ML20032E728
| ML20032E728 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 11/17/1981 |
| From: | OMAHA PUBLIC POWER DISTRICT |
| To: | |
| Shared Package | |
| ML20032E722 | List: |
| References | |
| NUDOCS 8111200850 | |
| Download: ML20032E728 (22) | |
Text
{{#Wiki_filter:. 2.0 LIMITING CONDITIONS FOR OPERATION 2.14 Engineered Safety Features System Initiation Instrumentation Settings (Continued) (3) Containment High Radiation (Air Monitoring) (Continued) The set points for the isolation function have been selected 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 separat9 instrument a.c. bus and each auxiliary relay requires power to operate. On fa' lure of a single a.c. suoply, 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 rede: tion 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 safety 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 pres-sure setting. This is done to minimize the temperature reduction in the reactor coolant system in the event of a main steamline break. The setting of 488 psia includes a +53 psi uncertainty; therefcre, 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 switch-over point of 16 inches above tank bottom is set to prevent the pumps 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. (6) Low Steam Generator Water Level As part of the AFW actuation logic, a signal is provided to initiate AFW flow to two steam generators upon sensing a low water level in the steam generator if the 8111200950 8113gg-DR ADOCK 05000285 2-62 ATTACHMENT A p PDR
2.0 LIMITING-CONDITIONS FOR OPERATION 2.14 Engineered Safety Features System Initiation Instrumentation Settings (Cortinued) (6) Low Steam Generator Water Level absolute steam generator pressure criteria or differential steam generator pressure criteria are satisfied. This function ensures adequate steam generator water level is maintained in the event of a failure to deliver main feedwater to either steam generator. The setting of 60.0% of wide range tap span includes a +25.0: uncertainty; therefore, a setting of 35% of wide range tap span was used in the safety analysis. (7) Hjgh 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 pressures are less than 488 psia. If the differential pressure between steam generators is.less than the setting neither steam generator is supplied with AFW in the presence of a low level signal. The setting of 88 psid includes a -47 psi uncertainty; therefore, a setting of 135 psid was used in the AFW safety analysis. 2-62a
[0) ' ~ Q.] 50 ! TABLE 2-1 ' Engineered Saraty Fantures System Initiation Inctrument Sntting Li tits g a Functiona? Unit Channel Setting Limit E ~ 5 psig Safety Injection (3) = 1. High Containment Pressure a. Containment Spray b. 4:- c. Containment Isolation. d. Containment Air Cooler DBA Mode II) 2. Pressurizer Lov/ Low Pressur9 n. Safety Injection 1 1600 psia 3) b. Containment Spray c. Containment Isolation d. Containment Air Cooler DBA Mode Containment Ventilation Isolation (h) 1 RM-050, 9.6 x 10-2 pei/sec ~ 3 Containment High Radiation < RM-051, 1.5 x 10-3 pei/ce ? 7 RM-060, 9.6 x 10-2 pei/sec 9 5RM-061,9.6x10-2 pei/sec i RM-062', 1.5 x 10-3 pei/cc I4 Lov Steam Generator Pressure
- a.. Steam Line Isolation 1 500 psia (2) b.
Auxiliary Feedvater Actuation > h88 psia l 5 SIRW Low Level Svitches Recirculation Actuation 16 inches +0, -2 in. above tank bottom 6. k.16 KV E=argency Bus Lov a. Loss of Voltage (2995 2 + 10h) volts ] Trip 20.8 Voltage 15 9(5) seconds s b. Degraded Voltage
- 1) Bus lA3 Side
> 3825 52 volts Trip Th.8 +.5) seconds S e* e
l h I TABLE 2-1 (Continu;d) g ~ Engineered Safety Features System Initiation Instrument Setting Limits a P Functional Unit Channel Setting Limit 6. (continued) b. (continued)
- 11) Bus 1Ah Side
> 3724.08 volts Trip Th.8 +.5) seconds T. Low Steam Generator Water Level Auxiliary Feedwater Actuation 1 60.0% of vide range tap span 8. High Steam Generator Delta Auxiliary Feedwater Actuation. ji 83 psid Pressure Y9 m (1) May be bypassed below 1700 psia and is automatically reinstated above 1700 psia. (2) Psy be bf;assed below 550 psia and is automatically reinstated above 550 psia. Simultaneous high containment pressure and pressurizer lov/ low pressure. (3) (b) RM-050 and RM-051 may be inoperable or out of service with respect to containment monitoring, provided that the containment ventilation isolation valves are closed. RM-061 and RM-062 may be inoperable, provided that RM-050 and RM-051 are monitoring the ventilation stack. RM-060 may be inoperable, pro-vided that (1) iodine samples are taken from the ventilation stack and analyzed each eight hours and (2) gas decay tank releases are not made. (5) Applicable for bus voltage < 2995 2 - 20.8 volts only. (For voltage > (2995.2 - 20.8) volts, time delay shall be > 5.9 seconds.)
TABLE 2-3 Instrument Oneratine Requirements for Engineered Safety Features Minimum Minimum Permissible Operable Degree of Bypass No. Functional Unit Channels Redundancy _ Conditions 4 1 Safety Injection A Manual 1 None None B High containment Pressure A 2(a) (d) 1 During leak Test B 2(a) (d) C Pressurizer Icv / Low A 2(a) (d) 1 Reactor Coolant Pressure B 2(a) (d) 1 Pressure Less Than 1700 psia (b) 2 Containment Spray A Manual 1 None None 2((a)(c)(d) B High Containnent Pressure A 1 During leak Test 2 a)(c)(d) y B C Pressurizer Lov/ Icv A 2(a)(c)(d) 1 Reactor Coolant B 2(a)(c)(d) 1 Pressure pess Than 1700 psiatD) 3 Recirculation A Manual 1 None None B SIEW Tank Low Level A 2(a) (d) 1 None B 2(a) (d) i k Emergency Off-Site Power Trip A Manual 1(*) None None B Emercency Bus Iov Voltage (Each Bus) - Loss of voltage 2(d). 1 Reactor Coolant - Degraded voltage 2(a)(d) 1 Temperature Less Than 3000F Amendment No. 41 2-68
S TABLE 2-3 (Continued) Instrument Operating Requirements _for Engineered Safety Features Minimum Minimum Permissible Test, Maintenance, Operable-Degree of Bypass- & Inoperability No. Functional Unit Channels Redundancy Conditions Bypass 5 Auxiliary Feedwater A Manual 1 None None N/A B~ Auto. Initiation A 0'perating B Modes 3, 4, and 5 - Steam Generator Low Level 3(a)(f) 1 (h) - Steam Generator Low Pres-3(a)(g) 1 (h) sure - Steam Generator Differen-tial Pressure 3(a)(g) 1 (h) a A and B actuation circuits each have 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. 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 cae 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. h One channel may be bypassed for up to 48 hours from time of initial loss of operability. 2-68a
TA3LE 2-5 '1.
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Inst unentation Creratine Recuirements for Other Safety Feature Pmetiens M.inimum Mini =um Per=1ssible Operable Degree of Bypass 1. P.:netional Unit Channels Redundancy Cenditiens 1 CEA Positica Indicatien 1 Ncne None Syste=s 2 Pressurizer Level 1 None Not Applicable 3 Subecoling Margin 1 None Not Applicable Monitor 4.. FORY Accustic Position 1 ac None Not Applicable Indication-Direct 5 Safety Valve Acoustic 1 ac None Not Applicable Position Indication 6 PCRV/ Safety Valve Tail 1 db None Not Applicable {::, Pipe Te=perature NOTES:
- a. One channel per valve, b One RC for both PORV's; two RD 's, one for each code safety, c If ite= - 61s operable, require =ents of specification 2.15 are modi-fied for items 4 and 5 to " Restore inoperable channels to operability vithin 7 days or be in hot shutdown within 12 hours."
d If ite=s 4 and 5 are operable, requirements of specificatica 2.15 are =odified for ite= 6 to " Restore inoperable channels to oper-ability within 7 days or be in hot shutdown within 12 hours," i I k.. A sad ent No. 5k 2-70 ,, _ ~ _ _.., _. _ _ _.
t ~ j TABLE 3-2 (continued) !i MInittuM FacouMNcIES Fon CnecxS. estionATroriS AND TESTIrlG OF 1 EllGINEERED SAFETY FEATlHIES. INSTHilt!KHTATION AND CONTROLS Q 4 S Surveillance u Channel Description Function Frequener Surveillance Method 1 sn. 22. Auxiliary Feedwater v a. Steam Generator Water a. Check S 'a. Compare independent level { 3 Level low (Wide Range) readings. b. Calibration R b. Known signal applied to senc7r. b. Steam Generator a. Check S a. Compare independent pressure Pressure Low readings. I y b. Calibration R b. Known Signal applied to sensor. i c. Stea5 Generator a. Calibrate R a. Known signal applied to sensor. l P l Differential Pressure High j d. Actuation Circuitry a. Test M a. Functional check of initiation-n. circuits. i i b. Test R b. System functional test of AFW l initiation circuits. 41 i S - Each Shift j D - Daily 4 M - Monthly Q _ Quarterly l R - 18 Months P - Prior to,Each Start-Up if Not Done Previous Week j MP - Monthly during designated modes and prior to taking the reactor critical if not completed within the previous 31 days (not applicable to a fast trip recovery) i i 3 /
\\_ 6.0 I'CERIM SPECIAL TECH'IICAL SPZCIFICATIO:IS - 3. yy o.3 Auxiliart Feed'.ater Automatic Initiation Settoint (This Specification is Deleted - Page Intentionally Left Blank) 6 , 3m
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ATTACHMENT B Discussion Reference 1 contains the NRC requirements for the auxiliary feedwater system at the Fort Calhoun Station Unit No.1. Included in this docu-ment was requirement GL1 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 Fort Calhoun Station including a conmitment 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 Reference 6. The Fort Calhoun 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 employed in the safety grade actuation circuitry for the auxiliary feedwater system. These setpoints are incorporated into the proposed Technical Specification changes. The analyses conducted for th.e 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 Specifica-tion system setpoints by insuring that the resultant conditions were less severe than those shown in previous analyses. Figure 1 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 488 psia. If both steam generator pressures fall below 488 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 accept-ance criterion for the steam generator level analysis was that the steam generator tube sheet remain 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 differen-tial 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 Fort Calhoun Station. Therefore, the acceptability of the steam generator low pressure and differential pressure setpoints was analyzed based on the steamline break.
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 generators' mass versus time. The inventory in the intact steam generator reaches a minimum of approximately 28,000 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 oscillations (in Figure 3) represent the opening and closing of the intact steam generator's first safety valve, because the steam 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 7.3 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 four feet below the AFAS level setpoint and that auxiliary feedwater flow is adeouate 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 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 fed if its level and pressure were to decrease below the low level and low pressure setpoints. 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 steam line break with concurrent loss of offsite power was less severe than the case without the loss of offsite power. The District has concluded that the auxiliary feedwater 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 auxiliary feedwater system functioned properly such that the broken steam generator was differentiated from the intact generator.
1 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 differ-entiate between the intact and broken steam generators in the steamline break events to assure that the results of the FSAR are conservative. 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 ditterent 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 auto-matic auxiliary feedwater system, be deleted. It is the District's intent to operate with these proposed Technical Specifications com-mencing with Cycle 7 startup, because the control grade actuation system will be removed and the setpoints 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 60% wide range level actuation setpoint and no time delay, the safety grade actuation system provides conservative pro-tection 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 Parrell G. Eisenhut to W. C. Jones dated October 22, 1979 2. Letter from W. C. Jones to Darrell G. Eisenhut dated January 14, 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, USNRC Office of Inspectica and Enforcement, Region IV, dated May 15, 1980 8. Letter from Robert A. Clark to W. C. Jones dated February 20, 1981
SG-A SG-B I P,>Pb P 'P b a SG-A SG-A SG-B SG-B ~. f'a Level level P b t,ow Low Low Low Y ~- OR CR \\ ~. s ARID AtiD %) %) Remote Remote Manual Manual OR OR Fec I Fe:c SG-A 50-3 Figure 1 OPPD AFW Logic Diagram
I I I I I i i i i Feed SG Only i Feed Both A I L i SG and SG A B I I I I 488 PSIA ,r. Feed SGA y Only a. cgc ys Feed SGB m 4 d9 Only 9 sP Feed SGB PSID Only 0 488 PSIA SG Pressure B
- Actuation only occurs if low SG level and pressure signals are concurrent Figure 2 AFW Steam Generator Actuation
- Logic
Figure 3 Steam Generator Total Mass versus Time for the Feedwater Line Break with Offsite Power Available 900C0 s 8C000 I i 7CCCC 1 h 60C00 l Intact Steam Generator 5 C 000 )~ r \\ o i [ 40C00' E e n 30000 20000-- Affected Steam Generator f -(- 10000 0 r i_ 0 100 200 300 400 500 600 70C 8C0 TIME. SECCt!CS
Figure 4 Steam Generator Pressure versus Time for the Feedwater Line Break with Offsite Power Available 1:00 Intact Steam Generator p -{%waywwwvwnvntNrnwrrenvvvtra
- 000 c
I G (; I acc f,o ..G iV = v> h SCC $ c. e E U 400 200 Affected Steam Generator C O 100 200 300. 400 500 600 700 8C0 T!"E. SECCt103 i e ano' o,+_ a w.
FIGURE 5 STEAM GENERATOR TOTAL MASS vs. TIME FULL POWER MAIN STEAMLINE BREAK S.G. TOTAL MASS (LBm)' 100000 ~ INTACT S.G. 90000 / / l r /N l 80000 b 70000 ~ 60000 g ~ y 50000 N N 40000 \\ 30000 N N 20000 s 10000 ,,,,,,,,,,,,w. O., 0.0 10.0 20.0 30.0 40.0 50.0 60.0 70.0 80.0 90.0 100.0 110.0 120.0 130.0 140.0 TIME (sec) ( l
FIGURE 6 STEAM GENERATOR PRESSURE vs. TIME FULL POWER MAIN STEAMLINE BREAK S.G. PRESSURE (psia) 900.0 800.0 ~ [N 700.0 s { - A N 600.0 x x 500.0 N t N ~ 400.0 \\ x 300.0 ^ s 200.0 N N R PTIE D S.G. 100.0 ~f f f I f f f f I f f f f I f f f f f f f I f f f f f f I f f f f f f f f f f f f f f f f f f f f f f f f I f f 0.0 10.0 20.0 30.0 40.0 50.0 60.0 70.0 80.0 90.0 100.0 110.0 120.0 130.0 140.0 TIME (sec)
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s FIGURE 8 STEAM bENERATOR PRESSURE vs. TIME ZERO POWER MAIN STEAMLINE BREAK S.G. PRESSURE (psia) 900.0 800.0 700.0 - l 600.0 ~ 500.0 _~ N N N 400.0 N N N 300.0 ^ INTACT S.G. 200.0 s N~ 100.0 RPTWED S.6. ~ ' 0.0 - 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)
JUSTIFICATION FOR FEE CLASSIFICATION The proposed amendment is deemed to be Class III, within the meaning of 10 CFR 170.22, in that it involves a single safety con-cern. ATTACR4ENT C}}