ML20209D418

Informs That Util Has Changed Listed TS Bases Pages Attached for NRC Use.Util Made No New Commitments in Ltr

ML20209D418
Person / Time
Site:	Prairie Island
Issue date:	07/07/1999
From:	Schuelke D NORTHERN STATES POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9907130133
Download: ML20209D418 (4)

Text

ov .

Northern States Power Company Prairie Island Nuclear Generating Plant 1717 Wakonade Dr. East Welch. Minnesota 55089 July 7,1999 U S Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Docket Nos.50-282 License Nos. DPR-42 50-306 DPR-60 Technical Specification Bases Chang ,

Cooling Water Supply to Safeguards Bay and Reactor Core Safety Limits In accordance with the Prairie Island Bases Control Program we have changed the following Technical Specification Bases pages which are attached for your use:

B.2.1-3, NSP Revision 145, Dated 4/22/99 B.3.3-4, NSP Revision 142, Dated 11/21/98 in this letter we have made no new Nuclear Regulatory Commission commitments.

Please contact Jeff Kivi (612-388-1121) if you have any questions related to this letter.

. < b Donald A. Schuelke Plant Manager '

Prairie Island Nuclear Generating Plant 1 h

9907130133 990707 xg C)

DR ADOCK0500g22 s l 130022

USNRC NORTHERN STATES POWER COMPANY July 7,1999 l Page 2 .

-Attachments c: Regional Administrator - Region lil, NRC Senior Resident inspector, NRC NRR Project Manager, NRC J E Silberg transmittal revs l'1 and 145. DOC

B.2.1-3 REV 145 4/22/99 A. Reactor Core Safety Limits Bases continued required that the steam generator safety 'ralves protect the pressure from exceeding 110% of design pressure so using 1195 psig in the calculations is conservative. Thus. the reactor is protected from violating the safety limits by the physical limit of the AT trips and the opening of the steam generator safety valves.

As an example, all the. limits for the 2235 psig curve are plotted in Figure B.2.1-1 along with the aT trips and the locus of points where the steam generator safety valves open. This plot demonstrates that the AT trips and the steam generator safety valves do protect the reactor from exceeding the safety limits, Note, however, that the OTAT trip locus on that plot is for steady state conditions and that the locus will drop in response to the rate at which the Tave is increasing. In addition, f(AI) increasing will also l lower the OTAT trip locus.

The safety' limit curves are plotted with AT on the x-axis for the following two reasons: 1.) the full power AT is different at different temperatures and pressures because water properties are nonlinear. This makes it difficult to plot the curves at each pressure using the same scale for the percent power axis. 2.) the AT trip setpoints which the reactor protection system actually calculates is based on the AT. not the percent power.

Except for special-tests. POWER OPERATION with only one loop or with natural circulation is not allowed. Safety limits for such special

' tests will be determined as a part of the test procedure.

The curves are conservative for the following nuclear hot channel factors:

F" = FS"(1 + PFDH(1 - P)) ; and F" = F""

a o where:

1 F""

Q is the Fa limit at RATED THERMAL POWER specified in the CORE l

{

OPERATING LIMITS REPORT. )

l 1

Fnn M is the Fu limit at RATED THERMAL POWER specified in the CORE OPERATING LIMITS REPORT.

- PFDH is the Power Factor Multiplier for F"3 specified in the CORE OPERATING LIMITS REPORT Use of these factors results in more conservative safety limits than would result from power distribution limits in Specification TS.3.10.

i l

v O

t .

f B.3.3-4

(

• REV 142 11/21/98 l 3.3 ENGINEERED SAFETY FEATURES Bases continued The Safeguards Traveling Screens and Emergency Intake Line together with l the Intake Canal and Traveling Screens are designed to provide a supply of screened cooling water to the safeguards bay in the event of a design

, basis earthquake. The design basis earthquake is postulated to destroy Lock & Dam No. 3. River level decreases over time, making the Intake Canal unavailable. The Safeguards Traveling Screens and Emergency Cooling Water Supply line provide an alternate supply of water to the Safeguards Bay, which contains the two diesel driven and the one vertical motor driven cooling water pumps. Their normal supply is from the Cire Water Bay thru one of two sluice gates. Either one of the two sluice gates or one of the two Safeguards Traveling Screens will adequately supply any of the three cooling water pumps. The Safeguards Traveling Screens are not considered part of the " engineered safety features associated with the operable diesel-driven cooling water pump" for determination of operability of diesel-driven cooling water pumps.

The component cooling water system and the cooling water system provide water for cooling components used in normal operation, such as turbine generator components, and reactor auxiliary components in addition to supplying water for accident functions. These systems are designed to automatically provide two separate redundant paths in each system following an accident. Each redundant path is capable of cooling required components in the unit having the accident and in the oper-ating unit.

There are several manual valves and manually-controlled motor-operated valves in the engineered safety feature systems that could, if one valve is improperly positioned, prevent the required injection of emer;ency coolant (Reference 7). These valves are used only when the reactor is suberitical and there is adequate time for actuation by the reactor eperator. To ensure that the manual valve alignment is appro- priate for safety injection during power operation, these valves are tagged and the valve position will be changed only under direct administrative control. For the motor-operated valves, the motor control center supply breaker is physically locked in the open position to ensure that a single failure in the actuation circuit or power supply would not move the valve.

References

1. USAR, Section 3.3.2

2. USAR, Section 14.6.1

3. USAR, Section 5.3.2

4. USAR, Section 6.3

5. USAR, Section 10.4.2

6. USAR, Section 10.4.1

7. USAR, Figure 6.2-1 USAR, Figure 6.2-2 USAR, Figure 6.2-5 USAR, Figure 10.2-11 l

l

\

L_