ML20024H408
| ML20024H408 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 05/24/1991 |
| From: | Creel G BALTIMORE GAS & ELECTRIC CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20024H409 | List: |
| References | |
| NUDOCS 9106030273 | |
| Download: ML20024H408 (4) | |
Text
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13 ALTIMORE GAS AND ELECTRIC CHARLES CENTER
- P.O. BOX 1475
- BALTIMORE, MARYLAND 21203-1475 GroRGE C CRtet hiay 24,1991 VaC E Pe( S cthf NvCitam rhtnGe i a no r e c-. s s U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION:
Document Control Desk
SUBJECT:
Calvert Cliffs Nuclear Power Plant Unit Nos.1 & 2; Docket Nos. 50-317 & 50-318 Reauest for Amendpent: Reactor Vessel Soccimen Withdrawal Schedule
REFERENCE:
(a)
Generic Letter 91-01," Removal of the Schedule for the Withdrawal of Reactor Vessel hiaterial Specimens from Technical Specifications,"
dated January 4,1991 Gentlemen:
The Baltimore Gas and Electric Company hereby requests an Amendment to its Operating License Nos. DPR-53 and DPR-69 for Calvert Cliffs Unit Nos.1 & 2, respectively, with the submittal of the proposed changes to the Technical Specifications.
DESCRIPTION The proposed amendment would revise the Technical Specification for both Units 1 and 2 to remove the schedule for the withdrawal of reactor vessel material specimens.Section II.B.3 of Appendix 11 to 10 CFR Part 50 requires the submittal to, and approval by, the Nuclear Regulatory Commission of a proposed withdrawal schedule for material specimens before impicmentation.
llence, the placement of this schedule in the Technical Specifications duplicates the control on changes to this schedule that have been established by Appendix IL IMCKGROUND Technical Specifications include limiting conditions for operation that establish pressure and
- temperature limits for the reactor coolant system. The limits are defined by Technical Specification figures that provide an acceptable range c,f operating temperatures and pressures for heatup, cooldown, criticality, and inservice leak and hydrostatic testing. These limits are generally valid for a specified number of effective full power years. A program for reactor vessel material surveillance ensures the availability of data to update the inservice operating temperature and pressure limits.
This program assists in fulfilling the requirements of Appendix 11 to Part 50 of Title 10 of the Code of Federal Regulatio: - (10 CFR) to prevent brittle fracture of the reactor vessel.
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Document Control Desk hiay 24, IW1 Page 2 The surveillance requirements associated with these limits specify the withdrawal schedule for the reactor vessel material specimens. Recently, the staff of the U. S. Nuclear Regulatory Commission (NRC) approved a request to remove this schedule from the Technical Specification for another i uclear power plant. This change was requested because the Technical Specincation was a duplicate of requirements within Section 11.B3 of Appendix 11 of 10 CFR Part 50.
Duplication of requirements is unnecessary and provides a basis for removal from the Technical Specification as a line item improvement consistent with the Commission Policy Statement on Technical Specification Improvements. The NRC recognized the generic applicability of this request and issued Generic Letter 91-01 to provide guidance to other licensees for deleting the withdrawal schedule.
REOUESTED Cil ANGE Change pages 3/4 4-23 and 3/4 4-25 of the Unit 1 Technical Specifications, pages 3/4 4-24 and 3/4 4-26 of the Unit 2 Technical Specifications, and page B3!4 4 5 of the associated Unit I and Unit 2 Bases as shown on the marked up pages in Attachments A and B.
SAFINY ANALYSES! JUSTIFICATION Limiting condition for operation (LCO) 3.4.9.1 for the reactor coolant system includes operating limits on pressure and temperature, deGned in Figure 3.4 2. that provide an acceptabic region for operation during heatup, cooldown, criticality, and inservice leak and hydrostatic testing. An associated suneillance requirement, SURVEILLANCE REQUIRENIENT 4.4.9.1.1, addresses the frequency of verifying that operation is within the speciGed limits during these operating conditions.
In addition, a separate surveillance, SURVEILLANCE REQUIREh1ENT 4.4.9.1.2, includes the requirement that reactor vessel material suneillance specimens be removed and examined to determine changes in material pioperties, as required by 10 CFR Part 50, Appendix II, and in accordance with the schedule in Technical Specification Table 4.4-5. The Technical SpeciGcation table and the reference to this table providing the schedule for withdrawal of the reactor vessel material surveillance specimens are requested to be removed from this surveillance requirement.
This surveillance requirement also specifies that the results of these examinations shall be used to update the Technical Specification Ugures for the pressure and temperature operating limits. This requirement is being retained.
The Bases Section for this Technic:.1 Specification is being revised to provide a detailed description of the bases for this LCO and the associated surveillance requirements. These bases state that the heatup and cooldown curves are recalculated when data from the suncillance specimens indicate a change in material properties that exceed the limiting value of those properties that were used to develop the existing pressure and temperature limits. These bases also provide information on the use of the data obtained from material specimens. This information defines the purpose and relationship of this information to the requirements included in the regulations and the American Society of hicchanical Engineers (AShtE) Code. Therefore, the removal of the schedule for specimen withdrawal from the Technical Specification will not result in any loss of clarity related to regulatory requirements of Appendix H to 10 CFR Part 50.
Finally, the Updated Final Safety Analysis Report (UFSAR) will be revised to incorporate the related information in accordance with 10 CFR 50.71. Additional information will be included to incorporate a table of the limiting values of reactor vessel material properties, and to update the description of the methods used to predict the effect of neutron radiation on reactor vessel materials.
The UFSAR currently includes the NRC approved reactor vesset material irradiation surveillance
Document Control Desk May 24,1991 Page 3 schedule in Table 413 of Section 4,1.5.3. This schedule will be updated to reGect future NRC approved changes to the specimen withdrawal schedule.
1)LTERMIN ATION OF SIGNil ICANT ll AZARI)S The proposed change has been evaluated against the standards in 10 CFR 50.92 and has been determined to not involve a significant hazards consideration, in that operation of the facility in accordance with the proposed amendment:
(1)
IVould not irn'ohe a signi.ficant increase in the probability or consequences of an accident previously evaluated.
The removal frem the Technical SpeciGcation of the schedule for the withdrawal of reactor vessel materic.1 surveillance specimens will not result in any loss of regulatory control because changes to this schedule are controlled by the requirements of Appendix 11 to 10 CFR Part 50 which require NRC approval and are maintained in the Updated Final Safety Analysis Report. In addition, the surveillance requirements on pressure and temperature indicate that the specimens must be removed and examined to determine changes in their material properties, and these results must be used to update the pressure and temperature limits. Therefore, no actual change in the required actions would occur and the change would not involve a signincant increase in the probability or consequences of an accident previously evaluated.
(2)
IVould not create the possibility of a new or different type of accident from any accident previously evaluated.
The requested change would not result in any change to the plant design, hardware or procedures for operation. Therefore, the requested change would not create the possibility of a new or different type of accident from any accident previously evaluated.
(3)
IVould not involve a significaru reduction in a margin of safety.
The requested change would neither result in fewer not less frequent removal and examination of the reactor vessel material irradiation surveillance specimens, nor would it result in any change to the required use of the results of the examinations.
Therefore, the pressure and temperature requirements for the reactor coolant system would be maintained in the same manner and the change would not involve a significant reduction in a margin of safety.
SCllEI)ULE This change is requested to be approved and issued by August 30,1991. Ilowever, issuance of this amendment is not currently identiGed as having an impact on outage completion or continued plant operation.
Document Control Desk hiay 24,1991 Page 4 SAFl?IY CONihil'ITEE REV1EW These proposed changes to the Technical Specifications and our determination of significant hazards have been reviewed by our Plant Operations and Safety Review Committee and Off. Site Safety Review Committee, and they have concluded that implementation of these changes will not result in an undue risk to the health and safety of the public.
Very truly yours,g '/$
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STATE OF MARYLAND 1
TO WIT COUNTY OF CALVERT t
I hereby certify that on the 8Y day of
/71t 41
,191, before me, the subscriber, a Notary Public of the State of hiaryland in and for Ca/gr/ duidt/
personally appeared George C. Creel, being duty sworn, and states that he is Vice President of the Baltimore Gas and Electric Company, a corporation of the State of hlaryland; that he provides the foregoing information for the purposes therein set forth; that the statements made are true and correct to the best of his knowledge, information, and belief; and that he was authorized to provide the information on behalf of said Corporation.
MTlNESS my lland and Notarial Scal:
N M
. c/bk Notary Public h!y Commission Expires:
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"/ (/ h L6 ate GCC/ERGlerg/dlm Attachment cc:
D. A. Brune, Esquire J. E. Silberg, Esquire R. A. Capra, NRC D. G. hicDonald, Jr., N RC T. T. h1artin, NRC L E. Nicholson, NRC R. I. NicLean, DNR J. H. Walter, PSC
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