ML20237B859

Suppl 3-A to Rev 1-P to Post-Test Analysis Semiscale Test S-UT-8,Response to NRC Conditional SER Issued 850620 on Justification of C-E Small Break LOCA Methods

ML20237B859
Person / Time
Site:	Maine Yankee
Issue date:	12/31/1985
From:	ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY
To:
Shared Package
ML19302D186	List: ML20237B859 ML20237B871
References
CEN-203-P-S03-A, CEN-203-P-S03-A-R-1P, CEN-203-P-S3-A, CEN-203-P-S3-A-R-1P, NUDOCS 8712170203
Download: ML20237B859 (198)

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? . . POST-TEST ANALYSIS OF

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SEMISCALE TEST S-UT-8..

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") RESPONSE TO NBC'S CONDITIONAL SER ISSUED JUNE 20,1985 DN THE ag ', .-

JUSTIFICATION OF C-E SMALL BREAK h].g '

..,, LOCA METHODS .

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LEGAL NOTICE ,

THIS REPORT BY COMBUSTION WAS PREPARED ENGINEERING, INC. AS AN ACCOUNT OF WORK SPONSORED e NOR ANY PERSON ACTING ON ITS BEHALF:NEITHER COMBUSTION EN 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 RIGHTS;ORDISCLOSED IN THIS REIVRT MAY NOT INFRINGE PRIVATELY B. ASSUMES ANY LI ABILITIES WITH RESPECT TO THE USE OF, OR FOR DAMAGES RESULTING FROM THE USE OF, ANY INFORMATION, APPARATUS, METHOD OR PROCESS DISCLOSED IN THIS REPORT.

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. 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 Prepared for the C-E OWNERS GROUP By

. TRANSIENT METHODS and LOCA NUCLEAR FUEL ENGINEERING e

DECEMBER 1985 cotusustion)musinsensua

f# 4 UNITED STATES I't i

NUCLEAR REGULATORY COMMISSION

• wasHmGToN.D. C 20655 .

February 11, 1987

.... 4 2

Dr. J. K. Gasper, Chairman

, Combustion Engineering Owners Group .

1623 Horney Street l Omaha, 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 S-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 CEOG and/or the ap-plicants referencing the topical report will be expected to revise and e

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Dr. J. K. Gasper February 11, 1987 resubmit their respective documentation, or submit justification for the ,

continued effective applicability of the topical report without revision of their respective documentation.

Sincerely.

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

Enclosure:

Safety Evaluation ,

cc: See next page D

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SAFETY EVALUATION SUPPLEMENT RELATING TO VERIFICATION OF ANALYSIS METHOD 5 FOR SMALL BREAK LOCAs TMI ACTION ITEM II.K.3.30 FOR COMPUSTION ENGINEERING PLANTS FOST TEST ANALYSIS OF  ;

SEMI 5CALE TEST 5-UT-08 CEN-203-p, PEVISION 1-P, SUPPLEMENTS 3 AND 4

.0. INTRODUCTION l 5

In its safety evaluation (Referer.ce 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 water level was attributed to liquid holdup in the steam generator, i.e., the liquid briefly accumulated in the U-tubes of the intact loop steam generator.

To satisfy 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

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part of the CEFLASH-4AS licensing model, the CEFLASH-4AS model could rot o be used directly to calculate results that would agree with the experimen-tal data. A "best estimate" (BE) version of CEFLASH-4AS was devised for  ;

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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 centained modeling features which would predict the drop in

b 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").

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

1. The EM version of CEFLASH-4AS was modified to a best estimate (BE) program for calculating the S-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 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 Section 7 of Reference 2.

3. The factor: that caused the core uncovery spike in S-UT-08 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 Reference 3.

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4. Those items of the EM that were not used in the BE/EM

. analysis were reviewed to assure that they do not affect 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 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

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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 calculate 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 reouested the CEOG to evaluate its cause. ,

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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 drops 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.

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 portions of the core periphery directly beneath them. This is such a 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 CEFLASH-4AS.

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

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. The component models (e.g., reactor kinetics model) 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.

O 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 CEFLASH-4AS computer program might not be able to calculate the initial rapid drop in water lev'el experienced in the simulated reactor vessel in the Semiscale Test S-UT-08. The staff has reviewed these submittals. It finds that the

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CEFLASH-4AS program can acceptably calculate this test. Therefore, as stated in NRC's safety evaluation (Reference 1) the requirements to per-form plant specific analyses, per Action Item II.K.3.31, will no longer

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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 S-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 1506.

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ABSTRACT In response to the NRC's conditional Safety Evaluation Report. issued June 20,

.. 1985 on TMI Action Plan Item II.K.3.30 - Justification of Small Break LOCA Methods,-this report provides a post-test analysis of Semiscale Test S-UT-8.

,, The results of the post-test analysis confirm that C-E's small break LOCA thermal-hydraulic computer code, CEFLASH-4AS, 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. This report also fulfills a C-E Owners Group commitment to NRC, made prior to the issuance of the Safety Evaluation Report, to submit a post-test analysis of Test S-UT-8.

The post-test analysis included first a best estimate. analysis, which shows overall excellent agreement with the Test S-UT-8 data.. Then, specific best estimate component models important for steam generator liquid holdup and pre-loop seal clearing core uncovery were replaced by their C-E Small Break LOCA Evaluation ModelLeounterparts. The results of this best estimate /

evaluation model analysis compare well to the depth and duration of the core uncovery data and conservatively underpredict the core coolant levels after loop seal clearing. These results demonstrate that the C-E Small Break LOCA Evaluation Model incorporates component models which permit acceptable prediction of Test S-UT-8 type core uncovery. Thus, this report satisfies the one condition in the Safety' Evaluation Report and provides the information

• required for final resolution of TMI Action Plan Item II.K.3.30.

Like the reports submitted earlier in response to TMI Action Plan Item II.K.3.30, the analysis provided in this report continues to show that

. using the currently approved C-E Small Break LOCA Evaluation Model results in conservatively high cladding temperatures for the break spectrum analysis of c an NSSS.

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Tjf; 0F CONTENTS

'Section Title. . Page Abstract i .

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1.0 Introduction- 1 2.0 Sunnary and Conclusions 2-1 3.0 Semiscale Facility and Experiment Description 3-1 3.1 Facility Description 3-1 3.2 Test S-UT-8 Description 3-2 3.3- Test S-UT-8 Observations 3-4 4.0 Discussion of Procedures 4-1 4.1 Method of Analysis 4-1 4.2 Computer Program 4-3 4.3 Model Setup 4-4 4.4 Initial Conditions and Boun'ary d Conditions 4-5 5.0 Model Modifications for Analysis of Test S-VT-8 5-1 5.1 Improvements to BE Steam Generator Component 5-1 Models 5.2 Modification to Address Core Rewet Steam 5-8 Production 5.3 Representation of Semiscale Design Features 5-9 -l 5.4 Changes to Eliminate Numerical Difficulties 5-13 6.0 Results of BE Analysis 6-1 l 6.1 Comparison of BE Analysis Results and Test Data 6-1 6.2 BE Analysis Conclusions 6-6 ii

TABLE OF CONTENTS (Continued) ,

Section Title . Page 7.0 Results of BE/EM Analysis 7-1 e 7.1 BE Component Model Replacements for 7-1 BE/EM Analysis 7.2 Comparison of BE/EM Analysis Results 7-4 and Test Data 7.3 BE/EM Analysis Conclusions 7-9 8.0 Impact on NSSS Analyses 8-1 <

1 8.1 Impact of Model Modifications on NSSS Analyses 8-1 8.2 Implications of S-UT-8 Analyses on Current 8-7 SBLOCA Licensing Analyses G

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1.0 INTRODUCTION

In the summer of 1979, C-E prepared for the C-E Owners Group two reports to i

). the NRC,.CEN-114-P (Reference 1-1) and CEN-115-P (Reference 1-2), which use'

-C-E's Small Break LOCA Evaluation Model (Reference 1-3) t'o predict typica'l PWR behavior following a small break LOCA. These submittals were prepared in response to NRC-requests following the TMI-2 accident. . After review of.these l documents, the NRC. identified a number of questions with the small break model l and requested a response to these questions via the NRC TMI Action Plan, NUREG-0737, Item II.K.3.30 (Reference 1-4).

At a meeting held on January 26,.1981 (Reference 1-5) with members of the NRC staff and representatives of the C-E Owners Group and C-E, the NRC described seven specific questions concerning the C-E Small Break LOCA Evaluation Model.

The NRC staff also indicated that responding to these seven questions would fulfill the response to Item II.K.3.30 of the NRC TMI Action Plan. In March 1982, report CEN-203-P, Revision 1-P (Reference 1-6), which contained the answers to the seven NRC questions, was submitted to the NRC. This report provided justification for maintaining approval of the C-E Small Break LOCA Evaluation Model. In August 1983, the NRC issued a letter to the C-E Owners Group (Reference 1-7) asking eight additional questions with the subject title

. " Request Number 1 for Additional Information on CEN-203-P". Responses to five of the eight additional questions were provided in Supplement 1-P to

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CEN-203-P, Revision 1-P (Reference 1-8). Responses to the three remaining questions dealing with steam generator modeling were provided in Supplement 2-P to CEN-203-P, Revision 1-P (Reference 1-9). In keeping with the intent of 1-1

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e report CEN-203-P, these supplementary responses continued to justify the conservatism of the currently approved C-E Small Break LOCA Evaluation Model.

During a meeting held on October 4, 1984 in Windsor with C-E Owners Group, .

- NRC, and C-E representatives, the.NRC indicated that without a post-test

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analysis of Semiscale Test S-UT-8 the responses to the steam generator modeling questions were insufficient. Subsequently, in March, 1985, the C-E Owners Group committed to provide to NRC a confirmatory post-test analysis of Semiscale Test S-U1-8 to show the acceptability of the steam generator thermal hydraulic models'of CEFLASH-4AS to calculate the Test S-UT-8 pre-loop seal clearing core level depression (Reference 1-10). Submittal of this report to the NRC satisfies that commitment.

NRC completed the review of CEN-203-P and Supplements 1-P and 2-P in June, 1985, and' issued a conditional Safety Evaluation Report (Reference 1-11). The NRC staff concluded that the material submitted, which justified the conserva-tism of the C-E Small Break LOCA Evaluation Model, was acceptable pending one condition. The one condition required "... confirmation that the CEFLASH-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,"

Prior to the issuance of the Safety Evaluation Report and before starting the post-test analysis of Test S-UT-8, a meeting was held on December 10, 1984, in C-E's Bethesda Office with C-E Owners Group, NRC, and C-E representatives.

I The scope of the post-test analysis work and the planned method of analysis were presented to the NRC at the meeting. Details of the analysis method are 1-2 ,

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discussed in Section 4.0. The NRC did not object to the analysis approach, l l but emphasized that any model change made to improve comparison of the l

l analysis to the test data must be assessed for its importance or impact on l

. NSSS calculations. Accordingly, the model modifications made for the analysis of Test S-UT-8 are described in Section 5.0, and their impact on NSSS analyses is discussed in Section 8.0. The overall results of C-E's post-test analysis are summarized in Section 2.0 and presented in detail in Sections 6.0 and 7.0.

The analysis begins with a review of Semiscale Test S-UT-8 given in Section 3.0.

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l References for Section 1.0 1-1 CEN-114-P (Amendment 1-P), " Review of Small Break Transients in Combustion Engineering Nuclear Steam Supply Systems", July 1979. -

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1-2 CEN-115-P, " Response to NRC IE Bulletin 79-06C, Items 2 and 3 for C-E Nuclear Steam Supply Systems", August 1979.

1-3 CENPD-137P, " Calculative Methods for the C-E Small Break LOCA Evaluation Model," August, 1974.

1-4 NUREG-0737, " Post TMI Action Plan Requirements", October 1980.

1-5 Letter, K. P. Baskin (C-E Owners Group) to P. S. Check (NRC), " Planned Response to NRC TMI Action Plan Requirement II.K.3.30", July 14, 1981m 1-6 CEN-203-P, Revision 1-P, " Response to NRC Action Plan Item II.K.3.30, Justification of Small Break LOCA Methods," March 1982, 1-7 Letter, C. O. Thomas (NRC) to R. W. Wells (C-E Owners Group), "NRC Request Number 1 for Additional Information on CEN-203-P., Revision 1-P, Small Break LOCA Methods", August 8, 1983. -

1-8 CEN-203-P, Revision 1-P, Supplement 1-P, " Responses to NRC Request Number i for Additional Information on C-E Report CEN-203-P, i

Revision 1-P," February,1984.

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References-for Section 1.0 (Continued)'

1-9 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.- i 1-10 Letter, R. W. Wells '(C-E.0wners Group) to C. O. Thomas (NRC), "Small o Break LOCA Methods Verification ( Action Plan Item II.K.3.30), Comitment to Perform Post-Test Analysis of Semiscale Test S-UT-8," March 11, 1985..

1-11 Letter, C. O. Thomas'(NRC) to R. W. Wells (C-E Owners Group),

" Conditional Acceptance for Referencing of Licensing Topical Report CEN-203(P) Rev. 1, ' Response to NRC Action Plan Item II.K.3.30 Justification of Small Break LOCA Methods'," June 20, 1985.

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2.0

SUMMARY

AND CONCLUSIONS 1

This report provides a post-test analysis of Semiscale Test S-UT-8 in response- )

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to the NRC's conditiorsal Safety Evaluation Report issued June 20, 1985 on TMI Action Plan Item II.K.3.30, Justification of Small Break LOCA Methods. The objective of the post-test analysis of Semiscale Test S-UT-8 is to satisfy the one condition of the Safety Evaluation, Report, namely to confirm that C-E's thermal-hydraulic small break LOCA computer code,.CEFLASH-4AS, can acceptably calculate pre-loop seal clearing ' core uncovery as observed in the data from Test S-UT-8.

Review and analysis of the Semiscale Test S-UT-8 data revealed several phenomena which proved to be important for the post-test analysis. The test demonstrated that steam generator liquid holdup contributed to a rapid, spike-like, complete core uncovery prior to loop seal clearing for a 5% break area small cold leg break loss-of-coolant accident (i.e., equivalent to a break area of 0.2 fte in a typical C-E NSSS). More detailed examination of the data from Test S-VT-8 showed that the adverse pre-loop seal core uncovery behvior was not entirely dependent on steam generator liquid holdup. During the liquid level depression, thermocouple data from the upper four feet of the heater rods showed no cladding heatup due to liquid flowing back into the vessel from the hot legs and rewetting the heater rods. Steam produced by rewetting the heater rods increased the duraticn of core uncovery prior to loop seal clearing by forcing steam to flow from the vessel into the downcomer. This bulk steam flow displaced liquid from the downcomer and partially refilled the pump suction loop seal, thus delaying loop seal clearing and prolonging core uncovery.

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The post-test analysis of Test S-UT-8 was divided into two phases: best-estimate (BE) analysis and best estimate / evaluation model (BE/EM) analysis.

A. BE Analysis The BE analysis used the most up-to-date best estimate ,

component models and code options to benchmark the CEFLASH-4AS code against Test S-VT-8 data. Modifications were made to best estimate component models to represent specific Semiscale design features and to improve the prediction of individual models relative to observations from j the test data. Improvements were made to BE steam generator component models for condensation heat transfer and countercurrent flooding in the U-tubes. A core rewet steam production model was added. Also, modifications were made to the downcomer and vessel bubble release rate models to allow for the slug-like passage of bulk steam from the vessel to the downcomer. Finally, the calculation of countercurrent flow in the small diameter Semiscale hot leg was modified to include the effects of interfacial drag on steam generator draining.

The results of the BE analysis of Test S-UT-8 showed overall excellent agreement between CEFLASH-4AS predictions and test data. In particular, the BE analysis accurately predicted the timing, depth, and duration of the liquid level depression in the vessel prior to loop seal clearing.

The BE analysis also accurately predicted liquid accumulation in the .

U-tubes and the timing of loop seal clearing. With the incorporation of the model modifications for representing the reverse flow of steam into the downcomer, the BE analysis accurately predicted the downcomer and 2-2

loop seal liquid levels. These results demonstrate that the BE'model is capable of accurately predicting the effects of steam generator liquid holdup on core uncovery prior to clearing of the loop seals as observed in Test S-VT-8.

B. BE/EM Analysis' In the BE/EM analysis, selected BE component models important for steam generator liquid holdup and core uncovery were replaced by their EM component model counterparts. The steam generator component model replacements included no countercurrent flooding in the U-tubes and EM condensation heat transfer models in the U-tubes. Core rewet steam production is not an EM component model, therefore, it was not used-in the BE/EM analysis. Upper head to downcomer core bypass was not represented in the BE/EM analysis and the two node BE vessel model was replaced with a single node EM vessel model. Initialization of vessel wall heat in the BE/EM analysis followed the Evaluation Model prescription. The purpose of the BE/EM analysis was to demonstrate the capability of the selected EM component models to predict the core' level depression observed in the data of Test S-UT-8.

The results of the BE/EM analysis of Semiscale Test S-VT-8 showed acceptable agreement with the initial core uncovery data prior to loop seal clearing. Due to the absence of modeling core bypass flow, the timing of core uncovery, by the mixture level, was predicted earlier by the BE/EM analysis and the predicted rate of vessel inventory loss was greater than the Test S-UT-8 data. The BE/EM analysis predicted complete core uncovery prior to loop seal clearing as observed in the test in spite of lower steam generator liquid levels and earlier loop seal 2-3 h

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clearing produced by the replacement of BE steam generator component models with EM component models. The core level depression in Test S-UT-8, prior to loop seal clearing, is dependent on the hydrostatic pressure imbalance between the vessel and downcomer, which in turn, is -

largely dependent on the difference between the collapsed liquid levels in the upflow side and the downflow side of the U-tubes, i.e., the steam generator pressure difference from inlet to outlet. Even though the predicted steam generator liquid levels were lower than the measured levels due to the absence of a countercurrent flooding model, the steam generator pressure difference was the same in the BE/EM analysis as measured in the test. Therefore, the effects of steam generator liquid holdup on core level depression in the BE/EM analysis were very similar to those observed.in Test S-UT-8. Without the core rewet steam production model, the BE/EM analysis did not predict the duration of complete core uncovery as seen in the test data. However, the BE/EM analysis predicted conservatively low vessel inventory at the time of core recovery after loop seal clearing and during the subsequent ,

inventory boiloff period. This predicted vessel inventory transient produced conservatively high cladding temperatures during the initial core uncovery and later core boiloff periods.

-l These BE/EM analysis results show that the EM component models important to steam generator liquid holdup and core uncovery adequately predict the vessel liquid level depression observed in Test S-UT-8. Thus, the BE/EM analysis provides the confirmation that the CEFLASH-4AS computer program can acceptably calculate the pre-loop seal clearing core water level depression of Test S-UT-8.

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A comparison of the BE and BE/EM analysis results'for Semiscale indicates that the impact of the CEFLASH-4AS model modifications (made for the analysis of i i

Test S-UT-8) on NSSS analyses is minimal because of 1) basic design

. differences between.Semiscale and a C-E NSSS and 2) other conservative aspects of the C-E Small Break LOCA Evaluation Model. The peak cladding temperatures for the EM break spectrum analysis of an NSSS would not be increased by use of the model modifications developed for analysis of Test S-UT-8.

l This report satisfies the one condition in the NRC's Safety Evaluation Report

. of June 20, 1985, and therefore, provides the information required for final resolution of TMI Action Plan Item II.K.3.30. Like.the reports submitted earlier in response to TMI Action Plan Item II.K.3.30, this report continues to show that using the currently approved C-E Small Break LOCA Evaluation Model results in conservatively.high cladding temperatures for the break spectrum analysis of an NSSS.

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f i 3.0 SEMISCALE FACILITY AND EXPE'RIMENT DESCRIPTION 3.1 Facility Description Semiscale Test S-UT-8 was performed in the Semiscale MOD-2A facility. The.

configuration of the M00-2A facility is shown in Figure 3-1 and is described.'

in detail in Reference 3-1. The M00-2A facility is a scaled two-loop simulation of a W four loop PWR. The core power and primary coolant volume scaling ratio is'about 1/1700. Geometric similarity, elevations of majo'r components and hydraulic resistances are preserved from the full-scale system.

The core simulator is also full-length (12 ft) and electrically heated.

The facility consists of a vessel, external downcomer, an intact loop

- (representing 3 loops of the PWR), and a broken loop with break orifice assembly. Both loops have coolant pumps which are controlled in a manner to simulta e ear ly reac or l pump trip and subsequent pump head degradation t coo ant due to coastdown. Both loops have full elevation steam generators with scaled secondary coolant volumes. The intact loop steam generator has six full-size U-tubes, and the broken loop has two. The facility utilizes external-loop-piping and vessel guard heaters to minimize the effects of ambient heat loss.

One additional feature important for Test S-UT-8 is the simulation of core l- bypass flow between the cold leg inlet to the downcomer and the vessel upper head.

During Test S-UT-8, several transient events were initiated by a low pressurizer pressure trip signal of 1827.5 psia (12.6 MPa). These events 3-1

.4 included core power ~ decay, pump coastdown, steam generator isolation (i.e.g feedwater and steam-line valve closure), and high pressure safety injectfah actuation. Appropriate time delays for each event initiation were .s l automatically incorporated in the transient simulation. -

l 3.2 Test S-VT-8 Description Semiscale Test 5-UT-8 was run on December 9, 1981, at Idaho National Engineer-ing Laborator'y. The experimental operating specifications for Test S-UT-8 are described in Reference 3-2. Test S-UT-8 was a 5% small cold leg break with the break orifice located on the side of the pipe at the pipe centerline elevation. The transient was initiated from steady state full-power operating conditions by opening a valve downstream of the break orifice.

The hydraulic response immediately following break initiation cru characterized by rapid voiding of the vessel upper plenum as the primary system pressure decreased. The low pressurizer pressure trip signal was generated at 18 seconds and thereby initiated (after suitable' time deldfi) core power decay, pump coastdown, steam generator isolation, and HPSI pump discharge. Core boiling and coolant flashing rapidly decreased the vessel coolant inventory over the first 120 seconds while the system pressure leveled off at roughly 1000 psia. -

The vessel coolant inventory continued to decrease after 120 seconds due to e

The break area was scaled to represent 5% of the cold leg flow area in a full-scale W PWR. By preserving break area to total primary system volume, this translates to a break size of roughly 0.2 fta for a C-E 2700 MW plant.  ;

3-2

~

c y b

, riet

, , r k

\

t  :)

the. development of a differential pressure between the core and the downcomer

~

acting .to. push the vessel fluid int'o the downcomer and out the break. This sa is differentia 7" pressure resulted from pump suction liquid seal formation, which .

. blocked the path for steam to reach the break, and from steam generator liquid p r ..

holdup, which provided additional elevation head. Steam generator liquid holdup was the result of slowing the draining of liquid by countercurrent flowing steam, resupplying the liquid by condensing the steam in the U-tubes and by carrying over liquid into'ihh U-tubes' frord the hot legs.

Aftir200 seconds,thevessellowerplenuminletfromthedowncomeruncovered.

v Bulk steam from the vessel entered the bottom of the downcomer and displaced liquid fron the downcomer which partially. refilled the loop seal. At 240pkonds,loopsealclearingallowedforrapidpartialrefillingofthe J

vessel by downcomer fluid. Also, the break uncovered discharging higher quality fluid, and increasing the, system depressurization rate. Core liquid

(

bciTcff continued to deplete the vessel inventory un'til accumulator discharge occurred at shortly after 500 seconds when the pressure reached 600 psia. The test was' ter'ninated at 735.2'Oconds as the core coolant inventory slowly increased due to accumulator discharge.

Many modifications were made to the M00-2A facility for Test S-UT-8 that were uniquely different from those in previous tests. Some of these differences complicated the specification of code /model input for the post-test analysis.

~

For example,dn the upper head of the vessel, changes to the support column configurati.on crolted in unwanted draining passages and atypical hydraulic resistances in the upper regions of the vessel. A " frozen" turbine meter in i

, 3-3 NI 3

- _ _ - _ - _ _ _ _ _ . ~ l- _

w}g ,

y >

., . A y + $g y , ,

dj jg thelohcomer produced a downcomer to vessel initial pressure differentia 1'  ;!

y/[

+

thape'sexcessivebyafactoroftwo. A modification to the pressuMzer- u H .ig (, t surg @ine produced an' increase in its' hydraulic resistance which war

)

m y ,

Q#

• exceusive by a factor:of ten.. Also, power to the intact loop suction leg -.

y, .f f. ~

l

extstal heater was excessive by a factor of five tending to increase system J 1 pressure and vessel inventorf loss. ~

, . ti <

N 5- _J si f ~

The Semiscale facility was fully instrumented with pressure taps, turbine h . .

meters % nd densitometers. During Test S-UT-8 some instrumentation failed.

j. . The - f?o measureit4nt in.the bypass line failed thus complicating the speci-  !

b f, fit 5$10n of the .cial bypass flow percentage for the post-test analysic.P k g,

't e > '

~

Also,Jflow measurements of high pressure safety injection and accumulator

\' J, injeftion failed. All pressure differential readings on the broken,Acop steam ,

generator failed which prevented the comparison of post-test predictions to ,

^* collapsed liquid levels in that steam generator.

, i' .

3

+

/ 3.3 Test S-UT-8 Observations h

4h

,{ In Test S-VT-8, liquid holdup in the steam generator U-tubes prior to pump ,

sugti dioop seel clearing caused a rapid (roughly 150 sec) water level

.. Jj ,1 deyhession which extended to th$ hottom of the heated bundle resulting in a

7. ,  !

/ #

modgst cladding temperature excursjon to about 800*F. However, the upper .

/

s n, I appruimately four feet of t'na"neatte rods showed no cladding heatup during /

the water level cepression prob' ably due to 71guid flowing back into the vessel '

4 from the hot legs and rewetting the heater rods. Steam produced by rewetting g the upper portions of the heater rods and hot surfaces in the top of the ,,

A \

x f -

/

t C

,fh '

3-4

b. q,

's vessel increased the duration of core uncovery prior to loop seal clearing by forcing steam into the downcomer. The bulk steam was driven to the top of the downcomer in a slug-like manner, displacing liquid from the downcomer into the cold legs. Liquid displaced from the downtomer partially refilled the pump suction loop seal thus delaying the loop seal clearing process and prolonging core uncovery.

The results of Test S-UT-8 have been shown to be more severe than other Semiscale 5% cold leg break tests with similar test conditions. The companion Test S-UT-6 in the MOD-2A facility, showed much less core level depression than Test S-UT-8 due to simulation of larger bypass flow and the use of larger diameter piping in the hot leg which contributed to less steam generator liquid holdup (Reference 3-3). The most recent Tests S-LH-1 and S-LH-2 in the MOD-2C facility, also showed less core level depression than Test S-UT-8.

This was due to improvements in the test simulation of realistic NSSS conditions such as more rapid pump coastdown, reduced external heater power in the intact loop and reduced downcomer-to-vessel hydraulic resistance (References 3-4 and 3-5).

1 I

a 3-5

References for Section 3.0 3-1 System Description for the MOD-2A Semiscale System, Addendum I, MOD-2A Phase I Addendum to M00-3 System Design Description, EG&G Idaho, Inc., December, 1980.

I 3-2 EGG-SEMI-5685, " Experiment Operating Specification for Semiscale i MOD-2A 5% Break Experiment S-UT-8," W. W. Tingle, EG&G Idaho, e

Inc., December, 1981.

3-3 EGG-SEMI-6010 " Vessel Coolant Mass Depletion D6 ring a Small Break LOCA," September, 1982.

3-4 EGG-SEMI-6813. " Experiment Operating Specification for Semiscale  ;

MOD-2C 5% Small Break Loss-of-Coolant Experiment S-LH-1,"

February,1985.

3-5 EGG-SEMI-6884, " Quick Look Report for Semiscale MOD-2C Experiments S-LH-1 and S-LH-2," June, 1985.

• l 3-6

l FIGURE 3-1 1

loop

. N .

Special Features:

Instrumented SG

• Honeycome insulation P!pe heat tracing

• New core Presounzer q

Vessel BypBSS

- Pump h ,  :

, Hotleg Hot leg Q d! ' '

I'I

, Steek assemety "si'

, p Pump suction Pump J f [

Blowdown valve Condensing coils

, Puma suction # .

jp u ..

- y Condensate measunng tanks Vesses downcomer b y tNewsr m Semiscale Mod-2A Facility 3-7

______m_--.-.- ---"-

4.0 DISCUSSION OF PROCEDURES 4.1 Method of Analysis The objective of the post-test analysis of Semiscale Test S-UT-8 is to confirm that C-E's thermal-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 purpose of the analysis-is to validate the CEFLASH-4AS component models important for representing the integral effects of steam generator liquid holdup on core uncovery prior to clearing of the loop seals. This analysis does not constitute a validation of the entire small break code or of the Evaluation Model (EM), but rather a validation of selected EM component models and their integral behavior limited to the specific conditions of Test S-VT-8.

l The post-test analysis is divided into two phases: best estimate (BE) analysis and combined best estimate / evaluation model (BE/EM) analysis. These two phases are described as follows:

I 13- 1) BE Analysis The BE analysis benchmarks the CEFLASH-4AS code against Test S-UT-8 data using the most up-to-date BE component models and code options. The analysis includes modifications to the BE code version which represent specific Semiscale design features or improve the prediction of individual component models relative to observations from Test S-UT-8.

4-1

i-l l

ii) BE/EM Analysis Using the BE analysis results for guidance, selected BE component models important for steam generator liquid holdup and i core uncovery are replaced by their EM component model counterparts.

This combined best estimate and evaluation model analysis demon- '

strates the capability of the selected EM component models to predict the core level depression observed in the data of Test S-UT-8.

The two-step procedure described above is consistent with the approach used previously to respond to the issues raised in TMI Action Plan Item II.K.3.30.

This approach consisted of justifying the continued acceptance of the q CEFLASH-4AS computer program for small break LOCA evaluation based on the conservative calculation of peak cladding temperature.

As described above, the post-test analysis of Test S-UT-8 includes some code modifications. Those modifications made to represent specific Semiscale design features were included in both the BE and the BE/EM analyses, but have no applicability to the small break LOCA evaluation of C-E NSSS's. Those modifications made to improve the BE model comparison to the Test S-UT-8 data were not included in the BE/EM analysis. The applicability of these code improvements to the small break LOCA evaluation of C-E NSSS's is assessed, first, by comparison of the BE and BE/EM analysis results for Semiscale and, -

second, by estimation of their impact in break spectrum analyses for an NSSS.

4-2

4.2 Computer Program The post-test analysis of Test S-UT-8 was performed using the most up-to-date

BE version of the.CEFLASH-4AS code (Reference'4-1). Modifications have been incorporated in the BE version as a result of the continuing development of a best estimate small break model. BE versions were used previously in the following analyses which contributed to the best estimate model development

.a) CEN-114, Review of small break transients (Reference 4-2) b) CEN-115, Response to NRC IE Bulletin 79-06C (Reference 4-3)-

c) LOFT Test L3-1, (Reference 4-4) d) Semiscale Test S-07-108, (Reference 4-5) e) LOFT Test L3-6, (Reference 4-6) f) CEN-203, Justification of small break methods (Reference 4-7) g) CEN-268,' Justification of RCP trip strategy (Reference 4-8)

This BE version of CEFLASH-4AS was also the basis for the BG&E reactor coolant system simulator code, and has been used in other NSSS simulation applications such as the CENTS code (Reference 4-11).

The BE version of CEFLASH-4AS with its BE component models is easily applied to Semiscale, while the EM version of CEFLASH-4AS is not adaptable to Semiscale because it contains various component models designed specifically for.NSSS representation in licensing calculations. It is for this reason that validation of the EM component models for steam generator behavior was done

. with the BE/EM analysis approach described in Section 4.1. In addition to the 4-3 f i

I best estimate component models, the modified version of CEFLASH-4AS includes the implementation of an improved integration technique and automatic time step procedure which combine to significantly reduce code running time with no significant'effect on the transient results, (Reference 4-9).

4 4.3 Model Setup ,

The nodalization of the Semiscale MOD-2A facility for Test S-UT-8 is based on the'nodalization used for the C-E analysis of Semiscale Test S-07-100. Test analyses of Semiscale Test S-07-10D were initiated but were. terminated in May, ,

1981 when the NRC dropped the requirement to include this test in the C-E model verification (Reference 4-10). The Test S-07-10D nodalization (i.e.,

control volumes, flow paths) was modified to reflect hardware changes between the MOD-3 and M00-2A Semiscale facilities. These hardware changes included ,

smaller diameter hot and cold leg piping, a new intact loop steam generator, modifications to the upper head geometry, and addition of external guard heaters on the piping, downcomer, and vessel.

Nodal diagrams for the BE and for the BE/EM models are shown in Figures A-1 and 4-2. The model setup includes individual representation of the intac'. loop and broken loop. In each loop, dual flowpaths are utilized for the hot and cold legs to allow for the occurrence of countercurrent flow of steam and two-phase mixture. Safety injection delivery to each of the intact loop and broken loop cold legs is modeled. Each steam generator secondary region is ,

modeled as a single volume node. Feedwater flow and steam-line flow are handled by code input tables. All nodes except the steam generator U-tubes have wall heat slabs to represent metal heat and the external strip heaters.

i 4-4 l _ _ _ __ -

The BE nodalization in Figure 4-1 differs from the BE/EM nodalization'in Figure 4-2 by the representation of the vessel. :The BE analysis separately nodalizes the upper head region of the vessel and explicitly models the core flow bypass line, guide tube, and support column fluid passages. The BE/EM analysis nodalizes the vessel as a single node without representation of core flow bypass.

4.4 Initial Conditions and Boundary Conditions The measured initial- conditions for Test S-UT-8 were simulated in the post-test analysis as accurately as possible. The initial pressure dis-tribution, coolant enthalpy distribution, loop and core flow rates and core power were represented. The flow rate and pressure changes around the loops were used to specify initial friction and geometric loss coefficients. The bypass line initial flow rate and loss coefficient were based on 1.1% of the initial total loop flow. The loss c6 efficient in the pressurizer surge line was adjusted to reproduce the measur:d pressurizer depressurization rate, thus accurately predicting the time of ths low pressurizer pressure trip signal.

^ The core power decay, HPSI flow, and pump-coastdown speed, which are all tripped by the low pressurizer pressure signal, were specified as boundary conditions for this analysis. Since only one set of pump homologous curves are input to CEFLASH-4AS, the initial speed of the broken loop pump was adjus.ted to reproduce the measured pump head using the homologous curves for the intact loop pump.

4-5

L Prior to steam generator isolation which also occurs on the low pressurizer pressure signal, measured feedwater and steam-line flows were specified as boundary conditions. After steam generator isolation, the actual feedwater coastdown was modeled. To simulate the observed slow decrease in steam .

generator secondary side pressure after isolation, steam-line leakage was allowed at an average value of approximately 2% of the initial steam flow for 1

both steam generators.

The external strip heaters were modeled as heat slabs using the CEFLASH-4AS

-lumped parameter wall heat model. Values of the heat capacity and overall

. heat transfer coefficient for individual heaters were selected to simulate the 1

measured power histories.

In both thE BE and the BE/EM analyses, all metal walls were initialized at the initial coolant temperature. However, in the BE/EM analysis with the single node representation of the vessel, the initialization of the vessel walls followed the Evaluation Model prescription (Reference 4-7).

Flow out the break in Test S-UT-8 was calculated with the homogeneous equilib-rium critical flow model. An adjustment to the discharge coefficient was made for the calculation of break flow during the period of subcooled discharge to compensate for non-equilibrium effects in the break region associated with the .

Test S-UT-8 break size..

I 4-6

- _ ____.__.___..____m__.__._-__ _ - . _ _ _ _ _ _ -

I l

l References for Section 4.0 4-1 CENPD-133, Supplement 1. "CEFLASH-4AS, A Computer Program for the Reactor l* Blowdown Analysis of the Small Break Loss-of-Coolant Accident",

August, 1974.

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

January, 1977.

l 4-2 CEN-114-P, Amendment 1-P, " Review of Small Break Transients in C-E Nuclear Steam Supply Systems", July, 1979.

4-3 CEN-115-P, " Response to NRC IE Bulletin 79-06C, Items 2 and 3 for C-E Nuc' lear Steam Supply Systems", August,1979.

4-4 " Combustion Engineering Analysis of LOFT Test L3-1", February, 1980.

4-5 Letter, George Liebler (C-E Owners Group) to Dr. Denwood F. Ross (NRC), i

Subject:

Semiscale S-07-10B Small Break Test, November 30, 1979.

4-6 Letter, K. P. Baskin (C-E Owners Group) to P. S. Check (NRC), l

Subject:

Combustion Engineering Analyses of LOFT Test L3-6, March 31, 1981. i 4-7 CEN-203-P, Revision 1-P, " Response to NRC Action Plan Item II.K.3.30, Justification of Small Break LOCA Methods," March, 1982.

4-7

y

)

J 4-8 CEN-268, " Justification of Trip Two/ Leave Two Reactor Coolant Pump Trip Strategy During Transients," March, 1984.

l 4 CENPD-133, Supplement 5, '!CEFLASH-4A, A FORTRAN 77 Digital- Computer -

Program for Reactor Blowdown Analysis," June, 1985.

4-10 Letter, Dr. J. K. Gasper (C-E Owners Group) to Dr. B. Sheron (NRC),

l

" Post-Test Analysis of Semiscale Test S-07-100," May, 1981.

4-11 Espinosa, R. J. , Doherty, P. K. , McBeth, R. L. , " Application. of Thermal-Hydraulic Design Codes to Real Time PWR Simulation,"' presented at the Fifth Power Plant Dynamics, Testing and Control Symposium, Knoxville, Tenn., March 21-27, 1983.

i 4-8

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l 5.0 MODEL MODIFICATIONS FOR ANALYSIS OF TEST S-UT-8 The post-test analysis of Test S-UT-8 includes the implementation of severai

~

code modifications into the best estimate version of CEFLASH-4AS. These modifications are organized into four categories:

i) . improvements to BE steam generator component models, ii) -modification to address core rewet steam production',

iii) representation of Semiscale design features, and iv) changes to eliminate numerical difficulties.

The following four sections will discuss the nature of each model modification, its justification, and its effect on the analysis of Test S-UT-8.

5.1 Improvements to BE Steam Generator Component Models The data from Test S-UT-8 demonstrates the importance of steam generator thermal-hydraulic behavior on pre-loop seal clearing core uncovery. For the BE analysis, improvements were made to four steam generator component models I related to the phenomena of liquid holdup and primary-to-secondary heat

\

transfer. Since the Test S-UT-8 data were used as a basis, certain elements '

of these code improvements apply only to Semiscale and not to a C-E designed NSSS. None of these steam generator model modifications were utilized in the BE/EM analysis which used only EM steam generator component models.

5-1

Flooding in Steam Generator U-Tubes The BE model. in CEFLASH-4AS for limiting countercurrent flow in the steam generator U-tubes by a flooding correlation is described in Reference 5-1. -

Experiments with countercurrent flow in vertical tubes have shown that there is a limit to the downward flow rate of water in a tube which contains an upward flow of a vapor. The limit is called the flooding limit and has been correlated by the following expression (Reference 5-2): ,

1

.ek ,<%

gg +

jf =C (5-1).

W h = V hl where 4 TgDig-6) f g; Ljn-e,)

e n

For analysis of Test S-UT-8, an improvement to this model was made by the addition of a deflooding and liquid penetration model. Experiments, started  :

from an initially flooded state, indicate that the unrestrained draining of water occurs at a slightly . lower point than the flooding limit (Reference 5-3). This deflooding point is modeled in CEFLASH-4AS using equation (5-1) -

but with a different value of C:

For Flooding C = .725 For Deflooding C = .7 5-2

The actual calculated flow rate of liquid leaving the bottom of the U-tube during the period of countercurrent flooding is based on a value of C = .7.

The numerical solution technique during countercurrent flooding was also revised to enhance the code's numerical stability without reducing time steps.

The CEFLASH-4AS BE analysis results indicate that countercurrent flow limited by flooding occurs for only a short time period during Test S-VT-8. For the intact loop steam generator, flooding begins at roughly 55 seconds and ends 40 seconds later. During this time period, liquid draining from the U-tubes into the hot leg is limited by the liquid downflow model. Thus,1) liquid carried into the U-tubes from the hot leg by the countercurrent flow of steam and 2) condensate produced by primary-to-secondary heat transfer accumulate in the U-tubes. This accumulation is referred to as steam generator liquid holdup. Liquid in the U-tubes resists and restricts the flow of steam produced in the core which is flowing toward the break. This produces a pressure difference between the hot side of the system (core outlet) and the cold side (downcomer/ break) which acts to depress the level in the vessel during the time period prior to loop seal clearing. The difference in liquid column height between the upside and the downside of the steam generator also adds to the hydrostatic pressure differences acting to depress the liquid ,

level in the core. While the time period of flooding is relatively short compared to the overall S-UT-8 transient, the vessel level depression due to liquid holdup in the steam generators leads to greater inventory loss from the vessel.

5-3

I BE Condensation Heat Transfer in U-Tubes

~

T'he reason for this model modification is to improve the prediction of conden-sation heat transfer in the steam generator U-tubes. The existing conden-sation heat transfer model, which is the EM component model described in Reference 5-1, intentionally predicts low heat transfer coefficients. -Since a low heat transfer coefficient would tend to produce higher values of the primary-to-secondary temperature differential for a given steam generator heat I

load, this EM condensation model leads to prediction of higher primary system pressure which means higher inventory loss out the break, less high pressure safety injection and conservatively high cladding temperatures.

The condensation heat transfer coefficient in the EM is calculated using the high-flow correlation of Akers, Deans, and Crosser (Reference 5-4).

A e.t y3 h = . o2(. /94 Pr (5-2) h where Pe, = (G,+G( y (5-3)

This C-E model is compared to the unified theory of heat transfer across a falling liquid film by Dukler (Reference 5-5) and the water data of Carpenter (Reference 5-6) in Figure 5-1. For the BE model, a factor of 1.5 has been applied to Equation (5-2) to bring the predicted heat transfer coefficients  ;

1 into better agreement with these two data sets. '

l 5-4

The Reynolds number, given in Equation (5-3), defines the " equivalent" single-phase (liquid) flow, which would produce the same shear force at the liquid film surface as does the flowing vapor in the true two-phase annuiu flow regime. In order to apply Equation (5-3) to the two-phase column of fluid in the U-tubes during countercurrent flooding or refluxing when downward liquid flows are small and the flow cegime is bubbly or slug-like, the equivalent shear stress method is also applied to the liquid continuum region of the U-tube. Equation (5-3) is rewritten as follows: .

l u

Re e = (G,4 G,(f)*( J) (5-4) l .

1 where b' = l & Coo ('l- 6) (Reference 5-3) (5-5) b Equation (5-5) limits this BE model to the flow regime conditions and void fractions of Test S-UT-8 and therefore is not generally applicable withcut further specification for other flow regimes. Also, heat transfer coefficients derived by using Equations (5-2) and (5-4) in the two-phase +

region are limited to a minimum'value based on convection to liquid for the average flow conditions of Test S-UT-8.

The overall primary-to-secondary heat transfer by condensation is determined by applying Equation (5-2) both above the two-phase level (with Equation 5-3) and below the two-phase level (with Equation 5-4). The appropriate level dependent U-tube surface area is utilized in each region.

5-5

Condensation of Bubbles in Steam Generator U-Tubes When condensation fluid conditions exist in the U-tubes, condensate in the CEFLASH-4AS Evaluation Model is produced from the steam region of the node. -

This condition is appropriate prior to the influx of large quantities of steam into the U-tubes from the hot leg. With 'large steam flows into the U-tubes producing countercurrent flooding, condensate should also be produced by condensing bubbles in the two-phase region. This BE model modification utilizes the improved condensation heat transfer model described above to calculate the mass of bubbles condensed:

my = K, ( (1 ) Q g,3 [ hg (5-6)

Since the CEFLASH-4AS model assumes that heat is removed from the U-tubes uniformly, the coefficient, K,, is used to represent flow regime dependent non-uniformities. For application to Test S-UT-8, the coefficient was spec-ified as unity, in recognition of high mixing and turbulance during flooding and bubbly flow. Therefore, Equation (5-6) is not generally applicable without further specification of K, for other flow regimes.

The combined influence of the changes to the condensation heat transfer model (BE condensation heat transfer and bubble convection) is illustrated in Figure 5-2, which shows the integral condensate produced in the intact loop steam generator U-tubes (upflow side). The improved BE condensation models produce 25% more total condensate than the previous model. Most of the increase in condensate occurs early in the transient when primary-to-secondary temperature differences are the greatest. During and after the time period of 5-6

countercurrent flooding, the rate of condensate produced (i.e. the slope of

\

, the integral mass curve) is not significantly different for the two models, about 0.1 lbm/sec or less. It should be noted that this calculated rate of condensate production during and after flooding is an order of magnitude less than the calculated rate of liquid carry-over from the hot legs into the U-tubes, about 3 lbm/sec. Thus, steam generator liquid accumulation in Test S-UT-8 is more dependent on liquid carry-over from the hot legs than on condensation in the U-tubes.

BE Bubble Convection in U-Tubes The BE bubble convection and disengagement model was developed for use with LOCA transients in which the reactor coolant pumps were allowed to continue running (Reference 5-7). In this model prior to pump head degradation, the homogeneous node quality is used in the calculation of the mass of bubbles convected from one node to another. After pump head degradation, the model switches to the heterogeneous node quality. The switch in node quality occurs when the upward fluid velocity in the U-tubes drops below the calculated bubble rise velocity. For analysis of Test S-UT-8, a modification was made to allow application of this BE model to the steam generator upside U-tube nodes alone. The use of this model provides for a more realistic prediction of

~

liquid carry-over at the top of the U-tubes prior to pump head degradation.

The CEFLASH-4AS input description requires that the path connecting the upflow side and the downflow side of the steam generator U-tubes be given a heterogeneous path height or window for coolant carry-over. If the two-phase level is above the top of the window, the two-phase mixture is carried over.

6-7 I

i If the two-phase level is below the bottom of the window, steam is carried over. For two-phase. levels occurring within the window, carry-over is weighted in proportion to the fluid height. The generic NSSS model for specifying this window height uses the diameter of a circular path with the -

area of the combined areas of all the tubes. For Semiscale with six U-tubes, this would be a height of only 0.16 ft, which results in an underprediction of liquid / bubble convection at the top of the U-tubes. Therefore, for this Semiscale analysis (both BE and BE/EM analyses), the path height at the top of the U-tubes was specified to cover the maximum and minimum heights of the U-tubes or 2.37 ft. The cross-sectional area of the path is unchanged. This results in a more accurate prediction of two-phase convection at the top of the U-tubes prior to pump head degradation.

5.2 Modification to Address Core Rewet Steam Production During the core uncovery period prior to loop seal clearing in Test S-UT-8, the cladding thermocouple data for the upper approximately four feet of the heater rods showed no cladding heatup. This is exhibited by Figure 5-3 which -

compares the peak cladding temperatures at various elevations for time periods before and after loop seal clearing. The Figure illustrates that the peak temperature profile after loop seal clearing exhibits cladding heatup in the upper parts of the heater rod, but the data before loop seal clearing does -

not. Rewet of the heater rods by liquid flowing back into the vessel from the hot leg is probably the explanation for this observation.

Rewet of the heater rods impacts the system response through the additional.

steam produced by vaporization. Additional steam production prolongs core uncovery relative to a system that otherwise would simply stop making steam 5-8

. _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ ___J

L 1

due to uncovery of the core. Use cf the rewet model in the BE analysis, ,

resulted in the accurate prediction of the duration of core uncovery and produced the observed bulk flow of steam from the vessel into the downcomer l- prior to loop seal clearing. The rewet model was not used in the BE/EM l

analysis because core rewet steam production is not part of the EM.

l The core rewet steam production model' uniformly removes heat from the upper L

half of the heater rods. A fixed fraction of liquid returning to the vessel from the hot leg is vaporized by heat from this portion of the core. The

' total integrated heat removed from the core is based on a simple energy balance between the heater rod and the coolant.

5.3 Representation of Semiscale Design Features The best estimate models in CEFLASH-4AS were primarily developed for analysis  ;

of C-E designed NSSS's. Although the majority of these models are applicable to the Semiscale design, additional models for certain Semiscale design features were needed. These unique design features pertain to the Semiscale vessel, downcomer, hot legs and lower plenum.

In Semiscale, the vessel and downcomer are separate components and, therefore, vessel wall heat is only added directly to the core region. Alto, the downcomer is circular in cross-section for the majority of its length and not annular. These differences affect the bubble release rate models in the downcomer and vessel. The unique vertical section of hot leg in Semiscale, the " pant leg", strongly affects the countercurrent flow frictional losses during steam generator draining. And the unique lower plenum cross-section 5-9 W*e**

1 I

design in Semiscale affects the depth of liquid level in the vessel during an extremely ~ deep level depression. Each of these aspects of the Semiscale design are addressed in the three model modifications discussed below.

1 l These model modifications representing unique Semiscale design features were used in both the BE and the BE/EM analyses. These model modifications do not i l apply to small break LOCA evaluation of C-E designed NSSS's.

Modified Bubble Release Rate in Vessel and Downcomer During Test S-UT-8, the core uncovery spike prior to loop seal clearing was deep enough that bulk steam from the vessel entered the downcomer. The lower plenum densitometer reading, shown in Figure 5-4, clearly exhibits a sudden drop in mixture density at 210 seconds due to the extensive inventory loss from the vessel. Bulk steam rises to the top of the downcomer in a slug-like manner, displacing liquid from the downcomer into the cold legs. The collapsed liquid level data in the downcomer, also shown in Figure 5-4, shows this downcomer inventory loss. Liquid displaced from the downcomer begins i refilling the loop seal thus delaying the clearing process and prolonging core uncovery. Lower liquid levels in the downcomer also result in a lower core recovery after loop seal clearing. ,

In order to simulate bulk flow of steam into the downcomer, the CEFLASH-4AS

]

models for bubble release rate in the lower plenum and downcomer were mod-ified. In the lower plenum of the CEFLASH-4AS model, when the path to the downcomer uncovers, the bubbles produced by lower plenum wall heat are released quickly to the steam region of the vessel to simulate the ,

J 5-10

t reverse flow and suction of steam and/or two-phase into the downcomer. In the downcomer, when bulk steam enters at the bottom, the bubble release rate is lowered to account for the slug-like passage of steam in the circular

~'

cross-section vertical pipe which carries inventory out of the downcomer.

In the analysis of Test S-VT-8 after loop seal clearing, bubbles produced by the lateral wall heat in-the Semiscale vessel are released quickly to the steam region.- This change modifies the CEFLASH-4AS EM model which places these bubbles in the. lower plenum where most of the wall heat is assumed to

exist. This EM assumption is valid for modeling an NSSS which does not have a separate vessel and downcomer, but is not accurate for Semiscale where the vessel wall heat adds directly to the-fluid in the core region. The modification to the vessel bubble release rate improves the two-phase level prediction in the vessel.

Modified Two-Phase Frictional Losses in Hot Leg The key geometric differences between the Semiscale and the C-E NSSS hot leg designs are illustrated in the following sketch:

• C-E NSSS SEMISCALE M00-2A

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Apart from the Semiscale hot leg inside diameter being more than a factor of- )

twenty smaller than the one for a C-E NSSS, the Semiscale hot leg contains a six foot vertical section, called the " pant leg", which connects the j

horizontal portion of the hot leg to the steam generator inlet plenum. The .{

CEFLASH-4AS dual flow path model for countercurrent flow in the hot leg was not developed to model vertical countercurrent flow or countercurrent flow in ~

q small diameter horizontal pipes. However, the model can be applied to Semiscale with a minor modification to the representation of the frictional losses.

The two-phase frictional losses during countercurrent flow in the hot leg were modified in the CEFLASH-4AS model to account for an additional loss due to interfacial drag in the small diameter Semiscale pipes. .For those flow paths containing the vertical pant-leg, the interfacial friction correlation of Bharathan and Wallis was used (Reference 5-3) to calculate a friction factor for the average vertical countercurrent flow conditions of Test S-UT-8.

Similarly, for those flow paths containing the horizontal part of the hot leg, the Wallis correlation for stratified countercurrent horizontal flow was used (Reference 5-2) to calculate a friction factor for the average horizontal countercurrent flow conditions of Test S-VT-8.

The effect of this model change is illustrated in Figure 5-5 which compares. _

the draining rates for the steam generator U-tubes. The use of modified frictional losses increases the steam generator draining time by roughly 40 seconds and brings the prediction in closer agreement with the test data.

5-12

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Variable Area Vessel Lower Plenum I

This model change incorp) rates a Semiscale specific geometry modification to represent multiple cross-sectional areas in the vessel lower plenum. This improves the calculation of the liquid level in the lower plenum when the core ,

5 , I is fully uncovered. (This geometry modification is illustrated in' ,j j Figure 5-6). The Semisca'te Tower plenum is modeled with two cross-sectional i

areas, preserving height and volume. , ,

I 5.4 Changes to Eliminate Numerical Difficulties l Even though Semiscale is a scaled system with very small fluid' volumes, the 1

heat transfer rates and fluid velocities are of the same order of magnitude as the full-scale system. This combination cf thermal-hydraulic conditions 'C created minor numerical difficulties in CEFLASH-4AS. The changes made to eliminate these numerical difficulties are described below. The I

implementation of these changes did not affect the calculated results in either the BE or the BE/EM analyses. These code changes do not apply to the analysis of C-E designed NSSS's.

I Stabilize Steam Generator Inlet Flow Path Quality This model modification eliminates CEFLASH-4AS numerical instabilities during I later stages.of steam generator draining when the node quality is predicted to be greater than 0.5. The revised model switches from heterogeneous.to homogeneous quality definitions in the flow paths at the inlet to the steam generators to avoid " overtraining" of the node.

5-13 l

l l _ . .

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Stabilize Wall Heat in the Hot Legs for High Ouality Conditions u .O

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This model.r. modification eliminatesi aoother source of numerical instability during later stages of het leg draininti'when node quality is calculated to be .

t+

greater than 0.5. The revised model sw' itches to level dependent wall heat J

', in the hot legs and, in addition, adds wall heat to the steam region of the ,

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node to avofd " overheating" of the node.

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7 l ReferencestfGe Section 5.0 ,

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y 5-1 CEN-203-P, Revision 1-P, " Response to NRC Action PTan It.em II.K 3.30, Justification of Small fqeak LOCA Methodt.," March, 1982.

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5-2 G. Wallis, "One Dimensiorad,bi.. Phase Flow," McGraw-Hill Book Company 9 .;

(1969)!.} j i

l' ,

• 5-3 NUREG/CR-3060, "A Critici Review of the FlooMag Literature," by s

Bankoff a Lee (Nor thweytera University), July,1983. <

,r ,

'l 5-4 Mers, W. W., Deans, H. L and Crosser ( 0.K., " Condensing Heat Trans.fer Nithin Horizontal Tubes," Proc. 2nd Nat. Heat Transfer Conf.tASME/AI_CHE, hgust, 1958.

( r 5-5 ' Dukler, A.E., "Viuid 11echenf cs and Heat Transfer in Vertical Falling-

' FNm Systems ," Chem. E' rig.~ Prog. Sym. Series , No. 30, Vol . 56,1960.

o

/ 5-6 Carpenter,F.G.,Ph.D.IN(is,'U.OfDelaware,1984.

t

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5-7 CEN-115-P, "F.esponse to VaC IE Bulletin 19-06C, Item 12 ar.d 3 for l

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Combustion Engineering Nuclear f.tum Supply Systems," August,1979.

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6.0 RESULTS OF BE ANALYSIS The results of the BE post-test an61ysis of Semiscale Test S-UT-8 are presented in this section. Comparisons are made between the measurements of

. Test S-VT-8 and the predictions of the BE version of CEFLASH-4AS. This analysis includes all the model modifications for Semiscale described in Section 5.0. The analysis time-frame covers 500 transient seconds. This-extends the CEFLASH-4AS calculations beyond the time of pre-loop seal clearing core uncovery and includes the boiloff period up to the. time of accumulator discharge.

6.1 Comparison of BE Analysis Results and Test Data Pressure Figure 6-l' compares the upper-plenum pressure history calctisted by the BE model analysis with the measured val'ues from Test S-UT-8. The predicted pressure compares very well with the data for most of the transient.

This overall good agreement indicates that the system boundary conditions are accurately represented in CEFLASH-4AS, particularly the modeling of break flow, steam generator heat transfer, wall heat and core power. The accurate prediction of the pressure plateau prior to loop seal clearing (i.e., the time period between 100 seconds and 240 seconds) also indicates that the improved BE condensation heat transfer model for the steam generator U-tubes, described in Section 5.1, is appropriate for Semiscale Test S-UT-8.

Figures 6-2 and 6-3 compare the pressures in the secondary side of the intact and broken loop steam generators, respectively. The predicted pressure 6-1

responses at the time of secondary side isolation (20 seconds), as well as the predicted peak pressures, compare well with the data for both steam generators. This indicates that the steam generator heat transfer models and

~

the representation of the secondary side region in CEFLASH-4AS are adequate -

for the analysis of. Test S-UT-8. The gradual decrease in secondary side

-pressure after 40 seconds is modeled by allowing a small leakage of steam from each steam generator. The leakage rate averages to about 2% of the initial steam flow for the remainder of the transient.

Vessel Co11aosed Liquid Level The collapsed liquid levels in the vessel are compared in Figure 6-4. The BE model prediction of the timing, depth, and duration of the core liquid level depression prior to loop seal clearing at 240 seconds, is excellent. In the test, the rate of inventory loss from the vessel prior to loop seal clearing is dependent on steam production in the core and the displacement of fluid from the vessel into the downcomer. The good agreement between predicted and measured rates of vessel inventory loss indicates that core heat transfer, core rewet steam production, steam generator liquid holdup, and loop resistances to steam flow are all adequately represented by the BE models in CEFLASH-4AS. The good prediction of collapsed liquid level below the bottom elevation of the core is the result of the modified lower plenum geometry model described in Section 5.3. The accurate prediction of the duration of core uncovery is the result of the core rewet -

steam production model described in Section 5.2.

The abrupt recovery in collapsed level starting at 240 seconds, is caused by loop seal clearing which significantly reduces the resistance to steam venting to the break. The BE model predicts the timing of loop seal clearing very accurately. This is due in part to the model modifications described in 6-2

b Section 5.3 for representing the bulk ficw of steam from the lower plenum into the downcomer. Also, the collapsed level in the vessel during this recovery period js predicted accurately, thus indicating that the distribution of liquid between the hot legs and the downcomer (i.e., the liquid sources during core recovery) was properly represented by CEFLASH-4AS.

The gradual decrease in vessel inventory after 270 seconds is the result of liquid boiloff in the core region and insufficient liquid make-up by the high pressure safety injection. Again.the predicted level by CEFLASH-4AS is close to the data.

Upper Head Collapsed Liquid Level Figure 6-5 compares the collapsed liquid I?vels in the vessel upper head. The predicted draining of liquid from the upper head agrees very well with the data. This good agreement indicates the i

Capability of the.CEFLASH-4AS BE model to adequately simulate the upper head l

draining characteristics of Test S-UT-8. This also verifies the code input {

1 specification of the loss coefficients for the core bypass line, guide tube, j and support columns. The representation of the bypass flow of steam from the I vessel to the downcomer is important for correctly predicting steam generator liquid holdup and core uncovery prior to clearing of the loop seal.

I l

Steam Generator Collapsed Liquid Level Figure 6-6 compares the collapsed liquid levels in the upflow side of the U-tubes of the intact loop steam generator. The BE model prediction of the timing and extent of liquid accumulation starting at 50 seconds compares very well with the data. This indicates that the improved steam generator component models for countercurrent flooding and condensation heat transfer described in 6-3 i

l Section 5.1, along with the dual flow path model for the hot' legs which allows for the carry-over of liquid from the hot leg into the U-tubes, are adequate for calculating steam generator liquid holdup in Test S-UT-8. Liquid draining from the U-tubes beginning at 100 seconds marks the end of countercurrent -

flooding, which is accurately predicted using the deflooding model described l

~

in Section 5.1. After 100 seconds, the predicted draining rate agrees very well with the data due to the model modification for countercurrent two-phase interfacial friction in the hot leg described in Section 5.3.

Figure 6-7 compares the collapsed liquid levels in the downflow side of the U-tubes of the intact loop steam generator. The prediction of the overall trend of the data is good. The accumulation of liquid in the U-tubes starting at 50 seconds is due to the prediction of condensation of steam and by the momentary carry-over of liquid from the pump suction loop seal node. The loop seal piping has atypically large wall heat addition to the coolant from the external guard heater, which, together with the steam generator outlet plenum walls, creates a steam flow high enough to carry liquid into the U-tubes for a short time after pump head degradation. Starting at 60 seconds the accumulation of liquid in the U-tubes is predicted to be less than the data due to early stagnation of the upward flow into the U-tubes and initiation of draining into the suction leg. The close correspondence of predicted collapsed level to data after 120 seconds is important for obtaining the -

observed vessel level depression due to steam generator liquid holdup in Test S-UT-8..

6-4

Suction Leg Collapsed Liquid Level Figure 6-8 compares the collapsed liquid levels in the intact loop pump suction leg, steam generator outlet side.

After the steam generator downside U-tubes have fully drained at 180 seconds, the predicted collapsed level agrees very accurately with the data. This verifies that the CEFLASH-4AS BE model correctly predicted the total mass of liquid in the downflow side of the U-tubes and loop s'eal and insures an accurate prediction of the timing of loop seal clearing. ~ Prior to 180 seconds differences between the predicted and measured levels show the effects of i differences in liquid and steam separation (" layer-caking") in the U-tubes and j the suction leg.

4 Beginning at 210 seconds, the collapsed liquid level increase seen in Figure 6-8 is produced by the bulk flow of steam from the vessel to the  !

downcomer. As described in Section 5.3, the passage of a slug of steam in the downcomer displaces liquid from the downcomer, reverses the flow in the cold leg, and partially refills the loop seal. Accurately predicting this system response is important to correctly predicting the timing of loop seal clearing.

Figure 5-9 compares the collapsed liquid levels in the intact loop suction leg, pump inlet side. The CEFLASH-4AS BE model correctly predicts the timing and extent of liquid clearing from the loop seal at 240 seconds. This indicates that the CEFLASH-4AS models adequately represented the liquid

' distribution. in the system prior to clearing and also predicted the correct steam flow needed to partially clear the liquid from the vertical loop seal pipe.

6-5

Downcomer Collapsed Liquid Level Figure 6-10 compares the collapsed liquid levels in the downcomer. The BE analysis correctly predicts the trend and magnitude of the measured downcomer liquid level. As described in Section 5.3, the NSSS BE models were not equipped to model the bulk flow of -

steam into the downcomer from the vessel during Test S-VT-8. As Figure 6-10 indicates, the modifications made to model this S-UT-8 characteristic are adequate. This good agreement with the measured level in the downcomer is important for predicting the observed core recovery after the loop seal clears at 240 seconds.

Vessel Mixture Level Figure 6-11 compares the mixture levels in the vessel.

The data points for mixture level were obtained by examination of the time of heatup of cladding thermocouple at various heater rod elevations (ranging from 4.1 ft. to 14.5 ft. relative to the bottom of the vessel). The CEFLASH-4AS BE model accurately predicts the timing, depth, and duration of the mixture level transient prior to the time of loop seal clearing at 240 seconds. After 240 seconds, the liquid boiloff portion of the test was

.i adequately predicted by CEFLASH-4AS.

s 6.2 BE Analysis Conclusions The overall agreement between the BE analysis predictions and Test S-VT-8 data -

is quite good, thus benchmarking the S-VT-8 best estimate C-E small break LOCA model. The BE model, with the modifications to represent Semiscale incor-porated, is capable of accurately predicting the effects of steam generator liquid holdup on core uncovery prior to clearing of the loop seals.

6-6

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TIME IN SEC 6-11

\

FIGURE 6-6 l BE AllALYSIS Of SEf!ISCALE TEST S-UT-0 liiTACT LOOP STEAfi GEllERATOR U-TUBE UPFL01: SIDE COLLAPSED LICUID LEVEL BE 10 DEL PREDICTION

---

S-UT-8 DATA 36.000

! l I

, l l

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TIME IN SEC l l

6-12

i 1 l .  !

1 l

i, FIGURE E-7

BE AilALYSIS OF SEHISCALE TEST S-LT-E liiTACT LOOP STEAM GEt'ERATOR U-TUBE D0\lllFL0tl SIDE COLLAPSED LICUID LEVEL 1

I I- BE f0 DEL PREDICTION S-UT-8 DATA 3S.000 ,

! l I

l l i

30.000 '

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• i O O C O O O C O O O C l O - N m v  :.O TIME IN SEC l l

6-13 I

]

)

FIGURE f>-C BE AllALYSIS OF SEMISCALE TEST S-UT-3 lilTACT LOOP PUMP SUCTI0li LEG D01lI!FL01: SIDE COLLAPSED LICUID LEVEL BE 10 DEL PREDl %^N

---

S-UT-8 DATA -

18.000 15.000 \l '

's 12.000 , y j t

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O o o o o o o a S E E S TIME IN SEC 6-14 .

~~~

FIGURE 6-9 BE AliALYSIS OF SEf!ISCALE TEST S-UT-8 IllTACT LOOP PUMP SUCTI0il LEG UPFLOW SIDE COLLAPSED LIQUID LEVEL BE MODEL PREDICTION

-- S-UT-8 DATA 18.C00 I l. l I

j l 15.000 '

i l

i i

l 12.000 '

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t_ _ _ - - - _- - -- --------_-------- _ O

l 1

FIGURE E-10 '

BE AllALYSIS OF SEtllSCALE TEST S-UT-S j D0llliC0flER COLLAPSED LICUID LEVEL BE 10 DEL PREDICTION

---

S-UT-8 DATA 30.000 , , ,

25.000  ; i 20.000 I

[ -

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y 15 000  ;

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l l

I FIGURE 6-11 BE AllALYSIS OF SENISCALE TEST S-UT-8 VESSEL TU0-PHASE F1IXTURE LEVEL BE tiODEL PREDICTION

---

S-UT-8 IATA 30.000 l

l 25.000 20.000 -- -

s L.o u

15 000 "

,\ " "l g . ,

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6-17 w-

7.0 RESULTS OF BE/EM ANALYSIS The results of the BE/EM post-test analysis of Semiscale Test S-VT-8 are presented in this section. Comparisons are made between the measurements of Test S-UT-8 and the predictions of the BE/EM version of CEFLASH-4AS. The BE/EM model is described in Section 7.1 and the results are discussed in Section 7.2. Like the BE analysis, the BE/EM analysis time-frame covers 500 transient seconds, which includes pre-loop seal clearing core uncovery and core boiloff up to the time of accumulator discharge.

7.1 BE Component Model Replacements for BE/EM Analysis The BE analysis results, presented in Section 6.0, benchmark the CEFLASH-4AS code against Test S-UT-8 data using options and component models designed to give optimum agreement. However, the post-test analysis of Test S-UT-8 is also being used to validate the EM component models important for steam

]

generator liquid holdup. Therefore, in this second phase of the post-test analysis, selected BE component models, which are important for the calculation of core uncovery due to steam generator liquid holdup, are replaced by their EM component model counterparts or are climinated if they are not part of the EM. The purpose of this combined best estimate / evaluation model (BE/EM) analysis is to demonstrate the adequacy of these selected EM component models to predict Test S-UT-8 type core uncovery. The list of component model replacements used in the BE/EM analysis is as follows:  !

7-1 i

1. Flooding in steam generator U-tubes not used

2. BE' condensation heat transfer in U-tubes replaced by EM condensation heat transfer

3. Condensation of bubbles in U-tubes not used

4. BE bubble convection in U-tubes replaced by EM bubble convection

5. Core rewet steam production not used

6. BE vessel model replaced by-EM vessel model:

a. 'Two node vessel replaced by single node

b. Upper head to downcomer bypass not used

c. BE wall heat in vessel replaced by EM wall heat.  !

7. Non-equilibrium modeling during HPSI replaced by equilibrium i modeling The first four model replacements in the above list are the EM component model counterparts to the BE steam generator model improvements described in Section 5.1. . The adequacy of these EM steam generator models, particularly.as they relate to the representation of the effects of liquid holdup on' core uncovery is the principal focus of the BE/EM analysis of Test S-UT-8. The impact of each of these model- replacements in the BE/EM analysis will be discussed in Section 7.2.

i The fifth model replacement in the list above, not using core rewet steam *l 1

production, is consistent with the C-E Small Break Evaluation Model. In the j BE analysis,.the core rewet. steam production model was effective in prolonging core uncovery prior to loop seal clearing by forcing bulk steam into the 7-2 l

1

rX

.\ r downcomer and partially refilling the loop seal. Elimination of this component model-from the BE/EM analysis means that these aspects;of; ,

Test S-UT-8 will not be predicted.'. Also, without the'@ ore.rewet steam (y

production, the BE/EM analysis prediction of stear . flow produces more gradual *

.4 loop seal clearing than observed in the test data which leads to prolonged core recovery rather than abrupt recovery as seen in the test 'due to loop seal clearing. t

?

l s k l

The vessel noding change for the BE/EM analysis, the sixth'model replacement, 1 '

s c l.L was described in Section 4.3. The key feature of this change is the p3 elimination of core bypass which means all steam produced in the core must .'^

flow through the steam generators to reach the break. Theeffectofcorehil bypass has been experimentally examined in Semiscale (Reference 7-1). Tests with more core bypass flow than Test S-UT-8 have exhibited significantly lower (

inventory loss prior to clearing of the loop seals. Thus, the elimination of bypass in the BE/EM analysis should tend to increase,the predicted rate of j inventory loss based on the experimental" evidence.

l .

Replacement of the BE vessel model with the EM vessel model also means imple- l l

mentation of the EM wall heat model described in Reference 7-2. The EM wall heat model tends to produce more wall to coolant heat tran!;fer in the BE/EM i- analysis than in the BE analysis.

Equilibrium modeling in the cold legs during high pressure safety injection, the seventh model replacement for the BE/EM analysis, has a minor effect on downcomer and loop seal liquid levels prior to loop seal clearing. However, the magnitude of HPSI delivery in Test S-VT-8 is negligibly small and 7-3

- _ _ - - - _ - . _ _ _ _ _ _ _ _ - - _ _ _ _ - _ _ _ _ _ . _ _ _ _ _ . - . _ _ . - - _ - - _ _ _ - =

ge ygy y 3

- g4 % 'o 0 4 *, 9 cm ( ('

e therefore thjs model replacement is of little consequence to the differential  ;

4 GJ;/ .

pressure,between

,s .

the hot side and the cold side of the system acting to I depress the core liquid level.

7.2 Comparison of BE/EM Analysis Results and Test Data Pressure The upper plenum pressures are compared in Figure 7-1. The use of theEM'condensationheattransfermodelintbsteamgeneratorU-tubesleads to a 50 psia higher predict" ion of primary side pressure than measured in Test S-VT-8. As described in References 7-2 and 7-3, the EM condensation heat transfer model deliberately underpredicts the heat transfer coefficient, which y leads to a higher primary pm.sure. This .in turn yields a conservatively high rediction of' inventory pr.s out the break and a conservatively low high gpressure safety injection flow.

• E 4 x- ,

.p -

37he secondary side preisures for both steam generators are compared in Figures 7.2 and 7.- 3. Since the overall steam GWerator heat removal from the primary to the secondary side is not significantly affected by the steam generator component model replacements, the BE/EM analysis results show the same good agreement with the measured secondary side pressures as the BE analysis.

Vessel Collapsed Liquid Leve,1, The collapsed liquid levels in the vessel are i9 compared in Figure 7-4. In general, the BE/EM analysis provides adequate agreement with the observed trend of the vessel inventory. Due to the absence  !

of modeling core bypass flow, the timing of core uncovery is predicted earlier by the BE/EM analysis and the predicted rate of vessel inventory loss is y

7-4

,j j3 g' ,P

, * /

,y '

/  : ,

6, e 7 s

^! ,

/ greater than the Test-5-UT-8 data. Also,theBE/ENanalysisI predicts complete *

\y +

core uncovery orier to loop seal clearing as observed in tiis test. . Therefore,

, in spite of. changes in steam generator behavior produced- by the replalemhnt of

. BE steem ordrator component models with EM compcNat ithde?s) the BE/EM 3( .

/ > +

,,.  !\

analysis stfl1 produces the, extreme l core uncovery observed ifTast S-UT-8.

i (The changes'in steam generator thermal-hydraulic behnior bill be discussed

(

,i, in the next 'section.) Without the representation of core f ewet steam

,b j

I production, the dE/EM analysis does not: produce the bulk flow of steam from i

'}

l the vessel into the downcomer and thhofore, does not predict the duration of

.s i complete' sore uncovery as seen in the test data. The rate of c' ore recovery at '

I170secondsintheBE/EManalysis:isslowertnanthetestd{adJetoa <

prediction of more gradual icop sesig clearing. The recovery level is significantly lower than the data at'210 ' seconds which leads to conservittively i

\

lower lev 61s during the boiloff petiod beyond 270 seconds. This BE/EM '

prediction of greater vessel inventory loss is related to the prediction of 3, . lower liquid levels in the downcomer, which are discussed below.

1

,, Steam Gdr.erator Collapsed Liquid Level' The collapsed liquid levels in the upflow side of the intact loop steam generator U-tubes are compared in Figure 7-5. The use of the EM steam generator component mode's without countercurrent flooding and bubble condensation in the BE/EM analysis produces significantly lower liquid levels than measured during the t"t and no liquid L

g accumulation. Tht. dual flow path model is still carrying liquid into the U-tubes fro'm- the hot leg as it; the BE analysis, however, in the absence of countercurrent flow limited by f moding, the liquid drains out as quickly as it is carried in, limited only by the resistance created by countercurrent 1 flow interfacial drag in the hot leg. '

7-5

l 8

Figure 7-6 shows the collapsed liquid levels in the downflow side of the inter.t loop steam generator U-tubes. The prediction of liquid level in the

. BE/Ed ana' lysis is significantly lower than the test data due to continuous '

draining of the condensate from the U-tubes into the pump suction loop seal, i

)

Unlike the BE analysis which predicted a momentary carry-over of liquid into 1 j

I the U-tubes from the loop seal, the BE/EM analysis predicts earlier uncovery L of the steam generator outlet plenum walls than the BE analysis which results in less wall heat transfer to the coolant, less steam production, no upward I flow of liquid into the U-tubes, and reduced liquid accumulation in the U-tubes.

[

^ In spite of'the low prediction of steam generator U-tube liquid levels in the BE/EM analysis, the EM component models for steam generator behavior still 1

produce the effects of liquid holdup on core uncovery prior to loop seal clearing, especially the depth of core uncovery. First, the presence of

1. liquid in the U-tubes still restricts or resists the flow of steam to the

- 4 break which adds to the differential pressure between the core and downcomer acting to depress the level in the core. Second, the hydrostatic pressure differential produced by liquid held in the U-tubes is dependent on the difference in liquid level between the upside and downside and not on the i

level in the upside alone. In this regard, the BE/EM analysis yields as much hydrostatic pressure differential between the core and downcomer due to liquid holdup as the test. In view of thesc two aspects related to the effects of liquid holdup, it is not surprising that the BE/EM analysis produces the same extreme extent of core uncovery prior to loop seal clearing as observed in Test S-UT-8.

7-6 .

l 1

- _ - - _ - _ _ _ _ _ _ _ _ . _ - - _ _ _ _ k

Suction Leg Collapsed Liquid Level The collapsed liquid levels in the steam h generator outlet portion of the suction leg are compared in Figure 7-7. After 120 seconds, the BE/EM analysis predicts lower liquid levels than observed in Test S-UT-8 due to the lack of liquid accumulation in the downflow side of the U-tubes. Also, the reverse flow of liquid back into this portion of the loop seal at 210 seconds pr'oduced by the flow of bulk steam into the downcomer from the vessel is not predicted due to the absence of core rewet steam production in the BE/EM analysis.

Figure 7-8 shows the collapsed liquid levels in the pump inlet portion of the suction leg. The BE/EM analysis predicts significantly earlier loop seal clearing starting at 170 seconds than observed in Test S-UT-8 (240 seconds). ,

Less total steam flow due to the absence of the core rewet steam production model leads to a prediction of a more gradual loop seal clearing process in the BE/EM analysis. This produces the reduced rate of core recovery prior to 270 seconds seen in Figure 7-4.

l Downcomer Collapsed Liquid Level The collapsed levels in the downcomer are 1

compared in Figure 7-9. The BE/EM analysis underpredicts the liquid levels j measured in Test S-UT-8 due to less inventory transfer from the loop seal to the downcomer because of the reduction of liquid accumulation in the downflow

~

side of the U-tubes. Underpredicting the downtomer level in the BE/EM analysis leads directly to lower liquid levels in the vessel during the recovery phase and the subsequent boiloff phase of the test.

7-7

I Vessel Mixture Level The vessel two-phase mixture levels are compared in Figure 7-10. The BE/EM analysis results show acceptable agreement with the 1 Test S-VT-8 data. The trend of the predicted core uncovery transient compares well with the test observation. The depth and timing of the. predicted mixture -

level are the result of the same physical arguments described above for the vessel collapsed liquid level, i.e., the effects of steam generator liquid holdup, no bypass flow, and no core rewet steam production. After. loop seal clearing, .the core mixture level recovery and subsequent boiloff beginning at 280 seconds are predicted conservatively low by the BE/EM analysis due primarily to the lower liquid levels in the downcomer.

Cladding Temperature Using the CEFLASH-4AS prediction of vessel two-phase mixture level as a forced boundary condition, the C-E Evaluation Model hot rod heatup code, PARCH (Reference 7-4), was run for the BE/EM analysis. The predicted peak cladding temperatures'are compared to asured cladding temperatures of the heater rods in Figure 7-11. The results shown are the peak temperatures versus elevation in the core for two time periods: prior to loop seal clearing and after loop seal clearing. The comparison shows that the BE/ZM analysis leads to 300 F higher predicted peak temperature than the test measurements before loop seal clearing and roughly 500*F higher peak temperatures after loop seal clearing because of the underprediction of the two-phase mixture level and conservatism of the heatup code. -

7-8

7.3 BE/EM Analysis Conclusions The BE/EM analysis results show acceptable agreement with the collapsed water

< level data in the vessel. The BE/EM analysis adequately predicts the trend of l

the core uncovery data both before loop seal clearing and later during the subsequent core inventory boiloff period. Before loop seal clearing, the depth of the vessel liquid level depression is adequately predicted by the BE/EM analysis, and also the duration of core uncovery for the upper portion of the heater rods is adequately represented. After loop seal clearing, the BE/EM analysis conservatively underpredicts the vessel liquid level recovery.

This leads to a conservatively low prediction of the vessel inventory during the subsequent core boiloff period. The BE/EM analysis results also produce conservatively nich predictions of cladding temperature during both the pre-loop seal clearing core uncovery period and the later core boiloff period.

These BE/EM analysis results validate the acceptability of the EM component models important for steam generator liquid holdup and core uncovery to predict S-UT-8 type vessel liquid level depressions and to produce conservative peak cladding temperatures for licensing calculations.

7-9

i Reference for Section 7.0 7-1 EGG-SEMI-6010. " Vessel Coolant Mass Depletion During a Small Break i

LOCA," September, 1982. -

7-2 CEN-203-P, Revision 1-P, " Response to NRC Action Plan Item II.K.3.30, Justification of Small Break LOCA Methods," March, 1982.

7-3 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, Revision 1-P," November, 1984.

7-4 CENPD-138, " PARCH, A FORTRAN-IV Digital Program to Evaluate Pool Boiling, Axial Rod and Coolant Heatup," August, 1974.

Supplement 1, February 1975.

Supplement 2, January 1977.

7-10

______- -_-_ - i

L l

l l

FIsuRe 7-1 i EE/Efi AllALYSIS OF SENISCALE TEST S-UT-S VESSEL UPPER PLENUFi PRESSURE BE/EM PREDICTION S-UT-8 DATA 2400.C

! l l  ;

i B

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j 2000.c  :

i i

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TIME IN SEC i

- . . . _ . .. i 7-11

I FIGURE 7-2 i BE/Eii Ai1ALYSIS OF SEllISCALE TEST S-UT-S IllTACT LOOP STEAJi GEliERATOR SEC0ilDARY SIDE PRESSURE .

BE/81 PREDICTION .

S-UT-8 DATA 1200.0 , ,

i j i 20C0.0 ,

lbl , _w - ~_._  : ,

l l -

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I  : i 800.0 ,

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l TIME IN SEC '

l 7-12 I

FIGURE 7-3 t

BE/Eii AilALYSIS OF SElilSCALE TEST S-UT-E BROKEfi LOOP STEN 1 GEliERATOR SEC0liDARY SIDE PRESSURE BE/EMPREDICTION

. --- S-UT-8 DATA 1200 0 j 3  ;

i

.  ; i l I l

' l i i l l  ! j t i

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O o o o o o O o o o a . 1 o - a m e a i TIME IN SEE -

7-13

FIGURE 7-4 bE/EM AiiALYSIS OF SEHISCALE TEST S-UT-6 VESSEL COLLAPSED LICUID LEVEL .

BE/EMPREDICT10f4

- -- S-UT-8 DATA 30.000 g  ; j 25 000 i

20 000 I i

t

.H i W 's uJ s TOP OR CORE 15.000 a l

> \

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l N, .' ,

0 000 O O O O O o 9 9 9 9 9 O O O O O O

• O O O O O O ~ N cn v U")

TIME IN SEC 7-14

i 1

FIGURE 7-5 BE/EM Ai4ALYSIS OF SBilSCALE TEST S-UT-S IliTACT LOOP STEAll GEliERATOR U-TUBE UPFLOU SIDE COLLAPSED LIOUID LEVEL BE/EMPREDICTION

- -- S-UT-8 DATA i 36 000  ;  ;

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TIME IN SEC 7-15

FIGURE 7-6 BE/Eh Ai1ALYSIS OF SEMISCALE TEST S-bT-E IliTACT LOOP STEAM GEkERATOR U-TUBE DOWl4 FLOW SIDE COLLAPSED LIQUID LEVEL BE/EMPREDICTION ,

---

S-UT-8 DATA I 36 000 , ,

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! TIME IN SEC L

7-16 w

I l

l 1

FIGURE 7-7 BE/EM AliALYSIS OF SEHISCALE TEST S-UT-E IliTACT LOOP PUMP SUCTION LEG DOWiiFLOW SIDE COLLAPSED LIQUID LEVEL BE/EMPREDICTION

~

S-UT-8 DATA 18 000 15.000 \\ i I

's 12 000 1

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u; $

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. ',' i 0 000 o a o a o a 9 9 9 9 9 9 8 8 8 8 8 o - OJ M v LD TIME IN SEC 7-17

i FIGURE 7-8 l EE/B1 AllALYSIS OF SB11 SCALE TEST S-UT-E li4 TACT LOOP Pui1P SUCTI0li LEG UPFLOW SIDE COLLAPSED LIQUID LEVEL .

BE/EM PREDICTION .

---

S-UT-8 DATA 18 000 l l I i  !

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I 15.000 l

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+

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o o o o o .

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. I TIME IN SEC l l

/-10

,I -

FIGURE 7-9 BE/ Ell ANALYSIS OF SElilSCALE TEST S-UT-E

, D0l!WC0f!ER COLLAPSED LIQUID LEVEL

. BE/EMPREDICTION S-UT-8 DATA -

30.000 ,

i i i I $

l ,  !

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i FIGURE 7-10 BE/EH AWALYSIS OF SEfilSCALE TEST S-bT-E VESSEL TWO-PliASE MIXTURE LEVEL .

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COMPARIS0il Of PEAK TEMPLRATLRES FOR VARIOUS AXII.L ELEVAT!0liS PRIOR TO LOOP SEAL CLEARING 12 PARCH HOT R0D CODE

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l 3.0 IMPACT ON NSSS ANALYSES I

- As described in Section 5.0, the post-test analysis of Semiscale Test S-UT-8 included a number of modifications to CEFLASH-4AS. In Section 8.1 below, the  !

l impact of these code modifications on NSSS analyses is discussed. This is j

~

followed by a discussion of the implications of the Test S-UT-8 analyses on current small break LOCA lice'nsing analyses.

I

> 8.1 Impact of Model Modifications on NSSS Analyses I

In this section, each of the CEFLASH-4AS model modifications described in Section 5.0 is reviewed for its effect on the analysis of Test S-UT-8. Then, j the applicability of the model modifications to the analysis of C-E designed NSSS's is assessed, first, by comparison of the BE and BE/EM results for I Semiscale and, second, by estimation of the models' impact in break spectrum a analyses for an NSSS.

Floodino in Steam Generator U-Tubes In the U-tubes prior to the time of loop seal clearing, countercurrent flooding produced liquid accumulation by limiting the rate of liquid downflow from the steam generator. The effects of liquid accumulation in Test S-UT-8 are discussed below for the BE and BE/EM analyses.

Comparison of the BE and BE/EM analysis predictions of collapsed level in the U-tubes (Figures 6-6, 6-7, 7-5, and 7-6) shows that without countercurrent flooding the BE/EM analysis produces significantly lower liquid levels in the 8-1 i

- - _ _ _ _ _ _ _ _ - _ _ a

U-tubes than the BE analysis. However, the difference.in liquid level between the upside and downside of the U-tubes is very similar for the two analyses. I I

Therefore, the absence of flooding in the BE/EM analysis did not influence the l net hydrostatic pressure difference producing the core level depression prior to loop seal clearing. Even though the steam generator U-tube levels were different, the U-tube draining times were similar in the BE and BE/EM analysis due to the influence of countercurrent flow interfacial drag on the Semiscale hot leg flow rates.

l In the break spectrum analysis of an NSSS, the effects of countercurrent flooding on the core uncovery phenomenon would be similar to the effects discussed above for the BE analysis. However, the impact of these flooding effects would be less than seen in Semiscale Test S-UT-8 due to differences in U-tube, loop seal, and hot leg geometry. The effects of these geometric differences are discussed below.

C-E steam generator U-tubes are on the average, approximately 20% shorter than Semiscale and the. maximum height variation (from shortest tube to longest).is  !

about twice that of Semiscale. This means liquid in C-E NSSS steam generators will not accumulate to levels as high as in Semiscale, will drain earlier than test S-VT-8, and contribute less to the primary system hydrostatic imbalance acting to depress the vessel liquid level prior to loop seal clearing. -

In C-E designed NSSS's, the loop seal elevation is higher relative to the position of the core than in Semiscale. This means that clearing of the NSSS loop seals will occur with greater inventory in the core for the C-E design than in Semiscale.

8-2 '

l

l i

l The Test S-UT-8 analyses showed that steam generator draining was closely related to hot leg hydraulic conditions. For the same break size, the flow regime dependent frictional losses in the hot legs would be much less in the  !

large diameter NSSS hot leg than in the small diameter Semiscale hot leg.

Also, the NSSS hot leg does not have a vertical section as does the Semiscale

" pant-leg." The Semiscale hot leg design contributed significantly to the delay in draining of the U-tubes in Test S-UT-8. Similar delays would not be expected in the draining of the NSSS steam generator U-tubes.

Therefore, for NSSS analyses with break sizes similar to Test S-UT-8, the effects of countercurrent flooding may possibly lead to deeper core uncovery than cases without flooding due to the effects of steam generator liquid holdup. But, the depth of core uncovery in this case would be less than  ;

observed in Test S-UT-8 due to beneficial aspects of the NSSS U-tubes, loop seal, and hot leg designs. The possible decrease in core level produced by j l

the addition of a flooding model for this NSSS analysis would not affect the i peak cladding temperature which, like Test S-UT-8, occurs during the boiloff phase of the transient where peak temperatures are at least 150"F higher than temperatures occurring prior to loop seal clearing.

The impact of countercurrent flooding in the U-tubes on the analysis of the NSSS worst break was documented in Reference 8-1. The worst break sizes in I C-E small break LOCA calculations (0.05 - 0.1 ft2 ) are considerably smaller than the equivalent break size of Test S-UT-8 (0.2 ft2 ). The slower leak rates in the limiting break size case allow time for the core inventory loss due to steam generator liquid holdup to be replenished by fluid draining back from the hot legs before core uncovery occurs. As shown in Reference 8-1, 8-3

i countercurrent flooding in the worst case analysis delayed steam generator draining (by reducing liquid downflow) but did not produce core uncovery ,

before. loop seal clearing. Subsequent core uncovery during the boiloff portion'of the transient was delayed due to countercurrent flooding but there -

was no change-in the extent.of uncovery. Therefore,~ the effects of 3

.i countercurrent flooding on core uncovery are insignificant in C-E NSSS limiting case analyses.

BE Condensation Heat Transfer ahd Bubble Condensation in U-Tubes Improvements to the CEFLASH-4AS condensation heat transfer models used in the BE analysis of Test S-UT-8 produced roughly 25% more total condensate in the U-tubes and 50 psia lower primary system pressure prior to loop sea 1' clearing than the models used in the BE/EM analysis. The effects of increased condensate and lower pressure on NSSS analyses are discussed below.

Most of the increase in integral condensate in the U-tubes due to improving the steam generator heat transfer models occurs before the onset of countercurrent flooding, see Figure 5-2. During and after countercurrent flooding, the condensation rates are very similar for the two Semiscale analyses; therefore, the effect of improved condensation modeling on liquid accumulation in the U-tubes is minimal for Test S-UT-8. The BE analysis -

showed that the liquid accumulation in Test S-UT-8 is more dependent on liquid i carry-over from the hot legs than on the condensation rate. In this regard, liquid carry-over in the BE/EM analysis, with its EM component model l replacements, was virtually the same as liquid carry-over in the BE analysis.

8-4

_ _ _ _ _ _ _ _ _ _ _ _ 1

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These analysis results for Test S-VT-8 are indicative of the'effect of using improved condensation models in NSSS analyses of a similar break size. That is, for an NSSS, an increase in integral condensate produced by' improved heat transfer methodology would not significantly affect steam generator liquid accumulation which depends more on liquid carry-over from the hot leg than on condensation in the U-tubes.

The primary system pressure results shown, in Figure 6-1 for the BE analysis and Figure 7-1 for the BE/EM analysis, illustrate that the EM condensation models used in the BE/EM analysis lead to higher primary system pressures.

This effect for Test S-UT-8 would be the same in an NSSS analysis. As .

i described in Reference 8-2 for an NSSS analysis, higher pressure produces more inventory loss out the break and less HPSI flow. Thus for the spectrum analysis of an NSSS, the EM condensation models lead to more conservat've core inventory results than the improved condensation mod'Is.

BE Bubble Convection in U-Tubes The BE bubble convection model described in Section 5.1 provided a slight increase in liquid carry-over at the top of the U-tubes compared to the BE/EM analysis prior to pump head degradation. In NSSS analyses, this model is only important during pump operation and was included in the justification of RCP l i

trip strategy documented in Reference 8-3. The model has no significant {

~

affect on NSSS analyses in which early pump trip is assumed, i

8-5

i Core Rewet Steam Production l

i The core rewet steam production model, described in Section 5.2, was implemented into CEFLASH-4AS to reproduce a specific Test S-UT-8 observation.

In the BE analysis, the core rewet model contributed to a prolonged core I

uncovery before loop seal clearing, produced the observed bulk flow of steam from the. vessel into the downcover during the deep vessel level depression, i

and produced lower core recovery after loop seal clearing. In some respects this model modification is Semiscale design specific, because liquid draining from the hot legs appears to uniformly rewet the upper regions of all the heater rods. Whereas in an NSSS, the rewet of fuel rods would be localized in peripheral regions of the core below the hot leg nozzles. After sorre initial steam production,-the regions of the NSSS core affected 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 the smaller, more limiting, break sizes in an NSSS, core uncovery prior to loop seal clearing does not even occur until after the hot legs have drained and this rewet model would not be used.

Representation of Semiscale Design Features The modifications to CEFLASH-4AS described in Section 5.3 for representation -

of specific Semiscale design features, do not apply to NSSS designs. The modified bubble release rate models for the vessel and downcomer apply to the i prediction of bulk flow of steam into the bottom of the downcomer. Since the geometry of the lower plenum, flow skirt, ard annulus are significantly different in an NSSS than in Semiscale, the slug-like displacement of fluid from the downcomer would not occur. The modification to the frictional losses 1

8-6 l

in the hot leg applies only to the small diameter Semiscale hot leg w'ith. its vertical section and not to the large diameter horizontal NSSS hot leg. The variable area vessel lower plenum model. modification, again, applies only to the Semiscale vessel design.

~

Changes to Eliminate Numerical Difficulties The numerical-difficulties in CEFLASH-4AS occurred due to modeling Semiscale Test S-UT-8 and would not occur in NSSS analyses.

8.2 Implications of S-UT-8 Analyses on Current SBLOCA Licensing Analyses-The observations from Semiscale Test S-UT-8 revealed several concerns which are significant for small break LOCA licensing analyses. One of the concerns was that the rapid' water level depression in the vessel prior to loop seal clearing extended below the lowest elevation of the loop seal to the bottom of the heated bundle. Using selected EM component models important for steam generator thermal-hydraulic behavior and vessel core uncovery behavior, the BE/EM analysis of Test S-UT-8 presented in Section 7.0, was capable of-producing adequate predictions of the observed water level depression prior to loop seal clearing, including liquid levels below the loop seal elevation.

i. Another concern with the rapid core coolant level depression prior to loop l

l' seal clearing in Test S-UT-8 was not only the depth of the depression and the accompanying core dryout and core rewet, but also the loss of inventory from the vessel into the downcomer and, finally, out the break during the core uncovery period. The excessive loss of inventory prior to loop seal clearing 8-7

subsequently lead to more severe core heatup during the coolant boiloff phase prior to actuation of the Semiscale accumulators. The BE and BE/EM analyses of Test S-UT-8 were continued beyond the time of loop seal clearing in recognition of this feature of the test. The results of the' hydraulics and. .

cladding temperature calculations of the BE/EM analysis demonstrate that the

-EM component models important for steam generator liquid holdup and core uncovery yield conservative predictions of the vessel inventory and cladding temperatures during the boiloff phase. Since peak cladding temperature occurs during this portion of the transient for C-E licensing calculations, as well as for Test S-UT-8, these BE/EM results demonstrate that the use of.these EM component models in licensing analyses of an NSSS is conservative.

Small break LOCA licensing calculations of C-E designed NSSS's show both of the effects of core uncovery and vessel inventory loss demonstrated by Test S-UT-8. As an example, Figure 8-1 presents the EM predictions of' two-phase level and collapsed liquid level in the reactor vessel for a 0.35ft2 cold leg break for the System-80 NSSS design, (Reference 8-4). The figure illustrates that the vessel collapsed liquid level prediction extends well below the bottom elevation of the pump suction leg loop seal during the time period of steam generator liquid holdup prior to loop seal clearing. After loop seal clearing, the collapsed level prediction shows a sizable reactor vessel inventory loss produced by the pre-loop seal clearing level depression. .

Thus, post-test analyses of Semiscale Test S-UT-8 show that the EM component models important for steam generator and core uncovery behavior can adequately predict S-UT-8 type core uncovery. Furthermore, licensing analyses show that the Evaluation Model predicts Test S-UT-8 core uncovery behavior for C-E designed NSSS's.

8-8

__A _ _ _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ - _ - _

Reference 8-2 shows'that more detailed modeling than the current EM of steam generator liquid holdup effects would not change the limiting break size of a C-E NSSS break spectrum, analysis. The worst break sizes in C-E small break- l l

LOCA calculations (0.05-0.lft2) are considerably smaller than the equivalent break size of Test S-UT-8 (0.2 ftz),

In order for break sizes similar to Test S-UT-8 to become limiting due to more detailed modeling of liquid holdup, the change in pre-loop seal clearing core uncovery must produce increases in the pre-loop seal clearing temperature spike of at least 550*F. Conservative modeling of SIT actuation (see References 8-2 and 8-5) produces peak cladding temperatures during boiloff which are at least 150*F greater than the cladding temperature spike produced by the effects of steam generator liquid holdup on pre-loop seal clearing core uncovery. Also, these larger break sizes with SIT injection have peak cladding temperatures already at least 400*F lower than the presently limiting NSSS break. Based on the discussions of Section 8.1, the effects of more detailed modeling of liquid holdup will not increase the pre-loop seal clearing temperature spike in an NSSS analysis by 550*F, therefore, the limiting break size is unchanged.

For break sizes equal to or less than the limiting break size, core uncovery

. occurs after the steam generators are drained. Thus, the core uncovery behavior is dominated by the capabilities of the high pressure safety injection flow to make up system inventory loss out the break and not by inventory distribution (liquid holdup) prior to clearing of the loop seals.

For conservatism in EM analyses, C-E models the minimum guaranteed high pressure safety injection flow from one pump.

8-9

l 1

These results demonstrate that the EM is adequate for predicting the type of vessel liquid level depression observed in the data of Test S-VT-8. Also,

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more detailed modeling of steam generator liquid holdup would not change the

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limiting break size or limiting peak cladding temperature of NSSS break spectrum analyses. k

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References for Section 8.0 >>

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., 8-1 CEN-203-P, Revision-1-P, " Response to NRC Action Plan Item II.K.3.30, Justification of Small Break LOCA Methods," March 1982.

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f[.p 8-2 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. c

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, ,.e 8-3 CEN-268, " Justification of Trip Two/ Leave Two Reactor Coolant Pump Trip s,

4 Strategy During Transients," March 1984.

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8-4 CEN-203-P, Revision 1-P, Supplement 1-P, " Response to NRC Request Number 1forAdditionalinformationonC-EReportCEN-203-P,Rev.1-P,"

b < <

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Teb ruary, 1984. I\[

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'!.,, 8-5 CENPD-137P, " Calculative Methods for the C-E Small Break LOCA Evaluation Model," August, 1974.

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