ML18054A910
| ML18054A910 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 08/10/1989 |
| From: | BERRY K W CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 8908180078 | |
| Download: ML18054A910 (50) | |
See also: IR 05000255/1989007
Text
- . "' * G11neral 1946 West Parn11ll Road, Jackson, Ml 49201 * (6171 788-1638 ** -----August 10, 1989 Nuclear Regulatory
Commission
Document Control Desk Washington, DC 20555 DOCKET 50-255 -LICENSE DPR-20 -PALISADES
PLANT -RESPONSE TO INSPECTION
REPORT 89007 NOTICE OF VIOLATION
Kenneth W Berry Director Nuclear Licensing
Nuclear Regulatory
Commission
Inspection
Report 255/89007, dated June 28, 1989, identified
strengths
in inservice
testing programs and weaknesses
relative to design control. These weaknesses
resulted in three violations
supported
by numerous examples.
None of these examples were safety cant, but collectively
they indicated
a need for programmatic
refinements
and additional
communication
of management's
expectations.
The NRC required a written response to be provided within 30 days, however, discussion
between respective
members of our staffs extended the due date to August 10, 1989. This letter. summarizes
the actions to be taken. Details pertaining
to the specific items are provided in the Attachments.
Since 1986 significant
efforts have been undertaken
by Consumers
Power Company to provide for effective
control of Plant design change activities.
These efforts have resulted from evaluation
of performance
by Plant Engineering
and Corporate
Engineering
personnel, Quality Assurance
personnel, NRC and the Institute
of Nuclear Power Operations.
In achieving
an effective
design control process; procedures
governing
modification
control activities
have been revised, a single design authority
has been established, changes to the facility are. being effected through a single unified approach and expectations
and standards
have been communicated
to Design Engineering
personnel.
Procedural
upgrades have focused on translation
of design input to the desired output, controlling
and implementing
the design change in the field and providing
close coordination
of the design with the needs of the Plant. In the past, the design authority
for "minor" modifications
has resided at the Plant while offsite engineering
organizations
retained the design
for "major" modifications.
Establishing
the Plant as the design authority
for all changes to the facility has been effected by Plant sponsorship
of all design control procedures, Plant approval for assignment
of design individuals
and Plant review of all work completed
by non-Plant
organizations.
Further, OC0889-0167-NL04
8908180078
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- * * Nuclear Regulatory
Commission
Palisades
Plant Response to IR 89007 August 10, 1989 semi-annual
design seminars and monthly design supervisor
meetings which include Engineering, Construction
and Testing and Quality Assurance
personnel
are being conducted
to facilitate
communication
of procedural
changes, standards
and expectations.
2 Consumers
Power Company believes, and as recognized
within the Inspection
Report, these efforts have resulted in programmatic
strengths.such
as; good design procedures, improved equipment
performance
and competent, knowledgeable
personnel.
However, Consumers
Power Company also recognizes
that as industry performance
standards
are increased, weaknesses
in established
programs may develop which require additional
effort. NRC violation
255/89007-01
presented
19 examples of inadequate
design control related to design changes implemented
at the Plant. The first seven of these examples were related to the failure to correctly
translate
design bases into drawings, procedures
and instructions.
Five of the examples are acknowledged
as presented
and are attributed
to the failure to; 1) follow established
procedures, 2) provide adequate justification
and documentation
within cation packages or 3) provide for adequate technical
reviews of installation
efforts. Also, certain areas were identified
where procedural
enhancements
and improved design guidance would preclude recurrences. er, the remaining
two examples, 255/89007-0ld
and Olg, are not acknowledged
as* presented
within the Inspection
Report. For these two* examples we believe the design intent of the modification
was preserved
and verified by testing and that record drawings utilized reflect the as-built condition
of the Plant. The
nine examples were related to the failure to adequately
verify and check design. Eight of the examples are attributed
to the failure to; 1) follow established
procedures, 2) document engineering
decisions
or 3) provide for adequate technical
reviews. Also, certain areas were fied where procedural
enhancements
would preclude recurrence.
However, Consumers
Power Company does not acknowledge
the remaining
example 255/89007-011.
For this example, the Inspection
Report noted that a setpoint change was implemented
without assuring the design intent of the system had not been compromised.
In review of the documentation
supporting
the design change, it was verified that design intent of the system was considered
and documented
within the modification
package and had not been compromised.
The remaining
three examples were identified
as non-compliances
for the
to adequately
delineate
acceptance
criteria.
Two of these examples are attributed
to a lack of procedural
guidance within modification
procedures.
Consumers
Power Company does not believe example 255/89007-0lq
is valid as presented
in that appropriate
equipment
selection
criterion
were applied during design and documented
within the modification
package * OC0889-0167-NL04
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- Nuclear Regulatory
Commission
- Palisades
Plant Response to IR 89007 August 10, 1989 3 In an effort to ensure the accuracy of the existing plant design basis is maintained, discrepancies
identified
within analyses supporting
the cited design changes have been or will be dispositioned
and documented.
As an effort to collectively
utilize auditing agencies appraisals
of our past performances, the identified
deficiencies
were presented
to Design Change Engineers
with emphasis placed on strict adherence
to established
procedures
and the concept of Plant based modification
engineering.
Enhancements
being made to design change procedures
regarding
documentation
of engineering
judgement, substantiating
input assumptions
and* thorough technical
reviews will be presented
to design change engineers
via personal letters, performance
seminars and continuing
training programs.
Enhanced design guidance is being developed
for weld engineering.
Specifically, code training for weld neers is being conducted
as well as design change procedure
revision to "prompt" the use of existing weld engineering
guidelines
for proper code selection
and specification.
In addition, as part of the Configuration
Control Project, additional
engineering
guidance regarding
cable sizing and raceway fill, designing
fire barriers and fire stops, evaluating
station and emergency
power* system.component
loads and cable routing including
the effects of cable submergence, is being developed.
Additionally, more engineering
guidance in the form of an engineering
specification
will be developed
for the civil/structural
discipline.
This specification
will be developed
by July 1990. . NRC violation
255/89007-02
presented
two examples where socket fillet welds were not-verified
to be in conformance
with weld size requirements
provided in welding specifications.
These examples are attributed
to a failure to meet current expectations
for the control of design change implementation.
To . avoid further non-compliance, design change procedures
are being revised to present welding specifications
input checklists
and implementation
drawings, and to provide for technical
reviews of weld requirement
inputs by Maintenance
Planners.
Additionally, Design, Engineers
and Quality Assurance
personnel
are.being
provided with training on structural
and welding codes and their application
to weld installation
and examination.
NRC violation
255/89007-03
was issued for a failure to implement
and maintain Technical
Specification
low temperature
overpressure (LTOP) setpoints
which were changed through the specification
change process. The violation
is attributed
to poor
within the Technical
Specification
Change Request development
process. When the LTOP setpoints
were derived, Plant personnel
failed to identify that the value included in the Technical cation did not account for calibration
tolerance.
A letter of interpretation
has been submitted
to the NRR which documents
ou"' '1-'osition
and commits to revising the setpoints
in a forthcoming
Technical
Specification
Change quest. In the interim, surveillance
procedures
which provide for setting and verifying
the LTOP setpoints.have
been revised to remove the positive tion tolerance.
An evaluation
will be conducted
to determine
where ments in the Technical
Specification
Change Request process can be made to preclude recurrence.
OC0889-0167-NL04 . t -. * ..... :*. : . *:: ... '*, ,;**.*:.r
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- * Nuclear Regulatory
Commission
Pal:isades
Plant Response to IR 89007 August 10, 1989 4 The Inspection
Report additionally
requested
a written response be provided for certain, specific examples of programmatic
weaknesses.
The first weakness cited involved the addition of zener diodes in the safety injection
tank pressure transmitter
power supply without analyzing
potential
failure modes and without checking diode input voltage after installation.
The failure to fully analyze potential
failure modes is attributed
to personnel
error. Administrative
Procedures
currently
require that .a failure modes and effects analysis (FMEAs) be performed
as part of the safety evaluation
process. The periodic*
refresher
training program for design engineers
will include emphasis on FMEAs. The next weakness cited pertained
to the backup nitrogen supply modification.
Specifically, an unauthorized
design change was implemented
when field nel implemented
their own weld requirements
after identifying
that an priate weld was specified
by the design engineer.
The condition
is attributable
to the fact that welding maintenance
procedures
are not. ly integrated
with design control procedures, thus assuring that changes. in the field will be approved by engineering
before they are undertaken.
The welding maintenance
procedures
will be better integrated
with the design control procedures.
The third weakness pertained
to utilization
of different
editions of the ASME Code relative to stress intensification
factors utilized in analyses.
In summary, usage of the later addition of the ASME Code, as currently
described
in the Palisades.
Final Safety Analysis Report (FSAR), was discussed
in an April 1980 meeting between Consumers
Power Company and the NRC and found to be acceptable.
Our interpretation
of the results of this meeting was submitted
to the NRC in the draft form, revised FSAR pages in our Final Response to IE Bulletin 79-14 dated September
26, 1980. As indicated
in our submittal
to the NRC dated October 24, 1980, the use of different
code editions was found to be acceptable, reviewed in accordance
with 10CFR50.59
and placed in the Palisades
FSAR. Therefore, usage of different
code editions as presented
in the FSAR currently
represents
our position and is believed to be acceptable.
The last weakness cited pertains specifically
to the Engineering
Design Change (EDC) form utilized to revise facility changes not listing calculations
which may be affected by the particular
EDC. Therefore, it was unclear whether technical
reviewers
had considered
the effects of the EDC on the original analyses.
Consumers
Power Company believes that existing procedural ments direct the EDC initiator
to "reflect" the change in all affected tailed design documents;
the engineering
analysis was clearly identified
in the procedure
as being a detailed design document.
However, "engineering
analyses" will be specifically
added to the EDC form to ensure that technical
reviewers
consider effects on engineering
analyses and provide documentation
of this consideration
- OC0889-0167-NL04
I
- * Nuclear Regulatory
Commission
- **Palisades
Plant Response to IR 89007 August 10, * 1989 5 The Inspection
Report also requested
that specific discussion
be provided regarding
unresolved
items pertaining
to welding. This discussion
is ed on page 41 of Attachment
1. In summary, we acknowledge
that no corrective
actions have yet been directed towards reviewing
previously
made socket fillet welds for compliance
with code requirements.
Consumers
Power Company plans, however, to select an appropriate
sample.of
as-built welds and inspect the * welds during the 1989 maintenance
outage. The sample will be chosen to include a range of weld types. The purpose of the inspection
will.be to verify that the weld characteristics (type and size) conform to requirements
set forth in the repair inspection
checklist
and/or applicable
welding code. Kenneth W Berry Director, . Nuclear Licensing-
CC Administrator, Region III, USNRC NRC Resident Inspector
-Palisades
Attachments
OC0889-0167-NL04
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- * * .*_.* *; ATT0889-0167-NL04
ATTACHMENT
1 Consumers
Power Company Palisades
Plant Docket 50-255 DETAILED RESPONSES
TO INSPECTION
REPORT 89007 August 10, 1989 45 Pages . ' . * ,. . ..* *. , .. *** .. * *. ** ** .. *---"""--'---'----'-'-'""'"--=...;.......*
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- Violation
(255/89007-0!A-S)
1. lOCFRSO, Appendix B, Criterion
III, as implemented
by the Palisades
Operations
Quality Assurance
Program requires, in part, that the design bases be correctly
translated
into
drawings, procedures, and instructions;
that the design control measures provide for verifying
or checking.the
adequacy of the design; and that design control measures be applied to the delineation
of acceptance
criteria for inspections
and tests. Contrary to the above, the following
instances
of inadequate
design control were identified:
This is a Severity Level IV Violation.
This violation
is sustained
by 19 examples.
Though Consumers
Power Company believes four of these are not supportive
examples.
We do acknowledge
the violation.
Our detailed response to each example follows: MI0789-1683A-TC01-NL02
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NRG Violation
255/89007-0la:
EA-FC-789-07, "Seismic Analysis of Auxiliary
Feedw'ater
Control ESSR 88714, 11 Revision l, August 24, 1988. [Refer to page 9 of NRG Report 50-255/89007 (DRS).] Example FC-789 contained
multiple dimensional
differences
between the analysis model and the installation
drawings.
The following
examples are provided:
-The location of new support 8224 was analyzed at 6 11 from the 45° elbow. The piping drawing (M-101 Sheet 5113) *used to install the support specified
a dimension
of l'-7 1/2" from the elbow. This difference
was not noted in the calculation.
-The length of pipe between Model Nodes 6276 and 6282 was analyzed as 5'-10" long. The installation
drawing specifies
S'-6" long. This difference
was not noted in-the calculation.
Several additional -dimensional .discrepancies
on the. new. bypass piping were . also noted between the analysis and installation
drawing. These discrepancies
ranged from 1 11 to 2-1/4" and were considered
minor by the inspector.
none of these discrepancies
were noted in the calculation.
Reason for Violation
During the evaluation*
of the design of the bypass piping system numerous changes in design dimensions
were encountered
due--to pipe, support and valve operator-interferences.
At a certain point in the analysis process, it was decided to build* the design *to. the drawing and* effect the final analysis reconciliation*when
the as-built data were recorded on a marked-up
drawing. The analysis reconciliation
with the as-built was never made. This violation
was due' to inadequate
documentation
of the* justification
for analytical
input and failure to follow established
procedures.
Corrective
Action Taken, and** Results Achieved-All engineering
groups have been briefed as to the results of this inspection.
These briefings
were completed
on August 2, 1989. The above noted
cies have been satisfactorily
dispositioned
and the finite element piping analysis model has been updated. Corrective
Actions to be Taken to Avoid Further Non Compliance
Interim All design change engineers
will be briefed as to the reported violations
by personal letter. These letters will require that all engineers
involved in design changes scheduled
for installation
in 1989 review existing design ages for similar problems and correct any identified
problems.
MI0789-1683A-TC01-NL02
2
... * ... *. -*. : .r;. : ...
I
.-:*** , .. ,,. .. Long-Term
Enhancements
will be made to plant administrative
design control procedures
to further clarify the requirements
that strict alignment
between engineering
analyses, associated/accompanying
drawings, and as-built condition
must be verified and documented
prior to declaring
modified systems/equipment
In additionf
a program will be developed
to provide periodic refresher
training to all design change engineers
on design change-related
administrative dures. Date When Full Compliance
Will be Achieved The personal briefings
by letter will be issued by September
1, 1989. dural enhancements
will be completed
by January 1 1 1990. The program for periodic training will be in place by March 1, 1990. NRC Violation
255/89007-0lb:
EA-FC-789-07, "Seismic Analysis of Auxiliary
Control ESSR 88714" Example b.l -For the south bypass loop, the Young's Modulus was specified
as 27.4 E6 psi instead of 27.9 E6 psi. This is equivalent
to analyzing
this portion of pipe with properties
at 300° instead of 70°. This discrepancy
was not noted in the analysis.
Reason for Violation
The use of the.27.4 E6 psi value for the Young's Modulus represents
a 1.8 cent error with *regard* to the correct value of 27 .9 E6* psi value. The impact of such an error is expected to be an underprediction
of thermal expansion
stress of no more than 1.8 percent. This resulted from inadequate
tion of technical
review and failure to follow existing procedures.
Corrective
Action Taken. and. Results= Achieved*
All engineering
groups have been* briefed as to the results of the inspection.
These briefings
were completed
on August 2, 1989. Corrective
Actions to be Taken to Avoid Further Non Compliance
Interim All design change engineers
will be briefed as to the reported violations
by personal letter. These letters will require that all engineers
involved in design changes scheduled
for installation
in 1989 review existing design ages for similar problems and correct the problems.
MI0789-1683A-TC01-NL02
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- ** Long-Term
to plant administrative
design control procedures
will be made tog -Provide the technical
reviewer a review checklist
with a "prompt" to justify the numerical
values of all constants
and variables
utilized as inputs to the analysis (the checklist
will provide a comprehensive
set of "prompts" to ensure an overall accurate, thorough and auditable
analysis). -A mechanism
for the reviewer to note minor errors which would not necessitate
a reanalysis.
In addition, a program will be developed
to provide periodic refresher
training to all design change engineers
on design change-related
administrative dures. Date When Full Compliance
Will be Achieved The personal briefings
by letter will be issued by September
1, 1989e The procedural
enhancements
and training on the enhancements
will be completed
by January 1, 1990. The program for periodic training will be in place by March 1990. Example b.2 The location of the center of gravity (CG) for the new bypass valves was analyzed at 19 11 from the pipe centerline.
The location specified
on the vendor-drawing
was 22 11* This represents
a 15% increase in the moment arm which was not noted in the calculationo
Reason for Violation
The piping analysis was set up from preliminary
data. The valve assembly weight was included in the model. However, the weight placement
was not sistent with-*the
final drawing received.from
the vendor. The existing mentation
does not indicate whether or not the analyst reviewed the center of gravity data from the vendor drawing. The analysis certainly
was not run to accommodate This violation
occurred due to failure to account for vendor information
as analytical
input and failure to follow established
procedures.
Corrective
Action Taken and Results Achieved All engineering
groups have been briefed as to the results of the inspection.
These briefings
were completed
on.August
2, 1989. The calculation
was revised to incorporate
the correct vendor data and was found to be acceptable.
Corrective
Actions to be Taken to Avoid Further Non Compliance
Interim Same as that required for Violation
l.a. MI0789-1683A-TC01-NL02
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. . * Long-Term
Enhancements
to plant procedures
will be made Ensure that vendor information/recolillllendations
are accounted
for ical input and that justification
be provided for departure
from information/recommendations, as such Provide the technical
reviewer a review checklist
with a 11 prompt" to assure that vendor information/recommendations
are appropriately
accounted
for. A program will be developed
to provide periodic refresher
training to all design engineers
on design change-related
plant administrative
procedures.
A 11 punch 11 list or equivalent
will be developed
to track items requiring
verification
when data becomes available.
Date When Full Compliance
Will be Achieved The personal briefings
by letter will be issued by September
1, 1989. These procedural
enhancements
will be in-place by January 1, 1990 as will required training on these enhancements.
The program to provide refresher
training will be in place by March 1, 1990. Example b.3 In addition to the above noted discrepancies
for modeling the bypass piping, other dis.crepancies
were noted in the model of the original auxiliary
piping. The inspector
could not determine
whether these discrepancies
were inherent in the original data or whether they occurred during the transcription
of the original model into the current piping analysis.
However, notes in the piping model stated the following: "Bechtel analysis is a bit off from ISO here." -"Bechtel has modeled elbows only with SIFs. Elbows are used here." -"Review ISO for pipe schedule change." These notes led the inspector
to question the validity of the assumption
made in the calculation
concerning
the correctness
of the original input data. CPCo Response The three notes recorded by the inspector
do not necessarily
imply errors in the original input analysis.
The notes reflect free text written into the ADLPIPE computer model by the translator
of the ME101 Bechtel model for the review by the piping analyst. The specific analysis model/ISO
discrepancy
was small. However, the note advised the analyst that a choice needed to be made for analysis record runs. MI0789-1683A-TC01-NL02
5 . :.'.-.** -
- There is nothing wrong with modeling elbows with SIFs and flexibility
characteristics.
However, the note merely advises the analyst that comparing
ADLPIPE elbows and ME101 elbows for counting of elbows for model benchmarking
will not yield consistent
results and that the MElOl model will require more review to ensure model consistency.
The note with respect to pipe schedule change is again for the benefit of the analyst. No error is implied. No corrective
action is required.
Example b.4 The additional
discrepancies
in the mod*el of the auxiliary
piping were as follows: -For flow element FE-0736, the weight of 192 lbs was modeled at node 211 instead of node 205 *. Although this was only a 4-1/2" error on a 6 11 pipe, the flange pair was analytically
modeled with the weight concentrated
at one edge instead of at the middle of the flanges. For Valve M0-0754, the 460 lb weight was modeled at the centerline
of the pipe at node 267. The weight should have been specified
at the valve CG at node 268, 18" out from the pipe centerline.
The horizontal
response spectra used in the analysis was inconsistent
with the spectra given in Specification
C-175. The spectra used was lower and not as broad as those given in the Specification.
-Piping .between the nodes 252 and 253 was modeled as 4", schedule 40, instead of 6 11 , schedule 80. The above discrepancies
are further examples of violation
of 10 CFR 50, Appendix B, Criterion
III in that the licensee failed to correctly
translate
the design into the drawing (255/89007-0lb).
Reason for Violation
The placement
of the flow element weight, the placement
of the valve operator weight and the pipe schedule discrepancy
constitute
discrepancies
which should be picked up in the review process. The reason for the violation
has been attributed
to an inadequate
technical
review and failure to follow established
procedures.
The horizontal
response spectra employed in the original IE Bulletin 79-14 analysis of the Palisades
piping systems were based upon the Taft 1952 record. The digit,ized
data and a straight-edged
set of plots from those data were
.. to Consumers
Power Company by Bechtel in 1976. The horizontal
response spectra used in the piping analysis were derived from these digitized
data. The straight-edged
plots were used for building and equipment
tion seismic work * MI0789-1683A-TC01-NL02
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Because the straight-edged
plots were very difficult
to read and because it was desired to incorporate
building and equipment
spectra in a single seismic ification
specification, the straight-edged
plots were redrawn and incorporated
into Specification
C-175. It is expected that the horizontal
spectra of C-175 could be slightly higher and broader than the straight-edged
spectra. However, that was not the purpose for drawing them. Although the C-175 horizontal tra should be very similar to the straight-edged
horizontal
they should be used for building analysis and equipment
qualification
only. They should not be used for piping analysis.
The correct horizontal
response spectra for safety related piping systems at Palisades
which use the initial plant seismic design basis are those included in the stress packages as developed
from the digitized
spectra._
New piping systems or modifications
involving
substantial
changes to existing systems will employ the spectra and procedures
in Specification
M-195. Corrective
Action Taken and Results Achieved All engineering
groups have been briefed as to the results of the inspection.
These briefings
were completed
on August 2, 1989. Corrective
Actions to be Taken to Avoid Further Non Compliance
Interim Same as for Violation
Item l.a. Long-Term
Enhancements
to plant procedures
will be made to: -Provid*e the technical
reviewer a checklist
with a comprehensive
set of "prompts" to ensure an overall accurate, thorough and auditable
analysis.
These "prompts" will specifically
require that the reviewer check the validity of all analytical
input and assumptions.
-Provide. the basis for the selection
of design
as governing, and -Provide a technical
review checklist
with.a prompt to concur that governing
design criteria (input) have been justifiably
selected.
-Identify applications
in which C-175 or M-195 would be used. Furthermore, a program witl be developed
to provide periodic refresher
training to engineering
personnel
on design change related plant administrative dures. Date When Full Compliance
Will be Achieved The personal briefings
by letter will be issued by September
1, 1989. dural enhancements
will be made by January 1, 1990 as will all required training on the enhancements.
MI0789-1683A-TC01-NL02
7 ::*:-";**
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.! * NRC Violation
255/89007.0lc:
Consumers
Power Company Drawing M-101 Sheet 5113, Revision O, "Piping Isometric, Auxiliary
Control Valve CV-0736A and CV-0737A Bypass Piping." [Refer to page 12 of NRC Report 50-255/89007 (DRS)o] Example -The size of the fillet weld was determined
by the requirements
of Welding Specification
WPS-11.21, Revision 2; however, for the socket welded fittings, the size of the fillet weld was not specified
on this drawingo In reviewing
the Repair Inspection
Checklist (RIC) for the welds in question, the weld size specified
is 1 1/2 11* This is misleading
in that this is the size of the pipe and not the size of the fillet weld. In order for the welder to determine
the size of the fillet weld, the pipe wall thickness
must be obtained and a calculation
of 1.09 times the wall thickness
must be per-. formed. Although this is a relatively
simple calculation, it is a design function and* as such must be controlled.
There is no documentation
to demonstrate
that this design activity was performed.
In addition, there are *no controls in place to check and verify this design activity.
Reason for Violation
Specifying
welding requirements (such as applicable
code, weld material, weld type and weld size) is an engineering
function.
If properly administered
by procedure, the maintenance
planner can (and has) effectively
prescribe
welding details* for the field provided that adequate input from engineering
exists as a basis. In the past, engineering
input has been limited to welding
tion and/-0r structural
analysis engineering
sketches.
which have lacked size dimensions
for the welds *. As a result, the planner has failed to provide the proper size on the Repair Inspection
Checklist (RIC) thereby requiring
the field welder to determine
and install the proper weld size. This practice fails to meet current expectations
for control of design change implementation.
The plant administrative
design control procedures
required and currently
require that the design change project engineer determine
code requirements
for assigned projects (Reference
4), and plant maintenance
procedures
required and currently
require that the maintenance
planner specify applicable
code and weld parameters
after consultation
with the Engineering
Department (Reference
3). These procedures
have not been effectively
integrated
to support one another to ensure that weld specifications
from engineering
were accurately
translated
into installation
planning, installation, and post-installation
The following
actions have been/will
be taken to ensure the administrative dures relating to weld specifications
are properly integrated
with the Maintenance
Department.
Prior to actions taken as a
of recent self-identified
failures to verify weld size (Reference
7), no specific requirements
existed to verify characteristics (weld, type, size contour) of installed
welds. Although Nuclear Operations
Department
Standards
suggest inspection
hold points for weld installation
verification, working level administrative
procedures
did not specify a hold point requirement
except for fit up. MI0789-1683A-TC01-NL02
8 .. '. **. _.\**.**.*.:
- ... . *.* ... _;**-::* .
- .* -.,:* * .*.... -*. *;.-* .
- * ---------Corrective
Action Taken and Results Achieved -All engineering
groups have been briefed as to the results of this inspectiono
The briefings
were completed
on August 2, 1989. The Inservice
Inspection (ISI) Section of the Plant Projects Engineering
Department
has effected the role of Design Authority
for weld engineering
by revising the RIC to identify critical weld parameters
and require ISI cal review of the maintenance
planner's
specifications.
The purpose of the review is to ensure that appropriate
welding codes are complied with in the areas of weld installation
and post-installation
examination.
Revision to the RIC was completed
as part of the revision to the plant administrative
procedure
for control of special processes (Reference
3). -The ISI Section (as well as planners, welders and welding supervisors)
has received specific training with respect to welding codes and technology
to augment their existing collective
knowledge.
-In addition, the RIC .. was revised to issue. the. weld minimum leg length to the field. This will eliminate
the need for the field welder to calculate
the length. The aforemenqoned
ISI review will assure that this specification
is provided.
-Finally, the RIC has been revised to require verification
of weld size * (RIC now requires that weld is inspected
for size, porosity, undercut,-etc.)
Training materials
for the welder tra1n1ng progression
course have been revised-to emphasize
fillet weld terminology
and conformance
of the completed
weld to the.design
specification.
Corrective
Actions to be Taken to Avoid Further Non Compliance
Interim -* Same as that required for Violation
Item l.a. Long-Term
-Enhancements
to plant design control and maintenance
procedures
will be made to more effectively
integrate
engineering
into weld specification
and mately into weld planning and verification:
Appropriate
welding codes will be included in the Design Input Checklist (Reference
2) to "prompt" the design engineer to specify appropriate
weld requirements (for installation
and examination)
in the facility change package as part of both conceptual
and detailed engineering.
In addition, a generic guideline
will be developed
to support the design engineer throughout
the weld design process. MI0789-1683A-TC01-NL02
9 . :*.-...
....... :. .\ :*_ .. *** '!: *-..... : .: . ' ,\ .. ,.' **::.**.*:-:
...
- * Design control procedures
related to engineering
analyses (Reference
1) will explicitly
require that all drawings accompanying
structural/seismic
analyses provide detailed weld information (type, size, material)
for input to the planner. The procedures
will also require that sizing calculations
be* performed
as part of the analysis.
Finally, a technical
review checklist
will be provided to require that the reviewer ensures that weld information
be accurately
represented
on the analysis drawings.
-Plant maintenance
procedures (Reference
3) will require that the maintenance
planner utilize the contents of the facility change package to complete the RIC in specifying
for the field weld installation
and examination ments. The procedure
will require that the planner consult the Design Input Checklist
and structural/seismic
engineering
analyses.
Relative to weld verification, the design control program and related welding program will be evaluated
and enhancements
developed
as necessary
to ensure that administrative
and quality verification
controls exist to consistently
verify that field installation
satisfies
design requirements (ie, input vs output). Interim actions related to changes to the RIC and !SI group review of the RIC (as described
above) will remain in effect * Design and quality assurance
engineers
will be trained on the appropriate
structural
and piping weld codes and their application
to weld installation
and examination.
The engineers
will also be trained on the above procedural
enhancements.
Finally, a program will be developed
to periodically
train design and quality assurance
engineers
on the aforementioned
codes and their application, and on the weld-related
design control and maintenance
procedures.
In summary, it is expected that these actions will ensure that proper welding requirements (type, material, size) are specified
by engineering, planned by maintenance (with a check on planning by engineering), and in turn, verified by quality control. Date When Full Compliance
Will be Achieved The engineering
group briefing has been
The personal briefings
by letter will be issued by September
1, 1989. Procedure
enhancements
and required training on the enhancements
will be completed
by January 1, 1990. The program for periodic refresher
training will be developed
by March 1, 1990. NRC Violation
255/89007-0ld:
EA-T-FC-722-501-01 "Calculation
of Acceptance
Criteria for Modification
Test Procedure
T-FC-722-501," January 13, 1987. [Refer to page 16 of NRC Report 50-255/89007(DRS).]
MI0789-1683A-TC01-NL02
10 *:**.
.* ..
.*. .. . ; . .
- * Example The calc.ulation
on page 2 of the engineering
analysis states that the total volume of gas contained
in the nitrogen bottles at 2000 psig is 209 scf. This value is incorrect
in that it is the usable cylinder volume as given in lation EA-FC-722-02.
The actual volume is approximately
228 scf. By using the incorrect
value, the calculated
acceptance
criteria for pressure drops were higher and, therefore, were nonconservative.
CPCo Response CPCo does not acknowledge
this example as a of violation
of 10CFR50, Appendix B, Criterion
III for the following
reasons. 1. As indicated
by EA-FC-722-02, the design intent of this modification
is to supply a nitr.ogen
header pressure from an initial minimum bottle pressure of 2,000 psig down to 150 psig to ensure that the associated
control valves would be brought to their safety-related
position and maintained
in that position for the -required
time period. * 2. In accordance
with the design intent of this modification, the usable volume of nitrogen is that volume contained
in the bottle from 2,000 psig to 150 psig or 209 scf as calculated
by EA-FC-722-02, Sheet 10 of 13. The usable volume of 209 scf is utilized as a conservative
value to establish
the number of nitrogen bottles required for each station to meet system design requirement.
3. Although not specifically
stated in *the body of EA-T-FC-722-501-01, the value of the "usable" volume of nitrogen (209 scf) was utilized in lishing test acceptance
criteria rather than the "total" volume of nitrogen (228 scf) to confirm the design intent, verify estimated
leakage rates, and confirm system margins. The test procedure
clearly tests the design intent of this modification.
Based up_on the above, we feel that this example does not support a violation
of lOCFRSO, Appendix B, Criterion
III has occurred.
However, certain actions will be undertaken
to remedy this minor deficiency
and prevent its recurrence:
Interim -All design change engineers
will be briefed as to the reported violation
by personal letter and by engineering
group presentation.
The letter briefings
will be completed
by September
1, 1989. The group presentations
were pleted on August 2, 1989. -EA-T-FC-722-Ji
will be revised to clearly indicate that "useable" volume has been utilized to calculate
the acceptance
criteria rather than "total" volume. MI0789-1683A-TC01-NL02
11 . **-.*****:*:
--:*;*-: ... :**:*,'. ... . . .. :-. ... ;-: . .;. :: ;*.* .. ...... _ ...... . ':. :*;*.-**
Long-Term
The actions identified
as being taken in the interim are considered
complete and effective
in responding
to this identified
condition;
no further action is required.
Date When Full Compliance
Will be Achieved The engineering
analysis will be revised by September
1, 1989. NRC Violation
255/89007-0le:
FC-756 11 HPSI Pump Miniflow Bypass Modification.
19 [Refer to page 18 of NRC Report 50-255/89007 (DRS).] Example Input into the AOLPIPE, Inc (AOL) piping stress analysis, contained
in FC-756, contained
multiple dimensional
differences
from the as-built dimensions.
Bechtel's
stress.isolmetric
drawing 03378, sheet 4 of 5, Revision 1, and drawing
Revision 4, showed a dimension
of 29 7/8 inches between pump 66A and the elbow. The as-built dimension
is 13 1/2 inches. Both (ADLPIPE, Inc.) AOL's and B.echtel's
stress analyses used 27 7/B inches. This dimensional
discrepancy
was not documented
during the NRC IEB 79-14 program, nor was it corrected
in Bechtel's
and AOL's stress analyses.
Further, this discrepancy
is in conflict with the assumptions
contained
in analysis No CS-ESSR 87-144 that purportedly
demonstrated
that the Bechtel drawings are correct. The inspector
also noted that the input data used in the modification
portion of the piping system was inconsistent
with as-built drawing No 03378, Sheet 4 of 5, Revision 2. The licensee reviewer was not aware of the above dimensional
discrepancies.
Failure to correctly
translate
the design into the drawings is considered
an example of violation
of 10CFR50, Criterion
III. Reason for Violation
The dimensional
discrepancy
associated
with the 27 7/8 versus 13-1/2 inch lengths was a result of the analyst relying on data being transmitted
from the field and not checking the installation
personally.
The smaller discrepancies
between the ADL and as-built drawing records were recognized
by the analyst when he was provided a marked-up
drawing of the as-built configuration.
The analyst acknowledged
receipt of the as-builts
via memo and stated that the as-built configuration
was acceptable
and no reanalysis
was required.
The reason for the violation
was inadequate
analytical
assumption
resulting
from a failure to perform a system walkdown and failure to follow established dures. Corrective
Action Taken and Results Achieved All engineering
groups were briefed on the results of this inspection.
The briefings
were completed
on August 2, 1989. The dimensional
discrepancies
noted have been satisfactoril*y
dispositioned
and documented.
MI0789-1683A-TC01-NL02
12 *::**. -.* ...... .
... :* ..... **', ..
- Corrective
Actions to be Taken to Avoid Further Non Compliance
The following
corrective
actions will be taken to prevent
Interim Same as that required for Violation
Item 1.a. Long Term Procedural
enhancements
will be made to ensure
-The analyst "walks down" the area of interest *to confirm all as-built (or intended as-built)
data is utilized in the analysis.
This confirmation
must be made prior to declaring
modified structures
or equipment
-By approval of the facility change "Responsible
Engineer, 11 the above bility for as-built data confirmation
may be delegated
to field construction
by controlled
procedure
or work order instruction.
-In the event the analyst concludes
that no further "analysis" is necessary, the reconciliation
of such shall be documented
as part of a controlled
analysis revision which ensures technical
review. A program will be developed
- to provide refresher
training on design change related prQcedures.
This training will be directed towards all design change engineers.
_ Finally, a portion of the Configuration
Control Projec.t involves the walkdown and field verification
of piping as-built dimensions
to confirm the accuracy of our stress isometric
drawings.
Verification
of the stress isometric ings for a sample system is planned for 1990 to assess theneed and extent of further verification
activities.
CPCo will perform any required walkdowns
by no later than the 1990 refueling
outage. Date When Full Compliance
Will be Achieved Personal briefings
by letter will be issued* by September
1, 1989. Procedural
enhancements
and required training on the enhancements
will be completed
by January 1, 1990. The periodic training program will be in place by March 1, 1990. Walkdown and field verification
of stress isometric
drawings requiring
verification
will be completed
by the 1990 refueling
outage. ' NRC Vio*lation
255/89007-0lf:
FC-756 "HPSI Pump Miniflow Bypass Modification." [Refer to page 19 of NRC Report 50-255/89007 (DRS).] Example The as-built sketch used in the analysis for FC-756 contained
a nine inch dimensional
error. MI0789-1683A-TC01-NL02.
13 *:.:*:: ** . *' ... . i.: .*
_ .. * . ,, .. *-* :-*._:: ... ; . . ' . : :: ... :. ' ........
The as-built sketch for the modification
near pump 66A was sent from the site to the engineering
office for review. The inspector
noted that this sketch contained.a
dimensional
error. the 2 1-6 1/2" dimension
was incorrectly
marked on the sketch. This dimension
was off by nine inches. Failure to correctly
translate
the design into the drawing is considered
an example of violation
of lOCFRSOP Appendix B, Criterion
III. Reason for Violation
As a result of required piping changes for this modification, a seismic analysis and Stress Package 03378 update were requested
by the site. Included with the request were M-107 Sh 2247/2248
which indicated
the existing configuration, and proposed modification.
Using the drawings as input 1 the system was modeled on ADLPIPE to generate the system stresses after the modification.
The existing drawings (sent as part of the request) were marked "Issued As-Built per NRC IE Bulletin 79-14.11 After the analysis was performed, a pre-installation
walkdown was performed.
During the walkdown the referenced
dimensional
discrepancy
was noted. The seismic analyst was contacted
to evaluate the change. As a resultp the analyst issued a letter stating *that since stresses in the area were low, based on his judgement, the change was acceptable.
When the construction
was complete, the seismic analyst compared the as-built to the dimensions
used in the preliminary
analysis.
It was determined
the analysis was acceptable
with the dimensional
variance .... Stress Package 03378 was annotated
to reflect this information.
The above-information
describes*
the circumstances
surrounding.the
modification
however does not indicate a root cause. The discrepancy
is not directly related to the modification
except that the modification
brought a previous error to light. That is, the drawings used were certified
as being dimensionally
correct per Bulletin 79-14, when in reality there was an error. Corrective
Action Taken and Results Achieved The engineering
groups were briefed as to the inspection
results. These ings were completed
on August 2, 1989. The above noted discrepancy
has been satisfactorily
- dispositioned
by analysis.
Corrective
Actions to be Taken to Avoid Further Non Compliance
The. following
corrective
actions will be taken to prevent recurrence:
Interim Same as that required for Violation
Icem l.a. Long-Term
- The "long-term" actions prescribed
for Violation
Item l .e will prevent rence. MI0789-1683A-TC01-NL02
14 .
..... ,, --.
- . -
- Date When Full Compliance
Will be Achieved The dates established
for.actions
related to Violation
Item l.e apply here as well. NRC Violation
255/89007-0lg:
FC-756 "HPSI Pump Miniflow Bypass Modification.eu
[Refer to page 19 of NRC Report 50-255/89007 (DRS).] Example Pipe support drawings in p1p1ng support Calculation
No 03378 of FC-756 did not adequately
describe the required weld sizes. Pipe support drawings DCl-8198.1
and DC1-Hl96.2
contained
in support tion No 03378 were reviewed.
The inspector
found that one drawing showed fillet welds at the structural
joints but no weld sizes were specified.
The other drawing showed a 3/16 inch fillet weld with a note "assumed." As a result, the design bases of the welds were not adequately
translated
into the drawings.
CPCo Response As part of the evaluation
of this example, M-107 Sh 2254/2255
were reviewed which are detail drawings for the subject hangers. The two ports *cited were not modified or installed
as part of FC-756. The supports were only evalua.ted
regarding
stresses in relation to the modification.
In both cases, the_drawings
are Rev 0 and are issued as-built per IE Bulletin 79-14. It appear-s that this is a situation
where documentation
from the 79-14 effort may not be completely
However, when past discrepancies
were identified, there was no signficant
impact on analytical
conclusion.
Neither drawing DC1-H198.l
nor DC2-Hl96.2
were utilized as design input to FC-756. After further discussion
on this issue with NRC Region III via telecon on July 26, 1989 and review of the drawings referenced
by the inspector, it was determined
that these drawings were initial IEB 79-14 calculation
file ings of preliminary
status. These drawings do not represent
the final hanger detail drawings referenced
above. Since these calculation
file drawings are not "record" drawings reflecting
as-built condition, and are not referenced (by intent) in our Facility Change Design Document Checklist, they are not input to our facility change process. No further action is required since neither a design change control deficiency
nor inaccurate
record (as-built)
document exists. Therefore, CPCo does not acknowledge
this example. However, reference
example e. for actions to be taken to ensure accurate dimensions
are utilized as* analysis inputs. NRC Violation
255/87007-0lh:
FC-731 "Regulatory
Guide 1.97 Transmitter
Replacement." [Refer to pages 19 and 20 of NRC Report 50-255/89007 (DRS).] Example The seismic stress calculation
assumed an incorrect
center of gravity which was not identified
during the checking process.
15 . *.*:.,*-*
-.-.:*-**
- .** '* .:: *:_.: '! ... ,' . _..,,-. '* .* .*.,** .:*::.'.-
- The analysis criteria shown on page 3 required the center of gravity (CG) of the instruments/equipment
to be considered
in the seismic stress calculationso
A review of the rack support bent plate on page 27 found that the CG of the instruments
was not considered
in the seismic stress calculations.
As a
the forces and moments at the rack support attachment
were inadequately lated. Reason for Violation
The analysis addresses
the adequacy of instrument
racks inside the containment
building.
For the GWO 7906, FC-731 job, the work involved modifying
all four instrument
racks. Three of the racks are tied together while the fourth one is by itself. The racks are made out of Unistrut attaching
to the containment
liner plate using bent plates. The instruments
are mounted on the mounting plate which in turn is* bolted to the Unistrut.
Analytical
error based on the failure to consider the center of gravity is acknowledged.
The reason for this
is an error made by the analyst, inadequate
technical
review and.failure
to follow established
procedures.
Corrective
Action Taken and Results Achieved and the The analysis has been revised to include the center of gravity analytical
results represent
an acceptable
as-built condition.
groups have been briefed as to the results of this* inspection.
were completed
on August 2, 1989. All engineering
These briefings
Corrective,Actions
to be Taken to Avoid.Further
Non Compliance
To prevent recurrence
of this or similar discrepancies, the following
corrective
actions will be taken: Interim Same* as* that required for Violation
Item* La. Long-Term
The Plant Administrative
Procedure
will be enhanced by the incorporation
of a technical
review checklist
consisting
of a comprehensive
set of review "prompts." One of the "prompts" will require that the reviewer ensure that all analysis objectives
be carried through to completion.
' In addition, a program will be developed
to provide periodic refresher
training to all design engineers
on design change-related
administrative
procedures.
Date When Full Compliance
Will be Achieved The personal briefings
letter will be issued by September
1, 1989. Procedural
enhancements, as well as required training on the enhancements, will be pleted by January 1, 1990. The program for periodic refresher
training will be in place by March 1, 1990. MI0789-1683A-TC01-NL02
16 *.:.
- NRC Violation
255 /89007-0li:
FC-731 "Regulatory
Guide 1. 97 Transmitter
Replacement." [Refer to page 20 of NRC Report 50-255/89007 (DRS).] Example The calculated
bending stress "fbx" shown on page 27 of the analysis was in error. The 5,645 psi should be 5,976 psi. The checker did not identify this calculational
error. Reason for Violation
Analytical
error based on the inaccurate
bending stress is acknowledged.
The analysis has been revised to incorporate
the accurate "fbx" value and the analytical
results represent
an acceptable
as-built condition.
Corrective
Action Taken and Results Achieved All engineering
groups have been briefed as to the results of this inspection.
These briefings
were completed
on August 2, 1989. Corrective
Actions to be Taken to Avoid Further Non Compliance
To prevent recurrence
of this or similar discrepancies, the following
corrective
actions will be taken: Interim Same as that required for Violation
Item La. Long-Term*
Same as that required for Violation
Item l.h with the exception
that a "prompt" will be included on the technical
review checklist
to require that the reviewer verify the accuracy of all analysis calculations.
Date When Full Compliance
Will be Achieved The dates specified
for Violation
Item l.h apply to this item also. NRC Violation
255/89007-0lj:
FC-567 "Core Cooling Instrumentation
Modification." [Refer to page 22 of NRC Report 50-255/89007 (DRS).] Example FC-567 did not address the impact of the increased
load on the inverters, bypass regulators
on the battery chargers.
The inspector
observed that the licensee performed
calculations
to analyze the impact of the increased
loading on the preferred
AC bus supply breakers, cabling to the preferred
busses from their respective
inverters
and on the DC batteries.
However, no calculations
or analyses were evident which addressed
MI0789-1683A-TC01-NL02
17 . . :_ . =** *; ..... ... : . . .. **. ::: :.
. :.' ..... -. the impact on the inverters, bypass regulator
or the DC system battery chargers.
This resulted in a concern for the capability
and capacity of these Class lE systems to perform their safety-related
functions.
The inspector
concluded
that the licensee had failed to employ adequate design controls during the design stage of the facility change in that the full impact of the increased
loading was not analyzed.
In response to the inspector's cern, the licensee verified the present loading on the respective
inverters
and battery chargers which includes the increase resulting
from the instrumentation
additions.
The inspector
concurs that based on the licensee's
reported inverter and battery charger outputs, plus the anticipated
emergency
loading, per the Design Basis document, the inverters, bypass regulator
and battery chargers will not be overloaded.
However, the licensee failed to employ adequate design controls which would have included analyses of all impacted components.
Reason for Violation
Facility Change FC-567 (Core Cooling Instrumentation)
added a Reactor Vessel Level Monitoring
System (RVLMS) to the plant design. Addition of this system resulted in an increased
load of 600VA on each of preferred
busses, YlO and Y20, the associated
DC to AC inverters, bypass regulator
and DC system. In reviewing
this design change, the inspector
identified
that, although the effect of the increased
load on the batteries
was determined, the facility change did .not. address the impact of the increased
load on the inverters, bypass regulator
or the battery chargers * . * . The apparent failure to adequately
verify and check design resulted from inadequate
documentation
of assumptions
and engineering
judgement
utilized to determine
the impact of the load additions
to the preferred
busses. The effect of the load increase on the batteries
was determined
based on the undocumented
assumption
that the batteries
were the limiting component.
In order to mine the effect of the increased
load on the batteries, the new loading on each of the preferred
buses and thus the loading on each of the inverters
was determined.
No documentation
was provided, however, comparing
the revised load on the invertors
against their design. rating. A similar situation
existed for the battery chargers.
The new battery load profile was determined
based on the increased
loads, however, no documentation
of the effect of the new load profile on the battery charges was provided.
Subsequent
evaluations
have been performed
to document that the load additions
to the preferred
buses performed
by FC-567 did not result in overloading inverter, battery charger or bypass regulator.
The results of these evaluations
are summarized
below: 1. The maximum loadings on the YlO and Y20 buses during emergency
conditions
are 4378VA and 5456VA respectively.
This includes the loads added by FC-567. The design rating of the invertors
is 6000VA and thus the tors are not overloaded.
MI0789-1683A-TC01-NL02
18 ., .... .._: .*
- *. * .....
.. *.-.,: .. *;-*: **.***.***
.. ** .. *. *.".: '*':.':-__ ._ .. _* . ....... *\ '.: .. ;.*-o*.** . ,, . : .... **_ cl I .!
. ' * -* _._ .. _. . ... ..... * ... 2. The steady state constant DC current requirements
during emergency tions is 253 amps for the most heavily loaded battery (Battery No 2) after approximately
ten minutes. This is less than the 400 amp combined rating of the two battery chargers connected
to each DC bus. The battery chargers thus have sufficient
capacity to provide the DC steady state load with capacity remaining
for restoration
of the batteries
following
the discharge
during the first ten minutes. 3. The bypass regulator
is utilized to provide temporary
power to a preferred
bus from a non-class
lE source to allow maintenance
to be performed
on an inverter.
The initial response made to the inspector
regarding
operation
of the bypass regulator
was incorrect.
The bypass regulator
is not shed during accident conditions
and could be subject to the emergency
load. Operation
with the bypass regulator
energizing
the preferred
buses is, however, restricted
by Administrative
Procedures
to less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (eight hours for some buses). This restriction
minimizes
the amount of time that the bypass regulator
would be subject to providing
power to a preferred
bus during accident conditions.
The limiting component
of the bypass regulator
is the isolation
transformer*
This transformer
is rated at 5000VA. As discussed
earlier, the maximum loading on
bus Y20 is 5456VA. Thus the load on the bypass regulator
could be exceeded if it were connected
to bus Y20 during an emergency
condition.
This discrepancy
had been previously
identified
by the Configuration
Control Project and Discrepancy
Report F-CG-88-002
was initiated.
This discrepancy
was quently closed out by assuring that the output voltage of the bypass regulator
will be maintained
at acceptable
levels at up to 150% of the nameplate
rating of the tr...an*sformer.
Corrective
Action Taken and Results Achieved All engineering
groups havebeen briefed on the results of this inspection.
These briefings
were completed
on August 2, 1989. -An engineering
analysis was per.formed
documenting
that the inverter and * battery charger were not overloaded
as a result of this modification.
-The Configuration
Control Project had.previously
identified
the concern with the bypass regulator
and has subseq'uently
resolved and closed out the crepancy.
Corrective
Actions to be Taken to Avoid Further Non Compliance
To prevent recurrence
of this or similar discrepancies, *the following
corrective
actions have or will be taken: Interim Same as that required for Violation
Item l.a * MI0789-1683A-TC01-NL02
19 *.*:-: * .. " .-* *,::: >*'. -:-,,. .. ........ **. *'* ----
- * Long Term Upgrades have been initiated
to our station load analysis program to account for full aystem impact of load additions.
In the
the load carry1ng
ability of load carrying components
will be assessed in addition to assessing
power supplies.
Specifically, the load carrying capability
of the battery chargers and preferred
power inverters
will be assessed, along with battery capacity whenever load is added to the 120V preferred
AC system. Periodic training as proposed for Violation
Item l.a will feature the ities of modifications
support groups such as: Power Resources
and Systems Planning (for load addition
and -Systems Protection
and Planning (for breaker
and -Energy Supply Services Civil Section (for structural
analyses).
It is expected that this training wil-1 maintain the design engineer's
awareness
as to what must be taken into account when adding electrical
or mechanical
load to plant systems. Date When Full Compliance
Will be Achieved Personal briefings
letter will be issued by September
1, 1989. The station load analysis program upgrades will be completed
by September
1, 1989. A gram for the periodic training on the capabilities
of support groups will be in place by -March 1, 1990. NRC Violation
25S/89007-0lk:
FC-760-02 "Control Room Emergency
Lighting." [Refer to pages 23.and 24 of NRC Report 50-255/89007 (DRS).] Example This FCcontained
an unverified
assumption
in that the assumption
that emergency
lighting fixtures were rigit was never proven. Engineering
Analysis EA-FC-760-2-001
was performed
to analyze the mounting of the lighting fixtures to be installed.
Section V of this document, referring
to the DC lighting fixtures, states in part "Assume the lighting fixture is rigid **** " This assumption
is not justified
in the analysis document and, in fact, the fixture (McMasters-Carr
Lampholder, Catalog No 1700Kl2) employs a swivel joint. The lighting fixtures are not safety-related, but mounting is considered
critical since they are in the control room and failure could endanger personnel
or safety-related
devices * MI0789-1683A-TC01-NL02
20 : * .. *:-. -**-. . .. .
-*. ' **:*:.-:***
Reason for Violation
The McMasters-Carr
Lampholder, Catalog No 1700Kl2 fixture has been used for the control room emergency
lighting design associated
with
The fixture employs a swivel joint for adjusting
only. The adjustment
is made in one plane only. The mechanism
used is a bolted connection
and the lamp tion is fixed in place by the friction from tightening
the bolt. Tightening
the bolt keeps the joint tight in service and keeps it from swiveling.
The assumption
of rigidity of the fixture service was based upon the analyst's
interpretation
of catalog data. That assumption
is considered
appropriate.
Plant administrative
design control procedures
required, and currently
that all analytical
assumptions
be documented, acknowledged
in terms of icance and technically
reviewed (Reference
1). The identified
discrepancy
results from failure to implement
this procedural
requirement.
Corrective
Action Taken and Results Achieved All e.ngineering
groups have .. been briefed as to the results of this inspection.
The briefings
were completed
on August 2, 1989. Corrective-Actions
to be Taken to Avoid Further Non Compliance
Interim * Same* as that required for Violation
Item 1.a. * Long-Terni-
-Develop a program to provide periodic refresher
training on "the requirements
of plant administrative
design change procedures
related to engineering
analyses.
Date When Full Compliance
Will be Achieved The personal briefings
letter will be issued by September
1, 1989. The program for periodic refresher
training will be in place by March 1, 1990. NRC Violation
255/89007-011:
Water Leak Detection
Set Point. [Refer to page 27 of NRC Report 50-255/89007 (DRS).] Example Specification
Change No 87-090 changed the Service Water (SW) leak detection
set point from 75 gpm to 300 gpm
verifying
what size of SW piping break in the containment
air coolers would result in a 300 gpm delta-flow
alarm * MI0789-1683A-TC01-NL02
21 -.... * *** *i,.. :. * .. *:. -... . -.
- CPCo Response The containment
SW leak detection
system monitors SW flow into and out of the reactor building and provides an alarm in the control room when a preset differential
flow is exceeded.
changed the differential
flow alarm set point from 75 gpm to 300 gpm. The instrumentation
loops for the leak detection
system consist of flow elements 1 differential
pressure transmitters
with square root output and a differential
flow switch with a time delay output. A time delay of approximately
15 seconds is incorporated
to eliminate
nuisance alarms due to flow noise spikes and still allow timely indication
of leakage. The SW leak detection
system is utilized as a post accident monitor. During accident conditions, without all control rods
water leaking inside the containment
building can dilute the containment
building sump water to a boron concentration
low enough to allow the reactor to return to a power state. As noted in Engineering
Analysis EA-SC-87-090-1, the basis for the original alarm set point of 75 gpm was engineering
judgement.
Further, the new 300 gpm set.point.was
selected based on the total inaccuracies
of the instrumentation
loop, times the full scale flow of the transmitters.
Use of instrument acies within the engineering
analysis provides a conservative
determination
based on instrument
capabilities.
As noted in the inspection
report, the engineering
analysis did not provide justification
that the set point meets the design intent of the SW leak tion systeqi..
However, the adequacy of the set point with respect to the tion system.design
intent was presented
and evaluated
as part of the written l0CFR50.5-9
.. (Safety Evaluation)
analysis for the SC. The safety evaluation
is part of the SC package and was reviewed with other supporting
documentation
comprising
the SC package by the Plant Review Committee (PRC) on March 2, 1987. Therefore, Consumers
Power Company does not acknowledge
this example as a lation of 10CFR50, Appendix B, Criterion
III. NRC Violation
255/89007-0lm:
Flow Transmitters." [Refer to pages 27 and 28 of NRC Report 50-255/89007 (DRS).] Example Specification
Change No 87-163 added a series voltage
zener diode to the feedwater
flow transmitter
instrument
loop for Transmitter
Nos FT-0701 and FT-0703 without specifying
the required zener diode design parameters.
Reason for Violation
upgraded FW flow transmitters
FT-0701 and FT-0703 to Rosemount
units. The supply voltage requirements
for an 1151 DP transmitter
is 12 Vdc to 45 Vdc (4 mA to 20 mA current loop). The transmitter
will operate within this voltage range as a function of load resistance.
The load resistance
for the FW flow transmitters
is approximately
300 ohms. The nominal supply voltage requirements
for the transmitter
as determined
from the Rosemount
functional
specifications
was approximately
19 Vdc. MI0789-1683A-TC01-NL02
22 *__:_-.*-:-
- .. ,._ *. ,. : o:. *..*.. * .. **:
- .* , ..... *. :, .. ,._ *,*
c* * ., .-.
- * As part of the SC, a zener diode was installed -in the series current loop to lower the power supply output voltage to the operating
voltage of the Rosemount
flow transmitter.
During development
of the SC, the design criteria for the zener diode, that is the required voltage was determined
to be 11 Vdc. This design criteria is shown on Drawing F-69 Sh 1, Rev 22 of SC-87-163.
As a result of this criterion
being stated within the SC package, the proper zener diode was installed
and as stat-ed in the inspection "the zeners were performing
their function." Therefore, Consumers
Power Company does not specifically
acknowledge-this
example as stated. While the design criterion
was detailed sufficiently
within the SC to provide for installation
of the proper zener diode, Consumers
Power Company acknowledges
the need for design packages to contain documentation
which provides the bases for engineered
changes. The failure to include the required enigneering
analysis which served as the basis for the design criterion
presented
within SC-87-163
has been attributed
to a weakness within the SC process regarding
documentation
of engineered
decisions.
Corrective
Actions Taken and Results Achieved In that the proper zener diode was prescribed
and installed, and resulted in the equipment
affected by the modification
being capable of performing
their design function, no immediate
corrective
actions have been undertaken.
All engineering
groups were briefed on the results of this inspection.
The briefings
were completed
on August 2, 1989. Correctiv.e .Actions to be Taken to Avoid Further Non Compliance
Interim Same as that required for Violation
Item 1.a. Long-Term
To ensure that adequate bases are developed
to justify the change and that these bases are technically
reviewed and documented
within the specification
change package, plant *administrative
procedures (Reference
5) will be revised either to require that a formal engineering
analysis (per Reference
1) or a new SC change justification
form be utilized for the following:
To provide a reason for the change (in part by describing
why the existing condition
is less than desired and why the change will improve as-built dition), ., *:ra describe the design basis function of the system within which this change is being made and justification
that this function will be maintained, -To identify the full impact
change will have on the system within which this change is being made and on potential
interfacing
systems, MI0789-1683A-TC01-NL02
23 ;---* . *:..:.::--.**
.*--... :. .. :.
,, .. ,, .. -....... ;:., :*. * .. **.:::;1-'* . :"'.*
- .. * *.*. .... :. -To identify critical functional
or physical features that must be met by the change to achieve the desired as-built condition (this may require formal engineering
analysis per Administrative
Procedure
9.11), and -To describe how these critical features will be verified (eg, inspection
or test). Date When Full Compliance
Will Be Achieved The personal briefings
letter will be issued by September
1, 1989. The revision to administrative
procedures
will be completed
by January 1, 1990. In addition, a program will be developed
by March 1, 1990 to provide engineers
with periodic refresher
training on SC-related
administrative
procedures.
NRC Violation
255/89007-0ln:
SC-88-069 "Upgrade Safety Injection
Tank Pressure Transmitters." [Refer to pages 29 and 30 of NRC Report 50-255/89007 (DRC).] Example Specification
Change No 88-069 added a series voltage regulating
zener diode to the safety injection
tank. pressure transmitter
instrument
loops for Transmitter
Nos PT-0361, 0367 , 0369, and 0371 without specifying
the required zener diode design parameters.
Reason for_Violation
safety injection (SI) tank pressure transmitters, PT-0363, PT-0367, ..PT-0369
and PT-0371 to Rosemount
units. This modification, like SC-87-163, introduces
a zener diode in series current loop to lower the power supply output voltage to the operating
voltage of the Rosemount
pressure mitter. During development
of the SC package for this modification, engineering
analyses.
were performed
to* determine
the design criterion
for the zener diode. However, as evidenced
by the transmitter
voltage measurements
taken during the inspection, an error was made .in the analysis.
This error was not identified
during design reviews of the modification
package due to the lack of a mented engineering
analysis within the SC package. Further, after modification
installation, no preoperational
testing specific to transmitter
operating age was conducted.
Therefore, the failure to attain a completed
modification
with all equipment
operating
within manufacturer
prescribed
operating
ranges has been attributed
to weaknesses
within the Specification
Change process regarding
documentation
of engineered
options and adequate preoperational
testing. Corrective
Action Taken and Results Achieved The power supply output voltage, zener diode vuic:age and transmitter
voltage for all the upgraded Rosemount
transmitters
associated
with SC-88-069
were measured.
As indicated
within the inspection
report, the transmitters
were found to be operating
outside their nominal operating
of 14 Vdc to 45 Vdc by.up MI0789-1683A-TC01-NL02
24 .. * '* .,. **:*
- to 12.62 Vdc. As a result of this finding, all other installed
transmitters
having zener diodes in their circuit had power supply, zener diode and mitter voltages measured.
From these measurements, two additional
non-safety
related transmitters (PT-5117 and PT-0927) were identified
to be operating
outside their prescribed
nominal* operating
range. Due to these findings, SC-89-162
was generated
to replace the improper zener diodes. As part of this modification
package, an engineering
analysis was completed
and technically
reviewed to assure proper zener diode selection
and to provide documentation
of design criterion.
The analysis was completed
on August 1, 1989. Additionally, work orders were generated
on June 5, 1989 to inspect the transmitters
that were operating
outside their nominal operating
range. Presentations
to all engineering
groups have been conducted
to. brief engineers
as to the NRC engineering
team inspection
results. These presentations
were completed
on August 2, 1989. Corrective
Actions to be*Taken to Avoid Further Non*Compliance
Interim Personal letters will be sent to all engineers
by September
1, 1989 describing
the NRC observed weaknesses
and requiring
that the engineer look at SC's rently being engineered
for similar problems.
Long Term -The plant administraive
procedure (Reference
5) revisions
described
for tion l.m apply as do the following:
-Revise plant administrative
procedures (Reference
1) to provide the technical
reviewer of an engineering
analysis a checklist
to assure a thorough, accurate and auditable
analysis.
The checklist
would feature a set of "prompts" in part to verifyall
analytical
input, assumptions
and calculation.
-Revise administrative
procedures (Reference
5) to require that pre-operational
testing be specified
as part of SC engineering
either in a work request or test procedure
prior to technical
review of the SC engineering
package. In addition, require that the test specification
align with the critical features identified
as part of the documented
change basis (see procedure
changes identified
for Violation
Item l.m). Date When Full Compliance
Will be Achieved Administrative
procedures
will be revised by January 1, 1990. Training on the procedure
revisions
will also be complete on January 1, 1990. In addition, a program will be in place by March 1, 1990 to provide periodic refresher
training on SC-rela.ted
procedures.
will be performed
by November 15, 1989. The work orders to inspect the affected transmitters
will be completed
by December 1, 1989. MI0789-1683A-TC01-NL02
25 ; .....
NRC Violation
255/89007-0lo:
SC-88-069 "Upgrade Safety Injection
Tank Pressure Transmitters." [Refer to pages 29 and 30 of NRC Report 50-255/89007 (DRS).] Example Specification
Change No 88-069 did not consider the effect of instrument
loop loading on the power supply; as a result, the load adjustment
resistor setting which matches impedance
for maximum power transfer was not specified
or adjusted.
Reason for Violation
upgraded safety inJection (SI) tank pressure transmitters, PT-0363, PT-0367, PT-0369 and PT-0371 to Rosemount
units. This modification, like SC-87-163, introduces
a zener diode in series current loop to lower the power supply output voltage to the operating
voltage of the Rosemount
pressure mitter. While reviewing
this SC the inspector
reviewed the SI tank pressure loop power supply manual. As-stated
intheinspectionreport; "the Foxboro Model 610A power supply is designed to furnish power to a single electronic
transmitter.
The nominal DC output voltage is 80 volts. The manual also states that the output load resistance
must be 600 ohms +10; -20 percent. The SC package did not determine
the load resistance.
The manual provided detailed instructions
to sum the input resistances
of all the receivers
in the loop (excluding
the
and to adjust the load adjustment
dial on the power supply to the difference,,between
the loop resistance
and 600 ohms. Subsequentcto
the inspection
on July 25, 1989, plant engineering
personnel
contacted
the power supply vendor to discuss the inspector's
concern regarding
the affects of increased
load resistance
on the power supply. During this conversation
the vendor noted that the specific requirement
for a load tance of 600 ohms applies only to Foxboro transmitters
connected
to Foxboro power supplies and that applied power supply load resistance
is based on the voltage requirements
of the associated
transmitter.
The voltage requirements
of the Rosemount
transmitters
installed
under SC-88-069
are addressed
in the modification
package, however, documentation
was not provided regarding
resultant.
power supply l6ad resistance.
Failure to include applicable
documentation
within the modification
package has been attributed
to a lack of guidance being provided within Administrative
Procedure
9.04, fication Changes." Corrective
Action Taken and Results Achieved Presentations
of the inspection
results were made to all affected engineering
groups. These presentatioris
were completed
on August 2, 1989. Corrective
Actions to be Taken to Avoid Further Non Compliance
Personal letters will be sent to all engineers
describing
the NRC engineering
inspection
results by September
1, 1989. The letters will require that neers review SC packages currently
being engineered
for similar problems.
MI0789-1683A-TC01-NL02
26 -:* : .. \_._ .:: ;*'.. ,. -** :=. * .. : ** _.-:-: .*
- . '. *--.*. -* .""; . . *.* .. '<' * .* , *' . ** ...
.. *. . '*
- -The plant administrative
procedure
revisions (and training)
described
for lation Items l.m and l.n effectively
respond to this item also. Date When Full Compliance
Will be Achieved Administrative
procedures
will be revised by January 1990. Training in the procedure
revisions
will also be complete on January 1, 1990. In addition, a program will be in place by March 1, 1990 to provide periodic refresher ing on SC-related
procedures.
NRC Violation
255/89007.0lp:
SC-88-102 "Upgrade Containment
Pressure Transmitter
PT-1812." [Refer-to
pages 31 and 32 of NRC Report 50-255/89007 (DRS).] Example Specification
Change No 88-102 installed
a different
model containment
pressure transmitter
for Transmitter
No PT-1812 without performing
a seismic analysis to determine
the acceptability
of installing
the new transmitter
on the old mounting.
Reason for-Violation
upgraded containment
building pressure transmitter, PT-1812 to a Rosemount
pressure transmitter.
The pressure loop affected by the modification
provides indication
only and is not required to be operable for any analyzed event. The pressure transmitter
is mounted off piping associated
with
ment Penetrcation
MZ-17 and is physically
located between the manual instrument
isolation
valve and the manual containment
isolation
valves. The manual instrument
isolation
valve is maintained
open to allow pressure transmitter
operation.
Therefore, the primary containment
boundary includes PT-1812. While processing
SC-88-102, engineering
personnel
- failed to identify that the pressure transmitter
constituted
part of the containment
boundary.
This ure is attributed
to the following
factor: The administrative
procedure
for Specification
Changes (Reference
5) requires that the engineer consult the Equipment
Data Base (EDB). The EDB-Q-Listing
identifies
the pressure retaining
and structural (seismic)
requirements
to be met by the equipment.
The existing Q-Listing
in the EDB for PT-1812 indicates
that the transmitter
function is not safety-related, there are no pressure retaining
requirements, and that the structural
mounting is not safety-related.
This specific Q-Listing
needs to be reviewed and revised as necessary.
Given accurate EDB information, the existing_
SC checklist "prompts" which also existed at the time this deficiency
occurred, are sufficient
to identify the governing
design codes, standards
and regulatory
guides to be complied with. Corrective
Actions Taken and Results Achieved A formal seismic engineering
analysis has been initiated
to document the adequacy of the existing transmitter
mounting and the associated
tubing. MI0789-1683A-TC01-NL02
27 ;..&.. ',' :*,* : .:*:
- . ..... , .* *:: :.:;.: *;* .. .. . , .. :" : .
--;_,.
The results of the inspection
have been presented
to all engineering
groups. These presentations
were completed
on August 2, 1989. Corrective
Actions to be Taken to Avoid Further Non Compliance
The existing Q-List interpretation
for PT-1812 will be reviewed for accuracy and revised as necessary.
In addition, if it is determined
that the tation is in error, other interpretations
will also be reviewed to identify the breadth of the discrepancy.
These additonal
reviews will cover, as a minimum, interpretation
for other instrumentation
serving pressure retaining
functions.
If additional
reviews indicate the need, additional
clarification
in
tive P.rocedures
related to Q-List interpretation (Reference
6) will be provided and engineers
will be trained. Further, a review will be conducted
to ensure the seismic qualification
of other similar configurations.
In addition, a program to provide periodic refresher
training on procedures
related to Q-Listing
will be developed.
Finally, a portion of the Configuration
Control Project involves the tion of the Q classification
for approximately
16,000 components
in the Plant's equipment
data base. This activity is currently
scheduled
to be completed
by the end of-1990 and will provide a sound technical
basis for future tions. Date When F.ull Compliance
Will Be Achieved The existing Q-List interpretation
for PT-1812 will be reviewed for accuracy and revised necessary)
by September
15, 1989. If it is concluded
that the PT-1812 interpretation
is in error, interpretation
for other similar tions will be completed
by November 1; 1989. If these additional
reviews tate the need for procedural
clarification, the procedures
will be enhanced by January 1, 1990 and all engineers*
will be trained on the enhancements
by this date. The program for periodic refresher
training on Q-Listing
will be in place by March 1, 1990. The additional
seismic review will be completed
by October 1, 1989. NRC Violation
255/89007-0lg:
EA-FC-722-10 "N2 Backup Test Evaluation
for Station 5," February*21, 1987. [Refer to page 15 of NRC Report 50-255/89007 (DRS).] Example The
stated that the nitrogen usage rate was 32.5 psig AP/hour based on the test results from Functional
Test T-FC-722-501-01.
However, the test results failed to account for the post test calibration
shift of 5 psig for of the pressure gauges. By incorporating
this additional
factor, the usage rate is increased
to 33.75 psig AP/hour. MI0789-1683A-TC01-NL02
28 . *. . -:* ** '7. *. ,-'* .. .*,* . ,* ' *.*.:* ... * *,'* *. **-:*** ."'-*,' .. * .** .... *.,_ ..
.; .....
. ' * * Using the above rate in the calculation
reduces the "actual operating
period" from 10.3 days to 9.93 days. This is below the assumed acceptance
limit given in the original calculationo
No safety significance
was attributed
to this occurrence;
however, the instrument
accuracy requirements
specified
in the test procedure
were inadequate
as noted belowo -Procedure
No T-FC-722-0501, "CV Air Supply -N2 Backup Performance
Test," Revision O, February 6, 1987. Under Special Tools/Equipment, a 0-3000 psig pressure gauge is called for. The accuracy specified
is +/- 2% minimum. This equates to a +/- 60 psig accuracyo
The acceptance
criteria for three of the four nitrogen stations ranged from 24 psig to 68 psig over the four hour span of the performance
test. CPCo Response CPCo does not acknowledge
this example as a violation
of
Appendix Criterion
III Design Control," based upon the following.
1. Page 6 of 32 of "Palisades
Nuclear Plant Modification
Procedure
No T-FC-722-501," and "Temporary
Change to a
Change No FFC-87-006, specified
calibrated
analog pressure gauges, 0-3000 psig, +/- 2% minimum accuracy and that these gauges shall be calibrated
in accordance
with 2.4, reference
paragraph
6.1.5. 2. The intent of specifying
a minimum accuracy of the test gauges was to allow qualified
test personnel
the. flexibility
to utilize test gauges of a higher degree"of
accuracy if available.
3. The intent of Reference
2.4 (Palisades
Nuclear Plant Administrative dure S.07, "Control of Measuring
of and Test Equipment"), paragraph
6.1.5, is to require performance
of pre-and post-calibrations
of the test gauges. These calibrations
were performed
as
Pre-and Post-Calibrations
of the gauges are utilized to determine/verify
the actual gauge accuracy as utilized during the test. 4. As stated in paragraph
1 of page 16 of NRC Report No 50-255/89007 (DRS), "Additional
reviews by the inspector
disclosed
that the pressure gauges actually used has a specified
accuracy of +/- 1%. In addition, pre-test and post-test
calibration
data indicated
that the actual accuracy was closer to +/- 0.1%." This statement
reinforces
the intent of specifying
and the requirement
to perform pre-and post-calibrations (reference
Item 83) of the gauges. 5. Acceptance
criteria for Palisades
Nuclear Plant Modification
Procedure
No T-FC-722-501
are established
via calculation
and are not affected by gauge inaccuracies
which are linear and constant throughout
the test range *
29 ... .: . ' . . ...... *--*
- * Based upon the above the specification
of test gauges, 0-3000 psig, +/- 2% accuracy was appropriate
and in accordance
with Palisades
Nuclear Plant Administrative
Procedures--.
Plant administrative
design control procedures (Reference
2) required, and currently
require, that modification
test procedures
feature requirement
-The use of calibrated
test equipment
of the proper range and accuracy to determine
conformance
to specified
acceptance
criteria, -Test equipment
be identified
along with its calibration
status, and -Acceptance
criteria (with appropriate
tolerances)
be specified
to effectively
determine
whether critical design requirements
have been satisfied.
Thus, no corrective
action is deemed necessary.
NRC Violation
255/89007-0lr:
Flow Transmitters." [Refer to pages 27 and 28 of NRC Report 50-255/89007 (DRS).] Example Specification
Change No 87-163 added a series voltage regulating
zener diode to the FW flow transmitter
loop for Transmitter
Nos FT-0701 and FT-0703 without specifying
__ the measurement .of. the power supply, zener, and transmitter
voltage as acceptance*
criteria to determine
if the transmitter
loop was operating
within its-design
limits. Reason for Violation
upgraded FW flow transmitters
FT-0701 and FT-0703 to Rosemount
units. The supply voltage requirements-
for a 1151 DP transmitter
is 12 VDC to 45 VDC (4 mA to 20 mA current loop). The transmitter
will operate within this voltage range as a function of load resistance.
The load resistance
for the FW flow transmitters
is approximately
300 ohms. The nominal supply voltage requirement
for the transmitter
as determined
from the Rosemount
functional
specifications
was approximately
19 Vdc. As part of the SC a zener diode was installed
in the series current loop to lower the power supply output voltage to the operating
voltage of the Rosemount
flow transmitter.
During the inspection, the NRC inspector
identified
that the SC package did not contain post installation
power supply output voltage urements.
Further, it did not contain zener diode and transmitter
operating
voltages following
modification.
The failure to adequately
specify necessary
preoperational
testing requirements
on the work orders which implemented
the SC has been attributed
to weaknesses
within Administrative
Procedure
9.04. Currently, no guidance exists as to the type of
which may be appropriate, nor does the procedure
specify the need to document testing performed
on implementing
work orders or within the SC package.
30 . *** ..............
- .*:***:_-
.. *. . .. ., .. ... ......
Corrective
Actions Taken and Results Achieved As noted within the inspection
reportp the power supply output voltage, and the zener diode and transmitter
operating
voltages were measured.
From these urements it was determined
that all components
were performing
their design function within manufacturer
specifications.
Presentations
have been made to engineers
discussing
the results of the recent NRC engineering
inspection.
These presentations
were completed
on August 2, 1989. Corrective
Action to be Taken to Avoid Further Non Compliance
Personal letters will be sent to all engineers
on or before September
lp 1989 describing
the results of the NRC inspection
and requiring
that SC's currently
being managed be reviewed for similar problems.
Date When Full Compliance
Will be Achieved The procedure
revisions
for Violation
Items l.m and l.n will effectively
respond to this item. NRC Violation
255/89007-0ls:
SC-88-069 "Upgrade Safety Injection
Tank Pressure Transmitters." [Refer to pages 29 and 30 of NRC Report 50-255/89007 (DRS).] NRC Identi£ied
Discrepancy
Specificai:ion
Change No 88-069 added a series voltage regulating
zener diode to the safety injection
tank pressure transmitter
loops for Transmitter
Nos PT-0363, 0367, 0379, and 0371 without specifying
the measurement
of the power supply, zener, and the transmitter
voltage as acceptance
criteria to determine
if the transmitter
loop was operating
within its design limits; and also did not specify acceptance
criteria for determining
the acceptability
of changing the load adjustment
resistor in the power supply. Reason for Violation
Consumers
Power Company's
response regarding
the failure to specify acceptance
criteria to determine
if the transmitter
loop was operating
within its design limits in the preoperational
stage is provided in our response to Violation
Item l.m. In regard to the post modification
stage of this SC, the failure to establish
a program to periodically
measure the pressure transmitter
loop voltages has been attributed
to plant personnel
not considering
all potential
failure modes and effects in the circuit design. Acceptance
criterion
for determining
the acceptability
of changing the load adjustment
resistor in the power supply were not specified
in the SC package. The manual for the Foxboro 610A power supply stated that the output load resistance
for the power supply must be 600 ohms + 10; -20 percent. In matory conversations
with the vendor on July 25, 1989, the requirement
for load resistance
was said to be based on transmitter
limitations, not power supply limitations.
The new Rosemount
transmitters
installed
per SC-88-069
do MI0789-1683A-TC01-NL02
31 . . : ' :* -. . : -. *-: ... ... , .... ....
not have this load restriction
and hence do not have acceptance
criteria as delineated
in the manual. Therefore
this item by itself is not a violation
of 10CFR50-, Appendix B, Criterion
III. It is noted however that the new Rosemount
transmitters
have voltage limitations
and this is discussed
in our response to Violation
Item l.n. Corrective
Actions Taken and Results Achieved Same as that taken for Violation
Item l.n. Corrective
Actions to be Taken to Avoid Further Non Compliance
Procedural
revisions
and tra1n1ng described
for Violation
Item l.n will ively respond to this item. Additionally, preplanned
and periodic control sheets (preventive
maintenance
activities)
will be established
to provide for periodic measurements
of loop voltages.
Date When Full Compliance
Will be Achieved The control sheet program will be established
by October 1, 1989. Violation
'255/87007-02a-b)
lOCFRSO, Appendix B, Criterion
X as implemented
by the Palisades
Operations
Quality Assurance
Program requires, in part, that a program for inspection
of activities-,affecting
quality be established
and executed by or for the zation performing
the activity to verify conformance
with the documented
instructions, procedures, and drawings for accomplishing
the activity and that examinations, measurements, or tests of materials
or products processed
be performed
for each work operation
where necessary
to assure quality. Contrary to the above: This is a Severity Level IV Violation.
NRC Violation
255/89007-02a:
CPCo Drawing M-101 Sheet 5113, Revision O, "Piping Isometric, Auxiliary
Control Valve CV-0736A and CV-0737A Bypass Piping." [Refer to pages 12 and. 13 of NRC Report 50-255/89007(DRS).]
Example A secondary
aspect, associated
with the socket welds, pertains to the quality control (QC) inspection
of the completed
fillet welds. The RIC forms have a column for "QC verification" but for the socket welds in question, the size of the fillet welds was not inspected
by QC. Line No 16 of the RIC form, which specifies
the weld, size, gap, and type of joint was marked "NA" (not applicable)
for all the welds in question under the QC Verification
column. Although all of the welds received a Nondestructive
Testing (NDT) Visual Examination (VT), it is not clear if the size of the welds was verified during these examinations.
Since the size of the socket fillet welds was not specified
on the drawing, nor noted on the RIC form, the NDT examiner would MI0789-1683A-TC01-NL02
32 ** : * '."'!'* *.* :: * .. * ..... ' ..... : ..... '.'-:_: ....... *.** -:-*:**.**
- . ,
- .-*.'* .. * ;, ..... * ::* : : .. *: .* ***.* ....... * .. : ........ .
- have had to determine
the required size in the same manner as previously
described
for the welder. No notation of size nor record of the size calculation
was
in the documentation
provided with the NDT-VT data. In addition, the VT report did not list fillet weld gauges under "Visual Aids Used" giving further indication
that the size of the welds was not checked. As a point of clarification, it should be noted that the VT performed
on the socket fillet welds was in accordance
with American Welding Society (AWS) Dl.l requirements.
This is a structural
welding code and allows portions of fillet welds to be undersized
by 1/16". This is inconsistent
with the requirement
of ANSI 831.1, Power Piping Code which specifies
minimum fillet weld sizes. If the size of the-socket fillet welds was verified by the stated VT examinationp
it cannot be assured that the weld meets the ANSI 831.1 Code requirements.
Reason for Violation
The failure to merit conformance
of the size of the socket fillet welds has been attributed
to a lack of engineering
input to and technical
review of the maintenance
planning for the welding process. Prior to actions taken as a result of recent self-identified
failures to verify weld size (Reference
7), no specific requirements
existed to verify characteristics (weld, type, size contour) of installed
welds. Although Nuclear Operations
Department
Standards
suggest inspection
hold points for weld installation
verification, working level administrative
procedures
did not specify:a
hold point requirement
except for fit up. Corrective'"Action-Taken
and Results Achieved Presentations
to all engineering
groups have been conductep
to review the results of this inspection.
These presentations
were completed
on August 2, 1989. -The Inservice
Inspection (ISI) Section oP the Projects Engineering
Department
has assumed the role of Design Authority
for weld engineering
by revising the RIC to technically
review the maintenance
planner's
specifications.
The purpose of the review is to ensure that appropriate
welding codes are complied with in the areas of weld installation
and post-installation
examination.
-The RIC has been revised to issue the-weld minimum leg length to the field. This will eliminate
the need for the field welder to calculate
the length. The aforementioned
ISI review will assure that this specification
is provided.
-Reference
Violation
255/89007-0lc
for other applicable
actions being taken. Corrective
Actions to be Taken to Avoid Further Non Compliance
Specifying
welding requirements (such as applicable
code, weld material, weld type and weld size) is an engineering
function.
If properly administered
by procedure, the maintenance
planner can (and has) effectively
prescribe
welding MI0789-1683A-TC01-NL02
33 :. . . ' . : . -; : . ':* *:-. . . ': . *. *: **.::-. ,._ *. ,*,_ ... , ..
details for the field provided that adequate input from engineering
exists as a basis. In the past, engineering
input has been limited to welding
tion and/or structural
analysis engineering
sketches which have lacked size dimensions
for the welds. As a result 11 the planner has failed to provide the proper size on the Repair Inspection
Checklist (RIC) thereby requiring
the field welder to determine
and install the proper weld size. This practice fails to meet current expectations
for control of design change implementation.
Although plant administrative
design control procedures
required and currently
require that the design change project engineer determine
code requirements
for assigned projects (Reference
4), and plant maintenance
procedures
required and currently
require that the maintenance
planner specify applicable
code and weld parameters
after consultation
with the Engineering
Department (Reference
3), these procedures
had not been effectively
integrated
to support one another to ensure that weld specifications
from engineering
were accurately
translated
into installation
planning, installation, and post-installation
verification.
As a result, the following
actions have been/will
be taken to prevent rence: Interim Same as that required for.Violation
Item.l.a.
Long-Term
-Enhancements
to .plant design.control
and maintenance
procedures
will be made to more effectively
integrate
engineering
into weld specification
and mately -into weld planning and verification:
Appropriate
welding codes will be included in the Design Input Checklist (Reference
2) to prompt the design engineer to specify appropriate
weld requirements (for installation
and examination)
in the facility change package as part of both conceptual
and detailed engineering.
-Design control procedures
related to engineering
analyses (Reference
1) will explicitly
require that all drawings accompanying
structural/seismic
analyses provide detailed weld information (type, size, material)
for input to the planner. In addition, the procedures
will require that sizing culations
be performed
as part of the analysis.
Finally, a technical
review checklist
will be provided to require that the reviewer ensure that weld information
be accurately
represented
on the analysis drawings.
Plant maintenance
procedures (Reference
3) will require that the maintenance
planner utilize the contents of the facility change package to complete the RIC in specifying
for the field weld installation
and examination ments. The procedure
will require that the planner consult the Design Input Checklist
and structural/seismic
engineering
analyses.
Interim actions related to changes to the RIC and ISI group review of the RIC (as described
above) will remain in effect. MI0789-1683A-TC01-NL02
34 *. *, -;.*; .. *:. ' ....... .. *' .... -. . *.::*****
,: .... . :*
.. ;*,
.. .a.* ' *-**.*. :-*: -.
- -Design and quality assurance
engineers
will be trained on the appropriate
structural
and piping weld codes and their application
to weld installation
and examination.
The engineers
will also be trained on the above procedural
enhancements.
A program will be developed
to periodically
train design and quality assurance
engineers
on the aforementioned
codes and their application, and on the related design control and maintenance
procedures.
In summary, it is expected that these actions will ensure that proper welding requirements (type, material, size) are specified
by engineeringp
planned by maintenance (with a check on planning by engineering)p
and in turn verified by quality control. Date When Full Compliance
Will be Achieved The personal briefings
by letter will be issued prior to September
lp 1989. Procedure
enhancements
and required training on the enhancements
will be pleted by January 1, 1990. The program for periodic refresher
training will be developed
by March lp 1990. NRC Violation
255/89007-02b:
SC-89-072 (Deviation
Report D-PAL-89-043).
[Refer to page 32 of NRC Report 50-255/89007 (DRS).] Example This
report documented
the undersized
fillet welds on socket welded fittings -for SC-89-072.
This specification
change was necessary
to provide an interim solution to primary coolant system leakage from cold leg drain valves. The change required the
of a new length of two inch schedule 160 pipe with a socket welded cap on each of the four loop drains. Inspection
of all eight socket fillet welds indicated
that none of them met the Code required size of 3/8 inch. During the inspector's
review* of the deviation
report, there were several concerns that apparently
were not addressed.
First, although the corrective
actions appear to recognize
that the current RIC form does not give the welder sufficient
information (specifically
the size of the fillet weld), there was no recognition
that QC did not and was not required to verify the size of the fillet weld. The.undersized
condition
was not discovered
until the authorized
inspector (AI) pointed it out to the licensee.
All of the welds had been reviewed and
by the licensee's
program and yet the size had never been verified.
This is considered
another example of violation
of 10CFR50, Appendix 8p Criterion
X, in that the size of the socket fillet welds was not verified (255/89007-02b).
Reason for Violation
Specifying
welding requirements (such as applicable
code, weld material,
type and weld size) is an engineering
function.
If properly administered
by procedure, the maintenance
planner can (and has) effectively
prescribe
welding MI0789-1683A-TC01-NL02
35 . *.".,'T:'
- .*. ': .. *: ***-._ .. ___
- .,. . ... . * .. *.:* *. '. . .. --. _.,*.
- details for the field provided that adequate input from engineering
exists as a basis. In the past, engineering
input has been limited to welding
tion and/or structural
analysis engineering
sketches which have lacked size dimensions
for the welds. As a result, the planner has failed to provide the proper size on the Repair Inspection
Checklist (RIC) thereby requiring
the field welder to determine
and install the proper weld size. This practice fails to meet current expectations
for control of design change implementationo
Corrective
Action Taken and Results Achieved -Presentations
to all engineering
groups were conducted
to brief engineers
as to the results of this inspection.
The presentations
were completed
on August 2, 1989. -The Inservice
Inspection (ISI) Section of the Projects Engineering
Department
has assumed the role of Design Authority
for weld engineering
by revising the RIC to technically
review the maintenance
planner's
specifications.
The purpose of the review is to ensure that appropriate
welding codes are complied .with in the areas of weld installation
and post-installation
examinationm
-The RIC has been revised to issue the weld minimum leg length to the fieldo This will eliminate
the need for the field welder to calculate
the length. The aforementioned
ISI review will assure that this specification
is provided.
Corrective
Actions to be Taken to Avoid Further Non Compliance
Although .plant administrative
design control procedures
required and currently
require that the design change project engineer determine
code requirements
for assigned projects (Reference
4), and plant maintenance
procedures
required and currently
require that the maintenance
planner specify applicable
code and weld parameters
after consultation
with the Engineering
Department (Reference these procedures
had not been effectively
integrated
to support one another to ensure that weld specifications
from engineering
were accurately
translated
into installation
planning,.
installation, and post-installation
verification.
As a result, the following
actions have been/will
be taken to prevent rence: Interim Same as that required for Violation
Item l.a. Long-Term
-Enhancements
to plant design control and maintenance
procedures, and to ESS Departmental
guidelines
will be ***ade by January 1, 1990 to more effectively
integrate
engineering
into weld specification
and ultimately
into weld ning and verification:
-Appropriate
welding codes will be included in the Design Input Checklist (Reference
2) to prompt the design engineer to specify appropriate
weld requirements (for installation
and examination)
in the facility change package as part of both conceptual
and detailed engineering.
MI0789-1683A-TC01-NL02
36 . , ... 'i--: ' .. . .. ' ....... ;** _. **:*: I
. '* ,. ' .. Design control procedures
related to engineering
analyses (Reference
1) will explicitly
require that all drawings accompanying
structural/seismic
analyses provide detailed weld information (type, size, material)
for input to the planner. In addition, the procedures
will require that sizing culations
be performed
as part of the analysis.
Finally, a technical
review checklist
will be provided to require that the reviewer ensure that weld information
be accurately
represented
on the analysis drawings.
-Plant maintenance
procedures (Reference
3) will require that the maintenance
planner utilize the contents of the facility change package to complete the RIC in specifying
for the field weld installation
and examination ments. The procedure
will require that the planner consult the Design Input Checklist
and structural/seismic
engineering
analyses.
-Interim actions related to changes to the RIC and ISI group review of the RIC (as described
above) will remain in effect. -Design and quality assurance
engineers
will be trained on the appropriate
structural
and piping weld codes-and
their application
to weld installation
and examination.
The engineers
will also be trained on the above procedural
enhancements.
A program will be developed
to periodically
train design and quality assurance
engineers
on the aforementioned
codes and their application, and on the related design control and maintenance
procedures.
In
is expected that these actions will ensure that proper welding requirement-s (t-ype, material, size) are specified
by engineering, planned by maintenance (with a check on planning by engineering), and in turn verified by quality control *. Date When Full Compliance
Will be Achieved *The personal briefings
by letter will be issued prior to September
1, 1989. Procedure
enhancements
and required training on the enhancements
will be pleted by January 1, 1990. The program for periodic refresher
training will be developed
by March 1, 1990. NRC Violation
255/89007-03:
Low Temperature
Over Pressure Set Points. [Refer to page 28 of NRC Report 50-255/89007 (DRS).] Technical
Specification (TS) No 3.1.8.1.a
requires a low temperature sure (LTOP) power operated relief valve (PORV) lift setting of < 310 psia for Tc < 300°F and TS 3.1.8.1.b
requires a LTOP PORV lift setting 575 psia for Tc < 430°F. Contrary to the above, between August 9, 1988 and February 27, 1989, the PORV as-left setting exceeded the TS requirement
on 17 occasions.
This is a Severity Level IV violation.
MI0789-1683A-TC01-NL02
37 *.-* ***-**-;* . * '* .* .. *.** ...... :* :,-* ;;;. ..... : .. :* .. , .
., I .. * * :* .. * .. Reason for Violation
changed the LTQP protection
system set points for temperature
switches TS-0115 and TS-0125. The LTOP system provides primary coolant system {PCS) overpressure
relief capability
to protect the reactor vessel from the potential
for brittle fracture.
The Palisades
LTOP system is a two channel system which relieves PCS pressure through either of two PORV's. Channel A relieves through PRV-1042B
and channel B relieves through
The system is enabled at two settings.
When the PCS cold leg temperature
is less than or equal to 300°F, the lift set point for the PORV is less than or equal to 310 psia. When the PCS cold leg temperature
is greater than 300°F but less than 430°F, the set point for PORV opening is less than or equal to 575 psia. Above 430°F the LTOP system is not required to be enabled. The LTOP system set points are derived from plant heatup and cooldown limits specified
in Plant Technical
Specifications.
The set points reflect the ature and pressure limits calculated
according
to the requirements
of Appendix G to 10CFR50, using the methodology
provided in Regulatory
Guide 1.99, Revision 2. These set points were enacted with the issuance of Amendment
117 to the Palisades
operating
license on November 14, 1988. At the time the 310 and 575 psia LTOP PORV set points were proposed on the Technical
Specification
change request which resulted in the issuance of Amendment
117, existing Technical
Specifications
did not recognize
the need for LTOP above_300°F.
Instrumentation
existing at this time did not operate above 600 psia had a recognized
accuracy of +/- 22 psia. Therefore, the 310 and 575 psia points were selected to provide the maximum practical
operating
window allawed by exi.sting
plant components
while remaining
bound by 10CFR50 Appendix G limits. The proximate
cause of this condition
is that the set point value which results from the addition of instrument
inaccuracies
is not conservative
with the lift point specified
in Technical
Specifications.
This condition
has been attributed
to poor documentation
within the Technical
Specifications
regarding
the fic lift point value. When the technical
specification
value was derived, Engineering
personnel
subtracted
instrument
inaccuracies
from the 10CFR50 Appendix G limit and arrived at the 310 and 575 psia set points found in Technical
Specifications.
The intent of the Technical
Specification
lift point value is to ensure compliance
with Appendix G. The typical set point methodology, if applied to this situation, would be to provide the applicable
Appendix G limit in TS and then control the actual set point, adjusted for instrument
inaccuracies, through Technical
Specification
Surveillance
dures. As noted in the inspection
report, the issue was identified
in parallel by both the and plant personnel.
At the plant, the issue was identified
during a review of the set point methodology
process utilized at Palisades.
Plant Engineering
personnel
identified
that the PORV lift point had been set at the technical
specification
values of 310 and 575 psia. Setting the lift points at the technical
specification
value, neglecting
instrument
accuracies, could result in the actual lift points being 332 and 597 psia when maximum instrument
inaccuracies
are accounted
for. A review of past performances
of MI0789-1683A-TC01-NL02
38 * '
- .. * * ,.. .-I -* <.. . / * r * ;:: ** **,: *. . "' .._<:*
. : , ........ ,-:*. Technical
Specification
Surveillance
Procedures
M0-27A through D which provide for functional
testing of the LTOP system, revealed that 29 of the 31 times lift set points (310 or 575 psia) were checked, the set point was greater than the technical
specification
limitc While the lift point did exceed the technical
specification
limit, it was within the acceptance
values provided by 10CFR50 Appendix Ge Corrective
Actions Taken and Results Achieved Plant Engineering
personnel
reviewed the basis for Technical
Specification
3.1.8.1 and Technical
Specification
Surveillance
Procedures
which set the PORV lift points and verified that even if the largest positive instrument
inaccuracy
was added to the technical
specification
lift point, the 10CFR50 Appendix G limit would not be exceeded.
Upon further review it was additionally
identified
that the curve utilized in defining the Appendix G limit has incorporated
a 30 psia measurement
inaccuracy.
In that a Technical
Specification
change request is being prepared for submittal
in support of LTOP protection
system modifications
to be performed
during an upcoming maintenance
outage, a letter of interpretation
was submitted
to the NRC on July 12, 1989 which presented
Consumers
Power Company's
position regarding
continued
compliance
with 10CFR50 Appendix G. Technical
Specification
Surveillance
Procedures
M0-27C and M0-27D 9 which provide setting and
the PORV lift set points were revised on May 11, 1989 to remove the + 22 psia tolerance.
Corrective_Actions
to be Taken to Avoid Further Non Compliance A Technical
Specification
change request will be submitted
which delineates
the requi.red
PORV lift set points to assure continued
compliance
with 10CFR50 Appendix G limits following
LTOP protection
system modifications.
- An tion of the Technical
Specification
change request development
process is being undertaken
to determine
where enhancements
in the review process are required to preclude future occurrences.
Date When Full Compliance
Will be Achieved Continued
compliance
with the lift set point value specified
in the Technical
Specifications
has been assured by submittal
of Consumers
Power Company's
letter dated July 12, 1989 and the rev1s1ons
to M0-27C and M0-27D. The cal Specification
change request supporting
the planned LTOP protection
system modifications
will be submitted
by October 1, 1989. The evaluation
of the Technical
Specification
change request development
process will be completed
by November 1, 1989. NRC Open Item 255/89007-04:
Consumers
Power Company Drawing M-101 Sheet 5113, Revision O, "Piping Isometric, Auxiliary
Control Valve CV-0736A and CV-0737A Bypass Piping, 11 [Refer to page 13 of NRC Report %-255/89007
(.DRS).] Example 'An additional
aspect was associated
with the size of socket fillet welds: The inspector
noted that the current design practice used by the licensee is sistent with the original Code of construction.
The current practice utilizes MI0789-1683A-TC01-NL02
39 '* ... ; .... "' ,****:::. *:;* .... -,, . ' . "*;. :.-: .. ..;.-.: ,;.* ... -.. . ..... .. .. .... :*"'* '/' --
- later editions of 831.1 Code which specify the lo09 times the nominal p1p1ng wall thickness.
The original Code of construction
required 1.25 times the nominal wall thickness. -From a technical
standpoint
the current practice is acceptable;
however, this inconsistency
has not been delineated
by the licensee in the FSAR. Pending revision of the
this item is considered
open (255/89007-04).
Reason for Violation
Construction
codes related to 831.1 have not been reconciled
1n a document useable to the modifications
engineer.
Corrective
Action Taken and Results Achieved Presentations
have been made to all engineering
groups on the results of this inspection.
These presentations
were completed
on August 2, 1989. Corrective
Actions to be Taken to Avoid Further Non Compliance
Interim Same as that required for Violation
l.a. Long-Term
Palisades
&taff will complete a reconciliation
of all construction
codes to the latest edit:,.ion
of 831.1. This. action would provide for standardization
of code usage-and
simplify the determination
of code requirements.
This effort will also address the structural
welding code AWS Dl.l. Such reconciliation
will be documented
in plant administrative
design control procedures ence 4). In addition, a periodic training program covering procedural
welding requirements
will be developed.
Upon completion
of the reconciliation
the FSAR will be updated to* identify applicable
codes and standards
and their application.
Date When Full Compliance
Will be Achieved The personal briefings
letter will be issued by September
1, 1989. The ciliation
of construction
codes will be completed
and implemented
into plant. design control procedures
by January 1, 1990. Training on these procedural
revisions
will also be complete by January 1, 1990. The periodic training program will be in place by March 1, 1990. The FSAR will be updated in the next revision following
January 1, 1990. NRC Unresolved
Item 255/89007-06:
SC-89-072 (Deviation
Report D-PAL-89-043).
[Refer to page 32 of NRC Report 50-255/89007 (DRS).] MI0789-1683A-TC01-NL02
40 .--' ...
.. '.* .:-.. --*<.
Example The second concern pertains to the generic aspect of the problem. The licensee appeared to recognize
the programmatic
weakness which contributed
to the problem by revising the RIC form to include the specific weld size. However, there appeared to be no corrective
actions directed toward reviewing
previously
made socket fillet welds for compliance
with Code requirements.
Based on the added complication
that the sizes of fillet welds in general apparently
have not been verified under the licensee's
program, reviews of past work may not be sarily limited to socket welded fittings.
Pending a review of the licensee's
justification
as to why additional
inspection
of previous fillet welds is not required, this is considered
an Unresolved
Item (255/89007-06).
CPCo Response CPCo acknowledges
that no corrective
actions have yet been directed towards reviewing
previously
made socket fillet welds for compliance
with code ments. CPCo plans, however, to select an appropriate
sample of as-built welds and inspect the-welds
during the 1989 maintenance
outage. The sample will be chosen to include a range of weld types. The purpose of the inspection
will be to verify that the weld characteristics (type and size) conform to requirements
set forth in the Repair Inspection
Checklist
and/or applicable
welding code. These field verifications
and resulting
report will be completed
by December 1, 1989. NRC Unresolved
Item 5: Consumers
Power Company Drawing M-101 Sheet 5113, Revision O,. "Piping Isometric, Auxiliary
Control Valve CV-0736A and CV-07-3JA
Bypass Piping." [Refer to page 14 of NRC Report 50-255/89007 (DRS).] NRC Identified
Discrepancy
A further concern associated
with the p1p1ng installation
drawing pertains to the attachment
weld for a bypass piping fitting onto the existing run pipe. For this situation, the drawing did not specify the type of joint nor the weld reinforcement
required.
However, the specified
fitting is a "Weldolet" and as such has an exisitng weld prep on it and requires no additional
design work. Also, the size of the fillet weld cover is specified
in the welding procedure
for this type of full penetration
branch line connection.
The problem arose during the review of the RIC forms for the four branch connection
welds. Although these are full penetration
single bevel groove welds, with fillet weld reinforcement, the RIC form labels these welds as "F.W." indicating
a fillet weld. For Gap Thickness, the RIC form specifies "NA" which would be appropriate
for a fillet weld but not for a full penetration
weld. Since this attachment
must be a full penetration
weld, there was no documentation able to assure that the proper penetration
has been achieved using the fied fillet weld. Additional
review by the inspector
of the NDT Examination
Reports revealed another deficiency.
According
to liquid penetrant (PT) examination
report sheet No MKV-01, welds No 2 and No 13 on line
1/2 did not receive a PT examination
as required by
Specification
M-152(Q) "Field Fabrication
and Installation
of ASME Section Xi Piping fication in a Nuclear Power Plant," Revision 14, September
30, 1986, paragraph
MI0789-1683A-TC01-NL02
41 .... ,.. .,.. . .... '* .. -. . --.. *.**-:**.'... . **.* .. -.
.. *. :*'. "*. -* ,c;, ' *.** '"' ,' * -' ' --
9.1.1. Pending verification
that all four branch attachment
welds are full penetration
welds and resolution
of the PT
this is considered
an Unresolved
Item (255/89007-05).
CPCo Response Reference
NRC Violation
255/89007-02a.
MI0789-1683A-TC01-NL02
42 . -=--***_µ_*
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ATTACHMENT
2 Consumers
Power Company Palisades
Plant Docket 50-255 LIST OF REFERENCES
August 10, 1989 1 Page ' ... *.* .. *.* .. ' .: ... :***_-..
References
.. lo Plant Administrative
Procedure (AP) 9.11 "Engineering
Analyses" --i I 2. AP 9.03 "Facility
Change" 3. AP 5.06 "Control of Special Processesn
4. AP 9.06 "Code Requirements
for Maintenance
and Modifications" 5. AP 9.04 "Specification
Changes" 6. AP 9.30 "Q-List" 7. Deviation
Report D-PAL-89-43
-..*. MI0789-1683A-TC01-NL02
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