ML20203H787

Ei Hatch Nuclear Plant Units 1 & 2 Single-Loop Operation

ML20203H787
Person / Time
Site:	Hatch
Issue date:	08/31/1979
From:	GENERAL ELECTRIC CO.
To:
Shared Package
ML20203H785	List: ML20203H780 ML20203H787
References
79NED289, NEDO-24205, TAC-61900, TAC-61901, NUDOCS 8608050092
Download: ML20203H787 (33)

Text

. - . . _. .-. - - - _ - -

• NEDO-24205 79NED289 CLASSI AUGUST 1979 n l

)

l l

l l EDWIN 1. HATCH NUCLEAR PLANT UNITS 1 AND 2 l

SINGLE-LOOP OPERATION l

l  :

0 jeA*ASSESIS!!!

GENER AL h ELECTRIC

w-NEDO-24205 79NED289 Class I August 1979 EDWIN I. HATCH NUCLEAR PLANT UNITS 1 AND 2 SINGLE-LOOP OPERATION NUCLE AR POWER SYSTEMS DIVISION e GENER A L E LECTRIC COMPANY SAN JOSE, CA LIFORNIA 96125 GENERAL $ ELECTRIC L. - . . - . - - . - . - - - - - - - ----- - - ~ - ~ ~ ~ ~ ^ ~ ~ ~ ~

....y i

f i

l i

DISCLAIMER OF RESPONS181LITY This document was prepared by or for the General Electric Company. Neither the General Electric Company not any of the contnbutors to this document:

A. Makes any warranty or representation, express or implied, with respect to the accuracy. completeness, or usefulness of the information containedin this docu-ment, or that the use of any information disclosed in this document may not unfringe privately owned rights; or B. Assumes any responsibility for loability or damage of any kind which may result

from tha use of anyinformation disclosed in this document.

f i

I i

l l

l l

l l

l l

l I

i l

l l

[

L2 l

NEDO-24205 TABLE OF CONTENTS Page 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 2-4 1

3. MCPR OPERATING LIMIT 3-1 1 3.1 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. ACCIDENT ANALYSES 5-1

! 5.1 Loss-of-Coolant Accident Analysis 5-1 5.2 One-Pump Seizure Accident 5-3

6. REFERENCES 6-1 lit /iv

t NEDO 24205 LIST OF ILLUSTRATIONS i

Figure Title Page 2-1 Illustration of Single Recirculation Loop Operatien 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 Hatch 1 Suction Break Spectrum Reflood Times 5-6 5-2 Hatch 1 Discharge Break Spectrum Reflood Times 5-7 5-3 Hatch 2 Suction Break Spectrum Reflood Times 5-8 5-4 Hatch 2 Discharge Break Spectrum Reflood Times 5-9 5-5 Hatch 1 Discharge Break Spectrum Uncovered Times 5-10 5-6 Hatch 1 Suction Break Spectrum Uncovered Times 5-11 5-7 Hatch 2 Discharge Break Spectrum Uncovered . Times 5-12 5-8 Hatch 2 Suction Break Spectrum Uncovered Times 5-13 J

LIST OF TABLES Table Title Page 5-1 MAPLHGR M11tiplier Cases 5-5 5-2 Limiting MAPLHGR Reduction Factors 1

5-5 I

v/vi

NBDO-24205

1. INTRODUCTION AND SIMfARY The current technical specifications for the Edwin I. Hatch Nuclear Plant Units 1 and 2 do not allow plant operation beyond a relatively short period of time if an idle recirculation loop cannot be returned to service. Unit 1 (Technical Specification 3.6.J) is permitted 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of operation with a single recircula-tion pump. If the pump cannot be made operable af ter this period of time, the plant must be placed in cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Unit 2 (Techni-cal Specification 3.4.1.1) is allowed 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of operation with one recircula-tion pump arti 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to place the plant in a hot shutdown condition should the secorvi pump not be made operable.

The capability of operating at reduced power with a single recirculation loop is highly desirable, from a plant availability / outage planning standpoint, in the event maintenance of a recirculation pump or other component renders one loop inoperative. To justify single-loop operation, the safety analyses documented in the Final Safety Evaluation Reports and Reference 1 were reviewed 3

for one-pump operation. Increased uncertainties in the core total flow and TIP readings resulted 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-rate dependent rod block and somm setpoint equations given in the technical specifications are adjusted for one-pump operation. The least stable power / flow condition, achieved by trip-l ping both recirculation pumps, is not affected by one-pump operation. Derived MAPUIGR reduction factors are 0.83, 0.85, and 0.75 for the 7x7, 8x8, and 8x8R fuel types in Unit 1 and 0.84 for Unit 2.

's i

J J

1-1/1-2

NED0-24205

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 cladding integrity safety limit are not dependent on 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 65 core flow measurement uncertainty has been established for single-loop operation (compared to 2.5% for two-loop operation). As shown below, this value conservatively reflects the one standard deviation (one sigma) accuracy of the core flow measurement system documented in Reference 2. The random noise component of the TIP reading uncertainty was revised 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 process computer uncertainty of 6.8% for initial cores and 9.1% for reload cores. Comparable two-loop process computer uncertainty values are 6.3% for initial cores and 8.7% for reload cores. The net effect of these two revised uncertainties is a 0.0l 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. For 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 measured flow in the active loop. In add ition , 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), . ( Active Loop ) )

-C l (I""

• O E l

j i

Plow 4 (Indicated Flowj ( Flow j l 2-1

NED0-24205 where C (= 0.95) is defined as the ratio of " Inactive Loop True Flow" to "Inac-tive Loop Indicated Flow," and " Loop Indicated Flow" is the flow indicated by the jet pump " single-tap" loop flow summers and indicators, which are set to indicate forward flow correctly.

The 0.95 factor was the result of a conservative analysis to appropriately modify the single-tap flow coefficient for reverse flow.s If a more exact, less conservative core flow is required, special in-reactor calibration 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 AP and the loop flow indicator readings.

2.1.2 Core Flow Uncertainty Analysis The uncertainty analysis procedure used to establish the core flow uncertainty for one-pump 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 aingle-loop operation, the total core flow can be expressed as follows (refer to Figure 2-1):

WC WA-WI where WC = total core flow, WA = active loop flow, and WI = inactive loop (true) flow.

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

"Ihe expected value of the "C" coefficient is f0.88.

2-2

NED0-24205 2 I 2 f a 2

%C 1

NA 2+ NI 2

1-a) 1-aj

- where ogC = un ertainty in total core flow (%),

WA = un ertainty in active loop flow (5),

wy = uncertainty in inactive loop flow (5), and a = WI/WA -

The uncertainty of ogA was analyzed to be 2.85. A conservative, bounding value of 3.0% was used for AW in the total flow uncertainty variance calculation. The uncertainty, ogy is comprised of the uncertainty in the "C" coefficient and ran-dom uncertainties such as jet pump AP measurement uncertainty and instrumentation errors. The bounding value of 3.75% for og y was used in the detemination of agC*

Based on the above uncertainties and a bounding value of 0 36 for a, the variance of the total flow uncertainty is approximately:

WC

• 1-0.36 3.0 M + ,0

- 6 (3.755)2

= (5.0%)2, When the effect of 4.1% core bypass flow uncertainty at 125 (bounding case) by-pass flow fraction is added to the above total core flow uncertainty, the active coolant flow uncertainty is:

O.

ctive = (5.0%)2 (4,39)2 = (5 7%)2 coolant which is less than the 65 core flow uncertainty assumed in the statistical analy sis .

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

2-3 i

NEDO-214205 i

2.2 TIP READING UNCERTAINTY 1

To ascertain the TIP noise uncertainty for single recirculation loop operation, a test was performed at an operating BWR. The test was performed at a power i

level 59.3% of rated with a single recirculation pug in operation (core flow 146.3% of rated). A rotationally symmetric control rod pattern existed prior to the test.

Five consecutive traverses were made with each of five TIP machines, giving a total of 25 traverses. Analysis of their data resulted in a nodal TIP noise of 2.855. Use of this TIP noise value as a component of the process computer total uncertainty results in a one-sigma process computer total uncertainty value for single-loop operation of 6.8% for initial cores and 9.1% for reload cores.

I i

1 1

I 4

4 I

2 li

NEDO-24205 CORE l

l L >

k WC

/

Wg A

WC =

TOTAL CORE FLOW WA =

ACTIVE LOOP FLOW a

W, INACTIVE LOOP FLOW Figure 2-1 nlustration of Single Reciret:1ation Loop Operation Flows 2-S/2-6

NEDO-21:205

3. MCPR OPERATING LIMIT 3.1 CORE WIDE TRANSIENTS Operation with one recirculation loop results in a maximum power output which is 20% to 30% 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 /FSAR results bound both the thermal and overpressure conse-quences of one-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 one-loop operation are considerably less because of the associated reduction in operating power level.

The consequences from flow decrease transients are also bounded by the full power analysis.

A single pump trip from one-loop operation is less severe than a two-pump trip from full power because of the reduced initial power 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 heater. The Kr factors are derived assuming that both recirculation loops increase speed to the maximum permitted by the M-0 set scoop tube position.

This condition produces the maximum possible power increase and hence maximum AMCPR for transients initiated from less 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 restart of the idle reirculation pump would result in a neutron flux transient which would exceed the flow reference scram. The resulting scram is expected to be less severe than the rated power / flow case documented in the FSAR.

The latter event, loss of feedwater hosting, is generally the most severe cold water increase event with respect to increase in core power.

This event is caused by positive reactivity insertion from core flow inlet subcooling; therefore, the 3-1

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

NEDO-24205 event is primarily dependent on the initial power level. 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 one-pump cold water increase case is conservatively bounded by the full power (two-pump) analysis .

From the above discussions, it can be concluded that the transient consequence from one-loop operation is bounded by previously submitted 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 dependent reload supplemental submittals. These analyses are per-formed to demonstrate that, even if the operator ignores all instrument indica-tions and the alann which could occur during the course of the transient, the rod block system will stop rod withdrawal at a minimum critical power ratio which is higher tnan the fuel cladding integrity safety limic. Correction of the rod block equation (below) and lower power t.ssures that the MCPR safety limit is not violated.

One-pump operation results in backflow through 10 of the 20 jet pumps while the flow is being supplied into the lower plenum from the 10 active jet pumps.

Because of the backflow through the inactive jet pumps, the present rod block j equation was conservatively modified for use during one-pump operation because l the direct active-loop flow measurement may not indicate actual flow above about l 2

i 33% drive flow without correction.

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

This proaerves the original relationship between rod block and actual effective

{ drive flow when operating with a single loop. ,

i The two-pump rod block equation is:

RB mW + [RBjoo - m(100) ]

i l

l 3-2 i

, _ . . _ . . _ _ _ _ . _ _ _ _ . _ _ . , ~ , _ _ _ _ _ _ _ , _ _ _ . _ . , _ . _ _ _ _ _ _ . . . _ _ _ , _ _ , _._ _. _ _ _ _ _

NEDO-216205 The one-pump equation becomes:' .

RL = mW + [RBjon - m(100)) - mM i

where i

i AW = difference, determined by utility, between two-loop and single-loop effective drive flow when the active loop indicated flow is the same; RB = power at rod block in %;

i m = flow reference slope for the rod block monitor (RBM), and W = drive flow in 5 of rated.

3 RBjoo = top level rod block at 100% flow.

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

4 l

1 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 the same procedural changes as the rod block monitor trip setting discussed i above.

1 3.3 OPERATING MCPR LIMIT I

i For single-loop operation, the rated condition steady-state MCPR limit is j

', increased by 0.01 to account for the increase in the fuel cladding integrity safety limit (Section 2).

i 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 i

always satisfied for any postulated abnomal operational occurrence.

1 I

1 i

3-3

- - _ _ . _ _ . . _ . . - _ _ . _ - ._ _ _ _ _ _ _ _....___- . . ~ _ _ _ _ _ _ _ _ _ _ _ _ _ - _

NEDo-24205 1160 1140 -

5 1120 -

200 g e

5 d

2 iim -

i 3

= 8 z

a 5

100 s 1080 -

E o 5 8 p:

E 5 w

=

giOc0 w

[

w

$ 5 5 e a -

$ 1040 -

E 1020 1000

+

w 980 -

RANGE OF EXPECTED  ? <

MAXIMUM 1 LOOP POWER OPERATION i I l 1 go 0 20 43 60 80 100 120 140 POWER LEVEL (% NUCLEAR BOILER RATED)

Figure 3-1 Hsin Turbino Trip with Bypass ntnual Flow Control 3-4

NEDO-24205 4 STABILITY ANALYSIS The least stable power / flow condition attainable under normal conditions occurs at natural circulation with the control rods set for rated power and flow. This 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 at minimum speed. Because of the increased flow fluctuation during one-recircu-lation-loop operation, the flow control should be left in manual operation to preclude unnecessary wear on the automatic controls.

r i

I.

i f

i l

i i

i i

4-1

NEDO-24205 1.2 ULTIMATE STABILITY LIMIT 1.0 - - - - - - - - - - - - - - - - - *

---

SINGLE LOOP, PUMP MINIMUM SPEED

- ammmmm. BOTH LOOPS, PUMPS MINIMUM SPEED 0.8 -

~o b

(

c:

0.6 -

6 NATURAL R ATED FLOW w CIRCULATION  ! CONTROL U

LINE /e LINE

//

/

04 -

/

HIGHEST POWER ATTAINABLE FOR SINGLE

1. LOOP OPERATION 0.2 -

0 0 20 40 60 80 100 POWER N Figure 4-1 Decay HatiO Versus Power Curve for Two-Loop and Single-Loop Operation 4-2

NEDO-24205

5. ACCIDENT ANALYSES The broad spectrum of postulated accidents is covered by six categories of design basis events. These events are the loss-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 given below. The results of the two-loop analysis for the last four events are conservatively applicable for one-pump operation.

5.1 LOSS-OF-COOLANT ACCIDENT ANALYSIS A single-loop operation analysis utilizing the models and assumptions documented in Reference 3 was performed for each Hatch unit. Using this method, SAFE /REFLOOD computer code runs were made for a full spectrum of break sizes for both the auction and discharge side breaks. Because the reflood minus uncovery time for the single-loop analysis is similar to the two-loop analysis, the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) curves currently applied to each unit were modified by derived reduction factors for use during one recirculation pump operation.

5.1.1 Break Spectrum Analysis A break spectrum analysis for each unit was performed using the SAFE /REFLOOD 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-tion in Figures 5-1 and 5-2, respectively, for Unit 1. Suction and discharge break spectrum reflood time comparisons for Unit 2 are shown in Figures 5-3 and 5-4.

The uncovered time (reflood time minus recovery time) for the Unit 1 dis-charge and suction break spectrum and the Unit 2 discharge and suction break spectrum is compared in Figures 5-5, 5-6, 5-7, and 5-8, respectively.

For Unit 1, the maximum uncovered time for the standard two-loop analysis is 205.40 seconds occurring at 80% of the DBA discharge break which is the most limiting break for two-loop operation. For the singlo-loop analysis, the maxi-mum uncovered times are 206.10 seconds at 60% DBA discharge, and 205.45 seconds 5-1

NEDO-24205 l

at 80% DBA discharge break. These uncovered times are almost equal (only 0.65 l second difference). Hence the larger break (80% DBA) will be more limiting because of the earlier uncovery and corresponding higher decay heat during the uncovered period. Consequently, for both the single- and two-loop analysis, the limiting break is the 80% DBA discharge break.

For Unit 2 the maximum uncovered time for the standard two-loop analysis is 219.9 seconds occurring at 86.5% DBA discharge break which is the most limiting break for two-loop operation. For single-loop analysis, the maximum uncovered times are 221.7 seconds at 87.5% DBA discharge break, and 220.9 seconds at 86.5% DBA discharge break. These uncovered times are almost equal (less than 1 second difference) and at 86.5% DBA discharge break the reflood time for single-loop analysis shows a less than 1 second difference when compared with the most limit-ing break for the standard two-loop analysis.

Comparison of the suction and discharge break spectrum reflood times between the single- and two-loop analysis for both units shows that the reflood times are similar. For the suctien break spectrum, the reflooding times for one-loop operation are within 1 second of the two-loop operation reflooding times. In the discharge break spectrum, the single-loop reflood times are approximately equal to or less than the two-loop reflood times for breaks greater than 70% DBA.

5.1.2 Single-Loop MAPLHGR Determination The sna11 differences in uncovered time and reflood time for the limiting break si e in both Units 1 and 2 would result in a less than 200F increase in the cal-culated peak cladding temperature. Therefore, as noted in Reference 3, the one-and two-loop SAFE /REFLOOD results can be considered similar and the generio alternattve procedure described in Section II.A.7.4 of this reference was used to calculate the MAPLHGR reduction factors for single-loop operation.

MAPLHGR reduction factors were determined for the cases given in Table 5-1. The most limiting reduction factors for esch fuel type is shown for beth units in Table 5-2. Slightly longer calculated boiling transition time for the Hatch 1 8x8R fuel required use of curve 3 on Figure II. A.7.4-1 of Reference 3 rather than curve 2 used for the other fuel types and for Hatch 2. One-loop operation MAPLHOR values are derived by multiplying the current two-loop operation MAPLHOR values 5-2

m . . _ _ _ _ .__ _ _ - _ _ _ . .___._ _ _ _ . _ _ _ _ _ .

NEDO-24205 i

by the reduction factor for that fuel type. As discussed in Reference 3, single recirculation loop MAPLHGR values are conservative when calculated in this manner.

! 5.1 3 Small Break Peak Cladding Temperature 4

Section II.A.7.4.4.2 of Reference 3 discusses the anall sensitivity of the calou-lated peak clad temperature (PCT) to the assumptions used in the one-pump opera-tion analysis and the duration of nucleate boiling. As this slight increase I

(J500F) in PCT is overwhelmingly offset by the decreased MAPLHGR (equivalent to 3000 to 5000F # PCT) for one-pump operation, the calculated PCT values for small 1

breaks will be well below the 12600F and 14600F small break PCT values previously i reported for Units 1 and 2, respectively, and significantly below the 22000F l j 10CFR50.46 cladding temperature limit.

i i

! 5.2 ONE-PUMP SEIZURE ACCIDENT '

1 The one-pump seizure accident is a mlatively mild event during two recirculation  ;

i pump operation as documented in References 1 and 2. Similar analyses were per-k formed to determine the impact this accident would have on one recirculation pump operation. These analyses were perfonned with the models documented in Reference 1 for a large core BWR/4 plant (Reference 4). The analyses were initialized from 1

steady-state operation at the following initial conditions, with the added condi-tion of one inactive recirculation loop. Two sets of initial conditions were

{ assumed:

{ 1.

Thermal Power = 755 and core flow = 58%

i l' 2. Thermal Power = 825 and core flow = 565 1'

These conditions were chosen because they represent reasonable upoer limits of 4

single-loop operation within existing MAPLHOR and MCPR limits at tha same maxi-mum pump speed. Pump seizure was simulated by setting the single operating pump k sped to zero instantaneously.

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

l 5-3 U---- -- -- - ----- -

NEDO-24205

1. The recirculation loop flow in the loop in which the pump seizure occurs drops instantaneously to zero.

2 Core voids increase which results in a negative reactivit/ insertion and a sharp decrease in neutron flux.

3. Heat flux drops more slowly because of the fuel time constant.

4 Neutron flux, heat flux, reactor water level, steam flow, and feedwater flow all exhibit transient behaviors. However, Lt is not anticipated that the increase in water level will cause a turbine trip and result in scram.

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

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

5-4

NEDo-24205 1

Table 5-1 j

MAPLHGR MULTIPLIER CASES Unit Fuel Type Cases Calculated

. 1 7x7 1005 DBA Suction Break j 100% DBA Discharge Break 80% DBA Discharge Break i

8x8 80% DBA Discharge Break' 8x8R 80% DBA Discharge Break' 2 8x8R 100% DBA Suction Break 100% DBA Discharge Break 86.5% DBA Discharge Break

1. Most limiting break for 8x3 and 8x8R fuel, Unit 1 1

1 1

Table 5-2 4

LIMITING MAPLHGR REDUCTION FACTORS Unit Fuel Type Reduction Factor 1

7x7 0.83 i

8x8 0.85 8x8R 0.75 2 8x8R 0.84 5-5

l lll' l1l ll l1 lll \\ ll1llI ll ,llll l,1'll 1ll . !l1 I

s zrpTZ8*

e 0

9 0 1

N N OO I 0 TI TI 9 A A

- R R E E

. P P OO P P OO OO L- -L 2 1 0

OO i 8

s e

m i

T d

o o

l 0 f O l 7 e R

m

) u A r B t D c e

( p A S 0 E I

6 R k A a K

e r

A B E

R n B o i

t c

u I

0 S 5

1 h

c t

a

- H I 0 1 4 -

5 e

r g

u i

F

- I 0 3

~ - - - - 0 _

2 0 0 0 1

1 m

i m 8 7

?51G oo3 b*

y* _

11

• li ,'l  ! l:I l ' 1l 'l1' I l;11- l .

e rE ? Ewo*

0 I 1 1

NN OO I I TT AA RR EE PP OO 0 PP I 0 OO 1 OO LL- -

/

21 '

/

OO /

/

lIlII hT 0 9 s e

~ m i

T

- U d

o I 0 o

_ 8 l

_ f

- _ e R

m

) u A r B t I 0 D c N 7

(

e p

N A S E

\ R k

\ A a K

e A

r A I 0 6

E R

B B

e g

/ h r

a

/ c s

i 0

D I

5 1

- h c

- t 0 H a

I 0 .

4 2

5 e

r

_ u

- g I 0 i 3 F

. I 0 2

0 0 0 o 6 5 4 0 0 2 2 2 n 2 2

1 2 " 0 9

1

?[2 OO3my T~

) ; j!' l ll 1 i . 1'! 1 jilI llli ,

! l

I NEDO-24205

\

k<

\

h _

s

\

\ .

2 2 6 99

EE: \\ O OO g [

8 oo x e 8

e t

\ E 8.

5

• i fs

, 4 8 i e I

s -

e t N

ka b

A A

a 8 E

C I 1 I o B 2 8 8 8 (oas) 3W110001338 5-8

ll l.ll il z@o8Y8u a 0 0

1 g

N I 0

\ 9

\

s e

m

/ i T

a_ d o

o I

0 l 8 f e

R m

u r

t NN ) c A e OO w

I B p TT D S AA  %

RR ( k a

OOI EE PP PP OO I 0 E 7 R A

A B e

r

- oO K A

e

- tL 21 E

g r

R a

- OO B h

- c s

- i D

2 h

o, I 0

6 t c

a I

/ I 4

5

/ e

/ r u

/ i g

/ i 0 5

F

/

f

/

y

/

0

- - 0 8 0 0 0 0 4 7

0 0 0 0 g 2 2  ! 4 3 2 2 2 2 2

2 1

2 g

_ 3 w2[ooiE u6

(

2M8Ls- ou 0

2 1

NN O O I

I TT AA RR 0 E E 1 P P I 1

OO P P

OC OO L L -

21 O0 0 s

, 0 e m/ I 1

i m

/ T

/ d

/ e

/ r

1. e 0 v 1 \[ h l 9 o iI$ c il]IjlI II n U

m u

r t

0 c l

8 e p

) S A

B k D a

% e

( r A B 0 E l

7 R e g

A K r A a E h

% R c

% B s i

D b l 0

6 1

/, h

' c t

/ b

' I

/ 0 .

l 5

5 5

e r

u D g i

0 l

4 F i

0 3

- - - - - - 0 2

0 0 0 0 0 0 0 g 1 0 9 8 7 6 5 2 2 1 1 1 1 1 efE$w58aj w4O l

= @ P 2' &v, o

o f

, 0

9

~  :

y E 9 m

" 1 T

- 4,

, p g

y y o y

y o

c n

y

/ r g

t c

p e

0 S g

g y g a e

Y E R r A B

' K n

- A E

i t

o c

I O S u S

/ 1 h

c t

a

/ M 1 0

• 4 6,

5

~ e P

~

D I

~

P

\

LL I 0 3

\

m ~ ~ - 0 2

0 o 0 7 @ = a 3 ga*o$r85 f=

N$ ~&

0

• 0 1

\

N y 0 9

\

\

s e

m

/ T A 0 4,

e r

y 8 v o

c m

u r

t g c g e O D l

3 A k I 0 E a 7 R e A P K E A

E '

g r

a

• • • h c

• s i

y g

N y 7 *

• h c

O / t a

I T

A f H R g E g

/

P p OO / 7_

P P OO

/ S O D L A

/ e 2 ' / r u

O / O R

I

/ I S '

/

/

/

/

p _

~ , , , *

- ~  %

• 0 0 0 0 0 o 0 0 s

4 2

3 2

2 2

2 ~ 9 1 1 7

1 6

1 1% a oy

,LN

1?$8w 5

0 0

1

/

• /

l 0

8

/ s

/ e m

f i T

c/ d e

r

/ e v

/ o c

/O l 0 U

n 6

m u

r

/ )

A B

t c

/ D e

p

( S A k E

R a

/ A e r

/ K A

B E

R n B o i

l 0 t 4 c u

S 2

h c

t a

N N H 01 O I T T .

A A 8 R A -

E E P P 5 OO e P P 0 OO l

2 r a

OO t g

L- -L 2 1 i

F OO

" o s 0 4

0 iE 555 TEY;

NEDO-24205 8

~

\

\

k

\

\

k\ -

8 e

\ a

\s N #

\ 8 NO s 8

\  !

N g t N e m 8.

3 x N i E N u s 5 c 5 3 3 -

e t

\ s N

.c 0

e 55 55 m' 55 E OO o C mm 1 -

~ c

&2 = u.

OO

• r P

• e l- l l o

R' 8 8 e (25) 3Wil 03H3 AODNO 5-13/$-14

NED0-24205

6. REFERENCES

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

2. General Electric BWR Theriaal 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 Company, Revision 1, July 1978 (NEDO-20566-2).

4 Enclosure to Letter #TVA-BFNP-TS-117, O. E. Gray III to Harold R. Denton, Sep tember 15, 1978.

6-1/6-2

NUCLEAR ENERGY DIVISIONS e GENERAL ELECTRIC COMPANY SAN JOSE. CALIFORNIA 95125 GENER AL h ELECTRIC TECHNICAL INFORMATION EXCHANGE TITLE PAGE AUTHOR SUBJECT TIE NUMBER 79NED289 Nuclear Science g ,7g J. Charnley and Technology July 1979 TITLE GE CLASS Edwin I. Hatch Nuclear Plant I Units 1 and 2 Single-Loop GOVERNMENT CLASS Operation -

REPRODUCISLE COPY FILED AT TECHNICAL NUMBER OF PAGES SUPPORT SERVICES, R&UO. SAN JOSE.

CALIFORNIA 96125 (Mail Code 211) 30

SUMMARY

The capability of operating at reduced power with a single recirculation loop is highly desirable, from a plant availability / outage planning standpoint, in the event maintenance of a recirculation pump or other component renders one loop inoperative. To iustify single-loop operation, the safety analyses documented in the Final Safety Evaluation Reports were reviewed for one pump operation. Increased uncertainties in the core total flow and TIP readings resulted in an 0.01 incremental increase in the MCPR fuel cladding integrity safety limit during single-loop operation.

This limit.

0.01 increase is reflected in the MCPR operating 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-rate-dependent rod block and scram setpoint equations given in the technical specificationn are adjusted for one-pump operation.

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

DOCUMENT NUMBER NEDO-24205 INFORMATION PREPARE D FOR ear Power hstems MWsion SECTION _ Safety and Licensing Operation BulLDING AND ROOM NUMBER K, Rn. 2606 M AIL CODE 682 WEO-914 (6/77)