ML20210P645

Draft Evaluation of Confinement Environ Temps Following High Energy Line Breaks Proposed for Fort St Vrain Environ Qualification Program, Technical Evaluation Rept

ML20210P645
Person / Time
Site:	Fort Saint Vrain
Issue date:	01/31/1987
From:	Wheeler C, White M Battelle Memorial Institute, PACIFIC NORTHWEST NATION
To:	NRC
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ML20210P526	List: ML20210P524 ML20210P645
References
TAC-42527, NUDOCS 8702130429
Download: ML20210P645 (14)

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DRAFT TECHNICAL EVALUATION REPORT EVALUATION OF CONFINEMENT ENVIRONMENTAL TEMPERATURES FOLLOWING HIGH ENERGY LINE BREAKS PROPOSED FOR THE FORT SAINT VRAIN ENVIRONMENTAL QUALIFICATION PROGRAM H.D. White C.L. Wheeler January 1987 Prepared for the U.S. Nuclear Regulatory Commission represented by Norm Wagner APPROVED:

l C.W. Stewart, Manager Fluid and Thermal Analysis Section l Engineering Sciences Department l

BATTELLE PACIFIC NORTHWEST LABORATORY RICHLAND, WASHINGTON 99352 8702130429 870129 PDR ADOCK 05000267 P PDR I

l This report is a working paper intended for the support and other contributors l to the program. Do not reference in open literature at this time.

ABSTRACT 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 generally fall within the Sargent and Lundy composite profties used for equipment qualification.

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NOMENCLATURE at,a2 coefficients A surface area b self-broadening coefficients Eb black body emissive power h heat transfer coefficient Le mean beam length p pressure q heat transfer rate T temperature V confinement volume X emissivity equation parameter e emissivity I

a Stefan-Boltzmann constant 4

Subscripts ai r- air g gas HO 2 water vapor s heat sink surface i

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l INTRODUCTION This report documents the evaluation of environmental conditions within confinement st'ructures 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 Battelle PNL (1) due to amendments in the confinement structural descriptions. The original analyses of these high energy line break scenarios performed by Gulf Atomic, representing Public Service of Colorado, differed with the results generated by Battelle 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 l determined, by Battelle PNL, to be non-conservative (2). The environmental temperatures calculated by Battelle PNL exceeded the limits for the Environmental Qualification Program, while the temperatures and pressures  ;

calculated by Gulf Atomic were within qualification limits. Following arguments supporting the opposing views in regard to the most appropriate natural convection heat transfer coefficients, the NRC ruled that the evaluations would be repeated with the more conservative heat transfer coefficients, but with c the inclusion of previously unaccounted heat sinks within the confinement structures. In addition Public Service and Battelle PNL agreed that radiation from the steam environment to the confinement surfaces should also be considered.

T' bis Ydiu Nn"t M d Ns t N SNu M o E CO N ~N b imu5 5 ons on the four most severe high energy line break scenarios entitled HRH-1, CRH-15, HRH-2, ,

1 and CRH-19. The first two consider line breaks within the turbine building and the others line breaks within the reactor building. All of the scenarios i were simulated with the amended confinement structure descriptions. The  !

i 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 Battelle's original report, therefore, such discussions will be forgone. The methodology for inclusion of the area and volume increases will, d

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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 '

scenarios 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, therefore, were categorized and simulated as one of the four heat sink types.

For the Turbine Butiding these four generic heat sinks are as follows: 1),

concrete exposed solely to the confinement environment, 2) structural steel exposed soldly to the confinement environment, 3) concrete partitions exposed to both the confinement and other passive interior surfaces, 4) composite steel partitions exposed to both the confinement and the external ambient, f 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 surfaces,

4) composite steel' partitions exposed to both the confinement and the external ambient. The total areas calculated for each of the four categories are

indicated below (Table 1).

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RadistioMtieatiYransfe fbE dieE g'asEs and W iou Eheat sinks normally is not encoded in COBRA-NC. A back of the envelope type calculation revealed the potential importance for this mode of heat transfer when addressing the Fort

Saint Vrain line break scenarios. In order to simulate this mode of heat j 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 l- 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.

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.- c TABLE 1. Wall Type Descriptions Turbine Building Assumed Height = 89.5 ft.

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Heat Sink Surface Type Average Thickness (in.) Total Area (ft2) 1)concreteexposedsolely 28.18 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 and interior ambient temperatures Reactor Building Assumed Height = 233 ft.

Heat Sink Surface Type Average Thickness (in.) Total Area (ft2) ,

1)concreteexposedsolely 28.59 83,260 to confinement

2) steel exposed solely 0.197 247,250 to confinement

- 3)' compost te esteel_ exposed r r - 5.25- t u4-n ca r :=

50,840 to confinement and ambient 4)concreteexposedtoconfinement 0.06 17,600 j and interior ambient

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temperatures l Addressing the first,. requires an initial assumption about the makeup of l the confinement gases. Both carbon dioxide and water vapor contribute to the i 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 l

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pressure, the partial pressure of steam, and the me,an beam length of the enclosure as(3):

e=a g [1-exp(-a /A)3*

2 II) where al 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=pg ote(300/T)(p ir+ bpH 0}'

a (}

2 where T is in degrees Kelvin, pressures in atms, and the mean beam length'in meters. The self-broadening coefficient b fdr water vapor is expressed as:

b = 5.0(300/T)1/2 + 0.5. (3) l 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 line break scenarios the confinement void volume and the sum of the heat sink areas is used with the above expressions to compute the mean beam length. The mean beam lengths calculated for the turbine and reactor buildings, respectively equal 14.'71and'12.61ft."Withinownlia'leigihsIa'n'd'va'portemperaturesand m

! pressures calculated during each simulation time step, the gas total emittance

! may be computed by applying equations 1 through 3. An expression for the net heat transfer between the gas and the enclosure may oe obtained by considering i a radiation network for a gray enclosure surrounding a gray gas (4). The following equation is appropriate for cases of an entire gas volume radiating to its entire boundary:

A gus = (Ebg -Eb)A's'g/EII-E)'g*'s' s s s 3 I4)

With some rearrangement equation 4 may be converted into a heat transfer

-coefficient expression as:

4

hrad = g(T2+Tf)D+T) g s

1. 8 g 's M Is)'g + 's). (5)

Equation 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 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.

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 for the four different line break scenarios at selected points throughout the transient.

TABLE 2. Convection and Radiation Conductances Simulation time Convection Radiation conductance conductance Scenario _(sec) (hours) (8tu/hr ft*F) (8tu/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 1.053 CRH-15 0.311 17.96 0.300 1.000 CRH-15 0.801 59.79 1.00 1.000 0.698

4 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 1-4, respectively. The plots represent temperature history results for COBRA-NC simulations with the amended turbine and reactor building heat sink volumes and areas. Ea h plot displays the results without radiation heat transfer (w/o radiation) considered between the gas and the surfaces, with radiation heat transfer (w/ radiation) and the Sargent and Lundy composite (S&L composite DBE) profile used for equipment qualification. In all cases the temperature profiles are substantially reduced compared with those profiles generated using the original heat sink volumes and areas. The magnitude of the reduction in peak temperatures between the previous simulations and the present simulations with the augmented areas and volumes in shown in Table 3.

The effect of radiation heat transfer generally appears to reduce the peak temperatures and post-peak temperatures.

TABLE 3. Peak Simulation Temperatures Peak Temperatures *F l

original building amended building description description Scenario s--si-w/o-radiation-4 ' w/ radiation HRH-1 463.7 270.8 265,.3

( HRH-2 492.1 262.0 257.6 CRH-19 320.6 197.8 191.7 l

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REFERENCES 1.

Wheeler, C.L.,-R.E. Dodge, and J.R. Skarda, " 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, Battelle, Pacific Northwest Laboratory, August (1986).

2. White, M.D., " Review of Convection Heat Transfer Coefficients Utilized in the Fort Saint Vrain Main Steam Line Break Anal Battelle, Pacific Northwest Laboratory, December (yses" 1986). FATE-86-117,

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

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

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