Headquarters Daily Report JUNE 22, 1998 *************************************************************************** REPORT NEGATIVE NO INPUT ATTACHED INPUT RECEIVED RECEIVED HEADQUARTERS û REGION I û REGION II û REGION III û REGION IV û PRIORITY ATTENTION REQUIRED MORNING REPORT - HEADQUARTERS JUNE 22, 1998 Licensee/Facility: Notification: Part 21 Database MR Number: H-98-0090 Goulds Pumps Date: 06/22/98 Subject: P21-ADDT'L DISCREPANCIES FOUND IN A VENDOR'S DEDICATION PROCESS Discussion: The Clinton licensee makes a potential Part 21 report of its findings, in addition to those found in NRC Inspection Report 99900732/97-01, concerning the dedication process and quality assurance program of Goulds Pumps. In particular, replacement parts purchased from Goulds Pumps did not contain the required material for application in safety-related systems. Specifically, the licensee found a safety-related water leg pump and a coupling kit in stores with SAE Grade 2 bolts instead of the required SAE Grade 5 bolts. The licensee evaluated this situation to be acceptable but replaced the bolts to restore design. Contact: Jerry Carter, PECB/NRR 415-1153 E-mail: tjc@nrc.gov _ HEADQUARTERS MORNING REPORT PAGE 2 JUNE 22, 1998 Licensee/Facility: Notification: Part 21 Database MR Number: H-98-0091 Siemens Date: 06/22/98 Subject: P21-DEVIATIONS IDENTIFIED IN STEAM LINE BREAK ANALYSIS METHOD. Discussion: Siemens Power Corporation (SPC), a nuclear fuel vendor, reports three deviations that apply either to the SPC steam line break analysis methodology or to application of this methodology to individual plants. SPC is evaluating whether these defects represent reportable defects under 10 CFR Part 21, to be completed by December 1, 1998. 1. Certain power distributions have been identified in steam line break analyses for some plants that may not bound potential plant conditions. This results from (a) variations in cycle to cycle radial peaking factors due to loading pattern changes that were not accounted for in the steam line break analysis and (b) the iteration between the XCOBRA-IIIC thermal-hydraulic code and the XTGPWR neutronics code may not ensure calculation of a conservative power distribution when the pumps are not operating. 2. Iteration between XCOBRA-IIIC and XTGPWR may result in nonconservative reactivity calculations. 3. Use of the ANF-RELAP HFP systems analysis code may result in an underprediction of the positive reactivity insertion due to the cooldown of the fuel during a steam line break event for some plants. This is because reactivity control system inputs were constructed for some plant such that a step change occurs in the Doppler reactivity when switching between two reactivity models. Contact: Edward F. Goodwin, PECB/NRR 415-1154 E-mail: efg@nrc.gov _ HEADQUARTERS MORNING REPORT PAGE 3 JUNE 22, 1998 Licensee/Facility: Notification: Part 21 Database MR Number: H-98-0092 Siemens Date: 06/22/98 Subject: p21-CALCULATION OF PEAK CLADDING TEMPERATURE Discussion: Siemens Power Corporation (SPC), a nuclear fuel vendor, reports that the locus of peak cladding temperature for an axial power shape skewed toward the top of the core may be unrealistically close to the top of the fuel (above 10.5 feet), depending on the axial nodalization used. SPC reports this potential result for the code TOODEE2 (part of the NRC-approved EXEM/PWR LOCA evaluation model). SPC is evaluating whether this behavior represents a reportable defect under 10 CFR Part 21, to be completed by September 30, 1998. SPC states that there is no indication at this time that this problem will result in peak cladding temperatures exceeding 10 CFR 50.46 limits. Standard practice at SPC is to use 3 inch nodes near the peak power node and 6 inch or 12 inch nodes near the top of the core where the power is lower than the axial peak value. With this nodalization, the peak cladding temperature for an axial power shape skewed toward the top of the core generally occurs at or slightly above the peak power node between 9 and 10.5 feet for a 12-foot core. If the larger nodes near the top of the core are replaced by 3 inch nodes, the EXEM/PWR model may calculate an even higher peak cladding temperature near the top of the fuel (above 10.5 feet), even though power at this elevation is less than that at the peak power node. Contact: Edward F. Goodwin, PECB/NRR 415-1154 E-mail: efg@nrc.gov _ HEADQUARTERS MORNING REPORT PAGE 4 JUNE 22, 1998 Licensee/Facility: Notification: Part 21 Database MR Number: H-98-0093 Valcor Date: 06/22/98 Subject: P21-FAILURE OF AIR PILOT VALVE TO CLOSE Discussion: Valcor Engineering Corporation (Valcor), the vendor, reports that Valcor Model V70900-65-11 air pilot valves failed to close immediately on deenergization. The valves failed after being continuously energized for 6-18 months. Closing action may be delayed by such extended service and subsequent compression setting of the O-ring seat because the air gap between the plunger and its stop may become too small and thus subject the valve to residual magnetic forces, which tend to prevent closure on deenergization. Opening operation is not affected. Valcor states that this model valve, an alternating current version, is the only version of the V70900-65 series of air pilot valves subject to this phenomenon. Only 46 units of this particular version have been delivered, all to the Susquehanna nuclear power plant. Valcor will modify all the units to correct against this effect. Contact: Rick Khan, PECB/NRR 415-1152 E-mail: tmk@nrc.gov _ HEADQUARTERS MORNING REPORT PAGE 5 JUNE 22, 1998 Licensee/Facility: Notification: Part 21 Database MR Number: H-98-0094 Duane Arnold/Hiller Date: 06/22/98 Subject: P21-EVIDENCE OF INTERNAL LEAKAGE IN MAIN STEAM ISOLATION VALVE Discussion: The Duane Arnold licensee notified its vendor, Ralph A. Hiller Co., that a Hiller Model SA-A101 main steam isolation valve actuator showed evidence of internal leakage, in both directions, past the pneumatic piston. The licensee returned the actuator to the vendor for evaluation. The vendor verified the excessive leakage and discovered a fracture of the pneumatic piston. The licensee found another similar actuator with a similar crack but no leakage. The vendor states that the safety function of the actuator would not have been prevented but is continuing to evaluate this potential defect for reportability under 10 CFR Part 21. Contact: John Tappert, PECB/NRR 415-1167 E-mail: jrt@nrc.gov _ HEADQUARTERS MORNING REPORT PAGE 6 JUNE 22, 1998 Licensee/Facility: Notification: Part 21 Database MR Number: H-98-0095 Cooper Energy Date: 06/22/98 Subject: P21-POST-INSTALLATION TESTING OF EMERGENCY DIESEL GENERATORS Discussion: Cooper Energy Services, an emergency diesel generator vendor, reports that post-installation testing of Enterprise DSR-4 and DSRV-4 diesel engines revealed that high electromagnetic fields affected new controllers but not original controllers. Another report is scheduled for no later than July 31, 1998. Contact: Nick Fields, PECB/NRR 415-1173 E-mail: enf@nrc.gov _ HEADQUARTERS MORNING REPORT PAGE 7 JUNE 22, 1998 Licensee/Facility: Notification: Part 21 Database MR Number: H-98-0096 Ginna/Westinghouse Date: 06/22/98 Subject: P21-FAILURE OF A SAFETY-RELATED DB-25 CIRCUIT BREAKER Discussion: The Ginna licensee reports failure of a Westinghouse DB-25 circuit breaker to close during testing. The breaker was installed as the power supply feed breaker for a service water pump, a safety-related function. The licensee found that a pin was missing in the operating mechanism. In 1992, Westinghouse Replacement Components Services, Monroeville PA, refurbished the breaker. Contact: David Skeen, PECB/NRR 415-1174 E-Mail: dls@nrc.gov _ REGION III MORNING REPORT PAGE 8 JUNE 22, 1998 Licensee/Facility: Notification: Commonwealth Edison Co. MR Number: 3-98-0046 Dresden 2 Date: 06/20/98 Morris,Illinois ENS PHONE CALL 34418 Dockets: 50-237 BWR/GE-3 Subject: UNIT 2 REACTOR SCRAM Discussion: On June 20, the licensee reported that Dresden Unit 2 scrammed from 100 percent power when the electro-hydraulic control (EHC) system lost pressure resulting in a turbine trip and reactor trip. The EHC system lost pressure when the standby EHC pump was started. The licensee ran multiple tests on the system, but could not duplicate the trip conditions. This led the licensee to conclude that air may have been entrained in the system and caused a low pressure condition which cleared afterward. Following the scram, equipment performed as expected, including the feedwater control system which in the past had overfilled the reactor vessel. The licensee has restarted Unit 2, entering Mode 2 "Startup" at 0149 a.m. on Monday, June 22, 1998. Regional Action: The resident inspector responded to the site and restart follow-up activities are in progress. Contact: M. RING (630)829-9703 _