ML20024C169

Suppls Info to 821230 Application for Amend to Licenses NPF-4 & NPF-7 Allowing Operation at RCS Temp of 587.8 F,Per NRC 830427 & 0621 Questions

ML20024C169
Person / Time
Site:	North Anna
Issue date:	07/06/1983
From:	Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	Clark R, Harold Denton Office of Nuclear Reactor Regulation
Shared Package
ML20024C170	List: ML20024C169 ML20024C173
References
726D, NUDOCS 8307120356
Download: ML20024C169 (4)

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VIRGINIA ELECTRIC AND POWER COMPANY RxcunoNn, VIRGINIA 202G1 W.L.STEWANT woe [.""[.'"1 oi . July 6, 1983 Mr. Harold R. Denton, Director Serial No. 726D Office of Nuclear Reactor Regulation GLD/RWC:brh:0555C rtn: Mr. Robert A. Clark, Chief Docket Nos. 50-338 Operating Reactors Branch No. 3 50-339 Division of Licensing License Nos. NPF-4 U. S. Nuclear Regulatory Commission NPF-7 Washington, D. C. 20555 Gentlemen:

SUPPLEMENT TO AMENDMENT TO OPERATING LICENSES NPF-4 AND NPF-7 NORTH ANNA POWER STATION UNIT NOS.1 AND 2 REACTOR COOLANT SYSTEM TEMPERATURE OF 587.8vF In our letter dated December 30,1982 (Serial No. 726), Vepco requested an amendment to Operating Licenses NPF-4 and NPF-7 to allow operation of North Anna Unit Nos. I and 2 at a reactor coolant system average temperature of 587.80F This letter provides in Attachment 1 supplemental infomation in answer to questions discussed with members of the Staff's Reactor Systems Branch on April 27, 1983 and June 21, 1983.

Should you have any further questions, please contact us at your earliest convenience.

_Very truly yours, 8307120356 830706 , ,

PDR ADOCK 05000338 W. L. Stewart P PDR Attachment (1) Response to Reactor Systems Branch for North Anna Operation at RCS Average Temperature of 587.80F cc: Mr. James P. O'Reilly Mr. Charlie Price Regional Administrator Department of Health Region II 109 Governor Street Richmond, Va. 23219 l

Mr. Richard Barrett i Reactor Systems Branch 00 Mr. M. B. Shymfock

, NRC Resident Inspector l \

l North Anna Power Station

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i' ATTACHMENT 1 e

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RESPONSE TO REACTOR SYSTEMS BRANCH

] FOR NORTH ANNA OPERATION

! AT RCS AVERAGE TEMPERATURE OF 587.8 F i

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l ATTACHMENT 1 RESPONSE TO NRC REACTOR SYSTEMS BRANCH QUESTIONS .

ON NORTH ANNA 7.50F UPRATING SUBMITTAL NSSS TRANSIENT ANALYSES Q1. Loss of Normal Feedwater (FSAR p.15.2-50, UFSAR p.15.2-40)

Assumption 4 in the existing FSAR states that the following was used in the calculation:

"A heat transfer coefficient in the steam generator associated with reactor coolant system natural circulation."

Why did the uprating submittal delete this assumption?

A1. The original FSAR analysis assumed a loss of offsite power occurred simula-taneously with the loss of normal feedwater, so the SG heat transfer co-efficient used was one appropriate for natural circulation conditions. A later reanalysis in response to an NRC question assumed offsite power was available. The 7.5 F0 uprating analysis employed current methodology, which is to analyze the loss of feedwater with and without offsite power available.

In each case, the SG heat transfer coefficient is calculated as a function of local SG conditions.

Q2. Loss of Normal Feedwater (FSAR p.15.2-52, UFSAR p.15.2 41)

The FSAR states that at no time during the transient does the water level in the steam generators receiving auxiliary feedwater uncover the tubesheet and there is no water relief from the pressurizer. There are figures of pressurizer and steam generator water level response in the FSAR, but no steam generator water level figure in the uprating submittal. What happens to SG water level during the transient?

A2. The loss of normal feedwater accident analyzed for the FSAR was simulated using the BLK0VT Code (Ref. WCAP-7501). The digital program computes pertinent variables including the SG level, pressurizer water level, ar.d reactor coolant average temperature; however, for the 7.50F uprating analysis, the LOFTRAN code (Ref. WCAP-7378, Rev. 3) was used to simulate the loss of normal feedwater accident. The LOFT 4 digital program computes pertinent variables including the steam generator mass versus time, pressurizer water volume, and reactor coolant average temperature. The capacity 0.f the auxiliary feedwater pumps are such that the water level in the fed steam generators _does not recede below the lowest level at which sufficient heat transfer area is available to dissipate core residual heat without water relief from the reactor coolant system relief or safety valves.

Q3. Excessive Load Increase (FSAR Fig.15.2-33, UFSAR Fig.15.2-46)

The existing UFSAR results (Fig.15.2-46) for an excessive load increase at BOL without automatic rod control show the nuclear power increasing from 1.02 to approximately 1.08 times nominal. Explain the difference between this and the uprating analysis, in which the power increases from 1.02 to 1.05 times nominal.

A3. The UFSAR analysis was performed by changing only the moderator density coefficient in the BOL and EOL cases. Current methodology analyzes the excessive load increase event for minimum and maximum reactivity feedback conditions. This requires minimum feedback values for the Doppler tempera-ture coefficient, the Doppler power defect, beta and lambda star; in addi-tion to the moderator density coefficient. Using these minimum reactivity feedback coefficients yields a lower steady state power level than that predicted in the UFSAR.

. Attached are revised FSAR Figures 15.2-33, -34, -35, 37, -38, -39, and

-40, and Table 15.2-1 which have been relabeled to reflect the minimum and maximum feedback conditions included in the 7.50F reanalyses.

Q4. Accidental Depressurization of the Main Steam System The uprating submittal includes neither a reanalysis of this event, nor a statement explaining why it was not reanalyzed.

A4. In FSAR Section 15.2.13, the existing analysis of this event assumes a steam release equal to the maximum capacity of any single steam dump or safety valve. The transient is assumed to be initfated at no load plant -

conditions, which are unchanged by the proposed 7.5 F increase in RCS average temperature. It is typically analyzed as a special case of the main steamline break accident, which is listed as one of the zero power transients requiring no reanalysis at the uprated conditions (enclosure 1, Page 14 of 7.50F uprating submittal).

QS. Main Feedline Rupture (proposed Fig.15.4.2-7A, 7B)

Why are the results of the uprating analysis (Figs.15.4.2-7A, 7B), e.g.,

pressurizer pressure and cors average temperature, less severe than the existing FSAR results (UFSAR Figs. 15.4-30,31)?

A5. The change in the florth Anna feedwater system from a headered (which was assumed in the FSAR analysis) to a one-on-one arrangement.(which was assumed in the '

-7.50F uprating analysis) is the major reason for the reduced severity of the main _feedline rupture results.

! Also, the results presented in F5AR are calculated using the MARVEL code, I whereas the results presented in the 7.50 F uprating analysis are italculated by the LOFTRAN code, using assumptions consistent with the Westinghouse Feedline Break topical WCAP-9230.

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