L-96-112, Forwards 10CFR50.61 Evaluation of PTS of Rv Beltline Matls, Completing Commitment in .Proprietary & Nonproprietary Versions of CEOG Rept, Application of Rv Surveillance Data... Also Encl.Proprietary Rept Withheld
| ML17228B497 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 05/14/1996 |
| From: | Bohlke W FLORIDA POWER & LIGHT CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML17228B498 | List: |
| References | |
| L-96-112, NUDOCS 9605210538 | |
| Download: ML17228B497 (38) | |
Text
CATEGORY 1 REGULATO INFORMATION DISTRIBUTION SYSTEM (RIDS)
ACCESSION NBR:9605210538 DOC.DATE: 96/05/14 NOTARIZED: YES DOCKET FACIL:50-335 St. Lucie Plant Unit I, Florida Power' Bight Co.
05000335 50-389 St. Lucie Plant, Unit 2, Florida Power 6 Light Co.
05000389 AUTH.NAME AUTHOR AFFILIATION BOHLKEPW.H.
Florida Power E Light Co.
RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk) p~sAZ
SUBJECT:
Forwards 10CFR50.61 evaluation of PTS of RV beltline matls, C
completing commitment in 951124 ltr.Proprietary 6
nonproprietary versions of CEOG rept, "Application of RV A
Surveillance Data..." also encl. Proprietary rept withheld.
T DISTRIBUTION CODE:
APOID COPIES RECEIVED:LTR g ENCL /
SIZE:
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~ 7A TITLE: Proprietary Review Distribution Pre Operating License 6 0 crating R E NOTES:
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INTERNAL: ACRS OGC/HDS3 EXTERNAL: NRC PDR COPIES LTTR ENCL 1
1 1
6 6
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PLEASE HELP US TO REDUCE WASTE!
CONTACT THE DOCUMENT CONTROL DESK, ROOM OWFN SD-5(EXT. 415"2083)
TO ELIMINATE YOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED1
<\\.
TOTAL NUMBER OF COPIES REQUIRED: ITTR 12 ENCL
Florida Power &Light Company, P.O. Box 128, Fort Pierce, FL 34954-0128 May 14, 1996 U.
S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.
C.
20555 L-96-112 10 CFR 50.4 10 CFR 50.61 10 CFR 2.790 RE:
St. Lucie Units 1 and 2
Docket No. 50-335 and 50-389 10 CFR 50.61 Evaluation of Pressurized Thermal Attachments A and B provide the 10 CFR 50.61(b)(1) pressurized thermal shock submittals for the St. Lucie Units 1 and 2 reactor vessel beltline materials.
The evaluations determined the projected reference temperature (RTpta) at end of license (EOL) for the reactor vessel beltline materials of each reactor vessel.
The EOL RTpga values were compared against the regulatory limit of 270 degrees Fahrenheit
('F) and determined to be acceptably below the limit.
This submittal completes the commitment in our letter, L-95-315, dated November 24,,1995.
In addition, one (1)'opy each of the proprietary and non-proprietary versions of Combustion Engineering Owners Group report, Application of Reactor Vessel Surveillance Data for Embrittlement Management, is enclosed.
CEN-405-P Revision 2 is the proprietary version and CEN-405-NP is the non-proprietary version of the report which is referenced in our submittal.
CEN-405-P contains proprietary information, the disclosure of which would comprom'ise trade secrets and commercial information considered by Combustion Engineering, Inc.
as privileged or confidential.
Pursuant to 10 CFR 2.790(a)(4),
FPL requests that the enclosed report CEN-405-P be withheld from public disclosure.
The affidavit, required by 2.790(b),
supporting this request and executed by an authorized representative of Combustion Engineering, Inc., is provided as Attachment C.
NRC approval of this material is requested by April 1, 1997, to support design of the fuel for St. Lucie Unit 1 Cycle 15. Please contact us if there are any questions.
Ve g rul
- yours, W.H. Bohlke Vice President St. Lucie Plant WHB/GRN cc:
Stewart D. Ebneter, Regional Administrator, Region II, USNRC Senior Resident Inspector, USNRC, St. Lucie Plant r Fn4.
nY an FPL Group company
L-96-112 Attachment A
Page A-1 ST LUCIE UNIT 1 REACTOR VESSEL BELTLZNE MATERIALS END OF LICENSE ASSESSMENT OF RT-PTS Z.
Discussion The purpose of this assessment is to show that the St. Lucie Unit 1 projected reference temperature (RT>>,) at end of license (EOL) remains below the PTS screening criteria when using the methodology in the most recent 10 CFR 50.61'.
The revised rule states that "licensees shall consider plant specific information that could affect the level of embrittlement including data from any related surveillance program results."
This submittal provides updated projections of RTpyg for the St.
Lucie Unit 1 reactor vessel beltline materials with improved accuracy due to incorporation of available surveillance data from the St. Lucie Unit 1 and Duquesne Light Co. Beaver Valley Unit 1 surveillance programs corresponding to these materials.
Using this surveillance data it will be shown that the St.
Lucie Unit 1
reactor vessel limiting beltline material, axial.welds, 3-203 A, B and C, EOL RTpgg is 213 F which is 57'F below the 270'F screening limit. All other materials are more than 100'F below their respective screening limits.
ZZ.
Introduction The initial RT<~, chemistry values and end of license RTpyg values have been reported for the St. Lucie Unit 1 reactor vessel beltline materials".
The NRC has incorporated these initial values into its Reactor Vessel Integrity Database (RVID).
The St. Lucie Unit 1 initial RT>>~ and chemical composition are summarized in Table 1
for the beltline plate and weld materials.
For projections of EOL RTpgg there exist additional information for the St. Lucie Unit 1 reactor beltline controlling weld, girth weld and controlling plate material from reactor vessel material surveillance programs.-
St. Lucie Unit 1 Surveillance Program The St. Lucie Unit 1 (SL-1) reactor vessel and surveillance program materials were fabricated by Combustion Engineering (CE).
The surveillance program was designed to meet ASTM E185-73.
Two capsules have been removed and tested as part of the original surveillance program' The data from these capsules provides information on the embrittlement behavior of the SL-1 reactor vessel lower shell beltline plate (Heat C-5935-2, Code C-8-2) and the intermediate to lower shell girth weld 9-203 (heat 90136).
The data indicates that these materials embrittle in a manner that is within 1@~ of their calculated best fit projections as outlined in
L-96-112 Attachment A
Page A-2 10 CFR 50.61.
Therefore, we propose using this surveillance data to determine bRT>>, and RT>>~ in accordance with 10 CFR 50.61.
Surveillance data for the controlling lower shell axial weld seams 3-203A, B
and C
(heat 305424) is not available in the SL-1 surveillance program.
- However, in addition to fabricating vessels for its Nuclear Steam Supply System (NSSS),
CE also fabricated vessels for Westinghouse (g)
NSSS designed vessels.
Therefore, surveillance materials from CE fabricated, g designed NSSS vessels may represent additional data applicable for the CE fabricated St. Lucie Unit 1 vessel.
The Duquesne Light Company, Beaver Valley Unit 1 is a g designed NSSS with a CE fabricated reactor vessel.
The Beaver Valley Unit 1 (BV-1) and SL-1 vessels were both fabricated during the same period by CE in Chattanooga, Tennessee.
The weld selected by Q for the BV-1 surveillance program is identical to the SL-1 lower shell axial welds.
Both welds were fabricated using the submerged arc weld process with weld wire heat 305424 and Linde 1092 flux lot 3889.
Use of the BV-1 surveillance data for the SL-1 welds 3-203A, B and C will permit a more accurate assessment of ART~ and RT>>,.
Beaver Valley Unit 1 Surveillance Program The Duquesne Light Co.
Beaver Valley Unit 1 reactor vessel and surveillance weld were fabricated by Combustion Engineering (CE).
The BV-1 surveillance program was designed to meet ASTM E185-73 Three capsules have been removed and tested as part of the original surveillance program"'.
Updated fluence values reported with the most recent capsule analysis'ill be used with the BV-1 capsule data.
The data from these capsules provides information on the embrittlement behavior of the SL-1 reactor vessel lower shell axial weld seams 3-203A,B and C (Heat 305424).
The data indicates that
,this material embrittles in a manner that is within 1o, of their calculated best fit projections as outlined in 10 CFR 50.61.
Therefore, we proposes to use this surveillance data to determine ART>>g and RT>>g in accordance with 10 CFR 50.61.
III. Method of Calculation of RT>>,
Calculation equations from 10 CFR 50'1 for determining limiting adjusted reference temperatures (RT>>g) are as follows:
Equation 1: Determination of RT>>g RT>>g = RTN~<<) + DRT>>g + Margin RT>>><<> =
Reference temperature
('F) for the unirradiated material or initial RT>>g.
L-96-112 Attachment A
Page A-3 Equation 2: Determination of margin value to apply to RT>>,
Margin = 2 [ (a~
+ a~~ ) ]
Margin
= The quantity (4F) to be added to obtain a
conservative upper bound value of RT>>g o = standard deviation for RT~~U~
'F for measured test
- values, 174F for generic data.
a~ = standard deviation for bRT >>~.
17 F for plate and 28'F for welds.
Equation 3:
Determination of transition temperature shift in RT>>g due to irradiation.
(CF) f
{o.28 - 0.10 log f)
CF = Chemistry factor ('F) is a function of Cu and Ni content of the weld or base material and is determined from tables in 10 CFR 50.61. It can also be calculated from actual surveillance data (Equation 4).
f {'
'g 's referred as the fluence factor (ff) f =
Best estimate peak fluence in units of 10E19 n/cm (E >
1MeV), at the clad base metal interface on the vessel ID surface of the material being evaluated.
Equation 4:
Sum of the squares method for determining CF with surveillance data.
/A~ x ff~
CF g ffg ff<= Fluence factor f "'" " '" "
of the actual surveillance capsule result.
A~=
Measured DRTgpg of the actual surveillance capsule data.
Note:
Were the Cu and Ni values of the surveillance material differ from the vessel material, A< must be adjusted by the ratio of the tabulated CF values.
IV. Criteria to Include Surveillance Data into Determination of Calculation of RT, In order to determine if the surveillance material from either of these programs are deemed credible for integrating into RT>>~
L-96-112 Attachment A Page A-4 estimates, the criteria stated in 10 CFR 50.61 paragraph (c) (2)(i) sections (A) through (E) must be addressed.
The five criteria are as follows:
(A)
The material in the surveillance capsules must be those which are the controlling materials with regard to radiation embrittlement.
(B)
Scatter in the plots of Charpy energy versus temperature for the irradiated and unirradiated conditions must be small enough to permit the determination of the 30 ft-pound temperature unambiguously.
(C)
Where there are two or more sets of surveillance data from one
- reactor, the scatter of ART>>T values must be less than 28'F for welds and 17'F for base metal.
Even if the range in the capsule fluence is large (two or more orders of magnitude),
the scatter may not exceed twice those values.
(D)
The irradiation temperature of the Charpy specimens in the capsule must equal the vessel wall temperature at the cladding/base metal interface within + 25'F.
(E)
The surveillance data for the correlation monitor material in the capsule, if present, must fall within the scatter band of the data base for the material.
In addition, the CE Owners Group (CEOG) produced a report" that provides guidance on the five credibility criteria above and developed additional criteria for using surveillance material from a host reactor of a different design.
The report also compares 146 irradiated surveillance data measurement, representing 27 reactor vessels and 52 capsules from CE and Westinghouse NSSS's and concluded that no significant bias exist between CE and Westinghouse surveillance data for CE fabricated vessels.
This report was submitted to the NRC and had received partial review The criteria above with the additional recommendations of the CEOG report'ill be used to determine the credibility of the BV-1 surveillance weld material for.incorporation into the SL-1 surveillance program and determination of RTpTg.
V.
Incorporation of All Applicable Surveillance Data to Calculate Chemistry Factor.
The St. Lucie Unit 1 reactor vessel surveillance program contains material samples from the lower shell plate C-8-2, heat C-5935-2 and the intermediate to lower shell girth weld seam, heat 90136.
Although these materials are not controlling, incorporation of the
~ surveillance data will improve the accuracy of RTpgs estimates
~
In addition, the Duquesne Light Co. Beaver Valley Unit 1 surveillance program has been identified as having the same weld heat (305424) and flux type/lot (Linde 1092/3889) as the SL-1 controlling lower shell axial seam welds (3-203 A, B, C).
L-96-112 Attachment A
Page A-5 The determination of the applicability of these surveillance materials and the determination of the chemistry factor for each material is provided below:
A)
Plate C-8-2( heat C-5935>>2, Cu = 0.154( Ni = 0.57%, tabulated Chemistry Factor (CF)
= 108.35OF.
The C-8-2 plate irradiation data is from the St Lucie Unit 1 surveillance program and the results are shown in Table 2.
This surveillance material is considered applicable since it is an extension of the vessel shell and was irradiated in the SL-1 reactor vessel under the same conditions (temperature and fluence) as the vessel.
The 30 ft-lb temperature was determined unambiguously.
The scatter of this ART>>~ plate data is within 17 F (la,) of the best fit of the surveillance data
,and the one correlation monitor plate or standard reference material (SRM) from the SL-1 surveillance program is within the scatter band (2@~) for plate material.
Therefore this surveillance material meets the credibility requirements stated in 10 CFR 50.61 section (c)(2)(i) in all respects except that this plate material is not the limiting vessel material. It is proposed that the calculated CF of 79.53'F be used in determinations of RT>>~ and RT>>,.
B)
Plate C-8-1, heat C-5935-1, Cu = 0.154, Ni = 0.564, tabulated Chemistry Factor (CF)
= 107.8 F.
This material is the same heat (C-5935) as the SL-1 surveillance plate.
The extension
(-1) indicates that the heat of plate was sectioned into multiple pieces.
Using the ratio procedure to adjust A< in Eq.
4 by 107.8'F/108.35'F the calculated value of CF for plate C-8-1, heat C-5935-1 is 79.13'F. It is proposed that this calculated CF of 79.13oF be used in determinations of RT>>~ and RT~~.
C)
Plate C-8<<3, heat C-5935-3, Cu = 0.124, Ni = 0.584, tabulated Chemistry Factor (CF)
= 82.6'F.
This material is the same heat (C-5935) as the SL-1 surveillance plate.
The extension
(-3) indicates that the heat of plate was sectioned into multiple pieces.
Using the ratio procedure to adjust A, in Eq.
4 by 82.6 F/108.35'F the calculated value of CF for plate C-8-3, heat C-5935-3 is 60.634F. It is proposed that this calculated CF of 60.63oF be used in determinations of RTND, and RT,.
Weld heat 90136/girth weld 9-203, Cu
= 0. 234, Ni = 0. 114, tabulated Chemistry Factor (CF)
= 109.8'F.
The 90136 weld irradiation data is from the St Lucie Unit. 1 surveillance program and the results are shown in Table 2.
L-96-112 Attachment A
Page A-6 This surveillance material is considered applicable since it is identical to the vessel girth weld (heat 90136/code 9-203),
and was irradiated in the SL-1 reactor vessel under the same conditions (temperature and fluence) as the vessel.
The 30 ft-lb temperature was determined unambiguously.
The scatter of this QRTgpg weld data is within 28 F (1a,) of the best fit of the surveillance data and the one correlation monitor plate or standard reference material (SRM) from the SL-1 surveillance program is within the scatter band (2o,) for plate material.
Therefore, this surveillance material meets the credibility requirements stated in 10 CFR 50.61 section (c)(2)(i) in all respects except that this weld material is not the limiting vessel material.
Zt is proposed that the calculated CF of 84.354F be used in determinations of RT~, and RTpgg ~
Weld heat 305424/welds 3>>203A, B
8 C, flux type/lot Linde 1092/3889 (controlling material for embrittlement),
Cu 0.284, Ni = 0.634, tabulated Chemistry Factor (CF)
= 191.654F.
This weld heat 305424 irradiation data is from the Beaver Valley Unit 1 surveillance program and the results of the 3
capsules are shown in Table 2.
This weld material is considered applicable to the SL-1 since it was made by the same submerged arc weld process, by the same fabricator (CE),
with the same heat of weld wire and flux lot as the SL-1 controlling lower shell axial welds (3-203 A, B
& C).
The 30 ft-lb temperature was determined unambiguously.
The scatter of this DRTgpg weld data is within 28'F (1cr~) of the best fit of the surveillance data and the calculated CF is 191.33 F.
The BV-1 surveillance program does not have correlation monitor plate, however the CEOG report" recommends the host vessel (BV-1) surveillance plate be compared to be within the scatter band (2a,) for plate material.
The BV-1 surveillance plate material is shown in Table 2 and meets this 344F (2a,)
test for credible surveillance program data.
The CEOG report'ecommends that, the fast neutron fluence be compared between the BV-1 surveillance capsules and the SL-1 capsule and vessel.
The CEOG report'oncluded that differences within a factor of 10 will result in comparable irradiation behavior.
Table 3 shows the fast neutron fluence of the BV-1 capsules to be within the a factor of 10 of the fluence received by SL-1 capsules and the limiting SL-1 weld that the capsule material represents.
The irradiation behavior of the BV-1 surveillance specimen are therefore comparable to the SL-1 limiting weld material.
The CEOG report'ecommends irradiation temperatures of the surveillance specimens be compared by evaluating the
0
L-96-112 Attachment A
Page A-7 temperature monitors inside the surveillance, capsule and the vessel cold leg temperatures to meet the + 25 F criteria.
The BV-1 capsules are mounted on the thermal shield.
The SL-1 capsules are mounted on the vessel wall.
Both capsules are subjected to the inlet temperature water during operations.
The BV-1 surveillance capsules have thermal monitor melt wires with temperatures of 579OF and
- 590oF, none of which were reported melted in the 3 surveillance capsule reports ".
The SL-1 surveillance capsules also have thermal monitor melt wires, with temperatures of 536'F, 558'F, 580'F and 590'F.
The two capsules that have been removed and tested in the SL-1 surveillance program had melted 536'F monitors and deformed 5584F monitors indicating temperatures had exceeded 5364F and were below 5584F.
The BV-1 design nominal inlet temperature was listed in Table 3.1-1 of the Beaver Valley Power Station Unit 1 Updated FSAR as 542.5'F.
This nominal temperature has been unchanged since Revision 0 (1/82) through Revision 8 (1/90) which covers the period through the last BV-1 capsule removal date.
Review of the inlet temperature data from available BV-1 control room logs" from cycles 4-7 indicate that the inlet temperature at 954
- power, ranged from 541.5'F to 547.8'F (544.6'F mathematical average) with the majority of the data clustered around 544'F and 545'F.
By comparison, the St. Lucie Unit 1 design nominal cold leg (inlet) temperature was listed in Table 5.1-1 in the St. Lucie Updated FSAR as 548.4 F from initial issue through the current Amendment 14.
- However, Technical Specification DNB Margin Limits for cold leg temperatures were set at
< 542'F from start up (Amendment
- 2) through cycle 4
(Amendment
- 27) which consisted of approximately 32.8 khrs of operation.
The St.
Lucie Unit 2 Updated FSAR compares several plants design data and lists the SL-1 'cycle 1
nominal inlet temperature as 538.94F.
From cycle 5 to present (cycle 13) the SL-1 Technical Specification DNB Margin Limits for cold leg temperatures (Amendment 48 and 130) have been 549'F.
Examination of available control room cold leg temperature logs from cycle 8 to through the current operating cycle'13, indicates the nominal inlet temperature to be 548.8 F with individual channel readings ranging between 547.9OF and 549.5'F.
This nominal inlet temperature should also be comparable for cycles 5-'7.
Assuming a cold leg temperature of 538.9'F for 32.8 khrs (Cycle 1-4) and 548.8 F for the remaining 85.6 khrs (Cycle 5-12) the time weighted average SL-1 nominal cold leg inlet temperature is 546.1'F which. is almost identical,to the 544.6'F BV-1 cold leg data and should not require any temperature correction for use.
The inlet temperature comparison between the BV-1 surveillance capsule specimens and the SL-1 reactor vessel therefore meets the
L-96-112 Attachment A
Page A-8 credibility criteria for
+
25'F irradiation temperature requirement stated in 10 CFR 50.61.
The BV-1 surveillance capsule specimen chemistry is a single measurement from the surveillance capsule program'f
.26%
Cu and'624 Ni which correspond to a CF determined from the table of 183.24F.
The average chemistry value for the same weld in the SL-1 and BV-1 vessel is reported as
.284 Cu and
.634 Ni with a CF of 191.654F.
Using a ratio of CF (191.654F/183.24F) and Eq.
4 above the calculated CF applicable to the SL-1 reactor vessel lower shell axial welds (3-203 A, B
& C) is 200.15OF.
The BV-1 weld surveillance material meets the credibility requirements stated in 10 CFR 50.61 section (c)(2)(i) in all respects and it is the controlling material for the SL-1 reactor vessel.
It is proposed that the calculated CF of 200.15oF be used in.determinations of RTND, and RT>>,.
VI.
Fluence Projections.
The cumulative vessel ma ximum fluence at the end of license (EOL) of March 1, 2016, is 3.415 E19 neutrons/cm'E
>1.0 MeV) and is applicable to all the beltline plates and the girth weld (9-203).
The azimuthally adjusted maximum EOL fluence for the axial weld is 2.272 E19 neutrons/cm'E
>1.0 MeV).
These best estimate fluence projections are based on accumulated fluence to date and projections based on current core loading patterns and a capacity factor of 89.74 from the approved, operating schedule.
This capacity factor should result in a
conservative EOL fluence projection since it is higher than the average capacity factor of the past cycles.
VII. EOL Calculation of RT>>,.
The EOL RT>>~ values for the St.
Lucie Unit 1 reactor vessel beltline materials are shown in Table 4.
The calculation were determined using Eq.'
through 4 above.
Actual surveillance data is used in the determination of chemistry factor and RT>>g for: the limiting axial welds 3-203 A, B and C; the girth weld, 9-203;.and the lower shell plates C-8-1, -2 and -3.
The margin term for these materials with credible surveillance data is adjusted by a factor of 2 as specified in 10 CFR 50.61.
Table 4
shows that the St.
Lucie Unit 1 reactor vessel limiting beltline material, axial welds, 3-203 A, B and C
EOL RT>>g is 213 F
which is 57'F below the 270OF screening limit. All other materials are more than 100'F below their respective screening limits.
L-96-112 Attachment A
Page A-9 VIII' REFERENCES 1 ~
2 ~
Title 10 Code of Federal Regulations, Section 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events",
December 19,
.1995, Federal Register Vol. 60, No. 243.
FPL Letter, L-86-20, "St. Lucie Unit 1 Docket No. 50-335, 10 CFR 50.61 (b)
(1) Report",
C.
O. Woody to NRC, January 23, 1986.
3 ~
FPL Letter, L-94-169, "St. Lucie Units 1 and 2, Docket No.
50-335 and 50-389, Generic Letter 92-01 Revision 1
Response
to Request for Additional Information",
D. A.
Sager to NRC, July 1, 1994 4 ~
5.
6.
7 ~
"Florida Power
& Light Co.
St.
Lucie Unit 1
Post Irradiation Evaluation of Reactor Vessel Surveillance
,Capsule W-97",
Combustion Engineering, Inc.,
December
- 1983, TR-F-MCM-004 "Analysis of the Capsule at 104'rom the FPL St. Lucie Unit 1 Reactor Vessel Radiation Surveillance Program",
Westinghouse Electric Corp.,
November 1990, MCAP-12751.
"Duquesne Light Co. Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program",
Westinghouse Electric Corp.,
October
- 1974, MCAP-8457.
"Analysis of Capsule V from Duquesne Light Co.
Beaver Valley Unit 1
Reactor Vessel Radiation Surveillance Program",
Westinghouse Electric Corp.,
January
- 1981, HCAP-9860.
8 ~
"Analysis of Capsule U from Duquesne Light Co.
Beaver Valley Unit 1
Reactor Vessel Radiation Surveillance Program",
Westinghouse Electric Corp.,
September
- 1985,
+CAP-10867.
9.
10.
"Analysis of Capsule W from Duquesne Light Co.
Beaver Valley Unit 1
Reactor Vessel Radiation Surveillance Program",
Westinghouse Electric Corp.,
November
- 1988, JgCAP-12005.
"Application of Reactor Vessel Surveillance Data for Embrittlement Management",
Combustion Engineering Owners Group, July 1993, CEN-405-P Revision 2.
a
L-96-112 Attachment A
Page A-10 CEOG
- Letter, CEOG-93-252,
'!Submittal of CEN-405-Pg Revision 2, Application of Reactor Vessel Surveillance Data for Embrittlement Management",
R. F. Burski to NRC, August 6, 1993.
Duquesne Light
- Letter, ND1DMS:0370, "Operating Temperature Data for Beaver Valley Unit 1", R. A. Hruby Jr. to R.
S.
- Boggs, FPL, April 4, 1996.
L-96-112 Attachment A
Page A-11 TABLE 1:
St. Lucie Unit 1 Reactor Vessel Beltline Material Initial Properties MATERIAL LOCATION
'(CODE NO.)
Intermediate Shell Plate (C-7-1)
Intermediate Shell Plate (C-7-2)
Intermediate Shell Plate (C-7-3)
Lower Shell Plate (C-8-1)
Lower Shell Plate (C-8-2)
Lower Shell Plate (C-8-3)
Intermediate Shell Axial Seam Welds(2-203 A, B, C)
Lower Shell Axial Seam (3-203 A,B,C)**
Intermediate to Lower Shell Girth Seam (9-203)
- HEAT NO A4567-1 B9427-1 A4567-2 C5935-1 C5935-2 C5935-3 A8746/
34B009 305424 90136 FLUX TYPE/LOT NA NA NA NA NA NA Linde 124/
3878&3688 Linde 1092/3889 Linde 0091/3999 0.11 0.11 0.11 0.15 0.15 0.12 0.19 0.28 0.23
>o Ni 0.64 0.64 0.58 0.56 0.57 0.58 0.10 0.63 0.11 INITIAL RTNDT (oF)
-10
+10
+20
+20
-56 (Generic)
-60
-60 Monitored in the St. Lucie Unit 1 reactor vessel surveillance program.
- Monitored in the Duquesne Light Company Beaver Valley Unit 1
reactor vessel surveillance program.
L-96-112 Attachment A
Page A-12 Table 2:
Surveillance Data for St.Lucie Unit 1 Beltline Material Plant/
Capsule Matl Measured Capsule
/Code
~T Fluence (n/cm )
Calculated Best Fit CF (Eq. 4)
Best fit Measured CF x ff AT~
(Eg 3) minus Best fit SL-1
/97'L-1
/104'eld/
74'F 90136 Weld/
73'F 90136 5.50 E18 84.35'F 7.16 E18 58.5'F 69.3'F 15.5 F
3.7 F
SL-1
/97'L-1
/97'-8-2 70'F Trans.
C-8-2 68'F Long.
5.50 E18 79.53'F 5.50 E18 55.2'F 55.2'F 14.8 F
12.8 F
SL-1
/104'-8-2 67'F Long.
7.16 E18 65.3'F 1.7'F SL-1
/104 SRM-HSST-01MY 110'F 7.16 E18 136.1 F
(Table
.18 Cu,.66 Ni) 123.3'F
-13.3'F BV-1 /V Weld/
150'F 305424 2.91 E18 191.33'F 126.8 F
23.2'F BV-1 /U BV-1 /W Weld/
155'F 305424 Weld/
185'F 305424 6.54 E18 9.49 E18 168.6'F
-13.6 F
188.5 F
3.5 F
BV-1 /V BV-1 /V BV-1 /U BV-1 /U BV-1 /W BV-1 /W plate 130'F
/long plate 140'F
/trans plate 120'F
/long plate 135 F
/trans plate 150'F
/long plate 185'F
/trans 2.91 E18 167 '
F 2.91 E18 6.54 E18 6.54 E18 9.49 E18 9.49 E18 111.2'F 18.8 F
111.2'F 28.8 F
147.8 F
-27.8 F
147.8'F
-12.8 F
165.4'F
-15.4 F
165.4 F
19.6 F
SL = St. Lucre BV = Beaver Valley Trans.
= transverse orientation Long.
= longitudinal orientation ff= fluence factor froh Eq.
3
L-96-112 Attachment A
Page A-13 Table 3:
Beaver Valley Unit 1 Surveillance Capsule and St. Lucie Unit 1 Vessel Fast Neutron Fluence (E> 1 MeV) Comparison Plant Capsule Fluence (f) ~
n/cm~
EFPY EFPS
- Fluence, Z/EFPY
- Fluence, f/EFPS BV-1 BV-1 BV-1 SL-1 SL-1 SL-1 97'04 Vessel limiting welds 2;91 E18 6.54 E18 9.49 E18 5.50 E18 7.16 E18 1.20 E19 1.16 3.59 5.89 4.67 9.515 11.27 3.66 E07 2.51 E18 7.95 E10 1.13 EOS 1.82 E18 5.78 E10 1.86 EOS 1.61 E18 5-.11 Elo 1.47 EOS 1.18 E18 3.73 E10 3.00 EOS 7.52 E17 2.39 E10 3.55 EOS 1.07 E18 3.39 E10 EFPY= Effective full power years EFPS= Effective full power seconds
L-96-112 Attachment A
Page A-14 Table 4!
St. Lucie Unit, 1.Reactor Vessel Beltline Material EOL RT>>> Values.
LOCATION Heat ID P Cu%
Ni%
CALCU'ABLE TED. CF CF INITIAL RTndt EOL PEAK
- FLEUNCE, FLUENCE Odta MARGIN E19 n/cm"2 FACTOR (ff) RTpts EOL RT ts PTS UMIT Lower shell plati (C-8-1 C-5935-1 0.15 0.56 79.1 3 17 'F 3 42E+1 9 1.32 105 'F 142 'F 270 'F Lower shell plate (C-8-2 C-5935-2 D.1 5 D.57 79.53 20 F 17 'F 3 42E+19 1.32 105 'F 142 'F 270 'F Lower shell late ~3 C-5935-3 0.1 2 0.58 60.63 0'F 17 'F 3 42E+19 1.32 80 'F 97 'F 270 'F Int. shell plate C-7-1 A-4567-1 Int. shell plate G-7-2 B-9427-1 Int shell plate C-7-3 A-4567-2 Lower shell axial welds (3 203AB.C 305424 Int. shell axial welds (2-A-8746 &
203A.B.C 34B009 0.11 0.11 0.11 0.28 0.19 0.58 0.63 200.1 5 0.10 74.6 73.&
91.5 0'F
-10 F 10 'F
-60 'F
-56 F 34'F 3 42E+19 1.32 99 F 133 F 34 'F 3 42E+1 9 1.32 99 'F 123 'F 34 'F 3 42E+19 1.32 97 'F 141 'F 28.0 F 2.27E+19 1.22 245 F 213 'F 65.5 'F 2.27E+1 9 1.22 112 F 121 'F 270 'F 270 'F 270 'F 270 'F 270 'F Int to Lower girth welds (9-203) 90136 0.23 0.11 84.35
-60 'F 28 'F 3 42E+19.
1.32 111 'F 79 'F 300 'F
L-96-112 Attachment B
Page B-1 ST.
LUCIE UNIT 2 REACTOR VESSEL BELTLINE MATERIALS END OF LICENSE ASSESSMENT OF RT-PTS I.
Discussion The purpose of this assessment is to show that the St. Lucie Unit 2 projected reference temperature (RT>>,) at end of license (EOL) remains below the PTS screening criteria when using the methodology in the most recent 10 CFR 50.61'.
This submittal provides updated projections of RT>>g for the St.
Lucie Unit 2
reactor vessel beltline materials.
The St. Lucie reactor vessel beltline is plate limited.
The beltline materials are low in copper and nickel, and therefore, are relatively more resistant to embrittlement.
Sufficient surveillance capsule data is not yet available, therefore projections of EOL RT>>g will be made based on initial chemistry values and the projection methodology of 10 CFR 50.61.
The St.
Lucie Unit 2 reactor vessel limiting beltline material,
- plate, M-605-2 has an EOL RT>>~ of 160'F which is 110'F below the 270'F screening limit. All other beltline materials are more than 110'F.below their respective screening limits.
II.
Introduction The initial RT~, chemistry values and previous end of license RT>>,
values have been reported for the St. Lucie Unit 2 reactor vessel beltline materials".
The NRC has incorporated these initial values into its Reactor Vessel Integrity Database (RVID).
The St.
Lucie Unit 2 initial RT~ and chemical composition are summarized in Table 1 for the beltline plate and weld materials.
III. Method of Calculation of RT>>~
Calculation equations from 10 CFR 50.61 for determining limiting adjusted reference temperatures (RT>>g) are as follows:
Equation 1: Determination of RT>>,
RTPTS RTNDT(U) + QRT>>g
+ Margin RTNpg{U) =
Reference temperature
('F) for the unirradiated material or initial RTNpg.
4
L-96-112 Attachment B
Page B-2 Equation 2: Determination of margin value to apply to RT>>,
Margin = 2[(0 '
a
')]'argin
= The quantity
('F) to be added to obtain a
conservative upper bound value of RT>>,
a= standard deviation for RTgpT<g>
~
0 F for measured test
- values, 17'F for generic data.
o0 = standard deviation for ART NpT.
17'F for plate and 28'F for welds
-Equation 3:
Determination of transition temperature shift in RT>>,
due to irradiation.
(CF) f t0.28 - 0.10 log f)
>>S CF = Chemistry factor ('F) is a function of Cu and Ni content of the weld or base material and is determined from tables in 10 CFR 50.61.
f "" '" 'g 's referred as the fluence factor (ff) f =
Best estimate peak fluence in units of 10E19 n/cm'(E 1MeV), at the clad base metal interface on the vessel ID surface of the material being evaluated.
IV.
Fluence Projections.
The cumulative vessel maximum fluence at the end of license (EOL) of April 6, 2023 is 2.76 E19 neutrons/cm'E
>1.0 MeV) and is conservatively applied to all the beltline plates and welds.
The fluence projections are based on current fluence to date and projections based on current core loading patterns.
These best estimate fluence projections are based on accumulated fluence to date and projections based on current core loading patterns and a
capacity factor of 96.0%
from the approved operating schedule.
This capacity factor should result in a conservative EOL fluence projection since it is higher than the average capacity factor, of the past cycles.
V.
EOL Calculation of'T>>s.
The EOL RT>>s values for the St.
Lucie Unit 2 reactor vessel beltline materials are shown in Table 2.
The calculation were determined using Eq.
1 through 3 above.
L-96-112 Attachment B
Page B-3 Table 2
shows that the St.
Lucie Unit 2 reactor vessel limiting beltline material, lower shell plate M-605-2, EOL RT>>~ is 160'F which is 110'F below the 270'F screening limit.
All other materials are in excess of 110'F below their respective screening limits.
VI.
REFERENCES 2.
Title 10 Code of Federal Regulations, Section 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events",
December 19, 1995/
Federal Register Vol. 60, No. 243.
FPL Letter, L-86-25, "St. Lucie Unit 2 Docket No. 50-389, 10 CFR 50.61 (b)
(1) Report",
C.
O. Woody to NRC, January 23, 1986.
FPL Letter, L-94-169, "St. Lucie Units 1 and 2, Docket No.
50-335 and 50-389, Generic Letter 92-01 Revision 1
Response
to Request for Additional Information",
D. A.
Sager to NRC, July 1, 1994
L-96-112 Attachment B
Page B-4 TABLE 1:
St. Lucie Unit 2 Reactor Vessel Beltline Material Initial Properties MATERZAL LOCATZON Ec (CODE NO ~ )
Lower Shell Plate(M-4116-1)
Lower Shell Plate(M-4116-2)
Lower Shell Plate(M-4116-3)
Intermediate Shell Plate (M-605-1)
Intermediate Shell Plate (M-605-2)
Intermediate Shell Plate (M-605-3)
Intermediate Shell Axial Seam Welds (101-124 A,B)
Intermediate Shell Axial Seam Welds (101-124C)
Lower Shell Axial Seam (101-142 A,B,C)
Intermediate to Lower Shell Girth Seam (101-171)
HEAT NO B-8307-2 A-3131-1 A-3131-2 A-8490-2 B-3416-2 A-8490-1 83642 83642
/83637 83637 83637
/3P7317 PLUX TYPE/LOT NA NA NA NA NA Linde 0091
/3536 Linde 0091/
3536&1122 Linde 0091/1122 Linde 124/0951
% Cu 0.06 0.07 0.07 0.11 0.13 0.11 0.04 0.04 0.05 0.07 w Ni 0.57 0.60 0.60 0.61 0.62 0.61 0.07 0.07 0.10 0.08 ZNZTZAL RTNDT (o~)
+20
+20
+20
+30
+10
-80
-50
-50
-70
L-96-112 Attachment B
Page B-5 Table 2:
St. Lucie Unit 2 Reactor Vessel Beltline Material EOL RT~s Values.
MATERIAL LOCATION R
{CODE NO.)
Chend.stxy Factor (CF)
Initial RTNDT Margin EOL Fluence n/cm~
EOL Res PTS Limit Lower Shell Plate(M-4116-1)
Loper Shell Plate(M-4116-2)
Lower Shell Plate(M-4116-3)
Intermediate Shell Plate (M-605-1)
Intermediate Shell Plate (M-605-2)
Intermediate Shell Plate
{M-605-3)
Interm. Shell Axial Welds (101-124 A,B)
Interm. Shell Axial Welds (101-124C)
Lower Shell Axial Welds (101-142 A,B,C)
Inter. to Lower Shell Girth Weld(101-171)
B-8307-2 A-3131-1 A-3131-2 A-8490-2 B-3416-2 A-8490-1 83642 83642
/83637 83637 83637
/3P7317 0.06 0.57 0.07 0.60 0.07 0.60 0.11 0.61 0.13 0.62 0.11 0.61 0.04 0.07 0.05 0.10 0.07 0.08 0.04 0.07 37.0 44.0 74.15 91.5 74.15 30.7 30.7 37.5 41.2
+20 F
+20 F
+20'F
+30'F
+10'F O'
-80 F
-50 F
-50 F
-70 F 34 F
2.76819 47 F
34 F 2.76E19 56'F 34 F
2.76E19 56'F 34 F
2.76E19 94'F 34 F
2.76819 116 F
34 F
2.76819 94'F 56 F
2.76819 39'F 56 F
2.76819 39'F 56 F 2.76819 48 F 56'F 2.76E19 52 F
101 F
110 F
110 F
158 F
160'F 128'F 15'F 45 F
54 F
38'F 270 F
270 F
270'F 270 F
270 F
270'F 270'F 270 F
270 F
300 F
t L-96-112 Attachment C
AFFIDAVITPURSUANT TO 10 CFR 2.790 I, I.C. Rickard, depose and say that I am the Director, Operations Licensing of Combustion Engineering, Inc., duly authorized to make this affidavit, and have reviewed or caused to have reviewed the information which is identified as proprietary and referenced in the paragraph immediately below.
I am submitting this affidavit in conformance with the provisions of 10 CFR 2.790 of the Commission's regulations for withholding this information.
/
The information forwhich proprietary treatment is sought is contained in the following documents:
1)
CEN405-P Revision 2, "Application of Reactor Vessel Surveillance Data for
~
Embrittlement Management," 'July 1993.
Enclosure II to CEOG-93-252, Formal Response to NRC Request for Additional Information on CEN-405-P, "Application of Reactor Vessel Surveillance Data for Embrittlement Management," August 1993.
These documents have been appropriately designated as proprietary.
I have personal knowledge of the criteria and procedures utilized by Combustion Engineering in designating information as a trade secret, privileged or as confidential commercial or financial information, Pursuant to the provisions of paragraph (b) (4) of Section 2.790 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure, included in the above referenced document, should be withheld.
1.
The information sought to be withheld from public disclosure, is owned and has been held in confidence by Combustion Engineering.
It consists of the details concerning the fabrication process, material properties, and surveillance data used to develop an approach to ascertain the embrittlement of reactor vessels.
2.
The information consists of test data or other similar data concerning a process, method or component, the application ofwhich results in substantial competitive advantage to Combustion Engineering.
3.
The information is of a type customarily held in confidence by Combustion Engineering and not customarily disclosed to the public. Combustion Engineering has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence.
The details of the aforementioned system were provided to the Nuclear Regulatory Commission via letter DP-537 from F. M. Stern to Frank Schroeder dated December 2, 1974. This system was applied in determining that the subject document herein is proprietary.
The information is being transmitted to the Commission in confidence under the provisions of 10 CFR 2.790 with the understanding that it is to be received in confidence by the Commission.
The information, to the best of my knowledge and belief, is not available in public sources, and any disclosure to third parties has been made pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence.
6.
Public disclosure of the information is likelyto cause substantial harm to the competitive position of Combustion Engineering because:
a.
A similar product is manufactured and sold by major pressurized water reactor competitors of Combustion Engineering.
b.
Development of this information by Combustion Engineering required hundreds of thousand of dollars and thousands of manhours of effort. A competitor would have to undergo similar expense in generating equivalent information.
In order to acquire such information, a competitor would also require considerable time and inconvenience to develop a similar approach to ascertain the embrittlement of reactor vessels given detailed knowledge of the fabrication process, material properties, and surveillance data.
The information consists of the details concerning the fabrication process, material properties, and surveillance data used to develop an approach to ascertain the embrittlement of reactor vessels, the application ofwhich provides a competitive economic advantage.
The availability of such information to competitors would enable them to modify their product to better compete with Combustion Engineering, take marketing or other actions to improve their product's position or impair the position of Combustion Engineering's product, and avoid developing similar data and analyses in support of their processes, methods or apparatus.
In pricing Combustion Engineering's products and services, significant research, development, engineering, analytical, manufacturing, licensing, quality assurance and other costs and expenses must be included. The ability of Combustion Engineering's competitors to utilize such information without similar expenditure of resources may enable them to sell at prices reflecting significantly lower costs.
Use of the information by competitors in the international marketplace would increase their ability to market nuclear steam supply systems by reducing the costs associated with their technology development.
In addition, disclosure would have an adverse economic impact on Combustion Engineering's potential for obtaining or maintaining foreign licensees.
Further the deponent sayeth not.
I..
ickard Director, Operations Licensing Sworn to before me this 4 "
day of
, 1996 Notary, Public Mycommission expires:
P 3 I/95 E
h'. il
0
~'
I