ML20148K651
| ML20148K651 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 06/09/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20148K635 | List: |
| References | |
| 50-348-97-05, 50-348-97-5, 50-364-97-05, 50-364-97-5, NUDOCS 9706180232 | |
| Download: ML20148K651 (44) | |
See also: IR 05000348/1997005
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U.S. NUCLEAR REGULATORY COMMISSION (NRC)
REGION II
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Docket Nos:
50-348 and 50-364
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License Nos:
Report No:
50-348/97-05 and 50-364/97-05
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Licensee:
Southern Nuclear Operating Company Inc.
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Facility:
Farley Nuclear Plant (FNP), Units 1 and 2
Location:
7388-North State Highway 95
Columbia, AL 36319
Dates:
March 30 through May 10, 1997
Inspectors:
T. Ross. Senior Resident Inspector
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J, Bartley, Resident Inspector
R. Ca Mwell, Resident Inspector (In training)
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G. Kuzo. Health Physics Ins)ector.(Sections
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R1.1, R1.2 R2.1, R5.1,
R8.1-8,5)
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W. Kleinsorge Reactor Inspector (Sections
M1.10 - M1.12)
R. Chou, Reactor Inspector (Section E1.1)
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M. Ernstes. Operator Licensing Examiner
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(Section 0.5)
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Approved by:
P Skinner. Chief, Projects Branch 2
Division of Reactor Projects
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Enclosure 2
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9706180232 970609
ADOCK 05000348
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EXECUTIVE SUMMARY
Farley Nuclear Power Plant. Units 1 and 2
NRC Inspection Report 50-348/97-05, 50-364/97-05
This integrated inspection included aspects of licensee operations,
engineering, maintenance, and plant support.
The report covers a 6-week
period of resident inspections.
Ooerations
0)erator attentiveness to MCB annunciator alarms and response to
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clanging plant conditions were prompt. Interviews with members of the
operating crew revealed that they were consistently aware of plant
conditions and ongcing activities (Section 01.1).
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Unit 1 refueling activities were well-controlled (Section 01.2).
Housekeeping and physical conditions were adequate, although certain
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areas remained poor.
Licensee efforts to improve targeted areas was
evident.
Improvement in overall plant appearances and material
conditions has increased (Section 02.1).
Safety system walkdowns and tours verified accessible portions of
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selected systems were well maintained and operational (Sections 02.1 and
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02.2).
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Tag orders were properly executed, with one identified exception
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(Section 02.3).
A violation was identified for failure to notify the NRC of a change in
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a licensed Senior Reactor Operator's medical status (Section 05.1).
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Licensee efforts to identify, resolve, and prevent problems remained
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effective, with one identified exception (Section 0/.1).
Maintenance
Maintenance and surveillance testing activities were generally conducted
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in a thorough and competent manner by qualified individuals in
accordance with plant procedures and work instructions (Sections M1.1
through M1.7).
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Over the years, numerous foreign objects and materials have been allowed
to enter the emergency core cooling system containment sumps.
Licensee
efforts to clean, inspect, and repair the sumps have been very thorough
to date (Section M1.7).
Enclosure 2
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A violation was issued for failing to include several specific Technical
. Specifications (TS) surveillance requirements in the Surveillance Test
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Program.
Examples of other missed TS surveillance requirements have
been identified in recent reports.
Load rejection testing of emergency
diesel generators (EDGs) was inconsistent with the licensing basis
(Section M1.8).
The inspectors * issues regarding compliance and conformance with the
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licensing basis of TS 4.8.1.1.2.e are identified as unresolved item
(URI) 50-348, 364/97-05-04 EDG 50% Load Reject Surveillance Testing.
pending the NRC's response to the licensee's TS interpretation request
and proposed TS amendment (Section M1.8).
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Inservice inspection activities observed / reviewed were conducted in
accordance with procedures, licensee commitments and regulatory
requirements (Section M1.10).
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A violation was identified associated with the control of the special
process of welding (Section M1.11),
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A weakness was identified associated with the licensee's control of
contracted welders (Section M1.11).
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Maintenance was subjected to independent audits, with appropriate action
generally taken for identified weaknesses (Section M1.12).
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The issue of control of painting inside the penetration room boundary
(PRB) being limited to less than 1000 ft' in any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period was not
documented by an existing analysis. This is identified as URI
50-364/97-05-06, Painting Effects On PRF Operability pending review of
the licensee's analysis (Section M8.1).
Enaineerina
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Calculations analyzing the SG loop C narrow support gap on Unit I were
acceptable.
Licensee promptly evaluated the potential for past damage
and restored the intended conditions by modifications (Section E1.1).
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Penetration room filtration (PRF) system licensing bases operability is
identified as URI 50-348, 364/97-05-07. Licensing Basis for PRF System
During Post-LOCA Recirculation, pending additional review by the NRC
(Section E1.2).
Plant Sucoort
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Radiological controls were good for routine operations and U1RF14 outage
activities. All personal exposures were within 10 CFR Part 20 limits.
Implementation of established ALARA program activities was verified
(Section R1.1).
Enclosure 2
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Service air compressor system supplied Grade D respirable air in
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accordance with 10 CFR 20 Appendix A requirements (Section R1.2).
Training and medical certifications for personnel using respiratory
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protective equipment were conducted in accordance with the licensee
procedures and met the a)plicable requirements of 10 CFR Part 19 and
10 CFR Part 20 (Section
RS).
Health Physics control over the radiologically controlled area, and the
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work activities conducted within it, was good.
Some contaminated areas
were cramped and physically restricted removal of anti-contamination
clothing (Section R2.1).
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Security activities continued to be performed in a conscientious and
capable manner, assuring the physical protection of protected and vital
areas.
Problems with improper display of security badges by workers in
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containment were resolved promptly (Section S1.1).
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Significant weakness identified in the implementation of Open Flame
Permits (Section F1.1).
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Life safety exits from the Turbine Building were inadequate (F2.3)
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Enclosure 2
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Report Details
Summary of Plant Status
Unit 1 was shutdown for its 14th refueling outage (U1RF14) during the entire
inspection report period.
U1RF14 was rescheduled to be completed in 65 days
instead of the original 55 days due to increased scope of steam generator (SG)
tube inspection and repair work.
Unit 2 operated continuously at 100% power for the entire inspection period,
except for a brief power reduction to 85% power during the weekend of May 3.
1997. for main turbine generator (MTG) governor valve testing.
I. Operations
01
Conduct of Operations
01.1 Routine Observations of Control Room Ooerations
a.
Insoection Scoce (Insoection Procedure (IP) 71707)
Inspectors conducted frequent inspections of ongoing plant operations in
the Main Control Room (MCR) to verify proper staffing, operator
attentiveness, adherence to approved operating procedures,
communications, and command and control of operator activities.
Inspectors reviewed operator logs and Technical Specifications (TS)
Limiting Condition of 0]eration (LCO) tracking sheets, walked down the
Main Control Boards (MC3), and interviewed members of the o)erating
shift crews to verify operational safety and compliance witi TSs.
The
inspectors attended morning plant status meetings and shift turnover
meetings to maintain awareness of overall facility operations,
maintenance activities, and recent incidents.
Morning reports and
Occurrence Reports (ors) were reviewed on a routine basis to assure that
the licensee properly reported and resolved potential safety concerns.
b.
Observations. Findinas and Conclusions
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Overall control and awareness of plant conditions during the inspection
period remained adequate.
Inspectors observed that the Unit 2 MCBs and
emergency power board were nearly " blackboard" most of the inspection
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period, with only a couple of persistent annunciators in alarm that were
recognized deficiencies.
Efforts to maintain MCB deficiencies at low
levels continued.
MCB deficiencies on Unit 2 increased to 15 or more
and have remained there.
Almost all of them involved nonsafety-related
instrumentation or equipment.
Operator attentiveness to MCB annunciator
alarms and response to changing plant conditions were promot. Interviews
with members of the operating crew revealed that they were consistently
aware of plant conditions and ongoing activities.
Pre-shift briefs of
the operating crews by the shift supervisors (SS) were generally concise
and informative.
Operator logs generally were of sufficient detail and
scope, with one notable exception.
Surveillance test procedure (STP)
FNP-1-STP-40.1. Revision 27. "B1F and B1H Sequencer Load Shedding
Enclosure 2
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Circuit Test." was performed on April 12. 1997, but was not mentioned in
the Unit 1 reactor operator (RO) logs.
This omission was discussed with
Operations management.
01.2 Unit 1 Refuelina (IP 60710)
A resident inspector observed refueling activities from the MCR, Spent
Fuel Pool (SFP). and MCR balance of plant area, on 18 April,1997.
The
Westinghouse Refueling Manual FP-ALA-R14. Revision 0. J.M. FARLEY
Unit 1 Cycle XIV - XV Refueling, conducted under WA# WOO 475930, was
reviewed.
Refueling activities observed by the resident inspectors were
performed in a well controlled and methodical manner in accordance with
FNP-1-U0P-4.1
" Controlling Procedure for Refueling and the Westinghouse
Refueling Manual." Revision 9.
Communications between the various
stations were clear and concise.
The subcritical multiplication plot
(1/M plot) was accurate and timely.
Personnel were very familiar with
the procedure and knowledgeable of the process and systems.
No
significant incidents occurred during fuel handling and all observed
fuel assemblies were landed in their appropriate locations.
The
inspector concluded that fuel handling was accomplished in a
professional and competent manner.
02
Operational Status of Facilities and Equipment
02.1 General Tours of Soecific Safety-Related Areas (IP 71707)
General tours of safety-related areas were performed by the inspectors
to examine the physical condition of plant equipment and structures, and
to verify that safety systems were properly aligned.
These general
walkdowns included the accessible portions of safety-related structures,
systems, and components in the following areas:
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Unit 1 containment
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Unit 1 and 2 SFP, SFP heat exchangers (HXs), and SFP cooling pump
rooms
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Unit 1 and 2 main steam valve rooms
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Unit 1 and 2 piaing penetration rooms (PPR) on 100 foot elevation
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Unit 1 and 2 PPRs on 121 foot elevation
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Service water intake structure (SWIS)
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Unit 1 component cooling water pump and heat exchanger (HX) rooms
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Unit 1 and 2 vital 125 volt direct current (VDC) switchgear and
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battery rooms
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Unit 1 and 2 new fuel storage areas
Unit 1 and 2 service and instrument air compressors, dryers and
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receivers
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Emergency diesel generator (EDG) building
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Unit 1 and 2 containment spray (CS)
aump rooms
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Unit 1 and 2 residual heat removal (RHR) HX rooms
Unit 1 and 2 RHR pump rooms
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Unit 1 and 2 charging pump rooms and hallway
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Enclosure 2
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Turbine building
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Unit 1 and 2 penetration room filtration.(PRF) system rooms-
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Unit 2 hot shutdown panel (HSDP) rooms
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Unit 1 filter gallery on 139-foot elevation
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Unit 1 vital 4160 volt alternating current (VAC) switchgear rooms,
trains A and B, including vital 600 VAC load centers
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General material conditions and housekeeping for Unit 2 were adequate.
-Areas were generally clear of trash and debris.
Attempts to control the
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impact of Unit 1 outage activities were obvious, as were additional
efforts at housekeeping in the outage unit.
Considerable effort to
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improve physical appearances of plant areas and equipment was in
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progress, 3rimarily in the form of extensive painting in the EDG and
auxiliary )uildings. These efforts have dramatically improved the
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appearances of rooms, structures and equipment in targeted areas.
Minor
equipment and housekeeping problems identified by the inspectors during
their routine tours were reported to the responsible SS and/or
maintenance department for resolution.
Maintaining all critical areas
of the auxiliary building well lamped remained an ongoing challenge.
None of these problems represented operability concerns.
02.2 Biweekly Insoections of Safety Systems (IP 71707)
Inspectors used IP 71707 to verify the operability of the following
selected safety systems and/or equipment:
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Unit 2 Hot Shutdown Panels
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Unit 1 Accumulators
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Unit 1 Containment Spray. Trains A and B
Accessible portions of the systems listed above were verified to be
properly aligned and ap) eared to be well maintained and in good
operating condition.
T1e inspectors did not identify any significant
issues that adversely affected system operability.
In addition,
accumulator pipe supports and components were walked down using
isometric system drawings.
02.3 Tao Orders (IP 71707)
During the course of routine inspections, portions of the following tag
orders (TO) and associated equipment clearance tags were examined by the
inspectors:
e TO# 97-0757-1: RHR
e TO# 97-1059-1: IB PRF EQ Upgrade
e TO# 97-1058-1: 1A PRF EQ Upgrade
e TO# 97-442-1: Service Water System (SWS) pumps 1C. 10, and 1E
e TO# 97-178-1: CS System and Chemical Addition Tank
e TO# 97-1165-1: SW to Containment Coolers 1A and IB
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Enclosure 2
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All tags and TOs examined by the inspectors were pro)erly executed and
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implemented. However, one . tag order was identified )y the inspectors;
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that used an air-operated valve, which fails to the open position, as
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.part of the clearance boundary.
While walking down TO 97-1059 on May 8, 1997, the inspector noted that
01V48HV3538B, SFP.to 1B PRF Supply Damper, was tagged shut per the tag
order. This valve is air-operated and thus depended on air pressure to
maintain itself in the tagged closed position.
Furthermore, because of
the type of work being done on the IB PRF filter housing, this valve was
required to remain closed to ensure the 1A PRF system was not rendered
inoperable by short-circuiting the suction flow path from the s)ent fuel
pool.
The inspector interviewed the T0 author and determined t1at he
was not aware that HV3538B was a fail open valve nor was he aware of the
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potential for compromising the operability of the other PRF train.
During discussions with additional operations personnel, the inspector
determined that other operations staff were also not aware of the
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)otential for short-circuiting the SFP supply to the 1A PRF system if
iV3538B had failed open.
When informed of the inspector's concern
regarding opeNbility of the 1A PRF system, the SS premntly revised the
T0 to jack shut HV3538B and initiated OR 1-97-191.
Ine inspector
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determined, through interviews with licensee personne . ,nat loads were
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not moved in the spent fuel pool area while either 1A he or 1B PRF were
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tagged out for the environmental qualification upgrade. Therefore, no
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TS LC0 action statements were required to be entered while the PRF was
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potentially inoperable.
02.4 TS LCO Trackina (IP 71707)
The inspectors routinely reviewed the TS LC0 tracking sheets filled out
by the shift foremen.
All tracking sheets for Unit 1 and 2 reviewed by
the inspectors were consistent with plant conditions and TS
requirements.
05'
Operator Training and Qualifications
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05.1 Conditions of Ooerator Licenses
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a.
Scooe (IP 71001)
The inspectors reviewed 10 CFR 55 and NRC Information Notice 94-14.
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" Failure to implement requirements for biennial medical examinations and
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notification to the NRC of changes in licensed operator medical
conditions." which was issued to remind licensees of the requirements to
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notify the NRC of changes in a licensed operators physical or mental
condition.
10 CFR 55.21, Medical examination, requires an NRC-licensed
operator to be examined by a physician every 2 years to determine if the
individual meets the requirements of 10 CFR 55.33(a)(1).
If, during the
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term of the license the operator develops a permanent physical or
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mental condition that causes the operator to fail to meet the
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Enclosure 2
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requirements of 10 CFR 55.21. the facility licensee shall, in accordance
with 10 CFR 55.25. notify the Commission within 30 days of learning of
the diagnosis.
b.
Observations and Findinas
On April 9,1997, an NRC inspector reviewed applications for renewal of
Senior Reactor Operator licenses. The NRC Form 396. " Certification of
Medical Examination by Facility Licensee." of one of the applicants
indicated that he needed corrective lenses to meet the requirements of
However, the individual's current Senior Reactor
Operator license contained no restrictions.
Based upon a telephone
conversation with Farley Training Department personnel, it was
determined that the individual had been diagnosed as needing glasses
during a biennial physical conducted on March 21, 1995.
However, the
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facility licensee failed to notify the NRC of the change. The licensee
conducted a review of all other licensed operators' medical conditions
to determine if any other changes in medical conditions had failed to be
reported.
None were found.
c.
Conclusion
The licensee's failure to notify the NRC within 30 days of a permanent
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change in a licensed operator's medical status is identified as
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violation (VIO) 50-348, 364/97-05-01. Failure to Notify NRC of Change of
Licensed Operator Medical Status.
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07
Quality Assurance in Operations
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07.1 Effectiveness of Licensee Control in Identifyina. Resolvino. and
Preventina Problems (IP 71707 and 40500)
The inspectors briefly reviewed all newly initiated ors and completed
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ors approved during the inspection period to ensure that plant incidents
which affect or could potentially affect safety were properly documented
and processed in accordance with Administrative Procedure (AP)
FNP-0-AP-30. " Preparation and Processing of Incident Reports."
Revision 22.
Selected ors that had been completed were reviewed in
detail.
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The inspectors concluded that the licensee's program for identifying and
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resolving problems remained effective and was being accomplished in
accordance with FNP-0-AP-30.
Plant personnel and management exhibited
an appropriate threshold for identifying problems, initiating ors. and
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assigning formal root cause determinations.
Each new OR received prompt
attention and was discussed in the morning status / plan of the day
meeting.
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Enclosure 2
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Inspectors reviewed the following ors for accuracy, completeness and
reportability. and adequacy of corrective actions:
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OR #1-9'7-132:
Non-temperature compensated Heise gauges
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OR #1-97-191:
PRF T0 deficiency
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OR #1-97-198:
DG 1B Rollup door stuck open
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These ors were properly processed to completion.
In general. licensee
corrective actions for resolving problems continued to remain effective.
However. OR 97-132 was initiated to address an instance where approved
corrective actions were not fully implemented.
On April 9. 1997, while
touring the Unit 1 Auxiliary Building, the inspector found four non-
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temperature compensated 0 - 100 psig Heise test gauges which were not
labeled with a restricted use tag. The gauges were staged for use in
local leak rate testing. The use of non-temperature compensated Heise
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gauges without restricted use tags was contrary to the corrective
actions identified in Corrective Action Request (CAR) 2173 for
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VIO 50-348, 364/95-18-03.
Licensee staff immediately impounded the four
gauges when informed by the inspector and initiated OR 1-97-132.
A
re-audit identified four additional non-temperature compensated Heise
test gauges in the Calibration Lab.
The licensee investigation
determined that these eight gauges, which were used by the Systems
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Performance Group for local leak rate testing (LLRT), were stored under
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a green tarp in a mechanical room. They were overlooked when the
actions for CAR 2173 were performed.
As further corrective action for
this event, the licensee: 1) performed two independent audits which did
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not identify any more test equipment re
to lack of temperature compensation 2) quiring restricted use tags due
modified calibration equipment
checkout procedures to check for temperature compensation, and 3)
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evaluated all work activities which used any of these eight gauges. The
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inspectors independently reviewed work activities which used these
gauges and determined that they were not invalidated by the use of a
non-temperature compensated gauge.
Failure to fully implement all of
the corrective actions identified by CAR 2173 and NOV response letter
dated December 19, 1995 constitutes a minor violation consistent with
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the guidelines of Section IV of the NRC Enforcement Policy.
This
non-cited violation (NCV) is identified as NCV 50-348, 364/97-05-09.
Failure to Fully Implement Corrective Actions.
08
Hiscellaneous Operations Issues (92901)
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08.1
(Closed) LER 50-348/97-002: Safety-Related 4160 Volt AC Breakers Not
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Seismically Oualified
(Closed) LEP 50-348/97-004: Safety-Related 600 Volt AC Breakers Position
Sensitive Seismic Oualification
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Inspectors verified that 4160 and 600 Volt AC safety-related breakers
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that can not be maintained in their seismically qualified positions were
removed from their respective cubicles. placed an appropriate distance
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Enclosure 2
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away, and properly secured. A few minor discrepancies (e.g., lack of
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blocking bolts on 4160 VAC breaker assembly wheels) were initially
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identified by the inspectors and promptly corrected.
No repeat problems
were found.
The inspectors also reviewed applicable operating procedure
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FNP-0-SOP-36.6. " Circuit Breaker Racking Procedure." Revision 12 and
Temporary Change Notice (TCN) 12A and interviewed responsible personnel
regarding proper handling and positioning of these breakers. One minor
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procedure deficiency was identified and discussed with Operations
management regarding . inadequate instructions for removing 600 volt
breakers and placing them a prescribed distance away. Although
different than the corrective actions described in LER 97-004
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Operations has decided it would rather place 600 volt load center
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breakers in spare cubicles rather than on the floor, whenever possible.
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By the end of the report period, applicable procedures were being
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revised to allow this option. These LERs are closed.
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08.2 (Closed) LER 50-364/96-001: Reduction /Restmotion of 2B Diesel Generator
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Soeed Caused By Inadeauate Procedural Guicance
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This LER was considered a minor issue and closed.
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II. Maintenance
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M1
Conduct of Maintenance
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M1.1 General Comments
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a.
Insoection Scoce (IP 61726 and 62707)
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Inspectors observed and reviewed portions of various licensee corrective
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and preventive maintenance activities, and witnessed routine
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surveillance testing to determine conformance with plant procedures.
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work instructions, industry codes and standards. TSs. and regulatory
requirements.
The inspectors observed all or portions of the following
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maintenance and surveillance activities, as identified by their
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associated work order (WO). work authorization (WA), or surveillance
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test procedure (STP):
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o WO# M96004128
ICCMS Probe Changeout
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e WA# 00467851
01P16MOV3019B MOVATS testing
Sequencer Panel B2H Sequence Start Undervoltage
Relay 0A-C
Sequencer Panel B1H Sequence Start Undervoltage
Relay OB-C
Sequencer Panel B1J Sequence Start Undervoltage
Relay OB-C
e WO# S00080418
e FNP-1-STP-80.11
DG 1B 1200 (W Load Rejection Test. Revision 7
Enclosure 2
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e FNP-2-STP-16.2
Containment Spray Pump 2B Inservice Test,
Revision 23
e FNP-0-MP-14.19
Emergency Diesel Generator 1-2A, 18, and 2B
Removal and Inspection of Air Start Check
Valves, Revision 2
e FNP-0-IMP-226.7B
Diesel Generator Single Circuit Emergency Start
Test, Revision 7
e FNP-0-SOP-38.0
Diesel Generators, Revision 56
J.M. FARLEY Unit 1. Cycle XIV - XV Refueling,
Revision 0, FCR 08 Westinghouse Refueling
Manual
e WO #S00080409
1A PRF E0 Upgrade. DCP No. B97-1-9157
e WO #S00080408
1A PRF EO Upgrade, DCP No. B97-1-9157
e WO #S00080411
Rewire 1A PRF heater coil, DCP No. B97-1-9157
e WO #S00080410
Rework 1A PRF fan motor 01E15M001A(2A) DCP
No. B97-1-9157
e FNP-1-STP-228.5
NIS Power Range Channel N41 Calibration
N1C55NE0041. Revision 38
e FNP-1-STP-40.0
Safety Injection with Loss of Off-Site Power
Test, Revision 31
e FNP-1-STP-40,0
Retest of 01P16MOV3134, SW from RCP MTR CLRS
(Appendix S)
under WO #97003589
e FNP-1-STP-256.15
Loss Of Offsite Power Response Time Test,
Train A. Revision 14 (TCN 14B) under
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e FNP-1-STP-16.1
Containment Spray Pump 1A Inservice Test,
Revision 29 under WO #0079666
e FNP-0-MP-14.1
Emergency Diesel Generator 1-2A, 1B & 2B Refuel
(18 Month) Inspections Revision 23
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e FNP-0-MP-84.4
Emergency Diesel Generator Vibration
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Measurements, Revision 4
e FNP-0-SOP-42.0
Diesel Generator Fuel Oil Storage & Transfer
System, Revision 2C
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e FNP-2-STP-24.1
2A, 28, and 2C Service Water Pump Quarterly
Inservice Test. Revision 23
e WO #S0068708, 11
SW Supply and Return Valve Replacements for
and 12
Containment Coolers IA and 10
e FNP-1-STP-80.14
Diesel Generator A Train Loss of Offsite Power
Test
e FNP-1-STP-80.15
Diesel Generator B Train Loss of Offsite Power
Test
e WA #W00477033
Preventive Maintenance (PM) Lubrication of 2C
Charging Pump
b.
Observations. Findinas and Conclusions
All of the maintenance work and surveillance testing observed by the
inspectors was performed in accordance with work instructicos,
procedures, and applicable clearance controls.
No adverse findings were
identi fied.
Safety-related maintenance and surveillance testing
Enclosure 2
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evolutions were well planned and executed.
Personnel demonstrated
. familiarity with administrative and radiological controls. . Surveillance
tests of safety-related equipment were consistently performed in a
deliberate step-by-step manner by personnel in close communication with
the MCR.
Overall, operators and technicians appeared knowledgeable.
experienced. and well trained for the tasks they performed.
M1.2 IB D/G Outaae Maintenance
'The inspector observed portions of the 18 Month 1B EDG maintenance and
testing work from March 30 - A)ril 4.1997. Specific evolutions
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observed include portions of tle following:
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Removal and inspection of air start check valves.
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Replacement of Lube Oil HX and Jacket Water HX tube bundles due to
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inlet tube sheet erosion concerns,
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Diesel Generator circuit emergency start inspection done under WO #470331.
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Various other refuel (18 Month) inspections.
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Emergency Diesel Generator vibration measurements.
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Inspection of the Diesel Generator fuel oil storage and transfer
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system.
All work was completed per procedure in a professional manner.
Post
maintenance testing was completed satisfactorily.
M1.3 Manual Safety In.iection (SI) Hand Switch Test
The inspectors observed the test in the MCR and at the reactor trip
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breakers.
The test was performed in accordance with FNP-1/2-STP-38.3.
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" Verification of Manual SI Actuation Input." Revision 1.
All components
responded as required.
Refer to section M8.2 for more details.
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M1.4 Unit 1 Gammametrics Probe Reolacement
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The inspectors observed maintenance activities involving the pre-job
planning, removal, and installation of the Gamma-Metrics Detector.
01C55NE0048. This work was performed under WA #00464676. . Pre-job
planning was thorough.
Licensee personnel were knowledgeable and
sensitive to ALARA planning.
Maintenance and HP interfaced well to
minimize radiation exposures. Work was conducted per the WA and the
staff responded well to unexpected situations such as the detector being
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much more activated than expected.
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Enclosure 2
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M1.5 Unit 1 Containment Cooler Coil Reolacement
An inspector observed portions of the work to remove all the original
containment cooler coils and replace them with new coils according to
design change package (DCP) 96-1-9054. All work witnessed by the
inspector was accomplished in accordance with authorized work orders.
Good cleanliness control was exhibited to prevent introducing foreign
material into the cooler piping.
During modification work, the
licensee's activity task manager identified two problems that warranted
being documented as ors: 1) Original cooler coils were missing mounting
bolts that attached the coils to the cooler frame (OR# 97-94) and 2)
Baffle gaskets were missing from some of the replacement coils
(OR# 97-151).
The inspector discussed each of these problems in detail
with the activity manager. After the containment cooler coils were
replaced, and the system returned to service, the inspector conducted a
walkdown of all four containment coolers. The inspector also reviewed
the post-modification testing (PMT) performed in accordance with
FNP-1-STP-17.0, " Containment Cooling System Train A(B) Operability
Test." Although both trains of containment coolers met the acceptance
criteria of STP-17.0, the inspector discussed testing the coolers
individually in order to verify sufficient SW flow through each one.
After further review, the licensee expanded the PMT to measure flow
through each cooler per WO #S97004055.
The inspector reviewed the
results which verified adequate flow.
M1.6 Unit 1 CS Addition Tank Deletion and Trisodium Phosohate Basket
Installation
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On May 8 a resident inspector observed the visual examination (i.e..
VT-2) performed on welds associated with the Unit 1 "A" train Spray
Additive Tank removal design change, under WO# 0079666. The 1A
Containment Spray pump was started per FNP-1-STP-16.1, " Containment
Spray Pump 1A Inservice Test." Revision 29, and the weld checked at
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operating pressure. A quality control (OC) ins)ector conducted a
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detailed visual check of the weld area and blanc flange.
The OC
inspector and resident inspector observed no leakage.
A resident ins)ector conducted a walkdown of the new trisodium phosphate
(TSP) baskets Juilt on the containment basement floor in accordance with
DCP 95-1-8931.
The ins)ector also observed licensee efforts on May 13
to load 70 fifty pound Jags of TSP into each of the three baskets.
Upon
the completion of this work the inspector noticed and discussed the
following concerns with the responsible design change engineer:
1) All
three TSP baskets were slowly leaking at the bottom corners. 2) Actual
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level of TSP in the baskets did not rise to the minnmum TS level marked
on the baskets, and yet the quantity loaded should have been sufficient,
3) Accuracy of TSP bag weight, and 4) Maintaining inventory control of
material lost during basket leak / repair.
Each of these observations was
,
adequately addressed by the licensee.
Subsequent walkdowns of the
Enclosure 2
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finished and filled baskets by the inspector did not identify any
additional concerns.
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M1.7 Emeraency Core Coolina System Containment Sumo Insoections and Foreian
Material Retrieval
On April 24, 1997 an SAER auditor and PMD engineer discovered that the
Unit 1 emergency core cooling system (ECCS) sump screens were not built
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per design. Two of the four ECCS sumps in containment were missing a
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structural, flat bar of. steel. OR #97-172 was initiated. Thorough
walkdowns of the Unit 1 ECCS sumps conducted by an inspector and
licensee personnel also identified numerous gaps in the sump screens
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created by poor fitup. Many of the gaps were on the order of 1" x 1,"
which are considerably larger than the 1/8" x 1/8" screen mesh openings.
On May 2, the licensee made a containment entry into Unit 2 to examine
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its ECCS sumps. Although gaps of comparable size were also discovered
in the Unit 2 screens, they were much fewer in number.
This inspection
also included the interior of the Unit 2 sumps and identified two pieces
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of foreign material that were removed.
OR #97-183 was initiated for
Unit 2.
The Unit 2 ECCS sum) screen gaps were promptly repaired.
returning the ECCS sumps bacc to original design. An operability
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determination (0D) 97-10 was developed by the licensee and reviewed by
the inspector.
It determined that operability of ECCS equipment had not
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been adversely affected.
The inspector discussed with the licensee the
aotential that foreign material may have also fallen down inside the
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ECCS sump suction piping for the CS and RHR systems.
Beginning May 8. the inspector observed maintenance personnel conduct
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internal visual inspections of the Unit -1 ECCS sump, both the interior
grating and suction piping. A boroscope was used to inspect inside the
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suction )iping.
Numerous articles of foreign debris were discovered
inside t1e interior grating (e.g. , many different lengths of duct tape,
metal washers, marking pen, weld rod segment, etc.) and sump suction
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piping (e.g.. many different lengths of duct tape, short piece of
string, small strip of plastic, etc.).
The boroscope used by the
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maintenance personnel could only inspect about six feet of the
hori2ontal run of piping.
By the end of the-inspection period, the
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licensee was still cleaning out the foreign debris from the Unit 1 sumps
and enlisting the aid of a contractor who could exam the entire run of
suction piping.
Operations was also pursuing an OD to address the
operability impact of the discovered material in Unit 1.
Plans for
,
conducting internal visual examination of Unit 2 interior grating and
sump suction piping was still being discussed.
In order to track the
licensee *s ongoing corrective actions and operability evaluations this
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issue is-identified as inspector followup item (IFI) 50-348,
364/97-05-02. Foreign Material In Containment ECCS Sumps.
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M1.8 Failure to Meet Multiole TS Surveillance Recuirements
During the current Unit 1 outage (U1RF14), the licensee identified the
following TS Surveillance Requirements that were not being adequately
implemented and/or missed entirely:
a)
TS Table 4.3-1. Reactor Trip Instrumentation Surveillance
Requirements. Functional Unit 6 requires a quarterly channel
,
functional test and shiftly channel checks of the nuclear
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instrumentation system (NIS) source range (SR) channels while in
Modes 2. 3. 4. and 5.
On March 19. 1997, during a review of its
Improved TS submittal, the licensee determined that when Unit 1
was shutdown on March 15 for U1RF14 the unit had entered Mode 2.
and then several minutes later Mode 3. without accomplishing the
surveillance requirements for NIS SR.
The TS surveillance
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recuirements were not met until March 18 while the unit was in
Moce 5 after NIS SR N-32 was functionally tested.
OR# 97-139 was
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initiated and LER 97-05 was written to document, resolve and
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report the incident.
Since the root cause was considered to be
inadequate test procedures, the licensee also concluded that the
failure to meet these TS surveillance requirements has occurred
numerous times in the past during previous Unit 1 and 2 shutdowns.
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b)
TS Table 4.3-1. Functional Unit 2.B. for NIS power range (PR).
neutron flux-low, requires quarterly channel calibration while in
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Mode 2.
On April 8,1997, while conducting a review specifically
looking for TS mode change difficulties, the licensee determined
that Unit 1 entered Mode 2 on March 15 without accomplishing the
surveillance requirements of TS Table 4.3-1.
This incident was
also reported in LER 97-05.
Unit 1 was only in Mode 2 for about
seven minutes until it entered Mode 3.
Due to inadequate test
procedures and TS inconsistencies, the licensee concluded this TS
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surveillance requirement had been missed during other Unit 1 and 2
shutdowns.
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c)
on April 23, 1997, while conducting a special, comprehensive
review of TS mode change requirements, the licensee determined
that the 18 month TS surveillance requirement 4.8.1.1.2.c.8 had
not been performed on Unit 2 for EDG 1-2A or 1C.
OR# 97-162 was
initiated and LER 97-03 (Unit 2 only) was written to document,
resolve and report the incident.
TS 4.8.1.1.2.c.8 requires
verifying a simulated SI signal will override the test mode of an
operating EDG connected to its bus returning it to standby
operation. This surveillance is a TS requirement for both units.
However. FNP-0-STP-80.8. " Diesel Generator 1-2A 1000 KW Load
Rejection Test." and FNP-0-STP-80.9. " Diesel Generator 1C 1000 KW
Load Rejection Test." which perform the required surveillance
test, were only done on Unit 1.
Credit was then given for meeting
the TS surveillance requirement for both units. On April 24. the
1-2A EDG was successfully tested while connected to its vital 4160
Enclosure 2
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VAC bus on Unit 2 (i.e.
Bus 2F). The 1C EDG was successfully
tested on Unit 2 the following day.
Although these were identified by the licensee, these are being cited as
a violation due to their repetitive nature and inadequate licensee
corrective action. These examples of failure to follow TS surveillance
requirements constitute a violation identified as VIO 50-348,
364/97-05-03, Failure To Follow Multiple TS Surveillance Requirements.
On April 25, 1997, the licensee also examined its compliance with TS
surveillance requirement 4.8.1.1.2.e which requires conducting a load
rejection test of 1200 - 2400 KW every five years without tripping the
EDG. and verifying that all fuses and breakers on the energized
emergency bus (es) are not tripped.
Similar to example c) above. the
licensee discovered that the applicable FNP-0-STP-80.11, " Diesel
Generator 1-2A 1200 KW Load Rejection Test." Revision 8. and
FNP-0-STP-80.12. " Diesel Generator 1C 1200 KW Load Rejection Test.'
Revision 6, were not being performed on both units.
Instead, STP-80.11
was routinely conducted only on Unit 1 and STP-80.12 was being conducted
on' Unit 2.
But, after additional evaluation, the licensee concluded
that TS 4.8.1.1.2.e could be satisfied for the shared A train EDGs by
only testing them on one unit and taking credit for both. A resident
inspector reviewed the licensee's evaluation and disagreed with its
conclusion. The inspector read the statement in TS 4.8.1.1.2.e. for
both units, that says " Verify that all fuses and breakers on the
energized emergency bus (es) are not tripped" seems to imply this
surveillance test is not just an EDG test.
Rather. each unit's
emergency bus (es) are also required to be tested during an EDG load
rejection to ensure their respective breakers and fuses do not trip.
The inspector conferred with NRC technical staff regarding this
difference in opinion, and then notified the licensee that their
position appeared inconsistent with TS requirements.
Following
discussions with the inspector. SNC concluded that a differing opinion
continued to exist. To reconcile this difference SNC formerly requested
a TS interpretation from the NRC by letter dated May 22, 1997.
In
addition, SNC also conducted a successful 1220 KW load rejection test of
the 1C EDG tied to its Unit 1 emergency bus: and depending on Unit 2
availability, the licensee was considering testing the 1-2A EDG. While
awaiting NRC respense to the May 22 letter, the licensee also has
submitted a TS amendment request dated May 28, 1997, that clarifies
TS 4.8.1.1.2.e.
In addition to the compliance implications of TS 4.8.1.1.2.e discussed
above the inspector also questioned the licensee on the adequacy of the
surveillance testing being conducted from a technical and licensing
basis.
Both STP-80.11 and 80.12 only require their respective EDG to
undergo a 1200 KW load rejection test, which is the minimum allowed by
TS. The TS load rating of the 1C EDG is 2850 KW and the 1-2A EDG is
4075 KW. According to the NRC's safety evaluation report (SER) that
approved the TS 4.8.1.1.2.e surveillance requirement, "the licensee has
Enclosure 2
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devised a test that, by manually tripping two circuit breakers (leaving
the diesel generator output breaker closed), a load approaching 50% of
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rating is rejected . ." The SER also stated. "the confidence gained
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from... testing of loss of. half the rated load every five years is
sufficient to provide reasonable assurance on a continuing basis that
the diesel generator will not be lost due to a load rejection..." A 50%
load rejection for the 1C and 1-2A EDG would be 1400 KW and about
<
2030 KW, respectively. TS 4.8.1.1.2.e establishes a range that
specifically allows for testing each EDG at their own 50% load rating.
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However, contrary to statements in the SER, the licensee's STPs only
test the 1-2A and 1C EDGs at 29% and 42% of their load rating.
respectively.
STP-80.11 and 80.12 only tri) one circuit breaker to
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initiate the load rejection vice two descri)ed in the SER. The
inspec'or discussed these concerns at length with SNC management. The
licensee will evaluate the adequacy of surveillance testing methodology
in light of the current licensing basis and technical arguments.
These concerns are identified as unresolved item (URI) 50-348,
364/97-05-04. EDG 50% Load Reject Surveillance Testing pending NRC's
response to the licensee's TS interpretation request and proposed TS
amendment.
M1.9 SI/LOSP Intearated Test
On April 25, 1997, a resident inspector reviewed the results of
FNP-1-STP-40.0.
" Safety Injection with Loss of Off-Site Power Test."
Revision 31.
The test appeared to be challenging and complex. Test
completion was considered satisfactory with several relatively minor
exceptions, which were noted and rescheduled.
Test signoffs were
current and the pre-job checklist was com)lete. Test data packages
appeared to be complete and results met t1e defined acceptance criterta.
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On May 5. a resident inspector observed FNP-1-STP-40.0 Appendix S.
" Retest of 01P16MOV3134. SW from RCP MTR CLRS " conducted under WO 97003589. The pre-job brief conducted by the Shift Supervisor was
clear, concise and identified potential areas of concern. The retest
was conducted satisfactorily after a procedural inaccuracy was
reconciled.
Overall. the test and retests appeared to adequately verify
SI operation with a loss of offsite power.
M1.10 Inservice Insoection
a.
Insoection Scooe (IP 737521
To evaluate the licen_ce's Inservice Inspection (ISI) program and the
program's implementation, the inspectors reviewed 3rocedures, observed
work in progress and reviewed selected records.
0)servations were
compared with ap)licable procedures, the Final Safety Analysis Report
(FSAR) and ASME 3&PV Code Sections V and XI,1903 Edition, Summer 1983
Addenda (83S83).
Enclosure 2
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Specific areas examined included the following:
observation of Liquid
Penetrant (PT) examination of Item Nos. ALA2-4517-37 and ALA2-4516-1;
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manual Ultrasonic (UT) examination of Item No. ALA2-4516-1: data
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acquisition and analysis activities associated with Eddy Current (ET)
examinations of Steam Generator (S/G) tubing; data acquisition and
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analysis activities associated automated UT examinations of reactor
ve.;sel welds: review of video tape of the remote Visual (VT) examination
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of the reactor vessel internals: direct VT examination of support -
ALA2-4516-SI-R206: Review of selected completed examination reports: and
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review of the Repair and Replacement Program.
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The inspectors performed an independent evaluation of indications to
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confirm the licensee's ISI examiners' evaluations.
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The inspectors reviewed records for the Nondestructive Examination (NDE)
personnel and ecui) ment utilized to aerform ISI examinations. The
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. records includec : 1DE equi) ment cali) ration and materials certificationi
and records attesting to N)E examiner qualification. certification and
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visual acuity.
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The inspector:; scrutinized FNP Occurrence Report OR No. 1-97-115.
b. ' Observations and Findinos
ISI examinations observed / reviewed were conducted in accordance with
properly approved procedures, by cualified and properly certified
examiners using properly certifiec/ calibrated equipment and materials.
The licensee had implemented the Containment Inspection Rule repair and
replacement (R/R) program by revisions to FNP-1-M-043 "Second Ten-Year
Inservice Inspection Program For Class 1. 2. and 3 Components" and
FNP-0-GMP-0.2. " Repair and Replacement Instructions for ASME Class 1. 2.
3. and MC Components.~ and the issuance of FNP-0-GNP-0.4. " Repair and
Replacement Instructions for ASME Class CC." FNP-0-100 .24. Visual
Examination VT-1 For IWE Components." FNP-0-100 .25, " Visual Examination
VT-3 For IWE Components." FNP-0-100 .26. " Visual Examination VT-1C."
FNP-0-100 .27. " Visual Examination VT-3C."
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FNP-1-M-043 paragraph 1.9 states, in part. " Code Case N-416-1 (RR-47)
may be used in place of hydrotest in some situations. The NRC SER
recuires a surface examination on the root pass layer of Class 3 butt
anc socket welds on pressure retaining boundaries." The NRC SER had not
been identified and the root inspection option had not been addressed in
lower tier documents. The licensee indicated that they would take
appropriate action.
The issue documented in OR No. 1-97-115, was properly identified,
evaluated and closed out.
Enclosure 2
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c.
Conclusion
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ISI activities observed / reviewed were conducted in accordance with
procedures, licensee commitments and regulatory requirements.
M1.11 Weldina
a.
Insoection Scoce (IP 62700)
To evaluate the licensee's welding program and the program's
implementation, the inspectors reviewed procedures, observed work in
progress, and reviewed selected records. Observations were compared
with applicable procedures, the FSAR and ASME B&PV Code Sections V and
XI. 83S83 and Section IX latest at the time of qualification.
Specific areas examined included the following: observation of welding
activities in the containment associated with the replacement of the
containment coolers; observation of welding activities in the auxiliary
building associated with the bypass of the Boron Injection Tank, and
observation of welding activities in the auxiliary building associated
with the replacement of service water valves; inspection of the welding
material issue station: and inspection of the welder qualification test
facility, The inspectors scrutinized Work Order (WO) Nos. 11276 and
11284 for the replacement of valve Nos. V044C and V0010C.
The inspectors reviewed records for welders. Quality control (OC)
inspectors, and materials utilized in the W0s. These records included:
Welding Procedure Specifications (WPS) and their supporting Procedure
Qualification Records (POR): Welder Performance Qualification (WP0)
records: records attesting to the maintenance of welder qualification:
receiving inspection reports and Certified Material Test Reports (CMTR)
for welding filler materials; and records attesting to QC inspectors
qualification, certification, and visual acuity.
b.
Observations and Findinas
The inspectors noted mechanical force in the form of a chain fall was
used to hold Service Water piping in place during fit-up activities
prior to welding.
This indicated that the pipe was flexed or cold
sprung into place, thereby inducing stresses in the piping.
Investigation by the inspectors revealed the following:
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Programmatically the only guidance for cold spring is contained in
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FNP-0-SPP-GW-002, Revision 18. " General Welding Standard For
Pressure Boundary Applications," paragraph 8.6 b., which states.
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"There are no indications of excessive cold spring at the time of
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joint fit-up." The licensee had no guidance to define
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" indications of excessive cold spring".
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The pipe deflections were not documented at the time of fit-up.
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Enclosure 2
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Prior to the application of the mechanical force (cold spring), no
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formal analysis had been conducted to determine whether the stress
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levels induced in the piping exceeded Code allowable levels.
The licensee informed the inspectors that they had approved the
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use of cold spring at'a number of locations in the SWS, because
each instance did not appear visually to be " excessive." It was
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recommended by 3rofessional pi)e fitters that aipe spring was
censidered by t1e licensee to 3e within the "scill of the craft"
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for pipe fitters.
The introduction of unknown stress levels into safety-related piping
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systems as shown above indicated a lack of control of the special
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process of welding, and is identified as an exam)le of VID
50-348/97-05-05: Failure to Control the Special 3rocess of Welding. The
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licensee documented this issue in OR 1-97-130.
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ASME B&PV Code.Section IX,
requires the thickness of side bend
specimens, used for the evaluation of welder qualification test
assemblies. to be 3/8-inch thick.
Procedure FNP-0-SPP-WP-030. Revision
15. " Specification for Welder Qualification for Pressure Boundary
Applications " requires the thickness of side bend specimens to be
3/8-inch thick with no tolerance specified.
An inspection of bend
specimens used to evaluate welder performance test assemblies revealed
several specimens that were 1/32 to 1/16-inch under 3/8-inch in
thickness.
Because the side bend thickness is
stress applied to the outer fibers of the bend, proportional to the
undersized specimens
constitute a less rigorous test than intended by ASME B&PV Code
Section IX.
The licensee's failure to conduct bend testing on welder
test assemblies in accordance the ASME B&PV Code Section IX, indicated a
lack of control of the special process of welding, and is identified as
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an example of VIO 50-348/97-05-05. The licensee documented this issue
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in OR 1-97-150.
FNP-0-SPP-WF-001 Revision 12. " Procedure for Welding Filler Material
Control " Paragraph No. 8.4. states, " Work areas shall be kept clear of
unauthorized, unidentified or discarded welding filler materials.
FNP-0-SPP-WF-001, Revision 12. " Procedure for Welding Filler Material
Control," Paragraph No. 8,1. states in part,
"...it is the
responsibility of the welder to maintain control of filler materials
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until used, discarded or returned to the storeroom." Further, the
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licensee stated that it is their expectation that at the end of shift,
work areas are free of all welding filler materials.
Contrary to the
above on April 11, 1997. between the night shift and the day shift. the
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inspectors found a significant quantity of partially used bare welding
filler material abandoned in the area of the Boric Acid Injection Tank.
Although some of the rods had flag tags that identified them as to ty)e,
there was no traceability to heat or batch identification. Many of tie
rods were of a length suitable for continued use.
The licensee's
failure to control welding filler materials indicated a lack of control
Enclosure 2
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of the special process of welding, and is identified as an example of
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VIO 50-348/97-05-05. The licensee documented this issue in OR 1-97-149.
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The licensee's introduction of unknown stress levels into safety-related
pioing systems resulting from their failure to provide guidance to
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define " indications of excessive cold spring:" failure to conduct bend
testing on welder test assemblies in accordance the ASME B&PV Code.
Section IX resulting from their failure to provide tolerances on the
bend specimen thickness; and failure to control welding filler materials
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resulting from their failure to adequately communicate their procedural
requirements and. expectations for welding filler material control to
their contract welders, demonstrates less than effective control of the
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special process of welding and is a violation of Title 10 CFR 50
Appendix B. Criterion IX. which requires that measures be established to
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assure that special processes including welding be controlled. This
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violation is identified as VIO 50-348/97-05-05.
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The licensee's Welding Manual procedures incorporates code requirements
without providing guidance for implementation, some examples are as
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follows:
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FNP-0-SPP-GU-003, paragraph 13.la states. " Undercut shall not
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exceed 0.01 inch deep when its direction is transverse to a
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primary tensile stress in the part that is undercut, nor more than
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1/32-inch for all other situations." No guidance is provided to
determine the direction of primary tensile stress.
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FNP-0-SPP-WP-030 requires the thickness of side bend specimens to
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be 3/8-inch thick with no tolerance specified.
As discussed
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above, the failure to specify a tolerance caused some specimens to
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be made under size.
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e
FNP-0-SPP-GW-002 states, "There are no indications of excessive
cold spring at the time of joint fit-up." As discussed above, the
,
failure to provide quantitative guidance for cold spring resulted
in the introduction of unknown stress levels into safety-related
4
.
piping systems.
j
l[
The Authorized Nuclear Inservice Inspector (ANII) determined that the
i
contract welders were uninformed concerning a number of requirements of
'
the licensee's Welding Manual and welding compliance issues.
In
i
addition structural contract welders were not provided with the means to
i
measure base metal temperature prior to the resumption of welding, to
assure compliance with interpass temperature requirements of the WPS.
l
This issue was documer*.ed in OR 1-97-092.
i
i
A licensee OC inspecte identified an instance where a contract welder
was welding prior to the completion of the welder's Performance
i
Qualification Test Record.
This issue was documented in OR 1-97-083.
i
Enclosure 2
!
1-.
-
_ _
-
-
. . - -
-
,
_
-
,-
. - .
,
-
,
F
19
A licensee QC ins]ector identified an instance where a contract welder
had made an unautlorized base material repair without benefit of a
repair procedure.
The contract welder burned through the material on
which he was welding, and subsequently made a unilateral decision to
effect a repair. The licensee viewed this decision and subsequent
unauthorized repair as a coverup. This issue was documented in
OR 1-97-071.
The inspectors identified a Welder Qualification Test (WOT) Record for
welder 420-04-1378 on test WOT No. ST-6, Revisicn 2. completed on March
i
6.1996, that was missing the certifying signature. .The licensee
indicated that they would conduct a records review to determine whether
'
welder 420-04-1378 3erformed any welding on or after March 6,1996 that
was only supported )y WOT No. ST-6, Revision 2 test.
If so, they will
take a)propriate actions. The licensee subsequently determined that the
'
test tlat the welder in question had taken was a licensee specific test
for "T." "K." and "Y" connections. They further determined that
although the welder had welded on 13 W0s since taking the test, those
i
W0s contained no "T." "K," or "Y" connections.
!
The inspectors identified no discrepancies associated with welding being
performed by the licensee's permanent employees.
Except as noted above the welding activities examined, were conducted by
properly qualified and certified welders, using correct and certified
welding filler materials in accordance with qualified Welding Procedure
Specifications. Procedure Qualification Records were reviewed and
determined to be adequate. Quality Control inspectors associated with
the repair and replacement activities were properly qualified and
certified.
c.
Conclusions
'
A violation was identified associated with the control of the special
process of welding. A weakness was identified associated with the
licensee's control of contracted welders.
M1.12 Audits
a.
Insoection Scooe (IP 62700)
To evaluate the licensee's Audit Program as it relates to maintenance,
the inspectors requested all the audits and self-assessments conducted
in the maintenance area during the previous 12 months. The inspectors
reviewed the two audits provided (96-STPm/34-1 Surveillance Testing -
Maintenance and 96-MAINT/15-1 Maintenance Department, Routine
Scheduled).
Enclosure 2
n_-
.
--. .
-.
. .
_ - .
. - -_- - - -.-._.- - - .-
_
.
.
'
.
.
t
20
b.
Observations and Findinas
i
Audit 96-STPm/34-1 contained no findings. Audit. 96-MAINT/15-1 - findings
included weaknesses related to: meteorological tower instruments not in
agreement with the FSAR: PMT guidelines not consistently followed: smoke
detector procedures performed without proper release; personal hold tag
discrepancies: hand-operated hoists used with out-of-date color codes:
-
and requirements of Purchase Orders not met. Appropriate corrective
actions were taken or planned.
The audit 96-MAINT/15-1 finding related to post maintenance testing
guidelines not consistently being followed, was addressed in Corrective
Action Report (CAR) No. 2201. Revision 1.
CAR 2001. Revision 1. did not
include an adequate job of determining the extent of the problem, as
only the specific examples identified by the auditors were corrected.
c.
Conclusions
i
The area of maintenance was subjected to independent audits, with
'
appropriate action generally taken for identified weaknesses.
l
M8
Miscellaneous Maintenance Issues (IP 92902)
M8.1 Paintina
a.
Insoection Scooe (62707)
The inspectors observed Jainting activities, and reviewed procedures and
paint data sheets from FiP-0-CP-MD-801, " Coating Program." Revision 7.
Also, the inspectors interviewed various licensee personnel including
'
supervisors, engineers, painters, and the painting foreman.
b.
Observations and Findinas
l
The licensee commenced a major 3ainting effort throughout the )lant this
!
spring. This effort included tie Auxiliary Building (AB) Tur)ine
Building, and Diesel Generator Building.
Overall the Sainters appear to
,
I
be doing a good job.
However, several problems with tie EDGs and an
issue with the Unit 2 PRF systems were identified as a result of the
painting.
e
Licensee personnel identified that the fusible links for the 1-2A
and 28 DG roll up doors had been painted on one side. The
I
licensee promptly instituted the appropriate compensatory actions
Jer the Fire Protection Program until the links were replaced.
ollowup testing of the painted links identified that the paint
raised the actuation point approximately 10 degrees F.
The slight
increase in actuation temperature was insignificant.
!
i
!
Enclosure 2
. - .
.-
j
,
.
l
21
e
On April 24. 1997, during a routine tour of the AB. the inspectors
noticed very strong paint fumes from recent painting efforts of
the RHR HX room (inside the 3enetration room boundary (PRB)).
The
paint was a modified epoxy-plenolic product with 56% solids and
thinned with up to 1 pint of thinner per gallon of paint.
One
'
gallon of paint would cover approximately 225 ft' if applied at
the recomended thickness of four mils.
The inspectors were
concerned about the potential operability effects of the volatile
organic compounds (VOC) on the PRF system charcoal filters.
The inspector discussed the control of painting inside the PRB
'
with licensee personnel.
The licensee staff stated that Sainting
inside the PRB was limited to less than 1000 ft' in any 2L hour
period which is about four gallons if applied per the paint data
sheet.
Neither the inspector nor the licensee staff were able to
find this limit proceduralized to control the painting. Although
the licensee did not have any direct procedural controls, the
inspector determined that this criteria was being followed. based
on interviews with o]erations personnel, the painting foreman, and
'
various painters.
T1is criteria was identified obliquely in the
" Precautions and Limitations" sections of FNP-1/2-STP-20.0.
" Penetration Room Filtration System Train A(B) Quarterly
Operability and Valve Inservice Test." Revisions 25 and 15.
FNP-1/2-STP-20.2. " Penetration Room Filtration System Train A(B)
Monthly Operability Test," Revisions 7 and 7. and
FNP-1/2-SOP-60.0. " Penetration Room Filtration System."
Revisions 10 and 12.
The procedures all stated "Do not run
Penetration Room Filtration within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following significant
painting (>1000 ft') in the penetration rooms."
The licensee was not able to provide an existing analysis
2
documenting the 1000 ft limit.
On May 9. the inspectors
'
discussed the continuing painting efforts of rooms within the PRB
without an analysis with senior licensee management.
Licensee
management stated they would stop painting in rooms served by a
safety-related charcoal unit until the issue was resolved.
This
is identified as URI 50-364/97-05-06. Painting Effects On PRF
Operability pending review of the licensee's analysis.
e
On April 25. 1997, the ins)ectors toured the EDG building to
specifically inspect the EE fuel racks. valves, and other
components for paint related problems. The inspectors found no
overspray on the fuel racks.
However. some paint was found on the
valve stems for the air header drains on one EDG.
This condition
was identified to the EDG S0 and immediately corrected.
Enclosure 2
. _ _
_ _ - _ . _ _ _ . - - . . - - -
-
. - - - . - - .
- -
-
.
'
.
.
i
22
!
c.
Conclusions
Painting was generally well-controlled, although, some minor
deficiencies which did not affect operability were identified with
overpainting of components in the EDG building.
However, the painting
of rooms in the PRB, while controlled, did not appear to have an
analysis to document continued operability of the PRF charcoal filters.
M8.2 (Closed) VIO 50-348. 364/96-006-01: Failure to Perform Surveillance Test
of SI Handswitch Inout
(Closed) LER 50-348. 364/96-004: Surveillance Reauirements Not Met for
Manual Safety In.iection Inout into the Reactor Trio System
(Closed) LER 50-348. 364/96-004-01: Surveillance Reauirements Not Met
l
for Manual Safety In.iection Inout Into the Reactor Trio System
l
,
The licensee's review of Generic Letter 96-01 identified that the
18-month surveillance test for the manual SI input into the reactor trip
system had not been performed since initial preoperational startup
t
testing. The licensee implemented FNP-1/2-STP-38.3. " Verification of
'
Manual SI Actuation Input to Reactor Trip," to 3rovide guidance and a
means to document completion of the test.
The Jnit 2 SI Handswitch was
satisfactorily tested prior to Mode 2 entry after the Fall 96 outage.
The -inspectors observed the satisfactory performance of the Unit 1 SI
.
Handswitch test on April 23, 1997. The licensee also performed a
broadness review of events from January 1993 to August 1996 which
i
resulted in an LER or were identified as a near miss.
This review was
documented in Corrective Action Report 2208 and concluded that the
'
primary root cause appeared to be a failure of personnel to adequately
self-check. The inspectors verified the licensee's corrective actions
were appropriate and complete.
M8.3 (Closed) IFT 50-364/96-013-02: Increased Freauency Test Proaram for
Charaina Pumos Due to Claddina Crackina
,
This item was opened pending formal recommendations and actions for
performing increased frequency testing of the charging pumps due to
!
cladding cracks. Southern Company )rovided the formal recommendations
1
for increased frequency testing to INP by-letter dated April 9. 1997.
,
The recommendations were:
o
2A CCP: UT every six months for three cycles then UT every
36 months. VT after three cycles then every five years (actually
'
will be done every 54 months).
J
l
e
All other CCPs:
UT every 36 months and VT upon disassembly.
Based on a review of the available data and verification of the testing
incorporated into the PM data base, the inspectors concluded these
Enclosure 2
- -
- _ _
.-
--
.
.
'
.
.
1
23
intervals would be adequate to ensure that pump degradation would be
identified prior to any operability concerns.
,
1
M8.4 (Closed) LER 50-348. 364/96-002: Technical Soecifications Surveillance
Reauirements Not Met and Common Cause Failure Identified
This issue was discussed in detail in IR 50-348, 364/96-04 and closed
'
out as NCV 96-04-03. Failure To Adequately Test RCP Underfrequency
Reactor Trip Relays. No new issues were revealed by the LER.
III. Enaineerina
El
Conduct of Engineering (IP 37551)
i
E1.1 Overstressed Unit 1 Reactor Coolant looD oue To Inadeauate Gaos 8etween
'
SG Lateral Suocorts
a. Insoection Scooe (37700)
The inspectors reviewed the pipe stress and support calculations.
modification packages and inspection records and discussed the SG
support gap on SG loop C with the engineers from Westinghouse Electric
Company. Southern Nuclear Operating Company (SNC), and Southern Company
Services (SCS) to determine if licensee activities complied with
industrial standards. regulatory requirements, licensee commitments, and
American Society of Mechanical Engineers (ASME) and American Institute
of Steel Construction (AISC) codes.
b. Observations and Findinas
i
The licensee identified that excessive vibration had occurred on the RCP
in Loop C of Unit 1 during normal operation. The licensee 3erformed an
inspection at cold shutdown condition during the current scleduled
refueling outage and found that several of the lower lateral supports on
reactor coolant loop (RCL) C piping had excessive gaps and that support
LS-12 had a gap 0.7-inch smaller than designed.
The designed gaps are
established to allow for SG thermal growth to close toward, but not
touch the supports during normal operation temperature of 610 degrees F
( F).
The licensee determined that the excessive gaps would cause no
i
problem.
The smaller gap would cause binding to occur in Support LS-12
l
when the operating temperature exceeded 445 *F. which would induce
thermal expansion stress in Loop C piping, components, and supports.
l
The thermal expansion stress could represent a situation that caused
damage to the piping or supports.
!
i
I
Enclosure 2
..
.-
_.
.-
-
_ . _ . . _ _ . . _
._
..
.
.
.
24
j
The licensee took the following actions to correct the prot,lem:
-
Performed stress analyses and evaluation for the piping,
components,.and supports based on the ASME Code Section III and
AISC Code.
-
Inspected piping.. components, and sup] orts for cracks,
deformations, and distress based on tie ASME Code Section XI.
-
Modified the existing shims by providing adjustable shims to
adjust the gaps to the desired dimensions.
The ins]ectors reviewed the following calculations performed by
Westinglouse and SCS for the licensee:
j
1)
Westinghouse Calculation W-SMT-97-067. Farley 1 RCL Support
Interference Thermal Pipestress Run, Rev. 0
2)
Westinghouse Calculation W-SMT-97-066. 3D Finite Element Analysis
of the Hot Leg 50 Degrees Elbow Using Jamming Loads, Rev. 0
3)
Westinghouse Calculation W-SMT-97-061. Farley Unit 1
Uprating/ Jamming - RCL Fatigue Evaluation Rev. O
4)
Westinghouse Calculation CSE-04-97-0036, Analysis of Steam
T.nerator Lower Support Strut Jamming Condition. Rev. 1
5)
SCS Calculation SC-97-1-9162-001. Evaluation of LS-12 Embed Steam
Generator IC, Rev. O
Calculation W-SMT-97-067.used a 3D computer model and pipe stress
computer program PIPESTRESS for the system analysis.
The model included
three loops, a reactor vessel, and three steam generators. The model
also considered gaps, the local shell flexibility of the reactor vessel
nozzle shell junction, and the horizontal stiffness of the reactor
vessel supports.
The results indicated that the critical stress was 67
kips per square inch (ksi) at Node 3216 of the elbow area, which
exceeded the ASME Code allowable stress of 51 ksi.
Calculation W-SMT-97-066 used a 3D finite element program WECAN and the
loads from the calculation W-SMT-97-067 to get more accurate results
based on a finite element model.
The critical stress was 45.3 ksi at
Node 15792 at the safe end of the reactor nozzle.
The critical stresses
at the weld between the elbow and the SG and at the elbow were 43.2 ksi
i
and 40.6 ksi respectively. Both stresses were within the allowable
stress of 51 ksi and were acceptable.
Calculation W-SMT-97-061, performed for the fatigue analysis, was based
on the past plant operating data such as start-up cycles, transients,
and predicted future operations.
The critical accumulated usage factor
Enclosure 2
i
_
. _ _ _ _
_ . _ .
._ _____
__ ._
_ . _ . _
'
.
.
.
25
was 0.974 at the outlet nozzle of the reactor vessel. The calculated
usage factor was less than 1.0 and therefore the current piping
condition is acceptable.
Calculation CSE-04-97-0036 was performed to assess the condition of the
steel portion of support LS-12 utilizing a calculated direct force of
2494 kips and a shear force of 113 kips.
The stress ratio for the
postulated seismic event upset and faulted conditions were 1.22 and 0.98
respectively.
The 1.22 ratio exceeded the allowable value of 1.0.
Thus, the steel could have been overstressed during a seismic event.
The licensee concluded that the support was acceptable as is, though,
since no deformations or cracks were found during the licensee's
walkdown inspection.
Calculation SC-97-1-9162-001 evaluated the embedded steel and concrete
and determined that the concrete was acceptable. The stress in the
embedded steel was 12 Jercent over the allowable stress for the faulted
,
condition.
However, t1e licensee concluded this condition was also
l
acceptable because no cracks or deformations were found during the
walkdown inspection.
The inspectors-concluded that the calculations were adequate.
Based on
discussion with the inspectors, the licensee planned to incorporate into
the calculations and plant procedures several changes as listed below:
-
Add the cross reference for the qualification of the critical
stress 45.3 ksi at the nozzle of the SG to Calculation
W-SMT-97-066.
1
-
Clarify the critical node numbers selected for the fatigue
analysis in Calculation W-SMT-97-061, which were incorrect and did
i
not match the stresses in the analysis.
The selected stresses
were correct.
i
1
-
Develop a procedure for monitoring the support gaps during the
subsequent heatup and cooldown.
,
Design Change Package DCP 97-1-9162-0-001 was reviewed for adequacy of
its 10 CFR 50.59 evaluation and the shim modification.
The inspectors
!
concluded that the design change was acceptable.
4'
c.
Conclusions
The inspectors concluded that the calculations were acceptable based on
i
their review of portions of the calculations.
The licensee promptly
'
evaluated the potential of past damage to the piping or supports and
q
restored the intended conditions through the modifications.
i
.
'
.
Enclosure 2
<
i
!
1
.
.
. ._ .
__.
. _ ._ . _____._ __._.___.__
_
'
.
.
.
26
E1.2 Licensino Basis of PRF System
As documented in IR 50-348, 364/97-04, dated April 2.1997, the
'
inspectors concluded that the PRF system was required to operate during
post-LOCA recirculation for'any size LOCA and proposed various apparent
violations.
During the subsequent pre-decisional enforcement conference
on April '18. '1997. SNC management disagreed with the NRC inspectors and
stated the PRF system was only required to operate under large break
LOCA conditions (i.e.. Condition IV events).
Following the enforcement
conference the NRC determined that several violations of NRC
requirements had occurred and issued a Notice of Violation (NOV) by
letter dated May 6, 1997.
In this letter the NRC acknowledged that SNC
did not agree with the inspectors' conclusions regarding the licensing
,
basis for the PRF system and stated that the matter was under review by
the Office of Nuclear Reactor Regulation (NRR).
The issue is identified
'
as URI 50-348, 364/97-05-07. Licensing Basis for PRF System During Post-
LOCA Recirculation, pending additional review of this issue by the NRC.
i
E8
Miscellaneous Engineering Issues (92903)
,
'
E8.1
(Closed) URI 50-348. 364/96-013-04: Common Tao for SG Steam Flow
Transmitter and Narrow Ranae Water Level
(Closed) LER 50-348/96-007: IEEE-279 Reauirements Not Met for Protection
Channel III
By letter dated April 29. 1997, the licensee submitted a request for a
" proposed alternative" to Section 4.7.3. Control and Protection System
Interaction-Single Random Failure, of IEEE 279-1971 pursuant to 10 CFR 50.55a(a)(3).
SNC implemented interim administrative controls which
provide an acceptable alternative until necessary protection / control
system hardware changes can be im)lemented during the next Unit 1 and 2
refueling outages in 1998 (i.e.
J1RF15 and U2RF12).
The NRC has
reviewed SNC's administrative controls and considers them acceptable as
j
an interim measure until such time as it can review the licensee's
request. Resident inspectors have verified implementation of the-
interim administrative controls. interviewed responsible operators on
their knowledge of these controls, and reviewed applicable procedural
requirements in FNP-0-SOP-0 " General Instructions To Operations
Personnel." Revision 47.
This URI and LER are closed.
Enclosure 2
- ..
. . _ , . __ . _ _ _ _ _ _ _ _ . _ _ _ . _ _ _ _ . . _ . .
. _ _... _ . _
. _ ._
A
1
,
,
27
)
IV. Plant Support
R1
Radiological Protection and Chemistry (RP&C) Controls
R1.1 Radioloaical Controls
a.
Insoection Scoce (IP 83750)
Radiological controls associated with ongoing Unit 1 (U1) Refueling
!
j
Outage Number 14 (U1RF14) activities and with Unit 2 (U2) routine
operations were reviewed and evaluated by the inspectors.
Reviewed
i
program areas included area postings, radioactive waste (radwaste)
'
container bibels, controls for high and locked-high radiation areas,
i
i
arocedurf and radiation work permit (RWP) guidance and general
I
l
lousekee)ing and cleanliness.
Established controls were compared
1
against rinal Safety Analysis Report (FSAR) details and documented
!
procedural requirements to meet applicable sections of Technical
Specifications (TSs) and 10 CFR Part 20.
'
l
l
The insaectors made frequent tours of the radiologically controlled
l
areas (RCAs).
Radiation work permit guidance and selected survey
results were reviewed and discussed with responsible Health Physics (HP)
l
,
staff and supervisors. The inspectors directly observed worker and HP
technician performance and discussed results of radiation and
contamination surveys conducted for selected equipment and facility
locations. Specific radiological controls and adequacy of surveys
associated with movement of materials out of the U1 Containment
equipment hatch. a U2 power entry and for a piece of loose metal
retrieved from the U1 reactor cavity were reviewed and discussed in
detail.
Further, the inspector reviewed and discussed occurrence
reports for three personnel contamination events (PCEs) associated with
outage activities.
For the PCEs reviewed,
the inspectors evaluated and
discussed licensee assumptions, dose methods and skin dose results in
detail.
1
The irispectors discussed and reviewed "As low as Reasonably Achievable"
(ALARA) program implementation. individual worker doses, and dose
expenditures associated with the following U1RF14 outage job evolutions,
e
RWP 197-1435. Incore Drive Work
l
!
e
RWP 197-1438, Seal Table / Thimble Cleaning
e
RWP 197-1439. Gamma Metrics Detector Work
<
l
e
RWP 197 1465, Lower Internals Movement
.
e
RWP 197-1480, Scaffolding - Containment
i
e
RWP 197-1730. Nozzle Dam Removal
l
Enclosure 2
,
l
4
i
. _ , _ _ . _ _- --
-
. _ . - -
._
. .
.
.
I
28
a
b.
'"orvations and Findinas
dnd locked-high radiation area controls were verified to be
splemented in accordance with TS requirements.
Postings for
radiologically controlled areas were proper and in accordance with TS or
10 CFR 20 Subpart J requirements.
Containers holding radwaste,
contaminated materials and equipment were labeled in accordance with
i
10 CFR 20.1904 requirements.
In general, workers followed proper radiological controls.
Radiological
controls and surveys associated with the U2 power entry to inspect
j
i
containment coolers and for a piece of metal having contact dose rates
i
of 1630 rem per hour, which was retrieved from bottom of +he U1 reactor
i
cavity, were conducted in accordance with approved procedo,es.
However,
i
several instances of individuals reaching across radiation control
'
boundaries during movement of equipment into and out of the U1 equipment
hatch were identified on April 7,1997.
'
For individuals involved in licensed activities, year-to-date (YTD) dose
estimates were within regulatory limits.
The maximum total effective
dose equivalent (TEDE) value reported was approximately 1510 millirem
(mrem).
Licensee skin dose evaluations for the PCEs reviewed were
thorough and technically adequate. Assumptions and details regarding
physical location, length of exposure and isotopic characteristics of
'
particles or contamination were appropriate. All skin doses were within
regulatory limits with a maximum exposure of 10.7 rem to the skin of the
whole body for a worker installing service water valve parts in the U1
pipe penetration room on the 121 foot elevation.
From discussion with responsible staff and from review of planning
4
documents and dose expenditures, the inspectors verified implementation
of ALARA program activities in accordance with FNP-0-Radiation Control
Procedure (RCP)-19. Pre an( Post Jcb Planning for Work in Radiation
Controlled Areas of the Plant. Revision 10. dated January 9, 1997.
In
<
particular, the inspectors reviewed and discussed planning for
replacement of the gamma metrics detector and the removal and
reinstallation of the reactor vessel lower internals.
Significant
-
reduction in dose expenditure was identified for tasks associated with
the lower internals.
The current dose expenditure of approximately
807 millirem (mrem) was reduced significantly relative to previous dose
expenditures of 6095 mrem and 2179 mrem for the same tasks conducted
during the U1RF8 and U2RF8 outages, respectively.
c.
Conclusions
i
Radiological controls for routine U2 operations and U1RF14 outage
activities were good.
Several minor isolated instances of poor
J
radiation control practices were identified.
Personal doses were
Enclosure 2
.
-
-.
. . -
- - .
=
. - - _ - ._-
--
.
- -.
.
.
.
29
maintained within regulatory limits.
ALARA program activities were
implemented effectively.
R1.2 Internal Exoosure
a.
Insoection Scoce (IP 83750)
The inspectors discussed program guidance for monitoring and evaluating
possible internal ex)osures, and reviewed in detail licensee results for
investigative whole Jody counts conducted during the current Unit 1
outage.
'
In addition, guidance for testing and test results to ensure quality of
supplied breathing air for respiratory protective equipment were
reviewed and discussed.
'
b.
Observations and Findinas
As of April 11. 1997, six investigative whole-body counts associated
with events which could indicate potential internal exposure during
,
U1RF14 outage activities were conducted.
The maximum u)take was
,
approximately 1.6 derived air concentration-hours (DAC-irs) resulting in
a committed effective dose equivalent (CEDE) of 3.11 mrem.
Because the
'
doses did not exceed 10 mrem. i.e.
0.2 percent of the annual limit of
intake (ALI), the resultant internal exposures were not added to the
individuals' official exposure records,
i
l
The inspectors verified that the compressor systems used to supply
breathing air were tested to certify Grade D air for potential use
during outage activities.
Breathing system air samples were collected
quarterly in accordance with FNP-0-RCP-110. " Radiation Control and
Protection Procedure. Sampling of Service Air to Meet Res)iratory
Limits." Revision. 4. and FNP-1-RCP-1112. " Operation of t1e Containment
,
Preathing Air System." Revision.12.
Sample results collected in March
'
1997 verified that the supplied breathing air quality exceeded the
established limits for Grade D air specified in the Compressed Gas
Commodity Specification G7.1.1973.
c.
Conclusions
Controls for minimizing internal exposure were effective.
Potential
,
!
uptake of radionuclides were evaluated appropriately.
Licensee tests
'
verified that the service air compressor system supplied Grade D
respirable air in accordance with 10 CFR 20. Appendix A requirements.
.
Enclosure 2
.,
,
-
.
30
l
R2
Status of Radiological Protection Facilities and Equipment
R2.1 Tours of the Unit 1 and 2 Radioloaically Controlled Areas (RCAs)
(IP 71750)
1
During the course of the inspection aeriod, the inspectors conducted
tours of the Unit 1 and 2 auxiliary auilding RCAs.
In general, health
physics (HP) control over the RCA, and the work activities conducted
within it, were good.
R5
Staff Training and Qualifications in Radiation Protection and Chemistry
R5.1 Resoirator Trainina and Fit Testing
l
a.
Insoection Scoce (IP 83750)
The inspectors reviewed and evaluated General Employee Training (GET)
provided to meet the requirements of 10 CFR Part 19, and the specific
.
training and medical certification requirements specified by
j
10 CFR Part 20. The frequency of training and fit testing was compared
to the guidance listed in American National Standards Institute
(ANSI) Z88.2. Practices for Respiratory Protection, May 19, 1992.
Current training, fit testing and medical certification for selected
contractor and licensee personnel who used or were designated to use
i
respiratory protection equipment were reviewed and discussed with
!
licensee representatives.
n
i
In addition, the frequency of training and fit testing was compared to
j
the guidance listed in ANSI Z88.2. Practices for Respiratory Protection,
1
l
May 19, 1992.
!
b.
Observations and Findinas
'
.
The inspectors verified that GET. respiratory protection training, and
respiratory medical certifications were conducted in accordance with the
2
requirements of 10 CFR 19.12, 10 CFR 20.1703 and licensee 3rocedure
FNP-0-RCP-101, "Use and Testing of Respiratory Protection Equipment "
-
Revision 24.
From review of training records of selected individuals
'
!
listed on " Respiratory Protection Record," HP Form-257, issued during
l
March and April 1997, the inspectors verified that persons who used
respiratory protection equi) ment were trained and medically certified in
j
.
accordance with the applica)le procedures.
!
The frequency of verifying medical certification met recuirements
specified in 10 CFR 1703.
However, the inspectors notec that
FNP-0-RCP-101 only s)ecified training and fit testing to be conducted
,
every five years ratler than annually as recommended by ANSI Z88.2.
'
'
Licensee representatives informed the inspectors that revisions to the
procedure would require annual training and fit testing at a five year
i
Enclosure 2
,
i
.-
_
_
_
.
31
interval.
Following these discussions, licensee management stated that
the fit testing frequency would be reviewed.
c.
Conclusions
Training and medical certifications for personnel using respiratory
protective equipment were conducted in accordance with the licensee
procedures and met the applicable requirements of 10 CFR Part 19 and
R8
Miscellaneous RP&C Issues (IP 83750, IP 84750)
R8.1
(Closed) VIO 50-348. 364/96-10-02: Failure to Follow a March 14. 1983
Order to Imolement and Maintain Commitments for Soecial Calibration of
the Containment Hiah Radiation Monitors (CHRMs)
Licensee STPs for the loop calibration of the CHRMs did not include
in-situ calibrations using electronic signal substitution for all range
decades above 10 Roentgens per hour (R/hr).
From review of
FNP-2-STP-227.18 and FNP-2-STP-227.19. the inspectors verified that
guidance was revised to include an electronic calibration check for each
decade of the required response range. The inspectors also verified
from review of data for completed U2 CHRM STPs conducted in November
1996 that the required electronic calibration was completed
satisfactorily. Completion of the in situ electronic calibration was
scheduled for the U1 CHRMs during the current outage.
This VIO is
closed.
R8.2 (Closed) VIO 50-348. 364/96-10-03: Failure to Label Casks of
Contaminated Resins in Accordance with 10 CFR 20.1904(a) Reauirements
Licensee documents and training emphasized labeling recuirements based
on dose rates rather than radionuclide quantities and cid not require
specific information detailed in 10 CFR 20.1904.
The ins)ectors
verified that FNP-0-RCP-57. " Radioactive and Potentially Radioactive
Material Handling " Revision 23. required appropriate labeling
information to be provided based on specific quantity of radioactive
material rather than measured dose rates.
From review of licensee
records the inspectors verified that management expectations regarding
container labeling requirements were verbally communicated in August
1996 to site health physics (HP) personnel and formal training regarding
the procedural revision was conducted during December 1996.
In December
1996. HP personnel conducted walk-downs of the RCA to verify compliance
with labeling requirements.
In addition, licensee representatives
stated that additional training regarding labeling requirements for
specific situations, e.g., liquids and alpha-emitting radionuclides or
liquids would be reviewed and discussed with HP staff and personnel.
l
The inspectors toured the RCA and verified implementation of the current
procedural requirements.
This violation is closed.
Enclosure 2
l
. _ . _ . _ .
._ .
_
... _ _ _ _ _ _ _ _ .. _ . _ _ _ . _ - _ _ _ _ . _
.
.
.
.
)
32
.R8.3 .(Closed) VIO 50-348. 364/96-10-04: Failure to Follow Procedures for
,
Prooer Personal Dosimetry Use
l
Observations by both NRC inspectors and licensee HP staff identified
'
individuals within established RCAs not adhering to HP manual
i
requirements nor training guidance for use and placement of personal
dosimetry including thermoluminescent dosimeters (TLDs) and digital
alarming dosimeters (DADS).
From review of documents and direct
observations, the inspectors verified implementation of licensee
corrective actions.
On August 27, 1996, a memorandum was issued from
the HP staff to all su)ervisors detailing requirements for use of
personal dosimetry. T1e inspectors also reviewed and discussed results
of periodic HP staff dosimetry observations conducted between October 29
and November 25. 1996.
By November 25. the documented error rate for
use of personal dosimetry was less than one percent.
On November 27,
1996 memorandum documented a request to HP personnel to increase
vigilance of dosimetry use by facility personnel.
During tours of-the
RCA, the inspectors did not identify programmatic problems associated
with use of personal dosimetry. This violation is closed.
R8.4 (Closed) VIO 50-348. 364/96-10-05: Failure to Have Adeauate Procedures
for Liauid Effluent Comoosite Samole Storaqe
,
Licensee procedures for storing composite liquid effluent samples
collected for quantification of non-gamma emitting radionuclides did not
require use of standard methods such as acidification to prevent
i
plate-out of radionuclides on the storage container. The inspectors
verified that applicable procedures involving the storage of composite
4
samples were revised to require proper acidification of the liquid
i
samples. The inspectors reviewed and discussed results of a Chemistry
'
Incidert Report (CIR) completed to determine the effect of
non-preservation of affected liquid sam)1es.
Results of the study were
documented in the Section 6.3. Program Jeviations of the FNP Annual
Radioactive Effluent Release Report, dated April 21, 1997.
Results of
the study indicated that for the worst case assumptions regarding
plate-out, historical doses were understated from 3 to 7 percent but
were within limits specified in the Offsite Dose Calculation Manual
,
!
(ODCM).
Based on licensee actions and documentation, this violation is
closed.
R8.5 (Closed) VIO 50-348. 364/96-13-05: Failure to Follow Radiation Work
l
Permit for Use of Prooer Protective Clothina
!
The ins)ectors observed personnel performing selected tasks within the
i
RCA witlout use of the protective clothing specified by the respective
RWPs. The inspectors reviewed and discussed the results of surveys and
immediate corrective actions as documented in Occurrence Reports (ors)
96-1002 and 96-1003.
In addition. the inspectors verified that
j
personnel were retrained in RWP adherence and that additional emphasis
!
to RWP adherence would be stressed in general employee training.
In
l
Enclosure 2
-
. -. .
-.
.
.
..
..
. _ _ _ . _ _ _ _ _ _ _ _ .
__ . _ _ . .
___
.
.
7
.
.
i
33
addition, an HP memorandum, dated October 25, 1996, to site personnel
detailed the poor radiological practices observed and importance of
. proper radiological practices.
The inspectors also reviewed and.
i
discussed initial implementation of a licensee initiative to track,
trend and take more effective corrective actions regarding poor
radiation worker practices.
During tours of the RCA, the ins)ectors
verified that selected tasks were conducted in accordance witi
established RWPs. This violation is closed.
S1
Conduct of Security and Safeguards Activities
i
S1.1 Routine Observations of Plant Security Measures (IP 71750)
i
During routine inspection activities, inspectors verified that
of site security program plans were being properly implemented. portions
This
was generally evidenced by: proper display of picture badges by plant
i
personnel: appropriate key carding of vital area doors: adequate
stationing / tours in the protected area by security personnel; proper
>
searching of packages / personnel at the primary access point and service
water intake structure: and adequate maintenance of security systems.
Security personnel activities observed during the inspection period were
l
performed well.
Site security systems were adequate to ensure physical
'
protection of the plant.
However, on April 29, 1997, the inspector
)
observed numerous individuals in Unit 1 containment who were not
}
displaying their security badges.
Certain of these individuals were
challenged by the inspector regarding their badges, all of whom had
their security badges inside their outer anticontamination clothing.
When questioned by the inspector, they expressed a belief that security
badges were not required to be displayed in containment.
FNP-0-AP-42,
" Access Control." Revision 25, requires " security badges will be
)rominently displayed in plain view at all times." But Section 8.3 of
NP-0-AP-42 does allow wearing the badge beneath outer anticontamination
clothing when working in hiah contamination areas. Although the
containment was posted as a contaminated area, the inspector realized
that the vast majority of Unit 1 containment floor space was clean or
only slightly contaminated.
Discussions were held with the Security
Chief and HP Superintendent regarding plant worker misconceptions and
the intent of FNP-0-AP-42. The Security Chief and HP Su3erintendent
3romptly corrected the situation.
Subsequent tours by tie inspectors
lave not identified any repeats problems.
F1
Control of Fire Protection Activities
F1.1 Weldina and Grindina - Unit 1
During the week of April 7,1997, an inspector ooserved numerous
examples of poor fire prouction practices in the Unit 1121 foot
i
elevation PPl.
Work activities during this week were extremely
i
intensive in the PPR, as was the quantity of tools, equipment and
material brought in to support the work. Welding and grinding was
!
Enclosure 2
<
,.-4
_
- , . - ,
i%
+ , . - .
.
.
.
34
pervasive throughout the PPR as part of many of the ongoing jobs
(e.g. , SWS valve replacements high head safety injection discharge line
replacement and reroute). All welding and grinding observed by the
inspector was accomplished by contractors.
During tours of the PPR and
discussions with craft personnel, the inspector ascertained the
following: 1) WO's did not identify applicable Open Flame Permits:
2) Open Flame Permits were not routinely posted for each job, inclusion
of many work activities on one Open Flame Permit made for confusion:
3) control of combustibles within a 35-foot area of the hot work was
very poor (in one instance the inspector observed a burlap bag ignite
and catch fire). 4) fire watches were tasked with ancillary duties,
which in a few cases recessitated the fire watch to leave the immediate
area 5) inadequate fire extinguishers (e.g., beyond inspection date,
partially discharged), 6) fire extinguishers were not in the immediate
area, and difficult to iocate, and 7) apparent discharging of fire
extinguishers without notifying Operations or supervision.
Furthermore,
during several containment tours in the month of April, the inspector
observed additional poor fire protection practices especially for
welding and grinding conducted in elevated areas.
These poor practices
involved floors not swept clean within 35 feet; and inadequate or
non-existent covers beneath work to collect sparks.
The quantity and
widespread nature of the aforementioned problems suggests a weakness in
the implementation of the licensee's fire protection program for
controlling open flame work.
Upon notification of the findings made
during the week of April 7. licensee management took prompt actions.
The inspectors will review the actions taken by the licensee to correct
this weakness.
F2
Status of Fire Protection Facilities and Equipment
F2.1 Half-Hour Kaowool Fire Barrier
Inspection Report 50-348, 364/96-09, issued November 8, 1996, documented
i
the failure to install Appendix R required fire barriers on various
electrical raceways on Unit 1 including BDE-15 (Train B charging pump
l
power cables) and BHF-24 in room 160. The failure to install an
A)pendix R required fire barrier on BDE-15 was cited as an example of
EEI 96-410/01013.
On November 14, 1996, while touring the AB. the
inspectors observed the installation of Kaowool on raceways BDE-15 and
BHF-24 in room 160. The wraps consisted of a 1-inch layer of Kaowool on
the sides and bottoms of the trays. A one-hour rated barrier requires
two 1-inch layers on all four sides.
l
!
The inspectors identified this discrepancy to the licensee on
'
November 14.
The licensee provided a co)y of Fire Protection Program
Reevaluation. Amendment 5 (precedes the :SAR documented Fire Protection
'
Program), which documented the installation and basis of " half-hour
barriers" on BDE-15 and BHF-24.
The inspectors reviewed FSAR
Appendix 98, Fire Protection Program, and found no mention of half-hour
'
Enclosure 2
.
-
=
---
_
.
.
--
.
.
.
35
barriers in it or Attachment B.10 CFR 50 Appendix R Exemptions.
Appendix 98. Attachment B. Section 21.3 on page 9B.B-91 states that:
"The redundant charging Jump power cables are provided with a
barrier (two 1-inch thicc wra)s of Kaowool blanket) having a fire
rating greater than that of t1e projected fire in the following
rooms in fire area 1-004:
train A in rooms 161. 162. 163. and
168; train B in rooms 175. 160 and 159."
This issue was discussed further with the licensee on November 15. 26.
December 9. and March 20, 1997.
The licensee's )lanned corrective
action was to modify the FSAR Appendix 9B. Attac1 ment B to " clarify"
that these raceways were wrapped with half-hour barriers (1-inch Kaowool
fire wrap). The inspectors informed the licensee that this
" clarification" would misrepresent FSAR Appendix 9B. Attachment B, to
'
read as though the " clarification" was part of the original NRC-approved
exemptions.
The licensee's position was that there is a licensing basis for the
half-hour barriers based upon SERs and NRC inspections pre-dating
10 CFR 50. Appendix R requirements. The inspectors disagreed because
the use of half-hour barriers was not identified as a s)ecific exemption
from Appendix R requirements in FSAR Appendix 98. Attac1 ment B.
As
interim corrective action the licensee has taken action to ensure that
a one-hour roving fire watch is maintained on room 160 until this issue
is resolved.
This is identified as URI 50-348/97-05-08. Installation Of
Half-hour Kaowool Fire Barriers Without Appendix R Exemption, pending
NRR review.
,
F2.2 (Ocen) IFI 50-348. 364/96-006-07:
Fire Main Failures
l
This item was opened pending metallurgical analysis of the failed piping
and implementation of long term corrective actions.
Southern Company
Services (SCS) provided the results of the metallurgical analysis and
recommendations for action via letter dated December 5, 1996. A ten-
inch cast iron pipe failed due to a pressure excursion at a degraded
section of pipe.
A four-inch pipe failed as the result of localized
exterior corrosion (due to wet insulation).
SCS stated that there was
little in the way effective ins)ection that could be performed to
determine the integrity of the Juried cast iron-lined pipe.
However.
regarding the four-inch pipe failure, they did recommend visual
,
examination of readily accessible pi)e in areas where water would tend
l
to accumulate. As of May 7, 1997, tais recommendation had not been
'
implemented. This IFI will remain open pending resolution of necessary
corrective actions.
1
Enclosure 2
.._. ,---.
_-. - -
.
- . ..
- -
- - ._ - - - . - . - .-
.-
,
l
.
.
.
36
F2.3 Inadeauate Life Safety Exits from Turbine Buildino in Case of Evacuation
On May 1.1997, and then again on May 2. with an industrial safety
specialist. an inspector walked down the TB exits.
The inspector
identified numerous deficiencies with the exit signs and evacuation
design scheme for the turbine building from a Life Safety perspective.
Examples of these deficiencies were: 1) Missing Exit signs. 2) Burned
out and/or broken Exit signs. 3) Nonvisible Exit signs. 4) Exit sign for
,
a nonexit door. 5) Inadecuate fire barriers for allowing personnel to
)
safely exit the south enc of the TB. and 6) No directions or signs in
the 155-foot stairwells to guide personnel on how and where to evacuate
the TB. These problems were pointed out to the safety specialist and
discussed with plant management, who were evaluating necessary
corrective actions.
V. Manaaement Meetinas and Other Areas
X1
Review of Updated Final Safety Analysis Report (UFSAR) Commitments
A recent discovery of a licensee o)erating its facility in a manner
contrary to the UFSAR description lighlighted the need for a special
focused review that compares plant practices, procedures and/or
parameters to the UFSAR descriptions.
While performing the inspections
discussed in this report. the inspectors reviewed the applicable
portions of the UFSAR that related to the areas inspected.
The
inspectors verified that the UFSAR wording was consistent with the
observed plant practices, procedures and/or parameters, except for the
following discrepancies:
(a)
Table 7.3-1. FUNCTIONS INITIATED BY ENGINEERING SAFETY FEATURES
ACTUATION SYSTEM. This table implies that PRF will actuate
simultaneously with equipment that is started on a SI signal.
Phase A and phase B actuation equipment is identified separately
and later in the table and PRF is not identified as actuating
equipment.
(b)
Table 7.3-9. FAILURE MODE AND EFFECTS ANALYSIS. PENETRATION ROOM
FILTRATION SYSTEM.
This table lists the analysis concerning the
effect on the system if a failure occurs and a PRF component is
not automatically aligned during a Phase A CTMT isolation signal.
(This represents an identification problem.)
i
(c)
The Staff's SER. NUREG-75/034. SAFETY EVALUATION REPORT JOSEPH M.
l
FARLEY NUCLEAR PLANT UNITS 1 AND 2. dated May 2. 1975, states that
the PRF will actuate on a Safety Injection signal and this
statement has not been incorporated into the FSAR or into plant
design.
2
Enclosure 2
- .- _ . - . _ ._ _ _ _ _ _ . _ _ ____ __
~ _ _ . _ . _ _
.
.
.
.
37
(d)
FSAR Chapter 1.3, COMPARISON TABLES, Table 1.3.1. DESIGN
COMPARISON identifies that FNP system functions are similar to
those of the North Anna and Surry Nuclear Power Plants.
However,
with regard to the PRF system _(Section 7.3 of the FSAR) the FNP is
not similar.
(e)
FSAR page 7.3-11 section 7.3.2.1.1. Single Failure Criteria,
sentence 6 states that CTMT s) ray is activated on high-high
containment pressure signal, lowever it actuates on high-high-high
j
containment pressure.
(f)
FSAR page 6.2-84. section 6.2.3.3.2 references paragraph
,
j
15.4.1.3.4. which does not exist.
l
4
Responsible licensee management were informed on each of the above
discrepancies, many of which the licensee had already identified as part
a
of its FSAR Reverification Program.
Several of the discrepancies are
!
involved with the NRC/SNC reviews regarding PRF system design and
j-
licensing basis (Section E1.2).
X2
Exit Meeting Summary
J
The inspectors presented the inspection results to members of licensee
i
management on May 15. 1997, after the end of the inspection period.
The
i
licensee acknowledged the findings presented.
The inspectors asked the licensee whether any materials examined during
the inspection should be considered proprietary.
No proprietary
i
information was identified.
f
PARTIAL LIST OF PERSONS CONTACTED
)
l
Licensee
!
M. Ajluni, SNC (Corporate) Licensing Manager - Farley Project
-
j
W. Bayne, Chemistry Superintendent
C. Byrd, Southern Company Services (SCS) Support Design Engineering Manager
!
S. Casey. Engineering Su3 port Supervisor - Steam Generators
R. Coleman Maintenance Manager
!
J. Fridrichsen. SNC (Corporate) Senior Project Engineer
S. Fulmer. Technical Manager
J. Garlington. Nuclear Support General Manager
1
!
D. Graves. Health Physics Supervisor
l
D. Grissette. Operations Manager
j
P. Harlos Plant Health Physicist
i
T. Harrison. Williams Power Corporation (WPC) Site Manager
J. Hayes, Fire Marshall
'
2
R. Hill. General Manager - FNP IP 73753:
Inservice Inspection
C. Hillman, Security Chief
T. Liu. Westinghouse Electric Company (WEC) Farley Special Project Manager
,
i
l
Enclosure 2
4
I
4
,
-
, - - . - -
- - - -
- - . , -
-
-
-
.-
---
.
.
,
'
.
.
!
i
38
G. Lofthus. NDE Level III Inspector
i
R. Martin. Su)erintendent Operations Support
'
A. Maze. NDE project Supervisor
J. McGowan SNC (Cor) orate) SAER Manager
l
M. Mitchell. Health 3hysics Superintendent
'
B. Moore. SNC (Corporate) Nuclear Support Manager
C. Nesbit. Assistant General Manager - Support
j
J. Odom. Superintendent Unit 1 Operations
J. Powell. Su]erintendent Unit 2 Operations
C. Sterzil. WEC Farley Special Project Support Design Engineer
{
L. Stinson.-Assistant General Manager - Plant Operations
j
J. Thomas. Engineering Support Manager
i
J
P. Webb. Technical Training Supervisor
'
d
R. Winkler. Plant Modifications and Design (PMD) Supervisor
i
R. Yance. PMD Manager
i
!EC
J. Zimmerman Project Manager - Farley Nuclear Plant
-
i
INSPECTION PROCEDURES USED
i
IP 37551:
Onsite Engineering
'
i
IP 37700:
Design Changes and Modifications
j
IP 40500:
Effectiveness of Licensee Controls in Identifying. Resolving and
!
Preventing Problems
j
IP 60710:
Refueling Activities
)
IP 61726:
Surveillance Observations
1
IP 62700:
Maintenance Implementation
j
IP 62707:
Maintenance Observations
i
IP 71707:
Plant Operations
!
IP 71750:
Plant Support Activities
!
IP 73753:
Inservice Ins)ection
i
j
IP 83750:
Occupational
Radiation Exposure
?
IP 84750:
Radioactive Waste Treatment, and Effluent and Environmental
Monitoring
IP 92901:
Followup - Operations
IP 92902:
Followup - Maintenance
IP 92903:
Followup - Engineering
Enclosure 2
4
, _ _ .
..
- _ . . -
.
- . . _
_
_ _ __ _ _ __
_ _ . _ _ . . _ .
.
.
.
.
l
39
ITEMS OPENED CLOSED, AND DISCUSSED
Opened
[
IYgg Item Number
Status
Description and Reference
VIO- 50-348, 364/97-05-01
Open
Failure to Notify NRC of Change of
Licensed Operator Medical Status
(Section 05.1).
,
l
l
IFI
50-348, 364/97-05-02
Open
Foreign Material in Containment ECCS
l
Sumps (Section M1.7).
l
>
!
50-348, 364/97-05-03
Open
Failure to Follow Multiple TS
!
Surveillance Requirements
(Section M1.8).
50-348. 364/97-05-04
Open
EDG 50% Load Reject Surveillance
Testing (Section M1.8).
50-348/97-05-05
Open
Failure to Control the Special
,
Process of Welding (Section M1.11).
l
50-364/97-05-06
Open
Painting Effects on PRF Operability
!
(Section M8.1).
50-348, 364/97-05-07
Open
Licensing Basis for PRF System
During Post-LOCA Recirculation
(Section E1.2).
50-348/97-05-08
Open
Installation of Half-hour Kaowool
Fire Barriers Without Appendix R
,
Exemption (Section F2.1).
l
'
50-348, 364/97-05-09
Open
Failure to Fully Implement
Corrective Actions (Section 07.1)
i
Closed
.
Iygg Item Number
Status
Description and Reference
LER
50-348/97-002
Closed
Safety-Related 4160 Volt AC Breakers
Not Seismically Qualified
- -
!
(Section 08.1).
LER
50-348/97-004
Closed
Safety-Related 600 Volt AC Breakers
Position Sensitive Seismic
Qualification (Section 08.1).
.
,
i
LER
50-364/96-001-00
Closed
Reduction / Resumption of 28 Diesel
l
Generator Speed (Section 08.2).
i.
Enclosure 2
<
. _ . _ _ _ _ _ . _ . _
_ _ _ . . _ _
.. . _ _ . _ . _ . _ . _
'
.
.
.
40
-VIO
50-348, 364/96-006-01
Closed
Failure to Per'orm Surveillance Test
of SI HandsC tch Input
(Section M8.2).
LER
50-348, 364/96-004
Closed
Surveillance Requirements Not Met
for Manual Safety Injection Input
into the Reactor Trip System
(Section M8.2).
LER
50-348, 364/96-004-01
Closed
Surveillance Requirements Not Met
For Manual Safety Injection Input
into the Reactor Trip System
(Section M8.2).
IFI
50-364/96-013-02
Closed
Increased Frequency Test Program For
Charging Pumps Due to Cladding
Cracking (Section M8.3).
LER
50-348. 364/96-002
Closed
Technical Specifications
,, 4
Surveillance Requirements Not Met
and Common Cause Failure Identified
(Section M8.4).
50-348, 364/96-013-04
Closed
Common Tap for SG Steam Flow
Transmitter and Narrow Range Water
Level (Section E8.1).
LER
50-348/96-007
Closed
IEEE-279 Requirements Not Met for
Protection Channel III
(Section E8.1).
50-348, 364/96-10-02
Closed
Failure to Follow a March 14, 1983
Order to Implement and Maintain
Commitments for Special Calibration
of the Containment High Radiation
Monitors (CHRMs) (Section R8.1).
50-348, 364/96-10-03
Closed
Failure to Label Casks of
Contaminated Resins in Accordance
with 10 CFR 20.1904(a) Requirements
(Section R8.2).
50-348, 364/96-10-04
Closed
Failure.to Follow Procedures for
Proper Personal Dosimetry Use
(Section R8.3).
50-348, 364/96-10-05
Closed
Failure to Have Adequate Procedures
for Liquid Effluent Composite Sample
Storage (Section R8.4).
Enclosure 2
,
._ .
. . . . _ _ .
. ~ . _ - . . . . . - . - . _ . . _ . _ - . _ . _ -
_.__.____ . .
_ _ _ . - -
_ . . . _ .
.
.
,
'
-
!
.
.
i
41
50-348, 364/96-13-05
Closed
Failure to Follow Radiation Work
Permit for Use of Proper Protective
Clothing (Section R8.5).
50-348, 364/97-05-09
Closed
Failure to Fully Implement
Corrective Actions (Section 07.1)
~
Discussed
i
Iygg Item Number
Status
Description and Reference
. IFI
50-348, 364/96-006-07
Open
Fire Main Failures (Section F2.2).
4
.
5
2
-
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