ML070520383

Administrative Control of Penetrations During Refueling (License Amendment Request No. 06-0002)

ML070520383
Person / Time
Site:	Davis Besse
Issue date:	02/12/2007
From:	Bezilla M FirstEnergy Nuclear Operating Co
To:	Document Control Desk, NRC/NRR/ADRO
References
33011, LAR 06-001
Download: ML070520383 (23)

Text

FENOC

""tt5501 North State Route 2 FirstEnergy Nuclear Operating Company Oak Harbor, Ohio 43449 Mark B. Bezilla 419-321-7676 Vice President - Nuclear Fax: 419-321-7582 Docket Number 50-346 10 CFR 50.90 License Number NPF-3 Serial Number 3301 February 12, 2007 United States Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

Subject:

Davis-Besse Nuclear Power Station Docket Number 50-346 Administrative Control of Penetrations during Refueling (License Amendment Request No. 06-0002)

Ladies and Gentlemen:

Pursuant to 10 CFR 50.90, the FirstEnergy Nuclear Operation Co. (FENOC) hereby requests amendment of the Operating License (NPF-3) for the Davis-Besse Nuclear Power Station, Unit 1 (DBNPS). The proposed amendment would revise Technical Specification (TS) 3/4.9.4, "Containment Penetrations," to allow containment penetrations that provide direct access from the containment atmosphere to the outside atmosphere to be open during refueling activities if appropriate administrative controls are established. The proposed changes are consistent with NRC-approved Technical Specification Task Force (TSTF) Traveler TSTF-312-A, Revision 1. describes the proposed TS changes, and includes the supporting technical and regulatory evaluations. Proposed changes to the TS Bases are also provided for information only.

Approval of the proposed amendment is requested by December 13, 2007, to support implementation and scheduling for the next DBNPS refueling outage. Once approved, the amendment will be implemented within 60 days.

A list of regulatory commitments made in this letter, including the Enclosure, is included in. If there are any questions or if additional information is required, please contact Mr. Henry L. Hegrat, Supervisor - Fleet Licensing, at (330) 315-6944.

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Docket Number 50-346 License Number NPF-3 Serial Number 3301 Page 2 The statements contained in this submittal, including its associated enclosures, are true and correct to the best of my knowledge and belief I am authorized by the FirstEnergy Nuclear Operating Company to make this submittal. I declare under penalty of perjury that the foregoing is true and correct.

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• - TS Page Markups

• - Changes to TS Bases

• - Retyped TS Pages : Commitment List cc:

Regional Administrator, NRC Region III NRC/NRR Project Manager Executive Director, Ohio Emergency Management Agency, State of Ohio (NRC Liaison)

NRC Region IIl, DB-1 Senior Resident Inspector Utility Radiological Safety Board

Docket Number 50-346 License Number NPF-3 Serial Number 3301 LICENSEE EVALUATION LICENSE AMENDMENT REQUEST NUMBER 06-0002

Subject:

Administrative Control of Penetrations during Refueling Technical Specification (TS) 3/4.9.4, Refueling Operations - Containment Penetrations 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

4.0 REGULATORY EVALUATION

4.1 Significant Hazards Consideration 4.2 Applicable Regulatory Requirements/Criteria 4.3 Precedent 4.4 Conclusions

6.0 REFERENCES

ATTACHMENTS

1.

Proposed Mark-up of Technical Specification Pages

2.

Technical Specification Bases Page

3.

Proposed Re-typed Technical Specification Pages

Docket Number 50-346 License Number NPF-3 Serial Number 3301 Page 2 of 12 1.0

SUMMARY

DESCRIPTION This evaluation supports a request to revise Operating License Number NPF-3 for the Davis-Besse Nuclear Power Station, Unit Number 1 (DBNPS). The request is applicable to containment penetrations capable of closure by a manual or automatic isolation valve, blind flange, or equivalent, and would allow the applicable containment building penetrations that provide directaccess from the containment atmosphere to the outside atmosphere to be open during refueling activities if appropriate administrative controls are established. Currently, TS 3.9.4(c) requires that the affected penetrations be closed during core alterations or movement of irradiated fuel within the containment.

The proposed change would revise TS Limiting Condition for Operation (LCO) 3.9.4, "Containment Penetrations," and is consistent with NRC-approved Technical Specification Task Force (TSTF) Traveler TSTF-312-A, Revision 1, which was approved on August 16, 1999, and the corresponding note to the Babcock and Wilcox Owners Group Standard Technical Specification (STS) 3.9.3, "Containment Penetrations."

The proposed amendment would add a note applicable to Technical Specification (TS) 3/4.9.4, Containment Penetrations, to allow penetrations included under TS 3.9.4(c) to be opened during core alterations or movement of irradiated fuel, under administrative controls. This class of penetrations includes those penetrations which provide direct access from the containment atmosphere to the atmosphere outside containment. The purpose of this change is to allow local leak rate testing of containment penetrations to be performed during refueling outages, while fuel movement is in progress. The FirstEnergy Nuclear Operating Company (FENOC) believes that this change will result in a reduction in future outage durations.

The dose consequences of this change are bounded by the dose consequences of TS 3.9.4.b, which allows both personnel air lock doors to be open during core alterations or movement of irradiated fuel within the containment, and therefore are well within the radiological dose guidelines of 10 CFR 100.11, "Determination of exclusion area, low population zone, and population center distance," and General Design Criteria (GDC) 19, "Control Room."

2.0 DETAILED DESCRIPTION Containment is a barrier to the release of fission products that breach the fuel cladding and reactor coolant system during a core-damaging accident. The containment barrier, including penetrations, acts to limit the release of fission products and the resulting that offsite radiation exposure.

TS 3.9.4 currently precludes opening containment penetrations during core alterations or fuel movement inside containment. The proposed change affects TS 3.9.4(c), which is part of

Docket Number 50-346 License Number NPF-3 Serial Number 3301 Page 3 of 12 TS 3.9.4, Containment Penetrations. TS 3.9.4(c) requires that during core alterations or movement of irradiated fuel within the containment: "Each penetration providing direct access from the containment atmosphere to the atmosphere outside containment shall be (1) closed by a manual or automatic isolation valve, blind flange, or equivalent, or (2) capable of being closed from the control room by an OPERABLE containment purge and exhaust valve upon receipt of a high radiation signal from the containment purge and exhaust system noble gas monitor."

FENOC maintains a surveillance test procedure that establishes specific closure controls for containment penetrations. This procedure is performed within 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to the start of core alterations or movement of irradiated fuel within containment, and is repeated at least once per seven days as long as core alterations or movement of irradiated fuel within containment continue. The procedure requires a review of containment status, and verifies completion of various activities that ensure appropriate containment penetration configuration.

The proposed change would allow containment penetration flow paths controlled by TS 3.9.4(c) to be open during operations involving core alterations or fuel movement inside containment if appropriate administrative controls are established and maintained. Specifically, the proposed change would add the following note, applicable to TS 3.9.4(c):

"NOTE: Penetration flow path(s) providing direct access from the containment atmosphere to the outside atmosphere may be unisolated under administrative controls."

As part of implementation of the proposed amendment, applicable procedures will be revised to permit opening the affected penetrations during fuel movement as long as administrative controls are in place to ensure.prompt closure of the affected penetration in the event of a Fuel Handling Accident. These controls will entail assignment of a designated individual who will be readily available to isolate the flow path in the event of a Fuel Handling Accident. This individual may be a part of the team assigned to perform local leak rate testing of the associated penetration.

The language in the proposed change is identical to corresponding language contained in STS 3.9.3, and is consistent with Revision 1 of TSTF-312-A, "Administratively Control Containment Penetrations." TSTF-312-A was approved based on the condition that licensees utilizing the TSTS (1) demonstrate acceptable radiological consequences from a Fuel Handling Accident (FHA) inside containment, and (2) commit to implementation of administrative procedures that ensure that open containment penetrations can and will be promptly closed in the event of an FHA inside containment.

3.0 TECHNICAL EVALUATION

The containment isolation system is designed so that each of its components requiring testing can be tested periodically. Pressure retaining components of the containment isolation system, including piping and valves, undergo periodic leak testing. Leakage rate testing of components

Docket Number 50-346 License Number NPF-3 Serial Number 3301 Page 4 of 12 requiring a local leakage rate test (Type B or C) is done in accordance with 10 CFR 50 Appendix J, Option B, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors. Type B tests include tests intended to detect local leaks and to measure leakage across each pressure-retaining or leakage limiting boundary for penetration containing resilient seals, gaskets, sealant compounds, piping penetrations fitted with expansion bellows, electrical penetrations fitted with metal seal assemblies, and air lock door seals. Type C tests include tests intended to measure containment isolation valve leakage rates. Type B and C tests are typically performed during reactor shutdown for refueling. Usually testing is performed by measuring the rate of pressure loss (or corresponding make-up flow rate) in a test chamber that includes the containment penetration, pressurized with air or nitrogen. Activities associated with connecting the test equipment, pressurizing the test chamber before the test, and venting the test chamber after the test require brief exposure of the test chamber to the outside atmosphere.

Compliance with TS 3.9.4, "Containment Penetrations," ensures that the consequences of a postulated Fuel Handling Accident inside containment during core alterations or fuel handling remain within acceptable limits. The TS Limiting Condition for Operation (LCO) requires that at least one barrier to the release of radioactive material be operable at all times. Exceptions are provided for the containment equipment hatch and the containment personnel airlock. LCO 3.9.4(c) requires each penetration providing direct access from the containment atmosphere to the atmosphere outside containment to be either (1) closed by a manual or automatic isolation valve, blind flange, or equivalent, or (2) be capable of being closed from the control room by an operable containment purge and exhaust system valve upon receipt of a high radiation signal from the containment purge and exhaust system noble gas monitor.

The changes proposed by this license amendment request are consistent with TSTF-312-A, Revision 1, and the Improved Standard Technical Specifications for Babcock and Wilcox Plants, LCO 3.9.3. The proposed administrative controls include assignment of specified individuals designated and readily available to isolate the flow path, who would close the penetration upon notification of a Fuel Handling Accident. These individuals may be part of the team assigned to perform local leak rate testing of the associated penetration. The NRC has already approved similar controls for the personnel airlock; for example, refer to LCO 3.9.4(b).

Updated Safety Analysis Report (USAR) section 15.4.7, "Fuel Handling Accident," provides the results of the Fuel Handling Accident analysis, performed in accordance with the assumptions and guidelines of Safety Guide 25, "Assumptions used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors," dated March 23, 1972. No credit is taken for containment isolation, and the activity from one fuel assembly with a bumup of 60,000 megawatt-days per metric ton uranium (MWD/MTU) was assumed to be released over a two hour period. The results meet the acceptance criteria provided in NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, and therefore are well within the dose guidelines of 10 CFR 100.11.

Docket Number 50-346 License Number NPF-3 Serial Number 3301 Enclosure I Page 5 of 12 TS Amendment 202 permitted both doors of the containment personnel air lock to be open during core alterations or movement of irradiated fuel in containment. Analyses performed in support of Amendment 202 are summarized in USAR section 15.4.7.3. These analyses calculated the control room dose for a Fuel Handling Accident inside containment, assuming fuel activity consistent with a fuel assembly with 60,000 MTD/MTU burnup, 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of fuel assembly decay, and no credit for removal of iodine activity by the control room emergency ventilation system. Although the DBNPS construction permit predates the General Design Criteria contained in 10 CFR 50 Appendix A, the results were compared with, and found to be well within the control room dose acceptance criteria of GDC 19, "Control Room." GDC 19 requires that adequate radiation protection be provided to permit access and occupancy of the control room under accident conditions, without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident.

In addition to the current USAR analysis, during its review of TS Amendment 202, the NRC staff completed an independent analysis of the potential radiological consequences of a Fuel Handling Accident, with the containment personnel air lock open, to determine conformance with the requirements of 10 CFR 100 and GDC 19. The staff's analysis utilized the accident source term given in Regulatory Guide 1.4, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors," the assumptions contained in Regulatory Guide 1.25, and the review procedures contained in Standard Review Plan (SRP) Sections 15.7.4 and 6.4. The staff assumed an instantaneous puff release of noble gases and radioiodine from the gap and plenum of the broken fuel rods. LCO 3.9.10 requires that, as a minimum, 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated within the reactor pressure vessel during movement of fuel assemblies or control rods within the reactor pressure vessel while in Mode 6. Therefore, the gas bubbles will pass through at least 23 feet of water covering the fuel prior to reaching the containment atmosphere. The guidance of NUREG/CR-5009, "The Assessment of the Use of Extended Burnup Fuel in Light Water Power Reactors, "was utilized to calculate thyroid dose based on use of extended burnup fuel at DBNPS. All airborne activity reaching the containment atmosphere was assumed to exhaust to the environment within two hours. As stipulated by LCO 3.9.3, the reactor must be sub-critical for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> during movement of irradiated fuel in the reactor pressure vessel, and therefore, the gap activity was assumed to have decayed for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The computed offsite doses and control room operator doses were within the acceptance criteria given in SRP Section 15.7.4 and GDC 19.

FENOC notes that the assumptions used in the above analyses were extremely conservative, because during refueling activities there is very little driving force in containment to cause the release to occur as quickly or completely as assumed. In addition, the assumption of a total release over a two hour period is independent of the release path, so that the results (at a minimum) are bounding for the proposed note for LCO 3.9.4(c): Furthermore, it is anticipated that the release path for a penetration opened under 3.9.4(c) would be much smaller than the potential release path already approved under LCO 3.9.4(b). Therefore, allowing penetration

Docket Number 50-346 License Number NPF-3 Serial Number 3301 Page 6 of 12 flow paths to be un-isolated during core alterations or movement of irradiated fuel inside containment will not invalidate the conclusion that the potential dose consequences from a Fuel Handling Accident are below 10 CFR 100.11 and GDC 19 limits. FENOC understands that the term "un-isolated" as used in TSTF-312-A refers to the condition of the penetration flow path rather than any action taken by the operator, since the analysis results are the same regardless of whether core alterations or fuel movement begin before or during opening of the penetration flowpath.

Based on the Fuel Handling Accident analyses and the administrative controls specified for the proposed allowance to un-isolate containment penetration flow paths, the proposed changes are acceptable. With respect to the proposed administrative controls, the proposed license amendment provides assurance that offsite dose levels associated with a Fuel Handling Accident inside containment will be maintained well within applicable regulatory limits.

4.0 REGULATORY EVALUATION

4.1 Significant Hazards Consideration An evaluation has been performed to determine whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1.

Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change would allow containment penetrations identified under Technical Specification 3.9.4(c) to remain open during fuel movement and core alterations. These penetrations are normally closed during this time period to prevent the release of radioactive material in the event of a Fuel Handling Accident inside containment. These penetrations are not initiators of any accident. The probability of a Fuel Handling Accident is unaffected by the status of these penetrations.

The Fuel Handling Accident analyses demonstrate that the maximum offsite dose is well with the acceptance limits specified in SRP 15.7.4, and the control room dose is within the acceptance criteria specified in GDC 19. Furthermore, the existing analysis results are independent of the containment release path, and therefore are unaffected by the proposed change.

Docket Number 50-346 License Number NPF-3 Serial Number 3301 Page 7 of 12 Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2.

Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change does not involve the addition or modification of any plant equipment. Also, the proposed change will not alter the design, configuration, or method of operation of the plant beyond the standard functional capabilities of the equipment. The proposed change involves a Technical Specification change that will allow containment penetrations identified under Technical Specification 3.9.4(c) to remain open during fuel movement and core alterations. Open penetrations are not accident initiators, and will not create the possibility of a new kind of accident. Administrative controls will be implemented to ensure the capability to close the affected containment penetrations in the event of a Fuel Handling Accident.

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

3.

Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed change has the potential to slightly increase the post-Fuel Handling Accident dose at the site boundary and in the control room. However, the existing analyses take no credit for containment of the release, so that the existing analysis results will remain bounding.

Based on the above responses, FENOC concludes that the proposed amendment does not involve a, significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and accordingly, a finding of"no significant hazards consideration" is justified.

4.2 Applicable Regulatory Requirements/Criteria The following lists the regulatory requirements and plant-specific design bases related to the proposed change.

• The regulatory basis for TS 3.9.4, "Containment Penetrations," is to ensure that containment is capable of containing fission product radioactivity that may be released from the fuel

Docket Number 50-346 License Number NPF-3 Serial Number 3301 Enclosure I Page 8 of 12 following a Fuel Handling Accident inside containment.

• At the time 10 CFR 50, Appendix A was published, it was not the intent of the Nuclear Regulatory Commission to make it retroactive to plants that had received their Construction Permits prior to May 21, 1971. This position was supported in a Memorandum from S. J.

Chilk (Secretary) to J. M. Taylor (Executive Director for Operations), September 18, 1992, "SECY-92-223 - Resolution of Deviations Identified During the Systematic Evaluation Program" that stated, in part:

The Commission (with all Commissioners agreeing) has approved the staff proposal in Option 1 of this paper in which the staff will not apply the General Design Criteria (GDC) to plants with construction permits issued prior to May 21, 1971. At the time of the promulgation of Appendix A to 10 CFR Part 50, the Commission stressed that the GDC were not new requirements and were promulgated to more clearly articulate the licensing requirements and practice in effect at that time. While compliance with the intent of the GDC is important, each plant licensed before the GDC were formally adopted was evaluated on a plant specific basis, determined to be safe, and licensed by the Commission.

Furthermore, current regulatory processes are sufficient to ensure that plants continue to be safe and comply with the intent of the GDC.

Because the DBNPS construction permit was issued before the GDC were incorporated into the regulations and became effective, the DBNPS is not required to meet the literal wording of the GDC. However, the DBNPS USAR provides that "the design of the Davis-Besse Nuclear Power Station meets the intent of Appendix A, 10 CFR 50, the General Design Criteria for Nuclear Power Plants as published in the Federal Register on February 20, 1971, and as amended in the Federal Register on July 7, 1971." The degree of conformance to relevant GDC is as identified below:

o GDC 16, "Containment Design." Reactor containment and associated systems are provided to establish an essentially leak tight barrier against the uncontrolled release of radioactivity to the environment and to ensure that the containment design conditions important to safety are not exceeded for as long as postulated accident conditions require.

o GDC 19, "Control Room." A control room is provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions. Adequate radiation protection is provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole-body, or its equivalent to any part of the body, for the duration of the accident.

Safe occupancy of the control room during abnormal conditions is provided for in the

Docket Number 50-346 License Number NPF-3 Serial Number 3301 Enclosure I Page 9 of 12 design. The Control Room Emergency Ventilation System (CREVS) is provided with radiation detectors and appropriate alarms. When CREVS is operating, control room air is recirculated (with or without minimum makeup) through HEPA filters and charcoal adsorbers. When CREVS is operating with makeup air, a positive control room pressure is maintained to minimize in-leakage.

o GDC 54, "Piping Systems Penetrating Containment." Piping systems penetrating containment are provided with leak detection, isolation, and containment capabilities having redundancy, reliability, and performance capabilities which reflect the importance to safety of isolating these piping systems. These piping systems are designed with a capability of testing periodically the operability of the isolation valves and associated apparatus and to determine if valve leakage is within acceptable limits.

All piping penetrations, with the exception of a few that are part of the Safety Features Actuation System, are provided with double barrier isolation systems, so that no single credible failure or malfunction of one active component can result in loss of isolation capability or intolerable leakage. The installed double barriers take the form of closed piping systems inside and outside of the containment vessel and various types of isolation valves. Containment and all piping penetrations are designed so that all isolation systems can be tested independently for leakage.

o GDC 56, "Primary Containment Isolation." Each line that connects directly to the containment atmosphere and penetrates containment is provided with containment isolation valves; unless it can be demonstrated that the containment isolation provisions for a specific class of lines, such as instrument lines, are acceptable on some other basis.

o GDC 61, "Fuel Storage and Handling and Radioactivity Control." The fuel storage and handling, radioactive waste, and other systems which may contain radioactivity are designed to ensure adequate safety under normal and postulated accident conditions.

The parameters of concern and the acceptance criteria applied are based on the requirements of 10 CFR 100.11 with respect to the calculated radiological consequences of a Fuel Handling Accident, and GDC 61 with respect to appropriate containment, confinement, and filtering systems.

Regulatory Guidance

• Safety Guide 25, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors," March 25, 1972, provides the basis for the assumptions used in USAR 15.4.7, "Fuel Handling Accident."

" The results meet the acceptance criteria provided in NUREG-0800, Standard Review Plan

Docket Number 50-346 License Number NPF-3 Serial Number 3301 Page 10 of 12 for the Review of Safety Analysis Reports for Nuclear Power Plants, and therefore are well within the dose guidelines of 10 CFR 100.11.

4.3 Precedent The proposed change is consistent with NRC-approved Technical Specification Task Force (TSTF) Traveler TSTF-312-A, Revision 1, which was approved on August 16, 1999, and the corresponding note to the Babcock and Wilcox Owners Group Standard Technical Specification (STS) 3.9.3, "Containment Penetrations." The proposed change is similar to changes made August 30, 2002 under Amendment Nos. 184 and 127 for Facility Operating License Nos. DPR-67 and NPF-16 for St. Lucie Plant, Units 1 and 2 (TAC Nos. MB5188 and MB5189), as well as changes made on September 11, 2002 under Amendment No. 144 to Facility Operating License No. NPF-4 and Amendment No. 144 to Facility Operating License No. NPF-51 and Amendment No. 144 to Facility Operating License No. NPF-74 for the Palo Verde Nuclear Generating Station, Units 1, 2, and 3, respectively (TAC Nos. MB5128, MB5129, and MB5130). The proposed DBNPS change only affects TS 3.9.4(c), while the precedent amendments also included changes to TS sections corresponding to TS 3.9.4(a) and TS 3.9.4(b). Changes to TS 3.9.4(a) and TS.3.9.4(b) are not being requested at this time, because these paragraphs were addressed by Amendment No. 202 to License No. NPF-3, dated November 17, 1995.

4.4 Conclusions The dose consequences of this change are within the limits of 10 CFR 100.11 and GDC 19, and require no change to the existing Fuel Handling Accident Analysis. Therefore, the proposed license amendment is in compliance with Regulatory Guide 1.25, NUREG/CR-5009, and Section 15.7.4 of the SRP (NUREG-0800).

Based on the considerations discussed above, (1) there is reasonable assurance that the health and safety or 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) issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

FENOC has determined that the proposed license amendment would change requirements with respect to the installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. FENOC has evaluated the proposed change and has determined that the change does not involve, (i) a significant hazards consideration, (ii) a significant change in the types of or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in

Docket Number 50-346 License Number NPF-3 Serial Number 3301 Page 11 of 12 individual or cumulative occupational radiation exposure. As discussed above, the proposed change does not involve a significant hazards consideration and the analysis demonstrates that the consequences from a Fuel Handling Accident are well within the 10 CFR 100.11 limits.

Implementation of administrative controls precludes a significant increase in occupational radiation exposure. Accordingly, the proposed change meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51, specifically 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), an environmental assessment of the proposed change is not required.

6.0 REFERENCES

1.

Technical Specification Task Force, TSTF-312-A, Revision 1, "Administratively Control Containment Penetrations," July 17, 1999.

2.

DBNPS License No. NPF-3, Amendment No. 202, Dated November 17, 1995.

3.

DBNPS Updated Safety Analysis Report Section 3D, "Conformance With the NRC General Design Criteria, Safety Guides, and Information Guides."

Section 15.4.7, "Fuel-Handling Accident."

4.

Code of Federal Regulations, Title 10, Energy 0

10 CFR 20, Standards for Protection Against Radiation.

0 10 CFR 50, Appendix A, "General Design Criteria for Nuclear Power Plants."

10 CFR 51, "Environmental Protection Regulations for Domestic Licensing and Related Regulatory Functions."

10 CFR 100.11, "Determination of Exclusion Area, Low Population Zone, and Population Center Distance."

5.

Regulatory Guides Regulatory Guide 1.4, Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors," Revision 2, June 1974.

Docket Number 50-346 License Number NPF-3 Serial Number 3301 Page 12 of 12 Safety Guide 25, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors," March 23, 1972.

6.

NUREGs NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, Section 15.7.4, "Radiological Consequences of Fuel Handling Accidents."

NUREG/CR-5009, Assessment of the Use of Extended Burnup Fuel in Light Water Power Reactors.

Docket Number 50-346 License Number NPF-3 Serial Number 3301 - Attachment 1 PROPOSED MARK-UP OF TECHNICAL SPECIFICATION PAGES (2 pages follow)

REFUELING OPERATIONS CONTAINMENT PENETRATIONS LIMITING CONDITION FOR OPERATION 3.9.4 The containment penetrations shall be in the following status:

a. The equipment hatch cover closed and held in place by a minimum of four bolts, except the equipment hatch may be open provided the requirements of Specification 3.9.12 are satisfied,

b. A minimum of one door in each air lock closed, but both doors of the containment personnel air lock may be open provided that at least one personnel air lock door is capable of being closed and a designated individual is available immediately outside the personnel air lock to close the door, and

c. Each penetration providing direct access from the containment atmosphere to the atmosphere outside containment shall be either:

1. Closed by a manual or automatic isolation valve, blind flange, or equivalent, or

2. Be capable of being closed from the control room by an OPERABLE containment purge and exhaust valve upon receipt of a higher radiation signal from the containment purge and exhaust system noble gas monitor.

NOTE---------------

Penetration flow path(s) providing direct access from the containment atmosphere to the outside atmosphere may be unisolated under administrative controls.

APPLICABILITY: During CORE ALTERATIONS or movement of irradiated fuel within the containment.

ACTION:

a. With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or movement of irradiated fuel in the containment.

b. With the requirements of Specification 3.9.4.c not satisfied for the containment purge and exhaust system, close at least one of the isolation valves for each of the purge and exhaust penetrations providing direct access from the containment atmosphere to the outside atmosphere within one hour.

c. The provisions of Specification 3.0.3 are not applicable.

DAVIS-BESSE, UNIT I 3/4 9-4 Amendment No. 186, 202, 221, 251

REFUELNGINFORMATION ONLY CONTAINMENT PENETRATIONS SURVEILLANCE REQUIREMENTS 4.9.4 Each of the above required containment penetrations shall be determined to be either in its required condition or capable of being closed by an OPERABLE containment purge and exhaust valve, within 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to the start of and at least once per 7 days during CORE ALTERATIONS or movement of irradiated fuel in the containment, by:

a. Verifying the penetrations are in their required condition, or

b. Verifying that with the containment purge and exhaust system in operation, and the containment purge and exhaust system noble gas monitor capable of providing a high radiation signal to the control room, that after initiation of the high radiation signal, the containment purge and exhaust isolation valves can be closed from the control room.

DAVIS-BESSE, UNIT I 3/4 9-5 Amendment No. 186, 202, 224

Docket Number 50-346 License Number NPF-3 Serial Number 3301 - Attachment 2 TECHNICAL SPECIFICATION BASES PAGE (1 page follows)

Notes: The Bases page is provided for information only. The entire section has been reformatted. Only the page with technical changes has been provided.

___FOR INFORMATION ONLY 3/4.9 REFUELING OPERATIONS BASES 3/4.9.4 CONTAINMENT PENETRATIONS (Continued)

The individual designated to be continuously available to close the air lock door must be stationed at the auxiliary building side of the air lock. A containment personnel air lock door is considered capable of being closed if the door is not blocked in such a way that it cannot be expeditiously closed, and any hoses and cables running through the airlock employ a means to allow safe, quick disconnect or severance, and are tagged at the airlock with specific instructions to expedite removal. The LCO 3.9.10 requirement to maintain a minimum of 23 feet of water over the top of irradiated fuel assemblies seated within the reactor pressure vessel during movement of fuel assemblies within the reactor pressure vessel while in MODE 6 ensures that sufficient water depth is available to remove 99% of the assumed iodine gap activity released from the rupture of an irradiated fuel assembly.

Further, sufficient time is available to close the personnel air lock following a loss of shutdown cooling before boiling occurs.

Regarding LCO 3.9.4.c, the phrase "atmosphere outside containment" refers to anywhere outside the containment vessel, including (but not limited to) the containment annulus and the auxiliary building. The LCO requirements are referred to as "containment closure" rather than "containment OPERABILITY." Since containment pressurization is not expected, the 10 CFR 50 Appendix J leakage criteria and tests are not required. The LCO is modified by a Note allowing penetration flow paths with direct access from the containment to the outside atmosphere to be open under administrative controls.

Admistrative controls ensure that 1) appropriate personnel are aware of the open status of the penetration flow path during CORE ALTERATIONS or movement of irradiated fuel assemblies within the containment, and 2) specified individuals are designated and readily available to isolate the flow path in the event of a Fuel Handling Accident.

The containment equipment hatch cover may be off during CORE ALTERATIONS or movement of irradiated fuel in containment provided the requirements of Specification 3.9.12 are satisfied. The requirements of Specification 3.9.12 ensure that the emergency ventilation system servicing the storage area is OPERABLE with the ability to filter any radioactive release through the containment equipment hatch following a fuel handling accident. Since containment closure is not credited for mitigating the consequences of the fuel handling accident as described in the Updated Safety Analysis Report, the equipment hatch cover need not be installed to ensure adequate protection of the public health or safety.

DAVIS-BESSE, UNIT 1 B 3/4 9-2 Amendment 202, 251 Revised by NRC letter dated March 19, 1998

Docket Number 50-346 License Number NPF-3 Serial Number 3301 - Attachment 3 PROPOSED RETYPED TECHNICAL SPECIFICATION PAGES (2 pages follow)

REFUELING OPERATIONS CONTAINMENT PENETRATIONS LIMITING CONDITION FOR OPERATION 3.9.4 The containment penetrations shall be in the following status:

a. The equipment hatch cover closed and held in place by a minimum of four bolts, except the equipment hatch may be open provided the requirements of Specification 3.9.12 are satisfied,

b. A minimum of one door in each air lock closed, but both doors of the containment personnel air lock may be open provided that at least one personnel air lock door is capable of being closed and a designated individual is available immediately outside the personnel air lock to close the door, and

c. Each penetration providing direct access from the containment atmosphere to the atmosphere outside containment shall be either:

1. Closed by a manual or automatic isolation valve, blind flange, or equivalent, or

2. Be capable of being closed from the control room by an OPERABLE containment purge and exhaust valve upon receipt of a higher radiation signal from the containment purge and exhaust system noble gas monitor.

NOTE Penetration flow path(s) providing direct access from the containment atmosphere to the outside atmosphere may be unisolated under administrative controls.

APPLICABILITY: During CORE ALTERATIONS or movement of irradiated fuel within the containment.

ACTION:

a. With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or movement of irradiated fuel in the containment.

b. With the requirements of Specification 3.9.4.c not satisfied for the containment purge and exhaust system, close at least one of the isolation valves for each of the purge and exhaust penetrations providing direct access from the containment atmosphere to the outside atmosphere within one hour.

c. The provisions of Specification 3.0.3 are not applicable.

DAVIS-BESSE, UNIT 1 3/4 9-4 Amendment No. 186, 202, 221, 251

REFUELING OPERATIONSONLY CONTAINMENT PENETRATIONS SURVEILLANCE REOUIREMENTS 4.9.4 Each of the above required containment penetrations shall be determined to be either in its required condition or capable of being closed by an OPERABLE containment purge and exhaust valve, within 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to the start of and at least once per 7 days during CORE ALTERATIONS or movement of irradiated fuel in the containment, by:

a. Verifying the penetrations are in their required condition, or

b. Verifying that with the containment purge and exhaust system in operation, and the containment purge and exhaust system noble gas monitor capable of providing a high radiation signal to the control room, that after initiation of the high radiation signal, the containment purge and exhaust isolation valves can be closed from the control room.

DAVIS-BESSE, UNIT I 3/4 9-5 Amendment No. 186, 202, 224

Docket Number 50-346 License Number NPF-3 Serial Number 3301 COMMITMENT LIST THE FOLLOWING LIST IDENTIFIES THOSE ACTIONS COMMITTED TO BY THE DAVIS-BESSE NUCLEAR POWER STATION (DBNPS) IN THIS DOCUMENT. ANY OTHER ACTIONS DISCUSSED IN THE SUBMITTAL REPRESENT INTENDED OR PLANNED ACTIONS BY THE DBNPS. THEY ARE DESCRIBED ONLY FOR INFORMATION AND ARE NOT REGULATORY COMMITMENTS. PLEASE NOTIFY THE SUPERVISOR - FLEET LICENSING (330-315-6944) OF ANY QUESTIONS REGARDING THIS DOCUMENT OR ANY ASSOCIATED REGULATORY COMMITMENTS.

COMMITMENTS DUE DATE As part of implementation of the proposed amendment, applicable procedures will be revised to permit opening the affected penetrations during fuel movement as long as administrative controls are in place to ensure prompt closure of the affected penetration in the event of a Fuel Handling Accident. These controls will entail assignment of a designated individual who will be readily available to isolate the flow path in the event of a Fuel Handling Accident. This individual may be a part of the team assigned to perform local leak rate testing of the associated penetration.

Concurrent with implementation.