ML20198J083
| ML20198J083 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 01/08/1998 |
| From: | Stanley H COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20198J090 | List: |
| References | |
| NUDOCS 9801140035 | |
| Download: ML20198J083 (63) | |
Text
Osmnumw calth libwn Osmpany I 400 Opm lige ihm ner, Ormr.11 (drili January 8,1998 U. S Nuclear Regulatory Commission Washington, D. C 20555 Attention: Do:ument Control Desk
Subject:
Supplemental Information Pertaining to Technical Specification Amendment Regarding Pressure Temperature Curves Byron and Braidwood Nuclear Power Stations NRC Docket Numbers: 50-454.53-455. 50-456 and 50-457
References:
- 1. J. Hosmer letter to the Nuclear Regulatory Commission dated May 21,1997, transmitting Technical Specification Amendment Request,
- 2. J. Hosmer letter to the Nuclear Regulatory Commission dated November 18,1997, transmitting Supplement to Technical Specification Amendment Request.
- 3. J. Hosmer letter to the Nuclear Regulatory Commission dated December 3,1997, transmitting WCAP-14824, Rev. 2.
- 4. December 10,1997, December 12,1997, and December 30,1997, Teleconferences between the Commonwealth Edison Company and the Nuclear Regulatory Commission Regarding the Pending Technical Specification Amendment.
Reference I transmitted the Technical Specification Amendment regarding the Pressure Temperature Curves for Braidwood and Byron Units 1 and 2. Subsequently, Reference 2 transmitted a Supplement to the Technical Specification Amendment Request. Reference 3 transmitted WCAP-14824, Rev. 2 " Byron Unit i Heatup and Cooldown Limit Curves for Normal Operation and Surveillance Weld Metal Integration for Byron and Braidwood." During the Referenced Teleconferences and subsequent teleconferences, the Nuclear Regulatory Commission (NRC) questioned this material. In response to those questions the Commonwealth Edison Company (Comed) provides the following:
. Attachment A documents the issues discussed and Comed's response.
. Attachment B contains the errata to WCAPs -14824 Revision 2, -14940 and -14970, specifically:
- Westinghouse Letter CAE-97-233/CCE-97-316, " Comed Transmittal of Updated Tables to WCAP-14824 Rev. 2, and :
Westinghouse Letter CAE-97-232/CCE-97-315, " Comed Transmittal
- Tables to WCAP-14940 and WCAP-14970." I of Upda
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NRC Document Control Desk 2 January 8,1998
- Attachment B also contains Westinghouse Letter CAE 97 231/CCE 97 314, which'
' provides responses to NRC questions regarding the aforementioned WCAPs. ,
- - Attachment C contains the revised Pressure Temperature Limit Reports (PTLR),
which were previously transmitted in Reference 1; e Attachment D contains docum?nts referenced in Attachment A
- ITI letter INS-97-2526, dated June 30,1997, and e FTl letter INS 97-4954, dated December 17,1997.
Comed appreciates the Staf1's prompt review of this information and request issuance of _
the pending Technical Specification amendment by January 16,1998, in rerder to avoid-delay of the restart of Byron Unit 1. Please address any questions that you may have on thir, correspondence to this oflice.
Sincerely, f e , /:
f$
ene tadley PWR Vice Preside [nt Attac* ments cc- Byron /Braidwood Project Manager-NRR Braidwood Senior Resident Inspector-Braidwood Byron Senior Resident inspector-Byron Regional Administrator-Rill Office of Nuclear Safety-IDNS d
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_ _ . _ _ .--_ - __-_ - -__-_-____-_ _ - _ _ _ _ - -]
1 Attachment A
- 1. Ilow was the Low Temperature Over Pressure (LTOP) enable temperature determined? Account for the instrument uncertainty and temperatur e difference between the coolant and ti.e 1/4T location in the vessel wnu.
LTOP Enable temperature was determined using the standard equation from ASME ,
Section XI, Appendix G,1996 Addenda:
RTwr + 50*F + max (AT,noui @ 1/4T)
Westinghouse performed an analysis by calculating the 1/4T AT from the "inside wetted surface" per the 1996 Addenda to ASME Section XI. This was compared to the values calculated from the clad base metal interface and it resulted in a negligible effect. The current Byron /Braidwood (B/B) Technical Specifications (Section 3/4.7.2 and associated bases) have an assumed 10*F conservatism added to account for instrument uncertainty. As documented in the Westinghouse analysis that evaluated the Tosi, values, adding 10*F to the calculated values for T si., results in temperatures 1:ss than 200*F for all four B/B units.
- 2. Section 2.3 of the Pressure Temperature Limit Report (PTLR) identifies the ,
enable temperature as non-Technical Specification, please justify?
The B/B TS require LTOPs to be operable anytime temper .ure is less than 350 F.
The T si, temperatures calculated for all 4 B/B units are $200*F, and therefore the TS requirement of 350*F remains the governing value. On Page 2 cithe Byron Unit 1 PTLR, the "Non-Technical Specification" will be struck from the title of Section 2.3 to make it consistent with the other B/B PTLR attachments.
- 3. Provije a summary of the analysis of both the heat and mass addition to support LTOPs curves.
The baseline analyses for Byron and Braidwood Stations was performed by Westinghouse using the LOFTRAN computer Code, consistent with the WCAP-14040 -NP-A methodology as referenced in our submittal and PTLR. The following summarizes the assumptions:
- The steady-state Appendix G curves at the applicable burnup were generated and the 10% relaxation for Code Case N-514 and/or 1996 Addenda of Section XI, Appendix G was applied, as applicable l
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- The allowable pressure limits were adjusted downward by the appropriate pressure in each temperature range for the pressure difference between the pressure transmitter in the RCS loop piping and the Rx vessel midplane.11n addition, a constant limit of 800 psig across the entire range is imposed to protect the Power Operated Relief Valve
_ (PORV) downstream piping.
Temnerature Range Limitaftons .Ab T > 350*F 4 RCPs,0 Rli pumps (72 psig) 350*F > T > 120*F 4 RCPs,2 Rii pumps (78 psig)
T 1120*F 1 RCP,2 R11 pumps (34 psig)
- PORV stroke times assumption of 2.4 seconds to open and 2 35 seconds to close vias assumed consistent with previous analyses . Solenoid delay of 0.3 seconds to open and clese was also included.
- For the heat injection transients, the setpoint development accounts for a thermal ,
transport of 50*F to account for the assumed maximum primary to secondary temperature differential which was then used to determined the maximum pressure overshoots. The typical ranges for the PORV overshoots is 6 - 67 psig for the mass injection (MI) cases and 11 - 92 psig for the heat injection (Ill) cases
- The MI cases and the til cases were then evaluated as a function of temperature and the most limiting values between the M1 and ill cases at each temperature were determined without instrument uncertainties.
- The maximum allowable PCRV setpoints, without instrument uncertainty, were then transmitted to Comed by Westinghouse.
- Comed performed instrument uncertainty calculations consistent with ISA-S67.04.
Note that the Comed calculations were performed consistent with the 1982 and 1994 versions of the ISA standard. The calculated instrument uncertainties were then applied by Comed to the Westinghouse provided values. This resulted in the LTOP curves that are contained in the Fig. 2.3 and Table 2.2 in the PTLRs. The maximum calculated instrument uncertainty for the PCV-456 channel is +/-112 psig and the uncertainty for PCV-455A is +/-129 psig.: The difTerence i. the uncertainties exists because there are different type transmitters on the channels.
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e For Byron Unit 1 and Braidwood Unit 1, Framatome (FTI) and Comed Nuclear Fuel :
Services performed analyses that verified that the Westinghouse calculated setpoints remained valid assuming replacement steam generators. This analysis was performed using an NRC approved version of RELAP. This is noted as one of our exceptions in Attachment E of our May 21,1997, submittal. (See Reference 1)
- 4. Account for the dynsimic and static nad instrument uncertainties in the analysis (for dynamic head, account for Reactor Coolant Pump (RCP) and Residual Heat Removal (RH) pumps running). Also, please address the valve stroke time assumption and overshoot should be addressed?
See discussion in 3 above.
- 5. Justify the drop in the LTOP urves between 100*F and 120*F?
The drop in the LTOP curve between 100"F and 120*F is due to the allowable pump combinations that are permitted over the givca temperature range where the . Appendix G limit is a constant value. Below 120"F, only 1 RCP + 2 RH pumps is allowed.
Above 120"F, 4 RCPs + 2 RH pumps are allowed. The limitation on dlowable cperating configuration is controlled via operating procedures at the stations. The mass injection case is the more limiting case at the lower temperature range.
Therefore, at temperatures 5120*F, more operating room is acceptable due to the lower AP possible from a Mi transient (i.e. AP = 34 psig for 1 RCP + 2 RH pumps versus AP = 78 psig for 4 RCPs + 2 RH pumps).
- 6. PTLR Section 2.2, and the LTOP PORY maximum lift settings in Figure 2.3 and Table 2.3 account for appropriate instrument errort why does Figure 2.3 identify the limita as nominal?
This is the standard practice for identifying setpoints at Comed. The instrument uncertainty associated with the setpoints is +/- around the values in the figure.
Therefore, the setpoints represent nominal values. As discussed in Section 2.2 of the PTLR documents, the data contained in Figures 2.3 and Table 2.3 account for appropriate instrument error consistent with ISA-S67.04-1982 and 1994. Note that the curves are used by plant operators to control reactor manipulations. From a
- human factors standpoint, it was decided that the format and content of the curves were appropriate.
- 7. Which setting curves were used in the LTOP analysis, the higher curve for PCV-456 or the lower curve for valve PCV-455A?
Neither I TOP setpoint curve was used in the analysis. The LTOP setpoints are based on Appendix G steady state limits, with Code Case N-514, AP efTects, PORY overshoot /undershoot values, and 50*F shift for the HI cases. This combination knla bybwdsgrppinai. doc 3
_ yielded a PORV setpoint curve. Westinghouse provided Comed with this PORV l setpoint curve and no instrument uncertainties were induded. Comed then applied appropriate instrument uncertainties baed on calculations consistent with ISA S67.04-1994. The spread between the curves is due to the fact that there are pressure transmitters of two different mn.nufactured types. This arrangement provides a nominal separation between the PORV setpoints that reduces the probability of cimultaneous PORV operation.-
- 8. Does Comed rely on the Ril suction relief valves for LTOPs? If so, address this from an analysis and a PTLR methodology perspective. WCAP-14040-NP-A j does not address these valves.
Both the current TS (CTS) and the proposed Improved TS (ITS) require Ril suction reliefs in the event pressurizer PORVs are not available. The Rif suction relief setpoints are 450 psig which protects the vessel over the entire range of LTOP operation. As described ir. B/B UFS AR section 5.4.7.2.3, the Rli suction relief valves are sized to provide overpressure protection for two centrifugal charging pumps compared to the design basis for LTOPs which assumes one charging pump in the analysis. The NRC reviewed the acceptability of Low-Temperature Operation in tne Byron SER, Chapter 5 (RCS), Section 5.2.2.2," Low Temperature Operation." The relieving capacity of each saction relief valve is more than adequate to relieve the combined flow of the 2 centrifugal charging pump. In addition, the NRC accepted the administrative control of the accumulator isolation valves and the restrictions on RCP operation via operating procedures. Therefore, based on the significant relieving capacity of the Ril suction relief valves, the substitution of the Rii suction reliefs in the event that one or more pressurizer PORVs are inoperable remains acceptable and ensures that the Appendix G requirements contimte to be met. None of the information in the PTLR changes any of the current licensing basis for the RII system or overpressure protection suction rehef valves as documented in the B/B UFSAR, the CTS, or the proposed ITS.
- 9. Where are the TS restrictions on accumulators in the LTOP region?
There are no TS restrictions on accumulator operation in the LTOP region. As discussed in #8 above, Comed has administrative control on accumdator operation for pressures less than 1000 psig and temperature less than 425'F. The NRC accepted, as described in response to Question 8, this administrative control in the current licensing basis for Byron and Braidwood Stations. None of the information in the PTLR amendment request requires a change to this current licensing basis.
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- 10. - l Analysis was confirme l with RELAP for the replacement S/Gs. Are the D-4 S/Gs bounded by the new analysis?
The process for evaluating the LTOP setpoints is discussed in #3 above. In summary, the LTOP setpeints were developed for both Byron and Braidwood Stations based on RCS volumes and other assumptions associated wi:h the original S/Gs. Subsequent to this analysis, RELAP was used to validate that the setpoints developed for the original S/Gs were conservative for the replacement S/Gs.
Therefore, the LTOP setpoints that are contained in the PTLR are accurate for either the original or replacement S/Gs.
- 11. For Response 1, provide the actual delta-T a:nd instrument uncertainty.
Using the standard equation from the ASME Section XI, Appendix G,1996 Addenda:
RTwr + $0*F + max (ATw @ 1/4T)
The following summarizes the values for the four B/B units:
Plant ART Values 1/4T AT Metal Instrument T,..w.
Uncertainty
- Hyron Unit i 70.0*F 29.254*F 15.0'F 164.3"F Hyron Unit 2 N7.6'F 29.254*F 15.0'F 181.9'F Braidwomi Unit i 76.6*F 29.254*F 15.0*F 170,9'F Hraidwomi Unit 2 66.9'F 29.254*F 15.0'F 161.2*F Instrument uncertainty of 15"F was used here which conservatively bounds the value of 14.3'F which was determined in B/B LTOP setpoint uncertainty calculation.
All the calculated values are less than 200*F which validates the B/B PTLR statement in Section 2.3 for LTOP T w. Temperature.
- 12. Was static head included in bullet 2 for response 3?
The AP values reported in Response 3 include static head correction for the difference between the middle of the reactor core and the pressure transmitter in the RCS loop.
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- 13. For Response 3, bul!ct 6, what plant configurations were considered for 111 and MI? Provide a summary cf the data obtained from analysis?.
As document d in the NRC cpproved methodology for the PTLR, WCAP 14040 NP-A, the design basis transient for determi1ation of LTOP setpoints is a mass injection (MI) transient postulated with the simultaneo.is isolation of the RiiR and letdown systems coupled with full flow from one charging pump. For the heat injection cases, the assumed configuration is a primary to secondary temperature differential of 50 F. In the NRC SER for WCAP-14040-NP-A, this assumed 50*F
<- is considered conservative if the Technical Specificatioris restrict the startup of an RCP when the S/G secondary side temperature is more that S0T higher than the RCS temperature. This requirement is implemented by B/B Stations in TS 3/4.4.1.3.
. For selected temperatures, the maximum allowable PORV setpoints are determined for the mass injection and heat injection transients from a detailed modeling of the PORVs and RCS.
- As discussed previously, the 111 overshoots (calculated by Westinghouse for B/B using LOFTRAN) range from 17 psig, to 92 psig. The following summarizes the 111 overshoots used in the LTOP analysis.
I IIcat injection Temperature (RCS)'F Maximum Overshoot (psig) k 70 17 100 25 120 29 150 37 200 59 250 80 300 92 ,
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- The pump APs which are used in the Mi cases were calculated from l LOITRAN for 11/11 Stations are comprised of the following:
l Trmaciatuts_Bange ._._ Limitations APs '
i T > 350*F 4 RCPs,0 Ril pumps (72 psig) 350*F > T > 120*F 4 RCPs,2 Rif pumps (78 psig)
T 1120*F 1 RCP,2 Ril pumps (34 psig) l The following example summarizes the Maximum Allowable PORV Setpoints for the Mi and 111 cases (without instrument uncertainties applied) for Ilyron Unit 1 12 EFPY.
Summary of Masimum Allowable PORY Setpoints for Hyron Unit 1 12 EFPY nr ; TEMP MIMas til Max 70 620 638 100 620 625 ,
120 576 577 150 576 569 200 772 748 250 772 727 300 772 712 Thabove points for the M1 and lit cases were then platted The til cases were shilled up by 50*F to account for the assumed RCS to S/G temperature differential. The limiting data points between the Mi and ill cases were then determined which provides a setpoint without instrument uncertainties. Note that WCAP.14040-NP.A specifies a constant limit for Ilyton and Braidwood 800 psig, to protect the PORV piping over the entire LTOP temperature range.
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l The following table summarizes the LTOP setpoint values for Ilyron Unit 1:
1 tilOPs PORY Setpoints for Hyron Unit 1 - 12 EFPY ;
Temperature No Unc. Setpoint PCV 455a seapoint PCV-456 setpoint ;
w/ uncertainty w/ uncertainty l i
sn 620 497 514 1 70 620 497 514 100 620 497 514 .
1 120 576 (569) 446 462 150 569 446 162 200 569 446 462 250 727 (712) 587 604 300 712 587 604 350 712 587 604 Note that the "no uncertainty setpoints" above were conservatively adjusted downward at 120T from 576 to 569 and at 250"F from 727 to 712. This was done to allow the PORV curve to have three distinct temperature ranges that are flat. Then the station applicable instrument uncertainties over tiie appropriate temperature range were applied to determine the PCV-455A and PCV-456 setpoint values listed above.
The same methodology and analysis was used for the other B/I3 units to develop their respective setpoint curves Subsequent to the Westinghouse analysis to generate applicable setpoint curves, FTl performed analyses using the NRC approved code RELAP5M3 to verify that the LTOP setpoints calculated by Westinghouse for the original S/Gs remain valid for the replacement S/Gs. The only cases that were reanalyzed by FTl were the til cases because the ill cases are the only ones impacted by the replacement S/Gs. This is due to the difTerent heat transfer areas and rates for the replacement S/Gs compared to the original S/Gs. The Mi cases are not impacted because LTOP Mi cases are assumed to occur at water solid conditior.s in the RCS as documented in WCAP 14040-NP.A.
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- 14. For Response 8, will the I. TOP function of the RilR suction relief valves he evaluated each time the PT limit curves are changed?
Comed has added a statement to Section 2.2 of the PTLRs to address the LTOP function of the RilR suction relief valves and a confirmation that the LTOP rnalysis is evaluated with changes in subsequent PTLRs.
- 15. In Comed response to Question 8, Comed rei<t red to the UFSAR discussion that addresses the RilR system pressure limits and the PT Curve Limits.
Art they the same? Are they related? Provide an analysis to address this.
The following provides the technical evaluation of the acceptability of substituting the Ril suction reliefs for the PORVs.
The Ril suction relief valvei, for 11/Il Stations are set at 450 psig +/ 3% setpoint drill with an allowance for pressure accumulation of 10%, as specified in the ASME Iloiler and Pressare Yessel Code, Anicle NC 7000. Therefore, the maximum pressure at which the relief valve would lifl is:
450 psig a 1.03 = 463.5 psig to account for setpoint drift 463.5 psig a 1.10 = 509.9 psig to account for pressure accumulation Therefore, the maximum pressure that the RCS would encounter when depending on an Ril suction relief valve for overpressure protection is 509.9 psig. It was
. verified that this absolute value protects all four 11/11 PTLR steady state P-T curves over the entire range of LTOP operability. The lowest Appendix G limit in the LTOP range of temperatures is 620.3 psig at 60*F. With the Code Case N-
$14/Section XI 1996 Addenda 10% allowance, this corresponds to a limit of 620.3
- 1.1 = 682,3 psig. Furthermore, the Ril suction relief valves are sized to relieve the capacity from 2 charging pumps compared to the design basis for LTOP setpoints, which is I charging pump. Therefore, based on this evaluation , the Ril suction relief valves for all four 11/I1 units will act as acceptable substitutes to the PORVs for low temperature overpressure protection per the proposed TS amendment.
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- 16. The following provides information on the Safety injection Pump Evaluation for low Temperature Overpressure Protection to support T.S. 3/4.5,4:
With reduced RCS inventory conditions during Modes 5 and 6 and pressurizer ,
level 55% and RCS T., 52007, the Byron /Braidwood Technical Specifications, (TS), Section 3/4 5.4.2, require that at least one Safety injection (SI) pump and injection flowpath be available to mitigate a loss of decay heat removal event. l Although the TS and operating procedures do not preclude the potential for both Si pumps to be available, typical operating practices at Byron and Braidwood ,
ensure that only one Si pump is available. Ilowever, to prevent an i overpressurization event, administrative controls are implemented by the sites to ;
ensure that an inadvertent actua: ion of one or both SI pumps does not occur. This administrative control consists of placing the Main Control Room hand switch for j the Si pumps in the Pull to Lock position in addition to the closure of the isolation valves in the Si pump discharge lines (1/2S18802A/B &1/2518835). These administrative controls ensure that a Si pump cannot be inadvertently started by an electronic signal. In addition, it would require a conscious act by plant operators to start a Si pump by physically moving the hand switch for the Si pump from the Pull-to-Loch position to the Run position and the deliberate opening of one or more discharge isolation valves. For the circumstance where both SI pumps are available, it would require at least three independent actions by the plant operators moving the hand switch for each Si pump from the Pull to Lock position to the Run position and opening of the Si pump cold leg discharge isolation valve >
(1/2S18835). Therefore, from an overpressure protection standpoint, the potential for an inadvertent start of both Si pumps is not considered a credible event. In the case of the inadvertent start up of one Si pump, suflicient time exists for plant operators to take steps to evaluate plant conditions and secure the Si pump from injecting into the reactor vessel. Note that Comed's proposed improved Technical Specification (ITS) Amendments for Byron and Braidwood Stations address this specific configuration in the Bases for ITS SR 3.4.12.1,3.4.12.2, and 3.4.12.3.
These surveillar.ce requirements ensure that there are two independent means to prevent an overpressurization event.
The conclusion that the insdvertent start of a single SI pump would not lead to an overpressurization event is supported by an evaluation of the volumetric capacity of the pressurizer compared to the maximum fir,wrate from an SI pump. One SI pump is capable ofinjecting 650 gpm to the re actor vessel under rtmout conditions. The capacity of the pressurizer bet ween 5% to 100% level is >l2,000 gallons of water. Therefore, in the event of en indvertent start of a single Si pump '
with pressurizer level 55%, it would take l',000 gallons /650 gpm = 18.4 minutes before the pressurizer would go water soli! and potentially overpressurize the RCS. This is significantly more than 10 nimutes for assumed operator action time and therefore, it is reasonable to assume that suflicient time exists for plant operators to take actions to mitigate this postulated event.
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In conclusion, the inadve tent start of two Si pumps is considered a non credible event and therefore, is no concern for overpressurintion of the RCS during low ,
temperature operation. Adr itionally, an evaluation of the inadvertent start of one Si pump demonstrates that there is sufficient time to allow plant operators to take actions to mitigate the event before overpressuriration would occur. lote that the requirements in Byron and Hraidwood TS Section 3/4 5 4.2 were subrnitted as part of Comed's rerponse to NRC Generic Letter 8817. " Loss of Decay lient -
Removal." This particular Technical Specification was reviewed and approved for Byron and Braidwood Stations by the NRC via an Safety Evaluation Report dated August 31,1990. In addition, Comed's proposed ITS Amendments for Byron and Braidwood Stations specifically address the administrative control for Si pump availability in the SR Dases.
- 17. Regarding the ART / PTS values for Nonle Shell Forging and Upper Cire Weld, Comed needs to add a statement that says fluences there are consen ative (i.e. these regions are above active fuel and therefore would experience fluences that are less than Ix10" n/cm').
See Westinghouse Letter CAE 97 231/CCE-97 317 dated January 6,1998.
- 18. Explain why Comed did not address the lower shell circumferential welds for Hyron I & 2 and Hraldwood I & 2.
The lower shell circumferential welds at Byron I & 2 and Braidwood 1 & 2 are all approximately 4 feet below the bottom of the active fuel and will not see a fluence 2
higher than 1 x 10 "n/cm . Therefore they are not considered in the evaluations for PTS or pressure temperature limit curves. See Westinghouse Letter CAE 231/CCE-97-317 dated January 6,1998.
- 19. For Hyron Unit I lower nonle belt forging. #123J218, Hyron UFSAR Table 5.3-7 provided an IRTsnr value of +20F. This differs from the value provided in the Comed response to GL 92 01 Supplement 1. Which is correct?
After conservative review of the information in the original Cenified Material Test Report (CMTR), Com Ed has determined that the Byron Unit I nonle shell forging (heat #123J218 ) IRT wnr is +30'F. This is consistent with the value reported in GL 92 01 Supplement 1.
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" For the Neutron Absorber /Henector/ Shield in the core barrel. how thick is i the shield? Ilow thick is the core barrel?
I
- The core barrel is 2.5 inches thick, the shield is 2.7 inches thick.
! 21. For the surveillance capsules which are part of data integration, what were
- the EFPY values?
For Ilyron Unit 1: capsule U - 1.15 EFPY, capsule X 5.64 EFPY.
For Ilyron Unit 2: capsule U 1.15 EFPY, capsule W. 4.634 EFPY. i For Ilraidwood Unit 1: capsule U 1.10 EFPY, capsule X - 4.234 EFPY. ;
For liraidwood Unit 2: capsule U- 1.15 EFPY, capsule X - 4.215 EFPY.
- 22. What was the EFPY when Comed went to the low leakage core?
i All four 11/11 units went to low leakage cores afler the end of their first fuel cycle.
For Ilyron Unit 1: Approximately 1.19 EFPY.
For Ilyton Unit 2: Approximately 1.20 EFPY.
For Iltaldwood Unit 1: Approximately 1.17 EFPY. ,
For llraidwood Unit 2: Approximately 1.19 EFPY. l
- 23. What is the expected EFPY at the upcoming end of cycles?
l For Ilyton Unit 1, End of Cycle 9 (3/99) Projection - 10.3 EFPYs.
For Ilyron Unit 2, End of Cycle 7 (3/98) Projection 9.9 EFPYs.
For liraidwood Unit 1, End of Cycle 7 (9/98) Projection - 8.06 EFPYs.
For Ilraidwood Unit 2, End of Cycle 7 (5/99) Projection . 8.68 EFPYs.
- 24. Where are the references to all capsule reports in the PTLR?
References to the surveillance capsule reports are p. wided in the unit specific PTLR and the WCAP for that unit.
- 25. What are the bases for the heatup rate restriction of to 'F/hr in 2.1.1.c?
The 10 'F/hr temperature rate of change limit for insenice hydrostatic and leak testing operations was established as an operationallimit to ensure these would be isothermal events. This is a reasonable temperature rate of change limit that would produce an insignificant throughwall thermal transient stress and consequently an insignificant Krr.
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- 26. For Hraldwood Unit 1, Table 4.3 footnote (b) is used twice for the 1/4T and 1 3/4T. It should be used only once. Footnote (c)is not clear. Are the curves good for 16 EFPY or 27.9 EFPY7 Note (b) was intended to distinguish bety een the RG 1.99 Rev. 2 position I and por,ition 2 ART values for the lower shell forging and the circumferential weld metal that were considered as input to the pressure temperature curve calculations.
This note was not intended to identify the single ART values used for the heatup and cooldown composite curves This determination is made in the WCAP 14243.
Note (c) was intended to clarify the intent to submit pressure temperature limits and LTOP set points for 16 EFPY The evaluation in WCAP-14824 Rev. 2 Appendix 11, integrating ilraidwood Units 1 and 2 weld metal, determined that this weld metal ART would actually be applicable to 27.9 EFPY, however, the pressure temperature limits and LTOP setpoints are submitted for 16 EFPY. .
- 27. Hegarding the Hyron 2 and Braldwood Units 1 & 2 PTLRs, what is the basis for, and the meaning, of the statement in Section 2.1.2 that " Uncertainties need not be considered since appropriate station operating procedures ensure that the limits contained in the figures and table are not exceeded."
This statement was removed in the revised Ilyron 2 and 11raidwood Units 1 and 2 i PTLRs.
- 28. The following provides information on Hyron Unit I and Unit 2 Limiting Weld Metal Initial RTsm Variation Ref: WCAP-14C24 Revision 2 Byron Unit I and Unit 2 vessel beltline region and surveillance program welds WF-336 and WF-447 were produced using the same heat of weld wire,442002, and different lots of Linde 80 weld flux,8873 and 8064 respectively. The initial RTun values are -30 F for WF.336 and +10 F for WF-447, based on Babcock and Wilcos weld qualification data. These substantially differing values ofinitial RTer are utilized in the respective vessels for adjusted reference temperature calculations for P-T limits and pressurized thermal shock evaluations. Absent other significant differences in weld production, Comed believes that the difTering initial RTun values may be explained on the basis of the different flux lots used and should be applied to the respective vessels.
Ilowever, WF 336 and WF 447 have performed very similarly under irradiation in the Byron i and 2 surveillance programs For example, the shill for both of the first capsules from Byron 1 and 2 is 0 F. For the second capsules removed from 11yron 1 and 2, the measured shills are equal to 30 F and 35 F, respectively, which Lmia b>bwd agrp plnai doc 15
isjudged to be very close. It is irradiation shifi, relative to the measured initial value for the given weld wire / flux combination, that is used in the integration of ,
surveillance data from more than one vessel. With this in mind, this data suppons the integration of the sun eillance program test repons for these welds, even !
though the initial RTm alues v difTer.
29 The following provides information on Byron Units 1 & 2 and Braldwood Units 1 & 2 Data Integra lon The attached Westinghouse Letter Repon PAE-97 231/CCE 97 314 dated January 6,1998, addresses the question of other welds and the nozzle shell course in the 11yron Units 1 and 2 and Ilraidwood Units 1 and 2 reactor vessels which are near the active fuel zone. Utilizing the Regulatory Guide 1.99 Revision 2 Position I chemistry factors for these welds, it is demonstrated that the limiting materials for the vessels do not change.
This result also indicates that, while there is surveillance data from Ilraidwood Units I and 2 that is applicable to nonlimiting welds in the active fuel zone of Ilyron Units 1 and 2, the integration of that data is unnecessary due to the fact that the welds remain nonlimiting using the more conservative Position I chemistry factors. For consistency, Comed will pursue data integration for Ilyron Units I and 2 and Iltaidwood Units 1 and 2 in a future submittal.
This same Westinghouse letter confirms that the Upper Cire Welds and Nozzle Shell Forgings are above the active fuel. Using the fluence values at the top of active fuel is conservative because the 11uence will be lower at elevations above the active fuel. Also, this letter confirms the Lower Cire. Weld is four feet below the bottom of the active fuel and therefore will not be subject to a fluence greater than the 1 x 10" n/cm' threshold for embrittlement.
- 30. The following provides information on Hyron Units I & 2 and Braidwood Units 1 & 2 Heltline Chemistry Data Ref: 1) WCAP-14824 Revision 2
- 2) ITI letter INS-97 2526. June 30,1997 in WCAP-14940, for Ilyron Unit 2 lower shell forging 49D330/49C298 1-1 MK 24 3, Comed has determined that the second set of 0.05% Cu,0.73% Ni values and the 0.05% Cu,0.75% Ni set of values were inappropriately included in the database for that forging, beginning with WCAP-14063. These two sets of values were obtained from ladle analyses found on the original Japan Steel Works test report. Ilowever, since two separate ladle melts of unknown relative volumes were poured to make a single forging ingot, the individual ladle analyses are not representative of the finished product. For this reason, only the check analysis L nta t@wd agrp polnai ex16
)
I which was originally reported in WCAP.10398 should be used. The list of data for this forging provided in 11AW 2261 is correct, and this will be documented in an errata to WCAP 14940. There is no adverse impact on RTm or adjusted
. reference temperature values for this forging. Addition information is provided in errata letter CAE-97 232/CCE 97-315 (see Attachment B).
I in WCAP 14824 Revision 2, Table D-4, the 0.04% Cu,0 67 Ni data set from l WCAP 9807 was indeed included in the analysis. Based on communications with l FTI, this set of values was obtained from an original weld qualification block, i separate from both of the individual Braidwood Unit I and Braidwood Unit 2 weld ,
blocks. For this reason, the 0.04% Cu,0.67 Ni data set was included as the third subset average value on the list near the top of Table B-4. The same value was, perhaps confusingly, reported for completeness on the list of Braidwood Unit I surveillance chemistry results; the asterisk next to those values was intended to indicate that those values were not used in the Braidwood Unit I surveillance material average calculation. Ilowes er, those values were used in the overall best estimate calculation.
- 31. WCAP-14970 Table I, contains a typo on the Cu content from Reference 17.
The typo was corrected in an errata to the WCAP (Westinghouse Letter CAE 232/CCE-97 315)
- 32. Regarding Braldwood Unit i PTLR, Figure 2.1 appears to include instrumen: uncertainty, Furthermore it appears that the noncritical curve is not conservative (by a few degrecs/psig)in the 565 psig and 120 F range, (i.e., the minimum temperature above 20% of the preservice hydrostatic pressure test should be 110*F ( 10+120)). F; ease explain.
Figure 2.1 of the PTLR does not include instrument uncertainty, flowever, the noncritical curve is conservative for pressures less than 1000 psig because of conservative inputs, i c. defmition of Tima.i for different modes, to the program used to calculate this curve. The methodology used to calculate the pressure-temperature limits in this figure is the same as the methodology in WCAP 14040-NP A. The limiting pressure of $65 psig at 60 F is not caused by the flange requirement in 10CFR$0, Appendix G. The vessel flange requirement is 621 psig at 110*F, L nla 13 tmd opp ptinai dx 17
I l
- 33. The following provides information on Hyron 2 and Braidwood 2 Surveillance Capsule Data Credibilityt Byron 2 and Braidwood 2 lower shell forging surveillance capsule data does not meet the credibility criteria. This was determined using the methodology provided in Appendices A and 11 of WCAP 14824 Rev. 2, which meets the requirements of 10CFR$0.61. WCAP 14824 Rev. 2, page 12 provides a detailed explanation of the methodology used to define the ca margin term for RG 1.99 Rev. 2, positions 1.1 and 2.1. The RG 1.99 Rev. 2, position 2.1 defmition is provided for credible surveillance data, however pages A9 and B9 provide a explanation of how the o4 i margin factor was dermed for the base metal surveillance data not meeting the l credibility criteria. As defmed in the 11/12/97 industry meeting with the NRC, the ormargin factor for non-credible rarveillance data was determined by using the RG 1.99 Rev. 2, position 1.1 definition, i c. the position 1.1 osmargin factor was not divided by 2 as permitted in position 2.1. The ART for the non-credible surveillance data was determined in this manner to ensure that the data was not limiting. Table 4.3 in the Byron Unit 2 and Braidwood Unit 2 PTLR documents efers to "Using credible surveillance capsule data" for the lower shell forging material description. This is a misleading ucscription and will be revised to identify this as non-credible capsule surveillance data. It was reported to document that it was not the limiting material. See Westinghouse Letter CAE-97 232/CCE t 315.
- 34. The following provides an explanation of the best estimate chemistry calculation for weld wire heat No. 442002:
Ref: 1) FI'l letter INS-97-2526, June 30,1997
- 2) FTl letter INS-97 4954, December 17,1997
- 3) WCAP-14824 Revision 2, Table 2, with errata (CAE-97 220/CCE-97-309 and CAE-97-233/CCE-97-316)
Comed has reviewed the weld chemistry data provided as a result of an NRC inspection at FTl in May 1997. Originally reported to the Babcock & Wilcox Owners Group, Reactor Vessel Working Group, in Reference 1, the complete underlying database was provided to Comed in Reference 2. Comed has taken into account the additional"round robin" test results reported by FTl for weld wire heat 442002. Comed has determined that the additional chemistry data made available, when accounted for using the " average of averages" approach, hes no impact on the "best estimate" chemistry value and resulting Regulatory Guide 1.99 Revision 2 Table 1 chemistry factor for weld wire heat number 442002 previously reported in WCAP-14824 Revision 2 Table 2. The average of averages approach is used as a weighting process, since the number of measurements from some separately identified sources, such as the Byron Unit 1 and Unit 2 surveillance bla by%d yp pinai av18
I l
blocks, is much greater than from other sources, such as the individual weld ,
qualification blocks. j As confirmed by FTI, the 21 additional round robin data points for wire heat 442002 provided in Reference 2 were obtained from material taken from the same weld block. Since the source material for the 21 data points is the same weld block, a simpic average of all 21 data points was calculated, with the copper for the 2 Oak Ridge National Laboratory Y 12 analyses conservatively set at 0.10%.
That simple average value became the 10* subset average value, identified as "Round Robin WF 336," for the best estimate calculation shown near the top of Reference 3, Table 2. The best estimate chemistry factor remained unchanged at
- 68. This has been documented in an errata to WCAP-14824 Revision 2.
Following is a description of the assumptions and judgements applied to each of the 12 subset average value; listed near the top of Reference 3, Table 2, which are used in the calculation of the best estimate chemistry value for weld wire heat number 442002:
- 1) Ilabcock & Wilcox (Il&W) Weld Qualification,0.024% Cu,0.7% Ni.
Identified in Reference 2 only as " Sample," in the absence ofinformation explicitly linking it to data from other sources, this was presumed to be taken from a separate block of WF.336.
- 2) Il&W Weld Qualification,0.031% Cu,0.46% Ni. Identified in Reference 2 as "Mt. Vernon WQ Lab No.13762,"in the absence ofinformation explicitly linking it to data from other sources, this was presumed to be taken from a separate block of WF 336.
- 3) Il&W Weld Qualification,0.03% Cu,0.72% Ni. Identified in Reference 2 as "Mt. Vernon WQ Lab No. 24923," in the absence ofinformation explicitly linking it to data from other sources, this was presumed to be taken from a separate block of WF-407,
- 4) Il&W Weld Qualification,0.068% Cu,0.48% Ni identified in Reference 2 as "Mt. Vernon WQ Lab No.16115,"in the absence ofinformation explicitly linking it to data from other sources, this was presumed to be taken from a separate block of WF 407.
- 5) 11&W Weld Qualification,0.114% Cu,0.54% Ni. Identified in Reference 2 as "Mt. Vernon WQ Lab No.16348,"in the absence ofinformation explicitly linking it to data from other sources, this was presumed to be taken from a separate block of WF-421.
- 6) Il&W Weld Qualification,0.148% Cu,0.6% Ni. Identified in Reference 2 as "Mt. Vernon WQ Lab No.17543," in the absence ofinformation explicitly linking it to data from other sources, this was presumed to be taken from a separate block of WF-446.
- 7) B&W Weld Qualification,0.053% Cu,0.62% Ni. Identified in Reference 2 as "Mt. Vernon WQ Lab No. 20369," in the absence ofinformation k nla bybwd gp pinsi dx 19
i explicitly linking it to data from other sources, this was presumed to be l
taken from a separate block of WF-447,
- 8) Il&W Weld Qualification,0.059% Cu,0.62% Ni. Identified in Reference 2 as "Mt. Vernon WQ Lab No.17544,"in the absence ofinformation explicitly linking it to data from other sources, this was presumed to be tak n from a separate block of WF-447.
- 9) Il&W Weld Qualification,0.029% Cu,0.65% Ni Identified in Reference 2 l only as " Sample," in the absence ofinformation explicitly linking it to data from other sources, this was presumed to be taken from a separate block of WF 336.
- 10) Round Robin WF-336,0.038% Cu,0.658% Ni. See explanation above for treatment of the 21 data points used as input to this subset average value.
I1) Ilyron 1 Surveillance Data,0.022% Cu,0.690% Ni, Simple average of the 21 data points available from testing of the 11yron I surveillance materials, presumed to be taken from a separate block of WF-336.
- 12) llyron 2 Surveillance Data,0.023% C,0.712% Ni Simple average of the 31 data points available from test:ng of the Ilyron 2 surveillance materials, presumed to be taken from a separate block of WF 336.
When the above 12 subset values of copper and nickel (now including the additional subset values consisting of the simple mean of the 21 round robin data points) are summed and divided by the number of subset values, the "best estimate" chemistry for weld wire heat 442002 changes from 0.055% Cu and 0.617% Ni to 0.053% Cu and 0.621% Ni. When rounded in accordance with the American Society for Testing and Materials (ASTM) E29 Rounding Method the result remains 0.05% Cu and 0.62% Ni, and the best estimate chemistry factor using Regulatory Guide 1.99 Revision 2 Table 1 remains the same.
- 35. Ilyron Units 1 & 2 and liraldwaod Units I & 2 Fluence Methodology Justification The fast neutron exposure methodology documented in WCAP-14040 NP A,
" Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Ileatup and Cooldown Limit Curves,"is consistent with the requirements of Drall Regulatory Guide DG 1053," Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," and references neutrca transport cross sections derived from the ENDFAl VI data base. Ilowever, the exposure evaluations documented in WCAPs -13880, -14064, 14241, and -14228 for the 13yron Units 1 and 2 and Iltaidwood Units 1 and 2 reactor vessel PTLR submittals were completed prior to the release of the ENDF/Il VI-based Light Water Reactor neutron transport cross-section library. Consequently, the neutron transport calculations performed as an integral part of these evaluations were based on the then currently available ENDFal IV based transport cross-section and ENDFal-V dosimetry cross section libraries in all respects other than the ENDFAl VI versus k nts bytmd sgrp ptirtai Ac 20
t i
ENDF/Il IV & -V cross section issue, the mrhodology applied to the Ilyton Units 1 and 2 and Ilraidwood Units 1 and 2 Cuene evaluations in their respective PTLR .
submittals is identical to the approved methodology described in WCAP-14040 NP A.
Comed is proposing the use of the ENDF/Il IV & V based Quence estimates, in the revised pressure-temperature limit and LTOP calculations of the Ilyron Units 1
& 2 and Ilraidwood Units 1 & 2 PTLR submittals, only until the next surveillance capsules for each unit are withdrawn and evaluated.
Commonwealth Edison will re evaluate all applicable, previous surveillance capsules and vessel Duence estimates utilizing ENDF/Il-VI neutron cross section libraries in accordance seith WCAP 14040 NP A at the next scheduled capsule withdrat ' Sr each unit. The current capsule withdrawal schedule is capsule W for Ilyron ' I during IllR08 (Fall 1997), capsule X for Ilyron Unit 2 during Il2R07 (Spring 1998), capsule W for Braidwood Unit I during AIR 07 (Fall 1998), and capsule W for Ilraidwood Unit 2 during A2R07 (Spring 1999). If this schedule changes, Comed will assess the impact on fluence estimate margins and notify the NRC.
This re-evaluation will impact the manner in which materials data are utilized, and therefore constitutes a change in PTLR methodology. For this reason, all revised values of ART resulting from the new fluence values along with an evaluation of their impact on pressure temperature limits will be submitted to the NRC for review and approval. These re-evaluations will be completed and .ubmitted to the NRC within one year of the date of capsule runoval at each unit.
Until the capsule and vessel Duence estimates can be brought into accordance with WCAP 14040 NP A, the use of ENDF/B IV & Y based Guence estimates in the revised pressure-temperature limit and LTOP calculations of the Ilyron Units 1 &
2 end Ilraidwood Unh I & 2 PTLR submittals isjustified based on the following f'icts:
- 1. Since the evaluation of the current Huence estimates was performed, the actual inner wall Duence levels at Ilyron Units 1 & 2 and Ilraidwood Units 1 & 2 have been reduced by low leakage fuel management, beginning with the second fuel cycle of each unit. Low leakage fuel management has resulted in signi0 cant per cycle Quence reductior.s in similar units. Low leakage cores in place at Ilyron Units 1 & 2 and Ilraidwood Units 1 & 2 are estimated to produce at least a 5% total Cuence reduction as of the proposed EFPY applicability date of the four Ilyron and Ilraidwood Units.
Recognizing that low leakage fuel management benefits are realized gradually over a number of cycles, this Duence reduction is conservatively estimated from the aux values obtained from the cycle immediately preceding the most recent capsules withdrawn, as documented in WCAPs -
- k nla twind spp pinal dos 21
l 1
l 13880, 14064, 14241, and -14228 for Ilyton Units I and 2 and Iltaidwood Units I and 2. Flux levels from subsequent cycles would be expected to be even lower, resulting in an even greater actual Duence reduction. The estim:4ed minimum 5% fluence reduction as a result oflow leakage fuel management has not been considered in the ENDF/ll IV & -V based fluence estimates used in the revised pressure-temperature limit and LTOP calculations of the Byron Uni ts 1 & 2 and 13raidwood Units 1 & 2 PTLR submittals-
- 2. At the end of the nominal 18 month fuel cycles following the upcoming capsule pulls, none of the Hyron and Braidwood units wili achieve an ,
operating time, in effective full power years, greater than approximately 85.8% of that used to establish the current Duence estimates. This margm is based on the most limiting of the following: !
11yron 1 projected EFPY at BlRO9, which is the end of the cycle following the next capsule pull, is 10.3; Quence applicability date of proposed P-T limits is 12 EFPY; 10.3/12 = 85.8%
llyron 2 projected EFPY at 112RO8, which is the end of the cycle following i the next capsule pull, is 9.9, fluence applicability date on proposed P-T '
limits is 12 EFPY; 9.9/12 = 82.5%
liraidwood 1 projected EFPY at AIRO8, which is the end of the cycle following the next capsule pull, is 9.56; Ouence applicability date of proposed P T limits is 16 EFPY; 9.56/16 = $9.8%
Braidwood 2 projected EFPY at A2 tO8, which is the end of the cycle following the next capsule pull, is 10.18; fluence applicability date of proposed P T limit is 12 EFPY; 10.18/12 = 84.8%
The reduction in projected operating time, which is at least 14% for Braidwood Unit I and 2 and Byron Unit 1 and 2, will compensate for at least 14% of any potential underprediction caused by the use of ENDF/B-IV & -V based cross sections in the Buence evaluation.
- 3. In the evaluation of the current fluence estimates, it was recognized that the use of transport calculations based on ENDF/B IV & -V transport cross-sections results in an underprediction of the fluence at the vessel L nla byted agrp gttrraidoc22 -
l l
I wall. Therefore, the evahiations performed for the 13yron units 1 & 2 and ;
liraidwood Units 1 & 2, fluence calculations were increased based on the I comparision of capsule dosimetry measurements with calculations. This i renormalization of the calculations resulted in increases of at least 6% in I the projected vessel fluence over that which would be predicted based on calculation alone.
According to DG 1053, Section 1.1.2, footnote 2, the possible maximum underprediction of calculated values of vcssel inner wall fluence when ENDF/Il IV & V based fluence estimates are utilized is conservatively ,
estimated as 20%, due to the preser.:e of stainless steel reactor internal cornponents such as the thermal shield and neutron pads located between active fuel and the vessel inner wall. :
The combination of reductions in projected fluence of approximately 14%
due to a decrease in irradiation time,5% due to the implementation oflow leakage fuel management, and the 6% increase applied to the calculations for the current fluence estimates, provides a fluence margin at the end of the fuel cycles following the upcoming capsule pulls of at least 25%. This ;
is suflicient to compensate for the potential 20% underprediction of the ENDF/Il IV & V based fluence estimates, with a remaining conservatism of at least 5%. Ilased on this margin, the commitment to have fluence evaluations and associated P-T limits and LTOP calculations updated within one year of the applicable capsule withdrawal will ensure the vessel fluence estimates are in compliance with WCAP 14040.NP-A before reaching the end of the following fuel cycle.
s nla bytmd app pinal doc.23
A'ITACilMENT 11 WESTINGilOUSE LETTERS CAE-97 231/CCE-97 314 CAE-97 232/CCE 97 315 CAE-97 233/CCE-97 316 L nta inted agrp pilnai dr24 i
O CAE.97-231 CCE.97 314 Westinghouse Electric Corporation P. O. Box 355 Energy Systems Pittsburgh, Pennsylvania 15230-0355 January 6,1998 (412)374 6788 hir. Guy DeBoo Commonwealth Edison Company I400 Opus Place Downers Groove,IL 6051$
Comed Response to NRC Question to WCAP 14940, WCAP 14970 and 14824 Rev. 2 Dear hit. DeBoo For your information and use are attached responses to NRC question regarding Westinghouse WCAP rtports WCAP 14824 Rev. 2, WCAP.14940 and WCAP.14970. The following summarizes the attachments:
Attachment 1: Westinghouse response to NRC questions.
Attachment 2: Sample Data from pressure temperature curves run for Draidwood Unit 2 (Reference WCAP 14970) using the methodology from the 1996 Addenda to Section XI App.G.
This letter and its attachments must be inserted into all known copies of WCAP-14824 Rev. 2, WCAP.14940 and WCAP.14970. If you have any questions or need additional information, please contact the untersigned or Thomas J. Laubham at 412 374 6788.
V nily your wv. e o West nghou e . ectric Company C.S. Hauser, Project hianager Comed Project Attachment Nuclear Services cc: Drad Adams Dyron hiike Oorski- Draidwood
I l
NITACilMENT tr WESTINGliOUSE RESPONSE TO NRC QUESTIONS ON WCAP-14824 REV. 2, WCAP 14940 & WCAP 14970
BACKOROUND; j i
Several questions were raised during a telephone conference call between Westinghouse, Comed and the NRC on Decembw 10,1997. De subjc:t of the qeestions was the PTLR methodology described in ,
WCAP 14040 NP+A (Rev. 2) and exceptions taken to this methodology in recent submittals for Byron !
and Braidwood, he purpose of this reponse is to address these questions. ;
RESPONSE
One question involved identification of the methodology actually used to calculate strus intensity ;
factors for the prnsure temperature (P.T) limit curves (Byron Unit 2 and Braidwood Unit 2 Only). This calculation used the equations provided in the 1996 Addenda of Section XI Appendix G, for calculating both the thennal and pressure strus intensity factors. His exception to WCAP 14040 PTLR ;
methodology was explicitly stated in the submittal to the NRC by Commonwealth Edison. ;
For Byron Unit 2 and Braidwood Unit 2, where the 1996 Addenda is being initially implemented, a .
. ammary of the chanan in the ASME Section XI 1996 Addenda to Appendix 0 is provided beginning }
on page $ of WCAP 14940 (Byron 2) and WCAP 14970 (Braidwood 2). De stress intensity factor for -
membrane strus used equation (3) on page 6, and the calc lation for thermal stress intensity factor used i equations (4) and (5) on page 7. nose formulas were implemented in the OPERLIM computer code ,
used for the P.T limit calculations. No other changes were made to the OPERLIM code. Therefore, the :
P.T curve methodology is unchanged from that dc:eribed in WCAP 14040 Section 2.6 (equations 2.6.2 l 4 and 2.6.3 1) with the exceptionsjust described. j An example of this calculation. is shown in attachment 2. Included in attachment 2 are the results for i cooldown at 100*F/ hour, as a function of time. For och time step, the time, water tempwature, rate and i r
thermal strees intensity factors are tabulated, along with the four coefficients of the least squares cun e fit used to calculate the thermal stras intensity factor, KIT. Note that the output column header states l
- Cl, C2, C3 and C4. His corresponds in order to Co, Cl, C2 and C3 from the 1996 Addenda to Section XI. For two time steps (1440 sec. and 3240 sec), detailed stress distributions are also provided so that ,
theKIT can be independently vwified.
i r
With regard to welds and nonle shell plates that boarder the effective height of the fuel, they are
. addrased as follows:
Byron Units I & 2: (Ref. Per Comed: NDIT No. BRW DIT 97 391 Rev. 0) he weld seam in question is weld seam WF 501 (Unit 1) and WF 562 (Unit 2). For Byron Unit 1, the ;
t weld was made with wire heat # 442011 Flux Typa Linde 80, Lot # 8086. For Byron Unit 2, the weld ;
was made with wire heat # 442011 Flux Type Linde 80,' Lot # 8061. Due welds were fabricated with .
the same wire heat and flux type as the Braidwood Unit I and 2 weld seam WF 562 (Hat # 442011, Flux Type Linde 80,14 # 8061). Derefore, the best estimate chemistry factor, which is 41'F (based on Cu A Ni content), can be used for detamining the ART values for these Byron welds for position 1 1 1 of Reg. Guide 1.99 Rev. 2.~
i y 3- =e b y e-m=--e- w m n - --t4er c-- , = - .e---c- 4 --,e., - e m v-,- r w a -,-,a, w-,-+-n---- -.-r.,-,bwe,----y ye--% we %,=,~w. - -
e-
- m -- -- -
l The lower noule belt forging, otherwise known as the noule shell, is heat # 123J2'8 (Unit 1) and 4P.
6107 (Unit 2). Per Westinghouse and Comed records, only one chemistry test was performed on this material, and it shows the weight percent of copper and nickel to be Cu = 0.05,10 = 0.72 for Byron Unit I and Cu = 0.05, Ni = 0,74 for Byron Un!t 2. Hus, per Table 2 of Reg Guide 1.99 Rev. 2 the CF's would be 31 for each unit.
ne Duences for each Byron Unit at the top of the active fuel height for 12 EFPY is as follows: (NOTE: l These Huences are conservative for evaluating material above the active height of the fuel.) ,
TABLF1: Summary of Fluences and initial Propenies UrJt / EFPY Fluence Weld IRTm Nonle Shell IRT.
Byron Unit 1/12 EFPY 2.11 x IC'8 n/cm8 10' 30 :
Byror, Urs i /11. EFPY 1.73 x 10d n/cm' ,
40 10
- Nota. that Byton I has a different flux lot vs, the Byron 2 and Braidwood WeJds i ne ART Calcula'.lon is as follows: (Vessel nicknesi = f. $ inches)
Byron 1: (We;d: WF 501) 1/4T Cuence = 1.267 x 10 n/cm' 3/4T fluence =
, 4.569 x 10" n/cm' 1/4T FF = 0.466 3/4T FF = 0.279 1/4T ART,,o, a CF 'FF = 19.1 3/4T ARTmT
= CF'FF= 11.4
+ 9.55 = 5.7 1/4T oA 3/4T oA 1/4T Margin = 19.1 3/4T Margin = 11.4 1/4T ART = 10 + 19.1 +19.1 3/4T ART = 10 + 11.4 +11.4
= 48 = 33 I'
Byron 1: (Nonle Shell: 123J218) 1/4T Ouence = 1.267 x 10 n/cm' 3/4T Ouence = 4.569 x 10" n/cm8 1/4T FF = 0.466 3/4T FF = 0.279 1/4T ARTm = CF'FF= 14.4 3/4T ARTm = CF 'FF = 8.6 1/4T oA
= 7.2 3/4T oA
" 43 1/4T Margin = 14.4 3/4T Margin = 8.6 1/4T ART = 30 + 14.4 +14.4 3/4T ART = 30 + 8.6 +8.6
= 59 = 47 4
I f:
. , - ,-,--,,,w,..,--- a...... -
, Byron 2: (Weld: WF.562)
)
1/4T Duence = 1.039 x 10" n/cm' 3/4T fluence = 3.746 x 10" n/cm' '
1/4T FF = 0.425 =
3/4T FF 0.250 !
I/4T ART. = CF 'FF = 17.4 3/4T ART. = CF 'FF = 10.3
8.7
1/4T ca 3/4T oA 5.15 1/4T Margin = 17.4 3/4T Margin = 10.3 !
l/4T ART = 40 + 17A +17.4 3/4T ART = 40 + 10.3 +10.3
= 75 = 61 Byron 2: (Nonle Shell: 4P.6107) 1/4T fluence = 1.039 x 10" n/cm' 3/4T fluence = 3.746 x 10" n/cm' I/4T FF = 0.425 3/4T FF = 0.250 1/4T ART. - CF 'FF = 13.2 3/4T' ART,m = CF 'FF = 7.8 .
^
= 6.6 = 3.9 1/4T cA 3/4T ca 1/4T Margin = 13.2 3/4T Margin = 7.8 1/4 TART = 10 + 13.2 +13.2 3/4T ART = 10 + 7.8 +7.8
= 36 = 26 TABLE 2: ART's from Tables $ & 6 of WCAP.14824 Rev.2 and WCAP.14940 Byron Unit i Byron Unit 2 1/4T 3/4T 1/4T . 3/4T Intermediate Shell Forging 78 66 12.1 1.7 (R0 Position l
. __ __ '.") ____ . _ _ _ .. _ _ _ _ .. _ _ _ _ _ .. . _ _ _ _ _ .
using surveillance capsule data 70 60 31.7 14.9 (R0 Position 2.1)
Lower Shell Forging 52 38 11.8 1.5 (R0 Position 1.1)
Circumferential Weld WF.336 79 43 119.2 83.7
& 447 respectively (R0 Position li.".
) . __ .. .. _ . .. . _ _ ...
using credible surveillance 47 31 87.6 71.5 capsule data (R0 Position 248)
It can been seen in Table 2 that the welds and the nor21e shell forgings in question are well below the limiting material for their respect'n units and will not become limiting for their given fluences, s
e e - . -m -
, , n .,- -- -,-w ..
i
. i For completeness, the une of the Braidwood CF determined from suneillance capsule data (Position [
2.1) will be investigated, if the 12 criteria from WCAP.14824 Rev. 2 for data integration v.ere to be j applied for Byrvn and Braidwood, and for the purposes of thisjustification, it was determined that data i
, integration made sense between Byron and Braidwood for weld heat 442011, then the following ART values for Byron Unit 1 & 2 would be determined using a CF of 16.7'F per WCAP.14824 Rev. 2 (Appendix B) and WCAP 14970.
Byron 1: :
I/4T fluence = 1.267 x 10 n/cm8 3/4T fluence = 4.569 x 10" n/cm' :
1/4T FF = 0.466 3/4T FF = 0.279 ,
1/4T ART , = CF'FF= 7.8 3/4T ART,., = CF 'FF = 4.7 _ !
= 3.9 = 2.3 l I/4T oA -
3/4T ca I/4TMargin = 7.8 - 3/4T Margin = 4.7 1/4T ART = 10 + 7.8 +7.8 3/4T ART ' = 10 + 4.7 +4.7 !
26 -
_ 19 j Byron 2: :
s 1/4T fluence = 1.039 x 10* n/cm' 3/4T fluence = 3.746 x 10" n/cm8
= 3/4T FF = 0.250 I/4T FF 0.425 . i 1/4T ART,., = CF'FF= 7.1 _ 3/4T ART,., = CF 'FF = 4.2
= 3.55 = 2.1 1/4ToA 3/4T oo 1/4TMargin = 7.1 3/4T Margin = - 4.2 1/4T ART = 40 + 7.1 +7.1 3/4 TART = ,
40 + 4.2 +4.2
=- M = M ;
Again, these ART values are well below the limiting vessel material for their respective units.
Braidwood Units I & 2: (Ref. Per Comed: NDIT No. BRW.DIT 97 391 Rev 0) _
For Braidwood Units 1 & 2, the weld seam in question are WF.645 (Heat # H4493, Linde 80, Flux 0261). Per Comed reconis, the best estimate percent copper and nickel are 0.042 Cu and 0.46 Ni for both Braidwood I and 2. Dus the best estimate chemistry factor from Table 1 of Reg. Guide 1.99 Rev.
2 is $7'F. De fluences for each Braidwood unit at the top of the active fuel height for 16 EFPY
- (Braidwood 1) and 12 EFPY (Braidwood 2) is as follows:
The noule shell forgings for Braidwood I and 2 are heat #'s SP 7016 (Unit 1) and SP 7056 (Unit 2). ;
Per Westinghouse and Comed records, only one chemistry test was performed on this material, and it 3
- show the weight percent of copper and nickel to be Cu = 0.04, Ni = 0.71 for Braidwood Unit I and Cu = ;
- 0.01, Ni = 0.90 for Braidwood Unit 2. Thus, per Table 2 of Reg. Guide 1.99 Rev. 2 the CF's would be _;
. 26 for each unit. .
Y-
- f. -
f l
+-w. ~ ~ v , % -- -ae -,ww ve,.,,,-o.,m .-.n w,-,m,,,,. -n,,-,-e.g.,,-- wr, , - - - - - , , , . --m --+,a Nr. w -- m-,.=g-,- .w , y g n ,..v, -
I L
The Quences for each Braidwood Unit at the top of the active fuel height for 16 & 12 EFPY respectively, is as follows: : (NOTE: These Cuences are conservative for evaluating material above the active height of the fue;.) _
TABLE 3: Summary of Fluences and initial Prewrties Unit / EFPY Fluence Weld IRTm Nonle Shell IRTm Braidwood Unit I /16 EFPY 2. ;9 x 10 n/cm' 25 10 Braidwood Unit 2 /12 EFPY 1.63 x 105 n/cm' 25 30 i
i The ART Calculation is as foSows:(Vessel Thickness = 8.5 inches)
Draldwood 1: (Weld: WF 645)
!/4T fluence = 1.495 x 10" n/cm8 3/4T duence = 5.392 x 10" n/cm8 1/4T FF = 0.502 3/4T FF = 0.305 CF 'FF = =
- FF = - 17.4 1/4T uA
= 14.3 3/4T oA " 87 1/4T Margin = 28.6 3/4T Margin = 17.4 1/4T ART = 25 + 28.6 +28.6 3/4T ART = 25 + 17.4 +17.4
= 32 = , 10 Braidwood 1: (Nonle Shell: SP 7016)
!/4T Duence = 1.495 x 10" n/cm' 3/4T fluence = 5.392 x 10" n/cm' I/4T FF = 0.502 3/4T FF = 0.305 1/4T ART = CF 'FF = 13.1 3/4T ART. = CF 'FF = 7.9
= 6.6 = 4.0 1/4T oa 3/4T oo
= 36 = 26
- - . . . - , - - , - -,.- -..-.-~__.-- - .-.- -
Braidwood 2: (Weld: WF.645) 1/4T Ouence = 9.788 x 10" n/cm' 3/4T iluence = 3.530 x 10" n/cm' 1/4T FF = 0.413 3/4T FF = 0.241 1/4T ART, = CF 'FF = 23.5 3/4T ART, = CF 'FF = 13.7
= 11.8 = 6.9 1/4T oA 3/4T oA 1/4T Margin = 23.5 3/4T Margin = 13.7
= 22 = 2 Draidwood 2 (Nonle Shell: $P.7056) 1/4T Auence = 9.788 ~ 10" rt'em' 3/4T fluence = 3.530 x 10" n/cm' 1/4T FF = 0.i 3/4T FF = 0.241 1/4T ART, = CF = 10.7 3/4T ARTm = CF *FF = 6.3
= 5.4 = 3.2 1/4T oA 3/4T oA 1/4T Margin = 10.7 3/4T Mirgin = 6.3 1/4T ART = 30 + 10.7 + 10.7 3/4T ART = 30 + 6.3 +6.3
= 51 = 43 TAELE 4: ART's from Table 5 of WCAP.14243 and Tables 5 & 6 of WCAP.14970 Braidwood Unit I (16 EFPY) Braidwood Unit 2 (12 EFPY) 1/4T 3/4T 1/4T 3/4T Intermediate Shell Forging 25.1 8.2 2.2 8.3 (RO Position li")
Lower Shell Forging 26.2 12.1 29.5 10.1 (RO Position 1.1) using surveillance capsule data (RO 13.4 3.2 8.6 15.6 Position 2.1)
Circumferential Weld WF.336 & 112.9 90.5 105.9 84.4 447 respectively(RO Position li )
using credible surveillance capsule 76.6 65.4 66.9 58.1 data (RO Position 2'd)
It can been seen in Table 4 above, that the welds and the nonle shell forgings in question is well below the limiting material and will not become limiting.
I
{
With regards to the Lower Shell Circumferential welds at Byron Units 1 & 2 and Braidwood Units I and i l
2, they are all - 4 feet below the bottom of the attive fuel and will not see a fluence higher than 1 x 10".
nerefore they are not considered in the evaluations for PTS or pressure temperature limit cun es .
Author: enu ~
- Verifier:
nomas J. Laubha:V Ed Terek l Engineering & Materials Technology Mechanical Systems Integration ;
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A'ITACllMENT 2: Sample Data from prenure temperature curves run for tiraidwood Unit 2 t
e - -
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4 W4TER -
i ' TOE .. .TEOF . ' RATE IGT C1 C2 C3 -C4
- l. l15A08 .SE5ABB .-1 nab 5 ; , .303 .102E*04 - .1485E44 JISOE43 .73T4E42 ,
1 .
30BAIB Sama n ; 100.000 1 2.400 .2100E44 .2EBE+04 SBM43 .100E45 f :- Sa9ABB . SEGOB : -10BAIB - 3.848 75m -- 5R003 -10BAID : ' 5.212
.3OE44
.MOE*04
.3305E44
.4031E+04
.983E*03
.105E44
.0002E*32 '
.90eE+02
^
9WAOS $25 m '-10B m . 6.448 .4702E*04 .4512E+04 .1STE*04 .BUE+02
- 1. .. _
10 5.888 ' 5 3 088 -108. W B 7.003 .542E*04 .501E*04 - .112E+04 -A00E 42
!I i 1mL900 515AIS ' -105.33 ' BA48 .000E+04 .5412E*04 .1127E+04 -JWE+02 ;
140BAOD 510ABB -100.000 ' 9.812 .OSIE*04 .5B M +04 .1157E+04 .7322E42 iL '10 E 000'. M A N '-198 2 10A02 .7157E*04 .0144E+04 .1102E*04 -AET1E+02 l
[ 150.000 50B083 -.-1 E 080 '11.3 4 300E*04 .047E*04 .1100E*04 -A327E*02 ;
I s 19BBAIB 405m . -100m 12AST .8BFE*04 ' .ET45E+04 .118E+04 .5700E+02 l t~ 210BAOB 400.000 -100.008 .12.871 AISE404 .701E-M .1200E*04 .5479E*02 i 2300080 405ABB -100m ' 132 .W WE*04 .7237E*04 .1287E+04 .5030E*02 f
- j. - < : 25mm - 4E0BB -100m 13A15 2700.000 .475.000 -100.000 14.315
.9171E*04
.90$E*04
.74EE*04 703E+04
.122E44
.1222E*04
.470E*02
.4377E+02 i
[ 2 W 000 .470A00 ' -100.000 14.754- .973E+04 .781E*04 .1231E46 .4151E+02 ;
l: .: 3088.000 ; 405AOB -10BAOB 15.195 .9572E*04 .790E+04 .1232E*04 .302E*02 i j:
330B085 49.000 -100000 15.520 .1019E*05 -A1 M *04 .12EE*04 -JE3E+02 1 4
3420.3 0 . 405 008 -100.000 15A45 .10 M +05 -J227E+04 .12M .330E*02 ~i
[: : MOOD 450.0W -10BAOO ' 16.138 -
.1M E+05. -A3 M +04 .1204E+04 unN+02 -
!- 3700 ABO 445.000 : -10BADO ' 16AOO .1072E*05 .943eE*04 .12 M+04 .2570E+02 1 j JOBOABO 480.000 -100.000 16A35 .100E405 .052E+04 .1247E*04 .2835E*0Z
[
E 4140.000 435 ADO -100.000 . 16A82 .100E*05 -AES7E*04 .1245E*04 -m t t
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e THisISCOOLDOWNPROFILE4 (100DEGEMRCOOLDOWN )
IRRADIATION TIME = 12000 YEARS REFERENCE TEMPERATURE (RTNDT) = 66.900 DEGF FIAWLOCATION = .250T = AOWIN T WALL THERMAL STRESS FRESSURE STRESS RESULTNG WA'ER TEMPERATURE NTENSmf FACTOR NTENSITY FACTOR PRESSURE CALCUI.ATED TEMPERATURE AT AOWN T AT AOWN T AT AOWIP. T STRESS PRESSURE (DEGF) (DEGF) (KSI' .T.IN ) (KSISQRT.N) (PSI) (PSK;)
1 545 000 549225 9.59715E-01 6.76348E+03 00000E+00 145825E+05 2 540.000 547.015 145809E+00 6.54988E+03 00000E+00 138061E+05 3 535 000 544.187 3.84920E+00 6.28670E+03 00000E+00 228496E+05 4 530.000 541.144 5.21237E+00 6.015M. 00000E+00 2.18630E+05 5 525.000 537.817 6 44606E+00 5.73194E+03 00000E+00 2.08332E+05 6 520.000 534.365 7.60309E+00 5.45221E+03 00000E+00 1.98166E+05 7 515.000 530.743 8.64561E+00 517333E+03 00000E+00 1.88029E+05 8 510.000 527.012 9.61167E+00 /..W100E+03 00000E+00 1.78131E+05 9 505.000 523.158 1.04820E+01 4.63481E+03 00000E+00 1.68456E+05 10 500.000 519 205 1.12843E+01 4.37681E+03 00000E+00 1.59079E+05 11 495.000 515.158 120069E+01 4.12751E+03 00000E+00 1.50018E+05 12 490.000 511.022 126708E+01 3.887#9E+03 .00000E+00 1.41294E+05 13 485.000 506.810 1.32678E+01 3.65739E+03 00000E+0n 1.32931E+05 14 480.000 502.522 1.38148E+01 3.43714E+03 00000E+00 124926E405 15 475.000 498.170 1.43055E+01 322722E+03 00000E+00 1.17296E+05 16 470.000 493.754 1.47538E+01 3.02729E+03 00000E+00 1.10029E+05 17 465.000 489.285 1.51546E+01 2.33759E+03 00000E+00 1.03135E+05 18 460.000 484.760 1.55196E+01 2.65769E+03 00000E+00 9.65959E+04 19 455.000 480.192 1.58445E+01 2.48760E+03 00000E+00 9.04141E+04 20 450.000 475.576 1.61391E+01 2.32686E+03 00000E+00 8.45719E+04 21 445.000 470.924 1.63999E+01 2.17534E+03 00000EG 7.90648E+04 22 440.000 466.232 1.66352E+01 2.03257E+03 00000E+00 7.38754E+04 23 435.000 461.508 1.68418E+01 1.89830E+03 00000E+00 6.89955E+04 2
THIS IS COOLDOWN PROFILE 4 (100 DEG-FMR COOLDOWN )
TIME = 1.44000E+03 (SEC.) WATER TEMPERATURE = 510.000 (DEG.F)
CURRENT COOLDOWN RATE = 1.00000E+02(DEG.F/HR.) AVERAGE VESSEL TEMPERATURE = $34120 (DEG F)
FOR FLAW LOCATION = .250T BENDING STRESS = 9.52193E+03 (psi)
MEAN STRE0S = 2.86505E+03 (PSI)
VESSEL VESSEL THER'AAL THERMAL STRESS MESH RADIUS TEMPERATURE STF. DSS INTENSITY FACTOR LLNE (INI EG F) (PSn (KSt&Q.RT.lN )
1 86.625 510.389 6.65688E+03 9.61167E+00 2 86.710 511.187 6.42650E+03 9.61167E+00 3 86.795 511.980 6.in785E+03 9.61167E40 4 86.880 512.765 5.97163E+03 9.61467E+00 5 86.965 513.541 5.74834E+03 9.61167E+00 6 87.050 514.306 5.52834E+03 9.61167E+00 7 87.135 515.059 5.31188E+03 9.61167E+00 8 87.220 513.799 5.09913E+03 9.61167E+00 9 87.305 516.527 4.89018E+03 9.61167E+00 10 87.3 % 517.242 4.68509Et03 9.61167E+00 11 87.475 517.943 4.48385E+03 9.61167E+00 12 87.560 518.632 4.28646E+03 9.61167E+00 13 87.645 519.308 4.09287E+03 9.61167E+00 14 87.730 519.971 3.90304E+03 9.61167E+00 15 87.815 520.622 3.71690E+03 9.61167E+00 16 87.900 521.260 3.53440E+03 9.61167E+00 17 87.985 521.886 3.35548E+03 9.61167E+00 18 88.070 522.501 2.18006E+03 9.61167E+00 19 88.155 523.103 3.00808E+03 9.61167E+00 20 88.240 523.695 2.83949E+03 9.61167E+,00 21 88 325 524 275 2.67423E+03 9.61167E+00 22 88.410 524.844 2.51223E+03 9.61167E+00 23 88.495 525.402 2.35345E+03 9.61167E+00 24 88.580 525.949 2.19783E+03 9.61167E+00 25 88.665 526.486 2.04534E+03 9.61167E+00 26 88.750 527.012 1.89591E+03 9.61157E+00 27 88.835 527.528 1.74952E+03 9.61167E+00 28 88.920 528.034 1.60612E+03 9.61167E+00 29 89.005 526.530 1.46567E+03 9.61167E+00 30 89.090 $29.016 1.32814E+03 9.61167E+00 31 89.175 529.492 1.19348E+03 9.61167E+00 32 89.260 529.958 1.06166E+03 9.61167E+00 33 89.345 530.415 9.32648E+02 9.61167E+00 34 89.430 530.862 8.06412E+02 9.61167E+00 35 89.515 531.301 6.82917E+02 9.61167E+00 36 89.600 531.729 5.62129E+02 9.61167E+00 37 89.685 532.149 4.44018E+02 9.61167E+00 38 89.770 532.560 3.28551E+02 9.61167E+00 39 89.855 532.962 2.15696E+02 9.61167E+00 40 89.940 $33.355 1.05421E+02 9.61167E+00 41 90.025 533.739 2.30319E+00 9.61167E+00 42 90.110 534.114 1.07509E+02 9.61167E+00 43 90.195 534.482 2.10228E+02 9.61167E+00 44 90.280 534.840 3.10489E+02 9.61167E+00 45 90.365 535.191 -4.08324E+02 9.61167E+00 46 90.450 535.533 5.03763E+02 9.61167E+00 3
l 47- 90.535 - 535.867 5.96837E+02 9.61167E+00 48 90.620 536.193 6.87574E+02 9.61167E+00 -
49 90.705 - 536.511 7.7600$E+02 9.61167E+00 50 90 790 536 821 -8.62159E+02 9.61167E+00 51 00.875 537.123 44600$E+02 9.6)id7D00 '
$2 90.960 537.418 1.02775E+03 -
53 91.045 537.706 1,10725E+03 54 91.130 537.985 1.18458E+03 55 91.215 539.258 1.25978E+03 56 91.300 533.523 1.33287E+03 57 91.385 538.781 140388E+03 58 - 91,470 539.032 - 1.47284E+03 59 91.555 539.276 153976E+03 60 91.640 - 539.513 1.60469E+03 61- 91.725 539.743 1.66763E+03 62 91.810 539.966 1.72862E+03 63 91.895 540.183 1.78768E+03 64 91.980 540.393 1.84484E+03 65 92.065 - 540.596 1.90011E+03 66 92.150 540.793 1.95352E+03 .
67 92.235 540.984 2.00510E+03 68 92.320 541,168 2.05485E+03 69 92.405 541.346 2.10281E+03 70 92.490 541.51b 2.14900E+03 71 92.575 54147.4 2.19342E+03 72 92.660 541.844 2.23611E+03 73 92.745 541.997 2.27709E+03 1 74 92.830 542.145 2.31636E+03 -
75 92.915 542.287 2.35395E+03 76 93.000 542.423 2.38987E+03 '
77 93.085 542.553 2.42414E+03 78 93.170 542.678 2.45679E+03 79 93.255 542.796 2.48781E+03 80 93.340 542.910 2.51723E+03 81 93 425 543.017 2.54506E+03 82 93.510 543.119 2.57132E+03 83 93.595 543.216 2.59602E+03 84 93.680 543.307 2.61916E+03 -
85 93.765- 543.393 2.64078E+03 86 93.850 543.473 2.66086E+03 87 93.935 543.548 2.67943E+03 88 94.020 543.617 2.69650E+03 89 94.105 543.682 2.71207E+03 90 94.190 543.741 2.72616E+03 91 94.275 543.795 2.73878E+03 92 94.360 543.843 2.74992E+03 93 94.445 543.887 2.75961E+03 94 94.530 543.925 2.76784E+03 95 94.615 ' 543.958 2.77463E+03 96 94.700 543.986 2.77998E+03 97 94.785 544.009 2.78389E+03 98 94.870 544.027 -2.78637E+03 99 94.955 544.040 2.78743E+03 100 95.040 544.047 2.78707E+03 101 95.125 544.050 2.78529E+03 4
t V
f-a., ;
- THISISCOOLDOWNPROFILE4 (100DEGFMRC00LDOWN ): _
- TIE = 3 24000E+03(SEC.) t _
WATER TEMPERATURE = 460.000(DEGE) '
CURRENT C00LDOWN RATE = 1.00000E+02 (DEG FMR ' AVERAGE VESSEL TEMPERATURE = .497.082 (DEG.F) .- .} #
FOR FLAW LOCATION = .250T - BENDING STRESS = 1.38325E+04 (PSI) :
~_
MEAN STRESS = = 3.63288E+03 (PSI) / l
)
cVESSEL' - VESSELE THERMALi - THERMAL STRESS '!
MESH RADIUS: . TEMPERATURE -- STRESS _ INTENSITY FACTOR j
. LINE : (IN) - l(DEG.F) - (PSI) : (KSI SQ.RT.lN.)
I
- 1 86.625 460.549 1.01997E+04 1.55190E+01 2 - 86.710 461.665 9.b7817E+03 1.55196E+01
- 3; 86.795 - 462.774 - - 9.55890E+03' 1.55196E+011
' 4 :- . 86.800 - : 463.874 . 9.24293E+03 :1.55196E+01 = ,
- 5 86.965- 464.961 8.93050E+03 - 1.55196E+01 L6 87.050 . 466.035 8.62199E+03 1.55196E+01 1 7- 87.135 467.096 8.31757E+03 1.55196E+01 ..
l- :8- - 87.220 468.144 -- 8.01729E+03 . 1.55196E+01 2
, t 9 87.305 : ' 469.177 ' 7.72116E+03 : 1.55196E+01 10 87.390 470.197 - 7.42915E+03 1.55196E+01 - {
11- 87.475 471.203 7.14122E+03 1.55196E+01
- 12. - 87.500' 472.196 6.85731E+03 - 1.55196E+01 i 13- 87.645 - 473.175 - = 6.57737E+03 - 1.55196E+01
~14- 87.730 474.142- 6.30136E+03 - 1.55196E+01
'15- 87.815 -475.096 6.02922E+03 1.55196E+0i 16 87.900 476.037 5.76091E+03 .1.55196E+01
-17 87.985 ' 476.965 5.49641E+03 1.55196E+01 i 18 88.070 477.881 5.23568E+03 1.55196E+01 19 88.155 - 478.784 ~ 4.97871E+03 1.55196E+01 20 88.240 479.675 4.72546E+03 = 1.55196E+01
-21 ' 88.325 ~ 480.553 4.47503E+03 1.55196E+01 22 - 88.410 481.419 4.23000E+03 1.55196E+01 23 4 88.495 482.273 3.98793E+03 - 1.56196E+01 -
24 : :. 88.500 483.114 3.74943E+03 1.55196E+01 25 - 88.665 483.943 3.51459E+03 1.55196E+01 i 26 88.750 484.760 - 3.28339E+03, 1.55196E+01 27 ~88.835 485.565 3.05582E+03 1.55196E+01 ,
28 = 88.920 486.358 2.83185E+03 ' 1.55196E+01 29 - 89.005 487.138 2.61149E+03 1.55196E+01 30 80.000 - 487.907 2.39470E+03 1.55196E+01 31- 80.175 488.663 2.18149E+03 1.55196E+01 32 89.200 489.400 1.97182E+03 1.55196E+01 33 - 89.345 490.140 1.76569E+03 1.55196E+01
- 34 89.430-- 490.861 1.56300E+03 1.55196E+01 ,
i 35 89.515 491.570 - 1.36399E+03 1.55196E+01
.~ 36 '80.000 492.267. 1.16837E403 1.55196E+01 37- 89.685 - - 492.952 9.76232E+02 - 1.55196E+01 38 - 89.770 . 493.625 > < 7.87548E+02 1.55196E41
~ 30 L 89.855 . 494.287 6.02304E+02 1.55196E+01
. 89.940 . 494.937 4.20483E+02 1.55196E+01
- 41 90.025 - 495.576 2.42070E+02 1.55196E+01
- 42 = l 90.110f - 496.202 C.70494E+01 .1.55196E+01 5
7- -_ s
_m _-7 ,
43 90.195 496.818 1.04596E+02 1.55196E+01 44 90.280 497.422 2.72881E+02 1.55196E+01 45 90.365 498.015 4.37822E+02 1.55196E+01 46 90 450 - 498.596 5.99433E+02 1.55196E+01 47 90.535 499.166 7.57731E+02 1.55196E+01 48 90 620 499.724 9.12730E+02 1.55196E+01 49 90.705 500.271 1.06445E+03 1.55196E+01 50 90.790 500.808 - 1.21289E+03 1.55196E+01 51- 90.875 501.332 1.35808E+03 1.55196E+01 52 90.960 501.846 1.50003E+03 53 91.045 502.349 1.63876E+03 54 91.130 502.841 1.77427E+03 55 91.215 503.321 -1.90658E+03 56 91.300 503.791 2.03571E+03 57 91.385 504.250 2.16167E+03 58 91.470 504.697 2.28447E+03 59 91.555 505.134 2.40413E+03 60 91.640 505.560 2.52065E+03 61 91.725 505.976 2.63406E+03
- 506.380 2.74436E+03 S2 91.810 63 91.895 506.774 2.85157E+03 64 91.980 507.157 2.95669E+03 65 92.065 507.529 3.05675E+03 66 92.150 507.891 3.15475E+03 67 92.235 508.242 3.24971E+03 68 92.320 508.583 3.34163E+03 69 92.405 508.913 3.43053E+03 70 91490 509.232 3.51642E+03 71 92.575 509.541 -3.59931E+03 72 92.660 509.840 3.67921E+03 73 92.745 510.128 3.75614E+03 74 92.830 510.405 3.83010E+03 75 92.915 510.673 3.90110E+03 76 93.000 510.930 3.96916E+03 77 93.085 511.176 4.03427E+03 78 93.170 511.413 -4.09647E+03 79 93.255 511.639 -4.15574E+03 80 93.340 511.855 -4.21211E+03 81 93.425 512.060 4.26558E+03 82 93.510 512.256 4.31617E+03 83 93.595 512.441 4.36387E+03 84 93.680 512.616 4.40870E+03 85 93.765 512.781 -4.45066E+03 86 93.850 512.936 4.48978E+03 87 93.935 513.081 -4.52604E+03 88 94.020 513.215 -4.55947E+03 89 94.105 513.340 4.59006E+03 90 94.190 513.454 -4.61783E+03 91 94.275 513.559 -4.64279E+03 92 94.360 - 513.653 -4.66493E+03 93 94.445 513.738 -4.68428E+03 94 94.530 513.812 4.70082E@
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O CAE-97-232 CCE-97 315 Westinghouse Electric Corporation P. O. Box 355 Energy Systems Pittsburgh, Pennsylvania 15230-0355 January 6,1998 (412)374 6788 Mr. Guy DeBoo Commonwealth Edison Company 1400 Opus Place ,
Dokners Groove,IL 60515 Comed Transmittal of Updated Lebles to WCAP-14940 and WCAP-14970 Dear Mr. DeBoo For your information and use are the attached tables from WCAPs-14940 and WCAP-14970 which entain the following updated information:
Attachment 1: 1.) Table 1 (page 9) of WCAP-14940 contains two additional copper and nickel data points which were taken out of the overall average. The overall average now becomes 0.05 % Copper and 0.71% Nickel. 2.) Table 2was revised to add additional weld copper / nickel values from round robin testing. 3.) Table 3 was revised to incorporate the new averages and add a note for explanation. 4.) Table 5 was revised to incorporate the new averages. 4.) Table 6 was revised to match the surveillance capsule data for the forging with the 1/4T ART & 3/4T ART values of 11.8 and 1.5 respectively, and to retract the word credible under the lower shell forging material type for surveillance data.
A sentence was added to note 'a' for clarification.
' Attachment 2: Table 1 (Page 9) of WCAP-14970 was revised to correct a two typos. 1.) Copper value for the Lower Shell Forging 50D102/50C97-1 should read 0.049. 2.) The standard deviation for the L: ter Shell Forging $0D102/50C97-1 should read 0.0046. 3) Table 6 was revised to retract the word credible under the lower shell forging material type for surveillance data. A sentence was added to note 'a' for clarification.
. ,. . . ~ _. . . -. . _ _ _ . _ . , > _ . . _ . _ _ _ _ _ _ . . - , . . _ _ _ _ _ _ _ . _ _ _ _ , _ _ . . . . . . _
i
- ~ '-
Mr. Guy DeBoo . 2- CAE-97 232:
. CCE.97 315 ~ -
This letter and its attachments must be inserted into all known copies of WCAP 14940 and WCAP 14970.L If you
- have any questions or need additional information, please contact the undersigned or Thomas J. Laubham at 412- .;
^
374-6788i Very truly yours,
- Westinghouse Electric Company
- ri n v C.S. Hauser, Project Manager h
Comed Project Attachment Nuclear Services
'- cc: . Brad Adams Byron
- Mike Gorski Braidwood -
J a
j
-7 1
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ATTACIIMk.NT 1: UPDATES TO WCAP-14940 s
. 9 All materials in the beltline region of the Byron Unit 2 reactor vessel were considered in determining the limiting material. The calculations to determine the ART values for beltline materials at 12 EFPY are shown in Table 5. The resulting ART. values for all beltline region materials at the 1/4T and 3/4T locations are summarized in Table 6, where it can be seen that the limiting materialis the circumferential weld (based on credible surveillance capsule data).
The 1/4T and 3/4T ART values for circumferential weld (based on credible integrated surveillance capsule data) were used in the generation of heatup and cooldown curves applicable to 12 EFPY.
TABLE 1 Calculation of Average Cu and Ni Weight % Values for the Byron Unit 2 Base Metals Inter. Shell Forging Lower Shell Forging 49D329/49C297 1 MK 24-2 49D330/49C298 1 MK 24-3 Reference Cu % Ni% Cu% Ni%
WCAP -14063 0.01 0.70 0.07 0.65 0.006 0.70 0.022 ,
0.689 Byron Unit 2 0.006 0.71 0.05 0.73 Heatup & Cooldown 0.05* 0.73*
for Normal 0.05* 0.75*
Operation I"1 0.067 0.772 Average 0.01 0.70 0.05 0.71 Standard Deviation 0.002 0.01 0.02 0.05 These chemistry values are from the ladle test performed on the individual heats 490330 and 49C298 which F mprises the material in the lower shell forging, and are not included in the overall averag< J Cu & Ni. Two separate ladie melts of unknown relative volumes were pourod to make a single forging inget. The individualladle chemistries are upstream process analyses which are not representative of the final solidified ingot. Therefore, only the check chemistry is included in the average. Reference The Japan Steel Works Material Test Results under contract No. 640-0012, JSW Job No. FN3-4268, dated July 1974.
Byron Unit 2 Heatup and Cooldown Limit Curves October 1997
10 TABLE 2 Calculation of Average Cu and Ni Weight Percent Values for the Byron Unit 2 !
Weld Material (Using Byron 1 & 2 Chemistry Test Results)(4 I l
Weld Type Cu% Ni% ;
1 B&W Weld Qualifications 0.024 0.70 (BAW-2261) _ 0.031 0.46 0.03 0.72 0.068 0.48 0.114 0.54 0.148 0.60 0.053 0.62 0.059 0 62 B&W Weld Qualification'* 0.029 0.65 Round Robin Data'" 0.038 0.658 Byron 1 Surv. Weld Data'* 0.022 0.690 Byron 2 Sury. Weld Data
- 0.023 0.712 Best Estimate Chemistry 0.053 0.621 Standard Deviation 0.040 0.087 NOTES:
(a) The weld materialin the Byron Unit i surveillance program was made of the same wire and flux as the reactor vessel intermediate to lower shell girth seam weld. (Weld seam WF 336, Wire Heat No. 442002, Flux Type Linde 80, Flux Lot No. 8873)
(b) The Byron Unit 2 surveillance weld is identical to that used in the reactor vessel core region girth seam (WF-447). The weld wire is type Linde MnMoNi(Low Cu-P), heat number 442002, with a Linde 80 type flux, lot number 8064.
(c) Cu & Ni values obtained from WCAP-14824 Rev.10* ,
(d) Obtaine(f from Reference 17.
(e) Obtaineo from Reference 18.
TABLE 3 Byron Unit 2 Reactor Vessel Material Properties Material Description CU (%) Ni(%) Chemistry initial Factor
- RTnor ( F)*
Closure Head Flange Not Reponed 0.74 -- 0*
Vessel Flange Not Reported 0.73 -
30'4 Intermediate Shell Forging 0.01 0.70 20 -20 49D329/49C297 1 Lower Shell Forging 0.05 0.71 32.2* -20 49D330/49C298 1 Circumferential Weld 0.05 0.62 68.0 10 NOTES:
(a) Chemistry Factors are calculated from Cu and Ni values per Regulatory Guide 1.99, Revision 2.
(b) Initial RT,,oy values are measured values.
(c) Closure head and vessel flange Initial RTuo7 values are used for considering flange requirementsm for the heatup/cooldown curves.
(d) This chemisty factor was determined using a Cu value of 0.052 and is considered conservative.
Byron Unit 2 Heatup and Cooldown Limit Curves October 1997
14
~
TABLE 5 Calculation of Adjusted Ret .ence Temperatures (ART) at 12 EFPY for all Byron Unit 2 Reactor Vessel Material (based on credible st rveillance capsule data)
Reactor Vessel Belthne Material i @ 12 Region Locaban identification Cu% Ni% CF" EFPY %-tf4 %-t FF I ART,o188 o, c. M ART *
(x 10") Etf Wt FF
% T Calculation intermediate Shet Forging 490329 11 0.01 010 20.0 0.822 0.494 0.803 -20 16.06 0 8.03 16.06 12.1 49C297-1 Lower SheB Forgog 490330-11 0.05 0 71 32.2 0.822 0.494 0.803 -20 25.86 0 12.93 25.86 311 49C298-1 Lower sheH Forgog 19.8 0.822 0.494 0.803 -20 15.90 0 7.95 15.90 11.8
- using S/C Data Wek;;Aetal WF-447 0.05 0.62 68 0 0.822 0.494 0.803 10 - 54.60 0 27.30 54 60 119.2 (Heat 442002)
Weld Metal 61.3 0.822 0.494 0.803 10 49.63 0 14.00 28.00 87.6
-+ using S/C Data
% T C3ndannn Intermediate SheB Forging 490329-11 0.01 0.70 20.0 0522 0.178 0.542 -20 10.84 0 5.42 10.84 1.7 49C297-1 Lower Shen Forging 49D330-11 0.05 011 32.2 0.822 0.178 0.542 -20 17.45 0 8.725 17.45 14.9 49C298-1 Lower sher Forging 19.3 0.822 0.178 0.542 -20 10.73 0 5.375 1013 1.5
-4 using SIC Data Weld Metal WF-447 0.05 0.62 68.0 0.822 0.178 0.542 10 36.86 0 18.43 36.86 83.7 (Heat 442002)
Weld Metal 61.8 0.822 - 0.178 0.542 10 33.50 0 14.00 28.00 71.5
- using SIC Data NOTES:
(t) The Byron Unit 2 reactor vessel wall thickness is 8.5 inches at the beltline region.
(b) ART = 1 + ART,eg + M (This value was rounded per ASTM E29. using the " Rounding Method".)
- FF (d) The CF is integ ated betweer- :he Byron 1 Weld (WF-336, heat # 442002) and the Byron 2 Weld (WF-447. Heat # 442002).
Byron Unit 2 Heatup and Cooldown Limit Curves October 1997 l
_ 15 TABLE 6 Summary of Adjusted Reference Temperatures (ART) at 1/4T and 3/4T Locations for 12 EFPY Material 12 EFPY 1/4T ART 3/4T ART truarmediate Shell Forging 12.1 1.7 49D329/49C297 1 (RG Position 1(a))
Lower Shell Forging 31.7 14.9 49D330/49C298 -1 1 (RG Position 1(a))
using surveillance 11.8 1.5 capsule data (RG Position 2(a))
Circumfe.'ential Weld 119.2 83.7 (RG Position 1(a))
~ ~
using credible surveillance 87.6M 71.5M capsule data (RG Position 2(a))
NOTES:
(a) Calculated using a chemistry factor based on Regulatory Guide (RG) 1.99, Ravision 2, Positions 1 and 2. Note: Surveillance data was determined to be NOT credible per WCAP-14824 Rev. 2, however, the results are reported here for completeness.
(b) These ART values were used to generate the Byron Unit 2 heatup and cooldown curves.
Byron Unit 2 Hestup and Cooldown Limit Curves October 1997
= . - -. . -.. . . -. - -, , .- . . . . . . . . - _ .-.--. .._ ~ -_
l i 27 )
13L Babcock & Wilcox drawing numbers 185265E, Rev. 2; " Reactor Vessel General Outline".
WCAP 14824 Rev.1, " Byron Unit 1 Heatup and Cooldown Limit Curves for Normal j Operation and Surveillance Wold Metalintegration for Byron and Braidwood" P. A. -
i Grendys, April 1997.
1 15 WCAP-14063, " Byron Unit 2 Heatup and Cooldown Limit Curves for Normal Operation",
P. A. Peter, November 1994.
16 l.S Raju and J.C. Newman, Jr., Stress intensity Factor influence Coefficients for Intemal ;
and External Surface Cracks in Cylindrical Vessels", in Aspect of Fracture Mechanics in l
Pressure Vessels and Piping, ed. S.S. Palusamy and S.G. Sampath, PVP-Volume 58, ASME 1982.
17, . Nuclear Design Information Transmittal, NDIT No. BYR97-346, Rev.0, " Additional data .
point for weld wire heat number 442002 for incorporation into Table 2 of WCAP-14824 ;
Rev.1", Dated 9/9/97.
- 18. NDIT No, BRW-DIT-97-391,~ Rev. O, " Byron and Braidwood Stations Units 1 & 2 3- Supplemental Reactor Vessel Materials Data", M.A. Gorski of Comed Braidwood, Dated 1 5-98.
4 4
e Byron Unit 2 Heatup and Cooldown Limit Curves - October 1997 py - - -
,,y .6, .,w-- + r , < , . - , _ _ r__.,- %- r-.-- . ,e_,, -.y. r-, - - . _ . ---- - < .~~.,m.. . , . - ,, --,. ---w m. --
A
- ATTACHMENT 2: UPDATES TO WC/.P-14970 4
9 w
a materials at the 1/4T and 3/4T locations are summarized in Table 6, where it can be seen that
- the limiting material is the circumferential weld (based on credible surveillance capsule data).
The 1/4T and 3/4T ART values for circumferential weld (based on credible integrated surveillance capsule data) were used in the generation of heatup and cooldown curves applicable to 12 EFPY.
TABLE 1 Calculation of Average Cu and Ni Weight % Values for the Braidwood Unit 2 Base Metals Upper Shell Forging Lower Shell Forging 49D963-1/49C904-1 -
50D102-1/50C97J1 Reference Cu % Ni% Cu% Ni%
Ref.18 0.03 0.71 Ref. 4 0.057 0.77 Ref.17 0.049 0.745 Ref.19 0.06 0.75 Ref. 20 0.056 , 0.804 Average 0.03 0.71 0.06 0.77 Standard Deviation 0.00 0.00- 0.0046 0.027 Note: Averages were originally documented in WCAP-14230nsi, i
s Braidwood Unit 2 Heatup and Cooldown Limit Curves October 1997
..._ .--m_______-.__.___m ____.s._ _
10 TABLE 6 Summary of Adjusted Reference Temperatures (ART) at 1/4T and 3/4T Locations for 12 EFPY Material 12 EFPY 1/4T ART 3/4T ART Upper Shell Forging 2.2 -8.3 49D963-1/49C9041 (RG Position 1(a))
Lower Shell Forging 29.5 -
10.1 50D1021/50C97-1 (RG Position 1(a))
using surveillance -8.6 -15.6 capsule data (RG Position 2(a))
Circumferential Weld 105.9 84.4 (RG Position 1(a)) ~ ~
using credible surveillance 66.95I 58.1
- capsule data (RG Position 2(a))
NOTES:
(a) Calculated using a chemistry factor based on Regulatory Guide (RG) 1.99, Revision 2, Positions 1 and 2. Note: Surveillance data was determined to be NOT credible per WCAP 14824 Rev. 2, however, the results are reported here for completeness. -
(b) These ART values were used to generate the Braidwood Unit 2 heatup and cooldown curves.
l f
, Braidwood Unit 2 Heatup and Cooldown Limit Curves October 1997 l
\. .
O CAE.97 233 CCE-97-316 Westinghouse Electric Corporation P. O. Box 355 Energy Systems Pittsburgh, Pennsylvania 15230-0355 January 6,1998 (412)374 6788 Mr. Guy DeBoo Commonwealth Edison Company 1400 Opus Place Downers Groove,IL 60515 -
Comed Transmittal of Updated Tables to WCAP-14824 Rev. 2 Dear Mr. DeBoo .
For your information and use are the attached tables from WCAPs 14824 Rev. 2 which contains the following updated infctmation:
Attachment 1: Table 2 (page 9) of WCAP-14824 Rev. 2 contains one additional copper and nickel data ,
point. The overall average remains at 0.05 % Copper and 0.62 Nickel. Note a new reference was added, therefore, page 26 was revised. >
Attachment 2: Figures 1 and 3 (pages 18 and 20) of WCAP-14824 Rev. 2 were revised to correct the criticality plot.
Attachment 3: Tables 7 and 8 (pages 22 and 23) of WCAP-14824 Rev. 2 were revised to correct the criticality data points.
i e - Mr. Guy DeBoo CAE-97-233 !
.CCE 97 316
- This letter and its attachments must be inserted into all known copies of WCAP-14824 Rev. ?.. If you have any questions or need additional infornation, please contact the undersigned or Thomas J. Laubham at 412 374-6788.
Very truly yours, Westyi house Electric Com any
<%se . *-
C.S. Hauser, oj t Manager Comed Pro' ect j Attachment Nuclear Services 4
= Lcc:' Brad Adams - Byron Mike Gorski- Braidwood I s L._ _
e 1
i i
l ATTACHMENT 1: UPDATES TO TABLE 2 l
i 1
e I
l 1
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1
1 2
9
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TABLE 2 Calculation of Average Cu and Ni Weight Percent Values for the Byron Unit 1
' Weld Material (Using Byron 1 & 2 Chemistry Test Results)
Best-Estimate Reference Cu Ni B&W Weld Qualification BAW 2261 0.024 0.7 B&WWekt Dualifcation 0.031 0.46 B&W Weld Qualification 0.03 0.72 B&WWeld Qualification 0.060 0.48
- 0,114 B&W Weld Qualification 0.54 B&W Weld Qualification 0.148 0.6 B&WWeld Qualifcation 0.053 0.62 B&W Weld Qualifcation 0.059 0.62 B&WWold Qualificaton Ref. 23 0.029 0.65 Round Robin WF-336 Ref. 28 0.038 0.658
, Byron i Survedlance Data See Below 0 022 0.690 -> 0.02 0.69 Surv. CF = 27 Byron 2 Surveillance Data See Below 0.023 0.712 -> 0.02 0.71 Surv. CF = 27 Best-Estimate Chemistry ($; 0.053 0.621 -> 0.05 0.82 Best Est. CF = 68 Standard Deviation: 0.040 0.087 Byron 1 & 2 Ratio = 2.5"'
Surveillance Data Chemistry Results: Byron Unit 2 Syron Unit f Reference Cu Ni Reference Cu Ni WCAP 10398ld) 0.03 0.65 WCAP-9517I3) 0.026 0.71 WCAP 12431[22] 0.024 0.740 WCAP 11651[21] 0.023 0.67 0.024. 0.786 0.022 0.665 0.022 0.704 0.021 0.714 0.020 0.681 0.021 0.741 0.021 0.706 0.022 0.713 0.020 0.697 0.021 0.714 0.019 0.668 0.020 0.704 0.022 0.759 0.020 0.694 0.021 0 714 0.020 0.706 0.020 0.678 0.021 0.677 0.020 0.695 0.023 0.677 0.019 0.689 0.021 0.680 0.021 0.744 0.021 0.680 0.022 0.738 0.021 0.667 0.022 0.771 0.024 0.677 WCAP 14064(11] 0.024 0.705 0.022 0.697 0.023 0.706 0.021 0.634 0.023 0.698 WCAP 13880l9] 0.024 0.682 0.024 0.696 0.022 0.678 0.023. 0.711 0.025 0.70$ 0.024 0.708 Avera9e 0.022 0.sW0 0.024 0.716 0.024- 0.715 0.024 0.707 0.024 0.720 0.024 0,717 0.024 0.711 0.024 0.706 0.024 0.707 0.025 0.717 Average 0.023 0.712 Byron Unit i Heatup and Cooldown Limit Curves November 1997
26
- 26. WCAP 9807, " Commonwealth Edison Co. Braidwood Unit 1 Reactor Vessel Radiation Surveillance Program", S.E. Yanichko, et. al., February 1981.
- 27. WCAP 11188, " Commonwealth Edison Co. Braidwood Unit 2 Reactor Vessel Radiation Surveillance Prograni", L. R. Singer, December 1986.
- 28. NDIT No. BRW DIT 97-391, Rev. O, " Byron and Braidwooo Stations Units 1 & 2 Supplemental Reactor Vessel Materials Data", M.A. Gorski of Comed Braidwood, Dated 1 5-98.
Byron Unit i Hostup and Cooldown Limit Curves November 1997
ATTACIIMENT 2: UPDATES TO FIGURES I AND 3 D
18 MATERIAL PROPEP.TY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL FORGING 5P 5933 (using surv. capsule data)
LIMITING ART VALUES AT 12 EFPY: 1/4T, 70'F 3/4T, 60*F d)3,d% ;
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0 1 1 I I i 0 50 100 150 200 250 3$0 3$0 4$0 4$0 500 Indicated- Temperature (Deg.F)
FIGURE 1 Byron Unit 1 Reactor Coolant System Heatup Lirnitations (Heatup Rates up to 10#F/hr) Applicable for the First 12 EFPY (Without Margins for Instrumentation Errors)
. Byron Unit 1 Heatup and Cooldown Limit Curves November 1997
20 4
MATERIAL PROPERTY BASLS-
- LIMITING MATERIAL: INTERMEDIATE SHELL FORGING SP 5933 (using surv.capsvie data)
LIMITING ART VALUES AT 12 EFPY: 1/4T, 70*F 3/4T, 60'F s.ws ,
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. FIGURE 3 . Byron Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rates up to 100*F/hr) Applicable for the First 12 EFPY (Without Margins for '
Instrumentation Errors; Margin of 74 psig for Pressure Difference Between Pressure Instrumentation atgi the Reactor Vessel Beltline Region)'
Byron Unit 1 Heatup and Cooldown Limit Curves November 1997 k
s.
r ATTACHMENT 3: UPDATES TO TABLES 7 AND 8 0
4 l
I I
22 f
TABLE 7 Byron Unit i Heatup and Cooldown Data at 12 EFPY Without Margins for Instrumentation Errors includes 1) Vesset flan 9e requirements of 180'F and 621 psig per 10CFR50.
I Cooldown Curves - Heatup Curve j Steady State 25F SOF 100F 100F Criticality. Limit Leak Test Limit T P T P T P T P T P T P T P l 60 621 60 595 60 554 60 470 60 621 203 0 182 2000 65 621 65 610 65 570 65 489 65 621 203 0 203 2485 70 621 70 621 70 587 70 509 70 621 203 620 75 621 75 621 75 605 75 531 75 621 220 620 80 G21 80 621 80 621 80 554 80 621 220 671 85 621 85 621 85 621 85 579 85 621 220 657 90 621 90 621 90 621 90 607 90 621 220 646 95 621 95 621 95 621 95 621 95 621' 220 639 100 -621 100 621 100 621 100 621 100 021 220 634 105 621 105 621 105 621 105 621 105 621 220 632 110 621 110 621 110 621 110 621 110 621 220 633 115 621 115 621 115 621 115 621 115 621 220 637 120 621 120 621 120 621 120 621 120 621 220 642 125 621 125 621 125 621 125 621 125 621 220 651 130 621 130 621 130 621 130 621 130 621 220 661 135 621 135 621 135 621 135 621 135 621 220 674 140 621 140 621 140 621 140 621 140 621 220 689 145 621 145 621 145 621 145 621 220 707 150 621 150 621 150 621 22C '17 155 621 155 621 220 749 160 621 160 621 220 774 165 021 165 621 220 801 170 621 170 621 220 8:1 175 621 175 621 220 864 180 621 180 621~ 220 900 180 1483 180 000 225 938 185 1559 185 938 230 980 190 1640 190 980 235 1026 195 1728 195 1026 240 1075 200 1821 200 1075 245 1128 205 1921 205 1128 250 1186 210 2020 210 1186 255 1247 215 2143 215 1247 260 1313 220 2266 220 1313 265 1385 225 2397 225 1385 270 1461 230 1461 275 1543 235 1543 280 1630 240 1630 285 1724 245 1724 290 1825 250 1825 295 1933 255 1933 300 2048 260 2048 305 2171 265 2171 310 2302 270 2302 315 2441 275 2441 (Configuration #9393315685880 for Cooldown. #2756858809292 for Heatup)
Byron Unit 1 Heatup and Coo!down Limit CuNes November 1997
. 23 TABLE 8 Byron Unit i Heatup and Cooldown Data at 12 EFPY Without Margins for instrumentation Errors inc8udes 1) Vessel Aan9e requirements of 180"F and 621' poig per 10CFR50, and 2) Pressure adjustment of 74 '
psig to account for pressure difference between the wide-range pressure transmitter and the limiting beltline region of the reactor vessel.
Cooldown Curves Heatup Curve -
- Steady State - 25F SOF 100F 100F Criticality. Limit Lenk Tes; ' %it T P T P T P T P T P T P T P 60 547 60 521 60 480 60 396 60 547 203 0 182 2000 65 547 65 536 65 496 65 415 65 547 203 0 203 2485 70 547 70 547 70 513 70 435 70 547 203 546 75 547 75 547 75 531 75 457 75 547. 120 546 80 547 80 547 80 547 80 480 80 547 220 597 85 547 85 547 85 547 85 505 85 547 220 583
-90 547 90 547 90 547 90 533 90 547 .220 572 95 547 95 547- 95 547 95 547 95 547 220 565 100 547 100 547 100 547 100 547 100 547 220- 560 105 547 105 547 105 547 105 547 105 547 220 558 110 547 110 547 110 547 110 547 110 547 220 559 115 547 115 547 ' 115 547- 115 547 115 547 220 563 120 547 120 547 120 547 120 547 120 547 220 568 125 547 125 547 125 547 125 547 125 547 220 577 130 - 547 130 547 130 547 130 547 130 547 220 587 135 547 135 547 135 547 135 547 135 547 220 600 140 547 140 547 140 547 140 547 140 547 220 615 i 145 547 145 547 -145 547 145 547 220 633 150 547 150 547 150 547 220 653 155 547 155 547 220 675 160 547 160 547 220 700 165 547 165 547 220 727 170 547 170 54T 220 757 175 547 175 547 220 790 180 547 180 547 220 826 180 - 1409 180 826 225 864 185 1485 185 864 230 906 190 1568 190 906 235 952 195 1654 195 952 240 1001 200 1747 200 1001 245 1054 205 1847 205 1054 250 1112 210 1955- 210 1112 255 1173 215 2069 215 1173 260 1239 220 2192 220 1239 265 1311 225 - 2323 225 1311 270 1387 230 1387 275 1469
-235 1469 280- 1556 240 1556 285 1650 245 1650 290 1751 250 1751 295 1859 255 1859 300 1974 260 1974 305 2097
. 265 7097. 310 2228 270 & 12 8 315 2367-(ConAgurshon #9396838589093 for Cooldown, #9291115685880 for Heatup) 1
- ' Byron Unit 1 Hestup and Cooldown Limit Curves - November 1997
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