ML20196L628
| ML20196L628 | |
| Person / Time | |
|---|---|
| Site: | Mcguire, Catawba, McGuire, 05000000 |
| Issue date: | 05/20/1988 |
| From: | DUKE POWER CO. |
| To: | |
| Shared Package | |
| ML20196L539 | List: |
| References | |
| NUDOCS 8807080079 | |
| Download: ML20196L628 (63) | |
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ATTACHMENT I OUKE POWER COMPANY CATAWBA NUCLEAR STATION MCGUIRE NUCLEAR STATION 1
FINAL DESIGN DESCRIPTION ATWS MITIGATION SYSTEM ACTUATION CIRCUITRY "AMSAC" ORIGINAL ISSUE:
JANUARY 23, 1987 8807080079 880601
- PDR ADOCK 05000369 P
PDC i
l QQCUMENT REVISION TABLE Revision Number Date 0
January 23, 1987 1
April 24,1987 2
May 20, 1988 1
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l TABLE OF CONTENTS Section Title Page
- 1. 0 INTRODUCTION 1
1.1 BACKGROUND
INFORMATION 1
2.0 FINAL DESIGN DESCRIPTI0i1 2
2.1 DESCRIPTION
OF THE SYSTEM 2
2.1.1 DESIGN DESCRIPTION 2
2.1.2 LOSS OF MFWPT ACTUATION 3
2.1.3 LOSS OF FEEDWATER-VALVE CLOSURE 4
l 3.0 RESPONSES TO NRC APPROVAL SER OF TOPICAL REPORT 10 l
l (WCAP-108S8) - PLANT SPECIFIC INFQg &IlQN l
4.0 RESPONSES TO APPENDIX A. AMSAr, ISOLATION DEVICES 22 5.0 JMPLEMENTATIONSCHEDULES 23 l
Rev. 1
E5 6.0 ATTACHMENT 1. RESPONSES TO NRC RE0 VEST FOR ADDITIQNAL 31 INFORMATION DATED APRIL 9. 1987 7.0 ATTACHMENT 2, RESPQN11S_TO NRC RE0 VEST FOR ADDITl0HAL 38 INFORMATION DATED JUNE 18. 1987 8.0.
ATTACHMENT 3, RESPONSES TO NRC RE0 VEST FOR ADDITIONAL 43 LNEQRMATION DATED OCTOBER 20, 1987 9.0 ATTAChiMENT 4. LETTER TO NRC OUTLINING CHANGES 49 MADE TO AMSAC FOR MC301RE AND TRANSMITTING A REVISED FINAL DESIGN DESCRIPTION Rev. 1
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1.0 INIR0RUCTION e
'In response to the July 16, 1986 NRC Staff Safety Evaluation Report for WCAP-10858 "AMSAC Generic Design Package",-Duke Power Company rubmits'the following Final Design Description for the Catawba and McGuire Nuclear Stations.
Plant specific
-inform: tion is contained in the final design description for each of the plants.
I
1.1 BACKGROUND
INFORMATION
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On July 26, 1984 the Code of Federal Regulations (CFR) was amended to include Section 10 CFR 50.62, "Requirements for Reduction of Risk from Anticipated Transients Without Scram (ATWS) Events for Light-Water-Cooled Nuclear Power Plants" (known as the "AiWS Rule").
Ai, ATWS is an expected operational transient (such as loss of feedwater, loss of condenser vacuum, or loss of offsite power) which is accompanied by a failure of the reactor trip system (RTS) to shut down the reactor.
The ATWS rule requires specific improvements in the design and operation of commercial nuclear power facilities to reduce the likelihood of failure to shut down the reactor following anticipated transients, and to mitigate the consequences of an ATWS event.
ED40151W/1 Rev. 0
2.0 FINAL DESIGN DESCRIPTION The basic requirements for Westinghouse plants is specified in Paragraph (c)(1) of 10 CFR 50.62, "Each pressurized water reactor must have equipment from sensor output to final actuation device, that is diverse from the reactor trip system, to autcmatically initiate the auxiliary (or emergency) feedwater system and initiate a turbine trip under conditions indicativ of an ATWS.
This equipment must be designed to perform its function in a reliable manner and be independent (from sensor output to the final actuation device) from the existing reactor trip system."
2,1 DESCRIPTION OF THE SYSTEM The AMSAC system that will be installed at the Catawba and McGuire Nuclear Stations is based upon the Westinghouse Owners Group (WOG) WCAP-10858 "AMSAC Generic Design Package" generie, design 3.
The following sections describe the plant specific design for the Catawba and McGuire stations.
Design differences for Catawba Unit 2, which has model D5 steam generators, are specifically identified.
2.1.1 DESIGN DESCRIPTION i
The AMSAC design for the Catawba and McGuire stations is based on conditions that are indicative of an ATWS event.
The system will monitor the main feed-water control valves, main feedwater control bypass valves, and the main feed-water isolation valves for position and will monitor both main feedwa~
oumps for operating status.
ED40151W/J Rev. 1
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Actuation of AMSAC will occur whenever:
0 Both main feedwater pumps are tripped
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0 When main feedwater flow to tne steam generaters is blocked due to inadvortent valve closure Failures of other Condensate - Feedwater system components or equipmert upstream of the main feedwater pumps which could result in the loss of feedwater to the steam generators will not be monitored.
This is because those events will result in loss of the main feedwater pumps due to low suction flow or pressure.
2.1.2 LOSS OF MAIN FEEDWATER PUMP TURBINES ACTUATI0h The actuation of the ANSAC system on loss of both main feedwater punips will be by three pressure switches monitoring the hydraulic control oil pressure to the stop valves for ea:h turbine.
Each of the feedwater pump turbine stop valves will close whenever there is s trip of the turbine.
These pressure switches will monitor the hydraulic oil prea.;ure holding the stop valves open.
Whenever the pressure switches sense a loss of pressure, indicative of a turbine trip, a two-out-of-three logic circuit will actuate.
If both pumps are tripped then the AMSAC circuitry will ;.orform the following:
E040151W/3 Rev. 0
i 1)
Trip the main turbine 2)
Start both motor driven auxiliary feedwater pumps 3)
Close the steam generator blowdown and sampling valves
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The actuation of the main turbine trip is performed.via the Turbine Supervisory Instrumentation system which is independent and separate from the Reactor Protection system.
Actuation of the motor driven auxiliary feedwater pumps and closure of the olowdown and sampling valves occurs as part of the normal control features for each system.
See Functional Logic / Block Diagram 1 for the above circuits.
2.1.3 LOSS OF FEEDWATER - VALVE CLOSURE 2.1. 3.1 Loss of Feedwater for 01, 02, 03 Steam Generators The following describes conditions and equipment to be monitored for the McGuire 1 and 2 and Catawba 1 units.
The differances between the D1, 02, D3 steam generators and the 04 and D5 models deals mainly with feedwater flow conditions during normal power operations.
01, D2 and D3 models do not use the j
split flow to the upper nozzle at power as ao the 04 and D5 models.
For plants with D1, D2 and 03 models it will be necessary to monitor the feedwater control EL40151W/4 Rev. 1 i
I valves and feedwater isolation valves for closure conditions which could lead to a loss of feedwater condition.
Additionally for Westinghouse plants which operate with the feedwater control bypass valve in the open position, it will also be necessary to monitor these valves for full open position.
McGuire 1 and 2 will monitor these valves for position status.
The actuation of the AMSAC system on closure of the feedwater flowpath valves position will be by limit switches monitoring the valves position.
Closure of 3 out of the 4 flow paths will cause the following to occur:
1)
Trip the main turbine 2)
Start both motor driven auxiliary feedwater pumps 3)
Close the steam generator blowdown and sampling valves McGuire 1 and 2 The following description applies to the McGuire Nuclear Station feedwater system for feedwater flowpath monitoring.
McGuire operates above the 50% power level with the feedwater control bypass valves fully open in order to make the feedwater system more perturbation tolerant.
Perturbations in the feedwater control system can be tolerated and responded to in a less operator burdensome manner with the feedwater control bypass valves open.
EL40151W/5 Rev. 1
Limit switches on the feedwater control bypass valves will be set to actuate thenever the valve is not fully open.
This typically corresponds to a 90%
open setting.
Feedwater control valve limit switches used for AMSAC applications will be set at a 25% open setting which relates to a 15% main feedwater flow setting to account for valve regulating movement.
Feedwater Isolation valve limit switches will be set at the full closed valve position since these valves are operated as open-closed valves only.
The feedwater ccntrol and control bypass valves position signal will be delayed 30 seconds prior to actuating the valve closure portion of the AMSAC logic to prevent potential trip conditions when steam generator level perturbations cause rapid feedwater conurol valve movements.
These valve fluctuations could cause unwanted unit trips when the valve status circuits have been reset by the operator or the unit is above the 40% power level and the circuits have been reset either manually or automatically.
The actuation of the turbine trip on feodwater flow path valve ci sure is performed via the Turbine Supervisory Instrumentation system which is indepen-dent and separate from the Reactor Protection system.
A "First-out" annunci-ator will be installed in the control room to alert the operator that the turbine has tripped due to the feedwater valves AMSAC logic.
An annunciator alarm already exists for the loss of both main feedwater pumps condition.
Status lights are provided to indicate when a feedwater control valve is less than 25% open.
A status light is also provided to indicate when a feedwater f.ow path has been closed for 30 seconds.
EL40151W/6 Rev. 1
Computer points will be provided for the feedwater control valves to indicate when these valves are in the 25% or less open condition.
Computer points are also provided for the feedwater control bypass valves to indicate when they are not fully open.
Actuation of the motor driven auxiliary feedwater pumps and closure of the steam generator blowdown and sampling valves occurs as part of the normal control features for each system.
The actuation of AMSAC due to feedwater flow path valve closure can be bypassed whenever the unit is below a 40% power level.
This is necessary to allow proper start up of the unit due to steam generator pre-heating requirements and feedwater control valve operational characteristics.
Normal startup calls for the feedwater control valves to remain closed unti! approximately 15% power and the feedwater isolation valves to remain closed until steam generator preheat-ing requirements are met, The use of a 40% load value is based upon the WCAP setpoint submitted by the j
l WOG in the February 27, 1987 Addendum.
See Section 3.6 for a detailed discussion of the operating bypass and reset l
mechanism.
See Functional Logic / Block Diagram 1 for McGuire 1 and 2.
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EL40151W/7 Rev. 1 1
Catawba Unit 1 The following description applies to the Catawba Nuclear Station Unit 1 feedwater system for feedwater flow path monitoring, Catawba Unit 1 presently operates with the feedwater control bypass valves closed above very low power levels (typically 0-15%).
Feedwater control valve limit switches used for AMSAC applications will be set at a 25% open setting to account for valve regulating movement.
Feedwater Isolation valve limit switches will be set at the full closed valve position since these valves are operated as open-closed valves only.
The feedwater control vahes position signal will be delayed 30 seconds prior to actuating the valve closure portion of the AMSAC logic to prevant potential trip conditions when steam generator level perturbations cause rapid feedwater control valve movements. These valve fluctuations enuld cause unwanted unit trips when the valve status circuits have been reset by the operator before reaching 40% power or the unit is above the 40% power level and the circuits have been reset automatically.
The actuation of the turbine trip on feedwater flow p th valve closure is performed via the Turbine Supervisory Instrumentation system which is independent and separate from the Reactor Protection system.
A "First-out" annunciator will be installed in the control room to alert the operator that EL40151W/8 Rev. 1
?q the turbine has tripped due to the feedwater valves AMSAC logic.
An annunciator alarm already exists for the loss of both main feedwater pumps condition.
Status lights are provided to indicate when a feeswater flow path is closed for 30 seconds.
Computer points will be provided for'the feedwater control valves to indicate when the:e valves are in the 25% or less open condition.
Computer points are also provided for the feedwater isolation valves to indicate full open or full closed.
Actuation of the motor driven auxiliary feedwater pumps and closure of the steam generator blowdown and sampling valves occurs as part of the normal control features for each system.
The actuation of AMSAC due to fecdwater flow path valve closure can be bypassed whenever the unit is below a 40% power level.
This is necessary ta allow proper start up of the unit due to steam generator pre-heating requirements and feedwater control valve operational characteristics.
Normal startup calls for the feedwater control valves to remain closed until approximately 15% power and the feedwater isolation valves to remain closed until steam generator preheat-ing requirements are met.
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The use of a 40% load value is based upon the WCAP setpoint submitted by the WOG in the February 27, 1987 Addendum.
l See Section 3.6 for a detailed discussion of the operating bypass and reset mechanism.
EL40151W/9 Rev. 1
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See Functional Logic / Block Diagram 2 for the Catawba 1 Logic Arrangement.
2.1.3.2 Loss of Feedwater for D4 and DS Steam Generators Catawba Unit 2 s
The following describes conditions and equipment to be monitored for the Catawba Unit 2 which uses model 05 steam generators.
D5 steam generators utilize a split flow feedwater arrangement where approximately 10% of the normal full power feedwater flow is diverted through the feedwater pre-heater bypass valve (Westinghouse notation FPBV).
The remainder of the flow follows the normal pathway through the feedwater isolation valve.
Closure of the isolation valves would not result in a loss of all feedwater flow to the steam generator.
Feedwater flow to the upper nozzle would provide protection and mitigation against reactor coolant system overpressurization.
The actuation of the AMSAC system on closure of the main feedwater control valves will be by limit switches monitoring the valves position.
Closure of 3 out of the 4 valves will cause the following to occur:
1)
Trip the main turbine 2)
Start both motor driven auxiliary feedwater pumps EL40151W/10 Rev. 1
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3)
Close the steam generator blowdown and sampling valves Feedwater control valve limit switches will be set at a 25% open setting to ac-s count for valve regulating movement.
The feedwater control valves logic will be delayed 30 seconds prior to actuating the valve closure portion of the AMSAC logic to prevent potential tr p conditions when steam generator level perturbations cause rapid feedwater control valve movements.
These valve fluctuations could cause unwanted unit trips when trying to bring the unit up in power and the valve status circuits have been reset by the operator.
The actuation of the turbine trip on feedwater valve closure is performed via the Turbine Supervisory Instrumentation system which is independent and separate from the Reactor Protection system.
A "First-out" annunciator will be installed in the control room to alert the operator that the turbine has tripped due to the feedwater valves AMSAC logic.
An annunciator alarm already exists for the loss of both main feedwater pumps condition.
A status light is provided to indicate when a feedwater flow path is closed for 30 saconds.
Computer points will be provided for the feedwater control vsives to indicate when these valves are in the 25% or less open condition.
Actuation of the motor driven auxiliary feedwater pumps and closure of the steam generator blowdown and sampling valves occurs as part of the normal control features for each system.
EL40151W/11 Rev. 1
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The actuation of AMSAC due to'feedwater control valve closure can be bypassed whenever the unit is below a 40% power level.
This-is necessary to allow proper start up of the unit due to steam generator pre-heating requirements as described in Section 2.1.3.1.
4 See Section 3.6 for a detailed discussion of the operating bypass and reset mechanism.
See Functional Logic / Block Ciagram 3 for the Catawba 2 logic arrangement.
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EL40151W/12 Rev. 1
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- 3. 0 RESPONSES TO NRC SER:
TOPICAL REPORT (WCAP-10858)
PLANT SPECIFIC INFORMATION 3.1 DIVERSITY The AMSAC design for the Catawba and McGuire Nuclear Stations is designed to maximiza the diversity between equipment used for AMSAC and the Reactor Protec-tion System (RPS).
The equipment used to detect conditions indic".ive of an ATWS as described in Sections 2.1.2 and 2.1.3 is independent of all equipment used in the Reactor Protection System (RPS).
The pressure switches which detect trip conditions of the main feedwater pumps will provide only AMSAC related functions.
These sensors will not have any interface with the RPS.
The limit switches used to detect valve position for the main feedwater isola-tion valves, the main feedwater control valves, and the feedwater control bypass valves.can provide AMSAC signals or other non-safety related applica-tions.
Ths valve limit switches used for AMSAC signals will not provide any signals to the RPS.
- 3. /.
LOGIC POWER SUPPLIES L
Power supplies for AMSAC logic power will be selected from existing plant sources,which will provide the maximum available independence from power supplie[usedbytheReactorProtectionSystem(RPS).
EL40151W/13 Rev. 1 A
m
9 Highly reliable non-interruptible non-safety power sources will be utilized for the \\MSAC design since the parameters being monitored are also non-safety.
The AMSAC design will utilize the 125 VDC station auxiliary batteries for the power. supply.
Power will be distributed to the circuits through existing distribution centers.
f Catawba will'use Distribution Centers CDA and CDB which are shown in Catawba Nuclear Station FSAR Figure 8.3.2-1.
McGuire will use Distribution Centers DCA and DCB or DCA-1 and OCB-1 depending upon present loading.
These distribution centers are shown in McGuire Nuclear Station FSAR Figure 8.3.2-2, 3.3 SAFETY RELATED INTERFACES The proposed AMSAC design does not have any interfaces with the existing RPS.
Therefore the RPS v;.11 continue to meet the existing safety criteria.
Auxiliary Feedwater, Steam Generator Blowdown and Steam Generator Sampling are systems which are safety related or.3v.' safety related components which will receive AMSAC inputs as described generally in the '.iCAP.
Interfaces with safety related systems will be designed such that the safety related system bwillperformitsfunctioncoincidentwithapostulatedfailureofthe non-safety AMSAC input.
EL40161W/14 Rev. O
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3.4 QUALITY ASSURANCE In response to Generic Letter 85-06, "Quality Assurance Guidance for ATWS Equipment that is not Safety Related", the existing Duke quality programs were reviewed to determine the need for any necessary changes or additions.
This s
review indicated that no new or separate quality program was needed to adequ-ately cover non-safety related ATVS equipment.
However, based on the eighteen criteria of the NRC quality assurance guidance, some adjustments were required
.to the implementing practices and procedures in order to clearly apply these to ATWS items.
The results of the review are described below on a criterion-by-criterion basis.
I.
Organization - the existing Duke organization meets the guidance of Generic Letter 85-36.
II.
Program - a new and separate quality program for non-safety related ATWS equipment was not developed.
The existing Duke practices and procedures were determined to be adequate in overall content to cover ATWS items.
However, minor changes to the existing Duke practices and procedures nave been made as described for each appropriate criterion.
III.
Design Control - the existing Duke Design Enginearing Department prore-dures and the Nuclear Station Modification Program were determined to meet J
i IV.
Procurement Document Control - the existing Duke Design Engineering Department orocedures and the Nuclear Production Department Administrative 1
s EL40151W/15 Rev. O tt
' ' l Policy Manual for Nuclear Stations were determined to meet Generic Letter 85-06.
V.
. Instructions, Procedures, and Drawings - a requirement for the development and use of plant procedures on ATWS items was added to the Nuclear Produc-tion Department Administrative Policy Manual for Nuclear Stations.
This was the only change to Duke practices and procedures determined to be Fj necessary to meet Generic Letter 85-06.
VI.
Document Control - the existing Duke practices and procedures were, deter-mined to meet Generic Letter 85-06.
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VII.
Control of Purchased Items and Services - the requirement to control purchased items and services for ATWS equipment including receipt inspec-tions,:as added to the Nuclear Production Department Administrative Policy Hanual for Nuclear Stations.
o VIII.
Identification and Control of Purchased Items - station specific listings of ATWS related systems and components will be added to each station's Quality Standards Manual to facilitate identification.
Otherwise, the existing Duke practices and nrocedures were determined to meet Generic Letter 85-06.
IX.
Centrol of Special Processes - the requirement to control special pro-cessori for ATWS equipment was added to the Nuclear Production Department Administrative Policy Manual for Nuclear Stations.
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EL40151W/16 Rev. 0 3
X.
Inspection - the inspection of ATWS items was added to the Nuclear Production Department Administrative Policy Manual for Nuclear Stations.
XI.. Testing - the testing of ATWS items was added to the Nuclear Production DepartmentAdmi[istrative,PolicyManualforNuclearStations.
XII.
Control of Measuring and Testing Equipment - the control of measuring and test equipment for ATWS items was added to the Nuclear Production Department Adm_inistrative; Policy Manual for Nuclear Stations.
XIII.
Handling, Storage and Shipping - the existing Duke practices and proce-dures were determined to.eet Generic Letter 85-06.
XIV.
Inspection, Test, and Operating Status - the existing Duke practices and procedures were determined to meet Generic Letter 85-06.
XV.
Non Conformances - the existing Duke practices and procedures were deter-mined to meet Generic Letter 85-06.
l XVI.
Corrective Action System - the existing Duke practices and procedures were i
determined to meet Generic Letter 85-06.
XVII.
Records - the existing Duke practices and procedures were determined to meet Generic Letter 85-06.
l XVIll.
Audits - the existing Duke practices and procedures were determined to 1
b meet Generic Letter 85-06.
EL40151W/17 Rev. 0
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HAINTENANCE' BYPASSES' i
The omponents selected for the AMSAC design will be of reliable design and
.in~ stalled in such a manner to enhance preventive and scheduled maintenance.
Where reaintenance may be required at power, design features will be provided to assist the maintenance technicians in t iat performance.
d Features that ass.fst in maintenance at power can be described as adequate p[anningofsystemdesignconfigurationtoenable.thetechniciantoservice
' individual components without undue hazard and jeopardy to plant operation.
/
Examples of the nature of the features are as follows.
1)
Isolation valves in hydraulic lines such that a pressure switch that has failed may be isolated for removal and replacement.
2)
Test switches, fuses, and the locating of sliding link terminals such that electrical isolation of the component can also be accomplished.
3)
Proper notification through plant monitoring systems to the operators that a failure of a component has occurred such that prompt corrective maintenance can be taken.
4)
Location of equipment and components with maintenance requirements in mind.
A human factors review will be performed on all controls and indications to ensure that they can be utilized in an efficient and readily understood manner.
EL40151W/18 Rev. 1 1
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n Indication of any bypass will be provided in the main control room and will be part of the human factors review.
3.6 OPERATING BYPASSES The AMSAC design for both the Catawbh and McGuire plants does use an operating bypass.
The purpose of this bypass is to allow the operators to bring the plant up in power using alternate flow paths to the steam generators and to meet steam generator preheating requirements.
The normal flow paths through the feedwater control valves and the feedwater isolation valves cannot be used because of pre-heating requirements associated with the steam generators.
3.6.1 MCGUIRE UNITS 1 AND 2 For McGuire, the feedwater flow paths are aligned in different ways depending upon power level and steam generator pre-heating requirements.
Up to 15%
The feedwater control bypass valve is open to allow forward feedwater flow.
The feedwater control valve is closed and the feedwater isolation valve is closed.
Above 15%
The feedwater control bypass valve is closed while simultaneously opening the feedwater control valve. When the preheating requirements have been met for EL40151W/19 Rev. 1
s the steam generators, the feedwater isolation valve is opened.
This typically occurs around 20-30% load.
Above 50%
The feedwater control bypass valves are reopened at 50% power level to provide an additional flow path to the steam geners rs.
This parallel operation of the bypass valves provides increased flexibility in dealing with feedwater control valve upsets.
Therefore, the operator must bypass the valve status inputs into the AMSAC system below 40% unit load.
In addition, for low load values typically between 30-50%, the feedwater control valves can modulate into the 25% open range.
A status light is provided in the control room for the operator to indicate when any of tho four feedwater control valves is less the 25% open for 30 seconds.
Computer inputs are also provided to determine which valve is below 25% set-point. When the unit is in a stable operating condition, the operator will remove the bypass and the valve status actuation will be armed.
The bypass is automatically removed at 40% load by the turbine impulse chamber pressure switches.
3.6.2 CATAWBA UNITS 1 AND 2 X-For Catawba, the feedwater flow paths are aligned in two ways depending upon power levels and steam generator pre-heating requirements.
EL40151W/20 Rev. 1 a
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Up to 15%
Catawba is aligned similarly to McGuire for this oower range.
The feedwater control bypass valves are open and the feedwater control valves are closed.
The feedwater isolation valves are also closed and feedwater is aligned to the steam generators through the feedwater preheater bypass valves.
Above 15%
The feedwater control bypass valve is closed while simultaneously opening the feedwater control valves.
When the pre-heating requirements have been met for the steam generators, the feedwater isolation valves are opened.
This typically takes place around 20-30% unit load.
From 15% power on up to 100%, the feedwater control valves through the feed-water isolation valves is the primary flow path.
The 05 model steam generators used on Catawba Unit 2 do have a 10% bypass as described in Section 2.1.3.2.
Therefore, the operator must bypass the valve status inputs into the AMSAC system below the 40% unit load.
In addition, for low load values typically between 30-50%, the feedwater c: 1 trol valves can modulate into the 25% open range.
A status light is provided in the control room for the operator to indicate when any of the four feedwater control valves is less than 25% open for 30 seconds.
Computer inputs are also provided to determine I
which valve is below 25% setpoint. When the unit is in a stable operating i
condition, the operator will remove the bypass and the valve status actuation l
will be armed.
The bypass is automatically removed at 40% load by the turbine impulse chamber pressure switches.
EL40151W/21 Rev. 1
3.7 MEANS FOR BYPASSING 3.7.1 MCGUIRE UNITS 1 AND 2 The means for bypassing the valves' status input into the AMSAC system is by a control switch mounted in the main control room and by automatic circuit logic 120 seconds after the turbine impulse chamber pressure has dropped below the 40% unit luad as allowed in the WCAP.
Both manual and automatic bypass cap-ability is provided.
The operators have the capability to manually bypass the feedwater_ valves portion of the AMSAC circuitry before the 120 seconds have elapsed in order to gain unit control during load reductions and prevent spurious trips.
This control switch is provided with an indicating light to indicate whether the valves' status portion of the system is bypassed.
An illuminated indicating light will continuously indicate when the system is reset.
The reset feature of the AMSAC design for the operating bypass will also be actitated automatically by two pressure switches (in a two-out-of-two logic) monitoring first stage turbine impulse chamber pressure.
These pressure switches will be set to initiate the automatic resetting of the bypass when turbine loading reaches a point coincident with a plant loading of 40%.
This will provide automatic resetting of the bypass independent of the operator controlled bypass switch.
The pressure switches which monitor turbine load will not have any interfaces with the RTS.
A failure in the RTS cannot prohibit this reset.
t EL40151W/22 Rev. 1 1
The automatic bypass is also performed by the turbine impulse chamber pressure switches.
These switches activate a timer which will instate the bypass of the AMSAC feedwater valve circuitry 120 seconds after the pressure has dropped below the 40% setting.
A human factors review will be conducted on the bypass controls provided in the main control room.
The disallowed methods of bypassing will not be utilized in the Duke AMSAC design.
3.7.2 CATAWBA UNITS 1 AND 2 The means for bypassing the valves' status input into the AMSAC system is by a control switch mounted in the nain control room.
This control switch is provided with an indicating light to indicate whether the valves' status portion of the system is bypassed.
An illuminated indicat-ing light will continuously indicate the system is reset.
The reset feature of the AMSAC design for the operating bypass will also be activated automatically by two pressure switches (in a two-out-of-two logic) monitoring first stage turbine impulse chamber pressure.
These pressure switches will be set to initiate the automatic resetting of the bypass when l
turbine loading reaches a point coincident with a plant loading of 40%.
This i
l EL40151W/23 Rev. 1
will provide aucor.latic resetting of the bypass independent of the operator controlled bypass switch.
The pressure switches which monitor turbine load will not have any interfaces with the RTS.
A failure in the RTS cannot prohibit this reset.
Ahumanfactorsreviewwillbecondudtedonthebypasscontrolsprovidedin the main control room.
The disallowed methods of bypassing will not be utilized in the Duke AMSAC design.
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l EL40151W/24 Rev. 1 V..
- 3. 8 MANUAL INITIATION Manual controls are available in the control room to perform a turbine trip and start auxiliary feedwater flow.
These controls were reviewed as part of the control room review described in Duke Power Company's response to Supplement 1, NUREG-0737 for both Catawba and McGuire.
These controls are conveniently located near each other enabling the operator to quickly initiate manual actions if they are required.
The controls provided for the main turbine consist of pushbuttons, which energize the trip solenoids which in turn close the main turbine stop valves.
Indication is provided which accurately provides feedback of the turbine trip condition.
The operator can start the motor driven auxiliary feedwater pumps from the control room by pressing their control pushbutton which in turn will close the breaker in the switchgear and start the motor.
Indicating lights and flow indicators are available to indicate a successful start of the motor driven pump.
3.9 ELECTRICAL INDEPENDENCE FROM EXISTING REACTOR PROTECTION SYSTEM (RPS)
Electrical independence for the Catawba and McGuire AMSAC design with regards to the RPS is achieved by complete system separation.
The AMSAC design has no common control devices, power supplies or sensors with the RPS.
EL40151W/25 Rev. 0 1
The Catawba and McGuire AMSAC design and their respective RPS's are completely
(
separate systems and have no interfaces in common.
The AMSAC system will perform its designed function regardless of failures within the RPS.
3.10 PHYSICAL SEPARATION FROM EXISTING REACTOR PROTECTION SYSTEM The Catawba and McGuire AMSAC system designs will utilize standard Duke separa-tion philosophy and criteria.
Since no interface is planned between the AMSAC design and the existing RPS, complete separation is achieved.
Existing separation between the RPS and non-safety related circuits will not be violated by the AMSAC design.
3.11 ENVIRONMENTAL QUALIFICATION The AMSAC system equipment will be located in areas of the plant that are considered a mild environment.
3.12 TESTABILITY AT POWER MSAC equipment will be tested prior to installation to ensure functionality.
The proposed AMSAC design is similar to the NRC accepted design option 3 as listed in WCAP-10858.
The portion of the design which uses main feedwater pump status as an input is fully testable at power.
The portion of the design that utilizes main feedwater control and isolation valvt position as an input is not fully testable at power.
EL40151W/26 Rev. 0
The main feedwater control and isolation valve status is determined by valve stem limit switch inputs.
These valve stem limit switches cannot be tested for status change at power without adversely affecting the operating status of the unit.
It is proposed that the limit switches for these valves be tested at each refueling outage during the valve stroke tests.
These limit switches can be tested during the stroke test to verify functionability.
Stroke testing the limit switches at power would likely trip the unit due to system instability when one steam generator flow path is blocked.
The main feedwater pump turbine trip sensing devices can be tested at power.
These devices are pressure switches which sense oil pressure on the stop valve control oil line.
The proposed AMSAC design will utilize three pressure switches per turbine to sense oil pressure and will provida a trip signal upon two-out-of-three actuation.
A selector switch and indicating light will be used to test each switch individually.
It is proposed that each pressure switch will also be tested at each refueling cycle.
Controls and indications used for testing purposes will include a human factors examination to ensure proper operation.
3.13 COMPLETION OF MITIGATIVE ACTION The actuation of plant turbine trip, auxiliary feedwater start, closure of the blowdown valves and closure of the sampling valves, once actuated, will continue to completion as part of the AMSAC system design.
EL40151W/27 Rev. 0 1
This is commensurate with Duke's design philosophy for important plant functions.
Operators can reset at the system level the AMSAC initiated signal to the auxiliary feedwater system with the signal present in order to modulate auxiliary feedwater flow.
This reset allows the operator an additional method to control steam generator level with the blowdown valves to prevent any possible overfilling.
This reset is provided as part of the existing normal auxiliary feedwater system controls.
The AMSAC signal itself can only be reset by either 1) resetting the main feedwater pump turbines and 2) opening two feedwater control and isolation valves or placing the valves' status bypass switch in the bypass mode below 40%
load.
3.14 TECHNICAL SPECIFICATION By letter dated September 15, 1986, Duke Power proposed a means by which items that do not meet the proposed technical specification selection criteria may be maintained.
Duke considers that the AMSAC system does not meet any one of three criteria proposed by the AIF (and endorsed by the Staff in SECY-66-10) in that ATWS is an event that is beyond the Design Basis of the plants.
Due to the high degree of reliability of the components of the system, Duke considers that testing on an 18 month frequency is sufficient.
EL40151W/28 Rev. 1 I
4.0 RESPONSE TO APPENDIX A. AMSAC ISOLAIION DEVICES The present design concept for ATWS/AMSAC does not call for the use of isolators between it and'the existing RPS.
The use of an isolator to access available sensors also utilized by the RPS would require a detailed response to
=,
Appendix A.
Because our design concept does not use those sensors, no Appendix A response is provided.
EL40151W/29 Rev. 0
5.0 IMPLEMENTATION SCHEDULE Potential outages have already been identified which would allow installation prior to the July 1989 deadline on each unit as outlined in the SER.
The most current forecasted dates for these proposed outages is as follows:
McGuire Unit 2 E004 - June 1988
-McGuire Unit 1 E0C5 - November 1988 Catawba Unit 1 E0C3 - February 1989 Catawba Unit 2 E0C2 - March 1989 NRC approval is needed by May 1, 1987 to accommodate the proposed installation schedule.
EL40151W/30 Rev. 0 a
6.0 AIIACBMENT 1. RESPONSES TO NRC REQUEST FOR ADDITIONAL INFORMATION DATED APRIL 9. 1987 1.
Identify the types of isolation devices used to isolate the AMSAC from Class IE circuits.
Response
The isolation devices employed to isolate the AMSAC signal from the Class 1E circuits are existing in place devices.
For the Catawba Nuclear Station, the isolation device used is an Optical Isolator manufactured by E-MAX Incorporated.
These isolators are in wide spread use throughout the piant for applications where Class 1E to Non-1E isolation is required.
Information on the analog model optical isolator was provided in H.B. Tucker's Letter to E.G. Adensam dated November 1, 1985.
These devices were found acceptable for Class 1E to Non-1E isolation as described in Section 18.3.6 of the Catawba Safety Evaluation Report-NUREG-0954 Supplement No. 5 dated February 1986.
The digital optical isolator has generically the equivalent isolation cap-
~
abilities and has been tested for its plant application.
For the McGuire Nuclear Station, the isolation device used is the coil to contact separation provided oy the system relays.
Relay isolation is applied at the McGuire Nuclear Station for most all intersystem signal transfer.
Relays used in this application are Cutler-Hammer Type 026.
EL40151W/31 Rev. 1
The Duke design for the ATWS/AMSAC system does not utilize sensors common to the Reactor Trip System (RTS).
Therefore no isoistors are required for sensor isolation.
The control interfaces between the non-safety ATWS/AMSAC system and the already existing plant systems described in Section 3.3 are the only places where isolation occurs and this employs existing 'and accepted devices.
2.
Provide a block diagram of the ATWS/AMSAC design.
As a minimum, the block diagram should show the Class IE to non-Class 1E interfaces and the classification and point of application of the power source for the AMSAC equipment and isolation devices.
Response
A functional logic / block diagram of the ATWS/AMAC design is enclosed to replace the logic drawings submitted in the original Design Description of January 23, 1987.
3.
The WCAP-10858P-A report, as approved by the Commission, specifies a 30-second time delay upon the initiation of the AMSAC signal and a 120-second time delay upon the termination of the AMSAC signal.
The proposed system design does not provide for these two time delays.
Discuss why they are not included in the system design.
Response
The 30 second time delay specified in the WCAP report is not a required time delay for AMSAC signals in Design Option 3 except for the trip signal which goes to the turbine.
The turbine trip signal originated in AMSAC will be delayed by the WCAP specified amount when that value is determined by the Westinghouse Owner's Group.
The 120 second time delay upon termination of the AMSAC signal as described and shown in the WCAP is provided to delay automatic removal of the signal when the plant load decreases below 40% to allow for completion of the AMSAC actions following a turbine trip.
EL40151W/32 Rev. 1
.m The Duke design does not use automatic removal of the AMSAC load permissive signal.
The Duke design has a manual removal of the load bypass to the valve's status portion of the AMSAC logic.
This manual bypass does not become automatically installed at any time during a trip event.
It is provided to allow the operators the ability to bypass the valve's closed input signal during start-up conditions.
The feedwater pump's status signal is always present in the AMSAC logic and is not operator bypassed at any time.
4.
Expand upon the technical merits of inhibiting AMSAC below the 56% power level rather than the 70% power level specified by WCAP-10858-A.
Justify the 56% power level.
Response
Based upon the Westinghouse Owner's Group Letter "Addendum 1 to WCAP-10858-P-A and WCAP-11293-A:
"AMSAC Generic Design Package",
dated February 26, 1987 the AMSAC manual bypass permissive will be changed to 40% power level.
This change reflects the load changes identified in the WCAP addendum.
5.
Section 3.5 of your submittal states that FEATURES will be provided to assist in maintenance.
Discuss the nature of these features.
Response
Features that assist in maintenance at power can be described as adequate planning of system design configuration to enable the technician to service individual components without undue hazard and jeopardy to plant operation.
Examples of the nature of the features are as follows.
- 1) Isolation valves in hydraulic lines such that a pressure switch that has failed may be isolated for removal and replacement.
EL40151W/33 Rev. 1 i
- 2) Test switches, fuses, and the locating of sliding link terminals such that electrical isolation of the component can also be accomplished.
e
- 3) Proper notification through plant monitoring systems to the operators that a failure of a component has occurred such that prumpt corrective maintenance can be taken.
- 4) Location of equipment and components 'ith maintenance requirements in mind.
6.
Your submittal did not verify that the disallowed methods of bypassing are not utilized.
Under what conditions are the actions of lifting leads, pulling fuses, tripping breakers, or pnysically blocking relays employed?
Response
As described in Sections 3.6, 3.7 and 3.12 means for bypassing the feedwater valves input below 40% power and testing the Main Feedwater Pump Turbine trip oil pressure switches at any power level is provided.
These provisions do not employ any of the disallowed methods of bypassing.
Lead lifting, fuse pulling and breaker tripping are not allowed as means for providing any bypass.
7.
Section 3.13 implies that the operator can abort the mitigative action of AMSAC before it goes to completion.
This is contrary to the approved design described in WCAP-10858P-A.
Clarify this section of your submittal and justify the proposed design.
Also, can the operator abort the AMSAC initiated turbine trip signal before the turbine trips?
EL40151W/34 Rev. 1 1
Response
The design of the ATWS/AMSAC circuits and their interfaces with the mitigative systems already installed at the plant is in conformance with the approved design described in the WCAP.
As described in Section 2.0 of the WCAP under design criteria item
- 14) "AMSAC shall be designed so that, once actuated, the completion of mitigating action shall be consistent with the plant turbine trip and auxiliary feedwater circuitry."
\\
The Duke design for AMSAC is consistent with our plant turbine trip and auxiliary feedwater system circuitry.
These systems and their design have been approved through the licensing review process.
The last paragraph of Section 3.13 will be revised to change "or" to "and" as printed prior to 2).
The operator can only reset or bypass the AMSAC signal as described in Sections 3.6 and 3.13.
He cannot abort the AMSAC initiated signal for a turbine trip unless the unit is below 40% power levek The WCAP does not require automatic initiation of AMSAC below 40% power.
The operator cannot block the mitigative actions of the plant systems required by AMSAC.
AMSAC level bypass of the feedwater valve status signal can only be done manually below 40% unit load.
8.
The logic diagrams with your submittal do not show all of the operator I
initiated resets, inhibits, blocks, and bypasses described in the l
submittal.
Re-submit or supplement the logic diagrams showing the logic i
function of all operator initiated actions.
EL4G151V/35 Rev. 1
s Respon'se:
The Duke design submittal for ATWS/AMSAC did not intend to show all of the existing system level resets, inhibits, blocks and bypasses.
The ATVS/AMSAC logic diagrams were provided ennsistent with the level of detail provided in the WCAP to enable review of the specific AMSAC design logic developed by Duke.
The brief description of the existing system level operator initiated action was provided to assist in the review of plant shutdown conditions once the reactor had been manually tripped by the operator and the control of steam generator water level was required.
I Supplemental logic diagrams are enclosed for the mitigative actions required by the ATWS/AMSAC to show operator initiated actions which would impact the performance of the Auxiliary Feedwater Motor Driven Pumps, Steam Generator Blowdown Isolation Valves and the Steam Generator Sampling Valves.
The main terbine cannot be reset with any trip signal present.
The AMSAC signal will maintain the turbine tripped until the AMSAC initiating conditions have been cleared.
The operator cannot bypass the ATWS/AMSAC signal to trip the turbine as above 40% unit load as described in Section 3.6 and 3.13, and the response to question 7.
The operator also cannot abort the ATWS/AMSAC signal above 40% un.
load.
9.
The Westinghouse Owner's Group (WOG) plans to transmit to the NRC a supplement to WCAP-10858P-A containing a new permissive power level (40%)
1 EL40151W/36 Rev. 1 j
for AMSAC and new actuation setpoints for the optional designs previously discussed in the WCAP.
Discuss whether the new permissive and new setpoints have been taken into account for both the McGuire ard Catawba AMSAC designs.
If not, state whether you intend to provide ravised permissive level and actuation setpoints and when, i
Response
Yes, the Duke ATWS/AMSAC design will take into account, for both McGuire and Catawba, the n,ew information contained in the WCAP addendum.
The new power level permissive (40%) has been integrated into the design.
Also a review of the impact of the turbine trip time delays will be undertaken once that information is available from the WOG.
The original submittal of January 23, 1987 will be revised to reflect any WOG changes that impact the Duke design.
EL40151W/37 Rev.
7.0 ATTACHMENT 2. RESPONSES TO NRC REQUEST FOR ADDITIONAL INFORMATIONDAT[D_dVNE 17. 1987 1.
The response to Question 3 of the April 9, 1987 letter stated that the WCAP recommended a time delay of 30 seconds for the initiation of the AMSAC signal is not required.
The staff needs additional information/
justification regarding the omission of the 30 second time delay as recommended by the WCAP.
Response
The response to Questien 3 of the April 9, 1987 letter stated that the 30 second time delay utilized in the WOG Design Option 3, contained in WCAP-10258P-A Section 4.0, is not required for the McGuire and Catawba AMSAC designs.
The justification for the omission of the 30 second time delay as recommended in the WCAP is based upon the original design of the Auxiliary Feedwater system interfaces with the Main Feedwater system concerning loss of the main feedwater pumps.
The existing design of the McGuire and Catawba plants utilize starting the motor driven Auxiliary Feedwater (CA) upon loss of both Main Feedwater Pump Turbines (MFWPTs).
When both-MFWPTs trip, the CA motor driven pumps automatically start without any designed-in time delay.
This feature is described in both the Catawba and McGuire FSAR Sections 7.4.1 and 10.4.7.
The existing design starts CA anticipatory to the low-low steam generator level set point used in the protection system to start auxiliary feedwater.
Experience has shown that when forward feed-water flow is stopped, the low-low setpoint is reached very quickly.
EL40151W/38 Rev. 1
l The omission of the 30 second AMSAC signal time-delay allows the original design of the auxiliary feedwater motor driven pumps to be maintained with respect to starting directly upon loss of all (both) the main feedwater 4
pumps.
2.
The response to the RAI (April 9, 1987) is not clear as to why the C-20 T000 provision is not being implemented.
This TD0D ensures that the AMSAC mitigative action once started goes to completion.
Provide the details of how the AMSAC initiation signal is maintained upon AMSAC actuation.
Response: The C-20 TD00 provision is not being implemented in the same auto-matic fashion as contained in the WCAP.
The Duke design of the C-20 permissive is electrically latched in place thus allowing the AMSAC mitigative action once started to go to completion and requires manual action to place the valve position portion of AMSAC in bypass below 40%.
As described in the response to Question 3 of the April 9, 1987 letter, the 120 second time-delay to remove tne C-20 permissive signal (i.e., put in a bypass) is based upon an automatic system.
The WCAP requires the time delay to allow for completion of the AMSAC initiated actions followirig turbine trip.
The Duke design does not use an automatic feature for instating a total system bypass below 40% load.
The Duke system is entirely manual.
A control room located switch is used to instate the bypass EL40151W/39 Rev. 1
t below the 40% (C-20) load value.
The bypass is automatically removed when unit load increases above 40%.
The feedwater valve position AMSAC signal cannot be bypassed above 40% load and it can only be bypassed manually below 40% load.
The loss of both main feedwater pumps portion of the AMSAC logic in not interlocked with 4
the C-20 permissive.
It is functional at all unit lo;d conditions.
The AMSAC signal is maintained upon AMSAC actuation as described aoove and as contained-in the responses to questions 3 and 7 of the April 9, 1987 letter.
3.
The submittal of January 23, 1987 discusses system instabilities below the 56% power level.
Addendum No. 1 to the WCAP calls for arming AMSAC at the 40% level.
Discuss how the steam generator preheating requirements and the feedwater control valve operational characteristics are being handled at this lower AMSAC arming set point.
Response: The January 23, 1987 submittal briefly discussed feedwater system operating characteristics while at low power levels.
Low power (0-60%) has typically been more sensitive to feedwater system upsets.
Steam generator preheating requirements are associated with steam generator water hammer prevention during startup.
A Feedwater Bypass System (FBS) is provided to allow proper preheating of the feedwater piping downstream of the feedwater isolation valve.
In order to warm this line, the feedwater control valves (FCV) and feedwater isolation valves (FIV) are closed and the FBS used to warm the line.
EL40151W/40 Rev. 1 2
.]
The discussions in the January 23, 1987 submittal reflected a design which allows manual control of bypassing the FCV and FIV inputs to the AMSAC logic for less than 56% power level.
This value was chosen because it is where we typically have both Main Feedwater Pumps (MFWPS) operating.
The 56% value was also more conservative than the 70% power level previously required by the WCAP.
Since the addendum to the WCAP changed the power level for the C-20 permissive from 70% power to 40% power the Duke design was changed accordingly.
This power level change (70% to 40%) required additional system changes. Minor changes in the AMSAC system design developed by Duke are being made.
The preheating of the feedwater lines is typically completed between 25% and 30% power.
This condition should not be affected by the power level change.
The statements provided in Sections 2.1.2. page 6 and 2.1.3.2 page 8 reflected additiona' Duke and Westinghouse steam generator requirements that justified the design of a manual bypass switch below a power level which supported the Duke design of a manual bypass switch.
This switch would allow manually closing the FCVs and FIVs below the WCAP specified power level and not actuating the AMSAC signal.
With respect to feedwater control valve operational characteristics, 3
the power level change did necessitate an additional change in the Duke AMSAC logic.
In response to some plant transient conditions, 3 out of the 4 FCVs may close below the 25% open setpoint required in 3
the WCAP and cause generation of an AMSAC signal.
In order to reduce the possibility of this occurring, Duke installed a 10 second 1
i 3
- l EL40151W/41 Rev. 1 i
I time delay in the'FCV signal to the AMSAC logic which would activate when a FCV reached the 25% open position.
Duke now plans to in-crease the time delay to the WCAP allowed 30 seconds.
The 30 second time delay allowed for AMSAC signal delay will be utilized on the FCV position signal (25% open) only.
This time-delay should prevent FCV fluctuations fro'm causing unwarranted AMSAC initiations.
FCV fluctuations will still occur during startup of the standby main feedwater pump at around 50% - 60% power but the 30 second time delay will allow sufficient time for the operators to control the FCVs and adjust main feedwater pump turbine speed accordingly.
F t
EL40151W/42 Rev. 1
8.0 ATTACHMENT 3. RESPONSES TO NRC REQUEST FOR ADDITIONAL INFORMATION DATED OCTOBER 20_. 1987 1.
Discuss the changes made in Revision 1 to the WCAP-10858P-A with respect to their impact on the Duke Power AMSAC design.
Specifically, discuss (a) the turbine trip variable time-delay as specified in the revised WCAP and (b) the C-20 permissive mechanis,ns.
Response: Specific items which affect the Duke Power Company AMSAC design and are contained in the WCAP revision are:
(a)
The Turbine Trip variable time-delay based upon unit load and, (b) The C-20 permissive arming adjustment with respect to the turbine trip variabh time-delay.
Duke Power Company intends to delete the turbine trip variable time delay from its logic diagrams submitted to the NRC and to maintain the plant designs for Catawba and McGuire in their original design status.
Section 10.2.2 of each plant's FSAR describes the Turbine Trip inputs.
We feel that it is conservatively desirable to main-tain the plant design as close to that which was originally licensed and analyzed.
Tripping the turbine directly (without time delay) upon loss of both Main Feedwater Pump Turbines (MFWPT) and pumps is anticipatory to a low-low steam generator level initiated trip through the Reactor Protection System (RPS).
The Catawba and EL40151W/43 Rev. 1 2
i McGuire plant designs included the loss of both MFWPTs as part of the main turbine trips in their original designs.
A complete loss of feedwater event which would develop from loss of both pumps is not an event which allows continued plant operation even at extreme-ly low power levels.
The most prudent and conservative approach is that upon loss of both feedwater pumps, the main turbine should trip to avoid unnecessary depletion of steam generator inventories and reduce the amount of relatively cold auxiliary feedwater required to stabilize steam generator levels.
For loss of load events, of which the most severe is a loss of condenser vacuum event, the main turbine would trip and the main feedwater pump turbines would also trip.
Loss of both main feed-water pumps would not be required to trip the main turbine because it would already be tripped and a time-delay in the turbine trip circuit would not provide any input anyway.
Loss of both main feedwater pumps will initiate auxiliary feedwater as required.
Loss of load (i.e., turbine trips without loss of vacuum) does not cause a loss of feedwater and therefore, operator action would be relied upon to trip the reactor, trip both main feedwater pumps and verify initiation of auxiliary feedwater.
The Westinghouse prescribed turbine trip variable time delay pro-vides no beneficial actions because of the existing plant designs for Catawba and McGuire based on the above explanations.
EL40151W/44 Rev. 1
The C-20 permissive as described in the revision to the WCAP adds a variable timer whose time-delay is based upon the same turbine loads associated with the turbine trip time-dalay.
In fact, these two timers are varied according to turbine load jointly.
The same argument for the Duke AMSAC design exclusion of these time delays as described above applies to the C-20 time-delay.
The Duke AMSAC design for Catawba and McGuire will maintain the original turbine trip upon loss of both main feedwater pump turbines and therefore, no turbine load input is planned for the C-20 permissive circuitry to automatically instated a bypass of the AMSAC belcv 40% load as in the WCAP.
The Duke AMSAC design incorporates an administratively controlled C-20 bypass below 40% load.
The Duke bypass only affects the valve position portion of AMSAC.
Feedwater pump's status cannot be bypassed.
The Westinghouse prescribed C-20 time-delay variance upon unit load between 40% and 100% provides no beneficial actions because the Duke AMSAC design includes a manual bypass below 40% unit load.
2.
Discuss what aaministrative procedures will be used to direct operators regarding the use of the C-20 permissive (reset) below the 40% load value.
Response: The administrative procedures used to direct the operators regarding use of the C-20 permissive below the 40% load value are developed as part of the Duke Nuclear Station modification process.
The EL40151W/45 Rev. 1
Administrative Policy Manual for the Nuclear Production Department requires that as part of the Safety Evaluation process per 10 CFR 50.59, each modification is evaluated for any required changes or additions to the station procedures.
Informaticn pertaining to the precautions governing use of the C-20 permissive bypass below 40% unit load prior to the timing out of the 30 second time-delay in the Feedwater Control valve status logic will be included in the Final Scope Document for the modification.
The Final Scope Document is used as the collection document for all significant design inputs into the modification.
This document is transmitted to the station for review and used in developing the implementation package for the modification.
The Final Scope Document is also used in the development of the 10 CFR 50.59 safety evaluation performed by the Design Engineering Safety Evaluation group prior to development of the implementation package at the station.
3.
Discuss the turbine trip on (a) loss of both main Feedwater Pump Turbines and (b) closure of the MFW control / isolation valves with respect to the ATWS scenarios contained Section 2.0, Design Bases of the WCAP (!.oss of Main Feedwater and Loss of Load).
Response
(a) The Turbine Trip on loss of both main feedwater pump turbines is part of the originally licensed Catawba and McGuire plant designs.
l EL40151W/46 Rev. I
The response to Item 1 discusses much of the relationship regarding the above trip and the two ATWS scenarios as described in the WCAP Section 2.0.
The only additional discussions to add to those previously given are that the original plant designs for Catawba and McGuire previously contained much of the ATWS rule required.
The Duke design for AMSAC adds 'the feedwater valve status inputs, yet does not implement the various time delays also described in Item 1 because of the previously discussed reasons.
There is no value gained in trying to continue main turbine operation when the water source for the turbine steam requirements does not function.
Tripping the turbine during either of the two scenarios results in an anticipatory response to low-low steam generator water level.
Maintaining the steam generator inventory and not continuing turbine operation to the low-low level set point during loss of feedwater events is conservatively mere desirable.
(b) The main turbine trip on closure of the MFW control or isolation valves is initiated by the AMSAC circuits directly as contained in the Final Design Description.
With respect to the ATWS scenarios contained in Section 2.0, Design Bases of the WCAP the turbine trip occurs as a direct result of either scenarios.
Closure of the valves is a direct trip input.
Whenever three out of four feedwater control valves are closed for 30 seconds or three out of four feedwater isolation valves are closed, the main turbine will trip.
This response is the loss of feedwater scenario described in the WCAP.
EL40151W/47 Rev. 1
The loss of Load scenario results in no action by the feedwater valves.
The feedwater control valves may open up more in response to loss of feedwater due to a main feedwater pump trio on loss of vacuum assuming the Feedwater Control System is functional.
The turbing, trip will still occur due to loss of vacuum.
For loss of load events where the turbire trips but there was no loss of vacuum, there is no loss of feedwater, the feedwater control / isolation valves do not close, and operator actions is relied upon to trip the reactor, trip both main feedwater pumps and verify initiation of auxiliary feedwater.
4.
Provide updated Logic Diagrams showing:
(a) The time-delay increases to 30 seconds in the Feedwater Control Valve signal circuit.
(b) The deletion of the time-delay in the Turbine Trip circuit.
Response
The two updated logic diagrams are encloseo with this submittal.
EL40151W/48 Rev. 1
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-and 50-413, 50-414 Mr. H. B. Tucker, Vice President Nuclear Production Department Duke Power Company 422 South Church Street Charlotte, North Carolina 28242
Dear Mr. Tucker:
Subject:
ATWS Rule (10 CFR 50.62) for McGuire and Catawba Nuclear Stations, Units 1 and 2 (TACS 59081/59111/59112/64535)
The NRC staff, with the technical assistance of EG&G, has reviewed your proposed implementation of the ATWS Rule (10 CFR 50.62) for the McGuire and Catawba Nuclear Stations.
Enclosed is our safety evaluation report in which we conclude that, pending further staff review of the appropriateness of technical specifications for ATWS requirements, the ATWS design proposed by Duke Power Company for the McGuire and Catawba units is, otherwise, in comoliance with the ATWS Rule of 10 CFR 50.62, paragraph (c)(1). Our decision regardirig technical specifications will not affect design and we recountnd that you proceed with i
installation as scheduled. We understand that implementation is scheduled for completion during the 1988 refueling outages for the McGuire units and during the 1999 refueling outages for the Catawba units.
Should you have questions, contact me at (301) 492-8961 or K. Jabbour at (3011 492-7367.
Sincerely, M. Hood, P N
l Darl S Project Manager Project Directorate II-3 Division of Reactor Projects, I/II
Enclosure:
As stated cc: See next page D ev,
oatJJ V
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3 ENCLOSURE SAFETY EVALUATION REPORT REGARDING COMPLIANCE WITH ATWS RULE 10 CFR 50.62 MCGUIRE AND CATAWBA NUCLEAR STATIONS, UNITS 1 AND 2 00CKET NOS. 50-369/370 AND 50-413/414
1.0 INTRODUCTION
26, 1984, the Code of Federal Regulations (CFR) was amended to include On July Section 10 CFR 50.62, "Requirements for Reduction of Risk from Anticipated Transients Without Scram (ATWS) Events for Light-Water-Cooled Nuclear Power Plants" (known as the ATWS Rule). The requirements of Section 10 CFR 50.62 apply to all cormercial light-water-cooled nuclear power plants.
An ATWS is an anticipated operational occurrence (such as loss of feedwater, loss of condenser vacuum, or loss of offsite power) that is accompanied by a failure of the Reactor Trip System (RTS) to shut down the reactor. The ATWS Rule requires specific improvements in the design and operation of connercial nuclear power facilities to reduce the probability of failure to shut down the reactor following anticipated transients and to mitigate the consequences of an ATWS event.
Paragraph (c)(1) of 10 CFR 50.62 specifies the basic ATWS mitigation system requirements for Westinghouse plants. Equipment, diverse from the RTS, is required to initiate the auxiliary feedwater (AFW) system and a turbine trip for ATWS events.
In response to paragraph (c)(1), the Westinghouse Owners Group (WOG) developed a set of conceptual ATWS mitigating system actuation WOG utilizes circuitry (AMSAC) designs generic to Westinghouse plants.
Westinghouse Topical Report WCAP-10858, "AMSAC Generic Design Package,"
which provides information on the various designs.
The staff reviewed WCAP-10858 and issued a safety evaluation of the subject In this safety evaluation, the topical report on July 7,1966 (Ref.1).
staff concluded that the generic designs presented in WCAP-10858 adequately meet the requirements 10 CFR 50.62. The approved version of the WCAP is labeled WCAP-10858-P-A.
During the course of the staff's review of the proposed AMSAC design, the WOG26 issued Addendum 1 to VCAP-10858-P-A by letter dated February This Addendum changed the setpoint of the C-20 AMSAC permissive signal from 70%
In addition, for those plants selecting the feed-reactor power to 40% power.
water flow or the feedwater pump / valve status logic options, a variable delay timer is to be incorporated into the AMSAC actuation logics. On August 3, 1987, the WOG issued Revision 1 to WCAP-10858-P-A (Ref. 9) which incoroorated The staff Addendum 1 changes and orovided details on the variable timer.
considers the Revision 1 changes to be acceptable.
IQAREO O /4 / I J)g
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. Paragraph (c)(6) of the ANS Rule requires that detailed fr.fomation to demon-strate compliance with the requirements be submitted to the NRC.
In accordance with paragraph (c)(6) of the ATWS Rule. Duke Power Company)(the licensee) pro-vided information by letter dated January 23, 1987.(Ref. 2. The letter forwarded the detailed design description of the AMSAC proposed for installation at the McGuire and Catawba Nuclear Stations.
The staff held a conference call with the licensee on February 10, 1987, to discuss the AMSAC design. As a result of the conference call, a request for additional infomation was sent to the licensee on February 27,1987(Ref.4).
The licensee responded to the request on April 9,1987 (Ref. 5). The licensee's response raised additional questions. Therefore, the staff held a second con-ference call with the licensee on June 8, 1987. This conference call resulted in a second request being sent to the licensee on June 17,1987(Ref.6). The licensee responded to the second request by letter dated June 18,1987(Ref.7).
Following the release of Revision 1 to WCAP-10858-P-A, the staff discussed the impact it would have on the licensee's proposed AMSAC design. This discus-sion by conference call with the licensee on October 5, 1987, was followed b13, 1987 (R another NRC request for additional information on October The licensee responded by letter dated October 20, 1987 (Ref. 11).
P.0 REVIEW CRITERIA The systems and equipment required by 10 CFR 50.62 do not have to meet all of the requirements nomally applied to safety-related equipment. However, the equipment required by the.ATWS Rule should be of sufficient quality and reli-ability to perfom its intended function while minimizing the potential for transients that challenge other safety systems, e.g., inadvertent scrams.
Th' following review criteria were used to evaluate the licenset's submittals:
e 1.
The ATWS Rule, 10 CFR 50.62.
2.
"Considerations Regarding Systems and Equipment Criteria,"
published in the Federal Regiscer, Volume 49, No.124, dated June 26, 1984 3.
Generic letter 85-06, "Ouality Assurance Guidance for ATWS Equipment That Is Not Safety Related," April 16, 1985.
4 Safety Evaluation of WCAP-10858 (Ref. 1).
5.
WCAP-10858-P-A, Revision 1 (Ref. 9) 3.0 DISCUSSION AND EVAL.UATION To determine that conditions indicative of an ATWS event are present, the licensee has elected to implement the WCAP-10858-P-A AMSAC design associated
\\ with monitoring the main feedwater (MFW) control and isolation valves position and the main feedwater pug operating status. However, many details and inter-faces associated with the implementation of the final AMSAC design are of a plant-specific nature. The following paragraphs provide a discussion on com-pliance with respect to each of the fourteen plant specific elements in Reference 1 and the deviations from the generic WCAP design.
3.1 Plant-Soecific Elements 1.
Divergity The plant design should include a<1 equate diversity between the AMSAC equipment ano the existing Reactor Protection System (RPS) equipment.
Reasonable equipment diversity, to the extent practicable, is required to minimize the potential for coninon cause failures.
The AMSAC will not use any sensors or comoonents coninon to the RPS. The AMSAC logic circuits will be diverse from the RPS in both equipment and design. The switches usad to detect pumo and valve status will not inter-face with or orovide signals to the RPS. Therefore, we find the design acceptable with respect to diversity.
2.
Logic Power Sucolies logic power suoplies need not be Class IE, but must be capable of per-forming safety functions upon a loss of offsite power.
The logic power must come from a power source that is independent from the RPS power supplies.
The licensee has provided infonnation that verifies that the power supplies selected for the McGuire and Catawba AMSAC logic circuits will provide the ma* ' a i available independence from the RPS cower supplies.
The AMSAC wil' oe powered from non-safety power supplies capable of operating ur a loss of offsite powar. From our review of this infor-mation, we
<nd chat the logic power supplies meet the above requirements and are ar sotaale.
3.
Safety l'
<ated Interface The ir lementation of ATWS Rule shall be such that the existing Reactor Prot' ; tion System continues to meet all applicable safety criteria.
L e proposed McGuire and Catawba AMSAC design will have no interfaces with the RPS.
Interfaces with safety-related systems such as the Auxiliary Feedwater, Steam Generator Blowdown and Samling Systems will be designed such that the safety-related system will perfonn its function coincident with a postulated failure of the AMSAC. We, therefore, find the design acceptable with respect to interfacing.
Refer to item 9 below for more discussion.
4.
Quality Assurance The licensee shall provide information regarding compliance with Generic Letter (GL) 85-06, "Ouality Assurance for ATWS Equipment That Is Not Safety Related."
_ _ The 18 criteria of the NRC quality assurance guidance (GL 85-06) were re-viewed by the licensee. This resulted in some minor adjustments to the existing McGuire and Catawba quality assurance programs. These adjustments incorporated the ATWS equipment's specific requirements into the licensee's quality assurance practices and procedures for McGuire and Catawba. We find that the licensee has satisfactorily implemented the guidance of GL 85-06.
5.
Maintenance Bypasses Maintenance bypass indications should be incorporated into the continuous indication of bypass status in the control rooms.
The licensee provided infomation showing how maintenance is accomplished at power. Maintenance which can be perfomed at power will be accomplished utilizing various equipment such as isolacion valves, test switches, and sliding link terminals. Continuous indications of maintenance bypasses will be provided in the main control room.
The licensee connitted to implement a human factors review to ensure the effectiveness of the bypass indications and to ensure that they can be utilized in an efficient and readily under-stood manner. Ne find that these provisions meet the requirements regarding maintenance and that they are acceptable, subject to completion of the licensee's human factors review.
6.
Operating Byoasses 1N operating bypasses should be indicated continuously in the control row. The independence of the C-20 pemissive signal should be addressed.
The i cGuire and Catawba AMSAC operating bypass will be used to enable the l
d l
operators to bring the unit up in power during startup. These operating l
bypasses All defeat the MFW control valves status inputs to the AMSAC, as the control alves can modulate into and below the AMSAC setpoints during plant startup. A status light will be provided in the control room when any one of the four MFW control valves is less than 25% open.
Operating bypass controls, including continuous bypass status indication, will be provided in the main control room.
The licensee will conduct a human factors review of the bypass controls consistent with the plant's detailed control room design process.
The C-20 pemissive signal will originate from two pressure switches that monitor first-stage turbine impulse chamber pressure. This signal will not interface with the RPS.
We find the operating bypass controls and indications and the C-20 per-missive signal to be scceptable, subject to the satisfactory completion of the licensee's human factors review.
7.
Means for Bypassing The means for bypassing shall be accomplished by the use of a pemanently installed, human-factored, bypass switch or similar device. Disallowed methods for bypassing mentioned in the guidance should not be utilized.
_ _ _ _ The licensee's response stated that a control / selector switch with an indicating light will be used at both the McGuire and Catawba Nuclear Stations for the operating bypass function.
This switch will be used to manually instate the bypass when the reactor is below 40% power (the C-20 permissive setootnt). The operating bypass will be automatically removed when reactor power increases above the C-20 setpoint.
The disallowed methods for bypassing, such as lifting leads, pulling fuses, blocking relays, tripping breakers, will not be used. Therefore, we find the means for bypassing to be acceptable.
8.
Manual Initiation In the plant-specific submittal, the licensee discussed how manual turbine trio and auxiliary feedwater actuation are accomplished by the operator.
The operator uses existing manual controls to perforin a turtine trip and to start auxiliary feedwater flow. These controls were reviewed during the detailed control room design review. We find the provisions for manual initiation to be acceptable.
9.
Electrical Indeoendence From Existing Reactor Protection System Electrical independence is required from the sensor output to the final actuation device, at which point nonsafety-related circuits must be isolated from safety-related circuits by qualified Class IE isolators.
The licensee discussed how electrical independence is to be achieved.
The proposed AMSAC design will have no control devices, power supplies, or sensors in connon with the Reactor Protection System.- The existing safety systems with which the AMSAC interfaces will be the Auxiliary Feedwater Steam Generator Blowdown and Sampling systems. These inter-faces will be made by using existing, previously accepted, isolation devices. We, therefore, find the design provisions for electrical independence to be acceptable.
10.
Physical Separation From Existing Reactor Protection System The implementation of the ATWS' mitigating system must be such that the separation criteria applied to the existing Reactor Protection System are not violated.
The proposed AMSAC design will have no interfaces with the existing Reactor Protection System. Therefore, existing separation between the Reactor Protection System and nonsafety-related circuits will not be violated by the installation of the AMSAC equipment, and the above requirement is met.
11.
Environmental Qualification The olant-specific submittal should address the environmental qualification of ATWS equipment for anticipated operational occurrences.
The licensee stated that AMSAC equipment at McGuire and Catawba will be located in areas that are considered a mild environment.
A mild environ-ment is an environment that would at no time be significantly more severe
L m
than the environment that would occur during normal plant operation, including anticipated operational occurrences.
Environmental qualifi-i cations of electric equipment important to safety located in a mild environment are not included within the scope of 10 CFR 50.49.
12.
Testability at Power Measures to test the ATWS mitigating system before installation, as well as periodically, are to be established. Testing of the system may be performed with the system in the bypass mode.
.i -
The licensee stated that the AMSAC equipment will be tested prior to installation. After installation, the portion of the AMSAC design which uses MFW pump ttatus as an input will be fully testable at power. The MFW pump status-signal is generated by the pressure switches monitoring oil pressure on the stop valve control oil line. A selector switch and indicating light will be used to test each pressure switch individually.
The portion of the AMSAC design that utilizes MFW control and isolation valve position as an input will not be fully testable at power. Testing the control and isolation valve limit switches during power would adversely affect the operating status of the plant. This part of the AMSAC design will be tested at each refueling outage. The complete end-to-end testing of the AMSAC system from senso-through final actuation device will be perfonned with the plant shut down.
The licensee will conduct a human-factors review of the controls and indications used for testing purposes consistent with the plant's detailed control room design process.
We find these testing measures, including the above noted exception, to be acceptable, subject to satisfactory completion of the licensee's human factors review.
13.
Comoletion of Mitigative Action The licensee shall verify)that (1) the protective action, once initiated, goes to completion and (2 the subsequent return to operation requires deliberate operator action.
The licensee responded that, once initiated through AMSAC, the completion of mitigating action will be consistent with the existing plant turbine trip and auxiliary feedwater pump trio control circuitry.
This circuitry will automatically lock in upon initiation and must be manually reset. The C-20 pennissive signal will be electrically latched in niace automatically above the 40% power level and will require manual action for reset (unlatched) when the C-?O signal falls below the 40% power level setooint.
This latching feature will allow the AMSAC mitigative action to go to completion in the event of a turbine trip and the turbine first stage impulse pressure is lost. We find these design provisions for completion of mitigative action and return to operation to be acceptable.
~
7-14.
Technical Specifications The equipment required by the Al'WS Rule to reduce the risk associated with an. ATWS event must be designed to perfom its functions in a reliable manner.
A method acceptable to the staff for demonstrating that the equipment satisfies the reliability requirements of the ATWS Rule would be to pro-i vide appropriate AMSAC technical specifications. Such technical specifications would include operability and surveillance requirements.
In its interim Connission Policy Statement of Technical Specification Improvements for Nuclear Power Plants (52 Federal Reflister 3788 February 6,1987), the Comnission establisheo a spec' fic set of objective criteria for detemining which regulatory rtquirements and operating restrictions should be included in technical specifications. The staff is presently reviewing ATWS requirements to criteria in this Policy Statement to detemine whether and to what extent technical specifications on the ATWS Rule are appropriate.
The licensee's position, by letter dated September 15, 1986 (Ref. 8),
is that the AMSAC requirements do not meet the proposed selection criteria for items to be included in technical specificatiods and, therefore, should be maintained in administrative 1y controlled documents other than technical specifications. The licensee would rely on compliance to Generic Letter 85-06, "Quality Assurance Guidance for ATWS Equipment That Is Not Safety Related," and controlled documents subject to the Duke Power Company manual, "Nuclear Production Department Administrative Policy Manual for Nuclear Stations."
Accordingly, this aspect of the staff review remains open pending completion of, and subject to the results of, the staff's further review. 'The staff will provide guidance regarding the technical specification requirements for ANSAC at a later date.
3.2 Deviations from Generic Design The staff's review of the plant specific submittals revealed that the licensee is implementing an AMSAC design which deviates from the approved WCAP. First, the AMSAC response generated by the MFW pump trip signal for initiation of the AFW pumps and tripping the main turbine will not be interlocked with any time delay or the C-20 pemissive. Thus, upon loss of the MFW pumps at any power level, the AFW pumps will be started and the turbine will be tripped. The licensee has informed the staff that the AMSAC design is consistent with the original design bases for the Catawba and McGuire Nuclear Stations which call for starting the auxiliary feedwater pumps and tripping the main turbine j
innediately upon loss of main feedwater. This action is anticipatory to low-low Further, the licensee has stated that the most prudent steam generator level.
and conservative approach is to trip the main turbine upon loss of both MFW pumps to avoid unnecessary depletion of steam generator inventories and to reduce the amount of relatively cold auxiliary feedwater required to stabilize steam generator levels. We have reviewed this feature of the design and agree that i
the licensee's approach is conservative and, therefore, acceptable.
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8-Second, the licensee has chosen not to in'terlock the C-20 permissive with a i
1 time delay on de-energization. The C-20 circuitry to be installed at Catawba and McGuire will electrically latch in above 40 percent reactor power (see J
Item 13 above). Once latched, it will require deliberate operator action to fnstate a bypass of the C-20 signal. Such manual action will be allowed only when the reactor is below the 40 percent power level. Thus, should the first-stage turbine impulse pressure be lost for any reason, the C-20 permissive will remain latched independent of reactor power, and the AMSAC 2
logic will remain anned at all power levels until the operator manually instates the bypass as discussed above. The incorporation of a C-20 time delay would be redundant to the existing latching feature. Accordingly,we find the licensee's position acceptable. Administrative procedures used to direct the operators regarding usa of the C-20 pennissive when below 40 percent power will be develcoed by the licensee as part of the nuclear station modification process.
4.0 CONCLUSION
S The staff cc,ncludes, pending and subject to final resolution of the technical specification issue, that the AMSAC design, as proposed by Duke Power Company for the Catawba and McGuire Nuclear Stations, is acceptable and is in compliance with the ATWS Rule,10 CFR 50.62, paragraph (c)(1). Staff conclusion is further subject to the successful completion of certain noted human factors engineering reviews to which the licensee has comitted. Until staff review is completed regarding use of technical specifications for ATWS requirements, the licensee should continue with the scheduled installation and implementation (olanned operation) of the ATWS design utilizing administratively controlled t
procedures.
5.0 REFERENCES
1.
~ Letter, C. E. Rossi (NRC) to L. D. Butterfield (WOG), "Acceptance for Referencing of Licensing Topical Report," July 7,1986.
2.
Letter, H. B. Tucker (Duke) to U.S. NRC, "ATWS/AMSAC Design Description," January 23, 1987.
3.
Letter, R. A. Newton (WOG) to J. Lyons (NRC), "Westinghouse Owners Group Adde.dum 1 to WCAP-10858-P-A and WCAP-11293-A:
AMSAC General l
Design Package," February 26, 1987.
I 4,
Letter, D. Hood (NRC) to H. B. Tucker (Duke), "Request for Additional Information Conceraing the ATWS/AMSAC Design at the McGuire and Catawba Nuclear Stations," February 27, 1987.
l 5.
Letter, H. B. Tucker (Duke) to U.S. NRC, "ATWS/AMSAC Design l
Description," April 9,1987.
I 6.
Letter, D. Hood (NRC) to H. B. Tucker (Duke), "Request for Additional Infonnation Concerning the ATWS/AMSAC Design at the 1
McGuire and Catawba Nuclear Stations," June 17, 1987.
7.
Letter, H. B. Tucker (Duke) to U.S. NRC, "ATWS/AMSAC Design Description," June 18, 1987, i
l(- -. -
t,
8.
Letter H. B. Tucker (Duke) to H. R. Denton (NRC), "McGuire Nuclear Station Docket Nos. 50-369, 50-370," September 15, 1986.
9.
Letter, R. A. Newton (EOG) to J. Lyons (NRC), "Westinghouse Owners Group Transmittal of Topical Report WCAP-10858-P-A, Revision 1: AMSAC Generic Design Package," August 3, 1987
- 10. Letter, K. N. Jabbour (NRC) to H. B. Tucker (Duke), "Request for Additional Inforwation Concerning the ATWS/AMSAC Design at the McGuire and Catawba Nuclear Stations," October 13, 1987.
- 11. Letter, H. B. Tucker (Duke) to USNRC, "ATWS/AMSAC Design Review,"
October 20, 1987.
e
s Mr. H.,B. Tucker Duke Power Company McGuire Nuclear Station cc:
Mr. A.V. Carr, Esq.
Dr. John M. Barry Duke Power Corgany Department of Environmental Health P. O. Box 33189 Mecklenburg County 422 South Church Street 1200 Blythe Boulevard Charlotte, North Carolina 28242 Charlotte, North Carolina 28203 County Manager of Mecklenburg County 720 East Fourth Street Charlotte, North Carolina 28202 Chaiman, North Carolina Utilities Cosmission Mr. Robert Gill Dobbs Building Duke Power Company 430 North Salisbury Street Nu' clear Production De vr m it Raleigh, North Carolina 27602 P. O. Box 33189 Charlotte, North Carelh.: 28242 Mr. Dayne H. Brown, Chief Radiation Protection Branch J. Michael McGarry, III Esq.
Division of Facility Services Bishop, Liberman, Cook, Purcell Department of Human Resources and Reynolds 701 Barbour Drive 1200 Seventeenth Street, N.W.
Raleigh, North Carolina 27603-2008 Washington, D. C.
20036 Senior Resident Inspector c/o U.S. Nuclear Regulatory Comission Route 4, Box 529 Hunterville, North Carolina 28078 Regional Administrator, Region II U.S. Nuclear Regulatory Commission, 101 Marietta Street, N.W., Suite 2900 Atlanta, Georgia 30323 L. L. Williams Area Manager, Mid-South Area ESSD Projects l
Westinghouse Electric Corporation MNC West Tower - Bay 239 l
P. O. Box 355 l
Pittsburgh, Pennsylvania 15230 l
i i
l
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Mr. H. B. Tucker Duke Power Company Catawba Nuclear Station CC' A.V. Carr, Esq.
North Carolina Electric Membership
'Ouke Power Company Corp.
422 South Church Street 3400 Sumner Boulevard Charlotte, North Carolina 28242 P.O. Box 27306 Raleigh, North Carolina 27611 J. Michael McGarry, III, Esq.
Bishop Libennan, Cook, Purcell Saluda River Electric Cooperative, and Reynolds Inc.
1200 Seventeenth Street, N.W.
P.O. Box 929 Washington, D. C.
20036 Laurens, South Carolina 29360 North Carolina MPA-1 Senior Resident Inspector Suite 600 Route 2, Box 179N 3100 Smoketree Ct.
York, South Carolina 29745 P.O. Box 29513 Raleigh, North Carolina 27626-0513 Regional Administrator, Region II U.S. Nuclear Regulatory Connission, L.L. Williams 101 Marietta Street, NW, Suite 2900 Area Manager, Mid-South Area Atlanta, Georgia 30323 ESSD Projects Westinghouse Electric Corp.
MNC West Tower - Bay 239 P.O. Box 355 Pittsburgh, Pennsylvania 15230 Mr. Hepard G. Shealy, Chief Bureau of Radiological Health South Carolina Departme:nt of Health and Environmental Control 2600 Bull Street Columbia, South Carolina 29201 County Manager of York County York County Courthouse Karen E. Long York South Carolina 29745 Assistant Attorney General N.C. Department of Justice Richard P. Wilson, Esq.
P.O. Box 629 Assistant Attorney General Raleigh, North Carolina 27602 S.C. Attorney General's Office P.O. Box 11549 Spence Perry, Esquire Columbia, South Carolina 29211 General Counsel Federal Emergency Management Agency Piedmont Municipal Power Agency Room 840 100 Memorial Drive 500 C Street Greer, South Carolina 29651 Washington, D. C.
20472 Mr. Michael Hirsch Federal Emergency Management Agency Office of the General Counsel Room 840
$.5 20472 hn o Brian P. Cassidy, Regional Counsel Federal Emergency Management Agency, Region !
J. W. McConnach POCH
.