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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

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Start Preamble

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.

Start Further Info

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.

End Further Info End Preamble Start Supplemental Information

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 Start Printed Page 63936adoption 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 Start Printed Page 63937with 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.

Start Signature

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.

End Signature

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 Start Printed Page 63938Specifications 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 Start Printed Page 63939collet, 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 Start Printed Page 63940more 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 Start Printed Page 63941CLARIFICATION 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 commitmentsDue 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].
Start Printed 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.

Start Signature

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.

End Signature End Supplemental Information

[FR Doc. E7-22159 Filed 11-9-07; 8:45 am]

BILLING CODE 7590-01-P