ML17262A455
| ML17262A455 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 04/24/1991 |
| From: | Mecredy R ROCHESTER GAS & ELECTRIC CORP. |
| To: | Andrea Johnson NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9105010079 | |
| Download: ML17262A455 (7) | |
Text
ACCELERATED DISTSJBUTION DEMONST TION SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
ACCESSION NBR:9105010079 DOC.DATE: 91/04/24 NOTARIZED: NO FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G
AUTH.NAME AUTHOR AFFILIATION MECREDY,R.C.
Rochester Gas 6 Electric Corp.
RECIP.NAME RECIPIENT AFFILIATION JOHNSON,A.R.
Project Directorate I-3 DOCKET 05000244
SUBJECT:
Clarifies util position re two items addressed in QA program I
App B dtd 890720 covering records 6 repts 6 inservice insp program.
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S TITLE: QA Topical Report, Amendment, or Correspondence (Docket/Utxlzty Speci NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72).
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89 EAST AVENUE, ROCHESTER N.Y. 14649-0001 ROBERT C MECREDY Vice President Cinna Nuciear Production April 24, 1991 TElEPHONE AREACODE 7tB 646 2700 U.S. Nuclear Regulatory Commission Document Control Desk Attn:
Allen R. Johnson Project Directorate I-3 Washington, D.C.
20555
Subject:
Quality Assurance Manual.,
Appendix B Clarification R.E. Ginna Nuclear Power Plant Docket No. 50-244
Dear Mr. Johnson:
The purpose of this letter is to clarify Rochester Gas and Electric's (RG&E's) position relative to two items addressed in the Quality Assurance Program, Appendix B, dated July 20, 1989 for R.E.
Ginna Nuclear Power Plant.
The Quality Assurance Program, Appendix B describes RG&E's Inservice Inspection Program for R.E.
Ginna Nuclear Power Plant.
RG&E,'s Quality Assurance
- Program, Appendix B was approved by the Nuclear Regulatory Commission (the Commission),
as documented in the Commission's Safety Evaluation Report. issued August 6, 1990.
Based on an RG&E internal audit and an internally-generated comment, it was determined that two separate areas of our submittal required clarification.
Those two areas are listed below:
1.
Section 1.9.4 of Appendix B deals. with the records and reports and states:
"Reports and records to the-extent required. by the construction.
code and IWA 7520(a)
(1 thru 7) as applicable for the replacement shall be completed for all replacements.."
The intent of Section 1.9.4 was to commit to all requirements of ASME Section XI IWA 7520(a) except Item 8 which requires the use of the NIS-1 and NIS-2 Forms.
The NIS-1 Form is an owner's report used to document inspections and the NIS-2 Form is an owner's report used to document repairs or replacements.
These. forms and other Reporting System Controls are intended. to be replaced. with equivalent and applicable controls prescribed in the Ginna Station Quality Assurance Program.
Paragraph 1.1.3 of the submittal and Relief Request 3
describes the exclusion
- and, administrative alternatives to the use of NIS-1 and NIS-2 Forms.
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lOOKOQQ.'n lieu of the use of these
- forms, governing procedures require that a
90 day report:
be submitted to the NRC following each scheduled outage.
This 90 day report. includes a
summary of the previous year's activities, pertinent information required, by the
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NIS-1 and NIS-2 Forms and statements of certification similar to those included on the NIS-1 and NIS-2'orms.
This format: was used for the 90 day submittal following the 1990 Outage.
The 1990 report was reviewed by NRC staff during routine safety inspections as documented in NRC Inspection Report 50-244/90-12 dated September 5, 1990.
- Further, the Authorized Nuclear Inservice Inspector (ANII) from Hartford Steam Boiler signs the 90 day report. to document ANII's review.
RG&E believes.
that. the use of the 90 day report is consistent with Relief Request 3 and the Commission's review of RG&E's Quality Assurance
- Program, Appendix B.
2.
Section 1.10.3.2 of Appendix.
B deals with the: insulation removal at bolted joints during pressure testing and states.:
"Insulation removal during the VT-2 examination is generally not required, however, in accordance with IWA-5242(a) "systems:
borated for the purpose-of controlling reactivity" shall have insulation removed at bolted connections during. conduct of: the VT-2 examination."
At Ginna, this requirement to remove insulation is considered. to be:
applicable only to those lines in. the primary boric acid flow path, from the boric acid supply and, CVCS. Charging lines to the. Reactor Coolant and the return lines via CVCS Letdown.
Pursuant to IWA-5242(a),
1986 Edition, for other. components, visual examination VT-2 may be conducted without removal. of insulation by examining. the accessible and exposed surfaces and joints of the insulation.
Therefore, removal of insulation. is not applicable to branch lines.
connecting to the primary flow path.
Furthermore, because boric acid is not considered.
corrosive to austenitic stainless or high alloy materials, insulation removal is only applicable to bolted connections using carbon steel or low alloy bolting materials.
- Finally, this requirement is only applicable to those VT-2 examinations performed during a hydrostatic pressure. test, since 1) the
- Leakage, Functional and Inservice. tests, often. done at. high temperature and pressure, are intended to be non-intrusive type
- tests, and
- 2) there is an increased. potential of damage. to heat tracing during removal of insulation which could render the. system inoperable.
As stated earlier, the purpose of this letter is to clarify our intentions relative to RG&E's. Appendix B Program.
RG&E does not expect'
- response, unless further clarification or action. is required.
Very truly yours, GAL/kaw/152 Attachment Robert C. Mecr dy
zc: Mr. Allen R. Johnson (Mail Stop 14D1)
Project Directorate I-3 Washington, D.C.
20555 U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 Ginna Senior Resident Inspector
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