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Home My WebLink About851180.tiff SSINS No. : 6835 IN 85-85 UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C. 20555 October 31, 1985 IE INFORMATION NOTICE $5-85: SYSTEMS INTERACTION EVENT RESULTING IN REACTOR SYSTEM SAFETY RELIEF VALVE OPENING FOLLOWING A FIRE-PROTECTION DELUGE SYSTEM MALFUNCTION Addressees: All nuclear power reactor facilities holding an operating license (OL) or a construction permit (CP). Purpose: This notice is provided to alert licensees of a serious systems interaction event involving the fire-protection deluge system located in the control room ventilation charcoal filter housing. Following inadvertent actuation of this system, an analog transient trip system (ATTS) panel was sprayed with water causing malfunctions in certain safety system components. It is expected that recipients will review this notice for applicability to their facilities and consider actions, if appropriate, to preclude a similar problem occurring at their facilities. However, suggestions contained in this notice do not constitute requirements; therefore, no specific action or written response is required. Description of Circumstances: On May 15, 1985, at Georgia Power Company's Hatch Unit 1, personnel manually scrammed the reactor from 75% power because of a stuck open low-low-set safety relief valve (LLS-SRV). Shorting of one of the two redundant power supplies and/or possibly intermittent shorting of logic system contacts in the ATTS panel is believed to have caused the stuck open LLS-SRV. The panel is one of two redundant panels located in the control room. The cause of the electrical shorts in the affected panel was water intrusion into the panel . The event began about 8:35 p.m. when an instrument water supply vent valve was damaged, apparently by dragging of a crane hook along the line. The instru- ment water supply line eventually depressurized causing a portion of the fire- protection deluge system to actuate. The water supply line is located above the control building and the deluge system is located in the control room charcoal filter housing. Following actuation of the deluge system, approximately 15 to 25 gal of water backed up into the ventilation header before the system could be secured. The 8510290039 851180 cz,4 rouil3PBS IN 85-85 October 31, 1985 Page 2 of 3 backup was caused by plugged drains 'in the charcoal filter housing. Water eventually leaked through a hole in the ventilation piping that was located above the ATTS panel in the control room. When the water sprayed onto the panel , one of two redundant panel power supplies apparently shorted because of water intrusion into the panel . As a result, a LLS-SRV valve began to cycle open and closed. The SRV cycled three times and then opened and remained open. The operator manually scrammed the reactor from 75% power. A false turbine high exhaust pressure trip signal also was generated', temporarily disabling the high pressure core injection (HPCI) system: The reactor core isolation cooling (RCIC) system was inoperable at the time, so neither HPCI nor RCIC was imme- diately available for use. Fortunately, neither system was needed during the event. This is because the water level was restored and maintained by the . reactor feedwater system until the MSIVs were shut. Subsequent to MSIV closure, water level was maintained by the control rod drive (CRD) system with thee, excess water being dumped to the condenser via the reactor water cleanup ,system. The LLS-SRV closed without operator action at 9:52 pm. Discussion: The event is of considerable concern because of the potential for multiple safety system failures through unanalyzed systems interactions. In this event, the water from the fire-suppression deluge system in the control room caused opening of a safety relief valve and loss of primary system inventory. The event could have been seriously aggravated by the spurious HPCI turbine high exhaust pressure .trip that was received, also apparently as a result of the water intrusion. Because the RCIC system was inoperable at the time of the event, no safety-related high pressure injection system would have been imme- diately available to restore water level should that have been necessary. ' The HPCI turbine trip signal was reset shortly after it occurred, however, and the system was returned to operability. Perhaps more serious is the potential effect the water could have had on numerous other safety systems. The ATTS panels have permissive and arming . logic and trip logic for various safety systems, 'as well as water level inputs to the HPCI, RCIC, core spray (CS) , automatic depressurization system (ADS), residual heat removal (RHR) system, and diesel activation logic. It is hard to predict the anomalous behavior that could occur if ,both power supplies had been lost, or if other portions of the logic had been shorted; but -quite possibly, several safety systems could have malfunctioned, seriously handicapping' the operators during their efforts to stabilize the unit. Prior to this event, no procedures were in place at Hatch Unit 1 for adequately cleaning the ventilation plenums or drains , in ,the charcoal filter units. Had . these procedures been prepared and implemented, the drains would have functioned as designed with no serious adverse effects. In response tp this event, the licensee cleaned and inspected drains in the remaining filter units andAs preparing cleanout and inspection procedures to be added to the maintenance schedules. IN 85-85 October 31, 1985 Page 3 of 3 Another example of a design feature which could cause potential adverse system interactions was recently found at Unit 1 of the South Texas Project. A non- seismic, non-category I potable water line was found to pass through the control room envelope via a relay room next to the control room. This could subject the solid-state protection system cabinets and the Westinghouse 7300 process control system located nearby to water damage following a seismic event. Although this unit is under construction, it does point out that these problems can occur. Also, IE Information Notice 83-41, "Actuation of Fire Suppression System Causing Inoperability of Safety Related Equipment," was issued on June 22, 1983. That notice identified a number of instances in which automatic actuation of fire suppression systems degraded or jeopardized the operability of safety- related equipment. No specific action or written response is required by this information notice. If you have any questions regarding this matter, please contact the Regional Administrator of the appropriate NRC regional office or the technical contact listed below. lwarrdan�or Divis n of Emergency Preparedness and Engineering Response Office of Inspection and Enforcement Technical Contact: David R. Powell , IE (301) 492-8373 Attachment: List of Recently Issued IE Information Notices Attachment 1 IN 85-85 October 31, 1985 LIST OF RECENTLY ISSUED IE INFORMATION NOTICES Information Date of Notice No. Subject Issue Issued to 85-84 Inadequate Inservice Testing 10/30/85 All power reactor Of Main Steam Isolation Valves facilities holding an OL or CP 85-83 Potential Failures Of General 10/30/85 All power reactor Electric PK-2 Test Blocks facilities holding an OL or CP 85-82 Diesel Generator Differen- 10/18/85 All power reactor tial Protection Relay Not facilities holding Seismically Qualified an OL or CP 85-81 Problems Resulting In 10/17/85 All power reactor Erroneously High Reading facilities holding With Panasonic 800 Series an OL or CP and Thermoluminescent Dosimeters certain material and fuel cycle licensees 85-80 Timely Declaration Of An 10/15/85 All power reactor Emergency Class Implenienta- facilities holding tion Of An Emergency Plan, an OL or CP And Emergency Notifications 85-17 Possible Sticking Of ASCO 10/1/85 All power reactor Sup. 1 Solenoid Valves facilities holding an OL or CP 85-79 Inadequate Communications 9/30/85 All power reactor Between Maintenance, facilities holding Operations, And Security an OL or CP; research Personnel and nonpower reactor facilities; fuel fabrication and processing facilities 85-78 Event Notification 9/23/85 All power reactor facilities holding an OL or CP OL = Operating License CP = Construction Permit SSINS No. : 6835 IN 85-84 UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C. 20555 October 30, 1985 IE INFORMATION NOTICE NO. 85-84: INADEQUATE TESTING OF MAIN STEAM ISOLATION VALVES Addressees: All nuclear power reactor facilities holding an operating license (OL) or a construction permit (CP). Purpose: This notice is being provided to alert recipients of a potentially significant problem concerning the possible failure of main steam isolation valves (MSIVs) to close under low steam flow conditions and the testing of these valves with non-safety-related motive power in place. It is expected that recipients will review the information for applicability to their facilities and consider actions, if appropriate, to preclude a similar problem occurring at their facilities. However, suggestions contained in this information notice do not constitute NRC requirements; therefore, no specific action or written response is required. Past Related Correspondence Information Notice 85-21, "Main Steam Isolation Valve Closure Logic" , March 18, 1985. Description of Circumstances: During inspections at Robinson Unit 2 in November 1984 and at Turkey Point Units 3 and 4 in February 1985, NRC inspectors noted that MSIV surveillance testing procedures did not call for securing the instrument air supply to the MSIV control system during a test. Recognizing this as contrary to the objec- tive of operational verification of the MSIVs, the NRC cited these plants for violating 10 CFR 50.55a(g). After reviewing the matter to determine the corrective action, Florida Power & Light Co. , the licensee for Turkey Point Units 3 and 4, reported to the NRC on July 23, 1985, that a deficiency existed concerning the ability of MSIVsto close under low steam flow conditions. The safety-related air supply, stored in accumulators, was not adequate to close the valves in the event of loss of the non-safety-related instrument air system. This had not been discovered during routine testing because that testing had been performed improperly using the non-safety-related instrument air to achieve closure. 8510250546 1�d m-nali )%6 IN 85-84 October 30, 1985 Page 2 of 3 Operating air for the MSIVs is stored in accumulators mounted on the valve assembly; the non-safety-related plant instrument air system provides addition- al supply. During normal operation the MSIVs at Turkey Point are held open against steam flow by air pressure acting on the bottom of the actuator operat- ing piston. When a closing signal is received, air is directed to the top of the piston while air is vented from the bottom of the piston. Closure of each MSIV is assisted by a spring that moves the piston part way, by steam flow in the steam line, and by gravity. Assuming a loss of the instrument air system, the air stored in the safety-related accumulators may not be adequate to close the MSIV without sufficient assistance from steam flow. The Turkey Point MSIVs are required to close within 5 seconds to mitigate the consequences of a large main steam line break accident. In the event of such an accident, the high steam flow rate would assist in closing the MSIVs. However, MSIV closure also is required for other events in which large steam flow may not exist. Under these conditions and a loss of instrument air pressure, the accumulator air volume may not be sufficient to close the MSIVs. In the regulations, 10 CFR 50. 55a(g) requires that inservice testing to verify operational readiness of pumps and valves whose function is required for safety be accomplished in accordance with Section XI of the ASME Boiler and Pressure Vessel (BPV) Code. The ASME BPV Code, Section XI, 1980 edition through winter 1980 addenda, Paragraph IWV-3415, requires that fail-safe valves be tested by observing the operation of the valves upon loss of actuator power. Since the MSIVs have been identified as fail-safe valves they should have been tested with the instrument air supply, as well as electric power, removed. Proper testing would have revealed the inadequate accumulators much earlier. Discussion: The practice of performing inservice testing of components, which are relied on to mitigate the consequences of accidents, with sources of power not considered in the safety analyses is not in keeping with the objective of periodic test- ing. This objective is to test equipment to verify operational readiness under conditions that reasonably duplicate the design basis. When such testing was performed at Turkey Point, it was shown that with low or no steam flow, MSIV closure could only be assured with instrument air powering the actuator. Continued operation at Turkey Point has been justified by the availability of two instrument air systems as backups and by procedures that require plant shutdown if the instrument air supply is lost. In addition, design modifica- tions are being implemented on an expedited basis that will ensure MSIV closure in 5 seconds without steam flow assistance or non-safety-related instrument air power. These modifications also will resolve the testing deficiency noted above. IN 85-84 October 30, 1985 Page 3 of 3 No specific action or written response is required by this information notice. If you have any questions about this matter, please contact the Regional Administrator of the appropriate regional office or this office. `i. ordan, Director Division f Emergency Preparedness and Engineering Response Office of Inspection and Enforcement Technical Contact: Vern Hodge, IE (301) 492-7275 Attachment: List of Recently Issued IE Information Notices Attachment 1 IN 85-84 October 30, 1985 LIST OF RECENTLY ISSUED IE INFORMATION NOTICES Information Date of Notice No. Subject Issue Issued to 85-83 Potential Failures Of General 10/30/85 All power reactor Electric PK-2 Test Blocks facilities holding an OL or CP 85-82 Diesel Generator Differen- 10/18/85 All power reactor tial Protection Relay Not facilities holding Seismically Qualified an OL or CP 85-81 Problems Resulting In 10/17/85 All power reactor Erroneously High Reading facilities holding With Panasonic 800 Series an OL or CP and Thermoluminescent Dosimeters certain material and fuel cycle licensees 85-80 Timely Declaration Of An 10/15/85 All power reactor Emergency Class Implementa- facilities holding tion Of An Emergency Plan, an OL or CP And Emergency Notifications 85-17 Possible Sticking Of ASCO 10/1/85 All power reactor Sup. 1 Solenoid Valves facilities holding an OL or CP 85-79 Inadequate Communications 9/30/85 All power reactor Between Maintenance, facilities holding Operations, And Security an OL or CP; research Personnel and nonpower reactor facilities; fuel fabrication and processing facilities 85-78 Event Notification 9/23/85 All power reactor facilities holding an OL or CP 85-77 Possible Loss Of Emergency 9/20/85 All power reactor Notification System Due To facilities holding Loss Of AC Power an OL or CP OL = Operating License CP = Construction Permit SSINS No. : 6835 IN 85-83 UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C. 20555 October 30, 1985 IE INFORMATION NOTICE NO. 85-83: POTENTIAL FAILURES OF GENERAL ELECTRIC PK-2 TEST BLOCKS Addressees: All nuclear power reactor facilities holding an operating license (OL) or a construction permit (CP). Purpose: This information notice is to alert recipients of a potentially significant problem involving fractures of PK-2 test block terminal posts that could lead to inoperability of essential electrical equipment. It is expected that recipients will review this information for applicability to their facilities and consider actions, if appropriate, to preclude a similar problem occurring at their facilities. However, suggestions contained in this information notice do not constitute NRC requirements; therefore, no specific action or written response is required. Description of Circumstances: The Tennessee Valley Authority (TVA) reported to General Electric (GE) and the NRC that the Sequoyah Nuclear Power Plant experienced fracture failure of terminal posts on some PK-2 test blocks during routine testing of a circuit that was not safety related. TVA subsequently tested PK-2 blocks at the Watts Bar Nuclear Plants and found that terminal post fractures could be induced, in some cases, by wiggling by hand. No failures were identified by TVA on PK-2 safety-related applications. GE is investigating to determine the root cause of the failures. However, because of the broad usage of the test blocks, GE has notified the NRC that GE is unable to determine all of the possible PK-2 class IE installations and therefore is unable to assure specific notification of the problem to each utility. Discussion: Failure of PK-2 test blocks could occur either during testing or at other times. Failure during circuitry testing could result in the inoperability of essential electrical equipment. In this case, the loss of the electrical equipment would be detected during the testing. The possibility also exists for the fracture failure to occur at the completion of the circuit testing or as the result of bumping during other maintenance. Such an occurrence would 8510250543 IN 85-83 October 30, 1985 Page 2 of 2 result in an open circuit and the unavailability of the associated electrical equipment could go undetected in certain applications. Though the failures noted by TVA were not safety related, the usage of PK-2 test blocks includes safety related equipment such as emergency diesel generator relay boards. In addition to the possible loss of safety related equipment there is a potential personnel safety concern if an open circuit is developed on a current trans- former circuit during testing. General Electric has recommended that, in addition to visual inspection, a force of about five pounds in any direction perpendicular to the terminal posts can be applied to detect incipient failures. No specific action or written response is required by this information notice. If you have any questions about this matter, please contact the Regional Administrator of the appropriate regional office or this office. Edward Jordan, for Divisi of Emergency Preparedness and gineering Response Office of Inspection and Enforcement Technical Contact: James C. Stewart, IE (301) 492-9061 Attachment: List of Recently Issued IE Information Notices Attachment 1 IN 85-83 October 30, 1985 LIST OF RECENTLY ISSUED IE INFORMATION NOTICES Information Date of Notice No. Subject Issue Issued to 85-82 Diesel Generator Differen- 10/18/85 All power reactor tial Protection Relay Not facilities holding Seismically Qualified an OL or CP 85-81 Problems Resulting In 10/17/85 All power reactor Erroneously High Reading facilities holding With Panasonic 800 Series an OL or CP and Thermoluminescent Dosimeters certain material and fuel cycle licensees 85-80 Timely Declaration Of An 10/15/85 All power reactor Emergency Class Implementa- facilities holding tion Of An Emergency Plan, an OL or CP And Emergency Notifications 85-17 Possible Sticking Of ASCO 10/1/85 All power reactor Sup. 1 Solenoid Valves facilities holding an OL or CP 85-79 Inadequate Communications 9/30/85 All power reactor Between Maintenance, facilities holding Operations, And Security an OL or CP; research Personnel and nonpower reactor facilities; fuel fabrication and processing facilities 85-78 Event Notification 9/23/85 All power reactor facilities holding an OL or CP 85-77 Possible Loss Of Emergency 9/20/85 All power reactor Notification System Due To facilities holding Loss Of AC Power an OL or CP 85-76 Recent Water Hammer Events 9/19/85 All power reactor facilities holding an OL or CP OL = Operating License CP = Construction Permit OMB No.: 3150-0011 Expiration Date: 9/30/86 IEB 85-01 UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C. 20555 October 29, 1985 IE BULLETIN 85-01: STEAM BINDING OF AUXILIARY FEEDWATER PUMPS Addressees: For Action - Those nuclear power reactor facility licensees and construction permit (CP) holders listed in Attachment 1. For Information - All other nuclear power reactor facilities. Purpose: The purpose of this bulletin is to inform licensees and CP holders of a poten- tially serious safety problem that has occurred at certain operating facilities involving the inoperability of auxiliary feedwater (AFW) pumps as a result of steam binding. Certain PWR licensees and all PWR CP holders are requested to take further action to prevent similar events from occurring at their facili- ties and to document those actions taken or planned. Description of Circumstances: Numerous events have been reported where hot water has leaked into AFW systems and flashed to steam, disabling the AFW pumps. Events at Robinson 2 in 1981 through 1983, Crystal River 3 in 1982 and 1983, and D. C. Cook 2 in 1981 were summarized in IE Information Notice (IN) 84-06, issued in January 1984. Also in January 1984, the Institute of Nuclear Power Operations (INPO) issued Significant Event Report (SER) 5-84 detailing events at Robinson 2 and Farley. In April 1984, INPO issued Significant Operating Experience Report (SOER) 84-3 that discussed another event at Surry 2 in 1983. The NRC's Office for Analysis and Evaluation of Operational Data (AEOD) issued a case study report entitled "Steam Binding of Auxiliary Feedwater Pumps" in July 1984. This study identified 22 events since 1981; 13 of these occurring in 1983. Based on operating experience, it appears that backleakage into AFW could occur in any PWR. In a number of plants, the two motor-driven pumps feed into a single pipe which feeds into the steam generator; therefore, a leaking valve in that pipe increases the probability of steam binding in both trains of AFW. Also, multiple AFW pumps often take suction from a common manifold; therefore, if one pump becomes steam bound because of leaking check valves, the steam can heat the common suction and cause other pumps to become steam bound. 8510250539 IEB 85-01 October 29, 1985 Page 2 of 4 AFW capability is needed for normal shutdown and transient conditions as well . as for accident mitigation. The AEOD case study examined the effects of steam binding on a sequence in which there was a loss of the power conversion (steam generation) system after a transient other than loss-of-offsite power. A probabilistic risk analysis had previously shown this sequence to be a dominant contributor to the core-melt risk for a sample plant (Sequoyah) . The case study indicated that unavailability of the AFW system as a result of steam binding contributes significantly to the risk of core melt in PWRs. Monitoring AFW to detect backleakage and to promptly correct the situation if backleakage occurs would reduce the probability of steam binding. Since the AEOD study was issued, a series of events involving backflow of hot water into the AFW system occurred at McGuire 2 over a period of 7 days in August 1984, before effective corrective action was taken. One of these events involved overpressurization of the suction line and damage to instruments. In November 1984, Catawba 1 experienced backflow of hot water into AFW resulting in fumes from insulation and blistering of paint. In December 1984, the NRC's Office of Nuclear Reactor Regulation (NRR) determined that steam binding of AFW was a generic issue and assigned it a high priority (Generic Issue 93, "Steam Binding of Auxiliary Feedwater Pumps") . To determine the extent of the safety issue and the need for short-term correc- tive actions, the NRC's regional offices conducted a survey in April and May of 1985. Of the 58 operating reactors surveyed, 39 had temperature monitoring of AFW piping at least once per shift. Of the remaining 19, 17 had normally closed gate or globe valves in the pump discharge path in addition to check valves, or some unique feature such as complete separation of trains that made serious safety problems unlikely. The remaining 2 licensees have subsequently decided to monitor AFW piping temperature. Although some degree of action has been taken at all units, many have not incorporated these actions into procedures to detect or correct steam binding. Without these provisions, there is little assurance that effective actions will continue. For this reason, the addressees are requested to take the following actions: Action for Addressees Listed in Attachment 1 1. Develop procedures for monitoring fluid conditions within the AFW system on a regular basis during times when the system is required to be operable. This monitoring should ensure that fluid temperature at the AFW pump discharge is maintained at about ambient temperature. Monitoring of fluid conditions, if used as the primary basis for precluding steam binding, is recommended each shift. This item is not intended to require elaborate instrumentation. A simple means of monitoring temperature, such as touching the pipe, is a satisfac- tory approach. 2. Develop procedures for recognizing steam binding and for restoring the. AFW system to operable status, should steam binding occur. IN 85-84 October 30, 1985 Page 3 of 3 No specific action or written response is required by this information notice. If you have any questions about this matter, please contact the Regional Administrator of the appropriate regional office or this office. edY. ordan, Director Division f Emergency Preparedness and Engineering Response Office of Inspection and Enforcement Technical Contact: Vern Hodge, IE (301) 492-7275 Attachment: List of Recently Issued IE Information Notices Attachment 1 IN 85-84 October 30, 1985 LIST OF RECENTLY ISSUED IE INFORMATION NOTICES Information Date of Notice No. Subject Issue Issued to 85-83 Potential Failures Of General 10/30/85 All power reactor Electric PK-2 Test Blocks facilities holding an OL or CP 85-82 Diesel Generator Differen- 10/18/85 All power reactor tial Protection Relay Not facilities holding Seismically Qualified an OL or CP 85-81 Problems Resulting In 10/17/85 All power reactor Erroneously High Reading facilities holding With Panasonic 800 Series an OL or CP and Thermoluminescent Dosimeters certain material and fuel cycle licensees • 85-80 Timely Declaration Of An 10/15/85 All power reactor Emergency Class Implementa- facilities holding tion Of An Emergency Plan, an OL or CP And Emergency Notifications 85-17 Possible Sticking Of ASCO 10/1/85 All power reactor Sup. 1 Solenoid Valves facilities holding an OL or CP 85-79 Inadequate Communications 9/30/85 All power reactor Between Maintenance, facilities holding Operations, And Security an OL or CP; research Personnel and nonpower reactor facilities; fuel fabrication and processing facilities 85-78 Event Notification 9/23/85 All power reactor facilities holding an OL or CP 85-77 Possible Loss Of Emergency 9/20/85 All power reactor Notification System Due To facilities holding Loss Of AC Power an OL or CP OL = Operating License CP = Construction Permit Hello