ML19094A144

Supplemental Information to Amendment to the Operating License Technical Specifications Change No.47

ML19094A144
Person / Time
Site:	Surry
Issue date:	03/04/1977
From:	Stallings C Virginia Electric & Power Co (VEPCO)
To:	Reid R, Rusche B Office of Nuclear Reactor Regulation
References
Download: ML19094A144 (29)

Text

{{#Wiki_filter:e VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND,VIHGIN.IA 23261 March 4, 1977 Mr. Benard.C. Rusche, Director Office of Nuclear Reactor Regulation Attn: Mr. R. W. Reid, Chief Serial No. 219/082776 Operating Reactor Branch No. 4 U. S. Nuclear Regulatory Commission Washington, D. C. 20555 FR/RWC:agp Docket Nos. License Nos. 50-280 50-281 DPR-32 DPR-37

Dear Mr. Rusche:

SUPPLEMENTAL INFORMATION TO AMENDMENT TO THE OPERATING LICENSE TECHNICAL SPECIFICATIONS CHANGE NO. 47 SURRY POWER STATION UNITS 1 AND 2 We have updated our previously submitted report, "Large Break LOCA-ECCS Reanalysis for Surry Units No. 1 and 2," which was transmitted in our letter dated October 29, 1976 (Serial No. 219/082776) to include an additional LOCA-ECCS accident case in which the same assumptions as in our previous limiting case were used except that a uniform steam generator tube plugging level of 20 percent (instead of 15 percent) was assumed. contains the results of this additional case and the associ-ated updated.text (on a one-for-one replacement page basis)*for.the above reference. It should be noted that the results of this additional case meet, with margin, the limiting criteria required by 10CFRS0.46. Should you have any questions or comments, we would.be most happy to meet with you at your earliest convenience. Attachment (40 copies) cc: Mr. Norman C. Moseley Very truly yours, ._Jl J/)ri C r Lt},//tJ( >C)7d&:,,~~ (/ C. M. Stallings ... Vice President-Power Supply and Production Operations '2495

-I VIRGINIA ELECTRIC AND POWER RICHMOND,VIBOIN.IA 23261 March 4, 1977 Mr. Benard C. Rusche, Director Office of Nuclear Reactor Regulation Attn: Mr. R. W. Reid, Chief . Operating Reactor Branch No. 4 U. S. Nuclear Regulatory Commission Washington, D. C. 20555

Dear Mr. Rusche:

Serial No. 219/082776 FR/RWC :_agp Docket Nos. License Nos. 50-280 50-281 DPR-3"2 DPR-37 SUPPLEMENTAL INFORMATION TO AMENDMENT TO THE

• OPERATING LICENSE TECHNICAL SPECIFICATIONS CHANGE NO, 47 SURRY POWER STATION UNITS 1 AND 2 We have updated our previously*submitted report, "Large Break LOCA-ECCS Reanalysis for Surry Units No. 1 and 2, 11 which was transmitted in our letter dated October 29, 1976 (Serial No. 219/082776) to include an additional LOCA-ECCS accident case in which the same assumptions as in our previous limiting case were used except that a uniform steam generator tube plugging level of 20 percent (instead of 15 percent) was assumed. contains the results of this additional case and the associ-ated updated text (on a one-for-one replacement page basis) for the above reference, It should be noted that the results of this additional case meet, with margin, the limiting criteria required by 10CFRS0.46.

Should you have any questions or comments, we would be most happy to meet with you at your earliest convenience. Attachment (40 copies) cc: Mr. Norman C, Moseley Very truly yours, ~)1?. ~tM/~= C. M. Stallings Vice President-Power Supply and Production Operations 2495

e ATTACHMENT 1 REPLACEMENT PAGES AND SUPPLEMENTAL TABLES AND FIGURES FOR THE REPORT "LARGE BREAK LOCA-ECCS REANALYSIS FOR SURRY UNITS NO. 1 AND 2"

e e Instructions for the update of the previously submitted report. "Large Break LOCA-ECCS Reanalysis for Surry Units No. 1 and 2" which was submitted October 29, 1976, Serial No. 219/082776.

1.

Pages 6, 7, and 8 are submitted as one for one replacement pages for the existing pages 6, 7, and 8 of the report.

2.

Table 1 (continued), Table 2e, Table 3e, Table 9a, Table 9b and Figures lg to 17g are an update of the report and supple-ment the respective existing tables and figures.

e SATAN-VI and WREFLOOD codes. In addition, conservatively chosen initial containment conditions and an assumed mode of operation for the containment cooling system are input to COCO. 4.0 RESULTS Based on the results of the LOCA sensitivity studies, (7, 12, l4) the limiting large break was found to be the _double-ended cold leg guillotine (DECLG) break of the RCS. Therefore, only the DECLG break results are reported. The results of five sets of initial operating conditions are pro-vided for the present reanalysis. Table 1 defines Cases A, B, C, D, and E for the five sets of initial operating conditions that have been assumed for this reanalysis. Tables 2a, 2b, 2c, 2d, and 2e present the time sequence of events for Cases A, B, C, D, and E, respectively. Tables 3a, 3b, 3c, 3d, and 3e present the results and the major initial conditions for the five cases which are discussed in more detail below. The base case, Case A, assumed a steam generator tube plugging level of 7 percent per steam generator and minimum accumulator water level of 975 ft 3* Table 3a presents results for the DECLG for the values of three discharged coefficients (Cn), This range of discharge coefficients was determined to include the.limiting case for peak clad temperature from sensitivity studies reported in WCAP-8356(7), WCAP-8572(lZ), and WCAP-8853(l4), The limiting base case (Case A) break, as in the previous analysis, was found to be the 0 Cn = 0. 4 break and resulted in a peak clad temperature of 2074.,_F, a maximum local metal-water reaction of 5.6 percent, and total core metal-water reaction of less than 0.3 percent. An additional analysis was then conducted with the limiting (Cn = 0.4) break size in order to determine the sensitivity of increased steam generator tube plugging on peak clad temperature. The Case B assumptions were the same as for Case A except that the steam generator tube plugging level was assumed to be 10 percent. These results are presented in Table 3b and indicate a peak clad temperature of 2091°F, a maximum local metal-water.reaction of 5.9 per-cent, and a total core metal-water reaction of less than 0.3 percent. Case C, assumed a change in accumulator water volume from 975 ft3 to 1075 ft3

• This was done because it was found that for steam generator tube plugging approaching 11 percent and an accumulator water volume of 975 ft3 the peak clad temperature slightly exceeded 2200°F.

In order to obtain more margin in peak clad temperature to accommodate the potential need for higher steam generator tube plugging levels, the effect of an increase in accumulator water volume was investigated and found to be beneficial. Table 3c provides results for Case C which assumed a steam generator tube plugging level of 12 percent and an accumulator water volume of 1075 ft3

• The results indicate a peak clad temperature of 2107°F, a maximum local metal-water reaction of 6.2 percent, and a total core metal-water reaction of less than 0.3 percent.

Case D was the same as Case C except that the steam generator tube plugging level was assumed to be 15 percent. The results indicate a peak clad temperature of 2120°F, a maximum local metal-water reaction of 6.7 per-cent, and a total core metal-water reaction of less than 0.3 percent. Case E was the same as Cases C and D except that the steam generator tube plugging level ~as assumed to be 20 percent. The results indicate a peak clad temperature of 2186°F, a maximum local metal-water reaction of 7,9 percent, ~nd a total core metal-water re~ction of less than 0.3 percent. Finally, for information purposes, an analysis was conducted to determine the impact of asymmetric steam generator tube plugging on peak clad temperature. Limiting asymmetric conditions were investigated and the result indicated no adverse impact on the limiting peak clad temperature. Currently, less than 20% of the steam generator tubes are plugged in the Surry units and e e the plugging distribution is essentially symmetric. The detailed results of the LOCA reanalysis for Cases A, B, C, D, and E are provided in Tables 1 through 9 and Figures 1 through 18.

5.0 CONCLUSION

S For breaks up to and including the double ended severance of a reactor coolant pipe and for the operating conditions specified by Cases A, B, C, D, and E, the Emergency Core Cooling System will meet the Acceptance Criteria as presented in 10CFRS0.46. That is:

1.

The calculated peak fuel element clad temperature is below the requirement of 2200°F.

2.

The amount of fuel element cladding t~at reacts chemically with water or steam does not exceed 1 percent of the total amount of Zircaloy in the reactor.

3.

The clad temperature transient is terminated at a time when the core geometry is still amenable.to cooling. The localized cladding oxidation limits of 17% are not exceeded during or after quenching.

4.

The core remains amenable to cooling during and after the

break,

5.

The core temperature is reduced and decay heat is removed for an extended period of time, as required by the long-lived radioactivity remaining in the core. Case D Case E Table 1 (continued) This case extended the results of Case C to indicate the effect on peak clad temperature of increased steam generator tube plugging. The following assumptions were made:

1.

total peaking factor (Fq) of 2.0

2.

minimum temperature of the containment of 90°F

3.

uniform steam generator tube plugging of 15 percent

4.

accumulator water volume of 1075 ft 3 *(per accumulator)

5.

temperature of the fluid in the reactor vessel upper head region equal to 100 percent of THOT* This case extended the results of Cases C and D to indicate the effect on peak clad temperature of increased steam generator tube

plugging, The following assumptions were made:

1.

total peaking factor (Fq) of 2.0

2.

minimum temperature of the containment of 90°F

3.

uniform steam generator tube piugging of 20 percent

4.

accumulator water volume of 1075 ft 3 (per accumulator)

5.

temperature of the fluid in the reactor vessel upper head region equal to 100 percent of THOT

TABLE 2e TIME SEQUENCE OF EVENTS START Reactor Trip Signal S, I. Signal Acc. Injection End of Bypass End of Blowdown CASE E Bottom of Core Recovery Acc. Empty Pump Injection DECLG (Cn=0.4) o.o 0.648 2.23 15.8 23.69

26. 77 37.38 55.31 27.23

TABLE 3e RESULTS - CASE E Results Peak Clad Temp., OF Peak Clad Location, ft Local Zr/H20 Rxn(max), % Local Zr/H20 Location ' ft. Total Zr/H20 Rxn, % Hot Rod Burst Time, sec. Hot Rod Burst Location, ft. Initial Conditions Core Power, Mwt, 102% of Peak Linear Power kw/ft 102% of Peaking Factor Accumulator Water Volume (ft3) Most Limiting Fuel Region Unit 1 Unit 2 2441 12.49 2.00 DECLG (Cn=0.4) 2186 9.0 7.932 9.0 <0.3

27. 0 6.0 1075 (per accumulator)

Cycle All All Region 4 4

Time (sec) 37.38 38.30 38.45 43.81 51.4 51.46 53.51 53.55 67.66 84.66 103.67 124.26 170.26 225.26 294.16 392.76 e TABLE 9a REFLOOD MASS AND ENERGY RELEASES FOR THE DECLG (Cn=0.4) BREAK AND 20% STEAM GENERATOR TUBE PLUGGING Total Mass Flow Rate Total Energy Flowrate (lbm/Sec) (105 Btu/Sec) 0.0 0.0 o.o 0.0

1. 076 0.014 33.69 0.438 33.69 0.438 2894.

3.07 2894. 3.07 250.8 1.42 250.8 1.42 261.8 1.39 268.9 1.36 275.1 1.32 291.4 1.19 308.8 1.06 232.9 0.925 337.4 0.827

• 1

t, e TABLE 9b BROKEN LOOP ACCUMULATOR FLOW TO CONTAINMENT FOR LIMITING CASE AT 20 PERCENT PLUGGING - DECLG (CD= 0.4) Time (Sec) 0.0 1.0 3.0 5.0 7.0 10.0 15.0 20.0 23:01 32.27 Mass Flowrate* (Lbm/Sec) 4107 3738 3232 2891 2639 2355 2024 1800 1703 o.o

• For energy mass flowrate multiply mass flowrate by a constant of 58 Btu/Lbm.

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