ML20054K062

Amend 65 to License DPR-40,adding to Tech Specs Setpoint, Operability & Surveillance Requirements for safety-grade Auxiliary Feedwater Automatic Actuation Sys

ML20054K062
Person / Time
Site:	Fort Calhoun
Issue date:	06/18/1982
From:	Clark R Office of Nuclear Reactor Regulation
To:
Shared Package
ML20054K061	List: ML20054K060 ML20054K062 ML20054K063 ML20054K064
References
NUDOCS 8206300365
Download: ML20054K062 (14)

Text

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'g UNITED STATES

,  ! p .r,( g NUCLEAR REGULATORY COMMISSION f* '&;. "

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W ASHINGTON. D. C. 20555

%; *** *vf DESIGNATED ORIGIILG Certified By btW b <

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,v, OMAHA PUBLIC POWER DISTRICT DOCKET NO. 50-285 FORT CALHOUN STATION. UNIT NO. 1

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AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 65 License No. DPR-40

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

A. The application for amendment by the Omaha Public Power District (the licensee) dated November 17, 1981, as supplemented by la.tter dated March 22, 1982 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in. conformity with the application, the provisions.of the Act, and the rules and regulations of the Comission; C. There is reasonable assurance (1).that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; l

D. The issuance of this amendment will not be inimical to l

the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

l 8206300365 820618 PDR ADOCK 05000285 P PDR

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

B. Technical Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 65, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION k- Ghrt y Robert A. Clark, Chief 6 Operating Reactors Branch #3 Division of Licensing

Attachment:

Changes to the Technical Specifications

• Date of Issuance: June 18, 1982

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ATTACHMENT TO LICENSE AMENDMENT NO. 65 FACILITY OPERATING LICENSE NO. DPR-40 DOCKET NO. 50-285 Revise Appendix "A" Technical Specifications as indicated below. The revised pages are identified by amendment number an'd contain vertical lines indicating the area of change.

Remove Pages Insert Pages 2-62 2-62 2-63 2-63 ,

2-64 2-64 2-64a , 2-64a 2-65 2-65 2-68 2-68

- 2-68a 2-70 2-70 2-70a 2-70a 3-12a 3-12a 6-3 6-3 e

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2.0 LIMITING CONDITIONS FOR OPERATION 2.14 Engineered Safety Features System Initiation Instrumentation Settings (Continued)

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

The set points for the isolation function have been se-lected to li=it radioactivity concentrations at the boundary of the restricted area to approximately 0.25 of 10 CTR 20 li=its, assu=ing existence of annual average meteorology.

1 Each channel is supplied from a separate instrument a.c.

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

(4) Low Steam Generator Pressure l

A signal is provided upon sensing a low pressure in a steam generator to close the main steam isolation valves in order to minimize the temperature reduction in the reactor coolant j system with resultant loss of water level and possible i addition of reactivity. The setting of 500 psia includes

! a +22 psi uncertainty and was the setting used in the safety analysis.(3)

As part of the AFW actuation logic, a separate signal is l

, provided to terminate flow to a steam generator upon l sensing a low pressure in that steam generator if the other -

! steam gen &rator pressure is greater than the pressure setting. This is done to uinimite the temperature re-duction in the reactor coolant system in the event of' a main steamline braak. The ietting of 466.7 psia includes a

+31.7 psi uncertainty; therefore, a setting of 435 psia was used in the safety analysis.

(5) SIRW Tank Low Level Level switches are provided on the SIRW tank to actuate the valves in the safety injection pump suction lines in such a manner so as to switch the water supply from the SIRW tank to the containment sump for a recirculation mode of operation after a period of approximately 24 minutes following a safety injection signal. The switchover point of 16 inches above tank bottom is set to prevent the pumps frem running dry during the 10 seconds required to stroke the valves and to hold in reserve approximately 28,000 '

gallons of at least 1700 ppm borated water. The ySAR lose of coolant accident analysis l') assumed the recirculation started when the minimum usable volume of 283,000 gallons had been pumped from the tank.

(6) Low Steam Generator Water Level As part of the AFW actuation logic, a signal is provided to initiate AFW flow to one or two steam generators upon sensing a low water level in the steam generator (s) if the 2-62 .

A=endmenton . 5, 32, 43, 65

2.0 LD!ITU:G CONDITIONS FOR OPERATION 2.14 Engineered Safety Features System Initiation Instrumentation Settings (Continued)

(6) Low Steam Generator Water Level (Continued) absolute steam generator pressure criteria ~or differencial steam generator pressure criteria are satisfied. This function ensures adequate steam generator water level is maintained in the event 'of a f ailure to deliver main feedwater to either steam generator. The setting of 28.2% of wide range cap span includes a +13.2% uncertainty; therefore, a setting of 15% of wide range cap span was used in the safety analysis.

(7) Hieh Steam Generator Dalta Pressure As part of the AFW actuation logic, a high steam generator differential pressure signal is generated to provide AFW to the higher pressure steam generator with a concurrent icw level signal if both steam gen'erator pressures are less than 466.7 psia. If the differential pressure between stern *,enerators is less chan the setting,neither steam genera.cor is supplied with AFW in the presence of a low level signal. The setting of 119.7 psid includes a -15.3 psi ancertainty; therefore, a setting of 135 psid was used in the AFW safety analysis.

References (1) FSAR, Section 14.1.3 . .

'(2) FSAR, Section 11.2.3.2 (3) FSAR, Section 14.12 (4) FSAR, Section 14.15 (5) -FSAR, Section 7.4.6

l. (6) FSAR, Section 7.5.2.5 (7) FSAR, Section 14.4.1 ,-

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Amendment No. 65 2-63 l

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TAhlE 2-1 Enrineered Safety Features System Int.tiation Instrument S*tting I.inits ,

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9 Setting I.imit Punctional Unit Channel-3 < 5 psis

1. High Containment Pressure a. Safety Injection * ~

Containment Spray I3)

.4

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a-

c. Containment Isolation .

" d. Contairusent Air Cooler DBA Hode m

a.

' > 1600 psia

2. Pressurizer inv/ Low Pressure Safety Injection (3)

Contaisunent Spray

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b.

g c. Containment Isolation

d. Containment Air Cooler DilA Hode

3. Containment High Radiation Containment Ven'11ation t Isolation IkI i HH-050, 9 6 x 10-2 peg /3ec

< HH-051, 1.5 x 10-3 pel/cc 7 131-060, 9 6 x 10-2 pel/sec y 5 Hil-061, 9.6 x 10-2 pci/sec p 1 HH-062', 1.5 x 10-3 pci/cc

a. . Steam Line Isolation > 500 usia(2)

Is . Low Steam Generator Pressure [466.7 psia l

b. Auxiliary Feedwater Actuation -2 in, above Hecirculation Actuation 16 inches +0, 5 SIRW Low Level Switches tank bottom

6. ,4.16 KY Emergency Bus Low s. Iose of Voltage (2995 2- + lois) volte ]-Trip 20.8 -

Voltage s 15 9(5) seconds

b. Degraded Voltage

1) Bus IA3 Side > 3825.52 volts Trip D.B i .5) seconds e

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s TABLE 2-1 (Continutd)

G Engineered Safety Fea(ures System Initiation Instrument Setting Limits ,

k 3

x Setting Limit Functional Unit Channel _

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sr

6. (Continued) b. (Continued) ,

7#

cn

11) -Bus lAh Side > 3724.08 volta 7,gp Th.8 + .5) seconds

> 28.2% of wide range tap span 7 Low Steam Generator Water Level Auxiliary Feedwater Actuation

8. liigh Steam Generator Delta Auxiliary Feedwater Actuation . _1 119 7 P81d Pressure

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(1) Hay be bypassed below 1700 psia and is automatically reinstated above 1700 psia. ,

(2) !!ay be typassed below 550 psia and is automatically reinstated above 550 pala. .

(3) Simultaneous high containnhnt pressure and pressurizer low / low pressure. , '

(4) RH-050 and RH-051 may be Inoperable or out of service with respect RH-061 andto containment RM-062 monitoring, provided may be inoperable, that the containment ventilation isolation valves are closed. RH-060 may be inoperable, pro-provided that RM-050 and RH-051 are monitoring the ventilation stack.vided that (1) iodi (2) gas decay tank releases are not made. (For voltsge > (2995.2 - 20.8) volta, time (5) Applicable to bus voltsgo < 2995 2 - 20.8 volts only.

delay shall be > 5 9 seconds.)

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e 2.0 LI:t!*!?iG CC'ID--'IC:IS FOR CPE*ATIC:(

2.15 Instr =entatics and centrol systems Atelicabilitr

. Applies to p' * ' a str=enta.ics syste=s.

% 'ective To delineate the ccaditicas of the plant instru=entaticn and centrol syste=s necessary to assure reactor safety.

Scecificaticas The operability of the plant instreanent and centrol syste=s shall be in accordance with Tables 2-2 through 2-5. g In the event the nu=ber of channels of a particular system in service falls below the l' 'ts given in the colu==s entitled

" iini=u= Operable Channels" or " Mini:r.un Degree of Redundancy",

except as conditioned by the colu=n entitled " Permissible By-pass Cenditions", the reacter 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 contain=ent isolation sig=als available if the ventila-tien isolation valves are closed. If =inimum conditions are not =et 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.

If, during power operation, the rod block function of the second-ary CEA position indication system and rod block circuit are-inoperable for more than 24 h.ours, or the plant computer PDIL alar =, CIA group deviatien alar = a=d the CIA sequencing functics are ineperable f:r = ore than kS tcurs , the CIA's shall be with-drawn and --4"tained at fully withdrawn and the control rod drive syste= mode switch shall be maintained in the off posi- ,..

tion except when~=anual =otion of CIA Group h is required to l centrol axial power distribution.

Basis i During plant operat' ion, the co=plete instrumentation systems vill nor= ally be in service. Reactor safety is provided by the reactor protection system, which automatically initiates appropriate action to prevent exceeding established 11=its. ,

Safety is not compremised, however, by continuing operation vith certain instru=entation channels out ,of service since provi-sions were made for this in the plant design. This specifi-cation outlines limiting conditions for operation necessary to preserve the. effectiveness 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 sufficien1; redundancy to provide the espability for channel test at power, except for backup channels such as. derived circuits in engineered safeguards control system.

• Anend=ent :fo. 3, 20, ')l, f 65 2-65 ,

TABLE 2-3 Instrument Oceratine Recuirements for Enrineered Safety Festures Mininus Minicus Per=issible Operable Oegree of 37 pass no.  ?.nettenal Unit Channels ?edundane'! Cenditions 1 Safety In.iectien A Manual 1 Ncne Ncne 3 High Contain=ent Pressure A 1 During Leak Test 3 2((a) (d) 2 a) (d) ,

C Pressurizer kv/kv A 2(a) (d) 1 Reacter Coolant Pressure 3 2(a) (d) 1 Pressure Less Tha 1700 psia (b) 2 Centainment Sersy A Manual 3 1 None None 3 High Centain=ent Pressure A 2(a)(c)(d) 1 During Imak Test 3 2(a)(c)(d) 1 C Pressurizer k v/Lov A 2(a)(c)(d) 1 Reactor Coolanti 3 2(a)(c)(d) 1 Pressure Less Than

,' 1700 psia (b) 3 Recirculatien A Manual 1- Ncne None 3 SI?.W Tank M v Level A 2(a) (d) 1 None 3 2(a) (d) I h Energency off-Site Power Trio A Manual 1(*} None None B E=ercency Bus kw Voltage (Esch Bus) - M ss of voltage 2(d) 1 Reactor Coolant

- Degraded voltage 2(a)(d) 1 Tec:perature Iess Than 2000F A=end=ent No. M; 65 2-68 . .

l TABLE 2-3 (Continued)

Instrument Ooeratina Recuirements for Encineered Safety Features

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Minimum Minimum Permissible Operable Degree of Bypass flo . Functienal Unit Channels Redundancy Conditions 5' Auxiliarv Feedwater A Manual 1 None None .

B Auto. Initiation A Operating B

Modes 3, 4, and 5

- Steam Generator Low Level 3(a)(f) )

- Steam Generator Low Pres-sure 3(a)(g) 1

- Stsam Generator Differen-tial Pressure i 3(a)(g) 3 a A and B actuation circuits each hav'e 4 channels. .

b Auto removal of bypass above 1700 psia.

c Coincident high containment pressure and pressurizer low / low pressure signals required for initiation of containment spray.

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

e Control switch on incoming breaker.

f Two channels lire allowed to be inoperable provided that one and only one is in the low level actuation permissive condition.

g Three channels required because bypass or failure results in auxiliary feedwater actuat. ion block, in the affected channel.

Amendment'No. 65 2-68a  :

TA3I.I 2-5 Ins: : .e .11 tic C:er11tne ?ec itrer.ents te: other Saten.r Test.re yuncti:ns F4 3 u= 244 d-.::s Per=issible-Oper2ble ters se of - 37; ass

.  ?:.-citen ti *: .it C'.1-m el s  ? e - ' , - ~r Cen'itters 1 C.~:.A Positic: b dicatien 1 Ncce Uc=e syste=s ,

2 Press ' -- 7avel 1 Ncce Not Applicabla 3 subecoli:g Ma.cgin 1 Ncce Not Applicable Mc=itor .

4 .. PCRV Acoustic Position 1 ac , gene Not Applicable

!=dicatics-Direct i

5 Safety Valve Acoustic 1 ac None Not Applicable Positien Indication 6 PcRV/ safety Valve Tail 1 db None Not Applicable Pipe Te=perst ce 3C"TE: L a One channel per valve.

b Cne RD for both PCRV's; two RD's , one for each code safety, c If'ite= 6 is operable, require =ents of specification 2.15 are : modi-l fled for items 4 a=d 5 to " Restore inoperable , k=n-els to cpershility vithin T days or be in het shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />."

! d It items 4 and 5 are operable, requirements of specification 2.15 are = edified for item 6 to " Restore i=cpersble channels to oper-ability withis 7 days or be in het shutdovn 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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A:::d:ent :fo. J'y, 65 2-70 e

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TABLE 2-6 DELETED l

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l Amendment No. 7t 2-70a f , 65

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TAlltE 3-2 (Continued)

HillilRIM FilEQtlENCIFS F0ll CilECKS. CAI.IDllATIONS AND TESTTilG OF k'.

" ~ EllGillEEllED !1 AFRIT l'EATilillTi. INSTiltillEllTATinti AllD COfftflalS Surveillance Clintuiel Description reinction Frequency !Inrveillance liethod 4

sa 22. Auxiliary'Feedwater se l

a. Check S

a. Coispare independent level cn a. Steam Generator Water readings.

f "' level low (Wide Range)

b. Calibration H b. Known signal applied to sensor.

a. Check 8 a. Compare independent pressure

b. Steam Generator readings.

Pressure Low

b. Calibration R b. Kr.own Signal applied to sensor.

Y' a.' Calibrate H -

m. Known signal applied to sensor.

c. Steam Generator Differential Pressure .

liigh Actuation Circuitry a. Test M A. Functional check or initiation

d. ~ circuits.

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b. Test y R b. System functional test of AFW .

)

i initiation circuits. 1 L

S - Each Shift .

D - Daily H - Monthly Q - Quarterly R - 18 Honths P - Prior to,Each Start-Up if Not Done Previous Week HP - Monthly during designated modeu and prior to taking the reactor critical if not completed within the previous 31 days (not upp11 cable to a fast trip recovery)

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6.3 Au niis- Feed ea e- Automatic !.-itt a-10 '. Settoint (nis specificatics is Deleted - Page != e=tional17 Left 31a=k) e l

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a A=en-h:s :o. Jh, 65 6-3

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