ML043200515
| ML043200515 | |
| Person / Time | |
|---|---|
| Site: | River Bend  (NPF-047) |
| Issue date: | 10/15/2004 |
| From: | Michael Webb NRC/NRR/DLPM/LPD4 |
| To: | Hinnenkamp P Entergy Operations |
| Webb M, NRR/DLPM, 415-1347 | |
| References | |
| TAC MC1094 | |
| Download: ML043200515 (13) | |
Text
October 15, 2004 Mr. Paul D. Hinnenkamp Vice President - Operations Entergy Operations, Inc.
River Bend Station 5485 US Highway 61N St. Francisville, LA 70775
SUBJECT:
RIVER BEND STATION, UNIT 1 - ISSUANCE OF AMENDMENT RE:
DELETION OF SHIELD BUILDING ANNULUS MIXING SYSTEM TECHNICAL SPECIFICATION (TAC NO. MC1094)
Dear Mr. Hinnenkamp:
The Commission has issued the enclosed Amendment No. 143 to Facility Operating License No. NPF-47 for the River Bend Station, Unit 1. The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated October 21, 2003, as supplemented by letters dated February 10, 2004, and August 24, 2004.
The amendment modifies the TSs to delete TS 3.6.4.4, Shield Building Annulus Mixing System and a reference to TS 3.6.4.4 within TS 3.10.1, Inservice Leak and Hydrostatic Testing Operation, and revise TS Surveillance Requirement 3.6.1.3.10 main steam isolation valve (MSIV) leakage limits.
A copy of our related Safety Evaluation is enclosed. The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.
Sincerely,
/RA/
Michael Webb, Project Manager, Section 1 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-458
Enclosures:
- 1. Amendment No. 143 to NPF-47
- 2. Safety Evaluation cc w/encls: See next page
October 15, 2004 Mr. Paul D. Hinnenkamp Vice President - Operations Entergy Operations, Inc.
River Bend Station 5485 US Highway 61N St. Francisville, LA 70775
SUBJECT:
RIVER BEND STATION, UNIT 1 - ISSUANCE OF AMENDMENT RE:
DELETION OF SHIELD BUILDING ANNULUS MIXING SYSTEM TECHNICAL SPECIFICATION (TAC NO. MC1094)
Dear Mr. Hinnenkamp:
The Commission has issued the enclosed Amendment No. 143 to Facility Operating License No. NPF-47 for the River Bend Station, Unit 1. The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated October 21, 2003, as supplemented by letters dated February 10, 2004, and August 24, 2004.
The amendment modifies the TSs to delete TS 3.6.4.4, Shield Building Annulus Mixing System and a reference to TS 3.6.4.4 within TS 3.10.1, Inservice Leak and Hydrostatic Testing Operation, and revise TS Surveillance Requirement 3.6.1.3.10 main steam isolation valve (MSIV) leakage limits.
A copy of our related Safety Evaluation is enclosed. The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.
Sincerely,
/RA/
Michael Webb, Project Manager, Section 1 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-458
Enclosures:
- 1. Amendment No. 143 to NPF-47
- 2. Safety Evaluation cc w/encls: See next page DISTRIBUTION:
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ENTERGY GULF STATES, INC. **
AND ENTERGY OPERATIONS, INC.
DOCKET NO. 50-458 RIVER BEND STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 143 License No. NPF-47 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Entergy Gulf States, Inc.* (the licensee) dated October 21, 2003, as supplemented by letters dated February 10, 2004, and August 24, 2004, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; and
- Entergy Operations, Inc. is authorized to act as agent for Entergy Gulf States, Inc., and has exclusive responsibility and control over the physical construction, operation and maintenance of the facility.
- Entergy Gulf States, Inc., has merged with a wholly owned subsidiary of Entergy Corporation.
Entergy Gulf States, Inc., was the surviving company in the merger.
E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Facility Operating License No. NPF-47 is hereby amended to read as follows:
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 143 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. EOI shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
The license amendment is effective as of its date of issuance and shall be implemented within 60 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
Michael Webb, Acting Chief, Section 1 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: October 15, 2004
ATTACHMENT TO LICENSE AMENDMENT NO. 143 FACILITY OPERATING LICENSE NO. NPF-47 DOCKET NO. 50-458 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by Amendment number and contain marginal lines indicating the areas of change.
Remove Insert 3.6-19 3.6-19 3.6-53 3.6-53 3.6-54 3.6-54 3.10-1 3.10-1
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 143 TO FACILITY OPERATING LICENSE NO. NPF-47 ENTERGY OPERATIONS, INC.
RIVER BEND STATION, UNIT 1 DOCKET NO. 50-458
1.0 INTRODUCTION
By application dated October 21, 2003 (Accession No. ML033030535), as supplemented by letters dated February 10, 2004 (Accession No. ML040500292) and August 24, 2004 (Accession No. ML042440630), Entergy Operations, Inc. (Entergy or the licensee), requested changes to the Technical Specifications (TSs) for River Bend Station, Unit 1 (RBS). The supplement dated August 24, 2004, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the Nuclear Regulatory Commission staff's original proposed no significant hazards consideration determination (which referenced the application dated October 21, 2003, and the supplemental "RBG-46222, Supplement to Amendment Request, Deletion of Technical Specification 3.6.4.4, Shield Building Annulus Mixing System; and Revision of [[system" contains a listed "[" character as part of the property label and has therefore been classified as invalid. (MSIV) Surveillance Requirement SR 3.6.1.3-10|letter dated February 10, 2004]]) as published in the Federal Register on May 25, 2004 (69 FR 29764).
The proposed change would revise the RBS TS to delete TS 3.6.4.4, Shield Building Annulus Mixing System, and its associated Bases section in their entirety; and to revise main steam isolation valve (MSIV) leakage limits contained within TS Surveillance Requirement (SR) 3.6.1.3.10. The licensee also requested changes to delete reference to TS 3.6.4.4 within TS 3.10.1, Inservice Leak and Hydrostatic Testing Operation.
The proposed deletion of TS 3.6.4.4 for the shield building annulus mixing system would allow the licensee to disable the shield building annulus mixing system fan motors, thus eliminating forced mixing of the shield building annulus atmosphere. The licensee stated that the motors used to power the shield building annulus mixing fans are obsolete, and procuring new motors is costly in engineering time and materials. The licensee believes therefore, that the elimination of the shield building annulus mixing function would result in money and labor savings.
The licensee stated that the proposed deletion of the SR 3.6.1.3.10 limit on MSIV leakage through a single valve would relax an overly burdensome requirement that may require reworking the MSIVs more frequently than strictly necessary to meet dose requirements.
SR 3.6.1.3.10 would continue to include the existing limit on total MSIV leakage through all four steam lines.
2.0 REGULATORY EVALUATION
The licensee identified the applicable regulatory requirements in Sections 4.0 and 5.1 of its application dated October 21, 2003. The regulatory requirements on which the staff based its acceptance of the proposed TS changes are found in Section 50.36 of Title 10 of the Code of Federal Regulations (10 CFR), "Technical specifications."
The TSs ensure the operational capability of structures, systems and components that are required to protect the health and safety of the public. The regulations require that the TSs include items in the following specific categories: (1) safety limits, limiting safety systems settings, and limiting control settings (50.36(c)(1)); (2) LCOs (50.36(c)(2)); (3) SRs (50.36(c)(3)); (4) design features (50.34(c)(4)); and (5) administrative controls (50.36(c)(5)). In general, there are two classes of changes to TSs: (1) changes needed to reflect modifications to the design basis (TSs are derived from the design basis), and (2) voluntary changes to take advantage of the evolution in policy and guidance as to the required content and preferred format of TSs over time. This amendment deals with the first class of changes.
This safety evaluation (SE) addresses the impact of the proposed changes on previously analyzed design basis accident radiological consequences and the acceptability of the revised analysis results. The regulatory requirements on which the staff based its acceptance are the accident dose criteria in 10 CFR 50.67, Accident source term, as supplemented in Regulatory Position 4.4 of Regulatory Guide 1.183 (RG 1.183), Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, and 10 CFR Part 50 Appendix A, General Design Criterion 19 (GDC-19), Control Room, as supplemented by Section 6.4 of NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants (SRP). Except where the licensee proposed a suitable alternative, the staff utilized the regulatory guidance provided in SRP Section 15.0.1, Radiological Consequence Analysis Using Alternative Source Terms, and RG 1.183 in performing this review. The staff also considered relevant information in the RBS updated safety analysis report and TSs.
3.0 TECHNICAL EVALUATION
3.1 Evaluation of Proposed TS Changes The NRC staff has reviewed the licensees regulatory and technical analyses in support of its proposed license amendment, which are described in Attachment 1 of the licensees submittal.
The detailed evaluation of each TS change requested by the licensee is discussed below.
TS 3.6.1.3, Primary Containment Isolation Valves Description of Change:
Revise SR 3.6.1.3.10 to remove the single main steam line leakage limit of 50 standard cubic feet per hour (scfh) when tested at > Pa.
The licensee performed a revised dose consequences analysis of the design basis loss of coolant accident (LOCA), which included the assumption that the leakage through the failed MSIV is equal to the SR 3.6.1.3.10 total leakage limit of 150 scfh. The results of the licensees dose analysis with the higher MSIV leakage assumption remain within the 10 CFR 50.67 dose criteria. The staffs review of the licensees analysis is discussed in this SE under Section 3.2, Radiological Consequences Analysis.
TS 3.6.4.4, Shield Building Annulus Mixing System Description of Change:
Delete TS 3.6.4.4 and its associated Bases section.
The licensee provided the following justification:
The current design basis for the shield building annulus mixing system is to mitigate the consequences of a LOCA. The shield building annulus mixing system is automatically initiated to provide mixing in the shield building annulus volume to dilute the radioactive material concentration in the annulus and reduce the quantity of radioactive material processed by the standby gas treatment system (SGTS). This would result in reducing the radioactive material released to the environment. The current licensing basis dose analyses assume that the shield building annulus mixing system provides for mixing and dilution into 50% of the shield building annulus volume.
The licensee performed a test during refueling outage RF11, in the spring of 2003, to demonstrate that the SGTS could meet its existing TS draw on requirements without the support of the shield building annulus mixing system. Test results indicated that the SGTS can meet all its acceptance criteria with the shield building annulus mixing system fans disabled.
The licensee performed a revised dose consequences analysis of the design basis LOCA, which included the assumption that mixing or dilution into the shield building annulus volume is not credited. The results of the licensees dose analysis without credit for mixing or dilution in the shield building annulus volume remain within the 10 CFR 50.67 dose criteria. The staffs review of the licensees analysis is discussed in this SE under Section 3.2, Radiological Consequences Analysis. Because the shield building annulus system is no longer credited as an accident mitigation function during MODES 1, 2, and 3, and does not satisfy any of the criteria of 10 CFR 50.36(c)(2)(ii) for TS limiting conditions for operation, the licensee proposes to delete TS 3.6.4.4 and its associated Bases section.
TS 3.10.1, Inservice Leak and Hydrostatic Testing Operation Description of Change:
Revise TS 3.10.1 to remove reference to TS 3.6.4.4, Shield Building Annulus Mixing System.
This proposed change acknowledges the proposed deletion of TS 3.6.4.4.
TS Changes
Conclusion:
The proposed TS changes are consistent with the revised alternative source term (AST) radiological consequences analysis. There are no other safety considerations involved; the proposed numbers are assumptions used in the AST radiological consequences analysis, and, insofar as the staff finds the AST radiological consequences analysis to be acceptable, the proposed TS changes are also found acceptable. TS 3.6.4.4 no longer fulfils the criteria in 10 CFR 50.36(c)(2)(ii) that denotes that information is placed in TSs.
3.2 Radiological Consequences Analysis The NRC staff reviewed the regulatory and technical analyses, as related to the radiological consequences of design basis accidents, performed by Entergy in support of its proposed license amendment. Information regarding these analyses was provided in Attachment 1 of the application dated October 21, 2003, and in the supplemental "RBG-46222, Supplement to Amendment Request, Deletion of Technical Specification 3.6.4.4, Shield Building Annulus Mixing System; and Revision of [[system" contains a listed "[" character as part of the property label and has therefore been classified as invalid. (MSIV) Surveillance Requirement SR 3.6.1.3-10|letter dated February 10, 2004]].
The staff reviewed the assumptions, inputs, and methods used by Entergy to assess the impact of the proposed changes and performed independent calculations to confirm the conservatism of the licensees analyses. However, the findings of this SE section are based on the descriptions of the licensees analyses and other supporting information docketed by the licensee.
By "RBG-46222, Supplement to Amendment Request, Deletion of Technical Specification 3.6.4.4, Shield Building Annulus Mixing System; and Revision of [[system" contains a listed "[" character as part of the property label and has therefore been classified as invalid. (MSIV) Surveillance Requirement SR 3.6.1.3-10|letter dated February 10, 2004]], the licensee provided a copy of their analysis of the design basis LOCA incorporating the increased MSIV leakage assumption and no credit for mixing in the shield building annulus. The staff verified that those were the only changes made to the LOCA dose analysis, which was previously reviewed and found acceptable by the staff in Amendment No. 132 to NPF-47, dated March 14, 2003, for full-scope implementation of an alternative source term (AST). The licensees assumption that leakage through one MSIV is equal to the total MSIV leakage limit for all four main steam lines (150 scfh) from SR 3.6.1.3.10, is consistent with the guidance in RG 1.183 for modeling of MSIV leakage. The staff also confirmed that the licensees modeling of the LOCA did not credit any mixing in the shield building annulus airspace. The licensees revised analysis of the radiological consequences of a postulated LOCA, with the above noted changes, shows an increase in the calculated doses, but they remain within the dose criteria in 10 CFR 50.67 and GDC-19. The results of the licensees dose analysis, expressed as total effective dose equivalent (TEDE), at the exclusion area boundary (EAB), low population zone (LPZ) boundary, and in the control room are:
Dose Criteria (rem)
EAB 16.4 25 LPZ 8.2 25 Control Room 3.5 5
The staff performed independent confirmatory analyses and verified the licensees results.
Table 1 at the end of this SE, provides the licensees analysis assumptions, which the staff also used in its confirmatory analyses. Therefore, the staff concludes that the radiological consequences analyzed and submitted by the licensee are acceptable.
Control Room Habitability In its revised analyses, the licensee assumed a control room unfiltered inleakage of 300 cubic feet per minute (cfm) for the duration of the accident. This assumption is unchanged from the previously approved analysis. Entergy has not performed integrated leakage testing at RBS to confirm this leakage value. On June 12, 2003, the staff issued Generic Letter 2003-01, "Control Room Habitability." This generic letter identifies staff concerns regarding the reliability of current surveillance testing to identify and quantify control room inleakage, and requests licensees to confirm the most limiting unfiltered inleakage into their control room envelope.
Entergy was requested by the generic letter to respond within 180 days of its issuance. Entergy submitted a 60-day response, by letter dated August 11, 2003, in accordance with the provisions of the generic letter to inform the NRC staff that they could not meet the 180-day schedule. Entergy committed in that letter to prepare a report and a plan for resolution of any significant discrepancies or conditions adverse to quality by January 31, 2005.
In its supplemental letter dated August 24, 2004, Entergy provided results of control room envelope inleakage testing performed for the RBS control room. This testing was completed on May 3, 2004. The licensee stated that these tests were performed using the methods of American Society of Testing and Materials Standard E741, Standard Test Method for Determining Air Change in a Single Zone by Means of a Tracer Gas Dilution. This test is referred to as a tracer gas test in the following discussion. The licensee performed two tracer gas tests, each with the Control Room Fresh Air (CRFA) system in isolation and either of the two divisions of the CRFA system alone operating in the emergency operation mode. The tracer gas test results for both tests demonstrated that the measured control room unfiltered inleakage was statistically zero. The licensees analyses performed in support of this amendment request assumed a control room unfiltered inleakage of 300 cfm, which results in a higher calculated control room dose than would be calculated using the results of the tracer gas test.
The staff has determined that there is reasonable assurance that the RBS control room will be habitable during a design basis LOCA with the proposed changes in place, and this amendment may be approved prior to the staff completing review of Entergys response to the generic letter for RBS. The staff bases this determination on the relative magnitude of the control room infiltration assumed in the RBS analyses, the results of control room envelope inleakage testing at RBS, and the relative impact of the proposed changes on the radioactive source term released to the environment. The staff's approval of this amendment does not relieve Entergy of addressing the information requests in Generic Letter 2003-01 and does not imply that the staff would necessarily find the analysis in this amendment acceptable as a response to information request 1(a) in Generic Letter 2003-01.
3.3 Technical Conclusion As described above, the NRC staff reviewed the assumptions, inputs, and methods used by Entergy to assess the radiological impacts of proposed changes to TS 3.6.4.4 and SR 3.6.1.3.10 at RBS. The staff finds that Entergy used analysis methods and assumptions consistent with the conservative regulatory requirements and guidance identified in Section 2.0 above. The staff compared the doses estimated by Entergy to the applicable criteria identified in Section 2.0. The staff finds reasonable assurance that the licensees estimates of the EAB, LPZ, and control room doses will continue to comply with these criteria. Therefore, the deletion of TS 3.6.4.4 and its reference in TS 3.10.1, and elimination of the single valve MSIV leakage limit from SR 3.6.1.3.10 is acceptable with regard to the radiological consequences of postulated design basis accidents. Furthermore, the deletion of TS is acceptable because the licensee will no longer credit the Shield Building Annulus Mixing System as an accident mitigation function during MODES 1, 2, and 3, so the system no longer satisfies the criteria of 10 CFR 50.36(c)(2)(ii) for inclusion in TS LCOs.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Louisiana State official was notified of the proposed issuance of the amendment. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding published May 25, 2004 (69 FR 29764). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor: M. Hart Date: October 15, 2004 Table 1 LOCA Dose Analysis Parameters and Assumptions Parameter Value Reactor power 3100 MWt Building volumes Containment 1.19E+6 ft3 Drywell 2.36E+5 ft3 Auxiliary (50% of total volume) 5.80E+5 ft3 Suppression pool 1.20E+5 ft3 Drywell particulate iodine natural deposition 10% Powers Model Drywell elemental iodine removal coefficient 0 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.01 hr-1
>24 hours 0
Leakage from drywell to containment based on core steaming 23 minutes to 1.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> 3000 cfm Flow between drywell and containment from hydrogen mixing 23 minutes to 1.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> 600 cfm 1.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> to 30 days 1.0E+8 cfm Containment leak rate 0 to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 0.325% per day 1 to 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 0.179% per day Secondary containment bypass leak rates 0 to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 0.341 cfm 1 to 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 0.188 cfm Main steam line leak rates 0 to 25 minutes 150 scfh 25 minutes to 30 days 0
Secondary containment positive pressure period 30 minutes SGTS filter efficiencies Particulate 99%
Iodine 90%
SGTS flow rates Annulus building release 2.5E+3 cfm Auxiliary building release 1.0E+4 cfm Emergency Core Cooling System leak rate 0 to 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 1 gallon per minute Iodine partition factor 10%
Control room Volume 1.88E+5 ft3 Filtered makeup air flow 1700 cfm Filtered Recirculation air flow 2000 cfm Unfiltered air inleakage rate 300 cfm Filter efficiencies Aerosol 99%
Elemental iodine 98%
Organic iodine 98%
June 2004 River Bend Station cc:
Winston & Strawn 1400 L Street, N.W.
Washington, DC 20005-3502 Manager - Licensing Entergy Operations, Inc.
River Bend Station 5485 US Highway 61N St. Francisville, LA 70775 Senior Resident Inspector P. O. Box 1050 St. Francisville, LA 70775 President of West Feliciana Police Jury P. O. Box 1921 St. Francisville, LA 70775 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064 Ms. H. Anne Plettinger 3456 Villa Rose Drive Baton Rouge, LA 70806 Mr. Michael E. Henry, State Liaison Officer Department of Environmental Quality Permits Division P.O. Box 4313 Baton Rouge, Louisiana 70821-4313 Wise, Carter, Child & Caraway P. O. Box 651 Jackson, MS 39205 Executive Vice President and Chief Operating Officer Entergy Operations, Inc.
P. O. Box 31995 Jackson, MS 39286-1995 General Manager - Plant Operations Entergy Operations, Inc.
River Bend Station 5485 US Highway 61N St. Francisville, LA 70775 Director - Nuclear Safety Entergy Operations, Inc.
River Bend Station 5485 US Highway 61N St. Francisville, LA 70775 Vice President - Operations Support Entergy Operations, Inc.
P. O. Box 31995 Jackson, MS 39286-1995 Attorney General State of Louisiana P. O. Box 94095 Baton Rouge, LA 70804-9095 Brian Almon Public Utility Commission William B. Travis Building P. O. Box 13326 1701 North Congress Avenue Austin, Texas 78701-3326