ML20062B610

Amend 133 to License DPR-40,modified Tech Specs to Increase Refueling Boron Concentration to 1900 Ppm & Revises Storage Requirements of Spent Fuel Pool Region 2

ML20062B610
Person / Time
Site:	Fort Calhoun
Issue date:	10/12/1990
From:	Quay T Office of Nuclear Reactor Regulation
To:
Shared Package
ML20062B607	List: ML20062B604 ML20062B610 ML20062B616
References
NUDOCS 9010250258
Download: ML20062B610 (18)

Text

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. umit] ;'ATES i E W 4 r#\,a NUCLEAR RECULATORY COMMISSION  !

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l OMAHA PUBLIC POWER DISTRICT .

l DOCKET NO. 50-285 FORT CALHOUN STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.133 t License No. DPR-40 ,

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1. The Nuclear Regulatory Commission (the Commission) has found that:

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A. The application for amendment by the Omaha Public Power District (the licensee) dated January 26, 1990, as supplemented May 10 June 18, and August 2, 1990, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in.10 CFR Chapter I; s-B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; l '

C. There is reasonable assurance. (i) that the activities authorized l

by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this license amendment will not be inimical to the '

common defense and security or to the health and safety of the public; and ,

E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied, i

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2. Accordingly, Facility Oper6 ting License No. DPR-40 is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B. of Facility Operating License No.

DPR-40 is hereby amended to read as follows: I B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.133 , are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

The license amendment is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

$ 6 A. 49 Theodore R. Quay, Acting Director Project Directorate IV-1 i

• Division of Reactor Projects - III, IV, V and Special Projects l Office of Nuclear Reactor Regulation i

Attachment:

l Changes to the Technical Specifications ,

Date of Issuance: October 12, '.990 l

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l ATTACHMENT TO LICENSE AMENDMENT NO.133 FACILITY OPERATING LICENSE NO. DPR-40 DOCKET NO. 50-285 i

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Revise Appendix "A" Technical Specifications as indicated below. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.

REMOVE PAGES INSERT PAGES 2 2 2-20 2-20 2-22 2-22 2-37 2-37 2-38 2-38 2-39 2-39 2-62 2-62 2-89 2-89 2-91 2-91 2-92 2-92 3-18 3-18

.. 3-19 3-19 i 4-4 4-4 5-15 5-15 Fig 2-10 Fig 2-10 1

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DEFINITIONS '

REACTOR OPERATING CONDITIONS (Continued)

Cold Shutdown Condition (Operating Mode 4) i The reactor coolant Tcold islessthan210'Fandthereactorcoola$tisat '

shutdown boron concentration.

Refuelina Shutdown Condition (Operating Mode 5)

The reactor coolant is at refueling boron concentration and Tcold is less than 210'F.

Refuelina Operation Any operation involving the shuffling, removal, or replacement of nuclear fuel, CEA's, or startup sources.

• g . ,

The Refuelina Boron Concentration A reactor coolant boron concentration of at least 1900 ppm, which corresponds l to a shutdown margin of not less than 5% with all CEA's withdrawn.

Shutdown Boron Concentration The boron concentration required to make the reactor suberitical by the amount defined in paragraph 2.10.

Refuelina Outaae or Refuelina Shutdown A plant outage or shutdown to perform refueling operations upon reaching the planned fuel depletion for a specific core.*

Plant Ooeratina Cvele The time period from a Refueling Shutdown t'o the next Refueling Shutdown.

Amendment No. Jf,JJ,ff,p),JpJ,133 2

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.2.0 L'IMITING CONDITIONS FOR OPERATION i' 2.3 Imeraency Core Coolina System I

Aeolicability '

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Applies to the operating statur. of the emergency core cooling iystem.

Objective To assure operability of equipment required to remove decay heat from the coro.  ;

i Specifications 1

(1) Minimum Reauirements The reactor shall not be made critical unless all of the following  ;

conditions are met: '

a. The SIRW tank contains not less than 283,000 gallons of water

' with a boron concentration of at least 1900 ppm at a temperature l not less than 50'F. ,

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, b. One means of temperature indication (local) of the SIRW tank is

,. operable,

c. All four safety injection tanks are operable and pressurized to i

at least 240 psig with a tank liquid of at least 116.2 inches j b(oronconcentration.67%) and a n.aximum level- of 128.1 inches (74%) wit 1

d. One level and one pressure instrument is operable on each safety injection tank,

e. One low pressure safety injection pump is operable on each bus.

f. One high pressure t.afety injection pump is operable on each bus.

g. Both shutdown heat exchangers and three of four component cooling heat exchangers are operable.

I h. Piping and valves shall be operable to provide two flow paths from the SIRW tank to the reactor coolant system.

1. All valves, piping and interlocks associated with the above components and required to function during accident conditions are operable. HCV 2914, 2934, 2974, and 2954 shall have power removed from the motor operators by locking open the circuit breakers in the power supply lines to the valve motor operators.  ;

FCV-326 shall be locked open, t

2-20 Amendment No. ,JJ,5J,19),JJ7,JJJ i

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2.0 LIM 111NG CONDITIONS FOR OPERQ1QN 2.3 fmycency Core Coolina System (Continued) t (3) Protection Aaainst low Temoerature Overoressurization i

The following limiting conditions shall be applied during scheduled heatups and cooldowns. Disabling of the HPSI pumps need 60t be  ;

required if the reactor vessel head, a pressurizer safety valve, or a i PORY is removed.

I Whenever the reactor coolant system cold leg temperature is below' i

320'F, at least one (1) HPSI pump shall be disabled. t i

Whenever the reactor coolant system cold leg temperature is below '

312'F, at least two (2) HPSI pumps shall be disabled. .

i Whenever the reactor coolant system cold leg temperature is below i 271'F, all three (3) HPSI pumps shall be disabled. ,

In the event that no charging pumps are operable, a single HPSI pump l

' may be made operable and utilized for boric acid injection to the core.

BALil The normal procedure for starting the reactor is to first heat the reactor coolant to near operating temperature by running the reactor coolant pumps. The reactor is then made critical by withdrawing CEA's and diluting i

boron in the reactor coolant. With this mode of start up, the energy stored in the reactor coolant during the approach to criticality is substantially equal to that during power operation and therefore all l engineered safety features and auxiliary cooling systems are required to be fully operable. During low power physics tests at low temperatures, there is a negligible amount of stored energy in the reactor coolant; therefore, an accident comparable in severity to the design basis accident is not possible and the engineered safeguards systems are not required.

The SIRW tank contains a minimum of {83,000 gallons of usable water containing at least 1900 ppm boron'II . This is sufficient boron '

l concentration to provide a shutdown margin of 5%, including allowances for uncertainties, w temperatureof60*F.g, all control rods withdrawn and a new core at a +

The limits for the safety injection tank pressure and volume assure the required amount of water injection during an accident and are based on ,

values corresponds used for to athe accident volume analgses.

of 825 ft The minimum 116.2 inch level corresponds to a volume of 895.5 ft d the maximum 128.1 inch level Prior to the time the reactor is brought critical, the valving of the '

safety injection system must be checked for correct alignment and appropriate valves locked. Since the system is used for shutdown cooling, the valving will be changed and must be properly aligned prior to start up of the reactor.

2-22 Amendment No. 17,U ,f),f7,5f,7f, 77,Jpp,JM,133

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i 2.0 LIMITING CONDITIONS FOR OPERATlQN 2.8 Refuelino Operations Applicabilit y i Applies to operating limitations during refueling operations.

• Obiective  !

To minimize the possibility of an accident occurring during refueling operations that could affect public health and safety.  ;

Specifications '

The following conditions shall be satisfied during any refueling j operations:

s (1) The equipment hatch and one door in the air lock shall be properly  !'

closed. In addition, all automatic containment isolation valves shall be operable or at least one valve in each line shall be closed.

(2) The five containment atmosphere and plant ventilation duct radiation monitors that initiate closure of the containment pressure relief, air sample, and purge system valves shall be tested and verified to be t oserable immediately prior to refueling operations. The five monitors '

sla11 employ one out of five logic from separate contact outputs for VIAS.

(3) Radiation levels in the containment and spent fuel storage areas shall be monitored continuously.

i (4) Whenever core geometry is being changed, neutron flux shall be .

continuously monitored by at least two source range neutron monitors,  ;

with each monitor providing continuous visual indication in the control room. When core geometry is not being changed, at least one ,

source range neutron monitor shall be in service. 4 (5) At least one shutdown cooling pump and heat exchanger shall be in operation. However, the pump and heat exchanger may be removed from operation for up to one hour per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period during the performance of core alterations in the vicinity of the reactor coolant hot leg l loops or during manipulation of a source, i

k 2-37 Amendment No. U ,# ,133 i

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2'0 LIP 1JLNG CONDITIONS FOR OPERATIONS 2.4 Refuelino Doerations (Continued)

(6) Direct communication between personnel in the control room and at l the refueling machine shall be available whenever changes in core geometry are taking place.

(7) When irradiated fuel is being handled in the auxiliary building, l the exhaust ventilation from the spent fuel pool area will be diverted through the charcoal filter.

(8) Prior to initial core loading and prior to refueling operations. -l a com)1ete check out, including a load test, shall be conducted on fuel aandling cranes that will be required during the refueling operation to handle spent fuel assemblies.

(9) A minimum of 23 feet of water above the top of the core shall be l maintained whenever irradiated fuel is being handled.

(10) Storage in Region 1 and Region 2 of the spent fuel racks shall be l restricted to fuel assemblies having initial enrichment less than or equal to 4.0 weight percent of U 235.

(11) Storage in Region 2 of the spent fuel racks shall be-restricted to l those assemblies whose parameters fall within the ' acceptable" region of Figure 2 10, if any of the above conditions are not met, all refueling operations shall cease immediately, work shall be initiated to satisfy the required conditions, and no operations that may change the reactivity of the core shall be made. However, refueling operations may commence and continue with less than 5 containment atmosphere and plant ventilation duct radiation monitors provided that gross, particulate and iodine monitors are monitoring the stack effluent. These three plant ventilation duct radiation monitors will initiate closure of the containment pressure relief, air sample and purge system valves and shall employ a one-out-of three logic for the initiation of VIAS.

The spent fuel assembly may be transferred directly from the reactor core to the spent fuel pool Region 2 provided the independent verification of assembly burnups as defined in Special Procedure SP-BURNUP 1 has been completed and the assembly burnup meets the acceptance criteria . identified in Technical Specification Figure 2-10.

Irradiated fuel movement shall not be-initiated before the reactor core has decayed for a minimum of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if the reactor has been operated at power levels in excess of 2% rated power.

h111 The equipment and general procedures to be utilized during refueling operations are discussed in the USAR. Detailed instructions, the above specifications, and the design of the fuel handling equipment incorporating.

built-in interlocks and safety features provide assurance that no 2-38 Amendment No ),Jf,JJ,pJ,75,133 1

2.0 11MITING CONDITIONS FOR OPERATION 2.8 Refuelino Operations (Continued) ,

incident could occur during the refueling operations that would result in a  ;

hazardtopublichealthandsafety.(1) Whenever changes are pot being  :

made in core geometry one flux monitor is sufficient. This permits '

maintenance of the instrumentation. Continuous monitoring of radiation t levels and neutron flux provides immediate indication of an unsafe condition. The shutdown cooling pump is used to maintain a uniform boron concentration. '

The shutdown margin as indicated will keep the core subcritical even if all CEA's were withdrawn from the core. During refueling operations, the reactor refueling cavity is filled with approximately 250,000 gallons of borated water. The boron concentration of this water (at least 1900 ppm l l boron) is sufficient to maintain the reactor suberitical by more than 5%

including allowance for uncertainties, in the cold condition with all

rodswithdrawn.(2)hutdownmargin.

ensure the proper s Periodic checks Communication of refueling requirements allow the water boron co control room operator to inform the refueling machine operator of any ,

impending unsafe condition detected from the main control board indicators during fuel movement.

' In addition to the above engineered safety features, interlocks are l utilized during refueling operations to ensure safe handling. An excess weight interlock is provided on the lifting hoist to prevent movement of more than one fuel assembly at a time. In addition, interlocks on the auxiliary building crane will prevent the trolley from being moved over storage racks containing irradiated fuel, except as necessary for the handling of fuel. The restriction of not moving fuel in the reactor for a period of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after the power has been removed from the core takes advantage of the decay of the short half-life fission products and allows for any failed fuel to purge itself of fission gases, thus reducing the consequences of fuel handling sceident.

The ventilation air for both the containment and the spent fuel pool area flows through absolute particulate filters and radiation moniters before discharge at the ventilation discharge duct. In the event the stack discharge should indicate a release 'n excess of the limits in the technical specifications, the containment ventilation flow paths will be closed automatically and the auxiliary building ventilation flow paths will be closed manually. In addition, the exhaust ventilation ductwork from the spent fuel storage area is equipped with a charcoal filter which will be manually put into operation whenever irradiated fuel is being handled.(1)

References 1 USAR, Section 9.5 2 USAR, Section 9.5.1.2 2-39 Amendment No. JA,7J JpJ.JJA

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2.0 'llM111NG CONDITIONS FOR OPERATIONS 2.14 Enoineered Safety Features System Initiation Instrumentation Settinos (Continued)

(3) Containment Hioh Radiation (Air Monitoring) (Continued)

The setpoints for the isolation function will be calculatid in accordance with the ODCM.

Each channel is supplied from a separate instrument A.C. bus and each auxiliary relay requires power to operate. On failure of a single A.C. supply, the A and B matrices will assume a one-out of-two logic.

(4)LowSteamG_eneratorPressure A signal is provided upon sensing a low pressure in a steam generator to close the main steam isolation valves in order to minimize the temperature reduction in the reactor coolant system with resultant loss of water level and possible addition of reactivity. The setting of500psiaincludesgi22psiuncertaintyandwasthesettingusedin the safety analysis.

Closure of the MSIVs (and the bypass valves, along with main feedwater is accomplished by the steam generator isolation isolation andwhich signal bypass is a valves) logical combination of low steam gen pressure or high containment pressure.

As part of the AFW actuation logic, a separate signal is provided to terminate flow to a steam generator upon sensing a low pressure in that steam generator if the other steam generator pressure is greater than the pressure setting. This is done to minimize the temperature reduction in the reactor coolant system in the event of a main steamline break. The setting of 466.7 psia includes a +31.7 pst uncertainty; analysis. therefore, a setting of 435 psia was used in the safety (5)SIRWTanktowlevel level switches are provided on the SIRW tank to actuate the valves in the safety injection pump suction lines in such a manner so as to switch the water supply from the SIRW tank to the containment sump for a recirculation mode of operation after a period of approximately 24 minutes following a safety injection signal. The switchover point of 16 inches above tank bottom is set to prevent the pumps from running dry during the 10 seconds required to stroke the valves and to hold in reserve approximately 28,000 gallons of at leas 4 L900 ppm borated l water. The FSAR loss of coolant accident analysis 4/ assumed the recirculation started when the minimum usable volume of 283,000 gallons had been pumped from the tank.

2-62 Amendment No. 5,JJ,JJ,$J,5%,Jp),Jpg ,

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f.0 LIMITING CONDITIONS FOR OPERATION 2.19' U re Protection System t

i Applicability 4

Applies to fire detection and fire extinguishing subsystems in nuclear  ;

safety related areas. .

Ob.iective '

j To define the degree of operability of the fire protection system necessary to provide the capability for detecting, alarming, and extinguishing plant fires and to specify corrective actions required when operability requirements are not met.

Soecification (1) As a minimum, 50% of the fire detection instrumentation of zones shown in Table 2 7 (areas outside of containment) shall be I operable. With more than 50% inoperable detector (s) in a zone in safety related areas outside of the containment or with two adjacent detectors in a zone inoperable:

a. Within one hour, establish a fire watch patrol to inspect the zone with the inoperable instrument (s) at least once every hour, and b.

Restore within the inoperable 14 days. If the instrument instrument s are(s))to operable not restored to status

, operable status within 14 days, pre (pare and submit a report t

to the Nuclear Regulatory Commission, pursuant to Section 5.9.3 of the Technical Specifications, within an additional l 30 days, outlining the cause of the malfunction and the plans for returning the Instrument (s) to operable status.

(2) All but one (1) fire detection zone must be operable, with 50% or more of the detectors in a zone operable, in the containment building, except as of the Technical

)rovided Specifications. Witi more thanby one Section 2.19(3) inoperable fire rane:

a. Within one hour, establish a fire watch patrol to inspect the zone with more than 50% inoperable instrument (s) at least once per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> operating shift, and b.

I Restore within the inoperable 14 days. If the instrument instrument s are(s))to operable status not restored l to operable status within-14 days,(prepare and submit a report to the Nuclear Regulatory Commission, pursuant to Section 5.9.3 of the Technical Specifications, within an ( -

additional 30 days, outlining the cause of the malfunction and the plans for restoring the instrument (s) to operable status, t

2-89 Amendment No. R133

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I I.0 L]U111NG CONDITIONS FOR OPERATION 2.19 Fire Protection System (Continued) 4 (3) The provisions set forth in Section 2.19(2) do not apply to time periods during which Containment integrated Leak Rate Tests are '

being performed.

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(4) Fire suppressions water system shall be operable, except during  !

system testing, jockey sump maintenance or training (n5t to exceed i 7 consecutive days) witi both fire pumps, each witl) a minimum  ;

capacity of 1800 gpm, with their discharge aligned to the fire  ;

suppression header and automatic initiation logic for each fire pump,

a. With less than the above required equipment:

(1) restore the inoperable equipment to operable i status within 7 days, 4

(ii) if equipment is not restored to operable status '

within 7 days, prepare and submit a report to the i Nuclear Regulatory Commission, pursuant to Section  :

5.9.3 of tlie Technical Specifications, within the l l next 30 days, outlining the plans and procedures to be used to provide for the loss of the system and -

the cause of the malfunction.  :

I b. With no fire suppression water system operable:

l (1) establish a backup fire suppression water system a within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

I (ii) notify the Nuclear Regulatory Commission, pursuant to Section 5.9.2 of the Technical Specifications, outlining the cause of the malfunction, the actions taken, and the plans and schedule for restoring the system to. operable status.

(iii) If (1) above cannot be fulfilled, place reactor in Hot Standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in Cold Shutdown within the following thirty (30) hours.

(5) The sprinkler system in the Diesel Generator Rooms, the sprinklers above the steam driven auxiliary feedwater pump the sprinkler deluge sy/ stem spray in thenozzle system personnel in between corridor the compressor, fire areas 6room, and 20and the shall be operable except during system testing. If, inoperable:

a. Within one hour establish a continuous fire watch with backup fire extinguishing equipment for those areas in which redundant systems or components could be damagedi for other areas, establish an hourly fire watch patrol.

l 2-91 Amendment No. Jp,y ,pp,yp 133

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2.0 11MITING CONDITIONS FOR OPERATION i 2.19 Fire Protection System (Continued)

( b. Restore the system to operable status within 14 days or l' l

prepare and submit a report to the Nuclear Regulatory i

Commission, pursuant to Section 5.9.3 of the Technical t i Specifications, within an additional 30 days, outlining the l t 1

cause of the inoperability and the plans for restoring the system to operable status.  :

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l (6) The fire hose stations designated in Table 2-8 shall be operable.  !

With a hose station inoperable, provide a hose of equivalent capacity which can service the unprotected areas from an operable hose station within one hour from the time that a hose station is i determined to be inoperable if the inoperable fire hose station is the primary means of fire suppression; otherwise, route the >

additional hose within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

(7) All penetration fire barriers protecting safety-related areas shall be functional (intact). With a penetration fire barrier nonfunctional, within one hour, either establish a continuous fire watch on at least one side of the affected penetration, or verify the operability of fire detectors on at least one side of the penetration and establish an hourly fire watch patrol. Restore the nonfunctional senetration to functional status within 7 days, or prepare and suamit a report to the Nuclear Regulatory Commission,

• ' pursuant to Section 5.9.3 of the Technical Specifications, within

• an additional 30 days outlining the action taken, the cause, and the plans and schedule for restoring tN penetration to functional status.

l (8) The control room, switchgear room and cable spreading room halon systems shall be operable with the storage tanks having at least

) 90% of full charge pressure and 95% of full charge weight or I l evel . With a halon system inoperable, establish a cuntinuous fire watch with backup fire suppression equipment. Restore the system to o)erable status within 14 days,-or prepare and submit a report to t1e Nuclear Regulatory Commission, pursuant to Section 5.9.3 of the Technical Specifications, within 30 additional 30 days, outlining the cause of the inoperability and the plans for '

restoring the system to operable status.

EiLill The fire arotection system provides a means for detecting, alarming, and extinguisiing plant fires. The system is divided into the fire detection subsystem and fire extinguishing subsystem.-

The fire detection subsystem is an instrumentation system which alarms control room operators of a fire, indicating fire location on a panel in  ;

the control room and providing a local indication from the detector in i the affected zone.

2-92 Amendment No. 75, JJ,133 i

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r TABLE 3 4 MINIMUM FRE0VENCIES FOR SAMPLING TESTS Type of Measurement Sample and Analysis and Analysis Frecuency

1. Reactor Coolant *

(a) Power Operation (1) Gross Radioactivity 1 per 3 days (OperatingMode1) (Gammaemitters)

(2) Isotopic Analysis for (i) I per 14 days DOSE EQUIVAl.ENT 1 131 (ii) 1 per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (1) whenever the radioactivity exceeds 1.0 WC1/gm l DOSE EQUIVALENT I-131.

(iii)1 sample between 2 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following a thermal power change ;

exceeding 15% of the rated thermal poser i

's , change exceeding 15% l of tae rated thermal power within a 1-hour period.

(3)iDetermination 1 per 6 months (2)

(4) Dissolved oxygen 1 per 3 days and chloride (b)HotStandby (1) Gross Radioactivity 1 per 3 days-(OperatingMode2) (Gammaemitters)

Hot Shutdown (2) Isotopic analysis for (OperatingMode3) (1) I per hours (1)

DOSE EQU1 VALENT l-131 whenever the radio-activity exceeds 1.0u1 C1/gm DOSE EQUIVALENT I-131.

(ii) I sample between.2-8 hours following a '

thermal power change exceeding 15% of the rated thermal power within a 1-hour period.

(3) Dissolved oxygen 1 per 3 days and chloride .

Amendment No. JJ,57,JJA,133 3-18 l

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, , TABLE 34(Continued)

MINIMUM FRE0VENCIES FOR SAMPLING TEST a 1 Type of Measurement and Analysis

),

1. Reactor Ceolant  ?

(Continued) ,

(c)ColdShutdown (1) Chloride 1 per 3 days (OperatingMode4) .

1 (d) Refueling Shutdown (1) Chloride 1 per 3 days (3)

(Operating Mode 5) (2)BoronConcentration 1 per 3 days (3) l (e) Refueling Operation (1) Chloride 1 per 3 days (3) / -

l (2) Boron Concentration 1 per shift (3) / ,

! 2. SIRW Tank Boron Concentration 1 per 31 days

3. Concentrated Boric Acid Boron Concentration 1 per 31 days Tanks t

l 4. Si Tanks Boron Concentration 1 per 31 days T 5. Spent Fuel Pool Boron Concentration 1 per 31 days (1) Until the radioactivity of the reactor coolant is restored to i 19 C1/gm- /

DOSE EQUIVALENT l-131. ,

(2) Sample to be taken after a minimum of 2 EFPD and 20 days of power operation have elapsed since reactor was suberitical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or ,

longer, (3) Boron and Chloride sampling / analyses are not required when the core has been off loaded. Reinitiate boron and chlorirl sampling / analyses one I shift prior to reloading fuel into the cavit'- assure adequate shutdown /

margin and allowable chloride levels are met. -

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l Amendment No.15, 57. SE JJE.133 3-19 1

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4.0' DESIGN FEATURFJ 4.4 fuel Storaae ~

4.4.1 New Fuel Storaae The new unirradiated fuel bundles will normally be stored in the dry new fuel storage rack with an effective multiplication. factor of less than 0.9. The open grating floor below the rack and the covers above the -

racks, along with generous provision for drainage, precludes flooding of the new fuel storage rack.

New fuel may also be stored ~in shipping containers or in the spent fuel '

! pool racks which have a maximum effective multiplication' factor of 0.95-l with Fort Calhoun Type C fuel and unborated water.

L The new fuel storage racks are designed as ~a Class.I structure, i

4.4.2 Soent ' Fuel Storace >

Irradiated fuel bundles will be stored prior to off-site shipment in the stainless steel lined spent fuel pool. The spent fuel pool is normally filled with borated water'with a concentration of at least .

1900 ppm.

l The spent fuel racks are designed as a Class I~ structure.

Normally the spent fuel 2001 cooling system will maintain the bulk water temperature of the pool )elow 120*F. Under other conditions of fuel discharge, the fuel pool water temperature is. maintained below'140*F.

The spent fuel racks are designed and will be maintained such that the calculated effective multiplication factor is no greater than 0.95 (including all known uncertainties) assuming the pool .is flooded with unborated water. The racks are divided into 2 regions. Region I racks are surrounded by Boraflex; Region 2 racks have no poison. Acceptance criteria for fuel storage in. Regions 1 and 2 are delineated in Section 2.8 of the:e Technical Specifications.

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4-4 - Amendment No. JJ,fJ,7),Jp) ,133

S.9.3 S.pecial Reports Special reports shall be submitted to the Regional' Administrator of the appropriate NRC Regional Office within the time period specified for '

each report. These reports shall be submitted covering the activities "

identified below pursuant to the requirements of the applicable reference specification where appropriate: ,  ;

a. In. service inspection report, reference 3.3.

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b. Tendon surveillance, reference 3.5.

c. Containment structural tests, reference 3.5.

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d. Special maintenance reports,

e. Containment leak rate tests, reference 3.5.

f. Radioactive effluent releases, reference 2.9.

g. Materials radiation surveillance specimens reports, reference 3.3.

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h. Fire protection equipment outage, reference 2.19. :l

[ 1. Post-accident monitoring instrumentation,- reference 2' 21..

l 5.9.4 Uniaue Reportina Reauirements

a. Radioactive Effluent Release Renort A report covering the operation of the Fort Calhoun' Station d' uring' the previous six months shall be-submitted within 60 days after January 1 and July I of _each year per the requirements of 10 CFR 50.36a.

j The radioactive effluent release report shall' include a summary of the quantities of radioactive liquid and gaseous effluents and' solid waste released from the plant as outlined in Regulatory Guide' l.21, Revision 1.

The radioactive effluent release report shall include a summary of-the meteorological conditions concurrent with the release of '

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gaseous effluents during each quarter as outlined in~ Regulatory

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Guide 1.21, Revision 1.

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The radioactive effluent-release report shall' include an assessment i of radiation doses from the radioactive liquid and-gaseous - '

effluents released from the unit during each calencar quarter asL outlined in Regulatory Guide l.21, Revision 1. In addition, the:  !

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. unrestricted area boundary maximum. noble gas gamma air and beta air 1 doses shall be evaluated.: The meteorological conditions concurrent, with the 5-15 A'mendment No. J,7f,JS,p),5%,JJp,JJJ 133-l \

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i >. ,

O SPENT FUEL POOL REGION 2 STORAGE CRITERIA l '

Minimum Reovireo fuel Assemoly Exposure os o Function of l Initial Enrienment' to Permit Storage in' Region 2 l

40.0 3 37.5 2

N 35.0 - 1 I

32.5 Ac eptoble W/0 CEA j in Region 2 .

a 27.5 1 I ,e 1 1

25.0 -

, c. /

a- 0 22.5 / '

$ 20.0 17.5 Acceptoble WITH CEA u

l 15.0 in Region 2

.6 y 12.5 h 10.0 ,

e 7.5 /

5.0 NOT occeptoble

, in Region 2 2.5 ,

O.0 i 2.00 2.25 2.50 2.75 3.00 - 3.25 ~ 3.50. '3.75 4.00 4.25 4.50 Assembly Initial Enrichment, w/o U-235 1

'[

$ SPENT FUEL POOL REGt0N 2 OMAHA PUBLIC POWER DISTRICT. FIGURE '

STORAGE CRITERIA FORT CALHOUN STATION-UNIT No.1 2-10 h Amendment No. /6,133 ,

1

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