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Evaluation of Confinement Environ Temp Following High Energy Line Breaks Proposed for Fort St Vrain Environ Qualification Program, Informal Technical Evaluation Rept
	ML20214Q541
Person / Time
Site:	Fort Saint Vrain
Issue date:	03/31/1987
From:	Wheeler C, White M; EG&G IDAHO, INC., IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY
To:	; NRC
Shared Package
ML20214Q522	List: ML20214Q519; ML20214Q541;
References
	CON-FIN-D-6023 EGG-NTA-7564, PLN-6182, NUDOCS 8706050066
	Download: ML20214Q541 (61)
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' EGG-NTA-7564 March 1987 l

INFORMAL REPORT TECHNICAL EVALUATION REPORT g L /daho - EVALUATION OF CONFINEMENT ENVIRONMENTAL National . TEMPERATURES FOLLOWING HIGH ENERGY LINE a Engineering BREAKS PROPOSED FOR THE FORT SAINT VRAIN ENVIRONMENTAL QUALIFICATION PROGRAM Laboratory Managed by the U.S. M. D. White Department C. L. Wheeler of Energy .

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DISCLAIMER This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, or any of ths-ir employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for any third party's use, j of any information, apparatus, product or process disclosed in this report or represents that its use by such third party would not infringe privately owned rights.

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e TECHNICAL EVALUATION REPORT 9 EVALUATION OF CONFINEMENT ENVIRONMENTAL TEMPERATURES FOLLOWING HIGH ENERGY LINE BREAKS PROPOSED FOR THE FORT SAINT VRAIN ENVIRONMENTAL QUALIFICATION PROGRAM Published March 1987 J

M. D. White C. L. Wheeler Pacific Northwest Laboratory Richland, Washington 99352 1

Prepared for the

- U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Under DOE Contract No. DE-AC06-76RLO 1830 FIN No. D6023 4

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. I ABSTRACT j l

l COBRA-NC simulations were performed of the high energy line break scenarios '

HRH-1, CRH-19, HRH-2, and CRH-15 in conjunction with the Pubic Service of Colorado Company's Fort Saint Vrain Environmental Qualification Program. The simulations comprise the amended heat sink areas and volumes specified for the Turbine and Reactor buildings. Consideration of radiation heat transfer processes between the confinement gas and heat sinks was incorporated into the scenario simulations. The confinement environment average temperature history plots never exceed the Sargent and Lundy composite profiles used for equipment qualification.

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4 CONTENTS INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I MODEL DESCRIPTION . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 RESULTS 6 REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 APPENDIX A . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A.1 o

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FIGURES 1 Turbine Building, Steam Release Scenario HRH-1 (Double Offset) ... 8 2 Turbine Building, Steam Release Scenario CRH-1 (25% Leak Area) ... 9 3 Reactor Building, Steam Release Scenario HRH-2 (Double Offset) ... 10 4 Reactor Building, Steam Release Scenario CRH-19 (10% Leak Area) . . . 11 e

TABLES 1 Wall Type Descriptions . . . . . .................. 4 2 Convection and Radiation Conductances . . . . . . . . . . . . . . . . 6 3 Peak Simulation Temperatures . . .................. 7 l

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i NOMENCLATURE at,a2 coefficients j A surface area b self-broadening coefficients Eb blackbodyedissivepower h heat transfer coefficient Le mean beam length a p pressure q heat transfer rate

, T temperature I V confinement volume X emissivity equation parameter e emissivity a Stefan-Boltzmann constant !

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INTRODUCTION This report documents the evaluation of environmental conditions within confinement structures of the Fort Saint Vrain Nuclear facilities, following several proposed high energy line break scenarios. The present evaluation differs from the previously submitted documentation by Pacific Northwest Laboratory (a) due to amendments in the confinement structural descriptions.

The original analyses of these high energy line break scenarios performed by a Gulf Atomic, representing Public Service of Colorado, differed with the results generated by PNL, representing the U.S. Nuclear Regulatory Commission (NRC), because of differences in natural convection heat transfer coefficients. The heat transfer coefficient determination methodologies used by Gulf Atomic were reviewed and determined, by PNL, to be nonconservative(U) .

The environmental temperatures calculated by PNL exceeded the limits for the Environmental Qualification Program, while the temperatures and pressures calculated by Gulf Atomic were within qualification limits. Following discussion of the opposing views in regard to the most appropriate natural convection heat transfer coefficients, the NRC decided that the evaluations would be repeated with the more conservative heat transfer coefficients, but with the inclusion of previously unaccounted heat sinks within the confinement structures. In addition Public Service and PNL agreed that radiation from the steam environment to the confinement surfaces should also be considered.

This document presents the results of COBRA-NC (Wheeler, Thurgood, Guidotti, and DeBellis,1986) simulations on the four most severe high energy line break scenarios entitled HRH-1, CRH-15, HRH-2, and CRH-19. The first n

(a) The initial Pacific Northwest Laboratory evaluation of Fort Saint Vrain

. LOCA was documented in: Wheeler, C. L., R. E. Dodge, and J. R. Skarda, 1986. " Independent Calculation of Pressure and Temperature Profiles for a High Energy Line Break Outside Containment Fort Saint Vrain Nuclear

. Generating Station Unit 1, FATE-86-114, Pacific Northwest Laboratory, Richland, Washington.

(b) White, M. D., 1986. " Review of Convection Heat Transfer Coefficients Utilized in the Fort Saint Vrain Main Steam Line Break Analyses",

FATE-86-117, Pacific Northwest Laboratory, Richland, Washington.

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d two consider breaks within the turbine building and the others breaks within the' reactor building. All of the scenarios were simulated with the amended confinement structure descriptions. The specific confinement heat sink modifications were reported in two letters between Public Service Company of

Colorado (PSCC) and NRC (see appendix A). In addition to these heat sink modifications sensitivity studies were performed to quantify the effects on

, the environmental temperature profiles of thermal radiation exchange between the gas and heat sink surfaces.

The COBRA-NC program and input structure for these scenarios was previously described in PNL's original report -therefore, such discussions will be forgone. The methodology for inclusion of the area and volume increases will, however, be addressed. In order to include gas radiation effects the COBRA-NC code was modified to specifically address the Fort Saint Vrain Scenarios. Since this represents a variation to the COBRA-NC code a brief description of the radiation model is presented below.

MODEL DESCRIPTION In order to simplify the resimulation of the high energy line break scenacios with the amended areas and volumes, the original number (i.e., 4) of heat sink types was maintained. The added heat sink areas and volumes, were categorized and simulated as one of the four existing heat sink types.

For the Turbine Building these four generic heat sinks are as follows: 1) concrete exposed solely to the confinement environment, 2) structural steel exposed solely to the confinement environment, 3) concrete partitions exposed l to both the confinement and other passive interior surfaces, 4) composite

, steel partitions exposed to both the confinement and the external ar.bient.

Similarly for the Reactor Building the generic heat sinks are described as follows: 1) concrete exposed solely to the confinement environment, 2) structural steel exposed solely to the confinement environment, 3) steel partitions exposed to both the confinement and other passive interior j surfaces, 4) composite steel partitions exposed to both the confinement and i

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Radiation heat transfer between gases and various heat sinks normally is not considered in COBRA-NC. In order to simulate this mode of heat transfer in a method consistent with the single volume-uniformly distributed area assumptions used for the convection heat transfer, a simplified uniform surface and gas enclosure problem structure was used. With this approach the

, gas is considered gray and its emissivity dependent on the mean beam length of the gas. The mean beam length in turn relates to the enclosure volume and areas. The problem may be divided into two distinct parts; the first pertains to the determination of the gas emissivity, and the second deals with the radiation heat transfer equation.

Addressing the first, requires an initial assumption about the makeup of the confinement gases. Both carbon dioxide and water vapor contribute to the thermal radiation exchange between the gas and its surrounding surfaces. As a slight conservatism to this analysis the carbon dioxide contribution to the gas emissivity will be ignored. The sole participating gas constituent, therefore, becomes the water vapor. The gas total emittance for water vapor may be expressed as a function of the absolute vapor temperature, the system pressure, the partial pressure of steam, and the mean beam length of the enclosure as (Siegel and Howell,1981):

e=a 3 [1-exp(-a2 /X)]. (1) where at and a2 are functions of the absolute vapor temperature. An expression relating the parameter X for water vapor-air mixtures to the Independent variables is expressed as:

[ X = pH 0Le(300/T)(p,yp+ bpH O). (2) 2 2 3

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TABLE 1. Wall Type Descriptions l

Turbine Building Assumed Height = 89.5 ft.

Heat Sink Surface Type Average Thickness (in.) Total Area (ft )

. 1) concrete exposed solely 28.1.8 46,710 to confinement

. 2) steel exposed solely 0.286 261,990 to confinement

3) composite steel exposed 2.922 45,610 to confinement and ambient

4) concrete exposed to 11.383 54,310 interior ambient temperatures Reactor Building Assumed Height = 233 ft.

Heat Sink Surface Type Average Thickness (in.) '.otal Area (ft2)

1) concrete exposed solely 28.59 83,260 to confinement

2) steel exposed solely 0.197 247,250 to confinement

3) composite steel exposed 5.25 50,840 to confinement and ambient

4) concrete exposed to 0.06 17,600

, interior ambient temperatures

. where T is in degrees Kelvin, pressures in atms, and the mean beam length in meters. The self-broadening coefficient b for water vapor is expressed as:

b = 5.0(300/T)1/2 + 0.5. (3) 4

The mean beam length while tabulated for simple enclosure geometries may be approximated by 0.9 4V/A for complex enclosures where the entire gas volume gas volume radiates to its entire boundary. For these Fort Saint Vrain scenarios the confinement void volume and the sum of the heat sink areas is used with 1

the above expressions to compute the mean beam length. The mean beam lengths calculated for the turbine and reactor buildings, respectively equal 14.71

. and 12.61 ft. With known beam lengths, and vapor temperatures and pressures calculated during each simulation time step, the gas total emittance may be

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. computed by applying Equations 1 through 3. An expression for the net heat q transfer between the gas and the enclosure may be obtained by considering a radiation network for a gray enclosure surrounding a gray gas (Welty, Wicks, andWilson,1969). The following equation is appropriate for cases of an entire gas volume radiating to its entire boundary:

i gg *+s = (Eb g - Ebs )A s 's'g/E(I~8s )"g* 'sl * (4)

With some rearrangement Equation 4 may be converted into a heat transfer coefficient expression as:

hrad = (T g + Tf)(Tg +T) s #'g 's /((1-es )'g + 's). (5) r

Equaticn 5 describes the heat transfer coefficient which is added to the existing convection heat transfer coefficient in COBRA-NC to determine the overall conductance between the gas and the surface.

The physics of an actual high energy line break in terms of the radiation heat transfer would be overwhelmingly complex. Beyond the gray gas and gray ,

surface assumptions all of the heat sink surfaces would be at different l

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temperatures radiating both to the gas and other enclosure surfaces. One would need to consider the positions and temperature distributions of the enclosure surfaces along with the temperature distribution throughout the gas.

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The COBRA-NC model ' described above, although simplified, is consistent with the single volume-uniform gas temperature model used for convection heat transfer. For comparison purposes the convection and radiation effective heat transfer conductances are tabulated below (see Table 2) for the four different line break scenarios at selected points throughout the transient.

. TABLE 2. Convection and Radiation Conductances

. Convection Radiation Simulation Time Conductance Conductance Scenario (sec) (hours) (8tu/hr-ft*F) (Btu /hr-ft*F)

HRH-1 0.20 0.0033 1.408 0.316 HRH-1 13.96 0.233 1.000 0.775 HRH-1 59.79 1.00 1.000 0.581 CRH-19 1.00 0.0167 1.000 0.285 CRH-19 29.79 0.500 1.000 0.641 CRH-19 179.79 3.00 1.000 0.495 HRH-2 0.50 0.0083 1.941 0.321 HRH-2 11.46 0.191 1.000 0.718 HRH-2 59.79 1.00 1.000 0.530 CRH-15 0.30 0.005 1.J53 0.311 CRH-15 17.96 0.300 1.000 0.801 CRH-15 59.79 1.00 1.000 0.698 I

RESULTS The confinement average environmental temperatures are plotted versus time for the high energy line break scenarios HRH-1, CRH-15, HRH-2, and CRH-19 in Figures 1 through 4, respectively. The plots represent temperature history results for COBRA-NC simulations with the amended turbine and reactor building heat sink volumes and areas. Each plot displays the results without radiation heat transfer (without radiation) considered between the gas and the surfaces, with radiation heat trar fer (with radiation) and the Sargent and Lundy composite (S&L composite LdE) profile used for equipment qualification. In 6

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TABLE 3. Peak Simulation Temperatures Peak Temperatures 'F Original Building Amended Building Description Description Scenario w/o radiation w/ radiation HRH-1 463.7 270.8 265.3 HRH-2 492.1 262.0 257.6 CRH-15 --

213.1 208.4 CRH-19 320.6 197.8 191.7 1

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I REFERENCES i

Siegel, R., and J. R. Howell, 1981. Thermal Radiation Heat Transfer, Second Edition, McGraw-Hill Book Company, pp. 619-627.

i i Welty, J. R., C. E. Wicks, and R. E. Wilson, 1969. Fundamentals of Momentum Heat and Mass Transfer, John Wiley & Sons Inc., pp. 431-436.

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Wheeler, C. L., M. J. Thurgood, T. E. Guidotti, D. E. DeBellis. 1986.

" COBRA-NC: A Thermal-Hydraulic Code for Transient Analysis of Nuclear Reactor l Components", NUREG/CR-3262, PNL-5515, Vol .1-7, Pacific Northwest Laboratory, Richland, Washington.

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E APPENDIX A REVIEW INFORMATION SUPPLIED BY NRC 4

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0Puolic Service ~ ~.

Company of Colorado i

2420 W. 26th Avenue, Suite 1000, Cenver, Colorado 80211 l

December 19, 1986 Fort St. Vrain

' Unit No. 1 P-86673 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 l i

Docket No. 50-267 i

SUBJECT:

Turbine Building Temperature Profiles

REFERENCE:

1) PSC letter, Warembourg i to Berkow, dated '

February 28, 1986 (P-86120)

2) PSC letter, Warembourg to Berkow, dated December 12, 1986 (P-86664)

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 l by Battelle/PNL and discussion in our meeting with the NRC on l November 20, we were requested to analyze three scenarios for the !

Reactor Building and three scenarios for the Turbine Building using a l convective heat transfer coefficient of 1.0. The value of 1.0 is

, consistent with PNL's approach. We were advised to add additional l, volumes we could defend and remove any other justifiable items from

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the conservatisms. The Reactor Building profiles were submitted by Reference 2.

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P-86673 Page 2 December 19 0 1986 Enclosed for your review is our second formal submittal in resporte to the above request. Information on the following Turbine Building steam line break scenarios is enclosed:

HRH-1, Hot Reheat Steam Leak In Turbine Building (Offset Rupture)

CRH-15*, Cold Reheat Steam Leak In Turbine Building (25% Leak Area)

CRH-13E, Cold Reheat Steam Leak In Turbine Building (0.18% Leak Area)

Attachment 1 provides temperature profiles, tables, and figures in the same format as submitted by the. Reference 1 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 operating floor and the area around the condensate storage tanks. These areas have adequate communication with the Turbine 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.

Additional conservatisms that were evaluated include:

- Orifice coefficient for steam pipe during blowdown

- Revaporization of condensate

- Radiation heat transfer The effect of the orifice coefficient and revaporization of the condensate has previously been analyzed for the Reactor Building breaks in Reference 2. . The Turbine Building breaks are equivalent and due to the magnitude of the benefit to the temperature profiles, we have elected to leave these as conservatisms 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 i

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 Turbine Building scenarios.

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P-86673 Page 3 -

December 19, 1986 It is noted'that the new HRH-1 curve has a significantly lower peak l temperature, but has shifted slightly to the right as one would expect given consideration to a larger volume. We have evaluated this curve shift with the determination that the new curve does not represent a more severe equipment qualification profile, especially ,

when additional conservatisms are considered. Therefore, we conclude that the original profiles generated for our EQ Program can be considered conservative even when an ultra-conservative value of one is assumed for the convective heat transfer coefficient. When the i

more realistic but still conservative values proposed by PSC are used, in addition to the larger volumes and heat sinks, the original profiles represent even more margin of conservatism.

Overall, based on the Reactor and Turbine Building reanalysis, we believe the composite temperature profile curves originally submitted and utilized as the basis of our EQ Program are conservative and remain appropriate without further changes. .

As you are aware, the profiles are a very critical part of our existing program. Please keep us abreast of your review activities.

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

1 Very truly yours, A0 W /Yw d w D. W. Warembourg, Manag,er Nuclear Engineering Division DWW/KD:pa Attachments 9

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ATTACHMENT 1 FORT ST. VRAIN ENVIRCNMENTAL QUALIFICATION PROGRAM TURBINE BulLDING TEMPERATURE PROFILES RESULTING FROM STEAM LINE BREAKS EVALUATION OF LARGER VOLUMES WITH 1.0 CONVECTIVE HEAT TRANSFER COEFFICIENT I

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v 0 0 0 C.1 2.7775 x 1C5 13.645 x 106 27.51- x !O9 0.2'O 5.5c33 x 10-5 10.923 x 100 15.4-3 x !09 0 361 1.0553 x 10-" 6.599 x 1C6 9,,73 x :;9 0.5 0999 1.5533 x 10-" 5.676 x 106 3,3 5 x :;9 0.920997 2.5583 x 10-" 4.729 x 106 7,037 x :;9 1.-6099 4.0563 x 10-' 4.040 x 106 6.115 x 109 2.40229 6.6730 x 10-" 3.468 x 106 5.307 x 109 4.00036 1.1112 x 10-3 2.901 x 106 g,gu6 x 109 5.00057 2.4224 x 10-3 2.427 x 100 3.723 x 109 10.0012 2.7761 x 10~1 -

2 323 x 106 3.565 x 102 11.002 3.0559 x 10-3 1.909 x 106 2.940 x 109 12.0003 3.3335 x 10-3 a43.0 x 103 1.294 x 109 13.0 3.6111 x 10-3 391.1 x 103 606.4 x 106 :

14.0017 3.8894 x 10-3 136.3 x 103 289.8 x 106 !

15.0012 4.1670 x 10-3 ,

142.8 x 103 218.0 x 106 l 16.0187 4.4496 X 10-3 0 1 0

= = l 0 0 O

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Gescription: Cold renea- lea < in Srtine Builc ng nrougn 0.18 leax area Reference Input: Hanc Calculation, GA Dec. 908336 Time Flow Rate En:P.alpy Rate (sec) (It/hr> (Stu/hr) 0.00 0.00 0.00 0.01 1.00 + C5 1 375 - 05 4500 1.00 - 05 1 375 + Oc 5364 0.00 0.00

=

0.00 0.00 4

l l

3 i

1 i

1

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RH 15 FL3 A:C I::EE;E RE* EA5E 'iER5';5 7:ME

escrip: 10n: Col 0 reneat leak in 7ar01:e 5al; ing nicugn 255 lea < area Referer.ce Input: Run 57 2216, (Flasn/0A)

Te Flew Rate In:haley Rate (sec) (lb/hr) (Stu/hr) 4 Sh/lb

,0.00 0.00 0.00

'4.49 -02

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' O.13 _

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1.00 9.75 + 06;7py,3 1 30 + 10 isss.S 2.14 7.50 + 06gggg3 9.71 + 09 ,12.'I S. S 3 31 5.66 + 06 jgj3,3 7.37 + 09 4.77 + 06 jygg y__.79 + 09 1263 4

. _"__ "I _ _ _ . . . _

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09

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/S.91 1.52 + 06 g 1.98 + 09

_..I27 %. L . _

3 3'o g , 9 9.71 1.23 + 06 ggy,p 1.6 4 + 09_ f333,3

<11.41 9.00 - 05 2C's . o 1.24 + 09 j $ 77,3 l

, 13.41 7.00 + 05 f94 g 9.84 + 08 gog,9 15.41 5.80 + 05 fg, f 8.31 + CS i.c3 2,7 17.41 5.05

05 7.28 + 08 19.42

/.90..-). _ . . _ _

li4I.la

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30.91

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_ g4qo,3

TABLE e ( Cor.: .+.'.e c ,

Time F10's Rate 2nt.9al;y Rate I

(sec) (1b/hr) (Stu/hr) 32.92

- 2.2i - 05 p.e 3 29 + : i4ssy 33.;i 1.e4 - 05 g.f 2.ri + 03 1472.1 42.91 2. 3 5

05 1.5 5 - 05 d.3,y _

gg. 3 46.91 1.41

- 05 gy,p 2.11 + 08 g 4 ql, . 5 50.91 1 30 + 05 36.I 1*95

CS 1500..0 54.91 1.16

05 32.L 1.76 + 05 56.91 1517.'2.

1.06 + 05 g , y 1.59 + OS l5 o o,0 62.91 9.75 + 04 7 7, f 1.46 + 03 l4q7 4

-66.91 9.00 + 04 2f 0 1 34 + Ob j.4 g3 , ']

70.91 8.22 + 0" %2 7 1.23 + 06 gqg,4 74.91 7.73 + 04 gf,( 1.16

03 j_ggo,,4 78.91 7.09 + 04 jg, 7 1.07 + 08 l$og 2 82.91 6.54 + 04 ;g, z 9.39 - 07 66.91 15A 2.

5.98 + 04 ft,, c 9.12 + 07 90.91 1525.I 5.87 + 04 fg, p 94.92 8.97 {07

_ [$ gg , ( _ ;

6.37

04 j7,f 9.70 + 07 ,jg g 7,_,g 98.92 6.6 4 + 04 jg,y 1.00 + OS 390.90 g g, o j 6.64 + 04 f f. </ 1.00 + 08 _gggg, o 791.00 0.00 g,ao 0.00

l30h. O

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4 3. ;0ncrete Partition 12 a 600 i Walls an F10 rs 2 *. Piping 0 375 3C,540 0 3 5. C::; site Steel Wall 5.25 7,93c 6 2 6. Stec-1 Oecking 0.0936 13.220 C 2 7. Structural Steel 0 375 67,410 C anc Equipnent ,

1 6. Electrical Concuits 0.0312 57,970 0 anc Caole Trays

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ATTACHi! Erit 2 FORT ST. VRAIN ENVIRONMENTAL QUALIFICATION PROGR TURSINE BUILDING TEMPERATURE PROFILES RESULTING FROM STEAM LINE EVALUATION OF LARGER VOLUMES WITH 1.0 CONVECTIVE HEAT TRANSFER COEFFICIENT !

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Company of Colorado 2420 W. 26th Avenue, Suite 1000, Denver, Colorado 80211 December 12, 1986 0

, Fort St. Vrain Unit No. 1 P-86664 Director of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission )

Washington, D.C. 20555 i Attention: Mr. H. N. Berkow, Director Standardization and Special Projects Directorate ,

l Docket No. 50-267 !

SUBJECT:

Reactor Building Temperature Profiles i

REFERENCE:

1) PSC letter, Warembourg to Berkow, dated l 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 Battelle/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:

P-86664 Page 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 famat 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 o an increased building volume as defined in the enclosure. The increased building volume includes the area above the refueling 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 parameters discussed in the NRCf 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: l

- Orifice coefficient for steam pipe during blowdown !

- Revaporization of condensate ,

1

- Radiation heat transfer !

The orifice coefficient used in the reference cases (i.e., per the

' 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 l 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 l

coefficient would result in a reduction of approximately 2 degrees.

It is therefore concluded that the conservatism of 1.0 vs. 0.8 for i the orifice coefficient has no significant impact in conjunction with ;

our reanalysis utilizing the increase in building volume. l

P-86664 Page 3 Oscember 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 have not factored this benefit directly into our reanalysis because of the magnitude of the benefit. On this basis, we have

. elected to leave this as a conservat. ism 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 independen'.ly 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. Hcwever, 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, i

um D. W. Warbbour[, Manager Nuclear Engineering Division DWW/KD:pa Attachments

j l

s ATTACHMENT 'l FCRT ST. VRAIN ENYlRCHENTAL QUALIFICATION PROGRAM REACTOR BUILDING TEW ERATURE PROFILES RESULTING FRCM STEAM LINE BREAKS EVALUATION OF LARGER VOLUES WITH 1.0 CCNVECTIVE HEAT TRANSFER COEFFICIENT i

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TABLE 1 DATA FOR PIPE BREAKS IN Tile REACTOR BUILDING 4

Data / Case llRif-2 CRil-19 CRii- 14 E 4

1. Broken pipe data

a. Type of fluid Steam Steam Steam

b. Temperature (*F) 1000. 740. 740.

3

c. Pressure (psia) 587.6 895. 895.

d. Source of the fluid S.G. E 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,1 Initial temp. -

90*F 90*F 90*F b,il Initial pressure 12.3 psi 12.3 psi 12.3 psi b,ill Initial humidity 70% 70% ,

70%

. b,1v Floor area Table 5 Table 5 Table 5

~

b,v Number of vents & vent creas (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. liest transfer coeffici~ent for walls Table 5 Table 5 Table 5 i

3

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.

TABLE 2 HRH-2 FLOW AND ENERGY RELEASE VERSUS TI!'I

Description:

Hot Reheat Steam Leak in Reactor Building Reference Input: Run ST8680, 9/4/85 at 19:24:57 (Flash /GA run) 1 Time Flow Rate Enthalpy Race s (sec) (hr) (1b/hr) (Btu /hr) j 0 0 0 0 0.1 2.7778 x 105 10.020 x 106 14.853 x 109 i 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 i 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 10'3 547.8 x 106 12.0016 3.3338 x 10-3 131.3 x 103

) 202.6 x 106 i 13.0049 3.6125 x 10-3 22.47 x 103 34.05 x 106 13.2170 3.6714 x 10-3 0 0

" # 0

, 0 4

l 1

1 IABLE 3 I CRH-19 FLOW AND ENERGY RELEASE VERSUS TIME

Description:

Cold Reheat Steam Leak in Reactor Building through 10* Leak Area

, Reference Input: Run ST6965, (Flash /GA)

Time Flow Rate Enthalpy Rate (sec) (1b/hr) (Btu /hr) i 0.00000 0.00000 0.00000 4.31952 - 02 6.76724 + 05 9.19545 + 45- os 8.80165 - 02 1.61093 + 06 0.105?: 00 - 2.l18 fc +oq l

.13349 1.99831 + 06 2.70948 + 09

.17932 2.00681 + 06 2.72152 + 09

.38290 2.00069 + 06 2.71502 + 09 i

.96548 1.96029 + 06 2.65647 + 09 l

1.5455 1.92229.+ 06 2.60334 + 09 2.1238 1.88949 + 06 2.55776 + 09 l 2.7011 1.85917 + 06 2.51666 + 09 3 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 6.8904 1.74870 + 06 2.37501 + 09 7.2904 1.71957 + 06 2.33178 + 09 7.6904 1.68288 + 06 2.27819 + 09 i

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 4

9.6904 1.50897 + 06 2.04338 + 09 4

,y-- --- - - . +---as a- --y e-7 v-iy -- m -- y v - *---e-- A-- w-P g -r --

Y Il TABLE 3 (Continued) 1 Time Flow Rate Enthalpy Rate l (sec) (lb/hr) (Beu/hr) I i

10.390 1.44849 + 06 1.96533 + 09 1

' i 11.390 1.37051 + 06 1.86566 + 09 I I

12.390 1.30115 + 06 1.77778 + 09 13.390 1.24720 + 06 1.71168 + 09 a 14.390 1.19969 + 06 1.65181 + 09 15.390 1.16019 + 06 1.59658 f 09 s n; 16.390 1.12334 + 06 1.5'3799 + 09 '

s, 17.390 1.08396 + 06 1.47114 + 09 18.390 1.04987 + 06 1.40969 + 09 s

19.390 1.02313 + 06 1.35618 + 09 (

20.890 9.91221 + 05 1.28767 + 09 s 22.00 9.60 + 05 1.23 + 09 l f.

112.00 0.00 0.C0 m 0.00 0.00

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)

TABLE 4 CRH 14E FLOW AND ENERGY RELEASE VERSUS TIME

Description:

Cold reheat leak in Reactor Building i through 0.25% leak area

, Reference Inpuci Hand Calculatien, GA Doc. 908838 a .

, .t .c -

Time Flow Rate Enthalov Rata (sec) (lb/hr) (Bru/hr) 0.00 0.00 0.00 N 0.01 5.15 + 04 -7.00 + 07 - \

(

4500 5.15 + 04 7.00 + 07 6178 0.00 0.00 -

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TABLE 5-REACTOR BUIt M WL-Hengld s 235 ft Volume = 1,396,980 ft3

-* M 81 6 ft flest Sink Ileat Transfer Wall Thick. Area Coefficient Outside(a)

Surface (in.) (ft2) (Btu /h-ft2_or) el 1. Concrete walls and floor, PCRV 36. 51,670 0

/ 2. PCRV support ring 21. 9,870 0 8t 8 3. Concrete partition walls, floors 12. 9,880 0 ft/ 4. Thin steel wall 0.06 17,600 2

3 5. Composite steel wall 5.25 50,240 6 42 6. Steel decking 0.09 6 50,840 0

2. 7. Structural steel and equipment 0.375 47,060 0

2 8. Ducting, electrical conduits 0.0312 80,950 0 Cable trays

2 9. Piping 0.375 62,520 0 g s ' 10. Concrete partition walls, floors 6. 6,280 0

11. Concrete partition walls, floors 18. 3,440 0 45 12. Concrete sealed rooms, regions 36. 5,940 2 -

/ 13. Keyway walls 24. 2,120 0

2. 14. Steel wall, 4a 0.05 5,880 0

(*) Locations that signify the inside and outside heat transfer coefficients are illustrated in Fig. 1.

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NRc EOR = n. u s. NuCuAR Raout ArORv COmuiss>ON . i RerOar Nuu.EM #Augeses Oy ffDC d@d VW #e, af espyd ,

I

[*, /E BIBUOGRAPHIC DATA SHEET EGG-NTA-7564 ,

so .NsrRuCriON. ON 1 . aen Ru PLN-6182

, nru ANo$u.riva e aan .LAN. )

Evaluation of Confirement Evnironmental Temperatures ~l Following High Energy Line Breaks Proposed For the Fort Saint Vrain Environmental Qualification Program , ','""""'""'"",,,

. AurRomisi January 1987 !

M. D. White . oAu Rw0ar muao '

C. L. Wheeler =oNr= "^a '

a March l 1987 i F. f ERFORMv4G ORGAN 62 AYlON NAME AND MasLING ADDRESS ##secessele Codes . PROJECr/T ASKnWORK u48 T NUM.ER EG&G Idaho, Inc.

. Idaho Falls, Id 83415 * * ' ' "'a^~'"*"""

D6023 l IE SPONSORING ORGANi2 ArtON NAME ANO M AILING ADDRESS (seechase le Cases tis rVPE OF REPORr Division of PWR Licensing ~

Informal Office of Nuclear Regulatory ammission I Washington, D. C. 20555 ana'= ana to "-~~ ~-e i i

N/A 12 SuPPLLMENr ARY NOrE$

,, A.,, R Ae, an r . ,

COBRA-NC simulations were perfonned of the high energy line break scenarios HRH-1, CRH-19, HRH-2, and CRH-15 in conjunction with the Public Service of Colorado Company's Fort Saint Vrain Environmental Qualification Program. The simulations a

comprise the amended heat sink areas and volumes specified for the Turbine and Reactor buildings. Consideration of radiation heat transfer processes between the confinement gas and heat sinks was incorporated into the scenario simulations. The !

confinement environment average teraperature history plots never exceed the Sargent and Lundy composite profiles used for equipment qualification.

i ;

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O le DOCUMENr ANALYSil e Et **ORDS OgscRiPf]ns ig avangag,yg y

$1ArtMENr

Unlimited s

14 SECumirY CLAlli8tCAr10%

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. .e Nr. s as 0,iN .No o NRM. Unclassified tihe reearto Unclassified

, , , ~uM. R O, , AG u 18 PRsCE 1

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ML20214Q541

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