NL-05-0252, Technical Specifications Bases Revisions for 2004

Technical Specifications Bases Revisions for 2004

ML050610161
Person / Time
Site:	Hatch
Issue date:	02/24/2005
From:	Sumner H Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-05-0252
Download: ML050610161 (5)

Text

H. L Sumner, Jr. Southern Nuclear Vice President Operating Company, Inc.

Hatch Project Post Office Box 1295 Birmingham, Alabama 35201 Tel 205.992.7279 SOUTHERN February 24, 2005 COMPANY Energy to Serve Your World">

Docket Nos.: 50-321 NL-05-0252 50-366 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant Plant Hatch Technical Specifications Bases Revisions for 2004 Ladies and Gentlemen:

Enclosed you will find the 2004 Technical Specifications (TS) Bases revisions implemented per 10 CFR 50.59 at Plant Hatch Units I and 2. This is supplied per the requirements of the Bases Control Program, TS section 5.5.1 1.

This letter contains no NRC commitments. If you have any questions, please advise.

Sincerely, H. L. Sumner, Jr.

HLS/OCV/daj

Enclosures:

Technical Specifications Bases Changes for 2004 cc: Southem Nuclear Operating Company Mr. J. T. Gasser, Executive Vice President Mr. G. R. Frederick, General Manager - Plant Hatch RTYPE: CHAO2.004 U. S. Nuclear Regulatory Commission Dr. W. D. Travers, Regional Administrator Mr. C. Gratton, NRR Project Manager - Hatch Mr. D. S. Simpkins, Senior Resident Inspector - Hatch ADDI

Enclosure Edwin I. Hatch Nuclear Plant, Units 1 and 2 Technical Specifications Bases Changes for 2004 Description of Bases Chances Each of the below Bases changes are applicable to both Hatch units unless otherwise noted.

Licensing Document Change Request (LDCR) 2004-066 Section B 3.7.1. "RHR SW System" This change was made to clarify the operability requirements for RHRSW with respect to the required flow rate per pump to support post-accident containment cooling.

Specifically, an engineering calculation was done, SMNH-04-010, which verified that 3500 gpm per pump, with two pumps operating in one loop, are required to support the required suppression pool heat load removal following a LOCA.

The Unit 2 Applicable Safety Analysis section of B 3.7.1 now states that two RHRSW pumps in one loop must run, each with a flow rate of 3500 gpm to support the assumed heat removal rate in the safety analysis. The final analyzed suppression pool temperature and pressure of 210 F and 37 psig, respectively, were also included in this Bases section.

The values for the Unit I pumps are 4000 gpm per pump for a final temperature and pressure of 281 F and 62 psig, respectively.

LDCR 2004-056 Section B 3.3.2.1. "Control Rod Block Instrumentation" This change was made to clarify when a reactor start-up may be performed with the Rod Worth Minimizer (RWM) system out of service.

Technical Specifications (TS) Required Action Statement (RAS) C.2 states that control rod movement during a reactor startup may continue provided that greater than 12 control rods have been withdrawn. If 12 rods have not been withdrawn, startup is permitted provided the RWM has not been inoperable within the last "calendar year" and a second licensed operator verifies compliance with the Banked Position Withdrawal Sequence.

The Bases section for RAS C.2 was revised to provide clarification on the meaning of "last calendar year". Specifically, the Bases section was revised to include the phrase,

"...in the last 12 months". Therefore, this provision may be used if no startup had been performed with the RWM inoperable in the last 12 months.

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Enclosure Edwin I. Hatch Nuclear Plant, Units 1 and 2 Technical Specifications Bases Changes for 2004 Description of Bases Changes LDCR 2004-032 Various Bases Sections This change was made to several Reference sections in the Bases to amplify Technical Requirements Manual (TRM) references. Before the change, most references to the TRM only indicated the TRM without listing a specific section.

For example, after this change, the Bases Reference section to B 3.3.1.1, "RPS Instrumentation" lists "Technical Requirements Manual Table T5.0-1", rather than just "Technical Requirements Manual", as it did prior to the change. Similar changes were made to other Bases Reference sections.

LDCR 2004-052 Section B 3.6.1.4. "Drowell Pressure" The Applicable Safety Analysis of this Bases section was changed to reflect the new primary containment peak post accident pressure. The new pressures were changed as a result of a TS amendment request approved by the NRC under Amendments 241 and 184 for Hatch Units 1 and 2 respectively. The TS change was made in section 5.5.12, "Primary Containment Leak Rate Testing Program". The change in the peak post accident drywell pressure came about from the Hatch measurement uncertainty recapture power uprate (MURPU). This uprate ultimately resulted in an increase in RPV pressure by 10 psi. The increase in pressure required an NRC approval separate from the power uprate program because of the increase in post accident drywell pressure.

LDCR 2004-018 Unit 1. Section B 3.7.1. "RHRSW System" The Applicable Safety Analysis Bases section was changed to correct a reference. The previous version stated that the ability of the RHRSW system to support long term cooling is discussed in FSAR section 5.2. This was incorrect. The Bases were therefore changed to indicate the correct FSAR section, 10.6.

LDCR 2004-041 and 2004-036 Section B 3.3.6.1.1.c. "Main Steam Line Flow High Instrumentation" This Bases section provides the differential pressure (DP) setting on the instruments measuring main steam line flow. These instruments initiate a closure of the main steam isolation valves on a value of 138 % of rated steam flow, which is the TS Allowable Value. Hatch Unit 1 and 2 recently received NRC approval for the measurement E-2

Enclosure Edwin I. Hatch Nuclear Plant, Units 1 and 2 Technical Specifications Bases Changes for 2004 Description of Bases Changes uncertainty recapture power uprate (MURPU). This uprate ultimately resulted in an increase in RPV pressure by 10 psi. (This increase in pressure required an NRC approval separate from the MURPU approval.)

As a result of that increase in pressure, the steam mass corresponding to 138 % steam flow changed, resulting in 140 PSID being representative of 138 % steam flow as opposed to 142 PSID for Unit 1. For Unit 2, the value changed to 173 PSID from 175 PSID. These DP values were therefore changed in the above referenced Bases section.

The TS Allowable Values for high steam line flow did not change.

LDCR 2004-049 Section B 3.1.8. "Scram Discharge Volume (SDV) Vent and Drain Valves" This Bases section was changed to support implementation of NRC approved TS Amendment 240 and 183 for Hatch Units 1 and 2 respectively. This TS amendment allows operating for an indefinite period of time with one scram discharge volume vent or drain valve inoperable. It further allows periodic opening of the valve, under administrative controls, to insure the volume is capable of retaining the volume of water expected during a reactor scram. Previously, a Required Action Statement with a seven day Completion Time was entered when an SDV valve was inoperable. The ACTION section in the Bases was revised to reflect the new TS guidance.

LDCR 2003-082 Section B 3.8.1, "AC Sources - Operating" The Bases for Required Action Statement B.4 was revised to clarify and remind operators that restrictions for using the 14 day Completion Time (CT) are in effect for the swing diesel generator, as well as for the dedicated diesel generators. The restrictions involve a list of equipment that must be functional prior to using the 14 day CT on the diesel generators. If the restrictions are not in effect, the 14 day CT may not be used, and the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> CT would be in effect. The specific components are listed in Hatch maintenance procedure, 90AC-OAM-002-0.

LDCR 2003-076 Section B 3.7.5, "Main Control Room Environmental Control (MCREC) System" This section was revised to remove a discussion of a Plant Service Water (PSW) system design feature dealing with an automatic PSW divisional transfer for cooling water to the "B" Main Control Room air conditioner. The automatic feature was removed via a design change.

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Enclosure Edwin 1.Hatch Nuclear Plant, Units I and 2 Technical Specifications Bases Changes for 2004 Description of Bases Changes Cooling water to the 'B' air conditioner can be supplied via Division I or Division II of the Plant Service Water (PSW) system. Previous to the design change, the transfer of the cooling water supply to the "B" air conditioner could be accomplished manually, and was also accomplished automatically on a low PSW flow signal coincident with a Loss of Coolant Accident or Loss of Offsite Power. That automatic logic was deleted via a design change, and therefore the discussion was removed from the Bases. The automatic swap was a design feature of the system that was unnecessary for the operability of the air conditioners. Nonetheless, the transfer of cooling water can still be accomplished manually via the operation of remote manual switches located in the control room. This manual operation is contained in the Unit 1 PSW system operating procedure.

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