ML20247F257
| ML20247F257 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 09/08/1989 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20247F226 | List: |
| References | |
| NUDOCS 8909180117 | |
| Download: ML20247F257 (12) | |
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SAFETY EVALUATION EY THE OFFICE _0F NUCLEAR REACTOR REGULATION EVALUATION OF COMPLIANCE WITH THE ATWS RULE: 10CFR50.62-REQUIREMENTS FOR REDUCTION OF RISK FROM ANTICIPATED TRANSIENTS i
WITHOUT SCRAM (ATWS) EVENTS FOR LIGHT-WATER COOLED NUCLEAR POWER PLANTS LOUISIANA POWER AND LIGHT COMPANY WATERFORD STEAM ELECTRIC STATION, UNIT 3 DOCKET NO. 50-382
1.0 INTRODUCTION
On July 26, 1984, the Code of Federal Regulations (CFR) was amended to include the "ATWS Rule" (Section 10 CFR 50.62, " Requirements for Reducticr of Risk from AnticipatedTransientsWithoutScram[ATWS]EventsforLight-Water-Cooleo t'uclear Power Plants"). An ATWS is an expected operational transient (such as loss of feedwater, loss of condenser vacuun., or loss of offsite power), which is accompanied by a failure of the reactor trip system (RTS) to shut down the reactcr. The ATWS Rule requires specific improvements in the design and uperation of comercial nuclear power facilities to reduce tre likelihood of failure to shut down the reactor following anticipated transients and to n.itigate the consequences of an ATWS event.
The 10 CFR 50.62 requirements applicable to pressurized water reacters manufactured by Combustion Engineering, such as Waterford, Unit 3, are:
(1)- Each pressurize water reactor must have equipment fror, serscr output to final actuation device that is diverse from the reactor trip system, which will autocratically initiate the auxiliary (or emergency) feedwater systen and initiate a turbine trip under conditions indicative of an ATWS. This equipment must be desi perform its function in a reliable manner and be independent (gned to from sensoroutputtothefinalactuationdevice)fromtheexisting reactor trip system.
(2) Each pressurized water reactor must have a diverse scram system from the sensor output to interruption of power to the control rods. This scram l
system must be designed to perform its function in a reliable manner and be independent from the existing reactor trip system (from sensor output l.
to interruption of power to the control rods).
In summary, the ATWS Rule requirements for Waterford are to install a diverse scramsystem(DSS))(thelicenseesubmittaldesignateditthediversereactor trip system (DRTS), diverse circuitry tc initiate a turbine trip (DTT) ano diverse circuitry for initiation of auxiliary feedwater (the licensee submittel cesigrated it the diverse emergency feedwater actuation system (DEFAS)).
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2.0 BACKGROUND
Paragraph (c)(6) of the ATWS Rule requires that detailed information to I
demonstrate compliance with the requirements of the Rule be submitted to the Director, Office of Nuclear Reactor Regulation (NRR).
In accordance with Paragraph (c)(6) of the ATWS Rule, Combustion Engineering Owners Group (CEOG) i provided information to the staff by letter dated September 18, 1985 (Ref. 1).
l The letter forwarded CEN-315. " Summary of the Diversity Between the Reactor Trip System and the Auxiliary Feedwater Actuation System for CE Plants," for staff review.
The staff reviewed CEN-315 and, by letter dated August 4, 1986 (Ref. 2),
forwarded its conclusion to the CEOG. The staff concluded that sufficient oiversity did nct exist between the reactor trip system (RTS) and the auxiliary feedwater actuatico system (AFAS). To achieve the degree of reduction in potential common mode failure (CMF), hardware diversity is required by the ATWS Rule. This decision affected San Onofre Nuclear Generating Station, Units 2 and 3 (SONGS-2, -3), Arkansas Nuclear One, Unit 2 (ANO-2), Palo Verde Units 1, 2 & 3, and Waterford Steam Electric Station, Unit 3 (WSEC-3).
In response to the staff's evaluation of CEN-315,,CE0G submitted CEN-349 to the staff by letter dated December 30, 1986 (Ref. 3). CEN-349 provided additional information to support the CEOG position stated in CEN-315. The staff reviewed CEN-349 and, by letter dated January II,1988 (Ref. 4), again rejected the CEOG position that the existing diversity between the RTS and the AFAS reets the requirements of the ATWS Rule.
Louisiana Power and Light Company (LP&L), licensee for Waterford, by letter dated October 7, 1988 (Ref. 5), submitted a plant-specific request for an exemption from the portion of the ATWS Rule that requires equipment diverse f rom the RTS to initiate the AFAS under conditions indicative of an ATWS. The submittal also provided detailed design information on the DSS and DTT. The staff reviewed the submittal and, by letter dated December 20, 1988 (Ref. 6),
forwarded a Request for Additional Information (RAI) on the licensee's proposed DSS /DTT design.
In addition, the staff denied the licensee's request for an exemption by letter dated March 8, 1989 (Ref. 7).
A meeting was held with the CEOG on May 1 and July 12, 1989, to discuss the conceptual design of the Diverse Emergency feedwater Actuation System (DEFAS).
The staff cocunented during the July meeting that the proposed DEFAS design was in general compliance with the ATWS Rule (Ref. 8). The licensee responded to the RAI by letter dated July 17, 1989 (Ref. 9).
This Safety Evaluation (SE) addresses the licensee's conformance to the ATWS Rule at Waterford, as detailed in References 5, 8 and 9.
3.0 CRITERIA The purpose of the ATWS Rule, as documented in SECY-83-293, " Amendments to 10 CFR 50 Related to Anticipated Transients Without Scram (ATWS) Events," is to
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require equipment / systems that are diverse from the existing reactor trip l '
system (RTS) and capable of preventing or mitigating the consequences of an ATWS event. The failure mechanism of concern is a common mode failure (CMF) of l
identical' components within the RTS (e.g., logic circuits; actuation devices; and instrument channel components, excluding sensors).
l The hardware / component diversity required by the ATWS Rule is intended to ensure that CMFs that could disable the electrical portion of the existing reactor trip system will not affect the capability of ATWS prevention / mitigation system (s) equipment to perform its design functions. Therefore, the similarities and differences in the physical and operational characteristics of these components must be analyzed to determine the potential for CMF mechanisms that could disable both the RTS and ATWS prevention / mitigation functions.
The systems and equipment required by 10 CFR 50.62 do not have to meet all of the stringer.t requirements normally applied to safety-related equipment.
Fewever, this equipment if a part of the broader class of structures, systems and components important to safety defined in the introduction to 10 CFR Part 50, Appendix A (General Design Criteria [GDC]). GDC-1 requires that " structures, systems, and components important to safety shall be designed, fabricated,
-erected, and tested tc quality standards commensurate with the importance of the safety functions to be performed." The criteria used in evaluating the licensee's submittal included 10 CFR 50.62, " Rule Considerations Regarding Systems and Equipment Criteria," published in the Federal Register, Volume 49, No. 124, dated June 26, 1984. Generic Letter No. 85-06, dated April 16, 1985,
" Quality Assurance Guidance for ATWS Equipment that is not Safety Related,"
details the quality assurance requirements applicable to the equipment installed per ATWS Rule requirements.
To minimize the potential for common mode failures, diversity is required for diverse scram system (DSS) equipment from sensor output to, and including, the components used tc interrupt control rod power. The use of circuit breakers from different manufacturers is not, by itself, sufficient to provide the required diversity for interruption of control rod power.
For mitigating systems (i.e., diverse turbine trip and diverse emergency feedwater actuation system), oiversity is required from sensor output up to, but not including, the final actuation device.
Electrical independence between ATWS circuits (i.e.,
DSS, DTT, and DEFAS) and the existing RTS circuits is considered desirable to prevent interconnections between systems that could provide a means for CMFs tc potentially affect both systems. Where electrical independence is not provided between RTS circuits and circuits installed to prevent / mitigate ATWS events, it must be demonstrated that faults within the DSS, DTT, or DEFAS actuation circuits cannot degrade the reliability / integrity of the existing RTS below an acceptable level.
It must also be demonstrated that a CMF affecting the RTS p(ower distribution systems, including degraded voltage and frequency conditions the effects of degraded voltage conditions over time must be considered if such ccnditions can go undetected), cannot compromise both the RTS and ATWS prevention / mitigation functions.
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Electrical independence of non safety-related ATWS circuits from safety-related circuits is required in accordance with the guidance provided in IEEE Standard 384, "IEEE Standard Criteria for Independence of Class IE Equipment and Circuits," as supplemented by Regulatory Guide (RG) 1.75, Revision 2, " Physical Independence of Electric Systems."
The equipment requireo by 10 CFR 50.62 to reduce the risk associated with an ATWS event must be designed to perform its functions in a reliable manner. The DSS, DTT, and DEFAS circuits must be designed to allow periodic testing to verify operability while at power. Compliance with the reliability and testability requirements of the ATWS Rule must be ensured by technical specification operability and surveillance requirements or equivalent means that govern the availability and operation of ATWS equipment; thereby ensuring that the necessary reliability of the equipment is maintained.
The ATWS prevention and mitigation systems should be designed to provide the operator with accurate, complete, and timely information that is pertinent te system status. Displays and controls shculd be properly integrated into the main control room and should conform to good human-engineering practices in design and layout.
4.0 DISCUSSION AND EVALUATION The following is a discussion on the licensee's compliance to the guidance contained in the Federal Reaister, " Statement of Considerations" (Ref. 10) and to the requirements of the MW5 Rule as discussed in Section 3 of this SE.
4.1 Diverse,5c_ ram System (Diverse _ Reactor Trip System (DRTS))
A.
_ System Description The Waterford diverse scram system is e dual channel control-grade system.
The existing pressurizer pressure transmitters on existing taps are used to provide signals to the DSS in a two-out-of-two trip logic. High pressurizer pressure will be used as the parameter indicative of an ATWS. The DSS will have two measurement channels. Each measurement channel consists of a pressure transmitter sensor with signal conditioner, an isolator and an A/D converter.
The DSS trip setpoint will be set greater than the RTS high pressurizer pressure trip setpoint (2420 psig vs 2350 psig) and less than the primary safety valve relief pressure setpoint (2500 psig). The two-out-of-two trip logic output signal will open motor-generator (MG) sets A and B output load contactors.
The reactor trip switchgear is powered by two MG sets connected in parallel with two output contactors in the feeder lines to the reactor trip switchgear.
a The MG set output contactors are held in the closed position by the energized l
holding coils. De-energization of the holding coils will cause the opening of the MG set output contactors and consequently trip the reactor.
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Two pushbuttons are installed on the main control board to provide the means l
of manual initiation of the DSS. The contacts of these two pushbuttons are wired in the same configuration as the contacts of the A/D converter.
. Depressing both pushbuttons simultaneously will result in de-energization of the MG set output contactor coils and subsequently trip the reactor.
B.
Diversity Hardware / component diversity is required for all diverse scram system (DSS) equipment from sensor outputs to, and including, the components used to interrupt control rod power. The use of circuit breakers from different manufacturers is not, by itself, sufficient to provide the required diversity for interruption of control rod power. The DSS sensors are not required to be diverse from the RTS sensors. However, separate sensors are preferred to prevent interconnections between the DSS and the existing reactor protection system (RPS or RTS).
The Waterford DSS instrument channels consist of transmitters, channel test cards, loop power supply with isolated output cards, signal isolator cards, and A/D converters. The pressure transmitter is an ITT Barton Mode 763 and the circuit cards are manufactured by Westinghouse and are diverse from the existing RTS components. The analog to digital converter is a Rochester Instrument System Model ET-1215 which is diverse from the existing RTS converter.
The actuatior. relays are powered from a nor-Class IE instrument AC-power panel.
The parallel device in the RTS is a mechanical circuit breaker powered by a Class IE vital bus.
Based on the above, the staff concludes that the level of hardware / component diversity provided between the DSS circuits and the existing RTS circuits at Waterford is sufficient to comply with the requirements of ID CFR 50.62 (the ATWS Rule) and is, therefore, acceptable.
C.
DSS Electrical Indep_endence/ Power Suppl _ies c
The purpose of the electrical independence requirements of the ATWS Rule is to prevent interconnections between the DSS and RTS (thereby reducing the potential for CMFs that could affect both systems) and to ensure that faults within the DSS circuits cannot degrade the RTS. Electrical independence of the DSS circuits from the RTS circuits should be maintained from sensor outputs up to the final actuation devices. The use of a common power source for the DSS and RTS senscrs is acceptable because, in accordance with the ATWS Rule, the senscrs can be shared between these two systems.
The sensors used in the DSS are separate from the existing RTS pressure transmitters. They do, however, share existing pressure sensing lines through instrument valves. The DSS transmitter circuits are completely independent from the existing RTS instrument loops.
The DSS relay logic will receive power from a separate non-Class IE power distribution panel, PDP 396AB, which is independent of the RTS. The DSS power supply is capable of performing its required design function upon loss of offsite power.
The pressure sensor instrumentation loops of the DSS are safety-related instruments which are located in the safety related cabinets LCP-61 and LCP-62.
These two cabinets are powered by the safety related power distribution panels 3MA-S and 3MB-S. The licensee has performed a power source common mode failure analysis and concluded that no common mode failures exist that could affect the RTS and DSS simultaneously. The staff finds this acceptable.
D.
DSS Quality Assurance, To ensure thet the DSS circuits perform their safety functions wher callec on, the Commission issued Generic Letter (GL) 85-06, " Quality Assurance Guidance for ATWS Equipment that is net Safety Related," which details the quality assurance requires for equipment installen per ATWS Rule requirements.
In addition, the ATWS Rule guidance stated tnat the DSS should be testable at
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power.
The licensee stated that the DSS circuits will be treated under a quality assurance program that is consistent with the guidance contained in Generic Letter 85-06. Safety related interfaces and other safety related portions of the DSS will be contro11ec in accordance with the approved 10 CFR Part 50, Appendix B Quality Assurance Program.
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DSS Testability The DSS input instrument loops and relay logic can be tested during plant cperation. A bypass switch (Enable / Disable) will temporary oisable the DSS system during the test. When the system is in a bypass condition an indicating light will be lit on the main control board.
Procedures will be prepared to guide the operator when the DSS is bypassed. The operational test of the MG set output contactors will only be performed during plant shutdown. The safety related instrument channels will be tested at power. The frequency of the channel test will be consistent with the RTS/ESFAS testing. An End-to-End test will be performed at each refueling outage.
The DSS maintenance and test bypasses will be built-in and will be part of the DSS circuits. Temporary modifications of the circuits for testing and maintenance will not be required. When a protection action is activated, and if any part of the DSS is placed in a bypass condition, an alarm annunciator is actuated in the main control room.
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7 Based on the above, the staff concludes'that the DSS surveillance testing proposed by the licensee, the means used to bypass the DSS for test and maintenance purposes, and the indication of the bypass condition are in accordance with good design practices and the requirements of 10 CFR 50.62 (the ATWS Rule) and are, therefore, acceptable.
F.
Other DSS Considerations Other system design considerations that enhance the DSS include:
1.
The energize-to-trip circuits will be used to exclude the activation of a trip by component failure.
2.
The DSS equipment will be qualified for the environment in which it will be installed.
3.
Once initiated, the DSS will seal-in and require deliberate r,anual operator action'to reset the system.
4 The DSS alarms will be consistent with the plant's control room design review process and good human-engineering practices. As a minimum the following will be annunciated:
- Diverse Reactor Trip
" Diverse Reactor Trip Active / Trouble G.
Conclusion Based on the above evaluation, the staff concludes that the prepcsed design cf the Diverse Scram System for Waterford, Unit 3 conforms to the requirements of 10 CFR 50.62 (the ATWS Rule) and is, therefore, acceptable.
4.2 Diverse Turbine Trip.
A.
General When the DSS causes a reactor scram, power is interrupted to the control element drive mechanism (CEDM) coils upstream of e rod power bus undervoltE5e relays. The de-energizing of these undervoltage relays actuates the turbine trip solenoid that results in a turbine trip.
The DTT design shares all circuit components with the DSS up to, but not including, the final turbine trip device. Those components are unique to the DTT (i.e., undervoltage relays, and the turbine trip solenoid) and do not appear in any of the' RTS trip paths. All of the information that is applicable to the DSS components and system, as discussed in Section 4.1 of this SE,
-is also applicable to the DTT components.
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Conclusion Based on the DSS evaluation and the above discussion, the staff concludes that the proposed design for the Diverse Turbine Trip for Waterford conforms to the requirements of 10 CFR 50.62 (the ATWS Rule) and is, therefore acceptable.
l 4.3 Diverse _ Emergency Feedwater Actuation System (DEFAS)
A.
DEFAS System Description The DEFAS will be a non-safety related system.
It will utilize the existing safety related (for Post-accident monitoring) steam generator (SG) level sensors and the existin equipment (pumps and valves) g safety related emergency feedwater systems to provide energency feedwater to the steam generator to mitigate the consequence of an ATWS event. Qualified isolators will be provided for the interface between the safety and the non-safety related components. The initiation logic will be a two-out-of-two on low SG 1evel. The trip setpoint will be at ES% of the SG wide range level which is lower than the existing emergency feedwater initiation setpoint.
The DEFAS initiation ~ signals cause actuation of the emergency feedwater pumps and valves only if there is a demand for emergency feedwater and if the actuation signal has not been generated by the existing emergency feedwater actuation system (EFAS).
The functional requirements for the DEFAS incluce:
- DEFAS must initiate emergency feedwater flow for ccnditions indicative of an ATWS where the EFAS has failed.
- The DEFAS will not be required to provide mitigation of an accident such as isolating feedwater flow to a ruptured steam generator.
- DEFAS will stop feedwater to the affected steam generator after reaching a pre-determined level setpoint at about 30 minutes after actuation; thereafter, manual operator intervention will control the system.
- DEFAS will utilize logic and redundancy to achieve a two-out-of-two initiation;
- DEFAS will utilize steam generator water level as the parameter indicative of the need for EFW actuation.
- DEFAS will interface with the actuated components via the existing safety related circuitries.
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- DEFAS will be blocked by the EFAS to prevent control / safety interactions when the EFAS actuates.
- DEFAS will be blocked by the steam generator pressure signal which indicates conditions of a Main Steam line break.
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- DEFAS will be enabled by a signal from the DSS indicating DSS actuation.
- DEFAS will include capabilities to allow testing at power.
- DEFAS will include features that provide alarms, plant computer data and other operator interfaces to indicate system status and reet 4
operability requirements.
- DEFAS setpoints will be coordinated with the existing FPS setpoints so that a competing condition between the PPS and DEFAS will be prevented.
- DEFAS equiptr.ent will te qualified for anticipated operational occurrences.
- DiFAS will be cesigned under the Quality Assurance procedures censistert with the requirements and clarifications contained in j
- DEFAS logic power will be separate ar.d iridependent frora the existirig RTS pcwer.
Each DEFAS logic power supply is capable of providirg 120 VAC uninterruptable pcwer for up to one hcur following the loss of its power bus.
Two pushbuttons, will be provided en the main control board for mar.ual initiation of the DEFAS.
A two-position " Disable - Enable", selector switch will be provided
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on the main control board to allow the operator to disable the operation of the DEFAS for maintenance or testing. Automatic indication will be lit when the DEFAS is disabled.
DEFAS testing can be performed during plant operation. To test the relay logic, the " Enable / Disable" selector switch should be placed at the
" Disable" position. Test procedures will guide the verification of the contacts being energized for each associated equipment (pump)or valve but notboth). A four-position switch (off, pump, normal, valve will be provided in each safety related emergency feedwater actuation channel to accomplish the test.
B.
Djversity The Waterford DEFAS design will utilize the existing safety related wide range steam generator level instruments as the sensor. The input signal l
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will be isolateo by a Westinghouse isolator card NLP3 before the signal actuates the non-safety related DEFAS bistable. The DEFAS bistable is a Rochester Model ET 1215 which is diverse from the RTS/EFWS bistables.
The DEFAS initiation relay is an Allen Bradley relay model 700 which is diverse from the RTS/EFWS initiation relay. The DEFAS actuation relay is a Potter Brumfield Model MDR-163-1 which is diverse from the existing RTS/EFWS actuation relays.
Based en our review of the DEFAS design, the staff concludes that the level of hardware / component diversity between the DEFAS design and the existing RTS/EFWS design is acceptable.
C.
DEFAS Electrical Independence / Power Supplies The safety related input signal will be isolated by a qualified isolator before it feeds into the non-safety related DEFAS. The output actuation signal from the DEFAS will be isolated by a cualified isolator before it actuates the emergency feedwater system equipment. The physical separation between the non-safety related DEFAS and the safety related systems will be properly maintained.
The DEFAS relay logic will receive power from a separate non-Class IE power distribution panel, PDP396 AB, which is independent of the RTS. The battery supply for panel PDP 396 AB is not physically connected to any of the reactor protection system (RPS) batteries. The DEFA5' power supply is capable of performing its safety function upon loss of offsite power.
The wide range steam generator level sensors are safety related instruments which are located in the safety related cabinets LCP-61 and LCP-62. These twc cabinets are powered by the safety related power
. distribution panels 3MA-S and 3MB-S. The licensee has performed a power source corinon mode failure analysis, and concluded that no common mode failure exist that could affect the RTS and the DSS /DEFAS simultaneously.
The staff fincs this acceptable.
D.
DEFAS Testability DEFAS testing can be performed during at-power operation. A bypass switch (Enable / Disable) will disable the DEFAS system during the test. Automatic indication will be continuously indicated on the Main Control board when the r,ystem is in a bypass position. The safety related instrument channels wili be tested at power and the (requency of the channel test will be consistent with the RTS/ESFAS tosting. The End-to-End test will be performed at each refueling outage.
The DEFAS maintenance and test bypasses will be built-in and will be part of the circuits. Temporary modifications of the circuits for testing and I
maintenance will not be required. When a protection action is activated, or when any part of the DEFAS is placeo in a bypass condition, an alarm l
annuciator is actuated in the main control room.
Based on the above, the staff concludes that the DEFAS surveillance testing proposed by the licensee, the means used to bypass the DEFAS for test and l
maintenance purposes, and the indication of the bypass condition are in accordance with good design practices and the requirements of 10 CFR 50.62-(the ATWS Rule) and are, therefore, acceptable.
E.
OtherlEFASConsiderations The interlock from the DSS allows the DEFAS to initiate feedwater flow only if a DSS actuation has occurred. The CEOG provided an analysis demonstrating that the DSS interlock would not affect the peak reactor vessel pressure for the limiting ATWS event. The DSS interlock would minimize inadvertent actuation of the DEFAS under non-ATWS accident ccnditions. The staff conclude that this is acceptable.
F.
Conclusion Based on the above evaluation, the staff concludes that the proposed design of the diverse emergency feedwater actuation System for Waterford Unit 3 conforms to the requirements of 10 CFR 50.62 (ATWS Rule) and is, therefore, acceptable.
5.0 Technical Specification Requirements The staff is presently evaluating the need for technical specification operability ano surveillance requirements, including actions considered appropriate when operability requirements cannot be met (i.e., limiting conditions for operation) te ensure that equiptrent installed per the ATWS Rule will be maintained in a operable condition.
In its Interim Commission Policy Statement on Technical Specification Improvements for i
Nuclear Power Plants [52 FR 3778. February 6,1987), the Commission established a specific set of objective criteria for determining which regulatory requirements and operating restrictions should be included in Technical Specifications.
This aspect of the staff's review of the Waterford ATWS design compliance I
with the ATWS Rule remains open pending completion of the staff's review to determine whether and to what extent Technical Specifications are appropriate. The staff will provide guidance regarding the Technical Specification requirements for DSS, DTT, and DEFAS at a later date.
Installation of ATWS prevention / mitigation system equipment should not be delayed pending the development or staff approval of operability and sune111ance requirements for ATWS equipment.
Principle Contributor:
H. Li Dated:
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REFERENCE 1
1.
Letter, R. G. Wells (CEOG) to F. Rosa (NRC), "CEN-315 Summary of the Diversity Between the Emergency Feedwater Actuation System for C-E Plants," September 18, 1985.
-2.
. Letter, D. M. Crutchfield (NRC) to R. W. Wells (CEOG), " Staff Evaluation I
of CEN-315," August 4, 1986.
3.
Letter, M. O. Medford (SCE) to G. W. Knighton (NRC), " San Onofre Nuclear Generating Station, Units 2 and 3 (Submittal of CEN-349)," December 30, 1986.
4 Letter, G. W. Knighton (NRC) to K. P. Baskin (SCE) and J. C. Holcombe (SDG&E), "NRC Evaluation of CEN-315 and CEN-349," January 11, 1988.
5.
Letter, R. F. Burski (L'P&L) to NRC Document Control Desk " Compliance with 10 CFR to 50.62: Reduction of Risk from Anticipated Transients Without Scram Events," October 7, 1985.
6.
Letter,D.Wigginton(NRC)toJ.G.Dewease(LP&L),"Requestfor Additional Information," December 20, 1988.
7.
Letter, G. M. Holahan (NRC) to J. G. Dewease (LP&L), " Nuclear Reactor Regulation Response to the Louisiana Power and Light Request for Partial Exemption from the Requirements of 10 CFR 50.62 for Waterford, Unit 3 "
March 8, 1989.
8.
Sumary of meeting with CEOG regarding the DEFAS Features to be installed per 10 CFR 50.62, dated August 15, 1989.
9.
Letter, R. F. Burski (LP&L) to NRC Document Control Desk, " Compliance with 10 CFR 50.62 Recuction of Risk From Anticipated Trentient Without Scram Event," July 17, 1989
- 10. Statement of Considerations, Federal Register, Vol 49, No. 124, June 26, 1984