ML040510064

Response to Request for Additional Information on the Mixed Oxide Fuel Lead Assemblies (Environmental, Radiological, and Materials)

ML040510064
Person / Time
Site:	Catawba
Issue date:	02/02/2004
From:	Mccollum W Duke Energy Corp
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
-nr, TAC MB7863, TAC MB7864
Download: ML040510064 (43)

Text

Duke Duke Power

• rPower.

Energy Center A Duke Ein Company P.O. Box 1006 Charlotte, NC 28201-1006 February 2, 2004 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555

Subject:

Duke Energy Corporation Catawba Nuclear Station Units 1 & 2, Docket Nos. 50-413, 50-414 Response to Request for Additional Information (TAC Nos. MB7863, MB7864)

Mixed Oxide Fuel Lead Assemblies (Environmental, Radiological and Materials)

By letter dated February 27, 2003 Duke Energy submitted an application to amend the licenses of McGuire and Catawba to allow the use of four mixed oxide fuel lead assemblies. As part of the review of this application the Nuclear Regulatory Commission staff requested that Duke provide additional information related to the application in letters dated December 16, 2003 for radiological and materials related information, environmental information in a letter dated November 21, 2003 and QA related information in a letter dated December 24, 2003.

The environmental responses are included in Attachment 1, the radiological responses are included in Attachment 2, the materials responses are included in Attachment 3 and the QA information in included in Attachment 4. is the only information that is proprietary, an affidavit requesting withholding is included in the attachment. Inquiries on this correspondence should be directed to M.T. Cash at (704) 382-5826.

W.R. Mc Collum Senior Vice Pres ent - Nuclear Generation Duke Energy Corporation attachments

Oath and Affirmation I affirm that, WrR Mc Collurn, am the person who subscribed my name to the foregoing, and that all the matters and facts set forth herein are true and correct to the best of my knowledge.

"'VR Mc Collum Subscribed and sworn to before me on this Zh J day of r bV(

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Notary Public My Commission expires:

I/22 0

Date

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

cc: w/attachments L. A. Reyes U. S. Nuclear Regulatory Commission Regional Administrator, Region II Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, GA 30303 R. E. Martin (addressee only)

NRC Project Manager U. S. Nuclear Regulatory Commission Mail Stop 0-8G9 Washington, DC 20555-0001 E. F. Guthrie Senior Resident Inspector U. S. Nuclear Regulatory Commission Catawba Nuclear Station J. B. Brady Senior Resident Inspector U. S. Nuclear Regulatory Commission McGuire Nuclear Station Diane Curran Harmon, Curran, Spielberg & Eisenberg, LLP 1726 M Street, N.W.

Suite 600 Washington, DC 20036 Mary Olson Director, Southeast Office Nuclear Information and Resource Service P.O. Box 7586 Asheville, NC 28802 H. J. Porter, Director Division of Radioactive Waste Management Bureau of Land and Waste Management Department of Health and Environmental Control Columbia, SC 29201

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bcc: w/attachments Richard Clark-DCS Patrick Rhoads-DOE David Alberstein-DOE Don Spellman-ORNL NCMPA-1 NCEMC PMPA SRE bcc: w/attachment (via email)

S. P. Nesbit M. T. Cash F. J. Verbos J. L. Eller S. P. Schultz L. F Vaughn M. W. Scott L. J. Rudy J. Hoemer - Framatome ANP G. A. Meyer - Framatome ANP bcc: wv/attachments (paper copy)

NRIA File/ELL - ECO50 MOX File 1607.2304 Catavba Document Control File 801.01-CN04DM Catawba RGC Date File (J. M. Ferguson - CN01 SA)

%:1 Responses to Environmental Questions Submitted by the Staff on November 21, 2003 Introductory Response relevant to all eleven environmental requests for additional information:

It is important to briefly outline the scope of the relevant application under review for the purposes of answering all 11 requests for additional information (RAI) in the environmental area. This will bring into focus the depth of information needed to answer each specific question. The license application tinder review by the Nuclear Regulatory Commission (NRC) staff is for the proposed insertion of four lead assemblies (LAs) at Catawba Nuclear Station. The physical design and material composition of each LA is identical (within manufacturing tolerances), the physical design is based on the Framatome Advanced Mark BW design. The fuel assembly upper and lower nozzles are 304L stainless steel. The lower nozzle has a debris filter which is A-286 steel alloy.

The grid straps located axially along the fuel assembly are either Inconel 718 or M5T1 Zirconium alloy.

The hold down springs on the fuel assembly top nozzle are Inconel 718.

The fuel rod cladding is M5T^1 zirconium alloy as well as the rod upper and lower end caps. The fuel rod is filled with helium gas and contains a plenum spring manufactured from either 302 or 304 stainless steel.

With the exception of the M5Tb1 cladding the materials used in the fuel assembly structural components are typical of those currently or previously in use at Catawba Nuclear Station.

Although Catawba has not previously used the M5T^' alloy, the alloy has been used in at least 4 other pressurized water reactors. Each of these reactors has had an exemption granted (Exemptions listed as reference ) for the use of this specific material. An exemption request was submitted with this license amendment application and can be found in Attachment 6 of that submittal.

Each of the exemption requests relies on the NRC approval of the Framatome Topical Report (Reference 2) submitted and approved by the staff. In the Oconee "Environmental Assessment and Finding of No Significant Hazards" the staff noted that the "M5T11!

alloy is a proprietary zirconium based alloy, composed primarily of zirconium and niobium, that has demonstrated superior corrosion resistance and reduced irradiation growth relative to both standard and low tin zircaloy." The fuel vendor estimates that clad integrity (Reference 3) under normal operating conditions should be consistent with the Mark BW design. The Mark BW design is cited as having a failure rate of less than one rod per 100,000 rods, from all manufacturing related causes, since 1987.

The fuel pellet contains a mixture of U0 2 and PUO2 manufactured through a sintering process like that used for the current fuel which consists of only U02. The current fuel is referred to as low enriched uranium (LEU) fuel.

The fuel proposed in this application is referred to as Mixed Oxide (MOX) and has only been used in a number of limited applications in pressurized water reactors in the United States. However, I

European Reactors have more than 35 years of experience with MOX fuel. As of 1988 three European fabrication plants have produced more than 435,000 MOX fuel rods as which have been used in 35 different pressurized water reactors.

The plutonium for use in the Catawba fuel will be obtained from decommissioned nuclear weapons material blended down to a fissile content useful for reactor operations. By contrast, the European MOX fuel is recycled from commercial operating reactor fuel.

The Catawba fuel has been referred to as weapons grade and the European fuel as reactor grade, however this is somewhat of a misnomer with respect to the presence of impurities. The Catawba fuel will be chemically polished to meet specifications for reactor operations.

During manufacturing the uranium in the LEU fuel is enriched in the U-235 isotope to approximately 3 % to 5% with the balance of the uranium almost completely consisting of the U-238 isotope. During reactor operations a substantial portion of the uranium in LEU fuel is converted into plutonium. The MOX Fuel Design Report contains a figure (Reference 4) which demonstrates the buildup of plutonium in LEU fuel as a function of burnup. At the established burnup limit of 50 MWD/MTU (megawatt day/metric ton uranium) a nominal "weapons grade" MOX assembly is estimated to contain approximately 13 kg of plutonium whereas a LEU assembly would contain approximately 6 kg of plutonium. Therefore, even with current LEU fuel in Catawba and all operating reactors plutonium exists in substantial quantities.

No other primary or secondary plant structures, systems or components are affected by this application.

None of the plants structures, systems or components including waste systems will be modified and none of these systems will be operated in a different manner or with different operating limits. The introduction of the MOX assemblies does not represent the introduction of any new sources of compounds, materials or elements beyond the new clad alloy or the mixed oxide fuel.

In addition, Duke is not requesting any changes to Technical Specifications 3.4.16 "RCS Specific Activity" or 5.5.6 "Radioactive Effluent Controls Program" or planning any changes to the detailed Radioactive Effluent Controls in Selected Licensee Commitments (Chapter 16 of the UFSAR) Section 16.11.

1. Describe any change to the types, characteristics, or quantities of non-radiological effluents discharged to the environment as a result of the proposed change.

Response

There are no expected changes in the types, characteristics, or quantities of non-radiological effluents discharged to the environment associated with the proposed change.

As noted above this application is associated with the use of lead assemblies located in the reactor. There are no materials or chemicals introduced into the plant that could affect the characteristics or types of non-radiological effluents.

In addition, the method of operation of non-radiological waste systems will not be affected by this change.

There are no known mechanisms associated with a change in fuel isotopic content that would alter the non-radiological effluent quantity.

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2. Describe any changes to liquid radioactive effluents discharged as a result of the proposed change.

Response

There are no expected changes to the liquid radioactive effluents discharged as a result of this change. Duke provided a previous discussion of fuel integrity affects on effluent discharges in response to staff questions. The outer surfaces of the fuel assemblies which are exposed to the reactor coolant system are the same materials which have been used at Catawba nuclear station for many years. The exception is the introduction of the M5T~1 alloy. This material is a zirconium based alloy and is more corrosion resistant than currently used zirconium based alloys. Therefore, the fuel assembly surfaces exposed to reactor coolant should not interact to product any different quantity or type of radioactive material in the reactor coolant system.

As noted in the previous response (Reference 5) pages 113 to 117, the cladding performance of M5T^1 is expected to meet or exceed that of the current zircaloy cladding therefore there is not expected to be any increase in the quantity of failed fuel rods. In the event of failed fuel rods the MOX fuel could release fission products from the gap into the reactor coolant system. However, the chemical volume and control system and radioactive waste systems are designed to cope with fuel rod failures. The same fission products present from the failure of a LEU fuel rod would be present for the failure of a MOX fuel rod. A sample of slight differences in curie content of respective isotopes is presented in the prior response.

Therefore, based on the materials and performance capabilities of the fuel and plant systems there is no basis to expect any change in liquid effluent characteristics typical of normal plant operations.

In addition, Duke is not requesting any changes to Technical Specifications 3.4.16 "RCS Specific Activity" or 5.5.6 "Radioactive Effluent Controls Program" or planning any changes to the detailed Radioactive Effluent Controls in Selected Licensee Commitments (Chapter 16 of the UFSAR) Section 16.11. These requirements and commitments place limits on various isotopes and specify requirements for monitoring and surveillance. These requirements and commitments limit the release of gaseous and liquid radioactive effluents.

3. Describe any changes to gaseous radioactive effluents discharged as a result of the proposed change.

Response

For the same reasons as described in number 2 above this change would have no affects on the characteristics of gaseous radioactive effluents.

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4. Describe any change in the type or quantity of solid radioactive waste generated as a result of the proposed change.

Response

The introduction of the four lead assemblies should have minimal impact on solid waste.

There would be no expected impact on primary system filters or resins associated with normal plant operations. The information supplied in number 2 above provides the basis for reaching this conclusion.

There is a pool side post irradiation examination program which will be conducted for the MOX fuel assemblies. The quantity of waste associated with such a campaign is minimal and consistent with other post irradiation examinations performed during refueling outages. This waste would be small volumes of low level waste such as disposable portions of anti-contamination clothing.

5. What is the difference in source characteristics of mixed oxide (MOX) fuel, compared to the low enriched fuel that is considered in the accident analysis?

Response

Duke provided substantial information associated with radiological consequences in a prior submittal to the staff (Reference 5). Pages 101 to 107 of Reference 5 provide a detailed description comparing MOX and LEU fuel assembly isotopics for accident analysis purposes. Tables on pages 103 and 104 provide isotope by isotope curie values for certain representative LEU and MOX fuel assemblies.

Please refer to Attachment 2 of this submittal for additional information associated with design basis accident dose calculations.

6. What is the expected change in occupational dose as a result of the proposed change under normal and design basis accident conditions?

Response

Under normal power operation there would be no expected radiological impact on either the workforce or the public.

During fuel receipt operations the MOX fuel assemblies present a higher dose rate that affects workers performing the fuel receipt. Section 3.6.4 of the application estimates that approximately 20 to 42 mrem will be accumulated during fuel receipt and inspection. There are no other expected changes in normal occupational operating doses.

Control room dose is the only occupational dose that has been previously considered for design basis accident conditions. Duke provided substantial information associated with radiological consequences in a prior submittal to the staff (Reference 5). Pages 96 through 97 of Reference 5 provide tables which summarize the control room dose for the accidents of interest. The tables provide information for both LEU fuel as well as for MOX fuel that can be used to evaluate the associated change. Please refer to Attachment 4

2 of this submittal for additional information associated with design basis accident dose calculations.

7. What is the expected change in the public dose as a result of the proposed change under normal and DBA accident conditions?

Response

Dose to the public will not be changed by the use of four lead assemblies at Catawba Nuclear Station during normal operations.

As noted in items 2,3 and 4 above there is no basis to contemplate an increased source of liquid, gaseous or solid radiological effluents that could contribute to increased public exposure during normal operations.

This is consistent with the Department of Energy (DOE) EIS (Environmental Impact Statement) for the Surplus Plutonium Disposition Program (Reference 6). Page 2-100 states that "No change would be expected in the radiation dose to the general public from normal operations associated with disposition of MOX fuel at the proposed reactors".

In addition, page 2-101 includes Table 2-10 which demonstrates no incremental change in doses for 16 years of reactor operation.

The public accident doses are summarized for the EAB (Exclusion Area Boundary) and LPZ (Low Population Zone) on pages 96 and 97.

Please refer to Attachment 2 of this submittal for additional information associated with design basis accident dose calculations.

8. What are the performance characteristics of the packages that will be used to ship irradiated assemblies offsite?

Response

It is Duke understands that this question refers to any rod or rods that may be shipped offsite for post irradiation examination. Environmental affects of shipping fuel assemblies for permanent repository storage would be governed by the Yucca Mountain proceedings. Shipping of spent nuclear fuel is governed by 10 CFR Part 71 and the specific shipping package must be in compliance with the relevant sections of the regulations. A shipping container must have a certificate of compliance (COC) issued by the Nuclear Regulatory Commission staff. As specified in 10 CFR Part 71 Subpart D the applicant for the COC submits a Safety Analysis Report (SAR) which the staff then reviews against a number of standards. Upon staff approval the NRC staff issues a safety evaluation report (SER) describing the basis of approval.

Nuclear Regulatory Guide NUREG-1 616 "Standard Review Plan for Transportation Packages for Spent Nuclear Fuel" specify the numerous performance characteristics and acceptance criteria for package design.

The performance standards include standards for structural, thermal, containment, shielding and criticality as well as specific tests specified in 10 CFR 71.71 and 10 CFR 71.73 to demonstrate satisfaction of these performance standards.

The regulatory review for the issuance of a certificate of compliance addresses normal conditions of transport as well as hypothetical accident conditions. Specific acceptance criteria are 5

provided for normal and accident conditions.

Although a specific shipping package has not been selected for shipping the MOX rods the package must be have a COC issued by the staff and the performance standards of Part 71 or any requested exemptions will be granted based on a case specific basis for that shipping package.

Duke suggests that the appropriate environmental review for shipping these irradiated rod(s) is the staff review performed for the certificate of compliance along with the finding of No Significant Environmental Impact associated with the promulgation of 10 CFR Part 71. The rule does not contemplate a case by case environmental review, but compliance with the regulation and the issued COC.

9. What are the expected impacts of transporting the fresh MOX assemblies (to workers/drivers and to the public) under normal and accident conditions?

Response

It is Duke's position that the transportation of the MOX fuel assemblies is outside of the scope of the subject application and does not require an environmental review beyond that already performed by the Department of Energy. The transportation of the MOX fuel assemblies is the responsibility of the Department of Energy (DOE). For information purposes however this subject has been addressed by the DOE in the Supplement Environmental Impact Study DOE/EIS-229-SA3 "Supplement Analysis Fabrication of MOX Fuel Lead Assemblies in Europe" November 2003 (referred to herein as SA3).

SA 3 does provide a description of truck transportation risks from ports to Catawba Nuclear station in section 5.2 which describes the methodology and summarizes the results as well. Table 2 on page 17 of that document provides a summary of information responsive to this request for additional information.

Table 2 indicates that for incident free transportation the radiological risk to the crew is a maximum of 4.0 x 10-06 which corresponds to shipping from the Naval Station Norfolk port. The maximum radiological risk to the public for incident free transportation is 3.2 x 1006 also associated with shipping from Norfolk. The radiological risk is an estimate of the number of latent cancer fatalities (LCF), which are much less than one for the public and the driver. For accidents Table 2 provides and estimate of the radiological risk (LCFs) and for non-radiological risks stated as expected number of accident fatalities from non-radiological factors. The analysis does not separate the crew from the public.

For the case of accidents, the radiological risk is a maximum of 2.1 x 10-07 which corresponds to shipping from the Naval Station Norfolk port or Yorktown Naval Weapons Station. The maximum non-radiological risk is 1.7x 1 04 which also corresponds to shipping from the Naval Station Norfolk port or Yorktown Naval Weapons Station.

For both normal and accident conditions the fatalities associated with incident free or accidents during transportation are far less than one.

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10.

What are the expected impacts of transporting the irradiated MOX assemblies (to workers/drivers and to the public) under normal and accident conditions?

Response

As noted in number 8 above it is Duke Energy's understanding that this refers to the potential shipment of an irradiated rod or rods for the post irradiation examination process. Environmental affects of shipping fuel assemblies for permanent repository storage would be governed by the Yucca Mountain proceedings. The specific package chosen would determine the specific consequences in response to this question.

However, as noted above there has not been a specific package chosen at this time. The regulatory process would be the same as that outlined in number 8. Duke suggests that the appropriate environmental review for shipping these irradiated rod(s) is the staff review performed for the certificate of compliance along with the finding of "No Significant Environmental Impact" associated with the promulgation of 10 CFR Part 71.

The rule does not contemplate a case by case environmental review, but compliance with the regulation and the issued COC.

For informational purposes the Surplus Plutonium Disposition EIS (Reference 6) provides information of relevance to this question.

Specifically, Section 4.27.6.3 provides a bounding evaluation for eight shipments from McGuire Nuclear Station to Oak Ridge National Laboratory (ORNL). The post irradiation examination dose to workers is estimated at 0.2 person-rem and 1.2 person-rem to the public. The incident free transportation of the results in 6.7x1 0-5 LCFs for transportation workers and 5.9x10-04 LCFs in the total affected population for the duration of this assumed transportation activity. Estimated non-radiological fatalities from vehicular emissions are estimated at 3.7x106. Impacts from transportation accidents are estimated to be 1.2x10 4 LCF from radiation and 1.4x 10-04 from traffic fatalities.

11. Provide an assessment of the occupational doses resulting from post irradiation examinations following each cycle of irradiation of the lead assemblies.

Response

For informational purposes the Surplus Plutonium Disposition EIS (Reference 6) provides information of relevance to this question.

Specifically, Section 4.27.6.3 provides estimates of the radiological consequences for the hot cell examination of fuel assemblies at Oak Ridge National Laboratory. There are an estimated 10 workers associated with the hot cell examination work each estimated to accumulate approximately 177 mrem which translates into an estimated 7.1x10-4 LCFs.

References:

1. Oconee Nuclear Station, March 17, 2000 ADAMS ML003693687; Davis Besse Nuclear Station, March 15, 2000 ADAMS ML003693687; Three Mile Island 7

Nuclear Station, May 8, 2001 ADAMS MLO1 1280063 and Sequoyah Nuclear Station, July 29, 2000 ADAMS ML003736764.

2. BAW 10227-A, SER Issued on 12/14/1999 and SER revised on 2/4/2000, ADAMS ML003686365.

3.

BAW-10238(NP), MOX Fuel Design Report, May 2003 See Page 7-2 referring to fuel integrity.

4. BAW-10238(NP), MOX Fuel Design Report, May 2003 See Page 3-17 Figure 3.3

5. Duke Energy Response to Request for Additional Information Regarding the Use of Mixed Oxide Lead Assemblies, November 3, 2003

6.

Surplus Plutonium Disposition Final Environmental Impact Statement, DOE/EIS-0283, November 1999 8

Responses to Radiological Ouuestions Submitted by the Staff on December 16. 2003 Radiological Consequences By letter dated February 27, 2003, Duke requested a license amendment related to the use of four MOX lead test assemblies. The staff requested additional information (RAT) by letter dated July 25, 2003 and Duke responded to that RAI on November 3, 2003. In reviewing the responses to the radiological consequences questions, the NRC staff has identified some areas where additional clarification or information is required to enable the staff to make its requisite safety findings.

1.

With regard to Footnote 3 to Table Q12-3 (cited in the response to RAI Question 3),

Section 4.5 of Regulatory Guide 1.195, "Methods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light-Water Nuclear Power Plants," does state that 50 rem thyroid may be used as the acceptance criterion. However, Section B of the guide states: "The guidance contained in this regulatory guide will supersede corresponding radiological analysis assumptions provided in other regulatory guides when used in coniunction with guidance that is in Regulatory Guide 1. 196,

'Control Room Habitability at Light-Water Nuclear Power Reactors."' Please provide a commitment to the guidance of Regulatory Guide 1. 196, identifying proposed alternatives to the guidance, if any, that are proposed by Duke.

Response

Duke provided a 180 day response to Generic Letter 2003-01 on December 9, 2003 (Reference QI-1). Duke plans to conform to Regulatory Guide 1.195 (Reference Q1-2) and Regulatory Guide 1.196 (Reference Q1-3) once approval for full implementation of Alternative Source Term (AST) is received via a Safety Evaluation Report (SER) issued for the Reference Q1-4 submittal, and agreement is reached on implementation of the Control Room Habitability technical specification (TSTF-448). Currently, the additional margin provided by the acceptance criteria in Regulatory Guide 1.195 is not needed. Therefore, Table Q12-3 from Reference Q1-5 has been updated and is attached.

References Ql-l McCollum, W. R., December 9, 2003, Letter to USNRC, "Response to NRC Generic Letter 2003-01, Control Room Habitability."

QI-2 U. S. Nuclear Regulatory Commission, Regulatory Guide 1.195, "Methods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light-Water Nuclear Power Plants," May 2003.

Q1-3 U. S. Nuclear Regulatory Commission, Regulatory Guide 1.196, "Control Room Habitability at Light Water Nuclear Power Reactors," May 2003.

Q1-4 Peterson, G. R., December 20, 2001, Letter to US Nuclear Regulatory Commission, "Proposed Amendment for Partial Implementation of Alternate Source Term and

Proposed Amendment to Technical Specifications (TS) 3.7.10, Control Room Area Ventilation System, TS 3.7.11, Control Room Area Chilled Water System, TS 3.7.13, Fuel Handling Ventilation Exhaust System, and TS 3.9.3, Containment Penetrations."

QI-5 Barron, H. B., November 3, 2003, Letter to US Nuclear Regulatory Commission, "Response to Request for Additional Information Regarding the Use of Mixed Oxide Lead Fuel Assemblies."

Table Q12-3 Regulator Dose Criteria For Accidents with MOX Fuel Lead Assemblies Accident Classc Source Reference Alternative Reference Termnenec Souc Term eeec Offsite Doses (EAB ndL )_________

LOCA 300 Rem Thyroid 1 OCFR1 00.11 25 Rem TEDE RG' 1.183 25 Rem WB SRP' 15.6.5 App. A ROCFR50.67 Steam Generator Tube Rupture 300 Rem Thyroid I OCFRI00.11 RG 1.183 with fuel failure or pre-incident 25 Rem WB SRP 15.6.3 25 Rem TEDE 10CFR50.67 Iodine spike Steam Generator Tube Rupture 30 Rem Thyroid SRP 15.6.3 2.5 Rem TEDE RG 1.183 with concurrent Iodine spike 2.5 Rem WB Main Steam Line Break 300 Rem Thyroid IOCFR100.11 RG 1.183 with fuel failure or pre-incident 25 Rem WB SRP 15.1.5 App. A 25 Rem TEDE 10CFR50.67 Iodine spike.

Main Steam Line Break 30 Rem Thyroid SRP 15.1.5 App. A 2.5 Rem TEDE RG 1.183 with concurrent Iodine spike 2.5 Rem WB Locked Rotor Accident 30 Rem Thyroid SRP 15.3.3 2.5 Rem TEDE RG 1.183 2.5 Rem WB Rod Ejection Accident 75 Rem Thyroid SRP 15.4.8 App A 6.3 Rem TEDE RG 1.183 6 Rem WB Fuel Handling Accident 75 Rem Thyroid SRP 15.7.4 6.3 Rem TEDE RG 1.183 Control Room oses 3ORem Thyroid SRP 6.4 All I5 Rem WB IOCFR50/

5RemTEDE RG 1.183 30 Rem skin Appendix A/

ROCFR50.67 30________skin GDC 19 I

WB= Whole body, RG=Regulatory Guide, SRP= Standard Review Plan

2.

Table Q3(a)-I states that the control room X/Q value from Reference Q3(a)-I is "increased" to a value of 1.04E-3 seconds/cubic meters (sec/m3) for unit vent releases.

The citation to Reference Q3(a)-1 appears to be in error. The NRC staff believes that the most recently approved value for this parameter is 1.74E-3 sec/M3 (See Catawba Amendments 198 and 191 dated April 23, 2002). Please resolve this apparent difference in values. If the 1.04E-3 sec/M3 value is a newly calculated value, please provide a revised response to RAI Question 3.c.

Response

The atmospheric dispersion factor (X/Q) used was calculated based upon the methodology in Reference Q2-1. Duke agrees that the most recently approved base X/Q for the flow path from the unit vent to the control room is 1.74E-3 sec/M3. Since the plant possesses dual intakes, the effective X/Q has typically been calculated by multiplying this value by 0.5, resulting in a value of 8.70E-4 sec/M3 in accordance with Reference Q2-1. This methodology inherently assumes perfect flow balance between the two intakes. Recent Duke testing indicates that perfect flow balancing is difficult to achieve and maintain. Therefore, Duke has assumed a bounding flow distribution whereby 60% of the air flow is taken from the contaminated stream and 40% from the uncontaminated stream. Application of Reference Q2-1 results in multiplying the base x/Q by 0.6 (vice 0.5) resulting in a value of 1.04E-3 sec/M3.

The characterization of an "increase" in the value was made to emphasize that the value used in this analysis assumes more contaminated airflow than the typical dual intake model since perfect ventilation balancing was not modeled. Thus, the value used (1.04E-3 sec/i 3) is "increased" and conservative relative to the value calculated by assuming perfect ventilation flow balancing (8.70E-4 sec/m3).

Reference Q2-1 U. S. Nuclear Regulatory Commission, Regulatory Guide 1.194, "Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants," June 2003.

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3.

In response to RAI Question 3.b on Page 93, Duke indicates that a loss of coolant accident (LOCA) is the most restrictive accident and its thyroid dose is more restrictive than its whole body dose. Please explain whether control room dose was considered in this conclusion and, if not, how this omission would impact this conclusion.

Response

Control room dose was considered in concluding that the loss of coolant accident (LOCA) is the most restrictive accident and the thyroid dose is more restrictive than the whole body dose. Duke has recently been performing consequence analyses in support of Alternative Source Term (AST) submittals which include control room doses. These analyses were performed in accordance with the current licensing basis of full low enrichment uranium (LEU) cores and include those submitted in References Q3-1 and Q3-2 as well as some secondary side accidents which are being analyzed in preparation for the adoption of AST. The release points for some secondary accident scenarios are closer to the control room intakes than the LOCA release points, but other aspects of the LOCA (such as source term) make its results the most limiting for control room doses as well as offsite doses. The limiting organ dose has historically and typically been the thyroid dose rather than the whole body dose for all accidents in the control room and offsite.

As was discussed above, and in the response to question 3 in Reference Q3-3, the models and analyses which form the basis for the determination of the limiting scenario and the limiting organ dose did not involve MOX fuel. These models were based upon all LEU cores. Given the results of the analyses and evaluations provided in Reference Q3-3, there is no reason to suspect that the MOX lead assemblies will alter these conclusions.

References Q3-1 Peterson, G. R., December 20, 2001, Letter to US Nuclear Regulatory Commission, "Proposed Amendment for Partial Implementation of Alternate Source Term and Proposed Amendment to Technical Specifications (TS) 3.7.10, Control Room Area Ventilation System, TS 3.7.11, Control Room Area Chilled Water System, TS 3.7.13, Fuel Handling Ventilation Exhaust System, and TS 3.9.3, Containment Penetrations."

Q3-2 Peterson, G. R., November 25, 2002, Letter to US Nuclear Regulatory Commission, "Proposed Technical Specifications and Bases Amendment: Technical Specification and Bases 3.6.10 Annulus Ventilation System, Technical Specification and Bases 3.6.16 Reactor Building, Technical Specification and Bases 3.7.10 Control Room Area Ventilation System (CRAVS), Technical Specification and Bases 3.7.12 Auxiliary Building Filtered Ventilation Exhaust System (ABFVES), Technical Specification and Bases 3.7.13 Fuel Handling Ventilation Exhaust System (FHVES), Technical Specification and Bases 3.9.3 Containment Penetrations, Technical Specification and Bases 5.5.1 Ventilation Filter Testing Program (VFTP)."

Q3-3 Barron, H. B., November 3, 2003, Letter to US Nuclear Regulatory Commission, "Response to Request for Additional Information Regarding the Use of Mixed Oxide Lead Fuel Assemblies."

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4.

A sentence in the last paragraph on page 94 states that: "Included in these tables is an evaluation of the results for the Catawba AST [accident source term] LOCA including MOX lead assemblies." That sentence could imply that the NRC staff is currently reviewing an analysis of a LOCA with AST and MOX lead assemblies for Catawba. This would be in error since Duke did not address MOX fuel in the AST amendment currently under review. Please clarify.

Response

It was not the intent of this response to imply that Duke has a submittal under NRC review to evaluate a Loss of Coolant Accident (LOCA) with mixed oxide (MOX) lead assemblies utilizing Alternative Source Term (AST) technology. No such submittal exists.

It was intended to communicate that there is a submittal under NRC review which models a LOCA accident and utilizes AST (Reference Q4-1). This submittal models cores consisting entirely of low enrichment uranium (LEU) fuel assemblies. This base model was then used to evaluate the expected impact to consequences including four MOX lead fuel assemblies. This evaluation was performed in a manner similar to other core wide impact accidents discussed in the response to question 3 in Reference Q4-2. This evaluation was performed by Duke for the purpose of providing supporting information for the review of the MOX lead assembly submittal (Reference Q4-3).

References Q4-1 Peterson, G. R., November 25, 2002, Letter to US Nuclear Regulatory Commission, "Proposed Technical Specifications and Bases Amendment: Technical Specification and Bases 3.6.10 Annulus Ventilation System, Technical Specification and Bases 3.6.16 Reactor Building, Technical Specification and Bases 3.7.10 Control Room Area Ventilation System (CRAVS), Technical Specification and Bases 3.7.12 Auxiliary Building Filtered Ventilation Exhaust System (ABFVES), Technical Specification and Bases 3.7.13 Fuel Handling Ventilation Exhaust System (FHVES), Technical Specification and Bases 3.9.3 Containment Penetrations, Technical Specification and Bases 5.5.1 Ventilation Filter Testing Program (VFTP)."

Q4-2 Barron, H. B., November 3, 2003, Letter to US Nuclear Regulatory Commission, "Response to Request for Additional Information Regarding the Use of Mixed Oxide Lead Fuel Assemblies."

Q4-3 Tuckman, M. S., February 27, 2003, Letter to U.S. Nuclear Regulatory Commission, "Proposed Amendments to the Facility Operating License and Technical Specifications to Allow Insertion of Mixed Oxide Fuel Lead Assemblies and Request for Exemption from Certain Regulations in 10 CFR Part 50."

6

5.

The NRC staff finds the response to RAI Question 3.g. to be inadequate. This question asked for ajustification for the continued use of the gap fractions in Table 3 of RG 1.183, "Alternative Radiological Source Terms For Evaluating Design Basis accidents At Nuclear Power Reactors." Duke responded to the staff's question by proposing to increase the Table 3 values (except alkali metals) by a multiplier of 1.5 without an adequate explanation of why the Table 3 values, adjusted by the multiplier, would reasonably bound the expected gap fractions for weapons-grade MOX assemblies with burnups to 60 Gigawatt days/Metric ton heavy metal (GWd/MThm). Duke is requested to provide the NRC staff with a basis for concluding that this multiplier is acceptable for all isotopic groups, including alkali metal, for weapons-grade MOX.

The NRC staff believes that the discussion provided in the response could be considered to provide qualitative support for Duke's conclusions regarding LOCA release fractions for the gap phase, but it does not establish the adequacy of the gap fraction multiplier for use with Table 3 for non-LOCA events. The expert panel identified in Duke's response was impaneled by the NRC in 2001 to address (1) whether or not the data in NUREG-1465, "Accident Source Terms for Light-Water Nuclear Power Plants," would apply to high burnup LEU fuel; (2) and to reactors using mixed oxide fuel. The panel reviewed available data and made recommendations for changes to the NUREG-1465 release phase fractions. Duke attempted to use the insights of the expert panel to address the adequacy of the Table 3 values as stated in the following paragraph from page 106 of Duke's November 3, 2003 response:

Since [Regulatory Guide 1.183] Table 3 is based upon expert panel work which was published in [NUREG-1465] and the panel saw similarities in gap release rates between LEU and MOX fuel, it could be inferred that the gap release rates in

[Regulatory Guide 1.183] Table 3 should also be valid for MOX fuel gap releases.

However, (1) the data in Table 3 were not derived from NUREG-1465. These data were generated by the NRC staff in recognition of the fact that the core average gap release fractions in NUREG-1465 were inappropriate for use with non-LOCA events since the gap fraction of many assemblies in the core could exceed the core average value; (2) the expert panel's deliberations were limited to LOCAs and other severe accidents involving a substantial portion of the core, since this was the direction given to the panel; (3) section 3.4.2 of the panel report tabulated the MOX fuel characteristics considered by the panel. This included a maximum burnup on an assembly basis of approximately 46 Gigawatt days/ton (GWd/t). Duke has requested MOX lead test assembly (LTA) burnups to 60 GWd/t; and (4) The panel's conclusions are not directly applicable to a comparison of an LEU assembly and a MOX assembly since the panel considered core-average releases from the LEU core and core-average releases from the core containing 40%

MOX assemblies.

7

Response

Reference Q5-5 describes and summarizes European experience with mixed oxide (MOX) fuel and the post irradiation examination (PIE) programs performed on this fuel. This testing includes both destructive and non-destructive examinations performed in spent fuel pools and hot cells.

The PIE data showed comparable fission gas releases for MOX and low enrichment uranium (LEU) fuel with differences in the fission gas release (FGR) rate attributed to the differences in the fuel/clad gap dimension and irradiation history. Rod puncture data showed a higher FGR for MOX compared to LEU. This behavior was attributed to higher linear heat rates associated with the MOX assemblies. MOX material properties were felt to also play a role in the increased MOX FGR rate, but to a lesser extent. Additionally, hot cell data on four cycle fuel rods did not show a relationship between increased FGR and burnup. Since the linear heat rates of the final cycle were low for these rods, it was concluded that FGR increases for MOX fuel are determined by the rate of power production at the end of life instead of being determined solely by burnup. It was concluded in Reference Q5-5 that the overall performance of MOX fuel is similar to that of equivalent LEU fuel, but with a somewhat higher fission gas release. It was also reported that LEU fuel FGR models have been used for MOX fuel FGR modeling with satisfactory results. Additionally, References Q5-3 and Q5-4 document the work of two expert panels related to the expected behavior of LEU from which some conclusions can be drawn about the general expected behavior of MOX fuel. These documents infer that the gas release processes are similar for LEU and MOX. Additionally, there are no discussions or data in these documents which conclude or suggest any significant deviations in the behavior of these two fuels. It is recognized that the body of work related to MOX fuel leaves some uncertainty regarding the absolute expected behavior of some characteristics. Because of this uncertainty, a conservative release model was derived (see response to Question 6 below), to offset these uncertainties in performance parameters.

In deriving an isotopic release scheme to apply to MOX fuel, Duke desired to model conservative release fractions that would bound those currently applied to limiting scenarios for LEU fuel and to provide additional conservatism and margin for the four MOX lead assemblies. Since the particular accidents that are most susceptible to an impact in consequences from the four MOX fuel lead assemblies are those with a low population of affected fuel assemblies (as discussed in Reference Q5-1), these assemblies would need to be modeled as limiting, rather than average, fuel assemblies. Their releases would then be based upon limiting modeling and assumptions.

Thus, as a starting point, existing data were reviewed for release modeling which is applicable to events where the gap fraction could exceed core average values. Table 3 of Reference Q5-2 provides LEU release fractions for more limiting releases scenarios, and serves as a basis for the derivation for MOX release fractions.

The response to question 6 below provides a discussion of fission gas release data from European MOX fuel as it is applied to the analyses in this application. The 50% increase applied to the Reference Q5-2 Table 3 release rates (except for alkali metals) is made to provide conservatism.

Alkali metals are released in particulate form and retained (partitioned) in the spent fuel pool water during a fuel handling accident. Therefore, the consequences (results) for this scenario are insensitive to this parameter. Thus, the 50% increase could also be applied to this isotopic group with little impact on the dose results.

8

See the response to Question 6 below for further details.

References Q5-1 Barron, H. B., November 3, 2003, Letter to US Nuclear Regulatory Commission, "Response to Request for Additional Information Regarding the Use of Mixed Oxide Lead Fuel Assemblies."

Q5-2 Regulatory Guide 1.183, "Alternative Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000.

Q5-3 ERINRC 02-202, "Accident Source Term for Light-Water Nuclear Power Plants: High Burnup and Mixed Oxide Fuels," November 2002.

Q5-4 USNRC, NUREG-1465, "Accident Source Terms for Light-Water Nuclear Power Plants," February 1995.

Q5-5 International Atomic Energy Agency, Technical Report Series No. 415, "Status and Advances in MOX Fuel Technology," May 2003.

6.

In the response to Question 3.g, Duke provided a graph of fission gas data for European reactor-grade MOX fuel and LEU fuel. With regard to this graph, please provide the following information:

a An explanation of the data sets represented on this graph. For example, what fuel configurations are included, what plutonium concentrations, what LEU enrichments, PWR / BWR / MAGNOX / etc. How were the data obtained?

b.

Was the linear heat generation rate (LHGR) for these assemblies comparable to what Duke proposes for the MOX LTAs?

c.

How do the data showing a nearly vertical rise in the MOX fission gas release (FGR) that occurs about 43,000 MWd/MThm support Duke's planned bumup to 60,000 MWd/MThm?

d.

Taking the four highest MOX data points at about 43,000 MWd/MThm, the MOX FGR values range from 4.5 to 7, implying an uncertainty of nearly 50% at this burnup. How is this uncertainty addressed?

Response (a)

Most of the low enriched uranium (LEU) data were obtained from 17x17 matrix fuel rods irradiated in Electricit6 de France (EDF) pressurized water reactors (PWRs) operated in load following conditions. These fuel rods were used in the following reactors: Bugey 3, Chinon B3, Cruas 2, Fessenheim 1, Fessenheim 2, Gravelines 2, Gravelines 3 and Gravelines 5. Typical enrichment varies between 3.25 and 4.50%. Two fuel rods in the data base were irradiated in the German Brokdorf (KBR) reactor using a 16x16 fuel matrix. Like most of the reactors involved in producing the European data, Catawba Nuclear Station (CNS) uses 17x17 fuel.

9

The mixed oxide (MOX) data were obtained from fuel rods irradiated in EDF PWRs operated in base-loaded or in load following conditions. These fuel rods were used in the following reactors:

St. Laurent B1, St. Laurent B2, Gravelines 4, and Dampierre 2. The fuel pellets were fabricated from depleted uranium and reactor grade plutonium using the MIMAS process. MOX fuel assemblies are characterized by radial zoning, so they contain fuel rods with three separate concentrations of plutonium. Typical initial plutonium (total) concentration in the rods ranged from 2 to 6%.

This data base is essentially the same as the fission gas release data that was used to develop and qualify the COPERNIC fission gas release model. Further information on the data is provided in Reference Q6-1, Sections 3 and 5, and Reference Q6-2.

Both LEU and MOX data were obtained through standard techniques. The fuel rods were shipped to hot cells, the rods were punctured, and the gas collected and analyzed in order to determine the helium, xenon and krypton contents as well as the xenon and krypton isotopic composition. The fractional fission gas release is the ratio of the collected Xe and Kr to the calculated production of Xe and Kr during irradiation.

Response (b)

The axially averaged linear heat generation rate (LHGR) for the MOX data base fuel is comparable to the projected rates in the Duke MOX lead assembly core designs. Comparisons based upon this data are felt to be appropriate since the LHGR of the fuel in this database encompasses the expected range of operation. The axially averaged linear heat rate in the Catawba core is 5.58 kW/ft. The predicted peaking factors for the peak rod in a representative MOX fuel lead assembly core design were provided in Table 2 of Reference Q6-3. The peaking factor and linear heat rate ranges for three cycles of lead assembly use are shown in Table Q6-1.

The maximum linear heat generation rate (axially-averaged) for the peak fuel rod is 7.9 kW/ft for a short time at the beginning of the first cycle. Fuel rod fission gas release is generally insensitive to power peaking of this magnitude that occurs early in fuel lifetime.

For the MOX fuel rod fission gas release data, the maximum linear heat generation (axially-averaged) during irradiation ranged from 4.7 kW/ft to 7.4 kW/ft. While the representative MOX fuel lead assembly LHGR briefly exceeds the range of the data, this occurs only for a short time at the beginning of the cycle. The average power for the peak rod is below the top of the data set range.

Note that the LHGR data in Table Q6-1 are for the peak rod. The average LHGR for these rods is typically lower by 10%.

10

Table Q6-1 Representative MOX Fuel Lead Assembly Peaking and Linear Heat Generation Rates (Axially-averaged)

Radial Peaking Peak Rod Linear Peak Rod Average Average Assembly Factor Heat Rate Cycle Power Power Cycle (Peak Rod)

(kW/ft)

(kW/fl)

(kW/fl) 1 1.22 - 1.42 6.8 - 7.9 6.99 6.42 2

1.14 - 1.36 6.4 - 7.6 6.76 6.12 3

0.59 - 0.68 3.3 - 3.8 3.43 1.89 Response (c)

Reference Q6-1, Section 5.1 discusses the phenomenon of fission gas release (FGR). FGR is a complex phenomenon involving mechanisms that are classified in two main groups: athermal release and thermal release. Athermal release is always present and is the result of the displacement of gas atoms by recoil and knockout and the subsequent escape of those atoms from the pellet through free surfaces. Thermal release occurs above a temperature threshold which decreases with burnup. Section 5.3.1 of Reference Q6-1 notes that for MOX fuel, the effective burnup for fission gas release is higher than the average burnup due to MOX heterogeneities. As a result, thermal release would be expected to become significant at a lower burnup for MOX fuel than for LEU fuel.

The data reflect a substantial increase in MOX fuel release fraction at a burnup of around 40,000 MWD/MThm. The LEU data on the same figure could be interpreted to involve a similar "nearly vertical rise" at somewhat higher burnups.

This effect is attributed to enhanced thermal release as fuel temperature increases and the thermal incubation threshold decreases with burnup (see Figure 5-2 of Reference Q6-1). The MOX fuel data around 50,000 MWD/MThm indicate that the MOX fuel release fraction does not continue an asymptotic increase from 43,000 MWD/MThm. Thus, fission gas release rates are not determined solely by burnup. Increases in the fission gas release rate are associated with increased thermal release by the pellet as the thermal incubation threshold is crossed. For CNS MOX fuel lead assemblies in their third cycle, the linear heat rate is reduced (relative to the second cycle) as the fuel assembly is moved to a lower power location. This lower heat rate drops the assembly below the incubation curve and reduces the driving thermal release mechanism, resulting in a lower fission gas release in the third cycle. Finally, it should be noted that there is some scatter in the experimental database for both MOX fuel and LEU fuel, as shown in Figures 5-6 and 5-7 of Reference Q6-1, which is attributable to the different linear heat rates experienced by the examined fuel rods. The scatter contributes to the perception of a "nearly vertical rise."

11

Response (d)

As noted in the response to Question 6(c), there is scatter in the experimental database for fission gas release (FGR) for both MOX and LEU fuel. The scatter (mainly for the MOX fuel rods) is due to the different heat rates (during second or the third irradiation cycle) experienced by the fuel rods (Reference Q6-5). However, all of the MOX data are below the "Raised AST" values that have been chosen to model the radiological consequences analyses for MOX fuel lead assemblies (Reference Q6-4, response to Question 3.g).

A review of the dose results for the fuel handling accident analysis performed in support of the MOX lead assembly submittal shows that 1-131 drives the dose result. It results in over 97% of the thyroid dose, and the thyroid dose is approximately 98% of the TEDE dose to the control room operators and over 86% of the offsite TEDE dose. These models include the effect of the MOX lead assembly in the scenarios, subject to the "Raised AST" release fractions. The 50%

increase over Regulatory Guide 1.183 values resulted in release fractions for all of the applicable fuel handling accident isotope groups which were greater than the highest fission gas release rate in the MOX data. In the case of I-13 1, which accounts for the vast majority of the dose in the accident scenarios with small populations of affected fuel assemblies, the release fraction used (12%) is almost twice that of the worst MOX data point.

Any time experimental or testing data is taken in the field, there will be some uncertainty and some data scatter. This will occur whether the fuel being tested is LEU or MOX. Duke has adopted release fractions for the MOX lead assemblies that bound the expected performance of the MOX assemblies relative to the LEU assemblies and also include margin for uncertainties and data scatter. These conservative input values were used in the dose analyses, along with other conservatisms in the dose methodology (e.g. bounding materiel and power history parameters used in deriving the assembly isotopic inventory; cladding failure of all fuel pins in the affected assembly; ventilation system modeling which compresses the release into a short time frame, does not credit fuel building ventilation filtration and does not assume perfect control room ventilation intake flow balancing).

Taking the overall conservative nature of design basis accident analyses into account, Duke concludes that the results of the dose analyses presented in Reference Q6-4 are conservative relative to actual doses that would be expected to occur in the unlikely event of an accident involving a radionuclide release from MOX fuel. Accordingly, these analyses demonstrate reasonable assurance that the health and safety of the public will be adequately protected.

References Q6-1 BAW-1023 1P, "COPERNIC Fuel Rod Design Computer Code," Framatome-ANP, September 1999 (Proprietary).

Q6-2 Letter, U. S. Nuclear Regulatory Commission to Mallay, J. (FANP), "Final Safety Evaluation for Topical Report BAW-1023 1P 'COPERNIC Fuel Rod Design Code,'

MOX Applications (TAC No. MB 7547)", January 14, 2004.

Q6-3 Letter, Canady, K. S. (Duke) to U. S. Nuclear Regulatory Commission, December 10, 2003, "Duke Energy Corporation Catawba Nuclear Station Units 1 & 2, Docket Nos. 50-12

413, 50-414 Response to Request for Additional Information dated November21, 2003 Regarding Mixed Oxide Fuel Lead Assemblies (TAC nos. MB7863, MB7864)."

Q6-4 Letter, Barron, K. S. (Duke) to U. S. Nuclear Regulatory Commission, November 3, 2003, "Catawba Nuclear Station Units I & 2, Docket Nos. 50-413, 50-414, McGuire Nuclear Station Units I and 2, Docket Nos. 50-369, 50-370, Response to Request for Additional Information Regarding the Use of Mixed Oxide Lead Fuel Assemblies."

Q6-5 P.Blanpain et al; "MOX Fuel Experience in French Power Plants," ANS International Topical Meeting On Light Water Reactor Fuel Performance, West Palm Beach, Florida, April 17-21, 1994.

13 Responses to Material Questions Submitted by the Staff on December 16, 2003 Request:

Materials Engineering Section 3.6.1 of Attachment 3 to the Duke Power (licensee or Duke) letter dated February 27, 2003, indicates that the fast flux impacting the reactor vessel will be virtually identical to that for a reactor core consisting entirely of low enriched uranium (LEU) fuel.

The licensee states that the Reactor Vessel Integrity Program will manage the reduction in fracture toughness of the reactor vessel beltline region so that the function of the vessel is maintained. The licensee states that the existing pressure-temperature curves in the Catawba Technical Specifications will remain valid with the use of four mixed oxide (MOX) lead test assemblies.

The Nuclear Regulatory Commission (NRC) staff requests that the licensee identify the capsules, dosimetry, capsule withdrawal schedule, and projected neutron fluence for the capsules that will be in the vessel during the period of time that the MOX lead test assembly fuel will be utilized. The test results from the reactor vessel material samples should be compared to the results predicted using Regulatory Guide (RG) 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," and dosimetry should be evaluated in accordance with RG 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence." Provide a basis for concluding that there is no change required in the withdrawal schedule for the capsules in the vessel during the period of time that MOX lead test assembly fuel will be utilized.

Response

There will be no surveillance capsules in a Catawba reactor vessel during use of MOX fuel lead assemblies. The purpose of surveillance capsules is to irradiate reactor vessel material samples to designated fluence levels prior to reaching those fluence levels on the reactor vessels themselves. Once the material samples in the final surveillance capsules have reached fluence levels corresponding to vessel fluence at end of life, the capsules should be withdrawn. In accordance with 10 CFR Part 50 Appendix H, the Catawba Reactor Vessel Material Surveillance Program has followed American Society for Testing and Materials guidelines (Reference 1) with respect to withdrawal schedules for surveillance capsules. Duke withdrew the last surveillance capsule from Catawba Unit 1 at the end of cycle 14 in the fall of 2003 when the capsule reached approximately 55 effective full power years (EFPY) of irradiation.

Duke will withdraw the last surveillance capsule from Catawba Unit 2 at the end of cycle 14 in the spring of 2006, with an estimated fluence equivalent to 55 EFPY. These fluence levels bound the fluence levels that are anticipated for sixty calendar years of operation (forty years on the initial operating license plus a twenty year license extension period).

I

The use of four MOX fuel lead assemblies will have no significant impact on the end-of-life fluence experienced by a Catawba reactor vessel. While the neutron energy spectrum from plutonium fissions is slightly higher than the spectrum from uranium fissions, the four MOX fuel lead assemblies represent only about 2% of the 193 fuel assemblies in the core.

Duke plans to use the MOX fuel lead assemblies for three operating cycles. For the first two cycles, the MOX assemblies will be loaded in the interior of the core (e.g., core location C8). For the third cycle, one or more MOX fuel lead assemblies will most likely be loaded in a core location at or near the core periphery (e.g., core location C14). A representative core loading map for the first cycle is shown in Figure QI 1-1 of Reference

2. It should be noted that the actual MOX fuel assembly core locations have not been finalized and will be determined as part of the cycle specific reload design. As discussed below, the incremental impact of the four MOX fuel lead assemblies on reactor vessel fluence will be insignificant.

In Reference 2, Response to Question 11, Duke showed that using four MOX fuel assemblies during the first cycle of operation will have a negligible impact on the fast flux in the core. At the beginning of the first cycle, Figure QI 1-2 of Reference 2 shows that the maximum calculated impact is a fast flux increase of 6.4% in the MOX fuel location itself (C8). Peripheral core locations are the most important with respect to the leakage of neutrons out of the core, and the maximum increase in fast flux in a peripheral fuel assembly is only 1.6% at the beginning of the first cycle. The small incremental impact of using MOX fuel on fast flux decreases further with burnup, because conventional LEU fuel assemblies produce more and more of their power from plutonium fissions as their burnup increases. Figure Qi 1-3 shows that at the end of the first cycle the impact of using four MOX fuel lead assemblies on the fast flux is less than 1% in all core locations.

Burup effects will make the incremental impact of using MOX fuel during the second cycle even smaller than during the first cycle. In the third cycle, with MOX fuel loaded in an exterior core location, any MOX fuel-related increase in fast flux would have more potential to affect the fluence at the vessel. However, the difference between a twice-burned MOX fuel assembly and a twice-burned LEU fuel assembly is very small.

As noted in Reference 3, at a burnup of 50 gigawatt-day per ton, "... only 36% of LEU fuel fissions are in uranium, so most of the power is coming from plutonium fissions. At this burnup the characteristics of LEU fuel have become very similar to those of MOX fuel."

Accordingly, during the third cycle of irradiation there will be little difference between the neutron energy from a MOX fuel assembly and the neutron energy from a twice-burned LEU fuel assembly that would otherwise be loaded at the expected location on the core periphery.

Therefore, the impact of four MOX fuel lead assemblies on vessel fluence should be negligible during all three cycles of operation.

The above conclusion that the use of four MOX fuel lead assemblies will have no significant impact on the end-of-life fluence experienced by a Catawba reactor vessel is 2

supported by bounding analyses of large-scale MOX fuel use. On the behalf of Duke Power, Westinghouse performed a study of the impact of large-scale MOX fuel use at McGuire and Catawba on reactor vessel integrity. The work is summarized in Reference

4. Extremely conservative boundary conditions were assumed, such as operation for twenty years per unit with cores comprised exclusively of MOX fuel. Even so, the end-of-life reactor vessel fluence increased by only 4-6% relative to a case assuming exclusive use of LEU fuel. With a more realistic loading pattern assumption (40% MOX fuel cores, with the MOX fuel loaded in interior core locations), analyses indicated that twenty years of operation with MOX fuel would result in a reactor vessel fluence increase of less than 1%. These large-scale MOX fuel use analyses assumed a throughput of hundreds of MOX fuel assemblies at each unit over many years. It is apparent that a limited MOX fuel deployment (four MOX fuel lead assemblies for less than five years) would have a negligible impact on end-of-life vessel fluence. No change in the capsule withdrawal schedule is needed.

It should be noted that an ex-vessel cavity dosimetry program is being implemented at both Catawba Unit 1 and Catawba Unit 2. This program will supplement the surveillance capsule program and monitor the reactor vessel fluence.

Ex-vessel dosimetry was installed in Catawba Unit 1 in 2003 and will be installed in Catawba Unit 2 in 2004.

The ex-vessel cavity dosimetry program will confirm that the predictions of vessel fluence used to assess vessel embrittlement are conservative.

Finally, it should be noted that the beltline regions of the Catawba reactor vessels were constructed with relatively low copper materials. The Nuclear Regulatory Commission staff calculated the highest Catawba end-of-life nil ductility temperature (RTpTS) to be 133F at 54 effective full power years (Reference 5, Section 4.2.2.2), as compared to the 10 CFR 50.61 screening criteria of 2707. It is evident that substantial margin is present at Catawba in the area of reactor vessel beltline region fracture toughness.

References

1. E-185-82, Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, American Society of Testing Materials.

2. Letter, Tuckman, M. S. (Duke Power) to U. S. Nuclear Regulatory Commission, Partial Response to Request for Additional Information Regarding the Use of Mixed Oxide Lead Fuel Assemblies, October 3, 2003.

3. 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, American Nuclear Society, Hilton Head, SC, October 2003.

4. Anderson, S. L., et. al., "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, American Nuclear Society, Charleston, SC, September 2002.

5. NUREG-1772, Safety Evaluation Report Related to the License Renewal of McGuire Nuclear Station, Units I and 2, and Catawba Nuclear Station, Units 1 and 2, U.S Nuclear Regulatory Commission 2003.

3 Quality Assurance Plan In Response to Staff Request of December 24, 2003 Non-Proprietary Version

AFFIDAVIT COMMONWEALTH OF VIRGINIA

)

) ss.

CITY OF LYNCHBURG

)

1.

My name is Gayle F. Elliott. I am Manager, Product Licensing in Regulatory Affairs, for Framatome ANP (FANP"), and as such I am authorized to execute this Affidavit.

2.

1 am familiar with the criteria applied by FANP to determine whether certain FANP information is proprietary. I am familiar with the policies established by FANP to ensure the proper application of these criteria.

3.

1 am familiar with the Quality Plan for European Fabrication of Mark-BW/MOX Lead Assemblies" referenced in a letter from H. B. Barron, Duke Energy Corporation, to the U. S. Nuclear Regulatory Commission, dated January 30, 2004, and referred to herein as "Document." Information contained in this Document has been classified by FANP as proprietary in accordance with the policies established by FANP for the control and protection of proprietary and confidential information.

4.

This Document contains nformation of a proprietary and confidential nature and is of the type customarily held in confidence by FANP and not made available to the public.

Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.

5.

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

6.

The following criteria are customarily applied by FANP to determine whether information should be classified as proprietary:

(a)

The information reveals details of FANP's research and development plans and programs or their results.

(b)

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

(c)

The information Includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for FANP.

(d)

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

(e)

The information is vital to a competitive advantage held by FANP, would be helpful to competitors to FANP, and would likely cause substantial harm to the competitive position of FANP.

7.

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

8.

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

9.

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

SUBSCRIBED before me this o?'

day of 2004.

I V

Ella F. Carr-Payne NOTARY PUBLIC, STATE OF VIRGINIA MY COMMISSION EXPIRES: 8/31/05 ELLA F. CARR-PAYNE Notary Public C

commonwealth of Vlrgina I

MY Commision Evs.Aug. 31,2005

MOX QA Plan Non-Proprietary Ver.

lof 8 January 29, 2004 Quality Plan for European Fabrication of Mark-BW/MOX Lead Assemblies

- NON-Proprietary Version -

1.0 Scope

Cogema has been selected by DCS as the European fabrication vendor for the four (4),

MOX Lead Assemblies. Cogema's scope consists of the following tasks:

1.) Receive US government-furnished, polished, plutonium dioxide powder 2.) Blend it with depleted uranium dioxide powder, using the MIMAS process 3.) Press the blended powder into MOX pellets 4.) Load the sintered and ground MOX pellets into fuel rods 5.) Pressurize and weld the fuel rod assemblies 6.) Load the fuel rod assemblies into, FANP designed, Advanced Mark-BW/MOX fuel assemblies.

Requirements for performing these tasks are specified in FANP specifications, provided in the MOX Lead Assembly technical data file (TDF)..

At the Cadarache facility, Cogema will receive 125 kg of US Government furnished, polished plutonium oxide powder from the Los Alamos National Laboratory (LANL). This plutonium oxide powder was polished by LANL, under a QA program audited and approved by FANP. Cogema will blend this powder with Cogema-procured depleted U02, using the proprietary"MIMAS" process of blending, milling, and sieving, which will be qualified for this application. Cogema will proceed to press, sinter, and grind the MOX pellets to meet FANP-provided MOX design requirements. The MOX pellets will be loaded into FANP-supplied M5TM cladding tubes and then be seal welded at Cadarache using qualified and approved welding procedures. Completed MOX fuel rod assemblies will be shipped from Cadarache to Cogema's MELOX fabrication facility for assembly into four (4) MOX lead assemblies.

At the MELOX facility, Cogema will assemble the MOX lead assemblies, using Advanced Mark-BW structural components, provided by FANP, into fuel bundles using qualified tooling and procedures approved by FANP, which meet the MOX lead fuel assembly design requirements supplied by FANP. Cogema will also provide all traceability and design records from the fabrication with a certificate of conformance to allow FANP to certify the MOX lead assemblies for delivery to Duke Power.

2.0 Quality Plan:

Qualification of Cogema as the MOX TEurofab" Lead Assembly fabrication vendor will occur in several phases. These phases consist of System audits conducted prior to production, implementation audits performed just prior to the start of fabrication, once all MOX specific qualifications have been submitted and approved by FANP, and on-going surveillances of production operations during the actual fabrication campaign. These audits and surveillances will be conducted and documented as described in the Framatome-ANP (FANP) Fuel Sector Quality Management Manual (FQM Revision 1, US Version - Applicable July 2003.

4 MOX QA Plan Non-Proprietary Ver.

2of 8 January 29, 2004 2.1 System Audits:

The Cogema Cadarache and MELOX fabrication plants will be audited for compliance with 1 OCFR5O Appendix B, prior to the start of production. This audit will be a system audit, on site at each facility, to review the quality systems in place at each facility in accordance with 1 OCFR5O Appendix B and the Framatome ANP QA Manual. The tentative schedule for these system audits at Cadarache and MELOX is shown in Table

1. The system audits will be performed by FANP-Fuel America (FANP-FA) auditors, with the participation of FANP QA personnel from FANP Fuel France (FANP-FF). After successful resolution of any audit findings, the Cogema facilities will be placed on FANP-FA's Approved Suppliers List (ASL). (Note: Cogema - Cadarache and MELOX are currently on the ASL for FANP-FF).

2.2 Implementation Audits:

After successful completion of the FANP Quality systems audit for compliance with 10CFR50, App. B, and subsequent inclusion on FANP-FA's ASL, each Cogema facility will undergo an 'Implementation Audit against specific requirements in the FANP ordering documents and in compliance with the DCSICogema fabrication contract. This audit will verify that specific work instructions, procedures, etc. in place for these tasks, impose the FANP requirements and are followed, such that the product produced will meet all design and quality requirements.

2.3 Surveillance

A third aspect of the FANP-FA quality plan for MOX lead assembly fabrication involves in-process surveillance of manufacturing processes and inspections FANP will perform surveillances of various production and inspection operations at both Cogema facilities to assure that Cogema is using approved processes, personnel, and materials and that the product is meeting design requirements. These surveillances will utilize both FANP-FA and FANP-FF quality personnel. FANP-FF personnel are intimately familiar with the Cogema operations and processes, and once trained in details of the Mark-BW/MOX design specifics and the FANP-FA quality system, will be utilized to provide the weekly surveillance of the fabrication operations FANP-FA quality auditors will periodically review the Cogema activities and the activities of the FANP-FF surveillance personnel.

At MELOX, during bundling of the MOX lead assemblies, FANP-FA auditors along with other FANP-FA technical personnel will be on site to perform the surveillances 2.4 Documentation Review Another part of FANP quality surveillance is review of documentation required by the MOX lead assembly technical data file (TDF).. The MOX technical data file and ordering documents require various documents to be generated by Cogema and approved by FANP prior to use for MOX fabrication. These documents include such items as; process qualification reports, various manufacturing/inspection process procedures, applicable visual standards, certification files, required archive files, and non-conformance reports. FANP auditors will verify that these documents have been generated and have been approved by FANP per applicable design requirements specified in the TDF.

MOX QA Plan Non-Proprietary er.

3of 8 January 29, 2004 TABLE -1 FRAMATOME ANP - EUROFAB CAMPAIGN QUALITY PLAN FOR MOX PELLETS, RODS, AND ASSEMBLIES ITEM PRODUCT INSPECTION CHARACTERISTICS SAMPLING FREQUENCY METHOD TIME

, (0/1)_.

CADARACHE 1

Pellets and System Audit Compliance with N/A Prior to On Site Audit February 2004 Rods 10CFR50 Appendix B Production 2

Pellets and Implementation Ordering Document N/A Pellet/Rod Verify Work Tentative for Rods Audit Requirements Overlap Instructions October 2004 impose Framatome requirements and are followed.

3 Pellets and Surveillance Visual and Dimensional As Specified Pellet/Rod Visual exam and

_ Tentative for Rods Inspection Inspection required by below Overlap measuring October 2004 Quality Plan

~ ~~~See N ote 2 PELLETS 4

Pellet Document

- Qualification

[

]

[

[

Review Plan

]

Report 5

Pellet Laboratory

- Material/ Chemical

[

]

Requirements

]

]

]

6 Pellet Laboratory

- Fissile Content C

]

l C

]

7 Pellet Laboratory

- Pellet Physical C

]

C C

Analyses__

]

8 Pellet Dimensional

-Pellet Dimensional C

C C

C

.I MOX QA Plan Non-Proprietary Ver.

4of8 January 29, 2004 TABLE -1 FRAMATOME ANP - EUROFAB CAMPAIGN QUALITY PLAN FOR MOX PELLETS, RODS, AND ASSEMBLIES ITEM PRODUCT INSPECTION CHARACTERISTICS SAMPLING FREQUENCY METHOD TIME Attributes

]

9 Pellet Visual

- Pellet Visual exam

[

[

1

]

l

]

1 10 Pellet Document

- Certification Report N/A Initial Lot of Review Pellets Certification

]

Document 11 Pellet Document

- Non Compliance N/A All Deviation Verify Framatome _ [

Reports Approval of External deviations 12 Pellet

Document,

- Process documents N/A Prior to Verify Framatome Tentative for Data, or Required Approval Production Approval of items October 2004 Standards

- Data or standards CADARACHE FUEL RODS 13 Rod Document

- Qualification N/A Prior to Verify Framatome

_ Prior to Release Production Approval of for Production Qualification 14 Rod Laboratory

-Helium C

]

1 1

15 Rod Visual

- Fabrication l

I C

C C

Requirements l

Ml MOX QA Plan Non-Proprietary Ver.

5of 8 January 29, 2004 TABLE -1 FRAMATOME ANP - EUROFAB CAMPAIGN QUALITY PLAN FOR MOX PELLETS, RODS, AND ASSEMBLIES ITEM PRODUCT INSPECTION CHARACTERISTICS SAMPLING FREQUENCY METHOD TIME 16 Rod Visual

- Weld Inspection

[

]

l

[

]

[

(Corrosion) 1 17 Rod Visual

-Internal Defects

[

[

[

[

-Helium Leak l

]

18 Rod Visual

-Corrosion Test

_ [

-Metallographic Exam 1

19 Rod Visual

- Plutonium and

[

[

[

[

Uranium Loading l

- Fuel Stack Inspection

]

]

]

- Presence of Plenum Spring 20 Rod Visual

- Handling & Storage

[

]

_ [

]

21 Rod Dimensional

- Dimensional

[

]

C

]

l Inspection per drawing 22 Rod Document or

- Process Documents N/A Prior to Verify Framatome Tentative for Standards Required Approval Production Approval of items October 2004

I.

MOX QA Plan Non-Proprietary Ver.

6of 8 January 29, 2004 TABLE -1 FRAMATOME ANP - EUROFAB CAMPAIGN QUALITY PLAN FOR MOX PELLETS, RODS, AND ASSEMBLIES ITEM PRODUCT INSPECTION CHARACTERISTICS SAMPLING FREQUENCY METHOD TIME (Oil) 23 Rod Document

- Non Compliance N/A All Deviation Verify Framatome

_ Tentative for Reports Approval of October 2004 External deviations 24 Rod Document

- Certification N/A Initial Lot of Review Tentative for Rods Certification October 2004 Document At MELOX Fuel Assembly 25 Assembly System Audit Compliance with N/A Prior to On Site Audit May 2004 10CFR50 Appendix B Production 26 Assembly Implementation Ordering Document N/A One Verify Work Tentative for Audit Requirements Assembly Instructions December 2004 impose Framatome requirements and are followed.

27 Assembly Surveillance Visual and Dimensional

[

1

[

Inspection Inspections required by

]

]

1 Quality Plan See Note 2 28 Assembly Dimensional

-Dimensional 1

]

C

]

Inspection

]

MOX QA Plan Non-Proprietary Ver.

7of 8 January 29, 2004 TABLE -1 FRAMATOME ANP - EUROFAB CAMPAIGN QUALITY PLAN FOR MOX PELLETS, RODS, AND ASSEMBLIES ITEM PRODUCT INSPECTION CHARACTERISTICS SAMPLING FREQUENCY METHOD TIME (0/1)

(Sampling) 29 Assembly Dimensional

- Dimensional

[

]

1 1

Inspections in-process

]

]

30 Assembly Dimensional

- Final Dimensional 1

]

1

]

Inspections

]

l 31 Assembly Visual

- Spedific Attribute C

CC IC Inspections

]

32 Assembly Visual

- Dimple Joint 1

]

1

]

_ 1 Inspection Samples

]

]

33 Assembly Visual

- Cleanliness 1

]

C

]

34 Assembly Visual

- Handling& Storage C

]

1 1

]

]

35 Assembly

Document,

- Reports Required N/A Prior to Verify operating Tentative for Data, or Production instructions December 2004 Standard require items 36 Assembly Document Non Compliances N/A All Deviation Verify Framatome _ Tentative for Reports Approval of December 2004 External deviations 37 Assembly Document Certification N/A Initial Review Tentative for Assembly Certification December 2004 Document

MOX QA Plan Non-Proprietary Ver.

8of 8 January 29, 2004 TABLE -1 FRAMATOME ANP - EUROFAB CAMPAIGN QUALITY PLAN FOR MOX PELLETS, RODS, AND ASSEMBLIES ITEM PRODUCT INSPECTION CHARACTERISTICS SAMPLING FREQUENCY METHOD TIME (0/1 38 Assembly Records Complete and Legible N/A One Review Tentative for Assembly December 2004 FA = Fuel America (Framatome ANP - Lynchburg Personnel - Trained and/or Qualified)

FF = Fuel France (Framatome - France Personnel - Trained and/or Qualified)

Note 1 - [

Note 2- [

1.

AFFIDAVIT COMMONWEALTH OF VIRGINIA

)

) Ss.

CITY OF LYNCHBURG

)

1.

My name is Gayle F. Elliott. I am Manager, Product Licensing in Regulatory Affairs, for Framatome ANP 'FANP"), and as such I am authorized to execute this Affidavit.

2.

1 am familiar with the criteria applied by FANP to determine whether certain FANP information is proprietary. I am familiar with the policies established by FANP to ensure the proper application of these criteria.

3.

1 am familiar with the "Quality Plan for European Fabrication of Mark-BW/MOX Lead Assemblies" referenced in a letter from H. B. Barron, Duke Energy Corporation, to the U. S. Nuclear Regulatory Commission, dated January 30, 2004, and referred to herein as "Document." Information contained in this Document has been classified by FANP as proprietary in accordance with the policies established by FANP for the control and protection of proprietary and confidential information.

4.

This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by FANP and not made available to the public.

Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.

5.

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

6.

The following criteria are customarily applied by FANP to determine whether information should be classified as proprietary:

(a)

The information reveals details of FANP's research and development plans and programs or their results.

(b)

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

(c)

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

(d)

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

(e)

The information is vital to a competitive advantage held by FANP, would be helpful to competitors to FANP, and would likely cause substantial harm to the competitive position of FANP.

7.

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

8.

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

9.

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

SUBSCRIBED before me this !?

day of a7

,2004.

Ella F. Carr-Payne NOTARY PUBLIC, STATE OF VIRGINIA MY COMMISSION EXPIRES: 8/31/05