ML20234C535

Forwards Supplemental Response to NRC 870910 Request for Further Clarification of Util 870818 Response to NRC Questions Re Util Proposed Tech Spec Change to Increase Steam Generator Tube Plugging Limit to 10%

ML20234C535
Person / Time
Site:	Farley
Issue date:	09/16/1987
From:	Mcdonald R ALABAMA POWER CO.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
TAC-62283, TAC-62284, NUDOCS 8709210328
Download: ML20234C535 (7)

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j Alabama Power Company  !

600 North 18th Street Post Othce Box 2641 Girmingham, Alabama 35291-0400 Telephone 205 250-1835 L

R. P. Wponald Senior Wce Presiaent Alabama Power the Southern elmtrK: System 10CFR50.90 September 16, 1987 Docket Nos. 50-348 l 50-364 J

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U. S. Nuclear Regulatory Commission l~

ATTN: Document Control Desk Washington, D. C. 20555 i

j Gentlemen: l Joseph M. Farley Nuclear Plant - Units 1 and 2 j Steam Generation Tube Plugging Limit Technical Specification Change N

by letter dated August 18, 1987, Alabama Power Company responded to several "

questions raised by the U. S. Nuclear Regulatory Commission regarding tne 1 proposed technical specification change that would increase the allowed steam generator tube plugging limit to ten percent. In a subsequent phone conversation held on September 10, 1987, t'ne U. S. Nuclear Regulatory Commission requested further clarification of the response provided by Alabama Power Company concerning the ef fects of increased tube plugging on the Farley small break LOCA analysis. Additionally, a minor clarification  ;

concerning the effects on reactor coolant flow was requested. The attachment to this letter supplements Alabama Power Company's previous responses to U. S. Nuclear Regulatory Commission questions.

Respectf ly suf nitte ,

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R. P. Mcdonald RPM /BHW: dst-D-T.S.7 Attachment 1

cc: Mr. L. B. Long l Dr. J. N. Grace l Mr. E. A. Reeves Mr. W. H. Bradford 8709210328 870916 PDR ADOCK 05000348 Opf P PDR l

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ATTACHMENT NRC Request Provide additional clarification of the effects of 10% steam generator tube plugging on the Farley small break LOCA analysis.

APCo Response INTR 00UCl10N A large break ECCS (LOCA) analysis is currently under review by the NRC for {

the Farley units. The analysis is based upon the recently approved BASH l evaluation model and shows continued compliance to acceptance criteria in  !

10CFRSO.46 assuming 10% steam generator tube plugging.

In that submittal a short evaluation was provided stating that the proposed ,

steam generator tube plugging would not impact the current Farley licensing basis analysis for small break LOCA. The Farley small break analysis of ,

record was performed with the WFLASH evaluation model in 1974. The limiting )

break from that analysis was the 6 inch break with a peak cladding l temperature of 1703 F. For Unit 1, a sensitivity study was performed to assess the ef fect of upflow conversion. The PCT of this case was 1820 F for the same 6 inch break size. That analysis supports a peaking factor (Fq ) of 2.32, l l

The effect of 10% Steam Generator Tube Plugging (SGTP) on the small break LOCA analysis for the Farley units was considered and determined to be very similar to an evaluation of the same type performed previously for the Almaraz plants, which also are 3 loop PWRs of Westinghouse design (Reference 1). The technical merits of that evaluation are outlined below, followed by a justification of the applicability of those conclusions on the Farley small break analysis, and a discussion of the continued validity of the Farley current small break analjsis of record.

DISCUSSION OF SGTP EFFECTS ON SBLOCA l l

The Reference 1 evaluation was developed by considering those plant and l model parameters which could possibly be affected by SGTP in a small break transient. Three specific ef fects were identified and evaluated: ,

1 1 The in; pact of the reduction of the SG tube area on the smal', break l transient as it relates to the ability to transfer heat from primary to secondary and, thus, dissipate core stored energy and decay heat.

2. The effect of changes to operating temperatures (primary and secondary) as a result of SGTP.

3. The effect that could be exerted on the draining of the steam generator tubes as this has a direct effect on water inventory in the vessel and potential for core uncovery, b

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Attachment Page 2 l

The Reference evaluation concluded that no effect would be expected in the i small break analysis that was then reported for plugging levels up to 20%. '

The justification for this conclusion is summarized below.

1. It was concluded that the effect of an appreciable amount of tube plugging would not inhibit the transfer of heat to the secondary side. Analyses have been performed (Reference 2) from which an l assessment of required heat sink to mitigate the ef fects of a small '

break LOCA was investigated.

In Reference 2, various small break analysis cases were performed I to deduce natural circulation effects in the RCS in a small break j scenario., As pc-t of this study, several cases were performed which specifically reduced secondary heat sink, by not modeling auxiliary feed flow, allowing secondary level to be reduced to j within 6 f t. of the bottom of tube plate, etc. )

l The resulting comparison concluded that with a smaller amount of I heat sink the primary side response was still similar to test I f acility response, full heat sink assumed, implying that a plant f can withstand severe degradation in heat sink capacity and still l not af fect expected transient response. {

In the Almaraz evaluation, the plugging of tubes was extrapolated l to be directly applicable to this reduction of heat sink in terms I of the reduced number of tubes availaole for heat transfer. It was concluded that the plant could remove 10% of the tubes from service and the reduction of heat transfer would still be bounded by the reduction of heat transfer demonstrated in Reference 2. Therefore, it was concluded that the first effect, the ability to transfer heat from the core, would not be impacted by the steam generator tube plugging due to the large amount of remaining heat transfer area.

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2. As a result of SGTP in steady state operation, the temperature dif ference across the SG tubes between primary and secondary will increase. This is due to the need to transfer the same amount of heat across a reduced net tube area. In the accident scenario, as is indicated earlier, the heat transfer requirement is reduced to only the core stored energy and decay heat after shutdown.

Nevertheless, the temperature gradient will af fect the heat transfer capacity, and was investigated as another change that should be addressed in an evaluation of SGTP.

The nature of the small break transient modeling is such that generation of a Icw pressure or reactor trip signal would be coincident with a lois of offsite power assumption. As a result, feedwater isolation would be indicated. Heat transferred from the primary would heat up the secondary inventory, forcing an increase 1

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Attachment Page 3 in pressure and driving the secondary to approach the SG safety valve setpoint. This happens very shortly after the accident. The increased initial temperature difference between the primary and I the secondary side disappears right after the break as the

. econdary side pressure stabilizes at the steam generator safety s

valve setpoint and acts as a governing influence on the continued primary blowdown. Considering the rapidity with which this occurs, the engineering evaluation for Almaraz concluded that this increased temperature dif ference will not affect the transients as '

i a result of SGTP.

3. Due to the size of the hypothesized small break, the small break LOCA accident is characterized as a draining process, with very discrete vessel and loop water levels marking the stages of the accident. Oraining the steam generator tubes is a prominent phase in the scenario. This was considered in the reference Almaraz evaluation postulating that hydraulic changes as a result of tube plugging could influence steam generator tube draining in the small break LOCA. This has a direct effect on water inventory in the vessel and potential for core uncovery.

For a break in the cold leg, as the RCS pressure decreases, a two-phase flow circulation is established. When the two phase circulation stops, the mixture level in the steam generator tubes drops steadily on both the up and down sides of the tube.

During this tube draining period, the steam generated in the core flows to the steam generators, and part of it will be condensed in the steam generator tubes and flow back to the core with the draining water. This countercurrent steam flow affects the draining rate from the steam generators and directly affects the ,

degree of core uncovery (inventory in the vessel) and consequences l of a small break LOCA. 4 investigations have been made for typical PWR geometry and i conditions existing during a small LOCA event. Reference 3 I documents a study performed to assess this countercurrent flow l effect, taking the assumption that there would be some location i between the vessel upper plenum and the steam generator which would restrict this flow to the greatest extent. One case investigated j was the limitation to countercurrent flow presented by the I hydraulic resistance of steam generator tubes themselves. A second case in that study considered the countercurrent flow restriction present in the hot leg. The limitation to countercurrent flow or countercurrent flow limit (CCFL) in the horizontal piping of the hot leg was found to be larger than that presented in the tubes.

Further, it was determined that additional hydraulic resistance would be present because of the bends in the inclined pipe from the J l

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Attachment Page 4 hot leg to the steam generator inlet and that this would be the limiting case of the the three. The conclusion was that the limiting countercurrent flow resistance (limit) in the inclined pipe would govern the draining of the tubes and most directly affect core level.

Considering the effect of steam generator tube plugging, the Almaraz evaluation acknowledged that the limiting countercurrent flow conditions for minimum water backflow from the steam generator to the vessel would occur in the steam generator inlet elbow. Tube plugging may affect this conclusion depending on the severity of the plugging. At a lower plugging percentage (up to 20%), the evaluation concluded that the limiting CCFL discussed above (Reference 3) would still occur at the inclined pipe connecting the steam generator inlet plenum and hot leg. Therefore, at a low plugging level (10%), the small break LOCA transients would not be af fected by the tube plugging.

On the other hand, at a higher percentage of tube plugging (approaching or greater than about 20%) the CCFL expected in the tubes may increase to the extent that the steam generator tubes would exhibit the limiting CCFL characteristics of the systen and govern the core uncovery resulting in a peak cladding temperature penalty.

Based on the above, it was concluded that the small break LOCA is of no concern at all for the steam generator tube plugging limits considered (up to 10%) for the Almaraz units.

JUSTIFICATION OF SIMILARITY Design information has been referenced for both the subject nuclear reactors. It has been confirmed that Almaraz and Farley are similar in design and would be so regarded with respect to inferring trends regarding plant specific small break LOCA analyses.

Both plants are of identical Westinghouse vessel design. Common features include: 3 loops,167 fuel assemblies, standard fuel design (.374 inch 00),

48 control rods,144 inch active fuel length, Model 93 reactor coolant p ump s . Both are analyzed with WFLASH as their small break analysis of record. A few small deviations exist with respect to operating parameters, but these are not significant in projecting the similarity of the plants in the small break accident scenario. For example, reactor power is 2686 MWt ,

for Almaraz whereas Farley's reactor power is 2652 MWt. l It is concluded that evaluations based on the Almaraz plant for purposes of I establishing trends and sensitivities would be equally applicable for the Farley units.

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Attachment Page 5 CONCLUSIONS AS TO VALIDITY OF CURRENT ANALYSIS Based on the foregoing technical argunents, it was concluded that the ef fect of 10% SGTP on the Farley small break analysis of record would be minimal.

The results of the BASH analysis show a limiting peak cladding temperature of 2013 F, indicating as much as 300 F margin over the small break results reported in Reference 4.

WCAP-8340 (Reference 5) presented the results of large break and small break LOCA analyses for typical Westinghouse designed two-loop, three-loop, and four-loop plants. The report results indicate that the large break LOCA peak cladding temperatures bound the small break LOCA peak cladding temperatures due to higher inventory depletion at higher core decay heat levels. A plant specific WFLASH analysis incorporating SGTP modelling would not result in a significant change to calculated PCT. Moreover, it would continue to be bounded by the Large Break analysis presented in the Farley submittal based upon the BASH Evaluation Model and would continue to conform j to the acceptance criteria of 10CFRSO.46. '

With the licensing of NOTRUMP as a new small break Evaluation Model, some ccasideration can be taken as to potential SGTP effects if reanalyzed on the i basis of the newer model. l l

Following the incident at Three Mile Island Unit 2, Westinghouse and the Westinghouse Owners Group developed the NOTRUMP computer code (References 6 and 7) as the new small break LOCA evaluation model, which was approved in May 198S, to meet the requirements of NUREG-0737 (Reference 8),Section II.K.3.30. Small break LOCA analyses performed using NOTRUMP (Reference 9) l demonstrated that, in general, the NOTRUMP evaluation model calculated lower peak cladding temperatures than the WFLASH evaluation model. This allowed l the WFLASH analyses contained in the Joseph M. Farley Final Safety Analysis Report (FSAR) to remain the licensing basis analysis of record in accordance with NRC Generic Letter 83-35.

REFERENCES

1. Risher, D. H., et. al . , "An Evaluation of up to 10% Steam Generator Tube Plugging on CN Almaraz Units 1 and 2," WENX-86-12, March 1986.

2. Ciana, S., Patti, B. and Lee, N., " Simulation of Small Break Type Behavior of PUN and SPES using NOTRUMP Code," Proceedings of the Specialists Meeting on Small Break LOC A Analyses in LWR's, Pisa, Italy, June (1985).

3. Lee, N., " Limiting Countercurrent Flow Phenomena in Small Break LOCA Transients," Proceedings of the Specialists Meeting on Small Break LOCA Analyses in LWR's, Pisa, Italy, June (1985) .

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1 Attachment Page 6 i

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4. Final Safety Analysis Report, Farley Units 1 & 2, Section 16.3.

S. WCAP-8340, PWR Systems Division, Nuclear Safety, " Westinghouse Emergency ,

Core Cooling System - Plant Sensitivity Studies", July 1974.

6. WCAP-10079-P- A, ( Proprieta ry) , Meyer, P. E. , "NOTRUMP : A Nodal Transient l Small Break General Network Code", Augus t 1986. j i

7. WCAP-100S4-P-A, (Proprietary), Lee, N. , Tauche, W. D. , et al . , " Westinghouse I l

Small Break ECCS Evaluation Model using The NOTRUMP Code", August 1985.

8. WCAP-1114b-P- A, (Proprietary), Rupprecht, S. D. , Osterrieder, R. A. , et al . ,

" Westinghouse Small Break LOCA ECCS Evaluation Model Generic Study With The NOTRUMP Code", October 1986.

9. NUREG-0737, United States Nuclear Regulatory Commission, " Clarification of j TMI Action Plan Requirements", November 1980. l l

l NRC Request Provide additional information regarding the effects of 10% steam generator tube plugging on calculated RCS flow as compared to 0% tube plugging.

1 APCo Response The calculated RCS flowrates for Farley with 0% steam generator tube plugging I are 96,400 gpm/ loop (289,200 gpm total) for Unit 1 and 95,700 gpm/ loop (287,100 l gpm total) for Unit 2. These values are based on best estimate flow models.

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