ML20207H750
| ML20207H750 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 07/12/1999 |
| From: | Levis W COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20207H753 | List: |
| References | |
| BYRON-99-0095, BYRON-99-95, NUDOCS 9907210201 | |
| Download: ML20207H750 (6) | |
Text
Coramonwcalth Edium Compan)
,15) ron Generating 5tation 4450 North Gcrm.m Church Riud llyron. IL 610l49794 Tel H15 2.4 4 5 4 41 i
July 12,1999 1
LTR:
BYRON 99-0095 FILE:
2.01.0320 l
U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 l
Byron Station, Units 1 and 2 l
Facility Operating License Nos. NPF-37 and NPF-66 NRC Docket Nos. STN 50-454 and STN 50-455
Subject:
Pressure Temperature Limit Report (PTLR)
/
References:
(1) Letter from William Levis (Comer.o NRC," Reactor Vessel Material Surveillance Capsule W i est Results and Schedule l
For Completing Assessment of Reactor Vessel Materials Data,"
dated January 11,1999.
(2) Letter from William Levis (Comed) to NRC, " Reactor Vessel Material Surveillance Capsule X Test Results and Schedule Fo'r Completing Assessment of Reactor Vessel Materials Data,"
j dated April 12,1999.
(3) Letter from R.A. Capra (U.S. NRC) to O.D. Kingsley (Comed),
" Byron Station, Units 1 and 2 and Braidwood Station, Units 1 and 2, Acceptance for Referencing of Pressure Temperature Limits Report," dated January 21,1998.
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Pursuant to Byron Technical Specification 5.6.6, " Reactor Coolant System Pressure and Temperature Limits Report," this letter transmits the revised PTLRs for Byron Station i
Units 1 and 2 (i.e., Attachments A and B). Our intent is to implement these revised PTLR's by May 15,2000. This letter also transmits revised pressurized thermal shock
- }
(PTS) evaluations for Byron Station Units 1 and 2 (i.e., Attachments E and F).
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The PTLR is based on surveillance capsule testing for Byron Station Units 1 and 2 (i.e.,
References 1 and 2), and credibility analyses of the data from the surveillance capsule test results (i.e., Attachments G and H). The surveillance capsule report and credibility reports also provided inputs for the PTS evaluations.
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U.S. Nuclear Regulatory Commission Page 2 The methodology used to develop the pressure temperature limits reported in the PTLR is in accordance with Technical Specification 5.6.6.b and Reference 3. In addition, the NRC review comments from Reference 3 have been addressed as described below.
Neutron Fluence The fluence evaluation in the capsule reports (i.e., References 1 and 2) are based on the ENDF/B-VI database. In addition, previously tested capsules' fluence analyses performed with ENDF/B-IV wsre updated to the ENDF/B-VI cross-sections. The fluence calculated using the ENDF/B-VI data base were used consistently to evaluate all of the surveillance data as well as the adjusted reference temperature used in the PTLR and PTS reports for Byron Unit 1 and Unit 2 reactor vessels.
Charpy V-Notch Data All Charpy V-notch data were replotted using a symmetric hyperbolic tangent curve
. fitting program.. Previously the Charpy data were based on hand fit curves using engineering Judgement. The CVGRAPH Version 4.1 was used to generate new curve fits for all charpy data. The new curve fits were used to determine the measured shift,
- ARTe, for each capsule.
Best Estimate Chemistry The applicable Byron welds are WF336 (i.e., Unit 1 girth weld - weld wire 442002) and for weld WF447 (i.e., Unit 2 girth weld - weld wire 442002). The best chemistry values for welds WF336 and WF447 were updated. The chemistry results from the latest capsule tests (i.e., References 1 and 2) were included in the best estimate chemistry using the mean of the means technique. Two weld qualification records were removed from the best estimate chemistry calculation. This was based on original fabrication records from Babcock and Wilcox (B&W) showing welds WF336 and WF447 were fabricated using non-cooper coated weld wire from heat 442002. See Attachment D, "The B&W Owners Group Reactor Vessel Working Group Byron and Braidwood RV Wald Chemistry initial RTc 51-5002206-00," from Framatome Technologies, in this report, B&W demonstrated two purchases of weld wire 442002 were made, one copper coated and one not. The non-copper coated wire was used to fabricate the welds WF336 and WF447 in the beltline region as defined by 100FR 50.61. The two weld
. qualifications that were removed from the best estimate chemistry were produced with copper coated wire. The mean of the means technique war, used to obtain the best estimate chemical composition.
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' Unirradiated Reference Temperature -
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?The weld metal unirradiated reference temperatures (RTw) used to calculate the adjusted reference temperatures (ART) have different values for the same heat of j
s material. Specifically,c j
Material rte Byron 1 RTw Byron 2 442002
-30.0 *F (WF336) 10.0 *F (WF447) 442011 10.0 *F (WF501) 40.0 *F (WF562) 31401 10.0 *F (WF472) 40.0 *F (WF614) 1 The RTa values presented in the table above are the measured values from the
. fabrication records (i.e., Attachment C). The fabrication records showed that the test temperatures were not the same and the tests reported were not all carried out to failure.
Examination of the data for the RTa values indicates that the +10*F value for 442002 and the +40 *F value for 442011 were obtained as a result of not continuing the drop weight tacting to lower temperatures until a break occurred. The Certified Material Test
-l Report (CMTRs) for heat 442002 in Attachment C show no break for drop weight testing
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of Unit 1 material fo -20 *F, resulting in an RTa -30.0 *F, with no break and a break at
--30.0 *F. However the unit 2 materiai was tasted to +10.0 *F with no break. No test results of a break were reported since there were no tests carried out at lower temperaturesc Since material was not tested to failure the values actually measured for I
the Unit 1 and Unit 2 materials have been applied. The impact of using the actual measured values is minimal. The Unit 1 forging 5P-5933 will be the controlling material; there is only a three degree difference in the ART ultimately.
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. Similarly, for heat 442011, testing was also not carried out to failure for the drop weight l
. tests in Attachment C, WF501 (Unit 1) had a drop weight test performed at +20 *F and
- demonstrated no break. Additional testing shown in the revised CMTR shows additional l
drop weight testing down to -40 *F with one break. This additional testing demonstrating the +10*F for Unit 1 is conservative. Weld wire heat 442011 for Unit 2 only performed a l
- drop weight test at +50 *F with no breaks. Since Unit 2 material was not tested to failure, the values actually measured for the Unit 1 and Unit 2 materials have been applied.
i The properties for 31401 will not be a controlling material. This material is approximately 4 feet below the bottom of the active fuel. Unit 1 will see fluence values of (calculated) ll 2.67 X 10 n/cm '(32 EFPY) and 4.47 X 10 n/cm (54 EFPY), E > 1MeV. Unit 2 will 2
2 see fluence values of (calculated) 1.99 X 10 n/cm (32 EFPY) and 3.35 X 10 n/cm 8
2 (54 EFPY), E > 1MeV.
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U.S. Nuclear Regulatory Commission Page 4 '
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' Chemistry Factor When the ART for the forging SP-5933 was calculated for Unit 1 in WCAP 15124 rev 0
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(Attachment G) an overly conservative mar i
margin value using the entire margin of 17* gin factor was used. An overty conserva F was applied; not one-half the value of the ARTeer (25.8'F/2) per the guidance outlined in previous reactor vessel integrity
- workshops. The impact of this is a conservative ART value which was applied to the
- 16 EFPY PTlimits for Byron Unit 1.
' Credibility Analysis-i Attachments G and H contain the surveillance data credibility analysis for both Byron
" capsules W(Unit 1) and X (Unit 2).
Pressure Temperature Curves The PTLR, Attachments A and B contain the heatup and cooldown curves calculated for.
16 EFPY. The current PTLRs are valid for 12 EFPY for each unit. These revised 1
PTLRs will be implemented by May 15,2000, prior to exceeding this limit on either unit.
The methodology used was from 1996 ASME B&P Vessel Code,Section XI, Appendix G. The methodology used to develop the pressure temperature limits reported in the PTLR is the same as reviewed and approved in WCAP 14040-NP-A with SER dated October 16,1995, to Mr. Roger A Newton from Christopher 1. Grimes titled " Acceptance for Referencing of Topical Report WCAP-14040, Revision 1, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves (TAC #M91749)" with the stress intensity tactors determined using 1996 ASME B&PV code section XI appendix G. This was also approved by the NRC in an SER, NRC letter from R. A. Capra, NRR, to O. D. Kingsley, Commonwealth Edison Co., " Byron Station,' Units 1 and 2, and Braidwood Station, Units 1 and 2, Acceptance for Referencing of Pressure Temperature Limits Report (TAC Numbers M98799, M98800, M98801, and M98802)," dated January 21,1998. The heatup and cooldown curves were generated without margins for instrumentation errors. The curves include a hydrostatic leak test limit curve from 2485 to 2000 psig and pressure-temperature limits for the vessel flange regions per the requirements of 10 CFR Part 50,.
Appendix G.'
Pressurized Thermal Shock Attachments E and F are the PTS evaluations for Units 1 and 2. The conclusion of the report is that both units will remain significantly below the screening criteria for 32 EFPY and 48 EFPY (10 CFR 50.61).
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[ Byron Ltr. 99-0095 1 July 12,'1999..
U.S. Nuclear Regulatory Commission Page 5 L
If you have any questions regarding this correspondence, please contact Mr. B.J.
l Adams, Regulatory Assurance Manager, at (815) 234-5441, extension 2280.
i
- Respectfully, 1
) C ( $
Ll 2 William Levis Site Vice President-Byron Station
.WLlJD/dpk 1-Attachments:- A - Unit 1 PTLR dated June 28,1999 B - Unit 2 PTLR dated June 28,1999 C1 - CMTR's for weld wire heats, Unit 1 C2 - CMTR's for weld wire heats, Unit 2 D "The B&W Owners Group Reactor Vessel Working Group Byron and Braidwood RV Weld Chemistry Initial RTm 51-5002206 00,"
Framatome Technologies.
E - WCAP-15125, Rev. O, " Evaluation of Pressurized Thermal Shock for -
Byron Unit 1".
F - WCAP-15177, Rev. O, " Evaluation of Pressurized Thermal Shock for i
Byron Unit 2".
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G - WCAP 15183, Rev. O, " Commonwealth Edison Company Byron Unit 1
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Surveillance Program Credibility Evaluation".
1 H - WCAP 15180, Rev. O, " Commonwealth Edison Company Byron Unit 2 Surveillance Prograni Credibility Evaluation".
1 i-WCAP 15124, Rev. O, " Byron Unit 1 Heatup and Cooldown Limit Curves for Normal Operation".
i J - WCAP 15178, Rev. O, " Byron Unit 2 Hestup and Cooldown Limit Curves for Normal Operation".
cc:
Regional Administrator-NRC Region 111
. NRC Senior Resident inspector-Byron Station 1
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Attachment A Byron Station j
Unit 1 PTLR dated June 28,1999 i
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