Text

Duke Energy Corporation'S Response to Blue Ridge Environmental Defense League'S First Discovery Request
	ML041130392
Person / Time
Site:	Catawba
Issue date:	04/14/2004
From:	Repka D; Duke Energy Corp, Winston & Strawn, LLP
To:	; Blue Ridge Environmental Defense League, Office of Nuclear Reactor Regulation
	Byrdsong A T
References
	50-413-OLA, 50-414-OLA, ASLBP 03-815-03-OLA, RAS 7635
	Download: ML041130392 (75)
	v • d • e

--RAs 763

fTriAED OORRESPONDENM April 14, 2004 UNITED STATES OF AMERICA USNRC NUCLEAR REGULATORY COMMISSION April 20,2004 (4:06PM)

BEFORE THE ATOMIC SAFETY AND LICENSING BOARD OFFICE OF SECRETARY RULEMAKINGS AND In the Matter of: )ADJUDICATIONS STAFF DUKE ENERGY CORPORATION )

) Docket Nos. 50-413-OLA (Catawvba Nuclear Station, ) 50-414-OLA Units I and 2) )

)

DUKE ENERGY CORPORATION'S RESPONSE TO BLUE RIDGE ENVIRONMENTAL DEFENSE LEAGUE'S FIRST DISCOVERY REOUEST In accordance with the schedule established by the Atomic Safety and Licensing Board ("Licensing Board") in its March 30, 2004 "Order (Confirming Matters Addressed at March 25 Telephone Conference)," Duke Energy Corporation ("Duke") hereby files its Response to the March 31, 2004 "Blue Ridge Environmental Defense League's First Set of Discovery Requests Directed to Duke Energy Corporation." Consistent with the direction of the Licensing Board in its "Order (Confirming Matters Addressed at April 6 Telephone Conference)," dated April 8, 2004, this response does not address Contention 111. (Order, at ¶8.)

This response includes certain documents that are proprietary and are governed by the Protective Order issued by the Licensing Board on April 8, 2004.1 Such documents have been identified as proprietary in the index of documents (Attachment 3) included with this "Memorandum and Order (Protective Order Governing Non-Disclosure of Proprietary Information)," dated April 8, 2004.

emptotez-S6c Y- c35 Oa*

response and the copies of the documents provided have been marked accordingly. These documents must be handled by BREDL in accordance with the Protective Order.2 I. RESPONSE TO BREDL GENERAL DISCOVERY REOUEST A. Duke Response To General Interrogatories GENERAL INTERROGA TORY NO. 1: State the name, business address, and job title of each person who was consulted and/or who supplied information for responding to each of the interrogatories, requestsfor admission, and requests for the production of documents posed by BREDL herein. Specifically note for which interrogatories,requestsfor admissions and requestsfor production each such person was consulted and/orsupplied information.

If the information or opinions of anyone who was consulted in connection wvith your response to an interrogatory or request for admission differs from your written answer to the discovery request, please describe in detail the differing information or opinions, and indicate why such differing infonnation or opinions are not your official position as expressed in your written ansiver to the request.

Duke Response to General Interrogatory 1:

The name, business address, and job title of each person who "was consulted and/or who supplied information" for Duke in response to each BRMDL interrogatory and document production request is set forth in Attachment I to this Response. Attachment 1 also lists for each person the interrogatories and document requests for which they were consulted.

Among those persons who supplied information for Duke's responses, no information or.

opinions received concerning the interrogatories were different from the information and opinions provided herein.

2 Counsel for BREDL and BREDL's technical consultant executed a Confidentiality and Non-Disclosure Agreement in accordance with the Protective Order on April 13, 2004.

2

GENERAL INTERROGA TORY NO. 2: For Contentions I, II, and III, give the name, address, profession, employer, area of professional expertise, and educational and scientific experience of each person whom Duke expects to call as a fact or expert witness at the hearing. For expert witnesses, provide a list of all publications authored by the witness within the preceding ten years and a listing of any other cases in which the witness has provided fact and/or expert testimony and/orsubmitted affidavit(s) or declaration(s) within the precedingfour years. For purposes of answering this interrogatory, the educational and scientific experience of expected witnesses may be provided by a resume of the person attached to the response. Fact and expert witnesses should be distinguished.

Duke Response to General Interrogatory 2:

For Contentions I and II, the requested information for Duke's expected witnesses is set forth in Attachment 2 to this Response. All wvitnesses are considered to be expert witnesses. Resumes for each prospective witness are included in Attachment 2. Note that Duke has not yet prepared testimony on Contentions I and II. If and when additional witnesses are identified, Duke will supplement its response to General Interrogatory 2 as appropriate.

GENERAL INTERROGATORYNO. 3: For each witness identified in response to General InterrogatoryNo. 2 above, describe the facts and opinions to which each witness is expected to testify, including a sumnmary of the groundsfor each opinion, and identify the documents (including all pertinent pages or parts thereof), data or other information which each witness has reviewed and considered, or is expected to consider or to rely onfor his or her testimony.

Duke Response to General Interrogatory 3:

The facts and opinions as to which each Duke witness is expected to testify are summarized below:

> Steven P. Nesbit Mixed oxide ("MOX") fuel is a well-established technology with widespread use in European nuclear power reactors. Duke's safety and environmental analyses submitted to the NRC have demonstrated that the MOX fuel lead assembly program can be conducted at Catawba without undue risk to the health and safety of the public. The substitution of four MOX fuel 3

assemblies for four low enriched uranium ("LEU") fuel assemblies will have no significant impact on Emergency Core Cooling System ("ECCS") performance, design basis event radiological consequences, or the probability of a severe accident at Catawba. In addition, changes in severe accident consequence impacts - as estimated based on various sources-should be less than such changes associated with power uprates. Power uprates are commonly approved by the Nuclear Regulatory Commission without requiring probabilistic risk assessments from the applicant.

Grounds for opinion:

Duke Power analyses, as documented in the February 27, 2003 Duke Energy Corporation MOX fuel lead assembly license amendment request ("LAR") and subsequent related submittals (e.g., responses to NRC requests for additional information).

Reviews of assessments and sensitivity studies related to severe accident consequences involving MOX fuel lead assemblies.

NRC regulatory reviews associated with power uprates.

Professional experience in nuclear safety analysis.

)> Robert C. Harvey The AREVA Loss of Coolant Accident ("LOCA") analyses for the MOX fuel lead assemblies provide a conservative estimation of ECCS performance at Catawba. The AREVA analyses are consistent with approved 10 CFR Part 50, Appendix K methodologies, adjusted appropriately to reflect the properties of MOX fuel.

Grounds for opinion:

Duke Power analyses, as documented in the February 27, 2003 Duke MOX fuel lead assembly LAR and subsequent related submittals.

Professional experience in nuclear safety analysis.

4

> Bert Dunn The AREVA LOCA analyses for the MOX fuel lead assemblies provide a conservative estimation of ECCS performance at Catawba. The AREVA analyses are consistent with approved 10 C.F.R. Part 50 Appendix K methodologies, adjusted appropriately to reflect the properties of MOX fuel. The LOCA analyses are documented in the February 27, 2003 MOX fuel lead assembly LAR and subsequent related submittals.

M5 cladding has been previously approved by the NRC and is currently used in the nuclear industry. The M5 cladding response is modeled in a conservative manner, and cladding ballooning does not pose a threat to core coolability. Clad ductility is not a MOX fuel effect and is not impacted by MOX fuel characteristics. In any event, the MOX fuel LOCA analyses assume the worst case clad ductility in the blockage model.

MOX fuel characteristics should not exacerbate any adverse fuel relocation effects in a LOCA or severe accident in a significant manner. VERCORS test results do not alter the conclusion that the MOX fuel lead assemblies can be safely utilized. One VERCORS test indicated fuel relocation, but that test was at temperatures far in excess of design basis LOCA temperatures. Moreover, any relocation at design bases temperatures is bounded by conservatisms inherent in the Appendix K criteria and the ECCS analysis models.

Grounds for opinion:

Duke Power analyses as documented in the February 27, 2003 MOX fuel lead assembly LAR and subsequent related submittals.

M5 Topical Report (BAW-10277-A), "Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel" (AREVA/Framatome ANP), and related NRC submittals and safety evaluations.

36 years experience in the evaluation of accidents in nuclear reactors, including experimental evaluation, calculations, and licensing.

5

Consultations with the international nuclear safety expert community through membership in several forums.

> H. Duncan Brewer The substitution of four MOX fuel assemblies for four low enriched uranium fuel assemblies will have no significant impact on the probability of a severe accident at Catawba.

Four MOX fuel lead assemblies will have no significant impact on the progression of a hypothetical core melt at Catawba.

Duke also appropriately estimated severe accident consequence impacts in the Environmental Report that was included with the February 27, 2003 MOX fuel lead assembly LAR. The estimated increase in public health consequences - assuming the radiological characteristics of four MOX fuel lead assemblies - is well within the uncertainties inherent in the analysis and is not significant in the context of a probabilistic risk assessment. For example, in NUREG-1 150, "Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants" (December 1990), the NRC studied risks from severe accidents at five commercial nuclear power plants, including Sequoyah, an ice condenser containment plant similar to Catawba. As part of the work, NRC generated quantitative risk values, and it also examined the uncertainty inherent in the results. It is apparent that the results include considerable uncertainty, with a range of two orders of magnitude between the 5th and 95th percentile results, and a difference of approximately a factor of 3 between the mean and the median. In this context, the small change in public health risk associated with MOX fuel lead assemblies is much smaller than the uncertainty inherent in the calculations and therefore is not significant. With four MOX fuel lead assemblies, the risk due to operation of Catawba is well below the NRC's Safety Goals.

Duke's conclusions regarding the small increase in public health risk associated with four lead assemblies are being confirmed by Catawba-specific analyses, based on Duke's 6

Level 3 Probabilistic Risk Assessment ("PRA"), of severe accidents involving the four MOX fuel lead assemblies.

Grounds for opinion:

Duke Power analyses, as documented in the February 27, 2003 MOX fuel lead assembly LAR and subsequent related submittals.

Risk analysis work prepared by the Department of Energy

("DOE") in connection with the plutonium disposition program.

Level III Probabilistic Risk Assessment analysis for MOX fuel lead assemblies at Catawba.

Review of published risk analysis uncertainty analyses such as NUREG-1 150.

NRC Safety Goal Policy and other NRC guidelines and guidance documents on risk assessment.

. Professional experience in nuclear severe accident and probabilistic risk analysis.

Robert Henry The substitution of four MOX fuel assemblies for four low enriched uranium fuel assemblies will have no significant impact on the probability of a severe accident at Catawba.

Four MOX fuel lead assemblies will have no significant impact on the progression of a hypothetical core melt at Catawba. Postulated MOX fuel assembly-specific radionuclide release fractions and release timing should not significantly perturb the overall plant response to design basis events or severe accident scenarios, or the resulting offsite consequences.

Grounds for Opinion:

Duke Power analyses, as documented in the February 27, 2003 MOX fuel lead assembly LAR and subsequent related submittals.

Level III Probabilistic Risk Assessment analysis of MOX fuel lead assemblies at Catawba.

7

Professional experience in nuclear safety and severe accident analyses.

Documents That May Be Relied Upon:

Further in response to General Interrogatory 3, the documents, data or other information that each prospective witness may to rely upon in his testimony are listed below:3

1. Duke Energy Corporation MOX Fuel Lead Assembly License Amendment Request ("LAR") dated February 27, 2003, and related submittals and responses to NRC requests for additional information

("RAIs"). The LAR is available through ADAMS (ML030760734).

BREDL has previously received copies of Duke's RAI Responses submitted in connection with the LAR, except for those that include NRC Safeguards Information.

2. BAW-10238 (NP), Rev. 1, MOX Fuel Design Report (AREVA/Framatome ANP), and related submittals and responses to NRC requests for additional information. The non-proprietary Fuel Design Report is available through ADAMS (ML031550349).

3. Technical Report No. 415, "Status and Advances in MOX Fuel Technology," International Atomic Energy Agency (2003). This document is publicly available through IAEA.

4. Edwin Lyman, "Public Health Risks of Substituting Mixed-Oxide for Uranium Fuel in Pressurized-Water Reactors," Science and Global Security, 2000. BREDL is in possession of this document.

5. ERIINRC 02-202, "Accident Source Terms for Light-Water Nuclear Power Plants: High Burnup and Mixed Oxide Fuels," October 2002. This document is available through ADAMS (ML023500093).

6. DOE/EIS-0229, DOE Storage and Dispositionof W~eapons-Usable Fissile MaterialsFinalProgrammaticEnvironmentalImpact Statement ("PEIS'),

December, 1996. BREDL is in possession of this document.

7. DOE/EIS-0283, Surplus Plutonium Disposition Environmental Impact Statement ("SPDEIS'), Department of Energy, 1999. BREDL is in possession of this document.

3 If additional responsive documents or other information are identified, Duke will supplement its Response to this interrogatory as appropriate.

8

8. October 23, 2003 Presentation by IRSN (Institut de Radioprotection et de Suret6 Nuclkaire) to the NRC. This document is available through ADAMS (ML032970624).

9. Malgouyres, Ferroud-Plattet, Ducros, Poletiko, Tourasse, and Buolard, "Influence of MOX Fuel in Fission Product Release Up to Meltdown Conditions," American Nuclear Society Nureth-9 Conference, San Francisco, CA, October 3-8, 1999. This document is available through the American Nuclear Society.

10. DCS-FQ-1999-001, Rev. 2, Fuel Qualification Plan (AREVA/Framatome ANP), April 2001. This document is available through ADAMS (ML013390597).

11. Travers, W. D., Memorandum to the Nuclear Regulatory Commission, "Agency Plan for Confirmatory Research Associated with the Use of Mixed-Oxide Fuel in Commercial Light Water Reactors," February 11, 2000. This document is available through ADAMS (ML003676642).

12. Travers, W. D., Memorandum to the Nuclear Regulatory Commission, "Mixed-Oxide Fuel Use in Commercial Light Water Reactors," April 14, 1999. This document is available through ADAMS (ML993620025).

13. 66 Federal Register 1,158 (January 5, 2001), "Nuclear Regulatory Commission - Commonwealth Edison Company, Byron Station, Units I and 2, Braidwood Station, Units I and 2; Environmental Assessment and Finding of No Significant Impact."

14. BAW-10277-A, "Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel" (AREVA/Framatome ANP), and related submittals and responses to NRC RAIs, February 2000. This document is available through ADAMS (ML003686367).

15. Catawba Updated Final Safety Analysis Report, Chapter 15.6 (Loss of Coolant Accident Analysis).

16. NUREG-I 150, "Severe Accident Risks: An Assessment for Five U.S.

Nuclear Power Plants" (December 1990). This document is available through ADAMS (ML040140729).

17. 51 Federal Register 30,028 (Aug. 21, 1986), NRC "Safety Goals for the Operations of Nuclear Power Plants; Policy Statement."

18. Anderson, S.L., Laubbam, T.J., Nesbit, S.P., and Gilreath, J.D., "Mixed Oxide Fuel Effects on the Integrity of the McGuire and Catawba Reactor Vessels." (This paper, presented at the Sept. 2002 American Nuclear Society (ANS) 5th Topical Meeting on U.S. Department of Energy Spent Nuclear Fuel and Fissile Materials Management, Charleston, S.C., is 9

included in the proceedings for that meeting, which are publicly available from ANS.)

19. Eller, J.L., "Reactor Core Model Benchmark for Partial MOX Fuel Cycles." (This paper, presented at the Oct. 2003 ANS meeting on Advances in Nuclear Fuel Management III, Hilton Head, S.C., is included in the proceedings for that meeting, which are publicly available from ANS.)

20. Nesbit, S.P., and Eller, J.L., "Basis for the Design of Reactor Cores Containing Weapons Grade MOX Fuel." (This paper, presented at the Oct. 2003 ANS meeting on Advances in Nuclear Fuel Management III, Hilton Head, S.C., is included in the proceedings for that meeting, which are publicly available from ANS.)

21. NUREG/CR-5249, "Quantifying Reactor Safety Margins - Application of Code Scale, Applicability, and Uncertainty Evaluation Methodology to a Large-Break, Loss-of-Coolant-Accident," U.S. Nuclear Regulatory Commission, December 1989. This document is available through ADAMS (ML030380473 and ML030380495).

In addition, as described above, a Level 3 Probabilistic Risk Assessment analysis of MOX fuel lead assembly use at Catawba is being performed by Duke. Duke's responses to document requests described below include a final calculation of radionuclide inventories for a core that is all LEU and for a core with MOX fuel lead assemblies. This is a final document used as input to the Level 3 analysis. Duke 'vill supplement its response with the final Level III analysis report when completed. Preliminary drafts and work papers are being treated as privileged, because the work is being prepared at the direction of counsel for preparation of testimony in this proceeding.

B. Duke Response To General Document Production Requests REOUESTNO. 1: All documents in yourpossession, custody or control that are identified, referred to or used in any way in responding to all of t/e above general interrogatories and the following interrogatories and requests for admissions relatingto specific contentions.

10

Duke Response to General Document Production Request 1:

See Response to Gcncral Interrogatory 3 above. Duke believes that all of the cited documents are publicly available or already in the possession of BREDL.

REOUESTANO. 2: All documents in your possession, custody or control relevant to each BREDL admitted contention, and to the extent possible, segregated by contention and separatedfromalreadyproduced documents.

Duke Response to General Document Production Request 2:

Duke's document production necessarily focused on documents that form the basis for Duke's position and that may be relied upon, and on issues raised in Contention I and II that are also addressed in other specific document requests. The documents identified in response to General Interrogatory 3 above, and in response to Specific Document Requests Il-I through 11-6 below, appear to comprehensively respond to this very general request for "relevant" documents. A list of all documents being provided in response to BREDL's discovery requests is included as Attachment 3. That list notes specific discovery requests to which each document appears to be responsive. A Privilege Log is provided as Attachment 4.

Consistent with the Licensing Board Order of April 8, 2004, Duke has not construed documents related to dose analyses of non-LOCA or non-core melt design basis events to be within the scope of either Contention I or II. Duke also understands Contention I to cover only LOCA design basis accident analyses related to the criteria of 10 C.F.R. § 50.46, and not to cover analyses related to releases. Contention II is limited to core melt events, whether resulting from LOCA or another initiator, and resulting consequences.

Duke also has not included in its response analyses (e.g., LOCA analyses) that are: (1) related only to potential batch-scale use of MOX fuel; or (2) related to LOCA analysis generally, but do not pertain to or involve differences between MOX fuel and LEU fuel.

11

Duke has not attempted to identify every pleading filed in this proceeding or in the completed McGuire/Catawba license renewal proceeding, or every draft of such pleading, whether including comments or not. While perhaps in a broad sense "relevant" to Contentions I and II, Duke believes that these documents are either publicly available (final pleadings) or are privileged under the attorney-work product and attorney-client communication privileges. These documents are not always shown on the Privilege Log, simply because of the number of documents involved and the time available to respond to the discovery requests.

Duke's responses to this and the other BREDL document requests also comport with the discussions of e-mail during the April 6, 2004 telephone conference and the Licensing Board's Order of April 8, 2004. (Order, at 1 4). Duke has limited its review of documents to relevant MOX Project files and files of other relevant Duke organizations and personnel responsible for input to the MOX Project on the lead assembly license amendment request.

REOUEST N1O. 3: All documents (including experts' opinions, workpapers, affidavits, and other materials used to render such opinion) supporting or othervise relating to testimony or evidence that you intend to use in the.hearing on each BREDL admitted contention.

Duke Response to General Document Production Request 3:

See Response to General Interrogatory 3 and Request 2. Duke has not yet prepared testimony on Contentions I or 11. If and when additional responsive documents are identified, Duke will supplement its response to General Document Production Request 3 as appropriate. In this regard, see the discussion above regarding Duke's ongoing Level 3 plant-specific risk assessment of MOX fuel lead assemblies.

12

HI. RESPONSE TO BREDL SPECIFIC DISCOVERY REQUESTS A. Response To Specific Interrogatories SPECIFIC INTERROGATORY NO. -1: Identify and describe in detail all experimental data and analysis relevant to your claim that the digferences between MOX and LEU fiel performance during a design-basis LOCA arefully accountedfor in your application and RAI responses, includingyour claim in the 11/4/04 RAI response that the methods employed by Duke for performing LOCA calculations, including calculations offutel clad ballooning, "are applicable to both MOX and LEUfifel because they are independent of the pellet type." RAI response at 8. Address this subject in the context of the NRC's statement that "chemical bonding between the pellets and the cladding, which may be different for MOX pellets and U02, may affect the ballooningprocess and hence the fuel behavior." Memorandum from William. Travers, Executive Director for Operations, to NRC Commissioners re: Agency Plan for Confirmatory Research Associated With the Use of Mixed-Oxide Fuel in Commercial Light Water Reactors, Attachment at 2 (February11, 2000).

Duke Response to Specific Interrogatory 1-1:

Provided there is reasonable similarity in the fuel pin design, the fuel pin (except for power and decay heat), is not dominant in determining the outcome of a LOCA. Reference 1 captures the importance of the various phenomena impacting the results of a LOCA. The identified controlling phenomena are listed in 13 categories, only one of which includes some phenomena that relate to the nuclear material. Within this latter category, fuel pellet enthalpy at operating conditions, fuel decay heat, gap conductance, and cladding oxidation are listed as significant contributors, and only the first three of those relate to nuclear material. All 12 of the remaining categories are related to the reactor system and independent of the fuel type. Thus, fuel pellet materials (e.g., MOX ceramic pellets) with approximate equivalence to uranium dioxide ceramic pellets in operational enthalpy, fuel decay heat and gap conductance will perform equivalently during a LOCA.

To provide further assurance that MOX fuel would not differ significantly from LEU fuel during a LOCA, AREVA reviewed MOX fuel programs in the U.S., France, and 13

Britain for LOCA-significant phenomena. Areas of concentration during the review were reactor kinetics, decay heat, and thermal and mechanical properties. Reference 2, Section 3.7.1.1, documents (i) the phenomena that were given specific consideration for the MOX fuel lead assembly LOCA evaluation, and (ii) the disposition that was given to each phenomenon. It should be noted that Duke has chosen not to take credit for MOX-LEU differences such as lower decay heat that would be beneficial to the LOCA analysis for MOX fuel lead assemblies. This provides added assurance that the licensing basis analysis for the lead assemblies is conservative.

Concerning pellet-cladding bonding and the potential impact on cladding swelling, both MOX and LEU fuel can develop pellet-cladding bonding during a design basis LOCA. The differences between MOX and LEU fuel in this area, if any, lie in the strength of the bond and the integrity of the pellet surface-body connection. Duke is aware of no specific experimental data directly relevant to differences between MOX and LEU fuel with respect to pellet-cladding bonding. As described below, the MOX fuel lead assembly LOCA evaluation treats cladding swelling and rupture in a manner such that any postulated small difference in the degree of pellet-cladding bonding does not impact the conservative nature of the calculation.

The established requirements for deterministic LOCA calculational approaches, adopted in Appendix K to IOCFR50.46, require that the degree of cladding swelling and incidence of rupture not be underestimated. The general impact of pellet-cladding bonding is to provide some additional support or effective strength to the cladding, potentially reducing the occurrence of cladding rupture and limiting cladding strain. In order to ensure that cladding swelling and rupture are modeled conservatively, consistent with Appendix K requirements, deterministic LOCA evaluations are typically based on data taken from unirradiated cladding 14

with no provision for pellet-cladding bonding. This conservative approach was used for the MOX fuel lead assembly LOCA evaluation as described in Reference 2.

In addition, it should be noted that the phenomena of cladding swelling and rupture, although they occur during design basis evaluations, were not identified as being of high level importance in Reference 1. Therefore, differences in these phenomena, which could be postulated to occur because of alternative pellet or cladding materials, should not be expected to produce significant differences in LOCA performance.

Finally, this evaluation of the potential impact of MOX fuel on LOCA performance is consistent with the assessment performed by the NRC and documented in Reference 3 (see page 11 of the attachment, Fuel Behavior section). The NRC noted the importance of initial stored heat to the LOCA analysis; the AREVA MOX fuel lead assembly LOCA analyses factored this effect into account as documented in Reference 2, Attachment 3, Section 3.7.1.1.4. Concerning chemical bonding between the pellets and the cladding, the NRC noted that the effect ". . . may be different" and ". . . could be investigated." The NRC further observed that ". . . a major effect is not expected." As described above, any pellet-cladding bonding effect, if present, was accounted for conservatively in the AREVA calculations and should at any rate be minor relative to the overall evaluation of conformance with 10 C.F.R. 50.46.

References:

1. NUREG/CR-5249, "Quantifying Reactor Safety Margins - Application of Code Scale Applicability, and Uncertainty Evaluation Methodology to a Large-Break, Loss-of-Coolant-Accident," U.S. Nuclear Regulatory Commission, December 1989.

2. Duke MOX Fuel Lead Assembly LAR, February 27, 2003.

15

3. Travers, W.D., Memorandum to the Nuclear Regulatory Commission, "Mixed-Oxide Fuel Use in Commercial Light-Water Reactors," April 14, 1999.

SPECIFIC INTERROGATORY NO. 2: Identif and describe in detail all experimental data and analysisjustifying your omission of a design-basis LOCA consequence analysis in the license amendment application. In particular,please address the extent to which you took into account in your assessment the wide uncertainty bands for AfOX early in-vessel release fractions for most fission product categoriespresented in the expert panel report "Accident Source Terms for Light-Water Nuclear Power Plants: High-Burnup and Mixed-Oxide Fuels, Energy Research, Inc., ERI/NRC 02/0202 at 38 (November 2002).

Duke Response to Specific Interrogatory 2:

Duke objected to this interrogatory and any associated document production request, on grounds that dose impacts from a design basis LOCA is an issue beyond the scope of any admitted contentions. In its April 8, 2004 "Order (Confirming Matters Addressed at April 6 Telephone Conference)," the Licensing Board addressed this question at ¶ 6. In its Order, the Licensing Board directed that Duke respond to this interrogatory in the context of Contention II as it relates to any consequences resulting from LOCAs, design basis accidents, or severe accidents. Accordingly, Duke is providing the following response.

Contrary to BREDL's assertion in this interrogatory, Duke's February 27, 2003 MOX fuel lead assembly LAR did not "omit" evaluation of dose consequences following a design basis LOCA. The dose impacts are addressed in the LAR safety analysis, Attachment 3, Sections 3.7.2 and 3.7.3. Duke provided additional, revised assessments of dose consequences following a LOCA in responses to NRC Requests for Additional Information dated November 3, 2003, and March 16, 2004 (already provided to BREDL).

Concerning the impact of "wide uncertainty bands for MOX early in-vessel release fractions," the referenced report ERIJNRC 02-202 did not include quantitative uncertainty estimates. Therefore, there are, in that report, no "wide uncertainty bands" as 16

referred to in the interrogatory. BREDL is perhaps referring to the differences among values provided by different experts for some parameters. However, such differences are not unusual for expert elicitations, particularly in areas such as severe accidents that necessarily involve considerable inherent uncertainty. In addition, it should be noted that the ERI/NRC 02-202 report is relevant only to an Alternative Source Term ("AST") methodology for calculating doses following a design basis event. Catawba's current licensing basis for dose following a design basis LOCA is the TID-14844 source term, not AST. The TID-14844 source term is generally recognized to be more conservative than AST. Furthermore, any postulated "MOX fuel effect" on early in-vessel release fractions would be confined to 2% of the fuel in the core, and should therefore be minimal relative to offsite doses, as reflected in the Duke analyses. For these reasons Duke considers the Duke assessment of MOX fuel impact on dose following a LOCA to be suitably conservative to support the conclusion that public health and safety are adequately protected.

Duke Response To Specific Document Requests REQUEST NO. I-1: Any and all documents containing data or analyses describedin response to InterrogatoryI-1 above.

Duke Response to Specific Document Request 1-1:

The references listed in the response to Interrogatory I-I are publicly available documents.

REQUEST NO. II-1: Any and all documents containing results of severe accident consequence assessments that have been conducted by Duke for the Catawba nuclearpowerplant during the pastfive years. If documents responsive to this request include documents that have already been provided to BREDL in the license renewal casefor Catawba andMcGuire, it will be sufficient to identify the documents.

17

Duke Response to Specific Document Request I- 1:

The documents produced are listed in Attachment 3 to this response. Documents related to Generic Safety Issue ("GSI") 189 and severe accident mitigation alternatives

("SAMAs") for Catawba that were submitted to the NRC, or that are NRC documents, have not been provided unless they include specific substantive comments regarding severe accident consequences. Drafts of certain Duke documents related to GSI-I 89 and responsive to specific document requests below are provided, except to the extent they are privileged.

REQUEST NO. II-2: Any and all MACCS2 input files that Duke has used in consequence assessments for the Catawba nuclear power plant during the last five years, including the meteorological data file and the source term release fractionsfor all severe accidents considered.

Duke Response to Specific Document Request 11-2:

The documents produced are listed in Attachment 3 to this response.

REQUESTNO. 11-3: With respect to any consequence assessments identified in response to Request No. 1, any and all documents containing technical justifications for any assumptions made with respect to the quantities and characteristicsofradionuclidesreleasedfrom core to containment.

Duke Response to Specific Document Request 11-3:

The documents produced are listed in Attachment 3 to this response. In this response, and the response to Request II4, Duke is providing information on the Catawba Level 2 and Level 3 PRAs which contain the requested information and which references the MAAP code that is commonly used in the industry for these applications. The Level 2 and Level 3 documentation contains descriptions and assumptions about the core damage progression sequences and how they are used to select releases that are based on the MAAP analyses.

REQUESTNO. 11-4: With respect to any consequehce assessments identified in response to Request No. 1, any and all documents containing technical justifications for any assumptions made with respect to the quantities and characteristicsof radionuclidesreleasedfrom containment to the environment.

18

Duke Response to Specific Document Request 11-4:

The documents produced are listed in Attachment 3 to this response.

REQUEST NO. 11-5: Any and all documents containingany analysis by Duke of the potentialfor or consequences of severe accidents wven using MOX LTAs at the Catawba nuclearpowerplant.

Duke Response to Specific Document Request 11-5:

The documents produced are listed in Attachment 3 to this response.

REQUEST NO. II-6: Any and all documents discussing research regardingthe potentialfor increasedprobabilities or consequences of severe accidents during use of MOXfiuel at nuclearpower plants, including but not limited to studies by the Institut de Radioprotection et de Szorete Nucliaire ("IRSN') or its predecessor, IPSN.

Duke Response to Specific Document Request 11-6:

The documents produced are listed in Attachment 3 to this response.

Respectfully submitted, David A. Repka Anne W. Cottingham WINSTON & STRAWN, LLP 1400 L Street, NW Washington, D.C. 20005-3502 Lisa F. Vaughn DUKE ENERGY CORPORATION 422 South Church Street Mail Code: PB05E Charlotte, N.C. 28201-1244 ATTORNEYS FOR DUKE ENERGY CORPORATION Dated in Washington, District of Columbia This 14th day of April 2004 19

April 14, 2004 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of: )

DUKE ENERGY CORPORATION )

) Docket Nos. 50-413-OLA (Catawba Nuclear Station, ) 50-414-OLA Units I and 2) )

)

AFFIDAVIT OF BERT M. DUNN Bert M. Dunn hereby declares under penalty of perjury that the following statements are true and correct of his own knowledge:

I. In the captioned proceeding, I have supplied information in response to Specific Interrogatory No. I-1, as set forth in Blue Ridge Environmental Defense League's First Set of Discovery Requests Directed to Duke Energy Corporation, dated March 31, 2004.

2. For 34 years, I have been employed by AREVA, working in areas related to nuclear energy. I currently hold the title of Advisory Engineer in the NSSS Engineering department.

DC:353049.1

3. The information in these responses is true and correct to the best of my knowledge and belief.

Bert M. Dunn Subscribed to and Sworn before me personally, on this 14 day of April, 2004.

My Commission expires: -',I -/SlZ c

2 DC:353049.1

April 14, 2004 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of: )

)

DUKE ENERGY CORPORATION )

) Docket Nos. 50-413-OLA (Catawba Nuclear Station, ) 50-414-OLA Units I and 2) )

)

AFFIDAVIT OF ROBERT C. HARVEY Robert C. Harvey hereby declares under penalty of perjury that the following statements are true and correct of his own knowledge:

I. In the captioned proceeding, I have supplied information in response to Specific Interrogatory No. I-I, as set forth in Blue Ridge Environmental Defense League's First Set of Discovery Requests Directed to Duke Energy Corporation, dated March 31, 2004.

2. For 25 years, I have worked in areas related to nuclear energy. I am responsible for loss of coolant accident analyses for the Catawba Nuclear Station. I currently hold the title of Senior Engineer in the Duke Energy Nuclear Generation Department.

DC:353050.2

3. The information in these responses is true and correct to the best of my knowledge and belief.

p

[Name] 1*R4 Subscribed to and Sworn before me personally, on this NMI day of April, 2004.

Notary Public My Commission expires: ainuaff V2 } 26o v MICHAEL T. CASH Notary Public Lincoln County, North Carolina Commission Expires January 22, 2008 2

DC:3530502

April 14, 2004 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of: )

)

DUKE ENERGY CORPORATION )

) Docket Nos. 50-413-OLA (Catawba Nuclear Station, ) 50-414-OLA Units I and 2) )

)

AFFIDAVIT OF STEVEN P. NESBIT Steven P. Nesbit hereby declares under penalty of perjury that the following statements are true and correct of his own knowledge:

I. In the captioned proceeding, I have supplied information in response to Specific Interrogatory No. 1-1 and No. 2, and General Interrogatories Nos. 2 and 3, as specified in Blue Ridge Environmental Defense League's First Set of Discovery Requests Directed to Duke Energy Corporation, dated March 31, 2004.

2. For 22 years, I have been employed by Duke Energy Corporation, working in areas related to nuclear energy. I currently hold the title of Engineering Supervisor II in the Nuclear Generation Department.

DC:35304I.1

3. The information in these responses is true and correct to the best of my knowledge and belief.

Subscribed to and Sworn before me personally, on this l./Mday of April, 2004.

lt"d 611-Notary Public My Commission expires: ?4 pv7 ' "r JcnUi Z2, 2 6 o &

MICHAEL T. CASH Notary Public Lincoln County, North Carolina Commission Expires January 22, 2008 2

DC:353041.1

Attachment 1 General Interrogatorv No. 1- Respondents to BREDL Discovery Request Michael J. Barrett Duke Energy Senior Engineer General Request 2 526 S. Church Street Specific Requests II-1 through II-6 Charlotte, NC 28202 H. Duncan Brewer Duke Energy Engineering General Request 2 526 S. Church Street Supervisor II Specific Requests II-1 through 11-6 Charlotte, NC 28202 Bryan C. Carroll Duke Energy Engineer General Request 2 526 S. Church Street Specific Requests II- I through II-6 Charlotte, NC 28202 Michael T. Cash Duke Energy Consulting Engineer General Request 2 526 S. Church Street Specific Requests 11 II-6 Charlotte, NC 28202 Robert C. Harvey Duke Energy Senior Engineer General Request 2 526 S. Church Street Specific Requests II-I through II-6 Charlotte, NC 28202 Specific Interrogatory I-I Melissa S. Moran Duke Energy Associate Engineer General Request 2 526 S. Church Street Specific-Requests II-1 through II-6 Charlotte, NC 28202 1

Steven P. Nesbit Duke Energy Engineering General Interrogatories 2 and 3 526 S. Church Street Supervisor II General Request 2 Charlotte, NC 28202 Specific Interrogatories I-l and 2 Specific Requests I-1, II-1, II-4, II-6 P. Doug Paul, Jr. Duke Energy Engineer General Request 2 526 S. Church Street Specific Requests II-I through II-6 Charlotte, NC 28202 Stephen P. Schultz Duke Energy Nuclear Engineering General Request 2 526 S. Church Street Manager Specific Requests II-1 through II-6 Charlotte, NC 28202 Michael W. Scott Duke Energy Senior Engineer General Request 2 526 S. Church Street Specific Requests II-1 through II-6 Charlotte, NC 28202 Gregg B. Swindlehurst Duke Energy Nuclear Engineering General Request 2 526 S. Church Street Manager Specific Requests II- 1 through 11-6 Charlotte, NC 28202 F. Jay Verbos, Jr. Duke Energy Senior Engineer General Request 2 526 S. Church Street Specific Requests II-1 through II-6

-Charlotte, NC 28202 Robert Henry Bert M. Dunn AREVA Advisory Engineer Specific Interrogatory I-1 3315 Old Forest Road Lynchburg, VA 24501 2

Attachment 2 General Interrogatory No. 2 - Prospective Witnesses (Contentions I and II)

e. 2;al-Nam

,Addiess Pitf;io .Ej;

'e; iP e f ioii;;is !4K Steven P. Nesbit Duke Energy Nuclear Duke Energy MOX fuel 526 South Church Street Engineering Nuclear safety Charlotte, NC 28202 analysis Robert C. Duke Energy Nuclear Duke Energy Nuclear safety Harvey 526 South Church Street Engineering analysis Charlotte, NC 28202 Loss of coolant accident (LOCA) analysis Bert M. Dunn AREVA Nuclear ARE-VA Nuclear safety 3315 Old Forest Road Engineering analysis Lynchburg, VA 24501 Loss of coolant accident (LOCA) analysis H. Duncan Duke Energy Nuclear Duke Energy Nuclear safety Brewer 526 South Church Street Engineering analysis Charlotte, NC 28202 Severe accident analysis Robert Henry Fauske & Associates, Inc. Mechanical Fauske & Nuclear safety 16W 070 West 83rd St. Engineering Associates, Inc. analysis Burr Ridge, IL 60527 Severe accident analysis l

STEVEN P. NESBIT Duke Power 526 South Church Street Charlotte, NC 28202 QUALIFICATIONS:

AMr. Nesbit has 24 years nuclear engineering and management experience in the commercialsector and oi DepartmlenatofEnergy (DOE) projects. iHe is the.Airxed Oxide ("AIOX") Fuel Project Managerfor Duke Power, which is playing a key role in the DOEprogram to dispose of surplus weapons plutonium, lie has 22 years experience with Duke Power. In addition, Air. Nesbit has managed activitiesfor the Alanaging and Operating Contractorto DOE's Office of Civilian Radioactive Wfaste AManagement lie also has arpertisein nulclearsafety analysis technologj. Air. Nesbit has extensive experience interacting with the Nuclear Regulatory Comnzission and he has authored unuerous topical reports and technicalpapers.

EDUCATION/TRAINING:

ME, Nuclear Engineering, University of Virginia, 1982 BS, Nuclear Engineering, University of Virginia, 1980 Graduate course work, Environmental Science Supervisory Development Program, Duke Power PROFESSIONAL AFFILIATIONS/CERTIFICATIONS:

Registered Professional Engineer, North Carolina Registered Professional Engineer, South Carolina American Nuclear Society EXPERIENCE:

3/99-Present Engineering Supervisor II- Duke Power Manages Duke Power's activities as part of the project to dispose of surplus United States weapons plutonium using mixed oxide (MOX) fuel. Directs technical, licensing, and business activities. Serves as a public spokesperson on the MOX fuel project.

09/96-3/99 Consulting Engineer - Duke Power Led Duke Power's feasibility investigations regarding using MOX fuel at the company's three nuclear plants to support DOE's surplus weapons I

Steven P. Nesbit plutonium disposition program. Served as a representative on the Nuclear Energy Institute's Working Group on Surplus Weapons Plutonium Disposition. Interacted with external groups (Congress, DOE, and the public) in support of the MOX fuel project.

11/95-09/96 Engineering Supervisor II - Duke Engineering & Services (DE&S)

Supervised the Design Basis and Project Integration Section of the DOE Office of Civilian Radioactive Waste Management (OCRWM) Management and Operating Contractor. Developed environmental design criteria and performed design basis accident evaluations for an interim storage facility for spent nuclear fuel.

05/94-11/95 Manager, Regulatory Interactions Section - DE&S Manager of the Las Vegas Regulatory Interactions Section of the Regulatory and Licensing Department of the Management and Operating contractor for the DOE OCRWM. Responsibilities of the seven-person section included interactions with the Nuclear Regulatory Commission (NRC) staff and on-site representatives, the Advisory Committee on Nuclear Waste, and the Nuclear Waste Technical Review Board; development of regulatory positions; regulatory reviews; Site Characterization Analysis comment responses; regulatory commitments; and NRC issue resolution activities.

12/92-04/94 Engineering Consultant - DE&S Licensing Engineer in the Las Vegas Regulatory and Licensing Department of the Management and Operating contractor for the DOE OCRWM.

Provided nuclear power plant licensing experience and general support to the DOE Yucca Mountain Site Characterization Office. Assisted with interactions between the DOE, the National Academy of Sciences, the Environmental Protection Agency, and the NRC, related to the development of an environmental standard for the potential repository at Yucca Mountain.

1991-1992 Utility Engineering Group (UEG) Site Engineer- DE&S Site Engineer in Washington, D.C., for the DE&S Utility Engineering Group. Provided utility perspective and experience to the DOE for the New Production Reactor Project. Served on the staff of the Chief Engineer of the project. Provided day-to-day liaison with the various project areas. Served as Project Engineer for the UEG. Managed the DE&S Washington, D.C.,

office.

2

Steven P. Nesbit 1990-1991 Senior Engineer - Duke Engineering & Services Worked in the safety review area of the UEG. Provided utility perspective and experience to the New Product Reactor Project in the area of nuclear reactor safety.

1988-1990 Design Engineer- Duke Power Lead engineer in the area of nuclear safety analysis technology, a work group comprised of five engineers. Worked on developing mass and energy release analysis capability for high energy line breaks at Oconee, McGuire, and Catawba Nuclear Stations. Used the RELAP5/MOD002 transient analysis computer code and wrote in-house analytical codes. Worked to develop reactor building analysis capability for large dry and ice condenser containments, including applications of the FATHOMS (COBRA-NC) and CONTEMPT computer codes. Tested the upgraded Oconee training simulator and evaluated vendor perfornance. Represented the Babcock and Wilcox Owners Group (B&WOG) on the Project Management Group of the Multi-Loop Integral System Test Facility, a thermal-hydraulic research project sponsored by the B&WOG, the Electric Power Research Institute (EPRI) and the NRC. Served on the Duke Power Crisis Management Team.

1982-1988 Design Engineer/Assistant Engineer/Junior Engineer - Duke Power Lead safety analysis engineer for the Oconee Nuclear Station, a work group of up to five engineers. Served as Duke Power representative on the B&WOG Analysis Committee. Participated in the Technical Advisory Group, a committee comprised of B&WOG, EPRI and NRC representatives, which evaluated the need for thermal hydraulic testing related to once-through steam generators. Helped develop symptom-oriented emergency procedures for Oconec. Performed extensive RETRAN benchmarks of plant transients and helped prepare a safety analysis methods topical report for submission to the NRC. Served as one of 12 auditors for the inaugural Duke Power Self-initiated Technical Audit, patterned after the NRC Safety System Functional Inspections. Participated in fuel loading and start-up physics testing at McGuire Nuclear Station. Participated in zero power physics testing at Oconee. Performed system and containment analyses of the Oconee plant. Prepared technical justifications for emergency Technical Specification changes which prevented unnecessary unit shutdowns.

3

Steven P. Nesbit 1979-1982 Reactor Operator/Reactor Operator Trainee - University of Virginia Reactor Facility Reactor Operator Trainee and licensed Reactor Operator for the 2-MW research reactor in Charlottesville, Va. Duties included shift operation work, training and fuel handling.

AWNARDS/hONORS:

"Doer of Deeds," Yucca Mountain Site Characterization Office, February 2, 1994.

Newcomb/Thomton Fellowship, University of Virginia, 1980-198 1.

Bachelor of Science with Highest Distinction, University of Virginia, 1980.

PUBLICATIONS:

Nesbit, S. P., Scott, M. W., Eller, J. L., Verbos, F. J., and Costello, M. V., "Non-LOCA Safety Analysis for Operation with Weapons Grade MOX Fuel Lead Assemblies," American Nuclear Society Winter Meeting 2003, New Orleans, LA, November 2003.

Nesbit, S. P. and Eller, J. L., "Basis for the Design of Reactor Cores Containing Weapons Grade MOX Fuel," Advances in Nuclear Fuel Management III, Hilton Head, SC, October 2003.

Anderson, S. L., Gilreath, J. D., Nesbit, S. P., and Laubam, T. J, "Mixed Oxide Fuel Effects on the Integrity of the McGuire and Catawba Reactor Vessels," Fifth Topical Meeting on Spent Nuclear Fuel and Fissile Materials Management, Charleston, SCj September 18, 2002.

Buckner, M. R., Bengelsdorf, H. D., and Nesbit, S. P., "American Nuclear Society Nonproliferation Position Statement," Fifth Topical Meeting on Spent Nuclear Fuel and Fissile Materials Management, Charleston, SC, September 18, 2002.

Clark, R. H., Dziadosz, D., and Nesbit, S. P., "MOX Fuel Irradiation Program for Disposition of Surplus United States Plutonium," Fourth Topical Meeting on Department of Energy Spent Nuclear Fuel and Fissile Materials Management, San Diego, SC, June 7,2000.

Nesbit, S. P. and Bengelsdorf, H. D., "A Comparison of Surplus Weapons Plutonium Disposition Technologies," Third Topical Meeting on Department of Energy Spent Nuclear Fuel and Fissile Materials Management, Charleston, SC, September 9,1998.

S. P. Nesbit, "A Utility Perspective on Surplus Weapons Plutonium Disposition in Existing United States Light Water Reactors," Advances in Nuclear Fuel Management II, Myrtle Beach, S.C.,

March 1997.

S. P. Nesbit, S. J. Brocoum, M. A. Lugo, J. A. Duguid, P. M. Krishna, "Regulatory Perspective on NAS Recommendations for Yucca Mountain Standards," 7th Annual International High-Level Radioactive Waste Management Conference, Las Vegas, NV, May 1, 1996.

4

Steven P. Nesbit J. Carl Stepp, Silvio Pezzopane, Quazi Hossain, Michael Hardy, Steven P. Nesbit, "Criteria for Design of the Yucca Mountain Structures, Systems, and Components for Fault Displacement,"

FOCUS '95 - Methods of Seismic Hazards Evaluation, Las Vegas, NV, September 20,1995.

J. Carl Stepp, Michael P. Hardy, Quazi A. Hossain, Steven P. Nesbit, J. Timothy Sullivan, "Seismic Design Methodology for a Geologic Repository at Yucca Mountain," 6th Annual International High-Level Radioactive Waste Management Conference, Las Vegas, NV, May 4,1995.

D. Stahl, S. P. Nesbit, L. Berkowitz, "Approach to Compliance with the NRC Substantially Complete Containment Requirement at the Potential Repository at Yucca Mountain," 6th Annual International High-Level Radioactive Waste Management Conference, Las Vegas, NV, May 3, 1995.

S. P. Nesbit, S. J. Brocoum, "New Public Health and Safety Standards for Yucca Mountain and Their Impact on the Carbon-14 Issue," Waste Management '95 Conference, Tucson, AZ, February 26, 1995.

S. P. Nesbit, R. J. Gerling, and G. B. Swindlehurst, "Qualification of the Oconee RETRAN Model by Comparison with Plant Transient Data," Nuclear TechnologU, Volume 83, December 1988.

TOPICAL REPORTS:

DPC-NE-1005P, "Duke Power Nuclear Design Methodology Using CASMO-4/SIMULATE-3 MOX," Duke Energy, August 2001.

YMP/TR-003-NP, "Seismic Design Methodology for a Geologic Repository at Yucca Mountain,"

U. S. Department of Energy, October 1995.

DPC-NE-3003-P, "Mass and Energy Release and Containment Response Methodology," Duke Power Company, August 1993.

BAW-2079, "Technical Advisory Group Investigation of Once-Through Steam Generator Thermal-Hydraulic Data Requirements," Babcock and Wilcox, March 1989.

DPC-NE-3000, "Thermal-Hydraulic Transient Analysis Methodology," Duke Power Company, July 1987.

SECURITY CLEARANCE:

DOE "L" Clearance (active)

REFERENCES:

DOE and commercial references available upon request 5

ROBERT C. HARVEY Duke Power 526 South Church St Charlotte, NC 28202 QUALIFICATIONS:

25 years of Thermal Hydraulic and Safety Analysis experience supporting the reload and licensing of pressurized water reactors. Mr. Harvey has performed numerous safety analysis calculations using the RELAP4, RELAP5, RETRAN, TOODEE-2, CONTEMPT-LT, and MAAP computer codes.

EDUCATION:

Nuclear Engineering Graduate Studies, University of Lowell (1 980-1982)

BS, Nuclear Engineering, University of Lowell, 1979 Supervisory Training, Yankee Atomic Electric Company, 1994 MAAP Code Utilization and Phenomena seminar, Fauske & Associates Two-Phase Gas-Liquid Flow Seminar, University of Houston Nuclear Power Reactor Safety Seminar, Massachusetts Institute of Technology (MIT)

Simulator Training, Combustion Engineering (CE)

Two-Phase Flow and Heat Transfer, Rensselaer Polytechnic Institute EXPERIENCE:

Senior Engineer - Duke Power Company 2/99 - present Lead Engineer responsible for the LOCA analysis supporting the Oconee, McGuire, and Catawba nuclear plants. Responsibilities include providing interface and oversight of the vendor analyses. In addition, performs LOCA mass and energy release calculation used as input to the containment analysis and performs UFSAR Chapter 15 non-LOCA safety analysis. Specific accomplishments include supporting the Oconee reanalysis to support steam generator replacement and the transition to best-estimate LOCA analysis methods for the McGuire and Catawba units. Serves as a member of the Emergency Operations Facility (EOF) in the position of Accident Assessment Manager.

Provided an independent assessment of the Texas Utilities LOCA analysis supporting the transition to Westinghouse fuel.

I

Robert C. Harvey Engineer - Duke Engineering & Services - (12/97 - 1/99); Senior Nu11clear Enlgilleer -

Yankee Atomic Electric Company (5/91 - 11/97)

Lead Engineer for pressurized water reactor (PWR) LOCA analyses supporting licensing for the Yankee Rowe, Maine Yankee and Seabrook Nuclear Power Stations. Areas of involvement included LOCA and containment analyses and severe accident analyses related to Individual Plant Evaluations (IPEs). Specific accomplishments included supporting the Maine Yankee small break loss of coolant accident (SBLOCA) analysis to justify a return to 2440 MWth operation, and providing oversight of Siemens Power Corporation SBLOCA re-analysis. Served as a response team member to Maine Yankee 1996 Independent Safety Assessment.

In addition, supported General Electric (GE) in severe accident analysis for simplified boiling water reactor (SBWR) certification and provided consulting to the Siemens fuel user group in the area of LOCA analysis.

Provided support to Northeast Utilities on severe accident management guidelines (SAMGs) for Millstone Units 2 & 3 and the Seabrook Nuclear Power Station and performed a technical review of ABB/CE reload analysis of St. Lucie Unit 2 for FP&L.

SeniorEngineer - Yankee Atomic Electric Company 5/88 - 5/91 Lead Engineer for Yankee Rowe and Seabrook LOCA analysis related activities and for all severe accident analysis activities. Duties involved reload licensing analysis for the Yankee Rowe plant and vendor oversight of Seabrook LOCA analysis activities.

Supported the plant specific model development and certification of the Yankee Rowe plant simulator. Participated in the Yankee Rowe plant life extension (PLEX) effort providing support in severe accident evaluations and pressurized thermal shock (PTS) analysis. Also, provided training to Texas Utilities personnel in LOCA analysis method applications.

Nuclear Engineer- Yankee Atomic Electric Company 5/85 - 5/88 Lead Engineer for Yankee Rowe LOCA analysis related activities. Activities included large break loss of coolant accident (LBLOCA) model development and applications related to reload licensing, steam generator tube rupture (SGTR) analysis and plant request responses. Participated in the development of Yankee Rowe, plant specific emergency operating procedures (EOPs) based on the generic Westinghouse Owners Group (WOG) emergency response guidelines (ERGs).- Performed a plant specific analysis to support deviations from the generic WOG guidelines. Also, provided training to Korean Power (KEPCO) engineers in LOCA analysis methods.

2

Robert C. Harvey Engineer - Yankee Atomic Electric Company 6/79 - 5/85 Performed LBLOCA analyses in support of reloads for the Yankee Rowe and Maine Yankee plants. Contributed to model enhancements of the LBLOCA methods used for Yankee Rowe. Participated in developing and assessing the RELAPSYA computer code used for PWR SBLOCA analysis.

PROFESSIONAL AFFILLIATIONS/CERTIFICATIONS:

American Nuclear Society (ANS), Member The Research Society of Sigma Xi, Associate Member Registered Professional Engineer North Carolina (Registration # 027387)

South Carolina (Registration # 22237)

SELECTED PUBLICATIONS:

1. Maine Yankee Steam Generator Tube Sleeving Thermal-Hydraulics and Safety Analysis Impacts, co-authors K. R. Rousseau, S. Palmer, P. A. Bergeron, presented at the American Power Conference, Chicago, III., 1995.

2. Maine Yankee Cycle 15 Core Performance Analysis, YAEC-1 907, co-authors, January 1995.

3. Yankee Rowe Pressurized Thermal Shock, Thermal-Hydraulic Analysis, International Heat Transfer Conference, co-authors P. A. Bergeron, N. Fujita, August 1993.

4. Maine Yankee Level 11 PRA Results, ASME/JSME International Conference on Nuclear Engineering, co-author K. E. St. John, March 1993.

5. Thermal Hydraulics Analysis of the Yankee Plant Due to a Stuck Open PORV Using RELAP5/MOD3 Computer Code, RELAP5/TRAC-B International Users Seminar, co-authors W. S. Yeung, R. K. Sundaram, November 1991.

6. Yankee Plant Small Break LOCA Analysis, YAEC-1732, co-authors S. Mihaiu-Westerlind, R. K. Sundaram, July 1990.

7. Yankee Nuclear Power Station Core 21 Performance Analysis, YAEC-1731, co-authors, July 1990.

8. Yankee Nuclear Power Station Severe Accident Closure Submittal, YAEC-171 1, co-authors, December 1989.

3

Robert C. Harvey

9. Plant-Specific Analysis to Support the Yankee Emergency Operating Procedures, YAEC-1 663, co-authors, April 1989.

10. Seabrook Station Risk and Plant Response for Low Power Operating Conditions, YAEC-1 623, co-authors, March 1988.

11. RELAP5YA Simulation of LOFT Small Break Experiments L3-6 and L5-1, Transactions American Nuclear Society, Volume 55, co-authors, L. Schor, November 1987.

12. Estimate of Peak Clad Temperature and Its Uncertainty in a Large Break LOCA at Yankee Nuclear Power Station, YAEC-1431P, co-authors, R. K. Sundaram, K. E. St.

John, May 1984.

13. RELAP5YA - A -Computer Program for Light Water Reactor System Thermal-Hydraulic Analysis, YAEC-1300P, co-authors, R. T. Fernandez, R. K. Sundaram, J.

Ghaus, A. Husain, J. N. Loomis, L. Schor, R. Habert, October 1982.

14. RELAP4 and RELAP5 Calculation of LOFT L3-5 and L3-6 Experiments: Comparison to Data, ANS Specialists Meeting on Small Break Loss-of-Coolant Accident Analyses in LWRs, co-authors, L. Schor, J. N. Loomis, A. Husain, August 1981.

15. RELAP4 Analysis of CREARE Flashing Transients with Reverse Core Steam Flow, Transactions American Nuclear Society, Volume 38, co-authors G. J. Brown, A.

Husain, August 1981.

16. Applications of a Lower Plenum Phase Separation Model to Yankee Rowe Large Break LOCA Analysis, YAEC-1 231, Revision 1, co-authors, March 1981.

17. Maine Yankee Cycle 5 Core Performance Analysis, YAEC-1202, co-authors, December 1979.

4

BERT INI. DUNN AREVA 3315 Old Forest Road Lynchburg, VA 24501 EDUCATION: BS in Physics, Washington State University, 1968 MS in Physics, Lynchburg College, 1973 EXPERIENCE: AREVA Framatome ANP Inc. (formerly Framaitome Nuclear Technologies Inc., Nuclear Power Division of the Babcock & Wilcox Company), Lynchburg, Virginia 3/84-Present Advisory Engineer AREVA representative to high burnup industry forums responsible for providing recommendations on licensing criteria to NRC. Lead on AREVA consultation with NRC on irradiated fuel LOCA testing at Argonne National Laboratory.

Member of U.S. NRC High Burnup Fuel Design Basis Accident Technical Evaluation PIRT (Phenomena Identification and Ranking Table) Panel.

Technical lead for the development of LOCA and Safety Analysis techniques for the licensing of new fuel cladding materials.

Technical lead for the development of evaluation techniques to determine the outcome of inherent boron dilution events.

Technical co-lead for the evaluation of best estimate LOCA licensing techniques.

Project lead for development of Loss-of-Coolant-Accident analysis capability for Westinghouse and Combustion Engineering plants.

2/83-3/84 Senior Product Manager Project Manager for Pressurized Thermal Shock (PTS) B&W Owners Group Task Force. Lead development of a probabilistic risk assessment for PTS in B&W plants.

1975-1980 Unit Manager. Emergengv Core Cooling System Analysis Responsible for all ECCS evaluations of the performance of B&W-designed nuclear power plants.

I

Bert M. Dunn 1970-1975 Engineer/Supervisory Engineer Licensing of B&W's ECCS evaluation models and development of techniques for evaluating reactor building subcompartment pressure forces.

1968-1970 Engineer - Douglas United Nuclear Corporation. Richland. Washington Reactor physics, fuel engineering, operations.

PUBLICATIONS: Major contributor or the principal author of:

a. BAW-10034, "Multinode Analysis of B&W's 2568 MWt Nuclear Plants During a Loss-of-Coolant Accident," October 1971.

b. BAW-10045, "Multinode Analysis of B&W's 205-Fuel Assembly Nuclear Plants During Loss-of-Coolant Accident,"

May 1972.

c. BAW-10052, "Multinode Analysis of Small Breaks for B&W's 2568 MWt Nuclear Plants," September 1972.

d. BAW-10064, "Multinode Analysis of Core Flooding Line Break for B&W's 2568 MWt Internals Vent Valve Plants," April 1973.

e. BAW-10091, "B&W's ECCS Evaluation Model Report With Specific Application to 177-FA Class Plants With Lowered-Loop Arrangement," August 1974.

f. BAW-10091, Supplement 1, "Supplementary and Supporting Documentation for B&W's ECCS Evaluation Model Report With Specific Application to 177-FA Class Plants With Lowered-Loop Arrangement," December 1974.

g. BAW-10102, "ECCS Evaluation of B&W's 205-FA NSS," June 1975.

h. BAW-10104, "B&W's ECCS Evaluation Model," May 1975.

i. BAW-10106, "- QUENCH - Digital Program for Analysis of Core Thermal Transients During Loss-of-Coolant Accident,"

May 1975.

2

Bert M. Dunn

j. Presentation for ANS Annual Meeting, Las Vegas, Nevada, June 8-12, 1980, "RC Pump Trip and Small Break LOCA," invited paper.

k. BAW-10227, "Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel," September 1997.

3

HARRY DUNCAN BREWVER Duke Power 526 South Church Street Charlotte, NC 28202 QUALIFICATIONS:

Mr. Brewer's nuclearpower experience includes all aspects ofprobabilisticrisk assessment, including basemodel development, project inanagem eut and applicationsofProbabilistic Safet Analysis(PRA) to resolve regulatory and operationalproblenms. lie has 25 years experience withs Duke Power. Air. Brewer is currently the Section Alazalger of the Severe Accident Analysis Section, the group which provides PSA analyses and applicationssupport for Duke's three nuclearstations.

EDUCATION/TRAINING:

BS, Nuclear Engineering, North Carolina State University, 1979 ME, Mechanical Engineering, University of South Carolina, 1983 Attended many continuing education courses, including leadership and professional development classes, decision analysis courses, and Advanced PRA Training.

PROFESSIONAL AFFILIATIONS/CERTIFICATIONS:

Registered Professional Engineer, NC #11574, SC #10040 Member of American Society of Mechanical Engineers EXPERIENCE:

Title of Position: Section AManager Department/Company:Nuclear GenerationDepartmentIDukePowser Company From: 6/99 To: Present Section Manager (first line supervisor) for nuclear plant Probabilistic Risk Assessment group which provides PRA models and PSA applications support for Duke Power's three nuclear power stations.

Title ofPosition: Senior Engineer Department/Company':Nuclear GenerationDepartmeunt/DukePower Company From: 6/90 To: 6/99 Lead engineer for nuclear plant Probabilistic Risk Assessment group which provides PSA applications support and develops system and accident models. Team Leader for Catawba Probabilistic Risk Analysis project and IPE Submittal. Member of working group that developed the NUMARC Guidelines for Implementing the Maintenance Rule. Prepared and taught sessions of the NUMARC Maintenance Rule Workshpps. Lead PRA analyst supporting Maintenance Rule implementation at Duke nuclear stations.

I

H. Duncan Brewer Title ofPosition: Design Engineer Department/Company:Design EngineeringDepartmenzt/DiikePowver Companyj From: 6/84 To: 6/90 Lead engineer for group responsible for severe accident consequence analysis for McGuire, Catawvba and Oconee Nuclear Stations. Performed extensive thermal/hydraulic computer analyses using the computer code MAAP to determine primary system and containment response to degraded core accidents. Also, have used CRAC2 and CRACIT to perform severe accident consequence modeling. Provided technical support to Industry Degraded Core Rulemaking Effort (IDCOR).

Title ofPosition: EngineerAssociate Department/Compan.y:Design EngineeringDepartmentIDuke Pow'er Company From: 6/81 To: 6/84 Responsible for integrated safety analysis for McGuire, Catawba and Oconee Nuclear Stations.

Coordinated review of nuclear power plant safety systems with Electrical, Mechanical and Civil engineering personnel to ensure an integrated, complete and safe design.

Title ofPosition: AssistantEngineer Department/Conmpany:Design EngineeringDepartmentIlDukePower Company From: 7/79 To: 6/81 Responsible for effluent and radiation shielding analysis for McGuire, Catawba and Oconee Nuclear Stations. Designed radiation shielding for components and systems for nuclear plant systems. Performed radioactive effluent release analyses to ensure compliance with I OCFR100 and I OCFR50, Appendix I requirements.

Title ofPosition: EngineeringCo-Op Student Department/Company:Design EngineeringDepartment/Duke Power Company From: 8/75 To: 8/78 Assisted a group of engineers in performing effluent release and radiation dose analysis for McGuire, Catawba and Oconee Nuclear Stations.

2

H. Duncan Brewer Industry Support Activities Sponsor Description Responsibilitics Dates AIF Sub-committee on Emergency Preparedness Member 1984 and Siting_.

AIF Source Term Task Force Chairman 1987-1988 EPRI MAAP Users Group Duke 1987-1989 Representative/

Chairman 1990 EPRI Safety Technology Task Force Duke Representative 1989-1992 EPRI Severe Accident Sub-Committee Chairman / Duke 1989-1992 l_ Representative NUMARC Accident Management Ad Hoc Advisory Member 1989 Committee NUMARC Maintenance Rule Working Group PRA Representative 1992-1996 EPRI Risk and Reliability Workstation Users Group Chairman / Duke 12/93 -

Representative 12/95 B&WOG Risk Informed Applications Committee Duke Representative 1995-Present WOG Risk Based Technology Working Group Past Chairman / 12/95 - 3/00 Duke Representative NEI Risk Applications Task Force WOG 3/96 -

Representative Present WANO International Technical Exchange Peer Peer Reviewer 1/6/97-1/10/97 INPO Events Database Working Group Member 4/97-12/97 INPO/MIT Risk Informed Operational Decision Guest Instructor 1/97-Management Course 7/2001 ASME BNCS Task Group on Risk Management for Project Team 1/98 -

Nuclear Facility Application (Project Team) Member 4/2002 ANS *PSA'99 Technical Program Committee Committee Member 1998-1999 ASME Committee on Nuclear Risk Analysis Committee Member 1/98-Present ASME Sub-committee on PRA Technology Chairman 3/2002-Present ASME Task Group on PRA Generic Database Chairman 9/2002-Present ANS PSA '02 Technical Program Committee Committee Member 2001-2002 EPRI Risk/Safety Management Committee Member 2004 -

._ Present ANS Utility Working Conference Organizing Track Organizer 2004 Committee II 3

H. Duncan Brewer Papers and Presentations Sponsor Description Format Date EPRI Conference "MAAP-GRAAPH: A Nuclear Power Presentation and June14,1988 of Power Plant Plant Severe Accident Simulation and Published Paper Simulators and Training Tool" Modeling NRC "Design Consideration for Severe Co-Author, June 15, Containment Accident Containment Performance" Published Paper 1988 Integrity Workshop IAEA/ANL "Use of PSA Optimization of Safety in Classroom 1/21/93 Interregional Operation of Nuclear Power Plants", Lecture Training Course Module - "Use of PSA in Operator Training" ANS Annual "The Role of PRA in Implementing the Presentation and 6/21/93 Meeting Maintenance Rule" Published Paper ANS Annual "Influence of Refueling Water Storage Co-Author, 6/22/93 Meeting Tank Draining on Performance of the Presentation and Catawba Nuclear Station Ice Condenser Published Paper

._ Containment" NUMARC -"Establishing Risk Significant Criteria" Presentation / 8/5/93 and Industry and "Risk Significant Systems Breakout Training 8/23/93 Workshop on Session" Implementing the NRC Maintenance Rule ANS Utility "Operating Cycle Risk Management" Panel Member 8/14/95 Working Panel Discussion Conference EPRI ORSERG "Applications of the IPE to ISEG Presentation 9/28/95 Objectives" ACRS PSA Sub- "PSA Support for Maintenance Rule Presentation 10/26/95 Committee Implementation" ACRS PSA Sub- "PSA Model Methods and Results Presentation 7/18/96 Committee Comparison Program" ANS Utility "Risk Informed Regulation" Panel Discussion 8/5/96 Working Panel Discussion Organizer and Conference Leader Risk Informed "Application of PSA in Maintenance Rule Guest Instructor 1997-2001 Operational Implementation", and "Standard for Decision Probabilistic Risk Assessment for Nuclear Management Power Plant Applications" Course Sponsored by INPO and MIT 4

H. Duncan Brewer Reliability "Probabilistic Safety Assessment Support Scientific 1998 Engineering and for the Maintenance Rule at Duke Power Journal System Safety Company" PSA '99 "On-line Maintenance Issues and Panel Discussion 8/1999 Insights" Organizer and Panel Discussion Leader ANS Utility "Maintenance Rule" Panel Discussion 8/2000 Working Panel Discussion Organizer and Conference Leader ANS Utility "License Renewal Requirements to Panel Discussion 8/2001 Working Perform Severe Accident Mitigation Organizer and Conference Alternatives Analysis" Leader Panel Discussion . -

ANS Utility "The Risk-Informed Environment: Do We Panel Discussion 8/2002 Working Think the Language" Organizer and Conference Panel Discussion Leader PSA '02 "Nuclear Plant Safety Risk Management: Co-Author, 10/2002 A Case Study" Published Paper and Presentation PSA '02 "PSA Standards" Panel Member 10/2002 Panel Discussion PSA '02 "Present and Future PSA Challenges" Panel Member 10/2002 Panel Discussion ANS Utility "Progress Toward a Risk-informed Panel Discussion 8/3/2003 Working Environment" Organizer and Conference Panel Discussion Leader 5

ROBERT E. HENRY EDUCATION:

University of Notre Dame B.S. Mechanical Engineering, 1962 University of Notre Dame M.S. Mechanical Engineering, 1964 University of Notre Dame Ph.D. Mechanical Engineering, 1967 EXPERIENCE:

Fauske & Associates, Inc., (1980 to present), Burr Ridge, Illinois.

Senior Vice President and co-founder of FAI in 1980, Dr. Henry was involved in the Zion and Indian Point Probabilistic Safety Studies as well as the Limerick Probabilistic Risk Assessment. More recently, Dr. Henry made numerous contributions to the Industry Degraded Core Rulemaking Program in which FAI was responsible for initiating the development of the Modular Accident Analysis Program (MAAP) codes to analyze the response of pressurized and boiling light water reactors during severe accidents. The MAAP code has gained widespread acceptance in the domestic and foreign nuclear industry. This computer code specifically treats such phenomena as concrete attack, hydrogen formation, distribution and combustion within the containment, debris distribution, debris bed coolability, and fission product behavior. Dr. Henry was also the Manager for the Issue Resolution and Individual Plant Evaluation Tasks in the IDCOR Phase II Program which culminated with the Individual Plant Evaluation Methodology (IPEM). During the IDCOR Program, he was one of the industry representatives in the U.S. delegation to IAEA/Vienna to evaluate the Russian interpretation of the Chernobyl Unit 4 accident. Following Chernobyl, Dr. Henry served on review panels for both the N-Reactor and the Savannah reactors with respect to potential severe accident conditions.

Dr. Henry developed the Technical Basis Report for the Severe Accident Management Guidelines (SAMGs) that arc being developed by the four owner's groups in the United States. As part of this, Version 4 of the MAAP code (MAAP4) was developed to provide analytical support for SAMG evaluations.

He has also served on several NRC ad hoc committees for resolution of issues such as scaling, steam explosions, core melt relocation and direct containment heating. He is the author of approximately 90 journal articles as well as numerous other reports and abstracts.

Recently Dr. Henry has promoted dynamic benchmarking of computer codes as a means of preserving insightful experimental data in a form that will remain abreast of the changing computer technology. Through such dynamic benchmarking with the large number of existing experiments, computer codes can be compared on a common basis and thoroughly qualified for their uses in supporting engineering analyses.

1

Robert E. Henry Dr. Henry has also initiated development of the MAAP5 code. This is focused on providing realistic, yet conservative, evaluations of the core, Reactor Coolant System and containment responses under postulated accident conditions.

Argonne National Laboratory, (1969 to 1980), Argonne, Illinois.

Dr. Henry held a number of responsible research and development positions at ANL. In March, 1979, he was appointed Associate Director of the Reactor Analysis and Safety Division, and was involved in the evaluation of the Three Mile Island-2 accident as part of the group formed by the Electric Power Research Institute's Nuclear Safety Analysis Center (NSAC).

OTHER EXPERIENCE:

Member: American Nuclear Society Chairman: Mechanical Engineering Department Midwest College of Engineering (1973-1978)

Dean: Graduate School Midwest College of Engineering (1978-1979)

Non-Resident Research Associate: Massachusetts Institute of Technology TECHNICAL AWARDS:

"Award for Outstanding Engineering Accomplishment", College of Engineering, University of Notre Dame (1990).

"Tommy" Thompson Award: The highest honor the American Nuclear Society gives in the field of reactor safety (1985).

OPEN LITERATURE PUBLICATIONS IN THlE LAST 10 YEARS:

"Response of the Davis-Besse Backup Boron Mixing System," Paper presented at the 14th Pacific Basin Nuclear Conference, Honolulu, Hawaii, March 23-25, 2004.

"CVTR Revisited," Paper presented at the 14th Pacific Basin Nuclear Conference, Honolulu, Hawaii, March 23-25, 2004.

"Benchmarking of LOFT ATWS Experiments With MAAP4 Using a Point Kinetics Model,"

Paper presented at the 14th Pacific Basin Nuclear Conference, Honolulu, Hawaii, March 23-25, 2004.

"Waterhammer Experiments in Support of a Feedwater Leakage Control System," Paper presented at the 16th International Conference on Structural Mechanics in Reactor Technology (SMIRT-1 6), Washington, DC, August, 2001.

2

Robert E. Henry "The Role of Expanded Metal Network in Preventing BLEVES," Paper presented at the AIChE 2001 Spring National Meeting, Fifth Bi-Annual Process Plant Safety Symposium, April 22-26.

"Best Estimate Containment Assessment," Paper presented at the International Meeting on "Best-Estimate" Methods in Nuclear Installation Safety Analysis (BE-2000),

Washington, DC, November, 2000.

"Dynamic Benchmarking of TREMOLO - A Program for Pipeline Two-Phase Flow Transient Analysis," Paper presented at the Ninth International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-9), San Francisco, California, October 3-8, 1999.

"Data Preservation and Dynamic Benchmarking," Paper presented at the ANS 1998 Winter Meeting and the Meeting of the Americas, Washington, DC, November 15-19, 1998.

"Waterhammer in Horizontal Lines During the Voiding Phase," Paper presented at the ANS 1998 Winter Meeting and the Meeting of the Americas, Washington, DC, November 15-19, 1998.

"Two-Phase Critical Flow Through Control Valves," Paper presented at the ANS 1998 Winter Meeting and the Meeting of the Americas, Washington, DC, November 15-19, 1998.

"An Assessment of a Fuel Channel Rupture Within a Calandria Tank in Candu Reactors," Paper presented at the 5th International Conference on Nuclear Engineering, ICONE5, May 26-30, 1997, Nice, France.

"Is Fuel Fragmentation Needed to Understand the KROTOS Experiments?," Paper presented at the 5th International Conference on Nuclear Engineering, ICONE5, May 26-30, 1997, Nice, France.

"Dynamic Benchmarking of Simulation Codes," Paper presented at the Canadian Nuclear Society Fifth International Conference on Simulation Methods in Nuclear Engineering, Montreal, Canada, September 8-11, 1996.

"Quenching of Metal Surfaces in a Narrow Annular Gap," Paper presented at the Canadian Nuclear Society Fifth International Conference on Simulation Methods in Nuclear Engineering, Montreal, Canada, September 8-11, 1996.

"Application of Uncertainty Analyses with the MAAP4 Code," Paper presented at the Canadian Nuclear Society Fifth International Conference on Simulation Methods in Nuclear Engineering, Montreal, Canada, September 8-11, 1996.

"Water Level Instrumentation Simulation," Paper presented at the Canadian Nuclear Society Fifth International Conference on Simulation Methods in Nuclear Engineering, Montreal, Canada, September 8-11, 1996.

3

Robert E. Henry "MAAP4 Uncertainty and Sensitivity Analyses," Paper presented at the 1996 ANS Annual Meeting, Reno, Nevada, June 16-20, 1996.

"MAAP4 Dynamic Benchmarking," Paper presented at the 1996 ANS Annual Meeting, Reno, Nevada, June 16-20, 1996.

"A Different Approach to Fragmentation in Steam Explosions," Paper presented at the Fourth International Conference on Nuclear Engineering (ICONE-4), New Orleans, Louisiana, March 10-14, 1996.

"How to Use Expert Judgement to Assess Uncertainties," Paper presented at PSA'95, Seoul, Korea, November, 1995.

"Formulation and Use of Uncertainty and Sensitivity Analyses," Paper presented at PSA'95, Seoul, Korea, November, 1995.

"MAAP4.0 Benchmarking with the TMI-2 Experience," Paper presented at PSA'95, Seoul, Korea, November, 1995.

"MARS Application and Benchmarking," Paper presented at PSA'95, Seoul, Korea, November, 1995.

"Overview: Uncertainties Remaining in Severe Accident Phenomenology," Paper presented at the OECD Specialist Meeting on Severe Accident Management Implementation, Niantic, Connecticut, June 12-14, 1995.

"Experiments on the Lower Plenum Response During a Severe Accident," Paper presented at ANS Topical Meeting on Thermal-Hydraulics, Taipei, Taiwan, 1994.

"Water in the RPV: A Mechanism for Cooling Debris in the RPV Lower Head," Paper presented at the OECD Specialist Meeting on Selected Severe Accident Management Strategies, Stockholm, 1994 and also submitted to Nuclear Engineering and Design for publication.

"Cooling of Core Debris Within a Reactor Vessel Lower Head with Integral Support Skirt," 9th Proc. of Nuclear Thermal-Hydraulics, 1993 ANS Winter Meeting, pp.92-100.

"External Cooling of a Reactor Vessel Under Severe Accident Conditions," Nuclear Engineering and Design, 1993, Vol. 139, pp. 31-43.

"Externally Triggered Steam Explosion Experiments: Amplification or Propagation?," Paper presented at the CSNI-FCI Specialist Meeting, Santa Barbara, CA, 1993.

4

Robert E. Henry "Integrated Modeling of Debris Bed Heat and Mass Transfer in the PWR Lower Plenum for Severe Accident Analysis," 9th Proc. of Nuclear Thermal-Hydraulics, 1993 ANS Winter Meeting, pp. 38-49.

81 Publications Prior to the Above IDCOR Reports "MAAP User's Manual," IDCOR Technical Tasks 16.2 and 16.3, (January, 1985).

"MAAP Uncertainty Analyses," IDCOR Technical Task 23.4, (March, 1985).

"Evaluations of Containment Bypass and Failure to Isolate Sequences for the IDCOR Reference Plants," IDCOR Technical Task 23.5, (April, 1985).

"Sequoyah Nuclear Power Plant, Integrated Containment Analysis," IDCOR Technical Task 23.1, (1985).

"Zion Nuclear Generating Station, Integrated Containment Analysis," IDCOR Technical Task 23.1, (1985).

"Peach Bottom Atomic Power Station, Integrated Containment Analysis," IDCOR Technical Task 23.1, (1985).

"Grand Gulf Nuclear Station, Integrated Containment Analysis," IDCOR Technical Task 23.1, (1985).

"Technical Support for Issue Resolution," IDCOR Technical Task 85.2, (1985).

"The Attainment of a Safe Stable State for a Damaged Core," IDCOR Technical Task 22.1, (January, 1984).

"FAI Aerosol Correlation," IDCOR Technical Task 11.6, (August, 1984).

"Final Report on In-Vessel Core Melt Progression Phenomena," IDCOR Technical Task 15.1, (January, 1983).

"Final Report on Debris Coolability, Vessel Penetration and Debris Dispersal," IDCOR Technical Task 15.2, (January, 1983).

"Final Report on Core-Concrete Interactions," IDCOR Technical Task 15.3, (March, 1983).

"Final Report of Key Phenomenological Models for Assessing Explosive Steam Generation Rates," IDCOR Technical Task 14.1, (May, 1983).

5

Robert E. Henry "Final Report of Key Phenomenological Models for Assessing Non-Explosive Steam Generation Rates," IDCOR Technical Task 14.1, (June, 1983).

"Hydrogen Generation During Severe Core Damage Sequences," IDCOR Technical Task 12.1, (June, 1982).

6

Attachment 3 Index to Documents Produced in Response To BREDL Request Dotef ,Descrition (letter, . "a'pe ,aofp.ns ve Isotopics Summary File for FHA BOC Fuel Excel 000001 000036 I-l Assembly Activities Radionuclide Source Terms for an Environmental Critique Article 000037 000053 11-4 of Reactor-Based Plutonium Disposition draft NRC RAI Word 9/16/2003 000054 000058 I-l Response Documents F-ANP input in LAR Word LOCA section Documents 5/1/2002 000059 I-1000068 source team panel E-mail 3/20/2002 000069 000073 11-6 meeting Hardcopy E-mail 2/20/2002 000074 000078 11-6 Hardcopy MOX Fuel Project PowerPoint 12/13/2001 000079 000119 I-1 Overview Handout Overview of NRC Word Fission Product Documents 9/23/2002 000120 000132 II-5 Research MOX LEU Differences Articles noted on 000133 000197 1-1 doWument Draft RAI Response Documents 000198 000211 I-l Framatome Letter Letter 5/30/2003 000212 000213 1-1 Draft RAI Response Word 000214 000235 1-1 Documents__ _ _ _ _ _ _ _ _ _

Draft RAI Response Word 000236 000266 I-1 Word Draft RAI Response l Do e 7/25/2003 000267 000296 I-1 Documents _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

Draft NRC Submittal Word 3/16/2003 000297 000303 1-1 Documents Supplement to the US DOE Surplus Plutonium document Apr-99 000304 000343 II-6 X Disposition .

1

-. wv ;,rff-;,--.-D. ;]**s-l'Dcumenteginnini", -~> df;--nsugEb1 ~_ ho#Zs v s ueiitDeseipxtidi' Pl.Qtt'eiyA;a of',' Btes, :- Bates"1 e R j2 ponsivii

- By~~7 Ad -ike; ,,v t.) -. ;'

-~-=F-=:-,

.r>o-t&.)Wal, -~- .- pNumbcr--cNumber -.

.*,- - .I- .'

draft NRC RAI Word 7/25/2003 000344 000349 1-1 Response Documents Engineering Calculation Form i/72004 000350 000399 1I-4 SBO Uncertainty E-mail and S taty document 10/10/2002 000400 000409 11-1 X stapled to e-mal E-mail 6/27/2002 000410 000411 11-6 Hardcopy E-mail Hardcopy and 2/22/2002 000412 000421 11-6 Hand written notes disposition of comments PowerPoint 6/18/2002 000422 000451 11-6 Handout comments on draft Word alternate source term Word 27-Mar-02 000452 000456 11-6 report comments on accident Excel 000457 000460 11-6 source terms draft panel report E-mail 3/15/2002 000461 000546 11-6 Hardcopy Phone Conversation Nv/ Hand written 10/24/2002 000547 000547 11-6 Schaperow Notes MOX Fuel and Cladding Temperatures During LOCA Compared to Chart 000548 000548 I-i Severe Accident Fuel Melt Temperatures MOX Monthly Word Jan-00 000549 000559 1-1 Management Report Documents MOX 4/9 Telecon Hand written 000560 000560 1-1 Notes I MOX Quarterly Review PowerPoint 12/17/2002 000561 000563 1-1 Mtg Handout .

to John R. Biller Letter 8/20/2004 000564 000568 I-l Questions 14, 15 & 16 Word 000569 000572 I-l Whord Questions 14, 15 & 16 Do e 000573 000576 1-1 Documents Questions 14, 15 & 16 Doduments 000577 000580 1-1 2

D '-ReBatsnsj eg Doceument^Description<,:^<iR Typeetr,-

'--Doc~~~~umet'~3,;einn~aEdiung>e'if 9 Act a>t~ ;< arm

.; .Bre>J~~ts^ < T6 ;Pro .; e

. 'eaic,cltC erGumNi er .

Questions 14, 15 Question & 16 47Wordet Word 000581 008 000584 008- 1-1 Question 47 Word 05 058 -

Documents Questions 14, 15 & 16 Doumnt 000590 000594 I-1 Attachment I Partial Response to NRC Staff Word 7/25/2003 000595 000595 1-1 Request for Additional Documents Information to Jason Schaperow Letter 3/13/2001 000639 000642 1I-6 Technical Paper: EdF Position on Radiological Word 1999 000643 000648 11-6 Consequences of MOX Documents Fuel Accidents NRC Confirmatory PowerPoint 10/4/2001 000649 000658 11-6 Research Handout Potential Research Word 2/20/2002 000659 000661 II-6 Needs Documents draft notes on PRA work Word 11/17/2003 000662 000665 1I-6 Documents . -

draft to Jason Letter 3/31/2001 000666 000669 1I-6 Schaperow Comments on Draft of Word NRC Request for Documents 9/30/2003 000670 000682 1-1 Additional Information Isotopics Summary File for FHA BOC Fuel Excel 000683 000718 I-I Assembly Activities Isotopics Summary File for FHA BOC Fuel Excel 000719 000754 1-1 Assembly Activities Isotopics Summary File for FHA BOC Fuel Excel 000755 000790 I-1 Assembly Activities Letter from WR Letter 3/16/2004 000791 000796 1-1 McCollum to NRC CCI16 Core Design PowerPoint 3/25/2004 000797 000807 I1- X

_ _ _ _ __ _ _ _ _ _ _ _ H an d o u t _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

3

guff ing )t - ,- ,...

V'pre

*...,.. ~ocument er f -- t "I.I>~',m-,-,.:"~.__'__._,_i.,w,;__;___c-etM -, Number;- Number<.

Draft Attachment 3 Word Description and Documents 000808 000862 I-1 Technical Justification Draft Attachment 3 Word 0 Description and 000863 000916 1-1 Technical Justification Documents Attachment 3 d Description and uort 000917 000953 I-I Technical Justification Documents Analysis Input Summary Word Jan-03 000954 000973 I-1 Documents.

to U.S. NRC Letter 2/27/2003 000974 000990 1-1 Outline for Landry Word 000991 000995 I-l Telephone Call Documents To Steve with Charts Letter 000996 000999 I-l Comments on Section Word 001000 001002 I-1 6.1.1 Documents 0 Question 24 Word 001003 001015 1-1 Draft ANP 02 Section Word 001016 001025 I-l 6.1.3 Documents Questions 14, 15 & 16 Documents 001026 001028 1-1 Word 001029 001045 1-1 Documents Questions 14,15 & 16 Word 001046 001050 11 Documents____

Questions 14, 15 & 16 Word 001051 001055 I-l Documents 0 draft NRC RAI Word 10/15/2003 001056 001058 I-l Response Documents draft NRC RAI Word 10/15/2003 001059 001066 1-1 Response Documents 1 0 draft NRC RAI Word 10/16/2003 001067 001072 I-1 Response Documents draft NRC RAI Word 10/30/2003 001073 001083 1 Response Documents draft NRC RAI Word//2003 I-001084 001094 Response Documents draft NRC RAI Word 9/22/2003 001095 001099 1-1 4

.Docuiment DocumentDes'cr-iptioin Decrpio1'--ype ,

(Ietter,js .,F;

.{_;Stt.ae

';-'-. eginninge-Eliidieeg

t'Bates r Propetr

~*t; e-a>__. &6..B!f-itrs

'u- pietar-Iy Response Documents draft NRC RAI Word 9/22/2003 001100 001104 I I Response Documents draft NRC RAI Word 9/19/2003 001105 001109 1-1 Response Documents draft NRC RAI Word 9/19/2003 001110 001110 I-l Response Documents draft NRC RAI Word 9/15/2003 001111 001115 1-1 Response Documents Mixed Oxide Fuel Word Project Probabilistic Dord 11/17/2003 001116 001119 11-1 X Risk Assessment Work Documents 5.0 Environmental NUREG-1437, Impacts of Postulated Supplement 9 May-02 001120 001150 I-I Accidents Splmn Document of Numbers Spreadsheet 9/27/2000 001151 001168 II-I X Severe Accident Release Word 1/15/2004 001169 001170 11-2 Fractions Documents Assessment of Differences Between Word 1-1 MOX Fuel and Dord Mar-00 001171 001200 Contention Conventional Uranium D n Fuel Rough Calculations by Hand witten0 Verbos & Nesbit related Notes 001201 001208 Contention to LOCA dose analysis II TypedI-Ginna MOX SER pages document 001209 001211 Contention ocumentI I Comments on Draft of Word NRC Request for 9/1/2003 001212 001215 1-1 Additional Information Documents From W.F. Jones, MOX Fuel Contractor --

LOCA Section for MOX Letter 9/5/2002 001216 001239 1-1 LAR with Proprietary Info removed Notes from initial Duke- Hand written 10/30/2000 001240 001256 I-1 F-ANP LOCA meeting Notes 5

eginnin-, n ng Responsive Dbocument" ei escipn:typoh.tr jj2

-, tl 1 Bates'!J?

ietPdPr1 t~~oNutm it_ urnhae'r~g,. 4ts.<A art egss: s } i * ~--. cact;.-

-s---*. '.- Nui j*A. Number ^ --;;-a;;

w.

Skip Copp Notes from Hand written Duke-FNP Meeting fr Notes on Word 10/31/2000 001257 001269 1 Duke-F-ANP Meeting Document draft NRC RAI Word 9/29/2003 001270 001276 I-l Response Documents 001270 001276 draft NRC RAI Word Response Document ResponseDocuments 10/10/2003 001277 001283 I-l draft NRC RAI Word 10/14/2003 001284 001289 Response Documents Figure Q 11-3 Fast Hand written Neutron Fluence Notes on Word 001290 001304 I-l Comparison Document Attachment 1 Partial Response to NRC Staff Word 001305 001346 I-l Request for Additional Documents information Word Mox Quarterly Meeting Documents 8/3/2000 001347 001349 I-I MOX Quarterly Review PowerPoint 12/17/2002 001350 001352 I-I Mtg Handout Reactor Systems Draft Word 001353 001406 I-l Documents 001353 001406____

ReactorReatorSytem Systems Drft Draf Word Documents 001407 001460 I-I Draft to US N.R.C. with attachments Letter Sep-00 001461 001594 1-1 Attachment 6 Request for Exemptions from Word Certain Provisions of 10 Documents 001595 001605 1-1 CFR Part 50 Draft license Word 5/6/2002 001606 001775 I-I amendment submittal Documents 560216 075_

Word Reactor Systems Draft Documents 001776 001829 I-l Certification of Engineering Calculation Form 2/22/2002 001830 002176 II-6

-- MOX Isotopic Inventory Calculation Cr2b MAAP Runs for Notebook 002177 002898 II-4 X Quantification _

6

,A ,.- . ,\. .:' :

.Do'cu'ment"Deiripti6n' Calculation Summary Sheet Attachment I Partial Response to NRC Staff Request for Additional information Responses to Harvey 7

Docunient Description Ap Dt egn,, g ResBe

-aeiteo I-I e-mail heard 0/24/2003 002998 002998 Contentionn Contention e-mail hard 1/9/2002 002999 002999 I e-mail hard I-1 1/92/2002 003000 00300 Contention I e-mail hard 1/9/2002 003002 003003 c Contention I e-mail hard 1/9/2002 003004 003005 I-1 copy Contention I e-mail hard 1/16/2002 003006 003006 I-I copy cop__Conte Contention ntion I e-mail hard 2/25/2002 003007 003007 I-1 e-mail hard 3/6/2002 003008 003009 I-1 copy .Contention I e-mail hard 3/8/2002 003010 003010 I-1 copy coyContention Contention I e-mail hard 4/2/2002 003011 003011 I-l e-mail hard 4/3/2002 003012 003012 I-I copy Contention I MOX Fuel and Cladding Temperatures During LOCA Compared to Chart 003013 03013 I-1 Severe Accident Fuel Melt Temperatures e-mail hard 4/9/2002 003014 003014 co__20_ntContention I e-mail hard 10/16/2003 003015 003018 I-I copy License Amendment Letter 2/27/2003 003019 003156 Contention I Request and II Westinghouse Fuel and LOCA Evaluation of article Dec-79 003157 003166 I-l R.E. Ginna Contention I 8

Draf RI Response ocumen Wordt

eg n 003167~-

E 0 7 1-pons iDocuneDt Deseriptin 'pee(ntton I Draft RAI Response Word 003167 003207 Documents Contention I Draft RAI Response Word 003208 003270 I-l Documents Contention I Draft RAI Response Word 003271 003314 1 Documents _____________Contention I MOX Lead Assembly Word 003315 003315 I-I Licensing Premise Documents Contention I Word 003316 003318 I-1 Documents Contention I Word I-l Documents 003319 003349 Contention DocuentsII Question 25 Word 003350 003356 I-l Documents Contention I Question 35 Word 003357 003358 I-l Documents Contention I Question 47 Word 003359 003363 I-l Documents Contention I Response to Request for Word I-I Additional Information Documents 003364 003484 Contention I

_____and II Response to NRC Staff Word I-I Request for Additional Documents 003485 003599 Contention I Information and II Attachment 3 Description and Word 003600 003609 I- I Technical Justification Attachment 6 Request Word 003610 003620 I-for Exemption Documents 003610 Contention I Attachment 3 Word I-1 Description and Dord 003621 003707 Contention I Technical Justification Documents and II, II-5 Attachment 3 1-1 Description and Documents 003708 003794 Contention I Technical Justification and II, 11-5 Draft Responses to Word I-l Request for Additional Documents 003795 003930 Contention I Information Documents and II, 11-5 C IC 16 Core Design PowerPoint 3/25/2004 003931 003965 C eo with MOX LTA Handout I___________ ___ Contention 11_____

9

Oc -Er

';J~~ b 'a Te2 i, Documrnnt Desription. ype(Ietter ;.

'P 9

-- - - atesP'

___ ___ ____ _ _ . . - ___________ -. Nhiiber; Num ber -,-: i __-_'_-

MOX Fuel Lead Assembly Safety and PowerPoint 4/21/2004 003966 004007 Contention I Environmental Handout and II Evaluations andII Word 1-1 Draft RAI Response Documents 3/16/2003 004008 004014 Contention Word I-I Draft MOX LAR Documents 004015 004103 Contention I

____and ____ II, 11-5 Severe Accident Notebook 004104 004396 11-2 X Analysis Report SAAG File NO. 726 Report Binder 004397 004429 II-I X Documents of Numbers Excel 004430 004580 II-1, I-5 X

-- CNS MOX Analysis Barrett Misc. Files and SNS PRA Rev. 2b Risk Disc 004581 004581 11-1 X Spreadsheet Word Rem Percentiles Documents 004582 004586 II-1 X To Steve Nesbit Letter 1/31/2003 004587 004600 1-1 X Calculation Summary Form 7/30/2003 004601 004629 I-1 X Sheet _ _ _ _ _ _ _ _ _ _ _ _ _ _

Draft Analysis Input Word Sep-02004630 004654 I-1 X Summary Documents Draft Duke Responses Word to NRC RAIs for Documents 1/4/2002 004655 004682 II-1 X CatawbaSAMA__ _ _ __ _ _ _ _ _____

From HB Barron Attachments 11/3/2003 004683 004803 Contentions X Attachments ~~~and II _______

Summary of North Anna Bullets, notes 004804 004820 I-l X Licensing Needs MOX Quarterly Review PowerPoint 12/17/2002 004821 004823 I-1 X Mtg Handout 004823 _ __ _

Partial Response to Word NRC Staff Request for Documents 004824 004867 1-1 X Additional Information Document Summary table for Excel 004868 004872 II-1 X various SAMA 10

A;:~'~ocu et ; ~ K *eginninigl n. ig

.Document Description ,4TypeQ(IettersI. *S' OWfl2.tC <t.*'2oNihf^.i Baer --. :-tcBaTes;_ 3:. rp e ax ~~~- . ;;,asI;<ct,--'alc

- ) Ad ____i___. . Number-ubr ,{*^az.

comparisons e-mail 9/7/2001 004873 004873 II-I X hardcopy Effect of High Bum-up and MOX Fuel on PowerPoint Jun-03 004874 004899 II-6 X Fission Product Release Handout from VERCORS Tests Word 004900 004907 Contention I X Summary of Word Modification Sensitivity Documents 004908 004911 11-I X Study Results Documents Summary of Word 0 0 Modification Sensitivity Documents004912 004914 - x Study Results Fission Product Release Article 004915 004930 II-6 and Fuel Behavior _

Values for sob Contribution to Core Table 004931 004943 II-I X Damage Frequency Engineering Information Record: LOCA Section Form 004944 004966 I-l . X for MOX LAR to NRC with to Letter NR 12/10/2003 withI-i 004967 004992 Contention X attachments II ______

I-I Reactor System Word 0-1 Questions Documents 004993 005107 Contentions X

_____________ I and _I11_ __ _

Duke Response to NRC Letter with 11/3/2003 005108 005344 Contentions X RAI Attachments I and II Duke Response to NRC Letter with I-I RAI Attachments 11/3/2003 005345 005453 Contentions X

_____AttachmentsI and II, 11-5 E-mail 2/16/1999 005454 005454 II-5 X HardcopyI LOCA Section for MOX Letter 3/22/2002 005455 005490 I-I X L AR _ _ I__ _ _ I__ _ I__ _ I__ _ I_

11

'4.....,'r;jocum en t '<z-Documen,Decrpto 'Tp (iet..r. -W i~fit>sss 7.> ., Baltes Ad~ - .._+P,;piitoi i;'~~cac~t. .' ->; ;,;=.vi,--reap 7of 1Nmbr < Nvumbr J A -,U.htNbfi CluainSummar Form 7/30/2003 005491 005519 I- 1 X MOX Fuel Data for Letter 3/7/2002 005520 005521 I-l X LOCA Calculations Physics Parameters Memo 1/24/2002 005522 005525 I-l X

. Documents Com parison__ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _

Certification for Engineering Calculation Proposed Procedure for Calculating Wr Radionuclide Wr.1/2200058 059 I-4X Inventories for PRA Dom 1/12/2003 005588 005589 114 Input B Estimate uel est LMemo 12/3/2003 005590 005599 11-4 X Batch Dataion eShin etn Calculation_ e-mail __ __ _ __ __ _

hardcopy and 11/24/2003 005600 005618 I-4 X

.n o tes__ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _

Release fraction study Hand written 005619 005621 II-4 X N otes _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _

from W.F. Jones Letterwith 9/25/2002 005622 005650 I-l X A ttachm ents __ _ _ _ _ _ _ __ _ _ _ _ __ _ _ _ _ _

Calcula atonSm r Form 1/16/2004 005651 005686 I-I X Severe Accident Word 5/22/2001 005687 005705 II-1 X Analysis Report Documents __

Information Spreadsheets 005706 005926 II_ X Accident Analysis Doumnt Attachments 005927 006208 II-I X Duke Response to NRC Word 1/4/2002 006209 006236 II-I X RAIDa Documents Conditional Person Rem Do nd 006237 006240 II-I X Hand written 0061 00621 11-4 X note and article from W.F. Jones Letter ac ith 3/22/2002 006317 006347 II-I X CR2B2OAS.Sum Notebook 006348 006865 II-I X A _Analysis WNotebookrd 006866 007500 II-I X 12

t Ji Docuument`, r.eginning

. -Xi.-ding o un sp n - -

Daes . Bat~~~~es"R~iis

, ~ Propiet

~ -~-

ORNI, and SAIC Re-Work of Accident Internal Memo 007501 007516 II-I X Consequences e-mail 2/16/1999 007517 007519 II-I X hardcopy Risk Impact of MOX Word 007520 007520 II-1 X Fuel Lead Assemblies Documents Draft LAR Word 5/6/2002 007521 007690 II-I X Word Documents 007691 007693 II-1 X Word 079 066 i-Documents 0 Documents Word 007697 007698 11-1 X Duke Responses to NRC Word 007699 007726 II-I X RAIs Documents CNSState.xls;Land Excel 4/7/2004 007727 007751 11-1 X Fraction Excel 4/7/2004 0077 0771 I- _

CNS Pop Data Excel 4/7/2004 007752 007765 II-I X cnscalc.xls:State Excel 4/7/2004 007766 007778 II-1 X Information CNSSFRAC-- State Excel 007779 007875 II X sector fractions Excel 007779 00775____

CNS ISO.xls Excel 007876 007877 lI-1 X Release Category Excel 007878 007916 II-1 X Frequency Matrix Excel 007878 0096___

Summary of Risk Excel 007917 007949 II- I X Results _ _ _ _ _ _

Rem Chart Excel 007950 007950 II-I X McGuire and Catawba Nuclear Station PRA Notebook 007951 008504 Il-l X System Documentation Level 2 & 3 Analysis ..

Public Health risks of Substituting Mixed-Oxide for Uranium Fuel Article 008505 008528 11-6 in Pressurized Water Reactors 13

..Dd'6imiiWe, " Beginningi~ Ending.,

'C-UenDDescription ocuien aTy c m e h.bT6st j....... Rsphnsive PBtc roprietary

--."calcvt. . ° N 4- b Nm . - ._

Public Health risks of Substituting Mixed-Oxide for Uranium Fuel Article 1/20/1999 008529 008582 11-6 in Pressurized Water Reactors Hand written note and article 008583 008596 11-6 D o entR008597 nd 008618 II-6 Impact of MOX Fuel Word Lead Assemblies Le CA on Wr Documents 008621 008663 Impact of MOX Fuel Word Lead Assemblies on Documents 008664 008674 I-1 LOCA__ _ _ ___ _ _

MOX Fuel Lead Assembly License Word 10/9/2002 008675 008682 Amendment Request Documents Article 2/1/1999 008679 008620 II-6 Contention 10 Word 008683 008683 I-1 Documents 008683 008683 Draf DocmentWord Draft Document Documents 008684 008698 I-1 Comments on Section Word 6.1.1 Documents 008699 008701 6.1 Impact of MOX Fuel W d Lead Assemblies on ord LOCA Analyses Documents 6.1 Impact of MOX Fuel Word Lead Assemblies on o008711 008719 11 LOCA Analyses Documents008711 00871 Attachment 3 Description and Word Technical Justification Documents 008720 008722 6.1 Impact of MOX Fuel Word LOCA Analyses Documents 008723 008730 I-1 Attachment 3 Description and Word Technical Justification Documents 008731 008739-14

Documentesc ~ i D te of A.. ~ . ~nding, Responsive

- **.,--.. PiX A

'Vf',;:,'.:-,';_-.. 1., _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ ._1 . :di-Dcmn.:

io-.-.i.;-- r .' ;'t.0J;t;

.;cu-Destc.)7 f . Number -

6.1 Impact of MOX Fuel Word Lead Assemblies on Documents 008740 008765 I-I LOCA Analyses Word Draft Document Word 008766 008818 I-1 Documents Attachment I Partial Request to NRC Staff Word 008819 008848 Contention Request for Additional Documents 0I Information Cr2b MAAP Runs for Disc 002177- Disc Diskette 002177- II-4 X Quantification 002898 00289 _

SAAG No.0726 Disc 004397- Disc Support SAAG Disc 004429-004397- 11-1 X Suppote 0049 004429 CNS MOX Analysis Disc 004430- Disc 1999 Disc 004580 004430- lI-1, II- X 006208_ 004580 Disc 004582- Disc GSI 189 Discs Disc 0056 004582- lI-I X 004586 004586 _ _ _ _ _ _ _ _ _

Disc 005526- Disc DPC- 1607.00-00-0008 Disc 0057 005526- 1-1 X 005587 005587 _ _ _ _ _ _ _ _ _

CNSFTBCost SAAG Disc 005687- Disc S Disc 005687-2I-I X

669 005705 005705 SAA #86 NSDisc A Disc A SAAGS #ModeCN Diskette 004104- 004104- I1-2 X MACCS2 Model004396 004396 ____ _____

SAMAR~l SuporingDisc A Disc A SA A fils uprig Disc 005927- 005927- lI-i X fies006208 006208 _____

SAA #86 NSDisc B Disc A SAAGS #586dCN Diskette 004104- 004104- 11-2 X MACS2Moel004396 004396 _____

Disc B Disc B Same Files Disc 005927- 005927- lI-iIX

____ ________ _____ __ ____ ____ ___ ____ ____ 006208 006208 _ _ _ _ _ _ _ _ _ _

15

in egnninv,--

Do un n ~' §ter ! nEiing, Documen S-- s-DocumDnf ., Yat -

__ _ __ _ _ __ _ _ .. ; .-- tc.)

,> ; - ^.cal4;

'y -- r'i2- J. N

- 4 ub e m r umbier sN - S;-g f-.

Disc Disc A SAAG #586 CNS Diskette C004104- 004104- II-2 X MACCS2 Model 004396 004396 _____

Technical Reports Series Public Not Produced 1-1 No. 415 Document 2003 -- Publicly Contention I Available and II Not Produced 1-1 License Amendment Public 2/27/2003 -- Publicly Contention I Request Document Available and 1I Letter from Rn Public Not Produced Availablean 1-1 II Mcttollum to NRn Doument 3/26/2004 -- Publicly Contention to NRC Document Avial.I _____

Available I MCollum to NRC Document 3/16/2003 -- Publicly Contention Letter from WR Public Not Produced Available 1-1 11 Leterfrm R ubicNot Produced I-1 ro W Doce ument 2/2/2004 -- Publicly Contention McCollum to NRC Avalaleument____

Accident Source Terms for Light-Water Nuclear Public Not Produced II-6, I-1 Power Plants: High Dec-02 -- Publicly Contention Burnup and Mixed Document Available II Oxide Fuels Request for Additional Information on Public Not Produced I-1 Application for MOX -- Publicly Contention Lead Test Assemblies Document Available II I

Duke Power Co.

NRC Memo on Plan for Confirmatory Research Public 2/28/2000 -- Publicly Contention I Associated with the Use Document 2800lPbl Cnt I of MOX Fuel of MO FuelAvailable and II NRC Memo on Plan for Not Produced 11-6,1-1 Confirmatory Research Public 2/11/2002 -- Publicly Contention Associated with the Use Document Available II of MOX Fuel Summary December 12, Public Not Produced 1-1 2000 Meeting Memo Document 1/27/2001 -- Publicly Contention I from NRC ___Available.and II 16

. Doc

,-.Ae4-.d. , S ,t i niB g E ni

-g pocumient D&cripiidn' T;ype getter,-sE,*

'g, ^>

i..¢,Bates

'~tStv-r" ' v W@A;~T t -i.'..roret

~cale 7~ ~Niibibe f* ube Not Produced

-- Publicly Available.

This copy contains a few Draft NUREG-1437, Public May-00 identifying II-I Supplement 9 Document marks and minimal underlining, none of which reflect substantive thoughts 2/28/00 memo transmitting 2/11/00 NRC Memo on Agency Not Plan for Confirmatory Public 2/28/00 and Produced--

Research Associated Document 2/11/00 Available II-6 with the Use of Mixed- Publicly Oxide Fuel in Commercial Light Water Reactors Not Technical Report Series Public Produced-- 11-5 NO. 415, IAEA Document Available Publication Publicly NRC Panel on Source Not Term--IPSN Public Produced--16 presentation Part 1: Pument 12/11/2001 Produce II-6 Overview on French D1ble Studies on Source Term Publicly Note To File:

Information Provided by Duke Energy corporation Related to Not Severe Accident Public Produced-- IIl Mitigation Alternatives Document 3/14/2002 Available in its License Renewal Publicly Application for the Catawba Nuclear Station, Units 1 and 2 17 DC:353813.1

Attachment 4 PRIVILEGE LOG I

2 I reference Ireference 3

, reference 4

5 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of: )

)

DUKE ENERGY CORPORATION )

) Docket Nos. 50-413-OLA (Catawba Nuclear Station, ) 50-414-OLA Units 1 and 2) )

)

CERTIFICATE OF SERVICE I hereby certify that copies of "DUKE ENERGY CORPORATION'S RESPONSE TO BLUE RIDGE ENVIRONMENTAL DEFENSE LEAGUE'S FIRST DISCOVERY REQUEST," in the captioned proceeding, have been served on the following by deposit in the United States mail, first class, this 14th day of April, 2004. Additional e-mail service, designated by **, has been made this same day, as shown below.

Documents being provided in connection with this response have been delivered directly from Duke Energy Corporation by overnight courier for receipt this same day by counsel for Blue Ridge Environmental Defense League and the NRC Staff (identified by t).

Ann Marshall Young, Chairman** Anthony J. Baratta**

Administrative Judge Administrative Judge Atomic Safety and Licensing Board Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Washington, DC 20555-0001 (e-mail: AMY~nrc.gov) (e-mail: AJB5@nrc.gov)

Thomas S. Elleman** Office of the Secretary **

Administrative Judge U.S. Nuclear Regulatory Commission 5207 Creedmoor Road, #101 Washington, DC 20555 Raleigh, NC 27612 Attn: Rulemakings and Adjudications Staff (e-mail: ellemangeos.ncsu.edu) (original + two copies)

(e-mail: HEARINGDOCKET~nrc.gov)

Office of Commission Appellate Adjudicatory File Adjudication Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington, DC 20555 Washington, DC 20555 I

Susan L. Uttal, Esq. **t Mary Olson Antonio Fernandez, Esq. ** Director, Southeast Office Margaret J. Bupp** Nuclear Information and Resource Office of the General Counsel Service U.S. Nuclear Regulatory Commission P.O. Box 7586 Washington, DC 20555 Asheville, NC 28802 (e-mail: slu(nrc.gov)

(e-mail: axf2@nrc.gov)

(e-mail: mjb5@nrc.gov)

Diane Curran **t Harmon, Curran, Spielberg &

Eisenberg, LLP 1726 M Street, N.W.

Suite 600 Washington, DC 20036 (e-mail: dcurraneharmoncurran.com)

David A. Repka Counsel for Duke Energy Corporation 2

ML041130392

Text

References:

REFERENCES:

Navigation menu

Search