ML20148F879
Jump to navigation
Jump to search
| ML20148F879 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 06/03/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20148F862 | List: |
| References | |
| 50-445-97-09, 50-445-97-9, 50-446-97-09, 50-446-97-9, NUDOCS 9706050082 | |
| Download: ML20148F879 (17) | |
See also: IR 05000445/1997009
Text
{{#Wiki_filter:.-
'
1 e l l
-
ENCLOSURE 2
i
'
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
i
Docket Nos. 50-445
50-446 i
1
License Nos.. NPF-87 !
NPF-89 )
Report No.: 50-445/97 09 I
50-446/97-09
Licensee: TU Electric
1
Facility: Comanche Peak Steam Electric Station, Units 1 and 2 !
l
Location: FM-56
Glen Rose, Texas
Dates: March 17-21,1997
Inspector: 1. Barnes, Technical Assistant
Accompanied By: Dr. C. V. Dodd, NRC Consultant
Approved By: Arthur T. Howell, Director ;
Division of Reactor Safety
ATTACHMENT: Supplemental information
l l l
'
9706050082 970603
PDR ADOCK 05000445
0 PDR ,,
r ! *
.
-2-
!
EXECUTIVE SUMMARY
Comanche Peak Steam Electric Station, Units 1 and 2
l NRC inspection Report 50-445/97-09; 50-446/97-09
Maintenance
*
Independent review of eddy current data from Refueling Outage 1RF05 concurred
with the final disposition call on all but one tube examined. The difference was
relatively minor (i.e., inside diameter defect location versus outside diameter
location) and of significance only because it would signify the probable presence of
a primary water stress corrosion crack (Section M1.1).
Enaineerina
*
The licensee performed comprehensive steam generator tube exarrinations during
Refueling Outage 1RF05, with the use of conservative examinatior expansion
criteria and adoption of new eddy current examination technology considered
indicators of management support for steam generMor tube. integrity initiatives
(Section E1.1).
*
The total number of tubes plugged during Refueling Outage 1RF05 was small
(i.e.,19), but considered of some significance because of the first time detection of
defect indications that are normally found on laboratory examination to be
secondary side stress corrosion cracking (Section E1.1).
*
The licensee has developed comprehensive steam generator eddy current program
requirements since the initial 1994 NRC inspection. Areas of program strengthening
included use of two separate parties for data evaluation, a requirement to use
Appendix H qualified examination techniques, use of quantitative data quality
evaluation criteria, and development of a steam generator health report for
identifying program and degradation status, chemistry trends, visual inspection
results, and future plans (Section E1.2).
*
Areas noted where program improvements could be made pertained to requirements
for monitoring and assessing effects of loose parts and the current lack of use of
eddy current graphics for depicting characteristics of degradation modes that could
be potentially encountered. A violation was identified with respect to the approval
of an acquisition technique specification sheet for plus point data, without requiring
l technique requalification, which changed the qualified Appendix H essential variable '
for probe and extension cable (Section E1.2).
)
_ .-. - .-
.
.
.
-3-
*
The programmatic requirement for analysts to be certified as qualified data analysts
.was considered a program strength. The licensee had compiled an appropriate pool
of eddy current test data for verification of individual analyst ability to make correct
interpretations of data (Section E1.3).
*
The licensee appropriately responded to the results of external assessments of eddy
current program requirements (Section E1.4).
, i
s
_ - . . .. _. _ . . _ _ .. . _ _ . - _ _ - . _ _ _ . _ _ _ _ _ . . .__ ._ *
u l '
\
I
-4- )
'
i
'
l
l
1
Report Details i
This inspection was performed as a followup to an initial baseline inspection of steam
generator programs, history, and material condition, which was documented in NRC
' Inspection Report 50-445;-446/94-01. The initialinspection' identified examples of where
the established program requirements, for examination of steam generator tubing, did not
fully reflect applicable industry guidance and generic communications. The current
inspection utilized Inspection Procedure 92903 and technical requirements contained in
!
Inspection Procedure 50002.
I
Summary of Plant Status !
Units 1 and 2 were at 100 percent power during the inspection period.
II. Maintenance
M1 Conduct of Maintenance
l M1.1 Review of Tube Examination Data 1
a. Inspection Scope .
The inspector selected a limited sample of Unit 1 eddy current data from the last : i
refueling outage, Refueling Outage 1RFOS, for independent assessment by the NRC )
consultant. Included in the assessment scope were defective calls by the primary 1
and/or secondary analyst which were overruled by the resolution analysts, data
anomalies that were initially identified during Refueling Outage 1RF04 examinations,
and Refueling Outage 1RF05 defective calls that were plugged. i
i
I
b. Observations and Findinas
The NRC consultant reviewed the plus point and pancake coil eddy current data that
was obtained during Refueling Outage 1RFOS from the following tubes in Steam
Generator 3: Row (R) 2, Column (C) 21; R23, C45; R9, C30; R2, C34; R3 C34; l
l
and R2, C35. These tubes were identified as containing circumferential defect
! indications by the primary and/or secondary analyst, with the final call by the l l resolution analysts being no detectable degradation. The NRC consultant did not l ! disagree with any of the calls by the resolution analysts, and noted that the defect l ,
calls by the production analysts appeared to result from the intersection of a non- '
j defect signal with the tube expansion transition region.
During the Refueling Outage 1RF04 eddy current examinations in 1995, a number
- of tubes were identified which exhibited signal anomalies in the data produced by a j
' 4 Zetec Delta probe from the top of the tube sheet region. This type of probe
,
contained a 0.115-inch pancake coil, an axial sensitive coil, and a circumferential
.
'
i
.
___ . _ -. . - . - . . - -
I
*
,
I
-
l
(
.s.
1
i
!
sensitive coil. These tubes showed long signals in the axial direction that had
" bumps" at regular intervals. The NRC consultant reviewed the 1995 data from the
following Steam Generator 2 tubes that exhibited this anomaly: R45,C43;R47, I
C55; R49, C51; and R49, C61. The phase of these signals was observed to not )
rotate like a defect above or below the tube sheet as the frequency was changed.
However, the intersection of these signals with the expansion transition region did
produce a signal on some of the frequencies that could be interpreted to represent a
defect with an approximate inside diameter location. The licensee concluded in i
1995, after detailed review and performance of supplementary ultrasonic
examinations, that the signal anomalies were not evidence of the presence of
defects. The NRC consultant reviewed the data produced for the same tubes in the
1996 Refueling Outage 1RF05 examinations. These examinations were performed
using a probe containing a plus point coil and two pancake coils. (See Section E1.1
below for additional probe information). The signal was noted to be present in the
data from the pancake coils, and greatly reduced in the plus point data ( particularly
away from the expansion transition region). Figure 1 shows an example of the
signal anomaly as exhibited by the plus point coil. Figure 2 shows the phase for an
anomaly at different frequencies, in some instances, a small, low-voltage
defect-like signal could be found on the plus-point channel in this region, but it was
well within the noise. The NRC consultant concurred with the licensee position that
the available information did not support classification of these signal anomalies as
defects. The inspector considered the only feasible method for determination of the
reasons for the anomalies would be removal of a tube sample from a steam
generator for laboratory examination.
The NRC consultant reviewed the eddy current data from the following tubes, which
had been called as containing defect indications and subsequently plugged: Steam
Generator 2 - Tubes R23, C44, R43, C47, and R1, C98; Steam Generator 4 - R26,
C104 and R34, C98. All of the calls appeared to be correct with one exception.
The NRC consultant determined that Tube R34 C98 of Steam Generator 4 appeared
to contain an inside diameter axial crack rather than the outside diameter axial crack
that was called. Eddy current information for this defect indication is shown in
Figures 3 and 4. There were two axial cracks noted to be present, with one of
them close to 100 percent through the wall. Most of the data indicated an inside
diameter indication, with the 0.080-inch pancake coil further confirming inside
diameter cracking by exhibiting a larger signal at a 800 kHz frequency than that
shown at 600 kHz.
+
e
-6-
.t. * * * un t I * ut r un * * *
2 l Luttrnst lft151 l Stat 14RY l tit %N&t l Stat tuint l u n'.a v w.e
1.51 8 1 400 Rhs 3D McT at ste '. Jd 3' J SLAT ated 0 CF f 5t ia l34' th i D 9tl=3 /
C1 l Sie il RDW 43 C OL 51
PP5: Stu 38/ 33 1 ?!CF5/5CHN Pl ! C H = 0. 040 L I RC *2. sf 06 I
g, ,
247 kftt 5f T _2f B0 SUB SCHN OF F 500 TRhCE OFF
(SLHN PLOT vtRT t tttP OLLf M*l.ee v 11Lk THH 164
~ICk
i CH. 8 R3LfW CFF 'FC* 72' #D=236' CIR STEPsiO* I NTE IN
F t *64 f vPE ## F L att Hng l e a Los t D;M. 100 MEH5UHE Of f [[gyggg lgyg l gig l{ g igy
SCAN TRr0 ON SET TPlG OF F fx_C I RC.0f f '$PffD ***** s a/ ses TRAlp
.N
Vpp 67$8 DIMI of
% /
[ ch
- s \
+2.57 131.9 3 *
N ' ~
-
+4. 92 187. 3
s
O
,
1
+ 2. 62 187.1 l
Figure 1 Anomaly at the top of the tube sheet for tube Row 49 Col 51 ,
1
1
.
.
.
-7-
.
l
i
I
, , . . . . . . . . . . . , , . . .. . . . . . . . . . .
.
.
-
-
i
inni ,m st enont ona l orinnt j vrn ;
i ,,,,,... i ,,,,, , , - . . . p. ,m. rs< ii i i. i-i>.<,--
-
.... .
., . . 4. ..,,
.., . . ,.
,, ,, ,, ,, , , , ,,
, .m . ..,. . ,,,, .
,
-
S/. 2 unif B , NL L T
>
/ pl
n la
v
[iitarl
p av,
in
a a"n
>
,,, . . i..... .I .,n ...
n .....
.
.
.
v,ro, ,rr ,r- .r v,rtn ,rn .r
f f
- %
~.s. % 3, ,
% l
l ., . . .
.,
, ,, ,. . .
J '
... .. .... ,
C$ C I
.
9
3
s{
.- 1
1.x.z. ss 1
v., , or or n rs a wrri. m >ii .- ;;; ., ..,;;;;; ,;;;
('
.
/ *t 7 9;:
M'
w
,
sx .cr
c
y
n
;
' = -
;
.
-
;
.
-
1
f -p
~
C
i{, gc Jg ,r,
-
I
-
Figure 2 Phase at different frequencies for a anomaly
l
,
1
I
l
;
!
4 y -a,a,1J , - -w m -m- - = m 4.- ---.- e a
l 8 ! l I l I l l i !
l Statte l M fvP l AUTB_RNat fill STament ant ePfitf t ( UtiL
fY l AtFDPI i l 8trott # l (8HPRWE l filif l SMd7 4$Y l MRttuim6 l s e u wwo % ** * Up t f * en tuu = ==
, 83W LP 3 V 4 b) 8 8 688 the its et in/ *, Jdt's M.ML t sceu # W i % i at 31' ubiu lepa/5
2; 6 25 C I tilf
g y ,, 34 ggg 9,
_ ,,g; fg,16/ 79 i flCh 5s W eme Pl f C*it. e U C I RC +2. 26 y , ,
315 88*9 51 f _2E RO SU6 St rue DF F S28 'hACE ON
(5CAN PLOT wtwr (p DE t t h= t ,00 v ilCE foM leg
_ is th (n. 6 RM EH Of f NJe 3h* NGalhr llh STEPulea jg } g. gg
Fl*h IvPE uh F L 44 #as te a Cm l OlN. In MLem Of F jattui nTt
- -
nit )t im5
5tHN TRAL'n UN
7eE lotees $18
'
ce sgf Tal6 Wf #M_C I RC J3F F SPf f B t
l _
f k
I t
l -
I
'
l > ,
~
v., i .T. eini .er
L
,_J
r
i
_
j Q
- - - _
,_
A
~
l
~8
~
. i -
-~ ' -
-,
*5.67 219.7
f
" '
*l.87 226.2
l *i.91 238.4
_.
_
_
- . - - - _- _ _ _ _ _ _ _
Figure 3 Defect at the top of the expansion transition showing an id phase at 600 kHz
i l I, l I l i
e
o
9
. . , o ., s .., ,,. s , ..,,,m,,..
l Mstle j st rur j auf s.,nwat v5:5 timment ent sr fit RL l ul L
l nae l af reet t l af ront t l (entnet l f tist l sumns? l ni na,,( l Maf f al.E j w.a en < t. ..s i,,i t . ni rna se
j, a3v tr I w d el e J Det she )* o f .i ti f
.15 " >4ixr;at 3t coa as e or e st i . is' wk. L Intes. /5
2: S 2: 6 C 5 Sif f _ pp.g fp yg , 3,3,,n { P1Trweop L j a( .2. eet,
% H m M m n
,
[ * ' " ' ~ *' *
J i'r Pm I it f 2E PO i sue stah of f suli 'kWE Of F
' '* j '
,
' DEL' A. t se y *
'
L5cy P ct l
i vtp? ty f i ta T** 164 I" '
*I '., I
_
f!Cs (d. E] *g d8 7 Wa* 6 7' ftc.14' CI A $7EPele' gyg g g g
3'u f *PE un }8rm 49 ee H l Con t 04 les ME PSoNt OF F
('( sig st uff NTI 1,mik
._ ,_ ,_ _ _
M aN f *M F. On j St i IRif, UP F l Ha_LikC JIP F 5pg g s .~ , W l,f..,~ ygMi
-
g - . _ . _ . . _ _ _ _ . . . _ _ _ _ .
t ! !'
-
4 '
.
,
. .n an ,
sn-
,
y / /. w,:.c
7 : e.%7am%
s e ~a
1 . Gcidf9eE53hhh
m;::%.fMS.yg; I
Q
'~
A .kA*$ht y:?
)$
'&:.
['
,
t
. . ,6sg~e;nwA
-
s
e w
pansd,.,g
ws -
k m; y:t y:r -wm
p a wd.~ 41%@
cs s i
r
& $ :h':kh khkj h b$$f $h}5 ''5fM
~
Y
<
2. - ' b[
'
[ } #
* N #'
~ ?)':hQhW
j'& & ;: h ik : \[_'?'KhSi$RRfNW$5ll{}ll'
~ N
f(Qj
z - ' ' *[Qf %Q5-h,'N'& ~ '?{?;hWyhW?
' -
W:'
_
^
-
.: :h'2m C %.s %f' Ti, 'dO'0'S $h .
V:WlklND'l$ Mc
"2 M '
-
-
'
Nkh[i' NNhkk$h'dh
~
y J'%'DN nd3'ijsy3:y N :
4.hW'iNUgh(-' kNh * <sM Se,
N
._ .s.se tu g- +
.
g;g%pywy&
-
Qd;kh![fQM.phP'~
- y :.d h@SN-
-
-
6,46 28.I
~
. I
Figure 4 Defect at the top of the tube support showing an id phase at 300 kHz
c. Conclusions
Independent review of eddy current data from Refueling Outage 1RF05 concurred
with the final disposition call on all but one tube examined. The difference was
relatively minor (i.e., inside diameter defect location versus outside diameter
location) and of significance only because it would signify the probable presence of
a primary water stress corrosion crack.
t
l
.
O
- 10-
111. Enaineerina
E1 Conduct of Engineering
E1.1 Review of Refuelino Outaae 1RF05 Tube Examination Scope and Results
a. Inspection Scoce (92903)
The inspector reviewed the tube examination scope and methods that were used in
Refueling Outage 1RF05 with respect to Technical Specification requirements,
industry guidance, and as a result of emerging degradation modes. A review was
also performed of the tube plugging data for this outage and the reasons for tube
plugging.
b. Observations and Findinas
The inspector was informed that the licensee initially planned a full-length bobbin
coil examination of 100 percent of the active tubes in Stearn Generators 1 and 4
and 20 percent of the active tubas in Steam Generators 2 and 3. The initial scope
,
of planned motorized rotating pancake coil examinations (using a probe containing a
plus point coil, a 0.115-inch pancake coil, and a 0.080-inch high frequency pancake
'
coil) included: (1) 100 percent of the active tubes in Steam Generators 1 and 4 and
i 20 percent of the active tubes in Steam Generators 2 and 3 at the top of the tube
sheet on the hot-leg side, (2) 20 percent of the expanded B and D baffle plate
l
intersections in the cold-leg preheater in all four steam generators, (3) all bobbin coil l
identified dents 2 3 volts at the first hot-leg side tube support plate (i.e., H3) and a l
sample of bobbin coilidentified volts 2 3 volts at higher elevation hot-leg side tube
support plates, and (4) characterizing ambiguous bobbin coil indications. The
adoption of a probe containing a plus point coil, a relatively new eddy current
technology, for examination of the hot-leg side expansion transitions, low radius !
U-bends, dents, and preheater intersections, was considered both proactive and a
further indicator of management support for steam generator tube integrity
initiatives. The selection of a 0.080-inch high frequency pancake coil was also
considered an excellent decision in that it increased the capability of the probe to l
detect the presence of defects located at the inside diameter of the tube. In
! addition, the planned motorized rotating pancake coil scope included an examination l by a probe containing a plus point coil of 20 percent of the low radius Rows 1 and
2 U-bends in all four steam generators.
Upon discovery of a circumferential indication in a tube at the top of the tube sheet
on the hot-leg side of Steam Generator 4, which was the first evidence found during
Unit 1 commercial operation of the probable development of a stress corrosion
crack (i.e., a tube sample was not removed to provide laboratory confirmation of the
nature of the degradation), the licensee expanded the planned scope of hot-leg top
of tube sheet examinations in Steam Generators 2 and 3 from 20 percent to
100 percent using the three coil probe referenced above. The planned scope of full-
i
)
_
e
.
-11
length bobbin coil examinations was also expanded to 100 percent of the active
tubes in all four steam generators. The inspector viewed the licensee actions as
both conservative and an indicator of management support for steam generator tube
integrity initiatives.
The licensee plugged a total of 19 tubes during Refueling Outage 1RF05 (Steam
Generator 1, O tubes; Steam Generator 2,11 tubes; Steam Generator 3,2 tubes;
and Steam Generator 4,6 tubes). The respective numbers of tubes plugged
because of detection of single circumferential tube indications on the hot-leg
side at the top of tube sheet were nine in Steam Generator 2 and three in Steam
Generator 4. All of these indications were determined to be locaicd ca ihe outside
diameter of the steam generator tubes. The licensee plugged a tote' of three tubes
as a result of the detection of axial tube indications, a single indication in Steam
Generato" 2 and single and multiple indications in Steam Generator 4. The Steam
Generator 2 single axial indication and the Steam Generator 4 multiple axial
indications were located on the hot-leg side at the top of tube sheet, with the
Steam Generator 4 single axial indication present at the H3 (i.e., first) hot-leg side
tube support plate. Two tubes were preventively plugged because denting
restricted passage of a bobbin coil, one in Steam Generator 2 and one in Steam
Generator 3. Two tubes (one in Steam Generator 3 and one in Steam Generator 4)
were plugged as a result of the identification of volumetric indications. The
inspector considered the number of plugged tubes to be small, but of some
l significance because of the first time detection of defect indications that are
- normally found on further investigation to be secondar/ side stress corrosion
cracking.
c. Conclusions
The licensee performed comprehensive steam generator eddy current examinations
,
during Refueling Outage 1RF05, with the use of conservative examination
j expansion criteria and adoption of new eddy current technology considered l indicators of management support for steam generator tube integrity initiatives. The I
total number of tubes plugged during Refueling Outage 1RF05 was small (i.e.,19),
but considered of some significance because of the first time detection of defect
indications that are normally found on laboratory examination to be secondary side I
stress corrosion cracking, j
E1.2 Review of Steam Generator Tube Examination Proaram Reauirements j
i
a. Inspection Scope (92903)
1
1
The inspector compared the steam generator eddy current examination program l
requirements for Refueling Outage 1RF05 against regulatory requirements, industry
guidelines and qualification criteria, and specific commitments made in response to
Generic Letter 95-03, "Circumferential Cracking of Steam Generator Tubes." l
r l* 1:
.
( - 12- I
b. Observations and Findinns
During review of the steam generator eddy current examination program
requirements for Refueling Outage 1RF05, the inspector noted that Section 6.1.3
l '
in Procedure STA-733, " Steam Generator Tube Examination," Revision 2, required
data evaluation by two separate parties. This practice was considered a program
improvement over the prior use by the licensee of the same contractor for
performing primary and secondary analysis. Westinghouse and Duke Power,
respectively, performed the primary and secondary analysis in Refueling
Outage 1RF05. The use of ANSER software by Westinghouse and Eddynet 95 by
Duke Power allowed two very different views of the data, which the NRC
consultant considered a positive in the inspection. The inspector considered
the requirement (contained in Section 6.1.1 of STA-733, Revision 2) to perform
eddy current examinations using techniques qualified in accordance with
Appendix H of Electric Power Research Institute Publication TR-106589-V1, "PWR
Steam Generator Examination Guidelines," Revision 4, to be a program strength.
This requirement was also noted to have been included as a commitment in the
- licensee response (TXX-96020) to Generic Letter 95-03 dated January 18,1996.
The inspector reviewed the " Unit 1 Steam Generator Health Report," dated
September 23,1996, which was prepared by a joint licensee / Westinghouse steam
generator management team. This report contained useful reference information in
i regard to eddy current history, chemistry trends, visual inspection results, sludge
removal, and future plans. The inspector considered the report to be an excellent
managernent tool, if maintained current, for determining appropriate mitigation ;
strategies for detected degradation.
The inspector noted that Section 4.5 in Procedure NDE 7.01, " Steam Generator
Eddy Current Analysis," Revision 1, required data quality to be controlled in
accordance with Westinghouse Procedure PRO-CHG-TUE-001, " Probe Change
Guidelines for Eddy Current Bobbin Probes." Revision 0 of the latter document was
ascertained to contain quantitative noise criteria for determining when probes
should be replaced. The inspector considered this approach to be commendable.
The program was also noted to include both appropriate provisions for disposition of
manuf acturer's burnish marks and restrictions on assignment of through-wall depths
i from bobbin coil data. 1
Two areas were noted where the current examination program requirements could
'
be further strengthened. The only programmatic requirement noted during the
review pertaining to handling of loose parts was contained in Section 6.3 of
Procedure NDE 7.01, Revision 1. The requirement simply stipulated that the top of
the tube sheet be monitored for loose parts using a low frequency. Factcrs, such
as, incorporation of information from foreign object search and retrieval activities,
examination requirements for characterizing wear in tubes abutting a loose part, and
evaluation and monitoring requirements for lodged foreign objects did not appear to
i ! ! I
o
.
-13-
be addressed. The other area pertained to the current absence of applicable eddy
current graphics (in Procedure NDE 7.01, Revision 1, or a training procedure for
analysts) depicting degradation that could be encountered at Comanche Peak Steam
Electric Station. Licensee personnel indicated they were working on incorporating
suitable graphics.
The inspector t erified that the eddy current contractor responsible for data
acquisition and nrimary analysis, Westinghouse Nuclear Services Division, had
prepared an Acq iisition Technique Specification Sheet (ACTS) and Analysis
Technique Specification Sheet (ANTS) for each of the eddy current techniques
utilized in Refueling Outage 1RF05. These documents, which are required by
Appendix H (Performance Demonstration for Eddy Current Examination) of Electric
Power Research Institute Publication TR-106589-V1, Revision 4, define the
Appendix H qualified parameter values and ranges. Changes to the qualified
parameters (i.e., essential variables) require requalification of the technique.
Approval signatures from licensee personnel were noted on each of the ACTS
(TBX 01-196, Revision 2; TBX-02-196, Revision 2; and TBX-03-196, Revision 2) !
and ANTS ( ANTS 001, Revision 3; ANTS 002, Revision 2; and ANTS 003, I
'
Revision 3) documents.
The inspector requested to see the Westinghouse Appendix H qualification l
information that was applicable to acquisition of plus point data from the tube ]
expansion transitions at the top of the_ tube sheet. This request was made as a '
result of the observation that the applicable ACTS sheet for this examination,
TBX-02-196, specified the use of Type RG 174/U or equivalent extension cable
rather than the low capacitance extension cable (i.e., Zetec low loss) that the
inspector had previously seen used at other facilities for examinations of this type.
The inspector had previously ascertained that Type RG174/U cable has an i
approximate capacitance of 26 picofarads/ foot versus approximately 16 picofarads/
foot for a low loss cable. Licensee personnel obtained the following information
from Westinghouse: (a) the original Electric Power Research Institute Appendix H
qualification for the plus point coil utilized a 50-foot long Zetec high performance
probe cable and a 50-foot long 7.etec low loss extension cable; and (b) the
Westinghouse equivalent qualification specified a 50-foot length of Zetec high
performance probe cable, with either a 100-foot long Zetec low loss extensin, cebie
or a 110-foot long Westinghouse cable. The inspector was informed thet tae latter
extension cable was a type that had a modified capacitance. The inspector noted
that ACTS TBX-02-196, Revision 2, stipulated the use of a 83-foot maximum length
of Type RG 174/U (or equivalent) probe cable and a 110-foot maximum length of
Type RG 174/U (or equivalent) extension cable,
i
l
Section H.2.1.1 of Appendix H of Electric Power Research Institute
Publication TR-106589-V1, Revision 4, identifies that probe and extension
cable type and length are an esscntial variable. Section H.3.3 of this document
requires requalification of an acquisition technique if a change in acquisition
l technique causes an essential variable to exceed the qualified range. The cable l l
*
J
.
- 14-
essential variable change introduced by ACTS TBX-02-196, without performing a
requalification of the acquisition technique, is a violation of Criterion IX
of Appendix B to 10 CFR Part 50 (50-445/9709-01). The inspector was informed
by licensee personnel that their review of this issue (resulting from the
inspector's questions) determined that the actual technique used during Refueling
Outage 1RF05 for acquisition of plus point data did not comply with the probe and
extension cable requirements of the applicable ACTS TBX-02-196. Actual
acquisition was found to have used an 83-foot long Zetec high performance probe
cable and a 50-foot long Zetec low loss extension cable. This technique, although
contrary to the governing ACTS TBX-02-196, was very close to the original Electric
Power Research Institute qualification use of a 50 foot Zetec high performance
probe cable and 50-foot Zetec low loss extension cable. The inspector concluded
that use of an additional 33 feet of Zetec high performance probe cable, to that
used in the Electric Power Research Institute qualification, should have no
discernible effect on data quality.
c. Conclusions
The licensee has developed comprehensive steam generator eddy current program
requirements since the initial 1994 NRC inspection. Significant areas of program
strengthening included use of two separate parties for data evaluation, a
requirement to use Appendix H qualified examination techniques, use of quantitative
data quality evaluation criteria, and development of a steam generator health report
for identifying program and degradation status, chemistry trends, visual inspection
results, and future plans. Areas noted where program improvements could be made
pertained to requirements for monitoring and assessing effects of loose parts and
the current lack of use of eddy current graphics for depicting characteristics of
degradation modes that could be potentially encountered. A violation was identified
with respect to the approval of an acquisition technique specification sheet for plus
point data, without requiring technique requalification, which changed the qualified
l Appendix H essential variable for probe and extension cable. l
E1.3 Reouirements for Trainina and Testina of Data Analysts
i
a. Inspection Scope (92903)
! The inspector reviewed the training and testing requirements for data analysts that l were established for Refueling Outage 1RF05.
.
b. Observations and Findinas
The inspector ascertained that Procedure NDE 7.03, Revision 1, required that data
analysts be certified as qualified data analysts in accordance with Appendix G of
Electric Power Research Institute TR-106589-V1, Revision 4, be certified to at least
Level llA in accordance with American Society of Nondestructive Testing
Recommended Practice SNT-TC-1 A, and must successfully pass a site-specific
.
l .
-15-
performance demonstration test prior to analyzing any data. The inspectors
considered the requirement for analysts to be certified as qualified data analysts to
be a program strength. The inspector verified that all analysts had successfully
passed the site-specific performance demonstration test and that the pool of eddy
current test data included an appropriate scope of eddy current probes (i.e., bobbin
coil, rotating pancake coil, and plus point coil) and tube degradation modes and
locations (i.e., no detectable degradation, outside diameter stress corrosion
cracking - freespan, top of tube sheet, and tube support plate; inside diameter
stress corrosion cracking - tube support plate; and wear at anti-vibration bars).
c. Conclusions
The programmatic requirement for analysts to be certified as qualified data analysts
was considered a program strength. The licensee had compiled an appropriate pool
of eddy current test data for verification of individual analyst ability to make correct
interpretations of data.
l
E7 Ouality Assurance in Engineering Activities !
E7.1 External Assessment of Licensee Eddy Current Examination Proaram
a. Inspection Scope (92903)
The inspector compared the results of 1996 external assessments of the licensee's
eddy current examination program by a peer utility and an industry group against
the current eddy current examination program content.
b. Observations and Findinas
The licensee was noted to have revised program requirements to appropriately
address all substantive comments.
c. Conclusions
The licensee appropriately responded to the results of external assessments of eddy
current program requirements.
V. Manaaement Meetinas
X1 Exit Meeting Summary
The inspector presented the inspection results to members of licensee management at the
conclusion of the inspection on March 21,1997. The licensee acknowledged the findings
presented. No proprietary information was identified.
i
'
e
, '
1
'
ATTACHMENT
!
SUPPLEMENTAL INFORMATION
l
PARTIAL LIST OF PERSONS CONTACTED l
! 5 1
!
Licensee l
I
J. Barker, Engineering Overview Manager l
O. Bhatty, Licensing Engineer, Regulatory Affairs !
T. Broughton, Operations Work Control !
T. Daskam, Nuclear Overview Department l
E. Dyas, Nuclear Overview Department ,
D. Goodwin, Operations Support Manager l
T. Hope, Regulatory Compliance Manager i
J. Hoss, Senior Engineer, Engineering Programs i
J. Kelley, Vice President, Nuclear Engineering
F. Madden, Engineering Technical Support Manager ]
B. Mays, Engineering Programs Supervisor l
'
D Snow, Licensing Engineer, Regulatory Affairs
M. Sunseri, Nuclear Training Manager
S. Swilley, Senior Engineer, Engineering Programs
C. Terry, Group Vice President, Nuclear Operation
R.Theimer, Chemistry Manager i
R. Walker, Regulatory Aff airs Manager
i
Hartford Steam Boiler Insoection and Insurance Comoany I
J. Hair, Authorized Nuclear Inservice Inspector )
Westinnhouse
H. Gutzman
D. Obenazu
A. Sagar
NRC
H. Freeman, Resident inspector
V. Ordaz, Resident inspector
i '
INSPECTION PROCEDURES USED
IP 50002 Steam Generators
IP 92903 Followup <
l
i
F ( tl'
o
i -2- ,
,
ITEMS OPENED
Opened
l
50-445/9709-01 VIO Failure to follow procedure (Section E1.2)
LIGT OF DOCUMENTS REVIEWED
Procedures / Documents
STA-733, " Steam Generator Tube Examination," Revision 2, and Procedure Change
Notices STA-733 R2-1 through 5
NDE 7.01, " Steam Generator Eddy Current Analysis," Revision 1
" Unit 1 Steam Generator Health Report," dated September 23,1996
CLI-704, " Determination of Primary to Secondary Leakrate," Revision 4
EPRI TR-106589-V1, "PWR Steam Generator Examination Guidelines," Volume 1
PRO-CHG-TUE-001, " Probe Change Guidelines for Eddy Current Bobbin Probes," .
Revision 0 (Westinghouse Procedure) l
I
Unit 1, Fifth Refueling Outage, Steam Generator Inservice Inspection Tube Plugging Special
Report, dated January 16,1997
Acquisition Technique Specification Sheet TBX-01-196, Revision 2
!
Acquisition Technique Specification Sheet TBX-02-196, Revision 2 1
Acquisition Technique Specification Sheet TBX-03-196, Revision 2
Analysis Technique Specification Sheet ANTS 001, Revision '3
Analysis Technique Specification Sheet ANTS 002, Revision 2
Analysis Technique Specification Sheet ANTS 003, Revision 3
l
l
1
l !
!
)
}}