ML20199D456

Advises That Wording Change to Clarify Bases Section of Tech Spec 4/5.2.2.5 Re Safety Valves & Reidentification of Turbine Trip W/O Bypass Transient as Postulated Most Limiting Pressure Transient Acceptable

ML20199D456
Person / Time
Site:	La Crosse File:Dairyland Power Cooperative icon.png
Issue date:	06/05/1986
From:	Zwolinski J Office of Nuclear Reactor Regulation
To:	Taylor J DAIRYLAND POWER COOPERATIVE
References
TAC-61137, NUDOCS 8606200322
Download: ML20199D456 (3)

Text

y . q o-  % ,

' Docket No. 50-409 .

Mr. James W. Taylor General Manager June 5, 1986 b h Dairyland Power Cooperative 2615 East Avenue South d

U

-La Crosse, Wisconsin 54601

SUBJECT:

WORDING CHANGE IN BASES TO SECTION 4/5.2.2 c REGARDING SAFETY VALVES AND REIDENTIFICATION OF THE TURRINE TRIP WITHOUT BYPASS TRANSIENT AS THE POSTillATED MOST LIMITING PRESSURE TRANSIENT

'(TAC 61137)

Re: La Crosse' Boiling Water Reactor In your application for amendment to license, LAC-11430, dated February Pl..

1986, you requested six specific changes or related actions to your current Technical Specifications. One of your requests was-to clarify the bases section to specification 4/5.2.2.5 regarding safety valves, and the reidentification of the turbine . trip without bypass as the most limiting. pressure transient.

According to 10 CFR 50.36(a), however, "A sumary statement of the bases or reasons for such specifications, other than those covering adminstrative con-trols, shall-also be included in the application, but shall not become part of-the technical specifications." 'Thus, changes to the bases portion of associated technical specifications are not considered technical specification changes,-and hence need not be analyzed using no significant hazard considerations.

. The staff has, however, reviewed your request and concludes that the wording change concerning the safety valves, and reidentifying the most limiting pressure

. transient as the turbine trip without bypass in the bases of Section 4/5.2.2.5 are appropriate changes-to clarify the bases. A copy of the revised page is i enclosed.

, Sincerely,

/S/

D 00h $bboOko9 _ John A. Zwolinski, Director P PDR BWR Project Directorate #1 1

Division of BWR Licensing I

Enclosures:

. Bases pace cc w/ enclosures:

See next pane b

r .

i DISTRIRIITION Docket File BGrimes (

d ,f i p/'

NRC PDR JPartlow L

. Local PDR

-PD#1 Reading' RRuck JStang C k 1

p[,)

RBernero CJamerson i

~

"0 ELD EJordan JZwolinski I j h. ,/

ACRS (10)

-La Crosse File E. Butcher T. Barnhardt (4)

W. Jones /

'DB PkN1

- . DBL:PD#I I D#1 DB DR #

CJ c RBuck' t Mg ilI s JZwolinski f/ 86 / /86 6 b /86 / /86 h/6 /8 b 2

i Mr. James W. Taylor .

Dairyland Power Cooperative la Crosse Boiling Water Reactor cc:

Fritz Schubert, Esquire Clarence Riederer, Chief Engineer Staff Attorney . Wisconsin Public Service Dairyland Power Cooperative Commission  :

2615 East Avenue South Post Office Box 7854 La Crosse, Wisconsin 54601 Madison, Wisconsin 53707 Roy P. Lessy, Jr.

O. S. Heistand Morgan, Lewis & Bockius 1800 M Street, N.W.

7th Floor North Receptionist Washington, D.C. 20036 Mr. John Parkyn, Plant Manager la Crosse Boiling Water Reactor Dairyland Power Cooperative P. O. Box 275 Genoa, Wisconsin 54632

. U.S. Nuclear Regulatory Commission Resident Inspectors Office Rural Route #1, Box 276 Genoa, Wisconsin 54632 Town Chairman

. Town of Genoa Route 1 Genoa, Wisconsin 54632

~ Chairman, Public Service Commission of Wisconsin Hill Farms State Office Building Madison, Wisconsin 53702 Regional Administrator, Region III U.S. Nuclear Regulatory Commission 799 Roosevelt Road Glen Ellyn, Illinois 60137 l

l l

l i

l l

l

, *6

. REACTOR COOLANT SYSTEM

^

BASES 4.2.2.5 SAFETY VALVES The safety valves are designed to meet the requirements of the ASME Boiler and Pressure Vessel Code. The reactor primary system overpressurization protection is sufficient to limit the pressure within the pressure-retaining boundaries to less than 1540 psig, which is less than 110% of the vessel design pressure of 1400 psig.

The safety valves have a minimum stamped relieving capacity of 294,612 lb. per hour at a relief pressure of 1390 psig and 302,160 lb per hour at a relief pressure of 1426 psig. Three safety valves are installed. The relieving capacity with one valve inoperable is sufficient to limit the primary system t pressure to less than 110% of the vessel design pressure during an abnormal transient with the highest pressure, which is the MSIV closure. A high pressure scram is initiated at 1325 psig and no credit is taken for the MSIV l closure scram signal, the power-flow scram, the overpower scram nor the pressure reduction due to automatic operation of the shutdown condenser heat sink.

During the postulated most limiting pressure transient, caused by a turbine j trip without bypass valve operation with full scram on high power (120%), I l . reactor pressure would not reach 1390 psig, the lowest safety valve set point.

l [\

The safety valve function is therefore not expected to be required under the most limiting operational transient.

l The testing frequency applicable to the safety valve function is provided to ensure operability and demonstrate reliability of the valves. The required testing interval varies with observed valve failures. Set point drift within 1 3% of the setpoint is not considered to be valve failure for the l putposes of this test schedule. Setpoint drif t > 13% of the setpoint will be

! cause to test additional valves in accordance with ASME Section XI test schedule. The popping setpoints are significantly below the 110% primary system design pressure safety limit. Therefore, adequate margin exists between the setpoint and the safety limit of 1540 psig.

For the purposes of establishing the test frequency, a valve shall be considered to have failed to function properly if the test relief pressure is

! determin<td to be outside of the allowable setpoint tolerance specified in the ASME Code to which the valve was constructed. For the LACBWR spring loaded valves which are constructed to the ASME Code Section VIII, 1962, and Nuclear Code Case N-1271,1962, and which must be removed from the primary steam system to conduct the test, the allowable setpoint tolerance is + 3% of the

~

set pressure. However, when the safety valve relief pressure is set prior to installing the valve on the reactor system, the maximum deviation of the test relief pressure from the specified set presure shall not exceed i 1.0% of the required set pressure.

l 30h WPl.6.8 Revised 06/ 5 /86

. _ _