IR 05000255/1987015

Insp Rept 50-255/87-15 on 870602-0706.No Violations Noted. Major Areas Inspected:Followup of Previous Insp Findings, Operational Safety,Maint,Surveillance,Physical Security & Radiological Protection

ML18052B218
Person / Time
Site:	Palisades
Issue date:	07/24/1987
From:	Burgess B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML18052B213	List: IR 05000255/1987015 ML18052B210
References
50-255-87-15, NUDOCS 8707310110
Download: ML18052B218 (12)

Text

U.S. NUCLEAR REGULATORY COMMISSION REGION I II Report No. 50-255/87015(DRP)

Docket No. 50-255 Licensee:

Consumers Power Company 212 West Michigan Avenue Jackson, MI 49201 Facility Name:

Palisades Nuclear Generating Plant Inspection At:

Palisades Site, Covert, Michigan Inspection Conducted: June 2 through July 6, 1987 Inspectors:

E. R. Swanson C. D. Anderson T. V. Wambach Inspection Summary License No. DPR-20 Date Inspection on June 2 through July 6, 1987 (Report No. 50-255/87015(DRP))

Areas Inspected:

Routine, unannounced inspection by resident inspectors of followup of previous inspection findings; operational safety; maintenance; surveillance; physical security; radiological protection; engineered safety features walkdown and repbrtable event Results:

Of the areas inspected no violations were identifie One unresrilved item was identified involving the adequacy of the licensee's root cause determination and corrective action regarding links found open in the circuitry for the atmospheric dump valve One open item was also identified relating to periodic maintenance procedure development for the switchyard seismic monitor and evaluation of the containment seismic monitors sensitivity and rang PDR ADOCK 05000255 G

PDR

(Open) Unresolved Item 255/87014-02:

The hot and cold primary coolant temperature indicators were found to be wired ih reverse to the remote shutdown panel. 'Two junction boxes had been installed to facilitate the installation of the remote shutdown panel in 198 Wiring mislabelling in both junction boxes caused a double reversal which gave correct control room indication though reversed remote shutdown panel: indicatio The wiring has been corrected arid tested satisfactorily~ The licensee plans to have.an independent review of modifications involved with the installation and testing of the remote shutdown panel fo~ any-similar problem Th.is item will remain open pending satisfactory completion of the review as documented in Licensee Report No. E-PAL.;.87-02 No violations or deviations were identifie.

Ope rat ion a 1 Safety The inspectors observed control r6om activities, discussed these activities with plant operators, and reviewed various logs and other operations records throughout the inspectio Control room indicators and alarms, log sheets, turnover sheets, and equipment status boards were routinely checked against operating requirements; Pump and valve controls were verified *to be proper for applicable plant condition On several occasions, the inspect6r observed shift t~rnover activities and shift briefing meeting *

Tours were conducted in the turbine and auxiliary buildings, and central alarm station to observe work activities and testing in progress and to observe plant equipment condition, cl~anliness~

fire safety, health physics and security measures, and adherence to procedural and regulatory requirement An ongoing review of all licensee corrective action program items at the Event Report level was performe At about 8:00 p.m. on June 10, 1987, ground motion was detected by operators in the control roo Safety Injection Tank levels fluctuated causing intermittent alatm An Unusual Event was declared, notifications were made, and operators and engineers conducted extensive plant walkdowns to identify any possible effect No damage was reporte Seismic instrumentation was inspected and it was found that the switchyard Strong Motion

• Accelergra.ph (SMA) was nonfunctional due to a dead battery, caused by failure of the charging circuitry. This monitor was repaired and returned to servic The Peak Recording Accelerometer (PRA) on the 590 level of the Reactor Building was retrieved anq found to be reading approximately O.lg in all th.ree axe The.data obtained from this 11 scratch plate*1. represented 5.0% of the 2g sensitivity rang The licensee postulated that the reading may have been due to motion during installation or physical contact with the device and may not accurately reflect the actua 1 ground motion *during the even The sensitivity of the instrument was det.ermined not to be appropriate for measuring 0.lg since that value falls within the*

vendor 1s total error band and coincides with the value of the

DETAIL.

Persons Contacted Consumers Power Company (CPCo)

• D. P. Hoffman, General Manager

.*J. G. Lewis, Technical Director

• R. D. Orosz, Engineering and Maintenance Manager

• R. M. Rice, Operations Manger

• D. W. Joos, Administrative and:Planning Manager

• W. Beckm~n, Radiological Services Manager

• D. J. Malone, Licensing Arialyst

• R. E. McCaleb, Quality Assurance Director R. Brzezinski~ Instrument and Control Superintendent

• R. A. Fenech, Operations Superintendent S. C. Cote, Property Protection.Supervisor

• J. S. Erickson, Genera 1 Engi,neer, Pl ant Safety Engineering

• Denotei those present at the Management Interview on July 6, 198 Other members of the Plant staff, and several members of the Contract Security Force, were also contacted briefl.

Followup on Previous Inspe~tion Findings

. (Closed) SEP Item "(S4028-0l): The licensee performed an additional analysis related to the environmental qualification of the components located in the Auxiliary Feedwater (AFW) pump roo This review focused on inspector concerns over the previous analysis technique demonstrating the adequacy of the AFW pumps, motor and turbine driver lubrication at the elevated temperatures caused by the loss of room ventilation. It was determined that the rule of thumb used in the prior analysis showed a larger conservatism than the current analysis for the oil lifetime at

• elevated temperature The reanalysis demonstrated adequate margin and conservatism with respect to the lubricant and hydraulic fluid utilized in the AFW pump roo A concern was expressed at the exit meeting on*

July 6, 1987 for the adequacy of the existing room ventilatio The system engineer was already aware of and pursuing this issu (Ope_n) Unresolved Item 255/87008-01: The licensee failed to reduce power to less than 85% within two hours as required by Technical Specification (TS) 3.23.1 when both the excore and incore monitoring systems were inoperable for linear heat rate determinatio Nd violation will be issued for this i tern as a 11 owed by 10 CFR 2, Appendix C:

the vi o 1 at ion was identified by the licensee, would be a Severity Level IV or V, was reported and could not have been expected to have been prevented by previous corrective action In regard to corrective action~ to prevent recurrence, the licensee plan~ to submit a TS change request and revise their Standing Order 54 to clarify this T This item.will remain open until these actions are take Palisades Plant design earthquak The licensee is evaluating replacement of the PRAs with more sensitive (e.g., O.Sg) devices to provide better resolution of expected sei~mic event This evaluation wi 11 be tracked by Open Item 255/87015-01 along with the develop~ent bf a procedure t6 perform preventive maintenance on the switchyard SM *On June 18, 1987, during a routine tour of the auxiliary building, the inspector observed that the watertight door to the East Engineered S~feguards room was improperly dogge Only one of the eight dogs was properly secure This door*provides fire and flood protection to one train of.safeguards equipment:* The licensee does

• not have a procedural control for this door, onl~ a painted message*

on the doo The inspector properly dogged the door upon leaving~

reported the 5ituatibri to the Shift Supervisor and discussed the issue at the management interview.

. After reducing power for governor valve t~sti~g on June 20, 1987, the turbine was rapidly taken off lirie at 11:40 p.m. in response to an electro~hydraulic control (EHC) fluid leak on the No. 4 Governor Valve Supply lin Reactor power was reduced te less than 2% (Hot*

Standby) and maintained there until the generator was synchronized on June 25, 198 A 2/3 circumferential through..:.wa 11 EHC tubing c.rack was found, at the base of the flared fitting and was caused by fatigu The licensee dye penetrant tested (DPT) the supply lines on the other governor

~alves, stop valves, intercept valvei, and reheat stop valves:

One questionable indication was found in another tube and it and the failed t~be were replace The system was tested on June 23, 1987, when the valves were stroked and vibration monitored; all of which revealed no abnormalities..

All lines similar to the one that failed were instrumented and monitored during pl~nt startup and power escalatio At about 27% power, some high amplitude low frequency vibration was identified associated with the No. 3 Governor valv Power was increased to 35% where further stroking of the valve could not reproduce the vibratio During power escalation at about 92%

power, high frequency vibration was identified on the No. 4 Gcivernor valve as it was starting to ope Po~er was reduced and the valve was isolated, stopping the vibratio This problem had also been seen in 1985 when the internal. dump valve in.~ne bf the governor valves was found leakih All the dump,valves were sub~equently replaced.. The licensee plans tb replace the governor valve as soon*

as parts and plant conditions permi As a.long term corrective measure, the licensee is planning to modify the EHC tubing in the October 1987 maintenance outage either by welding the tubing or adding flexible connection * Following the rapid power reduction, a number of equipment problems were identified and evaluated including a high level in the 11A

Steam Generator (SG).

The high level occurred when the 11A11 Main Feedwater Regulating Valve (FRV) failed to close due to a plugged air relay in the pneumatic positioner. *It is*postulated that *. *

co~rosion.particle from the carbon steel air manifold detac~ed and plugged the 3-5 mil orific The licensee had replaced the*

instrument air dryer and added low point blowdowns during the last refueling outage to eliminate past moisture and desiccant carryove The licensee replaced the carbon steel manifolds and added fine mesh filters for the two FRVs, Corrective actions for other valves are also being evaluate Foilowing the power reduction, the 11A11 Main Feedwater (MFW) puinp turbine Trip and Throttle (T&T) valve failed to close when the FW turbine was tripped due to suspected crud on the stem or dis This

~id not contribute to the overfeeding of the SG since the discharg pressure was less than SG pressur The lic~nsee plans tb put a protectfve cover over the valve stem, and instruct operators to verify T&T closure after MFW pump trips. The sticking condition was not reproducibl On the main turbine trip, the Moisture Separator Reheat~r (MSR)

steam admission valves failed to close as expecte Upon investigation, the licensee discovered an unknown portion of the circuitry that removed manual control and the expecte.d automatic closure on turbine trip when the low pressure turbines* inlet temperature exceeds the set point of 490 degrees.* The set point should have been between 525 degrees -.575 degfee The licensee has set the set point at. 550 ~egrees and is investigating the need and desirabilityof maintaining this feature which was designed to slowly close the MSR valves on sensing high temperatur A Sigma SG level indicator was discovered failed after the shutdown due to a failed servo-drive mechanism and was replace These indicators are currently scheduled to be replaced at a rate of 25%

per year, beginriing this yea The resident inspector is further pursuing their failure rates and correctiv~ action pla On June 21, 1987, the Diesel Generators (DG) auto-started when the main turbine was latche Troubleshooting has shown that the

• Mercoid pressure switches are susceptible to mercury movement (bounce) at the set point range due to the filling and pressurizing of the auto stop oil syste The licensee plans to proceduralize the holding out of one DG (with the other already running) during the latching of the turbine to prevent inadvertent DG start Additional long term corrective actions are planned to evaluate the spurious relay actuation and an alternate type of rela The NRC staff was concerned whether these failures reflected adversely on the effectiveness of th~ material.condition improvement program instituted after the.May 19, 1986, reactor trip and used a~ a ba~is for resta~t in March 198 As a result, the NRR Project Manager was dispatched ~o assist in an onsite review of the ci~cumstances of these failures.

5 *

The EHC system was included in the Mat~rial Condition Task Force improvement program (Observation N TGS-01)~ Work was performed to stop internal leaks (governor ctintrol and dump valves) and external leaks at the flared fitting In 1985, another EHC tube ruptured at the control block for a different governor *valve due to vibration produced by excessive leakage past the associated dump valv At th~t time, all the flared ends of tubing leading to the No. 1 a~d No. 4 governor valves and, in addition, the high pressure tubing to all other valves were* inspected using DP The tubihg

.that failed on June 20, 1987, showed no crack indications in 1985. *

During operation since the maintenance outage, the EHC system was monitored by system walk-down at least once a shift.. The failed tubing was replaced and the system has now been instr.umented to allow detection and improved monitoring of vibrations during return to operation as noted above in Subparagraph The Main* Feedwater Regulating Valves were also included in the improvement program (Observation No. FWS-01).

The failure of.*

the valve was attributed to a blockage of a 3-5 mil nozzle in the pneumatic controls for the valve by foreign matte During the maintenance outage prior to this restart, the air system including the lines. for th~ valves, was systematically blown out and checked for* foreign materia The licensee now believes that a carbon steel valve harness around a filter in the air supply to this pneumatic control system is a source of continuing contamination to this system and that the filter itself is too coarse for the small nozzle in this appliC:atio They therefore have replaced the carbon steer with copper and i nsta 11 ed a five micron filte The remaining three failures (i.e., the T&T valve, MSR steam

• admission valves and SG level indicator) were in components that did not fall under the purview of t~e Material Condition Task Force because there were no outstanding work orders or previous failure histories on those component The licensee had previously planned replacement of the Sigma type indicators used for the steam generator lev~l oveF a four-year period based on a~ undesirable

fail-as-is" feature and anticipated maintenance pr6blems with the servo-drive featur The T&T valves for the. main feedwater pump turbines and the moisture separator reheater would not have fallen under the *scope of the Task Force as being components important to reliable plant operation or safety-rel~te Based on the above considerations, the NRC concluded that these failures would not invalidate the conclusion on plant readiness for operatio On June 23, 1987, ~ technician who was pteparing to check the calibration of the Atmospheric Dump Valve (ADV) program; found links open which disabled the ADV "quick-open" featur A review of previous manipulation of these links determined that they were apparently opened on April 12, 1987, while troubleshooting an apparent inadvertent quick-opening of the ADV A procedure was issued for a.similar calibration but not completed (leaving the

4.

links open) when troubleshooting efforts were directed to another are The procedure was not made a part of the work package and therefore reviews were not able to identify the incomplete procedur Due.to plant conditions, post maintenance testing was not done to verify the operation of the quick-open feature, although the normal ADV mode was verified. *

No safety concerns result from the operation with this feature disabled.* During the turbine trip on June 20, 1987, ADV response was not ~oted as sluggis The quick-open feature ramps the ADVs full open on a turbine trip where otherwise the valves ~ould open to 25% open and then modulate from that pofnt, in response to pr~ssure/temperature. The liten~ee has identified pe~sonnel error as the root cause and has i dent ifi ed the event for review by the Human Performance Evaluation System (HPES) Coordinator. A review of all links in the control room was completed on June 24, 1987, with no other misalignments. identifie The inspector will review the HPES recommendations when they are availabl Additional

investigation of this event is required to determine whether a violation should be issued (Unresolved Item 255/87015-02). During troubleshooting of a reported early opening of the ADVs (535-537 degrees vs. 540 degrees design),

several secondary Code safety valves opene While in hot shutdown on June 24, 1987, a controlled increase in primary coolant temperature to test the ADV setpoint was terminated when the ADVs had not opened at 541 degree Temperature increased to 542 degrees at which point the safeties *

lifted after which they properly reseated. *This Value is within the tolerance of the combined temp.erature instrument and safety setpoint The.ADVs did not open because the arming signal (relay) had been defeated by the operators* attempt to ensure that they were armed~

After review of the logic, the relay was properly reset and the ADV setpoint of 540 degrees verified. A concern raised by this event is that operator knowledge of this logic circuit resulted in operation with the ADVs unknowingly disable Challenges to th safeties are considered highly undesirable by the NRC due to the

• potent i a 1 for malfunctionin Aside from this, no safety concern exist Planned licensee corrective action~ include isiuance of a memorandum to all Senior Reactor Operators, operator training, and proceduralizing the preplanning and troubleshooting of similar activities in the futur ~o viol~tions or deviations were identifie Maintenance

~he inspector reviewed and/or observed the following selected work activities and verified whether appropriate procedures were ~n effect

controlling removal from and return to service, hold points, verification testing, fire prevention/protection, radiological controls, and cleanliness where applicable: Instrument air tubing and filter replacement for "A 11 feedwater regulator valve CV-0701 (WO FWS 24703913). Gas effluent program controller troubleshooting (WO RIA 24703920). Testing of AFW flow control valves (FWS 24702764).

No violations or deviations were ident1fie.

Surveillance The inspectors reviewed surveillance -activities to ascertain compliance with scheduling requirements and to verify compliance with requirements relating to procedures, removal from and return to service, personne qualifications, and documentatio The following test activities were inspected: * M0-7A-2 Emergency Diesel Generator 1-2 monthly surveillanc DW0-1

. Weekly Control Rod Exercisin DW0-13 Personnel Airlock Local Leak Rate Tes ME-12 Monthly Battery Check N6 violations or deviations were identifie.

Physical Security The in~pectors observed physical security activities at various locations through out the protected and vital areas including the Central and Secondary Alarm Station Periodic observations of access control activities including proper personnel identification, badging and searches of personnel, packages ~nd vehicles were conducte The inspectors verified appropriate security force staffing and operability of search equipmen Protected and vital area boundaries were toured to verify maintenance of integrity. Ill umi nation was verified to be adequate to support patrol and Closed Circuit Televisi9n (CCTV) monitor observation CCTV monitor clarity and resolution was also observe The inspectors periodically verified that appropriate compensatory measures were taken for degraded or inoperable equipment and breached boundarie No violations or deviations were identifie.

Radiological Protection The inspectors made observations and had discussions concerning radio-logical safety practices in the radiation controlled areas including:

verification of radiation levels and proper posting; accuracy and

currentness of area status sheets; adequacy of and compliance with selected Radiation Work Permits and high radiation procedures; and the ALARA (As Low As is Reasonably Achievable) progra Implementation of dosimetry requirements, proper personnel survey (frisking) and cbntaminatton cohtrol (st~p-off-pad) practices were observe Health Physics logs and dose records were routinely reviewe.

.

No violations or deviations were identHie.

Engineered Safety Feature Walkdown The inspectors performed a walkdown of the 1-1 Emergency Diesel Generator and verified that e~ch accessible valve in the flowpath w~s in its required position and operable, that power was aligned for components that activate on an initiation signal, that essential instrumentation was operable, and that no conditions existed which would adversely affect system operation. *Several minor discrepancies (checklists, labeling, etc.) were reported to the operations departmen No vioiations or deviations were identifie.

Licensee Event Reports Thr6ugh direct observations, discussions ~ith litensee personnel, and review of records, the inspectors examined the following reportable events to determine whether:

reportability requirements were met; immediate ~orrective action was accomplished as appropriate; and torrective action to prevent recurrence has been accomplished.per the Technical Specification ( C'l osed) LER 255/87005:

The failure of the 11A 11 Steam Generator north secondary manway relief valve created a containment leakage pat This situation was identified during the plant heatup on February 16, 1986, and as a result, the plant was placed in cold ~hutdow An Unusual Event was declared based on the conclusion that the leakage constituted a violation of containment integrit The measured leakage, when combined with other known leakage exceeded the maximum leakage rat Repairs to the leaking inner manway and relief valve were completed and tested and the plant returned to powe.

.

A series of evaluations as to whether the secondary manway constituted a

• containment isolation boundary ensued over the next nfoe month With the involvement of the Palisades Corrective Action Board, the Plant Review Committee (PRC), the Nuclear Safety Board (NSB), the Nuclear

• safety and Licensing group and the Plant Safety Engineering group, it was not concluded until November 9, 1986 that the penetration was a containment-isolation boundar Feedback to the licensing group of the conclusion was not effective in starting the 30 day reporting clock until February 6,*

198 Although not contained in the LER dated March 9, 1987, corrective actions taken with res~ect to the late rep9rting include the Plant Licensing Engineer attending NSB meetings and receiving copies of the minutes to preclude a recurrence of PRC and NSB disagreement without

fe~dback to the correct group fo~ actio since the issues were licensee identified;

. violations; were reported; were corrected;

• been prevented by corrective action from a No violation will be cited did not constitute serious and were not expected to hav~

a ~reviou~ violatio (Closed) LER 255/87009:

While in hot shutdown, an inadvertent auxiliary feedwater (AFW) system actuation signal started the 11A 11 AFW pump due to an actual low level in the 11811 steam generator (SG).

The LER states in error that the 11C 11 AFW pump started while in fact the 11A 11 pump started as designe The cause of the low level was attributed to shrinkage when the atmo~pheric dump valves were tested while maintaining SG levels at

.approximately 30% for chemistry control:

All licensed operations personnel will be informed of this event through Continuing Trainin The licensee has revised the testing procedure to include the prerequisite of having SG levels greater than 45% prior to testing'.

(Closed) LER 255/87011: On April 12, 1987, at 5:24 a.m. the reactor was manually tripped from 75% by operators who identified the impending failure of the turbine Electro-Hydraulic Control (EHC) fluid syste The

_inspector witnessed the eveht from the control roo An unusual noise emanating from the EHC System and the starting of the standby pump alerted operators to a proble A small leak was found and the EHC fluid pump P~19B was vibrating and making excessive nois Operators turned off the B pump and attempted to tighten what appeared to be a fitting lea Since EHC pressure continued to-fall and the governor valves were beginning to close the operators tripped the plan Operator actions were-prudent and timel Follow* up to the trip utilizing the new Emergency Operating Procedures went smoothly, all systems functioning as designe Cause of the event was attributed to failure of the turbine EHC pump impeller which had evidently stress fracture The vibration of the failed pump caused both the fluid leak and the pump suction valves t vibrate closed. It was not the leak but the loss of pump suction which caused the EHC pressure to detay since the reservoi~ had not em~tie The P-19A pump discharge valve was also found to be partially open, but*

was not a prime contributor to the even The pump impeller was found to have failed due to cracking which was caused by improper heat treatmen The impellers on the other pump and the spares did not exhibit any cracks under dye penetrant testin The suction valves are now lock wired open to prevent undesired movement, and

- the tubing which cracked was replace The filters were inspected and metallic particles were found downstream of the filters, which were due to a broken fittin As a pr:ecautionary measure a flush of the EHC s~stem was performe No concerns remain regarding this event, and in fact it represents an example of good work by the operations, maintenance, engineering and management tea The LER form was noted as incomplete in that the pump impeller failure was not coded in block No. 1 (Closed) LER 255/87012:

While paralleling the main generator on April 17, 1987, at 8:18 a.m., a turbine generator trip on reverse power resulted in the automatic starting of diesel generator 1-1. This event was witnessed by the inspecto The 1-2 generator was already operating and the start

feature functi6ned as designed; The reactor wai at approximately five per cent power and did not trip since the turbine-reactor trip is not enabled until power is greater than 15%.

An EHC system fluid leak.from the turbine interface valve was believed to be the cause of the governor valves not responding properl The leak was repaired and a second. attempt was made. with the same result The reverse power trip was determined *to be caused by improper function of the turbine control syste Both diesels were already running during the second tri Licensee investigation into the Westinghouse control system determined that a voltage s-etting determines the initial valve setting and therefore, the amount of load that the turbine/generator picks up *

wh~n parall~led. A prior problem with this setting, which had evidently drifted, caused the generator to pick up 22.5 per cent Toad instead of the five per cent expecte Having found the voltage at 2.5, they determined from system prints that 0.5 volts would give the proper respons No testing or verificatioh of this setting was done, and based on a later inquiry with Westinghouse, the licensee learned that the value can only be determined accurately by measurement of the voltage at the desired power level and valve positio By comparison of the reference counter numbers displayed on the turbine controls on prior startups t the voltage it was determined that 1.6 volts was closer to five percent powe Adjustments were made and synchronization was successfully completed at 3:05 p.m. on the same day.* The diaphragm on the interface valve was replaced after the second trip and also functioned properl The licensee further plans to check the initial turbine gbvernor valve setpoint voltage during each refueling outage to ensure its correct -

setting. Th~ licensee was informed that the LER was incomplete in not having completed block No. 13 of the LER form for the de~cribed component failure (Closed) LER 255/87014:

The 1 icensee failed to reduce power to less than 85% in the two hours required by Technical Specification (TS) 3.23.1 for having inoperable incore alarm and ~xcore nuclear instrumentation system As discussed in Inspection Report No. 255/S7008(DRP) and Paragraph 2 of this report, Unresolved Item 255/87008-0l(DRP) is being used to track resolution of the apparent violation and corrective actions including submittal of a TS change reques (Closed) LER 255/87016:

Due to an Auxiliary Operator (AO) valving error while redirecting moisture separator reheater drains, a manua 1 reactor.

trip was initiated from approximately 35% powe The secondary side AO received valve lineup instructions from the control room and then requested assistance from an additional A The secondary side AO pointed to a set of valves for the additional AO to clos The additional AO closed the turbine driver exhaust valve which he thought had been pointed to~ This incorrect closure caused the turbine driver exhaust line rupture disc to operate: _Upon evaluation of the situation, the Shift Supervisor ordered a turbine and reactor trip. The two AOs were given disciplinary time off for this even Additionally, these AOs are

speaking to each shift of AOs describing the event, their failure to properly execute the valve lineup and the significance of the even All operations personnel will receive training on effective communications technique These actions should prevent recurrence of a similar even No violations or deviations were identifie.

Unresolved Items Unresolved items are matters about which more info~mation is iequired ih order to ascertain whether they are acceptable it~ms, violations or deviation An Unresolved Item disclosed during the inspection is discussed in Paragraph 3~ *

1 Open Items Open items are matters which have been discussed with the.licensee, which will be reviewed further by the inspectors, and which involve some action on the part of tha NRC 6r licensee or bot An open item disclosed during the inspection is discussed in Paragraph.

Management Interview A management interview was conducted on July 6, 1987, at the end of the inspectio The scope and findings oft.he inspection were discusse The inspector also discussed the likely information content of the inspection report with regard to documents or processes reviewed by the inspectors during the *inspectio The licensee did not identify any such documents/processes as proprietar