ML17264B112
| ML17264B112 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 11/21/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML17264B111 | List: |
| References | |
| NUDOCS 9711250277 | |
| Download: ML17264B112 (14) | |
Text
~R RE00
~o Cy F
n C
O IVl e
0 Y/
~O
++*++
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCL AR REACTOR REGULATION OF THE THIRD EN-YEAR INTERVAL INSERVICE INSPECTION OGRAM PLAN RE UEST FOR RELI F NO 1
~FO ROCHESTER GAS AND ELECTRIC CORPORATION R.
E.
GINNA NUCLEAR POWER PLANT DOCKET NO. 50-244
1.0 INTRODUCTION
The Technical Specifications (TSs) for R.
E. Ginna Nuclear Power Plant, Unit 1, state 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 (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'sed, 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 difficulty without 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 pre-service examination requirements, set forth in the ASHE 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 edition of Section XI of the ASME Code for the current Ginna Nuclear Plant, ten-year inservice inspection (ISI) interval is the 1986 Edition.
Pursuant to 10 CFR 50.55a(g)(5), if the 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 97fiQ~p~77 psppppgg PDR gDQCK 0
@DR P
Enclosure
Commission in support of that determination and a request made for relief from the ASHE 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, will not 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 if the requirements were imposed.
By letter dated April 9,
- 1997, Rochester Gas and Light Corporation (licensee),
submitted Request for Relief No.
17 from the ASHE Section XI requirements for the Ginna Nuclear Power Plant.'he licensee provided additional information in its letters dated Hay 9, 1997 (Revision 1), July 8,
- 1997, and July 18, 1997.
- 2. 0 EVALUATION The NRC staff, with technical assistance from its contractor, the Idaho National Engineering and Environmental Laboratory (INEEL), has evaluated the information provided by the licensee in support of its Third Ten-Year Inservice Inspection Interval Program Plan Request for Relief No.
17 for Ginna Nuclear Power Plant;.
Based on the results of the review, the staff adopts'he
, contractor's conclusions and recommendations presented in the Technical Letter Report (TLR) attached.
Relief Request No.
17 (Rev 1):
ASME Code, Table IWB-2500-1, Examination Category B-J, Item B9. 10, Reactor Coolant Cast Pump Terminal End-to-Cast Elbow Circumferential Welds and Associated Longitudinal Seam Welds requires 100X surface and volumetric examinations on pressure retaining welds in piping 4" nominal piping size or larger each inspection interval as defined by Figure IWB-2500-8.
These examinations are to be performed on essentially lOOX of the weld length for circumferential welds, and include at least one pipe-diameter
'length but not more than 12 inches of each longitudinal weld intersecting the selected circumferential welds.
Code Case N-460 specifies that a reduction in examination coverage on any Class 1 or Class 2 weld may be accepted provided the reduction in coverage for that weld is less than lOX.
Pursuant to 10 CFR 50.55a(g)(5)(iii), the licensee has requested relief from volumetrically examining, to the extent required by the Code, the following Reactor Coolant Cast Pump Terminal End-to-Cast Elbow Circumferential Welds and associated Longitudinal Seam Welds:
PL-FW-XIII PL-FW-XII-LR (XIII)
PL-FW-XV PL-FW-XI-LR (XV)
Elbow-to-Pump "A" weld Longitudinal seam weld Elbow-to-Pump "B" weld Longitudinal seam weld In the licensee's July 8, 1997 submittal, the licensee deleted six welds from Relief Request No. 17.
Therefore, this relief request evaluation applies only to the four Examination Category B-J welds listed above.
The Code requires 100X volumetric examination of the subject Reactor Coolant Pump-to-Cast Elbow welds.
However, the acoustical properties of the cast stainless steel components (A-351 Gr.
CFSH) (i.e., the highly attenuative characteristics of the austenitic grain structure) make ultrasonic examination coverage to the extent that is specified by the Code, and further defined in Code Case N-460 impractical.
In the letter dated July 8, 1997, the licensee stated that an. optimized ultrasonic technique utilizing a refracted longitudinal wave is being used to examine the subject welds, and that the following coverages are expected:
Welds PL-FW-XII-LR XIII 8 PL-FW-XI-LR XV :
The licensee stated that for these two longitudinal welds, they are expecting
<90X coverage based on the uncertainty of the sound penetrating the weld or following the weld fusion line to the ID surface.
Welds PL-FW-XIII & PL-FW-XV:
The licensee stated that these two elbow-to-pump circumferential welds will receive a one sided examination.
Coverage of approximately 50X will be obtained with the axial scan, and 75X with the circumferential scan.
The staff has reviewed the licensee's submittals and determined that the licensee is performing a best effort examination using state-of-the-art techniques.
Due to the highly attenuative characteristics of the austenitic grain structure, ultrasonic examination coverage to the extent required by the Code is impractical.
Compliance with the Code would require that the affected components be redesigned or replaced, thus causing a burden on the licensee.
The limited volumetric examinations, along with the Code-required surface examinations and system leakage tests, provides a reasonable assurance of continued structural integrity.
Therefore, the licensee's request for relief is granted pursuant to 10 CFR 50.55a(g)(6)(i).
3.0 CONCLUSION
The staff has reviewed the licensee's submittals and concludes that certain inservice examinations cannot be performed to the extent required by the Code at R.
E. Ginna Nuclear Power Plant.
For Relief Request No. 17, the licensee has demonstrated-that the Code coverage requirements are impractical.
In addition, imposition of the requirement would result in a burden upon the licensee.
Based on the Code-required surface examinations, the system leakage
- tests, and the best effort volumetric examinations using state-of-the-art techniques, the staff concludes that the licensee's proposed alternative provides reasonable assurance of continued structural integrity for the subject system components.
Therefore, Request for Relief No.
17 is granted pursuant to 10 CFR 50.55a(g)(6)(i).
The relief granted and alternative imposed are authorized by law and will not endanger life or property or the common defense and security and are otherwise in the public interest given due
consideration to the burden upon the licensee that could result if the requirements were imposed on the facility.
Attachment:
Technical Letter Report Principal Contributor:
T. McLellan Date:
TECHNICAL LETTER REPORT THIRD 10-YEAR INSERVICE INSPECTION INTERVAL RELIEF REQUEST NO. 17 FOR ROCHESTER GAS AND ELECTRIC CORPORATION R. E. GINNA NUCLEAR PLANT UNIT 1 DOCKET NUMBER 60-244
1.0 INTRODUCTION
By letter dated April 9, 1997, Rochester Gas and Electric Corporation (RGSE) submitted Relief Request No. 17.
Relief Request No. 17 (Rev. 1), was submitted by RGRE on May 9, 1997.
Following the initial review of these documents, additional information regarding Relief Request No. 17 was requested by the Nuclear Regulatory Commission (NRC) in a letter dated June 3, 1997.
The licensee responded to the NRC request for additional information in a letters dated July 8, 1997, and July 18, 1997.
The Idaho National Engineering and Environmental Laboratory (INEEL) staff has evaluated the licensee's submittals in the following section.
2.0 EVALUATION The third 10-year inservice inspection (ISI) interval for R. E. Ginna Nuclear Plant, Unit 1, ends in December 1999.
The Code of record for the third 10-year interval is the 1986 Edition of Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code.
The information provided by the licensee in support of the request for relief has been evaluated and the basis for disposition is documented below.
2.1 Relief Re uest No. 17 Rev 1
Examination Cate or B-J Reactor Coolant Cas Pum Terminal End-to-Cast Elbow Circumferential Welds and Associated Lon itudinal Seam Welds ATTACHMENT
Code Re uirement: Table IWB-2500-1, Examination Category B-J, Item B9.10, requires 100% surface and volumetric examinations on pressure retaining welds in piping 4" NPS or larger each inspection interval as defined by Figure IWB-2500-8.
These examinations are to be performed on essentially 100% of the weld length for circumferential welds, and include at least one pipe-diameter length but not more than 12 inches of each longitudinal weld intersecting the selected circumferential welds.
Code Case N-460 specifies that a reduction in examination coverage on any Class 1 or Class 2 weld may be accepted provided the reduction in coverage for that weld is less than 10%.
Licensee's Code Relief Re uest: Pursuant to 10 CFR 50.55a(g)(5)(iii), the licensee has requested relief from volumetrically examining, to the extent required by the Code, the following Reactor Coolant Cast Pump Terminal End-to-Cast Elbow Circumferential Welds and associated Longitudinal Seam Welds:
c$:: ". " ~:i::8ikWeld:IDN':c.'8(:;.jap'@44'Q;:.~@;.Pi;.a,Type'.:j's'~Y>rip" gg"":~ <'~
PL-FW-XIII PL-FW-XII-LR(XIII)
PL-FW-XV PL-FW-XI-LR (XV)
Elbow-to-Pump "A"weld Longitudinal seam weld Elbow-to-Pump "B" weld Longitudinal seam weld Note:
In the licensee's July 8, 1997 submittal, the licensee deleted six welds from Relief Request No. 17. Therefore, this relief request evaluation applies only to the four Examination Category B-J welds listed above.
Licensee's Basis for Re uestin Relief (as stated):
"Supplement 4 (b)(4) of Appendix III identifies ultrasonic scanning requirements on Austenitic and Dissimilar Metal Welds.
This paragraph further identifies that Cast items such as fittings, valve bodies, and pump casings may preclude meaningful examinations because of geometry and attenuation variables.
"At R. E. Ginna Nuclear Power Plant, the Reactor Coolant Pump, Westinghouse Model 93, is cast stainless (A-351 Gr CFSM). The associated fittings (elbows) are also cast stainless (A-351 Gr CFSM) and contain longitudinal seam welds."
... "When employing optimized ultrasonic techniques on these welds, the techniques may detect large flaws (25% or greater through wall) but are relatively ineffective in detecting smaller flaws. Therefore, the sensitivity is less than that required by the Code.
"Cast Stainless base metal and associated welds contain large grain structures, attenuation variables impact on performance of ultrasonic examination.
Experience has shown that these materials are not always amenable to ultrasonic examination and does not produce reliable and meaningful results.
Currently, the industry's Performance Demonstration Initiative (PDI) is not addressing Cast Stainless Ultrasonic Examinations.
"Due to the highly attenuative characteristics of the austenitic grain structure, ultrasonic examination coverage to the extent that is specified within Code Case N-460 may not always be achievable.
"Radiography is an impractical technique to use and, if applied, the alternative volumetric examinations are not expected to provide any meaningful increase in benefit beyond the alternative presented due to the high levels of background radiation emitting from these areas."
Licensee's Pro osed Alternative (as stated);
"None. Applicable Code-required volumetric examination will be completed to the maximum extent practical (a best effort ultrasonic examination of the cast stainless welds based on state-of-the-art techniques and associated achievable examination coverage)."...
"We willcontinue to evaluate new emerging inspection technology as they become available.
The code required surface examinations and system leakage tests will be performed."
Evaluation: The Code requires 100% volumetric examination of the subject Reactor Coolant Pump-to-Cast Elbow welds.
However, the acoustical properties of the cast stainless steel components (A-351 Gr. CFSM) make the Code coverage requirements impractical for these welds.
Due to the highly attenuative characteristics of the austenitic grain structure, ultrasonic examination coverage to the extent that is specified by the Code, and further defined in Code Case N-460, is not always achievable.
In the letter dated July 8, 1997, the licensee stated that an optimized ultrasonic technique utilizing a refracted longitudinal wave is being used to examine the subject welds, and that the following coverages are expected:
a
~ Welds PL-FW-XII-LR XIII 5 PL-FW-XI-LR XV: The licensee stated that for these two longitudinal welds, they are expecting (90% coverage based on the uncertainty of the sound penetrating the weld or following the weld fusion line to the ID surface.
~ Welds PL-FW-XIII & PL-FW-XV: The licensee stated that these two elbow-to-pump circumferential welds willreceive a one sided examination.
Coverage of approximately 50% will be obtained with the axial scan, and 75% with the circumferential scan.
The INEEL staff has reviewed the licensee's submittals and determined that the licensee is performing a best effort examination using state-of-the-art techniques.
Due to the highly attenuative characteristics of the austenitic grain structure, ultrasonic examination coverage to the extent required by the Code is impractical.
The limited volumetric examinations, along with the Code-required surface examinations and system leakage tests, will provide reasonable assurance of the continued structural integrity.
3.0 CONCLUSION
The INEEL staff has reviewed the licensee's submittals and concludes that certain inservice examinations cannot be performed to the extent required by the Code at R. E. Ginna Nuclear Plant, Unit 1.
For Relief Request No. 17, the licensee has demonstrated that the Code coverage requirements are impractical.
Based on the Code-required surface examinations, the system leakage tests, and the best effort volumetric examinations using state-of-the-art techniques, reasonable assurance of the continued structural integrity will be provided.
Therefore, it is recommended that for Relief Request No. 17 relief be granted pursuant to 10 CFR 50.55a(g)(6)(i).
Dr. Robert C. Mecredy ~
Vice President, Nuclea~perations-.
Rochester Gas and Electric Corporation 89 East Avenue Rochester, NY 14649
SUBJECT:
REQUEST FOR RELIEF NO.
17 FROM THE AMERICAN SOCIETY OF MECHANICAL ENGINEERS (ASME)
CODE SECTION XI REQUIREMENTS FOR THE R.
E.
GINNA NUCLEAR POWER PLANT-THIRD TEN-YEAR INSERVICE INTERVAL (TAC NO. M98377)
Dear Or. Mecredy:
By letters dated April 9,
- 1997, as supplemented May 9, 1997, July 8,
- 1997, and July 18,
- 1997, you requested'elief from the surface and volumetric examinations on pressure retaining welds in selected piping 4" nominal pipe size or larger since an acceptable Code examination is impractical.
The NRC staff concludes that certain inservice examinations cannot be performed to the extend required by Code at the R.E.
Ginna Nuclear Power Plant.
For Relief Request No.
17, the licensee has demonstrated that the Code coverage requirements for the R.
E. Ginna Nuclear Power Plant are impractical.
The staff has determined that relief be granted and imposes an alternative inspection method pursuant to 10 CFR 50.55a(g)(6)(i) which is authorized by law and will not endanger life or property, or the common defense and security and is otherwise in the public interest, giving due consideration to the burden that could result if the requirements were imposed on the facility.
The staff's evaluation and conclusions are contained in the enclosed safety evaluation.
Therefore, the subject request for relief from the Code requirement is granted for the third ten-year inspection interval.
Sincerely, Docket No. 50-244
Enclosure:
Safety Evaluation cc w/encl:
See next page DISTRIBUTION:
S. Singh Bajwa, Director Project Directorate I-l Division of Reactor Projects I/II Office of Nuclear Reactor Regulation Docket File PUBLIC PDI-1 R/F B. Boger S.
Bajwa S. Little G. Vissing T. Harris (SE only)
OGC G. Hill (2 copies)
ACRS SECY B. McCabe C. Hehl, Region I P. Patniak T. McLellan DOCUMENT NAME:
G:)GINNA)M98377.REL
- See Previous Concurrence To receive a copy of this document, indicate in the box:
"C" Copy without attachment enclosure "E"
Co ith attachment enclosure "N" - No co OFFICE NAME DATE PM G
g/rsI 11/ 0 /97 E
LA:PDI-1 SLIttI 11/'7 D:PDI-1 Sgajva 11/
/97 Officia Recor Copy MYoung 11/05/97
0
~
1