ML19309F614
| ML19309F614 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 04/07/1980 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19309F609 | List: |
| References | |
| NUDOCS 8004300174 | |
| Download: ML19309F614 (6) | |
Text
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT N0. 27 TO FACILITY OPERATING LICENSE NO. DPR-66 DUQUESNE LIGHT COMPANY OHI0 EDIS0N COMPANY PENNSYLVANIA POWER COMPANY BEAVER VALLEY POWER STATION, UNIT NO. 1 DOCKET NO. 50-334 Introduction By letter dated February 15, 1980, Duquesne Light Company (the licensee) requested relief fron certain requirements of the 1974 Edition through Summer 1975 Addenda of Section XI of the ASME Boiler and Pressure Vessel Code.
This is in addition to relief that was granted by NRC letter dated December 4,1979.
The licensee also applied in its February 25, 1980 letter for a change to the Appe'ndix A Technical Specification that would eliminate Table 4.4-4 in its entirety.
As discussed below, based on a further revision to The Beaver Valley Inservice Inspection Program as documented in the licensee's letter of February 21, 1980, the Staff is now h agreement that all reference to Table 4.4-4 may be removed from the Technical Specifications as an administrative matter.
In fact, all but the last page of Table 4.4-4 (page 3/4 4-30h) should have been rmoved from the Technical Specifications on December 4, 1979 when the NRC Staff approved the licensee's October 10, 1977 ISI program, since that approved program now addresses the surveillance requirements of that table.
Discussion and Evaluation Request for Relief The licensee requests relief from performing visual and surface or volumetric examination of the reactor vessel closure head cladding and visual examination of the reactor pressurizer, and steam generator vessel cladding.
Code Recuirement Visual and surface or volumetric examination of the reactor vessel closure head cladding shall include at least six patches (each 36.sq. in.) evenly distributed in the closure head.
Visual examination of the reactor vessel cladding shall include at least six patches (each 36 sq.' in.) evenly distributed-in accessible sections of the vessel shell.
The examinations perfonned durinn each inspection interval shall cover 100% of the patch areas.
j
-2 Visual examination of the pressurizer and steam generator vessel cladding shall include at least one patch (36 sq. in.) near each manway in the primary side of the vessel.
Se examinations perfomed during each inspection inter-val shall cover 100% of the patch areas.
The examination of the patch areas in the pressurizer and steam generator may be perfomed at or near the end of the inspection interval.
Licensee Basis for Requesting Relief The design of these vessels accounts for the stress loads to be adequately accomodated by the ferritic base material which is examined volumetrically as required by other examination categories. Additional technical guidance is provided by later editions of Section XI where cladding examinations are no longer required.
Evaluation Examination of the reactor vessel head cladding patches and the pressurizer and steam generator vessel cladding is impractical to perfom because of the relatively high radiation levels present in the areas required to be visually examined.
Other examinations which will be performed on these components will give more meaningful data about the structural acceptability of the components.
These examinations have been found to be suitable alternatives for the visual exant; nations of the vessels cladding and the visual and surface or volumetric examination of the vessel head cladding.
Examination of the reactor vessel cladding is also impractical to perform because of the necessity to remove the fuel and core barrel to accomplish the visual examination required.
Examination Category B-N-1, which the licensee has committed to perform, requires visual examination of the reactor vessel interior in accessible areas above and below the reactor core during normal refueling outages at approximately three-year intervals. This examination in conjunction with the volumetric examinations perfomed on shell and nozzel welds will provide adequate assurance of the structural integrity of the reactor vessel.
The Staff finds that the examinations which will be performed by the licensee on the reactor vessel head, reactor vessel, pressurizer, and steam generator will provide data necessary to determine-the structural integrity of these components and concludes that relief from the required examinations of the components cladding may be granted as requested.
Table 1 lists the specific items for which we find relief should be granted along with the alternative examinations required.
This table supplements Table 1 in The Staff Safety Evaluation dated December 4, 1979 which supports The the relief granted at that time from certain other ASME Code requirements.
page number and note number are selected to allow the table in this report to be incorporated into the table in our December 4, 1979 Safety Evaluation Report.
Addition to the Beaver Valley Unit'No.1 ISI Program - Augmented Program for Reactor Vessel Nozzle Safe Ends The licensee in its February 15, 1980 letter noted that the inspections of the reactor vessel nozzle safe end welds to detemine if corrosion is developing in the sensitized zones had been perfomed in accordance with the Technical Specifications.
Therefore, the Notation to Table 4.4-4 (corrected from 4.4-1)
on page 3/4 4-30h of the Technic-1 Specifications (covering three nozzle inspec-tions) should be deleted.
Howeve", in discussions with the licensee, it was noted that when the Technical Sp. ifications were first drafted (see Safety Evaluation Report (SER-OL) for the Facility Operating License DPR-66 dated October 1974 as supplemented May 1975 and October 1975), the five year duration cited in the Table 4.4-4 Notation was considered sufficient time for the Beaver Valley Unit No. I to have reached its third refueling outage.
Due to a number of technical difficulties, the Unit has only canpleted one operating fuel cycle and is not scheduled to return to power from its first refueling outage until July 1980. The requirements of the SER-OL are not tied to the five year period but rather to the first refueling and "the following three and one-third year interval coinciding in the scheduled inservice inspec-tions".
By letter dated February 21, 1980, the licensee requested the deletion of the Notation to Technical Specification Table 4.4-4 on the basis that the ISI pro-gram will be augmented to include the remaining provisions of the reactor vessel nozzel safe end examinations now specified in that Notation.
The augmented ISI program is to include the following inspections.
1.
At the first refueling outage, dye penetrant inspection shall be performed on all six reactor vessel nozzles and volumetric inspection shall be per-formed on at least two nozzles in the same manner as that conducted in the 1
baseline inspection (completed).
2.
At the second and third refueling outages dye penetrant inspections shall be performed on all six nozzles and volumetric inspections shall be performed on at least two nozzles in the same manner as the baseline inspections so that all six nozzles shall be vole. metrically inspected during the first 3
three refueling outages.
In the event that defects are indicated during this period of operation, the inspection program shall require an evaluation to determine.their significance and shall be reported as required to the Commission pursuant to Technical Speci fi cation 6.9.,1.JB,.c.
In the event that defects are not indicated or considered significant during this period of operdtion, the inspection program shall be performed thereafter in accordance with the requirements of Section XI of the ASME Code.
While we agree that the above augmented program should be included in the Beaver Valley Unit No.1 ISI program, the Notation to Technical Specifica-tion Table 4.4-4 should be mai.ntained in the Technical Specifications pending the examinations following the second and third refueling outages.
TABLE 1 BEAVER VALLEY UNIT 1 INSERVICE INSPECTION oo0 GRAM ASME CODE CLASS 1 COMPONENTS TABLE CODE TABLE
.IWB-2500 APPLICABLE CODE ALTERNATIVE
, IWB-2600-EXAMINATION TO EXAMINATION EXAMINATION ITEM NO.
CATEGORY SYSTEM OR COMPONENT CONSTRUCTION AREA TO BE EXAMINED REQUIREMENT REQUIRED Visual Note 18 Bl.14 B-I-l Reactor Vessel Internal Cladding' Cladding Bl.13 B:1-1 Closure Head Cladding Internal Cladding.
Visual and Surface or Volume Note 18 B2.9
-B-I-2 Pressure Vessel Internal Cladding Visual Note 18 Cladding B3.8 B-I-2 Steam Generator Internal Cladding Visual Note 18 Vessel Cladding
-3b-TABLE 1 NOTES (CONTINUED) 2 18.
General vessel cladding examination will be performed when vessel interiors are accessable. -These examinations will be performed once. in a ten year interval -(consistent with reactor vessel discharge of full core for examination and whenever the pressurizer and steam generator may be made accessable for examination or other repairs).
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Environmental Consideration We have determined that neither the relief requested nor the snendment auth-orizes a change in effluent types or total amoun' nor an increase in power level and will not result in any significant en ironnental impact.
Having made this determination, we have further concluded that the relief and the anendment involve an action which is insignficant from the standpoint of environmental impact and, pursuant to 10. CFR 151.5(d)(4), that an environ-mental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of this action.
Conclusion We have concluded, based on the considerations discussed above, that:
(1) because the relief and the amendment do not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, this action does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by 1
l operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's-regulations and the granting of relief and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Date: April 7,1980 t
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