ML20236G767
| ML20236G767 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 10/30/1987 |
| From: | Cockfield D PORTLAND GENERAL ELECTRIC CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| NUDOCS 8711030266 | |
| Download: ML20236G767 (12) | |
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. Pbrtland General ElectricCormany
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A David W. Cockfield Vice President, Nuclear -
october 30.-1987 Trojan Nuclear Plant.
Docket 50-344L License NPF-1 U.S. Nuclear. Regulatory Commission
-ATTN:. Document Control Desk
.)
Washington DC 20555 l
l
Dear Sir:
Plant-Specific Response'to-.the.
Nuclear-Regulatory Commission (NRC) Safety Evaluation of l
j WCAP-10858 " Anticipated Transient Without Scram (ATWS)
J Mitigating System Actuation Circuitry (AMSAC) Generic Design Package">
s i
References:
1.
NRC Let'ter to Portland General 1 Electric-Company (PCE) Dated September 23, 1986. Transmitting Safety Evaluation'of Topical-Report (WCAP-10858), "AMSAC Generic Design Package".
j 2.
PGE Letter to NRC Dated' October-29, 1986, " Response to NRC
)
Safety Evaluatioa on AMSAC Generic Design Package".
l to this letter transmits Plant-specific-design' features for AMSAC requested by and discussed in Reference 1, as committed to in Reference-2.
Because the Trojan AMSAC design is still undergoing development,. additional l
updates of specific design.fcatures will be necessary
_In particular,' specific information concerning the fault testing of isolation'dovices and testability at.
power is not presently available because an.AMSAC vendor has not.been. selected.
Thir, information will be provided to the NRC by September 30, 1988.
Sincerely,-
Attachment c:
Mr. John B. Martin Regional Administrator, Region V U.S. Nuclear Regulatory Commission
-Mr. Dave Yaden, Director State of' Oregon Department of Energy Mr. R. C. Barr NRC Resident Inspector Trojan Nuclear Plant i.4 E
8711030266 871030
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PDR ADOCK 05000344
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' 121 S.W Salmon Street, Portand, Oregon 97204 u_=_ __2. _ _ __z__
O Trojan Nuclear Plant Document Control Desk Docket 50-344 October 30, 1987 License'NPF-1 Page 1 of 9 PLANT-SPECIFIC RESPONSE TO NUCLEAR REGULATORY C0KMISSION (NRC) SAFETY EVALUATION OF WCAP-10858, " ANTICIPATED TRANSIENT WITHOUT SCRAM ( ATWS)
MITIGATING SYSTEM ACTUATION CIRCUITRY (AMSAC) GENERIC DESIGN PACKAGE" Portland General Electric Company (PGE) plans to meet the requirements of.
Title 10 Code of Federal-Regulations, Part 50.62(c)(1) by implementing the Logic 1 conceptual' design described in WCAP-10858P-A: Revision 1,.
"AMSAC Generic Design Packago".
.The. Logic 1 design provides for AMSAC actuation on low steam generator water level.
The AMSAC system is currently in the preliminary design stage..The following responses to the Plant-specific design elements denoted in the NFC'u Safety Evaluation on WCAP-10858 nre provided.
1.
NRC Position - Diversity The plant specific subtaittal should indicate the degree of diversity that exists between the AMSAC equipment and the existing Reactor Protection System. Equipment diversity to the extent reasonable.and practicable to minimize the potential.for common cause failures is required from the sensors output to, but not including, the-final actuation device, eg, existing circuit breakers may be used for the auxiliary feedwater initiation. The sensors need not be of a diverse design or manufacture.
Existing protection system instrument-sensing lines, sensors, and sensor power supplies may be used. Sensor and instrument sensing lines should be selected such that adverso interactions with existing control systems are avoided.
PCE Response - Diversity Trojan's Rcactor Protection System (RPS) and the proposed AMSAC equipment are tabulated below (ie, l'nput, output, and logic).
Diverse Equipment Existing Proposed (Yes/No)
Sensor Barton 763 and 764
.Same as existing No Instrument sensing Class 1E instrument Same as existing No lines lines Sensor power Class 1E Same as existing No j
supply
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- Trojan Nuclear Plant' LDocket 50-344'
.' October 30,'1987L License NPF-1' '
Page 2 of-9' i
'. Diverse 1
Equipmen+
Existing Proposed (Yes/No)-
Input isolation Westinghouse-(W)/
Energy. Inc.-(EI).
Yes' Hagan 7100
'Feries FCA 300 or;
'SC 993:
l'
. Input-icolation W/Hagan 7100
' Class 1E, 120-V-acM zYes power supply **
.to 115;V dc*:
Logic system
'W/Hagan 7100 Programmable
.YesJ controlleror i
microprocessor
- Output isolation Potter-Brumfield NotiaLPotter-r Yes 1
rotary-type MDR
'BrumfieldLrotary relay relay *.
l Actuating circuits Auxiliary feedwater A Train Turbine controls Same as. existing
.No B Train.
Diesel controls;
.Same as/ existing; nod
.-Main turbine
. Alarm'and trip; Same as existing.
No
. control
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- Actual manuftetut er/ type is to 'be determined' af terfvendor bid-acceptance.
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-** Power'cupply may be integral'with the isolation devic'e.
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i-The W/Hagan 7100 equi.pment is hardwired.1 as comparedito:EIL Series FCA 300 or SC 993 which use' optical l isolation and:the' proposed i
programmable controller-based -(or microprocessor) : logic. The outputi isolation devices will be relays which utilize coil to contact' isola-
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tion. All of the.AMSAC components will~be of a different manufacturer l
and design principir; than the W/Hagan 7100 equipment?
H 2.
NRC Position - Logic Power Supplies l
The plant specific submittal should discuss.the logic power supply.
' design.. According'to the rule, the1AMSAC. logic powertsupply is'not; required to be~ safety-related (Class 11E).: However,. logic power shduld
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be from an: instrument power' supply that~is independent-from'the reactorz a
L
. protection system l(RPS) power supplies..lOer review:of additional information' submitted by WOG indicated!that' power.to the" logic" circuits 1 will' utilize RPS batteries and inverters.
The' staff finds this portion u
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t a
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- Trojew Nuclear Plant" Document Control Desk ~
Dockot'50-344 October 30.'1987-License NPF Attachment 1-Page 3 of 9 of'thedesignunacceptable,therefore[ independent lpowersupplies should be provided.
I PGE Response " Logic Power Suppli
-j The logic. power supply which is:being considered is;the Plant. computer j
uninterruptible power supply (UPS). This UPS is'a Lortec Model 153CRH which has a sealed lead-calcium cell and can-be powered from one of two Claus 1E/ emergency diesel generator (EDG) backed' sources. :The power supply'itself is not Class 1E and is properly' isolated from the q
RPS power supply system.
)
3.
{l3C Position - Safety-Related Interface The plant' specific submittal should show that the implementation is-such that the existing protection system continues. to meet all applicable safety criteria.
PGE Response
' Safety-Related Interface l
1 The isolation device-is the interface that separates AMSAC.from Class 1E equipment. The-isolation devices will be designed'and qualified consistent with the Institute of Electrical and Electronics l
Engineers (IEEE) Standard Criteria for Independence'of' Class.1E Equipment ~and Circuits, IEEE 384-1977.
Environmental qualification:
j will follow PGE Topical. Report PGE-1025, " Environmental-Qualification Program Manual". Adherence to-these guides will' ensure'that the existing RPS will continue.to meet all applicable safety criteria.
1 4.
NRC Position - Quality' Assurance 1
~The plant specific submittal should provide information'regarding
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compliance'with Generic Letter 85-06, " Quality Assurance. Guidance for ATWS Equipment that is not' Safety-Related".
PGE Response - Quality Assurance The Trojan Nuclear Plant Quality Assura' ?.e (QA) Program complies with'the guidance-in NRC G2neric Letter 85-06, " Quality. Assurance Guidance for ATWS Equipment That is'Not Safety-Related". Appendix G.
l to PGE Topical Report PGE-8010, " Nuclear Quality Assurance Program",-
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describes:the QA and administration controls. implemented for non-safety-related ATWS equipment.
If'the.AMSAC vendor will supply i
equipment to"the:more' stringent requirements of the Trojan Nuclear Plant 1QA Program which'is in full. compliance with'10 CFR 50, Appendix B,.then such requirements,may be used...In all cases,-the requirements of Appendix G.to'PCE-8010 will.be~ met for non-safety-related ATWS equipment.
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-Trojan Nuclear Plant Document Control Desk Docke6 56 T44 October 30, 1987 Licenac at/-l Page 4 of 9 5.
NRC Position - Maintenanew Bypasses-The plant specific submittal should discuss how maintenance at power is accomplished and how good human factors engineering per'tice is incorporated into the continuous indication of bypass status in the control room.
PCE Response - Maintenance Bypasses Maintenance bypasses will not involve lifting leads, pulling fuses, tripping breakers, or physically blocking rolays.
I Four permanently installed steam generator level input channel bypass switches or similar devices will be provided to enable one-channel to be bypassed at a given time. When one of these switches or devices is enabled, then the bypassed channel will be placed in the tripped mode. The "three-out-of-four"' input trip logic will then become a "two-out-of-four".
There will be local AMSAC cabinet.
Indicating lights that will show bypass status' Contacts will be provided by the vendor for wiring to the' control room annunciator system.
Two permanently installed turbine load channel bypass switches or similse devices will be provided.which will operate as described in.
1 the previous paragraph. However, their permissive trip logic will
.j change from a "two-out-of-two" to a "one-out-of-two".
)
AMSAC bypass is required for testing and maintenance not accessible by the channel bypass.. Control room annunciation of the AMSAC bypass shall be consistent with PGE Topical Report PGE-1041,
" Detailed Control Room Design Review Report". P9E-1041 includes the human factors design aspects of. control room. indication. Because Trojan's annt:nciator system will be replaced in 1989, a revision to PGE-1041 will be made and the new revision followed.
6.
NRC Position - OperatinR Bypasses The plant specific submittal should state that operating bypasses are continuously indicated in the control room; provide the basis for the 70 percent or plant specific operating bypass. level; discuss the human factors design aspects of the continuous indication; and discuss the. diversity and independence of the C-20' permissive signal (DefeatH the block of AMSAC).
PGE Response - OperatinR Bypasses The indication of operating bypasses will be consistent. with the future revicion of PGE-1041 discussed;,in Item 5.
The C-20 per-missive signal defeats.the block of AMSAC, and its setpoint will be-
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q Trojan Nuclear Plant Document Control Desk
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Docket 50-344 October 30, 1987 j
License NPF-1 Page 5 of 9
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at 40 percent nominal turbine load. The basis for this setpoint is
.I contained in a Westinghouse Owners Group (WOG) letter to the NRC,
' 0G-87-10, dated February 26, 1987.
i Diversity and independence of the C-20 permissive signal are tabulated in Item 1.
Existing sensors, instrument sensing lines, and sensor power supplies will be used. However, diverse isolators will be used for the input from the turbine impulso pressure and steam generator level sensors.
Independence is achieved through the use of diverse Class lE isolating devices.
H I
i 7.
NRC Position - Means for Bypassing The plant specific submittal should state that the means for
]
bypassing is accomplished with a permanently installed, human i
factored, bypass switch or similar devico, and verify that disallowed methods mentioned in the guidance are not utilized.
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PQE Response - Means for Bypassing Means for bypassing and disallowed methods are discussed in Item'5.
Bypassing AMSAC will be accomplished with a permanently installed,-
human-factored bypass switch or similar device.
I i
8.
NRC Position - Manual Initiation The plant specific submittal should diccuss how a manual turbine trip and auxiliary feedwater actuation are accomplished by the operator.
1 PGE Response - Manual Initiation Manual initiation of turbine trip and auxiliary feedwater (AFW) are both initiated from Panel CD5 in the control room.
Turbine trip 18 accomplished by depressing the trip push button.
Each safety-related AFW pump is actuated individually.
TLa diesel-driven AFW pump requires placing the control switch'in the start position. The steam-driven AFW pump requires opening at least one steam supply-valve prior to placing'the pump control switch in start.
Emergency l
Instruction EI-0, " Reactor Trip Safety and Injection and Diagnosis",
provides the required instructions.
9.
'NRC Fosition - Electrical Independence From Existina P."C The plant *occific submittal should show that' electrical indepen-dence is scaleved. This'is required from the sensor output to the-
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Docket 50-344 October-30. 1987 License NPF-1 Page 6 of 9
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final actuation device at which point non-safety-related circuits must be isolated from safety related circuits by qualiff.ed Class 1E isolators. Use of existing isolators is acceptable. However, each plant specific submittal should provide an analysis and tests which' j
demonstrates that the existing isolator will function under the
'l maximum worst case fault conditions.
The required nothod for quali-fying either the existing or diverse isolators is presented in 1
Appendix A.
.PCE Response - Electrical Independence From Existing RPS
]
AMSAC electrical independence from the existing RP.S will he accom-plished by the use of qualified Class 1E isolation devices for interface with Class IE. input and output circuits. These isolators will be class IE devices and will isolate Class IE circuits from credible voltage / current faults which are postulated to exist in non-Class 1E circuits.
Item 3 describes the critoria documents that will be followed. The specific information on isolation devices requested by Appendix A of the NRC's Safety Evaluation will be provided after an AMSAC vendor is selected.
10.
NRC Position - Physical Separation From Existing RPS
.j Physical separation from existing reactor protection system is not required unless redundant divisions and channels in the existing reactor trip system are not physically separated.
The implementa-tion must be such that separation' criteria applied to the exirting protection system are not violated.
The plant specific submittal should respond to this concern.
PCE Response - Physical Separation From Existing RPS Trojan's RPS has physical separation between redundant divisions and channels.
AMSAC will be located whero interactions with the protec-tion set cabinets will not occur.
Isolation of inputs will be accomplished in the associated protection set cabinet to prevent the reduction of RPS reliability. Routing of cables from the protection set cabinets, where the input isolation devices are located..to AMSAC will be independent of the protection system cable routing.
Separation criteria will be applied to the RPS por IEEE Standard 384-1977.
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4 Trojan Nuclear Plant Document Control Desk $
Docket 50-344
-October 30, 1987 Attachment.1 License NPF-1 C
Page 7 of 9 3
11.
NRC Position - Environmental Qualification The plant specific submittal shbuld' address the environmental j
qualification of ATWS equipment for anticipated operational z
occurrences only, not for accidents.
PGE Response - Environmental Qualification Sensors and support cable systems are environmentally qualified for normal as well as design basis accidents. The AMSAC_ logic unit and isolation devices will be 1ocated in a mild environment.
The only
~
anticipated operational occurrences that would affect AMSAC are loss-of all offsite power and loss of the heating and ventilation system.
.]
Since there will be a UPS system for AMSAC which is EDG-backed, loss 1
of offsite power will not prohibit AMSAC's protective function, j
Also, AMSAC will be located where there is a redundant-train, i
l Claso 1E heating and ventilation system; therefore, loss of offsite power will not affect AMSAC's operation due to high-temperature' l
effects.
AMSAC equipment will be designed to perform its function during j
normal operating conditions as speciflod in PGE Topical Report J
PGE-1025, " Environmental Qualification Program Manual".
12.
NRC Position - Testability at Power Measures are to be established to test, as' appropriate, non safety related-ATWS equipment prior to installation and periodically.
Testing of AMSAC may be performed with AMSAC in bypass. Testing of
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AMSAC outputs through the final actuation devices will be performed.
with the plant shutdown. The plant specific submittals should present the test program and state that the output signal is indicated in the control room in a manner consistent with plant practices including human factors.
PGE Response - Testability st Power Specific information pertaining to AMSAC testing will be provided after an AMSAC vendor is selected.
l l
l 13.
NRC Position - Completion of-Mitigative Action AMSAC shal1 4e designed so that, once actuated,Tthe completion of l
mitigating action shall be consistent with the plant turbine trip and auxiliary feedwater circuitry. Plant "pocific submittals should verify that the protective' action, once. initiated, goes'to comple-tion, and that the-subsequent return to operation: requires.
deliberate cperator action.
L
5 j-y Trojan Nuclear / Plant-DocumentControlDeslb
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-Docket'50-344.
l0ctober 30', 1987 License'NPF-1
. Attachment 1 Page 8 of;9-PGE Response' - Completion of Mi laative* Action
~Once'AMSAC.is actuated:
'a ; The completion of.the' mitigating.' action will be consistentLwith main turbine ~ trip"and'AFW: circuitry.,
b.
The protective action will go.to' completion.
<c. ! Subsequent-return to operacion will requiro deliberate ~ operator-
- action.-
Mainturbine,andgeneratortrikisaccomplishedbyenergizingthei
-master trip bus'as shown on-Figure 1.f.The output _ isolation device will be diverse from the. existing'. turbine. trip de' vices.and' separated from other:AMSAC tripLdevices that provide Class.1E isolation,:sincef the master trip bus'circult:is non-Class 1E.'
AFW is divided into Trains A and B which'are both Class 1E.
Thel
- output isolation devices will-be.' diverse and separated from"each.
.other as well as from.the non-Class'1E" isolation device.: E An ' initi-- :
~
ating contact will be installed.in parallel with the existing: steam :
generator low-low' level.RPS initiatingicontactifor each. channel g
(Figure 2).
The A-channel turbine-driven : AFW pump. circuit performs the following' on " Auto-Start":
a.
Opens l air-operated steam supply valves,'one:for each of-the"four
[.
- steam generators, CV-1451 through CV-1454.
b.
Opens the turbine trip and throttle valve',jM0 3'071.'.
1 Isolates steam generator blowdown and ' sampling, c.
d.
Activates an alarm for " AUX FW AUTO START".
4 x
The B-channel diesel'-driven'AFW pump circuit performsfa'similar l
function during " Auto Start":-
a.
Energizes diesel-st' art relay'which. starts the' diesel'with a.
- normal. control' switch lineup.
b.
' Opens service water isclation~. valve,;MO-3060B, which suppliest l
g cooling water to-the diesel l1ube o111. cooler, pump 11ubeloilf 2
cooler, diesel' engineer jackett cooler, and speedLincrea'ser.!1ube' oil: cooler.
X
.c.
Isolates' steam. generator blowdown'and" sampling.
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Trojan Nuclear 'Plcnt Document Control Desk Docket 50-344-October 30, 1987 License NPF-1 Page 9 of 9 i
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Actuates an alarm for " DIESEL AUTO START".-
I Once the initiating signal is received by.the latch-type isolator i
device, the initiating contact will remain closed until the initi-ating circuit is reset by the operator. This~ design will ensure that the protective action will go to completion and that subsequent i
return to operation will require deliberate operator action.
14.
NRC Position - Technical Specifications Technical specification requirements related to AMSAC will have to be addressed by plant specific submittals.
PCE Response - Technical Specifications For the reasons presented in the WOG letter to the NRC dated February 10, 1986 (OG-171), Technical Specifications for'AMSAC are not considered to be necessary. Normal Trojan Nuclear Plant administrative controls will be utilized to address AMSAC operability and surveillance requirements.
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