ML19325C729
| ML19325C729 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 10/11/1989 |
| From: | Keuter D SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | Knighton G NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| AGM-NUC-89-136, NUDOCS 8910170221 | |
| Download: ML19325C729 (9) | |
Text
'
<d y
JSMUD' SACAAMENTO MUN6Cl*AL UTILITY DISTRICT D 6201 S Street, P.C1 Box 15830, Sacramento CA 958521830.1916) 4b2 3211
- AN ELECTRIC SYSTEM SERVING THE HEART OF CALIFORNIA AGM/NUC 89-136 l
October 11, 1989 U.!S.' Nuclear Pegulatory Commission in Attn: Document Control Desk Washington, DC 20555
-Docket No. 50-312 y
' Rancho Seco Nuclear Generating Station eL
' License No. OPR-54 SECOND 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAM PLAN, REQUEST FOR
[
ADDITIONAL.INFORMATION Attention: George Knighton
Reference:
NRC to SMUD letter dated ~ July 7, 1989, Second 10-Year Interval t
Inservice Inspection Program Plan, Request for Additional Information:
The NRC requested ~ additional information on the Second 10-year Interval Inservice' Inspection (ISI) Program in the referenced letter.
The attachment to this letter;provides responses.to questions raised within the referenced letter.
Rancho Seco is_ currently in a cold shutdown condition and is proceeding toward defueling the reactor. -Pursuant to the requirements of IHA-2400(c), tha District will take an extension to the second 10-year program for ISI equivalent l
to'the length of the current outage. As such, the second 10-year ISI program will not begin unti_1 the reactor is refueled and taken to the heatup-cooldown
. condition.
Therefore, the' inspections included in the second 10-year ISI H
3 Program will not be performed unless Rancho Seco is to return to power.
This L*
schedule meets the requirements of 10 CFR 50.55a and ASME Section XI.
u
. Members of your staff with questions requiring additional infor:r.ation or L
clarification may contact Dave Sa nk at (209) 333-2935, extension 4920.
l
-Sincerely, V/
8 -
hr
[
c Dan R. Keuter posfl
- Assistant General Manager i}\\
Nuclear Attachment cc w/atch:
J. B. Martin, NRC, Halnut Creek A. D' Angelo, NRC, Rancho Seco 8910170221 891011 PDR ADOCK 05000312 9
Q FDC RANCHO SECO NOCLEAR GENERATING STATION D 1444o Twin Cities Road, Herald, CA 95638-9799;(209) 333 2935 i
p.
m m
~'
c.
SACRAMENTO MUNICIPAL UTILITY DISTRICT RANCHO SECO NUCLEAR GENERATING STATION DOCKET NUMBER 50-312 a
Add i t i on gLlaf.omt i on /Cl a ri fi ca t i on R e qu.i.ted L
A.
Provide a listing of all ASME Nuclear Component Code Cases being used during second 10-year interval ISI examinations at Rancho Seco.
N.
Reipus.g - The scope of subjects to be addressed in Inspection Plans R
are specified in IHA-2420. Inspection Plans and Schedules.
The extent of Code Cases to be used is not available at this time and will only be ascertained when the inspection vendor is determined.
Any' Code Cases to be used during this ten year interval would be listed as accepted by the NRC in Regulatory Guide 1.147.
B.
Augmented examinations have been established by the NRC when added I
assurance of structural reliability is deemed necessary.
Examples of documents which may require augmented examination are:
(1) 'High Energy Fluid Systems Protection Against Postulated Piping Failures in Fluid Systems Outside Containment, Branch Technical Position ASB 3-1; Response - The Technical Opecifications address this subject as follows:
"4.13 AUGMENTED INSERVICE INSPECTION PROGRAM FOR HIGH ENERGY LINES OUTSIDE OF CONTAINMENT A.
For the 41 welds identified on Figures 4.13-1, 4.13-2 and 4.13-3:
1.
9rior to initial power operation (9. eater than 5 percent) a volumetric examiriation will be performed with 100 percent inspections of welds in accordanr.e with the requirement of ASME Section XI Code, Inservice Inspection of Nuclear Power Plant Componer ts, to establish system integrity and baselir,e
!!/
dr.t a.
L i
L,
p
m pf-^. y
?
f *g y
+
l L
, _t F
^~
2."
- TheLinservice inspection.at each weld will be performed in-R accordance with the= requirements of ASME Section XI Code,
. Inservice Inspection of Nuclear Power Plant Components,-with
- the following schedule:
(The inspection intervals i
identified below-sequentially follow the baseline-k; examination of Specification 4.13 A.1 abcVe):
{,
Successive'Insoection Intervals
.Every 10 years thereafter (or Volumetric inspection of-7; c
nearest refueling outage) 1/3 ofxthe welds at thet
" expiration of each 1/3 of the' inspection interval with a cumulative 100 percent coverage of all welds.
. him The welds selected during each ins'pection period-shall be distributed.among the total number to be' examined to provide a representative sempling of the conditions of the welds.
[
g 3;
. Examinations:that reveal unacceptable structural defects in a weld during an inspection:Under 4.13 A.2 shall be extended
- to require an additional inspection of another 1/3 of the L
If further unacceptable defects are detected in the
'second sampling, the remainder of the welds shall be
' ~
inspected.
4.
In the event' repairs of any welds are required following any examination during successive inspection intervals, the inspection schedule for the repaired welds will revert back E
to the first 10 year-inspection program."
n J
The 41 welds specified in this section of Technical 'f x lications o
are listed in;the-Inspection Plan, Section 1. Under
- m.n C5.51-XX
- on pages 37 through 42 and' item C5.81-XX on page 44.
m (2)., Regulatory Guide 1.150, Ultrasonic Testing of Reactor Vessel Helds
.During Preservice and Inservice Examinations; s
?
- Resoonse - Attachment 1 shows the quantity of circumferential and
' longitudinal welds in the Reactor Vessel.
They are labeled to
=
show those'to be examined due to ASME Code requirements (noted C),
and!those also to be examined as stated in the Program Plan as t
augmented (noted A).-
The Code requires three of the six circumferential and one of the four longitudinal welds to be n;
. examined. Our augmented program picks up the remaining three L;1, circumferential and three longitudinal welds. Thus, all reactor vessel welds will be 100% volumetrically examined from the inside surface, as'well as all eight nozzle welds.
~E~
lt:
l
Wf: #
,f f Q"[;y '
QQ;p L j i
y?, >
+
lyh;j[
iG
- Further specifics addressing Regulatory Guide 1.1S0 cannot be 1.
-clarified until the inspection contractor to perform the hs,,R examinations: has baen established.
y
"(3) ' Regulatory Guide.1 14. Reactor Coolant Pump Flywheel Integrity; F
and:
r, h
Address.these and any other augmented examination which may have been incorporated in the Rancho Seco Nuclear Generating Station Second-10-Year Interval Inservice Inspection Prcgram Plan.
Resoonse - The Technical Specifications' address the motor flywheel E
as-follows:
~
"4.2.2. Inservice Insoettion Uip
,1
.4.2.2.2,
...each reactor coolant pump motor flywheel will be
. inspected volumetrically during the ten-year inspection yi :
interval. One hundred percent of the. flywheel will be s
, 5. [
examined..All flywheels received a one hundred percent 9
ultrasonic examination prior. to installation on the motor."
These exah dations are listed in the Plan on page 76,'Section 1, under item X.1-X.
_ Another augmented series of examinations under item X.2-X is included as a result of the report, "B&W Owners Group Safe End Task Force Report on Generic Investigation of HPI/MU Nozzle
-Component Cracking;"
C.
Prcvide a listing of all Class 2 Residual Heat Removal (RHR), Emergency
' Core Cooling (ECC), and Containment Heat Removal (CHR) systems at Rancho-Seco and include the total number of welds in each of these
@X 1 systems..
wJ
. Staff review of all Class 2 piping welds receiving volumetric
- P examinations during the second 10-year inspection interval at Rancho
-Seco shows the following:
o-y.
[
Pipe Sizes >4" NPS and Hall Thicknesses 13/8" V;
l ',
Volumetric L
~'
System and_ Surface
[y
' Aux FH 5
Decay Heat A 5
1 "5
Decay Heat B 3
m Total Helds 13 W*,
C
'4 p
l-l':( -^. +
y
, i:
y ;,
ic L}i
\\
>nn a-
, a 37
,o..
Pipe. Sizes 22" and 14" NPS and Hall Thickness.>l/5" 4-?
^'
F Volumetric-System and Surface
, ;W '
7,y Makeup Discharge-8 di HPI-A Discharge 7
D HPI B Discharge 8'
HPI Mini Flow 3
Total Helds 26-a A' representative. sampling of welds in the RHR, ECC, and CHR systems should receive inservice volumetric examinations. The staff has previously determined that a 7.5% augmented volumetric-sample constitutes'an acceptable resolution at.similar plants.
Discuss the impact.of performing volumetric exanination of at least a 7.5% sampling M'
b of the; Class 2 piping welds in these systems, w.
gj l Response -1The submitted Plan istwritten to ensure at least a 7.5%
R~
sampling.of;all Class 2: required welds. Hence, there is no impact of C
performing volumetric examinations of the'7.5% sampling of the Class 2 welds.
-The following table ~shows the. number of welds by size.in~ column C, the calculated'7.5% sample size in column D'and the number of welds to be examined in column E.
A' B
C D
E
. Pipe Size
. System (Stainless 5tl.
No. of Helds 7.5% Sample "No. of Helds to be Unless Noted)
Insoected in Plan w
Aux-FN' 6"
65 4.9
'5 1
6"-
CS 16 1.2 2
j Decay.
12" 46 3.45 4
Plus 8
' Heat A 210" 17 1.3 1
Augmented Decay 12"_
11 0.8 1
Plus 5 Heat.B.
10" 20-1.5 2
Augmented i
'MU'Disch.-
4" 53 3.98 4
2 1/2" 49 3.7 4
L
,HPI A!
4" 58 4.35 4
l Di s ch ~.
3" 13 0.98 1
i
. 2 1/2" 27 2
2 fHPI B 4"'
57 4.3 4
'Disch.
3" & 2 1/2" 45 3.4 4
HPI Mini 2 1/2" 5
0.4 1
Flow-2" 29 2.2 2
m I
f
-4_
- i J/
m
ff y
,,m y
.4 y^
.DL Review of~ Section 1 of the ISI' Plan: listing the NDE examinations being 7
performed'during the second-interval and the calibration block drawings-i
"~
in Section 7 shows that some of the calibration-blocks may not: meet the applicable Code requirements.
Examples are as follows: Calibration 4
Block #26-(10-inch diameter,1.125-inch thick) is being used for ISI:
- examinations of Item B09.11-27 (12.8-inch diameter,1,3-inch wall-
' thickness) and Item B09.11-31 (14-inch diameter, 1.4-inch wall
.[
thickness).
Calibration Block #27 (3-inch thick flat block) is being J
used for the examination of-Items B09.31-1 throuah B09.31-4 (small U',
diameterbranchconnections-to-largediameter'primarycoolantsystem
~
piping). Calibration Block #23 (stainless steel) is being used a p
examine-Item B05.040-4 (carbon steel-to-inconel dissimilar metal weld).
It-is also noted that many of the calibration block drawings in gp Section 7 of the Plan have been reduced in size and are illegible with regard _ to dimensions and material specifications.
B
' Appendix III, " Ultrasonic Examination of Piping Systems," of Section XI P,
'of the Code requires that-basic calibration blocks be made from material of the same. nominal diameter hnd nominal wall thickness or I:
pipe schedule as the pipe to be examined. The calibration blocks for similar metal welds shall be fabricated from the material specified for the piping being joined by the weld. Calibration blocks for dissimilar-
- metal welds shall be fabricated from the material specified for the side of the weld from which.the examination will be conducted.
If the l
examination will be. conducted from both sides, calibration reflectors shall be provided in both materials.
p*
'The' staff considers inservice volumetric examinations of Code Class.1 and-2 systems crucial to plant safety and, therefore, feels that proper calibration standards should be obtained and utilized for all ISI examinations.
o Provide a discussion of the calibration, blocks 1being used for ISI
-examinations-during the second 10-year-interval at Rancho Seco and l
eitMr confirm that all calibration blocks meet or exceed the Code
- requirements or provide-technical justifications in the form of R
requests for relief for the continued use.of any' blocks which do not meet the Code requirements.
ResDonse - Calibration Block #26.is used for both 10" and 12" diameter pipe examinations since the block wall thickness is within the tolerance of the piping. wall thicknesses. However, all 14" diameter 1.'4" thick pipe welds will be examined using Block-#37 which was originally mounted for remote examinations.
The Plan will be revised to reflect the designated calibration block change.
t t
6.
^
y, 9 '
+
'M
- Calibration Block #27 is'being used for-branch connections in the Hot-i#
. and; Cold Leg Reactor Coolant piping.
While the piping diameter may
,m seem small, the' size-of the weld in the RC piping is not small.,as y
noted 1n'_the following.-table.
[
HM NOMINAL PIPE HELD DIA. IN RC PIEE ~
l E
9.31-1 10" 21 1/2" w
9.31-2 12" 22" 9.31 2 1/2" 10".
9.31-4' 2 1/2" 9 1/2" Larger. more:1egibic drawings of Calibration Blocks are available at the-site along with nozzle.and'other branch connection weld' i.
. configuration details.
D o
.The notch configurations of Appendix III will be added to all piping calibration blocks where space permits.-
E.
Relief Request'#2: - Relief is requested from the ASME Code-requirea a
surface examination of RPV core flood nozzle safe end welds.
The Licensee has prcposed performing a volumetric examination of 100% of r
- the pipe thickness with automated inspection equipment from the nozzle ID.
The. proposal couldLbe considered acceptable provided that the Licensee ments the following' conditions:
(1) The remote volumetric examination includes the entire weld volume and heat affected zone instead of only the inner one-third of the weld as required by the Code, t
(2). The ultrasonic testing instrumentation and procedure are demonstrated to be capable of detecting OD surface-connected
' defects, in the circumferential orientation, in a laboratory test
' block. The defects should be cracks and not machined notches.
- Provide a-discussion of the above conditions and verify that they will
-be. met.
l 2
h Reagante (1) The volumetric examination will include the entire welo heat affected zone as well as 100% of the weld.
(2) Currently, two B&W designed units in Regions II and III have developed an automated inspection technique to identify OD surface defects while examining from the ID surface..A longitudinal beam is transmitted by surface contact'and does locate small EDM i
h notches.
He intend-to use this same procedure.
?
l l
l m
p,
K
{
{,
E-
+
p t
[
N
~
f.
Relief Request #3:
Relief is requested from performing the'ASME Code-required-volumetric examination of Primary Coolant Pump casing welds and visual (VT-3) examination of_the' pump casing. internal i
surfaces.
The Licensee has proposed performing a visual examination of-100*4 of: the external surfaces of the welds in~ lieu of the Code-required
- volumetric examination.
l Other plants with similar pump configurations have committed'to performing ~ surface examinations of the exterior surfaces of the welds M,
once per inspection interval and, if the pumps:are disassembled for
~
naintenance
- a. surface examination of the accessible interior surfaces ofl the wlds.
Discuss the impact of performing surface examinations as.
f described above in lieu of the proposed visual examination for these-welds.
READAD11t - Performing internal surface examinations on RC Pumps manufactured by Ringham Willamette is impractical because of internal contamination and. interference from quadranc volutes.
He will perform a surface examination on the exterior surface of the L
-weld.
G.
Relief Request #4 (Class 2 hydrostatic test) and Relief Request #5-(Class 3 hydrostatic test):- Discuss.the operating and design pressures of the affected components as compared to the Code-required hydrostatic test pressure. As it is noted that the proposed substitute examination
.(a leak check during normal system operation) is a Code requirement and not a substitute examination, include a discussion of the design pressure of the affected pump seals giving consideration as to what the maximum alternative test pressure could be in order to meet the intent of the. Code.-
-Reiponse _ Isolating the pump casings from the piping hydrostatic tests
-includes a'mi..uscule portion of the system.
The maintenance isolation valvescare closely located to the pump suction and discharge.
Since-the maximum pressure the pumps can experience occurs while running, leak-tests on these pumps are accomplished during the normal quarterly pump. testing surveillances. Thus, the checks for maximum leakage is performed far more frequently than the hydro requirements.
In ASME Section XI, the Special Working Group on Pressure Testing is currently working to reduce and/or eliminate inservice hydrostatic p
tests.
Y ?
~ - -
i a
ATTACNMENV.I J}
4 u -.
gl
- p' :
u n
.)
~ '
L' C,
1 j.
l f
-/
n g-
[
1 ;k I
'l
)
-- ~:
w 1A p
C w
c l:u N
W k
~
l' -
?.
4 g
s Cp k,
i a
1 g__
7 g
r K
s N
fl!
I l
e M
I s
l:.
r- -
I, p
t F
7 s
y _)
./
i I
l t
4 I
C = CODE RE0MTS A = AUGMENTED.
..