ML20215E408

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Forwards Analyses of Three Steam Line Break Scenarios for Reactor Bldg & Three Scenarios for Turbine Bldg Using Convective Heat Transfer Coefficient of 1.0,per NRC 861120 Request
ML20215E408
Person / Time
Site: Fort Saint Vrain Xcel Energy icon.png
Issue date: 12/12/1986
From: Warembourg D
PUBLIC SERVICE CO. OF COLORADO
To: Berkow H
Office of Nuclear Reactor Regulation
References
P-86664, TAC-42527, NUDOCS 8612220264
Download: ML20215E408 (20)


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2420 W. 26th Avenue, Suite 100D, Denver, Colorado 80211 December 12, 1986 Fort St. Vrain Unit No. 1

.P-86664 Director of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Mr. H. N. Berkow, Director Standardization and Special Projects Directorate Docket No. 50-267

SUBJECT:

Reactor Building Temperature Profiles

REFERENCE:

1) PSC letter, Warembourg to Berkow, dated February 28, 1986 (P-86120)

Dear Mr. Berkow:

Reference 1 submitted temperature and humidity profiles for three steam line break scenarios (HRH-1, HRH-2, and CRH-19) associated with the Fort St. Vrain Environmental Qualification Program. After review by Battello/PNL and discussion in our meeting with the NRC on November 20, we were requested to analyze three scenarios for the Reactor Building and three scenarios for the Turbine Building using a convective heat transfer coefficient of 1.0. The value of 1.0 is consistent with PNL's approach. We were advised to add additional volumes we could defend and remove any other justifiable items from the conservatisms.

Enclosed for your review is our first formal submittal in response to the above request. Information on the following Reactor Building steam line break scenarios is enclosed:

8612220264 e61212 gen Anock0500$7 hgo\

. .P-86664 Paga 2 December 12, 1986 HRH-2, Hot Reheat Steam Leak In Reactor Building (Offset Rupture)

CRH-19, Cold Reheat Steam Leak In Reactor Building (10% Leak Area)

CRH-14E, Cold Reheat Steam Leak In Reactor Building (0.25% Leak Area Attachment 1 provides temperature profiles, tables, and figures in the same format as submitted by the referenced letter. Humidity profiles are not included since they were not a subject of discussion and they are less severe than in the earlier submittal. This data uses a convective heat transfer coefficient of 1.0 as requested and an increased building volume as defined in the enclosure. The increased building volume includes the area above the re_ fueling floor and the area east of the 4A wall up to grade level. These areas have adequate communication with the Reactor Building and are justifiable to use without plant modification. The temperature profiles have been plotted with the following three curves for comparison:

1) Sargent & Lundy (S&L) composite profile used for equipment qualification, 2) Reference Case with variable heat transfer coefficients and old volumes, and 3) New curves with h = 1.0 and enlarged new volumes.

In addition to the changes in the par m ters discussed in the NRC meeting, there is also a change in the blowdown data for CRH-19.

Between the time the reference letter was submitted and the time the released engineering documents were completed, the termination time for the CRH-19 blowdown was revised. It is now 112 seconds rather than the previous value of 67 seconds. This is reflected in the Attachment I data.

Additional conservatisms that were evaluated include:

- Orifice coefficient for steam pipe during blowdown

- Revaporization of condensate

- Radiation heat transfer The orifice coefficient used in the reference cases (i.e., per the i information in the referenced letter) and in the reanalyzed cases is 1.0. This was believed to be conservative in that a value of 0.8 might be more realistic. Using 0.8 for the reference case of HRH-2 would reduce the peak temperature by approximately 7 degrees Fahrenheit. For the reanalyzed case, the same change in orifice

( coefficient would result in a reduction of approximately 2 degrees.

,s It is therefore concluded that the conservatism of 1.0 vs. 0.8 (or F, the orifice coefficient has no significant impact in conjunction with our reanalysis utilizing the increase in building volume.

t l

. P-86664 Page 3 D:cemb:r 12, 1986 Revaporization of the condensate was considered by calculating the effect on peak temperature for scenario HRH-2 of 8% revaporization as suggested in NUREG-0588. This factor reduces the peak temperature by approximately 17 degrees Fahrenheit using the reference case building volume and h = 1.0. For the reanalyzed case, the reduction is approximately 3 degrees Fahrenheit. We have concluded that there is some benefit that could be realized by the effects of revaporization, but we hu a not factcred this benefit directly into our reanalysis because of the magnitude of the benefit. On this basis, we have elected to leave this as a conservatism in our analysis.

The effect of radiant heat transfer was not considered during preparation of the reference case information. At the November 20 meeting, ORNL and Battelle/PNL suggested this factor could result in a radiation heat transfer coefficient on the order of 1.0 Btu /h-ft2-degrees Fahrenheit. We have independently evaluated this and concur with their conclusion. The effect on the temperature profiles utilizing a radiant heat transfer component in our analysis is shown in Attachment 2 for all three scenarios.

Again, PSC would like to stress that a heat transfer coefficient of 1.0 is an ultra-conservative value and we continue to believe that based on the information presented in our November 20, 1986 meeting that a heat transfer coefficient of 5.0 is conservative, and would be more appropriate. However, as the attached curves show, we can compensate for the ultra conservative value for the convective heat transfer coefficient by considering other factors such as building volume and radiation heat transfer, such that the original temperature profiles used for the FSV EQ Program remain appropriate.

The turbine building scenarios that are being evaluated are HRH-1, CRH-13E and CRH-15. These are scheduled to be submitted to the NRC .

by December 19, 1986 and will be forwarded by a separate submittal when completed.

If you have any questions, please contact Mr. M. H. Holmes at (303) 480-6960.

Very truly yours, f

D. W Warembour . Manager l 'l :' , -

) Nuclear Engineering' Division DWW/KD:pa Attachments l

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ATTACHMENT 1 FORT ST. VRAIN ENVIRON E NTAL QUALIFICATION PROGRAM REACTOR BUILDING TEW ERATURE PROFILES RESULTING FRCM STEAM LINE BREAKS EVALUATION OF LARGER VOLUES WITH 1.0 CCNVECTIVE HEAT TRANSFER COEFFICIENT l

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TABLE 1 DATA FOR PIPE BREAKS IN THE REACTOR BUILDING Data / Case HRH-2 CRH-19 CRH-14E

1. Broken pipe data
a. Type of fluid Steam Steam Steam
b. Temperature (oF) 1000. 740. 740.
c. Pressure (psia) 587.6 895. 895.
d. Source of the fluid S.G. & pipes S.G., pipes & S.G., pipes &

auxiliary steam auxiliary steam

e. Flow rate versus time Table 2 Table 3 Table 4
f. Enthalpy rate versus time Table 2 Table 3 Table 4
2. Compartment data
a. Number of comparments 1 1 1 b,i Initial temp. 90*F 900F 90aF b,il Initial pressure 12.3 psi 12.3 psi 12.3 psi b,ili Initial hgmidity 70% 70% 70%

b,1v Floor area Table 5 Table 5 Table 5 b,v Number of vents & vent areas (a) (a) (a) b,vi Wall height Table 5 Table 5 Table 5

c. Simple compartment diagram Figs. 1,2,3 Figs. 1,2,3 Figs. 1,2,3
3. Assumptions used:
a. Orifice coefficient 1 1 1
b. Fluid expansion factor  ; e e a
c. Heat trans'fer~ coef fici'ent for walls. Table 5 Table 5 Table 5
4. Utility analysis results:
a. Temperature versus time Fig. 4 Fig. 5 Fig. 6
b. Pressure versus time (a) (a) (a)

(a)An "open building" calculation was performed, meaning that the building pressure was held constant (12.3 psia) and, at each time step, an appropriate mass of mixed air and steam exchange with the environment was calculated to maintain that pressure.

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TABLE 2 HRH-2 FLOW AND ENERGY RELEASE VERSUS TIME

Description:

Hot Reheat Steam Leak in Reactor Building Reference Input: Run ST8680, 9/4/85 at 19:24:57 (Flash /GA run)

Flow Rate Enthalpy Rate

_(sec) (hr) (lb/hr) (Btu /hr) 0 0 0 0 0.1 2.7778 x 105 10.020 x 106 14.853 x 109 0.12 3.3333 x 10 5 10.244 x 106 15.087 x 109 0.201 5.5833 x 10-5 9.081 x 106 13.332 x 109 0.381 1.0583 x 10-4 6.739 x 106 9.817 x 109 0.560999 1.5583 x 10-4 -5.487 x 106 8.062 x 109 0.920997 2.5583 x 10-4 4.320 x 106 6.432 x 109 1.46099 4.0583 x 10-4 3.548 x 106 5.317 x 109 2.40184 6.6718 x 10-4 3.093 x 106 4.660 x 109 4.00008 1.1111 x 10-3 2.906 x 106 4.403 x 109 8.00052 2.2224 x 10-3 2.433 x 106 3.715 x 109 10.0016 2.7782 x 10-3 1.144 x 106 1.762 x 109 11.0008 3.0558 x 10-3 348.8 x 103 547.8 x 106 12.0016 3.3338 x 10-3 131.3 x 103 202.6 x 106 13.0049 3.6125 x 10-3 22.47 x 103 34.05 x 106 13.2170 3.6714 x 10-3 0 0 m = 0 0 t

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TABLE 3 -

CRH-19 FLOW AND ENERGY RELEASE VERSUS TIME

Description:

Cold Reheat Steam Leak in Reactor Building through 10% Leak Area Reference Inputs Run ST6965, (Flash /GA)

Time Flow Rate Enthalpy Rate (sec) (lb/hr) (Btu /hr) 0.00000 0.00000 0.00000 4.31952 - 02 6.76724 + 05 9.19545 + 05 8.80165 - 02 1.61093 + 06 9.19545 + 08

.13349 1.99831 + 06 2.70948 + 09

.17932 2.00681 + 06 2.72152 + 09

.38290 2.00069 + 06 2.71502 + 09

.96548 1.96029 + 06 2.65647 + 09 1.5455 1.92229 + 06 2.60334 + 09 2.1238 1.88949 + 06 2.55776 + 09 2.7011 1.85917 + 06 2.51666 + 09 3.3001 1.83479 + 06 2.48449 + 09 3.8767 1.81768 + 06 2.46315 + 09 4.4532 1.80464 + 06 2.44811 + 09 5.0300 1.79510 + 06 2.43856 + 09 5.6257 1.78525 + 06 2.42817 + 09

-6.0904 1.77845 + 06 *

,-2.41940 + 09 6.4904 1.76862 + 06' 2.40478.+ 09 l

, 6.8904~ 1.74870 + 06 2.37501 + 09 '

7.2904 1.71957 + 06 2.33178 + 09

, 7.6904 1.68288 + 06 2.27819 + 09 8.0904 1.64815 + 06 2.22895 + 09 8.4904 1.61247 + 06 2.17977 + 09 8.8904 1.57663 + 06 2.13143 + 09 9.2904 1.54341 + 06 2.08798 + 09 9.6904 1.50897 + 06 2.04338 + 09

TABLE 3 (Continued)

Time Flow Rate Enthalpy Rate (sec) (1b/hr) (Btu /hr) 10.390 1.44849 + 06 1.96533 + 09 11.390 1.37051 + 06 1.86566 + 09 12.390 1.30115 + 06 1.77778 + 09 13.300 1.24720 + 06 1.71168 + 09 14.390 1.19969 + 06 1.65181 + 09 15.390 1.16019 + 06 1.59658 + 09 16.390 1.12334 + 06 1.53799 + 09 17.390 1.08396 + 06 . 1.'47114 + 09 18.390 1.04987 + 06 1.40969 + 09 19.390 1.02313 + 06 1.35618 + 09 20.890 9.91221 + 05 1.28767 + 09 22.00 9.60 + 05 1.23 + 09 112.00 0.00 0.00 a 0.00 0.00 t

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TABLE 4 CRH 14E " LOW AND ENERGY RELEASE VERSUS TIME

Description:

Cold reheat leak in Reactor Building through 0.25% leak area Reference Input: Hand Calculation, GA Doc. 908838 Time Flow Rate Enthalpy Rate (sec) (1b/hr) (Btu /hr) 0.00 0.00 0.00 0.01 5.15 + 04 7.00 + 07 4500 5.15 + 04 7.00 + 07 6178 0.00 0.00 m 0.00 0.00

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TABLE 5 REACTOR BUILDING Volume = 1,396,980 ft3 Heat Transfer Heat Sink Wall Thick. Area Coefficient Outside(a)

Surface (in.) (ft2) (Btu /h-ft2 op)

1. Concrete walls and floor, PCRV 36. 51,670 0
2. PCRV support ring 21. 9,870 0
3. Concrete partition walls, floors 12. 9,880 0
4. Thin steel wall 0.05 17,600 2
5. Composite steel wall 5.25 50,240 6
6. Steel decking 0.0936 50,840 0
7. Structural steel and equipment 0.375 47,060 0
8. Ducting, electrical conduits 0.0312 80,950 0 Cable trays
9. Piping 0.375 62,520 0
10. Concrete partition walls, floors 6. 6,280 0
11. Concrete partition walls, floors 18. 3,440 0
12. Concrete sealed rooms, regions 36. 5,940 2
13. Keyway walls 24. 2,120 0 j 14. Steel wall, 4a 0.05 5,880 0 a

(a) Locations that signify the inside and outside heat transfer coefficients are

illustrated in Fig. 1.

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