B130202, Revised Surveillance Specimen Program Evaluation for TVA Browns Ferry Unit 3
| ML18039A776 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 04/30/1999 |
| From: | Branlund B, Frew B, Tilly L GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML18039A775 | List: |
| References | |
| GE-NE-B1302026, GE-NE-B1302026--1, GE-NE-B1302026-00-01, NUDOCS 9905120297 | |
| Download: ML18039A776 (46) | |
Text
GE. Nuclear Energy Engineering &-Technolo gy General Electric. Company.
175'Curtner Avenue, San Jose, CA 95125 GE-NE-B 1302026-00-01 April 1999 Surveillance'Specimen Program Evaluation for Tennessee Valley Authority Browns Ferry Unit 3 Prepared by:
L.J.'illy, Senior Engineer Structural Assessment 2 Mitigation Verified by B.D. Frew, Senior Engineer Structural Assessment k Mitigation Approved by.
B. J. Branlund, Technical Leader Structural Assessment 2 Mitigation 9'P05i20297 990430 PDR ADOCK 05000296 P
el GE Nuclear Energy GE-NE-B1302026-00-01 IMPORTANTNOTICE RX(rA)RDINGCONTENTS, OF THIS REPORT
'lease Read Carefiillv The only undertakirigs of'eneral Electric Ch'an) resp'ecting. information in this document are contained in the contract, between Tennessee Valley Authority (TVA) and General Electric Company, and'nothing in this document sha)1 be construed as changing, the purchase order.
The use of this information by anyone other than TVA, or for any purpose other than that for which it is intended, is not authorized; and with respect to-any unauthorized use, General Electriic Company makes nd representation or warranty, and assumes no liability as to the cornplIeteness, accttraOy,,'or',usefulness of the infdrrnatibn contained in this document, or that its use may nott in&inge privately owned rights.
" GE Nuclear Energy GE-NE-'B1302026-00-01 Table OfContents P~ae 1.
ABSTRACT 2.
INTRODUCTION 3.
COMPARISON WITHOTHER SURVEILLANCE.DATA 4'.
PRESSURE-TEMPERATURE (P-'T) CURVES 5.
SUPPLEMENTAL'URVEILLANCEPROGIVBvf (SSP) 6.
REVISED SURVEILLANCESCHEDULE 7.
CONCLUSIONS 8.
'REFERENCES APPENDIX A 15 21 25 28 30 111
~I GE. Nuclear Energy GE-NE-81302026-00-01 Table OfFig:mes P~ae 2-1 Measured Sliiftvs. Predicted Shift for Base Metal 2-2 Measured Shift vs. Predicted'Shift for Weld Metal 3-1 Measured Stuft vs. Predicted Shift for Base Metal 3-2 Measured'Stuft vs. Predicted Shift for Weld Metal 4-1
.Comparison ofKzand Kq, 6-1 K~ vs. EE.'PY for Browns Ferry Unit 3 Platje Material
'-2 Predicted Shift'vs. EFPY,:Browns Ferry Unit 3 Surveillance
.Capsule Plate 7
i13'4'7
'23'4 A-1 ART vs. EFPY 35 Table CtfTables P~ae 3-1 BWR Surveillance Progmn Results for Base Metal 3-2 BWR Surveillance Proymn Results for Weld Metal A-1 Browns Ferry Unit 3 RPV Material Data 11 12'4
GE Nuclear Energy GE-NE-B1302026-00-01 ACKNOW'LEDGMENT The author would like to thank Sam Ranganath for valuable technical input to this document.
Cl 0
l
I GE Nuclear Energy GE-NE-B 1302026-00-01 1'. ABSTRACT Browns'Ferry Unit 3 (BFN-3).has maintained a vessel surveillance program. to meet the intent of 10CFR50, Appendix H.[1].
The current Browns, Ferry Unit 3 surveillance program schedule requires that the first surveillance capsule be removed at eight (8)Effective 'Full Power Years (EFPY) for BFN-3.
The original licensed schedule required, removal of the first capsule at six (6) EFPY, however this was subsequently changed'to eight EFPY by.a revision.to the BFN-3 Technical Specification [2].
The current Browns Ferry Unit 3 schedule was developed in accordance with the, intent of 10CFR50, Appendix H, and did not incorporate BFN-3 specific conditions listed below:
~
Plate and weld chemistry (copper content from 0;09% - 0.24%)[3,4];
~
LowRPV 1/4T 32 EFPY beltline fluence (<<'5 x 10'/cm fluence) [5];
~
Resulting low predicted shift in the capsule material reference nil-ductility temperature (RT~T); < 30'F at 32 EFP Y.
Ifthe current schedule is executed, the measured data for the first capsule material,may not be useful, as the -expected shift in RT~T (hRTiiDT) is 'low.
In addition, the data provided by the first capsule can 'be replaced by information from other sources.
Therefore, the surveillance program's withdrawal schedule can be extended.
The extended schedule can be justified because:
~
Actual industry BWR data shows predicted BFN-3 dXT~T+ margin values. based on Regulatory Guide 1.'99 Revision 2 (Rev 2) [6] are expected to bound'he measured d,RT~T values;
~
There is inherent, conservatism present, in the pressure-temperature (P-T) curves for BWRs;
0 GE Nuclear Energy GE-NE-81302026-00-91
~
The derived fracture toughness values are lower'bound values and are based on crack arrest (Kq) rather than the higher crack initiation (KI,)touphness
~
Data from other plants can be used to predict the~ behavior'f the material early in plant life.
Based on the evaluation presented in tjhis report!, thee recommended'ithdrawal schedule'or the first surveillance capsule'for Browns Ferry Unit 3 is 24 EH'Y. 'This new schedule meets the intent of ASTM E185-82 [5], as the first capsule would be removed with the capsule fluence being less than 5 x 10'/cm anti the value ofBRT~~ would'be less th~an
~
50OF 2
'E Nuclear Energy GE-NE-B1302026-00-01
- 2. INTRODUCTION Vessel fracture toughness is a major consideration for nuclear vessels; irradiation is known to decrease the fracture toughness of vessel materials.
Therefore, measurement of the long term effects of vessel irradiation is a key component, of surveillance programs.
Tennessee Valley Authority. (TVA) maintains a vessel'urveillance program at Browns Ferry Unit 3 (BFN-3) meeting the intent of 10CFR50, Appendix.H to monitor for changes in fracture toughness ofvessel beltline materials as required by the NRC.
The BFN-3 surveillance program meets the intent of 10CFR50, Appendix H and ASTM E185-73 (for design) for the following reasons:
~
The selected base and weld metals are representative of the vessel 'beltline materials;
~
The capsule materials have a similar fabrication history to the vessel;
~
The
- number, type, and design of capsule specimens are equivalent to ASTME185-73..
The surveillance program implemented at BFN-3 consists of three specimen holders installed in.the reactor during vessel construction.
The number ofholders was determined per ASTME185-66.
The three specimen holders were designed, built, and"analyzed.to ASME Section III, 1965 Edition, with Addenda through Summer 1966.
The selection ofholder location was established to duplicate as closely as possible the temperature history, neutron flux
Oi GE Nuclear Energy
'E-NE-81302026-00-01 spectrum, and maximum accumulated RPV beltline fluence, considering:
~
interference/accessIibility with other reactor hardware (e.g,',, jet pumps);
~
peak fluence as a function ofheight;
~
peak fluence as a function ofradial position.
Using these criteria, the capsules were located at the vessel inner diameter at core mid-height at the 30', 120'nd 300'essel azimuths (available areas considering jet pumps).
In 1989, when the v,withdrawal schedule of 8 EFPY for the first capsule was appiov4d by the NRC [4], ASTM E185-82 was in eFect.
Withdrav'val schedule requirements per
- 10CFR50, Appendix H and ASTM E185-82, state that the erst specimen holder be removed at 6EFPY (or when the accumulated fluence of the capsule exceeds Sx10'/cm or when the highest predicted L&'TNT of the capsule materials is
'pproximately 50"F, whichever comes first) and <he lsecbnd be retno0ed at 15 EFPY. All testing and reporting (regardless ofwithdrawal. schedule) is to be performed in accordance with ASTME185-82.
This capsule withdrawal schedule was recommended for two reasons:
1.
Data would be provided for future pressure-temperature (P-T) curve calculations.
The data woulld be used td remove conservatisrn present in the (P-T) callculations.
The P-T curves would be recalculated after the first capsule had been removed, using the capsule flux wire measurements instead of-the.,conseivative. callculated.fluence The data.
obtained fiom the first capsule would be used to identity any anomalous conditions, i.e. a greater than expected shift in RTNDr.
'GE Nuclear Energy GE-NE-B1302026-00-01 However, withdrawal at eight (8) EFPY of the Browns Ferry Unit 3 capsule is not essential for continued safe operation for the followingfour reasons:
1.
The BFN-3 fluence [7] used for shift predictions in.accordance with Rev 2 is based upon a conservative calculation, and willbound the actual fluence.
2.
Predicted shifts 'bound the measured results based on review of predicted RTNDY shifts and measured RTNDT shifts from other BWR surveillance capsules.
Figure 2-1 is a plot of actual shift measurements versus predicted shifts (calculated per Rev 2) for base material.
This figure shows that the predicted shift plus margin conservatively bounds the actual shifts measured from BWR surveillance specimen data.
The same plot for weld material (Figure 2-2) again shows the predicted shift plus margin term bounds the measured shift.
3.
Based on actual ART calculations performed in accordance with Rev 2 (see Appendix A), the shift (dRTNDT+ margin) for the Browns Ferry Unit 3 surveillance weld is calculated to be 60'F at 32 EFPY. Ifthe first capsule is removed at 8 EFPY, the actual shift (predicted to be 13'F) may not be large enough to,be differentiated from.the data scatter, since the predicted fluence of the capsule at 8EFPY (1.85 x10'/cm) is low, and the chemistry of the 0
BFN-3 capsule weld'material is good (0.11% copper).
Thus, the data, obtained may not be 'useful for predicting the material
- behavior, as it may be indistinguishable from the unirradiated data.
4.
Supplemental Surveillance Program (SSP) specimens will provide early test data for a weld similar to the BFN-3 surveillance weld; the weld is the material of concern, as the vessel weld material is limiting throughout plant life. This program supplements the BFN-3 surveillance program by providing timely detection of anomalous RTNDT shifts, should any occur.
The fluences on the
GE Nuclear Er)ergy GE-NE-.B 1302026-00-01 SSP capsules are comparable to the fluence for the BFN-3 vessel wall in, the'ime frame. ofinterest.
This report supports the extension of tlhe surveillance capsule testing. schedule for. BFN-3 for the following rea,soris:
~
The fluence. experienced by the BFN-3 vdssdl %all is low;
~
The BFN-3 capsule plate and weld material has good alloy chemistry (i.e., low copper in the range of0.10% -'.11%)I'3,',4];
~
The actual shifIt in the BFN-3 weld material may not bi'. distinguishable from'he data scatter with early testing.
The justification for extend!ing, the schedule is 'based on the following reasons:
~
Predicted s?iifts boun'd tlhe actual BWR industry surveillance results, and are expected to bound the BjFN-3 shifts as well;
~
~
The P-T curve calculations are inherently. conservative; The supplemental surveillance program willsupplement tlhe BFN-3 surve'illance program by providing for the timely detection ofanomalous RT~q shifts, Ll Extension ofthe surveillance progr'am schedule v>ill ensure'hat use6~1'data is obtained and
~
continued safe operation of Browns Ferry Unit 3 is ensured by using the SSP data and'aintaining the BFN-3.P-T curves in accordance with Rev 2.
GE Nuclear Energy GE-NE-B1302026-00-01 120 60 M
40 20
- - - -Predicted+
Margin
" " - - Predicted - Margin Measured
Predicted 0
0 10 20 30 40 Predicted Shift, OF>>
70 Figure 2-1 Measured ShiA vs. Predicted ShiA for Base Metal
GE.Nucif ar Energy GE-NE-B1302026-00-01 140 120 100
~ '
~e 80 60 CI7 dA
'dV CI
'I 20 t
0 Predicted Pre dfirtndl + Mnroin
- - - freafcrea-Margm k
Measured
-20
~,
AA ev
-60 0
~
ee 10 20 30 40 prd fllefelJ Sillff oF 50 60 70 80 Figure 2-2 Measured Shift vs. Predicted Shift for Weld Metal
GE Nuclear Energy GE-NE-B1302026-00-01
- 3. COMPARISON WITHOTHER SURVEILLANCEDATA The evaluation of the shiQ in the RT~ for Browns Ferry Unit 3 (see Appendix A) was performed using the techniques ofRev 2 for vessel material and the predicted fluence (i.e:,
no additional surveillance data).
These predicted values of RTiiDY shift indicate that the BFN-3 vessel will not experience a large shift over vessel life.
To confirm the conservative predicted shift plus margin values (used to modify the surveillance program schedule),
a comparison has been made between calculated shift and fluence values, and actual measured surveillance data from other BWRs.
A significant number ofsurveillance capsules from-BWRs have been tested.
Table 3-1 is a tabulation ofthe base metal results from'these surveillance programs.
The most significant feature, for a range of material chemistries and fluences, is that the expected shift is bounded by the calculated Rev 2 shift plus margin.
For example, the measured BWR/4, 251" vessel (similar to Browns Ferry Unit 3) shifts are less than the. predicted Rev 2 shift plus margin values by an average of 28'F (based upon the five complete data sets).
For BWR/4-251 capsules, the average first capsule shift observed was 17'F, while the average predicted shift plus margin was 45'F.
This data indicates that the BFN-3 capsule shift (predicted to be 13'F at 8 EFPY) willbe small and may not be distinguishable from data scatter.
Similarly, Table 3-2 lists surveillance capsule data for weld material.
The meas'ured shifts are bounded by the predicted shift plus margin values.
BWR/4-251 weld data (for the six complete data sets) shows the predicted shift plus margin to exceed the measured values by an average of47'F.
The average shift observed was 20'F, while the predicted shift plus margin was 67'F.
The predicted shift values are plotted against the measured shifts in Figures 3-1 and 3-2 for all BWR data available; the data is from Tables 3-1 and 3-2, respectively.
These graphs show that the measured shifts are bounded by the predicted shift + the margin
41 GE Nuclear, Energy
'E-hJE-B1302026-00-01 term [5]. Based on these data, thie measured shift for,BFN-3.would be conservktiviely bounded by the Rev 2 piredliction.
Since fluence has a significant effect on the Rev 2. calculation, use of an appropriate fluence value is essenti;al for,accurate shift prediction.
The shift + margin predictions in Tables 3-1 and 3-2 utilize fluence valu'es determined from flux. wiires removed early in plant life. In the case of BH.'l-.3, hiowever, a conservativie estimate of the fluence [7] is used which will bound the actual fluencie.
Therefore, the fluence used for the,ART'alculations (as described in Appendix A) forBFN-3 is considered conservative.
Other than, fluence, the most significant effect oui the ART i0 the chemistry factor (CF).
The CF is determined, from the copper and nickel levels, copper having the m6re significant effect.
A study has been perfoimed [8] on the copper levels preserit in BWR beltline materials, in response to NRC letter 92,-01, Supplement 1.
The intent was to id.entify the plants with significant variatiion in the rieported copper levels. For the electroslag weld material, recently available information in an NRC SER fot CdmIiioiiiwealth Edison (which is a.best
'stimate chemistry)[9] has been usedi,in determini'ng CF.'and'ART values.for BFN-3 [10].
Based on the evaluation of pirevious surveillance data of actual shifts and fluences, the expected measured fluence, for BjFN-3 and the chemistry ofthe BFN-3 vessel material, the actual shift for BFN-3 is expeicted tci be conservatively bounded by the calculaited value of shift + margin.
10
GE Nuclear Energy GEONE-B1302026-00-01 I'LANT R I'V BIYR ID
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. 40.8 42.0'.2 Table 3-1 BWR Surveillance Program Results for Base Metal
GE Nuclear Energy'LANT 8WR/l AC AS BS R/3 H
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>I hlcV FLI)ENC E IBEFPY DEITA DELTA+
RPV Ccpsalc TEST (sl0")1)
Bs R
R D
I.D.
Ca CF ARGIN SHIFT NI (a/ccc "1)
(c/c)
(Ia)
(Ic!)
(%)
(c/c) 3.6 2)3 l9.8 30 8l N/A O. I1 0.07 1.98 4.78
'23.l N/A 300 19.I O.I) 0.01
- LIS, 41.6 1.5 2l)
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0.35 0.28
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Table 3-2 BWR Surveillance Program Results for Weld Metal 12
GE Nuclear Energy GE-NE-B1302026-00-01 120 S4 60 C6 40 0
kk kk k
kk
- - - - Predicted+
Margin
- - - - Predicted - Margin Measured Predicted 0
10 20 30 40 Predicted Shift, ~F 70 Figure 3-1 Measured ShiA vs. Predicted ShiA for Base Metal 13
GE Nuclear Energy GE-NE-B $302026-00-01 140 120 t00
~en
~
en n
0 80 60 lO 40 0
-20 AA VV
-60 0
~a a
k 4
~ n 10 20 30 n
k 40 60 70 80 Prcdic(cd n
a n h A:
~ 3 a Ll I LMllhlWI~ lVldlglll
- " - Predicted-Margin Measured Dvedintnd Shift oP Figure 3-2 Measured Shift vs. Predicted Shift for Weld Metal 14
" 'GE Nuclear Energy GE-NE-B 1302026-00-01
- 4. PRESSURE-TEMPERATURE (P-T) CURVES The shift in RTgpr obtained from surveillance testing is used to evaluate the long term effects of irradiation on the fracture toughness of the vessel.
The reference fracture toughness (K~) is determined using the shift in RT~~,'~ is part ofthe calculations ofthe P-T curves performed in accordance with ASME Section XI, Appendix G.
The current Browns Ferry Unit 3 P-T curves were calculated with the shift in RT~~ corresponding to 20 EFPY.
The K~ correlation was.developed from several sets of material data on pressure vessel steel [11]. The KiR curve was drawn to bound the available data.
Thus the correlation has inherent conservatism.
- In addition, operation of BFN-3 follows the steam saturation
- curve, therefore, the operating temperatures are expected to be well in excess of the minimum required temperature.
During normal and accident conditions, the BFN-3 vessel maintains more than adequate margins.
The operational issues of Pressurized Thermal Shock (PTS) and Low Temperature Over Pressurization (LTOP) are not applicable to BFN-3. The limiting case for BFN-3 is the pressure test.
The P-T curve associated with the pressure test is calculated using the crack arrest fracture toughness, K~ (Kt.).
The static crack initiation fracture toughness, K~, is significantly higher than K~ in the temperature range of interest [12].
Therefore, use of K~ conservatively. bounds the fracture toughness of,the vessel.
Figure 4-1 is a plot ofKt. and K~, as a function ofT-RTmz [13]. The Kr. curve is shown to be lower than the Ki, curve, conservatively bounding the fracture toughness.
For
- example, at a
pressure test temperature of 221'F and a
vessel ART of 112'F (corresponding to 20 EFPY for BFN-3), the &acture toughness for initiation and arrest are estimated to be:
15
GE Nuclear Energy
'E-NE-B1302026-00-01 Ki,= 200.0 ksiIin K~ = 87.2 ksiIin Thus the K~, value is approximately 2.3 times the Kt, value, clearly showing Kt, to conservatively bound the calculations.
The combination of lower bound &acture toughness, the Browns Ferry Unit 3 operatiIng'haracteristics and the conservative fracture toughness values indicate that the BFN-3 vessel fracture toughness is not a significant concern over the Ilifeofthe plant.
16
GE Nuclear Energy GE-NE-B1302026-00-01 280 240 iK'fc,:;...26,0.:~0j~g 200 160
Kla
- - " Klc T-RTndt g 120-80 40 W
A W
0
-250
-200
-150
-100
-50 0
50 Temperature Relative to RTndt (T-RTndt), 'F 100 150 200 Figure 4-1 Comparison ofKt, and K>,
17
Oi
.GE Nuclear Energy GE-NE-B1302026-00-01
- 5. SUPPLEMENTAL SURVEILLANCEPROGRAM The BWRVIP is in the midst of executing a
supplemental test program being
,administrated by EPRI (and originated by the BWR Owner's Group [BWROG]) that is designed to significantly increase the amount of BWR surveillance data in a systematic manner which should permit the development ofa BWR-specific equivalent to Rev 2.
Description The Supplemental Surveillance Program (SSP) was begun in the late 1980s when the BWROG concluded'from their review ofBWR surveillance data the following:
~
Due to the smaller number. ofcapsules per plant and the relatively fewer number of BWRs than PWRs, there is limited BWR.surveillance data at higher fiuences available to analyze;
~
The ARTs associated with Rev 2 imposed some hardships on pressure testing for BWRs, some of which might be relieved ifbetter predictive models of the BWR embrittlement phenomenon were obtained.
In light of these
- issues, the BWROG prepared supplemental capsules which were installed in Cooper and Oyster Creek.
One capsule from Oyster Creek was withdrawn in 1996, with additional withdrawals planned for 2000 and 2002.
The results of the SSP will be the equivalent of 84'dditional surveillance capsules, compared to about 35 which have been tested to date.
These capsules were designed to systematically evaluate embrittlement trends in BWRs. For example:
18
GE Nuclear Energ!y
,GE-NE-81302026-00-01
~
The capsules are: positioned so that flux differs by a factor of 2.
Also, irradiiation times differ by a factor of 2. In this way, s()m(: c~ipsirles, have matching flux but with different fluence, while some have matching fluence and a differing flux leve'I;
~
The materials used were selected to bouncl the range of chemistries,in:B~
beltline materials, and in most cases are BWR beltline riaaterials;
~
Irradiations are being done.in BWRs to correctly simulate conditions like temperature, neutron spectrum and trarrsient operation.
Relationship to Browns Ferry Unit 3 The SSP does iilot contain: BFN-3 specific material among the materials irii the capsules.
However, the: SSP contains material similar tIa the BFN-3 limitingmeld (for BFN-3 the weld material is the liirniting material. throughout plant life, so the plate is not a significant concern).
The SSP Quad Cities '2 Eleictroslag Weld 'material contained in the program has a-composition similar to the BFN-3 surveillance program material, and was made by the same manufa'ctuter'(B'kW) in the same time perio (Quad Cities 2, 1970; BFN-3, 1971).
The copper content of'the Quad Ciities 2 weld material is higher than the BFIM-3 surveillance weld (O.Il8% v!>>. 0.11%)>> and the Quad Cities 2 nicke1l content is lower (0.18% vs. 0.28%).
The resultant chemistry fa'ctors
'CF)
(per Rev2) aire 93 and 81 for Quad Cities 2 and BFN-3, respectivelyYn
- addition, the BFN;3 surveillance plate is lrepres4nt6d by materials with similar chemistry in eachi of the SSP capsules.
The SSP results willbe applicable to BFN-3 for two reasons:
~
Generically, the SSP resullts,will be from representative environmental condiitions
'n materials representative of'all BWRs, including BFN-3;
~
Specifically, results willibe, developed which willprovide ihformation on a ma'terial
'hich is expected to respond to iirradiatio@ sir!nilar:t5 the Weld in the Br'owns Ferry Vnit 3 surveillance program.
19
GE Nuclear Energy GE-NE-B $302026-00-01 The SSP capsules, when tested, will have collected between 5 x 10'/cm (20;5 EFPY for BFN-3 at 1/4T) and 2 x 10'/cm (82;1 EFPY for BFN-3 at 1/4T) fluence.
Thus, the results of the SSP are complementary to the BFN-3 surveillance, program such that postponement ofthe capsule withdrawal willhave minimal'impact on the understanding of irradiation sects'on the BFN-3. vessel.
20
4I 0
1
'J
~GE Nuclear Energy GE-NE-B'l302026-00-01
,6. REVISED SURVEILLANCESCHEDULE The surveillance program is intended to characterize the vessel properties as a function of irradiation over the life ofBrowns Ferry Unit 3. The Charpy impact energy obtained from the prescribed testing is used to evaluate the reference fracture toughness of the BFN-3 vessel (K~) in accordance with ASME Section III, Appendix G.
The schedule for the surveillance program-testing should be designed to obtain the best data, while maintaining safe operation.
The expected change in fracture toughness of the BFN-3 weld material (the limiting beltline material) as a function ofEFPY is plotted:in Figure 6-1.
Since the pressure test is
,the limiting case, the calculated K~ is for a 1140 psig pressure test.
The pressure test temperature was modified at selected intervals for illustration purposes.
This figure demonstrates.that the K~ used to calculate the P-T curves is expected to conservatively bound the required vessel fracture toughness.
Since the K~ is considered a conservative prediction, and the SSP willidentify a greater than expected shift relative to BFN-3, the first surveillance capsule testing should be at the time at which a majority ofthe shift in the vessel RT~T has been achieved, consistent with the intent ofASTME185-82.
Early testing ofthe surveillance weld specimens may result in:the measured shift being less than the data scatter (sometimes resulting in negative shifts in RT~T).
Correct selection of the removal time will ensure useful data from all specimens.
If the shift is greater than expected, then the margin present in the P-T calculations together with the limiting fracture toughness represent an added margin of safety.
Since the SSP can be used to identify anomalous shifts, the first surveillance capsule testing schedule should be developed. to measure a significant portion of the fracture toughness change, as measured by dRT~T.
Since the limitingweld material for the vessel has a low expected bRT~T of 52'F (at 1/4T) over the life ofthe plant, the recommended 21
O GE Nuclear Energy GE-NiE-81302026-00-0 l.
schedule should bie designed to measure a majority ofdRTrrpy ofthe plate material.
(riven the low expected shiit, a criteria of 75% ofthe expected shift iin RTr,~T of the vessel weld material was selected to deternrine thie revised schedule.
For BFN-3, '75"./<i c)f.the expected 52'F shift is 39'F.
.Figure 6-2 is a plot ofth.e capsule shift in RT~ as a function ofEFPY.
The surveillance caps'ule material willexperience a shiA of30'F in <&.TrrpTover the lifi'eofthe plant.
TJsirrg a criteria of 75% ofthe expected shiift of the limiting vessel material (39'F), the capsule will experience this shifl; for the weld mate'rial at approximately 59 EFPY.
The removal schedule may be set to 24 EFPY in order to provide a reasonable schedule where a
significant shift wi,ll hiave occurred in the capsule karierikl. IIn support of this schedule, at 24 EFPY an. expected.25'F sliift will have occurred in'he surveillance capsule likutirig
'aterial, which will,provide sufficient data to detierniinel the r'eqirired vessel materi'al
'roperties.
In adciition, upon evaluation ofthe SSP materials, the BFN-3 'schedule can be further evaluated.
The fluence data, as determined from the surveillance capsule flux,wires at 24 EFPY, will provide an accu'rate indication. of nieutron fluence. As noted in Section 3, the current predicted fluence is.conservative.
The flux wires in the capsule withdrawn at 24 EFPY will be used to modify the prediicted fluence, rrieetiing the requirements of the BFN-3 Technical Specifications.
The use ofthe fluxwires at 24 EFPY willmeet the, requirements of 10CFR50 Appendix H and ASTME185.
22
GE Nuclear Energy GE-NE-B 1302026-00-01 Brooms Ferry 3 1/4T RPV Plate KIvs EFPY forPressure of 1140 psig 140 120 a
0 A a a
10EFPYP-T Curves 16EFPYP.T Curves
. 24EFPYP-T Curves 32EFPYP-T Curves Pressure Test 197'F Pressure Test 212'F Pressure Test 228'F Pressure Tcst242'F 40 20
-Required KI Weld KIR 0
0 12 16 24 32 Figure 6-1 KR vs. EFPY for Browns Ferry Unit 3 Weld Material 23
GE Nuclear Energy.
GE-NE-B1302026-00-01 45.000 40.000 75% ofvessel B,RTndt 139'F) 35.000 --
30.000 25.000-I-
Oz
~n nnn CV VVV CI
.1 A TlM A4. /A<OT.'X capsule L1WJ ILUl(4J I )
COrreSnOndi n n SurVei1 1nrlCe capsule wlthdIawal schedule 15.000 10.000 5.000 908AtYlwlArlfla11nlinrail nanna LXVVVllllllVJI\\k'VVi)L4l Y'VlllClll'4V
~
>1
~
~~- capsute wtmarawat schedule:
24 EFPY 0.000.
0.
12 16 20 24 28 32 36 40 44 4'2 56i 6i0 EFPY Figure 6-2 Predicted ShiA vs. EFPY, Browns Ferry Unit 3 Surveillance Capsule Weld 24
GE Nuclear Energy GE-NE-B1302026-00-01
- 7. CONCLUSIONS The purpose ofthe vessel surveillance program is to characterize the vessel properties as a function of irradiation. The current schedule for Browns Ferry Vnit 3 is a withdrawal schedule ofeight (8) EFPY for the'first surveillance capsule.
Schedules developed according to
- however, are general guidelines for all reactor pressure vessels.
The schedules do not take into account some specific characteristics of BFN-3 such as low fluence and good alloy chemistry for the capsule materials (0.10% - 0.11% copper), which results. in a low shift in RT~r. Ifthe first capsule is removed and tested according to the current schedule (8 EFPY), the data obtained for the plate specimens may be heavily affected by the scatter in Charpy results.
~ - Since early information on.a material similar to the limiting BFN-3 weld material can be obtained from the SSP to identify anomalous shifts, the BFN-3 surveillance schedule should be extended.
The schedule can be extended for the following reasons:
1.
Evaluation of similar data obtained from actual surveillance programs has shown that the measured
- fluence, shift and chemistry are bounded by expected values.
In particular, the BWR/4 data has.shown small RT~Y shifts for capsules removed from vessels similar to Browns Ferry Unit 3. Therefore, the surveillance capsule withdrawal schedule should be extended based on the conservatism in the calculated shift of RTm~
- 2. In addition, the P-T curves contain inherent conservatism, as noted in Section 4.
The fracture toughness values used for these calculations, are considered to be lower bound values and are significantly less than the crack initiation fracture toughness in the temperature range ofinterest. At operating temperatures, BFN-3 maintains more than adequate margins; the limiting condition is the pressure test.
This conservatism 25
0 GE Nuclear Energy GE-NE-81302026-OD-Q'I provides an added margin of safety'; therefore, the capsule withdrawal schedule can be modified.
3.
In addition, the SSP data willcomplement the available data, on surveillance specimens and also identify any anomalous information in the predicted values.
~ This
~
characterization will enhance the understanding of vessel embrittlement'ssues and' provide data for BFI'4-3 usiing a weld similar to the limiting weld material.
Hence the change in schedule for the BFN-3 surveillance specimens will n'ot have a significant effect on the understanding ofvessel irradiation'issues.
These reasons justify extending the withdrawal schedule while maintaining reactor safety margins, and provide for more accurate measured data neai EQL.. Therefore, the BFN-3 surveillance schedule can be extended.
The material property of most coricern is the fracture toughiiess of the vessel,",
the surveillance schedule should be based on evaluation of thit pi'op0rty.
Since the fracture toughness (KiR) is dependent on the shift in RT~z, the optimum EFPY for removal oftike capsule ensures useful data (measuring sigriificant shift)while identifying any anomalous conditions. Ifsuch an anomalous shiflt were to occur (which is unlikely), the margin between Ki (primary membrane fracture toughiiess) and KiR, as well as the inherent conservatism ofthe calculations, can provide a sufficie safety margin for extending the surveillance schedule.
In addition, operation ofBFN-3 follows the steam saturation curve; the operating temperatures are expected to be well~ in ~excess of the minimum required temperature.
As.shown. in Sectiion 6, the appropriate.dRTmi value sI:lee;ted was 75/o of the predicted beltline material 32 EE'PY change in hRT~.
Using this value to determine the appropriate shift in the capsule (hence the appi'opriate EFPY),,
the recommended withdrawal schedule for the first Browns Ferry anil 3 su*eillance capsule is 24 EFPY.
This proposed sch,edule meets the intent ofASTM E185-82, as the first capsule would be 26
'. GE Nuclear Energy GE-NE-B1302026-00-01 removed with the fluence being less than S.x 10" n/cm and the value of~TNDT would be D
less than 50'F; Removal of the capsule at the appropriate EFPY will provide. more meaningful data for,fracture toughness predictions.
27
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GE Nuclear Energy GE-NE-B1302026-00-01
- 8. REFERENCES "Reactor Vessel Material Surveillance Program Requirements,"
Appendix H to Part 50 ofTitle 10 ofthe Code, ofFederal Regulations, December 1995.
[2]
USNRC Docket Nos.
50-259, 50-260, and 50-296, "Revision to Technical Specifications Pertaining to Surveillance Requirement 4.6.A.3 and Bases Section 3.6/4.6 - (TAC 73141, 73142, 73143) (TS 270) - Browns Ferry Nuclear Plants, Units 1, 2, and 3", 8/3/89 Letter, P. Salas (TVA)to USNRC, "Browns Ferry Nuclear Plant (BFN) - Units 1, 2, and 3'- Generic Letter (GL) 92-01, Reactor Vessel Structural Integrity - Update to the 'Initial Reference NilDuctility Temperature (RTi~T) Chemical Composition and Fluence Values", 3/27/95
[4]
Letter, T. Abney (TVA).to USNRC, "Browns Ferry Nuclear Plant (BFN)-
Units 1, 2, and 3 - Generic Letter (GL) 92-01, Revision 1, Supplement 1, Reactor Vessel Structural Integrity
Response
to NRC Request for Additional Information", 9/3/98
[5]
ASTME185-82.
[6]
"Radiation Embrittlement of Reactor Vessel Materials," U.S. NRC Regulatory Guide 1.99, Revision 2, May 1988.
[7]
Letter PRGC-9803, Ray Carey (GE) to HL Williams (TVA), "New Bounding EFPY for Previously Generated P-T Curves Considering Power Uprate for Browns Ferry Units 2 2 3 Using Calculated Fluence and Estimated ESW Information" (DRF B13-02002-00), 12/11/98
[8]
"Bounding Assessment of BWR/2-6 Reactor Pressure Vessel Integrity Issues,"
BWR VIP-08NP, November 1995.
[9]
Letter, USNRC to I. Johnson (Commonwealth Edison Company), "Issuance of Amendments", 2/28/97
[10]
Letter PR92 981209 917,.RD Ryan (TVA) to L. Eichenberger (TVA), "Browns Ferry Nuclear Plant (BFN) - Work Impact,- Scope ofResponse to NRC Request for Additional Information (RAI) Regarding Pressure-Temperature (P-T) Curve Update for BFN Units 2 and.3.", 12/9/98 S.T. Rolfe and J.M. Barsom, Fracture and Fati ue Control in Structures, Prentice-Hall, Inc., New Jersey, 1977, p. 447.
28
GE Nuclear Energy II
. GE-'hJE-B1 302026-00-01
[12]
Ibid., p.:455.
[13]
ASME Section XI, Appendix A, 1992 Edition through Sumnser 1993 Addenda.
[14]
Letter, BH Shell (TVA) to USNRC, "Brogans Ferry Nuclear Plant (BFIQ, Sequoyah Nuclear Plant (SQiN), and Watts Bar Nuclear Plant (WBN) - Response to Generic -Letter 92-01 (Reactor Vessel Structural Integrity", Docket Nols. 50-259, 50-260, 50-296, 50-327', 50-328; 50*390, 50-391,', 7/7/92 29
- GE:Nuclear Energy GE-NE-B1302026-00-01 APPENDIXA ADJUSTED REFERENCE TEMPERATURE,(ART) CALCULATION 30
~I GE Nuclear Energy GE-NE-81,'302026-OO-Q'1 The ART is, according to Rev 2, a function of the initial RTNDq, the shift, and a margin term.
The shift iin.RTqtDq is dependent on the chenustry (speciflcally copper and nick'el) and fluence.
The: methods of Rev 2 are used to detern~ne the PBBS.
the procedure used depends on whether or not sur veillan.ce,specimen data is available.
In order to re-evaluate the surveillance, specimen program schedule, the ART for both tahe
~
vessel itself and the specimens must be'calculated.
-For Browns Feriy Unit 3, surveillance
.specimens have not been tested, which requires the procedUre of evaluating ART without surveillance specimens, as described below.
The ART for each beltline material is given by the following eqiuation:
ART =. InitialRT~DT + dlilTlqr+ iVlarg~in
'nitial RTNDq is the reference temperature deterrnin~ed ~according to ASME Sectio~ III, Paragraph NB-2331 for the, unirradiated material.
The shift in the reference temperature, bRTqtDY, is determined by a combination of tlhe chemistry and fluence as shown by. Equation (2):
CF g p(0.28-0.10log f)
NDT (2)
The CF.is the chemistry factor (dependent on the copper andI nickel content) and is
'etermined from the tables for weld and base miatel.ial in Rev 2,.
'the.fluence, f, at any depth in the vessel wall, is determined by Equation (3),
(3) 31
0 "GE:Nuclear Energy GE-NE-B1302026-00-01 where f~ is the calculated neutron fluence at the vessel ID, and x is the depth into the vessel measured from the inner (wetted) surface.
For these. calculations, the value off(at 1/4T) used was 7;8 x 10'/cm, obtained frofn [7].
The Margin term is'included to obtain'the upper bound values of the ART.
Since the Margin term provides upper bound values of the ART (which is a function of CF and fiuence), it is unnecessary to add extra conservatism by using the upper bound fluence.
Any:uncertainty in the fluence is captured by the Margin term.
The Margin term is given by Equation (4):
Margin = 2ger.', + cr~
(4) where err= standard deviation ofthe initial RTNDT.
a~ = standard deviation for dXTNDY The standard deviation for BRTNDy, an, is assumed to be 28'F for welds and 17'F for base metal, except that a~ need not exceed 0.50 times the, mean lKTNDT[5]; The conservative nature. ofthe initial RTNDT determination generally results in a~ being equal'o zero.
Using Equations (1) through (4),
the ART can be calculated for plants with.no surveillance data, including Browns'Ferry Unit 3.
EX4MPLE CALCULATION To better illustrate the ART methodology, the following calculation was,performed for the limiting BFN-3 vessel. weld material (Electroslag, Weld); this, material's chemistry and initial RTNDT bound that for the weld material in the surveillance capsule. '(The %Cu is 0.24 compared to the capsule material 0.11%; the %Ni is 0.37 compared to the capsule 32
Cl GE.Nuclear Energy
'E.-NE-81302026-00-01
.material 0.28%, and the initial RTNDT is 23.1'F compared to the c'apsul'e 10'F.) The data was obtained from,[7] and [14]:
InitialRTNDT.
Nickel:
Copper:
Peak Fluence:
Wall Thickness:
23.1'F 0.37%
0 24ogo 7.8 x 10 n/crn (32.EFPY at l/4T) 6.13 inc:hes From Table.2 ofRev 2, the chemistry factor for this heat of material is 141.
The fluence at the 1/4T depth, 7.8 x 10'/cm', was used.
The change irt reference temperature, hRTNDT, is calculated ac:cording to Equation.(2):
+ 0 19(0.28%.10log0.19) dRTgDr = 141* 0.166
== 51'.8"F -": 52'F For the margin term, the standard deviation of the initial RT~~, ar, is 13'F as protidied by [10]. The standard deviation for LKTND>,.cr>, is 25'Ft ak it is v'reld metal.
Therefore, using equation (1), the ART,at 32 EFP Yfor the Electroslag Weld is:
ART = 23 + 52 + 58 =
133oF'his calculation was relpeated for all of the vessel beltlline materials.
The results of the calculations for all the,beltliine materials and tlhe surveillance capsule materials are shown in Table A-1.
Figure A.-l is a plot.of,the ART against EFPY for the expected plant',
lifetime for the linuting (at EOL) vessel lplat.e and Weld r@atdrialls.
'3
GE Nuclear Energy GE-NE-B1302026-00-01 Low-Int Shell Thickness =
6.13 inches Low-Int Shell:
32 EFPY Peak I.D. fluence =
32 EFPY Peak 1/4 T fluence =
1.1E+18 n/cm 7.8E+17 n/cm Lower Shell Thickness =
6.13 inches Lower Shell:
32 EFPY Peak I.D. fluence =
32 EFPY Peak 1/4 T fluence =
1.1E+18 n/cm 7.8E+17 n/cm COMPONENT PLATES:
I.D.
HEAT
%Cu
%Ni CF INITIAL 32 EFPY RTndt 6 RTndt 32 EFPY 32 EFPY MARGIN SHIFT, ART Lower Shell Lower Shell Lower Shell Low-Int Shell Low-Int Shell Surveillance Low-Int Shell 6-145-4 6-145-7 6-145-12 6-145-1 6-145-2 6-145-6 C3222-2 C3213-1 C3217-2 C3201-2 C3188-2 C3188-2 B7267-1 0.15 0.13 0.14 0.13 0.10 0.10 0.13 0.52 0.58 0.66 0.60 0.48 0.51 0.51 106 90 101.5 91 65 65 88 10
-20 4
-20
-20
.-30
-20 39.1 33.2 37.4 33.5 24.0 24.0 32.4 17.0 16.6 17.0 16.8 12.0 12.0 16.2 34.0 332 34.0 33.5 24.0 24.0 32.4 73.1 66.3 71.4 67.1 47.9 47.9 64.9 83.1 46.3 67.4 47.1 27.9 17.9 44.9 WELDS:
Longitudinal Surveillance Circumference ESW'SW" D55733 0.24 0.11 0.09 0.37 0.28 0.66 141 81 117 23.1 10
-40 51.8 29.8 43.1 13 0
0 25.9 14.9 21.6 58.0 29.8 43.1 109.8 59.7 86.2 132.9 69.7 46.2 ESW chemistry based on NRC SER that was issued in February 1997 for Commonwealth Edison Co.
Specific weld heat chemistries are not available.
" Chemistry based upon NRC Letter 92-01 Table A-1 Browns Ferry Unit 3 RPV Beltline Material Data 34
- GE Nuclear Energy GE-NE-B $302026-00-01 140 120
Piaie
~weia 40 20 0
.0 12 16 EFPY 20 24 28 32 Fieure A-1 A4T vs EFPY 35
ENCLOSURE 2
TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)
UNIT 3 PROPOSED REVISION TO THE UNIT 3 REACTOR PRESSURE VESSEL MATERIAL SURVEILLANCE PROGRAM COMMITMENT Material testing results from the SSP and/or the first Unit 3 capsule will be used to develop an appropriate schedule for the second surveillance capsule.
4i l'
~
l