ML081690074

License Amendment, Control Rod Notch Surveillance Test Frequency

ML081690074
Person / Time
Site:	FitzPatrick
Issue date:	08/07/2008
From:	Bhalchandra Vaidya NRC/NRR/ADRO/DORL/LPLI-1
To:	Entergy Nuclear Operations
muniz A, ADRO/DORL/415-4044
References
TAC MD8048
Download: ML081690074 (14)

Text

August 7, 2008 Vice President, Operations Entergy Nuclear Operations, Inc.

James A. FitzPatrick Nuclear Power Plant P.O. Box 110 Lycoming, NY 13093

SUBJECT:

JAMES A. FITZPATRICK NUCLEAR POWER PLANT - ISSUANCE OF AMENDMENT RE: CONTROL ROD NOTCH SURVEILLANCE TEST FREQUENCY (TAC NO. MD8048)

Dear Sir or Madam:

The Commission has issued the enclosed Amendment No. 291 to Facility Operating License No. DPR-59 for the James A. FitzPatrick Nuclear Power Plant. The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated February 7, 2008.

The amendment revises the Technical Specifications (TS) Surveillance Requirement (SR) frequency in TS 3.1.3, Control Rod OPERABILITY, by decreasing the frequency of SR 3.1.3.2.

The amendment also revises one example in Section 1.4 Frequency to clarify the applicability of the 1.25 surveillance test interval extension (NUREG-1433). These changes were made consistent with Revision 1 to TS Task Force (TSTF) Standard Technical Specification Change Document TSTF-475, Control Rod Notch Testing Frequency and SRM Insert Control Rod Action. A notice of availability for this TS improvement using the consolidated line item improvement process (CLIIP) was published in the Federal Register on November 13, 2007 (72 FR 63935).

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/

Bhalchandra Vaidya, Project Manager Plant Licensing Branch I-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-333

Enclosures:

1. Amendment No. 291 to DPR-59

2. Safety Evaluation cc w/encls: See next page

August 7, 2008 Vice President, Operations Entergy Nuclear Operations, Inc.

James A. FitzPatrick Nuclear Power Plant P.O. Box 110 Lycoming, NY 13093

SUBJECT:

JAMES A. FITZPATRICK NUCLEAR POWER PLANT - ISSUANCE OF AMENDMENT RE: CONTROL ROD NOTCH SURVEILLANCE TEST FREQUENCY (TAC NO. MD8048)

Dear Sir or Madam:

The Commission has issued the enclosed Amendment No. 291 to Facility Operating License No. DPR-59 for the James A. FitzPatrick Nuclear Power Plant. The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated February 7, 2008.

The amendment revises the Technical Specifications (TS) Surveillance Requirement (SR) frequency in TS 3.1.3, Control Rod OPERABILITY, by decreasing the frequency of SR 3.1.3.2.

The amendment also revises one example in Section 1.4 Frequency to clarify the applicability of the 1.25 surveillance test interval extension (NUREG-1433). These changes were made consistent with Revision 1 to TS Task Force (TSTF) Standard Technical Specification Change Document TSTF-475, Control Rod Notch Testing Frequency and SRM Insert Control Rod Action. A notice of availability for this TS improvement using the consolidated line item improvement process (CLIIP) was published in the Federal Register on November 13, 2007 (72 FR 63935).

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/

Bhalchandra Vaidya, Project Manager Plant Licensing Branch I-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-333

Enclosures:

1. Amendment No. 291 to DPR-59

2. Safety Evaluation cc w/encls: See next page Package No.: ML081690087 Amendment No.: ML081690074 Tech Spec No.: ML081690094

(*) - no substantial changes to SE Input Memorandum OFFICIAL RECORD COPY OFFICE LPL1-1\\PM LPL1-1\\LA ITSB\\BC(*)

OGC LPL1-1\\BC NAME BVaidya SLittle RElliot MSimon MKowal DATE 7/24/08 7/24/08 07/23/08 8/05/08 8/07/08

DATED: August 7, 2008 AMENDMENT NO. 291 TO FACILITY OPERATING LICENSE NO. DPR-59 FITZPATRICK DISTRIBUTION:

PUBLIC LPL1-1 R/F RidsAcrsAcnw_MailCTR Resource RidsNrrDirsItsb Resource RidsNrrDorlDpr Resource RidsNrrDorlLpl1-1 Resource RidsNrrItsb Resource RidsNrrPMBVaidya Resource RidsNrrLASLittle (paper copy)

RidsOgcRp Resource RidsRgn1MailCenter Resource RGrover, NRR/ITSB GHill, OIS cc: Plant Mailing List

FitzPatrick Nuclear Power Plant cc:

Senior Vice President Entergy Nuclear Operations, Inc.

P.O. Box 31995 Jackson, MS 39286-1995 Vice President Oversight Entergy Nuclear Operations, Inc.

P.O. Box 31995 Jackson, MS 39286-1995 Senior Manager, Nuclear Safety &

Licensing Entergy Nuclear Operations, Inc.

P.O. Box 31995 Jackson, MS 39286-1995 Senior Vice President and COO Entergy Nuclear Operations, Inc.

440 Hamilton Avenue White Plains, NY 10601 Assistant General Counsel Entergy Nuclear Operations, Inc.

440 Hamilton Avenue White Plains, NY 10601 Manager, Licensing Entergy Nuclear Operations, Inc.

James A. FitzPatrick Nuclear Power Plant P.O. Box 110 Lycoming, NY 13093 Mr. Charles Donaldson, Esquire Assistant Attorney General New York Department of Law 120 Broadway New York, NY 10271 Mr. Paul Tonko President and CEO New York State Energy, Research, and Development Authority 17 Columbia Circle Albany, NY 12203-6399 Mr. John P. Spath New York State Energy, Research, and Development Authority 17 Columbia Circle Albany, NY 12203-6399 Mr. Paul Eddy New York State Dept. of Public Service 3 Empire State Plaza Albany, NY 12223-1350 Oswego County Administrator Mr. Steven Lyman 46 East Bridge Street Oswego, NY 13126 Supervisor Town of Scriba Route 8, Box 382 Oswego, NY 13126 Resident Inspector's Office James A. FitzPatrick Nuclear Power Plant U. S. Nuclear Regulatory Commission P.O. Box 136 Lycoming, NY 13093 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406

ENTERGY NUCLEAR FITZPATRICK, LLC AND ENTERGY NUCLEAR OPERATIONS, INC.

DOCKET NO. 50-333 JAMES A. FITZPATRICK NUCLEAR POWER PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 291 License No. DPR-59

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Entergy Nuclear Operations, Inc. (the licensee) dated February 7, 2008, 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-59 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 291, 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 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Mark G. Kowal, 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: August 7, 2008

ATTACHMENT TO LICENSE AMENDMENT NO. 291 FACILITY OPERATING LICENSE NO. DPR-59 DOCKET NO. 50-333 Replace the following page of the License with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Remove Page Insert Page 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 Pages Insert Pages 1.4-5 1.4-5 3.1.3-2 3.1.3-2 3.1.3-4 3.1.3-4 3.1.3-5 3.1.4-3 3.1.4-3

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 291 TO FACILITY OPERATING LICENSE NO. DPR-59 ENTERGY NUCLEAR OPERATIONS, INC.

JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333

1.0 INTRODUCTION

By letter dated February 7, 2008, (Agencywide Documents and Management System (ADAMS)

Accession No. ML080510730), Entergy Nuclear Operations, Inc., (the licensee) submitted a license amendment request (LAR) regarding the James A. FitzPatrick Nuclear Power Plants (JAFNPP) Operating License. The proposed amendment adopts Technical Specifications Task Force (TSTF)-475, Revision 1 Control Rod Notch Testing Frequency and SRM Insert Control Rod Action. The proposed changes would: (1) revise the Technical Specifications (TS) control rod notch surveillance frequency in TS 3.1.3, Control Rod OPERABILITY, (NUREG-1433 and NUREG-1434) and (2) revise one Example in Section 1.4 Frequency to clarify the applicability of the 1.25 surveillance test interval extension (NUREG-1430 through NUREG-1434). The licensees letter states, Entergy is not proposing any variations or deviations from the applicable TS changes described in the modified TSTF 475, Revision 1 and the NRC staff's model safety evaluation dated November 13, 2007. During the JAFNPP conversion to Improved Standard Technical Specifications, TS Section 3.3.1.2, required Action E.2, Source Range Monitoring Instrumentation and associated bases were clarified to state fully insert all insertable control rods for the limiting condition for operation (LCO), therefore we are not requesting a change to this section of the Technical Specifications.

These changes are based on the Nuclear Regulatory Commission (NRC) approved TSTF change traveler TSTF-475, Revision 1, that revised the reference Standard Technical Specifications (STS) by: (1) revising the frequency of Surveillance Requirement (SR) 3.1.3.2, notch testing of each fully withdrawn control rod, from 7 days after the control rod is withdrawn and THERMAL POWER is greater than the Low Power Setpoint (LPSP) of the Rod Worth Minimizer (RWM) to 31 days after the control rod is withdrawn and THERMAL POWER is greater than the LPSP of the RWM (NUREG-1433 and NUREG-1434) and (2) revising Example 1.4-3 in Section 1.4 Frequency to clarify that the 1.25 surveillance test interval extension in SR 3.0.2 is applicable to time periods discussed in NOTES in the SURVEILLANCE column in addition to the time periods in the FREQUENCY column (NUREG-1430 through NUREG-1434).

The purpose of the surveillances is to confirm control rod insertion capability which is demonstrated by inserting each partially or fully withdrawn control rod at least one notch and observing that the control rod moves. Control rods and the control rod drive (CRD) mechanism (CRDM), by which the control rods are moved, are components of the CRD system (CRDS),

which is the primary reactivity control system for the reactor. By design, the CRDM is highly reliable with a tapered design of the index tube which is conducive to control rod insertion.

A stuck control rod is an extremely rare event and industry review of plant operating experience did not identify any incidents of stuck control rods while performing a rod notch surveillance test.

The purpose of these revisions is to reduce the number of control rod manipulations and, thereby, reduce the opportunity for reactivity control events.

The purpose of the change to Example 1.4-3 in Section 1.4 Frequency is to clarify the applicability of the 25% allowance of SR 3.0.2 to time periods discussed in NOTES in the SURVEILLANCE column as well as to time periods in the FREQUENCY column.

2.0 REGULATORY EVALUATION

Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix A, General Design Criterion (GDC) 29, Protection against anticipated occurrence, requires that the protection and reactivity control systems be designed to assure an extremely high probability of accomplishing their safety functions in the event of anticipated operational occurrences. The design relies on the CRDS to function in conjunction with the protection systems under anticipated operational occurrences, including loss of power to all recirculation pumps, tripping of the turbine generator, isolation of the main condenser, and loss of all offsite power. The CRDS provides an adequate means of inserting sufficient negative reactivity to shut down the reactor and prevent exceeding acceptable fuel design limits during anticipated operational occurrences. Meeting the requirements of GDC 29 for the CRDS prevents occurrence of mechanisms that could result in fuel cladding damage such as severe overheating, excessive cladding strain, or exceeding the thermal margin limits during anticipated operational occurrences. Preventing excessive cladding damage in the event of anticipated transients ensures maintenance of the integrity of the cladding as a fission product barrier.

3.0 TECHNICAL EVALUATION

The NRC staff previously reviewed the following information provided by the TSTF to support the staffs review and approval of TSTF-475, Revision 1. Specifically, the following documents were reviewed during the NRC staffs evaluation:

TSTF letter TSTF-04-07 (Reference 1) - Provided a description of the proposed changes in TSTF-475 that changes the weekly rod notch frequency to monthly and clarify the applicability of the 25% allowance in Example 1.4-3.

TSTF letter TSTF-06-13 (Reference 4) - Provided responses to NRC staff request for additional information (RAI) on (1) industry experience with identifying stuck rods, (2) tests that would identify stuck rods, (3) continue compliance with General Electric (GE)

Services Information Letter (SIL) 139, (4) industry experience on collet failures, and (4) applying the 25% grace period to the 31 day control rod notch SR test frequency.

Boiling Water Reactor Owners Group (BWROG) letter BWROG-06036 (Reference 5) -

Provided the GE Nuclear Energy Report, CRD Notching Surveillance Testing for Limerick Generating Station, in which CRD notching frequency and CRD performance were evaluated.

TSTF letter TSTF-07-19 (Reference 6) - Provided response to NRC staff RAI on CRD performance in Control Cell Core (CCC) designed plants, including TSTF-475, Revision 1.

The CRDS at JAFNPP is the primary reactivity control system for the reactor. The CRDS, in conjunction with the Reactor Protection System, provides the means for the reliable control of reactivity changes to ensure under all conditions of normal operation, including anticipated operational occurrences that specified acceptable fuel design limits are not exceeded. Control rods are components of the CRDS that have the capability to hold the reactor core subcritical under all conditions and to limit the potential amount and rate of reactivity increase caused by a malfunction in the CRDS.

The CRDS consists of a CRDM, by which the control rods are moved, and a hydraulic control unit (HCU) for each control rod. The CRDM is a mechanical hydraulic latching cylinder that positions the control blades. The CRDM is a highly reliable mechanism for inserting a control rod to the full-in position. The collet piston mechanism design feature ensures that the control rod will not be inadvertently withdrawn. This is accomplished by engaging the collet fingers, mounted on the collet piston, in notches located on the index tube. Due to the tapered design of the index tube notches, the collet piston mechanism will not impede rod insertion under normal insertion or scram conditions.

The collet retainer tube (CRT) is a short tube welded to the upper end of the CRD which houses the collet mechanism which consist of the locking collet, collet piston, collet return spring and an unlocking cam. The collet mechanism provides the locking/unlocking mechanism that allows the insert/withdraw movement of the control rod. The CRT has three primary functions: (a) to carry the hydraulic unlocking pressure to the collet piston, (b) to provide an outer cylinder, with a suitable wear surface for the metal collet piston rings, and (c) to provide mechanical support for the guide cap, a component which incorporates the cam surface for holding the collet fingers open and also provides the upper rod guide or bushing.

The NRC staff approved TSTF-475 which revised the TS SR 3.1.3.2, Control Rod OPERABILITY in the STS (NUREG-1433 and NUREG 1434) from 7 days to monthly based on the following: (1) slow crack growth rate of the CRT; (2) the improved CRT design; (3) a more reliable method (scram time testing) to monitor CRD scram system functionality; (4) GE chemistry recommendations; and (5) no known CRD failures have been detected during the notch testing exercise. The NRC staff concluded that the changes would reduce the number of control rod manipulations thereby reducing the opportunity for potential reactivity events while having a very minimal impact on the extremely high reliability of the CRD system. The following paragraphs describe the bases for the staffs approval of TSTF-475:

According to the Boiling Water Reactor Owners Group (BWROG), at the time of the first CRT crack discovery in 1975, each partially or fully withdrawn operable control rod was required to be exercised one notch at least once each week. It was recognized that notch testing provided a method to demonstrate the integrity of the CRT. Control rod insertion capability was demonstrated by inserting each partially or fully withdrawn control rod at least one notch and observing that the control rod moves. The control rod may then be returned to its original position. This ensures the control rod is not stuck and is free to insert on a scram signal.

It was determined that during scrams, the CRT temperature distribution changes substantially at reactor operating conditions. Relatively cold water moves upward through the inside of the CRT and exits via the flow holes into the annulus on the outside. At the same time, hot water from the reactor vessel flows downward on the outside surface of the CRT. There is very little mixing of the cold water flowing from the three flow holes into the annulus and the hot water flowing downward. Thus, there are substantial through wall and circumferential temperature gradients during scrams which contribute to the observed CRT cracking.

Subsequently, many BWRs have reduced the frequency of notch testing for partially withdrawn control rods from weekly to monthly. The notch test frequency for fully withdrawn control rods are still performed weekly. The change for partially withdrawn control rods was made because of the potential power reduction required to allow control rod movement for partially withdrawn control rods, the desire to coordinate scheduling with other plant activities, and the fact that a large sample of control rods are still notch tested on a weekly basis. The operating experience related to the changes in CRD performance also provided additional justification to reduce the notch test frequency for the partially withdrawn control rods.

In response to NRCs RAIs and to support their position to reduce the CRD notch testing frequency, the BWROG provided plant data and a GE Nuclear Energy report entitled, CRD Notching Surveillance Testing for Limerick Generating Station (CRDNST). The GE report provided a description of the cracks noted on the original design CRT surfaces. These cracks, which were later determined to be intergranular, were generally circumferential, and appeared with greatest frequency below and between the cooling water ports, in the area of the change in wall thickness. Subsequently, cracks associated with residual stresses were also observed in the vicinity of the attachment weld. Continued circumferential cracking could lead to 360 degree severance of the CRT that would render the CRD inoperable which would prevent insertion, withdrawal or scram. Such failure would be detectable in any fully or partially withdrawn control rod during the surveillance notch testing required by the TSs. To a lesser degree, cracks have also been noted at the welded joint of the interim design CRT but no cracks have been observed in the final improved CRT design. Neither the BWROG nor the NRC staff was able to find evidence of a collet housing failure since 1975. To date, operating experience data shows no reports of a severed CRT at any BWR. No collet housing failures have been noted since 1975. On a numerical basis for instance, based on BWROG assumption that there are 137 control rods for a typical BWR/4 and 193 control rods for a typical BWR/6, the yearly performance would be 6590 rod notch tests for a BWR/4 plant and 9284 for a BWR/6 plant. For example, if all BWRs operating in the U.S. are taken into consideration, the yearly performances of rod notch data would translate into approximately 240,000 rod notch tests without detecting a failure.

In addition, the IGSCC crack growth rates were evaluated, at Limerick Generating Station, using GEs PLEDGE model with the assumption that the water chemistry condition is based on GE recommendations. The model which is based on fundamental principles of stress-corrosion cracking can evaluate crack growth rates as a function of water oxygen level, conductivity, material sensitization and applied loads. It was determined that the additional time of 24 days represented an additional 10 mils of growth in total crack length. The small difference in growth rate would have little effect on the behavior between one notch test and the next subsequent test. Therefore, from the materials perspective based on low crack growth rates, a decrease in the notch test frequency would not affect the reliability of detecting a CRDM failure due to crack growth.

Also, the BWR scram system has extremely high reliability. In addition to notch testing, scram time testing can identify failure of individual CRD operation resulting from IGSCC-initiated cracks and mechanical binding. Unlike the CRD notch tests, these single rod scram tests cover the other mechanical components such as scram pilot solenoid operated valves, the scram inlet and outlet air operated valves, and the scram accumulator, as well as operation of the control rods. Thus, the primary assurance of scram system reliability is provided by the scram time testing since it monitors the system scram operation and the complete travel of the control rod.

Also, the Hydraulic Control Units (HCUs), CRD drives, and control rods are tested during refueling outages, approximately every 18-24 months. Based on the data collected during the preceding cycle of operation, selected control rod drives, are inspected and, as required, their internal components are replaced. Therefore, increasing the CRD notch testing frequency to monthly would have very minimal impact on the reliability of the scram system.

Entergy stated in their application that they have reviewed the basis for the NRC staffs acceptance of TSTF-475, Revision 1, and concluded that the basis is applicable to JAFNPP, and supports their adoption of the TSTF-475 changes into the JAFNPP TS. The staff also reviewed the TSTF-475, Revision 1 basis, and similarly concluded that the basis for the TSTF is applicable to JAFNPP, and therefore, the TSTF is appropriate for adoption by the licensee. In addition, the NRC staff reviewed the licensees proposed changes against the corresponding changes made to the STS by TSTF-475, Revision 1, which the staff has found to satisfy applicable regulatory requirements, as described above. The proposed changes would: (1) revise the TS control rod notch surveillance frequency in TS 3.1.3, Control Rod OPERABILITY, and (2) revise one Example in Section 1.4 Frequency to clarify the applicability of the 1.25 surveillance test interval extension. The staff found that the proposed changes are consistent with the changes approved by the staff in TSTF-475, Revision 1. The NRC staff, therefore, finds these changes acceptable.

3.1 Summary The NRC staff has reviewed the licensees proposal to amend existing JAFNPP TS sections SR 3.1.3.2, Control Rod OPERABILITY, and Example 1.4-3, Frequency applicable to SR 3.0.2.

The NRC staff has concluded that the TS revisions will have a minimal effect on the high reliability of the CRDS while reducing the opportunity for potential reactivity events; thus, meeting the requirements of 10 CFR Part 50, Appendix A, GDC 29, and will clarify the applicability of the 1.25 provision in SR 3.0.2. Therefore, the staff concludes that the amendment request is 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 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 (73 FR 18008). 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.

7.0 REFERENCES

1. Letter No. TSTF-04-07 from the TSTF to the NRC, TSTF-475, Revision 0, Control Rod Notch Testing Frequency and SRM Insert Control Rod Action, August 30, 2004, ADAMS accession number ML042520035.

2. NUREG-1433, Standard Technical Specifications General Electric Plants, BWR/4, Revision 3, August 31, 2003.

3. Letter No. TSTF-07-19, Response from the TSTF to the NRC, Request for Additional Information (RAI) Regarding TSTF-475, Revision 0, Control Rod Notch Testing Frequency and SRM Insert Control Rod Action, dated February 28, 2007, (TSTF-475, Revision 1 is an enclosure), ADAMS accession number ML071420428.

4. Letter No. TSTF-06-13 from the TSTF to the NRC, Response to NRC Request for Additional Information Regarding TSTF-475, Revision 0, dated July 3, 2006, ADAMS accession number ML061840342.

5. Letter No. BWROG-06036 from the BWROG to the NRC, Response to NRC Request for Additional Information Regarding TSTF-475, Revision 0, dated November 16, 2006, with enclosure of the GE Nuclear Energy Report, CRD Notching Surveillance Testing for Limerick Generating Station, dated November 2006, ADAMS accession number ML063250258.

6. Letter No. TSTF-07-19 from the TSTF to the NRC, Response to NRC Request for Additional Information Regarding TSTF-475, Revision 0, dated May 22, 2007, ADAMS accession number ML071420428.

Principal Contributor: Ravinder P. Grover Date: August 7, 2008