IR 05000213/1987008
| ML20214K059 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 05/21/1987 |
| From: | Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20214K045 | List: |
| References | |
| TASK-05-10.B, TASK-08-02, TASK-1.C.1, TASK-5-10.B, TASK-8-2, TASK-RR, TASK-TM 50-213-87-08, 50-213-87-8, IEIN-87-015, IEIN-87-15, NUDOCS 8705280309 | |
| Download: ML20214K059 (12) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report N /87-08 Docket N License N DPR-61 Licensee: Connecticut Yankee Atomic Power Company P. O. Box 270 Hartford, CT 06101 Facility: Haddam Neck Plant, Haddam, Connecticut Inspection at: Haddam Neck Plant Inspection conducted: March 19 to May 5, 1987 Inspectors: Andra A. Asars, Resident Inspector Stephen M. Pindale, Resident Inspector, Beaver Valley 1 Paul D. Swetland, Senior Resident Inspector Approved by: $ S/zt/r7 E. C. McCabe, Chief, Reactor Projects 3B Date Summary: Inspection 50-213/87-08 (3/19 - 5/5/87)
Areas Inspected: This was a routine safety inspection (240 hours0.00278 days <br />0.0667 hours <br />3.968254e-4 weeks <br />9.132e-5 months <br />) by the resident inspectors. Areas reviewed included plant operations, radiation protection, fire protection, security, maintenance, surveillance testing, licensee events, open inspection items, IE Information Notices, Emergency Preparedness Drills, Systematic Evaluation Program findings, and temporary procedure change Results: Two unresolved items related to Nuclear Review Board audits (see detail 4.1) and low temperature overpressure protection (see detail 4.2) were close No violations were identified. Two new unresolved items were opened regarding the adequacy of licensee commitment control (see detail 4.2) and invalid reactor cool-ant temperature charts due to using chart paper with the wrong scales (see detail 4.2).
8705280309 PDR 8705 ] 3 ADOCK 0500 O
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TABLE OF CONTENTS P_ age Summary of Facility Activities....................................... 1 Review of Plant Operations........................................... 1 2.1 Plant Operations Review Committee................................ 2 Observation of Maintenance and Surveillance Testing.................. 2 Followup on Previous Inspection Findings............................. 2 4.1 Adequacy of Nuclear Review Board Audit Methods................... 2 4.2 Low Temperature Overpressure Protection System Inadequacies...... 3 Followup on Information Notice 87-15, Posting Requirements........... 4 Followup on Events Occurring During the Inspection................... 5 6.1 Licensee Event Reports........................................... 5 l 6.2 Excessive Reactor Coolant System Leakage......................... 5 6.3 Plant Trip Due to Turbine Control Valve Failure.................. 6 6.4 Spurious Turbine Load Runback.................................... 7 l Review of Periodic and Special Reports............................... 7 t Emergency Preparedness Dri11s........................................ 7 Followup on Systematic Evaluation Program Findings................... 8
9.1 Onsite Power Systems............................................. 8 9.2 Safe Shutdown Systems............................................ 9 1 Temporary Procedure Changes.......................................... 10
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1 Unresolved Items..................................................... 10 i
j 1 Exit Interview....................................................... 10 f
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DETAILS Summary of Facility Activities At the beginning of the inspection period, the plant was operating at full power. On March 29 a load reduction to 50% was conducted to investigate an increased reactor coolant system (RCS) leakage from the RCS loop isolation valve stem packing. An Unusual Event (UE) was declared because the leakage rose above the 10 gpm limit allowed by Technical Specifications (TS). The leakage stopped when the valves were back-seated, and the UE was terminate The plant returned to full pcwer. On March 30, power was reduced to 92% to correct feedwater heater level control valve deficiencies. Subsequent full power operation continued until April 16 when the reactor tripped due to high steam flow in the No. 3 main steam line. The high flow was created when the No. 4 turbine control valve failed open. Repairs were made to the turbine control system and after restart, the licensee began a load increase on April 18. However, power was limited to 16% due to turbine control system corro-sion/ contamination. After system flushing on April 20, power was increased to 90%. On April 21, the plant experienced a spurious turbine runDack of about 5% power due to a negative rate spike on the channel 33 power range nuclear instrument. Power operation was returned to 95% and remained there for the duration of the inspection period while turbine control oil system performance was being evaluate . Review of Plant Operations The inspectors observed plant operation during regular tours of the following plant areas:
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Control Room --
Security Building
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Primary Auxiliary Building --
Fence Line (Protected Area)
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Vital Switchgear Room --
Yard Areas
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Diesel Generator Rooms --
Turbine Building
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Control Point --
Intake Structure and Pump Building Control room instruments were observed for correlation between channels and for conformance with Technical Specification requirements. The inspector observed various alarm conditions which had been received and acknowledge Operator awareness and response to these conditions were reviewed. Control room and shift manning were compared to regulatory requirements. Posting and control of radiation and high radiation areas was inspecte Compliance with Radiation Work Permits and use of appropriate personnel monitoring devices were checked. Plant housekeeping controls were observed, including control and storage of flammable material and other potential safety hazards. The inspector also examined the condition of various fire protection system During plant tours, logs and records were reviewed to determine if entries were properly made and communicated equipment status / deficiencies. These records included operating logs, turnover sheets, tagout and jumper logs,
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=? " process computer printouts, and Plant Information Reports. The inspector observed selected aspects of plant security including access control, physical barriers, and personnel monitoring. No abnormal conditions were identifie .1 Plant Operations Review Committee (PORC)
The inspector attended Plant Operations Review Committee (PORC) meetings on April 9 and 23,1987. Technical Specification (TS) 6.5 member attend-ance requirements were met. The agendas included procedural approvals and changes, proposed TS changes and field changes to design change packages. The meetings were characterized by frank discussions and questioning of the proposed changes. In particular, consideration was given to assure clarity and consistency among procedures. The inspector had no further comment . Observation of Maintenance and Surveillance Testing The inspector observed various maintenance and problem investigation activi-ties for compliance with requirements and applicable codes and standards, QA/QC involvement, safety tags, equipment alignment and use of jumpers, per-sonnel qualifications, radiological controls, fire protection, retest, and reportabilit Also, the inspector witnessed selected surveillance tests to determine whether properly approved procedures were in use, test instrumenta-tion was properly calibrated and used, technical specifications were satisfied, testing was performed by qualified personnel, procedure details were adequate, and test results satisfied acceptance criteria or were properly dispositione The following activities were reviewed:
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PMP 9.5-23, Preventative Maintenance of Reactor Trip and Isolation Breakers
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PMP 9.5-40, Periodic Functional Test of Reactor Trip Breakers No inadequacies were identifie . Followup on Previous Inspection Findings
. Two NRC open items were reviewe Licensee actions were sufficient to close 1 these item Details follow:
4.1 (Closed) Unresolved Item (213/86-05-01): NRC evaluation of licensee methodology for audits of the QA program effectiveness by the Nuclear Review Board (NRB). The NRB conducts detailed reviews of the QA program with audit attributes which are similar to those used by NRC in the Systematic Assessment of Licensee Performance. The NRC licensing staff reviewed this methodology and found it to be acceptable for compliance with TS 6.5.2.6. Therefore, this item is close . - _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ - - _ _ _ - _ - _ _ _ _ _ _ - _ _ _ _ _ _ _
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4.2 (Closed) Unresolved Item (213/87-06-01): Reactor vessel integrity can be threatened by sudden pressure increases at low reactor coolant system (RCS) temperatures. This concern was addressed with the installation of a low temperature overpressure protection system (LTOP) relief system which is placed inservice whenever the RCS temperature is below 340 de-grees License requirements regarding the LTOP system were incorpor-ated into the plant operating license by Amendment 33. The analysis which demonstrates the adequacy of LTOP to mitigate such overpressure was based on reactor vessel pressure and temperature limits calculated for 14 Effective Full Power Years (EFPY) of reactor operation. In 1985, the licensee re-evaluated this analysis when the operating limits were extended to 22 EFPY. During a previous evaluation of LTOP (see NRC In-spection Report 50-213/87-06, detail 8), a concern was raised whether the current analysis allows adequate margin to the 22 EFPY limits with one train of LTOP operable and the potential flashing within the LTOP syste A review of the mass and energy addition transient analyses was performed by licensee engineering. The LTOP spring-loaded relief valve flow rates used in these analyses are based on a 100 psig backpressure. This cor-responds to the saturation pressure for the highest RCS temperature at which the LTOP system operates. Therefore, the assumed flow rates ade-quately model the system operation, including flashing considerations, because lower backpressures will produce higher flows even with flashin However, in order to assure that the LTOP system provides adequate reac-tor vessel protection based on the revised pressure / temperature curves for 22 EFPY of operation, unnecessary conservatisms were removed from the LTOP analysis, and more restrictive heat-up rates were assumed. The inspector concluded that, based on the licensee's revised analysis and more restrictive operating assumptions, the LTOP system provides adequate reactor vessel protection to 22 EFPY. This item is close During this inspection, the inspector also reviewed implementation of Technical Specification (TS) Amendment 70, dated October 1985, which updated the pressure / temperature curves to 22 EFPY. In particular, the inspector noted that TS Figures 3.4-8 and 3.4-9 were changed to reflect the more restrictive heat-up and cooldown rates specified in the new analyses (heat-up rate limited to 50 degrees F/ hour when the RCS tempera-ture is less than 200 degrees F and cooldown rate limited to 30 degrees F/ hour when RCS temperature is between 200 and 70 degrees F). However, TS 3.4.A.3. was not revised to reflect these changes. Currently, it states that the average rate of RCS temperature change during normal heatup or cooldown shall not exceed 100 degrees F/ hour. In addition, i Normal Operating Procedure (NOP) 2.1-1, Plant Startup - Cold Shutdown to Hot Standby, and N0P 2.3-4, Plant Shutdown - Hot Standby to Cold Shutdown had not been revised to include the new limits. Administrative limits of 50 degrees F/ hour had existed on heat-up and cooldown rate When the inspector brought this error to the attention of the licensee, temporary procedure changes were implemented to revise NOPs 2.1-1 and 2.3-4 to include the more restrictive heat-up and cooldown limit The
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inspector noted that other recent instances of inadequate implementation of Itcense commitments include fire protection requirements cited in NRC Inspection Report 50-213/86-17, a 3-loop operation prohibition documented in Licensee Event Report 86-047, radiological TS surveillances cited in NRC Inspection Report 50-213/86-15 and LTOP isolation valve commitments identified in NRC Inspection Report 50-213/87-06. These examples indi-cate inadequate communication and implementation of license commitments within the licensee's organization. This was discussed with station management on May 4, 1987. The licensee stated that this concern was recognized and that corrective actions were being evaluated. This item will remain unresolved pending licensee formalization and NRC review of these actions (UNR 213/87-08-01).
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The inspector reviewed plant operating records to verify whether the non conservative heat-up and cooldown rates specified in NOPs had re-sulted in exceeding TS requirements. Due to physical limitations of the plant, the heat-up rate limitations cannot be exceeded under normal operating conditions. However, the cooldown rate limit is low enough to be exceeded unless monitored closely. The inspector reviewed four separate RCS average temperature charts and the residual heat removal system (RHR) temperature charts for the cooldown for an unplanned eini-outage conducted on July 16, 1986. The inspector determined that the amended TS limits were met during this cooldown. However, the inspector noted that there was a discrepancy of about 80 degrees F between the RCS cold leg temperature indication and the residual heat removal system (RHR) temperature indications. Based on the reactor coolant system cold leg temperature (Tc) charts, a 35 degree per hour cooldown rate existed at abcut 6 a.m. on July 16, 1986. Tc was 190 F at that time. This con-dition appeared to violate the TS cooldown limit of 30 degrees per hou The RHR charts however, indicated that the RCS temperature at that time was about 270 F. Initially, the licensee could not account for this discrepancy. It later became evident that the chart paper for the RCS cold leg temperature did not have the correct scale for this recorde The recorder operates with a scale from 50 to 750 degrees F and the paper scale is from 0 to 600. Since the actual RCS temperature was above 200 F, no violation occurred. Nevertheless, this occurrence indicates that operators were not aware of the actual cooldown rate limit and were not careful to assure that TS compliance was achieved based on the most
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conservative main control board indication. Licensee determination and NRC review of why the Tc recorder scales did not match, how long this condition has existed and what corrective actions will be taken including correction of recorder charts transmitted to nuclear records will remain unresolved pending licensee review (UNR 213/87-08-02).
5. Followup on Information Notices (ins)
IN 87-15 Compliance With the Posting Requirements of Subsection 223b This notice informed licensees holding a construction permit of posting re-quirements in the Atomic Energy Act which have on occasion been omitted from site posting Specifically, Section 223b of the Act was identified as not
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being prominently posted. This section extends criminal penalty provisions to any director, officer, or employee of a firm licensed under Section 103 or 104b. The licensee closely evaluated the applicability of this notice because of ongoing construction activities for the new switchgear buildin A determination was made that, since the switchgear building construction is being handled as a design change in accordance with 10 CFR 50.59, the posting requirements highlighted by this notice are not applicable. This position was verified by discussions with the responsible NRC technical contac In relation to this notice, the inspector reviewed Administrative Control Procedure 1.0-14, Implementation of the Requirements of 10 CFR Part 21: Re-porting of Defects and Noncompliance. This procedure provides for posting as well as reporting in accordance with Part 21. The inspector verified that the required postings are prominently displayed. Currently, the licensee is evaluating the locations of these postings to determine if they should be updated or relocate In addition, Administrative Procedure 1.1-106, Employee Nuclear complaints and concerns, is being revised to include guidelines for posting required documents. The inspector had no further concern . Followup on Events Occurring During the Inspection 6.1 Licensee Event Reports (LERs)
The following LERs were reviewed for clarity, accuracy of the description of cause, and adequacy of corrective action. The inspector determined whether further information was required and whether there were generic implications. The inspector also verified that the reporting require-ments of 10 CFR 50.73 and Station Administrative and Operating Procedures had been met, that appropriate corrective action had been taken, and that the continued operation of the facility was conducted within Technical Specification Limit Technical Specification Fire Watch Improperly Secured Due to Personnel Error
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- 87-02 Turbine Load Runback Caused by Faulty Potentiometer in Nuclear Instrumentation System 87-03 Wide Range Stack Monitor Failed Due to Faulty Parts
- Event detailed in NRC Inspection Report 50-213/87-0 No unacceptable conditions were identified.
l 6.2 Excessive RCS Leakage from Loop Stop Valve Packing
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On March 29, 1987, leakage to the RCS leak-off collection system was i observed to be increasing over a period of about eight hours. Themajor '
contributor to the increased RCS leakage was identified as packing leak-age on the reactor coolant loop isolation valve A work procedure was
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prepared to manually backseat the valve A plant load reduction was initiated to minimize radiation exposure to the work party. During this period the RCS leakage rose to 13 gpm. The licensee declared an Unusual Event in accordance with the emergency plan at 4:58 a.m. on March 29; due to RCS leakage exceeding the TS limit of 10 gpm. Containment par-ticulate monitors indicated increasing radioactivity in the containment and containment air samples indicated a rise in containment activity of twice the normal values. The work party entered containment with
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respiratory protection and successfully backseated three of the loop isolation valves. One of these, the loop four hot leg isolation valve, had a visible plume of steam emanating fro:n the packing gland. This valve opened about three inches before backseating. Following the valve backseating, identified RCS leakage decreased to less than 2 gpm and containment radioactivity began t.' decline. Following completion of
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routine turbine valve te tinu, pawer ascension began and the plant i reached full power the same da In response to this incident, the licensee initiated shutdown work orders (CYWO 87-3433 through 3440) to repair the loop isolation valves, verify the operability of the operators, and position limit switches to assure that the valves stroke properly from the main control board. This will be reviewed during routine inspection of the upcoming refueling outage activitie The inspector had no further questions at this tim .3 Plant Trip Due to Turbine Control Valve Failure The plant automatically tripped from 100% power on April 16, 1987 due to high differential pressure (steam flow) in the No. 3 steam line. The high flow was created when the No. 4 turbine control valve failed ope The resulting high steam flow signal caused main steam isolation valve closure and a turbine trip / reactor trip. All plant systems responded normally and the licensee made the appropriate notifications to state, local and NRC officials. The resident inspector observed the trip re-covery and achievement of stable hot standby condition Earlier in the inspection period, the No. 4 turbine control valve had failed closed during load manipulations to stabilize feedwater regulating
valve oscillations. Preliminary evaluation of the cause of these prob-lems indicated corrosion contamination buildup in the control oil syste Critical components were cleaned and the oil system was flushed. After a period of low power operation to evaluate turbine control system stability, power was increased to 90% on April 20. Power was raised to 95% on April 21 and held for the duration of the inspection period to allow evaluation of control oil system performanc No unacceptable conditions were found in review of this event and the licensee's respons . _ _ - - -
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l 6.4 Spurious Turbine Runback With the unit at 90% power on April 21, 1987, Dropped Rod / Rod Stop, Overpower Rod Stop and Overpower Trip alarms were received from one of four Nuclear Instrumentation System (NIS) Channels (No. 33). The nega-tive rate Dropped Rod / Rod Stop signal initiated a turbine load runback to about 85% power. Operators over-rode the runback after they deter-mined that the actuations were spurious. The licensee installed a jumper on the turbine load runback signal for the affected channel because the spurious NIS fluctuations continued. Such jumpering is acceptable as long as rod out motion remains inhibited, and that was the case. Prompt notifications were made to local, state and NRC officials. A spare drawer was installed on April 22. The jumper remained for an additional day while operations monitored the channel for additional spikin Anomalies such as this have been a continuing problem in the NIS pri-marily due to aging components. The licensee has developed a long term corrective action plan to replace the NIS because of aging, replacement part procurement considerations, and previous similar events on other NIS channels. In the interim, the licensee closely monitors the NIS performance through weekly verification of system operability. The inspectors had no further questions at this tim . Review of Periodic and Special Reports Upon receipt, periodic and special reports submitted pursuant to Technical Specification 6.9 were reviewed. This review verified that the reported in-formation was valid and included the NRC required data; that test results and supporting information were consistent with design predictions and performance specifications; and that planned corrective actions were adequate for resolu-tion of the problem. The inspector also ascertained whether any reported information should be classified as an abnormal occurrence. The following periodic report was reviewed:
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Monthly Operating Report 87-03, plant operations from March 1-31, 1987 No unacceptable conditions were identifie . Emergency Preparedness Drills The licensee conducted two practice Emergency Preparedness drills and three drill training sessions during this inspection period. The inspectors ob-served the two practice drills on March 24 and April 30, 1987 and attended the post drill critiques. Control Room and Emergency Operations Facility activities were observed to ensure that exercise objectives were being ade-quately tested and that previous drill performance weaknesses were correcte The primary objectives of these exercises included demonstration of the acti-vation of the emergency response facilities, notification of key officials in the emergency organizations, utilization of communications systems, deter- i mination of emergency action levels, and technical evaluation of plant condi-
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tions to develop strategies for accident mitigation and recover In addition, the drill on March 24 was an unannounced exercise which challenged the licen-see's ability to fully staff the emergency response facilities within one hou The onsite and offsite organizations' abilities to handle medical emergencies involving extreme personnel external and internal overexposures were tested in the April 30th exercis For the March 24 drill, on-call personnel responded within the one hour time limit with the exception of three individual In response, the licensee conducted retraining and re emphasized the importance of responding promptl Also noted during that same drill was that the control room data coordinator (CRDC) was unfamiliar with the data sheets and the control board. This re-sulted in untimely delivery of data to the Emergency Operation Facilit During the April 30 drill, the inspector noted that the CRDC was able to promptly gather the necessary data and deliver it to the EOF. While the off-site portion of the medical emergency drill successfully met its objectives, weaknesses in the administration of first aid onsite were noted. Exercise control problems contributed to this situatio The licensee also noted this weakness in the post-drill critique and has agreed to conduct an additional medical emergency drill later this yea The post-drill critiques conducted by the licensee were frank and open dis-cussions of the weaknesses and strengths noted during the drills, and included most of the inspector's observations. Constructive criticism was aired in many area Resolution of noted weaknesses will be reviewed during NRC ob-servation of the annual site emergency exercise in May 198 . Followup on Systematic Evaluation Program Finding .1 Topic VIII-2. Onsite Emergency Power Systems - Diesel Generator I
This topic addresses the automatic trip features of the Emergency Diesel '
Generators (EDGs) under accident conditions. Current licensing criteria require that, under accident conditions, the EDG systems retain only the engine overspeed and generator differential trips. All other trips are ;
bypassed during an emergency EDG start. An acceptable alternative is ;
to provide two or more independent measurements of any other trip not !
, bypassed under emergency conditions. Such trip logic must sense an existing potential for EDG failure (no precautionary trips) prior to an l EDG tri The Systematic Evaluation Program found the plant in compli- '
ance with these criteria except that the EDG overcurrent, reverse power, and loss of field trips were not bypassed during emergency operatio The licensee proposal to either bypass these trips or install coincidence logic was approve The licensee upgraded their EDG systems to meet these criteria with Plant i
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Design Change Requests (PDCRs) 636 and 791 during the last two refueling l outages. PDCR 636 was incorporated during the 1984 refueling outag l With this PDCR, equipment was installed to bypass the reverse power and loss of field EDG trip These trips still annunciate in tne control i
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room to call operator attention to the condition to permit any necessary manual action Coincidence logic was initially intended to be installed
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for the overcurrent trip but suitable relays were unobtainable. There-i; fore, bypass circuitry for the overcurrent trip was installed under PDCR 791 during the 1986 refueling outage. As with the reverse power and loss of field trips, the overcurrent condition is alarmed in the control room even though the trip is bypassed.
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Both completed PDCR packages were reviewed to verify adequate documenta-tion of safety evaluations, ALARA reviews, post modification testing, and necessary procedural change The inspector noted an error in post modification testing of PDCR 791 discovered in 1986 during testing of Appendix R control circuit modification This error was documented in Plant Information Report (PIR) 86-91. Corrective action and retesting were completed in accordance with the PIR Program. Licensee corrective
, actions regarding the adequacy of post-modification testing were dis-cussed in NRC Inspection Report 50-213/87-0 Periodic testing of these modifications is performed each refueling outage in accordance with PHP 9.8-10, EDG Breaker Trip Test, SUR 5.1-18, Test of EDG 2A with Partial Loss of AC Coincident with Core Cooling Ac-tuation, and SUR 5.1-19, Test of EDG 2B with Partial Loss of AC Coinci-dent with Core Cooling Actuation. The inspector reviewed SUR 5.1-18 and SUR 5.1-19 performed on April 9 and 14, 1986, respectively. Bypass trip system testing methods and results were found to be adequat Based on the above. review, this' item is close .2 Topic V-10.8 - Use of Safety-Grade Systems for Safe Shutdown This topic addressed the ability to perform a plant shutJown and cooldown i
using only the systems identified in the Minimum System List (MSL). The licensee implemented the Westinghouse generic Emergency Operating Proce-dures (E0Ps) in September 1986 for compliance with TMI Action Plan Item
. I.C.1. These E0Ps utilize the systems listed in the MSL to perform a ,
plant shutdown and cooldown, including provisions for using backup sys- l tems if the primary systems are unavailable. Procedures also exist for l
conducting plant shutdowns with a total loss of AC power, a station l
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blackout, and operations from outside the control room. The Safety Evaluation Report, issued in August 1986, concluded that the E0Ps developed and implemented in accordance with the licensee's program would adequately provide control room personnel with guidance to effectively
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mitigate the consequences of a broad range of transients and accident An inspection of the E0Ps was conducted by NRC during the week of April 12, 1987 (NRC Inspection Report 50-213/87-10). This inspection focused on verifying that the E0Ps were prepared in accordance with the NRC-l approved Procedures Generation Package (PGP) and are technically adequate.
Based on the satisfactory results of this inspection, this topic is closed.
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l 10. Temporary Procedure Changes
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The inspector reviewed the implementation of temporary procedure changes (TPCs) in accordance with TS 6.8 and Administrative Control Procedure 1.2-6.4,
- Temporary Procedure Change. Ten TPCs were reviewed, including two changes j which were disapproved by the PORC because the changes did not meet the -
i specified criteria for temporary changes. In general, the TPCs were processed
in accordance with the requirements of ACP 1.2-6.4. The inspector found that i the licensee cancelled-the disapproved TPCs and notified the initiators that the TPC did not comply with the procedural requirements. However, the basis '
! of the disapproval and discussions of necessary corrective actions addressing j these errors were not documente A disapproved TPC may identify a noncom-
- pliance from the administrative requirements of TS 6.8, in which case correc-i tive action to prevent recurrence is required. In the cases reviewed, the
, disapproved TPCs reflected licensee identified administrative matters which >
l did not involve _ compliance with Limiting Conditions for Operation. To assure a
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that such corrective actions are evaluated, implemented and documented, the l licensee revised procedure ACP 1.2-6.4 to require documentation of the basis !
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for disapproval in PORC meeting minutes and initiation of corrective action review and tracking when necessary. The inspector verified the implementation of these changes in revision 20 to the ACP. The inspector had no further questions in this are . Unresolved Items Unresolved items are matters about which more information is required in order
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to determine whether they are acceptable items or violations. Unresolved items identified during this inspection are discussed in Paragraph . Exit Interview !
j j During this inspection, meetings were held with plant' management to discuss
, the finding No proprietary information related to this inspection was
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