ML20198T228
| ML20198T228 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 06/06/1986 |
| From: | Owen W DUKE POWER CO. |
| To: | Harold Denton, Youngblood B Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20198T233 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, RTR-REGGD-01.097, RTR-REGGD-1.097 GL-82-33, NUDOCS 8606110378 | |
| Download: ML20198T228 (12) | |
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DUKE POWER GOMPANY Powan BurLntsro, Box 30189, GnAmLorTE. N. G. asa4a
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June 6,1986 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Attention:
Mr. B. J. Youngblood, Project Director PWR Project Directorate No. 4 RE:
Catawba Nuclear Station Docket Nos. 50-413 and SC-414 Operating License and Technical Specifications Amendments
Dear Mr. Denton:
Attached are proposed license amendments to the Catawba Nuclear Station, Unit 1 operating license, NPF-35, and two sets of proposed changes to the Catawba Units 1 and 2 Technical Specifications. These changes will be i
required prior to startup following the first refueling outage at Catawba j
Unit 1.
Attachment I contains a proposed amendment to license condition 16 of operating license NPF-35. Attachment 2 contains a proposed amendment to license condition 12(a) of operating license NPF-35. Attachment 3 contains Technical Specification changes that will be required to reflect the j
addition of the Boron Dilution Mitigation System. Attachment 4 contains a Technical Specifications change that will be required to reflect the upgrade of the Unit 1 Reactor Coolant System Power Operated Relief Valves to safety grade. Additional Technical Specification changes as well as the Reload Safety Evaluation Report will be submitted at a later date.
i This request is applicable to the Catawba Unit 1 operating license and the Catawba Units 1 and 2 Technical Specifications. Accordingly pursuant to 10 i
CFR 170.21, a check for $150.00 is enclosed.
Pursuant to 10 CFR 50.91(b)(1) the appropriate South Carolina State official is being provided a copy of this amendment request.
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Very truly yours, bW h
gh Warren H. Owen
'Nf 8606110378 860606 WI,H:slb PDR ADOCK 05000413 P
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Mr. Harold R. Denton, Director June 6, 1986 Page Two xc:
Dr. J. Nelson Grace, Regional Administrator U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 Mr. Heyward Shealy, Chief Bureau of Radiological Health South Carolina Department of Health
& Environmental Control 2600 Bull Street Columbia, South Carolina 29201 INPO Records Center Suite 1500 1100 Circle 75 Parkway Atlanta, Georgia 30339 American Nuclear Insurers e/o Dottie Sherman, ANI Library The Exchange, Suite 245 270 Farmington Avenue Farmington, CT 06032 M&M Nuclear Consultants 1221 Avenue of the Americas New York, New York 10020 Mr. P. H. Skinner NRC Resident Inspector Catawba Nuclear Station
Mr. Harold R. Denton, Director June 6, 1986 Page Four Warren H. Owen, being duly sworn, states that he is Executive Vice President of Duke Power Company; that he is authorized on the part of said Company to sign and file with the Nuclear Regulatory Constission these amendments to the Catawba Nuclear Station Unit 1 operating license NPF-35 and this revision to the Units 1 and 2 Technical Specifications, Appendix A to License Nos. NPF-35 and NPF-52; and that all statements and matters set forth therein are true and correct to the best of his knowledge.
$tY3d Oldi Warren H. Owen, Executive Vice President Subscribed and sworn to before me this 6th day of June, 1986.
tif vt Notary Public My Commission Expires:
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Duke Power Company Catawba Nuclear Station, Unit 1 Proposed License Amendment to Facility Operating License NPF-35 License Condition (16)
Steam Generator Tube Rupture
1.
Requested Amendment Amend Facility Operating License NPF License Condition (16) to read:
Prior to startup following the second refueling outage, Duke Power Company shall subrait for NRC staff review and approval an analysis which demonstrates that the steam generator single-tube rupture analysis presented in the FSAR is the most severe case with respect to the release of fission products and calculated doses. Consistent with the analytical assumptions, Duke Power Company shall propose any necessary changes to Appendix A to this license.
2.
Discussion As a result of the January 1982 Steam Generator Tube Rupture.(SGTR) at the R. E. Ginna Plant, the NRC had questioned the assumptions used_in the safety analysis of a design basis SGTR, including the operator action time assumed in terminating leakage from the primary to the secondary coolant systems, and the qualification of the equipment assumed to be used in the SGTR recovery. The issue is further discussed in supplement 2 to the Catawba Safety Evaluation Report.
In response to these concerns, a subgroup of utilities in the Westinghouse Owners Group ( E0G ) was formed to address the licensing issues associated with an SGTR event on a generic basis.
Based on Duke Power Company's participation in the WOG effort, the Catawba Unit 1 operating license was conditioned to require that agreed upon analyses and NRC Staff approval be made before startup following the first refueling outage.
In December 1984, the subgroup submitted WCAP-10698, SGTR Analysis Methodology to Determine the Margin to Steam Generator Overfill, which presented the development of a design basis SGTR analysis methodology. On May 24, 1985, the subgroup submitted Supplement I to WCAP-10698, Evaluation of Of fsite Radiation Doses for a SGTR Accident, which presented an evaluation of potential offsite doses for a design basis SGTR in the absence of steam generator overfill. On February 28, 1986, the subgroup submitted WCAP-11002, Evaluation of Steam Generator Overfill due to a SGTR Accident.
WCAP-10698 and WCAP-11002 are under review by the Staff. The staff's Safety Evaluation Report on Supplement 1 to WCAP-10698 was transmitted to Mr. A. E. Ladieu, Chairman, SGTR Subgroup by letter dated December 17, 1985 from Mr. Herbert N. Berkow, NRC/0NRR.
Although significant progress has been made in addressing the SGTR issue, additional time is needed for the Staff to complete its reviews of submittals to date.
It is expected that additional plant specific submittals will be needed in order to demonstrate that the generic WOG submittals are applicable to Catawba.
3.
Safety Analysis It is Duke Power Company's conclusion that extension of the completion date for reanalyses of SGTR until startup following the second refueling outage does not involve any adverse safety considerations. Analyses submitted todate indicate:(1) that the operators can respond to a design basis SGTR and perform the required recovery actions to terminate the primary to secondary leakage before steam generator overfill occurs, and (2) that the offsite radiation doses for a design basis SGTR will be less than the allowable limits.
4.
Analysis of Significant Hazards Consideration As required by 10 CFR 50.91, this analysis is provided concerning whether the proposed amendment involves significant hazards considerations, as defined by 10 CFR 50.91.
Standards for determination that a proposed amendment involves no significant hazards considerations are if operation of the facility in accordance with the proposed amendment would not: 1) involve a significant increase in the probability or consequences of an accident previously evaluated; or create the possibility of a new or different kind of accident from any accident previously evaluated; or 3) involve a significant reduction in a margin of safety.
The proposed amendment would not involve a significant increase in the probability or consequences of accidents previously evaluated because the extension of time for resolution of the SGTR issue is administrative in nature and has no effect on cause mechanisms.
The proposed amendment would not create the possibility of a new or different kind of accident than previously evaluated since the time extension applies to the completion of the design basis SGTR accident analysis.
In addition, no changes to margins of safety are involved in extending the completion date.
Based upon the proceeding analysis, Duke Power Company concludes that the proposed amendment does not involve a significant hazards consideration.
Duke Power Company Catawba Nuclear Station, Unit 1 Proposed License Amendment
. to Facility Operating License NPF-35 License Condition (12)(a)
F M
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1.
Requested Amendment Amend Facility Operating License NPF-35 License Condition (12)(a) to read:
Prior to startup following the third refueling outage, Duke Power Company shall upgrade the accumulator tank level or pressure consistent with the guidance of Regulatory Guide 1.97, Revision 2 unless prior approval of an alternate design of these items is granted by the NRC Staff.
2.
Discussion Supplement 1 to NUREG - 0737 - Requirements for Emergency Response Capability (Generic Setter 82-33) included additional clarification regarding Regulatory Guide 1.97, Revision 2.
By letter dated September 26, 1983, Duke Power Company provided the information concerning the expectations to conformance to the regulatory guide.
Pending completion of the Staff's review of the Catawba design for conformance to the guidance of the regulatory guide, the operating license for Catawba Unit I was conditioned to require that modification be completed to provide compliance with the regulatory guide unless the exceptions were reviewed and approved by the staff before startup following the first refueling outage. The items identified were:
(a) Reactor coolant system cold leg water temperature (b) Containment sump water level (c) Residual heat removal heat exchanger outlet temperature (d) Accumulator tank level and pressure (e) Steam generator pressure (f) Containment sump water temperature (g) Chemical and volume control system makeup flow and letdown flow (h) Emergency ventilation damper position (i) Area radiation (j) Plant airborne and area radiation Ms. Elinor G. Adensam's letter of August 6,1985 transmitted a draft Technical Evaluation Report (TER) regarding Catawba's conformance to Regulatory Guide 1.97, Rev. 2.
The TER also requested additional justification for some of the exceptions taken by Duke. By letter dated October 22, 1985, Duke provided the requested information.
In Supplement 5 to the Catawba Safety Evaluation Report, dated February 1986, the Staff approved all of the exceptions except for accumulator level and pressure, requiring that Duke designate either level or pressure as the key variable to be upgraded.
By letter dated March 25, 1986, Duke requested additional technical justification from the Staff in order for Duke to be able to evaluate the merits of the Staff's requirement. Based on the receipt of these responses, it is anticipated that Duke would either begin implementation or pursue the issue further through the NRC's appeal process.
Assuming that Duke ultimately agrees to upgrade either the accumulator level or pressure instrumentation, it is estimated that approximately 23 months lead time would be required for implementation during a refueling i
l outage. This would coincide with the end of cycle 3 refueling outage currently scheduled to begin in January 1989.
3.
Safety Analysis The primary function of the accumulator pressure and level instrumentation is to monitor the pre-accident status of the accumulators to assure that the passive safety system is in a ready state to serve its safety function. Accumulator tank level and pressure are not referenced in any emergency procedure covering design basis events which may cause a harsh environment. No operator actions in these procedures are based on accumulator indications.
It is therefore Duke Power Company's conclusion that extension of the date for upgrading the accumulator pressure or level instrumentation until startup following the third refueling outage does not involve any adverse safety considerations.
Deletion of the items other than accumulator tank level and pressure has no safety implications, since such a change simply removes those items which have been reviewed and approved by the Staff in accordance with the license condition.
4.
Analysis of Significant hazards Consideration As required by 10 CFR 50.91, this analysis is provided concerning whether the proposed amendments involve significant hazards considerations, as defined by 10 CFR 50.91.
Standards for determination that a proposed amendment involves no significant hazards considerations are if operation of the facility in accordance with the proposed amendment would not: 1) involve a significant increase in the probability or consequences of an accident previously evaluated; or 2) create the possibility of a new or different kind of accident from any accident previously evaluated; or 3) involve a significant reduction in a margin of safety.
The proposed amendment would not involve a significant increase in the probability of an accident previously evaluated because the accumulator level and pressure are provided for pre-accident monitoring of the status of the cold-leg accumulators and as such have no effect on cause mechanisms.
The proposed amendment would not create the possibility of a new or different kind of accident than previously evaluated since the design and operation of the unit will not be af fected.
The proposed amendment would not cause a significant reduction in a margin of safety. The extension of time in which to upgrade the accumulator level or pressure instrumentation if necessary, would have no impact on safety margins since the instrumentation is fully qualified for its intended function of pre-accident monitoring of the cold-leg accumulators.
Deletion of the remaining items would have no impact on safety margins since the current qualification of these items has been found acceptable by the Staff as documented in SSER-5.
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ATTACENINT 3 DUKE POWER COMPANY CATAWBA NUCLEAR STATION, UNIT 1 PROPOSED AMENDMENT TO THE TECHNICAL SPECIFICATIONS BORON DILUTION MITIGATION SYSTEM I
O
Discussion of Amendment Request The requested changes to the Technical Specifications would change the specifications for systems that are affected by the addition of the Boron Dilution Mitigation System. Section 15.4.6 of the Standard Review Plan (SRP) requires that at least 15 minutes is available from the time the operator is made aware of an unplanned boron dilution event to the time a loss of shutdown occurs during power operations, startup', hot standby, hot shutdown and cold shutdown. A 30 minute warning is required during refueling. In accordance with the SRP, the NRC staff required Duke Power Company to provide control room alarms to alert the operating staff to boron dilution events in all modes of operation as stated in the Safety Evaluation Report, Section 15.2.4.2.
Prior to the first cycle of operation of each unit, Duke Power Company installed the necessary alarms to satisfy the requirements of the SER.
Duke now plans to install a Boron Dilution Mitigation System to provide automatic action for mitigation of a boron dilution event. This system will be installed during the first refueling outage for each unit. A request for similar changes to the Technical Specifications for Catawba i
Unit 2 will be submitted at the appropriate time.
Technical Specification 4.9.1.3 has been changed to allow isolation of Reactor Makeup water by verifying that valve W-230 will be closed and secured in position. This valve is located upstream from valves NV-231, NV-237, NV-240, NV-241 and NV-244, which were previously used to isolate the Reactor Makeup water supply.
The Boron Dilution Mitigatica System satisfies the requirements of the SER as well as making a positive contribution to the operation of :.he plant.
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JUSTIFICATION AND ANALYSIS OF NO SIGNIFICANT HAZARDS CONSIDERATION The proposed changes to the Technical Specifications are required to ensure proper operation and surveillance of the Boron Dilution Mitigation System at Catawba Nuclear Station, Unit 1.
Since Unit 2 will not be provided with this system until its first refueling outage, it is necessary to provide separate specifications for Units 1 and 2.
The proposed changes would provide the correct specifications required by the addition of the boron Dilution Mitigation System while retaining the current Technical Specifications that will be applicable to Unit 2 only.
10 CFR 50.92 states that a proposed amendment involves no significant hazards considerations if operation in accordance with the proposed amendment would not:
(1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.
The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated as they are necessary to reflect the addition of a system for which credit is taken in the Chapter 15 analysis for Cycle 2 operation.
Similarly, the proposed changes do not create the possibility of a new or different kind of accident from those previously evaluated since the addition of the Baron Dilution Mitigation Sy. stem will not affect plant 4
systems other than those required to mitigati a boron dilution event.
Finally, since the proposed changes are necessary to reflect the addition of a system which will mitigate boron dilution events, the proposed changes do not involve a reduction in a margin of safety.
Based upon.the above discussion, Duke Power concludes that the proposed changes to Technical Specifications do not involve significant hazards I
considerations.
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