ML060550473
| ML060550473 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 02/24/2006 |
| From: | Terao D Plant Licensing Branch III-2 |
| To: | Edington R Nebraska Public Power District (NPPD) |
| Benney B, NRR/DLPM, 415-3764 | |
| References | |
| TAC MC8512 | |
| Download: ML060550473 (13) | |
Text
February 24, 2006 Mr. Randall K. Edington Vice President-Nuclear and CNO Nebraska Public Power District P.O. Box 98 Brownville, NE 68321
SUBJECT:
COOPER NUCLEAR STATION RE: RELIEF REQUESTS RI-21, REVISION 2, AND RI-36 AND WITHDRAWAL OF RELIEF REQUEST RI-20 (TAC NO. MC8512)
Dear Mr. Edington:
By letter dated September 28, 2005, as supplemented by letter dated December 8, 2005, Nebraska Public Power District submitted requests for relief from certain inservice inspection (ISI) code requirements, and the withdrawal of Relief Request RI-20 for the third 10-year interval ISI program at the Cooper Nuclear Station.
Nebraska Public Power District requested the Nuclear Regulatory Commission (NRC) staff review and approval of Relief Requests RI-21, Revision 2, and RI-36. Relief Requests RI-21, Revision 2, and RI-36 address the volumetric examination coverage for certain reactor pressure vessel nozzle welds. You requested to withdraw Relief Request RI-20 because it is no longer applicable after the subject weld has been selected for the risk-informed ISI program.
Based on the information submitted, the NRC staff concludes that Relief Requests RI-21, Revision 2, and RI-36 are acceptable, and Relief Request RI-20 may be withdrawn.
Sincerely,
/RA/
David Terao, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-298
Enclosure:
Safety Evaluation cc w/encl: See next page
February 24, 2006 Mr. Randall K. Edington Vice President-Nuclear and CNO Nebraska Public Power District P.O. Box 98 Brownville, NE 68321
SUBJECT:
COOPER NUCLEAR STATION RE: RELIEF REQUESTS RI-21, REVISION 2, AND RI-36 AND WITHDRAWAL OF RELIEF REQUEST RI-20 (TAC NO. MC8512)
Dear Mr. Edington:
By letter dated September 28, 2005, as supplemented by letter dated December 8, 2005, Nebraska Public Power District submitted requests for relief from certain inservice inspection (ISI) code requirements, and the withdrawal of Relief Request RI-20 for the third 10-year interval ISI program at the Cooper Nuclear Station.
Nebraska Public Power District requested the Nuclear Regulatory Commission (NRC) staff review and approval of Relief Requests RI-21, Revision 2, and RI-36. Relief Requests RI-21, Revision 2, and RI-36 address the volumetric examination coverage for certain reactor pressure vessel nozzle welds. You requested to withdraw Relief Request RI-20 because it is no longer applicable after the subject weld has been selected for the risk-informed ISI program.
Based on the information submitted, the NRC staff concludes that Relief Requests RI-21, Revision 2, and RI-36 are acceptable, and Relief Request RI-20 may be withdrawn.
Sincerely,
/RA/
David Terao, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-298
Enclosure:
Safety Evaluation cc w/encl: See next page DISTRIBUTION:
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OFFICE LPL4/PM LPL4/LA CFEB/BC OGC LPL4/BC NAME BBenney LFeizollahi KGruss JZorn DTerao DATE 2/24/06 2/24/06 02/14/06 02/23/06 2/24/06 Official Record Copy
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUESTS RI-21, REVISION 2, AND RI-36 AND WITHDRAWAL OF RELIEF REQUEST RI-20 THIRD -YEAR INSERVICE INSPECTION INTERVAL COOPER NUCLEAR STATION NEBRASKA PUBLIC POWER DISTRICT DOCKET NO. 50-298
1.0 INTRODUCTION
By letter dated September 28, 2005, as supplemented by letter dated December 8, 2005, Nebraska Public Power District requested staff review and approval of Relief Requests RI-21, Revision 2, and RI-36, and the withdrawal of Relief Request RI-20 for the third 10-year interval inservice inspection program at the Cooper Nuclear Station. By letter dated October 23, 1997, the NRC had previously approved Relief Request RI-21, Revision 1, with associated Table RI-21-1, Revision 1, for the third 10-year inspection interval to allow limited coverage for the volumetric examination of welds due to restrictions and accessibility, based on the examinations performed in the second 10-year inspection interval (Agencywide Documents and Access Management System Accession Number 9711030046).
Relief Requests RI-21, Revision 2, and RI-36 address the volumetric examination coverage for certain reactor pressure vessel nozzle welds, as required by the American Society of Mechanical Engineers (ASME) Code. The licensee requested to withdraw Relief Request RI-20 because it is no longer applicable after the subject weld has been selected for the risk-informed inservice inspection program.
2.0 REGULATORY EVALUATION
The inservice inspection of the ASME Code Class 1, Class 2, and Class 3 components is to be performed in accordance with the ASME Code Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," and applicable edition and addenda as required by Title 10 of the Code of Federal Regulations (10 CFR), Part 50 (specifically 10 CFR 50.55a(g)), except where specific relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i).
Paragraph 10 CFR 50.55a(a)(3) states, in part, that alternatives to the requirements of paragraph (g) may be used, when authorized by the Nuclear Regulatory Commission (NRC), if the applicant demonstrates either that: (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.
Paragraph 10 CFR 50.55a(g)(6)(i) states that...The Commission will evaluate determinations under paragraph (g)(5) of this section that code requirements are impractical. The Commission may grant such relief and may impose such alternative requirements as it determines 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 upon the licensee that could result if the requirements were imposed on the facility...
Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1, 2, and 3 components (including supports) will meet the requirements set forth in the ASME Code Section XI 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) twelve months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein. The inservice inspection code of record for the third 10-year inservice inspection interval at the Cooper Nuclear Station is the ASME Code,Section XI, 1989 edition, with no addenda.
3.0 RELIEF REQUEST RI-21, REVISION 2
3.1 ASME Code Components Affected
The affected components are reactor pressure vessel nozzle-to-vessel welds, NVE-BD-N9 and NVE-BD-N4A&C, in the control rod drive return line and feedwater line, respectively. They are Class 1 welds and follow the requirements of Examination Category B-D, Full Penetration Welded Nozzles in Vessels - Inspection Program B, Item Number B3.90, in Table IWB-2500-1 of Section XI of the ASME Code.
3.2 Applicable Code Edition and Addenda
The code of record for the third 10-year inservice inspection interval is ASME Code,Section XI, 1989 edition, with no addenda.
3.3 Applicable Code Requirement
ASME Code,Section XI, Table IWB-2500-1, Examination Category B-D, Item B3.90, requires 100 percent volumetric examination of the reactor vessel nozzle-to-vessel welds as defined by Figure IWB-2500-7(a).
3.4 Licensees Proposed Alternative and Basis for Use Pursuant to 10 CFR 50.55a(g)(5)(iii), the licensee has determined that compliance with the ASME Code requirements of achieving essentially 100-percent coverage of the welds listed in Table RI-21-1 of Relief Request RI-21, Revision 1, is impractical due to the configuration of the nozzles, the design of the vessel insulation support rings and the nozzle access hatches, interferences from thermocouple pads, and instrument lines. In lieu of performing the code-required examinations, the licensee proposes to examine the accessible portions of reactor vessel nozzle-to-vessel welds to the extent practical.
The licensee stated that examinations conducted in the third 10-year inspection interval met or exceeded the coverage requirements for the nozzle-to-vessel welds listed in Table RI-21-1, Revision 1, with the exception of weld NVE-BD-N9. Examinations performed on weld NVE-BD-N9 achieved 40 percent coverage as compared to 70 percent required by Relief Request RI-21, Revision 1. The basis for Relief Request RI-21, Revision 2, therefore, is to change the examination coverage of weld NVE-BD-N9 from 70 percent to 40 percent (40.99 percent actual) as well as remove the reference to the inner radius.
The licensee examined weld NVE-BD-N9 based on the Performance Demonstration Initiative (PDI) in accordance with ASME Code,Section XI, Appendix VIII, 1995 edition with editions and addenda through 2000. The primary reason for reduction in coverage is due to differences in procedure and equipment qualifications as a result of the PDI, as compared to requirements for examinations before the PDI was implemented. For example, the requirement to use only PDI-qualified transducers limited the examination coverage and contributed to a reduction of coverage as compared to previous examinations in the outer 85-percent volume.
While the total coverage achieved during the third interval examination was 40 percent, the inner 15 percent of the weld volume received 100-percent coverage with the overall reduction of coverage present in the outer 85 percent. The licensee noted that the examination of the inner radius of weld NVE-BD-N9 in the third 10-year inspection interval achieved 100 percent of the required examination volume. Therefore, the licensee determined that relief is not required for the inner radius examination (Item B3.100). Based on Electrical Power Research Institute (EPRI) modeling of the examination volume as guidance for the field examination and demonstrated qualification of the examiners through the PDI, a higher quality of examination was achieved even though less overall coverage was achieved.
Because relief is not required for the inner radius examination of nozzle NVE-BD-N9, the licensee proposed to remove the words "and inner radius" from the "N9" entry in proposed Table RI-21-1, Revision 2, for Nozzle NVE-BD-N9. In addition, the table has been administratively updated to reflect Instrument Nozzles N11A&C in the Access Restrictions column for Nozzle Numbers N4A&C.
3.5 Duration of Proposed Alternative
The licensee requested that the proposed alternative be applicable for the remainder of the third 10-year interval of the inservice inspection program, which ends on February 28, 2006.
3.6 Staff Evaluation The proposed relief request, RI-21, Revision 2, with associated Table RI-21-1, Revision 2, is to update the examination coverage of nozzle NVE-BD-N9 and access restriction of nozzle NVE-BD-N4A&C, which the staff approved in Relief Request RI-21, Revision 1, in the October 23, 1997, letter. By letter dated November 21, 2005, the staff requested that the licensee provide additional information to clarify the technical basis of the relief request. The licensee responded by letter dated December 8, 2005. The licensee stated that examination of weld NVE-BD-N9 was first performed in the first 10-year inspection interval in 1985 with no coverage percentage reported. The second examination of the weld was performed in the second 10-year inspection Interval in 1993 with 70-percent coverage. Both examinations predate the requirement to use PDI techniques. No recordable indications were detected in the weld during those inspections.
The licensee stated that, in the third 10-year inspection interval, the total volumetric examination coverage of weld NVE-BD-N9 was 40 percent, which was based on the ultrasonic examination volume of the outer 85-percent thickness and the inner 15-percent thickness as measured from the inside diameter surface. The 60-degree refracted longitudinal scans were used to examine the entire weld volume with the exception of the lower 15 percent in the circumferential direction. The EPRI Model parameters were used to inspect the lower 15 percent of the examination volume using the required shear wave scans. The 60-degree refracted longitudinal scans, along with the combined model scans, were used to calculate the total examination coverage.
In its November 21, 2005, letter, the NRC staff asked the licensee why the examination of the inner radius and the inner 15-percent of the nozzle NVE-BD-N9 achieved 100-percent coverage, but only 40-percent coverage was achieved for the overall nozzle volume. By letter dated December 8, 2005, the licensee responded that the nozzle is divided into two distinct examination zones: the nozzle-to-shell weld zone (ASME Code Item B3.90) and the inner radius zone (ASME Code Item B3.100). Each zone for the NVE-BD-N9 nozzle has two distinctive component identifiers in the inservice inspection program. The nozzle-to-shell weld is identified as NVE-BD-N9, and the inner radius zone is identified as NVIR-BD-N9. The nozzle-to-shell weld zone and the inner radius zone each have their own respective code volumes that are required to receive a volumetric examination.
Relief Request RI-21, Revision 1, identified the coverages for both zones on NVE-BD-N9 as 70 percent (i.e., 70 percent for nozzle-to-shell weld zone and 70 percent for inner radius zone).
All the other nozzles identified in Relief Request Table RI-21-1, Revision 1, referred only to the nozzle-to-shell weld zones and did not include the inner radius zone as NVE-BD-N9 did. All the other examinations for the nozzle inner radius zones achieved the required inner coverage.
Therefore, their respective coverages were not listed.
The 40-percent coverage stated in proposed Relief Request RI-21, Revision 2, only applies to the nozzle-to-shell weld zone NVE-BD-N9. Within the required examination volume associated with NVE-BD-N9, that zone or volume is divided into the lower 15-percent region and upper 85-percent region. The licensee stated that 100-percent coverage was achieved for the lower 15-percent region, which is the region of most concern for flaw initiation and that the upper (outer) 85 percent region is less of a concern.
The inner radius zone (NVIR-BD-N9) is a different examination volume than the inner 15 percent region of weld NVE-BD-N9. Different techniques and different data sheets are used for each zone. The licensee stated that in its most recent examination, it achieved the 100-percent coverage of the inner radius zone compared to the 70-percent coverage in the second 10-year inspection interval in 1995. The licensee proposed to remove the words and inner radius from the listing in proposed Table RI-21-1, Revision 2, for nozzle NVE-BD-N9 because it has satisfied the 100 percent code-required coverage. Therefore, no relief is needed for the inner radius zone.
In its November 21, 2005, letter, the NRC staff asked the licensee to discuss how the EPRI modeling of the examination volume leads to a higher quality of examination. By letter dated December 8, 2005, the licensee responded that the proposed relief request should have stated that a more reliable examination was performed based on the EPRI modeling and PDI techniques that were utilized. According to the licensee, the EPRI modeling provides the geometric inputs and parameters so the PDI-qualified examiner can more reliably perform the examination and have more confidence that if a flaw were present, it would be detected.
The licensee stated that the use of PDI-qualified transducers limited the examination coverage of nozzle NVE-BD-N9 and contributed to a reduction of coverage as compared to previous examinations in the outer 85-percent volume. The NRC staff asked the licensee to determine whether an examination based on less coverage but with a higher quality examination is appropriate in terms of detecting flaws in nozzle NVE-BD-N9. The licensee responded that because the NRC regulations require the use of Appendix VIII to the ASME Code,Section XI, for detecting flaws in the nozzles, only qualified PDI procedures and specific qualified transducers may be used for this examination. Previous examinations may have reported additional coverage. However, the licensee explained that the previous examiners and equipment were not qualified by performance demonstration, so the coverage reported during that period cannot be considered as reliable as it would be today. Nozzle modeling in conjunction with PDI examination methods maximizes the coverage that can be reliably obtained, even though PDI-qualified transducers, procedures, and examiners limit how much coverage an examiner is able to attain.
In the safety evaluation approving Relief Request RI-21, Revision 1, dated October 23, 1997, the staff discussed a visual examination, VT-2, associated with inservice inspection of nozzle NVE-BD-N9. The NRC staff asked the licensee whether the visual examination, VT-2, has been conducted and will be performed for nozzle NVE-BD-N9 in conjunction with the required volumetric examinations in the future. The licensee responded that the VT-2 examination of this nozzle is performed every refueling outage in accordance with the vessel pressure test per Category B-P of ASME Code,Section XI. The VT-2 examination was last performed in Refueling Outage 22 in 2005, and, no visible leakage was detected.
The NRC staff finds that (1) the licensee has performed volumetric examinations of nozzle weld NVE-BD-N9 to the extent practical (i.e., 40-percent coverage), (2) the licensee followed the qualified PDI technique in accordance with Appendix VIII to the ASME Code,Section XI, and (3) it is acceptable to eliminate the reference to an inner radius of weld NVE-BD-N9 in Relief Request RI-21, Revision 2, because the licensee is able to meet the 100-percent examination coverage requirement of the inner radius of the weld. The licensee will perform the VT-2 visual examination every refueling outage, which provides additional monitoring of the subject weld to assure its structural integrity. Therefore, the NRC staff concludes that the licensee has demonstrated that it is impractical to comply with the ASME Code required volumetric examination coverage of the welds as described in Relief Requests RI-21, Revision 2.
Pursuant to 10 CFR 50.55a(g)(6)(i), the staff grants Relief Requests RI-21, Revision 2, for the remainder of the third 10-year inservice inspection interval.
In Relief Request RI-21, Revision 1, the restrictions to the volumetric examination of nozzle NVE-BD-N4A&C were identified as the nozzle configuration, insulation support frame, and thermocouple pads. For Relief Request RI-21, Revision 2, the licensee proposed to add "Instrument Nozzles N11A&C" to the access restriction listing for nozzle N4A&C in the proposed Table RI-21-1, Revision 2. This is an administrative change to reflect an additional restriction which the NRC staff finds acceptable.
4.0 RELIEF REQUEST RI-36
4.1 ASME Code Components Affected
The affected components are welds RVD-BJ-17 and RVD-BJ-18 in the reactor recirculation system. They are Class 1 pressure retaining welds and follow Examination Category B-J, Item Number B9.20, in Table IWB-2500-1 of the ASME Code,Section XI.
4.2 Applicable Code Edition and Addenda
The code of record for the third 10-year inservice inspection interval is ASME Code,Section XI, 1989 edition, with no addenda.
4.3 Applicable Code Requirement
ASME Code,Section XI, 1989 edition, Table IWB-2500-1, requires a surface examination for pressure retaining welds in piping less than nominal pipe size (NPS) 4. However, the subject welds were recently included in the Risk-Informed Inservice Inspection Program with Category R-A in accordance with Relief Request RI-34 as approved by the NRC on December 9, 2004.
These welds are now required to receive volumetric examinations with a required coverage of greater than 90 percent. If 90-percent volumetric coverage cannot be obtained, the process outlined in EPRI Technical Report (TR)-112657, Revision B-A, "Revised Risk-Informed Inservice Inspection Evaluation Procedure (PWRMRP-05)," Final Report, December 1999, must be followed.
4.4 Licensees Proposed Alternative and Basis for Use Pursuant to 10 CFR 50.55a(g)(5)(iii), the licensee has determined that compliance with the code requirements of achieving essentially 100-percent coverage of the piping welds is impractical. In lieu of performing the code-required examinations, the licensee proposes to examine the accessible portions of pressure retaining welds in piping to the extent practical.
The construction permit for the plant was issued before the effective date of implementation for the ASME Code,Section XI. Thus, the plant was not designed to meet the ASME Code requirements of inservice inspection. Therefore, 100-percent compliance is not feasible or practical.
In 2005, the licensee performed volumetric examination on welds RVD-BJ-17 and RVD-BJ-18, and achieved a coverage of 84 percent (84.21 percent actual) and 84 percent (84.85 percent actual), respectively. The "as-welded" weld crown height for each weld prevented greater than 90-percent coverage. These welds are located on two-inch, schedule-80 piping that is thin-walled material where a slight mismatch at the weld joints is possible. The licensee stated that weld conditioning was not considered or used, to prevent the inadvertent removal of base metal from one or both piping segments. Additionally, interferences were also a general concern.
Therefore, alternative welds were not selected for examination. The licensee used alternate angles of an ultrasonic scan. However, due to high signal-to-noise ratios of the ultrasonic technique, additional coverage could not be achieved.
4.5 Duration of Proposed Alternative
The licensee requested that the proposed alternative be applicable for the remainder of the third 10-year interval of the inservice inspection program, which ends on February 28, 2006.
4.6 Staff Evaluation The licensee performed the volumetric examination on welds RVD-BJ-17 and RVD-BJ-18 in refueling outage 22 in 2005, with no recordable indications. Prior to the application of the Risk-Informed Inservice Inspection Program, volumetric examination of these welds was not required. These welds were recently added to the population of welds to be scheduled for volumetric examinations when the Risk-Informed Inservice Inspection Program was implemented.
In its request for additional information, dated November 21, 2005, the NRC staff asked the licensee to demonstrate how Relief Request RI-36 satisfies the relief request guidelines in EPRI TR-112657, Revision B-A. In its December 8, 2005, letter, the licensee responded that in accordance with EPRI TR-112657, Revision B-A, Section 6.4, if greater than 90-percent coverage cannot be obtained for the selected welds, and none of the other welds in that segment would yield a greater coverage, the utility is directed to submit a relief request. Welds RVD-BJ-17 and RVD-BJ-18 were selected because they gave the best access for examination.
These welds are located in 2-inch small bore piping. These two welds are grouped with fifteen other small bore welds that are either butt welded or socket welded. These welds are classified as risk category 4 with a risk characterization ranking of medium, a consequence ranking of high, and a failure potential ranking of low. The licensee stated that reduction in coverage from the required 90 percent to slightly greater then 84 percent has little or no impact to the overall risk contribution.
In the December 8, 2005, letter, in response to the NRC staffs question about the examination coverage, the licensee stated that for welds RVD-BJ-17 and RVD-BJ-18, the ultrasonic examination was performed manually. Therefore, only the beam angle is known. The licensee used 45-and 60-degree transducers with acceptable signal-to-noise ratios. The signal-to-noise ratio for the 70-degree transducer was not acceptable, so the 70-degree transducer was not used. The examination coverage was limited by the "as-welded" crown. Conditioning the crown may have achieved more examination volume coverage. However, the inadvertent removal of weld metal below code minimum on a thin-wall piping material may have occurred on one or both of the segments and, therefore, was not performed.
The NRC staff finds that (1) the licensee has performed the volumetric examination using the appropriate ultrasonic technique, to the extent practical, (2) the ultrasonic examination coverage of 84 percent of the weld volume should have insignificant impact to the overall risk contribution, as compared to the required 90-percent coverage, and (3) conditioning the weld crown on a thin-wall pipe to achieve the required 90-percent coverage would be impractical and it may affect the structural integrity of welds RVD-BJ-17 and RVD-BJ-18. Therefore, the NRC staff concludes that the licensee has demonstrated that it is impractical to comply with the ASME Code required volumetric examination coverage of the welds as described in Relief Request RI-36. Pursuant to 10 CFR 50.55a(g)(6)(i), the staff grants Relief Requests RI-36 for the remainder of the third 10-year inservice inspection interval.
5.0 RELIEF REQUEST RI-20 Relief Request RI-20 pertains to partial surface examination coverage of weld RVD-BF-14 in the original Inservice inspection program, even though it was subsequently incorporated into the Risk-Informed Inservice Inspection Program. As a result, this weld has been selected for volumetric examination instead of surface examination. Therefore, relief from surface examination requirements for this weld becomes moot. The licensee stated that a separate relief request for volumetric examination of this weld will be submitted, when needed. The licensee proposed to withdraw Relief Request RI-20 because it is no longer applicable to the inspection of weld RVD-BF-14 in the Risk-Informed Inservice Inspection Program. The NRC staff finds the withdrawal of Relief Request RI-20 acceptable because relief from the ASME Code requirements for the surface examination of weld RVD-BF-14 is not needed.
6.0 CONCLUSION
On the basis of the information submitted, the NRC staff finds that the licensee has demonstrated that it is impractical to comply with the ASME Code required volumetric examination coverage of the welds as described in Relief Requests RI-21, Revision 2, and RI-36. The staff concludes that the proposed alternative to volumetric examination coverage in Relief Requests RI-21, Revision 2, and RI-36 will provide adequate examination coverage to ensure structural integrity for the subject welds. Pursuant to 10 CFR 50.55a(g)(6)(i), the staff grants Relief Requests RI-21, Revision 2, and RI-36 for the remainder of the third 10-year inservice inspection interval at the Cooper Nuclear Station.
The NRC staff concludes that granting Relief Requests RI-21, Revision 2, and RI-36 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 upon the licensee that could result if the requirements were imposed on the facility.
The NRC staff also concludes that the withdrawal of Relief Request RI-20 is acceptable because relief from the ASME Code required examination for the subject weld is no longer needed.
All other ASME Code,Section XI, requirements for which relief was not specifically requested and approved in Relief Requests RI-21, Revision 2, and RI-36 remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.
Principal Contributor: J. Tsao Date: February 24, 2006
February 2006 Cooper Nuclear Station cc:
Mr. William J. Fehrman President and Chief Executive Officer Nebraska Public Power District 1414 15th Street Columbus, NE 68601 Mr. Gene Mace Nuclear Asset Manager Nebraska Public Power District P.O. Box 98 Brownville, NE 68321 Mr. John C. McClure Vice President and General Counsel Nebraska Public Power District P. O. Box 499 Columbus, NE 68602-0499 Mr. Paul V. Fleming Licensing Manager Nebraska Public Power District P.O. Box 98 Brownville, NE 68321 Mr. Michael J. Linder, Director Nebraska Department of Environmental Quality P. O. Box 98922 Lincoln, NE 68509-8922 Chairman Nemaha County Board of Commissioners Nemaha County Courthouse 1824 N Street Auburn, NE 68305 Ms. Julia Schmitt, Manager Radiation Control Program Nebraska Health & Human Services R & L Public Health Assurance 301 Centennial Mall, South P.O. Box 95007 Lincoln, NE 68509-5007 Mr. H. Floyd Gilzow Deputy Director for Policy Missouri Department of Natural Resources P. O. Box 176 Jefferson City, MO 65102-0176 Senior Resident Inspector U.S. Nuclear Regulatory Commission P. O. Box 218 Brownville, NE 68321 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011 Director, Missouri State Emergency Management Agency P. O. Box 116 Jefferson City, MO 65102-0116 Chief, Radiation and Asbestos Control Section Kansas Department of Health and Environment Bureau of Air and Radiation 1000 SW Jackson Suite 310 Topeka, KS 66612-1366 Mr. Daniel K. McGhee Bureau of Radiological Health Iowa Department of Public Health Lucas State Office Building, 5th Floor 321 East 12th Street Des Moines, IA 50319 Mr. Keith G. Henke, Planner Division of Community and Public Health Office of Emergency Coordination 930 Wildwood P.O. Box 570 Jefferson City, MO 65102 Jerry C. Roberts, Director of Nuclear Safety Assurance Nebraska Public Power District P.O. Box 98 Brownville, NE 68321 Mr. John F. McCann, Director Licensing, Entergy Nuclear Northeast Entergy Nuclear Operations, Inc.
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