Headquarters Daily report DECEMBER 01, 1994 *************************************************************************** REPORT NEGATIVE NO INPUT ATTACHED INPUT RECEIVED RECEIVED HEADQUARTERS X REGION I X REGION II X REGION III X REGION IV X *************************************************************************** PRIORITY ATTENTION REQUIRED MORNING REPORT - REGION DECEMBER 1, 1994 MR Number: H-94-0108 NRR DAILY REPORT ITEM SIGNIFICANT EVENTS EVENTS ASSESSMENT BRANCH DIVISION OF OPERATING REACTOR SUPPORT Subject: REACTOR SCRAM FOLLOWED BY COMPLICATIONS The NRR/AEOD Events Assessment Panel on November 22, 1994, classified the scram and subsequent series of unexpected events and equipment failures as a Significant Event. The classification was based upon the extent of the complications that occurred following the reactor scram. On September 8, 1994, the River Bend Station experienced an automatic reactor trip on a reactor water level high (Level 8) trip signal, but the operators verified that the level was and had been stable at 36 inches (setpoint is 53 inches). After implementing the appropriate scram recovery procedures, the operators observed that an expected turbine trip had not occurred; therefore, they tripped the turbine manually about 10 minutes into the event. The operators also opened the generator output breakers which had failed to open automatically. After the output breakers were opened, electric power to a number of systems and components was unexpectedly lost. In addition, due to the loss of power to the Reactor Protection System, a containment isolation occurred which isolated the main steam and feed water systems. During the recovery period, the operators manually opened one safety/relief valve for pressure control and initiated operation of the reactor core isolation cooling system (RCIC); however, the RCIC pump turbine tripped on overspeed. No emergency action levels were reached. The complications were initiated by having a different actuation logic for a high water level scram (one-out-of-two taken twice) and the turbine generator trip (one-out-of-three). The resulting manual trip of the generator and con- comitant slow power transfer led to an unexpected loss of power to non-safety components and the RPS. Automatic fast transfer did not occur because of the unusually high inductive load on the generator prior to the trip effectively increased the reverse power trip setpoint of the generator. In addition, a number of independent component failures involving mitigating systems failed, thereby further complicating the recovery. A significant aspect of this event was the degradation of water make up capability after the scram. Feedwater was lost and the main steam line isolation valves closed about 10 minutes after the scram when the main generator output breakers were opened. RCIC was initiated and tripped. This left the operators with only one path of makeup water - the high pressure core spray system. The conditional core damage probability was estimated to be 1.2E-5. The event was briefed September 14, 1994, Operating Reactor Events Briefing 94-33, "Reactor Scram with Complications." CONTACT: Jerry Carter, NRR/DOPS/OECB (301) 504-1153 PRIORITY ATTENTION REQUIRED MORNING REPORT - REGION DECEMBER 1, 1994 MR Number: H-94-0109 NRR DAILY REPORT ITEM SIGNIFICANT EVENTS EVENTS ASSESSMENT BRANCH DIVISION OF OPERATING REACTOR SUPPORT Subject: INADVERTENT PRESSURIZER DRAIN-DOWN AND PARTIAL DE-PRESSURIZATION The NRR/AEOD Events Assessment Panel meeting on November 22, 1994, has classified inadvertent pressurizer drain-down and partial de-pressurization at Wolf Creek, Unit 1, as a Significant Event for the Performance Indicator Program. The significant event classification was based on the apparent programmatic weaknesses in the licensee's management of the plant. On September 17, 1994, with the plant in Mode 4 at 350 psig and 300 degrees F, operators were conducting motor operated valve testing on one of the isolation valves in the cross-connect line between the A and B trains of the residual heat removal (RHR) system. Train A of RHR was in service providing shutdown cooling. The plant had been shutdown less than 24 hours. The valve was initially stroked during the preceding evening shift. The results of the initial stroke indicated a need for a packing adjustment. Following the packing adjustment during the following shift, a request was made of control room operators to again stroke the valve. When the request was made, the balance of plant (BOP) control room operator (BOP operator) was preparing to adjust the boron concentration of RHR Train B to within that of the refueling water storage tank (RWST). He intended to recirculate the B train of RHR back to the RWST. (The boron concentration in Train B had been reduced during the operating cycle because of reactor coolant system (RCS) back-leakage through two check valves that isolate the discharge of the B train from the RCS). The BOP operator had briefed an auxiliary operator (AO) on the valve manipulation necessary to accomplish the RHR B train recirculation effort. The AO was dispatched to manually open the valve that isolates the common recirculation line for both trains of RHR. The recirculation line interconnects with the cross-connect line between the A and B trains of RHR. Around the time that the recirculation isolation valve was being opened, the BOP operator opened the cross-connect isolation valve at the request of maintenance personnel. With the cross-connect and recirculation isolation valves open, a flow path was created allowing RCS inventory to be transferred from RHR Train A to the RWST. With this flow path established, the pressurizer drained from nearly full to nearly empty. The RCS de-pressurized from 340 psig to approximately 225 psig, and approximately 9,200 gallons of water were transferred from the RCS to the RWST. The RWST overflowed and approximately 650 gallons of water were transferred to the waste holdup tank. The RCS remained sub-cooled, 225 psig and 307 degrees F (approximately 136 degrees subcooling). A supervisory operator recognized the cause of the event and directed the BOP operator to close RHR cross-connect isolation valve. This action terminated the drain-down in 66 seconds. Operators entered the abnormal operating procedure for responding to a loss of coolant accident while shutdown. This event was caused by deficiencies in human performance. The event is of concern because it indicates that, with the reactor in a shutdown condition, personnel appeared to have had a decreased awareness of the safety consequences of their actions and may not have realize the importance of careful control of plant activities to properly safeguard RCS inventory. Because plant systems are not in usual operating configurations while the plant is in a shutdown condition, proper human performance is important. Prior to this event, an operations outage coordinator had reviewed the outage schedule and communicated a concern to outage management and shift supervisors regarding the potential for draining the RCS to the RWST. The decision was made to conduct valve testing if the shift supervisor and Operations Outage Coordinator were "comfortable" to do so. The decision to proceed should have been based on some assurance that appropriate barriers were in place to prevent a RCS drain down and not based on a subjective "comfort" level. The information disseminated in NUMARC 91-06, "Guidelines for Industry Actions to Assess Shutdown Risk," should have sensitized the licensee to the potential for drain-down and appropriate preventive measures. The licensee was aware of other industry events, to the extent that in the control room they installed a mimic of the flow path that caused the drain down. They apparently missed the opportunity to ensure that their procedures adequately addressed the potential for drain down. The event was quickly terminated only because a supervising operator who was assisting in the shutdown activities (and not a member of the regular on-shift staff) recognized the cause of the event. A similar event at Wolf Creek had occurred in 1983 prior to fuel load. For these reasons, this event is -being classified as a Significant Event for the Performance Indicator Program. The Events Assessment and Generic Communications Branch is evaluating the risk implications of this event. CONTACT: Nick Fields, NRR/DOPS/OECB (301) 504-1173 PRIORITY ATTENTION REQUIRED MORNING REPORT - REGION DECEMBER 1, 1994 MR Number: H-94-0110 NRR DAILY REPORT ITEM SIGNIFICANT EVENTS EVENTS ASSESSMENT BRANCH DIVISION OF OPERATING REACTOR SUPPORT Subject: UNANALYZED CONDITION FOR EMERGENCY SERVICE WATER SYSTEM The NRR/AEOD Events Assessment Panel on November 22, 1994, classified the unanalyzed condition for Emergency Service Water (ESW) system as a Significant Event for the Performance Indicator Program. The significant event classification was based on the programmatic problems of poor communication between operators and maintenance and inadequate procedures. On August 3, 1994, with Unit 2 at 82 percent power (end of cycle coastdown) and Unit 3 at 100 percent power, the licensee left the ESW system in an unanalyzed condition for approximately 50 minutes. During Valve Operation Test and Evaluation System (VOTES) testing, the licensee failed to adequately control the positioning of an ESW valve that provided a flow path for ESW return to the ultimate heat sink. The licensee did not have the required procedures to ensure that personnel at all times were able to control the valve. Also, the control room supervisor and the reactor operator briefed the wrong maintenance crew on the proper control actions to operate the valve during testing. Due to the poor communications and lack of procedures, the ESW system was left in an unanalyzed condition with the valve shut and operator control/indication removed. Approximately 50 minutes later, the technicians reopened valve MO-498 on control room command due to overflowing of the emergency cooling tower by the normal service water. The licensee performed a detailed analysis of the limiting design basis accident assuming three of four emergency diesel generators (EDGs) operable during a loss-of-coolant accident with loss of offsite power. The ESW flow would be reduced approximately 40 percent from normal flow levels due to the ESW booster pumps being inoperable (as part of testing procedure). All emergency core cooling systems room coolers and equipment coolers would remain operable throughout the event. All EDGs would remain operable during the first 10 minutes without operator action. All EDGs would remain operable following the first 10 minutes if operators balanced EDG loads to below the continuous load rating of 2600 KW per their training and procedures (noting rising EDG temperatures). This event resulted in a Notice of Violation and proposed imposition of civil penalty of $87,500, for a Severity Level III violation. The violation was cited because the ESW system, designed to prevent or mitigate a serious safety event, was degraded to the extent a detailed evaluation was required to determine its operability. The item was escalated 150 percent due to NRC identification (+50 percent) and prior opportunity to identify (+100 percent). The item was mitigated 75 percent due to corrective actions (-25 percent) and licensee performance (-50 percent). Probabilistic risk assessment was not performed on this event because the short period of time the ESW was inoperable does not justify it as a significant event based on risk. However, the programmatic problem of poor communication between operators and maintenance and the inadequate procedures elevated it to a significant event. Peach Bottom has instituted several corrective actions to prevent future occurrences. In the short term, motor operated valves will be considered inoperable during VOTES testing and equipment subject to other maintenance activities will be considered inoperable unless work is adequately controlled. In the long term, enhanced controls will be placed on VOTES procedures and other work processes will be reviewed to ensure adequate control. The event was briefed November 30, 1994, Operating Reactors Events Briefing 94-40, "Unanalyzed Condition in Emergency Service Water." CONTACT: Thomas Greene, NRR/DOPS/OECB (301) 504-1175 PRIORITY ATTENTION REQUIRED MORNING REPORT - REGION II DECEMBER 1, 1994 Licensee/Facility: Notification: Florida Power & Light Co. MR Number: 2-94-0104 Turkey Point 4 Date: 12/01/94 Miami,Florida Dockets: 50-251 PWR/W-3-LP Subject: AUTOMATIC REACTOR TRIP DUE TO GENERATOR GROUND Reportable Event Number: 28091 Discussion: Turkey Point Unit 4 experienced an automatic reactor trip from 100% power due to actuation of two main generator ground relays and the lockout relay at 3:42 p.m. on November 30, 1994. Further investigations revealed that a generator flexible link to the "B" phase of the isolation phase bus had come loose causing the link to short to the bus duct thereby causing a ground. The flexible links associated with the main generator bus had been removed during the recent refueling outage. Apparently, several bolts associated with the flexible links were not properly torqued at that time. Post trip response was as expected. Following the reactor trip, a low-low steam generator water level caused all three auxiliary feedwater pumps to start. Auxiliary feedwater, combined with main feedwater, recovered steam generator water levels. As directed by the EOPs, operators closed the MSIVs to prevent RCS cooldown, thus transferring from condenser steam dumps to the atmospheric dumps the removal of the relatively low reactor decay heat. This response is normal at Turkey Point. Currently, the unit is in mode 3 and repairs are in progress. Restart is scheduled for today. Regional Action: The resident inspectors were in the control room at the time of the trip and are onsite following licensee activites. Contact: K. LANDIS (404)331-5509 PRIORITY ATTENTION REQUIRED MORNING REPORT - REGION IV DECEMBER 1, 1994 Licensee/Facility: Notification: Texas Utilities Electric Co. MR Number: 4-94-0137 Comanche Peak 1 Date: 12/01/94 Glen Rose,Texas Senior Resident Inspector Dockets: 50-445 PWR/W-4-LP Subject: TURBINE/REACTOR TRIP ON LOW COOLING WATER FLOW IN MAIN TURBINE GENERATOR Reportable Event Number: 28087 Discussion: This morning report provides an update to the 10 CFR 50.72 report made by the licensee on November 30, 1994, regarding a turbine trip and subsequent reactor trip on November 29, 1994, because of an indicated low primary cooling water flow in the main turbine generator when Unit 1 received a 2 out of 2 main generator stator low flow trip. A similar event occurred on February 1, 1994, in which the indicated low flow condition on primary cooling water to the main generator stator caused a turbine/reactor trip. It was determined that no actual flow loss to the main generator stator occurred. However, the precise cause of the event was indeterminate. As a result of the November 29, 1994, event, the licensee along with the turbine/generator supplier (Siemens) performed troubleshooting to determine the cause of the event. The licensee concluded that no actual flow loss to the main generator stator occurred. Although the actual cause of the event is indeterminate, the licensee believes that the spurious trip signal was generated from gas coming out of solution in the common transmitter sensing line. The licensee is continuing with evaluation of design modifications that could potentially improve the reliability of the turbine generator, which may include providing separate sensing lines for the stator cooling flow transmitters. The licensee plans to perform a reactor startup on December 1, 1994. The licensee will conduct additional review and monitoring during the turbine startup to attempt to further analyze and prevent the cause of the trip which will include venting of the primary water system. Regional Action: The resident inspector responded to the site on November 29, 1994, and Region IV has monitored licensee actions. The residents will continue to follow the licensee actions. Contact: Dwight Chamberlain (817)860-8249 PRIORITY ATTENTION REQUIRED MORNING REPORT - REGION IV DECEMBER 1, 1994 Licensee/Facility: Notification: Pacific Gas & Electric Co. MR Number: 4-94-0138 Diablo Canyon 2 Date: 12/01/94 Avila Beach,California Resident Inspector Dockets: 50-323 PWR/W-4-LP Subject: FAILURE OF UNIT 2 REACTOR TRIP BYPASS BREAKER B Reportable Event Number: N/A Discussion: During the performance of Train A SSPS Actuation Logic testing, a reactor trip bypass breaker failed to open when testing the breaker undervoltage trip function. The reactor trip breaker which was tested at the same time tripped satisfactorily. Following the failure of the reactor trip bypass breaker undervoltage coil trip function, the licensee opened the reactor trip bypass breaker manually using the local shunt trip. The reactor trip bypass Breaker B was considered inoperable following the breaker malfunction during testing. The licensee has removed the breaker and is performing troubleshooting to determine the cause of the failure. This bypass breaker will be tested once the root cause of the failure on the removed breaker has been determined, but before the next required surveillance test on the reactor trip breakers. The spare breaker has not been installed as of the morning of December 1, 1994, because the spare does not have the required seismic clips. Regional Action: The resident inspector and Regional management are monitoring the licensee's action in response to this event. Contact: D. Kirsch (510)975-0290 M. Tschiltz (805)595-2354 G. Johnston (510)975-0304