ML20247L327

Proposed Tech Specs Re Limiting Conditions for Operation of Pressurizer & Main Steam Safety Valves

ML20247L327
Person / Time
Site:	Fort Calhoun
Issue date:	09/08/1989
From:	OMAHA PUBLIC POWER DISTRICT
To:
Shared Package
ML20247L313	List: ML20247L310 ML20247L316 ML20247L327
References
NUDOCS 8909220184
Download: ML20247L327 (20)

Text

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.i TECHNICAL SPECIFICATIONS TABLE OF CONTENTS Pace DEFINITIONS........................................................... 1 1.0 S AFETY LIMITS AND LIMITING SIFETY SYSTD4 SETTINGS. . . .1-1 ........ [

1.1 S afety Limit s .- Re acto r Co re . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 i

1.2 l Safety Limit , Reactor Coolant System Pressure. . . . . . . . . . . . . . 1L 13 Limiting Safety Sye^em Settings , Reactor f P ro t e c t ive Sys t ua . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-6 {

1 2.0 LD1ITING CGNDITIONS FOR OPERATION................................ 2-0

2. 0.1 General Requirements....................... 1

........ 2-0 {

2.1 Reactor Coolant i System..................................... 2-1 2.1.1 Ope rab le Compone n ts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2- 1 1

2.1.2 'l He atup and Co oldown Rate . . . . . . ~ . . . . . . . . . . . . . . . . . . . . 2- 3 2.1.3 Be acto r Coolant Radioactivity. . . . . . . . . . . . . . . . . . . . . . .

2.1.4 Reactor Coolant System Leakage Limits. . . . . . . . . . . . . . . 2-8 2.1.5 Maximum Reactor Coolant Oxygen and Halogens 2-11

- Concentrations g...t 2.1.6 Pressurizer and Steam Oyr in........................... 2-13 s Se fety Valves . . . . . . . . . . 2-15 2.1.7 Pres s uriz e r Ope rability . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-16a l 2.1.8 Re actor Coolant Syst em Vents . . . . . . . . . . . . . . . . . . . . . . . .

2-1cb 2.2 Chemical and Volume Control 0ystem.........................

2.3 E ergency Core Cooling System................... 2-17 2.h Containment Cooling........................................ 2-20 2.5 ............ 2-2L S t e am and Feedvate r Systems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-2 8 2.6 Containment System.................................

2.7 2- 30 Ele c t ri cal Sys t ems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

2.8 Re fuelin g Ope rat i ons . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ....... .....

2- 3 2 2.9 Radioactive 2-37 Effluents...................................... 2-k0 2.9.1 Liquid and Gas eous Effluents . . . . . . . . . . . . . . . . . . . . . . . . 2 LO 2 9.2 Solid Radi oactive Was te . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-h 7a p 2.10 Re a c t o r Co re . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .2.L. 3. . .

2.10.1 Minimum Conditions for Criticality. . . . . . . . . . . . . . . . . . 2 h8 2.10.2 Reactivity Control Systems and Core Physics Parameter Limits................................ 2- 50 2.10. 3 In-Co re Instrumentation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-5L 2.10. L Powe r Dist rib ution Limits . . . . . . . . . . . . . . . . . . . . . . . . .2-56 ..

2.11 Contain; nt 3uilding and Fuel Storage B uilding Crane. . . . . . . . 5 8 -

) i Amendment No. 37, 33, 52, fs, 57, 67 F M , SI s 3h 8009220184 890908 FDR ADOCt: v5000283 F PDC

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.. . n. o. gpr9A. . n..I 2.1 Feaet:r ': elan: 5. stem (Continued) 2.1. o. W , ., 3 re.o .n Pressu-1:er ind steam- cvste r Safetv Valves Aculiesb!litr WY6 % % VI Applies to the status of the pressurizer and steam valves. A system safety Obieetive, To specify r.ini=um requirements pertaining to the pressurizer 2nd a stez= s m en safety valves.

.n . .a Seee:fiestiens To provide 2decua e overpressure protection far the resator c0 clan:

systen and steam syste=, the folleving safety valve rec.u reser.:: shall te met:

(1)

' he rese pressur :err shall safe:-not be me critical unless the.. .t,v,o

/ valves are operacle with their4_iftyt settings 11.'us ,ed to ensure valve opening between 2500 ps ia. .* J %

ps:z a.c. 25.!.5 psia +1",.(1) T h e. v a in.s h a <-. a- deer a 4,,. .m t4.- n.m.e.i 3 e Tr'*b S d 0 ' '"5 o f " * ' ' " -

u -w a nc. e. o 9 M (2} 'icenever there is fuel in the reactor, and the reactor vessel head-is installed, a mini =u: of cce operable si' sty 7117e : hall be installed on the pressurizer.

B.ever, v.en in at least the cold shut'evn :endition, sa: *:y valve nettles may be open to containment atmos-

he re during perfer. ance of safety valve tests or mainte-ner.

e o satisfy this specification.

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-nere;tgr ten o s.es:

u.ne reactor is in pcVer operaticn, eight ofghp,,3 safety valves shall be operable vith their, lift settings te:veen 1000 psia and 1050 psia with a t9 1grance of +1.1 of the nc=inal nameplate set point values. \lJ % e. 'lal 'eS 6 '8 e A J c',Or c. li ..o.nc e of- t 3 & - 2 7. G,,-om ::Le 'v- no dw.1 3e rr.-.p Juc;,3 (1)

?cth pressurizer power-operated relief valves (PORV's) 9 m E*a.

sh:11 he Operable during scheduled heatup snd cocidown prevent violatica of the pressure-temperature limits desi.nated by Figures 2-1A and 2-13. Oce PC3V may be incpers is 0;er:ble.

le f r up to 7 days, previaed the remaining FORV

• f the sbcVe conditi:ns of this paragraph Ottnt, be net, the primary system must be depressurized an: 7ented.

(3} :v,: :ver-cpers:n relief valves (?C?.7's ) and their 1s-j se:i:.:et b:.::r. valves shall be operable in Moces 1, 2 A tr. : 2.

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, f 260 LIMITI30 00NDITICNS FOR OPERATION A, ~2. l '

3eseter Ccolant Erstem (Continuedl 2.".5 .

?ressuri er and. Steam System Safety Valves .(Continued) i r 't h  !

a.

'41th one or more FORV(s) inoperable, within 1 hour:

- either restore the FORV(s) to operable status or close the associated block valve (s);' otherwise, be in at.least HOT STANDBY vithin the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOW vithin the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. ,

b.

' lith one or more block valvels) ineperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the block valve (s) to oper-able status or close the block valve (s). Other-vise, be in at least HOT STAUEBY vithin the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDCW within the following 2h hours.

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The _ h:.;; hest \ reactor coolant /

the accident 3 analy ec wast system pressure reached in.any of EL and resulted from a com-plete loss of turbine gener a @ tor load vijhout simultaneous re-actor trip vhile operating at 1500 M'4t. (2) The reactor is assuned to trip on a "Righ Pressurizer Pressure" trip signal.

The pover-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 pr: vide a positive shutoff espability should a re-lief valve bee me inoperable. The electrical power for both the re;ief nives and the block valves is capable of being l supplied fr:e sn emergency power source to ensure the ability to seal this possible RCS leakage path.

To determine the max 1=um steam flov, the onlygher pressure relieving system assumed operational is thensteam syetim safe.y valves.

Cctservative values for all systems parameters, de-lay times and core moderator coefficients n.re assumed. Over-pressure protection is provided to portions of the reactor coolant system which are at the highest pressurre considering pu=p head, flov. pressure d;rops and elevation heads.

M If ne resiinal heat vere removed by any of the meens avail-

• - able, the amount of steam which could be generr.ced at safety i

valve .

of :ne safety'ift pressurevalve. would be less than half :s the capacity j This specification, tt erefore, prc~ ides

\y adec.uate defense against overpressurization when the reactor y '

is suberitical.

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2.1 Reacter Coolunt System LContinued) 2.1.6 ?reszurt er enc Steam Svstem Safet7 Valves (Continued)

Perf:r:ance of certain calibration and maintenance procedures on safety valves requires removal from the pressurizer. Should a safety valve be removen, either operability of the other safety valve or =aintenance of at least ene no::le open to atnosphere vill assure that sufficient relief capacity is avail-able. Use of plastic or other similar material to prevent the entry of f: reign =cterial into the open no::le vill not te construed to violate the "cpen to atmosphere" provision, since the presence of this esterial vould not significantly restrict the discharge of reacter ecolant.

ns . n The total relief capacity f the ten steam n syste= safety valves is 6.Sh x 100 lb/hr. At the power of 1500 !Nt, sufficient relief valve capacity is available to prevent overpressuriza-tien Of the stee: syste= n loss-of-load ccnditions. l")

The p:ver-cperated relief valve lov setpoint vill be adjusted to pr vide sufficient margin, when u:ed in conjunction with Technical Specificati:n Sectict.s 2.1.1 and 2.3, to prevent the design basis pressure transients free causing an over-pressuri:ation incident. Limitation of this requirement to scheduled cooliovn ensures that, should e=ergency conditions dictate rapid cocidevn of the reactor coolant syste=, inoper-

)

' ability of the lov te=pers are overpressure protection syste=

vould not prove to be an inhibiting factor.

Eeteval cf the react:r vessel head proviaes sufficient expan-sien vo,1t=e to li=it any of the design basis pressure tran-sients.' Thus, no additional relief capacity is required. (

References (1)

ArticleSecti:n Cede, 9 of the:::

1968 ASMI 3 oiler and Pressure Vessel U

(2) /SAR Secticn 14 9 4 J

(;I) /SAR, 2ections h.3.5, L.3 9 5 i

3} v5g,5 g Jsw 1 'l . I O i

?. enizen: ::. p/'. 2 7 5L -;6 L

4 4 P #

e i ATTACHMENT B

4 JUSTIFICATION, DISCUSSION, AND NO SIGNIFICANT HAZARDS CONSIDERATIONS FOR REVISION OF THE MAXIMUM ALLOWABLE SETP0lN1 DRIFT FOR THE PRIMARY AND SECONDARY SAFETY VALVE SETPOINTS The proposed amendment to the Technical Specifications would increase the maximum allowable setpoint drift for the primary and secondary safety valve setpoints from.+_ 1% to +3% / -2%.

This amendment was initiated in response to LER's87-003 and 87-014 which resulted from safety valve setpoint drifts in excess of that allowed by Technical Specifications 2.1.6(1) and 2.1.6(3). To support this amendment, the NRC-approved CESEC-III transient analysis code and analysis methods were used to determine the acceptable setpoint drift for the primary and secondary safety valves.

Acceptable setpoint drift is defined as the maximum allowable drift in the primary and secondary safety valve setpoints which would ensure that the peak RCS pressure and peak secondary pressure do not exceed the design basis acceptance criteria (i.e. 110% of design) of 2750 psia and 1100 psia, respectively, and 14.10. The as specified in the Updated Safety Analysis Report, Sections 14.9 wording in the Basis for Technical Specification 2.1.6 has been revised and no longer contains the highest primary pressure calculated in the transient analyses.

The revised wording specifies that the highest primary pressure reached in any of the accidents analyzed is less than 2750 psia, which is the Safety Technical Limit for the Specification Reactor Coolant System Pressure as specified in L2. Future transient analyses will continue to demonstrate that the highest reactor coolant system pressure is below 2750 psia, value.but a facility license change will no longer be necessary to update this Based on past experience and review of the NRC approved OPPD reload licensing methodology, the Loss of Load and Loss of Feedwater events were found to be most limiting.

ient for determining the maximum steamThe Loss of Load event was analyzed as th While the analysis was performed with Cycle 11 conditions, generator pressure.the results bound Cycl LOSS OF LOAD EVENT The Loss of Load event was reanalyzed using the same assumptions as Reference 1, except that the allowable setpoint drift for the primary and secondary safety valves was changed from +1% to +3%. The Loss of Load analysis was initiated at the conditions, and for the combination of parameters shown in Table 1-1 to maximize the calculated peak RCS pressure and resulted in a high pressurizer pressure trip signal at 8.9 seconds. At 11.3 seconds, the primary pressure reaches its maximum value of 2649 psia. The increase in secondary pressure is limited by the opening of the main steam safety valves, which open at 6.8 seconds. The secondary pressure reaches its maximum value of 1078 psia at 14.4 seconds after initiation of the event.

Table 1-2 summarizes the sequence of events for this transient. Figures 1-1 through 1-4 show the transient behavior of power, RCS pressure, RCS coolant temperatures, and steam generator pressure.

LOSS OF FEEDWATER FLOW EVENT The total Loss of Main Feedwater Event was reanalyzed using the same assumptions in Reference 1, except that the primary safety valves were modeled as inoperable and the allowable drift for the secondary safety valves was changed from +1% to +5%. This was a bounding case used in the analysis, however, an allowable drift of +3% is proposed for this amendment to ensure ,

margin. The Loss of Feedwater Flow analysis was initiated at the conditions,  !

and for the combination of parameters shown in Table 2-1 to maximize the calculated peak RCS pressure and peak secondary pressure and resulted in a high pressurizer pressure trip signal at 34.9 seconds. At 37.9 seconds, the primary pressure reaches its maximum value of 2515 psia. The increase in secondary pressure is limited by the opening of the main steam safety valves, which open at 38.7 seconds. The secondary pressure reaches its maximum value of 1099.6 psia at 42.5 seconds after initiation of the event.

Table 2-2 summarizes the sequence of events for this transient. Figures 2-1 through 2-4 show the transient behavior of power, RCS pressure, RCS coolant temperature and steam generator pressure.

MANUAL REACTOR TRIP A manual reactor trip case with no setpoint drift was analyzed to determine the maximu.n allowable negative setpoint drift that would not challenge any of the safety s A -4% drift was determined to be the maximum allowable, however,ystems.

an allowable drift of -2% is proposed for amendment to ensure adequate margin is maintained to prevent challenge of the ESF system or components.

Based on these analyses, the enclosed application for amendment to the Technical Specifications requests revision of the allowable setpoint drift for the primary and secondary safety valves from +1% to +3% / -2%.

Basis for No Significant Hazards Considerations The proposed amendment to the Technical Specification does not involve a significant hazard consideration because the operation of the Fort Calhoun Station in accordance with this amendment would not:

(1) Involve a significant increase in the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report. The proposed revision to the allowable setpoint drift was conservatively analyzed using the NRC approved transient analysis methodology and computer code (CESEC-III). The results demonstrate that during a severe transient the peak RCS pressure and peak steam generator pressure would fall significantly below the Safety Limit and design basis acceptance criteria of 2750 psia and 1100 psia, respectively, as specified in the Updated Safety Analysis Report Sections 14.9 and 14.10. Since the safety valves function to control transient events, revision of the allowable setpoint drift would not increase the probability of occurrence of such events. Therefore, this amendment would not significantly increase the probability of occurrence or consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report.

_p.

i Basis for No Significant Hazards Considerations (cont.)

(2) Create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report.

safety valves function to control transient events. Analysis of theThe the NRC approved transient proposed allowable analysis methodology setpoint and computer drift using(CESEC-III) demonstrates code during the limiting overpressure transients, peak RCS pressure and peak steam generator pressure would be significantly below the design basis acceptance criteria of 2750 psia and 1100 psia, respectively, as specified in the Updated Safety Analysis Report Sections 14.9 and 14.10.

No new or different kind of accident is crehted because actual operation of the plant remains unchanged. Therefore, the possibility of an accident or malfunction of a different type than any evaluated previously in the safety analysis report would not be created.

(3) Involve a significant reduction in the margin of safety as defined in the basis for any Technical Specification. This revision only increases the allowable primary and r.econdary safety valve setpoint drift within .

safety limits as demonstrated using the NRC approved transient analysis  !

methodologyandcomputercode(CESEC-III). Therefore, the margin of safety as defined in the basis for any Tecnnical Specification is not reduced.

Based on the above considerations, OPPD does not believe that this amendment involves a significant hazards consideration.

REFERENCES 1.

Fort Calhoun Station, Unit No. 1, Amendment No. 109 to Facility Operating License No. DPR-40 (TAC Nos. 64434 and 64486) from NRC (Walter A. Paulson) to OPPD (R. L. Andrews) dated May 4, 1987.

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Table 1-1 FORT CALHOUN CYCLE 11 KEY PARAMETERS ASSUMED IN THE LOSS OF LOAD TO BOTH STEAM GENERATO Parameter Units Value Initial Core Power Level MWt 1530 (102%)

Initial Core Inlet Temperature *F 547 Initial Pressurizer Pressure psia 2,053 Initial Steam Generator Pressure psia 815 Initial RCS Flow Rate gpm 196,000 Moderator Temperature coefficient 10~0 Ap/*F + 0.5 Fuel Temperature Coefficient 10 ap/*F Least negative predicted during core life.

Fuel Temperature Coefficient Multiplier 0.85 CEA Time to 100% Insertion (Including Holding coil Delay) sec. 3.1 Scram Reactivity Worth %Ap -6.65 Kinetics Parameters S .004696 Allowable Primary and Secondary Safety Valve Setpoint Drift  % +3 l

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Table 1-2 FORT CALHOUN CYCLE 11 SEQUENCE OF EVENTS FOR THE LOSS OF LOAD EVENT TO MAXIMIZE CALCULATED RCS PEAK PRESSURE Time (sec) Event Setooint or Value 0.0 Loss of Secondary Load ----

6.8 Steam Generator Safety Valves Open 1066 psia 8.9 High Pressurizer Pressure Analysis 2422 psia Trip Setpoint is Reached 10.2 CEAs Begin to Drop Into Core ----

10.6 Pressurizer Safety Valves Open 2575 psia  :

11.3 Maximum RCS Pressure 2649 psia 14.4 Maximum Steam Generator Pressure 1078 psia i

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i Table 2-1 FORT CAL 110UN CYCLE 11 KEY PARAMETERS ASSUMED IN Tile LOSS OF FEEDWATER FLOW ANALYSIS Parametel Units Value Initial Core Power Level MWt 1530 (102%)

Initial Core Inlet Temperature *F 547 Initial Pressurizer Pressure psia 2,053 Initial Steam Cenerator Pressure psia 815 Initial RCS Flow Rate gpm 196,000 Moderator Temperature Coefficient 10~0 ap/*F + 0.5 Fuel Temperature Coefficient 10'O ap/*F Least negative predicted during core life.

Fuel Temperature Coefficient Multiplier 0.85 CEA Time to 100% Insertion (Including Holding Coil Delay) rec. 3.1 Scram Reactivity Worth tap -6.65 Kinetic = Faiaeters S .004696 Allowable Primary Safety Valve Setpoint Drift

% Inoperable Allowable Secondary Safety Valve Setpoint Drift t +5

Table 2-2 FORT CALHOUN CYCLE 11 SEQUENCE OF EVENTS FOR THE LOSS OF FEE 0 WATER FLOW EVENT TO MAXIMIZE CALCULATED RCS PEAK PRESSURE Time (seql Event Setooint or Value 0.0 Loss of Secondary Load ----

34.0 High Pressurizer Analysis Trip Setpoint is Reached 2422 psia 36.2 CEAs Begir. to Drop Into Core ----

37.9 Maximum RCS Pressure 2515 psia 38.7 Steam Generator Safety Valves Open 1066 psia 42.5 Maximum Steam Generator Pressure 1099.6 psia i

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