ML20237C076

Forwards Rev to Proposed Change for FSAR Page 15.4.7 & Listing of FSAR Chapter 15 Accidents to Suppl Util 870722 Tech Spec Changes to Support Deletion of Resistance Temp Detector Bypass Manifold Sys

ML20237C076
Person / Time
Site:	Catawba
Issue date:	12/14/1987
From:	Tucker H DUKE POWER CO.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
NUDOCS 8712210009
Download: ML20237C076 (4)

Text

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, i DUKE POWER GOMPANY P.O. DoX 33180 CIIAltLOTrE, N.O. 28242 IIAL H. TUCKER TELEPIEONE vior ensumarr (704) 373-4531 trtMal.EAS Peopt:UTION December 14, 1987 I

U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D. C. 20555

Subject:

Catawba Nuclear Station, Units 1 and 2 Docket Nos. 50-413 and 50-414 Proposed Technical Specification Changes RTD Bypass Manifold Deletion

Dear Sir:

By letter dated July 22, 1987, Technical Specification changes, WCAPs 11308 and 11309, FSAR changes, and appropriate justification were provided to support the I deletion of the RTD Bypass Manifold System at Catawba Units 1 and 2. Additional information was transmitted via letters dated August 31, 1987, October 1, 1987 .

and October 30, 1987.

Please find attached a revision to the proposed change for FSAR page 15.4-7 and also a listing of FSAR Chapter 15 accidents. The Chapter 15 listing indicates which accidents were affected by the proposed changes and which accidents weren't.

Since this letter supplements a previous request, no license fee is required.

Pursuant to 10 CFR 50.91 (b) (1) the appropriate South Carolina State Official is being provided a copy of this amendment request.

Very truly yours (A..

Hal B. Tucker RWO/1095/sbn Attachments K

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e

I, e a U. S. Nuclear Regulatory Commission

-December 14, 1987 Page Two xc: Dr. J. Nelson Grace, Regional' Administrator U. S. Nuclear Regulatory Commission ,

Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323

, Mr. Heyward Shealy, Chief l Lureau of Radiological Health South Carolina Department of Health &

Environmental' Control

! 2600 Bull Street Columbia, South Carolina 29201 l American Nuclear Insurerc l c/o Dottie Sherman, ANI Library l' The Exchange, Suite 245-270 Farmington Avenue Farmington, CT 06032

.M&M Nuclear Consultants l 1221 Avenue of the Americas New York, New York 10020 INPO Records Center Suite 1500 1100 Circle 75 Parkway Atlanta, Georgia 30339 l Mr. P. K. Van Doorn NRC Resident Inspector Catawba Nuclear Station l

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r-CNS i This accident is analyzed with the Improved Thermal Design Procedure as described in Reference 3. Plent characteristics and initial conditions are discussed in Section 15.0.3. In order to obtain conservative results for an uncontrolled rod withdrawal at power accident, the following assumptions are made:

f

1. Initial reactor power, pressure, and RCS temperatures are assumed to l l '

I' be at their nominal values. Uncertainties in initial conditions are j included in the limit DNBR as described in Reference 3.

< T M P* M d-

2. Reactivity Coefficients - Two cases are analyzed: V '~ ' ' '

resd.s V

a. Minimum Reactivity Feedback. A 4eee& r. w t M moderator coefficient of reactivity is assumed corresponding to the beginning of core A variable Doppler pow e ficient wit in the analysis. A conserv ively eme41 (in absolute' power is used life. agnitude) value is assumed, f isrmll 6-Maximum Reactivity Feedback. A atively lar itive b.

moderator density coefficient and a large (in absolute magnitude) l negative Doppler power coefficient are assumed.

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3. The reactor trip on high neutron flux is assumed to be actuatedThe AT attrips a-conservative value of 118 percent of nominal full power. The delays for include all adverse instrumentation and setpoint errors.

l trip actuation are assumed to be the maximum values.

I 4. The RCCA trip insertion characteristic is based on the assumption that

'the highest worth assembly is stuck in its fully withdrawn position.

5. The maximum positive reactivity insertion rate is greater than that for the simultaneous withdrawal of the combinations of the two control ba having the maximum combined worth at maximum speed. ,

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The limiting

6. Pressurizer pressure control is assumed to be available. Maintaining lower RCS criterion for this transient is minimum DNBR.Therefore pressure control pressures will result in a lower DNBR.

availability is conservative.

The effect of RCCA movement on the axial core power distribution is accounted for by causing a decrease in overtemperature AT trip setpoint proportional to a decrease in margin to DNB.

Plant systems and equipment which are available to mitigate theNoeffects of the accident are discussed in Section 15.0.8 and listed in Table 15.0.8-1.

single active failure in any of these systems or equipment will adversely offset the consequences of the accident.

is presented in Reference 5.

1 R*V- 14 15.4-7

The primary impact of the RTD Bypass Elimination is the increase in the RTD response time. Thus, only.those events which rely on the Overtemperature and Overpower Delta-T (OTDT and OPDT) reactor trips are impacted. For Catawba these events include the Uncontrolled RCCA Withdrawal at Power, the Uncontrolled Boron Dilution at Power, and the Steamline Rupture at Power events. For the remainder of the FSAR events, listed below, the OTDT and OPDT do not provide primary reactor protection. Therefore, for these events, the results remain unchanged and the conclusions of the FSAR remain valid.

15.1.1 Feedwater System Malfunctions that Result in a Decrease in Feedwater Temperature 15.1.2 Feedwater System Malfunctions that Result ir. an Increase in Feedwater Flow 15.1.3 Equipment Malfunction or Operating Failure that Results in Increasing Steam Flow 15.1.4 Inadvertent Opening of a Steam Generator Relief or Safety Valve 15.1.5 Spectrum of Steam Piping Failures Inside and Outside Containment 15.2.2 Loss of External Electrical Load 15.2.3 Turbine Trip 15.2.4 Inadvertent Closure of MSIVs 15.2.6 Loss of Non-Emergency A-C Power to the Station Auxiliaries 15.2.7 Loss of Normal Feedwater 1

15.2.8 Feedwater System Pipe Breaks Inside and Outside Containment 15.3.1 Partial Loss of Forced Reactor Coolant Flow 15.3.2 Complete Loss of Forced Reactor Coolant Flow 15.3.3 Reactor Coolant Pump Shaft seizure 15.3.4 Reactor Coolant Pump Shaft Break 15.4.1 Uncontrolled RCCA Bank Withdrawal from a Suberitical or Low Power Startup Condition 15.4.3 RCCA Misalignment 15.4.4 Startup of an Inactive Reactor Coolant pump (RCP) with Low Hot Leg Temperature 15.4.8 Spectrum of RCCA Ejection Accidents 15.5.1 Inadvertent Operation of Emergency Core Cooling System (ECCS) During Power Operation 15.6.1 Accidental Depressurization of the Reactor Coolant System J

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