IR 05000361/1998013
| ML20239A446 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 09/01/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20239A442 | List: |
| References | |
| 50-361-98-13, 50-362-98-13, NUDOCS 9809090182 | |
| Download: ML20239A446 (24) | |
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! ENCLOSURE i
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U.S. NUCLEAR REGULATORY COMMISSION-
REGION IV
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Docket Nos.: 50-361 50-362
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License Nos.: NPF-10 NPF-15 Report No.: 50-361/98-13 .
I 50-362/98-13 Licensee: Southern Califomia Edison Co.
j Facility: San Onofre Nuclear Generating Station; Units 2 and 3 ~
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l ' Location: 5000 S. Pacific Coast Hw San Clemente, California j Dates: July 12 through August 22,1998
- inspectors
- J. A. Sloan, Senior Resident inspector '
t J. G. Kramer, Resident inspector J. J. Russell, Resident inspector l.
l - Approved By: Dennis F. Kirsch, Chief, Branch E Division of Reactor Projects i
ATTACHMENT: SupplementalInformation L ,
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9809090182 980901 -
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l t-2-l EXECUTIVE SUMMARY San Onofre Nuclear Generating Station, Units 2 and 3 NRC Inspection Report 50-361/98-13; 50-362/98-13 This routine announced inspection included aspects of licensee operations, maintenance, engineering, and plant support. This report covers a 6-week period of resident inspectio !
Ooerations
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Licensed operators failed to recognize all applicable technical specification (TS) limiting conditions for operation (LCOs) when an emergency chiller failed, indicating incomplete l review of applicable TS. Also, control room logc did not reflect late entry into an ;
applicable TS LCO after the inspectors identified the error. The operators did enter the most direct TS LCO, such that TS LCO actions times would not have been exceeded (Section 01.2).
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The licensee's controls for the infrequently performed evolution of removing both trains of spent fuel pool (SFP) cooling from service were excellent. A control operator demonstrated attentiveness in identifying the potential for slightly diluting the SFP, which would have been contrary to the licensee's procedures. Monitoring of the SFP was l
axcellent (Section 01.3). '
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Licensee maintenance of radioactive floor drains in the common radwaste building and ,
some of the Unit 3 emergency core cooling system (ECCS) pump rooms was deficien l Rubber devices used to prevent gas from exiting the drains had deteriorated and debris had been allowed to accumulate in the drains. There was no safety consequence to the !
partially blocked drains (Section O2.1).
. i The licensee had generally been proactive in preserving the firemain system i (Section O2.1).
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The 480 volt circuit breaker procedure was not written in the format as recommended by the goveming procedure for writing procedures and the action request (AR) written to address an incorrectly racked out 480 volt breaker failed to identify the procedure weaknesses. This breaker was seismically restrained while in the incorrectly racked out position. Corrective actions to address operator performance when racking in and out breakers were thorough and included supervisory oversight of breaker racking evolutions (Section 08.1). ,
Maintenance
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Machinists' attention to procedural direction while adjusting charging pump crosshead bearings and then repacking the charging pump was weak. The applicable procedure contained a note that appeared to prohibit the sequence of work as performed. The work was actually performed technically correctly, and the procedural note was in error.
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-3-However, the procedural error was not noted or corrected until pointed out by the inspectors (Section M1.3).
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Operations test group personnel performance during a low pressure safety injection (LPSI) pump inservice test was good. The personnel were knowledgeable about the test process and methodology to perform the test. In addition, good ALARA and industrial safety practices were~ utilized. A test weakness was identified, in that the LPSI inservice test procedure criteria for declaring the pump inoperable based on seal leakoff was nonconservative and did not match the analysis assumed in the updated final safety analysis report (UFSAR). This had minimal safety consequence (Section M1.4).
Maintenance personnel displayed a knowledge weakness when installing an SFP cooling system valve actuator 180 degrees off from the required position as stated in the maintenance order. This was identified by an NRC ' inspector while work was in progress (Section M4.1).
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All 41 recently completed operability assessments (OAs) reviewed adequately demonstrated the basis for operability. Recently implemented expanded guidance for completeness of the OAs resulted in general improvements in the quality of toe OA However, some OAs contained minor deficiencies that required further explanation to complete the logical basis for operability (Section E2.1).
Engineering support for the core operating limits supervisory system (COLSS) software change to implement a reactor power calculation based on a steam calorimetric was excellent, including contracting vendor support to address a potential problem, and developing test cases to ensure the validity of the software. The faciiity change evaluation and the software modification report fully addressed all relevant aspects of the change. Finalimplementation was successful (Section E2.2).
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Nuclear Oversight demonstrated aggressive, independent oversight of the fuel fabrication process, and took appropriate action in issuing a stop work order to the fuel j vendor pending resolution of a quality problem that resulted in stainless steel
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contamination of welds in the Zircaloy grids (Section E7.1). 1
A noncited violation of 10 CFR Part 50, Appendix B, Criterion Ill, was identified by the ;
licensee as the result of discovering that the voltage at three 120 VAC loads did not
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comply with the design information in the UFSAR. The condition existed since original I plant construction, but the licensee missed an opportunity to identify the condition when I performing calculations in support of design basis reconstitution in 1996. The licensee's l efforts in ultimately identifying, and correcting, the deficiency were excellent !
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4- 1 Plant Suonort
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Health Physics support of maintenance work on a charging pump was excellent (Section M1.3).
Health Physics support of the maintenance evolutions during an SFP cooling system outage was excellent in that the technicians aggressively monitored the activities and established appropriate contamination boundaries (Section 4.1).
The licensee's conduct of the after-hours emergency preparedness drill was excellen One exception to staffing the technical support center (TSC) within the timeliness goals of the drill was identified by the licensee. Overall communications in the TSC were good (Section P1.1). - -- - - - - -
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Report Details Summary of Plant Status Unit 2 began this inspection period returning to full operating power after operating at 80 percent during heat treatment of the circulating water system. The unit remained at essentially full power from July 13,1998, until the end of this inspection perio Unit 3 operated at essentially 100 percent power throughout this inspection period, with the exception of July 25-26,1998, when the unit operated at 80 percent power during a heat -
treatment of the circulating water syste . Operations '
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01.1 General Comments (71707) j The inspectors observed routine and nonroutine operational activities throughout this l inspection period. Some of the activities observed included:
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control element assembly position verification surveillance (Unit 3) !
preparation for isolating both trains of SFP cooling (Unit 3)
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special briefing for control of an infrequently performed evolution during operation with both trains of SFP cooling out of service (Unit 3)
Operators were thorough and methodicalin preparing for and conducting routine evolutions. Close management and supervisory oversight of operational activities were evident. Procedure use and operator communications were consistent with written performance expectations. Specific comments on activities are discussed belo i O1.2 Recognition of TS LCO Acolicability - Units 2 and 3
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The inspectors reviewed TS LCO entry when control room operators declared Emergency Chiller E336 inoperable on July 22,1998, due to low reservoir oil temperature with the chiller operatin l Observations and Findinas
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At 5:00 p.m. on July 22,1998, control room operators declared Emergency Chiller E336
!- inoperable and entered TS LCO 3.7.10 for both units, which requires a unit shutdown in 7 days when one chiller is inoperable. The inspectors reviewed TS 3.i'.11, which requires a unit shutdown in 7 days when one control room essential air cleanup system (CREACUS) unit is inoperable. The TS bases for LCO 3.7.11 requires that LCO 3.7.11 f
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-2-be entered when the chiller that supplies chilled water to the CREACUS unit is inoperable. Emergency Chiller E336 Supplies chilled water to CREACUS Train B. The operators had not entered TS LCO 3.7.11. The inspectors brought this to the attention of the control room operators, and TS LCO 3.7.11 was entered for Unit 2 around 9:00 a.m. on July 23,1998. A 9:45 a.m. common log entry on July 23,1998, indicated that Emergency Chiller E336 was returned to service and operable. Unit 3 operators had insufficient time to update the paperwork to indicate entry into TS LCO 3.7.11. The
_ inspectors found that the failure to recognize all applicable TS LCOs, when Emergency
_ Chiller E336 was declared inoperable, was contrary to licensee Operations management's expectation for licensed operators.
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The Unit 2 operators indicated entry into LCO 3.7.11 by writing this LCO in on the LCO l action required sheet that documented LCO 3.7.10 entry. No record indicated that LCO 3.7.11 had been entered late, and the pen and ink change to the LCO action
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required sheet was made after the sheet had been reviewed by two different senior reactor operators and control operators. The inspectors found that log keeping, in this instance, was weak because an accurate reflection of the time of LCO entry was not recorde Conclusions-l Licensed operators failed to recognize all applicable TS LCOs when an emergency chiller failed, indicating incomplete review of applicable TS. Also, control room logs did L
not reflect late entry into an applicable TS LCO after the inspectors identified the erro The operators did enter the most direct TS LCO, such that TS LCO actions times would not have been exceede .3 SFP Coolino Outaae - Unit 3 Insoection Scone (71707 and 86700)
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On August 17,1998, the inspectors observed the "lT-1" prejob briefing for an infrequently performed evolution, the removal from service of both trains of SFP cooling to allow corrective maintenance to be conducte i l Observations and Findinos .
The evolution was controlled by Procedure SO123-IT-1, " Infrequently Performed Tests .
and Evolutions Controls Program," Revision 4. As directed by the procedure, the I licensee implemented special controls, including having a designated management l representative, an evolution director with no other concurrent responsibilities, and
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predetermined contingency plans and back-out criteria. The details of these provisions-. l were documented in a written pla One of the considerations in the licensee's procedures for having both trains of SFP ,
cooling out of service was to not allow any activities that would dilute the SFP boron concentration. The plan for restoration of the system included refilling the SFP cooling L
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. of SFP cooling from service were excellent. A control operator demonstrated attentiveness in identifying the potential for slightly diluting the SFP, which would have been contrary to the licensee's procedures. Monitoring of the SFP was excellen O2 Operational Status of Facilities and Equipment O2.1 Firemain and Sumo Drain System Walkdowns - Units 2 and 3 Insoection Scone (71707)
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The inspectors walked down portions of the Units 2/3 firemain system and the Units 2/3 drain and sump system. The inspectors reviewed valve alignments completed for these ,
systems, and reviewed piping and instrument Drawings (P&lDs) 40117. 40118,40119, {
. and 40120 (Sump and Drain System); and 40198 (Firemain System). The inspectors i also reviewed applicable portions of the Units 2/3 UFSAR and Calculation M-120.15, Revision 6, " Plant Flooding Analysis Review." The inspectors also discussed system performance with applicable cognizant engineers and reviewed documentation of portions of maintenance that had been performed on these system ; Observations and Findinos
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l The inspectors observed that every floor drain checked (a total of 12) in the Unit 2/3 radwaste building and the Unit 3 Train A ECCS pump room contained debris such that the drains would probably not have passed design flow as described in the UFSAR. All floor drains checked (a total of eight) in the Unit 3 penetration building and the Unit 3 Train B ECCS pump room were free of debris. Many radwaste building floor drains also ,l
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- contained " duckbills" that had been installed in the late 1980s. The " duckbills" were essentially rubber funnels, attached to the drain opening, and designed to prevent gas i from passing from the respective sump to the space, but allowing liquid to pass from the ;
space to the respective sump. The inspectors observed that most of the " duckbills" had '
become loose from the drain opening, fallen into the drain pipe, and were restricting flow and causing debris to accumulate. The " duckbills" were not maintained by any
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-4-preventive maintenance program. In response, licensee Operations and Station Technical personnel generated ARs 980800433 and 980800435 to resolve the problems described above. Although the UFSAR describes a design flowrate for radioactive floor drains, the inspectors determined that drains were not credited in the design basis flooding analysis for Units 2/3 and, consequently, the drains being partially blocked had no safety consequence. The partially blocked radioactive floor drains could have, however, complicated cleanup from any radioactive spills in the areas the floor drains serviced. Based on the above, the inspectors found that licensee maintenance of some of the radioactive floor drains had been deficient, but that this represented a potential radiological concern and not a safety concer The inspectors did not identify any discrepancies between the as-built drain and sump system, the system P&lD, and the valve alignmen .
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The inspectors observed minor material deficiencies including a rusted limit switch, externally rusted isolation valve bodies and connectors, and an externally rusted jockey pump body and connectors. These deficiencies were evaluated by the licensee for repair. Major portions of the firemain system are exposed to the environment, causing the exterior rusting. The licensee had recently preserved portions of the system, with those portions not recently preserved exhibiting the material deficiencies. The inspectors were unable to determine the actualinterior condition of the piping. Licensee technical personnel stated that licensee initiatives were in place to determine if any problems existed with pipe wall thickness and material condition. The inspectors observed that the licensee had been proactive in preserving significant portions of the firemain system, but that occasional failures did occur, normally associated with valve operability. The inspectors found that the licensee was generally proactive in preserving the firemain system, and that exterior material condition of the system was reflective of the preservation effor The inspectors did not identify any discrepancies between the as-built firemain system, the system P&lD, and the valve alignmen Conclusions Licensee maintenance of radioactive floor drains in the common radwaste building and some of the Unit 3 ECCS pump rooms was deficient. Rubber devices used to prevent gas from exiting the drains had deteriorated and debris had been allowed to accumulate in the drains. There was no safety consequence to the partially blocked drain j i
The licensee had generally been proactive in preserving the firemain syste .
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l l-5-08 Miscellaneous Operations issues
08.1 (Closed) Unresolved item 362/98007-01: circuit breaker incorrectly racked ou l Insoection Scooe (92901. 92903)
i The inspectors reviewed Procedure SO123-VI-0.9, " Author's Guide for the Preparation i of Orders, Procedures, and instructions," Revision 5 (the procedure in effect at the time j of the incorrectly racked out breaker); Procedure 8023-6-3.1, "lTE 480 Volt Air Circuit '
Breakers," Reyhion 5; Calculation C-258-7.04, " Control Area - Class 1 Equipment Supports;" and AR 980603290. In addition, the inspectors discussed the event with Engineering and Operations personne Observations and Findinas -
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On June 27,1998, the inspectors had observed that a 480 volt load center breaker was incorrectly racked out in the disconnect position; the racking shutter was not in the lowered position. The licensee initiated AR 980603290 to evaluate the event. The unresolved item was initiated to evaluate the licensee's seismic analysis and corrective actions of the circuit breaker incorrectly racked ou i Seismic Evaluation The inspectors reviewed Calculation C-258-7.04. The initial calculation included an evaluation that indicated that the 480 volt breakers were restrained seismically if the breakers were in the " connect,"" test," or " disconnect" position. The evaluation did not evaluate a breaker configuration racked out beyond the disconnect positio Engineering evaluated the as-found configuration and revised the calculation to state 1 that in the case where the 480 volt breaker was left in a configuration between the j
" disconnect" and * removed" positions inside the cubicle, the switchgear would still be l'
seismically qualified as long as the cubicle door was closed and properly fastened. The inspectors observed that the cubicle door was closed and properly fastened when they had observed the incorrectly racked out breaker. The inspectors concluded that the breaker was in a seismically restrained position.
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Corrective Actions The licensee initiated several corrective actions as a result of the incorrectly racked out 480 voit breaker and a previously inspector-identified incorrectly racked out 4160 volt breaker (NRC Inspection Report 50-361; 362/98003). The Operations superintendent directed that the shift managers ensure that for the next three breaker racking ,
evolutions on their shift, the operator performing the evolution would have the procedure I in hand and check it off step by step as the breaker racking evolution was performed. If i the procedure did not work, the operator was to stop and change the procedure. In addition, a supervisor was directed to witness and document the procedure compliance and usage aspects on a field observation form for at least one of the evolutions. The l licensee is also developing an on-the-job training package on the racking in and out of
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6-4160 volt and 480 volt breakers that all operators will perform. Operations management designated breaker racking as one of the " spotlight" areas for the third quarter to increase operator awareness of the issu The failure of operators to correctly rack out a 480 volt breaker in accordance with j
Procedure SO23-6-3.1 was a violation of TS 5.5.1.1.a. This failure constitutes a i violation of minor significance and is not subject to formal enforcement actio l Procedure Format The inspectors reviewed Procedure SO123-VI-0.9, " Author's Guide for the Preparation of Orders, Procedures, and Instructions," Revision 5 (the procedure in effect at the time of the incorrectly racked out breaker), and Procedure SO23-6-3.1, "ITE 480 Volt Air Circuit Breakers," Revision 5, to determine the adequacy of the procedure change that
. addressed ensuring that the breakers were in a seismically qualified positio .l The inspectors observed that the caution added to Procedure SO23-6-3.1 was not written in accordance with proceduralized guidance. Procedure SO123-VI-0.9, Step 6.6.12.3, states, in part, that cautions should not direct action in place of procedure steps, and Step 6.6.12.4 states, in part, that cautions should provide the reason for, or consequences of, not following the caution. The caution prior to Step 6.1.1 of Procedure SO23-6-3.1 was written such that it directed action to close the racking shutter window, and the caution did not provide a consequence for not following the cautio The inspector discussed the weakness of Procedure SO23-6-3.1 with the Manager, Operations Procedures Group. The manager assessed the inspectors' concerns and determined that procedure enhancements were required and planned to revise the procedure. The inspector also observed that AR 980603290 did not address the procedure weaknesse Conclusions The 480 volt circuit breaker procedure was not written in the format recommended by the governing procedure for writing procedures and the AR written to address an incorrectly racked out 480 volt breaker failed to identify the procedure weaknesses. The breaker was seismically restrained while in the incorrectly racked out positio Corrective actions to address operator performance when racking in and out breakers were thorough and included supervisory oversight of breaker racking evolutions.
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-7-II. Maintenance M1 Conduct of Maintenance j M1.1 General Comments InsoeEtion Scone (62701J ;
The inspectors observed all or portions of the following work activities:
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Component Cooling Water P24 emergency air conditioning unit V-belt
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adjustment (Unit 3) - -
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Charging Pump 3P190 feeder breaker control circuit relay preventive maintenance (Unit 3) 1
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EDG 3G003 Radiator Fan E549 breaker control circuit and relay preventive maintenance (Unit 3) i
- l Charging Pump 3P109 bearing crosshead adjustment (Unit 3) i
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The inspectors found the work performed under these activities to be thorough. All work observed was performed with the work package present and in active use. Technicians were knowledgeable and professional. The inspectors frequently observed supervisors i and system engineers monitoring job progress, and quality control personnel were present whenever required by procedure. When applicable, appropriate radiation controls were in plac In addition, see the specific discussions of maintenance observed under Sections M and M4.1, belo M1.2 General Comments on Surveillance Activities Insoection Scoce (61726)
The inspectors observed all or portions of the following surveillance activities:
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LPSI Pump 2P016 inservice test (Unit 2)
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Engineered safety features subgroup Relay K-4018 semiannual test (Unit 2)
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EDG 2G002 monthly surveillance (Unit 2)
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EDG 2G003 monthly surveillance (Unit 2)
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EDG 3G002 monthly surveillance (Unit 3)
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-8- Observations and Findinas The inspectors found all surveillance performed under these activities to be thoroug All surveillance observed were performed with the work package present and in active use. Technicians were knowledgeable and professional. The inspectors frequently observed supervisors and system engineers monitoring job progress, and quality control
. personnel were present whenever required by procedure. When applicable, appropriate radiation controls were in plac In addition, see the specific discussions of a surveillance observed under Section M1.4, belo M1.3 Charaina Pumo 3MP190 Corrective Maintenance - Unit 3 - - Insoection Scone (62707)
On July 28,1998, the inspectors observed Maintenance machinists adjusting the crosshead bearing clearance for the three plungers of Unit 3 Charging Pump 3MP19 The work was performed using Procedure SO23-I-8.25, " Charging Pump Repack, Lubrication and Crosshead Adjustment," Revision 3. 'A Health Physics technician provided continuous job coverage.
" Observations and Findinas The crosshead bearing adjustment was performed correctly. The Health Physics technician performed numerous contamination surveys of the work area while the job was ongoing, in order to continuously verify that appropriate radiological precautions for the contamination levels were in place. The workers were familiar with the results of the surveys. The inspectors consequently found the actual work was performed technically ,
correct with good Health Physics suppor During the recording of actual torque values for the three crosshead bearing set screws, ;
the inspectors observed that the machinist supervisor recorded data for the west !
plunger in the procedural blanks for the east plunger. The craft then corrected the erro '
This maintenance activity included adjusting the crosshead bearing clearance, and then (
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installing new packing on each plunger. A note following Step 6.4.11 of ;
Procedure SO23-1-8.25 stated that if the pump was to be repacked, then N/A
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Section 6.4.12 and proceed to Step 6.4.13. Section 6.4.12 directed the crosshead
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bearing adjustment, and Step 6.4.13 directed lubricating plunger threads that were l exposed during the performance of Section 6.4.12. Consequently, the note appeared to
!. prohibit both repacking and adjusting the crosshead bearing clearance during implementation of the procedure (as the craft performed during this maintenance), and the note also directed a procedural transition that was awkward (lubricating plunger threads that were not exposed). The inspectors later discussed this with the machinist supervisor who had originated the note during the most recent procedural revision, and
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-9-who had supervised this maintenance. The machinist supervisor generated l AR 980807172 to correct these procedural problems, and stated that it was not the intent of the note to prohibit both repacking and adjusting crosshead bearing clearanc Based on the above, the inspectors found that attention to the procedure during the performance of the maintenance was weak, and that the procedure was awkward to implement as writte Conclusions Machinists' attention to procedural direction while adjusting charging pump crosshead i bearings and then repacking the charging pump was weak. The applicable procedure contained a note that appeared to prohibit the sequence of work as performed. The work was actually performed technically correctly, and the procedural note was in erro However, the procedural error was not noted or corrected until pointed out by the inspectors. Health Physics support of the maintenance was goo M1.4 LPSI Pumo inservice Test - Unit 2 Insoection Scoce (61726)
The inspectors observed operators perform an inservice test of LPSI Pump 2MP016 and reviewed Procedure SO23-3-3.60.2, " Low Pressure Safety injection Pump and Valve Testing," Revision Observations and Findinos On July 27,1998, the inspectors obscwed Operations test group performance during an inservice test of LPSI Pump 2MP016. The operators conducted a prejob briefing in the control room prior to the evolution that included an overview of the test, expected pump performance based on orevious tests, and personnel precautions. During the test, operators used good ALARA practices, displayed a good knowledge of the test equipment and good methodology upon conducting the test. The operators used good independent verification when realigning the system upon completion of the test. The inspectors discussed the operators' performance with a nuclear oversight representative observing the test, who indicated that the operators' performance was good. The inspectors agreed with the oversight assessmen During the performance of the test, the operators recorded the pump sealleakoff at zero drops per minute. Procedure SO23-3-3.60.2 indicated that if sealleakoff of less then 13 drops per minuts was satisfactory and with a leakoff greater then 500 cc per minute to the sump, then the pump was inoperable. The inspectors observed the operators record the leakage 12 minutes after the pump was started. The inspectors questioned Engineering about the adequacy of the 13 drops per minute criteria, l because the leakoff drain piping transverses approximately 20 feet horizonal and 7 feet
! vertical with seven elbow bends. The engineer indicated that it would be a better
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practice to record the leakoff prior to securing the pump and initiated AR 980801274 to f change the procedure. In addition, the engineer indicated that the procedure was
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I-10-acceptable as written, but could be improved. The inspectors agreed with the engineer that the initial procedure was acceptabl *
l The inspectors questioned the engineer about the 500 cc per minute criteria. The engineer indicated that the 500 cc per minute criteria was referenced in Section 15.6.3.3.5.1.B.3.A of the UFSAR. The inspector reviewed the UFSAR and identified that for gross seal failure leakage, the seal vendor determined by test that leakage was less than 500 cc per minute. In addition, the vendor determined that 87.5 cc per minute would go to atmosphere and 412.5 cc per minute would go to the building sump. The inspectors informed Engineering that the 500 cc per minute criteria for declaring the pump inoperable as stated in Procedure SO23-3-3.60.2 would not be achieved since the vendor determined the maximum amount to the sump was 412.5 cc per minute. The engineer included a review of the acceptance criteria for declaring the ;
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10 CFR Part 50, Appendix B, Criterion V, requires, in part, that procedures shall have appropriate quantitative acceptance criteria. The failure of the licensee to provide j appropriate criteria for declaring the LPSI pump inoperable for gross seal failure leakage '
was a violation of 10 CFR Part 50 Appendix B Criterion V. This failure constitutes a violation of minor significance and is not subject to formal enforcement actio Conclusions
OperaNns test group personnel performance during a LPSI pump inservice test was good. The personnel were knowledgeable about the test process and methodology to perform the test. In addition, good ALARA and industrial safety practices were utilize A test weakness was identified, in that the LPSI inservice test procedure criteria fr declaring the pump inoperable based on seal leakoff was nonconservative and did not match the analysis assumed in the UFSAR. This had minimal safety consequenc M4 Maintenance Staff Knowledge and Performance M4.1 SFP Coolina System Maintenance - Unit 3 Insoection Scoce (62707)
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The inspectors observed maintenance activities on several components of the SFP cooling system. The inspectors discussed the mechanics performance with Maintenance managemen Observations and Findinas On August 17,1998, the inspectors observed Maintenance personnel perform work on l SFP system components. The inspectors observed calibration of the SFP Pump P009 discharge pressure Indicator 3Pl7707, packing adjust on the SFP Heat Exchanger 3E005 outlet Valve 3HCV7748, and actuator replacement of the SFP Heat Exchanger 3E006 outlet Valve 3HCV7747.
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i-11-l The inspectors observad mechanics remove the actuator from Valve 3HCV7747 and l
replace it with a modified actuator. The inspectors observed that the original actuator l was positioned with the handwheel on the west side of the valve. When the mechanics i installed the' modified actuator, they were observed installing it with the handwheel on i the east. With the handwheel on the east, component interferences would create further I
difficulty for operators to manipulate the valve. The inspectors reviewed Maintenance 1 Order 94030543002 and determined that the mechanics were installing the valve .
incorrectly. Step 10 stated, in part, to install the actuator with the handwheel accessible - 1 from the wes The inspectors discussed the valve orientation with a general foreman who agreed that the valve was incorrectly oriented. The general foreman indicated that the technicians had discussed the orientation of the valve but incorrectly determined what direction was
. wes The inspectors reviewed the maintenance order and determined that the licensee had an opportunity to identify the incorrectly oriented valve since the cognizant engineer j walkdown to verify the field change implementation had not been completed. However, j the failure of the mechanics to follow Step 10 of the maintenance order was a violation i of TS 5.5.1.1.a. This failure constitutes a violation of minor significance and is not j subject to formal enforcement actio The inspectors observed Health Physics support of the maintenance on the SFP cooling system. The technicians were aggressive in monitoring the work as components were ;
opened, which exposed contaminated surfaces. The technicians established excellent :
contaminated control area boundaries that allowed workers ample access to '
components and provided supervisors with a viewing and communications area that was
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free of contaminatio . Conclusions l Maintenance personnel displayed a knowledge weakness when installing an actuator l 180 degrees off from the required position stated in the maintenance order. This was l
identified by an NRC inspector while work was in progress. Health Physics support of j the maintenance evolutions was excellent in that the technicians aggressively monitored l the activities and established appropriate contamination boundarie !
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'.-12-111. Engineering
- E2 Engineering Support of Facilities and Equipment E2.1 ' Ooerabilitv Assessments
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- Insoection Scooe (37551)
l The inspectors reviewed the OAs for the following ARs:
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980100040 980502770 980600386 980601063 l 980100176 980501772 980600388 980601233 L .980100258 980502779 980600398 980601336-980100377 -980600142 980600468 980601497 980100395 980600181 980600483 980601508 98010039 , 980502587 980600361 980600838 .980601627
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980502707 980600385- 980600889 980601722 980502763
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!- Six OAs generated in January 1998 were reviewed and found to be acceptable.
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I 35 OAs generated in the first part of June 1998 were reviewed. 28 of these were clearly acceptable.
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L Between January and June, the licensee implemented revised guidance for OAs,
! expanding the list of aspects that were to be considered in the OA discussion. The
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guidance was provided in the instructions imbedded in the electronic documentation L system (MOSAIC). The guidance provided improved consistency and completeness of the OAs, although some of the OAs did not follow the format or answer all the questions.
!. Seven of the OAs contained minor deficiencies that did not invalidate the operability
- determination, such as not specifying the requirement or safety function that applied to
[ . the degraded condition, so that the basis for operability could be clearly establishe [ For example, one OA did not follow the format of the guidance and was poorly focused, !
and consisted of six pages of text that ultimately provided an adequate basis for ;
operability. Another OA omitted background information (valve size) needed in order to '
complete the logic for determining that the safety function was met. Another OA did not correlate various units of measure (level, percent volume, and gallons) used in the OA !
- to provide the logic necessary for understanding the operability basis. The inspectors discussed these issues with the cognizant Station Technical personnel, who i
- acknowledged the opportunities for improved clarity and completenes I I
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-13- Conclusions All 41 recently completed OAs reviewed adequately demonstrated the basis for operability. Recently implemented expanded guidance for completeness of the OAs resulted in general improvements in the quality of the OAs. However, some OAs contained r inor deficiencies that required further explanation to complete the logical basis for operabilit E Implementation of Steam Calorimetric Power Calculation in COLSS - Unit 3 Insoection Scoce (37551)
The inspectors reviewed Facility Change Evaluation (FCE) 3-98-001, " Implementation of Steam Calorimetric Power Calculation in COLSS," dated June 23,1998; Software Modification Record PMS-98-004; and monitored Unit 3 performance before and after implementation of the change, Observations and Findinas in July and August 1998 the licensee performed testing and finalimplementation of a change to the secondary calorimetric power calculations. The change, described in FCE 3-98-001, provided for the use of a steam flow based calorimetric to be used at high reactor power levels (greater than 95 percent). The steam calorimetric was considered more accurate than the feed water calorimetric because, unlike the feed water flow rate sensor (a venturi-type detector), the steam flow sensor does not become significantly fouled over the course of an operating cycle. Elimin'ation of the additional margin required to compensate for error introduced by the feed water flow venturi fouling was expected to allow the reactor to be operated at a slightly higher power level, resulting in a smallincrease in electrical output of the main generato Operators would be able to manually select use of the steam calorimetric when reactor power was greater than 95 percent, and the feed water calorimetric would automatically be selected when power was reduced to less than 95 percen FCE 3-98-001 described the tests to be performed for validating the steam calorimetri In essence, the tests involved off-line comparisons, using test cases, to show that the calculations performed by COLSS were acceptable, and on-line testing, using the backup COLSS system, to allow validation using real-time dat The FCE included an evaluation in accordance with 10 CFR 50.59. This evaluation concluded that the change did not represent an unreviewed safety question, based on COLSS being nonsafety-related and having no control or plant protection functions. The FCE also included the proposed revisions to the UFSAR and design basis l document implementation of the change in Unit 3, on August 7,1998, resulted in an increase in the electrical output of approximately 10 megawatts.
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-14-Software Modification Record PMS-98-004 included well-documented test results for the change, demonstrating that the revised calculations performed by the software were within the predicted tolerance The inspectors discussed with Reactor Engineering the validation and performance problems that occurred at another facility (Arkansas Nuclear One) that had implemented a similar change. The reactor engineer was fully aware of the other facility's problem, and stated that Southern California Edison had contracted Asea Brown-Boveri (ABB) to evaluate the susceptibility of the proposed software change to the problem. Because the software implemented at Unit 3 allowed use of the steam calorimetric only at reactor power levels greater than 95 percent, the coefficients used in the calculation were able
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to be adjusted to avoid the sensitivity problems experienced at the other facility. The -
licensee performed special test cases, some developed by the licensee, to ensure that the calculations were valid over the entire rang . . Conclusions Engineering support for the COLSS software change to implement a reactor power calculation based on a steam calorimetric was excellent, including contracting vendor support to address a potential problem, and developing test cases to ensure the validity of the software. The facility change evaluation and the software modification report fully addressed all relevant aspects of the change. Finalimplementation was successfu E7 Quality Assurance in Engineering Activities E7.1 Stoo Work Order for Fuel Fabrication - Unit 2 insoection Scoce (37551)
The inspectors reviewed the circumstances surrounding the licensee's July 7,1998, stop work order to ABB for its fuel manufacturing proces Observations and Findinos During visualinspections of seam and lap welds on Zircaloy grids, ABB inspectors identified differences in the appearance of some welds. The differences were later determined to have been caused by stainless steel inclusion in the Zircaloy, apparently from improperly focused lasers drawing material from the stainless steel weld fixture (jig)
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into the welds. ABB personnel examined the grids and classified the weld '
contamination as light, moderate, or heavy, based on subjective criteria. Subsequently, objective criteria for the classifications were develope The grids provide structural support for the fuel pins in the fuel assemblies. Ten grids ,
are installed in each fuel assembly. Each grid included dozens of welds, and only a few welds on some grids were contaminated.
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-15-ABB informed the licensee of the contamination problem. The licensee sent personnel !
to ABB's Windsor, Connecticut, facility to independently inspect the ? rids and assess the condition. The licensee personnelidentified that a large percentage of the grid >
classifications were less conservative than classifications based on the objective criteria that had been developed after the original classification. The licensee determined that j ABB had not confirmed that the pre /iously completed classifications were consistent l with the objective criteria. Furthermore, a number of grids had been shipped to ABB's Hematite, Missouri, facility to be installed i1 fuel assemblies. The licensee personnel
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went to Hematite to inspect them and found that many of those assemblies had also {
been misclassified. Some of the grids could not be inspected because they had already been installed in fuel assemblie I i
On July 7,1998, based on these discrepancies, the licensee issued a stop work order to ABB pending resolution of the quality issues. The conditions for rescinding the stop !
work order were for ABB to propose, and the licensee to evaluate and accept, acceptance criteria for the contamination in the seam and lap welds and the repair plan for unacceptable welds, and disposition of the 20 loaded fuel assemblie ABB subsequently performed structural testing of some grids with the contaminated welds removed, demonstrating that the welds were not structurally necessary. ABB also examined the welds from a microscopic perspective. Acceptance criteria for application j to future work were developed, and modified acceptance criteria for evaluation of the almdy manufactured grids was being develope ABB could not definitively determine when the welding problem began, but considered it possible that the problem was present since ABB implemented the laser welding proce,s in 1992. The earliest that San Onofre could have been affected was 199 However, based on its inspection program and the recent testing results, ABB did not consider that any significant nonconformances in fuel assemblies in service, at San Onofre or elsewhere, were likel The licensee Nuclear Fuels manager stated that the stop work order would remain in effect until ABB completes the development of inspection criteria, to be documented in a
" Deviation from Contract Requirements" report. However, ABB has resumed its i manufacturing process "at risk." l c. Conclusions I
Nuclear Oversight demonstrated aggressive, independent oversight of the fuel
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fabrication process, and took appropriate action in issuing a stop work order to the fuel vendor pending resolution of a quality problem that resulted in stainless steel contamination of welds in the Zircaloy grid I l
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-16-E8 Miscellaneous Engineering issues (92700)
E (Closed) Licensee Event Reoort (LER) 361/98-005-00: inadequate voltage to 120 VA0 circuit On November 17,1997, the licensee initiated AR 971101015 after questioning the adequacy of the assumptions used in the 120 VAC electrical distribution system design calculation. The licensee determined that the calculation evaluated 120 VAC circuits by allowing a 10 percent voltage drop in the cables and assumed nominal voltages of 480 volts at each of the upstream motor control centers instead of the worst case minimum voltage. An initial operability assessment concluded that the loads off the 120 VAC circuits were operable as long as both units remained on-lin On November 18,1997, the licensee initiated further evaluation that conc!uded that the 120 VAC system was operable during the winter months of October through May when the worst case minimum switchyard voltage was 222 kV (nominal switchyard voltage -
was 230 kV). Further evaluation with the switchyard voltage at 218 kV (which represents the minimum voltage at the site with one unit off-line and with a heavy J system load during the months of June through September) determined that several 120 VAC circuits did not meet their voltage acceptance criteria with the buses aligned to their alternate regulated AC sourc On January 22,1998, the licensee initiated design changes to improve the 120 VAC circuits. The primary tap setting for the common panel distribution transformer was changed, power feed cables to some components were rerouted to decrease the circuit length and voltage loss, and vital bus voltage regulators were adjusted. All the changes were completed prior to the summer month On March 5,1998, the licensee determined that the units had previously operated in a condition in which the 120 VAC circuits may not have performed their intended design function under worst case electrical loading and voltage condition During the review of the 120 VAC problem, the licensee determined that the engineer oerforming Calculation E4C-120 in March 1996 used incorrect assumptions and methodology and that an independent review engineer did not identify the error Therefore, the licensee included corrective actions to review calculations of a similar I type and to verify that the appropriate assumptions had been used. In addition, the !
licensee provided training to engineering personnel on technical review, cross checking, and the need to review all possible scenarios for electrical alignments. The inspectors concluded that the licensee's design changes and additional corrective actions were acceptabl In the LER, the licensee stated that the cause of the event was personnel error. During l' a design basis reconstitution of the electrical systems in March 1996,120 VAC control circuit voltage calculations were prepared and nonconservative assumptions were l introduced. The inspectors concluded that the licensee missed an opportunity to identify ,
the inadequate voltages of the 120 VAC circuits in March 199 '
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-17-The licensee identified three instances in which equipment did not comply with the design information in the UFSAR: 1) inadequate terminal voltage was found for specific loads supplied from common 120 VAC Panel QO33 that was fed from Motor Control Center BQ through Transformer T033, 2) inadequate terminal voltage was found for the excore startup and wide range monitor signal processor snd indicator on the remote shutdown panels,' and 3) inadequate voltage was found for Foxboro panel (nuclear steam supply system loop monitoring) when its 120 VAC vital bus was powered from the alternate regulated source. =10 CFR Part 50, Appendix B, Criterion 111, requires, in part, that. measures shall be established to assure that the design basis is correctly translated
'into specification, drawings, procedures, and instructions. The failure of the licensee to ensure the design basis was correctly translated into specifications was a violation of 10 CFR Part 50, Appendix B, Criterion 111. This nonrepetitive, licensee-identified and
- corrected violation is being treated as a noncited violation, consistent with Section Vll.B.1 of the NRC Enforcement Policv (NCV 361; 362/98013-01).
., IV. Plant Sunoo P1 Conduct of Emergency Preparedness Activities P Emeroency Recall Drill- Units 2 and 3 insoection Scone (71750)
The inspectors observed the August 18,1998, emergency preparedness drill that included an after-hours emergency recall of emergency response personne Observations _ad Findinas -
The event was declared at 4:23 a.m. on August 18,1998. The licensee's goal for the emergency recall was to have'the TSC, operations support center, and emergency operations facility activated within 75 minutes of event declaration. This goal allowed for 15 minutes to accomplish the notifications, and 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> following notification for personnel to assume their positions. With the exception of one position in the TSC, the licensee's goal was achieved. This last position, the Emergency Advisor for Notifications, was required for TSC activation and was filled approximately 4 minutes late. The licensee's Emergency Plan stated that these personnel can "be onsite normally within one hour following notification." The inspector determined that the general intent of the Emergency Plan requirement was me Communications within the TSC were generally good. The Site Emergency Director conducted periodic briefings to keep TSC personnelinformed of the overall situatio The main status board in the TSC was kept current, although the inspectors identified one salient omission from the posted status. The initial event was an anticipated transient without scram, but the status board did not ever reflect the fact that the reactor was successfully tripped.
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-18-Event simulation within the TSC was adequate. The station blackout resulted in a loss of all the not.nal lighting in the TSC, forcing personnel to use the available flashlights and portable lantems. The containment high range radiation monitor data, which could be viewed and trended in the TSC on the drill computer systems, did not model the effects of temperature, which presented some initial confusion for the TSC staf Response to the event was appropriate. The TSC staff appropriately assessed -
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conditions, including predicting the time to core uncovery and assessing fuel damag Although the drill scenario precluded implementation of successful mitigative actions, j such as cross-connecting electrical' distribution systems between units, potential i mitigative actions were considere '
c. . Conclusions The licensee's conduct of the afte hours emergency preparedness drill was excellen One exception to staffing the TSC within the timeliness goals of the drill was identified by the licensee. Overall communications in the TSC were goo V. Manaaement Meetinas l X1 Exit Meeting Summary The inspectors presented the inspection results to members of licensee management at the exit meeting on August 26,1998. The licensee acknowledged the findings presente The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.
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ATTACHMENT SUPPLEMENTAL INFORMATION PARTIAL LIST OF PERSONS CONTACTED Licensee -
' J. Fee, Manager, Maintenance G. Gibson, Manager, Compliance D. Herbst, Manager, Site Quality Assurance J. Hirsch, Manager, Chemistry R. Krieger, Vice President, Nuclear Generation J. Madigan, Manager, Health Physics
' O. Niebruegge, Manager, Station Technical (Acting) - --
D. Nunn, Vice President, Engineering and Technical Services -
T. Vogt, Plant Superintendent, Units 2 and 3 R. Waldo, Manager, Operations INSPECTION PROCEDURES USED IP 37551: Onsite Engineering IP 61726: Surveillance Observations IP 62707: Maintenance Observations IP 71707: Plant Operations IP 71750: Plant Support Activities IP 86700: Spent Fuel Pool IP 92700: On Site LER Review IP 92901: Followup - Operations IP 92902: Followup - Maintenance IP 92903: Followup - Engineering ITEMS OPENED AND CLOSED Ooened and Closed 50-361;362/98013-01 NCV inadequate voltage to 120 VAC circuits Closed 50-361; 362/98005-00 LER inadequate voltage to 120 VAC circuits 50-362/98007-01 URI circuit breaker incorrectly racked out I
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, LIST OF ACRONYMS USED
ABB Asea Brown-Boveri
. AR action request
, COLSS core operating limits supervisory system CREACUS control room essential air cleanup system ECCS- emergency core cooling system
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EDG emergency diesel generator j FCE' facility change evaluation LCO limiting condition for operations LER licensee event report -
LPSI' low pressure safety injection O operability assessment P&lD - piping and instrument drawing -
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PDR Public Document Room - -
SFP spent fuel pool
- TS - technical specifications TSC technical support center UFSAR updated final safety analysis report -
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