ML20059E743
| ML20059E743 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 08/30/1990 |
| From: | Tucker H DUKE POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20059E748 | List: |
| References | |
| NUDOCS 9009100302 | |
| Download: ML20059E743 (7) | |
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Duke herr Carvany M N h'*"
y a y 33;pg lice hesident Charlotte, N C 2824y Nuclear Production r
(704)373 453) l DUKEPOWER August 30, 1990 U.S. Nuclear Regulatory Commission ATTN: ')ocument Control Desk Washington, D.C.
20555
Subject:
McGuire Nuclear Station, Docket Nos. 50-369 and 50-370 i
Proposed Technical Specification Amendment to Revise the Reactor Vessel Pressure-Temperature Limits i
Purr.uant to 10 CFR 50.90, this letter contains a proposed amendment to the Technical Specifications (TS) for McGuire Nuclear Station (Facility Operating License Nos. NPF-9 and NPF-17). This proposed amendment request socks the following revisions:
1.
Replace the existing Unit 1 Reactor Coolant (NC) system heatup and cooldown limit curves with new curvest 2.
Replace the existing Unit 2 Reactor Coolant (NC) system heatup cnd cooldown limit curves with new curvest and, 3.
Revise the reactor vessel surveillance capsulu withdrawal schedule.
T Attachment No. 1 provides a technical discussion, no significant hazards analysis, and environmental impact analysis supporting these proposed revisions. The proposed changes to the station TSs in the form of pen and ink marked pages, and replacement curves are provided in Attachment No. 2.
Attachment No. 3 provides the supporting analysis report prepared by Westinghouse on behalf of Duke Power Company.
Based on Generic Letter 88-11, the existing Unit 2 pressure-temperature limits TS was approved for the first five effective full power years of operation, et until the end of fuel cycle six. Fuel cycle six will end with the beginning of the scheduled September 1990 refueling outage. This outage is currently scheduled for approximately 72 days. Based on this schedule.
please provide NRC approval of the proposed amendment by November 1, 1990.
Pursuant to 10 CFR 50.91(b)(1), a copy of this amendment request has been provided to the appropriate North Carolina official.
Very truly yours, chi A$.- /)
Ital B. Tucker SEL559 O h,,,. c,
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U.S. Nuclect R:gul ttry Commicsion ATTN: Document Control Desk August 30, 1990 Page 2
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Mr. S. D. Ebneter, Administrator U. S. Nuclear Regulatory Cemmission Region II i
101 Marietta Street, NW, Suite 2900
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Atlanta, Georgia 30323 l
Mr. Dayne Brown, Chief Radiation Protection Branch Division of Facility Services Department of Human Resources 701 Barbour Drive
. Raleigh, N.C.
27603-2008 Mr. D.S. Hood, Project Manager Office of Nuclear Reactor Regulation, USNRC i
Washington, D.C.
20555
'l' Mr. P.K. Van Doorn NRC Senior Resident Inspector McGuire Nuclear Station l
tir. T. Reed, Project Manager i
Office of Nuclear Reactor Regulation, USNRC Washington. D.C.
20555 3
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U.S. IMc1ctr R:gulcttry Commicsion ATTN: Dw eent Control Desk August 30, 1990 Page 3 j
l ITAL 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 U.S Nucicar Regulatory Commission this revision to the McGuire Nuclear Station Technical Specifications License Nos. NPF-9 and NPF-17: and, that all statements and matters set forth therein are true and correct. to the best of his knowledge.
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Ab.' A $. W k Jfal B. Tucker, Vice President. '
Subscribed and sworn to before me this thirtieth day of August 1990.
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- U.S. Nuclear Regulatory Commiccion ATTN
- Document Control Desk l
August 30, 1990 Attachment No. 1 I
Duke Power Company i
McGuire Nuclear Station Technical Discussion No Significant Hazards and Environmental Analysis
-l Technical Discussion
-This purpose of the proposed amendment request for McGuire Nuclear Station is to 1.
Replace the existing Unit 1 ReattGr Coolant (NC) system heatup and cooldown limit' curves with new curves (TS Figures 3.4-2 and 3.4-4);
2.
Replace the existing Unit 2-NC system Heatup and cooldown limit curves-with new curves (TS Figures 3.4-3 ~and 3,4-5); and, 3.
Revise the reactor vessel surveillance capsule withdrawal schedule (TS I
Table 4.4-5).
By letters dated January 22, 1989 and supplemented May 17, and June 19, 1989. Duke Power Company (DPC) proposed TS-amendments to'the McGuire Unit 1
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heatup and cooldown limit curves.
This proposal was approved as transmitted by NRC letter dated July 18,~1989 (Amendment Nos.~ 100-and 82). The existing curves are based upon a Westinghouse Report entitled, "McGuire. Unit l' Reactor Vessel Heatup > 4 Cooldown Limit Curves for Normalt Operation" dated November 1988 and provi*. as Attachment No. 5 of our. January 22, 1989 submittal. Additionally, the associated maximum cooldown rate during normal.
operation was decreased from 100 degrees-F.to 60 degrees-F.
Althoughtthe i
previously mentioned Westinghouse analysis demonstrated the 100 degrees-F
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maximum cooldown rate was acceptable. DPC requested the decrease to 60' degrees-F based upon an associated administrative-cooldown limit. %ha continue to have an administrative cooldown limit of 60 degrees-Fi however, at this time we request the cooldown limit. curve (TS Figure 3.4-4) be..
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revised to utilize the 100 degree-F maximum cooldown rate during normal l
operation as supported by the Westinghouse analysis. This submittal also
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proposes a revision to the Unit 2 heatup and cooldown limit curves based the last Unit 2 reactor surveillance capsule (X) analysis which-resulted-in a i
maximum cooldown rate during normal operation ofL100 degrees-F. By revising the Unit 1 curve to 100 degrees-F we will maintain consistency between Unit 1 and 2, and thereby reduce the probability of human performance errors with respect to the use of the cooldown curves.
The method for predicting radiation embrittlement in the Westinghouse L
analysis is based upon Regulatory Guide 1.99, Revision 2; " Radiation Embrittlement of Reactor Vessel Materials" and as before the new Unit.1 heatup and cooldown limit curves contain margins of 10 degrees-F and 60 psig for possible instrument ere and are applicable for a service period of up to ten effective full power years (EFPY). !
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U.S. Nuclear Regulatory Commission ATTN Document Control Desk August 30, 1990 Attachment No. 1 l
The Unit 1-heatup limit curve (fS Figure 3.4-2) is being. revised to correct data at the bottom of the curve that describes the. material _ basis for che curve.- The actual curve has not been revised, only the material basis data which is described in the supporting Westinghouse analysis previously hentioned. When this curve was last revised, the data was incorrectly updated.
For McGuire Unit 2, the propoaed revision to the heatup and cooldown curves (TS Figures 3.4-3 and 3.4-5) is needed to reflect adjustments to existing limits based upon analysis of the last Unit-2 reactor vessel surveillance-i capsulo (X). The supporting Westinghouse analysis (WCAP-12556) is provided
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in Attachment No. 3.
As before, the new curves contain margins of 10 1
degrees-F and 60 psig for possible instrument error.
The new curves are-applicable for 10 EFPY. The method fer predicting radiation embrittlement In the Westinghouse report, and the development of the new curves is: based i
upon Regulatory Guide 1.99, Revision 2.
As with the Unit 1 curves, the i
maximum cooldown rate during normal operation is being changed from 60 degrees-F to 100 degrees-F per the capsule analysis. liowever, please note that the McGuire administrative limit of 60 degrees-F also applies to Unit 2.
The Unit 2 reactor vessel surveillance capsule withdrawal _ schedule has been revised to reflect new capsule withdrawal. dates and the Unit 2 lead factors have been updated. 'The lead factors represent the relationship between the fast neutrons density at the capsule location and the inner wall of the pressure vessel, and are used to predict future radiation damage to the pressure vessel material. These lead factors have been revised to be consistent with current Westinghouse analysis and to meet NRC requirements.
The TS Bases for these TSs have been reviewed and do not need to be revised.
No Significant llazards Analysis 10 CFR 50.91 requires that the following analysis be provided-concerning whether the proposed amendment request involves a significant hazards y
consideration as defined in 10 CFR 50.92.
Standards for determination that i
an amendment request does not involve a significant hazards consideration 4
are if operation of the facility in accordance with the proposed amendment vould 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 previously.
evaluated; or, 3) Involve a significant reduction in a margin of safety.
The proposed revision to replace the existing Unit-1 NC system cooldown curve with a new curve (Figure 3.4-4) that allows a maximum cooldown rate of 100 degrees-F is consistent with the Westinghouse analysis provided by Attachment No. 5 of our letter to the NRC dated January 22, 1989. l
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- U.S.-Nuclear Rsgulatory Commisaien ATTN:-Document Control Desk August 30,-1990 Attachment No. 1 The new curve is based on the results of the reactor vessel surveillance capsule analysis which was performed in accordance with NRC methodology q
contained in Regulatory Guide 1.99, Revision 2 " Radiation Embrittlement of' Reactor Vessel Materials". The revision to correct the material basis data on the Unit I huatup limit curve is an administrative change to' correct a previous revirica error. This revision is also based on data provided in the capsula analysis. Implementation of the proposed Unit 1 cooldown'11mit curve and the corrected heatup limit curve in accordance with the proposed amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated,'nor will it create the possibility of a new or different kind of accident from any previously evaluated, nor will a margin in safety te reduced because the curves are _
based on analysis performed in compliance with Regulatory Guide 1.99, Rev 2.
Compliance with this NRC approved methodology when calculating the new pressure-temperature limits ensures reactor vessel integrity is maintained.
Shifting the cooldown'11mit to reflect the change in material toughness due tn irradiation using NRC approved methodology meets the requirements of 10 CFR 50 Appendix G. " Fracture Toughness Requirements"'to provide adequate margins of safety during any condition of normal operation, including operational occurrenres and system hydrostatic tests, to which the pressure
'puu. dary may be subjected over its service lifetime..
The proposed' revision to replace the existing Unit 2 NC system heatup and cooldown limit curves with new curves (Figures.3.4-3 and 3.4-5) incorporates new pressuro-temperature operating limits based on the Unit 2 reactor surveillance capsule _(X) analysis. These curves are conservative with respect to the existing pressure-temperature operating limits. The new curves reficct the change in material toughness of the reactor vessel due to irradiation effects. Additionally, the. curves are based on the results of-reactor surveillance capsule analyses performed byLWestinghouse using_ the NRC approved methods described in Regulatory Guide'1.99, Revision-2.
Implementation of the proposed curves in accordance with_the. proposed amendment would not involve a significant increase in the probability.or consequences of an accident previously evaluated,-nor will it create the possibility of a new or different kind of accident from any previously evaluated, nor will a margin in safety be reduced,because the' curves are based on analysis performed in compliance with Regulatory Guide 1.99, Rev 2.
Compliance with this NRC approved methodology when calculating the new pressure-temperature limits ensures reactor vessel integrity is maintained.
i Shifting the heatup and cooldown limit to' reflect t?e change in material tougnness due to irradiation using h20 aporo" a metlodology meets the.
requirements of 10 CFR 50 Appendix'G " Fracture Touginess Requirements" to provide adequate %argins of safety during any condi'. ion of normal operation, including oporr,cional occurrences and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime.
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U.S. Nuclear Regulatory Commission-
-ATTN Docueent Control Desk.
August 30, 1990-Attachment No. 1 4
.The proposed revision to TS Table 4.4-5 will update the reactor vessel surveillance capsule withdrawal schedule which-identifies the capsules already removed and future capsule withdrawal dates. The table has also bcon updated-to inclu6. '.he Unit 2 lead factors based-on the Westinghouse analysis of the Unit 2 reactor surveillance capsule-(X). Updating the-capsule removal schedule would not involve a significant increase in the probability or consequences of an accident previously evaluated, nor will it create.the possibility of a new or different kind of accident'frcm any l
c' previously evaluated, nor will a' margin in safety be reduced because this
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revision complies with ASTM E185-82, " Standard Practice _for Conducting Surveillance Tests for Light Water Cooled Nuclear Powered Reactor Vessels" jl g
which' is ~ approved by NRC as described in 10 CFR 50 Appendix 11. " Reactor Vessel Material Surveillance Program Requirements".
As. discussed above, we have determined the-proposed amendment request does not involve a significant hazards consideration as defined by 10 CFR 50.92.
Environmental Impact Analysias i
The proposed TS amendment has been reviewed against the criteria of 10 CFR j
51.22 for environmental considerations. The proposed amendmenu does=not involve a significant hazards consideration, nor increase the types and amounts of effluents that may be released offsite, nor increasu individual or cumulative occupational radiation exposures.
Thereftfe, ths proposed TS t
amendment meets the criteria given in 10 CFR 51.22(c)(9) for a categorical-exclusion from the requirement for an Environmental Impact Statement.
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