IR 05000244/1999007
| ML20212F669 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 09/20/1999 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17265A752 | List: |
| References | |
| 50-244-99-07, NUDOCS 9909280314 | |
| Download: ML20212F669 (28) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Docket No: 50-244 i
I License No: DPR-18 L
l Report No: 50-244/99-07 Licensee: Rochester Gas and Electric Corporation (RG&E)
l Facility: R. E. Ginna Nuclear Power Plant l
l l Location: 1503 Lake Road Ontario, New York 14519 Dates: July 19 - 23,1999, and August 2 - 6,1999 Inspectors: Leonard Cheung, Team Leader Ram Bhatia, Reactor Engineer Thomas Burns, Reactor Engineer Roy Fuhrmeister, Senior Reactor Engineer Kenneth Kolaczyk, Reactor Engineer l
Approved by: Lawrence T. Doerflein, Chief Engineering Programs Branch Division of Reactor Safety l
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9909290314 990920 PDR ADOCK 05000244 G pm
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EXECUTIVE SUMMARY R. E. Ginna Nuclear Power Plant NRC Inspection Report 50-244/99-07 Introduction i An onsite engineering team inspection was conducted at the R. E. Ginna Nuclear Power Plant (Ginna) during the period of July 19 to August 6,1999. The overall objective of the inspection was to determine whether engineering was providing proper support for safe plant operations. The inspection included evaluation of 1) the circuit breaker maintenance program at Ginna; and 2) the implementation of the 10 CFR 50.59 safety evaluation program relating to changes, tests or experiments at the plan Maintenance j l
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The failure rate of the low-voltage circuit breakers at Ginna was higher than the generic industry failure rate due to previous inadequate root cause analyses and corrective actions to address the repeated breaker failures. The licensee recognized these breaker l issues and had established reasonable goals to improve breaker performance. The !
preventive maintenance procedures were improved. The material condition of the {
breakers was good. (Section M2.1)
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The licensee completed extensive troubleshooting for the recent circuit breaker failure The corrective actions taken and the root cause evaluations completed for the breaker failures were appropriate. (Section M2.2)
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The preventive maintenance program procedures for low voltage DB series circuit breakers had been improved and were found satisfactory. The licensee had made good progress in the breaker preventive maintenance activities, and had developed an appropriate plan for future breaker maintenance. (Section M2.3)
Enaineerina
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Design change modifications were properly designed and implemented. Affected documents were appropriately updated to capture and preserve the changes in the design basis documents. The safety evaluations provided sufficient bases to demonstrate that no unreviewed safety questions were involved in the modification The design change documents were well written and thorough. Supporting calculations generally presented good technical bases. The setpoint changes were adequately i evaluated and properly implemented. However, two NCVs were identified by the NRC, I one involved an incorrect design input to an motor-operated valve weak link analysis, l while the other involved not promptly implementing setpoint changes. (Section E1.1) l-ii-
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The temporary modifications were properly prepared and documented in accordance with the station procedures. The evaluation, installation, post-modification test requirements and safety reviews provided by engineering presented adequate technical basis for the modifications. There were no longstanding temporary modifications at Ginna. (Section E1.2)
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The design data for the systems and components and for the design change modifications were consistent with the Ginna licensing and design bases as specified in the technical specifications, and the UFSAR. The design data for the design change modifications were controlled, documented and incorporated into the appropriate design documents. Set point evaluations were thorough, in-depth, and technically soun (Section E1.3)
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Engineering had been effective in identifying and properly resolving technical issues. In addition, the design control, temporary modification, and corrective action procedures provided appropriate guidance for identification and resolution of technical issue (Section E2.1)
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The plant change process procedures provided appropriate guidance to the engineers for dissemination of design information. The implementation of the modification-follow-meetings and the system engineering group had resulted in improved communication of engineering information to other departments at the site. (Section E2.2) {
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Engineering backlogs were being managed effectively. The new work planning and tracking program being implemented was an enhancement for engineering workload management. (Section E2.3)
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Safety evaluation / review procedures were well-written documents that provided adequate guidance to determine if a proposed activity could be implemented without prior NRC approval. With some minor exceptions, the completed safety evaluations were comprehensive and thorough. The safety evaluation training provided to technical personnel was good. The plant operation review committee's review of a safety evaluation was also good. Process controls were in place to ensure changes to the plant were reflected in design documents. (Section E3.1)
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The quality assurance audits were thorough and in-depth, and resulted in good findings regarding engineering and procurement activities. The licensee adequately addressed the audit findings. (Section E7.1)
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The self-assessment performed by Ginna personnel in the area of records management produced meaningful findings for improvement. The independent assessments performed by outside contract organizations contributed significant findings regarding specific technical areas. The assessments were in-depth and of high technical quality, l
and were conducted using developed plans, with corrective actions initiated on all findings and recommendations. Overall, the licensee had a good self assessment program. (Section E7.2)
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The original auxiliary building post-accident environment calculation used non-l conservative assumptions and was a non-cited violation (NCV) of 10 CFR 50, Appendix l B, Criterion lil, Design Control, which requires measures to provide for verifying and checking the adequacy of the design. (Section E8.1)
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The failure to establish appropriate administrative controls over the supply of design l inputs to vendors was a NCV of design control requirements contained in 10CFR 50,
! Appendix B, Criterion 111. The licensee's immediate and long term corrective actions were comprehensive and were either completed or appropriately scheduled for completion in a reasonable time. (Section E8.5)
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TABLE OF CONTENTS
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EXEC UTIVE S U M MARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -ii-TAB L E O F C O NTE NTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -v-I I . M a i nte n a nce . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 M2 Maintenance and Material Condition of Facilities and Equipment (IP 92903) . . . . . . . 1 M2.1 Circuit Breaker Material Condition . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 M2.2 Corrective Actions and Root Cause Evaluations for Circuit Breaker Failures . 2 M2.3 Review of Maintenance Procedures and Maintenance Activities . . . . . . . . . . 4 Ill . E nginee rin g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 E1 Conduct of Engineering . . . . . . . . . . . . . . . . . . . ... .................. ..... 5 E Plant Design Change Modification Reviews . . . . . . . . . . . . . . . . . . . . . . . . . . 5 E1.2 Temporary Modifications Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .8 E1.3 Design Bases Reviews . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 E2 Engineering Support of Facilities and Equipment . . . . . . . . . . . . . . . . . . . . . . . . . . 10 l E Technical issues identification and Resolution . . . . . . . . . . . . . . . . . . . . . . 10 i
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E2.2 Engineering Interfaces With Other Departments . . . . . . . . . . . . . . . . . . . . . 11 E2.3 Engineering Backlog and Prioritization . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 E3 Engineering Procedures and Documentation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 E Review of the 10 CFR 50.59 Safety Evaluation Program . . . . . . . . . . . . . . . . 13 E7 Quality Ascurance in Engineering Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 E Quality Assurance (QA) Audits of Engineering Activities . . . . . . . . . . . . . . . 15 i E7.2 Self-Assessments of Engineering issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 I
E8 Miscellaneous Engineering issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 E (Closed) IFl 97-201-13: Auxiliary Building Post-Accident Environment ................................... .............. 17 E8.2 (Closed) IFl 98-13-01: Total Instrument Uncertainty (TIU)
C alculation s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 E8.3 (Closed) IFl 97-201-06: Electrical Calculation Discrepancies . . . . . . . . . . . . 18 E8.4 (Updated) IFl 50-244/98-07-01: Evaluation of the Licensee's 10 CFR50.54(F)
, Review Project . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 l
E8.5 (Closed) LER 1999-001-01: " Deficiencies in NSSS Vendor Steamline Break Mass and Energy Release Analysis Results in Plant Being Outside ofits Design Basis," Supplement 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 V. M anagement M eeting s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 X1 Exit Meeting Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21-v-
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Report Details
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l 11. Maintenance l M2 Maintenance and Material Condition of Facilities and Eauloment (IP 92903) 1 l
M2.1 Circuit Breaker Material Condition Insoection Scope (92903) ,
l The team reviewed the failure history and the sub-component replacement data of safety-related low-voltage circuit breakers to assess the breakers' material conditio The team also physically inspected selected breakers to determine the installed breaker conditions in the safety-related buse Observations and Findinos The 480 Vac safeguards buses at Ginna station were equipped with 600 Vac circuit !
breakers for safety and non-safety related applications. Three types of Westinghouse DB series (DB-75, DB-50, and DB-25) circuit breakers were installed on four safety related buses. There were a total of 52 safety related circuit breakers installed on these buses; some of these breakers also supply non-safety related load During the January 1999 NRC breaker inspection (98-13), the NRC noted that a total of eighteen breaker failures had occurred in the past five years. Eleven out of these failures had occurred on three specific breakers (five on the B-emergency diesel generator (EDG) output breaker (DB-75), and three each on the A and B service water (SW) pump breakers (DB-25)). These three breakers were removed from service and sent to Westinghouse for refurbishment. The licensee also significantly improved their breaker inspection and preventive maintenance (PM) procedures, including the troubleshooting and root cause analysis processes, through consultation with the vendor and other industry breaker personnel. At the conclusion of Inspection 98-13, approximately seventeen breakers had been inspected and had undergone PM using
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the newly improved procedures. No significant findings were identified, and therefore, the licensee believed that the breakers at Ginna in general were reliable and planned to inspect and perform PM on more breakers during the March / April 1999 outag During this inspection, the team noted that the three breakers that had been refurbished were re-installed in their respective cubicles and had been functioning properly, The team also noted that the licensee had experienced three additional breaker functional failures this year. In all cases the breaker failed to close on demand. Two of these breaker failures were due to failures of a sub-component (trip bar switch, discussed in the next section) which were unique at Ginn To determine the failure rate of each type of DB breaker, the licensee collected data of all functional failures of breakers and the number of demands imposed on all breaker The calculated functional failure rates were 1.74% for DB-75,0.13% for DB-50,0.70%
for DB-25, and 0.39% for all breakers. The team compared Ginna's breaker failure rate with the general industry (=0.2%), and found that the breaker failure rate at Ginna
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was higher than the industry average, especially for the Westinghouse DB breakers, which were supposed to be more reliable when compared with other breakers. The team also observed that approximately half of the breaker failures were attributed to three specific breakers that had been recently refurbishe The licensee expected that the implementation of the new inspection /PM program would significantly improve breaker performance. The licensee established an official goal (incorporated into their business plan) for future breaker reliability of 99.8% (consistent with the industry average) with no repeated breaker failure within five year The team reviewed the licensee's sub-component inspection and replacement data between January 1997 through April 1999 and found that the licensee had adequately replaced components such as secondary contact blocks, insulating links, control relays and some other components in all types of breakers. The team noted no unusual sub-components or mechanism being found out-of-tolerance that indicated wom-out mechanical or aged electrical component The team physically inspected a selected sample of breakers installed in the safety-related switchgear rooms and identified no concerns. The switchgear areas were found clean, well maintained, and the breakers were properly installed, with no broken or missing part Conclusions The failure rate of the low-voltage circuit breakers at Ginna was higher than the generic industry failure rate due to previous inadequate root cause analyses and corrective i actions to address the repeated breaker failures. The licensee recognized these breaker l Issues and had established reasonable goals to improve breaker performance. The 3 preventive maintenance procedures were improved. The material condition of the j breakers was goo M2.2 Corrective Actions and Root Cause Evaluations for Circuit Breaker Failures Insoection Scooe (92903)
The team reviewed the corrective actions taken by the licensee to resolve the breaker failures which occurrod this year to determine their adequacy. The team also reviewed I the licensee's root cause evaluations for these circuit breaker failures to assess their qualit Observations and Findinas The team noted that there were three circuit breaker functional failures which occurred in 1999. The first breaker (DB-50) failure occurred on March 31,1999, when the safety injection pump "1C2" breaker failed to close on demand during the refueling outage. The licensee's initial investigation found no abnormalities in this breaker; several subsequent tests showed that the breaker closed successfully. After extensive testing and consultation with Westinghouse, the licensee found that the cause of this breaker failure
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l was due to the improper adjustment of a trip bar switch (TBS) in the breaker. The )
licensee learned that this breaker configuration (with TBS) was of an old breaker design i and was unique to Ginna, and the TBS adjustment requirement was not described in the vendor manua While th9 investigation of the first failure was still ongoing, the second breaker failed on i April 6,1999, when the Bus 17-18 bus-tie breaker (52/17-18, also a DB-50) failed to
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due to an out-of-adjustment TBS. Further discussions with Westinghouse revealed that Ginna had two additional breakers (one DB-50 and one DB-75) with this unique TBS configuration. The licensee obtained an acceptable tolerance for the TBS for proper breaker operation and included this adjustment requirement in the PM procedures.
l After replacing the defective TBS in one case, and adjusting the other, the licensee found the two failed breakers functioned properly. The licensee inspected the other two breakers and found their TBS settings within the acceptable toleranc The third breaker failure occurred on June 8,1999, when the emergency diesel generator (EDG) "A" output breaker (DB-75) failed to close on demand. In spite of numerous tests (all successful), the licensee was unable to replicate the failure. During these tests, the technician observed a sluggishness in the movement of the trip bar. The technician also found that the trip bar did not consistently reset to the same position. The licensee finally found that the cause of this breaker failure was due to a slight l misalignment of the trip bar resulting from improperinstallation of the end supports. The licensee had previously (in 1998) hired Westinghouse to perform a PM (involving lubrication of the trip bar mechanism) at the Ginna site on this breaker. This PM activity (performed by Westinghouse personnel) required the disassembly (and later reinstallation) of the trip bar end supports. As this activity was typically performed by the vendor, the vendor manuals did not provide instructions for end support installation and verification. The licensee corrected the problem and returned the breaker to servic The licensee also revised their DB-75 PM procedure to include instructions for the proper installation of the trip bar end support In the ACTION Reports for these failures, the licensee attributed the first two breaker failures to be lack of adjustments of the TBS because the breaker PM procedure (and the vendor manuals) did not specify this requirement. For the third breaker failure, the licensee determined that the DB-75 breaker preventive maintenance procedure did not contain instructions nor were similar instructions specified in the vendor manual Therefore the slight misalignment of the trip bar was not observed by the technician The corrective actions taken by the licensee to prevent recurrence included: 1) updating breaker PM procedures to include trip bar switch adjustments and the trip bar misalignment concerns; 2) inspecting all other similar breakers to ensure no similar
, problems existed; and 3) performing electric shop training to review the failure mode of these breakers and corrective actions taken. For the four breakers that contained TBS, the licensee had initiated PCR 99-039, to replace the TBS with external relays to make them consistent with other breaker l Conclusion The licensee completed extensive troubleshooting for the recent circuit breaker failure l The corrective actions taken and the root cause evaluations completed for the breaker failures were appropriate.
M2.3 Review of Maintenance Procedures and Maintenance Activities Insoection Scooe (92903)
The team reviewed the station procedures for circuit breaker maintenance to determine their adequacy. The team also reviewed maintenance activities performed by the licensee on safety related breakers to evaluate the status of their maintenance progra Observations and Findinas The licensee established an inspection, a troubleshooting, and three PM procedures (one for each breaker type, GME-50-02-DB25, DB-50, and DB-75) for breaker maintenance activities at Ginn The team noted that these procedures were routinely updated to include known issues )
from the station and the industry. The team's review of these procedures indicated that i detailed instructions existed for all key components, such as verification of a center pad, l'
a weight requirement for the trip bar, mechanism operation, lubrication, and the high and reduced voltage testing of control coils. Attachment 3 of each procedure appropriately included the acceptable tolerance of each component in accordance with Westinghouse recommendation The team also verified that the licensee had appropriately updated the maintenance procedures to include all recently identified breaker-related maintenance issues. For example, steps 7.42.8,9 and 10 of DB-75 PM procedure, Revision 4, provided the required instructions to assure that the trip bar was properly aligned. In addition, the licensee also revised their EDG surveillance test procedures to include the requirements to locally observe proper closing, opening, and the trip bar reset position for all EDG breakers during surveillance testing by breaker maintenance personnel to ensure proper breaker operation. The team's review of the test results performed on June 9,1999, on I
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the EDG 'A" output breaker, indicated that the breaker was operated five times and the steps were checked each tim In addition, the licensee contracted Westinghouse for reviewing all DB series PM procedures for accuracy, including specified tolerances. The review was expected to be i
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completed by September 199 At the time of this inspection, the licensee had completed the inspection /PM on 45 of the 1 52 safety-related breakers. The remaining seven safety-related breakers were scheduled to be inspected and, if necessary, to complete the PM by the end of November 199 The team's review of selected work orders indicated no material related concerns. The licensee stated that after these inspection /PM activities were completed, the future PM j
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for the breakers would be as follows: both EDG output breakers would be scheduled on an 18 month basis; all motor breakers on a three-year basis except that the service I water pump breakers would be scheduled on a 12-month basis; and all feeder breakers and bus-tie breakers on a six-year basi i
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The team also reviewed the direct current (de) system voltage calculations (DA-EE-99-047 dated August 2,1999) to verify that the available voltage at each breaker control circuit was sufficient to operate the trip coil and the closing coil during the normal and abnormal condition I Conclusions I l
The preventive maintenance program procedures for low voltage DB series circuit breakers had been improved and were found satisfactory. The licensee had made good progress in the breaker preventive maintenance activities, and had developed an appropriate plan for future breaker maintenanc Ill. Enaineerina E1 Conduct of Engineering E1.1 Plant Desian Chanoe Modification Reviews Insoection Scooe (37550)
The team reviewed the modification procedures and the preparation and implementation of nine permanent plant modifications installed during outage and non-outage periods to verify conformance with the applicable requirements. The team also observed portions of the completed installation wor Observations and Findinas Permanent plant design changes at the Ginna facility were governed by interface procedure IP-DES-2, " Plant Change Process," which covered the processing of plant changes from design through close-out of the plant change record (PCR) package. The team's review of the modification procedures indicated that the procedures were clearly written and of adequate detail to provide appropriate guidance for the engineering organization to develop, document and capture the essential elements of the modificatio !
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The team reviewed the following nine safety-related modifications:
- 96-086, Modification to Reduce Potential for Pressure Locking of RHR Valves;- 97-014, MOV 4007 and MOV 4008 Actuator Replacements;- 98-034, Reactor Vessel Baffle Former Bolt Replacement:
- 98-041, Main Steam Hi Steam Flow Bistable Setpoint Change;
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_ SEV 1129, Control Room HVAC Upgrade - Phase I;- 98-043, Modify Diesel Generators A&B Voltage-Controlled Overcurrent Relays for Optimal Fault Detection;- 95-022, MOV 3996 Control Circuit Modification;- 96-081, MCC L Molded Case Circuit Breaker Replacements - 1999 Outage;- 97-063, Reactor Coolant Flow Transmitter Replacemen The team identified an error in the weak link analysis (Calculation No. 90170-07)
performed to support the increase in motor operator size for motor operated valves (MOV) 4007 and 4008 in the auxiliary feedwater system. The replacement of the motor operator required a calculation to determine the valve weak link component and thrust capacity based on a stress evaluation of all components significantly loaded during valve operation. The licensee used an incorrect bolt circle dimension in the calculation to evaluate the operator adapter flange bolting under thrust, torque, weight and seismic conditions. This error was not identified by the independent reviewer during the revie In addition, during the course of correcting this error, the licensee discovered that the bolt diameter used in the calculation was also incorrect. This failure to ensure control of design input for a safety related analysis is a violation of 10 CFR 50, Appendix B, Criterion lil, " Design Control." However, this Level IV violation of NRC requirements is being treated as a non-cited violation (NCV) consistent with Appendix C to the NRC Enforcement Policy (NCV 50-244/99-07-01). This condition was promptly captured by the licensee in their action reporting system. The licensee immediately performed an operability determination of the affected valves and corrected the calculation. The revised calculation confirmed the operability of the components. In addition, the responsible design engineer and the independent reviewer were counseled regarding this issue. The licensee stated that they planned to use this issue as an example in future training sessions. The team determined that the licensee's corrective actions were adequat PCR 98-041 was used to change the main steam high steam flow (total four instrument channels) setpoints, which did not have sufficient margin to incorporate the instrument uncertainties. The team's review of the modification package indicated that all setpoint changes were adequately evaluated and properly implemented, and that the affected test and calibration procedures were updated to reflect the new setpoints. However, the I licensee's resolution for this setpoint issue, which was initially identified in October 1994, was not timely. Although the licensee promptly documented this setpoint issue in a 1994 Potential Condition Adverse to Quality (PCAQ) report (94-067), the licensee failed to take prompt corrective actions at that tim in 1997, this setpoint problem was discovered again. The licensee's review of this issue determined that the bistable instrument trip setpoint (plus instrument uncertainty) could cause a condition prohibited by Ginna technical specifications. The licensee subsequently issued Licensee Event Report (LER)97-003 on September 29,1997, to report this issue. This LER was closed by the NRC in inspection Report 97-09. The immediate corrective action taken at that time was to place the bistables (providing permissive interlock functions only) in the tripped configurations, and the setpoints were finM/ corrected on August 8,199 The team's review of other 1994 PCAQ records indicated that there were other cases (such as94-051, EOP setpoint too low;94-063, high power trip design basis not known; and 94-073, flux deviation uncertainty not included) involving instrument setpoint issues where the licensee failed to take prompt corrective actions. This constituted a violation of 10 CFR 50, Appendix B, Criterion XVI,' corrective actions, which requires activities affecting quality to be promptly identified and corrected. However, this Level IV violation of NRC requirements is being treated as a non-cited violation (NCV) consistent with Appendix C to the NRC Enforcement Policy (NCV 50-244/99-07-02).
The licensee attributed the above deficient condition to be the inaction of an individual who had left RG&E According to the licensee, this individual was assigned to resolve numerous PCAQs at that time but failed to process them for years. The licensee stated that appropriate processes to prevent recurrence of this deficient condition were already in place: 1) the PCAQ process was replaced in 1995 by the Nuclear Operation Group Action Reports, which included an operability and reportabilky review by a senior reactor operator; 2) extensions of any due dates for the disposition or completion of an Action l Report require documented approval; 3) any overdue Action Reports are identified to management and the status of key items are periodically reported to the plant operations review committee (PORC). The team determined that the licensee's completed corrective actions and actions to prevent recurrence for this violation were adequate.
c. Conclusions
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Design change modifications were properly designed and implemented. Affected documents were appropriately updated to capture and preserve the changes in the design basis documents. The safety evaluations provided sufficient bases to demonstrate that no unreviewed safety questions were involved in the modification The design change documents were well written and thorough. Supporting calculations generally presented good technical bases. The setpoint changes were adequately evaluated and properly implemented. However, two NCVs were identified by the NRC, one involved an incorrect design input to an MOV weak link analysis, while the other involved not promptly implementing setpoint change .
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E1.2 Temoorary Modifications Review Insoection Sggen (37550)
The team reviewed a selected sample of temporary modifications (TM 4 tr Jetermine whether the TMs were properly designed and implemented and in compliance with regulatory requirements. The TMs were also reviewed to determira the extent of l
engineering involvement, quality of design inputs, implementatio. of safety evaluations, and post installation testing requirement Observations and Findinas The team reviewed thirteen TMs, including both open and closed TMs. The TMs reviewed were found to be properly designed and implemented, and in compliance with the Ginna procedure (IP-DES-3), which govemed the TM program. Appropriate conditions for removal of the TMs, and correction of the conditions being compensated were provided. Evaluations of the modifications were performed to determine their effect on plant safety, and licensing requirements. Appropriate post-modification testing, where necessary, was specified. The team also reviewed the status of all open TMs. As of July 1999, there were eight TMs open, one had been installed for 18 months, and the remaining were less than a year. The team noted that the open TMs were of minor safety significance. The team also physically inspected the TM for " Auxiliary Building Negative Pressure Control (99-002)" to verify compliance with the modification ,
requirements. The team observed that the installed condition of this TM was in compliance with the instructions provided in the temporary modification package and the testing requirements specified were verified each shift and documented by auxiliary operator Conclusions The temporary modifications were properly prepared and documented in accordance with the station procedures. The evaluation, installation, post-modification test l requirements and safety reviews provided by engineering presented adequate technical basis for the modifications. There were no longstanding temporary modifications at Ginn E1.3 Desian Bases Reviews Insoection Scope (37550)
The team reviewed three risk-significant systems and sub-systems, and selected safety-related design change modifications to verify conformance with the Ginna design and licensing bases as specified in the technical specifications (TS) and the Updated Final Safety Analysis Report (UFSAR).
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b. Observations and Findinas I i
I The three risk-significant systems and subsystems reviewed were: the residual heat reemoval system, components affecting pressurizer power operated relief valves (PORV)
closure following steam relief, and the refueling water storage tank (RWST) and its level instruments. Four design change modifications were also included in this revie The team's reviews included the piping and instrument drawings (P&lD), control schematics, wiring diagrams, and instrument data sheets to verify the design requirements specified in the UFSAR and the TS. The team selected design parameters for verification during a walkdown of the applicable portions of the systems associated with those parameter The design data for the three systems / subsystems reviewed were found to conform, in all respects, to their design bases and licensing requirements as specifitnj in the technical specifications. In addition, the team independently calculated the RWST volume, and the setpoint data for the PORVs and for the RWST instruments, based on the data provided by the P&lDs, vendor drawings, and design specifications. These calculations confirmed that the as-built conditions and the instrument setpoints were consistent with the TS, and the UFSAR. The team also reviewed four instrument setpoint evaluations and noted that these set point evaluations were thorough, in-depth, and technically soun For the design change modifications, the team examined the change impact evaluations, design verification data sheets, safety evaluations, drawing update notices and associated revised drawings, and found that these design changes conformed with the UFSAR and the licensee commitments. The team also found that the design data for the modification changes were controlled, documented and incorporated into the appropriate design documents c. Conclusions The design data for the systems and components and for the design change
'aodifications were consistent with the Ginna licensing and design bases as specified in .
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the technical specifications, and the UFSAR. The design data for the design change modifications were controlled, documented and incorporated into the appropriate design documents. Set point evaluations were thorough, in-depth, and technically soun _- ---
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E2 Engineering Support of Facilities and Equipment
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E2.1 Technical Issues identification and Resolution 4 Insoection Scoos (37550)
The team reviewed corrective action program procedures to evaluate controls on non-conforming items and problem resolution. The team also reviewed several permanent and temporary modifications to evaluate the manner in which they resolved the original technicalissue.
. Observations and Findinas IP-CAP-1, Rev. 9, " Adverse Condition or Tracking Initiation or Notification (ACTION)
Report," provides the requirements for the Nuclear Operations Group corrective action program. The procedure provides the method for identification, documentation, notification, segregation, evaluation, disposition, correction, trending and reporting conditions, events, activities, and concems that have the potential for affecting the safe and reliable operation of the plan IP-DES-2, Rev 6, " Plant Change Process," provides guidance for the design, implementation, testing, and administrative control of permanent modifications to the statio IP-DES-3, Rev. O, " Temporary Modifications," provides the process for evaluation, installation, and removal of temporary, minor alterations to plant equipment, components, or systems which do not conform with approved drawings or other design document The team reviewed the following permanent and temporary modifications to evaluate the manner in which the originating technical issue was resolve Permanent Modifications98-043, Modify Diesel Generator Voltage-Controlled Overcurrent Relays for Optimal Fault Protection 95-022, I Aour-Operated Valve 3996 Control Circuit Modification 97-063, Reactor Coolant Flow Transmitter Replacement Temoorary Modifications99-022, Change Rod 111 Data Path to use Train Bus B 97-024, Rev 5, Disable Core Exit Thermocouple Point T23/H13 ~
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The team determined that not all of the technical issues addressed by the modifications were deficiencies. Several were enhancements to the plant design to provide greater flexibility to the operating staff or to limit financial consequences of an electrical faul Nonetheless, for the modifications reviewed, the team determined that the licensee's engineering organization performed appropriate evaluations of the original technical
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issue, developed adequate corrective or compensatory measures, and correctly implemented those measures in the plan Conclusions Engineering had been effective in identifying and properly resolving technical issues. In addition, the design control, temporary modification, and corrective action procedures provided appropriate guidance for identification and resolution of technical issue E2.2 Engineerina Interfaces With Other Departments Ln13pection Scope (37550)
The team reviewed work orders and plant configuration change folders associated with several permanent modifications to evaluate engineering interfaces with other departments regarding design information, in addition, the team interviewed a planner (Maintenance Department) and a member of the training department regarding their interfaces with engineerin , Qh.servations and Findin.gg Procedure IP-DES-2, " Plant Change Process," specified the duties and responsibilities of the responsible engineer with regard to: 1) determining the interfaces and support required for the design development; 2) ensuring that support personnel were involved in the design; 3) supporting the implementing activities for the change; 4) supporting the development of the modification package; 5) reviewing instructions for performing the work; and 5) coordinating with the Training Department for resolving training issue Discussions with a planner and a member of the Training Department revealed that the transfer of information had improved since the reorganization of the engineering department. They also stated that the implementation of the modification-follow-meetings had greatly improved the transfer of information. Engineers routinely contacted the work planners early in the development of the modification to determine what guidance needs to be used to do the field wor Conclusion The plant change process procedures provided appropriate guidance to the engineers for dissemination of design information. The implementation of the modification-follow-meetings, and the system engineering group had resulted in improved communication of engineering information to other departments at the sit l
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E2.3 Enaineerina Backloa and Prioritization
. Inspection Scooe (37550)
The team reviewed information, including backlogs, regarding the performance of the engineering department, and interviewed a computer tracking specialist to determine the status of the new work planning and scheduling process being implemented in the engineering departmen Observations and Findinas The team determined that since January 1996, the licensee has achieved a reduction of more than one third in the number of open engineering work items. This reduction occurred in spite of a reduction in staffing in the engineering departmen The new work planning and scheduling program, " Work Management 1.84," had not been fully implemented at the time of the inspection, implementation in all the work groups within the department was expected to be complete by the end of August 199 The new planning and scheduling program is integrated with the licensee's 12-week work schedule. This allows easier determination of due dates based on when work is expected to be performe Priorities are currently assigned by the engineering manager. The licensee has recently instituted bi-weekly meetings of department managers for the purpose of reviewing and prioritizing outstar. ding engineering work. The new work planning and tracking program simplifies the prioritization by tying "needed by" dates to the 12-week work schedul The new program also enables engineers and supervisors to evaluate the projected work load based on estimated time needed to complete an item. This also allows re-evaluation of priorities on a more realistic basi Conclusion !
Engineering backlogs were being managed effectively. The new work planning and l tracking program being implemented was an enhancement for engineering workload 1 management.
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E3 Engineering Procedures and Documentation E Review of the 10 CFR 50.59 Safety Evaluation Proaram Insoection Scone (37001)
The team reviewed the Ginna Safety Evaluation program, including safety evaluation training. The team also examined the procedures that described how safety evaluations and safety reviews should be prepared, developed, and approved. A sample of completed safety evaluations was also examined, along with plant changes that the licensee had implemented without a formal safety evaluation. Lastly, the team attended two Plant Operations Review Committee (PORC) meetings to determine how safety evaluations were reviewed and accepted by the committe Observations and Findinas Processes. Procedures and Trainina The Ginna safety evaluation program utilized a two-step screening process to determine if a proposed change would require prior NRC approval before implementation. First, using the guidance contained in Procedure IP-SEV-1, " Preparation, Review and Approval of Safety Reviews," Revision 4, dated October 15,1998, the proposed change was evaluated against screening criteria to determine if a safety evaluation should be prepared. If a safety evaluation was not required, NRC pre-approval of the change was also not required. If a safety evaluation was required, IP-SEV-1 directed the preparer of the change to perform the requisite evaluation, using procedure IP-SEV-2 " Preparation, Review, and Approval of 10 CFR 50.59 Safety Reviews," Revision 5, dated October, 1998. Procedure 1P-SEV-2 had criteria, similar to the wording contained in 10 CFR 50.59, that the preparer of the safety evaluation used to evaluate whether the proposed change could be implemented without prior NRC approva Both initial screening and safety evaluations required independent reviews. In addition, as required by the Ginna Quality Assurance program, each safety evaluation required a review by the POR Both procedures were well-written, easy-to-follow documents, that had captured the intent of 10 CFR 50.59 and other NRC guidance regarding performance of safety evaluations. If the proposed change would effect the description of the plant as outlined in the Updated Final Safety Analysis Report (UFSAR), both procedures had appropriate controls to ensure the UFSAR was revise Six months following each refueling outage, as required by 50.59(b)(2), the licensee submitted to the NRC a report containing a brief description of all changes, tests and experiments performed at Ginna along with a summary of the pertinent safety evaluations. The last report was submitted on schedule on May 21,199 _ _ _ _ _ _ _ _ _ _ _ _
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The licensee provided training to technical personnel on how to correctly perform a safety evaluation / review. The program was similar to training programs provided at other facilities in the industry in that it included both initial and refresher training module The initial training was reinforced by blannual four-hour refresher courses. Augmenting the training program was a series of " read and sign" newsletters that informed safety evaluation preparers of industry developments in the safety evaluation program are Safety Evaluations Thirteen safety evaluations that discussed proposed changes of varying complexity were reviewed by the team. The team verified that plant design documents, including the Technical Specifications (TS) and UFSAR had been updated to reflect the changes brought about through implementation of the safety evaluation The team found the safety evaluation which accompanied the proposed change was generally thorough and complete, and the text which had accompanied the evaluation <
provided adequate justification as to why the change did not involve an unreviewed l safety question (USQ). However, some discrepancies were noted by the team. For example, safety evaluation SEV-1133, "Minium Auxiliary Feedwater Temperature of 32 Degrees," which proposed reducing the minium temperature of water in the condensate storage tank from 50 to 32 degrees Fahrenheit, did not describe why freezing of the tank and accompanying branch lines was not a possibility. Although a subsequent followup of this issue by the team revealed this possibility was unlikely due to the location of the tank (in a large room next to a hot water heater), mentioning this fact in the description of the change would have improved the overall quality of the evaluation. The licensee stated that 32*F was chosen to be consistent with the other related analyse Plant Operations Review Committee Activities The team attended two PORC meetings covering primarily the reviews and dispositions of completed ACTION Reports (AR). Only one safety evaluation was reviewed by the PORC. The team note that the PORC's review of the safety evaluation was goo Conclusions Safety evaluation / review procedures were well-written documents that provided adequate guidance to determine if a proposed activity could be implemented without prior NRC approval. With some minor exceptions, the completed safety evaluations were comprehensive and thorough. The safety evaluation training provided to technical 1 personnel was good. The PORC's review of a safety evaluation was also goo i Process controls were in place to ensure changes to the plant were reflected in design document E7 Quality Assurance in Engineering Activities E7.1 Quality Assurance (QA) Audits of Enaineerina Activities Inspection Scope (37550)
The team reviewed two QA audits in the engineering and the procurement areas to evaluate the effectiveness of these audits related to the engineering activities. The team also interviewed four auditors to assess their knowledge in the audited area Observations and Findinas The team reviewed Nuclear Directive Procedure ND-ASU, " Audits and Surveillance,"
Revision 4, dated April 14,1999, which delineated the requirements of the QA audit activities, personnel qualification, audit frequency, and the process to follow-up actions and review of responses from other organizations to assure that the identified issues were properly resolved and the corrective actions were tracked to ensure completion by the target date The team reviewed the two QA audits performed in 1998. The first QA Audit (AINT-1998-0008-JMT) which was conducted from May 11,1998, to June 15,1998, covered the engineering and configuration control activities. The audit team consisted of several QA personnel and a technical specialist from another utility. This audit identified five areas that needed improvement: 1) updating of design basis information into the UFSAR; 2) quality of license amendments; 3) quality of safety evaluations; 4) late verification of completed tests; and,5) updating the configuration control documentation record. The corrective actions for three of the findings were completed, and the corrective actions for the other two were being evaluated by QA department and were being tracked for closure upon completio The second QA audit (AINT-1998-0005-TGT) was conducted from April 13-30,1998, on procurement activities. The team noted that in this audit, a technical specialist familiar with the procurement process from another utility participated with six in-house QA members. The audit assessed the effectiveness and adequacy of the procurement processes, such as procurement evaluations, document control, control of purchase items and services, including the handling shipping and storage of procured items. The purpose of this audit was to determine whether the procurement department at Ginna was capable of handling the responsibilities, since they had recently acquired the full control from their contractor Global Supply Group for these responsibilitie This QA audit resulted in the issuance of six Action Reports, indicating weaknesses in the areas of chemical control, procedural adequacies, and concems of storage of unidentified materialin the store rooms. The team reviewed selected licensee corrective actions for these findings and determined that the licensee had appropriately resolved the issues with the procurement proces ,f '
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l 16 Conclusions The quality assurance audits were thorough and in-depth, and resulted in good findings regarding engineering and procurement activities. The licensee adequately addressed the audit finding E7.2 Self-Assessments of Enaineerina Issg3 Inspection Scope (37550)
The team reviewed the self-assessment procedure to determine if adequate guidance was provided for an in-depth self-critical examination of engineering issues. The team also reviewed completed self-assessments of engineering issues and the licensee responses to the assessment findings to determine the effectiveness of the assessments. The review covered one peer-assisted self-assessment and supplemental assessments completed by contractors engaged by the licensee to address specific technicalissue Observations and Findinas The responsibilities and process guidance for Ginna assessment teams is clearly delineated in self-assessment procedure IP-SEP-2. Detailed guidelines are provided for the preparation and conduct of the assessment such that the process was used to assure the area assessed meets quality and performance standards. The NRC inspection team determined that the guidance provided in the self-assessment procedure was adequate for the implementation of the progra i The licensee utilizes contractors to a large extent to supplement assessment activitie The team reviewed one self-assessment performed by licensee personnel and eight independent assessments performed by contractors. The one self-assessment was of !
the records management area (retention and retrievability). Independent assessments ,
were performed in the engineering area. The independent assessments were performed I without active participation by Ginna personnel. The review was conducted to determine the extent and effectiveness of self-assessments in identifying areas of strength and weakness and, the disposition of any findings. The team noted that the self-assessment and the independent assessments performed were conducted in detail and that numerous findings were identified. The team selected two independent assessment reports considered significant to evaluate the licensee corrective action on the findings identified. All findings identified in both assessments were placed in the licensee's action tracking system for determination of priority, screening for safety evaluation and disposition for corrective action and closur L
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The team noted that the independent assessments performed by contractors were focused on specific technical issues rather than on the'overall performance of department activities. As a result of this intense evaluation of a specific technicalissue, the assessment did not reveal the status of department compliance with standards, practices and expected performance levels. The assessments produced significant findings in the specific technical are Conclusions The self-assessment performed by Ginna personnelin the area of records management produced meaningful findings for improvement. The independent assessments performed by outside contract organizations contributed significant findings regarding specific technical areas. The assessments were in-depth and were of high technical quality, and were conducted using developed plans, with corrective actions initiated on all findings and recommendations. Overall, the licensee had a good self assessment progra E8 Miscellaneous Engineering lasues E8.1 (Closed) IFl 97-201-13: Auxiliary Building Post-Accident Environment. This item was updated in October 1998 inspection (98-10). During the August 1997 design inspection, the NRC identified several non-conservative assumptions in the post-accident thermal environment calculation, " Engineering Evaluation of R. E. Ginna Nuclear Power Plant Ventilation System," Revision 1. These non-conservative assumptions included: the initial auxiliary building temperature of 85*F was used instead of the maximum design temperature of 104*F; the water temperature of 80*F for the refueling water storage tank (RWST) was used instead of the design temperature of 104*F; and the calculation did not considered the effect of the design basis 50 gpm seal leak from the residual heat removal (RHR) pump at 155'F. In December 1998, the licensee subcontracted Altran Engineering to perform the new calculation using the conservative input data. The preliminary calculation results indicated that the RHR pit could have a post-accident temperature of 165.5'F. There were two RHR pump motors, a sump pump motor and a level switch used to control the sump pump that required environmental qualification (EQ) in the RHR pit. The team's review of the EQ files indicated that the level switch and one of the RHR motors were not qualified to 165.5*. However, there was sufficient evidence that these two components could be qualified to the required temperature through analyses. The licensee agreed to complete the new auxiliary building post-accident thermal environment calculation by October 15,1999 and revise their EQ files accordingly by November 12,1999. The licensee also agreed to review other EQ files j and to correct the aging calculations for any affected EQ components in the auxiliary building by March 31,2000. The completion of these actions was being tracked by the licensee in their Commitment and Action Tracking System (Tracking Nos. R06468, R08000, and R08001).
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The team considered this item closed. The team determined that the original auxiliary building post-accident environment calculetior: was in violation of 10 CFR 50, Appendix B, Criterion lil, Design Control, which requires design control measures to provide for verifying and checking the adequacy of the design (calculation). However, this Level IV violation of NRC requirements is being treated as a non-cited violation (NCV) consistent with Appendix C to the NRC Enforcement Policy (NCV 50-244/99-07-03)
E8.2 (Closed) IFl 98-13-01: Total Instrument Uncertainty (TIU) Calculations. During the December 1998 inspection, the NRC inspectors observed that there were several instruments important to safety that did not have TIU calculations, and therefore, the instrument setpoints did not have sufficient margins to account for the TIU. One example was the sodium hydroxide spray additive spray tank level transmitter which had a as-found setting of 4.25% out of specification at the high end of its bank. The licensee completed an evaluation and determined the instrument to be operabl Following the inspection findings, the licensee hired a consultant team to perform an independent assessment of Ginna's overall Tlu program. The assessment results were documented in a report entitled " Independent Assessment of Total Instrument Uncertainty," dated May 14,1999. The assessment identified several areas (instrument drift methodology, surveillance test acceptance criteria, additional personnel training, and program plan development) that needed improvement. The assessment team also developed a list entitled "Ginna Total Instrument Uncertainty Matrix", which contained about 400 instrument data points for evaluation. The team's review of the assessment report and the instrument matrix found them to be extensive. The team also reviewed four instrument loop accuracy calculations (one for pressurizer level, one for reactor coolant flow, and two for refueling water storage tank levels) that were completed in 1998, and found the calculations technically soun The licensee completed a project plan entitled "Setpoint Control and Verification Project,"
dated July 17,1999. This project plan identified in detail the tasks and sub-tasks that needed to be completec , and contained a schedule for the completion of each tas Section 1,5,1 established the criteria for setpoints that required TIU calculations. Table 3 identified the setpoints that required new TIU calculations or justifications why calculations were not needed. The licensee assigned a project manager to oversee the completion of this project, and retained the consultants who had performed the independent assessment to work on this project, which is scheduled to be completed by mid-January 2000. The team's review of the project plan indicated the plan to be logical and thorough, and the schedule to be reasonable. The team determined that the licensee had established an appropriate instrument setpoint evaluation project and schedule. This item is close E8.3 (Closed) IFl 97-201-06: Electrical Calculation Discrepancies. This item was updated in a previous NRC inspection (98-13). At that time, the licensee was in the process of developing the de load study to address the calculation discrepancie I l
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During this inspection, the team noted that the licensee had completed the de system <
! load calculations and voltage analyses (DA-EE-99-047 dated August 3,1999) of de l l components. The team reviewed a sample of input data of cable impedances, components load requirements, and the assumptions made in these analyses and found them to be conservative and consistent with the components' vendor requirements . The team also verified that the licensee had appropriately used the minimum battery voltage
, of 108.6 Volts that was available during the last period of a postulated ste%n blackout l condition to demonstrate the required component performance. The ove V results l indicated that all safety-related loads and control components would have Wequate l voltage to perform their intended design function.
l The licensee also completed the de system fault current analysis (DA-EE-99-013) on February 18,1999. The team's review of this analysis indicated that the de system components were capable of withstanding the maximum short circuit currents available in class 1E de distribution system The licensee also completed a preliminary analysis of de system fuse coordination analysis (DA-EE-99-066). The preliminary results indicated that all fused non-class 1E l loads were properly isolated from the class 1E de distribution system. The team l determined that the licensee had revised all major de system calculations, indicating good progress. The minor analyses that had not yet been completed were being adequately tracked by the licensee, and were scheduled to be completed by September 1999. This item is close E8.4 (Uodated) IFl 50-244/98-07-01: Evaluation of the Licensee's 10 CFR50.54(F) Review Project. This item was opened to track the status of the licensee's design management program whose goals included optimizing their engineering configuration management / control processes, and validating current engineering assumptions contained in design documents. In keeping with these goals, eleven projects were underway, including a UFSAR verification effort, calculation control verification and validation initiative, and development of plant system design basis documents. The projects hac varying end-dates for completion with some projects, specifically the calculation control project, not finishing until January 200 At the time o) thr, inspection, some of the projects, including the UFSAR verification projects, were behind schedule. This fact was identified by the licensee management,
.and initiatives were underway to get the projects back on schedule. The effectiveness of these actions had yet to be determine The NRC has not yet established a definitive date when the design basis verification !
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efforts underwcy in the industry should be completed, nor has the NRC established criteria that could be used to measure the progress of these programs. However, at a ,
June 3,1999, " Licensing issues Workshop" meeting held in Philadelphia, PA., the NRC project manager for the UFSAR update program indicated that the NRC is considering establishing a March 30,2001, completion date for the design basis verification i initiative U , . . .
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l If this date is in fact adopted by the NRC, some of the design validation work that is l underway at Ginna may need to be accelerated if the March 2001 completion date is going to be met. This item remains open pending further review of licensee's progress in this are E8.5 (Closed) LER 1999-001-01: " Deficiencies in NSSS Vendor Steamline Break Mass and l Energy Release Analysis Results in Plant Being Outside of its Design Basis," I Supplement 1. LER 1999-01 identified there were modeling errors in the accident analysis for fuel cycle 27. These errors effected the predicted peak containment pressure following a main steam line break (MSLB) inside the containment, with an assumed failure of a main feedwater regulating valve (FRV). One modeling error involved failing to account for approximately 1000 gallons of high temperature feedwater in a length of pipe between the FRV and feedwater block valve. The second error concerned under-predicting the total time to isolate the feedwater syste The errors were discovered by Westinghouse, the supplier of the accident analysis, and
- were officially reported to the licensee in a February 22,1999, Westinghouse letter to i RG&E. The licensee's corrective actions were first discussed in NRC inspection report 50-244/99-02 and later in NRC inspection report 50-244/99-03. In the supplement to LER 1999-001, the licensee noted that in addition to cycle 27, these errors also affected the accident analysis for fuel cycles 12 through 2 !
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The team noted the modeling errors were caused by the failure of Westinghouse and licensee personnel to ensure the modeling code used correct inputs. For example, when i modeling the Ginna feedwater system, the Westinghouse analyst did not use the Ginna ;
plant specific feedwater isolation times. Absent use of plant specific information, when !
the analyst ran the code, it automatically defaulted to less conservative generic isolation I values. Similarly, the licensee did not supply Westinghouse with the correct volume of water in the feedwater system to use in the analysi To ensure vendors receive and use correct design inputs in the future, the licensee strengthened the process controls regarding how information was supplied to vendor The process by which information was supplied to a vendor was proceduralized, and the design inputs supplied by the licensee to the vendor required an independent review before being sent to the vendo The failure to establish appropriate administrative controls over the supply of design inputs to vendors is a violation of design control requirements contained in 10CFR 50, Appendix B, Criterion Ill, Design Control. This issue was previously entered in the corrective action program and the licensee's immediate and long term corrective actions were comprehensive and were either completed or appropriately scheduled for completion in a reasonable time. This Severity Level IV violation is being treated as a Non-Cited Violation consistent with Appendix C of the NRC Enforcement Policy (NCV 50-244/99-07-04). This LER supplement is close _
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V. Manaaement Meetinas (
X1 Exit Meeting Summary j The team met with the licensee personnel at the conclusion of the inspection on August 6,1999, ,
and summarized the scope of the inspection and the inspection results. The licensee did not !
dispute the inspection findings at the meetin PARTIAL LIST OF PERSONS CONTACTED Licensee P. Bamford Manager, Reactor Engineering and Analysis R. Davis QA Engineer K. Deisencoth Electrical /HVAC G. Graus Manager, l&C/ Electrical Maintenance !
J. Guider I&C/ Electrical System Engineer l
T. Harding Senior Licensing Engineer T. Marlow Manager, Nuclear Engineering Services R. Meeredy Vice President, Nuclear Operations T. Miller System Engineer K. Muller Primary System Engineering J.Pacher Manager, l&C/ Electrical Engineering J. Smith Maintenance Superintendent J. Traynor Senior QA Analyst J. Widay Plant Manager G. Wrobel Nuclear Safety & Licensing Manager NRC B. Holian Deputy Director, DRS I C. Osterholtz Resident inspector J. Yerokun Chief, Engineering Support Branch, DRS INSPECTION PROCEDURES USED IP 37001: 10 CFR 50.59 Safety Evaluation Program IP 37550: Engineering IP 92903: Follow-up -Engineering i
ITEMS OPENED, CLOSED, AND DISCUSSED Opened NCV 50-244/99-07-01 Incorrect input for MOV Weak Link Analysis NCV 50-244/99-07-02 Untimely Corrective Actions for Instrument Setpoint Changes NCV 50-244/99-07-03 Auxiliary Building Post-Accident Environment NCV 50-244/99-07-04 Steamline Break Mass and Energy Release Analysis Closed IFl 50-244/97-201-13 Auxiliary Building Post-Accident Environment IFl 50-244/98-13-01 Total Instrument Uncertainty (TIU) Calculations
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IFl 50-244/97-201-06 Electrical Calculation Discrepancies i NCV 50-244/99-07-01 Incorrect input for MOV Weak Link Analysis NCV 50-244/99-07-02 Untimely Corrective Actions for Instrument Setpoint Changes I NCV 50-244/99-07-03 Auxiliary Building Post-Accident Environment NCV 50-244/99-07-04 Steamline Break Mass and Energy Release Analysis LER 1999-001-01 Steamline Break Mass and Energy Release Analysis Updated IFl 50-244/98-07-01 Evaluation of the Licensee's 10 CFR50.54(F) Review Project i
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LIST OF ACRONYMS USED 50.59 Title 10 of the Code of Federal Regulations, Section 50.59 ACTION . Adverse Condition Tracking Initiation or Notification AR ' ACTION Report CFR Code of Federal Regulations
.CTEs Changes, Tests and Experiments de Direct Current EDG Emergency Diesel Generator EQ Environmental Qualification EWR Engineering Work Request F Fahrenheit FRV Feedwater Regulating Valve l&C Instrumentation and Control IFl Inspection Follow-up item IST In-service Test LER Licensee Event Report LOCA Loss of Coolant Accident MOV Motor Operated Valve MSLB Main Steamline Break NRC Nuclear Regulatory Commission P&lD Piping and instrument Drawing PCAQ Potential Conditions Adverse to Quality PCR Plant Change Record PDR Public Document Room PM Preventive Maintenance PORC Plant Operations Review Committee PORV Power Operated Relif Valve PSIG Pounds per Square Inch, Gage QA Quality Assurance RG&E Rochester Gas and Electric Corporation RHR Residual Heat Removal RV Relief Valve RWST Refueling Water Storage Tank SW Service Water TBS Trip Bar Switch TIU TotalInstrument Uncertainty TMs Temporary Modifications TS Technical Specifications UFSAR Updated Final Safety Analysis Report URI Unresolved item USQ Unreviewed Safety Question Vac Volts - Alternating Current i
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