ML19321B153
| ML19321B153 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 07/10/1980 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Linder F DAIRYLAND POWER COOPERATIVE |
| References | |
| TASK-03-03.B, TASK-03-10.A, TASK-05-11.A, TASK-06-07.C1, TASK-3-10.A, TASK-3-3.B, TASK-5-11.A, TASK-6-7.C1, TASK-RR NUDOCS 8007280190 | |
| Download: ML19321B153 (18) | |
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,g UNITED STATES NUCLEAR REGULATORY COMMISSION y
Q fe E WASHINGTON, D. C. 20555
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Ju.ly 10, 1980 Docket No. 50-409 Mr. Frank Linder General Manager Dairyland Power Cooperative 2615 East Avenue South La Crosse, Wisconsin 54601
Dear Mr. Linder:
RE: SEP TOPICS III-10.A, V-11. A, VI-7.C.1, and III-3.B (La Crosse Boiling Water Reactor)
Enclosed is a copy of our current evaluation of Systematic Evaluation Program Topics III-10.A, Thermal-0verload Protection for Motors of Motor-Operated Valves; V-11. A, Electrical, Instrumentation, and Control Features for Isolation of High and Low Pressure Systems; VI-7.C.1, Independence of Redundant Onsite Power Systems; III-3.B DC Power System Bus Voltage Monitoring and Annunciation.
This assessment comares your facility, as described in Docket No. 50-409 with the criteria currently used by the regulatory staff for licensing new facilities.
Please inform us if youras-built facility differs from the licensing basis assumed in our assessment within 90 days of receipt of this letter.
These evaluations will be basic inputs to the integrated safety assessments for your facility unless you identify changes needed to reflect the as-built conditions i
at your facility. The topic assessments may be revised in the future if your
~l facility design is changed or if NRC criteria relating to the topics is modified before the integrated assessments are completed.
Sincerely, enn s M. Crutchfield, Chie Operating Reactors Branch #5 Division of Licensing
Enclosures:
Cog leted SEP Topics III-10.A, V-ll.A, VI-7.C.1, and III-3.B cc w/ enclosure:
See next page l
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.8007280 l 10
Mr. Frank Linder July 10, 1980 1
cc w/ enclosures:
Fritz Schubert, Esquire Director, Technical Assessment Staff Attorney oivision Dairyland Power Cooperative Office of Radiation Programs 2615 East Avenue South (AW-459)
La Crosse, Wisconsin 54601 U. S. Envirormental Protection
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Agency g.:
- 0. S. Heistand, Jr., Esquire Crystal Mall #2 E
Morgan, Lewis & Bockius Arlington, Virginia 20460 1800 M Street, N. W.
Washington, D. C.
20036 U. S. Environmental Protection Agency Mr. R. E. Shimshak Federal Activities Branch La Crosse Boiling Water Reactor Region V Office Dairyland Power Cooperative ATTN:
EIS COORDINATOR P. O. Box 135 230 South Dearborn Street Genoa, Wisconsin 54632 Chicago, Illinois 60604 Coulee Region Energy Coalition Charles Bechhoefer, Esq., Chairman
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ATTN: George R. Nygaard Atomic Safety and Licensing Board P. O. Box 1583 U. S. Nuclear Regulatory Comission La Crosse, Wisconsin 54601 Washington, D. C.
20555 La Crosse Public Library Dr. George C. Anderson 800 Main Street Department of Oceanography La Crosse, Wisconsin 54601 University of Washington Seattle, Washington 98195 Mrs. Ellen Sabelko Society Against Nuclear Energy Mr. Ralph S. Decker 929 Cameron Trail Route 4, Box 1900 E
Eau Claire, Wisconsin 54701 Cambridge, Maryland 21613 F
Town Chairman Dr. Lawrence R. Quarles Town of Genoa Kendal at Longwood, Apt. 51 Route 1 Kenneth Square, Pennsylvania 19348 Genoa, Wisconsin 54632
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Chairman, Public Service Comission Thomas S. Moore, Esq.
of Wisconsin Atomic Safety and Licensing Appeal Board Hill Farms State Office Building U. S. Nuclear Regulatory Comission u..
Madison, Wisconsin 53702 Washington, D. C.
20555 g
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Alan S. Rosenthal, Esq., Chairman Mr. Richard E. Schaffstall Atomic Safety and Licensing Appeal Board XMC Incorporated U. S. Nuclear Regulatory Comission 1747 Pennsylvania Avenue, NW Washington, D. C.
20555 Washington, D. C.
20006 b
4 SEP TECHNICAL EVALUATION TOPIC III-10.A THERMAL-0VERLOAD PROTECTION FOR MOTORS OF MOTOR-OPERATED VALVES LACROSSE TOPIC III-10.A Thermal-Overload Protection for Motors of Motor-Operated Valves The objective of this review is to provide assurance that the appli-cation of thermal-overload protection devices to motors associated with safety-related motor-operated valves do not result in needl'ess hindrance of the valves to perform their safety functions.
In accordance with this objective, the application of either one of the two recommendations contained in Regulatory Guide 1.106, " Thermal-Overload Protection for Electric Motors on Motor-Operated Valves," is ade-quate. These recommendations are as follows:
(1) Provided that the completion of the safety function is not jeopardized or that other safety systems are not degraded, (a) the thermal-overload protection devices should be continucusly bypassed and temporarily placed in force only when the valve motors are undergoing periodic or maintenance testing, or (b) those thermal-overload protection devices that are normally in force during plant operation should be bypassed under acci-dent conditions.
(2) The trip setpoint of the thermal-overload protection devices should be established with all uncertainties Tesolved % favor of completing the safelsted action. With respect to those unces:ainties, consider-ation should be given to (a) variations in the ambient temperature at the installed location of the overload 1
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protection devices and the valve motors, (b) inaccura-cies in motor heating data and the overload protection device trip characteristics and the matching of these two items, and (c) setpoint drif t.
In order to ensure continued functional reliability and the accuracy of the trip point, the thermal-overload protection device should be periodically tested.
In addition, the current licensing criteria require that:
(3) In MOV designs that use a torque switch to limit the opening or closing of the valve, the automatic opening or closing signal should be used in conjunction with a corresponding limit switch.
DISCUSSION Review of Lacrosse drawings shows that the only motor-operated valves supplied power from ESF buses are the alternate core spray valves.' All
,of these valves have thermal-overload protection devices which are not bypassed; there is no docketed information to ' indicate that TOL trip set-points have been set to comply with Criterion 2,\\above. Additionally,
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automatic valve operate commands are terminated by corque switches rather chan limit switches.
EVALUATION 1
Thermal-overload protection for motor-operated valves at Lacrosse does not comply with current licensing criteria. Thermal-overload devices are not bypassed, no information is available to support adequacy of trip set-points, and torque switches rather than limit switches are used to terminate valve travel.
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d REFERENCES 1.
IEEE Standard 179-1971, " Criteria for Protection Systers for Nuclear Power Generating Stations."
2.
Branch Technical Position EICS'3-27, " Design Criteria for Thermal Over-load Protection for Motors of Motor-Operated Valves."
3.
Regulatory Guide 1.106, " Thermal Overload Protection for Electric Motors on Motor-Operated Valves."
4.
Lacrosse Drawing 503775, Revision 10, dated 5-2-78.
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s TOPIC V-ll.A SEP TECHNICAL EVALUATION REPORT ELECTRICAL, INSTRUMENTATION, AND CONTROL FEATURES FOR ISOLATION.0F HIGH AND LOW PRESSURE SYSTEMS LACROSSE NUCLEAR SIATION
1.0 INTRODUCTION
The purpose of this review is to determine if the electrical, instrumentation, and control (EI&C) features used to isolate systems with a lower pressure racing than the reactor coolant primary system are in compliance with current licensing requirements as outlined in SEP Topic V-11A.
Current guidance for isolation of high and low pres-sure systems is contained in Branch Technical Position (BTP) EICSB-3, BTP RSB-5-1, and the Standard Review Plant (SRP), Section 6.3.
2.0 CRITERLA 2.1 Residual Heat Removal (RER) Systems.
Isolation requirements for RER systems contained in BTP RSB-5-1 are:
(1) The suction side must be provided with the following isolation features:
(a) Two power-operatad valves in series with posi-tion indicated in the control room.
(b) The valves must have independent and diverse interlocks to prevent opening if the reactor coolant system (RCS) pressure is above the design pressure of the dHR system.
(c) The valves aust have independent and diverse interloc?ts to ensure at least one valve closes upon an increase in RCS pressure above the design pressure of the RER system.
(2) The discharge side must be providad with one of the following features:
(a) The valves, position indicators, and interlocks described in (1)(a) through (1)(c) above.
(b) One or more check valves in series with a nonnally-closed power-operated valve which has 1
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its position indicated in the control room.
If this valve is used for an Emergency Core Cooling System (ECCS) function, the valve must open upon receipt of a safety injection signal (SIS) when RCS pressure has decreased below RER system design pressure.
(c) Three check valves in series.
(d) Two check valves in series, provided that both may be periodically checked for leak tightness and are checked at least annually.
2.2 Emergency Core Cooling System. Isolation requirements for ECCS are contained in SRP 6.3.
Isolation of ECCS to prevent overpres-surization must meet one of the following features:
(1) One or more check valves in series with a normally-closed motor-operated valve (MOV) which is to be opened upon receipt of a SIS when RCS pressure is less than the ECCS design pressure (2) Three check valves in series (3) Two check valves in series, provided that both may be periodically checked for leak tightness and are checked at least annually.
2.3 Other Systems. All other low pressure systems interfacing with the RCS must meet the following isolation requirements from BT? EICSB-3:
l (1). At least two valves in series must be provided to isolate the system when RCS pressure is above the system design pressure and valve position should be provided in the control room (2) For systems with two MOVs, each MOV should have independent and diverse interlocks to prevent opening until RCS pressure is below the system design pressure and should es mmarically close when e
RCS pressure increases above system design pressure (3) For systems with one check valve and a MOV, the MOV should be interlocked to prevent opening if RCS pressure is above system design pressure and should l
i automatically close whenever RCS pressure exceeds system design pressure.
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3.0 DISCUSSION AND EVALUATION There is one system at the LaCresse Nuclear Station which has a direct interface with the RCS and has a lower design pressure racing for all or part of the system than RCS design pressure. This system is the Core Spray (CS) system.
3.1 Core Spray System. The CS system consists of a high pressure subsystem and two low pressure systems for providing emergency core cooling. One low pressure subsystem provides water by gravity feed from an overhead tank to the high pressure core spray line, while the other subsystem provides water from diesel-driven pumps to the core through a separate line.
Isolation of the line from the overhead tank to the high pressure core spray line is provided by a check valve in series with an air-operated valve (A0V). A solenoid valve controls the air to the ADV which has position indicated in the control room. The A0V cannot be opened unless reactor pressurn is below system design pressure and two low level sensors indicate low reactos ester level. The ADV will auto-matically shut if either reacto' a*.e increases increases above system design pressure or eithet of the low level indications are cleared. The valve fails open upon loss of power to the solenoid valve.
Isolation of the line from the diesel-driven core spray pumps to i
che RCS is provided by two series check valves at the* point where the piping changes from a high pressure to a low pressure system. The check valves are not individually testable since there are no connec-tions between the valves to determine leakage from the inboard (closest to RCS) valve or to pressurize the outboard check valve.
t The CS system is in compliance with current licensing requirements for isolation of high and low pressure systems except for the non-testability of the two series check valves in the spray line from the diesel-driven core spray pumps required by SRP 6.3.
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4.0
SUMMARY
The Lacrosse Nuclear Station has one system with lower design pressure rating chan the RCS which is directly connected to the RCS.
The CS system meets current licensing criteria contained in SRP 6.3 for isolation of high and low pressure systems except for the testability requirements for the series checic valves in the line from the diesel-driven core spray pumps.
5.0 REFERENCES
1.
NUREG-075/087, Branch Technical Positions EICSB-3, RSB-5-1; Stan-dard Review Plan 6.3.
2.
LAC 3WR Drawings 410200-237, 41-300-080, and 41-503-770.
3.
Safeguards Report for Operating Authorization, Lacrosse Boiling Water Reactor.
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TECHNICAL E7ALUATION TOPIC VI-7.C.1 INDEPENDENCE OF REDUNDANT ONSITE POWER SYSTEMS LA CROSSE
1.0 INTRODUCTION
The objective of this review is to determine if the onsite elec-trical power systems (AC and DC) are in compliance with current lican-sing criteria 5or electrical independence between redundant standby (onsite) power sources and their distribution systems.
General Design Critorion 17 requires that the onsite electrical power supplies and their onsite distribution systems shall have suffi-cient independence to perform their safety function assuming a single failure. Regulatory Guide 1.6, " Independence Betwen Redundant Standk (Onsite) Power Sources and Between Their Distribution System," and IEEE Strudard 308-1974, "IEEE Standard Criteria for Nuclear Power Cen-erating StJtions" provide a basis acceptable to the NRC staff for meeting CDC L) i n regards to electrical independence of onsite power systems.
2.0 CRITERIA 2.1 AC Supplies. When operating frem standby sources, redundant load groups and redundant standby sources should be independent of each other at least to the fol' lowing extent.
(1) The standb-Jou-e of o, load group should not be automati 411y paa 'lete with the standby source of another load gre-und accident conditions 6
(2) No provisions should exist for automatically trans-l ferring one load group to another load group or 454[
loads between redundant power sources L:
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'(3) If means exist for manually connecting redundant load groups together, at least one interlock should be provided to prevent an operator error that would parallel their standby power sources.
2.2 DC Supplies. Each DC load group should be energized by a bettery and charger. The battery-charger combination should have no automatic connection to any other redundant DC load group.
3.0 DISCUSSION AhD EVAI.UATION 3.1 AC Supplies 3.1.1 Discussion. The La Crosse onsite AC power system consists of two redundant diesel generator (DG) supplied power trains.
During LOCA conditions DG 1A supplies the 480-V essential (ESS) bus IA, the " Turbine Building 430-V MCC 1A," and the " Turbine Building 120-V bus."
The " Turbine Building 120-V Regulated Bus" is supplied from the
" Turbine Building 120-V bus" through an isolation transformer. The DG 1B supplies the "480-V ESS bus 13," the " Reactor Building 480-V HCC 1A," and *.he " Diesel Building 480-V MCC."
There is no automatic paralleling of the two power sources or transfers of loads or load groups between the sources.
Essencial buses 1A and 1B may be manually connected by closing breaker 452 TBA on bus 1A and 452 TBB cn bus 23.
There are no interlocks to prevent paralleling the two diesel generators when utilizing this manual connection.
It is also possible ;o parallel the two DCs through the "120 V AC Non-Interruptab.? Bu?
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This bus is normally supplied power
Upon the failure of the inverter, a staric switch in the inverter autematically transfers the bus to a 480/l20 Volt transformer supplied from the " Diesel Building 480 V Motor 7ontrol Center" (DG IB). The "120 V AC Non-interruptable Bas 1B" can be 2
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simultaneously connected to the " Turbine Building 120 V Regulated Bus" (DG 1A) '.f a normally-open circuit breaker on the regulated bus is also closid. No interlocks are provided to prevent paralleling the two sources. However, as described above, the " Turbine Building 120-V Regulated Bus" is isolated from DG 1A through the use of an isolation transformer.
3.1.2, Evaluation. There are no automatic parall. ling of sources or automatic transfers of loads and/or load groups between the redun-dant sources. This complies with the single failure criteria. How-ever, the design allows the manual tying of essential 480-Volt bus LA to IB, and the connection of two trains through the "120 V AC Non-interruptable Bus IB" without interlocks to prevent paralleling of the redundant onsite sources. Therefore, the independence of the La Crosse onsite AC power trains do not comply with the recommendations of R.G. 1.6.
3.2 DC Supplies 3.2.1 Discussion. The La Crosse onsite DC systems consists of threg 125 V DC trains. The Diesel Building (D.B.) standby bus, the Reactor Plant (RX) bus, and the Generator Plant (Gen.) bus are all supplied from separate battery-charger sources. The charger for the RX bus is supplied from the DGlA through motor control center (MCC)lA; the Gen. bus is supplied from the offsite power through MCCID; and the D.B.
standby bus charger is supplied from DG 1B through the D.B. 480-V MCC.
There are no automatic connections between the three buses. The Gen. bus may be tied to the RX bus by manually closing the normally open bus tie circuit breaker which is located on the RX bus. The RX and D.B. standby buses may be connected by manually closing of normally open bus tie circuit breakers located on the D.B. standby and on the RX buses. The Technical Specifications, Section 4.2.3.2.3, require as a 4
" Limiting Condition for Operation" that the tie breakers be open during operating modes 1, 2, 3, 4, and 5.
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4 3.2.2 Evaluation. Each of the three DC load groups are energized by a battery and charger. There are no autanatic connections l
between battery-charger combinations and other redundant DC load groups. The DC systems comply with present licensing requirements.
4.0
SUMMARY
In two cases, as described in Section 3.1, the design of the La Crosse onsite AC power systems allow the manual connection of redun-dant load groups without interlocks to prevent paralleling of redundant AC. sources. This does not comply with the staff recommendations in-cluded in R.C. 1.6.
s The review of referenced information and drawings indicates that, with the above exceptions, the La Crosse onsite AC and DC redundant power sources and their distribution systems comply with the single failure criteria as outlined in R.C. 1.6.
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5.0 REFERENCES
1.
Letter DPC (Shimshak) to NRC (Shea) dated July 25, 1974.
2.
Letter DPC (Madgett) to NRC (Giambusso) dated January 21, 1975.
3.
Letter DPC.(Madgett) to NRC (Reid) dated December 23, 1975; Attached Sargent and Lundy Report No. SL-3232, " Description of Diesel Generator Building and Secondary Emergency On-Site Electri-cal Power System, LACBWR - Genoa Station Unit 2," dated December 12, 1975.
4.
Letter DPC (Madgett) to Zieman, LAC-3554, dated December 18, 1975; Attached Report No. NES 81A0019, " Single Failure Analysis of the La Crosse Boiling Water Reactor Emergency Core Cooling Systems,"
Revision 2, dated November 24, 1975.
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3 argent and Lundy drawings:
(a) #503678, " Wiring Diagram RX Plant 125 V DC MCC, Revision B.
dated March 2, 1976 (b) #503679, " Wiring Diagram Cen. Plant 125 V DC MCC, Revision 7, dated April 21, 1978 (c) #503627, " Single Line Diagram Unit 1 - Part 2," Revision D, dated November 11, 1977 (d) #503628. " Key Diagram Diesel Building 480 V ESS. Swgr Bus 1B," tevision E, dated November 9, 1977 (e) #503753/pg AP01, " Schematic Diagram Auxiliary Power System (AP)," Revision A, dated October 10, 1975. PCo, Big Rock Point, dated.May 26, 1976, page 9 (f) #503753/pg APG2, " Schematic Diagram Auxiliar? ?ower System 4
(AP)," Revision C, dae:4 May 7, 1976 (g) #503837, " Wiring Diagrac D.B. 125 V DC Non-Iter. Power Supply MCC," Revision D, dated February 28. 1978 (h) #503629, 'PKey Diagram Dies. Bldg. 480 V AC MCC and 125 V DC Disc. Bus," Revision H, dated February 23, 1978.
s 6.
General Design Criterion 17, "Elsetrical Power System," of Appen-dix A, " General Design Criteria of Nuclear Power Plants," to 10 CFR Part 50, " Domestic Licensing of Production and Utilization 1
Facilities."
7.
" Independence Between Redundant Standby (Onsite) Power Sources and Between Their Distribution Systems," Regulatory Guide 1.6.
"IEEE Standard Criteria for Nuclear Power Generating Stations,"
8.
IEEE Standard 308-1974, The Institute of Electrical and Elsr.ronic Engineers, Inc.
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f SEP TECHNICAL EVALUATION TOPIC VIII-3.B DC POWEt SYSTEM BUS VOLTAGE MONITORING AND ANNUNCIATION LACROSSE
1.0 INTRODUCTION
The objective of this review is to determine if the DC power sys-tem bus voltage monitoring and annunciation are in compliance with current licensing criteria for Class IE DC power systems.
The specific requirements for DC power system monitoring derive from the general requirements embodied in Sections 5.3.2(4), 5.3.3(5),
1 and 5.3.4(5) of IEEE Standard 308-1974, and in Regulatory Guide 1.47.
In summary, these general requirements simply state that the DC system (batteries, distribution systems, and chargers) shall be monitored to the extent that it is shown to be ready to perform its intended function.
2.0 CRITERIA As a minimum, the following indications and alarms of the Class
- .E DC power system (s) status shall oe provided in the control rooms Battery current (anunater-charge / discharge) e 2-ttery charger output current (ansnecer) e e C; bus voltage (volt:neter)
Battery charger output voltage (voltmeter) e e Battery high discharge race alarm o DC ous undervoltage and overvoltage alarm DC tune ground alarm (for ungrounded ryotes) o Battery breaker (s) or fuse (s) open alarm e
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4 Battery charger output breaker (s) or fuse (s) open e
alarm Battery charger trouble alarm (one alarm for a number e
of aonormal conditions whien are usually indicated Ic. ally).
3.0 DISCUSSION AND EVALUATION 3.1 Discussion. Ihree 125 V oacteries, three battery chargers, and three DC buses comprise the Lacrosse Class IE DC power systems.
Control room indication consists of bus voltage (one bus only), bus undervoltage and ground alarms (each bus), and charger trouble :larms (each charger).' Local indication consists of bus current ammeters, charger output ammeters, bus voltmeters, and charger output volcmeters.
3.2 Evaluation. The Lacrosse control room has no indication of bettery current, charger output current, bus voltage (two buses),
charger output voltage, battery high discharge rate, bus overvoltage, battery breaker / fuse status, or charger output bresker/ fuse status.
Therefore, the Lacrosse DC power systems monitoring is not in compli-ance with current licensing criteria.
4.0
SUMMARY
Of 11 parameters currently required to be indicated or alarmed in the control room, only four are indicated or alarmed for one bus and only three for the other two buses in the Lacrosse control room. There-fore, the Lacrosse DC power systems are not monitored in compliance with current licensing criteria.
5.0 REFRENCES 1.
- DuuL. Standard 308-1974, " Standard Criteria for Class II Power Systems for Nuclear Power Generating Stations."
2.
Regulatory Guide 1.74, "3ypassed and Inoperable Status Indi-cation for Nuclear Power Plant Safety Systems."
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3.
NRC Memorandum, PSB (Rosa) to SEPB (Crutchfield), "DC System Monitoring and Annunciation," dated October 16, 1979, 4.
Letter, Dairyland Power Cooperative (Linder) to NRA (Ziemana),
"SEP Topic VIII-3.3, DC Pwer System Bus Volcage Monitoring and Annunciacion," dated July 11, 1979 i
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4.0 S"MMARY Yhe Ya'nkee Rowe accumulator isolation valve power and control sys:em 4
design does no: seet current licensing cri:eria, in tha: con::al roem valve posi:icn indica:ica is nei:her redundan: nor single-failure free.
5.0 RITIRINCES 1.
IIII standard 279, " Criteria for P : cec:icn Syste=s for Nuclea Power Generacing Stations."
2.
3:anch Technical Posi:ica ICS3 L, "Requirenents of Mo:or-Operated Valves in the ICCS Accu =ula:or Lines.'"
3.
Branch Technical Posicion IOS3 18, "Applica:icn of :he Single Failure Criterica :o Manually-Con:::lled Ilec::ically-Opera:ed Valves."
4 Yankee A::mic Ilec: i: Company Drawing 9699-FM-83A, No Revisien, da:ed Nove=ber 10, 1977.
5.
" Yankee Rowe Nuclear Power 5:ation, Techni:al Specifica:i:ns,"
paragraph 4.5.2.b.
6.
Yankee A:cmic Ilectric Coc;nay Drawing 9599-ISK-6AD, Revision 9, dated March 31, 1979.
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