ML20204E658

Return to Power Evaluation

ML20204E658
Person / Time
Site:	Beaver Valley
Issue date:	07/17/1986
From:	WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
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ML20204E643	List: ML20204E636 ML20204E658
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NUDOCS 8608010199
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BEAVER VALLEY UNIT 1 RETURN TO POWER EVALUATION

- July 17, 1986 -

I Prepared by: WESTINGHOUSE ELECTRIC CORPORATION I

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BEAVER VALLEY UNIT 1 RETURN TO POWER EVALUATION C

SUMMARY

An evaluation has been performed by Westinghouse to support the Beaver Valley f Unit I return to full power operation following the recent (July 1986) steam l generator tube plugging. The evaluation was based on a tube plugging level of 2.5% which conservatively bounds the actual plugging level. Based on engineering experience, the evaluation results at the 2.5% tube plugging level demonstrate that the safety of the plant is not impacted as a result of the plugging and that the plant can be safely operated at rated thermal power.

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I. INTRODUCTION An evaluation has been performed by Westinghouse to support the Beaver Valley Unit 1 plant return to full power operation following the recent (July 1986) steam generator tube plugging. The purpose of this evaluation was to perform a conservative assessment on an expedited schedule which provides a high level of confidence and to estimate the safety and operational impact on the plant, associated with returning to full power at the present plugged conditions. A plugging level of 2.5% was assumed for the evaluation. The 2.5% level conservatively bounds the actual plugging level.

The plugging of any steam generator tubes is expected to result in proportionate effects on plant characteristics including; a) reduction of steam generator heat transfer area between primary and secondary side, b) reduction in RCS (reactor coolant system) flow due to the increase of pressure drops in the steam generator tubes, and c) reduction in RCS active volume.

Such aspects were taken into account in the assessment of safety and operational impact of the Beaver Valley Unit 1 steam generator tube plugging.

The scope of this assessment is limited to using the existing data and calculation results. A detailed LOCA analysis will be completed as part of a more extensive tube plugging analysis program. Assumptions have been made regarding RCS flow, steam generator outlet pressure and steam generator moisture carryover. Evaluationresultshavebeenpredictedbasedonknown conservative sensitivities and use of available generic margins. Where appropriate, documents such as stress reports, design transient analysis reports and equipment specifications have been examined to determine whether the plugged conditions are bounded by existing analyses for relevant components. Finally, control and protection system setpoints have been reviewed.

II. ASSUMPTIONS AND PARAMETRIC DATA i

The evaluation of the Beaver Valley Unit I return to power operation with 2.5%

steam generator tube plugging has been performed utilizing the existing 5017e:1d/071786 1

licensing basis and industry codes and standards for Beaver Valley Unit 1 at the current fuel cycle with 17x17 standard fuel core. The major assumptions involved in the evaluation are as follows:

a) The level of tube plugging in each steam generator is assumed to be 2.5% (which is in excess of the highest plugging level experienced for Beaver Valley Unit 1).

b) The best estimate reactor coolant flow is assumed to be 96,000 gpm/ loop. This flow is based on previously measured Beaver Valley Unit I data (96,600 gpm/ loop for 0% plugging) and predicted flow reduction due to 2.5% tube plugging. The thermal design flow used for the evaluation remains at the currently assumed value of 88,500 gpm/ loop, based on the assumed best estimate flow.

c) The steam generator outlet pressure at rated power is assumed to decrease to 782 psia as a result of the plugging. In reality, based on operational data, it appears a steam pressure of at least 790 psia can be maintained, d) Vessel average coolant temperature is maintained at its design value of 576.2*F at rated power.

e) For the components evaluation, it is assumed that the design transients would not change as a result of the plugging.

Table 1 shows the Beaver Valley Unit 1 design power capability parameters used for the evaluation.

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TABLE 1 BEAVER VALLEY UNIT 1 DESIGN POWER CAPABILITY PARAMETERS Parameters 0% Plugging 2.5% Plugging

1. Tubes 3388 3388

1. Tubes Plugged (per S/G) 0 85 Best Estimate Flow (gpm/ loop) 96600 96000 Thermal Design Flow (gpm/ loop) 88500 88500 NSSS Power (MWt) 2660 2660 Reactor Coolant Pressure (psia) 2250 2250 RCS Temperatures (*F):

Reactor Loop Inlet 542.3 542.3 Reactor Loop Average 576.2 576.2 Reactor Loop Outlet 609.9 609.9 Core Inlet 542.5 542.5 Core Average 579.3 579.3 Core Outlet 612.8 612.8 Steam Temperature (*F) 516.8 515.7 Steam Pressure (psia) 790 782 Steam Flow (10E6 lb/hr total) 11.61 11.61 Feed Temperature (*F) 437.5 437.5 l

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III. EVALUATION RESULTS

1) General: Plugging of steam generator tubes generally affects a) primary to secondary heat transfer (reduction in heat transfer area in the steam generators), b) reactor coolant flow (reduction in flow due to the increase of friction in steam generators) c) reactor coolant volume due to steam generator tube volume reduction and d) loop flow balance.

Plugging of each steam generator to a level of 2.5% causes a reducticn of primary to secondary heat transfer area in the steam generators, and affects the heat transfer capability. With the reactor operating at the rated thermal power with the reduced heat transfer area without changing coolant Tavg, the steam generator steam pressure will decrease from the previous operating condition. The extent of steam pressure reduction is strongly affected by the steam generator fouling factor. Generally, the fouling factor margin (approximately 12% of the total resistance to heat transfer across the steam generator tubes) included in the design of steam generator is sufficient that the steam pressure may not go below the initial design pressure (790 psia). However, this can only be proven by in plant measurement, and for that reason an accurate measurement of the steam pressure is strongly recommended after return to full power.

With the above listed general aspects taken into consideration related to the steam generator tube plugging, evaluations and analyses have been performed in the following areas: a) LOCA analyses, b) non-LOCA analyses, c) NSSS design transients, d) system and component analyses, and

] e) protection and control system setpoints.

2) Summary of LOCA Analysis Impact: Based on the recently completed Eddy-Current tests at Beaver Valley Unit 1, Duquesne Light Co. predicts the highest level of steam generator tube plugging (SGTP) to be 1.06% (in steam generator A). The large break LOCA analysis for Beaver Valley Unit I was redone in 1979 assuming 1% SGTP, using the February 1978 Evaluation Model; Westinghouse small break LOCA analysis of record is tnat presented in the FSAR.

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Steam Generator Tube Plugging (SGTP) is a large break LOCA ECCS performance penalty, because of the flow restriction it produces within the steam generators. This restriction inhibits mass flow through the coolant loops, which affects core flow during blowdown and hinders the reflooding of the reactor core following a large break LOCA.

SGTP levels affect the performance of calculated small break LOCA in three important aspects namely, the reduced heat transfer area, the increased initial temperature difference between the primary and the secondary side, the countercurrent flow limit (CCFL).

Small Brea(L0 y nalysis Jmoact: Duping smuil breat LUCA only a small portion of the steam generator heat transfer area is required to provide an effective heat sink. Thus the availability of adequate steam generator heat transfer area is unaffected because of the tube plugging.

In Appendix K small break LOCA analysis the increasing temperature difference between the primary and the secondary side will disappear immediately after the break, because the secondary side pressure reaches the steam generator safety valve set point. SGTP may affect the CCFL l

characteristics depending on the severity of the plugging. At a 2.5%

plugging level, the limiting CCFL is still at the inclined pipe connecting the steam generator inlet plenum and hot leg. Therefore at low plugging

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levels, small break LOCA transients would not be affected by the tube plugging and remain non-limiting.

Large Break LOCA Analysis Impact: The 1% SGTP large break LOCA analysis using the February 1978 evaluation model resulted in a PCT of 2123.7'F for a double-end cold leg guillotine break (DECLG), with discharge coefficient, CD=0.4 (limiting case). Results of sensitivity studies reported in WCAP-8986 show that the relationship between % tubes plugged and increase in peak clad temperature (PCT) is linear for low levels of uniform steam generator tube plugging. A sensitivity study to tube plugging performed for a 3 loop plant predicts an increase of 12.4*F in 5017e:1d/071786 5 y-wmye- - , -cs-- c

PCT for 1% increase in SGTP. Using this result the increase in PCT was estimated for Beaver Valley Unit 1 for 2.5% SGTP.

1 PCT (*F)

At 1% tube plugging 2123.7 At 2.5% tube plugging 2142.3 (estimated)

The result of this evaluation indicates that increasing the SGTP level up to 2.5% at Beaver Valley Unit 1, would not impact the small break LOCA analysis results, and will still maintain the large break LOCA analysis calculated PCT below the 2200*F regulatory limit.

Summary of Results for S/G Tube Runture Event: Steam generator tube plugging levels up to 5% are being considered as a conservative upper bound for a 2.5% case. Plant measurements indicate that there is sufficient margin in thermal design flow such that the initial design value (88,500 gpm/ loop) can be maintained for tube plugging beyond 5%.

Thus, the design RCS flows and temperatures will not change from the initial design values.

An evaluation has been performed to assess the effect of up to 5% tube plugging on the SGTR analysis in tne FSAR. It was determined that up to 5% tube plugging will not result in an increase in the calculated mass of reactor coolant transferred to the faulted steam generator or the steam released to the atmosphere which are reported in the FSAR for an SGTR.

I Thus, it is concluded that the consequences of an SGTR will not be increased as a result of the proposed 2.5% steam generator tube plugging.

4) Non-LOCA Analysis: As shown in Table 1, the primary side design operating parameters for 2.5% tubes plugged are identical to those currently used in

'the Beaver Valley Unit 1 non-Loss of Coolant Accident analysis. However, there is a slight decrease in the secondary side steam generator temperature and pressure. Although not reflected in Table 1, there is a slight reduction in RCS active volume.

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Of the non-LOCA transients discussed in Chapter 14 of the BVPS-1 FSAR, only the loss of normal feedwater, loss of offsite power to the station auxiliaries, and major rupture of main feedwater pipe (FSAR Sections 14.1.8, 14.1.11, and 14.2.5.2) could be sensitive to secondary side conditions at power. The decreased steam pressure and temperature cause the nominal steam generator inventory and the inventory at time of reactor trip to increase slightly. These increases are a slight benefit. Thus, the conclusions in the FSAR remain valid.

For the Uncontrolled Boron Dilution event (FSAR Section 14.1.4) a reduction in active RCS volume could decrease the time available to terminate the event. For the case at refueling, there is no change to the current analysis since the RCS is drained below the steam generator nozzles. The case at Startup is covered by the Interim Operating Procedure. Analysis to support this procedure is based upon the Residual Heat Removal flow path which does not include the steam generators. Thus the results of that analysis are independent of the level of steam generator tube plugging. For the case at Power, the 2.5% reduction in steam generator tube volume results in an approximately 0.8% reduction in RCS active volume which can be absorbed by the conservatism in the current analysis. Therefore, the conclusions of the FSAR remain valid.

For all other non-LOCA transients, since the primary side parameters do not change, there is no impact upon the analyses as presented and the conclusions remain valid. In summary, the conclusions stated for non-LOCA transients in Chapter 14 of the Beaver Valley Unit 1 FSAR remain valid for a steam generator tube plugging level of 2.5%.

5) NSSS Design Transients: NSSS component design transients have been reviewed to determine their applicability to the Beaver Valley Unit 1 plant with 2.5% tube plugging. A comparison has been made between the conditions used for the original design transients and new plant operating conditions associated with 2.5% steam generator plugging for power, coolant flow, primary temperatures, and secondary pressure and temperature. Based on the evaluation, it is judged that for 2.5% tube 5017e:1d/071786 7

plugging and associated assumed steam pressure reduction, current SSDC (Systems Standard Design Criteria) 1.3, Rev.1, design transients are bounding for the new operating conditions due to the conservatisms assumed in the existing transient analyses.

6) NSSS Components: A review was conducted in order to evaluate the impact of 2.5% tube plugging on the Reactor Coolant System (RCS) components with respect to operation, structural integrity, performance and safety. As shown in Table 1, the Reactor Coolant Pressure, thermal design flow, and primary system temperatures are not affected by the 2.5% plugging.

Further, it was determined that the design transients for the primary system with 2.5% tube plugging are bounded by the original design transients. Therefore, the effects of 2.5% tube plugging on the Reactor Vessel, Reactor Internals, Control Rod Drive Mechanisms, Reactor Coolant Pumps, Pressurizer and Loop Isolation Valves as well as auxiliary equipment including auxiliary pumps, valves, tanks and heat exchangers were considered to be insignificant due to the fact that the original design parameters are unchanged with 2.5% tube plugging. The slightly reduced steam temperature and steam pressure at the 2.5% plugging level required a more detailed evaluation of the Steam Generators.

The Beaver Valley Unit 1 Steam Generators were evaluated for the 2.5% tube plugging condition in order to determine the effects of plugging on their performance.

The results of the evaluations are as follows:

Calculations were performed to determine if the Thermal-Hydraulic parameters resulting from 2.5% plugging would be acceptable with respect to Steam Generator degradation due to corrosion. The analysis results indicated that 2.5% plugging is acceptable.

All thermal-hydraulic operating characteristics were found to be acceptable. Adequate margins in steam pressure and moisture carryover continue to exist at a plugging level of 2.5%. Steam Generator stability and circulation ratios were found to be not significantly affected by a plugging level of 2.5%.

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Structural analysis of components affected by the increase in primary-to-secondary pressure resulting from 2.5% plugging show that stress levels and fatigue usage for the tubes, tubesheet, and divider plate remain within acceptable limits.

7) Protection and Control System Setpoints: An evaluation performed for Beaver Valley Unit 1 has indicated that no protection system setpoint revisions are required. The performance of the various control systems will not be materially affected, and there will be no loss in plant operability or degradation in the plant response to normal expected transients. No control system setpoint changes are required. This is based both on an overall review of the control systems, and on experience based on other plants which have undergone tube plugging.

Revision of protection system setpoints are not required since there is no change in RCS operating parameters such as RCS flow and reactor vessel coolant average temperature and AT for a 2.5% plugging level. Also, loop flow reduction could have a potential effect on the loop transit lag for dynamic compensation for the setpoints. Since there is no change to the nominal primary side operating parameters, revision of these setpoints is not required.

The effect of reduced reactor coolant flow on the response time of the reactor coolant temperature detectors has also been reviewed. There will be no change in response time since thermal design flow is maintained at its original value.

The effect of the unequal thermal loads in the steam generators on steam generator level swings has also been evaluated. The maximum level swings occur during large load rejections. Operating experience has shown, however, that sufficient margin exists on Beaver Valley 1 such that the change in level response due to the unequal thermal loads will not result in level swings which cause steam generator level trips outside the nominal operating design basis.

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8) Results:

The Beaver Valley Unit I return to full power evaluation results reveal: l The Beaver Valley Unit 1 LOCA analysis result satisfies the LOCA acceptance criteria with a Peak Clad Temperatura of 2142.3*F; that is far below 2200*F.

- Non-LOCA evaluation results confirm that Beaver Valley Unit 1 can operate at 100 percent of rated reactor power at the current steam

- generator plugging conditions.

Based on the NSSS design transient evaluation, for the 2.5% level of plugging, the existing systems design transients remain bounding.

Evaluation predicts that the impact of 2.5% plugging on the NSSS components with respect to operation, structural integrity, performance and safety are insignificant.

- Evaluation results indicate no required protection and control system setpoint changes from present values.

IV. CONCLUSIONS Based en the evaluation results, it is concluded that there is no safety problem associated with returning to full power operation on Beaver Valley Unit 1 with a 2.5% level of steam generator plugging.

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