ML18068A718
| ML18068A718 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 10/23/1998 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML18068A717 | List: |
| References | |
| NUDOCS 9811020065 | |
| Download: ML18068A718 (6) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO THE ALTERNATIVE TO 10 CFR 50.55ACg)(6)(iil(A)
AUGMENTED REACTOR PRESSURE VESSEL EXAMINATION
. COMMONWEAL TH EDISON COMPANY QUAD CITIES NUCLEAR POWER STATION. UNITS 1 AND 2 DOCKET NOS. 50-254 AND 50-265
1.0 INTRODUCTION
The Technical Specifications (TS) for Quad Cities Nuclear Power Station, Units 1 and 2, state that the inservice inspection (ISi) 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 (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(6)(g)(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 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 (inqluding supports) shall meet the requirements, except the design and access provisions and the pre-service examination requirements, set forth in the ASME Code,Section XI, "Rules for lnservice 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 10-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) 12 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 Quad Cities Nuclear Power Station, Units 1 and 2, third 10-year ISi interval is the 1989 Edition.
Pursuant to 1 O 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 Commission in support of that determination and a 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 determin.ed to be authorized by law, will not endanger life, 9811020065 981023 PDR ADOCK 05000254 P
PDR ENCLOSURE 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.
Pursuant to 10 CFR 50.55a(g)(6)(ii)(A), a licensee must perform an augmented volumetric examination of essentially 100 percent of each of the Item B 1.1 O shell welds of the Reactor Pressure Vessel (RPV). Essentially 100 percent is defined as greater than 90 percent of the examination volume of each weld. The licensee may submit an alternative pursuant to 10 CFR 50.55a(g)(6)(ii)(A)(5) for welds that have not been inspected essentially 100 percent In addition, all previously granted requests for relief under §50.55a, to licensees for the extent of volumetric examination RPV shell welds specified in Item B 1.1 O of Examination Category B-A, "Pressure Retaining Welds in RPV," in Table IWB-2500-1 of subsection IWB in applicable edition and addenda of Section XI, Division 1, of the ASME Code, during the ISi interval in effect on September 8, 1992, are revoked, subject to the specific modification in
§50.55a(g)(6)(ii)(A)(3)(iv) for licensees that defer the augmented examination in accordance with §50.55a(g)(6)(ii)(A)(3).
By letters dated November 22, 1996, and September 18, 1997, Commonwealth Edison Company (ComEd, the licensee) submitted alternatives pursuant to 10 CFR 50.55a(g)(6)(ii)(A)(5) on augmented RPV Examination for Quad Cities Nuclear Power Station, Units 1 and 2. The licensee provided a response to staffs request for additional information on April 14, 1997, for Unit 1. Subsequent to NRC staffs' telecon dated June 3, 1998, the licensee provided supplemental information on August 21, 1998, for both units of Quad Cities. The staffs evaluation and conclusion are stated below..
2.0* DISCUSSION
'COMPONENT IDENTIFICATION:
Code Class:
Reference:
Examination Category:
Item Numbers:
==
Description:==
ASME Code,Section XI, Class 1 Table IWB-2500-1 B-A B 1. 11 and B 1. 12 Limited volumetric examination of the following RPV shell welds Unit 1 Weld Numbers RPV-CW-C2C3 (Circ.)
RPV-CW-C1C2 (Circ.)
RPV-CW-LHC-1 (Circ.)
RPV-VSC1-77 (Axial)
RPV-VSC1-197 (Axial)
RPV-VSC1-317 (Axial)
RPV-VSC1-55 (Axial)
RPV-VSC2-22 (Axial)
RPV-VSC2-261 (Axial)
% Volumetric Coverage 71 52.7 0
0 0
85.7 85.7 55.6 55.6 RPV-VSC3-77 (Axial)
RPV-VSC3-197 (Axial)
RPV-VSC3-317 (Axial)
Unit 2 Weld Numbers RPV-CW-C2C3 (Circ.)
RPV-CW-C1C2 (Circ.)
RPV-CW-LHC-1 (Circ.)
RPV-VSC1-77 (Axial)
RPV-VSC1-197 (Axial)
RPV-VSC1-317 (Axial)
RPV-VSC2-22 (Axial)
RPV-VSC2-261 (Axial)
RPV-VSC3-77 (Axial)
RPV-VSC3-197 (Axial)
RPV-VSC3-317- (Axial)
RPV-VSC4-60 (Axial)
RPV-VSC4-219 (Axial)
EXAMINATION REQUIREMENT:
22.2 23.2
- 72.
% Volumetric Coverage 54.4 58.5 0
0 0
80.2 5.2 50.1 19.6 14.4 72.4 89.1 78.9 10 CFR 50.55a(g)(6)(ii)(A)(2) states that all licensees shall augment their reactor vessel examinations by implementing the examination requirements for RPV shell welds specified in Item 81.10 of Examination Category B-A, "Pressure Retaining Welds in Reactor Vessel," in Table IWB-2500-1 of Subsection IWB of the 1989 Edition of Section XI, Division I, of the ASME Code, subject to the conditions specified in 10 CFR 50.55a(g)(6)(ii)(A)(3) and (4 ). For the purpose of this augmented examination, essentially 100 percent as used in Table IWB-2500-:1
'means more than 90 percent of the examination volume for each weld. Additionally, 10 CFR 50.55a(g)(6)(ii)(A)(5) requires licensees that are unable to completely satisfy the augmented RPV shell weld examination requirement to submit information to the U.S. Nuclear Regulatory Commission to support the determination, and propose an alternative to the examination requirements that would provide an acceptable level of quality and safety.
LICENSEE'S BASIS FOR RELIEF:.
Pursuant to 10 CFR 50.55a(g)(6)(ii)(A)(5), relief is requested on the basis that the alternate method of examination and the achieved weld coverage provide an acceptable level of quality and safety.
The RPV was examined from the internal surface to the extent practical with an alternate method which is qualified to the highest standard available. Further examination from the inside surface is not practical without disassembly of vessel internal components. For weld areas that could not be examined from the inside surface, ex~mination from the outside surface was evaluated. The exterior vessel surface is covered with permanent'insulation located in close proximity to the ~PV outside surface. The lower exterior vessel surface is also covered with a structural bioshield wall. Access is limited to nozzle and inspection port penetrations in the i.nsulation and bioshield wall. It was determined that supplemental manual examinations from the outside surface were not practical due to the bioshield wall, insulation, and dose considerations.
The RPV shell weld examinations confirmed no unacceptable flaws in the vessel, even though greater than 90 percent coverage was not attained for all RPV shell welds.
Performing additional examinations to achieve the greater than 90 percent coverage presents hardship and would incur unnecessary radiological exposure and requires RPV internal, RPV bioshield, or insulation disassembly. Since the examination results conclude that there are no unacceptable flaws, the underlying objeGtives of the augmented examination requirements have been met. Additionally, a VT-2 examination performed on the RPV during system pressure test per category B-P each refueling outage, provides additional assurance that the integrity of subject welds is maintained.
For these reasons, ComEd requests relief from additional augmented examination of the
LICENSEE'S PROPOSED ALTERNATE EXAMINATION:
The shell welds of the RPV were ultrasonically examined remotely from the inside of the vessel to the extent practical. Personnel performing the examinations were* certified to the requirements of the 1989 Edition of ASME Section XI. The remote examinations were performed in accordance with an examination procedure developed by General Electric. This procedure was demonstrated at the "Performance Demonstration Initiative" (PDI) qualification in accordance with the 1992 Edition, 1993 Addenda of ASME Section XI, Appendix VIII requirements. The procedure does.not comply with ASME Section XI 1989 Edition, paragraph IWA-2232; ASME Section V, Article 4; or NRC Regulatory Guide 1.150 and, as such, is considered an alternative examination technique.
The use of PDI qualified procedures results in a more sensitive examination for the detect.ion of potential flaws than the Code described techniques. The error band of sizing has been established within the limits of Appendix VIII. This examination method's capability to reliably detect flaws in areas of restricted access was satisfactorily demonstrated at PDI Session No. 61-02.
Table 1 presents the weld specific examination coverage achieved and associated interferences. Drawing QC2-0001 shows the weld seams examined by GERIS 2000.
The examination identified 77 flaw indications, all of which were within the applicable acceptance standards of subarticle IWB-3500 of ASME Section XI, 1989 Edition. These indications are concluded to be from small acceptance flaws created during the RPV fabrication.
v
.;. 3.0 EVALUATION The licensee conducted an augmented examination of the RPV shell welds for both units of Quad Cities, pursuant to 10 CFR 50.55a(g)(6)(ii)(A)(2) and was unable to obtain the Code-required 90 percent volumetric coverage of some of the welds identified above due to obstructions in scanning of the welds. The staff, therefore, requested the licensee to provide the following additional information on the subject welds that did not meet the Code examination requirement.
- 1.
A detailed description of the physical constraints that inhibited the inspection.
- 2.
Any alternative examination technique or other information that could be used to expand the scope of inspection to ensure that there are no service induced cracking.
- 3.
Any alternative examination to determine if cracks in the vessel clad that could propagate into the shell welds.
- 4.
The schedule for any additional examinations being planned by the licensee.
The staff has evaluated the information provided by the licensee in response to the staff's request for additional information on the above and the licensee's proposed alternative to the augmented reactor vessel examination for Quad Cities Nuclear Power Station, Units 1 and 2.
To comply with the augmented RPV examination requirements of 10 CFR 50.55a(g)(6)(ii)(A),
licensees must volumetrically examine essentially 100 percent of each of the Item B 1.10 shell welds. Essentially 100 percent is defined as greater than 90 percent of the examination volume of each weld. As an alternative to the regulations, the licensee proposed that the examinations that were performed be deemed to satisfy the augmented reactor vessel examination requirement.
At Quad Cities Nuclear Power Station, Units 1 and 2, the augmented examination was*
performed with the General Electric GERIS 2000 inspection tool from inside the vessel. For Unit 1, the total coverage obtained for the circumferential welds was 54.9 percent and the total coverage obtained for the longitudinal welds was 65.1 percent. Two of the longitudinal welds (RPV-VSC1-77 and RPV-VSC1-197) received no examination coverage. The total combined examination coverage for circumferential and longitudinal welds was 58.8 percent. For Unit 2, out of four circumferential welds, only one weld met the Code-required volumetric coverage of greater than 90 percent. Two welds received an average volumetric coverage of 56.5 percent and one weld received no examination coverage. Also, out of 13 longitudinal welds, only three welds received examination coverage required by the Code. Eight welds received an average examination coverage of 51.3 percent ranging from 5.2 percent to 80.2 percent. Two longitudinal welds were not examined at all.
The staff has determined that based on its review of the information provided by the licensee concerning the examination coverages and associated restrictions, the licensee has performed
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______,.. __________ _ a best-effort examination of the RPV welds and has obtained optimum coverage for subject welds. However, the Boiling Water Reactor Vessel and Internals Project (BWRVIP), a technical committee of the BWR Owners' Group, has prepared a report, "BWR Vessel and Internals Project, BWR Reactor Pressure Vessel Shell Inspection Recommendations (BWRVIP-05),"
that proposed reductions in the scope of inspection of the BWR RPV welds to examine only longitudinal welds. The Nuclear Regulatory Commission has completed the evaluation of this document with its proposed examination requirements. This safety evaluation (SE) presented an assessment of the proposal submitted by the BWRVIP to reduce the scope of ISi of BWR RPV circumferential welds. Based on this assessment, the staff has concluded that the change in risk associated with elimination of BWR RPV circumferential weld examinations is negligible and that the BWRVIP proposal is, therefore, acceptable. Nevertheless, licensee's completed examination of circumferential welds indicate an average volumetric coverage of greater than 50 percent with no service-induced degradation found. Therefore, the licensee's proposed alternative to accept current results of circumferential welds, is acceptable. In regard to the examination coverage for axial welds, the licensee obtained an average volumetric coverage of 65 percent in Unit 1 and 55 percent in Unit 2. The staff further noted that the licensee implemented the "Performance Demonstration Initiative" program based on the criteria of Appendix VIII, ASME Code,Section XI, 1992 Edition, 1993 Addenda, for ultrasonic examination of the reactor yessel which provides a higher degree of reliability for detection and characterization of flaws when compared to the conventional amplitude-based ultrasonic technique required by the applicable Code. However, the results of these examinations did not reveal any service-induced degradation. The staff believes that if any active degradation mechanism were to exist in the subject welds, this amount of examination coverage should have detected it. Therefore, the change in risk associated with the reduced examination coverage, is negligible. The staff concludes that the licensee's proposed alternative to examination proyides an acceptable level of quality and safety.
4.0 CONCLUSION
The staff has reviewed the licensee's submittal on the proposed alternative to the augmented examination of the RPV required by 10 CFR 50.55a(g)(6)(ii)(A) for Quad Cities Nuclear Power Station, Units 1 and 2. The licensee has examined approximately 58.8 percent in Unit 1 and 53 percent in Unit 2 of the required length of welds (including circumferential and longitudinal welds); flaws found were those associated with fabrication, but within acceptable limits requiring no additional evaluation. No service induced degradation was found. The staff noted that reduction in examination coverage was solely due to physical constraints inside the reactor vessel and the licensee conducted a best-effort examination. The staff has determined that the change in risk associated with the reduced examination. coverage is negligible and hence, the licensee's proposed alternative provides an acceptable level of quality and safety.
Therefore, the proposed alternative to the augmented examination of the RPV is authorized, pursuant to 10 CFR 50.55a(a)(3)(i) for Quad Cities Nuclear Power Station, Units 1 and 2.
Principal Contributor: P. Patnaik Dated: October 23, 1998