ML18019A415

Forwards Draft Changes to FSAR Section 14.2 Re Initial Startup Testing Program.Changes Concern Control Rod Reactivity Worth & Pseudo Rod Ejection Tests.Changes Will Be Incorporated Into Future Amend to FSAR

ML18019A415
Person / Time
Site:	Harris
Issue date:	10/03/1985
From:	Zimmerman S CAROLINA POWER & LIGHT CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
NLS-85-291, NUDOCS 8510080439
Download: ML18019A415 (15)

Text

REGU TORY INFORMATION DISTRIBU "N SYSTEM (RIDS)

P

, ACCESSION NBR;8510080439 DOC ~ DATE ~ 85/10/03 NOTARIZED:

NO

.'- FACIL:50-400 Shearon Harris Nuclear Power Pl anti, Unit i~ Carolina AUTH,NAME AUTHOR AFFILIATION ZIMMERMANg8.R~

Car ol ina Power

~ L Light Co, RECIP,NAMEl RECIPIENT AFFILIATION DENTON HER Office~ of Nuclear Reactor", Regulationi Director DOCKEiT 05000400 SUBJECT;. Forwards dr aft changes to FSAR Section 14,2 re initial

-startup testing progr am.Changes concern control rod reactivity worth 8 pseudo rod ejection, tests. Changes" will be incorporated into future amend to FSAR ~

DISTRIBUTION CODE:

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T'ITLE'. Licensing Submittal:

PSAR/FSAR Amdts 8 Re.lated Correspondence",

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'tI Carolina Power & Light Company OCT 0 S >985 SERIAL: NLS-85-291 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation United States Nuclear Regulatory Commission Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT UNIT NO.

1 - DOCKET NO.50-000 CONTROL ROD REACTIVITYWORTH AND PSEUDO ROD E3ECTION TESTS

Dear Mr. Denton:

Carolina Power dc Light Company (CPdcL) submits for your review two changes to the Shearon Harris Nuclear Power Plant (SHNPP)

Initial Start-up Testing Program as described in FSAR Section 10.2.

The change to the Control Rod Reactivity Worth Test Summary eliminates the requirement of measuring the N-1 rod worth during Initial Start-up testing and allows the use of either the boron dilution or the rod swap technique for measuring control rod worths.

The other change requests deletion of the pseudo rod ejection tests from the Initial Start-up Program.

Historical calculations and

tests, along with SHNPP being a standard Westinghouse three-loop plant with a typical first-core fuel design, indicate the need for these tests and measurements to verify core physics models does not exist.

Deletion of these tests is also responsive to NRC, INPO, and CPdcL concerns over needlessly placing the plant in abnormal operating configurations which may potentially lead to severe radial power tilting.

Carolina Power 2 Light Company has discussed these revisions with the NRC Staff, and received concurrence with these changes.

Draft FSAR changes are provided for your information.

These changes will be formally incorporated into the FSAR in a future amendment.

~so 85~ooO'ga 8Saoos oc@ osonq~

A 411 Fayettevilte Street o P. O. Box 1551

~ Raleigh, N. C. 27602

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~ Qr. Harold R. De NLS-85-291 / Page&

If you have any questions, please contact Mr. Gregg A. Sinders at 919-836-8168.

Yours very truly, GAS/rtj (1836GAS)

Attachment S.. Zim erman ger Nuclear Licensing Section CC:

Mr. R. A Becker (NRC-PSRB)

Mr. B. C. Buckley (NRC)

Ms. M. Chatterton (NRC-CPB)

Mr. D. B. Fieno (NRC-CPB)

Mr. G. F. Maxwell (NRC-SHNPP)

Dr. 3. Nelson Grace (NRC-RII)

Mr. H. 3. Richings (NRC-CPB)

Mr. G. A. Schwenk (NRC-CPB)

Mr. Travis Payne (KUDZU)

Mr. Daniel F. Read (CHANGE/ELP)

Wake.County Public Library Mr. Wells Eddleman Mr. 3ohn D. Runkle Dr. Richard D. Wilson Mr. G. O. Bright (ASLB)

Dr. 3. H. Carpenter (ASLB)

Mr. 3. L. Kelley (ASLB)

Mr. H. A. Cole

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SHNPP FSAR 14.2.7 CONFORMANCE OF TEST PROGRAMS WITH REGULATORY GUIDEShe foLlowing appLicabLe regulatory guides wilL be used as guidance in development of the initiaL test program'.

a)

Regulatory Guide 1.20, Rev. 2, May,

1976, Com rehensive Vibration Assessment Program for Reactor Internals Durin Prep erational and InitiaL Startu Testin Re uirements for CLeanin of Fluid S stems and Associated Com onents of Mater-Cooled Nuclear PLants.

c)

Regulatory Guide 1.41, Rev. 0, March,

1973, Prep erational Testing of Reduncant On-Site Electric Power S stems to Verif Pro er Load Grou Assignments.

d)

Regulatory Guide 1.52, Rev. 2, March,

1978, Design, Testing, and Maintenance Criteria for En ineered Safet Feature Atmos here CLeanu S stem Air FiLtration and Absor tion Units of Li ht-Water-Cooled NucLear Power Plants.

e)

Regulatory Guide 1.68, Rev. 2, August, 1978, Initial Test Programs for Mater-CooLed NucLear Power PLants

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ReguLatory Guide 1.68.2, Rev.

1, Suly, 1978, Initial Startu Test I 22 Program to Demonstrate Remote Shutdown Ca abiLit for Water-CooLed Nuclear Power Plants.

g)

Regulatory Guide 1.79, Rev.

1, September,

1975, Preo erationaL Testin of Emergenc Core Coolin S stems for Pressurized Water Reactors with the following clarifications exceptions.'e

. Position CLarifications/Exce tions C.L.b.(2)

The capabiLity to reaLign vaLves for recirculation shaLL be tested for the pLant.

Test of a recirculation sump to demonstrate vortex control, acceptable pressure drops across suction lines and valves, and adequate NPSH wiLL be conducted for the plant by modeL tests.

CP&L will verify by appropriate physical examination and fLow demonstration test that recirculation sump suction Lines are not obstructed and that valves are properly installed.

14.2.7-1 Amendment No.

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SHNPP FSAR Pow<<r Co<<ff icient and Power Defect Measurement Test Summary Control Rod Reactivity Worth Test Sum)nary 13.

Boron React/.vity Worth Test Summary 14.

Automatic Rod Control Test Summary 15.

16.

Steam Generator Moisture Carryover Test Summary 17.

Load Swing Test Summary 18.

Large Load Reduction and Generator Trip From 100 Percent Power Test Summary 19.

20.

Turbine Trip From 100 Percent Power Test Summary Remote Shutdown Test Summary 21.

-Loss of Offsite Power Test Summary

" '22.

23.

24.

25.

Pressurizer Heaters and Spray Valves Capability Test Summary Failed Fuel Detection Test Summary Pressurizer Continuous Spray Flow Verification Test Sunnnary Reactor Coolant 'System Leakrate Test Summary 26.

27.

Natural Circulation Test Summary Main Steam and Feedwater Systems Test Summary 28.

Shield Survey.Test Summary 29.

Loss of Feedwater Heaters Test Summary 30.

31.

Main Steam Isolation Valve Test Summary Steam Generator Test for Condensation-Induced Water Hammer 8

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Amendment No 17t'v

SHNPP FSAR 14.2.12.2.12 Control Rod Reactivity Worth Test Summary a)

Test Oh)ective 1)

To determine the Integral Rod Reactivity Worths

+ha shut'ua of~control rods and ~~ rod banks, and to verify by analysis that the Rod Insertion limits will be adequate to ensure a shutdown margin consistent with accident analysis assumptions with the greatest worth control rod stuck out of the core.

b)

Prerequisites 1)

The reactor is critical at zero power.

2)

The general prerequisites are met.

c)

Test Method co+~i t od ~or+h5 wyatt be delev~iTl+8 olhd cowt 4vcd ~ Qggqg p ~dicHp~s b

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Acceptance Criteria 1)

The rod worths are determined to be within the limits of FSAR Table 4.3.2-3 r

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t I'0 7 lo 4K /Y 14.2.12.2.13 Boron Reactivity Worth Test Summary a)

Test Objective 1)

To determine the boron reactivity worth over the boron concentration ranges in which the reactor may be taken critical.

b)

Prerequisites c) 1)

The reactor is critical at zero power.

2)

The general prerequisites are met.

Test Method 1)

Determine the reactivity worth of the boron in solution by diluting the boron concentration and compensating for the reactivity effect by movement of control rods.

14 2.12-86 Amendment No.

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- 'f Sh.HAPP PSAR Acceptance Criteria 14.2. 12.2. 15

~ud~~~et-M ey Dele.he 3 1)

The Rod Control System naintains stability under steady state conditions in accordance with the precautions, limitations and setpoint document.

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To rify at the hot encl factors resultin~ from a sinulat ect rod cont" er assenb n accept s ~

ereq isit

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Excore an encore nuclear inst umentation and incore t ouple a e operabl 2)

Reac or is rit cal at zero power level.

3) e ge ral prerequ re met.

c) st eth d 1

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> imul wit 'drawal s perfo ed at approxima ely i

e a od e e tion a ident at n RCCA.

ero and 30 percent power.

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ptance Criteria 1

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sem safety a alys qP han el ly are s) Hesti house mclear Desi n nual.

acr rs esu' ng "om the rawn ro c

ster o &iste t w yth valises as Gmed i the plant t

14.2. 12.2. 16 Steam Generator Moisture Carryover Test Summary a)

Test Objective t

1)

To determine the moisture carryover performance of the stean generators.

b)

Prerequisites 1)

Power level is established as required by test procedure.

2)

The genera1 prerequisites are met.

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c)

', Test Method

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Inject a radioactive tracer.into the steam geenerator and perform

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.* activity analysis of selected water and steam samples.

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SHNPP FSAR generators No credit is taken for the possible pressure reduction caused by the assumed failure of the control rod pressure housing.

15.4.8.2.2

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Calculation of Basic Parameters Input parameters for the analysis are conservatively selected on the basis of values calculated for this type of core.

The more important parameters are discussed below.

Table 15.4.8-2 presents the parameters used in analysis-E 'ected Rod Worths and Hot Channel Factors The values for ejected rod worths and hot channel factors are calculated using either three dimensional static methods or by a synthesis method employing one d&ensional and two dimensional calculations.

Standard nuclear design codes are used in the analysis.

No credit is taken for the flux flattening effects of reactivity feedback.

The calculation is performed for the maximum allowed bank insertion at a given power level, as determined by the rod insertion 1&its.

Adverse xenon distributions are considered in the calculation.

Appropriate margins are added to the ejected rod worth and hot channel factors to account for any calculational uncertainties, including an allowance for nuclear power peaking due to densification.

Power distributions before and after e)ection for a "worst case" can be found configuratio~nd-eom pared-to-valses-used-in the-anabys~ Xt has-been'hat-tk~eee~d-wor th-end-powerpeakingMac4:ors-ave-consistently, 4 ~

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i3'gee Reacti~vit Feedback Ref htin Factors s4arii' f evi geng+

0 ad 5 8N pl Novi' reac 4ot'o<<

4 i'>ch is ntvMnicelhj si nai lioi'0 a4~ ujts4iqhog5< p-ioyp pjaq+g na d rrq4 ~eri pi'<<4.-

o p $ 4 t The largest temperature

rises, and hence the largest reactivity-feedbacks design prod:c4'~5.

occur in channels where the power is higher than average.

Since the weight od~"'n') "~ '

region is dependent on flux, these regions have high weights.

This means ~

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y that the reactivity feedback is larger than that indicated by a simple channel analysis.

Physics calculations have been carried out for temperature changes with a flat temperature distribution and with a large number of axial and radial temperature distributions.

Reactivity changes were compared and effective weighting factors determined.

These weighting factors take the form of multipliers which when applied'o single channel feedbacks correct them to effective whole core feedbacks for the appropriate flux shape.

In this

analysis, since a one dimensional (axial) spatial kinetics method is employed, axial weighting is not necessary if the initial condition is made to match the ejected rod configuration.

En addition, no weighting is applied to the moderator feedback.

A conservative radial weighting factor is applied to the transient fuel temperature to obtain an effective fuel temperature as a function of time accounting for the missing spatial dimension These weighting factors have also been shown to be conservative compared to three dlmensional analysis (Reference 15.4. 8-1) ~

15.4.8-6