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 l

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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1 O 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):

C/1r/lib aWNm of 40cndnrus of Um/ aL4o' W C 6 ls h in & /$as /en 4/ sl fdhW u /

(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 k er Yn clnwse2$ /3 /

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:

sto fec l m bo l .3 h ec h ir a b m C /Ianne U S ne&^k

/

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:

dee rM d C S 3lY

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

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i Form 34634 (R8-85)

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:

dee xia. a e s 3 I' 'l

/ 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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3 lY O Yes C(No Will the consequences of a malfunction of equipment important to safety previously evaluated in the FSAR beincreased? Explain:

dee Daoes J VY

/ 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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dlV O Yes o Will the margin of safety as defined in the bases to any Technical Specification be reduced?

Explain:

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s o 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.

(6) Prepared by: 4 9 Date: / 22' 88d J J (7) Reviewed by: Date: Y, I (Qualified Reviewer)

(8) Page 2 of Y J

Duka Power Company

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ADDRESS FROM he4er Le R-J Sus;EcT Sa#ek J

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Duka Power Company MEMORANDUM

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ss.Yackme f Table 14.2.12-2 (Page 23)

DOPPLER ONLY POWER COEFFICIENT VERIFICATION e @n, l Abstract Only )

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

cdkckmW la.-

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 reactor coolant system. To provide operator training to satisfy NUREG 0737 l 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 been reduced to approximately 7% rated thermal power. UHI isolation valves C 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.

Rev. 10

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Figure 14.2.11-1 i

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

Fuel Hot Precritical Initial Physics Test Post-Physics Testinn Power Testina Criticaffty toadina

1. Controlling Proc- 1. Radiation Shielding 1. Loss of Control Initial Fuel 1. Moveable Incore 1. Initial Criticality cedure for Zero Survey Room Test (Note 1)

Loading Detector Functional Test Power Physics Testing: 2. Natural Circulation 2. Station Blackout Vert (ication Test (Note 1)

2. Incore Thermo-couple Functional (a) Nuclear Instrua ( Alote .5 Test mentation Over- 3. Unit Load toady-lap Verification State Test

.) .

3. Incore Thereo-couple and RTD (b) Onset of Nuc- *4 Process and Ef fluent 3. NIS LiiTs Cross Calibration lear Heat Radiation Monitor Test '

(Optional)

5. NIS Initial Calibra-Qljh (c) All Rods Out

4. Rod Position Critical Boron tion Indication Check

5. Rod Control (d) Isothermal Temperature

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Cluster Assembly Coefficient

$egeyggp J Orop Time Test Test

6. Rod Control (e) Dif ferential and Integral yM System Alignment

' Test Worth of Se*

O ro e 7 Full Length Rod Drive Mechanise (f) Differential 7if.3 Timing Test Soron wortn at Hot Zero

8. Reactor Coolant Power System Flow Test (g) Integral Con-trol Rod worth

9. Reactor Coolant With one Stuck System Flow Coastdown Test Rod gQ 3 (h) Pseudo-Eject *

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

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.

NOTE 2: Test will be completed prior to exceeding the 75% testing plateua.

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afabed .s s30% F. P.

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1. Unit Load .

1. Unit Load Steady 1. Unit Lead 5teady 1. Unit Load Steady 1. Unit Load Steady Steady State State Test State Test State Test State Test

2. Radiation 2. Radiation Shielding 2. Radiation Shielding 2. NIS Initial Calibra- 2. Radiation Shielding Shielding Survey Survey tion Survey Survey ,

3. NIS Initial Calfbra- 3. NIS InittaI Calibra- 3. Core Power Distribu- 3. N!5 Initial

3. Rod Control tion tion tion Calibration System at Power Test 4. Core Power 4. Core Power Distri- *4. Feedwater Tempera- 4. Core Power Of s-Distribution Test bution Test ture varia lon Test tribution Test

4. N!S [nitial (fletc J Calibration 5. 7- C '; a; 4 5. ~.- . a. . .c ha; 5. Doppler on y Power

-4 Tv-. Afect- h -Defest Coef ficient verf f t-

5. Core Power Distributton

"; ; _ _ -- 7 ";;;.. _ cation g 5. Unit Load Tran-sient Test

6. Unit Load 6. Unit Load Tran-

6. Psuedo Ejec- Transient Test sient Test 6. Unit Loss of lion Rod Jest ,. . Electrical Load (M*fe 3) 7. 8elow Ban'k Iest 7. Incore and Nuc- Test

7. -8omeMos** /Jefe .I) lear Instrumen-2"' ' "-- 8. P(rocess and Ef- tation System 7. Process and Effluent

" - ^:" Hb fl ent Radiation Detector Correla- Radiation Monitor

.meeneremen- Monitor Test tion Test

8. Unit Load 9. Support Systems l 8. Turbine Trip 8. Support Systees Transient Verification Test (power Verification Test Test just below

9. Pressurizer P-9 setpoint)

Level and (Note 2)

Pressure Cont-ol Test W a>4eaccetoe

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s Form 35283 (R8-85) (1)ID No.M ONhg /

DUKE POWER COMPANY PROCEDURE MAJOR CHANGE Change No. y g/d PROCESS RECORD Festricted To (2) STATION Cd d h/ A (3) PROCEDURE TITLE Ba rahin,; Pen,,),,re fa c Pe nn Es< aMi,u (

(4) SECTION(S) OF PROCEDURE AFFECTED I 3. 4. 2 10, /2, [ 3. d (5) DESC IPTION F CHANGE: (Attach u.) ce a a s y a.v. w. e ~ n u,,ditional in u.paees,if

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($)hne si!ft in ys/a hir{oranwr o f TP/J/Abiro)s Dopp4c ci,/yp,g,- (soMcisd krificabra, (6) REASON FOR CHANGE g7, y 4 !Q d YMbik $br1h3 90 Yr flsike (7) PREPARED BY - Le " DATE N (8) SAFETY EVALUATION This change:

(A) X Yes O No Represents a change to the station or procedures as described in the FSAR, or a test or experiment not described in the FSAR?

(B) O Yes %No Requires a change to the station Technical Specifications?

(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 l1 A Date 20 YC (9) REVIEW BY h -

DATE 7 b Cross Disciplinary Review By N/R (10) TEMPOR ARY APPROVAL (if Necessary)

By (SRO) Date By 0 f Date (11) APPROVED BY DATE

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(12) MISCELLANEOUS ()

Reviewed / Approved By Date Reviewed / Approved By Date AM (13) Page 1 of l

Form 34895 (6 82)

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l'0 s'44 Formusu(ms4s)

DUKE POWER COMPANY NUCLEAR SAFETY EVALUATION CHECKLIST UNIT: 1 2 X 3 Catawba (1) STATION:

OTHER:

(2) EVALUATION APPLICABLE TP/2/A/2100/01.

TO (DESCRIPTION AND NUMBER OF NSM, PRO 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:

3 Yes O No A change to the station or procedures as descnbed Table in the FSAR:

14.2.7-1 or a test (oage 31. or experiment not d Figure 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-Q Yes G No tion (s)are:This item does not require a chance to the Station Technical 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 O Yes 00 No Will the probabilit of an accident previously evaluated in : e FSAR be increased? Explain p a Nb b oO o akkkn h'fhkk tN ok k If k d

!hN 2*M!!/L*si'X*1^t*& 9J'Mbd"?n idtMRRT 19Aoulffm ideptAcgb Anfi sis aluestMMtne of DoyNbinkre$shkiplain:_lerO O Yes 00No dM*cbatween uMojSAR Msgfyk' val j 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 in Unit no 1 festing.

reason for Unit By2virtue

~

of essentially measurements toidenti differ signifal ican y rom c%Te pesigOn 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.

(6) Prepared by:

VV V

J Date: N#C

'(

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(7) Reviewed by: ~A b / Date: b (Ouhlified Reviewer)

(8)Page 2 of

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^

TABLE 14.2.7-1 (Page 3)

COMPLIANCE WITH REGULATORY GUIDES Affected Justification Compliance Section(s) Exception Taken Regulatory Guide Tests and acceptance criteria will be Control system testing should verify proper 1.68 Rev. 2 Partial App. A 5 contre.1 of process variables within the design developed to demonstrate the ability control deadband, not over the range of design of major principal plant control values of process variables. Proper control systees to automatically control pro-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.

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

.Pf I. .#W 'e*t t * *

• JV.1 O ak-bDi '.{,8abg t..,t .a t N Departure rom nucleate boiling ratio

• f 3 .

Axial, Radial, and Total Peaking will be App. A 5.b (DNBR), samlaus average planar linear directly measured and verified during power escalation testing and will be used to verify

[

heat generation rate (MAPLHGR), and einimum critical power ratio (MCPR) DNBR and linear heat rate margin by analysis.

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

App. A 5.f Core thermal and nuclear parn sters The reactor core will be under menon transient Partial conditions at this time. There would be in-will not be demonstrated * . De in.

sufficient time to gather data under transient accordance with predictions following a return of the rod to its bank position. conditions. There are no NSSS vendor predictions for this configuration.

Special testing to demonstrate control Refer to q640.52 itse 4.1 response.

App. A 5.g rod sequencers/ withdrawal block funtions operation will not be per-formed.

Rod drop times will not be measured 14easuring rod drop times at power would re-

' App. A 5.h quire disabling all position indication for at power. the rods in violation of plant Technical 'C Specifications. k From vendor predictions the Xenon and power App. A 5.1 Test to demonstrate incore/excore 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  %

or less than Technical Specifications.

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Figure 14.2.11*1 TE571W FOLL0w!E INIt!al FUEL LOACIC Mot Procritical Initial Zero Power 0%

• 55 Powee 10% - 254 Fuel Physics fest Post-Paysics restfaa power festfaa Criticality Loadina Z

1. Controlling Proc- 1. #adiation $nleiding 1. Loss of Control Initial Fuel 1. Moveaole incore 1. Initial cedure for fore Survey Acom Test (Note I Loading Detector Criticality functional fest Power Physics Testing: 2. Natur al Circulation 2. Station Sinckout

2. Incore thermo- verification fest (Note 1) couple Functional (a) Nuclear Instru-mentation Over- 3. Unit Load Steady-C' .,:-4,.

• r l, Test leo verification State fest 3, -

3. Incore thermo- i J b l' ' ' I ' ! 8 couple and AfD (c) Onset of Nuc- *4. Process and Ef f teent Cross Calibration lear Meat Radiation Monitor fest (Optional) 9

• 3 +' '. g , n (c) All tods out 5. Nts initial Catinra-

4. Rod Position Critical 8eron tion h f 9F'84'

pgjp e- ffA irlP+f' Indication Chect 1, (d) Isothereal

5. Rod Control feeperature [#5 I Cluster assembly Coefficient Drop fine fest fest

6. Rod Control (e) Olf ferential System Alignment and [ntegral fest worth of Se-quenced Coe-trol Sanus

7. Full Length Rod Drive me chanism (f) Dif ferential iioing fest Soron worta at Hot loro

8. Reactor Coolant Power System Flow Test (g) Integral Caa-trol tod acrth

9. Reactor Coolant with One Stuca Systee Flow tod ( g *

• 5 ' ,

Coastdown fest (h) Pseudo

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10. AfD Oypass Flow ed aCCA worta Vertf tcation at Not loro Power ('fg * .)

11. Pressuriter Funct-tional Iest

• The Completion of this test is not required before Initial escalation to the neat power *esting plateau. ,

NOTE 1: fests will te completed prior to escoeding the 305 testing plateau.

NOTE 2: fest will be casoleted prior to esceeding the 755 testing plateua.

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$305 F. P 1 Unit Load steady 1 Unit Load steady State feet 1 Unit Load Steady State fest

1. Unit Load Steady State fest Jnit Load State fest 2 Radiation Sh{elding 5teady State Radiation Shielding 2. NIS Initial Calibra- Survey ,

2. tion

2. Radiation Shieldig ,$urvey Radiation Survey Core Powee Ofstribu-3 Nts Initial Sht 1 ding 3. Calibration Su r rey 3 415 Initial Calibra- tion 3 est$ Initial Calf tra- tien 4. Core Power 01s-tion

• 4. Feedntee feepera- tribwtfen fest Rod Control Systes ct 4. Core Power Otstet- ture v4.ington e fest i Power T:st

4. Core Power Distribution fest bution fest t hfTt- is

5. Doppler only Power 2!$ initial 5 Power Coefficient Calibr:tf on

5. Power Coef tfetent and Power Defect and Power Defect measurement ( h;ff h cattonh'y*yt, verit)t- 5 Unit Coefficient Lead fran-sient fest Measurement (gfy*h Core Power 6. Unit Loos of

6. Unit load fran. tiectrical Load Distributton 6. Unit Lead tient fest franstant fest Test Psuedo (jec-tion Rod fist f. Incore and Nuc- 7. Peocess and (ff tvent

7. Selow Bank fest lear Instrumen-

' /. ?I f i) tation Systee Radiation penntter Power Coef- Peocess and (f- feet ficient and 8 Oetector Correla-fluent Radiation tion Support Systems Power Deftet Monitor fest 8. j Meat". resent Tur bine fetp vertf tcation fest iflit $) 8.

9. Support Systaes fest (power Unit Load vertf tcation just below fransient fest P-9 setootat) pr,gg.,.,q ::p (Note 2) te%l and PressIre Contrst fest .

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