ML20205C156

Forwards Description of Change to Initial Startup Test Program,Deleting Doppler Only Power Coefficient Verification Tests

ML20205C156
Person / Time
Site:	Catawba
Issue date:	08/04/1986
From:	Tucker H DUKE POWER CO.
To:	Harold Denton, Youngblood B Office of Nuclear Reactor Regulation
References
NUDOCS 8608120265
Download: ML20205C156 (18)

Text

_ _ _ _

t-O DUKE POWER GOMPANY P.O. BOX 33189 C HARLOTTE, N.C. 28242 HAL H. TUCKER Tex.zenown vnon ramminaa7 (704) 373-4831 atma.aAa emonverson f

August 4, 1986 Mr. Harold R. Denton, Director Office of Nuclear keactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C.

20555 Attention:

Mr. B. J. Youngblood, Project Director PWR Project Directorate No. 4 Re: Catawba Nuclear Station, Unit 2 Docket No. 50-414

Dear Sir:

In accordance with License Condition 3 of Facility Operating License NPF-52 and 10 CFR 50.59(b), please find attached the description of a change that has been made to the Initial Startup Test Program for Catawba Unit 2.

This change would delete the Doppler Only Power Coefficient Verification tests as was previously done on McGuire Unit 2.

Very truly yours, d

Hal B. Tucker ROS/06/ sib Attachment xc:

Dr. J. Nelson Grace, Regional Administration U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 NRC Resident Inspector Catawba Nuclear Station

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Form 34634 (R8-85)

DUKE POWER COMPANY NUCLEAR SAFETY EVALUATION CHECKLIST x

(1) STATION:

(L dM d2 UNIT: 1 2

3 OTHER:

(2) EVALUATION APPLICABLE TO (DESCRIPTION AND NUMBER OF NSM, PROCEDURE, PROCEDURE CHANGE, OR TEST / EXPERIMENT):

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(3) SAFETY EVALUATION - PART A The item to which this evaluation is applicable represent:

dyes O No A change to the station or procedures as described in the FSAR: or a test or experiment not de-scribed in the FSAR? Affected FSAR Section(s) are:

[tkanlen N/ bre e er Yn clnwse2$

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k if the answer to the above is "Yes," identify the affected section(s) of the FSAR. Attach additional sheets as necessary.

(4) SAFETY EVALUATION - PART B O Yes [No Will this item require a change to the station Technical Specifications? Affected Tech. Specs. Sec-tion (s)are:

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U If the answer to the above is "Yes," identify the specification (s) affected and/or attach the applicable pacp(s) with the change (s) indicated. Tech. Spec changes require NSRB and NRC approval pnor to use.

(5) SAFETY EVALUATION - PART C As a result of the item to which this evaluation is applicable:

0 Yes E[No Will the probability of an accident previously evaluated in the FSAR be increased? Explain:

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V O Yes O'No Will the consequences of an accident prev!ously evaluated in the FSAR be increased? Explain: _

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DUKE POWER COMPANY NUCLEAR SAFETY EVALUATION CHECKLIST O Yes dNo May the possibility of an accident which is different than any already eva!uated in the FSAR be cre-ated? Explain:

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V O Yes INo Will the probability of a malfunction of equipment important to safety previously evaluated in the FSAR beincreased? Explain:

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U O Yes C(No Will the consequences of a malfunction of equipment important to safety previously evaluated in the FSAR beincreased? Explain:

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O O Yes INo May the possibility of malfunctions of equipment important to safety different than any already evalu-ated in the FSAR be created? Explain:

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U Will the margin of safety as defined in the bases to any Technical Specification be reduced?

O Yes o

Explain:

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Justification for the answers above (Yes or No) must be provided in the above spaces (attach additional sheets as necessary).

An unreviewed safety question is involved if any answer to Part C above is "Yes" and NRC authorization is required.

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Table 14.2.12-2 (Page 23) l DOPPLER ONLY POWER COEFFICIENT VERIFICATIONe @n, Onl )

y Abstract Purpose l

To verify the nuclear design predictions of the doppler only power coefficient.

Prequisites l

The reactor is at a stable power condition with rods in the specified maneuver-ing band.

The instrumentation necessary for collection of data is installed, calibrated and operable.

Test Method Initial data is taken.

With the turbine and reactor controls in manual, the turbine load is decreased then increased.

Data is recorded during and after the load maneuver and used to infer a measured doppler coefficient verifica-tion factor.

This factor is compared to a vendor supplied predicted doppler verification factor.

Acceptance Criteria The inferred measured doppler coefficient verification factor agrees with pre-dicted values as specified by the vendor.

Rev. 10 1

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Table 14.2.12-2 (Page 35)

NATURAL CIRCULATION VERIFICATION TES Abstract h"!Y I O Purpose To demonstrate the capability of the NSSS to remove sensible heat by natural circulation flow in the primary loop.

To verify that pressurizer pressure and level control systems can respond automatically to a loss of forced circulation and can maintain reactor coolant pressure within acceptable limits.

To verify that steam generator level and feedwater flow can be maintained under natural circulation conditions in order to maintain effective heat transfer from the l

reactor coolant system.

To provide operator training to satisfy NUREG 0737 requirements.

Prerequisites The reactor is critical at a power level of approximately 3% full power with all reactor coolant pumps in operation.

Rod control is in manual with Bank D positioned to maintain a slightly negative isothermal temperature coefficient.

Pressurizer pressure and level control are in automatic.

Steam dump control is in the pressure control mode.

Steam generator level is being maintained through use of the auxiliary feedwater header.

The intermediate and power range (low setpoint) high level reactor trips have C

been reduced to approximately 7% rated thermal power.

UHI isolation valves have been gagged.

Overtemperature and overpower AT reactor trip signals have been blocked.

Various Technical Specifications test exemptions are required for the conduct of this test.

These special test exemptions are provided in Technical Spec-ifications.

Special operator action guidelines are provided by the test pro-cedure to compensate for the blocking of various safety injection functions and reactor trips.

The test is required to be performed at core burnups which ensure that no signifigant core decay heat levels are present.

Test Method The test will be initiated by tripping all operating reactor coolant pumps.

The establishment of natural circulation will be verified by observing the response of wide range hot and cold leg temperatures as well as core exit thermocouples.

The response of pressurizer level and pressure will be ob-served.

Steam generator level and pressure response will be monitored.

Dur-ing the performance of this test on Catawba Unit 1 only, the test will be re-peated for each operating shift at Catawba or suitable simulator facility, for the purpose of initial operator training.

Each R0 and SRO will observe or participate in the initiation, detection and maintenance of natural circula-tion conditions during at least one of the test runs.

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TESTING FOLLOWING INIf!AL FUEL LCA0!NG Fuel Hot Precritical Initial Zero Power 0% - 52 Power 10% - 25%

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1. Moveable Incore

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Functional Test Power Physics Testing:

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10. RTO Bypass Flow ed RCCA worth Verification at Not Zero

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11. Pressurizer funct-tional Test The completion of this test is not required before initial escalation to the next power testing plateau.

NOTE 1: Tests =111 be completed prior to exceeding the 30% testing plateau.

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.meeneremen-Monitor Test tion Test 8.

Unit Load 9.

Support Systems l

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(C) Oyes 3?No involves an unreviewed safety question?

(D) XYes CNo Requires completion of a NUCLEAR SAFETY EVALUATION CHECK LIST ?

If the answer to any of the above is YES, attach a detailed explanation. As appropriate attach a completed NUCLEAR SAFETY EVALUATION CHECK Lif7 form. If the answer to (B) or (C)is YES the change must be approvM by the NSRB and NRC prior to implementatio By D

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DUKE POWER COMPANY NUCLEAR SAFETY EVALUATION CHECKLIST Catawba UNIT: 1 2

X 3

(1) STATION:

OTHER:

(2) EVALUATION APPLICABLE TO (DESCRIPTION AND NUMBER OF NSM, PRO TP/2/A/2100/01.

OR TEST / EXPERIMENT):

of steos to oerform Doooler Only Power Coefficient Verification at 50 and 90% Power.

(3) SAFETY EVALUATION -PART A The item to which this evaluation is applicable represent:

A change to the station or procedures as descnbed in the FSAR: or a test or experiment not d Table 14.2.7-1 (oage 31. Figure 3 Yes O No scnbed in the FSAR? Affected FSAR Section(s) are:

14.2.11-1 (marked uo cooies attachedl.

If the answer to the above is "Yes," identify the affected section(s) of the FSAR. Attach additional sheets as necessary.

(4) SAFETY EVALUATION - PART B Will this item require a change to the station Technical Specifications? Affected Tech. Specs. Sec-tion (s)are:This item does not require a chance to the Station Technical Q Yes G No Specifications.

If the answer to the above is "Yes," identify the specification (s) affected anc er attach the applicable pag change (s) indicated. Tech. Spec. changes require NSRB and NRC approva onor to use.

(5) SAFETY EVALUATION -PART C As a result of the item to which this evaluation is applicable:

The performan Will the probabilit of an accident previously evaluated in : e FSAR be increased? Explain akkkn h'fhkk tN ok k If k

!hN a Nb b oO O Yes 00 No 2*M!!/L*si'X*1^t*& 9J'Mbd"?n idtMRRT 19Aoulffm p

o d

alues of DoyNbinkre$shkiplain:_lerO idept An i sis gfyk' val dM*cbatween jSAR Acgb tMMtne j

f Ms uMo O Yes 00No Deletion of Doppler Only Power Coefficient Measurements will not increase the severity of accidents previously evaluated in the FSAR since the tran-sient analysis uses very conservative values that were never approached no reason for Unit 2 measurements to differ signifal c%Te pesigOn in Unit 1 festing. By virtue of essentially identi

~

ican y rom 1 results.

/f0dV Form 34634 (R845)

DUKE POWER COMPANY NUCLEAR SAFETY EVALUATION CHECKLIST O Yes G No May the possibility of an accident which is different than any already evaluated in the FSAR be cre-ated? Explain: No new accidents not evaluated in the FSAR will become nossible.

O Yes GINo Will the probability of a malfunctiorlof eauioment important to safety previously evaluated in the FSAR beincreased? Explain: Doppler Only Power Coefficient Measurements do not verify or affect performance of safety-related equipment. Therefore, deletion of these measurements will not increase probability of safety-related equipment malfunction.

O Yes Gi!No Will the consequences of a malfunction of equipment important to safety previously evaluated in the FSAR beincreased? Explain: nalatinn nf nnnnlor nnly pnwar enafficiant Maacnrements will not degrade safety-related equipment or further promote a previously degraded condition.

O Yes Gi!No May the possibility of malfunctions of equipment important to safety different than any already evalu-ated in the FSAR be created? Explain: No safety-related equipment malfunctions not evaluated in the FSAR will become possible as a result of measurement deleticn O Yes 00 No Will the margin of safety as defined in the bases to any Technical Specification be reduced?

Explain:

There are no bases in the Toch Snect which wnuld ho affected by the deletion of Onnnier Only pnwor Cnofficient Maaenrpmonte.

Justification for the answers above (Yes or No) must be provided in the above spaces (attach additional sheets as necessary).

An unreviewed safety question is involved if any answer to Part C above is "Yes" and NRC authorization is required.

N#C (6) Prepared by:

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(8)Page 2 of

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TABLE 14.2.7-1 (Page 3)

COMPLIANCE WITH REGULATORY GUIDES Affected Justification Regulatory Guide Compliance Section(s)

Exception Taken 1.68 Rev. 2 Partial App. A 5 Tests and acceptance criteria will be Control system testing should verify proper developed to demonstrate the ability contre.1 of process variables within the design of major principal plant control control deadband, not over the range of design systees to automatically control pro-values of process variables. Proper control cess variables within design limits of process variables will be demonstrated around the nominal reference value.

during power escalation over the range of 0 to 1005 F.P.

Partial App. A 5.a Power coef ficient measurements will NSSS vendor does not recommend performing this not be performed at 1005 power but test at 100K power due to potential of violating axla Hux eMnce eC ca Pec W cation.

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App. A 5.b Departure rom nucleate boiling ratio Axial, Radial, and Total Peaking will be

.a t

(DNBR), samlaus average planar linear directly measured and verified during power

[

heat generation rate (MAPLHGR), and escalation testing and will be used to verify einimum critical power ratio (MCPR)

DNBR and linear heat rate margin by analysis.

will not be directly verified dur-ing power escalation testing.

Partial App. A 5.f Core thermal and nuclear parn sters The reactor core will be under menon transient will not be demonstrated *. De in conditions at this time. There would be in-accordance with predictions following sufficient time to gather data under transient a return of the rod to its bank position.

conditions. There are no NSSS vendor predictions for this configuration.

App. A 5.g Special testing to demonstrate control Refer to q640.52 itse 4.1 response.

rod sequencers/ withdrawal block funtions operation will not be per-formed.

App. A 5.h Rod drop times will not be measured 14easuring rod drop times at power would re-quire disabling all position indication for at power.

the rods in violation of plant Technical

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Specifications.

App. A 5.1 Test to demonstrate incore/excore From vendor predictions the Xenon and power instrimentation sensitivity to distributions at SOE and 100K are sieflar.

detect rod alsalignment w111 not be The performance of this test at SOE should performed at full power.

adequately demonstrate the capability and sensitivity of incore/excore instrumentation to detect control rod misalignments equal to 2 h. i Msinfiell draf.d f**l or less than Technical Specifications.

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