Headquarters Daily report APRIL 25, 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 - HEADQUARTERS APRIL 25, 1994 MR Number: H-94-0040 NRR DAILY REPORT ITEM GENERIC COMMUNICATIONS BRANCH/EVENTS ASSESSMENT BRANCH DIVISION OF OPERATING REACTOR SUPPORT OFFICE OF NUCLEAR REACTOR REGULATION Subject: N/A NRC Administrative Letter 94-06, "Visits by Members of the Public to Nuclear Power Plants," will be issued April 27, 1994. NRC is issuing this administrative letter to inform addressees of the existing NRC policy regarding visits by the public to nuclear power plants. Contact: Gene McPeek, NRR (301) 504-3210 PRIORITY ATTENTION REQUIRED MORNING REPORT - REGION I APRIL 25, 1994 Licensee/Facility: Notification: Boston Edison Co. MR Number: 1-94-0048 Pilgrim 1 Date: 04/22/94 Plymouth,Massachusetts SRI PC Dockets: 50-293 BWR/GE-3 Subject: MULTIPLE SCRAM SOLENOID PILOT VALVE FAILURES Reportable Event Number: N/A Discussion: On April 22, 1994 and after about one week of scram time testing, the licensee decided to shutdown the plant since they lost confidence in their data that satisfactory scram time test results reflected a good condition for the BUNA-N diaphragm in the scram solenoid pilot valves (SSPV). The diaphragms associated with 5 rods with acceptable scram times were examined and found to be deteriorated. On April 18, control rod (CR) 10-31 exceeded the 7 second full insertion time during periodic scram time testing. The licensee declared the control rod inoperable and deenergized it in the fully inserted position. The direct cause of the slow scram time was determined to be a ruptured SSPV diaphragm for the 118-E exhaust port. The 117 and 118 valves on each hydraulic control unit are associated with the "A" and "B", respectively, reactor protection systems. The ruptured SSPV diaphragm is made of BUNA-N material. Continued scram time testing identified several additional control rods whose 10 percent rod insertion times had notably degraded since the latest baseline test in June 1993. The licensee expanded the sample set and scram time tested all 145 control rods. Slow 10 percent rod insertion times on three control rods caused their respective 2X2 arrays to exceed Technical Specification prescribed limits (0.58 seconds). In each case the control rod was fully inserted, deenergized, the SSPV rebuilt, and the control rod was successfully retested. The licensee was attempting to establish criteria to predict scram time degradation, in order to identify individual SSPVs for rebuild in advance of actual rod insertion time testing failures. As an initial preemptive action, the licensee established SSPV rebuild criteria of (1) >.60 second 10 percent scram insertion time or (2) > .09 second increase in the 10 percent scram insertion time since the last baseline surveillance on the specific control rod. Using the above criteria approximately 20 control rods were considered to have degraded sufficiently to warrant rebuild of the SSPVs. Although the 118 exhaust port is most susceptible to diaphragm failure, the licensee concluded it was prudent to rebuild both the 117 and 118 SSPVs for each degraded control rod. Visual inspections indicated that a few of the 117 SSPV diaphragms and more of the 118 SSPV diaphragms had begun to harden. Material analysis (Shore A hardness test) indicated a durometer reading of 60 to 65 for most of the diaphragms. A few diaphragms measured at 85 to 95 units. The majority of the replaced diaphragms still felt flexible to the touch. Cracks were identified on at least two diaphragms. Evaluation of degradation due to potentially elevated temperature environment has not yet been conclusive. The licensee noted that the individual diaphragms within each SSPV (2 per SSPV) did not necessarily indicate similar hardness. Therefore concerns regarding potential use of locktight compound on the instrument air system (GE RICSIL 067) do not appear applicable. Licensee root cause evaluation was in progress and an accelerated surveillance interval was being considered. GE SILs 128 and 575 recommend a 7 year SSPV diaphragm total life (shelf life plus installed operating life) and a 4 year operating life respectively. Pilgrim rebuilt the SSPVs for all 145 control rods in June 1991. Material records indicate that all of the SSPV (143 verified, records for the remaining 2 are still being located) rebuild kits used in 1991 are still within the recommended 7 year total life. The majority (107) of these SSPV kits were received from GE in March 1991 (packaging date of August 1990). Fifteen of the first 16 degraded PSSVs were tracked back to the 1991 GE shipment. The licensee is in contact with GE and WNP-2 to discuss the issue further. Diaphragms from several of the degraded control rods and some from successfully timed control rods are being shipped to GE for further materials analysis and causal assessment. GE is currently performing testing on degraded diaphragms from WNP-2 at their King of Prussia, PA facility. Pilgrim was in cold shutdown as of 11:00 a.m., April 23, 1994. The licensee anticipated a 5 day outage to replace all BUNA-N diaphragms. A long term corrective action is tentatively to replace the elastomer with a new type of material in the upcoming May 1995 outage, about a year from now. Regional Action: NRC resident inspectors are on-site, monitoring scram time testing, root cause analysis, and repairs. Material tracking information including GE P.O. number, PQC number and requisition numbers have been forwarded to the NRR vendor inspection branch for further review at GE facilities. Contact: Rich Conte (610)337-5183 David Kern (508)747-0565 PRIORITY ATTENTION REQUIRED MORNING REPORT - REGION III APRIL 25, 1994 Licensee/Facility: Notification: Commonwealth Edison Co. MR Number: 3-94-0082 Quad Cities 1 Date: / / Cordova,Illinois Dockets: 50-254 BWR/GE-3 Subject: CONTAMINATION OF WORKER Reportable Event Number: N/A Discussion: The licensee reported that on April 20, 1994, a contract worker became highly contaminated during the machining of a disc from a Reactor Water Cleanup System valve. The disc was being machined to correct problems with the wedge and seat mating surface. The worker was not wearing a respirator, in accordance with radiation protection department instructions, but was wearing a dust mask and a face shield. In addition, a portable ventilation system was in use during the job. Whole body count data for the worker from April 20 indicated the presence of approximately 1.5 microcuries (55.5 kiloBecquerels) of cobalt-60 and manganese-54. By April 22, the contamination was reduced to approximately 145 nanocuries (5.5 kiloBecquerels). This reduction is consistent with the elimination of the radioactive material through the gastro-intestinal tract and the removal of external contamination through the normal loss of skin and hair. The licensee will be conducting additional whole body counts to assess the long term retention of the contamination. Regional Action: A senior radiation specialist was sent to the plant from the Region III office to review the event. Licensee whole body count and air sample data indicated the contamination did not exceed the Annual Limit of Intake of 10 CFR 20. Contact: JOHN GROBE (708)829-9837 PRIORITY ATTENTION REQUIRED MORNING REPORT - REGION IV APRIL 25, 1994 Licensee/Facility: Notification: Entergy Operations Inc. MR Number: 4-94-0035 Arkansas Nuclear 1 2 Date: 04/25/94 Russelville,Arkansas Dockets: 50-313,50-368 PWR/B&W-L-LP,PWR/CE Subject: ENFORCEMENT DISCRETION AND STARTUP FROM REFUELING OUTAGE Reportable Event Number: N/A Discussion: On April 22, 1994, during startup from refueling outage 2RF010, the licensee requested and was granted enforcement discretion from the requirements of Technical Specification (TS) 3.7.1.2(b) concerning the operability of the turbine-driven auxiliary feedwater pump which was modified during the outage. Postmodification testing of the pump resulted in an overspeed trip condition occurring. The cause of the overspeed was the malfunctioning of Valve 2SV-205, which functions to automatically open upon receipt of a start signal to initiate turbine roll prior to the opening of main steam turbine admission Valve 2CV-0340. After the failure, Valve 2SV-205 was replaced during the allowed outage time provided by TS 3.7.1.2 for the turbine-driven pump being out of service. Subsequent testing failed and the Limiting Condition for Operation was to expire at 9 p.m. (CDT) on April 22, 1994. On April 22, 1994, Region IV and NRR discussed the licensee's safety evaluation for extending the allowed outage time for the inoperability of the turbine-driven pump for 7 days beyond the TS allowed 3 days. The licensee's written safety evaluation and formal request were also received the afternoon of April 22. The licensee's justification for extending the allowed outage time while in Hot Standby (Mode 3) versus returning to Hot Shutdown (Mode 4) was considered valid from a safety perspective and the enforcement discretion was verbally granted by the Regional Administrator on April 22. During the evening of April 22, the licensee in conjunction with the valve vendor determined that some internal valve parts were causing excessive friction in the valve seat and solenoid sleeve area and required machining. The valve internals were brought within nominal tolerances, installed, and tested satisfactorily. The licensee completed retesting of the turbine-driven auxiliary feedwater pump and the TS was exited at 4:24 a.m. on April 23. The plant startup then continued and the reactor went critical at 2:46 a.m. Saturday, April 23. The licensee will evaluate the as-found valve condition and determine if reporting requirements per 10 CFR Part 21 are applicable. As of the morning of April 24, the licensee could not accurately state if the as-found internal valve parts were within manufacturers tolerances. Regional Action: Region IV will monitor the licensee's actions in this area. Contact: T. Stetka (817)860-8247 PRIORITY ATTENTION REQUIRED MORNING REPORT - REGION IV APRIL 25, 1994 Licensee/Facility: Notification: Southern California Edison & San MR Number: 4-94-0036 Diego Gas & Electric Co. Date: 04/25/94 San Onofre 2 Senior Resident Inspector Telecon San Clemente,California Dockets: 50-361 PWR/CE Subject: DROPPED CONTROL ELEMENT ASSEMBLY Reportable Event Number: N/A Discussion: At 12:59 p.m. (PST) on April 23, 1994, a control element assembly (CEA) dropped fully into the core during routine CEA exercising while the unit was operating at 98 percent power. The dropped CEA resulted in a power reduction to 91 percent. Operators initiated a downpower to 78 percent, in accordance with licensee procedures, as they were recovering the CEA. The CEA, CEA 20 in regulating Group 6, was successfully recovered and fully withdrawn at 1:19 p.m. The reactivity changes from borating and withdrawing the CEA offset each other so that the minimum power indicated that it remained at 91 percent. Following restoration of the CEA to its normal position, power was increased back to 98 percent. Licensee maintenance and engineering personnel responded to the site but could not identify the cause of the dropped CEA. Additionally, the licensee confirmed that no changes in primary or secondary activity resulted from the event. Unit 1 was shut down permanently in 1992, and Unit 3 was operating at 96 percent power at the time of the event. Regional Action: The Senior Resident Inspector discussed the event with the operators involved and reviewed the licensee's actions. The resident inspectors will monitor the licensee's followup activities. Contact: J. A. Sloan (714)492-2641 H. J. Wong (510)975-0296 PRIORITY ATTENTION REQUIRED MORNING REPORT - REGION IV APRIL 25, 1994 Licensee/Facility: Notification: Arizona Public Service Co. MR Number: 4-94-0037 Palo Verde 1 2 3 Date: 04/25/94 Wintersburg,Arizona Senior Resident Inspector Telecon Dockets: 50-528,50-529,50-530 PWR/CE80,PWR/CE80,PWR/CE80 Subject: RAYCHEM SPLICE DEFICIENCIES Reportable Event Number: N/A Discussion: On April 22, 1994, the licensee identified a potential generic problem with the environmental qualification of certain design Raychem splices used to seal 4160V electrical connections. During site training in which electrical maintenance personnel were being trained on site and vendor instructions on the installation of Raychem splices, an instructor opened a completed connection and discovered that the 2-inch wide adhesive strip applied over a joint had not fused to the outer casing. The licensee discovered that, in some instances, even when the vendor's instructions were strictly adhered to, the splices may not completely seal. Further testing demonstrated that a complete seal would occur if the outer casing was heated until the surface develops a glossy appearance. The licensee discussed the reheating technique with the vendor, who found it to be acceptable. The licensee conducted a Plant Review Board meeting to discuss this matter. The licensee subsequently inspected all field installed splices of this size in Units 1 and 2 and reheated those which they could not positively identify as properly sealed. The licensee plans to perform similar inspections in Unit 3, which is currently defueled, prior to refueling. Although the vendor has subjected these splices to environmental qualification testing, the licensee suspects that the splices may have self-sealed when they were exposed to the high temperatures associated with these tests. The licensee concluded that the splices may be vulnerable to high humidity, low heat environments. The vendor and licensee are evaluating Part 21 reporting requirements. Regional Action: The resident inspectors will be monitoring the licensee's followup activities. Contact: H. Wong (510)975-0296 K. Johnston (602)386-3638 PRIORITY ATTENTION REQUIRED MORNING REPORT - REGION IV APRIL 25, 1994 Licensee/Facility: Notification: Texas Utilities Electric Co. MR Number: 4-94-0038 Comanche Peak 2 Date: 04/25/94 Glen Rose,Texas Senior Resident Inspector Telecon Dockets: 50-446 PWR/W-4-LP Subject: INAPPROPRIATE USE OF 10 CFR 50.54(x) Reportable Event Number: 27135 Discussion: On April 22, 1994, around 12:15 a.m. (CDT), Comanche Peak, Unit 2, was in Mode 3, borated to cold shutdown conditions with shutdown banks withdrawn, making preparations to begin a midcycle outage. During the performance of a solid state protection system test, source range Channel N-32 spiked high and generated a trip signal on Train A. All rods fully inserted and reactor trip Breaker B remained closed because that channel was in bypass for the test. The emergency operating procedure in response to this event stated that all reactor trip breakers should be opened, but the licensee elected to invoke 10 CFR 50.54(x) and leave the breaker closed in order to facilitate troubleshooting of the spurious signal. Region IV considered this an inappropriate use of 10 CFR 50.54(x) and discussed this concern with licensee management. Licensee management considered the decision to leave the breaker closed a correct one, but agreed that this should not have been done under the provisions of 10 CFR 50.54(x). Licensee management also agreed that 10 CFR 50.54(x) should only be invoked in rare and serious circumstances and has initiated an investigation of the event. The licensee subsequently informed the HOO that invoking 10 CFR 50.54(x) for this minor procedure modification was not required. Regional Action: Region IV has initiated a special inspection of this event led by the Senior Resident Inspector with assistance from the Chief, Operations Branch, Division of Reactor Safety, Region IV. The entrance for the special inspection was held at 2 p.m. at the site on April 22, 1994. The Senior Resident Inspector will also be monitoring the results of the licensee investigation. Contact: Larry Yandell (817)860-8182