ML20199L895

Proposed Tech Specs,Reflecting Relocation of pressure-temp Curves,Predicted Radiation Induced NDTT Shift Curve & LTOP Limits to FCS Unit 1 RCS PT Limits Rept

ML20199L895
Person / Time
Site:	Fort Calhoun
Issue date:	01/30/1998
From:	OMAHA PUBLIC POWER DISTRICT
To:
Shared Package
ML20199L879	List: LIC-98-0013, Forwards Application for Amend to License DPR-40,relocating pressure-temperature Curves,Predicted Radiation Induced NDTT Shift Curve & LTOP Limits,Per GL 96-03.Document Titled RCS Unit 1 RCS PT Limits Rept & CE NPSD-683,Rev 2 Encl ML20199L891 ML20199L895 ML20199L909 ML20199L918
References
NUDOCS 9802100064
Download: ML20199L895 (31)

Text

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TABLE OF CONTENTS (Continued)

Nuclear Steam Supply System ( 'SK........,..................PREt 4.3 43 1

.4.3.1 Reactor Coolant Sy em (RCS).m)...........................43 4.3.2 Reactor Core and C

.........................43 4.3.3 Engtency Core C mg 4-3 4.4 Fuel Storage

.....,,.......................................4-4 4

4.4.1 New Fuel Storage........................

.............,4-4 4.4.2 Spent Fuel Storage.....,...............................

4-4 4.5 Seismic Design for Class I Systems................................

4-5 5.0 ADMINISTRATIVE CONTROLS

.........51 5.1 Responsibility...

5-1 5.2 Organization

..............................................51 5.3 Facility Staff Qualifications....................................5.la 5.4 Trai nin g.................................................

5 -3 5.5 Review and Audit 5-3 4

5.5.1 Plant Review Committee (PRC) 5-3 5.5.2 Safety Audit and Review Committee (SARC).,.......,,...,...

55 5.6 Reportable Event Action.......................................

5-9 5.7 Safety Limit Violation 59 5.8 Procedures................................................

5 9 9

Reporting Requirements 5 10 5.9.1 Routine Reports............

5 10 5.9.2 Reportable Events...................................... 5 12

.9.4 niqu chrt' equiremen[

'l..... 'l[.ll 5.9.5 Core Operating Limits Report.......

........ 5 17a 5.9.6J URCS Press _ure-Temperatntcl Limits;Reportf(PTLR)). W.,....@.MW5+17b 5.10 e'enf in n

-....... A....h......

5.11 Radiation Protection ?itism 5-19 5.12 DELETED 5.13 Secondary Water Chemistry..................................... 5-20 f.14 Systems integrity....

5 21

". : i Post-Accident Radiological Sampling and Monitoring 5-21 5'6 Radiological Effluents and Environmental Monitoring Programs.............., 5-22 5.16.1 Radioactive Effluent Controls Program.......

5-22 5.16.2 Radiological Environmental Monitoring Program................... 5-23 5.17 Offsite Dose Calculation Marmal (ODCM).

5-25 5.18 Process Control Program (PCP) 5-26 6.0 INTERIM SPECIAL TECHNICAL SPECIFICATIONS 6-1 6.1 Limits on Reactor Coolant Pump Operation 6-1 6.2 Use of a Spent Fuel Shipping Cask 6-1 6.3 Auxiliary Feedwater Automatic Initiation Setpoint

. 6-1 6.4 Operation With Less Than 75% of Incore Detector Strings Operable

. 6-1 iii Amendment No. 32,31,'?,51,55.57,

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9002100064 980130 PDR ADOCR 05000285~

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TECIINICAL Sl'ECIFICATIONS FIGURES TABLE OF CONTENTS PAGE WillCII FIGURE DESCRIITION -

FIGURE FOLI.OWS l1

.P Safet its 4 Pump -

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DEFINITIONS

' Dose Equivalent I 131 (pCi/gm)

= pCi/gm ofI 131

+ 0.0361 x pCi/gm ofI-132

+ 0.270 x pCi/gm ofI-133

+ 0.0169 x pCi/gm ofI-134

+ 0.0838 x pCi/gm ofI 135 B - Averane Disintegration Enc.tgX B is the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration, in MEV, for isotopes, other than iodines, with half lives greater than 15 -

minutes making up at least 95% of the total non-iodine radioactivity in the coolant.

'Offsite Dose Calculation Manual (ODCM)

The document (s) that contain the methodology and parameters used in the calculations of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent radiation monitoring Warn /High (trip) Alarm setpoints, and in the -

conduct cf the Environmental Radiological Monitoring Program. The ODCM shall also contain:

1)

The Radiological Effluent Controls and the Radiological Environmental Monitoring Program required by Specification 5.16.

2)

Descriptions of the information that should be included in the Annual Radiological Enviro _nmental Operating Reports and Annual Radioactive Effluent Release Reports required by Specifications 5.9,4.a and 5.9.4.b.

- Unrestricted Area i

Any area at or bevond the site boundarv access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials.-

Core Operating I imits Reoort (COLR)

The Core Operating Limits Report (COLR) is a Fort Calhoun Station Unit No. I specific document that provides core operating limits for the current operating cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Section 5.9.5, Plant operation within these operating limits is addressed in the individual specifications.

RCS W=%TA-MLhnhi RanaW(PTIJO l

The RCS' PRESSURE-TEMPERATURE LIMITS REPORT (PTLR) isla EnceTdependent report'providing Limiting Conditions 1for Operation for heatup, cooldosnzinservice 2

~

i hydrostatic?and' leak testing?and core critical _ity(limits in the form of Perssure Temperiturej (P-T) limits lto ensufe ~ ievention of brittle fracture.s;In additl6n,-lthisLreport establishes -

p l

1 hhia= Conditions for. Operation which provide: Low Terr.perature Overpressure; Protection (LTOP) to assure the P *1' limits are not exceeded'during the most limiting;LTOP;. event &The P-T limits and LTOP criteria;;in'the PTLR are.applicablefthroughithe.. Effective Full Power Years (EFPY) specified_ in the PTLRN NRC?and;ASMF[ approved. 'methodologie: fare usedias

% basis; for the. LCOslprovided[in the PTLRj]

Re ences A

(1) USAR, Section 7.2 (2) USAR, Section 7.3 8

Amendment No. 67,SS,141,152,164 l

l

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2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued)

I 2.1,1 Operable Componenti (Continued)

(c)

For the purposes of items (a) and (b) above, the containment spray pumps can be considered as available shutdown cooling pumps only if both of the following conditions are met:

i 1

(i)

Reactor Coolant System temperature is less than 120*F.

1 (ii)

The Reactor Coolant System is vented with a vent area equal to or greater than 47 in2 Exceptions All decay heat removal loops may be made 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 (1) no operations are permitted that would cause dilution of the reactor coolant system boron concentration, (2) no refueling operations are taking place, and (3) all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere are closed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

(5)

At least one reactor coolant pump or one low pressure safety mjection pump in the shutdown cooling mode shall be in operation whenever a change is being made in the boron concentration of the reactor coolant when fuel is in the reactor.

(6)

Both steam generators shall be filled above the low steam generator water level trip set point and available to remove decay heat whenever the average temperature of the reactor coolant is above 300 F. Each steam generator shall be demonstrated operable by performance of the inservice inspection program specified in Section 3.17 prior to exceeding a reactor coolant temperature of 300*F.

(7)

Maximum reactor coolant system hydrostatic test pressure shall be 3125 psia.

A maximum of 10 cycles of 3125 psia hydrostatic tests are allowed.

/

(8)

Reactor co( ant system leak and hydrostatic test (all be conducted within the limitations the PTLR-Figure 21 A and 2 IB.

(9)

Maximum secon$ry hydrostatic test pressure shall not exceed 1250 psia. A minimum measured temperature ot 73 F is required. Only 10 cycles are permitted.

(10)

Maximum steam generator steam side leak test pressure shall not exceed 1000 psia. A minimum measured temperature of u

(11) ylreactor coolant pumps are operating, on-operhting reactor coo ant pump). hall not be started while T, is belos 3 the temperatdie listed in the PTLR untess at least one of the following c Ps e nh 7 h 2-2a Amendment No.

6,66,71,119,136,

,v v

- 2,0 LIMITING CONDITIONS FOR OPERATION 2,1 Reactor Coolant System (Continued) -

2.1.1 Operable Components (Continued)

(a)

A ininlinuni pressurizer steam space as'specified~in"ths PTLR ef-534 s

by"vclu:nfor gr=ter (50.65? cr !=netual-levelfexists, or '

" ~

(b)

The temperatute differeride between'ths steam generator secondary side and'the reactor. coolant system cold le magnitudep:;;ified in th'e PTLR 30'g tc:nperature is les

~

. above th:t of the reactciecclant syst =.t c...

(12)

Reactor Coolant ystem Pressurc Isolation Valves (a)

The integrity of all precure isolation valves listed in Table 2.9 shall be demonstrated, except as specified in (b). Valve leakage shall not exceed the amounts indicated.

(b)

In the event that the integrity of any pressure isolation valve specified i

in Table 2-9 cannot be demonstrated, reactor operation may continue, provided that at least two valves in each high pressure line having a nonfunctional valve are in and remain in the mode corresponding to the isolated condition. Manual valves shall be locked in the closed po;.ition; motor operated valves shall be placed in the closed position and power supply deenergized.

(c)

If Specifications (a) and (b) above cannot be met an orderly shutdown shall be initiated and the reactor shall be in the cold shutdown condition L

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

Basis The plant is designed to operate with both reactor coolant loops and associated reactor coolan, pumps in operation and maintain DNBR above 1.18 during all normal operations and anticipated transients.

In the hot shutdown mode, a single reactor coolant loop,provides sufficient heat removal capability for removing decay heat; however, smgle failure considerations require that two loops be operable.

l In the cold shutdown mode, a smgle reactor coolant loop or shutdown cooling loo?

provides sufficient heat removal capability for removing decay heat, but single fai ute considerations require that at least two loops be operable. Thus, if the reactor coolant loops ur not cperable, this specification requires two shutdown cooling pumps to be operable.

The requirement that at least one shutdown cooling loop be in operation during refueling ensures that: (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the r: actor pressure vessel below 210"F as required during the refueling mode, and (2) sufficient coolant circulation is maintained through the reactor core to minimize the effects of a boron dilution incident and prevent boron stratification.

2-2b Amendment No. 56, Order 4/20/81,70, 77,92,161

- ~ - - ' --- - - - ~ - - - - - - - - ' - - - " ~ -

~

U 2.0' LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.1 Ooerable Comnonents (Continued)

The requirement to have two shutdown cooling pumps perable when there is less than 15 feet of water above the core ensures that a s, gle fai ure of the operating shutdown m

cooling loop will not result in a complete loss of decay heat removal capability. With j

the reactor vessel head removed and 15 feet of water above the core, a large heat sink is available for core cooling; thus, in the event of a failure of the operating shutdown cooling loop, adequate time is provided to initiate emergency procedures to cool the Core.

The restrictiore on availability of the containment spray pumps for shutdown cooling service ensure that the SI/CS pumps' suction header pip, g is not subjected to an m

unanalyzed condition in this mode. Analysis has determined that the minimum required RCS vent area is 47 in2 Thi; requirement may be met by removal of the pressurizer manway which has a cross-sectional area greater than 47 in2 When reactor coolant boron concentration is being changed, the process must be uniform throughout the reactor coolant system volume to prevent stratification of reactor coolant at lower boron concentration which could result in a reactivity insertion. Sufficient mixing of the reactor coolant is assured if one low pressure safety inhetion pump or one reactor coolant pump is in operatian. The low pressure safety injection pump will circulate the reactor coolant system volume in less than 35 minutes when operated at rated capacity. The pressurizer volume is relatively inactive; l

therefore, it will tend to have a boron concentration higher than the rest of the reactor coolant system during a dilution operation. Administrative procedures will provide for use of pressurizer sprays to maintain a nominal spread between the boron concentration i

)

in the pressurizer and the reactor coolant system during the addition of boron.m

. Both steam generators are required to be filled above the low steam generator water level trip set point whenever the temperature of the reactor coolant is greater than the design temperature of the shutdown cooling system to assure a redundant heat removal s stem for the rt actor, wN The LTOP enable temperature is' documented in the PTLR h:: been : tdliched at T,-

3862F. The pressure transient analyses demonsfrate that a single PORV is capable of mitigating overpressure events. Additional uncertainties have been applied to the Pressure-Temperature (P-T) limits to account for the case where a PORV is not

. availablei} (T,> 385 F) which is ec scesen for ec disceminshy in thc P-T rigsrcs.

m.s. m., ~.... m..

m..~

,m.. w....

m.m

.m. ur. o_.

mm The design cyclic ansients for the reactor system are given m USAR Section 4.2.2.

In addition, the steam generators are designed for additional conditions listed in USAR Section 4.3.4. Flooded and pressurized conditions on the steam side assure minimum tube sheet temperature differential during leak testing. The minimum temperature for pressurizing the steam generator steam side is 70 F; in measuring this temperature, the instrument accuracy must be added to the 70 F; limit to determine the actual measured limit. The measured temperat"re limit will be 73 F based upon use of an instrument with a maximum inaccuracy of.i.2*F and an additional 1"F safety margin.

2-2c Amendment No. 55,4/SI/Ordct.71,136,M1-

q 2.0 LIMITING CONDITIONS FOR OPERATION i

2.1 Reactor Coolant System (Continued) 2.1.1 Operable Comnonents (Continued) 7 Formation o a M cam spa of the magnitude specified in the PTLR e ures that the resulting p re increase Id not esult 'n an ryc.

ssur' < tion ould the nrst reactor coolant pump be starte en t.

geiefato econ side temperature is greater than that of the RCS cold leg. The steam space requirement is not applicable to the start of a reactor coolant pump if one or more pumps are in o rat n.

For the case in which the pressurizer steam space is less than the magnitude specified I

in the PTLR M. limitation of the steam generator secondary side /RCS cold leg AT to less than thi *nagnitude specified in the PTLR-302F ensure that

(

P V woul prevent a!Loverpre?surization et n

e first reactor c,oolant pun s

5tM nn' i

ta start of a reactor coolant pump if one or more pumps are operating.

The exception to Specincation 2.1.l(4) requiring all containment penetrations providing direct access, from the containment to the outside atmosphere be closed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> requires that the equipment hatch be closed and held in place by a minimum of four bolts.

References (1) USAR Section 4.3.7 2-2d Amendment No. 55,4/81 Order,71,125,151

2.0 LIMITING CONDITIONS FOR OPERA flON 2,1 Reactor Coolant System (Continued) 2.1.2 IIeatup and Cooldown Rate Annlicability o the temperature change rates and pressure of the reactor coolant system O nective To specify limit!.g conditions of the reactor coolant system heatup and cooldown rates.

Specification Ti s or

>la+

ressure-shall-be-limi:cd during-plant-operation in neeerdanee with-Figure 2:1 A and 2-1B and as fellowe The combination of RCS pressure, RCS.

temperature 'and RCS heatup and cooldown rates shall be maintained within the limits specified in the PTLR and as designated below;

(+)a, llowable co inations of pressure and temperature (T)it lines as shosfor a specific

~

a si lo to de right of the applicable lim in t ie PTLR oiwFigure-24A.

9)b llowah r

. of essure and tem $erature (T,) for a s ecific auld e righ of the applicable hmit lines as show infthe PTLR on-Figure 2 !B.

I

_J GE:'^ 1 e heatup rate of the pressunzer stiall not axceed 100'F in any one hour riod.

(4)d.1T e cooldown rate of the pressurizer shall not exceed 200*F in any one hour

~ ' '

riod.

Beauired7Abtions 61)) When any of the above limits are exceeded, the following corrective actions shall be taken:

(a)

Immediately initiate action to restore the temperature or pressure to within the limit.

(b)

Perform an analysis to determine the effects of the out of limit condition on the fracture toughness prope ties of the reactor coolant system.

(c)

Determine that the reactor coolant system remains acceptable for continued operation or be in cold shutdown within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

(6)

Before4he-rmliatietwxposure of-the reac:cr vessel execc2-the exposure for the-fellowing crb", ria-ar:d proc":rwFigures 21 A and 2 IB cha!! be upd which th3y-app!

e

-3 Amendment No. 22,71,.1

7 2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.2 Heatun and Cooldown Rate (Continued)

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All components in the reactor coolant system are designed to withstand the effeds of cyclic loads due to reactor coolanc system temperature and pressure changes.;) These cyclic loads are introduced by normal unit load transients, reactor trips and startup and shutdown operation.

m huring start and si down, e rates of temperature and pressure changes are limited. The design number of cycles for heatup and cooldown is based upon i

aliowable heatup.'cooldown rates and cyclic operation. Cycle ~ dependent inf6rmati6n sucl7as the pressure-temperature"(P-T) limit curves and low temperature: overpressure 3rotection (LTOP) system limits are contained in the Fort Calhoun Station RCS 1(

hessurer, Temperature Limits Report (PTLR)nwhich_was developed.using the'

[

methodologies of CE NPSD-683;Rev 029%

u

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2-4 Amendment No. 22,47,51,74,,i,100,114,161 i

. C..

i 2.0 LlhilTING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.2 Heatup and Cooldown Rate (Continued) mt_

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FORT CALHOUN STATIO4N UNIT 1 P/T LBIITS,20 EFFY 2500" 2500 75'F/HR%

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LOWEST y

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'F RCS Pressure-Temperature Omaha Public Power District Limits for Hestup Fort Calhoun Station-Unit No. I 2-

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e Limits for Cooldown Fort Calhoun Station-Unit No. I 2-Amendment No. 74, 77, 100, 176 1G1 l

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1nn If1 2-7 Amendment No.' S.9.,A ',T, v,A,",?,A,avu,av.

~

2.0 ilMITING CONDITIONS FOR QPERATION 2.1 Reactor Coolant System (Continued) d) 2.1.2 leatun and Cooloown Rate (Continue q!;::mp;m:ure is b;;d en :revicus NDIT method. Thi :::npen:are

_^

cnt ectr =b,jee: :e:c the :::=ured ;

0'F NDTT cf-the-rete :ct veze! flange, whieh-is no: r.u radictica inn measurem:nts, p!= 12'F i=g:, p!= 5'#Niata r,ca::er in NDTT

rumen: error..

I E.

Th: : ;n-m: = :: which :he hen:up and eccidown m::: chnge in Fi ures 21 A crf21B refice:: :he coc!dcwn m::: with :=pce:pein :: which the =c : !!miting heatup-ona

c the in! : ::mpem:u : (T,)..ange.

References:

(1)

USAR, Section 4.2.2 (2)

ASME Boiler and Pressure Vessel Code,Section III (3)

USAR, Section 4.2.4 (4)

USAR, Section 3.4.6 l

(5)

Omaha Public Power District, Fort Calhoun Station Unit No.1, Evaluation of Irradiated Capsule W-225, Revision 1, August 1980.

(6)

Technical Specification 2.3(3)

(7)

Article IWB-5000, ASME Boiler and Pressure Vessel Code, Section' XI (8)

Omaha Public Power District, Fort Calhoun Station Unit No.1, Evaluation of Erradiated Capsule W-265, March 1984.

(9)- s2Forf Calhoun Station'Unif N651 RCS Pressdrs7Tdnperssre LimitsRsport (10)') TCE NPSD483'ReW023"DeslopmsifofiRCS PresssFe'and Tejriperatdie

' Limits Report for the Removal of P-T. Limits and LTOP Requirements from the Techmcal;Sgcificationsf" December;1997; ' "

~ ~ ~ ~ '

2-7a Amendment No. 22,17,51,71,100,161 1

.r 2.0 alh1LTING CONDITIONS FOR OPERATION

- 3.1 3eactor Coolarit System (continued) Valves 2.1.6 2ressurizer and Main Steam Safety Acclit@ility Applies to the status of the pressurizer and main steam safety valves.

Obiective To specify minimum requirements pertaining to the pressurizer and main steam safety valves.

Specifications To provide adequate overpresst.c protection for the reactor coolant system and steam system, the following safety valve requhements shall be met:

(1)

The reactor shall not be made critical unless the two pressurizer safety valves are operable with their lift settings adjusted to ensure valve opening at 2500 psia 11% and 2545 psia 11 %.m (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 nozzled may be open to containment atmosphere during l

performance of safety valve tests or maintenance to satisfy this specification.

l (3)

Whenever the reador is in power o ration, eight of the ten main steam safety valves shall l

be operable with their lift settings a usted to ensure valves on each header opening at 1000-psia +3/-2%,1015 psia +3/-2%,1 25 psia +3/-2%,1040 psia +3/-2%, and 1050 psia t

l

+ 3/-2 %.m (4)

Two power-operated relief valves (PORVs) shall be operable during heatups and cooldowns when the RCS. temperature is less t'lan 51$ F, and in Modes 4Ed 5 whenever head is ctor vessel and the RCS is not vented through a 0.f 4 stl on em arge

, to L

rev,e,n,t s

'on of the pressure-temperature limits designate,4 by,in the PT.LR Fign = 21 A o

..,w.

a.

ith one PORV inoperable during heatups and cocidowns when the RCS tem rature is less than 515'F, restore the ino wrable P'RV to operable within 7 days or in cold shutdown within the next 36 Tours ano w>ressurize and vent the RCS through at least a 0.94 square inch or larger vent within the following 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, b.

With both PORVs inoperable during heatups and cooldowns when the RCS temperature is less than 515"F, be m cold shutdown within the n( tt 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> and depressurize and vent the RCS through at least a 0.94 square inch or larger vent within the following 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

T With one P'ORVMe in Modes 4 or 5, within one hour ensure the pressurizer t

PTLR pace is greater than1the: minimum: volume foERCP startup as specified in the steam s 3% volume (50.6Tc 1c= ac:udidt!)"and res ~ the moperable PORV to

~

rahle w,ithi ~/ days ate mm t be estabhh mu o

1 4 GFu ws or e inoperable V to operable within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. If the PORV cannot be restored in the required time, depressurize and vent the RCS through at least a 0.94 square inch or larger vent within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

2-15 Amendment No. 39,47,51,145,161 o

2.0 LIMITING CONDITIONS FOR OPERATION 2.2 Chemical and Volume Control System 2.2.1. Boric Acid Flow Paths - Shutdown Apolicabilily Applies to the operational status of the boric acid flow paths in MODES 4 and 5 when fuel is in the

reactor, Objective To assure operability of equipment required to add negat:ye reactivity.

Specification As a minimum, one of the following boric acid flow paths from an OPERABLE borated water source shall be OPERABLE:

a.

A flow path from boric acid storage tank Cil-11 A via cither a boric acid transfer pump or a i

gravity feed connection and a rht.rging pump to the Reactor Coolant System.

b.

A flow path from boric acid storage tank Cil 11B via either a boric acid transfer pump or a gravity feed connection and a charging pump to the Reactor Coolant Syctem.

A flow path from both boric acid storage tanks (Cil-1l' A and Cll-11B) via either a boric acid c.

transfer pump or gravity feed connection and a charging pump to the Reactor Coolant System.

w d.

A flow path from the SIRW tank via either a charging pump or a high pressure safety injection-pump to the Reactor Coolant System. The flow path from the SIRW tank 15 the Reactor Coolant System via a single.IIPSI pump shall only be crablished if the requirements in the PTLR.are

met, ni (1)

With none of the above boric acid flow paths OPERABLE, suspend all operatior.s involving CORE ALTERATIONS or positive rc.:tivity changes.

2-17 Amendment No.m. W

~.

i 2.0 LIMITING CONDITIONS FOR OPERATION 2.3 Chemical and Volume Control System (Continued) 2.2.3 Charging Pumps - Shutdown Applicability i

Applies to the operational status of charging pumps in MODES 4 and 5 when fuel is in the reactor.

- Obiective To assure operability of equipment required to add negative reactivity.

J-T At least one charging pump or one high pressure safety injection pump in the boric acid flow path required to be OPERABLE pursuant to Specification 2.2.1 shall be OPERABLE. The flow path from the SIRW tank to the Reactor Coolant System via a single _HPSI pump shall.only be established if.the requirements in the PTLR are met:

(1)

With no charging pump or high pressure safety injection pump OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

2-19 Amendment No. 131,172

a a-=.,

2.0 LIMITING CONDITIONS FOR OPERATION 2.2 Chemical and Volume Control System (Continued)

Basis (ContiDRdl Charging IMmps

Whenever the reactor coalant temperature (T d is greater than or equal to 210"F, two charging pumps must o

be operable in order to ensure it is po,sible to inject concentrated boric acid into the reactor coolant system with an assumed single failure. With only one pump operable, 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is allowed to restore the system to two operable charging pumps. This is consistent with the allowed outage time for the borated water sources and flow paths required during these modes.

In Modes 4 and 5 when fuel is in the reactor, Only one charging pump or high pressure safety injection pump must be operable. This is consisten. with the number of operab:e borated water sources and flow paths required during these m '. A nump is required in order to complete an operable flow path to the reactor c

a are additional

'ctions on the use of high pressure safety injection pumps contained the FI'LR Tcchnical SpecM";caticr. 2.3 t( ensure that the reactor vessel is not overpressurized.

Figure 12 contains a F bias to account for temperature measurement uncertainty An administrative procedure to monitor the temperature of the BASTS and boric acid system piping in the Auxiliary Building ensures that the temperature requirements of Figure 2-12 are met. Should the system temperature be unacceptable for operation at the current boric c::id concentration, steps will be taken to reduce the boric acid concentration or raise the temperature of the system such that the concentration is within the acceptable range.

of Figure 2-12.

The limits on component operability and the time periods for inoperability were selected on the basis of the redundancy indicated above and NUREG-0212 Revision 2. The allowed outage times for the various comnonents are consistent such that a support system has the same allowed outage time as the supported system.

-References (1) USAR Section 9.2 2-19h Amendment No. H2

2.0 LIMITING CONDITIOh'S FOR OPERATION 2.3 Emergency Core Cooling Sys1gni(Continued)

(3)

Protection Acainst Low Tgmnerature Overpressurization h _

Limits Re,oit-(PTLR) at.d shall be Adhered to.Dif not in compl Operabilitp?requirsments for HPSI pumps'are provided3n the RCS Pressurs-Teinpe'rature~

shall be tahn immediately to restore com i

e~._.u.a..., x._u.. u: u.,.,,.,.u... u :.. 2s..u:.u.pl,iance to the,PTLR; The e!!st n# !!=i:insw... u.

,,,e. s.. u..o e,

r t._=.. :., = =

.. ~.

m.

..~ _ u_ -.

pumpr n=d net be required if t'r RCS is vented Wrcugh : 1:=: 0.91 sq=r: inch or larger veed -

"!he=_ver :he reae:ct ecc!=: sy,:::n-ccid !cg ::mpera:ure is 5:!c./ 385'F, a: ic=: cne (13 u..n.,e,r...n p.... n.. s...u.,_ u..a..

a.

m

"'h=:ver me :=c:cr coc!=: sy ::= celd !:g ::=pem:ur: !- 5:!cw 320 F, a: !=:: two (2)

HPS! pumps da!! be di';;6!cd.

"/h= ver Se re=:ct ecc!=: :y;:= cold !cg : mpera:ure i: below 270'F, a!! crec (3) "PS!

pt:mp'; da!! h d!=b!:dr l-In 1: ev=: that no ch=ging pumps =: t.pemb!: when 6: reacter cec!=: cy :=: cc!d 1:g

/

mpera:ure is below 270*F, a sing!c "PS! pump may be made apea6!c and,.:ili=d for baric neid inje;;ica :c $c core, cie ik ; raic re:.'rie:cd to ac grea:ct in 120 gpm.

(4)

Tristdium ioso inte r 1 Dodecahydrate During operating Modes 1 and 2, the TSP baskets shall contain a 110 ft' 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 within 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 heat the reactor coolant to near operating temperature by running the reactor coolant pumps. The reactor is then made critical by withdrawing CEA's and diluting boron in the reactor coolant. With this mode of start-up, the energy stored in the -

reactor coolant during the a pproach 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, t

2-22 Amendment No. 47,39,43,47,64,7L-77,

, us, a ss, a n a, a s e,1/1a v a,.,7 n (M 1 S *]

141 1 C *7 1*

e

I i

3.0 LIMITING CONDITIONS FOR OPERATION 2.3 timergewv Core coollem Sv3km (Continued)

References (1) USAR Section 14.15.1 (2) USAR. Section 6.2.3.1 (3) USAR, Section 14.15.3 (4) USAR, Appendix K (5) Omaha Public Power District's Submittal, December 1,1976

),.," '

" DELETED (7) USAR, Section 4.4.3 i

9 4

e 2-23b Amendment No. 47,64,M, +79

(.,-9

-wwy w.-

y 9

w y+7--M-'

N

-r~

W

• • M Y-T V

3.0 SURVEILLANCE MEOUIREMENTS 3.3 Reactor Coolant System and Other Components Sublect to ASME XI Iloller & Pressure Vessel Code Inspection and Testine Surveillance

~

~

Applicability Applies to in service surveillance of primary system components and other components subject to inspection and testing according to ASME XI lloller & Pressure Vessel Code.

Oblective To ensure the integrity of the reactor coolant system and other components subject to inspection and testing according to ASME XI Holler & Pressure Vessel Code.

1 Soccifications (1)

Surveillance of Le ASME Code Class 1,2 and 3 systems, except the steam generator tubes -

inspection, should be covered by ASME XI Boller & Pressure Vessel Code, n.

In-service insyction of ASME Code Class 1, Class 2, and Class 3 components, including app icable supports, and in-service testing of ASME Code Class 1, Class 2.

and Class 3 pumps and valves shall be performed in accordance with Section XI of the ASME Holler and Pressure Vessel Code, as required by 10 CFR Part 50, Section 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50, Section 50.55a (g)(6)(1),

b.

Surveillance of the reactor coolant pump flywheels shall be performed as indicated in Table 3-6.

c.

A surveillance propam to monitor radi

-induce s in leal and impact properties of the reactoyftes mate

. a i be mamtamed in acc wi 10 CFR Part 50 Appendix 11 "tExamination results shall be.ust.d to update the PTLRi (2)

Surveillance of Reactor Coolant Systen

/

a.

Periodic leakage testing

• on each valve listed in Table 2-9 shall be accomplished prior to entering the power operation mode every time the plant is placed in the cold shutdown

• To satisfy ALARA requirements, leakage may be measured indirectly (as from the performance of pressure indicators) if accomplished in accordance with approved procedures and supported by co,mputations showing that the method is capable of demonstrating valve compliance with the leakage criteria.

I 3-21 Amendment No. 45,75,104,142,157r1%

l

l v

5.0 ADMINISTRATIVE CONTROLS h

5.9.6 Reactor coolant System (RCS) Pressure - Temocrature Limits Report (PTLR)

I a.

Reactor Coolant System pressure and tem wrature limits for heatu, cooldown, low

[

temperature operation, criticality, and hyt rostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following: Technical Specifications 2.1.1, 2.1.2, 2.1.6, 2.2.1, 2.2.3, and 2.3.

b.

The analytical methods used to determine the RCS pressure and temperature limits and predicted radiation induced NDTT shift shall be those previously reviewt $ and approved by the NRC, specifically those described in the following documents:

1.

10 CFR 50.61, " Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events" 2.

10 CFR Part 50 Appendix G " Fracture Toughness Requirements" 3.

10 CFR Part 50, Appendix II, " Reactor Vessel Material Surveillance Program Requirements" 4.

Regulatory Guide 1.99, " Radiation Embrittlement of Reactor Vessel Materials,"

Revision 2, 05/88 5.

ASME Boller and Pressure Vessel Code Section 111, Appendix G, " Protection Against Nonductile Failure," 1986 Edition

.6.

CE NPSD-683, Rev. 02, " Development of a RCS Pressure and Temperature Limits Report for the Removal of P-T Limits and LTOP Requirements from the Technical Specifications," December 1997 c.

The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period

- (i.e., the number of EFPY used in the P 'I Limit /LTOP analyses) and for any revision or supplement thereto.

=

517b Amendment No.

7 U.S. Nuclear Regulatory Comission LIC-98-0013 ATTACHMENT B

,-----u--

DISCUS $10N, JUSTIFICATION AND NO SIGNIFICANT HAZARDS CONSIDERATION DISCUSSION AND JUSTIFICATION Omaha Public Power District (OPPD) is proposing revisions to the Fort Calhoun Station Unit No. 1 Technical Specifications (TS) in accordance with Generic Letter (GL) 96-03, "Relocatf on of the Pressure Temperature Limit Curves and Low Tempera *.ure Overpressure Protection System Limits," dated January 31, 1996.

The proposed changes are consistent with the recommendations of Combustion Engineering Owners Group (CE0G) Task 942, " Development of a RCS Pressure and Temperature Limits Report for the Removal of P-T Limits and LTOP Requirements from the Technical Specifications," CE NPSD-683, Rev. 02 December 1997.

Accordingly, OPPD is submitting Topical Report CE NPSD-683.

Rev. 02 for NRC approval.

The pressure-temperature (P-T) curves (Hgures 2-1A & 2-1B), the predicted radiation induced NDTT shift curve (Figure 2-3) and the low temperature overpressure protection (LTOP) limits (TS 2.3(3)) are proposed for relocation to a document entitled " Fort Calhoun Station Unit No. 1 Reactor Coolant System (RCS) Pressure-Temperature Limits Repor+ (PTLR)." To ensure that the RCS is not overpressurized when the plant is in Mode 4 or Mode 5 with fuel in the reactor, TS 2.2.1.d and TS 2.2.3 are seing revised so that PTLR requirements must be met when a flowpath from the safety injection refueling water (SIRW) tank to the RCS is established via a single HPSI pump.

A definition of the RCS PTLR is being added to the TS as well as a new administrative control (TS 5.9.6).

TS 5.9.6 establishes the scope of the PTLR, the NRC and ASME approved analytical methods utilized in the PTLR, and requirements for submitting PTLR revisions. Additional administrative revisions are also proposed, which include relocating certain specific values (e.g., minimum pressurizer steam space) to the PTLR, relocating mn t of the Basis of TS 2.1.2 to the PTLR, and adding the FCS PTLR and Topical Report CE NPSD-683 Rev. 02 as references in TS 2.1.2.

It should be noted that Topical Report CE NPSD-683, Rev. 02 references the use of the ABB/CE codes ROCS /MC.

OPPD used the ROCS /MC codes during the time period (i.e.,1990) in which the attached PTLR heatup/cooldown limit curves were generated.

However, OPPD has upgraded to the CASM0/ SIMULATE codes as contained in Topical Report OPPD-NA-8302-P, Rev. 04, dated May 1994 (Reload Analysis Neutronics Methodology).

NRC approval of Topical Report OPPD-NA-8302-P, Rev. 04 is documented in a letter dated December 16, 1994 from S. D.

Bloom (NRC) to T. L. Patterson (OPPD).

Therefore, future use of the CASM0/ SIMULATE codes by 0 PPD is considered equivalent to the use of the ROCS /MC codes described in Topical Report CE NPSD-683, Rev. 02.

As required by GL 96-03, OPPD utilizes NRC approved methodology (ASME Section 111. Appendix G) to derive the parameters used to construct the P-T curves and LTOP setpoints.

The NDTT shift curve is also derived using NRC approved methodology (Regulatory Guide,1.99, Revision 2).

1

~.

LISCUSSION AND JUSTIFICATION (Continued)

The proposed amendment will reduce the burden on OPPD and NRC resources by eliminating the necessity of processing an amendment request each time a change is made to P-T limits, NDTT shift curves or LTOP setpoints.

Future changes to the PTLR will be controlled by the requirements of 10 CFR 50.59 (similar to the Core Operating Limits Report (COLR)) and will not normally require a license amendment to be effective.

2

BASIS FOR NO SIGNIFICANT HAZARDS CONSIDERATION:

The proposed changes to the Fort Calhoun Station (FCS) Unit No. 1 Technical Specifications (TS) are in accordance with Generic Letter (GL) 96-03,

Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits " dated January 31, 1996.

The proposed changes are also consistent with the recommendations of Combustion Engineering Owners Group (CEOG) Task 942, Development of a RCS Pressure and Temperature Limits Report for the Removal of P-T Limits and LTOP Requirements from the Technical Specifications," CE NPSD 683, Rev. 02, December 1997.

The proposed changes do not involve significant hazards consideration because operation of FCS in accordance with these changes would not:

(1)

Involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed changes relocate the reactor coolant system (RCS) pressure-temperature (P-T) curves, the predicted radiation induced NDTT shift curve and the low temperature overpressure protection (LTOP) limits to the Fort Calhoun Station Unit No. 1 RCS Pressure-Temperature Limits Report (PTLR).

Compliance with these curves and limits continues to be required by the Technical Specifications.

Changes to the curves and linnts will be controlled by TS 5.9.6, and must be in accordance with the NRC and ASME approved methodologies listed there and with 10 CFR 50.59.

The FCS PTLR in combination with the limitations imposed by the TS, will ensure the integrity of the reactor vessel pressure boundary.

Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

(2)

Create the possibility of a new or different kind of accident from any accident previously evaluated.

There will be no physical alterations to the plant configuration (no new or different equipment is being installed).

No changes in operating modes or limits are proposed.

The TS retain requirements to maintain the RCS within acceptable operational limits established in accordance with NRC and ASME approved methodologies and assure operability of the LTOP system. As such, the TS will continue to require compliance with the limitations being relocated to the FCS PTLR.

Therefore, these proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

3

I

l BA$15 FOR N0 5IGNIFICANT HAZARDS CONSIDERATION (Contiteued):

(3)

Involve a significant reduction in a margin of safety.

This proposed change to the FCS TS is administrative in nature

(

i relocating the P-T curves NDTT curve LTOP limits and associated TS requirements to the FCS PTLR in accordance with GL 96-03.

Future updates of the FCS PTLR will be conducted under the-10 CFR 50.59 process-utilizing NRC and ASME approved methodologies (as described in FCS Unit i

No. 1 PTLR, Rev. O and CEOG Task 942, Report CE NPSD-683, Rev 02)..

j Therefois. the proposed changes do not involve a significant reduction i

in a margin of safety.

Based on the above considerations, the proposed amendment does not involve significant hazards considerations as defined by 10 CFR 50.92 l

and the proposed changes will not result in a condition which significantly alters the impact of the Station on the environment.

Thus, the proposed changes meet the eligibility criteria for categorical j

exclusion'setforthin10CFR51.22(c)(9)andpursuantto10CFR 51.22(b) no environmental assessment need be prepared.

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O.S. Nuclear Regulatory Comission LIC-98-0013 ATTACHMEhT C

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