSC. The LAR states this is the most limiting accident. With respect to the accident analysis provide the following:

a. A description of the controls that limit misleading event to one assembly.

b. A description of the analysis and results that led to the conclusion that the remaining accidents are bounded by the misloading accident.

Response to Question 5a Page 9 of Attachment 3 in the Reference 7 submittal states the following, with regard to misloading fuel assemblies in the Oconee SFP storage racks:

"Reference 4 [Kopp letter] requires that only a singlefiel assembly misload be analyzed unless there are circumstances that make multiple loading errorscredible. Redundant checks and proceduralverifications of eachfitel assembly movement within the Oconee spent fiuel pools preclude the occurrenceof multiple fitel assembly loading errorsin any storage region. "

The same procedural verifications and redundant checks that are used with spent fuel movements are also in effect during NUHOMSO-24P/24PHB DSC loading operations. These include:

1. Fuel movement instructions prepared and independently reviewed by qualified engineers in accordance with approved technical procedures.

2. Bridge positioning for fuel movement performed and independently verified by qualified fuel handlers in accordance with approved technical procedures and the fuel movement instructions. Each fuel move is independent of any other fuel move.

The General Office Spent Fuel Management group generates the list of fuel assemblies to be loaded into the DSC. This list is created and independently verified through the use of an approved procedure that follows the fuel selection process as set forth in the ISFSI Technical Specifications. This list is formally transmitted to the Reactor Engineering group at ONS.

The Reactor Engineering group uses this list to generate the fuel move sheets that are provided to the fuel handling group. The fuel move sheets are created and independently verified using an approved procedure.

NUHOMIS Certificate of Compliance (72-1004) Technical Specification 1.2.1 states.....

"Immediately, before insertion of a spent fuel assembly into a DSC, the identity of each fuel assembly shall be independently verified and documented." The controlling procedure for loading fuel into the DSC requires this as a blind verification performed independently by twc individuals. This verification is made using an underwater camera. Additionally all fuel moves are made using a blind verification technique. The fuel bridge operator has a copy of the fuel

Enclosure 3 - Dulke Response to NRC Request for Additional Information - Non-proprietary License Amendment Request No. 2005-09 April 2(i, 2006 Page 10 move sheet showing the step number, fuel assembly ID, withdraw location, and insert location.

The fuel bridge operator is directed to perform a step number by the procedure controller. When the fuel bridge is over the location called for by the move sheet, a fuel bridge spotter verifies the fuel bridge is in the proper location. During this process the step number is the only information that is verbally communicated. The fuel bridge spotter does not have access to the fuel move sheets, and does not know where the fuel bridge is supposed to be. The procedure controller verifies that the spotter has called out the proper location, and then directs the fuel bridge operator to withdraw, or insert the fuel assembly.

These barriers, in and of themselves, are sufficient to preclude a single fuel assembly misload and render multiple loading errors non-credible.

Response to Ouestion 5b From Section 6.5 of Enclosure 3 of Reference 6, a misloaded MkB 11 fuel assembly requires 630 ppm soluble boron credit in order to maintain DSC system keff under 0.95. The following description of the remaining credible accident scenarios demonstrates that the misloading accidenti requires the greatest quantity of soluble boron to remain under regulatory limits.

• Seismic Events Per Refierence 9, the analysis must "consider the effect on criticality of natural events (e.g.

earthquakes) that may deform, and change the relative position of, the storage racks and fuel ill the spent fuel pool." The mechanical uncertainty calculation performed in support of the Reference 6 analysis considered the placement of fuel assemblies into the most optimum possible pitch (consistent with the Reference 10 analysis); that is, each assembly is positioned as close to one another as possible within their respective storage cells). The maximum reactivity impact of such a transient (from Table 4) is 0.0077Ak for 0 ppm soluble boron credit. From Section 6.5 of of Reference 6, the maximum 95/95 keff for the DSC system with 430 ppm soluble boron credit is 0.9264. With the aforementioned reactivity increase, the maximum 95/95 keff for the DSC system during a seismic transient is 0.9341 with 430 ppm soluble boron credit, resulting in a 0.0Ol59Ak margin from the regulatory maximum of 0.95. Thus, this accident scenario is bounded by the misload presented in Enclosure 3 of Reference 6.

• Abnormal Water Temperatures Referenze 9 states that "abnormal temperatures (above those normally expected) and the reactivity consequences of void formation (boiling) should be evaluated." The criticality analysis determined that a water temperature of 150'F is more reactive than the lower nominal temperature limit, 680 F. Thus, a water temperature of 150'F is assumed when calculating the minimum burnup requirements for DSC loading. In order to fully analyze all criticality consequences for abnormal temperature conditions, temperatures both above and below those considered by the nominal analysis were evaluated as well as voiding effects at higher temperatures. The maximum 95/95 k~ff of 0.93365 occurs at 2120 F with 0 percent water voiding with credit taken for 430 ppm soluble boron (i.e. the partial boron credit assumed in the nominal analysis), resulting in a 0.01635Ak margin from the regulatory maximum of 0.95. Clearly, this

.Enclosure 3 - Duke Response to NRC Request for Additional Information - Non-proprietary License Amendment Request No. 2005-09 April 26, 2006 Page II accident scenario is bounded by the misload presented in Enclosure 3 of Reference 6.

• Fuel Assembly Drop In considering the consequences of a drop of a fuel assembly, several types of drop accidents are postulated. A drop resulting in an assembly residing immediately adjacent to the NUHOMS transfer cask would be essentially neutronically decoupled from the cask, as the thickness of the shielded transfer cask would provide for at least 17 inches of spacing between the dropped assembly and the loaded assemblies.

A drop resulting in an assembly falling on top of the DSC could result in one of several outcomes. The most likely outcome is the fuel assembly coming to rest on top of the DSC upper spacer disk. In such a scenario, a sufficient amount of spacing is present between the active fuel regions of the dropped assembly and those of the loaded assemblies to preclude any neutronic interaction between the dropped and stored fuel. Were a dropped assembly to land on an already-stored assembly, the impact would, at worst, slightly compress the stored assembly; however, the distance between the dropped and loaded assemblies would still remain sufficient to preclude any interaction, and the compression of the stored assembly would, at worst, lend itself to a slight reactivity change with the change in the water-to-fuel ratio. The worst-case misload accident - a fresh, maximum enrichment fuel assembly loaded into a DSC storage cell - clearly bounds any fuel dro-?/misplacement scenario, as a fuel assembly dropped anywhere on/outside the shielded canister would be essentially isolated from the active fuel in the DSC.

• Spent Fuel Pool (SFP) Dilution Accident A dilution accident concurrent with loading a DSC in the SFP, while highly unlikely, is a credible transient; however, the presence of a DSC or related activities does not create any additional initiating mechanisms for such an event. Thus, the dilution analysis currently supporting the Oconee SFPs (submitted to NRC in Reference 7) also supports cask loading operations in the SFPs. The dilution analysis concluded that, for both spent fuel pools, at least.

32.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> must pass before the spent fuel pool boron concentration is reduced to the credited quantity of 430 ppm. This amount of time is more than sufficient for site personnel to initiate action to mitigate the situation. Thus, this accident poses no risk to criticality safety.

-Enclosure 3 - Duke Response to NRC Request for Additional Information - Non-proprietary License Amendment Request No. 2005-09 April 26, 2006 Page 12 References

1. 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.

2. Critical Separation Between Subcritical Clusters of 2.35 wt % U-235 Enriched U02 Rods in Water with Fixed Neutron Poisons, PNL-2438, October 1977.

3. Criticality Experiments to Provide Benchmark Data on Neutron Flux Traps, PNL-6205, June 1988.

4. "Validation of YAEC Criticality Safety Methodology," D. Napolitano and F.

Carpenito, ANS 1988 Annual Meeting Transactions, Vol 56, pp 325-327.

5. Factors for One-Sided Tolerance Limits and for Variables Sampling Plans, SCR-607, Sandia Corporation, March 1963.

6. "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" - Transmittal from B. Hamilton (Duke Power) to U.S. NRC, March 1, 2006.

7. "Duke Energy Corporation, Oconee Nuclear Station, Units 1, 2 and 3; Docket Numbers 50-269, 50-270 and 50-287; Response to Request for Additional

]nformation - Proposed Technical Specification Amendment; Generic Letter 96-04

-- Spent Fuel Storage Racks (TSCR 2000-01)" - Transmittal from W. McCollum (D ike Power) to U.S. NRC, October 31, 2001. (Technical Specification amendment request approved by NRC via SER dated April 22, 2002)

8. Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel, BAW-1484-7, July 1979.

9. "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.

10. "Joseph M. Farley Nuclear Plant, Technical Specification Revision, Spent Fuel C.ask Loading Requirements" - Transmitted from L. Stinson (Southern Company) to U.S.

NRC, May 17, 2005. (Technical Specification amendment approved by NRC via SER dated June 28, 2005)