ML20044G950
| ML20044G950 | |
| Person / Time | |
|---|---|
| Issue date: | 01/29/1992 |
| From: | Richardson J Office of Nuclear Reactor Regulation |
| To: | Hutchinson J C-E OPERATING PLANTS OWNERS GROUP |
| References | |
| NUDOCS 9306070082 | |
| Download: ML20044G950 (20) | |
Text
.
w 9
)
og JAN 19191E j
u Mr. John J. Hutchinson, Chairman f
C-E Owners Group tr d
c/o' Florida Power & Light Company i
700 Universe Boulevard
)
Juno beach, FL 33408 q
Dear Mr. Hutchinson..
(
By letter dated December 6, 1991, you submitted for NRC staff review copies of Combustion Engineering Owners Group Topical Report, i
" Application of Reactor Vessel Surveillance Data for Embrittlement Management," CEN-405-P, Revision 1-P.
- If j
q The staff has initiated the review of this report under the NRC
{
Technical Assignment Control (TAC) No. M82272.
Enclosed is our r
$p request for additional information.
This request relates tot 1) the proof that the host reactor material is equivalent to the controlling j
material in the subject reactor; 2) the technical basis for the margin
]
[
reduction approach; 3) the statistical analysis of surveillance L }
materials data; and 4) the comparison between the staff surveillance
~
materials data and materials data in the report.
i We estimate that our review could involve 2 to 3 man-months of effort.
Our target date for completion of the review is December 31, 1992, 4
which is dependent upon timely response to our request for additional information.
Your response should be forwarded to the Document Control Desk, USNRC, Washington, D.
C.
20555.
If you have any questions, please contact John C. Tsao of my staff at 301-504-2702.
Sincerely, Original fgM Ly Y
hy James E. Richardson, Director h
B. D. Liaw Division of Engineering Technology i
Office of Nuclear Reactor Regulation l
I k
Enclosure:
. i As stated cc:
J. W.
- Pfeifer, ABB C-E Central File DLurie P. Nagata, EG&G, Idaho EMCB RF KBohrer DET RF JWiggins TMurley KWichman d
FMiraglia JTsao JPartlow ana YT 209
'l FGilles aul
- 91-73
- R BDLia WRussell z
- see previous concurrence DET:EMCB*
DET:EMCB*
DET:EMCB*
DET. -
JTsao:jt KWichman JWiggin.
JRic dson j
1 /29/92 1 /29/92 1 /29/92 4 /
/92 Official Record Copy Filename: Disc 5:
EN405M 9306070082 920129 L
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1 John J. Hutchinson ENCLOSURE REOUEST FOR ADDITIONAL INFORMATION REVIEW OF COMBUSTION ENGINEERING OWNERS GROUP REPORT APPLICATION OF REACTOR VESSEL SURVEITTANCE DATA-FOR EMBRITTTTMENT MANAGEMENT CEN-405-P AX~
MATERIiTS AND CHEMICAL ENGINEERING BRANCH DIVISION OF ENGINEERING TECHNOLDGY QIIICE OF' NUCTFAR REACTOR REGULATION In the Combustion Engineering Owners Group (CEOG) report, l
" Application of Reactor Vessel Surveillance Data for Embrittlement Management,"
CEN-405-P, the CEOG proposed two i
alternatives.to calculate the adjusted nil-ductility reference
{
temperature (ART) as. described in Regulatory Guide (RG) 1.99, Rev. 2;~namely,'the integrated surveillance approach (ISA) and margin reduction approach (MRA)..
l The IS~. addresses the case in which the controlling material is not:in the subject reactor surveillance capsule but is in another CE fabricated vessel' (the host reactor) surveillance capsule.
The ISA'uses the surveillance data from the host reactor to calculatefthe ART for the subject reactor controlling material using Regulatory Position 2.1-of RG 1.99 The MRA addresses the case in'which the controlling material is not-in,the subject reactor and host. reactor. surveillance capsules.. The MRA uses plant-specific surveillance data'to calculate the ART.but the margin in the RG 1.99 equation is reduced, depending on the measured RTndt shift relative'to the predicted RTndt' shift.
The following are the staff comments and request for clarification:
1.
There may be a. case in which either the ISA or MRA'could be used: however, the report is not clear about which approach ~
a-licensee should use since one approach would give lower (less 1
restrictive)LARTs than'.the other approach.
The CEOG should i
include a decision tree or? flow chart ~1n the' report to provide a
s:
4 3
processed using the same procedures as the surveillance plate."
The underlined sections-should be added for. clarification.
8.
Page 18, Criterion-2 reads "... Determination of the 30'ft-lb index temperature and the upper shelf, energy.shall be done unambiguously for both the irradiated and unirradiated-Charpy data."
How will these determinations be done--with^the hyperbolic tangent curve fit?. Clarify.
9.
Page 18, Criterion 3 reads "...If the fluence range.is two orders of magnitude or greater, the measurements must be within 2a."
Provide the basis for the factor of 2.
If the factor of 2 comes from RG:1.99, then reference RG.1.99 in the sentence for clarification.
Criterion 3 should also be revised to read "...The measurements shall be within tia of the~mean curve-of the actual surveillance data."
The underlined section should be added-for clarification.
10.
Page 18, Forgings in the reactor vessel are not discussed in the report.
The forgings, in general,chave lower copper contents than plates'and therefore.do not exhibit as large an increase in RTndt shift as do plates.
The Kewaunee reactor was fabricated by CE and was made fromJforgings.
The:CEOG'should include credibility criteria for the forgings in the' report.
11.
Page 18, Criterion 4 specifies that the irradiation temperature of the Charpy specimens.shall be within. 25 degrees F of the vessel. inlet temperature.
Some licensees record the temperature taken from thermocouples located at.the cold leg and some from the vessel' wall.. RG.l.99 specifies that the irradiation temperature.should be compared tolthe vessel wall temperature at~the cladding / base metal interface.
The CEOG needs to clarify why the vessel. wall temperature as specified in RG 1.99 was not used.
12.
Page 18, Criterion 5 reads "...One of. the surveillance.
capsules used in the evaluation should include Charpy specimens from a standard reference material..."
The staff infers that this standard reference material is one of the HSST01, HSST02, and HSST03 materials mentioned in' Figure 2 on page 20.
The CEOG should include the identification of the reference materials in^
the above statement for clarification.
13.
The purpose of the reference material:is to compare the material response from irradiation (e.g. RTndt shift) in order to j
determine any anomaly.
Any anomalous material response will-be-reflected in both the reactor material data and the reference material data.
However, there may be.a situation intwhich the-response data from the. reactor materials are anomalous but the y
response data from the reference material are normal.
How would L
CEOG resolve the differences?
r
_ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - - _ _ - _. _ _ = _ _ - _ _ - _. _ _ _ _ = _ _ _
= -
.~
i 5
chemical content, microstructure were similar for each plate." initial strength and toughness, content and initial toughness that were made under sim The staff has process and from the same supplier.
to support the above statement.The CEOG should provide document (s) 23.
Page 25, Paragraph A reads:
specimens from a plate,for C-E designed reactor vessel includes encaps weld, heat-affected zone reference material..."
and standard C-E designed reactors contain data from standard referencNot all surve material.
If the reactors did contain reference material e
were not reported to the NRC. specimens, the specimens were not tested or Clarify.
24.
Page 27 single measur,ement is the most restrictive and differsThe third para
"...in the event that a substantially from the other measurements, single measurement." evaluation may be performed to decide whether or not a supplemental the outlier in the measurement?Does this mean that the CEOG will discard a
significant for the ART calculation.Sometimes the outlier is discussion on the supplemental evaluation. Provide a detailed The Statistical Analysis in Appendices A & B:
25.
Page A-2 The CEOG states that "..The data in Table A-2 are the as-repo,rted values from ORNL-EDB or the individual post irradiation evaluation report; one exception is that SA 3 RG 1.99 recommends a value of 1.0% if the nickel contwere 02B data e."
reported.
Provide the basis for the 0.2% value.
ent was not 26.
Page A-3, one-way analysis of variance using the F test,The CEOG performed fo a) analysis using Student's t test, c) b) two sample Kolmogorov-Smirnow two sample test. Kruskal-Wallis analysis, d) and data are not "sufficiently different to indicate a siWestinghou ce plate difference in response to fast neutron irradiation " gnificant more detailed discussion and calculations for each of the above Provide tests.
The CEOG should input values, equations) provide calculations (e.g. assumptions, page A-3 were obtained.
to show how the statistical values on This would expedite the staff review.
- 27. Each of the four tests requires that the observations be independent (i.e. uncorrelated).
It appears that the values of the RTndt Shift in Tables A-1 and A-2 independent readings as multiple readings from the same plate a are not comprised of recorded.
Even though these readings represent different re orientations (i.e., transverse and longitudinal), these readings
o y
1 l
l Table A-1 (continued) i l
t Reactor Vessel Material Spec.
Chemistry Fluence Neutron Neutron Cansula Identity Heat ID Orien.
Guill N1f11 110 'n/cm 1 fl08'n/cm eect
- F..
Flux Shift 3
8 8
Fort Calhoun WU
'O U-225 FFCIO1 LT
.10
.48 W-225 MS-~ 4.5 5.60 SMS301 LT
.8
.6p
-Grt) t,i 5.60 60 U 265 FFC101 LT
.10
.48 8.30 r
124 W-265 FFt101 TL
.10
.48 8.70 4.78 74 5.01 70 Millstone Unit 2 aktfEb /cR - 4tib d 'sv o c R rw a t<3 (..de,d 3 hv thex sinodara W{emm mcdermi.
Wh et o.ve these t
W-97 vchei4come from t FML201 LT
[
W-97 r
.61 3.75 PML201 TL
.14
.61 3.67 3.87 3.96 70 96
(
n.i.,e T.
A-25 PHYO1 LT
.15
.59
-4t:60 13.0 43.0 A-25 4
SHSS01 LT A-35 QS 1 h t0 43.0 43.0 120 PMYO1 LT
.15
.59 7Jr30 93.'(
61.4 150 A-35 PMY01 TL
.15
.59 185 W-263 77:30 8f,'2.
61.4 FMYO1 LT
.15
.59 SM 195 W-263 FMYO1 TL
.15
.59 5-61 6, e 4.70
- 6. 6 4.70 gy 93 NOTE:
At) C hage,3 en t.he (g t[,iv.nj pages arehirt t b
-4 rI6 Onlos Ollew.w noled.
i C..
L 5
chemical' content, initial strength and toughness, and microstructure were similar for each plate."
The staff has reviewed plates in surveillance reports having different chemical-content and-initial toughness that were made under similar process and from the same supplier.
The'CEOG should provida document (s) to support the above' statement..
23.
Page 25, Paragraph A reads: "...Each' surveillance. program for C-E designed reactor vessel' includes encapsulated Charpy specimens from a plate, weld, heat-affected zone,.and standard reference material..."
Not all surveillance capsule. reports for C-E designed reactors contain data from standard' reference material.
If the reactors did contain reference material specimens, the specimens. wore not tested or data from those tests' l
were not reported to the NRC.
Clarify.
24.
Page 27, The third paragraph reads
"'...in the event'that a single measurement is the most restrictive and differs substantially from the other measurements, a supplemental evaluation may be. performed to decide whether or not to use that single measurement."
Does this mean-that the CEOG will discard-the outlier in the measurement?
Sometimes the outlier is l
significant for the ART calculation.. Provide a detailed discussion on the. supplemental evaluation.
The Statistical Analysis in Appendices A & B:
25.
Page A-2, The CEOG states that "..The data in Table A-2 are.the as-reported values from ORNL-EDB or the: individual ~ post-irradiation evaluation report; one' exception'is that SA 302B data were assigned a nickel content of~0.2% if no value was reported."
RG 1.99 recommends a value of 1.0% if the nickel content was not reported.
Provide the basis.for the 0.2% value.
26.
Page A-3, The CEOG performed four statistical tests: a) one-way analysis of variance using the F test, b).two' sample analysis.using Student's t test, c) Kruskal-Wallis analysis, and d) Kolmogorov-Smirnow two sample test.
These tests show that Westinghouse surveillance plate data and CE surveillance plate data are not "sufficiently different to indicate.a-significant
~
difference in response to fast neutron irradiation."
Provide more detailed discussion and calculations for each of the above tests.
The CEOG should provide calculations (e.g..assumptionsi input values, equations)-to show how the statistical values on page A-3 were obtained.
This would expedite the staff-review.
- 27. Each of the four tests requires that~the observations be independent (i.e. uncorrelated).
It appears that the values of the-RTndt Shift in Tables ~A-1 and A-2 are not comprised of independent readings as multiple readings from the same plate are recorded.. Even though these readings represent different orientations (i.e., transverse and longitudinal), theseireadings l
_ _ ____ _ _ ___ _ _ _ _ _ _ _ A _-_
)
Table A-1 (continued) l l
Neutron Neutron f
Reactor Vessel Material Spec.
Chemistry Fluence Flux Shift l
Cassule. Identity Hemt ID Orien.
Cuf11 Nifn) 8 8
8 i10 'n/cm )
(108'n/cm aec)
- F Ib Fort Calhoun W-225 FFC101 LT
.10
.48
-5:43' 4.5 5.60 60 W-225 SHS501 LT
.8
.6D 4r#3 i,i 5.60 124 W-265 FFC101 LT
.10
.48 8.30 4.78 74 W-265 FFC101 TL
.10
.48 8.70 5.01 70 2 EWLR-$lb d No c6<,w;st<3 C..,icst s Millstone Unit 2 hv ther sia.id#ra W{< rem m ujer m t.
Wh et a.w. these come fro m 1 t
vcituei4 7
W-97 FML201 LT
~
.61 3.75 3.96 70 W-97 FML201 TL
.14
.61 3.67 3.87 96 Maine Yankee A-25 FMY01 LT 15
.59 4h80 13.o 43.0 120 l
A-25 SHS501 LT QS
. 6p>
1h40 13.0 43.0 150 A-35 FMY01 LT
.15
.59 23r30 93.'t 61.4 185 A-35 FMYO!
TL
.15
.59 77-30 35,7.
61.4 195 W-263 kMY01 LT
.15
.59 5'"
- 6. 6 4.70 97 W-263 FMYO1 TL
.15
.59 5-67
- 6. 6 4.70 93 Nom : Att chavvj3 en m f,t(,w oj p,j,,
3,, g"gt,g cqj,, _ q,3 on i o, gg,,,,,q.
,, ;,3,
l
Table A.2 REACTOR VESSEL PLATE SURVE1LLANCE DATA UESTINGHOUSE NSSS Reactor Vessel Haterial Cansule Identity Spec.
Neutron Neutron Chemistry Fluence Heat ID_
Orien._
Cuul ELitl Il0"n/cm]
Shift Flux 8
Beaver Valley Unit 1 Il0"n/cm seel 2
"F.
NORE6k/t - 4gg (Jug U
PBV101 U
LT
.20
.54 PBV101 6.54 V
TL
.20
.54 5.79 PBV101 6.54 120 V
LT
.20
.54 5.79 135 PBV101
.e-tt 2,5 $
7.92 W
TL
.20
.54 130 PBV101 4r94 4,5$
7.92 s
W LT
.20
.54 9.49 140 Y
PBV101 TL
.20
.54 5.11 9.49 150 D. C. Cook Unit 1 5.11 185 T
PCK101 T
LT
.14
.49 PCK101 4:Yi-1. 3 6.79 T
TL
.14
.49 60 CHSS02 4:f t--
1 8 6.79 Y
LT
.14
.68 PCK101
-4dt-l.f 6.79 70 Y
LT
.14
.49 13.40 60 PCK101 Y
TL
.14
.49 10.60 105 8.59 SHSS02 LT
.14
.68 12.00 115 6.80 callaway Unit 1 7.69 3
110 U
I PCL101 U
LT
.07
.59 3.27
\\
PCL101 TL 12.1
,. 07
.59 3.27 0
12.1 30
bb V h khkRRSkk 0
oE00 ao n
z~
=
0.
C 4
N N ^n eoem--oo
~~~~s
.o o 4
.c
- as
.... - s.e.n gg o
eM Q* W WW
=
J O J@
D ce nt5 d ti i I.
a $ ()v e
oa oo ge 0 0. e*.*.
1-
~ ~
3 e
N N
,e
$Lo
~
4 4 M
ouo 0
0 I
i acao eaAa es.
~.
.E x..a lil9 99~
"9 be a
M E C
S 8.
~
i
~.
d e.
2 2 2. *.
"2828222 22 U
i t
s
\\
d!
bNbN bbbbbbbb bb 1:
mo I
.O
- N m m
n m
0h$h
=esmaa?h ff
- U N
4:
22gg
mn Ecc5EMEN tg
==
N l
M N
b b
5
\\
j i
c e u n
o s
l
= c
=
u i
eb b
>M C
A O
h b
k o o,
Ae g
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=u w
==
=
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ss>>
- r
- A _ _ _ _ _ _ _ _ _ _ _ _ _. _ _... _ _ _ - _ _ _ _ _ _
1 j
e en Table A-2 (continued)
{
Neutron Neutron Reactor Vessel Material Spec.
Chemistry Fluence Flux Shift Cansule Identity Heat ID Orien.
Cuft)
Niit)
(10"n/cm )
(10"n/cm sec)
- F a
2 Salem Unit 1 l
T PSA101 LT
.22
.53 Aref 7 fb 8.29 100
(
T PSA102 LT
.23
.54 2 4 2g(b 8.29 100 T
PSA103 LT
.22
.52 311I4 7 g6 8.29 75 T
SHSS02 LT
.14
.68 Mp, f6 8.29 60 Y
PSA103 LT
.22
.52 8.91 8.33 110 Y
SHS502 LT
.14
.68 8.91 8.33 125 e
3 Salem Unit 2
- b. '
T PSA201 LT
.10
.61 2.56 6.09 50 T
PSA201 TL
.10
.61 2.56 6.99 70 San Onofre Unit 1 A
PS0103 LT
.18
.20 28.60 49.1 100 A
SASTM LT
.20
.18 28.60 49.1 120 D
PS0101 LT
.17
.20 56.20 63.3 140 D
PS0102 LT
.18
.20 56.20 63.3 110 D
PS0103 LT
.18
.20 56.20 63.3 130 D
SASTM LT
.20
.18 56.20 63.3 150 F
PS0102 LT
.18
.20 57.30 23.5 120 F
SASTM LT
.20
.18 57.30 23.5 130 A
Table B-1 REACTOR VESSEL WELD SURVEILIANCE DATA CONBUSTION ENGINEERING NSSS l
Neutron Neutron Reactor Vessel Naterial Chemistry Fluence Flus Shift Caosule Identity Heat ID Cuft)
N1(1)
(10'*n/cm )
a 1
8 fl0'n/cm sec)
- F l
Arkansas Nuclear One l
Unit 2 W-97 UAM201
.04
.08 3.34 6.26 10 Calvert Cliffs Unit 1 1
sn W-263 WCC101
.24
.18 6.10 6.58 59 e
Calvert Cliffs Unit 2 W-263 WCC201
.20
.04 7.97 5.52 69 Fort Calhoun W-225 WFC101
.30
.60
+t51 i.2 5.60 p 2Jf W-265 WFt101
.30
.60 8.00 4.61 221 Millstone Unit 2 W-97 WNL201
.30
.06 3.77 3.98 76
.s Table 8 2 REACTOR VESSEL SURVEILLANCE DATA WESTINCHOUSE NSSS Neutron Neutron Reactor Vessel Material Chemistry Fluence Flux Shift Cansule Identity Heat ID_
Cuft)
N1(t)
(108'n/cm )
(10n/cm sec)
- F 8
8 Beaver Valley Unit 1 U
WBV101
.26
.62 6.54 5.79 155 V
W8V101
.26
.62
.2.At '/, T 5 7.92 150 W
WBV101
.26
.62 9.49 5.11 185 D. C. Cook Unit 1 inO T
WCK101
.27
.74
.3rft" 4 8 6.79 80 Y
WCK101
.27
.74 10.60 6.80 200 callaway Unit 1 U
WCL101
.06
.07 3.27 12.1 70 Haddam Neck A
WCTY01
.22
.046 M 2 O ')
6.04 95 s
i D
WCTY01
.22
.046 22.20 6.68 110 Diablo Canyon Unit 1 S
WDC101
.21
.98 2.98 7.51 110 l
n
?
Table B-2 (continued)
Neutron Neutron Reactor Vessel Material Chemistry Fluence Flux Shift Cansule Identity Heat ID Cuft)
Mift)
(10n/cm )
(10n/cm aec)
- F 2
8 Indian Point Unit 3 T
WIP301
.15 1.02 3rty 2.9 2 7.67 143 Y
WIP301
.15 1.02 8.05 8.15 180 Z
WIP301
.15 1.02 10.70 6.11 220 Kewaunee R
WKWE01
.20
.77 20.70 14.5 235 V
UKWE01
.20
.77 4541 S.59 15.8 175 a
s McGuire Unit 1 i
U WMC101
.21
.88 4.14 14.2 160 K
WMC101
.21
.88 13.80 10.1 165 Salen Unit 1 l
Y WSA101
.16 1.26 8.91 8.33 165 l
Salem Unit 2 l
T WSA201
.23
.71 2.56
'6.99 155 1
-