ML043440114
| ML043440114 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 01/25/2005 |
| From: | Tam P NRC/NRR/DLPM/LPD1 |
| To: | Spina J Nine Mile Point |
| References | |
| TAC MC2734 | |
| Download: ML043440114 (14) | |
Text
January 25, 2005 Mr. James A. Spina Vice President Nine Mile Point Nine Mile Point Nuclear Station, LLC P.O. Box 63 Lycoming, NY 13093
SUBJECT:
NINE MILE POINT NUCLEAR STATION, UNIT NO. 1 - ISSUANCE OF AMENDMENT RE: INTERMEDIATE RANGE MONITOR AND CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION (TAC NO. MC2734)
Dear Mr. Spina:
The Commission has issued the enclosed Amendment No. 186 to Facility Operating License No. DPR-63 for the Nine Mile Point Nuclear Station, Unit No. 1 (NMP-1). The amendment consists of changes to the Technical Specifications in response to your application transmitted by letter dated April 19, 2004, as supplemented on July 16, 2004.
The amendment revised Section 3/4.6.2, Protective Instrumentation, to establish a 24-month-operating-cycle calibration frequency for the intermediate range monitor instrumentation. In addition, the amendment authorized relocation of the limiting conditions for operation and surveillance requirements for certain control rod withdrawal block instruments from Section 3/4.6.2 to the Updated Final Safety Analysis Report.
A copy of the related Safety Evaluation is enclosed. A Notice of Issuance will be included in the Commission's next regular biweekly Federal Register notice.
Sincerely,
/RA/
Peter S. Tam, Senior Project Manager, Section 1 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-220
Enclosures:
- 1. Amendment No. 186 to DPR-63
- 2. Safety Evaluation cc w/encls: See next page
January 25, 2005 Mr. James A. Spina Vice President Nine Mile Point Nine Mile Point Nuclear Station, LLC P.O. Box 63 Lycoming, NY 13093
SUBJECT:
NINE MILE POINT NUCLEAR STATION, UNIT NO. 1 - ISSUANCE OF AMENDMENT RE: INTERMEDIATE RANGE MONITOR AND CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION (TAC NO. MC2734)
Dear Mr. Spina:
The Commission has issued the enclosed Amendment No. 186 to Facility Operating License No. DPR-63 for the Nine Mile Point Nuclear Station, Unit No. 1 (NMP-1). The amendment consists of changes to the Technical Specifications in response to your application transmitted by letter dated April 19, 2004, as supplemented on July 16, 2004.
The amendment revised Section 3/4.6.2, Protective Instrumentation, to establish a 24-month-operating-cycle calibration frequency for the intermediate range monitor instrumentation. In addition, the amendment authorized relocation of the limiting conditions for operation and surveillance requirements for certain control rod withdrawal block instruments from Section 3/4.6.2 to the Updated Final Safety Analysis Report.
A copy of the related Safety Evaluation is enclosed. A Notice of Issuance will be included in the Commission's next regular biweekly Federal Register notice.
Sincerely,
/RA/
Peter S. Tam, Senior Project Manager, Section 1 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-220
Enclosures:
- 1. Amendment No. 186 to DPR-63
- 2. Safety Evaluation cc w/encls: See next page Distribution:
PUBLIC PDI R/F R. Laufer S. Little P. Tam OGC G. Hill (2)
T. Boyce H. Li ACRS G. Matakas, RI Accession Number: ML043440114 OFFICE PDI-1\\PM PDI-1\\LA EEIB\\SC IROB OGC PDI-1/SC NAME PTam SLittle EMarino*
TBoyce SLewis PTam for RLaufer DATE 12/28/04 12/27/04 11/18/04 1/10/05 1/18/05 1/21/05
- SE transmitted by memo dated 11/18/04.
OFFICIAL RECORD COPY
NINE MILE POINT NUCLEAR STATION, LLC (NMPNS)
DOCKET NO. 50-220 NINE MILE POINT NUCLEAR STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 186 License No. DPR-63 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Nine Mile Point Nuclear Station, LLC (the licensee) dated April 19, 2004, as supplemented on July 16, 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, 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 amendment will not be inimical to the common defense and security or to the health and safety of the public; and 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. DPR-63 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A, which is attached hereto, as revised through Amendment No. 186, is hereby incorporated into this license.
Nine Mile Point Nuclear Station, LLC shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance and shall be implemented within 60 days.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
Richard J. Laufer, Chief,Section I Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: January 25, 2005
ATTACHMENT TO LICENSE AMENDMENT NO. 186 TO FACILITY OPERATING LICENSE NO. DPR-63 DOCKET NO. 50-220 Replace the following pages of 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 Pages Insert Pages 202 202 203 203 226 226 228 228 229 229 230 230 231 231 232 232 233 233
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 186 TO FACILITY OPERATING LICENSE NO. DPR-63 NINE MILE POINT NUCLEAR STATION, LLC NINE MILE POINT NUCLEAR STATION UNIT NO. 1 DOCKET NO. 50-220
1.0 INTRODUCTION
By letter dated April 19, 2004, as supplemented on July 16, 2004, Nine Mile Point Nuclear Station, LLC (the licensee) submitted an application for amendment to the Nine Mile Point Nuclear Station, Unit No. 1 (NMP1) Technical Specifications (TSs). The amendment would revise Section 3/4.6.2, Protective Instrumentation, to establish a 24-month-operating-cycle calibration surveillance frequency for the intermediate range monitor (IRM) instrumentation. In addition, the amendment would relocate the limiting conditions for operation (LCOs) and surveillance requirements (SRs) for certain control rod withdrawal block instruments from Section 3/4.6.2 to the NMP1 Updated Final Safety Analysis Report (UFSAR), a licensee-controlled document.
The July 16, 2004, supplement 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 (NRC) staffs original proposed no significant hazards consideration determination as published in the Federal Register on May 25, 2004 (69 FR 29769).
2.0 REGULATORY EVALUATION
The NRC staff used the guidance in Generic Letter (GL) 91-04, Changes in Technical Specification Surveillance Intervals to Accommodate a 24-month Fuel Cycle, dated April 2, 1991, for evaluating the acceptability of the proposed changes to the surveillance test intervals.
The criteria for including the LCOs and SRs are provided in 10 CFR 50.36, Technical Specifications. Specifically, 10 CFR 50.36(c)(2)(ii) requires that an LCO of a nuclear plant must be established for each item meeting one or more of the following criteria:
Criterion 1.
Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.
Criterion 2.
A process variable, design feature, or operating restriction that is an initial condition of a design basis accident (DBA) or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
Criterion 3.
A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
Criterion 4.
A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.
Accordingly, a licensee seeking to delete the LCO of a functional unit from the TSs must demonstrate that the functional unit no longer satisfies any of these four criteria.
3.0 TECHNICAL EVALUATION
3.1 Background
GL 91-04 provides generic guidance for amendments that change the TS surveillance intervals to accommodate a 24-month fuel cycle. In accordance with GL 91-04, the licensee should provide the following information to justify increasing the calibration intervals for instruments used to perform safety functions:
(1)
Confirm that instrument drift as determined by as-found and as-left calibration data from surveillance and maintenance records have not, except on rare occasions, exceeded acceptable limits for a calibration interval.
(2)
Confirm that the values of drift for each instrument type (make, model, and range) and application have been determined with a high probability and a high degree of confidence. Provide a summary of the methodology and assumptions used to determine the rate of instrument drift with time based upon historical plant calibration data.
(3)
Confirm that the magnitude of instrument drift has been determined with a high probability and a high degree of confidence for a bounding calibration interval of 30 months for each instrument type (make, model number, and range) and application that performs a safety function.
Provide a list of the channels by TS section that identifies these instrument applications.
(4)
Confirm that a comparison of the projected instrument drift errors has been made with the values of drift used in the setpoint analysis. If this results in revised setpoints to accommodate large drift errors, provide proposed TS changes to update trip setpoints. If the drift errors result in a revised safety analysis to support existing setpoints, provide a summary of the updated analysis conclusions to confirm that safety limits and safety analysis assumptions are not exceeded.
(5)
Confirm that the projected instrument errors caused by drift are acceptable for control of plant parameters to effect a safe shutdown with the associated instrumentation.
(6)
Confirm that all conditions and assumptions of the setpoint and safety analyses have been checked and are appropriately reflected in the acceptance criteria of plant surveillance procedures for channel checks, channel functional tests, and channel calibrations.
(7)
Provide a summary description of the program for monitoring and assessing the effects of increased calibration surveillance intervals on instrument drift and its effect on safety.
3.2 Evaluation of 24-Month IRM Calibration Frequency The licensee performed a safety assessment for the proposed changes to the surveillance test interval of the IRM instrumentation in accordance with the GL 91-04 guidance stated above.
The proposed 24-month IRM calibration frequency is supported by plant-specific analyses. To do the assessment, the licensee reviewed the historical maintenance and surveillance test data at the bounding surveillance test interval limit, performed an evaluation to ensure that a 24-month surveillance test interval would not invalidate any assumption in the plant licensing bases, and determined that the effect of the surveillance interval extension was small.
The licensee contracted with Alion Science and Technology Corporation (Alion) to statistically analyze 9 years of the NMP1 plant-specific IRM historical calibration data (April 11, 1994 to April 19, 2003) for drift. The IRM drift analysis uses the methodology of Electric Power Research Institute (EPRI) Report TR-103335, Guideline for Instrument Calibration Extension/Reduction, Revision 1, October 1998.
The Alion report concluded that the NMP1 IRM calibration data had been analyzed to determine the instruments statistically based drift values. The drift and as-found data in graphs show that the magnitude of drift is independent of the calibration interval (the as-found values are within the expected range). The analysis of IRM calibration data clearly indicates that instrument performance will support 24-month calibration intervals (30 month with the 25% grace period).
A calibration is not necessary before startup or shutdown as long as the IRM instruments are calibrated every 30 months. It is important to note that panel checks and functional tests are required to support the predicted performance of the IRM instruments.
In Attachment 1 of the April 19, 2004, application, the licensee addressed all the concerns related to GL 91-04. Attachment 6 of the application provides detailed NMP1 plant-specific IRM calibration data. Attachment 5 of the application provides the IRM setpoint calculation. The NMP1 trip setpoint methodology is based on General Electric Topical Report NEDC-31336, General Electric Instrument Setpoint Methodology, which has been approved by the NRC staff. Recently, the NRC staff has expressed concern with Method 3 of the setpoint methodology described in the Instrument Society of America (ISA) standard ISA RP67.04, Part II, however, the GE Topical Report NEDC-31336 used Method 2 of ISA RP67.04, Part II, which is acceptable to the NRC staff.
The NRC staff determined that the licensee has addressed the issues identified in GL 91-04 and has provided an acceptable basis for increasing the calibration interval for IRM instruments used to perform safety functions. On the basis of its evaluation, the NRC staff concludes that the licensee has confirmed that the safety limits and safety analysis assumptions will not be exceeded after worst-case drift is considered for the instruments for which the surveillance intervals will be extended to 24 months.
3.3 Evaluation of Proposed Technical Specification Changes The NRC staffs detailed review of the proposed TS changes is set forth below.
3.3.1 Table 4.6.2a, Parameter (9)(a): Neutron Flux - IRM Upscale Scram, and Neutron Flux -
IRM Inoperable Scram, Surveillance Requirement The function of the IRM upscale neutron flux scram is to limit maximum reactor power to prevent fuel damage in the intermediate power range. The IRM upscale scram function is required to be operable when the reactor mode switch is in the startup and refuel positions.
The IRM upscale scram provides a diverse protection function from the rod worth minimizer (RWM), which monitors and controls the movement of control rods at low power. The RWM prevents the withdrawal of an out-of-sequence control rod during startup; such a withdrawal could result in an unacceptable neutron flux excursion. The IRM upscale scram provides a diverse protection against local control rod withdrawal errors and the continuous rod withdrawal transient in the startup range, and provides backup protection for the average power range monitors (APRMs).
The function of the IRM inoperative scram is to provide assurance that a minimum number of IRMs are operable. An IRM inoperative scram signal is initiated whenever: (1) the mode switch is placed in a position other than operate; (2) the detector voltage drops below a preset level; or (3) one of the plug-in modules is removed or disconnected. This function is required to be operable when the reactor mode switch is in the startup and refuel positions. Since only one IRM channel in each trip system may be bypassed, only one channel in each trip system can be inoperable without resulting in a scram trip signal.
The current sensor check requirement per Note (f) to...check once per shift... is replaced with the equivalent tabular frequency of Once per shift(f). To conform with this change, the current Note (g) source range monitor (SRM)/IRM and IRM/APRM overlap requirements are moved to Note (f). This is consistent with the current boiling-water reactor (BWR) Standard TS requirements.
The current instrument channel test requirement per Note (f) to...test once per week... is replaced with the equivalent tabular frequency of Once per week(g). The current Note (g) requirements are moved to Note (f). Therefore, the instrument channel test frequency is modified by a new Note (g), which requires the test to be performed...within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> before startup, if not performed within the previous 7 days, and allows the performance of the test to be delayed...during shutdown until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after entering startup from run. This change is intended to reduce the burden and distraction to the operators during the start up and shutdown processes. The NRC staff believes that this change facilitates safe operation of the unit, and that it poses no undue risk to the health and safety of the public. Furthermore, this change is consistent with the current BWR Standard TS requirements.
The current instrument channel calibration requirement in Note (f) to calibrate prior to startup and normal shutdown... is replaced with the tabular frequency of Once per operating cycle(n).
The calibration frequency is modified by the addition of Note (n), which reflects the current industry practice to exclude neutron detector calibration, because there is no practical method to calibrate neutron detector other than laboratory calibration. This change is consistent with current BWR Standard TS requirements.
In summary, the instrument channel calibration frequency in TS Table 4.6.2a for the IRM upscale scram and the IRM inoperative scram are being changed from...prior to startup and normal shutdown... to Once per operating cycle. An IRM calibration is a complete check of the instrument loop (excluding the neutron detectors), and verification that each channel, including the associated actuation, alarm, and trip functions, responds with the necessary range and accuracy. As stated above in Section 3.2, the NRC staff found acceptable the proposed changes to extend the IRM calibration frequency to 24 months. The other changes related to the sensor check and the instrument channel test are consistent with the current BWR Standard TS requirements. The NRC staff finds these changes acceptable.
3.3.2 Table 4.6.2a, Parameter (9)(b)(i): Neutron Flux - APRM Upscale Parameter, Surveillance Requirement The current instrument channel calibration tabular frequency is modified by the addition of Note (n). Note (n) is added to reflect the current industry practice and is consistent with the Standard TS exclusion of the neutron detector calibration. The NRC staff agrees that it is not practical to calibrate the neutron detectors, and finds the proposed change acceptable.
3.3.3 Table 3.6.2g, Instrumentation that Initiates Control Rod Withdrawal Block, LCOs Table 4.6.2g, Instrumentation that Initiates Control Rod Withdrawal Block, SRs A.
Relocation of Parameter (1) to the UFSAR The function of the SRM control rod withdrawal block is to prevent control rod withdrawal errors during reactor startup, utilizing SRM signals to initiate the rod block signal. SRM signals are used to monitor neutron flux during refueling, shutdown, and startup conditions. The SRM performs no automatic safety function. Consequently, no design-basis accident (DBA) or transient credits the rod block signals initiated by the SRMs.
The licensee applied the screening criteria in 10 CFR 50.36(c)(2)(ii) to the Table 3/4.6.2g control rod withdrawal block LCOs for the SRM. The licensee did the evaluation by comparing the LCOs to the screening criteria, and determined that the SRM instrumentation LCOs do not satisfy any of the four screening criteria for retention in 10 CFR 50.36(c)(2)(ii). Accordingly, the licensee proposed to relocate the SRM LCOs and associated SRs to the NMP1 UFSAR. The relocation results in future changes to the LCOs and SRs being controlled by the 10 CFR 50.59 process.
The NRC staff has reviewed the licensees justification for this relocation. The NRC staff concurs with the licensees justification, and finds that the proposed change is consistent with the current BWR Standard TS, in conformance with requirements in 10 CFR 50.36(c)(2)(ii), and therefore, is acceptable.
B.
Relocation of Parameter (4) to the UFSAR An increase in reactor recirculation flow causes an increase in neutron flux, which increases reactor power. The increase in neutron flux is monitored by the neutron monitoring system, which initiates a reactor protection system (RPS) scram if the trip setpoints are reached.
However, no DBA or transient credits the rod block signals initiated by the recirculation flow instrument.
The licensee applied the screening criteria in 10 CFR 50.36(c)(2)(ii) to this parameter. The licensee did the evaluation by comparing the LCOs to the screening criteria, and determined that the recirculation flow instrumentation LCOs do not satisfy any of the four screening criteria for retention in 10 CFR 50.36(c)(2)(ii). Accordingly, the licensee proposed to relocate the recirculation flow LCOs and associated SRs to the UFSAR. The relocation results in future changes to the LCOs and SRs being controlled by the 10 CFR 50.59 process.
The NRC staff has reviewed the licensees justification for this relocation. The NRC staff concurs with the licensees justification, and finds that the proposed change is consistent with the current BWR Standard TSs, in conformance with requirements in 10 CFR 50.36(c)(2)(ii),
and therefore, is acceptable.
C.
Relocation of Parameter (8) to the UFSAR The scram dump volume (SDV) water level scram bypass control rod withdrawal block functions to prevent control rod withdrawals during resetting of an RPS-initiated scram. An RPS scram closes the SDV drain and vent valves. During a scram, the SDV partially fills with water that will cause an SDV high water level scram signal to be initiated. The SDV high-level sensors must be bypassed to allow resetting of the scram signal. This in turn allows the drain and vent valves to be opened to drain the SDV. The sensors can only be bypassed if the reactor mode switch is in the refuel or shutdown position. Thus, the SDV water level scram bypass control rod block is redundant to the shutdown and refuel mode switchs control rod withdrawal blocks. No DBA or transient credits the rod block signals initiated by the SDV water level scram bypass instrumentation.
The licensee applied the screening criteria in 10 CFR 50.36(c)(2)(ii) to this parameter. The licensee did the evaluation by comparing the LCOs to the screening criteria, and determined that the SDV volume water level scram bypass instrumentation LCOs do not satisfy any of the four screening criteria for retention in 10 CFR 50.36(c)(2)(ii). Accordingly, the licensee proposed to relocate this parameter to the UFSAR. The relocation results in future changes to the LCOs and SRs being controlled by the 10 CFR 50.59 process.
The NRC staff has reviewed the licensees justification for this relocation. The NRC staff concurs with the licensees justification, and finds that the proposed change is consistent with the current BWR Standard TSs, in conformance with requirements in 10 CFR 50.36(c)(2)(ii),
and therefore, is acceptable.
3.3.4 Table 4.6.2g, Instrumentation that Initiates Control Rod Withdrawal Block, SRs The licensee proposed to modify Parameter (2), regarding IRM detector not in startup position, inoperative, downscale, and upscale. The current instrument channel test requirements per Note (g) specifies...test prior to startup and normal shutdown and...once per week... and
...test once per week... are to be replaced with the equivalent tabular frequency of Once per week. To conform with this change and the other IRM parameter changes, the current Note (g) calibration and testing requirements are replaced with a new Note (g) which requires the instrument channel test to be performed...within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> before startup, if not performed within the previous 7 days, and allows the performance of the test to be delayed...during shutdown until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after entering startup from run. The current instrument channel calibration in Note (g) to Calibrate prior to startup and normal shutdown... is replaced with the tabular frequency of Once per operating cycle(j). The calibration frequency is modified by the addition of a new Note (j) which reflects the current industry practice of excluding calibration of neutron detectors.
The licensee proposed to modify Parameter (3), regarding the APRM, by the addition of the new Note (j) which reflect the current industry practice of excluding calibration of neutron detectors.
The NRC staff finds these changes acceptable for the reasons given in Section 3.3.1 above.
3.4 Summary of Review Based on the review of the licensees submittals referenced above, the NRC staff finds that the proposed amendment to establish an operating cycle (24-month) calibration surveillance frequency for the IRM instrumentation is in conformance with guidance in the GL 91-04 and, therefore, is acceptable. The licensees proposed relocation of the LCOs and SRs for certain control rod withdrawal block instrumentation from Section 3/4.6.2 to the NMP1 UFSAR is consistent with the current BWR Standard TS requirements, is in conformance with 10 CFR 50.36(c)(2)(ii), and is also acceptable.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the New York State official was notified of the proposed issuance of the amendment. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes surveillance requirements. 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 (69 FR 29769). 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 amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor: H. Li Date: January 25, 2005
Nine Mile Point Nuclear Station, Unit No. 1 cc:
Mr. Michael J. Wallace President Nine Mile Point Nuclear Station, LLC c/o Constellation Energy Group 750 East Pratt Street Baltimore, MD 21202 Mr. Mike Heffley Senior Vice President and Chief Nuclear Officer Constellation Generation Group 1997 Annapolis Exchange Parkway Suite 500 Annapolis, MD 21401 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Resident Inspector U.S. Nuclear Regulatory Commission P.O. Box 126 Lycoming, NY 13093 Charles Donaldson, Esquire Assistant Attorney General New York Department of Law 120 Broadway New York, NY 10271 Mr. Paul D. Eddy Electric Division NYS Department of Public Service Agency Building 3 Empire State Plaza Albany, NY 12223 Mr. Peter R. Smith, President New York State Energy, Research, and Development Authority 17 Columbia Circle Albany, NY 12203-6399 Mark J. Wetterhahn, Esquire Winston & Strawn 1400 L Street, NW Washington, DC 20005-3502 Supervisor Town of Scriba Route 8, Box 382 Oswego, NY 13126 Mr. James M. Petro, Jr., Esquire Counsel Constellation Energy Group, Inc.
750 East Pratt Street, 5th Floor Baltimore, MD 21202 Ms. Deb Katz, Executive Director Nuclear Security Coalition c/o Citizens Awareness Network P.O. Box 83 Shelburne Falls, MA 01370