IR 05000212/2003018
| ML17321A534 | |
| Person / Time | |
|---|---|
| Site: | Cook, 05000212  |
| Issue date: | 04/01/1985 |
| From: | Wright G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML17321A533 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-1.C.1, TASK-TM 50-315-85-09, 50-315-85-9, 50-316-85-09, 50-316-85-9, IEB-82-01, IEB-82-1, NUDOCS 8504110294 | |
| Download: ML17321A534 (19) | |
Text
U.
S.
NUCLEAR REGULATORY COMMISSION
REGION III
Reports No. 50-315/85009(DRP);
50-316/85009(DRP)
Docket Nos. 50-315; 50-316 Licenses No. DPR-58; DPR-74 Licensee:
American Electric Power Service Corporation Indiana and Michigan Electric Company Columbus, OH 43216 Facility Name:
Donald C.
Cook Nuclear Power Plant, Units 1 and
Inspection At:
Donald C.
Cook Site, Bridgman, MI Inspection Conducted:
February 12, 1985 through March 18, 1985 Inspectors:
B. L. Jorgensen J.
K. Heller Approved By:
G.
.
right Chief Reactor Projects Section 2A 4/i F~
Date Ins ection Summar Ins ection on Februar
throu h March
1985 Re orts No. 50-315 85009(DRP '0-316 85009 DRP Areas Ins ected:
Routine unannounced inspection by the resident inspectors of licensee actions on previous inspection findings; operational safety; maintenance; surveillance; Licensee Event Reports; IE Bulletins and Circulars and NUREG-0737.
A management meeting to review Regulatory Performance Inprovement Program status was also conducted.
The inspection involved a total of 274 inspector-hours by two NRC inspectors including 32 inspector-hours off-shift.
Results:
No items of noncompliance or deviations were identified in any of the areas inspected.
8504110294 850401 PDR ADOCK 05000315
f I
DETAILS 1.
Persons Contacted a.
Personnel attending March 5, 1985 Regulatory Performance Improvement Program meeting.
Indiana and Michi an Electric Com an J.
E.
M. P.
W.
G.
J.
G.
P.
A.
Dolan, Vice Chairman, Engineering Ec Construction (AEPSC)
Alexich, Vice President, Nuclear Operations (AEPSC)
Smith, Jr., Plant Manager Feinstein, Manager, Nuclear Safety and Licensing Barrett, Senior Licensing Engineer (AEPSC)
Other members of the corporate and plant staffs were also present.
USNRC Personnel J.
G. Keppler, Regional Administrator C. E. Norelius, Director, Division of Reactor Projects W. D. Shafer, Chief, Projects Branch
G.
C. Wright, Chief Projects Section 2A C. J. Paperiello, Chief, Emergency Preparedness and Radiological Protection Branch J.
K. Heller, Resident Inspector B. L. Jorgensen, Senior Resident Inspector b.
Ins ection Februar 12 throu h March
1985
- W B.
~>A.
- >K.
~"N.
- J G. Smith, Jr., Plant Manager Svensson, Assistant Plant Manager Kriesel, Technical Superintendent-Physical Science Blind, Technical Superintendent-Engineering Baker, Operations Superintendent Williams, Maintenance Superintendent Stietzel, Quality Control Superintendent Beilman, Quality Assurance Superintendent The inspector also contacted a number of licensee and contract employees and informally interviewed operation, technical and maintenance personnel during this period.
<<Denotes personnel attending exit interview on March 19, 198 l
2.
Licensee Actions on Previousl Identified Items a.
(Open)
Open Item (315/83-11-02; 316/83-12-02):
The facility organi-zation as identified in Technical Specification 6.2.2., Figure 6.2-2 does not reflect the current staffing.
The inspector reviewed a
recent Technical Specification Change Request (AEP:NRC:0659C) which updates the PSNRC membership to reflect the currect staffing plan, but it does not request a change to Figure 6.2-2.
This was discussed with the AEP Senior Licensing Engineer.
b.
(Open)
Open Item (315/83-02-02; 316/83-02-02):
The licensee redefined the plant fire zones on February 10, 1983 but had not resolved the question of the acceptability of previously unsealed small diameter conduit penetrating fire seals.
A subsequent licensee evaluation determined that previously unsealed conduit penetrating fire seals will not be sealed unless the conduit is modified, requiring rework of the seal.
The inspector has asked RIII's assistance in closeout of this item.
C.
(Closed)
Noncompliance (315/82-04-02):
Revision of plant drawings to reflect "as-built" conditions was not timely.
Revision 6 to PMI 5040 "Design Change Control Program" appears to resolve the concern by requiring that all affected
"OP" drawings be updated prior to changing to an applicable mode.
d.
(Closed)
Open Item (315/84-02-02; 316/84-02-02):
The procedure for installation and repair of fire seals did not require use of cali-brated scales when performing foam quality checks and did not require posting of no smoking areas.
The current revisions of QHP 2270.QC.001 and
<<:~MHP 5020.001.031 were reviewed; the concerns were addressed.
e.
(Closed)
Noncompliance (316/84-06-01):
Failure to maintain a fire barrier.
The licensee's action identified in his response dated June 22, 1984 (AEP:NRC:0891)
appears adequate to resolve this item.
(Closed)
Noncompliance (316/84-14-01):
The procedure for initiation of Recirculation Phase did not provide the instructions necessary to establish flow to the residual heat removal pumps.
The inspector verified the licensee's corrective actions identified in his September 28, 1984 (AEP:NRC:0899) response letter.
These actions appear adequate to resolve this issue.
(Closed)
Open Item (315/83-13-02):
Clarification of P.T. Procedure.
The inspector reviewed attachments 1 and 2 to 12 QHP 5050 NAR.001,
"Liquid Penetrant Examination for Nuclear and Non-Nuclear fields and Components" and found that the concerns were addressed.
No items of noncompliance or deviations were identifie erational Safet Verification The inspector observed control room operation including manning, shift turnover, approved procedures and LCO adherence; and reviewed applicable logs and conducted discussions with control room operators during the inspection period of February 12, through Harch 18, 1985.
Observations of control room monitors, indicators, and recorders were made to verify the operability of emergency systems, radiation monitoring systems, and nuclear and reactor protection systems.
Reviews of surveillance, equip-ment condition, and tagout logs were conducted.
Proper return to service of selected components was verified.
Tours of the auxiliary building, turbine building, and screenhouse were made to observe accessible equipment conditions, including fluid leaks, potential fire hazards, and control of activities in progress.
By observation and direct interview it was verified that the physical security plan was being implemented in accordance with the station security plan.
The inspector performed a review of the Unit 2 4KV, 600V and 480V ESS electrical lineup using licensee drawings (OP 2-1200A-4, and OP 2-1200G-4)
to verify that:
power (visual breakers and fuses)
was aligned to actuate on automatic signal; essential instrumentation was operable; and no condi-tion existed that degraded the system.
Flow path'verifications were per-formed by walkdown of the common Component Cooling Mater (CCM) emergency ventilation system (the licensee had earlier found both suction dampers isolated)
and the Unit 1 Control Room pressurization and recirculation ventilation system.
No discrepancies were noted.
Licensee testing during the inspection period, relating to ongoing Control Room "habitability" studies, twice resulted in short-term declarations of system "inoper-ability".
On both occasions the inspector verified prompt licensee corrective actions to restore the system within Technical Specifications.
The licensee's evaluation program is continuing.
During a review of Control Room logs, the inspector identifed an apparent noncompliance with Technical Specification 3.3.3.10, in that no documen-tation existed to show steam jet air ejector flow had been estimated each four hours with SFR 401 "inoperable".
Specifically, the February 10, 1985 logs showed estimates at 1000 and 1800 hours0.0208 days <br />0.5 hours <br />0.00298 weeks <br />6.849e-4 months <br />, but not at 1400 hours0.0162 days <br />0.389 hours <br />0.00231 weeks <br />5.327e-4 months <br />.
Upon identification of this matter, the licensee initiated Condition Report No. 1-2-85-345 to evaluate the situation.
The preliminary review suggests SFR 401 was never in fact "inoperable",
though it had been declared so; rather, the local reading which disagreed with SFR 401 may have been in error.
This is an Unresolved item pending completion and inspector review of the Condition Report evaluation (315/85009-01).
During facility tours, the inspector identified and referred a couple of minor problems for licensee corrective action as follows:
a.
As of March 1, 1985 a few local portable fire extinguishers had not been subjected to their "February" monthly quality checks.
Upon notification to the QC Department, the checks were completed that same day.
b.
A small pool of potentially contaminated water spreading across a
contamination boundary containing equipment (vessel head stud tensioning)
on the spent fuel pool deck was immediately cleaned up when pointed out to RP personnel.
No items of noncompliance or deviations were identified.
4.
Monthl Maintenance Observation Station maintenance activities of safety related systems and components listed below were observed and/or reviewed to ascertain that they were conducted in accordance with approved procedures, regulatory guides and industry codes or standards and in conformance with Technical Specifications.
The following items were considered during this review:
The limiting conditions for operation were met while components or systems were removed from service; approvals were obtained prior to initiating the work; activities were accomplished using approved procedures.
The following maintenance activities were observed and/or reviewed:
>>>>12 MHP 5021.087.001 Maintenance Inspection and Repair Procedure for 4KV Power Circuit Breakers
>>>>THP IMP.014 Time Overcurrent Relay Testing of The Hotwell Pump 12 THP 6030 IMP.026 Calibration of Diesel Fire Pump Pressure Switch 12 THP 6030 IMP ~ 028
>>>>2 MHP SP,087 Calibration of Diesel Fire Pump Temp Switch Special Procedure for Correcting "2AB" Battery Electrolyte Specific Gravity On February 28, 1985 the inspector observed a maintenance crew performing preventive maintenance on the Unit 2 "AB" Battery by making electrolyte specific gravity adjustments to selected cells per Job Order No. 035366 using Procedure
>>>>2 MHP SP.087.
The engineer at the job site stated surveillance testing had shown the electrolyte specific gravity in each cell was above the minimum required by Technical Specifications but
preventive maintenance was required to assure further testing would not find a cell inoperable due to low specific gravity.
The inspector observed the crew remove a predetermined amount of electrolyte from a cell and then add electrolyte with a higher specific gravity.
The inspector inquired if the battery became inoperable when the electrolyte was reduced below the lowest red mark on the cell casing.
The crew considered the battery funtionally operable as long as the electrolyte was above the plates.
The inspector reviewed the procedure, Job Order, and Unit 2 Control Room Logs, and found that the plant did not consider the battery inoperable when the electrolyte level was reduced below the minimum indication mark.
Units 1 and 2 Technical Specifica-tions, at paragraph 4.8.2.3.2.a.l, state that the 250 volt battery bank shall be demonstrated operable at least once per seven days by verifing that the electrolyte level of each pilot cell is between the minimum and maximum level indication marks.
In addition, paragraph 4.8.2.3.2.b.3 requires that the 250 volt battery bank shall be demonstrated operable at least once per 92 days by verifying that the electrolyte level of each connected cell is between the minimum and maximum level indication marks.
The inspector discussed electrolyte specific gravity adjustment with the Maintenance Superintendent, identifying that reducing the electrolyte level below the minimum indication mark apparently makes the battery
"inoperable".
The next day the Maintenance Superintendent informed the inspector that Condition Report 2-03-85-431 had been written, a Job Order had been written to verify electrolyte level, and the job had been secured pending resolution of the Condition Report.
At the exit interview the inspector discussed this item, noting that due to the short duration of specific gravity adjustment the licensee appeared to be in accidental com-pliance, not planned compliance, with the limiting condition for operation concerning battery operability.
On March 8, 1985 the inspector observed a technician calibrating a tempera-ture and a pressure switch for the Unit 2 diesel-driven fire pump.
The procedures (12 THP 6030 INP.026 and INP.028) were not present at the job site, nor were they required so.
Selected data is required to be recorded in satisfaction of these procedures, however, and the technician involved was entering this data on a "used" (i.e. previously completed)
data sheet.
As this was considered a questionable practice, it was discussed with Technical-Engineering management, and at the Management Interview.
No items of noncompliance or deviations were identified.
Monthl Surveillance Observation The inspector reviewed Technical Specifications required surveillance testing on the systems listed below and verified that testing was per-formed in accordance with adequate procedures, that test instrumentation was calibrated, that limiting conditions for operation were met, that removal and restoration of the affected components were accomplished, that test results conformed with Technical Specifications and procedure requirements and were reviewed by personnel other than the individual
directing the test, and that deficiencies identified during the testing were properly reviewed and resolved by appropriate management personnel.
The following surveillance activities were observed/reviewed.
>>>>2 MHP 4030 STP.035 Quarterly Surveillance Test Procedure for Plant 2AB Battery
>>>>1-OHP 4030 STP.007
>>>>12-MHP 4030 STP.013 Containment Spray System Operability Test Maintenance Surveillance Test Procedure for Inspection of 250 Volt Plant Batteries 1AB, 1CD Ec 2CD
>>>>2 MHP 4030 STP.036 Meekly Surveillance Test Procedure for Plant 2AB Battery
>>>>'12 THP 4030 STP.362
>>>>'12 THP 4030 STP.056 Incore-Excore Detector Calibration Radiological Chemistry Laboratory Area Monitor (R-3) Surveillance In following up station battery electrolyte level control, after the events discussed in Paragraph 4 above, the inspector reviewed completed surveillance data sheets (STP.013 atachments 1 and 2) and found the weekly 'GD" battery pilot cell readings for October 31, 1984 to January 23, 1985 documented a number of examples where the electrolyte level of pilot cells 111 and 132 were found to be above the maximum indication line and no corrective action was taken.
The inspector discussed this with the Maintenance Superintendent, identifying that finding the electro-lyte level above the maximum indication mark on the "1CD" battery and not taking corrective action is a violation of Technical Specification 4.8.2.3.2.a.l.
Subsequently, the Quality Control Supervisor provided the inspector with a copy of NSDRC Audit 114 dated February ll, 1985.
Finding 1 documents that STP.013 does not identify electrolyte level as an acceptance criterion.
The licensee corrective action was to update STP.013 and
.036 by April 15, 1985.
At the exit, interview the inspector discussed this item and stated, in accordance with NRC enforcement policy, an item of noncompliance was not being issued because this was identified by, the licensee, it fit a Severity Level IV or V, and apparently will be corrected within a reasonable time.
The inspector considers this an Unresolved Item pending completion of the licensee's corrective action and inspector review of completed STP.013 and
.036.
(Unresolved Item 315/85009-02; 316/85009-01)
The inspector reviewed STP.362 and found that Step 5.2 requires recording the time Delta I is out of the target band on form No. 362-2.
Alternately, P-250 computer address U9909 may be used to record the time out of the target band.
This should be kept below 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> per Technical Specifica-tion 3.2.1.a.2.b (not applicable below 50% rated thermal power).
The
inspector discussed this with the Nuclear Engineer, noting that Technical Specification 3.2.1.a.2.b is also not applicable above 90K, power.
The Nuclear Engineer agreed to make appropriate changes, which were completed prior to the conclusion of the inspection.
No items of noncompliance or deviations were identified.
6.
Licensee Event Re orts Through direct observation, discussions with licensee personnel, and review of records, the following event reports were reviewed to determine that reportability requirements were fulfilled, immediate corrective action was accomplished, and corrective action to prevent recurrence had been accomplished in accordance with Technical Specifications.
The follow-ing LERs are considered closed.
Unit
RO 82-050/03L-0 and RO 82-025/03L-0 A containment isolation valve for the containment air particulate detector would not close due to a defective solenoid.
RO 82"097/03L-0
"W" RHR pump made inoperable to repair oil leakage.
RO 83-004/03L-0 Containment air recirculation fan breaker breaker tripped during a surveillance test; event could not be duplicated.
RO 83-023/03L"0
"A/B" diesel generator was rendered inoper-able due to a failed inverter.
The inverter failure was caused by high ambient room temperatures due to improperly positioned ventilation dampers.
RO 83-055/03L-0 Rev.
6
"W" motor driven auxiliary feedwater pump was made inoperable to repair low suction pressure switch.
RO 83-086/03L-0 The low pressure CO-2 fire protection system was made inoperable to repair a
pilot valve on the CO-2 hose reel header system.
RO 84-015-0 Failure to complete daily survey of spent fuel storage area with monitor 12 R"5 inoperable.
Radiation Protection Section practices have been upgraded to include logging and posting of special "Action Statement" survey requirement RO 84-018-0 Reactor trip and safety injection on loss of control room instrument distribution (GRID) No. IV.
Plant response to the trip and SI was reviewed in Inspection Report 50-315/84-15.
Surveillance procedures for eyewash station testing were revised to preclude water leakage such as contributed to this event.
RO 84"023-0
>iissed steam generator blowdown composite sample, with the blowdown system out of service for repairs such that a sample could not be collected, the sample already taken for other purposes proved too small for use in compositing.
The minimum sample volume for all samples was increased.
RO 84-031-0 A SRO re-initialized the plant computer, not recognizing this caused both the computer and AFD monitor alarm to be inoperable for about 12 minutes.
Thus, hourly monitoring of indicated AFD for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter was not performed.
Operator training was supplemented with a module on plant computer "operability".
Unit 2 RO 81-061/03L-0 and RO 82-005/03L-0 Channel number 1 steam flow for steam generator number 3 failed low due to air trapped in the sensing line.
RO 83-011/03L-0 A pressurizer power operated relief valve closed during surveillance testing due to low control air pressure.
RO 84-004-0 and Rev
Surveillance of ice-condenser baskets during a refueling outage identified some baskets averaging below 1220 pounds/basket.
The accessible baskets were emptied, refilled with an average net ice addition of 300 pounds/basket, and reweighed prior to return to service from the outage.
RO 84"020-0 Reactor trip on loss of control room instru-ment distribution (GRID) No. II.
This event was previously reviewed in Inspection Report, 50"316/84"1 RO 84"021"0 Containment samples were being collected via valves permitted for use on U-1 but prohibited for U-2.
Alternate sampling (local) is being performed while a change to U-2 Technical Specifications is sought.
RO 84-028"0 Erroneous assumptions in auxiliary feedwater hydraulic analysis.
The licensee reduced power on both units commensurate with the discovered reduction in assumed AFW flow pending NRC review.
Full power was restored on issuance of the NRC Safety Evaluation Report, validating system adequacy.
See Inspection Reports 50-315/84-20 and 316/84-22.
RO 84-033"0 Safety injection pump ventilation ductwork found obstructed.
No determination could be made as to the duration of the conditon.
The obstruction was cleared and operator tour procedures modified to include airflow checks at least daily.
RO 84-035-0 Erosion of body-to-bonnet studs on RTD manifold isolation valves.
This was a
voluntary report documenting findings similar to (but in smaller lines than)
those addressed in IE Bulletin 82-02.
New carbon steel studs were installed pending replacement with stainless steel studs under safety Design Change RFC 12-2718 planned for the next refueling outage.
No items of noncompliance or deviations were identified.
7.
IE Circular Findin s For the Circulars listed below, the inspector verified that the Circulars were reviewed by licensee management, that a review for applicability was performed; and, if a Circular was applicable, that appropriate action was taken.
Units 1 and
IEC 81-01
"Design Problems Involving Honeywell Push Button Switches."
IEC 81-02
"Performance of NRC Licensed Individuals while on Duty."
IEC 81-03
"Inoperable Seismic Monitoring Equipment."
IEC 81-04
"The Role of the Shift Technical Advisor and Importance of Reporting Operational Events."
IEC 81-05 IEC 81"06
"Self-Aligning Rod and Bushing For Pipe Supports."
"Potential Deficiency Affecting Certain Foxboro 10 to 50 Milliampere Transmitters."
IEC 81"07 IEC 81-08 IEC 81-09
"Control of Radioactivity Contaminated Material."
"Foundation Materials."
"Containment Effluent Water That Bypasses Radioactivity Monitors."
IEC 81-10
"Steam Voiding in the Reactor Coolant System During Decay Heat Removal Cooldown."
IEC 81"11
"Inadequate Decay Heat Removal." Not applicable to PWR's.
IEC 81"12
"Inadequate Periodic Test Procedure of PWR Protection System."
IEC 81-13
"Torque Switch Electrical Bypass Circuit for Safeguarding Service Valve Motors."
IEC 78-08
"Environmental Qualification of Safety Related Equipment."
All actions were transfered to IE Bulletin 79-01.
No items of noncompliance or deviations were identified.
8.
IE Bulletin Followu IEB 82-01 and Revision
"Alteration of Radiographs of Welds in Piping Assemblies."
The actions of the Bulletin were applicable only to construction sites, except for Revision 2 items.
Revision 2 remains open.
9.
No items of noncompliance or deviations were identified.
NUREG-0737 TMI 0 en Items Status a.
(Closed)
Items I.C.1.2.B and I.C.1.3 '
originally addressed emergency procedure revisions for inadequate core cooling and for transients and accidents, respectively.
Both items were subsequently incorpor-ated into an Emergency Operating Procedure (EOP) Upgrade Program pursuant to Generic Letter 82-33.
As indicated in the licensee's letter AEP:NRC:0773E dated February 10, 1984 the subject procedures
'I I
II 1>
II
have been redrafted.
This was accomplished by adaptation of Owners Group guidelines prepared by Westinghouse into plant-specific proce-
'dures.
Verification and validation have also been completed, though comments derived therefrom have not all been resolved.
Operator training on the redrafted procedures has begun.
Pursuant to the licensee's letter referenced above, completion of training and full implementation of approved revised procedures is scheduled by January 1986.
b.
(Closed)
Item II.F.2.3.B involves reactor vessel level measurement system implementation.
The system is installed but, as addressed in the licensee's letter AEP:NRC:0761C dated August 20, 1984, will be slightly modified by relocation of associated wide-range pressure transducers outside containment during the upcoming outages in 1985.
Further, full "operability" of the system, as established in Generic Letter 82-28, is conditioned on completion of the EOP Upgrade Program discussed above.
The slight system modifications contemplated are expected to result. in minor changes to EOP instructions (setpoint guides)
and must be addressed, if required, in completion of that program by January 1986.
c>>
(Closed)
Item II.K.3.5.B addressed automatic reactor coolant pump trip systems for accident conditions.
Subsequently, Generic Letter 83-10 addressed acceptable criteria for exemption from this item, substituting operator manual trip instead.
The licensee's letter AEP:NRC:0785A dated May 30, 1984 requested exemption pursuant to the Generic Letter, and referenced Owner's Group documents OG-110 dated December 1,
1983 and OG-117 dated March 12, 1984.
Determination whether the submittals satisfactorily establish acceptable conditions for substituting manual trip for all owner's group plants is being done within the NRC Office of Nuclear Reactor Regulation.
Meanwhile, the manual trip criteria and instructions also must be included in the EOP Upgrade Program, to be completed and implemented by January 1986.
The various NUREG-0737 Action Items discussed above have all either evolved into or become a necessary part of the EOP Upgrade Program.
The items are therefore considered closed; but, for tracking purposes, a single new Open Item (for each Unit) will be substituted.
That is, completion of the EOP Upgrade Program, including appropriate instructions for reactor vessel level instruments in assessing core cooling adequacy, and criteria for manual RCP trip, is considered an Open Item.
(315/85009-03; 316/85009-02).
No items of noncompliance or deviations were identified.
10.
Hana ement Heetin
- Re ulator Performance Im rovement Pro ram RPIP A Management meeting was held on March 5, 1985 at the Donald C.
Cook Plant site to update the status of the RPIP, as signed on February 23, 1984 (AEP:NRC:0625F).
Attendance at the meeting was as described in Paragraph l.a above.
Mr. Alexich provided an introduction which included informa-tion on general facility safety and productivity achievements; and discussed plans for continued investment in the plants (including construc-tion of a new training facility with a plant-specific simulator)
as an indication of corporate intent in keeping the plant's performance at a
high level.
Mr. Barrett then presented the current status of the RPIP items contained in Appendix C to the February 23, 1984 letter.
Action on several of these items has been completed, and most are nearing completion.
NRC staff questions on a few items were addressed satisfactorily by the licensee.
Additional licensee/NRC items of interest, not specifically covered by the RPIP, were also discussed.
No items of noncompliance or deviations were identified.
11.
~Oen Items Open items are matters which have been discussed with the licensee, which will be reviewed further by the inspector, and which involve some action on the part of the NRC or licensee or both.
Open items disclosed during the inspections are discussed in Paragraph 9, above.
Unresolved items Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, items of noncompli-ance, or deviations.
Unresolved items are discussed in Paragraphs 3 and 5.
13.
Mana ement Heetin
"The inspector met with the licensee representatives (denoted in Paragraph 1.b) throughout the inspection and at the conclusion of the inspection on Harch 19, 1985 and summarized the scope and findings of the inspection.
More specifically:
a.
Previously identified matters to be "closed" on the basis of this inspection were reviewed (Paragraph 2).
b.
The unresolved item concerning actions taken with monitor SFR-401 apparently inoperable was discussed (Paragraph 3).
c.
Station battery electrolyte level control during maintenance (Para-graph 4) and surveillance (Paragraph 5) was discussed, certain surveillance questions being identified as an unresolved item.
d.
Licensee practices in recording data in the field during testing were reviewed.
The licensee indicated steps were underway to address the inspector's concern (Paragraph 4)
~
e.
Licensee Event Reports (Paragraph 6), I.E. Circulars (Paragraph 7)
and I.E. Bulletins (Paragraph 8) to be "closed" on the basis of this inspection were specifically identified.
f.
NUREG-0737 (Three Nile Island) items reviewed and "closed" were iden-tified by the inspector, with the "open item" created to track ongoing activities noted (Paragraph 9).
The inspector asked those in attendance at the meeting whether they con-sidered any of the matters discussed to contain proprietary information or other information exempt from disclsure.
No such information was identified.
I