ML20113G552
| ML20113G552 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 12/04/1989 |
| From: | Bell H GEORGIA POWER CO. |
| To: | |
| Shared Package | |
| ML20092F288 | List:
|
| References | |
| CON-IIT05-191-000B-90, CON-IIT5-191-B-90, RTR-NUREG-1410 LO-HO-12101-002, LO-HO-12101-2, NUDOCS 9202210472 | |
| Download: ML20113G552 (25) | |
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p GEOROIA POWER POWER GENERATION DEPARTMENT V0GTLE ELECTRIC GENERATING PLANT TRAINING STUDENT liANDOUT ITLE:
LOSS OF RHR - INDUSTRY EVENTS NUMBER:
LO-HO-12101-002-C 10 GRAM: LICENSED OPERATOR TRAINING REVISION:
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BELL DATE:
11/30/89
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FERENCES o
NOP.464 SER 86.035 IEN 86.101 SOER 85.004 IEB 80.012 IEN B"J.023
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LO-HO-12101-002-04-C 4
Excerrt From NOP_-Aff 1
MID-LOOP OPERATION CONCERNS i
1.
BACKGROUND On July 9, 1987, the Office of Nuclear Reactor Regulation issued Generic Letter 87-12, " Loss of Residual Heat Removal (RHR) While the Reactor Coolant System (RCS) is Partially _ rilled". This letter deals With NRC concerns related to operation of a Pressurizer Water Reactor (PWR) when the RCS water level is below the top of the reactor vessel, in particular when the water-level is near the mid-loop elevation. At the Westinghouse owner's Group-general session meeting in late September of 1987, Westinghouse was authorized to perform analysis and testing related to Item _5 and Enclosure 1 of the NRC Generic Letter 87-12.
Westinghouse has completed the analysis and testing and-has issued a draf t report that is currently in review by the Mid-Loop Operations Working Group.
The final report is: scheduled to be completed and available to the full Woo by mid-July following review and comment.
The. intent of this letter is to summarice come important preliminary
.results that should be considered by utilities when performing certain types'of maintenance while operating at or near mid-loop conditions.
In
.particular, there is a concern with a loss of RHR cooling: scenarios in which there.is a large cold leg side opening.
This postulated-scenario is worse if the loop lwith the opening is isolated on the hot leg side of the opening.
A number of cautions for RHR operation, level indication, and vortex formation:are also described.
2.
CONCERNS REGARDING LOSS OF RHR SCENARIOS
-2.1. Scenario 1: Large Cold Side Opening, Loop Isolated The scenario considered here involves loss-of-RHR cooling when there is a large cold. side opening and the loop with the opening is isolated on the-hot leg side of'the opening. The cold side opening could be caused by removal of an So manway for so tube inspection or maintenance.or removal of a large cold leg check valve or loop isolation valve for repair'and
. inspection. The loop would be considered isolated due to installation of SG nozzleidame or closure of the loop isolation valves.
r The specific cases of-primary concern include loss of RHR when either'an
-opening in.the SG manway with the hot side SG nozzle dam installed.(or hot leg loop isolation valve closed); or an opening in the cold leg (e.g.,
check valve-opening)-with either the hot side or cold side or both SG dams installed'(either or-both hot leg and cold leg loop isolation valves closed).
Under this postulated condition, the RCS will pressurize faster
-at the core exit than at the cold leg, following the loss of RHR cooling.
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p4 LO-HO-12101-002-04-C RCS inventory will then be forced out-of the cold side opening at a rapid rate.
Typically, the core will become uncovered within several minutes af ter tne onset of boiling.
Because the SG nozzle dams (or rioned loop isolation valver) do not allow a vapor vent path to the opening, the core.
will remain uncovered for a prolonged period of time pnless actions are L
taken to restore RCS inventory in a-timely manner.
For mid-loop operation prior to a typical refueling, the RCS would be expected to reach saturation in about 30 minutes after the loss of RHR cooling.
However, at more limiting conditions for mid-loop operation (e.g., 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after reactor shutdown, 1400F initial RCS temperature), the RCS could reach saturation in less than 10 minutes.
In view of this potentially short time to the onset of boiling and core uncovery, it is important to emphasize the need to prevent 'ose of RHR cooling for this scenario.
If RHR cooling is lost, recovery actions should-be taken before boiling occurs to minimize the possibility of a prolonged core uncovery.
To avoid prolonged core uncovery for this scenario, it was found that hot leg injection at a sufficiently high rate would be effective in suppressing boiling and refilling the RCS.
The hot leg injection flowrate is cons'idered high enough if the core residual heat is less than the sensible heat. required to raise the temperature of the makeup water to saturation.
Determination of this condition is explained in the following example.
For the 2-loop plant used in WoG analysis, the injection flow required to match the core decay heat at 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> shutdown time following extended operation at 1520 MWt would be 62 lbm/sen or 450 gpm. -This estimate is based on a core decay heat of 0.48% of-full power or 7.3 MWt (6915 BTU /sec).
The injection water is assumed to be heated from 100 F (a typical RWST 0
temperature) to 2120F.
This flow is typically within the capacity of one high-head SI pump for a 2-loop plant. One or possibly-two high-pressure SI pumps would be required for comparable conditions in most 3-loop and 4-loop plants.
The analysis for the 2-loop plant demonstrated that the recovery with hot
' leg injwetion was successful even at-a slightly lower flowrate (50 lbm/sec
= 360 gpm).
It was not.possible to-demonstrate successful recovery using cold leg-injection at a comparable flowrate since the amount of cold water reaching:the core was not adequate to suppress boiling.
s For the case where the opening is in the SG manway and the loop with the opening is likely to be isolated, the scenario daccribed above would be c
made less likely and less severe if the cold side 41 nozzle dams are installed first and removed last. This simple change to the order in which the nozzle dams are installed will reduce the potential for core uncovery if RHR is lost during the installation, by keeping an open vent path for the longest period of time.
If RHR is lost after the cold side SG nozzle dam has been installed, the time to_ core uncovery for this case will.be 3
prolonged by preventing release of liquid inventory to the manway until the i
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nozzle dam fails.
Nottle dam failure would be cauted by system pressuritation due to boiling.
It should be noted that nottle dams could fail arove pressures as low as 26 pulg (typical test pressure).
They can nct te counted on to prevent liquid release to the 50 manway opening abovas their cosign pressure, during a loss of RHR event.
Therefore, if RHR is lost when '.stle dams are installed, hot leg injecticn should be initiated prior to or within minutes after boiling starts.
Also, it is strongly recommended that the RCS level should be raised above mid+ loop af ter nostle dams have been anstalled to minimite the potential for loss of PHR under this configuratict.
It i s important to emphasire, once again, that utilities should take all precautions to prevent lose of RHR cooling during the configuration when there is a large cold side opening and the nostle dams or the loop isolation valves (in the same loop with the opening) are installed or closed.
If it becomes necessary to enter this configuration, the operator ohould be prepared to initiate hot leg injection early as descrihed above.
2.E 9"enario 2t Large Hot /Ccid side Opening, Loop Not isolated for the case of the 50 manway opening without the installation of the hot side _3c rostle dams (or closure of the hot leg loop isolation valves), the manway openings on either hot or cold sides of the 50 would provide a large vent path for the air and steam.
The tine to core uncovery would be signi.ficantly greater than thu previous case with the loop isolated.
For all 2, 3, and 4-loop cases studied, thw tiras to core uncovery, after accounting for potential spill, all exce6d forty minutes based on bo11+ott of the water above the top of the fuel.
For the case of a large cold leg opening (e.g., 4" or 12" check valve opening) without any of the so nottle dams installed (or loop isolation valves closed), the RCS will pressurize until the water in the pump suction (loop seal) piping is expelled.
The core may uncover briefly during this translent but not long enough for fuel temperatures to become excessive.
After the loop seal clears, a vent path to the opening is provided and the core level will stab 111:e above the top of the active fuel. The RCS inventory will then be depleted at the boil-ofi rate.
Establishing charging flow to an intact loop cold leg withit.10 minutes following loss of FHR cooling at a rate exceeding the boil-off cate will typically be sufficient to pruvent a subsequent core uncovery for this caso.
A makeup rate two to three times the boil-off rate is recommended to allow faster uncovery.
This rate is typically within the capacity of two positive-displacement charging purps for most low-pressure plants or within the capacity of one centrifugal charging /s! pump injecting in the normal charging mode for most high-pressure plants.
If the cold leg openi Q is in the loop with the charging connection, the operator should be instructed to use alternate charging, Alternatively, hot leg injection (as descrfoed in i
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4 LO-Ho-12101-002-04-c the Trevious scenario for loop isolated) could be used to restore RCS invent - ry.
c However, if the opening is in the hot leg (e.g., due to removal of an,50 manway or loop isolation valve), it is recommended th,at the operator use only normal or alternate charging to the cold leg.
Not leg injection for tiis latter case could be less effective as liquid niay spill to containmont.
- 3. CAUTIONS REGARDING RHR OPERATION WITH A PARTIALLY TALLED RCS Work done in the testing phare of the WOG program indicates several points on which operations personnel should be informed.
- 1. As expected, vortex formation and consequential air entrainment is a functio.x.f RCS level (at the RHR inlet) and RHR flowrate.
However, it was observed that once vortexing commences, return to level and flow operating conJitions that preclude initial vertex formation ray not be adequate to terminate an established vortex.
The quickest way to eliminate the vortex would be to significantly reduce flow (if plant conditions permit) while increasing level.
For plants with two RHR lines from the RCS, the operating RHR pump should be stopped; the other pump should be started at a low flowrate.
For plants with a single RHRS inlet line, venting should be accomplished prior to second pump start.
- 2. The actual level at the RHR inlet connection has a significant effect on vortex formation.
It is known that lovel gradients will exist in the RCS due to fluid momentum and density effecto.
Level measurement error can also be increased due to density effects.
Therefore, the type and location of level 1netrumentation must be considered when operating near level and flow conditions known to be unacceptable.
- 3. Once air is entrained in the RHR system, it may be very difficult to eliminate.
This is largely due to long hoeirontal piping from the RCS to the RHR pumps.
In this piping, it takes a long tims for trapped air to migrate back to the RCS.
In addi'Sion, operation of an RHR pump at low flow rates will sweep only a minimal amount of air through the inlet pipe.
If rapid removal of air from the RHR pipe is required, the RCS should be raf t11ed (to the top of the hot leg), then one RHR pump operated at normal flows.
This will sweep air through the pump.
Cautions pump performance indicators should be monitored closely af ter the RHR pump is started.
End of Excerpt from NOP-464 1-:
e e
4 LO-HO-12101-002-04-C
SUMMARY
OF SER 86.035 EXTENDED LCSS OF SHUTDOWN COOLING DUE TO STEAM DINDING While in cold chvtdown, operators at Waterford 3 (CE PWR) were draining the RCS in preparatien for replacement of a reactor coolant pump seal.
Cooling flow was being provided by the low pressure safety injection (LPSI) pump D.
The PCS was being drained via the CVCS to the Holdup Tank.
In addition, the LPSI pamp B mini-flow recite line had been opened to drain water to the Refueling Water 4torage Pool.
This was not per procedura.
The nitrogen gas being added to maintain RCS pressure et approximately atmospherie pressure was not added rfpidly enough to compensate for the drain rate, which created a slight vacuum in the RCS.
This collapsed a tygon tube standpipe being monitored locally for RCS level, resulting in an inaccurate level indication.
Level was also being monitored from the control room using the RVLIS (which does not have the securacy required to control level for shutdown cooling operations while partially drained).
Wher the optrators isolated tha drain path to CVCS, they failed to isolate the additional drain via the pump miniatlow.
When the RCS wqs vented to
'orrect problems with the tygon tubing level indication, the tygon tubing level inaication decreased to read offueale low.
RVLIS indicated level was in the vicinity of the hot legs and LPSI pump B was operating autisfactorily, su maintenance was allowed to begin.
RCS continued to drain via the pump mini-flow path.
The operating pump began to cavitate. Operators isolated the mini-flow recirc line and refilled the RCS until standpipe level indicated above the hot leg centerline.
- Less than an hour after cavitation began, core exit thermocouple readings of 223 degrees F indicated boiling was occurring in the core.
Attempts to restore shutdown cooling had been unsuccessful due to persistent cavitation, from air binding initially and then steam binding.
A steam bubble was trapped between the shutdown cooling line loop seal And the LPS!
pump suction. The steam bubble could not be condensed by reactor coolant flow because this water war at saturation.
The capacity of the vacuum priming system was too small to promptly remove the accumulated steam.
The LPSI pumpt were eventually started by opening warmup valves downstream of the shutdown cooling heat exchangers to return cooled water back to the pump suctions and condense the steam in those lines. (It was necessary for the optrators to open the valve motor operator supply breakers to allow-manual throttling of these valves to tneresse flow to the RCS).
Core exit 1-n l
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tt.ormo:caple temperatures decreased and were below saturation approximately i
an hour later.
Norroai shutdown cooling flow was then established.
Based on RVLIS, the core remained fully covered throughout the event.,
l COMMENTS 1.
Level should be monitored continuously when draining the RC3.
Frequent comparisons between independent level indications should be made.
Draining of the RCs should be secured at once if *evel indication is lost or becomes suspect.
RCS level should be raised if necessary to restore redun lant level indication 1
2.
Opa'raura should be conscious of all nctor coolant system drain paths in operation at any time.
A status board or other means should be used to keep operators aware of each drain path.
3.
Reactor level indicators should be capable of operation under all anticipated conditions including a vacuum. Operators should be aware of limitations on indicating systems.
For example, tygon tubing systems, such as werts used here, can collapse and give inaccurate
' indications when subjected to a vacuum.
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LO-HO-12101-002-09-c f
SUMMARY
OF ITN 86.101, ATTACH. 1 LOSS OF RHR EVENTS AT PWRS J
San Onofre Prior to partially draining the RCS for SG maintenance at Sen Onofre 2, a CE designed reactor, wide-and narrou-range RCS level inntruments were put in earvice by installing their temporary connections and calibrating them.
Tygon tubing was also installed temporarily to provide a slight gauge for monitoring RCS water level.
Thus, three devices were available for monitoring water level in the system.
i To parmit repairs, personnel began draining vessel water level to 17.5 inches above the bottom of the 42-inch diameter hot lege.
One of the hot legs suppline water to the inlet side of the shutdown cooling system (SDCS) through a connection in the bottom of the pipe. While the water level was being lowered, a vortex formod on the suction side of the low pressure (LPSI) pun p.
The vortex entrained air causing the pump to become air bound, loss of SDCS flow, and thus loss of decay heat removal.
The pump was secured and the redundant pump was started.
It, too, became air bound and was secured.
To reestablish flow through the SDCS, the system was vented, and the water level in the reactor vessel was raised.
Seventy minutes after the first indication of vorteming, decay heat removal was again established.
Meanwhile, the hot leg temperature increased from 114 to 210 degrees r, and local boiling occurred in the reactor core.
Steam and two curies of radionuclides were released to containment.
The operators did not trust installed narrow and wide-range level instrumentatioe because of its tendency to oscillate during the use of certain equipmant, and were relying on the temporary tygon tubino sight gauge for level indication.
However, during installation and filling of the tygon tubing, an air bubble was inadvertently trapped in the tubing causing it to read high by 10.5 inches.
Further, the reference scale for the tubing was displaced by 2.5 inches in the upward direction causing a total error of 13 inches (high).
- Thus, lowerAng level to 17.5 inches as indicated on sight gauge would have lowered actual level to 4.5 inches.
( Although the operator did not have confidence in the narrow range instrument, its reading was approximately correct at that time).
Vortexing started at-an actual level of about 9.5 inches.
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- ack of knowledge about the perforicance of the system at low water Ivvels and unrallable instrumentation for monitoring watet level bore the principle cauenu of this avont.
Sequoyah 1 Sequoyah 1, a Westinghouse reactor, was in cold shutdown with the water level in the reactor vesnel 4 inches below the centers of the hot leg nortles.
R4tR Tr&in h was in service for removal of decay heat.
Curing an evolution to put Train A in service, RHR pump A wcs started and then Pump B was secured.
Running both pumps s
simultaneously with low reactor vossol water level caused initiation of vortexing and air binding in Pump A.
The pump was secured, and Pump 1 was restarted and operatved norme11y.
The alignment of Train A was 'terified and the pump was vented.
Pump D was secured, and Pump A was restarte'
~* this time it became air bound and was secured.
Vassel love war
~ ed
.ad abnuh 43 minutse nfter loss of decay heat removal, Pur e & u t + ~ ' +d i returned tb service.
Pump 8 was vented, demo,irrn ao t.p5h, anet doenergized.
Both pumps take sues.on from the same hot leg, whose level would not i
support operation of both pumps simultaneously.
The proceduto for operating RHR with pa tt ially dr&ined vessel did not a6equately reflect the relationship between RHR flow rate and water level for t'ho onset of vortexing in the auction line for the RHR pumps.
Catawna 1 Catawba 1 (Westinghouse) was in cold shutdown with RHR B in service to remove decay heat.
Although RHR w..a inopergble due to maintenance, the licensee started to lower the water level in the reactor vessel fo; maintenance. While deatning was in progress, erratic performant's of RHR Pump 8 indicated that vortexing, air sfstrainment, and air-binding were occurring.
The pump was secured and vessel level was raised using a charging pump aligned with RWST.
RHR Pump B wan returned to servics.
Temperature of the RCS peaked at 177 degrees T.
l It is believed that information obtained from inaccurate twvel instrumentation contributed to loss of RHR at this plant, whose procedure for lowering water level in the vessel does limit RHR flow as a function of level, apparently to preclude the onset of vortening.
Further, ths licensee incurred an increased riek of loss of RHR flow by lowering water level with one train of RHR cooling out of service.
A Tsch Spec LCO requires that one RHR trgin be operating and that the other be operable under the conditions present.
The operatore concluded incorrectly that water level could be lowered if corrective action had been initiated to comply with th& action statement for that limiting condition for operation.
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LO-HO-12101-002-04-C i
i e
SUHMARY OF IE INFORMATION NOTICE 80-20 LOSS OF DECAY HEAT REMOVAL CAPABILITY AT DAVIS-BESSE UNIT 1 WHILE IN A REFUELING MODE Deveription of circumstancese On April 19, 1980, decay heet removal capability was lost at Davis-Basse Unit 1 for approximately two and one-half hours.
At the time of the event, the Unit was in a refueling mode (e.g., RCS temperature was 90 F; decay 0
heat was being removed by Decay llent Loop No. 2; the vessel head was detensioned with bolts in places the reactor coolant level was elightly below the vessel head flanges; and the manway covers on top of the once through steam generators were removed). (S6a Enclosure A, status og Davis-Besse 1 Prior to Loss of Power to Busses E-2 and F-2 for additional details regarding this event).
Since the plant was in a refueling mode, many systems or components were out of service for maintenance or testing purposes.
In addition, other systems and components unre deactivated to preclude their inadvertent actuation while in a refueling mode.
wystems and components that were not in service or deactivated included:
Containment Spray System; liigh Pressure Injection System; Sourew Range Channel 2; Decay Heat Loop No.1; Station Battery IP and IN; Emergency Ciesel-Generator No. 1; 4.16 KV Essential-Switchgoar Bus c1; and 13.8 KV Switchgear Bus A (this bus was energized but not aligned)
In brief, the event was due to the tripping of a non-safeguards feodor breaker in 15.8 KV Switchgear Bus S.
Because of the extensive maintenance and testing activities being conducted at the time, Channels 1 and 3 of the PeactGr Protection System (RPS) and Safety Features Actuation System (SFAS) were being energized from only one source, the source emanating from the tripped breaker.
Since the SFAS logie used at Davis-Besse is a two-out-of-four input scheme in which the lons (or actuation) of any two input signale results in the actuation of all four output channels (i.e..
-Channels 1 and 3, and Channels 2 and 4), the loss of power to channels 1 end 3 bistables also resulted in actuation of SFAS Channels 2 and 4.
The actuation of SFAS Channels 2 and 4, in turn, affected Decay Heat Loop No.
2, the operating loop.
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since the initiating event was a loss of power event, all five levels of 4
51As were sctuated (i.e., Level 1 - High Radiation; Level 2 - High Pressure In3eetion; Level 3 - Low Pressure injections Level 4 - Containment Sprayt
[
and Level 5 - ECCS Recirculation Mode).
Actuation of SFAS Level 2 and/or 3 resulted in containment isolation and Icon of r;ormal. decay heat pump suct ion f rom RCS het leg No. 2.
Actuation of SFAS Level 3 aligned the becay Heat Pump No. 2 suction to the Dorated Water Storage Tank (BWST) in i
the low pressure injection mode.
Actuation of SFAS Level S represente a low level in the BWST; therefore, upon its actuation, ECCS operation was automatically transferred from the injection Mode to the Recirculation Mode.
As a retelt, Decay Heat Pemp No. 2, the operating pump, was I
automatically aligned to take surtion from the containment sump rather.han from the UWST or the reactor coolant system.
Since the emergency centainnent sump was dry, suction to the operating decay heat pump was lost.
As a result, the decay heat removal capability was lost for approximately two and one-half hours, the time required to vont the system.
Furthermore, since Decay Heat Loop No. I was down for maintenance, it was net available to reduce the time required to restore decay heat cooling.
MAJOR CONTRIBUTORS TC THE EVENT The rather extended lows of decay heat removal capability at Davis-Besse Unit I was due to three somewhat independent factors, any one of which, if corrected, could have precluded this event.
These three factors ares 1)
Inadequate procedures and/or administrative controls; 2)
Extensive maintenance activitieur and 3
3)-
The two-out-of-four SFAS logic.
Regarding inadequate procedures and/or administrative controls, it should be noted that the High Pressure Injection Pumps and the Containment Spray l
Pumps were deactivated to preclude their inadvertent actuation while in the refueling mode.
In a similar vein, if the SFAS Level 5 scheme had been by-passed or deactivated while in the refueling mode, or if the emergency sump isolation valves were closed and their breakers opened, this event would have been, at most, a minor interruption of decay heat flow.
Regarding the extensive maintenance activities, it appears that'this event would have been precluded, or at least ameliorated, if the maintenance I
activities were substantially reduced while in the refueling mode.
For example, if the maintenance-activities hrd been restricted such that two SFAS channele would not be lost by a sir le event (e.g.,
serving Channels 1 and 3 from separate sources), this even' would have been precluded.
Likewise, if maintenance activities had been planned or restricted such i
that a backup decay heat removal system would have been readily available, I
the consequences of the loss of the crerating decay heat removal loop would have been ameliorated.
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LO-HO-12101-002-04-C i
Regarding the two-out-of-four SFAS logie used at Davis-Basse, even under n rmal ':, n d i t i t n s, it appears that this type of logic is somewhat more surreptiole to spurious actions than other logic schemes (e.g., a ene-out-9f two taken-twice scheme).
This susceptibility is amplified,when
- wo srAS channels are served from one source. Consequently, when the source feeding SFAS Channels 1 and 3 was lost, all five levels of SFAS were actuated.
As stated previously, this particular event would have been I
precluded if SFAS Channels 1 and 3 were being served frnm separate and independent eeurces.
In a elinilar vein, this specific event would have been precluded by a one-out-of-two taken twice type of logic that requires l
the coincident actuation of or loss of powwr of an even numbered SFAS Channel and an odd numbered SFAS Channel.
Since each LWR can be expected to be in a refueling mode many times during its lifetime, licensees should evaluate the susceptibility of their plants to losing decay heat removal capability by the causes described in this Information Notice.
No specific action or response is requested at this time.
Licensees having questions regarding this matter should contact the director of the appropriate NRC Regional Office, Enclosures i
Davis-Besse Event of April 19, 1980 i
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LO-HO-12101-002 04-c SOER 85.004 (ABBREVIATED)
LOSS OR DEGRADATION OF RESIDUAL HEAT REMOVAL CAPAHILITY IN PWRS j
Events involving brief Icss or degradation of residual heat removal (RHR) capability have occurred frequently over the last eight years.
Though such events have not resulted in major consequences, probabilastic risk studien have identified extended loss of RHR capability as a significant contrioutor to the potential for core damage, Improvement in a few key areas could reduce the frequency and severity of these RHR events.
I Lpw Ptactor Vessel Lgy11 At Zion 1, the RCS was being drained in preparation for steam generator primary-to-secondary leak testing with the plant in cold shutdown.
The first thift had lowered the RCS level and had made the valve line-up for nitrogen purging, During shift turnover the off-going shift expressed concern that they may not have drained the RCS sufficiently.
The second shift continued the draining, and the RCS level dropped below the auction
)
line to the RHR pump.
The BHR pump was stopped when it was noticed that the motor amperage was fluctuating.
An investigation determinod that a valve in the line between the reactor head and pressuriter was open allowing purge gas to cause an observed level higher than the actual reactor vessel level.
0 The RCS temperature increased 37 F during the 4b minutes without RHR cooling.
The 10*F minimum subcooling margin was always maintained.
Some actions to prevent recurrence were to retrain the operators on proper valve line-up prt;edures, to prohibit reactor vessel draining while purging, and to require th:: all actions that could reduce RCS level be stopped until proper level is verified if any doubt exists aLwut the correct level.
Several other events have occurred in the industry that demonstrate the difficulties that plants have had in controlling a partially drained reactor vessel level within inches of the RHR suction line conntetion to the RCS, Reliable, sensitive, and accurate RCS level indication is required to maintain proper RCS water level control.
Automatic RHR Suetion Valve.Clogggg Maintenance personnel at V.c. Summer were performing a cold overpressuritation protection system surveillance test with the plant in cold shutdown.
The test simulated a high RCS pressure that closes an RHR t rain header _ suction isolation valve to protect the low pressure RHR piping from overpressurization.
During the test, the RHR pump had to be tripped by the operator due to the unexpected closure of the pump suction valve.
The surveillance test procedure did not indicate that the suction valve would actually close.
1-13
LO-HO-12101-002-04-c Ho equipment damage resulted from this event since the RHR pumps were equilged with a minimum flow recirculation line, and the operator took prompt action to secure the pump. The duration of the loss of RHR flow was minimal.
Corrective action was to revise the tGet procedure to require deenergitation of the suction valve in the operating.RHR train when the asucciated pressure transmitter is being tested.
Several other industry events involving loss of RHR capability resulting from automatic RHR suction valve closure have occurred.
At Ginna, a closed signal was
- sealed in' while the motor control center was out of service for inspection.
When power was restorel to the motor control centor, the valve closed and the closure went unnoticed by the operators.
Calvert Clif f s and Diablo Canyon 1 both reported event s resulting f rom the interlock that isolates the PHR f rom the RCS on htgh RCS pressure.
The event at Diablo Canyon 1 resulted in PHR pump seal damage and a bowed shaft.
The Calvert Cliffe event resulted in the RCS temperature inerracing from 130*F to 195 r in 40 minutes.
L25e of R atilna Pump W
At Beaver Valley 1, while performing a design change with the plant in a refueling outage, a wire on the terminal block for an emergency bus supply breaker was lifted prematurely prior to establishing an electrical clearance.
Lifting the wire caused a phase unbalance in the overcurrent circuitry which tripped the breaker and deenergitud the bus supplying the running RHR pumps.
The diesel restored power to the bus and the RHR pump was started in less than two minutes.
A breakdown in communications between shifts resulted in construction personnel being unaware of equipment status.
Air platino Due to Vortexina Durino Pumo shif t A control room operator at D.C. Cook started a second RHR pump in preparation _for removing the operating pump from service during a cold shutdown. The resulting flow was excessive for single RHR loop operations and caused -vortuxing resulting in cavitation and air binding of the RHR pumps.
Both pumps were vented and the RHR was returned to service within 2$ minutes.
S!GNIFICANCE-Analyses have identified the loss of RHR capability as a significant contributor to the potential for core damage.
In several events over the a
last several yeare, the temperature in an opened RCS has approached the
-boiling-point-during-the loss of RHP rapability.
The probability of uncovering the cote by the loss of inventory due to boiling is greatly increased upon the loss of RHR.
1 1 14
.I
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LO-HO-12101-002-04-C 4
In ad dit ion, the loss of RHR capability can lead to a release of airborne i
radicactivity, loss of cose shielding, equipment damage, degradation of RCs tenterature monitoring, schedute delays, and additional costs.
ANALYSIS / DISCUSSION i
The RHR system normally is used to cool the RCS from approximately 320 F to a desired temperature range below 200 F.
Although redundant RHR trains l
usually are available, events have occurred where decay hoat removal has ceased. The extent of any resulting RCS heatup is dependent upon the decay heat rate, available makeup, and duration of the loss of RHR.
Analyses have shown that.under adverse conditions (high power history, high decay heat rate, partially drained reactor vessel water level, no makeup and available vent path) it is possible to expose the core within 15 to 30 minutes due to boiling off the coolant.
Loss of Inventory Approximately one-third of the loss of RHR capability events were caused by low reactor vessel level (loss of inventory) during partially drained vessel operations that resulted in the loss of RHR pump suction.
controlling the vessel level in the required narrow range is a difficult evolution. There is a strong need for reliable reactor vessel water level information in the control room.
Frequently, only a temporary clear plastic tube standpipe is used to monitor the partially drained reactor vessel water level.
Problems that have occurred include root valves being left shut, open root valves being obstructed by debris, lack of monitcring by operations, lack of information in the control room, and constriction of the plastic level monitoring tube.
Loss of Runnino Pumo Hany of the loss.of RHR capability events involved air binding of t!.e RHR pumps.
The repriming of the pumps can be difficult in plants that have a portion of the RHR pump suction line at a higher elevation that the RCS line to which it is connected.
One particular plant that actually had such an event now utilizes vacuum priming to obtain and maintain the prime.
Venting and filling with the RHR system isolated from the RCS could also be used under certain circumstances if required.
- In many loss of RHR pump suction events the immediate response of the operator was to start a second pump.
This has resulted in the loss of suction to the second pump in those car a where there was a common cause, thus complicating the recovery.
In most events, sufficient time exists to investigate and correct the cause prior to starting a second pump.
Automatte Valve closure
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4 LO-HO-12101-002-04-C f
In su h events, not only is RHR flow to the RCS interrupted, but damage may c
occJr t? the PHR pumps unless there is either an interlock to automatically j
stcp the pun-po or immediate operator action to protect the system.
Because of tne potential for loss of RHR through inadvertent isolation of suction valves, sone case studies recommend disabling the suction valve closure interlocks during certain phases of RHR operation.
Disabling of suction valve closure interlocks requires careful analysis of low tenperature overpressurization and containment integrity concerns.
Additional relief capacity may be necessary.
Removing power from the i
suction valve motor operator is an undesirable method of avoiding inadvertent auction valve closure, nince such action comprises the ability to quickly isolate RHR suction from the RCS in the event of an RHR system LOCA.
Plants that have their own letdown connected to the RHR system should be eencorned with cold overpressuriwation protection during solid plant operations.
Inadvertent suction valve tlocure-in these conditions will result in a rapid pressuro increase due to continued charging without-eenpensating letdown.
Several preventive actions could be used to minimite the potential for a cold overpressurirution event.
These include testing of PORVs immediately before going solid and mairtaining the normal letdown path open such that the relief valve'lu this line could provide a redundant relief path.
I Lona-term Considerj u p,ng It i s possible that a long-term degradation or loss of RHR capability might occur that can complicate other emergency events.
In an Oconee 2 SOTR further plant cooldown, and consequently the primary and secondary s
- event, pressure equalization, was delayed nearly a day because an RHR suction valve failed to open.
Consideration should be given to alternate heat removal schemes such as I
feed-and-bleed operations and use of ARVs or steam dumps.
Steaming the i
generatore requires inventory makeup sources.
Alternate cooling flow paths should be identified (such as portions of CVCS or the SFp cooling system).
Continued RHR flow, even without the RHR HX, will provide mixing of the BCS and will extend the time to bulk boiling in the reactor vessel.
If the RCS is open and cannot be closed upon loss of RHR,- the containment should be evacuated prior to bulk boiling because of the release of radioactivity and eventual loss of core shielding.
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Lo-Ho-12101-002-04*C 1
rnelos.grqth Davis-BEssE EVENT OP.AEhlt 19.__1980 STATUS or DAVIS-BEsst 1 PR!oR To Loss or POWER 70 Busses E-2 AND r-2:
1.
Rufueling mode with RCS temperature at 900r and level slightly below vessel head flange.
Head detensioned with bolts in place.
Manway cover on top of otso removed.
tygon tubing attached to lower vents of I
RCS hot leg for RCS level indication.
Decay heat loop 2 in service for RCS cooling.
]
2.
All non-nuclear instrument (NN!) power and Static Voltage Regulator YAR supplied from 13.8 KV Bus B via HBBF2.
13.8 K Bus A energized but not connected.
RPS and SFAS Channele-1 and 3-being supplied from YAR.
- 3.
Equipment Out of service--
a.
Source Range Channel 2 - surveillance b.
Emergency Diesel Generator 1 - Maintenance c.
Decay Heat Loop 1 - Maintenance 4.-
Breakers for containment spray and HP! pumps racked out.
1-;'
_ _. _. _. _... _. - _.. _ _ _ _. _ _.. _ _ _ _ _. _ _. - _ _ _ _ _. - _. _ _.. ~ _. _ _ _
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LO-HO-12301-002-04-C i
l SEQUENCE OF EVENTS l
4
____Iltr EVENT CAUSE/COKMENTS t
f f
2:00 p.m.
Loss of power to Ground short on 13.8 KV Dreaker HDBF2 i
Buseos E-2 and F-2 which caused braaker to open. This (non-essential 480 interrupted power to busses E-2 and F-2 VAC) which were supplying all non-nuclear
[
instrument (NNI) power, Channels 1 and 3 of the Reactor Protection System (RPS)
I and the safety Features Actuation Signal (SFAS), the computer and much of the i
control room 1.ndicatore 2:00 p.m.
SFAS Level $
Two out of four logic tripped upon lose (recirculation of Dusses E-2 and F-2.
Actuation caused mode) actuation RCCS. pump suction valves from i
containment sump to open and ECOS pump
{
suction valves from Borated Water Storage Tank to close.
During valve travel times, gravity flow path existed j
from BWST to containment sump i
2 02 p.m.
Decay Hoat (low operator turned off only operating DH pressure safety pump to avoid spillage of RCS water to injection) flow containment via the tygon tubing for RCS l
secured by level indication and open So manway operator
?
2:33 p.m.
Partial _
Power to Bus E-2 and SFAS Channels 1 and
{
restoration of 3 restored along with one channel-of N!
2:44 p.m.
Attempt to Started DH pump 1-2 then stopped it when reestablish DH it was determined that air wac in flow suction line.
Pump secured to pr9 vent damage i
3:34 p.m.
Source Range Channel 2 I
energized 4:00 p.m.
. Restoration of Dusses restored sequentially to-offortu -
to Dusses (480 VAC).
trogressed to isolate ground fault 4:06 p.m.
F-2, F*21, F-22, and F=23-1-18 i
. _., _ _... _. _ _... s. _. _. _., _. _ _._.._. _.._ _ _ _._ _ _
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I l
LO-HO-12101-002-04-C I
.__ IJ M E EYINT CAUSE/cQPfigtifs 4:25 p.m.
DH flow restored DH pump 1-2 started after venting.
RCS 0
temperature at 170 F DH ficw bypasping j
cooler.
Incore TC's being taken and 0
maximum is 170 F 4:46 p.m.
Containment sump Precautionary measure to assure pump breakers containment sump water from BWST opened remained in containment.
Incore TC*s 0
range from 161 F to 164 F
$140 p.m.
Computer returned Incore TCs range from 158"F to 160 F-to service 6:24 p.m.
DH flow directed RCS cooldown established at less than 0
through cooler 2$ F per hour.
RCS temperature at 1$0 F.
Incore range from 151"r to 158 F 9:50 p.m.
Power completely RCS_ temperature at approximately 116 F 0
STATUS OF DAVIS-BESSE 1 AFTER RECOVFRY FROM LOSS OF POWER TO BUSSES E-2 AND F-2 1.
Refueling mode with RCS temperature at 115 F and level slightly below vessel flange.
Head detensioned with bolts in place.
Manway cover on top of OTSO removed. Tygon tubing attached to lower vents of RCS hot leg for RCS level indication.
Decay heat Loop 2 in service for RCS cooling.
2.
Bus E-2 being supplied from 13.8 KV Bus A via breaker HAAE2 and Bus F-2 being supplied from 13.8 KV_ Bus B via breaker HBBF2 r
3.
Decay heat-loop tilled, all tage clear. Maintenance work restricted so restoration of system will be lose than two hours.
4.
ECCS pump suction valves (DH-9A and DH-9B) from containment sump closed and breakers racked out.
This will prevent the suction of air into the decay heat loop.during a Level 5 actuation (recirculation
-mode) when there is no water in the sump.
5.
Equipment out of Service:
Emergency Diesel Generator 1 - maintenance 6.
Breakers for containment spray and HP1 pumps racked out.
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LO-HO-12101-002-04-0 t
CASE STUDY ON THE DIAB 1.0 CANYON 1.05S OF RHR XVENT DUIIING HID-LOOP OPFAATION (IEN 87.023)
INTR 0bOCTION 1.
This case rtudy material covers a loss of residual heat removal during midalcop operation and the phenomena influencing that behavior at P0&E's Diablo Canyon Unit 2.
4.
37 additional events have occurred that are attributed in inadequate PCS water level.
b.
Coro damage or a release to the environment could have occurred.
II.
SUMMARY
1.
P0&E's Diablo Canyon Unit 2 Pour loop Westinghouse 1119 MWe PWR - same as Vogtle a.
b.
Good initial operating history
?
2.
Reactor in MODE 5 a.
Seven days after shutdown for its first refueling outage b.
A loss of both RHR trains occurred for approximately 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> 3.
Complications affecting the loss of RHR cooling Pemoval of the containment equipment hatch (release path to a.
{
the environment) b.
RCS hot leg mid-loop level operation Steam generator manway removal in progress during the event c.
d.
The reactor coolant heated from 07 F to boiling 0
e.
Steam was vented from the RV head f.
Water spilled from the partially unsealed so manways I
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LO-HO-12101-002-04-c g.
Containment radiogas activity was observe to increase
!!!. OETAILED EVENT DESCRIPTION 1
Initial conditions a.
MODE $
b.
The containment building equipment hatch was removed c.
The personnel airlock was open d.
Containment purge was in progress Removal of steam generator manways was in progress a.
f.
Core exit thermocouples were decoupled g.
Local leak rate testing of containment penetration was in i
progress 1
h.
The RHR pump 2-1 was operating through both RHR heat exchangers, both trains were cross-connected 1.
The RCs was drained down to the mid-loop level f
-j.
RV level was being monitored by:
1)
A tygon tube manometer inside containment i
2)
Two electrical systema (a wide'and narrow range) 3)
Normal RVLIS.was out of service
?
k.
The RV was vented to the pressuriser 1.
The si pamps circuit' breakers were racked out m.'
RV level maintained by:
1) sending excess water to the RWST 2)- Makeup from the RWST IV.-EVENT INITIATION 1.
A plant engineer opened a valve to perform a local leak rate test creating a leak from the Pcs
e 4
LO-HO-12101-002-04-c Loss of RHR Cooling a.
THR pump began cavit ating b.
0[:erator shutdown the running pump Operator star ~ced and than shutdown the standby pump c.
1)
It also cavitated 1,
RHR cooling capability lost 1)
No method of monitoring incore temperaturen 2)
Validity of the temporary RV lovel indicat ion suspect ed a)
Operator dispatchod to check local RV t ygon t ut e indication e.
Operator attenpted to verify RCS integrity f.
Operators attempted to stop leak g
g.
NOUE declared 3.
Leak stepped after approximately 1.5 houro by Engineer Operators refilled the system from the RWST via an RHR pump a.
V.
FUNDAMENTAL CAUSES AND DISCUSSION TOPICS 1.
RV level indication system problems 2.
Improperly seated valve J.
Operator awareness of evoluticns in progress 4.
Containment integrity problems 5.
Mid-loop operation 6.
Instrumentation 7
Ccmmunications problems 8.
Event mioclassification VI. POST-EVENT TECHNICAL AND ADMINISTPATIVE INVESTIGATIVE ACTIONS TAFEN AT
LO-HO-12101-002-04-c PLANT VDGTLE 1.
Several procedore related deficiencies were identified and corrected 2.
Hardware changes b
v t
1-23
1 i
LO-HO-12101-002-04-C V111. Ai!ACHMENT ATTACHMENT A PLANT VOCTLE RESPONSE TO CENERIC LETTER 87-12 Some Post-event Technical and Administrative Investigative Actions Taken at Plant Vogtle The detailed event description performed after the Diablo Canyon loss of RHR cooling resulted in the NRC issuing Generic Letter 87-12 Lons of
~
Residual Hest Removal (RHR) While the Reactor Coolant System (RCS) is Partially Filled, and NUREG 1269, Loss of Residual Heat Removal System.
These reports identified factors that either contributed to or caused some aspect of the event.
Additionally, other problems were identified that had t5c potential to further complicate the event.
Generic Letter 87-12 requested answers to specific questions posed by the NRC on how a similar event occurring at Plant Vogtle is prevented from happening, or actions planned to prevent this event.
Some of these actions are provided below.
A.
Several procedure-related deficiencies were identified and corrected 1.
Procedures 12000 - Refueling Recovery, 12006 - Unit cooldown to Cold Shutdown, and 12007 - Refueling Entry have been revised to require at least two incore thermocouples to be maintained operable during periods of mid-loop operation.
If the RPV head is removed the disconnection of these thermocouples will be delayed until the last possible moment and restored at the first opportunity after the head is replaced.
2.
Guidelines for monitoring reactor vess'd level when draining or filling the RCS have been expanded.
More information concerning the parameters to be monitored is also given in Procedures 12000
- Refueling Recovery, 12006 - Unit Cooldown to Cold Shutdown, and 12007 - Refueling Entry.
a.
Continuous monitoring when changing levels when PZR level is
< 174.
b.
Periodic level che;ks are required between the control room indicators and the tygon tube every four hours.
c.
A continuous tygon tube watch is required if no control room 1-24
.f
t e-LO-HO-12101-002-04-C indicator is available.
d.
A continuous monitoring of the tygon tube level during mid-loop operation is required by 13005 - Reactor Coolant System Draining.
3.
RHR train operation guidance is given. One train in operation with a flow of 3000 gpm in Procedure 12000 - Refueling Recovery, 12006 - Unit Cooldown to Cold Shutdown, 12007 - Refueling Entry, 13011 - Residual Heat Removal system, and 13005 - Reactor Coolant system Draining.
4.
During the draining of the steam generator, U tube guidance is given for expected RPV level responses in Procedure 13005 -
Reacter Coolant system Draining.
S.
Minimum level of 180 feet is maintained whenever the RHR is in service per Procedure 12000 - Refueling Recovery, 12006 - Unit Cooldown to Cold Shutdown, 12007 - Refueling Entry, 13005 Reactor Coolant System Draining 6.
13005 - Reactor Coolant System Draining instructs the operator that only one drain path shall be used at a time and operators shall be aware of the path being used.
Log entries shall be made to keep personnel aware of drain paths.
7 13005 - Rosetor Coolaat system Draining instructs the operator that if draining via the RCDT, do not drain from the same loop (s) that are being monitored for level. Thus, filling or draining operations should not have an adverse affect on level indication.
B.
Hardware Changes 1.
Temporary reactor vessel level indicators are to be installed on the control board using $1 accumulator level instruments.
Both alarm functions and trending information will be available.
2.
An evaluation of the removal of interlocks associated,with the RHR loop suction valves is in progress.
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CEORGIA POWER
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POWER GENERATION DEPARTMENT V0GTLE ELECTRIC GENERATING Pl. ANT INSTRUCTIONAL UNIT 1
l TITLE:
RESPOND TO LOSS OF RESIDUAL HEAT NUHitER: LO-10-60315-001-02 i
REMOVAL
{
PROGRAM:
LICENSED OPERATOR TRAINING REVISION: 2 AUTHOR:
FITZWATER DATE:
8/9/89 APPROVED:
DATE:
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)
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REFERENCES:
I VEGP PROCEDURE 16019-C, REV 6.-LOSS OF RESIDUAL IfEAT REMOVAL i
MASTB 0:PY gwuswr
LO-1U-60315-001-02: Recpond to Loos of Residus! Haat Rtmoval PERFORMANCE OBJECTIVE Given indications of a loss or reduction of RRR capability. respond to lobs of RHR.
The LO must restore the faulted / failed RHR train (s) tb operablo status. It Pj a required RHR train cannot be restored, the LO must shut down the train.
I refer to Technical Specification for additional requirements, and initiate
-appropriate actions to restore the train. The to must comply with all applicable Technical Specifications. All communication and activities must be performed in accordance with current, approved procedures.
,1_N, FORMATION During a normal plant cooldown, the RER system transfers heat f rom the RCS to the Ccw system to reduce = the temperature of the reactor coolant to cold -
shutdown temperature. A loss of RHR capability could result in an uncontrolled RCS temperature rise and. over the long term, any of the following conditione:
- 1. RCS pressure increase
- 2. Thermal stresses
- 3. Reactor coolant boiling A rapid increase in RCS temperature is not expected if the reduction or loss of RER capability occurs early in the fuel cycle. However, if the reduction or loss occurs late in the fuel cycle, an immediate, uncontrolled RCS temperature increase-is probable because of the presence of a high rate of decay heat.
The rate of decay heat is not only dependent on time and fuel cycle but also on recent power history.
The highest rate of decay heat will be present durins a shutdown at the end of the fuel = cycle after=
extended operation at his power.
?
Loss of RNR also means that a major component of the ECCS will not be available to provide emergency core cooling in the event of a loss of cooling accident.
The most Itkely causes of a reduction or loss of residual heat removal are:
- 1. RHR pump trip
- 2. Failure of pressure transmitters in a RHR -train
- 3. RHR system breaks
- 4. Pump / flow problems due to
- a. Loos of/ inadequate level
- b. Vortexing i-1
l LO-It!-60315-001 -02 ; Resgend to loss of Residual llent Removal i
pump cavitation c.
An kHR pump t rip is most likely to occur during operation iti mode 5 kit h the RCS loops partially drained for maintenance, due to air / gas binding of t he RilR pump (s ).
!! this occuts, remot ely>opet at ed RHR piunp suct ion vent s (liv-10665. HV-10466) can be opened to release the air / gas.
Failure of pressure transmit ters in a RHR train may cause suctiori isolation valves to close, thus, causing a l o s t, of suction.
This in a low probability cause of reduction or loss of RHR.
RHR Syeem breaks are also unlikelyt the following components are most susce,otible to a break leading to a reduction or lor.s of RHR capability:
- 1. putap seals
- 2. valve racking glands
- 3. heat exchangers (tube leaks) 4 piping (failure)
The symptoms of reduction or loss of residual heat removal aret Unexplained decrease in RHR flow or discharge pressure Detected RHR system leakage An urwxpected rise in RCS temperature while RHR is in operation Any observed loss of RHR capability while RHR is in operation The following annunciators may be present on loss of RHRt RHR PMP OVERLOAD annunciator slarm CCW TRAIN A(B) RHR PMP SEAL LO FLOW annunciator alarm WSCW TRAIN A (B) RHR PHP MTR CLR LO FLOW annunciator alarm RHR MiP OVERLOAD will occur only with an RHR pump trip.
Symptoms 3 and 4 may also accompey a pump trip.
plan'. Vogtle Procedure 18019-C. " Loss of Residual Heat Removal." is entered when the symptoms of reduction or loss of RHR capability are present.
The goal of this procedure le to restore the faulted / failed RHR train (s) to operable status as quickly as possible.
The RHR system is one of the few plant systema thtt has Technical Speelfication requirements placed on it for all modes of operation. The response to loss or reduction of RHR o pability in Procedure 19019-C is highly dependent upon the mode of opriition and the limiting conditions of the applicable Technical specificattons-MODE T.S.
1-2
l LO-10-60315-001-02: Respond to Loss of Rocidual Hast Rsmoval 4
3.4.1.3 3
3.4.1.4.1 (reactor coolant loop 4 filled) 5 3.4.1.4.2 (reactor coolant loops not filled) 6 3.9.8.1 (water level above the top of the reactor pressure vessel flange greater than or equal to 23 feet) 6 3.9.8.2 (water level above the top of the reactor pressure vessel flange less than 23 feet)
Technical Specifiestions should be consulted for complete operating requirements.
Technical Specification requirements will vary depending on the mode of operation. Generally, a minimum of one RHR train must be operable.
Certain modes or plant conditions require two RHR trains to be operable. Every effort should be made in Plant Vogt10 Procedure 18019-C to restore faulted / failed RHR trains to operable status as quickly so possible.
There has been many Loss of itKR events in the industry over the past several years. These events are significant because core voiding and overheating can occur rather quickly.
Refer to Figure 2 " Time to Roll". Figure 3. " Time for Core Uncovery",
and Figure 4 "Heatup Rate" curves located in AOP 18019 for a better understanding of the urgency for recovery from loss of RHR events.
RESPOND TO LOSS OF RHR This Instructional Unit is divided into two sections:
Section A. Loss cf RHR in Modes 4 or 5. and Section B. Loss of RHR in Mode 6.
SECTION A - LOSS OF RER IN HODES 4 OR 5 CAUTION During midloop operatieskenth ut dans installed and inadequate RCS venting, a loss of RHR cooling wifi Jeesult in saturated RCS conditions within 10 minutes, subsequently resulting in core uncovery and requiring containment closure initiation.
Monitor / Maintain core cooling This must be done continuously until you exit the procedure. To do this-you will monitor the Core Exit TC less than 200 F and RRR cooling is restored.
If Core Exit TC is greater than 20d'F or RHR cooling is not restored, then initiate Procedure 910'>1-C.
" Emergency cLiswification and Implementing Instructions".
You must also evacuate non-- sential personnel and initiate 1-3^
i
i e
LO-Ill-60315-001-02 Respond to loss of Residus! Hsat R:moval containment isolation. Additionally verify the RCS is in tact and initiate charging flow to provide core cooling.
11 core exist ta perature is less than 20@ F. then continue to monitor the TCs.
Should you lose the TC indication while in midloop operation then rat a r,JI level to the top of the hot leg (188 feet 3 inches) and monitor temperature using RCS wide range indication Tgog.
If no RHR Train is operating. then suspend any RCS boron reduction and if the steam generators are available maintain RCS temperature below 35C F by use of steam dump or atmospheric relief valves (one SG IcVel should be filled and maintained in the narrow range).
Concurrently with these actions you must monitor operating RHR pumps for cavitation. Cavitation would be indicated by the discharge flow being lower than normal for the RCS pressure or by an unstabic discharge pressure.
These indications are located on the QMCB.
pt-0614 Train A - Pressure PI-0615 Train B - Pressure FI-0618A Train A - Flow FI-0619 Train B - Flow c
If at any time during performance of this procedure, cavitation occurs.
then stop the running RHR pump, realign misaligned valves. Vent the affected pump using HV-10465 Trn A or HV-10466 Trn B valves using the QNCB located hand switches.
The valves are physically located in sealed rooms and cannot be observed locally. If HV-10465 is not closed when venting is complete. the contaminated water from the vent will overflow from the vent room down into the vestibule R-131 level C.
Failure to close HV-10466 will result in contaminated water on the floor of the room containing the vent
-and the actuation of the room sump level switch light on the QPCP.
If Core Exit TCs are stable or lowing, then return to the appropriate t)0P.
If the TCa are rising, verify the af fected RHR and CCW flow norhal.
This may occur due to inadequate time from when the probles first occurred and the time that you are at this step.
If the TCs do not stabilite or begin to lower with normal flows. then an inadequate RCS inventory may exist.
Monitor / Reestablish Adequate RCS inventory To have adequate inventory you should have greater than 725 indication on the RVLIS FR or local or remote level indication greater than 187 ft. elev.
If this does not exist, then you will take actions to restore the level.
In doing so, you will disatch operators as necessary to locate and isolate any leaks, adjust charging flow to re t m n level or, if no charging is available. gravity fill the RCS from the NL 1-4 j
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1.0 -10-60315-001-02
Respond to Loso of Residual Heat Removal If these actions do not result in level accovery, a leak may exist.
locate the leak you will isolate the RHR pump suction valves one at a time To to (Solate the leak.
Additionally. you will be directed to AOP 18004-C i
"RCS teakage" to rentore the level.
After level is t' stored, then restore RHR.
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Restore RHR To restore RNR you will initiate procedure B011 "RHR System". Also see l
1,0-10-12101-002. "Startup RHR".
Once RHR is restored. check RCS lutact and determine is alternate RCS cooling is required.
8 Determine / Establish Alternate RCS Coolina The RCS temperature is the primary indicator of whether alternate is required.
cooling If RCS temperature is greater then 20d P.
required.
then alternate RCS cooling may be If the SCs are available, then use the steam dumps or ARVs for cooling, and maintain at least two SG levels greater than 17%.
Also start all CRDM fans which will also remove a significant amount of heat.
Start the containment cooling fans and maintain RCS pressure at 365 psig with PER heaters 'and spray (if bubble in PZR) or. charging and letdown solid).
(if Start an RCP is possible to provide forced flow for better heat transfer (rom the core to the stesa generator,.
Initiate Repairs This may take considerable time depending on the cause of the event.
i Additionally-Technical Specifications must now be consulted.
You will also return the system to. normal operation.
This will require consultation with the TSC for recovery actions.
SECTION B - LOSS OF RRR IN HODE 6 Check RCS level above RV flange w'th cavity filled.
If not, so to Section A Suspend all operations involving a reduction of RCS boron concentration.
Verify the loop suction valves are open.
l 1-5
._.2.
LO-IU-60315-001-02) R2spond to lot
" Rooidual Heat Removal Verify the following valves are open:
HV-8701A Train A
+
HV-8701B Train A
- IIV-8702A Train B HV-8702B Trair. B If any of these valves are closed, stop the affected RIDi pump (s) and open the affected train auction valves.
Verify reactor cavity level is greater than 23 feet.
Use lotil Tyson tubing or remote indication to verify reactor level.
If less than 23 feet, restore cavity level or fill the vessel to 98 percent RVLIS FR.
Suspend all fuc1 movement.
Determine if any leats exist.
Check the Auxiliar:' Building and Containment Leak Detection Syctem to determina if any leaks exist.
Dispatch an PE0 r.o locate and isolate any leaks.
If a break occurs on the RMR Syntem, stop the RER pumps.
Att~mpt to place either train of RHR i,n operation Initiate 13011, "RIIR System.* to place one train of RHR in operation.
Initiate repairn and applicable Technical Specificat4_on requirements.
When the RER System is in operation, return to the approprir.te UOP.
Start all containment cooling fans.
Place both trains of the Spent Fuel Fool Cooling and Purification System in se rvice.
Initiate 13719, " Spent Puel Pool Cooling and Purification System." to place both RHR trains in service.
Place FSB_HVAC in service.
Initiate 13T10. "FHE HVAC System," to place FHU HVAC in service.
Ensure dCS temperatcre is greater than 185, degrees F.
It RCS temperatura is less than 185 degrees F. return to the placing either trein of RRR in service step (Procedure step B6).
Establish an RCS feed path from the RWST.
At leest one CCP must be running.
Verify valve aligaments for the operacing pumps.
Manually start a CCP and align the valves as necessary.
If feed free the RWST cannot be established, return to the placing etther train of RHR in service step (Procedure step B6).
1-6 ta
~. _ _ _ _ - _ _ _ _. _.
L0-1U-60315-001-021 Respond to Lon of Residual H2at Renoval
=,
Establish an RCS bleed path.
Establish an RCS bleed p9th from at least one open loop.
Dispatch a PE0 to open the RCDT pump to RWST isolation valves 1901-114-041 and 1204-U4-002 and the RCS drain loop to RCDT pump (solation valve.1901-U4-242, Operate the RCDT pumps as necessary to maintain the reactor cavity level at greater than ?3 feet and the temg?rature at less than 185 degrees F.
Check RPR System status.
If the RHR System cannot be placed in ep. ration, rett n,, the placing either train of RHR in service step (procedura step BL)-
Terminate bleed and feed, Return to the UOP in effect, I
s 1-7 i
LO-IU-60315-001-02; Respond to Loss of Residual Hast Removal PERFORMANCE CUIDE The following steps are required to respond to loss of RHR:
-SECTION A - LOSS OF RHR IN MODES 4 OR 5 1.
Monitor /malatain core cooling 2.
Monitor / reestablish adequate RCS inventory C
3.
Restore RHR 4.
Determine / establish alternate RCS cooling 5.
Initiate repairs SECTION C ASS OF RHR IN HODE 6 1.
Check RCS level above RV flange with cavity filled.
If not, go to Section A.
2.
Suspend all operations involving' a reduction of RCS -boron
-concentration.
3.
Verify the' loop suction valves are open.
4.
Verify-reactor cavity level is greater than 23 feet.
5.
Suspend all fuel movement.
6.
Deter 1 sine if any leaks exist.
7.
Attempt to place either train of:RHR in operation.
8.
_ Initiate repairs and applicable Technical Specification requirements.
9.
Start all containment cooling fans.
10.
Place both trains of the Spent Fuel pool Cooling and Purification System in service.
11.
Place FHB HVAC in service.
12.
Ensure RCS temperature is greater than 185 degrees F.
13.
Establish an RCS feed path from the RWST.
14.
Establish _an RCS bleed path.
15.
Check RHR Systes status.
16.
Terminate bleed and feed.
- 17.. Return to the UDP in effect.
k 1-8 l
1
^
LC-IU-60315-001-021 Respond to Loss of Residuci Host Rimoval SELF-TEST Before proceeding to the Task Practice, answer the following questions as completely as possible.
1.
List the symptoms of loss of RHR.
2.
List three annunciators which may be present on loss of RHR.
3.
If a feed path is to be established during Mode 4. the ECOS pumps will provide feed.
State the feed path during Mode 6.
4.
Briefly state the implicationra of loss of RFIR ently in core life versas late in core life.
5.
State the most likely causes of a reduction or loss of residual heat removal.
1-9
. __.. -.., - _ _.. _ _ _. _. _. ~, - _ _ _ _ -.. _ _.. _... _ _ _ _. _ _. _ _ _. _. -. _. _ -.. _
o i
-LO-TU-60315-001-02: Respond to Loss'of Residual. Heat Removal I
w ANNERS 1.
Unexpla(ned damage in RHR flow or discharge pressure
- -Detected RilR system leakage' An unexpected rin in RCS temperature while RkR is'in operation Any observed loss of RHR capability while RHR is in operation i
l 2.
RRR PMP OVERLOAD annunciator alarm
~ CCW TRAIN A(B) RNR PMP SEAL LO. FLOW annunciator-alarm NSCW TRAlti A (R) RHR PMP P'R CLR LO FLOW annunciator alarm 3,
The RCDT pumps provide - feed flow, as required, through the RWST via-the CCPs back to the reactor vessel.
- 4.
A rapid increase in RCS temperature is not expected if the reduction or loss of RHR. capability occurs early in fe l cycle. However, if the reduction or loss occure late in fuel g ele,. an - immediate, uncontrolled RCS temperature increase is probable because of the presence of a high rate of decay heat. The rate of decay heat -is_ not -
only dependent on time 'and fuel cycle, but also ou~ recent-power history. The highest rate of decay heat will be present during a shutdown _at the end of fuel cycle after extended operation at hiah power.
5.
RER pump trip Failure of pressure transmitters in a RER train
-- RHR system breaks
_ Pump / flow problems due:to loss of/ inadequate-level, vortexing or pump cavitation:
Y h'S
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W 31 1-10
LO-!U-60315-001-02; Ra: pond to Losc of Residusi Hect Rtmoval TASI PRACTICE 1.
Review Procedure 28019-C, " Loss of RHR."
Be sure that you understand all precautions, limitations, and steps associated with responding to.
loss of RifR.
2.
Take this instructional unit and Procedure 18019-C " Loss of RRR" to the control room or simulator.
De sure that you can locate all instrumentation associated with responding to loss of RHR.
3.
In the control room or simulator, simulate responding to loss of RRR.
If possible, have a fellow trainee evaluate your performance using Procedure 18019-C, " Loss of RHR" and this instructional unit.
N'%
t E
l-11
4 LO-IU-60315-001-02; Respond to Loss of Residual Heat Removal FEEDBACK ON TASK PRACTICE 1.
If you have any questions about the precautions, limitations, or steps in Procedure 18019-C. " Loss of RHR". ask your instructor.
2.
You should have been able to locate all instrustentation associated with responding to loss of RHR.
If you had any difficulty, ask your instructor for help.
3.
You should have simulated the steps necessary to respond to loss of RHR.
If you had any difficulty, re-read the pertinent <ections of this instructional unit and the procedure. Resolve any questions with your instructor.
~
1-12
~_
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GEORGIA POWER POWER GENERATION DEPARTMENT V0GTLE. ELECTRIC GENERAT]NG PLANT TRAINING LESSON PLAN 4
TITLE:
LOSS OF RESIDUAL HEAT REMOVAL Nt'MBER :
LO-LP-60315-04-C PROGRAM:
LICENSED-OPER' TOR TRAINING REVISION:
4 AUTHOR:
L. FITZWATER DATE:
8/9/89 APPROVED:
)
DATE:
sk k d'~ M N
O' INSTRUCTOR GUIDELINES:
7 I. Lesson Format:
A.
Lecture With Visual Aids II.
Materials:
-A.
. overhead Projector
- 8.. Transparencies C.
White Board with Markers III.
Evaluation:
A.
Written or Oral exam in conjunction with other Lesson Plans IV.
Remarks A.
Performance-based instructioual units (IUs) are attached to the lesson plan as student hatulout s.
After the lecture on Loss of Residual Heat Removal, the student should be given adequate self-study time for the It's..The instructor should direct self-study activities and be aviiiihte to answer questions that may.
arise concerning the IU ni'ri iala After self-study. the student will perform, simulate. oh-orm, or discuss (as identified on the cluster signof f criteria i
.br task covered in the instructionni unit in the presence of rin
=iinitor L
- sc= were N ASTER COPY
~
tA,-Le-t>0313-04-C 1,
PliRPOSE STAILMENT!
FOLLOWINO COMPLETION OF THIS LESSON.
THE STUDENT WILL POSSESS THOSE KNOWLEDGES SYSTEMATICALLY IDENTIFIED FOR THE PERFORMANCE OF LOSS TASKS
)
4 II.
LIST OF OflJECTIVES:
1 1
1.
Describe factors that can lead to a loss of RHR.
2.
State the possible consequences of a sustained loss of RHR.
___---~~
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~
. LO-L P-6 0 31.A ~ u,- t.
REFERENCES:
1.
Plant Vogtle Procedures
-18019. " Loss of Residual Heat Removal" 2.
Technical Specifications: None 3.
Vogtle Training Text: None 4
Plant Manual: None S.
Design Manual None-6.
P& ids, Logics.'and other Drawings: None 7,
Vendor Manuals and other
References:
'None 8.
FSAR: None 9.
Commitments and other Requirements:
IEN-87.023 Diablo Canyon 1.oss of RHR IEN-86.101 Loss of RHR due to Loss of Fluid Levels in RCS OMR 324 CCW Inventory Losses Resulting in Loss of RHR GL-88.017 NRC Concerns and Actions on Loss of RRR SOER 88.003 Losses of RHR with reduced water level at PWRs 10.
Transparencies LO-TP-60315-001. Objectives 11.
Instructional Units LO-IU-60315-001 Respond to Loss of RHR 12.
Handouts LO-HO-60315-001 Industry Event Sumraries (SOER 88.003) e
..,,,a r
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,-v.
,n
Lu-Le-mo a 3-uw-c Illi LESSON OUTLINE:
NOTES l~
4 1.
1NTR3 DUCT 10N A.
The loss of Residual Heat Removal Procedure AGP-18019-C is to identify and correct a lows of RHR capability B.
Procedure covers Loss of RHR in:
1.
Modes 4 and 3 2.
Mode 6 (head removed)
C.
Present Lesson Objectives LO-TP-60313-001 II.
PRESENTATION A.
Symptoms:
1.
Unexplained decrease in RHR flow or discharge AOP-18019 pressure 2.
Detected RHR system excessive leakage while AOP-18019 RHR in operation 3.
Any unexplained raise in RCS temperature while AOP-IR019 RHR in operation 4.
Any observed loss of RHR capability while RER AOP-18019 in cperation 5.
GL-88.014 lists the following conditions in
-which entry to Loss of RHR Procedures may be required Accidental loss of-a system that is oper-a.
ating to cool the RCS b.
Unsuccessful attempt to start a-system when the system was to be used for RCS cooling and the RCS.was not being actively cooled by another DHR system c,
Uncontrolled and significant loss of RCS inventory d.
Uncontrolled and significant break in the RCS coolant boundary Any valid symptom of loss of control-of the e.
state-of the RCS. such as uncontrolled temp-ereture increase, uncontrolled pressurizn-tion or the attainment of values of these 4
l
LO-LF-69315-04-G 111.
LESSON OUTLINE:
NOTES-i parameters shich are sufficiently high that action is required that_is not contained within normal procedures f.
Significant core damage expected g.
Any valid symptom of significant core damage observed B.
Factors that can lead to a loss of RHR 1.
Improper valve lineups a.
Many projects / evolutions occurring simultaneously b.
Loss of control c.
Personnel in plant not keeping control or informed of changing conditions d.
Lack of coordination 2.
System leakage a.
Improperly seated valves b.
Improper valve lineups c.
Operatar not familiar with valve position verification methods for reach rod operated valves d.
Loss of control of evolutions e.
Inadequate tagging orders f.
Normal system leakage not considered 3.
Deficient procedures a.
Infrequently or first t ime used procedures
- untested b.
Abnormal condit ions raquiring special procedures or tempotary procedures 4.
Opening RCS or related >vstems a.
SG manway removal b.
Reactor vessel beoi il A
LO-1.p-60313-04-c III.
LESSON OUTLINE:
NOTES Reactor coolaat pump seal replacement c.
d.
Installation of local reactor vessel level standpipe e.
Venting CRDM's L
Starting idle RifR loop a,
Idle loop partially drains while shutdown due to gas coming out of solution and accumulating in unvented high points b.
Gas pocket shifts to pump suction - pump cavitation results c.
Gas pocket shifts to reactor vessel with subsequent decrease in level d.
Gas pockets shift to atmosphere through open reactor vessel or other system components, with a subsequent decrease reactor vessel level to fill the void left by the gases 6.
Poor communications a.
Inadequate / incomplete shift turn over b.
Information on phone system not cicar no repeat backs, misunderstood / wrong interruptions c.
Distracting activities in control room and/or other critical areas d.
Inadequate / improper documentation of problems or evolution tiuring shift 7.
Changing plant conditions s
a.
Pressure reduction
- 1) Gases coming out o solution r
b.
Lowering reactor vemsol level for reactor head removal /othei miutenance c.
Opening reactor < n i
r, fueling transfer gate valve d.
Surveillance is in
.~..
W
+ -
1.0-LP-60313 c
~
- 111, LESSON 0llTLINEi-NOTES
' 4 e.
Pressurization of RCS
- 1) Caused by steam formation-in unvented reactor vessel 4
- 2) Inappropriate use of SG nozzle dam can lead to cote' voiding within 1.5-20 min-utes following loss of RRR
.3)
Cold leg openis.2 can allow water to be ejected from vessel following loss of FJiR until suf ficient water is lost that
- 4) steam is relieved by clearing the crossover pipes 3)
Pressure difference within RCS may pre-vettt water f rora reaching the RV -
- 6) Rapid RCS pressuritation may prevent gravity feed from tanks anticipated to be available
- 7) Rapid pressurization may-cause instru-
-ments to malfunction or provide mis-leading indications 8)~ Rapid pressuri;:ation may-caisse the RCS to tespond in unanticipated ways
-unanticipated RCS responses
- 10) Large RCS pressure boundary openings-(SG manway. HCP sea 1s. pressurizer man-ways)- may lead to: instrument 'malfunc-
- tion or unant;icipated RCS responses
- 11) ~SG sceandarv side inventory and opening' may-influence Rcs behavior.
f, Tortexing.
I --
1). Small amount
- f. air.Into RHR pump suc+
tion may I.cl- 'i subt le changes that occur over 4 4 ie of minute to an hour-f2 or more
- 2) Large amoun><
, i it may cause immedi-ate. lost of i.
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~ i Lo-LP-60313-04-c-III..
LESSON 00illNE14 NOTE 5-l
- 3) Vortexing may occur at levels higher than anticipated
- 4) Vortexing may not be reflected by pump current and-flow rate instruments until it is sufficiently severe to cause a lose of RRR
- 5) Vortexing may cause RCS level indien-tion errors g.
SG Cube draining
- 1) Draining sc U-tube is frequently done by dratning t he RCS to the point whet e vortexing could occur h.
RCS level itifferences
- 1) Critical level parameter la in hot leg whete RHR takes suction
- 2) Level inst ruments connected at other points
- 3) 1.evel differences exist between level indicator and hot leg RHR connection point (may be several' inches difference) l '.
RHR system effects
- 1) Shif ting f rom one train to another may cause level changes due to differences in actual site etc.
i
- 2) Starttng one size while another is run-ning een increawe total flow thus in-crease vortesing'~
- 3) Operator t oponse t o loss of one RHR pump by stu ting the second pump may result in a he lo w ol the occond pump.
also if st.urml wIthout correcting the cause of the low of the first pump
- 4) Stopping nt 3rofing RHR may cause RCS-level chanr km to partially filled system or os imhrtinn into the RHR System J.
Instrumentation J
s A
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Lu-LP-60315-04-C-J III.
LESSDN OUTLINkt NOTES 6
e
- 1) May be in error by half a foot or more without detection of inaccuracies
- 2) Flow dynamics, entrapped air, and pres-surization may affect level indication, individually or 'all indicators simultaneously
- 3) Many normal instruments disconnected during RV head removal etc.
- 4) Remaining instruments may be inadequate 8.
Loss of CCW Begin OMR 124 a.
Loss could be from improper clearance on one RHR heat exchanger that comprises the operable heat exchanger b.
Loss of CCW to heat exchanger prevents heat removal from RCS c.
Care should be taken when tagging one heat exchanger so as not to compromise remaining heat-exchanger End OMR 324 C.
Consequences of Loss of RRR 1.
Core damage a.
Loss of core cooling 2.
Radioactive releases to the environment a.
Open RCS with containment-open'to atmosphere J.
RCS overpressurization due to heat up-from decay heat 4.
-Contamination of personnel 5.
Activation of emergency plan D.
Loss of RHR in Modes 4 or 3 1.
Use AOP-18019 to discuss this event a.
Ensure students pay rirticular attention
'to:
1)' Caution statement at beginning of 9
LO-LP-60315-0*-C 111.
LESSON OUTLINLi NOTES
+
Section A
+
- 2) The statement that steps Al thru A3 are to be performed continuously until exit from this procedure
- 3) Note before Step Al Discuss'each of these figures with the class
- 4) Note statement before step A3
- 5) Indications of RHR pump cavitation Quiz selected student to ensure the class has an unde rs tatuling
- 6) Note statement before step A5
- 7) Note statement before step A6
- 8) Note statement before step A7 b.
The procedure establishes the priorities ast
- 1) Monitor / Maintaining core cooling
- 2) Monitor / Reestablishing adequate RCS inventory
- 3) Restote RHR
- 4) Determine / establish alternate RCS cooling
- 5) Fix problems
- 6) Return to normal ops E.
Loss of RHR.- Modes 6 1.
Use AOP 18019 to discuss this event 1
i L This procedure provides the following ptiorities i)- RCS level above Rx vessel flange
- 2) Suspend boron reduction l
- 3) Verify RHR pump tIow path i
- 4) Level greates t 'i nt 21 feel ausve Rx e
.J
LO-L -60313-04-c III.
LESSON OUTLINEt NOTES i
vessel flange
- 5) No lors of inventory has occurred (It it has must return to Section A of 18019)
- 6) Repair problems
- 7) Be prepared to secure-the RHR train that may tr subsequently lost
- 8) Restore RNR
- 9) Initiate alternate cooling F.
Industry Events IEN 87.023 IEN 87,101 1.
Diablo Canyon GL 88.017 Event included in a.
Loss of both RHR trains while in Mode 5 1.0-HO-60315-001 SOER 88.003
.b.
idIR out for 1 - 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> c.
Containment equipment hatch was removed d.
S/G manway renioval was in progress with RCS inv. at mid-loop level of Th RCS temp increased from 87 F to boiling e.
(212 F) f.
Steam was vented from Rx vessel head g.
Water-spilled from partially removed manway
-to containment floor h.
Containment radiogas level begins to increase 1.
RHR restored 4
j.
Potential problems
- 1) Core damage due to boiling out RCS inv.
- 2) Rad release to "vir.
- III,
SUMMARY
A.
_ Review Objectives i
- s LO-LP-60315-04-C r r
- III,. LESSON OUTLINE:
NOTES 1.
DESCRIBE FACTORS THAT CAN LEAD TO A LOSS OF RHR a.
Improper _ valve lineup
-Refer._to LP Sectini II,B for examples b.
System leakage and discussion c.
Deficient procedures d.
Opening RCS or related systems e.
Starting Idle RHR loop
- f.. Poor communication g.
Changing plant conditions 2.
STATE THE-POSSIBLE CONSEQUENCES OF A SUSTAINED LOSS OF RHR a.
Core damage b.
Radioactive release to environment c.
RCS overpressurization d.
Personnel contamination e.
Activation of the emergency plan 1
3 -*
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LESSON OBJECTIVES REPLACE THIS PAGE WITH THE MOST RECENT COPY OF THE LESSON OBJECTIVES N
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CLOSE CLOSE BREAKER BREAKER RHR PUMP CONTROLS LO-TP-12101-007
5 BHP-3 400 2 - 30' M.
E 300" 20 a NPSHR 2000 - 10 500 2
90 450 EFFICIENCY 80 400 350 70 o
O
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$ 250 - e 50 S
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150
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500 1000 1500 2000 2500 3000 3500 4000 4500 5000-GAUMIN
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RHR PUMP CURVE (TYPICAL)
LO-TP-12101-008
= _.
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i i : I I LR 990 LR 764 LR.104 RW5T LEVEL CONT SUMP B.A.YK.1 LEVEL LEVEL I
H5 8812A H54W11 A H5 87018 HS-88128 H5-8811P H5,87028 RWST TO SUMP TO HOT LEG RWST TO CNMT LEG SUCT RNR $UCT RHR SUCT VLV A RHR SUCT SUMPTO VLV A 1TR.A PMP1 2 TR. 8 RHRPMP 2 TR. 8 HS 8701A HS 8702A HOT LEG LEG $UCT
$UCT VLV VLV H5 610 H5 620 H5411 H5 621 RHRPMP1 RHRPMP1 RHRPMP2 RHRPMP2 MINIFLOW TR.A MINIFL OW TR 8 150L VLV A 150L VLV 8 H5 8804A H5 8809A H5-8716A H5 88048 HS 88098 H5-87168 RHR TO 515 RHRPMP1 RHR PMP 1 RHR TO HL RHRPMP2 RHRPMP2 PMP VLV A TO CL 150L HL 150L 150L VI.V 8 TO CL 150L HL 150L VLV A VLV A VLV 8 VLv 8 i
H5 8840 H510465 H5 40082!
H5 40083/
H510466 RHR TO SIS RHR UNE RWSTA 51 R W 5TB SI '
RHR UNE PMP VLV 8 VENT TR. A RESET TR. A RESET TR. 8 VENT TR. 8 CONTROL BOARD SWITCHES LO TP-12101-009 l
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RHR HEAT EXCHANGER LO TP 12101011 40 De lat ti (
MCB (di SPRING RETURN TO NEUTRAL FROM OPEN NEUTRAL CtOSE BOTH SIDES 1(PT Mj RCS HIGH PRE 55U'F.(*)
RCS HIGH DRE55UREN RECIRCULATION LINE ISOLATION VALVE LOSED RHR PUMP /RWST ISOLATION VALVE CLOSED SUMP LINE ISOLATION VALVE CLOSED o,
, o u o o
o EINEMMRE M M M
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S i
l p
y OPEN VALVE CLOSED VALVE NOTE: lor-tC FOR VALVES IN EACH FLUID SYSTEM TRAIN 15 IDENTICAL
- a. Automat lC Close setpoint,
- b. Prevent open setpoint.
- c. PT-Pressure transmitter.
PT 408 in MOV 8701 B interlock (forinner valve).
l PT 418 in MOV 8702 A interlock (for outer valve).
PT 428 in MOV 8702 B interlock (forinner valve).
PT 438 in MOV 8701 A interlock (for outer valve),
- d. MCB Main control board (local panel not shown),
8701 A/87018 AND 5702A/8702B CONTROLS t
4 0 s. u n,,,
L
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l'0 OPEN LOOP SUCTION VALVES l
i 1.
RVLIS pressure < 365 psig.
2.
Train related RWST suction valve (8812) closed.
3.
Train related Sump suctron valve (8811) closed.
4.
Train related ECCS recirculation valve (8804) closed.
1 i
RHR LOOP-SUCTION-VALVES i
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Train related Sump suction valve (8811) closed.
i 2.
Train related ECCS recirculation valve (8804) closed.
OPEN RWST SUCTION VALVES LO TP 12101015A to o. um
N i
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Q Ost'
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8812A/88128 LO-TP-12101-0158
FIRST SECOND LOOP LOOP SUCT SUG NT AUTO SUCT GOSED GO5ED SIAS 04
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BEFORE AFTER SAFETYGRADE SAFETYGRADE COLD SHUTDOWN COLD SHUTDOWN REQUIREMENTS REQUIREMENTS j
G e,
87018 8701A 08 Qp M
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87018 8701A
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LO-TP-12101-021
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TOP OF PZR EL 247'( + 864 *)
6O i
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FRESS.
TOP OF SG TUBES EL 231*( + 672")
l SURGE UNE TOP OF n
PZR HEATERS l
EL 207'( + 384~)
,j g
EL 190'( + 180~)
80TTOM OF PZR l l *'j EL 1%'( + 252~) %
W EL 194'i + 223~)
g t
i SG TusE SHEET l
l/
c EL 187'( + 144~)
e SURGE UNE-CENTERUNE COLDlEG
~
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CL OFINTERMEDIATE LEG EL 177'( + 24~1
,k EL 173'(0~
J INSTR)
EL.172'(FLR. -36~)
f
' / / / / / / / / / // // / / / / ' / / / / //
/ ' '
80TTOM OFINTERMEDIATE LEG VOGTLE UNITil. STAND PIPE ELEVATIONS LO-TP-12101-024
^
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@m NUCLEAR
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TEMPORARY l00 PRE 550RE PRESSURIZER i
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TUBE l
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l.:d$h*.lE W
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1 02
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TYGON TUBE LEVEL INDICATOR LO TP 12101025
... '"S"ti
WW a
Th, WIDE E
f RAN8E E
O
[i i
l t
LI ~
9 g 57g LOWLEVEL NARROW ALARM b
RANoE
\\
p
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RYLIS J
i LEVEL E
TRANSMITTERS q
RVLIS l
k LEVEL
\\
L TRANSMITTERS i
) $ =,E 3g
_i Leap 4 WIDE RANGE t,,,
PRESSURE Het tag r
TRANSMITTERS y
WIDE RANGE J
PRESSURE 1
TRANSMITTERS l
L j
\\
i TEMPORARY LEVEL INDICATGRS LO-TP-12101-026 u,..
VESSEL FLANGE----- -4 94'
-- - 100 %
-- 90 %
-193' s
. 80%
-192'
-- 70 %
' 191'
-- 60 %
l RCP SEALS
= -
-190'
- 50%
-189' TOP OF HOT LEG--- --- 188' 3" -
- =30%
--100%
~90%
- = - -
-188' 00%
=70%
- 20%
-60%
l 187' 40 %
(AL806-03)
- 10%
-30%
-20%
-10%
' ' 186'
-0%
-- - -0 %
TYGON HOSE L I 957 L l 950 Ll 950 RCS Loop 1 Hot Leg Narrow Range Level LI 957 RCS 4 Hot Leg Wide P.:nge Level NOTE: This Operator Aid is to be used oni when in mid loop configuration an temporary level instrumentation is installed.
RCS LEVELS V5 INDICATORS MID LOOP LEVEL INSTRUMENTATION UNIT 1 LO TP.12101027 m-5:
~
13.0 12.0 11.0
\\
c 10.0
.E h
~
b
.0 9
\\
8.0-7.0 h
6.0 5.0 m
4.0
- L
- ' "" "C5 ' "" 5 ' ** '
O.0 20.0 40.0 60.0 80.0 100.0 120.0 140.0 160.0 180.0 200.0 Time After Reactor Shutdown (Hours)
RHR HEATUP RATE LO.TP 12101028 LO De litst
____-_,._,__...___._....m_
__,.___._ _-_._ _____ __.. _ ______ ~
...-.~.
v e
1000.0 900.0
[
/
0 800.0 s
f E
/
W 700.0 f
3
/
o 600.0 r
bp 500.0
)
400.0
[
300.0
- #' '"" "C5' ? S' **
O.0 40.0 80.0 120.0 160.0 200.0 20.G 60.0 100.0 140.0 180.0 Time After Reactor Shutdown (Hours)
~
1 TIME TO BOILING LO TP.12101029 JP**"" W
y 8000.0 7000.0 95 p
6000.0
.}
y
,o' 8
',,s cD 5000.0 2
e' _
O
,o' M
4000.0
",e E
e p
,s' 3000.0 m
r 2000.0 0.0 40.0 80.0 120.0
'160.0 200.0 20.0 60.0 100.0 140.0 180.0 Time After Reactor Shutdown (Hours)
TIME FOR CORE UNCOVERY LO TP 12101030 1
I I
J
y
',l1 o
SELECTED LICENSED OPERATOR ANNUNCIATOR TRAINING MATERERIAL J
l l
GENARIC AXNUNCIATOR RESPOXSE i
AND SPECIFIC DIESEL i
GENERATO.RAXNUNCIATOR RESPONSE L
j
=-
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