ML15113A378
| ML15113A378 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 09/04/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML15113A377 | List: |
| References | |
| NUDOCS 9709120295 | |
| Download: ML15113A378 (9) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REQUEST FOR RELIEF NO. 97-02 AUGMENTED EXAMINATION OF REACTOR VESSEL OCONEE NUCLEAR STATION, UNITS 1. 2, AND 3 DUKE POWER COMPANY DOCKET NOS. 50-269, 50-270, AND 50-287
1.0 INTRODUCTION
The Technical Specifications (TS) for Oconee Nuclear Station, Units 1, 2, and 3, 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 Code (ASME Code) and applicable addenda as required by Title 10 of the Code of Federal Regulations (10 CFR) Section 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i). Section 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 difficulties 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 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 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 Oconee Nuclear Station, Units 1, 2 and 3, third 10-year inservice inspection (ISI) interval is the 1989 Edition.
Pursuant to 10 CFR 50.55a(g)(6)(ii)(A), the Commission revoked all previous reliefs granted to licensees for the extent of volumetric examinations of reactor vessel shell welds, as specified in Section XI, Division 1, of the ASME Code.
9709120295 970904 PDR ADOCK 05000269 P
-2 The Commission further required that all licensees augment their reactor vessel examination by implementing once, as part of the ISI in effect on September 8, 1992, the Item 81.10 requirements (examine essentially 100 percent of the volume of each shell weld) of the 1989 Edition of the ASME Code.
Under 10 CFR 50.55a(g)(6)(ii)(A)(4), licensees may satisfy the augmented requirements by performing the ASME Section XI reactor vessel shell weld examinations scheduled for implementation during ISI intervals in effect on September 8, 1992. As a result, the licensee is required to submit both an alternative to 10 CFR 50.55a(g)(6)(ii)(A) and a request for relief pursuant to 10 CFR 50.55a(g)(5)(iii), or a proposed alternative pursuant to 10 CFR 50.55a(3),
for the same welds when the licensee obtains less than the required coverage (essentially 100 percent) during the examinations.
By letter dated March 13, 1997, Duke Power Company (the licensee), submitted to the NRC its Augmented Examination of the Reactor Pressure Vessel, Request for Relief No. 97-02 for the Oconee Nuclear Station, Units 1, 2, and 3.
2.0 EVALUATION The 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 Augmented Examination of the Reactor Pressure Vessel, Request for Alternative No. 97-02 for Oconee Nuclear Station, Units 1, 2, and 3. Based on the information submitted, the staff adopts the contractor's conclusions and recommendations presented in the Technical Letter Report.
To comply with the augmented reactor vessel examination requirement of 10 CFR 50.55a(g)(6)(ii)(A), licensees must volumetrically examine essentially 100 percent of each of the Item B1.10 shell welds. In accordance with the regulations, 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 proposes that the ultrasonic examinations performed to the extent possible provide an acceptable level of quality and safety.
At Oconee Nuclear Station, Units 1, 2, and 3, the augmented examination requirements cannot be met for shell welds WR18 and WR34 due to physical restrictions that limit ultrasonic scan coverage. For these two welds, the physical obstructions and/or geometric configuration limited volumetric coverage is 73.4 percent and 43.5 percent, respectively.
Additionally, Oconee, Unit 1, has weld WR17 joining the nozzle belt region to the upper shell cylinder (Units 2 and 3 vessels do not have a weld located at this transition). For weld WR17, the taper angle on the nozzle belt region restricts the volumetric coverage to 49.0 percent.
-3 As a result of the augmented volumetric examination rule, licensees must make a reasonable effort to maximize examination coverage of their reactor vessels. In cases where examination coverage from the inside surface (ID) is inadequate, examination from the outside surface (OD) using automated or manual inspection techniques is a potential option. However, at Oconee Nuclear Station, Units 1, 2, and 3, the design of the reactor building prevents access from the OD. The licensee has attempted to maximize coverage from the inside surface by optimizing transducer arrangements for scanning close to obstructions. Therefore, the staff determined that the licensee has made a reasonable effort to maximize examination coverage using ultrasonic techniques.
The use of radiography as an alternate volumetric examination method is not feasible due to component thickness and geometric configurations. Also, physical barriers both on the vessel OD and ID prohibit access for placement of source, films, etc.
The licensee has examined 49.0 percent, 73.4 percent, and 43.5 percent of the Code-required volume on reactor pressure vessel welds WR17, WR18, and WR34, respectively. These percentages, in combination with the complete volumetric examination of the remaining reactor pressure vessel shell welds and performance of the visual examination of the vessel interior, as required by the Code, should have detected inservice degradation, if present. Therefore, the staff concluded that the licensee's proposed alternative provides assurance of weld integrity and compliance would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
3.0 CONCLUSION
The staff has reviewed the licensee's submittal regarding the augmented reactor pressure vessel examination and concludes that the licensee has maximized examination coverage for the reactor vessel welds and the licensee's proposed alternative provides an acceptable level of quality and safety, and compliance would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Based on the coverages obtained, it can be concluded that degradation, if present, would have been detected. Therefore, the alternative contained in Request for Relief No. 97-02 is authorized pursuant to 10 CFR 50.55a(g)(6)(ii)(A)(5) and 50.55a(a)(3)(ii).
Principal Contributor: Tom McLellan Date:
September 4, 1997
TECHNICAL LETTER REPORT AUGMENTED EXAMINATION OF THE REACTOR PRESSURE VESSEL REQUEST FOR ALTERNATIVE NO. 97-02 FOR DUKE POWER COMPANY OCONEE NUCLEAR STATION, UNITS 1, 2, AND 3 DOCKET NUMBERS: 50-269, 50-270, AND 50-287
1.0 INTRODUCTION
By letter dated March 13, 1997, the licensee, submitted Request for Alternative No. 97-02 regarding the augmented reactor pressure vessel (RPV) examinations required by 10 CFR 50.55a(g)(6)(ii)(A). The Idaho National Engineering and Environmental Laboratory (INEEL) staff has evaluated this request in the following section.
2.0 EVALUATION In accordance with 10 CFR 50.55a(g)(6)(ii)(A), "Augmented Examination of Reactor Vessel", licensees' are required to implement an augmented inspection of the RPV to the provisions in the 1989 Edition of the American Society of Mechanical Engineers (ASME)
Boiler and Pressure Vessel Code,Section XI. Licensees are required to examine "essentially 100%" of the length of all reactor vessel shell welds. The information provided by the licensee in support of the proposed alternative to the regulatory requirements has been evaluated and the basis for disposition documented below.
(A) Request for Alternative No. 97-02 to the Auqmented Reactor Pressure Vessel (RPV)
Examinations per 10 CFR 50.55a(q)(6)(ii)(A)
Requlatory Requirement: In accordance with 10 CFR 50.55a(g)(6)(ii)(A),. all licensees must implement once, as part of the inservice inspection interval in effect Enclosure 2
on September 8, 1992, an augmented volumetric examination of the RPV welds specified in Item B1.10 of Examination Category B-A of the 1989 Edition of the ASME Code,Section XI. Examination Category B-A, Items B1.11 and B1.12 require volumetric examination of essentially 100% of the RPV circumferential and longitudinal shell welds, as defined by Figures IWB-2500-1 and -2, respectively.
Essentially 100%, as defined by 10 CFR 50.55a(g)(6)(ii)(A)(2), is greater than 90%
of the examination volume of each weld.
Requirement for Which Alternative is Requested: The essentially 100% coverage requirement could not be met for the Item B1.10 welds listed in the table below.
Essentially 100% of all other Examination Category B-A, Item B1.10 welds have been examined.
Reactor Vessel Shell Welds Unit 1 Unit 2 Unit 3 1RVP-WR17 2RVP-WR18 3RVP-WR18 1RVP-WR18 2RVP-WR34 3RVP-WR34 1 RVP-WR34 Licensee's Basis for the Proposed Alternative (as stated):
"If licensees make a determination that they are unable to completely satisfy the requirements for the augmented reactor vessel shell weld examination specified in 10 CFR 50.55a(g)(6)(ii)(A); then 10 CFR 50.55a(g)(6)(ii)(A)(5) requires the licensee to submit information to the Commission to support this determination and to propose an alternative to the examination requirements that would provide an acceptable level of quality and safety. The licensee may use the proposed alternative when authorized by the Director of the Office of NRR.
"Attachment 2' provides the calculations documenting the actual amount of ASME Section XI Code required examination coverage obtained. A combination of multiple angles and ultrasonic techniques was used to obtain maximum coverage possible.
The use of an alternate transducer head provided increased coverage through optimum transducer arrangement for scanning close to obstructions. During the ultrasonic examination of the welds shown below, the minimum 90% coverage Included in licensee's submittal but not in this report.
requirement of 10 CFR 50.55a(g)(6)(ii)(A)(2) could not be obtained due to part geometry and actual physical barriers.
"Reactor Pressure Vessel Weld WR1 72 (B01.011.001)
"This weld joins the nozzle belt region to the upper shell cylinder in the reactor vessel core region. The principal limitation for this weld is the presence of a taper at the ID surface starting at the upper edge of the weld and extending up on the nozzle belt section. The taper angle is approximately 18.40. Due to this taper, it was not possible to position the transducer contact head. This taper acts like a wedge and forces the contact head off of the reactor vessel. The amount of coverage achieved was 49.0%.
"Reactor Pressure Vessel Weld WRi 83 (1301.011.003)
"This weld is located between nozzles, 13" below the nozzle centerline. The nozzles themselves form obstructions to 100% examination coverage. Between the inlet nozzles, below the core flood nozzle (2 regions), 25.70 out of 32.50 was scanned. The weld extends out on either side of this region intersecting the inlet nozzle-to-shell welds. Between each inlet nozzle and outlet nozzle (4 regions) 19.60 out of 27.80 was scanned. The weld extends out on either side of this region intersecting the inlet and outlet nozzle-to-shell welds. Additional coverage was lost in these regions due to interference by the outlet nozzle lip. The amount of coverage achieved was 73.4%.
"Reactor Pressure Vessel Weld WR34 4 (1301.021.002)
"Due to the core catcher lugs, the entire circumference of the vessel could not be scanned. Scanning was conducted between each of the 12 lugs. The total obstructed area for each lug is 10.50 for the axial scans, excluding the near surface and 0o scans. This results in a total obstruction of 1260 for the axial scans.
"The total obstructed area for each lug is 14.70 for the circumferential scans, including the axial near surface and 00 scans. This results in a total obstruction of 176.50 for these scans.
"The actual circumferential scan volume examined between the core catcher lugs is limited due to the obstruction of the flow stabilizer stubs and the transition area between the shell and lower head which prevents increased scan coverage by the alternate transducer head. This results in a lower percentage of coverage, (31.7%)
from the straight beam or 00 transducer. Including the circumferential scan 2
Weld WR17 is applicable to Oconee Unit 1 only.
3Weld WR18 is applicable to Oconee Units 1, 2, and 3.
Weld WR34 is applicable to Oconee Units 1, 2, and 3.
coverage from the 450, 600, and 700 transducers, the percentage of coverage increases to 67% for the examination area between the lugs.
"The axial scans were performed from the shell side above the weld using a full node examination method due to obstructions from the flow stabilizer stubs on the lower head below the weld and the transition area between the shell and lower head. The full node examination yielded an average coverage of 95% in the axial direction.
"The partial aggregate coverage from the area examined between the lugs is 74.2%. The actual aggregate coverage reduces to 43.5% when the area obscured by the lugs is included.
"As a result of inspections performed, the 100% requirement has been determined to create a hardship for Oconee Nuclear Station. The reactor vessel 1.10 category welds were examined to the maximum extent practical in accordance with ASME Section V, Article 4, 1980 Edition, Winter 1980 Addenda, and the additional requirements of Regulatory Guide 1.150. To meet the 10 CFR 50.55a(g)(6)(ii)(A)(2) examination coverage requirements, design modifications would be necessary to gain access to the welds in order to obtain complete coverage. The design modifications are impractical due to the vast scope of work that would be required.
Imposition of this requirement would cause a considerable burden on Duke Power with no commensurate safety benefit realized."
Licensee's Proposed Alternative Examinations (as stated):
"In addition to the volumetric examination that has been performed on the Oconee reactor vessels, Duke Power has performed a visual examination of the internals and on the inside of the reactor vessel as required by ASME Section XI, Table IWB 2500-1. The visual examination did not identify any rejectable situations per ASME Section XI acceptance standards.
"The use of radiography as an alternate volumetric examination method is not feasible due to component thickness and geometric configurations. Other restrictions which preclude the use of radiography as an alternative are the use of double wall techniques and physical barriers which prohibits access for placement of source, film, number bands, etc.
"Performing the ultrasonic examination from the outside of the reactor vessel is not a viable option. The design of Oconee's reactor building prohibits access for the equipment and personnel.
"Duke Power Company will continue to perform ultrasonic examinations of all reactor vessel welds to the maximum extent practical in accordance with the requirements of ASME Section V, Article 4, 1989 Edition and Regulatory Guide 1.150, Revision 1, Appendix A. The application of Code Case N-460 will be utilized in all cases where less than 100% but greater than 90% weld coverage is obtained.
In cases where weld coverage of less than 90% is obtained, a request for relief from ASME Section XI Code requirements will be submitted.
"Although the coverage requirements of 10 CFR 50.55a(g)(6)(ii)(A)(2) could not be met, the amount of coverage obtained for these examinations provides an acceptable level of quality and safety. Based on these evaluations, the limited coverage will in no way endanger the health and safety of the general public."
Evaluation: To comply with the augmented reactor vessel examination requirement of 10 CFR 50.55a(g)(6)(ii)(A), licensees must volumetrically examine essentially 100% of each of the item B1.10 shell welds. In accordance with the regulations, essentially 100% is defined as greater than 90% of the examination volume of each weld. As an alternative to the regulations, the licensee proposes that the ultrasonic examinations performed to the extent possible provide an acceptable level of quality and safety.
At Oconee Nuclear Station, Units 1, 2, and 3, the augmented examination requirements cannot be met for shell welds WR18 and WR34 due to physical restrictions that limit ultrasonic scan coverage. For these two welds, the physical obstructions and/or geometric configuration limited volumetric coverage to 73.4%
and 43.5%, respectively. Additionally, Oconee, Unit 1, has weld WR17 joining the nozzle belt region to the upper shell cylinder (Units 2 and 3 vessels do not have a weld located at this transition). For weld WR1 7, the taper angle on the nozzle belt region restricts the volumetric coverage to 49.0%.
As a result of the augmented volumetric examination rule, licensees must make a reasonable effort to maximize examination coverage of their reactor vessels. In cases where examination coverage from the ID is inadequate, examination from the outside surface (OD) using automated or manual inspection techniques is a potential option. However at Oconee Nuclear Station, Units 1, 2, and 3, the design of the reactor building prevents access from the OD. The licensee has attempted to maximize coverage from the inside surface by optimizing transducer arrangements for scanning close to obstructions. Therefore, it is concluded that the licensee has made a reasonable effort to maximize examination coverage using ultrasonic techniques.
The use of radiography as an alternate volumetric examination method is not feasible due to component thickness and geometric configurations. Also, physical barriers both on the vessel OD and ID prohibit access for placement of source, films, etc.
The licensee has examined 49.0%, 73.4%, and 43.5% of the Code-required volume on RPV welds WR1 7, WR1 8, and WR34, respectively. These percentages, in combination with the complete volumetric examination of the remaining RPV shell welds and performance of the visual examination of the vessel interior, as required by the Code, should have detected inservice degradation, if present. Therefore, the licensee's proposed alternative provides an acceptable level of quality and safety.
3.0 CONCLUSION
The INEEL staff has reviewed the licensee's submittal on the augmented reactor pressure vessel examination and concludes that the licensee has maximized examination coverage for the reactor vessel welds. Based on the coverages obtained, it can be concluded that degradation, if present, would have been detected. Therefore, for Request for Alternative No. 97-02, the licensee's proposed alternative provides an acceptable level of quality and safety, and it is recommended that the licensee's proposed alternative be authorized pursuant to 10 CFR 50.55a(g)(6)(ii)(A).