ML20011D708
| ML20011D708 | |
| Person / Time | |
|---|---|
| Site: | Zion File:ZionSolutions icon.png |
| Issue date: | 12/19/1989 |
| From: | Danielson D, Schapker J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20011D707 | List: |
| References | |
| 50-295-89-28, IEB-88-008, IEB-88-8, IEB-89-001, IEB-89-1, NUDOCS 8912280297 | |
| Download: ML20011D708 (9) | |
See also: IR 05000295/1989028
Text
_ _ - - ,
- , , - , - - - - . .
~
-
,
,,
.
.
,
U. S. NUCLEAR REGULATORY COMMISSION
-
REGION III
,
Report No.:
50-295/89028(DRS)
Docket No.:
50-295
License No. DPR-39
,
Licensee: Commonwealth Edison Company
Post Office Box 767
Chicago, IL 60690
.
Facility Name: Zion Nuclear Power Station - Unit 1
Inspection At: Zion, IL 60099
Inspection Conducted: August 23-24 September 25-28, October 10-11, 26-27,
,
November 21-22, and 29, 1989
Inspector:
M
~
/.2// 9/d-f
J. F. SchapKer
Da'te '
Approved By* 2T/Wbhdhm
/kMlf?
D. H.
Janielson, Chief
Date
l
Materials and Processes Section
Inspection Summary
i
Inspection on August 23-24, September 25-28, October 10-11, 26-27,
November 21-22, and 29 1989 (Report No. 50-295/89028(DR5))
Areas Inspected: Routine unannounced inspection of inservice inspection
activities including review of program (73051); procedures (73052); observation
~
of work activities (73053); and data review and evaluation (73755); and licensee
action in response to NRC Bulletins (92703).
Results:
No violations or deviations were identified within the areas-
inspected.
The licensee's ISI program, procedures, work activities, nondestructive
examination results and interpretations were in compliance to the ASME.
Code,Section XI, 1980 Edition, Winter 1981 Addenda.
.
!
The licensee demonstrated a positive commitment to assure the
reliability of the steam generators.
The licensee's eddy current examination program exceeds the regulatory
requirements for sampling.
Steam generator transition girth weld examination (ultrasonic / magnetic
particle) detected ID cracks, which were dispositioned in accordance with
the ASME Code,Section XI, 1980 Edition, Winter 1981 Addenda.
8912280297 891220
{DR
ADOCK 05000295
_.
1
-
..
.
.
.
DETAILS
<
1.
Persons Contacted
Commonwealth Edison Company (Ceco)
'
- T. Rieck, Technical Staff Superintendent
- H. Peterson, Regulatory Assurance
- B. Wulf, ISI Group Leader
- B. Kurth, Production Superintendent
T. Saksefski, Regulatory Assurance
B. Wilson, SMAD Level III
J. Ramage, Planning Supervisor
A. Panagos, Nuclear Engineer
G. Olson, QA Engineer
R. Sumners, Technical Staff, 151
L. Laspisa, Assistant Technical Staff Supervisor
Westinghouse Electric Corporation (W)
,
B. Lefebure. ISI Coordinator
J. Delbusso, ISI Inspector
Combustion Engineering (CE)
E. Jackson, Steam Generator Service Manager
Conam Inspection Services (CIS)
M. Gortemiller, Level III Analyst (ET)
G. Herrera, Level IIA
U. S. Nuclear Regulatory Commission-(U. S. NRC)
- W. Kropp, Senior Resident Inspector (Byron)-
J. Smith, Senior Resident Inspector
R. Leemon, Resident Inspector
- Denotes those attending the exit meeting on November 29, 1989.
Other
station technical and administrative personnel were contacted during the
course of this inspection.
2.
Followup on NRC Bulletins
a.
(Closed) NRC Bulletin 295/88-08; 304/88-08: Thermal Stresses in
Piping Connected to Reactor Coolant Systems (RC5).
On December 9, 1987, while Farley 2 was operating at 33% power, the
2
i
!
'
..
.
-
.
.
.
licensee noted increased moisture and radioactivity within containment.
The unidentified leak rate was determined to be 0.7 gpm. The source
of leakage was a circumferential crack extending through the wall of
an unisolable section of the emergency core cooling system (ECCS)
piping that is connected to the cold leg of Loop B in the RCS. This
section of piping consists of a nozzle. two pipe spools, an elbow,
and a check valve. The crack resulted from high-cycle thermal
fatigue that was caused by relatively cold water leaking through a
closed globe valve at a pressure sufficient to open the check valve.
The leaking globe valve is in the bypass pipe around the boron
injection tank (BIT).
During normal operation this valve and others
isolate the ECCS piping from the discharge pressure of the charging
pumps. With a charging pump running and the valve leaking, temperature-
stratification occurred in the ECCS pipe.
In addition, peak-to-peak
amplitudes as large as 70*F occurred and with periods between two and
twenty minutes.
.
In response to this bulletin, the licensee took the following actions:
(1) The licensee reviewed the systems connected to the RCS for
unisolable piping that could be subjected to the thermal cycling
phenomenon for Zion Units 1 and 2.
These susceptible sections
of piping were identified as:
One (01) Alternate Charging Line, RC065'(3 inches nominal
l
pipe size);
One (01) Auxiliary Spray Line, RC145 (2 inches nominal pipe
size); and
Four (04) Charging Pump to Cold Leg Injection Lines: RC065,
RC038, RC071,_and RC079 (1.5 inches nominal pipe size).
(2) The licensee performed nondestructive examinations (NDE) on the
above piping, welds and heat affected zones (HAZ). NDE. included
ultrasonic examination (UT) (volumetric) and liquid penetrant
examinations (LPT). The UT examinations were performed on all
base _ metal, welds, and HAZ that were accessible for examination.
LPT was performed on sockolet welds.
(3) The licensee also installed temporary temperature monitoring
devices on the Alternate Charging and Auxiliary Spray: lines to
'
identify the temperature _ profiles downstream of-potential
leaking isolation valves. The temperature profiles are to be
taken immediately upon unit startup and after manipulation of
the subject valves.
The NRC inspector observed UT calibrations reviewed during the
examination procedure and reviewed the results of temperature
monitoring recordings installed on Unit 2 during the previous outage.
No indications of cracking were identified. The temperature
monitoring indications were within the designed material requirements.
3
._
_
_
1
.-
,
.
.
.
The licensee's actions taken in response to the bulletin are adequate
to assure the RCS attached piping has not degraded due to thermally
induced stresses.
b.
(0 pen) NRC Bulletin 295/89-01:
Failure of Westinghouse Steam
Generator Tube Mechanical Plugs
Background
Numerous plants have experienced primary water stress corrosion-
cracking (PWSCC) and leaks of Westinghouse mechanical plugs. On
February 25, 1989, North. Anna, Unit 1, experienced a mechanical
plug failure following a reactor trip during a feedwater isolation
transient. The plug failure caused a 75-gallon per minute (gpm)
,
primary-to-secondary leak and was the subject of NRC Information
,
Notice No. 89-33, " Potential Failure of Westinghouse Steam Generator
Tube Mechanical Plugs." The failure mechanism involved a full
circumferential severance of the top portion of the plug from the
i
body of the plug. The top portion of the plug was propelled up the
length of the affected tube by primary system pressure to a point
just above the U-bend tangent point where it impacted and punctured
the outer curvature of the tube. The top portion of the plug
subsequently impacted and dented an adjacent tube. The failed plug
was installed in November 1985.
Licensee Action
Steam generator maintenance records were reviewed to identify the
installation date, location, and heat number for all installed
mechanical plugs.
Using the methodology of WCAP-12244 Revision 1
and the benchmark degradation rate for Milstone 2, estimated plug
lifetimes were determined for the susceptible heats identified by
Plug lifetime estimates are determined for each plug
size as a function of the hot leg and cold leg operating temperatures.
-
Based on this estimated plug lifetime, Unit 1 susceptible hot leg
plugs were removed during this outage. Susceptible cold leg plugs
will be removed / repaired, if necessary, during.a future outage. The
projected life is beyond the year 2000 for these plugs.
Inspection
The NRC inspector observed removal and repair of the PWSCC
susceptible plugs installed in the hot leg tubesheets of Unit 1
steam generators. The inspector-informed the licensee'of the need
'
for a tracking system for the cold leg PWSCC susceptible plugs which
have not been removed, though these plugs could exceed the steam
generator life expectancy as malyzed.- The inspector will followup
on this item during the next Unit 2 refueling outage. The licensee's
corrective action for the hot leg plug repairs was satisfactory.
4
-
.
.
. _ .
.
.
.
3.
Inservice Inspection (ISI), Unit 1
a.
General
>
This is the second outage of.the second period of the second ten-year
plan. CECO contracted Combustion Engineering (CE) to perform the
eddy current examination (ET) of the steam generator (S/G) tubes.
,
Westinghouse Electric Corporation (W) performed the ultrasonic (UT),
liquid penetrant (PT), magnetic particle (MT), and visual (VT)
examinations for the ISI, in accordance with the rules and
requirements.of the ASME Code,Section XI, 1980 Edition, Winter 1981
Addenda.
The nondestructive examinations were performed in accordance with
.
approved procedures which were reviewed by the authorized nuclear
inservice inspector (ANII) and approved by the licensee's Level III
inspector who is also certified to the EPRI standards for UT.
b.
Programs and Procedures
The NRC inspector reviewed the following nondestructive examination
-
(NDE) procedures:
1
Procedures
Revision
Title
ISI-8
9
Visual Examination (VT)
ISI-11
11
Liquid Penetrant Examination (PT)
ISI-47
4
Ultrasonic Examination (UT) of
Vessel Welds
ISI-70
2
Magnetic Particle Examination (MT)
ISI-206
1
Manual Ultrasonic Examination of
151-10
6
Qualification of UT Equipment
ISI-41
5
151-88
5
Underwater Remote VT
CE/STD-410-049
1
' Eddy Current Examination (ET)
of Steam Generator Tubes
STD-500-002
4
Welding the 7/8" Steam Generator
Tube Sleeves
OPS-NSD-101
5
Preservice and Inservice
Documentation.
Zion 400-001
0
UT of Tube to Sleeve Upper Weld
Zion 400-002
1
Visual Examination of Steam
Generator Tube Sleeve Plug and
Procedures reviewed adequately described examination requirements
and complied with ASME Code Section XI, 1980 Edition, Winter 1981
Addenda.
l
5
. _ - .
. -
- - . ___
_ _- _ ___ _ _ _____ _ __________ _
-
.
.
.
c.
Review of Material, Equipment, and Personnel Certifications, and
NDE Data
The NRC inspector reviewed documents relating to the following:
Ultrasonic instruments, calibration blocks, transducers, and
UT couplant certification,
,
Certification of liquid penetrant and magnetic particle
materials.
Eddy Current (ET) equipment calibration.
NDE personnel certifications in accordance with SNT-TC-1A.
NDE reports for ISI performed this outage.
ET data reports.
ET analyst examinations.
ET report for previous outage.
No violations or deviations were identified.
d.
Observation of Work Activities
The NRC inspector observed work and discussed examinations with NDE
examiners. These activities included observation of calibrations,
examination performance, and review of documentation of the
following:
Liquid penetrant examination of charging pump aiping welds.
Ultrasonic examination of steam generator girt 1 weld.
Magnetic particle examination of the ID surface steam generator
'
girth weld.
Visual examination of the steam generator secondary side (ID)-
girth weld, feedwater ring and supports.
Ultrasonic examination of welded sleeves installed in steam-
generator tubes.
Work activities were performed with approved procedures, utilizing
calibrated NDE equipment. Detection and resolution of indications
disclosed by ISI procedures are discussed in Paragraphs 4 and 5 of
this report.
4.
Eddy Current Examination of the Steam Generator Tubing
The licensee employed CE to perform the eddy current examinations of the
steam generator-tubes as required by the-Zion Technical Specifications (TS).
CE utilized the MIZ 18 multi-frequency acquisition and DDA-4 analyzer
system to conduct the examinations. A computerized evaluation using
Zetec auto-evaluation programs was used followed by evaluations by
Level IIA or Level III evaluators. Sorts and parameters were set to
over-call indications with the auto-evaluation. All' relevant indications
were called by the auto-evaluation method. The initial evaluation
performed by CE and Zetec analysts (Levels IIA and III) re-evaluated the
entire tube with emphasis on the auto-evaluator's calls. A second
independent evaluation was performed by Conam level IIA and Level III
analysts without the aid of the auto-evaluator,
i
I
6
._.
_
'l
.
.
.
.
The inspection included 100% of all steam generator tubes fr,m the hot
leg tube end to the seventh support plate in the cold leg, a 25 tube
sample of the row two tubing U-bend utilizing a motorizing ritating
pancake coil (MRPC) probe, and the following full length (tube sheet to
>
tube sheet) ET's:
S/G "A"
S/G "B"
S/G "C"
S/G "D"
1,022
1,146
1,126
973
The licensee also performed fifty additional MRPC's to quantify
indications found by the bobbin coil examination. The licensee identified
42 tubes with indications in excess of TS requirements (more than 40%).
These tubes were plugged or sleeved.
Sleeving was performed on the tubes
which had degradation in excess of the plugging limit but were
technically capable of being sleeved.
Additional plugging and sleeving was performed on tubes having ET
-
indications which may exceed the plugging limit in the future. The total
of tubes plugged in each Unit 1 steam generator as a result of the ET is
as follows:
Reason for Plug /Siteg3
Total This Outage
Tech. Spec.
Plugged
Sleeved
Required
Preventive
(A) 1RC100
18
82
1
99
(B) 1RC400
50
138
28
160
(C) 1RC200
15
191
4
202
(D) 1RC300
11
34
9
35
The NRC inspector observed the eddy current examination (ET) of the steam
generator tubes in progress, verified cortification of ET equipment,.
calibration standards, probe travel speed qualification,.and reviewed
the qualifications and certification of the ET examiners, including site
specific analyst training and qualification.
The NRC inspector observed the plugging and sleeving of steam generator
tubes. The licensee installed welded sleeves fabricated by CE this outage
in accordance with Zion Technical Specification Appendix A, Section 4.3.6.
Observations of cleaning, inserting, expanding, welding, and ultrasonic
examination / visual examination of completed welds were performed to
approved procedure requirements. Review of pre-installation cleaning,
inspection, welding and NDE procedures was performed. The licensee's
preventive maintenance program included sludge lancing and maintaining
chemistry controls in accordance with prescribed guidelines. The eddy
current examination (100%) program exceeded Technical Specification
requirements, which requires sampling in accordance with;TS Table 4.3.B-1.
The licensee also performed the MRPC examination described above to
assure circumferential cracking in the U-bend interior rows (1 and 2)
are not developing. This examination was performed with no indications-
of circumferential cracking in the two U-bends (row one U-bends were
previously plugged in the Unit 1 steam generators).
7
-
--
.
.
.
.
The licensee's eddy current program exceeds TS requirements which
provides that a high confidence of steam generator integrity is
,
maintained for safe operation.
5.
Examination of the Steam Generator Transition Cone Girth Welds
!
The licensee performed volumetric examination of the steam generator
examination (UT) girth weld of steam generator"D", Unit 1.
transition cone
Ultrasonic
of this weld detected 34 indications (more than 20%
Distance Amplitude Curve (DAC)). Twelve of these indications were
verified as surface indications by internal magnetic particle examination
(MT) of the ID surface. Three of the subsurface indications exceeded the
50% DAC and did not meet the acceptance criteria of the ASME Code Section
XI,1980 Edition, Winter 1981 Addenda tables (IWB-3511-1). Consequently,
the licensee examined the three remaining steam generators' transition
cone girth welds. The results of the UT examinations were:
17 indications
in steam generator "A", 21 indications in steam generator "B", and 9
indications in steam generator "C".
All the UT surface indications confirmed by magnetic-particle testing
were removed by grinding and blending.
Complete removal of all surface
indications was verified by MT. The remaining imbedded indications which
exceeded 50% of DAC were dispositioned by fracture mechanics to
demonstrate their acceptance by the criteria of the ASME Code,Section XI,
1980 Edition, 1981 Winter Addenda, Paragraph IWB-3600.
The follcwing table described the bounding subsurface indication in each-
generator (all dimensions in-inches):
Generator
Ind
2a Dim
1 Dim
S Dim
Y-Dim
a/1
a/t(%)
A
13
0.06
0.80
0.58
1.0
0.38
8.2
8
2
0.37
1.70
0.52
1.0
0.11
5.1
C
9
0.35
0.50
1.48
1.0
0.35
4.7
0
32
0.47
0.625
0.45
1.0
0.38
6.4
The above indications as well as all embedded indications which exceeded
the standards of Table IWB-3511-1 were subjected to a fracture
evaluation, using the guidelines of Appendix A_of Section XI. The
analyses completed were specific to Zion steam generator girth welds.
The detailed technical basis for these analyses was the same as that used
for the Byron and Braidwood steam generators previously submitted to and
approved by the NRC (W WCAP-11063).
.
The NRC inspector observed the ultrasonic / magnetic particle examinations
of the transition cone girth weld (steam generator "D") including
calibration, setup and indication disposition and sizing. Visual
inspection of the ID of the girth weld and feedwater ring, nozzle and
supports was performed by the NRC inspector. The NRC inspector
identified one area on the feedwater support which was suspect due to
visual abnormalities.
The licensee magnetically examined the area with
no apparent defects observed.
Pitting and cracking was visually
8
. . -
.
.
,.
.
observable on the ID of the transition cone girth weld. These
indications (cracks) were removed and a metallurgical sample was taken
for root cause analysis from one of the crack indications. The worst case
crack was approximately 6.5 inches long and was excavated at .5 inches
in depth. Final magnetic particle examinations of the excavations
were acceptable.
The NRC inspector reviewed the summary report of the evaluation of the
steam generator girth weld UT indications, dated November 13, 1989
(DRAFT). The indications detected by ultrasonic examination were plotted
with respect to the known weld configuration and classified in accordance
with ASME ig_e Section XI, 1980 Edition, Winter 1981 Addenda.
The NRC inspector notified NRR/EMTB upon identification of the girth weld
indications and participated in conference calls with the licensee and
NRR concerning fracture analysis of subsurface indications which
exceeded-ASME Section XI standards, and cause analysis _of the ID surface
cracks.
.
The NRC inspector informed the licensee that the NRC will followup on
future examinations of these welds as determined from NRR review and
assessment of the flaw evaluations.
The analyses and corrective action taken by the licensee concerning the
cracks and embedded flaws complied with the applicable ASME Code and
regulatory requirements.
6.
Exit Interview
The NRC inspector met with site representatives (denoted in Person's
Contacted Paragraph) at the conclusion of the inspection. The inspector
summarized the scope and findings of the inspection noted in this report.
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
documents / processes as proprietary.
9
.
w -
, - ,