ML20154L797
| ML20154L797 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 03/06/1986 |
| From: | Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20154L776 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-1.A.1.1, TASK-1.A.1.3, TASK-2.G.1, TASK-3.A.1.1, TASK-3.A.1.2, TASK-3.D.3.3, TASK-TM 50-423-85-74, NUDOCS 8603120252 | |
| Download: ML20154L797 (32) | |
See also: IR 05000423/1985074
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No.
50-423/85-74
Docket No.
50-423
License No.
Category B
Licensee:
Northeast Nuclear Energy Company
P.O. Box 270
Hartford, CT 06101
Facility Name: Millstone Nuclear Power Station, Unit 3
Inspection At: Waterford, Connecticut
Inspection Conducted:
November 19, 1985-January 6, 1986
Inspectors:
T. A. Rebelowski, Senior Resident Inspector, Millstone 3
F. A. Casella, Resident Inspector, Millstone 3
J. T.
Shedlosky, Senior Resident Inspector, Millstone 1/2
H. H. Nicholas
R. J. Summers, Project Engineer
Approved by:
&bM
't / GI S6
E. C. McCabe, Chief, Reactor Projects Section 3B
Date
Inspection Summary:
Inspectio. 50-423/85-74, 11/19/85-1/6/86
Areas Inspected:
Routine onsite regular and backshift inspection by the Senior
Resident and Resident inspectors (381 hours0.00441 days <br />0.106 hours <br />6.299603e-4 weeks <br />1.449705e-4 months <br />) and an NRC Region I inspector and
contractor employee (36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />).
Areas inspected included: review of licensee ac-
tion on previous findings, review of NUREG 0737 action items, witnessing of system
and component testing, observation of the initial core load; inspection of sur-
ve111ance, maintenance, construction and plant physical protection activities; and
operational safety verification reviews.
Results:
Four violations were identified.
The first, (Appendix A, Item A, and
Detail 19), dealt with maintenance and construction personnel failing to maintain
cleanliness control on safety-related systems which were opened up for modification
or inspection. The second, (Appendix A, Item B, and Detail 18) was a failure by
operations personnel to follow a written procedure to restore a safety-related
system to its normal configuration after an infrequent evolution. The third and
fourth violations were isolated instances related to physical security,~(Appendix
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B and Detail 13).
Notable license strengths were found in the continued professionalism and high
morale of the startup organization and in the conservative and safe approach taken
by management and the operations staff during the initial fuel load.
8603120252 e40307
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TABLE OF CONTENTS
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1.
Persons Contacted.....................................................
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2.
Summary of Facility Activities........................................
1
3.
Licensee Action on Previous Inspection Findings......................
2
4.
Licensee Reports of Significant Deficiencies..........................
5
5.
Rev i ew o f Wo r ke r Co nce rns . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
6
6.
Licensee Actions Taken as a Result of TMI Action Plan Requirements
Specified in NUREG 0737.............................................
7
7.
Licensee Action on Safety Evaluation Report (SER) Open Items..........
9
8.
Visit with Waterford First Selectman.................................
11
9.
Witnessing of Initial Fuel
Load......................................
11
10.
Preoperational Test Results Review...................................
14
11.
Startup Test Program Review..........................................
15
12.
Test Observation.....................................................
16
13.
Physical
Security....................................................
17
14.
Reactor Coolant System Pressure Boundary Leak........................
18
15.
Reporting of a Construction Deficiency on a Reactor Coolant Pump
Snubber............................................................
18
16.
Stress Reconciliation Program Repairs................................
13
17.
Routi ne Periodic I nspecti ons. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
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18.
Review of Plant Events...............................................
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19.
Cleanliness Control
Anomalies........................................
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20.
Exit Meeting.........................................................
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DETAILS
1.
Persons Contacted
J. Ferland, President
J. Opeka, Senior Vice President, Nuclear Engineering and Operation
R. Werner, Vice President, Generation and Construction Engineering
W. Romberg, Station Superintendent
J. Crockett,' Unit 3 Superintendent
F. Rothen, Construction Superintendent
The inspector also contacted other licensee employees during the inspection,
including members of the Operations, Radiation Protection, Chemistry, Instru-
ment and Control, Maintenance, Reactor Engineering, Security and Training De-
partments.
2.
Facility Activities Summary
The NRC issued operating license NPF-44 to Millstone Point Unit 3 in the
afternoon on November 25.
The first fuel assembly was loaded in the reactor
core approximately 26 hours3.009259e-4 days <br />0.00722 hours <br />4.298942e-5 weeks <br />9.893e-6 months <br /> later; detailed technical specification review
and Mode 6 surveillance testing were completed during the interim.
Fuel load
proceeded for 6 days, with the final of 193 assemblies loaded by approximately
11PM on December 2.
There were delays in the fuel load process, principly due to difficulties with
the spent fuel pool crane and Sigma refueling machine.
The spent fuel pool
crane load limit device repeatedly exhibited an intermittent open circuit that
could not be identified.
As corrective action, the licensee plans to relocate
the device to a more vibration free area of the bridge and install a repeater
for operator interface.
The Sigma refueling machine had problems with index-
ing, vertical positioning, drive motor degradation, and disc brake flutter.
At times, it was necessary to position and lower fuel assemblies manually.
One fuel assembly (A35) was bowed approximately 1/4" over its twelve foot
length, preventing the adjacent assembly (B49) from landing on the core plate.
The licensee modified the loading sequence to " box in" the B49 location.
Subsequently, the assembly was properly landed.
After the initial fuel load, numerous maintenance and construction activities
were undertaken and pushed transition to Mode 4 and commencement of post-core-
hot-functional-testing beyond the period of this report.
As part of the
stress reconciliation program, the four Containment Recirculation System (RSS)
heat exchanger tube side' nozzles were strengthened and the Quench Spray System
Chemical Addition Tank (QSS-CAT) supports were upgraded.
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Other work included' modifications to Supplementary Leak Collection and Release
System (SLCRS) damper control circuits, motor control centers and fan control
breakers.
A number of SLCRS boundary door seals had to be improved before
the SLCRS draw down test was completed satisfactorily.
Both Emergency Diesel
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Generators (EDGs) underwent complete fuel system inspections and had fuel
return lines replaced.
A check valve in the Primary Grade Water System (PGS)
supply to the Charging and Volume Control System (CVCS) was replaced.
Component Cooling Water (CCP) system containment isolation check valves were
lapped in and upstream flanges were installed to allow Local Leak Rate Testing
(LLRT).
In addition, CCP containment isolation valve motor operators were
replaced to assure timely closure against full system flow.
These changes
were undertaken to safeguard against a loss of coolant accident outside con-
tainment in the event of a tube failure in the reactor coolant pump thermal
barrier heat exchangers.
A Reactor Coolant System pressure boundary leak occurred when the B loop RTD
bypass manifold vent line socket weld failed due to cyclical fatigue.
The
socket was rewelded and the line configuration was modified to prevent recur-
Further vibration testing of other small bore piping subject to Reac-
rence.
tor Coolant Pump (RCP) induced oscillations led to additional work in upgrad-
ing the supports to the B RCP seal injection line and reanalysis of.the seal
injection line supports on the remaining 3 pumps.
At the completion of this report period, the plant remained in Mode 5.
3.
Licensee Action on Previous Inspection Findings
. A.
(Closed) Unresolved Item (85-62-04), Calibration of CO2 Concentration
Measurement Instruments
The licensee was able to mix a 40% concentration by volume of CO2 in air.
The 2 Truure instruments were able to measure this known concentration
within the +/- 2% accuracy provided by the manufacturer.
In each of the
8 areas tested, the Truure instruments were used to take data.
These
data show conformance with acceptance criteria.
This item is closed.
B.
(Closed) Unresolved Item (85-54-01), Seismic Category 1 Acceptability
of Control Room Pressurization Air Storage Tanks
The inspector did not find the installation acceptable due to construc-
tion deficiencies, lack of QC acceptance criteria, and incomplete QC in-
spection.
A QC inspection was performed shortly after the above discrepancies were
made known.
The inspector reviewed Stone and Webster QA Inspection Re-
port 5A00154, originated 9/23/85, which performed the QC inspections on
the air storage tank supports required by E&DCR TR-02761.
The specific
construction deficiencies from this unresolved item were included in the
list of findings attached to the QC inspection report.
The findings were
specific and well documented and resulted in Nonconformance and Disposi-
tion Reports 15390, 15530, and 15381.
The inspector reviewed portions
of these N&Ds for rework and disposition.
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There was a questionable " accept as-is" disposition.
One QC finding was
that some fastener nuts did not have sufficient thread engagement because
the required 1/4" of bolt was not visible.
This was disposed " accept
as-is".
The inspection questioned the disposition.
The cognizant engi-
neer stated that threaded rods were the fasteners in question, and that
they had square ends.
The 1/4" visibility factor was based on ensuring
full thread engagement when chamfered bolt ends were in use.
The engi-
neer stated that all fasteners checked had full engagement of the threads
on the rods and nuts.
The inspector had no further questions on this
item.
All discrepancies listed as unresolved in the Seismic Category 1 inspec-
tion have been completed.
This item is closed.
C.
(Closed) Unresolved Item (423/85-06-01) Approved Procedures forwarded
to NRC for Review 60 days prior to Scheduled Test;
The licensee has scheduled draft procedures of the start-up and power
ascension program to meet the 60 day criteria.
During this inspection,
the NRC received draft start up and refueling procedures within the above
time frame.
This item is closed as being no longer applicable.
D.
(Closed) Unresolved Item (423/84-07-01) Review Plant Design Change
Request Control Document to Improve Management Control
This item is addressed in ACP-QA-3.10, Preparation, Review and Disposi-
tion of Plant Design Change Requests (PDCRs) (NEO).
An extensive revi-
sion of Nuclear Engineering and Operation Procedure NE0 3.03 based on
review of PDCRs at the Haddam Neck Plant.
This item is closed.
E.
(Closed) Unresolved Item (423/85-02-03) Review of Service Water Tran-
sient Test Results
The licensee has analysed and tested various piping configurations to
determine the induced pressure transients caused by the partial draining
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of service water, with an accompanying air gap, between restarts of ser-
vice water pumps.
In-line check valve closing propagates pressure waves
throughout the service water system.
The tests pinpointed the service
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water system supply to a Ventilation Water Chiller as a contributor to
these pressure transients.
The licensee modified piping to reduce the
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air gap formation.
Additional testing verified that pressure transients
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had been reduced to an acceptable operating condition. This item is
closed.
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F.
(Closed) Security Outstanding Items (85-64-01)
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The licensee completed closure of security plan commitments prior to
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issuance of the operating license.
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The resident inspector performed verification inspections in a number
- of areas including:
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Verifying barriers and vital area closures for thirty eight areas
including gates, fences, drains, inlet and outlet ventilation areas,
diesel enclosures, and witnessing the use of increased patrols and
additional CCTV monitoring.
Protected Area Barriers in areas of drainage were inspected and
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patrols were observed.
Intake structure and hatch protection concerns were resolved.
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Lighted areas including patrols, trailer lights, and areas that did
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not previously meet criteria.
Detection aids and compensatory measures.
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Upgraded training of supervision and security force.
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All of the above areas were found acceptable.
This item is. closed (85-
64-01).
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G.
(Closed) Violation (IV) (85-423/85-12-03) Flooding of Engineered
Safety Features Building (ESFB)
As documented in the licensee's June 26, 1985 response to the May 28,
1985 report, the licensee's actions included an investigation of the
cause.
Correction actions were reviewed by the Plant Operation Review
Committee (PORC) and the Joint Test Group (JTG).
Included in this review
was the Flush Program activities in progress and their controls, test
activities released and currently in progress, and maintenance activities
in progress.
The results of licensee review indicated a failure to fol-
low the Automated Work Order tagging requirements.
Startup Engineers
were briefed on requirements to follow administrative program controls.
The above actions did not prevent additional flooding incidents which
are discussed in NRC Reports 85-16 and 85-23.
A program review, per-
formed in response to the above incidents, resulted in a tightening of
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flush controls.
Those corrective actions did result in a clarification
of flush program interties.
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The resident inspector witnessed area cleanup, removal of all wetted in-
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sulation, examination of RHR bearings and retest of systems that were
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involved in flooding (electrical circuit retest, valve packing replace-
ment, etc.) No deficiencies were observed.
This item is closed.
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H.
(Closed) Violation IV (85-423/85-16-01) Flooding of a portion of Con-
trol Building
The licensee investigation indicated that tagging for the Automated Work
Order (AWO) was adequately performed and in place.
The licensee failed
to verify that the isolated service water system was drained prior to
removal of a valve for maintenance.
Shift Supervisors, Supervising Con-
trol Operators, and Maintenance department personnel were briefed on the
incident.
Corrective actions included an examination of flooded area
electrical cabinets and battery rooms.
Shutdown panels in the 4'6" level
appeared to have soaked control wire.
No wiring required replacement.
The resident inspectors witnessed the cleanup.
Two successful tests of
the remote shutdown system were performed.
The administrative controls
on verification of maintenance prerequisites did not have to be modified
because the event was due to failure to comply with established practices.
Maintenance practices will be reviewed further under the NRC operational
inspection program.
This item is closed.
I.
(Closed) Violation IV (85-423/85-23-01), Fire Protection Water Flushing
With Improper Lineup
The licensee failed to tag a boundary valve during flushing.
Licensee
investigation concluded that personnel error and a weakness in admini-
strative Startup Manual procedures were root causes.
The licensee re-
vised the Startup Manual (Rev 4) to ensure that interrupttd test condi-
tions, which was a contributing factor in this incident, will be reviewed.
Additional controls on documenting system changes are in place.
No fur-
ther flooding incidents have occurred.
This item is closed.
4.
Reported Significant Construction Deficiency (COR)
(Closed) CDR 84-00-08, Main Steam Line Break Outside Containment.
The poten-
tial for a high energy line break outside containment to affect environmentally
qualified equipment was identified by the licensee on June 5, 1984, with the
Main Steam Valve Building being the primary concern. As a member of the
Westinghouse Owner's Group, the licensee reviewed the basis for the postulated
breaks, temperature and pressure profile development criteria, resulting
temperature and pressure profiles, and consequent results.
To satisfy NPF-44
licensee condition 2.C(3), the. licensee presented to NRR an analysis conclud-
ing that there is adequate protection to satisfy concerns related to this
issue.
The NRC found this acceptable as documented in Paragraph 3.11 of Sup-
plement 5 to the Safety Evaluation Report (NUREG-1031).
This CDR is therefore
closed.
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S.
Review of Worker Concerns
A.
Allegation RI-85-A-47
An FQC inspector stated he had knowledge of inspection aspects that other
FQC inspectors did not have but was directed by his management to cease
instructing other workers. An on-site individual was identified as being
knowledgeable.
That onsite individual was contacted by the NRC Region
I specialist inspector who followed up on this alleger's concerns. Also,
the NRC Region I Projects Section Chief for Millstone 3 contacted that
on-site individual by telephone.
That individual stated that he con-
sidered the alleger to be knowledgeable and that, to his knowledge, the
individual had been instructed not to instruct other inspectors in the
areas involved because the individual's experience at another site may
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not be directly correlatable to Millstone 3.
The NRC concluded that it
is the licensee's prerogative to designate instructors and that no con-
struction inadequacy had been identified in this matter.
Adequacy of
construction and construction FQC has been routinely assessed during NRC
inspections.
The overall conclusion, as documented in NRC Region I Re-
port 50-423/85-67, is that there has been acceptable construction, qual-
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ity assurance, and quality control.
Therefore, no additional follow up
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on this allegation has been prescribed.
B.
Allegation RI-85-A-105
This allegation pertains to QC inspection activities.
The first concern was that management failed to take corrective actions
for concerns about drawing control.
The alleger identified drawings as
not being of the proper revision.
Drawing control had been a previous
NRC concern for which the licensee had implemented corrective actions.
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It was determined by the inspector that these actions had been imple-
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mented and that the drawings identified by the alleger were of the proper
revision.
Two other concerns, pertaining to improper storage of drawings
and non-traceability of inspection records of cable tray supports in the
control building, had been identified by the alleger to the licensee's
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quality concern organization.
The inspector determined that the licen-
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see was taking action in accordance with their allegation program.
The
licensee's preliminary findings for the above concerns were that:
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(1) the drawing revisions were proper; (ii) improper drawing storage was
not substantiated; and, (iii) the inspection records were, in fact,
traceable.
As noted in Inspection Report 50-423/85-67, the as-built
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Millstone 3 design has been found, through multiple inspections, to con-
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form to requirements.
The inspector had no further questions on drawing
control,
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The second concern pertains to discrepancies on support anchor bolts
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which had been previously inspected and found acceptable.
The inspector
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reviewed the alleger's inspection reports and determined that a total
of 11 anchor bolts were identified as having less than the specified
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embedment on 4 supports located in the turbine building on the outside
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wall of tie control building.
As alleged, these anchor bolts had been
previously accepted.
However, the check which identified the discrepant
condition was a scheduled final inspection.
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The nonconforming conditions identified by the alleger were dispositioned
in accordance with the licensee's procedures.
The support was classified
as QA Cat 1 because it was attached to the Seismic Category 1 control
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building wall; however, the cable trays supported were on the turbine
building side and are neither seismically qualified nor safety related.
The licensee had identified the cause of the erroneous earlier inspection
as inspector error in interpreting an E&DCR written for the supports.
The E&DCR addressed whether rebar could be cut in order to get the re-
quired embedment depth for the anchor bolts.
The licensee reviewed other
QC inspection reports by the individual who made the faulty inspection
to take corrective actions where the same error was made.
In this case,
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defects identified were not numerous and were properly dispositioned,
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affected cable trays were not safety-related, and the lack of embedment
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of these anchor bolts does not adversely affect the safety-related wall.
The inspector had no further questions on this item.
The third concern was that the alleger was terminated af ter identifying
these concerns to the NRC.
The inspector determined that the alleger
was laid off in a planned reduction in force based on seniority.
The
alleger had been employed for less than a year and this layoff included
QC inspectors with greater seniority than the alleger.
The inspector
had no further questions on this concern.
6.
Licensee Actions taken as a result of TMI Action Plan Requirements Specified
in NUREG-0737
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The NRR staff reviewed the licensee's submittals associated with these items
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and discussed the results of this review in the Safety Evaluation Reports
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(SER) related to the operation of Millstone Nuclear Power Station Unit 3.
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During this inspection, the licensee's actions as des:ribed in the SER were
verified to ho<c been taken.
A.
I.A.1.1.1 & 3 (85-TM-01) Shift Technical Advisors (STA) (Closed).
The licensee had numerous meetings with NRR on the use of STAS at Mill-
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stone 3 and has committed to provide " Shift Advisors" for shifts which
do not have the prescribed hot operating e>nerience.
On obtaining re-
quired operating experience, an individual who is SRO licensed on Mill-
stone 3 and who holds a bachelors degree in engineering, engineering
technology or a physical science may serve in a dual role of SRO/STA.
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Otherwise, a dedicated STA who meets the criteria of NUREG 0737 is to
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be on each operational shift.
The inspectors verified that STA and Shift
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Advisor staffing meets Technical Specification and Operating License re-
quirements, and have reviewed and witnessed training of Shift Advisors.
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No deficiencies were identified.
This item is closed.
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B.
1.3.2A & B (85-TM-03) Shift Manning (Closed).
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The shift manning as described in the Technical Specifications and the
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Shift Operating Experience have been subject to License NP-44 condition
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14, which required certain hot operating experience.
The licensee's
present four shifts, expanding to six with NRR approval of Shift Advisors
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based on training program and simulator training, have been monitored
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by NRR, NRC Region I and the resident inspectors.
Shift complement meets
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Technical Specification and operating license requirements.
This item
is closed.
C.
II.G.1 (85-TM-19) Emeraency Power for Pressurizer Equipment (Closed).
The licensee's plan to satisfy General Design Criteria 10,'14, 15, 17,
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and 20 of Appendix A to 10 CFR 50 for the event of loss of offsite power
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required implementation of circuit design to assure reliable power sup-
plies. Millstone 3 has one PORV receiving power from 125V vital dc, with
its' associated block valve receiving power from the same 480V ac emer-
gency bus.
The second PORV and block valve has a similar arrangement
but is powered from the opposite electrical division.
The review of
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drawings EE-1AQ-B and EE-1BS-8 verifies this arrangement.
Specific Configurations:
1) Orange Train
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Valve
Power
3RCS SOV 455A
(PORV)---------BATT 1 (0)
3 PYS-22F
3RCS MV 8000A
(Block Valve)--32(0)-2R
(RHF)
2) Purple Train
Valve
Power
3RCS SOV 456
(PORV)---------Batt 2 (P)
3RCS MV 8000B
(Block Valve)--32(P)-2W
(RHF)
The licensee addressed this item to the NRC in letter 811511, April 11,
1985 (Counsil to Youngblood).
This item is closed.
D.
III A.1.1 (85-TM-22) Emergency Preparedness (Closed)
The NRR review of the Technical Support Center discusses the location
and facilities.
The resident inspectors have witnessed the use of the
center during a Unit 3 walk-thru of a simulated nuclear incident.
The
support center has three separate work stations, one for each unit, with
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design drawings and CRT displays of plant system data.
NRC facilities
for 2 to 3 persons are provided.
Communications to the Control Room and
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Emergency Operating Facility were acceptable.
This item is closed.
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E.
III A.1.2 (85-TM-23) Uparade of Emeraency Support Facilities (Closed)
The licensee has a combined site upgraded Emergency Plan and a Corporate
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Nuclear Incident Plan.
NRC Emergency Preparedness Inspections (Reports
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50-423/85-39; 85-66) have found the licensee's emergency planning facili-
ties and controls to be acceptable under current standards.
This item
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is closed.
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F.
1I1.0.3.3 (85-TM-24) Improved Inplant Iodine Instrumentation under-
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Accident Conditions (Closed)
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The licensee has placed airborne monitors in areas that sample the air
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from the reactor containment, the engineering safety. features building,
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the control room, and locations in the reactor plant heating and venti-
lation air streams.
In addition, portable instrumentation is available
for use where workers may be exposed to a high radiation field.
Proce-
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dure AOP'3573, Radiation Monitor Alarm Response, and Procedure EPIP 4225,
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Containment Air Post Accident Sampling, were reviewed with the finding
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that the licensee has met the commitments documented in SER 12.3.4.2.
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This item is closed.
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7.
Licensee Action on Safety Evaluation Report (SER) Open Items
The licensee has presented to the Office of Nuclear Reactor Regulation (NRR)
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a number of physical and procedural items to resolve NRR concerns.
This in-
spection verified the following items.
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A.
(Closed) SER Section 7.4.2.4
Remote Shutdown Operation (85-SE-05) The
licensee has developed two tests for remote shutdown outside the control
room.
TP 5018 was performed during the preoperational test program.
TP 8029 will be performed during power ascension testing.
In addition,
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E0P 3505, " Shutdown Outside Control Room".and E0P 3504, "Cooldown Outside
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Control Room" have been written and are used during shutdown testing.
Regional inspectors witnessed TP 5018 testing and found the results ac-
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ceptable.
TP 8029 testing acceptability is an integral part of the NRC
field inspection program.. Based on the licensee having provided the
committed features and en acceptable testing program which will receive
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separate NRC review, this item is closed (85-SE-05).
8.
(Closed) SER Section 8.3.2.2 0.C. Monitorina (85-SE-06) The generic re-
quirements in IEEE STD 308-1974 state that the DC system (batteries,
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distribution systems and chargers) shall be monitored to the extent that
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it is shown to be ready to perform its intended function. The NRC staff
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has accepted some remote monitoring.
NRC Inspection Report 423/85-54
Page 15 thru 18 discusses the areas of concern.
In addition, Technical
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Specification 3/4.8.2, D.C. Sources, denotes requirements for these sys-
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tems.
The alarms for main board annunciators are found on Main Boards
1-MB10, 1-MBIE, 8-M88A, and 8-M88C.
Operating procedures address re-
sponses to alarms that indicate possible D.C. anomalies.
The inspector
determined that TS 3/4.8.2 was being met.
This item is closed (85-SE-06).
C.
(Closed) SER Section 8.3.2.3, Compliance to Generic Letter 81-04 (85-
SE-07) The license has developed procedures (3500 series) that address
the loss of power.
The inspector reviewed the licensee's program for
classroom training, Lesson Plan 3504, and Simulator Instructors Guide
- CLC(S-11).
The licensee has completed the commitments documented in
the Safety Evaluation Report.
This item is therefore closed (85-SE-07).
D.
(0 pen) SER Section 9.3.2.2, PASS System (85-SE-08) The regional inspec-
tion of PASS systems that verify the 11 criteria of NUREG 0737 Item
II.B.3 is scheduled when enough radionuclides are present in the coolant
to perform a meaningful test.
The resident inspectors witnessed the
licensee's ability to line up systems and obtain a representative PASS
sample.
Mechanical leaks were observed and corrected prior to conclusion
,
I
of this test.
EPIP 4226, Core Damage Estimate Procedure, addresses the
techniques and EPIP 4224, Post Accident Sampling System, address the
l
methods of sampling.
These include such parameters as preplanning, prior
to sampling, of stay times, routes, respiratory protection, and dosimetry.
These items will be inspected at a subsequent inspection.
This item re-
mains open (85-SE-08).
.
E.
(Closed) SER Section 10.4.9, Emeraency Procedure for Backup Water Supply
(85-SE-12) The licensee has described the various sources of water to
maintain the Steam Generator as a heat sink in operational procedure OP
3322.
Paragraph 7.6 describes shifting of auxiliary feedwater pump suc-
tion to the service water system.
Length of stay of the introduced. con-
taminents is addressed in the procedure.
The inspector has witnessed
the licensee's successful placement of the spool pieces to tie the feed-
water suction to the service water system.
The licensee has satisfac-
torily addressed the SER commitment.
This item is closed (85-SE-12).
F.
(Closed) SER Section 15.9.11, Report on Outaaes of ECCS (85-SE-13) The
licensee has issued N0P-2.12, Participation in Industry Data Programs,
such as Nuclear Plant Reliability Data System (NPRDS) and the Generating
Availability Data System (GADS) covering system and component failures,
outages and power reductions, licensee event reports, etc.
The licensee
has implemented the above systems, satisfying their commitment on this
item.
This item is therefore closed (85-SE-13).
G.
(Closed) SER Section 10.4.2, Mechanical Vacuum Pump Exhaust Monitorino
(85-SE-11) The licensee's commitment to continuous sampling for iodines
and particulates in the mechanical vacuum pump has been met.
Procedure
OP 3329, Condenser Air Removal, has been revised to require chemistry
to set up a monitoring system and to remove the monitoring system upon
conclusion of vacuum pump operations.
This item is closed (85-SE-11).
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8.
Contact with Waterford First Selectman
The Senior Resident Inspector met with the Waterford First Selectman in the
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Selectman's office on January 2, 1986.
The meeting was held to maintain com-
munication between town officials and the NRC resident office.
Discussions
centered around the status of Millstone Unit 3 and included the non-radioac-
tive steam discharges with high noise levels that would be associated with
hot functional testing and the power ascension programs.
The inspector asked
the Selectman to feel free to call on the resident staff if any questions
should arise.
9.
Witnessing of Initial Fuel Load
A.
Scope
The resident inspectors, assisted by regional office based inspectors
and resident inspectors from Millstone 1 and 2 and Haddam Neck, observed
the fuel load, which began on November 26 and ended on December 2.
,
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Test procedure 3-INT-4000, the overall controlling document for the fuel
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load process, was approved by the Plant Operations Review Committee (PORC)
I
on November 8, 1985.
This procedure referenced various plant operating
and surveillance procedures for completion of routine requirements as
well as five appendices for completion of specific fuel load tasks.
The
Appendices were: 4002, " Operational Alignment-Nuclear Instrumentation
Systems;" 4003, " Core Load Instruments and Neutron Source Requirements;"
4004, " Inverse Count Rate Ration Monitoring;" 4005, " Initial Core t.oading;" '
and 4006, " Core Map."
This entire 4000 series of procedures had been
reviewed and found acceptable by Region I specialists.
B.
Verification of Activities
The following activities were verified to be in accordance with the
I
facility Technical Specifications and the 4000 series procedures during
the six day fuel loading sequence.
1.
Communications were established via headsets between the spent fuel
'
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crane operator, the Sigma refueling machine operator, a control room
operator and the reactor engineer's staff.
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2.
All fuel movements were tracked on a status board'in the control
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room.
The board indicated fuel pit locations, core locations, and
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fuel-assembly-in-transit locations (i.e., spent fuel crane, fuel
,
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transfer tube, Sigma refueling machine).
3.
The reactor engineer directed fuel assembly movements with the con-
currence of the Shift Supervisor. Watchbills were in force.
Even
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distribution of experienced watchstanders over the 3 shift rotation
I
was verified. Qualified Plant Equipment Operators (PE0s) manipu-
lated the spent fuel bridge and transfer tube machinery.
Licensed
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reactor operators handled the Sigma refueling machine.
A Senior
Control Operator [ Senior Reactor Operator (SRO) License] was on the
refueling machine and in charge of refueling floor (Containment)
operations whenever core alterations were in progress.
4.
Source range neutron level was continuously monitored by audible
count rate in the control room and on the refueling floor.
5.
Containment integrity was set and maintained during core alterations
in accordance with Technical Specification Section 3/4.9.
The
equipment access hatch was closed, the transfer canal was flooded
above the transfer tube elevation, and at least one door to the
personnel access hatch was shut at all times.
The containment purge
and exhaust dampers were open when the exhaust radiation monitor
was in service and closed when the monitor was not in service in
accordance with Technical Specification 3.9.9.
6.
Primary sources (inserts A23PS1 and 823PS2) were inserted in the
correct assemblies (C30 and C04 respectively) which were loaded
first and in close proximity to the two source range excore detec-
tors to put the source range channels on scale.
Neutron dose rates
were monitored by Health Physics personnel during the movements of
these two assemblies.
7.
Class II Housekeeping zones were established and maintained in the
spent fuel pool area and on the refueling floor around and above
the reactor cavity.
8.
The reactor coolant system was flooded to above the loop nozzles
with water having a boron concentration of about 2050 ppm.
The 1.6%
shutdown margin was determined to be 1850 ppm boron.
Sampling and
analysis of boron concentration was performed at different locations
in the system periodically.
Both Residual Heat Removal (RHR) trains
were operable with at least 1 train always in service.
A boron flow
path with two sources of borated water (Refueling Water Storage Tank
and Boron Addition Tank) was always available.
9.
In addition to the permanent source range detectors, 3 portable
submersible " dunkers" were installed in the core and were shifted
around in accordance with the Appendix 4005 loading sequence to
monitor the changing core geometry.
The portable detectors output
went to signal processing equipment, counter / scalers, and strip
chart recorders on the refueling floor.
10.
The reactor engineering staff evaluated the response of each loaded
assembly by extrapolation of the inverse count rate ratio plot be-
fore allowing the loading of the next assembly.
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11.
Excore nuclear instrumentation surveillances were performed at re-
quired frequencies.
The "high flux at shutdown" alarms were in
nurmal for both source range channels for the entire time.
12.
Fuel load test procedure 4000 results reviews are documented in De-
tails llB and 11C.
C.
Problems encountered durina fuel load
1.
ySia Refueling Machine.
The licensee encountered numerous problems
with the refueling crane. While moving to pick up the second fuel
.
assembly, the operators shut the machine down because of a loud
noise whose source could not be determined. Maintenance determined
the cause to be excessive movement of the trolley disc brake mech-
anism causing a flutter when the crane was moved at high speed.
The brakes were adjusted and the operators were cautioned not to
travel at maximum speed.
While the sixth assembly was being lowered into the core, an opera-
tor noticed a loose 1/4 X 20 X 1 1/4" bolt resting on a cluster
plate in the crane insertion tube.
The assembly was lowered.
Fueling was halted for inspection of the Sigma machine.
The absence
of 2 bolts with nuts and washers from a junction box above the in-
sertion tube was revealed.
Engineering inspected the reactor vessel
with an underwater camera but did not find the missing fasteners.
It was noted that the fasteners could have been disengaged prior
'
to fuel load.
The licensee concluded that there was no safety con-
cern from the loss of the missing material.
The missing fasteners
were replaced.
Fuel load was restarted.
The resident inspector
witnessed the discovery of the bolt and examined the area of the
missing fasteners.
There were delays due to difficulties with the positioning and load
limit switches. On one occasion, while lowering a fuel assembly
into the core, the hoist height limit switch went out of adjustment,
,
causing the lower limit to actuate before the assembly was unloaded.
'
The Sigma logic concluded there was nothing for the assembly to land
on.
After visual verification of proper positioning, the assembly
was successfully manually lowered, unloaded, and ungrappled.
i
Finally, the trolley drive motor seriously degraded during the fuel
load.
Prior to core mapping, the drive motor was replaced.
2.
Spent fuel pool crene.
The other major cause of delay in the in-
itial fuel load was spent fuel crane breakdowns.
The major con-
tributor to these was an intermittent open in the load limit control
'
circuit that could not be identified through numerous troubleshoot-
ing attempts.
To correct this problem, the licensee intends to re-
locate the load limit circuitry to a more vibration free area of
the bridge and install a repeater for operator interface.
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3.
Bowed fuel assembly.
On November 30, the operators were unable to
land assembly B49 at core location E-4 due to interference with as-
sembly A35 (location F-4).
It was determined that A35 was bowed
approximately 1/4 inch over its 12 foot length, which was enough
'
.
to keep B49 from contacting its guide pins on the lower core plate.
!
The licensee called a Plant Operations Review Committee (PORC)
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meeting at 0300 on 30 November (meeting 3-85-399), with Westinghouse
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representation, to approve Change 5 to INT-4000, Appendix 4005,
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which modified the loading sequence.
Assembly B49 was temporarily
!
landed at the northern perimeter of the core while the sequence
continued and " boxed in" its E-4 location. On December 1, B49 was
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loaded successfully.
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D.
Conclusion
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Fuel load went well.
Actions were conservative and deliberate.
Operator
'
training and performance were appropriate,
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10.
Preoperational Test Results Review
A.
The inspector reviewed five completed preoperational test procedure
packages ready for licensee review, evaluation, and approval of test
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results as listed below.
(1) Preoperational Test Results Evaluation Reviewed
T3322-P Auxiliary Feedwater System
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T3313-AP Hydrogen Recombiner and Building Ventilation Systems
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T3314-EA Service Building HVAC System
T3314-FP Control Building HVAC System
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T3410-BP Reactor Vessel Level
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No discrepancies were noted and test deficiencies were documented.
B.
The inspector reviewed twenty-three completed preoperational test folders
for deficiencies, including tracking and resolution, as listed below.
(1) Completed Preoperational Test Package Test Deficiencies Reviewed
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T3315-BA Main Steam Valve Building HVAC
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T3316-AP001 Main Steam
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T3322-P Auxiliary Feedwater
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T3330-CP Reactor Plant Chilled Water
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T3330-EP Safety Injection Pump Cooling
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T3307-AP003 Safety Injection Accummulator Test
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T3307-AP001 Low Pressure Safety Injection
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T3306-P Containment Recire Spray
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T3313-FP Containment Vacuum
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3-INT-2001 APP R03 Data Reduction Grid Alignment
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T3308-P001 High Pressure Safety Injection Flow Balance
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T3340-BA Water Treating
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T3341-BP Fire Protection-Halon
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T3344-BA050 MCC-32-1k
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3-INT-2001 App J03 RCS Leakage
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3-INT-2001 Aoo J09 Rx Eng-NSS Data Sheet
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3-INT-2001 R02 Reasonability Check and Time Coverage
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T3308-P002 HP Safety Injection
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T3314-FP Control Building HVAC
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T3321-AP Feedwater and Recire
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3-INT-2001 App J12 RP Prerated. Water Inventory
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No inadequacies were noted in the review of these preoperational test
deficiencies.
The inspector verified the tracking and proper resolution
of the closed test deficiencies.
The remaining open test deficiencies
have been adequately re-tied to later modes and inspections.
11.
Startup Test Program Review
A.
The inspector reviewed three approved and one draft startup test proce-
dures for technical and administrative adequacy as listed below.
(1) Startup Test Procedures Reviewed
3-INT-5000 App 5033, RCS Loop Stop Valve and Pump Interlocks
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3-INT-5000 App 5017, Precritical RCS Flow Coasdown Measure.aent
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3-INT-6000, Initial Criticality-Controlling Procedure
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3-INT-6000 App 6001, Inverse Count Rate Ratio
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No discrepancies were noted in t.he review of these procedures.
B.
The inspector reviewed five completed startup test procedure packages
ready for licensee review, evaluation and approval of test results as
listed below.
(1) Startup Test Results Evaluation Reviewed
3-INT-4000, Initial Fuel Load-Controlling Procedure
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3-INT-4000 App 4003, Core Load Instruments and Neutron Source
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Requirements
3-INT-4000 App 4004, Inverse Count Rate Ratio Monitoring
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3-INT-4000 App 4005, Initial Core loading
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3-INT 4000 App 4006, Core Map
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No discrepancies were noted.
Test deficiencies were documented.
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16
C.
The inspector revit:wed five completed startup test folders for test da-
l
ficiencies, their tracking and resolutions, as listed below.
(1) Completed Startup Test Package Test Deficiencies Reviewed
3-INT-4000, Initial Fuel Load-Controlling Procedure
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3-INT-4000 App 4003, Core Load Init. and Neutron Source Req.
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3-INT-4000 App 4004, Inverse Count Rate Ratio Monitoring
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3-INT-4000 App 4005, Initial Core Loading
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3-INT-4000 App 4006, Core Map
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No inadequacies were noted.
The inspector verified the tracking and
l
proper resolution of the closed test deficiencies.
12.
Test Observation
A.
The inspectors witnessed portions of various Post Core Hot Functional
Tests.
These included INT-5000 Appendix 5008, " Rod Drop Testing," both
!
cold no flow and full flow; INT-5000 Appendix 5033, "RCS Loop Stop Valve
l
Interlocks and Pump Interlocks;" INT-5000 Appendix 5016. " Loose Parts
Monitoring System;" and INT-5000 Appendix 5031, " Chemical and Volume
Control System."
Test perfcrmance was monitored for conformance to test procedures, opera-
tion of equipment in accordance with plant' operating procedures, and good
l
engineering practices.
There were no discrepancies noted.
B.
Secondary Leak Collection and Recovery System (SLCRS) Phase II testing
was closely monitored by the inspectors.
The licensee had numerous
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problems getting SLCRS boundary door seals to function properly and
having fans and dampers operate as designed.
As a result, a number of
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tests failed to meet the acceptance criteria of 0.25" water gauge nega-
tive pressure throughout the containment enclosure and conti uous build-
0
ings within 50 seconds.
Extensive repair and upgrading of equipment was
.
accomplished between test failures.
Prior to the final successful performance of T3314IP, the inspector
walked down all 43 SLCRS boundary doors with the cognizant Startup Engi-
neer.
No discrepancies were noted.
The final test met the acceptance
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criteria.
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THE REMAINDER OF THIS PAGE
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THIS PAGE CONTAINS SAFEGUARDS
INFORMATION AND IS NOT FOR
PUBLIC DISCLOSURE, IT IS
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14.
Reactor Coolant System Pressure Boundary leak
A leak on the B reactor coolant loop RTD bypass manifold high point vent line
was discovered by a Plant Equipment Operator (PEO) on rounds on 1/3/86.
The
leak was very slow (a few drops per hour) and was detected by presence of
boric acid crystal growth on components below the manifold.
The leak was
located at the toe of the socket weld that attached a short run of 3/4" 316
SS pipe to a sockolet attached to the manifold.
The other end of the vertical
pipe has a manual isolation valve topped by a 1500 pst rated blank flange.
There are no supports connected to this pipe run.
Initial inspection revealed a clean crack at the toe of the weld against the
pipe.
It appeared to be fatigue induced.
Licensee vibration analysis indi-
cated that this section of unsupported pipe was in resonance with the B Reac-
tor Coolant Pump (RCP).
That was considered to be the cause of the cyclical
fatigue crack.
The remaining 3 bypass manifold vent lines were vibration
tested and found not to be in resonance with their respective RCPs.
(The B
loop vent line was longer than the other three.)
Corrective Action:
Automated Work Order (AWO) M3-85-00213 was issued to per-
form repairs.
The weld was ground out, with the crack preserved for further
failure analysis, and the 3/4" pipe was shortened by approximately 5 1/2" to
eliminate resonance. A new weld was completed in accordance with ASME Section
XI based repair program EM 31100 and Station Procedure DWP-101 " Fillet Welds
on SS plate, pipe and fittings using GTAW process," Revision 3, dated 2/12/85.
l
The new weld was nondestructively examined by visual and dye penetrant testing
with satisfactory results.
Quality control performed inspections at all
requisite hold points.
Hydrostatic pressure testing was not required because
piping of less that 1" minimal is exempted by ASME Section XI IWA-4000 Section
4400(b)5.
A visual leak test was performed at 400 psi and the AWO remains
open pending a system leak test at 2250 psi.
Vibration measurements were made
on the new configuration with satisfactory results.
The inspector will review
the final operational hydrostatic test results during a subsequent inspection
(IFI 85-74-06).
15.
Reactor Coolant Pump P-1 Snubber Support
Northeast Utilites informed the NRC staff of an error made in the calculations
associated with the design of the load path for the Reactor Coolant Pump P-1
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Snubber support.
The P-1 snubber is one of several large supports associated
I
with each of the four Reactor Coolant Pumps.
The NRC was first informed of
l
this design error during a September 12, 1984 meeting between the licensee
and NRR personnel.
The licensee provided additional details of the problem
l
in a letter to the NRC (Serial A04477) dated December 7, 1984.
In addressing
the issue of not providing a Construction Deficiency Report, the licensee's
stated position was that the error was discovered during the design verifica-
tion process and, since the design was not final, the error was not considered
reportable.
The inspector reviewed this event.
The error was found to have
occurred in a manual calculation which was made to determine loads at the
support during 1974.
It was discovered during re analysis of these loads in
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1981.
At that time, new analytical techniques were available. After correc-
tion of the original error, these new loads were found to exceed the design
,
l
loading of the concrete embedments associated with two of the four reactor
coolant loops.
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!
The error did not appear to be generic to other primary support members be-
cause it occurred during the manual transformation of load coordinates and
resulted from the change of an arithmetic sign for one component of displace-
ment.
The coordinate transformation was needed to allow loads generated by
the cold leg rupture restraint gap analysis to be used in the dynamic analysis.
l
The error was found using advanced computer techniques which had not been
l
available earlier.
These would have disclosed similar errors in other sup-
I
ports.
!
!
Although the new, correct loads were within the capacity of the P-1 snubber,
!
they exceeded the design capacity of the concrete embedments used in Reactor
l
Coolant Loops 1 and 2.
The concrete corbels used to transmit the loads from
l
support P-1 embedment to the containment structure are slightly different in
l
loops 1 and 2, and were found to be limiting.
l
The safety issue was resolved on October 10, 1984 when calculations were com-
i
pleted to show that, in their as-built condition, the loop 1 and 2 embedments
were capable of carrying the correctly calculated loads.
This was due to the
margin which existed in the original installation beyond the design minimum
of the corbel.
The calculations are stated in the " Qualification of the P-1
Snubber Support Bracket", (Calculation Number 12179-NS(B)-158-20C dated Octo-
ber 10, 1984.]
The licensee supported his position that the installed restraint design was
l
not final until 1984 by identifying the number of load calculations made for
l
the P-1 restraint.
From its initial conceptual design, which was made on
1
January 5, 1974, through final stress reconciliation on October 17, 1984,
there have been nineteen (19) load calculations made for the P-1 support,
Sixteen (16) of these were made before finding the error on November 6, 1981.
,
The significance of the problem was stated by the licensee in its letter to
the NRC (Serial B11295) dated September 12, 1984.
The issue of whether the licensee should have submitted a Construction De-
ficiency Report was discussed with the licensee at the Region I offices on
January 9, 1986.
Resolution of this consideration will be documented incident
to that meeting.
16. Observation of Stress Reconciliation Program Repairs
The inspector observed work in progress during the reinforcement of the nozzle
pads on the tube sides of the 4 Containment Recirculation System (RSS) heat
exchangers.
Reinforcement was determined to be necessary as a result of the
!
stress reconcillation program; the as-built loading of these nozzles exceeded
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design bases.
Work was being performed in accordance with E&DCR 1-5-07940
under AWO M3-85-38023.
The inspector observed gusset fltup, control of weld-
ing rod and welding in progress.
There were no discrepancies noted.
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17.
Routine Periodic Inspections
A.
Numerous plant tours were conducted during this inspection period to
i
observe activities in progress and verify compliance with administrative
i
requirements.
Systems and equipment in areas toured were observed for
l
fluid leaks and abnormal vibrations.
Snubbers and restraints were ob-
,
r
I
served for proper conditions.
Plant housekeeping conditions were ob-
f
j
served for cleanliness controls and fire hazard prevention.
'
During a tour of the auxiliary building on 12/20/85 at noon, the inspec-
tor noted 2 valves with significant leakage to atmosphere on the 4'8"
level.
One was the Charging and Volume Control System (CVCS) letdown
l
containment isolation valve (CHS*CV8152), the other was primary grade
I
water (PGS) relief valve (PGS*RV77).
It appeared that both valves had
been leaking for a significant period.
There was a large area of crys-
talline boric acid on the floor under CHS*CV8152 and a deep and wide-
,
spread puddle under the drain funnel leading away from the tailpiece on
,
,
PGS*RV77.
There were no existing trouble reports on either of these
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valves. With the unit in operation, leakage from CHS*CV8152 would be
potentially contaminated.
The inspector will follow this as IFI 85-74-05.
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,
During the period of this report, the inspector made numerous tours of
l
safety-related spaces to inspect work in progress.
He noted that the
auxiliary and service buildings exhibited numerous instances of grafitti
l
and poor cleanup practices.
Observations of painting of buildings and
.
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systems were included in these tours.
In all cases, the inspector ob-
served proper surface preparation prior to application of coatings.
B.
Shift logs and operating records were reviewed periodically to determine
the status of the plant as well as changes in operational conditions
I
since the last log review.
In addition, the following verifications were
'
made; selected Technical Specification limits were met, operating logs
and surveillance sheets were complete and log reviews were conducted by
operating staff, and operating and night orders did not conflict with
technical specification requirements.
No deficiencies were noted.
18.
Review of Plant Events
A.
Diesel Fuel Oil Spill
Approximately 40 gallons of diesel fuel oil were spilled onto concrete
and asphalt surfaces northeast of the emergency diesel generator building
at 2PM on 11/21.
The source of the oil was the flame arrester on B Fuel
!
011 Storage Tank as the tank was overflowed due to an incorrect valve
lineup.
The midshift on 11/21 performed a transfer from A to B storage
1
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tanks by pumping through a cross connect line and overflowing the B day
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tank back to the B storage tank.. This line-up was in accordance with
OP 33468 Rev 0, Change 2, " Diesel Fuel Oil System." Apparently, step
7.5.5 of that procedure was not completed because the system was not
restored to a normal line-up.
Early in the afternoon of that same day,
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Diesel Generator A was being run with Day Tank level control in Automatic.
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However, level in the day tank was constantly lowering, even though the
!
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transfer pump was running.
A Plant Equipment Operator (PEO) was dis-
.
patched to check the position of the transfer pump discharge valve.
He
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found it closed and opened it.
No further checks were made.
The opera-
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tors were confident that valve lineups performed on 11/19 (for B) and
,
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11/20 (for A) diesels were current. "A" day tank level was restored.
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Simultaneously, however, day tank B was overflowing to storage tank B,
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which eventually overflowed into the yard (Violation 85-74-02).
.
l
Immediate corrective action was witnessed by the inspector.
The storm
drains were unaffected-sandbags were quickly placed to minimize oil
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spread.
Fire hoses were broken out and the spill was covered with sand
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and swept up.
The storage tank vault was pumped out to waste drains.
The inspector was concerned that the flame arrestor on the storage tank
overflow might have been compromised.
The licensee showed that the ar-
I
restor was a stacked metal disc type and that saturation was not a
problem.
l
19.
Cleanliness control
A.
During a tour of the Emergency Diesel Generator (EDG) enclosures for
purposes of observation of inspection work being performed on EDG-A, the
inspector noted the absence of material and personnel accountability
while critical portions of the EDG were opened up.
Specifically, 4
,
i
[
crankcase inspection covers (2 on each side) had been removed to permit
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access to the crank shaft, journal bearings and cylinder liners.
The
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catwalks, below the open ports, contained loose tools and debris.
Fur-
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ther, access to the internal area of the engine was uncontrolled.
Main-
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tenance and contractor personnel were not taking the precautions required
i
to prevent items such as pencils and rulers from falling out of pockets
l
while leaning over the crank or looking up into the cylinder liner.
B.
During a routine tour of the lower level of the containment building on
l
November 25, 1985, the inspector noted that some of the Containment
l
Structure Sump (Engineered Safety Features Sump) protective deck plates
i
were out of position.
Further observation revealed that the sump was
i
completely opened and that there was an unobstructed pathway directly
!
to the Containment Recirculation System (RSS) pump suctions.
Various
i
tools and fasteners were randomly placed on the remaining horizontal
,
plates above the sump and on the sump floor.
There were no personnel
present at the sump at the time of this observation.
Licensee management
i
was immediately informed of this situation.
'
Work was being performed by construction forces under Construction Work
.
Permit (CWP) M3-85-36279, authorized for performance on 11/13/85, to re-
'
locate the sump level instrument 3RSS-LE49 as per Engineering and Design
Change Request (E&OCR) TC-05339.
That post-turnover work order did not
specify any housekeeping / material accountability requirements.
Stone
and Webster Specification 2200.000-914, " Mechanical Equipment Erection,"
r
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22
,
,
1
Section 3.4, " Housekeeping Requirements," states that housekeeping during
!
construction shall be in accordance with ANSI N45.2.3, " Housekeeping
' During the Construction Phase of Nuclear Power Plants." ANSI N45.2.3
spectried material and personnel accountability for Housekeeping Zone
III systems.
Station Administrative Control Procedure ACP-QA-2.02C,
" Work Orders" lists the Containment Recirculation System (RSS) as a Zone
l
III system.
Immediate corrective action was taken by the licensee to cover the RSS
pump suctions with plywood taped to the sump floor.
Subsequently, the
suction lines were inspected using a boroscope with no foreign materials
detected.
Later in the report period, the inspector again observed that work was
!
being performed in the RSS Sump.
New grating clamps were being installed
on the RSS Sump Vorte>: Grating under PMMS Work Order MS-85-39595.
This
order was written to maintain cleanliness per Specification M914, (Stone
and Webster Specification 2200.999-914), with specific instructions that
the sump be lined with poly with plywood covers over the suction pipes.
However,, worker conformance to these administrative stipulations did not
appear to be thorough.
The material accountability list, posted inside
th.e protective tent over the sump access, had a list of tools with no
indication of when they were checked in or out.
Further, there were rags
and knee p' adding in the sump that were not included on the list.
The
pump suction penetrations were covered with plywood.
The inspector ex-
pressed further concern to licensee management that cleanliness require-
ments were not being followed carefully enough.
The failure to maintain material and personnel accountability for both
'
the EDG sump and the RSS sump are a violation (85-74-01).
20.
Exit Meeting
At periodic intervals during the course of this inspection, meetings were held
with senior plant management to discuss the scope and findings of this in-
spection.
No proprietary information was identified as being in the inspec-
tion coverage. At no timeiduring the inspection was written material provided
to the licensee by the inspector.
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Transaction Typ2 M-3 NRC:1 Fcnn 6 Rev. Oct 80 (Std21) . , O N:w item p y/.
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