ML20204B100
| ML20204B100 | |
| Person / Time | |
|---|---|
| Issue date: | 10/31/1988 |
| From: | Padovan L NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD) |
| To: | |
| Shared Package | |
| ML20204B107 | List: |
| References | |
| TASK-AE, TASK-E806 AEOD-E806, GL-87-12, IEIN-86-101, IEIN-87-063, IEIN-87-63, NUDOCS 8810200150 | |
| Download: ML20204B100 (14) | |
Text
.'
r AEOD/EC06 ENGINEERING EVALUATION REPORT LOSS OF DECAY HEAT REMOVAL CAPABILITY DUE TO RAPID REFUELING CAVITY PUMPDOWil October 1988 Prepared by:
L. Mark Padovan Office for Analysis and Evaluation of Operational Cata U.S. Nuclear Regulatory Conrnission 8810200150 881014
~
PDR ORG NEXDp
s SUPRARY A loss of decay heat renoval (OHR) capability while the reactor coolant system (RCS) water level is below the top of the reactor vessel flange during refueling has been an item of concern over the past several years. A number of studies and generic conrunications, incluoing AE00 case study report C503 (Ref.1),
Engineering Evaluation Report E-710 (Ref. 2), Generic Letter 87-12 (Ref. 3).
Infomation Notice 86-101 (Ref. 4) and Information Notice 87-63 (Ref. 5) have been issued to licensees to address this issue. However, a review of plant operating events indicates losses of DHR capability have also occurred at plants that were in the refueling mode with the reactor vessel head removed, even thcugh water remained above the reactor vessel flange in the refueling t
cavity. The potential exists that a vortex may form in the suction of residual l
heat removal (RFP) pumps and possibly cause pump air binding and cavitation l
during pumodown of the refueling cavity at high flow rates.
19,1988 (Ref. 5)g of RHR Vortexin pump suctions cccurred at Byron Unit 1 on September
, Diablo I
Canyon Unit 2 on May 12,1987 (Ref. 6), and Wolf Creek on November 09, 1986 (Ref. 7), when operators rapidly pumped down the refueling cavities.
In these occurrences, operating parsonnel stationed inside containment visually observed water present in the refueling cavity during pumpdown, and thus assumed adequate suction head was available to the PHR pumps. However, as the only drain paths from the refueling cavity to the reactor vessel are through rod cluster control assembly guide tubes, head cooling flow holes in the upper support plate, and between the upper intervals package and the internals support ledge, drainage
]
flow frcm the refueling civity was insufficient to keep up with the large RHR 1
pump flow rates utilized. Accordingly, while water remained in the refueling cavity above the reactor vessel flange, water level in the reactor vessels had 1
been unknowingly lowered below the top of the reactor vessel hot legs causing vortexing of PHR pumps at each facility.
I Corrective actions are still under evaluation at the Byron plant. As corrective actions at the Diablo Canyon and Wolf Creek facilities, plant operating procedures were revised te define appropriate RHP flow rates when refueling cavity water j
level nears the reactor vessel flange, and the importance of operator reliance i
en the reactor vessel refueling level indication system (RVRLIS) was emphasized.
I CESCRIPTION OF OCCURRENCES 1.
Byren Unit 1 On September 19, 1988, the unit was in the refueling mode with the reactor l
1 vessel head removed. Operators were in the process of lowering refueling L
cavity water level to the reactor vessel flange to permit replacement of a reactor vessel head stud plug. The 1A RHR train was operating at 3200 gpm in the shutdown cooling mode, and the IB RHR train was being utilized to lower vessel water level at an approximately 1000 gpm flow rate. A non-licensed operator was inside containment directly observing water level from the operating deck above the reactor ccvity, because the licensee believeo direct observation of level was rm accurate than the installed temporary tygon tube level instrurentation RHR pump 18 was stopped when cavity water level neared the reactor veqel flange.
However, the pump was
i 2-restarted with about four inches of water remaining in the cavity, in order to further reduce water level for maintenance persennel to access the vessel flange stud hole area. When the apparent desired vessel level had been reached (399 feet 9 inches, which is 2 to 3 inches below the vessel flange),
the IB FHR Pump was shut down. Subsequently, a unit I reactor operatnr (RO) observed that amperage indication for the 1A RHR pump motor was oscillating between 20 and 60 amps, and discharge flow was decreasing and also oscil-lating. The F0 shutdcwn the 1A RHR pump, lined up the IB RHR train to the refueling water storage tank (RWST), and started gravity filling the RCS at a 1500 gpn rate.
Shortly, thereafter, the IB RHR pump was started. Peactor vessel level was then restored to 401 feet, and the B train was realigned to the shutdown cooling mode.
The licensee originally believed that refueling cavity / reactor vessel water f
level had been lowered to just relow the vessel flange. However, after i
further evaluation, the licensee. concluded that the actual water level had dropped below the elevation of the top of the reactor coolant hot legs where vortexing of RHR pump 1A could occur. Apparently, once refueling cavity water level dropped below the top of the upper intervals assembly, drainage from the refueling cavity to the reactor vessel eculd no longer
[
pass through the coolant flow windows in the rod cluster control assembly (RCCA) guide tubes or the penetrations in the top of the upper internals assembly. With these drainage paths removed the remaining drain paths could not pass sufficient water flow to keep up with the 1.000 gpm water removal flow rate provided by RHR pump 18.
Instead, air was sucked through i
the PCCA guide tubes causing air intrusion at the suction tn RHR pump 1A (which was operating at 3,200 gpm). No vortexing was observed at the i
suction to PHP pump 10, as the pump was operating at a reduced flow rate I
of abcut 1,000 gpm.
t 2.
Diablo Canyon Unit 2 In preparation to replace a source range detector, on May 12, 1987, with the unit in a defueled condition, operators initiated the process of pumping water from the refueling cavity to the PWST.
The operating 4
procedure (0P) utilized included a caution against reducing the water level in the reactor vessel to the point where suction to the PHP pumps could be potentially lost. The OP specified that once water level in the refueling cavity was lowered to about one foot above the reactor vessel flange, the fueling cavity pumpdown rate was to be throttled back. The OP further directed that when water level was lowered to about one foot below the reactor vessel flange (approximately 113 foot elevation on RVRLIS or tygon tube indication), PHR pump 2-2 was to be stopped and final vessel pumpdown j
was to be accerplished using the refueling water purificatien system. RHR pump 2-2 was aligned to take suction from the PCS loop 4 hot leg and discharge through RHR heat exchange" 2-1 utilizing the heat exchanger flow control valve to regulate the pumpdown rate.
1
~-n v-----.-.w e
e
. Initial cavity water level was about 138' in elevation (24 feet above the reactor vessel flange), with pumpdown continuing at a rate of approximately 3,000 gpm. The cavity water level was being observed by an operator stationed inside contair. ment to confirm wide range RVRLIS control room indications.
Disparities between wide range RVRLIS indications and visual level measurements were observed, but pumpdown flow rate was not throttled since water level visually appeared to be at the 116 foot elevation, and the operators questioned the accuracy of the RVRLIS.
Shortly thereafter, wide range and narrow range RVRLIS both indicated 1131' while the watch i
inside containment reported a 116' level. The RHR heat exchanger outlet flow control valve was then closed, autcmatically placing RHR pump 2-2 on recirculation. An auxiliary building auxiliary operator and a senior reactor operator (SRO) were then sent into containment to independently verify the RVRLIS lineup was correct.
At abnut 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> after the first discrepancy between RVRLIS and direct visual indication inside containment was noted, an approximately four-amp fluctuation of RHR pump 2-2 motor current was observed, and the pump was shut down.
RVRLIS indicated 107' while the operator inside containment reported the cavity water level to be 1161'. The SRO inside containment doubted cavitation of the RHR pump since water remained in the cavity, but verified that both wide and narrow range level indication agreed with tygen tube level indication. After reassessing the problem and changing valve alignments, the RHR pump was restarted, b'Jt again experienced vortexing and was shutdown, t
A decision was then made to reflood the reactor vessel from the RWST and observe RVRLIS irdication. Both wide and narrow RVRLIS showed an increase in level after about thirty seconds.
The operator inside centainment reported no change in cavity level, even though RVRLIS indicated a 5-foot level increase.
Eventually, the operator inside centainment reported water was coming out of the upper internal package rod cluster control guide tubes. At that time, reflood of the reactor vessel was terminated and RVFLIS stabilized at 115' elevation. Operators then concluded that even l
thougn abcut two and ene-half feet of water remined in the cavity above the vessel flange, the water level inside the reactor vessel had been inadvertently lowered to seven feet below the flange.
The licensee's review of the discrepancies between level indications concluded the discrepancies were due to the upper internals assembly forning a seal on the reactor vessel flange since no fuel elements were in the core. With fuel installeo in the core, the leaf springs on the fuel assembly top plate provided enough lifting force to unseat the upper internals package frem the upper internals support ledge when the reactor head was not installed. With the upper internals package seated on the support ledge, the only drain paths from the refueling cavity to the reactor vessel were through the rod cluster control assembly guide tubes
i
.4.
and the head cooling flow boles in the upper support plate. Since the l
guide tubes extend about two feet above the vessel flange, once water level in the refueling cavity dropped to the two-foot (116' elevation) level, flow from the cavity to the reactor vessel was limited to the capacity of the head cooling flow holes which is significantly less than RHR pump capability.
3.
Wolf Creek On November 29, 1986, operations personnel were purping down the refueling
~,
cavity, transferring water to the PWST via RHR pump B at a flow rate of 4
1,500 to 2,000 gpm. RHR pump A was providing RCS recirculation flow at 3,000 gpm. An operator was stationed on the upper edge of the refueling cavity to monitor cavity water level and to notify the control room when cavity level reached reactor vessel flange level.
With cavity water level about one foot above the vessel flange, a "FHR pump A low flow alarm" was received in the control room.
RHR pump B was realigned to the RCS loops, terminating the transfer of water to the RWST.
RHR pump A was shutdown due to the less of flow.
Flow on RHR pump B was subsequently also lost, and the B RHR pump was shut down. An attempt was made to restart RHR pump A, but pump discharge pressure fluctuated and the pump was stopped. RHR pump B was successfully restarted with a 500 to 1,000 gpm RCS recirculation flow. Operators vented "substantial amounts" of air from both pump suction lines.
3 The licensee determined that one of the most probable causes was the limited water ficw from the refueling cavity to the reactor vessel due to the upper internal assembly restricting drainage.
L Operators reflooded the reactor vessel with water from the RWST, and then re-established a 1,400 to 1,500 gpm flow to the RWST utilizing RHR pump B.
Peactor vessel water level was visually monitored at the top of the refueling cavity and using a temporary tygon tube. At slightly above flange 3
level, using the tygon tube, a rapid level drop to midloop level was observed and letdown was secured by switching pump B to RCS recirculation. PHR pump A, recirculating water to the RCS, again experienced air binding. The pump was secured, vented, and returned to service. After reflooding the reactor
~
vessel, pumperwn of the refueling cavity and reactor vessel was perfortree utilizing che chemical and volume control system at a flow rate of abcut 100 gpm.
No further pump vortexing problems were encountered.
ANALYSIS AND EVALUATION Plant Technical Specifications general'y require at least one RHR pump tn be in operation when water level in the refueling cavity is at least 23 feet above the vessel flange.
If water level is less than 23 feet above the flange, two RHP trains rust be operable with one train in service.
These requirements assure sufficient cooling capacity is available to remove decay heet and main-tain water temperature in the reactor vessel below specified limits during i
p refueling. Also, RHR pump flow assures sufficient coolant flow is maintained through the reactor core to minimize the effects of a boron dilution incident, and prevent boren stratification. Additionally, the PHR system provides a small reactor coolant flow ta the chemical and volume control system for purification and reactor coolant chemistry control. Air entrainment at the suction of a RHR pump could cause pump cavitation anc pump inoperability.
In the cases cited, water level in the reactor vessel could be restored through operator actions by starting available charging pumps, safety injection pumps, or by draining water from the FWST to the reactor vessel, since the PWST is locateo at a higher elevation.
This provided additional time for operators to restore a RHR train to service.
4 r
FINDINGS AND CONCLUSIONS The potential exists that a vortex would form in the suctions of RHR pumps during pumpdown of the refueling cavity at high flow rates, even with I
4 several feet of water remaining in the refueling cavity. Air intrucion was observed in three operational events (Byron 1, Diablo Canyon 2, and Wolf Creek) in the past two years.
Corrective actions to preclude occurrences of this natur.1 included licensees performing plant-specific evaluations to determine appropriate refueling cavity pumpdown rates and establishing corresponding minimum refueling cavity water levels at which reduced pumpdewn rates are required. This information was included in operating instructions and operator training programs.
When pumping dcwn the refueling cavity water level, reliance should be placed on reactor vessel refueling level indicating systems wnich measure the water level in the vessel as well as direct visual observation of refueling cavity water level by an operator inside centainment.
a A loss of RHR pumps due to air intrusion represents a si0nificant cperational situation.
I i
i i
1 i
i j
REFERENCES:
l'.
USNPC, Case Study Peport AE00/C503, "0ecay Heat Removal Problems at U.S.
Pressurized Water Reactors," December 1985.
2.
USNRC, Engineering Evaluation Report AEOD/E716, "Excessive Flow Rates in Low Pressure Safety Injection Systems of Westinghouse Plants," October 1987.
3.
USNRC, Ger.eric Letter 87-12. "Loss of Residual Heat Removal (RHR) While the Feactor Coolant System (FCSi is Partially Filled," July 9,1987.
l 4
USNRC IE Information Notice No.86-101, "Loss of Decay Heat Removal Due to Loss of Fluid Levels in Peactor Coolant System." December 12, 1986.
[
5.
USNRC Information Notice No. 87-63, "Inadeouate Net Positive Suction Head in Low Pressure Safety Systems."
6.
-R. Pleniewicz September 21, 1988 letter to K. L. Graesser, Commonwealth Edison, "Potentially Significant Event Preliminary Report," Byron Unit 1.
7.
USNRC Inspection Report No. 50-323/87-20, Diablo Canyon Unit 2. June 23, 1987.
8.
Wolf Creek Event Report No. 86-82, "Air Binding of RHR Pumps in Mode 6,"
Kansas Cas and Electric Company, November 29, 1986.
I t
f l'
l I
I i
c.
-c
.- y - >
3 i
.c INS R.:
" 86.'
i vp errent' 1 c...c0 STATES ag ar7ATORY CCFfSS?X
_ l10H T.30 ENP9'"'MEid TON,' DC 0."
Mber, i 18 IE INFORMA110N NC" 11, sui +1 VOiT 1:
LOSS 0F DECAY HEAT REMOVAL DUE TO LOSS OF FLUID LEVELS IN REACTOR COOLANT SYSTEM Addressees:
All helders of an operating license or a construction permit for pressurized-water reactor (FWR) facilities.
purpose:
This notice is intended to advise licensees of continuing problems during PWR outages in which vorti. ting of residua' heat retreval pump suctions is encountered during pumpoown of refueling cavity water level, prior to establishing mid-loop operation. Yhet a problems have resulted in temporary loss of decay heat removal capability.
It is expected that recipients will review this information for applicability to their reactor facilities and consider actions, if appropriate, to preclude recurrence of similar problems. Suggestions contained in this notice do not constitute NRC recuirements; therefore, no specific action or written response is recuired.
Description of Circumstances:
A typical FWR has a decay heat removal system with two redundant trains.
Generall;, both trains take suction from the same reactor coolant system (RCS) hot leg, and the connecting piping is attached to either the bottom or a lower quadrant of the hot leg. During refueling and certain maintenance activities, water level in the refueling cavity must be lowered below the top of the reactor vessel flange. Lowering the level too quickly can cause water level in he reactor vessel beneath the upper internals assembly to drop belcw the top of the pCS hot leg, while water level remains above the upper internals assembly in the refueling cavity. This occurs when refueling cavity water level drops below the top of the upper internals assembly, eliminating drain paths through the coolant flow windows in the rod cluster control assembly (pCCA) guide tubes and the head cooling flow holes in the upper support plate.
With pumpdown flew rate in excess of drainage flow rate from the refueling cavity, air can then be sucked into the reactor vessel through the RCCA guide tubes causing air instrusion and poter.tial vortex fonnation in the hot leg at the suction nozzle for the decay heat removal system, air entrainment in the water flowing to the operating decay heat reraval pumps, and air binding of
IN 86-101. Supplement 1 October
, 1988 Page 2 of 2 the pumps. Consequently, all decay heat removal capability could be lost until the water level in the reactor vessel is raised and the decay heat remov61 pumps are vented and restarted.
During outages over the last several year period, decay heat removal pumps at several PWRs lost suction because of vortexing encountered during pumpdown of the refueling cavity.
Three of these events are described in Attachment 1 to this infomation notice. Corrective acticns taken by certain licensees involved in these occurrences include 1) performing plant specific evaluations to determine appropriate refueling cavity pump down rates, 2) establishing corresponding minimum refueling cavity water levels at which reduced pumpdown rates are required, 3) incorporating this information into operating procedures and operator training programs, and 4) relying upon refueling reactor vessel water level indicating systems which measures the water level in the vessel during refueling cavity pumpdown, in addition te direct visual observation of cavity water level by an operator inside containment.
4 No specific action or written response is required by this information notice.
If you have any questions about this matter, please contact the Regional Administrator of the appropriate NRC regional office, or this office.
Charles E. Rossi, Director Divisicn of Operational Events Assessment Office of Nuclear Reactor Regulation Tech'eical
Contact:
L. Mark Padovan s AEOD (301) 492-4445 Attachments:
1.
Loss of Decay Feat Removal Capability Events (During Refueling Cavity Pumpdown).
2.
List of Recently Issued Information Notices.
SSINS No.:
IN 86-101, Supplement 1 UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMEhT WASHINGTON, DC 20555 October, 1988 IE INFORMATION NOTICE NO.86-101, SUPPLEMENT 1: LOSS OF DECAY HEAT REMOVAL DUE TO LOSS OF FLUID LEVELS IN REACTOR COOLANT SYSTEM Addressees:
All holders of an operating license or a construction permit for pressurized-water reactor (PWR) facilities.
Purpose:
This notice is intended to advise licensees of continuing problems during PWR outages in which vortexing of residual heat removal pump suctions is encountered during pumpdown of refueling cavity water level, prior to establishing mid-loop operation. These problems have resulted in temporary loss of decay heat removal capability.
It is expected that recipients will review this information for applicability to their reactor facilities and consider actions, if appropriate, to preclude recurrence of similar problems.
Suggestions contained in this notice do not constitute NRC requirements; therefore, no specific action or written response is required.
Description of Circumstances:
A typical PWR has a decay heat renoval system with two redundant trains.
Generally, both trains take suction from the same reactor coolant system (RCS) hot leg, and the connecting piping is attached to either the bottom or a lower quadrant of the hot leg. During refueling and certain maintenance activities, water level in the refueling cavity must be lowered below the top of the reactor vessel flange. Lowering the level too quickly can cause water level in the reactor vessel beneath the upper internals asserbly to drop below the top of the RCS hot leg, while water level remains above the upper internals assembly in the refueling cavity.
This occurs when refueling cavity water level drops below the top of the upper internals assembly, eliminating drain paths through the coolant flow windows in the rod cluster control assembly (RCCA) guide tubes and the head cooling flow holes in the upper support plate.
With pumpdown flow rate in excess of drainage flow rate from the refueling cavity, air can then be sucke<i into the reactor vessel through the RCCA guide tubes causing air instrusion and potential vortex fonnation in the hot leg at the suction nozzle for the decay heat removal system, air entrainment in the water flowing to the operating decay heat removal pumps, and air binding of
IN 86-101, Supplement 1 October
, 1988 Page 2 of 2 the pumps. Consequently, all decay heat removal capability could be lost until the water level in the reactor vessel is raised and the decay heat removal pumps are vented and restarted.
During outages over the last several year period, decay heat removal pumps at several PWRs lost suction because of vortexing encountered during purpdewn of the refueling cavity. Three of these events are described in Attachment 1 to this infomation notice. Corrective actions taken by certain licensees involved in these occurrences include 1) perfoming plant specific evaluations to detemine appropriate refueling cavity pump down rates, 2) establishing corresponding minimum refueling cavity water levels at which reduced pumpdown rates are required. 3) incorporating this infomation into operating procedures and operatur training programs, and 4) relying upon refueling rer: tor vessel water level indicating systems which measures the water level in the vessel during refueling cavity pumpdown, in addition to direct visual observation of cavity water level by an operator inside containment.
No specific action or written response is required by this information notice.
If you have any questions about this matter, please contact the Regional Administrator of the appropriate NPC regional office, or this office.
Charles E. Rossi, Director Division of Operational Events Assessment Office of Nuclear Reactor Pegulation l
Technical
Contact:
L. Mark Padovan, AE00 (301)492-4445 Attachments:
1.
Loss of Oecay Heat Removal Capability Events (During Refueling Cavity Pumpdcwn).
2.
List of Recently Issued Information Notices.
IN 86-101. Supplenent 1 October
, 1988 Page 1 of 3 LOSS OF DECAY HEAT REMOVAL CAPABILITY EVENTS (DURING CAVITY PUPPDOWN) 1.
Byron Unit 1 On September 19, 1988, the unit was in the refueling mode with the reactor vessel head removed. Operators lowered refueling cavity water level to the reactor vessel flange to pemit replacement of a reactor vessel head stud plug. The 1A RHR train was operating at 3200 gpm in the shutdown cooling mode, and the IB RHR train was being utilized to lower vessel water level at an approximately 1000 gpm flow rate. A non-licensed operator was inside containment directly observing water level from the operating deck abnve the reactor cavity, because the licensee believed direct observation of level was more accurate than the installed temporary tygon tube level instrumentation.
PHR pump 1B was stopped when cavity wate;* 1evel neared the reactor vessel flange. However, the pump was restarted with abcut four inches of watcr remaining in the cavity, in order to further reduce water When the apparent desired vessel level had been reached (ge stud hole area.
level for maintenance personnel to access the vessel flan 2 to 3 inches below the vessel flange), the 18 RHR Pump was shut down. Subsequently, a unit I reactor operator (RO) observed that amperage indication for the 1A RHR pump r:otor was oscillating between 20 and 60 amps, and discharge flow was decreasing and also oscillating. The R0 shutdown the 1A RHR pump, lined up the IB RHR train to the refueling water storage tank (RWST), and started gravity filling the FCS at a 1500 gpm rate. Shortly, thereafter, the IB RHR pump was started.
Reactor vessel level was then resu red to one foot above the vessel flange, and the B train was realigned to the shutdcwn cooling mude.
The licensee originally believed that refueling cavity / reactor vessel water level had been lowered to just below the vessel flange. However, af ter further evaluation, the licensee concluded that the actual water level had dropped below the elevation of the top of the reactor coolant hot legs where vortexing of RNR pump 1A could occur. Apparently, once refueling cavity water level dropped below the top of the upper internals assembly, drainage from the refueling cavity to the reactor vessel could no longer pass through the coolant ficw windows in the rod c. luster control assembly (RCCA) guide tubes or the penetrations in the top of the upper internals assembly.
With these drainage paths removed, the remaining drain paths could not pass sufficient weter flow to keep up with the 1,000 gpm water removal flow rate provided by RHR pump 1B.
Instead, air was sucked through the RCCA guide tubes causing air intrusion at the suction to RHR pump 1A (which was operating at 3200 gpm).
No vortexing was observed at the suction to RHR pump 18, as the pump was operating at a reduced flow rate of about 1,000 gpm.
IN 86-101, Supplement 1 October
, 1968 Page 2 of 3 2.
Diablo Canyon Unit 2 In preparation to replace a source range detector, on May 12, 1987, with the unit in a defueled condition, operators initiated the process of pumping water from the refueling cavity to the RWST. RHR pump 2-2 was aligned to take suction from the RCS loop 4 bot leg and discharge through RHR heat exchanger 2-1 utilizing the heat exchanger flow control valve to regulate the pumpdown rate.
Initial cavity water level was about 24' above the reactor vessel flarge, with purpdown continuing at a rate of approximately 3,000 gpm. The cavity water level was being observed by an operator stationed inside centainment to confirm wide range reactor vessel refueling level indication system (RVRLIS) centrol room indications.
Disparities between wide range RVRLIS indications and visual level measurements were observed, but pumpdown flow rate was not throttled since water level visually appeared to be two feet above the vessel flange, and the operators questioned the accuracy of the RVRLIS.
At about li hours after the first discrepancy between RVRLIS and direct visual indication inside containment was noted, an approximately four amp fluctuation of RHR pump 2-2 motor current was observed, and the pump)was shutdown.
RVRLIS indicated 107' (seven feet below the vessel flange while the operator inside contairment reported the cavity water level to be 1161'.
After reassessing the problen and changing valve alignments, the RPp pump was restarted, but again experienced vortexing and was shutdown.
The reactor vessel was reflooded from the RWST and both wide and narrcw RVRLIS showed an increase in level after about thirty seconds. The operator inside containment reported no change in cavity level, even though RVRLIS indicated a 5' level increase. Eventually, the operator inside contain-nent reported water was coming out of the upper internal package rod cluster control guide tubes.' At that time, reflood of the reactor vessel was tenni-nated and RVRLIS stabilized at 115' elevation. Operators then concluded that even though about two and one-half feet of water remained in the cavity I
above the vessel flange, the water level inside the reactor vessel had been inadvertently lowered to seven feet telow the flange.
The licensee's review of the discrepancies between level indications concluded the discrepancies were due to the upper internals package forming a seal on the reactor vessel flange since no fuel assemblies were in the cere. With fuel installed in the core, the leaf springs on the fuel assembly top plate provided enough lifting force to unseat the upper internals package from the upper internals support ledge when the reactor head was not installed. With the upper internals package seated on the l
support ledge, the cnly drain paths from the refueling cavity to the reactor vessel were through the rod cluster control assembly guide tubes e
IN 86-101, Supplement 1 Octcber, 1988 Page 3 of 3 and the head cooling flow holes in the upper support plate. Since the guide tubes extend about two feet above the vessel flange, once water level in the refueling cavity dropped to the two foot (116' elevation) level, flow from the cavity to the reactor vessel was limited to the capacity of the head ccoling flow holes which is significantly less than RHR pump capability.
3.
Wolf Creek On November 29, 1986, operations personnel were pumping down the refueling cavity, transferring water to the PWST via RHR pump 8 at a flow rate of 1500 to 2000 gpm. RHR pump A was providing RCS recirculation flow at 3000 gpm. An operator was stationed on the upper edge cf the refueling cavity to monitor cavity water level and to notify the control rocm when cavity level reached reactor vessel flange level.
^
With cavity water level about one foot above the vessel flange, a RHR pump i
A low flow alarm was received in the control rocm. RHR pump B was realigned to the RCS loops, terminating the transfer of water to the RWST.
RHR pump A was shutdown due to the loss of flow. Flow on RHR pump B was subsequently also lost, and the B RHR pump was shutdown. An attempt was made to restart RHR pump A, but pump pressure fluctuated and the pump was a
i stopped. RHR pump B was successfully restarted with a 500 to 1000 gpm RCS recirculation flow. Operators, dispatched to vent the pump suctions, vented "substantial amounts" of air from both pump suction lines.
4 The licensee determined that one of the most probable causes was the limited water flow from the refueling cavity to the reactor vessel due to the upper internal assembly restricting drainage.
i Operators reflooded the reactor vessel with water from the RWST, and then re-established a 1400 to 1500 gpm flow to the RWST utilizing RHR pump B.
t Reactor vessel water level was monitored at the top of the refueling cavity and using a temporarf tygon tube. At slightly above flange level, a rapid level drop to midloop level was observed in the tygon tube, and letdcwn was secured by switching pump B to RCS recirculation.
RHR pump A, recirculating water to the RCS, again experienced air binding. The pump i
was secured, vented, and returned to service.
After reflooding the reactor i
vessel, pumpdown of the refueling cavity and reactor vessel was performed i
utilizing the chemical and volume control system at a flow rate of about 100 gpm. No further air intrusion problems were encountered.
i i
(
i l
.