ML20041E472
| ML20041E472 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 05/31/1981 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML18025B735 | List: |
| References | |
| 81NED264, NED-24236, NEDO-24236, NUDOCS 8203100581 | |
| Download: ML20041E472 (28) | |
Text
NEDO-24236 81NED964 CLASSI MAY 1981 BROWNS FERRY NUCLEAR PLANTS UNITS 1,2, AND 3 SINGLE-LOOP OPERATION i
GENER AL h ELECTRIC sao w l
0,203100301- -m
NED0-24236 81NED264 Class I May 1981 BROWNS FERRY NUCLEAR PLANTS UNITS 1, 2, AND 3 SINGLE-LOOP OPERATION NUCLEAR POWER SYSTEMS DIVISION e GENERAL ELECTRIC COMP,iNY SAN JOSE, CALIFORNI A 95125 GEN ER AL h ELECTRIC
NED0-24236 IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT (Please Read Carefully)
This report was prepared by General Electr1c solely for the Tennessee l
Valley Authority (TVA) for TVA's use with the U.S. Nuclear Regulatory.
Commission (USNRC) for supporting TVA's operating license of.the Browns L
Ferry Nuclear Plants Units 1, 2 and 3.
The information contained in I
this report is believed by General Electric to be an accurate and true representation of the facts known, obtained or provided to General Electric at the time this report was prepared.
The only undertakings of the General Electric Company respecting infor-mation in this document are contained in the General Electric Company Single-Loop Operation Analysis Proposal 414-TY37-ERO (GE letter No.
G-ER-9-34, dated April 27, 1979) and TVA Purchase Contract 79P64-164185, dated August 31, 1979. The use of this information except as defined by said proposal and contract, or for any purpose 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 representation or warranty (express or implied) as to the completeness, 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, i
1 l
11
I l
CONTENTS fage 1.
INTROEUCTION AND SUIMARY 1-1 f
2.
MCPR FUEL CLADDING INTEGRITY SAFETY LIMIT 2-1 2.1 Core Flow Uncertainty 2-1 2.2 TIP Reading Uncertainty 2-4 3
MCPR OPERATING LIMIT 3-1 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 5-9 6.
REFERENCES 6-1
/
4 s.
iii/iv
NEDO-24236 ILLUSTRATIONS Figure Title Payge 2-1 Illustratica 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 Browns Fer ry 1 Suction Break Spectrum Reflood Times 5-3 5-2 Browns Ferry 1 Discharge Break Spectrum Reflood Times 5-4 o
5-3 Browns Ferry 2 Suction Break Spectrum Reflood Times 5-5 5-4 Browns Ferry 2 Discharge Break Spectrum Reflood Times 5-6 5-5 Browns Ferry 3 Discharge Break Spectrum Reflood Times 5-7 5-6 Browns Ferry 3 Suction Break Spectrum Reflood Times 5-8 TABLES Table Title Page 5-1 Limiting MAPLHGR Reduction Facters 5-2 r
a v/vi
. /
NEDO-24236 1.
INFMIDUCTIM AND SINSIARY The current technical specifications for the Browns Ferry Nuclear Plants, Units 1, 2, and 3 do not allow plant operation beyond 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if an idle recirculation loop cannot be roturned to service. Technical Specification 3.6.F.3 for each unit provides that if the pump cannot be nede operable after this period of time, the plant must be placed in hot shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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 takes longer than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and one loop is rendered inoperative. To justify single-loop operation, the safety analyses documented in the Final Safety Evaluation Reports and Reference 1 were reviewed for one-pump operation.
Increased uncertainties in the core total flow and TIP readings result in an 0.01 incremental increase in the MCPR fuel cladding integrity safety limit during single-loop operation.
These uncertainties are compensated for by adding 0.01 to the MCPR operating limit for single-loop operation. 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 specifications are adjusted for one-pump operation. The least stable power / flow condition (natural circulation) is not af fected by single-loop operation. Under single-loop operation, the flow control should be in manual, since control oscillations may occur in the recirc-ulation flow control system under abnormal conditions. Derived MAPLHCR reduction factors for single recirculation pump operation are tabulated below:
Fuel Type 7x7 8x8 8x8R P8x8R 0.70 0.83 0.82 0.82 The analyses were performed assuming the equalizer valve was closed, in accordance with normal valve lineup for operation at Browns Ferry. The discharge valve in the idle recirculation loop should be operable and able to close in order to main-tain the LOCA mitigating systems for Browns Ferry. Alternatively, suction valves in the idle recirculation loop may be closed to prevent the loss of low pressure coolant injection (LPC1) flow out of a postulated break in the idle suction line.
1-1/1-2
NEDO-24236 2.
MCPR FUEL CLADDIEG 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 Table 5-1 of Reference 1 for reloads. A 6% core flow measurement t:ncertainty 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 operating plant test results given in Subsection 2.2 below. This revision resulted in a single-loop operation process computer uncertainty of 9.1% for reload cores. A comparable two-loop process computer uncertainty value is 8.7% for reload cores. The net effect of these two revised uncertainties is a 0.01 incremental increase in s
the requimd MCPR fuel cladding integrity safety limit.
2.1 CORE ELOW UNCERTAIETY 4
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 banks of jet pumps are in forward flow; total core flow is t.he 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 punp 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:
l
[ActiveLoop Inactive Loop Total Core
-C
(
Flow /
(Indicated Flow)
(Indicated Flow /
2-1
NEDO-24236 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."
If a more exact, less conservative core flow measurement is required, special in-reactor cali-bration 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 bnsed 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-puun operation is essentially the same as for two-pump operation, except for some extensions. The core flow uncertainty analysis is described in Befer-ence 2.
he 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 l
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:
"The expected value of the "C" coe fficient is r0.88.
2-2
NEDo-24236 (1-a)2 esA. c a )2 se er/
esc = e3 1
e 1
WIrand sys 4 -a i
rand where uncertainty of total core flow; UWg
=
uncertainty systematic to both loops; Og
=
random uncertainty of active loop only; og
=
random uncertainty of inactive loop only.
ogI
=
rand uncertainty of "C" coefficient; and OC
=
ratio of inactive loop flow (W ) to active loop flow (W )
- I A
a
=
Resulting from an uncertainty analysis, the conservative, bounding values U
Of UW e
WArand' WIrand 3y3 respectively.
Based on the above uncertainties and a bounding value of 0.36 for a "a," the variance of the total flow uncertainty is approximately:
I
~
Wc = (1.6)2 [ 0.36 (2.6)2 0.36 (3.5)2 + (2.8)2
= (5.0%)2 1
2 e
(1-0.36j (1-j 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 uncertainty is:
= (5.0%)2 0.12 (4.1%) = (5.0%)2 O
ctive
_,2 coolant which is less than the 6% core flow uncertainty assumed in the statistical analysis.
2-3
NEDO-24236 In summary, core flow during one-pump operation is measured in a conservative way, and its uncertainty has been conservatively evaluated.
2.2 TIP READIEG UNCERTAINTY 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 level 59.3% of rated with a single recirculation pump in operation (core flow 46.3% 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 j
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 9.15 for reload cores.
t l
1 2-4
NEDO-24236 CORE
(
i
\\
i
/
C W g A
TOTAL CORE FLOW W
C ACTIVE LOOP FLOW W
=
A W,
INACTIVE LOOP FLOW
=
Figure 2-1.
Illustration of Single Recirculation Loop Operation Flows 2-5/2-6
EPR OPERATIBG LIMIT 3.1 core WIDE TRANSIENTS Operation with one recirculation loop results in a maximum power output which 4.s about 30% below that which is attainable for two-pump operation. Therefore, 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 the tran-sient during 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 thmi a two-puwp 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-G set scoop tube position.. This condition produces the maximum possibla power increase and hence ruaximum ACPR 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.
Inadvertant restart of the idle recirculation 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 heating, is gener-l ally the most severe cold water increase event with respect to increase in t
core power. This event is caused by positive reactivity insertion from core l
l 3-1 I
l
NEDO-24236 flow inlet subcooling; therefore, the 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 conser-vatively 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 analyses.
3.2 R(B 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 performed to demonstrate that, even if the operator ignores all instrument indications and the alarm 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 than the fuel cladding integrity safety limit. Correction of the rod block equation (below) and lower power assures 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 punps.
Because of the backflow through the inactive jet pumps, the present rod block equation was conservatively modified for use during one-pump operation because the direct active-loop flow measurement may not, without correction, indicate actual flow above about 35% total drive flow.
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 preserves the original relationship between rod block and actual effective drive flow when operating with a single loop.
The two-pump rod block equation is:
RB = mW + (RB100-m(100))
3-2
l i
NEDO-24236 The one-pump equation becomes:
RB = mW + (RB100-m(100))-MAW where AW = difference, determined by utility, between two-loop and single-loop effective drive flow when the activo loop indicated flow is the same; RB = power at rod block in 5; m = flow reference slope for the rod block monitor (RBM), and W = drive flow in 5 of rated.
RB100 = top level rod block at 100% flow.
If the rod block setpoint (RB100) is changed, the equation must be recalculated using the new value.
The APRM trip settings are flow biasad 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 above.
33 OPERATING MCPR LIMIT For single-loop operation, the rated condition steady-state MCPR limit is in-creased 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 require-ment is always satisfied for any postulated abnormal operational occurrence.
3-3
NEDD-215236 t
1100 1140 -
5 200 %
1120 -
E 5
l N
5 m
E 1100 g
a
=
Sz N
100
.E 1080 3
w u.
5 z
0 E
$w a.
d 1000 fz 8
e w
E<
z W
E E
< 1040 E
1020 2000 980 RANGE OF EXPECTED 7
4 MAXIMUM 1 LOOP POWER OPERATION l
I I
I gg O
20 40 60 80 100 120 140 POWER LEVEL (% NUCLEAR BOILER RATED)
Figure 3-1.
Main Turbine Trip with Bypass Manual Flow Control 3 14
NEDO-24236 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 pungs.
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. Under single-loop operation, the flow control should be in manual, since control oscillations may occur in the recirculation flow control system under abnormal conditions.
k 4-1
l 1.2 ULTIMATE STABILITY LIMIT 1.0 ------
= - = = SINGLE LOOP, PUMP MINIMUM SPEED
- BOTH LOOPS, PUMPS MINIMUM SPEED 08
^o n
Ey 0.6 tr 6
NATURAL R ATED FLOW w
CIRCULATION CONTROL
/
LIN E LINE#y' 0.4
/
HIGHEST POWER
/
ATTAINA8 LE FOR SINGLE LOOP OPERATION 0.2 i
l I
I g
O 20 40 60 80 100 POWER (%)
Figure 4-1.
Decay Ratio Versus Power Curve for Two-Loop and Single-Loop Operation 4-2
l NEDO-24236 l
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 pum seizure, control rod drop, main steamline break, refueling, and fuel assembly loading accidents. The analytical results for loss-of-coolant and recirculation pug seizure accidents with one recirculation pump operating are given below. The results of the two-loop analysis for the last four events conservatively bound one-pump operation.
5.1 IASS-OF-COOLANT ACCIDENT ANALYSIS Single-loop operation analyses utilizing the models and assumptions documented in Reference 3 were performed for Browns Ferry Units 1, 2, and 3 Using this method, SAFE /REFLOOD computer code runs were made for both the suction and dis-charge side breaks.
Figures 5-1 through 5-6 show the variation of reflooding time for the standard two-pump operation analysis and the one-pump operation analysis. The plots show that for Browns Ferry Unit 1, the one-pump operation analy sis indicates reflooding times that are similar to the reflooding times in two-pump operation.
Since the results are similar, the multipliers for the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) for Browns Ferry Unit 1 were determined for the discharge and suction break Design Dasis Accidents (DBA's) and for discharge breaks of maximum reflooding times as described in Section II.A.7.4 of Reference 3 The one-loop operation MAPLHGR's for each type of fuel can be determined by multiplying the standard two-loop operation MAPLHGR's by the appropriate multiplier. As discussed in Reference 3, the one-loop MAPLHGR's calculated in this manner result in conservative values.
For Browns Ferry Units 1, 2 and 3, the most limiting break for single-loop opera-tion was found to be a discharge line break at 65% of the DBA. For Units 2 and 3, the reflooding time was longer compared to the two-loop operation. Sinc e,
for these two units, the reflooding times for the limiting breaks were considered significantly longer than the two-loop reflooding times, an analysis of MAPLHGR using the approved CHASTE model was performed for each unit. The analyses proved the procedure described in Section II. A.7.4 of Reference 3 still is 5-1 b
NEDO-24236 ccriservatively applicable to all three units, and analyses for each type of fbel were made at 1005 DBA suction and discharge breaks and at 655 DBA discharge breaks using the methods of Section II.A.7.4 of Reference 3 The limiting MAPLHGR reduction factors are provided in Table 5-1.
The analyses were done assuming the equalizer valve was closed and with the LPCI modified configuration. The discharge valve in the idle recirculation loop is normally closed or operable, but if its closure is prevented, the suction valvc in the loop should be closed to maintain normal system alignment for low pressure coolant injection (LPCI).
l Table 5-1 LIMITING mal'LHGR REDUCTION FACTORS Fuel Type Reduction Factor 7x7 0.70 8x8 0.83 8x8R 0.82 P8x8R 0.82 l
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NEDO-24236 150 0
- 1. LOOP OPERATION O 2-LOOP OPER ATION 140 130
-5 i
P E 120 5
8a b
a:
110 100 i
l i
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I go 20 30 40 50 60 70 80 90 100 110 BREAK AREA t% OF DBA)
Figure 5-3 Browns Ferry 2 Suction Break Spectrum Reflood Times 5-5 I
290 280 0 1 LOOP OPERATION 270 O 2. LOOP OPERATION 200 250 2<o l
44J 3*
220 1
0 1
E LTo 3
210 e
a:
200 190 180 l
170 160 150 g
140 50 00 70 80 90 1%
110 BREAK AREA t% OF DBA)
Figure 5-4 Browns Ferry 2 Discharge Break Spectrum Reflood Times 5-6
NEDO-24236 290 280 -
270 -
260 -
0 1-LOOP OPER ATION 2%
O 2-LOOP OPER ATION 240 230 I
220 2*
a E
210 Q
8 g
m 1
x 190 180 170 too 150 140 2
I I
[
l I
3, so so 70 80 90 100 110 BREAK AREA (% OF DBA)
Figure 5-5.
Browns Ferry 3 Discharge Break Spectrum Reflood Times 5-7 L_ - _
NEDD-24236 150 0 1. LOOP OPERATION O 2.tOOP OPERATION 140 tw
_E E
P o 120 E
8 9
bc 110 100 I
I I
I I
I l
I 90 20 30 40 50 60 70 80 90 100 110 BREAK AREA t% OF OBA)
Figure 5-6.
Browns Ferry 3 Suction Break Spectrum Reficod Times 5-8 j
NEDo-214236 S.1.1 Small Break Peak Cladding Temperature Section 11.A.7.4.4.2 of Reference 3 discusses the relative insensitivity of the calculated peak clad temperature (PCT) to the assumptions used in the one-pump operation analysis and the duration of nucicate boiling. As this slight increase
(#500F) in PCT is overwhelmingly offset by the decreased MAPLHGR (equivalent 0
to #300 to 500 F PCT) for one-pump operation, the calculated PCT values for 0
small breaks will be significantly below the 2200 F cladding temperature limit specified in 10CFR50.46.
5.2 ONE-PUW SEIZURE ACCIDENT The one-pug seizure accident is a relatively mild event during two recirculation pump operation as documented in References 1 and 2.
Similar analyses were per-formed to determine the impact this accident would have on one recirculation pump operation. Rese analyses were performed with the models documented in Reference 1 for Browns Ferry Unit 1 (Reference 4).
The analyses were initialized from steady-state operation at the following initial conditions, with the added condition of one inactive recirculation loop:
(1) thermal power = 75% and core flow = 585; and (2) thermal power = 82% and core flow = 565.
These conditions were chosen because they represent reasonable upper limits of single-loop operation within existing MAPLHGR and MCPR limits at the same maximum pump speed. Pug seizure was simulated by setting the single operating pump speed to zero instartaneously.
The anticipated sequence of events following a recirculation pug seizure which occurs during plant operation with the alternate recirculation loop out of ser-vice is as follows:
1.
he recirculation loop flow in the loop in which the pump seizure occurs drops instantaneously to zaro.
2.
Core voids increase which results in a negative reactivity insertion and a sharp decrease in neutron flux.
3 Heat flux drops more slowly because of the fuel time constant.
5-9
4.
Neutron flux, heat flux, reactor water level, steam flow, and feedwater flow all exhibit transient behaviors. However, it 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 cir-culation and reactor operation will continue. There will be a small decrease in system pressure.
The minimum CPR for the pump seizure accident for Browns Ferry Unit 1 was deter-mined 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.
These results are applicable to Browns Ferry Units 2 and 3
\\
5-10
(
NED& 24236 6.
REFERENCES 1.
General Reload Fuel Application, General Electric Company, August 1979 (NEDE-24011-P-A-1).
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 Company, Revision 1, July 1978
( NEDO-20566-2 ).
4.
Enclosure to TVA Letter, J. E. Gilleland to T. A. Ippolito, dated September 28, 1978.
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