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| {{#Wiki_filter:UNITED STATESNUCLEAR REGULATORY COMMISSIONOFFICE OF NUCLEAR REACTOR REGULATIONWASHINGTON, D.C. 20555-0001May 5, 1999NRC INFORMATION NOTICE 99-14: UNANTICIPATED REACTOR WATER DRAINDOWNAT QUAD CITIES UNIT 2, ARKANSAS NUCLEAR ONEUNIT 2, AND FITZPATRICK | | {{#Wiki_filter: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 |
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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 | | 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 H ort<<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 WA RHR 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 |
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| | about 13 inches because of a misinterpretation |
| | |
| | of causes of the level decrease. |
| | |
| | After detecting |
| | |
| | the decrease, the operating |
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| | crew was slow to react, which allowed the level to decrease another 20 inches before the operators |
| | |
| | isolated shutdown cooling which terminated |
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| | the draindown. |
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| | The licensee estimated |
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| | that a total of 6000 to 7000 gallons was drained from the reactor to the suppression |
| | |
| | pool.Operations |
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| | staff practices |
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| | including |
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| | 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 |
| | |
| | ===Circuit Breaker Maintenance=== |
| | Programs for nuclear power reactors 99-12 Year 2000 Computer Systems Readiness |
| | |
| | Audits Incidents |
| | |
| | Involving |
| | |
| | the Use of Radioactive |
| | |
| | Iodine-131 |
| | 4/28/99 4/23/99 All holders of operating |
| | |
| | licenses or construction |
| | |
| | permits for nuclear power plants All medical use licensees 99-11 97-15, Sup 1 Reporting |
| | |
| | of Errors and 4/16/99 Changes in Large-Break/Small- Break Loss-of-Coolant |
| | |
| | Evaluation |
| | |
| | Models of Fuel Vendors and Compliance |
| | |
| | with 10 CFR 50.46(a)(3) |
| | All holders of operating |
| | |
| | licenses for nuclear power reactors, except those who have permanently |
| | |
| | cease operations |
| | |
| | and have certified |
| | |
| | that fuel has been permanently |
| | |
| | removed from the reactor 99-10 99-09 Degradation |
| | |
| | of Prestressing |
| | |
| | 4/13/99 Tendon Systems in Prestressed |
| | |
| | ===Concrete Containments=== |
| | Problems Encountered |
| | |
| | When 3/24/99 Manually Editing Treatment |
| | |
| | Data on The Nucletron |
| | |
| | Microselectron-HDR (New) Model 105.999 Urine Specimen Adulteration |
| | |
| | 4/1/99 All holders of operating |
| | |
| | licenses for nuclear power reactors All medical licensees |
| | |
| | authorized |
| | |
| | to conduct high-dose-rate (HDR)remote after loading brachytherapy |
| | |
| | treatments |
| | |
| | All holders of operating |
| | |
| | licensees for nuclear power reactors and licensees |
| | |
| | authorized |
| | |
| | to possess or use formula quantities |
| | |
| | of strategic |
| | |
| | special nuclear material 99-08 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 [3 /1/99 4 /4I9 1' /0g99 F .V. ...OFFICE PDI-1 IA .I PDIII-2 I C:PECB:DRIP |
| | |
| | I NAME 2 Jiiam 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 |
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| | REFERENCES: |
| | NRC Integrated |
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| | Inspection |
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| | Report No. 50-333/98-08, issued February 10, 1999 (Accession |
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| | 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 |
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| | *See previous concurrence |
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| 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
| | To receive a copy of this document. |
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| 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 events.
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| 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.Attachment: List of Recently Issued NRC Information Notices
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| ~~ 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
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| 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.Attachments: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
| | 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 |
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| IN 99-14May 5, 1999
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| | I-Marsh _ _ __ _DATE 04/27/9 .04/27/99 k,-u99 OFFICIAL RECORD COPY}} |
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Unanticipated Reactor Water Draindown at Quad Cities Unit 2, Arkansas Nuclear One Unit 2, & FitzPatrickML031040444 |
Person / Time |
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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 |
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Issue date: |
05/05/1999 |
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From: |
Marsh L B Division of Regulatory Improvement Programs |
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To: |
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References |
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IN-99-014, NUDOCS 9905070080 |
Download: ML031040444 (8) |
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Category:NRC Information Notice
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Underestimate of Dam Failure Frequency Used in Probabilistic Risk Assessments ML1007804482009-11-23023 November 2009 Email from Peter Bamford, NRR to Pamela Cowan, Exelon on TMI Contamination Control Event Information Notice 2009-11, NSP000059-Revised Prefiled Testimony of Northard/Petersen/Peterson-NRC Information Notice 2009-112009-07-0707 July 2009 NSP000059-Revised Prefiled Testimony of Northard/Petersen/Peterson-NRC Information Notice 2009-11 Information Notice 2009-10, Official Exhibit - NYS000019-00-BD01- NRC Information Notice 2009-10, Transformers Failures - Recent Operating Experience (Jul. 7, 2009) (NRC in 2009-10)2009-07-0707 July 2009 Official Exhibit - NYS000019-00-BD01- NRC Information Notice 2009-10, Transformers Failures - Recent Operating Experience (Jul. 7, 2009) (NRC in 2009-10) Information Notice 2009-09, Improper Flow Controller Settings Renders Injection Systems Inoperable and Surveillance Did Not Identify2009-06-19019 June 2009 Improper Flow Controller Settings Renders Injection Systems Inoperable and Surveillance Did Not Identify Information Notice 2008-12, Reactor Trip Due to Off-Site Power Fluctuation2008-07-0707 July 2008 Reactor Trip Due to Off-Site Power Fluctuation Information Notice 2008-11, Service Water System Degradation at Brunswicksteam Electric Plant Unit 12008-06-18018 June 2008 Service Water System Degradation at Brunswicksteam Electric Plant Unit 1 Information Notice 2008-04, Counterfeit Parts Supplied to Nuclear Power Plants2008-04-0707 April 2008 Counterfeit Parts Supplied to Nuclear Power Plants Information Notice 1991-09, Counterfeiting of Crane Valves2007-09-25025 September 2007 Counterfeiting of Crane Valves Information Notice 2007-28, Potential Common Cause Vulnerabilities in Essential Service Water Systems Due to Inadequate Chemistry Controls2007-09-19019 September 2007 Potential Common Cause Vulnerabilities in Essential Service Water Systems Due to Inadequate Chemistry Controls Information Notice 2007-29, Temporary Scaffolding Affects Operability of Safety-Related Equipment2007-09-17017 September 2007 Temporary Scaffolding Affects Operability of Safety-Related Equipment Information Notice 2007-14, Loss of Offsite Power and Dual-Unit Trip at Catawba Nuclear Generating Station2007-03-30030 March 2007 Loss of Offsite Power and Dual-Unit Trip at Catawba Nuclear Generating Station Information Notice 2007-06, Potential Common Cause Vulnerabilities in Essential Service Water Systems2007-02-0909 February 2007 Potential Common Cause Vulnerabilities in Essential Service Water Systems Information Notice 2007-05, Vertical Deep Draft Pump Shaft and Coupling Failures2007-02-0909 February 2007 Vertical Deep Draft Pump Shaft and Coupling Failures Information Notice 2006-31, Inadequate Fault Interrupting Rating of Breakers2006-12-26026 December 2006 Inadequate Fault Interrupting Rating of Breakers Information Notice 2006-29, Potential Common Cause Failure of Motor-operated Valves as a Result of Stem Nut Wear2006-12-14014 December 2006 Potential Common Cause Failure of Motor-operated Valves as a Result of Stem Nut Wear Information Notice 2006-29, Potential Common Cause Failure of Motor-operated Valves As a Result of Stem Nut Wear2006-12-14014 December 2006 Potential Common Cause Failure of Motor-operated Valves As a Result of Stem Nut Wear Information Notice 2006-13, E-mail from M. 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 (7008 PD(L H ort<<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 WA RHR 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
Circuit Breaker Maintenance
Programs for nuclear power reactors 99-12 Year 2000 Computer Systems Readiness
Audits Incidents
Involving
the Use of Radioactive
Iodine-131
4/28/99 4/23/99 All holders of operating
licenses or construction
permits for nuclear power plants All medical use licensees 99-11 97-15, Sup 1 Reporting
of Errors and 4/16/99 Changes in Large-Break/Small- Break Loss-of-Coolant
Evaluation
Models of Fuel Vendors and Compliance
with 10 CFR 50.46(a)(3)
All holders of operating
licenses for nuclear power reactors, except those who have permanently
cease operations
and have certified
that fuel has been permanently
removed from the reactor 99-10 99-09 Degradation
of Prestressing
4/13/99 Tendon Systems in Prestressed
Concrete Containments
Problems Encountered
When 3/24/99 Manually Editing Treatment
Data on The Nucletron
Microselectron-HDR (New) Model 105.999 Urine Specimen Adulteration
4/1/99 All holders of operating
licenses for nuclear power reactors All medical licensees
authorized
to conduct high-dose-rate (HDR)remote after loading brachytherapy
treatments
All holders of operating
licensees for nuclear power reactors and licensees
authorized
to possess or use formula quantities
of strategic
special nuclear material 99-08 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 [3 /1/99 4 /4I9 1' /0g99 F .V. ...OFFICE PDI-1 IA .I PDIII-2 I C:PECB:DRIP
I NAME 2 Jiiam 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
|
---|
|
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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