|
|
Line 14: |
Line 14: |
| | page count = 8 | | | page count = 8 |
| }} | | }} |
| {{#Wiki_filter:UNITED STATES NUCLEAR REGULATORY | | {{#Wiki_filter:UNITED STATES |
|
| |
|
| COMMISSION | | NUCLEAR REGULATORY COMMISSION |
|
| |
|
| ===OFFICE OF NUCLEAR REACTOR REGULATION===
| | OFFICE OF NUCLEAR REACTOR REGULATION |
| WASHINGTON, D.C. 20555-0001 May 5, 1999 NRC INFORMATION
| |
|
| |
|
| NOTICE 99-14: UNANTICIPATED | | WASHINGTON, D.C. 20555-0001 May 5, 1999 NRC INFORMATION NOTICE 99-14: UNANTICIPATED REACTOR WATER DRAINDOWN |
|
| |
|
| REACTOR WATER DRAINDOWN AT QUAD CITIES UNIT 2, ARKANSAS NUCLEAR ONE UNIT 2, AND FITZPATRICK
| | AT QUAD CITIES UNIT 2, ARKANSAS NUCLEAR ONE |
| | |
| | UNIT 2, AND FITZPATRICK |
|
| |
|
| ==Addressees== | | ==Addressees== |
Line 29: |
Line 30: |
|
| |
|
| ==Purpose== | | ==Purpose== |
| The U.S. Nuclear Regulatory | | The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice to alert |
| | |
| 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.
| | addressees to the potential for personnel errors during infrequently performed evolutions that |
|
| |
|
| It is expected that recipients
| | result in, or contribute to, events such as the inadvertent draining of water from the reactor |
|
| |
|
| will review the information | | vessel during shutdown operations. It is expected that recipients will review the information for |
|
| |
|
| for applicability
| | applicability to their facilities and consider actions, as appropriate, to prevent a similar |
|
| |
|
| to their facilities
| | occurrence. However, suggestions contained in this information notice are not NRC |
|
| |
|
| and consider actions, as appropriate, to prevent a similar occurrence.
| | requirements; therefore, no specific action or written response to this notice is required. |
|
| |
|
| However, suggestions
| | DescriDtion of Circumstances |
| | |
| 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 | | 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 | | at 131 'F and reactor water level at 80 inches indicated level (normal level during operations is |
| | |
| level (normal level during operations | |
| | |
| is 30 inches indicated | |
|
| |
|
| or 173 inches above the top of active fuel [TAF]). Core cooling was being maintained | | 30 inches indicated or 173 inches above the top of active fuel [TAF]). Core cooling was being |
|
| |
|
| 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 | | maintained in a band of 120 'F to 170 OF by the OA" loop of the shutdown cooling mode of the |
|
| |
|
| failed to close the OA RHR minimum flow valve as required by the procedure.
| | residual heat removal (RHR) system after being switched from the nB" loop at 12:32 a.m. |
|
| |
|
| Sometime later operators
| | During the switch over the licensee inadvertently failed to close the OA RHR minimum flow |
|
| |
|
| noted a decreasing | | valve as required by the procedure. Sometime later operators noted a decreasing reactor water |
|
| |
|
| 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
| | level and at about 1:02 a.m. secured the *2A RHR pump and isolated shutdown cooling. At |
|
| |
|
| restored the *2A' loop of shutdown cooling to the proper lineup and started the *2A RHR pump. Water level had decreased | | 1:55 a.m. operators restored the *2A' loop of shutdown cooling to the proper lineup and started |
|
| |
|
| to a minimum of about 45 inches indicated, and reactor water temperature | | 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 |
|
| |
|
| had risen to a maximum of about 163 OF. Forced circulation
| | reactor vessel water using a reactor recirculation pump remained in effect throughout the event. |
|
| |
|
| of reactor vessel water using a reactor recirculation | | On the basis of post event reviews, It appears that the minimum flow valve in the OA loop was |
|
| |
|
| pump remained in effect throughout
| | left open because the nuclear station operator failed to ensure that the tasks were performed in |
|
| |
|
| 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 | | the sequence specified in the operating procedures. The nuclear station operator who was |
|
| |
|
| in the sequence specified
| | (7008 PD(L Hort<<4qj-Oiif qqos(J5 C7Ffcj |
|
| |
|
| in the operating
| | ANW\\b |
|
| |
|
| procedures.
| | 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 |
|
| |
|
| The nuclear station operator who was (7008 PD(L H ort<<4qj-Oiif
| | to the required closed position. De-energizing the breaker also removed power to the valve |
|
| |
|
| qqos(J5 C7Ffcj ANW\\b
| | position indicator lights in the control room. Thus, when the nuclear station operator tried to |
|
| |
|
| IN 99-14 May 5, 1999 directing
| | verify that the valve was closed, there was no position indication in the control room to make |
|
| |
|
| the evolution | | that verification. The nuclear station operator made the incorrect assumption that the valve was |
|
| |
|
| from the control room gave the non-licensed
| | already closed and moved to the next step in the procedure. This failure to close the WAX RHR |
|
| |
|
| operator permission
| | minimum flow valve opened a drain path from the reactor to the suppression pool. To further |
|
| |
|
| to de-energize the breaker for the WA RHR minimum flow valve operator before the valve was taken to the required closed position.
| | complicate the event, the operating crew did not recognize that there was any problem until |
|
| |
|
| De-energizing
| | approximately 10 minutes had passed and the water level had decreased about 13 inches |
|
| |
|
| the breaker also removed power to the valve position indicator | | 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 |
|
| |
|
| lights in the control room. Thus, when the nuclear station operator tried to verify that the valve was closed, there was no position indication
| | before the operators isolated shutdown cooling which terminated the draindown. The licensee |
|
| |
|
| in the control room to make that verification.
| | estimated that a total of 6000 to 7000 gallons was drained from the reactor to the suppression |
|
| |
|
| The nuclear station operator made the incorrect
| | pool. |
|
| |
|
| assumption
| | Operations staff practices including poor communications, poor activity briefings for high-risk |
|
| |
|
| that the valve was already closed and moved to the next step in the procedure.
| | activities, lack of effective pre-shift briefings, inadequate supervision of important control room |
|
| |
|
| This failure to close the WAX RHR minimum flow valve opened a drain path from the reactor to the suppression
| | activities, inadequate monitoring of control room panels, and slow event response may have |
|
| |
|
| pool. To further complicate | | contributed to the event. Although the unintended loss of inventory to the suppression pool |
|
| |
|
| the event, the operating
| | highlighted significant weaknesses in plant operations, the safety significance was minimized by |
|
| |
|
| crew did not recognize
| | two features. First, a reactor recirculation pump remained in service throughout the event |
|
| |
|
| that there was any problem until approximately
| | which served to distribute decay heat. Second, an automatic isolation of shutdown cooling |
|
| |
|
| 10 minutes had passed and the water level had decreased
| | would have occurred at 8 inches indicated level which would have stopped the draining event. |
|
| |
|
| about 13 inches because of a misinterpretation
| | An indicated water level of 8 inches corresponds to approximately 151 inches of water level |
|
| |
|
| of causes of the level decrease.
| | above the TAF in the reactor core. |
|
| |
|
| After detecting
| | ===Arkansas Nuclear One Unit 2=== |
| | On February 2, 1999, at Arkansas Nuclear One Unit 2, the operators were draining the |
|
| |
|
| the decrease, the operating | | refueling canal in preparation for installing the reactor vessel head. Refueling was complete |
|
| |
|
| crew was slow to react, which allowed the level to decrease another 20 inches before the operators
| | and steam generator nozzle dams were installed. The operators were using the two low |
|
| |
|
| isolated shutdown cooling which terminated
| | 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 |
|
| |
|
| the draindown. | | approximately 3.3 Inches per minute. When the water level reached 105 inches, the reactor |
|
| |
|
| The licensee estimated
| | operator noted that level started to lower rapidly. Operators stopped one of the LPSI pumps |
|
| |
|
| that a total of 6000 to 7000 gallons was drained from the reactor to the suppression
| | and instructed a local operator to close the isolation valve to the refueling water tank. This |
|
| |
|
| pool.Operations
| | manually operated valve required 55 turns of the handwheel to fully close. Within |
|
| |
|
| staff practices
| | approximately 1.5 minutes, the reactor vessel level had dropped below the 65 inch level (where |
|
| |
|
| including
| | reduced inventory begins) and continued down to 56 inches before the valve could be fully |
|
| |
|
| poor communications, poor activity briefings
| | closed. (Reference zero on these level instruments is the bottom of the hot leg, with mid-loop |
|
| |
|
| for high-risk activities, lack of effective
| | being defined at approximately 24 inches.) The average rate of level decrease between 105 |
|
| |
|
| pre-shift
| | IN 99-14 May 5, 1999 inches and 56 inches was approximately 33 inches per minute. At its lowest level, 56 inches |
|
| |
|
| briefings, inadequate
| | indicated, there were still 93 inches of water above the TAF. Using the high pressure safety |
|
| |
|
| supervision
| | injection (HPSI) pump the operators brought the level back up to 90 inches. The plant was in |
|
| |
|
| of important
| | reduced inventory operations (below 65 inches) for approximately 7 minutes. During the event |
|
| |
|
| control room activities, inadequate
| | the level remained well above the point where LPSI pump cavitation would be expected. The |
|
| |
|
| monitoring
| | licensee concluded that the safety significance of the event was minimal because multiple |
|
| |
|
| of control room panels, and slow event response may have contributed | | sources of makeup water were available, redundant mitigation equipment was available, and |
|
| |
|
| to the event. Although the unintended | | the operators were quick to recognize and respond to the event. |
|
| |
|
| loss of inventory
| | On the basis of post event reviews, it was determined that the procedure used for draining |
|
| |
|
| to the suppression
| | down the refueling canal was inadequate in that it incorrectly stated that the draindown should |
|
| |
|
| pool highlighted
| | be secured at the 90-inch level. The procedure should have directed that the rate of draining |
|
| |
|
| significant
| | be secured at the 106-inch level so that appropriate precautions could be taken before |
|
| |
|
| weaknesses
| | resuming the draindown. These precautions should have Included reminders to the operating |
|
| |
|
| in plant operations, the safety significance
| | crew that below the 106-inch level the level will drop much more quickly due to the transition of |
|
| |
|
| was minimized
| | pumping from a large volume in the refueling canal to a small volume In the reactor vessel. |
|
| |
|
| by two features.
| | Therefore, in order to maintain control of the water level, the draindown rate should be |
|
| |
|
| First, a reactor recirculation
| | decreased and an operator should be stationed to directly monitor the level. |
|
| |
|
| pump remained in service throughout
| | Additional factors that contributed to this event include: the operators received little specific |
|
| |
|
| the event which served to distribute | | training on this evolution; the crew was inexperienced in performing this task; the task should |
|
| |
|
| decay heat. Second, an automatic
| | have been classified as an infrequent task requiring a more thorough briefing; and, operators |
|
| |
|
| isolation
| | failed to station an operator in a position where he could directly monitor the water level in the |
|
| |
|
| of shutdown cooling would have occurred at 8 inches indicated
| | refueling canal. Instead they monitored it remotely using a video camera that did not provide a |
|
| |
|
| level which would have stopped the draining event.An indicated | | clear picture of the water level. |
|
| |
|
| water level of 8 inches corresponds
| | FitzPatrick |
|
| |
|
| to approximately
| | On December 2, 1998, at the James A. FitzPatrick Nuclear Power Plant, the operators were in |
|
| |
|
| 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
| | the process of reassembling the reactor following refueling. Operators were controlling the |
|
| |
|
| were draining the refueling
| | reactor vessel water level at 357 inches above TAF by adjusting the water discharge rate to |
|
| |
|
| canal in preparation
| | compensate for the constant input from the control rod drive cooling water system. While in this |
|
| |
|
| for installing
| | condition, the licensees risk analysis requires that reactor vessel water level be monitored using |
|
| |
|
| the reactor vessel head. Refueling | | two independent level indicators. To meet this requirement, the licensee designated a wide |
|
| |
|
| was complete and steam generator
| | range indicator which provided Indication up to the top of the reactor vessel and an RHR |
|
| |
|
| nozzle dams were installed.
| | interlock level indicator which provided indication in the range from -150 inches to +200 Inches |
|
| |
|
| The operators
| | as the instruments to be used during this evaluation. |
|
| |
|
| were using the two low pressure safety injection (LPSI) pumps to drain the canal to the refueling
| | In order for the wide-range level Indicator to remain available with the reactor head removed, a |
|
| |
|
| water storage tank;one pump also served as the shutdown cooling pump. The rate of draindown
| | temporary standpipe and fill funnel were used to replace a portion of the reference leg. At the |
|
| |
|
| was approximately | | time of the event, the licensee was in the process of removing this temporary standpipe and |
|
| |
|
| 3.3 Inches per minute. When the water level reached 105 inches, the reactor operator noted that level started to lower rapidly. Operators
| | reinstalling the original reference leg components. As the water drained from the standpipe, it |
|
| |
|
| stopped one of the LPSI pumps and instructed
| | caused the wide-range level indicator to erroneously show an increasing water level. For a |
|
| |
|
| a local operator to close the isolation
| | period of approximately one hour the operators in the control room, unaware that the ongoing |
|
| |
|
| valve to the refueling
| | maintenance would cause an error in the indicated water level, compensated for the apparent |
|
| |
|
| water tank. This manually operated valve required 55 turns of the handwheel
| | increasing level by increasing the discharge rate. This action had the effect of reducing the |
|
| |
|
| to fully close. Within approximately | | IN 99-14 May 5, 1999 actual water level from 357 inches to 255 inches. During the same time period, the operators |
|
| |
|
| 1.5 minutes, the reactor vessel level had dropped below the 65 inch level (where reduced inventory
| | were also in the process of filling and venting the reactor feedwater piping, which could have |
|
| |
|
| begins) and continued
| | affected the reactor water level. Once the normal reference leg piping had been reinstalled and |
|
| |
|
| down to 56 inches before the valve could be fully closed. (Reference
| | the reference leg began to refill, the indicated level decreased from 357 inches to the actual |
|
| |
|
| zero on these level instruments
| | level of 255 inches. The second level instrument, which does not come on-scale until the level |
|
| |
|
| is the bottom of the hot leg, with mid-loop being defined at approximately
| | goes below 200 inches, remained off-scale high. |
|
| |
|
| 24 inches.) The average rate of level decrease between 105 IN 99-14 May 5, 1999 inches and 56 inches was approximately
| | When operators discovered the level discrepancy, they used a temporary pressure gauge |
|
| |
|
| 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
| | connected to the reactor vessel low-point tap to confirm the actual water level. After confirming |
|
| |
|
| brought the level back up to 90 inches. The plant was in reduced inventory
| | 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 |
|
| |
|
| operations (below 65 inches) for approximately
| | determined that the safety significance of this event was low since the reactor was in cold |
|
| |
|
| 7 minutes. During the event the level remained well above the point where LPSI pump cavitation
| | shutdown with low decay heat and the reactor water level remained well above the TAF. In |
|
| |
|
| would be expected. | | addition, the drain-down would have been limited by an automatic Isolation of the draindown |
|
| |
|
| The licensee concluded
| | path, which would have occurred prior to vessel level reaching 177 Inches above the TAF. |
|
| |
|
| that the safety significance
| | The licensee's post event review identified: weaknesses in the operator's knowledge of the |
|
| |
|
| of the event was minimal because multiple sources of makeup water were available, redundant | | 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 |
|
| |
|
| mitigation
| | designated reactor water level indicators were available, only one indicator, the wide-range |
|
| |
|
| equipment
| | 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 |
|
| |
|
| was available, and the operators | | unavailable. The second instrument was pegged off-scale high and remained that way |
|
| |
|
| were quick to recognize
| | throughout the event because the level never dropped below 200 inches. A post event review by |
|
| |
|
| and respond to the event.On the basis of post event reviews, it was determined
| | the licensee indicated that other reactor water level instruments, remained operable during the |
|
| |
|
| that the procedure
| | event but, apparently the operators did not rely on these other instruments or notice the |
|
| |
|
| used for draining down the refueling
| | discrepancy between them and the wide range Indicator. Proposed corrective actions included |
|
| |
|
| canal was inadequate
| | procedural enhancements to ensure that reactor level instrumentation credited by the outage |
|
| |
|
| in that it incorrectly
| | risk assessment remains available during reactor disassembly and reassembly. |
| | |
| 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 | | Discussion |
|
| |
|
| Personnel | | Personnel errors appear to have caused, or contributed to, these three inadvertent reactor |
| | |
| 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
| | vessel draindown events. The likelihood of personnel errors is dependent upon the operators |
|
| |
|
| activities
| | knowledge of the task gained through previous experience and training. It is also dependent |
|
| |
|
| might have prevented
| | upon the quality of the procedures used to perform the task, the level of supervision, the |
|
| |
|
| these events.
| | adequacy of pre-job briefings, fatigue, and distractions resulting from multiple tasks. In each of |
|
| |
|
| IN 99-14 May 5, 1999 This information
| | the events, the plant staff made errors during a seldom-performed evolution. Because it was a |
|
| |
|
| notice requires no specific action or written response.
| | seldom-performed evolution, more training, better pre-job briefings, closer supervision, and |
|
| |
|
| If you have any questions
| | procedures that contain more details than those for frequently performed activities might have |
|
| |
|
| about the information
| | prevented these events. |
|
| |
|
| in this notice, please contact the technical
| | IN 99-14 May 5, 1999 This information notice requires no specific action or written response. If you have any |
|
| |
|
| contact listed below, the appropriate | | 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. |
|
| |
|
| regional office, or the appropriate
| | Ledyard B. Marsh, Chief |
|
| |
|
| Office of Nuclear Reactor Regulation (NRR)project manager.Ledyard B. Marsh, Chief Events Assessment, Generic Communications
| | Events Assessment, Generic Communications |
|
| |
|
| And Non-Power | | And Non-Power Reactors Branch |
|
| |
|
| Reactors Branch Division of Regulatory
| | Division of Regulatory Improvement Programs |
|
| |
|
| Improvement
| | Office of Nuclear Reactor Regulation |
|
| |
|
| ===Programs Office of Nuclear Reactor Regulation===
| | Technical contact: Chuck Petrone, NRR |
| Technical | |
|
| |
|
| contact: Chuck Petrone, NRR 301-415-1027 E-mail: cdDRenrc.aov
| | 301-415-1027 E-mail: cdDRenrc.aov |
|
| |
|
| REFERENCES: | | REFERENCES: |
| NRC Integrated | | NRC Integrated Inspection Report No. 50-333/98-08, issued February 10, 1999 (Accession No. |
| | |
| 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 | | 9902170348) for the James A. FitzPatrick Nuclear Power Plant for the period November 22, |
| | 1998, through January 10, 1999. |
|
| |
|
| Iodine-131
| | Attachment: List of Recently Issued NRC Information Notices |
| 4/28/99 4/23/99 All holders of operating
| |
|
| |
|
| licenses or construction
| | ~~ Attachment 1 IN 99-14 May 5, 1999 Page 1 of I |
|
| |
|
| permits for nuclear power plants All medical use licensees 99-11 97-15, Sup 1 Reporting
| | LIST OF RECENTLY ISSUED |
|
| |
|
| of Errors and 4/16/99 Changes in Large-Break/Small- Break Loss-of-Coolant
| | NRC INFORMATION NOTICES |
|
| |
|
| Evaluation
| | Information Date of |
|
| |
|
| Models of Fuel Vendors and Compliance
| | Notice No. Subject Issuance Issued to |
|
| |
|
| with 10 CFR 50.46(a)(3)
| | 99-13 Insiahts from NRR Inspections 4129199 All holders of operatina licenses |
| All holders of operating | |
|
| |
|
| licenses for nuclear power reactors, except those who have permanently
| | of Low-and Medium-Voltage for nuclear power reactors |
|
| |
|
| cease operations
| | Circuit Breaker Maintenance |
|
| |
|
| and have certified
| | Programs |
|
| |
|
| that fuel has been permanently
| | 99-12 Year 2000 Computer Systems 4/28/99 All holders of operating licenses |
|
| |
|
| removed from the reactor 99-10 99-09 Degradation
| | Readiness Audits or construction permits for nuclear |
|
| |
|
| of Prestressing
| | power plants |
|
| |
|
| 4/13/99 Tendon Systems in Prestressed | | 99-11 Incidents Involving the Use of 4/23/99 All medical use licensees |
|
| |
|
| ===Concrete Containments===
| | Radioactive Iodine-131 |
| Problems Encountered
| | 97-15, Sup 1 Reporting of Errors and 4/16/99 All holders of operating licenses |
|
| |
|
| When 3/24/99 Manually Editing Treatment
| | Changes in Large-Break/Small- for nuclear power reactors, except |
|
| |
|
| Data on The Nucletron
| | Break Loss-of-Coolant Evaluation those who have permanently |
|
| |
|
| Microselectron-HDR (New) Model 105.999 Urine Specimen Adulteration
| | Models of Fuel Vendors and cease operations and have |
|
| |
|
| 4/1/99 All holders of operating
| | Compliance with 10 CFR 50.46(a)(3) certified that fuel has been |
|
| |
|
| licenses for nuclear power reactors All medical licensees
| | permanently removed from the |
|
| |
|
| authorized
| | reactor |
|
| |
|
| to conduct high-dose-rate (HDR)remote after loading brachytherapy
| | 99-10 Degradation of Prestressing 4/13/99 All holders of operating licenses |
|
| |
|
| treatments
| | Tendon Systems in Prestressed for nuclear power reactors |
|
| |
|
| All holders of operating
| | Concrete Containments |
|
| |
|
| licensees for nuclear power reactors and licensees | | 99-09 Problems Encountered When 3/24/99 All medical licensees authorized |
|
| |
|
| authorized
| | Manually Editing Treatment Data to conduct high-dose-rate (HDR) |
| | on The Nucletron Microselectron-HDR remote after loading |
|
| |
|
| to possess or use formula quantities
| | (New) Model 105.999 brachytherapy treatments |
|
| |
|
| of strategic | | 99-08 Urine Specimen Adulteration 4/1/99 All holders of operating licensees |
|
| |
|
| special nuclear material 99-08 OL = Operating
| | for nuclear power reactors and |
|
| |
|
| License CP = Construction
| | licensees authorized to possess |
|
| |
|
| Permit
| | or use formula quantities of |
|
| |
|
| IN 99-xx April xx, 1999 Page 5of 5 This information
| | strategic special nuclear material |
|
| |
|
| notice requires no specific action or written response.
| | OL = Operating License |
|
| |
|
| If you have any questions
| | CP = Construction Permit |
|
| |
|
| about the information
| | IN 99-xx |
|
| |
|
| in this notice, please contact the technical
| | April xx, 1999 Page 5of 5 This information notice requires no specific action or written response. If you have any |
|
| |
|
| contact listed below, the appropriate | | 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. |
|
| |
|
| regional office, or the appropriate
| | Ledyard B. Marsh, Chief |
|
| |
|
| office of Nuclear Reactor Regulation (NRR)Project Manager.Ledyard B. Marsh, Chief Events Assessment, Generic Communications
| | Events Assessment, Generic Communications |
|
| |
|
| And Non-Power | | And Non-Power Reactors Branch |
|
| |
|
| Reactors Branch Division of Regulatory
| | Division of Regulatory Improvement Programs |
|
| |
|
| Improvement
| | Office of Nuclear Reactor Regulation |
|
| |
|
| ===Programs Office of Nuclear Reactor Regulation===
| | Technical contact: Chuck Petrone, NRR |
| Technical | |
|
| |
|
| contact: Chuck Petrone, NRR 301-415-1027 E-mail: cdRDanrc.aov
| | 301-415-1027 E-mail: cdRDanrc.aov |
|
| |
|
| REFERENCES: | | REFERENCES: |
| NRC Integrated | | NRC Integrated Inspection Report No. 50-333198-08, issued February 10, 1999 (Accession No. |
|
| |
|
| Inspection
| | 9902170348) for the James A. FitzPatrick Nuclear Power Plant for the period November 22, |
| | 1998, through January 10, 1999. |
|
| |
|
| Report No. 50-333198-08, issued February 10, 1999 (Accession
| | Attachments: |
| | 1. List of Recently Issued NMSS Information Notices |
|
| |
|
| No.9902170348)
| | 2. List of Recently Issued NRC Information Notices |
| for the James A. FitzPatrick
| |
|
| |
|
| Nuclear Power Plant for the period November 22, 1998, through January 10, 1999.Attachments:
| | DOCUMENT NAME: G:ICDPDRAININ\DRAIN.0B.WPD |
| 1. List of Recently Issued NMSS Information
| |
|
| |
|
| Notices 2. List of Recently Issued NRC Information
| | To receive a copy of this document, Indicate In the box C=Copy w/o attachmentlenclosure E=Copy with attachment/enclosure N = No copy |
|
| |
|
| Notices DOCUMENT NAME: G:ICDPDRAININ\DRAIN.0B.WPD
| | OFFICE PECB:DRIP I Tech Editor l DRCH I PDIV-1 I |
|
| |
|
| To receive a copy of this document, Indicate In the box C=Copy w/o attachmentlenclosure
| | NAME CPetrone I_ RGallo 1 MNolangfarP. |
|
| |
|
| E=Copy with attachment/enclosure
| | DATE V/0199 F . |
|
| |
|
| N = No copy OFFICE PECB:DRIP
| | [3 /1/99 V. . |
|
| |
|
| I Tech Editor l DRCH I PDIV-1 I NAME CPetrone I_ RGallo 1 MNolangfarP.
| | 4/4I9 |
| | . . |
|
| |
|
| DATE V /0199 [3 /1/99 4 /4I9 1' /0g99 F .V. ...OFFICE PDI-1 IA .I PDIII-2 I C:PECB:DRIP
| | 1' /0g99 OFFICE PDI-1 IA .I PDIII-2 I C:PECB:DRIP I |
|
| |
|
| I NAME 2 Jiiam RPulsjier
| | NAME 2Jiiam RPulsjier LMarsh |
|
| |
|
| LMarsh DATE lf/499 I1'/t 99 I /99 OFFICIAL RECORD COPY
| | DATE lf/499 I1'/t 99 I /99 OFFICIAL RECORD COPY |
|
| |
|
| IN 99-14 May 5, 1999 This information | | IN 99-14 May 5, 1999 This information notice requires no specific action or written response. If you have any |
|
| |
|
| notice requires no specific action or written response. | | 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. |
|
| |
|
| If you have any questions
| | [arig sjid by] |
| | Ledyard B. Marsh, Chief |
|
| |
|
| about the information
| | Events Assessment, Generic Communications |
|
| |
|
| in this notice, please contact the technical
| | And Non-Power Reactors Branch |
|
| |
|
| contact listed below, the appropriate
| | Division of Regulatory Improvement Programs |
|
| |
|
| regional office, or the appropriate
| | Office of Nuclear Reactor Regulation |
|
| |
|
| Office of Nuclear Reactor Regulation (NRR)project manager.[arig sjid by]Ledyard B. Marsh, Chief Events Assessment, Generic Communications
| | Technical contact: Chuck Petrone, NRR |
|
| |
|
| And Non-Power
| | 301-415-1027 E-mail: cdr)ODnrc.gov |
| | |
| 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: | | REFERENCES: |
| NRC Integrated | | NRC Integrated Inspection Report No. 50-333/98-08, issued February 10, 1999 (Accession No. |
|
| |
|
| Inspection
| | 9902170348) for the James A. FitzPatrick Nuclear Power Plant for the period November 22, |
| | 1998, through January 10, 1999. |
|
| |
|
| Report No. 50-333/98-08, issued February 10, 1999 (Accession
| | Attachment: List of Recently Issued NRC Information Notices |
|
| |
|
| No.9902170348)
| | DOCUMENT NAME: S:XDRPMSEC\99-14.IN |
| 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 | | *See previous concurrence |
|
| |
|
| To receive a copy of this document. | | 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 |
| | |
| indicate in the box C=CoDv w/o attachment/enclosure | |
|
| |
|
| E=CoDv with attachment/enclosure
| | OFFICE PECB:DRlIP I Tech Editor l DRCH l-ii PDIV-1 lI |
|
| |
|
| N = No coov OFFICE PECB:DRlIP
| | NAME CPetrone* BCalure* RGallo* MNolan* |
| | DATE 04/27/99 .3/15/99 _________04128199 = 04/27/99 |
| | 1 . . . |
|
| |
|
| I Tech Editor l DRCH l-ii PDIV-1 lI NAME CPetrone*
| | OFFICE PDI-1 I PD111-2 C:PECB:DJRIP I |
| 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*
| | NAME JWilliams* RPulsifer' _I-Marsh _ __ _ |
| RPulsifer' | | DATE 04/27/9 . 04/27/99 k,-u99 OFFICIAL RECORD COPY}} |
| I-Marsh _ _ __ _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 |
---|
Category:NRC Information Notice
MONTHYEARInformation Notice 2020-02, Flex Diesel Generator Operational Challenges2020-09-15015 September 2020 Flex Diesel Generator Operational Challenges ML20225A0322020-09-0303 September 2020 NRC Choice Letter to NAC International with Attached Safety Inspection Report, IR 0721015/2020201, February 24-27, 2020 and July 22, 2020, Inspection of NAC International in Norcross, Georgia Information Notice 2012-09, PWROG-16043-NP-A, Revision 2, PWROG Program to Address NRC Information Notice 2012-09: Irradiation Effects on Fuel Assembly Spacer Grid Crush Strength for Westinghouse and CE PWR Fuel Designs.2019-11-30030 November 2019 PWROG-16043-NP-A, Revision 2, PWROG Program to Address NRC Information Notice 2012-09: Irradiation Effects on Fuel Assembly Spacer Grid Crush Strength for Westinghouse and CE PWR Fuel Designs. Information Notice 2011-20, NRC060 - NRC Information Notice 2011-20: Concrete Degradation by Alkali-Silica Reaction (Nov. 18, 2011)2019-07-24024 July 2019 NRC060 - NRC Information Notice 2011-20: Concrete Degradation by Alkali-Silica Reaction (Nov. 18, 2011) ML19196A2452019-07-15015 July 2019 Public Notice - Sequoyah Nuclear Plant, Unit 2 - Exigent Amendment to Facility Operating License Information Notice 2019-01, Inadequate Evaluation of Temporary Alterations2019-03-12012 March 2019 Inadequate Evaluation of Temporary Alterations ML16028A3082016-04-27027 April 2016 NRC Information Notice; IN 2016-05: Operating Experience Regarding Complications From a Loss of Instrumentation Air Information Notice 2015-05, Inoperability of Auxiliary and Emergency Feedwater Auto Start Circuits on Loss of Main Feedwater Pumps2015-05-12012 May 2015 Inoperability of Auxiliary and Emergency Feedwater Auto Start Circuits on Loss of Main Feedwater Pumps Information Notice 2015-05, Inoperability Of Auxiliary And Emergency Feedwater Auto Start Circuits On Loss Of Main Feedwater Pumps2015-05-12012 May 2015 Inoperability Of Auxiliary And Emergency Feedwater Auto Start Circuits On Loss Of Main Feedwater Pumps Information Notice 2013-20, OFFICIAL EXHIBIT - NYS000538-00-BD01 - NRC Information Notice 2013-20: Steam Generator Channel Head and Tubesheet Degradation (October 3, 2013) (ML13204A143)2013-10-0303 October 2013 OFFICIAL EXHIBIT - NYS000538-00-BD01 - NRC Information Notice 2013-20: Steam Generator Channel Head and Tubesheet Degradation (October 3, 2013) (ML13204A143) Information Notice 2013-20, Official Exhibit - NYS000538-00-BD01 - NRC Information Notice 2013-20: Steam Generator Channel Head and Tubesheet Degradation (October 3, 2013) (ML13204A143)2013-10-0303 October 2013 Official Exhibit - NYS000538-00-BD01 - NRC Information Notice 2013-20: Steam Generator Channel Head and Tubesheet Degradation (October 3, 2013) (ML13204A143) Information Notice 2013-11, OFFICIAL EXHIBIT - NYS000551-00-BD01 - NRC Information Notice 2013-11: Crack-Like Indication at Dents/Dings and in the Freespan Region of Thermally Treated Alloy 600 Steam Generator Tubes (July 3, 2013)2013-07-0303 July 2013 OFFICIAL EXHIBIT - NYS000551-00-BD01 - NRC Information Notice 2013-11: Crack-Like Indication at Dents/Dings and in the Freespan Region of Thermally Treated Alloy 600 Steam Generator Tubes (July 3, 2013) Information Notice 2013-11, Official Exhibit - NYS000551-00-BD01 - NRC Information Notice 2013-11: Crack-Like Indication at Dents/Dings and in the Freespan Region of Thermally Treated Alloy 600 Steam Generator Tubes (July 3, 2013)2013-07-0303 July 2013 Official Exhibit - NYS000551-00-BD01 - NRC Information Notice 2013-11: Crack-Like Indication at Dents/Dings and in the Freespan Region of Thermally Treated Alloy 600 Steam Generator Tubes (July 3, 2013) Information Notice 2010-12, Intervenors' Fifth Motion to Amend and/or Supplement Proposed Contention No. 5 (Shield Building Cracking). Appendix VI: NRC FOIA Responses (B-51 Through B-53); Turkey Point Event Report; NRC Information Notice 2010-12: Contain2012-08-17017 August 2012 Intervenors' Fifth Motion to Amend and/or Supplement Proposed Contention No. 5 (Shield Building Cracking). Appendix VI: NRC FOIA Responses (B-51 Through B-53); Turkey Point Event Report; NRC Information Notice 2010-12: Containment Liner Cor Information Notice 2010-12, Intervenors' Fifth Motion to Amend and/or Supplement Proposed Contention No. 5 (Shield Building Cracking). Appendix VI: NRC FOIA Responses (B-51 Through B-53); Turkey Point Event Report; NRC Information Notice 2010-12: Con2012-08-17017 August 2012 Intervenors' Fifth Motion to Amend and/or Supplement Proposed Contention No. 5 (Shield Building Cracking). Appendix VI: NRC FOIA Responses (B-51 Through B-53); Turkey Point Event Report; NRC Information Notice 2010-12: Containment Liner Cor Information Notice 2010-12, Intervenors' Fifth Motion to Amend And/Or Supplement Proposed Contention No. 5 (Shield Building Cracking). Appendix VI: NRC FOIA Responses (B-51 Through B-53); Turkey Point Event Report; NRC Information Notic2012-08-17017 August 2012 Intervenors' Fifth Motion to Amend And/Or Supplement Proposed Contention No. 5 (Shield Building Cracking). Appendix VI: NRC FOIA Responses (B-51 Through B-53); Turkey Point Event Report; NRC Information Notice 2010-12: Containment Liner Cor Information Notice 2012-13, Boraflex Degradation Surveillance Programs and Corrective Actions in the Spent Fuel Pool2012-08-10010 August 2012 Boraflex Degradation Surveillance Programs and Corrective Actions in the Spent Fuel Pool Information Notice 2012-13, Boraflex Degradation Surveillance Programs And Corrective Actions In The Spent Fuel Pool2012-08-10010 August 2012 Boraflex Degradation Surveillance Programs And Corrective Actions In The Spent Fuel Pool Information Notice 2012-11, Age Related Capacitor Degradation2012-07-23023 July 2012 Age Related Capacitor Degradation ML12031A0132012-02-0606 February 2012 U.S. Nuclear Regulatory Commission Investigation Report No. 2-2010-058, Cpn International, Inc Information Notice 2011-19, Licensee Event Reports Containing Information Pertaining to Defects to Basic Components2011-09-26026 September 2011 Licensee Event Reports Containing Information Pertaining to Defects to Basic Components Information Notice 2011-15, Steel Containment Degradation and Associated License Renewal Aging Management Issues2011-08-0101 August 2011 Steel Containment Degradation and Associated License Renewal Aging Management Issues Information Notice 2011-17, Calculation Methodologies for Operability Determinations of Gas Voids in Nuclear Power Plant Piping2011-07-26026 July 2011 Calculation Methodologies for Operability Determinations of Gas Voids in Nuclear Power Plant Piping Information Notice 2011-13, Official Exhibit - NYS000329-00-BD01 - NRC Information Notice 2011-13, Control Rod Blade Cracking Resulting in Reduced Design Lifetime (Jun 29, 2011) (NRC in 2011-13)2011-06-29029 June 2011 Official Exhibit - NYS000329-00-BD01 - NRC Information Notice 2011-13, Control Rod Blade Cracking Resulting in Reduced Design Lifetime (Jun 29, 2011) (NRC in 2011-13) Information Notice 2011-13, Official Exhibit - Nys000329-00-Bd01 - NRC Information Notice 2011-13, Control Rod Blade Cracking Resulting in Reduced Design Lifetime (Jun 29, 2011) (Nrc in 2011-13)2011-06-29029 June 2011 Official Exhibit - Nys000329-00-Bd01 - NRC Information Notice 2011-13, Control Rod Blade Cracking Resulting in Reduced Design Lifetime (Jun 29, 2011) (Nrc in 2011-13) Information Notice 2011-13, OFFICIAL EXHIBIT - NYS000329-00-BD01 - NRC Information Notice 2011-13, Control Rod Blade Cracking Resulting in Reduced Design Lifetime (Jun 29, 2011) (NRC in 2011-13)2011-06-29029 June 2011 OFFICIAL EXHIBIT - NYS000329-00-BD01 - NRC Information Notice 2011-13, Control Rod Blade Cracking Resulting in Reduced Design Lifetime (Jun 29, 2011) (NRC in 2011-13) Information Notice 2011-04, IN: Contaminants and Stagnant Conditions Affecting Stress Corrosion Cracking in Stainless Steel Piping in Pressurized Water Reactors2011-02-23023 February 2011 IN: Contaminants and Stagnant Conditions Affecting Stress Corrosion Cracking in Stainless Steel Piping in Pressurized Water Reactors Information Notice 2011-04, In: Contaminants and Stagnant Conditions Affecting Stress Corrosion Cracking in Stainless Steel Piping in Pressurized Water Reactors2011-02-23023 February 2011 In: Contaminants and Stagnant Conditions Affecting Stress Corrosion Cracking in Stainless Steel Piping in Pressurized Water Reactors Information Notice 2011-04, in: Contaminants and Stagnant Conditions Affecting Stress Corrosion Cracking in Stainless Steel Piping in Pressurized Water Reactors2011-02-23023 February 2011 in: Contaminants and Stagnant Conditions Affecting Stress Corrosion Cracking in Stainless Steel Piping in Pressurized Water Reactors Information Notice 2010-26, New England Coalition'S Motion for Leave to Reply to NRC Staff'S Objection to Nec'S Notification of Information Notice 2010-26 and Entergy'S Response to the Supplement to Nec'S Petition for Commission Review of LBP-10-2010-12-30030 December 2010 New England Coalition'S Motion for Leave to Reply to NRC Staff'S Objection to Nec'S Notification of Information Notice 2010-26 and Entergy'S Response to the Supplement to Nec'S Petition for Commission Review of LBP-10-19 Information Notice 2010-26, New England Coalition'S Motion for Leave to Reply to NRC Staff'S Objection to Nec'S Notification of Information Notice 2010-26 and Entergy'S Response to the Supplement to Nec'S Petition for Commission Review2010-12-30030 December 2010 New England Coalition'S Motion for Leave to Reply to NRC Staff'S Objection to Nec'S Notification of Information Notice 2010-26 and Entergy'S Response to the Supplement to Nec'S Petition for Commission Review of LBP-10-19 Information Notice 2010-26, 2010/12/21-NRC Staff'S Objection to Nec'S Notification of Information Notice 2010-262010-12-21021 December 2010 2010/12/21-NRC Staff'S Objection to Nec'S Notification of Information Notice 2010-26 ML13066A1872009-12-16016 December 2009 Draft NRC Information Notice 2009-xx - 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 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
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)
|
---|