ML19273B733

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Responds to NRC 790406 Ltr Requesting Addl Info to Verify That Only Modest Number of Fluctuations Would Be Initiated
ML19273B733
Person / Time
Site: Fort Saint Vrain Xcel Energy icon.png
Issue date: 06/01/1979
From: Justin Fuller
PUBLIC SERVICE CO. OF COLORADO
To: Gammill W
Office of Nuclear Reactor Regulation
References
P-79115, NUDOCS 7906120196
Download: ML19273B733 (12)


Text

.

pubile service company ce cchImde June 1, 1979 Fort St. Vrain Unit No. 1 P-79115 Mr. William P. Gamill Assistant Director for Standardization and Advanced Reactors Division of Project Management U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Docket #50-267

Subject:

Fort St. Vrain - Cycle 2 Fluctuation Testing

Reference:

NRC letter - T.P. Speis to J.K. Fuller dated April 6,1979 Gentlemen:

In prior comunications with the NRC, PSC has proposed that a series of core fluctuation tests be performed during the six-month period subsequent to reactor startup without insertion of the region constraint devices. The above referenced NRC letter requested additional infomation verifying that only a modest number of fluctuations would be initiated during the fluctuation test-ing. This infomation is submitted to the NRC in Attachment A.

In addition, at the PSC/NRC/GAC meeting on February 26, 1979 in Bethesda, the NRC requested infomation on expected core pcwer distribution and region outlet temperatures along with a definition of what constitutes a fluctuation event. These items are provided in Attachment B.

Please let us know if additional infomation on this subject is required.

Very truly yours ,

r s d.lc. F A 7. IN S J.K. Fuller, Vice President Engineering and Planning JKF/MLP:ler 2349 310 *I' Attachment p% gogq(o

A

%# Q Attachment A Page A-1 C0flSEQUEilCES OF PROPOSED FLUCTUATION TESTIrlG Ill CYCLE 2 PRIOR TO INSTALLATION OF REGION CONSTRAINT DEVICES During Cycle 1 operation at Fort St. Vrain a total of 30 fluctua-tion events were identified whicn comprised an accumulated operating time of about 65 hours7.523148e-4 days <br />0.0181 hours <br />1.074735e-4 weeks <br />2.47325e-5 months <br />. See Table 1 for a listing of the events, the maximum offset linear channel deviations and the maximum module main steam outlet temperature peak-to-peak variations. Of these 30 fluctuation events, five events (X, Y, Z,1, and 2) were judged to be mini-fluctuations based .

upon the small amplitude of recorded parameter deviations. These five events comprised about 61/2 hours of the total 65 hours7.523148e-4 days <br />0.0181 hours <br />1.074735e-4 weeks <br />2.47325e-5 months <br />.

In Reference ( A*), arialyses were presented that demonstrated the maximum expected loads on graphite internal core components were significantly less than the strength of these components considering up to 10,000 impact cycles. During Cycle 1, assuming a 10 minute period for the temperature fluctuations, a total of only 390 cycles occurred. Based on the above, no damage to internal core components was anticipated and has now been confirmed by the inspection during core unloading for insertion of Segment 7 and the test fuel elements. The detailed inspection of Region 35 and the core support block of Region 13 also confimed no structural damage due to fluctuations or any other causes (Ref. B*).

Additionally, tests and analyses were performed after the next to last fluctuation events of November 4,1978, which concluded there was no detectable changes in the core characteristics or operation based upon similar test data taken earlier during Cycle 1 operation. These latter tests and analyses were reported in Ref. C*.

Consequences to secondary coolant system components are inconsequential when module main steam outlet tercperatures are maintained within test limits permiting continuation of fluctuation testing. When this limit (60 F peak-to-peak amplitude) is exceeded, the test procedure requires the fluctuation 2349 311

O 2 O ^1T^ case"T ^

Page A-2 be stopped by a reduction in rear tor power. Potentially higher module main steam temperature swings can be experienced as occurred on November 4, 1978, but these are a less severe thermal transient than a reactor scram from power operation for which the plant has been designed.

During the approximately 6 months of Cycle 2 plant operation, prior to installation of the region constraint devices, only minimal fluctuation testing is planned (Ref. D*). Testing will be performed under RT-500. Issue E, to establish the fluctuating threshold line. The test procedures anticipates about 24 fluctuation events for both testing below and above 70% reactor power. The total accumulated time for operation in a fluctuating mode will be about 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />. At 10 minute periods for the fluctuation, the core and secondary system will experience about 84 fluctuation cycles.

It is, therefore, concluded that fluctuations conservatively anticipated for the first part of Cycle 2 operation will be a small fraction (14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> vs. 65 hours7.523148e-4 days <br />0.0181 hours <br />1.074735e-4 weeks <br />2.47325e-5 months <br />) of that which occurred during C. ;1e 1 and will be well within all endurance limits based upon analyses previously submitted.

REFERENCES

  • A. P-78137, " Safety Evaluation-Reactor Temperature Fluctuations", dated August 11, 1978.

B. To be supplied (In-Core Inspection Results).

C. P-79032, " Report on November 4 and December 12, 1978 Fluctuations",

dated January 29, 1979.

D. P-79094, "Startup Test Plans", dated May 11, 1979. .

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ATTACHMENT A Page A-3 TAELE 1-1

. FORT ST. VRAIN CORE TEMPERATURE FLUCTUATION EVENTS DEFINES MAXIMUM MAXIMUM 0FFSET FLUCTUATION ' PEAK TO PEAK

% POWER LINEAR CHANNEL EVENT MODULE MAIN DATE TIME START DEVIATION ", (LI!! EAR CliANNEL) STEAM - *F A* 10/31/77 2230-2315 60 - - -

B 11/23/77 1250-1440 59 , 7 Yes 44 C 11/23/77 '1730-2020 53 4 Yes 18 D 11/24/77 0044-0510 57 5 Yes 39 E 11/26/77 1550-1930 70 8 Yes 57 F 11/28/77 2040-2250 55 5 Yes 19

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G 4/22/78 0210-0540 61 10 Yes 66 H 4/22/78 2304-0028 60 9 Yes ' 52 J* 4/23/78 0038-0048 56 - - -

K/L 4/23/78 1152-1500 60 10/7 Yes/Yes '40/45 M 4/25/78 1156-1422 .65 9 Yes 62 N 4/26/78 1107-1300 55 5 Yes 82 P* 4/26/78 1356-1416 39 - - -

Q 4/26/78 1515-1705 50 3 Yes 91 R* 4/26/78 1750-1855 45 - - -

S* 5/08/78 0726-0301 68 - - -

T 5/19/78 2049-0032 66 6 Yes 66 U* 5/20/78 0655-0705 66 - - -

V 6/03/78 1000-1100 50 3 Yes 65 11 6/04/78 1140-1400 48 5 Yes 93 X 10/06/78 0255-0445 28 h No 35**(10)

Y 11/01/78 1223-1327 28 h No 38***(10)

Z 11/02/78 1407-1504 28 h No 8 1 11/02/78 2330-0127 28 h No 7 2 11/03/78 0203-0254 28 3/4 No 20** (5 )

3 11/04/78 0103-0120 43 - - -

4 11/04/78 0345-0540 52 3 Yes 97 5 11/04/78 0634-0750 39 2 Yes 29 6 12/12/78 2045-1230 67 3 Yes 22

  • Insufficient data available to determine peak to peak amplitudes.
    • Much of the change is due to an orifice adjustment at nearly the onset of the fl uctua tion. Values shown in brackets more nearly approximate the fluctuation effect.
      • Much of the 38'F change is due to trin valve adjustments. Value shown in brackets more nearly approximates the fluctuation effect.

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h ATTACHMENT B Page B-1 NRC INQUIRY l Provide a core map thowing expected power distribution and region outlet temperatures or mismatch for Cycle 2 operation.

RESP 0:lSE

. Core power distributions for Cycle 2 are shown in Figures 2-1 and 2-2.

Figure 2-1 shows the expected power distribution :d the rod positions at 64% power for 5,100 and 200 effective full power days (EFPDs).

Figure 2-2 shows the same data for Cycle 2 at 100% power. For comparison to Cycle 1, Figure 2-3 is the core power distribution and rod positicas on December 31, 1978, at '64.5% power and 158 EFPDs. In all cases t'le region power distributions were calculated using the GAUGE code.

Comparison of Cycle 1 and 2 core power distributions have been r:ade by calculating the average power of the inner, middle and outer ring of regions and by the 12 segment pie model described in response to Request for Additional Information tio. Se of the Reference *. Table 2-1 contains the data for power distribution by rings of regions. Because the partially withdrawn rod bank is shifted from the outer ring in Cycle 1 to the middle ring in Cycle 2, the outer ring of regions 20 through 37 during Cycle 2 will be at a higher average power which in turn is accompanied by a re-duction in the average power of both the inner and middle ring of regions.

Figure 2-4 shows the comparison of Cycle 1 and 2 for 64% power by the 12 segment pie model. The most significant observation is that the average power has been increased .in the lui core quadrant and decreased in the SE core quadrant for Cycle 2.

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  • G. L. L!cssman to G. Kuzmycz letter dated January 17, 1979.

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ATTACHMENT B Page B-2

. INQUIRY 1 (CONTINUED) .

Reg' ion outlet temperatures or region temperature mismatch cannot with reason-able accuracy be calculated in advance of actual reactor operation, therefore, is not being supplied. Region outlet temperatures are varied by orifice value adjustments as dictated by the need to limit core AP and to balance the steam generator modules. The region outlet temperature mismatch will be maintained within the limits of Technical Specifications LC0 4.1.7 and the power-to-flow ratio within S.L.'3.1 Figure 3.1- 2.

In general, it can be stated based upon past experience, that high powered regions will be run above core average outlet temperature and very low powered regions will run rignificantly below core average temperature.

2349 315 I

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h ATTACHMENT B Page 8-8 NRC Inquiry #2 Provide a definition of what constitutes a fluctuation event.

RESPONSE ,

There are a number of plant parameters which exhibit cyclic change; during fluctuations. The criteria used to select the parameters to

-define a fluctuation event are (1) the parameter consistently exhibits ch;nges during fluctuations, (2) the amplitude of the change is of sufficient magnitude to be observable and (3) the parameter data is displayed in the control room and, thus, available to the operators.

Based upon the above criteria and a review of 30 fluctuation events during Cycle 1 operation, a fluctuation event is best defined by cyclic changes in linear nuclear channels. The Fort St. Vrain plant is defined to be in a fluctuation operating mode when a single nuclear channoi exhibits offset deviation from the average equal to or greater than 1" of full scale reading on a cyclic basis not exceeding a 30 minute period.

Table 1, in Attach. A, is a listing of the thirty fluctuation events during Cycle 1 operation. There was sufficient data to define peak to peak amplitudes for 23 of the 30 events. Of these, all but Events X, Y, Z,1 and 2 would be classified as a fluctuation by the previously stated definition.

These five events were " mini" fluctuations with no real significance re-garding the safety of normal plant operation as evidenced by the small amplitude of module main steam outlet temperature deviations.

2349 jp7

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