Text

Summary of 960819-20 Meeting at Fauske & Associates,Inc in Burr Ridge,Il Re Review of SCDAP/RELAP5 Code Modeling of Natural Circulation Under Severe Accident Conditions
	ML20134K080
Person / Time
Issue date:	08/26/1996
From:	Richard Lee; NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
To:	Ader C; NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
References
	NUDOCS 9611180230
	Download: ML20134K080 (120)
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Augu st 26, 1996 i

MEMORANDUM T0:

Charles Ader, Chief Accident Evaluation Branch Division of Systems Technology l

Office of Nuclear Regulatory Research THRU:

Charles Tinkler Accident Evaluation Branch Division of Systems Technology Office of Nuclear Regulatory Research FROM:

Richard Lee Accident Evaluation Branch Division of Systems Technology Office of Nuclear Regulatory Research

SUBJECT:

SUMMARY

OF MEETING AUGUST 19-20, 1996 MEETING AT FAUSKE AND ASSOCIATES, INC., IN BURR RIDGE, ILLIN0IS.

Enclosed is a summary of the meeting on the review of the SCDAP/RELAPS i

code modeling of natural circulation under severe accident conditions, held on August 19-20, 1996, at Fauske & Associates, Inc., in Burr Ridge, Illinois.

Enclosure:

As stated Distribution:

DST Chron, AEB r/f, Lee r/f, Lee, Tinkler, Ader, King, Hodges CF

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DOCUMENT NAME: Ader3.mem DISK: Iae 3 To r:ceive e copy of this document. indient in the box: *c" copy withaut enclosure /*E" = Copy with enclosures *N" No copy 0FFICE AEB/ DST /RES J

AEB f T AEB \\ /l l

l NAME Lee EU Tinkler LW L Ader X DATE 08/;l/96 08/24t/96 08/ //96 N 08/ /96 08/ /96 0FFICIAL K CORD COPY (RES File Code) RES 2C-2 9611180230 960826 PDR ORG NRED PDR NRC FILE CENTER CGPY

I Enclosure Meeting Summary SCDAP/RELAP5 code modeling of natural circulation under severe accident conditions August 19-20, 1996 Fauske & Associates, Inc.

16WO7 West 83rd Street Burr Ridge, IL

Participants:

j R. Viskanta, Purdue University t

M. Ishii, Purdue University P. Griffith, Massachusetts Institute of Technology C. Ader, C. Tinkler, R. Lee, J. Donoghue, J. Staudenmeier, USNRC 1

D. Knudson, E. Harvego, P. Bayless, Idaho National Engineering 1

Laboratory I. Catton, ACRS-NRC i

M. Epstein, R. Henry, Fauske & Associates Note: Reviewers are Viskanta, Ishii and Griffith (consultants of Energy i

Research, Inc.).

Catton participated as an observer for ACRS.

R. Henry participatad as a representative of the nuclear industry.

Summary:

The meeting commenced at 9:00 a.m. on August 19, 1996, with the introduction of the agenda by R. Lee (Attachment 1). An introduction on l

the. purpose of the meeting, and background on the SCDAP/RELAP5 (SR5) i modelling of natural circulation under severe accident conditions was provided by C. Tinkler (Attachment 2).

I. Catton stated his views on the scaling of the Westinghouse 1/7 scale natural circulation experiments, and how uncertainties should be treated in estimating risk for steam generator tube failures.

P. Bayless presented the background on the development of the SR5 model (Attachment 3). Next, R. Henry gave a presentation on the Westinghouse 1/7 scale experiments (Attachment 4).

After, Henry's presentation, Bayless returned to present the benchmarking of SR5 against the Westinghouse 1/7 scale experiments, and discussed the use of SRS to analyze natural circulation in the Surry plant (Attachment 3).

Next, D. Knudson presented the most'recent SR5 analyses of the Surry plant (attachment 5). Throughout these presentations, discussions took place among the reviewers and participants on subjects presented.

On August 20, 1996, J. Donoghue discussed how the SR5 analysis results are being used in estimating the risk associated with steam generator tube failure (Attachment 6).

After some discussion, the reviewers caucused among themselves. Thereafter, P. Griffith provided their preliminary comments on the meeting.

The reviewers stated their view that (a) the experimental (Westinghouse 1/7

4 4

2 scale experiments) data was good and that the experiment was well.

designed, and additional experimental data was not needed; and (b) the code (SR5) was certainly adequate for the job (i.e., to calculate natural circulation under severe accident conditions), the implementation of the code was good and the constitutive relationships used by SR5 were adequate. They recommended additional activit for the purpose of demnnstrating more clearly the adequacy of the modeling (i.e., establishing a " figure of merit" by summarizing the experimental data and the analytical (SR5 calculations) results in a fashion (e.g.,

temperature vs. time) to show that the experimental data and the analytical results give the same systematic overall behavior in the reactor system being studied.

P. Griffith went on to state his additional view that once a " figure of merit" was established, it would be worthwhile to perform a few sensitivity calculations for parameters which may vary widely (e.g., a factor of 2, affecting the heatup rates of different reactor components (e.g., surge line, hot leg, steam generator tubes)). After a brief discussion to clarify some of the reviewers' comments, the meeting was adjourned around 1:00 p.m.

Attachments: As stated i

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ATTM PMEtet i, SCDAP/RELAP5 cods modeling of natural circulation under severe accident conditions August 19-20, 1996 l

Fauske & Associates,Inc.

16WO7 West 83rd Street Burr Ridge, IL AGENDA i

August 19,1996 1

a) 9:00 a.m.

Opening Remarks SRC b)

SCDAP/RELAP5 (SRS) m6deling sf naturht cirestation anld steam generator _(SG)~

tube heating in a PWRs

? Model development and'assessmentiand

. scallag isseen development of SR5 model 9:15 a.m.

in-vessel natural circulation hot leg countercurrent now SG inlet plenum mixing heat transfer modeling (in SG hot leg, surge line)

INEL 10:45 a.m.

' Break

_1/7 scale experiments FAI W

11:00 a.m.

11:45 a.m.

Discussion Reviewers 12:15 p.m.

Lunch SR5 assessment ISTL 1:15 p.m.

2:15 p.m.

Discussion Reviewers 3:00 p.m.

Break c)

Applicatida"of SR5 for PWR analyses Surry INTL 3:15 p.m.

modelling uncertainties INEL 4:15 p.m.

Discussion Reviewers 5:00 p.m.

Adjourn

\\

August 20,1996 d)

Application of SR5 for PWR analyses l(contfaned) i Most recent SRS application j

9:00 a.m.

i for P%Rs INEL 10:00 a.m.

Discussion Reviewers 10:30 a.m.

Break A perspective on the use of SR5 e) 10:45 a.m.

thermal hydraulic analyses (including affects of fission products transport)

NRC i

11:15 a.m.

Discussion / Comments Reviewers i

12:00 p.m.

Lunch 1:15 p.m.

Comments / Discussion All 2:30 p.m.

Adjourn i

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1 Review of SCDAP/RELAP5 Modelling for

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Assessment of Steam Generator Tube Integrity August 19-20, 1996 4

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REVIEW OF SCDAP/RELAP5 MODELLING OF NATURAL CIRCULATION UNDER SEVERE ACCIDENT CONDITIONS Objective:

Obtain independent assessment of the adequacy of SCDAP/RELAP5 modelling of l

natural circulation under severe accident conditions for the purpose of calculating the relative timing and failure of RCS components in order to evaluate the risk associated with thermally-induced steam generator tube ruptures Reviewers:

3 Consultants to Energy Research Incorporated (ERI):

Raymond Viskanta (Purdue University)

Mamoru Ishii (Purdue University) l Peter Griffith (Massachusetts Institute of Technology) l Schedule:

)

Review Group to provide its conclusion to ERI by 8/30/%.

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t REVIEW OF SCDAP/RELAPS MODELLING OF NATURAL CIRCULATIO l

SEVERE ACCIDENT CONDITIONS f

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1 BACKGROUND f

As part of the rulemaking on steam generator tube integrity, an issue has been raisl relative to the performance of flawed tubes and their likelihood of failure during a 9

i severe accident. The concern arises due to hot gases circulating through the steam j

generator tubes and inducing tube failure due to the elevated temperature of the r

Past SCDAP/RELAPS (SR5) analyses performed in conjunction with DCII issue resolution assessed the relative heatup and failure (using creep rupture models) of RCS O

components (i.e., hot leg, surge line, unflawed SG tubes and the RPV lower head)

The calculations considered a standard TMLB' sequence (with a pressurized secondary; system).

Current analyses focused on sequence with depressurized steam generator (on i

pressurizer h>op). Greater challenge to tubes but a lower probability due to S

failures required.

Additional analyses to address mixing /phenomenological uncertainties.

9 SR5 analyses are to assist in addressing the thermally-induced SGTR. Spontaneous 1

9 SGTR and pressure-induced SGTR are examined separately.

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i REVIEW OF SCDAP/RELAP5 MODELLING OF NATURAL CIRCULATION l SEVERE ACCIDENT CONDITIONS Past usage of SR5 calculations with natural c:rculation modelling:

SR5 use to address natural circulation in the reactor system as one of the major l

e areas of uncertainty identified in NURGE-0956 (" reassessment of the Technical Bases for Estimating Source Terms,7/86). Specifically, SR5 analyses were performed for a Surry TMLB' accident (NUREG/CR-5214,10/88) i SR5 was also used to assess failures of ex-vessel components (hot leg, surge line, r

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steam generator tube) vs. RPV lower head under the direct containment heating l

(DCII) issue resolution for PWRs. Results were peer reviewed.

Zion DCII issue resolution (NUREG/CR-6075, Supp.1,12/94),

Surry DCII issue resolution (NURGE/CR-6109, 5/95),

i DCII issue resolution for Westinghouse plants with large dry containments or subatmospheric containments (NUREG/CR-6338, 2/96)

Levy-LA, IIenry-FAI, Moody-GE, Modarres-Univ. of Md, Peer reviewers:

Sheppard-Cal. Tech, Ishii-Purdue i

SRS was also used to assist peer reviewers (Levy, IIenry, Ishii, Moody.

Corradini) in establishing initial conditions (e.g., melt mass and composition) for e

the CE DCII testing.

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i Summary of Results O

Plant analysis performed for representative designs Surry SV)

ANO-2 (CE)

Analyses consistently showed that for countercurrent flow severe accident conditions e

first failures occurred at surge line or hot leg nozzle 1

Surge line or hot leg failure occurred 20-40 minutes before SG tube railure Sensitivities done on T-II modelling did not alter finding on tube integrity i

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SCDAP/RELAP5 SEVERE ACCIDENT NATURAL CIRCULATION i

MODELING AND APPLICATIONS i

PAUL D. BAYLEss, INEL i

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NRC MEETING ON SCDAP/RELAPS CODE MODELING OF NATURAL CIRCULATION I

UNDER SEVERE ACCIDENT CONDITIONS August 19-20, 1996 BURR RIDGE, Il

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DUTLINE BACKGROUND SEVERE ACCIDENT NATURAL CIRCULATION FLOW DESCRIPTION SCDAP/RELAP5 INPUT MODEL DEVELOPMENT j

RELAPS ASSESSMENT WITH WESTINGHOUSE EXPERIMENTS PLANT APPLICATIONS t

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t WHY INVESTIGATE SEVERE ACCIDENT NATURAL CIRCULATION FLOWS?

NUREG-0956 IDENTIFIED NATURAL CIRCULATION AS A SEVERE ACCIDENT ISSUE.

ANALYSES FOR ORIGINAL NUREG-1150 DID NOT CONSIDER NATURAL CIRCULATION.

WESTINGHOUSE WAS PERFORMING NC EXPERIMENTS UNDER EPRI SPONSORSHIP.

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NRC WAS LOOKING AT THE PROBLEM ANALYTICALLY.

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t NATURAL CIRCULATION FLOWS TRANSFER HEAT FROM THE CORE TO OTHER RCS STRUCTURES.

UPPER PLENUM STRUCTURE MELTING i

RCS PIPING FAILURE HPME OR LPME (CONTAINMENT INTEGRITY CONSIDERATION) i r

ACCUMULATOR INJECTION t

SG TUBE FAILURE CONTAINMENT BYPASS i

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l FOLLOWS THE CORE LIQUID LEVEL DOWN ADDITIONAL COLD RETURN PATH (IN CORE BYPASS PLANTS) WHEN LEVEL DROPS BELOW BOTTOM OF CORE FORMER PLATES HOT LEG NATURAL CIRCULATION FLOW CONTROLLED BY MIXING IN THE SG INLET PLENUM.

PORY OPENING INTERRUPTS FLOW, BUT IT IS QUICKLY RE-ESTABLISHED WHEN VALVES CLOSE i

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DITTuS-BOELTER LAMINAR FLOW:

SELLARS, Nu = 4.36 i

FREE CONVECTION:

CHURCHILL-CHU (VERTICAL), MCADAMS (HORIZONTAL) f CODE USES THE MAXIMUM OF THE FORCED AND FREE CONVECTION HEAT TRANSFER L

COEFFICIENTS.

CONVECTION.)

(NEARLY ALL OF THE CALCULATIONS WERE IN TURBULENT FORCED t

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THE HOT LEG WAS SPLIT INTO TOP AND BOTTOM HALVES TO MODEL THE COUNTERCURRENT FLOW, SINCE THAT FLOW IS NOT POSSIBLE WITHIN A CONTROL VOLUME OF A 1-DIMENSIONAL CODE.

THE HOT / COLD FLOW STEAM GENERATOR TUBE SPLIT WAS SET TO 35/65%, BASED ON THE LOW PRESSURE EXPERIMENTS.

VARIOUS NODALIZATION SCHEMES WERE TRIED.

l THE NODALIZATION USED PRESERVED THE CHARACTER OF THE FLOW PATTERN WHILE MINIMIZING UNPHYSICAL BEHAVIOR.

THREE RADIAL RINGS WERE USED IN THE CORE AND UPPER PLENUM, CONNECTED BY CROSSFLOW JUNCTIONS.

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i HOT LEG MODEL BENCHMARKING FOR NATURAL CIRCULATION FLOW MODELING EPRI SPONSORED COMIX CALCULATIONS OF THE LOW PRESSURE TESTS, WITH GOOD l

AGREEMENT BETWEEN THE CALCULATED AND MEASURED RESPONSES.

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CONDITIONS.

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AND STEAM GENERATORS FOR A GIVEN HOT VAPOR TEMPERATURE ENTERING THE HOT LEG.

ALTERED THE SG INLET PLENUM VOLUME / FLOW AREA SPLIT, LOSS COEFFICIENTS THROUGH THE FLOW PATH l

DID NOT TRY TO MATCH FLOW RATES OR MIXING FRACTIONS I

WHEN REASONABLE AGREEMENT WAS REACHED, THE HOT LEG / STEAM GENERATOR INPUT MODEL WAS " FROZEN" AND USED IN SCDAP/RELAP5 SEVERE ACCIDENT CALCULATIONS.

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BALANCE.

THE SAME MODELING APPROACH USED IN THE PLANT CALCULATIONS WAS USED TO MODEL THE FACILITY.

LOSS COEFFICIENTS FOR HOT LEG NATURAL CIRCULATION WERE ADJUSTED BASED ON THE RESULTS FROM ONE TEST, THEN WERE LEFT UNCHANGED TO MODEL THE SECOND TEST.

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.e Topof esse l1 i

namen m ane y,,,,,,,

~0.0 8

0.5 0.4 0.2 0.0 0.2 0.4 0.6 Cistance from top of core (m)

Test $6: vessel outer channel 250.0

-l l

l

- "I i

6"M i 6 200.0 e

l 3

g x

,"r i*T B 150.0 A

Aa a

a 4

"3 5

.L

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a l. :t w

e 100.0 l.

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a F *

one 7,,,,,,,,,, a

! i 50.0-0.6

-0.4

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Figure D-12. Companson of expenmental and esiculated SF vapor temperatures in the core and upper 6

pienum for Test 5-6.

NUREG/CR 6235 D-25

I FLOW PARAMETERS FOR HOT LEG NATURAL CIRCULATION TEST S-7.

Prediction Percent error Experiment Left Right Left Right Left Right llot leg Flow parameters 110.7 114.9 128.3 128.3 15.9 11.7 Tri. oui ( C) 60.8 60.3 66.2 64.5 8.9 7.0 Tc,;n( C)

T.out. Tc,in ( C) 49.9

-54.6 62.1 63.8 24.4 16.8 3

110.4 108.6 104.3 104.3

- 5.5

- 4.0 ii pi,(kg/rn )

3 140.8 141.0 136.2 137.5

- 3.3

- 2.5 pc (kg/m )

pc _ pai (kg/m )

30.4 32.4 31.9 33.2 4.9 2.5 3

2.95 3.26 3.I6 3.33 7.1 2.I q58 (kW) 0.071 0.072 0.0631-0.0647

- 11.1

- 10.1 m,,i (kg/s),

t m.

HOT LEG NATURAL CIRCULATION FLOW PARAMETERS FOR TEST S-6.

t Predictiorr Percent error Experiment llot leg Left Right Left Right Left Right T. oui ( C) 129.7 134.9 137.2 137.2 5.8 1.7 Flow parameters 65.5 63.4 67.7 66.8 3.4 5.4 h

Tc,ic( C)

T,out - Ic.in 64.2 71.5 69.5 70.4 8.3

-1.5 h

3 77.I 75.8 75.5 75.5

-2.1

-0.4

( C) ph(kg/m )

Pc (kg/m )

98.9 99.6 97.6 98.0

- 1.3

- 1.6 3

pc - ph (kg/m )

21.8 23.8 22.1 22.5 1.4

-5.5 3

2.43 2.67 2.49 2.55 2.5

-4.5 qsg (kW) 0.0467 0.0461 0.0446 0.0451

-4.5

-2.2 rii (kg/s),

g t

STEAM GENERATOR. NATURAL CIRCULATION FLOW PARAMETERS FOR TEST S-7.

Prediction Percent error Experiment Left Riglit Left Right Left Right Steain generator Flow parameters 2.95 3.26 3.16 3.33 7.I 2.I (15g (kW)

Ntimber of hot tubes 72 NA 72" 728 Number of cold tubes 144 NA 144a g44a i

0.147 NA 0.131 0.132

-10.9 iii, (kg/s) 0.071 0.072 0.0631 0.0647

-11.1

-10.I iii,i (kg/s) b 2.05b 1.0 2.06 NA 2.08 i

in,/rin,,

0.89 NA 0.89b 0.89b 0.0 i

ffi2 69.5 NA 72.9 71.4 4.9 T a ( C) 45.9 NA 44.8 42.3

-2.4 i

Tci (*C) 23.6 NA 28.1 29.1 19.1 T a - Tct ( C) 67.I NA 72.2 70.9 7.6 i

Tm ( C)

Not predicted by code; input to code.

a.

Not predicted by code; stearn generator inlet plenurn loss coefficients and junction areas a b.

ing fractions and flow ratio.

N A = Data was not obtained in the experirnents.

s

STEAM GENERATOR NATURAL CIRCULATION FLOW PARAMETERS FOR TEST S-6.

Experiment Prediction Percent error Steam generator Left Right Left Right Left Right Flow parameters qsg (kW) 2.43 2.67 2.49 2.55 2.5

-4.5 Number of hot tubes 64 NA 72a 728 Number of cold tubes 152 NA 144" 144a iii, (kg/s) 0.0919 NA 0.0907 0.0891

- 1.3 siig (kg/s) 0.0467 0.0461 0.0445 0.0451

-1.7

-2.2 vii,/m W

M 2.@

1.9@

3.6 hi 0.85 NA 0.89b 0.89b 4.7 ff2 i

Thi ( C) 77.5 NA 76.4 75.9

- 1.4 Tci ( C) 44.5 NA 42.8 40.9

-3.8 Thi - Tci ( C) 33.0 NA 33.6 35.0 1.8 T n ( C) 73.2 NA 74.5 74.0 1.8 i

Not predicted by code; input to code.

a.

Not predicted by code; steam generator inlet plenum loss coefficients and junction areas adjusted to obtain mix-b.

ing fiactions and How ratio.

N A = Data was not obtained in the experiments.

m

I THE RELAP5 CALCULATIONS WERE IN REASONABLE l

AGREEMENT WITH THE MEASURED RESPONSE.

1 FLOW PATTERNS WERE THE SAME AS IN THE EXPERIMENT.

l l

TEMPERATURE PROFILES WERE THE SAME AS IN THE EXPERIMENT.

l CALCULATED HOT LEG AND STEAM GENERATOR MASS FLOW RATES WERE WITHIN 11%

l OF THE MEASURED VALUES.

j 1

VAPOR TEMPERATURES IN THE STEAM GENERATOR TUBES WERE WITHIN 5% OF THE MEASURED VALUES.

I VAPOR TEMPERATURES ENTERING THE HOT LEGS WERE OVERPREDICTED BY UP TO i

16%.

HEATING OF THE VAPOR IN DOWNFLOW IN THE CORE WAS OVERPREDICTED.

l 1

HEATING OF THE VAPOR IN UPFLOW IN THE CORE WAS UNDERPREDICTED.

t t

j SURRY STATION BLACKOUT (TMLB' SEQUENCE) l CALCULATIONS f

INITIAL CALCULATIONS WERE PERFORMED TO INVESTIGATE IF EX-VESSEL FAILURES

~

MIGHT OCCUR, WHERE THEY WOULD OCCUR, AND WHEN THEY WOULD OCCUR IN f

RELATION TO THE CORE DAMAGE PROGRESSION I

I s

t i

i s.

l

SCOPING CALCULATIONS PROGRESSIVE ADDITION OF NATURAL CIRCULATION FLOWS RESULTS AS EXPECTED:

MORE NATURAL CIRCULATION LED TO SLOWER CORE HEATUP j

t SURGE LINE FAILURE PREDICTED IN BOTH NATURAL CIRCULATION CASES I

f l

l l

'I l

l i

SENSITIVITY CALCULATIONS i

CODE /MODEL f

KNEW THERE WERE UNCERTAINTIES BECAUSE OF THE LIMITED DATA, l

RESTRAINTS, INABILITY TO BENCHMARK / ASSESS THE CODE AGAINST DATA.

[

CASES CONSIDERED:

- AXIAL POWER PROFILE CORE AND UPPER PLENUM CROSSFLOW RESISTANCE STEAM GENERATOR INLET PLENUM MIXING 1

HOT LEG / SURGE LINE PIPING HEAT LOSS i

HEAT TRANSFER COEFFICIENTS IN THE UPPER PLENUM, HOT LEG, AND STEAM i

GENERATOR TUBES I

l SIMULATED RADIATION HEAT TRANSFER BETWEEN THE HOT LEG FLOW STREAMS I

1

i Base case used best-estimate values l

for the sensitivity parameters.

i Surce

,ine f ai,ure a~: 246 min e

e initial fuel rod relocation at 248 min e

75% of core heat removed by coo ant 4% to hot legs 19% to steam generators LCOOO3??

i

-I

A Steam generator tubes were much cooler than the surge line and hot legs.

1500 creep O

1101 leg rupture

<D 5

O Surge line

' " I'""

a steam generator tubes m

1250 ba E*

~

i 1000 M

E r-A 750 w

m s

8 8

i i

d 500 160 180 200 220 240 260 Time After Initiation (min)

I rcooo4e I

Piping heat loss effects were primarily local.

Convection anc 1

Convection rac.iation 7 min 13 min Surge ine aiLure de,ay Fue rocL re ocation delay 2 min 4 min 75 %

76 %

Core heat removal 5%

5%

To hot ecs To steam generators 18 %

17 %

To containment 3%

4%

10000380 1

i 1

SIGNIFICANT REDUCTIONS IN THE STEAM GENERATOR l

1 INLET PLENUM MIXING HAD A SMALL IMPACT ON THE CALCULATED RESULTS.

MIXING FRACTION OF 0.7 IN THE PRESSURIZER LOOP, 0.3 IN THE OTHER TWO LOOPS I

SURGE LINE FAILURE AT 255 MIN l

l INITIAL FUEL ROD RELOCATION AT 254 MIN i

HOT LEG FLOW INCREASED 25% COMPARED TO BASE CASE 77% OF CORE HEAT REMOVED BY COOLANT i

3% TO HOT LEGS l

24% TO STEAM GENERATORS i

t

[

t r

i CONCLUSIONS FROM THE SENSITIVITY CALCULATIONS e

l IN ALL OF THE CASES, EX-VESSEL PIPING FAILURES WERE PREDICTED TO OCCUR ABOUT THE TIME OF INITIAL FUEL ROD RELOCATION WITHIN THE CORE.

THE ONLY CALCULATIONS THAT HAD A NOTICEABLE DIFFERENCE FROM THE BASE i

CALCULATION WERE THE INLET PLENUM MIXING SENSITIVITIES.

SIGNIFICANTLY REDUCING THE MIXING FRACTION (TO 0.7 IN THE PRESSURIZER LOOP AND 0.3 IN THE OTHER TWO LOOPS, FROM 0.87 IN THE BASE CASE) BROUGHT THE STEAM GENERATOR TUBE TEMPERATURES HIGHER, BUT THE SURGE LINE STILL WHEN THE MAXIMUM SG TUBE TEMPERATURE WAS ABOUT 360 K LOWER FAILED FIRST, THAN THE SURGE LINE TEMPERATURE; IT WAS ABOUT 410 K LOWER THAN THE SURGE t

3 LINE TEMPERATURE IN THE BASE CASE.

i h

i

Armcass, 4 WESTINGHOUSE 1/7TH SCALE EXPERIMENTS R. E. Henry and M. Epstein Presented for EPRI to NRC Review Group on Natural Circulation August 19, 1996

1 Approach to Scaling NRC Severe Accident Scaling Methodology (SASM)

Need to have an experiment which has all the physical processes even if the scaling is -

not perfect. This greatly aids the formation of Process Identification and Ranking Tables (PIRT). The W/EPRI experiments have all the dominant processes.

Two types of considerations:

Top Down Scaling, Bottom-Up Scaling.

Both of these approaches are used to evaluate the W/EPRI experiments and to apply the understanding to the reactor system.

i\\H VGilot195 A

i MAAP 4/3B HLNC l

r-i 4

i Steam Generator Wa s

(Total Flow)

N i

i t

i

~

/

"Out" Tube y

4 "Back" Tube 1

1 1

4 SeC s

Inlet Outlet j

Plenum '

co

' Plenum l

{

Tg Tc j'

)

T Wgt UP

[

_ ~ - ~

l, Wst

( Hot Leg -

1 i

1 I

i AHS25005 CDR 0105-94

),

1 i

)

Figure 1 Hot leg and steam generator natural circulation flow model.

I i

J DATE: 05/01/94 j

VOLUME II REVISION: 0.0 4 i

k Top Down Scaling Mixing in the SG Inlet Plenum i

Necessary Conditions To establish the appropriate mixing behavior in the inlet plenum, it is necessary to have the same geometry even though it may be in a scaled down system.

The EPRI/W experiments represent the RPV, hot leg, surge line and SG geometry.

i The ratio of the flow through the steam generator tubes to the hot leg circulation flow should be preserved in the experiment.

I M *G\\l0lt93.A

1

)

Natural Convection Flow In the Hot Legs i

Countercurrent natural circulation flow through the hot leg

.l WHL

CFC!E

- PUP P UP L

H l ' HVG'.10l t 95 A

Flow Through the SG Tubes The turbulent momentum equation for flow through the tubes is given by L

W*2 AP=Apgha=f 2

D 2pA e

1 g s a reference height.

i where h Solving for the flow through the tubes results in

'1/2 2,2D e

n(c -

n) 8 h-W=N D

a e

1 4

.fL I W VGilol195.A

Ratioing the two flow rates results in i

)

l

~

' /

f W,

pn (pc - pH)

D, '2 2 D, li r

g fLD

. UP H ~ PUP,

(

FCj (DHLj ut, W

P HL The density ratio term is of order unity.

The ratio of this flow should be preserved between the reactor system and the model, i.e.

the left hand margin can be assumed to be a constant.

Therefore:

D" 3 D, h ' M f

fD 32

'2 D li g

i i

g t

N,

=N (DHL>so HL.so (Dutj,jLDut[m 4

fLD SG - Steam Generator m - experimental model 1 Sci 101195.A

-......n.a

..w-.

.-...u.

-a a-.

s-e-

.s

.e-..

a It is further assumed that 1) the ratio of the effective height driving the natural circulation and the length of the tube is approximately the same in the scaled experiment.al model, and the steam generator, I

2) the frictional coefficient is similar for the two systems, which implies a similar Reynolds number in the steam generator tubes. Finally we arrive at the expression

- D,

2.5 D

HL SG N

1 N

=

sc D,

,5 m

r HL.m I'HYGil01195.A

Comparison 4

i

=

j Steam Generator N g = 3260 3

Tube inside diameter = 0.775 in.

i i

Diameter of the hot leg = 36 in.

2 Experimental Model I

Experimental model tube ID = 0.305 in.

4 Hot leg diameter = 5.14 in.

i N

= 258 (216 used m the model).

l m

Hence, it is reasonable to assume that the mixing behavior observed in the Westinghouse

~

experiments is representative of that which would be experienced in the reactor system.

l ' HVG'.101195. A

i 2

Bottom-Up Scaling 4

Major Elements of Plenum Mixing i

.+

1.

Hot fluid rising through a scaled steam l

generator inlet plenum geometry.

l Experiments have this.

j 2.

Entrainment of surrounding fluid with the i

appropriate (scaled) rise height.

Experiments have this.

l 3.

Spreading " ceiling plume" if the hot tube outflow cannot transmit the entire plume.

1 l

Experiments have this.

l:

4.

Mixing of returning cold tube flow and the j

excess flow of the " ceiling plume".

j' Experiments have this.

i I

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figure 4-71.

Iemperatures (*C) of SF in Steam Generator Channel Itcad and flot Leg. Iocations Shown Isometrically i

6 from Central Plane. Steady Cooling by Water on Secondary Side. Test SG-53.

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l Figure 4-69.

S. G. Inlet Plenum Mixing - Test SG-53.

Isotherms in a Vertical fh Plane Through the Hot leg Center Line.

.h.,

. b p-d'!

it; Q

t..

N O

i

y li i

.c 4-97 i l l

l

P

" Top Hat" Model Comparison Willi Westingliouse SF Test Data 6

Test Variable Units SG-S1 SG-S2 SG-S3 SG-S4 SG-T1 SG-T2 SG-T3 SG-T4 l

6768 6700 3582 3362 Time sec Power kw 22 22 30 30 22 22 30 30 Pressure har 20.7 27.6 20.7 27.6 20.7 27.6 20.7 27.6" T,

C 123.7 114.8 159.3 143.2 248.7 253.4 250.0 240.5 i

'I'c C

70.3 72.1 86.8 86.2 142.4 171.7 126.4 146 T,,

C 79.7 80.2 100.8 98.4 165.8 185.8 153.4 159.2 i

(I'i,,)iiiax,tiana 85.4 83.3 106.5 101.6 181.3 199.4 170.9 177.7 T,,,,,,,,ici C

84.8 84.2 105.5 101.5 168 192 15g 169 l

i E = 0.1 o

T,,,,,,,,ici

C 83 83 103 100 164.5 190 151.5 166 i

E = 0.116.

o T,,,,,,o,ici C

82.8 82.5 102.8 99.5 164 189 150.5 165.8 i

E = 0.12 n

t

! O h YO s't l 4k!I! h. U

--r a --

i Model Comparison With l

Westinghouse Water Test Data 1

H,0 E,0 Variable

-50

-51

~

Time Power 29 29 Pressure 1.0 1.0 T

38.0 26.2 h

T

.30.0 21.0 c

T 32.9 23.5 ht (T

) max, data 32.9 23.5 ht T

ht model E,= 0.1 o

T 32.5 22.8 ht model E

,= 0.116 o

T ht model E,= 0.12 o

T 32.9 23.2 ht Gaussian E

,= 0.085 n

l l

! 9fVC' TABLE C

e Gaussian Plume Model 2

r u=u exp m

R,

=

~

T-T, r 2

= exp 2

12 R_

T

- T, m

-u

- centerline velocity.

m T

- centerline temperature.

ni Entrainment coefficient = 0.082 for Gaussian distribution.

Boundary Conditions R(o) = R'o, u (o) = u, T (o) = T',

m o

m Where 1 + A2-T', = T

+ (T -T) o R'o = y/f Ro l

! H v G\\t01195 A

" Gaussian" Model Comparison With Westinghouse SF Test Data 6

Test Variable Units SG-Si SG-S2 SG-S3 SG-S4 SG-T1 SG-T2 SG-T3 SG-T4 Tiine sec 6768 6700 3582 3362 l'ower kw 22 22 30 30 22 22 30 30 -

l'ressure bar 20.7 27.6 20.7 27.6 20.7 27.6 20.7 27.6 T,,

C 123.7 114.8 159.3 143.2 248.7 253.4 250.0 240.5

'I'c C

70.3 72.1 86.8 86.2 142.4 171.7 126.4 146 T,,

C 79.7 80.2 100.8 98.4 165.8 185.8 153.4 159.2 i

(l'i,,)iiiiis,d,,.,

85.4 83.3 106.5 101.6 181.3 199.4 170.9 177.7 16h 176.57 T,,,c;,,,ss;a,,

C 89.5 88.0 112.0 106.4 176.5 199 i

E,, = 0.085 t

tiivut i a til i

PLUNE VERSUS STEAM GENERATDR FLOW RATES Plume Flow' Floiv in Steam Just Below Tube Generator, Test Sheet, kg s-1 kg s-1 SG-S1 0.250 0.114

~

SG-S2 0.345 0.146 SG-S3 0.243 0.120 SG-S4 0.330 0.137 SG-T1 0.208 0.036 SG-T2 0.239 0.083 SG-T3 0.220 0.101 SG-T4 0.283 0.136 H 0-20 0.289 0.024 i

2 Calculated with axisymmetric Gaussian plume model.

! ' HVC\\ TABLE b

)

4 i

l l

Mixing of Excess Plume j

Flow and Returning =

l Tube Flow

- T ) = n1, gc - T )

- NIht)

(b ht Plume c

3 ct L

I

.Hi ht R=

b

-b Plume ht i

T

+RT ht ct T' _

1+R

! \\HVC'.101195. A

Comparison of Measured and Calculated Plenuin Mixing Temperatures T

T c

c Test R

T T

(expt.)

(plume inodel) iit ct SG-SI 0.838 83.0 55.4 70.3 70.4 1

SG-S2 0.734 83.0 57.3 72.1 72.1 SG-S3 0.976 103.0 64.7 86.8 84.1 SG-S4 0.71 100.0 65.2 86.2 85.5 SG-Tl 0.209 164.5 115.8 142.4 156.1 SG-T2 0.532 190 152.0 171.7 176.8 SG-T3 0.849 151.5 103.3 126.4 129.4 SG-T4 0.925 166.0 120.9 146.0 144.3 II 0-20 0.0906 32.9 10.1 30.6 30.0 2

t I slit t al Al41 t.

?

.3 i

y SCDAP/RELAP5 SGTR Analyses i

Idah D.L.Knudson National t

Engineering L a b ora tory

]

SCDAP/RELAPS Natural Circulation Model Review 3l Burr Ridge, Illinois j!

August 19-20,1996

{

4 G,

i i

l Presentation Overview SCDAP/RELAP5 calculation oajectives SCDAP/RELAP5 Surry loop calculations Surry loop results SCDAP/RELAPS Surry plant calculations Surry plant results l

Conclusions i

1 SCDAP/RELAPS Calculation Objectives Evaluate variations in hot leg countercurrent natural circulation with respect to SG tube temperatures using a Surry (stand-alone) loop model Evaluate the potential for natural circulation-induced RCS pressure boundary failures, including SGTRs, using a Surry (full) plant model i

l i

SCDAP/RELAP5 Surry Loop Calculations Based on a stand-alone model of the Surry primary coolant loop containing the pressurizer j

Boundary conditions to drive the loop model were extracted from Surry plant results for a TMLB' transient without i

recovery; without operator action; without oxidation; and with modeling provisions to allow development of in-vessel, l

full loop, and hot leg countercurrent natural circulation All loop calculations initiated at the onset of countercurrent flow (9200 s) and extended ~5000 s (corresponding with surge line failure in the plant calculation) with variations in l

hot leg countercurrent natural circulation conditions

l

~

SCDAP/RELAPS Surry Loop Calculations Variations considered

-num 3er of tubes participating in forward (hot) flow (measured between 29 and 61% of the SG tube bundle)

-mixing fraction (measured between 0.76 and 0.89)

-recirculation ratio (measured between 1.69 and 2.39)

Separate loop calculations were included so that SG tube temperatures could be evaluated over the measured range i

All loop calculations were completed with variation of only one parameter at a time (all other paro oeters were he d l

constant at " average" conditions) i

/

Surry Loop Results 900.0 i

i g

O O 29% hot tubes s.

A A 35% hot tubes (Base) e y

e e 53% hot tubes

, s;-

800.0 A---A 61% hot tubes E

?

3 8

5 700.0 3

i g

i v) g s

600.0 9000.0 11000.0 13000.0 15000.0 Time (s) k

Surry Loop Results

~

900.0 i

/.t/

e

-y O

00.76 mixing fraction 3

A A 0.87 mixing fraction (Base) 3 V

7 0.89 mixing fraction q

a wo.

800.0 E

i s.

m

~

I. 3 'i..

~

t U

/

3 Yn 700.0 g

0 in 2

600.0 9000.0 11000.0 13000.0 15000.0 Time (s)

a b'

0 l

0 0

~

0 5

1

/

0 j

s 00 t

0 i

l 3

i u

1 s

e

)

/

s 8

R e

(

m p

T i

o

)e

)

o a

0 s

L B

0 n( n

' 0 y

i ono 0

ioi t

t 1

aia t

1 r

al l

u ul r

cuc r

r ci u

ici c r

ece S

r r e 5,

r 9

j 692 1 1 2 0A7 CAV 0

0 0

0 9

0 0

9 8

7 6

M $EgoE3 838i o.3 ow 2

l 4

i SCDAP/RELAP5 Surry Plant Calculations i

TMLB' transient in all calculations

-wit 7out recovery and without operator action i

-with creep rupture monitoring of the surge line, hot legs, and SG tubes

-with modeling provisions to allow development of in-vessel, full loop, and hot leg countercurrent natural circulation 1

i Benchmarked with

~9% of core energy deposited in structures in each loop

-35% of SG tube bundle available for forward (hot) flow l

-mixing fractions at ~0.87 7

,.I.

l

-recirculation ratio at ~1.9 I

Vessel Nodalization w~

190 i

k

, k

k

I s

c 174:

164:

1543 144:

1343 3

s s

s l:

2 g

100 1733 163!,,-.1533,

. 143;,->1333

'~

200.300.400 :

[,

[3 1

1, s

172!d 1623d 152Ed 142 132! T 222,322,422 -l 102 f

1, g

104 {

17 13, 161E 151E 141!

131!

a 1

T, s

o n

o L,__.

i i

t 3

3 s

s s

-+

t i

3, a

-+

s, s

t s

111 112

{;

E 115 --*

114 113 -*

.e s

gg

-+

s s

y s

-+

a t

3 i

d

,t

-+

s R

s i

3 3

-+

o n

a n

o a

if 3

d if l 5

108 s

o if

(

1%

N

F n

o s

m ne n

uu i

a.n t

eC a

9 z

44 i

E l

a r

E y5 d

a4 j

e s4

[

e P

[

o 2[,

jj t.

'l m

E 4

t a

i l

v#

e 3

e m'

n

, 3 S

N R

Y.

n e

, 4

[

?

l, 0

t 4

E 2

l p

E 5

l

'~

o 8

s 9

4 1

^

o 1

2 4

p s

1 d

L 4

3 I

n F

d,",.

r

?

Yi e

d 4

[4 z

m R.

i i

i TI

- z, I

I r

s

- m'>r i i

i 4

u

=

e 1

+ m 1

l I

4 "

u j

0 2

4 8

o s

[I 4

4 l

I i

l s

_,4 T

e mm m

rP

SCDAP/RELAPS Surry Plant Calculations Case Assumed Severe Accident Condition 1 2 3 4 5 6 7 x

x x x

x x x TMLB' w/o recovery and w/o operator actions Depressurization of pzr loop SG secondary via failed ADV x

x x

w/o RCS depressurization via pressure boundary failure x

x x

x x RCS depressurization via pressure boundary failure x

x x x x x

x 35% hot tubes /65% cold tube nodalization i

x x

53% hot tubes /47% cold tubes nodalization Depressurization of all SG secondaries via failed ADVs x

a

- ~

.e J

s a

a 4

4 m..

a n

am..

O e

l O.

O O

3 O

i O

l N

O G

d (r) 3

- 8 m

m W

O I

w O^

d

- 8$

~

G ss E

N y

C CU

~

O

- O O

f LO u

6 f

3

(

D

_0 O

O O

d d

0 0

E N

,~

'~

(edW) eJnsseJd

Surry Plant Results - Case 1 15.0 i

1 E

55 10.0 -

_m_

32 8

Top of fuel ue u>

Q.

m 5.0

=O Bottom of fuel O

m 0.0 O.0 5000.0 10000.0 15000.0 20000.0 Time (s)

~

Surry Plant Results - Case 1 30.0 i

b.

I; ljl i

l 20.0 l

b lj 7;

10.0 Me g

g d M_apflLJ

-f ?

d6 f -O--G -O o o _o o n

~

/>\\un\\%~t f

?

-10.0 7

8 G--O mflowj-111090000 0

A A mU wj-112090000

-20'0 V

7 mflowj-113090000 i

A C

C mflowj-114090000

-30.0 A---A mflowj-115090000 L

3 n

_49_9 9000.0 14000.0 19000.0 Time (s)

~

Po O

I (J

i O_

OO g

88 OO

,m OO OO OM 3

iT r

gg JL 88 m

- cE O <J g

O o <a

~

N

~

g w

r-oO 3

Y

_ 85 M

o m

Pr.E

~A L

,- 1--

O z:

[

R TI

~c 6F Je g

9A

-s J A1 g

n

/

- _=

M

==

=

L U

qLi.E _.

~

- _1

. =

c1 0

G g

o 9

9 9

98 o

o o

o N

w (s/ >t) ele; mol} 6e!loH 6

4 s

1 g

9 o

i i

i cj h

OO 88

/ #

OO i

b5 i

OO l

P TT gg G

22 1

W EE COQ O <3 O <3 W

~

-R, W w

J 97 U

c 1

_ 8 W

=

1 o g f

"A Z F-CE E

i d

t C

C d

t cg 4

L

_1 t

w C

J W

9 aa 9

9 9

9 98 o

m o

N v--

v-(s/6>i) ele; mol} eqnt es y 6elloH

-I I

'l s

C.

0 0

06 1

1 e

s 0

a 0

'0 C

04

~

1 s

)

t s

(

lu e

m s

T i

eR

['

0 0

t 0

n 0

2 a

1 lP en e

i gb l

y eeu gl t

r uoG r t r

SHS u

OAV S

OAV 0

0 0

01 0

0 0

0 5

2 1

1 2;53 oE$.23gN 8o_ENE b i

Surry Plant Results - Case 1 1.00 O

A i

i g

O O Surge line B

A A Hot leg

's V

V SG tube E

E m

u O

y 0.50 Sa E

-e a

E e

o_

E 0.00 vea OA ' Tba UAMA OA'OB-e v

10000.0 12000.0 14000.0 16000.0 Time (s)

- -, - - - +

4

a.--,_a,-

u

_nn-.

-.4 4

r-A 0

O O,

O O

O I

i I

ON 3

O6 G

8 cf) m

~

cgo I

~

=

O-O e OO.E

.: E l

'~ >

5 i$md

I Y

to i

O c

i L.

3 s

_ O, I

I O

O O

O l

d b

s-N (edlN) eJnsseJd

Surry Plant Results - Case 2 15.0 i

i

^

E 3

10.0

_a) o

.3 Top of fuel a-

=

oe u)o_

m 5.0

=oo Bottom of fuel

=

0.0 O.0 5000.0 10000.0 15000.0 20000.0 Time (s) i I

r oo o

?

o N

cio G

S W

CO O

~

M es a-

=3 ge

_ n.5 y)

-s g

CC

C CU q

0

- 8 1

e r

go

$2 3 6

moo w

wIe v

3 O <1 C>

O D >

O <3 C>

(9 o.o

~

o og o

o o

j o

o o

o (O

N CO v

i (y) eJn}EJedujOl OJnlonJ}S dOO] JOZpnSSOJd l

Surry Plant Results - Case 3 2000.0 i

i i

2 2

O O Surge line j

A A Hot leg (O-5 V

V SG tube

(

o_

E 1500.0 2

a E

S o

1000.0 B

-e a

u>

2 o_

500.0 10000.0 11000.0 12000.0 13000.0 14000.0 15000.0 16000.C Time (s) t-t n

r Surry Plant Results - Case 3 A

1.0 i

i i

t X

O O Surge line

.@ 0.8 A

A Hot leg

.E V

V SG tube m

O E

m o 0.6 o_

m O

Q.

_8 0.4 5

~

.$a 8 0.2 o_

0.0 -n-e-n o n-e-E-d-n-e-n-e-n O E4-10000.0 11000.0 12000.0 13000.0 14000.0 15000.0 16000.0 Time (s)

Differential pressure (MPa) o o

o 6

6 6

9 go N

m Od O

a i

i a

O" ea 5

M

~

E C

O q

4 o

oo O

e E

w

=

==

3 97 o-m

_ oe e

oE

-W 2

E l2 e

W E

\\

C C

c

i CU l ?

E o

""1 L

6

_ 6 oo m

66 3

C o,

l e

o o

d 6

6 6

d O

O o

N o

CD o

v s-v-

(y) eJnlEJedWel OJnlOnJ1S leay pe6eJOAe-awn l0A

a

,I e

C.

0 0

0 y-5 1

3g 5

es 0

a 0

0 C

i 03 1

s

)

t s

(

lu e

m s

T i

eR 0

0 t

i 0

n 0

i 1

a 1

lP 1 5 ee y

ss aa r

CC r

OA u

S CA 0

0 0

009 0

0 0

9 8

7 6

1

@ 83~$E ea 8 := l s n O8 l y ?n8

i l

i i

i a

Surry Plant Re.sults - Case 6

~

20.0 i

i i

.S I

C O Case 3 l

A A Case 6 15.0

^

1 q

O i

r G6 h

G 2

3 10.0 0

g( (g m

l i Win i

o j

ku u

$ lb eg g jk,nrd ;

\\~ s g 5.0 d

l L

0.0 t

9000.0 10000.0 11000.0 12000.0 13000.0 14000.0 15000.0 Time (s)

F

Surry Plant Results - Case 6 i

i i

20.0 O

O Case 3 A

A se 6 j

15.0 -

blI 1 fII f

I q

k

(

ve(gcn,r e,ow 10.0 1

a 3

\\

(g<'~nrf

(

t 1

3 1

r 0*

5.0 h

C l

9000.0 10000.0 11000.0 12000.0 13000.0 14000.0 15000.0 0.0 Time (s) t e

4 9o i

i i

l o

o o

4 o

(D o

i co i

(U O

o 6

s o

o N

W "g

w v

3 o

E y)

H 9

e o

E

- 8 wC g

e o

1

.E e

en.o O

O D a E" - O

- 8

~

ao o

y en r en 3

O <3 C>

3 Gi O

o < C>

t>1 o

O d

a 9

9 o.

9 9

98 o

o o

o LD C*)

o N

LD (y) eJnlEJOdWel OJnlOnJ)S dOOl JOZpnSSOJd l

i

=--.

90O I

i O

m 9

O O

O*

D G

O CO 5

(0O 8

m E

9s O*

O e Om.E_.

i (p)

,- i_.

G 9

O w

- 8 C

_U

~

(

k

~

me 5

88

_ O mm O

y 00 0

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W ee 9

O i

98 9

9 9

o O

O O

O cn s

m m

(y) eJnlBJedwel eJnlOnJls eqnl es

Surry Plant Results - Case 6

,n 1.0 C

O Case 3 x 0.8 -

A A Case 6

-8.s O

&y0.6 mo m

0 0.4

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M 0.2

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'^ ^

' ^ ^

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0.0 AD i

9000.0 10000.0 11000.0 12000.0 13000.0 14000.0 15000.0 Time (s)

Surry Plant Results - Case 7 20.0 i

i y

(/

15.0 -

-m O_2

~

Pn (D

CD 10.0 1

6 I

I 5.0 5000.0 10000.0 15000.0 Time (s)

.I I

I Surry Plant Results - Case 7 15.0 i

i

\\

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')

g 10.0 t

\\

32a Top of fuel a-teu)

O.

m 5.0

=oo Bottom of fuel e

0.0 O.0 5000.0 10000.0 15000.0 Time (s)

Surry Plant Results - Case 7 20.0 i

i i

i O

O mflowj-400010000 A

A mflowj-430030000 t

15.0 I

r t

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il n

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I

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Surry Plant Results - Case 7 C

S-1.0 i

i i

i

~

g O

O Surge line E 0.8 A

A H t ieg V

V SG tube m

?

E m

V 0.6 O

e o

1 8 0.4

]

~

$ 0.2 e

o_

00 m m

bX 3

3 4-N-G-d-G-M 5000.0 7000.0 9000.0 11000.0 13000.0 15000.0 Time (s) t 4

e

Conclusions l

l Maximum SG tube temperatures increase with

-a decrease in the number of tubes participating in forvvard (hot) flow

-a decrease in the mixing fraction

-an increase in the recirculation ratio t

SCDAP/RELAP5 results are consistent with the expected SG tube temperature effects for variations in hot leg countercurrent natural circulation Variations in hot leg countercurrent natural circulation (within the experimental range) have a negligible impact on l

SG tube temperatures

i i

i Conclusions Pre.ssurizer surge line creep rupture was the first RCS i

l pressure boundary failure in all Surry plant calculations l

considered The pressurizer surge line failed in the early phase of core I

damage (before the onset of fuel melting) in all Surry plant j

calculations considered If pressurizer surge line and hot leg failures are ignored AND l

a SG secondary ADV fails open, SGTR could occur ~15 to 20 min after the first RCS pressu.re boundary failure i

l

.l

i I

i t

USE OF SEVERE ACCIDENT THERMAL-HYDRAULIC ANALYSIS RESULTS lN RISK ESTIMATE FROM 1

STEAM GENERATOR TUBE FAILURE i

AUGUST 20,1996 t

1 v

i El l

X I

Joseph Donoghue, NRC Telephone: (301) 415-1131 l

k, l

i

BACKGROUND

~

SGTR risk contributions from spontaneous and induced tube failures

- Spontaneous failures often due to unknown mechanisms

- Induced Failures:

Mechanical - Circumferential Crack failures Pressure - ATWS, Secondary Depressurization Thermal - Severe Accident Previous studies concluded that thermally induced tube failure not a significant concern during severe accidents (e.g, NUREG 1150, DCH studies)

- Same conclusions reached for degraded and pristine tubes t

(NUREG/CR-4551, " Evaluation of Severe Accident Risks: Qdantification of Major input Parameters," December 1990) 1 4

SEVERE ACCIDENT TUBE CHALLENGE Event tree developed to include thermally induced SGTR resulting from o

core damage events i

e Considerations:

Events leading to high tube temperature Events resulting in high or intermediate RCS pressure and dry SGs Frequency range of E-5 for thermal challenge from NUREG 1150 analysis e

for Surry and comparisons with available information in IPE Database Starting an order of magnitude from surrogate safety goal Applicability to other designs considered:

f e

No PORVs in some CE plants Potential for RCP seal LOCAs, loop seal clearing Differences in severe accident progression l

1 SEVERE ACCIDENT TUBE CHALLENGE Thermal-Hydraulic Analyses:

Use representative plants to analyze most likely thermal challenge

[

scenario Surry Base Case: SBO, loss of AFW, One SG depressurized e

Arkansas Nuclear One, Unit 2

- CE design without PORVs e

Other cases:

i

- RCP seal LOCA

- Primary-to-secondary leakage

- PORV fails open

- All SGs depressurized r

i b

l TUBE FAILURE PROBABILITY l

Calculate probability of SG tube failure PRIOR TO surge line or hot leg o

for any given crack size. Result is conditional tube failure probability o

Steps:

- SCDAP/RELAP5 analysis provides temperature and pressure histories

- Calculate creep failure times for each component

- include material uncertainties

- Include effects of tube flaws i

- Include thermal-hydraulic uncertainties:

l

- Relative time to failure

- Conditions leading to tube failure

- Other factors:

Tube dimension uncertainties, Crack paranleter variability

y

'i i

[

Surry Case 3R j

i i

Variable Pressure-1.0

..- 7 il from RELAP/SCDAP il i

1 ij I

L f

0'8 i

CD Index.SM = 1.0 y

l t

Surge Line CD Index R/S l

i 7

'I x

-- Hot Leg CD Index R/S l

- Tube CD Index R/S l

E 0.6 I

[

e CD I

I m

i E

I m

O 0.4 1

l i

ii 1

I I

E I

b 0.2 i-i i

[

t g

f t

0.0 0.0 100.0 200.0 300.0 Time (min) l

Surry Case 3R Variable Pressure 1.0 from RELAP/SCDAP

,l i

,' l il il 0.8 ll

~

CD Index SM = 2.0 it Surge Line CD Index R/S ll

'l

- - - - Hot Leg CD Index R/S x

- Tube CD Index R/S I

.E 0.6 I

e 8'

E i

i.

5o 0.4 l'i ri i ll It 0.2 l,i I'

j t

0.0 0.0 100.0 200.0 300.0 Time (min)

Surry Case 3R 1.0 E

i lI Steam Generator Tube &

Surge Line l

'\\ \\

0.8 j l l

i li ANL Up Limit ll l CDI from R/S l! I l If f ll x 0~6

- - - - ANL Mid

-8

- -- - ANL Low Limit gj i dI e

i Surge li i o

E Line N

Tubes pl m

i o 0.4 il l l Il I '

q! I k

i i;

0.2 li Ili I

$ I).

ti l Ib '/

i i

0.0 O.0 100.0 200.0 300.0 Time (min)

Surry Case 3R 1.0

.1)

Ei I

Effect of Temperature on il i!'.

Steam Generator Tube Creep Rupture il 0.8

IIil Median ANL Equation

,l

+25 deg-K

'l

- - - - +50 deg-K ll g 0.6 u

I e

O si E

,':i m

o 0.4 i.

lilj tj

'!i; i

af ll 0.2 il lt

.i II I!

E!

I;,

r-0.0 0.0 100.0 200.0 300.0 Time (min)

O M

L

CONTAINMENT BYPASS PROBABILITY e

Apply conditional failure probabilities for each crack size to flaw distribution covering range of crack sizes Flaw distributions generated for three plant categories:

Good, Average, Poor Preliminary results for " Average" plant:

- Assuming nominal tube temperatures, Tube failure probability about 6%, Bypass < E-6

- Assuming higher tube temperatures, Tube failure probability about 20%, Bypass = E-6

" Poor" plant distribution being revised, but initial result yielded e

Tube failure probability of 100%, Bypass = E-5

RISK ESTIMATE KEY ISSUES Event Tree Quantification:

- Event frequency in E-5 range is consistent with range in IPE survey Thermal-Hydraulic Modeling:

- Creep failure prediction dependent on understanding temperature and pressure time histories Representative Flaw Distribution:

- Effort ongoing to revise poor plant distribution

- Also working to understand uncertainties Tube Performance Model:

- Based on high temperature tube tests

- Includes material uncertainties

- Workin'g to understand flaw characterization uncertainties RCPB Weak Points: Only qualitative treatment for potential of other component failures (other than hot leg, surge line, tubes) 6

.~

SG SEVERE ACCIDENT RISK ESTIMATE EVENT TREE KEY ASSOCIATED T-H

[

ANALYSIS

(

hs7o#'

SEQUENCES I 1

ry

~

\\ Uncertainty _ f,/

s x/

R tt TUBE Q

,x CONDITIONS N's.s

'N w RELATIVE FAILURE TIMES Jf TUBES /SL/HL q' /

CREEP FAILURE MODEL

'N/

TUBE FAILURE,,

/,

PROBABILIT(

' l'- ~

i L.i.

CONTAINMENT 4

' ~ "

FAILURE O

PROBABILITY Material

//

rN DISTRIBUTION (/

FLAW Uncertainty and

(

Crack Paramotor

/

Variability

//

Basis and v,

'[

Gmwth Rate /

SURROGATE z'

lyy#[j"fy SAFETY GOAL

\\

'x

._ f

-

ML20134K080

Contents

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