ML20237B871

Suppl 4-A to Rev 1-P to Response to NRC Request for Addl Info for Verification of Analysis Methods for Small Break Locas

ML20237B871
Person / Time
Site:	Maine Yankee
Issue date:	11/30/1986
From:	ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY
To:
Shared Package
ML19302D186	List: ML20237B859 ML20237B871
References
CEN-203-P-S04-A, CEN-203-P-S04-A-R1-P, CEN-203-P-S4-A, CEN-203-P-S4-A-R1-P, NUDOCS 8712170212
Download: ML20237B871 (88)

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> RESPONSE TO NRC REQUEST FO'R ADDIT 10NAL INFORMATION FOR ,

, VERIFICATION OF ANALYSIS METHODS

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k LEGAL NOTICE THIS REPORT WAS PREPARED AS AN ACCOUNT OF WORK SPONSORED BY COMBUSTION ENGINEERING, INC. NEITHER COMBUSTION ENGINEERING NOR ANY PERSON ACTING ON ITS BEHALF: l A. MAKES ANY WARRANTY OR REPRESENTATION, EXPRESS OR IMPLIED INCLUDING THE WARRANTIES OF FITNESS FOR A PARTICULAR PURPOSE OR MERCHANTABILITY, WITH RESPECT TO THE ACCURACY, COMPLETENESS, OR USEFULNESS OF THE INFORMATION CONTAINED IN THIS REPORT, OR THAT THE USE OF ANY INFORMATION, APPARATUS, METHOD, OR PROCESS OWNED DISCLOSED IN THIS REPORT MAY NOT INFRINGE PRIVATELY RIGHTS;OR B. ASSUMES ANY LIABILITIES WITH RESPECT TO THE USE OF, OR FOR DAMAGES RESULTING FROM THE USE OF, ANY INFORMATION, APPARATUS, METHOD OR PROCESS DISCLOSED IN THIS REPORT, e

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RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION FOR l

VERIFICATION OF ANALYSIS METHODS FOR SMALL BREAK LOCA's  ;

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Prepared for the C-E OWNERS GROUP By

• TRANSIENT METHODS and LOCA NUCLEAR FUEL ENGINEERING

. NOVEMBER 1986 9

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UNITED STATES

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NUCLEAR REGULATORY COMMISSION 1 wenwoTow, p. c.roess February 11, 1987 4.....

Dr. J. K. Gasper, Chairman Combustion Engineering Owners Group 1623 Horney Street h ha, Nebraska 68102-2247

SUBJECT:

ACCEPTANCE FOR REFERENCING OF LICENSING TOPICAL REPORT The Nuclear Regulatory Commission (NRC) staff has completed its review of Topical Report CEN-203-P, Revision 1-P, Supplements 3 and 4. " Post-Test Analysis of Semiscale Test 5-UT-8," and " Response to NRC Request for Addi-tional Information for Verification of Analysis Methods for Small Break ' .

LOCA's." i We find the report to be acceptable for referencing in license applications to the extent specified and under the limitations delineated in the report and the associated NRC evaluation, which is enclosed. The evaluation defines the basis for acceptance of the report.

We do not intend to repeat our review of the matters described in the report and found acceptable when the report appears as a reference in license applications, except to assure that the material presented is applicable to the specific plant involved. Our acceptance applies only to the matters described in the report.

In accordance with procedures established in NUREG-0390, it is requested that the Combustion Engineering Owners Group (CEOG) publish an accepted version of

', this report within three months of receipt of this letter. The accepted version shall incorporate this letter and the enclosed evaluation after the title page. . The accepted version shall include an -A (designating accepted) following the report identification symbol. .

Should our criteria or regulations change such that our conclusions as to

. the acceptability of the report are invalidated, the CE06 and/or the ap-plicants referencing the topical report will be expected to revise and

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_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ - . ^

Dr. J. K. Gasper February 11, 1987 resubmit their respective documentation, or submit justification for the i continued effective applicability of the topical. report without revision

• of their respective documentation.

Sincerely, h, '

ennis M. Crutch ield, s stant Director Division of PWR Licens g-B Office of Nuclear Reactor Regulation

Enclosure:

Safety Evaluation

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SAFETY EVALUATION SUPPLEMENT RELATING TO VERIFICATION OF ANALYSIS METHOD 5 FOR 5 MALL BREAK LOCAs l - TMI ACTION ITEM II.K.3.30 FOR COMPUSTION ENGINEERING PLANTS POST TEST ANALYSIS OF SEMI 5CALE TEST 5-UT-08 CEN-203-P, PEVISION 1-P SUPPLEMENTS 3 AND 4 l.

. .0 INTRODUCTION l

In its safety evaluation (Reference 1), the NRC staff found TMI Action Item II.K.3.30 to be resolved for all licensed Combustion Engineering (CE) plants with the condition that it be shown'that the computer program CEFLASH-4AS could acceptably calculate the results of Semiscale Test S-UT-08, a small-break loss-of-coolant accident (SBLOCA). During this test, the water level in the simulated reactor vessel dropped rapidly ,

prior to loop seal clearing. The rapid drop in the reactor vessel we b,-

level was attributed to liquid holdup in the steam generator, i.e., t%

liquid briefly accumulated in the U-tubes of the intact loop steam generator.

To sa~tisfy the condition in the staff's safety evaluation, the Combustion

-, Engineering Owners Group (CEOG) performed an analysis of the S-UT-08 test.

Because of the small scale of the Semiscale facility, unique phenomena which occured during the test and conservatism that are an integral e partoftheCEFLASH-4ASlicensingmodel,theCEFLASH-4ASmodelcouldnot beuseddirectlytocalculateresultsthatwouldagreewEththeexperimen-tal data. A "best estimate" (BE). version of CEFLASH-4AS was devised for -

the analysis. Since the BE version would not be used for licensing, it also had to be shown that the licensing " evaluation model" (EM) version of CEFLASH-4AS contained modeling features which would predict the drop in I

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l water level in the reactor vessel prior to loop seal clearing in a SBLOCA {

in a full-scale plant (i.e., the " core level depression" or the " core )

uncovery spike"). -

1 The procedure that the CEOG followed for doing this was as follows: i

1. The EM version of CEFLASH-4AS was modified to a best .

estimate'(BE)programforcalculatingtheS-UT-08 test results, and the test results were calculated. The results of these calculations are given and discussed in Section 6 . . ..

of Reference 2. *

2. The BE models of components that determine the steam S

generator liquid holdup and the core level depression in S-UT-08 were replaced by their EM counterparts or ,

eliminated if they were not part of the EM version of CEFLASH-4AS. This BE/EM version, as it was called, was used to calculate the results of the S-UT-08 test. The results of these calculations are given and discussed in v c

• Section 7 of Reference 2.

3. The factors that caused the core uncovery spike in S-UT-08 4 et .

were examined and the relative magnitudes of the effects of each of the factors were determined semi-quantitatively.

The results of these efforts are discussed in pages 4 to 23 of R'eference 3.

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4. Those iters of the EM that were not used in the BE/EM analysis were reviewed to assure that they do not affect i

the ability of the EM version of CE FLASH-4AS to conser- '

vatively account for the core uncovery spike in a full-scale plart. The.results of this review are discussed in pages 24 to 30 of Reference 3. *

.0 EVALUATION i In general, the BE analysis results are in excellent agreement with the -"

experimental values. However, in the 50 to 180 second time interval, there are differences between the calculated and S-UT-8 liquid levels in  !

the steam generator and pump suction leg in the intact loop. After this time interval, there is excellent agreement in the liquid levels; so the minimum liquid level in the simulated reactor vessel is accurately cal-culated by the BE program. On an overall basis,'the staff finds that the  !

BE analysis acceptably predicted the S-UT-08 experimental values", especially the core level response.

The agreement between BE/EM results and the S-UT-08 data was also generally good. However, the BE/EM analysis did not conservatively calculete the minimum liquid-level in the simulated reactor vessel during the experiment.

The staff was concerned that this non-conservatism might be present in the

. EM program and reovested the CEOG to evaluate its ecause.

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The CEOG determined that the primary cause for the non-conservative core liquid-level prediction was "rewet steam," 1.e., steam that is produced when water in the hot legs flows back into the reactor vessel and drop? on the fuel rods. When the S-UT-08 model of this "rewet steam" ,

production from the BE. program was put into the BE/EM program, the minimum liquid level in the simulated reactor vessel was conservatively calculated.

4 The phenomena of "rewet steam" is a Semiscale specific phenomena. In -

Semiscale, the simulated reactor vessel is only 3 inches in diameter.

During the S-UT-08 test, the water flowing back from the hot legs spread fairly uniformly over all of the 25 simulated fuel rods and produced a . - --

.significant amount of steam. However, since the diameter of a reactor vessel in a plant is about 12 feet instead of 3 inches, the water flowing back from the hot legs in a plant would only contact. the fuel elements in the porticns of the core periphery directly beneath them. Thisissudha

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small percentage of the total-number of fuel elements that "rewet steam" production is not expected to be a significant phenomenon in a SBLOCA in a plant. Hence, the staff finds that it does not have to be modeled in the EM version of CERASH-4AS.

( Based upon the analysis discussed above, the staff finds that the BE/EM analysis acceptably predicts the S-07-08 experimental values. Thus, the I staff concludes that the EM version of CERASH-4AS contains modeling ,

yp features which would conservatively predict the " core uncovery spike" in a plant. "

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The CEOG identified those items of the EM which were not used in the ,

BE/EM analysis and classified them into three categories:

1. The conservatism required by Appendix K of 10 CFR 50.

2. The boundary conditions that were not compatible with those of the S-UT-08 test.

3. Thecomponentmodels(e.g.,reactorkineticsmodel) that were not applicable to S-UT-08. - -

While these items are significant portions of the EM, the CEOG concluded that these items are not directly related to the phenomena which leads to the core level depression. Thus, the CEOG concluded that BE/EM analysis is sufficient for demonstrating that the EM version of CE FLASH-4AS will conservatively calculate the core level depression. The staff has reviewed the CEOG assessment and concurs with their conclusions. ,

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.0 CONCLUSION The CEOG submitted CEN-203, Revision 1-P, Supplement 3 and Revision 1-P, Supplement 4'in response to NRC's concern that the CEFLASF-4AS computer e

program might not be able to calculate the initial rapid drop in water lev'el ex'p erienced in the simulated reactor vessel in the Semiscale Test.

S-UT-08. The staff has reviewed these submittals.' It finds that the p

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I l CEFLASH-4AS program can acceptably calculate this test. Therefore, as stated in NRC's safety evaluation (Reference 1) the requirements to per-fom plant specific analyses, per Action Item II.K.3.31, will no longer be required. The staff finds topical report CEN-203-P, Revision 1-P, .

Supplements 3 and 4 acceptable for referencing by licensees of CE plants ,

for resolving TMI Action Item II.K.3.30.

. 0 REFERENCES

1. NRC Safety Evaluation Report, "TPI Action Item II.K.3.30 for Combustion - -

Engineering Plants," dated May 23, 1985.

2. ' Combustion Engineering Report, " Post-Test Analysis of Semiscale Test 5-UT-08;" CEN-203, Revision 1-P, Supplement 3; dated December 1985.

3. Combustion Engineering Report, " Response to NRC Request for Additional Information for Verification of Analysis Methods for Small Break LOCA's,"

CEN-203, Revision 1-P, Supplement 4 dated November 1986.

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TABLE OF CONTENTS l

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- Section Title Page l 1.0 Introduction 1 i

i 2.0 Response to NRC Request for Additional 3 Information on Post-Test Analysis of Semiscale Test S-UT-8  ;

l Question 1 4 Question 2 16 Question 3 24 Question 4 31 3.0 References 32 e

Appendix f A Presentation Slides From the A-1 August 12, 1986 Meeting 1

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1.0 INTRODUCTION

In lulfillment of the one pending condition of NRC's SER issued June 20, 1945 on the Justification of C-E Small Break LOCA Methods (Post-TMI Action Item II.K.3.30), the C-E Owners Group (CE0G) submitted in December,1985 Supplement 3 to CEN-203-P, Revision 1-P to NRC (Reference 1). This supple-ment presented a post-test analysis of Semiscale Test S-UT-8 which showed that CEFLASH-4AS (Reference 2) includes component models and features necessary to predict or conservatively bound the core level depression prior to clearing of the reactor coolant pump suction leg loop seals as observed in the data from Semiscale Test S-UT-8.

After transmittal of Supplement 3, several conference calls and a meeting took place to assist the NRC staff in its review. In particular, the NRC staff raised concerns about the verification of the LOCA Evaluation Model (EM) version of CEFLASH-4AS via the Best Estimate / Evaluation Model (BE/EM) approach. At a meeting on August 12, 1986, C-E and the CE0G presented additional information to support the BE/EM analysis. Someoftgis information had been developed earlier but was not documented in Supple-ment 3,andsomeinformationwaspreviouslydocument$dinSupplement3but <

was recast for clarification for the meeting. In a letter dated August 27, 1986 (Reference 3), the NRC requested further information -

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< pertaining to the material presented at the meeting. This information consisted of three questions concerning the BE/EM analysis method used in

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the Test S-UT-8 analysis and a request for formal transmittal of the ,

I presentation slides from the August 12th meeting.

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1 This present Supplement 4 to CEN-203-P, Revision 1-P presents the responses to the three NRC questiens and provides the presentation slides from the August 12, 1986 meeting.

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2.0 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION ON POST-TEST ANALYSIS OF SEMISCALE TEST S-UT-8 This section provides the written responses to the questions pertaining to the verification of analysis methods used in small break LOCA's for C-E designed NSSS's as demonstrated by the post-test analysis of Semiscale Test S-Vi-8.

Question 1 requests an explanation of the basis for the qualitative assign-ment of the relative importance of the several factors affecting the core uncovery spike in Test S-UT-8.

Question 2 requests additional information regarding2the effect of core rewet steam production on the core uncovery spike in Test S-UT-8.

Question 3 requests a description of the portions of the EM that were not used in the BE/EM analysis of Test S-UT-8.

Question 4 requests formal transmittal of the presentation slides from the August 12, 1986 meeting. '

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Question 1:

I Please provide the basis for each of the " QUALITATIVE EFFECT IN TEST S-UT-8" l designations of " BIG", " MEDIUM", and "SMALL" in Enciosure A. Where applic-able give a specific reference (page no. and paragraph to a previous submit- ,

tal for each item).

Response to Question 1:

As described in Section 3.2 of Reference 1, the core level depression or

" core uncovery spike" in Test S-UT-8, prior to loop seal clearing, is caused by the development of a hydrostatic pressure imbalance between the vessel upper plenum and the downcomer acting to push the vessel fluid down through the core, into the downcomer and out the break. This differential pressure exists until one or more of the pump suction leg liquid seals clears. These seals block the steam generated in the vessel from reaching the break.

Figure 1.2 shows that up to 160 seconds this differential pressure in Test S-UT-8 is essentially equal to the steam generator elevation head pressure l e difference from inlet to outlet; i.e., the difference between the collapsed liquid levels in the upflow side and"downflow side of the U-tubes. In Test S-UT-8, a higher collapsed liquid level in the steam generator upflow side ,

than the downflow side results from several factors: 1)liquidwhich C drains from the vessel upper head to the vessel upper plenum is carried over into the U-tubes, 2) rapid downflow side draining contributes to the a

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mass flowing out of the break, 3) countercurrent flowing steam and high frictional losses in the " pant leg" slows the upflow side draining, and

4) the rate of resupplying liquid by condensing steam is higher on the upflow side than the downflow side.

Steam generated in the vessel is the result of liquid vaporization from heat added by the heater rods and vessel walls. In particular for Test S-UT-8, liquid fallback from the hot legs rewets the heater rods producing steam even though the vessel collapsed level is near or below '

the heated length. A small fraction of the vessel steam flow passes to the downcomer through the upper head bypass line which tends to reduce the pressure differential acting to depress the vesselelevel; however, this t

path for steam is much more restrictive to flow than the vessel lower plenum to downcomer crossover leg where bulk steam is obgpeved to enter the downcomer at times of minimum vessel liquid level.

For the August 12, 1986 meeting, these factors affecting the core uncovery spike in Test S-UT-8 were summarized in an overhead slide, see Figure 1-1.

Furthermore, each factor was assigned a qualitative designation of " Big", >

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" Medium", or "Small", depending on its relative effect in Test S-UT-8.

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As part of the technical presentation of the August 12, 1986 meeti ng ,

. Figure 1-1 also indicated how the factors affecting the core uncovery

. spike were modeled in the BE/EM analysis of Test S-UT-8. In all cases except the " pant leg" friction model, which is a specific model for  %

Semiscale's unique design, the BE/EM modeling of the factors followed the EM prescription.

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The qualitative assessment in figure 1-1 shows that the BE/EM analysis modeled all the major factors affecting the core uncovery spike, thus 1

indicating the reasons why the BE/EM analysis resulted in an acceptable prediction of the Test S-UT-8 core uncovery spike prior to loop seal ,

clearing.

The quantitative basis for each factor's qualitative designation in Figure 1-1 is given in Table 1 and in the accompanying figures. The majority of the assessment is based on examination of Test S-UT-8 data and where necessary, the results of the Best Estimate (BE) analysis of Test S-UT-8 (see Section 6.0 of Reference 1) were used to provide more I

detailed information.

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Table 1 Quantification of Factors Affecting Core Uncovery Spike in Test S-VT-8 Prior to Loop Seal Clearing at 240 Seconds e For a spike to occur, the upper plenum pressure must be greater than the

, downcomer pressure during pump suction liquid seal existence:

AP = P(vessel upper plenum) - P(downcomer) > 0 V-D also AP SG = P(SG inlet plenum) - P(outlet plenum) >0 e The following factors cause the upper plenum Qualitative pressure to be higher than the downcomer Effect in pressure (except #5): Test S-UT-8

1. Higher inventor Big downflow side (y level in SG upflow side than liquidholdup)

AP SG increases from 1 psid to 3.5 psid-see Figure 1-2 AP increases from 0 psid to 5 psid-see Figure 1-2 APSG-D(Max) is 70% of APV-D(Max)

Therefore " Big" effect 1.A. Upper head draining into upper plenum supplies e Big inventory which is carried into SG upflow side Initial upper head liquid mass of 18 lbs reduces ,

to 9.5 lbs primarily due to flashing About threc-fourths of the upper head liquid mass ,

(9.5 lbs @ 50 sec) is carried into the intact SG upflow side along with liquid mass from the upper plenum and hot leg SG upflow side accumulates 9 lbs of liquid mass from '

upper head, upper plenum, hot leg, and steam condensation-see Figure 1-3 When upper head is empty, SG begins draining at 100 sec SG draining time is 200 seconds

. Delay in SG draining time is 110 seconds (50 sec delay due to liquid accumulation, and 60 sec delay due to draining of accumulated liquid)

Time delay is more than half of draining time Therefore " Big" effect m'

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6 Table 1-' - continued 1.B. 'SG downflow side looses inventory faster. than upflow side Medium Upflow side draining rate is .2 lbs/sec - see Figure 1-4

. Downflow side ' draining rate is .3 lbs/sec - see Figure 1-4 Resistance to downflow side draining, which supplles break .

flow, is less than resistance to upflow side draining with counter-flowing steam Produces 2 psid increase in AP until downflow side empties - see Figure 1-2 SG -

Therefore " Medium" effect-1.C. CCFL in SG slows up draining of upflow side Small Delay in SG draining time due to CCFL is 50 sec-see Figure 1-3 Delay is one-fourth SG draining time Produces .5 psid increase in AP3g - see Figure 1-2 Therefore "Small" effect 1.D More condensate is generated in upflow side'of SG Small i The change. in mass of condensate during the time period from 50 to 240 sec on the upflow side is

• 8 only 1.'25 times more than the downflow side-see Figure 1-5 (BE analysis result) m The change in mass of condensate between 50 and 240 sec is an order of magnitude less than the liquid mass in SG'U-tubes e Therefore "Small" effect e .

2. Frictional losses in " pant leg"-during counter- Medium

.e current flow add to the loop pressure losses ,

Delay in SG. draining time due to " pant leg" friction is 40 sec-see Figure 1-6 (BE analysis result) .

Upflow side draining rate with " pant "leg" friction is .2 lbs/sec o Upflow side draining rate without friction is .3 lbs/sec '

Produces 2 psid increase in APV-D - see Figure 1-2 Therefore " Medium" effect u

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3. Wall heat in core region Medium i

Core wall heat average is 20 Btu /sec (BE analysis result)

Total cold side wall heat average is 13 Btu /sec

. Core power average is 60 Btu /sec Difference in wall heat (core-cold side) is 12%

of core power

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4. Core rewet steam production Medium Total heat added to rewet fluid (Btu) is 3 times more than core-to-coolant heat during rewet period (BE analysis result)

After hot leg draining, a6P V-D is still 2.5 psid due to rewet steam production -

see Figure 1-2 (S-UT-8 data)

Therefore " Medium" effect

5. Bypass line provides path for steam from upper head Small-to downcomer Initial core bypass flow rate of 1.1% of core flow is small During the transient, the bypass flow is never more than 3% of the average break flow (BE analysis results)

Acts to reduce ALP V-D by allowing steam to flow to break but produces nearly the same result e as no bypass because flow resistance is very high Therefore "Small" effect

6. Equili$'rium fluid state in the cold legs 'and downcomer Small during HPSI flow

• Mass addition from HPSI flow is less than 7% of total mass out the break Therefore "Small" effect 4

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Figure 1-1 (from IrGsentation at August 12, 1986 Meeting)

FACTORS AFFECTING CORE UNC0VERY SPIKE IN TEST S-UT-8 e FOR SPIKE TO OCCUR, THE UPPER PLENUM PRESSURE MUST BE GREATER THAN THE DOWNCOMER PRESSURE DURING PUMP SUCTION LIQUID SEAL EXISTENCE. -

e THE FOLLOWING FACTORS CAUSE QUALITATIVE EFFECT .

THE UPPER PLENUM PRESSURE TO IN TEST S-UT-8 BE HIGHER (EXCEPT #5): MODELED IN BE/EM l'. HIGHER INVENTORY LEVEL IN SG UPFLOW SIDE BIG YES THAN 00WNFLOW SIDE (LIQUID HOLDUP) BECAUSE A. UPPER HEAD DRAINING INTO UPPER PLENUM BIG YES(EM)

SUPPLIES INVENTORY WHICH IS CARRIED INTO SG UPFLOW SIDE B. SG DOWNFLOW SIDE LOOSES INVENTORY MEDIUM YES(EM)

FASTER THAN UPFLOW SIDE C. CCFL IN SG SLOWS UP DRAINING OF SMALL N0(EM)

UPFLOW SIDE D. MORE CONDENSATE IS GENERATED IN UPFLOW SMALL N0(EM)

SIDE OF SG

2. FRICTIONAL LOSSES IN " PANT LEG" DURING MEDIUM YES(BE)

COUNTERCURRENT FLOW ADDS TO THE LOOP PRESSURE LOSSES 5

3. WALL HEAT IN CORE REGION ,

MEDIUM YES(EM)

4. CORE REWET STEAM PRODUCTION MEDIUM N0(EM) l .

5. BYPASS LINE PROVIDES PATH FOR STEAM FROM SMALL N0(EM) l j UPPER HEAD TO DOWNCOMER v; .

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6. EQUILIBRIUM FLUID STATE IN THE COLD LEGS SMALL YES(EM) l AND DOWNCOMER DURING HPSI FLOW I

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Figure 1-2 Comparison of Vessel-to-Downcomer Diff erential Pressure with Intact Loop Steam Generator U-Tube Diff erential Pressure

. as Measured During Test S-0T-8 4

APv-d = P (vessel upper plenum)-

P(top of downcomer)

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P(SG outlet plenum) 5 -

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'CCFL Period [ ' SG Differmtial ' ~ ' Pant Ieg' ' Bulk Icap SG Liquid Draining: Upflow Friction ard Steam to Seal W-1

• imt vs. Downflow Cbre Rewet Downcamer Clearing and (bre e i  !  : , Rewet i g 40 e 80 120 160 200 240 TIM E (SIC) s.

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Figure 1-3 Costparison of Measured Liquid Masses in-Test S-UT-8 for the Intact Loop Steam Generator-Upflow Side U-Tubes and the Vessel Upper Head .

IL SG Upflow Side U-Tubes

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~60 second delay in SG draining e

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Figure 1-4 l , Comparison of Measured Liquid Masses in Test S-DT-8 for the Intact Loop Steam Generator Upflow and Downflow Side U-Tubes 30 _

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Figure 1-5 Comparison of. Calculated Integral Condensate in the B-E Analysis of Test S-DT-8 for the .

- Intact Loop Steam Generator Upflow. and Downflow Side U-Tubes l CCFL Period for Upflow Side J

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Figure 1-C (Figure 5-5 from Reference 1)

EFFECT OF MODIFIED FRICTIONAL LOSSES

. IN THE SEMISCALE HOT LEG ON STEAM GENERATOR ORAINING END OF CCFL 24- WITH BE MODEL MOD.

- - WITHOUT MODEL MOD.

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Question 2: l In the August 12th meeting, the CE0G stated that the " foot" of the core

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level depression between approximately 200 and 240 seconds (see S-VT-8 data curve in Enclosure B) is primarily due to the core rewet steam produc- ,

tion in S-VT-8 and that an earlier calculation demonstrated this effect.

Please discuss the earlier calculation or provide some other basis for this conclusion.

Response to Question 2:

During the core uncovery period prior to loop seal clearing in Test S-VT-8, the collapsed level in the vessel falls below the bottom of the heated length at about 200 seconds and remains there till about 240 seconds forming the " foot" in the data curve. During this 30 to 40 second period, bulk steam flowed from the vessel into the downcomer causing the observed displacement of liquid from the downcomer. (See Figure 2-1.) This raises the question as to the origin of the steam flow at a time when there is essentially no water inventory in the vessel for steam production. j Examination of the cladding thermocouple data for the upper approximately four feet of the heater rods shows no cladding heatup while lower eleva-tions do show heatup. (See Figure 2-2.) Rewet of the heater rods by ,'

liquid flowing back into the vessel from the hot leg is believed to be responsible for this observation. Rewet of the heater rods impacts the transient core uncovery response of Semiscale through the additional steam 16

a e

l I

produced by vaporization of the liquid fallback. Vaporization of liquid '

fallback prolongs core uncovery relative to a system response without fallback vaporization that otherwise would simply stop making steam due .

to uncovery of the core and accumulate the liquid fallback from the. hot leg in the lower core region starting core recovery.

Based on this experimental evidence, a core rewet steam production model was used in.the Best Estimate (BE) analysis of Test S-UT-8, see section 6.0 of Reference 1. Use of the model resulted in the accurate prediction of the depth and duration of core uncovery prior to loop seal clearing. In particular, the " foot" of the core level depression between 200 and 240 seconds was accurately predicted. (See Figure 2-3.)

The rewet model was not used in the BE/EM analysis of Reference 1, where selected Evaluation Model (EM) component models related to steam generator behavior were used in place of their BE model counterparts, because core rewet steam production is not part of the C-E Small Break LOCA EM.

However, an earlier BE/EM calculation (August,1985) which included the core rewet steam production model is available for comparison to the BE/EM Reference Case. The vessel collapsed liquid level results for this earlier BE/EM calculation are shown in Figure 2-4, along with the Test S-UT-8 data and the BE/EM Reference case predictions. These results show that the core

. rewet steam production model is effective in prolonging core uncovery prior to loop seal clearing in a manner similar to the Test S-UT-8 observation.

17

l a

Prior to 150 seconds the results of both BE/EM cases are the same since

~

core rewet steam production is only a small part of the total steam produc- l tion. Beginning at about 170 seconds, the BE/EM Reference case without the core rewet model predicts a gradual core level recovery due to the accumulation of liquid fallback from the hot legs and the reduction in ,

differential pressure between the upper plenum and the downcomer caused by the initiation of loop seal clearing. Also, the calculated steam flow produces more gradual loop seal clearing than observed in the test data which leads to a slower core recovery rather than the abrupt recovery seen in the test. By comparison, the additional steam flow produced by the core rewet model used in the earlier BE/EM calculation produces enough steam to predict the observed bulk flow of steam from the vessel into the downcomer and then leads to abrupt loop seal clearing and rapid core recovery begin-ning at 220 seconds as observed in the test data.

The core recovery level at 270 seconds is. conservatively lower in the BE/EM Reference case due to the prediction of lower . liquid leve,ls g in the downcomer.

This difference is largely due to differences between the BE/EM Reference case and the earlier BE/EM calculation related to wall heat and bubble rise velocity in the downcomer which influence the core recovery period but have little effect on the " foot" of the core level depression.

As to the absence of a rewet model in the EM, the core rewet steam produc- ,

tion model was implemented into CEFLASH-4AS to reproduce a specific Test S-UT-8 observation. In Test S-UT-8 liquid draining from the hot legs appears to penetrate uniformly the 3 inch Semiscale vessel and rewet the 18

upper regions of all 25 heater rods. Whereas in an NSSS, liquid draining from the hot legs is not expected to penetrate the approximately 12 ft diameter reactor vessel. Rewet of the NSSS fuel rods would be localized in peripheral regions of the core below the hot leg nozzles. After some initial steam production, the regions of the NSSS core cooled by rewet (and also the annular gap between the core shroud and core support barrel inner surface) would channel the returning liquid to the lower plenum, adding little to the overall steam production. For these reasons the core rewet steam production mechanism is not considered important for NSSS analyses and is not modeled in the EM and was, therefore, not included in the BE/EM calculation of Test S-UT-8 documented in Reference 1 .

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' Question 3:

To demonstrate the adequacy of the EM model, the CE0G used a BE/EM model-

~

for S-UT-8. Please show why all of the portions of the EM model that were l not in the BE/EM model have~little or no effect on the core uncovery spike.- ,

Response to 0uestion 3:

As described 'in Section 4.1 of Reference 1, the objective of the post-test L-analysis of Semiscale Test S-UT-8 was to confirm that C-E's Small Break LOCA Evaluation Model (EM), specifically the themal-hydraulic Small Break LOCA computer. code CEFLASH-4AS, can acceptably calculate pre-loop seal clearing core uncovery as observed in the data from Semiscale Test S-UT-8. 'The calculational framework, commonly referred to as EM, contains component models and inputs (e.g. the Appendix K requirements) as well as calcula-tional procedures which contain many conservatism in order to yield conser-vatively high cladding temperatures. Also, the EM is designed for analyz- ..

ing large NSSS's and not a small scale test facility such as Semiscale (about 1/1700 scale of an NSSS). The combined effect of the conservative features of the EM can be justified for licensing calculations of an NSSS, but make it difficult, if not impossible, to verify the basic quality of a

specific EM component models against tests like Semiscale Test S-UT-8 unless test specific models are added to the EM. ,

24

_ ____ - __ a

In order to resolve this conflict, the approach was taken to use the Best Estimate .(BE) version of CEFLASH-4AS to a large extent in order to remove many of the EM conservatism. In order to obtain the required verification of.the EM, BE component models or calculational procedures (e.g. noding) were replaced by their EM counterparts and the resulting analysis is referred.to as the BE/EM anlaysis. Thus, all of those portions'of the EM related to steam generator modeling were included in the BE/EM analysis, particularly those portions affecting the core uncovery spike prior to clearing of the loop seals. Specific BE component models which are not

'used at all in the EM (like countercurrent flow limit in steam generator U-tubes.or vessel bypass) were left out in the BE/EM analysis.

This approach of using the BE/EM analysis for verifying the-EM was present-  !

ed to the NRC staff at a meeting on 12-10-84 and documented by a letter of 3-11-85, from R. W. Wells (CEOG) to C. O. Thomas (NRC), (Reference 4).

In this context, it is pointed out that the BE version of CEFLASH-4AS has ,

grown out of the EM version. Many of the BE component models were derived from EM component models with phenomenological mechanisms added to the physical description for improved realism or improved representation of boundary conditions and geometry.

In order to specifically answer Question 3, those portions of the EM that were not used in the BE/EM analysis of Test S-UT-B have been reviewed and the models which were left out have been categorized into four groups.

These groups are described below along with representative examples of models not used in the BE/EM analysis.

25

1 ll

i. Conservative Models Reauired by Appendix K

]

l Several component models required by Appendix K were not used in the BE/EM l analysis of Test S-UT-8 because their effect is known to be conservative.

• ii 4

For example, use of Moody break flow, the required multipliers on core heat, ,

and no return to nucleate boiling would increase the prediction of the depth and duration of core uncovery in a conservative but unrealistic manner and this would overshadow the effects of steam generator liquid holdup modeling on core uncovery. Also, there are several models related to U0 2 fuel required by Appendix K, such as cladding rupture, which simply do not apply ,

i to the Semiscale electrical heaters.

ii. Models Not Compatible with Test S-UT-8 Boundary Conditions As described in Section 4;4 of Reference 1, the measured initial conditions and transient boundary conditions of Test S-UT-8 were simulated in the j post-test analysis as accurately as possible. Therefore, all EM models, methods, and procedures related to conservative specification of loss coefficients, initial flow rates, initial fluid temperatures, and initial balancing of steam generator primary to secondary heat transfer were not used. For the transient boundary conditions, all EM models, methods and procedures related to core power, HPSI ficw, the low pressurizer pressure trip signal, and coolant pump coastdown speed were not used. The EM models ,

l for the secondary side of the steam generators were used in the BE/EM analysis; however, modeling of steam generator isolation followed measured feedwater and steam conditions and subsequent steam line leakage was 26

e allowed to' approximate the measured secondary' side pressure decay.. The EM procedures for modeling break geometry and break discharge flow were not used in favor of a realistic representation of the Test S-UT-8 brea'K configuration.

iii. NSSS Component Models Not Applicable to Semiscale Several EM component models developed for analysis of C-E designed NSSS's were not used in the BE/EM analysis of Test S-UT-8.due to unique design feature differences between Semiscale and a C-E NSSS pertaining to the vessel, downcomer, hot legs, and suction legs. EM component models related to the downcomer and vessel two-phase level were modified for Semiscale, as described in Section 5.3 of Reference l'. This was necessary because the vessel and downcomer are separate components in Semiscale where vessel wall heat is added directly to the core region and where the downcomer and lower plenum geometric cross-sections and inter-connections are uniquely different from the NSSS. The EM methods for defining flow path connections between nodes, and flow path geometric cross-sections were not used for all flow paths in the Test S-UT-8 analysis, due to the unique design of Semiscale, particularly for the hot leg with its unique vertical " pant leg" design, see Figure 3-1. The EM component model for loop seal clearing, which was developed for the C-E designed loop seal, was not used for the Test S-UT-8 analyses because the Semiscale suction leg piping is significantly different from the NSSS design, see discussion on page 3.1-7 of Reference 5. A mechanistic BE loop seal clearing model was used in the BE/EM analysis. Finally, the EM component model noding for NSSS safety injection was not used in the BE/EM analysis.

27

iv. Models Related to Numerical Methods The EM version of CEFLASH-4AS was not used for the analysis of Test S-UT-8 because it~ includes various numerical component models which are not applicable to Semiscale. The BE version of CEFLASH-4AS was used instead, ,

which includes the: implementation of improved integration techniques which

. reduce code running time without affecting the results and produce stable solutions for the relatively small fluid volumes in Semiscale. Even with the use of the BE version, additional models were needed to stabilize the numerical solution for Test S-VT-8, see Section 5.4 of Reference 1.

l l

In spite of the large number of EM component models not used in the BE/EM analysis, a side-by-side comparison of the BE/EM predictions for Semiscale 2

Test S-UT-8 and the EM predictions of System 80 0.35 ft cold leg break shows that these two analyses demonstrate analogous core uncovery phenomena, see Figure 3-2. This is not surprising since the EM component models for steam generator behavior do predict a hydrostatic pressure difference between core and downcomer and, thus, produce the effects of liquid holdup as observed in Semiscale Test S-UT-8.

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l Question 4:

Please provide the slides used in the August 12, 1986 presentation.

Response to Question 4:

The presentation slides from the August 12, 1986 meeting are given in Appendix A.

1 31

i i

3.0 REFERENCES

i

1. CEN-203-P, Revision 1-P, Supplement 3 " Post-Test Analysis of Semiscale Test S-UT-8, Response to NRC's Conditional SER Issued .

June 20, 1985 on the Justification of C-E.Small Break LOCA Methods", ,

~

December, 1985.

2. CENPD-133, Supplement 1, "CEFLASH-4AS, A Computer Program for the Reactor Blowdown Analysis of the Small Break Loss-of-Coolant Accident",

Janua ry, 1974. 1 CENPD-133, Supplement 3, "CEFLASH-4AS, A Computer Program for the Reactor Blowdown Analysis of the Small Break Loss-of-Coolant Accident",

January, 1977.

o

3. Letter, D. M. Crutchfield (NRC) to J. Holman (C-E Owners Group),  !

1

" Verification of Analysis for Small Break LOCA's; TMI Action Item II.K.3.30 for Combustion Engineering Plants; Post-Test Analysis of Semiscale Test S-UT-8", August 27, 1986.

l 1

4. Letter, R. W. Wells (C-E Owners Group) to C. O. Thomas (NRC), "Small BreakLOCAMethodsVerification(ActionPlanItemII.K.3.30), .

Commitment to Perform Post-Test Analysis of Semiscale Test S-UT-8",

March 11, 1985. l l

S. CEN-203-P, Revision 1-P, Supplement 2-P, "Further Response to NRC Request Number 1 for Additional Information on C-E Report CEN-203-P, Rev. 1-P", November, 1984.

32

________ - a

i

)

i i

Appendix A Presentation Slides From the August 12, 1986 Meeting 6

O i

I

. A-1

i Description of C-E Owners Group Response to NRC's Conditional SER Issued June 20,1985 '

On the Justification of C-E Small Break LOCA Methods Post-Test Anaysis of Semisca'e Test S-UT-8 C-E Owners Group /NRC/C-E Meeting August 12, 1986 f.- 2

OBJECTIVE OF MEETING 0 DISCUSS WITH NRC ADEQUACY OF POST-TEST ANALYSIS OF SEMISCALE TEST S-UT-8 (CEN-203, SUPPLEMENT 3) 0 DISCUSS WITH NRC CLOSE-0VT OF POST-TMI

^

ACTION ITEM II.K.3.30 AGENDA (1) BACKGROUND AND CHRONOLOGY OF CEOG RESPONSE TO II.K.3.30 AND S-UT-8 POST-TEST ANALYSIS ISSUE (2)

SUMMARY

OF S-UT-8 POST-TEST ANALYSIS RESULTS (3) ADDRESS NRC CONCERN ABOUT ADEQUACY I

OF S-UT-8 POST-TEST ANALYSIS 0 PHYSICAL PHENOMENA AFFECTING CORE UNC0VERY SPIKE IN SEMI-SCALE TEST S-UT-8 0 BE/EM ANALYSIS (BE WITH EM COMPONENT MODELS) 0F S-UT-8 0 IMPLICATIONS FOR NSSS LICENSING

^

ANALYSES (4) DISCUSSION OF CLOSE-0UT OF II K.3.30 9-3 l

l

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! l, 9

4 CEOG/C-E APPROACH FOR RESPONDING TO II.K,3,30 JUSTIFY ADEQUACY OF CURRENTLY APPROVED C-E SMALL BREAK LOCA EVALUATION MODEL t

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CHfiON0LOGY

SUMMARY

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NRC/CE0G ACTIONS ON II.K.3.30 01/26/81 CE0G/NRC meeting where seven questions on C-E's small break LOCA evaluation model were identified by.the NRC which would close out II.K.3.30. if answered. satisfactorily.

07/14/81 CEOG letter issued to NRC to document the seven questions and to provide a commitment to answer them.

03/82 Report CEN-203-P submitted by CE0G to the NRC providing responses to the seven questions.

08/83 NRC letter to CE0G asking eight additional questions: five on specific subjects in CEN-203-P, three new questions en steam generator modeling.

02/08/84 CEOG/NRC/ANL meeting on three steam generator modeling questions; agreement reached- that questions could be answered without analys'is of Semiscale Test S-UT-8.

02/28/84' Supplement 1-P to CEN-203-P submitted by CEOG to the NRC providing responses to the five of eight additional questions on specific subjects in CEN-203-P.

, 07/84 CE0G letter to NRC connitting to answer the three steam generator modeling questions in accordance with agreements reached at February 8, 1984 meeting.

09/84 CE0G/NRCconferencecallwher,esRCinputwasprovidedthatS-UT-8 analysis is required to successfully answer remaining three '

questions and close out II.K.3.30.

A-5

Page 2 Chronology 10/04/84 CE0G/NRC meeting where original CEOG approach for responding to three questions was presented. NRC did not object to submittal of ,

information presented, but stated that lack of model verification against Semiscale Test S-UT-8 prevents close out of II.K.3.30. ,

i 11/30/84 CE0G submits to NRC Supplement 2 to CEN-203-P which provides responses to the three remaining steam generator modeling questions based on the original CE0G approach which resulted from the February 8, 1984 meeting.

12/10/84 Meeting with NRC, presentation of S-UT-8 Post-Test analysis approach (BE and.BE/EM Analysis).

3/11/85 Letter by CEOG to NRC stating commitment to 5-UT-8 analysis by December, 1985.

Documentation of analysis approach which was presented at 12-10 84 meeting. .

6/20/85 Conditional SER issued by NRC:

CEN-203 and its supplements acceptable response to II.K.3.30.

Condition: " Confirmation that the CE FLASH-4AS computer program can acceptably calculate core level depression, prior to clearing of the reactor coolant pump loop seals, as observed in the data from Semiscale Test S-UT-8."

With acceptable confirmation: ,

II.K.3.30 Will be resolved II.K.3.31 (Plant specific analyses) No longer required A-6

I i l Page 3 Chronology l

l 10/31/85 Meeting with NRC, presentation of S-UT-8 analysis results. New NRC

, reviewer voiced some concerns with BE/EM analysis.

12/18/85 Submittal of S-VT-8 analysis (CEN-203, Supplement 3) by CE0G.

1 5/8/86 Conference call NRC/CE0G/CE to discuss NRC concerns about BE/EM {

analysis which are holding up approval of S-UT-8 analysis. I NRC concerns revolve around the adequacy of modeling aspects (CCFL and bypass) affecting the S-UT-8 core uncovery spike.

In subsequent telephone calls between NRC, CE0G and CE agreement was f reached for further discussions (present meeting).

1 l

9 A-7

SUMMARY

OF S-UT-8 POST-TEST ANALYSIS RESULTS l 8

ANALYSIS OBJECTIVES 0 THROUGH POST-TEST ANALYSIS DEMONSTRATE ABILITY OF SBLOCA EM (EVALUATION MODEL VERSION OF CEFLASH-4AS)

TO PREDICT S-UT-8 CORE UNC0VERY SPIKE FULFILL CONDITION OF SER 0 NOT A GLOBAL VALIDATION OF THE EM 0 RETAIN APPROVAL OF CURRENTLY APPROVED EM 4

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ANALYSIS APPROACH I

O PERFORM BE (BEST ESTIMATE VERSION OF CEFALSH-4AS) ANALYSIS HISTORICAL C-E APPROACH FOR TEST ANALYSES, -

BENCHMARK BE (WHICH WAS DEVELOPED FOR NSSS)

AGAINST THE 1/1700-SCALE SEMISCALE FACILITY .

4 0 PERFORM VALIDATION OF-EM VIA BE/EM ANALYSIS EM WAS DEVELOPED FOR CONSERVATIVE NSSS LICENSING CALCULATIONS. IS ILL-SUITED FOR POST-TEST ANALYSIS OF SEMISCALE i

USE BE CODE, BUT-REPLACE BE COMPONENT MODELS IMPORTANT FOR STEAM GENERATOR LIQUID HOLDUP BY EM COMPONENT MODELS - BE/EM ANALYSIS 1

0 THIS APPROACH PRESENTED TO NRC AT 12-10-84 MEETING AND DOCUMENTED IN CEOG LETTER TO NRC 0F 3-11-85 NRC CONCURRENCE WITH THIS APPROACH AT 12-10-84 MEETING AND IN SER.

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_ 0 0 C 0 0 0 0 0 0 0 0 0 0 0 0 0 0 5 0 5 0 3 2 2 1 1 b u

_ s _ iJ

_ ruul a

>u_ a ._ o u

!l

CONCLUSIONS o BE ANALYSIS SHOWS VERY GOOD AGREEMENT WITH COLLAPSED WATER LEVEL DATA IN VPSSEL o BE ANALYSIS SHOWS THAT BE SBLOCA MODEL INCORPORATES COMPONENT MODELS WHICH PERMIT ACCURATE PREDICTION OF .

S-UT-8 TYPE CORE UNC0VERY SPlKES o BE/EM ANALYSIS SHOWS ACCEPTABLE AGREEMENT WITH COLLAPSED WATER LEVEL DATA IN VESSEL:

, ADEQUATE TREND OF UNC0VERY SPIKE AND SUBSEQUENT B0ll-OFF

, ADEQUATE DEPTH OF UNC0VERY SPIKE

. CONSERVATIVELY LOW INVENTORY AT CORE RECOVERY AND DURING B0ll-OFF

. CONSERVATIVELY HIGH CLADDING TEMPERATURES DURING UNC0VERY SPIKE AND B0ll-OFF o BE/EM ANALYSIS SHOWS THAT SBLOCA EM INCORPORATES CGMPONENT MODELS WHICH PERMIT ACCEPTABLE PREDICTION OF S-UT-8 TYPE CORE UNC0VERY SPIKES o RESULTS OF BE AND BE/EM ANALYSES FULFILL PENDING CONDITION OF SER OF CONFIRMING THAT CE SBLOCA COMPUTER CODE ACCEPTABLY CALCULATES S-UT-8 TYPE CORE UNC0VERY, THUS, FINAL CLOSE-0VT OF ACTION ITEM II.K,3,30 EXPECTED

?,-12

, PRESENTATION OF POST-TEST ANALYSIS OF SEMISCALE TEST S-UT-8 MAJOR TOPICS e REVIEW OF TEST S-UT-8 e COMPARISON OF BE/EM ANALYSIS PREDICTIONS TO TEST DATA e IMPLICATIONS FOR NSSS LICENSING ANALYSIS A-10

REVIEW OF SEMISCALE TEST S-UT-8

.i s MOD-2A FACILITY I

e FACTORS AFFECTING CORE UNC0VERY SPIKE e TWO ADDITIONAL EXPERIMENTAL OBSERVATIONS CORE STEAM PRODUCTION (REWET)

CORE STEAM DISTRIBUTION e CONCLUSIONS FROM THE REVIEW 0F TEST S-UT-8 e

l j

A-14

~ '

t I

g  :

= .

Type !! steam genenseur e-  :

Type il steam generator (twoken loopp .

• (Intact loom Special l'estures:

Instrumented SG J

• Moneycorno insulation Moe heat trarJng New core Pressurtser veeses l

I h Bypass [

. Nmp Hot leg kh h

, Hot leg i g., , .

~~$ .

u .,s.

fp Pump suction e.m.,

Nmp e, l @

l[, g: Blowdown valve Condensing coils Nmp W # , llp y.,,

y Condensate measuring tanks Vessel downcomer dI L y mantr m V

i Semiscale Mod-2A Facility

~

A-15

FACTORS'AFFECTING CORE UNC0VERY SPIKE IN TEST S-UT-8 e FOR SPIKE TO OCCUR, THE UPPER PLENUM PRESSURE MUST BE GREATER THAN THE DOWNCOMER PRESSURE.DURING PUMP' SUCTION LIQUID SEAL' EXISTENCE. .

e THE FOLLOWING FACTORS CAUSE -QUALITATIVE EFFECT ,

THE UPPER PLENUM PRESSURE TO IN TEST S-UT-8 BE HIGHER-(EXCEPT #5):

1. HIGHER INVENTORY LEVEL:IN SG UPFLOW SIDE BIG THAN DOWNFLOW. SIDE (LIQUID HOLDUP) BECAUSE.

A. UPPER HEAD DRAINING INTO UPPER PLENUM BIG SUPPLIES INVENTORY.WHICH IS CARRIED INTO SG UPFLOW SIDE B. SG DOWNFLOW SIDE LOOSES INVENTORY MEDIUM FASTER THAN UPFLOW SIDE C. CCFL IN SG SLOWS UP DRAINING OF SMALL UPFLOW SIDE D. MORE CONDENSATE IS GENERATED IN UPFLOW SMALL i

SIDE OF SG

2. FRICTIONAL LOSSES IN " PANT LEG" DURING MEDIUM COUNTERCURRENT FLOW ADDS TO THE LOOP PRESSURE LOSSES

3. WALL HEAT IN CORE REGION MEDIUM

4. CORE REWET STEAM PRODUCTION MEDIUM

~

5. BYPASS LINE PROVIDES PATH FOR STEAM FROM SMALL UPPER HEAD TO DOWNCOMER

6. EQUILIBRIUM FLUID STATE IN THE COLD LEGS SMALL AND DOWNCOMER DURING HPSI FLOW A-16

EXPERIMENTAL FINDING RELATED TO CORE STEAM PRODUCTION e DURING CORE UNC0VERY SPIKE PRIOR TO LOOP SEAL CLEARING, UPPER APPR0XIMATELY'4 FT OF HEATER RODS SHOW NO-HEATUP, REWET BY LIQUID DRAINING FROM HOT LEGS SUSPECTED,

'S-UT-8 MEASUREMENTS OF PEAK HEATER R00 TEMPERATURES eDATA POINT ENVELOPE OF DATA TOP OF HEATED 12 - PRIOR TO LOOP SEAL LENGTH 0 CLEARING (240 SEC)

REGION OF REWET E

~

E

= > ,

g 4- ,

I 1

.. BOTTOM OF 0 HEATED LENGTH 500 700 900 TEMPERATURE (OF) 1 A-17 I

EXPERIMENTAL FINDING RELATED TO CORE STEAM DISTRIBUTION

^

~

e THE CORE UNC0VERY SPIKE WAS DEEP ENOUGH THAT BULK STEAM FROM THE VESSEL ENTERS THE DOWNCOMER AND RISES TO THE TOP IN A SLUG-LIKE MANNER. LIQUID DISPLACED FROM THE DOWNCOMER PRODUCES A COLLAPSED LEVEL SPIKE-LIKE DROP IN THE DOWNCOMER, FLOW REVERSAL IN THE INTACT COLD LEG, AND MOMENTARY REFILLING OF THE LOOP SEAL.

S-UT-8 LOWER PLENUM DENSITY S-UT-8 DOWiiC0i'ER LIQUID LEVEL

. 201RV'A8-6 50 - 29 1

"O " 20' l ll d --_

in 30 - 15<

! 1 '

0 20 -

5 10 ~f

t l - - 30(stc) h . _ 3nc3ge) ,

10 - 5<

0

' , , , , o

, i ,

o 200 400 600 0 200 400 600 200 soo TWE (SEC)

I f.-l 8 i

. CONCLUSIONS FROM THE REVIEW 0F SEMISCALE TEST S-UT-8 DATA

. 1. RAPID, COMPLETE CORE UNC0VERY PRIOR TO LOOP SEAL CLEARING (T pggg = 800*F) WAS CAUSED BY DIFFERENTIAL PRESSURE BETWEEN THE VESSEL AND THE DOWNCOMER,

2. THIS DIFFERENTIAL PRESSURE DURING PUMP SUCTION LIQUID EXISTANCE IN TEST S-UT-8 RESULTED FROM A NUMBER OF FACTORS, THE MOST IMPORTANT OF WHICH WAS A HIGHER INVENTORY LEVEL IN THE STEAM GENERATOR UPFLOW SIDE THAN~IN THE DOWNFLOW SIDE (LIQUID HOLDUP),

3. STEAM GENERATOR LIQUID HOLDUP RESULTED FROM SLOWING THE DRAINING 0F LIQUID FROM THE UPFLOW SIDE BY:

A) CARRYING OVER LIQUID INTO THE U-TUBES FROM THE HOT LEG AND UPPER HEAD, B) COUNTERCURRENT FLOWING STEAM, AND C) CONDENSING STEAM IN THE U-TUBES,

!4 , CORE REWET STEAM PRODUCTION AND BULK STEAM FLOWING INTO THE DOWNCOMER FROM THE VESSEL LOWER PLENUM ARE EFFECTS PRODUCED BY EXTREME CORE UNC0VERY IN TEST S-UT-8 WHICH DELAYED LOOP

- SEAL CLEARING AND PROLONGED CORE UNC0VERY.

5. SEMISCALE DESIGN FEATURES AND TEST S-UT-8 OBSERVED PHENOMENA NECESSITATE ALTERNATE MODELING OPTIONS IN THE C-E SBLOCA MODEL.

A-19

I i

BE/EM ANALYSIS RESULTS l

-l

~

OBJECTIVE: VALIDATION OF EM COMPONENT MODELS FOR SG BEHAVIOR AND PRE-LOOP SEAL CLEARING CORE UNC0VERY I

e RATIONALE FOR BE/EM ANALYSIS e BE MODEL REPLACEMENTS FOR BE/EM ANALYSIS e BE/EM N0 DING DIAGRAM e COMPARISON OF ANALYSIS RESULTS AND TEST DATA e CONCLUSION A-20

R.-

RATIONALE FOR BE/EM ANALYSIS EM WAS-DEVELOPED FOR ANALYSIS OF NSSS RESPONSE LEADING TO e

CONSERVATIVELY HIGH PCT AND IS NOT APPLICABLE TO SEMISCALE.

e EM HAS SEVERAL MODELS (E.G., MOODY BREAK FLOWT WHICH GIVE CONSERVATIVE RATHER THAN CORRECT RESULTS.

~

e USING THE BE MODEL AND THE BE ANALYSIS RESULTS FOR GUIDANCE, SELECTED BE COMPONENT MODELS IMPORTANT FOR STEAM GENERATOR LIQUID HOLDUP AND CORE UNC0VERY ARE REPLACED BY THEIR EM COMPONENT MODEL COUNTERPARTS.

o THIS PROCEDURE IS CONSISTENT WITH THE APPROACH USED PREVI-OUSLY TO RESPOND TO THE ISSUES RAISED IN ITEM II.K.3.30.

e THE BE/EM APPROACH CONSISTS OF JUSTIFYING THE CONTINUED ACCEPTANCE OF THE CEFLASH-4AS COMPUTER PROGRAM FOR SBLOCA EVALUATION BASED ON THE CONSERVATIVE CALCULATION OF PCT, I

1 e THIS BE/EM APPROACH WAS DISCUSSED WITH NRC ON 12-10-84, WITH NO SUBSTANTIAL OBJECTION, A-21  !

_ _ _ _ - - -__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. - _ _A

C-E SMALL BREAK LOCA HYDRAULICS CODE CEFLASH-4AS  ;

i EVALUATION MODEL (EM) VERSION LICENSING CALCULATIONS OF NSSS ONLY (NOT ADAPTABLE TO SEMISCALE)

BEST ESTIMATE (BE) VERSION BE CALCULATION OF LOFT AND SEMISCALE STANDARD PROBLEMS SPECIAL STUDIES FOR RESPONSE TO POST TMI ACTION PLAN ITEM 11 K 3.30 JUSTIFICATION OF RCP TRIP STRATEGY BASIS FOR BG&E SIMULATOR TRANSIENT BEHAVIOR SOFTWARE STARTING POINT FOR IMPROVED SMALL BREAK LOCA MODEL BE VERSION EASILY APPLIED TO SEMISCALE AND TEST S-UT-8 A-22

l MODIFICATIONS OF CEFLASH-4AS BE VERSION USED/NOT USED IN BE/EM ANAL,

.e IMPROVEMENTS'TO BE STEAM GENERATOR COMPONENT MODELS

]

. 1. FLOODING IN STEAM GENERATOR U-TUBES

2. BE CONDENSATION HEAT TRANSFER IN U-TUBES NOT (SED

3. CONDENSATION OF BUBBLES IN THE U-TUBES )

4. BE BUBBLE CONVECTION IN THE U-TUBES s MODIFICATION TO REPRODUCE SPECIFIC TEST S-UT-8 OBSERVATION

5. CORE REWET STEAM PRODUCTION .NOT USED i

e REPRESENTATION OF CERTAIN SEMISCALE DESIGN FEATURES

6. MODIFIED BUBBLE RELEASE RATE IN VESSEL j AND DOWNCOMER  :

7. MODIFIED TWO-PHASE FRICTIONAL LOSSES I USED IN HOT LEGS ,

8. VARIABLE AREA VESSEL LOWER PLENUM e CHANGES TO ELIMINATE NUMERICAL DIFFICULTIES

9. STABILIZE STEAM GENERATOR INLET FLOW 1 USED

. PATH QUALITY

10. STABILIZE WALL HEAT ADDITION FOR HIGH OVALITY IN HOT LEGS 1

A-23

BE COMPONENT MODEL REPLACEMENTS FOR BE/EM ANALYSIS

~

1. FLOODING IN SG U-TUBES NOT USED

2. BE CONDENSATION HEAT TRANSFER IN 0-TUBES REPLACED BY EM CONDENSATION HEAT TRANSFER

3. CONDENSATION OF BUBBLES IN 0-TUBES NOT USED

4. BE BUBBLE CONVECTION IN U-TUBES REPLACED BY EM BUBBLE CONVECTION

5. CORE REWET STEAM PRODUCTION NOT USED

6. NON-EQUILIBRIUM MODELING DURING HPSI REPLACED BY EQUILIBRIUM MODELING

7. BE VESSEL MODEL REPLACED BY EM VESSEL MODEL:

(A) TWO NODE VESSEL REPLACED BY SINGLE N0DE (B) UPPER HEAD TO DOWNCOMER BYPASS NOT USED (C) BE WALL HEAT IN VESSEL REPLACED BY EM WALL HEAT 4

A-24

g - 3 2

1 A

/ 4-1 6-1 2

/ 4 1

7 1

1 3 2 1 7

1 K

R 6 O 2 W o 9 T 3 E 1 1 N

H T

A P

W 8 O -

0 2 9 8 L T 2 1 I 1 F U

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E S D

0 T N S ~ _

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0 _ -

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S _

=

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/

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=

7 1

1 3 1

4 7 1

\ A 9 1

A

/ 4- 6- 7

/ 5 6

_ =

5 3h

BE/EM ANALYSIS RESULTS OF TEST S-UT-8 KEY TO FIGURES:

4 CEFLASH-4AS MODEL PREDICTIONS S-UT-8 DATA PRESSURE 1 VESSEL COLLAPSED LEVEL 1 VESSEL COLLAPSED LEVEL 4 INTACT LOOP STEAM GENERATOR UPSIDE COLLAPSED LEVEL 5 INTACT LOOP STEAM GENERATOR DOWNSIDE COLLAPSED LEVEL 6 INTACT LOOP SUCTION LEG STEAM GENERATOR OUTLET COLLAPSED LEVEL 7 INTACT LOOP SUCTION LEG PUMP INLET (LOOP SEAL)

COLLAPSED LEVEL 9 DOWNCOMER MIXTURE LEVEL 1 VESSEL O

e A-26

1 i

VESSEL  !

PRESSURE 1 SUT8 BE/EM l

. BE/EM PREDICTION 5-UT-8 DATA '

2400.0  ! - - - . .

. ,4: ,,m.

g _

211 T- * " "

2000.0 ii i -i 4 _ F wrnt -

i 3rr-

. e_ _

_1-mo m.

1600 0 -Q ir I

cr  !

D \

Q- i 1200.0

'N s x

~~Q w

5 N w

w 800 0 N -

w 's N' 's '

a_ -,

R 400 0 00 a a a a O O O O a . . . .

O O O O O

(.O OJ CD v C - CO v (.O

~

TIME IN SEC A-27

VESSEL COLLAPSED LEVEL 1 SUT8 B.E/EM .

ew 4

BE/EM PREDICTION

- - - - - S-UT-8 DATA 30 000 . ..

.. .h - , ,, ,

a'12 - a';t T. ~ "

25 000 4_.

' i

' "_ _ ).

~ -

nJ--& : '

r. tsy l ,

-D 20 000-s

& \

w ',

uJ LL '

TOP 0 CORE 15 000 '

uJ s

'g

_.J i n

J \

10 000 '

\ c- ,N

_J s CJ I / \

u 6 i x

's N -

5.000 ,

y (V t  ! .

%v BOTTOM 0F CORE 0 000 e a a a o o e a 1 a . . . .

O o o o o W N CD v a - m v W

' ~

l TIME IN SEC A-23 L.________-_____._

liiTACT LOOP STEAM CEiiERATOR UPSIDE COLLAPSED LEVEL 4 SUT8 BE/EM

=

BE/EM PREDICTION

- - - - UT-8 DATA 36 000

, ..= . h . . ,,

"1- -

211 4

}

30.000

~

>?" t :i i :

l s , . .

r1J--W~  : .,. ;ch . .

i ,,l \i l

m. n

_(j ,,

^-

mm 24 000 'l i

H ls. t

'dJ ie t W ir i 18 000 k y n

> t uJ t

! I

-.J L 12.000 '

_J l o i o i

\  ;

t  :

6.000 \\

\ .

i l I

0 000  !'~'--

• I O O O O o

O O O O O W N 03 v O - (D v W TIME IN SEC A-29 l

l

I!iTACT LOOP STEMi GEiiERATOR DOWESIDE COLLAPSED LEVEL 5 SUT8 BE/EM l

BE/EM PREDICTION l S-VT-8 DATA ,

36.000 - -. - -,. . ..

J" - i .

l ..

i , ,

-1 2 -

-22 y "

30.000 g _.

't :i i : ._ _ ,

~rh W .

: . ..  : is a- a-n

_9 , _

m 24.000 l',

,' 't w *t

• uJ uJ lI ','

LL ,' i 18.000 ',

6 i '

> i i ,

uJ t #

g

_) i i

' t

' i

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12.000 ',

J ,

a i LJ l n

\

n 6.000 s .

I l

L O.000 y a g 6 O O O O O . . . .

O O O O O (O N CD v

,C - M v w TIME IN SEC .

A-30

l l

l l __

Ii; TACT LOOP SUCTI0ii LEG STEAii GEi'ERATOR OUTLET COLLAPSED LEVEL 6 SUT8 BE/EM l

• BE/EM PREDICTION

- - - - - S-VT-8 DATA 18 000 , , , _ _ _ _ , , , , , ,

- L.,,m n'a'u - a'.;.

j .. T-15 000 -

c- " ' ' -

3 nW

-  : :..  : iTurr-a- =-

,s m c0 m 12.000 ,

i

& \ I i

'g i I LLJ i s' 's u_ , i 9.000 l 'i is i

> .. i w

i

__J i i

6.000 t a ,

O i U i i

3.000 i)i

,i n.

ill

. ie, i,,

[

'd !

0.000 a a a a O O O O g . . . .

O O O O O (D N CO V O - M v W

~

. TIME IN SEC 1 A-31 4 i

liiTACT LOOP SUCT10ii LEG PUHP IliLET COLLAPSED LEVEL 7 SUT8 BE/EM BE/EM PREDICTION

- - - - S-UT-8 3ATA ,

18.000 . . . . - . .

a: .. .. , ,,

s _

a T- "

15.000 i< iii 4__

_' _ _F ._

n,r Te

.:[._fu ler-

_z_

_2-

_q j _

mm o n, 12.000

+

uJ uJ LL.

9 000 y g- . - -r, uJ l -

i 6 000 '

i o >

LJ t i ,

14 3.000 .. s..

1 x .,x s -

0.000 o a o o a o a o a . . .

O O O O O

• (.O N CD v O - M v W TIME IN SEC -

MS."'

A-32 y 1

D0iGC0iER COLLAPSED LEVEL 9 SUT8 BE/EM BE/EM PREDICTION

- - - - UT-8 DATA 30.000 _ _ , , , _ _ , , ,

..- h m ,

lir -

T- . . .

25 000 .[: -

-3, rL M

~m -

2- _a_

..q j ,

mn m 20.000 o

uJ ,

u_ i 15.000 il, '

> i s i uJ ,

• \

A 1 \

._J 10.000 -

n

'j \\ i

_J \

o s U s

\

\

l i

5 000 i l

l l

I 0 000 a a a a l C O O O O . . . .

O O O O O.

f.D CNJ 00 v O - m v O

~

TIME IN SEC A-33 1

9 VESSEL MIXTURE LEVEL 1 SUT8 BE/EM

+

BE/EM PREDICTION

---

S-UT-8 DATA .

30.000 ,,.. ., _ ,_

-- 4: - - ,

a'su a'gu T- . . .

25 000 , ,[: -

n,rm c i-- . ter

- :). -

_2_

_m_

m m.  !

20 000 ' - I w

LLJ uJ 1 TOP 0F CORE g 15.000 , , ,

I o ',,

i i

\ s

- i i \

10.000 \ >

4 Q \ s x ', i g  ;

~ a g g

n i i s I

' l \,

i ,

\

5.000 , .

r , .

BOTTO 1 0F CORE 0 000 a a a a O O O O a . . . .

O O O O O

(.O N CO v O - m v (.O TIME IN SEC .

A-34

S-UT-8 BE/EM ANALYSIS C0liPARIS0W OF PEAK AXIAL TEl1PERATURE FROFILES PRIOR TO LOOP SEAL CLEARING 12 PARCH HOT ROD CODE

- S-UT-8 DATA ENVELOPE 4

A E 8' E

-es_ _9 o.

C o0 Q s i

$, 4 9

9

/

d 5d0 7d0 9d0 11'00 TEMPERATURE (OF, AFTER LOOP SEAL CLEARING 12- y

\

N

\

)

t8  %$

ny 5 * ,'

- E /

C s' /'

d. 4 , # #~ .

i 0

0- .

, , . 1 1 - - - t i 500 700 900 1100 1300 1500 TE;!PEPATURE (OF)

A-35

l l FACTORS AFFECTING CORE UNC0VERY SPIKE IN TEST S-UT-8 e FOR SPIKE TO OCCUR, THE UPPER PLENUM PRESSURE MUST BE GREATER THAN THE DOWNCOMER PRESSURE DURING PUMP SUCTION LIQUID SEAL EXISTENCE. -

o THE FOLLOWING FACTORS CAUSE QUALITATIVE EFFECT .

THE UPPER PLENUM PRESSURE T0 IN TEST S-UT-8 BE HIGHER (EXCEPT #5): MODELED IN BE/EM

1. HIGHER INVENTORY LEVEL IN SG UPFLOW SIDE ~ BIG YES THAN DOWNFLOW SIDE (LIQUID HOLDUP) BECAUSE A. UPPER HEAD DRAINING INTO UPPER PLENUM BIG YES(EM)

SUPPLIES INVENTORY WHICH IS CARRIED INT 0 SG UPFLOW SIDE B. SG DOWNFLOW SIDE LOOSES INVENTORY MEDIUM YES(EM)

FASTER THAN UPFLOW SIDE C. CCFL IN SG SLOWS UP DRAINING OF SMALL N0(EM)

UPFLOW SIDE D. MORE CONDENSATE IS GENERATED IN UPFLOW SMALL N0(EM)

SIDE OF SG

2. FRICTIONAL LOSSES IN " PANT LEG" DURING MEDIUM YES(BE)

COUNTERCURRENT FLOW ADDS TO THE LOOP PRESSURE LOSSES

3. WALL HEAT IN CORE REGION MEDIUM YES(EM)

\

4. CORE REWET STEAM PRODUCTION MEDIUM N0(EM)

5. BYPASS LINE PROVIDES PATH FOR STEAM FROM SMALL N0(EM)

UPPER HEAD TO DOWNCOMER  !

l

6. EQUILIBRIUM FLUID STATE IN THE COLD LEGS SMALL YES(EM)

AND DOWNCOMER DURING HPSI FLOW 4

n-36 l

l CONCLUSIONS FROM BE/EM ANALYSIS e THE BE/EM ANALYSIS RESULTS SHOW ACCEPTABLE AGREEMENT WITH THE INITIAL CORE UNC0VERY SPIKE PRIOR TO LOOP SEAL CLEARING. >

,e THISVALIDATESTHEMCCEPTABLILITYOFTHEEMCOMPONENT f

MODELSlKPOPTANTFORS'ilEAMGENERATORLIQUIDHOLDUPAND CORE UNC0VGY TO ADECM.TELY REPRESENT

. THE TIMING, DEPTH, AND DURATION OF THE INITIAL x, ,

CORE UNC0VERY SPIKE,

. THE INVENTORY RECOVERY AFTER LOOP SEAL CLEARING, L , 711E SilBSE0l!ENT INVENTORY B01LGFF IN THE CORE, AND

. THE RES'!LTMIT PFA CLTPDINC TErPFPETI.'PE, l k

( .p (CONTihuED)

['

~

. 1 l

l

j. s l l l

A-37

CONCLUSION FROM BE/EM ANALYSIS (CONTINUED) e THE'BE/EM ANALYSIS ADEQUATELY PREDICTS THE CORE UNC0VERY SPIKE FOR THE FOLLOWING REASONS:

4

1. 'THE' SINGLE NODE VESSEL WITH NO BYPASS FLOW TO THE DOWNCOMER PRODUCED ROUGHLY THE SAME STEAM AND LIQUID - _

CARRYOVER TO THE STEAM GENERATOR FROM THE VESSEL AS IN S-UT-8 (INFERRED FROM THE BE ANALYSIS WITH MULTI-N0DE l VESSEL).  ;

2. - THE CEFLASH-4AS MODELS AFFECTING LIQUID HOLDUP PRODUCED THE SAME DIFFERENCE IN LIQUID LEVEL.BETWEEN THE UPFLOW AND DOWNFLOW SIDES OF-THE STEAM GENERATOR IN THE BE/EM ANALYSIS AS IN TEST S-UT-8,THUS PRODUCING SIMILAR HYDROSTATIC PRESSURE DIFFERENTIALS ACTING TO DISPLACE LIQUID FROM THE VESSEL. .

'3. REDUCED. STEAM FLOW DURING LOOP SEAL CLEARING IN THE BE/EM ANALYSIS COMPARED T0. TEST S-UT-8 WITH CORE REWET (AS INFERRED FROM THE BE ANALYSIS) LED TO PROLONGED CORE RECOVERY AND' SUBSEQUENT LOWER CORE LEVELS DURING THE BOILOFF PERIOD.

4. USE OF THE EM WALL HEAT MODEL IN THE VESSEL PRODUCED MORE WALL TO COOLANT HEAT TRANSFER IN THE BE/EM ANALY-SIS THAN IN TEST S-UT-8 (AS INFERRED FROM THE BE ANALYSIS) CONTRIBUTING TO LOWER VESSEL LEVELS.

- i e THEREFORE, REPRESENTATION OF THE MOST IMPORTANT FACTORS l AFFECTING THE CORE UNC0VERY SPIKE (UPPER HEAD DRAINING AND ,

LIQUID CARRYOVER TO SG U-TUBES) AND CONSERVATIVE REPRESENTA-TION OF LESS IMPORTANT FACTORS (WALL HEAT AND CORE REWET)

RESULTED IN ADEQUATE PREDICTION OF THE CORE UNC0VERY PHENOM-ENA IN THE BE/EM ANALYSIS, A-38

IMPLICATIONS OF S-UT-8 ANALYSES ON SBLOCA LICENSING ANALYSES 1

CONCLUSIONS FROM OCTOBER 4, 1984 MEETING (SUPPLEMENT 2P T0 l CEN-203-P)

]

l

. e THE FEATURES OF THE EM IMPORTANT FOR MODELING STEAM l GENERATOR LIQUID HOLDUP DEMONSTRATE THE OBSERVED STEAM I I

GENERATOR AND CORE UNC0VERY PHENOMENA.

e THE INTEGRAL EFFECTS PREDICTED FOR A C-E NSSS ARE ANALOG 0US TO THE OBSERVATIONS OF TEST S-UT-8 CONCLUSIONS FROM THE BE/EM ANALYSIS PREDICTIONS e SIDE-BY-SIDE COMPARISONS OF THE BE/EM PREDICTION FOR SEMISCALE TEST S-UT-8 AND THE EM PREDICTION FOR SYSTEM 80 0.35 FT 2 COLD LEG BREAK SHOWS THAT THESE TWO ANALY-SES DEMONSTRATE ANALOGOUS CORE UNC0VERY PHENOMENA, e STEAM GENERATOR LIQUID HOLDUP AFFECTS CORE UNC0VERY TO LESSER EXTENT IN C-E NSSS THAN IN SEMISCALE TEST S-UT-8 DUE TO GE0 METRIC DIFFERENCES AND UNIQUE TEST CON-DITIONS, e EM COMPONENT MODELS IN THE BE/EM ANALYSIS DEMONSTRATED

. CONSERVATIVE CORE UNC0VERY DURING THE BOILOFF PERIOD WHERE PCT OCCURS FOR NSSS LICENSING ANALYSIS, A-39

0 P

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I NSSS BREAK SPECTRUM ANALYSES LIMITING BREAK SIZES 5.' s STEAM GENERATOR LIQUID HOLDUP PRECEDES BOILOFF CORE UNC0VERY BY 400 SECONDS; CONSERVATIVE MODELING OF HPSI FLOW OVERSHADOWS THE EFFECTS OF LIQUID HOLDUP PRIOR TO CLEARING OF THE LOOP SEALS, e IN EM ANALYSES, C-E MODELS THE MINIMUM GUARANTEED HPSI FLOW FROM ONE PUMP, NON-LIMITING BREAK SIZES e IMPACT OF LIQUID HOLDUP ON CORE UNC0VERY IS GREATEST FOR BREAK SIZES LARGER THAN THE LIMITING BREAK, e IN ORDER FOR LARGER BREAK SIZES TO BECOME LIMITING DUE TO MORE DETAILED MODELING OF LIQUID HOLDUP, THE CHANGE IN PRE-LOOP SEAL CLEARING CORE UNC0VERY MUST PRODUCE INCREASES IN THE PRE-LOOP SEAL CLEARING TEMPERATURE SPIKE OF AT LEAST 550*F.

e IN EM ANALYSES, C-E CONSERVATIVELY MODELS THE CORE

]

INVENTORY AT THE START OF REFLOOD (SIT ACTUATION) WHICH OVERSHADOWS LIQUID HOLDUP EFFECTS PRIOR TO LOOP SEAL

. CLEARING, A-41

CONCLUSIONS ON THE IMPACT FOR NSSS ANALYSES 9

e THE RESULTS OF THE POST-TEST ANALYSIS OF TEST S-UT-8 .

DEMONSTRATE THAT THE EM IS ADEQUATE FOR PREDICTING THE TYPE OF VESSEL LIQUID LEVEL DEPRESSION OBSERVED IN THE DATA 0F TEST S-UT-8.

e MORE DETAILED MODELING OF STEAM GENERATOR LIQUID HOLDUP WOULD NOT CHANGE THE LIMITING BREAK SIZE OR LIMITING PCT OF NSSS BREAK SPECTRUM ANALYSES.

l 1

A-42

n l CONCLUSIONS FROM POST-TEST ANALYSIS OF TEST S-UT-8 e THE BE/EM ANALYSIS PRODUCES ACCEPTABLE AGREEMENT WITH THE INITIAL CORE UNC0VERY SPIKE DURING TEST S-UT-8 PRIOR TO LOOP SEAL CLEARING.

e THE BE/EM ANALYSIS RESULTS DEMONSTRATE ACCEPTABLE CAPABILITY OF THE EM COMPONENT MODELS TO REPRESENT INITIAL CORE UNC0VERY SPIKES. THE. POST-LOOP SEAL' CLEARING CORE RECOVERY IS CONSERVATIVELY PREDICTED, e MORE DETAILED MODELING OF STEAM GENERATOR LIQUID HOLDUP WOULD NOT CHANGE THE LIMITING BREAK SIZE OR LIMITING PCT OF NSSS BREAK SPECTRUM ANALYSES, 1

A-43

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