ML050340223

License Amendments 228 and 210 Pressurizer Enclosure Hatch

ML050340223
Person / Time
Site:	Mcguire, McGuire
Issue date:	04/05/2005
From:	James Shea NRC/NRR/DLPM/LPD2
To:	Gordon Peterson Duke Energy Corp
SHea J, 415-1388, NRR/DLPM
Shared Package
ML050990004	List: ML050340223 ML050960423
References
TAC MB9523, TAC MB9524
Download: ML050340223 (16)

Text

April 5, 2005 G. R. Peterson, Vice President McGuire Nuclear Station Duke Energy Corporation 12700 Hagers Ferry Road Huntersville, NC 28078

SUBJECT:

MCGUIRE NUCLEAR STATION, UNITS 1 AND 2 RE: ISSUANCE OF AMENDMENTS (TAC NOS. MB9523 AND MB9524)

Dear Mr. Peterson:

The Nuclear Regulatory Commission has issued the enclosed Amendment No. 228 to Renewed Facility Operating License NPF-9 and Amendment No. 210 to Renewed Facility Operating License NPF-17 for the McGuire Nuclear Station, Units 1 and 2. The amendments consist of changes to the Technical Specifications (TSs) in response to your application dated June 3, 2003, as supplemented by letter dated July 29 and December 7, 2004, and January 18, 2005.

The amendments revise TS 3.6.14 to allow a pressurizer enclosure hatch between the upper and lower containment volumes to be open for up to 6 hours0.25 days <br />0.0357 weeks <br />0.00822 months <br /> to facilitate inspections of components such as the powered-operated relief valve block valves.

A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely,

/RA/

James J. Shea, Project Manager, Section 1 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-369 and 50-370

Enclosures:

1. Amendment No. 228 to NPF-9

2. Amendment No. 210 to NPF-17

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

G. R. Peterson, Vice President McGuire Nuclear Station Duke Energy Corporation 12700 Hagers Ferry Road Huntersville, NC 28078

SUBJECT:

MCGUIRE NUCLEAR STATION, UNITS 1 AND 2 RE: ISSUANCE OF AMENDMENTS (TAC NOS. MB9523 AND MB9524)

Dear Mr. Peterson:

The Nuclear Regulatory Commission has issued the enclosed Amendment No. 228 to Renewed Facility Operating License NPF-9 and Amendment No. 210 to Renewed Facility Operating License NPF-17 for the McGuire Nuclear Station, Units 1 and 2. The amendments consist of changes to the Technical Specifications (TSs) in response to your application dated June 3, 2003, as supplemented by letter dated July 29 and December 7, 2004, and January 18, 2005.

The amendments revise TS 3.6.14 to allow a pressurizer enclosure hatch between the upper and lower containment volumes to be open for up to 6 hours0.25 days <br />0.0357 weeks <br />0.00822 months <br /> to facilitate inspections of components such as the powered-operated relief valve block valves.

A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely,

/RA/

James J. Shea, Project Manager, Section 1 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-369 and 50-370 DISTRIBUTION:

PUBLIC

Enclosures:

PDII-1 R/F

1. Amendment No. 228 to NPF-9 RidsNrrDlpmLpdii (EHackett)

2. Amendment No. 210 to NPF-17 RidsNrrDlpmLpdii1 (JNakoski)

3. Safety Evaluation RidsRgn2MailCenter (MErnstes)

RidsAcrsAcnwMailCenter RidsNrrPMJShea RidsOgcRp cc w/encls: See next page RidsAcrsAcnwMailCenter DlpmDpr CHawes (Hard Copy)

Package: ML050990004 Amendment: ML050340223 Tech Spec Pages: ML050960423 NRR-058 OFFICE PDII-1/PM PDII-1/LA SPLB-A(A) SPSB-C IROB-A OGC PDII-1/SC NAME JShea CHawes SJones RDennig TBoyce DCummings JNakoski DATE 03/16/05 03/22 /05 03/16/05 03/25/05 03/25/05 03/24/05 03/ /05 OFFICIAL AGENCY RECORD

DUKE ENERGY CORPORATION DOCKET NO. 50-369 MCGUIRE NUCLEAR STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 228 Renewed License No. NPF-9

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

A. The application for amendment to the McGuire Nuclear Station, Unit 1 (the facility),

Renewed Facility Operating License No. NPF-9 filed by the Duke Energy Corporation (licensee) dated June 3, 2003, as supplemented by letter dated July 29 and December 7, 2004, and January 18, 2005, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-9 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 228, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of its date of issuance and shall be implemented within 30 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

John A. Nakoski, Chief, Section 1 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation

Attachment:

Technical Specification Changes Date of Issuance: April 5, 2005

DUKE ENERGY CORPORATION DOCKET NO. 50-370 MCGUIRE NUCLEAR STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 210 Renewed License No. NPF-17

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

A. The application for amendment to the McGuire Nuclear Station, Unit 1 (the facility),

Renewed Facility Operating License No. NPF-9 filed by the Duke Energy Corporation (licensee) dated June 3, 2003, as supplemented by letter dated July 29 and December 7, 2004, and January 18, 2005, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-17 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 210, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of its date of issuance and shall be implemented within 30 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

John A. Nakoski, Chief, Section 1 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation

Attachment:

Technical Specification Changes Date of Issuance: April 5, 2005

ATTACHMENT TO LICENSE AMENDMENT NO. 228 RENEWED FACILITY OPERATING LICENSE NO. NPF-9 DOCKET NO. 50-369 AND LICENSE AMENDMENT NO. 210 RENEWED FACILITY OPERATING LICENSE NO. NPF-17 DOCKET NO. 50-370 Replace the following page of the Appendix A Technical Specification and associated Bases with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert 3.6.14-1 3.6.-14-1 B3.6.14-1 B3.6.14-1 thru thru B3.6.14-5 B3.6.14-6

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 228 TO RENEWED FACILITY OPERATING LICENSE NPF-9 AND AMENDMENT NO. 210 TO RENEWED FACILITY OPERATING LICENSE NPF-17 DUKE ENERGY CORPORATION MCGUIRE NUCLEAR STATION, UNITS 1 AND 2 DOCKET NOS. 50-369 AND 50-370

1.0 INTRODUCTION

By letter to the Nuclear Regulatory Commission (NRC or the Commission) dated June 3, 2003, as supplemented by letters dated July 29 and December 7, 2004, and January 18, 2005, Duke Energy Corporation (Duke or the licensee), submitted a request for changes to the McGuire Nuclear Station (McGuire), Units 1 and 2, Technical Specifications (TSs). The requested changes would revise TS 3.6.14 to allow a pressurizer enclosure hatch to be open for up to 6 hours0.25 days <br />0.0357 weeks <br />0.00822 months <br />. The increase from the present 1-hour allowance would facilitate inspections of components such as the power-operated relief valve (PORV) block valves. The July 29, 2004, letter provided additional information requested by the NRC staff to clarify the type of work to be performed during hatch removal and to provide clarification on the licensees heavy load analysis. Additional information to support the licensees heavy load analyses was supplied in supplemental submittals dated December 7, 2004, and January 18, 2005. This clarifying information did not change the initial proposed no significant hazards consideration determination.

2.0 REGULATORY EVALUATION

Section 182a of the Atomic Energy Act of 1954, as amended (the Act) requires applicants for nuclear power plant operating licenses to include TSs as a part of the license. The TSs ensure the operational capability of structures, systems and components that are required to protect the health and safety of the public. The NRCs regulatory requirements that are related to the content of the TSs are contained in Title 10 of the Code of Federal Regulations (10 CFR)

Section 50.36. Section 50.36 of 10 CFR requires that the TSs include items in the following specific categories: (1) safety limits, limiting safety system settings, and limiting control settings (50.36(c)(1)); (2) limiting conditions for operation (50.36(c)(2)); (3) surveillance requirements (SRs) (50.36(c)(3)); (4) design features (50.36(c)(4)); and (5) administrative controls (50.36(c)(5)).

Pursuant to 10 CFR 50.90, a licensee may apply for an amendment to its license, including the TSs incorporated into the license. In determining the acceptability of the proposed changes,

the NRC staff interprets the requirements of the current version of 10 CFR 50.36. Within this general framework, licensees may revise the current TSs provided that a plant-specific review supports a finding of continued adequate safety whereas: (1) the change is editorial, administrative, or produces clarification (i.e., no requirements are materially altered); (2) the change is more restrictive than the licensees current requirement; or (3) the change is less restrictive than the licensees current requirement, but continues to afford adequate assurance of safety when judged against current regulatory standards.

In NUREG-0612, Control of Heavy Loads at Nuclear Power Plants, the NRC staff provided regulatory guidelines for control of heavy load lifts to assure safe handling of heavy loads in areas where a load drop could impact on stored spent fuel, fuel in the reactor core, or equipment that may be required to achieve safe shutdown or permit continued decay heat removal. In a letter dated December 22, 1980, as supplemented by Generic Letter (GL) 81-07, Control of Heavy Loads, dated February 3, 1981, the NRC requested that all licensees describe the extent that the guidelines of NUREG-0612 were satisfied at each facility and what additional modifications would be necessary to fully satisfy the guidelines. The responses to these GLs established a heavy loads control program, and changes to this program are subject to review pursuant to 10 CFR 50.59.

3.0 TECHNICAL EVALUATION

The NRC staff has reviewed the licensee's regulatory and technical analyses in support of its proposed license amendment. The evaluation below will support the conclusion that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

3.1 Ice Condenser Containment The ice condenser containment at McGuire Nuclear Station is divided in three major compartments: the upper compartment, the lower compartment, and the ice condenser compartment. A high-energy line break inside the containment, such as a loss-of-coolant accident (LOCA) or main steamline break, would occur in the lower compartment. The resultant high pressure would be relieved into the ice condenser compartment and then into the upper compartment. Air and steam would be cooled and the steam condensed by the ice bed prior to entering the upper compartment. The action of the ice keeps the peak containment pressure during an accident relatively low (less than 15 psig). If steam should pass directly from the lower to upper compartments, bypassing the ice condenser, it would not be condensed by the ice as designed. Too much steam bypass could cause the peak containment pressure to exceed the containment design capability. Therefore, the ice condenser bypass area (the total area of openings between the upper and lower compartments) must not exceed a certain limit during plant operation. McGuire is designed with necessary openings between the upper and lower compartments, such as the refueling canal drains that have a total area of 4.6 square feet for Unit 1 and 5 square feet for Unit 2.

The boundary between the upper and lower compartments is called the divider barrier or operating deck. It is for the most part horizontal and flat, but the tops of the steam generators (SGs) and the pressurizer is considerably higher than the main part of the operating deck. The

divider barrier bulges upward to cover the SGs and pressurizer and enclose these upper parts in doghouses or enclosures that essentially conform to the shapes of the components. This places the components in the lower compartment, to assure that any high-energy break occurs in the lower compartment, directing the released steam and heat through the ice condenser.

There are four hatches in the pressurizer enclosure to facilitate access to its interior from the upper compartment. Normally this is done only during a plant shutdown, but the licensee has at times found it necessary to access the pressurizer enclosure during plant operation. Situations arise where it is necessary to enter the pressurizer enclosure to perform inspections and maintenance. It is not feasible to enter the upper part of the pressurizer enclosure from the containment lower compartment while the plant is operating, because of heat and radiation hazards and the cramped spaces that one would have to pass. Thus, opening a pressurizer hatch is the only practical way to access the enclosure at power.

The pressurizer enclosure hatch is an opening between the upper and lower containment compartments. There is one pressurizer enclosure hatch at McGuire, Unit 1 and there are three at McGuire, Unit 2; the largest one has an area of 7.5 square feet.

3.2 Purpose for Amendment During plant operation, situations arise where it is necessary to enter the pressurizer cavity to perform inspections and maintenance. Entries into the pressurizer cavity are made during startup and shutdown to check for leaks. During operation, if a leak is suspected in the pressurizer cavity, it may be necessary to open the pressurizer enclosure hatch to perform an inspection and, if needed, repair. If repairs are to be made, or the inspections are time consuming, the pressurizer hatch would need to be open for longer than the TSs 3.6.14 allowed access time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Allowing up to 6 hours0.25 days <br />0.0357 weeks <br />0.00822 months <br /> would enhance personnel safety and radiation safety since access to the pressurizer cavity can be made from above instead of below. In June 1992, Catawba Nuclear Station was granted an amendment to increase the hatch allowed opening time to 6 hours0.25 days <br />0.0357 weeks <br />0.00822 months <br /> prompted by PORV valve stem problems that required periodic monitoring. More specific reasons for accessing the pressurizer enclosure were made in a supplemental letter submitted by the licensee on July 29, 2004, in response to additional information requested by the NRC staff.

Unplanned entries into the pressurizer enclosure include the following situations:

1) Upon entering Mode 3 at the beginning of every refueling outage, an inspection of the packing leakoff lines on the PORVs and the PORV block valves is performed. While in the cavity, a general visual inspection for any type of leak or other problem is performed. This inspection must be performed in Mode 3 because it may not be possible to detect leakage when the unit is cooled down and depressurized.

2) An inspection similar to the one described above is performed during Mode 3 following every refueling outage. The PORVs and PORV block valves are inspected for packing leakage.

3) If any valve work (seat, bonnet, packing, or removal) was performed during the refueling outage on the PORVs, block valves, or safety valves, a functional test

for external leakage must be performed at full temperature and pressure. The necessity to perform these activities usually occurs during Mode 2.

4) Surveillance is performed upon entering Mode 3 at the start of a refueling outage, or after a trip following a long run in order to check for boric acid leakages. Duke inspects all areas of containment, including the upper pressurizer enclosure, for signs of boric acid corrosion from leaks.

There are also several reasons why unplanned entries into the pressurizer enclosure would be made during Mode 1 operations. These reasons include the following:

1) Suspected instrument tubing leak affecting pressurizer level indication.

2) Confirm pressurizer safety valve(s) leakage.

3) Confirm pressurizer PORV seat leakage.

4) Investigate pressurizer safety valve relief line high temperature.

Since there is an ongoing need to enter the pressurizer cavity for more than 1-hour, and since, the licensee submits, removing a pressurizer enclosure hatch does not have a significant effect on safety, the licensee has requested that TS 3.6.14 be modified to allow a hatch to be open for up to 6 hours0.25 days <br />0.0357 weeks <br />0.00822 months <br />.

3.3 Safety Considerations The two safety concerns with opening the pressurizer hatch are as follows: the effect of increased bypass area on the containment pressure and temperature transient during an accident, and the possibility of dropping a hatch while lifting. These will be addressed in turn.

3.3.1 Increased Containment Bypass Area An ice condenser containment pressure during a LOCA has two peaks: one approximately at the end of the blowdown of the reactor coolant system (sometimes called the peak compression pressure), and the second sometime after all the ice melts (called the long-term peak pressure).

The long-term peak is usually higher than the compression peak.

Westinghouse analyzed the effects of divider barrier leakages for bypass areas of up to 50 square feet, as part of the original plant design. The results are presented in Final Safety Analysis Report (FSAR) Section 6.2.1.1.3.1, Table 6-20, and Figure 6-22. The results of this analysis show that the pressure peaks are below the design pressure. In fact, the NRC staff found, in its SER for McGuire (NUREG-0422, dated March 1978), the following: The applicant has also provided analyses that indicate that about 40 square feet of bypass area can be accommodated in the design without the design pressure of the containment being exceeded.

The licensee has provided more specific calculations to support the TS change request, discussed below.

The calculation of the new peak compression pressure consists of an extrapolation of Westinghouse results found in the McGuire FSAR, Section 6.2.1.1.3.1 (Loss of Coolant

Accident). The compression peak pressure during the blowdown phase of the accident was calculated by Westinghouse to be 7.8 psig that includes 0.4 psi for the effect of the operating deck bypass area that is assumed to be 5 square feet.

The effect of the potential deck leakage is expressed by the following equation that was derived by Westinghouse based on the Waltz Mill test results:

P = Bypass Flow Area x 0.080 Substituting the additional area of 7.5 square feet resulting from the open pressurizer hatch, in the above equation, the following increase in peak pressure is obtained:

P = 7.5 square feet x 0.080 = 0.6 psi Hence, the new compression pressure is 8.4 psig, that is well below the design internal pressure for containment of 15 psig and below the TSs acceptance criteria of 14.8 psig (TS Bases 3.6.2).

The open pressurizer hatch will not increase the long-term containment peak pressure of 13.21 psig (USFAR Section 6.2.1.1.3.1). The bypassed steam is condensed by the containment spray in the upper compartment, instead of by the ice, resulting in lower decay heat at the time the ice is fully depleted. The limiting case for containment temperature is a steamline break (SLB) with the peak occurring in the lower containment. Additional bypass area would result in a lower temperature peak, by directing part of the steam into the upper containment. However, the upper containment temperature is not a concern, since it is 150 to 200°F below the peak in lower containment (FSAR Figure 6-24 through 6-25). The upper containment pressure from a SLB is bounded by the peak containment LOCA pressure.

Based on the foregoing, the NRC staff finds that the removal of the pressurizer hatch for the purpose of performing work would not result in exceeding the containment design pressure should a LOCA occur while the hatch is removed.

3.3.2 Possible Dropping of the Hatch The second area of concern is the removal of the pressurizer enclosure hatch during Modes 1-4. Section 5.1.3 of NUREG-0612, provides guidelines for the control of heavy loads within pressurized-water reactor buildings. One acceptable method is to demonstrate that the consequences of a load drop will neither challenge the integrity of the reactor coolant pressure boundary nor result in the failure of redundant trains of safe-shutdown equipment. Appendix A to NUREG-0612 describes recommended assumptions for evaluation of the consequences of load drops.

In response to the NRCs letter regarding the control of heavy loads at nuclear power plants dated December 22, 1980, the licensee stated that no heavy loads would be lifted inside containment during Modes 1-4. Subsequently, the licensee identified a need to access valves and other components within the pressurizer enclosure during operation at power. The licensee performed a calculation to ensure that a drop of the largest pressurizer hatch plug on the pressurizer enclosure roof or operating floor, and a drop of the polar crane load block onto the operating floor, would not damage any equipment, components, or systems necessary for

safe shutdown. Based on this calculation, the licensee stated that the operating floor and the pressurizer enclosure roof can withstand a drop from the highest potential lift height of either the largest pressurizer hatch plug using the auxiliary hook or the polar crane load block. This evaluation supported a change to the licensees heavy load program that was implemented pursuant to the requirements of 10 CFR 50.59 to allow operation of the polar crane in operating Modes 1-4. The licensee informed the NRC staff of this change to the heavy load handling program at McGuire in response to NRC Bulletin 96-02, "Movement of Heavy Loads Over Spent Fuel, Over Fuel in the Reactor Core, or Over Safety-Related Equipment," by letter dated May 13, 1996. Because the change was implemented through 10 CFR 50.59, the staff neither has formally accepted this change to the design basis of the facility nor reviewed the supporting evaluation. However, by letter dated March 27, 1990, the staff accepted a change in the heavy load program at the similar Catawba Nuclear Station to allow operation of the polar crane in Modes 1 through 4 based on an evaluation of load block and pressurizer hatch drops on the operating floor.

In support of the increased time allowed for hatch removal, the licensee revised the heavy load drop analysis to ensure the calculation had enveloped the case of the largest pressurizer hatch plug dropping over the hatch plug opening. This additional heavy load drop case that was analyzed involved dropping the plug precisely back into the hole just as it was being lifted, and evaluating the effect of a subsequent drop of the auxiliary hook load block on the hatch plug. In supplemental letters submitted by the licensee on July 29 and December 7, 2004, in response to additional information requested by the NRC staff regarding the heavy load analysis, the licensee provided a report that summarized the analyses performed to support the safe handling of a pressurizer hatch during plant operation. The licensee described two calculations that were performed to ensure that a drop of the largest pressurizer hatch plug back into the hatch opening combined with a drop of the auxiliary hook onto the plug would not cause unacceptable damage to the hatch or hatch opening.

The pressurizer hatch plugs are constructed of concrete reinforced with reinforcing steel bars, nelson studs, steel frames, and steel plate. The center of the largest hatch plug is 2 feet thick, and both the seating ledge of the hatch plug and the seating ledge of the hatch opening are about 1 foot thick. The licensee stated that the hatch plug cannot fit through the opening in any of several orientations described and that a small (i.e., 3 degree) horizontal rotation of the hatch plug will cause the seating ledges to engage with the thickest parts of the hatch plug and pressurizer enclosure.

Important recommended assumptions from Appendix A to NUREG-0612 for this load drop scenario include: the load is dropped from the highest possible lift height, the load falls in the orientation with the most severe potential consequences, and the hook, rigging and load fall simultaneously. The licensee load drop analysis assumes that the hatch plug is dropped from a height of 1 foot above the hatch opening based on administrative limits on lift height. This limits the maximum drop distance to 3 feet 2 inches that assumes the plug fits precisely back into the opening. If the hatch plug were not oriented in this position, the maximum drop distance would decrease and, for most orientations, thicker and stronger portions of the hatch and hatch opening would absorb the energy of the drop.

If one sling breaks and the hatch rotated 90 degrees about one horizontal axis and then the second sling breaks, the hatch could potentially fit into the enclosure opening. The maximum drop height before the edge of the hatch makes contact with the enclosure is 1 foot. This would

be about 1/3 of the energy from a full 3 foot drop as assumed in the worst case analysis as described above. Due to the thickness of the hatch, the energy would be transferred to a significant section of the enclosure opening ledge, and the licensee concluded that the damage would be minor. Therefore, the licensee found the evaluation of the larger energy release associated with the drop of the hatch plug precisely back into the opening to be conservative.

In the case of a run away lift, while lifting the pressurizer hatch plug, the licensee has incorporated NUREG-0612 Phase-1 requirements. The licensee credits the crane preventive maintenance program, controlling procedures, and operator training to respond to any unplanned movement of the crane. In any lift using the polar crane/auxiliary hoist, the operator can stop unexpected movement by several means. The operator can reverse the controls, hit the emergency stop button, or release the dead-man foot switch. Finally, the crane is equipped with upper lift limits on the hoist, that should stop upward movement to prevent two blocking.

In a supplemental letter submitted by the licensee on January 18, 2005, in response to additional information requested by the NRC staff regarding the crane safety features, the licensee provided a description of the crane control circuitry redundancy features. The dead-man foot switch when released, removes power from the crane via the crane motor contactors.

The dead-man foot switch powers the motor contactor coil that must remain energized to maintain power to the crane. Upon release of the dead man foot switch, all crane motions are suspended. Depressing the reset button is required before power can be restored to the crane through the dead-man switch. Independent of the dead-man switch circuitry, the emergency stop button on the main control removes power from the crane via a shunt trip of the main circuit breaker. Both these switches are functionally verified during pre-operational inspections as part of the Duke administrative procedures for operation of the polar crane. Considering the limited time allowed for removal of the hatch by TSs and the resulting attention to the evolution, the procedural controls and methods available for stopping the crane provide reasonable assurance that the hatch plug will not be raised significantly more than 1-foot above the pressurizer enclosure either due to operator inattention or failure of the crane control system.

Therefore, the assumed drop height used in the load drop analysis for the hatch plug is acceptable.

The second drop analysis involving the auxiliary load block assumed a drop from the highest potential drop height. This drop height assumption is consistent with the potential of two-blocking of the unloaded load block and the guideline assumptions of Appendix A to NUREG-0612. As a result of the significantly greater load drop height, the postulated drop of the approximately 1500 pound auxiliary load block is the more limiting drop in terms of energy that must be absorbed by the hatch plug and the hatch plug opening ledges.

For both load drops, the licensee stated that conservative assumptions were used in the structural evaluation of the hatch plug and ledges. In evaluating the energy absorption capability, the licensee considered only the energy necessary to yield most of the reinforcing steel bars within the hatch plug and the hatch opening ledges. The analysis discounted the additional strength provided by the concrete itself and other reinforcement provided by Nelson studs, steel framing, and steel plate. The results of the analysis indicate that substantial margin to failure is maintained. The margin to failure and the conservative assumptions regarding the materials absorbing the energy of the drop overcome uncertainties regarding the methodology.

These considerations, the clearly robust construction of the hatch plug and the pressurizer

enclosure, and the configuration of the hatch and hatch opening provide adequate assurance that a credible drop of the hatch plug would neither challenge the integrity of the reactor coolant pressure boundary nor result in the failure of redundant trains of safe-shutdown equipment.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the North Carolina State official was notified of the proposed issuance of the amendments. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts and no significant change in the types of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (68 FR 43383). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributor: J. Shea Date: April 5, 2005

McGuire Nuclear Station cc:

Ms. Lisa F. Vaughn Ms. Karen E. Long Duke Energy Corporation Assistant Attorney General Mail Code - PB06E North Carolina Department of Justice 422 South Church Street P. O. Box 629 P.O. Box 1244 Raleigh, North Carolina 27602 Charlotte, North Carolina 28201-1244 Mr. R. L. Gill, Jr.

County Manager of Manager - Nuclear Regulatory Issues Mecklenburg County and Industry Affairs 720 East Fourth Street Duke Energy Corporation Charlotte, North Carolina 28202 526 South Church Street Mail Stop EC05P Mr. C. Jeffrey Thomas Charlotte, North Carolina 28202 Regulatory Compliance Manager Duke Energy Corporation NCEM REP Program Manager McGuire Nuclear Site 4713 Mail Service Center 12700 Hagers Ferry Road Raleigh, North Carolina 27699-4713 Huntersville, North Carolina 28078 Mr. Richard M. Fry, Director Anne Cottingham, Esquire Division of Radiation Protection Winston and Strawn North Carolina Department of 1400 L Street, NW. Environment, Health and Natural Washington, DC 20005 Resources 3825 Barrett Drive Senior Resident Inspector Raleigh, North Carolina 27609-7721 c/o U.S. Nuclear Regulatory Commission 12700 Hagers Ferry Road Mr. T. Richard Puryear Huntersville, North Carolina 28078 Owners Group (NCEMC)

Duke Energy Corporation Dr. John M. Barry 4800 Concord Road Mecklenburg County York, South Carolina 29745 Department of Environmental Protection 700 N. Tryon Street Charlotte, North Carolina 28202 Mr. Peter R. Harden, IV VP-Customer Relations and Sales Westinghouse Electric Company 6000 Fairview Road, 12th Floor Charlotte, North Carolina 28210