Federal Register: October 18, 2000 (Volume 65, Number 202)

DOCID: FR Doc 00-26645

SUBJECT CATEGORY:

Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations

DOCUMENT SUMMARY:

I. Background

Pursuant to Public Law 97415, the U.S. Nuclear Regulatory Commission (the Commission or NRC staff) is publishing this regular biweekly notice. Public Law 97415 revised section 189 of the Atomic Energy Act of 1954, as amended (the Act), to require the Commission to publish notice of any amendments issued, or proposed to be issued, under a new provision of section 189 of the Act. This provision grants the Commission the authority to issue and make immediately effective any amendment to an operating license upon a determination by the Commission that such amendment involves no significant hazards consideration, notwithstanding the pendency before the Commission of a request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or proposed to be issued from September 25, 2000, through October 6, 2000. The last biweekly notice was published on October 4, 2000.
Notice of Consideration of Issuance of Amendments to Facility Operating Licenses, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing

The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commission's regulations in 10 CFR 50.92, this means that operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. The basis for this proposed determination for each amendment request is shown below.

The Commission is seeking public comments on this proposed determination. Any comments received within 30 days after the date of publication of this notice will be considered in making any final determination.

Normally, the Commission will not issue the amendment until the expiration of the 30day notice period. However, should circumstances change during the notice period such that failure to act in a timely way would result, for example, in derating or shutdown of the facility, the Commission may issue the license amendment before the expiration of the 30day notice period, provided that its final determination is that the amendment involves no significant hazards consideration. The final determination will consider all public and State comments received before action is taken. Should the Commission take this action, it will publish in the Federal Register a notice of issuance and provide for opportunity for a hearing after issuance. The Commission expects that the need to take this action will occur very infrequently.

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Written comments may be submitted by mail to the Chief, Rules Review and Directives Branch, Division of Freedom of Information and Publications Services, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 205550001, and should cite the publication date and page number of this Federal Register notice. Written comments may also be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of written comments received may be examined at the NRC's Public Document Room, the Gelman Building, 2120 L Street, NW, Washington, DC through September 22, 2000. The NRC is relocating its Public Document Room to the NRC's headquarters building. Effective September 26, 2000, documents may be examined at the NRC's Public Document Room, located at One White Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. The filing of requests for a hearing and petitions for leave to intervene is discussed below.

By November 17, 2000, the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written request for a hearing and a petition for leave to intervene. Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the Commission's ``Rules of Practice for Domestic Licensing Proceedings'' in 10 CFR part 2. Interested persons should consult a current copy of 10 CFR 2.714 which is available at the Commission's Public Document Room, located at One White Flint North, 11555 Rockville Pike (first Floor), Rockville, Maryland 20852. Publicly available records will be accessible and electronically from the ADAMS Public Library component on the NRC Web site, http://www.nrc.gov (the Electronic Reading Room). If a request for a hearing or petition for leave to intervene is filed by the above date, the Commission or an Atomic Safety and Licensing Board, designated by the Commission or by the Chairman of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the designated Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order.

As required by 10 CFR 2.714, a petition for leave to intervene shall set forth with particularity the interest of the petitioner in the proceeding, and how that interest may be affected by the results of the proceeding. The petition should specifically explain the reasons why intervention should be permitted with particular reference to the following factors: (1) The nature of the petitioner's right under the Act to be made a party to the proceeding; (2) the nature and extent of the petitioner's property, financial, or other interest in the proceeding; and (3) the possible effect of any order which may be entered in the proceeding on the petitioner's interest. The petition should also identify the specific aspect(s) of the subject matter of the proceeding as to which petitioner wishes to intervene. Any person who has filed a petition for leave to intervene or who has been admitted as a party may amend the petition without requesting leave of the Board up to 15 days prior to the first prehearing conference scheduled in the proceeding, but such an amended petition must satisfy the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference scheduled in the proceeding, a petitioner shall file a supplement to the petition to intervene which must include a list of the contentions which are sought to be litigated in the matter. Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the petitioner shall provide a brief explanation of the bases of the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the petitioner intends to rely in proving the contention at the hearing. The petitioner must also provide references to those specific sources and documents of which the petitioner is aware and on which the petitioner intends to rely to establish those facts or expert opinion. Petitioner must provide sufficient information to show that a genuine dispute exists with the applicant on a material issue of law or fact. Contentions shall be limited to matters within the scope of the amendment under consideration. The contention must be one which, if proven, would entitle the petitioner to relief. A petitioner who fails to file such a supplement which satisfies these requirements with respect to at least one contention will not be permitted to participate as a party.

Those permitted to intervene become parties to the proceeding, subject to any limitations in the order granting leave to intervene, and have the opportunity to participate fully in the conduct of the hearing, including the opportunity to present evidence and cross examine witnesses.

If a hearing is requested, the Commission will make a final determination on the issue of no significant hazards consideration. The final determination will serve to decide when the hearing is held.

If the final determination is that the amendment request involves no significant hazards consideration, the Commission may issue the amendment and make it immediately effective, notwithstanding the request for a hearing. Any hearing held would take place after issuance of the amendment.

If the final determination is that the amendment request involves a significant hazards consideration, any hearing held would take place before the issuance of any amendment.

A request for a hearing or a petition for leave to intervene must be filed with the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 205550001, Attention: Docketing and Services Branch, or may be delivered to the Commission's Public Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, by the above date. A copy of the petition should also be sent to the Office of the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 205550001, and to the attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended petitions, supplemental petitions and/or requests for a hearing will not be entertained absent a determination by the Commission, the presiding officer or the Atomic Safety and Licensing Board that the petition and/or request should be granted based upon a balancing of factors specified in 10 CFR 2.714(a)(1)(i)(v) and 2.714(d).

For further details with respect to this action, see the application for amendment which is available for public inspection at the Commission's Public Document Room, located at One White Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. Publicly available records will be accessible and electronically from the ADAMS Public Library component on the NRC Web site, http:// www.nrc.gov (the Electronic Reading Room).
AmerGen Energy Company, LLC, Docket No. 50289, Three Mile Island Nuclear Station, Unit 1, Dauphin County, Pennsylvania

Date of amendment request: August 9, 2000.

Description of amendment request: The proposed amendment revises Sections 6.5.3 and 6.5.4 of the Technical Specifications to eliminate reference to the Independent Onsite Safety Review
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Group (IOSRG) and to redefine the performance of the IOSRG function by the nuclear quality assurance organization.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:

1. Operation of the facility in accordance with the proposed amendment would not involve a significant increase in the
probability of occurrence or the consequences of an accident previously evaluated. The proposed changes do not affect assumptions contained in plant safety analyses, the physical design and/or operation of the plant, nor do they affect Technical Specifications that preserve safety analysis assumptions. None of the proposed changes involve a physical modification to the plant, a new mode of operation or a change to the UFSAR [Updated Final Safety Analysis Report] transient analyses. No Technical Specification Limiting Condition for Operation, Action Statement, or Surveillance Requirement is affected by any of the proposed changes. The proposed changes do not alter the design, function, or operation of any plant component. Therefore, the proposed amendment does not affect the probability of occurrence or consequences of an accident previously evaluated.

2. Operation of the facility in accordance with the proposed amendment would not create the possibility of a new or different kind of accident from any previously evaluated. The proposed changes do not affect assumptions contained in the plant safety analyses, the physical design and/or modes of plant operation defined in the plant operating license, or Technical Specifications that preserve safety analysis assumptions. The proposed changes do not introduce a new mode of plant operation or surveillance requirement, nor involve a physical modification to the plant. The proposed changes do not alter the design, function, or operation of any plant components. Therefore, the proposed amendment does not affect the possibility of a new or different kind of accident from any accident previously evaluated.

3. Operation of the facility in accordance with the proposed amendment would not involve a significant reduction in a margin of safety. None of the proposed changes involve a physical modification to the plant, a new mode of operation or a change to the UFSAR transient analyses. No Technical Specification Limiting Condition for Operation, Action Statement, or Surveillance Requirement is affected. Therefore, the proposed amendment does not reduce the margin of safety.

Based upon the analysis provided herein [the licensee's August 9, 2000 application], the proposed changes will not increase the probability or consequences of an accident previously evaluated, create the possibility of a new or different kind of accident from any accident previously evaluated, or involve a reduction in a margin of safety. The performance of safety assessment and the IOSRG functions by a single qualified organization will lead to
efficiencies in the performance of both functions. The training and qualification of the personnel performing the IOSRG functions will be unchanged from the current requirements. Therefore, the proposed changes meet the requirements of 10 CFR 50.92(c) and involve no significant hazards consideration.

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: Edward J. Cullen, Jr., Esq., PECO Energy Company, 2301 Market Street, S231, Philadelphia, PA 19103.

NRC Section Chief: Marsha Gamberoni.
AmerGen Energy Company, LLC, Docket No. 50289, Three Mile Island Nuclear Station, Unit 1, Dauphin County, Pennsylvania

Date of amendment request: August 9, 2000.

Description of amendment request: The proposed amendment revises the Three Mile Island Nuclear Station, Unit 1 (TMI1), Updated Final Safety Analysis Report (UFSAR), Section 14.1.2.10, ``Steam Generator Tube Failure Analysis,'' to include the dose resulting from the postulated postaccident steam release through the main steam safety valves. The revised dose for the TMI1 steam generator tube failure analysis would be increased above the values previously reviewed and approved by the NRC, but would continue to be below the limits in Title 10 of the Code of Federal Regulations (10 CFR) Part 100. The proposed change to the UFSAR modifies the existing analysis to account for release of radioactivity to the atmosphere for the postulated tube rupture analysis. The existing dose calculations do not account for this release. Editorial and grammatical corrections are also made.

1. Operation of the facility in accordance with the proposed amendment would not involve a significant increase in the
probability of occurrence or the consequences of an accident previously evaluated. This change has no effect on structures, systems or components prior to the postulated steam generator tube failure accident or any other accident. The proposed change corrects the existing UFSAR Steam Generator Tube Failure accident analysis to account for the release to atmosphere through the main steam safety valves (MSSVs). The resulting revised radiological consequences for the postulated Steam Generator Tube Failure accident remain well below the 10 CFR 100 limits.

2. Operation of the facility in accordance with the proposed amendment would not create the possibility of a new or different kind of accident from any previously evaluated. This change has no impact on any plant structures systems or components. The only impact is the revised radiological consequences of the Steam Generator Tube Failure accident analysis to account for the release to atmosphere through the MSSVs. This change only corrects the existing TMI Unit 1 UFSAR.

3. Operation of the facility in accordance with the proposed amendment would not involve a significant reduction in a margin of safety. No change to any plant structure, system or component is being made or proposed by this change. This change does not involve any change to safety system setpoints for operation. The revised radiological consequences of the Steam Generator Tube Failure accident analysis remain well below 10 CFR 100 limits.

Attorney for licensee: Edward J. Cullen, Jr., Esq., PECO Energy Company, 2301 Market Street, S231, Philadelphia, PA 19103.

NRC Section Chief: Marsha Gamberoni.
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50317 and 50 318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, Maryland

Date of amendments request: September 14, 2000.

Description of amendments request: The proposed amendment revises the Unit 1 and Unit 2 heatup curves (Technical Specification Figures 3.4.31) and Unit 1 and Unit 2 cooldown curves (Technical Specification Figures 3.4.32) to increase the allowable heatup and cooldown rates. Use of stress intensity factor KIC, permitted by American Society of Mechanical Engineers (ASME) Code Case N640,
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made it possible to increase the heatup and cooldown rates without changing existing pressuretemperature (PT) limits. The existing PT limits were approved previously. Application of Code Case N640 to generate PT curves is not currently permitted by the regulations. Therefore, pursuant to 10 CFR 50.12, a separate request for an exemption to use Code Case N640 was submitted in a letter dated September 14, 2000.

1. Would not involve a significant increase in the probability or consequences of an accident previously evaluated.

In accordance with 10 CFR Part 50, Appendix G, the Calvert Cliffs pressure/temperature (PT) limits for material fracture toughness requirements of the reactor coolant pressure boundary materials were developed using the methods of linear elastic fracture mechanics and the guidance found in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, Appendix G. The Calvert Cliffs PT limits are based on fluence level. The fluence levels are determined in the same manner as the pressurized thermal shock (PTS) screening criteria defined in 10 CFR 50.61 for the critical elements. Methods described in the Nuclear Regulatory Commission Regulatory Guide 1.99, Revision 2, are used to predict the embrittlement effect of neutron irradiation on reactor vessel materials. Regulatory Guide 1.99 defines embrittlement effect in terms of adjusted reference temperatures, which depends on the material property of the PTS critical elements.

The proposed higher heatup and cooldown rates for the Technical Specification PT limits were made possible by the ASME Code Case N 640 which permits use of reference stress intensity factor KIC, in place of KIA. Use of KIC, for the maximum stress intensity factor that will not lead to failure, is the correct value to use. Although conservative in terms of developing PT limits, use of KIA results in a very restrictive heatup and cooldown rate that challenges plant safety. To bound the existing LTOP [lowtemperature overpressure protection] enable temperatures, while increasing the heatup and cooldown rates, the criteria described in ASME Section XI Code Case 514 is used. Code Case 514 is listed in Regulatory Guide 1.147 as acceptable to the Nuclear Regulatory Commission (NRC) for this application. With the new higher heatup and cooldown rates, the underlying intent of the 10 CFR Part 50, Appendix G, requirement for adequate margin to prevent brittle failure of the reactor coolant pressure boundary materials is maintained. Additionally, since the cooldown rates are not changed above 300 deg. F, the safety analyses and dose consequences in the Updated Final Safety Analysis Report are not affected.

Therefore the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Would not create the possibility of a new or different [kind] of accident previously evaluated.

The implementation of the proposed revision has no significant effect on either the configuration of the plant, or the manner in which it is operated.

Therefore, this proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Would not involve a significant reduction in a margin of safety.

As discussed above, although conservative in terms of developing PT limits, use of KIA results in a very restrictive heatup and cooldown rate that challenges plant safety. The insignificant margin reduction in PT limits is more than
compensated by the safety benefits that are realized in terms of plant component integrity as a result of the higher heatup and cooldown rates. With the proposed change, the underlying intent of the 10 CFR Part 50, Appendix G, requirement for adequate margin to prevent brittle failure of the reactor coolant pressure boundary materials is maintained, and there is a net gain in overall plant safety margin.

Therefore, this proposed change does not significantly reduce the margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendments request involves no significant hazards consideration.

Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.

NRC Section Chief: Marsha Gamberoni.
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50317 and 50 318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, Maryland

Date of amendments request: September 14, 2000.

Description of amendments request: The proposed amendment adds two analytical methods to the list of approved core operating limits analytical methods in the Technical Specifications (TSs) for Calvert Cliffs, Unit Nos. 1 & 2. In a letter dated March 16, 2000, from Mr. S. A. Richards, NRC to Mr. I. C. Rickard, ABB Combustion Engineering, the Nuclear Regulatory Commission approved the Topical Report CENPD387P, ``ABB Critical Heat Flux Correlations for [pressurizedwater reactor] PWR Fuel'' for referencing in licensing applications for Asea Brown Boveri, Inc. Combustion Engineering, Inc. (ABBCE) plants.

1. Would not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed change allows the use of the ABBNV and ABBTV CHF [critical heat flux] correlations in the thermal hydraulic analysis for Calvert Cliffs Nuclear Power Plant. The ABBNV is used for a nonmixing vane fuel assembly and the ABBTV correlations are used for Turbo mixing vane fuel assembly. The CHF correlations determine the departure from nucleate boiling ratio (DNBR). The specified acceptable fuel design limit for DNBR will change for ABBNV and ABBTV. The use of the ABBNV and/or ABBTV correlations with the appropriate DNBR limit provides additional operating margin for those analyses that presently use the CE1 correlation.

The use of a different CHF correlation will not increase the probability of an accident because the plant systems will not be operated outside of design limits, the plant equipment will not be operated in a different manner, and system interfaces will not change.

As Turbo fuel is introduced to reactor, transistion cores will exist in which Turbo mixing vane grid fuel assemblies are co residents with nonmixing vane grid fuel assemblies. The grid hydraulic loss coefficient in the Turbo grids is greater than the grid hydraulic loss coefficient for the nonmixing grids. The flow diversion that will result does not increase the probability of an accident previously evaluated because assembly flow has no impact on accident initiators, and because plant systems will not be operated outside of design limits, plant equipment will not be operated in a different manner, and system interfaces will not change.

The change in the CHF correlation was the subject of Topical Report CENPD387PA, which was reviewed and approved by the NRC. The use of a different CHF correlation will not increase the consequences of an accident because Limiting Conditions [for] Operation (LOCs) will continue to restrict operation to within the regions that provide acceptable results, and Reactor Protection System (RPS) trip setpoints will plant transients so that the consequences of accidents will be acceptable.

The transition cores that will exist as Turbo fuel is introduced to the reactor will not increase the consequences of an accident. The TORC code accurately predicts the flow conditions in adjacent fuel bundles that contain grids with different designs and coefficients. The flow diversion will be compensated for by DNBR margin gains. Operation within the LOCs and RPS setpoints will continue to restrict plant
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transients so that consequences of accidents will be acceptable.

Therefore, the proposed TS changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Would not create the possibility of a new or different [kind] of accident from any accident previously evaluated.

The proposed change does not add any new equipment, modify any interfaces with any existing equipment, alter the equipment's function or change the method of operating the equipment. The proposed change does not alter plant conditions in a manner that could affect other plant components. The proposed change does not cause any existing equipment to become an accident initiator. The Turbo grid design does not introduce features that could initiate an accident.

Therefore the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Would not involve a significant reduction in the margin of safety.

Safety Limits ensure that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences. One of the safety limits that accomplishes this is the DNBR limit. The CHF correlations that have been approved for ABBNV and ABBTV result in a DNBR limit that provides a 95% probability, at a 95% confidence, that the hot fuel rod in the core will not experience departure from nucleate boiling. The RPS in combination with the LCOs, will continue to prevent any anticipated combination of transient conditions for reactor coolant system temperature, pressure and thermal power level that would result in a violation of the Safety Limits.

Therefore the margin of safety is not significantly reduced by this proposed change.

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendments request involves no significant hazards consideration.

Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.

NRC Section Chief: Marsha Gamberoni.
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50317 and 50 318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, Maryland

Date of amendments request: September 14, 2000.

Description of amendments request: Calvert Cliffs Nuclear Power Plant, Inc. (CCNPPI) proposed an amendment to incorporate changes described below into the Technical Specifications (TSs) for Calvert Cliffs Units 1 and 2.

On September 9, 1996, a final rule amending 10 CFR 50.55a was issued requiring owners to implement, by September 9, 2001, the requirements of the 1992 Edition through the 1992 Addenda of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) Section XI, Subsections IWE and IWL, as modified and supplemented by 10 CFR 50.55a. CCNPPI has developed a program to effect the implementation of Subsections IWE and IWL. This submittal requests a license amendment in support of the program.

The TSs change replaces the reference to Regulatory Guide (RG) 1.35 with a reference to Section XI of the ASME Code, and deletes the applicability of Surveillance Requirement 3.0.2. Compliance with RG 1.35 is not sufficient to comply with 10 CFR 50.55a, as amended, and inspection frequencies will be in accordance with Subsection IWL of Section XI; therefore, Surveillance Requirement 3.0.2 will no longer apply.

1. Would not involve a significant increase in the probability or consequences of an accident previously evaluated.

The Containment Building is a passive safety structure that prevents the release of radioactive materials to the environment in postaccident conditions. The proposed Technical Specification change updates requirements of the Technical Specifications that have been made obsolete by the improvements of the Containment [B]uilding inspections required by the changes in the regulations. The improved inspections required by the American Society of Mechanical Engineers [Boiler and Pressure Vessel] Code serve to maintain containment response to accident conditions, by causing the identification and repair of defects in Containment Buildings.

Relocating existing requirements, eliminating requirements that duplicate regulations, and making administrative improvements provide Technical Specifications that are easier to use. Because existing requirements are controlled by regulation, there is no reduction in commitment and adequate control is still maintained. Therefore, the proposed change would not involve a significant increase in probability or consequences of an accident previously evaluated.

2. Would not create the possibility of a new or different [kind] of accident from any accident previously evaluated.

The Containment Building is a passive safety structure designed to contain radioactive materials released from the reactor coolant system. The performance of the Containment Building is not evaluated as the causal factor in any accident at Calvert Cliffs Nuclear Power Plant. The proposed Technical Specification change updates requirements of the Technical Specifications that were made obsolete by the improvements of the Containment [B]uilding inspections required by the changes in the regulations. Revising the Technical Specifications, to comply with current regulations and to eliminate duplication of requirements, does not create the possibility of a new or different [kind] of accident from any previously evaluated.

3. Would not involve a significant reduction in a margin of safety.

The safety function of the Containment Building is to provide a boundary to the release of radioactive material to the environment during postaccident conditions. The change to the Technical Specifications incorporate[s] improved inspection techniques and criteria to ensure optimum containment integrity and, therefore, optimum containment response in the event of an accident resulting in a release of radioactive material from the reactor coolant system. Optimizing containment integrity will result in maintaining the margin of safety allowed by the Containment Buildings. Therefore, the proposed change will not involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendments request involves no significant hazards consideration.

Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.

NRC Section Chief: Marsha Gamberoni.
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50317 and 50 318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, Maryland

Date of amendments request: September 15, 2000.

Description of amendments request: The proposed amendment revises the Unit 1 and Unit 2 Technical Specification Surveillance Requirement (SR) 3.1.7.2 which verifies that each control element assembly (CEA) not fully inserted is capable of full insertion when tripped from at least the 50 percent withdrawn position. Specifically, the proposed amendment adds a note to SR 3.1.7.2, which allows the SR to not be performed during initial power escalation following a refueling outage if SR 3.1.4.6 (CEA drop time test) has been met. In addition, ``once'' was added to the SR frequency
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as an administrative change to clarify that the SR is only performed once and not on a periodic basis. This proposed license amendment is consistent with Technical Specification Task Force (TSTF)134, Revision 1, which received Nuclear Regulatory Commission (NRC) approval on April 21, 1998.

1. Would not involve a significant increase in the probability or consequences of an accident previously evaluated.

A risk assessment was performed to support a prior license amendment request submitted to change Surveillance Requirement (SR) 3.1.7.2 frequency from 24 hours to 7 days. Results of a study performed in support of the risk assessment indicated no change in the geometry of those components utilized in control element assembly (CEA) insertion over the 7day period. The study also evaluated electronic/electrical failures that could cause a CEA to be stuck, concluding that the feature that controls the movement of the CEAs is not timerelated. Since there have been no modifications performed on the components analyzed or changes in the manner in which they are operated, it is reasonable to assume that the conclusions remain valid.

The CEA drop time test SR 3.1.4.6 proves that any work done during the refueling outage does not prevent the rods from tripping. Revising SR 3.1.7.2, such that it could allow more than seven days from successfully performing the CEA drop time test does not change this. However, as with any component, there will eventually be some timerelated degradation that may impact the ability of the CEAs to drop. Thus, when the seven days are exceeded, there is some negligible increase in the probability that a rod would fail to drop. This causes an insignificant increase in core damage frequency because it requires multiple rod failures to cause core damage in the event of an overcooling event (the most bounding accident for a stuck CEA during rod worth testing). This additional risk is believed to be small since the degradation is the result of core changes, which occur slowly, and not the result of maintenance. Thus the risk increase due to this Technical Specification change is considered to be negligible. The probability of an overcooling event is not changed by the proposed change.

Therefore the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Would not create the possibility of a new or different [kind] of accident from any accident previously evaluated.

The proposed change to the surveillance requirement for CEA trippability does not result in any change to the facility or the manner in which it is operated.

Therefore, this proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Would not involve a significant reduction in a margin of safety.

Operation of the facility in accordance with this proposed amendment does not involve a significant reduction in a margin of safety. Control element assembly trippability is still demonstrated via performance of SR 3.1.4.6. The risk increase due to this change is considered to be negligible. Thus, appropriate equipment continues to be tested in a manner and at a frequency necessary to provide reasonable assurance that the equipment can perform its assumed safety function.

Furthermore, this change is consistent with Technical Specification Task Force (TSTF)134, Revision 1, which has been approved by the Nuclear Regulatory Commission. Adopting testing practices consistent with those specified in TSTF134, Revision 1 are acceptable based on similar design, likecomponent testing for the system application and the availability of other Technical Specification requirements which provide regular checks to ensure limits are met.

Therefore, this proposed modification does not significantly reduce the margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendments request involves no significant hazards consideration.

Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, Potts and Trowbridge, 2300 N Street, NW, Washington, DC 20037.

NRC Section Chief: Marsha Gamberoni.
Duke Energy Corporation, Docket Nos. 50369 and 50370, McGuire Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

Date of amendment request: August 1, 2000.

Description of amendment request: The proposed amendments would provide revised spent fuel pool configurations, revised spent fuel pool storage criteria, and revised fuel enrichment and burnup requirements which take credit for soluble boron in maintaining acceptable margins of subcriticality in the spent fuel storage pools. Also, the proposed amendments would provide additional criteria for ensuring acceptable levels of subcriticality in the spent fuel storage pools.

1. Will the change involve a significant increase in the probability or consequence of an accident previously evaluated?

No, based upon the following:

Dropped Fuel Assembly

There is no significant increase in the probability of a fuel assembly drop accident in the spent fuel pools when considering the degradation of the or Boraflex panels in the spent fuel pool racks coupled with the presence of soluble boron in the spent fuel pool water for criticality control. The handling of the fuel assemblies in the spent fuel pool has always been performed in borated water, and the quantity of Boraflex remaining in the racks has no affect on the probability of such a drop accident.

The criticality analysis showed that the consequences of a fuel assembly drop accident in the spent fuel pools are not affected when considering the degradation of the Boraflex in the spent fuel pool racks and the presence of soluble boron.

Fuel Misloading

There is no significant increase in the probability of the accidental misloading of spent fuel assemblies into the spent fuel pool racks when considering the degradation of the Boraflex in the spent fuel pool racks and the presence of soluble boron in the pool water for criticality control. Fuel assembly placement and storage will continue to be controlled pursuant to approved fuel handling procedures to ensure compliance with the Technical Specification requirements. These procedures will be revised as needed to comply with the revised requirements which would be imposed by the proposed Technical Specification changes. Note that the proposed amendment would increase the number of different storage limits in Technical Specification 3.7.15. However, these revised storage limits were developed with input from station personnel. Their awareness, in conjunction with any procedure changes as described above, will provide additional assurance that an accidental misloading of a spent fuel assembly will not occur.

There is no increase in the consequences of the accidental misloading of spent fuel assemblies into the spent fuel pool racks because criticality analyses demonstrate that the pool will remain subcritical following an accidental misloading if the pool contains an adequate soluble boron concentration. Current Technical Specification 3.7.14 will ensure that an adequate spent fuel pool boron concentration is maintained in the McGuire spent fuel storage pools. A McGuire Station UFSAR change will revise Chapter 16, ``Selected Licensee Commitments'', to provide for adequate monitoring of the remaining Boraflex in the spent fuel pool racks. If that monitoring identifies further reductions in the Boraflex panels which would not support the conclusions of the McGuire Criticality Analysis, then the McGuire TS's and design bases would be revised as needed to ensure that acceptable subcriticality are maintained in the McGuire spent fuel storage pools.
[[Page 62386]]

Significant Change in Spent Fuel Pool Temperature

There is no significant increase in the probability of either the loss of normal cooling to the spent fuel pool water or a decrease in pool water temperature from a large emergency makeup when considering the degradation of the Boraflex in the spent fuel pool racks and the presence of soluble boron in the pool water for subcriticality control since a high concentration of soluble boron has always been maintained in the spent fuel pool water. Current Technical Specification 3.7.14 will ensure that an adequate spent fuel pool boron concentration is maintained in the McGuire spent fuel storage pools.

A loss of normal cooling to the spent fuel pool water causes an increase in the temperature of the water passing through the stored fuel assemblies. This causes a decrease in water density that would result in a decrease in reactivity when Boraflex neutron absorber panels are present in the racks. However, since a reduction in the amount of Boraflex present in the racks is considered, and the spent fuel pool water has a high concentration of boron, a density decrease causes a positive reactivity addition. However, the additional negative reactivity provided by the current boron concentration limit, above that provided by the concentration required to maintain keff less than or equal to 0.95 (1470 ppm), will compensate for the increased reactivity which could result from a loss of spent fuel pool cooling event. Because adequate soluble boron will be maintained in the spent fuel pool water, the consequences of a loss of normal cooling to the spent fuel pool will not be increased. Current Technical Specification 3.7.14 will ensure that an adequate spent fuel pool boron
concentration is maintained in the McGuire spent fuel storage pools.

A decrease in pool water temperature from a large emergency makeup causes an increase in water density that would result in an increase in reactivity when Boraflex neutron absorber panels are present in the racks. However, the additional negative reactivity provided by the current boron concentration limit, above that provided by the concentration required to maintain keff less than or equal to 0.95 (1470 ppm), will compensate for the increased reactivity which could result from a decrease in spent fuel pool water temperature. Because adequate soluble boron will be maintained in the spent fuel pool water, the consequences of a decrease in pool water temperature will not be increased. Current Technical Specification 3.7.14 will ensure that an adequate spent fuel pool boron concentration is maintained in the McGuire spent fuel storage pools.

2. Will the change create the possibility of a new or different kind of accident from any previously evaluated?

No. Criticality accidents in the spent fuel pool are not new or different types of accidents. They have been analyzed in Section 9.1.2.3 of the Updated Final Safety Analysis Report and in Criticality Analysis reports associated with specific licensing amendments for fuel enrichments up to 4.75 weight percent U235. Specific accidents considered and evaluated include fuel assembly drop, accidental misloading of spent fuel assemblies into the spent fuel pool racks, and significant changes in spent fuel pool water temperature. The accident analysis in the Updated Final Safety Analysis Report remains bounding.

The possibility for creating a new or different kind of accident is not credible. The amendment proposes to take credit for the soluble boron in the spent fuel pool water for reactivity control in the spent fuel pool while maintaining the necessary margin of safety. Because soluble boron has always been present in the spent fuel pool, a dilution of the spent fuel pool soluble boron has always been a possibility, however, a criticality accident resulting from a dilution accident was not considered credible. For the proposed amendment, the spent fuel pool dilution evaluation (Attachment 7) demonstrates that a dilution of the boron
concentration in the spent fuel pool water which could increase the rack keff to greater than 0.95 (constituting a reduction of the required margin to criticality) is not a credible event. The requirement to maintain boron concentration in the spent fuel pool water for reactivity control will have no effect on normal pool operations and maintenance. There are no changes in equipment design or in plant configuration. This new requirement will not result in the installation of any new equipment or modification of any existing equipment. Therefore, the proposed amendment will not result in the possibility of a new or different kind of accident.

3. Will the change involve a significant reduction in a margin of safety?

No. The proposed Technical Specification changes and the resulting spent fuel storage operating limits will provide adequate safety margin to ensure that the stored fuel assembly array will always remain subcritical. Those limits are based on a plant specific criticality analysis (Attachment 6) based on the
``Westinghouse Spent Fuel Rack Criticality Analysis Methodology'' described in Reference 1. The Westinghouse methodology for taking credit for soluble boron in the spent fuel pool has been reviewed and approved by the NRC (Reference 6). This methodology takes partial credit for soluble boron in the spent fuel pool and requires conformance with the following NRC Acceptance criteria for preventing criticality outside the reactor:
(1) keff shall be less than 1.0 if fully flooded with unborated water which includes an allowance for uncertainties at a 95% probability, 95% confidence (95/95) level; and
( 2) keff shall be less than or equal to 0.95 if fully flooded with borated water, which includes an allowance for uncertainties at a 95/95 level.

The criticality analysis utilized credit for soluble boron to ensure keff will be less than or equal to 0.95 under normal circumstances, and storage configurations have been defined using a 95/95 keff calculation to ensure that the spent fuel rack keff will be less than 1.0 with no soluble boron. Soluble boron credit is used to provide safety margin by maintaining keff less than or equal to 0.95 including uncertainties, tolerances and accident conditions in the presence of spent fuel pool soluble boron. The loss of substantial amounts of soluble boron from the spent fuel pool which could lead to exceeding a keff of 0.95 has been evaluated (Attachment 7) and shown to be not credible. Accordingly, the required margin to criticality is not reduced.

The evaluations in Attachment 7, which show that the dilution of the spent fuel pool boron concentration from the conservative assumed initial boron concentration (2475 ppm) to the minimum boron concentration required to maintain keff 0.95 (730 ppm) is not credible, combined with the 95/95 calculation which shows that the spent fuel rack keff will remain less than 1.0 when flooded with unborated water, provide a level of safety comparable to the conservative criticality analysis methodology required by References 2, 3 and 4.

Therefore the proposed changes in this license amendment will not result in a significant reduction in the facility's margin of safety.

Attorney for licensee: Ms. Lisa F. Vaughn, Duke Energy Corporation, 422 South Church Street, Charlotte, North Carolina 282011006.

NRC Section Chief: Richard L. Emch, Jr.
Indiana Michigan Power Company, Docket Nos. 50315 and 50316, Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

Date of amendment requests: September 26, 2000.

Description of amendment requests: The proposed amendments would revise the current licensing basis in the Updated Final Safety Analysis Report by requiring operator action to mitigate the effects of a loss of seal injection (LOSI) cooling to the reactor coolant pumps (RCPs).

1. Does the change involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated?

The proposed change to the licensing basis recognizes that if RCP Number 1 seal leakoff rates are low, continuous RCP operation following a sustained LOSI may no longer be permitted. Tripping the plant, securing the affected RCPs, and maintaining hot standby [[Page 62387]]
conditions following a sustained LOSI will permit adequate RCP seal cooling by readily achievable process controls. These actions ensure that the probability of developing excessive seal leakage equivalent to that of a previously evaluated loss of coolant accident (LOCA), has not been significantly increased. Plant and RCP tripping are anticipated transients that do not involve plant operation outside design limits.

The consequences of large and smallbreak (SB) LOCAs have been evaluated and it has been shown that the radiological consequences of these events do not result in unacceptable exposures to members of the public. Therefore, even if stopping of the RCPs following a LOSI and control of process parameters as described above does not preclude RCP seal failures, the consequences of such failure are bounded by the current accident analysis.

Therefore, the probability of occurrence or the consequences of accidents previously evaluated are not significantly increased.

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The leakage resulting from failed RCP seals may be large enough to be considered a SBLOCA and industry data on SBLOCA initiating frequencies includes the contribution from failed RCP seals. SBLOCAs are a previously evaluated class of accidents. There is no new or different kind of accident created as a result of this change.

Therefore, the change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the change involve a significant reduction in a margin of safety?

The original design objective for the controlled leakage seal assemblies in the RCPs was to permit sufficient controlled leakage following a LOSI, such that cooling of the leakage provided by the thermal barrier heat exchanger would be sufficient to continue RCP operation unabated without challenging seal integrity. This is an implied margin of safety for seal integrity, even if not explicitly defined in the basis for any Technical Specification. It has been postulated that the reduced seal leakoff will no longer permit continuous RCP operation following a LOSI. The proposed change to the licensing basis recognizes this condition and requires pump tripping if seal injection cannot be restored prior to receiving high temperature alarms in the leakoff return lines. Pump tripping reduces the heat generated in the pump and permits readily achievable process controls to maintain adequate seal cooling and an adequate margin to seal failure.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive, Buchanan, MI 49107.

NRC Section Chief: Claudia M. Craig.
Indiana Michigan Power Company, Docket No. 50316, Donald C. Cook Nuclear Plant, Unit 2, Berrien County, Michigan

Date of amendment request: September 30, 2000.

Description of amendment request: The proposed amendment would allow an extension of the steam generator tube inspection surveillance requirements of Technical Specification (T/S) Surveillance Requirement 4.4.5.3. The proposed amendment would prevent a midcycle shutdown to meet the required 40calendar month inspection interval of SR 4.4.5.3 and would allow the steam generator tube inspection to be performed during the refueling outage following the current operating cycle.

1. Does the change involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated?

The accident considered applicable to the proposed change is a steam generator tube rupture (SGTR). The precursors/initiators of a SGTR (degraded, defective, or leaking tubes) are not known or expected to be present in the CNP [Cook Nuclear Plant] Unit 2 steam generators. These steam generators were newly installed in 1988, and include corrosion prevention design features not included in previous generations of steam generators.

There are no active degradation mechanisms present in the Unit 2 steam generators. Any tube imperfections that may be present or that may be initiated during the current operating cycle are not expected to progress to the point of tube failure before the next refueling outage.

Considering the condition of the steam generators and the operational time between inspections, the proposed change will not significantly increase the probability of occurrence of an accident.

The proposed change will not affect the scope, methodology, acceptance limit, or corrective measures of the existing steam generator examination program.

Unit 2 recently completed an extended shutdown that effectively limited the operational time that is the basis for the surveillance frequency. When the reactor is shut down and the reactor coolant system is at a reduced temperature, the steam generators are not subject to conditions that lead to significant tube degradation. Based on power operation time, the proposed extension will not increase the operating interval between surveillances beyond that currently allowed by [the] T/S.

The steam generator tube inspection interval is not used in the SGTR accident analysis. The proposed change will, therefore, not affect the accident analysis or methodology.

The severity of an analyzed tube rupture event is not related to the time interval between inspections. The proposed change does not affect allowable leakage rates or source terms, and does not change the duration of an SGTR or the response to the event. Because the severity of an accident is not increased by the proposed change, there is no impact on offsite dose considerations.

Therefore, the probability of occurrence or the consequences of accidents previously evaluated are not significantly increased.

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change will not result in a change in plant configuration or operation. Plant systems and components will not be operated in a different manner because of this change. The proposed change does not affect or create new accident initiators or precursors.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the change involve a significant reduction in a margin of safety?

The T/S limit of one gallon per minute total steam generator tube leakage ensures the offsite dose from tube leaks is limited to a small fraction of 10 CFR 100 limits. The T/S leakage limit of 500 gallons per day in one steam generator is based on ensuring tube integrity in the event of a steam line rupture or loss of coolant accident. Because the offsite dose considerations from steam generator tube failures are limited by the primarytosecondary leak rate program and not the tube inspection program, the proposed change has no impact on offsite dose.

There are no tubes in service in any of the Unit 2 steam generators that were found to be degraded, and no active steam generator tube degradation is known to be occurring. Therefore, the available margin in tube wall thickness is not being significantly reduced. During the last inspection, 50% of the tubes were inspected (more than sixteen times the T/S requirement), and none were found to exceed the plugging limit, providing additional assurance that safety margins are not being reduced. The absence of tube
degradation, along with the material and design features and chemistry controls, provide reasonable assurance that tube repair limits will not be approached during the current operating cycle.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the [[Page 62388]]
amendment requests involve no significant hazards consideration.

Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive, Buchanan, MI 49107.

NRC Section Chief: Claudia M. Craig.
Northeast Nuclear Energy Company, et al., Docket No. 50423, Millstone Nuclear Power Station, Unit No. 3, New London County, Connecticut

Date of amendment request: June 30, 2000, as supplemented September 22, 2000.

Description of amendment request: The proposed changes would modify Sections 2.4.13.5, ``Design Bases for Subsurface Hydrostatic Loading'' 2.5.4.6.1, ``Design Basis for Groundwater'' 3.4.1.2, ``Permanent Dewatering System'' 3.8.1.6.4, ``Waterproofing Membrane'' 3.8.1.6.5, ``Steel Liner and Penetrations'' 9.3.3.1, ``Reactor Plant Vent and Drain Systems, Design Bases'' 9.3.3.2.4, ``Reactor Plant Aerated Drains System'' 9.3.3.2.4.1, ``SafetyRelated Containment Recirculation Cubicle Sump'' 9.3.3.3, ``Safety Evaluation'' 9.3.3.4, ``Tests and Inspections'' and 12.3.1.3.2, ``PostAccident Access to Vital Areas'' Tables 1.81, 3.21, 8.33, 12.33, and 12.34; and Figures 3.867 and 9.36 of the Final Safety Analysis Report (FSAR) to reflect the addition of the new subsystem and its impact on other safetyrelated systems. The new sump pump system creates the possibility of a malfunction of a different type than previously evaluated in the FSAR because of the system's dependence on electrical power; only one non environmentally qualified, nonsafetyrelated pump is provided; and portions of the Engineered Safety Feature Building structure are now credited with preventing Recirculation Spray System (RSS) cubicle flooding. Additionally, the proposed changes involve deviations from safety classification and ``code & standards,'' Standard Review Plan 3.4.1 and Regulatory Guide (RG) 1.26.

1. Involve a significant increase in the probability or consequences of an accident previously evaluated.

This license amendment request deals with changes in Millstone Unit No. 3 Final Safety Analysis Report (FSAR) due to the
installation of a new sump pump system in the Engineered Safety Features Building (ESFB). The sump pump system which prevents inleakage through the containment basemat is not connected to and is fully independent of the reactor coolant system. Therefore, the proposed changes to this system will not increase the probability of occurrence of a Loss of Coolant Accident (LOCA). The new system is a support system for the Recirculation Spray System (RSS) and containment protective boundary which are mitigation design features. Therefore, the new system does not increase the
probability of occurrence of accidents previously evaluated.

The proposed changes to the groundwater sump system separate the sump from the RSS pump cubicle. As such, the proposed changes would preclude flooding of the RSS cubicles and a potential malfunction of the RSS pumps. The RSS pumps function to provide containment and core cooling, as early as 11 minutes and 30 minutes, respectively, post LOCA. Operability of the RSS pumps is required long term. Since the changes do not affect the operation of the RSS pumps, they will not increase the consequences of a LOCA.

The new collection tank 3SRWTK1 will be installed in the location of the existing abandoned in place Chemical Addition Tank (CAT) 3QSS*TK2, by the Refueling Water Storage Tank (RWST). The tank will be seismically supported utilizing similar struts and attachments to the RWST as the removed CAT. A calculation has confirmed that there is no impact on the seismic qualification of the RWST as a res

SUMMARY:

Operating licenses, amendments; no significant hazards considerations; biweekly notices,

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