IR 05000413/1997005
| ML20138J468 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 04/21/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20138J464 | List: |
| References | |
| 50-413-97-05, 50-413-97-5, 50-414-97-05, 50-414-97-5, NUDOCS 9705080227 | |
| Download: ML20138J468 (27) | |
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U.S. NUCLEAR REGULATORY COMMISSION l
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REGION II
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i Docket Nos:
50-413, 50-414
License Nos:
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i Report Nos.:
50-413/97-05, 50-414/97-05
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l Licensee:
Duke Power Company
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Facility:
Catawba Nuclear Station. Units 1 and 2
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i Location:
422 South Church Street
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Charlotte, NC 28242
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Dates:
February 16 - March 22, 1997.
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j Inspectors:
R. J. Freudenberger. Senior Resident Inspector
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t P. A. Balmain. Resident Inspector
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R. L. Franovich Resident Inspector i
j D. B. Forbes, Reactor Inspector (Sections R1-R7)
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W. M. Sartor. Reactor Inspector (Sections P1-P8)
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Approved by:
C. A. Casto. Chief
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Reactor Projects Branch 1
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j Division of Reactor Projects
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9705080227 970421 PDR ADOCK 05000413 O
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EXECUTIVE SUMMARY Catawba Nuclear Station. Units 1 & 2 NRC Inspection Report 50-413/97-05. 50-414/97-05 This integrated inspection included aspects of licensee operations.
maintenance, engineering, and plant support.
The report covers a six-week period of resident inuection: in addition, it includes the results of announced inspections ]y regional reactor safety inspectors.
Operations Following a loss of indication for two control rods on the Digital Rod
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Position Indication Panel, the licensee and inspector ensured that the unit was operated in compliance with the technical specifications (TS)
and promptly took action to identify the root cause.
Efforts to restore the indications on the Digital Rod Position Indication Panel were timely. (Section 01.1)
Appropriate actions were taken to ensure the steam generator (SG) power
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o]erated relief valves (PORVs) could be manually operated locally within t1e time assumed Dy a revised accident analysis that had been submitted for NRC review.
(Section 02.1)
The Nuclear Safety Review Bo]rd's observations were consistent with
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recent inspection findings.
(Section 07.1)
Maintenance Corrective actions for a tagout error that caused a nonsafety-related e
service water pipe to rupture were effective.
Appropriate revisions to the tagout ]rocess were made to require the same level of review for expanding t1e scope of a tagout as the level of review required for initial tagouts. The safety significarce of the error was low: however, it resulted in work arounds and impeded the licensee's ability to make liquid radioactive waste releases. (Section M1.1)
Plant Operations Review Committee (PORC) review of extending scheduled
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maintenance for preplanned Unit 1 control area chiller work was detailed and focused on safety. (Section M1.2)
Recognizing recent failures of two Main Feedwater Isolation Valves
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(MFIVs) at the McGuire Nuclear Station (MNS), the licensee evaluated the MFIV solenoid valves at Catawba Nuclear Station (CNS) and determined that they were of sufficiently different design than those at MNS to preclude failures similar to those experienced at MNS. (Section M2.1)
Enaineerina A more limiting single failure than was considered in the UFSAR analysis e
of a steam generator tube rupture was appropriately determined to involve an unreviewed safety question. Administrative controls that ENCLOSURE
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l were implemented on an interim basis were appropriate to increase the existing safety margin.
Pending NRC approval of the license amendment
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to require four SG PORVs operable per unit and the use of manual i
o)erator action to mitigate a steam generator tube rupture accident, i
t11s issue was identified as an Unresalved Item.
(Section El.1)
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Plant Suooort
i The licensee effectively implemented a program required by NRC and
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Department of Transportation (DOT) regulations for shipping radioactive materials. (Section R1.1)
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Radiological facility conditions and housekeeping in radioactive waste
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storage areas were adequate. Observations and survey results revealed that the licensee was effectively controlling radioactive material.
(Section R1.2)
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The licensee's water chemistry control program for monitoring primary
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and secondary water quality was implemented, for those parameters reviewed, in accordance with the TS requirements and the Station i
Chemistry Manual for Pressurized Water Reactor (PWR) water chemistry.
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The licensee had maintained an overall high level of operability for-
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radiation monitors in 1996 and was effectively tracking monitor performance.
One Unresolved Item was identified to determine the L
applicability of monitoring requirements of Criterion 64 of 10 CFR 50
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Appendix A and reporting requirements of 40 CFR 190 and 10 CFR 50.36a regarding potential unmonitored release pathways. (Section R2.1)
The meteorological instrumentation had been adequately maintained and
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the meteorological monitoring program had been effectively implemented.
(Section R2.2)
The licensee was effectively conducting formal Radiation Protection and
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Chemistry audits as required by TS and completing corrective actions in a timely manner. (Section R7.1)
The emergency preparedness program was being maintained in a high state
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of readiness.
Programs strengths-included the highly motivated staff and their effectiveness in providing an Emergency Response Organization
(ERO) that was frequently exercised by challenging training. (Section i
Pl.1)
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Reoort Details Sunmary of Plant Status Unit 1 operated at or near 100% power during the inspection report period.
l Unit 2 operated at or near 100% power until February 28, when power was reduced to 48% for the inspection and re) air of the 2B main feedwater pump turbine condenser. which had a small leac. A failure of the condenser
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circulating water system outlet valve caused a delay in the unit's return to l
100% power. The unit returned to full power on March 7.
The unit operated at
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or near 100% power until March 21, when load reduction commenced for a l
refueling outage.
The unit was taken offline on March 21 and remained in the
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refueling outage for the remainder of the inspection report period.
l Review of UFSAR Commitments
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While performing inspections discussed in this report, the inspectors reviewed the applicable portions of the UFSAR that were related to the areas inspected.
The inspectors verified that the UFSAR wording was consistent with the observed plant practices. procedures, and/or parameters.
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I. Operations
Conduct of Operations 01.1 Loss of Diaital Rod Position Indication a.
Insoection Scooe (71707)
On March 4. Unit 1 Digital Rod Position Indication (DRPI) for two control rods was lost during troubleshooting to investigate and repair the cause of a control rod position indication discrepancy between the DRPI panel and the Operator Aid Computer (OAC).
The inspector reviewed TS 3.1.3.2. the UFSAR. and design basis documentation, visually inspected both the DRPI panel and the OAC indication in the control room, and discussed the loss of indication and actions taken with plant personnel.
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Observations and Findinas While performing the control board surveillance at shift turnover on March 4 the licensee discovered a control rod position indication discrepancy between the OAC and the DRPI panel in the control room for
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Control Rod M-4. located in Control Bank D.
The surveillance procedure required that DRPI indication be within one step of the control rod i
position indicated on the OAC.
Engineering personnel were contacted to determine the actual indicated control rod position.
The raw data, referred to as gray code, for Control Rod M-4 corresponded to an indicated control rod position of 210 steps, indicating that the OAC indication was accurate.
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' Work Request'97010094 was generated to investigate the cause of the inaccuracy in the DRPI anel.
During troubleshooting, manipulation of the Control Rod M-4 dis lay card resulted in a complete loss of DRPI
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panel light emitting di de (LED) indication for that control rod.
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supplies were verified to be functioning properly.
Because degraded l
display cards associated with individual control rod indications can l
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indications, the troubleshooting method. employed was to unseat one
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display card at a time until all control rod position indications were l
restored.
During that process. DRPI panel indication was also lost for control rod D-8. located in Shutdown Bank E.
The troubleshooting was terminated, and the Failure Investigation Process (FIP) was initiated.
The licensee consulted TS 3.1.3.2 to determine if the loss of indication
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l on the DRPI panel placed the unit in a TS Action Statement. Technical Specification 3.1.3.2 states that the Digital Rod Position Indication System shall be operable and capable of determining the control rod positions within +/- 12 steps.
Engineering personnel indicated that the OAC program that provides control' rod position indication receives grey code information for all 53 control rods from the DRPI Display Computer input / output card.
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program then processes that information independently of the DRPI l
display to indicate the position of each control rod as detected by the electronics located in containment.
The Digital Rod Position Indication
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System provided the recuired information via the OAC. which served as a
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l display location reduncant to the DRPI panel.
The licensee concluded I
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l The inspector also consulted the TS and other design basis documentation.
The use of the DRPI panel was not specified in the TS.
TS Basis or any other reference reviewed by the inspector. As a result, the inspector concluded that the unit was being operated in compliance with the TS.
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On March 5. the FIP team continued to troubleshoot the loss of control rod indication for Control Rod M-4 and Shutdown Rod D-8.
The FIP team systematically tested individual control rods: when the display card for Control Rod H-14 in Control Bank C was removed, all control rod position indications were corrected.
The FIP team attributed the loss of indications to a faulty display card for Control Rod H-14. The display card was removed and replaced with a new card under Work Order 97019770-01, and all indications returned to normal. On March 6. a functional test was performed and DRPI was determined to be fully functional.
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Conclusions The inspector concluded that the licensee correctly interpreted the TS.
ensured that the unit was operated in compliance with the TS. and promptly took action to identify the root cause.
The inspectcr j
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considered the decision to terminate troubleshooting after a second DRPI panel control rod position indication was lost, and initiate a FIP to
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continue the troubleshooting ap3ropriately conservative.
Efforts to restore all indications on the DRPI panel were timely.
Operational Status of Facilities and Equipment 02.1 Readiness to Manually Ooerate SG PORVs a.
Insoection Scooe (71707)
The licensee performed walk throughs of procedures for manual operation of the SG PORVs for both units after a design basis discrepancy was identified. The design basis discrepancy increased the importance of timely local manual operation of the valves.
Resolution of the design discrepancy involved an unreviewed safety question and as such necessitated NRC review.
The issue is described further in Section E1.1 of this report. The inspector performed walk downs of SG PORVs to evaluate accessibility and reviewed the licensee's readiness for local manual operation of toe valves.
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Observations and Findinos The licensee performed a walk through of the procedures to operate the SG PORVs and determined that local operation to open the valves could be i
accomplished in 7-8 minutes. The licensee walked down the access routes to each SG PORV and initiated actions to improve access by adding
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emergency lighting and prestaging ladders, toolkits and procedures.
The ins)ector walked down several of the SG PORV areas and verified
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accessi)ility.
The inspector identified several discrepancies including one SG PORV area for which a ladder had not been prestaged.
The licensee had also identified these discrepancies, and actions were taken to correct them.
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Conclusions The licensee took appropriate actions to ensure the SG PORVs could be manually o)erated locally within the time assumed by a revised accident analysis tlat had been submitted for NRC review.
Quality Assurance in Operations (40500)
07.1 Observation of Nuclear Safety Review Board Meetina On March 6 the inspector attended the portion of a Nuclear Safety Review Board Meeting that included a discussion of recent performance at Catawba.
The discussion covered plant events. NRC violations and trends status of Maintenance Rule implementation. status of Improved Standard Technical Specifications implementation. results of audits and in-plant reviews, and results of board member plant tours and ENCLOSURE
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observations.
The board members noted some areas for improvement in the conduct of critiques of Licensed Operator simulator training sessions.
Overall, the board members noted significant improvement in recent
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Jerformance at Catawba.
The inspector noted that the Nuclear Safety Review Board's observations were consistent with recent inspection-
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l 07.2 1996 INP0 Evaluation Review i-On February 28, 1997, the final report for the November 1996 Institute
of Nuclear Power Operations Evaluation was issued.
The inspector i
reviewed the report and determined that the results of the evaluation
were generally consistent with the results of recent similar NRC j
evaluations.
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Miscellaneous Operations Issues (92700)
08.1 (Closed) VIO 50-413.414/95-12-01: Human Performance Errors That
Challenged / Degraded Safety Systems.
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I This violation cited four examples whereby plant personnel failed to
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follow procedures, which resulted in challenges to plant safety systems.
i These violations, as well as other examples of personnel errors that occurred in close succession, revealed a continuing adverse performance l
trend in 1995 and a broader issue with human performance (refer to NRC i
inspection report 50-413.414/95-12).
Each of the violation examples was
documented in a Licensee Event Report (LER). The inspector reviewed the l
licensee's violation response, dated July 7. 1995, which addressed i
additional actions and initiatives to correct the broader scope human
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performance issue.
The inspector reviewed specific corrective actions for each example as part of the review of each respective LER (refer to i
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sections 08.'2. 08.3. 08.4. and M8.1).
l The licensee and NRC management held meetings on May 17, 1995, and June l
23, 1995, to discuss the licensee's initiatives to reduce human errors.
The licensee consolidated the initiatives into a long-term human error reduction plan.
Part of this plan included "all hands" meetings held periodically with the entire plant staff.
The inspector attended several of these meetings and verified that plant management reinforced expectations performance standards, and goals.
In addition, the
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licensee has implemented a program of flawless human performance i
standards that are reinforced frequently during working meetings at all levels of the plant organization. Since the issuance of this violation
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and the licensee's subsequent implementation of corrective actions.
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ersonnel errors have declined in frequency and safety significance.
j his item is closed.
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08.2 (Closed) LER 50-413/95-01: Technical Specification 3.0.3 Entry Due To Both Trains Of Control Room Ventilation Being Inoperable.
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The event occurred as a result of improper sequencing of a tagout.
restoration of the nuclear service water system, which briefly disabled the B train control room ventilation system while the A train system was
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out of service for maintenance activities.
The control room operator
recognized and corrected the error and restored the B train equipment to
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an operable status within six minutes. The licensee determined that the cause of the event was a result of a failure of the Unit 1 senior reactor operator (SRO) to verify that the tagout restoration was
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adequate prior to restoring the nuclear service water system.
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licensee's corrective actions included counseling the individual and i
holding time out sessions with all o)erations staff and work groups to explain management expectations for luman performance improvement.
The inspector attended several of the time out sessions and verified that plant management reviewed several similar events whereby operator actions were unacceptable, explained to the staff why these actions were unacceptable, and presented expectations for improving performance. The inspector reviewed revisions to Operations Management Procedure 2-18.
Tagout Removal and Restoration Procedure, and verified that the licensee completed actions to require that tagouts receive a second independent verification for accuracy by an onshift SRO before they are implemented.
This LER is closed.
08.3 (Closed) LER 50-414/95-03: Technical Specification Violation Involving Containment Isolation Valves.
.The TS violation involved a failure to comply with TS action requirements associated with containment isolation valves.
Specifically, four containment isolation valves were tagged closed to support maintenance activities but were not de-energized as required by TS to ensure the containment isolation function. This condition existed
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for several shifts before the licensee identified it.
These valves also
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were logged with an incorrect TS reference during this time in the
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licensee's TS item tracking system.
These errors were attributed to
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poor work practices.
Corrective actions included conducting time out
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sessions with the operations staff and revising Operations Management Procedure 2-18. Tagout Removal and Restoration, to require a second independent SRO verification as discussed in section 08.2.
The inspector verified that the licensee's TS item tracking process also was revised to include requirements for a second independent SR0 verification for initiating and closing all TS action items.
This LER is closed.
08.4 (Closed) LER 50-414/95-05: Manual Reactor Trip Due To Loss of Main Feedwater.
This event occurred as a result of a non-licensed operator misaligning inlet, outlet and the bypass valves associated with several low pressure ENCLOSURE I
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I feedwater heaters while attempting to place the heaters in service.
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misalignment resulted in isolating main feedwater flow. The cause of
the error was attributed to the operator's failure to implement proper
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self-checking prior to performing the evolution.
The licensee disciplined the operator, conducted time out sessions with the operations staff as discussed in section 08.2. and revised unit i
operating procedures and annunciator resBonse procedures for operation
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j of the low pressure feedwater heaters.
ihe inspector verified that
0)erating Procedure OP/1(2)/A/6100/03. Controlling Procedure for Unit
Slutdown, and the appropriate annunciator panel response procedures were revised to (1) incorporate steps to bypass the low pressure feedwater heaters at lower power levels to prevent heater level fluctuations, and (2) provide better guidance for realigning the heaters following an isolation. This LER is closed.
08.5 (Closed) Violation 50-413/95-10-01: Failure To Implement Reactivity Limits During Zero Power Physics Testing.
An excessive positive reactivity insertion during Zero Power Physics Testing (ZPPT) resulted from two identified root causes: (1) inadequate planning, preparation and job briefing for the evolution, and (2)
inadequate work practices by the control room operator.
In addition, management actions following the event did not demonstrate sufficiently conservative decision making. Multiple corrective actions were defined in the licensee's response to the violation, including programmatic and procedural improvements.
The inspector reviewed station Procedures PT/0/A/4150/01. Controlling
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Procedures for Startup Physics Testing PT/0/A/4150/11A. Control Rod Worth Measurement by Boration/ Dilution, and PT/0/A/4150/118. Control Rod Worth Measurement by Rod Swap, to verify that enhancements had been
made. The ins)ector reviewed Operations Management Procedure 1-8.
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Authority and Responsibility of Licensed Reactor Operators (R0s). and Licensed Senior Reactor Operators (SR0s). to verify that the oversight function of ZPPT by the Control Room SRO and management ex)ectations j
therein, were clearly defined.
Sections 7.2 and 8.2 were clanged on June 20. 1995, to more clearly define the responsibilities of SR0s and R0s and to emphasize the importance of reactivity management.
Operator proficiency training was provided to each crew during requalification training in October 1995.
The proficiency training consisted of in-class review and discussion of the event, which was documented in PIP 0-C95-0468, and a simulator session to denionstrate the
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occurrence.
To verify that a Quality Improvement Team (OIT) was initiated to provide recommendations for improving the ZPPT program, the inspector obtained a
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copy of PIP 0-C95-1227. in which the OIT's recommendations were i
documented. The recommendations were characterized as corrective i
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-actions and assigned to various organizations for resolution. The i
inspector. reviewed the PIP and determined that all corrective actions p
had been closed.
Specifically. (1) test equipment setup was evaluated.
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startup ZPPT: (2). alarms on startup rate have oeen established on the i
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Operator Aid Computer: (3) the nuclear engineering organization has
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received instruction on expected and unexpected plant response during
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responsibilities of the dedicated R0, SRO and test coordination
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3ersonnel during ZPPT have been pre)ared. reviewed and accepted by the i-31 ant Operations Review Committee (30RC) for ZPPT as an infrequently
performed control room activity.
In' addition, the power excursion event was reviewed by PORC on June 29. 1995. which served as a forum for site
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conservative decision making.
The inspector concluded the corrective l
actions appropriate.
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In previous inspections. improvements were noted in the conduct and i
control of ZPPT (see NRC Inspection Reports 413.414/95-24 and 413.414/96-16), although the inspector also noted that communications
techniques among the test coordinator, SRO. and R0 were not consistent j
across shifts. The inspector concluded that the licensee's corrective i
actions have been implemented and effective in improving the conduct and j
control of ZPPT.
This violation is closed.
08.6 (Ocen) LER 50-414/95-01:
Reactor Trip due to Closure of a Main Steam
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Isolation Valve.
l The event described in this LER involved an automatic reactor trip i
because of the failure of an optical isolator in the B main steani
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isolation valve control circuit that caused the valve to close. The inspector reviewed the status of the licensee's corrective actions in
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PIP 2-C95-0246. Although the licensee had completed actions to test and replace optical 1.solators throughout the facility, and the reliability of the optical isolators that have control functions was significantly improved as a result. planned corrective action #3 in the LER had not been completed. This planned corrective action involved the development of a preventive maintenance program to periodically monitor certain energized optical isolators in critical applications. According to the I
PIP. the planned completion date for the development of the program is June 1. 1997.
Since the failure of the optical isolator was the cause
of the reactor trip. the inspector considered the development and implementation of the preventive maintenance program. absent a design change to remove the single failure vulnerability, important to prevent recurrence of the event.
This LER remains open pending completion of planned corrective action #3 of the LER.
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i II. Maintenance i
M1 Conduct of Maintenance i.
l M1.1-Taaout Error Results In Crackina Nonsafety Service Water Pioina a.
Insoection Scone (62707.71707)
On February 10 the licensee made an error while expanding a tagout boundary, which resulted in cracking a portion of nonsafety-related
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buried service wat.er piaing near the auxiliary building.
The inspector l
discussed the event wit 1 operations management and reviewed licensee's
corrective actions and the associated PIP (0-C97-0369).
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Observations and Findinas
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The licensee expanded a tagout boundary to reduce low pressure service water system isolation valve leakby.
Valve leakby was preventing the licensee from draining a portion of the low pressure service water system to support the installation of a plant modification. The expanded tagout did not consider other system inputs (Unit 1 and Unit 2 containment chiller discharge) into the section of piping that was isolated. As a result, the system was pressurized and a portion of buried low pressure service water system piping near the auxiliary building was cracked. The licensee subsequently determined, based on the pioing design pressure being higher than the shutoff head of the low 3ressure service water pumps, that a preexisting defect or corroded area lad failed and resulted in the crack.
The licensee determined that the cause of the tagout error was the result of an inadequate review prior to expanding the original low pressure service water system tagout boundary.
The personnel involved did not review all necessary plant drawings and flow diagrams and also failed to recognize that all system inputs into the piping (containment chilled water system discharges) were isolated.
In addition to counseling the involved personnel, the licensee revised the portion of the tagout procedure that provides instructions for expanding tagout boundaries. The licensee revised the procedure to require a second onshift SRO to perform an independent verification when a tagout boundary is expanded.
This is the same level of review required for initial tagouts.
The most significant plant systems that utilize the failed )ortion of the low pressure service water system are the containment cailled water system discharges for both units, the Unit 2A and 2B diesel generator cooling water discharge to Lake Wylie, and the auxiliary building liquid radioactive waste discharge to Lake Wylie.
Following identification of the cracked piping, the licensee implemen'.'ed several actions and temporary modifications to isolate the failed area. The actions included redirecting the containment chilled water cooling discharges to ENCLOSURE
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the storm drain system, isolating the auxiliary building liquid radioactive waste discharge, and realigning the Unit 2 diesel generator
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cooling water discharge to the standby nuclear service water pond.
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These actions resulted in work arounds and impeded the licensee's j
ability to make liquid radioactive waste releases.
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The inspector reviewed the potential significence of isolating the 1.
containment chilled water systems and determined that operating j
procedures would have directed operators to restore cooling to important
equipment located inside containment before an im)act to plant operation (e.g. reactor coolant pump manual tri) due to higa motor stator
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j temperatures) would have occurred. T1e inspector verified that control i
room operators were familiar with symptoms of a loss of the containment
j chilled water system and the appropriate restoration procedures.
l The portion of piping that failed was not safety-related but repair i
activities involved excavating in the vicinity of buried safety-related
piping.
The licensee developed compensatory actions to )rovide for i
missile protection of affected safety-related piping. T1e inspector
observed part of the excavation activities conducted to support repair i
of the failed service water piping and the inspector verified that j
maintenance supervision directing work activities knew the requirements i
and time constraints of the compensatory action. The inspector verified
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equipment needed to perform the compensatory action was at the work j
site.
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Conclusions
The licensee's corrective actions for a tagout error that caused l
cracking of a nonsafety-related service water pipe were effective.
j Appropriate revisions.to the tagout 3rocess were made to require the
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j same level of review for expanding tie scope of a tagout as the level of
review required for initial tagouts. The safety significance of the
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i error was low; however, it resulted in work arounds and impeded the t
j licensee's ability to make liquid radioactive waste releases.
M1.2 Control Area Ventilation System Online Maintenance Time Extension j
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Insoection Scooe (62707.40500)
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On February 19 the inspector observed a Plant Operations Review l
Committee (PORC) meeting that was convened to review an extension of the
out of service time for the Unit 1 Control Area Ventilation System i
beyond the planned duration because leaking isolation valves that
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delayed performance of online maintenance activities.
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Observations and Findinos i
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Train A control area chiller preventive maintenance, chiller condenser j
tube cleaning, and replacement of a section of corroded service water
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.10 piping were beirc - formed.
Leaking valves in the service water system
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prevented the 7:
re-from draining the chiller and service water
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piping. As a re of the leakage the licensee anticipated exceeding l
50 )ercent of the J allowed 7 day out of service time.
The licensee's
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worc management controls require a PORC review prior to exceeding 50
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percent of the TS Limiting Condition for Operation time.
The PORC considered options to (1) discontinue maintenance and return
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the system to service, or (2) use a freeze seal and a blank flange to
complete maintenance. The discussions were detailed and focused on i.
safety. The PORC found alternate freeze seal and flanging methods
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acceptable and approved extending the scheduled maintenance time beyond
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50 percent of the LC0 time. The inspector witnessed final installation I
of the freeze seal and isolation of the service water system and did not identify any discrepancies. Work activities were appropriately controlled.
,
c.
Conclusions i
Plant Operations Review Committee review of extending scheduled
maintenance for preplanned Unit 1 control area _ chiller work was detailed i
and focused on safety.
l M2-Maintenance and Material Condition of Facilities and Equipment
M2.1 Main Feedu ter Isolation Valve Solenoid Failure at McGuire
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a.
Jnsoection Scooe (62707)
t During the Unit 1 Steam Generator Replacement Outage (1E0C11) at the
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,
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McGuire Nuclear Station (MNS). a malfunction of the hydraulic actuator
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i for Main Feedwater Isolation Valves (MFIVs) 1CF-28 and ICF-30 (see NRC i
Inspection Report 50-369.370/97-04).
Engineering staff at the Catawba
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Nuclear Station (CNS) evaluated the failure of the solenoids to l
determine if CNS's MFIV solenoids were susceptible to the same failure.
l The inspector obtained information on the MNS finding, discussed the 1ssue with the Catawba engineering staff, reviewed component drawings.
.
and inspected an MFIV to determine if the solenoid valves at CNS were similar in design and mechanical function to those at MNS.
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b.
Observations and Findinas During MFIV testing at the MNS in February 1997. safety solenoids
,
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malfunctioned and did not allow the MFIV's accumulator to dump pressure
!
to the hydraulic actuator to close the MFIV.
The licensee discovered that a number of variables caused the safety solenoids to fail to
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perform their function.
,
l A vendor representative was consulted during the root cause i
investigation and concluded that the hydraulic fluid in the solenoid ENCLOSURE
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valves had degraded as a result of high temperatures, high humidity. and stagnation of the fluid. The degraded fluid thickened in areas of the
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valves with tight clearances, inducing frictional forces that
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essentially bound the. plunging mechanism in the solenoids and prevented j-them from actuating;
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r To determine if the MFIV solenoid valves at CNS were susceptible to the
.
same degradation, the ins)ector reviewed valve diagrams and discussed the differences between t1e valves at the respective sites.
Valve
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I diagrams revealed that designs were dissimilar.
While the hydraulic fluid circulates around the coil plunger of the MNS solenoid valves, it
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is isolated from that region for the CNS solenoid valves. Hence, the
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hydraulic fluid is not exposed to the heat generated at the coils.
In
!
j addition, the licensee obtained temperatures at the solenoid body and i
mounting manifold of one solenoid valve and found that they were within i
the vendor-recommended range of 110 F to 130 F.
The inspector concluded
.
that the hydraulic fluid was not exposed to high temperatures that would
cause heat-related degradation.
.
The licensee indicated that desiccant dryers had been installed in the mid-1980s to remove moisture from the air space in the accumulators.
The licensee also had implemented a periodic maintenance program to inspect the dryers and replace the desiccant on a two-year interval.
The inspector verified that model (predefined, repetative) Work Orders 91010268 and 91010269 were provided to implement Procedure IP/0/A/3010/009E. Main Feedwater (CF) System Miscellaneous Maintenance of Feedwater Isolation Valve Actuators. for the inspection and replacement of the desiccant. The inspector also verified that the work
order has been generated every two years since 1991.
The inspector concluded that the presence of desiccant dryers. if functioning correctly, would prevent a high moisture environment that would cause humidity-related degradation of the hydraulic fluid.
The valve diagrams also revealed that the hydraulic fluid in the MNS solenoids collected in areas with tight clearances (around the coil plunger area) where the fluid was not effectively circulated during valve actuation. The stagnated fluid congealed and induced frictional forces that the coil plunger could not overcome when it actuated. At CNS the hydraulic fluid is isolated from the coil plunger area and
.
circulates easily in open spaces in the valve body so that stagnation
)
does not occur.
The inspector concluded that the hydraulic fluid in the solenoid valves at CNS was not susceptible to the stagnation.
As a result, the hydraulic fluid was not likely to congeal and adversely
'
affect the solenoid's ability to actuate.
c.
Conclusions i
The inspector concluded that the MFIV solenoid valves at CNS are of
sufficiently different design from those at MNS to preclude failures i
similar to those experienced at MNS.
In addition, the environmental
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ENCLOSURE
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conditions (temperature and moisture) are not likely to result in heat-and moisture-related degradation of the hydraulic fluid.
M8 Miscellaneous Maintenance Issues (92902)
M8.1 (Closed) LER 50-414/95-004: Reactor Trip Due To Component Failure And Inadvertent Feedwater Isolation.
,
l This event was caused by a failed reactor trip breaker cell switch or its associated electrical relay which generated a saurious reactor trip signal during breaker testing. The signal caused tie steam generator level control system to run both main feedwater pumps to minimum speed,
'
which resulted in a 10-10 steam generator level reactor trip.
The inspector verified that the licensee replaced the suspect components and sent them to the breaker vendor for analysis. The results of the analysis were inconclusive, and the failure could not be repeated under
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laboratory conditions (PIP 2-C95-0637).
Following the reactor trip a feedwater isolation engineered safety j
features actuation occurred as a result of the impro)er manipulation of plant equipment during reactor trip breaker troubleslooting activities.
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The failure investigation process team leader manipulated a reactor trip breaker cell switch outside of the troubleshooting plan, without authorization from IAE personnel, and the individual was not qualified to the troubleshooting 3rocedure in use. The licensee counseled the individual involved.
T1e inspector concluded the licensee's corrective actions were appropriate.
This LER is closed.
M8.2 (00en) LER 50-413/95-03. Rev. 1: Failure to Perform TS Surveillances Due to Unanticipated Interaction of Systems This item continues to remain open pending the completion of further NRC review of the licensee's position on containment integrity positions.
III. Enaineerina El Conduct of Engineering E1.1 Steam Generator Tube Ruoture Accident Analvsis Discreoancy a.
Insoection Scooe (37551)
On February 10. 1997. during a dose analysis review of Chapter 15 of the Catawba Updated Final Safety Analysis Report. the licensee concluded that the potential for a more limiting single failure with regard to the steam generator tuba rupture accident existed.
The inspector reviewed the circumstances surrounding the identification of this design discrepancy and licensee's actions in response to the issue.
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ENCLOSURE
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b.
Observations and Findinas The potential for a more limiting single failure with regard to the steam generator tube rupture accident analysis involved a single failure of vital power that would affect two of the four Steam Generator Power Operated Relief Valves (SG PORVs). This more limiting single failure affected the ability of opercars to cool the plant down in a timely manner using the SG PORVs from the control room.
Upon recognition of this issue, the licensee reported the condition in accordance with 10CFR 50.72.
Subsequent evaluation by the licensee revealed that an unreviewed safety question existed: TS 3.7.1.6 is non-conservative in that it requires only three of the four SG PORVs to be operable. With two SG PORVs inoperable from the postulated single failure and one SG PORV unavailable because it is on the SG with the ruptured tube, the fourth would be required operable to initiate the cooldown and mitigate the event.
Additionally, to meet the timeliness assumption of the cooldown local manual operation of at least one additional SG PORV would be necessary to maintain margin to overfill of the main steam lines.
Station PIP 0-C97-0233. LER 413/97-01, and a license amendment were initiated by the licensee.
Pending ap3roval of the license amendment.
the licensee considered both units to ]e in an o)erable but degraded condition. Administrative controls implemented Jy the licensee to maintain operability and increase safety margin on an interim basis were i
as follows:
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All four SG PORVs were required to be operable. Technical Specifications 3.7.1.6 action (a) was to be applied with one SG PORV inoperable, and action (b) was to be applied with two SG PORVs inoperable.
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Equilibrium reactor coolant specific activity was not to exceed 0.46 uCi/gm 1131 dose equivalent (TS 3.4.8).
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Transient reactor coolant specific activity was not to exceed 26uCi/gm 1131 dose equivalent (TS 3.4.8).
The inspector periodically verified that the administrative controls were implemented as stated.
c.
Conclusion The licensee appropriately determined that this issue involved an unreviewed safety question and initiated actions to resolve it.
Administrative controls that were implemented on an interim basis were appropriate to increase the existing safety margin.
Pending NRC approval of the license amendment to (1) require four SG PORVs operable per unit, and (2) allow for the use of manual operator action to mitigate a steam generator tube rupture accident, this issue is ENCLOSURE
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characterized as Unresolved Item 50-413.414/97-05-01: Non-Conservative SG PORV Technical Specification.
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IV Plant Support l
k1 Radiological Protection and Chemistry Controls R1.1 Transoortation of Radioactive Materials a.
Insoection Scooe (86750. TI 2515/133)
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i The inspector evaluated the licensee's transportation of radioactive
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j materials programs for implementing the revised Department of
Transportation (DOT) and Nuclear Regulatory Commission (NRC)
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transportation regulations for shipment of radioactive materials as
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required by 10 Code of Federal Regulations (CFR) 71.5 and 49 CFR Parts
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100 through 177.
l b.
Observations and Findinas
j The inspector reviewed procedures and determined that they adequately addressed the following: assuring that the receiver has a license to
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receive the material being shipped; assigning the form, quantity type.
and proper shipping name of the material to be shipped: classifying
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waste destined for burial: selecting the type of package required: '
t assuring that the radiation and contamination limits are met: and
preparing shipping papers.
l Licensee's records for the three shipments of radioactive material
performed since the last inspection of this area were reviewed and the j
inspector determined the shipping papers contained the required
information.
The inspector also determined the licensee had maintained
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records of shipments of licensed material for a period of three years
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after shipment as required by 10 CFR 71 91(a). In addition, the licensee possessed a current certificate of approval (NRC Form 311) for their l
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" Quality Assurance Program Description for Radioactive Material Shipping
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l Packages Licensed Under 10 CFR 71".
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c.
Conclusionji
i Based on the above reviews, the inspector determined that the licensee j
had effectively implemented a program for shipping radioactive materials j
required by NRC and DOT regulations.
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R1.2 Radioloaical Protection and Chemistry Control.ji i
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a.
Insoection Scooe (84750)
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The inspector reviewed implementation of selected elements of the licensee's radiation protection and chemistry program. The review
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ENCLOSURE
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included obser% tion of radiological protection activities for the
control of radioactive material as required by 10 CFR Parts 20.1801, e
1802, 1902. and 1904.
t b.
Observations and Findinas
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During tours of the Auxiliary Building and Radwaste Building facilities,
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the inspector reviewed survey data and performed selected independent i
i radiation and contamination surveys of radioactive material storage-areas. Observations and survey results determined the licensee was j
effectively controlling and storing radioactive material.
i The inspector also reviewed operational and administrative controls for
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controlling access of material to the Catawba Nuclear Station Landfill where some licensee material with low levels of radioactivity are
i allowed to be buried under a South Carolina State Industrial Waste Landfill Permit These low level materials allowed by permit for burial
l included plant pond sludge and secondary plant resins. The inspector determined the licensee was exercising operational and administrative
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controls to ensure plant byproduct material not authorized for landfill l
was not being transferred to the Catawba landfill for disposal.
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Conclusions
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Radiological facility conditions and housekeeping in radioactive waste storage areas were observed to be adequate. Observations and survey l
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result.s determined the licensee was effectively controlling radioactive
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material.
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j R1.3 Water Chemistry Controls
a.
Insoection Scooe (84710).
Tae inspector reviewed implementation of selected elements of the licensee's water chemistry control program for monitoring primary and
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j secondary water quality as described in the Technical Specification
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limits, the Station Chemistry Manual, and the Final Safety Analysis
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Report (FSAR). The review included examination of program guidance and i
implementing procedures and analytical results for selected chemistry i
parameters.
i b.
Observations and Findinas
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j The inspector reviewed selected analytical results recorded for Units 1, and 2 reactor coolant and secondary samples taken between November 1996 and February 1997.
The selected parameters reviewed for primary
chemistry included dissolved oxygen, chloride, and fluoride. The.
selected parameters reviewed for secondary chemistry included hydrazine, iron, and chloride.
Those primary parameters reviewed were maintained
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i well within the relevant TS limits for power operations.
Those j.
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secondary parameters reviewed were maintained according to station procedures.
j c.
Conclusions Based on the above reviews, it was concluded that the licensee's water chemistry control program for monitoring primary and secondary water quality had been implemented, for those parameters reviewed, in accordance with the TS requirements and the Station Chemistry Manual for PWR water chemistry.
R2 Status of Radiation Protection (RP) Facilities and Equipment R2.1 Process and Effluent Radiation Monitors a.
Insoection Scooe (84750)
The ins)ector reviewed and discussed licensee effluent release limits and pat 1 ways as described in the licensee's Offsite Dose Calculation Manual (0DCM) and in Chapter 16 of the Selected Licensee Commitments Manual (SLC).
b.
Observations and Findinas The inspector toured the facility to observe the physical operation of selected process and radiation monitors in use.
The most recent system status report available, which covered the period January through November 1996, indicated that the overall availability for the Radiation l
Monitoring System remained at greater than 90 percent operability. The inspector reviewed and discussed availability trending records for both safety related and non-safety related monitors with the radiation monitor system engineer and engineering management.
Radiation monitor l
availability had improved during 1996 and 1997.
However, monitor availability was tracking below the licensee's goal of 95 percent for monitors not required by TSs and below 98 percent for TS required monitors. The cognizant licensee engineer informed the inspector that the system focus was to continue the current improving trend in radiation monitor availability to achieve the established site goals.
The inspector determined that there were no effluent monitors and that procedures did not exist to monitor or obtain samples of secondary plant water for release quantification during some environmental discharge evolutions.
Examples of these evolutions included discharges from the
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steam generator atmospheric dump valves or the power operated relief valves. (PORVs) in situations other than an emergency procedure or when l
a known primary to secondary leak has been identified.
The licensee was performing weekly secondary plant chemistry samples.
However, the inspector was informed that there was not a procedure for notifying chemistry and cognizant personnel responsible for preparing the annual effluent release report during these unplanned /unmonitored releases.
ENCLOSURE i
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The inspector informed the licensee that a release through these pathways during periods for which radioactive material has not been accounted for may result in an unmonitored pathway to the environment.
The licensee was informed by the inspector that the applicability of monitoring requirements of Criterion 64 of 10 CFR 50 Appendix A, and the reporting requirements for effluents of 40 CFR 190 and 10 CFR 36a, regarding the doses from the unmonitored steam generator atmospheric dumps /PORVs, would be considered as an Unresolved Item (URI) pending further NRC review.
URI 50-413,414/97-05-02: Determine the Applicability of Monitoring Requirements of Criterion 64 of 10 CFR 50 Appendix A and Reporting Requirements of 40 CFR 190 and 10 CFR 50.36a
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Regarding Potential Unmonitored Release Pathways.
c.
Conclusions Based on the above reviews, the inspector determined the licensee had maintained an overall high level of operability fcr radiation monitors in 1996 and was effectively tracking monitor performance.
One URI was
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identified to determine the applicability of monitoring requirements of Criterion 64 of 10 CFR 50 Appendix A and reporting requirements of 40 CFR 190 and 10 CFR 50.36a regarding potential unmonitored release pathways.
R2.2 Meteoroloaical Monitorina Proaram (84750)
a.
Insoection Scooe (84750)
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The inspector reviewed the licensee's implementation of the operational and surveillance recuirements for the meteorological monitoring
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instrumentation as cescribed in Section 2.3.3 of the Final Safety Analysis Report (FSAR).
Equipment calibration and maintenance checks
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and semiannual calibration checks were required to be performed by
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prescribed statior procedures.
b.
Observations and Findinos The inspector reviewed the meteorological monitoring procedures that
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included provisions for performing the required surveillances on the i
meteorological monitoring instrumentation and the most recent records of calibration checks performed and semiannual calibrations of wind speed, wind direction and air temperature instrumentation.
The inspector determined that the calibration checks and calibrations were performed in accordance with the above procedures and at the required frequency.
The inspector toured the control room area and observed meteorological monitoring equipment and reviewed equipment operation with licensee personnel. At the time of the insaection, the meteorological instrumentation was operable and tlat data for wind speed, wind direction, air temperature, and precipitation were being collected as descrit.ed in the FSAR.
ENCLOSURE
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Conclusions 4.
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Based on the above reviews and observations it was concluded that the
meteorological instrumentation had been adequately maintained and that
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the meteorological monitoring program had been effectively implemented.
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R7.0 Quality Assurance in Radiation Protection and Chemistry R7.1 Licensee's Activities And Self Assessment Programs a.
Insoection Scone (84750)
<
The licensee's activities and self assessment programs were reviewed to determine the adequacy of identification and corrective action programs i
for deficiencies in the areas of RP and Chemistry.
b.
Observations and Findinas
Reviews by the inspector determined that Quality Assurance audits and Self Assessment efforts in the area of RP and Chemistry were accomplished by reviewing RP procedures, observing work, reviewing industry documentation, and performing plant walk downs to include surveillance of work areas by supervisors and technicians during normal work coverage.
Documentation of problems by licensee representatives was included in Quality Assurance Audits and Self Assessment Reports.
Corrective actions were included in the licensee's Problem Investigative Process and were being completed in a timely manner.
The inspector reviewed the licensee's corrective actions for an NRC Violation 50-413.414/96-13-05 which involved an individual entering the Radiological Control Area (RCA) of the facility without the proper training, thermoluminescent dosimeter, and an initial body burden count.
The inspector determined the licensee had updated training requirements, posted dosimetry sign requirements at RCA entrances, and updated visitor escort requirements to preclude reoccurrence of the violation.
The inspector informed the licensee that Violation 50-413.414/96-13-05 would be closed based on licensee corrective actions.
c.
Conclusions The inspector determined the licensee was effectively conducting formal RP and Chemistry audits as required by Technical Specifications and completing corrective actions in a timely manner.
i ENCLOSURE
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P1 Conduct of EP Activities Pl.1 Ooerational Status of the Emeraency Preoaredness Program f
.a.
Insoection Scone (82701)
The inspector reviewed day-to-day routine operations and program changes i
to assess the effectiveness of the licensee's implementation of their
Emergency Plan (EP) in meeting the regulatory requirements of emergency
preparedness. The following routine areas were reviewed:
t e
changes to the Emergency Plan and Implementing Procedures
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e maintenance of selected emergency facilities, equipment and
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supplies
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I e
changes to the emergency organization or management control systems e
review of the indeperident audit report conducted since the last inspection j
t e
effectiveness of licensee controls in the identification and
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resolution of issues identified in the area of emergency
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preparedne:s The inspector observed training conducted during the inspection week l
which focused on the Emergency Operations Facility (EOF) and Operational l
Support Center (OSC).
b.
Qbservations and Findinos The revision of the Emergency Plan currently in effect was Revision 97-1. effective January 8. 1997.
The changes addressed the minimum
staffing requirements for emergencies.
Only those changes that had been
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approved by the NRC staff in a September 17, 1996, Safety Evaluation had been implemented. The changes had been submitted to the NRC within 30 days after the changes were made.
The inspector reviewed random copies of the Emergency Plan and found them to contain the latest revisions.
The licensee had implemented its Emergency Plan twice since the last inspection, both times for a Notification of Unusual Event. The first was on February 15. 1995, for the transport of a contaminated injured i
worker. The second was the loss of offsite power event on Unit 2 from February 6-8, 1996. A review of properly implemented and required notifications indicated that they were made in a timely manner.
During the tour of the Emergency Res)onse Facilities which included the Technical Support Center (TSC) and tie OSC. the inspector observed that facilities were maintained in a state of operational readiness..
Random ENCLOSURE
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selection of instrumentation in wall-lockers found required equipment present, maintained, and properly calibrated.
Equipment upgrades made since the last inspection included new write-boards for-the TSC and OSC.
a new OAC for Unit 1 (new OAC for Unit 2 to be installed during outage beginning March 22, 1997), new Pentium computers (vice 386 computers)
for the E0F, and video conferencing between the TSC Radiation Protection (RP) and the EOF RP. Communication upgrades include two satellite tele hones for emergency communications, with one unit being maintained
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in t1e Control Room, and the other unit being a spare.
Irrproved overwatch to the dedicated TSC and OSC facilities is now present as a revised security plan had dedicated officers stationed in each of these facilities.
Changes to the organization and management control have been minimal but have provided for improved EP program.
In January 1996 the EP trainer was reassigned to the Training Department from the EP group.
The individual still provided full support to the EP training, but has the resources of the Training Department to assist in the training. This has resulted in significantly improved visual aids being used to support the EP training as well as improved training techniques.
The training program continued to be a strength of overall EP.
Besides the above improvements, the major contribution to program performance is the excellent training exercises.
The scenarios provided for an aggressive drill program. For example, during 1996 there were six integrated, simulator driven ERO drills conducted in conjunction with a 100 day S/G outage.
For 1997. their are four integrated drills that are
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of adequate scope that the offsite governmental agencies have been given
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the opportunity to participate in any or all of the four for training if they desire.
The inspector reviewed the independent audits required by 10 CFR f
50.54(t) since the last inspection. The audits identified four
strengths and one finding. The audits were thorough and provided for an independent EP technical expert from other utilities to augment the j
licensee's audit team.
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The inspector concluded the licensee was proactive in resolving an issue i
that could impact the effectiveness of the EP program.
This was based on the successful augmentation drill that was conducted during heavy traffic conditions with the new football stadium (Ericsson Stadium) to determine impact on the ability to staff the EOF. Another positive observation was the EP group had initiated a self assessment program which contributed to their effectiveness.
ENCLOSURE
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Conclusions
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The inspector concluded the emergency preparedness program was being l
maintained in a high_ state of readiness.
Programs strengths included
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the highly motivated staff and their effectiveness in providing an ERO
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that was frequently exercised by challenging training.
t P8 Miscellaneous Emergency Preparedness Items
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~P8.1 (Closed) IFI 50-413. 50-414 (96-04-01) Exercise Weakness - Incorrect and i
Untimely News Releases An Exercise Weakness was identified for failure to demonstrate the-I ability to provide accurate information to the news media in a timely manner.
The inspector verified the corrective action described in the
licensee's response letter, dated April 26. 1996, and modified by l
letter, dated August 14, 1996, to be reasonable and complete.
Additionally, the Duke Power Company Media Center and staff that support
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Catawba Nuclear Site also support the other Duke Power Company nuclear sites.
The Media Center and staff provided timely and accurate
!
information to the news media in support of the September 17. 1996.
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Oconee Nuclear Site Annual Emergency Response Exercise.
This exercise l
weakness is therefore closed, i
V. Manaaement Meetinas
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X1 Exit Meeting Summary i
i The inspector presented the inspection results to members of licensee i
management at the conclusion of the inspection on March 26, 1997. The
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licensee acknowledged the findings presented.
No proprietary information was j
identified.
ENCLOSURE
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PARTIAL LIST OF PERSONS CONTACTED Licensee Bhatnager. A.. Operations Su)erintendent Birch. M., Safety Assurance Manager Christopher. S., Emergency Planning Manager Coy. S.. Radiation Protection Manager Forbes. J., Engineering Manager Harrall. T.,
IAE Maintenance Superintendent Kelly. C.
Maintenance Manager Kimball. D., Safety Review Group Manager Kitlan. M.
Regulatory Compliance Manager Kuhr. E., Nuclear Emergency Planning Consultant McCollum W., Catawba Site Vice-President Mitchell. G.. Emergency Planning Senior Technical Specialist Nicholson. K., Regulatory Compliar.ce Technical Specialist Peterson. G.. Station Manager Prepst. R.
Chemistry Manager Rogers. D., Mechanical Maintenance Manager Smith. B., Emergency Planner Tower. D., Compliance Engineer l
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ENCLOSURE
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D
INSPECTION PROCEDURES USED IP 37551:
Onsite Engineering IP 40500:
Self-Assessment IP 61716:
Surveillance Observation IP 62707:
Maintenance Observation IP 71707:
Plant Operations IP 71750:
Plant Support Activities IP 82701:
Operational Status of the Emergency Preparedness Program IP 83750:
Occupational Radiation Exposure
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IP 84750:
Radioactive Waste Treatment and Effluent and Environmerital Monitoring IP 86750:
Solid RadWaste Management and Transportation of Radioactive Materials IP 92700:
Onsite Followup of Written Reports of Nonroutine Events IP 92901:
Followup - Operations i
IP 92902:
Followup - Maintenance IP 92903:
Followup - Engineering IP 93702:
Onsite Response to Events i
TI 2515/
l 133:
Implementation of Revised 49 CFR Parts 100-177 and 10 CFR Part 71 ITEMS OPENED. CLOSED AND DISCUSSED Ooened 50-413.414/97-05-01 URI Non-Conservative SG PORV Technical Specification. (Section El.1)
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50-413.414/97-05-02 URI Determine the Applicability of Monitoring Requirements of Criterion 64 of 10 CFR 50
Appendix A and Reporting Requirements of i
40 CFR 190 and 10 CFR 50.36a Regarding Potential Unmonitored Release Pathways (Section R2.1)
Closed 50-413.414/95-12-01 VIO Human Performance Errors That Challenged / Degraded Safety Systems (Section 08.1).
50-413/95-01 LER Technical Specification 3.0.3 Entry Due To Both Trains of Control Room Ventilation Being Inoperable (Section 08.2).
t INCLOSURE
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50-414/95-03 LER Technical Specification Violation
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Involving Containment Isolation Valves
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(Section 08.3).
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50-414/95-05 LER Manual Reactor Trip Due To Loss of Main Feedwatcr (Section 08.4),
50-413/95-10-01 VIO Failure To Implement Reactivity Limits
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During Zero Power Physics Testing (Section 08.5).
50-414/95-04 LER Reactor Trip Due To Component Failure And Inadvertent Feedwater Isolation (Section M1.2),
j 50-413.414/96-04-01 IFI Exercise Weakness - Incorrect and Untimely News Releases (Section P8.1).
50-413.414/96-13-05 VIO RCA Entry Without Dosimetry contrary to TS and Rad Protection Directive.
Person entered RCA without TLD, BBA or GET (Section R7.1).
i l
Discussed 50-413/95-03,Rev 1 LER Failure to Perform TS Surveillances Due to L
Unanticipated Interaction of Systems i'
(Section M8.2).
50-414/95-01 LER Reactor Trip Due to Closure of Main Steam l
l Isolation Valve (Section 08.6)
I LIST OF ACRONYMS USED
!
Body Burden Analysis BBA
-
!
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Main Feedwater l
CFR
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Code of Federal Regulations
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Catawba Nuclear Station DOT
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De)artment of Transportation DPC Duce Power Company
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DRPI
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Digital Rod Position Indication EOF
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Emergency Operations Facility EP Emergency Planning
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ERO Emergency Response Organization
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-
Failure Investigation Process FSAR -
Final Safety Analysis Report
!
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General Employee Training
'
IAE
-
Instrument and Electrical j
IFI
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Inspector Followup Item
'
ENCLOSURE
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,
.
s
,a
IR
-
Inspection Report
LER
-
Licensee Event Report
LED
-
Light Emitting Diode
LCO
-
Limiting Condition for Operation
MFIV -
Main Feedwater Isolation Valve
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McGuire Nuclear Station
-
Non Cited Violation
OAC
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Operator Aid Computer
ODCM -
Offsite Dose Calculation Manual
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Operational Support Center
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Problem Investigation Process
PORC -
Plant Operations Review Committee
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Pressurized Water Reactor
OlT
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Quality Improvement Team
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Radiation Control Area
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Radiation Protection
Steam Generator Power Operated Relief Valves
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Selected Licensee Commitments
SR0
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Senior Reactor Operator
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Thermoluminescent Dosimeter
TS
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Technical Specifications
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UFSAR -
Updated Final Safety Analysis Report
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Unresolved Item
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Violation
Work Order
-
ZPPT -
Zero Power Physics Testing
ENCLOSURE
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