ML20245D531
| ML20245D531 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 05/22/1989 |
| From: | Blake J, Crowley B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20245D457 | List: |
| References | |
| 50-321-89-02, 50-321-89-2, 50-366-89-02, 50-366-89-2, NUDOCS 8906270147 | |
| Download: ML20245D531 (81) | |
See also: IR 05000321/1989002
Text
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,, .. UNITED STATES -
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[ $
NUCLEAR REGULATORY COMMISSION
\ g 'Q' REGION ti
f7 o 101 MARIETTA STREET, N.W.
' In f ATLANTA, GEORGIA 30323 -
' '+, . . . . . de -
Report Nos.: 50-321/89-02 and 50-366/89-02
Licensee: Georgia Power Company '
-P. O. Box 1295
Birmingham, AL 35201
Docket Nos.: 50-321 and 50-366- License'Nos.: OPR-57 and NPF-5
Facility Name: Hatch 1 and 2
Inspection Conducted: February 27 -' March 17. 1989
Inspector: [. 5 22
B. R. Crowley (Team Lead y) Date Signed
Team Members
G. A. Hallstrem
M. D. Hunt l
P.. J. Fillion
F. N. Wright
S. S. Kir s
G. s
Approved by: .
i/ o
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22 7
J. J. ake, Chief Date Signed
M ter als and Processes Section
ng eering Branch
Division of Reactor Safety
SUMMARY
Scope: This special announced inspection consisted' of an in-depth team
inspection of the Hatch maintenance program and its implementation.
NRC-Temporary Instruction 2515/97, dated November 3,1988, was used
for guidance.
Results: Overall, the maintenance program was judged to be " Good" with " Good" ,
implementation. Areas of strength and weakness are highlighted in }'
the Executive - Summary with details provided in the report. Four
violations were identified: inadequte administrative procedure - l
paragraph 3.a.; failure to complete adequate corrective action - ;
paragraphs 3.b. and 3.c.; failure to take breathing air samples -
paragraph 3.d.; and failure to follow acceptance criteria for weld
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patch on reactor building roof drain paragraph 3.e. l
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8906270147 890615
PDR. ADOCK 05000321
0 PDC
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REPORT DETAILS
1. Persons Contacted
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Licensee Employees l
G. Brinson, Superintendent of QC
Y. Brown, Systems Engineer
H. Buchans, I&C Supe, visor
- G. Barker, Superintendent of I&C
J. Cameron, Senior Maintenance Plant Engineer
B. Coleman, Supervisor, Document Control
A. Cowan, I&C Supervisor
G. Creighton, Senior Regulatory Specialist
- S. Curtis, Supervisor -Shift Technical Advisor
J. Dawson, Maintenance Supervisor
D. Davis, Manager of General Support
- W. Drinkard, Manager, Safety Analysis and Engineering Review
W. Duvall, HP Chemistry Supervisor
L. E11 gass, NPRDS Coordinator
- P. Fornel, Manager of Maintenance
- 0. Fraser, QA Site Manager
G. Gill, Senior Maintenance Plant Engineer
- W. Glisson, Maintenance Engineering Supervisor
- R. Godby, Maintenance Superintendent
- M. Googe, Manager of Outages and Planning
F. Gorley, Operations Supervisor
R. Grover, Plant Engineer - Nuclear Safety and Compliance
- L. Gucwa, Manager, Nuclear Engineering and Licensing
J. Hadden, Supervisor, Plant QC
- J. Hammonds, Nuclear Safety and Compliance Supervisor
R. Hukill, Supervisor, Maintenance Support Group
B. Keck, Reactor Systems Engineering Superintendent
R. King, Discipline Engineering Supervisor
T. King, Maintenance Supervisor
W. Kirkley, Acting Manager of HP/ Chemistry
J. Lanier, Senior Systems Engineer - Reactor Control
- J. Lewis, Acting Operations Manager
M. Link, Supervisor, HP Operations
A. Manning, QA Auditor
D. Matthews, Systems Engineer - Nuclear Boiler
W. Metts, Maintenance Supervisor
E. Metzler, Nuclear Safety and Compliance Supervisor
D. Midlik, Senior Maintenance Plant Engineer
L. Mikulecky, Senior Plant Engineer - Regulatory
- C. Moore, Plant Support Manager
- H. Nix, General Plant Manager
G. O'Donnell, I&C Supervisor
R. Ott, Supervisor, Training
R. Pooni, Acting Supervisor, Reactor Protection Engineering
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h 4 ENCLOSURE 2
EXECUTIVE SUMMARY
Background '
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The Nuclear Regulatory Commission considers effective maintenance of equipment
and components a major aspect of ensuring safe nuclear plant operation and has
made this area one of the NRC's highest priorities. In- this regard, the
Commission. issued'a. Policy Statement dated March 23, 1988, that states, "it is
the objective of the Commission that all components, systems, and structures of
- 1 ear power plants be maintained so that plant equipment will perform its
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intended function when required. To accomplish this objective, . each licensee-
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should develop and implement a maintenance program which provides for the
l periodic evaluation, and prompt repair of plant components, systems, and
structures t4 ensure their availability." {
To- ensun.- effective implementation of the Commission's maintenance policy, the !
NRC staff is undertaking a major program to inspect and evaluate the
effectiveness of licensee maintenance activities. As part of this inspection 1
activity, the current inspection was performed in; accordance with guidance
provided in NRC Temporary Instruction'2515/97, Maintenance Inspection Guidance, ,
dated November 3, 1988. The temporary instruction includes a " Maintenance !
Inspection Tree" that identifies the major elements associated with effective 1
maintenance. . The tree was designed to ensure that all factors related to {
maintenance are evaluated. l
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Conduct of Inspection j
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The maintenance inspection at the Hatch Nuclear Station was initiated with a !
site meeting on January 24-26, 1989, where the inspection scope, including the i
maintenance inspection tree, was discussed. At that meeting, the licensee i
presented to the inspection team leader an overview of the site maintenance ;
program. In addition, a comprehensive package of material, as requested by NRC l
1etter dated January 10, 1989, was provided for inspection preparation. l
The inspection was conducted by a team consisting of a team leader and six .l
inspectors. Four of the inspectors were from RII and two were from NRR. The ,
team spent two weeks, February 27-March 3 and March 13-17, 1989, on site !
conducting the inspection. .l
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The inspection was performance based, directed toward evaluation of- equipment l
conditions; observation of in process maintenance activities; review of
equipment histories and records; and evaluation of performance indicators, :
maintenance control procedures, and the overall maintenance program. Based on i
known industry problems, plant specific problems, and discussions with the i
Hatch Resident Inspectors, the team selected five systems and directed the l
inspection toward determining whether these systems were being properly j
maintained and assessing if the current maintenance system would ensure proper j
maintenance in the future. The systems selected were: Ell (RHR), B21 (Main
Steam), N21 (Condensate and Feedwater), PS2 (Instrument Air), and E41 (HPCI).
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The team performed walkdown inspections'of portions of the selected systems to
, determine the material condition of the equipment. In addition, maintenance
history records for the last two years were obtained and reviewed for any
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adverse; trends. NPRDS data were also reviewed for the selected systems. In
review-of equipment history records, any questionable trends were examined in
detail to determine if equipment was being properly maintained. In the course
of the inspection, the team also observed general housekeeping and equipment
-condition for a large part of the plant.
Results
After completion of the inspection, the maintenance program was evaluated using
the.NRC TI and inspection tree as a tool. See paragraph 2 of the Inspection
Report for details of.the rating scheme.
The inspection results are presented pictorially in Figure 1 as the completed
inspection tree. As noted in Figure 1, overall, the Hatch program for
establishing and implementing an effective maintenance program was rated " Good"
both in program and implementation. For the three major areas: (1) Overall.
Plant Performance was rated " GOOD", (2) Management Support was rated
" SATISFACTORY" for program and implementation, (3) Maintenance implementation
was rated " Good" for program and implementation. These ratings were based on
specific strengths and weaknesses identified in the report details. The
following are the more significant strengths and weaknesses identified:
Strengths -
Overall, the training program for maintenance was very
strong. The facilities and the use of actual components as
training aids were outstanding.
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In general, plant housekeeping was good.
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The maintenance data base and equipment records (NPMIS)
were very good. The data base appeared to be user friendly
and records were readily retrievable.
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The overall maintenance staff was a strength. Staffing
levels appeared to be adequate. Team work was evident.
Management was well qualified and enthusiastic.
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The QC staff was well organized, qualified, and heavily
involved in the maintenance process.
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The licensee has a strong program for controlling the
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maintenance backlog. The backlog is low.
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The licensee makes good use of-performance indicators.
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Overall, plant equipment condition was good.
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A strong program (deficiency card system) for identifi-
cation of deficiencies and intiation of action was in place
and appeared to be working well.
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? Interfaces between maintenance organization and other-
organizations were well established and appeared' to be
working well. Daily planning meetings were well organized
, -and' appeared to be a strong point in the maintenance
process.
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Both clean and:" hot" machine shops were indicative of good
j' -maintenance facilities.
Weaknesses -
Weaknesses were identified in the PM program.for electrical
y' equipment in that vendor recommended PMs were not included'
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ini procedures' and no documented justification ' existed
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for excluding the recommendations - examples: 4160 : volt
i switchgear,.busbars and cable compartments not included in
procedures and' no requirement to check protection charac-
teristics. for molded-case circuit breakers.
-- Weaknesses . were identified in the root cause analysis
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program as follows: the procedure needs strengthening to
provide more. detail .on how to perform root cause analysis,
a' motor failure on a HPCI valve did not receive a root
cause analysis, excessive time was taken to determine cause
of Feedwater Pump-Seal leakage.
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Responsibilities for Systems Engineers were not well
' defined.
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Some procedural weaknesses were identified - examples: the
maintenance program procedure and predictive maintenance
vibration analysis procedure needs to include cross
reference to' ASME Section XI requirements for ASME
Section _XI components; the maintenance program procedure
needs to include additional detail relative to ensuring
proper functional / operability test when changes are made to
the MWO; and the procedure controlling the procedure update
program was inadequate to insure that vendor recommended
maintenance is included in maintenance procedures.
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Weaknesses in the program for personnel safety were
identified - examples: failure to have unique fittings
for connecting breathing air to instrument air and
procedure for maintenance of electrical equipment could
be strengthened by adding some safety precautions.
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The level and clarity of detail on some MW0s was poor
resulting.-in difficulty in determining details of work
performed - examples: MW0s 2-88-4862, 2-88-1906 and
2-88-3177.
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Weaknesses were identified relative to corrective action -
examples: failure te properly torque upper mounting bolts
on hydraulic control units (HCUs) and failure to ensure
that all fittings for connecting breathing air were unique
fittings.
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UNITED STATES
'* , [ Mar - o . NUCLEAR REGULATORY COMMISSION
g-
[ g
Igg
REGION ll .
.101 MARIETTA STREET, N.W.
- ATLANTA, GEORGI A 30323 '
'
$
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Report-Nos'.: 50-321/89-02 and 50-366/89-02
. Licensee: Georgia. Power Company
'P. 0. Box 1295
.
Birmingham, AL 35201
Docket.Nos..: 50-321 and 50-366 License Nos.: DPR-57 and^NPF-5~
Facility Name: LHatch 1 and E
Inspection Conducted: February 27 - March 17, 1989
Inspector: .[. 5 22 'f '
B. R. Crowley (Team Leadg") Date Signed
. Team Members
G. A. Ha11strom
M. D. Hunt
P. J. .Fillion
F. N. Wright
S. S. Kir s-
G. s
L , -Approved by: , 'u E2 S7
J. J. ake, Chief Date Signed
M ter als and Processes Section
ng eering Branch
- .
Division of Reactor Safety -
SUMMARY
Scope: This special announced inspection consisted of an in-depth team
inspection of the Hatch maintenance program and its implementation.
NRC Temporary Instruction 2515/97, dated November 3, 1988, was used
for guidance.
Results: Overall, the maintenance program was judged to be " Good" with " Good"
implementation. Areas of strength and weakness are highlighted in-
the Executive Summary with details provided in the report. Four
violations were identified: inadequte administrative procedure -
paragraph 3.a.; failure to complete adequate corrective action -
paragraphs 3.b. and 3.c,; failure to take breathing air. samples -
paragraph 3.d.; and failure to follow acceptance criteria for weld
patch on reactor building roof drain paragraph 3.e.
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REPORT DETAILS
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.1. Persons Contacted
Licensee Employees.
G. Brinson, Superintendent of QC
Y. Brown, Systems Engineer
H. Buchans, I&C Supervisor
- G. Barker, Superintendent of I&C
J- Cameron.. Senior Maintenance Plant Engineer
.
B. Coleman, Supervisor, Document Control
A. Cowan, I&C-Supervisor
G. Creighton, Senior Regulatory Specialist
- S. Curtis, Supervisor..-Shift Technical Advisor
J.-Dawson,' Maintenance Supervisor
D. Davis, Manager.of General Support
- W. Drinkard, Manager, Safety Analysis and Engineering Review
W.;Duvall, HP Chemistry Supervisor
L. E11 gass', NPRDS Coordinator.
- P. Fornel, Manager of Maintenance
- 0. Fraser, QA Site Manager
G. Gill, Senior Maintenance. Plant Engineer
- W. Glisson, Maintenance Engineering Supervisor
- R. Godby, Maintenance Superintendent
- M. Googe, Manager of Outages and Planning
F. Gorley, Operations Supervisor
R. Grover, Plant Engineer - Nuclear Safety and Compliance
.
- L. Gucwa, Manager, Nuclear Engineering and Licensing
'J. Hadden, Supervisor, Plant QC
- J. Hammonds, Nuclear Safety and Compliance Supervisor
R. Hukill, Supervisor, Maintenance Support Group
B. Keck, Reactor Systems Engineering. Superintendent .
R. King, Discipline Engineering Supervisor I
T. King, Maintenance Supervisor
W. Kirkley, . Acting Manager of HP/ Chemistry
J. Lanier, Senior Systems Engineer - Reactor Control.
- J. Lewis, Acting Operations Manager
M. Link, Supervisor, HP Operations
A._ Manning, QA Auditor
D. Matthews,- Systems Engineer - Nuclear Boiler
W. Metts, Maintenance Supervisor
E. Metzler, Nuclear Safety and Compliance Supervisor
l D. Midlik, Senior Maintenance Plant Engineer ,
l L. Mikulecky,- Senior Plant Engineer - Regulatory )
- C. Moore, Plant' Support Manager
- H. Nix, General Plant Manager
G. O'Donnell, I&C Supervisor j
R. Ott, Supervisor, Training '
R. Pooni, Acting Supervisor, Reactor Protection Engineering j
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-T. Powers, Engineering Support Manager
W. Porter,' Senior Maintenance Plant Engineer - Vibration
J_. Reddick, Supervisor, HP Support
P. Roberts, Plant' Project Superintendent
- W. Rogers, Chemistry Superintendent
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H. Scarbrough, Maintenance Supervisor
V. Shaw Senior Plant Systems Engineer
J. Sherman, Reactor Control Systems Engineering Supervisor
D.' Smith, HP' Superintendent
R. Staines, Training Coordinator
M. Sutton, Training Supervisor
L. Sumner., Plant Manager
- S.'Tipps, Nuclear Safety and Compliance Manager
J. Vaughn, Maintenance Supervisor
A. Vora, Senior Maintenance Plant Engineer
A. Wheeler, BOP Systems Engineering Supervisor
J. Wilkes, Superintendent of Planning and Control
D. Williams, Plant Systems Engineer'- ECCS
C. W111 yard, Senior Systems Engineer -ECCS
C. Wright, Shift Supervisor
R. Zorn, QC Supervisor
Other ; licensee' employees contacted during this -inspection included
craftsmen, engineers, operators, mechanics, security force members,
technicians, and administrative personnel.
NRC Personnel
- A. Herdt, Branch Chief, DRP:PB3, RII
J. Menning, Senior Resident Inspector
- E. Merschoff, Deputy Director, DRS, RII
- R. Musser, Resident. Inspector
- Attended exit interview
Acronyms and initialisms used throughout this report are listed in the
last' paragraph.
2. Inspection Methodology
The inspection was performance based, directed toward evaluation of plant
equipment condition and evaluation of maintenance for systems which have
had problems. The systems selected for evaluation were based on the
following:
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Known industry problems l
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Review of Hatch LERs - Site Specific Problems !
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Review of NRC Bulletins and Notices I
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Review of Hatch Deficiency Reports 1
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Discussions with Resident Inspectors {
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PRA information provided by NRR i
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Inspector's Experience t
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Based on the above criteria, the following systems were selected for the
inspection effort:
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Ell, Residual Heat Removal (RHR)
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B21, Nuclear Boiler (Main Steam)
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N21, Condensate and Feedwater
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PS2, Instrument Air (including P51)
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E41, High Pressure Coolant Injection (HFCI)
Maintenance for the selected systems was inspected by: observation of
equipment condition (walkdown inspections), observation of in process
maintenance activities, review of equipment history records (MWO and
NPRDS), and evaluation of performance indicators and trending data.
Based on the inspections performed, the maintenance program was evaluated
using the inspection tree from NRC TI 2515/97 (see Figure 1). As
indicated in Figure 1, three major areas of the licensee's maintenance
program were evaluated: (1) Overall Plant Performance Related to
Maintenance, (2) Management Support of Maintenance, and (3) Maintenance
Implementation. Under each major area, a number of elements were ;
evaluated, rated, and colored using the following guidelines:
" Good" Performance (Green) -
Overall, better than adequate; shows
more than minimal effort; can have a
few minor areas that need improvement
" Satisfactory" or " Adequate" -
Adequate, weaknesses may exist, could
Performance (Yellow) be strengthened
" Poor" Performance (Red) -
Inadequate or missing
(Blue) -
Not evaluated
In general, the top half of the box (element) was rated depending on
whether the element was in place and the bottom half was rated depending
on how well the element was being implemented.
3. Significant Issues Identified
a. Maintenance on the Indoor Metal-Clad Switchgear for the 4160 Volt
Distribur. ion System.
During tFe inspection detailed in paragraph 4 below, the team
identified issues regarding the recommended preventive maintenance,
the maintenance interval and the quality of the preventive
maintenance procedure for the 4160 volt metal-clad switchgear.
Procedure 52PM-R22-001-05, specifies preventive maintenance work for
the 4160 volt switchgear. The procedure covers verification of the
undervoltage trip attachments (UVTAs are incorporated into one
line-up per unit), breaker cleaning and inspecting (with breaker ,
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.. removed from the compartment), cell cleaning and inspecting, and
relay / control wiring compartment cleaning and inspecting. The
maintenance interval specified in the procedure was:
.(1) Recommended 18 months for the UVTAs
(2) ~ Required 60 months for four Unit 2 line-ups in Technical:
Specification 3/4.8.2.6.lb,.which is related to containment
penetration overcurrent protection.
(3) Recommended five years for.all other switchgear.
Observations
Procedure ~ 52PM-R22-001-05, Rev. 3, was reviewed in detail by the.
. team and all comments were discussed with Senior Plant Engineers (one
from the maintenance group and one from the systems engineering
group) and a Maintenance Foreman. One general comment made by the
- NRC, which applied to the circuit breaker portion. of the procedure
'(Step 7.5), was that the procedure lacked sufficient detail.
Relative to cleaning, inspecting and lubricating the breaker contact
assembly, procedure Steps 7.5.6.2, 7.5.6.3, 7.5.6.9, and 7.5.6.'11
apply. These steps do not adequately address'the following-PM items:
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Inspect all current carrying parts for evidence of overheating.
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Operate the breaker slowly, by using .the spring blocking device.
Check. for binding or friction and correct if necessary. The
manufacturer's instruction book gives detailed instructions on
this step.
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Inspect primary contacts for burns or pitting. Wipe contacts
with clean cloth. Replace badly burned or pitted contacts.
Rough or galled contacts should be smoothed with a crocus ~ cloth.
Resilver where necessary.
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Inspect arcing contacts for uneven wear or damage. 1
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Figure 2-C, Contact Dimensions, indicates six dimensions that
could be verified. I
Relative to Step 7.5.6, Breaker Contact Assembly, the team made the
following comments: j
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The contact resistance test criteria of 500 micro-ohms should be
50 micro-ohms. I
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In Step 7.5.6.5, the word "megger contacts to ground" should !
read "megger contacts to frame."
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The ' reference diagrams were difficult to read because of the
small print.
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Addition of QC hold points should be considered.
Step 7.5.9.2 simply states " Clean and. inspect all _ parts [of the
L _ stored energy mechanism]." The following PM items are not adequately
addressed:
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Remove spring charging motor brushes. Measure brush length and
compare to acceptance criteria. Replace brushes if necessary.
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Inspect motor support for loose or ' missing bolts and tighten or
replace as necessary. (Refer to NRC Information Notice 88-42)
In addition to' commenting on the level of detail in the procedure,
the team also commented relative to items not identified for
inspection that should be inspected. The program does not include
. periodic inspection and insulation resistance measurement of the
switchgear bus. . The outgoing cable compartment is not inspected,
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although the licensee stated that thermographic imaging of the cable
termination was included in the predictive maintenance program.
Procedure 52-R22-001-0S did not incorporate steps for inspection
of the ; potential transformer compartment, although apparently
maintenance 1 work orders to inspect the PT compartment were carried
out using that _ procedure. Furthermore, the NRC questioned the
five year maintenance interval since it was much longer than the
one-year interval recommended by the manufacturer in his instruction
book. (Refer to W Instruction Book S.0. 25-Y-9285-1, dated June
1975, page 48, Hatih No. SX-13698 )
Discussions were also held with key personnel in the training
department relative to preventive maintenance on the 4160 volt
switchgear. At present, there is no lesson plan but the licensee is
in the process of developing a lesson plan for that topic as part of
Phase V of the INP0 training program. The licensee also stated that
outside courses were not provided in .this area. Therefore, plant
electricians have not received training at Plant Hatch that could
offset the lack of detail in the 4160 volt PM procedure.
The licensee's response to the above comments was as follows. The
maintenance engineer who was involved in the discussions agreed to
review and upgrade procedure 52PM-R22-001-0$ with the objective of
providing a detailed inspection checklist appropriate to the i
circumstances. The maintenance Engineering Supervisor stated that
only four failures of the 4160 volt switchgear were reported to NPRDS
for the two-year period from January 1,1987, to December 31, 1988.
He also stated that a survey was conducted of (five) other nuclear
plants. Each of these plants reported using a PM interval greater
than one year. ,
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Conclusion
The NRC's position on the matter is as follows. The NRC's SER for
Generic Letter 83-28, Item 3.2.2, Check of Vendor and Engineering
Recommendations for Testing and Maintenance (All Other Safety-Related
Components), transmitted July 29, 1987, is relevant to these inspec-
tion findings. Page 4 of that SER states: " Item 3.2.2 requires
licensees and applicants to submit the results of their check of
vendor and engineering recommendations. The licensee's supplemental
responses dated August 21, 1986, and July 1, 1987, to Item 3.2.2
stated that a procedure upgrade program has been developed and
designed to provide assurance that appropriate vendor and engineering
information is either included or referenced in the procedures. The
licensee indicates that Hatch Procedure DI-ADM-05-1085, Rev. 2,
included a requirement to ensure that applicable vendor manuals
and vendor and engineering recommendations are reviewed and are
included in all procedures, not just test and maintenance procedures."
Procedure 52PM-R22-001-0S, Rev. 3, had been through the procedure
upgrade program; however, all the manufacturer's recommendations were
not incorporated into procedures nor was proper documented justifica-
tion provided for any deviation from the recommendations.
Procedure DI-ADM-05-1085, Rev. 2, was inadequate because it did not
contain the instructions that would ensure that applicable vendor
recommendations were included in the plant procedures as stated in
the correspondence described above. The licensee is responsible for
the maintenance program. Therefore, the licensee may, on occasion,
deviate from vendor recommendations, but any such deviation should
be justified by auditable documented analysis. The 4160 volt AC
switchgear PM procedure is an example of the inadequacy of the
controlling administrative procedure. Therefore, this matter
represents a violation of NRC requirements, and is identified as
Violation 321, 366/89-02-01, Inadequate Administrative Procedure.
b. Lack of Corrective Action on HCU Bolting
Background
The Team had reviewed NRC IN 87-56, " Improper Hydraulic Control Unit
Installation at BWR Plants," previous to this inspection. IN 87-56
provides details of inadequate bolting on HCUs at two BWRs and notes
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that:
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The CRD system controls the position of the control rods within
the reactor core either to change reactor core power or to
rapidly shutdown the reactor (scram). The HCU is a major
monent of the CRD system that incorporates all the hydraulic,
.ctrical, and pneumatic equipment necessary to move one CRD
mechanism during normal or scram operations. This equipment,
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which includes the accumulators, CRD insert lines, CRD withdraw
lines, and scram' valves, is supported by the HCU frames.
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If a sufficiently large number of HCU frame bolts are missing or
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loose, a 3afe Shutdown Earhquake (SSE) could result in damage
affecting the_ ability of the CRD system to control .the
l_ ' positioning of the control rods.
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In addition, damage to a CRD
l- withdraw- line could result in a small-break loss-of-coolant
accident in the area of the HCUs.
The Team completed an inspection of'the bolting for a majority of the
Unit I and Unit 2 HCUs and identified one case of partial bolt
engagement for an upper frame mounting bolt (Unit 2 HCU 46-23) and
several cases (more than a dozen randomly dispersed betwen both
units) where the upper frames of back-to-back HCUs appeared to
indicate inadequate bolt torquing (upper plates exhibiting a gap
rather'than continuous contact).
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Documentation Examined-
Cognizant licensee personnel informed the Team of previous NRC
' Violation- 50-321/86-20-02, regarding inadequacies.(missing lock
washers) from the Unit 1 bottom HCU mounting bolts. Corrective
actions' were to install the missing washers and verify torque of
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bottom bolts to 45-50 foot pounds for Unit 1 HCUs (MWO 1-86-7330) and
verify torque of bottom bolts to 45-50 foot ' pounds for Unit 2 HCUs
(MWO 2-86-3811).
The Team examined the documentation listed above and additional
supporting documentation as follows.
-
. December 17, 1987 Correspondence from S. B. Tipps to C. T. Jones
(Log: LR-REG-029-1287), regarding improper hydraulic control
unit installation at BWR Plants - This correspondence states
that in the process of resolving these NRC items, it was
determined that during the construction of both Units 1 and 2
the torque value for the HCU hold-down bolts was not specified.
The torque value information was subsequently obtained from GE
(letter G-GPC-6-266 of July 22,1986).
-
July 22, 1986 Correspondence from GE to GPC (G-GPC-6-266),
regarding Hatch I and 2 Hydraulic Control Unit Dynamic
Qualification - Regarding loose hold-down bolts and missing flat
washers, this correspondence states that, subject to the
conditions that no previous upset or faulted events have
occurred at the Hatch I and 2 site and that the six extreme
bolts in the eleven bolt hold-down pattern are in place and are
at least snug tight, the installed HCU's will remain operable
through at least one future faulted event.
_ _ _
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The.. six extreme bolts are the four -bottom bolts and two upper -
bolts. The bolt torque values are given in.tho_ Reference 2-test..
specification. The relevant pages of that specification : are
provided as Attachment'1.
LThe team noted that Note 1 of Attachment 1 to GE Correspondence
G-GPC-6-266.provided a limiting torque value of 50 foot pounds for
the.0.50 inch bottom bolts and 15 foot pounds for.the 0.375 inch top
capscrews. The Team requested occumentation verifying the torque for
the Unit 1 and Unit 2 top bolts.(capscrews),
Cognizant licensee personnel responded that the torque levels had not
been previously checked for the top bolts but would be ' accomplished
during. this inspection.
Licensee Action
The'. licensee' completed. activities to torque the HCU.back-to-back top
plate mounting capscrews to 15-25 foot pounds for all HCUs and
confirm full thread. engagement. . (MWO No. 1-89-010977 for Unit.1 and
MWO. No. - 2-89-00727 -' for. Unit 2). Results revealed excessively loose
bolts on 18 HCUs for Unit 2 and 5 HCUs for Unit 1.
Conclusion ~
After review of the above, the Team informed cognizant licensee
personnel . that this. issue was considered a lack of conformance to
-10 CFR 50, Appendix B, Criterion XVI- and_ would be identified as
Violation'321/366-89-02-02. Failure to Complete Adequate Corrective
Action (See paragraph 3.c. for an additional example of this
violation).
c. Failure to have unique fittings on; the plant service air system
(breathing air) outlets.
Background
The licensee is required to have unique. fittings on the service air
system to prevent inadvertent use of nonrespirable air when using
supplied-air respirators. The need to have unique fittings was
documented.in a study made in 1981. The licensee also identified
failure to have unique fittings for the Service Air System in 1987
and again in 1988. However, in 1989, the team determined that the
licensee still did not have unique fittings for the service air
system.as identical fittings were found on Instrument-Air and Service
Air Systems. See paragraph 4.m. of the report for further details.
l
L____ _- . _ - . . _ - _ _ _ - _ _ _ _ _ - _ - - - _ _ _ _ _ _
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Findi igs
l
10 CFR, Appendix B, . Criterion XVI, requires that measures shall be
established.to assure that conditions adverse to quality, such as
' deviations and nonconformances are promptly. identified and corrected.
The, team determined that the licensee'in the past had identified at
least three': examples of noncompliance relative to. breathing air
L fittings. and failed .to take adequate corrective action to preclude
repetition '(see report paragraph 4.m below for details). The team
stated that . failure to take prompt and adequate ' corrective action
for not having unique breathing air fittings'.is a violation of the
Quality Assurance. Program,10 CFR 50, Appendix B, Criterion XVI and
is another example of Violation 321,366/89-02-02, identified in
paragraph 3.b. above.
d. Failure to Sample the Plant's' Breathing Air System
Background
Administrative Control Procedure, 60AC-HPX-006-0S requires that
respirable air supplied by air compressors and cylinders meet the
minimum requirements of Grade D air as prescribed by the Compressed
- Gas - Association Commodity Specification G.7-1-1966. The procedure
further requires that respirable air be sampled at least quarterly.
Further details are provided in paragraph 4.m.
Finding
Contrary to the above, the licensee failed to follow procedure
60AC-HPX-006-0S during the fourth quarter of 1988, in that the
respirable air used to fill self-contained breathing apparatus was
not sample and analyzed. .This was identified as Violation
321,366/89-02-03, Failure to Take Breathing Air Samples.
e. Failure to Follow Acceptance Criteria for Weld Patch on Unit 2
Reactor Building Roof Drain
Background
During a general inspection of the 130 foot elevation of the Unit 2
Reactor Building, the Team noted a welded patch (approximately
3" x 4" x 3/8" plate) on the 20 inch schedule 10 Roof Drain (MPL
No. 2T55-RSD-5) which exhibited apparent welding discrepancies. The
1/4 inch fillet weld' attaching the patch to the drain pipe exhibited
poor weld profile and excessive grinding (more than 1/16 inch below
pipe' surface). Further, the welder's ID was not stamped on the pipe.
- - - - - - -
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The Team noted that this section of roof drain also served to main- i
tain isolation of secondary containment and questioned the licensee
as to whether adequate minimum wall thickness had been maintained
for the excessively ground area. The Team further requested for
review a copy of documentation showing acceptance of the present
condition.
Licensee Action
Cognizant licensee representatives completed a deficiency card
(No. 2-89-0505) during this inspection to accomplish ultrasonic
thickness measurements which indicated that the ground area was ,
reduced to 0.165 inch thick; i.e. less than the manufacturer's i
tolerance of 0.219 inch. However, the gouge did not violate the
0.145 inch design minimum wall thickness.
Documentation Reviewed
The Team reviewed documentation associated with MWO 2-85-1424 and QC
acceptance of the initial weld dated March 27, 1985. Cognizant
licensee personnel agreed that the initial acceptance had been in
error since the weld inspection plan imposed at that time (A-MB-01,
Rev. 1) prohibited excessive grinding and required an acceptable weld
profile and the welder's ID stamp on the pipe.
During this inspection, cognizant licensee personnel completed an
independent review of other welds accepted by the QC inspector 1
involved and conducted additional training on weld acceptance. QC
management personnel felt that the error on MWO 2-85-1424 had been an
isolated example.
The Team completed an independent verification by examination of the
QC program, interviews wie several QC inspectors and reinspection of
several welds recently accepted by the QC inspector involved.
Further details of this review are included in paragraph 4.k. below.
Conclusion
After completion of the above, the Team concluded that the error in
initial acceptance of the welded patch had been an isolated example,
and that overall, QC at Plant Hatch was a strength in both program
and implementation.
The Team informed cognizant licensee personnel that this issue would
be identified as Violation 366-89-02-08, Failure to Follow Acceptance
Criteria for Weld Patch on Unit 2 Reactor Building Roof Drain.
However, due to the low safety significance, isolated occurrence, and
previously completed licensee corrective actions, this violation will
not be cited.
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a
r ~f. Programmatic Link Cetween Maintenance Procedures and ASME B&PV
Section XI Requirements
t-.
Background
b During the Team's examination of the Maintenance Program, a need was
identified for additional requirements to ensure proper coordination
and testing for ASME Section XI .:omponents. An equivalent need was
also identified for the predictive maintenance program, particularly
for vibration analysis. Details are listed below:
[ -
Procedure 50AC-MNT-001-0S establishes the requirements and
b responsibilities for the control of maintenance activities at
Plant. Hatch. This procedure details requirements for intiating
'
and processing MW0s. Additional details for MWO processing are
included in procedure DI-0AP-10-0588N. Neither 50AC-MNT-001-0S
nor. DI-0AP-10-0588N clearly specify that for a Section XI
component, - Section ' XI . programs are to be reffered to for
.
determining post maintenance testing requirements. This
omission is of concern due to the potential differences in. post
,
maintenance tests (functional ~ tests) . required for Section XI
components'versus other components.
-
Preventive. Maintenance Proceduro 53PM-MON-001-0S describes the
- - ' method used to obtain and analyze vibration analysis data for -
the purpose of detecting. incipient failure of equipment. The
, program is intended to apply to. preventive maintenance only and
l not to interface with any Technical Specification requirements.
The procedure is applied by maintenance engineers and does not
necessarily require that an MWO be issued. Section 5.2.2
states:
"The vibration monitoring program governed by this proededure is
for preventive maintenance purpcses only. When actual vibration
levels exceed preidentified. suggested maximum recommended
levels, this does not necessarily mean that the associated
equipment is inoperabl.e, instead the information is intended for
use as a diagnostic tool to indicate the need to perform
additional testing, schedule future maintenance or do other
analysis of equipment condition."
The above is of concern since there is potential that the
referenced vibration analysis can apply to a Section XI pump.
In that case, if vibration results exceed the requirements of
ASME,Section XI, Subsection IWP, Table IWP-3100-2 of Section XI
must take precidence and proper actions taken to satisfy
Section XI requirements.
_ - _ _ _ _ _ _ . _ - _ _ _ - _ _ _ - _
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Conclusion
'
Cognizant licensee _ personnel agreed that correction to 50AC-MNT-003-0S
should be completed to require that an MWO for a Section XI component
be so identified. However, licensee personnel were concerned' that-
-a: predictive. maintenance vibration analysis not .be considered an
equivalent to the Section XI type test. The ' Team concurred with
'
this: reasoning but noted that a full spectrum vibration analysis was
. presently being taken when monthly readings indicated a potential
problem (1.e. a situation most apt to be equivalent to the Section XI
alert or action range) and that the computer comparison'to.Section XI
"
type vibrations in mils could be automatically accomplished. After
. . further. consideration the licensee agreed to the need for tying
vibration test results to Section XI requirements.
3
The . Team informed cognizant licensee personnel that NRC concern
,.
'
regarding' a programmatic link between Section XI requirements and
i
- procedures 50AC-MNT-001-0S and 53PM-MON-001-0S would be identified as
IFI 321,366/89-02-04, Programmatic Link Between Maintenance '
~ Procedures and ASME Section XI Requirements,
g. Inspection of RHR Hanger Weld Removal
'
l' Background
During a general inspection of QC activities, the Team became
involved in discussions between engineering and QC supervision.
regarding final review and close out of MWO 1-88-5022. This MWO
accomplished modifications of several RHR supports located in the
Unit 1 Reactor Building. Changes included installation of a new
,
embed support plate and adjustable rigid strut to the existing pipe
clamp.
Additional repairs to support E11-RHR-H293 were required due to slag
pockets in the existing welds which attached the pipe clamp support
lugs to the pipe. The support lugs were removed and the weld area
ground to sound metal. Field weld IE11-HFW-059 was made to repair
!: the ground area. However, there was no indication that a QC
inspection of the excavated area (fit-up inspection) was done.
Technical Requirements
The Team noted that ASME,Section XI, Subsection IWA-4130, requires
that:
Repair operations shall be performed in accordance with a program
delineating essential requirements of the complete repair cycle ...
including the flaw removal method, method of measurement of the
cavity created by removing the flaw, and dimensional requirements for
reference points during and after the repair ....
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l. The. Team also noted .that paragraph 7.1 of licensee procedure
42EN-ENG-014-0S, states'that:
~~
%' Documentation required by ASME Code is included in the scope of
Repair / Replacement Program and shall include the following as
applicable: a description of the flaw, the method which
^
- ... revealed the flaw and the location of the flaw; the flaw-removal-
method and the depth of excavation; and an evaluation of' the
g. flaw or failure to, ensure the selected repair or replacement is
F suitable prior to repair.
'
Documentation Review
! -
The . Team reviewed quality documentation associated with MWO
.1-88-5022 to re-verify the statements made above.
-
The Team reviewed additional quality documentation to verify
that:
-
A NDE (PT/MT) was intially required for the lug removal on
'
. Hanger- E11-RHR-H293 (Step 8A on Work Process Sheet No.
81-058-M105).
The. MWO was intially considered a Section XI replacement /
repair (R/R Applicability Checklist, dated September 8,
1988).
The repairs to 1E11-RHR-H293 received an engineering
exclusion (R/R Checklist, dated November 15, 1988) with
basis as follows:
"When original construction installed lugs, .2 slag pockets
were left in the pipe wall. This revision of the original
MWO is to base metal repair pipe wall. This part of MWO is
not in R/R program."
Licensee Response
Cognizant licensee personnel informed the Team that the engineering
decision to exclude the repair weld 'from the Section XI R/R program
also removed any requirements for NDE of the excavated area.
Cognizant licensee personnel were unable to provide any alternative
assurance that the flaw had been completely removed and were not
aware of any programmatic requirements for flaw removal evaluation
outside of those imposed for Section XI R/R components.
Conclusion
Cognizant licensee personnel informed the Team that a question
regarding omission of the fit-up inspection for weld IE11-HFW-059 had
been raised by the ANII during review of documentation for MWO {
4
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1-88-5022 and final resolution had not yet occurred.
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'At the end of the inspection, this_ problem was still being evaluated
by'the licensee. The Team informed cognizant licensee personnel that
NRC cencern - regarding programmatic requirements for examination /
evaluation of welding flaws would be reviewed during a future inspec--
tion andcident.ified as IFI 321,366/89-02-05, Inspection of RHR Hanger
l _.W eld Removal.
L
h. Design Verification of Containment Isolation Valves T48-F310 end F311
The Team examined licensee activities in response to NRC' Generic
- Letter _ (GL) 88-14. The purpose of GL 88-14 is to request that each
licensee review NUREG 1275, Volume 2 (Operating Experience Feedback
Report - Air System Problems) and then perform a design and operation
verification of their Instrument Air System (IAS). . Verification was
to include:
Item 1- -
Verification by test that actual instrument air
quality is consistent with the manufacturer's
recommendations for individual components served.
Item 2 -
Verification that . maintenance practices, emergency:
procedures, and training are adequate to ensure that
safety-related equipment will function as intended on-
loss of instrument air.
' Item 3 -
Verification that the design of the entire instrument
air system including air or other pneumatic accumula-
tors is in _accordance with its intended function,
including verification by test that air-operated
safety-related components will perform as expected in
accordance with all design basis events, including a
loss of the normal instrument' air system. This design
verification should include an analysis of current
air-operated component failure positions to verify
'that they are correct for assuring required safety
functions.
A final requirement, Item 4, was to provide a discussion of the
licensee's program for maintaining proper instrument air quality.
Background
The Team reviewed the licensee's initial response, dated February 10,
1989,'to GL 88-14 and noted licensee statements as follows: ,
-
The' reviews and/or investigations to date indicate that the
design, installation, testing, operation and maintenance of the
instrument- air systems at Hatch Nuclear Plant are adequate to
ensure the proper and reliable operation of pneumatically i
operated, safety-related equipment.
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Upon completion' of the . additional evaluations, a subsequent
report will - be - submitted. This report is scheduled to be
-provided by June 1, 1989. A final report will be issued upon
l- completion of all actions associated with GL 88-14.
Documentation Review
The . Team held discussions with cognizant licensee personnel and
reviewed additional documentation associated with GL 88-14 activities
as follows:
-~ Documentation associated with air sampling and station service-
air compressor (SSAC) maintenance and reliability.
3
--
. Documentation associated with design verification of MSIV,
'
Containment Vacuum Breakers, and Containment Isolation Dampers
(valves).
A ' complete. list of documentation reviewed is -included in
paragraph 4.1.
Conclusions
The. Team. noted that the licensee had completed comprehensive
activities. in response' to GL 88-14. However, design verification -
was not yet complete for critical . components (Valves T48-F310 and
F311). These valves are redundant to the torus vacuum breakers, and
use instrument air pressure' to maintain the valves in the closed
position. Upon loss of air pressure, these valves are designed to
open to ' allow the vacuum breakers to perform their safety function
of preventing containment implosion. When the valves fail open, the
isolation function of the valves is lost.
The Team further noted that the Unit 2-valves had failed LLRT testing
during the last refuel outage. The Team informed cognizant licensee
personnel that NRC concern regarding adequate design verification
would be identified as IF1 321,366/89-02-06, Design Verification of
Containment Isolation Valves T48-F310 and T48-F311. The resolution
of this matter by the licensee will be reviewed during a future NRC
inspection.
1. Failure to Have Adequate Procedures for Sampling Plant Breathing Air
Background
The licensee is required by 10 CFR 20 to sample respirable air to
meet Compressed Gas Association Commodity Specifications G.7-1. The
, specifications define limits for oxygen content, hydrocarbons, carbon
monoxide, and carbon dioxide. The licensee is required by licensee
Technical Specifications 6.11 to have procedures consistent with
= - _ _ _ . . - _ -
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10 CFR 20. The Technical Specifications also. requires that the
procedures be approved, maintained and adhered to for all operations
involving personnel' radiation exposure.
Finding
The team determined that a licensee audit of health physics was being
made during the inspection period and that the lack of procedures for
sampling and-analyzing Grade D air had.been identified by the Quality
Assurance Auditor. The. inspectors determined that the licensee had
also . initiated some corrective action. concerning sampling
requirements. In order to review the licensee's corrective actions
forf failure : to have ' a written procedure, IFl 321,366/89-02-07,
Written Procedure for Sampling Breathing Air, was identified.
4 :
4. Inspection Details
' The Team performed walkdown inspections, observed maintenance in process,
reviewed maintenance history records and MW0s, and reviewed . maintenance-
procedures to evaluate the overall maintenance program. The following
paragraphs summarize the details of the. inspections / reviews performed.
,
a. Wal kdown . Inspections
The Team . conducted a ~ general inspection of Units 1 and 2 turbine
buildings, control buildings and reactor buildings. The inspection
included observation of general equipment condition, housekeeping
practices, deficiency condition and control, and identification
. practices for permanent plant equipment. In addition to general
<
cleanliness, mechanical equipment was observed for . water Land oil
leaks, corrosion, lubrication, proper fasteners, evidence of
vibrations,.etc. Electrical equipment was observed for cleanliness
of equipment. and general area. (floor, etc.), painting, equipment-
grounding, corrosion, control wiring terminations, broke or _nissing
relays, meters, lamps,' etc. , proper labels, conduit and tray '/111 and-
l support, floor and wall penetration seals, bushing tightness,
L lighting, missing fasteners, cable tie wraps and supports, wire and
cable'nos., namplates, etc.
Appendix C is a list (not all inclusive) of the equipment and areas
observed.
The following is a list of deficiencies identified by the team:
-
Small leaks at valves 2C11-F0468, 2011-F005, IN21-N8178 . and
IN43-F138.
-
Small steam leak at 1 inch union, E41 System, Unit 1, HPCI Room
-
Missing insulation - About 2 feet of 2 inch diameter pipe near
bottom of Unit 2 Main HPCI pump
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17
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Conduit support and tubing on floor in demineralized valve nest
(Unit 1)
-
Tag for valve IN21-F447B laying on floor
-
Valve 2P51-F087 handwheel tied in place with cable tie wraps
-
At Unit 1 Main Generator Exciter Housing - conduit loose,
bushing loose
-
At Unit 1 Main Generator - stator cooling piping, insulation
covers broken, and lamp covers missing at diode indicators
-
IP63-B001A Turbine Building Central Water chiller - water on
floor
-
208V MCC IG 1R24-S0-45 Frame 10 (RFPT 1A Hi Pres
Steam) -corrosion on starter pan
-
Turbine building leak allows rain water to drip near or into
-
Stator Water Cooling System - turbine building ground floor
Small oil leak at pump B
- Valve IN43 F138 Y-61 stator cooling make-up inlet - leaking
-
H2 and stator cooling panel IN43-P001 - annunciator "VAC TR OIL
LEV HIGH-LOW" flashing
-
At 1R23-S002 600V SWGR IB - Some compartments have heavy dirt
inside, example - normal feeder to turbine building chiller 208V
SWGR 1A
-
4160-600V station service transformers have PCB insulation
-
Transformer 1R23-S001 Small leak at fill valve
-
Corrosion on diesel generator battery racks
-
At batte ry 2R42-S002A, battery 2A - scrap and debris in sump
drain may plug drain
-
At cooling tower electrical house near tower No. 4 - metal
building siding stored on vent fan enclosure
-
Cooling tower No. 4 - loose grounding wire on SW corner
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At feedwater. pump N21-C003A-
208 V panel - most cover bolts missing
condulet cover loose
'
scrap sheet metal stored in area
valve IN21-N817B leaking-
-
Welding problems associated with a weld patch on the Unit 2
Reactor Building roof drain (see paragraph 3.e.)
-
Redundant conductivity recorders (CE-N424A and CE-N424B) for
recording condensate conductivity downstream of two of the
demineralizers were inoperative. This was more a loss of
convenience than a problem with system operation or safety as
the conductivity is recorded by larger more accurate recorders
in the demineralized area.
The' Team reached the following conclusions from the above
examinations:
,
-
General Equipment Condition
Most of the maintenance items noted had not been identified by
the licensee. However, all of the items were .reselved during
the inspection period . by. issuance of work. . orders or other
acceptable means. The licensee presented similar punchlists of
items. identified durf tp their walkdowns. When compared with the
overall equipment conaition and plant maintenance, the various
housekeeping and equipment condition problems listed above were
considered relatively minor.
-
Housekeeping
During the above inspection, the Team observed general
housekeeping conditions. Programmatic Control of Housekeeping
is maintained by procedure 51GM-MNT-002-0S.
The Team noted a general high level of cicanliness within all
areas of the plant. The Team conse sus was that general plant
housekeeping is a major strengto.
-
Deficiency Identification and Control
The Team noted relatively few discrepancies without MW0s issued
for correction. Further, no major discrepancies were identified
without MW0s issued.
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.The Team consensus was that deficiency - identification and
control was a programmatic strength'at Plant Hatch.
-
ID of Permanent Plant Equipment
During the above inspection, the Team observed that
identification of perranent plant equipment was never in
question, due to the use of equipment identification tags,
decals, etc. which were prominately located, securely att' ached,
and of a size to be clearly legible.
The Team consensus was that identification of permanent plant
equipment was a programmatic strength at Plant 1.atch.
b. Repeat Failures - LPCI Inverters
During the evaluation of the maintenance program, various instances
of ' what appeared +o be repeat troubles / failures were examined.
Discussions were held with various licensee personnel concerning
repeat problems with LPCI inverters. The following summarizes the
discussions and examinations:
During a period of 17 days, LPCI inverter 1R44-S002 was found to have
blown fuses twice and LPCI inverter IR44-S004 had a blown fuse once.
After the third failure, an Event Review Team was organized to
examine the problem.
L
Root cause' analysis revealed that the failures were due to the
installation of incorrectly rated parts which were supplied by the
vendor. This problem was unique to. this plant in that the output
voltage for these irverters is 600V AC (Rather than the more common
480V) and the parts of the Plant Hatch .LPCI inverters must be
modified by the manufacturer. The parts were identified by the same
part number as the 480V part, however, and therein was the problem.
Failure to uniquely identify the modified parts led to the use of
- underrated parts. It should be noted further that there are. similar
'
inverters installed in other newly installed mystems wh'ch have 480V
AC as the required output.
-The reports reviewed were complete and indicated that good
engineering practices had been employed in solving this problem,
c. Feedwater Control System
Units 1 and 2 have experienced several feedwater control problems.
These problems were also investigated relative to repeat failures.
h Following is a summary of the licensee's approach to solving these !
problems:
-
It was determined by analysis of failure rate and consultation
with GE that a certain manufacturer's capacitors were failing in
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the GEMAC compone'nts. I&C. started ~a program to change out these '
!
type capacitors with more reliable ones. 1
--
Based on a GE SIL recommendation, a DCR was initiated to remove
the density correction instrumentation in the feedwater. control
loop. This DCR removed approximately .11 modules (GEMAC) which
f
made.the control loop more reliable due to a lower probability
of a component failure. Of the GEMAC components that were left
'
in the loop, the majority were replaced with TOSMAC components,
a GEMAC eauivalent made by Toshiba. Any components not having
TOSMAC units for replacement were replaced with components
l
(GEMAC) . refurbished by GE. This DCR was completed on Unit 1
' this past outage and will be completed on ' Unit 2 during tne fall
refueling outage.
-
A. problem was found with the cascade switch on the GEMAC
controllers. The switches were found to be intermittent. A DCR
was initiated to solder a jumper across the switch facilitating
a much more reliable continuity path. This DCR has been
' implemented on Unit 1 but not on Unit 2.
-
Recorders have been connected to various_ points in the feedwater
control loop so . that if a failure does occur, data can be
collected for an accurate determination of the failure mode.
-
Feedwater control problems on both Units have been reduced from
feedwater swings occurring frequently, including Unit trips, to
a feedwater level dip of approximately 4 inches at which time
the controller .immediately catches the decrease and compensates
for it. These fluctuations happen very infrequently. The
.overall performance .of the feedwater control system has been
vastly improved.
In the above. listed instances, the licensee solved their problems
using a variety of different methods.
During review of the RHR system, the Team examined the licensee's
responsive actions to NRC IN 87-30, Cracking of Surge Ring Brackets
in large General Electric Company Electric Motors. The RHR Pump
Motors 1E11-C002B and IE11-C002D had been modified by installing new
improved design surge ring brackets. The brackets for RHR Pump
Motors 1E11-C002A and IE11-C002C had been inspected and no problems
were found. The surge ring brackets for the A and C RHR Pump Motors
will be replaced during the next refueling outage. The work will be
performed under Design Change Request No.88-190, which covers the
four RHR pump motors and the two core spray pump motors. It was
further determined that the parts were onsite for the modifications.
_ - _ - _ _ - _ _ _ _
_ _ _ __
-
, ,
1..
,
E 4
For Unit 1, core spray pump motor IE21-C001B had been modified and- )
'
.
'
core spray pump motor 1E21-C001A is scheduled to be modified during ,
the next refueling outage. The motors for the core spray and RHR
pumps for _ Unit 2 are ' a different design and will _ not require
modification.
I
It appears that the licensee responded well to this industry / vendor j
-initiative'and the NRC Information Notice 87-30.
e. Observation of In-Process Maintenance
~(1) Repair _ of Intermittent Alarm on Station Service Battery Charger l
Observations
The team observed the performance of MWO 1-89-00722 which was
issued to repair an intermittent alarm condition on station
service. battery charger 10. The AC voltage failure relay which
was.specified as the part to be replaced was incorrect. The MWO
was revised and the under-voltage alarm relay was specified.
The steps required to revise the MWO were followed including QC
verifications.
The old under-voltage alarm relay was tested and the
l repeatability was cut of tolerance. A new relay was installed.
When the charger was re-energized, the AC voltage' failure relay
chattered. Voltage measurements taken indicated low output
voltage (84 VAC versus 125 VAC). During the troubleshooting to
determine the cause of the low AC voltage, it was discovered
that the control fuse holder cover was loose. When the cover
was fully in place, the AC voltage returned to normal. The fuse
holder was examined and all fuse clips and cover fingers were -
cleaned to ensure proper electrical contact. The charger was
returned to service. Later follow-up of the completion of this
MWO revealed that the battery charger was only tested for proper
operation. There was no evidence that the alarm function was
,
tested.
Conclusion
Additional examination of this MWO and the process by which MW0s
are revised revealed a procedural weakness associated with
proper review of nw:essary post-maintenance test changes for
revised MW0s. This is discussed in paragraph 5.c.(1).
_ _ _ _ __ _ _ _ - _ _ _ _ _ _ _ _ _ -
__ _ _ _
.
.,
, :> .
22
- (2)' Cooling Tower Motor Changeout
'
Observations-
The Team. observed portions of a cooling tower motor changeout,
' protective. relay calibration, recirculation pump motor generator
set-brush surveillance and 480 Volt circuit breaker trip device
calibration.
Conclusion
In. all cases, procedures were being followed and data carefully
.
. recorded.
(3) Replacement of Programmable Controller in Demineralized Building
Observations
Thisf activity was assessed with respect to the adequacy of the
mainte' nance effort, whether applicable procedures were followed,
and whether operations personnel were aware that the subject
maintenance was being performed.
During this ~ activity, an I/C technician was observed while
replacing a backup battery for the programmable controller in-
. the demineralized building. This individual appeared to be
well qualified for the. task. He had previously worked.for the
manufacturer of -the control _ler. The technician received the 1
'
folder for MWO 1-88-8411 from the shift foreman; obtained a
sign-off from the . shift supervisor in the control room; and
thereby informed operations personnel that the maintenance
effort was to be performed; obtained the ' spare part (battery)
from the warehouse; and replaced the battery. Proper installa-
tion of the battery was shown when the annunciator light for
the controller cleared.
Conclusion
The task was well performed, applicable procedures were .
'followed, and the control room personnel were aware that the I
activity was underway. !
(4) Operability Test for RHR Pump 2E11-C002A
Observations
This activity was assessed with respect to adequacy of the
maintenance effort, whether applicable procedures were followed,
and whether operations personnel were aware that the subject
maintenance was being performed.
i
s
._. _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . . . _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ . _ . _ _
,_ -_- - - _ _ _ - - _ _ _ - - - -- _.
_ - _ _ _ _ - - _ _
. _ - _ _ - - - - _
sv -
l 4 '.
, l
.l
23 l
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!. 1
During this activity, maintenance testing regarding RHR pump
2E11-C002A was observed. Procedure 34SV-E11-0012S was . used
to determine operability of_ the pump. The- following : actions ,
were performed during the operability test: (1) telephone -l
communication with control room personnel occurred, (2) oil i
level was checked,-(3) verification that the service water valve j
opened, (4) the pump ran for five minutes, (5) the discharge '
'
pressure of 190 PSIG was read from. the appropriate gauge, and l
(6)..the discharge pressure was conveyed by phone to the control ;
room.
Conclusion
.
. The task was well performed, the applicable procedures were l
followed, and 'the control room personnel were aware thet the
activity was' underway. !
(5) Preventive Maintenance on Fire Pump 1x43-C001 j
Observations .!
!
. The Team observed preventive maintenance on electric fire ,
protection pump 1x43-C001 (MWO 1-88-08388). l
l
The maintenance mechanics were working from a copy of l
Section 7.7 of procedure 52PM-X43-006-15. post-Maintenance i
testing of pump temperature and vibration as well as operability
tests were compitted.' j
In.the course of performing the above preventive maintenance,
the craftsmen noted.that the relief valve was lifting while the
pump.was running and' deficiency card 1-89-1209 was written. l
)
During this inspection, the Team also noted a small water leak i
from jockey fire pump IX43-C003. Deficiency card 1-89-1210 was i
written for correction. !
1
Conclusion ]
The Team concluded that the fire pump preventive maintenance and
post-maintenance testing were performed in accordance with the j
appropriate plant procedures. Deficiency cards were written
for the deficiencies found in the course of the maintenance
operations. No discrepancies were identified.
(6) Motor Shaft Pinion Key Replacement j
l
l
1
_ _ - - - _ _ _ _
_ _ _ _ . ___
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<
24
E
r
<i Observations
".
The Team observed performance of MWO 1-89-308 to replace the -
pinion gear key (Part Number S/N '87160-63368) in MOV 1E41-F011
in response to-NRCLIN 88-84. Procedure 52GM-MEL-022-05~was used
and a QC inspector was present.
A' functional test on' the reassembled valve was performed and
indicated proper operating characteristics.
Conclusion
L
h The Team concluded that the above corrective maintenance was
performed in accordance with appropriate procedures. No
discrepancies were identified.
. (7) ' Overhauling of Waste Collector Pump Bearing
f
Observations
i. Following the performance of MWO 2-88-4862 to change the oil in
Waste. Collector Pump 2G11-C016,- the plant ' equipment operator
-felt'the inboard bearing and thought it was too hot.
MW0' 2-89-400 'was issued to " Rebuild" the pump using procedures -
51GM-MME-0020 and 51GM-MNT-0020. Maintenance craftsmen ignored
the " Rebuild" order and instead began " troubleshooting" the
pump. The. team observed the troubleshooting of the' pump. The
craftsmen could detect no' sticking or grinding as the pump shaft
was turned by hand. A maintenance engineer determined that the
temperature and . vibration. of the pump while operating were
normal. With a laser device, the engineer checked the alignment
of the motor shaft 'with the pump shaft and found .them properly
. aligned. Since. the pump was operating normally, the MWO was
closed out without further work.
Conclusion
The Team noted the proper activity of the craftsmen in response
to the." trouble" involved. However, the Team consensus was that
the MWO should have been more definitive regarding tasks to be
accomplished.
(8) PM on Overcurrent Rela" Calibrations
Observation
The Team observed the overcurrent relay calibration for Conden-
sate Booster Pump 2A, Phase 3, per procedure 57CP-CAL-108-2S.
_ _ _ _ _ _ _ _ _ _ _ _ _ -
~
25
Conclusion
No problems were identified.
f. Electrical Maintenance
The Team reviewed the electrical PM procedures as detailed below.
This inspection effort was directed at answering two questions:
(1) Were there procedures in existence to cover all the normal
preventive maintenance activities that should be governed by
procedures?
<
(2) Were the procedures of sufficient quality to be considered
acceptable for maintenance work at a nue'aar power plant?
Question (1) was addressed by studying the index of procedures and
discussing with the Maintenance Engir.eers any apparent gaps that
could be discerned from the index. . There were about 27 electrical
preventive maintenance procedures that applied to each unit. Motor
maintenance was included in the maintenance of the driver equipment.
Comments resulting from the procedure index review were:
-
The fact that the maintenance program does not include
protective trip testing of molded-case circuit breaku (other
than containment penetration circuits) is considered a weakness.
-
The fact that the maintenance program does not include periodic
visual inspection of the 4160 volt current limiting reactors is
considered a weakness.
To address question (2), the preventive maintenance procedure for the
4160 volt switchgear was reviewed in detail. Several comments were
generated during review of this PM procedure which represent program
weaknesses. Refer to Section 3.a. for complete details.
In order to help evaluate the effectiveness of the maintenance
organization, the team reviewed the work history for the 4160 Volt
System and the High Pressure Coolant Injection System for Unit 1.
These reports gave maintenance work order details for at least the
last two years. It was concluded from this review that repetitive
failures of these two systems had not been a problem at Plant Hatch
and that root cause analysis of problems for these two systems was
satisfactorily carried out.
Trending reports, as an indicator of maintenance work control, were
reviewed. One report, dated March 1,1989, indicated that the total
electrical corrective maintenance backlog was 52 work orders, which
is relatively very low. Another report indicated that for 1989, all
periodic / planned MW0s were completed within the allowable time.
_ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ - _-
,. -
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- - ___ --
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it
't 26
Report No'. 41, " Equipment with greater than five corrective
L maintenance work orders for'1988," indicated that repetitive failures
>
of electrical components had not been a problem.
g. Machine Shop Facilities
Observations
The Team was able to observe . general conditions and specific
- activities during this inspection for both the- clean and " Hot"
machine _ shops. The shops are well laid out with adequate space,.
equipment, and partitioning to accomplish a variety of tasks
associated with machining, cutting and welding, troubleshooting and
assembly / disassembly bench work. The clean machine shop also has
adequate ' space .and. bench cabinetry for tool storage by individual
mechanics.
The hot . machine shop has less space and equipment than the clean
.
machine shop, but large machine tools are installed and the space
L appears adequate .for a variety of " Hot" machining tasks due to good
organization. of the space involved. A. special " Bailey"
building / facility is included for CRD repair. The atmosphere of~the
L " Hot" . shop is controlled, and radiation monitors, decontamination
~
i equipment and health physics support appeared adequate.
.
Conclusion.
'
After review of the above, the team consensus was: the " Hot" and
~ clean machine shop facilities were a strength in the maintenance
program.
h. Craft Personnel and Training
. Observations
The Team completed an overview examination of all phases of the
licensee's training program for mechanical / electrical and I&C craft
l
personnel by review of programmatic procedures, courses involved, and
discussions with maintenance management personnel.
The Team also completed a review of the current interim
classification matrix records for all mechanical and electrical
craft.
In addition, interviews were conducted with a random sample of
i mechanical craft. Those interviewed were asked specific questions
! related to methods to minimize and control hot particles (radioactive
l' particulate) during grinding, troubleshooting and repair of
centrifugal pumps, inspection and repair of valves (including seat
L
i
___
14:
sc
27L
,
'
lapping, troubleshooting and repair of MOVs). General questions were
also asked, regarding the.following procedures:
o
l. 50AC-MNT-001-0S, Maintenance. Program
!~ t
51GM-MNT-002-05, Maintenance, Housekeeping and Tool
,
Control
52CM-MME-001-0S, Repacking Valves and the Adjustment
of '/alve Packing'-
<
52CM-MME-005-05, Limitorque-Valve Operator Models
SMB-0 through SMB-4 Mechanical
Maintenance
h All. questions were satisfactorily answered.
[ Findings
The . Licensee's maintenance t raining program received INP0 accredita-
tion in April 1987. -The training has been fully implemented and
includes full time- training coordinators. The training is completed
in phases. with . an additional monetary incentive attached to each
phase which ensures craft interest in advancement. An overview is
as follows:
--
Mechanical
. The mechanical training program' consists of six phases.
The completion of phases 1 through 5 is mandatory for all
mechanics. Phase 6, however, consists- of specialized
skills ~ training. All mechanics are not required to
complete all courses in phase 6.
,
-
Electrical
The electrical training ' program consists of 6 phases. The
completion. of phases -1 through 5 is mandatcry for all
electricians. Phase 6, however, consists of specialized
'
skills training. All electricians are not required to
complete all courses in phase 6.
- . -
,
.The I&C technician training program consists of 4 phases. I
- The completion of phases 1 through 3 is mandatory for all
technicians. Phase 4, however, consists of specialized
, skills training. All technicians are not required to
complete all courses in phase 4.
i
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1
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28
9
~ -
Continuing Training
Continuing training modules for mechanical, electrical and
I&C have been presented twice .a year since these. programs
were accredited in April 1987. To date, four modules have
L .been presented for each area.
'
Prior , to completion of formal craining, craft are assigned tasks
using an ' interim qualifications matrix. The matrix was assigned to
reflect evaluation of each craftsman' by' a qualification committee
based 'on .the. applicant's prior job performance, knowledge,
,
proficiency- and training. Control i s -' maintained by procedures
'
DI-MNT-10-0287N, and DI-MNT-11-0278N. Craftsmen unable to.
satisfactorily complete the. formal training -course also loose their
interim qualification and are considered not qualified for the area
of concern.
Present maintenance management goals are to have all craftsmen fully.
certified (through phase 5 for mechanical and electrical and phase 3
for.I&C) by the end of 1989.
Maintenance supervisors provide surveillance to assure that craft are
adequately trained for the job assigned. The Team reviewed an
example where supervisor surveillance during this -inspection
identified need for additional training for the craftsman involved.
The Team did identi fy a training deficiency since the training
department did not provide any training for performing preventive-
maintenance .on 4 KV switchgear. At the time of this inspection, a
lesson plan for this was being developed as part of the phase V of
INP0 training program.
Conclusion
The Team consensus was that the Plant Hatch training program is a
strength.
1. Instrument Air System (IAS)
Observations
The Team reviewed a listing of open MW0s on the P51 Service Air
and PS2 Instrument Air Systems for Unit I and Unit 2 to identify
potentially recurring problems. The Team noted repeat problems with
the station service air compressors (SSAC) and cognizant licensee
personnel provided additional documentation as listed below.
The Team also examined licensee activities in response to GL 88-14
(see -paragraph 3.h.). Documentation associated with design
verification of MSIVs, containment vacuum breakers, and containment
isolation dampers (valves) was also reviewed as listed below.
- __ _. - - - _ -
- - _ _ _ _ _ _ _
,
};y l
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.
29
L
' Documentation Review i
i
The - Team held discussions with cognizant licensee personnel and
reviewed additional documentation associated with the above as
'follows:
-
Documentation associated with air sampling and station ' service
- air compressor (SSAC) maintenance and reliability
Air Compressor Replacement Plan
-
System Engineering Concerns Regarding SSAC, dated
September:20, 1988 (Log: LR-BOP-016-0988)
Management Action Plan regarding SSAC, dat'ed
i September '21,1988 (Log: LR-MGR-006-0988)
Management approved reliability improvement action,
dated September 27, 1988 (Log: LR-MGR-009-0988)
-
Laboratory Analyses of eight air samples taken November 23,
1988, [(includes sample location, dew point (- c), particle
size (micron), oil content (ppm), carbon monoxides.-(ppm),
carbon dioxide (ppm)]
ANSI.. Standard ISA-S7.3,. Quality Standard for Instrument Air-
'
Unit 1 PM Procedure.52PM-PSI-001-IS
Unit 1 PM Procedure 51PM-P51-001-IS
Unit 2 PM Procedure 52PM-P51-001-2S
..
---
Documentation associated with ' Design Verification of MSIVs,
Containment Vacuum Breakers, and Containment Isolation Damper
(valves) ,
1
- January 13, 15 t ,' Correspondence from G. A. Goode to
S. B.' Tipps (Lv _R-PES-016-0187) regarding testing for
Unit 2 - MSIV closure times .with and without air supply.-
(Note: MSIVs'B21-F022A-D and F028A-D met the 3 to 5 second
closing time.both with and without air supply)
~
- ~ June 11, 1987 correspondence from T. Powers to J. Kane
(Log: 'LR-ENG-011-0687) regarding relief from ASME
Section XI, Subsection IWV-3415 requirements to allow
continued testing of MSIVs without isolation of the gas
supply to the accumulators during normal surveillance
testing.
Follow on correspondence'of August 3, 1987, August 18, 1987
and September 4, 1987 regarding MSIV fail-safe testing
requirements.
I * March 2, 1989 correspondence from BPC to GPC regarding
design verification of drywell/ torus vacuum breakers
(T48-F323A-L); torus / reactor building vacuum breakers
(T48-F328 A & B; and Unit 1, 18 inch purge valves
(T48-F318, F320 and F326).
- - - _ _ _ _ _ _ _ _ _
___
.
.' O
30
,
March 9, 1989 correspondence from K. W. McCracken to
L..T. Gucwa regarding design verification of containment
isolation / vacuum relief valves T48-F310 and F311 and
T48-F328'A & B.
April 19, 1988 correspondence from GPC to NRC regarding LER
88-004-01 (LLRT failure of Unit 2 valves including T48-F310
and T48-F311).
Conclusion
The Team concluded that the licensee had completed comprehensive
activites in response to GL 88-14. However, the Team noted a
continuing concern regarding design verification of valves T48-F310
and F311. Discussion provided in paragraph 3.h.
J. Preventive Maintenance (PM) Program
Observations
The Team reviewed procedures, held discussions with maintenance
engineering personnel and observed activities associated with ;
predictive / preventive maintenance. '
The Team also- examined historic documentation demonstrating use of
predictive maintenance to prevent incipient failure on large rotating
equipment.
Findings
The licensee's predictive maintenance program includes vibration
analysis, lube oil analysis, equipment performance analysis,
infrared analysis, and analysis of maintenance history. The
program is run by maintenance engineers and controlled by procedure ,
50AC-MNT-007-05. Other procedures involved include 53PM-MON-002-0S !
and 53PM-MON-001-05.
An NRC concern was identified regarding need for a programmatic link
between preventive maintenance and ASME,Section XI requirements.
Details are i ~luded in pargraph 3.f.
Conclusion
The Team consensus was that the Preventive Maintenance Program was a
programmatic strength.
l
l
l
l
_ . _ _ _ _ _ _ _ _ _ . _ _ _ _
_ _- - _ _ .
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31
.
k. Quality Control (QC) Program
Observations
The' Team completed an examination of the QC program; interviews with
-QC management and several QC inspectors; and reinspection of several
welds 'recently accepted by .the QC inspector involved with the weld
~
patch- problem .on the Unit 2 reactor building roof drain -(see -
- paragraph 3.e. above). Further details on examination of the. QC
program' are listed below. (Note: The Team was aware of details
associated with previously identified NRC~ violation 321,366/88-31-01
of 'a related nature and responsive licensee. correction actions.
However,Lthose corrective actions were not examined in detail since -
' full compliance is not anticipated before September 1989).
Findings
The Plant Hatch QC Program provides-the following:
-
24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> shift coverage for I&C, . Electrical, and . Mechanical
-
Maintenance
-
Inspection of safety systems and selected Balance of Plant .
'
systems
-
NDE testing and evaluation
-
Monitor of welding qualification and performance
-
Final MWO elosecut reviews
-
Material Receipt Inspection
The following controlling procedures were reviewed and found to be
'
adequate:
GEN-12750
'
.c
The ANSI N45.2.6. and SNT-TC-1A certifications of all (24) QC
inspection personnel were reviewed and found to be current. All
inspectors are certified to visually inspect welding activities.
__ ___-_- __ - _ _ . -
_ ___ _
4 4 i
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32~
a
'
Interviews were held with six QC ' personnel. Specific questions were
i posed; relative to acceptance criteria and other details of the
procedures: li sted 'above. All questions were adequately answered with
clear and. specific detail.
< :
The following welds recently accepted by an inspector were re-
. examined and-verified as acceptable.
~
MWO No'. Weld No.
1-88-07300 FW001 4
2-89-0619- FW Nos. I through 8
Conclusion
' The Team consensus after examination of the above was that Quality
Control at Plant Hatch was a strength in both program and
implementation.
.1 . Engine'ering Support
Observations
1 The Team ' interacted with several systems engineering personnel
during . examination of potential repetitive failures- of critical
- components (see paragraphs 4.t .and 4.x below)';. IAS GL. 88-14 activi--
ties-(paragraph 3.h.); corrective actions associated with HCU bolting-
(paragraph 3.b.) and unique breathing air fittings (paragraph 3.c.).
The Team 'was favorably impressed with the . capability and enthusiasm
of the engineering personnel . involved and their strong cooperation
with the maintenance organization.
The Team completed additional inspections in two areas of engineering
support (duties of t systems engineers and DCR prioritization) by
review;of controlling procedures, discussions with. management and
- engineering personnel and review of documentation.
Findings ;
The Team identified a lack of procedural definition regarding the
i. duties and responsibilities of systems engineers. Some definition is
provided by procedure '10AC-MGR-001-0S, Plant Organization Staff
Responsibilities and Authorities. However, this upper-tier procedure
does not provide specifics related to systems Engineers.
l
In some cases, the implementation of Design Change Requests (DCRs)
'
has not been timely. An example is DCR 80-440, "RCIC low speed
bypass 'line" which has been implemented on Unit 2 for several years
but is not yet implemented on Unit 1. Implementation was given a low
priority since it was considered to have little impact on reliable
^
1 - - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ - - - _ _ - _ _ _ .
y ;
-- -
,
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.l
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L
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g 33
.
. plant. operation. Cognizant licensee personnel provided details on a
recently implemented DCR prioritization rationale, the downward trend
'
data'for open DCRs (22% reduction since March 1986),.and an informal ~
'
schedule of DCRs recommended for approval in 1990.
Conclusion-
The ' Team c'oncluded that additional specifics regarding systems
p
-
engineers duties and responsibilities should be added to the
procedures involved.
Th'e Team also ' concluded that' the presently implemented DCR
prioritization rationale and schedule were sufficient to resolved any
NRC concern.
The Team consensus was that technical rapport could be improved in
both program and implementation.
F m. Review of Licensee's Service Air System (Breathing Air System)
-
Requirements'
Licensee Technical Specification 6.11 states in part that
procedures for personnel radiation protection shall be prepared
consistent with the requirements of 10 CFR 20, and shall be
approved, maintained and adhered to for all operations involving
'
personnel radiation exposure.
10_ CFR 20, Apendix A, footnote (d), requires that' respirable air
shall be provided of the quality and quantity required in
- accordance with NIOSH/MSHA certification (described. in
, 30 CFR Part 11) for atmosphere - supplying respirators.
30 CFR, Part 11, Subchapter b, subparts H and J require that
< , breathing air meet the applicable minimum grade requirements for
Type 1 gaseous air set forth in the compressed gas association
commodity specification for AIR, G-7.1 (Grade D or higher
quality).
Occupational Safety and Health Administration (OSHA) 1910.134,
" Respiration Protection" and NUREG 0041, " Manual of Respiratory
7.
Protection Against Airborne Radioactive Materials," include .
requirements to have air line couplings that are incompatible i
with outlets for other gas systems to prevent inadvertent
servicing of air line respirators with non-respirable gases or
,
e- .: a
.; 4 ; ' r
,. . _
, 1. -
L;
.,
, 34
f
"
ANSI Z-88.2-1969, Practices for Respiratory Protection,
.Section 5.3, Respirable Air and Oxygen for Self-Contained
'
- Breathing Apparatus and Hose Type Respirators, also requires
air-line couplings to be incompatible with outlets for other gas
-sytems to prevent' inadvertent servicing of-air-line respirators
with nonrespirable gases or oxygen.
Unique Fittings
'
-
-Observations
The. team walked down the system from the air -intake to selected
systems outlets, and reviewed the following: historical back-
ground ilnformation for- the. system; operations. procedures .for
annunciwor response and abnormal operating' procedure; instru-
ment 'and service air maintenance; health physics procedures
~ relating to the supplied air respiratory protection program;
.
- and calibration records for the system's temperature monitors.
. Fin'ings
d
The licensee documented on Deficiency Card 2-87-659, October 6,
1987,.that quick disconnects on the Service Water-System outlets
were identical to those used on Service Air System outlets. The
root cause for the identified deficiency was "No gMdance .on
installation of quick disconnects." The ' licensee issued
guidance on the use of quick disconnects on December 11, 1987.
The guidance reported that quick disconnects were used on only-
the Demineralized Water System (P21), Service. Water System
.(P41), and the' Service Air System (p51). The guidance did not
address the. Instrument Air System (PS2).
The licensee issued another Deficiency Card 2-88-1452 on
March 16, 1988, identifying a Service Air System fitting on. a
Demineralized Water System outlet in Unit 2, High Pressure
Coolant- Injection ' (HPCI) ' room.' -The corrective actions taken *
referenced the Significant Occurrence Report (SOR) 2-87-659-185
that was written for the previously identified October 6,1987, .j
finding which'was closed in December 1987.
During tours of licensee's facilities on March 3, 1989, the
inspectors ' determined that identical fittings were on the
instrument air and service air lines. During the tours, the
team requested a health physics technician to accompany them.
When .the health physics technician was asked which system,
instrument air (P52) or service air (P51), should be utilized
to supply breathing air, the technician was unsure and reported
that he did not know. The service air lines (breathing air
lines) were not identified as service air-breathing air outlets
as recommended in the REA HT-0718 study in 1981.
= _ _ _ - - ._ _ N
- -_ ___ . _ - _ _ _ - _ _ _ _ _ _ _ -_ _ - _
_. . ..
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'
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35
i
Conclusions
'
The team informed licensee personnel that failure to have
- '
incompatible' fittings on the. Service Air. System (breathing air
system) was a violation of licensee Technical Specification 6.11
in.that the-licensee had failed to comply with the implementing.
references specified in licensee procedure 60AC-HPX-006-0S and
'that the. failure to implement- 60AC-HPX-006-0S occurred 1as a-
result of inadequate corrective actions for -deficiencies
identified by ~ the licensee in the HT-0718 study in 1981 and
two' Deficiency cards (2-87-659 in October 1987 and 2-88-1452
in ' March 1988). The inspectors stated that failure to take
u timely and adequate corrective' actions to prevent recurrence
was a violation of. the licensee's quality assurance program
Appendix B, Criterion XVI, Failure to Complete Adequate Correc-
tive Action 321,366/89-02-02.
'-
Failure to-Sample the Plant's Breathing A. System
'
Observations .
The team determined that licensee procedure 60AC-HPX-006-05
- required the licensee to provide Grade' D air or better as
prescribed by the Compressed Gas Association. The procedure
-also -requires that the respirable air be sampled monthly for
radioactivity. .The inspectors requested a review of the Grade D
and Radioisotopic Analyses made in the last 12 months. Licensee
procedure DI-RAD-03-1087N lists 'the locations and frequencies
for each sample. The licensee samples the respirable air
systems for Grade 0 air on a quarterly basis. The team
determined that the licensee had. completed the monthly isotopic
samples for radioactivity as required. However, the licensee
could not demonstrate that the Grade D sample on the air
compressor utilized to fill Self Contained Breathing Apparatus
(SCBA) had been made during the fourth quarter of 1988.
Conclusions
,
The inspectors informed licensee representatives that failure to
take a quarterly a1r sample and analyze it for Grade D air was a
violation of Technical Specification 6.11.
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Procedures for Sampling
Observations
The Team determined that the licensee's radiation protection
y procedures did not describe how the plant breathing air was
- sampled and analyzed to versfy that the plant breathing air
lL systems meet the minimum requirements for Grade D air. When
licensee management was notified that there appeared to be a
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violation of licensee Tehenical Specifications, for failure
to have written procedures,.the inspectors were informed that a
site Quality Assurance (QA) Auditor had already identified the
procedure problem in an ongoing QA audit. The inspectors
interviewed the QA auditor and ' determined that the auditor had
begun a radiation protection program audit March 7, 1989 and had
discussed -the finding with health physics personnel. The
inspectors reviewed a Procedure / Request Development form that
had already been completed to address the deficiency. The Team
stated that a review of the licensee's corrective action.s
concerning the sampling and analyses of the plant's breathing
air system to meet Grade 0 requirements would be performed and
identified as Inspector Followup Item (IFI) 321,366/89-02-07,
Written Procedure for Sampling Breathing Air.
-
Breathing Air System Instrumentation
During the review of Service Air System High Temperature
instrumentation, the Team determined that the licensee was
verifying correct operability every five years plus or minus
five years. The inspectors discussed the calibration frequency
with licensee management and licensee representatives agreed to
increase the frequency to every 18 months.
n. Radiological Protection Program Interfaces
The Team reviewed the method and degree of interaction between the
radiation protection staff and other plant groups. In addition,
craft and operations personnel were interviewed relative to support
they received from the radiation protection staff and found that in
general, there appeared to be a good working relation between the
health physics group and other plant sections. The licensee had
established a shift coverage schedule, in which, all of the people
working rotating shifts did so together. Through interviews with
various shift personnel, the inspectors determined that most people
interviewed like the idea of working together routinely and thought
the schedule enabled the various work groups on a shift to work
together more as a team.
The Team determined that the licensee had a radiation specialist
assigned to the planning / controls section.
o. Control of Radioactive Material, Contamination, Surveys, and
Monitoring
Reviews of records and observations during plant tours revealed no
instances in which unsatisfactory controls were being exercised over
radioactive material, contamination, surveys or personnel monitoring.
The licensee had made improvements in controlling radioactive
materials and in reducing the total area contaminated.
_ _ _ - _ - _ . b
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p. Radiation-Protection Audits
e The Team discussed the audit 'and surveillance program related to
radiation protection and control of radioactive material with
licensee representatives and reviewed the following audits:
Quality Assurance Audit of Health Physics Program (88-HP-1)
L Quality Assurance Audit of Health Physics (88-HP-2)'
The audit findings identified program strengths and weaknesses.
Examples of the audit findings documenting program weaknesses
included but were not limited to:
Failure to perform adequate surveys
. Poor documentation of ALARA activities
. Inaccurate man-rem estimates
Inadequate guidance to require ALARA review of plant documents
The audits were .goed health physics appraisals, in-depth, and
appropriate in scope. The 11censee's audit program for radiation
protection activities is a program strength.
-q. As Low As Reasonably Achievable (ALARA)
10 CFR 20.1c states that persons engaged in activities under licenses
issued by the NRC should make every. reasonable effort to maintain
radiation exposures as low as reasonably achievable. The recommended
elements of an ALARA program are contained in Regulatory Guide 8.8,
Information Relevant to Ensuring that Occupational Radiation Exposure
at. Nuclear Power Stations will be ALARA, and Regulatory Guide 8.10,
Operating Philosophy for Maintaining Occupational Radiation Exposures
As documented above licensee radiation protection audits had
identified needs for improvement in tha ALARA area. Program ,
weaknesses identified included: l
Initial man-rem estimates for radiation work permits are ,
inaccurate. Errors in both the projected man-hour and dose
estimates have contributed to the problem.
Some aspects of ALARA Program are not well understood by plant
personnel.
At the time of inspection, most of the corrective action as a result
of licensee audits had not been implemented, however, the licensee
was in the process of strengthening its ALARA program. The licensee
was reviewing an ALARA training program to give plant workers
additional training that would enable the staff to better understand
methods to reduce exposures. The licensee was also requiring more
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,
involvement from section supervisors in setting ALARA goals and
guidelines were being developed to strengthen the Plant ALARA Review
Committee (PARC). Implementation of the proposed corrective actions
should strengthen the licensee's radiation protection program.
The licensee's 1988 person-rem per unit for Boiling Water Reactors
(BWRs) was 701 versus the 1988 national average of 511 person-rem per
unit. The licensee's three year average is 619 person-rem per
unit versus the national average of 551 person-rem per unit.
r. Maintenance Related Data
Observations
The Team examined the following data associated with maintenance
at Plant Hatch. Most of the data showed an improving trend in the
years up to 1987. In that year, record performance was achieved for
availability factor (over 80%), consecutive days on-line (143), ,
electrical generation (10,832 gigawatt-hours), forced outage rate l
(3.0%) and industrial. safety (10,880,000 man-hour without a lost-time
accident). In 1988, as shown in Table 1, most usta continued to
show acceptable performance and showed improvement over 1986 but in
some. areas performance was not as good as in 1987. The number of
reactor trips and ESF actuations in 1988 were above the industry
average, but within the acceptable range.
Table I, Maintenance Related Data
1988 Industry
Indicator, both units 1988 1987 1986 Average per unit
Availability Factor, % 7 81 - 52f 77
Forced Outage Rate, % 12.4 3.0 9. 3 11
Reactor Trips 10 8 11 2
ESF Actuations 6 5 4 2 ,
TS Violations 24 27' 43 i
SALP Rating, Maintenance 2 2 2
LERs .
38 27 77
NPRDS Failure Reports 693 460 173
Significant Occurrence
Reports 534 827 NA
Work Orders Backlog,12/31 1259 2400 3144
Radiation Exposure, Man-rem l
per unit 383 431 742 521 (1987)
Absenteeism, % 1.8 1.9 2.1
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Conclusion
The Team concluded that although some. ofL the historic data showed
poorer ' performance in 1988 than in 1987, the long-term trend is
improving and on balance the data indicate good performance.
s. -Root Cause Analysis-
The root cause analysis program at Hatch Nuclear Plant was evaluated
' with respect-to training, procedures, and implementation. . Interviews'-
were conducted with individuals who were involved'in tne development
of 'the program. and who are currently responsible for its
implementation. Specific areas inspected were training materials and
their application, procedures, MW0s (2-88-4850, 2-89-65, 2-88-704,
2-88-1810, 'and 2-88-1788), and significant equipment failures.
_
Additionally, one , manager who completed ~ the 40-hour root cause
analysis cour'se was interviewed and observed applying principles and
techniques covered during the. course.
Observations and Findings
During 1988, some managers and engineers received initial training to
familiarize them with concepts and methods used to conduct root cause
analyses.- The methods included MORT, event and' causal factors
analysis, fault tree analysis, change analysis, barrier analysis, and
Kepner Tregoe's . problem . analysis. The training consisted of an
eight-hour course on' root cause analysis. The course materials
included an instructor handbook (IT-IH-21100-00) and a student text
(IT-ST-21100-00).
Currently, a 40-hour course on root cause analysis is taught by EG&G
Intertech. This course was introduced December 1988. The course-
materials also include an instructor handbook and a student text
(IT-21300-01). The stated goal is to ensure that about 50
individuals (including engineers, supervisors, manager, general ~
l support personnel, security personnel, and mainten_ance personnel)
.
L
receive training on root cause analysis. The stated -expectation is
that :about. ten new individuals will recaive training on root cause
analysis each year. Each department nominates candidates for the
root cause analysis course.
l
Hatch Nuclear Plant has several procedures in place to address root
cause analysis. Procedure AG-MGR-27-0687N provides guidance for
personnel reviewing events necessitating root cause determination. -,
I The Team identified a weakness in this procedure concerning the i
lack of details on how to conduct a root cause analysis. Other
procedures relevant to root cause analysis include 10AC-MGR-004-05,
40AC-REG-002-05, and 10AC-MGR-012-OS. The first procedure assigns )
responsibility for root cause determination and provides guidance on
identifying significant deficiencies. The determination that a-
deficiency is significant necessitates a root cause analysis. The ;
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- second procedure specifies significant events or conditions that
require reporting. The. third procedure ' provides specific guidance
for addressing significant or repetitive events (or conditions),
including the need for root cause determination.
Two approaches were used to assess the actual use of root cause
analysis in maintenance efforts at Hatch Nuclear Plant. In the
first approach, several pieces of equipment that had been previously
judged asi having'significant failures were the focal point.' These
significant equipment failures were investigated to determine if root
cause analyses were completed as required. The pieces of equipment
.were: 2E41-F006 (HPCI pump discharge valve), 2E51-F008 (RCIC Steam
Isolation. Valve), 2B31-R620 (master recirculation controller),
2B21-F013H . (safety relief valve), 2B21-F022B (air valve), and
2B21-F022C (air valve). For all of the equipment' failures except
one, it was .found that. root cause analyses had been completed and
were considered adequate. However, no root cause analysis was
conducted for valve 2E51-F003.
In the second approach, an individual was identified who had not only'
completed the 40-hour root cause analysis course, but also was
attempting to determine the cause of a failed pump. This individual
described and demonstrated principles and techniques taught in the
course- that were being applied to the failed pump problem. The
. observed process was considered a'dequate and seemed to reveal some
insights on the." weight"~that should be given to vibration, oil, and
wear particle analyses.
L0ne procedural weakness was noted regarding procedure
AG-MGR-27-0687N. The procedure lacks details on how to conduct a
root cause analysis. The Team further observed that an excessive
length of time was required to determine root cause of Feedwater Pump
leakage discussed in paragraph 4.x.
Conclusions
The root cause analysis program regarding - maintenance at Hatch
Nuclear Plant was adequately documented and seemed to. be well
implemented. LHowever, weaknesses were noted as discussed above.
Overall, the program was judged satisfactory,
t. Trending
The trending program at Hatch Nuclear Plant was evaluated regarding
established procedures and program implementation.
Observations and Findings
-
Hatch Nuclear Plant has two procedures in place to address trending
in the area of maintenance. The first procedure, DI-MNT-02-1085N, is
concerned with repetitive maintenance problems (for example, repeated
failure of the same piece of equipment) and is applicable to
maintenance engineering personnel.
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\. Trends were investigated for the main steam (B21) recirc (B31), CRD
filters (C11), HPCI (E41), and RCIC (E51), regarding equipment with
equal to (or greater than) five corrective work orders for the period
January 1,1988 to December 30, 1988. The equipment considered was
as follows: 2B21-F002A (4-way air valve), 2B21-F022B (2-way air
valve), 2B21-F022C (3-way air valve), 2B21-R614 (SRV temperature
h recorder), 2831-S001A and B (Recirc M-G Sets), 2C11-R003B (CRD
Filters), 1E41-FOC2 (HPCI Steam Supply Isolation Gate Valve),
IE41-C001 (HPCI Main and Booster Pump), 2E51-F007 (RCIC Steam Supply
Isolation Valve), 2E51-F045 (RCIC Steam Turbine Valve), and 1E51-F045
(RCIC Steam Turbine Valve).
The Team found that trend data for the equipment provided useful
information and except for the CRD Filters, the data indicated that
the subject equipment failed for a different reason each time. The
systems engineer pointed out that the problem with the high CRD
filter replacement rate during March 1988 on Unit 2 was found to be ,
related to start-up from a refueling outage. CRD takes suction from
the condensate system (carbon steel pipe). After a unit has been
shutdown two to three months, corrosion builds up in the condensate
system causing CRD filters to need replacing more of ten after an
1 outage.
The second procedure, DI-REG-08-1285N, describes the trending program
for deficiency cards (DCs), significant occurrence reports (SORS), -
and licensee event reports (LERs). The trend report for DCs and SORS
-
covering the period January 1, 1988 to December 31, 1988 was
examined. The equipment included the following: IC11-R018 (CRD
temperature recorder), IN21-C007 (condensate demineralized pump),
2W24-C021 (cooling tower fan), and 2N21-C002A (condensate booster
pump). It was found that the trend report was both adequate and ,
comprehensive, including a detailed breakdown of the type of
deficiency (e.g. , personnel related).
Although not explicitly covered by procedures DI-MNT-02-1085N and -
DI-REG-08-1285N, trending of NPRDS equipment failures were also
investigated. The NPRDS equipment failure analysis report for the
period January 1987 to June 1988 was examined. In addition to review
of the NPRDS report, a summary description was reviewed of all MWDs
for all systems with NPRDS component failure from Jar.uary 1,1988, to
December 30, 1988. The failed equipment included the following:
CRD-N26-23 (control rod), B31-K634A (controller), C11-R601 (pressure
indicator), C32-R607 (flow recorder), C32-K6008 (amplifier),
B31-N014D (transmitter), E11-C001A (pump), and B21-F010A (valve).
The subject trend report was considered a definite strength to the
overall Hatch Nuclear Plant trending program because it not only
provided useful data on specific equipment that had failed, but also
provided comparisons with the industry average.
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. Trends were~. investigated for the main steam (B21). recirc (B31), CRD
filters (C11)', HPCI (E41), and RCIC (E51), regarding equipment with
' equal to (or greater than)'five corrective work orders for the period
' January 1,1988 to. December 30, 1988. The equipment. considered 'was
as - .follows: 2B21-F002A '(4-way air valve); 2B21-F0228 (2-way air
"
valve) 2B21-F022C (3-way . air valve), 2B21-R614 (SRV temperature
recorder), 2B31-S001A and B (Recirc M-G . Sets), 2C11-R003B (CRD
Filters), 1E41-FOC2 (HPCI- Steam- Supply Isolation Gate. Valve),
IE41-C001f (HPCI Main and Booster Pump),. 2E51-F007 (RCIC Steam Supply
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Isolation Valve), 2E51-F045 (RCIC Steam Turbine Valve), and IE51-F045
(RCIC Steam Turbine Valve).
The Team found . that trend data for the equipment provided useful
information and except.for the CRD Filters, the data indicated that
the . subject equipment _ failed for a different reason each time. The
systems engineer pointed out that the problem with the high CRD
-filter replacement rate during March 1988 on Unit 2 was found to be
related to start _up from a refueling outage. CRD takes suction from
~
the condensate system (carbon steel pipe). After a unit has been
shutdown two to. three months, corrosion builds up in the condensate
"
system causing CRD . filters to need replacing more often after an
outage.
The second procedure, DI-REG-08-1285N, describes the trending program
for deficiency cards (DCs), significant occurrence reports. (SORS),
and licensee event reports (LERs). The trend report for DCs and SORS
' covering the. period Janusey 1, 1988 to December. 31, 1988 was
examined. The equipment included the following: IC11-R018 (CRD
temperature recorder), IN21-C007 (condensate demineralized pump),
2W24-C021 (cooling tower fan), and 2N21-C002A (condensate booster
. pump). It was found that the trend report was both adequate and
comprehensive,- including - a detailed . breakdown of the type of.
deficiency (e.g. , personnel related).
Although not expl.icitly covered by procedures DI-MNT-02-1085N and
DI-REG-08-1285N, trending of NPRDS equipment failures were also
investigated. 'The NPRDS equipment failure analysis report for the-
period January 1987 to June 1988 was examined. In addition to review
of the NPRDS report, a summary description was reviewed of all MW0s
for all systems with NpRDS component failure-from January 1, 1988, to
December 30, 1988. The failed equipment included the following:
CRD-N26-23 (control rod), 831-K634A (controller), C11-R601 (pressure
indicator), C32-R607 (flow recorder), C32-K6008 (amplifier),
B31-N014D (transmitter), E11-C001A (pump), and B21-F010A (valve).
The subject trend report was considered a definite strength to the
overall Hatch Nuclear Plant trending program because it not only
provided useful data on specific equipment that had failed, but also
provided comparisons with the industry average.
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Interviews with maintenance management revealed that functional (or
post maintenance) test data'. are . trended to identify any adverse
trends. The team pointed out to the licensee that . trending NPRDS
data = with L respect. to failed components that can be attributed to
personnel error during previous maintenance is another trend that is-
recommended. This trend 'is readily available and could serve to
augment functional test trends.
Conclusions
The Hatch trending program in the area of maintenance was
satisfactorily documented through procedures and appeared to be well
implemented. The overall program was judged " good." This judgment
-was based on adequate. procedures that were in place and appropriate
implementation of the program. One recommendation for enhancing the
trending . program was noted. The recommendation concerned trending
NPRDS data regarding failed components that can be ' attributed to
previous maintenance. The benefits of such a trend would be
L two-fold: to augment trend data on functional tests and to identify
b
any adverse trends in this area.
u. Spare Parts
During the inspection, two MW0s were identified that required spare
parts: for final resolution. .The fi rst MWO, F.-89-00536, concerned
obtaining a motor' for the MSIV leakage control system. The second
MWO,. 1-88-8411, involved' obtaining a backup battery for the
programmable. controller in the demineralized building.
Observations and Findings
The resolution of the spare part issue for the MSIV system was
evaluated .by monitoring morning management sessions and interviewing
.the maintenance manager concerning this issue. There was some
difficulty in obtaining a spare motor because the manufacturer is no
longer in business. During the week of March 6, the motor arrived at
the plant and was installed, returning the MSIV system to operability
and thereby resolving a LCO. Currently, the motor that failed is
being refurbished and will-serve as a spare. The maintenance manager
indicated that parts that are no longer manufactured are a problem
'for Hatch Nuclear Plant and the industry at large. He also noted
that' the corporate office is supportive in resolving issues of this
kind.
The second-MWO, 1-88-8411, concerned obtaining a backup battery for
the programmable controller in the demineralized building. The
technician who replaced the battery indicated that the subject
battery was ordered and promptly received.
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Conclusions
i
The resolutions of the above described spare ' part issues were
considered good based on the continuous attention given by plant and i
corporate management to obtain the spare motor and the prompt
acquisition of the backup battery. i
v Document Control System for Maintenance I
The document control system for maintenance utilized at Hatch' Nuclear ,
Plant was evaluated with respect to these criteria: established,- 1
proceduralized, maintained, traceable, and updcted. .;
Observations and Findings
The Nuclear Plant . Management Information System' (NPMIS) has been
established, maintained, and is continuously updated, providing ' a
computer-based control system for processing MW0s. Procedure
DI-0AP-10-0588N provides guidance _for processing MW0s. Specific MW0s j
that were examined (that is, from initiation to closeout) through the 1
. NPMIS included the following: 1-88-01297 (failed SRV temperature
recorder), 2-87-03973 (replaced ASCO solenoid valve), 2-88-02235
(valve failed to close), and 2-88-02240 (valve air leak). It was
- found that MWO history and status are easily traceable through the
NPMIS. Since. the plant does not employ .a " trouble tag system",
checks were made to ensure' that the NPMIS included MW0s for failed
equipment that was observed during plant walkdowns. .With some
exceptions, it was confirmed that the NPMIS did ' include the subject
MW0s. Numerous NPMIS computer terminals were located throughout the
-plant in areas that seemed convenient for management, system
engineers, mair,tenance engineers, and support personnel.
Conclusion
The NPMIS was an effective system for not only documenting the
history and status of maintenance on equipment but also for trending
failed equipment. Based on the above findings and observations, the
system was judged " Good".
w. Control Room Annunciator Alarms
Control room annunciator alarms were evaluated for both Units 1 and 2
regarding the number of annunciators that are continuously lighted
and whether annunciators that should be cleared are being addressed
by the maintenance program.
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-0bservaticas and Findings
The control room was inspected on two different occasions. During
the inspections, it - was noted that only a few annur.ciators were
continuously lighted. Of these annunciators, the SAFETY / BLOWDOWN
VALVE LEAKING-annunciator, was investigated to determine how it. was
, being resolved. It was determined that the annunciator was in alarm
. resulting from 'a problem in the drywell. The problem was c.cheduled
to be, fixed during the ,iext forced outage. During the daily' morning
management -meetings, the number and causes of lighted annunciators
were. discussed. On ' February 28, 1989, the following annunciators
were reported: 1H11-P651, COOLING , TOWER OR~ DEEP WELL PUMP BKR
TRIPPED (several cooling tower fans tripped locally); 1H11-P700, WGT l
BLDG CHILLER B TROUBLE (blown gasket); IN62-P600, ABSORBER VESSEL
. TEMP HIGH (MWO 1-89-912); 2H11-P657 and 2H11-P654, TORUS WATER HI/ LOW l
LEVEL (due to venting with low level present); and 2H11-P700, ;
REAC/RADW' BLDG COOLING TOWER BASIN HIGH LEVEL (operations l
investigating and deficiency card written).
{
Conclusion
Based Lon the 'above findings and observations, the number of
continuously ~ lighted annunicators were judged to be few and the l
subject annunciators appeared to be adequately addressed. ;
i
x. . Condensate and Feedwater System (N21) !
Obs'ervations
The. Team inspected ~ maintenance activities on the N21 system. The .
'
inspection included examination of a Summary List of 75 MW0s related. l
to repetitive corrective maintenace and 75 MW0s related to NPRDS I
equipment failures. Each of these - MW0s was discussed with the i
cognizant system engineer and a walkdown of the system, with emphasis i
on the' items requiring repeated corrective maintenance, was conducted j
with the system engineer. '
Findings, ,
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- Repetitive Tracking System
The system for tracking repetitive equipment failures was
developed by the Maintenance Engineering Supervisor to provide
guidance for prioritizing corrective maintenance work. The
tracking system uses the NPMIS te, sort those Master Parts List
(MPL) items for which five or more MW0s for corrective
maintenance (CM) were written in 1988. This list provided a
method for team inspectors to focus inspection effort on those
MPLs with potential maintenance problems.
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For the Ccndensate and Feedwater System (N23), the following
MPL's were listed:
MPL Number Number of MW0s Description
IN21 B006a 5 5th Stage Extraction Heater A
IN21.B006B 5 5th Stage Extraction Heater B
IN21 C005A 11 Reactor Feedwater Pump and
Turbine A
IN21 C005B 10 Reactor.Feedwater Pump and
, Turbine B
IN21 P001- 17 Condensate Polisher Control
Panel
2N21'C002A 7 Condensate Booster Pump
2N21 C005A 9 Reactor Feedwater Pump and
Turbine A
'2N21 C0058- 11 Reactor Feedwater Pump and
Turbine B
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Feedwater Pump Leaks
During the system walkdown, sizable saal leaks wera noted on the
shafts of each of the four feedwater pumps. A. tray to catch
this water was installed below each pump with.a drain. tube to a
50 gallon . drum. A drain tube lead from the bottom of the drum
to the floor. drain. Drain tubes also lead from the seal weep
holes to the 50 gallon drum. Plastic funnels were installed to
catch leak water 'from flanges and fittings on pipes in the
feedwater pump rooms. Drain tubes from the plastic funnels lead
to floor drains.
The' seal leakage observed was not considered to be normal and a
consultant from the pump vendor (Byron-Jackson) was called in on
March 6, 1989, to analyze the problem. According to the vendor
representative the root cause of the problem was that the normal
seal water flow was routed back to the condenser hot well rather
than to the booster pump intake. The low pressure in the
hot well caused the seal water to flash into vapor, thereby
restricting liquid flow to the hot well. The vendor representa-
tive suggested that the excessive leakage could be decreased by
rerouting the seal water flow to the booster pump intake or by
increasing the size of the piping. Until these design changes
are made, the leakage may be decreased by careful adjustment of
the seal water controls.
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Rebuild of Condensate Pump
The Team examined documentation associated with the repair of
Unit 2 condensate pump N21-C001B. The complete work order
oackages for the first and second rebuilds of the pump were
obtained. MWO 2-88-1906 was written on April 5, 1988, when high
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vibration was noted and the pump shaft appeared to be out of
alignment. Worn bearings and wear rings on the pump shaft
were found. The pump shaft and bearings were replaced.
MWO 2-88-3177 was written on July 6, 1988, when the pump failed
again shortly after startup following the first rebuild. The
lengthy MWO packages (55 pages for the second rebaild) did not
provide a clear picture of the root cause of either failure nor
of the actual repair operations.
A further explanation was provided by the maintenance manager.
After the first failure of the pump, cracks were noted in the
section of the pump casing (52" in ' diameter and 104" long)
containing the suction and discharge flanges. This casing
section was rebuilt. The root cause of the second failure was
that the flanges for attaching the new casing section to the
motor and to the pump casing section were not properly aligned,
because instruments were not available in the maintenance shop
for accurately aligning such large-sized sections of casing.
When it was deduced by the maintenance manager that alignment I
was the problem, special equipment was designed and built for
checking the alignment of the casing flanges. The flanges were
found to be out of line. The alignment was corrected and the
pump was reassembled. It has been running without problems
since September 1988. The Team observed this pump in operation
with a maintenance engineer familiar with checking vibration and
alignment. Wire leads from the pump shaft bearing area for
attachment to a vibration measuring instrument were visible.
The pump appeared to be running smoothly.
A representative of the pump vendor was present during the first
and second rebuilds of the pump. The representative did not
recognize the alignment problem with the first rebuiid and was
surprised by the subsequent failure.
Conclusions
-
Repetitive Tracking System
The Team consensus was that the NPMIS and repetitive tracking
system is a programmatic strength (also see paragraph 4.v
above).
-
Feedwater Pump Leaks
The temporary provisions to route the seal water leaks, and
other pump room leaks, to the floor drains are unsightly and
constitute poor housekeeping practice, but do not represent
significant contamination or safety hazards. The pump room
leaks, except the seal leaks, will be corrected at the next
outage. Hatch management is moving toward a long-term solution
__
_- -_ _
,
,
a ,
'
'
.1
, 4
,
47'
r
'
,,
of the seal leak problems-along the lines suggested by the pump.
vendor representative. No = definite . schedule' for corrective
action has.yet been established. Ideally, the licensee should
have discovered the' root.cause of. the leakage and corrc:ted it'
sooner, but the delay has not interfered with system operation .
nor resulted in a' safety hazard.
'
Rebuild of Condensate Pump
The Team consensus was -:that the . condensate ~~ pump rebuild ~
.
indicated la strong plant maintenance organization able to
"
~ analyze and. correct a' subtle and complex maintenance problem.
However, the . poor- description. of the -root cause analysis and
L corrective actions in the two " rebuild" MW0s is considered to
indicate. a weakness of the licensee's record keeping in the
maintenance area.
. S '. Evaluation of Maintenance Program
Based on the inspection ' details and.. inspection results of paragraphs 3
and 4 above, the. team' evaluated .the Maintenanc_e Program using the
guidance of NRC TI 2515/97. The below paragraphs detail the evaluation.
'a. Overall. Plant Performance Related to Maintenance - Direct' Measures
Rating - Good
Findings /0bservatio'ns
Review of Direct measures revealed an improving trend of most
performance indicators up to 1987. In that year, record erformance
was. achieved for availability - factor (over 80*0. consecut:vt v ..y s
v on-line (143), electrical generation (10,832 cigWatt-hours), forced
outage rate ( 3. 0*4) and industrial safety (10,880,000 man-hours
.without a lost-time accident).
Although some of the historic data showed poorer performance in 1988'
,
than 'in _1987, the long-term trend is improving and on balance the
data indicate good performance.
The general plant walkdowns found the plant to be in relatively good
material condition and L the team consensus was that the general
. quality :of housekeeping in the plant was good. As noted in
paragraph 4.a., some deficiencies were identified. On bal&nce, the
team 'does not regard the noted deficiencies as significant and
considers the overall condition of plant and housekeeping to be good.
--_ - _ _ __-
g -_ ____ _______ -
y :- 4 :
1 .h
1
48
p
Conclusion
,
L' .The overall plant performance related to maintenance as indicated by
l
historic data and observed in plant'walkdown inspections is good.
\ b. Ma'nagement Support of Maintenance
Rating -
7
Program: Satisfactory
Implementation: Satisfactory
Scope
. Management, support of maintenance was examined by reviewing and
, evaluating- (1) management- commitment to and involvement in
maintenance;. (2) management f organization and administration; and
(3) ' technical support provided to the maintenance organization.
f (1). Management Commitment and Involve 9ent
n Rating-
Program: Good
Implementation: Good
Findings / Observations
In general, the team found, during the inspections detailed in
paragraph 4. above, that the licensee had a good program 'for
application of industry intiatives. The inspection-revealed the
following examples of good application of industry initiatives:
few control room annunciator alarms that are continuously
lighted; the- NPRDS is used for trending' and equipment failure
history; motor operated valve motor' shaft keys are being
replaced; backseating of valves is no longer done on a routine
basis (IN; 87-40); policy has been established to remove PCBs
from 4160-600V . transformers by retrofilling with non-PCB
insulation; checks for silicon bronze carriage bolts (IN 88-11)
in - equipment identified in the IN as well as other related
equipment (e.g., SKV switchgear).
The following weaknesses were identified relative to application
of industry initiatives: Information Notice 88-42, " Circuit'
Breaker Failures Due to Loose Charging Spring Mounting Bolts,
was.not incorporated into the preventive maintenance-procedure;
the duties and responsibilities of the " systems engineer" is not
well defined; and the vendor's recommendation regarding PM on
- - _ _ _ _ _ _ _ _ - - _ _
_ _ _ _ _ ____-_
'.. - o ,
r'
tj,
t )
F 49'
4KV switchgear was not .followed. ' The sytems engineer related
- and switchgear-related weaknesses are discussed .further in
'
.naragraphs 4.l. and 3.a., respectively. 1
Investigation of. management vigor and example indicated the.
'
following: management performs systematic area inspections;-
various morning meetings conducted by upper management serve to
n identify important maintenancelissues, follcwed by meetings with-
departmental managers and foremen to resolve the identified
'
. issues; training is generally excellent except that PM training:
on_4KV switchgear is not yet complete; and feedwater flow system
maintenance -replacing GEMAC transmitters . indicate that plant -
L aging is being addressed. The latter . finding is discussed ;
.further in paragraph 4.c.
.
}
Conclusion
Based, on management's commitment to the application of industry
. initiatives, as . noted ..above, and observation of man'agement's
clear and active involvement in the maintenance program both the
program and its implementation were rated " good." Weaknesses in
'this inspection area were noted regarding .IN 88-42, the vendor :
recommendation on SKV switchgear, 'and incomplete PM training on
SKV'switchgear. ;
(2) Management Organization and Administration
.i
Rating -
.
Program: Not Evaluated
!
~ Implementation: Good ;
1 a
!
Findings / Observation 1
Maintenance staffing level seemed adequate,' including the amount I
of technical- support provided; no adverse indicators of material
problems were found in the MWO review; various types of mainte-
nance activities (e.g., ISI, surveillance testing, diagnostic,
preventive, predictive, and corrective) have been implemented i
in the maintenance process; walkdown inspections are completed ;
by management (e.g., Maintenance Superintendent and Plant l
Engineering Supervisor);' daily feedback is provided through j
morning meetings and staff meetings regarding maintenance issues ]
where improvement-is needed; numerous performance measurements i
(e.g., backlogs, reworks, and deferrals) are well identified 1
'
and implemented; and plant management appeared to be involved
in and aware of decisions regarding upgrades, plant agir.g, and l
work deferment.
i
l'
_ _ _ _-__ _ _ _ _ _ _ _ _ _ _ _ ._
_-_
_ _ _ _ _ _ _ , _ _
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,.
-
[. .
50
C
(
. The- computer-based maintenance work order system was well
4 implemented and judged to be a definite strength to the. overall
maintenance program. ' Specific- strengths included (a) the
capabilitymto trend repetitive equipment failures (b) the MPL
numbering system, and (c) the availability of numerous CRT
termi nal s_. This system is discussed in further detail in
paragraph 4.v.
Relative to definition of maintenance requirements, a weakness
was identified in that the . licensee was not following the vendor
technical manual regarding PM on 4KV switchgear (see paragraph-
3.a). . Additional . weaknesses were identified relative .to root
.cause analyses. The first concerned a feedwater. pump seal leak
and- the length of time required to- determine the cause of the
!
excessive' leakage (See paragraph 4.x.). The second involved a
MOV ; motor failure where root cause was not anlayzed (MWO
2-88-02612). Root cause analysis is discussed further in
paragraph 4.s.
Conclusion-
Based on the above inspection findings, management organization
and administration was' rated " Good." Some needed improvements
-
in this inspection area were noted with respect to following
,
"'
. vendor technical manuals and conducting root cause analyses.
The program for this inspection area. was not reviewed in
. sufficient detail and, therefore,'it was rated "Not Evaluated."
(3) Technical Supp' ort
Rating -
Program: . Satisfactory
Implementation: Satisfactory
q
Firidings/0 observations i
Formal and informal communication between technical support 'and
other organizations were not examined in detail. However,
indications are that communication is strong. Maintenance
information is communicated in the daily 7:30 AM meeting to 1
discuss the five principal operations to be performed that day. 1
A . weekly meeting projects maintenance work to be done in the )
next two weeks. These meetings are attended by about 30 :
maintenance, support and supervisory personnel including the
plant manager.
Good communication between maintensoce craftsmen, their foreman
and other support personnel was observed during the performance
of several maintenance jobs,
i
- _ _ _ _ _ _ _ - - - _ _
- ..
- .:,2
x
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$Y j
Examination l of. Engineering Support . revealed that: the - major
engineering support for maintenance at Hatch .comes from the
4 maintenance engineering group reporting to the maintenance
-
manager ' for engineering support. The maintenance engineering
.
' '
group resolves routine engineering problems related to
maintenance, monitors preventive and predictive maintenance, and
</ performs trending analyses on repetitive ~ failures and on the
ratio :of preventive. to total maintenance. Less routine-
maintenance problems are sent to the system engineers for
analysis and resolution.
L
\ Maintenance engineers appear to be competent and enthusiastic in
completing their assigned. tasks. The Team also interacted with
system engineers during this inspection. It was clear that the
system engineers were thoroughly familiar with the maintenance
problems of their assigned systems and took active part in their
resolution. However, a need for better definition of systems
engineers duties and responsibilities was identified.
Examination of: QC revealed that: criteria for inspection and
aucit are Established and implemented, inspection / verification
is scheduled and accomplished, and corrective actions are taken
as necessary. The Team noted that plant QC inspectors were
present. during the performance of maintenance. and that MW0s
specified holdpoints for QC checks. In one instance, the QC
inspector miscalled the acceptability of a faulty weld patch
operation (see pragraphs 3.e. and 4.k.). Overall, this element
is rated good in program and implementation.
Radiological controls were examined and found satisfactory with
exception of procedures .related to breathing air sampling (see
paragraphs 3.d. and 4.m.)
Examination of maintenance safety revealed that: some
electrical procedures lacked sufficient safety instructions,
although the electricians actually observed good safety
precautions in performing their work. However, . a violation
involving breathing air was noted by the Team. Safety codes
repire the use of unique fittings on breathing air lines to
preclude non-respirable gas. The same type of fitting was
observed on instrument air lines subject to nitrogen use.
No deficiencies were observed involving hazardous materials,
fire protection or confined spaces.
The integration of regulatory documents was examined and two
related violations noted. First, loose bolts were observed on
CRD Hydraulic Control Units (related to regulatory documents
IN 87-56 and violation 321/86-20-02) (sce paragraph 3.b.).
Second, an inadequate procedure for insuring incorporation of
_ _ _ - _ _
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t
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52
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f
..
r ' vendor information into. procedures was identified (related to
regulatory document Generic Letter 83-28) (see paragraph 3.a).
Conclusion
Based on the inspections and findings summarized above, the-
consensus of-the team for the Technical Support element is that
both the program and its implementa'. ion are satisfactory.
c. Maintenance Implementation
! ' Rating -
Program: Good
l Implementation: Good
Scope
The purpose of .this part ' of the . inspection was to determine the
,
quality. of the ' established controls and, more importantly, the
implementation of these~ controls. The controls established in four
areas were evaluated. These areas are (1)' Work Control, (2) Plant
Maintenance Organization, .(3) Maintenance Facilities Equipment and
-Materials Controls, and (4) Personnel Control. The effectiveness was
determined.through a review of completed work orders, procedures, and
other documentation associated with maintenance and training of
. maintenance personnel; physical observation of work in progress and
tools in stock; and discussions with all levels of personnel.
(1) ! Work Control
Rating -
Program: Good
Implementation: Good
Findings /0 observations
Review of work in progress in the field indicated that
appropriate authorizations were received; proper documentation
was issued; foremen observe the work in progress; personnel
appear competent- and properly qualified; procedures were
followed; and no major problems were identified during the
observation of work. However, a concern was identified
regarding lack of precise definition / details on MW0s 2-88-4862,
2-88-1906 and 2-88-3177.
.- - _ - - - _ _ _ _ - _ - - -
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,
This 'was aggravated by an . error regarding the type. of post-
+ maintenance testing required after revision of MW01-89-00722
on Estation service battery ' charge 10. These weaknesses are
further discussed in. paragraphs 4.e.(7), 4.x. and 4.e.(1).,
9 respectively.
e
I, ' Examination ~of the work order control system revealed a' program
in- place to . identify discrepancies - the Nuclear plant
Management Information System -(NPMIS). NPMIS is routinely
g updated by planner / schedulers during the MWO review and approval
cycle (see procedures DI-0AP-10-0588N and. 50AC-MNT-001-OS) and
'is an excellent tool for the analysis / trending / history of
maintenance activities. The NPMIS is further . discussed in
- paragraph 4.v.
Review of equipment and maintenance history. records indicated:
maintenance - history is easily retrievable through NPMIS; work'
history is updated at the completica (closure) of the MWO, the
Master Parts List'(MPL)/ Equipment' Locator Index (ELI) is being
. updated and expanded; repair. time is tracked for each MWO; root
cause analysis could .be improved .(see paragraph 4.s); NPRDS is
used but not to the extent to yield maximum benefit - (see
paragraph 4.t.) and the MWO data form includes an input- for
NPRDS. .On. balance, the Team considered the equipment and
maintenance history records program to be a strength.
An ins'pection of the conduct of job planning revealed: the
safaty significance of an item to be repaired / replaced is the
first consideration; LCO items are' worked until completion;
drawings / technical manuals / procedures are included on the MWO;
the planner coordinates work between disciplines on a MWO; spare
parts are identified on - the MWO, when possible; personnel
requirements / qualifications are well documented and are known
or easily available to the foreman assigning work; and program /
procedures promote coordination and teamwork with ' system
Engineering / Technical Support.
Examination of the licensee's work prioritization controls
revealed that safety significance and the effect on safety by
BOP is considered and no safety significant items were found
that were not included in the work schedule.
By review of the licensee's maintenance work scheduling, it was
determined that: the maintenance backlog is being trended and
this- backlog appears to be decreasing; personnel are organized
in teams and rotated between day and night shifts so that new
-personnel are evenly dispersed and compensation occurs for work
loads in the various areas; the planning and control group of
the outage and planning department determines the schedule for
maintenance (except emergency maintenance) to reduce conflicts
and the NPMIS provides for MWO tracking.
= _
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.
54
The licenseei s establishment of backlog controls was reviewed
s and it was determined that PM maintenance activities are
sometimes deferred based on " sufficient written technical basis
~ for the deferral "(see procedure 50AC-MNT-007-05). The Team
.
examined several of these deferred PMs and- noted that the
majority were assocated with complex and expensive PMs on
. rotating equipment (examples: RHR pump motors IA and ID and
-. makeup wate pump 1A) which were authorized after an engineering
analysis of other predictive maintenance. data (i.e. vibration,
lubrication,etc.). The Team _ concluded that deferred PM's were
.
not adversely affecting backlog controls. Backlog controls
f' appear - to adequately acknowledge work significance, ' BOP
concerns, estimated manhours, and contribution to ALARA.
Backlogs are measured and trended, considered to be below
' '
industry average, - and receive adequate attention from plant
management. The Team consensus was that back'og' controls were a
j. programmatic strength.
An examination of maintenance procedures . revealed that
procedures 'are generally well conceived, thorough, technically
adequate and easy to use. However, some procedural problems
1 were noted regarding adequate tie with Section XI requirements, .
conduct of root cause analysis, and 4160 Volt switchgear pMs.
These are further discussed in paragraphs 3.f. , 4.s. , and 3.a. ,
respectively.
The Team consensus regarding maintenance procedures was that
some improvements were necessary in this area.
An examination of post-maintenance testing revealed that
post-maintenance testing criteria- have been established,
documented and implemented. However, the Functional Tests
recommended . by procedure 951T-0TM-001-0S are directed toward
component post-maintenance testing and do not necessarily assure
operational readiness. The operations supervisor on shift
(OSOS) and shift supervisor (SS) review the MWO-and may accept
the FT as a satisfactory method of proving operability or may
require additional operability testing. The Team did not
consider the- above aspects of the post-maintenance testing
program to be of concern. However, the Team observed at least
one instance of incorrect functional testing after revision of
MWO 1-89-00-00722 (see paragraph 4.e.(1)) and identified the
need for programmatic assurance in procedure 50AC-MNT-001-0S
that post-maintenance tests as required by ASME,Section XI, are
correctly imposed. These items are further discussed in
paragraphs 4.e.(1) and 3.f, respectively. The Team consensus
regarding post-maintenance testing was that some improvements
were necessary in this area.
f
. - - . - . - _ . - . . _ - - . - - . . - - - _. . _ , - - -
_ _
_. __ .
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'55
l
J
"
The Team reviewed a sample of completed work control documenta-
-J
tion as. listed in Appendix B. This examination . established -
that a document review methodology is~ implemented and performed- i
in . a. timely manner. No' major anomolies/ discrepancies were
identified.
Conclusion
l'
Based on the inspection above, the consensus of the team, for
this element was'that the program and implementation were rated
good.
"(2) , Plant Maintenance Organization
Rating -
Program: Good
,
Implementation: Good
Findings / Observations
The review of the control' of. - Mechanical , Electrical, and
Instrumentation ~and ^ Control maintenance activities revealed
that a methodology has been established and implemented . to
identify the need for . maintenance and control. rework, vendor
technical manuals, procedures, materials, tools and personnel.
Items- such as assurance of system integrity, monitoring, use of
qualified parts, and personnel accountability are included. The-
Team noted general ' positive conditions such .as: small work
backlogs; PMs according to schedule; improving trends-on rework
. items; use of appropriate procedures; acceptable to good
equipment condition; and enthusiastic and well trained
personnel. Additional specific areas examined are as'follows:
-
Mechanical - work histories for several major - components
-(e.g. Unit 2 A and- B Recire M-G Sets, Unit 1 HPCI Main Pump
and Steam Supply Isolation Gate Valve, Units 1 and 2 RCIC
Steam to Turbine Valves -etc. (See paragraph 4.t.)) were
reviewed in detail. No major discrepancies were identified
with one ' exception - (failure to conduct a root cause
-
analysis after motor failure of RCIC Steam Supply Isolation
Valve MOV 2E51-F008). The Team also noted that an undue
length of time was required to reach an accurate root cause
analysis of Feedwater Pump excessive seal leakage. No
major repetitive failures were identified with exception of
multiple failures of CRD filters. These were adequately
explained as futher discussed in paragraph 4.t.
. _ _ _ _ _ _ _ _ _ _ _ . _
_ . _ _ . _. _-_
tg j
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56
L {
L
z
,
-- ' Electrical - Review of work . history reports (4KV) and
.
special . historic . trend reports - indicated .that repetitive
' failures were not a . problem. :Further, the licensee trends
.
the condition of station batteries' (not a Technical
Specification requirement). However, the Team identified
weaknesses in that:
h
testing. of' the overcurrent- characteristic of
molded-case circuit breakers is not done (except for
penetration circuits)
preventive maintenance on 4KV switchgear is not
adequate since all vendor recommendations are not
incorporated into PM procedures. (See paragraph 4.a.
for details)
-
Instrumentation and Control - The Team noted a weakness -in
the: control of calibration of measuring and test equipment.
The control.is accomplished by a manual, monthly review of'
equipment. calibration cards. The computerized I&C
Automated Tracking System for Measuring and_ Test Equipment
was saidsto be deficient in that it counted 30 days only in
each month and caused calibration dates to' vary by three to
four days. This program should be. corrected and the card
system used as a backup. However, no s' pecific problems
were identified associated with the manual system.
The licensee's deficiency identification and control system
was reviewed. During plant walkdown inspections, only m'nor
deficiencies were found (see paragraph 4.a.) .which were not
previously identified in an MW0. The deficiency identification
and control program was considered to be a strength.
The Licensee's performance trending was examined and the
determinations were: (1) root cause analysis is adequate but
should be improved; (2) performance indicators are trended and
the majority were found 'to be better' than industry standards.
These were in overtime work, percentage of non-outage MW0s
greater than three months old; the ratio of highest priority
non-outage corrective MW0s to total non-outage MW0s and overdue
PMs.
Other trend information examined included: deficiency cards,
LERs, SORS and NPRDS failed components. The Team identified no
discrepancies in the above but does recommend the augmentation of
the presently conducted functional test trending with
information from NPRDS (See paragraph 4.t.).
The Team consensus was that performance o? maintenance trending
could be improved, especially with regard to root cause
analysis.
- _ _ _ _ _ _ _ - -
m; '~..
-
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,
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57'
,
..
The support. interfaces -were reviewed and. noted to be strong.
.o _ The Team noted good-cooperation between maintenance and other
- organizations. Daily and 'long term. planning meetings appeared
, ,
to be a. strong point. However, the Team identified a lack of
procedural definition regarding the duties'and responsibilities
.of systems engineers. Procedure: 10AC-MGR-001-0S provides some
'
e definition but is 'not specific to systems engineers. The Team
consensus was that this level of deff nition should be added to
the program.
Conclusions
' Based on the inspection above,. the consensus of the team for
- this element' was. that. the program was . rated good and the
implementation was rated good.
.(3) Maintenance Facilities, Equipment and Materials Control
Rating -
Program: Good
Imp' lamentation: Good
' Findings / Observations
The Team found the following: maintenance facilities were
' located :as _ efficient 1y' 'as possible; maintenance supervisors'
offices were located close to the shops; maintenance shops
appeared to have most tools required for the work performed; and
parts and tools storage and: requisitions was well. organized and
_
calibration activities efficiently completed-(with exception of
the weakness.in calibration activities as discussed in paragraph
5.c.(2) . above); staging and laydown areas were adequate (with
exception of one area in the Un.it I turbine building); rigging
and scaffolding were' adequate; training and mockup facilities
.
appeared adequate; and the " Hot" and " clean" machine shops are
considered programmatic strengths. These are discussed further
in paragraph 4.g.
Conclusions
'
The Team consensus was that the program and implementation for
this area was rated good.
'
1
(5
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9 .f .
58
i
, .(4) Personnel Control
Rating -
.
Program: Good
. t. -
Implementation: Good
Findings /0 observations
Observations of' in process work and discussions. with craft
personnel indicated that the electrical, I&C, and mechanical
journeymen were knowledgeable and well trained for their jobs.
, TrainingL is discussed futher :in. paragraph 4.h. Staffing for
craftLpersonnel was considered adequate. Overtime work was
maintained within reasonable limits. 'The morale and atmosphere
of; teamwork displayed by craft personnel was considered above
,
average and is . reflected in low turnover. Maintenance
management .is ' considered qualified, . enthusiastic and
instrumental in' maintaining the. teamwork displayed by. craft.
Discussions with supervisors indicated that the maintenance
training program was in accordance with INPO requirements.
- Craftsmen were not' " grandfathered",. but interim qualifications
were maintained until ' formal training could be scheduled.
Craftsmen . unable to satisfactorily complete' the formal
qualification requirements are considered not qualified for the..
- area of concern.
However, the training department did not provide any training
for performing preventive maintenance of 4KV switchgear. At the
time of this inspection, a lesson plan for this was being
developed as part of the phase V of INPO. training program. This
omission was considered to be a fault, but training overall was
considered a programmatic strength.
System Engineers have 13-week Engineer-in-Training Systems
Course with four days per year followup. This was considered to
be adequate. However, as previously discussed, there is need
for improved definition of the duties r.nd responsibilities of
sytems engineers and additional training may be required.
,
Conclusion
Based on the inspection above, the team rated personnel control
" good" in both program and implementation.
_ _______
z - -- - - - - - -
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. .
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59:
F
u L6. Exit' Interview
The . inspection scope and results were summarized on April 4,1989, with'. .
b those persons indicated in. paragraph 1. The team leader described the-
.
'
areas inspected and discussed in detail the - inspection results listed
E , below.; Proprietary information is not ' contained in this report.
Dissenting comments were not received from the licensee.
(0 pen)' Violation. 321,366/89-02-01,. Inadequate Administrative
, Procedure - Paragraph 3.a.
p,
L (0 pen) Violation 321,366/89-02-02, Failure to Complete Adequate-
L ' Corrective Action Paragraphs 3.b. and 3.c.
p
.(0 pen) Violation 321,3'66/89-02-03, Failure to Take Breathing Air
- Samples - Paragraph 3.d.
(Closed)' . Violation " 366/89-02-08, Failure to Follow Acceptance
Criteria for Weld Patch on Reactur Building Roof Drain -
, , : Paragraph 3.e.
<
(0 pen) :IFI 321,366/89-02-04,- Programmatic Link' Between Maintenance
Procedures and ASME Section XI Requirements - Paragraph 3.f.
(0 pen) IFI 321,366/89-02-05, Inspection of RHR Hanger Weld Removal -
Paragraph 3.g.
'(0 pen) IFI 321,366/89-02-06, Design Verification .of Containment
Isolation Valves T48-F310 and T48-F311 - Paragraph 3.h.
(0 pen) IFI 321,366/89-02-07, Written' Procedure for Sampling Breathing
,
Air - Paragraph 3.1.
1
1
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pm j.s
,
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4
7. Acronyms and Initialisms
ALARA- -
As Low As Reasonably Achievable
ANI -
Authorized Nuclear Inspector
-ANII_ -
Authorized Nuclear Inservice Inspector
ASME -
American Society of Mechanical Engineers
BOP -
Balance of Plant
B&PV -
Boiler and Pressure Vessel Code
BWR -- Boiling Water. Reactor
CRD -
-Control Rod Drive
CM -
Corrective Maintenance
DC -
Deficiency Card
DCR -
Design Change Request
ELI -
Equipment Locator Index
FT -
Functional Test
General Electric
'
-
GL. --
Generic-Letter
HCU -
Hydraulic Control Units
HP ,- Health Physics
HPCI -
High Pressure Coolant Injection-
IAS -
Instrument-Air System
I&C. --
Instrumentation and Control
IFI -
. Inspector Followup Item
IN -
NRC Information Notice
INPO -
Institute of Nuclear Power Operations
ISI -
Inservice Inspection
- LCO -
Limiting Condition for Operations
Licensee Event Report
~
LER -
LLRT -
Local Leak Rate Test
LPCI -
Low Pressure Coolant Injection
MCC -
Motor Control Center
M-G -
Motor Generator
,
MPL- -
Master Parts Lis'
MSIV -
Main Steam Isolat'on Valve
.MT -
. Magnetic Particle Test
MWO- -
Maintenance Work Order
NDE. -
NPMIS -
Nuclear Plant Management Information System
NPRDS -
Nuclear Plant Reliability Data System
-050S -
Operations Supervisor on Shift
PARC -
Plant ALARA Review Committee
PM -
. Preventive Maintenance
PPM -
Parts Per Million
PRA -
-PSIG -
Pound Per Square Inch Gage
PT -
'
Liquid Penetrant Test
PT -
Potential Transformer
-
Quality Assurance
-
Quality Control
_ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . -_---- -_-_ _ _______-__ _ _ _ - -
. _ _ - _ _ - -
s
...
61
.R&R -
Repair and Rariac'.inent
'RCIC -
' Reactor Cm e Isolat tun Cooling
RHR -
.Re. ;oual Heat Removal
SALP -
Systematic Assessment of Licensee Performance
SCBA -
Self Contained Breathing Apparatus
'
SER -
Safety Evaluation Report
SIL --
' Service Information Letter
SOR -
Significant Occurrence Report
SOER -
Significant Operating Experience Report
SS - '
Shift Supervisor
SSAC -
Station Service Air Compressor
TI -
Temporary Instruction
UVTA -
Undervoltage Trip Attachment
W -
Westinghouse Electric Corporation
- _ _ _ - - - _ _ _ - _ _ _ _
__-
,; .-.,
y
3 >
APPE'NDIX A _
,
LIST OF LICENSEE PROCEDURES REFERENCED / REVIEWED
PROCEDURE NUMBER TITLE
52PM-X43-006-1S, Rev.'0 Electric Fire Pump Cleaning and Inspection
52GM-MEL-022-0S Motor Shaft. Pinion Key Replacement
51GM-MME-0020 Rabuild of Waste Collector Pump
b 51-GM-MNT-0020
51GM-MNT-002-0S Maintenance Housekeeping and Tool Control
l- .
52PM-R22-01-0S, Rev. 3. _4160 Volt AC Switchgear and Associated Electrical
Components Preventive Maintenance
52PM-X43-006-IS. Electric Fire Pump Cleaning Inspection
- 345V-E11-001- RHR Pump Operability,
- AG-MGR-27-0687N Root Cause Determination-
10AC-MGR-004-OS Deficiency Control System-
40AC-REG-002-0S Federal and State Requirements
10AC-MGR-012-0S' Plant Event Analysis and Resolution Program
- DI-REG-08-1285N DC, SOR and LER Trending Program
DI-MNT-02-1085N, Rev. 3 Maintenance. History and Trending Program
52CM-MME-001-05, Rev. 2 Repacking Valves and the Adjustment of Valve
Packing
52CM-MME-005-OS, Rev. 1 .Limitorque Valve Operator Models MB-0 through
SMB-4~
Mechanical Maintenance
- 52CM-MME-011-05, Rev. 2 Gate and Globe Valve Repair
52CM-MME-013-0S, Rev. O Purge and Vent Valve T-Ring Replacement
51GM-MNT-002-OS, Rev. 3 Maintenance Housekeeping and Tool Control
42SV-SUV-004-2S, Rev. 1 Safety Relief Valve ISI Test
10AC-MGR-003-0S, Rev. 9 Preparation and Control of Procedures
I
l
'
m_'_m___ l_ _ _ _ _ _
- -
- - - - - - - - - - - - - - -
7
,,
.,
'
a t
, Appendix'A 2 l
..
- 10AC-MGR-001-OS, Rev. 4 Plant Organization, Staff Responsibilities - and
Authorities
. AG-ENG-03-1185N, Rev. 1 GE Service Information Letters (SILs) and Rapid
Information ' Communication Service Information
Letters (RICSILS) Review and Tracking
'42EN-2NG-014-05, Rev. 1 ASME Section XI Repair Replacement Program
a 50AC-MNT-001-OS, Rev. 8 Maintenance Program
01-0AP-10-0588N, Rev. 0" Planning and Control Maintenance Work Order
Processing
' DI-MNT-10-0287N, Rev. O Int ,im Qualification Job Assignment
53PM-MON-001-0S', Rev. O Vibration Monitoring of Rotating Machinery
L 53PM-MON-002-05, Rev. O Lubrication Analysis
.50AC-MNT-007-OS, Rev.:1' = Preventive Maintenance Program
20AC-ADM-003-05, Rev. 2 Vendor Manual Control
01-TRN-29-0286N,-Rev. 0 Vendor Provided Training
26MC-MTL-003-0S, Rev. O Vendor Manual Review
295IT-0TM-001-OS, Rev. O .
Maintenance Work Order Functional Test Guideline
L AG-ENG-01-0786N, Rev. O Control for Technical Information Letters (TIL's)
'20AC-MTL-001-OS, Rev. O Procurement of Materials and Services
GEN-12750, Rev. 6 Qualification and Testing of nondestructive
Testing (NOT) Personnel
40AC-QCX-001-0S, Rev. 3 Quality Control Inspection Program
45QC-INS-004-OS,:Rev.1 Visual Examination Procedure, Piping and
Component
45QC-INS-005-0S, Rev. 1 Visual Examination Procedure for Structural Steel i
45QC-INS-006-OS, Rev. O Liquid Penetrant Examination Procedure
45QC-INS-008-OS, Rev. O Magnetic Particle Inspection
-45QC-QCX-002-05, Rev. 2 Quality Control Inspection Plans
!
i
1
- - _ - _ _ _ _ _ - _ _ __ i
_ _ - _ _ _ _ _
'
..
i ..
t
j Appendix A 3
,
450C-QCX-009-OS, Rev. O, Quality Control Document Review and Hold Point
Assignment
45QC-PQL-001-00S, Rev. 3 Qualification Of Inspection Personnel
A-MB-01, Rev. 1 Weld Inspection of B31.1 Component
QA-05-17, Rev. 4 QA Surveillance
31GO-OPS-006-05, Rev 1 Limiting Conditions For Operations (LCO)
42EN-ENG-010-OS, Rev. 2 Requisition Review for Quality Requirements
40AC-ENG-011-OS, Rev. 2 Environmental Qualification Program
55MC-PRO-001-OS, Rev. 2 Procurement Document Processing
26MC-MTL-001-0S, Rev. 2 Materials Receiving
'45QC-QCX-001-05, Rev. 2 Materials Receipt Inspection
50AC-MTL-002-0S, Rev. 2 Identification and Control of Material and
Equipment
55MC-MTL-003-05, Rev. 2 Material Identification and Issue Control
50AC-MTL-003-0S, Rev. 2 Warehouse Preservations, Handling, Shipping
Storage of Materials Equipment
26MC-MTL-002-05, Rev. O Preservation,. Storage and Handling of Material &
Equipment
42EN-ENG-009-05, Rev. 4 Equivalency Determination of Replacement Parts or
Materials
51GM-MNT-002-05, Rev. 3 Maintenance Housekeeping and Tool Control,
11/9/87
34AR-654-051-1, Rev. O Annunciator Response Procedure - Control Building
Service Air Trouble, 10/01/85
10AC-MGR-003-0S, Rev. 9 Preparation and Control of Procedures, 10/20/88
62RP-RAD-003-05, Rev. 1 Use and Care of Respirators
60AC-HPX-006-05, Rev. 3 Respiratory Protection Program
PROCEDURE NUMBER TITLE
62EV-SAM-005-05, Rev. 2 Monitoring Program for Detection of Releases Via
Unplanned Routes, 10/19/88
- - _ - -__ - - ____-_-. - i
._ . _ _ _ _ - _ . _ _ _ _ _ _ _ _ _ _ _-_ _ _ - _ _ - _ _ _ _ _ _ _ _ - _
_ _ _ _ _ - _ _ - _ ,
o
4
Appendix A 4
l
DI-RAD-03-1087N, Rev. I Survey / Inspection Frequency and Work Schedule,
8/8/88
34AR-700-040-2, Rev. O Annunciator Response Procedure, 8/23/85
34AB-0PS-020-2S, Rev. 4 Loss of Instrument and Service Air System,
12/12/88
- 34AB-0PS-020-IS, Rev 4 Loss of Instrument and Service Air System,
12/13/88
52PM-P51-001-1S, Rev. 1 Instrument and Service Air Maintenance, 2/1/89
i
I
i
,
l
l
.
.
I
- _ - _ _ _ _ _ _ - . - .
- _ _ _ _ .
.
APPENDIX B
List of Work Orders Reviewed
Work Order 'iumber Description
l
2-88-4802 Cot 11r.g Fan 2A2 Motor Changeout MPL #2W24-C002
2-88-4922 Inspect Brushes on Recirc Pump Motor Generator Sets
MPL # 2B31-S001A/B
2-88-1788 Valve Air Leak
2-88-01810 Valve Air Leak
1-88-08388 PM on Electric Fire Pump IX43-C001
2-89-00536 Repair MSIV
1-88-8411 Replace Battery for Demineralized Programmable
Controller
1-89-00227 Annunciator in Alarm For Safety / Blowdown Valve Leak
1-87-02710 Valve 2B21 - F019 Failed to Close
2-87-02714 Switches not Sensing Vacuum
2-88-01788 Valve Air Leak
2-88-01787 Turbine Building Vent Supply From Low Flow Alarm
2-88-02239 Valve Air Leak
2-87-04312 Water in Switch Internals
1-88-01286 Indicating Switch Failed to Respond
1-88-08388 Preventive Maintenance on Electric Fire
Protective Pump
1-89-308 Motor Shaft Pinion Key Replacement
1-89-309 Motor Shaft Pinion Key Replacement
2-89-400 Rebuild of Waste Collector Pump
2-88-1906 Rebuild Condensate Pump
2-88-1907 Rebuild Condensate Pump !
- - - - l
_ - _ - _ - . _ _
"
1
.] '
,
..,
n .
Appendix B 2'
1
4 -
' Work Order Number : Description
2-88-3177 Rebuild. Condensate Pump
2-89-540 Correct Seal Water Leak on Feedwater Pump
1-89-821. . Correct Seal Water Leak on Feedwater Pump
2-88-00460- Complete Rework on RCIC
steam supply Isolation valve 2E51-F007
2-88-01260 Changed torque switch settings on 2E51-F007.'
.2-88-01325 Changed Stem on Valve 2E51-F007
1-88-07113 . Change torque switch settings to repair cracked
yoke on HPCI Isolation Valve IE41-F002
2-88-04575_. Repair Coupling on Recirc M-G Set 2B31-S001A'
- 2-88-04345- ~ Repair oil mist eliminator.on Recirc M-G Set
,
2831-S001A
2-88-02384- Correct gage labeling on lube oil pump: headers on
Recirc M-G Set 2831-S001A-_
28-88-02226' Correct erratic tachomeier. on Recirc M-G Set :
2831-S001A l
'2-88-02156 Repair pump outboard oil seal leak on Recirc M-G Set
2831-S001A
2-88 102155
~
Repair oil leaks from fluid drive filters.on Recirc
M-G Set 2B31-S001A-
2-88-02151 Repair check valve bonnet leak on Recir M-G Set
2831-S001A'
-2-88-02149 Repair fluid drive oil leak from 2" flange on Recirc
M-G Set 2B31-S001A
2-88-02132 Repair oil leaks from bearing plugs and sightglasses
on Recirc M-G 2B31-S001A
2-88-02067 Repair Binding scoop tube actuator on Recirc.M-G
Set 2B31-S001A
2-88-02060 Correct lack of pump speed runback after trip on
Recirc M-G Set 2831-S001A
Work-Order Number Description
_ _ _ _ _ _ - _ _ _ - _
g . _ _ _ __ .______ _ _
<,
. d
7 Appendix B 3
L2 -88-02152 Repair' bonnet and packing leaks on valves F154A and
F157A on Recirc M-G Set 2B31-S001A
2-88-04845 Repair lack of scoop tube positioner reset on Recirc
M-G set 2831-S001B
2-88-03468- Repair field breaker on Recirc M-G Set 2B31-S001B
2-88-02688 Repair tachometer on Recirc M-G Set 2831-S001B
2-88-02384 Repair lube oil suction header gages on Recirc M-G
Set 2B31-S001B
2-88-02159 Repair. fluid drive flange leak downstream of valve
F106B on Recirc M-G Set 2831-S001B
2-88-02157- ' Correct fluid drive sightglass leaks on Recirc M-G
Set 2831-S0018-
L2-88-02154 Repair pump outboard oil seal leak on Recir M-G Set
2831-S0018-
2-88-01342- Repair fan in M-G Set room of Recirc M-G Set
.2-88-02069 '* Correct high DP on CRD Drive water filter 2C11-D003B
2-88-01694' * Correct high DP on CRD Drive water filter 2C11-D003B
2-88-01606 * Correct high DP on CRD Drive water filter 2011-D003B
2-88-01592- * Correct high DP on CRD Drive water filter 2C11-D003B
2-88-01439 * Correct high DP on CRD Drive water filter 2C11-D003B
'2-88-01162 * Correct high DP on CRD Drive water filter 2C11-D003B
..
2-88-00522 * Correct high DP on CRD Drive water filter 2C11-D003B
- Note: These were clearly identified as repetitive failures and were due to
initial poor water quality during start-up from refueling outage.
(
'l-88-67475 Repair bent coupling guard on HPCI pump IE41-C001
'
1-88-07248 Repair governor valve control circuitry for HPCI pump
.1-88-05104- HPCI pump IE41-C001 tripped on turbine exhaust
l pressure - Repair as necertary
Work Order Number Description
w____________ - _ _ _ -
__ __ _ _ - _ _ _ - . - - _ _ -
t
(4 ; , .,
Q lll '
Appendix B 4
'
- 1-88-02513 Repair leak in seal water -line' on suction side of
1-88-01733: Repair. gear box oil' leak on HPCI pump IE41-C001
"
1-88-01237- Replace missing alignment bolt on gearbox for HPCI
, pump 1E41-C001
2-88-00460 ~ Repair seat leakage on RCIC steam isolation valve
2-88-00775 Reset Limitorque motor operator using MAC tester on
'
RCIC steam isolation valve 2ESI-F007
'
'
.
'2-88-00760' Resplice-cable for RCIC' steam isolation valve-
i 2E51-F007 at Penet' ration 2T52-X1050
2-88-01260 . Increase torque switch setting for RCIC steam
isolation valve 2E51-F007
2-88-01277 . Rotate Limitorque operator on RCIC steam isolation
valve 2E51-F007
~ 2-88-01325' ' Troubleshoot and repair RCIC' steam isolation
valve 2E51-F007
2-88-02575: Repair packing leak on RCIC ' steam isolation' valve
2-88-02612 Replaced damanged motor on RCIC steam isolation valve
'
2E51-F008 (Note: motor meggered ok - root cause of
failure not completed)
2-88-01262 Repair damaged motor lead on RCIC steam isolation
valve 2E51-F008-
2-88-01240 Repair packing leak on RCIC steam isolation valve
2-88-01213- Increase torque switch setting and re MAC test RCIC
steam isolation valve 2E51-F008
..
2-68-00625 Repair broken flex cable EEA90738 on RCIC steam
isolation valve 2E51-F008
2-88-00605' Replace brittle control leads on RCIC steam isolation
valve 2E51-F008
.
Work Order Number Description
I
-
_ _ _ _ _
gr = . .. ,
,
p ' . x 1 .: e
4 , ,
-
'
'
Appendir. B 5.
t
i
' L2-88-02695- ; Repair faulty limit switch'on.RCIC steam to turbine.
valve 2E51-F045:
'
2-38-025811 Repair.mid position stop on closing for RCIC steam to-
turbine valve 2E51-F045
%i
12-88-01953 Repair leak-by on RCIC steam to. turbine valve
'
2-88-01025 7 Adjust limit switch per MAC test on RCIC steam to
, turbine. valve 2E51-F045
-2-88-00231 Repair / replace Bellville washer pack on RCIC steam
to turbine valve 2E51-F045
1-88-08045 . Correct Faulty trip on RCIC steam to turbine valve
l-88-04251 Adjust declutch. lever and fingers on RCIC steam to
. turbine' valve 1E51-F045
o 1-88-04248' Adjust operator to prevent coasting into backseat on
RCIC steam to turbine valve 1E51-F045-
'
- 1-88-01311 Perform MAC test at system flow on'.RCIC steam to:
turbine valve IE51-F045
1-88-01310 Perform static MAC test on RCIC steam to turbine
valve IE51-F045
~1-88-00589 Correct-seat-leakage on.RCIC~ steam to turbine valve
2-86-3811 Verify torque of bottom bolts to 45-50 foot pounds for
Unit 2 HCUs
(
1-86-7330 Install missing flat washers and verify torque to
45-50 foot pounds for Unit 1 HCUs
'
1-88-5022 Modify reactor building RHR supports E11-RHR-H33, 34,
35, 293 and 274
2-85-1424 Weld patch on Rx building roof drain 2T55-RSD-5
2-89-00724 Torque back-to-back top plate cap screws to 15-25
foot pounds and confirm full thread engagement
for all Unit 2 HCDs
7 1-89-01077 Torque back-to-back top plate cap screws to 15-20
foot pounds and confirm full thread engagement for
all Unit 1 HCUs
-___ - _ _ - __ _ - _ -
_ _ _ _ _ _ _ _ _ _ _ _ _ - _ -
!
..
1
Appendix B 6
1-88-07300 Weld patt:h to repair valve IN71-F200 (FW001) l
2-89-0619 Fabricate and install motor base for surge tank pump
.
_ - - _.
- __ _ _ _ _ _ _ _ _ _ _ _ _- _______-_ ___-___ _ __-___
+.
.s.,
APPENDIX C
List of Components / Systems / Areas
Inspected During Walkdown Inspections
Unit 1 Main Generator
4160V Station Service Switchgear
IG 1R24-SO45
IB 1R22-S002
1C 1R22-S003
' IDL 1R22-S004
2A 2R22-5001
2B 2R22-5002
2C 2R22-S003
2D 2R22-S004
IF 1R22-S006
IG 1R22-S007.
2E -2R22-S005
2F 2R22-S006-
2G 2R22-5007
600V Station Service Switchgear
IB' 1R23-5002
2A 2R23-S001
28 2R23-S002
Motor Control Center (MCC)
2D 2R24-S035
208V 1G 1R24-5045
250 VDC 2A 2R24-5021
600/208V 2A 2R24-S013
600/208V 2C 2R24-S011
600V 2E 2R24-5018B
600V 2E 2R24-S018A
250VDC 2B 2R24-S022
Five (5) diesel generators and associated batery rooms and electrical equipment
roomr.
Cooling tower electrical room
Cooling tower fan motor 2W24-C002 at tow 9r 4
l ' Inverters IR44-5002 and S003
Inverters 2R44-S002 and S003
l
l
t
b . ..
_--
- . _ . ._ . - . .
,' ;, (
- Appendix <C ,
>
2-
l>
< _ Units I and 2 control- room and back panel areas
Unit 2 Recirc pump motor generator sets 2B31-S001A and B
l Backwash-Pump 2N21-C008
Condensate. Pump'2d21-C001A
'
. Condensate. Booster Pump 2N21-C002A~
,
Recirc MG Set 2831-S001B
Valve 2P42-F3006B
Lube 011: Circ Pump 2B31-S001B3
EHC'011 Rooms'(Unit 1) including EHC Purr.ps . IN32-C001A and 1N32-C001B and EHC
Coolers-
. Reactor.SBuilding Closed Cooling Water Heat Exchangers 2P42-B001A and 2P42-B001B
and Pumps 2P42-C001A and 2P42-C001B-
Backwash. Pump1 1N21'C008
,
. Unit 1: demineralized Valve. Nest
Condensate Pumps IN21' C001A', ~ 1N21-C001B, and :IN21-C001C
-
Unit 1 NE diag. and HPCI Room' including Main Pump IE41-C001.and
. Booster Pump IE41-C001
Unit li SE ' diag. including Pump Motors '1E11-C002A,1E11-C002C, and 1E11-C001A-
Unit?2'SE diag. HPCI Roon. including Main Pump 2E41-C001 and Booster Pump
FW Pump N21-C003A and general area
Electric Fire Protection Pump -1X43-C001 and. general area
Fire-Protection Jockey Pump IX43-C003 and general area
Unit 2 Loop'A RHR Pump 2E11-C002A;
discharge pressure gage 2E11-R003A;
suction pressure cage 2E11-R002A;
and nearby area
'
Unit 2 Core Spray Pump 2E21-C003A and Valve 2E21-F126A
Control Room and annunciator system for main steam
,
Unit 1: Turbine Building
QL _ _ _- _ - . - - _ - . - - _. 1
--_ _
-
.
Appendix C 3
Unit 1 and Unit 2 HCUs
Unit I and Unit 2 Reactor Buildings at 130 foot elevation
Unit 1 and Unit 2 Turbine Building east cableways
Unit 1 and Unit 2 Service Air System P51
Clean and Hot machine shops
Valves as listed below:
IZ43-F45-1AB 2P51-F087
IZ43-F45-2AD 2U43-F309C
IP42-F057 2P51-F098
IP41-F368B 2U43-F091
m. ______