ML20004C376

Proposed Revisions to Tech Spec Sections 2.1.3,2.7,2.10.4, 4.3.2,5.5.2 & 6.4

ML20004C376
Person / Time
Site:	Fort Calhoun
Issue date:	06/02/1981
From:	OMAHA PUBLIC POWER DISTRICT
To:
Shared Package
ML20004C374	List: ML20004C373 ML20004C375 ML20004C376
References
NUDOCS 8106030289
Download: ML20004C376 (23)

Text

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• % c f' ABLE OF CONTENTS (Continu:d)

V Page

-3.0) LSURVEILLANCE REQUIREMENTS .....................................'........ 3 3 1'. Instrumentation and. Control ..................................... 3-1, 3.2 Equipment:and Sampling Tests ................................... 3-17

' 33, Resetor. Coolant System, Steam Generator Tubes, and Other Components Subject to'ASME XI Boiler & Pressure Vessel Code Inspection and Testing Surveillance .....................-3-21 3.h. Reactor Coolant System Integrity Testing:....................... 3-36 3.5 - Cont ainme nt Te st . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-37

. 3.6- Safety Injection and containment Coolins. Systems Tests ......... 3-Sh~

3.7 l Bnergency Power System Periodic Tests .' . . . . . . . . . . . . . . . . . . . . . . . . . 3-58

- 3.8 Main Ste mn Isolation - Valve s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ' 3 39' Auxiliary Feedvater System ..................................... 3-62:

3.10 Reactc' Core Parameters ........................................ 3-63 3 11 Environ 2 ental Radiological Monitoring ........................... 3-6h 3.12 . Radioactive Materials .......................................... 3-69 3 13- Radioactive Material Sources Surveillance ...................... 23-76 3 1h Shock Suppressors (Snubbers) ................................... 3-77 3.15 Fire Protection System ......................................... 3-80=

4.0' - DESIGN FEATURES ....................................................... 4-l'

- 4.1 Site ............................................................ 4-1.

h.2 Containment - De sign Fe ature s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . h-1 h.2.1' Containment Structure ................................. 4-1 4.2.2 Penetrations ........................................... 4-1 h.2.3 Containment Structure Cocling Systems .................. h-2 h.3 Nuclear Steam Supply System ( NSSS) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-3 h.3.1 Re actor Coolant System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - h- 3 h.3.2 Re actor Core and Control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-3 h.3 3 Emergency Core Cooling ................................ 4-3 h.4 Fuel Storage .......................................... .... 4-h 4.4.1 New Fuel Storage .............................. ....... h-4 h.h.2 Spent Fuel Storage .................................... 4 h 4.5 Seismic Design for Class I Systems ............................. 4-5 50 ADMINISTRATIVE CONTROLS ............................................... 5-1 5.1 Responsibility.................................................. 5-1 52 Organization .......................... ........................ 5-1:

5.3 Facility Staff Qualifications .................................. 5-la 5.4 Trainins ....................................................... 5-3 55 Revi e w and Aud it . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-3 5 5.1 Plant Review Connittee (PRC ) . . . . . . . . . . . . . . . . . . . . . . . . . . 5-3 552 Safety Audit and Review Committee (SARC) . . . . . . . . . . . . . . 5-5 553 Fire Protection Inspection ................... ........ 5-8 l 5.C Beportable Occurrence Action ................................... 5-8 5.7 Safety Limit Violation ......................................... 5-9 5.8 Procedures ..................................................... 5-9

' Amendment No. 35, X3, K6, 5h 11 A'I'fACFSENT A 8106 03 0j(j$d

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2.0 -LIMITIN3' CONDITIONS FOR OPERATION 2.li ; Reactor Coolant System (Continued)' .

'2.1 3: 7 Maximum-Reactor. Coolant Radioactivity'(Continued):

tube ~ rupture accident.(3) The potential dose at the. site boundary' '

. , ffor this' accident is. larger an'd'hence more-limiting than the--

i dose that would result from.'one' year of operation with the maximum reactor coolant activity' combined-with.the maximum-

-permissible unidentified leakage from the reactor coolant-system-

. '(Section 2.1.h-of'theseLSp'ecifications)..

~ The accidents -for which primary and:: secondary ' coolant ~ concen-

.trations.become the limiting parameters are a^ steam generator'

. tube rupture. and a steam line. break tpstream of< the isolatiori

. valves,'respectively, with the following basicLassumptions:

~

(1) 1Stean Generator Tube Rupture 1 a.. Maximum permissible two-hour doses at the: exclusion:-

distance of 1.5. rem to. the thyroid, and :0.5: rem whole

body.. ,

b. Loss. of offsite 'pover.

• L c ', 90,000.lbs.1 of primary coolant.is: released to the secondary. .

d. Identification of the accident and pressure equali-zation=between primary'and secondary occurs within 30.

minutes.

e. -Ten percent of the primary coolant entering the secondary. system through the ruptured tube flashes and:is-released directly to.the environs.

f. The quantity of steam released from the secondary.

system is' determined from mass and energy balances on-the primary and secondary systems.

gs Partitioning'of iodine in the. steam released from both steam generators by a factor of 0.1.

(2) -Steam Line Break

a. Maximum permissible two-hour doses at the exclusion distance of 1.5 rem to the thyroid, and 0.5 rem whole body.

~

b. Loss of offsite power.

c. Blowdown of the contents of one steam generator.

T 2-9

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L L t2.0'iLIMITING! CONDITIONS FOR OPERATION

2.7.; Electrical-Systems (Continued). n

g. . One of the:four a-e i'nstrument; buses may be ~ inoperable for

18. hours provided the-reactor, protective and engineered safe -

-guards: systems instrument channels, supplied by the remaining 1

.three buses are-all' operable.-

Two-battery chargers'may be inoperable for up to'8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> pro-

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vided battery charger lNo.1. or No. '2 is -operable.

~

.i. JEither one=of.the diesel generators may be inoperable for up

to seven days (total for both) during any month, provided the' -

other diesel'is started to verify operability,Lshutdown and controls are left in the automatic mode and'there are no in-

. operable' engineered safeguards ~ components associated with the

~

operable diesel generator.

.j. Island buses 1B3A hA,.1B3B hD,~and 1B3C-iC may be' inoperable.

for.up to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />' provided there are no inoperable ~ safeguards.

components associated with the-operable bus which are re-

dundant to the inoperable bus.

' k. .Either one of the DC buses (Panels EE-8F and EE-8G) may be.

inoperable for up:to'8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

-1. Either one of the DC Distribution Panels'AI 41A and AI-41B may-be inoperable for up to 8' hours.

m. AC Instrument Panel AI-42A or AI-42B may be inoperable for-up to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

n. The 161 kV transmission line may be out of service and unit operation may continue or the reactor may be restarted from a hot shutdown coridition if (i) operability of the remain-ing source is immediately verified and (ii) immediate noti-

+

fication is made by telephone or telegraph to the Director of the'NRC Regional Compliance office in Arlington, Texas l of. the loss and of the plans to restore the electric power system to its full capability.

Basis The electrical system equipment is arranged so that no single failure can-inactivate enough engineered safeguards to jeopardize the plant safety. The 480 V safeguards are arranged on nine bus sections. The 4.16 kV safeguards are supplied from two buses.

The normal source of auxiliary power with the plant at power for the safeguards buses. is from the house service power transfc,rmers being

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. fed from the 161 kV incoming line with on-site emergency power 2-34

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L g.i.  : DNB~ MARGIN ^-. * >

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.b et - - SPECIFIED' OPERATING LIMITS n .

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- OPERATING LIMIT -

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Monitored- ..

p .- Parameter. :h Pump'

, f Cold Ieg . Temperature <(545 F)*--  :.--

Pressurizer Pressure >(2075 c psia)* ' .

m <

Reactor Coolant Flow j >(195,700 gpm)**l

_ s _.

' Axial Shape'Index <(Figure 2-7). ,

to 4 '

h%

• Limit not applicable during either.a thermal power.. ramp.lnerease in excess.of.~5%.of rated thermal pove'r.-

. per minute or .a' thermal power step increas'e of greater than'10f of.' rated -thermal power. ,

• • This number is :an _ actual limit (not , including uncertainties). . All otherivalues in this. table are-Iindi--

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cated values-and include an allowance for measurement uncertainty (e.g.,-5450F, indicated, allow's'fc ,

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L an actual Te of 547 9).

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TABLE 2-2 L Instrument Operating Requirements for Reactor' Protecti re'~ Sys'temi

-Minimum. Minimum : Permissible

0perable _ ' Degree of- . Bypass 1 Functional' Unit

Channels.

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No '. Redundancy- Condition

!1: Manual'::(Trip - httons) ! '.1 'None None

2. High Power'. Level.L 2(bc) 1(c) - 'ThermalPowerInpug"'

Bypassed beloz 10-

-of Rated Power (a);(d) 3- 1 Thermal Margin /Lov: L2(b):. ;1 ~

-3elov 10-4;(a (d)

Pressurizer Pressure Rated Power

-h High Pressurizer Pressure 2(b) . 11 None .

5.

• Low R.C. Flov 2(b)? 'l E=.'_ov 10 hp(o l(d)

Rated Power a

'. 6 ^ Low Steam Generator 2/S ep 1/ Steam None Water Leve1~ Gen bi Gen

7.  : Low Steam Gentrator 2/S e 1/ Steam -Belov 550 psia (a) (d)

Pressure- Gen b Gens

8. 2(b) .During' Leak Test

- - Containment High - 1-Pressure 9: Axial Power Distribution 2(bc) 1(c) Below 15% o .

Rated Power 8).

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High Rate Trip-Wide Belov 10-b % and abov

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10- 2 1 Range Log Channels 15%ofrate'dpover(a$(eN

-11 Loss of Load 2(D) 1 Below  % of rated power l

a. Bypass autome.tically removed.

b .One of the inoperable channels must be in the tripped condition.

. c' If'two channels are inoperable, load shall be reduced to 70% or less of rat ed power ,

d 4

'For. low power physics testing this trip mcy.be bypassed up to 10-15 of rated-power.

e For.:each channel, the same bistable automatically activates.the Loss of Load and Axial Power Distribution (APD) trips and automatically bypasses

-the high rate-trip at 15% of rated power. Only the APD trip is a

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Limiting Safety' System Setting. Therefore, the bistable is set to actuate within the APD tolerance band.

2-67 Change No. 7 February 28, 1974

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TABLE ' 2 -7 n -

, . FIRE DETECTION ZONES' -

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S' 'Zonef .. .

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. No.~ Locatiori

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' Auxiliary Building, Elevations 971 and 989 West

• 2I Auxiliary Building, Elevation 989 East . . .

3 xAuxiliaryl Building, Elevation 989, Lower Electrical ~

Penetration ~ Room (Room 20)z

h. . Auxiliary Building, Elevation'989,1 Air Compressor L -

~ Room (Room 19)'

'5 ' Auxiliary Building, Elevation 1007, Corridor 26, Rooms

.58, 59',-and 60 6.2 -

~ Auxiliary Building, Elevations 1007 and'1011, Uncontrolled

7. Auxiliary Building, Elevation 1013,_ Upper Electrical

' Penetration Room (Room 57)

'8 . Auxiliary Building, Elevations 989 and 1007, Boric Acid Tank Area, Drumming Area, dev Fuel Area 9 Auxiliary Building,' Elevation 1036,_ Control Room Complex,

. Control Room Hallways

' 10 .  : Containment, Elavation 1013, RC Pump Cavities-11 Containment,. Elevation 99h .

12 . Containment, Elevation 10h5

13. Auxiliary Building, Elevation 1025 (Rooms 69 and Til lh' .-Turbine Building, Elevation 990-

'15 l Turbine Building, Elevction 1011 16 Tutb 4e Building, Elevation 1036

'17 c Containment Fans VA-3B and VA-7D L18 Containment Fans.VA-3A and VA-7C 19- Containment Fans VA-P.A and VA-2B' 20 . Control Room Panels CB-l'2/3 Return Air 21 Containment NSWC Fans VA-12A and VA-12B 22 Containment Purge Discharge Fans VA-32A and VA-32B 23 DG-2 Room Exhaust Fan, VA-52B 2h- Containment Purge Supply Fans VA-2hA and VA-2hE 25 Control Ecom and Hallway Ventilation Ducts 26 Auxiliary Building (Controlled) Supply Fans, VA-35A and VA-35B 27 Auxiliary Building (Controlled) Exhaust Fans, VA-h0A, VA 40B, and VA h0C

'28' Auxiliary Building (Uncontrolled) Supply Fans, VA h5A and VA-45B 29 Auxiliary Building (Uncontrolled) Exhaust Fan, VA-41

-30 Auxiliary. Building Elevator Shaft Fan, VA-51

.31- Control Room Air Conditioning Fans, VA-46A and VA 46B 32 DG-1 Room Exhaust Pan, VA-52A 33 Auxiliary Building, Elevation 1036 (Room 81) 3h Plant Sprinkler Flov '

'35

-Auxiliary Building, DG-2 (Room 6h)

36 . Auxiliary Building, DG-1 (Room 63)

Intake Structure Including Rav Water Pump Room 38_ Auxiliary Building Open Stairvell 39; -Auxiliary Building Open Hatchway

. Amendment No. 38 2-90

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. (Continuedl- ...

HALON' AREA FIRE ZONES--

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--Zone L.  :

" No. '

• a- ,

l11 Cable' Spreading Room. .

-Cable Spreading Room:

'2i .

3-  ! Control Boon Walk-In Cabinets' 3 - ~ fControl Room Walk-In Cabinets

-. 5 5 . iSwitchgear Room..--West

- 6' Switchgear' Room  : West-s7 JSwitchgear. Room Eastc 8- 'Switchgear. Room - East' t.

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s TABLE 2-8-- ,

. . FIRE HOSE STATION LOCATIONS'

--No. . Location- Elevation' Size

' 1. - FP hN'- -Intake' Structure 1012'-6"- .1 5"/2 5"'

2.. FP hP.  ! Intake Structure' 1012'-6" 1 5"/2.5"'

At grade level .

3.  ! FP-3C ' Yard ' Area : 2 5"-

4.1 FP-3B ' Yard Area' At grade level .2.5" 5.- FP-3A'. Yard Aren' At grade level 2.5"

- 6. FP-3F Yard' Area At grade'. level 2.5" 7 FP-3E. . Yard Area. At grade level- 2.5" 8.. FP-3D. Yard-Area. At grade' level 25" 9.- FP-TA . Auxiliary Building 989'-0" 1.5"/2.5"1 .

.10. FP-TB-- . Auxiliary' Building ~ 989'-0" 1.5"/2.5"

11. FP-TC Auxiliary Building 989'-0" -1.5"/2.5"

12. FP-7D Auxiliary Building 989'-0" 1.5"/2 5"

13. 'FP-TE~ Auxiliary Building 989'-0" 1.5"/2.5"

14. FP-7F Auxiliary Building 989'-0" 1 5"/2 5" 15 FP-TG Auxiliary Building 989'-0" 1 5"/2.5"

.16. FP-8A Auxiliary Building -1011'-0" 1.5"/2 5" 17 .FP-8B Auxiliary Building 1011'-0" 1.5"/2.5"

. 18 . FP-8C' Auxiliary Building 1011'_-0" 1 5"/2.5" 19 FP-8D Auxiliary Building 1007'-6" 1.5"/2 5" 2C. FP-8E . Auxiliary Building 1007'-6" 1.5"/2.5"

21. FP-8F. Auxiliary Building 1007'-6" 1.5"/2 5"

22. FP-8G- Auxiliary Building 1007'-6" 1.5"/2.5"

23. FP-8H Auxiliary Building 1007'-6" 1.5"/2.5" t.

AmendmentNo.38/53 '2-94 .f

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iTABLE 2-8 (Continued)

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. FIRE HOS5: STATION LOCATIONS .

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' No .' 5Locationf ' Elevatio'n I . Size; F

2k. .FP-9A . Auxiliary l Building 1025 '-d.

1.5"/2 5"'

[ , i 25'.- FP-9BS Auxiliary' Building 1025'-0" - 1 5"/2 5"- m p

f26.1 ! FP-9C ~ 1Auxi11ary Building .1025'-0" 1 5"/2 5"'

27. -. FP-9D. ' Auxiliary Building?

.1025 '-0"- 1 5"/2 5"-

4

, l 28.1  ; FP-10A . Auxiliary Building

. 11036'-0" 11 5"/2 5

_ -29 FP-10B.- -' Auxiliary = Building 1036'-0" 1.1"/2 5" '-

130.- FP-100- ' Auxiliary Building .1036'-0" 11.5"/2 5"'

31. ;. FP-10D Auxiliary Building .1036'-0" - 1. 5'_'/2. 5 " ,

^32. FP-10E - Auxiliary Building' 1036'-0"~ 1.5"/2 5"~

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UAmendment.No. 53: .

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-TAELE 3 '

MINIMUM FREQUEUCIES FOR CHECKS, CALIBRATIONS-AND TESTING ~OF REACTOR' PROTECTIVE'SYSTEN- h?~,d f

s Surveillance _ . .

-*- +

+

Channel Description Function . Frequency- Surveillance'Methodi -

1. ~ Power Range Safety .a. Check- -S- .a.' Compazison cf four'powerj. _ ~

. Channels channel' readings,'forlbothi - '

neutron. flux and thermal'. power. - y,

b. Adjustment 'DS)- lb. ChEnnel'. adjustment to , ,

agree;with heat-balance '

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calculation.- -

Int'ernal test. signal.to

c. Calibrate. M(2) c.-

.and Test verify. trips,. alarms, per-missives'and auctioneer ' ' '

circuits. ,

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2. Wide-Range Logarithmic' a. Check 'S a. Comparison of'fourl wide-range ~ ^

Neutron Monitors readings.

b.- Test (3) , p. 1b. Internal test signals to . ,

verify SUR 'ind' ication~ and ~:

~

trip, powerTlevel'permissives,;

instrument? accuracy.. -

3. Reactor Coolant Flow a. Check S ~a .: . Comparison of.fourfseparate , r

' total + flow indications.'

^

b. Calibrat'e ~ :R b. 'Known differential pressure

~

~ applied to sensors to cali ' ,  ;,'

. tbrate all:loopfdevices.; _

c. . Test , M(2)- .c...Bist$bleltrip;. tester.(1)

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[ (Continued) -

kD ' FSAR~Section: ,

Test Frequency Reference - ' 1 O

.At least once per plant-operating; x 10c. _( Continued): 4.'; Automatic.and/or Manual initia-y tion of't1e system shall'be de- cycle.

monstrated.

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11. ' Containment Cool- 1. Demonstrate damper action. 1 year, 2 years, 5 years,Jand . 9 10.

ing at.d Iodine .

every 5 years-theree.fter i.emoval Fuseable 2. Test a spare fuseable link. .

. Linked. Dampers

12. Fuel Elements Visually inspect fuel' elements During each refueling-outage: 3 removed from the reactor. ,

13 -Diesel Generator Calibrate During each refueling outage' i8.4.3.

y Under-Voltage g Relays p.

14. Motor' Operated Verify the contactor pickup value.' During each, refueling l outage '

' Safety Injection at <85% of 460 V.

• Loop Valve Motor Starters (HCV- "

311, 314, 317, 320, 327, 329, i

- 331, 333, 312, I

. 315, 318, 321) 15 Pressurizer Verify control circuits .During each refuel -

Heaters operation for post- ' ing~ outage.

accident heater use.

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'4[0$[DESIGNFEATURES-( ,s H M 3!} Nuclear Steam Supply System (Con $1nued).- ~

, .k.3.1l! Reactor' Coolant-System:(Continued)--

,_ > . .. ., . ~

y The* reactor coolant-system is designedLand; constructed in accordance. -

i v withTthe'ASME. Boiler and Pressure VesselECode, Sec51on'III; Rules .

1for Construction of Nuclear _ Vessels l including all. addenda through'

ftheLvintercof.1967 Land the1 Code for Preasure Piping USAS-B31.1.

The; reactor coolantisystemjis_ designed for a pressureToff2500f .. - '

~ fpsialard a temperature;of 650 F"except forLthe pressurizer which ihas.a, design l temperature of 7000F. The volume of, the' reactor?

Leoolantfsystemfis approximately 6,616. cubic' feet. 7 -

e 4.3.2 ; Reactor-Core and:Centrol- >

The'-reactor core shall approximate. a rightl circular cylirider with' an:equivalenti-diameterLof 106.5' inches and an: active height of.

1128Linches. The reactor' core shall normally consist of Zircaloy-h, clad fuel-rods _containing-slightly enriched uranium in the-form"of:

1. _-sintered UO~ , 2 pellets. - The fuel rods'shall normlly. be grouped ~into ,

.133 assemblies.-

The; core exceso reactivity shall be' controlled by a combination-

- of boric ~ acid chemical shim, contro1' element assemblies,- and -

[ mechanically fixed.. boron rods where required. Forty-ninefcon-

, , .-trol element assemblies are distributed throughout the core as shown ;in Figure '3.h-5 of the FSAR; four of..the CEA's :contain~

part-len'gth adsorbers.'

!h.3.3:

• Emergency core Cooling

. Emergency. core" cooling is.provided_by.'the' Safety Injection' System which' consists-of.various subsystems, each with internal redundancy.

Included in'the Safety Injection System are four safetyLinjection tanks, three high-pressure and two low-pressure safety' finjection pumps, aisafety' injection and refueling water storage tank, and l interconnecting piping as shown.in FSAR Section 6. ~

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50 ADMINISTRATIVE CONTROLS f

5-5 1 7 b. Render determinations in writing with regard to whether or not each item considered under 5.5.1.6(a) through (e) above constitutes an unreviewed safety question.

c. Provide immediate written notification to the Section Manager - Operations and the Safety Audit and Review Committee of disagreement between the Plant Review Committee and the Manager - Fort Calhoun Station; however, the Manager - Fort Calhoua Station shall have responsibility for resolution of such disagreements pursuant to 5 1.1 above.

Records 5 5.1.8 The Plant Review Committee shall maintain written minutes of each meeting and copies shall be provided to the Section Manager - Operations and Chairman of the Safety Audit and Review Committee.

552 Safety Audit and Review Committee (SARC'i Punction 5 5.2.3 The Safety Audit and Reviav Committee shall function to pro-vide the independent review and audit of designated activitiee in the areas of:

a. nuclear power plant operation

b. nuclear engineering

c. chemistry and radiochemistry

d. metallurgy

e. instrumentation and control

f. radiological safety

g. mechanical and electrical engineering

h. quality assurance Composition 5 5 2.2 The Safety Audit and Review Committee shall be composed of:

Chairman: Division Manager - Environmental and Regulatory Affairs Member

• Asriatant General Manager - Production Operations, Fuela, and ERA Member: Assistant General Manager - Electric Operations and Engineering Member: Division Manager - Engineering l Member: Division Manager - Production Operations Member: OPPD Operations, Engineering, and Technical Support Staff Member: Qualified non-District Affiliated Consultants r,s Required and as Determined by SARC Chairman Amen 9. ment No.Af, 19 5-5

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-' Alternates _

5 5 2.3i Alternate members shall be' appointed in writing by the Chair-

. man .of the Safety Audit and Review Committee to serve on a '

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' temporary basis; however,- no more than tvo aP.ernates may:.

pr.rticipate in the Safety. Audit and Review Committee acti-vities atiany'one time.:

' ' Consultants .

5.5.2.h~ : Consultants'shall be utilized as determined by the Safety Audit and' Review Committee Chairman to provide expert advise to the Safety Aud't .and Review Committee.-

Meeting Frequency 5.5.2.5 The Safety Audit and Review Committee shall meet at least once every six months.

t _ Quorum

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5 5 2.6 A quorum of the Safety Audit and Review Committee shall' con-  ;

sist..of the Chairman or his designated alternate and a major-ity of the Safety Audit and Review Committee members including alternates. No more than a minority of the~ quorum shall have line responsibility for the operation of the nuclear plant.

- R_eview 5.5.2.7 The' Safety Audit and Review Committee shall review:

a.

The safety evaluations for 1) procedures, equipment or.-

systems and 2) tests or experiments: completed under the provision of section 50 59, 10 CFR, to verify that such actions did not constitute an unreviewed safety question. .

b. Proposed changes to procedures,' equipment or systems which involve an unreviewed safety question as defined in section 50.39, 10 CFR. 6 Arendment:lo.)I,19L 5-6 1

g. -y I 25 0 ADMINIGTRATIVE CONTROLS I

5 5.'2.8 e. The Fort Calhoun Station Emergency Plan and implementing procedures at least once per two years.

1

f. The Site Security Plan and implementing procedures at least once per two years,

g. Any other area of facility operation considered appropriate by the Safety Audit and Review Committee or the Assistant General Manager -

Production Operations, Fuels, and Environmental & Regulatory Affairs.

Authority 5,5.2 9 The Safety Audit and Review Committee shall report to and advise the Assistant General Manager - Production Operations, Fuels, and Environ-mental & Regulatory Affairs on those areas of responsibility specified in Section 5 5 2.7 and 5 5 2.8.

Records

).5.2.10 Recordr of Safety Audit and Review Committee activities shall be pre-pared, approved and distributed as indicated below:

a. Minutes of eacL Safety Audit and Review Committee meeting ahall be prepared, approved and forwarded to the Assistant General Manager - Production Operations, Fuels, and Environmental &

Regulatory Affairs within 1h days following each meeting,

b. Reports of reviews encompassed by Section 5.5.2.7 e, r, g, and h above, shall be prepared, appr-aed and foruarded to the Assistant General Manager - Production Operations, Fuels, and Environmental & Regulatory Affairs within lh days following completion of the review.

c. Audit reports encompassed by Section 5 5 2.8 above, shall be forwarded to the Assistant General Manager - Production Oper-ations, Fuels, and Environmental & Regulatory Affairs and to the responsible management positions designated by the Safety Audit and Review Committee within 30 days after completion of the audit.

5.5.3 Fire Protection Inspection

a. An independent fire protection and loss prevention inspection and audit shall be performed annually utilizing either quali-fled off-site licensee personnel or an outside fire protection firm. The audit and inspecticn program responsibility shall reet with the Safety Audit and Review Committee.

b. An inspection and audit of the fire protection and loss pre-vention progran by an outside qualified fire consultant shal2 be performed at intervals no greater than 3 years.

5.6 Reportable Occurrence Action 5.6.1 The following actions shall be taken in the event of an REPORTABLE OCCURRENCE:

a. The Commission shall be notified and/or a report submitted pursuant to the requirements of Specification 5.9 Amendment No. 9, 19, 38 5-8

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6.0' ? INTERIM SPECIAL' TECHNICAL SPECIFICATIONS

[f y !_6. 41 Operation With Less Than 75% of.Incore Deteetor Strings Operable 1(Continued)

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-(d) If 'U 7% 'the Lt'ot 1 peaking uricertainty factor defined ' ' '

o as1 ($ +>,Uq ) shall be used in place of the measurement-.

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calculation factor of 1.07'in Specification .2.10.k(1). . ~

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..(e):. 'The maximum' local peak linear.f. eat' rate in the core, '

.-c 14 max; shall'be' determined'and the incore detectcr_ ' arms '

shall be. adjusted to-no. greater than the following:

-Alarm Setting = C'* eallowed 9 max -

j - where

- i C = The de* actor signal. converted.to' flux units )

when the reactor,is. operating at steady-state.

Gallowed = Linear Heat Rr/;e (kv/ft) allowed by Specifi-4

cation 2.10.h(1) and adjusted as required by Specification 6.3(1)(d).

.4 max =.The meximum local peak linear. heat rate (kw/ft)'

measured e.!, the same reactor conditions as.

C above.

L(2) If the incore detector system is not operable within the interval specified, the peak linear heat. rate shall be monitored by ex- ~

core detectors.per Specification 2.10.4(1)(c) and the surveillance l requirements of Specification 3.10(5) are deleted.

Basis

, Operation of the-incore detector '

monitoring and surveillance of F pstem for peak linear heat rate R and Fg Twith less than 75% of the strirgs operable requires additional measures to compensate for degradation of the incore instrument system. Periodic com-parisons between calculated and measured power distributions -are made to confirm the core is depleting as designed. . The measure- -

ment uncertainties are computed to assure the assumptions made in the setpoint analysis are valid. The uncertainties are computed using the methods given in the reference. ,

If the determined uncertainties exceed the uncertainties used'in the' setpoint and safety analysis, . the measured values of FR and

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F are augmented by the-appropriate uncertainty. These new values ' '

' oFF3 and r are then used to verify compliance with Specifications 2.10.4(2) and'2.10.4(3). This assures that the product of the '

radial. peaking factors and their appropriate-uncertainties are [

less than.the velues used in determining the setpoints. *

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j Amendment No. 55 6-5'

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a v , DISCUSSION

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0The; proposed licena'e change is submitted to:

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' , ((1)?CorrectthMTaNeiofContents:(pageii). -

U e ](2)fCorrectatypographical' error-(page2-9).

(3) ~ Correct a' table numbering problem-(duplicate Tables.2-6).

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(4) : Provide clarification of-bypass Leonditions- for three Reactor -

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' Protective System (RPS) trip. functions ~(. Table 2-2)..

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( (5) Delete turbine' runbacs ' test reference (Tab 1'e 1). '

(6) Correct a typogiaphical error (Table-3-5).-

(7)* Update reactor l core'and-control design' features to reflect l changes

- . fin fuel design with Cycle 6 reload'.

(8) > Update,0maha Public, Power. District's organizational charts (Figures' -

~5-1,:5-1A, and 5-2).

~ '(9) : Revise the~ Fafety Audit and Review Comittee membership. consistent 1with the new organization (.pagesl5-5, 5-6, and'5-8).

A more detailed discussion of each item above is provided. .

Table of Contents (page 11). The Table of Contents-has been revised

'to add section 5.5.3. Specification ~5.5.3 was added by Amendment No. 38.

Typographical Errors (page 2-9). .Two typographical errors were noted

.on page 2-9.. The first sentence should read: "The potential dose at the sita boundary..." Also, item (1) title should be: "(1) Steam Generator Tube Rupture". ' These two items have been corrected.

Typographical Errors-(page 2-3h). Change AEC to NRC. Also, change Denver to Arlington, Texas.

Table Number Change. Two tables' numbered 2-6 are presently in the 1

-Technical Specifications. The first is on page 2-57f and is referenced "in~. Specification 0.10.4 (pages 2-57c and 2-57d). The second Table 2-6 is on' pages 2-75 through 2-88 and is referenced in Specification 2.18 '

(pages 2-73 -and 2-74). . The ' table of page 2-57f has been renumbered to j

. Table 2-9 and references to it corrected accordingly.

L- : RPS Bypass Conditions. For each channel, a single bistable provides for: automatic activation' of the Axial Power Distribution ( APD) and Loss of  ;

! Load reactor' trips andideactivation (bypass)' of the high rate trip-wide ~

range log channels as reactor power increases above 15% + %. Table 2-2 lists . permissible bypass conditions as follows:

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F j t 9 9 Axial Power Distribution: Belov 15% of rated power.

10 High Rate Trip-Wide Range Long Channels: Belov'10-b5 and above 15% of rated power.

11. Loss of Lead: Below 15% of rated power.

Using the same bistable to provide the automatic bypass function, condition 10 cannot be met concurrently with items 9 and 11. In order to eliminate any confusion, note (e) has been added to Tsble 2-2.

This proposed change does not constitute an unreviewed safety question, as detailed in the attached evaluation (Form FC-154). -

Table Corrections. Tables 2-7 and 2-8 have been corr acted to agree with Fire Protection System since installation of fire protection modifications.

Turbine Runback Test (Table 3-1). Amendment No. 32 deleted the turbine runback requirement from the Fort Calhoun Station Unit No. 1 Technical Specifications. This editorial change updates the surteillance testing requirements.

Typographical Error (Table 3-5). Item 14, under the component eclumn, lists 12 valves; the ninth in succession being HCV-372. This is a typographical error and has been revised to the correct valve number, HCV-312.

Reactor Core Design (Specification 4.3.2). The reactor core design has changed vitt the Cycle 6 reload. Specifically, there are no burnable poison rods .nd the fuel distribution has been changed such that there are nov 23,408 fuel rods in the core and 120,610 pounds of UO2

• Organizational Charts. The District's Staff Support organizational charts (Figures 5-1, 5-1A, and 5-2) have been revised to reflect the District's current organization. Figure 5-1A has been deleted.

Safety Audit and Review Committee (SARC) Revision (Specification 5 5.2).

The proposed license change revises the SARC membership listing to reflect Omaha Public Power District's current organization and to specifically designate those consultants and staff members assigned to the SARC. This is strictly an administrative change and does not change the responsibilities or functions of the SARC. Also, 5 5.2.8.e. is revised to delete fire protection program, since this requirement is redundant to Technical Specification 5.5.3.a.

Interim Technical Specification 6.h. Typographical error in section 6.4(2). Change is made to reference correct Technical Specification section.

The eleven proposed license changes discussed above do not constJtute safety or enviroamental issues and do not constitute an unreviewed safety question. Additionally, all proposed license changes are consistent with the requirements of Combustion Engineering Standard Technical Specifications.

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,:(JUSTIFICATION FOR FEE CLASSIFICATION The proposed amendment is deemed to be a Class II Amendment, within the meaning of 10 CFR 170.22. The requested changes are >

strictly administrative .in' nature and _ represent no safety consider--

lations.

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