ML052850005

Revision to Safety Evaluation for Amendments No. 220 and 226, Dated September 23, 2005

ML052850005
Person / Time
Site:	Point Beach
Issue date:	01/12/2006
From:	Lyon C Plant Licensing Branch III-2
To:	Koehl D Nuclear Management Co
Muniz A, NRR/DLPM, 415-4093
References
TAC MC7650, TAC MC7651
Download: ML052850005 (7)

Text

January 12, 2006 Mr. Dennis L. Koehl Site Vice President, Point Beach Nuclear Management Company, LLC 6610 Nuclear Road Two Rivers, WI 54241-9516

SUBJECT:

POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 - REVISION TO SAFETY EVALUATION FOR AMENDMENT NOS. 220 AND 226 (TAC NOS. MC7650 AND MC7651)

Dear Mr. Koehl:

On September 23, 2005, the Nuclear Regulatory Commission (NRC) issued Amendment No. 220 to Facility Operating License No. DPR-24 and Amendment No. 226 to Facility Operating License No. DPR-27 for the Point Beach Nuclear Plant (PBNP), Units 1 and 2, in response to your application dated July 24, 2005, incorporating a PBNP, Unit 1 reactor vessel head (RVH) drop accident analysis into the PBNP Final Safety Analysis Report and revising the PBNP, Unit 2 RVH drop accident analysis. This letter transmits a revision to the NRC staffs safety evaluation (SE) associated with these amendments. describes the revisions to the SE. Copies of the revised SE pages are included in . The revisions do not change the conclusions of the original NRC staffs SE.

If there are any questions concerning this matter, please call me at (301) 415-4018.

Sincerely,

/RA/

Carl F. Lyon, Project Manager Plant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-266 and 50-301

Enclosures:

As stated cc w/encls.: See next page

ML052850005 *Previously concurred OFFICE LPLIII-1/PM LPLIII-1/PM LPLIII-1/LA BC/AADB OGC LPLIII-1/BC NAME AMuniz FLyon DClarke* MKotzalas* AHodgdon* TKobetz*

DATE 1/12/06 1/12/06 11/8/05 12/27/05 1/9/06 1/11/06 Point Beach Nuclear Plant, Units 1 and 2 cc:

Jonathan Rogoff, Esquire Mr. Jeffery Kitsembel Vice President, Counsel & Secretary Electric Division Nuclear Management Company, LLC Public Service Commission of Wisconsin 700 First Street P.O. Box 7854 Hudson, WI 54016 Madison, WI 53707-7854 Mr. F. D. Kuester Nuclear Asset Manager President & Chief Executive Officer Wisconsin Electric Power Company WE Generation 231 West Michigan Street 231 West Michigan Street Milwaukee, WI 53201 Milwaukee, WI 53201 Douglas E. Cooper Regulatory Affairs Manager Senior Vice President - Group Operations Point Beach Nuclear Plant Palisades Nuclear Plant Nuclear Management Company, LLC Nuclear Management Company, LLC 6610 Nuclear Road 27780 Blue Star Memorial Highway Two Rivers, WI 54241 Covert, MI 49043 Mr. Ken Duveneck Site Director of Operations Town Chairman Nuclear Management Company, LLC Town of Two Creeks 6610 Nuclear Road 13017 State Highway 42 Two Rivers, WI 54241 Mishicot, WI 54228 Chairman Public Service Commission of Wisconsin P.O. Box 7854 Madison, WI 53707-7854 Regional Administrator, Region III U.S. Nuclear Regulatory Commission 801 Warrenville Road Lisle, IL 60532-4351 Resident Inspector's Office U.S. Nuclear Regulatory Commission 6612 Nuclear Road Two Rivers, WI 54241

POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 CORRECTION TO THE NRC STAFFS SAFETY EVALUATION FOR AMENDMENTS NO. 220 AND 226 REASON FOR THE PAGE # ORIGINAL TEXT REVISED TEXT CHANGE

... factor of 8.8 and multiplying by the sump volume scaling factor of 0.81 to account for the change in volume from 197,000 13 ... factor of 8.8. gallons to 243,000 gallons. Clarify scaling factor

... results by 8.8 and multiplying

... results by 8.8 as discussed by the sump volume scaling 13 above ... factor as discussed above ... Clarify scaling factor 14 Unit 1 shall be operable Units 1 and 2 shall be operable Typographical error Enclosure 1

ENCLOSURE 2: REVISED SE PAGES sump fluid volume. The LOCA ECCS leakage pathway dose results reported in the PBNP FSAR were adjusted by this scaling factor, by dividing the values by the limiting I-131 factor of 8.8 and multiplying by the sump volume scaling factor of 0.81 to account for the change in volume from 197,000 gallons to 243,000 gallons. The licensees dose results at the exclusion area boundary are 5.3 rem thyroid and 0.022 rem whole body, and at the low population zone they are 3.4 rem thyroid and 0.006 rem whole body. The offsite dose results are well within the dose criteria in 10 CFR Part 100, i.e., they are within 75 rem thyroid and 6 rem whole body.

The PBNP FSAR LOCA analysis does not bound the recent results of control room envelope unfiltered inleakage tracer gas testing. Additionally, the control room analysis assumed an ECCS leakage rate half that assumed for the offsite dose analysis. NMC included the impact on the control room dose results of (1) increasing the assumed unfiltered inleakage from 10 cubic feet per minute (cfm) to 100 cfm to account for the testing results, and (2) increasing the ECCS leakage rate from 400 cubic centimeters per minute (cc/min) to 800 cc/min.

Considering these changes to the control room dose analysis assumptions, the licensee showed that the control room dose, estimated by dividing the FSAR LOCA results by 8.8 and multiplying by the sump volume scaling factor as discussed above, increases by a factor of 2.7.

The licensees calculated control room doses for the postulated RVH drop are 26.5 rem thyroid and 0.04 rem whole body. These are within the General Design Criteria-19, Control Room Habitability System, dose criteria of 5 rem whole body or its equivalent to any part of the body, given as 30 rem thyroid in SRP 6.4, Control Room.

3.5 Regulatory Commitments The licensee made the following commitments in a letter dated July 24, 2005:

1. The reactor has been shutdown for greater than 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />.

2. A Senior Reactor Operator will be stationed in containment during RVH lift activities and will have communications capability with the control room.

3. The containment sump screen shall be installed and the flowpath for aligning RHR pump suction to the containment sump is available.

4. A minimum borated water volume of 243,000 gallons shall be available for sump recirculation.

5. The containment equipment hatch will be on and bolted. Both personnel airlock door interlocks will be functional to ensure one door in each airlock is closed.

6. Containment purge supply and exhaust fans are off and associated containment isolation valves are closed when the RVH is suspended greater than 24 inches over the reactor vessel flange.

7. Other containment penetrations that allow containment atmosphere to communicate with the environment or the Primary Auxiliary Building atmosphere shall be closed.

8. The maximum allowable lift height for the RVH (i.e., 26.4 feet above the reactor vessel flange when over the fuel) shall not be exceeded.

Revised by letter dated January 12, 2006

9. Both SI trains shall be available.

10 Both RHR trains shall be operable.

11. Technical Specification Limiting Condition for Operation (LCO) 3.7.9, "Control Room Emergency Filtration System (CREFS)," and LCO 3.3.5, "CREFS Actuation Instrumentation," shall be met.

12. One standby emergency power source capable of supplying each 4.16 kV/480 V Class 1E safeguards bus on PBNP, Units 1 and 2 shall be operable.

13. The licensee will incorporate an analysis of the RVH drop into the PBNP FSAR.

14. The licensee will incorporate the PBNP method of NUREG-0612 Phase I compliance into the PBNP FSAR.

15. The Programmed and Remote reactor vessel inservice inspection device will not be lifted over a core containing fuel assemblies.

4.0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION

DETERMINATION The Commissions regulation at 10 CFR 50.92(c) states that the Commission may make a final determination that a license amendment involves no significant hazards consideration if 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) result in a significant reduction in a margin of safety. The NRC staff has made a final determination that no significant hazards consideration is involved for the proposed amendment and that the amendment should be issued as allowed by the criteria contained in 10 CFR 50.91. The NRC staffs final determination is presented below:

1. Would the proposed amendment involve a significant increase in the probability or consequences of any accident previously evaluated?

Response: No The proposed change incorporates a postulated RVH drop accident into the FSAR for PBNP Unit 1 and revises the PBNP Unit 2 accident analysis. This postulated accident involves the drop of the RVH over a reactor vessel containing fuel assemblies. Assuming that the BMI tubes are severed as a result of displacement of the reactor vessel, a decrease in reactor coolant inventory will occur. Thus, a RVH drop accident can be considered as a LOCA under shutdown conditions.

Revised by letter dated January 12, 2006