ML022110339
| ML022110339 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 05/31/2002 |
| From: | Charles Brown, Gresham J Westinghouse |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| AEP:NRC:2349-01, FOIA/PA-2005-0108 WCAP-13517, Rev 1 | |
| Download: ML022110339 (30) | |
Text
Westinghouse Non-Proprietary Class 3 Evaluation of Pressurized Thermal Shock for D. C.
Cook Unit 2 Westinghouse Electric Company LLC
2G/25/2002 14:47 412-374-?-'.
W Eo-LCENS:NG PAGE 02/e4 WESTINGHOUSE NON-FROPRIETARYCLASS 3 WCAP-13517, Revision 1 Evaluation of Pressurized Thermal Shock for D. C. Cook Unit 2 C. Brown Mlay 2002 Prepared by the Westinghouse Electric Company LLC for American Electric Power J. A. Gresham, Manager Engineering & Materials Technology Wcstinghouse Electric Company LLC Energy Systems P.O. Box 355
?ittsburgh, PA 15230-0335
©2002 Westinghouse Electric Company LLC A,1 Rights Reserved
Lu TABLE OF CONTENTS L IS T O F T A B L E S.................................................................................................................................
iv L IS T O F F IG U R E S.................................................................................................................................
v P R E F A C E............................................................................................................................................
v i EXECUTIV E SUM M ARY......................................................................................................
.. vii S
IN T R O D U C T IO N...................................................................................................................
1-1 2
PRESSURIZED THERPVLXL SHOCK RULE...........................................................................
2-1 3
M ETHOD FOR CALCULATION OF RTPTS..........................................................................
3-1 4
VERIFICATION OF PLANT SPECIFIC MATERLAL PROPERTIES......................................
4-1 5
NEUTRON FLUENCE VALUES.............................................................................................
5-1 6
DETERMINATION OF RTpTs VALUES FOR ALL BELTLINE REGION NIATERL-kS......... 6-1 7
C O N C L U S IO N.......................................................................................................................
7 - 1
...............................................................8 -1 8
R E F E R E N C E S....................
APPENDIX A PROJECTED UPPER SHELF ENERGY PTS Evaluation of the D. C. Cook Unit 2 Reactor Vessel
.A-1
.I
IV LIST OF TABLES D. C. Cook Unit 2 Reactor Vessel Beltline Unirradiated Material Properties.................. 4-3 Peak Fluence (E > 1.0 MeV) on the Pressure Vessel Clad/Base Interface for D. C. Cook Unit 2 at 32 (EOL) and 48 (Life Extension) EFPY.....................................
5-1 Interpolation of Chemistry Factors Using Tables I and 2 &f 10 CFR 50.61.......
.6-2 Table I Table 2 Table 3 Table 4 RTp-IS Calculation for D.
at EOL (32 EFPY).......
RTPTS Calculation for D Extension (48 EFPY)....
C. Cook Unit 2 Beltline Region Materials 6-4 C. Cook Unit 2 Beltline Region Materials at Life
.... 6-5 PTS Evaluation of the D. C. Cook Unit 2 Reactor Vessel Calculation of Chemistry Factors using Surveillance Capsule Data Per Regulatory Guide 1.99, Revision 2, Position 2.1.........
..... 6-3 Table 5 Table 6
LIST OF FIGURES Identification and Location of Beltline Region Materials for the D. C. Cook U n it 2 R ea cto r V esse l.......................................................
......................................... 4 -2 PTS Evaluation of the D. C. Cook Unit 2 Reactor Vessel Figure I V
PREFACE This report has been technically reviewed and verified by:
Reviewer:
T. J. Laubham Record of Revision Revision 1 The PTS evaluation m Revision 0 was based on 'best-estimate" fluences. WCXFP-135 15 was revised to update the fluence methodology and to include the "calculated" fluences. Thus, this report was revised to incorporate the "calculated" fluences into the D C. Cook Unit 2 PTS calculations.
PTS Evaluation of the D. C. Cook Unit 2 Reactor Vessel
Vii EXECUTIVE SUMIMNARY The purpose of this report is to determine the RTpTs values for the D C. Cook Unit 2 reactor vessel beltline materials based upon the results of the Surveillance Capsule U evaluation-The conclusion of this report is that all the beltlmne materials m the D. C. Cook Unit 2 reactor vessel have RTpTs values below the screening criteria of 270'F for plates, and 300'F for circumferential welds at EOL (32 EFPY) and EGLE (48 EFPY).
PTS Evaluation of the D. C. Cook Urnit 2 Reactor Vessel
INTRODUCTION A Pressurized Thermal Shock (PTS) Event is an event or transient in pressurized water reactors (PWRs) causing severe overcooling (thermal shock) concurrent with or followed by significant pressure in the reactor vessel. A PTS concern arises if one of these transients acts on the beltline region of a reactor vessel where a reduced fracture resistance exists because of neutron irradiation. Such an event may produce a flaw or cause the propagation of a flaw postulated to exist near the inner wall surface, thereby potentially affecting the integrity of the vessel The purpose of this report is to determine the RTpTS values for the D. C. Cook Unit 2 reactor vessel using the results of the surveillance Capsule U evaluation. Section 2.0 discusses the PTS Rule and its requirements. Section 3.0 provides the methodology for calculating RTPTS. Section 4.0 provides the reactor vessel beltline region matenral properties for the D C. Cook Unit 2 reactor vessel. The neutron fluence values used in this analysis are presented in Section 5.0 and were obtained from Section 6 of WCAP-13515, Rev. 1ý". The results of the RTPTS calculations are presented in Section 6.0. The conclusion and references for the PTS evaluation follow in Sections 7.0 and 8.0, respectively PTS Evaluation of the D. C. Cook Unit 2 Reactor Vessel
2-1 2
PRESSURIZED THERMAL SHOCK RULE The Nuclear Regulatory Commission (NRC) amended its regulations for light-water-cooled nuclear power plants to clarify several items related to the fracture toughness requirements for reactor pressure vessels, including pressurized thermal shock requirements. The latest revision of the PTS Rule., 10 CFR Part 50.6112], was published in the Federal Register on December 19, 1995, with an effective date of January 18, 1996.
This amendment to the PTS Rule makes three changes:
1 The rule incorporates in total, and therefore makes binding by rule, the method for determining the reference temperature, RTbrFT, including treatment of the unirradiated RTNDT value, the margin term, and the explicit defiition of "'credible" surveillance data, which is also described in Regulatory Guide 1.99, Revision 213.
2 The rule is restructured to Improve clarity, with the Requirements section giving only the requirements for the value for the reference temperature for end of license (EOL) fluence, RTPTS.
3 Thermal annealing is identified as a method for rmtigating the effects of neutron irradiation, thereby reducing RTpTs.
The PTS Rule requirements consist of the following:
For each pressurized water nuclear power reactor for which an operating license has been issued, the licensee shall have projected values of RTPTS, accepted bv the NRC. for each reactor vessel beltline material for the EOL fluence of the material.
The assessment of RTPTS must use the calculation procedures given in the PTS Rule, and must specify the bases for the projected value of RTpTs for each beltline material. The report must specify the copper and nickel contents and the fluence values used in the calculation for each beltline material.
This assessment must be updated whenever there is a significant change in projected values of RTprs or upon the request for a change in the expiration date for operation of the facility. Changes to RTpTs values are significant if either the previous value or the current value, or both values, exceed the screening criterion prior to the expiration of the operating license, including any renewal term, if applicable for the plant.
The RT~yrs screening criterion values for the beltline region are.
270'F for plates, forgings and axial weld materials, and 300'F for circumferential weld materials.
PTS Evaluation of the D. C. Cook Urut 2 Reactor Vessel
3-1 3
METHOD FOR CALCULATION OF RTpvs RTpTs must be calculated for each vessel beitline material using a fluence value, f, which is the EOL fluence at the clad/base metal interface for the material. Equation 1 must be used to calculate values of RT,0T for each weld and plate or forging in the reactor vessel beltline RTVD-T = RTvD.(u) + NM + ART,,vr (1)
- Where, RTN-T(OJ, Reference Temperature for a reactor vessel material in the pre-service or unirradiated condition M
=
Margin to be added to account for uncertainties in the values of RTZITŽU,, copper and nickel contents, fluence and calculational procedures. M is evaluated from Equation 2 NI
- o* +oT (2) a~u is the standard deviation for RTNDT(U).
au
=
0°F when RTND(u? is a measured value.
aj 170F when RT>,*TT*J is a generic value.
aY, is the standard deviation for RTNTDT.
For plates and forgings:
GA
=
17'F when surveillance capsule data is not used.
¢a
=
8.5°F when surveillance capsule data is used.
For welds:
CYA
=
28°F when surveillance capsule data is not used.
- T
=
14'F when surveillance capsule data is used.
7A not to exceed one half of,ARTN-DT ARTM-*DT is the mean value of the transition temperature shift, or change in RT,;DT, due to irradiation, and must be calculated using Equation 3.
ARTYDT = (CF)
- f
- s-o :0,ogf; (3)
PTS Evaluation of the D. C. Cook Unit 2 Reactor Vessel
3-2 CF ('F) is the chemistry factor, which is a function of copper and nickel content. CF is determined from Tables I and 2 of the PTS Rule (10 CFR 50.61). Sun eillance data deemed credible must be used to determine a material-specific value of CF. A material-specific value of CF, when using credible surveillance data, is determined using Equation 5.
The EOL Fluence (f) is the calculated neutron fluence, in units of 1019 n/cm' (E > 1 0 MeV), at the clad base-metal interface on the inside surface of the vessel at the location where the matenal in question receives the highest fluence. The EOL fluence is used in calculating RTpTs.
Equation 4 must be used for determining RTP-Ts using Equation 3 with EOL fluence values for determining RTPTS RTPTS = RT\\TT(U) + M + ARTTS (4)
To verifý that RT,>DT for each vessel beltline material is a bounding value for the specific reactor vessel, licensees shall consider plant-specific information that could affect the level of embrittlement. This information includes but is not limited to the reactor vessel operating temperature and any related surveillance program results. Results from the plant-specific surveillance program must be integrated into the RT,,DT estimate if the plant-specific surveillance data has been deemed credible.
A material-specific value of CF for surveillance materials is determined from Equation 5.
CF=
[A
- fi(O 0 101"P]
(5)
Y If," f1 5 -0 20 log fi5)
In Equation 5, "A," is the measured value of,XRT,,Tr and "f," is the fluence for each surveillance data point.
If there is clear evidence that the copper and nickel content of the surveillance weld differs from the vessel weld, i.e, differs from the average for the weld wire heat number associated with the vessel weld and the surveillance weld, the measured values of RT>9DT must be adjusted for differences in copper and nickel content by multiplying them by the ratio of the chemistry factor for the vessel material to that for the surveillance weld Irradiation temperature and fluence (or fluence factor) are first order environmental variables in assessing irradiation damage. To account for differences in temperature between surveillance specimens and vessel, an adjustment to the data must be performed. Studies have shown that for temperatures near 550F., a I°F decrease in irradiation temperature will result in approximately a 1°F increase in ART.iT. For capsules w.ith irradiation temperature of T, p,,i, and a plant with an irradiation temperature of Tp,, an adjustment to normalize ARTpT s......
d to Tplt is made as follows:
Temp. Adjusted A,_MRTprs = ARTPTS*
MeasLed +
1.0*( Tcapsule - Tpiant)
Note that the temperature adjust methodology has been reinforced by the NRC at the NRC Industry Meetings on November 12, 1997 and February 12, 13 of 1998.
PTS Evaluation of the D. C. Cook Unit 2 Reactor Vessel
4-1 4
VERIFICATION OF PLANT SPECIFIC MATERIAL PROPERTIES Before performing the pressurized thermal shock evaluation, a review of the latest plant-specific material properties for the D. C. Cook Unit 2 vessel was performed. The beltline region of a reactor vessel, per the PTS Rule, is defined as, "the region of the reactor vessel (shell material including welds, heat affected zones and plates or forgings) that directly surrounds the effective height of the active core and adjacent regions of the reactor vessel that are predicted to experience sufficient neutron radiation damage to be considered in the selection of the most limiting material with regard to radiation damage". Figure 1 identifies and indicates the location of all beltline region materials for the D. C. Cook Unit 2 reactor vessel, The calculated copper and nickel contents of the beitline materials were obtained from ATI report DIT-B 02230-0014.
The calculated copper and nickel content is also documented in Table 1 herein. The average values were calculated using all of the available material chemistry information. Initial RTNDT values for D. C. Cook Unit 2 reactor vessel beltline material properties are also showh in Table 1.
PTS Evaluation of the D. C. Cook Unit 2 Reactor Vessel
4-2 0"
270" 90" 10*
C5556-2 180 CORE 0
C 5540-2 270" 90" CC 180 Figure 1 Identification and Location of Beltline Region Matenials for the D. C Cook Unit 2 Reactor Vessel PTS Evaluation of the D. C. Cook Unit 2 Reactor Vessel
4-3 Table 1 D. C. Cook Unit 2 Reactor Vessel Beltlne Unirradiated Material Properties Surveillance Weld Notes:
(a)
Copper and nickel content values from Ref 1, Table 4-1 unless otherwise noted.
(b)
The Initial RTNDT values from reference 1, Table 5-1.
(c) Per ATI-01-024-T004 (Ref-5).
(d) Actual value is 0.125 and was conservatively rounded to 0.130. It should also be noted that Inter. Shell Plate 10-2 has credible surveillance data, overriding the weight percent Cu & Ni.
PTS Evaluation of the D. C. Cook Unit 2 Reactor Vessel Material Description Intermediate Shell Plate 10-1 Intermediate Shell Plate 10-2 Lower Shell Plate 9-1 Lower Shell Plate 9-2 Intermediate to Lower Shell Weld Intermediate Longitudinal Weld Lower Longitudinal Weld
5-1 5
NEUTRON FLUENCE VALUES The calculated fast neutron fluence (E > 1.0 MeV) values at the clad base metal interface of the D. C.
Cook Unit 2 reactor vessel for 32 and 48 EFPY are shown in Table 2. These values were projected using the results of the Capsule U analysis. See Section 6 of the revised D C. Cook Unit 2 Capsule U analysis report, WCAP-13515 Rev. II'].
TABLE 2 Peak Fluence (E > 1.0 MeV) on the Pressure Vessel Clad,/Base Interface for D.
at 32 (EOL) and 48 (Life Extension) EFPY C. Cook Unit 2 PTS Evaluation of the D. C. Cook Urnt 2 Reactor Vessel Material Location 32 EFPY Fluence 48 EFPY Fluence Intermediate Shell Plate 10-I 450
- 1. 625 x 10'9 n/cin 2.457 x 10'9 n/cm Intermediate Shell Plate 10-2 45" 1.625 x 1019 n/cm2 2.457 x 10" n./cm2 Lower Shell Plate 9-1 450
- 1. 625 x 10'9 n/cm' 2.457 x 1019 n/cm 2 Lower Shell Plate 9-2 450 1.625 x 10'9 n/cm2 2.457 x 1019 n/cm 2 Beltline Welds 450 1.625 x 1019 n/cm2 2.457 x i0 19 n/cm-
.I
6-1 6
DETERMINATION OF RTPTS VALUES FOR ALL BELTLINE REGION MATERIALS Using the prescribed PTS Rule methodology, RTTS values were generated for all beltlime region materials of the D. C. Cook Unit 2 reactor vessel for fluence values at the EOL (32 EFPY) and life extension (48 EFPY).
Per 10 CFR Part 50.61, each plant shall assess the RTpTs values based on plant-specific surveillance capsule data. The D_ C. Cook Unit 2 surveillance program data has been evaluated and shown to be credible in WCAP-15047 Rev. 2[']. The related surveillance program results have been included in this PTS evaluation.
As presented in Table 3, chemistry factor values for D. C. Cook Unit 2 based on average copper and Mickel weight percent values were calculated using Tables I and 2 from 10 CFR 50.6 12]. Additionally, chemistry factor values based on credible surveillance capsule data are calculated in Table 4 for D. C. Cook. Tables 5 and 6 contain the RTPTS calculations for all beltline region materials at EOL (32 EFPY) and life extension (48 EFPY).
PTS Evaluation of the D. C. Cook Unit 2 Reactor Vessel
6-2 TABLE 3 Interpolation of Chemistry Factors Using Tables I and 2 of 10 CFR Part 5061 Material Cu wt. %
Ni wt. %
Chemistry I
- _Factor,
_F Inter. Shell.Axial Welds 0.056 0.956 76.4 0F Inter. Shell Plate 10-1
- 0. 150 0.570 108.4 0F (C5556-2)
Inter. Shell Plate 10-2
- 0. 130 0.580 90.4 0F (C552 1-2)
Int./Lower Shell Circ. Weld 0.056 0.956 76.4 0F Lower Shell Axial Welds 0.056 0 956 76.4°F Lower Shell Plate 9-1 0.110 0.640 74.60F (C5540-2)
Lower Shell Plate 9-2
- 0. 140 0.590 99.5 0F (5592-1)
Surveillance Weld Metal 0.055 0.970 750F PTS Evaluation of the D. C. Cook Urut 2 Reactor Vessel
6-3 TABLE 4 Calculation of Chemistry Factors using Surveillance Capsule Data Per Regulatory Guide 1.99. Revision 2, Position 2. 1 Material Capsule Capsule F
FFb)
-T RT\\-D-co)
FF*ARTh-DT FF2 L
S T
0.238 0.612 55 33.66 0.375 Intermediate Shell Plate y
0.664 0.385 90 79.65 0 783 C -552 1-2 (Longitudinal)
X 1.019 1005 95 95.48 1.010 U
1.583 1.127 95 107.07 1.270 T
0.238 0612 80 48.96 0 375 Intermediate Shell Plate y
0.664 0,885 100 88.50 0.783 C-5521-2 (Transverse) x 1.019 I 005 103 103.52 1.010 U
1.583 1.127 130("
146.51 1 270 SUM:
703 35
- 6. 876
..F : (FF RTNrDT,)
T-( FF2)= (703.35)+ (6.876) =102.3 0F T
0.238 0.612 40.76 (4 0 )(d) 24.95
- 0. 375 Y
0.664 0 885 50.95 ( 5 0 )"d) 45 09 0783 Surveillance Weld Metal X
1.019 1.005 7 1. 3 3 ( 7 0 )(*d 71.68 1.010 U
1.583 1 127 76.43 (7 5)d) 86.14 1.270 SUM:
227 86 3.438
-CF
=
(FF* RTN)jr+
-'(FF
- 2) - (227.86) + (3.438) = 66.3"F Notes:
(a) f= Calculated fluence from the D. C-Cook Unit 2 capsule U dosimetry analysis results, (x 1019 n/cm2.
E > 1.0 MeV).
(b) FF = fluence factor = fo 2 8- 0 1 --
- o1)
(c) Data obtained from WCA-P-13515 Capsule U Analysis.
(d) The surveillance weld metal ARTý,Dr values have been adjusted by a ratio factor of 1.019. Original ARTN.,DT values are in parenthesis.
(e) WCAP-13517 Rev. 0 originally reported this value as 138. Thi s was an error, Per WCAP-13515 the value is 130 ft-lbs.
PTS Evaluuation of the D. C. Cook Unit 2 Reactor Vessel
RTPS Calculation for D. C.
TABLE 5 Cook Unit 2 Beltline Region Materials at EOL (32 EFPY)
Fluence b
.ARTPTS
Iarg in j RTDTUL' RTTS N E>1.0 1
Me,
(cF)
(F)
(OF)
(F)
('F)
Intermediate Shell Plate 10-1 1.625 1.134 108.4 122.9 34 58 215 Intermediate Shell Plate 10-2 1.625 1.134 90.4 102.5 34 38 175 Intermediate Shell Plate 10-2 --.
1.625 I.134 102.3 1 16.0 17 38 171 using S/C Data Lower Shell Plate 9-I 1.625 1.134 74.6 84.6 34
-2)0 99 Lower Shell Plate 9-2 1.625 1.134 99.5 12.8 34
-20 127 Beldline Welds 1.625 1.134 4 86.6 56
-35 108 Beltline Welds -- using S/C Data 1.625 1.134 66.3 75.2 28
-35 68 Notes:
(a) Initial RTNIZT values are measured values (b) RTpTS = RT,,DTlL,
+ ARTIrs + Margin jiF)
(c).-\\RTrs = CF - FF PTS Evaluation of the D. C. Cook Unit 2 Reactor Vessel 6-4
6-5 TABLE 6 RTI-TS Calculation for D. C. Cook Unit 2 Beltline Region Materials at Life Extension (48 EFPY)
SFluencC F
AýRTpTS Margin RTNEM u RTs' Maerial n/cm, FF (F)
(F)
F F
E> I0 MONe'V)
'F (F)
(F(F)
I)
Intermediate Shell Plate 10-I 2.457 1084 1346 34 Intermediate Shell Plate 10-2 1-2 90.4 112.3 34 38 184 Intermediate Shell Plate 10-2 21 2.457 1.242 102 3 127.1 17 38 182 uIsin SiC Data Lower Shell Plate 9-1 2.457 1242 792.7 34
-20 107 2..
99.5 12.63 138 Lower Shell Plate 9-2 2.457 242 9123.6 34
-20 Belimne Welds 2.4 5 I7.242 76.4 94.9 56
-35 116 Belilne Welds ---) using S/C Data 2.457 1242 66.3 82.38
-35 75 Noiesý (a) Intial RTIT values are measured values (bI RT1 TS = RTNI)TI* +.ART1,T5 + Margin (F)
(c)
,\\RTI-T,
= CF ý FF IPTS Evaluation of the D. C. Cook Unit 2 Reactor Vessel
'V 7-1 7
CONCLUSIONS As shown in Tables 5 and 6, All beltline matenrals in the D. C. Cook Unit 2 reactor vessel remain below the PTS screening criteria through the end of current license life (32 EFPY) and license renewal (48 EFPY)
PTS Evaluation of the D. C. Cook Unit 2 Reactor Vessel
8-1 8
REFERENCES I
WCAP-135 15, Revision 1. "Analysis of Capsule U from the Indiana Michigan Power Company D.C. Cook Unit 2 Reactor Vessel Radiation Surveillance Program", T.J. Laubham, et. al, dated December 2001.
10 CFR Part 50.61., "Fracture Toughness Requirements For Protection Against Pressurized Thermal Shock Events", Federal Register, Volume 60, No. 243, dated December 19, 1995.
3 Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials", U S.
Nuclear Regulatory Commission, May 1988 4
AEP Design Information Transmittal (DIT), DIT-B-02230-00, "Ma'terial Chemiustry of the Reactor Vessel Belt-line Materials for Cook Nuclear Plant Units I & 2", T.Satyan-Sharma, 10/23/01.
5 WCAP-15047, Revision 2, D. C. Cook Unit 2 WOG Reactor Vessel 60-Year Evaluation Minigroup Heatup and Cooldovn Lirmt Curves for Normal Operation," C. Brown, et al., December 2001 PTS Evaluation of the D. C. Cook Umt 2 Reactor Vessel
.I A-1 APPENDIX A PROJECTED UPPER SHELF ENERGY PTS Evaluation of the D. C. Cook Unit 2 Reactor Vessel
A-2 TABLE A-I Predicted End-of-License (32 EFPY) USE Calculations for all the Beltline Region Materials Weight % of 1/4T EOL Unirradiated Projected USE Projected Material Fluence USE(a)
Decrease EOL USE Cu (1019 n/cm2)
(ft-lb)
(%)
(ft-lb)
Intermediate Shell Plate 10-1 0.15
.975 90 24 68 Intermediate Shell Plate 10-2 0.13
.975 86 22 67 Lower Shell Plate 9-1
- 0. 11
.975 110 19 89 Lower Shell Plate 9-2 0.14
.975
.103 23 79 Beltlmne Welds 0.056
.975 77 10 69 Notesý (a) Matches RVID2.
(b) Determined using Figure 2 of Reg. Guide 1.99, Rev. 2 with the % Cu and 1/4T Fluence.
Hence, all beltline material USE values remain above 50 ft-lb, through EOL (32 EFPY).
PTS Evaluation of the D. C. Cook Umit 2 Reactor Vessel
A-3 TABLE A-2 Predicted End-of-License (48 EFPY) USE Calculations for all the Beltline Region Materials 1/4T EOL Unirradiated Projected USE Projected Material Weight % of Fluence USE (a)
Decrease EOL USE Cu (1019 n/cm2)
(ft-lb)
(%)
(ft-lb)
Intermediate Shell Plate 10-1 0.15 1.475 90 26 67 Intermediate Shell Plate 10-2 0-13 1.475 86 23 66 Lower Shell Plate 9-1 0.11 1.475 110 22 86 Lower Shell Plate 9-2 0.14 1.475
.103 25 77 BeltlMe Welds 0.056 1.475 77 12 68 Notes:
(a) Matches RVID2.
(b) Determined using Figure 2 of Reg. Guide 1.99, Rev. 2 with the % Cu and 1/4T Fluence.
Hence, all beltline material USE values remain above 50 ft-lb. through EOL (48 EFPY).
PTS Evaluation of the D. C. Cook Unit 2 Reactor Vessel
COPYRIGHT NOTICE The reports transmitted herewith each bear a Westinghouse copyright notice. The NRC is permitted to make the number of copies of the information contained in these reports which are necessary for its internal use in connection with generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.790 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding. With respect to the non-proprietary versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.
to AEP:NRC:2349-01 Request for Exemption from Requirements in 10 CFR 50.60(a) and 10 CFR 50, Appendix G
Background
Regulation 10 CFR 50.60(a) requires that operational nuclear reactors meet the fracture toughness requirements for the reactor coolant pressure boundary set forth in 10 CFR 50, Appendix G.
Regulation 10 CFR 50, Appendix G, Section IV.2.a states that appropriate requirements on both the pressure-temperature limits and the minimum permissible temperature must be met for all conditions. Further, 10 CFR 50, Appendix G, Section IV.2.b, and the associated table require that the limits be at least as conservative as limits obtained by following the methods of analysis and the safety margins of Appendix G to Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code.
Requested Exemption Indiana Michigan Power Company (I&M) requests an exemption from the requirements of 10 CFR 50.60(a), 10 CFR 50, Appendix G, Section IV.2.b, and the associated table, for Donald C.
Cook Nuclear Plant (CNP), Unit 2. I&M requests approval to use ASME Code Case N-641, "Alternative Pressure/Temperature Relationship and Low Temperature Overpressure Protection System Requirements,Section XI, Division I, approved January 17, 2000," in lieu of these requirements. ASME Code Case N-641 presents alternative procedures for calculating pressure temperature relationships and for calculating low temperature over pressure protection (LTOP) system enable temperatures and allowable pressures.
Application of Code Case N-641 The revised reactor coolant system (RCS) pressure-temperature curves proposed for inclusion in the CNP Unit 2 Technical Specifications, and the LTOP system enable temperatures have been developed in accordance with ASME Code Case N-641. ASME Code Case N-641 allows use of the lower bound K[c fracture toughness curve in lieu of the lower bound KIA fracture toughness curve.
10 CFR 50.60 and 10 CFR 50.12 Requirements Regulation 10 CFR 50.60(b) states that proposed alternatives to the requirements in 10 CFR 50, Appendix G, or portions thereof, may be used when an exemption is granted by the Nuclear Regulatory Commission (NRC) under 10 CFR 50.12. Regulation 10 CFR 50.12 states that the NRC may grant an exemption from requirements contained in 10 CFR 50 if certain criteria are met. These criteria are addressed below.
Page I to AEP:NRC:2349-01 The requested exemption is authorized by law: No law exists which precludes the activities covered by this exemption request. 10 CFR 50.60(b) allows the use of alternatives to 10 CFR 50, Appendix G when an exemption is granted by the Commission under 10 CFR 50.12.
As described below the other criteria of 10 CFR 50.12 have been met.
The requested exemption does not present an undue risk to the public health and safety: Use of the KIc curve as the basis fracture toughness curve for the development of RCS pressure temperature limits, LTOP system pressure setpoints, and LTOP system enable temperatures is more accurate technically than use of the KRA curve. The KIc curve appropriately implements a relationship based on static initiation fracture toughness behavior to evaluate the controlled heatup and cooldown process of a reactor pressure vessel (RPV), whereas the KIA curve was developed from more conservative crack arrest and dynamic fracture toughness test data.
The application of the KR, fracture toughness curve was initially codified in Appendix G to Section XI of the ASME Code in 1974, to provide a conservative representation of RPV material fracture toughness. As documented in an NRC letter approving a similar exemption for Arkansas Nuclear One, Unit No. 2 (see "Precedent Licensing Actions" below), this initial conservatism was necessary due to the limited knowledge of RPV material behavior at that time. The letter also documents that, since 1974, the level of knowledge about the fracture mechanics behavior of RCS materials has been greatly expanded, especially regarding the effects of radiation embrittlement and the understanding of fracture toughness properties under static and dynamic loading conditions.
As stated in the letter, this additional knowledge has demonstrated that the lower bound on fracture toughness provided by the KIA fracture toughness curve is beyond the margin of safety required to protect the public health and safety from potential RPV failure.
Additionally, use of pressure-temperature curves based on the Klc fracture toughness curve may enhance overall unit safety by enlarging the RCS pressure-temperature operating window, with the greatest safety benefit in the region of low temperature operations. The RCS heatup and cooldown operating window is defined by the maximum allowable pressure as determined by brittle fracture considerations, and the minimum required pressure for the reactor coolant pump seals adjusted for instrument uncertainties. A small operating window may have an adverse safety impact by increasing the possibility of inadvertent overpressure protection system actuation due to pressure surges associated with normal unit evolutions such as RCS pump starts and swapping operating charging pumps with the RCS in a water solid condition. By allowing an increased upper pressure limit that still provides adequate brittle fracture protection, application of ASME Code Case N-641 can result in a benefit to safety by precluding unnecessary overpressure protection system actuation.
The requested exemption will not endanger the common defense and security: The common defense and security are not endangered by this exemption request.
Page 2 to AEP:NRC:2349-01 Special circumstances are present which necessitate the request for an exemption: Pursuant to 10 CFR 50.12(a)(2), the NRC will consider granting an exemption to the regulations if special circumstances are present. The regulation lists the conditions that constitute special circumstances.
This requested exemption from requirements in 10 CFR 50.60(a) and 10CFR50, Appendix G, meets the special circumstances described in paragraph 10CFR50.12(a)(2)(ii) which states: "Application of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule."
The underlying purpose of the regulations in 10 CFR 50 Appendix G is to specify fracture toughness requirements for ferritic materials of the reactor coolant pressure boundary in order to provide adequate margins of safety under normal operating conditions, system hydrostatic tests, and during transient conditions to which the system may be subjected over its service lifetime. As described above, application of ASME Code Case N-641 to determine pressure temperature limits and LTOP system enable temperatures provides appropriate procedures to determine limiting maximum postulated defects and consider those defects in establishing the limits and enable temperature.
This application of the code case maintains an adequate margin of safety in the fracture toughness requirements for the reactor coolant pressure boundary as was originally contemplated in the regulations.
Accordingly, use of ASME Code Case N-641, as described above, achieves the underlying purpose of the associated NRC regulations regarding brittle fracture concerns.
Therefore, I&M considers that special circumstances are present as defined in 10 CFR 50.12(a)(2)(ii), in that application of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule.
Precedent Licensing Actions This exemption request is similar to exemption requests approved for Arkansas Nuclear One, Unit 2, North Anna Power Station, Units I and 2, and Turkey Point Units 3 and 4, as documented in the following letters:
Letter from S. P. Sekerak, NRC, to C. G. Anderson, Entergy Operations, Inc., "Arkansas Nuclear One, Unit No. 2 - Issuance of Amendment re: Reactor Vessel Pressure-Temperature Limits And Exemption From the Requirements of 10 CFR Part 50, Section 50.60(a) (TAC Nos. MB3301 AND MB3302)," dated April 15, 2002 Letter from S. R. Monarque, NRC, to D. A. Christian, Virginia Electric and Power Company, "North Anna Power Station, Units 1 and 2 - Issuance of Amendments and Exemption From the Requirements of 10 CFR Part 50, Section 50.60(a) re: Amended Pressure-Temperature Limits (TAC Nos. MA9343, MA9344, MA9347, and MA9348)," dated May 2, 2001 Page 3 to AEP:NRC:2349-01 Page 4 Letter from K. N. Jabbour, NRC, to T. F. Plunkett, Florida Power and Light Company, "Turkey Point Units 3 and 4 - Exemption from the Requirements of 10 CFR Part 50, Section 50.60 and Appendix G (TAC Nos. MA9504 and MA9505)," dated October 24, 2002 The acceptability of Code Case N-641 is also recognized in the proposed Revision 13 of R.G.
1.147 (Draft RG 1091, Reference 6 in Enclosure 2 to this letter).