ML20207S233

Safety Evaluation Re Composite Temp Profile for Qualification of Electrical Equipment.Profiles Satisfactory for Establishing Upper Bound for Expected Temps During Postulated Limiting High Energy Line Breaks

ML20207S233
Person / Time
Site:	Fort Saint Vrain
Issue date:	03/13/1987
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20207S229	List: ML20207S228 ML20207S233
References
TAC-42527, NUDOCS 8703190160
Download: ML20207S233 (2)

Text

.

j# UNITED STATES e" i%o NUCLEAR REGULATORY COMMISSION g WASHINGTON. D. C. 20555 Enclosure 1 SAFETY EVALUATION BY THE OFFICF 0F NUCLEAR REACTOR REGULATION PELATING TO COMPOSITE TEMPERATURE PROFILE FOR QUALIFICATION OF ELECTRICAL EQUIPPEhT PUBLIC SERVICE COMPANY OF COLORADO FORT ST. VRAIN NUCLEAR GENERATING STATION DOCKET NO. 50-267

1.0 INTRODUCTION

The Public Service Company of Colorado (PSC), the licensee for the Fort St. Vrain Nuclear Generating Station, has provided analyses to determine the maximum temperatures to be expected in the event of postulated high energy pipe breaks in the reactor and turbine buildings. The licensee has conducted these analyses in order to assure that the composite temperature profile being used to qualify electrical equipment will serve as an upper bound to temperatures resulting from the aforementioned pipe breaks.

2.0 DISCUSSION 2.1 Original Analyses PSC provided analyses (temperature profiles) in a submittal dated February 28, 1986 (amended by letter of March 14,1986), which concluded that the maximum temperatures resulting from the most limiting high energy line breaks, i.e., the hot reheat steam line in the reactor and turbine buildings were approximately 375*F and 365 F, respectively.

2.2 Analyses by Pacific Northwest Laboratory (PNL)

As an independent means for verifying the temperatures determined by PSC, the staff contracted with PNL to conduct analyses utilizing the computer code developed by PNL and satisfactorily utilized previously to confirm equipment qualification (EQ) temperatures at various other nuclear power plants. The PNL analysis determined that the peak temperature for the hot reheat steam line break in the reactor building was 487*F. This exceeded the value determined by PSC by approximately 110'F. Therefore, the staff questioned the validity of the analytical methodology employed by PSC. Further, on the basis of the PNL analysis, the composite temperature profile used by PSC would need to be raised by at least 110-120*F in order to provide an upper bound for electrical EQ in the reactor building.

8703190160 DR 870313 p ADOCK 05000267 PDR

Subsecuently, PNL reviewed the calculations generated by PSC and ioentified the errors ir, the licensee's methodology. However, modi-fication of the calculations by PFC with the errors corrected woulo not resolve the primary concern that the peak temperature used for equipment qualification was too low, since such a correction would increase the EQ temperature to the approximate value found by PNL.

2.3 Reanalysis by PSC PSC subsequently reviewed the abcve concern and met with the staff on November 20, 1986 to discuss the issue further. At the neeting, FSC pointed out that there were additional volumes contiguous to the reactor and turbine buildings which could be utilized to increase the volume in both buildings, thereby providing a significantly greater heat sink.

Utilizing these revised volumes and a corrected methodology, by letters dated December 12 and 19, 1986, the licensee presented new temperature profiles for the reheat steam line breaks in both the reactor building and the turbine building. These new analyses indicated that building temperatures are within the original composite temperature profile used to environmentally qualify electrical equipment.

2.4 Reanalysis by PNL PNL also developed new temperature profiles utilizing the increased volumes for the turbine and reactor buildings. In the Technical Evaluation Report (TER) dated January 1987, PNL reported that the revised temperature profiles for the hot reheat breaks in the reactcr and turbine buildings "---never exceed the Sargent and Lundy composite profiles used for equipment qualification." The TER is considered to be a part of this Safety Evaluation.

3.0 CONCLUSION

Based upon the foregoing, the staff cencludes that the composite tempera-ture profiles used by PSC to qualify electrical equipment are satisfactory for establishing an upper bound for the expected tenperatures during postulated limiting high energy line breaks in both the reacter and turbine buildings. The staff, therefore, considers this issue resolved.

Reviewer: N. Vagner, DPWRL-B Date: March 13, 1987

Enclosure 2 6

TECHNICAL EVALUATION REPORT EVALUATION OF CONFINEMENT ENVIRONMENTAL TEMPERATURES FOLLOW 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: '

OfW. Stewart, Manager "

Fluid and Thermal Analysis Section Engineering Sciences Department BATTELLE PACIFIC NORTHWEST LABORATORY RICHLAND, WASHINGTON 99352 This report is a working paper intended for ',he support and other contributors to the program. Do not reference in open literature at this time, fD b i lY

• ABSTRACT 6

COBRA-NC simulations were performed of the high energy ifne 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.

?

e h

NOMENCLATURE --

ai,a2 coefficients '

a 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 a Stefan-Boltzmann constant Subscripts air air g gas HO2 water vapor s heat sink surface 4

i

l INTRODUCTION This report documents the evaluation of environmental conditions within confinement structures of the Fort Saint Vrain Nuclear facilities, followin several proposed high energy line break scenarios.

The present evaluation differs from the previously submitted documentation by Battelle PNL (1) d amendments in the confinement structural descriptions.

The original analyses of these high energy line break scenarios performed by Gulf Atom Public Service of Colorado, differed with the results generated by Battelle PNL, representing the U.S. Nuclear Regulatory Commission (NRC), becaus differences in natural convection heat transfer coefficients. The heat transfer coefficient determination methodologies used by Gulf Atomic were reviewe 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 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, b with the inclusion of previously unaccounted heat sinks within the confinem structures.

In addition Public Service and Battelle PNL agreed that radiation from the considered. steam environment to the confinement surfaces shoul This document presents the results of COBRA-NC simulations on the four most severe high energy line break scenarios entitled HRH-1, CRH-15, HRH-2 and CRH-19. ' ,

The first two consider breaks within the turbine building and the others breaks within the reactor building.

All of the scenarios were simulated wit,h the amended confinement structure descriptions. The specific confinement heat sink modifications were reported in two letters between Pu Service Company of Colorado (PSCC) and NRC (see appendix A).

In addition to these heat sink modifications sensitivity studies were performed to quan 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 p described forgone.

in Battelle's original report, therefore, such discussions will be 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 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, 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 to both the confinement and other passive interior surfaces, 4) composite steel partitions exposed to both the confinement and the external ambient. 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).

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.

.n,-- - , - ---

\

J TABLE 1. Wall Type Descriptions Turbine Building ,

Assumed Height = 89.5 ft.

Heat Sink Surface Type Average Thickness (in.) 2 Total Area (ft )

1) concrete exposed solely 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) 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 co'nfinement 0.06 17,600 and interior ambient temperatures 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(3): 4 5

e=a 1 [1-exp(-a2 yX)]. '

, (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=p g0 Le(300/T)(pa ir+ bpH O).

2 (2) 2 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)

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 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 prer,sures '

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 thq enclosure may be obtained by considering a radiation n,etwork for a gray enchosure surrounding a gray gas (4). The following equation is appropriath for cases of an entire gas volume radiating 7.o its entire boundary:

e

+

9++s-(Eb 9 g - Ebs)As 's'g/[(1-es )'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-E3 )C g t C3 ). (5) 6 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) (Btu /hr ftfF) (Btu /hr ftfF) i HRH-1 0.20 0.0033 1.408 0.316 HRH-1 13.96 0.233 1.000 HRH-1 0.775 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 0.311 CRH-15 17.96 0.300 1.000 0.801 CRH-15 59.79 1.00 1.000 0.698 l

l

RESULTS -

The confinement average environmental temperatures are plotted versus time for the high energy line break scenarios HRH-1, CRH-15, HRH-2, and CR in Figures 1-4, respectively. The plots represent temperature history results for COBRA-NC simulations with the amended turbine and reactor building hea sink volumes and areas. Each 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 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-19 , 320.6 197.8 191.7 t

1 k

d 0

0 1

o l .

r a l l E 0 0

n l l

B

.D. 1 e l c n l

e.

Do n l

S l N L. o. is.

t.

Et e G i i.' t o E d o. p l

sa )t L r aC i

do m.

e s e I l

r 0 l L 1 l l

/ / a-e ff l w wS

)

R O I

s s

' e.

m e t a N u l

e. b / s x

n .

t L S u .- y t 1

m 1

o uu '

'-T .

( t a

,D

\\

1

\'

.t ' u g( c

\

i -k4 e i r

o 1

_ ~

m.

L - j I / ,t d H- I '-

/ ,

I

/

/ 1 T

l ,

l R 0

u H r ,

) ,

i

/

B / '

e , ,

n z -

L b

r

/

/ ' 1 0

_ u ,

0

- T /

L 1

/ -

N 5:

1 0 P 8 0:

0

/s e 2 O l l 1 0 0 0 0 0 0 e

2 5 8 1 4

0 t 7

4 3 2 7 t 2 1 a 8

/

_ egL _ B 6

-L g2 gLU>.oteoeeH )e.ce)o_eT -

/

2

, Turbine BulldLng, Steam Release Scenarlo CRH-15 (25% Leak area) 420 l_ ._

I

. l. } l ll lIl q) LEGDJD

' a) 350 '

-'- w/o radi,ot.i.on O w/ radi.at.i.on e ,'

m , ,

_5f.C C' 'o3po'sU. t'$ ~D. lif '.

i e -

2 280 '

) .)\

L a / '

s 210 o

L -

/

s si u

\

D '

/ ^ \

O. s

.- \ \

./

140 g

,/ x d ~

c G 70 g ..>

. o E

G~

0 t 0.001 0.01 0.1 1 10 100 1000 Time (minutes)

Battelle PNL 0***

1/29/87 11:4 3:40

Reactor BulldLng, Steam Release Scenarlo HRH-2 (Double Offset)

, 420 L.

LEGEND l g e 350

"'o ' odlot.Lon

@ --- d ERQ(o,t._i_q9,, _ , ,,,

l ,

S&L Composi.te DBE L ,

i m

i i

/

(D

' \

2 280

/

k g

I 'h

& l' i L / l 'R ^ '

U f I ,

t 4 210 < / m o '

/

, n 4

L '

e i / j. -

, Q. / '

> r y n __- --

E '

,s e 140 7 ,  ;

E--. .

/ 'N -

~

-.3 , , ~-

j c - -

l 0 70

! _o E

l C i

0 r

0.001 0.01 0.1 1 10 100

. 1000 r TLme (mtnutes) i FIGURE 3.

i Battelle PNL t

l

2/ 6/87 14

02:52 t

O s

o O

o O

.o L  :

O  := 'u o C --

-- :m

,o.

o e --

U C .e (D -- a zJ

-- .O C .)

0,l.

a> . >ee .J.O em ~~ 0 .O.) O 0, mO O e --

JEL 48- Oc /

eL J O 22 N O ': e e --

> N:'d

>m r e .x

' m

~~ i O  ! O ee \ > '

OJ \

D e  ; / C .

) ,\*

cn .-<

a

, ./

md q r

f .e m gw ce

.- -' \' D a3 '

x e e C 0)

.a -

N s e '

, J T l \ b

.>r \ .

.a e s o

ao '

s s

s CD  %

._ - ss \,

L '

s' O \

s O _

O s o -

g ..

O e s

.a s z 0 1

s s

a o_ s,

\ \

o 2

R 0 $ $ $ $ R S

• s (J seea6ep) aan30aedwel queqqwg n

v o

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 , FATE-86-117, Battelle, Pacific Northwest Laboratory, December (yses"1986).

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).

APPENDIX A Appendix A consists of letters from Public Service Company of Colorado dated December 12 and 19, 1986 (P-86664 and P-86673) submitting the revised temperature profiles for the Fort St. Vrain Reactor and Turbine buildings.

b b

e

_ _ . _ , _ ,- ..,.- - - - - e