IR 05000220/1979021
| ML17053B216 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 09/10/1979 |
| From: | Architzel R, Briggs L, Markowski R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17053B209 | List: |
| References | |
| 50-220-79-21, NUDOCS 7911230028 | |
| Download: ML17053B216 (32) | |
Text
U.
S.
NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT
REGION I
Report No.
50-220/79-21 Docket No.
50-220 License No.
DPR-63 Priority Category Licensee:
Nia ara Mohawk Power Cor oration 300 Erie Boulevard West S racuse New York 13202 Facility Name:
Nine Mile Point Station Unit 1
~
Inspection At:
Scriba New York Inspection'onducted:
June 11-15 nd Jul 16-20 1979 Inspectors:
R. Archstzel, Reactor Inspector
.7u~uf
. Markowski, Reactor Inspector, June 11-15 only eeA~5 L., Briggs, Reactor Inspector, June 11-15 only Approved by:
C. < ~~
3c-
. E.
C.
McCabe, Jr., Chief, Reactor Projects Section No. 2, R08NS Branch date e@b~
date el mhs date
~l(c la9, date Ins ection Summar
Ins ection on June ll-l5 and Jul 16-20 1979 (Re ort No. 50-220/79-2l)
Areas Ins ected:
Routine, unannounced inspection by three regional based inspectors 135 hours0.00156 days <br />0.0375 hours <br />2.232143e-4 weeks <br />5.13675e-5 months <br />) to review actions taken by Niagara Mohawk Power Cor-poration in response to the nuclear incident at Three Mile Island (TMI), to verify compliance with facility Technical Specifications and to assure that factors contributing to the incident at TMI do not exist at Nine Mile Point.
In addition, a portion of the Containment Integrated Leak Rate Test and licensee actions relating to IE Bulletin 79-02 (Concrete Expansion Anchor Bolts) were observed.
~Nli:1 (if
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1 1;lf piping and instrumentation drawings inaccurate; Infraction - valves not labeled; and, Infraction - valve lineup check not accurate, all Paragraph 5.c).
Region I Form 12 (Rev. April 77)
~ 9~~ gs002
DETAILS l.
Persons Contacted The following technical and supervisory level personnel were contacted:
J. Aldrich, Supervisor, Training T.
Lempges, General Superintendent, Nuclear Generation
"D. Palmer, equality Control - Operations, Supervisor
'T. Perkins,'Station Superintendent M. Sill'iman, Results Superintendent
~C. Stuart, Supervisor - Operations Other licensee employees were contacted during the inspection.
These included engineering personnel, administrative personnel, reactor operators, shift supervision, health physics personnel, and security personnel.
"denotes those present at the exit interview.
2.
Licensee Action on Previous Ins ection Findin s
(Closed)
Followup Item (79-14-01):
The inspector verified that formal training pursuant to IEB 79-08 was completed prior to plant restart.
(Closed)
Unresolved Item (79-14-03):
The licensee issued Licensee Event Report 79-07 on May 18, 1979.
NRC evaluation of the licensee's response to IE Bulletin 79-07 is continuing.
NRC review/action in this area will be documented in a future NRC inspection report.
3.
IE Bulletin 79-02 Pi e
Su ort Base Plate Desi n Usin Concrete The inspector had previously inspected some of the licensees action on this Bulletin (NRC Inspection Report 220/79-14).
As a result of the test program initiated pursuant to IEB 79-02, the following types of deficiencies were identified relative to the subject bolts:
a.
Bolts loose, missing or cut and welded to the under side of plates.
b.
Oversized plate holes.
c.
Skewed bolts.
d.
Other bolts and plate installation deviations.
e.
Missing seismic restraint P
Based on the results of the test program, the licensee submitted LER 79-12 in accordance with TS 6.9.2(a)(9).
This LER addressed the deficiencies and described NHPC's course of correctiv'e action.
The licensee committed to take the following actions prior to plant startup:
a.
Repair all deficiencies in areas which are inaccessible during plant operation.
b.
Concrete anchor bolts on inaccessible base plates to be pull tested to assure adequate factors of safety exist in accordance with Bulletin 79-02.
c.
Repair all deficiencies in the Core Spray, Emergency Condenser, and Control Rod Drive Systems.
d.
Seismic restraints to be added to the Containment Spray System to assure stresses are within allowable values.
In addition, before December 31, 1979, Niagara Hohawk will complete the inspection, testing and repair program for all concrete anchor bolts on Class I seismic piping systems.
The licensee based justification of continued operation upon the above delineated actions, substantial margins incorporated in both the original base plate and piping system designs, and the licensee's belief that flexible plate effects had been overstated in IEB 79-02.
The inspector reviewed the status of the anchor bolt testing program, as of June 14, 1979.
Restraints
~ld
, Anchors Tensioned Anchor Repairs Passed Fai led Not Tensioned To Be Done 252 308
855 120 In addition, 841 anchors required washer installation due to oversized burn holes for the bolts.
The "repairs to be done" number includes those bolts not installed, welded to plate and cut, missing, pulled loose during nut removal, etc.
The inspector reviewed procedure (CB8I) ATR1N, Revision 0, Pipe Restraint Anchorage Testing and Repair Procedure, dated May ll, 1979.
In addition, the inspector reviewed selected test reports and the Nonconformance Control List generated as a result of the testing program to dat During the plant tours to inspect valve positions,.the inspectors noted that the anchor bolts in the supports for the Emergency Condensor Condensate Return Isolation Valves had apparently not been tested, nor had U-clamps been installed on makeup water piping (for the Emergency Condensor System).
The licensee stated that these would be tested/repaired prior to startup.
This item is unresolved (220/79-21-01).
IE Bulletin 79-02 remains open pending completion of the licensee's program and reinspection by the NRC.
4.
Containment Inte rated Leak Rate Test CILRT conducted following receipt of the final report by t ceptable conditions were identified.
5.
Ins ection of En ineered Safet Features (IEB 79-08)
a.
General I
An inoffice and onsite inspection of Engineered Safety Features was conducted which encompassed the following areas:
A review of System Operating Procedures to determine that valve/ breaker/switch alignment was specified; and was consistent with piping and instrument diagrams; Direct observation of valve positions utilizing the valve check list (Table I) of the appropriate operating procedure; Direct observation of major component power supplies, on a sampling basis, utilizing the check list (Table II) of the appropriate operating procedure; Review of procedures to determine that "return to service" provisions, subsequent to maintenance and testing, surveillance testing and extended outages, are defined; and, During this inspection the inspector observed portions of the conduct of the CILRT.
On June 12, 1979, the initial testing attempt was declared a
failure.
The previous CILRT conducted at Nine Mile Point in November, 1975, initially failed (refer to NRC Inspection Report 220/79-,03).
Since the present test also initially failed 10 CFR 50, Appendix J, Section III.A.6.b requires that a CILRT be run each refueling outage until two consecutive tests are successful.
The licensee acknowledged the inspector's comment.
The licensee located the source of the leak (Containment Spray-Raw Mater Intertie Blocking Valves) and repaired their seating surfaces while at test pressure.
The proper lineup was reestablished and the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> CILRT successfully completed.
LER 79-17 was submitted. describing the test failure, however, final review of CILRT test results will be he NRC.
No unac-
Review of surveillance test results.
The findings in the above areas are discussed in the following paragraphs.
b.
Review of 0 eratin Procedures and Pi in and Instrument Dia rams PAID s The below,listed procedures were initially reviewed against the referenced P&ID s (identified in parenthesis)
in the Region I office.
The review was conducted to verify, on a sampling basis, that the Valve Check List (Table 1) and the Major Component Power Supply List (Table 2) and/or the procedure body specified the required operating valve/ breaker/switch alignment for the system as depicted on the PAID'.
The procedures (and/or associated drawings)
reviewed were:
NI-OP-02, Core Spray System, Revision 4 (C-18007-C, Reactor Core Spray, Revision 9).
NI-OP-04, Shutdown Cooling and Head Spray, Revision 3 (C-18018-C, Reactor Shutdown Cooling, Revision 6).
NI-OP-05, Control Rod Drive, Revision 7 (C-.18016-C, Control Rod Drive, Revision 3).
NI-OP-10, Reactor Building Heating, Cooling and Ventilation System, Revision 2, (C-18013-C, Reactor Building Heating, Cooling and Ventilation Systems, Revision 5).
NI-OP-12, Liquid Poison System, Revision 5 (C-18019-C, Reactor Liquid Poison System, Revision 6).
NI-OP-13, Emergency Cooling System, Revision 5 (C-18017-C, Emergency Cooling System, Revision 7).
NI-OP-14, Containment Spray System, Revision 5 (C-18012-C, Reactor Containment Spray System, Revision 7).
. NI-OP-45, Emergency Diesel Generators, Revision 3 (B-18040-C, Fuel Oil Handling and Storage for Emergency Diesel Generator, Revision 2; Figures OP-45-1 and 2, Diesel Generator Air Start System; and Figures OP-45-3 and 4, Diesel Generator Cooling Water).
The inspectors'indings in this area are incorporated in the following paragraph.
Direct Observation of Valve and Breaker Ali nment The inspectors toured plant areas at various times during this inspection to directly observe the status of valves and breakers.
The Valve Check Lists of the appropriate system opei ating procedures referenced in S.b above were utilized.
Findings in this area are delineated on a per system basis in the subparagraphs below.
(1)
Emer enc Coolin S stem Two check valves shown installed on the loop
and
vent lines to the Main Steam system labeled SS-C-4 between valves IV-05-02 and 03 were not installed.
Two manual blocking valves were installed on this line.
They were not listed on She valve check."-off list nor on the P8ID.
Drain valves EC 701., 703, 705, and 707 required to be locked closed on the VCOL (valve checkoff list) were closed but not locked.
The drain valve configuration on the reactor side of IV's 39-09 and 10 was not as built on the P8ID (2 valves installed, 1 shown, neither on VCOL).
Valve FS-45 (Emergency Condensor Makeup Tank Supply from Fire Protection System)
was labeled FS-44.
The following valves were not incorporated in the VCOL:
FCV 39-15 and 16, Steam Line Drain Path Flow Control Valves EC 748, 749, 751, and 752 (Four Drain Valves)
Emergency Condensor's overflow loop seal vent and drain valves FS 44, 45, EC Makeup Mater Tank Fire Protection System Supplies Table II (component power supplies) did not incorporate checks f'r:
IV 05-01 through 04, Steam Line Vents IV 39-11 through 14, Steam Line Drains (2)
Li uid Poison S stem The "normal" position of LP 711 and LP 712 (utilized during monthly flow test) were not specified in the body of NI-OP-12 no)
an. the valve check list.
The power supply for heat tracing was not included in Table II.
P8ID, C-18019-C, Revision 6, April 6, 1978, did not show the flow switch upstream of LP 12.
Two valves, PAID No. 41-15 (Test Tank Drain) and the drain downstream of PAID No. 42-07 (LP 705), were assigned the same number, LP 711.
The demineralized water supply to the pumps and the test tank are assigned two different operating numbers:
PAID No.
41"03 41-12 41-13 (3)
Diesel Generators LP 15 LP 17 DM ¹13 DM ¹12 LP 16 DM ¹ll P8 ID OP No.
Check List OP No.
The position of DGA-34, Air Supply Pressure Regulating Valve, was listed as open on the VCOL.
Opening this valve could jeopardize the integrity of the air start motors.
Pressure should be regulated by the setting of DGA-34.
DG Air Valves were not labeled; DG Fuel Oil Cooling Mater Valves were labeled with manila tags (only).
(4)
Core S ra S stem A 1" vent valve was installed on piping connected to the primary containment (torus) directly (no isolation valves).
The valve was closed but not locked and the vent was not capped.
This valve was not shown on the P8 ID nor VCOL.
(Located between torus and PCV 81-,85.)
A
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CRS 15 and 16, check valves on the pump discharge headers, were shown in the incorrect position on the P&ID.
These check valves function as vacuum breakers following system operation allowing containment atmosphere to. replace water in the discharge piping.
This flow path bypasses the system isolation valves.
The requirement to set PCV 81-85 and 86 was not specified in VCOL.
Three additional drain valves were installed on CRS-11 Bypass IV.
(Not on P&ID nor VCOL.
This discrepancy was noted by licensee in performing recent valve lineups.)
(5)
Containment S ra CTN-SP The body of operating procedure specified opening the CTN-SP and Raw Water (RW) flow rate set valves.
These are required to be locked set by the P&ID.
When inspected, the RW rate sets were locked set, the CS rate sets were set with the handwheels removed.
(CTN-SP-5, 6, 17, 20, and CTN-SP-RW-7, 8, 15, 16. )
The VCOL lists 10 drain valves per loop (four loops)
and six vent valves (in one loop seven)
and assigns specific numbers.
All 65 valves are specified as being closed.
These valves could not be directly correlated to installed valves because none were labeled in the plant.
In addition, the P&ID requires locking of some of these valves (at least 12).
None were locked.
Also, the P&ID is not as built in that eight vent yalves are shown installed on IV operator metal shafts.
The power supply for the 80-21 valve (No.
112 CTN-SP Suction IV) was not included in Table II.
(6)
Reactor Bui1din Venti1ati on The valves/dampers were not labeled.
The valving arrangement in the supply to the airborne monitor was not as built.
The P&ID showed 1 inlet isolation valve.
The VCOL listed two inlet isolation valves.
Four inlet isolation valves were installed,
each in parallel path ~ <<
(7)
Reactor Shutdown Cool in SC Pumps Nos.
11, 12, and 13, Casing Vent Valves SC-303, 304, 305, 306, 307, and 309 were required to be closed by the P8ID and the VCOL.
These valves did not exist in the system, however, were documented as checked closed when the system valve lineup was completed on June 17, 1979.
SC Pump Seal Water Vent Valve for No.
SCP (SC-757)
was not incorporated on VCOL.
'(8)
Control Rod Drive CRD CRD 10 and ll were required to be closed per the VCOL.
These are the Drive Water Filter Outlet Valves, closing both eliminates the system flow path.
"When checked, they were open.
CRD 8 and 9, the drive water filter supply valves, are shown with one open, one locked closed on the PAID and VCOL.
Both were open when checked, and in addition, two manual block valves, not shown on the PAID nor VCOL, were installed'between CRD 8 and 9 and their respective filters.
These are process flow path blocking valves.
The VCOL requires air to CRD Flow Control Valves A and B
to be isolated.
This would make the FCV's inoperable.
The air supply valves were open.
The vents on the CRD cooling, exhaust drive, and charging water headers were closed but not capped.
(9)
General Comments - Ma or Com onent Power Su lies Table II of the various system VCOL's does not check air supplies available to the various air operated valves.
The licensee stated that the VCOL's would be revised to check the air supplies.
This item is unresolved (220/79-21-02).
(10) Instrument Isolation Valves The inspector reviewed the instrument sensing lineups specified in Instrument Surveillance Procedure NI-ISP-IC-21, Revision 2.
In addition, the valve lineups were checked in the North and East Instrument Rooms.
The inspector noted that for instru-mentation and sensors not located in the instrument rooms, the
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system procedure checklists verified the positioning of isolation and root valves on a haphazard basis.
In addition, for those listed on the VCOL's, for example, various CRD system pressure and flow transmitter, the valves were not labeled and did not necessarily reflect the installed configuration.
The licensee stated that instrument isolation valving checks would be reviewed and incorporated in the VCOL's.
This item is unresolved (220/79-21-03).
Mith regard to the findings delineated above, the inspector stated that there were four items of noncompliance:
(a)
Piping and Instrument Drawings did'ot accurately reflect the as built plant (220/79-21-04).
(b)
Valves required to be locked by system operating and PAID's not locked (220/79-21-05).
(c)
Valves documented as closed but not installed (220/79-21-06).
(d)
Valves not labeled as required (220/79-21-07).
Additional comments made regarding these valve lineups are unresolved (220/79-21-08)
pending appropriate licensee actions.
d.
Review of Surveillance Test Procedures and Results The below listed surveillance procedures and associated data sheets were reviewed to verify:
Return to service provisions were incorporated which assure that the system will be returned to its standby/automatic initiation configuration if the test is performed during normal operation; and, The data sheet, associated with those tests performed subsquent to March 2, 1979, provided documentary evidence that the test was performed satisfactorily and the system was retured to normal.
It is noted that with respect to the data sheet review, the inspector utilized the acceptance criteria defined within the appropriate procedure and verified the technical adequacy of the procedure, the frequency of performance of the surveillance and the consistency of the acceptance criteria with respect to the Technical Specification requirements, on a sampling basi The surveillance procedures and data sheets reviewed were (" indicates procedure review only):
NI-ST-M1, Liquid Poison Pumps Flow Rate and Discharge Pressure Check, June 7, 1979; NI-ST-M2, Emergency Cooling System Makeup Tanks Level Control Valves Exercise, June 7, 1979; NI-ST-M3, Suppression Chamber to Drywell Vacuum Relief Valve Exercise, June 7, 1979; NI-ST-M4, Emergency Diesel Generator Manual Start and 1 Hour Rated Load Test, June 8, 1979; NI-ST-M8, Emergency Ventilation System Operability Test, June 8,
1979; NI-ST-gl, Core Spray Pumps and Motor Operated Valves Operability Test, May 20, 1979; NI-ST-(2, Control Rod Drive Pumps Flow Rate Test, May 21, 1979; NI-ST-(3, High Pressure Coolant Injection Pump Operability Test, Revision 2; NI-ST-(4, Reactor Coolant System Isolation Valves Exercise Test, Revision 1; NI-ST-(5, Primary Containment Isolation Valves Exercise, May 20, 1979; NI-ST-(6, Containment Spray and Raw Water Pumps Operability Test, June 5, 1979; NI-ST-(10, Suppression Chamber to Reactor Building Vacuum Breaker Valves Exercise, May 20, 1979; NI-ST-R2, Loss of Coolant and Emergency Diesel Generator Simulated Automatic Initiation Test, Revision 2; NI-ST-R3, Drywell to Suppression Chamber and Suppression Chamber to Reactor Building Opening Force Test, May 16, 1979; NI-ST-R5, Primary Coolant Manual Isolation Instrument Channel Test, Ma'rch 3, 1979;
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NI-ST-R6, Primary Containment Manual Isolation Instrument Channel Test, March 3, 1979; NI-ST-R7, High Radiation Emergency Cooling System Vent Monitoring Instrument Channel Test, March 4, 1979; NI-ST-R8, Reactor Coolant and Primary Containment Isolation Valve Timing Test, Revision 2; NI-ST-R9, Core Spray System Operability Test Using Demineralized Water, March 7, 1979; NI-ST-Cl, Liquid Poison System Functional Test Using Demineralized Water with Squib Valves, March 24, 1979; NI-ST-C2, Manual Opening of Solenoid Activated Pressure Relief Valves and Flow Verification, Revision 3; NI-ST-C3, Auto Startup of High Pressure Coolant Injection System, March 3, 1979; NI-ST-C4, Containment Spray Air Flow for Spray Header and Nozzles Test, June 10, 1979; NI-ST-C5, Secondary Containment Leak Rate, Reactor Building Emergency Ventilation System Operational Test, Revision 3; ISP-17, Auto Depressurization System Operability Test, June 2, 1979 ISP-18, Electromatic Relief Valve Operability Test, June 2, 1979; ISP-77.2, Recombiner Hydrogen Analyser Instrument Check and Calibration, June 4, 1979; ISP-201.2, Torus Water Level Instrument, May 4, 1979; NI-ICP-80, Containment Spray System Flow and Pressure, June 2, 1979; NI-ICP-201.2, Containment H2-02 Analyzers Check and Calibration, May 30, 1979; NI-IST-RPS-XMTR, Reactor Protection System - Transmitter Calibration, June 2, 1979.
This procedure included the following instruments:
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(1)
Main Steam Break, Instrument Nos.01-26A,through H;
(2)
Reactor Level High and Low, Instrument Nos.36-03A through D;
(3')
Low-Low Reactor Level; Instrument Nos.36-04A through D;
(4)
Low-Low-Low Reactor Level, Instrument Nos.36-05A through D;
(5)
Emergency Cooling High Flow, Instrument Nos.36-06A through D;
(6)
Reactor Pressure High and Low, Instrument Nos.36-07A through D;
(7)
High/Lo Lo Reactor Pressure, Instrument Nos.36-08A through D
~
(8)
Drywell High Pressure, Instrument Nos.
201.2-476A through D
~l NI-ISP-RD-08, High Level Scram Discharge Volume Instrument Channel Calibration, June 4, 1979; and, NI-ISP-1C-21, Pre-Startup Valve Lineup Check, Revision 2.
During the review the inspector noted that surveillance testing for the Drywell to Suppression Chamber Vacuum Breakers did not include a
physical verification (by observation) that the breakers were closed.
This item is unresolved (220/79-21-09).
d.
Review of Administrative and General 0 eratin Procedures The inspectors reviewed the below listed procedures to verify that procedural controls were specified which provide for return to service of safety systems subsequent to extended outages and performance of maintenance during. normal operations.
The procedures reviewed were:
APN No. 7, Procedure for Control of Equipment Mapkups, Etc.,
Revision'.
APN No. 8, Test and Inspection Program, Revision APN No.
13, Procedure for Control of Station Corrective Repair and Maintenance, Revision 0.
NI-OP-43, Startup and Shutdown Procedure, Revision 11.
NI-ISP-IC-21, Pre-Startup Valve Lineup Check, Revision 2.
Specifically, the inspectors observed that:
The valve lineup lists associated with each system operating procedure were identified as the record of the status of the system; Subsequent to each extended outage, a new valve lineup was required; Subsequent to each extended outage, power supplies, switch positions, and major system parameters were required to be checked; and, During the conduct of maintenance, changes to the plant configur-ation were maintained utilizing the maintenance request procedure.
The inspector noted that APN 13 did not specifically require operability testing of alternate systems prior to removal of a system from service for maintenance or repair.
All licensed operators interviewed stated that the alternate systems were always tested prior to removing a redundant system from service.
This item is unresolved (220/79-21-10)
pending revision of APN 13 to incorporate this requirement.
6.
0 erator Trainin During the weeks following the accident at TMI-2, the licensee conducted training for the licensed staff in both a reading file and formal lecture basis.
Documentation including the lecture plan has been retained by the licensee and was reviewed by the inspector.
Specific areas addressed by the licensee included:
a 0 Providing operators an awareness of the details of the Three Mile Island incident to the extent of information available at the time of this inspection.
Additionally, the licensee made'available Bulletins 79-05 and 79-05A in the control room for informational purposes.
The inspectors confirmed that copies were made ayailable to licensed personne b.
Reinstruction on specific measures which provide assurance that engineered safety features are available when required.
c.
Instructions on specific and detailed measures to assure that automatic actuations of emergency safety features are not overridden.
d.
Review of pl'ant automatic actions initiated by reset of engineered safety features that could. affect the control of radioactive liquids and gases.
e. 'ns ector Discussion with Licensed 0 erators In addition to training conducted by the licensee, the inspector held direct discussions with licensed operations personnel during day, afternoon, midnight, and training shifts (2 individuals each shift) with respect to details surrounding the events at TMI-2.
The following. topical areas were discussed:
l.
An evaluation of training received to date relative to the events at TMI.
2.
A discussion of the six specific contributing factors to the incident as described on Pages 1 and 2 of IE Bulletin 79-05/,.
3.
The seriousness and consequences of the simultaneous blocking of both auxiliary feedwater trains.
4.
The need for prompt reporting of serious events to the NRC and discussion regarding the licensee's reporting procedures.
5.
The necessity to avoid premature resetting of Engineered Safety Feature Systems, including core cooling systems and containntent isolation systems.
These discussions also encompassed resetting from spurious signals.
6.
The need to avoid premature tripping of Engineered Safety Feature Systems during any transients requiring flow.
Prom t Re ortin Re uirements Plant procedures requiring prompt notification were reviewed to ensure that provisions are established incorporating prompt (1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) notification 'any time the reactor is not in a controlled or expected condition of operation.
The procedure reviewed was APN 21, Procedure for Reporting Variations from Normal Plant Operations, Defects and Noncompliance, Revision The inspector noted that the licensee's procedure changes did not specify that a technically knowledgeable individual must be available within one hour of such a transient to maintain the line of communications with NRC.
Prior to the completion of this inspection a "Hot Line" was installed and tested directly connecting the Control Room, Emergency Communications Center (Lunch Room)
and the NRC Emergency Operations
.Center, Bethesda; Maryland.
Revision of the licensee's procedures to specify that a technically competant individual is available within one hour of an event requiring immediate notification per 3EB 79-08 is unresolved (220/79-21-ll).
7.
Unresolved'tems An item about which more information is required to determine acceptability is considered unresolved.
Paragraphs 3, 5, and 6 contain unresolved i tems.
8.
Exit Interview At the conclusion of the inspection the inspector held a meeting (see Paragraph 1 for attendees)
to discuss the inspection scope and findings.
The items of noncompliance and unresolved items were identifie nn
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