ML20101B530
| ML20101B530 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 06/01/1992 |
| From: | OMAHA PUBLIC POWER DISTRICT |
| To: | |
| Shared Package | |
| ML20101B505 | List: |
| References | |
| NUDOCS 9206040137 | |
| Download: ML20101B530 (54) | |
Text
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- .0
- .1 L Pli?'G C3DfiiMS Po oPEDAT10N
;eacter Coolant Svitem (Continueo)
- .l.1 Opencie Co-conent s (Continueo)
(c) For the curposes of items (a) ano (b) above, the containment i spray pumps can be considered as available shutoown cooling ; pumps only if both of the following conoitions are met: ! is (i) Rtactor Coolant System temperatureYless than 120*F. j t I (ii) The Reactor Coolant System is vented with a vent area 1 equal to or greater than 4 hat-of-the-pressueuen
.manway,. WT if, l l
Exceetions All decay heat removal loops may be made inoperable for up to 8 hours provided (1) no operations are permitted that would cause dilution of the reactor coolant system boron concentration. (2) no refueling ooerations are taking place, and (3) all containment penetrations providing direct access from the centainment atmosphere to the outside atmosphere are closed within 4 hours. (5) At least one reactor coolant pump or one low pressure safety injection pump in the shutdown cooling mode shall be in operation wnenever 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 coolant temperature of 300'F.procram specified in Section 3.17 prior to axceeding I (7) Maximum 2125 psia. reactor coolant system hydrostatic test pressure shall be are allowed. A maximum of 10 cycles of 3125 psia hydrostatic tests (8) Reactor coolant system leak and hydrostatic test shall be conducted within the limitations of Figures 2-1A and 2-18. (9) Maximum psia. secondary hydrostatic test pressure shall not exceed 1250 10 cycles are permitted.A minimum measured temperature of 73*F is required. Only (10) Maximum exceed 1000 steam psia.generator steam side leak test pressure shall not required. A minimum measured temperature of 73*F is . TP no rmcTor evomnT Pun;PS Gre ope @mG W)e Tc is (ll) W non-operating reactor coolant pump shall not be started unlessklaeae at least one of the following cond:tions is met:
)
2-2a Amendment No. 39.36,55,7J ,13. 96-
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20 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued)
, 2.1.1 Operable Components (Continued) 53'/a (a) A pressurizer steam space of-60% by volume or gre~ater exists, or (b) The steam generator tecondary side temperature is less than +0 Q above that of the reactor coolant system cold leg.T so*F (12) Reactor Coolant System Pressure Isolation Valves (a) The intogrity of all pressure isolation valves listed in Tabic 2-9 shall be demonstrated, except as specified in (b). Valve leakage shall not axceed 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 nonfunc-tional valve are in and remain in the mode corresponding to the isolated condition.* 4 (c) If Specifications (a) and (b) aoove cannot be me t, an orderly shutdown shall be initiated and the reactor shall I be in the cold shutdown condition within 24 hours. Basis The plant is designed to operate with both reactor coolant loops and associated reactor coolant pumps in operation and maintain DNBR above 1.18 during all normal operations and anticipated transients. l In the het shutdcwn mode, a single reactor coolant loop provides sufficient heat removal capability for :emoving 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 require 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 boron stratification. _ ' Manual valves shall be locked in the closed position; metor operated valves _ j shall be placed in the closed position and power supplied deenergized. Icend.nen: No.- ff, 0/#f U20/U, 70, 2-2b N/ , 92
2.0 tIPf71NG CONDIT10NS FOR OPfotTION 2.1 oeactor Coolant System IContinued) 2.1.1 Q erable Com onents (Cortinued) l 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 f ailure 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 core.
The restrictions on availability of the containment spray pumps for shutdown cooling service enst e that the S!/CS pumps' suct1m header piping is not subjected to an unanalyzed conaition in this mode. Analysi M ,.eevece schen s has determined that the minimum required RCS vent area is 47 The-p re s su r4 ze Hnanway4 s-s pec+f4 ed-as-t he-e k : mum vent-ar-ea-to- aWw vant4apthrouch the 1im444ng-cross-sect 4enabarea e of-the-pressur4,terquch %a, cf % pr sweger uru 34n Ws y- e*rW m{ Le met removy nne e w en W am g emTee w 4 hn,. 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 baron concentration which could result in a reactivity insertion. Sufficient mixing of the reactor cool nt is assured if one low pressure safety injection pump or one reactor coolant is assured if one low pressure safety injection pump or one reactor coolant pump is in operation. The low pressure safet;' 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 ter.d to have a boren concentration higher taan 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 concen' ration in the pr urizer and the reactor coolant system during the addition of boren.9{I \ 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 tnan the design temperature of the shutdown cooling system to assure a redundant. heat removal system for the reactnr. L eecT 1- > The design cyclic transients for the reactor system are given in USAR Section 4.2.2. In addition, the steam genera *. ors 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 tne actual measured limit. The measured i
temperature limit will'be 73*F based upon use- of an instrument with a maximum inaccuracy of : 2*F i.nd an additional l'F safety margin. 2 ?c Amendment No. 56, ,4/EI/p/dit, 71, t3G
T ,y z_ :r } .) The LTOP enable temperature has been established at T, = 385'F. The pressure transient analyses f demonstratef that a single i ORV is capable of mitigating overpressure events. Additional uncertainties t have been applied to the Pressure-Temperature (P-T) limits to account for the case where a PORV is not available (T,> 385"F3 which is the reason for the apparent discontinuity in the P-T Figures. F i
l 7 .
* .y tIMITING CONDITIONS rnR OPERATION
.the EU'il
- ff ector Coolant System (Continued) ~
'!, Deerable r emrenents (Continued)
SV/, Formation of a 404 stcam space ensures that the resulting pr essure increase would not result in any overpressurization should-0 reactor -
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coolant pump be started when the steam generator secondary side temperature is, so greater than that of the RCS cold leg. 7he -TwoSP'W> rwpo rm.nent y - '.g
.pw YOr t$e Ea's7 f ,wrogt.nole1esluf[$r"g$m,7 1n Eh' d M Ire # ste spaceofsaistu r e cu r cro b.if p us"r limitation of tee 7t~eam generator secondary side /RCS cold leg o.T that a single low setpoint PORV would prevent an ove to-69'f ensures rpressurization due to actuation of-t-reactor coolant pump.A Lac F L -+ h e {'icct The exception to Specification 2.1.l(4) requiring al l containment penetrations providing direct access from the contai nment to the outside atmosphere be closed within 4 hours requires that the equipment hatch be closed and held in place by a minimum of four bolts.
;5 ggg3 @<g M oeferences (1) USAR Section 4.3.7
- Tu re airement /5 ,
gny apphenble -fs +!* O* '? '*, a ,vo uTc c etelant fo'Y ' I #" o r- tvere puanfa n re of e'*T, *')
- 1 1
2 2d Amendment No f 6, .4/jtJ/pfdtt, /J . H6-l I _ - _ _ _ - . - - - - - - - - - - ~ - > - - - - -
2.0 Lili! TING CONDITIONS FOR OPERATION i 2.1 Reactor Coolant system (Continued) 2.1.2 Heatvo and Cooldown Rate golicability Applies to the temperature change rates and pressure of the reactor coolant system. Objective To specify limiting conditions of the reettor coolant system heatup and cooldown rates. Sneci fi ca tio n i The reactor coolant pressure shall be limited during pl' int operation in accordance with Figure 2-1 A and 2-10 and as follows : l j
~Jnse rt TieMm The-heetup-eet habl.-no t-ex c eed4000f-4 n9 nv-o ne--ho ur-pedsd . I (1) 1 (2) Allowable combinations of pressure and temperature (Tc) for a specific cooldown rate shall be below and to the right of the ,
applicable limit lines as shown on Figures f-1A-eM 2-1B. I 1 (3) The heatup rate of the pressurizer shall not exceed If.,00F in I any one hour period. (4) The cooldown rate of the pressurizer shall not exceed 2000F in any one hour period. (5) When any of the above limits are exceeded, the following cor. rective actions shall be taken: (a) Irlaediately initiate action to res tore the temperature i or pressure to within the limit. l (b) Perfom an analysis to determine the effects of the out of limit condition on the fracture toughness properties of the reactor coolant system. (c) Determine tha t the reactor coolant s)/s tem remains ac te?t-able for continued operation or be in cold shutdown with-in 36 hours. , (6) Before the radiation exposure oD the reactor vessel exceeds the exposure for which they apply, Figures 2-1 A and 2-13 shall be } l updated in accordance with the following criteria and procedures: l Anenement No. ??, f* 2-3
- 2. 0 LIMll]NG CONDITIONS FOR OPERATION I 2.1 Reactor Coolant System (Continued) 2.1.2 Heatup and Cooldown Rate (Continued)
(a) The curve in Figure 2-3 shall be used to predict the increase in transition temperature based on integrated fast neutron flux. If measurements on the irradiation specimens indicate a deviation from this curve, a new curve shall be cor.tructed. (b) The limit line on the figures shall be updated for a new integrated power period as follows: the total integrated reactor thermal power from startup to the end of the new period shall be converted to an equivalent integrated fast neutron exposure (Ei:1 Mev). For-this-plantr-based-upon_ surveitaance-mater 4al+-tests r-weld-chemicaMempos4t4on daterend-the-ef4ect-of-a-reduced-vessel fluence _r. ate provided-t>y-core-load-des 94 n64eginning-with-fue4-Cyc48 8,
- t he-p red i c ted--su r4 ace-f4 uenc e-e t-t he-4*4 (4 al-eeec tee-ve sse l 4#e44Line-we44 -ma te r4al--fo r-40-yea ns-e t-1500-HW t-and-e n-804-load 44ctor- 4 s 2. 55x1019 n/c,,2 The-f4ux-reduet4en-applied to-the-face nce - c a l c u l a tions-wa s4a sed-on-Cyc4 e-44--a verage a nd-Cyca e-8-a v era ge-a z4 mu tham 4u x-di s t ribu t4en-ple te generated-used-001-4r3. The predicted transition temperature shift to the end of the new period shall then be obtained from Figure 2-3.
g-r9 I (c) The limit lines in Figures 2-1A and 2-1B s hall be moved parallel to the temperature axis (horizont al) in the direction of increasing temperature a dist ance equivalent to the transition temperature shift durin{ the periot since the curves were last constructed. The boHup-temperature limit line shall remain at 82 F as it is set by the NOTT of the reactor vessel flange and not subject to fast neutron flux. The lowest service temperature shall remain at 182 F because components related to this temperature are also not subject to fast neutron flux. (d) The Technical Specification 2.3(3) shall be revised each time the curves of Figures 2-1A and 2-1B are revised. Basis All components in the reactor coolant system are designed to withstand theeffectsofcyclic({oadsduetoreactorcoolantsystemtemperature and pressure changes. These cyclic loads are introduced by normal unit load transients, reactor trips and startup and shutdown operation.
--During unit startup and shutdown,- the rates of tumperature and pressure changes are limited. The des'ign number of cycles for heatup and cool-down is based upon a-rate of 100 e 44-any-one-hour-per4od and 6 cyclic operation. b @oggge hecckp/coelckwn rcCtes
)
2-4 Amendment No. 22,47,64,74,77,200 cH4
2.0 LlHITING CONDITIONS FOR OPERATION
) 2.1 Reactor (colant System (Continued) 2.1.2 FeatupandCoc1downKate(Continueo) 1500 MWt and 80% load factor. The predicteo shift at this location at the 1/4t depth from the inner surft.ce is 332'F, including margin, and was calculated using the shif t prediction equ6 tion of the-proposed Regulatory Guide 1.99 Revision 2. The actual shift in T will be re-established
. periodically during the plant operatien by tesbg of reactor vessel material samples which are irradiated cumul6tively by securing them near the inside wall of the reactor vessel as describec in Section 4.5.3 ano Figure 4.5-1 of the USAR. To compensate for any increase in the T caused by irradiation, liraits on the pressure temperature relationNIp are periodically changed to stay within the stress limits during heatup and cooldown. Analysg)ofthesecondremovedirradiatedreactorvessel fluente- tyrve111ance specimen , combined with welo chemical composition data and reduced} core loading designs initiated in Cycle _8,_j_nd4cated that 20*0 the fluence at the eno-o 14WUfective FulTPower fears (MY) at j,yogo!? 1500 ~This kWt result will be ~s in+rB)xM-{9 n/cm8 vessel. a total shif t of the RTontheinsidesuifaceo@fthereactorgg,I; of 2 F, incluainc margin, for the area of greatest sensitivity (web, metal) at the 1/4t-location as determined from Figure 2-3.-- Operation thrcugh fuel- Cycle-B-/9 will result in less than -14e EFPY. 6anda sh 4 t of M "F. # 1he.
20* C 3/% leceWn. The limit lines in Figures 2-1A and 2-1B are based on the following: A. Heatup and Cooldown Curves - From Section III of the ASME Code, I Appendix G-2215. ' K;p = 2 Ky*tlT KIR = Allowance stress intensity factor at temperature related to RTNOT (ASME 111 Figure G-2110.1). X;g = Stress intensity factor for membrane stress (pressure). The 2 represents a safety factor of 2 on pressure. KIT = Stress intensity factor radial thermal gradient. The above equation is applied to the reactor vessel beltline. Fo r-p l a n t-he e tu p- t he-t hema l-s tre s s-i s-op pos i te-in449n-f rom-t he
'Inse.c't Treni2- geressure-stressand-considerat4cn higher pressurer-For-heetup-it isntthere+cre a heatg.,rgig_wogld_alhw_for conservative-to
, -con sd ev r--a r.-4sotte rmal-hea tup-en-Mp =-Or for plant cooldown thermal and ssute stress are additive.
2-6 Amendment No. 22.#7,H,7#,77, 199,224.121-
t RCS PRESS-TEMP LIMITS HEATUP 14EFPY / . 3 REAt10a not tn111tAL d500 ut 3200 m s ma m a press n 1A1 /
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) / RC 1R.El TDP lDEG f) Ic FORTCALHOUN FI URE TECHNICAL SPECIFICATIONS eeeeement no. rs,77,wn. )
- 2-tA x .
FORT CALIIOUN STATION UNIT 1 P/r LIMITS,20 EFPY llEATUP & CORE CRITICAL 2500 2500
- 00*F/IIR 75'F/11R 2000 ISOTimRFf A1^ 2000
/ 1 lC rd1500 .
1500
$ LOWEST CORE CRITICAL y SERVICIl a: TEMPERATURE l 182'F w ALLOWABLE liEKFUP RATES 1000 TEh1P. LIMIT 'F RATE 'F/HR
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$ I s335 75 g gOTl iERMALl > 335 100
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75"F/IIR 500 -
'd. - - . 500 10C "F/IIR N
MilN. BOLTUP TEMP. 82* F i 0 -0 0 100 200 300 400 500 600 Tc INDICATED REACTOR COOLANT SYSTEM TEMPERATURE, F l i RCS Prrssure-Temperature Omaha Public Power District Limits for 11eatup Fort Calhoun Station- Unit No.1 _$(*
\ RCS PRESS-TEMP LIMITS C00LDOWN 14 EFPY REACTORNOTcA111 cat 1500 Wt PRESSURHER PRESS (PS]Al 110 1000
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FORTCALHOUN FIGURE TECHNICAL SPECIFICATIONS Ame' dment 'io " 77 no 12 4 2-1B
FORT CAlllOUN STATION UNIT 1 P/r LIMITS 20 EFPY COOLDOWN AND INSERVICE TEST 2500 , 2500 INSERVICE (1YDROSTATIC TESTw N 2000 2000 f1500 M LOWEST SERVICE ! 1500 100'F/liR Tg ISOTHER stAL h c-TEMPERATURES ALLOWADLE COOLDOWN RATE 182*F N e d l TEMP. LIMIT 'F RATE 'F/HR E 1000 < < 135 10 ISOTilEli MAL y 100 a. 500 500 l 10*F/HR-:r?[ / 30'F/HR fMIN. BOLTUP TEMP,82 F 0 0 100 200 300 400 500 600 Tc INDICATED REACTOR COOLANT SYSTEM TEMPERATURE, 'F , RCS Pressure-Temperature Omaha Public Power District i Limits for Cooldown Fort Calhoun Station- Unit No. I _
PREDICTED RADIATION INDUCED NDTT SH1'T
\ FORT CALHOUN REACTOR VESSEL BELTLINE 500 400 N /
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h0RT CALHOUN TECHNICAL
\ FIGURE SPECIFICATIONS 2-3 Amendment No. /,4,77,200.JI,4,121
1 Predicted Radiation induced NDTT Shift Fort Calhoun Reactor st essel Beltline RTndt 500 , , 450 i .
! ! I 400 -
1 1.D. Sill including Margin I 350 l l I , I 1/4t SHIFT 300 "# " "E "'E " j - 250 [ i 3/4t SHIFT I Including Margin 200 - -
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l I too l 0.0 0.5 1.0 1.5 2.0 2.5 3.0 3.5 4.0 4,5 5.0 2 Neutron Fluence, IE19 n/cm Predicted Radiation Induced Omaha Public Power District Figure NDTT Lhltt Fort Calhoun Station-Unit No.1 2-3
- 2. 0 LIMIT 114G C0f1DIT10NS FOR OPERATION 2.1 Reactor Coolant System (Continued) i 2.1.2 Heatup and Cooldown Curves (Continued)
Kjg = Mgf MM = ASME III, Figure G-2214-1 P = Pressure, psia R = Vessel Radius - in, t = Vessel Wall Thickness - in. KIT " MTa Ty MT = ASME III Figure G-2214-2 LTy = Highest Radial Temperature Gradient Thrnur.,h Wall at End of Cooldown K;7 is therefore calculated at a maximum gradient and is considered a constant = A for cooldown and-zero for heatup. MM R is also a constant = B. T~ Therefore: KIR = AP + B I P=KIR-B A Lsca.T -Kg-is-t he n--va r4 ed-a s-+-f u nction-of t mmerature from 549ure G-211M-of. Iqem 3->ASHE-l+1-end-the-eHowaMe--pressum--cekulatedr-Hydr +st-at4o-head-448--pi-)
-and-instrumentation-errors (12 F and-32-ps44-are-cons 4 der +d-wre-plotting.
-t he-<u r+es .
Inser v'tce inse -hcc B. 4ystem-Hydrostatic Test - The 4ystera.hyirostatic test curve is developed in the same manner as in A above with the exception that a safety factor of 1.5 is allowed by ASME III in lieu of 2. C. Lowest Service Temperature = 50*F + 120'F + 12*F = 182'F. As indicated previously, an RTNDT for all material with the exception of the reactor vessel beltline was established at 50*F. 10 CFR Part 50, Appendix G, 1 IV a.2. requires a lowest service temperature of RTNDT + 120 F for i piping, pumps and valves. Below-this-temperaturc a pressure-of-20-per-InMRT <ent-o f-the-sys tem-hyd ros t a t4 c-te s t-pressu re-{ rE@342M 48 3Fpsv = Ttryn @AS-ps4e-sannot-be-eneeded, Sto9 D. Boltup Temperature = 10'F + 60*F + 12*F = 82*F. At pressure below +46-l psia, a minimum vessel temperature must be maintained to comply with j the manufacturer's specifications for tensioning the vessel head. 1 2-7 Amendment No. 27,/7,E/.7!,1 @
2.0 1.1MITING CONDITIONS FOR OPERATION 2.1 Reactor coolant System (Continued)
! 2.1.2 Heatup and Cooldown Rate (Continued) l This temperature is based on previous NDTT methods. This temperature corresponds to the measured 10*F NDTT of the reactor vessel flange, which is not subject to radiation damage, plus 60'F data scatter in HDTT measurements, plus 12 F instrument error.
E. Tie ~1emoearum xt whic h -H,e- hsang a nd ccetdown twTeo clAnge i., F$u ce s 2-I A a ut 2- 16 N 0lecTs -t he poi n-t et tohic h 4he most hm% rvited te r+h epe cT -te .he lin4Mn,g i,3tet ' fempe,gaTweecuyl (yy coelcb etwa
References:
$c. (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 (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 Irradiated Capsule W-265, March 1984
)
2-7a Amendment No. /2,f 7,f>f ,74,100 a?
Text to Be htsetted irl. Technical Specification 2.1.2 Item 1. Allowable combinations of pressure and temperature (Tc) for a specific heatup rate shall be below and to the right of the applicable limit lince :.s shown un Figure 2-1 A. Item 2. The above equation is applied to the reactor vessel beltline. For plant heatup the reference stress intensity is calculated for both the 1/'t and 3/4t locations. Composite curves are then generated for cach heatup rate by combining the most restrictive pressure-temperature limits over the completc temperature interval. Item 3. K ta is then varied as a function of temperature from Figure g-2110-1 of ASME-Ill and the allowable pressure calculated. Pressure correction factors for clevation and flow (-56 psia for Tc < 210 *F and -62 psia for Tc 2. 210 F) and temperature instrumentation uncertainties (+16 'F) are considered
- when plotting the curves. Pressure instrumentation uncertainty is also considered above the LTOP cnacle temperature of 385 *F. Pelow this
, temperature, pressure instrumentation uncertainty is accounted for in the LTOP PORV setpoints.
Item 4. 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, tnis pressure is 569 psia. t
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2.1 3eact:r '::: ant Syste.iCOntinued)
- 2. _. 6 Pretturi er and Eteam Svstem Sefety Valves Act11:1bilit t Ar.alies to the status of the pressurizer and steam system safety valves.
Objective TO specif;r .inimun requirements pertaining to the pressuricer 2.d stea . synen safety valves. Ececi' :niens To provide adequate overpressure protection for the renetor c: clin systen ind stcea system, the folleving safety valve requirerer.:s shall te met: (1) Tne reae: r shall not be .ade criti:al unless the two pre:suri:er saf ny valves are operable with their lift settings zijusted to ensure valve openir.g betveen 2500
;3ia ini 25 5 psia +15.(1) _
(2} Ynenever there is fuel in the reactor, and the reactor g' vessel nezd is installed, a minimum of cne cperable nfn; t1.ve shall be installed on the pressuriter, 2:vever, ran in at lesst the cold shutdevn condition, s a fe .; /s re nc::ler, nay be open to containment atmos-
;ne.e .! urin,; performs.nce of safety valve tests or mainte-nance to s nia ^/ this specificatien.
( :- ) ner.eter the reactor is in pcVer cperaticn, eight of the ten 5.es: safety valves shall be operable vith their lift ' se tings te;veen 1000 psia and 1050 psia with a t91erance cf +1., _ Of the ncminal nameplate set point values \1) g ( 1- ) M.? pressuriter pover-;perated relief valves (PCRV's) nocl. 6 thall be :;erible during scheduled hestup and eccidevn,, AncI In OdC5 N
- n pretent violatien of the pressure-temperstr e limits desi;r.ated by Figures 2-1A and 2-1B. One PC3V r.ay te
, it. :ersble Mr up :o days, previded the renaining ?0PM i is :;ct;tle. If the ateve cend ti:ns Of this paragraph
- ann n be net, Ge-pelmuy' -ayst,en.eue t- be d: presourf-ce 4 m e m ece. be,in CCL.D dHuTDowM w,' tin,) W }e n e x 7 20 4 % d 9'd
% s *C 2,
;ytuTe-}c> . e Pro n1 A ti trU s,Te m clePf'essatthed Ndis ,1n AC$ v enT a(
+kun e; t ..z ;cuer-etern cr-rel l p va t,ves to
. F3. '.m.n(,,(an: a fle** wsYhin their ss--f hef,//ge,nj y 9,.2g-ca:ir. ei t_c:k valres shall be operable in :.!odec 1, 2,
% e . s.
A eni ent :!: . h, W , $1r 2-15 i
2 , ') LD1IT!:in COWIT.I.CNFn .. POR
. - OWRf TI.O!I 2.1
? s int lco gr g yst n (Continued) 2.. 6 Pressuri:er ind Stesa System Zarety Valv_es (Continued)
'%s With cno er more ?CRV(s) incperable, within 1 hour Q N v(+{ ace restore the FORV(s) to operable status or j elos%the associated bicch valve (s); othervise.
.Dthch&O ke in IN er.st HOT STMfDBY vithin the next 12 hours and hyCI.D SHUTDOW11 vithin the fo.U.oving
'A n o ur s .
- . - ...pm
\
- b. 'Jith ent or more blo ck\ t Q ve's) incperable, within I hcur ei;her restore the 'bhek valve (t ) to oper-able atatus or close the 'cloch M ve(s). Other-vise, be J.n at least HOT STM.TSY dth '
12 hourc and in COLD SHUTDC.1 within t(n the nen htNfollowing D nours. N Ess,.is Tc.e hir,hes.t reacter coolant system pressure reached in n.ny of the acci:ents snalysed was 2LSO psia asd resulted from a ecc-plet e icss of turbine generator load viihout sirultaneous re-actor trip vnile operating at 1500 M'4t. (2) The reacter is n.stred to trip on a "High Pressurizer Pressure" trip signal.
'he pover-cperated relief valves (?0PV's) operate to relieve RCS pressure below the setting of the pressurizer code safety valve;. !bese relief velves have remotely operated block ve.1ves i pr: vide a positive shatoff ::apability should a re- j lief va vt beccee incperable. The electrical power fer both 2
the relief vsiv e a.d the block valves is capable of being
<g supplin fr:n in energency pover source to ensure the ability j M1hch6d to sea; thi: pcssid:e acs leskage path.
- D To deter tne the naximu:u steam flow., the only other pressure relievin;; system sosur.ed operational is the steam system safety valves. Conser rative values for all systems para::eters, de-lay times a:rd : re =oderator coefficients are assumed. Over-pressure pr:tection is provided to portions of the reactor cocian , sy.itet. which are at the highest pressure considering pur.p head, flow pr:ssure drops and elevation heads.
If no residual heat vere rer.cVed by any of the means avail-able , the u.ount of steam which could be generated at safety va've lift pressure vould be less than half of the capacity of one safety valve. This specification, therefore , provides adequate defense against overpressurization when the reactor is subcrttical. J i ;.en ent :;o . -% P '.5 a
'\
% -t to be ctdclecl % , . Speei Acattcu 2./4 05)
- 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 valvel,'htherwise, 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 d:te 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, e c. With both PORViinoperable due to causes other than excessive seat leakage, within I hont either restore at least one PORV to operable status or close both block valves, rem.4e power from the block valves, and be in HOT SHUTDOWN within the next 6 hours and its COLD SHUTDOWN within the following 36 hours,
- d. With one or 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 following 36 hours. Te n to be addc<l '/d bqsis ef Spec 4cgr% Q. I, G 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 valve (s) are only intended to permit operation of the plant for a limited period of time not to exceed the next refueling shutdown (Mode 5), te that maintenance can be performed on the PORV(s) to eliminate the seat leakage condition. l
2.0 1* 1 T' :0 00:' TIT!0!.5 TOR OF'O' ATIO!! 2.; Feset r reeltnt Sytten (0:ntinued) 2 1.6 ((ysturiter sna Stele Eyster Safet r Va'ves (Centinued) 1 Pert:r:ince cf certain calitratico and maintenance procedures on safety valves requires ra:cve.1 from the pressuri:er. Should a safety valve be removed, either cperability cf the other safety valve er =aintenance of at least ene no::le open to ainesphere vill assure that sufficient relief capacity is avail-able. Use of plastic cr ether sOnilar caterial to prevent the entry of fereign material into the open no::le vill net te ccnstrued *.o viciate the "cren to atecsphare" previsien, since the presence of this caterial vculd net significantly restrict the discharge of reacter coclant. The total re}ief capacity of the ten steam systas safety valves is 6 5L x 100 It/hr. At the pcVer of 1500 ff4t, sufficient relief valve capacity is available to prevent overpressuri:a-tien of the steam system en less-of-load conditicas. The pover-cperated relief valve lov setpoint vill be adjusted to provide sufficient margin, when used in conjunction with Technical Specification Secticas 2.1.1 and 2 3, to prevent the dent;n basis pressure transients from causing en over-pressurization incident. Limitation of this require =ent to sehenuled coolievn ensures that, should energency conditions , dictate rapid cocidevn of the reactor coolant syste=, inoper-abilit/ of the icv temperature overpressure protection system vould as c not prove to be an inhibiting factor. fif6dtn. Ell fkv m4 ti
?.ecyal rau of A the R Vreacter 4 0. w vessel in'. head provides sufficient expan-sien volume to liti% any of the design basis pressure tran-sients. Thus, no additional relief capacity is required. ,
References (1) Article 9 of the Code 3ecticn 1968 ASME Soiler and Pressure Vessel 111 LA (2) TSAR, Section ik.9 (3) YSA3,Eecticnsk.3.h.h.395 e Amendment J 'o. 3 7 , Sr" 2=16
~
2.0 14MITIMIQb31LILOSS FOR Ol'ERATION 2.3 Emfl.EtnCylMClooline System (Continued) (3) ETAc. tim 2 gainst low Tempenttore OvsmJfinitintha The following limiting conditions shall be applied during scheduled heatups and cooldowns. Disabling of the HPSI pumps need not be required if the reactor ves:,el head, a pressuriter safety valve, or a PORV is removed. 36GY Whenever the reactor coolant system celd leg temperature is below-32&F, at least one (1) HPSI pump shall be disabled. WP Whenever the reactor coolant system cold leg temperature is below 312'F; at least two (2) HPSI pumps shall be disabl d.
.29c'F Whenever the reactor coolant system cold leg temperature is below ?.719, all three (3) HPSI pumps shall be disabled, gm -Hw ma Ge cedo4 sy p m co ighpenda is In the event that no charging pumps are operable, a sirillNP pump may be made operable and utilized for bode acid injection to the core, wrA #lero rJe resccWred % n e y rwzer -f % n 120 gpm.
The normal procedure for staning 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 substantially equal to that during power operation and therefore all engineered safety features and auxiilary 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 enginected safegurids systems are not required.
The SIRW tank contains a minimum of 283,000 gallons of usable water containing a boron ecncentration of at least the refueling baron 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 60 F.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 ff and the maximum 128.1 inch level corresponds to a volume of 8E5 ff. Prior to the time the reactor is brought critical, the valving of the safety injection sj: tera must be checked for correct alignment and appropriate valves bcked. Since the system 6 used for shutdown cooling, the valving will be changed and must be properly aligned priar to stan up of the teat.or. 2-22 Amendment No. I1,M,39,43,47,64, t 74,77,400,443r&3,141- j i l
9O 2.3 peroency LIMITING Core, Cool CONDITIONS-n {,F_0_R ng_37 OPERATION stem (Continued)
) be available for emergency core cooling, but the contents of one of the tanks is assumed to De lost through the reactor coclant system. In addition, of the three high-pressure safety injection pumps and the two low-pressure safety injection pumps, for large break analysis it is assumed that two high pressure and one low pressure operate while 9 ply y one of each type is assumed to operate in the small break analysis W1; and also that 25% of their combined discharge rate is lost from the reactor coolant system out of the break. The transient hot spot fuel clad temperatures for the break sizes considered are shown on4SARrAppendix4 Tobles+19-( Amendment 4 tor 34)v id6 % S @ cn M.
ina dv e rt ent1ctuati rm-of-th ree @)diPSI-pumps -and-t bre e -(3 )-c ha rg4 ng-pumpf, coincident with the opening of one of the two PORV's, would result in V peak kir.ary system pressure of 1190 psia. 1190 psia corresponds wit 6 a qEP g"C C minimum pennissible temperature of 320'F on Figure 2-1B. Thus , at'least wMh one HpSI pumpsis disabled at 320'F. NC Inadvertent actua't' ion of two (2) HPSI pumps and threje- charging pumps, hxT coincident with the 6pening of one of the two PORV'A , would result in a peak primary system pressgre of 1040 psia. 10f4 psia corresponds with a e of 312'F on 49ure 2-1B. Thus, at least l two HPSI pumps will be disabl minimum permissible ttemperatu(eda 312'F
/
\
Inadvertent actuation of one (1) HJST9nd three (3) charging pumps, coincident with opening of orej f the tho PORV's, would result in a peak primary system pressure of 685 psia. 685s psja corresponds with a minin,am allowable temperature of)TI'F on Figure 2-1Bd Thus, all three HPSI - pumps will be disable 't 271
- F . I Inadvertent actuation of three (3) charging pumps, co1 ident with the opening of o g of the two PORV's, would result in a peak rimary system pressure ,or 160 psia. 160 psia would correspond with a mini allowable tempeyaturethatislessthanthe82*Fboltuptemperatureligito m Fi3ufe 2-18. Therefore, operation of the charging pumps need not restricted.-
Removal of the reactor vessel head, one pressurizer safety valve, or one p0RV provides sufficient expansion volume to limit any of the design basis pressure transients. Thus, no additional relief capacity is required. Technical Specification 2.2(1) specifies that, when fuel is in the reactor, at least one flow path snall be provided for boric acid injection to the core. Should boric acid injection become necessary, cnd no charging pumps are operable, operation of a single HfSI pump would provide the required flow path. The H PST pump +/cw 1%Te muST be- restricted f %qt of '4hree cha p fw 7S in order % minQe yhe emepwace s of A was n ddctice hsie.d whue ! a w s emes.e 2-23a Amendment No. 39, A7,6A ,7 A ,77, log. e
)
Tag _to Be inserted in Technical Specification 2.3 Basis The restriction on HPSI pump operability at low temperatures, in combination with the PORV setpoints ensure that the reactor vessel pressure-temperature limits would not be exceeded in the case >f an inadvertent actuation of the operable HPSI and charging pumps.
, _..______________-.m
~ ' - _ _ - = .
o J [I TABLE 3-3 (Continued)
- s
[I HINULH FEEQUENCIES FOR CHECKS. CALIBRATIONS AND TESTING OF HISCELLANEOUS INSTRUMENTATION AMD CONTHOUi Surveillance f Function Frequency Surveillance Method Channel Description m
~
Auxiliary Feedvater Flow Check M Channel cP ck. 19 E.
$ Calibrate R Known pressure inputs.
o Subcooled Margin Monitor Check M Channel eneck. l 20. Calibrate R Kncvn pressure inputs and known resistance substituted for RTD inputs. Y H -Channel check. g 21. FORV Operation and Acoustic Check
- Position Indication .
Calibrate R Apply acoustic input. Verify R- Operation on emergency power supply. Check Q Cycle valve. O !Ue IL 8AS "P t
- 22. PORV Block Valve Creraticn dwt '7%t?ng when ir% en cicted 46 an! Positicn Irdication Calibrate B Ccwply w M ACC ddf'4[fM7eN (f-hs bit ssitch position.
Verify R Operability on emer-ancy power j supply. Check M Circuit check. 23 Sarety Valve Acoustic Position Indication Apply acoustic input. Calibrate R Check M Circuit check. 2h. POHV/ Safety Valve Tall Pipe Temperature Calibrate R Apply known input. i ! -~ t__ _m
~ . .,
2.0 1.lMITING. CONI)1IlONSJOR OFER ATION
- 2. l' Reactor Cooktidiy11cm (Continued) 2.1.1 Doerable Comtonents (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) Reactor Coolant System temperature is less than 120'F. l (ii) The Reactor Coolant System is vented with a vent area equal to or greater than 47 in 2. IhCntlions All decay heat removal loops may be made inoperable for up to 8 hours 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 r ' providing direct access from the containment atmosphere to the outside atmosphere are closed within 4 hours. (5) At least one reactor coolant pump or one low pressure safety injection pump in the shutdown cooling mode shall be in operation whenever a change is being made in the boron concentration of the reactor cooiant when fuel b in the reactor. (6) Both steam generators shall be filled above tiac hw 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 coolant system leak and hydrostatic test shall be conducted within the limitations of Figures 2-1 A and 2-1B. (9) Maximum secondary hydrostatie test pressure shall not exceed 1250 psia. A minimum measured temperature of 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 73"F is required. (11) If no reactor coolant pumps are operating, a non-operating reactor coolant pump shall not be started while T is below 385 F unless at least one of the following conditions is met: 2-2a Amendment No. 39,%,66,H,+19rl%
2.0 LlhilTING CONIELQNSlOILQ1'EMllON
- 2. l' ikactor Coolant EyW.m (Continued) 2.1.1 Operable Compognts (Continued)
(a) A pressurizer steam space c. 03% by volume or greater exists, or l (b) The steam generator secondary side temperature is less ti..m 3lrF above 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 specined in (b). Valve leakage shall not exceed the amounts indicated. l l (b) In the event that the integrity of any pressure isolation valve specified in Table 2 9 cannot be demonstrated, reactor operation may contir e, provided that at least two valves in ea4 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
- p. 3ition: motor operated valves shall be placed in the closed position and the power supply deenergized.
(c) If Sped 6 cations (a) and (b) above cannot be met, an ordnly shutdown shall be initiated and the reactor shall be in the cold shutdown condition within 24 hours. l Ilash The plant is designed to operate with both reactm coolant loops and assoc.iated reactor coolant pumps in operation and maintain DNBR above 1.18 during all normal opeintions 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 loaps 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 require that at least two loops be operable. Thus, if the reactor coolant loops are not operable, this gecificanon requires l two shutdown cooling pumps to be operable. l ! The requirement that at leasi one shutdown cooling loop be in operation during refueling ensures that: (1) sufficicat cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 2101 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 stratiGeation. l I 2-2b amendment Nc. 56, Order-4/20/81,70, 7h92
, ,- -w.. . , , , - -,--c.. , s--v--- , - - - -- .- -
2.0 1 IMITING CONDITJONS FOR OPERATION
- 2. l' Reactor Coolant System (Continued) 2.1.1 Operable Components (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 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 omling loop, adequate time is provided to initiate emergency procedure; to cool _ the core. The restrictions on availability of the containment spray pumps for shutdown cooling service ensure that ! \ the SI/CS pumps' suction header piping 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 boron concentration is being changed, the pt, cess 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 injection pump or one reactor coolant pump is in operation. The low pressure safety injection pump will circulate the reactv 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 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 during the addition of boron.* 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 system for the reactor. The LTOP enabic temperature has been established at T, == 385 F. The pressure transient analyses g demonstrate that a single PORV is capable of mitigating overpressure events Additional uncertainties j have been applied to the Pressure-Temperature (P-T) limits to account for the case where a PORV is not available (T,> 385 F), which is the reason for the apparent discontinuity in the P-T Figures. 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 steam side is 70*F; in measuring this temperature, the instrument accuracy must be added to the 7(TF limit to determine the actual measured ilmit. 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. %,4/84/ Order,74 4%
2.0 LIMITING CONDITIONS FOR OPERATION
- 2. l' Reactor Coolant System (Continued) 2.1.1 Operable Comnonents (Continued)
Formation of a 53% steam space ensures that the resulting pressure increase would not result in any overpressurization should the first reactor coolant pump be started when the steam generator secondary 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 operation. For the case in which the pressurizer steam space is less than 53%, limitation of the steam generator secondary side /RCS cold leg AT to 30 F ensures that a single low setpoint PORV would prevent an overpressurization due to actuation of the first reactor coolant pump. This requirement is not applicable to the start of a reactor coolant pump if one or more pumps are operating. The exception to Specification 2.1.l(4) requiring all containment penetrations providing direct access from the containment '.o the outside atmosphere be closed within 4 hours 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. 56, 4/81/ Order,74dM - _ _ _ _ _ _ - _ _ _ \
1 2.0 I,lMITING CONDITIONS FOR OPERATION 2.1' Reactor Coolant System (Continued) 2.1.2 Heatun and Cooldown Rate Applicability Applies to the temperature change rates and pressure of the reactor coolant system. Objective To specify limiting conditions of the reactor coolant system heatup and cooldown rates. - Specification , The reactor coolant pressure shall be limited during plant operation in accordance with Figure 2-1 A and 2-1B and as follows: (1) Allowable combinations of pressure and temperature (T,) for a specine heatup rate shall be below and to the right of the applicable limh lines as shown on Figure 2-1 A. (2) Allowable combinations of pressure and temperature (T,) for a specific cooldown rate shall be below and to the rignt of the applicable limit lines as shown on Figure 2-1B. l (3) The heatup rate of the pressurizer shall not exceed 10&F in any one hour period. (4) The cooldown rate of the pressurizer shall not exceed 20&F in any one hour period. (5) 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 umit condition on the fracture toughness properties of the reactor coolant system. (c) Determine that the reactor coolant system remains acceptable for continued operation or be in cold shutdown within 36 hours. (6) Before the radiation exposure of the reactor vessel exceeds the exposure for which they apply, Figures 2-1 A and 2-1B shall be updated in accordance with the following criteria and procedures: t j 2-3 Amendment No. 2k74 f 4
2.0 1.IMITING CONDITIONS FOR OPERATION.
- 2. l' Reactor Coolant System (Continued) 2.1,2 Heatup and Cooldown Rate (Continued)
(a) The curve in Figure 2-3 shall be used to predict the increase in transition temperature based on integrated fast neutron Dux. If measurements on the irradiation specimens indicate a deviation from this curve, a new curve shall be constructed. (b) The limit line on the figures shall be updated for a new integrated power penod as follows: the total integrated reactor thermal power from startup to the end of the new period shall be converted to an equivalent integrated fast neutron exposure (Eh 1 MeV). l The predicted transition temperature shift to the end of the new period shall then be obtained from Figure 2-3. (c) The limit lines in Figures 2-1 A and 2-1B shall be moved parallel to the temperature axis (horizontal) in the direction of increasing temperature a distance equivalent to the transition temperature shift during the period since the curves were last constructed. The boltup temperature limit line shall remain at 82 F as it is set by the NDTP of the reactor l vessel flange and not subject to fast neutron Dux. The lowest service temperature shall remain at 182 F because components related to this temperature are also riot subject to fast neutron f'ux. (d) The Technical Specification 2.3(3) shall be revised each time the curves of Figures 2-1 A and 2-1B are revised. Ihlis All components in the reactor coolant system are designed to withstand the effects of cyclic loads due to reactor coolant system temperature and pressure changes.(" These cyclic loads are introduced by normal unit load transients, reactor trips and startup and shutdown operation. During unit startup and shutdown, the rates of temperature and pressure changes are limited. The design number of cycles for heatup and cooldown is based upon allowable heatup/cooldown rates and l cyclic operation. 2-4 Amendment No. 22,43,64,W,W,400;144
2.0 I,1511TJNG CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.0 IJeatuo and Cogidown Rate (Continued) 1500 htWt and 80% load factor. The predicted shift at this location at the 1/4t depth from the inner 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 Tunt will be re-estabiished periodica'ly during the plant operation by testing of reactor vessel material samples which are irradiated cumulatively by securing them near the inside wall c# the reactor vessel as described in Section 4.5.3 and Figure 4.5-1 of the USAR. To compensate for any increase in the Tsur caused by irradiation, limits on the pressure-temperature relationship are periodically changed to stay within the stress limits during heatup 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 htWt will be 1.50x10" n/cm' on the inside surface of the reactor vessel. This results in a total shift of the RTun1 ef 298 F, including margin, for the area of greatest sensitivity (weld metal) at the 1/4t location ' 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. The limit lines in Figures 2-1 A and 2-1B are based on the following: A. Heatup and Cooldown Curves - From Section 111 of the ash 1E Code, Appendix G-2215. Ki a = 2 Ki u + Kn K i a = Allowance stress intensity factor at temperature a related to RTsur (AShtE III Figure G-2110.1). Ku= i Stress intensity factor for membrane stress (pressure). The 2 represents a safety factor of 2 on pressure. Krr = Stress intensity factor radial thermal gradient. 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 for each heatup rate by combining the most restrictive pressure-temperature limits over the complete temperature interval. For plant cooldown thermal and pressure stress are additive. 2-6 Amendment No. N,17,64,-74,-77, 400,4u,4a
FORT CALHOUN STATION UNIT 1 Pfr LIMrrS,20 EFPY HEATUP & CORE CRITICAL 2500 2500 00*F/HR 75 F/HR 2000 IM"NT AI^ '
, 2000
/
s s uJ1500 1500 h cc LOWEST SERVICE
\ CORE CRITICAL
$ TEMPEdATURE
$ 182*F " ALLOWABLE HEATUP RATES TEMR LIMIT *F RATE *F/HR
!g1000 g 4 Eau ^ti 23E
'5 100 75"F/HR 500 '
- 500 10C 'F/HR N
\ MIN. BOLTUP TEMR 82 F 0 ! i0 0 100 200 300 400 500 600 Tc INDICATED REACTOR COOLANT SYSTEM TEMPERATURE, F RCS Pres.sure-Temperature Omaha Public Power District Figure
. Limits for Heatup Fort Calhoun Station- Unit No. I 2-1A Amendment No. 75,77,100, H 4
. I FORT CALHOUN STATION UNIT 1 Pfr LIMITS,20 EFPY COOLDOWN AND INSERVICE TEST 2500 ,
2500 INSERVICE LIYDROSTATIC TESTw i 2000 2000 EO N1500 1500 5 A LOWEST SERVICE
\ -100 F/HR Tg ISOTHER\1AL h TEMPERATURES ALLOWABLE COOLDOWN RATE c-. 182 F h N
I TEMR LIMIT. F g RATE *FJIR. e 1000 ,
< 135 10
, ISOTHERMAL j b,\ 10
/ ,
500 500 10* F/HR- '/ 30'F/HR 100'F/ hrs BOLTUP TEMP, 82"E p r MIN. 0 0 100 200 200 400 500 600 Tc INDICATED REACTOR COOLANT SYSTEM TEMPERATURE. 'F RCS Pressure-Temperature Omaha Public Power District F gure l Umits for Cooldown Fort Calhoun Station- Unit No. I 2-1B Amendment No. 7 4,77,100, H 4 i
Predicted R.adiation-Induced NDTT Shift Fort Calhoun Reactor Vessel Beltline ARTndt 500 450 , 400 e i ! - i _
!.D. SHIFT
'~
l Including Margin # 350 - ! l [_ If' I 3oo Inci d ng gin
/ -
250 200 /[ /
/
/[ Including Margin 150- ,-
100 0.0 0.5 1.0_ l.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0 Neutron Fluence, IE19 n/cm* Predicted Radiation Induced Omaha Public Power District Figure l NDTT Shift iFort Calhoun Station-Unit No.1 2-3 Amendment No. 74,77,100,114,121 l
2.0 LIMITING CONDITIONS FOR OPERATION
- 2. l' Reactor Coolant Sysicatt(Continued) 2.1.2 litalup and Cooldown Curves (Continued)
Ki u = Mu 13 t hiu = ASME 111, Figure G-2214-1 P= Pressure, psk R = Vestel Radius in, t - Vessel Wall Thickness - in. Krr " M a Tw i MT = ASME 111, Figure G-2214-2 ATw = Ilighest 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 R is also a constant = B. l t Therefore: K, = AP + B P = Km - B A ib is then varied as a function of temperature from Figure G-2110-1 of ASME III and the allowable pressure calculated. Pressure correction factors for elevation and flow (-56 psia for T, <210 F and -62 psia for T, d 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 j as in A above with the exception that a safety factor of 1.5 is allowed by ASME IIIin lieu of 2. C. Lowest Service Temperature = 50"F + 120"F + 12*F = 182 F. As indicated previously, an RTuur for all nitaterial with the exception of the reactor vessel beltline was established at 50'F. 10 CFR Part 50, Appendix G, IV.a.2. requires a lowest service temperature of RTsur + 120 F for pipin;;, pumps and valves. Below this temperature a pressure of 20 percent of the system hydrostatic test pressure can not be exceeded. "Iaking into account pressure correction factors for elevation and flow, this pressure is 569 psia. D. Boltup Temperature = 10 F + 60"F + 12 F = 82 F. At pr;ssure below 569 psia, a minimum l vessel temperature must l'e maintained to comply with the manufacturer's specifications for tensioning the vessel head. 2-7 Amendment No. 22,47,64,74;400
2.0 LIMITING CONDITIONS EOR OPERATION
- 2. l' Reactor Coolant System (Coltinued) 2.1.2 Heatup and Cooldown IWg (Continued)
This temperature is based on previous ND1T methods. This temperature co: responds to the measured 10"F NDTT of the reactor vessel Gange, which is not subject to radiation damage, plus 60*F data scatter in ND1T measurements, plus 12*F instrument error. E. The temperature at which the heatup and cooldown rates change in Figures 2-1 A and 2-1B reDects the point at which the most limiting heatup and cooldown rates with respect to the inlet temperature (T,) change.
References:
(1) USAR, Section 4.2.2 (2) ASME Boiler and Pressure Vessel Code, Section 111 (3) USAR, Section 4.2.4 (4) USAR, Section 3.4.6 (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) Artic!c IWB-5000, ASME Boiler and Pressure Vessel Code, Section XI (8) Omaha Public Power District, Fort Calhe'.m Station Unit No.1, Evaluation of Irradiated Capsule W-265, March 1984, 2-7a Amendment No. 22,+7,64,-74d00
2.0 LJMITING CONDITIONS _E01LOfERATION
- 2. l' Reactor Coolant System (Continued) 2.1.6 Pressurizer and Steam System Safety Valves Applicability Applies to the status of the pressurizer and steam system safety valves.
Obsclirc To 3pecify mimmum requirements pertaining to the pressurizer and steam system safety valves. Speci6catioru 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 between 2500 psia and 2545 psia il % 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, .,:n in at least the cold shutdown condition, safety valve nozzles may be open to containment atmosphere during performance of safety valve tests or maintenance to satisfy this speci6 cation. (3) Whenever the reactor is in power operation, eight of the ten steam safety valves shall be operable with : heir lift setta.gs between 1000 psia and 1050 psia with a tolerance of I% of the nominal nameplate setpoint values.m (4) Both pressu.izer power-operated relief valves (PORV's) shall be operable during scheduled heatup and cooldown and in Modes 4 and 5, to prevent violation of the pressure-temperature l limits designated by Figures 2-1 A and 2-1B. One PORV may be inoperable for up to 7 days, provided the remaining PORV is operable. If the above conditions of this paiagraph cannot be met, be in COLD SHUTDOWN within the next 36 hours and have the primary system 2 depressurized with an RCS vent of greater than or equal to 0.94 in within the following 36 j hours. (5) Two power-operated relief valves (PORV's) and their associated block valves shall be operable 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.
2-15 Amendment No. 39,4h54
2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor coolant Svam (Continued) 2.1.6 Pressuri7cundcam System SMety Valves (Continued)
- 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 PORV's 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,
- d. With one or 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 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 leasi HOT SHUTDOWN within the next 6 hours and in COLD SHUTDOWN within the following 36 hours. Basis The highest reactor coolant system pressure reached in any of the accidents analyzed was 2480 psia and resulted from a complete loss of turbine generator load without simultancom reactor trip while operating at 1500 MWt.* The reactor is assumed to trip on a "High Pressurizer Pressure" trip signal. The power-operated relief valves (PORV's) operate to relieve RCS pressure below the setting of the pressurizer code safety 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. 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 naintenance. 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 shutdown (bde 5), so that maintenance can be performed on the PORV(s) to eliminate the seat leakage condition. To determine the maximum steam flow, the only other pressure mlieving system assumed operational is the steam system safety valves. Conservative va!ues for all systems parameters, delay times and core moderator coefficients are assumed. Overpressure protection is proviJed 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 vahe lift pressure would be less than half of the capacity of one safety valve. This specification, therefore, provides adequate defense against overpressurization when the reactor is suberitical. 2-15a Amendment No. 54 l
2.0 1,IMITING CONDITIONS FOR OPERATIOE 2.1 Reactor Coolant System (Continued) 2.1.6 Pressurizer and Steam System Safety Valves (Continued) 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, s The total relief capacity of the ten steam system safety valves is 6.54 x 106 lb/hr. 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 adjested 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. 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, Sections 4.3.4, 4.3.9.5 1 ) 2-16 Amendment No. 39,4h54
1
, 2.Q LIMITING CONDITIONS FOR OPERKl'lDE 2.3 Emergency Core Cooling System (Continued)
(3) Protection Against Low Temnerature Overoressurizalign The following limiting conditions shall be applied during scheduled heatups and cooldowns. Disabling of the HPSI pumps need not be required if the teactor vessel head, a pressurizer safety valve, or a PORV is removed. 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 pumps 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 flow 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 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 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 60"F.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 i16.2 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 systera must be checked for correct' alignment and appropriate valves locked. Since the system is used-for shutdown cooling, the valving will be changed and n,ust be properly aligned prior to start-up of the reactor. 2-22 Amendment No. 44,M,39,43,47,64,74, 77,400,403,133, 144
2.() IJMITING CONDITIONS FOR OPERATIQH 2.3 Emergency Core Coolinn System (Continued) be available for emergency core cooling, but the contents of one of the tanks is assumed to be lost through the reactor coolant system. In addition, of the three high-pressure safety injection pumps and the two low-pressure safety injection pumps, for large break analysis it is assumed that two high pressure and one low pressure operate while only one of each type is assumed to operate in the small break analysis *; and also that 25% of their combined discharge rate is lost from the reactor coolant system out of the break. The transient hot spot fuel clad temperatures for the break sizes considered are shown in USAR Section 14. l The restriction on HPSI pump operability at low temperatures in combination with the PORV setpoints ensure that the reactor vessel pressure-temperature limits would net be exceeded in the case of an inadvertent actuation of the operable IIPSI and charging pumps. Removal of the reactor vessel head, one pressurizer safety valve, or or.. PORV provides sufReient expansion volume to limit any of the design basis pressure transients. Thus, no additional relief capacity is required. Technical SpeciRcation 2.2(1) specifies that, when fuel is in the reactor, at least one Dow path shall be provided for boric acid injection to the core. Should boric acid injection become necessary, and no charging pumps are operable, operation of a single HPSI pump would provide the required flow path. The HPSI pump flow rate must be restricted to that of three charging pumps in order to minimize the consequences of a mass addition transient while at low temperatures. 1 1 2-23a Amendment No. 39,47,64,-74,77dO9
s TABLE 3-3 (Continued) , MINIMUM FREOUENCIES FOR CHECKS. CAI!!1 RATIONS AND TESTING OF MISCELLANEOUS INSTRUMENTATION AND CONTROLS Surveillance Chancel Description Function Frequency Surveillance Method Auxiliary Feedwater Flow Check M Channel . heck. 19. Calibrate R Known pressure inputs. Subcooled Margin Monitor Check M Channel check. 20. Calibrate R Known preswre inpue and known resistance substituted for RTD inputs. PORV Operation and Acoustic Check M Channel check. 21.' Position Indication Calibrate R Apply acoustic input. Verify R Operation on emergency power supply. PORV Block Valve Operation Check Q Cycle valve. ' Valve is exempt from testing when it has been 22. ad Position Indication closed to comply with LCO action statement 2.1.6(5)a. t Calibrate R Check valve stroke against limit switch position. l Verify . R Operability on emergency nower supply. Safety Valve Acoustic Check M Circuit check 23. i Position Indication - apply acoustic input. l Caintnate R PORV/ Safety Valve Tail Check M Circuit check. 24. Pipe Temperature Calibrate R Apply known input. 3-16a Amendment No. 39,51,'!0 m
ATTACHMENT B , W b V I 3 i
Attachment B
.lustification, Discussion, and No Significant Hazards Considerations The Fort Calhoun Technical Specifications Sections 2.1.1,2.1.2,2.3(3), and Figures 2-1 A,2-IB and t 2--3 are being amended to update the current pressure-temperature (P-T) limits, as well as specifications related to the low temperature overpressure protection (LTOP) system, for continued safe operation beyond 14 Equivalent Full Power Years (EFPY). This application requests continued operation through 20 EFPY.
ABB Combustion Engineering has provided an analysis to develop reactor vessel beltline P-T limits and LTOP system requirements for Fort Calhoun Station for continued operation through 20 EFPY, c The P-T limitswere calculated to meet the regulations of 10 CFR 50 Appendix A, Design Criterion 14 and Design Criterion 31. The limits were developed using the requirements of 10 CFR 50 Appendix G and ASME Section 111, Appendix G. The P-Tlimits were based ca the irradiation damage prediction methods cf Regulatory Guide 1.99, Revision 2. This methodology was used to calculate the limiting material Adjusted Reference Temperatures (ART) for Fort Calhoun Unit I for 20 EFPY of operation. The neutron fluences utilized in these calculations were calculated based on surveillance capsule measurements and DOT 4.3 calculations. No credit was taken for neutron flux reduction with extreme low radialleakage fuel management strategies implemented in Cycle 14 The predicted fluence at the lower lungitudinal scam welds (3-410) was determined to oe 1.501 x 10W n/cm.2 (E>l MeV) for 20 EFPY, Using Regulatory Guide 1.99, Revision 2, the predicted ARTat the l/4t and 3/4t locations of the criticalwelds are 29S F and 241 F for 20 EFPY, Pressure instrument loop uncertainties are not included in the proposed P-Tlimits since these limits have been included in the LTOP PORV trip setpoints. Including the pressure instrument loop uncertainties in the P-T limits therefore would have resulted in a redundant summation of these uncertainties which would be overly conservative. The pressure loop uncertainties have been included in the portion of the P-T limits which is above the LTOP enable temperature,i.e.,385 *E The LTOP analysis was performed with an objective of evaluating and modifying,if necessary, the existing provisions for low temperature overpressure protection at Fort Calhoun Station to ensure that reactor coolant pressure boundary integrity will continue to be maintained in low temperature modes of operation. The primary objective of an LTOP system is to automatically prevent pressure excursions above the applicable P-T limits during pressurization events that could result from operator error or equipment malfunction. I 1
Technical Specifications which are affected by the LTOP system requirements determined for Fort Calhoun Station by the analysis are those which concern the requirements for starting the first idle reactor coolant pump (RCP) and the limits on high pressure safety injection (IIPSI) pump availability -
' during heatup and cooldown.
One of the bases of the iTOP analysis performed by ABB Combustion Engineering is a 4tcam generator temperature that is less than 30 "F above that of the reactor coolant system (RCS) ccidleg. This assumption reduces the severityof the pressere transient associated with the start of the first RCP when the steam generator temperature exceed that of the RCS ccid leg, and, therefore, allows for a greater operating window. It was also determined as part of the LTOP analysis that a pressurizer steam space of greater than or equal to 53% would ensure that the start of a RCP, when no other reactor coolant pumps are in operation,would not result in an overpressurizationi of the RCS if the secondary temperature exceeds that of the RCS cold leg by 30 F or more. The basis for "RCS venting" was also defined more specifically as an area equel to or greater than 47 in.2, removing the reference to pressurizer manway which was to reduce the potential for mi3 interpretation of the actual vent area requirement. These assumptions moQy the current Techr'. O Specification 2.1.1(11) which requires either a 60% pressurizer steam space or less than a 50*F secondary-to primary temperature differen*ial for the start of a non-operating reactor coolant pump. The proposed revisions to the Technical Specifications clarify that these requirements apply only to the case where no RCPs are currently in operation. Another assumption that was made in the LTOP analysis that must be translated to a Technical-Specification is the limitation on allowed high pressure safety injection (HPSI) pump operability durmg heatup and cooldown. Technical Specification 2.3(3) will be modified as follows:
~
Whenever the reactor coolant system cold leg temperature is below 385 'F , at least one (1) HPSI pump shall be disabled. Whenever the reactor coolant system cold leg temperature is below 320 'F, at least two (2) HPSI pumps shall be disabled. Whenever the reactor coolant system cold leg temperature is below 270
- F, all three (3) HPSI pumps must be disabled.
In the event that no charging pumps are operable, a single HPSI pump may be made operable with mass input restricted to that no greater than the three charging pump flow rate, and utilized for borie acid injection to the core. These limitations on HPSI pump operability either reduce or climinate the potential for some pressurization events. This allows for an expanded operating window and improved heatup and cooldown rates. 2 l
. . .. .. .. . .. .. . __ _ __ _ _-______a
. .. .. .. . .m The No Significant Hazards consideratius.s are discussed for each of the croposed 'llchnical .
d Specification changesin Attachments B.1 through B.3. The 'Rchnical Specification changes to pages 2-15, 2--15a, 2-16 and 3-16a were made to incorporate the requirements of Generic Letter 90-0& - The operating Modes 4 and 5 were added to sp:cify requirements in addition to the normal heatup and cooldown operations. A 72 hour time period was specified for the depressurization and venting to occur if both PORVs were inoperable. The time period for depressurization and venting is longer than that contained in the generic letter due to the safe shutdown mode for Fort Calhoun. Since Fort Calhoun was designed as a hot shutdown plant, it requires a longer time period to reach a cold, depressurized condition where personnel or reactor safety to initiate the venting operation would not be compromised. The definition of " venting" was also added to the basis to indicate an area greater than 0.94 in.2, 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 LCO statement was clarified by replacing "all" with "both". The requirement to maintain power to closed block' alve(s) was included because removal of power would render the block valve (s) inoperable, and the requirements of action statement c would apply. Power is maintained to the blockvalve(s) so that it is operable and may be subsequen tly opened to allow the PORV to be used to control reactor ccolant systern pressure. Closure of the block valve (s) establishes the reactor coolant pressure boundary integrity for a PORV that has excessive seat leakage. The integrity of the reactor coolant pressure boundary takes priority over the capability 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 valve at a time in which the PORV may not be closed due to maintenance to restore it to operable status. Action statement d. has been modified 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 blockvalve(s) cannot be restored te operable statuswithin 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 1 consistent with the use of the PORVs to control reactor coolant system pressure if the bh)ck valves are inoperable at a time when they have been closed to isolate PORVs with excessive seat leakage. 3-
~ _ _ _ _ _ _
The operating mode for meeting LCO commitments (IIOT 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 Fort Calhoun Technical Spe:ifications for safe shutdown of the unit. The safe shutdown design for Fort Calhoun is HOT SHUTDOWN. 'lechnical Specification section 2.0.1 allows 36 hours to achieve cold shutdown from a hot shutdown condition. The recommeodations in GL 90-06 were reviewed and the LCO action statements were modified to meet the intent of the G L, yet remain consistent with the other Fort Calhoun Technical Specification action statements. This ren'ains consistent with the design and operating license requirements for Fort Calhoun. i 1 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 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 bkwkvalve is closed to isolate a PnRV with excessive seatIcakage, the operability of the block valve is of;importance, because opening o r the block valve is necessary to permit the PORV to be used for manual control of reactor 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 ellowable limits (25 percent) to extend the block valve surveillance interval (92 days). Furthermore, these test requirements would be completed by the reopening of a recently closed block valve upon restoration of the PORY to an operable status. The times to complete the action statements have been conseivatively reduced to ensure prompt compliance with the requirements for safe operation. The No Sigriificant Hazards considerations are discussed for each of the proposed Technical Specification changes in Attachment B.4. r 4 _ _ _ _ _ - . - -_.J
~ ' - " ' -- '
I t Attachment B.1 Description of Amendment Requests to Change the Fort Calhoun Technical Specification Pressure-Temperature Limits: g The proposed amendment would modify 'Dechnical Specification 2.1.2, "licatup and Cooldown Rate; Figures 2- 1 A and 2-1B; and the associated Bases. The proposed amendment would replace Figures 2-1 A and 2-1B with pressure-temperature limi;s which would be valid through 20 EFPY of operation. The new pressure-temperature limit curves were developed in conjunction with the I f0P analysis pcrformed for Fort Calhoun by ABB Combustion Engineering and were used as the % .s to; the development of LTOP astem requirementr. for Fort Calhoon Station. Tbc proposed pressure-temperature limits do not contain pressure instrumentation loop uncertainties, since these $ uncertaintics are contained in the l!f0P PORV setpoints. The pressure instrumentation loop uncertainties are accounted for in the P- flimits above the LTOP enable temperature, though, since the PORV setpoint protection is not available in that range. Ilnsis for No Significant Hnzards Determinntion: This proposed amendment does not involve a significant hazards consideration because the operation of Fort Calhoun Station in accordance with this amendment would not:
- 1) Involve a significant increase in the probability or conscquences of an accident previour.ly evaluated.
The proposed change weald not increase the probability or consequence of r.ny accident since the curves are being updated for operation to higher reactor vessel neutron fluences.
- 2) Create the possibility of a new or different kind of accident from any accident -
previously evaluated. It has been determined that a new or different type of accident is not created because no newor different modes of operation are proposed for the plant. The continued use of the same Technical Specification administrative controls prevents the passibility of a new or different kind of accident.
- 3) Involve a significant reduction in a margin of safety.
The proposed curves do not constitute a significant reduction in the margin of safety since the uncertamties that are not accounted for in the P-T limits are accounted for in the LTOP PORV setpoints. This ensures that the actual reactor vessel pressure-temperature limits would not be exceeded during any postulated low temperature overpressure transient. Bawd en ine above considerations, OPPD doca not believe that this amendment involves a significant - hazards consideration as defmed in 10 CFR 50.92 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 exclusion set forth in 10CFR51.22(c)(9) and pursuant to 10CFR51.22(b) no envircnmental assessment need be prepared. 5-
, _---__a
Attachment B.2 Description of Amendment Requests to Change the Minimum Requirements for Starting a Non-Operating Reactor Coolant Pump: The proposed amendment would modify Technical Specification 2.1.1, " Operable Components", and the associated Bases on Page 2-2b and 2-2c of the Technical Specifications. The changes concern modifying the minimum requirements which must be met to start a non-operating reactor coolant pump. The current specifications state that a pressurizer steam space of at least 60% by volume, or a steam generator-to-RCS cold leg temperature differential of less than 50 'F must exist for a non-operating reactor coolant pump to be started. The proposed specification would change these requirements to a pressurizer steam space of at leaat 53% by volume, or a steam generator-to-RCS cold leg temperature differential of less than 30 F for the start of a reactor coolant pump when no other reactor coolant pumps are operating. A steam generator temperature ofless than 30 F above that of the RCS cold leg limits the severity of a reactor coolant pump start with a steam generator temperature higher than that of the RCS cold leg. The recent ABB Combustion Engineering analysis also showed that a pressurizer steam volume of greater than or equal to 53% is sufficient to allow a reactor coolant pump start without verifying the secondary-to piimary temperature differential. These limitations apply only to the start of the first reactor coolant pump when operating in the LTOP region (Tc below the LTOP enable temperature of 385 F) and no reactor coolant pumps are alreadyin operation. These requirements do not need to be raet for the start of subsequent reactor coolant pumps. liasis for No Significant Hnzards Determinntion: This proposed amendment does not involve a significant hazards consideration because the operation of Fort Calhoun Station in accordance with this amendment would not:
- 1) Involve a significant increase in the probability or consequences of an accident previously evaluated.
Limiting the secondary-to- primary temperature differential to less than 30 *F decreases the consequences of a RCP start transient, and is therefore conservative, since less energy would be added to the primary during such a transient. The consequences of a reactor coolant pump start would not be increased by changing the pressurizer steam volume requirement since the analysis has shown that the LTOP system would protect the vessel pressure-temperature limits.
- 2) Create the possibility of a new or different kind of accident from any accident previously evaluated.
It has been determined that a new or different type of accident is not ercated because no new or different modes of operation are proposed for the plant. The continued use of the same Technical Specification administrative controls prevents the possibility of a new or different kind of accident. 6
i Itasis for No Signifiennt IInznrds Determination: (Cont)
- 3) Involve a significant reduction in a margin of safety.
These changes will not reduce the margin of safetysince the LTOP system is designed such that the reactor vessel pressure-temperature limits will not be exceeded during a RCP startup associated pressure transient at low temperature. i Based on the above considerations, OPPD does not believe that this amendment involves a significant hazards consideration as defined in 10 CFR 50.92 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 exclusion set forth in 10CFR51.22(c)(9) and pursuant to 10CFR51.22(b) no emironmental assessment need be prepared. 7
l 1 1 Attachment B.3 Descripuon of Amendment Requests to Change the Requirements for Disabling IIPSI Pumps During Scheduled IIcatup and Cooldown Operations: The proposed amendment would modify Technical Specification 2.3(3), " Protection Against Low Temperature Overpressurization"; and the associated Bases on pages 2-22 and 2-23a of the Technical Specifications.The changes concern modifying the temperatures at which the high pressure safety injection (HPSI) pumps must be disabled during heatup and cooldown. A requirement is also placed on operation of one HPSI pump below 270
- F such that the mass addition from one llPSI pump would not exceed that from three charging pumps.
These changes are being made in order to ensure that an inadvertent starting of the enabled pumps would not cause the reactor vessel P-Tlimits to be exceeded given the LTOP PORV setpoints that will be incorporated in the Fort Calhoun Station Operating Manualin conjunction with this amendment. Ilasis for No Significant IInzards Determination: This proposed amendment does not involve a significant hazards consideration because the operation of Fort Calhoun Station in accordance with this amendment would not:
- 1) Involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed changeswould r, change the probabilityor consequences of a HPSI pump start transient since the proposed disable temperatures were chosen such that an inadvertent start of the available pumps would not cause the RCS pressure to exceed the vessel pressure-temperature limits.
- 2) Create the possibility of a new or different kind of accident from any accident previously evaluated.
It has been determined that a new or different type of accident is not created because no new or different modes of operation are proposed for the plant. The continued use of the same Technical Specification administrative controls prevents the possibility of a new or different kind of accident.
- 3) Involve a significant reduction in a margin of safety.
The margin of safety will not be reduced since an analysis has shown that the proposed changes will ensure the reactor vessel P-Tlimits would not be exceeded during an inadvertent start of enabled HPSI pumps. Based on the above considerations, OPPD does not believe that this amendment involves a significant hazards consideration as defined in 10 CFR 50.92 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 exclusion set forth in 10CFR51.22(c)(9) and pursuant to 10CFR51.22(b) no environmental assessment need be prepared. 8
{ Attachment B.4 Description on Amendment request to change the Fort Calhoun Tbchnical Specification Power Operated Relief Valve-(PORV) Limiting Conditions of Operation (LCO) and Surveillance Requirement. The proposed amendment would modify Tbchnical Specification 2.1.6 and Table 3-3 and th associated basis, as described in the discussions and justification section of the proposed amendment. The proposed amendment would modify the LCO action statements for the PORVs and associated block valves while providing an exception for surveillance testing of closed block valve (s). The LCO action statements and surveillance test were modified in response to Generic Letter 90-06. The changes implement the NRC staff position from the resolution of Generic Issue 70(GI-70). The objective of the changes is to enhance safety as associated with the PORVs und corresponding block valves. Ilasis for No Significant liar rds Determination: This proposed amendment does not involve a significant hazards consideration because operation of Fort Calhoun Station in accordance with this amendment would not:
- 1) Involve a significant increase in the probability or consequences of an accident previously evaluated.
Credit is not taken for the PORVs in transient overpressure analyses in the USAR The changes described earlier will reduce the probability .,J a RCS Deprenuri ation Event by clarifying position and electrical power requirements for the block valves lherefore, this change will not involve a significant increase in the probability or consequeno a af an accident previously evaluated.
- 2) Create the possibility of a new or different kuu of accHent from any accident previously evaluated.
The potential failure modes have been previously evaluated in the USAR. A failue of a PORV with the block valve open will result in a RCS Depressurization Event. Failure of a PORV with the block valve closed will not have any adverse consequences since the RCS pressure boundary is maintained. No credit is taken for the operation of the PORV(s)in any safety analysis. Therefore,it has been determined that a new or different kind of accident from any accident previously evaluated in the USAR will not be created. 9 _ _ - - _ _ _ _ _ _ . -a
Ilasis for No Significant flazards Deternhaation: (Cont.)
- 3) Involve a significant reduction in a margin of safety.
None of the changes will require a reduction in any margin of safety, since the analyses previously contained in the USAR remain valid for the changes described nbove. Therefore, no significant reduction in the margin of safety is required. Based on the above considerations, OPPD does not believe that this amendment involves a significant hazards consideration as defined in 10CFR50.92 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 exclusion set forth in 10CFR51.22(c)(9) and pursuant to 10CFR51.22(b) no emironmental assessment need be prepared. 1 4-3 e 10 _ - _ _ _ _ _ _ _ _}}