ML18093B407
| ML18093B407 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 01/17/1989 |
| From: | Blumberg N NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML18093B406 | List: |
| References | |
| 50-272-88-80, 50-311-88-80, NUDOCS 8901240164 | |
| Download: ML18093B407 (45) | |
See also: IR 05000272/1988080
Text
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report Nos.
50-272/88-80 and 50-311/88-80
Docket Nos.
50-272 and 50-311
License Nos.
DPR-70 and bPR-75
Licensee:
Public Service Electric and Gas Compahy
P.O. Box 236
Hancocks Bridge, New Jersey 08038
Facility Name:
Salem Units 1 and 2
Inspection At:
Hancocks Bridge, New Jersey
Inspection Conducted:
October 17-28, 1988
Inspectors:
D. Caphton, Sr. Technical Reviewer, Team Leader
J. Carrasco, Reactor Engineer
P. Drysdale, Reactor Engineer
B. Hughes, Operations Engineer
H. Kaplan, Sr. Reactor Engineer
R. McBrearty, Reactor Engineer
~pproved b~
_ ~
NOrman ~1Uffib¢.~*
Operational Pr{grams Section,
perations
Branch, Division of Reactor Safety
\\ lr1l<i 1
date ' *
Inspection Summary:
An announced outage team inspection on October 17-28, 1988
(50-272/88-80; 50-311/88-80)
Areas Inspected:
Inspection of refueling outage activities which included
design change modifications/installations; inservice,inspection; and licensee
action on previous inspection findings for Units 1 and 2.
The inspection also
included modifications to control panels at the training simulator in Salem,
Results: No violations were identified; however, two unresolved items were
identified.
Installation of modifications was determined to be adequate;
however, a number of concerns were identified with the licensee's management
controls relative to the design change and modification installation process.
The inspector noted' that in several cases NRC identified concerns were already
known to the site management; however, an apparent lack of direct management
action allowed these concerns to persist without a clear plan for resolution.
In some cases where the inspector identified inadequate work, concerns were
raised which reflect little or no documented evidence of an effective oversight
directly coupled to the work product (See Attachment C for a listing of
identified concerns).
8901240164 890118
ADOC:I< 05000272
Q
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DETAILS
1.0 Persons Contacted
The names and positions of individuals contacted during this inspection
are listed in Attachment A to this report.
2.0 General
2.1 Objective and Scope of Inspection
2.2
The objective of this inspection was to evaluate the licensee's
performance in implementing the Unit 2 outage activities with
particular emphasis placed on design change modifications*and
their installations.
Licensee corrective actions taken on
previous inspection findings were inspected with particular focus on
structural items.
The licensee's inservice inspection outage work,
including the progress being made by the licensee's plant piping
erosion/corrosion prevention and control program, was also inspecte~.
The licensee has implemented a new program for controlling design
change modifications/installations for Salem.
However, for this Unit
2 outage the majority of the modifications were performed under the
old program.
The inspectors selected three modifications being
.. performed under the.,new pr,ogram (this is annotated in the .tit 1 e for
these modifications in the pertinent paragraphs of this report) and
the remainder of the modifications inspected were under the old
program ..
Since some outage activities were still in progress at the conclusion
of the inspection, it was not possible to confirm final closeout of
each outage activity.
The inspection did examine the licensee's
controls to assure that each activity had progressed properly through
the system and appropriate controls and procedures existed to ensure
final closeout prior to plant restart.
Summary of Conclusions and Findings
The licensee is in a transition period of implementing a new design
change modification process. Modifications were being performed
under both the old and new process.
Overall, the installation work for modifications was found acceptable
whether under the old or new process.
However, certain management
controls were found for both processes to be lacking in attention to
details in a number of areas inspected by the 'f:eam (refer to
Attachment C for a listing of identified concerns).
The team
concluded that an increased level of management attention and
involvement is needed to improve effectiveness of the design
change/modification/installation process.
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The licensee's approach to handling 10 CFR 50.59 reviews
exhibited a lack of preciseness and attention to detail (Refer
to Attachment C).
Design analyses for potential consequences of
system or component failures was also noted to exhibit weak-
nesses, for example, during the inspection of the design change
involving the P-9 modification (2EC-2193, reference paragraph
3.9.c.), the'analysis failed to examine potential consequences
of system or component failures. *
The licensee's QA audits are capable of identifying program
problem areas to the plant management as noted in QA's audit
of the Engineering and Plant Betterment (E&PB) group's design
change/modification process.
These program audits are, however,
relatively infrequent (approximately on a two year cycle).
The
in~pection team concluded that without aggressive management
involvement to assure that corrective actions to audit findings
(including reaudit of deficient areas) are properly pursued and
resolved, QA's overall effectiveness in program and process
improvement will be limited.
The inspectors also noted during
this inspection that quality assurance of the ongoing work
exhibited lapses, i.e., where direct QA involvement was
absent.
The licensed'~ inser~ice inspection program (ISI) was effective
in meeting *appli"cable ASME .code and* regula.tory .requirements.
The licensee's plant piping erosion/corrosion prevention and*
control program is being implemented, however, it needs to be
strengthened in several areas to assure its effectiveness.
During the inspection of modifications to control room panels,
several questions arose regarding dual-licensed reactor
operators shifting work stations between Units 1 & 2 control
rooms without restriction.
These concerns were addressed
and resolved.
(See paragraph 3.1.b.)
3.0 Design Change/Modifications (Modules 37700, 37701, 37702, 37828, 55050,
57050, and 72701)
a.
Scope
The following is a list of the modifications inspected.
At the time
of the inspection, most of the work had been completed on the listed
modifications .
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Design Change Request (OCR)
Modification Description
Identffication No.
Correct Human Engineering Discrepancies
Install ATWS Mitigation System Actuation Circuitry
Incore Instrumentation Mods
-
Service Water Fan Coil Mods
-
Replace Service Water Butterfly Valves
- *Replace Service Water Expansion Joints
- *Diesel Cable Reroute
-
Reactor Control and Protection Mod, P-9
- Auxiliary Feed Water Pump 2 inch Bypass
-2EC-1915A
- Modifications worked under the new Engineering.and Plant Betterment
(E&PB) design change modification installation program. *
b.
Details of the Inspection Activities Performed
The inspection included specific observations concerning each of the
modifications and included:
Conducting system/equipment walkdowns in the field to confirm
as-built information per installation drawings.
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Verifying that installed conditions conformed to modification
specifications and drawings.
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Observing ongoing installation work, inspection and testing.
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Reviewing portions of the work that were already completed.
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Verifying that engineering work was technically sound.
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Verifying that the level* and type of verification of quality was
adequate for selected work.
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Determining proper classifieation of work according to standards,
e.g., ASME requirements.
Verifying that field changes were dispositioned properly .
Verifying that personnel were being trained as appropriate.
In addition to checking the above items on each of the selected
modifications, certain modifications were checked for the following:
That installation and inspection procedures.were adequate.
That onsite and offsite review committees performed their review
responsibilities concerning the modifications.
That there was proper level of QA/QC involvement in inspection
activities and problems.
Specific inspection findings and pertinent inspector observations
concerning each of the selected modifications are discussed below.
3.1 Correct Human Engineering Discrepancies (HED) in the Salem Unit 2
a.
b.
Scope .
Design change request 2EC-2151 made numerous changes to the
switch/control locations on the control room panels.
These
changes were generated during the Control Room Design Review
performed in accordance with NUREG-0700.
The inspection reviewed the design input and review process,
- the completed field installation, workmanship, training,
staffing documentation, housekeeping, fire barrier control,
welder qualifications, control room access, and various
work procedures.
The inspector interviewed craft superv1s1on, craft fire barrier
installers~ the Station QA manager, off-site review engineers,
control room operators, simulator instructors, emergency
procedure co-ordinator, contractor engineers, and the plant
operations engineer.
In addition, the inspector visually
inspected the Unit 2 control panels, the changes made inside
the Unit 2 control panels, the mockup facility, and the revised
control panels of the training simulator.
Findings
The design process utilized a full scale mockup and solicited
licensed operator feedback regarding improvement changes being
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proposed in addition to a detailed control room design review.
The simulator was modified prior to the control room and again
operator feedback was utilized for this design change.
The
workmanship was adequate.
Measuring and test equipment (M&TE)
was properly controlled and documented.
The fire barriers were
adequately controlled and found to be reinstalled. Operator
training was conducted and documented.
New control room panel
labels have been installed which enhance performance of the
emergency operating procedures.
Two discrepancies were identified.
The 50.59 review was not
properly executed inaccordance with procedure GM8-EMP-009 .. The
50.59 Safety Evaluation Form (VPN-030) was not signed by the
designated reviewer or by the department manager.
Visual
inspection beneath the Unit 2 control room.console panel
identified that the relocated recorder had a double nut
installed, which was contrary to the analyzed design.
The
unauthorized double riut arrangement was corrected promptly when
brought to the licensee's attention.
The inspector compared the Unit 2 control room changes to the
existing unchanged Unit 1 control room,* Due to the many
- observed differences between *the units., resulting from the
changes made to the Unit 2 control room, ~ concern developed
regarding whether or not dual~licensed reactor. operators should
be* restricted from rotating between the units. A meeting was
held between Salem Operations, Training, and Engineering staff at
the NRC Region I office to determine if dual-licenses should be
modified under 10 CFR 55.6l(b)(2).
Additional control room
inspections and interviews with control room personnel were
conducted, and a course of action was prepared to define new
requirements for dual-licensed operators. Additional NRC and
licensee activity regarding this matter was conducted outside
the scope of this inspection and results are detailed in NRC
Combined Inspection Report Nos. 50-272/88-19 and 50-311/88-20.
New staffing restrictions have been finalized in a letter:
Labruna, PSE&G to US NRC, dated October 28, 1988.
c.
Conclusion
This design change was extensive, involving approximately 10
volumes of documentation.
Operator feedback was acted upon
where possible.
Training was adequate for.both operators and
craft personnel.
Control room access was maintained in a
controlled manner during the installation.
The 50.59 signoffs
were missed by several reviews, indicating that the reviews,
including QA's, were not effective in this case.
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3.2
ATWS Mitigation System Actuation Circuitry (AMSAC) (OCR 2EC-2174)
a.
Scope
Design Change Request Package (OCR 2EC-2174, AMSAC) adds a
process cabinet, signal isolators and cables, and modifies
existing connections.
The inspector reviewed the completed installation, and visually
inspected the new process cabinet and interconnections in the
field.
The inspector interviewed the team leader and jointly
walked down the process cabinet wiring changes, fire barrier
installations, and evaluated the quality*of workmanship performed
in the field.
The inspector reviewed the 50.59 evaluation for
adequacy and completeness, and checked document control and
cable records.
b.
Findings
The inspector reviewed controlled prints for the cable pull cards
used *for the installati~n. The pull cards matched the
controlled drawings and Loop 529 was found correctly installed
in the field.
Workmanship wa*s adequate.
The -accessible field
run cable was visually inspected and found free of nicks,
- abra~ions, cuts .or any evidence of damage.
The 50.59 review*was
- properly executed .. Housekeeping was adequate on the new
installation but debris was found in the safety related
protection cabinets.
Further visual inspection of the nuclear
instrumentation (NIS) cabinets revealed a cigarette butt which
apparently had been extinguished on the fire stop inside the
cabinet. It was then identified that the rear doors to the
nuclear instrumentation cabinets are normally left open because
of interference problems.
Leaving the door open has potential
to compromise fire protection.
c.
Conclusions
Foreign material in the Reactor Protection and process cabinets
has been accumulating over a period of time.
QA reviewed the
installed wiring changes and either did not notice this debris
or accepted the condition. The cigarette butt in the NIS
Cabinet demonstrates a lack of adequate control and supervision
of personnel having access to this safety related equipment.
The Nuclear Instrumentation System rear doors are always open
due to a cable run petruding from the instrumentation inside the
cabinet.
Station Management was apparently aw~re of the
condition and had not taken action to correct the condition to
permit closure of the rear panel doors .
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The licensee took corrective action during the inspection to
clean out all the process and protection cabinets.
Four bags of
debris and a flashlight were removed during the licensee's
cleanup of the reactor protection and process cabinets;
Except as noted above, the AMSAC modification was found to be
installed in accordance with the design package.
3.3 Incore Flux Monitoring (OCR 2EC-2232) and Core Exit Thermocouple
(OCR 2EC-1915A) Systems Modifications
a.
Scope
This modification work involved removing the 64 top-mounted core
exit thermocouple assemblies and the 58 bottom-mounted incore
flux monitoring thimbles.
Both systems were then converted to
an integral bottom-mounted flux thimble thermocouple (FTTC)
'system which includes associated incore detectors, external
cabling, junction boxes, containment penetrations, signal
processors, and control room instrumentation, etc.
The
modifications are a design upgrade to make the new system
"Safety Related Equ.ipment. 11
The system now meets the Seismic
Class I *and Envirpnmental Class lE criterja, and also conforms
to the requirements .of Regulatory Guides. 1.89 and 1.97, and
NUREG-07~7.
.
The inspection effort in this area involved a review of the
OCR work packages to ascertain that these modification~ are
in conformance with the Technical Specification, 10 CFR 50.59
and other regulatory requirements; and that the licensee has
implemented a QA program to control these plant modifications.
At the time of this inspection, all installation work associated
with these modifications had been accomplished.
Functional and
operational system testing could not proceed until plant startup,
which would be subsequent to this inspection period.
Specific areas covered in the inspection of FTTC modifications
are as follows:
Review of detailed work instructions to ensure technical
adequacy and proper implementation of administrative
requirements.
Review of fire safety practices associated with FTTC
modification work.
Review of control room Emergency Operating Procedures
(EOPs) and Abnormal Operating Procedures (AOPs) affected by
FTTC modifications .
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Review of 10 CFR 50.59 safety evaluations performed on new
FTTC systems and equipment.
Direct inspection of installed equipment to ensure
conformance with procedure requirements and high quality*
work practices.
Review of QA/QC program to verify that appropriate controls
of modification work were executed in a satisfactory manner.
b.
Findings
1.
Procedures prepared for DCR Packages 2EC-1915A and 2EC-2232
for Unit 2 were virtually identical in content and
structure to the corresponding packages prepared for the
same modifications performed during the last outage of Unit
1.
The administrative controls over the preparation of the
Unit 2 procedures had not been officially superseded by new
administrative controls of the Engineering and Plant
Betterment Department.
New administrative requirements
were never the less imposed on the conduct of this
modifications work, and complete and timely documentation
of work accomplished was therefore cumbersome.
Further
review of detailed work instructions indicated that al~
procedures had* received appropriate reviews and approvals
prior to the start of modifications work.
In the areas
inspected, work instructions were determined t.o be
technically adequate, and were maintained current.
Procedure steps were observed to be verified by the
installation contractor 1.s supervision and by the PSE&G
Project Supervisors.
Deficient conditions encountered
during work performance were adequately documented and the
required engineering resolutions were obtained and approved
prior to continuing with further work.
2.
The inspector reviewed selected procedure sections which
invoked fire protection requirements and also interviewed
fire department supervisors to assess the fire safety
activities and controls imposed on FTTC modiffcations work.
Specific items inspected were fire seal impairment permits
and fire watch coverage for new and modified cable
penetrations in the Auxiliary Building.
The inspector
verified that the impairment of all fire seals and barriers
had prior approval and that the necessary permits were
issued by the station fire department.
The inspector also
verified that the necessary fire watches were provided
during the impairment periods. A review of selected fire
watch logs was conducted for six separate shifts during the
time that the open 5 inch penetration in the control room
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floor caused impairment of a fire barrier.
The inspector
confirmed that the fire watch required by technical
specification for this barrier had been provided on an
hourly basis for the necessary time period.
No discrep-
ancies were noted in this area.
3.
Review of control room Emergency Operating Procedures
(EOPs) and Abnormal Operating Procedures (AOPs) revealed
that six EOPs and no AOPs were affected by FTTC modifica-
tions.
These procedures were being revised during the
inspection period to reflect changes in operators
responses and instrumentation differences resulting from
these modifications.
The inspectdr.reviewed the revisions
with the cognizant operations staff engineer and determined
that the revisions were appropriate and technically adequate.
All revisions for Unit 2 EOPs reflect the corresponding
changes made to EOPs on Unit 1 for the same instrumentation
modifications.
The inspector verified that all revised
procedures subjected to this inspection were completed,
approved, and in place in the control room prior to
achieving Mode-4 plant conditions.
4:
The inspector r.eviewed.the principal engineering document,
DE-AP.ZZ-008(Q) (supersedes GM8-EMP-028), which provides
guidance for personnel conducting, reviewing, and approving
10 CFR 50.59 safety evaluations.
This procedure has
recently been implemented in 50.59 evaluations at the Salem
Station. It provides a systematic and logical approach for
performing these evaluations based on five different
categories of design changes, and provides a significant
improvement over the procedure it replaced.
The procedure
does not, however, provide a mechanism for dealing with
50.59 reviews which must be amended or revised by unfore-
seen field conditions that require a change in actual
modification designs or installation details.
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It was observed that the 50.59 evaluation for the 5 inch
core bore penetration in the Auxiliary Building (OCR
2EC-1915A) stated that no work would degrade the Seismic I
integrity of the building because no rebar would be
disturbed in the control room floor.
In fact, three
sections of rebar were cut during this operation, and an
engineering analysis was performed to accept the altered
condition.
The analysis concluded that the condition did
not affect the original 50.59 evaluation.
However, the inspector concluded that cutting the rebar did
affect the 50.59 evaluation.
The inspector reviewed the
engineering analysis with the cognizant civil engineer and
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found it to be technically sound.
The altered installation
condition was also discussed with the individual who
. prepared the original 50.59 evaluation and with the Station
Licensing Engineer. Both agreed that the original 50.59 *
evaluation now presents an incorrect conclusion because
that evaluation presumed that no rebar would be cut.
They
also agreed that the civil engineering analysis does not
validate the 50.59 evaluation, even though it does support
its conclusion.
Although the 50.59 evaluation does not
specifically prohibit cutting rebar, the resulting weakness
in that evaluation would have been prevented by a more
thorough review, acknowledging a highly probable condition
e.g. cutting rebar, and identifying existing engineering
controls and practices that deal with such conditions.
The inspector also noted that the 10 CFR 50.59 safety
evaluation performed on the monorail installation in the
seal table room (OCR 2EC-2232) did not account for the
trolley assembly suspended from the monorail beam.
The
evaluation concluded that the integrity of the primary
pressure boundary components and safety related equipment
located at the seal table could be violated or degraded
only through gross fa i 1 ure of the monorail.. The in specto*r
noted that although the trolley and monorail were
adequately load tesied after installation, the trolley
is not restricted in any way from motion along the rail.
Furthermore, the trolley is assembed from standard
commercial catalog components of significant mass which
reside approximately 20 feet directly above the seal table.
Based upon visual observation of the seal table area, and
review of NRC Information Notice IN-84-55 and PSE&G's
Safety Evaluation S-C-R200-MSE-0322, the inspector
.concluded that sensitive primary pressure boundary
components and safety related equipment would be in direct
jeopardy if the overhead trolley disassembled or failed and
impinged upon the seal table.
The inspector discussed this
situation with the FTTC modifications Project Manager who
agreed that a complete safety analysis should be performed
on the entire monorail and trolley system. It was further
agreed that plant maintenance procedures should include
appropriate instructions to inspect and restrain or remove
the trolley in Units 1 & 2 prior to plant operation.
This
is an unresolved item (50-272/88-80-01 and 50-311/88-80-01)
pending completion of an adequate safety analysis and
revision of applicable maintenance procedures to address
the above concerns .
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5.
Direct inspection .of installed FTTC equipment and
components was performed to ensure that the work met the
specified requirements in the design documents, and that
the work had been performed in accordance with approved
procedures and instructions.
The installations reviewed
appeared to have been performed with good quality workman-
ship and were in accordance with specified technical
requirements.
No deficiencies were noted in this area.
6.
The inspector reviewed the FTTC modification DCRs and
eight Station QA Surveillance Reports (SRs) to assess the
extent and adequacy of QA involvement in the modifications
work.
It was noted that Station QA engineers had reviewed
and concurred in these modification packages and had
incorporated necessary notifications and hold points,
however, the work instructions and design packages had not
received any QA review for technical adequacy.
Station
QA engineers interviewed indicated that limited time and
resources precluded technical reviews for these modifica-
tions.
The Station QA Manager stated that technical reviews of
maintenance and modification work packages are periodically
performed by his organization.
Selected QA SRs reviewed by
t.he .inspector revealed that adequ.ate oversight functions *
were performed by Station personnel to assure that*
contractor.work practices, procedure controls,*QC methods,
ind personnel qualifications were proper, effective, and in
accordance with PSE&G requirements.
Except for the concern
regarding lack Of technical review of work instructions and
design packages by QA, no discrepancies were noted in this
area.
c.
Conclusion
The FTTC modifications instal1ed during the current outage
have been accomplished using technically adequate design
practices.
Personnel accomplishing the modification work, and
the engineering and QA services supporting the work were deemed
to be adequate.
The concerns raised by the inspector over the
adequacy of 10 CFR 50.59 safety evaluations reflect a lack of
attention to detail and thoroughness in the design and design
review processes.
No adverse affects regarding FTTC functional
capability or plant startup were identified.
3.4 Auxiliary Feed Water Pump Bypass Line (OCR 2SX-2003)
Note:
This OCR was performed under the new Engineering and Plant
Betterment procedures.
a.
Scope
This modification installs a 2 inch recirculation line across
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the No. 23 turbine driven auxiliary feed water pump.
The new
recirculation line will permit achieving 25% of rated flow and
permit stable flow for conducting the technical specification
required inservice test on the pump.
The existing recirculation
line only permits 100 gpm flow and the licensee's representative
stated that the pump manufacturer recommends 245 gpm to achieve
stable flow conditions.
The modification required structural
changes to provide for seismic grade hangers for the piping and
included penetrations of the metal enclosure room surrounding
the turbine driven auxiliary feed water pump.
In addition, the
OCR involves installing a clamp-on type flow measuring trans-
ducer having a digital display of flow in the vicinity of the -
auxiliary feed water pump.
The new flow measuring instrument
(trade name is Controlotron) will be used for inservice testing
of the pump to determine operability under the technical
specifications.
At the time of the inspection the installation wai complete
except for the hydro testing of newly installed piping which was
scheduled to be accomplished during plant start up when steam is
available to operate the turbine driven auxiliary feed water
pump .
Findings
The inspector visually examined the newly installed p1p1ng, the
welds,--hangers, valves and the penetration through the auxiliary
feed pump room metal enclosure wall.
No deviations were noted
regarding the actual installation versus the OCR design.
The
workmanship appeared adequate.
QC hold points were utilized
during pipe fit up and welding of the piping.
The weld records
were included in the OCR package.
The inspector noted an incorrect checkoff on the mechanical
package OCR Exhibit 7, Internal Hazards Analysis Specialty
Review Checklist, Procedure DE-AP.ZZ-0007(Q).
Question 6 asked
11 *** does the DCP involve deletion or modification of the
structures? 11
This was checked No in the mechanical package.
However~ on the civil package exhibit 6, the question was
checked, Yes.
During the visual inspection of the piping and
hangers, the inspector noted that revisions had been made to
existing' pipe an~pipe hang~rs, in addition new hangers were
attached via welding to existing structures and piping and
hanger penetrations were made through the auxiliary feed water
pump room metal enclosure wall and ceiling.
The inspector
reviewed this finding with the cognizant engineering group and
it was acknowledged to be an incorrect mechanical package
checkoff in view of the changes made.
Detailed Technical
Standards for engineers using the new E&PB procedures had not
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14
b~en published. It was stated that engineers and project
managers were being provided training in the use of the new
procedures.
The inspector noted that the OCR procedure check off lis~
DE-AP.ZZ-OOOl(Q), Exhibit 3, item D.
11 Interface Review" was
checked
11 No 11 in the mechanical package to questions 14, 16, and
17 which related to the operability interface on the front end
of developing the design.
The result of checking
11No
11 took away
the operations and maintenance department interface with the OCR
on the front end of its development.
Question 17, which was
checked No, asked
11Are there any human factors considerations? 11
A human factor consideration question was raised by the
inspector and is discussed below.
Inspection of the
11Controlotron
11 installation (ultrasonic flow
measurement) noted that the electronic cabinet containing the
flow indicator would be observable by an operator at the
recirculation line throttle valve No. 146 by looking through the
turbine driven auxiliary feed pump room's door opening.
However, the elec~ronic flow cabinet was mounted next to a
hydrazine tank and hydrazine fumes were noticeably present at
the electronic c~binet while the inspection was ongoing.
The
OCR did not consider the potential* for -a hydrazine .environment
for the operators or equipmen~. The inspector noted that the
Controlotron manufacturer's installation manual
indi~ated that
an independent air source would be needed if the electronic
- cabinet would be subject to a corrosive environment.
No
independent source of air was provided by the design.
Licensee
representatives subsequently stated that it was planned to
relocate the hydrazine tank to another area.
The OCR included completed separate 50.59 reviews and safety
evaluations for mechanical, electrical and civil areas.
The
inspector noted that the mechanical 50.59 review did not discuss
the consequences of a malfunction of a different type, for
example, inadvertently leaving open valve 2AF~144, the recir-
culation line block valve.
The cognizant design engineers
stated that if the valve was left open adequate flow would still
be provided by the pump due to its large capacity and adequate
time would exist to permit manual closing of the valve.
As
previously stated, the inspector noted an apparent lack of
formal guidance for engineers in completing the check list
procedures that make up the OCR package.
The engineering mapual system is described in OA-AP.ZZ-0002(Q),
Revision 0, approved May 13, 1988.
The overall system consists
of five manuals:
Engineering, Project Management, Technical
Standards, Programmatic Standards and Design Basis Documenta-
tion.
The licensee refers to the new E&PB Engine~ring Manual
- --
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c.
15
System procedures as "new paper" and to the old system
procedures as "old paper".
Instructions for changing from "old
paper" to "new paper" were.covered by documented directives and
letters. However, several instances were noted where it was not
specifically documented as to e.g. which procedure was the
preferred procedure where a new procedure had been issued before
superseding the old procedure, indicating a lack of management
preciseness during the transition period.
The manual system is
still under development, e.g., Technical Standards not yet
developed.
A lack of guidance for insuring consistency in
performance of the new E&PB procedures is considered a weakness
based upon the inspector's observations.
The station's valve lineups for operation are placed on a
computer system TRIS (Tagging Request and Information System).
The system's print out for control of the three newly installed
auxiliary feed water system recirculation bypass line valves was
inspected.
The pri~t out showed only two of the.three valves
were entered into TRIS at the time of the inspection.
The 144
block valve was listed as locked closed and the 145 drain valve
was listed as closed.
The 146 throttle valve was not listed.
It was noted that there was no valve position dual verification
indicated for bloc~valve No. 144 on the TRIS:
An Inservice Testing-Auxiliary Feed Pump procedure
SP(0)4.05-D-AF(23), Re.vision 8, has been prepa_red for use in
conducting the pump operability testing using the new recircu-
lation line and clamp-on flow meter.
This procedure requires
that block valve 144 be locked closed upon completion of the
procedure.
It also requires independent verification of the 144
valve position.
This procedure also requires the 146 valve to
be locked i~ the throttled position (this was not shown as such
on the TRIS).
Conclusions
The auxiliary feedwater pump recirculation line installation
appeared to be installed in accordance with the DCR package.
Guidance for the engineers completing the DCR package is less
than adequate to achieve consistency during package preparation.
A number of specific instances were noted where a lack of
attention to detail, lack of preciseness and a general looseness
in the implementation of the DCR work existed.
The interface
between design and operations appears to need improvement.
Improvement is needed in the DCR package review process.
3.5 Service Water Fan Coil Modification (DPR 2EC-2270)
a.
Scope
As the result of serious corrosion and erosion problems
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b.
16
experienced in the Service Water System (SWS), the licensee
initiated Modification No. OCR 2EC-2270 to replace the existing
type 316 stainless steel and cement lined carbon steel piping
with AL-6XN, a new relatively highly corrosion resistant stain-
less steel material.
The modification covered piping systems
associated with three of five fan cooling units (FCU) namely
- 21, #22, and #23.
The remaining portions of the SWS will be
replaced during subsequent outages.
AL-.6XN is an austeniti c
stainless steel consisting of 20% Cr-24% Ni-6% Mo with nitrogen
addition.
The filler material for the girth welds was alloy
625, a 60% Ni-20%Cr-9% Mo alloy.
The system was being replaced
in accordance with USAS 831.7, 1969 Edition and 1970 Addenda.
Nondestructive examination (NOE) requirements included 100%
visual and 100% liquid penetrant inspection.
To provide control
of welder performance and to monitor corrosion behavior of
welded joints, the licensee voluntarily imposed a 10%
radiographic inspection requirement on 3 inch and 10 inch welds.
Findings
The inspector reviewed the basis for the selection of the new
materials as recommended by the licensee 1s consultants Stone &
Webster and MPR Assp'ci ates.
The inspector determined th~t the
selection of the new material$ was based on a comprehensive
corrosioTI prevention *and controi program including laboratory
testing and turbine building lpop tests covering*various flow
and temperature conditions.
In house development of both
automatic and manual welding procedures was perfGrmed in
parallel with the corrosion testing with the aid of information
obtained in visits to European manufacturers and installers.
The licensee informed the inspector that installation of the
SWS piping was being performed by Stone & Webster, hydrostatic
testing and review of Code packages by Bechtel, and NOE by
Magnaflux Quality Services (MQS-Wilmington, Delaware).
The inspector reviewed the manufacturing and fabrication history
of the ALX-6N piping components and obtained the following
information.
The pipe material was purchased by Connex, the
shop fabricator (formerly Dravo, Marrietta, Ohio) from Trent
Tube, a division of Crucible Steel. In accordance* with Stone
& Webster Specification No. 001-P-3010 Trent Tube, the pipe
manufacturer, produced the piping by rolling and welding plate
furnished by Alleghney Ludlum in accordance with the requirements
of S8688 (plate) SB675 (pipe), SA312 and Code Case N438.
The
inspector reviewed random Trent Tube certified mill test reports
(CMTRs) which showed acceptable mechanical properties and
chemistry results.
The CMTRs and attached furnace charts
indicated that the welded pipe had been solution annealed
at 2175°F and held a minimum of 15 minutes followed by water
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17
quenching.
The reports also indicated that the material
had successfully passed corrosion, liquid penetrant, x-ray,
hydrostatic testing, metallurgical testing, and macro/micro
examination.
The latter included checks for inclusions,
undesirable sigma phase, weld undercut, and heat affected
zone cracks.
The inspector verified by a review of licensee's
surveillance Report VS87-122 dated December 30, 1987, that
Donovan Co., a subcontractor of Connex, had solution annealed
pipe bends as required by Specification 001-P301D.
The inspector observed the automatic welding of a 3 inch
schedule 40 pipe girth butt weld for spool piece C-S2-SWP-569.
Welding was performed using ~he Diametric Gold Track II machine
in accordance with automatic Tungsten Inert Gas (TIG) Welding
Procedure NDWP-58.
The root pass had been deposited using
manual TIG procedure NDWP-46.
The inspector visually noted the
machine settings for the parameters employed during welding
including amperage, voltage and wire speed.
The heat input
based on these parameters was calculated to be 19,320 joules/in,
well below the licensee's self imposed value of 35,000
joules/in. The inspector verified that welding procedures and
automatic machine operators (P62 and P69) qualifications
conformed to ASME IX w~lding*procedure and perform~n~e
requirements.
The inspector visually examined the deposited
intermediate layers and found the welds to be free of*
discernible defects with good fusion along the side walls.
The inspector also reviewed other welding procedures used in the
replacement program utilizing various combinations of automatic
and manual welding processes, TIG and SMA (shielded metal arc),
open butt and consumable insert for root passes, and found them
to have been qualified in accordance with Section IX
requirements.
The inspector reviewed two final document packages representing
Test #2 and Test #14 field hydros in FCU-21 and FCU-22 systems
respectively.
The records showed that testing was performed
successfully in accordance with specified ~ngineering require-
ments of 300-315 psig for a minimum of 10 minutes.
Weld History
Records 4831 and 5078, representing welds C-52-SWP-556-1 and
C-S2-SWP-3291-l were selected from these packages for review.
The former weld (556-1) was welded with ASME IX qualified
procedure NDWP-47, the latter (3291-1) with NDWP-58 and 46.
The
records showed that the final weld layers had been subjected to
liquid penetrant and visual inspection.
The former by MQS
inspectors, and the latter by PSE&G inspectors as identified by
their initials. Base metal and filler metal heat/lot numbers
identified in these records were compared to appropriate CMTRs.
The AL-6XN CMTRs (pipe or fittings) were identified as Trent
Tube ht-821481, ht-711631, ht-LBVM, and WFl ht-628 PNEl.
The
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Alloy 625 filler material CMTRs were identified as Techalloy
VX-0160AY, Lehigh Test Lab ht-1X12 and ht-04647, Huntington
Alloy ht NX04E2AK and Acros YN5859.
No deviations to SA or SFA
material specifications were observed.
.
,
The inspector requested the licensee to provide the results of
the self imposed 10% radiographic sampling program.
The
licensee reported a significant rejection rate for the 3 inch
Only three 10 inch welds were
radiographed.
Of these, two were rejected.
The automatic
process was primarily used for the 3 inch welds, whereas the*
manual process was used for the 10 11 welds.
For the most part,
the majority of the defects in the 311 welds appeared to be due
to lack of fusion that occurred during automatic machine welding
of the intermediate fill passes.
Some minor root conditions
(e.g., lack of penetration) were observed in the root passes.
The inspector reviewed some of the rejectable radiographs and
concurred with the licensee's interpretation.
The licensee
attributed the lack of fusion to the unauthnrized use of higher
than normal travel speed.
The defects in the 10 11 welds were
attributed to tungsten inclusions and lack of fusion.
All of
the rejected welds were successfully repaired or cut out and-
.* replaced with new welds.
The licen~ee.chose not to expand: th~
10% radipgraphic sampling plan for the following reasons: (1)
radiography was not a Code requirement (2) excessive repair
could leap to undesirable sigma formation in the heat affected
zone and (3) the type of defects found would have minimal effect
on the serviceability of the SWS because of the low operating
temperature and pressure involved.
In addition the licensee
supported their decision not to expand the radiographic sampling
plan, providing the inspector with a fracture mechanic analysis
of 3 inch welds with an internal defect (2~
11 long x 1/32 11 deep)
that represented a flaw twice the size observed in the radio-
graphs.
The analysis showed that internal defects of the size
described would not initiate and grow into fatigue cracks of
critical size, and also would not result in structural failure
of the pipe because of a reduction in cross sectional area.
The
latter is supported by the fact that the joint is four times
thicker than required.
The additional thickness is intended for
corrosion resistance.
In addition,. it is noted that-all welds
were liquid penetrant inspected.
The "inspector agreed with the
licensee's conclusion regarding the major internal defects, but
expressed concern about the potential effects of root defects,
albeit minute, which could act as initiation sites for crevice
corrosion.
Because of this concern, the license decided to
manufacture weld coupons with intentionally induced root defects
to be placed in the presently operating SWS corrosion test loops
for subsequent inspection and evaluation .
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19
The inspector reviewed the licensee's QA surveillance program
which was employed during the manufacture of the spool pieces at
Connex.
The inspector concluded that after reviewing numerous
reports that the license had conducted an intensive surveillance
program covering all phases of fabrication including bending,
welding, CMTR review, heat treatment after bending, and NOE.
All deviations and findings were reportedly resolved.
It is
noted that radiographs of girth welds were reviewed by the
licensee during his surveillance activities at Oravo, whereas
radiographs of the longitudinal seams as produced by Trent Tube
were not reviewed by the licensee .. The inspector requested that
the licensee verify that these radiographs had been reviewed or
if they were not reviewed, initiate review of same.
On October 25, 1988, the licensee reported that four weld
history records showed evidence that signatures of MQS liquid
penetrant inspectors had been falsified in four instances. It
is noted the problem was discovered by Stone & Webster and
reported to the licensee.
The licensee investigated the
incident and reported in Memorandum NQ5-88-0006, dated November
1, 1988, that four (4) of six hundred and seventy three (673)
weld history records exhibited apparently falsified signatures.
The-se welds have.been r.einspec;:~ed. Also fifty fo"ur (54) weld .
records generated in OCR 2EC-2187 were reviewed by the Hcensee.
No suspect records were found.
The person responsible for the
apparently falsified signatures has not been identified. The
licensee's investigation in this matter is still in progress.
The apparent falsification of weld records will be an Unresolved
Item 50-311/88-80-02 pending the results of the licensee's
investigation.
c.
Conclusion
The work involving the SWS p1p1ng replacement was found to be
in accordance with specified Code requirements and performed
under a comprehensive QA program.
The licensee's decision not
to expand a self imposed radiographic sampling plan when
significant defects were found was supported with adequate
justification. In addition, the licensee plans to prepare
mockups with similar defects for corrosion testing.
The
incident involving apparent falsified signatures was immediately
reported by the licensee.
An intensive licensee investigation
ensued which preliminarily indicated that this was an isolated
incident.
3.6 Replacement of SWS Expansion Joint (OCR 2EC-2207)
a.
Scope
As the result of determining that seven existing rubber
- ..
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-*--*- *-'
I .:
20
expansion joints in the SWS Intake Structure were not required,
Modification 2EC-2207 was initiated to replace the joints with
Belzona (ceramic epoxy) carbon steel spool pieces.
The use of
carbon steel spool pieces is intended to reduce future material
and manpower costs.
b.
Findings
The inspector verified that seven Belzona lined spool pieces,
2 feet long and made of SA 106 Gr ~carbon steel (identified as
2-SW-P-133, 131, 135, 137 141, 139 and* 143), were installed in
the intake structure.
The pieces had been shipped with one
slip-on flange welded on and one-slip on flange shipped loose
for field fit-up and welding.
The inspector visually inspected
a flange to pipe fillet weld and found no discernable defects.
A review of weld history records showed that the welds had been
welded with a combination of TIG and SMA processes in accordance
with qualified Section IX welding procedures NPWP-13 and NPWP-2.
The record showed that the weld had been subjected to visual and
magnetic par~icle inspection.
c.
Conclusions
The work described in tne subject modification was found to have
been performed as specified.
No deficiencies or violations were
observed.
3.7 Replacement of SWS Butterfly Valves (OCR 2EC-2203)
a.
Scope
b.
As the result of deterioration of the rubber lining and
attendant corrosion, seven existing carbon steel butterfly
valves were replaced with new aluminum bronze valves in the
intake structure (OCR 2EC-2203).
Findings
The inspector verified that valves identified as 24SW20,
23SW2C, 21SW17, *21sw20, and 22SW20 were installed in the intake
structure.
The inspector reviewed the CMTR 1 s and verified* that
the properties conformed to the requirements of SA-148-Gr C
95400. The certification indicated that the valves were temper
annealed at 1175°F for 7 1/2 hours.
c.
Conclusion
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21
The work described in the subject modification was found to have
been performed as specified.
No deficiencies or violations were
observed.
3.8 Diesel Cable Reroute (OCR 2SC-2011)
Note: This OCR was performed under the new E&PB procedures
a.
Scope
A licensee's design review identified a design deficiency
relating to 10 CFR 50, Appendix R,Section III, G.3.
The
deficiency was that the emergency diesel generator cable
2CDC22-CT, which provides an alternate source of control and
field flashing power to the three diesel generators during a
postulated fire that requires alternate shutdown measures,
was not physically independent of the ceiling area for the
"zone under consideration." The licensee's corrective action
initiated by OCR 2SC-2011 was to remove* the cable and reroute
the cable to comply with the Appendix R criteria.
A new seismically mounted conduit run was required to be
installed by this modification.
No electrical loads or *
c-ircuitry changes were required.
The new 2 inch conduit run
was from an existing tray (2A089) in the auxiliary building
where the cable was interceptea, through a newly drilled
4 inch diameter concrete wall penetration to the 480 volt
switchgear room cabin~t. The cable terminated in the same
cabinet as did the original design.
This design change, including the installation work, was made
under the licensee's revised Engineering and Plant Betterment
procedures and program for implementing design changes.
b.
Findings
The inspector noted that the work order for this modification
had been signed off as complete at the time the inspection was
initiated.
On October 18, 1988, the inspector walked down the
revised conduct run inside the auxiliary building and the 480
volt switch gear room.
The 2-inch conduit run within the
auxiliary building was visually observed to be wrapped with
fire wrapping from the tray to the wall penetration as required
by the OCR.
The tray had also been restored to match the
appearance of the undisturbed tray run.
The new penetration
through the concrete wall was observed to have been grouted
around the conduit.
The seismic supports for the conduct run
were also inspected and noted to be installed per the OCR.
. .. * . - .
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22
Inspection of the 2COC22-CT cable inside the 2C125VOC bus panel
in the 480 volt switch gear room noted that the excess cable
resulting from the new shorter cable route was handled by making
large loops in the front of the cabinet.
Inspection of this
accessible cable showed what appeared to be some minor nicks and
abrasive damage to the insulation.
The inspector asked to see
the acceptance criteria used to assess the insulation damage and
was told that none existed.
An engineering group representative
stated that damage to insulation was normally determined by
meggering the cable.
However, at this time the work had been
completed and no *meggering of this cable run had been done.
There was no QC hold point in the modification
11step by step
11
instruction to witness the megger of the cable.
The QA manager
stated that it was not intended to be meggered. The project
manager stated that a megger test would be conducted in
accordance with the procedures specified in the Installation
Verification Procedure, Insulation Resistance, Continuity and
Integrity Checks Ml3-IVP-501, Revision 0.
The inspector noted
that this procedure was part of the OCR installation package,
however the installation instruction appeared to permit
interpretation regarding the intent to megger the cable.
The* 2COC22-CT cable was satisfactorily meggeredby the
electrical contractor as witnessed by the inspector.
The above.
stated procedure was used during the meggering. * This megger
test was witnessed by QA.
The inspector requested to see the
post calibration test of the megger instrument.
The inspector
witnessed the satisfactory post calibration test at the
calibration lab.
A check at the tool room which issues measuring and test
equipment (M&TE) found the control of M&TE equipment issued to
contractors to be under adequate procedural control with one
exception.
The M&TE control procedure requires that before an
M&TE item can be issued to a contractor, the contractor's name
shall be on the approved list. The M&TE supervisor was found to
have recorded on the M&TE issue log a megger EG-ZNM-0653 Serial
No. G4539 issued (contrary to procedure) to a person not on the
approved list to receive the device.
The person was verified
later by the inspector to have been subsequently added to the
approved list.
The inspector reviewed the licensee's approved 50.59 Review and
Safety Evaluation.
The inspector had no questions regarding the
50.59 review.
c.
Conclusion
The installation workmanship observed by this inspector
appeared to be adequate.
The OCR lacked specificity in the
I
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I
'
I
'
'-
23
installation procedure to clearly and precisely specify that the
cable run be meggered following installation.
The OCR package
provided no acceptance criteria for use by the craft or QA
personnel to assess potential cable insulation damage
although
the potential existed for such damage in that the old cable had
to be removed then rerun through new conduct.
Some apparent
minor surface damage to that portion of the cable insulation
visually accessible was noted by the inspector. A-satisfactory
megger test was subsequently conducted.
3.9 P-9 Modification (OCR 2EC-2193)
a.
Scope
b.
Design change request OCR 2EC-2193, the P-9 modification,
replaces the existing reactor trip on turbine interlock
permissive C-8, turbine trip with permissive P-4, reactor trip.
The inspector visually inspected the NIS drawer installation,
the soldered connections, the qualifications of craft persons
performing the work, the relays installed in the reactor trip
breakers, the connecting cabling, seismic installation of a
cable pull box, anc;I the torque wrench used in. the installation.
The inspector interviewed the cognizqnt lead enginee~, licensing
engineer,* system engineer, operations engineer and craft
supervision. A lamp test in the Unit 2 control was visually
inspected to verify indication.
At the training simulator, the inspector observed a Turbine Trip
at 25% power without a reactor trip, which verified that the
software was changed in the training simulator.
The station
Operations Review Committee Meeting Minutes were examined to
ensure that this modification had been reviewed by them.
The
50.59 review was inspected for adequate technical basis and
completeness.
Findings
This modification was found to be installed without a properly
executed 50.59 review in that Form VPN-030 was missing a
department head approval signature.
Torque wrench EG-ANM-0146
with a scale range of 25 to 250 ft-lbs was used a~ a 30 ft-lbs
setting. This is in violation of station maintenance procedure
M-23
11T6rquing Guidelines
11 , which states on page 7 of 41
1100 not
use a Torque wrench to apply values that are below 20% or above
100% of the torque wrench scale. 11
The inspector determined
that the wrong size torque wrench was used to install four
anchor bolts holding a seismically mounted pull box located
above the reactor trip breakers.
Post calibration of this
*-------------~--~-------.---- --*-*---- --- --------------------------_-_
-* -*-*-'.".'--'-_ __;_,'C.c __ _
24
torque wrench was performed at the inspectors reque-St.
The
inspector witnessed the post calibration test and observed that
this wrench failed.
During the post calibration lab's
inspection, it was found that the lead seal on the adjusting
screw was missing.
The licensee does have procedural provisions
requiring rewrirk'upon identification of failed M&TE equipment,
however, subsequent engineering evaluation determined that the
applied torque was adequate and no further action is ~equired.
The licensee's approach to 10 CFR 50.59 safety evaluations
placed significance on identifying potential failure modes
instead of examining the potential consequences of system or
component failures.
The review covered normal system operation
but not allowable system operation, e.g., control rods in manual
vs. failure of power mismatch circuit in the rod control system.
When this was brought to the attention of the licensee by the
inspector, changes were made to the operating procedures to
limit operation with control rods in manual.
Visual inspection of the NIS field change kits verified adequate
installation and workmanship.
Visual observation of the wiring
and relay in the Reactor Trip Breaker Cabinet indicated proper
installation.* M&TE control was not adequate in that when
requested to locate a stopwatch used during the post test it
took a day tb find it, even though i~ was supposed to be in
the issue room.
Procedures are in place to checkout M&TE on
the backshifts but in some cases the proper documentation is
not filled out or th~ wrong entry log is used.
c.
Conclusions
The licensee's review process including the QA review did not
identify the missing 50.59 approval signatures. There was a QA
holdpoint to verify torque on the anchor bolts but proper tool
usage was not evaluated by the QA inspector.
The failure of the licensee to examine the potential
consequences of system and component failures indicates an
inadequate review process. It appears that the vendor's
evaluation of the design change was accepted without a thorough
review of supporting materi~l or use of adequate independent
review.
Based on the inspection with the exception cited above, the P-9
modification installation was found to be installed in accordance
with the requirements established in the modification package.
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25
3.10 Design Change/Modification Overall Conclusion
The design change/modification process was* in the process of being
upgraded by a new matrix type system.
New procedures and management
controls were in the process of being implemented at the time of this
inspection.
Most design change/modifications for the outage were
being accomplished using the old process, however some were being
accomplished using the new matrix *process.
This performance oriented
inspection examined design changes/modifications performed under both
the old and new processes.
Overall, the installation work for
modifications was found generally acceptable whether under the old
or new process.
However, management controls fo~ both processes were
found to lack preciseness and attention to details in a number of the
areas inspected by the team (Refer to Attachment C for specific
concerns).
The team concluded that an increased level of management
attention is needed to improve the effectiveness of the design
change/mod~fication process.
4.0 Inservice Inspection (Modules 73755, 73051, 57080 and 73753)
a.
Scope of Inspection
The licensee perfotmed inservice i~spection during tbis outage to .
comply with requirements of the. ASME Boiler and Pressure Vessel Code,
Section XI,-and witn its inservice inspection schedule for the 1984
outage~ The licensee additionally performed examinations in
accordance with document S-2-VARX-MFD-0517, Revision 0, entitled
"Ultrasonic Thickn~ss Examination of Piping Systems with High Rate
Probability of Erosion - Salem Generating Station, Unit No. 2
11 *
The following areas were selected for inspection:
Examination data related to RPV 60° azimuth meridianal weld No.
2-RPVCH-1446 C, Head to Flange weld No. 2-RPVCH-6446A, and weld
No. 12-CF-1243-lA, 12 inch diameter chemical and volume control
system weld.
Control of ISI related nonconforming items.
ISI vendor visual examination personnel qualification/certifi-
cation records.
ISI implementing NOE procedures.
QA/QC involvement in IS!.
Facility's plant piping erosion/corrosion prevention and
control program.
,.
. **.1 *
- -*----'-"***-'-**------. -~* -- _, - . - ---- ..
-**** -- .:.. -.:::.:<.~*.: __ ..:-:.::-.:. '*
26
The above areas were inspected with regard to compliance with
applicable ASME Code and regulatory requirements and, in addition,
NDE procedures were considered with respect to technical adequacy.
Nonconforming !SI items were inspected with regard to proper closeout
based on technical justification, disposition and the adequacy of the
- 'tracking system.
The QA/QC involvement was examined by reviewing QA
surveillance reports of !SI activities which were performed during
the 1988 refueling outage ..
b.
Findings
Inservice inspection is mandated by the ASME B&PV Code,Section XI,
and the Code edition applicable to a specific facility is identified
by 10 CFR 50.55a(g) based upon the issuance date of its construction
permit.
The Salem Unit 2 facility is committed to the 1974 edition
through the Summer 1975 Addenda.
The inspector determined that the examinations represented by the
reviewed data met the applicable Code and regulatory requirements
regarding test method, recording, evaluation, plotting and reporting
of results.
The inspector further determined that each data sheet
was reviewed by the licensee.
The initialled stamp appearing on
each sheet indicated that the item represented by the data sheet was*
. acceptab~e and met applicable ASME Code requirement~ Licensee*
procedure*No. M9-11P-01C, Revision 0 entitled "Review and Acceptance
of NDE Data Result Records of !SI Long Term Plan Examinations
11 is in
the PSE&G procedure review process and is intended to govern the
revi~w and acceptance of !SI data when it is adopted.
'Visual examination methods are not included in SNT-TC-lA.
Section XI, IWA-2300 requires that personnel performing nondestruc-
tive examination methods not covered by SNT-TC-lA documents shall be
qualified to a program that follows the guidelines of SNT-TC-lA.
The
program must be established by the plant owner or his agent.
The !SI
vendor visual examination personnel were qualified and certified in
accordance with the vendor's program which was patterned after
SNT-TC-lA.
Visual examination personnel qualification/certification
records confirmed that applicable requirements of ASME Section XI,
1974 Edition through Summer 1975 Addenda were met.
Nonconforming items were documented, tracked and closed out in
- conformance with lhe licensee's program.
The inspector traced the
items, which were documented by South West Research Institute (SWRI)
Customer Notification Forms, through the system and verified that
dispositions were technically adequate and that the closeout of each
item was properly done.
-*.- .!_._
The licensee has incorporated its !SI vendor's procedures into its
program, renumbered them according to the PSE&G system, and has
approved each procedure for use at Salem Unit 2 .
--~~- :::.'* ... _._ ,.., __ .. .:*. -
.~-- . . .. :::.\\_*,~ :
. '
. .
,**.*-;.:*
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- .-.:**,.
-**-' **.****-**
27
ASME Section XI requires that NOE procedures must_be approved
and qualified.
Liquid penetrant examination procedure M9-ISV-01S,
Revision 1 (SWRI-MDT-200-1/68) was qualified in accordance with ASME
code requirements.
Deviation 4 was written to permit the use of
non-flammable penetrant materials which were different than the
materials originally specified by the procedure, therefore, in
accordance with ASME Section V, the procedure was required to be
requalified with the new materials.
The qualification record of the
new materials was not available and, at the inspector 1 s request, the
qualification examination was performed.
The licensee 1s test block
No. 08 was used and the qualification examination was performed by a
Southwest Research Institute Level II examiner in the-presence of the
licensee 1s NOE supervisor and the inspector.
By virtue of detecting
the defects in the test block the procedure deviation 4 was
considered acceptable and qualified to perform its intended function.
Based on the above, the inspector had no further questions regarding
this matter.
The inspector determined that the procedures were approved by the
licensee, met applicable Code and regulatory requirements, and were
technically adequate for their intended use.
Quality, Assurance involvement in ISI ac.tivities was inspected by
examining .selected QA surveillance reports issued during the period
from September.7, 1988 to October 12, 1988.
The.specific reports
selected by the inspector were from the licensee 1s Surveillance Log
which is maintained by the QA department.
The reports covered a
variety of activities including 10-year hydrostatic tests, personnel
qualification/certification records, steam generator activities (tube
plugging, J-nozzle replacement, cleanliness and tool control, helium
leak testing, tube sheet marking verification), calibration of
ultrasonic equipment, liquid penetrant examination, ultrasonic
thickness measurements and functional testing of mechanical and
hydrodraulic snubbers.
Failed snubbers resulted in the testing of
an expanded sample, leaks observed during hydrostatic tests were
corrected and each report indicated that applicable procedures were
followed, personnel performing the activities were properly qualified
and that the final results were acceptable in each case.
Concern regarding erosion/corrosion in balance of plant piping
systems has been heightened as a result of the December 9, 1986
feedwater line rupture that occurred at Surry Unit 2.
This
event was the subject of NRC Information Notice 86-106 issued on
December 16, 1986 and its supplement issued on February 13, 1987 and
The licensee 1 s plant piping erosion/corrosion prevention and
control program for this outage included approximately 160 scheduled
-*--****:*:-- ::-* :*-*-- .
- .*
....... -** ... ..,.:~~:.::.-....:.-z-.E: ... *-* .............. ----=-- ---
- --*. -***~-*-'--~. :, ... __ :
--* '\\
28
examinations,.all of which were completed at the time of this
inspection.
The document entitled,
11 Ultrasonic Thickness Examination
of Piping Systems With High Rate Probability of Erosion 11 identifies
the systems and areas which are included in the program and requires
that each area be scheduled for examination for three consecutive
refueling outages, including the current outage.
Data collected
from each examination will be used to calculate the rate of
erosion/corrosio~ of each area and to determine the schedule for
future examinations.
Information regarding nominal pipe wall thickness and design m1n1mum
thickness is included in the aforementioned document, and the senior
staff engineer in charge of the project has prepared an engineering
analysis logic diagram to control the evaluation and disposition of
examination data.
No areas requiring replacement have been
identified at this time.
The program at Salem Unit 2 is new and the examinations performed
during the 1988 refueling outage represent the first time the
specific areas have been examined for erosion/corrosion.
The
inspector noted that the responsibilities for evaluating and
dispositioning examination results and for the prioritization of
future examinations are not well defined in the'program.
c.
Conclusions
The licensee's IS! program is effective in meeting the applicable
ASME code and regulatory requirements.
Quality assurance
surveillances have included a wide enough variety of activities to
'provide assurance that the program is being properly implemented.
The licensee's plant piping erosion/corrosion prevention and control
program is new.
Responsibilities for evaluating and dispositioning
examination results and for the prioritization of future examinations
are not well defined in the program.
5.0 Quality Assurance Audits (Module 35701)
a.
Scope
An inspection was made of audits relating to the Unit*2 outage and
engineering/modification work for the outage.
b.
Findings
The QA Audit Group's Audit Schedule for 1987-1988, Revision 3 dated
September 8, 1988, listed one recent June through September 1988
audit, NM-88-35, for Engineering and Plant Betterment.
Another
earlier audit was performed for the Unit 1 outage during October
.. ~-.-* .. ..,~*** -.--~*~-~***-.. *--.
. ***- --*. -*.
-.-., *... ".**. -:- .. - -- :
- -**-:*:o- .
- .-.".** ...
--'-'---"----'-~~~.:........:.~.___c...~---'-~~**_.~*-**_*~~~~~~ ~:
c.
29
and November 1987.
No recent audit of the Unit 2 outage had been
performed.
The audit frequency was stated to be approximately two
years.
The audit NM-88-35 of Engineering and Plant Betterment (E&PB)
focused on technical and programmatic areas and concluded that E&PB
implementation of design controls had not been fully effective and
listed specific program findings.
Technical specialists were used
to supplement the QA staff auditors.
Corrective action was being
initiated by* E&PB to the audit findings.
Seven
11Corrective Action Requests
11 (CAR) were listed in the NM-88-35
audit package.
At the time of the inspection, one draft response
(CAR No. SA-88-Q044-0) was provided to the inspectors.
A letter from
the Nuclear Engineering Standards Manager to the Manager-QA Programs
and Audits, dated September 29, 1988, requested an extension in time
for due dates on four CARs C017, C019, Q045 and Q043.
These
responses are due after the conclusion of this inspection.
Review of the licensee's NM-88-35 audit of E&PB conducted June 20
through September 2, 1988, identified that the audit focus although
different from the focus of this inspection, identified s0me similiar
type findings thus indicating to the inspectors the relative
effectiveness of the licensees audit to provide to upper management
indications of problems, e.g., 11that E&PB implemel'iltation of design_
controls.has not been fully effective.
11
Conclusions
The licensees audit NM~88-35 of !&PB appeared effective in
identifying specific technical problems upon which to base a general
conclusion that E&PB implementation of design controls had not been
fully effective.
The audit findings were addressed to the General
Manager by letter dated September 27, 1988 and corrective action
responses were pending at the time of the inspection.
The audit
findings had similarities to findings by this inspection e.g.,
instances where no acceptance criteria were specified, no evidence
of QA involvement, lack of attention to detail, no QC hold point
specified, implementing procedures/instructions unclear or lacking
specificity and design review inadequacies.
The licensee's audit
results reinforce the conclusion that the E&PB implementation of
management controls on a broad base has not been fully effective.
6.0 Containment Spray Valve Cracking - Unit 2 (Module 57050)
During this inspection the licensee reported that two 8 inch containment
spray valves exhibited weepage and staining.
The valves are utilized for
test purposes.
The test valves were temporarily bypasse~ with installa-
tion of a new line using flange connections. A preliminary metallurgical
investigation indicated that the leakage was due to chloride stress
corrosion cracking starting on the outside surfaces apparently from
'
.-
---*.-**-:** .. -*
....
......... - --
' ..
!
.*
30
service water system leakage.
The licensee was in the process of cleaning
and liquid penetrant inspecting piping components which may have come in
contact with the SWS leakage.
The results of the cleanup, the liquid.
penetrant inspection and the metallurgical report of the failed valves
have been requested by the inspector. This incident will remain an open
item pending completion of licensee actions (50-311/88-01-03).
7.0 ,Status of Previously Identified Items (Module 92701)
7.1. (Closed) Violation (50-272/87-08-01 and 50-311/87-09-01) Wall Survey
Conducted Without Written Procedure
The inspector verified and reviewed a written procedure number
S-C-SOOO-SDM-0582-1 dated May 6, 1988 which established an annual
Civil Engineering inspection program to verify the continuous
structural integrity of masonry block walls in Salem safety
. related structures.
The inspector found this procedure adequate
and self contained.
'Inspection of block walls and their drawings identified specific
cases where improper labeling of the block walls on the drawings
as well as the lack of physicaJ labels o~ block walls existed.
This concern* was expressed to the licensee, .who acknowledged:the
comment and agreed to impl~ment further changes in order to
improve the already established system of *contro~ of block walls.
Based upon the licensees existing procedural controls and commitments
to further improve his controls, the violation 50-272/87-08-01 and
50-311/87-09-01 is closed.
7.2 (Closed) Violation (50-272/87-08-02 and 50-311/87-09-02) Wall
Calculations Were Not Recorded Nor Controlled to Demonstrate the
Structural Adequacy of the Modification
The inspector verified the existence of a well documented and
self contained Computech Engineering Report for the assessment
of the structural integrity/qualification of the masonry walls.
In addition, the inspector verified the design modification
(see reference on Attachment A) based on the Computech analysis.
The inspector found the analysis/design modification to be adequate
and properly controlled.
Therefore, violation 50-272/87-08-02 and
50-311/87-09-02 is closed.
7.3 (Closed) Open Item (50-272/87-08-03 and 50-311/87-09-03) Provide
Description of Analysis Techniques and Results for Cracked Block Wall
Upon examination of wall designated 2-4A (at elevation 100 1-0
11 ,
separating Units 1 and 2); the inspector determined the existence
--- --- --- . **---* ------- --**-- .......... -~
-
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____:_~ *.'..__:_'_" .,
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31
of a detailed.evaluation that demonstrates the adequacy of the wall
after eight supports were mounted on the Unit 2 (north) side of this
wall.
A PSE&G document titled
11Masonry Wall Evaluation, 11 clearly shows the
calculation for the most critical of the eight supports based on the
loads.
This support is labeled CTAT-11 23.
The calculation shows
that the overall structural integrity of the concrete block wall is
maintained and the actual stress values are within allowable limits
(this is based on block wall capacities).
In addition, the inspector
verified physically that the crack on wall 2-4A was properly repaired
in accordance with a technically adequate procedure.
Therefore~ item (50-272/87-08-03 and 50-311/87-09-03) is closed.
7.4 (Closed) Unresolved Item (50-272 and -311/87-02-01) Lack of Evaluation
of Pipe Supports for Seismic Stresses Induced by Self Weight Excitation
The inspector reviewed selected calculations covering the inclusion
of self weight excitation of pipe support frames prepared by CYGNA
Energy Service, and verified that this inclusion does not affect the
structural integrity of the support frames.
This conclusion is based
- on.the summary of stress in-teractions for critical members
(calculation package P-2110, Revision 1, dated 6-11-87 page 10).
7.5 (Open) Violation (50-272 and -311/87-02-02) Use of Uncontrolled,
Instructions in Performance of Piping and Pipe Support Design
Activities*
The inspector determined that the licensee did not incorporate the
U-bolt, strap load capacities and the requirements for evaluation of
locally induced stress at U-bolt anchor support in the master pipe
stress/pipe support specification. However, the licensee is taking
steps to correct this issue and for this purpose the licensee
prepared the drafts for Sargent and Lundy (contractor for
the procedure consolidation task).
Therefore, this item, Violation (50-272 and -311/87-02-02) will
remain open until the inspector verifies the final draft of the
consolidated pipe stress and pipe support specifications.
7.6 (Open) Violation (50-272 and -311/87-02-03) Lack of Documented
Procedures and Instructions in Piping and Pipe Support Activities
The inspector determined that the licensee did not include a
quantitative acceptance crite~ia to perform the check of pipe
support displacements and rotations under applied design loads
to insure acceptability. Also, the licensee did not show any
documented criteria for pipe supports.
- , -. - **- -:.~
._,*
'
- .. * .*
....
- \\
... --~---*--*-* **-
__ ,,._,, ________
---~- ..
-****-* ... -.**--**-. . .. **
32
However, the licensee has prepared a draft for Sargent and Lundy to
consolidate all issues on self ~ontained pipe stress and pipe support
specifications; meanwhile, item 50-272 and.-311/87-02-03 will remain
open.
7.7 (Closed) Unresolved Item (50-272 and -311/87-02-04) Technical.
Concerns Related to the Use of Infinitely Rigid Supports in Piping
Stress Analyses
This item consists of two parts which were resolved by the licensee
in adequate and acceptable fashion as follows:
The approach of considering support hangers, guides and anchors
as infinitely rigid in the restrain directions triggered a
safety concern of underestimation of seismic piping response.
The inspector verified the technical justification for using
rigid support models in piping design basis analysis, prepared
by Sargent and Lundy engineers and concluded to be acceptable
and technically adequate.
Therefore, this issue is resolved on
the conservatism of the design.
'The* flexibility and the stiffness *matrices for U-bolts, row 2,
.column 2 had zer~ value.
This was a mistake whic~ was corrected
by Report No.
s~c-MPOO-VDC-0133-0 ~repared by Franklin Research
Center.
Therefore, Unresolved Item (50-272 and -311/87-02-04) is closed.
7.8 (Closed) Unresolved Item (50-272 and -311/87-02-05) Failure to
Implement Design Interface Requirements Between Mechanical and
Civil/Structural Groups
7.9
The inspector verified the existence of stress directive No. 18,
which is the identification and control of ~esign activities
between participating design disciplines.
The inspector verified
the implementation by reviewing a design change No. 2SC-2003 package
1 of 3 which in exhibit 2 and 3 delineate the interdiscipline
interface record and design consideration check list respectively.
This verification is sufficient for the inspector to determine that
there is adequate communication among disciplines involved in design
activities.
Therefore, Unresolved Item (50-272 and -311/87-02-05) is closed.
(Closed) Unresolved Item (50-272 and -311/84-05-04) Justification
is Lacking for Utilization of U-bolt for Axial and Torsional Restrain
The inspector verified the existence of an established base line for
torque values for safety related U-bolts piping assemble, including
specific diameters of l~ inches and
l~ inches, which were pointed out
on a previous inspection.
..
. ;- ". ,. - -* -*- -* ... **- , :*~**
. :
--' --~ ----*
- -- --"*-----'~-
33
This is shown on Field Directive No. S-C-VAR-NFD-0460, Rev. 4.
The
specific torque values calculated for 1~ inches and 1~ inches
diameters are shown on document P-12SWA-5 and 2C-CVCA-518
respectively.
The inspector also verified the existing on-going program to evaluate
the locally induced stress on the pipe at U-bolt anchor locations.
The licensee informed the inspector that the large bore analysis is
completed.
The inspector verified selected calculations to be
adequate.
Nevertheless, the small bore p1p1ng remains to be completed.
For this
purpose, the licensee has committed resources and budget to complete
the program in its entirety.
Therefore, this Unresolved Item is closed.
7.10 (Closed) Open Item (50-272/85-08-01) Catalytic Welding Procedure
M13A-7 for Gas Tungstan Arc Did Not Include Three Non-essential
Variables Specified by ASME Section IX
The inspector reviewed Public Service Welding Procedure NDWP-7
(simi1~r to M13A-7), whi~h is pfesently contained in the Public
Service Welding and Brazing Manual, and v.erified that all non-
essential variables ~re included in the subject procedure.
The
licensee stated that all procedures presen~ly in the manual contain
non-essential variables listed in Section IX.
It is noted that at
the time the finding was reported the licensee was in the process of
upgrading the manual in anticipation of applying for National Board
11 R
11 and
11 NR
11 Certificates.
8.0 Unresolved Items
Unresolved items are matters about which more information is required
to ascertain whether they are acceptable or violations.
Unresolved
Items are discussed in paragraphs 3.3.b.4. and 3.5.b.
9.0 Management Meetings
Licensee management was informed of the scope and purpose of the
inspection at an entrance meeting conducted on October 17, 1988.
The
findings of the inspection were periodically discussed with licensee
representative during the course of the inspection.
An exit meeting was
conducted on October 21, 1988 for team members concluding their inspection
at that time and a final exit meeting was conducted on October 28, 1988,
at the conclu~ion of the inspection.
The findings of the inspection were
presented at the exit meetings.
See Attachment A for persons attending
the exit meetings .
- -**--:*-... -;:-:;--- ... ----"7. *-: -- ..
,\\.*.:-.*
- '
- J.
34
At no time during this inspection was written material concerning
inspection fin~ings provided to the licensee by the inspectors.
The
licensee did not indicate that any proprietary information was involved
within the scope of this i~spection.
\\
ATTACHMENT A
1.0 Persons Contacted
Public Service Electric and Gas Company (PSE&G) and Contractors
L. Adams, Senior Installition'Engineer
c
R. Burricelli, General Manager, E&PB
M. Bursztein, Principal Safety Review Offsite
P. Benini, Principal QA Engineer
H. Berrick, Principal Engineer
b
R. Best, Nuclear Training Supervisor
D. Bhavnani, Senior Staff Engineer
b
P. Cartellano, SW Project Engineer, Stone & Webster
.B. Connor, Operations Staff Engineer
C. Connor, ISi Supervisor
R. Connors, Mechanical Systems Engineer
b
J. Cortez, Staff Engineer
_L. Doyle, Calibration Coordinator, Bogan, Inc.
abc
R. Donges, Senior Staff Engineer
W. Denlinger, NOE Supervison, ISi
J. Elwood, Insulator, Bechtel
b
J. Gorga, Stress Supervisor
c
H. Gross, Team Leader, UE&C
M. _Gross, _Quality Assurance Engineer
b
J. Hawks, Project M9nager
b * J. Jackson, Tech Manager, Salem OPS
A Kao,
Civil/Structu~al Supervisor
G. Kapp, Project Manager
J. Kerin, Senior Fire Protection Supervisor
P. Kwok, Senior Staff Engineer
.J. Lark, Station QA Engineer
b
M. Leach, Technical Staff Engineer
c
S. Lehman, General Physics Craft Supervisor
b
L. Leitz, Project Manager
J. Lloyd, Principal Nuclear Training
b
D. Dongo, Stress Supervisor
T. Mc!vaine, fire Protection Supervisor
abc
L. Miller, General Manager Salem Operations
M. Morroni, Technical Engineer
V. Morton, NOE Level III, Southwest Research Institute
J. Musumeci, Salem Operations Engineer
D. Namit, Senior Staff Engineer
P. O'Donnell, Principal Engineer
be
A. Orticelle, Outage Manager
P. Ott, Technical Engineer
a
Denotes attendance at the entrance meeting on October 17, 1988
b
Denotes attendance at the exit meeting on October 21, 1988
c
Denotes attendance at the exit meeting on October 28, 19ga
- - .**.*.
. * ... *.
\\
- '.
... *.
Attachment A
2
Persons Contacted (continued)
Public Service Electric and Gas Company (PSE&G)
abc
D. Perkins, Salem QA Manager
be
M. Raps, Standards.and A~surance Supervisor
R. Raymond, Lead Civil Engineer
F. Ricart, Offsite Safety Review Engineer
D. Rice, Installation Engineer, M&M Contracts
A. Robinson, Nuclear Technician
abc
b
J. Rowey, Project Engineer
F. Saraceni, Electrical Systems Engineer
T~ Shome, Civil/Structural Lead Designer
W. Schultz, Manager QA & Audits
W. Straubmuller, Project Manager
R. Swartzwelder, Senior Licensing Engineer
D. Tauber, Quality Control Supervisor
be
F. Thompson, Supervisor Nuclear Licensing
D. Thompson, Field Superintendent, Combustion Engineering
W. Tomanek, Senior Design Engineer, General Physics
b
H. Trenka, Project M~nager
L .. Trow, Principal E~gineer, Atometrics Co.
.
~- Vorderbueggen, PE Project Director, General Physics
M. Wita, Station QA Engineer
T. Worrell, Station QA Engineer
United States Nuclear Regulatory Commission (U.S. NRC)
abc
R. Borchardt, Senior Resident Inspector, Salem
ab
K. Gibson, Resident Inspector, Salem
b
P. Swetland, Chief, Reactor Projects Section No. 28
The inspectors also contacted other administrative, operational,
technical and contractor personnel during the inspection.
a
b
c
. :*.* ..
Denotes attendance at the entrance meeting on October 17, 1988
Denotes attendance at the exit meeting on October 21, 1988
Denotes attendance at the exit meeting on October 28, 1988
- . -.**-
- . *:*
.*
'
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. .
1.0
2.0
ATTACHMENT B
Reference Documents
Organization/Administrative Procedures
Procedure Number
Revision
OA-AP.ZZ-OOOI(Q)
0
OA-AP.ZZ-0002(Q)
0
NA-AP.ZZ-0008(Q)
0
NA-AP.ZZ-OOOl(Q)
0
DE-AP.ZZ-OOOl(Q)
0
DE-AP.ZZ-0003(Q)
0
DE-AP.ZZ-0007(Q)
0
DE-AP.ZZ-0008(Q)
0
DE-AP.XX-0009(Q)
0
DE-AP.ZZ-OOIO(Q)
0
DE-AP.ZZ-0048(Q)
0
GM8-MSP-001
. 3
GM8;-MSP-003
.0
GM8-~MP-004
1
GMB-EMP-005
2
GMB-EMP-009
2
OA~PJ.ZZ-OOll(Z)
0
Engineering and Work Control Procedures
Procedure Number
Revision
GM8-EMP-007
0
GMB-EMP-008
1
GM8-EMP-010
- 2
DE-AP.ZZ-0017(Q)
0
DE-AP.ZZ-0018(Q)
1
. . ' .
Title
E&PB Organization
Engineering Manual System
Administrative Control of Design
and Configuration Change
Preparation and Use of Procedures
Design Bases/Input
Modification Walkdown Program
Speciality Review
10 CFR 50.59 Reviews and Safety
Peer Review
Design Verification
Control of Calibrated Measuring
and Test Equipment
E&PB *Manua 1 *
Indoctrination and Training
Design Drawing Control
Design Calculations
Operational Design Change
Control
Matrix Organization-A Project
Overview
Title
Document Identification
NRC Bulletin, Information
Safety Evaluations & Field
Directives
Modification Concerns and
Resolutions
Engineering Deficiency
Control
-* -
-;*-** - .-
:*'*-;--**-*,*-
-*---~-,7'~:--*:**
., *: . .'* .\\ . _*::;*
- .* -- ----__ . __ __. * ..
Attachment B
2
Procedure Number
Revision
Title
DE-CS.ZZ-0013(Q)
0
Contractor Use of M&TE
DE-CS.ZZ-0014(Z)
0
E&PB Contractor Electrical
Installation Verification
Procedures
M13-IVP-501
0
Installation Verification
Procedure, Insulation
Resistance, Continuity
and Integrity Checks
M3K
3
Electrical Cable
Installation/Pulling
S-C-EOOO-EFD-0438
0
Technical Requirements for
Construction of Electrical
Installations
S-C-ECOO-EFD-0384
0
Acceptance Criteria for
Crimp and Formed Wire
Hook Terminations
Specification 401-P301D
Stone and Webster
D
Specification for Shop
Fabricated Piping
3.0 Structural References
,,
Document Number
Report/Revision
Title
S-C-SOOO-SDM-0582-1
5-6-88
Design Memorandum S-C-SOOO-SDN
Engineering Department
Annual Inspection of IE
Bulletin 80-11 Masonry
Walls
Computech
1-30-88
Control Facility
Engineering Report No
Building/Walkway and
SOOO-VDC-0-0197
Truckbay - Assessment
of Structural
Integrity/Qualification
of Masonry Wa 11
N/A
11-28-80
PSE&G Report on Re-evaluation
of Masonry Walls for Salem
Generating Station Unit 1
.-- -
- . -*
- *: . .
. ,_ .. -. *:- ....... , :*: . ...
-*- .. -
Attachment B
Document Number
N/A
N/A
N/A
N/A
PSE&G Stress
Directive No. 18
PSE&G Report No.
s~c-MPOO-VDC-0133-0
PSE&G Stress
Directive No. 17
PSE&G Design
Modification
PSE&G Fie 1 d.
Directive
S-C-VAR-NFD-0460
3
Report/Revision
N/A
12-7-87
N/A*
5-18-87
3-20-87
N/A
Rev. 1
6-2-88
8-24-88
Rev. 4
*'-"**
Title
PSE&G Repair Procedure for
Cracked Masonry Wall on
Reference Line No. 14
PSE&G Masonry Wall Evaluation
(Wall 2-4) in Reference to
CYGNA Energy Services -
Calculation Package
P-2110 Multiple Support
Self Weight Excitation
CYGNA Energy Services -
Calculation for Pipe
Stiffness
Pipe Support Evaluation
"Identification and Control
of Design Activities
betwe*en participating
Design Disciplines Salem
No. *1 and 2 Units 11
Analysis and Testing of
U-bolt Anchor Assemblies
Criteria for Evaluation of
Directive No. 17 Locally
Induced Stress in U-bolt
Anchors and Welded
Attachments
Installation of 211 Diameter
Recirculation Line #23
Auxiliary Feedwater Pump
Torque Verification Program
for Safety Related -
Piping U-bolt
Anchor Assemblies
. ' .. :*: ..... ~~-~*--* -* .*- - .
- ~** .* .
- .
Attachment B
Document Number
Sargent and Lundy
Engineers Report
EMD-064314
Franklin Research
Center Report
4
Report/Revision
12-87
3-14-85
Title
Technical Justification for
Using Infinitely Rigid
Support Models in
Piping Design Basis
Analysis
Analysis and Testing of
U-bolt Anchor
Assemblies
4.0 Non-Destructive Examination Procedures/References
Document Number
Revision
M9-ISV-01S
Rev. 1
M9-ISV-02S
Rev. 0
M9-ISV-03S
Rev. 0
M9-ISV-05S
Rev. 0
M9-ISV-15S
Rev. 0
M9-11P-01C
Rev. 0
AP-9
Rev. 14
- - :-* ... *:*-:~-
--.*~.,. ...
. :'
.. . . .. : ' .. _ ;, ..
~ ..... :::;i . ~ ..
Title
Solvent-Removable
Liquid Penetrant Color
Contrast Examination
(SWRI-NDT-200-1/68, Rev. 4)
Visible Water-Washable
Examinations
(SWRI~200"."3/7)
Dry Powder Magnetic
Particle Examination
(SWRI-NDT-300-1/26,
Rev. 4)
Manual Ultrasonic Examination
of Pressure Piping Welds
(SWRI-ND1-600-3/~2, Rev. 5)
Visual Examination of Nuclear
Power Plant Components by
Direct or Remote Viewing
(SWRI-NDT-900-1/51, Rev. 1)
Review and Acceptance of
NOE Data Result Records
of ISI Long Term Plan
Examinations
Work Control Program
-***:. - - -. *-. *.:-- . . *.::***3*:;-:*--*
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Attachment B
Document Number
S-2-VARX-MFD-0517
5.0 Drawings
Document Number
201093A-8706
207076A-8798
245702A-1682
201061A-8705
207082A-8798
5
Revision
Rev. 0
Revision
3-11-87
Rev. 27
10-17-86
Rev. 17
4-30.:.84
Rev. 2
10-25-83
Rev. 20
1-28-87
Rev. 18
6.0
QA Surveillance Reports (SR)
Title
Ultrasonic Thickness
Examination of Piping
Systems with High Rate
Probability of Erosion
Salem Generating Station,
Unit No. 2
Title
Salem Nuclear Generating
Station No. 1 & No. 2 Units
Auxiliary Building, Section
X-X, Sheet 2
Salem Nuclear Generating
Station No. 1 Unit
Auxiliary Building Floor
Plan Elevation 64 1-0 11
Architectural
Salem Nuclear Generating
Station Controlled
Facilities Building Walkway
and Truck-Bay Roof Plan
Wall Sections & Details
Salem Nuclear Generating
Station No. 1 & 2 Unit
Auxiliary Building Section
F-F
Salem Nuclear Generating
Station No. 1 Unit
Auxiliary Building Floor
Plan Elevation 122'-0"
SR 88-0639; .SR for Installation of Reference Junction Boxes for the 78'
Elevation Penetration.
SR 88-0647; SR for Combustion Engineering Welder Certifications.
SR 88-0649; SR for Review of Data Sheets and Test Equipment Logs to
Verify Combustion Engineering, M&TE Program Compliance with
PSE&G M&TE Program.
- '*
........... -*** .... ~*--~*. ~-= **::-.--:-
.--~*-------.
- =-~ -* .
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.-1
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.... *., *.
Attachment B
6
SR 88-0665; SR for 511 Core Bore in Control Room Equipment Room Floor.
SR 88-0688; SR for Review of OCR 2EC-1915A, DCP No. 2.
SR 88-0733; SR for Review of OCR 2EC-1915A, DCP No. 2.
SR 88-0852; SR for Removal of Flux Thimbles Nos. 23: 31, 42, and 49.
SR 88-1093; SR for Assembly and Installation of FTTC Hoist Frame in Seal
Table Room.
7.0 Work Orders
WO 880511051; Erect Flux Thimble Frame and Hoist in the Seal Table Room
to Support the FTTC Installation.
WO 881002055; Repair Penetration Seal #F-15612-112.
8.0 Other Reference Documents
Fire Protection Permit #88-654; Permit for Penetration Seal #F-15612-112
Impairment.
MCR-2EC-1915-5; Modification Concern/Resolution. for Cut Rebar in 5
11 Core
Bore.
ANSI B30.ll-1980; Monorails and Underhang Cranes.
NRC Information Notice IE-84-55; Seal Table Leaks at PWRs.
S-C-R300-CDM-486-0; Design Memorandum on Bottom Entry In-Core
Instrumentation System, Core Exit Thermocouple Upgrade for NUREG-0737.
S-C-R300-CDM-0490-0; Design Memorandum on Core Exit Thermocouple Backup
Display - Upgrade of NUREG-0737.*
S-C-R200-MSE-274; Design Memorandum on Flux Thimble Ejection and Seal
Table Leak; Review of Westinghouse and NRC Documents.
Civil Engineering Directive No. 1, Rev. O; Instructions for Drilling
Holes and Core Bores in Concrete.
S-C-R200-MSE-0322; Safety Evaluation of the Flux Mapping System;
Potential for Interaction of the System with the Seal Table Due to
Seismic Loads.
. .. ****-. '* :* *:** -.
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ATTACHMENT C
Specific Concerns
The inspection team used the following evaluation criteria for assessing
management activities relative to the -inspect.ion findings and concerns:
Involvement~ Active management participation to ensure that engineering
design, analysis, and work packages are adequately prepared, reviewed, and
approved; including active participation in revie~ of results of ongoing
work .
Control:
Active management participation during the execution phases of
work to ensure that administrative controls exist and are fully
implemented both in work performance and in deficiency resolution.
Attention to Detail: Sufficient oversight to ensure that adequate detail
is considered to properly prepare engineering and work documents and to
provide for adequate and timely resolution of deficient conditions.
The specific concerns identified during the inspection are tabulated
below:
Inspection
Report
Aetention
Paragraph
Concern
Involvement Control
to Detail
3.1.b
50.59 Review not properly
executed.
x
x
x
Double nut installed, contrary
to seismic design specified.
x
x
Adequacy of restrictions for
dual-unit operators
x
x
3.2.b
Debris found in safety related
cabinets (including cigarette butt)
x
x
Rear doors are permanently.open
to nuclear instrumentation cabinets x
x
x
3.3.b.4
50;59 review presents incorrect
conclusion after rebar was cut
x
x
x
50.59 review failed to consider
trolley assembly
x
x
..
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- -- ___ .__:.,::.-
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Attachment c
-2-
Inseection
Reeort
Attention
Paragraeh
Concern
Involvement Control
to Detail
3.3.b.6
Work instructions and design
packages received no QA review
for technical adequacy'
x
x
3.4.b
Incorrect checkoff on Design
Change Request, Exhibit 7,
question 6, of Procedure
DE-AP.ZZ-0007(Q)
x
x
3.4.b
DE-AP.ZZ-OOOl(Q); Exhibit 3D,
operability questions 14, 16,
and 17 as checked removed the
operations interface with the
modification design on the
front end
x
x
Lack of detailed guidance for
engineers doing design work
x
~
x
"Contra l otron" ( ultra~onic fl O\\'{
measuring) electronic cabinet
installed in a potential hydrazine
environment.
Operating personnel
may be exposed to t~e hazardous
environment
x
x
50.59 review did not consider
the consequences of a malfunction
.
'-'
of a different type
x
x
x
Valve No. 146 not listed
in the Tagging Request and
Information System
x
x
x
3.5.b
Sampling plan was not
expanded for weld defects
and root defects may have*
potential for initiating
crevice corrosion
x
x
- .. *:**
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-* *
-3-
Inseection
Reeort
Attention
Paragraeh
Concern
Involvement Control
to Detail
3.8.b
No acceptance criteria for
craft or QC personnel for
assessing damage to
emergency diesel gen~rator
cable insulati.on.
No QC
hold point to witness
meggering of the cable
x
x
x
One Measuring & Test Equipment
controlled megger was issued to
- unauthorized person contrary to
procedure
x
x
3.9.b.
without a properly executed
50.59 review
x
x
x
Torque wrench of incorrect
size was used contrary to
procedure.
Torque wrench
...
failed post use calibration
test and lead seal was
missing
x
x
x
Measuring & Test Equipment
'
controlled stop watch was
found to be missing for a day
x
3.9.c
50.59 review failed to examine
'*"
potential consequences of the
I
~ '!
1.-.
allowable system operation,
indicating inadequacies in
. '
the review process
x
x
' i
. *:~
4.0.c
Plant *piping erosion/corrosion
prevention and control program
- . .;
needs improved definition
x
x
x
I
'
5.0.c
Engineering and Plant Betterment
implementation of management
controls has not been fully
effective
x
x
x