View source for W3F1-2003-0024, TS Bases Update to NRC for Period January 8, 2003 Through April 7, 2003

{{Adams
| number = ML031080424
| issue date = 04/11/2003
| title = TS Bases Update to NRC for Period January 8, 2003 Through April 7, 2003
| author name = Ridgel J
| author affiliation = Entergy Operations, Inc
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000382
| license number = NPF-038
| contact person = 
| case reference number = A4.05, W3F1-2003-0024
| document type = Letter type:W, Technical Specifications
| page count = 5
}}

=Text=
{{#Wiki_filter:Entergy Operations, Inc.
n     e Lhterg    FAh 17265 River Road Killona, LA 70066 Tel 504 739 6650 W3Fl -2003-0024 A4.05 PR April 11, 2003 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555

==Subject:==
Waterford 3 SES Docket No. 50-382 License No. NPF-38 Technical Specification Bases Update to the NRC for the Period January 8, 2003 Through April 7, 2003 Gentlemen:
Pursuant to Waterford Steam Electric Station Unit 3 Technical Specification 6.16, Entergy Operations, Inc. (EOI) hereby submits an update of all changes made to Waterford 3 Technical Specification Bases since the last submittal per letter W3F1-2003-0002, dated January 15, 2003. This TS Bases update is well within the update frequency listed in 10 CFR 50.71(e).
There are no commitments associated with this submittal. Should you have any questions or comments concerning this submittal, please contact Ron Williams at (504) 739-6255.
Very truly yours,
    . Ridgel Acting Licensing Manager JAR/RLW/ssf Attachment            Waterford 3 Technical Specification Bases Revised Pages cc:                   E.W. Merschoff (NRC Region IV), N. Kalyanam (NRC-NRR),
J. Smith, N.S. Reynolds, NRC Resident Inspectors Office

ATTACHMENT I TO W3F1-2003-0024 Waterford 3 Technical Specification Bases Revised Pages T.S. Bases   Implement       Affected TS Bases                Topic of Change Change No.       Date               Pages 18         3/3/03      B 3/4 4-5            Change to TS Bases section 3/4.4.7 B 3/4 4-6            implemented by ER-W3-2003-0059-000 concurrently with TS Amendment 184 that revised TS 3.4.7 to limit Reactor Coolant System activity permitted by ACTION STATEMENT 'a' to 60 microcuries per gram.

TECHNICAL SPECIFICATION BASES CHANGE NO. 18 REPLACEMENT PAGES (2 pages)
Replace the following pages of the Waterford 3 Technical Specification Bases with the attached pages. The revised pages are identified by Change Number 18 and contain the appropriate DRN number and a vertical line indicating the areas of change.
Remove                          Insert B 3/4 4-5                       B 3/4 4-5 B 3/4 4-6                       B 3/4 4-6

REACTOR COOLANT SYSTEM BASES CHEMISTRY (Continued) the chemistry within Steady State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant The associated effects of exceeding the oxygen, chloride and fluoride limits are time and temperature dependent. Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System. The time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within Steady State Limits The Surveillance Requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action.
3/4.4 7 SPECIFIC ACTIVITY
-   (DRN 03-173, Ch. 18)
The Code of Federal Regulations, 10 CFR 100 specifies the maximum dose to the whole body and the thyroid an individual offsite can receive during a design basis accident. The LCO contains specific activity limits for both DOSE EQUIVALENT 1-131 and gross specific activity.
The specific activity limits ensure that these doses are held within the appropriate 10 CFR 100 requirements (small fraction, well within, or within) during analyzed transients and accidents.
Operation with iodine specific activity levels greater than the LCO limit is permissible for up to 48 hours, provided the activity levels do not exceed 60 uCi/gm. A 48 hour limit was established because of the low probability of an accident occurring during this period. The dose consequences of an accident during this 48 hour period would not exceed the full 10 CFR 100 limits.
The surveillance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take corrective action.
 +- (DRN 03-173, Ch. 18)
WATERFORD - UNIT 3                               B 3/4 4-5                         AMENDMENT NO. 8 CHANGE NO. 18

REACTOR COOLANT SYSTEM BASES SPECIFIC ACTIVITY (Continued)
-  (DRN 03-173, Ch. 18) 4  (DRN 03-173, Ch 18) 3/4.4.8 PRESSURE/TEMPERATURE LIMITS All components in the Reactor Coolant System are designed to withstand the effects of cyclic loads due to system temperature and pressure changes. These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations. The various categories of load cycles used for design purposes are provided in Section 3.9.1.1 of the FSAR. During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.
During heatup, the thermal gradients in the reactor vessel wall produce thermal stresses which vary from compressive at the inner wall to tensile at the outer wall. These thermal induced compressive stresses tend to alleviate the tensile stresses induced by the internal pressure.
Therefore, a pressure-temperature curve based on steady-state conditions (i e., no thermal stresses) represents a lower bound of all similar curves for finite heatup rates when the inner wall of the vessel is treated as the governing location.
The heatup analysis also covers the determination of pressure-temperature limitations for the case in which the outer wall of the vessel becomes the controlling location. The thermal gradients established during heatup produce tensile stresses at the outer wall of the vessel.
These stresses are additive to the pressure induced tensile stresses which are already present.
The thermal induced stresses at the outer wall of the vessel are tensile and are dependent on both the rate of heatup and the time along the heatup ramp; therefore, a lower bound curve similar to that described for the heatup of the inner wall cannot be defined. Consequently, for the cases in which the outer wall of the vessel becomes the stress controlling location, each heatup rate of interest must be analyzed on an individual basis.
WATERFORD - UNIT 3                            B 3/4 4-6                           CHANGE NO. 18}}