IR 05000321/1998002
| ML20249A315 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 06/01/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20249A314 | List: |
| References | |
| 50-321-98-02, 50-321-98-2, 50-366-98-02, 50-366-98-2, NUDOCS 9806160299 | |
| Download: ML20249A315 (39) | |
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U.S. NUCLEAR REGULATORY COMMISSION
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REGION II
l Docket Nos: 50-321. 50-366 License Nos: DPR-57 and NPF-5 l Report No: 50-321/98-02, 50-366/98-02 l
L Licensee: Southern Nuclear Operating Company. Inc. (SNC)
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Facility: E. I. Hatch Units 1 & 2 l l
Location: P. O. Box 2010-Baxley. Georgia 31515 i-Dates: March 22, 1998 - May 2. 1998 Inspectors: B. Holbrook. Senior Resident Inspector J. Canady. Resident Inspector H D. Forbes. Health Physics Inspector. Sections R1.2. R7.1 and R Accompanying T. Fredette. Resident Inspector Personnel: L. Olshan. Project Manager - Hatch Approved by: P. Skinner. Chief..~ Projects Branch 2 Division of Reactor Projects
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l EXECUTIVE SUMMARY Plant Hatch. Units 1 and 2 L NRC. Inspection Report 50-321/98-02. 50-366/98-02
This' integrated inspection included aspects' of licensee operation engineering. maintenance. and plant support. The report covers a 6-week period of resident ~ inspection:. in. addition. it includes the results of an announced inspection by a regional health physics specialist.
L '00erations-The planning and coordinating activities performed by engineering in
._ support of. operations- for the movement of a potentially damaged fuel l . assembly were detailed and thoroug Heath Physics personnel provided L
the required radiological monitoring-(Section 01.2).
During the Unit 2 Reactor Core Isolation Cooling testing, operators.
, reviewed and used applicable procedures; were proactive in the use of.
! )eer checks" prior to opening electrical links that were identified in l tle procedure; and operators performing the test and other crew members l effectively used three-'part communications (Se: tion 04.1)
!- Operations personnel demonstrated a lack -of sensitivity for cnanging unclear steps in the Reactor Core Isolation Cooling surveillance procedur The procedure required operators to make independent-verifications that could not be made and required operators to take
' actions that were not.specifically identified in the procedure. The procedure.was later revised to correct-the' deficiencies (Section 04.1).
The Significant Occurrence Report associated'with a turbine load set L problem on both units was not thorough and detailed. The actions
- . completed _were not entirely effective to ensure that the load set was
! .properlyLadjusted by operators. The inspectors identified areas for.
D corrective action improvement (Section 08.1).
Maintenance L Poor. work planning resulted in a small breach of the Unit 1 secondary
~ containment associated with the ,titrogen relief valve-(Section M1.2).
Overall maintenance work activities associated with a Unit 1 Emergency
. Core Cooling System' area cooler were well organized and controlled
.(Section M1.3)..
-Licensee Event Report.(LER) 50-321/98-002 developed to report a blown
- . fuse and Primary Containment Isolation Signal on April was thorough with
' clear and detailed investigation results (Section M1.4).
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Enclosure
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Maintenance personnel correctly determined that the root cause of a blown fuse in the turbine combined intermediate valve trip test circuit on Unit 2 was due to a faulty solenoid. The reactor power reduction initiated by the operating staff demonstrated conservative decision making (Section M3.2).
- Maintenance personnel troubleshooting activities were effective in determining that the flow rate meter was the problem component associated with the unsatisfactory performance of the reactor water level keepfill system surveillance on Unit The inspectors did not view this intermittent problem with a sticking flow meter as a safety concern (Section M3.3).
A Non-Cited Violation (NCV) 50-321/98-02-02: Failure to Meet Technical Specification Requirements for a Unit 1 Reactor Building Ventilation Radiation Monitor Setpoint, was identified. The as-left setpoint exceeded the Technical Specification allowable value (Section M3.4).
A poor supervisory review of a completed surveillance procedure for the Unit 1 Reactor Building Ventilation System failed to identify that the Technical Specification acceptance criteria was not met and contributed to a violation of regulatory requirements. This was identified as an area for improvement (Section M3.4).
Maintenance personnel displayed a questioning and attention-to-detail attitude during the Unit 2 High Pressure Coolant Injection online maintenance outage. This resulted in the timely discovery of the damaged HPCI thrust bearing (Section M3.5).
The Unit 2 High Pressure Coolant Injection maintenance and testing activities were well-planned and coordinated. Maintenance and engineering support to operations were evident for the planning of the work activities and the subsequent return of the system to service (Section M3.5).
Operations supervision was actively 'nvolved in the Unit 2 High Pressure Coolant Injection maintenance and testing activities and provided continuous oversight and direction. Operators correctly used procedure displayed an attentiveness to datail, and effectively monitored system critical parameters. Technical Specification and surveillance testing acceptance criteria were met (Section M3.5).
Enaineerina The licensee had taken appropriate actions to review and evaluate the potential 10 CFR 21 issue associated with General Electric (GE) Type SBM Control Switches (Section E2.1).
Enclosure
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The failure to compare and document the closure time of the Unit 2 Reactor Core Isolation Cooling (RCIC) valve. 2E51-F003, to the inservice testing reference time was identified as NCV 50-366/98-02-03. Failure to Meet RCIC IST Valve Testing Requirements (Section E2.2).
An engineering inservice testing (IST) review of completed surveillance did not detect that IST requirements for RCIC valve 2E51-F003 were not met and was identified as an area for improvement (Section E2.2).
Operations and Maintenance personnel's prompt a'ctions in clam)ing a cut i
.made.on-the demineralized water piping during maintenance war (
effectively minimized the loss of glycol inventory from Emergency Diesel Generator expansion tanks due to siphoning action (Section E2.3).
The clearance writer and design engineer for work on the turbine building chiller on Unit 1 failed to consider the effects of siphoning on the emergency diesel generator expansion tanks. This siphoning. problem did
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not affect the EDG operability (Section E2.3).
The 10 CFR 50.59 Safety Evaluation Program was effectively implemente The detail and quality of the 50.59 evaluations improved over the last 18 months (Section E3.1).
Some 10 CFR 50. 59 evaluation questions were not answered in sufficient detail to ensure an independent reviewer could draw the same. conclusion (Section E3.1). I l
l Site and'cor) orate personnel had received the specific training to conduct 10 C R 50.59 evaluation The safety evaluation procedure-
! included appropriate and detailed guidance for the preparation and review '
of 10 CFR 50.59 evaluations (Section E3.1).
The Plant Review Board met the requirements specified in Section 1 of the Hatch Quality Assurance Manual anc Section 17 of the Updated Final Safety Analysis Report (FSAR). The~ Plant Review Board members conducted a detailed and thorough _ review of the'10 CFR 50.59 evaluations completed
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for procedure revisions (Section E3.1).
Audits in the area of the 10 CFR 50.59 Evaluation Program were thorough-and detailed. The audits were conducted by corporate and site personnel (- knowledgeable of the safety evaluation program. The response and corrective ' actions. to audit findings were timely and appropriate to l ensure program improvement (Section E3.1).
Non-Cited Violation (NCV) 50-321. 366/98-02-04. Plant Operation Outside of the Design _ Basis for an Engineered Safeguard System was identified following the NRC's review of Apparent Violation (EEI) 50-321. 366/98-01-08 (Section'E8.3).
Enclosure i
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Plant Sucoort Housekee)ing has generally-improved since the last report period. Plant Health Plysics has placed increased emphasis upon area decontamination and the removal of debris from contaminated areas Section R1.1).
The licensee was effectively maintaining controls.for personnel monitoring, control of radioactive material, radiological postings, radiation area controls, and high radiation area controls as required by 10 CFR Part 4 Efforts to reduce 3ersonnel contaminations during 1998 to date had oeen positive (Section 11.2)..
Chemistry's persistence in determining the root cause for the loss:of the primary Reactor Coolant System flow to the coritinuous in-line conductivity cell resulted in a detailed and timely solution to the problem- (Section R2.1).
-The inspectors determined that the licensee's most recent formal Quality-Assurance Audit identified items of substance and that auditors used checklists to effectively assess the' radiation protection program, as required by 10 CFR Part 20.1101 (Section R7.1).
Information communicated.to the control room during the medical emergency drill conducted on April 21 did not identify that the simulated injured individual was contaminated. Subsequently, information relayed to the hospital-was incorrect (Section P4 1).
The command and control responsibilities of the first responders for emergencies du_ ring the medical emergency drill were not clearly j delineated or defined. This contributed to a lack of command and control observed during the medical emergency drill conducted on April (Section P4.1).
The' fire protection personnel provided an effective drill for the fire brigade. Response and coordination objectives were me Minor
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deficiencies associated with a failure to initially establish a hydrant operator and sparse followup messages to the fire brigade leader were discussed with the fire drill coordinator. These deficiencies were discussed in a post-drill critique (Section F5.1).
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Report Details Summary of Plant Status Unit 1 operated at 100% rated thermal power (RTP) for the entire 1 reporting period, except during routine testing activitie ' Unit 2 began the report period at 100% RTP. Reactor Power was reduced to about 60% RTP on March 29 for condenser bay maintenance activities and replacement' of selected servo-strainers in the turbine Electro-Hydraulic
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Control System. Power was returned to 100% RTP.'on March 30. Reactor power was reduced again on April 5 to about 60% RTP to aerform a control rod. sequence exchange and repair. condenser bay leaks. )ower was returne to 100 RTP the following day. On April 19. reactor power was reduced to about 40%-RTP due to an unsatisfactory combined intermediate valve surveillance testing result. as required by the Technical Requirement 1 Manual. The unit was returned =to 100% RTP the same day, f(;10 wing troubleshooting and repair activities by maintenance. The unit remained at 100%' RTP for the remainder of the reporting period, except during routine testing activitie I. Operations 01 Conduct of Operation .1 General Comments (71707)
Using Inspection Procedure 71707 'the inspectors conducted frequent reviews of ongoing plant operations. The inspectors' also
.. verified on April 14 that the licensee had the current revision-(Rev.) of NRC Form 3. Notice to Employees. 3rominently posted in conspicuous : locations., as required by,10 CF119.11~. The
. inspectors observed that the forms were not defaced or altere .In general, the conduct of operations was professional and safety-conscious: specific events and observation are detailed in the sections below.-
01.2 Movement of Potentially Damaaed Soent Fuel Assembly in the Unit 1 Soent Fuel Pool-(37551) (71707) (71750).
On April 16. the inspectors observed the movement of a potentially
, damaged spent fuel assembly in the Unit 1 spent' fuel pool. The L . s)ent fuel _ assembly was potentially damaged on October 6.1984.
, w1en it was inadvertently released and dropped from a height of about twelve feet into its spent fuel pool rack cell. The s)ent fuel assembly had not been moved since the event. The assem)1y was moved to safely segregate.it with other damaged and/or defective' fuel assemblies and to make other cells available for spent fuel storag Enclosure
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The inspectors observed that the procedurally required number of l
operations personnel were on the refueling bridge for the assembly movement and health physics personnel were present for radiological monitoring. The inspectors also observed the presence of reactor engineering personnel who developed and coordinated the planning for the movement of the assembly. The fuel assembly was successfully moved by operations personne The planning and coordinating activities performed by engineering in su) port of the movement of the Potentially damaged fuel assem]ly were detailed and thoroug1. Heath Physics personnel provided the required radiological monitoring.
04.0 Operator Knowledge and Performance 04.1 Observation of Ooerator Performance Durina Unit 2 Reactor Core Isolation Coolina (RCIC) Test 1_ n Insoection Scooe (71707)
The inspectors observed control room operator performance during the performance of-procedure 34SV-E51-001-25. "RCIC Valve Operability." Rev. 11. on April 3. 199 Observations and Findina The ins)ectors observed that operators reviewed and used applica)le procedures. Operators used peer checks prior to opening electrical links that were identified in the procedur Operators performing the test and other crew members used three-part communication .When cuestioned, the operatcr performing the test stated that one
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procecural step that required independent verification was not i clear. In general, the procedural step required operators to stop the activity and independently verify that the RCIC room cooler had automatically started. However, the ste) required one l operator to depress a reset push-button whic1 stopped the cooling fan that had automatically st.arted. However, the procedure
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required another operator to independently verify that the cooling fan had automatically started. Following the reset action. there would be no indications for the second operator to independently
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. veri fy. . Additionally, the procedure did not= direct the operators to reset the automatic start signal to the cooling fan that is i normally operating. A reset of the automatic start signal for
' both fans is. required to reset the common control room alarm. The inspectors observed that.the Unit 1 procedure contained the same wording as the Unit 2 procedure. The inspectors discussed these observations with operations management. The inspectors were l- later informed that the orocedural steps would be revised. The
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inspectors observed that the unclear procedural steps did not l invalidate the surveillance tes Inspector observations of the Inservice Testing Requirements for
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the P.CIC valves are discussed in section E2.2 of this Inspection Repor Conclusions The inspectors concluded that, during the Unit 2 Reactor Core Isolation Cooling testing observed, operators reviewed and used applicable procedures. Operators were proactive in the use of
" peer checks" prior to opening electrical links that were identified in the procedure. Operators performing the test and other crew members used three-part communications. The current procedure revision was not clear for some actions needed to return the operating fan to normal alignment. The procedure was later revised to clarify the step Miscellaneous Operations Issues (92901\(92700)
0 (Closed) IFI 50-321. 366/97-02-04- Review of Goerator Performance Deficiencies and Licensee Corrective Action The inspectors reviewed Significant Occurrence Report (SOR) 97-83, dated July 3. 1997. As corrective actions, the licensee replaced the Unit I load set indicator and calibrated the load set indicator on both units. Operator responsibilities, the need for pre-job briefs. Seer checks, and having a questioning attitude i were also re-emplasize The inspectors were informed that prior to this problem operators were aware that the load set indicators were ,ut of calibration and generally indicated between 20 to 50 megewatts electrical (MWE) P.igher than actual turbine load. As a result, operators increased the load set high u remained on pressure control.pscale During thetotime ensure that the corrective actions turbine were being implemented, operations management recognized that the Unit 2 load set had not been corrected as stated in the SO Operations management responded to the SOR and documented in a memorandum that "the Unit 2 load set indication is presently off by as much as 20 to 50 MWE and thus represents an unacceptable challenge to the operating crews in performing main turbine activities."
The inspectors observed the load set indicator on both units on A)ril 21. 1998, and noted that both indicators were greater than t1e 905 megawatt setting per procedure 3450-N30-001-1S/2S " Main Turbine Operation" Enclosure I
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The inspectors discussed their observations of the load set indicators with operations personnel. The inspectors were later informed that the deficiency for Unit 1 had been worked and the setpoint was verified to be correct. To ensure that the load set was properly adjusted, operations management initiated actions to place a placard on the control panel indicating the correct setpoin The inspectors identified minor deficiencies associated with licensee actions. First, the closed-out SOR indicated that the load set indications on both units had been corrected. The Unit 2 load set had not been corrected. Second, operations management recognized that the SOR was not correct and communicated this observation by memo to Nuclear Safety and Compliance. Neither department took action to correct the 50 Conclusions The inspectors concluded that the Significant Occurrence Report process and corrective actions associated with a turbine load set problem on both units were not entirely effectiv II. Maintenance M1 Conduct of Maintenance M1.1 General Comments Insoection Scoce (62703) (62707) i l
The inspectors observed or reviewed all or portions of the I following work activities: 4
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. MWO 2-98-0899: Troubleshoot reason for blown fuse associated with 2N32-F004D e MWO 2-98-0148: Limitorque Valve Operator Inspection for MOV 2E41-F004
. MWO 2-97-3564: Change Circuit Breaker Settings on MOV 2E41-F004
. MWO 2-98-367: Repack Valve 2E11-F031B RHR Discharge Check Valve Observations and Findinas Work activities performed on valve 2E41-F004 Unit 2 HPCI Pump Condensate Storage Tank Suction Isolation, were scheduled as part of a HPCI system maintenance outage. The inspectors observed maintenance personnel conducting grease and electrical component inspections (including torque switch inspection) of the valve motor actuator Maintenance personnel were knowledgeable and Enclosure i
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attentive to appropriate acceptance criteria as part of their inspection The inspectors verified that recommended changes to the actuator motor circuit breaker trip settings for full load amperage ~ (FLA)
and locked rotor amperage (LRA) met design guidelines. The design guidelines provided recommended settings at the nearest trip setting at or above 1.5 times LRA or 9 times FLA, whichever is greater. Trip settings were verified for all HPCI system DC valve .
motor actuator No deficiencies were identified, Conclusions on Conduct of Maintenance Preventive and corrective maintenance activities were scheduled and coordinated. Maintenance personnel utilized correct procedures, and work packages contained all necessary documentation to cover the work activities. Maintenance foremen were present at each job site providing supervisory oversight as neede M1.2 Potential Breach of Unit 1 Secondary Containment Intearity Insoection Scooe (62707) (92902)
The inspectors reviewed Maintenance Work Order (MWO) 1-97-2506 associated with a five-year preventive maintenance (PM) work-activity for nitrogen relief valve 1T48-F025 Observations and Findinas On March 20, the licensee discovered that nitrogen relief valve 1T48-F025 had been removed on March 18, creating an opening to the atmosphere ~ A system engineer discovered the opening in the exhaust piping to atmosphere during a walkdown of the Nitrogen Inserting System. The system engineer initiated a deficiency card and reported the observed conditions to operations. The Secondary
- Containment bounuary would be breached when the inner door of the airlock was opened. This. relief valve is located inside a personnel airlock associated with maintaining secondary containment integrity for the reactor buildin The exhaust piping for the removed valve provided a one-inch diameter opening to the atmosphere from the airlock The inspectors reviewed Maintenance Work Order (MWO) documents associated with the work activity and discussed the work with
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maintenance personnel. The inspectors noted that the MWO and the associated clearance'did not contain any precautionary notes regarding the loss of the secondary containment boundary upon removal of the valve.
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i The. inspectors later observed that notes indicating that secondary
! containment boundary would be breached if the inner door of the l airlock was opened was added to the work package after the
. deficiency was. identified. Maintenance personnel also informed L the inspectors that the precautionary comments were added to the computer MWO data form and the associated clearance.
l As a corrective action, blank flanges were-installed on the l openings in the piping where the valve was removed. The
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inspectors observed the installed blank flanges and also observed that a danger tag had been attached. The' licensee informed the inspectors that procedural enhancements to the clearance procedure l> were being. implemented to prevent similar problems in the future.
L The inspecturs reviewed the data package for the most recent i results of the Secondary Containment Test and discussed the leak
!. tightness of the reactor building with the system engineer The i inspectors were inTormed by the system engineer that no
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operability concern existed for secondary containment. The i Standby Gas Tre.atment System was capable of maintaining the l operability of.the secondary containment with the one-inch
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boundary breach. -The inspectors concluded, based upon the results of the last secondary containment surveillance test, that the engineering assessment was reasonabl Conclusions
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The inspectors concluded that poor work planning resulted in a
! breach of Unit 1. secondary containment. The removal of the nitrogen ' relief valve and the subsequent one-inch opening of the exhaust piping to the outside atmosphere'did not adversely affect l secondary _ containment integrity. For prompt corrective actions.
l the-opening was closed with blank flanges with an attached danger
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- . M1.3 Review of Maintenance' Work and Post-Maintenance Testina of a 1
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Unit 1 Emeraency Core Coolina System Area Cooler !
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! Insoection Scooe (62707) (92901)
On February 18, operations personnel placed Unit 1 area cooler B003A under clearance for maintenance to clean the unit. The inspectors reviewed the applicsble maintenance and post- .
maintenance testing procedures and observed part of the work l activitie Enclosure l l
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f Observations and Findinas In September 1997 following the completion of procedure 421T-TET-014-1S, " Safeguard Equipment Room Coolers Data Collection." Rev. 0, the efficiency of Unit 1 area cooler B003A was determined to be degraded but operability was not a concer A MWO was developed to clean the internal cooling coils. The inspectors observed part of the work activities and observed that procedures were used and that the work was well organize Quality Control (OC) hold points were previously identified and were correctly met. Visual inspections of internal cooling coils were completed as required by procedure. Five studs were replaced in accordance with approved weld process sheets. OC personnel reviewed and approved the welding activity. After maintenance work and following fill and vent of the cooling unit, some leakage was observed. Bolts were retorqued to correct the leakage. The inspectors verified that the bolts were torqued to the required value The inspectors observed that operator logs and the MWO work package did not specifically identify which post-maintenance tests or operability checks were performed following the maintenanc Following discussions with maintenance and operations 3ersonnel, the inspectors concluded that proper checks were made aut were not clearly documerited. The inspectors discussed this observation with operations and planning and control managemen Conclusions Overall maintenance work activities associated with a Unit 1 Emergency Core Cooling System area cooler were well-organized and controlle M1.4 Enaineered Safety Feature (ESF) Actuation Durina Unit 1 Testina Insoection Scooe (62707)(92902)
The inspectors reviewed applicable procedures and discussed the circumstances surrounding an ESF actuation that occurred during the performance of procedure 57SV-D11-016-15. "MSL Radiation Monitor FT." Rev. 5. Edition (ED.) 1. on April 6. 1998 with operations and maintenance personne Observations and Findinas The inspectors reviewed the procedure in use and observed that procedural steps were being signed when completed. Electrical links were opened and closed as required by procedure. A discussion with maintenance personnel indicated that the technicians were experienced at performing the surveillance. The Enclosure
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inspectors did not identify any deficiencies in the performance of the procedur A partial Primary Containment Isolation Signal (PCIS) for Group 2 and Group 5 valves was received during ESF testing per procedure 57SV-D11-016-1S, on April The unit response due to the problemOperations procedure clearly identified personnel identified the that two Group 2 isolation valves did not have indication in the
,. control room. The valves were verified closed by operators using
! the Safety Parameter Display System.
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Maintenance personnel investigated the aroblem and identified an open fuse. The fuse was replaced and t1e PCIS valves were returned to their required position. Licensee management assigned i a team to investigate the problem te identify the possible root l cause and recommend corrective acti ms. The inspectors were later i
informed that a root cause of the p;>blem was not determined and no specific recommended corrective actions were identifie The inspectors verified that Technical S) edification Required Action Statement 3.0.3 was entered and tlat the correct 10 CFR 50.72 four-hour report was made. The inspectors reviewed Licensee Event Report (LER) 50-321/98-002 and observed that the licensee l~ conducted a thorough investigation and the documented results were
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! Conclusions l The inspectors concluded'that Unit 1 systems responded properly to L an open fuse and PCIS. The correct Technical Specification and 10 CFR 50.72 actions were completed. Documentation for the problem
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and corrective maintenance were clear and concise. LER 50-321/98-
, 02 which was developed to report the problem in accordance with l 10 CFR 50.73, was thorough with clear and detailed investigation l result M2 Maintenance and Material Condition of Facilities and Equipment M2.1 Review of Maintenance on Unit 2 Residual Heat Removal (RHR) Check j; Valve 2E11-F031 InSDection Scone (62707) (92902)
On April 23 operations secured the B loop of RHR from torus
- cooling and observed that check valve 2E11-F031B failed to clos The inspectors reviewed licensee actions taken to troubleshoot and repair the proble I Enclosure j
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b. Observations and Findinas The inspectors verified that the correct section of the T Required Action Statement (RAS) 3.5.1.A. was entered when 2E11-F0318 failed to close. Following the failure of the valve to fully close on April 23 maintenance determined that the torque on the gland nut should be reduced to ensure proper valve operation. The re-adjustment was made with the RHR pump in operation in accordance with the procedure. The as-left torque of 10 ft-lbs was the value prior to adjusting the packing to correct the leak on February 12, 1998. The valve continued to leak slightly. The inspectors verified that the as-left torque setting was within the procedural acceptance criteria. The inspectors were informed that maintenance )ersonnel also suspected that corrosion around the packing may lave contributed to the valve sticking problem and had documented this comment on the work package. The packing adjustment may not have caused the problem. The valve was scheduled to be repacked during a system outage scheduled to begin on May 5. A more informed decision would be made as to what actions would be taken following the valve maintenanc A performance history review of this valve by the inspectors indicated that the valve packing was adjusted for a slight leak on February 12, 1998. The leak was reduced to about 1 drop every minutes. The inspectors observed that procedural requirements )
were met for documenting the packing adjustmen The inspectors discussed with maintenance management the actions I which were normally completed following routine-check valve packirg adjustments. Following a discussion with maintenance personnel and a review of the work package and procedures, the inspectors did not identify that any post-maintenance or operability checks were taken to verify proper operation of the valve following packing adjustment. This is identified as Inspector Followup Item 50-366/98-02-01: Review As-Found Conditions, Repair Activities and Post-Maintenance Testing for 2E11-F031B. pending additional review of this issu The inspectors will review the as-found conditions of check valve 2E11-F0318. maintenance actions to repack the valve and actions i to verify proper check valve operation following packing )
adjustments. The inspectors did not identify any concern for
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proper valve operation following the packing adjustment, j c. Conclusions Maintenance personnel used ap)ropriate procedures and conducted detailed documentation of worc performe ;
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M3 Maintenance Procedures and Documentation M3.1 Surveillance Observations (61726)
The inspectors observed all or portions of seven Unit 1 and Unit 2 surveillance activities which included the following:
'42SV-FPX-003-05: Emergency Lighting Surveillance Rev. 8
' 42SV-FPX-004-05: Fire Pum) Test. Rev. 7 34SV-N30-003-25: Main Tur)ine Monthly Surveillance Test, i Rev. 4 34SV-5UV-018-1S: .ECCS Status Check Rev. 6 i
For the surveillance observed, all data met the required acceptance criteria. Troubleshooting and repair activities were necessary for the satisfactory completion of surveillance 34SV-N30-003-2S. as' discussed in section M3.2. The performance of the operators and crews conducting the surveillance was professional and competen M3.2 Power Reduction Due to Combined Intermediate Valve (CIV) Testina
' Problems on Unit 2 Insoection Scoce (62707) (61726) (71707)
The inspectors reviewed Surveillance Procedure' 34SV-N30-003-2 " Main Turbine Monthly Surveillance Test." Rev. 4. and. Technical Requirement Manual (TRM) specification T3.3.10. Turbine Overspeed Protectio Observations and Findinas On April 19, reactor power was reduced to about 40% RTP on Unit 2 when the No. 4 CIV for the main turbine did not close as required-by procedure 34SV-N30-003-2S during testing. The Technical Requirements Manual (TRM) limiting condition for operation (LCO)
required that the main turbine be isolated from its steam suppl , During troubleshooting activities maintenance personnel discovered
~
that a blown fuse in the trip test circuit was caused by a fault solenoid associated with the CIV. The faulty solenoid was replaced and the No. 4 CIV was satisfactorily tested prior to the expiration of the 6-hour time limit of the TR Conclusions
, Maintenance troubleshooting activities were effective for
'
determining the root cause of the blown fuse. The reactor power
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' reduction initiated by the operating staff demonstrated j conservative' decision makin Enclosure
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l l 11 M3.3 Review of Reactor Water level Keeofill System for Units 1 and 2 Insoection Scone (62707)(61726H37551)
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The inspectors conducted a two-year performance review of the keepfill systems. -This ' included the data packages associated with surveillance procedure 34SV-B21-007-0S, " Reactor Water Level Cold Reference Leg Keepfill' System Surveillance." Rev 1. Ed observations and Findinos The inspectors observed from the review of the data packages that l the acceptance criteria was met until November 11, 1997. At this l
time, the surveillance associated with Keepfill Subsystem Janel 2H21-P532B on Unit 2 was unsatisfactory. The failure of tie flow indicator to show a flow rate of less than or equal to 0.1 gph with the system isolated for six hours or greater was the criterion that was not me Following the failure in November 1997, a deficiency card (DC) was written and a MWO was implemente During troubleshooting activities, maintenance personnel discovered that the flow indicator was sticking and that a light tap would free it causing the flow indication to deflect to 0.1 gph or less with the system isolated., The flow indicator was replaced and the surveillance was satisfactorily performed on December 10. 199 The results of the surveillance for keepfill subsystem panel
'2H21-P532B performed on February 13. 1998, were again unsatisfactory due to a failure to meet the same acceptance criterion that caused the previous problem. Maintenance personnel freed the sticking flow indicator by lightly tapping on i The surveillance was subsequently performed satisfactoril The inspectors discussed this problem with operations engineering personnel. Operations engineering itated that the flow testing with the system isolated for determii ng whether the kee? filld system is masking a reference leg-le...: was not part of t7e original system design. This criterion was added'after the installation of the system, The flow meter is susceptible to inaccurate readings at flows approaching zero. Licensee personnel were unable to provide reasons for the problem not manifesting itself until November 1997 or why .it was only a problem on the B" channel of Unit 1. The inspectors did not view the intermittent problem with a sticking flow meter as a safety concer <
Enclosure
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.12 Conclusions Maintenance personnel troubleshooting activities were effective in determining that the flow rate meter was the problem component associated with the unsatisfactory performance of the Keepfill System surveillance. The inspectors did nat view this intermittent problem with a sticking flow meter as a safety concer M3.4 Review of Nonconservative Radiation Monitor Setooints for the Unit 1 Reactor Buildina Ventilation System a; Insoection Scone (61726)(92902)
The' inspectors' reviewed licensee actions following the performanc of procedure 57SV-D11-008-1S, " Reactor Building Exhaust Vent-Radiation Monitor Instrument FT." Rev. Two of the four radiation monitors were discovered to N ye nonconservative setpoint ' Observations and Findinas On April 8. following performance of procedure 57SV-D11-008-1 two rad monitors were found to have non-conservative setpoint The licensee made a 10'CFR 50.72 notification to report this issue. Due to'the radiation monitor setpoints, a delay in the reactor-building ~ isolation for a PCIS Group 2 would have occurre ' The inspectors reviewed TS section 3.3.~6.1 for both units for PCIS
. Instrumentation setpoints and observed that the correct high setpoint for Unit I was equal to or less than 20 nillirem per hour (mrem /hr) and Unit 2 was less than or equal to 60 nrem/hr The as-found condition for the Unit 1 "A" monitor was 55 mrem /hr and the "C" monitor was 51-mrem /hr. The inspectors discussed the as-found condition with maintenance instrumentation and control (I&C).
. personnel. The inspectors were informed that I&C suspected that the setpoints had drifted high. The setpoints were calibrated and
.left within the procedure and TS limit The inspectors discussed the difference in the Unit 1 and Unit 2 TS setpoints with' Nuclear Safety and Compliance (NSAC)' management-as to why Unit 1 setpoints were less-than'or equal to 20 mrem /hr and unit 2 setpoints were.less than or equal to 60 mrem /hr. The licensee did not offer any explanation for this differenc The inspectors reviewed a calculation and evaluation provided by
. corporate engineering which indicated that offsite doses would be restricted to well within 10 CFR 100 limits with radiation monitor setpoints up to 1200 mrem /h Following the com)letion of the evaluation. the licensee determined that the pro)lem did not Enclosure.
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require a 10 CFR 50 72 notification and on April 29 withdrew the previously made. report to the NR The inspectors reviewed completed surveillance procedure
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575V-011-008-1S/2S for both units back to June 1996, to gain an
! understanding of previous as-found radiation monitor setpoint
! conditions and to ensure that procedure and TS acceptance criteria'
! were met. The inspectors observed one other instance where the
!- as-found' condition of the radiation monitor had drifted above the
- setpoint. The instrument was calibrated and corrected. The
- . procedure and _TS acceptance. criteria were met for the procedures l which the. inspectors reviewed.
l-ihe inspectors observed that on June 28, 1996, the "D" radiation monitor for Unit I was found and left at 21 mrem /hr instead of the required TS: limit of 20 mrem /hr. The procedure had been reviewed L by supervision and signed that the surveillance )rocedure was.
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acceptable. The review failed to identify that tie surveillance l was not satisfactory. This problem was also identified by the l' iicensee during their review of the problem. The licensee
[ immediately informed the inspectors of the finding.
! 1
! The inspectors reviewed I&C performance for the past two years and l concluded that-performance had been satisfactory. Based upon the l inspectors * review of licensee actions, this issue constitutes a violation of minor safety significance and is being identified as .
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j Non-Cited Violation (NCV) 50-321/98-02-02: Failure to Meet l Technical Specification Requirements for a Ventilation Radiation Monitor Setpoint, consistent with Section IV of the NRC L . Enforcement Polic The inspect' ors observed that the surveillance procedures provided guidance on how to calculate the procedure acceptance criteria plus or minus 10% of the setpoint. For Unit 1,.the acceptance f criteria of plus 10% was above the TS allowable value' of 20 ;
MREM /hr. However, the procedure indicated that the TS acceptance i criteria was not to be exceeded. The inspectors observed that I&C l l routinely. calibrated the instrument and left the setpoint equal to
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the TS limit. This did not allow any margin for upward drift of the instrument without exceeding the TS limit. The inspectors discussed this observation with I&C.' chemistry, and NSAC
. personne c .- Conclusions -
!- Non-Cited: Violation (NCV) 50-321/98-02-02: Failure to Meet Technical Specification R;quirements for a Ventilation Radiation
- Monitor Setpoint; was identified. A direct contributor to the violation was a poor supervisory review of the completed l surveillance that: failed to identify ~that the TS acceptance j
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Enclosure i i L
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criteria was not mee This was an area identified for improvement. A poor work 3ractice for leaving instrument setpoints equal to the Tec1nical Specification limit, was identifie M3.5 Unit 2 Hiah Pressure Coolant In.iection (HPCI) Online Maintenance Outaae and 00erator Performance Durina Testina Insoection Scooe (61726) (62703) (71707)
The ins)ectors .erved maintenance activity associated with the Unit 2 iPCI online maintenance outage and discussed the work activity with maintenance and engineering personnel. In additio the inspectors reviewed procedures 34SV-E41-002-25. HPCI Pump Operability." Rev. 26. and 571T-CAL-001-2S. 'HPCI Turbine Control FT&C." Rev 5. and observed system component, and operator performance during the post-maintenance functional testing activities, Observations and Findinas The inspectors observed on April 21. that maintenance personnel had discovered a small piece of babbit material in the oil sump of the gear box located between the HPCI main and booster pumps. The babbit material was discovered during a planned pre-outage work activit Further investigation by maintenance personnel revealed )
that the babbit on the low speed thrust bearing nearest the main i HPCI pump was damaged. The damaged thrust bearing was replace The system engineer informed the inspectors that the damaged bearing did not affect operability of the system. This assessment was based upon the bearing temperatures and shaft vibration ;
reading for the system surveillance prior to the HPCI outag l These parameters were within the normal range. The inspectors concluded this assessment was reasonabl Following the scheduled HPCI outage. the operators performed the surveillance for system operability and Inservice Testing requirement Maintenance. Instrumentation and Control, and '
engineering personnel were present and provided good support to operations for the testing activitie The system engineer was also present in the control room and observed system and parameter respons ;
The inspectors reviewed completed test data and observed that all surveillance test and TS requirements were met. System operation and response was norma Enclosure 1 . _ _ _ _ _ _ _ _
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15 Conclusions
- < Maintenance personnel displayed a -questioning and attention-to-J ' detail attitude during the Unit 2 Figh Pressure Coolant Injection
! online maintenance outage. This resulted in the timely discovery of the damaged HPCI thrust bearin The maintenance and testing activities were well-planned and coordinated. Maintenance and engineering support to operations ~was evident for the planning of i the work activities and the subsequent return of the system to service M8 Miscellaneous Maintenance Issues (92700) (92902)
M8.1 -(Closed) LER 50-366/98-01. Hiah Pressure Coolant In.iection System
- nocerable Durina Maintenance Investigatio This LER is discussed in Section M3.2 of IR 50-321, 366/97-1 No l new information was revealed by the LER. This LER is close M8.2; (Closed)'LER 50-321/98-02: Blown Fuse Results in Unolanned Actuations of Enaineered Safety Feature This- LER is discussed in Section M1.4 of this Inspection Repor . l No new information was presente Based upon the inspectors review of licensee actions. this LER is close III. Enaineerina
'E2' Engineering Support of Facilities and Equipment
. E2.1' Review of Potential 10 CFR Part 21. Reoortino of Defects and Noncompliance
- In12ection Scooe (37551) (92903)
- The insperters teviewed a 30-day report for a potential 10 CFR
' Pert 21' defect identified at another. facility. The potential L defect was associated with a potential failure for spring retur binding in General Electric .(GE) type SBM Control Switches. The inspectors reviewed and assessed the licensee's evaluation for
' applicability at the Hatch facility in accordance with Regulatory
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Compliance procedure. 03RC-CPL-002-0S, " Defects and Noncompliance."
Rev. 2.
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I Observations and Findinas
- A potential 10 CFR Part 21 defect report from another facility i
identified a possible failure of certain GE Type SBM switches to reset properly after operation. The control switches are l manufactured by GE Electric Distribution and Control Power l Management as commercial grade item GE had supplied these switches to several licensees as basic components for unspecified safety-related a)plications. Based on the preliminary '
investigation t1e root cause was suspected to be " post-mold cure" shrinkage of the phenolic material used to construct some of the switch component Engineering 3ersonnel reviewed the potential 3roblem and determined tlat one switch was installed on t1e Unit 1 HPCI vacuum pump and that no previous problems had been identified with the switch. Another switch of concern had not been installed but was listed on a work order for future installation. The evaluation concluded that no safety concern existed for the installed switc The inspectors reviewed the evaluation and concluded that the licensee *s evaluation and assessment was reasonabl The inspectors observed that GE suggested in the 10 CFR Part 21 defect report which was issued in January 1998, that station personnel who operate these switches be advised to return the operating handle to the normal position after operation. The inspectors discussed this with operators and operations supervision in the control room for both units. None of the operators or the operations manager could recall any discussion or training about the hand switch problem. The inspectors were later informed that the suggestion had not been followed. The inspectors observed that 'in this case, there were no consequences for not following the GE recommendation to inform personnel of the
. potential proble Conclusions The inspectors concluded that the licensee had taken appropriate action in addressing the potential 10 CFR Part 21 defect associated with GE Type SBM Control Switche i E2.2 Review of Reactor Core Isolation Coolina (RCIC) Inservice Testina (IST) Requirements Insoection Scooe'(37551) (71707) I The inspectors reviewed procedures 40AC-ENG-001-0S "ASME Section XI-Program." Rev. 10, 42EN-INS-001'-0S, " Inservice Testing Program." Rev. 5. ED1. and 31G0-INS-001-0S, "ISI Pump and Valve j
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Enclosure
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Operability Tests." Rev. 9. ED 1, which identifies the requirements for RCIC valve testin Observations and Findinas In preparation to perform Unit 2 procedure 34SV-E51-001-2S. ~RCIC Valve Operability," Rev.11, operations aersonnel questioned if valve 2E51-F003 (pump suction - torus in)oard isolation) should-have an IST reference time and calculated allowable time for the closed direction. The IST data reference book maintained by control room operations personnel indicated that these times were not applicable. The IST engineer was contacted for assistanc The inspectors reviewed Table 11 for the Unit 2 IST Program and observed that valve 2E51-F003 had a safety function for.both the o)en and closed direction. The actual closing time of the valve slould have been compared to a reference tim Comparing valve stroke times to reference data was used as an early indictor for possible valve degradation. Procedure 345V-E51-001-2S was performed quarterly to satisfy the IST requiremen The inspectors reviewed surveillance results for Unit 2 valve data completed in 1997 and observed the reference and calculated allowable times were marked not applicable. -The inspectors observed that the valve closing time was compared to a maximum t closing time limit and the valve closing time met the surveillance-requirement of equal to or less than six seconds. The inspectors were informed that the IST reference time for valve 2E51-F003 to close was established in November 1996. and was 1.5 seconds The-inspectors observed that all IST-completed surveillance were forwarded to engineering for review and approval for IST acceptanc In this case, the deficiency was not identified. The inspectors concluded that a detailed review would have identified the deficienc The inspectors calculated the IST allowable time and observed that, for the surveillance previously completed, the actual valve closing times met the IST and procedure requirements. The failure
.to document and compare the closure time of RCIC valve 2E51-F003 to an IST reference time constitutes a violation of minor safety significance and is identified as Non-Cited Violation (NCV)
-50-366/98-02-03. Failure to Meet RCIC IST Valve Testing
. Requirements, consistent with Section IV of the NRC Enforcement
. Polic Conclusions The failure to document and compare the closure time of Unit 2 RCIC valve 2E51-F003 to an IST reference time was identified as Non-Cited Violation (NCV) 50-366/98-02-03. Failure to Meet RCIC Enclosure l
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~ IST Valve Testing Requirement Engineering review of completed surveillance to ensure that inservice testing requirements were met, did not detect deficiencies and was identified as an area for improvement.
E _oss of Glycol Volume in Emeraency Diesel Generators (EDG)
Exoansion Tanks
! Insoection Scooe (37551) (62707)
The inspectors reviewed Piping and Instrumentation Diagram (P&ID)
H-11038f Demineralized Water: the 10 CFR 50.59 evaluation: and clearance 1-98-106. This review was associated with the loss of l' glycol' coolant from the EDG ex)ansion tanks during the implementation of the. design clange package for the replacement of the Unit 1 turbine building chiller Observations and Findinas
! On March 23, while implementing a design change for the i installation of permanent turbine building chillers on Unit 1, a:
demineralized water line was cut by craft personnel. The
, demineralized water line that was cut provided makeup water to the glycol expansion tanks associated with the emergency diesel generators'(EDG). -
Craft personnel expected-water to flow from the cut in the piping due to entrained water. After making the initial cut, craft personnel observed the flow of glycol coming from the cut in the piping'following the drainage-of the clear demineralized wate The cutting of the piping was immediately stopped and maintenance and operation personnel placed a clamp on the cut to prevent further leakag The inspectors reviewed the clearance sheets and discussed the
. loss of glycol from the EDG expansion tank with the writer of the clearance. The clearance writer and engineering personnel informed the inspectors that the siphoning effect on the tanks had not been considered prior to the work activit The: inspectors discussed with maintenance and engineering personnel the im)act that the loss of inventory from the expansion i tanks had upon t1e operability of the EDGs. ~The inspectors were informed that the' inventory loss had no impact upon EDG operability, particularly, because there were no leaks in the EDG coolant' syste Enclosure m- --- - - - -- - - - - -- - - - - - -_ _ - - - - - , - - _ - - - - - - - - _ __ D
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Based upon a review of the P&ID and a walkdown of the demineralized water system at the EDG expansion tanks.,the L inspectors concluded that the licensee's engineering assessment which showed no impact on EDG operability, was reasonabl A nitrogen. freeze seal was subsequently used on the piping near the cut so corrective maintenance could be completed. The work activity was com)leted successfully with no further loss of inventory from t1e expansion tank.
! 'onclusions C
i 0)erations and Maintenance personnel's prompt actions in clamping l t1e cut made on the demineralized water piping effectively
- minimized the loss of glycol inventory from EDG expansion tanks l due to siphoning actio The clearance writer and design-engineering failed to consider the effects of siphoning on the I expansion tanks. This problem did not affect the EDG operability.
E3 Engineering Procedures and Documentation E3.1 Review of 10 CFR 50.59 Safety Evaluation Proaram I Insoection Scooe-(37001)-
- The inspectors assessed the licensee's program. procedures and-implementation of the requirements of 10 CFR 50.59 for conducting
'
safety evaluations related to facility changes as described in the UFSAR. A review was conducted of selected safety evaluations for both planned and completed design changes, procedure changes, and temporary modifications. Also evaluated were the licensee's safety evaluation 3rocedures, and training and qualification of personnel responsi)le for implementing the 50.59 safety evaluation program.
l- Chservations and Findinas L = Safety Assessment Training and Qualifications l
In late 1997, the licensee conducted training for all site
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personnel responsible for the preparation and review of ,
10 CFR 50.59 evaluations. The applicable procedures were 1 revised to include im !
reflected management' proved guidance and direction thats commitmI The inspectors observed an overall improvement in the detail and quality of recently performed 10 CFR 50.59 evaluations, as compared to those performed in mid-1997 and befor ;
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Enclosure
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l One of the licensee's training requirements was that all personnel '(including corporate personnel) who perform or review 10 CFR 50.59 evaluations were to read and understand the latest revision of the administrative procedure for the development and review of evaluations. The inspectors reviewed a site training list and observed that all applicable personnel had
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received the required training. The inspectors also selected twelve individuals from the site and corporate engineering that had developed or reviewed 50.59 evaluations between 1991 and
- 1998, to verify that required training had been met. All personnel selected had received trainin * 10 CFR 50.59 Assessment and Procedure Review The inspectors reviewed procedure 10AC-MGR-010-05. " Preparation and Ap3roval of 10 CFR 50.59 Evaluations." Rev. 3. and observed that t7e procedure had been revised and included appropriate- ,
and detailed guidance for the preparation and review of safety l evaluation I The inspectors assessed the licensee's use of the procedure in the evaluation process. The inspectors reviewed Licensing Document Change Request (LDCR)98-027. dated March 19, 1998, for changes made to sections 3.2.1.5. 5.4.3 and 11.0 of the site security plan. The inspectors also reviewed the 10 CFR 450.54(p) evaluation conducted to sup) ort the change The evaluation indicated that the. proposed clanges did not decrease the overall level of effectiveness of the plan. The inspectors observed that the evaluation was of sufficient detail to support the conclusions. Procedure requirements were me The inspectors reviewed six Unit 1 and six Unit 2 50.59
. evaluations for Design Change Requests (DCRs). Minor Design ,
Changes (MDCs), and procedure changes. In general, procedure ;
requirements were met. However, there were examples where the 3 answers to screening questions were not of sufficient detail such that an independent reviewer could arrive at the same conclusio For example:
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Unit 2 DCR 2H97-017: This DCR rewired motor operated valves susceptible to mechanical damage resulting from fire-induced hot shorts. The evaluation stated that the seismic capability of the supports for the raceways affected by the DCR has been evaluated and found to be acceptable. However, there was not sufficient detail in the 50.59 evaluation for the inspectors to review the licensee's seismic study, nor was there a specific reference to the seismic stud Enclosure
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Unit 1 DCR 97-20: This DCR eliminated a number of snubbers on various systems. The evaluation stated that a structural analysis had been completed. However, the evaluation did not provide information about the analysis nor was there a specific reference to the analysi The inspectors reviewed DCR 97-15 for Unit The DCR rerouted cables for a radiation monitor through a cable penetration into the drywell. The cable penetration used had been made available as a spare when DCR 94-13 was complete The inspectors reviewed plant drawings to verify that they accurately reflected the current plant condition. The inspectors observed that the applicable drawing indicated that the cable penetration was a spare even though DCR 97-15 was i completed. This discrepancy was brought to the attention of l maintenance supervision who initiated a review of the proble l The inspectors were later informed, and were presented ap3ropriate documentation, that indicated the as-built-notice (A3N) used to change )lant drawings for DCR 97-15 was '
completed before the A3N for DCR 94-13. This meant that the ABN for DCR 97-15 changed the drawing to reflect that the spare penetration was being used and the ABN for DCR 94-13 changed the drawing to reflect that the cable penetration was a spar The inspectors later verified that the drawing had been corrected. The inspectors did not view this discrepancy as significant, based upon the short time the problem existed and it appears to be an isolated exampl * Plant Review Board Meeting The inspectors attended a Plant Review Board (PRB) meeting to assess licensee management participation and review 10 CFR 50.59 evaluations for procedure revisions. Procedures from the operations and maintenance departments were included. The inspectors observed that the PRB requirements specified in Section 1 of the Hatch Quality Assurance Manual and Section 17 of the UFSAR were met. The inspectors observed that the procedure and 50.59 review completed by PRB members was thorough and detailed. The reviews identified minor deficiencies that were corrected prior to procedure approva * Audit Review of 10 CFR 50.59 Evaluation Program The inspectors reviewed audit report 96-SA-11. Audit of 10 CFR 50.59 Safety Evaluation Program for the site, and Audit 96- for Corporate, dated August 1996. The inspectors also reviewed audit 96-ADM-2. Audit of Administrative Controls, dated January 1997. The inspectors observed that the audit checklists were thorough and detailed. The audits were conducted by corporate and site personnel knowledgeable of the safety evaluation Enclosure l
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l l 22 program. ihe audit evaluation concluded that overall, the 10 CFR.50.59 evaluation process was being implemented effectively
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at Plant Hatch. However, the auditors identified several areas for improvement in the process and program. The inspectors observed that the. response and corrective actions to audit i
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findings.were timely and appropriate to ensure program improvemen Conclusion The inspectors concluded that the 10 CFR 50.59 Safety Evaluation Program was being effectively implemented. The detail and quality of the 50.59 evaluations improved over the last 18 months. Some evaluation questions were not answered in sufficient detail to ensure an independent reviewer could draw the same conclusion and l arrive at the same answe Site and corporate personnel received the required training to conduct 50.59 evaluations. The safety evaluation procedure included appro)riate and detailed guidance for'the preparation and review'of 10 C R 50.59 evaluation The PRB met the requirements specified in Section 1 of the Hatch Quality Assurance Manual and Section 17 of the UFSAR. Procedure and 10.CFR 50.59 reviews completed by PRB members were thorough.
L l Audits in the area of 10 CFR 50.59 reviews were thorough and
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detailed. The audits were conducted by corporate and site personnel knowledgeable of the safety evaluation program. The response and corrective actions to audit findings were timely and appropriate to ensure program improvemen E8 Miscellaneous Engineering Issues (92700) (92903)
!
E8.1 LClps_ed) VIO 50-366/97-03-04: Inadeauate Procedures for Testina t Activities - MultiDie Examole )
L The licensee responded to this violation in correspondence dated July 17, 1997. The licensee's corrective actions for the violation included-training, counseling individuals involve , issuance of departmental directive to communicate management l expectations, and procedure changes to clarify steps. Based upon
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the inspectors review of licensee actions, this VIO is close Enclosure
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E8.2 -(Closed) Licensee Event Deoort (LER) 50-366/98-02: Assembly Error Leads to Malfunctioning uovernor Causina Emeraency Diesel Generator Inocerability.
l-l l ' Subsequent vendor failure analysis and bench testing of the.
I malfunctioning governor revealed that the governor shutdown solenoid had been improperly assembled during governor refurbishment in 1988. Specifically, an air gap washer was
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l- mispositioned in the solenoid assembly, causing the solenoid to
" stick" in the shutdown position. Woodward communicated their
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findings.to.the licensee and concluded.that this assembly error
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was'an isolated occurrence. The inspectors reviewed the-l licensee's failure evaluation documentation and Woodward's rationale for concluding this issue to.be an isolated case and found it to be acceptable. The licensee has initiated corrective action to add EDG governors to a preventive maintenance schedule
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for refurbishment and/or replacement on an initial 9-year I frequency.-with the frequency to be adjusted based on refurbishment results. Based on a review of the licensee *s corrective actions, this:LER is close E8.3- (Closed) Apparent Violation (EEI) 50-321, 366/98-01-03, Plant Operation Outside of the Design Basis for an Engineered Safeguar : System Following the NRC's review of the Apparent Violation.-the NRC j concluded that the violation met the requirements of Section VII- {
of NUREG-1600, for Violations involving old design issues. As a result, NCV 50-321, 366/98-02-04, Plant Operation Outside of the Design Basis for an Engineered Safeguard Systen. was identifie Based upon the NRC's review and conclusion Apparent Violation (EEI) 50-321. 366/98-01-08 is closed.
l IV. Plant Support
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R1 Radiological' Protection and Chemistry Controls R1.1 The inspectors observed during routine plant tours that housekeeping has generally improved overall since the last report period. Plant Health Physics (HP) has placed increased emphasis upon area decontamination and the removal of debris from contaminated areas.
l Enclosure
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RI.2 Tour'of Radiological Protected Areas t Insoection Scooe (83750) 1 i
The inspectors reviewed implementation of selected elements of the l licensee's radiation protection program. The review included observation of radiological protection activities including control of radioactive material, radiological surveys / posting and radiation area /high radiation area centrol ' Observations and Findinas ;
During tours of the turbine building. Units 1 and 2 reactor buildings, radioactive waste building, and storage and handling facilities, the inspectors reviewed survey data and observed activities in progress. The licensee had effectively posted areas where radioactive material was stored and radioactive material ,
observed was labeled as required. During tours the ins)ectors
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also observed that Locked High Radiation Areas were locced as ;
required by licensee procedures. Radiological surveys reviewed I
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were well documente As of May 1.1998, approximately 15 Personnel Contamination Events (PCEs), defined as 1000 but less than 10.000 disintegrations per ;
minute (DPM). had occurred during 1998 which included both !
particles and dispersed contamination events for clothing and skin contaminations. Approximately 14 Personnel Contamination Reports (PCRs), defined as 10.000 DPM or greater had occurred in 199 The inspectors noted that licensee efforts during 1998 to date to reduce personnel contaminations had been positiv Radiation Work Permits (RWPs) established for Jerforming work were reviewed. These controls included the use of RWPs to be reviewed and understood by workers prior to entering the RCA. The inspectors reviewed selected RWPs for adequacy of the radiation protection requirements based on work scope. location, and I conditions. For the RWPs reviewed. the inspector noted that appropriate protective clothing and dosimetry were require During tours of the plant, the inspectors observed the adherence of plant workers to the RWP requirement During tours, radiation monitors were observed to have been calibrated to meet program calibration frequencie Conclusions Based'on observations and procedural rc .ews, the inspectors determined that the licensee was effectively maintaining controls for personnel monitoring, control of radioactive material, radiological postings, radiation area controls, and high radiation l
Enclosure l
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area-controls, as required by 10 CFR Part 20. Efforts to reduce personnel contaminations during 1998 to date had been positiv R2 = Status of RP&C Facilities and Equipment R2.1 Loss of Reactor Coolant System Primary Flow to Conductivity
. Element on Unit 1 Insoection Scooe (71750) (37551) (62707) (92904)
The inspectors reviewed the applicable piping and instrumentation diagrams (P&ID), the Technical Requirement Manual (TRM), and procedures associated with the loss of reactor coolant system (RCS) flow to the conductivity measuring element for Unit Observations and Findinos On April 13. a chemistry technicianwas not able to obtain l sufficient RCS flow for the performance of Chemistry procedure 64CH-0CB-008-05. "Iontrac System." Rev.'1. A deficiency card'(DC)
-was written and an initial investigation was performed by chemistry supervision. This initial investigation revealed that j-there was also no RCS flow to the reactor coolant continuous in-line conductivity element. The RCS conductivity and surveillance requirements were referenced in TRM section T3.4.1 RCS Chemistr Chemistry aersonnel discovered that pressure control. valve (PCV)
.1P33-F113 lad failed. Maintenance personnel replaced IP33-F11 During the troubleshooting activities, chemistry personnel identified minor deficiencies on a chemistry P&ID for abandoned equipment and initiated corrective action The inspectors observed portions of the ongoing troubleshooting ,
and repair activities and reviewed the proposed P&ID and procedure l changes to correct the minor deficiencie l Conclusions The. chemistry department's persistence in determining the root
- cause for the loss of primary RCS flow to the continuous in-line -
conductivity cell resulted in a detailed and timely solution to the problem. The inspectors concluded that effective corrective actions-were taken~for minor P&ID problem i I
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~ R7 Quality Assurance -in Radiation Protection (RP) and Chemistry
. R7.1 ' Quality Assurance Audits Insoection Scoce (83750)
10 CFR 20.1101 requires that the licensee periodically review the RP program content and implementation at least annually. Licensee periodic reviews of the RP program were reviewed to determine the
~a dequacy of problem identification and corrective action Observations and Findinas Reviews by the inspectors determined that Quality Assurance review
in the area of RP were accomplished by reviewing RP procedures, observing work. reviewing: industry documentation, and performing
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plant'walkdowns. including observance of work areas by supervisors and technicians during normal work. coverage. Documentation of problems by licensee representatives was included in Quality Assurance Audits. The inspector reviewed the licensee's most P recent audit in the area of radiation protection, audit report number 98-HP-1.. dated April 2<8. 1998, and the checklist used during the )erformance of the audit. The inspectors ~also discussed t1e audit findings with the auditors and RP managemen The audit identified two items of substance that required RP management respons Conclusions .
The inspectors determined that the licensee's most recent formal Quality Assurance Audit identified items of substance and that auditors,used' checklists to effectively assess the radiation protection' program as required, by 10 CFR Part 20.110 R8~ . Miscellaneous Radiation Protection and Chemistry Issues (83750, 84750)
- R (Closed) Insoector Followuo Item (IFI) 50-321. 366/96-10-12:
Verification of Adequate. Testing of the Kaman Effluent Accident Range Monito After further review ~of this issue, it was determined that the i valves used to switch filters are based on a? low flow alarm which was' tested by procedure 57SV-D11-021-1S, ~ Building Vent Radiation Monitor FT and C;" Revision 1. Ed. 1. Based on procedural information reviewed, this item is close Enclosure y
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P1 Status of CP Facilities. Equipment, and Resources Pl.1 Review of Emeraency Doeratina Procedure (EOP) Eouioment Availability Insoection Scooe (82301) (92904)
The inspectors reviewed procedure 341T-EOP-001-0S "EOP Equipment Checks, Rev. 12. and conducted a review of selected E0P equipment availability as delineated in the procedur i Observations and Findinas The inspectors conducted a walkdown of both units to' verify that selected E0P equipment was at the designated location and available for. use during emergency conditions. The inspectors '
observed that the E0P equipment cabinets were in their designated locations in the reactor building and diesel generator building and were secured and lockwired closed. as required by procedur i E0P jumper binders were in their designated storage location in both control rooms and were in good condition. Stepladders were properly stored near E0P cabinets. Special ecuipment, such as the modified flange connection for the control roc drive system suction filter, were in their designated location Conclusions The inspectors concluded that tools and equipment used to implement the Emergency Operating Procedures for both units were in their designated storage locations and were well-maintaine P4 Staff Knowledge end Performance in EP P4.1 Medical Emeroency Drill Insoection Scoce (82301)
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The inspectors observed licensee performance during a Medical Emergency Drill conducted on April 2 Observations and Findinas On April'21. the licensee conducted a Medical Emergency exercise to demonstrate that personnel assigned responsibilities in an
. emergency situation were adequately trained to perform in accordance with emergency preparedness plans and procedure ;
The drill consisted of two phases. Phase one was the on-site
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l response to a contaminated injured person that was transported to the Meadows Regional Center in Vidalia. Georgi Phase two was for the transfer of the injured individual to the Memorial Medical Enclosure j
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Center in Savannah, Georgia. The drill included elements of a l Communication drill, Radiological Monitoring drill and Radiological Emergency drill.
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l The inspectors observed the on-site response portion of the drill for the contaminated injured individual. The inspectors observed l that the drill controllers provided pertinent scenario information l
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for the drill participants to assess the situation and make decisions to correctly implement emergency procedures. Health l Physics (HP) personnel provided adequate coverage and took appropriate actions to prevent the spread of contamination. HP coverage included the necessary actions to move the simulated
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injured individual from the Unit 1 reactor building to a waiting
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ambulance'for transport to the Meadows Regional Hospita Security personnel provided the required support for onsite arrival of the ambulance. Emergency Medical Technicians provided
, medical attention and first ai The inspectors observed that, during the initial phase of the simulated injury, approximately 12 individuals responded to the scene. However, no one individual took charge of the scene for command and control functions and to ensure that pertinent information was communicated to the control room. The inspectors reviewed procedure 73EP-EIP-013-05. Step 4.1. of the procedure which stated, in part, that "the most senior (quali'ied)
' individual must take charge AND ensure that actions are performed and decisions are made." The inspectors discussed the requirements of the procedure step with EP personnel with respect to who determined who was the most senior (qualified) individual and what were management's expectations to meet the procedure step. The inspectors were informed that the procedure step was unclear to EP personnel and would be reviewed for improvement.
. In addition, the first responder organization including command and control responsibilities, had not been clearly delineated or define Information communicated to the control room did not identify that the simulated injured individual was contaminate Subsequently, information relayed to the hospital was incorrec The exercise controller intervened and corrected this problem. As a result, exercise objectives C1: Demonstrate that an Individual is Assigned and is in Charge of the Emergency Response, and E-2:
Demonstrate the Adequacy of Communications to Offsite Response Agencies (Meadows Memorial Hospital and Local Ambulance Service). ,
were not me )
i Operations correctly classified the exercise event as a l Notification of husual Event (NUE). in accordance with procedure !
73EP-EIP-001-05. The procedural time requirement for the :
notification was me Enclosure
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'One of the inspectors 3roceeded to the Meadows Hospital in L Vidalia. Georgia.and caserved'that HP coverage and support for arrival of the simulated contaminated injured individual was L provided. Hospital personnel expected the arrival of the
! ambulance. Site HP personnel provided radiological survey results L 'to hospital personnel. Proper radiological and contamination:
control barriers were already established. This included roped
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contamination and radiation signs. All individuals involved in the drill activity wore dosimetry and. adequate protective
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clothing The inspectors observed that hospital medical supervision, local Emergency Management, and law enforcetant personnel were presen Conclusions Information communicated to the control room during the medical emergency drill conducted on April 21 did not identify that the simulated injured individual was contaminated. Subsequently, information relayed to the hospital was incorrect. As a result, exercise objective E-e: Demonstrate the Adequacy of Communications to Offsite Response Agencies (Meadows Memorial Hospital and Local Ambulance Service), was not met. The first responder organization including command and control responsibilities, had not been clearly delineated or defined. As a result, exercise objective'C1: Demonstrate that an Individual is Assigned and is in Charge of the Emergency Response, was not-me Operations, Health Physics. Security and Emergency Medical responders provided the required support during the exercis .
S2 Status'of Security Facilities and Equipment (71750)
The inspectors. toured the protected area and observed'that the perimeter fence was intact and not compromised by erosion nor
' disrepair The fence fabric was secured and barbed wire was
~ angled as required by the licensee's Plant Security Progra Isolation zones were maintained on both sides of the barrier and were free of objects which could shield or conceal an individual The inspectors observed that personnel and packages entering the protected area were searched either by special purpose detectors or by a physical patdown for firearms, explosives, and contraban Badge issuance was observed, as was the processing and escorting of visitors. Vehicles were searched, escorted and secured as
~ described in applicable procedure The insp'ctors e concluded that the areas of security inspected met the applicable requirements.
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l 30 F5 ~ Fire Protection Staff Training & Qualification F5.1 Announced Fire Drill Insoection Scooe (71750)
The licensee conducted an announced fire drill on May 1,199 The inspectors monitored the activities of the fire brigade at the scen Observations and'Findinas Licensee fire protection personnel simulated an electrical fire in the 2G Emergency Diesel Generator-(EDG) switchgear room. The fire brigade responded as directed by the control room announcemen Fire brigade members and security personnel were prompt and efficient in their initial response and staging of fire fighting equipmen The insaectors observed that the actions taken by the fire brigade.~attact" team in initial entry to the switchgear room were closely directed by the brigade Incident Commander. The inspectors discussed the following observations with the plant <
fire marshall coordinating the drill:
. U on initial deployment of fire hoses from the nearest h drant station, no one from the fire brigade was posted at t e station to act-as hydrant operator. This delayed the initial attack team entry until a brigade member could be posted to operate the hydrant to supply water to the hose . After the attack team entry, status messages from the attack team leader to the Incident Commander were spars Specifically, when the attack team reported that the fire had spread to a switchgear room cable tray, several minutes elapsed before the next report was passed that the fire was contained. In general, the inspectors observed that there were very few reports of the fire status passed on to the Incident Commander Conclusions The fire protection personnel provided an effective drill for the fire brigade. Response and coordination objectives were me Minor deficiencies associated with a failure to initially establish a hydrant operator and sparse followup messages to the fire brigade-leader were discussed with the fire drill coordinator. These deficiencies were discussed in a post-drill critique and identified as areas to be improve Enclosure
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V. Manaaement Meetinas
. Review of UFSAR' Commitments A recent.-discovery of a licensee operating its facility in a manner contrary to the Updated Final Safety Analysis Report
'(UFSAR) description highlighted the need for a special focused review that compares plant practices, procedures and/or parameters to the' UFSAR description. 4hile performing the ins)ections discussed in this re) ort. the inspectors reviewed t1e applicable portions of the UFSAR that related to the areas inspected. The inspectors verified that the UFSAR wording was consistent with the observed plant practices procedures, and/or parameter Exit Meeting Summary The inspectors presented the inspection results to members of the licensee management at the conclusion of the inspection on May 14 1998. The licensee acknowledged the findings presented. An ;
interim exit was conducted on May 1.1998. The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identi fie Other NRC Personnel On Site On March 26 and 27 Mr. P. H. Skinner visited the site. 'He met with the resident inspector staff and discussed licensee performance _and regulatory issues. He toured the plant to observe equipment in service and overall plant conditions. He attended the morning management meeting and met with licensee managemen personnel. He also met with the Assistant General Manager -
Operations and discussed plant performance and regulatory issue From March 31 through April 2 Mr. L. Olshan visited the site to assist the resident inspector staff in conducting an inspection of the safety evaluation program under 10 CFR 59.5 Issues related
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to this inspection are documented in section E3.1 of this repor PARTIAL LIST OF PER30NS CONTACTED Licensee l
Anderson. J., Unit Su)erintendent Bennett. D.. Health 3hysics Superintendent Betsill.- J., Assistant General Manager - Operations ;
Breitenbach. K., Engineering Support Manager - Acting l Curtis. S. , Unit Superintendent 1'
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Davis, Plant Administration Manager Enclosure I
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Fornel, P.. Performance Team Manage Fraser. 0. . Safety. Audit and Engineering Review Supervisor Hammonds. J., Operations Support Superintendent Kirkley, W., Health Physics and Chemistry Manager Lewis, J., Training and Emergency Preparedness Manager Link, M., Health. Physics Supervisor Madison. D. , Operations Manager Moore. C., Assistant General Manager - Plant Support Reddick. R . Site Emergency Preparedness Coordinator Roberts. P., Outages and Planning Manager Smith, D,, Chemistry Superintendent Thompson, J. . Nuclear Security Manager Tipps' S., Nuclear-Safety and Compliance Manager Wells, P., General Manager - Nuclear Plant Other licensee employees contacted during the inspection included technicians,
. maintenance personnel and administrative personne INSPECTION PROCEDURES USED IP 37001: 10 CFR 50.59 Safety Evaluation Program IP 37551: Onsite Engineering IP 61723: Surveillance Observations IP 62707: Maintenance Observations IP 62703: Maintenance Observations IP 71707:~ Plant Operations IP 71750: P1 ant Support Activities IP'82301: Evaluation Of Exercises For Power Reactors
.IP 83750: Occupational Exposure IP 84750: Solid Radioactive Waste Management md Transportation of Radioactive Materials IP.92700: 'Onsite Follow-up of Written.. Reports of Nonroutine Events at Power Reactor Facilities IIP 92901: Follow-up - Operations IP 92902: Follow-up . Maintenance / Surveillance IP 92903: Follow-up - Follow-up Engineering-IP 92904: Follow-up -Plant Support ITEMS OPENED, CLOSED, AND DISCUSSED Ooened 50-366/98-02-01 IFI Review As-Found Conditions,
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Repair Activities, and Post-Maintenance Testing for-2E11-F031B (Section M2.1).
50-321/98-02-02 NCV Failure to Meet Technical Specification Requirements for L a Ventilation Raciation
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Monitor Setpoint '
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50-366/98-02-03 NCV Failure to Meet RCIC IST Valve Testing Requirements (Section E2.2).
50-321, 366/98-02-04 NCV Plant Operation Outside of the Design Basis for an Engineered Safeguard System (Section E8.3).
Closed 50-321, 366/97-02-04 IFI Review of Operator Performance Deficiencies and Licensee Corrective Actions (Section 08.1).
50-321/98-02 LER Blown Fuse Results in Unplanned Actuations of Engineered Safety Features (Section M8.2).
50-321/98-02-01 NCV Failure to Meet Technical Specification Requirements for a Ventilation Raciation Monitor Setpoint (Section M3.4).
50-366/98-01 LER High Pressure Coolant Injection System Inoperable During Maintenance Investigation (Section M8.1).
l l 50-366/97-03-04 VIO Inadequate Procedures for l Testing Activities - Multiple i
Examples (Section E8.1).
50-366/98-02 LER Assembly Error Leads to Malfunctioning Governor Causing Emergency Diesel ,
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Generator Inoperability 1 (Section E8.2).
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-34 50-321. 366/98-01-08 EEI Plant 0)eration Outside of the Design 3 asis for an Engineered Safeguard System q (Section E8.3)
50-321, 366/98-02-04 NCV Plant 03eration Outside of the Design 3 asis for an Engineered Safeguard System (Section E8.3) {
l 50-321, 366/96-10-12 IFI Verification of Adequate Testing of the Kaman Effluent Accident Range Monitor (Section R8.1).
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