Notice of Availability of Model Application Concerning Technical Specification Improvement To Revise Control Rod Notch Surveillance Frequency, Clarify SRM Insert Control Rod Action, and Clarify Frequency Example, 63935-63942 [E7-22159]

Download as PDF 63935 Federal Register / Vol. 72, No. 218 / Tuesday, November 13, 2007 / Notices need to prepare an environmental impact statement. amendment and supporting documentation, are available electronically at the NRC’s Electronic Reading Room at http://www.nrc.gov/ reading-rm/adams.html. From this site, you can access the NRC’s Agencywide IV. Further Information Documents related to this action, including the application for Document Access and Management System (ADAMS), which provides text and image files of NRC’s public documents. The ADAMS accession numbers for the documents related to this notice are as follows: Document ADAMS accession No. NUREG–1748, ‘‘Environmental Review Guidance for Licensing Actions Associated With NMSS Programs—Final Report,’’ Nuclear Regulatory Commission, Washington, DC. NUREG–1620, Rev. 1, ‘‘Standard Review Plan for Review of a Reclamation Plan for Mill Tailings Sites Under Title II of the Uranium Mill Tailings Radiation Control Act of 1978,’’ Nuclear Regulatory Commission, Washington, DC. Rio Algom Mining LLC, 2004, ‘‘Closure Plan-Lined Evaporation Ponds’’ ......................................... Rio Algom, 2005; Reclamation Plan for Disposal of Pond Sediments and Ancillary Materials, Tailings Cell 2 Expansion. Rio Algom 2007; Reclamation Plan for Disposal of Pond Sediments and Ancillary Materials, Tailings Cell 2 Expansion, Revision 1. Environmental Assessment for the Tailings Cell 2 Expansion Reclamation Plan, Rio Algom Mining LLC’s Uranium Mill Facility, Ambrosia Lake, New Mexico, Final Report. ML031000403 April 10, 2003. ML040560561 February 19, 2004. ML050240058 ML051290050 November 1, 2004. April 30, 2005. ML071790245 ML071790250 ML072670278 May 31, 2007. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the NRC’s Public Document Room (PDR) Reference staff at 1–800–397–4209, 301– 415–4737, or by e-mail to pdr@nrc.gov. These documents may also be viewed electronically on the public computers located at the NRC’s PDR, O1 F21, One White Flint North, 11555 Rockville Pike, Rockville, MD 20852. The PDR reproduction contractor will copy documents for a fee. Dated at Rockville, Maryland, this 2nd day of November, 2007. For the Nuclear Regulatory Commission. Keith I. McConnell, Deputy Director, Decommissioning and Uranium Recovery Licensing Directorate, Division of Waste Management and Environmental Protection, Office of Federal and State Materials and Environmental Management Programs. [FR Doc. E7–22114 Filed 11–9–07; 8:45 am] BILLING CODE 7590–01–P NUCLEAR REGULATORY COMMISSION Notice of Availability of Model Application Concerning Technical Specification Improvement To Revise Control Rod Notch Surveillance Frequency, Clarify SRM Insert Control Rod Action, and Clarify Frequency Example Nuclear Regulatory Commission. ACTION: Notice of availability. rfrederick on PROD1PC67 with NOTICES AGENCY: 15:30 Nov 09, 2007 Jkt 214001 The NRC staff issued a Federal Register notice (72 FR 46103; August 16, 2007) which provided a model SE, model application, and model NSHC related to BWR plant control rod notch surveillance frequency, BWR SRM control rod insertion action, and clarification of a surveillance frequency DATES: SUMMARY: Notice is hereby given that the staff of the Nuclear Regulatory Commission (NRC) has prepared a model safety evaluation (SE) relating to the revision of Standard Technical VerDate Aug<31>2005 Specifications (STS), NUREG–1430 (B&W), NUREG–1431 (Westinghouse), NUREG–1432 (CE), NUREG–1433 (BWR/4) and NUREG–1434 (BWR/6). Specifically the SE addresses: (1) The revision of the technical specification (TS) surveillance requirement (SR) 3.1.3.2 frequency in STS 3.1.3, ‘‘Control Rod OPERABILITY,’’ (NUREG–1433 and NUREG–1434), (2) a clarification to the requirement to fully insert all insertable control rods for the limiting condition for operation (LCO) in STS 3.3.1.2, Required Action E.2, ‘‘Source Range Monitor Instrumentation’’ (NUREG– 1434 only), and (3) the revision of Example 1.4–3 in STS Section 1.4 ‘‘Frequency’’ to clarify the applicability of the 1.25 surveillance test interval extension (NUREG–1430 through NUREG–1434). The NRC staff has also prepared a model license amendment request and a model no significant hazards consideration (NSHC) determination relating to this matter. The purpose of these models is to permit the NRC to efficiently process amendments that propose to modify TS control rod SR testing frequency, clarify TS control insertion requirements, and clarify SR frequency discussions. Licensees of nuclear power reactors to which the models apply can request amendments, confirming the applicability of the SE and NSHC determination to their plant licensing basis. PO 00000 Frm 00063 Fmt 4703 Sfmt 4703 Date September, 2007. example for all plant types. Similarly, the NRC staff herein provides a revised model SE, model LAR, and model NSHC incorporating changes based upon the public comments received. The NRC staff can most efficiently consider applications based upon the model LAR, which references the model SE, if the LAR is submitted within one year of this Federal Register Notice. FOR FURTHER INFORMATION CONTACT: Timothy Kobetz, Mail Stop: O–12H2, Technical Specifications Branch, Division of Inspection & Regional Support, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555– 0001, telephone: 301–415–1932. SUPPLEMENTARY INFORMATION: Background Regulatory Issue Summary 2000–06, ‘‘Consolidated Line Item Improvement Process for Adopting Standard Technical Specification Changes for Power Reactors,’’ was issued on March 20, 2000. The consolidated line item improvement process (CLIIP) is intended to improve the efficiency of NRC licensing processes by processing proposed changes to the standard technical specifications (STS) in a manner that supports subsequent license amendment applications. The CLIIP includes an opportunity for the public to comment on proposed changes to the STS following a preliminary assessment by the NRC staff and finding that the change will likely be offered for adoption by licensees. The CLIIP directs the NRC staff to evaluate any comments received for a proposed change to the STS and to either reconsider the change or to proceed with announcing the availability of the change for proposed E:\FR\FM\13NON1.SGM 13NON1 63936 Federal Register / Vol. 72, No. 218 / Tuesday, November 13, 2007 / Notices adoption by licensees. Those licensees opting to apply for the subject change to technical specifications are responsible for reviewing the staff’s evaluation, referencing the applicable technical justifications, and providing any necessary plant-specific information. Each amendment application made in response to the notice of availability will be processed and noticed in accordance with applicable rules and NRC procedures. This notice involves the modification of BWR TS control rod SR testing frequency, clarification of BWR TS control insertion requirements, and clarification of SR frequency discussions for all pant types. This change was proposed for incorporation into the standard technical specifications by the Owners Groups participants in the Technical Specification Task Force (TSTF) and is designated TSTF–475 Revision 1. TSTF–475 Revision 1 can be viewed on the NRC’s Web page at http:// www.nrc.gov/reactors/operating/ licensing/techspecs.html. rfrederick on PROD1PC67 with NOTICES *** Reviewer’s Note *** TSTF–475 involves three changes to the Standard Technical Specifications NUREGs that, depending upon the adopting plant, may or may not be adopted by a plant. The first changes the surveillance frequency for control rod notch testing from 7 to 31 days, and applies to BWR/4 and BWR/6 plants (NUREG–1433 & NUREG–1434). The second adds the word ‘‘fully’’ to a Required Action statement to clarify that control rods should be fully inserted, and applies to only the BWR/6 plants (NUREG–1434). The third change clarifies the usage of the 1.25 surveillance frequency interval extension, and applies to all plants (NUREG–1430 through NUREG–1434). The model application and model safety evaluation will need to be tailored (where brackets indicate) for plant specific applications. Applicability This proposed TS change modifies TS control rod SR testing frequency and clarifies TS control insertion requirements for BWR plants, and clarifies SR frequency discussions for all NSSS plant types. The CLIIP does not prevent licensees from requesting an alternative approach or proposing the changes without the attached model SE and the NSHC. Variations from the approach recommended in this notice may, however, require additional review by the NRC staff and may increase the time and resources needed for the review. To efficiently process the incoming license amendment applications, the staff requests that each licensee applying for the changes proposed in TSTF–475, Revision 1, include TS Bases VerDate Aug<31>2005 15:30 Nov 09, 2007 Jkt 214001 for the proposed TS consistent with the TS Bases proposed in TSTF–475, Revision 1 (note: the change to STS Section 1.4 does not entail a Bases change). The staff is requesting that the TS Bases be included with the proposed license amendments in this case because the changes to the TS and the changes to the associated TS Bases form an integral change to a plant’s licensing basis. To ensure that the overall change, including the TS Bases, includes appropriate regulatory controls, the staff plans to condition the issuance of each license amendment on the licensee’s incorporation of the changes into the TS Bases document and that the licensee control changes to the TS Bases in accordance with the licensees TS Bases Control Program. The CLIIP does not prevent licensees from requesting an alternative approach or proposing the changes without the requested TS Bases. However, deviations from the approach recommended in this notice may require additional review by the NRC staff and may increase the time and resources needed for the review. Significant variations from the approach, or inclusion of additional changes to the license, will result in staff rejection of the submittal. Instead, licensees desiring significant variations and/or additional changes should submit a LAR that does not request to adopt TSTF–475, Revision 1, under CLIIP. Public Notices The staff issued a Federal Register Notice (72 FR 46103, August 16, 2007) that requested public comment on the NRC’s pending action to approve the modification of BWR TS control rod SR testing frequency, clarification of BWR TS control insertion requirements, and clarification of SR frequency discussions for all pant types, as proposed in TSTF–475, Revision 1. The TSTF–475, Revision 1, can be viewed on the NRC’s web page at http:// www.nrc.gov/reactors/operating/ licensing/techspecs.html. TSTF–475, Revision 1, may be examined, and/or copied for a fee, at the NRC’s Public Document Room, located at One White Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records are accessible electronically from the ADAMS Public Library component on the NRC Web site, (the Electronic Reading Room) at http://www.nrc.gov/reading-rm/ adams.html. In response to the notice soliciting comments from interested members of the public about the modification of BWR TS control rod SR testing frequency, clarification of BWR TS control insertion requirements, and PO 00000 Frm 00064 Fmt 4703 Sfmt 4703 clarification of SR frequency discussions for all pant types, the staff received one set of comments (from the TSTF Owners Groups, representing licensees). The specific comments are provided and discussed below: 1. Comment: TSTF–475 contains three changes: The revision to SR 3.1.3.2 which is applicable to NUREG–1433 and NUREG–1434 (the Improved Standard Technical Specifications, or ISTS, for BWR/4 and BWR/6 plants), the change to Specification 3.3.1.2, Required Action E.2 which is applicable to NUREG–1434 (the ISTS for BWR/6 plants), and the change to Example 1.4– 3 which is applicable to NUREG–1430 through –1434 (the ISTS for all plant types). The applicability of the third change to all plant types is clearly indicated on the Traveler cover page and in the justification (last paragraph of Section 2.0, ‘‘Proposed Change.’’) However, the Notice for Comment, model Safety Evaluation, model application, and No Significant Hazards Considerations Determination (NSHC) incorrectly state that TSTF–475 is only applicable to BWR plants. The Notice, the model application, model Safety Evaluation, and NSHC should be revised to state that the change to Example 1.4–3 is applicable to all plant types. The model Safety Evaluation, model application, and NSHC should be revised to bracket (e.g., indicate as optional) the BWR/4 and BWR/6 specific changes so that the documents are applicable to a BWR/6 plant adopting all three changes, a BWR/4 plant adopting the SR 3.1.3.2 and Example 1.4–3 changes, or a pressurized water reactor (PWR) plant adopting only the Example 1.4–3 change. Response: The staff agrees with the comment and the model application, model Safety Evaluation, and NSHC have been revised accordingly. 2. Comment: In Section 3.0, ‘‘Technical Evaluation,’’ of the Notice, reference is made three times to the ‘‘BWROG TSTF’’ or ‘‘BWROG TSTF– 475.’’ The Technical Specifications Task Force (TSTF) is sponsored by the Boiling Water Reactor Owners Group and the Pressurized Water Reactor Owners Group. The proper designation is either ‘‘TSTF’’ or ‘‘Owners Group TSTF.’’ Response: The staff agrees with the comment and Section 3.0 of the model Safety Evaluation has been revised by removing explicit reference to the BWROG in referring to TSTF–475. 3. Comment: In Section 3.0, ‘‘Technical Evaluation,’’ the model Safety Evaluation states, ‘‘Therefore, the NRC staff finds the change acceptable E:\FR\FM\13NON1.SGM 13NON1 rfrederick on PROD1PC67 with NOTICES Federal Register / Vol. 72, No. 218 / Tuesday, November 13, 2007 / Notices with the commitment to implement GE water quality for the CRD system recommendations.’’ In the model application, a regulatory commitment is included which states, ‘‘[LICENSEE] will establish the water quality controls as recommended by SIL No. 148, Water Quality Control for the Control Rod System,’’ September 15, 1975.’’ This commitment should be removed. The TSTF’s justification for TSTF– 475 made no mention of and did not rely on water quality controls. The TSTF’s July 3, 2006 response to the NRC’s March 21, 2003 Request for Additional Information (RAI) did not credit water chemistry controls. As stated in the justification and the Staff’s model Safety Evaluation, 30 years of operating experience at BWRs without a control rod drive failure detected by the weekly notch testing is sufficient to demonstrate the acceptability of the change. The reference is technically incorrect. Supplement 1 to SIL No. 148 was issued in June 2004 and updates the SIL to bring it into alignment with current Electric Power Research Institute (EPRI) BWR water chemistry requirements, which were in conflict with the 1975 version of SIL. The NRC’s Technical Evaluation in the draft Safety Evaluation did not reference SIL No. 148 (either the 1975 version or the current version). It is not appropriate for the NRC to require commitments to documents that were not relied on in the licensee’s application, were not reviewed by the NRC, and were not discussed in the NRC’s technical evaluation. Therefore, the reference to water chemistry controls in the model Safety Evaluation and the commitment in the model application should be removed. Response: The staff agrees with the comment and the requirements for a commitment to establish water quality controls as recommended by SIL No. 148, Water Quality Control for the Control Rod System, in the model Safety Evaluation and in the model application have been removed. 4. Comment: Model Application: Attachment 5, ‘‘Proposed Technical Specification Bases,’’ should be marked as optional. There are no Bases changes associated with the PWR-applicable changes to Section 1.4. Furthermore, the Bases changes associated with TSTF– 475 simply reflect the changes made to the specifications. It should be left to the licensee whether to submit Bases changes with the amendment request. The third paragraph omits Attachment 5, which is shown in the list of attachments below the signature. Attachment 3, ‘‘Proposed Technical VerDate Aug<31>2005 15:30 Nov 09, 2007 Jkt 214001 Specification Pages,’’ should also be marked as optional as not all licensee’s submit retyped Technical Specification pages as attachments to their amendment requests. Response: The staff does not agree with the comment. For those sections of the technical specifications that are changed in accordance with TSTF–475 and that have Bases, the Bases must be changed to reflect the change in accordance with TSTF–475. TS Section 1.4, that does not have Bases, does not need to have Bases changes submitted, and for those plants that are only adopting the TS Section 1.4 change, the Model Application Attachment 5, ‘‘Proposed Technical Specification Bases,’’ will be revised to indicate that the submittal of revised Bases pages is optional in that case. The staff does not see a need to revise Model Application Attachment 3. The staff expects to see the licensee’s Bases changes associated with the adoption of TSTF–475. 5. Comment: Model Application: The Model Application states, ‘‘I declare under penalty of perjury under the laws of the United States of America that I am authorized by [LICENSEE] to make this request and that the foregoing is true and correct.’’ This statement is not consistent with the recommended statement given in RIS 2001–18, ‘‘Requirements for Oath or Affirmation.’’ RIS 2001–18 recommends the statement, ‘‘I declare [or certify, verify, state] under penalty of perjury that the foregoing is true and correct.’’ Note that RIS 2001– 18 states that this statement must be used verbatim. We recommend that the Model Application be revised to be consistent with RIS 2001–18. Response: The staff agrees with the comment and the requirement in the model application for oath or affirmation has been reworded to be consistent with RIS 2001–18. 6. Comment: Attachment 4: The regulatory commitment states ‘‘[LICENSEE] will establish the Technical Specification Bases for [TS B 3.1.3, TS B 3.1.4, and TS B 3.3.1.2] as adopted with the applicable license amendment.’’ This statement is incorrect as the Bases changes included for information with the license amendment request are not ‘‘adopted’’ with the license amendment. Bases changes are made under licensee control under the Technical Specification Bases Control Program. We recommend revising the commitment to state ‘‘[LICENSEE] will implement Technical Specification Bases for TS [3.1.3, 3.1.4, and 3.3.1.2] consistent with those shown in TSTF–475, Revision 1, ‘‘Control Rod Notch Testing Frequency and SRM Insert Control Rod Action.’’ PO 00000 Frm 00065 Fmt 4703 Sfmt 4703 63937 The commitment should also be marked as optional consistent with Comments 1 and 4, as the PWR-applicable change to Section 1.4 has no associated Bases changes. Response: The staff agrees with the comment in the sense that the Bases are not adopted as a license amendment is adopted, and therefore the wording of the commitment will be revised to state, ‘‘[LICENSEE] will establish the Technical Specification Bases for [TS B 3.1.3, TS B 3.1.4, and TS B 3.3.1.2] consistent with those shown in TSTF– 475, Revision 1, ‘‘Control Rod Notch Testing Frequency and SRM Insert Control Rod Action.’’ The staff does not agree with the comment with respect to the Bases being provided purely for information and that the commitment is optional. The staff will review the Bases changes to ensure they are acceptable. If a licensee is only adopting the TS Section 1.4 portion of the TSTF–475 change, then the commitment would not apply, otherwise it would apply. 7. Comment: Model NSHC: To be consistent with 10 CFR 50.91(a), the title of Criterion 2 should be revised to add the word ‘‘Accident’’ before ‘‘Previously Evaluated.’’ Specifically, it should state, ‘‘The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident from any Accident Previously Evaluated.’’ Response: The staff agrees with the comment and the model NSHC Criterion 2 statement has been reworded accordingly. For the Nuclear Regulatory Commission. Dated at Rockville, Maryland, this 5th day of November, 2007. Timothy J. Kobetz, Chief, Technical Specifications Branch, Division of Inspection and Regional Support, Office of Nuclear Reactor Regulation. Model Safety Evaluation, U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Consolidated Line Item Improvement, Technical Specification Task Force (TSTF) Change TSTF–475, Revision 1, Control Rod Notch Testing Frequency, Source Range Monitor Technical Specification Action to Insert Control Rods, and Surveillance Frequency Discussions 1.0 Introduction By letter dated August 30, 2004, the TSTF submitted a request (Reference 1) for changes to the Standard Technical Specifications (STS): NUREG–1430 Standard Technical Specifications B&W Plants (Reference 2); NUREG–1431 Standard Technical Specifications Westinghouse Plants (Reference 3); NUREG–1432 Standard Technical E:\FR\FM\13NON1.SGM 13NON1 rfrederick on PROD1PC67 with NOTICES 63938 Federal Register / Vol. 72, No. 218 / Tuesday, November 13, 2007 / Notices Specifications Combustion Engineering Plants (Reference 4); NUREG–1433, Standard Technical Specifications General Electric Plants, BWR/4 (Reference 5); and NUREG–1434, Standard Technical Specifications General Electric Plants, BWR/6 (Reference 6). The proposed changes would: (1) Revise the TS control rod notch surveillance frequency in TS 3.1.3, ‘‘Control Rod OPERABILITY,’’ (NUREG–1433 and NUREG–1434), (2) clarify the TS requirement for inserting control rods for one or more inoperable SRMs in MODE 5 (NUREG–1434 only), and (3) 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). These changes are based on Technical Specifications Task Force (TSTF) change traveler TSTF–475, Revision 1, that proposes revisions to the reference STS by: (1) revising the frequency of 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 LPSP of 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), (2) adding the word ‘‘fully’’ to LCO 3.3.1.2 Required Action E.2 (NUREG–1434 only) to clarify the requirement to fully insert all insertable control rods in core cells containing one or more fuel assemblies when the associated SRM instrument is inoperable, and (3) 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 control rod drive (CRD) Mechanism (CRDM), by which the control rods are moved, are components of the CRD System, 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 VerDate Aug<31>2005 15:30 Nov 09, 2007 Jkt 214001 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 an 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 In order to perform this SE, the NRC staff reviewed the following information provided by the TSTF to justify the submitted license amendment request to [revise the weekly control rod notch frequency to monthly (STS NUREG– 1433 and NUREG–1434)], [clarify the SRM TS action for inserting control rods (NUREG–1434 only), and] revise the discussion of the applicability of the 25% allowance in Example 1.4–3. Specifically, the following documents were reviewed during the NRC staff’s evaluation: • TSTF letter TSTF–04–07 (Reference 1)—Provided a description of the PO 00000 Frm 00066 Fmt 4703 Sfmt 4703 proposed changes in TSTF–475 that changes the weekly rod notch frequency to monthly, clarify the SRM TS actions for inserting control rods, and clarify the applicability of the 25% allowance in Example 1.4–3. • [TSTF letter TSTF–06–13 (Reference 8)—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 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. • BWROG letter BWROG–06036 (Reference 9)—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 10)—Provided response to NRC staff RAI on CRD performance in Control Cell Core (CCC) designed plants, including TSTF–475, Revision 1. The CRD System is the primary reactivity control system for the reactor. The CRD System, 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 CRD System 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 CRD System. The CRD System 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 E:\FR\FM\13NON1.SGM 13NON1 rfrederick on PROD1PC67 with NOTICES Federal Register / Vol. 72, No. 218 / Tuesday, November 13, 2007 / Notices 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. According to the 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 the weekly basis. The operating experience related to the changes in CRD performance also provided additional justification to VerDate Aug<31>2005 15:30 Nov 09, 2007 Jkt 214001 reduce the notch test frequency for the partially withdrawn control rods. In response to the NRC staff RAIs and to support their position to reduce the CRD notch testing frequency, the BWROG provided plant data and GE Nuclear Energy report, 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 Technical Specifications. To a lesser degree, cracks have also been noted at the welded joint of the interim design CRT but no cracks haven been observed in the final improved CRT design. In a request for additional information, BWROG response of being unable to find a collet housing failure since 1975 supported the NRC staff review of not finding a collet housing failure. 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 GE’s PLEDGE model with the assumption that the water chemistry condition is based on GE recommendations. The model is based on fundamental principles of stress corrosion cracking which 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 PO 00000 Frm 00067 Fmt 4703 Sfmt 4703 63939 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 HCUs, CRD drives, and control rods are also 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. The NRC staff has reviewed the TSTF–475 proposal to amend the (NUREG–1433 and NUREG–1434) TS SR 3.1.3.2, ‘‘Control Rod OPERABILTY’’ from seven days to monthly. Based on the following evaluation condition: (1) Slow crack growth rate of the CRT; (2) the improved CRT design; (3) a higher 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 utilities should consider the replacement of the CRT, when possible, with the GE CRT improved design. The NRC staff has reviewed the TSTF–475 proposal to amend the NUREG–1434, Specification 3.3.1.2, Required Action E.2 from ‘‘Initiate action to insert all insertable control rods in core cells containing one or E:\FR\FM\13NON1.SGM 13NON1 63940 Federal Register / Vol. 72, No. 218 / Tuesday, November 13, 2007 / Notices more fuel assemblies’’ to ‘‘Initiate action to fully insert all insertable control rods in core cells containing one or more fuel assemblies.’’ The NRC staff finds the revision acceptable because the requirement to insert control rods is meant to require control rods to be fully inserted and adding ‘‘fully’’ does not change but clarifies the intent of the action. The NRC staff has reviewed the TSTF–475 proposal to amend (NUREG– 1430 through NUREG–1434) Example 1.4–3 in Section 1.4 ‘‘Frequency,’’ to make the 1.25 provision in SR 3.0.2 to be equally applicable to time periods specified in the ‘‘FREQUENCY’’ column and in the NOTE in the ‘‘SURVEILLANCE’’ column. The NRC staff finds this change acceptable since the revision would make it consistent with the definition of specified ‘‘Frequency’’ provided in the second paragraph of Section 1.4 which states that the specified ‘‘Frequency’’ is referred to throughout this section and each of the Specifications of Section 3.0, Surveillance Requirement (SR) Applicability. The specified ‘‘Frequency’’ consists of the requirements of the Frequency column of each SR, as well as certain Notes in the Surveillance column that modify performance requirements.’’ rfrederick on PROD1PC67 with NOTICES 3.1 Conclusion The NRC staff has reviewed the licensee’s proposal to amend existing [(NUREG–1433 and NUREG–1434) TS sections SR 3.1.3.2, ‘‘Control Rod OPERABILTY,’’ (NUREG–1434) LCO 3.3.1.2 Required Action E.2, ‘‘Source Range Monitor (SRM) Instrumentation,’’ and] (NUREG–1430 through NUREG– 1434) Example 1.4–3, ‘‘Frequency’’ applicable to SR 3.0.2. The NRC staff has concluded that the TS revisions [will have a minimal affect on the high reliability of the CRD system while reducing the opportunity for potential reactivity events; thus, meeting the requirement of CFR, Part 50, Appendix A, GDC 29, and] will clarify the 1.25 provision in SR 3.0.2. Therefore, the staff concludes that the amendment request is acceptable. Based on the considerations discussed above, the Commission has concluded 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. VerDate Aug<31>2005 15:30 Nov 09, 2007 Jkt 214001 4.0 State Consultation In accordance with the Commission’s regulations, the [ ] State official was notified of the proposed issuance of the amendment. The State official had [(1) no comments or (2) the following comments—with subsequent disposition by the staff]. 5.0 Environmental Consideration The amendments change a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR part 20 and change surveillance requirements. The NRC staff has determined that the amendments involve 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 amendments involve no significant hazards considerations, and there has been no public comment on the finding [FR ]. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) [and (c)(10)]. Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments. 6.0 Conclusion The Commission has concluded, on the basis of 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. 7.0 References 1. Letter TSTF–04–07 from the Technical Specifications Task Force 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–1430, ‘‘Standard Technical Specifications Babcock and Wilcox Plants, Revision 3,’’ August 31, 2003. 3. NUREG–1431, ‘‘Standard Technical Specifications Westinghouse Plants, Revision 3,’’ August 31, 2003. 4. NUREG–1432, ‘‘Standard Technical Specifications Combustion Engineering Plants, Revision 3,’’ August 31, 2003. PO 00000 Frm 00068 Fmt 4703 Sfmt 4703 5. NUREG–1433, ‘‘Standard Technical Specifications General Electric Plants, BWR/4, Revision 3,’’ August 31, 2003. 6. NUREG–1434, ‘‘Standard Technical Specifications General Electric Plants, BWR/6, Revision 3,’’ August 31, 2003. 7. Letter TSTF–07–19, Response from the Technical Specifications Task Force 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. 8. Letter TSTF–06–13 from the Technical Specifications Task Force to the NRC, ‘‘Response to NRC Request for Additional Information Regarding TSTF–475, Revision 0,’’ dated July 3, 2006, ADAMS accession number ML0618403421. 9. Letter BWROG–06036 from the BWR Owners Group 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. 10. Letter TSTF–07–19 from the Technical Specifications Task Force to the NRC, ‘‘Response to NRC Request for Additional Information Regarding TSTF–475, Revision 0,’’ dated May 22, 2007, ADAMS accession number ML071420428]. THE FOLLOWING EXAMPLE OF AN APPLICATION WAS PREPARED BY THE NRC STAFF TO FACILITATE USE OF THE CONSOLIDATED LINE ITEM IMPROVEMENT PROCESS (CLIIP). THE MODEL PROVIDES THE EXPECTED LEVEL OF DETAIL AND CONTENT FOR AN APPLICATION TO REVISE TECHNICAL SPECIFICATIONS REGARDING REVISION OF CONTROL ROD NOTCH SURVEILLANCE TEST FREQUENCY, CLARIFICATION OF SRM INSERT CONTROL ROD ACTION, AND A CLARIFICATION OF A FREQUENCY EXAMPLE. LICENSEES REMAIN RESPONSIBLE FOR ENSURING THAT THEIR ACTUAL APPLICATION FULFILLS THEIR ADMINISTRATIVE REQUIREMENTS AS WELL AS NUCLEAR REGULATORY COMMISSION REGULATIONS. U.S. Nuclear Regular Commission Document Control Desk Washington, DC 20555 SUBJECT: PLANT NAME, DOCKET NO. 50— APPLICATION FOR TECHNICAL SPECIFICATION CHANGE REGARDING REVISION OF CONTROL ROD NOTCH SURVEILLANCE TEST FREQUENCY, CLARIFICATION OF SRM INSERT CONTROL ROD ACTION, AND A E:\FR\FM\13NON1.SGM 13NON1 Federal Register / Vol. 72, No. 218 / Tuesday, November 13, 2007 / Notices CLARIFICATION OF A FREQUENCY EXAMPLE USING THE CONSOLIDATED LINE ITEM IMPROVEMENT PROCESS Gentleman: In accordance with the provisions of 10 CFR 50.90 [LICENSEE] is submitting a request for an amendment to the technical specifications (TS) for [PLANT NAME, UNIT NOS.]. The proposed amendment would: (1) [revise the TS surveillance requirement (SR) frequency in TS 3.1.3, ‘‘Control Rod OPERABILITY’’, (2) clarify the requirement to fully insert all insertable control rods for the limiting condition for operation (LCO) in TS 3.3.1.2, required Action E.2, ‘‘Source Range Monitoring Instrumentation,’’ and (3)] revise Example 1.4–3 in Section 1.4 ‘‘Frequency’’ to clarify the applicability of the 1.25 surveillance test interval extension. Attachment 1 provides a description of the proposed change, the requested confirmation of applicability, and plant-specific verifications. Attachment 2 provides the existing TS pages marked up to show the proposed change. Attachment 3 provides revised (clean) TS pages. Attachment 4 provides a summary of the regulatory commitments made in this submittal. [LICENSEE] requests approval of the proposed License Amendment by [DATE], with the amendment being implemented [BY DATE OR WITHIN X DAYS]. In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated [STATE] Official. I declare [or certify, verify, state] under penalty of perjury that the foregoing is true and correct. If you should have any questions regarding this submittal, please contact [NAME, TELEPHONE NUMBER]. Sincerely, [Name, Title] Attachments: 1. Description and Assessment 2. Proposed Technical Specification Changes 3. Revised Technical Specification Pages 4. Regulatory Commitments 5. Proposed Technical Specification Bases Changes] cc: NRC Project Manager NRC Regional Office NRC Resident Inspector State Contact Attachment 1—Description and Assessment rfrederick on PROD1PC67 with NOTICES 1.0 Description The proposed amendment would: (1) [Revise the TS surveillance requirement (SR 3.1.3.2) frequency in TS 3.1.3, ‘‘Control Rod OPERABILITY’’, (2) clarify the requirement to fully insert all insertable control rods for the limiting condition for operation (LCO) in TS 3.3.1.2, Required Action E.2, ‘‘Source Range Monitoring Instrumentation’’, and (3)] revise Example 1.4–3 in Section 1.4 ‘‘Frequency’’ to clarify the applicability of the 1.25 surveillance test interval extension. The changes are consistent with Nuclear Regulatory Commission (NRC) approved Industry/Technical Specification Task Force (TSTF) STS change TSTF–475, Revision 1. The Federal Register notice published on [DATE] announced the availability of this TS improvement through the consolidated line item improvement process (CLIIP). 2.0 Assessment 2.1 Applicability of Published Safety Evaluation [LICENSEE] has reviewed the safety evaluation dated [DATE] as part of the CLIIP. This review included a review of the NRC staff’s evaluation, as well as the supporting information provided to support TSTF–475, Revision 1. [LICENSEE] has concluded that the justifications presented in the TSTF proposal and the safety evaluation prepared by the NRC staff are applicable to [PLANT, UNIT NOS.] and justify this amendment for the incorporation of the changes to the [PLANT] TS. 2.2 Optional Changes and Variations [LICENSEE] is not proposing any variations or deviations from the TS changes described in the modified TSTF–475, Revision 1 and the NRC staff’s model safety evaluation dated [DATE]. 3.0 Regulatory Analysis 3.1 No Significant Hazards Consideration Determination [LICENSEE] has reviewed the proposed no significant hazards consideration determination (NSHCD) published in the Federal Register as part of the CLIIP. [LICENSEE] has concluded that the proposed NSHCD presented in the Federal Register notice is applicable to [PLANT] and is hereby incorporated by reference to satisfy the requirements of 10 CFR 50.91(a). 3.2 Verification and Commitments As discussed in the notice of availability published in the Federal 63941 Register on [DATE] for this TS improvement, the [LICENSEE] verifies the applicability of TSTF–475 to [PLANT], and commits to establishing Technical Specification Bases for TS as proposed in TSTF–475, Revision 1. These changes are based on TSTF change traveler TSTF–475 (Revision 1) that proposes revisions to the STS by: (1) [Revising the frequency of SR 3.1.3.2, notch testing of fully withdrawn control rod, from ‘‘7 days after the control rod is withdrawn and THERMAL POWER is greater than the LPSP of RWM’’ to ‘‘31 days after the control rod is withdrawn and THERMAL POWER is greater than the LPSP of the RWM’’, (2) adding the word ‘‘fully’’ to LCO 3.3.1.2 Required Action E.2 to clarify the requirement to fully insert all insertable control rods in core cells containing one or more fuel assemblies when the associated SRM instrument is inoperable, and (3)] 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. 4.0 Environmental Evaluation [LICENSEE] has reviewed the environmental evaluation included in the model safety evaluation dated [DATE] as part of the CLIIP. [LICENSEE] has concluded that the staff’s findings presented in that evaluation are applicable to [PLANT] and the evaluation is hereby incorporated by reference for this application. ATTACHMENT 2—PROPOSED TECHNICAL SPECIFICATION CHANGES (MARK-UP) ATTACHMENT 3—PROPOSED TECHNICAL SPECIFICATION PAGES ATTACHMENT 4—LIST OF REGULATORY COMMITMENTS The following table identifies those actions committed to by [LICENSEE] in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments. Please direct questions regarding these commitments to [CONTACT NAME]. Regulatory commitments Due date/event [[LICENSEE] will establish the Technical Specification Bases for [TS B 3.1.3, TS B 3.1.4, and TS B 3.3.1.2] consistent with those shown in TSTF–475, Revision 1, ‘‘Control Rod Notch Testing Frequency and SRM Insert Control Rod Action.’’]. [Complete, implemented with amendment OR within X days of implementation of amendment]. VerDate Aug<31>2005 15:30 Nov 09, 2007 Jkt 214001 PO 00000 Frm 00069 Fmt 4703 Sfmt 4703 E:\FR\FM\13NON1.SGM 13NON1 63942 Federal Register / Vol. 72, No. 218 / Tuesday, November 13, 2007 / Notices ATTACHMENT 5—PROPOSED CHANGES TO TECHNICAL SPECIFICATION BASES PAGES [Not required for plants only adopting portion of TSTF–475 change pertaining to TS Section 1.4 that provides example to SR Frequency] rfrederick on PROD1PC67 with NOTICES Proposed No Significant Hazards Consideration Determination Description of Amendment Request: [Plant Name] requests adoption of an approved change to the Standard Technical Specifications (STS) for [General Electric (GE) Plants (NUREG– 1433, BWR/4 and NUREG–1434, BWR/ 6) and] plant specific technical specifications (TS), that allows: (1) [revising the frequency of SR 3.1.3.2, notch testing of fully withdrawn control rod, from ‘‘7 days after the control rod is withdrawn and THERMAL POWER is greater than the LPSP of RWM’’ to ‘‘31 days after the control rod is withdrawn and THERMAL POWER is greater than the LPSP of the RWM’’, (2) adding the word ‘‘fully’’ to LCO 3.3.1.2 Required Action E.2 to clarify the requirement to fully insert all insertable control rods in core cells containing one or more fuel assemblies when the associated SRM instrument is inoperable, and (3)] 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. The staff finds that the proposed STS changes are acceptable [because the number of control rod manipulations is reduced thereby reducing the opportunity for potential reactivity events while having a very minimal impact on the extremely high reliability of the CRD system as discussed in the technical evaluation section of this safety evaluation and] the discussion of the SR Frequency example provides clarification. Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration is presented below: Criterion 1—The Proposed Change Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated The proposed change generically implements TSTF–475, Revision 1, ‘‘Control Rod Notch Testing Frequency and SRM Insert Control Rod Action.’’ TSTF–475, Revision 1 modifies NUREG–1433 (BWR/4) and NUREG–1434 (BWR/6) STS. The changes: (1) revise TS testing frequency for surveillance requirement (SR) 3.1.3.2 in TS VerDate Aug<31>2005 15:30 Nov 09, 2007 Jkt 214001 3.1.3, ‘‘Control Rod OPERABILITY’’, (2) clarify the requirement to fully insert all insertable control rods for the limiting condition for operation (LCO) in TS 3.3.1.2, Required Action E.2, ‘‘Source Range Monitoring Instrumentation’’ (NUREG–1434 only), and (3) revise Example 1.4–3 in Section 1.4 ‘‘Frequency’’ to clarify the applicability of the 1.25 surveillance test interval extension. The consequences of an accident after adopting TSTF–475, Revision 1 are no different than the consequences of an accident prior to adoption. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated. Criterion 2—The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident from any Accident Previously Evaluated The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. The proposed change will not introduce new failure modes or effects and will not, in the absence of other unrelated failures, lead to an accident whose consequences exceed the consequences of accidents previously analyzed. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated. Criterion 3—The Proposed Change Does Not Involve a Significant Reduction in the Margin of Safety TSTF–475, Revision 1 will: (1) [revise the TS SR 3.1.3.2 frequency in TS 3.1.3, ‘‘Control Rod OPERABILITY’’, (2) clarify the requirement to fully insert all insertable control rods for the limiting condition for operation (LCO) in TS 3.3.1.2, ‘‘Source Range Monitoring Instrumentation,’’ and (3)] revise Example 1.4–3 in Section 1.4 ‘‘Frequency’’ to clarify the applicability of the 1.25 surveillance test interval extension. [The GE Nuclear Energy Report, ‘‘CRD Notching Surveillance Testing for Limerick Generating Station,’’ dated November 2006, concludes that extending the control rod notch test interval from weekly to monthly is not expected to impact the reliability of the scram system and that the analysis supports the decision to change the surveillance frequency.] Therefore, the proposed changes in TSTF–475, Revision 1 are acceptable and do not involve a significant reduction in a margin of safety. Based upon the reasoning presented above and the previous discussion of the amendment request, the requested change does not involve a significant hazards consideration. Dated at Rockville, Maryland, this 5th day of November, 2007. For the Nuclear Regulatory Commission. Timothy J. Kobetz, Section Chief, Technical Specifications Branch, Division of Inspection & Regional Support, Office of Nuclear Reactor Regulation. [FR Doc. E7–22159 Filed 11–9–07; 8:45 am] BILLING CODE 7590–01–P PO 00000 Frm 00070 Fmt 4703 Sfmt 4703 NUCLEAR REGULATORY COMMISSION NUREG–1556, Volume 21, ‘‘Consolidated Guidance About Materials Licenses Program-Specific Guidance About Possession Licenses for Production of Radioactive Material Using an Accelerator’’ Nuclear Regulatory Commission. ACTION: Notice of availability. AGENCY: SUMMARY: The Nuclear Regulatory Commission (NRC) is announcing the completion and availability of NUREG– 1556, Volume 21, ‘‘Consolidated Guidance About Materials Licenses, Program-Specific Guidance About Possession Licenses for Production of Radioactive Material Using an Accelerator,’’ dated October 2007. ADDRESSES: Copies of NUREG–1556, Volume 21, may be purchased from the Superintendent of Documents, U.S. Government Printing Office, P.O. Box 37082, Washington, DC 20402–9328; www.access.gpo.gov/su_docs, 202–512– 1800 or The National Technical Information Service, Springfield, Virginia 22161–0002; www.ntis.gov; 1– 800–533–6847 or, locally, 703–805– 6000. A copy of the document is also available for inspection and/or copying for a fee in the NRC Public Document Room, 11555 Rockville Pike, Rockville, Maryland. Publicly available documents created or received at the NRC after November 1, 1999, are available electronically at the NRC’s Electronic Reading Room at http://www.nrc.gov/ NRC/ADAMS/index.html. From this site, the public can gain entry into the NRC’s Agencywide Document Access and Management System (ADAMS), which provides text and image files of the NRC’s public documents. The ADAMS Accession Number for NUREG–1556, Volume 21 is ML072900058. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the NRC PDR Reference staff at 1–800–397–4209, 301– 415–4737, or by e-mail to pdr@nrc.gov. The document will also be posted on NRC’s public Web site at: http:// www.nrc.gov/reading-rm/doccollections/nuregs/staff/sr1556/ on the ‘‘Consolidated Guidance About Materials Licenses (NUREG–1556)’’ Web site page, and on the Office of Federal and State Materials and Environmental Management Programs’ NARM (Naturally-Occurring and AcceleratorProduced Radioactive Material) Toolbox Web site page at: http://nrc-stp.ornl.gov/ E:\FR\FM\13NON1.SGM 13NON1

Agencies

[Federal Register Volume 72, Number 218 (Tuesday, November 13, 2007)]
[Notices]
[Pages 63935-63942]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E7-22159]


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NUCLEAR REGULATORY COMMISSION


Notice of Availability of Model Application Concerning Technical 
Specification Improvement To Revise Control Rod Notch Surveillance 
Frequency, Clarify SRM Insert Control Rod Action, and Clarify Frequency 
Example

AGENCY: Nuclear Regulatory Commission.

ACTION: Notice of availability.

-----------------------------------------------------------------------

SUMMARY: Notice is hereby given that the staff of the Nuclear 
Regulatory Commission (NRC) has prepared a model safety evaluation (SE) 
relating to the revision of Standard Technical Specifications (STS), 
NUREG-1430 (B&W), NUREG-1431 (Westinghouse), NUREG-1432 (CE), NUREG-
1433 (BWR/4) and NUREG-1434 (BWR/6). Specifically the SE addresses: (1) 
The revision of the technical specification (TS) surveillance 
requirement (SR) 3.1.3.2 frequency in STS 3.1.3, ``Control Rod 
OPERABILITY,'' (NUREG-1433 and NUREG-1434), (2) a clarification to the 
requirement to fully insert all insertable control rods for the 
limiting condition for operation (LCO) in STS 3.3.1.2, Required Action 
E.2, ``Source Range Monitor Instrumentation'' (NUREG-1434 only), and 
(3) the revision of Example 1.4-3 in STS Section 1.4 ``Frequency'' to 
clarify the applicability of the 1.25 surveillance test interval 
extension (NUREG-1430 through NUREG-1434). The NRC staff has also 
prepared a model license amendment request and a model no significant 
hazards consideration (NSHC) determination relating to this matter. The 
purpose of these models is to permit the NRC to efficiently process 
amendments that propose to modify TS control rod SR testing frequency, 
clarify TS control insertion requirements, and clarify SR frequency 
discussions. Licensees of nuclear power reactors to which the models 
apply can request amendments, confirming the applicability of the SE 
and NSHC determination to their plant licensing basis.

DATES: The NRC staff issued a Federal Register notice (72 FR 46103; 
August 16, 2007) which provided a model SE, model application, and 
model NSHC related to BWR plant control rod notch surveillance 
frequency, BWR SRM control rod insertion action, and clarification of a 
surveillance frequency example for all plant types. Similarly, the NRC 
staff herein provides a revised model SE, model LAR, and model NSHC 
incorporating changes based upon the public comments received. The NRC 
staff can most efficiently consider applications based upon the model 
LAR, which references the model SE, if the LAR is submitted within one 
year of this Federal Register Notice.

FOR FURTHER INFORMATION CONTACT: Timothy Kobetz, Mail Stop: O-12H2, 
Technical Specifications Branch, Division of Inspection & Regional 
Support, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, telephone: 301-415-1932.

SUPPLEMENTARY INFORMATION: 

Background

    Regulatory Issue Summary 2000-06, ``Consolidated Line Item 
Improvement Process for Adopting Standard Technical Specification 
Changes for Power Reactors,'' was issued on March 20, 2000. The 
consolidated line item improvement process (CLIIP) is intended to 
improve the efficiency of NRC licensing processes by processing 
proposed changes to the standard technical specifications (STS) in a 
manner that supports subsequent license amendment applications. The 
CLIIP includes an opportunity for the public to comment on proposed 
changes to the STS following a preliminary assessment by the NRC staff 
and finding that the change will likely be offered for adoption by 
licensees. The CLIIP directs the NRC staff to evaluate any comments 
received for a proposed change to the STS and to either reconsider the 
change or to proceed with announcing the availability of the change for 
proposed

[[Page 63936]]

adoption by licensees. Those licensees opting to apply for the subject 
change to technical specifications are responsible for reviewing the 
staff's evaluation, referencing the applicable technical 
justifications, and providing any necessary plant-specific information. 
Each amendment application made in response to the notice of 
availability will be processed and noticed in accordance with 
applicable rules and NRC procedures.
    This notice involves the modification of BWR TS control rod SR 
testing frequency, clarification of BWR TS control insertion 
requirements, and clarification of SR frequency discussions for all 
pant types. This change was proposed for incorporation into the 
standard technical specifications by the Owners Groups participants in 
the Technical Specification Task Force (TSTF) and is designated TSTF-
475 Revision 1. TSTF-475 Revision 1 can be viewed on the NRC's Web page 
at http://www.nrc.gov/reactors/operating/licensing/techspecs.html.

*** Reviewer's Note ***

    TSTF-475 involves three changes to the Standard Technical 
Specifications NUREGs that, depending upon the adopting plant, may 
or may not be adopted by a plant. The first changes the surveillance 
frequency for control rod notch testing from 7 to 31 days, and 
applies to BWR/4 and BWR/6 plants (NUREG-1433 & NUREG-1434). The 
second adds the word ``fully'' to a Required Action statement to 
clarify that control rods should be fully inserted, and applies to 
only the BWR/6 plants (NUREG-1434). The third change clarifies the 
usage of the 1.25 surveillance frequency interval extension, and 
applies to all plants (NUREG-1430 through NUREG-1434). The model 
application and model safety evaluation will need to be tailored 
(where brackets indicate) for plant specific applications.

Applicability

    This proposed TS change modifies TS control rod SR testing 
frequency and clarifies TS control insertion requirements for BWR 
plants, and clarifies SR frequency discussions for all NSSS plant 
types. The CLIIP does not prevent licensees from requesting an 
alternative approach or proposing the changes without the attached 
model SE and the NSHC. Variations from the approach recommended in this 
notice may, however, require additional review by the NRC staff and may 
increase the time and resources needed for the review.
    To efficiently process the incoming license amendment applications, 
the staff requests that each licensee applying for the changes proposed 
in TSTF-475, Revision 1, include TS Bases for the proposed TS 
consistent with the TS Bases proposed in TSTF-475, Revision 1 (note: 
the change to STS Section 1.4 does not entail a Bases change). The 
staff is requesting that the TS Bases be included with the proposed 
license amendments in this case because the changes to the TS and the 
changes to the associated TS Bases form an integral change to a plant's 
licensing basis. To ensure that the overall change, including the TS 
Bases, includes appropriate regulatory controls, the staff plans to 
condition the issuance of each license amendment on the licensee's 
incorporation of the changes into the TS Bases document and that the 
licensee control changes to the TS Bases in accordance with the 
licensees TS Bases Control Program. The CLIIP does not prevent 
licensees from requesting an alternative approach or proposing the 
changes without the requested TS Bases. However, deviations from the 
approach recommended in this notice may require additional review by 
the NRC staff and may increase the time and resources needed for the 
review. Significant variations from the approach, or inclusion of 
additional changes to the license, will result in staff rejection of 
the submittal. Instead, licensees desiring significant variations and/
or additional changes should submit a LAR that does not request to 
adopt TSTF-475, Revision 1, under CLIIP.

Public Notices

    The staff issued a Federal Register Notice (72 FR 46103, August 16, 
2007) that requested public comment on the NRC's pending action to 
approve the modification of BWR TS control rod SR testing frequency, 
clarification of BWR TS control insertion requirements, and 
clarification of SR frequency discussions for all pant types, as 
proposed in TSTF-475, Revision 1. The TSTF-475, Revision 1, can be 
viewed on the NRC's web page at http://www.nrc.gov/reactors/operating/
licensing/techspecs.html. TSTF-475, Revision 1, may be examined, and/or 
copied for a fee, at the NRC's Public Document Room, located at One 
White Flint North, 11555 Rockville Pike (first floor), Rockville, 
Maryland. Publicly available records are accessible electronically from 
the ADAMS Public Library component on the NRC Web site, (the Electronic 
Reading Room) at http://www.nrc.gov/reading-rm/adams.html.
    In response to the notice soliciting comments from interested 
members of the public about the modification of BWR TS control rod SR 
testing frequency, clarification of BWR TS control insertion 
requirements, and clarification of SR frequency discussions for all 
pant types, the staff received one set of comments (from the TSTF 
Owners Groups, representing licensees). The specific comments are 
provided and discussed below:
    1. Comment: TSTF-475 contains three changes: The revision to SR 
3.1.3.2 which is applicable to NUREG-1433 and NUREG-1434 (the Improved 
Standard Technical Specifications, or ISTS, for BWR/4 and BWR/6 
plants), the change to Specification 3.3.1.2, Required Action E.2 which 
is applicable to NUREG-1434 (the ISTS for BWR/6 plants), and the change 
to Example 1.4-3 which is applicable to NUREG-1430 through -1434 (the 
ISTS for all plant types). The applicability of the third change to all 
plant types is clearly indicated on the Traveler cover page and in the 
justification (last paragraph of Section 2.0, ``Proposed Change.'') 
However, the Notice for Comment, model Safety Evaluation, model 
application, and No Significant Hazards Considerations Determination 
(NSHC) incorrectly state that TSTF-475 is only applicable to BWR 
plants.
    The Notice, the model application, model Safety Evaluation, and 
NSHC should be revised to state that the change to Example 1.4-3 is 
applicable to all plant types. The model Safety Evaluation, model 
application, and NSHC should be revised to bracket (e.g., indicate as 
optional) the BWR/4 and BWR/6 specific changes so that the documents 
are applicable to a BWR/6 plant adopting all three changes, a BWR/4 
plant adopting the SR 3.1.3.2 and Example 1.4-3 changes, or a 
pressurized water reactor (PWR) plant adopting only the Example 1.4-3 
change.
    Response: The staff agrees with the comment and the model 
application, model Safety Evaluation, and NSHC have been revised 
accordingly.
    2. Comment: In Section 3.0, ``Technical Evaluation,'' of the 
Notice, reference is made three times to the ``BWROG TSTF'' or ``BWROG 
TSTF-475.'' The Technical Specifications Task Force (TSTF) is sponsored 
by the Boiling Water Reactor Owners Group and the Pressurized Water 
Reactor Owners Group. The proper designation is either ``TSTF'' or 
``Owners Group TSTF.''
    Response: The staff agrees with the comment and Section 3.0 of the 
model Safety Evaluation has been revised by removing explicit reference 
to the BWROG in referring to TSTF-475.
    3. Comment: In Section 3.0, ``Technical Evaluation,'' the model 
Safety Evaluation states, ``Therefore, the NRC staff finds the change 
acceptable

[[Page 63937]]

with the commitment to implement GE water quality for the CRD system 
recommendations.'' In the model application, a regulatory commitment is 
included which states, ``[LICENSEE] will establish the water quality 
controls as recommended by SIL No. 148, Water Quality Control for the 
Control Rod System,'' September 15, 1975.'' This commitment should be 
removed.
    The TSTF's justification for TSTF-475 made no mention of and did 
not rely on water quality controls. The TSTF's July 3, 2006 response to 
the NRC's March 21, 2003 Request for Additional Information (RAI) did 
not credit water chemistry controls. As stated in the justification and 
the Staff's model Safety Evaluation, 30 years of operating experience 
at BWRs without a control rod drive failure detected by the weekly 
notch testing is sufficient to demonstrate the acceptability of the 
change.
    The reference is technically incorrect. Supplement 1 to SIL No. 148 
was issued in June 2004 and updates the SIL to bring it into alignment 
with current Electric Power Research Institute (EPRI) BWR water 
chemistry requirements, which were in conflict with the 1975 version of 
SIL.
    The NRC's Technical Evaluation in the draft Safety Evaluation did 
not reference SIL No. 148 (either the 1975 version or the current 
version).
    It is not appropriate for the NRC to require commitments to 
documents that were not relied on in the licensee's application, were 
not reviewed by the NRC, and were not discussed in the NRC's technical 
evaluation. Therefore, the reference to water chemistry controls in the 
model Safety Evaluation and the commitment in the model application 
should be removed.
    Response: The staff agrees with the comment and the requirements 
for a commitment to establish water quality controls as recommended by 
SIL No. 148, Water Quality Control for the Control Rod System, in the 
model Safety Evaluation and in the model application have been removed.
    4. Comment: Model Application: Attachment 5, ``Proposed Technical 
Specification Bases,'' should be marked as optional. There are no Bases 
changes associated with the PWR-applicable changes to Section 1.4. 
Furthermore, the Bases changes associated with TSTF-475 simply reflect 
the changes made to the specifications. It should be left to the 
licensee whether to submit Bases changes with the amendment request. 
The third paragraph omits Attachment 5, which is shown in the list of 
attachments below the signature. Attachment 3, ``Proposed Technical 
Specification Pages,'' should also be marked as optional as not all 
licensee's submit retyped Technical Specification pages as attachments 
to their amendment requests.
    Response: The staff does not agree with the comment. For those 
sections of the technical specifications that are changed in accordance 
with TSTF-475 and that have Bases, the Bases must be changed to reflect 
the change in accordance with TSTF-475. TS Section 1.4, that does not 
have Bases, does not need to have Bases changes submitted, and for 
those plants that are only adopting the TS Section 1.4 change, the 
Model Application Attachment 5, ``Proposed Technical Specification 
Bases,'' will be revised to indicate that the submittal of revised 
Bases pages is optional in that case. The staff does not see a need to 
revise Model Application Attachment 3. The staff expects to see the 
licensee's Bases changes associated with the adoption of TSTF-475.
    5. Comment: Model Application: The Model Application states, ``I 
declare under penalty of perjury under the laws of the United States of 
America that I am authorized by [LICENSEE] to make this request and 
that the foregoing is true and correct.'' This statement is not 
consistent with the recommended statement given in RIS 2001-18, 
``Requirements for Oath or Affirmation.'' RIS 2001-18 recommends the 
statement, ``I declare [or certify, verify, state] under penalty of 
perjury that the foregoing is true and correct.'' Note that RIS 2001-18 
states that this statement must be used verbatim. We recommend that the 
Model Application be revised to be consistent with RIS 2001-18.
    Response: The staff agrees with the comment and the requirement in 
the model application for oath or affirmation has been reworded to be 
consistent with RIS 2001-18.
    6. Comment: Attachment 4: The regulatory commitment states 
``[LICENSEE] will establish the Technical Specification Bases for [TS B 
3.1.3, TS B 3.1.4, and TS B 3.3.1.2] as adopted with the applicable 
license amendment.'' This statement is incorrect as the Bases changes 
included for information with the license amendment request are not 
``adopted'' with the license amendment. Bases changes are made under 
licensee control under the Technical Specification Bases Control 
Program. We recommend revising the commitment to state ``[LICENSEE] 
will implement Technical Specification Bases for TS [3.1.3, 3.1.4, and 
3.3.1.2] consistent with those shown in TSTF-475, Revision 1, ``Control 
Rod Notch Testing Frequency and SRM Insert Control Rod Action.'' The 
commitment should also be marked as optional consistent with Comments 1 
and 4, as the PWR-applicable change to Section 1.4 has no associated 
Bases changes.
    Response: The staff agrees with the comment in the sense that the 
Bases are not adopted as a license amendment is adopted, and therefore 
the wording of the commitment will be revised to state, ``[LICENSEE] 
will establish the Technical Specification Bases for [TS B 3.1.3, TS B 
3.1.4, and TS B 3.3.1.2] consistent with those shown in TSTF-475, 
Revision 1, ``Control Rod Notch Testing Frequency and SRM Insert 
Control Rod Action.'' The staff does not agree with the comment with 
respect to the Bases being provided purely for information and that the 
commitment is optional. The staff will review the Bases changes to 
ensure they are acceptable. If a licensee is only adopting the TS 
Section 1.4 portion of the TSTF-475 change, then the commitment would 
not apply, otherwise it would apply.
    7. Comment: Model NSHC: To be consistent with 10 CFR 50.91(a), the 
title of Criterion 2 should be revised to add the word ``Accident'' 
before ``Previously Evaluated.'' Specifically, it should state, ``The 
Proposed Change Does Not Create the Possibility of a New or Different 
Kind of Accident from any Accident Previously Evaluated.''
    Response: The staff agrees with the comment and the model NSHC 
Criterion 2 statement has been reworded accordingly.


    For the Nuclear Regulatory Commission.
    Dated at Rockville, Maryland, this 5th day of November, 2007.
Timothy J. Kobetz,
Chief, Technical Specifications Branch, Division of Inspection and 
Regional Support, Office of Nuclear Reactor Regulation.

Model Safety Evaluation, U.S. Nuclear Regulatory Commission, Office of 
Nuclear Reactor Regulation, Consolidated Line Item Improvement, 
Technical Specification Task Force (TSTF) Change TSTF-475, Revision 1, 
Control Rod Notch Testing Frequency, Source Range Monitor Technical 
Specification Action to Insert Control Rods, and Surveillance Frequency 
Discussions

1.0 Introduction

    By letter dated August 30, 2004, the TSTF submitted a request 
(Reference 1) for changes to the Standard Technical Specifications 
(STS): NUREG-1430 Standard Technical Specifications B&W Plants 
(Reference 2); NUREG-1431 Standard Technical Specifications 
Westinghouse Plants (Reference 3); NUREG-1432 Standard Technical

[[Page 63938]]

Specifications Combustion Engineering Plants (Reference 4); NUREG-1433, 
Standard Technical Specifications General Electric Plants, BWR/4 
(Reference 5); and NUREG-1434, Standard Technical Specifications 
General Electric Plants, BWR/6 (Reference 6). The proposed changes 
would: (1) Revise the TS control rod notch surveillance frequency in TS 
3.1.3, ``Control Rod OPERABILITY,'' (NUREG-1433 and NUREG-1434), (2) 
clarify the TS requirement for inserting control rods for one or more 
inoperable SRMs in MODE 5 (NUREG-1434 only), and (3) 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).
    These changes are based on Technical Specifications Task Force 
(TSTF) change traveler TSTF-475, Revision 1, that proposes revisions to 
the reference STS by: (1) revising the frequency of 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 LPSP of 
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), (2) 
adding the word ``fully'' to LCO 3.3.1.2 Required Action E.2 (NUREG-
1434 only) to clarify the requirement to fully insert all insertable 
control rods in core cells containing one or more fuel assemblies when 
the associated SRM instrument is inoperable, and (3) 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 control rod drive (CRD) 
Mechanism (CRDM), by which the control rods are moved, are components 
of the CRD System, 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 an 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

    In order to perform this SE, the NRC staff reviewed the following 
information provided by the TSTF to justify the submitted license 
amendment request to [revise the weekly control rod notch frequency to 
monthly (STS NUREG-1433 and NUREG-1434)], [clarify the SRM TS action 
for inserting control rods (NUREG-1434 only), and] revise the 
discussion of the applicability of the 25% allowance in Example 1.4-3. 
Specifically, the following documents were reviewed during the NRC 
staff's 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, clarify the SRM TS actions for 
inserting control rods, and clarify the applicability of the 25% 
allowance in Example 1.4-3.
     [TSTF letter TSTF-06-13 (Reference 8)--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 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.
     BWROG letter BWROG-06036 (Reference 9)--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 10)--Provided response 
to NRC staff RAI on CRD performance in Control Cell Core (CCC) designed 
plants, including TSTF-475, Revision 1.
    The CRD System is the primary reactivity control system for the 
reactor. The CRD System, 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 CRD 
System 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 CRD System.
    The CRD System 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

[[Page 63939]]

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.
    According to the 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 the 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 the NRC staff RAIs and to support their position to 
reduce the CRD notch testing frequency, the BWROG provided plant data 
and GE Nuclear Energy report, 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 Technical Specifications. To a lesser degree, cracks 
have also been noted at the welded joint of the interim design CRT but 
no cracks haven been observed in the final improved CRT design. In a 
request for additional information, BWROG response of being unable to 
find a collet housing failure since 1975 supported the NRC staff review 
of not finding a collet housing failure. 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 GE's PLEDGE model with the 
assumption that the water chemistry condition is based on GE 
recommendations. The model is based on fundamental principles of stress 
corrosion cracking which 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 HCUs, CRD drives, and control rods are also 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.
    The NRC staff has reviewed the TSTF-475 proposal to amend the 
(NUREG-1433 and NUREG-1434) TS SR 3.1.3.2, ``Control Rod OPERABILTY'' 
from seven days to monthly. Based on the following evaluation 
condition: (1) Slow crack growth rate of the CRT; (2) the improved CRT 
design; (3) a higher 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 utilities should 
consider the replacement of the CRT, when possible, with the GE CRT 
improved design.
    The NRC staff has reviewed the TSTF-475 proposal to amend the 
NUREG-1434, Specification 3.3.1.2, Required Action E.2 from ``Initiate 
action to insert all insertable control rods in core cells containing 
one or

[[Page 63940]]

more fuel assemblies'' to ``Initiate action to fully insert all 
insertable control rods in core cells containing one or more fuel 
assemblies.'' The NRC staff finds the revision acceptable because the 
requirement to insert control rods is meant to require control rods to 
be fully inserted and adding ``fully'' does not change but clarifies 
the intent of the action.
    The NRC staff has reviewed the TSTF-475 proposal to amend (NUREG-
1430 through NUREG-1434) Example 1.4-3 in Section 1.4 ``Frequency,'' to 
make the 1.25 provision in SR 3.0.2 to be equally applicable to time 
periods specified in the ``FREQUENCY'' column and in the NOTE in the 
``SURVEILLANCE'' column. The NRC staff finds this change acceptable 
since the revision would make it consistent with the definition of 
specified ``Frequency'' provided in the second paragraph of Section 1.4 
which states that the specified ``Frequency'' is referred to throughout 
this section and each of the Specifications of Section 3.0, 
Surveillance Requirement (SR) Applicability. The specified 
``Frequency'' consists of the requirements of the Frequency column of 
each SR, as well as certain Notes in the Surveillance column that 
modify performance requirements.''

3.1 Conclusion

    The NRC staff has reviewed the licensee's proposal to amend 
existing [(NUREG-1433 and NUREG-1434) TS sections SR 3.1.3.2, ``Control 
Rod OPERABILTY,'' (NUREG-1434) LCO 3.3.1.2 Required Action E.2, 
``Source Range Monitor (SRM) Instrumentation,'' and] (NUREG-1430 
through NUREG-1434) Example 1.4-3, ``Frequency'' applicable to SR 
3.0.2. The NRC staff has concluded that the TS revisions [will have a 
minimal affect on the high reliability of the CRD system while reducing 
the opportunity for potential reactivity events; thus, meeting the 
requirement of CFR, Part 50, Appendix A, GDC 29, and] will clarify the 
1.25 provision in SR 3.0.2. Therefore, the staff concludes that the 
amendment request is acceptable.
    Based on the considerations discussed above, the Commission has 
concluded 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.

4.0 State Consultation

    In accordance with the Commission's regulations, the [ ] State 
official was notified of the proposed issuance of the amendment. The 
State official had [(1) no comments or (2) the following comments--with 
subsequent disposition by the staff].

5.0 Environmental Consideration

    The amendments change a requirement with respect to the 
installation or use of a facility component located within the 
restricted area as defined in 10 CFR part 20 and change surveillance 
requirements. The NRC staff has determined that the amendments involve 
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 amendments involve no significant hazards 
considerations, and there has been no public comment on the finding [FR 
]. Accordingly, the amendments meet the eligibility criteria for 
categorical exclusion set forth in 10 CFR 51.22(c)(9) [and (c)(10)]. 
Pursuant to 10 CFR 51.22(b), no environmental impact statement or 
environmental assessment need be prepared in connection with the 
issuance of the amendments.

6.0 Conclusion

    The Commission has concluded, on the basis of 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.

7.0 References

    1. Letter TSTF-04-07 from the Technical Specifications Task Force 
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-1430, ``Standard Technical Specifications Babcock and 
Wilcox Plants, Revision 3,'' August 31, 2003.
    3. NUREG-1431, ``Standard Technical Specifications Westinghouse 
Plants, Revision 3,'' August 31, 2003.
    4. NUREG-1432, ``Standard Technical Specifications Combustion 
Engineering Plants, Revision 3,'' August 31, 2003.
    5. NUREG-1433, ``Standard Technical Specifications General Electric 
Plants, BWR/4, Revision 3,'' August 31, 2003.
    6. NUREG-1434, ``Standard Technical Specifications General Electric 
Plants, BWR/6, Revision 3,'' August 31, 2003.
    7. Letter TSTF-07-19, Response from the Technical Specifications 
Task Force 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.
    8. Letter TSTF-06-13 from the Technical Specifications Task Force 
to the NRC, ``Response to NRC Request for Additional Information 
Regarding TSTF-475, Revision 0,'' dated July 3, 2006, ADAMS accession 
number ML0618403421.
    9. Letter BWROG-06036 from the BWR Owners Group 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.
    10. Letter TSTF-07-19 from the Technical Specifications Task Force 
to the NRC, ``Response to NRC Request for Additional Information 
Regarding TSTF-475, Revision 0,'' dated May 22, 2007, ADAMS accession 
number ML071420428].
    THE FOLLOWING EXAMPLE OF AN APPLICATION WAS PREPARED BY THE NRC 
STAFF TO FACILITATE USE OF THE CONSOLIDATED LINE ITEM IMPROVEMENT 
PROCESS (CLIIP). THE MODEL PROVIDES THE EXPECTED LEVEL OF DETAIL AND 
CONTENT FOR AN APPLICATION TO REVISE TECHNICAL SPECIFICATIONS 
REGARDING REVISION OF CONTROL ROD NOTCH SURVEILLANCE TEST FREQUENCY, 
CLARIFICATION OF SRM INSERT CONTROL ROD ACTION, AND A CLARIFICATION 
OF A FREQUENCY EXAMPLE. LICENSEES REMAIN RESPONSIBLE FOR ENSURING 
THAT THEIR ACTUAL APPLICATION FULFILLS THEIR ADMINISTRATIVE 
REQUIREMENTS AS WELL AS NUCLEAR REGULATORY COMMISSION REGULATIONS.

U.S. Nuclear Regular Commission
Document Control Desk
Washington, DC 20555

SUBJECT: PLANT NAME, DOCKET NO. 50--APPLICATION FOR TECHNICAL 
SPECIFICATION CHANGE REGARDING REVISION OF CONTROL ROD NOTCH 
SURVEILLANCE TEST FREQUENCY, CLARIFICATION OF SRM INSERT CONTROL ROD 
ACTION, AND A

[[Page 63941]]

CLARIFICATION OF A FREQUENCY EXAMPLE USING THE CONSOLIDATED LINE 
ITEM IMPROVEMENT PROCESS

Gentleman:

    In accordance with the provisions of 10 CFR 50.90 [LICENSEE] is 
submitting a request for an amendment to the technical 
specifications (TS) for [PLANT NAME, UNIT NOS.].
    The proposed amendment would: (1) [revise the TS surveillance 
requirement (SR) frequency in TS 3.1.3, ``Control Rod OPERABILITY'', 
(2) clarify the requirement to fully insert all insertable control 
rods for the limiting condition for operation (LCO) in TS 3.3.1.2, 
required Action E.2, ``Source Range Monitoring Instrumentation,'' 
and (3)] revise Example 1.4-3 in Section 1.4 ``Frequency'' to 
clarify the applicability of the 1.25 surveillance test interval 
extension.
    Attachment 1 provides a description of the proposed change, the 
requested confirmation of applicability, and plant-specific 
verifications. Attachment 2 provides the existing TS pages marked up 
to show the proposed change. Attachment 3 provides revised (clean) 
TS pages. Attachment 4 provides a summary of the regulatory 
commitments made in this submittal.
    [LICENSEE] requests approval of the proposed License Amendment 
by [DATE], with the amendment being implemented [BY DATE OR WITHIN X 
DAYS].
    In accordance with 10 CFR 50.91, a copy of this application, 
with attachments, is being provided to the designated [STATE] 
Official.
    I declare [or certify, verify, state] under penalty of perjury 
that the foregoing is true and correct.
    If you should have any questions regarding this submittal, 
please contact [NAME, TELEPHONE NUMBER].

Sincerely,

[Name, Title]

Attachments:
1. Description and Assessment
2. Proposed Technical Specification Changes
3. Revised Technical Specification Pages
4. Regulatory Commitments
5. Proposed Technical Specification Bases Changes]
cc:
NRC Project Manager
NRC Regional Office
NRC Resident Inspector
State Contact

Attachment 1--Description and Assessment

1.0 Description

    The proposed amendment would: (1) [Revise the TS surveillance 
requirement (SR 3.1.3.2) frequency in TS 3.1.3, ``Control Rod 
OPERABILITY'', (2) clarify the requirement to fully insert all 
insertable control rods for the limiting condition for operation (LCO) 
in TS 3.3.1.2, Required Action E.2, ``Source Range Monitoring 
Instrumentation'', and (3)] revise Example 1.4-3 in Section 1.4 
``Frequency'' to clarify the applicability of the 1.25 surveillance 
test interval extension.
    The changes are consistent with Nuclear Regulatory Commission (NRC) 
approved Industry/Technical Specification Task Force (TSTF) STS change 
TSTF-475, Revision 1. The Federal Register notice published on [DATE] 
announced the availability of this TS improvement through the 
consolidated line item improvement process (CLIIP).

2.0 Assessment

2.1 Applicability of Published Safety Evaluation

    [LICENSEE] has reviewed the safety evaluation dated [DATE] as part 
of the CLIIP. This review included a review of the NRC staff's 
evaluation, as well as the supporting information provided to support 
TSTF-475, Revision 1. [LICENSEE] has concluded that the justifications 
presented in the TSTF proposal and the safety evaluation prepared by 
the NRC staff are applicable to [PLANT, UNIT NOS.] and justify this 
amendment for the incorporation of the changes to the [PLANT] TS.

2.2 Optional Changes and Variations

    [LICENSEE] is not proposing any variations or deviations from the 
TS changes described in the modified TSTF-475, Revision 1 and the NRC 
staff's model safety evaluation dated [DATE].

3.0 Regulatory Analysis

3.1 No Significant Hazards Consideration Determination

    [LICENSEE] has reviewed the proposed no significant hazards 
consideration determination (NSHCD) published in the Federal Register 
as part of the CLIIP. [LICENSEE] has concluded that the proposed NSHCD 
presented in the Federal Register notice is applicable to [PLANT] and 
is hereby incorporated by reference to satisfy the requirements of 10 
CFR 50.91(a).

3.2 Verification and Commitments

    As discussed in the notice of availability published in the Federal 
Register on [DATE] for this TS improvement, the [LICENSEE] verifies the 
applicability of TSTF-475 to [PLANT], and commits to establishing 
Technical Specification Bases for TS as proposed in TSTF-475, Revision 
1.
    These changes are based on TSTF change traveler TSTF-475 (Revision 
1) that proposes revisions to the STS by: (1) [Revising the frequency 
of SR 3.1.3.2, notch testing of fully withdrawn control rod, from ``7 
days after the control rod is withdrawn and THERMAL POWER is greater 
than the LPSP of RWM'' to ``31 days after the control rod is withdrawn 
and THERMAL POWER is greater than the LPSP of the RWM'', (2) adding the 
word ``fully'' to LCO 3.3.1.2 Required Action E.2 to clarify the 
requirement to fully insert all insertable control rods in core cells 
containing one or more fuel assemblies when the associated SRM 
instrument is inoperable, and (3)] 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.

4.0 Environmental Evaluation

    [LICENSEE] has reviewed the environmental evaluation included in 
the model safety evaluation dated [DATE] as part of the CLIIP. 
[LICENSEE] has concluded that the staff's findings presented in that 
evaluation are applicable to [PLANT] and the evaluation is hereby 
incorporated by reference for this application.

ATTACHMENT 2--PROPOSED TECHNICAL SPECIFICATION CHANGES (MARK-UP)

ATTACHMENT 3--PROPOSED TECHNICAL SPECIFICATION PAGES

ATTACHMENT 4--LIST OF REGULATORY COMMITMENTS

    The following table identifies those actions committed to by 
[LICENSEE] in this document. Any other statements in this submittal are 
provided for information purposes and are not considered to be 
regulatory commitments. Please direct questions regarding these 
commitments to [CONTACT NAME].

------------------------------------------------------------------------
           Regulatory commitments                   Due date/event
------------------------------------------------------------------------
[[LICENSEE] will establish the Technical     [Complete, implemented with
 Specification Bases for [TS B 3.1.3, TS B    amendment OR within X days
 3.1.4, and TS B 3.3.1.2] consistent with     of implementation of
 those shown in TSTF-475, Revision 1,         amendment].
 ``Control Rod Notch Testing Frequency and
 SRM Insert Control Rod Action.''].
------------------------------------------------------------------------


[[Page 63942]]

ATTACHMENT 5--PROPOSED CHANGES TO TECHNICAL SPECIFICATION BASES PAGES

    [Not required for plants only adopting portion of TSTF-475 change 
pertaining to TS Section 1.4 that provides example to SR Frequency]

Proposed No Significant Hazards Consideration Determination

    Description of Amendment Request: [Plant Name] requests adoption of 
an approved change to the Standard Technical Specifications (STS) for 
[General Electric (GE) Plants (NUREG-1433, BWR/4 and NUREG-1434, BWR/6) 
and] plant specific technical specifications (TS), that allows: (1) 
[revising the frequency of SR 3.1.3.2, notch testing of fully withdrawn 
control rod, from ``7 days after the control rod is withdrawn and 
THERMAL POWER is greater than the LPSP of RWM'' to ``31 days after the 
control rod is withdrawn and THERMAL POWER is greater than the LPSP of 
the RWM'', (2) adding the word ``fully'' to LCO 3.3.1.2 Required Action 
E.2 to clarify the requirement to fully insert all insertable control 
rods in core cells containing one or more fuel assemblies when the 
associated SRM instrument is inoperable, and (3)] 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. The staff finds that the 
proposed STS changes are acceptable [because the number of control rod 
manipulations is reduced thereby reducing the opportunity for potential 
reactivity events while having a very minimal impact on the extremely 
high reliability of the CRD system as discussed in the technical 
evaluation section of this safety evaluation and] the discussion of the 
SR Frequency example provides clarification.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident 
Previously Evaluated

    The proposed change generically implements TSTF-475, Revision 1, 
``Control Rod Notch Testing Frequency and SRM Insert Control Rod 
Action.'' TSTF-475, Revision 1 modifies NUREG-1433 (BWR/4) and 
NUREG-1434 (BWR/6) STS. The changes: (1) revise TS testing frequency 
for surveillance requirement (SR) 3.1.3.2 in TS 3.1.3, ``Control Rod 
OPERABILITY'', (2) clarify the requirement to fully insert all 
insertable control rods for the limiting condition for operation 
(LCO) in TS 3.3.1.2, Required Action E.2, ``Source Range Monitoring 
Instrumentation'' (NUREG-1434 only), and (3) revise Example 1.4-3 in 
Section 1.4 ``Frequency'' to clarify the applicability of the 1.25 
surveillance test interval extension. The consequences of an 
accident after adopting TSTF-475, Revision 1 are no different than 
the consequences of an accident prior to adoption. Therefore, this 
change does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.

Criterion 2--The Proposed Change Does Not Create the Possibility of 
a New or Different Kind of Accident from any Accident Previously 
Evaluated

    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or a change in the methods governing normal plant operation. The 
proposed change will not introduce new failure modes or effects and 
will not, in the absence of other unrelated failures, lead to an 
accident whose consequences exceed the consequences of accidents 
previously analyzed. Thus, this change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.

Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety

    TSTF-475, Revision 1 will: (1) [revise the TS SR 3.1.3.2 
frequency in TS 3.1.3, ``Control Rod OPERABILITY'', (2) clarify the 
requirement to fully insert all insertable control rods for the 
limiting condition for operation (LCO) in TS 3.3.1.2, ``Source Range 
Monitoring Instrumentation,'' and (3)] revise Example 1.4-3 in 
Section 1.4 ``Frequency'' to clarify the applicability of the 1.25 
surveillance test interval extension. [The GE Nuclear Energy Report, 
``CRD Notching Surveillance Testing for Limerick Generating 
Station,'' dated November 2006, concludes that extending the control 
rod notch test interval from weekly to monthly is not expected to 
impact the reliability of the scram system and that the analysis 
supports the decision to change the surveillance frequency.] 
Therefore, the proposed changes in TSTF-475, Revision 1 are 
acceptable and do not involve a significant reduction in a margin of 
safety.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.

    Dated at Rockville, Maryland, this 5th day of November, 2007.

    For the Nuclear Regulatory Commission.
Timothy J. Kobetz,
Section Chief, Technical Specifications Branch, Division of Inspection 
& Regional Support, Office of Nuclear Reactor Regulation.
[FR Doc. E7-22159 Filed 11-9-07; 8:45 am]
BILLING CODE 7590-01-P