ML19350C221
| ML19350C221 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 03/23/1981 |
| From: | Clark R Office of Nuclear Reactor Regulation |
| To: | Mayer L, Mayer L NORTHERN STATES POWER CO. |
| References | |
| TAC-11094, TAC-11095, NUDOCS 8103310616 | |
| Download: ML19350C221 (2) | |
Text
p neag%
<< [G j
]L l
4 UNITED STATES y 7, s-g NUCLEAR REGULATORY COMMISSION
- n e ;; e c.Asectos, c. c. csss e
. 5. m March 23, 1931 Docket !ios. 50-282 50-305
!!r. L. O. Payer, Panager Nuclear Support Services
?:crthern States Fewer Ccepany 414 Nicollet Fall - Sth Flccr Minneapolis, Minnesota 55401 Cear Mr. P.ayer:
In the Prairie Island Fire Prntecticn Safety Evaluatien Report we identified a nu-ber of items which could et be evaluated until additional information was provided. By letter dated October 22, 1979, the Northern States Fewer Company provided the results of an analysis of the ca: ability of the plant to achieve and raintain safe shutdown conditions given the existence of fires in various areas of the plant. The 3rcokhaven Naticnal Laboratory (BNL) pursuant to a technical assistance contract with the NRC has reviewed the SSP analysis. The SNL report is included as the Encicsure to this letter. We agree with BNL's conclusions that the proposed redifications do not reet our criteria. As you knew the Prairie Island plant will be
- required to meet Sectiens IIIG and L of Appendix R to 10 CFR Part 50 in this regard.
L'e understand from conversaticns with your staff that the requirements of Appendix R will require the preparation of a new re;crt on the plant's safa shutdcwn capabilities. Accordingly, since the cuestions in the Enclosure are based on the previous NSP report of October? 22, 1979, an explicit item-by-ite 1 response to these questions is not required by the staff. These questions should be considered in the developcent of the infcrmation to be contained in the new report. Our review of the new re:crt will include consideraticn of whether the issues represented by these questions are acceptably addressed.
Sincerely, mjg
.;, kf s$l Robert A. Clark, Chief U
l L-
,,".,o 7 193 W r Operating Reactors Branch !3
-i
}M o 5-
,y Division of Licensing e#
N Re rt by. g.
p(
- E' nclosure:
cc w/ encl:
See next page 81033106@
/
u
Northern States Power Corpany cc:
Gerald Charnoff, Esquire Bernard M. Cranum Shaw, Pittman, Potts and Trowbridge Bureau of Indian Affairs, DOI 1800 M Street, N.W.
831 Second Avenue South Washington, D. C.
20036 Minneapolis, Minnesota 55402 Ms. Terry Hoffman Director, Criteria and Standards Division Executive Director Office cf Radiation Programs (ANR 460)
Minnesota Pollution Control Agency U.S. Environmental Protection Agency 1935 W. County Road B2 Washington, D. C.
20460 Roseville, Minnesota 55113 U. S. Environmental Protection Agency The Environmental Conservation Library Federal Activities Branch Minneapolis Public Library Region V Office 300 Nicollet Mall ATTN:
EIS COORDINATOR Minneapolis, Minnesota 55401 230 South Dearborn Street Chicago, Illinois 60604 Mr. F. P. Tierney, Plant Manager Prairie Island Nuclear Generating Plant Northern States Power Company Route 2 Welch, Minnesota 55089 Chairman, Public Service Commission Joclyn F. Olson, Esquire of Wisconsin Special Assistant Attorney General Hill Farns State Office Building Minnesota Pollution Control Agency Madison Wisconsin 53702 1935 W. County Road B2 Roseville, Minneosta 55113 Robert L. Nybo, Jr., Chairman Minnesota-Wisconsin Boundary Area Corrd ssion
- 619 Second Street.
~
Hudson, Wisconsin 54016 U.S. Nuclear Regulatcry Commission Resident Inspectors.0ffice Route #2, Box 500A Welch, Minnesota 55089 Mr. John C. Davidson, Chairman Goodhue County Board of Commissioners 321 West Third Street Red Wing, Minn'esota 55066 e
h:
BROOKHAVEN NATIONAL LABC'RATORY
{ {} }
ASSOCIATED UNIVERSITIES, INC.
Upton. New York 11973 Department of Nuc!ect Energy (516) 345-2144 February 19, 1981 Mr. Victor Benaroya, Chief Chemical Engineering Branch U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Vic:
I am enclosing our final report covering the alternate shutdown capability for the Prairie Island Nuclear Power Station.
The report is an evaluation of the licensee's sub.mittal of October 22, 1979 on safe shutdown capabilities but, due to scheduling difficulties, does not have the benefit of the requested telephone communication with the licensee.
Respectfully yours, f
3j g,-
/,obert E. Hall, Group Leader Risk Assessment & Engineering Analysis REH:ROS:sd enclosure cc.:
R. Cerbone wo/att.
.R. Ferguson W. Kato wo/att.
V. Lettieri E. MacDougall P. Sears R.' Smith
~
INTERIM REPORT POST FIRE SHUTDOW CAPABILITY PRAIRIE ISLAND NUCLEAR GENERATING PLANT Section 3.2.1 of the SER, Safe Shutdown Systems states that the safe shutdown capability requirements will be re-evaluated.
Section 4.1 states that the licensee is continuing his analysis to determine whether safe shut-dcwn can be achieved in each plant area where the fire is postulated.
The licensee has addressed this requirement in his report of October 22, 1979 entitled, " Revised Fire Protection Safe Snutdcwn Analysis" fcrwarded with his letter of October 22,1979, " Fire Protection Safe Shutdown Re-analysis."
The critical areas of concern as identified by the licensee are the cable spreading and relay roans and the two auxiliary feed pump rooms, each of which contains a hot shutdown panel. The Prairie Island program for alter-nate shutdown capability relies on existing system equipment with manual realignments to achieve hot and cold shutdown. TS alternate shutdown capa-bility also relies heavily on existing or proposed fire protection methods which.are, in turn, dependant on a design basis (postulated) fire established in the Fluor-Pioneer Fire Hazards analysis of March 1977, Supplement, July 1977.
The following equipment is required to place and maintain Unit No.1 in a sub-critical hot shutdown condition.
Steam Safety Relief Yalves, Diesel Generators, Batteries Auxiliary Feedwater Pumps and Associated Valves Station Air Compressors Cooling Water Pumps Steam Generator Driven Operated Relief Valves Boration including Heat Trace Circuits Steam Generator Level, Reactor Coolant Pressure and Temperature and Pressurizer Level Indication.
l The following equipment is required to bring Unit No. I to a safe cold shutdown condition within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> using either onsite or offsite power l
sources. This equipment is in addition to that identified above for hot shut-l down.
Residual Heat Removal Systems Component Cooling System Component Cooling Heat Exchanger Supply Valves Reactor Coolant Temperature, Pressurizer Level Indication Seal Leak Off Isolation Valves We have evaluated the Prairie Island post fire shutdown capability using l
NRC guidelines'" Staff Position, Safe Shutdown Capability" dated June 19, 1979 and NRC requirements in Section III.L of Appendix R to 10 CFR 50. We have found that:
L-
1.
The licensee's procedures for hot shutdown are predicated on repairs in several instances. These are the disconnection of specific cables to disable the diesel safety trip function in the event of a relay and cable room fire, re-directing of air around four solenoid valves and providing a temporary jumper in cabinet 1134 for power to solenoid SV-33236, part of the boration flow path.
2.
The licensee's procedures for cold shutdown are also predicated on manual operations and repairs such as providing jumpers to operate the heat exchanger by-pass valve and to operate the component cool-ing pumps in the event of a fire in areas 18 and 18A. Operation of the seal leak off valves is dependent on electrical jumpers in cabinet 1138 or redirectng air around the solenoid. The licensee also specifies manual operation of valves associated with the re-sidual heat removal pumps and the conponent cooling heat exchanger supply valves.
3.
The licensee has not. addressed the alternate shutdown capability in each plant area where the fire is postulated.
4.
Critical fire areas described in the SER have not been satisfactor-ily addressed. The licensee states that a fire in the control room would have no effect on the operation of either diesel generator although controls for each are located here.
This statement is -
based on the Fluor analysis which classifies any fire in this area as trivial. Similar logic is applied to fire area 32, Auxiliary Feed Pump Room, through which cabies for both trains pass.
5.
The post fire support functions have not been shown to be capable of providing the lubrication necessary to permit the operation of equipment used for' shutdown functions.
6.
The post fire shutdown capability has not been shown to be isolated from associated circuits so that fire damage to associated circuits in a fire area may prevent the operation of shutdown equipment.
7.
The post fire process monitoring function has not been shown to be capable _ of. providing direct readings of the process variables neces-
~
sary to perform and control shutdown functions. The licensee has not-stated that pressurizer pressure and temperature indication or pump flow and radiation level would be provided.
8.
The licensee has not as yet demonstrated that repair procedures for cold shutdown systems are fully developed and material for repairs is maintained on -site.
- 9. -The post fire shutdown capability depends on fire protection measures in the following areas:
Fire Area 13 - Control 'Aso, Fire Area 18 and 18A - Rc'ay and Cable Room Fire Areas.31 and J' A.. :liary Feedwater Pump Rooms which each contain-a hot M u g hnel,
Fire Areas 58 and 69 - Aux;11ary Building Ground and Mezzanine Floors, _
Fire Area 60 - Auxiliary _ Building Operating Level.
r
These protection features should meet the NRC re,
nts of Sec-tion III.G of Appendix R to 10 CFR 50. The concerns and areas.
SER Section 3.2.1(1) should also be addressed and evaluated in regard to thc equirements of Sectioa HI.G.
We conclude that the proposed alternative shutdown capability for the Prairie Island Nuclear Power Station does not confom with MRC guidelines and requirements and, therefore, is unacceptable.
We recon 'end the following:
A.
The alternative shutdown capability should be S.odified to meet the requirements of Section III L of Aptendix R to 10 CFR Part 50, tak-ing into consideration the above findings.
B.
The supporting functions should be capable of providing the process lubrication necessary to pemit the operation of the Muipment used for safe shutdown by the systems identified as part c the alternate shutdown capability. All of the support functions sF Id be avail-able for the equipment used in the alternative shutd:
capability.
C.
The process monitoring should be shcwn to be capab' providing direct readings of the process variables necessary 2ntrol re-activity, reactor coolant makeup, and reactor heat tal. Perma-nently installed instruments should be used to pro-
.apability for reading pressurizer pressure, tenperature and ~
. reactor coolant loop temperature, stean generator level an
- .sure,
auxiliary feed water flow, and condensate storage -
level and radiation levels.
D.
All repair procedures should be fully developed and '. should be verified that the materials for the repairs are maint:ined on site.
E.
The manpower for these procedures should be shown to be available on site and the work to be perfomed should be reasonable in light of the manpower available.
F.
Repairs _to systems used to place the reactor in the hot shutdown operating node should not be incorporated in the procedures es-tablished to bring the plant to hot shutdown.
G.
The ability to achieve safe shutdown in each plant area where a fire is postulated must be shown.
H.
Section III.G of Appendix R to 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 Sec-tion III.G.? 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 separated by the equivalent of a three-hour rated fire barrier from the fire area.
In evaluating an alternative shutdown method, associated circuits are circuits that could prevent operation or cause malfunction of the alternative train which is used to achieve and maintain hot shutdown conditions due to fire induced hot shorts, open circuits, or shorts the ground.
Safety related and nonsafety related cables that are associated with the equipment and cables of the alternative or dedicated method of shut wn 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 comon cower source with the alternative 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 shut-down capability, e.g., RHR/RCS isolation valves or (3) a common en-closure, e.g., raceway, panel, junction box with alternative shut-down 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 10 CFR Part 50 is provided by proposed modification the following information is required to demonstrate that associated circuits will not prevent operation or cause nalfunction of the alternative or dedicated shutdown method.
(1) Provide a table that lists all equipment including instrumenta-tion and support system equipment that are required by the alternative or dedicated method of achieving and maintaining hot shutdown.
(2) For each alternative shutdown equipment listed in (1) above, provide a table that lists the essential cable (instrumenta-tion, control and power) that are located in the fire area.
(3) Provide a table that lists safety related'and nonsafety related cables associated with the equipment in cables constituting the alternative or dedicated method of shutdown that are located in the fire area.
(4) -Show that fire induced failures of the cables listed in (2) and
-(3) above will not prevent operation or cause malfunction of the alternative or dedicated shutdown method.
(5) For each cable listed in (2) above provide a detailed elec-trical schematic drawing that show how each cable is isolated from the fire area.
I. The residual heat removal system is generally a low pressure sys-tem that interfaces with the high pressure primary coolant system.
To preclude a LOCA through this interface, we require compliance with the recommendations 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 assoc'ated 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 protected #rca the effects of a sir.gle fire, sie require the fo!-
lowing infornation:
Identify each high-low pressure interface that uses redundant electrically ~atrolled devices (such as two series motor operated valves) to isolate or preclude rupture of any primary coolant boundary.
Identify the device's essential cabling (power and control) and describe the cable routing (by fire area) from source to temin-ation.
Iden' tify each location where the identified cables are separated by less than a wall having a three-hcur fire rating from cables for the redundant device.
For the areas identified above (if any) provide the bases and justification as to the acceptability of the existing design or any proposed modifications.
,