ML19318A827
| ML19318A827 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 06/10/1980 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19318A823 | List: |
| References | |
| NUDOCS 8006240208 | |
| Download: ML19318A827 (19) | |
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 76 TO FACILITY OPERATING LICENSE NO. DPR-57 GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA l
EDWIN I. HATCH NUCLEAR PLANT UNIT NO. 1 1
DOCKET NO. 50-321 i
Introduction i
By letter dated March 22, 1979, as supplemented by letters dated May 11 and 16, June 4,1979, and February 28, 1980, Georgia Power Company (the licensee) requested an amendment to the Technical Specifications appended to Facility Operating License No. DPR-57 for the Edwin I. Hatch Nuclear Plant Unit No. 1.
The request involved Cycle 4 operation of Hatch 1.
4 The licensee's analysis for Cycle 4 operation included the effect of an End-of-Cycle Recirculation Pump Trip (E0C-RPT) as a supplement. to the Reactor Trip System.
On August 6,1979, we issued Amendment No. 69 to DPR-57 for Cycle 4 operation.
However, that amendment included a restriction on Minimum Critical Power Ratios (MCPR) pending the comple-tion of our review of hardware implementation of the E0C-RPT feature.
This amendment completes our actions on the licensee's March 22, 1979, application and involves:
(1) removal of the restriction on MCPR, and (2) addition of limits and surveillance of the EOC-RPT feature.
Evaluation Various transient events can reduce'the MCPR from its normal operating level. To assure that the fuel cladding integrity Safety Limit MCPR would not be violated during any abnormal operational transient, the most limiting transients were reanalyzed by the licensee. The analysis presumed the beneficial effect of the E0C-RPT initiated by turbine stop valve closure or control valve fast closure.
Our detailed evaluation of the licensee's transient analysis results was provided in the Safety Evaluation supporting-Amendment No. 69 to DPR-57 and is incorporated i
herein by reference. That evaluation supports a single MCPR limit for-each type fuel for the entire cycle assuming an acceptable hardware implementation of the E0C-RPT.
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. The acceptability of the Hatch 1 E0C-RPT feature was reviewed for the NRC staff by Lawrence Livermore Laboratory (LLL). The basis for acceptance
's documented in " Technical Evaluation of the End-of-Cycle Recirculation Pu1p Trip" which is incorporated herein by reference.
We have reviewed the report and agree with its conclusion that the E0C-RPT feature for Hatch 1 meets the criteria of IEEE Std-279-1971, IEEE Std-323-1974, and General Design Criteria 13, 20 through 24, and 29 of Appendix A to 10 CFR 50 and is therefore an acceptable design. 41owever, during our review, we identified certain changes to the licensee's proposed Technical Specifi-cations that should be made to more clearly identify limiting conditions for operation (LCOs) action statements and surveillance requirements.
These changes involved:
(1) inclusion of action statements for inoperable E0C-RPT channels in Table 3.2-9 vice the licensee's recommendation for inclusion in Table 4.2-9; (2) designation of response time testing as a calibration vice a functional test; and (3) minor editorial changes.
Each of these NRC staff recommended changes was discussed with the licensee and he agreed.
We also discussed with the licensee the LLL conclusion that circuit breaker time-respoase tests should verify that the E0C-RPT response time is no greater than the response time used in the transient calculations for the corresponding fuel cycle. This verification includes a correlation bet-ween unloaded and loaded conditions for the EOC-RPT circuit breakers. The licensee agreed.
Based on the above, we find the design of the Hatch'l E0C-RPT system and the associated Technical Specifications, as amended by the NRC staff, are acceptable.
Environmental Considerations We have determined that the amendment does not involve a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact.
Having made this determination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact and pursuant to 10 CFR 551.5(d)(4) that an environmental impact state-ment or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of the amendment.
Conclusion We have concluded, based on the considerations discussed above, that:
(1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the amendment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Comission's regulations and the issu-ance of the amendmen'. will not be inimical to the common defense and security or to the health and safety of the public.
Dated: June 10,1980
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i TECHNICAL EVALUATION OF THE END -OF-CYCLE RECIRCULATION PUMP TRIP FOR EDWIN I. HATCH NUCLEAR PLANT UNIT I (Dacket.No. 50-321) i L. R. Peterson r
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ABSTRACT
' This report documents the technical evaluation of the end-of-
. cycle. recirculation pump trip for the Edwin I. Hatch Nuclear Plant Unit No. 1.
The review criteria are based on IEEE Std-279-1971, IEEE Std-323-1974, IEEE Std-338-1977, and General Design Criteria 13, 20 through 24, and 29 of the Code of Federal Regulations, Title 10, Part 50, Appendix A requirements for determing the acceptability of the proposed system.
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L FOREWORD This report is supplied as part of the Selected Electrical,
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Instrumentation, and Control Systems Issues (SEICSI) Program being con-ducted for the U. S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Division of Licensing, by Lawrence Livermore Laboratory, Engineering Research Division of the Electronics Engineering Department.
The L. S. Nuclear Regulatory Commission funded the work under the authorization entitled " Electrical, Instrumentation and Control System Support," B&R 20 19 04 031, FIN A-0231.
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4 il TABLE'0F CONTENTS 4
I Page 4
1.
- INTRODUCTION.
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2.
CORRESPONDENCE AND C0tDIUNICATIONS.
3 3.
DESIGN DESCRIPTION 5
7 4.
EVALUATION 5.
CONCLUSIONS 11 12
. REFERENCES 4
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i TECHNICAL EVALUATION OF THE END-OF-CYCLE RECIRCULATION PUMP TRIP FOR EDWIN I. HATCH NUCLEAR PLANT UNIT 1 (Docket No. 50-321)
L. R. Peterson Lawrence Livermore Laboratory, Nevada 1.
INTRODUCTION o
Georgia Power Company (GPC) by its letter dated May 11, 1979
[Ref. 1), requested approval of instellation of an end-of-cycle (EOC) recirculation pump trip (RPT) featur'e at Edwin I. Hatch Nuclear Plant, Unit 1 (Hatch 1) as an amendment to the Reload 3 Licensing Application for that unit which had been submitted March 22,1979 [Ref. 2].
As a result of NRC staff and consultant questions and the need for additional information from the licensee, NRC review of the EOC-RPT proposal was separated from the Reload-3 Licensing Application to prevent delay of plant restartup and fuel cycle-4 operation.
The EOC-RPT feature is installed to improve the thermal margin of a boiling. water reactor (BWR) near the end of each fuel cycle by reducing the ' severity of possible pressurization transients.
The two most limiting pressurization transients near the end-of-cycle would be produced by
- turbine trip without bypass and generator load rejection without bypass.
-The EOC-RPT rapidly cuts off power to the recirculation pump
=c tors during generator load rejection (turbine control valve fast closure) or turbine trip (stop valve closure).
This action results in a rapid reduction in recirculation flow and increases the core void content during a pressurization transient, thereby reducing the peak transient power and heat flux.
Operation of the E0C-RPT system reduces the change in reactor critical power ratio ( A CPR) that would be produced by a pressurization transient.
Analyses by General Electric Company (GE) indicate that adding the EOC-RPT feature will result in significant reduction in A CPR for pressurization transients involving turbine stop valve closures or turbine control valve closures with assumed bypass failure.
It should be noted that EOC-RPT is not related to the RPT that is associated with an anticipated transient without scram (ATWS-RPT).
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CORRESPONDENCE AND COMMUNICATIONS On Hay 11, 1979, Georgia Power Company amended its March 22, 1979, Reload-3 Licensing Application for Edwin I. Hatch Nuclear Plant Unit 1 (Refs. I and 2] to include the EOC-RPT feature.
A description of the Hatch 1 EOC-RPT [nef. 3] and proposed changes to the, Hatch-1 Technical Specifications.for operation with the EOC-RPT installed [Ref. 4] were included as enclosures in the thy 11, 1979, submittal.
4 As a result of discussions with GPC representatives the NRC staff requested further. information on the design logic and equipment qualification standards for the Hatch 1 EOC-RPT installation and requested a conparison of their EOC-RPT design with that of Brown's Ferry 1.
The Brown's Ferry 1 EOC-RPT had been epproved by NRC as the first design for an operating reactor where safety credit was given for the EOC-RPT protection feature.
The licensee's June 4, 1979 letter [Ref. 5] stated that the RPT logic design for. Hatch 1 was the same as that approved for Brown's Ferry 1.
GPC also stated that the equipment to be installed on Hatch I would be identical to that which is installed on Hatch 2.
The licensee stated that in their view the equipment is qualified, based on previous NRC approval of the Hatch 2 EOC-RPT design and that a comparison with the equipment installed on Brown's Ferry 1 was unnecessary.
The licensee also submitted an undated document, "Edwin I. Hatch Nuclear Plant Unit 1, Electrical Control and Tastrunentation Aspects of End of Cycle Recirculation Pump Trip" [Ref. 6].
NRC staff.and LLL consultant review of the Hatch 1 EOC-RPT plans found significant differences in RPT circuit design between the Hatch 1 and the Brown's Ferry 1 installations.
The most significant differences involved fusing-to provide reactor protective system isolation and equipment protection,
- neans of making the RPT systems inoperahle, and system status indications to the reactor operator.
Furthermore,-the proposed changes to the Technical Specifications were. judged to have an inappropriate EOC-RPT association
-ith A!WS-RPT, inadequate EOC-RPT surveillance requirements, inadequate definition of EOC-RPT system inoperable status, and inadequate limitations on ;lant operation when one or both EOC-RPT systems are inoperable.
A letter from NRC that detailed changes needed in the proposed Te:haical Specifications and requesting additional information on the Hatch 1 EGC-lPT trip breaker circuits and their operation was sent to the licensee August 17, 1979 [Refs. 7 and 8).
Georgia Power Company by its letter dated February 28, 1980
[P.ef. 9] submitted revised Ratch 1 Technical Specifications for operation with the EOC-RPT feature and responses to the NRC August 17, 1979 request for additional information on the Hatch 1 EOC-RPT installation.
A conference call on May 2,1980 (Ref.10] between the LLL consultant, the NRC ORB #3' Project Hanager, an NRC Plant Systems Branch reviewer, and CPC er.gineering staff discussed the fuse protection arrangement of the Hatch 1 ECC-lPT breaker control and trip circuits.
System time response measurements and verification that the EOC-RPT system is a Class lE installation were also discussed.
I On May 5, 1980, the NRC Project Manager for Hatch 1 informed LLL by telephone [Ref. 11] of Georgia Power Company's committment to make needed icprovements _in the protective fusing of the Hatch 1 EOC-RPT system at the first scheduled reactor downtime opportunity, to be no later than the next feel reload outage. i_
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DESIGN DESCRIPTION
-The generic design philosophy for the EOC-RPT feature is described
-in General Electric Company (CE) report NEDO-24119, Basis for Installation
.of Recirculation Pump Trip System, Brown's Ferry Nuclear Plant, April 1978
[Ref.-12].
The design of the Hatch 1 EOC-RPT installation is described in References 4 and 6 and shown on Hatch 1 drawing number H 17822 (Ref.13].
The EOC-RPT is part.of the reactor protection system (RPS) because it is an' essential supplement to the reactor scram system.
All components of the EOC-RPT system are Class lE.
To mitigate the pressurization transient that would be produced by a turbine trip without bypass or a generator load rejection without bypass the
-EOC-RPT is - required to quickly interrupt power to shut down both BWR coolant recirculation pumps when closure of all four turbine stop valves occurs, or when fast closure of all four turbine control valves occurs.
An EOC-RPT trip may occur, but is not required, when one turbine stop valve or one turbine contrcl valve remains open'.
To mitigate pressurization transient effects, the EOC-RPT aust shut down the recirculation pumps within approximately 175 ms after initial closure movement of either the turbine stop valves or the turbine control valves.
The.EOC-RPT installation is composed of sensors that detect closure of the turbine stop-valves or fast closure of the turbine control valves combined with relays,- logic circuits, and fast-acting circuit breakers that interrupt the-current ~from the recirculation pump notor generator set generators to the recirculation pump motors.
When the RPT breakers trip open,lthe' recirculation pumps coast down under their own inertia.
Jro-satisfy-the reactorLprotection syst'em single-failure' criterion, the
. EOC-RPT has. two almost identical divisions.that actuate RPT in a one-out-
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of-two configuration.
Either of the two RPT divisions operates independent breakers in the supply. circuits of both recirculation pumps motors.
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Turbine stop valve closure is detected by four position switches that open when the associated stop valves are less than 90 percent open.
Turbine control valve fast closure is detected by four pressure switches in the hydraulic control system for the valves.
The pressure. switches open when the hydraulic control fluid pressure decreases below the trip level.
The stop valve position sensors and the control valve hydraulic pressure sensors for RPT are the same ones used in the reactor scram system to initiate scram when turbine stop valve closure or turbine control valve fast closure occurs.
The actuation of any RPT sensor causes an associated electro-magnetic relay to de-energize.
The contacts of these relays are combined in logic circuits with contacts from an operating bypass and contacts from a key-controlled manual bypass switch.
The logic cirguits control current to the trip circuits of the RPT circuit breakers.
The operating bypass disables the RPT system when turbine first-stage pressure is less than that for 30 percent reactor power.
The same operating bypass concurrently disables the turbine inputs to the scram' system.
A manual bypass switch allows each RPT division to be disabled and placed out of service for maintenance or testing.
The fast-closure sensors from each of two turbine control valves provide inputs to one RPT division and the sensors from the other two turbine control valves provide inpats to the second RPT division.
Similarly, the position switches from each of two turbine stop valves provide inputs to one RPT division and position switches from the other two stop valves provide inputs to the other RPT division.
The sensor relay contacts for each RPT division are arranged to form a two-out-of-two logic f or the' f ast closure of control valves and a two-out-of-two logic for closure of the stop valves.
The operation of either logic in an,RPT division will actuate the EOC-RPT feature.
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EVALUATION The EOC-RPT feature is part of the reactor protection system and is an essential supplement to the reactor scram function.
The EOC-RPT is required to comply with the criteria of IEEE Std-279-1971 [Ref. 23], IEEE Std-323-1974 [Ref. 24}, and IEEE Std-338-1977 [Ref. 25] and with General Design Criteria 13, 20 through 24, and 29 of 10 CFR 50, Appendix A [Ref. 26].
The EOC-RPT system at Edwin I. Hatch Nuclear Plant Unit 1 is similar to that previously approved by NRC for Browns Ferry, Unit 1.
Th.e two RPT divisions are' physically and electrically independent.
The sensors and relays providing inputs to the RPT systems originate from separate Class lE scram channels.
The signal channels are properly grouped and separated to provide independence between the corresponding scram chan-nels and the associated RPT divisions.
The sensor relays which actuate both the scram logic and RPT logic are fail-safe and will go to the tripped state on loss-of power or loss-of-input signal from each sensor.
The RPT circuit breaker control and trip circuits will not trip on loss of power and thus are not fail-safe.
The RPT circuit breakers that interrupt the current to the racirculation pump motors require power to actuate.
For this reason, the RPT logic circuits, control circuits, and trip circuits operate on 125 Vdc.
Each RPT division is supplied by a separate Class lE-rated 125 Vdc battery power supply with 30 amp inline fuses for the positive and negative lines from the battery supply.
A relay in each RPT division senses loss of power to the trip circuit in that division and actuates an "RPT Out of Service or Loss of Control Power" annunciator and alarn for that RPT division in the control room'.
In addi-tion, indicating lights are provided in the control room to monitor the trip coil circuits and the position of the trip breakers.
The NRC has previously found that this departure from fail-safe design is acceptable..
r To better meet IEEE 279 Section 4.7 Control and Protection System Interaction criteria GPC has committed to install branch fuses in the EOC-RPT breaker closing circuits.
These branch fuses will isolate the EOC-RPT breaker closing circuits from the breaker trip circuits so that a short cir-cuit in the elevating or closing functions of the breaker during reactor operation will not disable breaker trip actuation.
GPC plans to install this improvement during the next scheduled reactor downtime, to be no
'ater'than the next refueling outage.
We consider this satisfactory.
There is one other interconnection between each EOC-RPT division and a non-safety system.
When each RPT breaker trips, auxiliary relay con-tacts in the RPT breaker actuate a control circuit for the recirculation pump motor generator (M-G) set to de-energize the M-G set after the RPT breaker interrupts the current from the M-G set to the recirculation pump motor.
This interlock is adequately isolated so that no credible failure can prevent proper RPT action.
An operating bypass automatically disables the RPT system when the reactor is operating at less than 30 percent power.
The operating bypass is annunciated automatically in the control room.
Each RPT division can be bypassed manually by use of an out-of-service keyswitch which is administratively controlled.
Use of the out-of-service keyswitch bypass produces a suitable annunciator indication in the control room when the keyswitch is turned to the "RPT SYS INOP" position.
The proposed technical specifications for the Edwin I. Hatch Nuclear Plant Unit No. 1 provide suitable restrictions to limit operating power when one or both of the EOC-RPT divisions are inoperable.
Capability to check the R?T sensors and logic is provided by operating each valve, one at a time.
Lights across the relay contacts in the logie indicate proper operation at that point.
The RPT divisions do not need to be bypassed to' conduct such' tests.
During the periodic testing -
- f. :he scram logic, when two valves are operated simultaneously, the affected RPT division must be bypassed briefly to prevent RPT actuation.
The bypass is acconplished by use of the EOC-RPT system out-of-service hay switch during the scram-logic test.
The proposed technical specifications for Hatch 1 specify monthly fun:tional checks of the EOC-RPT initiate logic.
We consider monthly testing of the EOC-RPT input sensors and logic circuits to be adequate for providing timely indications of component failure.
Although the purpose of the RPT is to mitigate a core-wide pres-surization transient, the desired thermal margin advantage can be realized only if. the initiating events are sensed on an anticipatory basis, rather than by monitoring reactor pressure directly.
The use of pressure switches to sense the loss of hydraulic control fluid pressure to each turbine con-trol valve is adequate to anticipate fast closure of those valves.
Simi-larly, position switches set to trip at 90 percent open.will adequately anticipate closure of the turbine stop valves.
The EOC-RPT*is not given safety credit for any other initiating events.
To be effective l the RPT must be initiated almost immediately.
GE states that their analysis shows that manual initiation of a prompt trig of the recirculation pumps, at any reasonable point after the time when automatic action should have occurred, will not produce a significant inprovement on the situation.
The power to the recirculation pump motor-ger.erator sets can be tripped manually from the control room.
Therefore,
. provisions fot.anual initiation of the EOC-RPT feature are unnecessary.
The NRC has previously approved this position.
l The RPT feature is required to reduce recirculation pump flow after either the turbine control valees or the stop valves start closing, and within a delay time assumed in the transient calculations for that operating cycle.
The licensee has specified that the RPT circuit breakers
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will have 175 ms naxinun response time measured fron initial movement of turbine _ control' valves or turbine stop valv2s until the RPT breaker inter-rupts current to the recirculation pump motor.
The. remainder of EOC-RPT system response tine will include the time for recirculation pump coastdown after current to the pump is interrupted by the RPT breaker.
We concur with the surveillance requirement in the proposed Edwin I.' Hatch Nuclear Plant Unit 1 Technical Specifications that func-tional and time response tests of the EOC-RPT circuit breakers be con-ducted.once per operating cycle.
The GPC plan to test the breakers separately in an unicaded condition is satisfactory for the functional and time response tests provided suitable correlation is made with a reference measurement of response time under load.
The GPC procedure to do the time response tecting in section; and then add the resulting times of sensor response, logic response, and system action is acceptable.
To meet the criteria of IEEE Std-338-1977, Section 6.3.4, suitable cor-relation of these separate time response measurements must be made to verify that the actual overall EOC-RPT system response time is no greater than the response time in the. transient calculations for the corresponding operating cycle.
The time response tests and correlation to transient calculations should be made prior to each operating cycle.
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CONCLUSIONS Considering the separation, independence, and isolation of the two EOC-RPT divisions and their respective inputs, circuits, and power supplies, the EOC-RPT feature for the Edwin I. Hatch Nuclear Plant Unit 1 meets the criteria of IEEE'Std-279-1971, IEEE Std-323-1974, and General Design CriteriaE13, 20 through 24, and 29 of 10 CFR 50, Appendix A.
We reconnend approval of the EOC-RPT system design as submitted by the licensee with the installation of the branch fuses in the EOC-RPT breaker closing circuits to be added during the next scheduled reactor downtime as proposed by the licensee.
We also recommend approval of the proposed change for the addition of an EOC-RPT feature to the Hatch 1 Technical Specifications.
To fulfill the criteria of IEEE Std-338-1977, the separate neasurcaents nade during the EOC-RPT circuit breaker time-response tests oust be correlated suitably to verify that the EOC-RPT overall system response time is no greater than the response time in the applicable transient calculations for the corresponding operating cycle.
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o REFERENCES 1.
Georgia Power Company letter (C. F. Whitmer) to Director of Nuclear Reactor Regulation NRC/NRR dated May 11, 1979.
2.
Georgia Power Company., "Edwin I. Hatch Nuclear Plant, Unit 1, Reload-3 Licensing Application," March 22, 1979 3.
Georgia Power Company, "Edwin I. Hatch Nuclear Plant Unit 1 Recirculation Pump Trip," enclosure 2 to GPC letter dated May 11, 1979.
4.
Georgia Power Company, "NRC Docket 50-321, Operating License DPR-57, Edwin I. Hatch Nuclear Plant Unit 1, Proposed Changes to Technical Specifica-tions," enclosure 3 to GPC letter dated May 11, 1979.
t 5.
Georgia Power Company letter (C. F. Whitmer) to NRC/NRR dated June 4,1979.
6.
Georgia Power Company, ",Edwin I. Hatch Nuclear Plant Unit 1, Electrical Control and Instrumentation Aspects of End-of-Cycle (EOC)' Recirculation Pump Trip (RPTR)," undated, received by NRR/PSB reviewer July 3, 1979.
7.
G. Lainas (DOR /PSB) memorandum to T. Ippolito (DOR / ORB #3), " Hatch-1 End-of-Cycle Recirculation Pump Trip," August 14, 1979.
8.
Nuclear Regulatory Commission letter (T. Ippolito) to Georgia Power Company (C. F..Whitmer) dated August 17, 1979.
9.
Georgia Power Company letter (W. A. Widner) to Director of Nuclear Reactor Regulation NRC/NRR dated February 28, 1980.
10.
Conference. call between D. Verrelli (NRC), J. T. Beard (NRC), L. R.
Peterson 1
(LLL), R. Baker (GPC), and G. Doyle (GPC), May 2, 198C.
l I 1
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2 Peterson-(LLL) 11.-
Telephone conversation D. Verrelli (NRC) and L. R.
My 5, 1980.
General Electric Company, Nuclear Power Systems Division, " Basis ' for
- 12. -
Installation of Recirculation Pump Trip System, Brown's Ferry Nuclear -
Plant," report NEDO-24119, 78 NED 261, Class 1, April 1978.
13.
Bechtel, Southern Services Inc. drawing for Georgia Power Co.,
"Edwin I, Hatch Nuclear Plant Unit No.1, Reactor Protection System C71,. Elementary Diagram, Sh. 17 of 17," drawing number 10-502 H17822, dated May 1, 1978, Revision OC dated June 18, 1979.
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