ML100090047

November 9, 2009, Summary of Telephone Conference Call Held on Between NRC Staff and Nebraska Public Power District, Related to a Clarification for Certain Responses to Requests for Additional Information for Cooper License Renewal

ML100090047
Person / Time
Site:	Cooper
Issue date:	01/21/2010
From:	Tam Tran License Renewal Projects Branch 1
To:
Tran T, NRR/DLR, 415-3617
References
Download: ML100090047 (17)

Text

January 21, 2010 LICENSEE: Nebraska Public Power District

FACILITY: Cooper Nuclear Station

SUMMARY

OF TELEPHONE CONFERENCE CALL HELD ON NOVEMBER 9, 2009, BETWEEN THE U.S. NUCLEAR REGULATORY COMMISSION STAFF AND NEBRASKA PUBLIC POWER DISTRICT, RELATED TO A CLARIFICATION FOR CERTAIN RESPONSES TO REQUESTS FOR ADDITIONAL INFORMATION FOR COOPER NUCLEAR STATION LICENSE RENEWAL The U.S. Nuclear Regulatory Commission staff and representatives of Nebraska Public Power District (the applicant) held a telephone conference call on November 9, 2009, to discuss clarifications for certain responses to requests for additional information for Cooper Nuclear Station license renewal.

Enclosure 1 provides a listing of the participants and Enclosure 2 contains a brief description of the conference call.

The applicant had an opportunity to comment on this summary.

/RA/ Tam Tran, Project Manager Projects Branch 1 Division of License Renewal Office of Nuclear Reactor Regulation Docket No. 50-298

Enclosures:

As stated cc w/encls: See next page

ML100090047 OFFICE PM:RPB1:DLR LA:RPOB:DLRPM:RPB1:DLRBC:RPB1:DLR PM:RPB1:DLR NAME T. Tran S. Figueroa B. Brady B. Pham T. Tran DATE 01/14/10 01/12/10 01/19/10 01/21/10 01/21/10 Letter to Nebraska Public Power District from Tam Tran dated January 21, 2010

SUMMARY

OF TELEPHONE CONFERENCE CALL HELD ON NOVEMBER 9, 2009, BETWEEN THE U.S. NUCLEAR REGULATORY COMMISSION STAFF AND NEBRASKA PUBLIC POWER DISTRICT, RELATED TO A CLARIFICATION FOR CERTAIN RESPONSES TO REQUESTS FOR ADDITIONAL INFORMATION FOR COOPER NUCLEAR STATION LICENSE RENEWAL DISTRIBUTION

HARD COPY: DLR RF E-MAIL: PUBLIC RidsNrrDlr Resource RidsNrrDlrRpb1 Resource RidsNrrDlrRpb2 Resource RidsNrrDlrRarb Resource RidsNrrDlrRasb Resource RidsNrrDlrRapb Resource RidsNrrDlrRerb Resource RidsNrrDlrRpob Resource RidsNrrDciCvib Resource RidsNrrDciCpnb Resource RidsNrrDciCsgb Resource RidsNrrDraAfpb Resource RidsNrrDraApla Resource RidsNrrDeEmcb Resource RidsNrrDeEeeb Resource RidsNrrDssSrxb Resource RidsNrrDssSbpb Resource RidsNrrDssScvb Resource RidsOgcMailCenter Resource ------------- T. Tran G. Pick, RIV B. Brady G. Miller, RIV F. Lyon I. Couret, OPA D. Roth, OGC A. Jones, OGC N. Taylor, RIV E. Collins, RIV C. Casto, RIV B. Maier, RIV V. Dricks, RIV D. Chamberlain, RIV A. Vegel, RIV W. Walker, RIV ENCLOSURE 1 LIST OF PARTICIPANTS TELEPHONE CONFERENCE CALL COOPER NUCLEAR STATION LICENSE RENEWAL APPLICATION November 9, 2009 PARTICIPANTS AFFILIATIONS T. Tran U.S. Nuclear Regulatory Commission (NRC) B. Brady NRC R. Auluck NRC G. Cheruvenki NRC C. Doutt NRC A. Obodoako NRC M. Yoder NRC A. Wong NRC J. Medoff NRC W. Smith NRC H. Ashar NRC E. Wong NRC S. Min NRC O. Yee NRC B. Lehman NRC D. Diercks NRC D. Bremer Nebraska Public Power District (NPPD) W.Victor NPPD K. Pounders NPPD K. Thomas NPPD T. McClure NPPD D. Lach Entergy Nuclear Operations, Inc. (Entergy) A. Cox Entergy PARTICIPANTS AFFILIATIONS J. Lingenfelter Entergy T. Ivy Entergy A. Taylor Entergy ENCLOSURE 2 COOPER NUCLEAR STATION LICENSE RENEWAL APPLICATION (Brief description of the conference call)

A conference call was conducted on November 9, 2009, between the U.S. Nuclear Regulatory Commission (NRC or the staff) and Nebraska Public Power District (NPPD or the applicant) to discuss follow-up questions relating to request for additional information (RAI) 3.3.1-1, B.1.7-5, B.1.10-2, and B.1.40-1 and to discuss license renewal application (LRA) Section 3.3.2.2.6.

A. Clarification for RAI 3.3.1-1 and RAI B.1.7-5 Responses In the teleconference call held on August 6, 2009, and follow-up communications, the applicant provided the following technical information in relation to RAI 3.3.1-1 and RAI B.1.7-5. Class 1 CSB-BF-1 weld was replaced with IGSCC resistant material (IGSCC: intergranular stress corrosion cracking). The base material for the weld replacement was 316L stainless steel. The welding was closely monitored to control interpass temperatures, heat input and deposition rate. In addition, induction heating stress improvement was performed.

The weld material was ER308 stainless steel.

The staff noted that in accordance with Generic Letter (GL) 88-01, the base material is an IGSCC-resistant material. However, the staff finds that the weld material for the weld replacement is non-resistant to cracking sensitization and IGSCC because the carbon content of the weld material (ER308 stainless steel) is higher than 0.035 wt percent.

The staff needed to further evaluate the IGSCC Category of the replaced weld in accordance with GL 88-01 because the applicant stated that the weld is a Category A weld (IGSCC-resistant) with the high carbon content.

In addition, the staff needed further clarification to determine whether the Category A welds are properly categorized in accordance with GL 88-01.

The concern regarding the base and weld materials of the CSB-BF-1 weld and stress corrosion cracking (SCC) susceptibility of the materials was resolved during the teleconference call held on September 2, 2009, as described in the conference call summary (Agencywide Documents and Management Access Accession Number ML092990400).

The staff requested further clarifications on the operating experience and IGSCC categorization of the stainless steel welds; and the applicant provided responses as follow:

1. In terms of the operating experience of the CSB-BF-1 weld, provide the following information.

a. What were the original base and weld materials of the CSB-BF-1 weld before the weld replacement? NPPD Response: The original base material of the safe end was 304SS. The original weld was Inconel 182.

b. Clarify whether the weld replacement was due to the stress corrosion cracking in the original weld or base material.

NPPD Response: The weld was replaced as part of a large project to replace IGSCC susceptible piping in the drywell. Cooper Nuclear Station (CNS) replaced the safe-ends on the Reactor Recirculation, Core Spray piping, and jet pump instrumentation. In addition, the Class 1 stainless steel piping was replaced with IGSCC-resistant materials as well (Ref. CNS Letter NLS8800803, dated August 3, 1988).

c. If applicable, provide the ferrite content in the original and replaced welds.

NPPD Response: Measurement of the ferrite content for the original welds was not required at the time of construction.

For the piping replacement project, NPPD required a maximum delta ferrite of 10 FN and a minimum of five FN for the piping inside primary containment. The ferrite content of the individual welds is not readily available (Ref. CNS Letter NLS8900265, dated July 24, 1989).

d. If applicable, clarify whether the inspection results for the replaced weld and its associated base material have indicated any occurrence of SCC.

NPPD Response: No recordable indications have been detected since replacement in 1985.

2. In terms of the weld and base materials of the 106 GL 88-01 IGSCC Category A welds described in the applicant's fourth interval inservice inspection (ISI) plan (NLS2008093, Enclosure 1 of Updated Safety Analysis Report [USAR], Revision XXIII), provide the following information:

a. List all the austenitic stainless steel base materials associated with the 106 Category A welds and clarify whether all the base materials are IGSCC-resistant in accordance with GL 88-01 (e.g., "The base materials are 316L SS, 316NG SS...").

NPPD Response: The pipe materials for the category A welds are 316NG stainless steel. The safe end materials are 316L stainless steel (Ref. CNS Letter NLS8800803, dated August 3, 1988). 2.a.1 In the applicant's letter dated January 30, 2008 (NLS2008012) regarding Title 10 of the Code of Federal Regulations Part 50.55a (10 CFR 50.55a) Request Number RI-35, Revision 0, the applicant stated that the safe-end material associated with the safe-end-to-nozzle welds described below in relation to Request 3 is SA403-WP316 (Page 4) in contrast with 316L stainless steel (from this teleconference call). The staff notes that SA403-WP316 described in the foregoing letter is different from 316L stainless steel that the applicant provided during this teleconference call. Explain why the discrepancy between SA403-WP316 (i.e. 316 stainless steel) and 316L stainless steel exists for the safe end material.

NPPD Response: The nozzle safe-ends were procured as SA403 type WP316 with modified chemistry that is equivalent to 316L.

The Relief Request did not provide the additional level of detail regarding the modified chemistry as it was not considered necessary and would not have changed the basis for the Relief Request. The CNS documentation refers to both the SA-403 and 316L for the safe-ends. The chemistry requirements for both the piping and forgings were as follows:

Carbon: 0.020% maximum Silicon: 0.75% maximum Manganese: 1.50-2.00%

Phosphorous: 0.030% Sulfur: 0.030% Nickel: 11.0 - 14.0% Chromium: 16.0 0% Molybdenum: 2.00 - 3.00%

Nitrogen: 0.06 - 0.12% Vanadium: 0.05% maximum Titanium: 0.02% maximum Niobium and Tantalum: 0.02% maximum Cobalt: 0.15% maximum Copper: 0.3% maximum Note: The nitrogen was added to increase the strength due to the reduced carbon.

2.a.2 The staff notes that as described in USAR Revision XXI, Enclosure 5 (NLS2005065 dated August 12, 2005), the applicant's third interval ISI Program indicates that the safe-end base material is "P20" (316NG stainless steel). Explain why the discrepancy between 316NG stainless steel and 316L stainless steel exists for the safe-end material.

NPPD Response: The piping was procured to SA-312, Type 316 with modified chemistry that our documentation refers to as 316NG. The piping was provided with the same chemistry specification as above also with the addition of nitrogen to increase the strength due to the reduced carbon.

b. In relation to the foregoing stainless steel base materials, list all the weld materials used for the 106 Category A welds and clarify whether all the weld materials are IGSCC-resistant in accordance with GL 88-01 (e.g., "The weld materials are ER308L SS, ER309L SS...").

NPPD Response: All the welds are IGSCC resistant. The weld material is ER316L. c. If a material with carbon content higher than 0.035 wt percent such as ER308 is used in any of the 106 Category A welds, provide what material is used in the weld. In addition, if applicable, provide the component ID, Examination Category (B-J or B-F) and ferrite content of the weld. The staff notes that the ID is described in the aforementioned ISI plan.

NPPD Response: Not applicable. ER316L weld material does not have a higher carbon content than 0.035 wt percent.

3. Relative to technical report NLS 2008012 (regarding Item RI-35), state whether any of the welds described in the reference (Page 1) has an indication of IGSCC.

NPPD Response: No, the relief request was a contingency in the event we found indications in RE24, April 2008.

4. Relative to technical report NLS 2008012 (regarding item RI-35), state whether any of the welds described in the reference (Page 1) has been reinforced by weld overlay. NPPD Response: No weld overlays were needed.

B. Clarification for RAI B.1.10-2 Response In NPPD letter NLS2009055, dated July 29, 2009, the applicant submitted response to RAI B.1.10-2 related to recoating the torus prior to or during the period of extended operation to reduce the corrosion rate. The applicant stated that NPPD does not have firm plans to recoat the torus. However, NPPD will continue to inspect the torus prior to and during the period of extended operation as required by American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, Subsection IWE. The indications found during the inspections will continue to be evaluated to determine the need for corrective actions including, but not limited to recoating. The last recoating of areas within the torus was the result of the 2005 torus inspection.

The staff has a concern that the large number of repairs, excessive zinc depletion, and pitting at thousands of locations to the torus during the last 35 years indicate that the integrity of the torus coating cannot be relied upon during the period of extended operation. Normal life of the torus coating is 15-20 years. In addition, CNS internal documents concerning self assessment of the torus coating have previously recommended recoating of the torus. The staff requests the applicant to provide detailed justification for not having any firm plans to recoat the torus. The staff needs the following information to confirm that the effects of aging of the torus will be adequately managed so that its intended function will be maintained consistent with the current licensing basis for the period of extended operation as required by 10 CFR 54.21(a)(3):

1. How many repairs have been performed to the CNS torus shell coating until now.

NPPD Response: The applicant does not have an exact number of coating repairs. In 2007, there were 77 deep pits; in 2005, there was 1; and in 2008, there were 25.

2. How many of these repairs were performed at locations where the torus base metal thickness had been reduced by greater than 10 percent.

NPPD Response: The applicant did not recall that there have been any repairs performed where the torus base metal thickness had been reduced by greater than 10 percent. However, this will be validated in the response for a subsequent RAI.

3. How many of the repairs required augmented inspection (including ultrasonic testing [UT] examination) in accordance with ASME Section XI, Subsection IWE Code, Table 2500-1. Articles IWE-3122.3.b, IWE 3200, IWE 3511.3 and IWE-2420 of the ASME Code requires augmented examination of the area containing flaw or degradation if the base metal thickness is reduced by greater than 10 percent.

NPPD Response: None, per the response to Question 2.

4. The bottom half of interior surface is not easily accessible for visual examination. Therefore, does CNS has any plans to perform UT examination from the torus exterior surface at a number of randomly selected locations to demonstrate with 95 percent confidence that 95 percent of the torus surface has base metal thickness greater than 90 percent of the nominal thickness?

NPPD Response: The applicant performs a 100 percent visual inspection of the torus for pitting, and measures the deep pits to affirm that there is no greater than 10 percent penetration. The applicant indicated that it was unaware of any pits that had exceeded that depth. The 95/95 criteria only applies to volumetric examinations.

Follow-Up Discussion The staff agreed to provide an RAI that clearly clarifies the information needed, such as, "How does NPPD meet the 95/95 criteria?"

C. Clarification for RAI B.1.40-1 Response The applicant responded by letter dated June 15, 2009, and repeated its position that performance monitoring is typically a part of an engineering program, which would not be a part of water chemistry. The applicant also again stated that functional and performance testing verifies that component active functions can be accomplished and is of little value in detecting loss of passive functions due to such aging effects as loss of material. The applicant further stated that passive intended functions of pumps, heat exchangers, and other components will be adequately managed by the Water Chemistry Control Closed Cooling Water and One-Time Inspection Programs through monitoring and control of water chemistry parameters and verification of absence of aging effects. The staff noted that Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants (SRP-LR) Table 3.1-2 states that the Closed-Cycle Cooling Water System Program relies on both preventative measures and non-chemistry monitoring, and in Table 2.1-3, it further recommends performance monitoring of heat exchangers as a part of aging management.

The staff further noted that GL 89-13 states, "a closed-cycle system is defined as a part of the service water system that is not subject to significant sources of contamination, one in which water chemistry is controlled, and one in which heat is not directly rejected to a heat sink. If all these conditions are not satisfied, the system is to be considered an open-cycle system in regard to the specific actions require by GL 89-13." In view of the extended lack of control over dissolved oxygen in the applicant's closed-cycle cooling water system discussed in RAI B.1.40-3, the staff determines that it is subject to the GL 89-13 guidelines.

One of these guidelines is to conduct a test program to verify the heat transfer capability of all safety-related heat exchangers cooled by service water. This program is to include both initial testing and periodic retesting of these components.

Based on its review, the staff finds that the applicant has not demonstrated that, with this exception, the applicant's Water Chemistry Control-Closed Cooling Water Program will be consistent with the program elements of Generic Aging Lessons Learned (GALL) aging management program (AMP) XI.M21; therefore, the staff's concern described in RAI B.1.40-1 remains unresolved.

NPPD Response: Regarding the SRP-LR Table 2.1-3, the applicant believes this statement is not a recommendation. It does not say performance monitoring being parts of an AMP but does say that programs, "in conjunction with existing procedures for monitoring heat exchanger performance." It simply states that heat transfer should be considered as an intended function. This was done for systems in the Closed Cooling Water Chemistry AMP in the LRA.

Heat exchangers cooled by service water (SW) are included in GL 89-13 testing or monitoring portion of the SW Integrity AMP, and therefore, the applicant believes that this has no relevance to Closed-Cooling Water Chemistry AMP. The applicant believes that it is inappropriate to apply GL 89-13 to the CNS closed cooling water systems. The applicant refers to the RAI response discussion of the effects of the high oxygen to support its belief. The applicant does not believe the intent of GL 89-13 was to bring in closed cycle systems following chemistry excursions. The effects of the oxygen concentration excursions were minimal and the One-Time Insection Program will provide further confirmation of the closed cooling water chemistry program effectiveness.

Follow-Up Discussion

The staff clarified the concern in the RAI is that since the REC system has had a history of dissolved oxygen excursions, the exception taken to GALL performance and functional testing would be acceptable only in concert with a periodic inspection regimen, versus the proposed One-Time Inspection AMP. The staff agreed the question should be revised to reflect the clarification above.

D. Clarification for RAI 3.5-1 Response RAI 3.5-1 response discusses the water cement ratio of safety-related structures. Please clarify with actual values of the ratio found in CNS concrete.

NPPD Response: The applicant indicated understanding of the request.

Follow-up Subsequent to the teleconference the following supplemental information was provided by the applicant.

The intent of CNS concrete mix design is to secure plastic, workable concrete of homogeneous structure which, when hardened, will have durability, impermeability and the strengths for the specified class of concrete. The proportions of concrete including entrained air, is as set forth in ACI-613 Recommended Practice for selecting proportions for concrete. Concrete mix design is specified for multiple compressive strengths, however for the two specific classes in question, 4000 psi and 3000 psi, CNS defines these as large (3LA and 4LA, <1-1/2") and small (3SA and 4SA, < 3/4") aggregates containing a water reducing retardant and air entrainment. The table below provides the final recommended batch proportions for CNS for 4000 psi and 3000 psi concrete.

The final batch proportions did not specify minimum or maximum air content, however the entrained air used in the actual trial mix is provided in the table. The specification states air entrainment to be between four percent and six percent. Class of Concrete 3SA 3LA Minimum 28-day Compressive Strength, psi 3000 3000 Yield, cu. yd 1.0 1.0 Portland Cement, Type II, sk 3.59 3.33 Portland Cement. Type II, lb 337 313 Fine Aggregate (sat. sur. dry), lb 1320 1290 Fine Aggregate (sat. sur. dry), % 40 38 Coarse Aggregate (sat. sur. dry), No. 4 to 3/4", lb 1980 1055 Coarse Aggregate (sat. sur. dry), 3/4" to 1-1/2", lb -- 1055 Air Entraining Agent (Daravair), fl, oz (Trial mix actual ) 4.9 5.3 Water-Reducer Retarder (Daratard), fl, oz 32 30 Total Water (free + added), gal 30.5 27 Water Cement Ratio, gal/sk 8.5 8.1 Water Cement Ratio 0.75 0.71 Class of Concrete 4SA 4LA Minimum 28-day Compressive Strength, psi 4000 4000 Yield, cu. yd 1.0 1.0 Portland Cement, Type II, sk 4.60 4.09 Portland Cement. Type II, lb 432 384 Fine Aggregate (sat. sur. dry), lb 1155 1160 Fine Aggregate (sat. sur. dry), % 35 34 Coarse Aggregate (sat. sur. dry), No. 4 to 3/4", lb 2145 1125 Coarse Aggregate (sat. sur. dry), 3/4" to 1-1/2", lb -- 1125 Air Entraining Agent (Daravair), fl, oz (Trial mix actual ) 5.1 5.8 Water-Reducer Retarder (Daratard), fl, oz 41 37 Total Water (free + added), gal 27.0 24.0 Water Cement Ratio, gal/sk 5.87 5.87 Water Cement Ratio 0.52 0.52 E. Clarification for RAI 3.5.2.2-1 Response Please discuss the containment bellows (containment penetration). The staff indicated the question relates to the main steam and feedwater containment penetration bellows; whether the bellows were single ply or two ply bellows. The staff referred to IEN 92-20 that described inadequacies in local leak rate test (LLRT) programs for single ply bellows that could experience cracking.

NPPD Response: The main steam and feedwater testable penetrations consist of double layered metal bellows which are locally leak tested by pressurizing the annulus between the double layers to five psig rather than Pa. They are also tested to Pa as part of the containment LLRT. These bellows are "2-ply" per Drawings 453012425 and 453012427. The design of the bellows does not permit testing at a higher pressure. The bellows are exposed to drywell atmosphere on one side and RB atmosphere on the other.

The feedwater and main steam penetrations are enclosed in a guard pipe that is attached to the process line through a multiple head fitting. The guard pipe and fluid head are designed to the same pressure requirements as the process line. The penetration nozzle is welded to the drywell and extends through the biological shield where it is welded to the expansion bellows assembly which in turn is welded to the fluid head fitting. As shown on USAR Figure V-2-3, each penetration assembly is also provided with a continuous annular plate, attached to the guard pipe and extending to the outside diameter of the penetration nozzle. This annular plate is located at the inside of containment and serves the double functions of a testing seal seat and a jet deflector plate, which prevents impingement of fluid from an adjacent pipe that is postulated to have ruptured.

F. Additional Discussion Item Associated With LRA Section 3.3.2.2.6 Please clarify whether the responses to RAIs regarding aging management review Section 3.3.2.2.6, "Reduction of Neutron-Absorbing Capacity and Loss of Material due to General Corrosion," apply to the corresponding AMP B.1.23, "Neutron Absorber Monitoring." Please discuss whether the AMP Section B.1.23 will be supplemented with additional information.

NPPD Response: The applicant indicated that it was in the process of developing the response to RAIs 3.3.2.2.6-3 and would address these elements in the response and would supplement B.1.23, if needed.

Cooper Nuclear Station cc: Mr. Ronald D. Asche 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. David Van Der Kamp 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 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 612 E. Lamar Blvd., Suite 400 Arlington, TX 76011-4125

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 Ms. Melanie Rasmussen Radiation Control Program Director 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 Drive P.O. Box 570 Jefferson City, MO 65102

Cooper Nuclear Station cc: Mr. Art Zaremba, 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. 440 Hamilton Avenue White Plains, NY 10601-1813 Mr. Mike Boyce Cooper Strategic Initiatives Manager Cooper Nuclear Station 72676 - 648A Avenue Brownville, NE 68321 Mr. Dave Bremer License Renewal Project Manager Cooper Nuclear Station 72676 - 648A Avenue Brownville, NE 68321

Mr. Bill Victor License Renewal Project Licensing Lead Cooper Nuclear Station 72676 - 648A Avenue Brownville, NE 68321

Mr. Garry Young License Renewal Manager Entergy Nuclear 1448 S.R. 333, N-GSB-45 Russellville, AR 72802 Mr. Alan Cox License Renewal Technical Manager Entergy Nuclear 1448 S.R. 333, N-GSB-45 Russellville, AR 72802

Mr. Dave Lach LRP Entergy Project Manager Entergy Nuclear 1448 S.R. 333, N-GSB-45 Russellville, AR 72802 Mr. Jerry Perry 500 S. Main Street Rock Port, MO 64482 Ms. Yolanda Peck 1008 Central Ave. Auburn, NE 68305

Ms. Kendall Neiman 830 Central Ave. Auburn, NE 68305 Ms. Annie Thomas 1522 I Street Auburn, NE 68305 Mr. John Chaney 1101 17 th Street Auburn, NE 68305 Mr. Darrell Kruse 2415 McConnell Ave. Auburn, NE 68305

Ms. Daryl J. Obermeyer 64381 727A Road Brownville, NE 68321

Ms. Sherry Black, Director Auburn Memorial Library 1810 Courthouse Ave. Auburn, NE 68305 Board of Brownville, NE Attn: Chairman Marty Hayes P.O. Box 67 223 Main Street Brownville, NE 68321

Cooper Nuclear Station cc: Mr. Bob Engles Mayor of Auburn, NE 1101 J Street Auburn, NE 68305 Ms. Jo Stevens Mayor of Rock Port, MO 500 S. Main Street Rock Port, MO 64482 Mr. John Cochnar U.S. Fish and Wildlife Service Ecological Services Nebraska Field Office 203 West Second Street Grand Island, NE 68801 Mr. John Askew Regional Administrator U.S. EPA Region 7 901 N. 5 th Street Kansas City, KS 66101 Ms. Joann Scheafer, Director Nebraska Department of Health & Human Services 301 Centennial Mall South Lincoln, NE 68509

Mr. Doyle Childers, Director Missouri Department of Natural Resources P.O. Box 176 Jefferson City, MO 65102

Mr. Mark Miles State Historic Preservation Officer Department of Natural Resources P.O. Box 176 Jefferson City, MO 65102 Mr. Michael J. Smith State Historic Preservation Officer Nebraska State Historical Society P.O. Box 82554 Lincoln, NE 68501 Mr. Robert Puschendorf Nebraska State Historical Society 1500 R Street, P.O. Box 82554, Lincoln, NE 68501-2554 Mr. Stewart B. Minahan Vice President Nuclear and Chief Nuclear Officer Cooper Nuclear Station 72676 - 648A Avenue Brownville, NE 68321