ML20205C243
| ML20205C243 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 08/04/1986 |
| From: | Tucker H DUKE POWER CO. |
| To: | Harold Denton, Youngblood B Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20205C249 | List: |
| References | |
| NUDOCS 8608120290 | |
| Download: ML20205C243 (7) | |
Text
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DUKE POWER GOMPANY P.O. BOX 33180 CHARLOTTE, N.C. 28242 HAL B. TUCKER
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333 August 4, 1986 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commissica Washington, D.C.
20555 ATIENTION: Mr. B.J. Youngblood, Project Director PWR Project Directorate No. 4 i
Re: Catawba Nuclear Station, Units 1 and 2 Docket Nos. 50-413 and 50-414 Technical Specifications Amendment i
Dear Mr. Denton:
This letter contains a proposed amendment to the Technical Specifications for Facility Operating License Nos. NPF-35 and NPF-52 for the Catawba Nuclear Station.
The attachment contains proposed changes necessary due to modificationr which are scheduled to be performed during the upcoming Unit I refueling outage (Please note that it is requested that the proposed change to Table 3.7-1 also be applicable to Unit 2).
In addition, the attachment contains a discussion of the justification and safety analysis which was conducted pursuant to 10 CFR 50.91 in support of these proposed changes.
It has been concluded from this analysis that the pro-posed changes do not involve significant hazards considerations.
These amendments are required to be in place prior to startup following the Unit I first refueling. The current Unit I shutdown date is August 29, 1986 and the Unit 1/ Cycle 2 initial criticality date is currently scheduled for November 2, 1986.
This request is a supplement to the June 6, 1986 and July 15, 1986 reload sub-cittals for Catawba Unit 1; therefore no additional fees have been remitted with this submittal.
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Mr. Harold R. Denton August 4, 1986 Page 2 Pursuant to 10 CFR 50.91(b)(1), the appropriate South Carolina State official is being provided a copy of this amendment request.
Very truly yours,
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.se Hal B. Tucker RWO/17/jgm Attachment Dr. J. Nelson Grace, Regional Administrator xc:
U.S. Nuclear Regulatory Commission - Region II 101 Marietta St., NW, Suite 2900 Washington, D.C.
20555 NRC Resident Inspector Catawba Nuclear Station Mr. Heyward Shealy, Chief Bureau of Radiological Health S.C. Dept. of Health & Environmental Control 2600 Bull Street Columbia, S.C.
29201 INPO Records Center Suite 1500 1100 circle 75 Parkway Atlanta, Geogia 30339 American Nuclear Insurers i
c/o Dottie Sherman, ANI Library i
The Exchange, Suite 245 270 Farmington Avenue Farmington, CT 06032 M&M Nuclear Consultants 1221 Avenue of the Americas N e York, NY 10020
Mr. Harold R. Denton August 4, 1986 Page 3 HAL B. TUCKER, being duly sworn, states that he is Vice President of Duke Power Company, that he is authorized on the part of said Company to sign and file with the Nuclear Regulatory Connaission this revision to the Catawba Nuclear Station, 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.
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<W Hal B. Tucker, Vice President Subscribed and sworn to before me this 4th day of August, 1986.
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Notary Public My Commission Expires:
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JUSTIFICATION AND ANALYSIS OF NO SIGNIFICANT HAZARDS CONSIDERATION The changes to Technical Specification Tables 3.6-1 and 3.6-2a are needed in order to include a penetration which is to be added during the Unit 1 first refueling outage.
The change to Technical Specification Table 3.7-1 is necessary in order to allow testing of the main steam line safety valves after the planned replacement of their springs. There are five valves per steam generator, each of which is to get a replacement spring during the upcoming refueling outage. After the modification, the valves are technically inoperable until they can be successfully tested and thus shown to be able to perform their intended safety function. Full system temperature and pressure is necessary to perform the tests.
Therefore, allowance for all main steam line safety valves to be inoperable, for the period of time needed to successfully test them, is necessary.
The proposed change to Technical Specification Table 3.7-1 would add the condition where if four or five of the safety valves are inoperable, the unit may remain in Mode 3.
By having to place the Power Range Neutron Flux High Setpoints at zero percent of Rated Thermal Power, the plant will not be able to go critical.
The proposed changes to the Catawba, Units 1 and 2 Technical Specifications have been reviewed pursuant to 10 CFR 50.91.
The following analysis provides a determination that the proposed amendments do not involve any Significant Hazards Considerations, as defined in 10 CFR 50.92, 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 i
accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.
The proposed amendment, do not involve a significant increase in the probability or consequences of an accident previously evaluated. The operation of the unit will not be affected by the addition of the containment penetration or by the change proposed to Table 3.7-1.
None of the prop 4 ed changes affect the operation of any safety system.
The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed changes will not result in a change to the station design or operation that has not already been evaluated. The addition of the containment penetration has been evaluated under the Nuclear Station Modification process and it has been determined that this does not involve an Unreviewed Safety Question.
The change to Table 3.7-1 would allow up to five main steam line safety valves to be inoperable it the power range trip setpoints are reduced to 0% Rated Thermal Power.
The OPERABILITY of the main steam line safety valves ensures that the Secondary System pressure will be limited to within 110% (1304 psig) of its design pressure of 1185 psig during the most
severe anticipated system operational transient.
The maximum relieving capacity is associated with a Turbine trip from valve wide-open condition coincident with an 4
assumed loss of condenser heat sink (i.e., no steam bypass to the condenser).
Resetting the setpoints will keep the unit from becoming critical, therefore, since the turbine cannot be put on line, there is no possibility of a new or different kind of accident created by this change.
i The proposed changes do not involve a significant reduction in a margin of safety.
The addition of the containment penetration is an administrative change to the Technical Specifications and does not affect the margin of safety built into the plant. The change to Table 3.7-1 does not significantly decrease the margin of safety since the power range trip setpoints must be set at 0% Rated Thermal Power if four or five of the main steam line safety valves per steam line are inoperable.
The unit will be kept from going critical, therefore there will not be a significant decrease in a margin of safety with the change to Table 3.7-1.
Based upon the above discussion, it can be concluded that the proposed changes do not involve a significant hazards consideration.
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JUSTIFICATION AND ANALYSIS OF NO SIGNIFICANT HAZARDS CONSIDERATION The changes to Technical Specification Tables 3.6-1 and 3.6-2a are needed in order to include a penetration which is to be added during the Unit I first refueling outage.
The change to Technical Specification Table 3.7-1 is necessary in order to allow testing of the Main Steam Coded Safety Relief valves after the planned replacement of their springs. There are five valves per steam generator, each of which is to get a replacement spring during the upcoming refueling outage. After the modification, the valves are technically inoperable until they can be successfully tested and thus shown to be able to perform their intended safety function.
Full system temperature and pressure is necessary to perform the tests.
Therefore, j
allowance for all Main Steam Coded Relief valves to be inoperable, for the period of time needed to successfully test them, is necessary.
The proposed change to Technical Specification Table 3.7-1 would add the condition j
where if four or five of the relief valves are inoperable, the unit may remain in Mode 3.
By having to place the Power Range Neutron Flux High Setpoints at zero percent of Rated Thermal Power, the plant will not be able to go critical.
The proposed changes to the Catawba, Units 1 and 2 Technical Specifications have been reviewed pursuant to 10 CFR 50.91.
The following analysis provides a determination that the proposed amendments do not involve any Significant Hazards Considerations, as defined in 10 CFR 50.92.
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 amendment do not involve a significant increase in the probability or consequences of an accident previously evaluated.
The operation of the unit will not be affected by the addition of the containment penetration or by the change proposed to Table 3.7-1.
None of the proposed changes affect the operation of any safety system.
The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed changes will not result in a change to the station design or operation that has not already been evalnated. The addition of the containment penetration has been evaluated under the Nuclear Station Modification process and it has been determined that this does not involve an Unreviewed Safety Question. The change to Table 3.7-1 would allow up to five Main Steam Relief Valves to be inoperable if the power range trip setpoints are reduced to 0% Rated Thermal Power. The OPERABILITY of the Main Steam Line Code Safety valves ensures that the Secondary System pressure will be limited to within 110% (1304 psig) of its design pressure of 1185 psig during the most
- severe anticipated system operational transient. The maximum relieving capacity is associated with a Turbine trip from valve wide-open condition coincident with an assumed loss of condenser heat sink (i.e., no steam bypass to the condenser).
Resetting the setpoints will keep the unit from becoming critical, therefore, since the turbine cannot be put on line, there is no possibility of a new or different kind of accident created by this change.
The proposed changes do not involve a significant reduction in a margin of safety.
The addition of the containment penetration is an administrative change to the Technical Specifications and does not affect the margin of safety built into the plant. The change to Table 3.7-1 does not significantly decrease the margin of l
safety since the power range trip setpoints must be set at 0% Rated Thermal Power if four or five of the Main Steam Coded Relief valves per Steam line are j
inoperable. The unit will be kept from going critical, therefore there will not be a significant decrease in a margin of safety with the change to Table 3.7-1.
Based upon the above discussion, it can be concluded that the proposed changes do not involve a significant hazards consideration.
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