ML20049J954
| ML20049J954 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 03/22/1982 |
| From: | OMAHA PUBLIC POWER DISTRICT |
| To: | |
| Shared Package | |
| ML20049J948 | List: |
| References | |
| NUDOCS 8203290253 | |
| Download: ML20049J954 (21) | |
Text
, .
2.0 LIMITING CONDITIONS FOR OPERATION 2.14 Engineered Safety Features System Initiation Instrumentation Settings (Continued) -
(3) Containment Iligh Radiation (Air Monitoring) (Continued)
The set points for the isolation function have been se-1ected to limit radioactivity concentrations at the boundary of the restricted area to approximately 0.25 of 10 CFR 20 limits, assuming existence of annual average meteorology.
Each channel is supplied from a separate instrument a.c.
bus and each auxiliary relay requires power to operate. On failure of a single A.C. supply, the A and B matrices will assume a one-out-of-two logic.
(4) Low Steam Generator Pressure A signal is provided upon sensing a low pressure in a steam generator to close the main steam isolation valves in order to minimize the temperature reduction in the reactor coolant system with resultant loss of water level and possible addition of reactivity. The setting of 500 psia includes a +22 psi uncertainty and was the setting used in the sa'fety analysis.(3)
As part of the AFW actuation logic, a separate signal is provided to terminate flow to a steam generator upon sensing a low pressure in that steam generator if the other steam generator pressure is greater than the pressure setting. This is done to minimize the temperature re-duction in the reactor coolant system in the event of a main steamline break. The setting of 466.7 psia includes a i +31.7 psi uncertainty; therefore, a setting of 435 psia was used in the safety analysis.
(5) SIRW Tank Low Level Level switches are provided on the SIRW tank to actuate the valves in the safety injection pump suction lines in such a manner so as to switch the water supply from the SIRW
! tank to the containment sump for a recirculation mode of operation after a period of approximately 24 minutes following a safety injection signal. The switchover point of 16 inches above tank bottom is set to prevent the pumps i from running dry during the 10 seconds required to stroke the valves and to hold in reserve approximately 28,000
! gallons of at least 1700 ppm borated water. The FSAR loss of coolant accident analysis (4) assumed the recirculation started when the minimum usable volume of 283,000 gallons
[
had been pumped from the tank.
1 (6) Low Steam Generator Water Level As part of the AFW actuation logic, a signal is provided to initiate AFW flow to one or two steam generators upon sensing a low water level in the steam generator (s) if the 8203290253 820322 i PDR ADOCK 05000285 12 , 4 3 2-62 ATTACHMENT A
-P PDR
~ . . - - -. .. .
._ _ . _ . _ ~
2.0 LIMITING CONDITIONS FOR OPERATION 2.14 Engineered Safety Features System Initiation Instrumentation Settings (Continued) -
(6) Low Steam Generator Water Level (Continued) absolute steam generator pressure criteria or differential steam generator pressure criteria are satisfied. This function ensures adequate steam generator water level is i maintained in the event of a failure to deliver main feedwater to either steam generator. The setting of 28.2% of wide range tap span includes a +13.2% uncertainty; therefore, a setting of 15% of wide range tap span was used in the
- safety analysis.
2 (7) High Steam Generator Delta Pressure As part of the AFW actuation logic, a high steam generator differential pressure signal is generated to provide AFW to the higher pressure steam generator with a concurrent low level signal if both steam generator pressurer are less than 466.7 psia. If the differential pressure between steam generators is less than the setting neither steam t generator is supplied with AFW in the presence of a low level signal. The setting of 119.7 psid includes a -15.3 psi uncertainty; therefore, a setting of 135 psid was used j in the AFW safety analysis.
1 i
e i
i c.
4 l
2-62a
TABLE 2-1 .
Enrineered Safety Fentures System Initiation Instrument Setting Limits u *
& Functional Unit Channel Setting Limit g
e w 1. High Contain=ent Pressure a. -< 5 Psig
. b.
Safety Injection (3)
Containment Spray s- c. Containment Isolation.
" d. Containment Air Cooler DBA Mode
- 2. Pressurizer Lov/Lov Pressure a. Safety Injection > 1600 psia (1)
- c. Containment Isolation
- d. Contain=ent Air Cooler DBA Mode Contain=ent High Radiation Containment Ventilation Isolation (h) 1 RM-050, 9.6 x 10-2 pei/see 3 < RM-051, 1 5 x 10-3 pei/ce 7 RM-060, 9 6 x 10-2 pet /see Y 7
_ RM-061, 9.6 x 10-2 pei/see 9 1 RM-062', 1 5 x 10-3 pei/cc
- k. Lov Stea:n Generator Pressure a. . Steam Line Isolation > 500 usia(2)
- b. Auxiliary Feedvater Actuation 466.7 psia l 5 SIRW Low Level Switches Recirculation Actuation [6 inches 1 +0, -2 in, above tank bottom b.16 KV zmergency Bus Lov a. Loss of Voltage (2995.2 + loh) volts ]5 Trip ,
- 6. ' Voltage - 20.8 15 9(5) seconds s
- b. Degraded Voltage
- 1) Bus lA3 Side > 3825.52 volts Trip D.8+.5) seconds
N TABLE 2-1 (Continued) g -
Engineered Safety Features System Initiation Instrument Setting Limits
@g r>
x Setting Limit
. Functional Unit Channel
- 6. (Continued) b. (Continued) ii) . Bus lAh Side > 3724.08 volts 7,gp Th.8 + .5) seconds Auxiliary Feedvater Actuation > 28.2% of vide range tap span 7 Low Steam Generator Water Level
< 119. 7 psid
- 8. High Steam Generator Delta Auxiliary Feedvater Actuation .
Pressure
?
D o
(1) May be bypassed below 1700 psia and is automaticc11y reinstated above 1700 psia.
(2) Pay be bypassed below 550 psia and is automatically reinstated above 550 psia.
(3) Simultaneous high contain=ent pressure and pressurizer low / low pressure..
(4) RM-050 and RM-051 may be inoperable or out of service with respect to containment RM-061 and RM-062 monitoring, provided may be inoperable, that the containment. ventilation isolation valves are closed. RM-060 may be inoperable, pro-provided that RM-050 and RM-051 are monitoring the ventilation stack.vided that (1) iodine (2) gas decay tank releases are not made. (For voltage > (2995 2 - 20.8) volts, time
($) Applicable for bua voltage < 2995 2 - 20.8 volts only. ,
delay shall be > 5 9 seconds.) ,
O
l 2.0 LI:1ITING CONDITICUS FOR OPERATICN 2.15 Instrumentation and Control Systems Applicability Applies to plant instrumentation systems.
Objective To delineate the conditions of the plant instrumentation and control systems necessary to assure reactor safety.
1 Specifications The operability of the plant instrument and control systems shall be in accordance with Tables 2-2 through 2-5. g In the event the number of channels of a particular system in service falls below the limits given in the columns entitled
";11nimum Operable Channels" 7r " Minimum Degree of Redundancy",
except as conditioned by the column entitled " Permissible By-pass Conditions", the reactor shall be placed in a hot shut-down condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; however, operation can continue without containment isolation signals available if the ventila-tion isolation valves are closed. If minimum conditions are not met vitbin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the reactor shall be placed in a cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
If, during power operation, the rod block function of the second-ary CEA position indication system and rod block circuit are inoperable for more than 24 hou s, or the plant computer PDIL alarm, CEA group deviation alarm and the CEA sequencing function are inoperable for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the CEA's shall be with-drawn and maintained at fdlly withdrawn and the control rod drive system mode switch shall be maintained in the off posi-tion except when manual motion of CEA Group h is required to control axial power distribution.
Basic During plant operation, the complete instrumentation systems vill normally be in service. Reactor safety is provided by the reactor protection system, which autcmatically initiates appropriate action to prevent exceeding established limits.
Safety is not compromised, however, by continuing operation with certain instrumentation channels out,of service since provi-sions were made for this in the plant design. This specifi-cation outlines limiting conditions for operation necessary to preserve the. effectiveness of the reactor control and pro-tection system when any one or more of the channels are out of service.
All reactor protection and almost all engineered safety feature channels are supplied with sufficient redundancy to provide the capability for channel test at power, except for backup channels such as derived circuits in engineered safeguards control system.
- Amendment No. 8, 20, Sh 2-65
TAllLE 2-3 Instrument Oneratine Retini rement s for Enrineered Safety Features Minimum Minimum Permissible Operable Degree of Bypass No .__ Functional Unit Chnnnels Redundancy Conditions 1 Safety In.iection A Manual 1 None None B High Containment Pressure A 2(a} (d) 1 During Leak Test 3 2(al (d)
C Pressurizer Lov/Iov A 2(a) (d) 1 Reactor Coolant Pressure 3 2(a) (d) 1 Pressure s Than 1700 psia 2 Containment Spray A Manual 1 None None B High Containment Pressure A 2(a)(c)(d) 1 During Leak Test 3 2(a)(c)(d) y C Pressurizer Inv/Lov A 2(a)(c)(d) 1 Reactor Coolant 3 2(a)(c)(d) 1 Pressure s 6 an 1700 psia 3 Recirculation A M:inual 1 None None B SIRV Tank lov Level A 2(a) (d) 1 None 3 2(a) (d) 1 4 Fmere,ency Orr-Gi te Power Trip A Manual 1(e) None None B Emergency Bus Iov Voltage
- (Each Dus) - Ioss of voltage 1 Reactor Coolant
! 2((d)
- Dc6raded voltage 2 a)(d) 1 Temperature less Than 300or i
Amendment No. 41 2-68 '
TABLE 2-3 (Continued)
Instrument Operating Requirements for Engineered Safety Features Minimum Minimum Permissible Operable Degree of Bypass No. Functional Unit Channels Redundancy Conditions 5 Auxiliary Feedwater A Manual 1 None None B Auto. Initiation A Operating B Modes 3, 4, and 5
- Steam Generator Low Level 3(a)(f) 1 l - Steam Generator Low Pres-j sure 3(a)(g) 1
- Steam Generator Differen-tial Pressure 3(a)(g) 1 a A and B actuation circuits each hav'e 4 channels.
b Auto removal of bypass above 1700 psia.
c Coincident high containment pressure and pressurizer low / low pressure signals required for initiation of containment spray.
4 d One of the inoperable channels must be in the tripped condition.
e Control switch on incoming breaker.
f Two channels are allowed to be inoperable provided that one and only one is in the low level actuation permissive condition.
g Three channels required because bypass or failure results in auxiliary feedwater actuation block, in the affected channel.
f i
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! 2-68a
TA3LE 2-5 Instrumentatien Operatine Fecuirsnents fer Other Safet r Feature Functions Minimu= Mini =um Permissible Operable Degree of Bypass No. Functional Unit Channels Redundancy Ccnditions 1 CIA Positica Indication 1 Ncne None Syste=s 2 Pressurizer Level 1 None Not Applicable 3 Subcooling Margin 1 None Not Applicable Monitor
- 4. PORV Acoustic Position l ac None Not Applicable Indication-Direct 6 Safety Valve Acoustic l ac None Not Applicable Position Indication 6 PORV/ Safety Valve Tail 1 db None Not Applicable Pipe Temperature NOTES:
[
a One channel per valve.
b One RTD for both PORV's ; two RTD's , one for each code safety.
C If iten 61s operable, requirements of specification 2.15 are modi-fled for items 4 and 5 to " Restore inoperable channels to operability within 7 days or be in hot shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />."
d If items 4 and 5 are operable, requirements of specification 2.15 are nodified for item 6 to " Restore inoperable channels to oper-ability within 7 days or be in hot shutdovn within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />."
i A: sad:ent :fo. 5L 2-70 i,
f TABLE 2-6 4- .
\
t DELETED Amendment No. 54 2-70a
[
TABLE 3-2 (Continued) i N
e HIIIIt!UM FREQUENCIES FOR Cl!ECKS, CALIDRATIO!iS AND TESTIflG OF
" E!!GIf3FERED SAFETY FEATIIRIS. IllSTRl!!!EIITATIOff A!!D CG!!'Ill3!E I
5 Surveillance u Channel Description Function Frequency Surveillance Method sn 22. Auxiliary Feedvater s-
! a. Steam Generator Water a. Check S 'a . Compare independent level
! Leve] low (Wide Range) readings.
l
- b. Calibration R b. Known signal applied to sensor. [
- b. Steam Generator a. Check S a. Compare independent pressure Pressure Low readings. ,
t Y
y b. Calibration R b. Known Signal applied to sensor. i U t
" c. Steam Generator a. Calibrate R -
- a. Known signal applied to sensor.
Differential Pressure High I
Test M a. Functional check of initiation i
- d. Actuation Circuitry a.
circuits. ,
1~
i b. Test R b. System functional test of AFW . l initiation circuits.
r S - Each Shift D - Daily M - Monthly Q - Quarterly R - 18 Months P - Prior to,Each Start-Up if Not Done Previous Week MP - Monthly during designated modes and prior to taking the reactor critical if not completed l within the previous 31 days (not applicable to a fast trip recovery)
J
6.0 I::Tra::4 sPzcIAL Trem!IcAL SPrcIFIcATIc:rs '
6.3 Auxilia:r 7 ateat er Au+.or.atic Initi att en Setr,oint (This Specification is Deleted - Page Intentionally Left 31arl) i l
l Acend:est .'Jo, S g,3
DISCUSSION Reference I contains the NRC requirements for the auxiliary feedwater system at the Fort Calhoun Station Unit No. 1. Included in this docu-ment was requirement CLI for the installation of safety grade actuation circuitry for the auxiliary feedwater system. Reference 2 contains the District's responses to the NRC requirements for the auxiliary feedwater system at the Fort Calhoun Station, including a commitment to install the safety grade actuation circuitry during the 1981 refueling outage.
Requirements for the auxiliary feedwater system were further amplified in Reference 3. In Reference 4, the District supplied the design of the safety grade auxiliary feedwater system to the NRC.
The District was required, in Reference 5, to reanalyze the steamline break event for the control grade auxiliary feedwater system. A generic analysis applicable to the Fort Calhoun Station was provided in Refer-ence 6. The Fort Calhoun Station specific analysis was provided in Reference 7 in response to IE Bulletin 80-04. The Commission's safety evaluation report for the auxiliary feedwater system at the Fort Calhoun Station is contained in Reference 8.
This document provides the safety analysis for the setpoints to be em-ployed in the safety grade actuation circuitry for the auxiliary feed-water system. These setpoints are incorporated into the proposed Techni-cal Specification changes. The analyses conducted for the safety grade system utilized the same computer codes and techniques as those used for the control grade system in References 6 and 7. These analysis methods are consistent with those used in the FSAR. The analyses were conducted to determine the adequacy of the proposed Technical Specification system setpoints by ensuring that the resaltant conditions were less severe than those shown in previous analyses.
Figure I shows the logic diagram of the auxiliary feedwater actuation signal (AFAS) circuitry. The requirements for initiating auxiliary feedwater flow to a steam generator are a concurrent low steam generator level signal and a permissive signal based on steam generator pressure.
Figure 2 shows a simplified logic diagram for the pressure dependent permissive signal. The diagram shows that in the presence of a low steam generator level signal, a generator will be fed if its steam generator pressure is greater than 466.7 psia. If both steam generator pressures fall below 466.7 psia, the differential pressure logic is used to determine which steam generator will be fed.
The three setpoints were assessed based upon two criteria. The ac-ceptance criterica for the steam generator level analysis was that the steam generator tube sheet r; main covered. To meet this criterion the limiting loss of primary heat sink events were analyzed. These included the loss of main feedwater and feedwater line break, both with and without a concurrent loss of offsite power. The combined low pressure and differential pressure logic of the auxiliary feedwater actuation circuitry was designed to prevent supplying feedwater to a ruptured steam generator. A ruptured steam generator is not fed in order to minimize the potential effects of additional overcooling of the reactor coolant system for a steamline break event at the Fort Calhoun Station.
Therefore, the acceptability of the steam generator low pressure and differential pressure setpoints was analyzed based on the steamline break.
ATTACHMENT B
i l
For the low level setpoint verification, it was determined that the feedwater line break with offsite power available results in the most .
rapid depletion of steam generator inventory. During this event, the auxiliary feedwater system is initiated in time to ensure that an adequate heat sink is maintained. Figure 3 shows a plot of both steam i generators' mass versus time. The inventory in the intact steam gener-ator reaches a minimum of approximately 18,646 lbs. mass. Only one auxiliary feedwater pump was credited, thus resulting in a slow recovery of level in the intact unit. Figure 4 shows a plot of steam generator pressure versus time. The pressure oscillations and the mass oscil-lations (in Figure 3) represent the opening and closing of the intact i
steam generator's first safety valve, because the stecm dump and bypass system is not credited in the analysis. An actuation signal to feed the broken steam generator does occur prior to the pressure dropping below the low pressure setpoint; however, the duration of this condition (i.e., pressure above the cutoff point) is only 6.8 seconds which is
- less than the delay time for flow to reach the ruptured unit. The results of the limiting loss of heat sink transient are that the steam generator level falls approximately two feet below the AFAS level set-point and that auxiliary feedwater flow is adequate for restoration of normal steam generator level and complete decay heat removal.
The steamline break was reanalyzed to verify that the ruptured steam generator would not be fed and that, if necessary, the intact steam generator would be fed by the auxiliary feedwater system. Figures 5 and 6 show plots of the total steam generator mass versus time and steam generator pressure versus time, respectively, for the full-power large main steamline break. Figures 7 and 8 show the same parameters plotted i
J for the zero-power case. Initially, both steam generator pressures quickly decrease because the single failure assumed is that the reverse flow check valve on the ruptured unit is stuck open. This results in a blowdown of the intact unit until the MSIV's close. During this time period, neither steam generator has a low level signal, so AFAS will not occur. Af ter the MSIV's close the low pressure and delta pressure setpoints prevent the broken steam generator from being fed. The delta pressure setpoint enables the intact unit to be ied if its level and pressure were to decrease below the low level and low pressure set-points. The intact unit is not fed (as can be seen in Figures 5 and 7) because the mass inventory remains high. These same trends can also be seen for the steamline break cases without offsite power. Based on these analyses, it can be concluded that the auxiliary feedwater system will not feed a ruptured steam generator and, therefore, this steamline
- break analysis is bounded by the analysis contained in the FSAR. In addition, in Reference 8 the NRC concluded that the return to power for a steamline break with concurrent loss of offsite power was less severe than the case without the loss of offsite power. The Distric.t has concluded that the auxiliary fecduater system will properly actuate for either case.
Because there is a possibility that smaller steamline breaks could
- become more limiting if the auxiliary feedwater system did not properly diagnose the broken steam generators, intermediate and small steamline breaks were examined. The examination of the smaller steamline breaks showed that the larger steamline break was conservative and the auxi-i liary feedwater system functioned properly such that the broken steam generator was dif ferentiated from the intact generator.
Based on these analyses, the District has concluded that the auxiliary feedwater system actuation logic is adequate to 1) maintain level in the steam generators for loss of primary heat sink events, and 2) to dif-ferentiate between the intact and broken steam generators in the steam-line break events to assure that the results of the FSAR are conserva-tive. Therefore, it can be concluded that the auxiliary feedwater system actuation logic does not involve an unreviewed safety question because 1) the probability of occurrence or consequences of an accident or equipment important to safety previously evaluated in the FSAR is not increased, 2) the possibility of accident or malfunction different than the type evaluated previously in the FSAR is not created, and 3) the margin of safety as defined in the basis of any Technical Specification is increased. Additionally, the proposed amendment to the Technical Specifications is consistent with the CE Standard Technical Specifi-cations.
These Technical Specification changes require that Interim Special Technical Specification 6.3, which addresses the control grade automatic auxiliary feedwater system, be deleted. The District is operating with these proposed Technical Specifications commencing with Cycle 7 startup, since the control grade actuation system has been removed and the set-points used with the safety grade system are conservative with respect to the control grade system. For loss of feedwater events, the control grade system could receive an actuation signal at 0% level in the narrow range level instrument and then, after a 180 second delay, flow would commence. The 0% steam generator narrow range level corresponds to 58.5% wide range level. With a 28.2% wide range level actuation set-point and no time delay, the safety grade actuation system provides conservative protection in comparison to the control grade system.
Additionally, the new safety grade system will not feed a ruptured steam generator. _
LIST OF REFERENCES
- 1. Letter from Darrell G. Eisenhut to W. C. Jones dated October 22, 1979
- 2. Letter from W. C. Jones to Darrell G. Eisenhut dated January 1h,1980
- 3. Letter from Darrell G. Eisenhut to All Licensees of Operating Plants and Applications for Operating Licenses (NUREG-0737) dated October 31, 1980
- h. Letter from W. C. Jones to Darrell G. Eisenhut dated December 30, 1980 5 Letter from Robert W. Reid to W. C. Jones dated December 21, 1979
- 6. Letter from W. C. Jones to Robert W. Reid, Director of Nuclear Reactor Regulation, dated January 10, 1980 7 Letter from W. C. Jones to K. V. Seyfrit, Director, USITRC Office of Inspection and Enforcement, Region IV, dated May 15, 1980
- 8. Letter from Robert A. Clark to W. C. Jones dated
. February 20, 1981 l
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- [SID Only 0 466.7 PSIA i
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ressure :-
- Actuation only occurs if low SG 1evel and pressure signals are i concurrent i
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Figure 2 AFW Steam Generator Actuation
- Logic
Figure 3 Steam Generator Total Mass Versus Time for the Feedwater Line Break
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- Intact Steam Generator tR 2"" -
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Affected Steam Generator 0
0 in 200 300 400 500 800 700 TN. SEEGG5
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FIGURE 8 STEAM GENERATOR PRESSURE vs. TIME ZERO POWER HAIN STEA. TINE BREAK S.G. FAESSURE (psia) 900.0 ,
800.0 700.0 ,
600.0 ,
500.0 N
400.0
- v x x N
N 300.0 _
s _ ^ I m CT S.6.
e 200.0 s 100.0
- x RLPTWED S.S.
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0.0 10.0 20.0 30.0 40.0 50.0 60.0 70.0 80.0 90.0 100.0 110.0 120.0 TIME (sec)
- - - - -