ML20237K103
| ML20237K103 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 09/01/1987 |
| From: | Murphy W VERMONT YANKEE NUCLEAR POWER CORP. |
| To: | Rooney V NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM), Office of Nuclear Reactor Regulation |
| References | |
| CON-#487-4392 FVY-87-87, OLA, NUDOCS 8709040194 | |
| Download: ML20237K103 (18) | |
Text
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VERMONT YANKEE NUCLEAR POWER CORPORATION FVY 87-87 RD 5. Box 169. Ferry Road. Brattleboro, VT 05301
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ENGINEERING OFFICE fx
' 671 WORCESTER ROAD i
FRAMINGH AM, MASSACHUSETTS 01701 TELEPHONE 617-8 72-8100 September 1, 1987 1
1 U.S. Nuclear Regulatory Commission Document Control Desk Washington, D.C.
20555 Attn:
Office of Nuclear Reactor Regulation i
Mr. V.L. Rooney, Senior Project Manager Project Directorate I-3
~
Division of Reactor Projects I/II
References:
a)
License No. DPR-28 (Docket No. 50-271) b)
Letter, VYNPC to USNRC, FVY 86-34, " Proposed Technical Specification Change for Spent and New Fuel Storage",
dated 4/25/86 c)
Letter, USNRC to VYNPC, NVY 87-120. " Request for Additional Information - Spent Fuel Pool Expansion" (TAG 61351),
dated 8/7/87 d)
Letter, VYNPC to USNRC, FVY 87-65, " Vermont Yankee Proposed Change No. 133, Spent Fuel Pool Expansion - Response to NRC Request for Additional Information", dated 6/11/87 e)
Letter, USNRC to VYNPC, NVY 87-115 " July 14, 1987 Meeting with the Vermont Yankee Nuclear Power Corporation - Spent Fuel Pool Expansion", dated 7/30/87
Dear Sir:
Subject:
Vermont Yankee Proposed Change No.133 -- Spent Fuel Pool Expansion Pursuant to your recent letter [ Reference c)], Vermont Yankee herein pro-vides, as Attachment 1 to this letter, the information requested.
By letter dated June 11, 1987 [ Reference (d)], Vermont Yankee responded to several Staff questions pertaining to our Spent Fuel Pool Expansion (Reference (b)]. With respect to the Staffs question regarding operational controls asso-clated with the Spent Fuel Pool Cooling System (SPPCS), Vermont Yankee committed to administratively implement certain license conditions enclosed with our June 11, 1987 letter prior to start-up from our 1987 refuel outage.
Vermont Yankee also commited to subsequently submit those license conditions to the NRC for approval as a separate Technical Specification amendment request.
/L gol 8709040194 870901 PDR ADOCK 05000271 0
P PDR g
VERMoldT YANKEE NUCLEAR POWER CORPORATION U.S. Nuclear Regulatory Commission August 24, 1987
)
Page 2 l
Following the July 14, 1987 NRC/VY meeting (Reference e)], which discussed the review status of Vermont Yankee's Spent Fuel Pool Expansion request, Vermont Yankee determined that incorporation of the abovementioned operational controls l
within the Technical Specifications is not necessary. Accordingly, Vermont Yankee will not be submitting the Spent Fuel Fool Cooling System Operational l
Control Technical Specification amendment as previously stated in our letter of l
June 11, 1987 [ Reference d)].
l l provides a summary of the administrative controls Vermont Yankee will procedurally implement prior to startup from the 1987 outage to address the staff's concerns regarding Vermont Yankee Spent Fuel Pool Cooling System operability.
In addition, Attachment 3 is provided to further clarify Vermont Yankee's evaluation of the spent fuel pool cooling system heat removal capabilities.
We trust that the information provided by this letter is satisfactory and, accordingly, completes all of the information requested of Vermont Yankee to date associated with the subject spent fuel pool amendment requested.
Should you have any questions or require further information regarding this matter, please contact us.
Very truly yours, VERMONT YANKEE NUCLEAR POWER CORPORATION ANV' --
s+<
Warren P.
urp'y Vice President an Manager of Operat l
/dm cc: ASLB Service List
1 l
4 1
I ATTACHMENT 1 l
RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION VERMONT YANKEE SPENT FUEL POOL CAPACITY EX,ANSION P
i l
QUESTION 1 The information provided by the licensee indicates that the Spent Fuel Pool (SFP) Cooling System is not designed to Seismic Category I criteria.
In order to meet the guidelines of the Standard Review Plan, Section 9.1.3, when the SFP Cooling System is not Seismic Category I, the licensee should provide a SeismicL Category I Pool Water Makeup System and Spent Fuel Storage Building Ventilation System.
Further, the Building Ventilation System should conform to the guidelines of Regulatory Guide 1.52.
Therefore:
a)
Provide a description of tho Seismic Category I make-up capability to the SFP, including the water source and deli-very system.
b)
Provide a description of the Seismic Category I Spent Fuel Storage Area Ventilation System, including a discussion of its compliance with the guidelines of Regulatory Guide 1.52.
Note that Regulatory Guide 1.52 specifies a maximum tem-perature for inlet air of 180*F.
Therefore, provide the results of an analysic that demonstrates that the maximum air temperature for the Spent Fuel Storage Building is less than 180'F, or provide the results of an analysis which spe-cifies the maximum temperature and verifies that the Ventilation System can process air of that temperature, or provide justification for a deviation from the above cri-teria.
Response 1 DESIGN
SUMMARY
)
The following is a summary of Vermont Yankee's design capability to main-l tain safe shutdown reactor conditions and remove decay heat from the spent fuel pool assuming a seismic event.
3 I.
Although the SFPCS design is not documented to satisfy Seismic Category 1 criteria, power plant piping systems designed to ANSI B31.1 (such as Vermont Yankee's SFPCS) are typically capable of withstanding earthquake loads much greater than the design basis earthquake loads for Vermont Yankee.
The seismic experience database compiled by the Seismic Qualification Utility Group documents the reality that standard, non-seismically designed power plant piping failure due to seismic events is not likely.
A
V
\\Yttachment1 aO PMe 2 II.
If total loss of both trains of the spent fuel pool cooling system is t
assumed, one train of RHR is capable of removing both the SFP decay heat and the reacter's decay heat auch that safe shutdown conditions are' assured i
given a seismic event, loss of offsite power..and a single failure of one.
train of RHR.
The following aequence of events summarizes this capability (refer to Sketch B and Figure 1).
. gg Initial Conditions:
t 1.
The reactor is operating at full power lwgediately before the seismic
- event, y
2.
The torus temperature is 100*F (Technical Specification limit).
3 3.
The Spent Fuel Pool temperature is 150*F, and the heat load in the pool'is conservatively assumed to be 9.1 MBTU/hr, which is the heat 4
removal capacity of one train of SFPC.
4.
Reactor decay heat is calculated in accordance with the 1971 ANS draft /
- standard, s
l t
5.
At time zero (seismic event):
$) the SFPCS is assumed lost; b) Loss I
of Offsite Power and reactor scram is assumed; and c) only one train '
of RHR is available.
I Sequence of Events:
I 1.
The RRR torus cooling mode is initiated within 10 minutes.
2.
Deactor cooldown commences with high pressure coolant injection (HPCI) system.
s i
3.
The torus temperature rises as reactor decay heat is transferred to the torus?.lue to.HPCI operation.
In addition, the spent fuel pool temperature is rising due to sp nt fuel heat load ir the pool and loss of the normal SFPC system. The apent fuel pool reaches 190*F in approximately 10' hours.
4.
During the first 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />, the reactor can be cocled down to 150 psig utilizing HPCI/RCIC to establish and maintain safe shutdown.
In addition, the RHR/SFP cross-connect is made ready for augumented fuel pool cooling with RHE by reversing the spectacle flanges in the reactor building.
5.
Manual operator action is utilized to transfer the one RHR train bet-ween torus cooling and augmented fuel pool cooling modes as necessary to maintain the fuel pool below 200*F and the torus below 150'F (heat capacity limit when reactor is pressurized).
Page 3 l
6.
Evaluation of this scenario over the first two days indicates that
.I sufficient time exists for the operators to successfully cope with this low probability scenario.
The operator action cycle entails approximately G hours of torus cooling followed by 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> of aug-umented fuel pool cooling with 20 minutes allowed for valve realign-ment at enc'n transfer (see Figure 1).
III. Additionally, numerous seismic sources of make-up water to the spent fuel pool exist as discussed below. Should all plant capability to cool the fuel por ? become unavailable, make-up will assure safe wat(c level is main--
l tained in the pool until cooling capability is restored.
]
4 RESPONSE TO QUESTION 1.a i
Seismic Makeup Description The augrented fuel pool cooling mode (AFPC) of RHR provides cooling and emergency watir make-up to the pool should the Spent Fuel Pool Cooling system be unavailable.
The AFPC mode meetu SRP 9.1.3 requirements by providing a seismic makeup water path to the SFP and exceeds the SRP requirements by also providing spent fuel pool cooling capabilities as described previously.
Vermont Yankee has three seismically qualified sources of makeup water to j
the fuel pool for use during a seismic event (see Sketch A).
1.
Torus water can be pumped to the fuel pool by utilizing an RHR pump and the Rl;R/ spent fuel pool cross-connect piping.
2.
River water from the Service Water intake structure can be pumped to the j
fuel pool by utilizing a service water or fire water system pump, the RHRdW/RHR cross-connect piping, and the RHR/ Spent Fuel Pool cross-connect piping.
3.
Water stored in Vermont Yankee's cooling Tower Deep Basin Alternate Cooling Cell can be pumped to the fuel pool by utilizing a RHRSW pump, the RHRSW/RHR cross-connect piping, and the RHR/ spent fuel pool c'ross-connect piping.
In addition, as further backup, temporary hose can be installed from seismic service water, RHRSW, or RHR system piping inside the reactor building and routed to the spent fuel pool, i
A radiological analysis has been performed assuming a fuel pool boiling l
situation and taking no credit for SBGT. The results show that off-site dosage due to the release will not exceed the limits of 10CFR20.
In addition, fuel
(
pool boiling will not make the reactor building uninhabitable, j
l L____-__
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Page 4 I
RESPONSE TO QUESTION 1.b Seismic Ventilation System Description Vermont Yankee's apent fuel storage area ventilation system is the standby l
gas treatment system (SBGT).
The spent fuel poo) is located within the reactor j
building, which serves as secondary container.nt for the plant.
The SBGT system is designed to operate during design basis accident eituations to filter the secondary containment ventilation exhaust. The SBGT exhaust flow rate is such that the secondary containment will be maintr.ined under a slight vacuum with respect to the ambient outside etmosphere.
The SBGT system is comprised of two redundant filter-trains and is designed to Seismic Category 1 ciiteria (reference FSAR, Section 5.3.4).
l Vermont Yankee's SBGT system complies with the environmental design criteria outlined in Section C.1 of Regulatory Guide 1.52, Rev. 2.
These criteria require consideration of pressure differential, radiation dose rate, relative humidity, maximum and minimum temperature, end any other conditions resulting from a postulated Design Basis Accider.t (DBA).
The SBGT system is included in the Vermont Yankee Environmental Qualification Program (EQ) for electrical equipment. Under this program, essential components of the SBGT system that could experience harsh accident conditions are demonstrated to be qualifico for such conditicus.
The accident conditions used at Vermont Yankee have been determined with plant specific analysis (as suggested in Section B of Regulatory Guide 1.52), in lieu of the " typical accident conditions" listed in Table 1 of Regulatory 1.52.
The plant specific analysis specifies a maximum reactor building temperature of 132*F, and 115'F in the area of the SBGT system.
This i
analysis assumes that the spent fuel pool cooling system (SFPCS) 18 maintained at the Technical Specification temperature of 150*F.
Therefore, Vermont Yankee's spent fuel pool coo'ing system is also ircluded in the EQ program to l
insure operation under anticipated DBA environments.
In summary, the EQ program insures that the SBGT system will function when required under all anticipated Design Basis LOCA conditions. As discussed pre-vlously for LOCA events involving fuel pool boiling, if no ventilation was assumed, the release limits specified in 10CFR20 would not be exceeded.
QUESTION 2 Specify the location at which the SFP Cooling Sytem return line will be cut off in the SFP 2n terms of the distance from the top of the spent fuel storage racks.
1 1
Responsa 2 The proposed modification includes removal of a portion of the two cooling water return spargers.
Each downcomer will be cut approximately 15 ft above the top of the fuel storage racks (approximately 8 ft below the surface of the fuel pool water).
The portion of each sparger removed will be shipped as low level waste to a low level waste burial site, l
l Page 5 I
J QUESTION 3 Provide the results of a light load drop analysis in accordance with the guidelines of the Standa"d Review Plan, Section 9.1.4, which demonstrates that the dropping of any light load will not result in a radiological release in excess of that determined for the design basis fuel handling accident.
I RESPONSE TO QUESTION 3 The results from the light load drop annlysis performed indicate that the radiological consequences from such an event are bounded by the existing design l
basis fuel handling accident.
This analysis reviewed those loads normally carried over atored spent fuel and the heights at which they are handled. Also included in the review were those tools and equipment used around the spent fuel pool in conjunction with refueling and the re-rack which would weigh less than 700 lbs.
The kinetic energy created by a drop of these loads was calculated and compared to that used in the design basis fuel handling accident. The kinetic energies were all less than that used in the design basis fuel handling acci-dent. Therefore, the radiological release would not exceed those caltalated for the design basis fuel handling accident (any load in excess of 700 lbs would be handled in accordance with NUREG 0612).
l QUESTION 4 The February 25, 1987 submittal provided a drawing of the safe lead path boundaries for the carrying of the spent fuel storage 1
racks. This drawing indicates that some of the racks will be I
carried over the reactor vessel.
This is not in conformance with the guidelines provided in NUREG 0612 " Control of Heavy Loads at Nuclear Power Plants".
Provide justification for carrying the racks over the reactor vessel, including a discussion of the measures provided to prevent a load drop or provide a revised drewing which shows that the safe loud paths to be used do not paso over the reactor vessel.
Response 4 Section 5.1.1(1) of NUREG 0012 calls for heavy loads to be carried along paths that minimize the potential for impacting (if dropped) Irradiated fuel or safe shutdown equipment. Most of the racks will be handled over load paths that avoid going over the reactor vessel.
However, the load paths described in our February 28, 1987 letter show that some of the spent fuel racks will be carried' over the reactor vessel, specifically, one (1) 20 x 13 NES rack and four (4) 10 x 10 par racks.
These racks will only be carried along this load path when the reactor vessel head, drywell head, and concrete shield blocks are in place.
Alternative load paths are considered less desirable since they would require passing over irradiated fuel assemblies in the spent fuel pool..
Also, the maxi-mum lift height will be less than 15 inches, and the point of closest approach to the reactor vessel is essentially the outer surface of the vessel shell (see Figure 2).
With respect to NUREG 0612, the rack handling activities will conform to the following:
Page 6 j
1)
Safe load paths have been defined to minimize the possibility of a heavy load, if dropped, from impacting irradiated fuel in the reactor vessel and 1
spent fuel pool.
I 2)
Procedures will control the load handling operations.
~
3)
Crane operators are trained and qualified in accordance with Vermont Yankee's program that is based on Chapter 2-3 of ANSI B30.2.
4Property "ANSI code" (as page type) with input value "ANSI B30.2.</br></br>4" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process.)
Special lifting devices and slings conform to ANSI N14,6 and B30.9, respec-tively.
5)
The reactor building crane is inspected and maintained in accordance with Vermont Yankee's program that is based on Chapter 2-2 of ANSI B30.2.
6Property "ANSI code" (as page type) with input value "ANSI B30.2.</br></br>6" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process.)
The reactor building crane design conforms to Chapter 2-1 of ANSI B30.2 and CMAA-70.
7)
The reactor building crane and all fuel rack handling tools used in the reactor building to carry racks. over the reactor vessel areas are single-failure proof, in accordance with the guidelines of NUREG 0612.
Based on the above, we conclude that our rack handling activities are in accordance with the guidelines of NUREG 0612 and that Vermont Yankee fixtures and procedures have been optimized to minimize spent fuel pool risks.
QUESTION 5 In the June 8, 1987 telecon, you committed to increase the sur-veillance of the SFP water temperature to at least once every four hours whenever one of the seven identified alarms annun-clated in the Control Room, and to maintain this increased sur-veillance until continued adequate cooling for the SFP was j
restored or until the pool water temperature stabilized. The June 11, 1987 submittal was to document this commitment; however, instead it related the increased surveillance period to the RHR Technical Specification that was being proposed. Your commitment a compromise, based on not having SFP water temperature van alarms in the Control Room, to ensure that the operators would have adequate time to isolate the SFP Cleanup System and restore adequate pool cooling. Therefore, provide Control Room alarms which annunciate when the SFP water temperatures reach 140'F and 150'F or provide a Technical Specification surveillance require-ment for monitoring the SFP water temperature at least once every fout hours upon annunciation of any one of the seven alarms listed in your June lith submittel.
Response 5 Vermont Yankee presently has control roor 4 usunciation on Control Room Panel 9-4 which alarms in the event the spent fuel pool water temperature increases to 125'F.
The 125'F temperature limit is an administrative limit
1 I
l Page 7 Imposed to ensure the Tech Spec limit of 150*F is not violated. This tem-perature alarm receives its input from three (3) different areas:
- 1) heat exchanger A outlet; 2) heat exchanger B outlet; and (3) heat exchanger common inlet. Any one of these inputs will annunciate in the control room on Panel 9-4.
Given that the inputs to the spent fuel pool water temperature alarm would not be representative if the fuel pool cooling system were inoperable (because of no flow through the heat exchanger), Vermont Yankee committed to increase the surveillance of the spent fuel pool water temperature to at least once every four hours whenever one spent fuel pool cooling subsystem becomes inoperable.
The seven alarms indentified in our June 11, 1987 letter are control room alarms which indicate a potential fuel pool cooling subsystem problem.
It is Vermont Yankee's position that any one of the listed alarms would identify any potential problem and that the control room operators would evaluate the nature of the alarm.
If the alarm was indeed identifying a problem that rendered a fuel pool cooling subsystem inoperable, then the administrative cor.crols would be invoked.
However, if the operator did not substantiate that a fuel pool cooling subsystem was inoperable, the surveillance requirement would not be invoked.
For example, if the control room were to receive a "SW pump trip" alarm, this might inuicate that only one of three running SW pumps had tripped and the fourth standby SW pump would automatically start. The fuel pool cooling subsystems would not be inoperable and therefore no increased surveillance would be initiated.
It is important to note that VY Tech Spm s define operable as "a system, subsystem, train, component or device shall be operable or have operability when it is capable of performing its specified function (s).
Implicit in this defini-tion shall be the assumption that all necessary attendant instrumentation, controls, normal and emergency electrical power sources, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, sub-system, train, component or device to perform its function (s) are also capable of performing their related support function (s)."
Based on that definition, it is necessary that all auxiliary equipment be capable of performing their related support functions in order to declare the system / subsystem operable.
It is therefore redundant and unnecessary to require a specific 4-hour surveillance based solely on the receipt of any of the seven alarms unless the actual / conditions warrant the system / subsystem to be declared inoperable.
1 In accordance with the above, administrative controls incorporated by operational procedures exist whereby the shift supervisor / control rcom operators would fully evaluate any of the seven alarms thtt would potentially affect the operability of the fuel pool cooling system and subsequently invoke the admi-nistrative surveillance requirement that would assure that the fuel pool cooling demineralizers could be isolated prior to the inlet water temperature exceeding 140*F.
l
l l
Page 8 QUESTION 6 What, if any, changes will be required in the NPDES permit as a result of your proposed SFP expansion?
Response 6 l
l Vermont Yankee's National Pollutant Discharge Elimination System (NPDES) permit is based on the temperature of the discharge to the river. As the pro-posed reracking will not measurably increase the temperature of the discharge, l
Vermont Yankee's NPDES permit is not impacted.
I l
l QUESTION 7 Describe the disposal plans for current fuel storage racks.
Response 7 The existing par spent fuel racks will be removed from the pool, mechanically volume-reduced to a burial volume of less than 3,300 tt, and 3
shipped to Barnwell, South Carolina, as low level waste, for burial.
QUESTION 8 Describe the means of transportation and the locations between which you propose to ship the new spent fuel racks.
Response 8 The new Nuclear Energy Services (NES) designed spent fuel racks, fabricated under subcontract to U.S. Tool and Die, Inc., will be shipped via truck from the fabrication shop in Allison Park, F* asylvania, to the Vermont Yankee site in I
Vernon, Vermont.
1 l
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l
ATTACHMENT 2
SUMMARY
OF ADMINISTRATIVE CONTROLS FOR SPENT FUEL POOL COOLING SYSTEM (SFPCS) i The following controls apply to the operational status and the periodic monitoring of the fuel pool cooling system.
Incorporation of these administra-l tive controls within Vermont Yankee operational procedures will assure that ade-quate cooling is available for heat removal in the spent fuel pool.
j l
I.
Demineralizers Vermont Yankee will administrative 1y ensure isolation of the spent fuel pool filter /demineralizers at 140*F inlet temperature by implementing operational procedures.
II.
Fuel Pool Cooling Operability a)
Both fuel poc1 cooling subsystems shall be operable and capable of maintaining the fuel pool temperature below 150*F prior to a reactor i
startup from the cold shutdown condition.
{
1 b)
From and after the date that one of the fuel pool cooling subsystems is made or found inoperable and the remaining subsystem is capable of j
maintaining the fuci pool temperature below 150*F, then the reactor shall be in the cold chutdown condition within thirty days unless such 1
subsystem is sooner made operable.
l c)
From and after the date that both fuel pool cooling subsystems are made or found inoperable or the fuel pool temperature cannot be main-tained below 150*F, the reactor shall be in a cold shutdown condition prior to the fuel pool temperature exceeding 200*F.
l III. Fuel Pool Cooling System Surveillance a)
Surveillance of the fuel pool cooling subsystem shall be as follows:
1.
fuel pool temperature and level shall be monitored and recorded I
once per eight hour shift; and 1
2.
fuel pool temperature and level instrumentation shall be calibrated ared functionally tested once per operating cycle.
b)
When it is determined that the insarvice fuel pool cooling subsystem is inoperable, fuel pool temperature shall be monitored and recorded once per four hours until fuel pool temperature stabilizes
Page 2 IV.
Bases In order to ensure that the spent fuel pool cooling system is available to remove decay heat from the spent fuel pool, the administrative controls require that both spent fuel pool cooling subsystems be available prior to startup and that the spent fuel pool temperature be maintained below 150*F.
This ensures that the spent fuel pool cooling system will remain operable even with a single active failure.
The Vermont Yankee fuel pool cooling system consi:+s of two (2) pumps and two (2) heat exchangers.
Each pump can be aligned to supply water through one or both heat exchangers.
Normal:v, one pump and one heat exchanger are aligned together, thus providing two sebsystems.
Both subsystems can be operated simultaneously or by cross-connecting the heat exchangars, one pump can supply both heat exchangers.
A single active tallure for Vermont Yankee wculd be the loss of one fuel pool cooling pump. Our analysis shows that one pump and two heat exchangers can supply the required heat removal capabilities for the fuel pool. The other pump would be available as a backup to the running pump.
Thus, Vermont Yankee meets single active failure criteria for spent fuel pool cooling.
After startup, if one cooling subsystem becomes inoperable, continued operation is permitted for up to 30 days as long as the spent fuel pool temperature can be maintained below 150*F.
If the subsystem cannot be restored to operation within the 30-day time period, the plant must be shut down.
In the unlikely event that both fuel pool cooling subsystems become in-operable or if the temperature cannot be maintained below 150*F, a reactor shutdown must be completed prior to allowing the use of the RilR system to cool the spent fuel pool.
In the event of loss-of-cooling capability, the requirement that the reac-tor be in the shutdown condition prior to the spent fuel pool temperature reaching 200*F ensures that the fuel pool cooling event will not compound other events involving plant safety.
In order to verify the operability of the fuel pool cooling subsystems and ensure that sufficient time is availahie to take corrective actions well before exceeding the spent fuel pool temperature limit of 150*F, fuel pool temperature and level shall be monitored and recorded once per eight-hour shift, and associated instrumentation calibrated and functionally tested once per operating cycle. Additionally, in the event that one of the fuel pool subsystems is inoperable, fuel pool temperature shall be monitored and recorded once per four hours until fuel pool temperature is stat 111 zed.
ATTACHMENT 3 VY SPENT FUEL POOL COOLING SYSTEM HEAT REMOVAL CAPABILITY - CLARIFICATION j
The design flexibility of the Vermont Yankee fuel pool cooling system allows two (2) heat exchangers to be lined up in parallel so that either fuel pool cooling pump can supply both heat exchangers simultaneously.
This align-ment results in a heat removal capacity such that, assuming a single active failure, the spent fuel pool cooling system is capable of handling Standard Review Plan (SRP) calculated heat loadings eleven (11) days following shutdown.
The NRC has previously approved such a design and allowed credit to be taken for this capability (reference Millstone 2 Nuclear Power Plant SER, dated January 15, 1986 and Norti, Anna Units 1 & 2 SER, dated 12/21/84),
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