ML20151U387

Revised Bases of TS Sections 1.3(8),2.0.1(2),2.1.6,2.3,2.4, 2.13,2.15,3.1 & 3.6

ML20151U387
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Site:	Fort Calhoun
Issue date:	09/04/1998
From:	OMAHA PUBLIC POWER DISTRICT
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ML20151U381	List: LIC-98-0067, Forwards Updated Pages of TS Bases of Sections Listed & Requests That NRC Adminstratively Incorporate Pages Into Master Copy of FCS TSs & Reissue Pages ML20151U387
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NUDOCS 9809100375
Download: ML20151U387 (14)

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1.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 1.3 Limiting Safctv System Settings. Reactor Protective System (continued)

(7)

Containment High Pressure - A reactor trip on containment high pressure is provided to assure that the reactor is shut down simultaneously with the initiation of the safety injection system. The setting of this trip is identical to that of the l

containment high pressure signal which indicates safety injection system operation.

(8)

Axial Power Distribution - The axial power trip is provided to ensure that excessive axial peaking will not cause fuel damage. The Axial Shape Index is determined from the axially split excore detectors. The set point functions, shown in the COLR ensure that neither a DNBR of less than 1.18 nor a maximum linear heat rate of more than 22 kW/ft (deposited in the fuel) will exist as a consequence of axial power maldistributions. Allowances have been made for instrumentation inaccuracies and uncertainties associated with the excore symmetric offset - incore axial peaking relationship. A variance of 5% between iT-Power and NI-Power is permitted due to the significant margins to local power density limits before calibration of NI-Power is performed at 30% power.

(9)

Steam Generator Differential Pressure - The Asymmetric Steam Generator Transient Protection Trip Function (ASGTPTF) utilizes a trip on steam generator differential pressure to ensure that neither a DNBR ofless than 1.18 nor a peak linear heat rate of more than 22 kW/ft occurs as a result of the loss ofload to one steam generator.

(10)

Physics Testing at Low Power - During physics testing at power levels less than 104 % of rated power, the tests may require that the reactor be critical. For these tests only the low reactor coolant flow and thermal margin / low pressure trips may be bypassed below 104 % of rated power. Written test procedures which are approved by the Plant Review Committee will be in effect during these tests. At reactor power levels less than 104 % of rated power the low reactor coolant flow and the thermal margin / low pressure trips are not required to prevent fuel element thermal limits being exceeded. Both of these trips are bypassed using the same bypass switch. The low steam generator pressure trip is not required because the low steam generator pressure will not allow ~a severe reactor cooldown if a steam line break were to occur during the tests.

References (1) USAR, Section 14.1 (2) USAR, Section 7.2.3.3 (3) USAR, Section 7.2.3.2 (4) USAR, Section 3.6.6 (5) USAR, Section 14.6.2.2,14.6.4 (6) USAR, Section 14.7 (7) USAR, Section 7.2.3.1 (8) USAR, Section 3.6 (9) USAR, Section 14.10 l-9 Amendment No. 7.32,70,77,92,109,170

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9009100375 980904 PDR ADOCK 05000295:

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i 2.0 LIMITING CONDITIONS FOR OPERATION 2.0.1 General Requirements (Continued)

. function and have at least one normal and one emergency power source OPERABLE). If they are not satisfied, shutdown is required i

in accordance with this specification.

As a further example, Specification 2.7(1)b requires in part that two phydeally indepcadent circuits bc;weca the sffsitc transmission network and the casite Class IE distribution sysicm that both. House Service Transformers TI A-3 and TI A-4 be OPERABLE. The spccincation Specification ~ 2.7(2)c provides,a 24 72-hour out-of-service time when both required o!Tsite circuits House Service Transformers TI Aj3 and TI Aj4 are not OPERABLE provided the operability,of l

both Diesel Generators is'immediately verified. If the defmition of OPERABLE L

- were applied without consideration of Specification 2.0.l(2), all systems, subsystems, trains, components, and devices supplied by the inoperable normal powcr sourcca, both of the ofTsite circuits, House.S.ervice Transformers TI A-3 l

and Tl AM would also be inoperable. This would dictate invoking the applicable measures for each of the applicable LCO's. However, the provisions of Specification 2.0.l(2) permit the time limits for continued operation to be l

consistent with the corrective measures for the inoperable normal power sources instead, provided the other specified conditions are satisfied.-hahis casc, this wculd mcan that for onc division hc cmcrgcacy power sourcc must bc OPERACLE (as must bc the compencats supplicd by :hc cmcrgency powcr scurec) and all redundan: sys: cms, subsys: cms, trains, components, and devicca in the other division must bc OPERABLE, or likcw;se satisfy Spccifica:ica 2.0.!(2)

(i.c., be capabic of perfomiing their design func !cas and have an emergency power source OPERABLE). In other words, both emergency power sources must be OPERABLE and all redundant systems, subsystems, trains, components, and devices in both divisions must also be OPERABLE. If these conditions are not l

satisfied, shutdown is required in accordance with this specification.

1 in Operating Modes 4 or 5, Specification 2.0.l(2) is not applicable, and thus the l

individual requirements for each applicable Limiting Condition for Operation in these modes must be adhered to.

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4 Amendment No. 52 2-Ob I

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2.0 LIETING CONDITIONS FOR OPERATION 2.1, Reactor Coolant System (continued)

- 2.1.6 Pgssurizer and Main Steam Safety Valves (continued)

Action statements (5)b. and c. include the removal of power from a closed block valve to preclude any inadvertent opening of the block valve at a time the PORV may not be closed due to maintenance.

However,'the applicability requirements of the LCO to operate with the block valve (s) closed with power maintained to the block valve (s)are only intended to permit operation of the plant for a limited period of time not to exceed the next refueling shutdown (Mode 5), so that maintenance can be performed on the PORV(s) to eliminate the seat leakage condition.

To determine the maximum steam flow, the only other pressure relieving system assumed operational is the main steam safety valves. Conservative values for all systems parameters, delay times and core moderator coefficients are assumed. Overpressure protection is provided to portions of the reactor coolant system v.hich are at the nighest pressure considering pump head, flow pressure drops and elevation heads.

If no residual hest were removed by any of the means available, the amount of steam which could be generated at safety valve lift pressure would be less than half of the capacity of one safety valve. This specifiestion, therefore, provides adequate defense against overpressurization when the reactor is subcritical.

Performance of certain calibration and maintenance procedures on safety valves requires removal from the pressurizer. Should a safety valve be removed, either operability of the other safety valve or maintenance of at least one nozzle open to atmosphere will assure that sufficient relief capacity is available. Use of plastic or other similar material to prevent the entry of foreign material into the open nozzle will not be construed to violate the "open to atmosphere" provision, since the presence of this material would not significantly restrict the discharge of reactor coolant.

The total reliefcapacity of the ten main steam safety valves is 6.54 x 106 lb/hr. If, following testing, the as found setpoints are outside +/-1% of nominal nameplate values, the valves are set to within the

+/-J% tolerance. The main steam safety valves were analyzed for a total lose of main feedwater flow while operating at 1500 MWtm to ensure that the peak secondary pressure was less than 1100 psia, the ASME Section III upset pressure limit of 10% greater than the design pressure. At the power of 1500 MWt, sufficient relief valve capacity is available to prevent overpressurization of the steam system on loss-of-load conditions.W These analyses,are based on a minimum of four-of-five operable main steam safety valvespn each mainham_ header.

The power-operated relief valve low setpoint will be adjusted to pmvide sufficient margin, when used in conjunction with Technica: Specification Sections 2.1.1 and 2.3, to prevent the design basis pressure transients from causing an overpressurization incident.

Limitation of this requirement to scheduled cooldown ensures that, should emerger.y conditions dictate rapid caoldown of the reactor coolant system, inoperability of the low temperature overpressure protection system would not prove to be an inhibiting factor. The effective full flow area of an open PORV is 0.94 in,

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Removal of the reacar vassel head provides sufficient expansion volume to limit any of the design basis pressure transients. Thus, no additional relief capacity is required.

References (1) Article 9 of the 1968 ASME Boiler and Pressure Vessel Code,Section III (2) USAR, Section 14.9 (3) USAR, Section 14.10 (4) USAR, Sections 4.3.4,4.3.9.5 2-16 Amendment No. 39,47,54,146,161 9

2.0 LIMITING CONDITIONS FOR OPERATION 2.3 Emergency Core Coolino System (Continued)

(3)

Protection Against Low Temperature Overpressurization The following limiting conditions shall be applied during scheduled heatups and cooldowns. Disabling of the HPSI pumps need not be required if the RCS is vented through at least a 0.94 square inch or larger vent.

WWnever the reactor coolant system cold leg temperature is below 385 F, at least one (1) HPSI pump shall be disabled.

Whenever the reactor coolant system cold leg temperature is below 320 F, at least two (2) HPSI pumps shall be disabled.

Whenever the reactor coolant system cold leg temperature i:; below 270 F, all three (3) HPSI pumps shall be disabled.

In the event that no charging pumps are operable when the reactor coolant system cold leg temperature is below 270 F, a single HPSI pump may be made operable and -

utilized for boric acid injection to the core, with flow rate restricted to no greater than 120 gpm.

(4)

Trisodium Phosphate (TSP) Dodecahydrate During operating Modes 1 and 2, the TSP basket shall contain 2110 ft3 of active TSP.

a.

With the above TSP requirements not within limits, the TSP shall be restored within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

' b.

With Specification 2.3(4)a required action and completion time not met, the plant shall be in hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and cold shutdown ivithin the following 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Basis The normal procedure for starting the reactor is to first heal the reactor coolant to near operating temperature by running the reactor coolant pumps. The reactor is then made critical by v.ithdmaing CEA's and diluting baron in the resctor coolant. We this mode of start-up, thcEIhe energy stored in the reactor coolant during the approach to criticality is substantially equal to that during power operation and therefore all engineered safety features and auxiliary cooling systems are required to be fully operable.

2-22 Amendment No. 17,3",43,47,64,74,77, 100,103,133,14 l,157,1M,179

2.0 LIMITING CONDITIONS FOR OPERATION 2.3 Emergency Core Cooling System (Continued)

The SIRW tank contains a minimum of 283,000 gallons of usable water containing a boron concentration of at least the refueling boron concentration. This is sufficient boron concentration to provide a shutdown margin of 5%, including allowances for uncertainties, with all control rods withdrawn and a new core at a temperature of 60"F 68 F.m 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 used for the accident analyses. The minimum 116.2 inch level corresponds to a volume of 825 ft) and the maximum 128.1 inch level corresponds to a volume of 895.5 ft) 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.

The operable status of the various systems and components is to be demonstrated by 4

periodic tests. A large fraction of these tests will be performed while the reactor is operating in the power range.

If a component is found to be inoperable, it will be possible in most cases to effect repairs and restore the system to full operability within a relatively short time. For a single component to be inoperable does not negate the ability of the system to perform its ftmetion. Ifit develops that the inoperable component is not repaired within the specified allowable time period, or a second component in the same or related system is found to be inoperable, the reactor will initially be put in the hot shutdown condition to provide for reduction of cooling requirements after a postulated loss-of-coolant accident. This will i

also permit improved access for repairs in some cases. After a limited time in hot shutdown, if the malfunction (s) is not corrected, the reactor will be placed in the cold shutdown condition utilizing normal shutdown and cooldown procedures. In the cold shutdown condition, release of fission products or damage of the fuel elements is not considered possible, i

The plant operating procedures will require immediate action to effect repairs of an inoperable component and therefore in most cases repairs will be completed in less than the specified allowable repair times. The limiting time to repair are intended to assure that operability of the component will be restored promptly and yet allow sufficient time to effect repairs using safe and proper procedures.

The requirement for core cooling in case of postulated loss-of-coolant accident while in the hot shutdown condition is significantly reduced below the requirements for a postulated loss-of-coolant accident during power operation. Putting the reactor in the hot shutdown condition reduces the consequences of a loss-of-coolant accident and also allows more free access to some of the engineered safeguards components in order to effect repairs.

Failure to complete repairs within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of going to the hot shutdown condition is considered indicative of a requirement for major maintenance and, th refbre, in such a case, the reactor is to be put into the cc

• mdown condition.

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_J Amendment No. 32,39,47,49774,179

2.0 LIMITING CONDITIONS FOR OPERATION 2.3 Emergency Core Cooling System (Continurd)

With respect to the core cooling function, there is functional redundancy over most of the range of break sizes.*

The LOCA analysis confirms adequate core cooling for the break spectrum up to and including the 32 inch double-ended break assuming the safety injection capability which most adversely affects accident consequences and are defined as follows. The entire contents of all four safety injection tanks are assumed to be available for emergency core cooling, but the contents of one of the tanks is assumed to be lost through the reactor coolant system. In addition, of the three high-pressure safety injection pumps and the two low-pressure safety injection pumps, for both large break analysis and small break analysis it is assumed that two onc high-pressure pump and one low-pressure pump operate whde only one of each type is assumed to operate 4n-the sma!! break analysisP; and also that 25%

of their combined discharge rate is lost from the reactor coolant system out of the break. The transient hot spot fuel clad temperatures for the break sizes considered are shown in USAR Section 14.

The restriction on HPSI pump operability at low temperatures, in combination with the PORV setpoints ensure that the reactor vessel pressure-temperature limits would not be excceded in the case of an inadvertent actuation of the operable HPSI and charging pumps.

Removal of the reactor vessel head, one pressurizer safety valve, or one PORV provides sufficient expansion volume to limit any of the design basis pressure transients. Thus, no additional relief capacity is required.

Technical Specification 2.2(1) specifies that, when fuel is in the reactor, at least one flow 3

path shall be provided for boric acid injection to the core. Should boric acid injection become necessary, and no charging pumps are operable, operation of a single HPSI pump would provide the required flow path. The HPSI pump flow rate must be restricted to that of three charging pumps in order to minimize the consequences of a mass addition transient while at low temperatures.

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i Trisodium Phosphate (TSP) dodecahydrate is required to adjust the pH of the recirculation water to 2 7.0 after a icss of coolant accident (LOCA). This pH value is necessary to prevent significant amounts of iodine, released from fuel failures and dissolved in the

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recirculation water, from converting to a volatile form and evolving into the containment atmosphere. Higher levels of airborne iodine in the containment may increase the releases of radionuclides and the consequences of the accident. A pH of 2 7.0 is also necessary to prevent stress corrosion cracking (SCC) of austenitic stainless steel components in containment. SCC increases the pmbability of failures of components.

The required amount of TSP is represented in a volume quantity converted from the Reference 7 mass quantity using the manufactured density. Verification of this amount i

during surveillance testing utilizes the measured volume.

2-23a Amendment No. 39,47,64,74,77,100,179

' 2.0 LIMITING CONDITIONS FOR OPERATION 2.4 Containment Cooling (Continued) l l

to function during accident conditions may be inoperable for a period of no more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. If operability is not restored within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the reactor shall be placed in a hot shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

l Basis A full capacity diesel-generator is connected to each of the two engineered safeguards 4.16-kV buses. Three engineered safeguards 480-Volt double-ended load centers are provided; of te six transformers, three are connected to each of the two 4.16-kV buses.

Two load centers are operated as two-bus-section units; the third is provided with a i

center bus manually transferable to either associated end section. The center bus section supplies HPSI Pump SI-2C, CS Pump SI-3C and Charging Pump CH-lC any of which can thus be supplied from eitner 4.16-kV bus if required. Thrce compcaca cooling heat e= hanger have safScient capacity ic ren=e 102 x 10* BTU /hr fc!!cwing a isss-of-ecc! ant accident."1 The containment sprays initially take coolant from the safety injection and refueling water (SIRW) tank. Before this supply of water is exhausted (at least 24 minutes)m the spray system is transferred to the recirculation mode and the pumps take suction from the containment sump. One shutdown cooling heat exchanger is sufficient to satisfy the spray system requirements during the long-term containment cooling period?) In addition, in the unlikely event of the component cooling water supply being lost, raw water can be utilized for direct cooling of certain engineered safeguard components?)

The containment spray system is redundant with independent from the containment air recirculation, cooling and iodine removal system for the containment cooling function?)

The spray sysicm is sizcd such tha two For the limiting, Loss ofdoolant Accident l

(LOCA) scenario, on_e of the three spray pumps would limit the containment pressure to I

below the design value fc!! awing a DBA without taking credit for the air coolers or the cooling capacity of the safety injection system?H7) Similarly, two cooling and Shcring l-units or cac cooling and Shcring unit and both cooling units have the capability of l

l limiting thc containment prcasurc undcr thc samc conditions as two spray pumps. For the limiting Main Steam Line Break (MSLB) scenario, a heat removal contribution is credited from the air coolers in the mitigation of containment peak pressure. m Thc redundant cooling equipment provided to limit the containment pressure following l

a DBA is divided between the independent power supply systems. The raw water and i

component cooling water pumps are similarly distributed on the 4.16-kV and 480 Volt

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buses is scrve the above cccling groups. Each cccling group has a dcsign capacny cqual l

to that required is restrict :he containment pressurc to bclow the dcagn value. In the event of a DBA, loss of normal power sources and failure of one diesel-generator to operate, bct:ct than a. minimum of at least one full group spray pump, Land two air coolers would be connected to the available diesel-generator, thus providing morc than ampic l

This;would. provide adequate.. containment cooling _cquipment to limit.the reserve containment pressure below the d_esign value for the limiting one pump, one spray header LOCA event.yThe; limiting MSLB cvent in which'off site power is available,'is not affected by the loss ofone, diesel generator (Any onc unit rcmoved from a given bus do;s not restrict thc groups which can bc connccica to cac dicsci-gencrator fica fttiftlling

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2.0 LIMITING CONDITIONS FOR OPERATION 2.4 Containment Cooling (Continued)

.;;endby condidon. Dudag no.w.el power cgre: ion clier two air cooling and Slicring

. units or two air cooling uni s and orc air cooling r.nd SI cring unit arc in agra:;ca to r:mov: h::: !O:;; from piping 2".d eg!; men: ' i in the con: in=0n: * 'In additica, if dudag ic po;; ccident phase ic cornponca: cooling water supply is 103:, centaiarncnt cooling could be raaintained until repairs arc cn'ecicd. The component cooling system pumps and heat exchanger, the spray pumps and the shutdown heat exchangers are located in the auxiliary building?H* The raw water pumps are located in the intake

- structure.(")

Analyses shows that after a high load accident such as a large break LOCA or Main Steam Line Break inside containment, three in service component cooling heat exchangers will maintain CCW return temperature in an analyzed range. This assumes all of the containment air cooling units are ot - % w'hich could create the maximum heat load on the CCW system. In order to en3.,

.nat three heat exchangers would be in service after a DBA in conjunction with an assumed lingle failure, four are required to be operable.

If the river temperature is below 70 F, the single failure of a component cooling heat exchanger's RTV valve to open would not raise CCW return temperature to an unanalyzed level. A single failure of a RW valve to auto-open will raise CCW retum temperature due to CCW passing.through a heat exchanger which has no flow. If the river temperature is greater than or equal to 70 F, a single failure of a heat exchanger's RW valve to open could raise CCW retum temperature to an unanalyzed level. Therefore, when the river temperature is greater than or equal to 70 F and RW is not in service to a heat exchanger, the heat exchanger is considered inoperable. Having RW in service eliminates the potential failure of a RW valve to open as a credible single active failure.

When a component cooling heat exchanger is inoperable for maintenance and the river temperature is greater than or equal to 70 F, at least one RW and CCW isolation valve is secured in the closed position. This prevents the valve from opening upon receipt of i

a Safety Injection Actuation Signal, and therefore prevents CCW passing through a heat exchanger that has no RW flow. If the opening stroke time of a RW or CCW isolation valve on a heat exchanger is such that the valve is considered inoperable, the heat exchanger associated with the inoperable valve is still operable as long as the affected RW or CCW valve is maintained in the open position. The RW and CCW isolation valves may be opened or closed intermittently under administrative control without entering the LCO action statement in order to conduct required testing of components.

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' TSI 98-03 0 2-26 Amendrnent No.175

l 2.0 LIMITING CONDITIONS FOR OPERATION

' 2.4 Containment Cooling (Continued)

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i References (1) USAR, Scctica 9.7.5 Deleted l

(2) USAR, Section 6.2.3.1 (3) USAR, Section 6.2.3.4 (4) USAR, Section 9.8.2 l

(5) USAR, Section 6.4.5 (6) USAR, Section 6.3.5 (7) USAR, Section 14.16.5 (8) USAR, Sectica 9.10.2.3 Deleted

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(9) USAR, Section 9.7 l

(10) USAR, Section 6.3 l

(11) USAR, Section 9.8 i

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l-i 2-27 Amendment No.175

l 2.0 LIMITING CONDITIONS FOR OPERATION 2.13 ' Nuclear Detector Cooling System Applies to the operational status of the Nuclear Detector Cooling System.

Objective To assure that the concrete in the biological shield surrounding the reactor vessel does not l

overheet.

Specifications i

l (1)

The annulus exit temperature from the nuclear detector cooling system shall not exceed a temperature found to correlate to 150 F concrete temperature.

(2)

There will be at least 2 temperature detectors in service to measure annulus exit air temperatures whenever the reactor is in service.

Basis The Nuclear Detector Cooling System is used to cool the air in the annulus between the reactor vessel and the biological shield. While the nuclear detcetors can withstand temperatures consecr61y higher than 150 F the elevered temperature could result in reduction in concrete strength through loss af meMure. Each nuclear detector well cooling unit is rated at 100% af the system deJgn capability of 173,00 Btu /hr.m A test webe was performed during Hot Functionals ana/or low Power Tests to determine (1) the correlation between annulus air temperature and concrete temperature; (2) rate at which the concrete will heat up if no cooling is available.

The results of these tests webe were used to provide control room indication of concrete l

temperatures (that is annulus air temperature) and allowable reactor operation time in the event both nuclear detector well cooling units were inoperable.

The objective for this specification is to hold the concrete bulk temperature to less no greater than 150 F. The annulus exit temperature which correlated to 150 F (concrete temperature) is being wasfdetermined by testing. Temperature sensors are installed in the concrete and in the annulus air discharge. The sensors in the concrete-webe are subjected to neutron flux during operation and thus may not bc availabic later are no longer functional. The value determined for annulus exit temperatures which co'rrelate with concrete temperatures webe were determitied and a maximum value used to comply;with as the Tech Spec limit.

References (1) FSAR, Section 9.10.2.3 1

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2-60 1

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LIMITING CONDITIONS FOR OPERATION 2.15 Instrumentation and Control Systems (Continued)

Basis (Continued).

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. The engineered safety features system provides a 2 of 4 logic on the signals used to actuate the equipment connected to each of the two emergency diesel generator units.

' The rod block system automatically inhibits all CEA motion in the event a Limiting Condition for Operation (LCO) on CEA insertion, CEA deviation, CEA overlapor CEA sequencing is approached. Theinstallation of the rod block system ensures thatno single failure in the control element drive control

. system (other than a dropped CEA) can cause the CEAs to move such that the CEA insertion, deviation, sequencing or overlap limits are exceeded.

Accordingly, with the rod block system installed, only the dropped CEA event is considered an AOO and factored into the derivation of the Limiting Safety System Settings and Limiting Conditions for Operation. With the rod block function out-of-service several additional CEA deviation events must be considered as AOOs. Analysis of these incidents indicates that the single CEA withdrawal incident is the most limiting of these events. An analysis of the at-power single CEA withdrawal incident was performed for Fort Calhoun for various initial Group 4 insertions, and it has been

- concluded that the Limiting Conditions for Operation (LCO) and Limiting Safety System Settings (LSSS) are valid for a Group 4 insertion of less than or equal to 15%.

The operability of the Alternate Shutdown Panel (AI-185), including Wide Range Lcgarithmic Power and Source Range Monitors on AI-212, and Emergency Auxiliary Feedwater Panel (Al-179) instrument and control circuits ensures that sufficient capability is available to permit entry into and maintenance of the Hot Shutdown Mode from locations outside of the Control Room. This capability is required in the event that Control Room habitability is lost due to fire in the cable spreading room or Control Room.

Mariances which'may;e_xist at startup between the. move accuratelT-Power and Nuclear Instrumentation Power (NI-Power) are not significant for enabling of the trip functions.1By 15%lof rated power as measured by,the uncalibrated NI Power, the Axial PoweC Distribution (APD) and Loss,of Load (LOL); trip, functions are enabled lwhile the High Rate of Change of Power trip is bypassed.

ThelAPD trip function? acts to. limit the axial power s_hape to the range assumed in the setpoint analysis.]Significant margins to local power density limits exist at;15% power, as wel[as pop, levels up to at least 3_0% (where NI calibration occurs).

The1LOUtrip function acts as an anticipatory trip for the high pressurinr pressure and_ high power trips in order to limit the severity,of a LOLLtransient2 This trip).5 not. credited in the USAR Chapter 14 Safety. Analyses _and_anyyariance between AT,-Po)yer and NI-Power has no effect on the safety, analysis; i

Migh Rate.of Change of PoweritripLucts to limit _ power excursions from' low power i

jevels and bypassing of this trip at a high power level is.conservctive.jT1 is trip is not faedipan the,USAR Chapter 14 SafetyLAnalyses for Mode l' operation.' Any variance 1

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.betweenMPower and NI-PowerAas no effect on'the safety analysis, 2-66b Amendment No.125 l

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3.0 SURVEILLANCE REOUIREMENTS s

3.1 Instrumentation and Control (Continued)

• Substantial calibration shifts within a channel (essentially a channel failure) will be j

revealed during routine checking and testing procedures.

The minimum calibration frequencies of once-pci-day (heat balance adjustment only) for 2

the power range safety channels, and once each refueling shutdown for the process system channels, are considered adequate.

The minimum testing frequency for those instrument channels connected to the Reactor Protective System and Engineered Safety Features is based on ABB/CE probabilistic rLk analyses and the accumulation of specific operating history. The quarterly frequency for the channel functional tests for these systems is based on the analyses presented in the NRC approved topical report CEN-327-A, "RPS/ESFAS Extended Test Interval i

Evaluation,"

as supplemented, and OPPD's Engineering Analysis EA-FC-93-064, "RPS/ESF Functional Test Drift Analysis."

The low temperature setpoint power operated relief valve (PORV) CIIANNEL FUNCTIONAL TEST verifies operability of the actuation circuitry using the installed test switches. PORV actuation could depressurize the reactor coolant system and is not required.

t CalibrationjofjnstrumentTchannels withjresistance.: temperature detectorf(RTD)'.or thermocouple; sensors may consist of an inplace.qualitati.ve. assessment of sensor behavior and; normal calibration:of the remaining.adjustab.le, devices in_the' channel. Whenever a sensing element,is replacedLthe next required _ CHANNEL CALIBRATION _shall include anginplace cross;calibraticn that coinpares theLother. sensing elements.with the recently y

installed sensi.ng element:(The CHANNEL ~ CALIB_ RATION may be performed by means of.any series of sequential; overlappi,ng, or total channel. steps so that the entire channel is calibr. ate. d; 1

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3-2 Amendment No. M3,182

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3.0' SURVEH1ANCE REOUIREMENTS 3.6 Safety Iniection and Containment Cooling Systems Tests (Continued) that th: spray piping and nozzles are open will be made initially by a smoke test or other

' suitably sensitive method, and at appropriate intervals thereafter. A single containment spray header flow rate of M551885. gpm of atomized spray is required to provide the m

containment response specified in Section 2.4 of the Technical Specification: To achieve the M551885 gpm flow rate, no greater than ten (10) spray nozzles may be inoperable of which no more than 6ne may be missing. Since the material is all stainless steel, normally in a dry condition, with no plugging mechanism available, retesting at appropriate intervals is considered to be more than adequate.

Other systems that are also important to the emergency cooling function are the SI tanks, the component cooling system, the raw water system and the containment air coolers.

The SI tanks are a passive safeguard. In accordance with the specifications, the water volume and pressure in the SI tanks are checked periodically. The other systems mentioned operate when the reactor is in operation and are continuously monitored for satisfactory performance.

The in-containment air treatment system is designed to filter the containment building atmosphere during accident conditions. Both in-containment air treatment systems are designed to automatically c, tart upon accident signals. Should one system fail to start, the redundant system D designed to start automatically. Each of the two systems has 100 l

percent capacity.*

High efficiency particulate air (HEPA) filters are installed before the charcoal adsorbers l

to prevent clogging of the iodine adsorbers. The charcoal adsorbers are installed to reduce the potential release of radiciodine to the environment. The laboratory carbon sample test results should indicate a radioactive methyl iodide removal efficiency of at least 85 percent. If the efficiencies of the HEPA filters and charcoal adsorbers are as specified, the resulting doses will be less than the 10 CFR Part 100 guidelines for the accidents analyzed.

Pressure drop across the combined HEPA filters and charcoal adsorbers of less than 6 inches of water will indicate that the filters and adsorbers are not clogged by excessive l

amounts of foreign matter.

If significant painting, fire or chemical release occurs in a ventilation zone communicating with the system that could lead to the degradation of charcoal adsorbers or HEPA filters, testing will be performed to assure system integrity and performance.

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3-57 Amendment No. 15,121,171, 179 I

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