IR 05000423/1988010
| ML20207B527 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 07/14/1988 |
| From: | Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20207B524 | List: |
| References | |
| 50-423-88-10, IEIN-84-68, NUDOCS 8808020362 | |
| Download: ML20207B527 (14) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
i Report No.
50-423/88-10 i
l Oocket No.
50-42,3 l
f License No.
NpF-49 i
licensee:
Northeast Nuclear Energy Company
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ETBox 270
j Hartford, CT 06101-027,0 l
I Facility Name: Millstone Nuclear Power Station, Unit 3 i]
Inspection At: Waterford, Connecticut i
j Inspection Conducted:
May 24 - July 5,1988
i ReportinC Inspector:
G. S. Barber, Resident Inspector
j Inspector:
W. J. Reymond, Senior Resident Inspector l
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Appioved by:
M C db A. b
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j E. C. Hitabe, Chief, Reactor Prdects Section IB-
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Inspect _1on Sumary: Inspection on 5/24 - 7/5/88
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j Areas Inspected: Routine onsite inspection (115 hours0.00133 days <br />0.0319 hours <br />1.901455e-4 weeks <br />4.37575e-5 months <br />) of: Plant Oper.'tions; Status l
of Previous Inspection Findings; Failure of Control Bank "A" to Move During Sur-
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i veillance Testing; Control Building Isolation Signals Caused by Radiation Monitor
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i Maloperation; Plant Incident Reports; Allegations; Ga'tma-Metrics: Post-Accident
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Monitoring Instru entation, 10 CFR 21 Report; Licensee Event Reports; Maintenance; I
I Surveillance; and Action on Information Notice (!N) 34-68.
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Results: No violations, deviations or unsafe pisnt conditions were ident,1fied.
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Licensee conduct of a Special Procedure written to correct a malfuncti.ntrg Control l
Bank was a notable strength.
The Integrated Safety Assessment generated prior to i
performance of the Special Procedure showed a thorough review of safety aspects, i
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8800020362 000714 POR ADOCK 05000423
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TABLE OF CONTENTS l
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1.0 Persons Contacted....................................................
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2.0 S umma ry o f Fa c i l i ty A c t i v i t i e s.......................................
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3.0 Status of Previous Inspection Findings...............................
i 3.1 (Closed UNR 87-17-01) Follow Actions to Correct Items Identified i
j During Plant Tour.............................................
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3.2 Followup on Unusual Event Dec1sssification......................
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i 4.0 Review of Facility Activities........................................
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4.1 Failure of Control Bank "A" to Move During Surveillance Testing.
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4.2 Control Building Isolation Signals Caused by Radiation Monitor
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Ma1 operation..................................................
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5.0 Plant Operational Status Reviews......................................
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5.1 Pl a nt Incident Report s ( PIRs)...................................
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6.0 Physical Security....................................................
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I 7.0 A l l e g a t i o n s ( 8 7 - A- 0113 )...............................................
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7.1 SFP Lighting Fixture............................................
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7.2 Solenoid Valves Environmental Qualification.....................
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i 8.0 Gamma-Metrics: Post Accident Monitoring Instrumentation, l
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(10 CFR 21 Report)..................
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9.0 Lt ee n see Ev en t Re p o rt s ( LER s)........................................
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11.0 Surveillance.........................................................
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j 12.0 Corrections to Previous Inspection Reports...........................
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13.0 Informatien Notice (1N)
84-68........................................
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i 14.0 Management Meetings.................................................
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DETAILS 1.0 Persons Contacted Inspection findings were discussed periodically with the supervisory and man-
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agement personnel identified below-
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S. Scace, Station Superintendent l
C. Clement, Unit Superintendent, Unit 3
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M. Gentry, Engineering Supervisor
R Rothgeb, Maintenance Supervisor
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K. Burton, Staff Assistant to Unit Superintendent J. Harris, Operations Supervisor
D. McDaniel, Reactor Engineer
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R. Satchatello, Health Physics Supervisor
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M. Pearson, Operations Assistant j
2.0 Surnmary of Facility _ Activities The plant began the inspection period at full power on 8:30 a.m., May 24.
Power was decreased to 90*4 due to a seal leak on the "B" turbine driven feed
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pump. Af ter repair, full power again was rcathed at 10:00 p.m.
Power was subsequently lowered to 90*4 at 8:03 a.m., May 26 for condenser backwashing, i
and was returned to 100*o at 3:35 p.m. that same day.
Power remained at 100*4 (
until 12:12 a.m., June 4 when it was reduced to 90*4 for condenser backwash.
Full power was achieved at 4:00 p.m.
On June 25, another thermal backwash
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required a power reduction to 90*. at 5:00 a.m., June 25, with a return to full
power at 6:14 p.m.
In addition, there were 2 short duration decreases (less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) of 2*i full power to perform surveillances.
The plant continued to operate at
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full power through the end of the inspection period.
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3.0 Status of Previous Inspection Findings 3.1 (Closed UNR 87-17-01) Iters identified Duriao Plant Tour The Regional Administrator toured all three Millstone Units on August 19, 1987.
During this tour,. 7 generic and 11 unit specific items were identified. Generic issues applied to all units and have been addressed by the licensee along with the Unit Specific items. The generic items i
are sur. arized below:
Overall material condition of the plant, including housekeeping and a.
fire prevention controls was very good. An especially notable finding was that there were no instances where damaged flexible conduit was observed. Damaged pipe lagging was observed in some locations.
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The lubrication program for rotating equipment should be reviewed i
to ensure that grease does not enter the motor windings.
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The program for using calibration stickers and trouble reports
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should be reviewed and revised to be implemented consistently
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throughout all three units.
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Review the use of nails in scaffold erection to ensure they do not l
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enter rotating machinery and sumps.
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Control of radiological areas was generally good except that drip i
pockets in the "C" charging cubicle were not collecting leakage.
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Review control of pipe caps on vent and drain valves to ensure they l
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j The licensee has provided the following resolutions:
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All Millstone 3 Department Heads reiterated to their respective de-
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j partment the need to exercise care when working in equipment spaces i
j and discussed the damaged pipe lagging.
The Millstone 3 Maintenance i
j Department has 2 individuals assigned full tine to lagging.
Damaged
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lagging is repaired as assigned by supervision when these workers i
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are not supporting lagging requirements of ongoing work. Routine l
resident inspector tours have noted improved addressal of damaged i
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A history report was submitted to the computerized maintenance man-
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agement system (PMMS) and no motor failures were attributed to i
grease in the motor windings. Millstone 3 uses lubrication tech-
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I nique sheats for motor bearings and no problems were identified l
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The observed instances of overgreas-
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ing occurred prior to system turnover and have since been corrected.
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A program now exists to inspect and clean these motors, as necessary, Millstone Station now uses a standard set of Treuble Report tags (
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and stickers which are stocked at the warehouse.
These tags and i
i stickcrs are fluorescent and include the applicable Trouble Report l
or AW number. ACP-QA-2.02C was revised to clarify the use of
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Trouble Report stickers.
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Representatives from all 3 unit's !&C department met to evaluate
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a the use of cslibration stickers. A review of applicable standards
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indicated that stickers should display only the actual calibration
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i date and the identity of the person performing the calibration;
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there is no "Date Due" requirement for installed plant equipment.
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Calibration frequency, grace periods, and scheduling dates for all Millstone equipment is controlled via the Production Maintenance l
Management System (FPMS), applicable FORC approved procedures, and l
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L the Unit Technical Specifications.
Calibration is also subject to l
system or unit availability.
The use of this block for installed i
j plant equipment is, therefore, not required or desirable.
It should be noted that rieasuring and Test Equipment (M&TE) does (
require the "Date Due" block and this practice is consistent within l
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the station M&TE Lab.
The differnce in the two types of equipment
is notable.
The sticker used for M&TE equipmert will remain unchanged and the j
"Otte Out' filled in as required.
The sticker for installed plant
equipmen: has been redesigned without the "Date Due" block and vill l
be utili4ed stationwide on all future equipment calibrations.
Both
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stickers will comply with the following documents: (1) NU Quality
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Assuran:e Program Topical Report (NVQAP); (2) ANSI N45.2.4-1972/IEEE i
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Std 336-1971; (3) P.egulatory Guide 1.30; (4) ACP-QA-9.04.
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new stickerf, are on order and will be stocked in the warehouse.
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Consistent implementation has since been observed at all three units.
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d.
The station's practice of using nails for constructing scaffolding i
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in equipment spaces was reviewed and determined to require clarifi-l cation via a procedure change.
The use of bailing wire as a sub-l
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stitute for nails was evaluated with the conclusion that bailing
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wire could create more safety and potential problems than nails.
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This would, therefore, not be an acceptable reslacement for r. ails.
i Additionally, as a rule it is not practical to unstruct scaffolding
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in outside areas and carry it into the work sites. Many plant areas l
have small access / egress paths (radiologically controlled areas, vital areas, etc.) and do not facilitate large pieces of equipment, j
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The use of nails for constructing scaffolding appears to remain the
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j most feasible fastening material.
To better control the use of
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rails ACP 2.19 "Scaffolding Program" was revised to delineate good l
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construction practices with nails, hazards associated with loose l
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J nails and necessary precautions to be taken. Loose nails are now l
j specifically required to be removed after scaffolding erection.
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Metal scaffolding is now in general use, j
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The collection bags were repositioned for the items noted and the
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Maintenance Department retightened valve packing to the extent
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i possible. MP3 has a weekly management inspection by Operations and
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j Health Physics to review the control of radiological areas.
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group will continue to look for and correct collection methods which k
i have become ineffective. Routine resident inspector tours have l
identified no recurrence of this problem.
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Units 1, 2, and 3 have determined that adequate controls are in I
place to ensure leakage boundaries are maintained.
The boundary
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valves are aligned by approved station procedures. No credit is l
taken in the FSARs for Unit 2 or Unit 3 for leakage protection
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utilizing pipe caps.
The drain or vent valve forms the actual cys-i tem boundary. Additionally, any system leakage is investigated in
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l response to radwaste drain collection increases or sump alarms.
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The inspector reviewed the licensee's corrective actions and noted during inspections that housekeeping was very good, with one exception being l
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the installation of conduit covers.
This problem was identified to the
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I licensee and improved performance has been observed.
Various rotating equipment was spot checked for grease in the motor windings with negative
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findings.
No loose nails have been observed around scaffolding.
During
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the Unit 3 outage, scaffolding inside containment was metal stock fas-l
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tened with clamps, minimizing the use of nails.
The inspector frequently t
accompanied the Operations Assistant and HP Supervisor on their weekly
j tours and noted prompt correction of deficient radiological conditions.
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The use of pipe caps was spot-checked with no negative findings.
No
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inadequacies were noted.
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There were 11 specific material discrepancies identified. All items have been corrected, l.icensee actions were comprehensive and responsive to
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NRC concerns. No inadequacies were noted.
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3.2 Followup on Unusual Event Declassification i
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Region ! Inspection Report 50-423/88-08 Detail 5.3 identified that the
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licensee failed to terminate or declassify an Unusual Event.
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tation of a procedure change to EPIP 4701 has been cortpleted and Step
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i 3.3.1 now requires termination of Unusual Events by the Shift Supervisor.
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)l The inspector had no further questions on this item.
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4.0 Review of Facility Activities
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l 4.1 Failure of Control Bank "A" to Move _0uring Surveillarice Testing l
During routine surveillance testing to verify control rod movement, con-l
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trol bank "A" (a non-controlling group) failed to respond to rod motion i
i coreands at 5:00 a.m., May 26.
Prior to and during attempted movement, I
j no urgent or non-urgent failure alarms were received. The licensee
entered technical specification (TS) 3.1.3.1.c. af ter verifying that all r
rods were trippable and aligned within 12 stops of their demanded post-
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tion (they were all full out).
TS 3.1.3.1.c. required the inoperable
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rods to be returr.ed to an operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, j
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The licensee began troubleshooting activities in accordance with Auto-
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mated Work Order (AWO) M3-85-08154.
The AVO required instrumentation
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I and control (!&C) technicians to determine if the failure to move was
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the result of a bad bank select switch or movemont logic card.
The in-J spector cbserved the technicians' activities in the control rod drive logic instrument racks.
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)l The technicians (techs) used a calibrated digital volt meter (DVM) (QA (
5059 Cal Due 10/22/88) to determine high (14.0-15.0 VOC) and low (0.5-1.0
VDC) voltages at various logic test points.
By using Westinghouse i
Manuals for Rod Control, Figure 4-65, Change 2, Sheets 8 and 9, they were j
able to determine that the master cycler select card (A106) was not pro-I cessing signals as required.
Before replacing the card, the techs had
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to ensure no adverse action would result from pulling the card, if a i
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1% cal output that was at a high voltsge went to a low voltage, a block
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j or inhibit signal could have been removed, and a reactor trip could occur.
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The tech identified 10 to 12 circuits that would be directly affected.
In his review of the first circuit, he discovered that eight branch cir-
i cuits were also affected. Because of the likelihood of rany branch cir-f cuits, the techtician discussed the drawing review with his supervisor
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and other management personnel. A decision was made to contact Westing-(
house (W) to have them check the results of pulling card A106 on their l
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full scale replica of the control rod drive system. W determined that L
q no adverse action resulted when the card was replaced! The licensee
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J reviewed this report and, even though the Unit 3 systems were similar l
(90's duplicate components) decided to write a Special Procedure to re-
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place the defective card. This Special Procedure (88-3-2) placed all
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i rod control power cabinets in urgent alarm.
This condition energizes l
both the stationary and movable grippers for each rod, holiing the rods
in place.
It allows maintenance on logic cabinet circuitry and portions f
j of the power cabinet circuitry not associated with the stationary grip-j pers.
It also requires a manual reactor trip if red motion is needed.
The inspector reviewed the integrated safety evaluation written to ensure L
g that the procedure ret 10 CFR 50.59 requirements and identified no in-l l
adequacies, i
t The procedure to replace the defective card was received and approved f
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The card was re-j
placed without incident, Some additional troubleshooting was necessary
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I before rod group "A" was returned to service.
Red group "A" was returned l
to an operable status after a satisfactory retest and TS 3.3.3.1.c was i
exited prior to expiration of the 72-hour action statement. No inade-l
j quacies were noted, j
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j 4,2 Control Building, isolation Signals Caused by Radiation Monitor Malepera-l J
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The licensee reported that a Control Building Isolation (CBI) was gene-t
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rated at 12:34 a.m., June 11.
The signal was generated when a radiation I
j monitor (3HVC*RE16A) was removed f rom service to correct a lockup problem (the display failed to update the previous readirg).
All equipment re-
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spended as required to the CB!. Control room pressurization was defeated
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within the 60 second time delay af ter verification of the invalid actu-l
atton signal. While resetting the monitor's conversion factor to correct i
j the earlier probiri, a second CBI was received. The plant responded as
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designed to the second CBI.
Proper setpoints were established and the radiation monitor was returned to service.
The first CBI was the result of deenergizing the monitor to reset the lockup problem.
The voltage transient caused a high alarm which generated the CBI. The licensee has cautioned operations and maintenance personnel regarding this equipment response.
The second CB! was caused by using an incorrect conversion factor in establishing an alarm setpoint.
The need to be attentive to detail was erphasized to the personnel involved and the procedure was reviewed to ensure it adequately detailed the required actions.
No in-adequacies were noted.
The inspector has no further questions on this issue.
5.0 plant _ Operational status Reviews The inspector reviewed plant operations from the control room and reviewed the operational status of plant safety systems to verify safe cperation of the plant in accordance with the technical specifications and plant operating procedures. Actions taken to meet technical specification requiremeats when equipment was inoperable were reviewed to verify the limiting conditions for operations were eet.
Plant logs and control room indicators were reviewe.i to identify changes in plant operational status sinct the last review and to verify that changes in the status of plant equipment was properly communicated in the logs and records.
Control room instruments were observed for correla-tien between channels, proper functioning and conformance with technical specifications. Alarm conditions in effect were reviewed with control room cperators to verify proper response to of f-normal conditions and to verify operators were knowledgeable of plant status. Operators were found to be cognizant of control room indications and plant status. Control room manning and shift staffing were reviewed and compared to technical specification re-quirements.
No inadequacies were identified.
The following specific activi-ties were also addressed.
5,1 Review of Plant Incident Reports
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The plant incident reports (PIRs) listed below were reviewed during the l
inspection period to (i) determine the significance of the events; (11) review the licensee's evaluation of the events; (iii) verify the j
licensee's response and corrective actions were proper; and (ty) verify that the licensee reported the events in accordance with applicable re-quirements, The PIRs reviewed were: n eber's 1 SS dated 1/5/SS, 37 83 dated 2/19/SS, 40-S3 dated 2/22/S3, 41-SS dated 2/23/SS, 42-SS dated 2/23/SS, 63-SS dated 3/27/SS, 64-SS dated 4/4/SS, 66-SS dated 3/25/63, 67-SS dated 4/12/33, 69-SS dat,ed 4/13/SS, 70-S3 dated 4/14/13, 73-SS dated 4/14/33, 74-58 dated 4/15/33, 76-SS dated 4/14/SS, 77-SS dated 4/20/SB, 78 SS cated 4/20/SS, 80-SS dated 4/22/58, 83-83 dated 4/25/SS, S4-SS dated 4/25/88, $5-SS dated 4/25/SS, 66-SS dated 4/23/SS S7-SS dated 4/27/88, SS-SS cated 4/26/S8, 89-SS dated 4/23/SS, 90 38 dated 4/2S/SS, 91-83 dated 4/23/33, 92-SS dated 4/29/SS. The following items warranted inspector followup:
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- 65-88 dated 4/7/88, 5purious OTdi and OPdi alarms.
See Inspection Report (IR) 50-423/88-05, Detail 4.2.
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68-88 dated 4/13/S8, Turbine / Reactor Trip on low Condenser Vacuum, See l
j IR 50-423/88 05, Detail 5.1.
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71-88 dated 4/14/89, Reactor Vessel inner 0-ring failure, 5ee IR 50 423/
88-05, Detail 5.2.
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l 72-SS dated 4/14/88, RC5 Unidentified Leakage, see IR 50-423/8S-05, Oe-
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tail 5.3.
6.0 physical Security
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Selected aspects of site security were verified to be proper during inspection
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tours, including site access controls, personnel and vehicle searches, per-l sonnel monitoring, placenent of physical barriers, compensatory measures, j
guard force staffing, and response to alarms and degraded conditions.
No j
j inadequacies were identified.
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7.0 Allegations (87-A-0113)
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During a meeting with the alleger on June 8,1928 to review to status of NRC
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findings on previously identified concerns (as addressed in NRC Inspection
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Report 50-336/88-13), additional concerns were identified.
Those issues are
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j 7.1 SFp Lighting Fistures f
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This issue invohed a concern that an electrician working at the site
for the Millstone 3 outage in November 1987 received excessive or need*
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less exposure because he had to replace lamps in the spent fuel pool
underwater lighting fixtures.
This ratter was referred to the licensee
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manage-ent for follewup and evaluation on Jure 9.
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Based on inspector discussions with the licensee and reviews of the i
j Millstone 3 spent fuel pool (5FP) design and radiological conditions,
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i this concern was not substantiated. Underwater lighting for the 5FP is I
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proviced by dual lamp fixtures attached to 8' long poles that are sus-l pended from the top of the pool.
There are 18 light fixtures in the 5FP.
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j The fiatures are shielded frou spant fuel st,ored in racks on the bottom
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j of the pool by about 20 f t. of water.
Based on reviews on June 13. t,he l
inspector noted that dose rates at the edge of the pool, which contained t
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the partial core discharge from cycle 1, was less than 0.2 mrem /hr.
Con-I
tinuous exposure to such a field during a 40-hour work week would result l
l in a cumulative quarterly dose of about. 104 nRen; tiRC regulations permit l
i radiation workers to receive from 1250-3000 rRes per quarter, depending
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on previous occupational exposure.
Therefore, replacecent of these lamps l
wculd r.ot result in a substantive radiation exposure or exposure rat.e.
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l The inspector noted further that radiochemistry results for the pool l
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e water showed low activity levels of 10-4 uC1/mi beta gamma, which would require that work on pool fixtures be conducted with some precautions to w:
rol the potential spread of minor amounts of radioactive contamin-ation.
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Light bulbs cre changed out by removing the lighting fixture from the pool and working on the deck along the side of the pool. The light fix-tures use high intensity bulbs that rely on water submersion for proper cooling.
The licensee stated that, prior to the Millstone 3 refueling outage, all lights in the pool had to be replaced after the lighting
i fixtures were mistakenly energized in a dry condition, which burned out the bulbs.
l Based on the above, the inspector concluded that no excessive exposure
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could have occurred to relamp the SFP lighting fixtures.
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7.2 Solenoid Valve Environmental Qualification
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This item involved a question on the design adequacy of 300 solenoid valves used in Millstone 3 that were found to be undersized and were seneduled for replacement during the first refueling outage.
The item was referred to licensee management on June 9 for evaluation and esponse.
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Based on the information contained in the allegation, the licensee con-
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cluded the item most probably referred to actions in progress to address concerns raised by the NRC in IE Information Notice (IN) 84-68, Potential Deficiency in Improperly Rated Field Wiring to Solenoid Valves (SOVs).
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Experience at another nuclear facility had shown that field run cabling l
to certain solenoid valves had an insulation temperature rating of 144 degrees F.
In certain SOVs where the field cable terminates at coil lugs i
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mcunted inside the valve housing, the insulation could be subjected to temperatures in the 250-280 degree F range, which could degrade the in-sulation prematurely and cause insulation failure.
Licensee reviews in
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response to IN 84-68 concluded the problem potentially applied to Mill-stone 3, although no failures of the type described had occurred. The i
present status of licensee actions and evaluations were described in a NUSCO Qualification Engineering memorandum GSP-88-022 dated February 3,
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1988.
The licensee concluded that about 300 SOVs used in various safety-related and non-safety-related applications possibly had the problem described in the information notice.
Subsequent review concluded that the Target
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i Rock valves installed in the plant were not of concern, and that certain i
ASCO valves were. Of 209 ASCOs in use, 117 ASCO Type NPK valves were deemed a potential concern since they were normally energized.
Data was obtained from the valve vendor to determine the temperature rise of the coil internals for various ambient temperatures.
Initial calculations
by the architect-engineer (AE) showed the valves htJ a qualified life i
j of about 6.3 years at the elevated temperature.
The AE recommended that l
j modifications be made at the first refueling outage (starting in November j
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1987).
The initial modification considered was to terminate the field cable at a junction box outside the SOV and to install cabling with a higher temperature rating from the junction box to the valve.
Plans to modify the valves were initiated at the start of the refueling outage through the issuance of work orders.
The plans to modify the valve wiring were deferred to the second refuel-ino outage based on updated field inspection data obtained during the first refueling outage and a revised engineering evaluation (thermal life calculation PA-78-236-743-GE), which showed that the cables installed had a longer qualified life than previously calculated.
The field in-spections were conducted to identify the specific type of cable (0konite, Kerite, etc.) associated with each 50V and to inspect the condition of the wires at the termination lugs.
No signs of burning, cracking or other evidence of heat stress were observed.
The revised thermal life calculations were based on actual ambient temperatures for each valve (which were generally less than those assumed in previous calculations)
and showed each valve had a qualified life good through the end of the second refueling outage.
Subsequent licensee action on this item is tracked by Commitment 3-88-0030 dated March 5, 1988.
The licensee plans to replace the valves on a phased basis according to the calculated end of qt611fied life.
The licensee plans to upgrade the 50V coils to an NP type, which is provided from the manufacturer with a high temperature pigtail.
The inspector identified no inadequacies with the licensee's evaluations or plans to disposition the issues raised by IN 84-68. Actions to replace components based on calculated qualified life as a preventive maintenance activity is an accepted industry practice.
Based on the above, concerns regarding the safety of the plant and the adequacy of plant systems to perform intended functions were not sub-stantiated.
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8.0 Gamma-Metrics: Post-Accident Monitoring Instrumentation (10 CFR 21 Report)
The licensee notified the inspector, May 23 that solder connections on Gamma-Metrics (G-M) cable assemblies may be susceptible to moisture intrusion during a design basis accident (DBA).
On February 19, 1988, G-M made a pre-liminary 10 CFR 21 report that identified recent environmental qualification tests of a solder joint on a metal hose of a G-M cable assembly which had failed to hold pressure at elevated temperatures.
The licensee was notified of this problem by letter from G-M on February 22. G-M also provided a letter to the licensee on May 10, to include guidance for inspection of neutron flux monitor cabling and evaluation of a retrofit to provide additional sealing of the conduit at specific connections to prevent moisture intrusion.
The G-M neutron flux monitor and cabling assemblies are used to provide the operator neutron flux indication from the source range to 150% power in Post Accident Monitoring Environments (Regulat)ry Guide 1.97). Millstone 3 tech-
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nical specification (TS) 3.3.3.6 specifies the operability requirements of wide range neutron flux monitors with respect to their post-accident monitor-
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ing function.
TS 3.3.3.6 requires that at least one of two channels of Post-Accident Neutron Monitorinn ')e operable in Modes 1 through 4.
If both chan-
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nels are inoperable, one <
nel must be restored to service or e shutdown l
must be begun within 7 days.
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The licensee completed their review of the environmental qualification of the I
G-M cable assemblies on May 27 after contacting the vendor. A phone conver-i
sattun was held on May 25 between the licensee's Qualification Engineering i
(QE) vendor department and the vice president (VP) of Gamma Metrics. The
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subject of this conversation was the February 22 and May 10 letters from j
Gamma-Metrics.
This problem was identified by the vendor during recent
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qualification testing of Gamma-Metrics Neutron Monitors for BWR applications.
The root cause of the problem was identified as voiding in the solder joints i
of in-containment cable assemblies,-allowing moisture to migrate to the vari-ous cable connectors.
The voiding results in voltage discharges across the
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c insulation in these connectors.
That results in signal degradation that looks
like increased neutron flux.
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In the licensee's discussion with the vendor VP, it was noted that Gamma-l
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Metrics had changed their shop fabrication procedure for performing the pre-
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tinning of the solder joints in question.
Prior to 1984, a solder pot dip
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process was utilized, and this was the method used to fabricate the cable
assemblies tested in the original qualification testing.
The latest qualifi-cation testing (the tests that failed) was done on cable assemblies fabricated using an iron-applied tinning. The G-M VP also stated that they have compared finished solder joints fabricated using both methods, and have observed voids
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similar to the failed cable assemblies only in those samples using the iron-t applied tinning method.
The G-M VP stated that Gamma-Metrics shipped the original cable assemblies
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j for Millstone Unit 2 and Millstone Unit 3 in December of 1982, and that the
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tinning procedure definitely changed after Dec W r 1982.
Therefore, the
licensee's QE concluded that the original cables are fully qualified and operable since they are identical in form, fit, function, materials and manu-
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facture to those which were successfully qualification tested in Gamma-Metrics
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i Report No. 010, Rev. O.
The inspector reviewed the issue and had no further
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9.0 Licensee Event Reports (LERs)
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Licensee Event Reports (LERs) submitted during the report period were reviewed f
to assess LER accuracy, the adequacy of corrective actions, compliance with
J 10 CFR 50.73 reporting requirements and to determine if there were generic
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implications or if further information was required.
Selected corrective J
actions were reviewed for implementation and thoroughness.
The LERs reviewed I
were:
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LER 88-14-00, Reactor /Turbirie Trip due to Low Condenser Vacuum.
See
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Inspection Report 50-423/88-08, Detail 5.1.
LER 88-15-00, RCS Unidentified Leakage Action Statement Improperly Ter-
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minated.
See Inspection Report 50-423/88-08, Detail 5.3.
LER 88-16-00, Mode Change with Action Statement in Effect due to Person-
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nel Error (NV4 88-10-01).
This licensee-identified item was evaluated
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as being of low safety significance, appropriately reported and corrected, and not a result of inadequate corrective action en a prior violation.
Therefore, no Notice of Violation was issued.
No inadequacies were noted.
10.0 Maintenance The inspector observed and reviewed selected portions of preventive and cor-rective maintenance to verify compliance with regulations, use of administra-tive and maintenance procedures, compliance with codes and standards, proper QA/QC involvement, use of bypass jumpers and safety tags, personnel protection, and equipment alignment and retest.
The following activities were included:
Main Generator Voltage Regulator, AWO M3-88-09068
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Particulate and Gaseous Radiation Monitoring Auto Filter Repairs.
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(3HVR-P3),M2-83-09138 Fire Protection Low Pressure C02 Inlet Isolation Valve Body-to-Bonnet
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Leak, M3-88-01000 No inadequacies were noted.
11.0 Surveillance
The inspector observed portions of surveillance tests to assess performance in accordance with approved procedures and Limiting Conditions of Operation, removal and restoration of equipment, and deficiency review and resolution.
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The following tests were reviewed:
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Auxiliary Building Filter System Operability Test, SP 3614A.1
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Emergency Generator Fuel Oil Particulate Sample, SP3646B 8
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Auxiliary Feed Pump (3FWA*P1A) Operational Readiness Test, SP3622.1
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No inadequacies were noted.
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12.0 Corrections to previous Inspection Reports Region I Inspection Report 50-423/88-02, Detail 6.1 (pg. 9) should read,
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"no inadequacies were noted" vice "inadequacies were noted."
Region I Inspection Report 50-423/88-02, Detail 7.0 (pg. 10) should read,
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"which was due on January 12, 1988" vice "which was due on January 10, 1988."
Region I Inspection Report 50-423/88-08, Detail 7.0, LER 88-12-00, should
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read "NV4 88-08-02" vice "NC4 88-08-02."
13.0 information Notice (IN) 84-68
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The inspector reviewed licensee actions taken to address the issues raised in IE Inforn.ation Notice (IN) 84-68, Potential Deficiency in Improperly Rated Field Wiring to Solenoid Valves.
The review was performed to determine whether the licensee actions taken or planned to address the NRC concerns were acceptable.
Licensee actions are discussed in Section 7.2 above and were found satisfactory. No inadequacies were identified.
14.0 Management Meetings Periodic meetings were held with station management to discuss inspection findings during the inspection period. A summary of findings was also dis-cussed at the conclusion of the inspection.
No proprietary information was covered within the scope of the inspection.
No written material was given to the licensee during the inspection period.
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