ML062200114
| ML062200114 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 09/01/2006 |
| From: | Geoffrey Miller NRC/NRR/ADRO/DORL/LPLI-2 |
| To: | Crane C AmerGen Energy Co |
| Miller G, NRR/DLPM, 415-2481 | |
| Shared Package | |
| ML062200104 | List: |
| References | |
| TAC MC8761 | |
| Download: ML062200114 (13) | |
Text
September 1, 2006 Mr. Christopher M. Crane President and Chief Executive Officer AmerGen Energy Company, LLC 4300 Winfield Road Warrenville, IL 60555
SUBJECT:
OYSTER CREEK NUCLEAR GENERATING STATION - ISSUANCE OF AMENDMENT RE: REVISION TO ELECTROMATIC RELIEF VALVE SURVEILLANCE REQUIREMENT (TAC NO. MC8671)
Dear Mr. Crane:
The Commission has issued the enclosed Amendment No. 260 to Facility Operating License No. DPR-16 for the Oyster Creek Nuclear Generating Station (Oyster Creek), in response to your application dated October 18, 2005, as supplemented by letter dated May 26, 2006.
The amendment revises the Oyster Creek Technical Specifications Surveillance Requirement (SR) 4.4.B.1 to provide an alternative means for testing the electromatic relief valves located on the main steam system. The revised SR allows demonstration of the capability of the valves to perform their function without requiring that the valves be cycled with steam pressure while installed.
A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commissions biweekly Federal Register notice.
Sincerely,
/RA/
G. Edward Miller, Project Manager Plant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-219
Enclosures:
- 1. Amendment No. 260 to DPR-16
- 2. Safety Evaluation cc w/encls: See next page
September 1, 2006 Mr. Christopher M. Crane President and Chief Executive Officer AmerGen Energy Company, LLC 4300 Winfield Road Warrenville, IL 60555
SUBJECT:
OYSTER CREEK NUCLEAR GENERATING STATION - ISSUANCE OF AMENDMENT RE: REVISION TO ELECTROMATIC RELIEF VALVE SURVEILLANCE REQUIREMENT (TAC NO. MC8671)
Dear Mr. Crane:
The Commission has issued the enclosed Amendment No. 260 to Facility Operating License No. DPR-16 for the Oyster Creek Nuclear Generating Station (Oyster Creek), in response to your application dated October 18, 2005, as supplemented by letter dated May 26, 2006.
The amendment revises the Oyster Creek Technical Specifications Surveillance Requirement (SR) 4.4.B.1 to provide an alternative means for testing the electromatic relief valves located on the main steam system. The revised SR allows demonstration of the capability of the valves to perform their function without requiring that the valves be cycled with steam pressure while installed.
A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commissions biweekly Federal Register notice.
Sincerely,
/RA/
G. Edward Miller, Project Manager Plant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-219
Enclosures:
- 1. Amendment No. 260 to DPR-16
- 2. Safety Evaluation cc w/encls: See next page DISTRIBUTION:
PUBLIC RidsOgcRp RidsNrrDorl LPLI-2 R/F GHill (2)
RidsNrrDorlLpl1-2 RidsNrrPMGMiller RidsNrrLACRaynor RidsAcrsAcnwMailCenter RidsRgn1MailCenter Package Accession Number: ML062200104 Amendment Accession Number: ML062200114 TS(s) Accession Number: ML062440411 OFFICE LPLI-2/PM LPLI-2/LA DCI/CPTB/BC(A)
OGC LPLI-2/BC (A)
NAME GEMiller:rsa CRaynor TLiu MYoung BPoole DATE 8/31/06 8/30/06 7/28/06 8/24/06 8/31/06 OFFICIAL RECORD COPY
AMERGEN ENERGY COMPANY, LLC DOCKET NO. 50-219 OYSTER CREEK NUCLEAR GENERATING STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 260 License No. DPR-16 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by AmerGen Energy Company, LLC, (the licensee), dated October 18, 2005, as supplemented by letter dated May 26, 2006, 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 Commissions 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 Commissions regulations and all applicable requirements have been satisfied.
2.
Accordingly, the Facility Operating License is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment.
(2)
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 260, are hereby incorporated in the license. AmerGen Energy Company, 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 of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
Brooke D. Poole, Acting Chief Plant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: September 1, 2006
ATTACHMENT TO LICENSE AMENDMENT NO. 260 FACILITY OPERATING LICENSE NO. DPR-16 DOCKET NO. 50-219 Replace the following pages of Appendix A, Technical Specifications, with the attached pages.
The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Remove Insert 4.4-1 4.4-1 4.4-3 4.4-3
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 260 TO FACILITY OPERATING LICENSE NO. DPR-16 AMERGEN ENERGY COMPANY, LCC OYSTER CREEK NUCLEAR GENERATING STATION DOCKET NO. 50-219
1.0 INTRODUCTION
By application dated October 18, 2005, as supplemented by letters dated May 26, 2006, AmerGen Energy Company, LLC, (AmerGen or the licensee) requested changes to the Facility Operating License for the Oyster Creek Nuclear Generating Station (Oyster Creek). The proposed amendment would revise the Oyster Creek Technical Specifications (TSs)
Surveillance Requirement (SR) 4.4.B.1 to provide an alternative means for testing the electromatic relief valves located on the main steam system. The revised SR would allow demonstration of the capability of the valves to perform their function without requiring that the valves be cycled with steam pressure while installed.
The supplement dated May 26, 2006, 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 or the Commission) staffs original proposed no significant hazards determination as published in the Federal Register, on December 20, 2005 (70 FR 75490).
2.0 REGULATORY EVALUATION
Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36, Technical Specifications, provides the regulatory requirements for the content required in a licensees TSs. Criterion 3 of 10 CFR 50.36(c)(2)(ii) requires a limiting condition for operation (LCO) to be established for 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. Section 50.36(c)(3) of 10 CFR specifies, among other things, that SRs should ensure that LCOs are met.
The NRC has previously approved similar TS changes at LaSalle, Clinton, Peach Bottom, Dresden, and Quad Cities nuclear power facilities.
3.0 TECHNICAL EVALUATION
3.1 Description of Proposed Change Currently, a footnote to SR 4.4.B.1 requires the valve operability tests to be performed at system operating pressure prior to exceeding 5% rated thermal power following a refueling outage. This test entails a manually-actuated exercise of the valve that utilizes the main steam pressure to lift the valve disc. The proposed revision would delete the footnote and revise SR 4.4.B.1 to read as follows:
Verify each relief valve actuator strokes when manually actuated.
The proposed change would allow a demonstration of valve operability by manually stroking the relief valve actuator during an outage every 24 months without lifting the main valve disc. The main valve disc would be tested as required by the Oyster Creek Inservice Testing (IST)
Program.
Additionally, the Bases for this SR would be revised to read as follows:
The operability of the Electromatic Relief Valves (EMRVs) is verified by a stroke test of its relief valve actuator as specified in TS 4.4.B.1, and by the Inservice Testing Program (IST).
The EMRV actuator stroke test is performed with the pilot valve actuator mounted in its normal position. The test checks the manual actuation electrical circuitry, solenoid actuator, pilot operating lever, and pilot valve assembly. This verifies pilot valve movement. However, since this test is performed prior to establishing the reactor pressure needed to overcome the main valve closure spring force, the main valve will not stroke during the test, thereby minimizing the potential for valve leakage.
3.2 AmerGens Basis for the Proposed Change AmerGen stated that there are five Dresser Model 1525VX EMRVs on the main steamlines between the reactor pressure vessel (RPV) and the main steamline isolation valve within the drywell. The EMRVs consist of a main valve assembly, pilot valve assembly, and a solenoid actuator. The EMRVs are opened by automatic or manual switch actuation of a solenoid actuator. When energized, the solenoid actuates the plunger, which pushes down the pilot valve operating level, thereby opening the pilot valve. When the pilot valve opens, pressure under the main valve disc is vented. This results in an unbalanced steam pressure across the main disc, which moves the main disc downward from its seat, opening the main valve.
The function of the EMRVs is described in the Oyster Creek Updated Final Safety Analysis Report (UFSAR), Section 6.3.1.2. The EMRVs are part of the automatic depressurization system (ADS), which supports the emergency core cooling system. The ADS is designed to depressurize the reactor during a small-break loss-of-coolant accident to permit the low-pressure core spray (CS) system to inject water into the reactor core. The EMRVs are actuated by simultaneous occurrence of triple low reactor water level, high drywell pressure, and indication that a CS booster pump has started and developed adequate differential pressure. The EMRVs also provide overpressure protection for the RPV as discussed in UFSAR Section 6.3.1.2. In the overpressure mode, the EMRVs are actuated by pressure switches that monitor RPV pressure.
AmerGen stated that operating experience at Oyster Creek has indicated that manual actuation of the EMRVs during plant operation can lead to main and pilot valve seat leakage. Leakage through the main valve results in increased suppression pool temperature and level, and pilot valve leakage results in unidentified drywell leakage.
AmerGens proposed revision to the SR provides an alternative method of verifying EMRV operability. Currently, SR 4.4.B.1 requires an operability test utilizing the main steam pressure to stroke the EMRV. The proposed alternative test would verify EMRV operability by stroking the EMRV actuator using the manual switch every 24 months. The alternative test would be performed with little or no reactor pressure on a 24-month frequency. Stroke testing of the EMRV main valve will continue to be performed in accordance with the IST program.
Additionally, each valve is removed, refurbished, and stroke tested every two refueling outages.
The licensee further stated that stroking of the EMRV actuator every 24 months, in combination with the removal, refurbishment, and stroke testing of the valve every two refueling outages will provide a complete verification of the EMRV functional capability.
The alternative test would be performed with the solenoid actuator mounted in its normal position. This would allow testing of the manual actuation electrical circuitry, solenoid actuator, pilot operating level, and pilot valve assembly. This test would verify pilot valve movement.
However, since this test will be performed in the absence of the reactor pressure needed to overcome the main valve closure spring force, the main valve will not stroke during the test.
Stroking the pilot valve in the absence of steam pressure is referred to as dry cycling.
AmerGen also stated that Nine Mile Point 1, Dresden Nuclear Power Station, Units 1 and 2, and Quad Cities Nuclear Power Station, Units 1 and 2, also utilize the Dresser EMRVs, model 1525VX. As described in an NRC Inspection Report dated December 22, 2000, Nine Mile Point Unit 1 experienced a spurious opening and failure to re-close for one of their EMRVs. As stated in the Inspection Report, the utility concluded that the event was probably caused by a pilot valve bent stem and partial disc-stem separation. The utility further concluded that dry cycling of EMRV pilot valves may result in partial disc-stem separation.
AmerGen stated that the proposed Oyster Creek valve actuator testing would include manual dry cycling of the pilot valve to verify that the stem travel and level arm adjusting screw gap are within limits. Following this verification, the EMRV solenoid would be energized to stroke the pilot valve. The stem travel and lever arm adjusting screw gap would then be rechecked to verify that these parameters are within limits following the dry cycling. The licensee stated that partial disc separation caused by dry cycling of the pilot valve would be detected during this re-check. The EMRV manufacturer, Dresser, also confirmed that this re-check would detect partial disk-stem separation caused by the dry cycling of the pilot valve. Additionally, dry cycling of the pilot valves has been performed on the EMRVs at the aforementioned stations for many years, with no signs of partial or full disc detachment.
Regarding the potential for a pilot valve bent stem, the licensee stated that the maintenance procedures for the EMRV pilot valves would include appropriate inspections of the stem, pilot valve bushing, and disc to identify any nicks, gouges, or other damage that could impair free movement. The procedure would check the gap at the end of the stem that has the thinnest cross section. The licensee determined that this is the area most likely to be bent if not properly handled. In addition, free movement of the stem in the bushing and of the disc-to-stem connection would be checked. The licensee stated that this check would assure that the stem is straight, the pilot can travel freely, and the pilot disc can seat properly.
Another Nine Mile Point 1 event, as described in NRC Event Notification 39779, was a failure of an EMRV to open when actuated. The failure was reportedly due to inadequate solenoid force, caused by high resistance in the cutout switch, such that the output force was not adequate to overcome the pilot spring force. The licensees proposed actuator testing for the Oyster Creek EMRVs will include manual actuation of the electrical circuitry, solenoid actuator, pilot operating lever, and pilot valve assembly. AmerGen stated that this test would demonstrate that the solenoid force is adequate to overcome the pilot spring force.
Recent operating experience at Quad Cities indicated that significant degradation of EMRVs and their actuators had occurred after only a few months of plant operation due to flow-induced vibration of the main steamlines at extended power uprate (EPU) conditions. Some parts of the EMRVs and their actuators were found severely degraded, such that they would not have performed their safety function. In response to NRC staff questions regarding the applicability of this operating experience at Oyster Creek, AmerGen reviewed the root cause of the Quad Cities failures and determined that Oyster Creek has not been licensed to operate at EPU conditions that could result in changes to main steamline vibrations similar to Quad Cities.
The licensee also found no significant wear of the Oysteer Creek EMRV actuators over the past 6-year period, which included three refueling outages, and found no vibration-induced degradation that would prevent the Oyster Creek EMRVs from remaining functional for the proposed 24-month interval between surveillances. Therefore, the licensee concluded that additional instrumentation to determine current vibration levels or modifications to the EMRVs to address excessive vibration was not necessary.
AmerGen stated that the relief valves will continue to be tested in accordance with the Oyster Creek IST Program, fourth 10-year interval, as required by TS 4.3.C. The current IST program for relief valves is based on the American Society of Mechanical Engineers Code for Operation and Maintenance of Nuclear Power Plants (ASME OM Code), 1995 Edition through OMa-1996 Addenda. As required by Appendix I,Section I 1.3.3, Class 1 PRVs are tested at least once every 5 years, with a minimum of 20% of the valves tested within any 24-month interval. This means that two of the five EMRVs will be tested every 24 months, with the other three EMRVs on the subsequent refueling outage. The licensee states that this would be accomplished by replacing the installed valves with new or refurbished valves that have been pre-tested. The ASME OM Code test would be performed at a steam test facility, where the valve (i.e., the main valve and pilot valve) and an actuator representative of the actuator used at the plant would be installed on a steam header in the same orientation as the plant installation. The test conditions in the test facility would be similar to those in the plant installation, including valve body temperature and steam conditions. The valve would then be leak tested and functionally tested to ensure the valve is capable of opening and closing (including stroke time), and leak tested a final time. Valve seat tightness would be verified by a cold bar test, and if not free of fog, leakage would be measured and verified to be below design limits. The licensee states that the storage requirements in effect ensure the valves are protected from physical damage and that, prior to installation and being electrically connected, the valves would again be inspected for foreign material and damage.
As part of the preventive maintenance program during each refueling outage, the licensee replaces the pilot valve assemblies in the EMRVs that are not scheduled for removal and testing. The replacement of the pilot valve assemblies does not involve removal of the EMRVs and does not affect the main valve disc. Additionally, all five solenoid actuators are refurbished on a refueling outage basis. Following replacement of the pilot valve assemblies and installation of the refurbished EMRVs, the proposed SR would require testing of the EMRV actuator without stroking the main valve. The licensee further states that this SR would ensure that the affected portion of the valve will be fully tested and that if other maintenance is performed, controls regarding testing requirements following maintenance ensure that appropriate post maintenance testing is performed. For example, if maintenance is performed that affects the main valve, the capability of the main valve would be tested at the testing facility or on the installed valve at the plant.
The maintenance is also performed on the solenoid actuator with specific attention given to maintenance and testing of the cutout contacts. The contacts are cleaned, the associated springs and mechanisms are inspected, and as-left contact resistances are verified.
Resistance checks and meggar tests are performed on both coils. During electrical actuation, operating currents are verified to be within acceptance criteria limits. The licensee states that these steps provide substantial indication that the solenoid operator is capable of functioning as designed.
The solenoid actuator is designed to operate the pilot valve under design conditions. The actuator includes two coils. One coil can be considered a pull-in coil and the second considered a hold-in coil. The pull-in coil provides sufficient force to actuate the pilot, while the hold-in coil provides sufficient force to maintain the pilot in an open position. Contacts designated as cutout contacts control the energization of these coils during solenoid motion.
AmerGen concluded that the combination of the test using steam at a test facility, and the proposed valve actuator testing at the site, will provide a complete check of the capability of the valves to open and close. Further, the licensee concluded that the proposed changes provide for the testing of the EMRVs such that full functionality is demonstrated through overlapping tests, without cycling the valves under steam pressure with the vavles installed. This approach would reduce the potential for valve seat leakage and the proposed alternative test for the EMRVs reflects the recommendations of NUREG-0737, Clarification of TMI [Three Mile Island]
Action Plan Requirements, Item II.K.3.16, Reduction of Challenges and Failures of Relief Valves, that the number of relief valve openings be reduced as much as possible, and unnecessary challenges should be avoided.
The licensee did indicate that the proposed EMRV actuator test, which avoids the discharge of steam through the valve discharge piping to the suppression pool, eliminates confirmation of discharge pipe blockage. AmerGen stated that, as implemented at Oyster Creek, the Foreign Material Exclusion (FME) program provides the necessary requirements and guidance to prevent and control introduction of foreign materials into structures, systems, and components and minimizes the potential for debris blocking a relief valve discharge line. The licensee concluded that, considering the size of the discharge pipe (8 inches), the energy associated with high-pressure steam, and the FME program, the probability of blocking a relief valve discharge line and preventing the valve function is remote.
As a result of deleting the requirement for full functional testing of the EMRVs and replacing these requirements with the proposed SR, the only change in the frequency of testing is that the main valve disc of the EMRVs will be lift tested every two operating cycles (approximately four years) compared to the current one operating cycle (approximately two years). The licensee stated that a review of the surveillance testing results for the past 10 years at Oyster Creek was performed for the EMRVs and checked for any failures of the main valve disc to stroke open. Based on this review, the licensee concluded that no failures of the valves to lift have occurred in the past 10 years and, therefore, extending the frequency of checking the function of the main disc as described is not expected to result in additional valve failures.
3.3
NRC Staff Evaluation
The NRC staff has reviewed AmerGens basis for the proposed TS change and finds that with the proposed testing, the functional capability of the EMRVs is adequately verified. A manual actuation and valve leakage test will be performed at a steam test facility using the conditions similar to those for the installed valves in the plant, including valve orientation, ambient temperature, valve insulation, and steam conditions. Following EMRV installation, the licensees proposed testing includes verifying electrical supply connections and actuator performance. It is noted that, although the tests of the EMRVs at the steam test facility are not performed with the actual valve solenoids at Oyster Creek, the In-Plant EMRV solenoids are adequately tested and verified by actuation during the proposed surveillance. Further, the NRC staff finds that the licensee has adequately considered the applicable operating experience regarding the necessary verification and testing of the EMRV solenoid capability and the prevention and detection of possible damage to the EMRV pilot valves during the proposed dry stroke testing following installation. Additionally, the licensee has adequately considered the operating experience regarding possible degradation due to excessive vibration.
Therefore, the NRC staff finds that all of the components necessary to actuate the EMRVs will continue to be tested as necessary to demonstrate the functional capability of the valves, without the need to stroke test the valves on-line with system pressure conditions. In addition, the NRC staff finds that the current testing requirements could result in seat leakage of the EMRVs during power operation. Excessive seat leakage could result in excessive suppression pool temperature and level, or in unidentified drywell leakage.
The NRC staff also finds that although the proposed alternative does not verify the absense of foreign material by passing steam, the licensees FME program can prevent and control introduction of foreign bodies and minimize the potential for debris blockage. Thus, the FME provides reasonable assurance that the EMRV discharge lines would remain unblocked and that foreign material would not interfere with valve operation.
The test frequency of the EMRV main valve discs from every cycle to every two cycles is acceptable since the licensee has had no failures of the valves to stroke open in the past 10 years. The maintenance and testing programs described by the licensee provide reasonable assurance that the operational problems would be detected and that corrective action would be taken.
Therefore, the NRC staff finds that the licensees proposed revision to SR 4.4.B.1 for stroke testing the Oyster Creek EMRVs is acceptable. Additionally, the NRC staff reviewed the proposed Bases changes and does not object to their inclusion in the TSs. The bases pages will be adequately controlled by AmerGens Bases Control Program.
4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the New Jersey State official was notified of the proposed issuance of the amendment. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
This 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 effluent that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has made a final finding that the amendment involves no significant hazards consideration. 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: C. G. Hammer Date: September 1, 2006
Oyster Creek Nuclear Generating Station Site Vice President - Oyster Creek Nuclear Generating Station AmerGen Energy Company, LLC P.O. Box 388 Forked River, NJ 08731 Senior Vice President of Operations AmerGen Energy Company, LLC 200 Exelon Way, KSA 3-N Kennett Square, PA 19348 Kathryn M. Sutton, Esquire Morgan, Lewis, & Bockius LLP 1111 Pennsylvania Avenue, NW Washington, DC 20004 Kent Tosch, Chief New Jersey Department of Environmental Protection Bureau of Nuclear Engineering CN 415 Trenton, NJ 08625 Vice President - Licensing and Regulatory Affairs AmerGen Energy Company, LLC 4300 Winfield Road Warrenville, IL 60555 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406-1415 Mayor of Lacey Township 818 West Lacey Road Forked River, NJ 08731 Senior Resident Inspector U.S. Nuclear Regulatory Commission P.O. Box 445 Forked River, NJ 08731 Director - Licensing and Regulatory Affairs AmerGen Energy Company, LLC Correspondence Control P.O. Box 160 Kennett Square, PA 19348 Manager Licensing - Oyster Creek Exelon Generation Company, LLC Correspondence Control P.O. Box 160 Kennett Square, PA 19348 Regulatory Assurance Manager Oyster Creek AmerGen Energy Company, LLC P.O. Box 388 Forked River, NJ 08731 Assistant General Counsel AmerGen Energy Company, LLC 200 Exelon Way Kennett Square, PA 19348 Ron Bellamy, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406-1415 Correspondence Control Desk AmerGen Energy Company, LLC 200 Exelon Way, KSA 1--1 Kennett Square, PA 19348 Oyster Creek Nuclear Generating Station Plant Manager AmerGen Energy Company, LLC P.O. Box 388 Forked River, NJ 08731