ML18039A388
| ML18039A388 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 06/12/1998 |
| From: | Abney T TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| TAC-MA1153, NUDOCS 9806220188 | |
| Download: ML18039A388 (32) | |
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CATEGORY j.
REGULAT INFORMATION DZSTRIBUTZO SYSTEM (RIDS)
ACCESSION NBR:9806220188 DOC.DATE: 98/06/12 NOTARIZED: NO DOCKET I FACIL:50-296 Browns Ferry Nuclear Power Station, Unit 3, Tennessee 05000296 AUTH.NAME AUTHOR AFFILIATION ABNEY,T.E.
Tennessee Valley Authority RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)
SUBJECT:
Provides response to staff request for addi info transmitted by NRC ltr dtd 980519 re plant request for revised relief request 3-ZSI-1.
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Tennessee Valley Authority. Post Office Box 2000. Decatur. Alabama 36609 June 12, 1998 U.S. Nuclear Regulatory Commission ATTN:
Document Control Desk Washington, D.C.
20555 Gentlemen:
In the Matter of
)
Tennessee Valley Authority
)
Docket No. 50-296 BROWNS FERRY NUCLEAR PLANT (BFN)
UNIT 3 REVISED RELIEF REQUEST 3 ISI 1 r REACTOR PRESSURE VESSEL (RPV)
SHELL WELDS r AUGMENTED EXAMINATIONAND AMERICAN SOCIETY OF MECHANICAL ENGINEERS (ASME)
SECTION XI INSPECTIONS
RESPONSE
TO NRC REQUEST FOR ADDITIONAL INFORMATION (TAC NO. MA1153)
This letter provides Tvh's response to the staff's request for
/
additional information (RAI) transmitted by NRC letter dated May 19,
- 1998, regarding BFN request for relief 3-ISI-1 (revised).
TVA letter to NRC dated January 22,
- 1997, contained the original relief request 3-ISI-1.
NRC denied this relief request by letter dated July 8,
- 1997, and documented four reasons for the denial.
TVA submitted revised request for relief 3-ISI-1 by letter dated February 17, 1998, that provided supplemental technical justification based on industry initiatives and information to assist NRC in reconsidering BFN's request for relief.
In addition, TVA responded to NRC's comments regarding the denial of the original request for relief.
'tt806220188 980612 PDR ADOCK 05000296 6
PDR~
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U.S. Nuclear Regulatory Commission Page 2
June 12, 1998 The enclosure to this letter provides additional background information and'ists the specific NRC questions and the corresponding TVA responses.
There are no new commitments contained in this letter.
If you have any questions, please telephone me at (256) 729-.2636.
S'e el T. E. Ab e Manager of
'-ensi an dustry Aff girs E
losure cc (Enclosure):
Mr. Michael
. Anderson Res ch Center 2151 North Boulevard P.O.
Box 1625 Idaho Falls, Idaho 83415-2209 Mr. Harold O. Christensen, Branch Chief U.S. Nuclear Regulatory Commission Region II 61 Forsyth Street, S.W.
Suite 23T85
- Atlanta, Georgia 30303 NRC Resident Inspector Browns Ferry Nuclear Plant 10833 Shaw Road
- Athens, Alabama 35611 Mr. Albert W.
De Agazio, Project Manager U.S. Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852
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ENCLOSURE TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)
UNIT 3 REACTOR PRESSURE VESSEL (RPV)
SHELL WELDS ~
AUGMENTED EXAMINATIONAND AMERICAN SOCIETY OF MECHANICAL ENGINEERS (ASME)
SECTION XI INSPECTIONS
RESPONSE
TO NRC REQUEST FOR ADDITIONAL INFORMATION FOR REVISED RELIEF REQUEST 3-ISI-1 Introduction The following are TVA's responses to questions asked by the staff in NRC letter to TVA, dated May 19, 1998..
These responses are in support of the NRC staff review of TVA's revised relief request 3-ISI-1.
This request for relief asks for a delay of one operating cycle in performing the first successive examination of 15 shell weld flaws found in the BFN Unit 3 reactor pressure vessel (RPV).
These flaws were originally identified during the conduct of the augmented RPV examinations performed during the extended Unit 3 Cycle 5
outage in the Fall of 1993.
The successive examinations are required to be performed during the first and second inspection periods of the second ISI program inspection interval.
The NRC questions and the corresponding TVA responses are provided below.
Provide justification that each of the 15 flaws found in the BFN-3 vessel during the Fall 1993 augmented examination are subsurface by considering the '(1) uncertainties of the non-destructive examination (NDE) method, (2) size, and (3) location of each flaw.
TVA Res onse As reported in TVA's original report on the RPV examination, the qualification demonstration was designed using available samples to match as closely as possible the requirements of Appendix VIII, Supplement 4 (for RPV clad-to-base metal interface region) and Supplement 6 (for shell welds other than clad-to-base metal interface).
As part of these requirements,
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the examination procedure protocol required that the flaw initial sizing acceptance criteria be as follows:
"A two-dimensional plot with the depth estimated by ultrasonics plotted along the ordinate and the true depth plotted along the abscissa shall be compiled.
A flaw shall be successfully sized when the following statistical parameters have been met.
(a) slope of the linear regression line is not less than 0.70 (b) the mean deviation of flaw depth is less than 0.25 inch (c) correlation coefficient is not less than 0.70 (d) flaw length estimated by ultrasonics shall be the true length minus
(-) 4 inch, plus
(+)
1 inch" For the 15 flaws reported and analyzed, the demonstration results for the Supplement "6 requirements are the most appropriate to use since none of the flaws are surface connected.
The flaw depth (through-wall dimension) and length measurement accuracies demonstrated during the examination qualification process are as follows:
Parameter:
Root Mean S uare (RMS) Error Flaw Depth (through wall)
Flaw Length 0.13 inches 0.44 inches An analysis of the 15 flaws by TVA, in accordance with the requirements of ASME Section XI, Subsection IWB-3500 and with the accuracy for flaw depth shown above included, indicates that the flaws can all be classified as subsurface.
Attachment A is a compilation of the analysis performed.
The computations for this table were performed in the most conservative manner so as to present the flaws in the most limiting configurations when compared with the ASME Code requirements.
It should be noted that an analysis performed in accordance with Subsection IWB-3600 (for flaws that exceed the IWB-3500 criteria) does not prescribe a method for addressing flaw sizing uncertainties.
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Part of the basis for the requested relief is that the General Electric (GE)
GERIS 2000 system that was used to perform the
.previous RPV examination is not available for the refueling outage scheduled to begin in October 1998.
TVA asserts that a modified system similar to the previously, utilized system would need'o be constructed due to availability issues.
GERIS 2000 is Performance Demonstration Initiative (PDI) qualified.
PDI qualification of equipment, procedures, and personnel demonstrates that examination and evaluation techniques are repeatable.
Please provide information on alternate vendors that are PDI qualified and have equipment that are available for the upcoming Cycle 8 outage.
TVA Res onse TVA has contacted two other NDE service vendors who are PDI qualified and are known to have the capability of performing these type of RPV examinations.
Southwest Research Institute and Framatome Technologies, Inc. were contacted for information on the availability of qualified individuals and NDE systems to perform the Unit 3 RPV flaw successive examinations.
Both vendors expressed interest in providing the service to perform the needed examinations.
- However, the availability of their respective RPV examination personnel and equipment was not readily determinable.
In addition, neither vendor could state whether or not a determination could be easily made that their specific RPV examination tooling and systems were. fully compatible with the BFN Unit 3 RPV and would provide the appropriate level of information and quality as that provided by the GERIS-2000 system.
An analysis, based upon specifications provided by TVA, would have to be performed by each vendor to determine if their available NDE system would meet BFN examination needs.
In addition, an evaluation would have to be made to determine whether special tool modifications would, have to be made to the vendor's equipment to allow for physical access to the previously identified flaw areas.
In order to maintain the ability to readily compare any successive examination results to the original examination results with the best possible correlation of the sets of
- data, TVA prefers to continue to plan for General Electric to perform RPV examinations.
As shown in TVA's analysis, the subject examinations are not technically warranted under the given circumstances.
Therefore, the need to procure costly E-3
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and exigent RPV examination services solely for the purpose of compliance with the ASME Code requirements is likewise unwarranted.
NRC Re est It is stated on Page 3 of "General Electric Corporation Flaw Evaluation Extending Service Lifetime of Browns Ferry Unit 3 Reactor Pressure vessel Flaws to Forty Years (Design Service Lifetime)" which is Attachment 3 to your February 17,
- 1998, letter:
"As operation continues past 12 EFPY, and vessel irradiation increases, the leak test temperature also increases as required to meet 10 CFR 50, Appendix G
requirements for the vessel P-T curves.
As a result, the (T-RT>D~) temperature difference increases with time for non-beltline weld locations such as are evaluated here."
To substantiate this statement, please (a) provide the leak test, temperature as a function of years of operation; (b) provide the fluence for Weld H23 to clearly rule out any meaningful embrittlement to Weld H23 due to fluence; and (c) confirm that the increase in T exceeds the increase in RT>D~
for Welds
- H34, H45, V4, and VFW during 40 years of operation.
"...(a) provide the leak test temperature as
.a function of years of operation..."
As part of a planned reactor power uprate initiative, TVA recently submitted a Technical Specification (TS)
Change (No.
393) which contained revised pressure-temperature (P-T),
operating curves for the Units 2 and 3 reactor vessels.
Attachment C shows the Unit 3 associated P-T curves.
The TS changes were submitted by TVA letter to NRC, dated March 3, 1998.
The revised P-T curves reflect the results of a recent analysis which extends the validity of the 'P-T curves to 32 Effective Full-Power Years (EFPY).
To date, Unit 3 has accumulated approximately 7.0 EFPY of operation.
The P-T
- curves, based upon the new calculations, will resu'lt in a minimum vessel leak test temperature of 210 degrees Fahrenheit
('F)
(Attachment B, Figure 1), at a vessel, pressure of 1000
- psig, when the Unit 3 limiting operating conditions are adjusted from the 12 EFPY based P-T curves to 32 EFPY based curves.
Plant operation and reactor coolant system pressure test temperatures will conform to the limits of these new EA
it
~ ~'r P
operating curves.
This will ensure that leak test temperatures will be adjusted to allow for shifts in RT>>~
of the Unit 3 RPV beltline materials in accordance with the requirements of NRC Regulatory Guide 1.99, Revision 2.
The P-T operating curves are periodically adjusted based upon calculations of accumulated fluence which are validated by testing of RPV material specimens.
The use of the 32 EFPY curves test temperatures will ensure that any needed compensation for the change in RPV RT>>> will be adequately accounted for by the limitations imposed with the use of the new curves.
"...(b) provide the fluence for Weld H23 to clearly rule out any meaningful embrittlement of Weld H23 due to fluence,..."
As stated in TVA's letter dated February 17,
- 1998, the calculated beltline region accumulated fluence for 12 EFPY of operation is 0.0401 x 10'/cm on the RPV wall.
Based upon the power uprate 32 EFPY operating P-T curve information, the expected fluence in the beltline region is calculated to be" 0.064 x 10" n/cm't the
~~ T location (vessel wall thickness).
The P-T curves proposed to be used with the power uprate are based on a limiting beltline Adjusted Reference Temperature (ART) of 113 'F.
In the beltline region the elevation of the bottom of the active fuel (BAF) is 216.3 inches above the bottom of the vessel.
As compared (Attachment B, Figure 1),
the calculated RPV fluence at an area 135 inches above the BAF (at 351.3 inches) will decrease to a level of approximately 50 percent (8) of the value at the peak in the beltline region.
RPV circumferential weld H23 (designated as C-2-3 in figures and tables previously provided) is located at an elevation of 391.5 inches above the bottom of the RPV.
The H23 weld is the circumferential weld located between the vessel shell courses Mk-58 and Mk-59.
This is the first circumferential shell weld located above the beltline region shell course.
This area is 175.2 inches above the BAF region and is located 19.5 inches above the top of the designated beltline region.
Consequently, the fluence at the H23 weld will be considerably less than that encountered in the beltline region and has been conservatively estimated for the 32 EFPY of plant operation to be between 10% and 20% of the beltline peak fluence value.
This corresponds to fluence values for the H23 (C-2-3) weld of approximately 0.0064 to 0.0128 x 10" n/cm'.
This fluence is 2.5 to 4.2 times less than the beltline region expected fluence.
Materials test report data for the Unit 3 RPV H23 shell weld fabrication filler wire indicates that the percent Copper and Nickel content of the filler wire is 0.09% and 0.72%, respectively.
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Calculations in accordance with the guidelines in Regulatory Guide 1.99, Revision 2, using the estimated fluence shown above result in an expected shift of the RT>>>
(hRTND~)
tempe'rature of 15.4 'F for the H23 weld.
A tabulation of the expected fluence embrittlement effects to the H23 (C-2-3) weld is shown in the following table.
These values are calculated in the same manner as the Table 1 information for the beltline region provided in TVA's February 17, 1998 letter for the revised request for relief.
BROWNS FERRY UNIT 3 RPV H23 SHELL WELD INFORMATION FOR 32 EFFECTIVE FULL-POWER YEARS (EFPY)
Neutron fluence for the H23 (C-2-3) weld at the end of 32 EFPY Beltline Region Circumferential Weld between Unit 3 RPV Shell Courses Mk-58 and Mk-59 0.0064-0.0128 x 10 n/cm~
Initial (unirradiated) reference temperature
-40 F
Weld Chemistry Factor (CF) 119
,Weld copper content 0.09 Weld Nickel content 0.72 0
Increase in reference temperature due to irradiation ~TNT) 15.4 F
Margin term 15.4 F
Mean adjusted reference temperature (ART)
-24.6 F
Upper bound adjusted reference temperature (ART)
-9.2 F
The estimated ART of -9.2 'F for the H23 weld is well below the beltline region ART.
The estimated shift for the H23 shell weld is insignificant when compared to the estimated shift for the beltline region welds.
BFN Unit 3 operation, in accordance with the limitations based upon the 32 EFPY P-T operating curves, will ensure that RPV leak test (hydrostatic
0
and/or system pressure tests) and operating temperatures will be sufficiently adjusted to accommodate shifts in the limiting beltline region embrittlement and will therefore exceed any expected shift in the H23 weld RTNpT.
As with the use of the 32 EFPY P-T curves in item 1 above, the value of the expected shift for the H23 weld RTNpT will be adequately bounded by the use of the 32 EFPY P-T curves.
"...,(c) confirm that the increase in T exceeds the increase in RT>>T for Welds H34, H45, V4, and VFW..."
BFN Unit 3 RPV shell course circumferential welds H34 (C-3-4),
H45 (C-4-5),
and VFW (C-5-FLG) are located at levels 524.5,
- 573, and 706 inches; respectively, above the bottom of the RPV.
Shell weld V4 (V-4-B) for the vessel Mk-59 shell course runs vertically from the level of 524.5 to 573 inches.
These four welds are located at distances between 308.2 to 489.7 inches above the BAF region.
Vessel welds in these areas are much less affected by fluence embrittlement because of the reduced direct exposure.
In addition, vessel internal apparatus, such as the steam dryer, attenuates the fluence and provides extensive shielding for the RPV wall materials in the upper regions of the RPV.
Fluence embrittlement of the RPV shell welds in these areas results in little or no significant shift in their individual RT>>T values.
Therefore, any adjustment in the P-T operating curves resulting from the RTNpT shift of the beltline region will significantly exceed any expected change in RT>>T for the H34,
- H45, VFW, and V4 welds.
Therefore, operation of the BFN Unit 3 RPV at the limits validated by the 32 EFPY operating curves (Attachment C) has compensated for any expected shift in the RPV RT>>T for all of the vessel welds.
On Page 11 of "Browns Ferry Unit III Flaw Evaluation Handbook" GENE-523-120-0992, (Reference 1 to Attachment 3 of your February 17, 1998 letter),
Equation (2-5) is reported to be cited from the paper by Paris and Sih (Reference 11).
Please verify its accuracy and provide the Equation No.
From Paris'aper that you cited.
TVA Res onse The GENE-523-120-0992 analysis Equation 2-5 is an appropriate interpretation of Equation 167 from the Paris and Sih paper (Reference 11).
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Equation (2-5) is derived from Equation 167 in Reference 11:
K1 P /
(2mc)'
where "P" is load per unit depth and "c" is the distance from the application of P to the crack tip.
The equation used in GENE-523-120-0992 (Equation 2-5) is:
K1 2@clad tc1ad /
(2zc) 1/2 The term a 1,d t 1,d is equivalent to P, and the factor of 2 is applied to account for loading on both crack faces.
The "c" term in Equation 2-5 is taken from the center of the clad
- depth, which is where the equivalent load P would be applied, to the crack tip.
Therefore, TVA considers that Equation 2-5 on page 11 of GENE-523-120-0992 is an appropriate interpretation of Equation 167 from the paper by Paris and Sih (reference 11).
Ci'n addition to the above information, TVA'has identified that flaw size inspection data values provided for the flaw designated 12-015, located in circumferential weld H34 (C-3-4),
need to be corrected.
These values are being amended to eliminate differences shown between the previously provided copies of the General Electric GERIS 2000 Examination Summary Sheets for RPV circumferential shell weld H34 (C-3-4) and vertical shell weld V4 (V-4-B)'.
These data sheets are shown as pages 6 of 9 and 9 of 9 in Attachment 2 of TVA's February 17, 1998 letter, transmitting the revised request for relief 3-ISI-1.
This specific flaw data was reported on both data sheets because the vertical shell weld intersects the circumferential shell weld and the flaw indication is encountered in both the circumferential and the vertical shell welds'xamination volumes.
This flaw is circumferentially oriented and is actually located in the H34 (C-3-4) circumferential shell weld.
The discrepancies in the values are those shown for the flaw's Z position (Z Pos),
the "S" value, and the "T wall" value.
The correct values are:
Z Pos:
1.16 inches 0
94 inches "T wall":
0.444 inches All other parameters for this flaw indication are correct and remain the same as originally reported.
These incorrect values were also reported in the original inspection data report for the ASME Section XI, NIS-1 Outage Summary E-8
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Report submitted following the Unit 3 Cycle 5 refueling outage by TVA letter to NRC dated March 6, 1995.
These differences were the result of typographical and data transposition errors made during the creation of the original Summary Report data sheets.
TVA is forwarding corrected'IS-1 Summary Report examination data sheets for these Unit 3 reactor pressure vessel shell welds in a separate correspondence.
The calculation results using this data, and,previously reported to you in TVA's letter of February 17,
- 1998, are not affected by these changes/corrections.
The calculations shown in Attachment A for the responses shown here have been verified to use the correct values for the H34 (C-3-4) flaw size parameters.
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~ 0 ATTACHMENT A BROWNS FERRY NUCLEAR PLANT UNIT 3 REACTOR PRESSURE VESSEL FLAW EVALUATION FOR SUBSURFACE DESIGNATION CONSIDERING NDE METHOD UNCERTAINTIES Indication ff Weld ID Indication No. T wall=2a 2a+.13 a
.4a s
Including Original Original Original RMS Error s/Aa Original Original Results "a" = (2a+.13)/2.4'[(2a+.13)/2]
s-.13 (s-.13)/(.4'[(2a+.13)/2))
Including Including Including Including RMS Error RMS Error RMS RMS Error Error New Results 1
2 3
4 5
6 7
8 9
10 11 12 13 14 15 C4-FLG C-5-FLG C-5-FLG C-5-FLG C-5-FLG CWS C~
C~
C~
C44 C~
CGA C~
C-24 C-24 20407 20412 20409 20411 20-008 1676 16475 12-148 12-116 12415 12-145 12C69 12-144 08-026 OIWI67 0.4 OAS 0.38 0.39 0.39 OA4 0.3 0.511 0.62 OA44 0.39 0.34 0.325 0.41 0.37 0.53 0.61 0.51 0.52 0.52 0.57 0.43 0.641 0.75 0.574 0.52 0.47 OA55 0.54 0.5 0.2 0.08 0.82 0.24 0.096 0.95 0.19 0.076 0.84 0.195 0.078 1.04 0.195 0.078 0.83 0.22 0.088 2.53 0.15 0.06 2.83 0.256 0.102 0,43 0.31 0.124 2.6 0.222 0.089 0.75 0.195 0.078 2.4 0.17 0.068 1.94 0.163 0.065 0.75 0.205 0.082 1.56 0.1 85 0.074 1.47 10.250 Subsurface 9.896 Subsurface 11.053 Subsurface 13.333 Subsurface 10.641 Subsurface 28.750 Subsurface 47.167 Subsurface 4.207 Subsurface 20.968 Subsurface SA46 Subsurface 30.769 Subsurface 28.529 Subsurface 11.538 Subsurface 19.024 Subsurface 19.865 Subsurface 0.265 0.305 0.255 0.26 0.26 0.285 0.215 0.321 0.375 0.287 0.26 0.235 0.228 0.27 0.25 0.106 0.122 0.102 0.104 0.104 0.114 0.086 0.128 0.15 0.115 0.104 0.094 0.091 0.108 0.1 0.69 0.82 0.71 0.91 0.7 2.4 2.7 0.3 2.47 0.62 2.27 1.81 0.62 1.43 1.34 6.509 6.721 6.961 8.750 6.731 21.053 31.395 2.340 16.467 5A01 21.827 19.255 6.813 13.241 13A00 Subsurface Subsurface Subsurface Subsurface Subsurface Subsurface Subsurface Subsurface Subsurface Subsurface Subsurface Subsurface Subsurface Subsurface Subsurface Reference ASME Section XI, Figures IWA-3320-1 and IWA-3380-1 Ifs is greater than or equal to.4a originally (i.e. the ratio is > 1.0); then the flawwas judged to be subsurface.
Ifthe conservatively adjusted 's,'s-.13), is greater than or equal to the conservatively adjusted '.4a,'(.4'[(2a+.13)/2]) (i.e. the ratio is > 1.0); then the flaw is stilljudged to be subsurface.
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ATTACHMENT 8 250 240 230 220 210 6 200
'Df 190 180 5
170 I-o 160 150 140 130 120 110 100 10 15 20 EFPY 25 30 35 Figure 1. Minimum Leak Test Temperature vs. EFPY for Browns Feny 3 I
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ATTACHMENTC 1600 COMBINED PRESSURE-TEMPERATURE CURVES FOR BROWNS FERRY UNIT 3 (32 EFPY) 1500 Curve 2 1400 1300 1200 1100 a.
1000 0
N 900 uj
, tr'.
O 800 I-O 700 I
600 o"D cn 500 Curve 1 - Pressure Test with Fuel in the Vessel Curve 2-Heatup/Cooldown Core Not Critical Curve 3-Heatup/Cooldown Core Critical Operation 400 300 200 100 0
50 100 150 200 250 300 350 400 MINIMUMREACTOR VESSEL METALTEMPERATURE ('F)
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