ML19347C085

Proposed Tech Specs,Sections 2.0,3.0 & 5.0 to Assure Control Isolation & Actuation Equipment Operability Per TMI-2 Lessons Learned Requirements.Safety Considerations, Discussion & Justification Encl

ML19347C085
Person / Time
Site:	Fort Calhoun
Issue date:	10/10/1980
From:	OMAHA PUBLIC POWER DISTRICT
To:
Shared Package
ML19347C076	List: ML19347C075 ML19347C080 ML19347C085
References
RTR-NUREG-0578, RTR-NUREG-578 NUDOCS 8010160555
Download: ML19347C085 (25)

Text

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2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 0 2.1.6 Pressurizer and Steam System Safety Valves Applicability Applies to the status of the pressurizer and steam system safety valves.

Objective To specify minimum requirements pertaining to the pressuriser and steam system safety valves.

Specifications To provide adequate overpressure protection for the reactor coolant system and steam system, the following safety valve requirements shall be met:

(1) The reactor shall not be made critical unless the two pressuriser safety valves are operable with their lift settings adjusted to ensur psia and 2545 psia 1,1%.ll)e valve opening between 2500 (2) Whenever there is fuel in the reactor, and the reactor vessel head is installed, a minimum of one operable safety valve shall be installed on the pressurizer.

However, when in at least the cold shutdown condition, safety valve nossles may be open to containment atmos-phere during performance of safety valve tests or mainte-nance to satisfy this specification.

(3) Whenever the reactor is in power operation, eight of the ten steam safety valves shall be operable with their lift settings between 1000 psia and 1050 psia with a tolerance of 1,1% of the nominal nameplate set point values.(1)

(h) Both pressurizer power-operated relief valves (PORV's) shall be operable during scheduled heatup and cooldown to prevent violation of the pressure-temperature limits designated by Figures 2-1A and 2-1B. One PORV may be inoperable for up to 7 days, provided the remaining PORV is operable. If the above conditions of this paragraph cannot be met, the primary system must be depressurised and vented.

(5) Two power-operated relier valves (PORV's) and their as-sociated block valves shall be operable in Modes 1, 2, and 3.

AmendmentNo.)[,LT 2-15 8 o 101erodi;$r;5r l

2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant Systen (Continued) 2.1.6 Pressurizer and Steam System Safety Valves (Continued)

a. With one or more FORV(s) inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORV(s) to operable status or close the associated block valve (s); otherwise, be in at least HOT STANDBY within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 2h baurs.

b. With one or more block valve (s) inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the block valve (s) to oper-able status or close the block valve (s). Other-vise, be in at least HOT STARD3Y within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SEUTDOWN within the following 2h hours.

Basis The highest reactor coolant system pressure reached in any of the accidents analyzed was 2h80 psia and resulted from a com-plete loss of turbine generator load %1 hgut simultaneous re-actor trip _while operating at 1500 M4t. 21 The reactor is assumed to trip on a "High Pressurizer Pressure" trip signal.

The power-operated relief valves (PORV's) operate to relieve RCS pressure below the setting of the pressurizer code safety valvt3. These relief valves have remotely operated block valves to provide a positive shutoff capability should a re-lief valve become inoperable. The electrical power for both the relief valves and the block valves is capable of being supplied from an emergency power source to ensure the ability to seal this possible RCS leakage path.

To futermine the maximum steam flow,'the only other pressure I relieving system assumed cperational is the steam system safety l valses. Conservative values for all systems parameters, de-lay times and core moderator, coefficients are assumed. Over- l pressure protection is provided to portions of the reactor

{

coolant system whichLare at the highest pressure considering l pump head, flow pressure drops and elevation heads.

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If no residual heat were removed by any of the means avail- l able, 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 specification, therefore, provides adequate defense against overpressurization when the reactor is suberitical.

2-lSa

2.0 LIMITING CONDITIONS FOR OPERATION 12.1 Reactor Coolant System (Continued) 2.1.6 Pressurizer and Steam System Safety Valves (Continued)

Performance of certain calibrution 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 avail-able. Use of plastic or other similar material to prevent the entry of foreign material into the open nozzle vill 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 re}ief caprcity of the ten steam system safety valves is 6.54 x 10 lb/hr. At the power of 1500 MWt, sufficient relief valve capacity is available to prevent overpressuriza-tion of the steam system on loss-of-load conditions.

The power-operated relief valve low setpoint will be adjusted to provide sufficient margin, when used in conjunction with Technical Specification Sections 2.1.1 and 2.3, to prevent the design basis pressure transients from causing an over-pressurization incident. Limitatior of this requirement to scheduled cooldown ensures that, should emergency conditions dictate rapid cooldown of the reactor coolant system, inoper-ability of the low temperature overpressure protection system would not prove to be an inhibiting factor.

Removal of the reactor vessel head provides sufficien- expan-sion volume to limit any of the design basis pressure tran-sients. Thus, no additional relief capacity is requi;ed.

References (1) Article 9 of the 1968 ASME Boiler and Pressure Ve tsel l Code,Section III I (2) FSAR, Section 14.9 (3) FSAR Sec. 'ns 4.3.h, 4.3.9.5

' Amendment No. M -47 2-16

2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.7 Pressurizer Operability Applicability Applies to tne status of the pressurizer and pressurizer heaters.

Objective Tc specify minimum requirements pertaining to the pressurizer water volume ar.1 availability of heaters for accident condi-tions.

! Specifications l

(1) The pressurizer shall be operable with at least 150 KW of pressurizer heaters , and pressurizer inventory shall be maintained in a range of level h0 5% to 69.2%.

. a. With the pressurizer inoperable due to an inoper-l able emergency power supply to the pressurizer j heaters either restore the inoperable emergency l power supply within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT SHUTDOWN within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. With the pressurizer l otherwise inoperable, be in HOT SHUTDOWN vithin l the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. This is applicable for i Modes 1 and 2

b. With the pressurizer level outside the above range,

( either' restore the level within the specified limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in HOT SHUTDOWN within the fol-l loving 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. This is applicable for Moden 1 and ,

2, except during mgnthly testing of the pressurizer l )

l level control circuit. i l I Basis l The requirement that 150 KW of pressurizer heaters and their associated controls be capable of being supplied electrical power from an energency bus provides assurance that these heaters can be energized during a loss of offsite power con- )

dition to maintain natural circulation at HOT STAUDBY. Either l

! diesel generator is equipped with 225 KW of heater capacity, i

[ Either diesel vill fulfill the minimum requirements of this i specification. The level should be maintained above the lower l limit to prevent heater cutoff and the upper limit should not

! be exceeded to prevent going solid or reducing the effective-ness of the pressurizer sprays by immersion during an RCS swell transier..

l 2-16a

2.0 LIMITING CONDITIONS FOR OPERATION 2.15 Instrumentation and Control systems Applicability Applies to plant instrumentation systems.

Ob.iective To delineate the conditions of the plant instrumentation and control systems necessary to assure reactor safety.

Specifications The operacility of the plant instrument and control systems shall be in accordance with Tables 2-2 through 2-6. l In the event the number of channels of a particular system in service falls below the limits given in the columns entitled "Ninimum Operable Channels" or " Minimum Degree of Redundancy",

except as conditioned by the column entitled " Permissible By-pass Conditions", the reactor shall be placed in a hot shut-down condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; however, operation can continue without containment isolation signals available if the ventila-tion isolation valves are closed. If minimum conditions tre not met 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 cold shutdown condition within 2h hours.

l If, during power operation, the rod block function of the second-l ary CEA position indication system and red block circuit are

! inoperable for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or the plant computer PDIL alarm, CEA group deviation alarm and the CEA sequencing function are inoperable for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the CEA's shall be with-l drawn and maintained at fully withdrawn and the control rod drive system mode switch shall be maintained in the off posi-i tion except when manual motion of CEA Group h is required to l control axial power distribution.

Basis During plant operation, the complete instrumentation systems will normally be in service. Reactor safety is provided by the reactor protection system, which automatically initiates appropriate action to prevent exceeding established limits.

Safety is not compromised, however, by continuing operation l with certain instrumentation channels out of service provi-l sions were made for this in the plant design. This specifi-cation outlines limiting conditions for operation necessary to preserve the effcetiveness of the reactor control and pro-tection system when any one or more of the channels are out of service.

All reactor protection and almost all engineered safety feature channels are supplied with sufficient redundancy to provide the capability for channel test at power, except for backup channels such as derived circuits in engineered safeguards control system.

AmendmentNo.[,.20 2-65

- TABLE 2-5 Instrumentation Operating Requirements for Other Safety Feature Functions Minimum Minimum Permissible Operable Degree of Bypass No. Functional Unit Channels Redundancy Conditions ,

1 UEA Position Indication 1 None None 3ystems 2 2 essurizer Level 1 None Not Applicable 3 Auxiliary Feedvater 1" None Not Applicable Flow / Steam Generator Level 4 Subcooling Margin 1 None Not Applicable Monitor 5 PORV Acoustic Positicn 1 None Not Applicable Indication-D! rect d

6 Safety Valve Acoustic 1 None Not Applicable Position Indication i

ec 7 PORV/ Safety Valve Tail l None Not Applicable Pipe Temperature NOTES:

" Auxiliary feedvater flow monitoring requirement of this specification is satisfied by one flow channel per pump or one of four level channels on each steam generator. Flow indication has one channel per pump and four level channels on each steam generator.

b One channel per valve.

c one RTD for both PORV's; two RTD's, one for each code safety.

d If item 7 is operable, requirements of specification 2.15 are modi-fied for items 5 and 6 to " Restore inoperable channels to operability within 7 days or be in hot shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />."

elf items 5 and 6 are operable, requirements of specification 2.15 are modified for item 7 to " Restore inopersble channels to oper-  ;

ability within 7 days or be in hot shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />."

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TABLE 2-6 Instrument Onerating Condition for Auxiliary Feedwater t

Minimum Minimum Permissible Operable Degree of Bypass No. Functional Unit Chonnels Redundancy Conditions

1. Auxiliary Feedwater A St'eam Generator Watel. 1 0 Eeactor coolant Level Lov less tha1 3000F.

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i 2-70a

TABLE 3-2 MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS AUD TESTING OF ENGIHEERED SA/ETY FEATURES, INSTRUMENTATION AND CONTROLS Surveillance Operating Mode Channel Description Function Frequency Required for Testing Surveillance Method

1. . Pressurizer Pressure Lov a. Check S 1, 2, or 3 a. Comparison of four separate pressure indications.

b. Calibrate R h or 5 b. Known pressure applied to sensors and PPLS actuation and blocking logic verified.

c. Test M(1)P 1, 2, or 3 c. Signal to meter relay adjusted with test device to trip one channel at a time.

[ 2. Pressurizer Lov Pressure a. Calibrate R 4 or 5 a. Part of 1(b) above.

Blocking Circuit 3 Safety Injection Actuation a. Test M 1, 2, or 3 a. Simulation of PPLS or CPBS 2/4 logic using built-in test-ing system. Both " standby power" and "no standby power" circuits will be tested for

.i A and B channels. Test will verify fanctioning of initi-ation circuits of all equip-4 ment normally operated by safety feature actuation signals.

b. Test R 5 b. Complete automatic test initiated sensor operation (Item 1(b) or h(t) and including all normal operation.

TABLE 3-2 (continued)

MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS AND TESTIUG OF ENGIHEERED SAFETY FEATURES, INSTRUMEUTATION AND CONTROLS Surveillance Operating Mode Channel Description Function Frecuency Reauired for Testing Surveillance Method

6. (continued) b. Calibrate R 1, 2, 3,14, b. Exposure to known external or 5 radiation source.

c. Test M 1, 2, 3, I 4, c. Remote operated integral or 5 radiation check source used to verify instrumentation, one channel at a time, and isolation lockout relay functional check.

y 7 Manual Safety Injection a. Test R 5 a. Manual initiation.

• Initiation

8. Manual Containment Isol- a. Test R 5 a. Manual initiation.

ation Initiation

b. Check R 5 b. Observe isolation valves closure.

9 Manual Initiation Con- a. Test R 5 a. Manual switch operation; tainment Spray pumps and valves terted

, separately.

es t:

c' P 25 10. Automatic Load Sequencers a. Test Q 1, -, or 3 a. Proper operation will be y3 verified during safety

'4 y feature actuation test of

. -a Item 3(a) above.

$ 11. Diesel Start a. Test M 1, 2, 3, or 14 a. Manual initiation followed y by synchronizing and loading.

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TABLE 3-2 (continued)

MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS AND TESTING OF ENGINEERED SAFETY FEATURES, INSTRUMENTATION AND CONTROLS Surveillance Operating Mode Channel Description Function Frequency Required for Testing Surveillance Method

22. Auxiliary Feedwater

a. Steam Generator Water a. Check S 1, 2, or 3 a. Compare independent level Level Low readings.

b. Test MP 1, 2, or 3 b. Functional check of initi-ation circuits.

c. Test R h or 5 c. System functional test of AFW initiation circuits.

Y

};j d. Calibration R h or 5 d. Known signal applied to P sensor.

S - Each Shift D - Daily M - Monthly Q - Quarterly R - 18 Months P - Prior to Each Start-Up if Hot Done Previous Week MP - Monthly during designated modes and prior to taking the reactor critical if not completed within the previous 31 days (not applicable to a fast trip recovery)

_k TABLE 3-3 (Continued)

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MINIt'UM FREQUENCIES FOR CHECKS, CALIBRATIONS AND TESTING f

g 0F MISCELLANEOUS INSTRUMENTATION AND CONTROLS f Surveillance Operating Mode j Channel Description Function Frequency Required for Testing Surveillance Method

.w 19 Auxiliary Feedvater Flow Check M 1, 2, or 3 Channel check.

(>300 F)

Calibrate R h or 5 Known pressure inputs.

20. Subcooling Margin Meter Check M 1, 2, or 3 Channel check.

Calibrate R h or 5 Known pressure inputs and known resistance substituted for RTD inputs.

Y ys 21. PCRV Operation and Acoustic Check M 1, 2, or 3 Channel check.

P Position Indication Calibrate R 4 or 5 Apply acoustic input.

Verify B 4 or 5 Operation on emergency power supply.

22. PCRV Block Valve Position Check Q 1, 2, or 3 Cycle valve.

Indication Calibrate R h or 5 Check valve stroke against limit switch position.

Verify R 4 or 5 Operibility on emergency power supply.

23. Safety Valve Acoustic Check M 1, 2, or 3 Circuit check.

Position Indication Calibrate R h or 5 Apply acoustic input.

24. PORV/ Surety Valve Tail Check M 1, 2, or 3 Circuit check.

Pipe Temperature Calibrate R 4 or 5 Apply known input.

TABLE 3-3 (continued) .

MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS AND TESTING OF MISCELLANECUS INSTRUMENTATION AND COUTROLS Surveillance Operating Mode Channel Description Function Frequency Required for Testing Surveillance Method

?

Q - Quarterly S - Eac; Shift D - Dai.

M - Men'hly A - Annually R 18 Months P - Prior to each startup if not performed within previous week.

PM - Prior to scheduled cold leg cooldown below 300 F; 0 monthly whenever temperature remains below y 3000F and. reactor vessel head is installed.

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IY in TABLE 3-5 (Continued)

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5 . Operating Mode FSAR Section n- Test Frequency Required for Testing Reference z- 10c. (Continued) k. Automatic ana/or Manual At least once per plant h or 5 H' initiation of the system operating cycle.

shall be denonstrated.

11. Containment Cool- 1. Demonstrate damper 1 year, 2 years, 5 4 or 5 9.10 ing and Iodine action. years, and every 5 Removal Fuseable years thereafter.

Linked Dampers

12. Fuel Elements Visually inspect fuel ele- During each refuel- 1, 2, 3, h , or 5 3 mt.nts removed from the re- ing outage.

actor.

Y o 13 Diesel Generator Calibrate During each refuel- 4 or 5 8.h.3

• Under-Voltage ing outage.

I Relays 1h. IJotor Operated Verify the contactor pick- During each refuel- h or 5 Cafety Injection up value at <85% of h60 V.

ing outage.

Loop Valve Motor Starters (iiCV-311, 31b, 317, 320, 327, 329, 331, 333, 372, 315, 318, 321) 15 Pressurizer Verify control circuits During each refuel- 4 or 5 IIeaters operation for post- ing outage.

accident heater use.

3.0 SURVEILLANCE REQUIREMENTS 39 Auxiliary Feedwater Systen Applicability Applies to periodic testing requirements of the turbine-driven and motor-driven auxiliary feedwater pumps.

Objective To verify the operability of the auxiliary feedwater (AFW) system and its ability to respond properly when required.

Specifications (1) The position of valves necessary to ensure auxiliary feedvater flow to the steam generators shall be veri-fled by a monthly inspection. Anytime maintenance is ,

performed on the auxiliary feedwater system which alters )

valve alignments, an operator shall check that the AFW system valves are properly aligned, to ensure AFW flow 1 to the steam generate rs , and a second operator shall independently verify proper valve alignment. Operating )

Mode Required for Testing: 1, 2, 3, or k. j (2) The operability of the motor-driven auxiliary feedwater i pump, the steam turbine-driven auxiliary feedwater pump, I and the auxiliary feedwater pumps' steam generator level regulating valves HCV-1107A, HCV-1107B, HCV-1108A, HCV-1108E, and auxiliary feedwater cross-tie valve HCV-138h shall be confirmed at least every three months. Operating Mode Required for Testing: 1, 2, or 3.

(3) The capabilities of the motor-driven and turbine-driven

! auxiliary feedwater pumps shall be verified by using local pressure indicators and flow indicators in the centrol room. The discharge pressure will be verified to be h0 psig above the steam 6enerator pressure at rated steam

, flow. Operating Mode Required for Testing: 1, 2, or 3.

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.(b) Following cold shutdown and prior to raising the reactor coolant temperature above 3000F, the motor-driven auxi-liary feedwater pump shall be tested to verify the nor-mal flow path for auxiliary feedwater to the steam gener-ators.

(S) At least once per 18 months during shutdown by:

a. Verifying that each automatic valve in the flow path actuates to its correct position upon receipt of each auxiliary feedwater actuation test signal.

b. Verifying that each auxiliary feedwater pump starts as designed automatically upon receipt of each auxi-liary feedwater actuation test signal. Operating Mode Required for Testi7g: h or 5 Amendment No. 49 3-62

3.0 SURVEILLANCE REQUIRE!ENTS 3.9 Auxiliary Feedwater System Applicability Applies to periodic testing requirements of the turbine-driven and motor-driven auxiliary feedvater pumps.

Objective To verify- the operability of the auxiliary feedwater (AFW) system and its ability to respond properly when required.

Specifications (1) The position of valves necessary to ensure auxiliary feedwater flow to the steam generators shall be veri-fled by a monthly inspection. Anytime maintenance is performed on the auxiliary feedwater system which alters valve alignments, an operator shall check that the AFW system valves are properly aligned, to ensure AFW flow to the steam generators, and a second operator shall independently verify proper valve alignment.

(2) The operability of the motor-driven auxiliary feedwater pump, the steam turbine-driven auxiliary feedwater pump, and the auxiliary feedvater pumps' steam generator level regulating valves HCV-1107A, HCV-1207B,- HCV-1108A, HCV-1108B, and auxiliary feedwater cross-tie valve HCV-138h shall be confirmed at least every three months.

(3) The capabilities of the motor-driven and turbine-driven auxiliary feedwater pumps shall be verified by using local pressure indicators and flow indicators in the control room. The discharge pressure will be verified to be 40 psig above the steam generator pressure at rated steam flow.

(h) Following cold shutdown and prior to raising the reactor coolant temperature above 3000F, the motor-driven auxi-liary feedwater pump shall be tested to verify the nor-mal flow path for auxiliary feedwater to the steam gener-ators.

(5) At least once per 18 months during shutdown by:

a. Verifying that each automatic valve in the flow path actuates to its correct position upon receipt of each auxiliary feedwater actuation test signal.

b. Verifying that each auxiliary feedwater pump starts as designed automatically upon receipt of each nuxi-liary feedwater netuation test signal.

3-62 Amendment No. 49

3.O SURVEILLANCE REQUIREMENTS 3.9 Auxiliary Feedwater System (Continued)

. Basis The valve position verifications performed monthly and fol-l lowind auxiliary feedwater system maintenance will confirm the availability of an auxiliary feedwater flow path to the steam generators.

The testing every three months and after cold shutdowns of the auxiliary feedvater pumps will verify their operability by recirculating water to the emergency feedwater storage tank and operating, one at a time, the regulating valves (HCV-1107B and HCV-1108B) to confirm a flow path to the steam generators and operability of the valves.

Proper functioning of the steam turbine admission valve and starting of the feedwater pump will demonstrate the integrity )

of the steam driven pump. Verification of correct operation  :

will be made both from instrumentation within the main con- )

trol room and direct visual observation of the pumps.

]

The operability of the auxiliary feedwater system ensures

! that the reactor. coolant system can be cooled down to less

, than 3500F from normal operating conditions in the event of l a total loss of off-site power.

l References l (1) FSAR, Section 9.4 l

(2) Technical Specification 2.5 l

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Amendment No. 49 3-62a

5.0 ADMINISTRATIVE CONTROLS

-5.3 Facility Staff Qualification 5.3.1 Each member of the plant staff shall meet or exceed the mini-num qualifications of ANSI N18.1-1971 for comparable positions, with the exception of the Supervisor - Chemical and Radiation Protection (SCRP) and the Shift Technical Advisor (STA). The SCRP shall meet the requirements set forth in Regulatory Guide 1.8 dated September, 1975, entitled " Personnel Selection and Training". The SCRP is considered to meet the educational and experience qualifications set forth in Regulatory Guide 1.8 with at least five years of experience in applied radi-ation protection and extensive formal training in radiation protection. The Shift Technical Advisor shall have a bachelor's dcgree or equivalent in a scientific or engineering discipline with specific training in plant design and response and analysis of the plant for transients and accidents.

e Amendment No. 38 5-la

TABLE 5.2-1

[1INIMT1 SHIFT CRE'd COMPOSITION License Cold Shutdown or Operating or Category Core Alteration Refueling Shutdown Hot Shutdown Modes Senior.

Operator License 1* 1 1 Operator License 2 1 2 Hon-Licensed (As required) 1 2 Shift Technical Hone None 1 Advisor 1

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• Does not include individual with Senior Operator License supervising l Refueling Operations. l l

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Amendment No. /,'24 5-2

6.0 IUTERIM SPECIAL TECICTICAL SPECIFICATIONS 6.5 Auxiliary Feedwater Automatic Initiation Setpoint Apulicability This specification applies to category A Lassons Learned Control Grade Auxiliary Feedwater Automatic Initiation System.

Objective To define the steam generator level range and time delay band in which automatic initiation of auxiliary feedwater must occur.

Specification Functional Unit Channel Setting Limits Steam Generator Auto Initiation 0% to 31.2% level with Level (Downcomer of Auxiliary time delay greater than level) and Time Feedwater or equal to 180 sec.

Delay Basis The level setpoint is to ensure autortatic initiation of AFW in j case of loss of the main feedwater (MFW) system. The range is provided to minimize initiation during operational occurrences with IJW available. The time ~ delay is provided to ensure that the FtCS is not overcooled during a main steam line break (MSLB) event.

6-5

DISCUSSION The Omaha Public Power District received a letter from the Com-

, mission, dated July 2,1980, requesting Technical Specification (T.S. )

changes. The T.S. changes are required to assure that plant operations are maintained within acceptable limits following the implementation of TMI-2 Lessons Learned Category "A" items. The specification proposals requested in the Commission's July 2, 1980, letter are listed below wit 2.

the Commission's- specific requests.

(1) Emercency Power Supply Reauirements The pressurizer water level indicators, pressurizer relief and block valves, and pressurizer heaters are important in a post-accident situation.

Adequate emergency power supplies insure post-accident functioning of these components. The enclosed specifications will satisfy our require-ments. Removal of the PORV block valve power supplies is not necessary.

No automatic opening signals are installed in the cor?*ol circuit.

(2) Valve Position Indication The installed system for indication of valve position is a diagnostic aid to the operator. Although the indicating system provides no auto-matic action, we believe that this system should be operable and that periodic surveillance should be performed.

(3) Instrumentation for Inadequate Core Cooline

(h) Containment Isolation We believe your specifications should include a Table of Contain-ment Isolation Valves which reflect the diverse isolation signals which your design currently provides. Sample specifica; ions and associated surveillance are included.

(5) Auxiliary Feedvater Systems Setpoint rance is consistent with the limits assumed in the safety analysis response to IE Eulletin 80-Oh,

(()) Shift Technical Advisor The specification related tq minimum shift manning should be re-vised _to reflect the augmentatior of a Shift Technical Advisor.

.Each specific proposal is discussed separately.

(1) Emergency Power Supply Recuirements The pressurizer water level indicators, pressuriner power operated relief and block valves, and h50 KW (225 KW per diesel) of pressurizer heaters are not powered from emer,3ency buses. Operability requirements for the_ level indicators are adequately provided for in proposed changes to T.G. 2.1.5, Table 2-5, an'd T.S. 2.0.1 in the Application for Amendment filed with the Commission on August 5, 1980. Surveillance requirements ATTACE'!ENT B

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)

I (1) Enerrency Power Supply Reauirements (Continued)

)

for the pressurizer level indicator are provided for in existing T.S. .

3.1 and Table 3-3. PORV and block valve operability is provided for  !

in the proposed T.S. 2.1.6 and surveillance is provided for in the i proposed Table 3-5 Pressurizer heater and inventory operability )

requirements are detailed in the proposed new Limiting Conditions l for Operation (LCO's), T.S. 2.1.7, and surveillance requirements in l Table 3-5 Since the reactor is suberitical, the pressurizer heaters  ;

and inventory are not required for Mode 3. The proposed inventories are those ranges assumed in the safety analysis. i In all cases, the proposed T.S. changes are consistent with the standard T.S. provided by the Commission as modified to conform in scope and content with Fort Calhoun's existing specifications.

(2) Valve Position Indication j l

LCO's and surveillance requirements for the safety and power operated l relief and block valve position indicators are proposed to assure accurate i status tracking of these valves. Revised T.S. 2.15, Table 2-5, and Table 1 3-3 assure the operability of the position indicators. Since the block

)

valves have no automatic closure, a quarterly check of the limit switch position indication along with valve operability is adequate. This also minimizes the time in which the valves are closed during power operation. It is felt that the 7 day LCO for valve indication by either of the systems of PORV/cafety valve is adequate due to the short i 7 day position indication exposure. I l

(3) Inatrumentation for Inadeauate Core Coolinz The aubcooling margin system provides an indication of the re-actor's approach to saturation conditions. In an accident situation, l I

this indication vill provide an important source of information to the l operator in 11aking his decisions. Accordingly, the accuracy and oper- I ability of the subcooling margin system is assured through the pro- ,

posed LCO's and surveillance requirements in T.S. 2.15, Table 2-5, l and Table 3-3. j l

(h) Containment Isolation I TMI-2_ related requirements required no changer to the existing con-tainment isolation actuation system or existing hardware. The CIAS is initiated by two diverce signals: low pressurizer pressure or high con-tainment pressure. All containment isolation valves assume their acci-dent position on actuatico and in all cases can only be manually reset.

Some valves have an override function such that the valve position can l be changed while the actuat~on signal is present; however, this is a manual function. Existing T.S. assure operability and accuracy of the CIAS and the containment isolation valves, and no changes to these T.S.

are proposed.

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l l (5) Auxiliary Feedvater Systems l

,.. As a result of the TMI-2 Lessons Learned requirements, the auxiliary l feedvater ( AFW) system was modified to actuate automatically. As e re-l sult of the AFW changes, changes to T.S. 2.15, 3 9, Table 3-2, and addition of Table ,2 6 are proposed to assure the operability of the AFW system.

The operability of the automated AFW system is critical to assuring an adequate heat sink during the initial stages of an accident similar to the TMI-2 event. For the purpose of testing, it is considered adequate to test the steam admission valves to the steam driven feed pump during the auto initiation of auxiliary feedvater circuit testing. The pump operability will be demonstrated prior to a return to power operation l following refueling.

(6) Shift Technical Advisor l

l T.S. 5 3 and Table 5.2-1 have been revised to include the require-ment for a Shift Technical Advisor (STA) on the operating shifts. The proposed T.S. comply with the standard T.S. provided with the Commission's letter _of July 2, 1980.

In addition to the T.S. changes required, the Commissicn in their July 2, 1980, letter reconmended that conditions be added to the Facility License for leak monitoring and upgraded iodine monitoring. The Facility License has been amended to add license conditions requiring programs to reduce leakage of radioactive systems outside containment and to upgrade iodine monitoring during accident conditions. The proposed conditions comply with the suggested format provided with the Commission's letter.

l All proposed Facility License and Technical Specification revisions l do not constitute an unreviewed safety question nor do they represent a l 'possible threat to the public. The changes provide procedures to assure l operability of systems used in mitigation of transients and possible

[ accidents, thereby improving the safety of the Fort Calhoun Station.

l The proposed T.S. revisions are also consistent with the Combustion l Engineering Standard Technical Specifications as modified to conform l vith the scope and content of existing T.S.

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E JUSTIFICATION FOR FEE CLASSIFICATION The proposed amendments are deemed to be Class III within the meaning of 10 CFR 170.22 because they have been identified acceptable by Comission positions. The Commission has identified the need and format for the proposed amendments by a letter dated July 2, 1980.

ATTACICENT C

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