ML17265A534
| ML17265A534 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 02/04/1999 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML17265A533 | List: |
| References | |
| NUDOCS 9902090208 | |
| Download: ML17265A534 (4) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATIONBYTHE OFFICE OF NUCLEAR REACTOR REGULATION ON THE THIRD 10-YEAR INSERVICE INSPECTION INTERVAL RE VEST FOR RELIEF NO. 36 ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR POWER PLANT DOCKET NO. 50-244
1.0 INTRODUCTION
The Technical Specification 4.2.1.5 for the R.
E. Ginna Nuclear Power Plant states that the inservice inspection of the American Society of Mechanical Engineers (ASME) Code Class 1, 2 and 3 components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code (ASME Code) and applicable addenda as required by 10 CFR 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i).
10 CFR 50.55a(a)(3) states that alternatives to the requirements of paragraph (g) may be used, when authorized by the NRC, if (i) the proposed alternatives would provide an acceptable level of quality and safety or (ii) compliance with the specified requirements would result in hardship or unusual difficultywithout a compensating increase in the level of quality and safety.
Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1, 2 and 3 components (including supports) shall meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code,Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," to the extent practical within the limitations of design, geometry, and materials of construction of the components.
The regulations require that inservice examination of components and system pressure tests conducted during the first ten-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(b) twelve months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein. The applicable ASME Code,Section XI, for the Ginna Nuclear Power Plant, third 10-year inservice inspection (ISI) interval is the 1986 Edition. The components (including supports) may meet the requirements set forth in subsequent editions and addenda of the ASME Code incorporated by reference in 10 CFR 50.55a(b) subject to the limitations and modifications listed therein and subject to Commission approval.
Pursuant to 10 CFR 50.55a(g)(5), ifthe licensee determines that conformance with an examination requirement of Section XI of the ASME Code is not practical for its facility, information shall be submitted to the Commission in support of that determination and a 9902090208 990204 PDR ADOCK 05000244 P,
PDR Enclosure request made for relief from the ASME Code requirement.
After evaluation of the determination, pursuant to 10 CFR 50.55a(g)(6)(i), the Commission may grant relief and may impose alternative requirements that are determined to be authorized by law, willnot endanger life, property, or the common defense and security, and are otherwise in the public interest, giving due consideration to the burden upon the licensee that could result ifthe requirements were imposed.
By letter dated December 18, 1998, Rochester Gas and Electric (RG&E) submitted to the NRC its requests for relief on reactor vessel nozzle-to'-safe end welds, Item No. B5.10 of Examination Category B-F of the ASME Code,Section XI, from the Code-required surface examination of essentially 100 percent of each weld (Relief Request (RR)-36) due to inaccessibility. The licensee has proposed alternatives to the Code-required surface examination of the same vessel welds identified in the RR-36 to be conducted for the Ginna plant during the spring 1999 outage.
The licensee's proposed alternative requires a limited surface examination of accessible portions of each weld along with a volumetric examination from the nozzle bore of the reactor vessel.
Pursuant to 10 CFR 50.55a(g)(6)(i), the staff has evaluated the information provided by the licensee in authorizing its proposed alternative stated in RR-36.
2.0 DISCUSSION RELIEF REQUEST RR-36 Alternative to Code-re uired surface examination of RPV nozzle-
"'" " 'I EXAMINATIONRE UIREMENT "Under Category B-F, Item Number B5.10, volumetric and surface examinations shall be performed with essentially 100% of the weld length to obtain code coverage.
ASME Section XI Code Case N-460 states that ifthe entire examination volume or area cannot be examined due to interference by another component or part geometry, a reduction in coverage is acceptable provided that the (lack of) coverage is less than 10%. Previous Codes utilized did not include this 90% coverage requirement and examinations were performed to the extent obtainable."
COMPONENTS FOR WHICH RELIEF IS RE VESTED The relief is requested for six (6) reactor pressure vessel nozzle-to-safe end butt welds.
Inspection of these welds is addressed under Class I, Category B-F, Item Number B5.10, nozzle-to-safe end weld surface examinations as identified below.
Weld Identification Covera e Obtained PL-FW-II PL-FW-v PL-FW-Iv PL-FW-vll AC-1003-1 AC-1002-1 7
70 76%
0% (*)
0% (*)
Note: (*) welds embedded in concrete.
LICENSEE'S BASIS FOR RELIEF "Relief is requested pursuant to the provisions of 10 CFR'50.55a(g)(5)(iii), the required examination coverage for the identified welds is impractical and would require redesign or replacement to obtain code required surface examination coverage.
R. E. Ginna Nuclear Power Plant was designed and constructed to the B31.1, 1955 edition Construction Code. This Code did not contain requirements to ensure that items be accessible for future examinations.
The above noted piping welds were installed utilizing this construction code, which did not provide for accessibility for future ISI non-destructive examination.
Due to the limited design accessibility, ISI surface examination coverage is below the Code percentage requirements as identified within this relief request.
The first four.(4) welds of this relief request are located in a "sand box" configuration. Within the "sand box", the welds are against the floor and one wall. The angled wall is joined to the floor and is against the weld. The surface examination of these welds is limited due to original construction code interferences of the floor and wall of the "sand box". The "sand boxes" would have to be redesigned to enable the welds to be inspected to obtain Code-required coverage for the surface examinations.
The last two welds of this relief request are embedded in concrete.
This concrete structure is the wall that surrounds the reactor pressure vessel.
ASME Section XI Class 1 system leakage examinations are performed. These leakage examinations demonstrate pressure boundary integrity and provide additional assurance in maintaining plant safety."
LICENSEE'S PROPOSED ALTERNATIVEEXAMINATION "R. E. Ginna Nuclear Power Plant proposes that the surface examination coverage identified for the first four (4) welds above be acceptable in fulfillingthe Code required examination coverage.
The actual physical configuration of the "sand boxes" is not conducive in obtaining the requirements specified within Code Case N-460 for acceptable coverage.
Volumetric examination of these welds is performed from the inside of the vessel, and willbe performed during the 1999 outage.
For the last two (2) welds, the Code surface examination requirements are impractical and cannot be examined due to them being embedded in concrete.
Volumetric examination of these welds is performed from the inside of the vessel, and willbe performed during the 1999 outage."
3.0 EVALUATION The staff finds that the design and construction Code for Ginna plant did not contain provisions for accessibility of welds for inservice inspection since the Code for inservice inspection of welds (ASME Code,Section XI) did not exist at the time of the design of the plant. The licensee has stated that the locations of four out of the six nozzle-to-safe end welds are in areas where each weld is located against a floor and at an angled wall. Therefore, the Code-
4 required surface examination of each weld is impractical due to limited access.
The remaining two welds are embedded in concrete and are inaccessible.
In order to perform the Code-required surface examination, the floor and the walls where the welds are located, need to be redesigned which imposes a burden on the licensee.
The licensee has to performed examination of at least 70% of the surfaces of four welds, but could not perform any surface examination for the remaining two welds that are embedded in concrete.
Alternatively, the licensee has proposed a full volume ultrasonic examination of each of the six welds from inside the nozzle bore which would be capable of detecting surface flaws, ifany, in the weld.
The proposed alternative provides reasonable assurance of structural integrity for the subject welds.
4.0 CONCLUSION
In regard to licensee's request for relief RR-36, the staff has evaluated the licensee's proposed alternative of a best-effort surface examination of four out of six nozzle-to-safe end welds along with a full volume ultrasonic examination of each of the six welds from the reactor vessel nozzle bore. The requirements of the applicable ASME Code,Section XI; in regard to surface examination of the subject welds is impractical due to lack of access and would cause burden on the licensee ifthe Code requirements were imposed.
The proposed alternative examination by the licensee, however, provides a reasonable assurance of structural integrity and, therefore, pursuant to 10 CFR 50.55a(g)(6)(i), relief is authorized and the alternative imposed for the R.E. Ginna Nuclear Power Plant for the third 10-year inservice inspection interval. The relief granted is authorized by law and willnot endanger life or property or the common defense and is otherwise in the public interest giving due consideration to the burden upon the licensee that could result ifthe requirements were imposed on the facility.
Principal Contributor:
P. Patnaik Date:
February 4, 1999