ML18058A569
| ML18058A569 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 06/22/1992 |
| From: | Schapker J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML18058A567 | List: |
| References | |
| 50-255-92-12, NUDOCS 9206290040 | |
| Download: ML18058A569 (26) | |
See also: IR 05000255/1992012
Text
U. S. NUCLEAR REGULATORY COMMISSION
REGION III
Report No.:
50-255/92012(DRS)
Docket No.:
50~255
Licensee:
Consumers Power Company
1945 West Parnall Road
Jackson, MI
49201
Facility Name:
Palisades Nuclear Power Plant
Inspection At:
covert, MI
49043
License No.:
Inspection Conducted:
February 25-27, April 3, 9, 27-28, 30, and
May 27-28, 1992 (Region III)
Inspector:
Inspection Summary
23 through April 3, 1992. (Region I)
es Section
6 ..-..,_ 'L- --?' .,__
Date
Inspection on February 25-27. April 3, 9, 27-28, 30, and May 27-
28, 1992 (Region III); March 23 through April 3. 1992 (Region I)
(Report No. 50-255/92012(DRS)).
Areas Inspected:
Routine announced .inspection of inservice
inspection (ISI) activities including review of program (73051),
procedures (73052), observation of work activities (73753), data
review and evaluation (73755), and inspection of the ventilated
concrete cask (VCC) fabrication performed on the Palisades site*
(37700).
Selected areas inspected by the NRC Mobile NDE
Laboratory were the safety injection and refueling line (SIRW),
main steam (MSS), feedwater (FWS), chemical volume control (CVC),
containment spray (CSS), long term cooling (LTC), and safety
injection (SIS) systems.
Results:
Of the areas inspected, two violations of NRC
requirements and eight unresolved items were identified in the
ISI areas.
Inspections of the VCC fabrication identified some
quality control/quality assurance weaknesses and lack of
contractor oversight resulting in an unresolved item.
9206290040 920622
ADOCK 05000255
G
.
The NRC inspectors noted the following:
Implementation of ISI was generally considered weak.
Examples of deficiencies include omission of a number of
.welds from the ISI program, incomplete documentation of
examination results, and some questionable examination
techniques.
-
0
Observations of the fabrication of the vcc identified
examples of work activities and QA/QC.implementation which
failed to meet applicable specifications and standards.
oversight of the contractor activities on site was weak .
2
': .-
--
DETAILS
1.
Persons Contacted
Consumers Power Company CCPCol
- J. Amthor, Construction Superintendent
- D. Engle, Senior Engineer
- R. Smedley, Staff Licensing Engine~r
- T. Newton, Senior Nuclear Operations Analyst
- T. Fouty, Senior Nuclear Operations Analyst
- V. Beilfuss, Assistant Outage Manager
R. Vanwagner, ISI Supervisor
s. Wellman, NDT Project Supervisor
Pacific Sierra Nuclear Corporation CPSN)
- W. Lee, Vice President
+J. Massey, Project Manager
J. A. Jones
F. Slatton, Project Superintendent *
- u. s. Nuclear Regulatory Commission CNRCl
- J. Heller, Senior Resident Inspector
- Denotes those attending the exit interview on May 28, 1992.
+Attended the exit interview via telecon.
Other licensee and contractor personnel were contacted
during the course of the inspection.
2.
Inservice -Inspection CISI) (73051, 73052, 73753, 73755)
a.
Eddy Current (ET) Examination of Steam Generator Tubes
The licensee replaced steam generators (SGs) in the
1990-91 outage. * This was the first examination for the
first operating cycle of the replacement SGs.
The
inspection included full length tube examination of
1,599 tubes in the "A" SG and 1,597 tubes in the "B" SG
utilizing a bobbin coil probe and multifrequency (MIZ-
18) eddy current examination data acquisition system.
3
Results of the examination identified no1 defective
tubes in either SG; however, degradation was
identified in 45 tubes.
While the results of the ET
examination appeared to exhibit substantial degradation
for the first operating cycle, comparison with the
baseline ET performed prior to installation revealed
that little, if any, degradation had actually occurred
during the operating cycle.
The degradation recorded
during this inspection was identified during the
baseline inspection and had not substantially changed.
The NRC inspector observed ET in progress, verified
calibration/certification of ET equipment and
calibration standards, and reviewed the
examiner/analyst qualification certifications.
The NRC
inspector also reviewed the ET Data Analyst Guidelines
and observed evaluations being performed by the
analysts.
The licensee has contracted with Westinghouse Electric
corporation (~) for a Steam Generator Reliability
Program.
This program includes ET services, chemistry
control, cleaning services, and general maintenance to
reduce and control degradation in the SGs.
The licensee's SG reliability program appears to be
conservative in assuring the safe operation of the SGs.
b.
Observation of ISI Work Activities (73753)
The NRC inspector observed the following ISI work
activities in progress:
UT examination of reactor vessel closure head
studs.
MT examination of reactor vessel stud nuts and
washers.
PT examination of SG feedwater piping.
1 Technical Specifications require those tubes with excess of
20% wall thinning (wastage) be reported as degraded, t_ubes in
excess of 40% are considered defective and must be repaired or
plugged.
2 ET work activities observed are described in Paragraph 2.a
of this *report.
4
Work activities were* performed in accordance with
approved procedures, utilizing calibrated NOE equipment
and certified personnel.
Detection: and evaluation of
flaws detected.by NDE procedures were completed in
accordance with ASME Code and regu-latory . requirements.
c.
ISI Documentation Review* (73755)
The *irac inspector reviewed documents relating to the
following:
0
0
0
0
0
NDE procedures for wor~ activities observed.
NOE personnel c.ertification:s in accordance with
SNT-TC-lA.
NOE equipment calibration and examination reports.
Liquid penetrant mat.erial certifications.
ET examination reports.
ET data analysis guidelines.
No violations or deviations were identified.
3.
Inspection of the Ventilated Concrete Cask Fabrication at
the Palisades Site
a*-
Background .
The licensee contracted with Pacific Sierra Nuclear
Corporation (PSN) to design and construct a dry cask
spent fuel storage facility to be partially constructed
- onsite for long term temporary storage of spent fuel.
The licensee i.s currently documenting a 10 CFR 50.59
evaluation as required by 10 CFR 72.212 (Subpart K),
showing that use of .the general license .for storage of
spent fuel at the power reactor site will not involve
an unreviewed safety question or Technical
Specification (TS) change.
The PSN cask desig~ consists of a steel multi~assembly
basket (MSB) which holds 24 spent fuel assemblies
(sealed) and a steel clad ventilated concrete cask
(VCC) which provides biological shielding and MSB
protection.
The PSN cask design has not yet been granted a
Certificate of Compliance by the NRC which would add it
to the list of approved spent fuel.storage casks in 10
CFR '}2.214.
This inspection was conducted in terms of
5
b.
the specifications, standards, codes, and commitments
described in the licensee's request for design
certification.
Inspection*
The NRC inspector's initial observation of work
activities onsite was the welding of the ventilation
ducts to the steel liner of the VCC.
The NRC inspector *
reviewed procedures, specifications, and qualification
records for the welding in progress.
Specification CVCC-89-001, Revision 1, dated February
1991, entitled "Fabrication for The Ventilated Concrete
_Cask" was reviewed.
Section 3.10, "Steel Fabrication
and Welding"'" required that all welding be perforpied to
approved drawings, and. that welders and weld inspectors
be qualified/certified to AWS D.1.1.
The NRC inspector's review confirmed the use of AWS
approved weld procedures and welders; however, no AWS
certified weld inspector was assigned onsite to perform
inspections.
The vendor (PSN) subsequently initiated a
Document Change Notice (CPC-041). dated April *14, 1992,
which deleted the AWS b1.1 requirement in the
fabrication specification.
The safety significance of this deletion was minimal
because the welds in question serve no structurally
significant purpose.
However, the inadequate
implementation of the specification requirements is
considered a weakness in both contractor implementation
and licensee oversight of contractor activities.*
Inspection of the. first concrete pour (April 3 O, -1.992)
identified wea.knesses in training of the craft in
performance of their assigned task.
The NRC inspector observed the placement of concrete in
the VCC forms.
The consolidation of the concrete was
being performed with mechanical vibrators.
Initial
placement/consolidation was being performed by placing
the vibrators in the concrete and vibrating to both
consolidate and move the concrete within the forms.
Using the vibrators to move the poured concrete
violates the requirements of the American Concrete
. Institute (ACI) Standard 301.
The NRC inspector
questioned the craft to determine the extent of their
knowledge and training in the use of vibrators.
The
craft was unaware of the ACI requirements with respect
to moving concrete with the vibrators *
6
The NRC inspector immediately informed the contractor
.
. superintendent that the craft were improperly using the
vibrators.
The contractor superintendent took prompt
corrective action and instructed the craft in the
proper method of consolidation and placement of
concrete in the form.
Due to the prompt. identification
of the improper practice and corrective action taken
.during the initial placement in the concrete forni of
the first VCC, the NRC inspector believes that no
segregation of concr.ete materials was likely and
therefore, the concrete placement was*acceptable.
The NRC inspector also observed placement of concrete
in the fourth VCC (May 28, 1992).
Prior to the day of
the pour, the contractor informed _the NRC.inspector
that the PSN representative had been reassigned to
another project.
The NRC inspector inquired as to who
the PSN inspector would be for the-.upcoming pour of the
fourth vcc.
No inspector was assigned at that time; .
however, because of the NRC's iriquiry, the PSN
inspector was required to report to the site to perform
inspections of the pour.
Discussions at the exit
meeting with the PSN Project Manager disclosed that .*the
inspections for the concrete pour we~e waived by PSN,
and the licensee and the fabrication contractor were to
perform the inspections. .However, the fabricator did
not have any individuals who were qualified or
sufficienctly independent to perform the inspection.
Subsequent discussions identified that the licensee had
no knowledge of this agreement; and, no one had been
_assigned by the licensee to perform the inspections.
Furthermore, the licensee had not, a$ of this
inspection, performed any onsite audits or
.. surveillances of the contractor* s work*.
The fabrication contractor's (J. A. Jones) standard
operating procedures for fabrication of the vcc were
reviewed by the NRC inspector prior to the first pour
of VCC #1.
These procedures describe the inspection
responsibilities of the contractor superintendent.
No
inspection procedures for PSN were available for review
onsite; however, the PSN inspector had completed an
inspection check sheet and recorded the appropriate
dimensions and inspection characteristics.
The NRC inspector also reviewed the Quality Assurance
Data Package for VCC #1, #2 and completed portions of
- 3.
CVCC-89-001, Paragraphs 4.2.3 and 4.2.4 required
the vendor to use fabrication travelers and record all
critical dimensions in the traveler however, the
process control sheets or travelers were not completed .
7
i
j
,.
Hold points and witness points were not signed by the
J. A. Jones superintendent or the .PSN inspector.
In summary, contractor oversight and QA/QC
implementation was weak.
Examples of weaknesses
identified during this inspection include failure' to
follow procedures, insufficient training of craft, and
inadequate definition of QA/QC responsibilities.
These
weaknesses are collectively considered an unresolved
item and a written response describing the licensee's
evaluation of the extent of the problems and corrective
actions taken is requested (50-255/92012-11)
Subsequent to the exit interview on May 28, 1992, the
licensee issued a Stop Work Order on May 29, 1992
(Enclosure 4).
No violations or deviations were identified .
.
'
4.
Unresolved Items
5.
Unresolved items are matters about which more.information is
required in order to ascertain whether they are acceptable
items; violations, or deviations. An unresolved item
disclosed during the inspection is discussed in Paragraph
3.b of this report *
Exit Interview
The inspector met with licensee representatives (denoted in
Paragraph 1) at the conclusion of the inspection on May 28,
1992.
The inspector summarized the scope and findings of
the inspection activities.
~he licensee acknowledged the
inspection findings.
The inspector also discussed the
likely informational content of the inspection report with
regard to documents. or processes reviewed by the inspector
during the inspection.
The licensee did not identify any
such document/processes as proprietary .
8
Docket No ..
License No.
Licensee:
ENCU'.>suRE 3
U.S. NUCLEAR REGULATORY COMMISSION
REGION 1
Consumers Power
212 West Michigan Avenue
. fackson. Michigan 49201
Facility Name:
Palisades* Power Station
Inspection At:
Covert. MI
Inspection Dates:
February 24 to March 6. 1992
Inspectors: . . * . Ham{
hmc1an
Mobile NDE Laboratory, EB, DRS
P. M. Peterson, Technician
Mobile NDE Laboratory, EB, DRS
. D. M. Wiggins; TET~ Inc.; Mobile, Alabama
W. M. Mingus; TET, Inc.; Mobile, Alabama
ApprOved by: ~
_,*
'
~NDE
Laboratory,
Engineering Branch, DRS
Dale
2
Inspectfon Summary and Conclusions: An announced inspection was conducted at the
Palisades NuCiear Power Station ~uring the period March 23 to* April 3, 1992, using the
NRC Mobile NDE Laboratory (Report 50-255/92~012). The purpose of this inspection was.
the independent nondestructive evaluation (NDE) of components and welds in order to
ascertain the accuracy of the NDE performed by the licensee. 'During this inspection forty-
eight (48) welds and/or components located in 7 pl3.!lt.systems were examined by
nondestructive methods by the NRC. J'he results obtained by the NRC during this
- evaluation, when compared with those of the licensee, were not consistent with the variations
- expeeted* in teehnique and revealed an ISi system with weaknesses.
Areas InspeCted: Selected areas from the safety injection and refueling line (SIRW), main
steam (MSS), feedwater (FWS); chemical volume control .(CVC), containment spray (CSS),
long term cooling (LTC)., and safety injeetion (SIS) system were independently examined by
the NRC utilizing various nondestructive methods. An evaluation of the licen~ee's quality
program, including NDE, was performed using NRC promulgated procedures iri conjunction
. with the licensee's approved procedures.
. Results: . The inspection. found a number of weaknesses in the ISi system at Palisades. These
weaknesses were contained in both the programmatic and testing portions of the ISi system. *
In addition to the violation *for failure to identify or examine welds and the viol~tion of.
inadequate volumetric coverage there are eight unresolved findings. * The vi?lations are *
discussed in Paragr~phs 2.3 and 3A.
3
1.0
Introduction
The Code of Federal Regulations (CFR), Title 10, Part 50.55a (10 CFR 50.55a),
requires inservice inspection (ISi) of safety related equipment to identify system*
degradation. Before the licensee generated program *of inspection is applied- to the
-equipment it must be approved by the Nuclear Regulatory Commission (NRC) under
the authority embodied in 10 CFR 50.55a (g) (4) (iv). The required inspections are
qetailed in the American Society of Mechanical Engineers .(ASME) Boiler and
Pressure Vessel Code,Section XI for Inservice Inspection as embraced in 10 CFR
50.55a (b). The NRC inspection described in this report was made using the Mobile
Nondestructive Examination (NDE) Laboratory. The Mobile NDE Laboratory is
capable of independently performing the examinations required of the licensee. This
capability provides the NRC with unique insights into the licensee's inservice -
inspection program and on a sampled basis, the adequacy and accuracy of the -*
licensee's specific NDE inspections.
The scope of this inspection was to review the administrative portions of the program
and to perform NDE of the systepls that were avfillable. *
2.0
Inservice Inspection Program Review
2.1 -
NDE Procedures (73052)
Before a license's program of inspection is used, it must be approved by the
Nuclear Regulatory Commission under the authority embodied in 10 CFR
50.55a (g) (4) (iv). The required inspections are detailed in the American
Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,
Section XI for Inservice faspection as embraced in 10 CFR 50.55a (b). For
any inspection program, the code edition and addenda used is determined in
- accordance with the requirements of 10 CFR 50.55a (g). For Palisades Power
Station the applicable code edition is 1983 with the Summer of 1983 _addenda
(a. k.a., 83S83). It was found during this inspection that the following
procedures had been revised to comply with the 1983 Edition with Winter of
1984 Addenda (a.k.a.,83W84) editic,m of the ASME code even though this
edition was not approved for use at the plant by the NRC: NDT-PT-01, Rev 9;
NDT-PT-02, Rev 5; NDT-RT-01, Rev 8; NDT-MT-01, Rev 7; NDT-UT-02,
Rev 4; NDT-UT-08, Rev 2. These procedures were revised in this manner in
-
-
order to be able to use them uniformly through out the Consumers Power
system; both at Big Rock Nuclear Power Station, Palisades Power Station in_
addition to their fossil fueled plants. No evidence was available showing that
a technical review of the procedures had been undertaken to determme if the
revision impacted any technical specification or quality commitment at the
plant. This is unresolved (50-255/92012-01) pending the licensees review.
Paragraph IWA 2120 and Mandatory-Appendix III, Paragraph 2300 (p) *
requires the approval of the ANI for all NDE procedures. This requirement
for approval includes, in the former paragraph, the requirement that each
'_,
procedure "shall include, as a minimum, the following information:* ... (p)
. approval of the procedure as reciuired by IW A 2120." This requires some
form of testimony on the procedure, that the ANI has approved the procedure.
The simplest way of fulfilling this obligation is by. having. the ANI sign* off the
procedure a~ a matter of course. Palisades. had no such testimony on the
procedures. The only proof that the ANI was involved in the process was a
. sheet with his signature listing a number of instances where each procedure*
had been used during an actual inspection. These inspections were dated .
approximately 11h to 5 months after the procedure was issued. Although the
code does not state specifically that the procedure must l:)e approved by the
ANI before use, the method submitted by Palisades does not preclude the use
of the procedure for 5 months without ANI approval. Had the ANI
disapproved the procedure it would fall to Palisades to prove that the*
procedure had not been used in the interim or, failing that, reinspect all the
items that had been inspected in the interim. Pending the licensees revision of
. the procedure to make_ them plant specific, this will remrun an unresolved
finding (50-233/ 92012-02).
The radiographic procedure at Palisades (NDT-RT-01, Revision 8) listed a . *
table in paragraph 10.6.3.l for double wall radiography. This procedure, as
noted earlier, . was revised to the Winter '84. Addenda to Section XI. The table
of paragraph 10.6.3.1, is a reflection of an earlier edition of ASME Section V,
Article 2.. The code has since dropped the table. It is no longer applicable to
Section V *(referred to by Section XI for radiography) either for the
Summer '83 or Winter '84 Edition of the code. According to the licensee the
table remained since the procedure was used in a fossil plant where it was still
applicable under ASNI B31.l for power piping. The table did not conform to
the current requirements of ASME Section XI and Section V, since it required
the use of a 25 penetrameter at 2T sensitivity for the thickness range of 1 to
and including 11/2 inches and the code requires that a 20 penetrameter at 4T
sensitivity be used in the range of 1 to and including l 1A inches, The licensee
was unable to give a satisfactory technical justification for this deviation from
the code and it remains as an unresolved item (50-255/92012-03).
5
The following procedures were reviewed by the inspection team:
Title
No./Rev.
Visual Examination
NDT-VT-01, Rev. 10
Liquid Penetrant Examination.
NDT-PT-01, Rev. 9
Liquid Penetrant Examination
NDT-PT-02, Rev. 5
Nonstandard Temperature
RadiOgraphic Examination of Welds
NDT-RT-01, Rev. 8 -
Ultrasonic Examination of Ferritic
NDT-UT-01, Rev. 8
and Austenitic Piping and Branch
Connection Welds
Magnetic Particle Examination
NDT-MT-01, Rev. 7
Ultrasonic Thickness Measurements
NDT-UT-02, Rev. 4
Ultrasonic Examination of Nuts, Studs,
NDT-UT-07, Rev. 4
Bolts and Pins
Ultrasonic Examination of Vessel Welds
NDT-UT-11, Rev. 1
Ultrasonic Examination of Nozzle-To-
NDT-UT-12, Rev. 1
Vessel Welds and Nozzle Inner Radius
Sections
Ultrasonic Examination of Reactor
NDT-UT-08, Rev. 2
Coolant Pump Flywheels of the Palisades
Nuclear Plant
Each procedure was reviewed for compliance with ASME Section XI and V
requirements to the 83983 edition. In addition, the results of examinations
were compared with these procedures for compliance with stated
commitments. With the exception of the previously noted findings, the
procedures were found to be in compliance with the requirements.*
6
2.2
ISI Program <73051)
It is the intent of ASME in formulating the requirements of Section XI for an
inservice inspection program to track iri.dications for trends; as well as detect
them ... This is reflected in ASME Section XI, Paragraph IWB 3131 (for class
1 and 3 piping) and Paragraph IWC 3131 (for class 2 welds).* Detailed in
these references is the stipulation to compare the current state of an indication
with its previous state. Volumetric and surface. examination results "shall be
compared with recorded results of the preservice examination and prior
. inservice examinations." Th.is is further strengthened in IWB 3131.1 and IWC
3131.1 when it'states that if the "volumetric or surface examination either
reconfirms the absence of flaw indications or reveals flaw indications that do
not exceed the acceptance standards listed . . .
11 the component shall be deemed *
acceptable for continued service. Palisades was not doing this analysis for all
indications; just those that were rejected. This makes Palisades far less able to
trend indications in the plant and develop an early warning system for any
generic category of indication. This is an unresolved finding pending the
licensees review of reports (50-255/92012-04).
The licensee has a written system that identifies a method of orientation for the
-- recording of indications on the final reports as required in ASME Section XI, *
Paragraph IW A-2610, and .this system meets the intent of the requirement. .
However the identification system does not include any requirement to
- permanently mark the weld centerline. This is an important basis for
repeatability of the NDE and is further required in mandatory* appendix III,
paragraph III 4320. It is important enough for the code to dedicate a full page *
supplement to appendix III describing suggested ways of accomplishing the
marking system. The licensee has centerline marked welds as part of
modifications in safety systems but does not mark existing welds. -Pending the
licensees evaluation of a method for conclusively determining the centerline of .
. existing welds this will remain an unresolved item (50-255/92012-05).
2.3
- ISi Data Review (73755)
The NDE evaluation of weld flaws, for acceptance, under Section XI is based
on the stress of the_ indication in a plane. If two flaws are located sufficiently
close to _each other they are considered to be one flaw, This evaluation is
required in Paragraph IW A 3300 as diagramed in Figure IW A 3330-1. The
final Palisades ISi penetrant report for weld ESS-12-SIS-2Bl-1 noted two, with
the assumption that round penetrant indications are as deep a 112 their
diameter). This is also an unresolved finding pending the licensees revision of
the report form and/or penetrant procedure to include the requirement to
determine flaw separation (50-255/92012-06).
Consumer Power NDE procedure NDT-UT-01 Rev 8 (not one of the -
procedures revised to 83W84) paragraph 10.6 requires the axial shear wave
scanning of welds-in two dire.ctions on the crown of and along the weld. The
7
procedure under paragraph 10.5, in compliance With Section XI para 111-4430
(a), requires the entire weld and HAZ be scanned by the longitudinal method
sides. According to the note on the ultrasonic report for weld ESS-12-SIS-2Bl
the weld could only be examined from one side due to restriction of a valve
taper on the opposing side. In addition the report notes that the_ longitudinal
scan could not be done due to weld crown configuration. If the longitudinal
scan could not be done with a 1/2" transducer it follows logically that a
transverse examination with the additional interference of a 45 degree
transducer shoe could not have been done. The failure to do the 45 transverse .
examination is not noted on the report nor is there any description as to what
is done with this weld since only one of four examinations was performed. It *
is very difficult to determine exactly what surface has been used to scan and
what direction since the requirement of ASME Section XI, para III-4500(t) is
not complied with i.e.: "The_following data shall be recorded on an
examination data sheet: ... (t) surface from which examination is conducted".
This weld examination had been. accepted by Consumers Power. Failure to
adequately document the inability to perform an inspection of the Code.
required weld volume is a violation of 10 CFR 50, Appendix B, Criterion
XVII (50-255/92012-07).
2.4
ISi Performance (75753)-
An 80% of DAC reflector (intermittent in length) was not recorded for weld
FWS-18-FWL-2Sl-243 per the requirements of ASME Section XI, para IWA-
2232 (b), CP Procedure NDT-UT-OL Rev 8, para 11.1 and ISi plan Rev 3,
Appendix 6A, Parts Band E. The licensee was unable to duplicate the NRC's
results in this regard. The equipment used by the licensee and the NRC was
essentially the same and the transducer was the same type, brand, size and
. frequency from the same manufacturer. The NRC Mobile NDE Laboratory*
subjected the two transducers to a frequency spectrum analysis. As can be
seen in Figure 1, below, the NRC transducer displays an evenly changing *bell
curve that contains no lobes, while in Figure 2 the licensee transducer shows
side lobes. The indication to the right in the diagrams is the amplitude display
from the same side drilled hole reflector used to obtain the spectrum scan.
..
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Figure 2
The
11 A
11 scan is the one normally seen on the ultrasonic instrument. As can
be seen there is no apparent difference in the A scan. However the spectrum
display clearly shows that the damping material in the licensee transducer has
loosened to such a degree as to create ringing. This ringing leads to a .
reduction of sensitivity of the transducer and explains why the licensee* was
unable to duplicate the results of the NRC. The spectrum analysis performed
by the NRC is available on the instrument used by the licensee. The NRC
considers this another unresolved item since the licensee needs to dispose of *
the indication in the above weld and determine what other transducers might
be in inventory that suffer from the same deterioration. (50-255/92012-08).
9
Consumer Power procedure NDT-UT-01 Rev 8,. para 7.2.2.2 D. states:
"Search unit size, frequency, and angle shall generally be selected according to*
Table 1." Table 1 for a nominal material thickness of 1.00" to 1.5" lists a
transducer size *of 3/8" to 1 /2" iri size at an angle of 45 ° or 60 °. The
thicknesses listed on the final ultrasonic examination reports for welds* MSS-
36-MSL-2S 1-219, MSS-36-MSL-2Sl-217 and MSS-36-MSL-2Sl-218, are less~ .
than 1.5". According to the Consumer Power procedure a 1/2" or less
transducer would have been used. There are no weld, geometry or
. interference restrictions which would have caused any other choice of
- transducer. However the welds were examined with a. l" diameter transducer.
The large 45° shoe of this transducer (2 1/8 ... long by 1 11/16" wide) along
with the wide crown of the minimally prepared surface of the* weld meant that
the required coverage of the 1/3 zone could .not be obtained on the first 1/2 *
node. Instead. of exercising the options contained in_ ASME Section XI Para
III-3230 (a) (1) and/or (2) along with.Consumer Power procedure NDT-UT-
01, Rev 8, para 7.2.3 A. and/or B., which would have guided them back to
the smaller transducer of their Table 1, they decided to exterid. the scan to a
full 1 and 1/2 nodes in order to obtain coverage. This means a sound path
that is 3 times longer than a smaller transducer with the inherent lose of
sensitivity to small indications.
The NRC examined weld 219 with a 1/2" diameter transducer using the first
. 1/2 node. This examination revealed an indication that is 26 '1 long and up to
100 % of the reference level. The nature of the indication seems to be
geometric but would require further examination to more fully categorize.
Consumer Power was made aware of this discovery and repeated the
- examination with the 1" transducer and the transducer that was later
.
.
determined by the Nl{C to suffer from a loss of sensitivity. This remains an
unresolved item pending further evaluation by the licensee (50-255/92012-09).
3.0
Nondestructive Examinatfon (NDE)
During this inspection 48 welds and/or components located in 7 plant systems were
examined by nondestructive methods by the NRC. Selected areas from the SIRW,-
MSS, FWS, CVC, CSS, SIS, and LTC were independently examined by the NRC
utilizing various nondestructive methods. An evaluation of the licensee's quality
program, including NDE, was performed using NRC promulgated procedures in
conjunction with the licensee's approved procedures .
10
3.1
,Visual Examination (57050)
3.2
Thirty-Nine (39) safety related pipe weldments and adjacent base material (1/2
inch on either side of the weld) were visually examined in accordance with
NRC procedure NDE-10, Rev. 0, Appendix A; and associated site procedure
NDT-VT-Ol_Rev.10, QC documents, isometrics and as-bui~t drawings.
Examined-during this inspection were ASME Class 1, and 2 pipe weldments
selected from SIS, MSS, CVC, FWS and CSS. Inspections were performed
specifically to identify any cracks or linear indications, gouges, leakage, -arc
strikes with craters, or corrosion, which may infringe upon the minimum pipe
-wall thickness and modifications to piping or components. Mirrors, flash
lights, and w_eld gauges were used to aid in the inspection and evaluation of _
the examination.
Results: The welding and overall workmanship inspected was satisfactory. No
cracks, gouges, corrosion or any indication that may be detrimental to the
operation of those systems insp~ted, were revealed. *
Inspection Hanger/Support (57050)
During this inspection fiftee~ (15) safety related hanger/supports were visually
inspected per NRC procedure NDE-:10, Rev.O, Appendix A and B in
conjunction with site procedure NDT-VT-oi-R.ev.10, and QC documents and
associated isometric/drawings. Included in this inspection were hanger/supports
selected from the SIS, and CVC system. In the area of welds, the accessible
surface area and adjacent base metal for a distance of one-half inch on either
side of the weld was examined. In the area of component integrity, specific
attributes examined were proper installation, configuration, or modification of
supports, evidence of mechanical or structural damage, corrosion, bent, and
missing or broken members. Table #2 is a list of sp~ific hanger/supports
inspected._
Results: The overall condition of hanger/supports inspected was satisfactory.
-
-
-
1
-
3.3
Liquid Penetrant Examination (57060)
Twenty-six (26) safety related pipe weldments and adjacent base material (1/2
inch on either side of the weld) were examined using the visible dye, solvent
removable method per NRC procedure NDE-9, Rev.O, in conjunction with the _
licensee's procedure NDT-'-PT-01 Rev.9. Included in this inspection was
-
ASME Class 1 and 2 carbon and stainless steel pipe, ~ite _field and vender shop -
weldments selected from the SIS, CSS, CVC and LTC.
Results: The surface areas examined were properly prepared for the
examination. Interpretation of indications located in the same area needed to
be strengthened.
-
-*
11
3 .4 . Ultrasonic Examination (57080)
Nineteen (19) safety related pipe weldments were ultrasonically examined using NRC
procedure NDE-1, Rev. 1 in conjunction with the licensee's procedure* NDT-UT-01,
Rev. 8 and associated isometric drawings and ultrasonic.data reports. included in this.
_examination were ASME Class 1 and 2 pipe weldments selected from the SIS, FWS
and MSS. To* obtain the greatest possible repeatability, in performing the NRC
independent measurements the examinations were. performed utilizing transducers and
cables that matched those used by the licensee as closely as possible. A distance
amplitude correction curve was established utilizing Consumer Power calibration
standards 21 PAL, 44 PAL, 50 PAL and 14A PAL per procedure NDT-UT-.01, Rev.
8.
.
Results: The ultrasonic examinations performed by the licensee were at best
minimum by the ASME Code standard with many of the examinations performed
from one side of the weld. To be ctble to perform the required volumetric
examination the technician had to inspect areas of the weld using the third leg (or * 1
112 node) of the ultrasonic sound path. This. may cause problems in highly
attenuative materials such as stainless steel.
Section XI of the Code requires that longitudinal welds be examined at the
circumferential weld intersection for a length of 2.5 times the wall thickness. Upon
examination of welds MSS-36-MSL-ISI-211, MSS-36-MSL-2Sl-212, and MSS-36-
MSC-2Sl-219 on the main steam system, it was noted that the intersecting
longitudinal welds had not been examined and were not included in the licensee's ISi
program. Failure to examine these welds as required by the Code is a violation of 10
CFR 50.55a (50-255/92012-10).
Subsequent to the NRC identification of this issue, the licensee identified an add~tional
27 longitudinal welds which had not been included in* the ISi program. The licensee's
immediate actions were to perform the required examination of these welds.
3.5
Radiographic Examination (57090) *
. Radiographs for nine (9) *safety related pipe weldments recently completed in the *
SIRW line *were reviewed by the NRC inspectors. These included welds FW-110 to
arid including FW-118. The procedure and technique used to make the radiographs
was in accordance with site procedure, NDE-:10, Rev. 18. Approximately one-
hundred and twenty radiographs were reviewed for weld integrity, technique and
- radiographic quality. Also reviewed were associated isometric/drawings and
radiographic data reports.
12
Results: Although no violations were identified and the radiographs and welds met the
stated requirements, it was noted that the reader sheets were very confusing. The *
industry practice of including one weld, and sequence of repairs, on one reader sheet
or radiographic report in one package of radiographs is not practiced at this plant. It
makes it very difficult to track the welds. In addition the welds are radiographed by
placing the film around the weld in keepirig with the fossil practice of 0, 60 and 120
degrees or 0, 120 and 270 degrees (depending on pipe size) even though the section
markers are in even increments of 1 inch. This leads to very odd sequences of for
example 4-12 on one weld and 2-10 on the next, since the 120 degree mark is not the
same as the 0 inch mark on each weld.
4.0
Erosion/Corrosion (49001)
Concerns regarding erosion/corrosion in balance of plant piping systems has
heightened as a result of the December 9, 1986 feedwater piping line rupture which
occurred at Surry. This event_ was the subject of the NRC Information No-tice 86-106,
issued December 16, 1986, and its supplement issued on February 13, 1987.
The licensee's actions with regard to the detection of erosion/corrosion in plant
components were reviewed with respect to NUREG-1344, "Erosion/Corrosion
Induced Pipe Wall Thinning in U. S. Nuclear Power Plants", dated April 1989,
Generic Letter 89.;08 issued May 2, 1989, and NUMARC Technical Subcommittee
_ Working Group on Piping and Erosion /Corrosion Summary Report, dated
June 11, 1987. The licensees -procedure, EM-09-08, Rev.2, outlines the minimum
requirements for the erosion/corrosion program. Tbe areas that were reviewed by the
NRC were system selection, component selection, ultrasonic inspection, and data
evaluation.
The selection of systems to be model~ were based on the systems susceptibility to
erosion/corrosion (flow velocity, steam quality, material of the components and
temperature). The parameters used to identify those systems are similar to those used
by EPRI for the CHECMATE program. The computer program used at Palisades to
model the systems is Krautkramers PIPE program. Fourteen systems were selected to
be modeled by the computer program. Modeling determines the ranking of
component susceptibility to erosion/corrosion. The component ranking goes from the -
highest projected wear rate to the lowest projected wear rate. Only the components
with the highest wear rates will be inspected for wall.thinning.
-
Ultrasonic thickness inspection is used to locate areas of wall thinning. -Five
components, (item numbers 6.8.BB, 5.2, 9.lB, 6.18B and 6.18Bl) were
independently inspected by the NRC to determine the accuracy of the licensees
measurements. The thickness readings the NRC obtained* matched those of the
licensee .
The ultrasonic data was entered into Krautkramers DA TAMATE computer program
for reporting and evaluation. If the component had a remaining wallthickness greater
than code minimum wall, it was accepted. If the remaining wall thickness was less
- than code minimum, engineering analyses is performed per Palisades J\\dministrative
13
Pr.ocedure 9.11. * Accordirig to the Palisades program, EM-09-08, if the highest
ranked components are found acceptable after three consecutive outages, the system
no longer needs to be inspected for wall thinning
Results The program has a potential to be comprehensive. . One weakness in the
program is the failure to update the computer model with current plant data. Not
updating the model will give false predication of time to thickne~s critical or T crit.
Another weakness is the failure to inspect components with low susceptibility ranking.
The use of engineering judgement and industry experience could highlight low
ranking components which have had problems in the past and may never be inspected
according to this program. .
An effective erosion/corrosion program should be a "living" procedure with_
continuously updating and inputting data to accurately predict component .failures.
Palisades erosion/corrosion program does not have the flexibility to change with the
plant changes. *
14
5. 0
Examination Results
I
NRC INDEPENDENT MEASUREMENTS
I
TABLE N0.1
WELD ID. No.
SYS
NONDESTRUCTIVE TEST I
SHT.# 1 I
on.
ISO/DRAWING
LIN
E
CL
RT EI:JWEJWEJ
ESS-6-SIS-lBl-4
1
x
x
x
x
ESS-2-SIS-lAl:.6
1
x
x
x
ESS-2-SIS-lAl-7
1
x
x
x
ESS-2-SIS-lAl-18
- l
x
x
x
ESS-(i-SIS-1B6-204
2
x
x
x
ESS-8-SIS-2B6-201
2
x
x
x
ESS-6-SIS-lHP-224
2
x
x
x
ESS-i2-SIS-1A5-218
2
x
x
x
ESS-12-SIS-1 eS-220
2
x
x
x
ESS-12-SIS-leS-221
2
x
x
x
ESS-12-SIS-2Bl-2
2
x
x
x
x
ESS-12-SIS-2Bl-1
2
x
x
x
x
ESS-2-LTe-1 B-32
LTe
. 1
x
x
x
ESS-2-LTe-lB-33
LTe
1
x
x
x
eVC-2-LDL-2B5-2
eve
1
,
x
x
x
eVC-2-LDL-2B5-1
eve
1
x
x
x
FWS-1~-FWL-lSl-244 *
FWS
2
x
x
x
FWS:-18-FWL-2S 1-243
FWS
2
x
x
x
MSS-36-MSL-lSl-211
MSS
2
x
x
x
MSS-36-MSL-2S 1-212
MSS
2
x
x
- X
MSS-36-MSL-2Sl-219
MSS
2
x
x
x
..
15
.,
NRC INDEPENDENT MEASUREMENTS
TABLE NO. 1
WELD ID. No
SYS
NONDESTRUCTIVE TEST
SHT.# 2
ISO/DRAWING
LIN
CL
M,
REJ
E
T
c*
ESS-14-CSS-IPA-201
css
2
x
x
x
~
ESS-14-CSS-IPA-202
css
2
x
x
x
...
ESS-l 4-CSS-IPA-203
css
2
x
x
x
ESS-14-CSS-IPA-'208
css
2
x
x
x
ESS-10-CSS-IPA-209
css
2
x
x
x
ESS-14-SIS-LPA-216
.2
x
.x
x
MSS-36-MSL-lSI-217
MSS
2
x
x
x
MSS-36-MSL-2SI-218
MSS
2
x
x
x
- MSS-36MSL-2SI-213
MSS
- 2
x
x
x
MSS-36-MSL-1SI-210
MSS
2
x
x
x
MSS-36-1 SI-MSL-216
MSS
2
x
x
x
'
FWS-l 8-FWL-lSI-242
FWS*
2
x
x
x
FWS-l 8-FWL-2SI-242
FWS
2
x
x
x
MSS-36-MSL-lSI-208
MSS
2
x
x
x
MSS-36-MSL-2SI-208
MSS
i
x
x
x
ESS-6-SIS-CHR-225
2
x
x
x
x
- ESS-6-SIS-CHR-226
2
x
x
x
x
- ESS-lO-CSS-lPA-227
css
2
x
x
x
16
,.
NRC INDEPENDENT MEASUREMENTS PROGRAM
HANGER/SUPPORTS
TABLE2
IDENTIFICATION.
SYS
CL
AC ' REJ
COMMENTS.
c
ESS-12-SIS-lBl-7PR
B
x
ESS~ 12-SIS-2Al-4PR
A
x
ESS-12-SIS-2Al-6PR.
A
x
ESS-12-SIS-2Bl-6PR
B
x
ESS-6-SIS-2HP-PR 1
B
x
ESS-6-SIS-2HP-PR2
B
x
ESS-6-SIS-2HP~ PR3 *
B
x
ESS~6-SIS-2HP-PR4.
- SIS
B
x
ESS-6-SIS-2HP-PR5
B
x
ESS-6-SIS-CHR-236PR
A
x
ESS-6-SIS-CHR-234-1
A
x
ESS-6-SIS-CHR-234-2
A
x
ESS-6-SIS-CHR-223PR
B
x
ESS-6-SIS~CHR-226PR
B
x
ESS-14-CSS-IPA-201PR
css
B
x
. *~
. '"
17
6.0
Persons Contacted-
Consumers Power Company
Tom Fouty
Tom Newton
Dennis Ziegler
Richard Humphrey
James Schepes
P. Donnelly .
R.Orosz
Richard Kasper *
. R. Van Wagner
James Kuemin
Systems Engineer
Nuclear Operations
Special Projects
NDT Supervisor
Assessment Specialist
Plant Safety and Licensing
Nuclear Engineer
Maintenance Manager
Systems Engineer
Licensing
U.S. Nuclear Regulatory Commission
R. Rodin
John Jacobson
Jerry Schapker
Jacque Durr*
Resident Inspector Riii
Chief, Engineering Brah ch Riii
Reactor Inspector Riii
Chief Engineering Branch RI
The above listed personnel were present at the.exit meeting. The inspector also
'
contacted other administrative and technical personnel during the inspection.
7.0
Unresolved Items
Unresolved items are matters about which more information is required in order to
ascertain whether they are acceptable items or violations.
8.0
Management Meetings
Licensee management was informed of the scope and purpose* of the inspection at the
entrance interview on March 23, 1992. The findings of the inspection were discussed
with the licensee representatives during the course of the inspection and presented to .
licensee management at the exit interview. The licensee did not indicate that
- proprietary information was involved within the scope of this inspection.
..
.
'**
..
..... *
- 'II
13:58
PRLI~;ADES RES.: DEl'ff WSPEC. CIFF,
consumers
Power
l'llWllUR
M1£11111All'I,.,.,,,,,...,
ENCLOSURE.4
- P*llHdH Nual11r Pte111: 27780 llut 9ttr Memori*I lilg!iwey, Ccwtn, Ml 49043
JCA 34*92
May 29, 1992
John V. Massey
Sierra Nuclear Corporation
S619 Scott* Vall*y Drive #240
Scotti Valley, CA
9'066
,
Subj*ct:
Palisade* Nuclear Plant
Dry Fuel Storage Projfllct
F'. 02
A.a a r**ult of the NRC audit exi.t meeting May 28, 1992, uveral i.uuu were
id*ntified regarding th* ~**pon~1bi11ti*1 and implementation of the Quality
Anul:'ance Plan for fabrication (if 'dry futl atorage ca*.ka.
Therefcu, in
acccrd.ance with our t*hphone c~:inve*uation today, .May 29, 1992 at 11:30 a.m.
EST, you are direot*d to.cea1e B11 work, effecti~e immediately, r~lating to
fab:i;'ication and con1t~uction of oa1ka until thas*. iaauea are 1atiefaccorily
~~*olved. You will' be notifi*d in writing by this office when work activities
may b* reconvened,
Sincer*ly,
John c. Amthar, P, E.
1
.
Canatruction S~p*r1nt*ndent
CC: ==.w~l
Rl>Oro1z, Pali1ade1
TJPalmi1ano 1 Pali1ad**
JPPomar&n1k1, Pali1ade1
OJVandaWalle, Pali1ad**
ERVanHoof, P26*302B