ML19351F358

Single Loop Operation.

ML19351F358
Person / Time
Site:	Peach Bottom
Issue date:	05/31/1980
From:	Herikson P GENERAL ELECTRIC CO.
To:
Shared Package
ML19351F346	List: ML19351F345 ML19351F352 ML19351F358
References
80NED270, NEDO-24229-1, NUDOCS 8101120309
Download: ML19351F358 (33)

Text

NED0-24229-1 80NED270 Class I May 1980 PEACH BOTTCM ATCMIC PC'n*ER STATICN UNITS 2 AND 3 SING 13-LOJP OPERATION NUCLEAR PCWER SYSTEMS DIVISION e GENERAL ELECTRIC COMPANY SAN JOSE. CALIFORNI A 95125 GEN ER AL h ELECTRIC g9

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

NEDO-24229-1 IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT (Please Read Carefully)

This report was prepared by General Electric solely for Philadelphis.

Electric Company (PECo) for PECo's use with the U.S. Nuclear Reg-ulatory Commission (LSNRC) for supporting PECo's operating license of the Peach Bottom Atomic Power Station Units 2 and 3. The infor-mation contained in this report is believed by General Electric to be an accurate and true representation of the facts known, obtained i or provided to General Electric at the time this report was prepared.

l The only undertakings of the General Electric Company respecting information in this document are contained in the General Electric Company Single-Loop Operation Evaluation Proposal 414-TY37-HE0 (CE letter No. G-HE-8-208A, dated September 7, 1978). The use of this information except as defined by said proposal, or for a ty purpose l other than that for which it is intended, is not authorized; and

, with respect to any such unauthorized use, neither General Electric Company nor any of the contributors to this document makes any repre-sentation or warranty (express or implied) as to the completeness, I

accuracy or usefulness of the information contained in this document or that such use of such information may not infringe privately owned rights; nor do they assume any responsibility for liability or damage of any kind which may result from such use of such information.

l l

NED0-24229-1 CONTENTS fagg

1. INTRODUCTION AND

SUMMARY

1-1

2. MCPR FUEL CLADDING INTEGRITY SAFETY LIMIT 2-1 2.1 Core Flow Uncertainty 2-1 2.2 TIP Reading Uncertainty 24 3 MCPR OPERATING LIMIT 3-1 31 Core-Wide Transients 3-1 3.2 Rod Withdrawal Error 3-2 3.3 Operating MCPR Limit 3-3 4 STABILITY ANALYSIS 4-1

5. 4CCIDENT ANALYSES 5-1 0.1 Loss-of-Coolant Accident Analysis 5-1 5.2 One-Pump Seizure Accident 5-3

6. REFERENCES 6-1 i

l l

l 2

iii/iv

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NEDO-24229-1 ILLUSTRATIONS Figure Title Page 2-1 Illustration of Single Recirculation Loop Operation Flows 2-5 3-1 Main Turbine Trip with Bypass Manual Flow Control 3-4 4-1 Decay Ratio versus Power Curve for Two-Loop and Single-Loop Operation 4-2 5-1 Peach Bottom 2 Suction Break Spectrum Reflood Times 5-6 5-2 Peach Bottom 2 Discharge Break Spectrum Reflood Times 5-7 5-3 Peach Bottom 3 Saction Break Spectrum Reflood Times 5-8 53 Peach Bottom 3 Discharge 3reak Spectrum Reflood Times 5-9 5-5 Peach Bottem 2 Discharge Break Spectrum Uncovered Times 5-10 5-6 Peach Bottom 2 Suction Break Spectrum Uncovered Times 5-11 5-7 Peach Bottom 3 Discharge Break Spectrum Uncovered Times 5-12 r

5-8 Peach Bottom 3 Section Break Spectrum Uncovered Times 5-13 TABLES Table Title Py 5-1 MAPLHGR Multiplier Cases 5-5 5-2 Limiting MAPLHGR Reduction Factors 5-5 v/v1

/

NEDO-24229-1

1. _ INTRODUCTION AND SD! MARY The current technical specifications for the Peach Bottom Atomic Power Station Units 2 and 3 require that the reactor shall not be operated for a period in excess of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with one recirculation loop out of service.

The capability of operating at reduced power with a single recirculation loop is highly desirable, from a plant availability / outage planning standpoint, in

',he event maintenance of a recirculation pump or other ccmponent renders one loop inoperative. To justify single-lovp operation, the safety analyses documented in the Final Safety Evaluation Reports and Reference 1 were reviewed for ene-pump operation. Increased uncertainties in the ccre total flow any TIP readings result in an 0.01 incremental increase in the MCPR fuel cladding integrity safety limit during single-loop operation. This 0.01 increase is reflected in the MCPR operating limit. No other increase in this limit is required as core-wide transients are bounded by the rated power / flow analyses performed for each

, cycle, and the recirculation flow-cate dependent rod block and scram setpoint l

equations given in the technical specifications are adjusted for one-pump operation.

The least stable pcwer/ flow condition, achieved by tripping both recirculation pumps, is not affected by one-pump operation. During single-loop operation, the flow control should be in master manual to avoid control oscillations in the recirculation flow control system under abnormal conditions. Derived MAPLHGR reduction factors are given in Table 5-2 for application during single-loop operation.

, The analyses were performed assuming the equalizer valve is closed. The dis-charge valve in the idle recirculation loop is normally closed, but if its

closure is prevented, the suction valve in the loop should be closed to prevent l the loss of Low Pressure Coolant Injection (LPCI) flow out of a postulated break in the idle suction line.

This analysis is applicable to future cycles as long as the basis is unchanged.

1-1/1-2

/

NEDO-24229-1

2. MCPR FUEL CLADDING INTEGRITY SAFETY LIMIT Except for core total flow and TIP reading, the uncertainties used in the statis-tical analysis to determine the MCPR fuel eladding integrity sefety limit are not dependent en whether coolant flow is provided by one or two recirculation pumps. Uncertainties used in the two-loop operation analysis are documented in the FSAR for initial cores and in Table 5-1 of Reference 1 for reloads.

A 6% core ficw measurement uncertainty has been established for single-loop operation (compared to 2.5% for two-loop operation). As shown belew, this value conservatively reflects the one standard deviation (one sigma) accuracy of the core flow measurement system documented in P.eference 2. The random noise component of the TIP reading uncertainty was reviaed for single recirculation loop operation to reflect the operating plant test results given in Subsection 2.2 below. This revision resulted in a single-loop operation TI? reading uncertainty of 9.1% for reload cores. Comparable two-loop TI? reading uncertainty value is 8.7% for reload cores. The net effect of these two revised uncertainties is a 0.01 incremental increase in the required MCPR fuel cladding integrity safety limit.

2.1 CORE FLOW UNCERTAINTY 2.1.1 Core Flow Measurement During Single-Loop Operation The jet pump core flow measurement system is calibrated to measure core flow when both sets of jet pumps are in forward flow; total core flow is the sum of the indicated loop flows. Fce single-loop operation, however, the inactive jet pumps will be backflowing. Therefore, the measured flow in the backflowing jet pumps must be subtracted from the meast.ed flew in the active loop. In addition, the jet pump flow coefficient is different for reverse flow than for forward flow, and the measurement of reverse flow must be modified to account for this difference.

For single-loop operation the total core flow is derived by the following formula:

Total Core I Active Locp I I Inactive Loop

• ~

( Flow j Indicated Flow Indicated Flowj 2-1

/

NECD-24229-1 where C (= 0.95) is defined as the ratio of " Inactive Loop True Flow" to "Inac-tive Loop Indicated Flow," and " Active Loop Indicated Flow" is the flow indicated by the jet pump " single-tap" loop flow summers and indicators, which are set to indicats forward flow correctly.

The 0.95 factor is the result of a conservative analysis to appropriately modify the single-tap flow coefficient for reverse flow.* If a more exact, less conservative core flow measurement is required, special in-reactor cal-libration tests would have to be made. Such calibration tests would involve calibrating core support plate AP versus core flow during two-pump operation along the 100% flow control line, operating on one pump along the 100% flow control line, and calculating the correct value of C based on the core flow derived from the core support plate 3P and the loop flow indicator readings.

i 2.1.2 Core Flow Uncertainty Analysis The uncertainty analysis procedure used to establish the core flow uncertainty for one-pucp operation is essentially the same as for two-pump operation, except for some extensions. The core flow uncertainty analysis is described in Refer-ence 2. The analysis of one-pump core flow uncertainty is summarized below.

For single-loop operation, the total core flow can be expressed as follows (refer to Figure 2-1):

WC=WA-WI where WC = total core flow, Wg = active loop flow, and

.WI = inactive loop flow.

By applying the " propagation of errors" method to the above equation, the vari-ance of the total flow uncertainty can be approximated by:

'The expected value of the "C" coe fficient is #0.88.

2-2

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W W -a '#W 1-a '

W C I sys A rand \ rand I where c = uncertainty of total core flow; WC eg = uncertainty systematic to both loops; sys eg ,

= random uncertainty of active loop only; cg = random uncertainty of inactive loop only; e = uncertainty of "C" coefficient; and c

a = ratio of inactive loop flow (W ) to active loop flow (W )g .

7 Based on an uncertainty analysis, the conservative, bounding values of cg , og , eg7 and o are 1.6%, 2.6%, 3.5%, and 2.8%, respectively, C

sys Arand rand Based on the above uncertainties and a bounding value of 0.36 for "a," the variance of the total flow uncertainty is approximately:

e = (1.6)2 , 0

[(3.5) + (2.8) ] = (5.0%)2 1-0.36 (2.6[+(1-0 6 When the effect of 4.1% core bypass flow split uncertainty at 12% (Bounding case) bypass flow fraction is added to the above total core flow uncertainty, the active coolant flow uncerta,inty is:

' ' 0*12 o'active =

(5.0%)' + 1-0.12 (4.1%)2 = (5.0%)2 coolant which is less than the 6% core flow uncertainty assumed in the_ statistical i

analysis. In summary, core flow during one-pump operation is measured in a conservative way, and its uncertainty has been conservatively evaluated.

2-3 m s

NEDO-24229-1 2.2 TIP READING UNCERTAINTY

' To ascertain the TIP noise uncertainty for single recirculation loop operation, a test was performed at an operating BWR. The test was perfor ed at a power level 59 3% of rated with a single recirculation pump in operation (core flow 46.3% of rated). A rotationally sy==etric control rod pattern existed prior to the tes t .

i-

'Five consecutive traverses were made with each of five TIP =achines, giving a total of 25 traverses. Analysis of their data' resulted in a nodal TIP noise of 2.85%. Use of this TIP noise value as a ec=ponent of the TIP total uncertainty results in a One-sig=a total uncertainty value for single-loop operation of 9.1%

for reload cores.

24 L -

NED0-24229-1 CCRE l

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k l / W C

WA l

l Wg = TOTAL CORE FM W

A

= ACTIVE LOOP FLOW W, = INACTIVE LOOP FLOW Figure 2-1 II .ustration of Single Recirculation Loop Operation Flows 2-5/2-6

r NED0-24229-1 3 MCPR OPERATING LIMIT 3.1 CORE-WIDE TRANSIENTS Operation with one recirculation loop results in a maximum power output which is 20% to 305 below that which is attainable for two-pump operation. There fore ,

the consequences of abnormal operational transients from one-loop operation will be considerably less severe than those analyzed from a two-loop operational mode.

For pressurization, flow decrease, and cold water increase transients, previously transmitted Reload /7SAR results bound both the thermal and overpressure conse-quences of ene-loop operation.

Figure 3-1 shows the consequences of a typical pressurization transient (turbine trip) as a function of power level. As can be seen, the consequences of the transient during one-loop operation are considerably less because of the asso-ciated reduction in operating power level.

The consequences from flow decrease transients are also bounded by the full power analysis. A single' recirculation pump trip fecm one-loop operation is less severe than a two-pucp trip frem full power because of the reduced initial pcwer level.

Cold water increase transients can result from either recirculation pump speedup or restart, or introduction of colder water into the reactor vessel by events such as loss of feedwater heating. The Kr factors are derived assuming that both recirculation icops increase speed to the maxi =um permitted by the M-G set scoop -

tube position. This condition produces the maximum possible power increase and hence maximum ACfR for transients initiated from f.ess than rated power and flow.

When operating with only one recirculation loop, the flow and power increase associated with the increased speed on only one M-G set will be less than that associated with both pumps increasing speed; therefore, the Kr factors derived with the two-pump assumption are conservative for single-loop operation. Inad-vertant festart of the idle reirculation pump would result in a neutron flux transient which would exceed the flow reference scram. The resulting transient with scram is less severe than the rated power / flow case documented in NEDE-24011-P-A. The loss of feedwater heating is generally the most severe cold water increase event with respect to increase in core power. This event is caused by positive reactivity insertica from core flow inlet subcooling; 3-1

.J

NED0-24229-1 ,

therefore, the event is primarily dependent on the initial power level. For low power conditions experienced during single-loop operation, the higher the initial power level, the greater the CPR change during the transient. Since the initial power level during one-pump operation will be significantly lower, the cold water increase transients during single-loop operation are conservatively bounded by the full power (two-pu=p) anclysis.

From the above discussicas, it can be concluded that the transient consequences from one-loop operation are bounded by full power analysis.

32 ROD WITHDRAWAL ERROR The rod withdrawal error at rated power is given in the FSAR for the initial core and in cycle d raendent relcad supplemental submittals. These analyses are per-formed to demonstrate that, even if the operator ignores all instrument indica-tiens and the alarm which could occur during the course of the transient, the red block system will stcp rod withdrawal at a minimum critical power ratio which is higher than the fuel cladding integrity safety limit. Correction of the rod block equation (below) and lower initial power for single-loop operation assures that the MCPR safety limit is not violated.

One-pump operation results in backflow through 10 of the 20 jet pu=ps while the flew is being supplied into the icwer planum from the 10 active jet pumps.

Because of the backficw through the inactive jet pumps, the present rod block equation was censervatively modified for use during cne-pump operation because the direct active-loop flew measurement may not indicate actual flow above about 357 recirculation drive flow without correction.

A procedure has been established for correcting the red block equation to account for the discrepancy between actual ficw and indicated flow in the active loop.

This preserves the original relationship between rod block and recirculation drive flow when operating with a single loop.

The two-pump rod block equation is:

RB = mW + RB100 - m(100 3-2

NEDD-24229-1 The one-pump equatien becomes:

RB = mW + (RB100 - m(100)) -Am W where AW = difference between two-loop and single-loop ef fective recirculation drive flow at the same core flow; SB = power at rod block in %;

3= flow reference slope for the rod block monitor (RSM), and W = recirculation drive flow in % o# rated.

RB3 co = top level rod block at 100% flew.

If the rod block setpoint (RB100) is changed, the equation must be recalculated using the new value.

The APRM trip settings are flow biased in the same manner as the rod block monitor trip setting. Therefore, the APRM rod block and scram trip settings are subject to th's same procedural changes as the rod block monitor trip setting discussed above.

33 OPERATING MCPR LIMIT For single-loop operation, the rated condition steady-state MCPR limit is increased by 0.01 to account for the increase in the fuel cladding integrity safety limit (Section 2). At lower flows, the steady-state operating MCPR limit is conservatively established by multiplying the rated flow steady-state ' limit by the Kr factor. This ensures that the 99.9% statistical limit requirement is always satisfied for any postulated abnormal operational occurrence.

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Figure 3-1 Main Turbine Trip with Bypass Manual Flow Control 34

NED0-24229-1

! 4 STABILITY ANALYSIS L__._--__ _. . _ _ - . . - . .

The least stable power / flow condition attainable under normal conditions occurs ,

1 i at natural circulation with the control rods set for rated power and flow. This j condition may be reached following the trip of both recirculation pumps. As shown in Figure 4-1, operation along the minimum forced recirculation line with one pump running at minimum speed is more stable than operating with natural cir-culation flow only, but is less stable than operating with both pumps operating l

at minimum speed. During single-loop operations, the flow control should be l in = aster manual to avoid control oscillations in the recirculation flow control f system under abnormal conditions.

I Since the stability analyses for each reload are performed at natural circulation c C icions, the reload stability results are bounding for single-loop operation.

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N'EDO-2422 9-1

._1.2 ULTIVATE STABILITY LIMIT 1.0 - -- ------ - -=

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- - BOTH LOCPS, PUMPS MINIMUM SPEED 0.8 -

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Figure 4-1. Decay Ratio Versus Power Curve for Two-Loop and Single-Loop Operation 1

4-2

NED0-24229-1

5. ACCIDENT ANALYSES _. ___ _

The broad spectrum of postulated accidents is covered by six categories of design basis events. Ihese events are the less-of-coolant , recirculation pump seizure, control rod drop, main steamline break, refueling, and fuel assembly loading acci-dents. The analytical results for the loss-of-coolant and recirculation pump seizure accidents with one recirculation pump operating are giver. below. The results of the two-loop analysis for the other four events centioned conserva-1 tively bound one-pump operation.

5.1 LOSS-OF-COOLANT ACCIDENT ANALYSIS A single-loop operstien analysis utilizing the models and assu=ptiens documented in Reference 3 was performed for each Peach Bottom unit. Using this method, SAFE /REFLCOD computer code runs were made for a full spectru= of break sizes for both the suction and discharge side breaks. Because the reflood minus uncovery time for the single-locp analysis is similar to the two-loop analysis, the Maxi-( mum Average Planar Linear Heat Generation Rate (MAPLHGR) curves currently applied

! to each unit can be modified by derived reduction factors for use during one recir-culation pump operation.

5.1.1 Break Spectrum Analysis A break spectrum analysis for each unit was performed using the SAFE /REFLCCD computer codes and the assumptions given in Section II.A.7.3.2 of Reference 3 The suction and discharge break spectrum reflood times for one recirculation loop operation are compared to the standard previously performed two-loop opera-tien in Figures 5-1 and 5-2, respectively, for Unit 2. Suction and discharge break spectrum reflood time comparisons for Unit 3 are shown in Figures 5-3 and 54 The uncovered time (reflood time minus recovery time) for the Unit 2 dis-charge and suction break spectrum and the Unit 3 discharge and suction break spectrum is compared in Figures 5-5, 5-6, 5-7, and 5-8, respectively.

For the Unit 2 standard two-loop analysis, the most limiting break was a 100%

discharge DBA with a total uncovered time shown in Figure 5-5 and boiling tran-sition times less than 17.6 sec for 7x7 fuel and all 8x8 fuel.

5-1

NEDO-24229-1 f

The 66% discharge DBA for Unit 2 two-loop analysis has a greater total uncovered time, but the later boiling transition time (about 4 sec later) and later uncovery time more than compensate for this longer total uncovered time. Thus, the 100% discharge DBA is limiting.

For Unit 2 single-loop analysis, a boiling transition time of 0.1 see is conser-satively assumed for all breaks larger than 1.0 ft , and the reflooding times i and total uncovered times are similar to the two-loop analysis. Because of the conservatively assumed 0.1 see boiling transition time, the most limiting break for single-loop analysis is the 66% discharge DBA. The single-loop reflooding time is less than the two-loop reflooding time for this break.

i Since the single-loop reflooding times for Unit 2 are less than or very close to two-loop reflooding times, the procedure described in Section II.A.7.4 of Reference 3 may be applied.

For the Unit 3 standard two-loop analysis, the most limiting break is also a 100% discharge DBA with a total uncovered time shown in Figure 5-7 and a boiling ,

transition time of less than 17.6 see for 7x7 fuel and all 8x8 fuel.

The 66% discharge DBA for Unit 3 two-loop analysis has a greater total uncovered time, but the later boiling transition time (about 4 sec later) and later uncovery time more than compensates for this longer total uncovered time. Thus, the 100% discharge DBA is limiting.

For Unit 3 single-loop analysis, a boiling transition time of 0.1 see is conser-vatively assumed for all breaks larger than 1.0 ft , and the reflooding times and total uncovered times are similar to the two-loop analysis. Because of the conservatively assumed 0.1 see boiling transition time, the most limiting break for single-loop analysis is the 66% discharge DBA. The single-loop reflooding time is less than the two-loop reflooding time for this break. .

. Since the single-loop reflooding times for Unit ) are less than or very close to two-loop reflooding times, the procedure described in Section II. A.7.4 of Refer-ence 3 may be applied.

l 5-2

NEDO-26229-1 5.1.2 Single-Loop MAPLHGR Deter =ination The small differences between single and two-loop operation in uncovered ti=e and reflood time for the limiting break size in both Units 2 and 3 would result in a less than 200F increase in the calculated peak cladding te=perature. There fore ,

as noted in Reference 3, the one- and two-loop SAFE /REFLOOD results can be considered similar and the generic alternative procedure described in Section II.A.7.4 of this reference was used to calculate the MAPLHGR reduction facters for single-loop operatien.

j MAPLHGR reduction factors were determined for the cases given in Table 5-1. The

=cs: limiting reducticn factors for each fuel type are shown for both units in Table 5-2. One-loop operation MAPLHGR values are derived by cultiplying the current two-loop cperation MAPLHGR values by the reduction facter fer that fuel type. As discussed in Reference 3, single recirculatien loop MAPLHGR values are conservative when calculated in this =anner.

5.1.3 Small Break Peak Cladding Temperature Section II.A.7.4.4.2 of Reference 3 discusses the s=all sensitivity of the calcu-lated peak clad temperature (PCT) to the assumptions used in the one-pu=p opera-tien analysis and the duration of nucleate boiling. As this slight increase

(*50cF) in PCT is overwhel=ingly offset by the decreased MAPLHGR (equivalent to 3000 to 5000F ~ PCT) for one-pu=p operation, the calculated PCT values for s=all breaks will be wcll below the small break PCT values previously reported and significantly below the 22000F 10CFR50.46 cladding temperature limit.

5.2 CNE-PUMP SEIZURE ACCIDENT The ene-pump seizure Occident is a relatively mild e;4nt during two recirculatien pump operation as documented in References 1 and 2. Similar analyses were per-fermed 'to determine the impact this accident would have en one recirculation pu=p operation. These analyses were performed with the =odels documented in Reference 1 for a large core BWR/3 plant (Reference 4). The analyses were initialized from steady-state operation at the following initial conditions, with the added condi-tien of one inactive recirculation loop. Two sets of initial conditions were assu=ed:

5-3

!!EDO-24229-1 (1) Thermal Power = 75% and core flow = 58%

(2) Thermal Power = 82% and core flow = 56%

These conditions were chosen because they represent reasonable upper limits of single-loop operation within existing MAPLHGR and MCPR limits at the same maxi-zum pu=p speed. Pump seizure was simulated 'by setting the single operating pump speed to zero instantaneously.

The anticipated sequence of events following a recircul? tion pump seizure which occurs during plant operation with the alternate recirculation loop out of ser-vice is as follows:

(') The recirculation loop flow in the loop in wnich the pump seizure occurs

! drops instantaneously to :ero.

i (2) Core voids increase which results in a negative reactivity inser tion and a shar p decrease in neutron flux.

(3) Heat flux drops more slowly because of the fuel ttne constant.

(4) Neutron flux, heat flux, reactor water level, steam flew, and feedwater flow all exh,1 bit transient behaviors. However, it is not anticipated that the increase in water level will enuse a turbine trip and result in scram.

It is expected that the transient will terminate at a condition of natural circu-lation and reactor operstion will continue. Thste will also be a small decrease in system pressure.

The minimum CPR for the pump seizur's accident fer the large core BWR/4 plant was determined to be greater than the IJel cladding integrity safety limit; therefore, no fuel failures were postulated to occur as a result of this analyzed event.

These results are also applicable to Peach Bottom Units 2 and 3.

5-4

NEDC-24229-1 Table 5-1 MAPLHGR MULTIPLIER CASES Unit Fuel Type Cases Calculated 2 7x7, 8x8, 100% DBA Suction Break 8x8R, and P8x82, 100% DBA Discharge Break and LTA 66% DBA Discharge Break

• 3 7x7, 8x8, 1005 DBA Suction Break l

8x8R, and P8x8R, 100% DBA Discharge Break and PTA 665 DBA Discharge Break

• l l

'Most limiting break.

Table 5-2 LIMITING MAPLHGR REDUCTICN FACTORS Unit Fuel Type Reduction Factor 2 7x7 0.71 l 8x8 0.82 8x8R and P8x8R 0.80 l

LTA 0.79 3 7x7 0.71 8x8 0.83 i 8x8R and P8x8R 0.81 PTA 0.81 l

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6. REFERENCES

1. Generic Reload Fuel Application, General Electric Company, August 1979 (NEDE-24011-P-A).

2. General Electric BWR Thermal Analysis Basis (GETAB): Data, Correlation, and Design Application, General Electric Company, January 1977 (NEDO-10958-A) .

3 General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFR50 Appendix K Amendment No. 2 - One Recirculation Loop Out-of-Service, General Electric 2cmpany, Revision 1, July 1973 (!!EDO-20566-2) .

4 Enclosure to Letter #TVA-BFNP-TS-117, O. E. Gray III to Harold R. Denton,

September 15, 1978.

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6-1/6-2

NUCLEAR ENERGY DIVISIONS

• GENERAL ELECTRIC COMPANY SAN JOSE. CALIFORNIA 95125 GENERAL h ELECTRIC TECHNICAL INFORMATION EXCHANGE TITLE PAGE AUTHOR SUBJECT TIE NUMBER

?. H. Henriksen Nuclear Science DATE and Technology May 1980 TITLE U " "#88 Peach Bottom Atomic Power I Station Units 2 and 3 UU"' * ""*" # C **

Single Loop Oper2ti.On _

REPRODUC 8LE CCPY FILED AT TECHNICAL NUMBER OF PAGES SUPPORT SERVICES. R&UO. SAN JcSE. ,0 CALIFORNIA 951:5 (Mail Code 211) l

SUMMARY

The capability of operating at reduced power with a l

single recirculation loop is highly desirable, from a plant availability / outage planning standpoint, in the event maintenance of a recirculation pu=o or I

ether ccmoonent renders one loop inoperative. To justify single-icop oceration, the safety analyses documented in the Final Safety Evaluation Reports were reviewed for one-puro coeration. Increased

! uncertainties in the core total flow and TIP readings l resulted in an 0.01 incremental increase in the MCPR l

fuel cladding integrity safety limit during single-loop ocerati6n. This 0.01 increase is reflected in I the MCPR operating limit. No other increase in l this limit is recuired as core-wide transients I are bounded by the rated power / flew analyses per-l formed for each cycle, and the recirculation-flow-rate-deoendent rod block and scram setpoint f

l equations given in the technical specifications l are adjusted for one-pump operation.

l l

By cutting out this rectangle and folding in half, the above information can be fit 1md into a standard card file.

DOCUMENT NUMBER NEDO-24229 l

1 iNFORMATION PREPARED FOR Nuclear Power Systems Division

. SECTION Safety and Licensing Operation BUILOING AND ROOM NUMBER K Rm. 2604 M AIL CODE 682 m o' o 3~

l J6e oJuuS..trum