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| | issue date = 05/05/1999 | | | issue date = 05/05/1999 |
| | title = Unanticipated Reactor Water Draindown at Quad Cities Unit 2, Arkansas Nuclear One Unit 2, & FitzPatrick | | | title = Unanticipated Reactor Water Draindown at Quad Cities Unit 2, Arkansas Nuclear One Unit 2, & FitzPatrick |
| | author name = Marsh L B | | | author name = Marsh L |
| | author affiliation = NRC/NRR/DRIP | | | author affiliation = NRC/NRR/DRIP |
| | addressee name = | | | addressee name = |
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| | page count = 8 | | | page count = 8 |
| }} | | }} |
| {{#Wiki_filter: | | {{#Wiki_filter:UNITED STATES |
| [[Issue date::May 5, 1999]]
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| NRC INFORMATION NOTICE 99-14: UNANTICIPATED REACTOR WATER DRAINDOWNAT QUAD CITIES UNIT 2, ARKANSAS NUCLEAR ONEUNIT 2, AND FITZPATRICK | | NUCLEAR REGULATORY COMMISSION |
| | |
| | OFFICE OF NUCLEAR REACTOR REGULATION |
| | |
| | WASHINGTON, D.C. 20555-0001 May 5, 1999 NRC INFORMATION NOTICE 99-14: UNANTICIPATED REACTOR WATER DRAINDOWN |
| | |
| | AT QUAD CITIES UNIT 2, ARKANSAS NUCLEAR ONE |
| | |
| | UNIT 2, AND FITZPATRICK |
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|
| ==Addressees== | | ==Addressees== |
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|
| |
|
| ==Purpose== | | ==Purpose== |
| The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice to alertaddressees to the potential for personnel errors during infrequently performed evolutions thatresult in, or contribute to, events such as the inadvertent draining of water from the reactorvessel during shutdown operations. It is expected that recipients will review the information forapplicability to their facilities and consider actions, as appropriate, to prevent a similaroccurrence. However, suggestions contained in this information notice are not NRCrequirements; therefore, no specific action or written response to this notice is required.DescriDtion of CircumstancesQuad Cities Unit 2On February 24, 1999, Quad Cities Unit 2 was in cold shutdown with reactor water temperatureat 131 'F and reactor water level at 80 inches indicated level (normal level during operations is30 inches indicated or 173 inches above the top of active fuel [TAF]). Core cooling was beingmaintained in a band of 120 'F to 170 OF by the OA" loop of the shutdown cooling mode of theresidual heat removal (RHR) system after being switched from the nB" loop at 12:32 a.m.During the switch over the licensee inadvertently failed to close the OA RHR minimum flowvalve as required by the procedure. Sometime later operators noted a decreasing reactor waterlevel and at about 1:02 a.m. secured the *2A RHR pump and isolated shutdown cooling. At1:55 a.m. operators restored the *2A' loop of shutdown cooling to the proper lineup and startedthe *2A RHR pump. Water level had decreased to a minimum of about 45 inches indicated,and reactor water temperature had risen to a maximum of about 163 OF. Forced circulation ofreactor vessel water using a reactor recirculation pump remained in effect throughout the event.On the basis of post event reviews, It appears that the minimum flow valve in the OA loop wasleft open because the nuclear station operator failed to ensure that the tasks were performed inthe sequence specified in the operating procedures. The nuclear station operator who was(7008 PD(L H ort<<4qj-Oiif qqos(J5C7FfcjANW\\b IN 99-14May 5, 1999 directing the evolution from the control room gave the non-licensed operator permission to de-energize the breaker for the WA RHR minimum flow valve operator before the valve was takento the required closed position. De-energizing the breaker also removed power to the valveposition indicator lights in the control room. Thus, when the nuclear station operator tried toverify that the valve was closed, there was no position indication in the control room to makethat verification. The nuclear station operator made the incorrect assumption that the valve wasalready closed and moved to the next step in the procedure. This failure to close the WAX RHRminimum flow valve opened a drain path from the reactor to the suppression pool. To furthercomplicate the event, the operating crew did not recognize that there was any problem untilapproximately 10 minutes had passed and the water level had decreased about 13 inchesbecause of a misinterpretation of causes of the level decrease. After detecting the decrease,the operating crew was slow to react, which allowed the level to decrease another 20 inchesbefore the operators isolated shutdown cooling which terminated the draindown. The licenseeestimated that a total of 6000 to 7000 gallons was drained from the reactor to the suppressionpool.Operations staff practices including poor communications, poor activity briefings for high-riskactivities, lack of effective pre-shift briefings, inadequate supervision of important control roomactivities, inadequate monitoring of control room panels, and slow event response may havecontributed to the event. Although the unintended loss of inventory to the suppression poolhighlighted significant weaknesses in plant operations, the safety significance was minimized bytwo features. First, a reactor recirculation pump remained in service throughout the eventwhich served to distribute decay heat. Second, an automatic isolation of shutdown coolingwould have occurred at 8 inches indicated level which would have stopped the draining event.An indicated water level of 8 inches corresponds to approximately 151 inches of water levelabove the TAF in the reactor core.Arkansas Nuclear One Unit 2On February 2, 1999, at Arkansas Nuclear One Unit 2, the operators were draining therefueling canal in preparation for installing the reactor vessel head. Refueling was completeand steam generator nozzle dams were installed. The operators were using the two lowpressure safety injection (LPSI) pumps to drain the canal to the refueling water storage tank;one pump also served as the shutdown cooling pump. The rate of draindown wasapproximately 3.3 Inches per minute. When the water level reached 105 inches, the reactoroperator noted that level started to lower rapidly. Operators stopped one of the LPSI pumpsand instructed a local operator to close the isolation valve to the refueling water tank. Thismanually operated valve required 55 turns of the handwheel to fully close. Withinapproximately 1.5 minutes, the reactor vessel level had dropped below the 65 inch level (wherereduced inventory begins) and continued down to 56 inches before the valve could be fullyclosed. (Reference zero on these level instruments is the bottom of the hot leg, with mid-loopbeing defined at approximately 24 inches.) The average rate of level decrease between 105 IN 99-14May 5, 1999 inches and 56 inches was approximately 33 inches per minute. At its lowest level, 56 inchesindicated, there were still 93 inches of water above the TAF. Using the high pressure safetyinjection (HPSI) pump the operators brought the level back up to 90 inches. The plant was inreduced inventory operations (below 65 inches) for approximately 7 minutes. During the eventthe level remained well above the point where LPSI pump cavitation would be expected. Thelicensee concluded that the safety significance of the event was minimal because multiplesources of makeup water were available, redundant mitigation equipment was available, andthe operators were quick to recognize and respond to the event.On the basis of post event reviews, it was determined that the procedure used for drainingdown the refueling canal was inadequate in that it incorrectly stated that the draindown shouldbe secured at the 90-inch level. The procedure should have directed that the rate of drainingbe secured at the 106-inch level so that appropriate precautions could be taken beforeresuming the draindown. These precautions should have Included reminders to the operatingcrew that below the 106-inch level the level will drop much more quickly due to the transition ofpumping from a large volume in the refueling canal to a small volume In the reactor vessel.Therefore, in order to maintain control of the water level, the draindown rate should bedecreased and an operator should be stationed to directly monitor the level.Additional factors that contributed to this event include: the operators received little specifictraining on this evolution; the crew was inexperienced in performing this task; the task shouldhave been classified as an infrequent task requiring a more thorough briefing; and, operatorsfailed to station an operator in a position where he could directly monitor the water level in therefueling canal. Instead they monitored it remotely using a video camera that did not provide aclear picture of the water level.FitzPatrickOn December 2, 1998, at the James A. FitzPatrick Nuclear Power Plant, the operators were inthe process of reassembling the reactor following refueling. Operators were controlling thereactor vessel water level at 357 inches above TAF by adjusting the water discharge rate tocompensate for the constant input from the control rod drive cooling water system. While in thiscondition, the licensees risk analysis requires that reactor vessel water level be monitored usingtwo independent level indicators. To meet this requirement, the licensee designated a widerange indicator which provided Indication up to the top of the reactor vessel and an RHRinterlock level indicator which provided indication in the range from -150 inches to +200 Inchesas the instruments to be used during this evaluation.In order for the wide-range level Indicator to remain available with the reactor head removed, atemporary standpipe and fill funnel were used to replace a portion of the reference leg. At thetime of the event, the licensee was in the process of removing this temporary standpipe andreinstalling the original reference leg components. As the water drained from the standpipe, itcaused the wide-range level indicator to erroneously show an increasing water level. For aperiod of approximately one hour the operators in the control room, unaware that the ongoingmaintenance would cause an error in the indicated water level, compensated for the apparentincreasing level by increasing the discharge rate. This action had the effect of reducing the IN 99-14May 5, 1999 actual water level from 357 inches to 255 inches. During the same time period, the operatorswere also in the process of filling and venting the reactor feedwater piping, which could haveaffected the reactor water level. Once the normal reference leg piping had been reinstalled andthe reference leg began to refill, the indicated level decreased from 357 inches to the actuallevel of 255 inches. The second level instrument, which does not come on-scale until the levelgoes below 200 inches, remained off-scale high.When operators discovered the level discrepancy, they used a temporary pressure gaugeconnected to the reactor vessel low-point tap to confirm the actual water level. After confirmingthe accuracy of the wide-range indicator, they restored the reactor vessel water level to 357inches. The 100-inch error represented approximately 14,000 gallons of water. The licenseedetermined that the safety significance of this event was low since the reactor was in coldshutdown with low decay heat and the reactor water level remained well above the TAF. Inaddition, the drain-down would have been limited by an automatic Isolation of the draindownpath, which would have occurred prior to vessel level reaching 177 Inches above the TAF.The licensee's post event review identified: weaknesses in the operator's knowledge of thereactor assembly process; lack of explicit detail in the reactor assembly procedure; and,weaknesses in the plant risk assessment process. Contrary to the assumption that twodesignated reactor water level indicators were available, only one indicator, the wide-rangeinstrument, was available in the range above 200 inches. When the reference leg on the wide-range instrument was disassembled and drained, the one usable indicator was renderedunavailable. The second instrument was pegged off-scale high and remained that waythroughout the event because the level never dropped below 200 inches. A post event review bythe licensee indicated that other reactor water level instruments, remained operable during theevent but, apparently the operators did not rely on these other instruments or notice thediscrepancy between them and the wide range Indicator. Proposed corrective actions includedprocedural enhancements to ensure that reactor level instrumentation credited by the outagerisk assessment remains available during reactor disassembly and reassembly.DiscussionPersonnel errors appear to have caused, or contributed to, these three inadvertent reactorvessel draindown events. The likelihood of personnel errors is dependent upon the operatorsknowledge of the task gained through previous experience and training. It is also dependentupon the quality of the procedures used to perform the task, the level of supervision, theadequacy of pre-job briefings, fatigue, and distractions resulting from multiple tasks. In each ofthe events, the plant staff made errors during a seldom-performed evolution. Because it was aseldom-performed evolution, more training, better pre-job briefings, closer supervision, andprocedures that contain more details than those for frequently performed activities might haveprevented these event IN 99-14May 5, 1999 This information notice requires no specific action or written response. If you have anyquestions about the information in this notice, please contact the technical contact listed below,the appropriate regional office, or the appropriate Office of Nuclear Reactor Regulation (NRR)project manager.Ledyard B. Marsh, ChiefEvents Assessment, Generic CommunicationsAnd Non-Power Reactors BranchDivision of Regulatory Improvement ProgramsOffice of Nuclear Reactor RegulationTechnical contact: Chuck Petrone, NRR301-415-1027E-mail: cdDRenrc.aovREFERENCES:NRC Integrated Inspection Report No. 50-333/98-08, issued February 10, 1999 (Accession No.9902170348) for the James A. FitzPatrick Nuclear Power Plant for the period November 22,1998, through January 10, 1999. | | The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice to alert |
| | |
| | addressees to the potential for personnel errors during infrequently performed evolutions that |
| | |
| | result in, or contribute to, events such as the inadvertent draining of water from the reactor |
| | |
| | vessel during shutdown operations. It is expected that recipients will review the information for |
| | |
| | applicability to their facilities and consider actions, as appropriate, to prevent a similar |
| | |
| | occurrence. However, suggestions contained in this information notice are not NRC |
| | |
| | requirements; therefore, no specific action or written response to this notice is required. |
| | |
| | DescriDtion of Circumstances |
| | |
| | Quad Cities Unit 2 On February 24, 1999, Quad Cities Unit 2 was in cold shutdown with reactor water temperature |
| | |
| | at 131 'F and reactor water level at 80 inches indicated level (normal level during operations is |
| | |
| | 30 inches indicated or 173 inches above the top of active fuel [TAF]). Core cooling was being |
| | |
| | maintained in a band of 120 'F to 170 OF by the OA" loop of the shutdown cooling mode of the |
| | |
| | residual heat removal (RHR) system after being switched from the nB" loop at 12:32 a.m. |
| | |
| | During the switch over the licensee inadvertently failed to close the OA RHR minimum flow |
| | |
| | valve as required by the procedure. Sometime later operators noted a decreasing reactor water |
| | |
| | level and at about 1:02 a.m. secured the *2A RHR pump and isolated shutdown cooling. At |
| | |
| | 1:55 a.m. operators restored the *2A' loop of shutdown cooling to the proper lineup and started |
| | |
| | the *2A RHR pump. Water level had decreased to a minimum of about 45 inches indicated, and reactor water temperature had risen to a maximum of about 163 OF. Forced circulation of |
| | |
| | reactor vessel water using a reactor recirculation pump remained in effect throughout the event. |
| | |
| | On the basis of post event reviews, It appears that the minimum flow valve in the OA loop was |
| | |
| | left open because the nuclear station operator failed to ensure that the tasks were performed in |
| | |
| | the sequence specified in the operating procedures. The nuclear station operator who was |
| | |
| | (7008 PD(L Hort<<4qj-Oiif qqos(J5 C7Ffcj |
| | |
| | ANW\\b |
| | |
| | IN 99-14 May 5, 1999 directing the evolution from the control room gave the non-licensed operator permission to de- energize the breaker for the WARHR minimum flow valve operator before the valve was taken |
| | |
| | to the required closed position. De-energizing the breaker also removed power to the valve |
| | |
| | position indicator lights in the control room. Thus, when the nuclear station operator tried to |
| | |
| | verify that the valve was closed, there was no position indication in the control room to make |
| | |
| | that verification. The nuclear station operator made the incorrect assumption that the valve was |
| | |
| | already closed and moved to the next step in the procedure. This failure to close the WAX RHR |
| | |
| | minimum flow valve opened a drain path from the reactor to the suppression pool. To further |
| | |
| | complicate the event, the operating crew did not recognize that there was any problem until |
| | |
| | approximately 10 minutes had passed and the water level had decreased about 13 inches |
| | |
| | because of a misinterpretation of causes of the level decrease. After detecting the decrease, the operating crew was slow to react, which allowed the level to decrease another 20 inches |
| | |
| | before the operators isolated shutdown cooling which terminated the draindown. The licensee |
| | |
| | estimated that a total of 6000 to 7000 gallons was drained from the reactor to the suppression |
| | |
| | pool. |
| | |
| | Operations staff practices including poor communications, poor activity briefings for high-risk |
| | |
| | activities, lack of effective pre-shift briefings, inadequate supervision of important control room |
| | |
| | activities, inadequate monitoring of control room panels, and slow event response may have |
| | |
| | contributed to the event. Although the unintended loss of inventory to the suppression pool |
| | |
| | highlighted significant weaknesses in plant operations, the safety significance was minimized by |
| | |
| | two features. First, a reactor recirculation pump remained in service throughout the event |
| | |
| | which served to distribute decay heat. Second, an automatic isolation of shutdown cooling |
| | |
| | would have occurred at 8 inches indicated level which would have stopped the draining event. |
| | |
| | An indicated water level of 8 inches corresponds to approximately 151 inches of water level |
| | |
| | above the TAF in the reactor core. |
| | |
| | ===Arkansas Nuclear One Unit 2=== |
| | On February 2, 1999, at Arkansas Nuclear One Unit 2, the operators were draining the |
| | |
| | refueling canal in preparation for installing the reactor vessel head. Refueling was complete |
| | |
| | and steam generator nozzle dams were installed. The operators were using the two low |
| | |
| | pressure safety injection (LPSI) pumps to drain the canal to the refueling water storage tank; |
| | one pump also served as the shutdown cooling pump. The rate of draindown was |
| | |
| | approximately 3.3 Inches per minute. When the water level reached 105 inches, the reactor |
| | |
| | operator noted that level started to lower rapidly. Operators stopped one of the LPSI pumps |
| | |
| | and instructed a local operator to close the isolation valve to the refueling water tank. This |
| | |
| | manually operated valve required 55 turns of the handwheel to fully close. Within |
| | |
| | approximately 1.5 minutes, the reactor vessel level had dropped below the 65 inch level (where |
| | |
| | reduced inventory begins) and continued down to 56 inches before the valve could be fully |
| | |
| | closed. (Reference zero on these level instruments is the bottom of the hot leg, with mid-loop |
| | |
| | being defined at approximately 24 inches.) The average rate of level decrease between 105 |
| | |
| | IN 99-14 May 5, 1999 inches and 56 inches was approximately 33 inches per minute. At its lowest level, 56 inches |
| | |
| | indicated, there were still 93 inches of water above the TAF. Using the high pressure safety |
| | |
| | injection (HPSI) pump the operators brought the level back up to 90 inches. The plant was in |
| | |
| | reduced inventory operations (below 65 inches) for approximately 7 minutes. During the event |
| | |
| | the level remained well above the point where LPSI pump cavitation would be expected. The |
| | |
| | licensee concluded that the safety significance of the event was minimal because multiple |
| | |
| | sources of makeup water were available, redundant mitigation equipment was available, and |
| | |
| | the operators were quick to recognize and respond to the event. |
| | |
| | On the basis of post event reviews, it was determined that the procedure used for draining |
| | |
| | down the refueling canal was inadequate in that it incorrectly stated that the draindown should |
| | |
| | be secured at the 90-inch level. The procedure should have directed that the rate of draining |
| | |
| | be secured at the 106-inch level so that appropriate precautions could be taken before |
| | |
| | resuming the draindown. These precautions should have Included reminders to the operating |
| | |
| | crew that below the 106-inch level the level will drop much more quickly due to the transition of |
| | |
| | pumping from a large volume in the refueling canal to a small volume In the reactor vessel. |
| | |
| | Therefore, in order to maintain control of the water level, the draindown rate should be |
| | |
| | decreased and an operator should be stationed to directly monitor the level. |
| | |
| | Additional factors that contributed to this event include: the operators received little specific |
| | |
| | training on this evolution; the crew was inexperienced in performing this task; the task should |
| | |
| | have been classified as an infrequent task requiring a more thorough briefing; and, operators |
| | |
| | failed to station an operator in a position where he could directly monitor the water level in the |
| | |
| | refueling canal. Instead they monitored it remotely using a video camera that did not provide a |
| | |
| | clear picture of the water level. |
| | |
| | FitzPatrick |
| | |
| | On December 2, 1998, at the James A. FitzPatrick Nuclear Power Plant, the operators were in |
| | |
| | the process of reassembling the reactor following refueling. Operators were controlling the |
| | |
| | reactor vessel water level at 357 inches above TAF by adjusting the water discharge rate to |
| | |
| | compensate for the constant input from the control rod drive cooling water system. While in this |
| | |
| | condition, the licensees risk analysis requires that reactor vessel water level be monitored using |
| | |
| | two independent level indicators. To meet this requirement, the licensee designated a wide |
| | |
| | range indicator which provided Indication up to the top of the reactor vessel and an RHR |
| | |
| | interlock level indicator which provided indication in the range from -150 inches to +200 Inches |
| | |
| | as the instruments to be used during this evaluation. |
| | |
| | In order for the wide-range level Indicator to remain available with the reactor head removed, a |
| | |
| | temporary standpipe and fill funnel were used to replace a portion of the reference leg. At the |
| | |
| | time of the event, the licensee was in the process of removing this temporary standpipe and |
| | |
| | reinstalling the original reference leg components. As the water drained from the standpipe, it |
| | |
| | caused the wide-range level indicator to erroneously show an increasing water level. For a |
| | |
| | period of approximately one hour the operators in the control room, unaware that the ongoing |
| | |
| | maintenance would cause an error in the indicated water level, compensated for the apparent |
| | |
| | increasing level by increasing the discharge rate. This action had the effect of reducing the |
| | |
| | IN 99-14 May 5, 1999 actual water level from 357 inches to 255 inches. During the same time period, the operators |
| | |
| | were also in the process of filling and venting the reactor feedwater piping, which could have |
| | |
| | affected the reactor water level. Once the normal reference leg piping had been reinstalled and |
| | |
| | the reference leg began to refill, the indicated level decreased from 357 inches to the actual |
| | |
| | level of 255 inches. The second level instrument, which does not come on-scale until the level |
| | |
| | goes below 200 inches, remained off-scale high. |
| | |
| | When operators discovered the level discrepancy, they used a temporary pressure gauge |
| | |
| | connected to the reactor vessel low-point tap to confirm the actual water level. After confirming |
| | |
| | the accuracy of the wide-range indicator, they restored the reactor vessel water level to 357 inches. The 100-inch error represented approximately 14,000 gallons of water. The licensee |
| | |
| | determined that the safety significance of this event was low since the reactor was in cold |
| | |
| | shutdown with low decay heat and the reactor water level remained well above the TAF. In |
| | |
| | addition, the drain-down would have been limited by an automatic Isolation of the draindown |
| | |
| | path, which would have occurred prior to vessel level reaching 177 Inches above the TAF. |
| | |
| | The licensee's post event review identified: weaknesses in the operator's knowledge of the |
| | |
| | reactor assembly process; lack of explicit detail in the reactor assembly procedure; and, weaknesses in the plant risk assessment process. Contrary to the assumption that two |
| | |
| | designated reactor water level indicators were available, only one indicator, the wide-range |
| | |
| | instrument, was available in the range above 200 inches. When the reference leg on the wide- range instrument was disassembled and drained, the one usable indicator was rendered |
| | |
| | unavailable. The second instrument was pegged off-scale high and remained that way |
| | |
| | throughout the event because the level never dropped below 200 inches. A post event review by |
| | |
| | the licensee indicated that other reactor water level instruments, remained operable during the |
| | |
| | event but, apparently the operators did not rely on these other instruments or notice the |
| | |
| | discrepancy between them and the wide range Indicator. Proposed corrective actions included |
| | |
| | procedural enhancements to ensure that reactor level instrumentation credited by the outage |
| | |
| | risk assessment remains available during reactor disassembly and reassembly. |
| | |
| | Discussion |
| | |
| | Personnel errors appear to have caused, or contributed to, these three inadvertent reactor |
| | |
| | vessel draindown events. The likelihood of personnel errors is dependent upon the operators |
| | |
| | knowledge of the task gained through previous experience and training. It is also dependent |
| | |
| | upon the quality of the procedures used to perform the task, the level of supervision, the |
| | |
| | adequacy of pre-job briefings, fatigue, and distractions resulting from multiple tasks. In each of |
| | |
| | the events, the plant staff made errors during a seldom-performed evolution. Because it was a |
| | |
| | seldom-performed evolution, more training, better pre-job briefings, closer supervision, and |
| | |
| | procedures that contain more details than those for frequently performed activities might have |
| | |
| | prevented these events. |
| | |
| | IN 99-14 May 5, 1999 This information notice requires no specific action or written response. If you have any |
| | |
| | questions about the information in this notice, please contact the technical contact listed below, the appropriate regional office, or the appropriate Office of Nuclear Reactor Regulation (NRR) |
| | project manager. |
| | |
| | Ledyard B. Marsh, Chief |
| | |
| | Events Assessment, Generic Communications |
| | |
| | And Non-Power Reactors Branch |
| | |
| | Division of Regulatory Improvement Programs |
| | |
| | Office of Nuclear Reactor Regulation |
| | |
| | Technical contact: Chuck Petrone, NRR |
| | |
| | 301-415-1027 E-mail: cdDRenrc.aov |
| | |
| | REFERENCES: |
| | NRC Integrated Inspection Report No. 50-333/98-08, issued February 10, 1999 (Accession No. |
| | |
| | 9902170348) for the James A. FitzPatrick Nuclear Power Plant for the period November 22, |
| | 1998, through January 10, 1999. |
| | |
| | Attachment: List of Recently Issued NRC Information Notices |
| | |
| | ~~ Attachment 1 IN 99-14 May 5, 1999 Page 1 of I |
| | |
| | LIST OF RECENTLY ISSUED |
| | |
| | NRC INFORMATION NOTICES |
| | |
| | Information Date of |
| | |
| | Notice No. Subject Issuance Issued to |
| | |
| | 99-13 Insiahts from NRR Inspections 4129199 All holders of operatina licenses |
| | |
| | of Low-and Medium-Voltage for nuclear power reactors |
| | |
| | Circuit Breaker Maintenance |
| | |
| | Programs |
| | |
| | 99-12 Year 2000 Computer Systems 4/28/99 All holders of operating licenses |
| | |
| | Readiness Audits or construction permits for nuclear |
| | |
| | power plants |
| | |
| | 99-11 Incidents Involving the Use of 4/23/99 All medical use licensees |
| | |
| | Radioactive Iodine-131 |
| | 97-15, Sup 1 Reporting of Errors and 4/16/99 All holders of operating licenses |
| | |
| | Changes in Large-Break/Small- for nuclear power reactors, except |
| | |
| | Break Loss-of-Coolant Evaluation those who have permanently |
| | |
| | Models of Fuel Vendors and cease operations and have |
| | |
| | Compliance with 10 CFR 50.46(a)(3) certified that fuel has been |
| | |
| | permanently removed from the |
| | |
| | reactor |
| | |
| | 99-10 Degradation of Prestressing 4/13/99 All holders of operating licenses |
| | |
| | Tendon Systems in Prestressed for nuclear power reactors |
| | |
| | Concrete Containments |
| | |
| | 99-09 Problems Encountered When 3/24/99 All medical licensees authorized |
| | |
| | Manually Editing Treatment Data to conduct high-dose-rate (HDR) |
| | on The Nucletron Microselectron-HDR remote after loading |
| | |
| | (New) Model 105.999 brachytherapy treatments |
| | |
| | 99-08 Urine Specimen Adulteration 4/1/99 All holders of operating licensees |
| | |
| | for nuclear power reactors and |
| | |
| | licensees authorized to possess |
| | |
| | or use formula quantities of |
| | |
| | strategic special nuclear material |
| | |
| | OL = Operating License |
| | |
| | CP = Construction Permit |
| | |
| | IN 99-xx |
| | |
| | April xx, 1999 Page 5of 5 This information notice requires no specific action or written response. If you have any |
| | |
| | questions about the information in this notice, please contact the technical contact listed below, the appropriate regional office, or the appropriate office of Nuclear Reactor Regulation (NRR) |
| | Project Manager. |
| | |
| | Ledyard B. Marsh, Chief |
| | |
| | Events Assessment, Generic Communications |
| | |
| | And Non-Power Reactors Branch |
| | |
| | Division of Regulatory Improvement Programs |
| | |
| | Office of Nuclear Reactor Regulation |
| | |
| | Technical contact: Chuck Petrone, NRR |
| | |
| | 301-415-1027 E-mail: cdRDanrc.aov |
| | |
| | REFERENCES: |
| | NRC Integrated Inspection Report No. 50-333198-08, issued February 10, 1999 (Accession No. |
| | |
| | 9902170348) for the James A. FitzPatrick Nuclear Power Plant for the period November 22, |
| | 1998, through January 10, 1999. |
| | |
| | Attachments: |
| | 1. List of Recently Issued NMSS Information Notices |
| | |
| | 2. List of Recently Issued NRC Information Notices |
| | |
| | DOCUMENT NAME: G:ICDPDRAININ\DRAIN.0B.WPD |
| | |
| | To receive a copy of this document, Indicate In the box C=Copy w/o attachmentlenclosure E=Copy with attachment/enclosure N = No copy |
| | |
| | OFFICE PECB:DRIP I Tech Editor l DRCH I PDIV-1 I |
| | |
| | NAME CPetrone I_ RGallo 1 MNolangfarP. |
| | |
| | DATE V/0199 F . |
| | |
| | [3 /1/99 V. . |
| | |
| | 4/4I9 |
| | . . |
| | |
| | 1' /0g99 OFFICE PDI-1 IA .I PDIII-2 I C:PECB:DRIP I |
| | |
| | NAME 2Jiiam RPulsjier LMarsh |
| | |
| | DATE lf/499 I1'/t 99 I /99 OFFICIAL RECORD COPY |
| | |
| | IN 99-14 May 5, 1999 This information notice requires no specific action or written response. If you have any |
| | |
| | questions about the information in this notice, please contact the technical contact listed below, the appropriate regional office, or the appropriate Office of Nuclear Reactor Regulation (NRR) |
| | project manager. |
| | |
| | [arig sjid by] |
| | Ledyard B. Marsh, Chief |
| | |
| | Events Assessment, Generic Communications |
| | |
| | And Non-Power Reactors Branch |
| | |
| | Division of Regulatory Improvement Programs |
| | |
| | Office of Nuclear Reactor Regulation |
| | |
| | Technical contact: Chuck Petrone, NRR |
| | |
| | 301-415-1027 E-mail: cdr)ODnrc.gov |
| | |
| | REFERENCES: |
| | NRC Integrated Inspection Report No. 50-333/98-08, issued February 10, 1999 (Accession No. |
| | |
| | 9902170348) for the James A. FitzPatrick Nuclear Power Plant for the period November 22, |
| | 1998, through January 10, 1999. |
| | |
| | Attachment: List of Recently Issued NRC Information Notices |
| | |
| | DOCUMENT NAME: S:XDRPMSEC\99-14.IN |
| | |
| | *See previous concurrence |
| | |
| | To receive a copy of this document. indicate in the box C=CoDv w/o attachment/enclosure E=CoDv with attachment/enclosure N = No coov |
|
| |
|
| ===Attachment:===
| | OFFICE PECB:DRlIP I Tech Editor l DRCH l-ii PDIV-1 lI |
| List of Recently Issued NRC Information Notices
| |
|
| |
|
| ~~ Attachment 1IN 99-14May 5, 1999Page 1 of ILIST OF RECENTLY ISSUEDNRC INFORMATION NOTICESInformation Date ofNotice No. Subject Issuance Issued to99-13 Insiahts from NRR Inspections 4129199 All holders of operatina licensesof Low-and Medium-VoltageCircuit Breaker MaintenanceProgramsfor nuclear power reactors99-12Year 2000 Computer SystemsReadiness AuditsIncidents Involving the Use ofRadioactive Iodine-1314/28/994/23/99All holders of operating licensesor construction permits for nuclearpower plantsAll medical use licensees99-1197-15, Sup 1Reporting of Errors and 4/16/99Changes in Large-Break/Small-Break Loss-of-Coolant EvaluationModels of Fuel Vendors andCompliance with 10 CFR 50.46(a)(3)All holders of operating licensesfor nuclear power reactors, exceptthose who have permanentlycease operations and havecertified that fuel has beenpermanently removed from thereactor99-1099-09Degradation of Prestressing 4/13/99Tendon Systems in PrestressedConcrete ContainmentsProblems Encountered When 3/24/99Manually Editing Treatment Dataon The Nucletron Microselectron-HDR(New) Model 105.999Urine Specimen Adulteration 4/1/99All holders of operating licensesfor nuclear power reactorsAll medical licensees authorizedto conduct high-dose-rate (HDR)remote after loadingbrachytherapy treatmentsAll holders of operating licenseesfor nuclear power reactors andlicensees authorized to possessor use formula quantities ofstrategic special nuclear material99-08OL = Operating LicenseCP = Construction Permit IN 99-xxApril xx, 1999Page 5of 5This information notice requires no specific action or written response. If you have anyquestions about the information in this notice, please contact the technical contact listed below,the appropriate regional office, or the appropriate office of Nuclear Reactor Regulation (NRR)Project Manager.Ledyard B. Marsh, ChiefEvents Assessment, Generic CommunicationsAnd Non-Power Reactors BranchDivision of Regulatory Improvement ProgramsOffice of Nuclear Reactor RegulationTechnical contact:Chuck Petrone, NRR301-415-1027E-mail: cdRDanrc.aovREFERENCES:NRC Integrated Inspection Report No. 50-333198-08, issued February 10, 1999 (Accession No.9902170348) for the James A. FitzPatrick Nuclear Power Plant for the period November 22,1998, through January 10, 1999.
| | NAME CPetrone* BCalure* RGallo* MNolan* |
| | DATE 04/27/99 .3/15/99 _________04128199 = 04/27/99 |
| | 1 . . . |
|
| |
|
| ===Attachments:===
| | OFFICE PDI-1 I PD111-2 C:PECB:DJRIP I |
| 1. List of Recently Issued NMSS Information Notices2. List of Recently Issued NRC Information NoticesDOCUMENT NAME: G:ICDPDRAININ\DRAIN.0B.WPDTo receive a copy of this document, Indicate In the box C=Copy w/o attachmentlenclosure E=Copy with attachment/enclosure N = No copyOFFICE PECB:DRIP I Tech Editor l DRCH I PDIV-1 INAME CPetrone I_ RGallo 1 MNolangfarP.DATE V /0199 [3 /1/99 4 /4I9 1' /0g99F .V. ...OFFICEPDI-1 IA .IPDIII-2IC:PECB:DRIPINAME 2Jiiam RPulsjier LMarshDATE lf/499 I1'/t 99 I /99OFFICIAL RECORD COPY IN 99-14May 5, 1999 This information notice requires no specific action or written response. If you have anyquestions about the information in this notice, please contact the technical contact listed below,the appropriate regional office, or the appropriate Office of Nuclear Reactor Regulation (NRR)project manager.[arig sjid by]Ledyard B. Marsh, ChiefEvents Assessment, Generic CommunicationsAnd Non-Power Reactors BranchDivision of Regulatory Improvement ProgramsOffice of Nuclear Reactor RegulationTechnical contact:Chuck Petrone, NRR301-415-1027E-mail: cdr)ODnrc.govREFERENCES:NRC Integrated Inspection Report No. 50-333/98-08, issued February 10, 1999 (Accession No.9902170348) for the James A. FitzPatrick Nuclear Power Plant for the period November 22,1998, through January 10, 1999. | |
|
| |
|
| ===Attachment:===
| | NAME JWilliams* RPulsifer' _I-Marsh _ __ _ |
| List of Recently Issued NRC Information NoticesDOCUMENT NAME: S:XDRPMSEC\99-14.IN*See previous concurrenceTo receive a copy of this document. indicate in the box C=CoDv w/o attachment/enclosure E=CoDv with attachment/enclosure N = No coovOFFICE PECB:DRlIP I Tech Editor l DRCH l-ii PDIV-1 lINAME CPetrone* BCalure* RGallo* MNolan*DATE 04/27/99 .3/15/99 _________04128199 = 04/27/991 ...OFFICEPDI-1IPD111-2C:PECB:DJRIPINAME JWilliams* RPulsifer' I-Marsh _ _ __ _DATE 04/27/9 .04/27/99 k,-u99OFFICIAL RECORD COPY}}
| | DATE 04/27/9 . 04/27/99 k,-u99 OFFICIAL RECORD COPY}} |
|
| |
|
| {{Information notice-Nav}} | | {{Information notice-Nav}} |
Unanticipated Reactor Water Draindown at Quad Cities Unit 2, Arkansas Nuclear One Unit 2, & FitzPatrickML031040444 |
Person / Time |
---|
Site: |
Beaver Valley, Millstone, Hatch, Monticello, Calvert Cliffs, Dresden, Davis Besse, Peach Bottom, Browns Ferry, Salem, Oconee, Mcguire, Nine Mile Point, Palisades, Palo Verde, Perry, Indian Point, Fermi, Kewaunee, Catawba, Harris, Wolf Creek, Saint Lucie, Point Beach, Oyster Creek, Watts Bar, Hope Creek, Grand Gulf, Cooper, Sequoyah, Byron, Pilgrim, Arkansas Nuclear, Three Mile Island, Braidwood, Susquehanna, Summer, Prairie Island, Columbia, Seabrook, Brunswick, Surry, Limerick, North Anna, Turkey Point, River Bend, Vermont Yankee, Crystal River, Haddam Neck, Ginna, Diablo Canyon, Callaway, Vogtle, Waterford, Duane Arnold, Farley, Robinson, Clinton, South Texas, San Onofre, Cook, Comanche Peak, Yankee Rowe, Maine Yankee, Quad Cities, Humboldt Bay, La Crosse, Big Rock Point, Rancho Seco, Zion, Midland, Bellefonte, Fort Calhoun, FitzPatrick, McGuire, LaSalle, Fort Saint Vrain, Shoreham, Satsop, Trojan, Atlantic Nuclear Power Plant |
---|
Issue date: |
05/05/1999 |
---|
From: |
Marsh L Division of Regulatory Improvement Programs |
---|
To: |
|
---|
References |
---|
IN-99-014, NUDOCS 9905070080 |
Download: ML031040444 (8) |
|
Similar Documents at Beaver Valley, Millstone, Hatch, Monticello, Calvert Cliffs, Dresden, Davis Besse, Peach Bottom, Browns Ferry, Salem, Oconee, Mcguire, Nine Mile Point, Palisades, Palo Verde, Perry, Indian Point, Fermi, Kewaunee, Catawba, Harris, Wolf Creek, Saint Lucie, Point Beach, Oyster Creek, Watts Bar, Hope Creek, Grand Gulf, Cooper, Sequoyah, Byron, Pilgrim, Arkansas Nuclear, Three Mile Island, Braidwood, Susquehanna, Summer, Prairie Island, Columbia, Seabrook, Brunswick, Surry, Limerick, North Anna, Turkey Point, River Bend, Vermont Yankee, Crystal River, Haddam Neck, Ginna, Diablo Canyon, Callaway, Vogtle, Waterford, Duane Arnold, Farley, Robinson, Clinton, South Texas, San Onofre, Cook, Comanche Peak, Yankee Rowe, Maine Yankee, Quad Cities, Humboldt Bay, La Crosse, Big Rock Point, Rancho Seco, Zion, Midland, Bellefonte, Fort Calhoun, FitzPatrick, McGuire, LaSalle, Fort Saint Vrain, Shoreham, Satsop, Trojan, Atlantic Nuclear Power Plant |
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Category:NRC Information Notice
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Mclaughlin on NRC, Regarding NRC Information Notice 2006-13: Groundwater Contamination2006-07-13013 July 2006 E-mail from M. Mclaughlin on NRC, Regarding NRC Information Notice 2006-13: Groundwater Contamination 2020-09-03 The following query condition could not be considered due to this wiki's restrictions on query size or depth: <code> [[:Beaver Valley]] OR [[:Millstone]] OR [[:Hatch]] OR [[:Monticello]] OR [[:Calvert Cliffs]] OR [[:Dresden]] OR [[:Davis Besse]] OR [[:Peach Bottom]] OR [[:Browns Ferry]] OR [[:Salem]] OR [[:Oconee]] OR [[:Mcguire]] OR [[:Nine Mile Point]] OR [[:Palisades]] OR [[:Palo Verde]] OR [[:Perry]] OR [[:Indian Point]] OR [[:Fermi]] OR [[:Kewaunee]] OR [[:Catawba]] OR [[:Harris]] OR [[:Wolf Creek]] OR [[:Saint Lucie]] OR [[:Point Beach]] OR [[:Oyster Creek]] OR [[:Watts Bar]] OR [[:Hope Creek]] OR [[:Grand Gulf]] OR [[:Cooper]] OR [[:Sequoyah]] OR [[:Byron]] OR [[:Pilgrim]] OR [[:Arkansas Nuclear]] OR [[:Three Mile Island]] OR [[:Braidwood]] OR [[:Susquehanna]] OR [[:Summer]] OR [[:Prairie Island]] OR [[:Columbia]] OR [[:Seabrook]] OR [[:Brunswick]] OR [[:Surry]] OR [[:Limerick]] OR [[:North Anna]] OR [[:Turkey Point]] OR [[:River Bend]] OR [[:Vermont Yankee]] OR [[:Crystal River]] OR [[:Haddam Neck]] OR [[:Ginna]] OR [[:Diablo Canyon]] OR [[:Callaway]] OR [[:Vogtle]] OR [[:Waterford]] OR [[:Duane Arnold]] OR [[:Farley]] OR [[:Robinson]] OR [[:Clinton]] OR [[:South Texas]] OR [[:San Onofre]] OR [[:Cook]] OR [[:Comanche Peak]] OR [[:Yankee Rowe]] OR [[:Maine Yankee]] OR [[:Quad Cities]] OR [[:Humboldt Bay]] OR [[:La Crosse]] OR [[:Big Rock Point]] OR [[:Rancho Seco]] OR [[:Zion]] OR [[:Midland]] OR [[:Bellefonte]] OR [[:Fort Calhoun]] OR [[:FitzPatrick]] OR [[:McGuire]] OR [[:LaSalle]] OR [[:Fort Saint Vrain]] OR [[:Shoreham]] OR [[:Satsop]] OR [[:Trojan]] OR [[:Atlantic Nuclear Power Plant]] </code>.
[Table view]The following query condition could not be considered due to this wiki's restrictions on query size or depth: <code> [[:Beaver Valley]] OR [[:Millstone]] OR [[:Hatch]] OR [[:Monticello]] OR [[:Calvert Cliffs]] OR [[:Dresden]] OR [[:Davis Besse]] OR [[:Peach Bottom]] OR [[:Browns Ferry]] OR [[:Salem]] OR [[:Oconee]] OR [[:Mcguire]] OR [[:Nine Mile Point]] OR [[:Palisades]] OR [[:Palo Verde]] OR [[:Perry]] OR [[:Indian Point]] OR [[:Fermi]] OR [[:Kewaunee]] OR [[:Catawba]] OR [[:Harris]] OR [[:Wolf Creek]] OR [[:Saint Lucie]] OR [[:Point Beach]] OR [[:Oyster Creek]] OR [[:Watts Bar]] OR [[:Hope Creek]] OR [[:Grand Gulf]] OR [[:Cooper]] OR [[:Sequoyah]] OR [[:Byron]] OR [[:Pilgrim]] OR [[:Arkansas Nuclear]] OR [[:Three Mile Island]] OR [[:Braidwood]] OR [[:Susquehanna]] OR [[:Summer]] OR [[:Prairie Island]] OR [[:Columbia]] OR [[:Seabrook]] OR [[:Brunswick]] OR [[:Surry]] OR [[:Limerick]] OR [[:North Anna]] OR [[:Turkey Point]] OR [[:River Bend]] OR [[:Vermont Yankee]] OR [[:Crystal River]] OR [[:Haddam Neck]] OR [[:Ginna]] OR [[:Diablo Canyon]] OR [[:Callaway]] OR [[:Vogtle]] OR [[:Waterford]] OR [[:Duane Arnold]] OR [[:Farley]] OR [[:Robinson]] OR [[:Clinton]] OR [[:South Texas]] OR [[:San Onofre]] OR [[:Cook]] OR [[:Comanche Peak]] OR [[:Yankee Rowe]] OR [[:Maine Yankee]] OR [[:Quad Cities]] OR [[:Humboldt Bay]] OR [[:La Crosse]] OR [[:Big Rock Point]] OR [[:Rancho Seco]] OR [[:Zion]] OR [[:Midland]] OR [[:Bellefonte]] OR [[:Fort Calhoun]] OR [[:FitzPatrick]] OR [[:McGuire]] OR [[:LaSalle]] OR [[:Fort Saint Vrain]] OR [[:Shoreham]] OR [[:Satsop]] OR [[:Trojan]] OR [[:Atlantic Nuclear Power Plant]] </code>. |
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C. 20555-0001 May 5, 1999 NRC INFORMATION NOTICE 99-14: UNANTICIPATED REACTOR WATER DRAINDOWN
AT QUAD CITIES UNIT 2, ARKANSAS NUCLEAR ONE
UNIT 2, AND FITZPATRICK
Addressees
All holders of licenses for nuclear power, test, and research reactors.
Purpose
The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice to alert
addressees to the potential for personnel errors during infrequently performed evolutions that
result in, or contribute to, events such as the inadvertent draining of water from the reactor
vessel during shutdown operations. It is expected that recipients will review the information for
applicability to their facilities and consider actions, as appropriate, to prevent a similar
occurrence. However, suggestions contained in this information notice are not NRC
requirements; therefore, no specific action or written response to this notice is required.
DescriDtion of Circumstances
Quad Cities Unit 2 On February 24, 1999, Quad Cities Unit 2 was in cold shutdown with reactor water temperature
at 131 'F and reactor water level at 80 inches indicated level (normal level during operations is
30 inches indicated or 173 inches above the top of active fuel [TAF]). Core cooling was being
maintained in a band of 120 'F to 170 OF by the OA" loop of the shutdown cooling mode of the
residual heat removal (RHR) system after being switched from the nB" loop at 12:32 a.m.
During the switch over the licensee inadvertently failed to close the OA RHR minimum flow
valve as required by the procedure. Sometime later operators noted a decreasing reactor water
level and at about 1:02 a.m. secured the *2A RHR pump and isolated shutdown cooling. At
1:55 a.m. operators restored the *2A' loop of shutdown cooling to the proper lineup and started
the *2A RHR pump. Water level had decreased to a minimum of about 45 inches indicated, and reactor water temperature had risen to a maximum of about 163 OF. Forced circulation of
reactor vessel water using a reactor recirculation pump remained in effect throughout the event.
On the basis of post event reviews, It appears that the minimum flow valve in the OA loop was
left open because the nuclear station operator failed to ensure that the tasks were performed in
the sequence specified in the operating procedures. The nuclear station operator who was
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IN 99-14 May 5, 1999 directing the evolution from the control room gave the non-licensed operator permission to de- energize the breaker for the WARHR minimum flow valve operator before the valve was taken
to the required closed position. De-energizing the breaker also removed power to the valve
position indicator lights in the control room. Thus, when the nuclear station operator tried to
verify that the valve was closed, there was no position indication in the control room to make
that verification. The nuclear station operator made the incorrect assumption that the valve was
already closed and moved to the next step in the procedure. This failure to close the WAX RHR
minimum flow valve opened a drain path from the reactor to the suppression pool. To further
complicate the event, the operating crew did not recognize that there was any problem until
approximately 10 minutes had passed and the water level had decreased about 13 inches
because of a misinterpretation of causes of the level decrease. After detecting the decrease, the operating crew was slow to react, which allowed the level to decrease another 20 inches
before the operators isolated shutdown cooling which terminated the draindown. The licensee
estimated that a total of 6000 to 7000 gallons was drained from the reactor to the suppression
pool.
Operations staff practices including poor communications, poor activity briefings for high-risk
activities, lack of effective pre-shift briefings, inadequate supervision of important control room
activities, inadequate monitoring of control room panels, and slow event response may have
contributed to the event. Although the unintended loss of inventory to the suppression pool
highlighted significant weaknesses in plant operations, the safety significance was minimized by
two features. First, a reactor recirculation pump remained in service throughout the event
which served to distribute decay heat. Second, an automatic isolation of shutdown cooling
would have occurred at 8 inches indicated level which would have stopped the draining event.
An indicated water level of 8 inches corresponds to approximately 151 inches of water level
above the TAF in the reactor core.
Arkansas Nuclear One Unit 2
On February 2, 1999, at Arkansas Nuclear One Unit 2, the operators were draining the
refueling canal in preparation for installing the reactor vessel head. Refueling was complete
and steam generator nozzle dams were installed. The operators were using the two low
pressure safety injection (LPSI) pumps to drain the canal to the refueling water storage tank;
one pump also served as the shutdown cooling pump. The rate of draindown was
approximately 3.3 Inches per minute. When the water level reached 105 inches, the reactor
operator noted that level started to lower rapidly. Operators stopped one of the LPSI pumps
and instructed a local operator to close the isolation valve to the refueling water tank. This
manually operated valve required 55 turns of the handwheel to fully close. Within
approximately 1.5 minutes, the reactor vessel level had dropped below the 65 inch level (where
reduced inventory begins) and continued down to 56 inches before the valve could be fully
closed. (Reference zero on these level instruments is the bottom of the hot leg, with mid-loop
being defined at approximately 24 inches.) The average rate of level decrease between 105
IN 99-14 May 5, 1999 inches and 56 inches was approximately 33 inches per minute. At its lowest level, 56 inches
indicated, there were still 93 inches of water above the TAF. Using the high pressure safety
injection (HPSI) pump the operators brought the level back up to 90 inches. The plant was in
reduced inventory operations (below 65 inches) for approximately 7 minutes. During the event
the level remained well above the point where LPSI pump cavitation would be expected. The
licensee concluded that the safety significance of the event was minimal because multiple
sources of makeup water were available, redundant mitigation equipment was available, and
the operators were quick to recognize and respond to the event.
On the basis of post event reviews, it was determined that the procedure used for draining
down the refueling canal was inadequate in that it incorrectly stated that the draindown should
be secured at the 90-inch level. The procedure should have directed that the rate of draining
be secured at the 106-inch level so that appropriate precautions could be taken before
resuming the draindown. These precautions should have Included reminders to the operating
crew that below the 106-inch level the level will drop much more quickly due to the transition of
pumping from a large volume in the refueling canal to a small volume In the reactor vessel.
Therefore, in order to maintain control of the water level, the draindown rate should be
decreased and an operator should be stationed to directly monitor the level.
Additional factors that contributed to this event include: the operators received little specific
training on this evolution; the crew was inexperienced in performing this task; the task should
have been classified as an infrequent task requiring a more thorough briefing; and, operators
failed to station an operator in a position where he could directly monitor the water level in the
refueling canal. Instead they monitored it remotely using a video camera that did not provide a
clear picture of the water level.
FitzPatrick
On December 2, 1998, at the James A. FitzPatrick Nuclear Power Plant, the operators were in
the process of reassembling the reactor following refueling. Operators were controlling the
reactor vessel water level at 357 inches above TAF by adjusting the water discharge rate to
compensate for the constant input from the control rod drive cooling water system. While in this
condition, the licensees risk analysis requires that reactor vessel water level be monitored using
two independent level indicators. To meet this requirement, the licensee designated a wide
range indicator which provided Indication up to the top of the reactor vessel and an RHR
interlock level indicator which provided indication in the range from -150 inches to +200 Inches
as the instruments to be used during this evaluation.
In order for the wide-range level Indicator to remain available with the reactor head removed, a
temporary standpipe and fill funnel were used to replace a portion of the reference leg. At the
time of the event, the licensee was in the process of removing this temporary standpipe and
reinstalling the original reference leg components. As the water drained from the standpipe, it
caused the wide-range level indicator to erroneously show an increasing water level. For a
period of approximately one hour the operators in the control room, unaware that the ongoing
maintenance would cause an error in the indicated water level, compensated for the apparent
increasing level by increasing the discharge rate. This action had the effect of reducing the
IN 99-14 May 5, 1999 actual water level from 357 inches to 255 inches. During the same time period, the operators
were also in the process of filling and venting the reactor feedwater piping, which could have
affected the reactor water level. Once the normal reference leg piping had been reinstalled and
the reference leg began to refill, the indicated level decreased from 357 inches to the actual
level of 255 inches. The second level instrument, which does not come on-scale until the level
goes below 200 inches, remained off-scale high.
When operators discovered the level discrepancy, they used a temporary pressure gauge
connected to the reactor vessel low-point tap to confirm the actual water level. After confirming
the accuracy of the wide-range indicator, they restored the reactor vessel water level to 357 inches. The 100-inch error represented approximately 14,000 gallons of water. The licensee
determined that the safety significance of this event was low since the reactor was in cold
shutdown with low decay heat and the reactor water level remained well above the TAF. In
addition, the drain-down would have been limited by an automatic Isolation of the draindown
path, which would have occurred prior to vessel level reaching 177 Inches above the TAF.
The licensee's post event review identified: weaknesses in the operator's knowledge of the
reactor assembly process; lack of explicit detail in the reactor assembly procedure; and, weaknesses in the plant risk assessment process. Contrary to the assumption that two
designated reactor water level indicators were available, only one indicator, the wide-range
instrument, was available in the range above 200 inches. When the reference leg on the wide- range instrument was disassembled and drained, the one usable indicator was rendered
unavailable. The second instrument was pegged off-scale high and remained that way
throughout the event because the level never dropped below 200 inches. A post event review by
the licensee indicated that other reactor water level instruments, remained operable during the
event but, apparently the operators did not rely on these other instruments or notice the
discrepancy between them and the wide range Indicator. Proposed corrective actions included
procedural enhancements to ensure that reactor level instrumentation credited by the outage
risk assessment remains available during reactor disassembly and reassembly.
Discussion
Personnel errors appear to have caused, or contributed to, these three inadvertent reactor
vessel draindown events. The likelihood of personnel errors is dependent upon the operators
knowledge of the task gained through previous experience and training. It is also dependent
upon the quality of the procedures used to perform the task, the level of supervision, the
adequacy of pre-job briefings, fatigue, and distractions resulting from multiple tasks. In each of
the events, the plant staff made errors during a seldom-performed evolution. Because it was a
seldom-performed evolution, more training, better pre-job briefings, closer supervision, and
procedures that contain more details than those for frequently performed activities might have
prevented these events.
IN 99-14 May 5, 1999 This information notice requires no specific action or written response. If you have any
questions about the information in this notice, please contact the technical contact listed below, the appropriate regional office, or the appropriate Office of Nuclear Reactor Regulation (NRR)
project manager.
Ledyard B. Marsh, Chief
Events Assessment, Generic Communications
And Non-Power Reactors Branch
Division of Regulatory Improvement Programs
Office of Nuclear Reactor Regulation
Technical contact: Chuck Petrone, NRR
301-415-1027 E-mail: cdDRenrc.aov
REFERENCES:
NRC Integrated Inspection Report No. 50-333/98-08, issued February 10, 1999 (Accession No.
9902170348) for the James A. FitzPatrick Nuclear Power Plant for the period November 22,
1998, through January 10, 1999.
Attachment: List of Recently Issued NRC Information Notices
~~ Attachment 1 IN 99-14 May 5, 1999 Page 1 of I
LIST OF RECENTLY ISSUED
NRC INFORMATION NOTICES
Information Date of
Notice No. Subject Issuance Issued to
99-13 Insiahts from NRR Inspections 4129199 All holders of operatina licenses
of Low-and Medium-Voltage for nuclear power reactors
Circuit Breaker Maintenance
Programs
99-12 Year 2000 Computer Systems 4/28/99 All holders of operating licenses
Readiness Audits or construction permits for nuclear
power plants
99-11 Incidents Involving the Use of 4/23/99 All medical use licensees
Radioactive Iodine-131
97-15, Sup 1 Reporting of Errors and 4/16/99 All holders of operating licenses
Changes in Large-Break/Small- for nuclear power reactors, except
Break Loss-of-Coolant Evaluation those who have permanently
Models of Fuel Vendors and cease operations and have
Compliance with 10 CFR 50.46(a)(3) certified that fuel has been
permanently removed from the
reactor
99-10 Degradation of Prestressing 4/13/99 All holders of operating licenses
Tendon Systems in Prestressed for nuclear power reactors
Concrete Containments
99-09 Problems Encountered When 3/24/99 All medical licensees authorized
Manually Editing Treatment Data to conduct high-dose-rate (HDR)
on The Nucletron Microselectron-HDR remote after loading
(New) Model 105.999 brachytherapy treatments
99-08 Urine Specimen Adulteration 4/1/99 All holders of operating licensees
for nuclear power reactors and
licensees authorized to possess
or use formula quantities of
strategic special nuclear material
OL = Operating License
CP = Construction Permit
IN 99-xx
April xx, 1999 Page 5of 5 This information notice requires no specific action or written response. If you have any
questions about the information in this notice, please contact the technical contact listed below, the appropriate regional office, or the appropriate office of Nuclear Reactor Regulation (NRR)
Project Manager.
Ledyard B. Marsh, Chief
Events Assessment, Generic Communications
And Non-Power Reactors Branch
Division of Regulatory Improvement Programs
Office of Nuclear Reactor Regulation
Technical contact: Chuck Petrone, NRR
301-415-1027 E-mail: cdRDanrc.aov
REFERENCES:
NRC Integrated Inspection Report No. 50-333198-08, issued February 10, 1999 (Accession No.
9902170348) for the James A. FitzPatrick Nuclear Power Plant for the period November 22,
1998, through January 10, 1999.
Attachments:
1. List of Recently Issued NMSS Information Notices
2. List of Recently Issued NRC Information Notices
DOCUMENT NAME: G:ICDPDRAININ\DRAIN.0B.WPD
To receive a copy of this document, Indicate In the box C=Copy w/o attachmentlenclosure E=Copy with attachment/enclosure N = No copy
OFFICE PECB:DRIP I Tech Editor l DRCH I PDIV-1 I
NAME CPetrone I_ RGallo 1 MNolangfarP.
DATE V/0199 F .
[3 /1/99 V. .
4/4I9
. .
1' /0g99 OFFICE PDI-1 IA .I PDIII-2 I C:PECB:DRIP I
NAME 2Jiiam RPulsjier LMarsh
DATE lf/499 I1'/t 99 I /99 OFFICIAL RECORD COPY
IN 99-14 May 5, 1999 This information notice requires no specific action or written response. If you have any
questions about the information in this notice, please contact the technical contact listed below, the appropriate regional office, or the appropriate Office of Nuclear Reactor Regulation (NRR)
project manager.
[arig sjid by]
Ledyard B. Marsh, Chief
Events Assessment, Generic Communications
And Non-Power Reactors Branch
Division of Regulatory Improvement Programs
Office of Nuclear Reactor Regulation
Technical contact: Chuck Petrone, NRR
301-415-1027 E-mail: cdr)ODnrc.gov
REFERENCES:
NRC Integrated Inspection Report No. 50-333/98-08, issued February 10, 1999 (Accession No.
9902170348) for the James A. FitzPatrick Nuclear Power Plant for the period November 22,
1998, through January 10, 1999.
Attachment: List of Recently Issued NRC Information Notices
DOCUMENT NAME: S:XDRPMSEC\99-14.IN
To receive a copy of this document. indicate in the box C=CoDv w/o attachment/enclosure E=CoDv with attachment/enclosure N = No coov
OFFICE PECB:DRlIP I Tech Editor l DRCH l-ii PDIV-1 lI
NAME CPetrone* BCalure* RGallo* MNolan*
DATE 04/27/99 .3/15/99 _________04128199 = 04/27/99
1 . . .
OFFICE PDI-1 I PD111-2 C:PECB:DJRIP I
NAME JWilliams* RPulsifer' _I-Marsh _ __ _
DATE 04/27/9 . 04/27/99 k,-u99 OFFICIAL RECORD COPY
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list | - Information Notice 1999-01, Deterioration of High-Efficiency Particulate Air Filters in a Pressurized Water Reactor Containment Fan Cooler Unit (20 January 1999)
- Information Notice 1999-02, Guidance to Users on the Implementation of a New Single-Source Dose-Calculation Formalism and Revised Air-Kerma Strength Standard for Iodine-125 Sealed Sources (21 January 1999, Topic: Brachytherapy)
- Information Notice 1999-03, Exothermic Reactors Involving Dried Uranium Oxide Powder (Yellowcake) (29 January 1999, Topic: Brachytherapy)
- Information Notice 1999-04, Unplanned Radiation Exposures to Radiographers, Resulting from Failures to Follow Proper Radiation Safety Procedures (1 March 1999, Topic: Brachytherapy)
- Information Notice 1999-05, Inadvertent Discharge of Carbon Dioxide Fire Protection System and Gas Migration (8 March 1999, Topic: Brachytherapy)
- Information Notice 1999-06, 1998 Enforcement Sanctions as a Result of Deliberate Violations of NRC Employee Protection Requirements (19 March 1999, Topic: Enforcement Discretion, Brachytherapy)
- Information Notice 1999-06, 1998 Enforcement Sanctions As a Result of Deliberate Violations of NRC Employee Protection Requirements (19 March 1999, Topic: Enforcement Discretion)
- Information Notice 1999-07, Failed Fire Protection Deluge Valves & Potential Testing Deficiencies in Preaction Sprinkler Systems (22 March 1999, Topic: Safe Shutdown)
- Information Notice 1999-08, Urine Specimen Adulteration (26 March 1999, Topic: Brachytherapy)
- Information Notice 1999-09, Problems Encountered When Manually Editing Treatment Data on the Nucletron Microselectron-HDR (New) Model 105-999 (24 March 1999, Topic: Brachytherapy)
- Information Notice 1999-10, Degradation of Prestressing Tendon Systems in Prestresssed Concrete Containments (13 April 1999)
- Information Notice 1999-11, Incidents Involving the Use of Radioactive Iodine-131 (16 April 1999, Topic: Brachytherapy)
- Information Notice 1999-12, Year 2000 Computer Systems Readiness Audits (28 April 1999, Topic: Brachytherapy)
- Information Notice 1999-13, Insights from NRC Inspections of Low-and Medium-Voltage Circuit Breaker Maintenance Programs (29 April 1999, Topic: Brachytherapy)
- Information Notice 1999-14, Unanticipated Reactor Water Draindown at Quad Cities Unit 2, Arkansas Nuclear One Unit 2, & FitzPatrick (5 May 1999, Topic: Reactor Vessel Water Level, Brachytherapy)
- Information Notice 1999-15, Misapplication for 10CFR Part 71 Transportation Shipping Cask Licensing Basis to 10CFR Part 50 Design Basis (27 May 1999, Topic: Brachytherapy)
- Information Notice 1999-16, Federal Bureau of Investigation'S Nuclear Site Security Program (28 May 1999, Topic: Brachytherapy)
- Information Notice 1999-17, Problems Associated with Post-Fire Safe-Shutdown Circuit Analyses (3 June 1999, Topic: Hot Short, Safe Shutdown, Temporary Modification, Emergency Lighting, Fire Protection Program)
- Information Notice 1999-18, Update on Nrc'S Year 2000 Activities for Material Licensees and Fuel Cycle Licensees and Certificate Holders (14 June 1999, Topic: Brachytherapy)
- Information Notice 1999-19, Rupture of the Shell Side of a Feedwater Heater at the Point Beach Nuclear Plant (23 June 1999)
- Information Notice 1999-20, Contingency Planning for the Year 2000 Computer Problem (25 June 1999, Topic: Brachytherapy)
- Information Notice 1999-21, Recent Plant Events Caused by Human Performance Errors (25 June 1999, Topic: Probabilistic Risk Assessment)
- Information Notice 1999-22, 10CFR 34.43(a)(1); Effective Date for Radiographer Certification and Plans for Enforcement Discretion (25 June 1999, Topic: Enforcement Discretion, Brachytherapy)
- Information Notice 1999-23, Safety Concerns Related to Repeated Control Unit Failures of the Nucletron Classic Model High-Dose-Rate Remote Afterloading Brachytherapy Devices (6 July 1999, Topic: Enforcement Discretion, Brachytherapy)
- Information Notice 1999-24, Broad-Scope Licensees' Responsibilities for Reviewing and Approving Unregistered Sealed Sources and Devices (12 July 1999, Topic: Enforcement Discretion, Brachytherapy)
- Information Notice 1999-25, Year 2000 Contingency Planning Activities (10 August 1999, Topic: Enforcement Discretion, Brachytherapy)
- Information Notice 1999-26, Safety and Economic Consequences of Misleading Marketing Information (24 August 1999, Topic: Enforcement Discretion, Brachytherapy)
- Information Notice 1999-27, Malfunction of Source Retraction Mechanism in Cobalt-60 Teletherapy Treatment Units (2 September 1999, Topic: Enforcement Discretion, Brachytherapy)
- Information Notice 1999-28, Recall of Star Brand Fire Protection Sprinkler Heads (30 September 1999, Topic: Enforcement Discretion, Brachytherapy)
- Information Notice 1999-30, Failure of Double Contingency Based on Administrative Controls Involving Laboratory Sampling and Spectroscopic Analysis of Wet Uranium Waste (8 November 1999, Topic: Enforcement Discretion, Brachytherapy)
- Information Notice 1999-30, Failure of Double Contingency Based On Administrative Controls Involving Laboratory Sampling and Spectroscopic Analysis of Wet Uranium Waste (8 November 1999, Topic: Enforcement Discretion, Brachytherapy)
- Information Notice 1999-31, Operational Controls to Guard Against Inadventent Nuclear Criticality (17 November 1999, Topic: Enforcement Discretion, Brachytherapy)
- Information Notice 1999-31, Operational Controls To Guard Against Inadventent Nuclear Criticality (17 November 1999, Topic: Enforcement Discretion, Brachytherapy)
- Information Notice 1999-32, Effect of Year 2000 Issue on Medical Licenseess (17 December 1999, Topic: Brachytherapy)
- Information Notice 1999-32, Effect of Year 2000 Issue on Medical Licensees (17 December 1999, Topic: Brachytherapy, Overdose, Underdose)
- Information Notice 1999-33, Management Of Wastess Contaminated with Radioactive Materialss (21 December 1999, Topic: Brachytherapy)
- Information Notice 1999-33, Management of Wastes Contaminated with Radioactive Materials (21 December 1999, Topic: Brachytherapy)
- Information Notice 1999-33, Management Of Wastes Contaminated with Radioactive Materials (21 December 1999, Topic: Brachytherapy)
- Information Notice 1999-33, Management Of Wastes Contaminated With Radioactive Materials (21 December 1999, Topic: Brachytherapy)
- Information Notice 1999-34, Potential Fire Hazard in the Use of Polyalphaolefin in Testing of Air Filter (28 December 1999)
- Information Notice 1999-34, PotentialPotentialPotential FireFireFire HazardHazardHazard ininIn thetheThe UseUseUse ofofOf PolyalphaolefinPolyalphaolefinPolyalphaolefin ininIn TestingTestingTesting ofofOf AirAirAir FilterFilterFilter (28 December 1999)
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