NRC Generic Letter 1981-12
| ML031080537 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley, Millstone, Hatch, Monticello, Calvert Cliffs, Dresden, Davis Besse, Peach Bottom, Browns Ferry, Salem, Oconee, Nine Mile Point, Palisades, Indian Point, Kewaunee, Saint Lucie, Point Beach, Oyster Creek, Cooper, Pilgrim, Arkansas Nuclear, Prairie Island, Brunswick, Surry, North Anna, Turkey Point, Vermont Yankee, Crystal River, Haddam Neck, Ginna, Duane Arnold, Farley, Robinson, San Onofre, Cook, Yankee Rowe, Maine Yankee, Quad Cities, Humboldt Bay, La Crosse, Big Rock Point, Rancho Seco, Zion, Fort Calhoun, FitzPatrick, 05000000, Fort Saint Vrain, Trojan, Crane |
| Issue date: | 02/20/1981 |
| From: | Eisenhut D Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| GL-81-012, NUDOCS 8103180264 | |
| Download: ML031080537 (13) | |
6e.te aL ARCS
I II ALL POWER REACTOR LICENSEES
Docket No. 50-348 Docket No. 50-3 Farley Unit 1 Indian Point Unit 1 Docket No. 50-313 Docket No. 50-247 Arkansas Unit 1 Indian Point Unit 2 Docket No. 50-368 Docket 50-286 Arkansas Unit 2 Indian Point Unit 3 Docket No. 50-317 Docket No. 50-155 Calvert Cliffs Unit 1 Big Rock Point Docket No. 50-318 Docket No. 50-255 Calvert Cliffs Unit 2 Palisades Docket No. 50-293 Docket No. 50-409 Pilgrim Unit 1 Lacrosse Docket No. 50-325 Docket No. 50-269 Brunswick Unit 1 Oconee Unit 1 Docket No. 50-324 Docket No. 50-270
Brunswick Unit 2 Oconee Unit 2 Docket No. 50-261 Docket No. 50-287 H. B. Robinson Unit 2 Oconee Unit 3 Docket No. 50-10 Docket No. 50-334 Dresden Unit 1 Beaver Valley Unit 1 Docket No. 50-237 Docket No. 50-302 Dresden Unit 2 Crystal River 3 Docket No. 50-249 Docket No. 50-335 Dresden Unit 3 St. Lucie Unit 1 Docket No. 50-254 Docket No. 50-250
Quad-Cities Unit 1 Turkey Point Unit 3 Docket No. 50-265 Docket No. 50-251 Quad-Cities Unit 2 Turkey Point Unit 4 Docket No. 50-295 Docket No. 50-321 Zion Unit 1 Edwin I. Hatch Unit 1 Docket No. 50-304 Docket No. 50-366 Edwin I. Hatch Unit 2 I
Zion Unit 2 Docket No. 50-213 Docket No. 50-315 IAe'
Connecticut Yankee (Haddam Neck) D. C. Cook Unit 1
810818 O PC 4 r
IN
I1 .
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Docket No. 50-344 Docket No. 50-305 Docket No. 50-316 Kewaunee D. C. Cook Unit 2 Trojan Docket No. 50-333 Docket No. 50-29 Docket No. 50-331 Yankee-Rowe Duane Arnold FitzPatrick Docket No. 50-267 Docket No. 50-339 Docket No. 50-219 Ndrth Anna 2 Oyster Creek Unit 1 Ft. St. Vrain Docket No. 50-272 Docket No. 50-311 Docket No. 50-309 Salem 2 Maine Yankee Salem Unit 1 Docket No. 50-289 Docket No. 50-244 Three Mile Island Unit 1 R. E. Ginna 1 Docket No. 50-320 Docket No. 50-312 Three Mile Island Unit 2 Rancho Seco Docket No. 50-298 Docket No. 50-206 Cooper Station San Onofre 1 Docket No. 50-220 Docket No. 50-259 Nine Mile Point Unit 1 Browns Ferry Unit 1 Docket No. 50-245 Docket No. 50-260
Millstone Unit 1 Browns Ferry Unit 2 Docket No. 50-336 Docket No. 50-296 Millstone Unit 2 Browns Ferry Unit 3 Docket No. 50-263 Docket No. 50-346 Monticello Davis-Besse 1 Docket No. 50-282 Docket No. 50-271 Prairie Island Unit 1 Vermont Yankee Docket No. 50-306 Docket No. 50-338 Prairie Island Unit 2 North Anna 1 Docket No. 50-285 Docket No. 50-280
Ft. Calhoun Surry Unit 1 Docket No. 50-133 Docket No. 50-281 Humboldt Bay Surry Unit 2 Docket No. 50-277 Docket No. 50-266 Peach Bottom 2 Point Beach Unit 1 Docket No. 50-278 Docket No. 50-301 Peach Bottom 3 Point Beach Unit 2
UNITED STATES
NUCLEAR REGULATORY COMMISSION
WASHINGTON, D. C. 20555 February 20, 1981 TO ALL POWER REACTOR LICENSEES WITH PLANTS
LICENSED PRIOR TO JANUARY 1, 1979 SUBJECT: FIRE PROTECTION RULE (45 FR 76602, NOVEMBER 19, 1980) -
Generic Letter 81-12 Paragraph 50.48(b) of 10 CFR Part 50, which became effective on February 17,
1981, requires all nuclear plants licensed to operate prior to January 1, 1979 to meet the requirements of Sections III.G, III.J and III.0 of Appendix R to
10 CFR Part 50 regardless of any previous approvals by the Nuclear Regulatory Commission (NRC) for alternative design features for those items. This would require each licensee to reassess all those areas of the plant "... where cables or equipment, including associated non-safety circuits, that could prevent operation or cause maloperation due to hot shorts, open circuits or shorts to ground or (sic) redundant trains of systems necessary to achieve and maintain hot shutdown conditions are located within the same fire area outside of primary containment ... "* to determine whether the requirements of Section III.G.2 of Appendix R are satisfied. If not, the licensee must provide alternative shutdown capability in conformance with Section III.G.3 or request an exemption if there is some justifiable basis.
Paragraph 50.48(c)(5) requires that any modifications that the licensee plans in order to meet the requirements of Section III.G.3 of Appendix R must be reviewed and approved by the NRC. This paragraph also requires that the plans, schedules and design descriptions of such modifications must be submitted by March 19, 1981. To expedite our review process and reduce the number of requests for additional information with regard to this review, we are enclos- ing two documents which specify the information that we will require to complete our reviews of alternative safe shutdown capability. Enclosure 1 is *Staff Position Safe Shutdown Capability". This document was originally sent to you in late 1979. Section 8 specifies the information required for staff review. If you have already submitted any of the information required, you need only reference that previous submittal. Enclosure 2 indicates the additional information needed to ensure that associated circuits for alter- native safe shutdown equipment is included in your reassessment and in our review. If you made no modifications that were required to provide alternative safe shutdown capability and if your reassessment concludes that alternative safe shutdown capability in accordance with the provisions of Section III.G.3 is not necessary, you do not have to provide the inf.ormation requested by these Enclosures.
- Quoted from Section III.G.2 of Appendix R to 10 CFR Part 50. Note that the 'or" preceding "redundant trains" is a typographical error and should read 'of redundant trains".
2- Fin4lly, we request that as part of your submittal of plans and schedules for meeting the provisions of Paragraphs (c)(2), (c)(3) and (c)(4) of 10 CFR 50.48 as required by Paragraph 50.48(c)(5), you Include the results of your reassessment of the design features at your plant for meeting the require- ments of Sections III.G, III.J and III.0 of Appendix R to 10 CFR Part 50.
This detailed information need not accompany the design description that must be submitted by March 19. 1981. However, we request that it be submitted as soon as possible, but no later than May 19, 1981.
This request for information was approved by GAO under a blanket clearance number R0071 which expires September 30, 1981. Comments on burden and dupli- catign may be directed to the U. S. General Accounting Office, Regulatory Repqrts Review, Room 5106, 441 G Street, N. W., Washington, 0. C, 20548.
Sincerely, rr G. eenoutr Division of icensing Office of Nuclear Reactor Regulation Enclosures:
1. Staff Position
2. Request for Additional Information cc w/enclosures:
See next page
STAFF POSITION Enclosure 1 SAFE SHUTDOWN CAPABILITY
Staff Concern During the staff's evaluation of fire protection programs at operating plants, one or more specific plant areas may be identified in which the staff does not have adequate assurance that a postulated systems.
fire will not damage both redundant divisions of shutdown resulted This lack of assurance in safe shutdown capability has from one or both of the following situaticns:
- Case A: The licensee has not adequately identifiedshutdown the systems and components required for safe and their location in specific fire areas.
- Case E: The licensee has not demonstrated that the fire protection for specific plant areas will prevent damage to both redundant divisions of safe shutdown components identified in these areas.
For Case A, the staff has required that an adequate safe shutdown analysis be performed. This evaluation includes the Identification of the systems required for safe shutdown and determinedthe location of the by this system components in the plant. Where it ofis both redundant divisions evaluation that safe shutdown components to demonstrate are located in the same fire area, the licensee is required or provide alternate that a postulated fire will not damage both divisions shutdown capability as in Case B.
shutdown For Case B, the staff may have required thatof anthealternatearea of concern capability be provided with is independent such a capability in lieu of or the licensee may have proposed in the area. The certain additional fire protection modifications concern along with
- pecific modifications associated with the area of the area form the other systems apd equipment already independent of the modifications needed and alternate shutdown capability. For each plant, shutdown functions may be the combinations of systems which providethetheshutdown functions provided unique for each critical area; however, of the limiting should maintain plant parameters within thethebounds design basis event.
safety consequences deemed acceptable for Staff Position Safe shutdown capability should be demonstrated (Case A) or alternate shutdown capability provided (Case B) in accordance with the guidelines provided below:
1. Desicn Basis Event
7he design basis event for considering the need for alternate shu.down is a postulated fire in a s;ecific fire area containing where redundant safe shutdown cables/equipment in means proximity cfose cannot assure
1: nas been deermirned that fire ProtectiCn 7wo cases shculd
- .2a: safe shutcwn capability will be preserved. and (2) offsite be considered: (1) offsite power is available;
power Is not available.
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2. Limiting Safety Consequences arid Required Shutdown Functions
2.1 No fission product boundary integrity shall be affected:
a. No fuel clad damage;
b. No rupture of any primary coolant boundary;
c. No rupture of the containment boundary.
2.2 The reactor coolant system process variables shall be within those predicted for a loss of normal ac power.
2.3 The alternate shutdown capability shall be able to achieve and maintain subcritical conditions in the reactor, maintain reactor coolant inventory, achieve and mairtain hot standby* conditions (hot shutdown' for a BAR) for an extended period of time, achieve cold shutdown* conditions within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and maintain cold shutdown conditions thereafter.
- As defined in the Standard Technical Specifications.
3. Performance Goals
3.1 The reactivity control function shall be capable of achieving and maintaining cold shutdown reactivity conditions.
3.2 The reactor coolant makeup function shall be capable of
- maintaining the reactor coplant level above the top of the core for BWR's and in the pressurizer for PWR's.
3.3 The reactor heat removal function shall be capable of achieving and maintaining decay heat removal.
3.4 The process monitoring function shall be capable of providing direct readings of the process variables necessary to perform and control the above functions.
3.5 The supporting function shall be capable of providing the process cooling, lubrication, etc. necessary to permit the operation of the equipment used for safe shutdown by the systems identified in 3.1 - 3.4.
3.6 The equipment and systems used to achieve and maintain hot standby conditions (hot shutdown for a BWR) should be
(1) free of fire damage; (2) capable of maintaining such conditions for an extended tine period longer than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if the equipment recuired to achieve and maintain cold shutdown is not available due to fire damage; and (3) capable of being powered by an onsite emergency power system.
3.7 The equipment and systems used to achieve and maintain cold shutdown conditions should be either free of fire damage or the fire damage to such systems should be limited such tha: repairs can be'nade and cold shutdown conditions 3cnleved within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Equipment and systems used prior to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after the fire should be capable of being powered by an onsite emergency power system; those used after 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> may be powered by
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Enclosure 2 REQUEST FOR ADDITIONAL INFORMATION
1.Section III.G of Appendix R to 10 CFR Part 50 requires cabling for or associated with redundant safe shutdown systems necessary to achieve and maintain hot shutdown conditions be separated by fire barriers having a three-hour fire rating or equivalent protection ( see Section III.G.2 of Appendix R). Therefore, if option III.G.3 is chosen for the protection of shutdown capability cabling required for or associated with the alternative method of hot shutdown for each fire area, must be physically spearated by the equivalent of a three-hour rated fire barrier from the fire area.
In evaluating alternative shutdown methods, associated circuits are circuits that could prevent operation or cause maloperation of the alternative train which is used to achieve and maintain hot shutdown condition due to fire induced hot shorts, open circuits or shorts to ground.
Safety related and non-safety related cables that are associated with the equipment and cables of the alternative, or dedicated method of shutdown are those that have a separation from the fire area less than that required by Section III.G.2 of Appendix R to 10 CFR 50 and have either (1) a common power source with the alternate shutdown equipment and the power source is not electrically protected from the post-fire shutdown circuit of concern by coordinated circuit breakers, fuses or similar devices, (2) a connection to circuits of equipment whose spurious operation will adversely affect the shutdown capability, e.g., RHR/RCS Isolation Valves, or (3) a common enclosure, e.g., raceway, panel, junction box, with alternative shutdown cables and are not electrically protected from the post-fire shutdown circuits of concern by circuit breakers, fuses or similar devices.
For each fire area where an alternative or dedicated shutdown method, in accordance with Section III.G.3 of Appendix R to 10 CFR Part 50, is provided by proposed modifications, the following information is required to demonstrate that associated circuits will not prevent operation or cause maloperation of the alternative or dedicated shutdown method:
A. Provide a table that lists all equipment including instrumentation and support system equipment that are required by the alternative or dedicated method of achieving and maintaining hot shutdown.
B. For each alternative shutdown equipment listed in l.A above, provide a table that lists the essential cables (instrumentation, control and power) that are located in the fire area.
c. Provide a table that lists safety related and non-safety related cables associated with the equipment and cables constituting the alternative or dedicated method of shutdown that are located in the fire area.
D.: Show that fire-induced failures of the cables listed in B and C above will not prevent operation or cause maloperation of the alternative or dedicated shutdown method.
E. For each cable listed in L.B above, provide detailed electrical schematic drawings that show how each cable is isolated from the fire area.
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2. The residual heat removal system is generally a low pressure system that interfaces with the high pressure primary coolant system. To preclude a LOCA through this interface, we require compliance with the recommenda- tions of Branch Technical Position RSB 5-1. Thus, this interface most likely consists of two redundant and independent motor operated valves.
These two motor operated valves and their associated cable may be subject to a single fire hazard. It is our concern that this single fire could cause the two valves to open resulting in a fire-initiated LOCA through the subject high-low pressure system interface. To assure that this interface and other high-low pressure interfaces are adequately pro- tected from the effects of a single fire, we require the following information:
A. Identify each high-low pressure interface that uses redundant electrically controlled devices (such as two series motor operated valves) to isolate or preclude rupture of any primary coolant boundary.
B. Identify the device's essential cabling (power and control) and describe the cable routing (by fire area) from source to termi nation.
C. Identify each location where the identified cables are separated by less than a wall having a three-hour fire rating from cables for the redundant device.
D. For the areas identified in item 2.C above (if any), provide the bases and justification as to the acceptability of the existing design or any proposed modifications.
-3- offsite power.
I
3.8 These systems need not be designed to (1) seismic category with criteria; (2) single failure criteria; or (3) cope valves other plant accidents such as pipe breaks or stuckthese
- Appendix A BTP 9.5-1), except those portions ofsafety systems.
systems which interface with or impact existing
4. PWR Eouipment Generally Necessary For Hot Standby
(1) Reactivity Control e.g.,
Reactor trip capability (scram). Boration capability pump charging pump, makeup pump or high pressure injection taking suction from concentrated borated dater supplies, and letdown system if required.
(2) Reactor Coolant Makeup Reactor coolant makeup capability, e.g., charging pumpsrelief or the high pressure injection pumps. Power operated of the valves may be required to reduce pressure to allow use high pressure injection pumps.
(3) Reactor Coolant System Pressure Control pumps Reactor pressure control capability, e.g., charging or pressurizer heaters and use of the letdown systems if required.
(4) Decay Heat Removal relief Decay heat removal capability, e.g., power operated for heat valves .(steam generator) or safety relief valves removal with a water supply and emergency or auxiliary Service feedwater pumps for makeup to the steam generator. water for auxiliary water or other pumps may be required to provide is feed pump suction if the condensate storage tank capacity not adequate for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
(5) Process Monitorinq Instrumentation and Process monitoring capability e.g., pressurizer pressure level, steam generator level.
(6) SuDoort.
above The equipment required to suppcrt operation of the water described shutdown equipment e.g. , cD=,,,cnent cooling ensize :ower sources (AC, DC) with service water, et:. and their associated eiectriCal 'distr vuti fn system.
t.
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5. PWR Eouioment Generally Necessary For Cold Shutdown*
(1) Reactor Coolant System Pressure Reduction to Residual Heat Removal System (RHR) Cabaoi ity Reactof coolant system pressure reduction by cooldown using steam generator power operated relief valves or atmospheric dump valves.
(2) Decay Heat Removal Decay heat removal capability e.g., residual heat removal system, component cooling water system and service water system to removal heat and maintain cold shutdown.
(3) Support Support capability e.g., onsite power sources (AC & DC)
or offsite after 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and the associated electrical distribution system to supply the above equipment.
Equipment necessary in addition to that already provided to maintain hot standby.
6. BWR Equipment Generally Necessary For Hot Shutdown
(1) Reactivity Control Reactor trip capability (scram).
(2) Reactor Coolant Makeuo Reactor coolant inventory makeup capability e.g., reactor core isolation cooling system (RCIC) or the high pressure coolant injection system (HPCI).
(3) Reactor Pressure Control and Decay Heat Removal Depressurization system valves or safety relief valves for dump to the suppression pool. The residual heat removal system in steam condensing mode, and service water system may also be used for heat removal to the ultimate heat sink.
(4) Suopression Pool Cooling Residual heat removal system (in suppression pool cooling mode) service water system to maintain hot shutdown.
(5) Process Monitorina Process monitoring capability e.g., reactor vessel level and pressure and suppression pool temoerature.
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(6) Support source (AC & DC) and Support capability e.g., onsite power to provide for the their associated distribution systems shutdown equipment.
For Cold Shutdown'
7. BWR Eouipment Generally Necessary for hot shutdown has reduced At this point the equipment necessary to where the RHR
the primary system pressure and temperature RIR cooling mode.
system may be placed in service in
(1) Decay Heat Removal cooling mode, service Residual heat removal system in the RXR
water system.
(2) Supoort
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Onsite sources (AC & DC) or offtite after to provide and their associated distribution systems for shutdown equipment.
for achieving hot shutdown.
- Equipment provided in addition to that
8. Information Recuired For Staff Review thereof used to (a) Description of the systems or portions required provide the shutdown capability and modifications if required.
to achieve the alternate shutdown capability normal and alternate (b) System design by drawings which show location of components, and shutdown control and power circuits, the wiring which is out that wiring which is in the area and system.
of the area that required the alternate will not (c) Demonstrate that changes to safety systems isolation switches degrade safety systems. (e.g., new criteria and and control. switches should meet design in the system standards in FSAR for electrical equipment that the that the switch is to be installed; cabinets also meet the same switches are to be mounted in should cabinets and criteria (FSAR) as other safety related from the control panels; to avoid inadvertent isolation be keylocked, or alarmed room, the isolation switches should or "isolated" position;
in the control room if in the "local'
To verify switch is in the periodic checks should be made and a single transfer proDer position for normal operation;
be a source for a switch or other new device should not safety systems).
single failure to cause 1css of redundant
'sourCesfor the (d) Demonstrate that wiring, including Dowerfor the alternate control circuit and equi;ment operationequipment wiring in shutdown metnod, is independent of the area to be avoided.
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sources, including (e) Demonstrate that alternate shutdown power circuits control all breakers, have isolation devices beon avoided, even if the that are routed through tho area to breaker is to be operated manually.
been developed (f) Demonstrate that licensee procedure(s) have to effect the shutdown which describe the tasks to be performed be submitted.
method. A summary of these procedures should
('g) Demonstrate that spare fuses are available for control in supplying circuits where these fuses may be required power to control circuits used for the shutdown :able spreading method and may be blown by the effects of a convenient room fire. The spare fuses should be locatnd should to the existing fuses. The shutdown procedure inform the operator to check these fuses.
to perform the (h) Demonstrate that the manpower required of (f) as well shutdown functions using the procedures the fire is as to provide fire brigade members to fighttechnical available as required by the fire brigade specifications.
tests are performed.
(1) Demonstrate that adequate acceptance operates from the These should verify that: equipment local control station when the transferthat or isolation switch and the equipment is placed in the "local" position that equip- room; and cannot be operated from the control cannot be operated ment operates from the control room but or isolation at the local control station when the transfer switch is in the "remote" position.
requirements Ci) Technical Specifications of the surveillance that equipment and limiting conditions for operation for For example, not already covered by existing Tech. Specs. added to a service if new isolation and control switches are surveillance require- water system, the existing Tech. Spec. a statement ments on the service water system should add similar to the following:
the pump
'Every third pump test should also verify that moving station after starts from the alternate shutdown to the local all service water system isolation switches control position."
adequate to perform (k) Demonstrate that the systems available are functions required t.he necessary shutdown functions. The possible (e.g..
should be based on previous analyses, if power or shutdown in the FSAR), such as a loss of normal a.c. required for the on a Group I isolation (BWR). The equipment alternate capability should be the same or equivalent to that relied on in the above analysis.
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(1) Demonstrate that repair procedures for cold shutdown systems are developed and material for repairs is maintained on site.