ML20247J237

Proposed Tech Specs Re Fuel Handling & Storage

ML20247J237
Person / Time
Site:	Fort Saint Vrain
Issue date:	09/14/1989
From:	PUBLIC SERVICE CO. OF COLORADO
To:
Shared Package
ML20247J229	List: ML20247J223 ML20247J237
References
P-89344, NUDOCS 8909200226
Download: ML20247J237 (35)

Text

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ATTACHMENT 2 TO P-89344 e PROPOSED CHANGES 8909200226 890914 oDR ADOCK 05000267

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. Fort St. Vrain #1 Technical Specifications Amendment No.

Page 2- 9 4

2.23 CALCULATED BULK CORE TEMPERATURE The CALCULATED BULK CORE TEMPERATURE shall be the calculated average temperature of the core, including graphite and fuel, but not the reflector, assuming a loss of all forced circulation of primary coolant flow.

2.24 CORE AVERAGE INLET TEMPERATURE The CORE AVERAGE INLET TEMPERATURE shall be the arithmetic average of the operating circulator inlet temperatures, adjusted for circulator power input, steam generator regenerative heat loads, and PCRV liner cooling system heat losses.

l 2.26 CORE ALTERATION tl l CORE ALTERATION (S) shall be the movement or manipulation l of any component within the PCRV that alters the core l reactivity (except for insertion of control rod pairs or l reserve shutdown material) or activities that could result i in damage to the core components, while fuel is in the l reactor vessel. Suspension of CORE ALTERATION (S) shall l not preclude completion of movement of a component to a l safe conservative position or condition.

l. 2.30 SHUTDOWN MARGIN l SHUTDOWN MARGIN shall be the instantaneous amount of I reactivity by which the reactor is subcritical or would be i subtritical from its present condition assuming that all l OPERABLE control rod pairs are fully inserted except for l the single control rod pair of highest reactivity worth I capable of being withdrawn, which is assumed to be fully l withdrawn.

.. Fort St. Vrain #1 Technical Specifications Amendment No.

Page 4.4-7b SPECIFICATION LCO 4.4.1 TABLE 4.4-4 (Part 2)

INSTRUMENT OPERATING REQUIREMENTS FOR REACTOR PROTECTIVE SYSTEM, R0D WITHDRAWAL PROHIBIT (RWP)

MINIMUM MINIMUM PERMISSIBLE OPERABLE DEGREE OF BYPASS NO. FUNCTIONAL UNIT CHANNELS REDUNDANCY CONDITIONS

1. Startup Channel - Low 2 1 Above 1.0E-03%

l Count Rate Rated Power (u) 2a. Linear Channel - Low 2 1 (g)

Power RWP (Channels 3, 4, and 5) 2b. Linear Channel - Low 2 1 (g)

Power RWP (Channels 6, 7, and 8) 3a. Linear Channel - High 2 (f) 1 None Power RWP (Channels 3, 4, and 5) 3b. Linear Channel - High 2 (f) 1 None Power RWP (Channels 6, 7, and 8)

Notes for Tables 4.4-1 through 4.4-4 are on Pages 4.4-8 and 4.4-9

.1 Fort St. Vrain 01 Technical Specifications Amendment No.

Page 4.4-9 NOTES FOR TABLES 4.4-1 through 4.4-4 (continued)

(r) -Separate instrumentation is pros'ded on each circulator for this functional unit. Only the affected helium circulator shall be shut down within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> if the indicated requirements are not met.

(s) Deleted.

(t) A prima ry coolant dew point moisture monitor shall not be considered operable unless the following conditions are met:

1) Reactor Power Range Minimum Sample Flow Startup to 2% 1 scc /sec.

> 2% - 5% 5 scc /sec.

> 5% - 20% 15 scc /sec.

>20% - 35% 30 scc /sec.

>35% - 100% 50 scc /sec.

2) Minimum flow of item 1) is alarmed in the control room and the alarm is set in accordance with the power ranges specified.

3) Deleted

4) Fixed alarms of I scc /sec and 75 scc /sec are operable.

l (u) When all control rod pairs have been withdrawn without l resulting in reactor criticality, as determined with a l calculated k(eff) not to exceed 0.95 assuming all conditions I specified in SR 4.1.4, the startup channel low count rate rod l withdrawal prohibit may be bypassed.

Fort St. Vrain #1

. Technical Specifications Amendment No.

Page 4.4-13 Basis for Specification LCO 4.4.1 (Continued) minimizing thermal cycling of plant and installed.

equipment.

Steam Leak Detection in the Turbine Building is required for equipment qualification of Safe Shutdown Cooling Systems. Thus, the limits and basis are the same as discussed in the basis for steam leak detection in the reactor building.

d. Rod Withdrawal Prohibit Inputs The termination of control rod withdrawal to prevent further reactivity addition will occur with the following conditions:

Startup Channel - Low Count Rate Start-up Channel -

Low Count Rate is provided to prevent control rod pair withdrawal and reactor startup without adequate neutron flux indication. The trip level is selected to be above the background l noise level. When all control rod pairs have been l withdrawn without resulting in reactor criticality, as l determined with a calculated k(eff.) not to exceed 0.95 l assuming the conditions specified in SR 4.1.4, the l startup channel monitors and scram function will l remain in service. However, the possibility of l reactor criticality will be so low as to permit the l bypass of the low count rate rod withdrawal prohibit l at that time.

Linear Channel - Low Power RWp Linear Channel (5% Power) directs the reactor operator's attention to either a downstale failure of a power range channel or improper positioning of the Interlock Sequence Switch. (FSAR Sections 7.1.2.2 and 7.1.2.8)

Linear Channel - High Power RWP Linear Channel (30% Power) is provided to prevent control rod pair withdrawal if reactor power exceeds the Interlock Sequence Switch limit for the

" Low Power" position. (FSAR Sections 7.3.2.2 and 7.1.2.8)

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L r. Fort St. Vrain #1

_ Technical Specifications Amendment No.

Page 4.5-3 i:

1.

Specification LCO 4.5.2 - Reactor Vessel Internal Maintenance, Limiting' Conditions for Operation (These requirements have been deleted. Refer to LCO 4.7.1 for Reactor Vessel Internal Maintenance Limiting Conditions for Operation.)

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Fort St. Vrain #1 gM. ,

. Technical Specifications b $h #- Amendment No.

0 l~ Page 4.7-1 py 4.7. FUEL HANDLING AND STORAGE SYSTEMS - LIMITING CONDITIONS FOR

' OPERATION

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Objectives To prevent an uncontrolled release of. radioactivity- during irradiated fuel handling and storage 'by defining the minimum operable. equipment and characteristics.

l' SPECIFICATION LCO 4.7.1 FUEL HANDLING AND MAINTENANCE IN' THE REACTOR 1 LIMITING CONDITION FOR OPERATION l The following reactor conditions shall be maintained:

-l a. The PCRV shall- be depressurized to atmospheric pressure

. -l -(< 1 psig) or slightly below,

- l b. The CORE AVERAGE INLET TEMPERATURE shall be 165 degrees F or l 1ess*, and I c. -The SHUTDOWN MARGIN requirements of LCO 3.1.4 shall be met.

l APPLICABILITY: Whenever both primary and secondary PCRV closures l of any PCRV penetration (s) are removed.

l l* Applicable only when the fuel handling machine or auxiliary l- transfer cask is mounted on the reactor vessel, with the cask l_ isolation valve and reactor isolation valve open.

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Fort St; Vrain.#1-

1. Technical Specifications.

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. Amendment No.

Page 4.7-2' 1

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SPECIFICATION LC0"4.7.1 (Continued)-

l ACTION:

l .a . 'With the conditions of a or b above.not met, restore the l . condition (s) to within the above. limits within .I hour, or l terminate _ ' fuel handling and resctor vessel internal l maintenance, retract the. fuel handling mechanism or any

_l other remotely operated mechanisms from the PCRV, and close l -the reactor isolation valve or opening through the PCRV as

-l' soon as practicable. Before the CORE AVERAGE INLET

.l , TEMPERATURE reaches 400 degrees F, a control rod drive and 1 orificei assembly shall be reinstalled; wherever both primary l and secondary PCRV closures are removed.

l b. - With the SHUTDOWN MARGIN requirements of LCO 3.1.4 not met, l comply with the ACTION items of LCO 3.1.4.

l-ASSOCIATED SURVEILLANCE REQUIREMENT: SR 5.7.3

-l BASIS FOR SPECIFICATION LCO 4.7.1 l To prevent the outleakage of primary coolant and potential release of l activity during fuel handling or maintenance in the reactor vessel, l the reactor must be depressurized and maintained within the required l conditions. The CORE AVERAGE INLET TEMPERATURE is limited to 165 l degrees F to prevent short-term pressurization of the fuel handling l equipment over 5 psig (the maximum allowable working pressure of the I-fuel handling equipment) as a result of accidental inleakage of water l into the vessel during fuel handling.

1 The. ACTION statements ensure that the reactor and Fuel Handling l Machine will be placed in the safest configuration as soon as

, l Practicable, if a required condition cannot be maintained. The

l. requirement to reinstall the control rod drive and orifice assembly

.l before the CORE AVERAGE INLET TEMPERATURE reaches 400 degrees F is l designed to avoid damage to the reactor isolation valve seal l material.

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. Fort St. Vrain #1 Technical Specifications Amendment No.

Page 4.7-3 t

l SPECIFICATION LCO 4.7.2 FUEL HANDLING MACHINE i

1 l LIMITING CONDITION FOR OPERATION l The Fuel Handling Machine (FHM) shall be OPERABLE with:

l a. A helium atmosphere

• in the FHM at atn,ospheric pressure (< 1 l psig) or slightly below, l b. The fuel har:dling purge system connected and OPERABLE, with a l supply of helium

• available, and the gas waste system available I to receive purge gas, l c. One cooling water coil operating with the outlet coolant water l temperature 150 degrees F or less, and i 1. An additional cooling water coil DPERABLE, or l 2. The backup fire water connections and hose OPERABLE, l d. The reactor building overhead crane attached, unless the FHM is l bolted to a reactor isolation valve over the reactor, a fuel I storage well, the original fuel shipping cask loading port, or l bolted to another seismically qualified locat'.on, and l e. Switches and alarms which verify correct orientation and I placement of fuel elements, reflector elements, or defueling

,  ! elements OPERABLE.

l APPLICABILITY: During use of the FHM for reactor internal I maintenance or any handling of IRRADIATED FUEL *.

I i_ l* A helium atmosphere is not required in the FHM when it is empty l

l or when handling nonirradiated fuel or handling IRRADIATED FUEL l that has undergone fission product decay for 100 days or more, or l when the FHM is not being used in fuel handling operations with j the reactar core or fuel storage wells.

. Fort St. Vrain #1 Technical Specifications Amendment No.

Page 4.7-4 l

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l l SPECIFICATION LCO 4.7.2 (Continued) l ACTION:

I a. With requirements a., or b., or d. above not satisfied, l' within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> initiate action to retract the fuel transfer l mechanism and close the reactor isolation valve and FHM cask l valve.

l b. With requirement a. above not satisfied when the FHM cask l valve is closed, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> initiate action to reduce gas l inflow to the FHM and/or increase gas outflow to the gas l waste system, as appropriate.

l l c. With requirement c. above not satisfied, within I hour  !

l initiate action to return the IRRADIATED FUEL elements I within the FHM to the reactor core, to the fuel storage -

l wells, or the fuel shipping casks, and terminate fuel I handling with the FHM.

l d. With requirement e. above not satisfied, terminate placement l of the fuel elements, reflector elements, or defueling l elements within the reactor vessel with the FHM.

l ASSOCIATED SURVEILLANCE REQUIREMENT: SR 5.7.1 l BASIS FOR SPECIFICATION LCO 4.7.2 l The objective of this specification is to ensure the prevention of I any uncontrolled release of radioactivity during handling of l IRRADIATED FUEL or reactor internal maintenance, and to ensure the l correct installation of fuel elements into the fuel storage wells and I installation of fuel elements, reflector elements, or defueling i elements into the reactor core. Defueling elements are described in j the Fort St. Vrain Defueling Safety Analysis Report, GA-C19694, j submitted to the NRC, August 16, 1989 (P-89287).

l The OPERABILITY requirements for the FHM ensure that:

l a. Failure of a primary seal at any boundary would result in l negligible outleakage or no more than slow inleakage of air, l since the machine is maintained at approximately atmospheric l pressure (< 1 psig) or slightly below,

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. Fort St. Vrain #1 Technical Specifications Amendment No.

Page 4.7-5 l BASIS FOR SPECIFICATION LCO 4.7.2 (Continued)

I b. The FHM is connected to the gas waste system, and can be vented I while in transit and/or when the cask valve is shut, giving l assurance that no pressure buildup in or escape of fission l products from the machine should occur, l c. Fuel elements contained in the FHM will be maintained at surface l l temperatures below 750 cegrees F to prevent significant graphi te l l oxidation in the event there is any inleakage of air. An outlet l cooling water temperature of less than or equal to 150 degrees F l from one of the two redundant cooling systems provides adequate l cooling to maintain the fuel elements below 750 degrees F ( FSAR l Sections 9.1.1 and 14.f;.3.1). Silencing of the low-flow alarm by l increasing the flow of cooling water ensures existence of cooling i I water flow, while the 150 degrees F limit ensures adequacy of l flow, and l d. A seismic event will not adversely affect the FHM when it is l bolted to a reactor isolation valve over the reactor, a fuel I storage well, the original fuel shipping cask loading pnrt, l bolted to another seismically qualified location, or supported by l the reactor building overhead crane.

l The ACTION statements ensure that the FHM will be promptly placed in l a safe configuration in the event any of the OPERABILITY requirements l can no longer be met. For example, if a switch or alarm necessary to l verify correct seating of fuel elements or other elements becomes l inoperable, the ACTION statement prohibits continuing seating of the l elements in the core. However, it does not prohibit unloading the l FHM into a fuel storage well where exact orientation is not a safety I concern. This would permit the FHM to be unloaded for easier repair I and minimizes radiation exposure to maintenance personnel.

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. Fort St. Vrain #I Technical-Specifications Amendment No.

Page 4.7-6 l SPECIFICATION LCO 4.7.3 FUEL STORAGE WELLS l LIMITING CONDITION FOR OPERATION l Each fuel storage well containing IRRADIATED FUEL shall have:

l a. A helium atmosphere maintained at approximately atmospheric l pr - ure (< I psig) or slightly below, and l b. Both cooling water coils OPERABLE with at least one coil in l operation, and the outlet cooling water temperature of each coil

.l operating shall be 150 degrees F or less, or I c. One cooling water coil operating with the outlet cooling water l temperature 150 degrees F or less, and the fuel storage well l emergency booster fan OPERABLE and capable of moving a minimum l- total air flow of 9000 cfm through the fuel storage vaults.

l d. No IRRADIATED FUEL located in the central column of a fuel l storage well.

l APPLICABILITY: When IRRADIATED FUEL is stored in the fuel storage l well and stored fuel element surface temperatures I can exceed 750 degrees F.

l ACTION: With less than the above required conditions satisfied,

i. l within I hour initiate corrective action to re-establish the l l required conditions within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or perform one of the l l following within the next 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s:

l a. Transfer IRRADIATED FUEL to a storage well or wells for l which the required conditions are met, or i b. Establish backup cooling to the affected fuel storage wells, l or l c. Perform an engineering evaluation to confirm that the l hottest fuel element surface temperature will not exceed 750 l degrees F.

l ASSOCIATED SURVEILLANCE REQUIREMENT: SR 5.7.2

- Fort.St. Vrain #1 Technical Specifications Amendment No.

Page 4.7-7 l BASIS FOR SPECIFICATION LCO 4,.13 l The storage well cooling water system is designed with two 100*4 l capability cooling coils supplied from independent water sources l (FSAR Section 9.1.2).

l The accident conditions described in the FSAR postulate the total l loss of cooling water to one of the nine fuel storage wells. If this l were to occur, adequate cooling could be achieved by an increase in l the normal ventilation air flow to cool the well by convection on the.

l external surface. The increase in air flow is supplied by the fuel l storage well emergency booster fan. This specific.ation is based on I the analysis in FSAR Section 14.6.3.2 which uses the conservative l assumption that a total flow of 9000 cfm would be drawn (equally I divided) through all three vault compartments of the fuel storage I wells, thus adequately protecting the affected storage well and fuel l within it from damaging temperatures. (There are three fuel storage l wells in each of the three fuel storage vaults).

l IRRADIATED FUEL elements are not stored in the central column of a l fuel storage well because they cannot be adequately cooled in that I location (FSAR Section 9.1.2.2.2).

I To prevent significant oxidation of the IRRADIATED FUEL, the fuel l storage wells are designed to maintain the IRRADIATED FUEL cool and l under a nominally dry atmosphere of helium. An outlet temperature of l 150 degrees or less from a single cooling coil ensures the hottest l IRRADIATED FUEL element surface temperatures will be maintained below l 750 degrees F, preventing any significant graphite oxidation in the j event of air inleakage into the storage well. The helium storage l system provides purified helium for this service, giving sufficient I protection against a moist atmosphere.

l The ACTION statements provide for corrective actions within en I adequate time to prevent the hottest stored fuel element surface l l temperature from reaching 750 degrees F, thus obviating any j significant oxidation in the unlikely cembined events of air l inleakage and loss of normal cooling. The applicability statement is I worded to indicate that this specification is not applicable if the I stored fuel element cannot reach a surface temperature of 750 degrees l F under any conditions.

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. Fort St. Vrain #1 Technical Specifications Amendment No.

Page 4.7-8 1'

l SPECIFICATION LCO 4.7.4 SPENT FUEL SHIPPING C'sSK .

I LIMITING CONDIi10N FOR OPERATION lA spent fuel shipping cask shall not be loaded or partially loaded i

l with IRRA01ATED FUEL having less than 100 days of fission product l decay.

l APPLICABILITY: At all times.

l ACTION:

l a. IRRADIATED FUEL with less than 100 days of fission product l decay shall be removed from the spent fuel shipping cask l . prior to lifting the cash.

i l b. The provisions of LCO 4.0.3 are not applicable.

1 ASSOCIATED SURVEILLANCE REQUIREMENT: .SR 5.7.4 I

l BASIS FOR SPECIFICATION LCO 4.7.4 _ _ _ _ .

1 The potential radiological consequences of an accident whereby the l spent fuel shipping cast breaks open while being lowered to the l truck, have been analyzed in FSAR Section 14.6.3.3. This analysis I assumes that the cask is loaded with the most radioactive IRRADIATED I FUEL elements contemplated to be shipped from the plant af ter 100 1 days of fission product decay.

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' Fort St. Vrain #1 s-Technical Specifications Amendment No.

Page 4.7-9

[ SPECIFICATION LCO 4.7.5 INSTRUMENTATION l LIMITING CONDITION FOR OPERATION l The reactivity of the core shall be continuously monitored by at

] least two startup channel neutron flux monitors in accordance with l LC0 4.4.1 (Tables 4.4-1 and 4.4-4).

l APPLICABILITY: REFUELING until all control rod pairs have been l withdrawn without resulting in reactor criticality l as determined with a calculated k(eff) not to l exceed 0.95 assuming all conditions specified in l SR 4.1.4. When the k(eff) determination has been l made, the startup channel rod withdrawal prohibit l of Table 4.4-4 is no longer required.

l ACTION:

l a. With one of the above required neutron flux monitors l l inoperable, or not operating, immediately suspend all involving ALTERATIONS, any evolution I operations CORE I resulting in positive reactivity changes, or movement of l IRRADIATED FUEL in the PCRV.

l b. With both of the above required neutron flux monitors l inoperable or not operating:

l 1. Immediately suspend all operations involving CORE l ALTERATIONS, any evolution resulting in positive I reactivity changes, or movement of IRRADIATED FUEL, l 2. Retract the fuel handling mechanism or any other i remotely operated mechanism from the PCRV, l 3. Close the reactor isolation valve or opening through l the PCRV as soon as practicable, and l 4. Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> evaluate the SHUTDOWN MARGIN per l SR 4.1.4.

j ASSOCIATED SURVEILLANCE REQUIREMENT: SR 5.7.5

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l , Fort St. Vrain #1 Lo

• Technical Specifications

, Amendment No.

Page 4.7-10 l

1 l BASIS FOR SPECIFICATION LCO 4.7.5 l l The OPERABILITY of the startup channel neutron flux monitors ensures l that redundant monitoring capability is available to detect changes l in the reactivity condition of the core. This specification also l applies to reactor internal maintenance that does not involve a CORE l ALTERATION, such as a helium circulator changeout or modifications or l repair of the primary coolant purification systems. For this

l. maintenance where no changes are being made to core reactivity, l activities are permitted in the event of inoperable startup channel

[ flux monitors, provided the other ACTION requirements are met.

l When all control rod pairs have been withdrawn without resulting in l reactor criticality, as determined with a calculated k(eff) not to

[ exceed 0.95 assuming all conditions specified in SR 4.1.4, the j startup channel monitors and scram function will remain in service, l but the' startup channel rod withdrawal prohibit of LCO 4.4.1 (Table l 4.4-4) will not be required. When this condition is reached, the

! probability of the reactor going critical will be very low (Reference l the Defueling Safety Analysis Report, GA-C19694, Section 3.1).

l The ACTION statement ensures that activities that could affect the l reactivity condition of the core are suspended whenever neutron flux l monitoring capabilities are degraded.

' Fort St. Vrain #1 s _

Technical Specifications i Amendment No.  !

Page 4.7-11

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l SPECIFICATION LCO 4.7.6 COMMUNICATIONS DURING CORE ALTERATIONS L

1 l LIMITING CONDITION FOR OPERATION l Direct two-way communications shall be maintained between operations l Control Room personnel and the Fuel Handling Machine (FHM) Control l Room personnel.

l APPLICABILITY: During CORE ALTERATIONS conducted from the l Refueling floor.

l ACTION: With no direct communications between the operations Control l Room personnel and the FHM Control Room personnel, suspend l all CORE ALTERATIONS.

l ASSOCIATED SURVEILLANCE REQUIREMENT: SR 5.7.6 l BASIS FOR SPECIFICATION LCO 4.7.6 l The requirement for communications ensure. that fuel handling l personnel can be promptly informed of significant changes in the l facility status or core reactivity conditions during CORE l ALTERATIONS, and operations Control Room personnel can be informed by l fuel handling personnel whenever CORE ALTERATIONS are being performed l so that core conditions can be monitored.

l The FHM Control Room and operations Control Room personnel must l coordinate control rod movements to ensure the required SHUTDOWN l MARGIN is maintained during CORE ALTERATIONS. Maintaining direct I communication also permits the operations Control Room personnel to l immediately request a stop of any movements causing excessive count l rate changes.

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. Fort St. Vrain #1

. Technical Specifications Amendment No.

-Page 5.7-1 l

l 5.7 FUEL HANDLING AND STORAGE SYSTEMS - SURVEILLANCE REQUIREMENTS l- Objective l To ensure the~ prevention of any uncontrolled release of l radioactivity during fuel handling and fuel storage by

-l establishing the minimum frequency and type of surveillance on l the equipment for the fuel handling and storage systems. (

1 l SPECIFICATION SR 5.7.1 FUEL HANDLING MACHINE (FHM) l The FHM shall be demonstrated OPERABLE whenever LC0 4.7.2 is l applicable:

l a. Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that:

l 1. The internal pressure is within the limit of LCO 4.7.2, and l 2. The cooling coil water outlet temperature is within the l limit of LCO 4.7.2.

I b. Within 31 days prior to beginning refueling operations and once i per 12 months thereafter by:

l 3. Functionally testing the FHM, FHM cask valve and associated l reactor isolation valve (s), and by performing a CHANNEL TEST l of the interlocks, limit switches, and alarms, l 2. Functionally testing the redundant cooling water coils and i verifying:

l a) The flow rates are sufficient to silence the low-flow l alarms, and l b) The outlet temperatures are less than 150 degrees F, l 3. Performing a CHANNEL CALIBRATION of the cooling water j temperature alarms, l 4. Functionally testing the FhM cooling water leak detector by l adding water to the drain header.

l 5. Visually inspecting the backup fire water connections and l hose to verify the hose and connector are in place and l undamaged, and

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Forc St. Vrain #1

. Technical Specifications Ar.,endment No.

Page 5.7-2 l SPECIFICATION SR 5.7.1 (Continued) l 6. Functionally testing the fuel handling purge system, l including availability of a supply of helium

• and the l availability of the gas waste system to receive purge gas by l evacuating and filling the FHM.

l c. Within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to initial attachment to the FHM, and once l per 31 days thereafter, demonstrate the reactor building ,

I overhead crane OPERABLE by: l l 1. Performing a visual inspection of crane components, l including cable and reeving, blocks and sheaves, and the "

l special lifting hook for carrying the FHM when it is l installed and secured on the crane, and l 2. Functionally testing the upper limit switch of the FHM l lifting hoist under no load by slowly running the block into l the limit switch.

l ASSOCIATED LCO: LCO 4.7.2 l

l* A helium atmosphere is not required in the FHM when it is empty l or when handling nonirradiated fuel or handling IRRADIATED FUEL l that has undergone fission product decay for 100 days or more, l or when the FHM is not being used in fuel handling operations l with the reactor core or fuel storage wells.

. . Fort St.'Vrain #1 L.'

Technical Specifications Amendment No. I Page 5.7-3 l EASIS FOR SPECIFICATION SR 5.7.1 1 The FHM provides for safe reactor fuel handling. To ensure the l reliability of the FHM during the fuel handling operations, the FHM j and the isolation valve (s) will be functionally tested prior to fuel I handling operations and once per 12 months thereafter, of extended l operation.

l A functional test of the fuel handling purge system and cooling water l system and an inspection of the backup fire water connections and l hase will be made. These checks ensure the capability to maintain l tie proper atmosphere environment within the FHM, to prevent any l uncontrollable release of activity, and the capability to maintain I the surface temperature of fuel elements within the FHM below 750 l degrees F.

l The reactor building overhead crane is inaccessible for visual l inspections and tests during operation with the FHM. It must be i moved to an accessible location to perform the surveillance l identified. Many of the crane operations involve several days when l it is attached to the FHM ar.d is not moved. Inspections once per 31 1 days are in additior, to other crane preventive maintenance l inspections in accordance with ANSI B.30.2 requirements and they are l appropriate for assuring safe operation of the reactor building i overhead crane.

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Fort St. Vrain #1 r

Technical Specifications Amendment No.

Page 5.7-4 i

l SPECIFICATION SR 5.7.2 FUEL STORAGE WELLS 3

l SURVEILLANCE REQUIREMENTS l The cooling water, purge, and ventilation systems for each fuel l' storage well containing IRRADIATED FUEL shall be demonstrated l OPERABLE whenever LCO 4.7.3 is applicable *:

l .a . Once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by:

l 1. Verifying that the outlet cooling vater temperature of l operating cooling coil (s) is 150 degrees F or less, with a l flow rate of greater than 7 gpm, and either:

I a) Both cooling water coils are OPERABLE with at least one l coil in operation, or l b) One cooling water coil is OPERAB E and in operation, I and the fuel storage well emergency booster fan is l OPERABLE.

l 2. Verifying that the pressure within the well is at l approximately atmospheric pressure (< 1 psig) or slightly l below.

l 3. Verifying that no IRRADIATED FUEL elements are inserted into l the central position of fuel storage wells.

l b. Once per 31 days by verifying that the fuel storage well l emergency booster fan operates upon manual initiation.

l c. Once per 18 months by verifying the capitbility of the fuel I storage well emergency booster fan to draw a .ninimum of 9000 cfm l of air through the fuel storage wells.

l ASSOCIATED LCO: LCO 4.7.3 l

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• Each of these surveillance shall be first performed within l

l 45 days after approval of this amendment or sooner.

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. . Fort St. Vrain #1

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Technical Specifications

(. Amendment No.

Page 5,7-5 L

l BASIS FOR SPECIFICATION SR 5.7.2 l

1 l Fuel ' elements contained in the fuel storage wells w111 be maintained l at surface temperatures below 750 degrees F to prevent significant l graphite oxidation in the event there is any inleakage of air. An u l outlet cooling water temperature of less than 150 degrees F in l l operating cooling coils provides adequate cooling to maintain the l fuel elements below 750 degrees F (FSAR Section 9.1.2).

l Verifying that the pressure within the well is at approximately I atmospheric pressure (< 1 psig) or slightly below ensures that.only r.

l negligible outleakage or no more than slow inleakage of air would l result from a failure in a primary seal.

l This specification is based on the analysis in FSAR Section 14.6.3.2 l which uses the conservative assumption that a total flow of 9000 cfm l would.be drawn (equally divided) tFrough all three vault compartments l of the fuel storage wells, thus adequately protecting the affected l storage well and fuel within it from damaging temperatures.

l The specified tests and testing frequencies are sufficient to I demonstrate the OPERABILITY of the fuel storage well emergency l booster fan, should it be called upon for performance of its required l safety function.

l IRRADIATED FUEL elements are not stored in the central column of a l fuel storage well because they cannot be adequately cooled in that l location (FSAR Section 9.1.2.2.2).

l The surveillance frequencies of this specification have been l established to be generally consistent with Standard Technical l Specification frequencies for LWR Fuel Storage Pool Air Cleanup l System surveillance.

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Technical Specifications Amendment No.

Page 5.7-6

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l SPECIFICATION SR 5.7.3 FUEL HANDLING AND MAINTENANCE IN THE REACTOR l a. The reactor pressure and temperature conditions shall be l determined to be within the limits of LCO 4.7.1 once per 12 l hours, I b. Verification of SHUTDOWN MARGIN shall be in accordance with l SR 4.1.4.

l ASSOCIATED LCO: LCO 4.7.1 l BASIS FOR SPECIFICATION SR 5.7.3 l The surveillance requirement frequency gives adequate assurance that l changes in reactor conditions will be detected in time to permit l corrective actions if required.

. Fort St. Vrain #1 o

Technical Specifications Amendment No.

Page 5.7-7 l SPECIFICATION SR 5.7.4 SPENT FUEL SHIPPING CASK l SURVEILLANCE REQUIREMENTS

{ Prior to loading the spent fuel shipping cask, the IRRADIATED FUEL j elements shall be determined to have undergone fission product decay l for 100 days or more.

l ASSOCIATED LCO: LCO 4.7.4 l BASIS FOR SPECIFICATION SR 5.7.4 1 Determination prior to loading the cask that the IRRADIATED FUEL l elements have undergone at least 100 days of fission product decay l ensures that a cask loading accident is within the assumptions used l in the safety analysis.

3 Fort St. Vrain'#1 L .

Technical Specifications

, Amendment No.

Page 5.7-8 L i l SPECIFICATION SR 5.7.5 INSTRUMENTATION l Each startup channel neutron flux monitor shall be demonstrated

-l OPERABLE whenever LCO 4.7.5 is applicable:

J a. Once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, by performance of a CHANNEL CHECK, I b. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to the initial start of CORE ALTERATIONS, l by performance of a CHtNNEL TEST, and l c. Once per 7 days, by performance of a CHANNEL TEST.

l ASSOCIATED LCO: LC0 4.7.5 l BASIS FOR SR 5.7.5 i The surveillance requirements ensure that the neutron flux monitors l are capable of detecting changes in reactor conditions in time to 1 permit corrective actions if required. These are in addition to the I calibration requirements of SR 5.4.1.

l When all control rod pairs have been withdrawn without resulting in l reactor criticality, as determined with a calculated k(eff) not to l exceed 0.95 assuming all conditions specified in SR 4.1.4, the l startup channel scram will remain in service, but the startup channel l low count rate rod withdrawal prohibit of LCO 4.4.1 (Table 4.4-4) l will not be required. At this point the probability of the reactor j going critical will be very low (Reference the Defueling Safety l Analysis Report, GA-C19694 Section 3.1).

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L: L* . Fort St. Vrain #1 Technical Specifications ]

-Amendment No. i Page 5.7-9 1 i

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] SPECIFICATION SR 5.7.6 COMMUNICATIONS'DURING CORE ALTERATIONS z l

l SURVEILLANCE REQUIREMENTS j l Direct communications between the operations Control Room personnel l and the FHM Control Room personnel shall be demonstrated within one I hour prior to the start of and once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during CORE l ALTERATIONS.

l ASSOCIATED LCO: LCO 4.7.6 i

l BASIS FOR SPECIFICATION SR 5.7.6 l The surveillance times specified give adequate assurance that l-communications will be available as needed.

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l k ATTACHMENT 3 TO P-89344' NO SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS

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Attachment 3 P-89344 Page 1 of 6 NO SIGNIFICANT HA7_ARDS CONSIDERATION ANALYSIS BACKGROUND:

In a letter dated August 6, 1989, (G-89261) the NRC has requested that Public Service Company of Colorado (PSC) submit these Technical Specification amendments to the fuel handling and fuel storage sections of the Fort St. Vrain Technical Specifications. The fuel handling sections, for the most part, are only applicable while the reactor is shutdown. The fuel storage sections can be applicable more frequently. The handling of defueling elements as described in the Fort St. Vrain Defueling Safety Analysis Report, GA-C19694 (Reference 7), has been added to LCO 4.7.2. These changes have been stimulated by the Technical Specification Upgrade Program (TSUP),

previously reviewed by the NRC, and by the Defueling Safety Analysis Report.

SYSTEM DESCRIPTION:

The Fort St. Vrain HTGR has 37 control rod pairs that enter the core from the top. For an accident such as loss of power the control rod drive motors and brakes are deenergized and gravity causes the control rods to fall into the core. During fuel handling, the reactor is shutdown with most of the control rods inserted. The Fuel Handling Machine is used to remove and insert fuel elements into the core during a refueling. The changes being included with this submittal include increased monitoring and surveillance of the Fuel Handling Machine operation.

Other changes being included involve the cooling and monitoring of spent fuel in storage wells. Each of the changes is addressed in the following evaluation.

l- EVALUATION:

PSC has evaluated the proposed amendment request for significant hazards consideration using the standards in Title 10, Code of Federal Regulations, Part 50.92. The proposed amendment request I involves no significant hazards, since the proposed amendment would not:

1. Involve a significant increase in the probability or consequences of an accident previously evaluated. j l

The revisions to Sections 4.7 and 5.7 of the Fort St. Vrain Technical Specifications contained herein are being made to 3 improve the specifications for maintaining safety and ensuring proper surveillance of the plant safety systems.

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Attachment 3 P-89344

.Page 2 of 6  !

Section 2.26 is added to include a definition of Core Alterations.

Section 2.30 is added to include a definition of Shutdown

. Margin.

LCO 4.4.1 has been modified to add Note (u) to Page 4.4-7b under permissable bypass conditions . This note identifies a condition where the startup channel. low count rate rod withdrawal prohibit can be bypassed because the probability of the reactor going critical is very low. This condition is discussed in more detail in the Fort St. Vrain Defueling Safety Analysis Report, Section 3.1.

LCO 4.5.2 is deleted as it is duplicated by the requirements of LCO 4.7.1.

LCO 4.7.1 adds a clarification of atmospheric pressure as being less than 1 psig, relocates the requirement for monitoring of neutron flux to LCO 4.7.5, adds a requirement to maintain the shutdown margin requirements of LCO 3.1.4, adds the Auxiliary Transfer Cask to the note, and adds an action statement to close up the reactor before the core average inlet temperature reaches 400 degrees F. Each of these changes is a format change or improves the specifications and safe operation of plant equipment. The clarification of atmospheric pressure to be less than 1 psig maintains the reactor pressure well within the pressure ratings of the associated equipment, reactor isolation va'lve (5 psig) and Fuel Handling Machine (5 psig). Maintenance of the shutdown margin during fuel handling and maintenance is a confirmation of that requirement from LCO 3.1.4. The addition of the Auxiliary Transfer Cask is included as it can be used in place of the Fuel Handling Machine for reactor maintenance. The change of " Specification 3.1.3" to "LCO 3.1.4" is made to agree with the new Reactivity

Technical Specifications. An applicability statement I

stating that applicability is whenever botn primary and secondary closures are open on any PCRV penetration is unchanged from the TSUP LCO 3.9.1. The action statement is very similar to the current requirement of LCO 4.7.1. The insertion of a control rod drive and orifice assembly (ies) to close up the reactor before the core average inlet temperature reaches 400 degrees F is designed to protect the seal material of the reactor isolation val ve( s) . None of these changes involve a significant increase in the probability or consequences of an accident previously evaluated.

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r Attachment 3

.P-89344 Page 3 of 6 LC0 4.7.2 adds a clarification at atmospheric pressure, adds a requirement for the gas waste system to be available, adds a requirement for an additional cooling water coil to be

. operable or to.have backup fire water' connections and hose operable, adds a requirement to keep the reactor building crane attached to the Fuel Handling Machine except when the Fuel Handling Machine is bolted to a seismically qualified location, and adds a requirement to maintain switches and alarms operable. An applicability Statement has been added which is very similar to the current requirements of LCO 4.7.2 and includes reactor internal maintenance. Action Item c. is modified to include the option of unloading fuel to the Fuel Storage Casks. Action Item d. is new and adds reflector elements and defueling elements to the statement.

Eech of .the changes maos is either a format change or improves the specifications ano safe operation of the equipment and maintains TSUP analysi:. Defueling elements are described in the Fort St. Vrain Defueling Safety Analysis Report, GA-C19694.

The LC0 4.7.3 title has been changed to Fuel Storage Wells, a clarification of atmospheric pressure is added, the requirement to have both cooling cails operable is changed, the requirement for emergency cooling air flow is changed to l 9000 cfm in accordance with FSAR section 14.6.3.2, a l prohibition from storing irradiated fuel in the central column of a fuel storage well is added and two action statements to incorporate TSUP and FSAR section 14.6.3.2 analyses are added. The Applicability Statement has been revised from the TSUP to adcress the possibility that the stored fuel elements can, at some time, maintain a temperature below 750 degrees F without the provisions of this LCO. The Basis is revised to discuss the fact that this LCO is only required to ensure that the fuel element l- surface temperatures do not exceed 750 degrees F. Each of the changes made is a format change or is a change in accordance with FSAR and/or TSUP analyses. The prohibition from storing irradiated fuel in the central column is addressed in FSAR Section 14.6.3.2 and is a new addition to the Technical Specifications.

LCO 4.7.4 and the Basis are idertical to TSUP LCC 3.9.5 which has previously been reviewed by the NRC.

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10 Attachment 3 P-89344 Page 4'of 6 LCO 4.7.5 adds the requirements for monitoring the core reactivity that were formerly included in LCO 4.7.1 and adds action statements that require the startup channels to be operable. A provision has been added that when all control rods have been withdrawn without re sulting in reactor criticality, as determined with a calculated k(eff) not to exceed 0.95 assuming all conditions specified in SR 4.1.4, the startup channel low count rate rod -withdrawal prohibit can be bypassed. These provisions are in agreement with the Defueling Safety Analysis Report, (Reference 7) Section 3.1.

LCO 4_7 6 is a new LCO addressing the maintenance of two-way communications between the Control Room and the refueling deck. It maintains the TSUP analyses and format intact. It also includes an applicability statement and an associated surveillance requirement reference. These additional requirements are considered an enhancement to the Technical Specifications.

SR 5.7.1 changes the current SR 5.7.1 with surveillance every 12 months instead of prior to use and every refueling period. The surveillance requirements of SR 5.7.1 are very similar to TSUP SR 4.9.3 except that some editorial changes are made and the frequency of item c., reactor building crane inspection, has been changed from every 14 days to every 31 days because movement of the reactor building crane is very infrequent. The 31 day interval, in addition to other preventive maintenance inspections required by ANSI B.30.2, adequately assures that the crane will perform properly. This surveillance is an enhancement of the current SR 5.7.1. .

SR 5.7.2 changed title of Surveillance from " Fuel Storage Facility" to "Fuei Storage Wells". This Specification is substantially revamped to incorporate the TSUP and FSAR Section 14.6.3.2 analyses. It also envelopes the current Technical Specification analysis. It includes an associated LCO statement. It replaces "once per REFUELING CYCLE" in item c. of the TSUP with "once per 18 months". REFUELING CYCLE is de(ined as 18 months. This surveillance is an enhancement of the current SR 5.7.2.

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i Attachment 3 P-89344 Page 5 of 6 l SR 5.7.3 is a new surveillance for the Technical Specifications and adds verification of reactor pressure and temperature every 12 F:urs. This surveillance and basis are identical to TSUP SR ' 9.1.

. These are new surveillance requirements and enhance the Technical Specifications.

SR 5.7.4 is a new SR added to. provide surveillance requirements not included in the current Technical Specifications for the Spent Fuel Shipping Cask. SR 5.7.4 and the Basis are identical to TSUP SR 4.9.5.

SR 5.7.5 is a new specification for the current Technical Specifications. The new SR 5.7.5 is very similar to TSUP 4.9.2. The Basis is revised to add reference to the condition where subtriticality has been demonstrated with all control rods withdrawn and with a calculated k(eff) not greater than 0.95.

This surveillance is new to the Technical Specifications, supplements the requirements of SR 5.4.1, and is considered an enhancement to the Technical Specifications. The permissible bypass conditions of the startup channel low count rate rod withdrawal prohibit are addressed by the LCD 4.4.1 and LCO 4.7.5.

l SR 5.7.6 is a new SR added to provide surveillance I requirements for the new Communications During Core l

Alterations LC0 (LCO 4.7.6). It also includes an associated LCO statement. It incorporates the TSUP analysis and is an enhancement to the Technical Specifications.

These changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

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O Attachment 3 P-89344 Page 6 of 6

2. Create the possibility of a new or different kind of accident from any accident previously evaluated.

As described in item I, no change is being made to plant

, operation or safety systems that has not been previously l analyzed in the FSAR, TSUP analyses, or the Defueling Safety Analysis Report, GA-C19694 (Reference 7).

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3. Involve a significant reduction in a margin of safety.

l No margins of safety are being affected other than those already analyzed in the FSAR or Defueling Safety Analysis Report, GA-C19694. The changes made are format and other changes which will better implement the conditions specified l

by the FSAR and Defueling Safety Analysis Report, GA-C19694 CONCLUSION:

Based on the above evaluation, it is concluded that operation of Fort St. Vrain, in accordance with the proposed changes, will involve no

significant hazards consideration. PSC considers the proposed I changes to be an improvement in the overall plant reliability i. n d documentation as the new limiting conditions for operation and surveillance are designed to improve fuel handling and storage performance. The majority of these changes have been previously reviewed by the NRC during the TSUP review process (Reference 3).

i Other changes have been presented in the Fort St. Vrain Defueling Safety Analysis Report, GA-C19694 (Reference 7).

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ATTACHMENT 4 TO P-89344 REFERENCES

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o Attachment 4 P-89344 Page 1 of 1 References

1) PSC Letter, Brey to Calvo, Dated May 27, 1988 (P-88184)

2) PSC Letter, Brey to Calvo, Dated June 14,1988 (P-88205) 3)' NRC Letter, Heitner to Williams, Dated April 19, 1989 (G-89145)

4) PSC Letter, Hock to Murley, Dated May 11, 1989 (P-89180)

5) PSC Letter, Williams to Hebdon, Dated May 15, 1989 (P-89183)

6) NRC Letter, Heitner to Williams, Dated August 8, 1989 (G-89261)

7) PSC Letter, Crawford to Weiss, Dated August 16, 1989 (P-89287)

(GA-C19694 Fort St. Vrain Defueling Safety Analysis Report)