ML20207L919

Discusses from Util Informing NRC That Bases References for TS 15.4.4 Revised to Correct References to Point Beach Nuclear Plant FSAR Related to Reactor Containment Design.Revised Bases Page 15.4.4-7 Encl

ML20207L919
Person / Time
Site:	Point Beach
Issue date:	03/11/1999
From:	Wetzel B NRC (Affiliation Not Assigned)
To:	Sellman M WISCONSIN ELECTRIC POWER CO.
References
NUDOCS 9903180315
Download: ML20207L919 (3)

Text

.-.. - . _ . =- . - . _. -.

March 11, 1999 Mr. Michael B. Sellman '

Senior Vice Presideni and Chief Nuclear Officer Wisconsin Electric Power Company 231 West Michigan Street Milwaukee,WI 53201

SUBJECT:

POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 - REVISION TO BASES PAGE 15.4.4 ON REACTOR CONTAINMENT DESIGN REFERENCES

Dear Mr. Sellman:

By letter of February 2,1999, Wisconsin Electric Power Company (WE) informed us that the bases references for Technical Specification 15.4.4 had been revised to correct references to the Point Beach Nuclear Plant (PBNP) Final Safety Analysis Report related to reactor containment design.

We have updated the PBNP Bases. A copy of the revised Bases page 15.4.4-7 is enclosed.

Sincerely, ORIGINAL SIGNED BY Beth A. Wetzel, Senior Project Manager Project Directorate ill-1 Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-266 and 50-301

Enclosure:

As stated cc w/ encl: See next page DISTRIBUTION:

'qDocket FHei G. Hill (4) 3 PUBLIC' ~ W. Beckner )jr PDill-1 Reading C. Berlinger II J. Zwolinski/S. Black B. Wetzel (2) C. Jamerson ,

f ,

OGC B. Burgess, Rlli ACRS \q')

DOCUMENT NAME: G:\PD3-1 \W PDOCS\PTB EAC H\ BAS 99022.W P D To recohn a copy of this document Indicate in the bor *C' = Copy without attachment /encbsure "E" = Copy with attachment /encbsure 'N' = No copy OFFICE PM:PD31, 3 lE LA:PD31 s E D:PD31 lJ NAME BWetzel W CJamerson OJ/ CACarpente(M DATE j /"/0 /99 3 /849 9 /1 l /99 OFFICIAL FMCORD COPY 9903190315 990311 - - - - - -

PDR P

ADOCK 05000266 pga I-.-

L ._ ._, .~-~--~J Ab

• . 4 4

i

- Mr. Michael B. Sellman Point Beach Nuclear Plant Wisconsin Electric Power Company Units 1 and 2 cc:

Mr. John H. O'Neill, Jr. Ms. Sarah Jenkins Shaw, Pittman, Potts & Trowbridge Electric Division 2300 N Street, NW ~ . Public Senrice Commission of Wisconsin i Washington, DC 20037-1128 P.O. Box 7854 Madison, Wisconsin 53707-7854 Mr. Richard R. Grigg ,

President and Chief Operating Officer i

. Wisconsin Electric Power Company 231 West Michigan Street Milwaukee, Wisconsin 53201 Mr. Mark E. Reddemann Site Vice President Point Beach Nuclear Plant Wisconsin Electric Power Company 6610 Nuclear Road Two Rivers, Wis.consin 54241 Mr. Ken Duveneck Town Chairman Town of Two Creeks 13017 State Highway 42 Mishicot, Wisconsin 54228 Chairman Public Service Commission l of Wisconsin P.O. Box 7854 Madison, Wisconsin 53707-7854 j i

Regional Administrator, Region ll1

' U.S. Nuclear Regulatory Commission 801 Warrenville Road  !

- Lisle, Illinois 60532-4351 '

Resident inspector's Office

').S. Nuclear Regulatory Commission .

6612 Nuclear Road  :

Two Rivers, Wisconsin 54241 l october 1see

. . ~- - - -- _ , . - - _ -

Safety analyses have been performed on the basis of a leakage rate of 0.40% by weight per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 60 psig. With this leakage rate and with minimum containment engineered safety  !

systems for iodine removal in operation, i.e. one spray pump with sodium hydroxide addition,

- the public exposure t;ould be well bslow 10 CFR 100 values in the event of the design basis accident."

The safety analyses indicate that the containment leakage rates could be slightly in excess of 0.75% per day before a two-hour thyroid dose of 300R could be received at the site boundary.

The performance of periodic integrated leakage rate tests during plant life provide a current assessment of potential leakage from the containment in case of an accident that would pressurize the interior of the containment. These tests are performed in accordance with the Containment Leakage Rate Testing Program.

Periodic visual and physical inspection of the containment tendons is the method to be used to determine loss of load-carrying capability because of wire breakage or deterioration. The tendon surveillance program specified in 15.4.4.11 is based on the recommendation of ,

Regulatory Guide 1.35 Rev. 3. Containment tendon structuralintegrity was demonstrated for r both units at the end of one, three and eight years fo! lowing the initial containment structural  ;

integrity test. I The pre-stress lift-off test provides a direct measure of the load-carrying capability of the tendon. A deterioration of the corrosion preventive properties of the uheathing filler will be I indicated by a change in the physical appearance of the filler. If the surveillance program indicates, by extensive wire breakage, tendon stress-strain relations, or other abnormal conditions, that the pre-stressing tendons are not behaving as expected, the abnormal conditions will be subjected to an engineering analysis and evaluation in accordance with Specification 15.4.4.II.D to determine whether the condition could result in a significant adverse impact on the containment structural integrity. The specified acceptance criteria are such as to alert attention to the situation well before the tendon load-carrying capability would deteriorate to a point that failure during a design basis accident might be possible. Thus, the cause of the incipient deterioration could be evaluated and corrective action studied without need to shut down'the reactor. If the engineering evaluation determines that the abnormal l

condition could result in a significant adverse impact on the containment structural integrity, an abnormal degradation situation will be declared and a report submitted to the NRC in

accordance with the specifications. l References (1) FSAR Section 5.1.2.2 I (2) FSAR Section 5.1.2 (3) FSAR Section 14.3.5 (4) FSAR Section 14.3.4 (5) Deleted (6) FSAR pages 5.1-61 and 5.1-62 l l

l Unit 1 - Amendment No.181 15.4.4-7 Unit 2 - Amendment No.185 Revised by NRC letter dated March II, 1999

_. ._, . - _ - , __