ML070650535
| ML070650535 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 03/28/2007 |
| From: | Pickett D NRC/NRR/ADRO/DORL/LPLI-1 |
| To: | Korsnick M Ginna |
| david marshall NRR/DORL 415-1547 | |
| References | |
| TAC MD0686 | |
| Download: ML070650535 (14) | |
Text
March 28, 2007 Mrs. Mary G. Korsnick Vice President R.E. Ginna Nuclear Power Plant R.E. Ginna Nuclear Power Plant, LLC 1503 Lake Road Ontario, NY 14519
SUBJECT:
R.E. GINNA NUCLEAR POWER PLANT - AMENDMENT RE: TECHNICAL SPECIFICATION 3.5.1 REGARDING EMERGENCY CORE COOLING SYSTEM ACCUMULATORS (TAC NO. MD0686)
Dear Mrs. Korsnick:
The Commission has issued the enclosed Amendment No. 101 to Renewed Facility Operating License No. DPR-18 for the R.E. Ginna Nuclear Power Plant. This amendment is in response to your application dated March 28, 2006, as supplemented by letter dated October 24, 2006.
The amendment revises Technical Specification Surveillance Requirement 3.5.1.4 to change the method and frequency for verifying emergency core cooling system accumulator boric acid concentration for the R.E. Ginna Nuclear Power Plant.
A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commission's biweekly Federal Register notice.
Sincerely,
/RA/
Douglas V. Pickett, Senior Project Manager Plant Licensing Branch I-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-244
Enclosures:
- 1. Amendment No.101 to Renewed License No. DPR-18
- 2. Safety Evaluation cc w/encls: See next page
March 28, 2007 Mrs. Mary G. Korsnick Vice President R. E. Ginna Nuclear Power Plant R.E. Ginna Nuclear Power Plant, LLC 1503 Lake Road Ontario, NY 14519
SUBJECT:
R.E. GINNA NUCLEAR POWER PLANT - AMENDMENT RE: TECHNICAL SPECIFICATION 3.5.1 REGARDING EMERGENCY CORE COOLING SYSTEM ACCUMULATORS (TAC NO. MD0686)
Dear Mrs. Korsnick:
The Commission has issued the enclosed Amendment No. 101 to Renewed Facility Operating License No. DPR-18 for the R.E. Ginna Nuclear Power Plant. This amendment is in response to your application dated March 28, 2006, as supplemented by letter dated October 24, 2006.
The amendment revises Technical Specification Surveillance Requirement 3.5.1.4 to change the method and frequency for verifying emergency core cooling system accumulator boric acid concentration for the R.E. Ginna Nuclear Power Plant.
A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commission's biweekly Federal Register notice.
Sincerely,
/RA/
Douglas V. Pickett, Senior Project Manager Plant Licensing Branch I-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-244
Enclosures:
- 1. Amendment No.101 to Renewed License No. DPR-18
- 2. Safety Evaluation cc w/encls: See next page Package No.: ML Amendment No.: ML070650535 Tech Spec No.: ML NRR-058 OFFICE LPLI-1\\PE LPLI-1\\PM LPLI-1\\LA SNPB\\BC APLB\\BC CGSB\\BC OGC LPLI-1\\BC(A)
ITSB/BC NAME MDavid DPickett SLittle JNakoski LMrowca*
AHiser SHamrick JBoska TKobetz DATE 3/14
/07 3/14
/07 3/14/07 3/14/07 1/26/07 3/8/07 3/22/07 3/26/07 3/26/07
- SE input provided by memo on date shown Official Record Copy
DATED: March 28, 2007 AMENDMENT NO. 101 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-18 R.E. GINNA NUCLEAR POWER PLANT PUBLIC LPLI-1 R/F Acting BC RidsNrrDorlLpl1-1 S. Little RidsNrrLASLittle D. Pickett RidsNrrPMDPickett T. Kobetz RidsNrrDirsItsb J. Nakoski RidsNrrDssSnpb L. Mrowca RidsNrrDraAplb A. Hiser RidsNrrDciCgsb B. Singal RidsNrrDpr M. David G. Hill (2)
OGC RidsOgcMailCenter ACRS RidsAcrsAcnwMailCenter B. McDermott, RI RidsRgn1MailCenter cc: Plant Service list
R.E. Ginna Nuclear Power Plant cc:
Mr. Michael J. Wallace President R.E. Ginna Nuclear Power Plant, LLC c/o Constellation Energy 750 East Pratt Street Baltimore, MD 21202 Mr. John M. Heffley Senior Vice President and Chief Nuclear Officer Constellation Generation Group 1997 Annapolis Exchange Parkway Suite 500 Annapolis, MD 21401 Kenneth Kolaczyk, Sr. Resident Inspector R.E. Ginna Nuclear Power Plant U.S. Nuclear Regulatory Commission 1503 Lake Road Ontario, NY 14519 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Peter R. Smith, President New York State Energy, Research, and Development Authority 17 Columbia Circle Albany, NY 12203-6399 Carey W. Fleming, Esquire Senior Counsel - Nuclear Generation Constellation Generation Group, LLC 750 East Pratt Street, 17th Floor Baltimore, MD 21202 Charles Donaldson, Esquire Assistant Attorney General New York Department of Law 120 Broadway New York, NY 10271 Ms. Thelma Wideman, Director Wayne County Emergency Management Office Wayne County Emergency Operations Center 7336 Route 31 Lyons, NY 14489 Ms. Mary Louise Meisenzahl Administrator, Monroe County Office of Emergency Preparedness 1190 Scottsville Road, Suite 200 Rochester, NY 14624 Mr. Paul Eddy New York State Department of Public Service 3 Empire State Plaza, 10th Floor Albany, NY 12223
R.E. GINNA NUCLEAR POWER PLANT, LLC DOCKET NO. 50-244 R.E. GINNA NUCLEAR POWER PLANT AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 101 Renewed License No. DPR-18 1.
The Nuclear Regulatory Commission (the Commission or the NRC) has found that:
A.
The application for amendment filed by the R.E. Ginna Nuclear Power Plant, LLC (the licensee) dated March 28, 2006, as supplemented by letter dated October 24, 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 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 Renewed Facility Operating License No. DPR-18 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 101, are hereby incorporated in the license. The licensee 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 45 days.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
Douglas V. Pickett, Acting Chief Plant Licensing Branch I-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the License and Technical Specifications Date of Issuance: March 28, 2007
ATTACHMENT TO LICENSE AMENDMENT NO. 101 RENEWED FACILITY OPERATING LICENSE NO. DPR-18 DOCKET NO. 50-244 Replace the following page of the Facility Operating License with the attached revised page.
The revised page is identified by amendment number and contains marginal lines indicating the areas of change.
Remove Insert 3
3 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 3.5.1-1 3.5.1-1 3.5.1-2 3.5.1-2
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 101 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-18 R.E. GINNA NUCLEAR POWER PLANT, LLC R.E. GINNA NUCLEAR POWER PLANT DOCKET NO. 50-244
1.0 INTRODUCTION
By letter dated March 28, 2006 [1] (Agencywide Documents Access and Management System (ADAMS) Accession No. ML060940329), as supplemented by letter dated October 24, 2006 (ADAMS Accession No. ML063030027), R.E. Ginna Nuclear Power Plant, LLC (the licensee) submitted a request to change Technical Specification (TS) Surveillance Requirement (SR) 3.5.1.4, related to verifying emergency core cooling system (ECCS) accumulator boric acid concentration for the R.E. Ginna Nuclear Power Plant (Ginna). In order to meet the current SR, samples from one of the two ECCS accumulators must be taken every 31 days on a staggered test basis, resulting in sampling frequency of once every 2 months for each individual accumulator. In the proposed SR, each ECCS accumulator will be sampled every 6 months, and water leakage into the accumulators will be monitored every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The licensee provided engineering analyses and risk insights in support of the requested change. In addition, the licensee stated that the revised SR will result in reduction of approximately 0.2 REM per year in exposure to the operators taking accumulator fluid samples.
The supplemental letter dated October 24, 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) staffs original proposed no significant hazards consideration determination as published in the Federal Register on April 25, 2006 (71 FR 23960).
2.0 REGULATORY EVALUATION
The NRC staff evaluated the licensees request utilizing Standard Review Plan (SRP) Section 6.3, Emergency Core Cooling System. The acceptance criteria in SRP 6.3 that pertain to boron concentration in the ECCS accumulators comprise the following General Design Criteria (GDC):
GDC 27, Combined reactivity control systems capability, the reactivity control systems shall be designed to have a combined capability, in conjunction with poison addition by the ECCS, to maintain the capability to cool the core with appropriate margin for stuck rods.
GDC 35, Emergency core cooling, requires that a system to provide abundant emergency core cooling shall be provided.
GDC 36, Inspection of emergency core cooling system, requires that the ECCS shall be designed to permit appropriate periodic inspection of important components.
GDC 37, Testing of emergency core cooling system, requires that the ECCS shall be designed to permit appropriate periodic pressure and functional testing.
The licensee provided risk insights but did not submit a risk-informed license amendment request. Therefore, the NRC staff reviewed the licensees request using guidance contained in SRP Chapter 19, Appendix D, Use of Risk Information in Review of Non-Risk-Informed License Amendment Requests.
3.0 TECHNICAL EVALUATION
The Ginna plant is equipped with two ECCS accumulators, each of which is connected to the cold leg of one of Ginnas two reactor coolant system (RCS) loops. Each accumulator has a capacity of 1750 ft3, which contains nitrogen gas, and between 1111 ft3 and 1139 ft3 of water [2].
The accumulators are maintained at a temperature of about 300 EF and a pressure of about 745 psig. Each accumulator connection to the RCS contains a motor-operated isolation valve and two check valves in series. The motor-operated valve is normally open, with power removed from the valve motor to prevent inadvertent closure prior to or during an accident.
The primary safety function of the accumulators is to inject large quantities of borated water into the reactor vessel during the blowdown phase of a large loss-of-coolant accident (LOCA) and to provide inventory to help accomplish the refill phase that follows the blowdown phase. Boron concentration is controlled in the accumulators to prevent either excessive or insufficient boron concentrations. Post-LOCA emergency procedures are based on the worst case minimum boron precipitation time. Maintaining the accumulator boron concentration within the upper limit ensures that the borated water sources used for injection during a LOCA would not result in boron precipitation earlier than predicted by the design-basis calculation. The minimum boron concentration requirement is based on beginning-of-life reactivity values selected to ensure that the reactor will remain subcritical during the reflood stage of a large-break LOCA. During a large-break LOCA, control rods are not assumed to insert into the core. Instead, the reactor is shutdown by void formation during blowdown. The concentration of boron in the accumulators must be high enough to prevent a return to criticality during the reflood stage of the large-break LOCA.
The licensee proposes to change the method for complying with the surveillance required by TS SR 3.5.1.4. SR 3.5.1.4 requires that the boron concentration in the accumulators be verified every 31 days on a staggered test basis (i.e., by sampling one of the two accumulators every 31 days). The proposed change would require that leakage into the accumulators be monitored every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and that a sample be taken from each accumulator every 6 months. The samples would be used to verify that the boron concentration, inferred from inleakage observation, remains within the TS limit. Water level in the accumulators would continue to be measured every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The proposed change would reduce radiation exposure by reducing the number of required containment entries during power operation for sampling. In addition to the proposed changes to the surveillance requirement, the licensee has also proposed changes to the TS Bases that would require an additional sampling and verification of boron concentration if inleakage monitoring indicates that the ECCS accumulator boron concentration falls to within 100 ppm of the TS limit.
Deterministic Evaluation To justify the proposed change, the licensee evaluated the measured inleakage actually occurring in both ECCS accumulators during an 8-month period between April 2005 and December 2005. During this period, the inleakage to the accumulators was very steady and there was no significant change in boric acid concentration due to feed and bleed or any other procedurally initiated operations. Also, no trends were evident that would suggest significant inleakage causing boron depletion. The licensees study indicated that the largest change in accumulator boron concentration was approximately 2 ppm over the period, which is less than 0.1 percent of the total boron concentration in an accumulator.
Inleakage past two check valves is a very slow process, and is the only means by which boron dilution could occur in the accumulators. The licensee noted that the water volume between the low accumulator level alarm setpoint (60% of span) and the high level alarm setpoint (75% of span) is 94 gallons. Therefore, if an inleakage flow is assumed to begin when the water level is just above the low level alarm setpoint, and continues, unnoticed, until the high level alarm setpoint is reached, then the accumulator water inventory will have increased by a maximum of 94 gallons. Since water level and inleakage is monitored at 12-hour intervals, the 94 gallons can enter the accumulator during a period of no more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, or at a rate of no more than 7.83 gpm. To attain a net flow rate of 7.83 gpm, the leakage past either of the two serial check valves cannot be less than 7.83 gpm. Higher inleakage flow rates could not add more than 94 gallons since the operator would be expected to correct the problem after the high level alarm is sounded. Lower inleakage flow rates will be detected at the next regular surveillance, per SR 3.5.1.2, before 94 gallons could be added. Typically, the inleakage that could be expected to occur during normal operation would not add more than about 0.06 gallons over a 12-hour period.
The licensee included a simple calculation to show that addition of 94 gallons of clean water to the accumulator, at the currently maintained boron concentration, would dilute the boron concentration by about 31 ppm. The NRC staff agrees with this calculation and its underlying assumptions.
As shown in the table below, the proposed TS SR is comparable to the Standard TS (STS) for Westinghouse plants [7]. The Ginna sampling interval is six times longer (1 month vs. 6 months) than the STS, but includes inleakage observations at 12-hour intervals. Other plants require boron concentration measurements after a specified change in accumulator volume, typically 1%
(or about 130 gallons). For Ginna, it is apparent that boron concentration would not be diluted by more than 31 ppm in any 12-hour period. Therefore, it would take more than three 12-hour surveillance intervals to dilute by more than 100 ppm.
Plant Surveillance of Boron Concentration in Accumulators Ginna (present) 2 mos (each accumulator)
Ginna (proposed) [1]
6 mos and monitoring inleakage every 12 hrs Catawba [3]
1 mo and # 6 hrs after volume increases by 75 gal Cook [4]
1 mo and # 6 hrs after volume increases by 1%
Millstone Unit 2 [5]
6 mos and # 6 hrs after volume increases by 1%
Calvert Cliffs [6]
12 hrs by inleakage monitoring and 6 mos and within 1 hr after volume increases by 1%
Westinghouse STS [7]
1 mo and # 6 hrs after volume increases by { } gal or { }%
The results of the licensees analysis indicate that the normal change of boric acid concentration due to inleakage is very low, but even with a high inleakage due to operational error causing fluid level in the accumulator to raise from low to high alarm levels, the change of measured boric acid dilution remains within the acceptable limit.
The NRC staff finds Ginnas proposed SR acceptable, because no more than 94 gallons can enter an accumulator before either an inleakage surveillance is required, or a high level alarm is actuated. In comparable plants, boron concentration surveillance is required when the accumulator volume changes by 1% (typically about 130 gallons).
Based upon the review of the licensees proposed change to the Ginna TS SR and supporting justification, the NRC staff concludes the proposal will provide protection against boron dilution in the accumulators that is equal to or better than the current SR.
Evaluation Based on Risk Insights The licensee considered two potential risk impacts resulting from extending the surveillance test sampling interval for the ECCS accumulator boron concentration:
1.
A higher likelihood of a boron concentration less than 2100 ppm, given the increase in sampling interval from 2 months to 6 months.
- 2.
Incremental conditional core damage probability (ICCDP) and core damage frequency (CDF) due to an inadequate accumulator boron concentration following the blowdown phase of a large-break LOCA.
The licensee did not perform a full scope risk assessment of the potential impact on initiating events nor a condition analysis given a failure of an accumulator or variation of accumulator boron concentration from an acceptable level. Instead, the licensee focused on the major risk contributors: the likelihood of LOCA and probability of potential decrease in the boron concentration due to the change in accumulator sampling interval. In the event the boron concentration of one accumulator decreases below 2100 ppm, the reactor may not be able to be shutdown following a LOCA.
The licensee estimated the probability of core damage for a sequence involving a large-break LOCA with a failure of the control rods to insert occurring in parallel with the boron concentration in an accumulator less than 2100 ppm, and a sequence involving large-break LOCA with the boron concentration in an accumulator less than 2100 ppm and no control rod failure. The licensee estimated the frequency of a large-break LOCA (0.1 ft2) by logarithmic interpolation using data from Table 7.1 of NUREG-1829, Estimated Loss-of-Coolant Accident (LOCA)
Frequencies through the Elicitation Process. The LOCA frequency was estimated as 2E-5/year.
The licensees review of data from previous surveillance tests indicated that while there had never been an observation of an accumulator with a boron concentration less than 2100 ppm, reductions and increases in accumulator boron concentration during operation had been observed. An analysis of this data indicated that the likelihood of a reduction below 2100 ppm is approximately 2E-2 given the increased sampling interval of every 6 months. The control rod failure probability was derived from Table 3 of NUREG/CR-5500, Reliability Study, resulting in a base failure likelihood that all control rods fail to insert due to mechanical binding of less than 1E-6. Failure likelihood was also considered due to other systematic causes as well as the potential impact of flow dynamics during large-break LOCAs; the failure likelihood of insertion for each break range was logarithmically interpolated. The licensee estimated the total likelihood of a LOCA with a rod failure, systematic and flow-induced, to be less than 1E-7.
Summarizing the licensees assessment of the risk metrics by increasing the surveillance interval to 6 months:
The estimated value for the probability that one accumulator's boron concentration is less than 2100 ppm at the time of a LOCA is 2E-2.
The assumed frequency of a large-break LOCA is 2E-5/year.
The estimated CDF as a result of a large-break LOCA coupled with accumulator boron concentration less than 2100 ppm is 4E-7/year (i.e., (2E-2)x(2E-5/year)).
The estimated frequency of having core damage due to a large-break LOCA coupled with accumulator boron concentration less than 2100 ppm and failure of control rods to insert is less than 1E-7/year.
The licensee did not estimate a value for the large early release frequency (LERF) impact of the proposed SR change. The licensee stated that a very small fraction of large-break LOCAs result in LERF events, and concluded that this coupled with the low change in CDF ensures that the increase in LERF is also extremely small.
The NRC staff finds that the input values for these estimates of risk metrics were determined in an acceptable fashion and are within the expected ranges.
Since the increase of surveillance interval is 4 months (from 2 months to 6 months), the risk metric, ICCDP, will be further decreased by a factor of:
(4 months)/(12months/yr)= 0.333yr and the ICCDP for the large-break LOCA coupled with an accumulator boron concentration less than 2100 ppm is:
(4E-7/yr)x(0.333yr) = 1.3E-8 This increase of 1.3E-08 is consistent with the ICCDP estimated by employing the NRCs simplified plant analysis risk (SPAR) model for selected large-break LOCA sequences. The incremental conditional large early release probability (ICLERP) can be estimated using a conversion factor for the ICCDP-to-ICLERP employing NRC Manual Chapter 0609, Appendix H, "Containment Integrity Significance Determination Process. The factor in Appendix H is less than 0.1 for the Ginna containment; therefore, the ICLERP will be on the order of 1E-9.
The estimates of the increase in CDF and LERF are well within the acceptance guidelines prescribed in Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis."
Furthermore, the ICCDP of 1.3E-8 and ICLERP of less than 1E-9 are also well within the acceptance guidelines of 5E-7 and 5E-8, respectively, in RG 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications." Therefore, the NRC staff concludes that the changes in CDF and LERF, as well as the increases in CCDP and CLERP due to the proposed change in surveillance interval, are very small and that the proposed change does not introduce a level of risk that could challenge the presumption of adequate protection when all of the Commissions regulations are being met.
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 a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and 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 (71 FR 23960). 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 NRC staff 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 Commissions 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.
7.0 REFERENCES
1.
Letter from M. G. Korsnick, Site Vice President, R.E. Ginna Nuclear Power Plant, LLC, to USNRC, R.E. Ginna Nuclear Power Plant, Docket No. 50-244, License Amendment Request: Emergency Core Cooling System (ECCS) Accumulator Boron Concentration Verification Frequency, March 28, 2006 (ADAMS Accession No. ML0609403292) 2.
Docket No. 50-244, Updated Final Safety Analysis Report, R.E. Ginna Nuclear Power Plant, Rev. 18, April 2004 3.
Letter from Kahtan Jabbour, USNRC, to H. B. Tucker, Vice President, Nuclear Production Department, Duke Power Company, Issuance of Amendment No. 31 to Facility Operating License NPF-35 and Amendment No. 22 to Facility Operating License NPF-52-Catawba Nuclear Station, Units 1 and 2, November 10, 1987 (ADAMS Accession No. ML0130402530) 4.
Letter from J. B. Hickman, USNRC, to E. E. Fitzpatrick, Vice President, Indiana Michigan Power Company, Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2 - Issuance of Amendment re: Accumulators, November 8, 1994 (ADAMS Accession No. ML0210505580) 5.
Letter from S. Dembek, USNRC, to M. L. Bowling, Jr., Recovery Officer - Technical Services, Northeast Nuclear Energy Company, Issuance of Amendment - Millstone Nuclear Power Station, Unit No. 2, December 17, 1998 (ADAMS Accession No. ML0129105220) 6.
Letter from D. Skay, USNRC, to P. E. Katz, Vice President, Calvert Cliffs Nuclear Power Plant, Inc., Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2 - Amendment re:
Safety Injection Tank Surveillance Frequency, December 19, 2002 (ADAMS Accession No. ML0235306420) 7.
NUREG-1431, Vol. 1, Rev. 3.0, Standard Technical Specifications, Westinghouse Plants, June, 2004 Principal Contributors: S. Miranda J. Chung K. Parczewski Date: March 28, 2007