Text

Proposed TS Figures 2-1A & 2-1B Re pressure-temp Limits for Heatup & Cooldown,Respectively
	ML20128C046
Person / Time
Site:	Fort Calhoun
Issue date:	02/01/1993
From:	; OMAHA PUBLIC POWER DISTRICT
To:	;
Shared Package
ML20128C034	List: LIC-93-0069, Forwards Suppl Application for Amend to License DPR-40, Changing TS to Implement Guidance in GL 90-06; ML20128C038; ML20128C046;
References
	NUDOCS 9302030292
	Download: ML20128C046 (30)
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{{#Wiki_filter:, U.S. Nuclear Regulatory Commission LIC-93 0069 ATTACHMENT A 9302030292 930201 PDR ADOCK 05000285 l P PDR

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i I. i . IEC11NICAL SPECIF.1 CAT 10NS - FIGURES ! TABLE OF CONTENTS i l o PAGE WlilCH FIGURE DESCRIlvrION FIGURE FOLLOWS i l t 11 TMLP Safety Limits 4 Pump Operations . . . . . . . . . . . . . . . . . . . . 1-3 12 Axial Power Distribution LSSS for 4 Pump Operations . . . . . . . . . . . 13 l ! Pressuce - Teme reahre udh & Meccm RCS Prest-Temp-Limitt-Heatup. . . . . . . . . . . . . . . . . . . . p...... 2-6 l 2-1A Pf t LSure - T.e mD e/ L te Limit 5 {crr Ceolc(eto q '

21.B RCS Press: Temp-Limits tooldown. . . . . . . . . . . . . . . . . . . . . . . . 2-6 23 Predicted RadiaJon Induced NDTT Shift . . . . . . . . . . . . . . . . . . . . 2-6 2 11 MIN BAST Level vs Stored BAST Concentration ............. 2-19 l

j 2-12 Boric Acid Solubility in Water - . . . . . . . . . . . . . . . . . . . . . . . . . 2 19 2 10 Spent Fuel Pool Region 2 Storage Criteria . . . . . . . . . . . . . . . . . . 2-38 l I j( 28 Flux Peaking Augmentation Factors . . . . . . . . . . . . . . . . . . . . . . 2-53 ,

-s.

4 I i e i l l 1 l I 1 i l 4 i i t viil Amendment No. H6,+26dM:444-5 _ - . - - ...- . - - - . - - . . . . - , - = . - . - - . . ~ - . . . - - - -

2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System

   /           2.1.1         Operable comoonents Tron          (Continued) tinued) 43 (a) g athl(atl)A pressurizer exists, or steam space ofM by volume or gre~steho.fo % or P 30,f (b) The steam generator secondary side temperature is less than,$B*Tabove that of the reactor coolant system cold leg.                                                   .

(12) Reactor Coolant System Pressure Isolation Valves (a) The integrity of all pressure 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 4 valve specified in Table 2-9 cannot be demonstrated. reactor operation may continue, provided that at least two valves in each high pressure line having a nonfunc-tional valve are in and remain in the mode corresponding totheisolatedconditionq (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 c.ondition within 24 hours.' Basis The plant is designed to operate with both reactor coolant loops and associated reactor coolant pumps in operation and maintain DN8R 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, single failure considerations require that two loops be operable. In the cold shutdown mode, a single reactor coo

ant loop or shutdown cooling loop provides sufficient heat removal capability for removing decay heat,

> but single failure considerations require that at least two loaps be operable. Thus, if the reactor coolant loops are not operable, this specification requires two shutdown cooling pumps to be operable. The requirement refueling ensuresthat that:at least one shutdown cooling loop be in operation during (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 210'F as required during the refueling mode, and (2) cuffisient coolant circulation is maintained through the reactor core to minimize the effects of a boron dilution incident and prevent boron stratification.

            ~
                 / Manual valves shall be locked in the closed position; motor operated valves shall be placed in the closed position and power supp1fyd deenergized.

Amendment No. 36, 9/#/ f/29/BI, 70, 2-2b by I// M

j . . l . . 2.0 LIMtTINr> CONDITTONS FOR OPEPATION 2.1 Reactor Coolant System (Continued) 2.1.1 grerable Comconents (Continued) i The requirement to have two shutdown cooling pumps operable when there is less than 15 feet of water above the core ensures that a single failure of the operating shutdown cooling loop will not result in a complete loss of decay heat removal capability. With 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 care. The restrictions on availability of the containment spray pumps for I shutdown cooling service ensure that the SI/CS pumps' suction header 1 piping is not subjected to an unanalyzed condition in this mode. I 2 Analysis has determined that the minimum required RCS vent area is 47 l i n . +quarS4nches,--4he-pres sur4aer-manwayespec4f44d-4 s--the-minimum vent-area 40-414ew-vent 4e -through4he4imit4ng-cross-sect 4enal-eree- ll o{ W prmuryer mewq wguch h% aef-the-pressteteeMur et cen h aT m t eene ine. T his re.yH Mnrquem# h, mq be When reactor coolant boron concentration is being changed,ghW the process must be uniform throughout the reactor coolant system volume to prevent stratification of reactor coolant at lower baron concentration which could result h a reactivity insertion. Sufficient mixing of the reactor coolant is assured if one low pressure safety injection pump or one reactor coolant ts-assured-4f-onc low-pressure-safety M3ee4en-pump-or-one-reneter-coolant. pump is in operation. The low (' pre!,$ure tafety 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; therefore, it will tend to have a boron concentration higher than the rest of the reactor coolant system durirg a dilution operation. Administrative procedures will provide for use of pressurizer sprays to maintain a-nominal the reactor:pread coolantbetween the boron system during concentration the addition of boron.\ -in the pr9p'urizer and Soth 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 Jhutdown cooling system to assure a redundant heat cemoval system for the reactor.

Tfd. SERT 1 --h The design cyclic transients for the reattor system are given in USAR Section 4.2.2. In adctition, the steam generators are designed for additional conditier..s listed in USAR Section 4.3.4 Flooded and pressurized conditions on the steam side assure minimum tube sheet temperature difierential during leak testing. The minimum temperature for pressuri7ing 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 temperature limit will be 73'F based upon usa of an instrument with a maximum inaccuracy of       2'F and an additional l'F safety margin.

2-2c Amendment No. EB, //EI/0t h t, 71, 136

                                                                      -                        . - _ -     _ _ _ _ _         i

s e e i i INSERT I FOR T.S. 2.1.1 BASIS The LTOP enable temperature has been established at T, - 385'F. The pressure transient analyses demonstrate that a single PORV is capable of mitigating overpressure events. Additional uncertainties have been applied to the Pressure- ; limits to account for the case where a PORY is not available Temperature (P- T)is the reason for the discontinuity in the P-T Figures. (T,>385'F) which The curves have been conservatively smoothed for operations use. i 4

i s i s' j FORT CALHOUN STATION (J ~ r 1 P/r LIMITS,20 EFPY i

COOLDOWN AND INSERVICE TEST l 2500 f j-2500 j INSERVICE LIYDROSTATIC TESTg .

l l i i 2000 2000 l ? ! e 1500 1500 } m LOWEST SERVICE ! . -100'F/HR 9 m wEaw RE i i i E a: 182'F \ ALLOWABLE COOLDOWN RATE i N I TEMR LIMIT. 'F - RATE.'F/HR I j E 1000 , < 135 10 0 135-285 30

                                                                                                                          > 285                   100-

{~ l j 500

                                               -A$                                                                                         ~

500 j 10'F/HR 30*F/HR

                                                           /
                                                                 /

100'F/HR s f fMIN. BOLTUP TEMP,82*F s 0 0 100 200 300 400 - 500 600 4 . Tc INDICATED REACTOR COOLANT SYSTEM TEMPERATURE, 'F -

                   ~ RCS Pressure-Tempemture                                            Omaha Public Power District                                                           Figure Limits for Cooldown                                          Fort Calhoon Station- Unit No. I                                                      2-1 B
                                               .,,.,#.       -.,           .a w.,   ,-.-,.,,-,,-i,;.-,-t--..r.                --w~.-,.r...   .%-.,,,,,-#...m._,m-w                   -.3 - , ,

y.

2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.2 Tiestup and Coofdown Curves (Continued)

$ PR Kg=Hg7 Mg = ASME III, Figure G-2214-1 P = Pressure, psia , R = Vessel Radius - in, t = Vessel Wall Thickness - in. , r,77 =-MT #~IE MT TV MT = ASMC III, Figure G-2214-2 ATg = Highest Radtal Temperature Gradient Through Wall at End of - Cooldown is therefore calculated at a maximum gradient and is considered K a [onstant = A for cooldwn and-:re for heatup. g l p R is also a constant = B. Therefore: XIR = AP + B

i P=XIR

B A

    .                   Esegy        Kyi s-then-ve ri ed-a s-a-fu nc tion-o f- tempe ra tu re-f rom 4( g u re-C - %10-1-of-ngM 3 *mE-III-and-the-alloweble-pressure-calculatedr-Hydrostati t-head-(48-ps44

! and-instrumentation-errors-(12'F-and-32-psi}-are-considered when-plotting-t the-curves. l Truuvid6

8. 4ystem intervl Hyilrostatic Test - The systew ky(d ostatic test curve is developed

! in the same manner as in A,tbove with the exception that a safety factor ! of 1.5 is allowed by ASME III in lieu' of 2. l C. Lowest Service Temperature = 50'F + 120*F + 12'F = 182*F. As indicated l previously, an RTNDT for all material with the exception of the reactor i vessel beltline was established at 50*F. 10 CFR Part 50, Appendix G, IV a.2. requires a lowest service temperature of RTNDT + 120'F for j piping, pumps and valves. Bel ow-thi s-tempera tu re-a-p res s u re-o f-20-pe r-7, peg cent-of-the-system-hydrostatic: test-pressure-{-20){S125) tS psi = Tren y "545-psia-cannot-be-taneeeded, 569 D. Boltup Temperature = 10*F + 60'F + 12*F = 82*F. At pressure below 445-psia, a minimum vessel temperature must be maintained to comply with the manufacturer's specifications for tensioning the vessel head. l l l 2-7 Amendment No. 22./7,5A 78,400-l l l i-

Text-to be Inserted in Technical Specification 2.1.2 Item 1. Allowable combinations of pressure and temperature (T,) for a specific heatup rate shall be below and to the right of the applicable limit lines as shown on Figure 21A. Item 2. For plant heatup the reference stress intensity is calculated for both the 1/4t and 3 locations. Composite curves are then generated for each he/4tatup rate by combining the most restrictive pressure-temperature limits over the complete temperature interval. Item 3. K, is then varied as a function of temperature from Figure G 21101 of ASME 111 and the allowable pressure calculated. Pressure correction factors for elevation and flow (-56 psia for T,<210'F and

              -62 psia for T,2210'F) and temperature instrumentation uncertainties

(+16'F) are considered when plotting the curves. Pressure instrumentation uncertainty is also considered above the LTOP enable temperature of 385'F . Below this temperature, pressure ' instrumentation uncertainty is accounted for in the LTOP PORV setpoints, item 4. Below this temperature a pressure of 20 )ercent of the system hydrostatic test pressure cannot be exceedec. Taking into account this pressure is pressure (.20)(3125 correction

                           - 56 = 569factors       psia, for                     elevation where   56 psiand  flow, hydrostatic head is the correction) factor.

i

2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (continued) 2.1.6 Pressurizer and MaJn Steam Safety Valves l Apolicability Applies to the status of the pressurizer and main steam safety valves. l Objective To specify valves. minimum requirements pertaining to the pressurizer and main Specifications To provide adequate overpressure protection for the reactor coolant system and steam system, the following safety valve requirements shall be met: (1) The reactor shall not be made critical unless the two pressurizer safety valves are operable with their lift settings adjusted to ensure valve opening at 2500 psia 11% and 2545 psiail%.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 nozzles m be open to containment atmosphere during performance of safety valve tests or maintenance to satisfy this specification. (3) Whenever the reactor is in power operation, eight of the ten main steam safet valves shall be operable with their lift settings adjusted to ensure valves on each header opening at 1000 psia +3/-2%,1015 psia +3/+2%,1025 ptla +3/ 2%, 1040 psia +3/ 2%, and 1050 psia +3/ 2%.m (4) Bothpressurieertower-operated-relief-valves-(PORVa s ) shall be operable-during

                       -scheduled-heatup-and<ooldown-to-prevenFv.lolation-of-thetressure-temperature 7tGCdT -->limitsdesignated-by-Figures 4-1A-and-2-1B-D]ni-PORV-mayhinoperable-for up--to+daysrprovided1heTemaining-PORV-is operablerlf thrabovrconditions of-thisTaragraph cannot-be metrthrprimarynystem mustkdepisned and-vented.

(5) Two power-operated relief valves (POR and their associated blcx:k valves shall be operable in Modes 1,2, and 3. 2 15 Amendment No. 39,4WS4-44b

   ',                           TNS6dT fCK               SP6CI f!M ricm) 2, /. 6 (4 )

(4)

Two power operated relief valves (PORVs) shall be operable during hcatups and cooldowns when the RCS temperature is less than $15*F, and in Modes 4 and 5 whenever the head is on the reactor vessel and the RCS is not vented through a 0.94 square inch or larger vent, to prevent violation of the pressure temperature limits designated by Figures 21 A and 2-18.

a. With one PORV inoperable during heatups and cooldowns when the RCS temperature is less than 515'F, restore the inoperable PORV to operable within 7 days or be in cold shutdown within the next 36 hours and depressurize and vent the RCS through at least a 0.94 square inch or larger vent within the following 36 hours.

b. With both PORVs inoperable during heatups and cooldowns when the RCS temperature is less than 515'F, be in cold shutdown within the next 36 hours and depressurize and vent the RCS through at least a 0.94 square inch or larger vent within the following 36 hours.

c. With one PORV inoperable in Modes 4 or 5, within one hour ensure the pressurizer steam space is greater than 53% volume (50.6% or less actual level) and rettore the inoperable PORV to operable within 7 days. If adequate steam space cannot be established within one hour, then restore the inoperable PORV to operable within 24 hours. 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 hours,

d. With both PORVs inoperable in Modes 4 or 5 depressurize and vent the RCS through at least a 0.94 square inch or larger vent within the next 36 hours.

{ 2.0 1AUTI.NG CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (continued) 2.1.6 Pressurizer and Main Steam Safety Valves (continued) l aQVith one or more PORV(s) inoperable, within I hour either restore the PORV(s) SCE tolperable status or close the associated block valve (s); otherwise, be in at least HOT STANDBY-within the next 12 hours and in COLD SHUTDOWN with'n the UTM following 24 hours.

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b. With one or tr. ore block valve (s) inoperable, withinQour either restore the block valve (s) to operable status or close the block valve (s)N>thentise, be in at least HOT STANDBY within the next 12 hours and in COLD SHUTDO%R within the following 24 hours.

Eull The highest reactor coolant system pressure reached in any of the accidents analyzed resulted from a complete loss of turbine generator load without simultaneous reactor trip while operating at 1500 MWt.* This pressure was less than the 27b0 psia safety limit and the ASME Section HI upset pressure limit of 10% greater than the design pressure.m The reactor is asst.med to trip on a "High Pressurizer Pressure" trip signal. ( The power-operated relief valves (PORV operate to relieve RCS pressu.re below the setting of the pressurizer code safety valves. These relief valves have remotely operated block valves to provide a positive shutoff capability should a relief valve become inoperable. The electrical power for both the relief valves and the block valves is capable of being supplied from an emergency power souren to ensure the ability to seal this possible RCS leakage path. Su ATn1CHED -- r To determine the maximum steam flow, the only other pressure relieving system assumed operational is the main steam safety valves. Conservative values for all systems l parameters, delay times and core moderator coefficient: are assumed. Overpressure protection is provided to portions of the reactor coolant system which are at the highest pressure considering pump head, flow pressure drops and elevation heads. If no residual heat were removed by any of the means available, the amount or steam which could be generated at safety valve lift pressure would be less than half of the capacity of one safety valve nis specification, therefore, provides adequate defense against ovemressurization when the reactor is suberitical. 2-15a Amendment No.14-146.

2.0 LI3flTING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (continued) 2.1.6 Pressurizer and Afain Steam Safety Valvn (continued) l Performance of certain calibration and maintenance procedures on safety valves requires removal from the pressurizer. Should a safety valve be removed, either operability of the other safety valve or maintenance of at least one nozzle open to atmosphere will assure that sufficient relief capacity is available. Use of plastic or other similar material to prevent the entry of foreign material into the open nozzle will not be construed to violate the "open to atmosphere

provision, since the presence of this material would not significantly restrict the discharge of reactor coolant.

The total relief capacity of the ten main steam safety valves is 6.54 x 10'lb/hr. If, following testing, the as found setpoints are outside +/.1 % of nominal nameplate values, the valves are set to within the +/.1% tolerance. The main steam safety valves were analyzed for a total loss of main feedwr.ter flow while operating at 1500 MWr* to ensure that the peak secondary pressure was less than '00 psia, the ASME Section III upset pressure limit of 1095 greater than the design pressure. At the power of 1500 MWt, sufficient relief valve capacity is available to prevent overpressurization of the steam system on loss-of load conditions.* The power-operated relief valve low setpoint will be adjusted to provide sufficient margin, when used in conjunction with Technical Specification Sections 2.1.1 and 2.3, to prevent the design basis pressure transients from causing an overpressurization incident. Limitation of this requirement to scheduled cooldown ensures that, should emergency conditions dictate rapid cooldown of the reactor coolant system, inoperability of the low temperature overpressure protection system would not prove to be an inhibiting factor. 77tc e &che O // /h; 6 nN o/ oo Pc gv is c. <t 4 ;,, .* of e ie Removal of the reactor vessel head provides sufficient exprnsion volume to limit any of the design basis pressure transients. Thus, no additional relief capacity is required. References (1) Article 9 of the 1968 ASME Boiler and Pressure Vessel Code, Section III (2) USAR, Section 14.9 (3) USAR Section 14.10 (4) USAR, Sections 4.3.4, 4.3.9.5 2 16 Amendment No. 39,4-7,W146-

72 ext T~o BE NDMO To SPL~cif~te a ncW 2,i. n (c)

a. With one or both PORV(s) inoperable because of excessive seat leakage, within 1 hour either restore the PORV(s) to operable status or close the associated block valve (s) with power maintained to the block valve (s); otherwise, be in at least HOT SHUTDOWN 1 within the next 6 hours and in COLD SHUTDOWN within the following 36 hours.

b. With one PORV inoperable due to causes other than excessive seat leakage, within 1 hour either restore the PORV to operable status or close its associated block valve and remove power from the block valve; restore the PORV to operable status within the following 72 hours or be in HOT SHUTDOWN within the next 6 hours and in COLD SHUTDOWN within the following 36 hours, i c. With both PORVs inoperable due to causes other than excessive seat leakage, within 1 hour either restore at least one PORV to operable status or close both block valves, remove power from the block valves, and be in HOT SHUTDOWN within the next 6 hours and in COLD SHUTDOWN within the following 36 hours.

l

d. With one or both block valve (s) inoperable, within 1 hour restore the block valve (s) to l operable status or place the associated PORV(s) in the closed position. Restore at leact one block valve to operable status within the next hour if both block valves are inoperablel restore the remaining inoperable block valve to operable within 72 hours, i Otherwise, be in at least HOT SHUTDOWN within the next 6 hours and in COLD I SHUTDOWN within the following 36 hours.

1bxT Yo BE ADDEt) 7o~ i E/?S/S Of" SREC/F(0AT/CH & /. (c

Action statements (5)b. and c. include the removal of power from a closed block valve } inadvertent opening of the block valve at a time the PORV may not be closed due to maintenance. However, the applicability requirements of the LCO to operate with the block valve (s) closed with power maintained to the block valve (s) are only intended to permit operation of the plant for a limited period of time not to exceed the next refueling sSutdown (Mode 5), so that maintenance can be performed on the PORV(s) to eliminate the seat leakage condition.

1 2.0 LIMITING CONDITIONS FOR OPERATION 2.3 Emercency Core Cooline System (Continued) l l (3) Protection Acainst low Temocrature Overoressurization j l The following limiting conditions shall be applied during scheduled heatups and ; cooldowns. Disabling of the HPSI pumps need not be required if the reaetor 1 a

                               -vessel-headrjrecturim       =fetyxalverw+PORV is4cmoved. RCS is verfr ed _           ;

4heogh

n eeT a cA4 space inch or larg e r var.

Whenever the reactor coolant system cold leg temperature is below32&F, at least l one (1) HPSI pump shall be disabled. L85'F Whenever the reactor coolant system cold leg temperature is below 3128F3 at least two (2) HPSI pumps shall be disabled. 3#04

                                                                                                        .2 'Io*F Whenever the reactor coolant system cold leg temperature is below Wl'P, all three (3) HPSI pumps slall be disabled.                   when the reacTer cehmT q cem <xW g gemper.date In the event that no charging pumps are                         operablf, gdNNI   pump may  a be l                               made operable and utilized for boric acid injection to the        f core dth f(cw mTe      restrETed to no 3recer              e r 4han e y m.

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 coclant. With this mode

              )       of start-up, the energy stored in the reactor coolant during the approach to criticality is   ,

substantially equal to that during power operation and therefore all engineered safety features and auxiliary cooling systems are required to be fully operable. During low power physics tests at low temperatures, there is a negligible amount of stored energy in the reactor coolant; therefore, an accident comparable in severity to tu design basis accident is not possible and the engineered safeguards systems are not required. The SIRW tank contains a minimum of 283,000 gallons of usable water containing a l boron concentration of at least the refueling boron concentration. This is sufficient boron i concentration to rovide a shutdown margin of 5%, including aliowances for uncertainties, with aJ wntrol rods withdrawn and a new core at a temperature of 60*F.m The limits for the safety injecuan tank pressure and volume assure the required amount of water injection during an accident and are based on values ; sed for the accident l analyses. The minimum 116.2 inch level corresponds to a ' alume of 825 ft' and the

maximum 128.1 inch level corresponds to a volume of 895.5 fe. Prior ta the time the

! reactor is brougcl critical, the valving of the safety injection system must be checked for ( correct alignment and appropnate valves locked. Since the system is used for shutdown I cooling, the valving will be changed and must be properly aligned prior to start-up of the reactor. 2-22 Amen <! ment No. I1,4-7,-39,+3,47,64, l 74,7-7,400,103,133rl4F-

l . l TECIINICAL SPECIFICATIONS - FIGURFS n i 1 TABLE OF CONTENTS J j i i . PAGE WIIICII l FIGURE DESCRIITLQE FIGURE FOLID%'S i 1 ! l-1 TMLP Safety Limits 4 Pump Operations . . . . . . . . , . . . . . . . . . . . . . . . . . . . . 1-3 1-2 Axial Power Distribution LSSS for 4 Pump Operations . . . . . . . . . . . . . . . . . . . . 1-3 2-1 A RCS Pressure-Temperature Limits for Heatup . . . . . . . . . . . . . . . . . . . . . . . . . . 2-6 l

j. 2-1B RCS Pressure-Temperature Limits for Cooldown . . . . . . . . . . . . . . . . . . . . . . . . 2-6 l i 2-3 Predicted Radiation Induced NDTT Shift . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-6 i .

i 2-11 MIN BAST Level vs Stored BAST Concentration ................ ,,.... 2-19

2-12 Boric Acid Solubility in Water . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-19 2-10 Spent Fuel Pool Region 2 Storage Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-38 2-8 Flux Peaking Augmentation Factors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-53 a

i a i i f i 4 i

viii Amendment No. 4 % ,126,131,141

                                                                                                                                +=rw+-

gf e 9 --?F-Nvmt-gty- -e e e

           .                                                                                                              I 2.0 '   LIMITING CONDITIONS FOR OPERATION 2.1     Reactor Coolant SystcB (Continued)                                                                        l 2.1,1 Operable Comoonents (Continued)

(a) A pressurizer steam space of 53% by volume or greater (50,6% or less actual level) l- y exists, or (b) The steam generator secondary side temperature is less than 30 F abou that of the l reactor coolant system cold leg. (12) Reactor Coolant System Pressure Isolation Valves (a) The integrity of all pressure 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 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 position; motor operated valves shall be placed in the closed position and the power supply deenergized. 2 (c) If Specifications (a) and (b) above cannot be met, an orderly shutdown shall be initiated j and the reactor shall be in the cold shutdown condition within 24 hours. Luis The plant is designed to operate with both reactor coolant loops and associated ieactor coolant pumps in operation and maintain DNBR above 1.18 during all normal operations and anticipated transients.' 4 In the hot shutdown mode, a single reactor coolant loop provides sufficient heat removal capability for removing decay heat; however, single failure considerations require that two loops be operable. In the cold shutdown mode, a single reactor coolant loop or shutdown cooling loop provides sufficient heat removal capability for removing decay heat, but single failure considerations reqeire that at least two loops be operable. Thus, if the reactor coolant loops are not operable, 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 reactor 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 i boron stratification. I 4 ,. 2 2b Amendment No. 56, Order 4/20/81,70, y , t

2.0 ' IJMITING CONDITIQNS FOR OPERATION 2.1 Reac19tCoolant System (continued) 2.1.1 Operable Comoone.0h (Continued) The requirement to have two shutdown cooling pumps operable when there is less than 15 feet of water above the core ensures that a single failure of the operating shutdown cooling loop will not result in a com;! te loss of decay heat removal capability. With 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 restrictions on evailability of the containment spray pumps for shutdown cooling service ensure that the SI/CS pumps' suction headt.r pipirig is not subjected to an unanalyzed condition in this mode. Analysis has determined that the minimum required RCS vent area is 47 in2 . This requirement may be met by removal of the pressurizer manway which has a cross-sectional area greater than 47 in2 . When reactor coolant boren concentration is being changed, the process must be uniform throughout 4 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 coolam is assured if one low pressure safety injection pump or one reactor coolant pump is in operation. The low l 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; therefore, it wiP. 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 in the pressurizer and the reactor coolant system & iring the addition of boron.m Both steam generators are required to be filled above the low steam generator water level trip set point l whenever the temperature of the reactor coolant is greater than the design temperature of the shutdown cooling system to assure a redundant heat removal system for the reactor. The LTOP enable temperature has been established at T, = 385'F. The pressure transient analyses demonstrate 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 available ( > 385 F), which is the reason for the discontinuity in the P-T Figures. The curves have been conservauvely smoothed for operations use. l The design cyclic transients for the reactor system are given in 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 steem 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 temperature limit will be 73 F based upon use of an instrument with a maximum inaccuracy ofi 2 F and an additional 1*F safety margin. 2-2c Amendment No. 56, 4/M/ Order,M FM t- .

i

2.0 LIMITING CONDITIONS FOR OPERATLQN 2.1 Reactor Coolant System (Continued)
2.1.2 Heatuo mad.Cooldown Rate (Continued) i

[ 1500 MWt and 80% load factor. The predicted shift at this location at the 1/4t depth from the inner j surface is 332*F, including margin, and was calculated using the shift prediction equation of Regulatory l . Guide 1.99, Revision 2. The actual shift in Tm will be re-established periodically during the plant > l operation by testing of reactor vessel material samples which are irradiated cumulatively by securing j them near the inside wall of the reacter vessel as described in Section 4.5.3 and Figure 4.5-1 of the ! USAR. To compensate for any increase in the Tm caused by irradiation, limits on the l pressure-temperature relationship are periodically changed to stay within the stress limits during heatup l and cooldown. Analysis of the second removed irradiated reactor vessel surveillance specimen *, ! combined with weld chemical composition data and reduced fluence core loading designs _ initiated in

Cycle 8, indicated that the fluence at the end of 20.0 Effective Full Power Years (EFPY) at 1500 MWt I will be 1.50x10" n/cm 2on the inside surface of the reactor vessel. This results in a total shift of the i RTum of 298"F, includiag margin, for the area of greatest sensitivity (weld metal) at the 1/4t location-l as determined from Figure 2-3, and a shift of 241*F at the 3/4t location. Operation through fuel Cycle

[ 19 will result in less than 20.0 EFPY. f The limit lines in Figures 2-1A and 21B are based on the fol!owing: e ' A. Heatup and Cooldown Curves From Section 111 of the ASME Code, Appendix G-2215. l l Km = 2 Km' + Krr L [ Ka = Allowance stress intensity factor at temperature

related to RTum (ASME III Figure G-2110.1).

4 i Km =. Stress intensity factor for membrane stress (pressure). i The 2 represents a safety factor of 2 on pressure. I Krr = Stress intensity factor radial thermal gradient. l The above equation is applied to the reactor vessel beltline. For plant heatup the reference stress ! intensity is calculated for both the 1/4t and 3/4t locations. Composite curves are then generated l for each heatup rate by combining the most restrictive pressure-temperature limits over the . j . complete temperature interval. . i For plant cooldown thermal and pressure stress are additive. t S I L 2-6 Amendment No. E,47,64,7'4,-7-7, li M0,114,121 4 i . . . . . . . . --

5 j . -. i ,. i a 1 i 4 I FORT CALHOUN STATION UNrr 1 P/r LIMITS,20 EFPY l 4 2500 2500 } 1.

00'F/HR~

j -75'F/HRN i I ) 2000 l 2000 , E c. 1500 f1500

o LOWEST SERVICE l l TEMPERA *IURE x

p 182'F - ' ALLOWABLE HEATUP RATES ! N 1000 TEMP. T TMIT'F RATE.*F/HR g l @ !. s335 75 100 y > 335 i x c. i / 75"F/HR i 500 500: ! 10C 'F/HR h' ' \

                                                           %. BOLTUP TEMR 82*F I'                                  0                                                                                                                                         0 0                  100-         200             300           400                          500                                 600 l                                                                                                                                                                                -

i . t-Te INDICATED REACTOR COOLANT SYSTEM TEMPERATURE. 'F i l, i , ' 1 l t l I t RCS Pressure-Temperature Omaha Public Power Distnct ~ ~ Figure Limits for Heatup Fort Calhoun Station- Unit No.1 2-1A-I ' Amendment No. 75, 77, 100,-fyg l

                                                                         ...s..       - -+ ._.            . s    _.-_a. . . . . . -       _ _ - .

}

  .n J

i FORT CALHOUN STATION UNIT 1 P/T LIMITS,20 EFPY

i. COOLDOWN AND INSERVICE TEST ,

1 2500 2500 { 7 i j INSERVICE FIYDROSTATIC TESTw a l- 2000 2000 i E LOWEST SERVICE [ X 100'F/HR- < $ TEMPERATURE

f cc

182'F \ ALLOWABLE COOLDOWN RATE I N ' TEMR I IMIT 'F RATE *F/HR l E 1000 '

                                                                                < 135                       10 U$                                                                      135-285                    30 1

8 cc j > 285 100 i l 500

                                   -d$                                                                            500-10*F/HR-30'F/IG "j        /

i 100'F/HR3

- 7 MIN. BOLTUP TEMP, 82

F l 0 -

l 0 LC0 200 300 400 500 600 Te INDICATED REACTOR COOLANT SYSTEM TEMPERATURE, 'F i r I i-l RCS Pressure-Tempemture Omah: Public Power District ! Limits for Cooldown Fort Calhoun Station- Unit No. I _Ik l Amendment No. 74, 77,100, f 7W j

2.0' LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.2 Heatun and Cooldown Curves (Continued) Ku = hiu 12 t hiu = ash 1E 111, Figure G 2214-1 P= Pressure, psia R = Vessel Radius - in. t = Vessel Wall Thickness - in. Krr = htTATw l h1T = AShlE III, Figure G-2214-2 AT, = Highest Radial Temperature Gradient Through Wall at End of Cooldown Krr is therefore calculated at a maximum gradient and is considered a constant = A for cooldown and heatup. l L R is also a constant = B. Therefore: Ku = AP t B P = K, - B A s 14 is then varied as a function of ternperature from Figure G-2110-1 of ASME III and the allowable pressure calculated. Pressuie correction factors for elevation and flow (-56 psia for T, <210 F and -62 psia for T, 2: 210'F) and temperature instrumentation uncertainties (+ 16*F) are considered when plotting the curves. Pressure instrumentation uncertainty is also considered above the LTOP enable temperature of 385 F. Below this temperature, pressure instrumentation uncertainty is accounted for in the LTOP PORV setpoints. B. Inservice Hydrostatic Test - The inservice hydrostatic test curve is developed in the same manner as in A. above with the exceptien that a safety factor of 1.5 is allowed by ASME III in lieu of 2. C. Lowest SerAce Temperature = 50 F + 120*F + 12*F = 182 F. As indicated previously, an RTwr for all material with the exception of the' reactor vessel beltline was established at 5&F. 10 CFR Part 50, Appendix G, IV.a.2. requires a lowest service temperature of RTwr + 120 F for piping, pumps and valves. Below this temperature a pressure of 20 percent of the system hydrostatic test pressure can not be exceeded. Taking into account pressure correction factors for elevation and flow., this pressure is (.20)(3125)-56=569 psia, where 56 psia is the hydrostatic head correction factor. D. Boltup Temperature = 10 F + 60 F + 12 F = 82 F. At pressure below 569 psia, a minimum vessel temperature must be maintained to comply with the manufacturer's spec tensioning the vessel head. 2-7 Amendment No. 24.47,64,-74dOO l

j . .- 4 i Predicted Radiation Induced NDTT Shift - i, e Fort Calhoun Reactor Vessel Beltline - i ARTndt

$00 3 1 i

i

! 430 i l l

                                        !.D. SHIFT l

Including Margin p # 350 /

If fI l

' /

                                                    /                              l/+t SHIFT Including Margin

.' 300 -

,/

250

                              -     2                                      #                         3/4t SHIFT f

j-

                                      /                         /r                                Including Marpn i                  200
                                 /               f I

A 1$0 , l 100 00 0.5 1.0 1.5 2.0 2.3 3,9 3,3 - 4g 4_3 3,9 2 Neutron Fluence, IE19 n/cm l l 1 i Predicted Radiation induced - Omaha Public Power District . Figure NOTT Shift Fort Calhoun Station-Unit No.1 2-3 Amendment No.- 74, 77, ygg,. yyg, yyy

4 1 . i 2.0 ' LIMITING CONDITIONS FOR OPERATION j' 2.1- Feactor Coolant System (Continued)- j- 2,1.6 - hessuriter and Main Steam Safety Valves l Applicability I j Applies to' the status of the pressurizer and main steam safety valves. 4 Obiective 4 , To specify minimum requirements pertaining to the pressurizer and main steam safety valves.

Spqgifications 6

! To provide adequate overpressure protection for the reactor coolant system and steam system, the - !- following safety valve requirements shall be met: (1) The reactor shall not be made critical unless the two pressurizer safety valves are operable with j their lift settings adjusted to ensure valve opening at 2500 psia i 1% and 2545 psia il%.* [ (2) Whenever there is fuelin the reactor, and the reactor vessel head is installed, a minimum of one i operable safety valve shall be installed on the pressurizer. However, when in at least the cold i shutdown condition, safety valve nozzles 'may be open to containment atmosphere during , performance of safety valve tests or maintenance to satisfy this specification. i (3) Whenever the reactor is in power operation, eight of the ten main steam safety valves shall be ! operable with their lift settings adjusted to ensure valves on each header opening at 1000 psia j +3/-2%,1015 psia +3/-2%,1025 psia +3/-2%,1040 psia +3/-2%, and 1050 psia +3/-2%.m (4) Two power-operated relief valves (PORVs) shall be operable during heatups and cooldowns , when the RCS temperature is less thart 515 F, and in Modes 4 and 5 whenever the head is on , the reactor vessel and the RCS is not vented through a 0.94 square inch or larger vent, to l prevent violation of the pressure-temper?ture limits designated by Figures'2-1 A and 2-1B.

a. With one PORV inoperable during heatups and cooldowns when the RCS temperature i is less than 515*F, restore the inoperable PORV to operable within 7 days or be in cold l shutdown within the next 36 hours and depressurize and vent the RCS throu:,h at least j a 0.94 square inch or larger vent within the following 36 hours, l b. With both PORVs inoperable during heatups ao cooldowns when the RCS temperature
is less than $15 F, be in cold shutdown wid.in the next 36 hours and depressurize and

!. vent the RCS through at least a 0.94 square inch or !arger vent within the following 36 i hours. l c. With one PORV inoperable in Modes 4 or 5,'within one hour ensure the pressurizer

                              - steam space is greater than 53% volume (50.6% or less actual level) and restore the f                                inoperable PORV to' operable within 7 days. If adequate steam space cannot be l                                established within one hour, then restore the inoperable PORV to operable within 24 l                                hours. If the PORV cannot be restored in the required time, depressurize t .8 vent the RCS through at least a 0.94 square inch or larger vent within the next 36 hours. _ _

2-15 Amendment No. 39,4-7,64,446, L l

2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Beactor Cholant System (continued) 2.1.6 Pressurizer and Main Steam Safety Valves (continued)

d. With both PORVs inoperable in Modes 4 or 5 depressurize and vent the RCS through at least a 0.94 square inch or larger vent within the next 36 hours.

(5) Two power-operated relief valves (PORVs) and their associated block valves shall be operable l in Modes 1,2, and 3.

a. With one or both PORV(s) inoperable because of excessive seat leakage, within I hour either restore the PORV(s) to operable status or close the associated block valve (s) with power maintained to the block valve (s); otherwise, be in at least HOT SHUTDOWN within the next 6 hours and in COLD SHUTDOWN within the following 36 hours,

b. With one PORV inoperable due to causes other than excessive seat leakage, within I hour either restore the PORV to operable status or close its associated block valve and remove power from the block valve; restore the PORV to operable status within the following 72 hours or be in HOT SHUTDOWN within the next 6 hours and in COLD SHUTDOWN within the following 36 hours,

c. With both PORVs inocerable due to causes other than excessive seat leakage, within I hour either restore at least one PORV to operable status or close both block valves, remove power from the block valves, and be in HOT SHUTDOWN within the next 6 hours and in COLD SHUTDOWN within the following 36 hours.

d. With one er both block valve (s) inoperable, within I hour restore the block valve (s) to operable status or place the associated PORV(s) in the closed position. Restore at least one block valve to operable status within the next hour if both block valves are inoperable; restore the remaining inoperable block valve to operable within 72 hours.

Otherwise, be in at least HOT SHUTDOWN within the next 6 hours and in COLD SHUTDOWN within the followig 36 hours. Basu The highest reactor coolant system pressure reached in any of the accidents analyzed resulted from a complete loss of turbine generator load without simultaneous reactor trip while operating at 1500 MWt.m This pressure was less than the 2750 ps'.a sofety limit and the ASME Section III upset pressure limit of 10% greater than the design pressure.m The reactor is assumed to trip on a "High Pressurizer Pressure" trip signal. The power-operated relief valves (PORVs) operate to relieve RCS pressure below the setting of the l pressurizer code safety valves. These relief valves have remotely operated block valves to provide a positive shutoff capability should a relief valve become inoperable. The electrical power for both the relief valves and the block valves is capable of being supplied from an emergency power source to ensure the ability to seal this possible RCS leakage path. 2-15a Amendment No. 54,446,

a . 2.0' IN11 TING CON 91TIONS FOR OPERATION 2.1 Reactor Coolant System (continued) 2.1.6 Pressurizer and Main Stearn_Sifety Valves (continued) Action statements (5)b and c. include the removal of power from a closed block valve to preclude any inadvertent opening of the block valve at a time the PORV may not be closed due to maintenance. However, the applicability requirements of the LCO to operate with the block valve (s) closed with power maintained to the block valvc(s) are only intended to permit operation of the plant for a limited period of time not to exceed the nex* refueling shutdown (Mode 5), so that maintenance can be performed on the PORV(s) to eliminate the seat leakage condition. To determine tl.e maximum steam Dow, the only other pressure relieving system assumed operational is the main steam safety valves. Conservative values for all systems parameters, delay times and core moderator coefficients are assumed. Overpressure protection is provided to portions of the reactor coolant system which are at the highest pressure considering pump head, How pressure drops and elevation heads, if no residual heat were removed by any of the means available, the amount or steam which could be generated at safety valve lift pressure would be less than half of the capacity of one safety valve. This specification, therefore, provides adequate defense against overpressurization when the reactor is suberitical. Performance of certain calibration and maintenance procedures on safety valves requires removal from the pressurizer. Should a safety valve be removed, either operability of the other safety valve or maintenance of at least one nozzle open to atmospl ere will assure that suf6cient relief capacity is available. Use of plastic or other similar material to prevent the entry of foreign material into the open nozzle will not be construed to violate the "open to atmosphere" provision, since the presence of this material would not significantly restrict the discharge of reactor coolant. The total relief capacity of the ten main steam safety valves is 6.54 x 106 lb/hr. If, following testing, the as found setpoints are outside +/-l% of nominal nameplate values, the valves are set to within the

              +/-l % tolerance. The main steam safety valves were analyzed for a tual loss of main feedwater flow while operating at 1500 MWP to ensure that the peak secondary passure was less than 1100 psia, the ASME Section ill upset pressure limit of 10% greater than the design pressure. At the power of 1500 MWt, suf6cient relief valve capacity is available to prevent overpressurization of the steam system on loss-of-load conditions.

The power-operated relief valve low setpoint will be adjusted to provide sufficient margin, when used in onjur.ction with TecFalcal Specification Sections 2.1.1 and 2.3, to prevent the design basis pressure transients from causing an overpressurizatiori incident. Limitation of this requirement to scheduled cooldown ensures that, should emergency conditions dictate rapid cooldown of the reactor coolant system, inoperability of the low temperature overpressure protection system would not prove to be an inhibiting factor. The effective full flow area of an open PORV is 0.94 in 2, j Removal of the reactor vessel head provides sufficient expansion volume to limit any of the design basis pressure transients. Thus, no additional relief capacity is required. References (1) Article 9 of the 1968 ASME Boiler and Pressure Vessel Code, Section III (2) USAR, Section 14.9 (3) USAR Section 14.10 (4) USAR, Sections 4.3.4,4.3.9.5 2-16 Amendment No. -39,4h54

2.0 ' LIN11 TING CO"DITIONS FOR OPERATION 2.3 Emergency Core Cooling System (Continued) i l (3) Protection Against Low Temperature Overoressurization The following limiting conditions shall be applied during scheduled heatups and cooldowns. Disabling of the HPSI pumps need not be required if the RCS is vented through at least a 0.94 square inch or larger vent. Whenever the reactor coolant system cold leg temperature is below 385'F, at least one (1) HPSI l pump shall be disabled. Whenever the reactor coolant system cold leg temperature is below 320'F, at least two (2) HPSI l pumps shall be disabled. Whenever the reactor coolant system cold leg temperature is below 270'F, all three (3) HPSI pumps shall be disabled. In the event that no charging pnps are operable when the reactor coolant system cold leg temperature is below 270*F, a single HPSI pump may be made operable and utilized for boric acid injection to the core, with Dow rate restricted to no greater than 120 gpm. 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 approach to criticality is substantislly equal to that during power operation and therefore all engineered safety features and auxiliary cooling systems are required to be fully operable. During low power physics tests at low temperatures, there is a negligible amount of stored energy in the reactor coolant; therefore, an accident comparable in severity to the design basis accident is not possible and the engineered safeguards systems are not required. The SIRW tank contains a minimum of 283,000 gallons of usable water containing a boron concentration of at least the refueling boron concentration. This is sufficient boron concentration to provide a shutdown margin of 5%, including allowances for uncertainties, with all control rods withdrawn and a new core at a temperature of 60aF.m The limits for the safety injection tank pressure and volume assure the required amount of water injection during an accident and are based on values used for the accident analyses. The minimum 116.2 inch level corresponds to a volume of 825 ft' and the maximum 128.1 inch level corresponds to a volume of 895.5 ft'. Prior to the time the reactor is brought critical, the valving of the safety injection system must be checked for correct alignment and appropriate valves locked. Since the system is used for shutdown cooling, the valving will be changed and must be properly aligned prior 'a start-up of the reactor. 2-22 Amendment No. 1-1,47,39,43,47,64,74, 77,400,103,133, 141

U.S. Nuclear Regulatory Commission LIC-93-0069 ATTACHMENT B i l i I I l t

l . The No Significant Hazards considerations are discussed for each of the proposed

Technical Specification changes in Attachments B.1 through B.3.: The Technical Specification changes to pages 2-15, 2-15a, 2-16 and 3-16a were made to incorporate the requirements for Generic Letter 90-06, Operating Modes 4 and 5 were added to specify requirements in addition to the normal heatup and cooldown operations. The applicability of the PORV operating requirements for the Modes 4 and 5 was also added. One PORV is allowed to be; inoperable for up to 7 days if the reactor coolant system is not water solid, i This allows for the expansion of the coolant during a heatup event. One PORV may only be inoperable for 24 hours if the reactor cooTant system is in a water solid condition. The time periods meet the intent of the GI 94 resolution contained in the GL 90-06. A 72 hour time period was specified to reach cold shutdown and complete depressurization and venting if both PORVs are inoperable. The time period for depressurization and venting is longer than that contained in the Since Fort generic letter due to the safe shutdown mode for Fort Calhoun.

4 Calhoun was desig'ned as a hot shutdown plant, it requires a longer time period , to reach a cold, depressurized condition without compromising plant or personnel safety. The definition of " venting" was also added to the basis to indicate an area greater than 0.94 in.8, which is equivalent to the cross sectional area of a PORV. The additions are consistent with the intent of GL 90-06, Attachment B-1. The action statements in Technical Specification 2.1.6(5)a. through d. were modified or added to ensure that the operability requirements of Generic Letter 90-06 were incorporated. The LC0 statement was clarified by replacing "all" with "both." The requirement to maintain power to closed block valve (s) was included inoperable, and the because removal of power would render theapply. block valve Power(s)is maintained to the requirements of action statement c. would 4 block valve (s) so that it is operable and may be subsequently opened to allow the PORY to be used to control reactor coolant system pressure. Closure of P e s block a PORV valve that (ha)s excessive seat leakage.The establishes integritythe reactor of the coolant reactor pressure boundar coolant pressure boundary takes priority over the caphbility of the PORV to mitigate an overpressure event. Action statements b. and c. include the removal of power from a closed block valve as additional assurance to preclude any inadvertent opening of the block i l valve at a time in which the PORV may not be closed due t'o maintenance to restore it to operable status. Action statement d. has been mcdified to established remedial measures that are consistent with the function of the block valves. The primary function is the capability to close the block valve to isolate a stuck-open PORV. Therefore, if the block valve (s) cannot be restored to operable status within one hour, the remedial action is to place the PORV in manual control to preclude its automatic opening for an overpressure event and to avoid the potential for a stuck-open PORV at a time that the block valve is inoperable. The time allowed to restore the block valve (s) to operable status is based upon the remedial action time limits for inoperable PORVs from action statement b. and c. since the PORVs are not capable of mitigating an overpressure event when placed in manual control. These actions are also consistent with the use of the PORVs to control reactor coolant system pressure if the block valves are inoperable at a time when they have been closed to isolate PORVs with excessive seat leakage. 3

1 k . 1 The operating mode for meeting LCO comitments (HOT STANDBY) in the section < 2.1.6 action statements a to d., described above, and the time to achieve cold shutdown conditions (24 hours) were changed to be consistent with the c .. ort Calhoun Technical Specifications for safe shutdown of the unit. -The safo j shutdown design for Fort Calhoun is HOT SHUTOOWN. Technical Specification :

2.0.1 allows 36 hours to achieve cold shutdown from a hot shutdown condition, i The recommendations in GL 90-06 were reviewed and the LC0 action statements
were modified to meet the intent of the GL, yet remain consistent with the other Fort Calhoun Technical Specification action statements. This renains consistent with the design and operating license requirements for Fort Calhoun.

4 When the block valve is inoperable, placing the PORV in manual control is

sufficient to preclude the potential for having a stuck-open PORV that could i not be isolated because of an inoperable block valve.

? Surveillance requirement 22 in Table 3-3 of the Technical Specifications is

proposed for modification to allow an exception for testing the block valves

! when they are closed for isolation of an inoperable PORV. If the block valve

is closed to isolate a PORV with excessive seat leakage, the operability of '

. the block valve is of importance, because opening of the block valve is necessary to permit the PORV to be used for manual control of reactor i pressure. If the block valve is closed to isolate an otherwise inoperable PORV, the maximum allowable outage time is 72 hours, which is well within the ! allowable limits (25 percent) to extend the block valve surveillance interval ! (92 days). Furthermore, these test requirements would be completed by the i reopening of a recently closed block valve u)on restoration of the PORV to an } operable status. The position of the PORV ) lock salves will be verified on a daily basis in response to the requirements of GL 90-06.

The times to complete the actions statements have been conservatively reduced l to ensure prompt compliance with the recuirements for safe operation. The No l Significant Hazards considerations are ciscussed for each of the proposed Technical Specification changes in Attachment B.4.

A number of items were not incorporated into the Technical Specifications as requested by Generic Letter 90-06 and are discussed below. The testing i requirement for the PORVs is covered in the OPPD inservice inspection testing (IST) program, which is approved by the NRC. Thus, a change in the testing of the PORVs would require NRC approval before the valves could be dropped from the IST program. The verification of the PORV block valve position is completed on the control room log (Form FC-75) once a day which is more frequent than the 72 hour requirement Specification.4.4.9.3.c contained in the Generic Letter. The reporting requirements in Specification 3.4.9.3.e are redundant to reporting recuirements of 10 CFR 50.73; thus, the reporting

f. requirement will not be acded to the Technical Specifications since the requirement to produce an account of the event-is redundant to the Code of Federal Regulations.

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