IR 05000423/1990019
| ML20058J121 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 11/14/1990 |
| From: | Haverkamp D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20058J115 | List: |
| References | |
| 50-423-90-19, NUDOCS 9011270242 | |
| Download: ML20058J121 (25) | |
Text
-'
j
,
.o
.
i U.S NUCL!AR REGULATORY COMMISSION
REGION I
Report No.:
50-423/90-19 Docket No.:
50-423 License No.
NPF-49 Licensee:
Northeast Nuclear Energy Company PTOT Box 27D ffa~rtford, fonnecticut 06141-0270 Facility Name: Millstone Nuclear Power Station, Unit No. 3 Inspection at:
Water' rd, Connecticut
- Inspection Conducted:
September 5 through October 15, 1990
'
)
Reporting Inspector:
Kenneth S. Kolaczyk, Resident Inspector, Millstone 3 Inspectors:
William J, Raymond, Millstone Senior Resident Inspector Kenneth S. Kolaczyk, Resident Inspector, Millstone 3 David H. Jaffe, Project Manager, Project Directorate 1-4, Office of Nuclear Reactor Regulation
<
//ddr W[- )
/t /
fo Donald R. Haverkamp, Ch e ate
~
Reactor Projects Section 4A Division of Reactor Projects Inspection Summary; Inspection on September 5, through October 15, 1990 (Inspection Report No. 50-423/90-19)
Arear, Ins ected:
Routine onsite inspection at Millstone 3 during normal and l
backshift work periods of operational safety including plant operations, radiological and chemistry controls, and security; maintenance and surveillance; engineering and technical support; and safety assessment and quality verification.
Results:
See Executive Summary i
co3i;/oro; 901114 pg ADOCK 0500
I.,
,
f
'E EXECUTIVE SUMMARY MILLSTONE NUCLEAR POWER STATION, UNIT NO. 3 NRC REGION I INSPECTION REPORT NO. 50-423/90-19 Mant Operations On September 14 and October 4, operators quickly recognized degrading condi-tions at the intake structure and rapidly downpowered the plant preventing a i
possible reactor trip.
In response to a violation (90-08-01), which concerned a failure to follow procedures, during meetings with the operating shift crews, the unit director and operations manager stressed management support for pro-cedural adherence. The meetings were constructive and judged to be acceptable,
'
however, there continues to be a concern regarding procedure.dherence as discussed in section 3.5 and the violation remained open. Another violation (88-23-01), which concerned establishing appropriate corrective actions when fire doors are blocked open for maintenance, is closed.
Radiological /ChemistryControh No significant observations were noted during the report period.
j Security No significant observations were noted during the report period.
Maintenance / Surveillance Inattention to detail by maintenance personnel resulted in a temporary loss of reactor coolant pump seal injection and a spill of 800-1000 gallons of reactor coolant to the auxiliary building, froubleshooting activities conducted on a failed control room chille'. breaker by electrical maintenance personnel were systematic and thorough.
Engineering / Technical Support The licensee has elected to foclow the guidance contained in Generic Letter 90-05 regarding temporary repairs to ASME Class III service water piping. An unresolved item (89-04-01), which tracked NRC
!
development of guidance on this issue, is closed.
However, a concern was generated in the methodology for repair of a failed bellows expansion joint. New fuel receipt and inspection was conducted in a safe and controlled manner during the report period,
y
.
-
- _ -
.
.
-
.
' ' '
s
- j
.
- - -
/
'-
. e; s
,
?
!
Safety Assessment / Quality Verification
A licensee-identified violation was cited regarding the failure to
.
post a compensatory fire satch when a fire barrier was taken out of service. Enforcement discrotion was not exercised as a result of the recurring violation.~ Poor coi.munications between fire watch personnel
.
have resulted in fire protection program lapses. A concern regarding
!
the adequacy of testing, verification and development of QA Category I computer software at Millstone was not substantiated.
Inspector and
!
-licensee review of the issue found that there are adequate procedural i
controis in place to ensure' Category I computer sof tware is tested to i
ensure it will perform its design function.
,
f
!
I p
!
~
.-
i
's c
-,
+-
!
>
i i i
' k Is-I
~'
l-
-
-mWe'--
y
-
emy-Tm
- Ty wyp i+M
-'e
i' *
,
,
,
r TABLE OF CONTENTS Page 1.0 Plant Operations Summary................................
I 2.0 NRC Inspection Summary..................................
3.0 Operational / Safety Verification (IP 71707/93702)*.......
3.1 Control Room Observations..........................
3.2 Plant Operations Tours.............................
3.3 Radiological Activities............................
3.4 Security...........................................
3.5 Review of Plant Incident Reports (PIRs)............
3.6 (0 pen) Violation 90-08-01:
Failure to Follow Procedures During Heatup/Cooldown of Plant.........
3.7 Control Room Chiller Breaker Failure...............
i 3.8 (Closed) Violation 88-23-01: Failure to Establish
'
Corrective Measures When Fire Doors Are Blocked i
Open for Maintenance...............................
4.0 Maintenance (IP61726)..................................
4.1 Observation of Maintenance Activities..............
4.2 Spill of Contaminated Fluids During Filter Changeout..........................................
5.0 Surveillance (IP 62703).................................
5.1 Observation of Surveillance Activities.............
6.0 Engineering / Technical Support (IP 90712/92701/60502C)...
6.1 (Closed) Unresolved Item 89-04-02 Temporary Non-Code Repairs to ASME Class III Piping...........
6.2 New Fuel Receipt Inspection........................
6.3 10 CFR-21 Report Submitted Concerning Veritrah j
Transmitters.......................................
i
.
iv l
t l
,
A
b
- ,
c>. 8u
!
e-c
, -. e-i t
..
, y -
.- -
s:
'
.i
,
p 7.0 Safety Assessment /0uality Verification (IP 71707//90712/92700)................................
7'1 Committee. Activities...............................
c
.
F 7.2 Periodic Reports...................................
7.3 Licensee Event Report Review.........'..............
7.3.1 LER 90-27, Blocked Open Fire Door...........
7.4 Quality Assurance Issues Closed....................
7.4.1 Adequacy of Computer Software...............
-
[
7.5. Management Meetings.................................
- The NRC inspection manual inspection procedure (IP) or temporary lc
' instructions (TI) that was used as _ inspection guidance-is listed for each applicable report'section.
f.
r-(E
,
b!!!i '
f
<
<
\\l
.
'.
h p-
!<
,
.;;
i (
i i
V i
If
.
e
.
,
,
,
'
DETAILS 1.0 plant Ooerations Summary Millstone Nuclear Power Station, Unit No. 3 (Millstone 3 or the plant)
began the report period at 100 percent of rated thermal power. On September 13, plant power was reduced to 50 percent when excessive loading on the intake structure caused by a jelly fish influx threatened to auto-
'
matica11y trip condenser circulators water pumps.
On September 14 while at reduced power, the "B" feedwater regulating valve began to operate erratically. Accordingly, the plant was further downpowered to 25 percent power. While the steam generator water level was being controlled on the feedwater regulating valve bypass valves, the positioner and booster relay on the B feedwater regulating valve were replaced which eliminated the erratic valve operations. A plant power increase was commenced on September 13 with full power obtained on September 16. On September 18 and October 4, plant power was reduced to 85 percent for condenser back-Wash operations. On October 4, while backwashing operations were in progress, stormy weather conditions caused debris to be blown into the intake structure which exceeded seaweed removal capability of the equip-ment. Accordingly, plant power was rapidly reduced to 32 percent which is below the turbine trip / reactor trip setpoint. Af ter completing storm related repairs to the intake system, traveling screens and screen wash system, plant power increase was commenced on October 6.
Full power was reached on October 7 at which the plant remained for the balance of the report period.
2.0 NRC Inspection Summary During the week of September 4, an evaluation of the Quality Assurance Program and the ef fectiveness of the progrim implementation was conducted at Millstone Unit 3 by regional inspectors.
Findings will be addressed in inspection report 50-423/90-17.
During the week of September 24, two separate inspections were performed.
The first inspection was conducted by regional physical security in-spectors in response to improper handling of two recent Fitness for Duty (FFD) events at the Connecticut Yankee and Millstone sites.
The team reviewed both events and conducted an evaluation of the licensee FFD program. At the September 28 exit meeting the inspectors informed the licensee that an enforcement conference would be scheduled in Region I to discuss the two FF0 cvents.
Findings will be documented in inspection report 50-427/90-22.
(
-
-
_
..
F'
i
.
,..
i
'
i
'
.
1 i
\\
\\
l
,
The second inspection conducted during the week of September 24 involved i
regional and resident personnel.
The inspection, which was conducted at the licensee corporate headquarters in Berlin, focused on the adequacy of J
the licensee program for timely identification and reportability of plant
!
design vulnerabilities.
Inspection results will be contained in report
,
50-423/90-83.
j Two inspections were also conducted during the week of October 1.
The first inspection consisted of a routine review of the licensee health l
physics program.
Findings will be documented in inspection report 50-423/90-23.
'
c
l-The second inspection included a review of the licensee program for handling nuclear safety concerns raised by employees, and an assessment of j
the overall safety ethic at Millstone. The inspection team consisted of
.
Region I inspectors, headquarters personnel and NRC consultants.
The team interviewed licensee personnel at the Millstone and Connecticut Yankee
,
sites and the corporate office.
Inspection findings are documented in l
inspection report 50-423/90-82.
On October 11, Commissioner Curtiss and a staff technical assistant accompanied by Charles Hehl, Director, Division of Reactor Projects,
'
Region I, toured the Millstone site.
During the tour, items of NRC interest at the Millstone Site were discussed with senior licensee
-
management.
The. inspection activities during this report period included 152 hours0.00176 days <br />0.0422 hours <br />2.513228e-4 weeks <br />5.7836e-5 months <br /> of inspection during normal activity working hours.
In addition, the review of_ plant operations was routinely conducted during periods of backshifts (evening shifts) and deep backshifts (weekends and midnight shirts).
i
-
Inspection coverage was provided for 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> during backshif's and 7 l
hours during deep backshifts. An exit meeting which provided the results of this inspection (50-423/90-19) was conducted on November 6, 1990.
'
3.0 Operational Safety Verification
3.1 Control Room Observations The inspector reviewed plant operations from the control room and reviewed the operational status of plant safety systems to verify safe operation of the plant in accordance with the requirements of technical specifications and plant operating procedures. Actions taken to meet technical specification requirements when equipment was inoperable were reviewed to verify the limiting conditions for operations were met.
Plant logs and control room indicators were reviewed to identify changes in plant operational status since the last review and to verify that the changes in the status of plant equipment were properly communicated in the logs and records.
.
,
,
,
'
.
Control room instruments were observed for correlation between
'
channels, proper functioning and conformance with technical speci-fications. Alarm conditions in effect were reviewed with control
,
room operators to verify proper response to off-normal conditions and to verify operators were knowledgeable of plant status.
Control roor manning and shift staffing were reviewed and compared to technical specification requirements.
'
Trainees who were manipulating reactor controls were under instruction by licensed operators. Operators were found to be cognizant of control room indications and plant status during normal working hours and backshift observations.
No significant observa-tions were noted.
3.2 Plant Operations Tours The inspector observed plant operations during regular and backshift tours of the following areas:
,
Control Room Vital Switchgear Rooms
,
L Diesel Generator Rooms iurbine Building l
Intake Structure Auxiliary Building Engineered Safety Feature Spent Fuel Building (ESF) Building Demineralized Water Main Steam Valve Building Storage Tank (DWST)
Of the above listed areas, the inspector observed good housekeeping
>
practices in that there was a notable lack of loose tools, equipment
,
and combustibles.
Radiological controlled areas were properly
'
posted. Contamineted floor space areas were modest in size and did not seem to be more extensive than observed in recent tours.
In plant areas where service water system equipment was located, ex-tensive surface corrosion uf pipes and valves was noted. One example is the service water piping and valves in the control room heating ventilation and air conditioning (HVAC) chiller cubicles. The 11-censee should monitor these areas and preserve equipment where pos-sible to prevent corrosion build-up which could affect component operability. The inspector noted that the licensee continues to make progress in the painting of bare concrete and support piping in areas
,
such as the "A" diesel generator building.
The appearance of that area as a result of the painting was substantially improved.
Con-stquently, surface corrosion buildup, which had been noted on piping and support structures in the diesel room, was eliminated.
During the plant tour, the inspector noted a structural element marked as a " temporary support" attached to an instrument line in the
"B" recirculation spray system (RSS) heat exchanger room. The in-spector questioned the licensee as to the purpose or function of the
I
.
I
i
.
,
-
,
[
support since the instrument line appeared to have adequate seismic support.
Licensee research concluded that the structural element was temporarily installed during construction to prevent damage to the
,
'
instrument line from foot traffic and that it did not perform any support function.
Since the instrument line was seismicly supported
,
and the mass of the " temporary support" was not reflected in the
'
associated pipe loading calculations, the licenne elected to remove the subject structural element.
The i.aspr%r confirmed removal of the structural element and has not observed any additional temporary supports on the RSS during walkdowns.
Therefore, the inspector concluded that the observation was an isolated incident of improper system turnover during construction and had no further questions.
3.3 Posting and Control of Radiological Areas During plant tours, posting of contaminated and high radiation areas were reviewed with respect to boundary identification, locking re-quirements, and appropriate control points.
No significant obser-vations were made. A spill of contaminated fluid during routine reactor coolant pump seal injection filter changeout and the sub-sequent radiological followup is discussed in section 4.2.
r
,
3.4 Security-Selected aspects of site security including site access controls,
personnel searches, personnel monitoring, placement of physical i.
barriers, compensatory measures, guard force staffing, and response to alarms and degraded conditions, were verified to be proper during inspection tours. No significant observations were made.
3.5 Review of Plant Incident Reports @lRs)
i The plant incident reports (PIRs) listed below were reviewed during the inspection period to (1) determine the significance of the
.
events; (ii) review the licensee's evaluation of the events; (iii)
i verify that the licensee's response and corrective actions were proper! and, (iv) verify that the licensee reported the events, if
'
required.
The PIRs reviewed were:
numbers 3-90-141 through 3-90-154.
PIR 3-90-151 warranted inspector followup and is discussed in section 3.7.
3.6 (0 pen) Violation 90-08-01:
Failure to Follow Procedures-During Heat _up/Cooldown of Plant This violation, as documented in inspection report 50-423/90-08, concerned the failure of operators to adhere to procedures during heatup/cooldown operations at Millstone Unit 3 during a forced outage period. The lack of procedural adherence during plant cooldown on May 12, caused a feedwater isolation to occur and on May 18, 1990,
,
t k
.
i y
.
\\
.
during piant heatup, rendered both trains of safety injection inopere'ile.
In a July 31 letter to the NRC Region I office, which responJed to the notice of violation, the licensee outlined the following corrective 3ction.
(1) Information on the events would be
!
routed i;; ;.11 licensed operators for review to emphasite the need to i
reference and use procedores. (2) The unit director and operations manager will emphasize following existing procedural requirements with operations perste :I during training sessions.
(3) Direction was provided to operators on the need to provide a detailed turnover when an " extra" operator is used to perform a task.
(4) Operating procedure 3260 " Conduct of Operation" was revised to provide detailed instructions on the use and responsibilities of a dedicated operator.
g
'(5) All operations department procedures were to be reviewed for determination of steps which require a dedicated operator.
Accord-ingly, those steps would be revised to include the additional guid-ance.
The inspector noted that licensed operators were required to review the, licensee event reports and NRC inspection report 50-423/90-08 which described the events.
The inspector reviewed the revised OP
.
3260, which provided guidance on the use and duties of a dedicated I
operator, and determined the additional instructions were satis-
!
factory.
The inspector attended two licensee meetings during which
!
shift crews heard a presentation by the Millstone 3 operations mana-ger and unit director concerning procedural compliance. The opera-tions manager talked at length on the theory and need for procedural compliance._ The unit director stressed:
(1) management support for
.any shift crew that determined the need to stop plant operations for V
clarification of procedures during plant operation, and (2) the need to perform plant operational tasks in a controlled manner properly the first time.
Following the presentation, suggestions on proce-dural and operational improvements were discussed.
,
The licensee's presentation was appropriate in tone and content. The
!
presence of the utit director was appropriate and demonstrated the licensee's support of, and commitment to, procedural compliance. The
,
inspector concluded that the licensee fulfilled its commitment con-i cerning: (1) meeting with shift crews to discuss procedural com-pliance; (2) providing information on the two events to licensed operators; and, (3) revising operathg procedure 3260. The effec-tiveness of these corrective actions will be reviewed during future resident inspections.
Examinations of the revised operations de-partment procedures will be commenced, when the licensee has com-pleted their review in December of 1990.
<
___ _ ____ _ ___ ___
__--__
l *s
.
.c
,
D
).
i 3.7 Control Room Chiller Breaker Failure On September 18, at 6:30 a.m., while a primary equipment operator (PEO) was preparing to lock out the carbon dioxide fire suppression system to the east switchgear area to support maintenance activities scheduled for that morning, an east switchgear zone panel fire alarm
';
was received.
The operator locked out the affected zone panel fire alarm, and while investigating the cause of the alarm identified j
smoke in the switchgear panel areas. The control room was notified and the unit fire brigadt responded in accordance with Millstone 3 procedures. The source of the smoke was traced to cubicle 34 C-2 which contains the breaker for the "A" control room air conditioning chiller. The breaker was subsequently manually tripped by electrical maintenance personnel and " racked down" for examination.
Inspection i
'
of the breaker revealed that the trip coil which opens the breaker l
did not deenergize when the chiller was secured from the control room at 6:16 a.m. to allow starting of the B unit.
Because the trip coil is not rated for continuous energiration, the resulting heat buildup caused the trip coil insulation to break down resulting in the smoke.
Damage was confined to the trip coil and neither the breaker nor any other safety-related equipment was affected.
Licensee Troubleshooting Efforts i
The licensee subsequently performed the following three actions to
)
identify the failure (1) Repeated cycling of the breaker with a new trip coil.
(2) Visual inspection of the breaker in an attempt to
.
identify loose or binding mechanisms.
(3) Disassembly of operating mechanism. Although no definitive failure mechanism was identified, licensee personnel did note that auxiliary contacts 52 2/2C and l
4/4C which supply power to the trip coil mechanism had varying resistance measurements.
It was postulated by the licensee personnel and a General Electric representative (the vendor of the 4160 breaker) that the varying resistance that existed across the contacts
.
(25-50 ohms) dropped the coil voltage below the 60 volts that is
required to engage the trip shaft which opens the breaker and de-
energizes the coil. Therefore, when operators attempted to secure the chiller, the normally deenergized coil continued to be energized;
!
,
this provided constant voltage across the coil which eventually led i
to failure of the mechanism.
!
This explanation was further supported by the sequence of events
{
printout vnich shows the "A" chiller breaker opening at 7:08 a.m.,
'
when it was tripped manually by maintenance personnel. This was i
approximctely 45 minutes after the chiller was placed into pull-to-lock by a control room operator.
Licensee corrective actions consisted, in part, of replacement of the auxiliary breaker contacts and modifying breaker preventive maintenance procedures to require measurement of the auxiliary contact resistance.
l i
,
,:
l Q
-):
a
,.
w Inspector Review
,
i The inspector reviewed the licensee response to the breaker failure and determined that operators responded properly to the east switch-
gear fire alarm when it was received in the control room.
Station
'
firefighting procedures were followed, proper notifications were made, and a due regard for personnel and plant safety was noted.
Troubleshooting activities conducted on the breaker mechanism by electrical maintenance personnel were thorough and systematic.
However, inspector review of operator actions prior to the event
-
revealed that the control room ventilation system operating p.ocedure OP 3314F, " Control Building Heating, Ventilation, Air Conditioning
,
and Chilled Water," was not followed when the control room air con-i ditioning units were " swapped".
apecifically, section 7.2.3.2 of OP i
3314F provides the following instructions to operators when switching control building air conditioning units.
"TilRN the operating train Chiibr switch to 0FF,
,
then to PULL-TO-~.00K and Observe the following:
1.
Chilled water pump 3HVK*P1A (P18) STOPS.
2.
The following Chilled Water Isolation Vanos
!
CLOSE:
3HVK*TV68A(68B) and HVK*TV72A(728)
--
3HVK*TV69A(69B) and 3HVK*TV73A(72B)
--
3HVK*TV70A(708)and3HVK*TV74A(74B)
i
--
3HVK*TV71A(718)and3HVK*TV75A(758)
--
3.
Control Room ACU 3HVC*ACLIA (B) STOPS.
4, Computer / Inst Rack Room ACU 3HVC*ACU2A (B)
i
- STOPS.
5.
Booster Pump 35WP*P2A (P2B) - STOPS.
Verify I
by Control BLDG CHLk CONDSR A'(B) SW
~
FLOW LO (VPIA # 3-3) alarm, 'ing received.
6.-
Chiller 3HVK*CHLIA (IB) - STOPS. This energizes the Program timer. Observe the following local indicating lights are LIT.
,
!
'
a.
01-Stop a
b.
Program Timer c.
Oil Pump d.
Power Avail e.
Low Water Flow
,
!
+
-
!
-
..
m
i 40 Seconds later, the Chiller Lube Oil Pump 3HVK*P3A (P3B) - STOPS and the Oil Pump light goes L'
.
7.
CONTROL BLOG CHW SYS TRAIN A (B) Bypassed Annun-iator Lit (MBID 6-1) (MBIE 6-1)"
The operator who performed the chiller "twap" verified that actions
1-5 and 7 were completed through observations of annunciators, changes in status lights at the main control board and hearir,,
variations in control room noise as the chilled water and service wator booster pumps, which are located one level above the control room, stopped. However, he did not verify that chiller 3HVK*CHVIA had stopped as required in Step 6.
The inspector noted that this action cannot be accomplished at the main control board and requires an individual to examine the chiller locally.
The inspector con-cluded that if the chiller was examined by an individual per step 6 of OP 3314F'after performing the " swap," he would have noted that the
!
chiller.was still in operation.
Therefore, actions could have been l
taken to' trip the breaker manually and possibly avert the subsequent
!
trip coil-failure.which occurred approximately 14 minutes after the-l
'
>
,
operator attempted to secure the chiller.
'Through conversations with the operator and. licensee personnel, the
inspector was informed that the majority of operators do not perform
!
step 6 of the procedure.
Additionally, the inspector was informed i
m that, because of difficulty experienced in operating the air condi-
'
tioning unit, the control room ventilation operating procedure OP
.;
3314F_is routinely-performed differently by different operators.
.
Variations in procedure implementation is a concern since operators
.were. instructed by licensee management to follow station procedures or revise them if they do not work or can be improved. This ex-pectation was expressed during recent training sessions concerning procedural compliance, as discussed in section 3.6' of this report. -
'
The inspector concluded that these expectations may need to be re-F inforced among operators since this crew-had already attended the sessions. Also of concern was the failure of-the licensee to iden-tify the-procedural comp 1'ance issue during the investigation of l
Plant Incident Report (PIR) 3-90-151, which was written to document-
the trip coil failure. Through conversations with an operations i
supervisor,_the inspector was informed that the licensee PIR inves-
^
tigation focused on the cause of the trip coil failure and the op-erator response to the resulting firr alarm.
Examination of events
prict to the trip coil failure wert "t considered by the operations department, since the initial repod i.om electrical maintenance was i
that the mainibreaker had opened as designed and the auxiliary con-tacts.had failed. It was only after subsequent investigation that the
'
it was determined that the main breaker had not opened. The inspector was concerned that_the failure to examine this aspect of the event E. -,
y
..
'
l lwould preclude-the licensee from recognizing a precurser to a simi-lar event.<Therefore, any guidance that could be provided to opera-tors if another breaker fails to actuate would never have been gen-
erated. The inspector discussed the procedural adherence issue with
'
licensee management who stated that this event was discussed during an operations department supervisor's meeting, and used as an example where procedural compliance could be improved. The operations man-ager stated-that through continual reinforcement of management's
.
expectations concerning procedural usage, improved results will be
!
obtained. The inrpector noted the operations manager's comments.
i i
The inspector considers the failure of the operators to follow OP 3314F step 6 in this instance to not be safety significant since it
,
cannot be determined whether the trip coil burnup could have been
!
prevented if an operator had followed the prot.edure and discovered that the chiller had not tripped.- Therefore, per the policy of 10~
-
l CFR.2 Appendix C this incident is considered a non-cited violation d
(50-423/90-19-01).
<
-).
The it.spector has noted that previous PIR investigations are thorough-
and complete; therefore, this occurrence would appear to be isolated.
'
'The inspector will continue to follow operator use and-adherence to-
-
,
procedures and licensee followup of events in future resident inspec-
!
tions.
I
~ 3.8- (Closed) Violation (88-23-01):
Failure to Establish
Corrective Measures When Fire Doors Are Blocked-Open For i
i Maintenance This item tracked development of im' proved procedures which would
outline compensatory actions that must be taken when fire doors are j
blocked open-for maintenance activities. Improved procedures have
'
been developed as discussed in section 7.3, and this item is closed.
,
4.0 Maintenance-
-i
' 4.1 Observation of Maintenance Activities i
The inspector observed and reviewed selected portions of preventive
!
and corrective maintenance to verify compliance with regulations, use~
'
of admiristrative 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:
!
a
.
.
s
...
- -
.
e
.
'
M3-90-15080, Service Water Leak Repair on the "A"
.
--
Control Room Chiller Spool Piece 3 SWP-006-035-003,
'
'
September 10, 1990 M3-90-16754, Pressurizer Pressure Alarm Indication 3
--
RCS*JK1 Gain Adjustment (Channel 1) Lead /l.ag, September 11, 1990
.
M3-90-17200, Control Building Water Chiller 3HVK*CHCIA
--
Breaker Troubleshooting Operations, September 18, 1990 M3-89-16512, Emergency Olesel Generator "B"
!
--
Maintenance, October 2, 1990 M3-90-03290, Seal Water Injection Filter Changeout Due E
--
to Clogging, October 9, 1990 AWO M3-90-03296 is discussed in section 4.1.1 of this report.
4.2 Spill of Contaminated Fluids During Filter Changeout On October 9, 1990, at 10:24 a.m., a spill of approximately 800 to 1000 gallons of primary-coolant to the auxiliary building resulted during a changeout of reactor coolant pump seal injection filter-
' 3CHS-3A.
The sp111'resulted when maintenance personnel inadvertently loosened the cover bolts for the "A" filter assembly while it was-still pressurized. The filter-that was scheduled (and tagged out) to be changed was the adjacent filter, 3CHS-38. Both filters 3CHS-3A
.
and 3CHS-3B.are located adjacent to each other and are separated by
- !
concrete barriers. ' Filter changeout is accom;.lished through removal of a concrete plug and. subsequent filter unbolting through use of
- long handled extension tools.
Through discussions-with licensee personnel, the inspector was in-formed that the mechanics who were performing the filter assembly c
~
changeout had loosened two of the four filter assembly retaining bolts when difficulty was. experienced in loosening the remaining-two.
~
Sensing that something was wrong, the' mechanics stopped trying to loosen the remaining bolts and then noted through observation of'a-
.
-label. located on the adjacent _ wall that they were working.on the
'
incorrect assembly.
The mechanics then reattached the two loosened
'
bolts and'while rebolting the assembly, the spill occurred.
Control room operators were alerted to the spill by a " Reactor
.!
Coolant Pump Seal' Water Injection Flow Low" annunciator, which illu-
'
minates when seal flow decreases to 6.5 gallons per minute vice the normal 8' gallons per minute. Through discussions with operators, reactor coolant pump seal flow eventually dropped to 0 gallons per minute and pressurizer level decreased about 7 percent during the
,
'i transient. Upon receipt of the alarm, operators were dispatched to the auxiliary building and af ter being ' informed of what happened by the mechanics, unisolated the B reactor coolant pump filter and
,
I
- 4 9'.L c
,
.+.,
3 isolated the A filter which stopped the leak.
During the. event, i
reactor coolant pump seal flow was maintained through thermal. shield I
backleakage. The total time of the event was nine minutes from
receipt of the low reactor coolant pump seal flow annunciator to j
restoration of normal flow.
Observation of seal parameters following
-
the event indicated that no seal damage resulted from the transient.
During isolation of the leaking filter assembly, four individuals were slightly contaminated.
The contamination was primarily confined t
,'
to the shoes of the personnel.
However, one individual's feet were l
wetted and required subsequent decontamination through use of soap i
and water.
l
,
Activity levels detected on the individuals were low with the grest-est at 3000 counts per minute.
The individuals who were contaminated were admini cered whole' body counts and their thermoluminiscent devices (TLDs) were read. Whole body results indicated that no radioactive intake was obtained, and,TLD readings did not indicate
,
any personnel exposure. These results were consistent with local area radiation monitors which did not show an increase in radiation levels had occurred as a result of the spill.
Inspector Assessment The1 inspector noted that operators responded quickly to restore-
-
,
. reactor coolant pump seal injection flow once it.was lost..The inspector noted that seal injection flow may have been restored even
-
earlier had several remote n1ve actuators not failed to operate.
-The failures of the remote'actua brs necessitated having the opera-tors enter confined spaces to perfo w the system isolations.
The inspector noted that prior to this event reactor coolant' pump filters had'been. changed approximately 60 times without-incident.
therefore, this event appears to be an isolated occurrence of in-attentiveness to detail. The' inspector also noted that the mechanics
-
who were performing the task showed a good. questioning attitude by.
stopping filter changeout when difficulty was encountered during v
tremoval of the two remaining bolts. This action: prevented a possible
,
personnel injury from occurring or a more serious spill. The in-
'
spector considered health physics cleanup and assessment of the event
'
to be comprehensive and thorough.
Planned corrective actions to prevent recurrence include improved filter labeling and discussion of-the event with station personnel.
The inspector has concluded, based upon review of the event, that the licensee corrective actions are adequate and has no further_ questions.
\\
l'
'
L'
L l:
It
e:
-.%,
,- :.
Ls
5.0 Surveillance i
5.1 Observation of Surveillance Activities The inspector observed and reviewed portions of completed surveil-lance tests to assess performance in accordance with approved pro-cedures.and Limiting Conditions of Operation, removal and restoration of, equipment, and deficiency review and resolution. The following
'
tests were reviewed:
SP 3616A.1, Main Steam Valve Operability Test,
--
September 6, 1990 SP 3622.3, Auxiliary Feed Pump 3FWA*P2, Operational
--
Readiness Test SP 3211A, New Fuel Assembly and RCCA Receipt and
--
Inspection, September 26, 1990 J
l No significant observations were noted.
6.0 Engineering / Technical Support 6.1 (Closed) Unresolved Item 50-423/89-04-02: Temporary Non-Code Repairs to ASME Class III Piping L
This item was. opened to track NRC establishment of guidance for
,
performance of non-code repairs to ASME code class III piping. On June 15,.1990, the NRC-issued generic letter 90-05 " Guidance for
' Performing Temporary Non-Code Repair of ASME Code Class.), 2 and 3 Piping", therefore this item is' closed. The licensee has elected te follow the guidance contained in the generic letter and two relief requests for ^ temporary non-code repairs for ASME class III-piping
'
.have been submitted to the NRC for review. As of October-10, 1990, one submittal had been approved, the other submittal is awaiting NRC-
~dispositioning'
The inspector has witnessed several non-code repairs.
.
to-service water ASME class-III piping.
Licensee actions to address a:1eak in a bellows expansion joint on the "A"' control room chiller
'
. unit are described below.
<
The inspector reviewed the temporary non-code repair of a service water system expansion joint on the inlet side of the train A control room chiller. The expansion joint had experienced leakage and-due:to
"
- a. lack of a suitable replacement part,.the licensee elected to per.
,
'
form a' temporary non-code repair. The non-code repair consisted of
'
the fabrication of a; replacement expansion joint using a rubber spool piece (General Rubber Company " Sound Zorber") and-flanges salvaged-from the original expansio,i joint, r
,
_
N
.
- *
..
,
s.
The inspector observed the hydrostatic test of the replacement ex-pansion. joint.
Some leakage was noted in the attachment flange but was under conditions permitted by the ASME code for hydrostatic test-
,
!
ing. With regard to hydrostatic testing, section XI of the ASME code-directs the user to the original construction code section if the
,
applicable rules are not addressed in section XI.
In this regard, paragraph ND6224 of section III permits slight leakage of temporary gaskets, installed for hydrostatic testing, provided the leakage does not exceed the capacity of the hydrcstatic test pump to maintain the
,
h test pressure. The li;ensee subsequently addressed the leakage in
[
nonconformance report 390-079. - The inspector viewed the temporary-non-code repair after installation and noted a loose tie bolt and
,
'
minor leakage at the attachment flange; both conditions were subse-quently corrected by the licensee.
-
l The inspector noted that by destroying the bellows expansion joint to perform the non-code repair, the licensee had performed an ir evers-
.
,
l ible repair action.
Performance of an irreversible repair prior to u'
receipt of NRC approval is not consistent with the NRC staff position
':
on-performance of non-code repairs.
This position was outlined in an August-16, 1990 NRC internal memorandum which stated that the NRC
- staf f will allow'only reversible types of non-code repairs to be undertaken prior to NRC approval.
Reversible repairs are viewed as temporary stop gap measures to-limit leakage (that is, the element in question can be ret'urned to.its flawed condition),
In this case, the
licensee destroyed the flawed expansion joint in order to salvage the flanges for use with the rubber spool piece.
In'the event that the
-
NRC.does not accept this temporary non-code repair,' an unacceptable situation would exist since no' code-acceptable replacement would be
,
available for approximately six months. This issue was discussed
"
with an engineering supervisor who noted the inspector's comments.
.
<The'NRC internal memorandum has been provided to the licensee for
"
guidance and is available in the Public Document Room for review.
- The. engineering supervisor stated that in the future, prior to per-
,
formance of a non-reversible repair, NRC approval'of the non-code
,
repair will be obtained.
q By' letter dated September 26, 1990, the licensee submitted a request for relief from 4SME code requirements in order to obtain NRC ap-proval for the subject temporary, non-code repairi This relief request is presently under review by the NRC staff.
6.2 New Fuel Receipt Inspection During the report period, new fuel assemblies were delivered on _ site in preparation for the February 1,1991 refueling outage period. The inspector-noted that, prior to fuel delivery, training was conducted for. personnel involved in the receiving, inspection, and handling of fuel assemblies.
The inspector attended one training session and
noted that, during the training session, the instructor covered the
.-
.
4..,,
significant' areas of SP 3211A "New Fuel Assembly and RCCA Receipt and Inspection", and problems that were encountered during previous hand-ling of fuel assemblies.
The inspector had no questions regarding the adequacy of the training, however, the inspector noted that fuel handling problems which other facilities experienced were not dis-cussed.
The inspector discussed this item with a licensee reactor engineer who_ indicated that he would review applicable data banks to determine if there are any events reported at other facilities which
Millstone could benefit from.
!
The inspector noted that, upon arrival of the fuel delivery vehicle e
at Millstone,' health physics personnel performed gamma and alpha surveys in accordance with station procedures. The inspector observed that, once the vehicle was surveyed by health physics-per-
!
sonnel and released for entrance into the facility, it was searched i
by security guards in accordance with licensee procedures.
'{
-The' in'spector observed fuel of fload in the fuel delivery area.
The
. inspector noted that appropriate radiological surveys were taken and fuel inspection was conducted in accordance with SP 3211A.
Based j
upon the inspector's obse tions, new fuel receipt, inspection and
'
offload was conducted in a safe and controlled manner. -The inspector
-
had no further questions on this item.
6,'3 10 CFR 21 Report Submitted Concerning Veritrak Transmitters By letter dated August 27, 1990, the licensee issued a 10 CFR Part 21
'
report to the NRC concerning Westinghouse Veritrak Transmitter Model i
No. 760P2. The subject transmitters are-used in nineteen applica-i tions at Millstone Unit 3 including steam generator level and pres-l surizer level inputs to the reactor protection system.~ The defect
associated with the' transmitters caused them_to experience drift
!
- during plant' operation. When the protective coves were removed.for
!
troubleshooting-operations' on the pressurizer. (3RC*LT459) and steam j
.
generator Veritrak (3FWS*LT528) level monitoring chai.,els which had
.;
experienced channel _ dri f t, signi ficant aaan.ic;..*1 dri f t was noted,
~
.This caused the licensee to hypothesize that_the drift e @ t was temperature related. These Veritrak transmitters were subsequently replaced during a May 1990 outage.
Prior to. replacing the. transmit-
~
ters, the replacement units were checked for temperature sensitivity
by testing transmitter output using a tape chart recorder over a
,
temperature range varying from 76 - 120 and 69 - 120 degrees in a
!
controlled environment with a:7300 power supply input.
Tested W
transmitters were found to respond properly.
.
Recent conversations with the licensee's personnel indicate that the
.h steam generator replacement transmitter 3FWS*LT528 is performing
?
properly while the pressurizer level transmitter 3RCS*LT459 continues
.{
-to experience some drift.
These observations were confirmed by the j
inspector via review of comparison of redundant control board indica-j tions channel check and from plant computer p-intouts.
The
i 1>
t
_
._
,
-
z.
l '. Q.
>
.y,'..
w
,
subsequent drift of the pressurizer level transmitter suggests that a
_ temperature related problem may not be the cause. Accordingly, dur-
.
ing the next refueling outage, the licensee intends to (1) calibrate the three pressurizer level (two Rosemount 'and one Veritrak) trans-
-mitters, and (2) perform an ultrasonic test of the pressurizer level sensing line to determine if the associated condensate pot is vapor-bound.
Licensee personnel believe that the presence of vapor in the pressurizer level condensate pot might explain the drift associated
with 3RCS*LT459.
Licensee engineers stated that if the transmitter
?
drift is caused by vapor buildup, replacement of the pressurizer condensing pots with a smaller size design which is less susceptible
-
to gas entrapment will be considered as a long term fix.
Based upon direct observations and conversations with licensee per-sonnel, the inspector concluded that, although the licensee has not determined the'cause of drift in the replacement transmitter for 3RCS*LT459, the licensee is proceeding at a reasonable pace with a valid method of inquiry to determine the cause of the problem.
It was further concluded that the performance of a channel check to determine transmitter operability is sufficient to identify addi-
- tional transmitter drift.
Further, the inspector determined that the 10 CFR 21 report criteria requirements were met.
7.0: Safety Assessment / Quality Verification 7.1 Committee Activities-The inspector attended meetings of the Plant Operations Review Committee (PORC). The inspector noted by observation that committee administrative requirements'were met for the meetings, and that the committees discharged their functions in accordance with regulatory
'~
requirements. The inspector observed.a thorough discussion of mat-
ters before the PORC and a good regard for safety in.the_ issues under
' consideration by the committee.
No significant observations were
'
-noted.
-7.2 Periodic Reports'
,
Upon receipt.: periodic reports submitted pursuant to= technical
- specifications were reviewed.
This review verified that the reported information was valid'and included the required NRC data. The in-
'
!
spector also ascertained'whether any reported information should be classified as an abnormal. occurrence, The following reports were reviewed:
'
August Monthly 0perating Report
--
,
September Monthly Operating Report
--
No significant observations were made.
p L
k
,
- .,
...
7-3 Licensee Event Report Review
.
Licensee Event Reports (LERs) submitted durb the report period were reviewed to assess LER activity, adequacy
.:orrective actions, compliance with 10 CFR 50.73 reporting ret ements, and determina-tion of generic implications if further 1r..vrmation was required.
Selected corrective actions were reviewed for implementation and thoroughness.
LER 90-29 was selected for in depth review as dis-cussed in the foilowing report section.
7.3;l LER 90-27, Blocked Open Fire Door On August 11, 1990, at 8:40 a.m. with the plant at 99 percent of rated thermal power, a primary equipment operator (PEO) discovered a technical specification (TS) fire door blocked open. An hourly fire watch had not been established for approximately three hours as re-quired by TS 3.7.13 " Fire Rated Assemblies." Upon discovery of the deficiency, the door was promptly closed, placing the plant in com-
.pliance with'TS 3.7.13, Licensee investigation of the event determined that the door was
opened and a fire watch established on August 8,1990, for personnel
safety reasons to' eliminate a high differential. pressure across the i
door while maintenance'was performed on the associated surplcmentary
- leak collection and release system (SLCRS) filter room.
C r.
!
' August 11, 1990, when the automated work order ( AW0) was ' completed at
!
approximately 4:15 a.m., the operations department shift supervisor
(SS) requested a PE0 to close the door.
The PE0 ciosed the door at-4:45 a.m.
At 5:10 a.m., the SS contacted the National Fire and Medical (NFM) fire watch supervisor which is the contracted site fire
,
watch _ organization and instructed him to remove the door from the hourly-fire watch schedule. At 5:18 a.m. and at 5:24 a.m., two sepa-rate fire patrols documented that the fire door was in the open posi-tion after already being closed by the PEO. The 5:24 a.m. watch-personnel removed the fire patrol sheet and returned it to the con-trol room even though it was open, Who opened the door after the PE0 closed it and why it was opened is not known, At 5:30'a.m., the
<
fire watch superviser verified that the fire wat.ch sheet nad been removed.
The supervisor, noting'that the fire door was open, ap-parently believed that' the PE0 had not yet been dispatched to close the fire door.
The licensee attributed the event to be caused by inadequate train-
,
ing, since the individual who removed the fire watch sheet did not understand why the-fire door should be closed. To prevent recurrence of the event, a memorandum was issued.to fire watch personnel empha-sizing the importance of. relaying relevant information to operations department personnel concerning fire watch degradation without appar-ent corrective actions being instituted. Additionally, the NFM fire watch manager was replaced due'to recognized administrative deficien-cie '
w l
Mi l,.zy.. * s
.#
1'
Inspector Review i
l Inspector followup of this event consisted of review of Millstone and L
the contractor's Nationel Fire and Medical (NFM) procedures, discus-sions with licensee and NFM personnel, and examination of recent i
licensee event report 5 and plant incident reports (PIRs). The in-spector noted that f'.*1 watch personnel are to be trained and quali-fied in accordance with NFM-1 "NFM Operating Procedure" which is a station operations review committee (SORC) approved procedure.
In section 6 of NFM-1, which provides guidelines for hourly fire watch L
patrols, step 6.2 requires hourly fire patrols to point out the
,
'
L reasons for each hourly fire watch post and report potential hazards l
l -:-
t'o the control room and NFM supervisor.
The individual who removed L
the fire watch sheet from the fire door while it was still open ap-parently did not effectively communicate this finding to the plant shif t supervisor. Additionally, the NFM supervisor did not communi-
=cate his observation concerning the fire door.
He thought the L'
blocked open fire door was going to be closed by the PE0 when. in fact the PE0 had already closed the door. The~ licensee was not able to determine who reopened the fire door after it had been closed by the PEO.
p Inspector Assessment I
i L'
Implementation of the fire protection program at Millstene Unit 3 J
'
since the start of commercial operation has been less than optimal.
This fact has been documented both in systematic assessment of li-censee performance (SALP) reports.and routine resident inspection reports. A notice of violation was. issued in inspection report 50-423/88-23 to document NRC concarns regarding fire protection
- implementation. Licensee corrective actions taken as a result of the
Lnotice of violation and previous fire protection events, as docu-
,
mented in Inspection Report 50-423/90-15, consisted of labeling fire doors and. improving'the specificity of fire protection procedures to
- reduce chance for interpretation'and possible error.
The inspector
- has also noted that the licensee conducts periodic reviews of fire-c w tches during Sorning meetings During tne meetings, the status of work that must.be completed' prior to dismissal of the fire watch is reviewed'and completion dates are obtained from the cognizant depart-ment. The inspector noted that a periodic review of firewatches is
werthwhile= since management wi'l be able: to (1) identify and correct
.leng scanding-deficiencies; ano (2) determine if a firewatch is the
'
most siable answer to address a te&porary deficiency, i.e., does a fire door have to be blocked open to accomplish a task? The inspec-
,
tor believes that minimal. use of compensatory firewatches will lessen
'
.the possibility for improper fire protection implementation in that
'
fewer opportunities will exist for human error in the progra _
._
m
,
.q e
- *
,
L l'
After-completing a review of-the Millstone 3 procedures which implement the fire detection program, the inspector determined that l-they provide sufficient guidance concerning the type of action that should be taken when fire detection and protection assemblies are
'
disabled. Therefore, NRC violation 88-23-01 which tracked licensee L
correction of procedural weaknesses in the fire protection program is
closed. The inspector determined that inadequate communication be-tween plant personnel and ineffective implementation of station pro-
>
p cedures continue to degrade the fire prctection program.
Inadequate
'
'
communication and procedure inviementation resulted in LERs 90-18 and
~
90-27,-and PIR 90-127.
In a February 21, 1990,. letter to the NRC in
,
which the licensee responded to the SALP report 88-99, the licensee stated that communications in tha fire protection area would be im-proved.
The-inspector believes'that improved communications between the operations department and the' fire watch patrols should be imple-
"
mented. The plant director informed the inspector that he has held
. discussions with the NFM supervisor concerning implementation of the fire protection program. During those meetings, the plant director
. emphasized that he expects fire watch personnel to ask questions if a y,,
",
' condition appears.to be abnormal and inform the plant shift supervi-
-
sor.- Additionally, he emphasized the importance of performing duties correctly the first time. The director indicated that a task force has been established to examine ways to improve communications be-tween.the operations department and fire watch personnel.
The recent failure to establish a fire watch within one hour when the-
'>
fire related assembly was disabled is a violation of TS 3.7.13.
,
.Although previous violations of this technical specification have been considered licensee identified, and per the policy of 10 CFR 2 LAppendix'C, no violation was issued, the recurrence of these events e
- suggests additional ~ corrective action should be taken by the license to prevent' recurrence. Accordingly, enforcement' discretion is not t
-
being exercised in this instance and:a violation (50-423/90-19-02) is
- being issued'as a result of continuing weaknesses in the fire pro-
'
- tection program with specific-. regard to the' poor-communications that
have existed between plant personnel and-inadequate implementation'of-i 7the fire protection procedures.
j 7.4'-Quality Assurance Issues Closed-7.4.1 Adequacy of Computer Software The inspector reviewed the'following three concerns of a former
~
licensee employee regarding the adequacy of computer software pro-
'
grams;at Millstone Unit 3.
(1) Nuclear Engineering and Operations-Procedure 2.24, " Quality Software Programs" is inadequate in that it does not fully implement recognized industry standards.
For example, NE0 2.24 does not fully' adhere to ANS 7.4.3.2, " Application Criteria for Programmable Digital Computer Systems in Safety-Systems of
~w
,
(b '
L icl,,, i se
,
Nuclear Power Generating Stations." (2) The verification / valida-
'
tion process for. computer software programs at the Millstone Site is
,
deficient since the verification / validation process does not include
,
a functional test with specific predetermined acceptance criteria.
~
(3) Personnel qualifications for review of software are not specified by procedure.
Therefore, the possibility exists for unqualified people to conduct reviews of software packages.
These concerns were outlined in a June 6,1990, letter which provided the concerns to the licensee for investigation.
In an August 3
.
>
1990, letter which responded to the issues, the licensee concluded
>
that the individual's concerns were not substantiated. The bases for this determination is provided as follows:
In response to the first concern regarding the technical adequacy of NE0 2.24, the licensee noted that revision 1 of NE0 2.24 dated October 1989 committed to regulatory guide 1.152 which endorses ANS 7.4.3.2.
To implement the requirements of ANS 7.4.3.2, QS-12, " Additional Requirements for
~ Category I Software" was written and and effective on June 12, 1989.
In response: to_ the second concern regarding the adequacy of the
_
verification / validation process of software at Millstone Unit 3, the licensee reviewed software implementation' packages (SIPS) for control
,
grade:and quality software instituted from 1987 to the present. The results concluded that the verification / validation performed was
' adequate based ~ upon the complexity of the changes performed.
In followup.)f the issue which concerned the adequacy of qualifica-tions of personnel who conducted review of software packages, the licensee con:1uded_that software packages were reviewed by personnel i.
who are qualified by experience, education or both.' This is in accordance with station procedure: ACP-QA-2,13A "Ccmputer Sof tware Implementatian "
Inspector Review of Licensee Response Inspector _ followup of this issue consisted of a review of station-
. procedures, software' packages, and-interviews with personnel.
,
The inspector reviewed QS-12 and concluded that the procedure fol-lowed the. intent of ANS:7.4.3.2 and regulatory guide 1.152. There-fore, the inspector concluded that procedures written in accordance with this standard would follow the standards that the NRC has re-
'
. cognized as acceptable for QA category I sof tware. The inspector
>
reviewed plant design change request (PDCR) 89-15 which modified the software program-for_the inadequate core cooling system (ICCS), which is the only-software program that falls within the scope of RG 1.152.
Through conversations with instrumentation and control technicians who performed the change, computer programmers and the system engin-eer who was responsible for implementation of the PDCR, and through independent review of the PDCR package, the inspector concluded that the functional test performed was sufficient to test the effect of the software change on the plan. _ _ - _.
_ -_ _____ _
,
i g
-
.[99
,
.;
,
The inspector reviewed ACP-QA-2.13C " Control of Systems Containing Category I Sof tware" and noted that the procedure which provides instructions for the control, implementation, and testing of category 1 software, requires personnel to conduct independent reviews in accordance with ACP-QA-3.13 " Preparation, Review, and Approval of Design 1 Analyses and Calculations." Additionally, ACP-QA-2,13C re -
quires personnel who conduct independent reviews to be qualified in accordance with ACP-QA-8.16 " Training, Certification and Identifi-cation'of Qualified Inspection, Examination and; Testing Personnel."
Based upon inspector review of the above listed procedures, the in-
spector concluded that Millstone Station procedures adequately identi-
fy the qualificatiors that an individual must have prior to perform-ance of an independent review. Based upon inspector review of.the
.
licensee investigation and independent inspector followup, the in-
-.
spector concluded that the individual's concerns regarding computer software were unsubstantiated. This matter is closed.
(
7.5 Management Meetings Periodic meetings were held with station management to discuss
' inspection findings during the inspection period. A summary of findings:was also discussed at the conclusion.of the inspection.
No-
. proprietary information was covered within the scope of the inspec-j tion. Written material available in the public document room, con-cerning the completion of non-code repairs to ASME piping syste;.s, i
was-given to the licensee during the inspection period.
l
!
y i
\\ t
.$
i
'
.