ML20207F102
| ML20207F102 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 03/03/1999 |
| From: | Ted Sullivan BOSTON EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20207F107 | List: |
| References | |
| RTR-REGGD-01.099, RTR-REGGD-1.099 2.99.022, NUDOCS 9903110165 | |
| Download: ML20207F102 (13) | |
Text
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Boston Edison A BEC ENERGY COMPANY 10CFR50.90 T.A. Sullivan Vice President Nuclear and Station Director March 3,1999 BECo Ltr. 2.99.022 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Docket No. 50-293 License No. DPR-35 REQUEST FOR TECHNICAL SPECIFICATION CHANGE CONCERNING VESSEL MATERIAL SURVEILLANCE INTERVAL (CAPSULE PULL)
Purpose Boston Edison Company (Pilgrim Nuclear Power Station) requests NRC review and approval for a change to Pilgrim Technical Specification Table 4.6-3. The proposed change substitutes "18 (approx)" under the column " Effective Full Power Years" for the current "15 (approx)." The attached "No Significant Hazards Considerations" evaluation is provided with the proposed change.
The current requirement of "15" EFPY was incorporated into Pilgnm's Technical Specifications by Amendment No.140, effective May 16, 1990. Amendment No.140 was developed in accordance with Regulatory Guide 1.99, Revision 2. Pilgrim proposes changing Amendment No.140 based on its fluence report { attached) and in conformance with the guidance provided in Regulatory Guide 1.99, Revision 2.
Pilgrim proposes this change because the second capsule withdrawal, currently scheduled for the ned refueling outage (RFO No.12), leaves only one surveillance capsule. The remaining capsule must remain in the vessel until the end-of-life, currently listed as 32 EFPY on Table 4.6-3.
Further, an industry effort (
Reference:
BWR Vessel and Intemals Project [BWRVIP]) has been initiated to develop an integrated surveillance program (ISP) which would incorporate exis'ing capsules along with supplemental capsules. The BWRVIP has committed to submit a program plan to the NRC by the end of 1999 (BWRVIP memorandum to its executive committee dated ;
January 21,1999). Pilgrim is likely to participate in the final ISP. To take full advantage of this ([
program Pilgrim must defer its next capsule pull, which is currently due during refueling outage 4 12, scheduled to commence in May 1999. /
9903110165 990303 ~
PDR ADOCK 05000293
'300068 ]
P PDR_ :(
Pilgrim Nuclear Power Station, Rocky Hill Road. Plymouth, Massachusetts 02360
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ckaroynd Data from the first capsule pull (4.17 EFPY) indicates Pilgrim's pressure-temperature curves l
(Technical Specification Tables 3.6-1, 3.6-2, and 3.6-3) represent conservative values and '
operation of Pilgrim consistent with the operating regime defined by these curves provides j assurance of reactor vessel integrity for the remainder of plant life. The attached information i and "No Significant Hazards Considerations" are provided to demonstrate deferring Pilgrim's j capsule withdrawal until 18 EFPY is justified and does not impact safe operation, j Please contact P.M. Kahler at (508) 830-7939 if you should require further information on this !
issue.
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'. 7v A. A. Sullivan !
l Commonwealth of Massachusetts) :
County of Plymouth .) l K
Then personally appeared before me, T. A. Sullivan, who being duly swom, did state that he is j Vice President Nuclear, Station Director of Boston Edison Company and that he is duly 1, authorized to execute and file the submittal contained herein in the name and on behalf of !
Boston Edison Company and that the statements are true to the best of his knowledge and r belief. '
My commission expires
/ DATE i/
JMC 'MW1_ MM (2 NOTAR UBLIC W
Attachments
- 1) Narrative on Proposed Change and "No Significant Hazards Consideration" (
- 2) Pilgrim Fluence Report
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- 3) Proposed Changed Pilgrim Technical Specification Page 3/4.6-13 ,
- 4) Marked-up Current Pilgrim Page 3/4.6-13 l i
TAS/PMK/cis 2.99.022
3 I I cc: Mr. Alan B. Wang, Project Manager Mr. Robert Hallisey ,
- Proje'ct Directorate 1-3 Radiation Control Program Office of Nuclear Reactor Regulation Commonwealth of Massacliusetts Mail Stop: OWFN 14B20 Exec Offices of Health & Human Services 1 White Flint North Dept. of Public Health 11555 Rockville Pike 174 Portland Street !
Rockville, MD 20852 Boston, MA 02114 U.S. Nuclear Regulatory Commission Mr. Peter LaPorte, Director Region i Mass. Energy Management Agency ;
475 Allendale Road 400 Worcester Road King of Prussia, PA 19406 P.O. Box 1496 Framingham, MA 01701-0313 Senior Resident inspector Pilgrim Nuclear Power Station I
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l ATTACHMENT 1 TO PILGRIM LETTER 99-022 l
Description of Proposed Chance l 1
Pilgrim Nuclehr Power Station (Pilgrim) Technical Specification Table 4.6-3, " Reactor Vessel Material Surveillance Program Withdrawal Schedule," currently gives the value of "15 (approx)" effective full l power years (EFPY) for the second capsule withdrawal. Pilgrim proposes to change "15 (approx)" to "18 approx)" EFPY for this capsule pull.
Reason for Proposed Chance i l
Pilgrim proposes this change because, after the second capsule withdrawal (i.e., pull number 2), there l is only one capsule left which must remain in the vessel until the end-of-life, currently listed as 32 l EFPY on Table 4.6-3. !
Data from the first pull (4.17 EFPY) indicates Pilgrim's pressure-temperature (P-T) curves (Technical ,
Specification Tables 3.6-1, 3.6-2, and 3.6-3) represent conservative values and operation of Pilgrim j consistent with the operating regime defined by these curves provides assurance of reactor vessel i integrity for the remainder of plant life. !
Further, an industry effort (
Reference:
BWR Vessel and Intemals Project [BWRVIP]).has been initiated to develop an integrated surveillance program (ISP) that would incorporate existing capsules
- along with supplemental capsules. The BWRVIP has committed to submit a program plan to the NRC
. by the end of 1999 (BWRVIP memorandum to its executive committee dated January 21, 1999).
Pilgrim is likely to participate in the final ISP. To take full advantage of this program Pilgrim must defer its next capsule pull, which is currently due during refueling outage 12, scheduled to commence in l May 1999.
Safety Evaluation and No Sionificant Hazards Considerations 10CFR50.91 requires that at the time a licensee requests an amendment, it must provide to the Commission its analysis using the standards in 10CFR50.92, to determine no significant hazards considerations. In accordance with 10CFR50.91, Pilgrim has performed an analysis for the proposed change to Technical Specification Table 3.6-3 that increases the second reactor vessel capsule pull from (approximately) 15 EFPY to 18 EFPY. It is provided below.
- The opr edon of Pilgrim in accordance with the proposed amendment will not involve a signlNcant i,ocrease in the probability or consequences of an accident previously evaluated.
Pilgrim meets the requirements of paragraph 50.55a and General Design Criteria 1,14,31 and 32 of Appendix A of 10 CFR Part 50 by providing assurance material of the reactor coolant pressure boundary (RCPB) possess adequate fracture toughness properties to resist rapidly propagating failure and act in a non-brittle manner when stressed under operating, maintenance, testing, and anticipated operational conditions. The requirement, in part, of General Design Criterion 32 is met by conducting a surveillance program to monitor the change in fracture toughness properties of the ferritic materials in the reactor vessel.
The fracture toughness requirements for ferritic materials in the pressure retaining components of the RCPB are specified for testing and operational conditions, including anticipated operational occurrences, in Section IV of Appendix G of 10 CFR Part 50. This appendix requires the acceptance I and performance criteria of Appendix G of Section lli of the ASME Boiler and Pressure Vessel Code. l Pressure-temperature calculation procedures are described in Appendix G of the ASME code while i the detailed technical basis for the ASME code requirement is provided by the Welding Research I
- Council (WRC) Bull: tin 175, "PVRC R:comm:nd: tion on Toughn ss R:quirem:nts for Ferritic Mit:ri Is." Ching s in th3 fr:ctura toughn:ss prop:rti:s of mit: rills in th3 b:ltlins rrgion, rssulting from neutron irradiation and the thermal environment, are monitored by a surveillance program in compliance to the requirements of Appendix H of 10 CFR Part 50. The effect of neutron fluence on the shift in the nil ductility temperature of pressure vessel steel is predicted by Regulatory Guide 1.99, "Effect of Residual Elements on Predicted Radiation Damage to Reactor Vesse! Materials." Pilgrim currently uses Regulatory Guide 1.99, Revision 2, in conformance with NRC guidance and an NRC Safety Evaluation Report (SER) dated January 29, 1992 for Amendment No.140 to Pilgrim's Technical Specifications.
Technical Specification section 3/4.6.A, " Thermal and Pressurization Limitations," applies to the periodic examination and testing of the reactor pressure vessel. Table 3.6-3 is part of this section and provides a schedule of capsule pulls governed by EFPY.
Pilgrim withdrew its first capsule during the October 1979 refueling outage (RFO-4). Southwest Research Institute (SWRI) was contracted to test the capsules and provide a report based on the results of the testing (SWRI report 02-5951 dated July 1981) which established the reactor vessel fluence up to December 31,1979. The results follow: l Location Fluence EFPY ARTuor l
Pressure Vessel I.D. Surface 2.6x10n/cm2 4.17 27 F l (Projected to END-OF-LIFE) 2.0x10'8n/cm* 32.0 78 F Pressure Vessel @ 1/4 T Loc Son 1.8x10"n/cm2 4.17 22 F ~ ,
(Projected to END-OF-LIFE) 1.4x10 n/cm 2 1
32.0 65 F I
Pilgrim performed calculations and developed new reactor vessel P-T curves, based on extrapolation ;
of the SWRI results. The new graphs were incorporated into Pilgrim's Technical Specifications via Amendment No. 82. These extrapolated values were subsequently found to be overly conservative and, in 1985, more realistic fluence calculations were performed by General Electric Company (GE) ;
(
Reference:
GE report No. 277-1285, dated November 7,1985) based on the DOT neutron transport '
methodology, the original SWRI data, and the core reload history through mid cycle number 7. The revised values are:
)
Location Fluence EFPY ARTuor 2
Pressure Vessel 1.D. Surface 4.24x10'7n/cm 8.08 36 F Pressure Vessel @ 1/4 T Location 2.84x10'7n/cm2 8.08 29 F in 1986 BECo contracted Teledyne Engineering Services (TES) to develop new Technical Specification P-T limit curves for the reactor vessel. As part of this project, Pilgrim calculated new end-of-life fluences based on the previously referenced SWRI and GE reports.(
Reference:
Pilgrim calculation M-256). The new fluence values (extrapolated to end-of-life) are:
Location Fluence EFPY (END-OF-LIFE) ARTuor 2
Pressure Vessel I.D. Surface 1.46x10'8n/cm 32 67 F !'
Pressure Vessel @ 1/4 T Location 0.98x10'8n/cm2 32 56 F To comply with the 10CFR50 Appendix H requirements for the second capsule withdrawal (i.e., next to last capsule), the capsule withdrawal time (in terms of EFPY) may occur when either the capsule accumulated neutron fluence corresponds to the approximate neutron fluence of the reactor vessel
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, inn:r will location (or 1.48x10 n/cm*) ct end-of-lifs or 15 EFPY which:ver com:s first (R:farance:
ASTM E-185, Tcbb 1, footnota b).
Pilgrim's withdrawal schedule for the second capsule is not technically limited to 32 EFPY because the cumulative neutron fluence for the capsule at end-of-life will not reach the cumulative neutron fluence of the vesselinside surface prior to end-of-life. This statement is also true for the most limiting material, which is the RPV lower intermediate shell longitudinal seam welds 1-338 A,B,C. Therefore, f considering the criteria of ASTM E-185, Table 1, footnote (b) the lesser of the two conditions would be i 15 EFPY. This is an administrative limit and not a technical restriction. l l
There are other considerations justifying defe tal of the second capsule withdrawal from RFO-12 to a later RFO:
. The BWRVIP in their January 21, 1999, memorandum recommended that nuclear plants scheduled for capsule withdrawal within the next 18 to 24 months postpone withdrawal.
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. Withdrawing the next-to-last capsule during RFO-12 would leave no capsules remaining for ,
l future evaluation except the last capsule that ASTM E-185 requires withdrawn at 32 EFPY. j l Therefore, the possibility of gaining future interim information from a capsule would be lost !
unless Pilgrim can defer withdrawal of the second capsule.
. . The current data provided in Regulatory Guide 1.99, Revision 2, conceming the effects of long term radiation embrittlement and the overall affects of neutron exposure of reactor pressure l vessel (RPV) metal and weld materials is still evolving for light water reactors (LWRs), and data 1 base scatter is significant; hence, the need for the large 2a i 'riation to accommodate the l
scatter. As time progresses and technological advances on ' 2 effects of neutron exposure become available, a better understanding of these effects ans more accurate predictions can be made. Therefore, deferring capsule withdrawal will allow Pilgrim to use advances in this technology. l l
. Pilgrim has maintained its inventory of previously tested capsules at Southwest Research l Institute. Pilgrim believes technology will be developed allowing re-installation of new capsules I made from these used specimens. The technology is not currently available, but it is reasonable to defer withdrawal at least one cycle to take advantage of any state-of-the-art development in this technological area. Such deferral does not impact safety because Regulatory Guide 1.99, Revision 2, and past specimen testing indicate Pilgrim's current P-T l
curves are conservative and ensure safe operation of the plant.
. Pilgrim's reactor vessel history of accumulated fluence from initial startup to projected end-of-i life (32 EFPY) is provided in attachment 2. This information shows that Pilgrim is not and is not expected to reach significant fluence values at the most limiting location (i.e., Weld 1-338 at the l
1/4 T ) prior end-of-life. A significant fluence value is considered fluence greater than 1x10 '
2 n/cm . Therefore the effect of radiation embrittlement is not a significant concem for the life of Pilgrim and is not a concem for the current estimated RFO-12 operating history of 15-16 EFPY. '
The deferral of the second capsule pull at Pilgrim does not challenge safety but does defer a Technical Specification surveillance. Pilgrim's configuration and operational practices are not changed by this proposed change. Pilgrim's current Technical Specification P-T curves are l conservative and are not changed by this proposed change. The existing P-T curves were reviewed I
and approved by the NRC in Technical Specification Amerdment No.140 dated January 29,1992. ,
Operation in accordance with the existing P-T curves ensures reactor vessel and cooling system 1 integrity. The capsule pull is a surveillance technique that provides data for modification of the curves. Since margins will not change absent data from the capsule pull, and since the methods used
- ' to dev: lop tha tImperatur:s cssoci ted with thiss curvss cro regtrded ts cons:rvrtiva and currantly reside in Pilgrim's Technical Specifications with NRC reviiw End approval, ops ution of Pilgrim in accordance with the proposed change continues to be conservative and safe.
For the reasons given above, deferral of withdrawal of Pilgrim's second capsule for at least one additio'nal cycle (or 3 EFPY) is justified and desirable, and operation of Pilgrim in accordance with the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
- The operation of Pilgrim in accordance wlth the proposed amendment will not create the posalbmty of a new or dlWorent kind of accident from any accident previously evaluated.
As discussed above, the deferral of the second capsule pull at Pilgrim does not challenge safety but does defer a Technical Specification surveillance. Pilgrim's configuration and operational practices are not changed by this proposed change. Pilgrim's current Technical Specification P-T curves are conservative and are not changed by this proposed change. The existing P-T curves were reviewed and approved by the NRC in Technical Specification Amendment No.140 dated January 29,1992.
Operation in accordance with the existing P-T curves ensures reactor vessel and cooling system integrity. The capsula pull is a surveillance technique that provides data for modification of the curves. Since margins will not change absent data from the capsule pull, 8nd since the methods used to develop the temperatures associated with these curves are regardsd as conservative and currently reside in Pilgrim's Technical Specifications with NRC review and approval, operation of Pilgrim in accordance with the proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated.
- The operation of Pilgrim in accordance with the proposed amendment will not involve a algnMcant reduction in the mergin of anfety.
As discussed above, the deferral of the second capsule pull at Pilgrim does not challenge safety but does defer a Technical Specification surveillance. Pilgrim's configuration and operational practices are not changed by this proposed change. Pilgrim's current Technical Specification P-T curves are conservative and are not changed by this proposed change; the existing P-T curves were reviewed and approved by the NRC in Technical Specification Amendment No.140 dated January 29,1992.
Operation in accordance with the existing P-T curves ensures reactor vessel and cooling system
. integrity. The capsule pull is a surveillance technique that provides data for modification of the curves. Since margins will not change absent data from the capsule pull, and since the methods used to develop the temperatures associated with these curves are regarded as cont.ctvative and currently reside in Pilgrim's Technical Specifications with NRC review and approval, operation of Pilgrim in accordance with the proposed change will not involve a sig ~ ant reduction in the margin of safety.
These changes have been reviewed and recommended for approval by the Operations Review Committee, and reviewed by the Nuclear Safety Review and Audit Committee.
Schedule of Channe This change will be implemented within 30 days following Pilgrim's receipt of its approval by the Commission.
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ATTACHMENT 2 TO PILGRIM LETTER 99-022 1 Pilarim Fluence Report i
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PNPS Fluence Renort ;
i l Pilarlm Fluence vs EFPY l Per Rors.1 &21 !
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References:
- 1. SWRI Report 02-5951 Dtd. 7/81 [1 EFPY= 31538000 sec/yrj j
- 2. GE Report 277-1285 Dtd.11/27/85 ' 3.15E+07 j
- 3. Pil0rlm Calc M-258 (&Att. A) Dtd.1/23/88
- 4. CE Dwg. E-232-351-3 Vessel intamal Attachments
- 5. Sudds 91-81 (TR 7487-1) PNPS RPV T.S. P/T Limit Curves Capsule locations are at 30,120,300 degrees from "0" azimuth . ,
Worst case weld location (Heat 27204/12006) is at 80,180 and 300 degrees fmm "0" azimuth l Peak flux occurs at the 24.5 degree azimuth, i
- E: The Azimuthal Lead Factor for the capsules is .87(Ref.1) or 1/.87= 1.15 . i The Azimuthal Lead Factor for the capsules is .93 (Ref. 2) or 1/.93= 1.08 !
0' cepeuse (Typ) !
179I ese 300* E _.
A-33sA,s.c .
- Wald (Typ) 270* 90' !
ICore/ l ReactorVesselWet .
r 120* i
-210* 80 t 180*
Fluence Azimuth Location iper Rors.264) Axial Location (per Ret 5) ,
peaks Aalnudh Land Fessors Capsule WeM (" Nodes per GE MDE277-1285)
Ashnuth SWIWpist GEptof.8) past No. ID leem Elev (in)
- Node 24.5 1 Bot. Lower Head 1 0 N/A 30 0.87 0.93 Bot. Active Fuel 211.1 1 80 0.87 0.93 79 1-338 A Girth Wid.1-344 242.5- 5 85.5 1 Lower int. Shell 242.5 to 5 to 114.5 1 Weld 1-338 A,B,C 398 24 155.5 1 Top Active Fuel 355.1 24 180 ,_ 0.39 .
204.5 1-210 0.87 0.93 80 1-338 B 245.5 1 294.5 1 300 0.87 0.93 81 1-338 C i
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- PNPS Fluence Ranort Per Ref.1(Unmodified) NOTE
- " Unmodified" refers to results recorded from Ref.1 (Flux rate is an Average per Ref.1)
, Determine vessel (& weld) lead factors and Dementages Ref.1 reported (pg.16) that the capsule to ve:ssel lead factor is .87 Ref.1 (pg.16) also reported that the " total" effect at the vessel 1/4 T and 3/4 T locations are 1.3 and 3.8 (Respective Lead Factors)
Ref.1 (pg. 31 reported the raduction in fluence at the 1/4 T and 3/4 T locations I i
as 67% and 23% respedively.-
These values are obtained as follows.
3 Maximum vessel exposure = Capsule Exposure / l.ead Factor = 1/.87 = 1.15 i Vessel Exposure @ 1/4 T x Max Vessel Exp. = (141m. 67% of Capsule j -
Vessel Exposure g 3/4 T = (143.ac1. 23% of Capsule NOTE: The SWRI report did not adjust the fluonae lead factor to acountfor the location 2 of the anM relative to the capseda. However, considering the common location -
- of the capsule and the longitudinal seam weld (1-338 C) at 300* Azimuth, the .
weld surface fluence would be comparable to the capsule (i.e. Lead Factor =1)
Also, SWRI did not mention (nor consider) an axial lead fador.
- Azimuthal Lead Facter = 0.87 (Capsule to Vessel) l Capsule / Weld fluxte,,,= 1.74E+09 n/cm'-se (Unmod)
Vessel flux <cycq= 2.00E+09 n/an'-se (Unmod) 1 Cycle = .18 Mos.' _ . Fluence per SWRI Report (Unmodiflod)
From Oct 1979 Base Metal Wold 1-338(Host 13s0013Fass)
- Cycle Capsule Plate & ID Plate 81/4T Wid.dR AD Wold. d t 1/4T
- (Est.) EFPY Fluence Fluence ARTsarr Fluence ARTaor Fluence ARTuor Fluence' ARTayr 1.1 1.0 5.5E+16 6.3E+16 10.00 4.2E+16 6.05 5.5E+16 19.62 3.7E+16 14.48 1.9 2.0 1.1E+17 1.3E+17 -17.42 6.5E+16 13.30 1.1E+17 31.76 7.4E+16 24.13
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4.0 4.2 2.3E+17 2.6E+17 27.64 1.8E+17 21.64 2.3E+17 50.83 1.5E+17 39.61 8.0 8.3 2.9E+17 3.3E+17 31.74 2.2E+17 25.06 2.9E+17 58.55 1.9E+17 46.00 l 6.0 8.6 3.6E+17 4.2E+17 36.13 2.8E+17 28.75 3.6E+17 66.82 2.4E+17 52.91 7.0 8.1 4.4E+17 5.1E+17 40.22 3.4E+17 32.22 4.4E+17 74.57 3.0E+17 59.45 1 8.0 - 9.3 5.1E+17 5.9E+17 43.30 3.9E+17 34.86 5.1E+17 80.41 3.4E+17 64.42
., 8.8 10.0 5.5E+17 6.3E+17 45.03 4.2E+17 36.34 5.5E+17 83.69 3.7E+17 67.22 9.0 10.8 5.8E+17 6.6E+17 46.13 4.4E+17 37.29 5.8E+17 85.79 3.9E+17 69.03 10.0- 11.7 6.4E+17 7.4E+17 48.76 4.9E+17 39.57 6.4E+17 90.80 4.3E+17 73.33 10.3 12.6 6.6E+17 7.6E+17 49.43 5.1E+17 40.15 6.6E+17 92.08 4.4E+17 74.43
-11.0 12.9 7.1E+17 8.1 E+17 51.22 5.5E+17 41.70 7.1 E+17 95.49 4.7E+17 77.38 12.0 14.8 7.9E+17 9.1E c17 54.26 6.1E+17 44.36 7.9E+17 101.31 5.3E+17 82.42 12.3 -18.0 8.2E+17 s.oE+17 55.20 6.3E+17 45.18 8.2E+17 103.10 5.5E+17 83.99 13.0 18.1 88t+17 1.0E+18 57.08 S.8E+17 46.83 8.8E+17 106.70 5.9E+17 87.13 14.2 18.0 9.9E+17 1.1 E+18 60.21 7.6E+17 49.60 9.9E+17 112.72 6.6E+17 92.40 17 19.3 - 1.1 E+18 1.2E+18 62.17 8.2E+17 51.33 1.1 E+18 116.48 7.1 E+17 95.71
~?E .; 20,0 1.1 C+18 1.3E+18 63.23 8.5E+17 52.28 1.1 E+18 118.52 7.4E+17 97.51 20.0 28.9 1.6E+18 1.8E+18 74.43 1.2E+18 62.32 1.6E+18 140.14 1.1 E+18 116.77 23.0 32.0 1.8E+18 2.0E+18 77.74 1.4E+18 65.32 1.8E+18 148.54 1.2E+18 122.55 I
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PNPS Fluence Report Per Ref. 2 (Isodified after Cycle 4)
NOTE:The cumulative fluence is the fluence at Cycle 4 plus the time x flux rate from Ref 2.
' (Flux rates are averages from the GE report)
Azimuthal Lead Factor = 0.93 (Capsule to Vessel) Axial Lead Factor = 1.02 Capsule / Weld flux <cyc s>= 1.27E+09 n/cm 2-sec -
Vessel flux <cycs>= 1.37E+09 n/cm2 -sec Fluence per GE Report (Modified)
Base Metal Wold 1-338(Heat 12000M204) !
Cycle ' Cer,sule Plate @ ID Plats @ 1/4T Wid.deID Weld. ib 1/4T
- (Est.) EFPY Fluence Fluence ARTm Fluence ARTm Fhence ARTm Fluence 1.0 ARTm
- 1.1. 5.5E+16 6.3E+16 10.80 4.2E+16 8.05 5. 5E+16 19.52 3.7E+16 14.48 l 1.9 2.0 1.1E+17 1.3E+17 17.42 8.5E+16 13.30 1. fE+17 31.76 7.4E+16 24.13 i 4.0 4.2 2.3E+17 2.6E+17 27.64 1.77E+17 21.64 _2.3E+17 50.63 1.5E+17 39.61 3 l 8.0 8.3 2.84E+17 3.06E+17 30.21 2.05E+17 23.78 2.9E+17 58.58 1.9E+17 46.03 6.0 6.8 3.36E+17 3.61E+17 33.26 2.43E+17 26.33 3.43E+17 64.54 2.3E+17 51.00 7.0 l 8.1 3.94E+17 4.24E+17 36.36 2.84E+17 28.94 4.02E+17 70.61 2.7E+17 56.11 8.0 9.3 4.4E+17 4.75E+17 38.74 3.19E+17 30.96 4.51E+17 75.27 3E+17 60.05 8.6 l
, _, 10.0 4.7E+17 5.07E+17 40.09 3.40E+17 32.11 4.81E+17 77.92 3.2E+17 61.65 9.0 10.8 4.9E+17 5.27E+17 40.96 3.54E+17 32.86 SE+17 79.64 3.3E+17 63.09 10.0 11.7 5.4E+17 5.79E+17 43.06 3.89E+17 34.65 5.49E+17 83.74 3.6E+17 66.57 10.3 12.0 5.5E+17 5.93E+17 43.60 3.98E+17 35.11 5.63E+17 84.80 3.7E+17 67.47 11.0 12.9 5.9E+17 6.31E+17 45.04 4.24E+17 36.35 5.99E+17 87.62 3.9E+17 69.87 12.0 14.8 6.5E+17 7.00E+17 47.52 4.70E+17 38.49 6.64E+17 92.49 4.4E+17 74.03 12.3 18.0 6.7E+17 7.23E+17 48.29 4.85E+17 39.16 6.85E+17 94.01 4.5E+17 - 75.33 13.0 16.1 7.2E+17 7.69E+17 - 49.84 5.16E+17 40.50 7.3E+17 97.05 4.8E+17 77.95 14.2 18.0 7.9E+17 8.52E+17 52.45 5.72E+17 42.78 8.08E+17 102.18 5.3E+17 82.38 18.0 19.3 8.4E+17 9.08E+17 54.09 6.09E+17 44.21 8.61E+17 105.42 5.7E+17 85.18 18.8 20.0 8.7E+17 9.39E+17 54.99 6.30E+17 45.00 8.9E+17 107.18 5.9E+17 86.71 20.0 28.9 1.2E+18 1.32E+18 64.61 8.87E+17 53.50 1.25E+18 126.14 8.3E+17 103.32 23.0 32.0 1.36E+18 1.46E+18 67.51 9.78E+17 56.09 1.38E+18 131.85 9.1 E+17 108.36
- 18 month cycle through RFO-10 & 24 Mo. cycle from RFO-11 on .
CC.T6 ns time for caosule to reach vessel EOL Fluence 7
Vessel Fluence = [ Peak Fluenco re ,.n + (EFPYp,g- EFPY c,n)*3.156e i (Sec/EFPY)* Peak Flux (n/cm*-sec)]
Capsule Fluence = Vessel Fluence
- Lead Factor (LF)
Conclusion No capsule will reach a fluence value comparable to the vessel wall during the life of the plant, which is currently 32 EFPY, because of the lead factor less than unity. 6 The capsule fluence (from the SWRI report) forms the basis for the Tech. Spec Ammendment withdrawal schedule. However, fluence values based on service time were significantly reduced per the GE Neutron Transport Analysis. Consequently, the fluence based on the GE analysis at 19.23 EFPY would be comparable to the 15 EFPY fluence based on the SWR 1 report.
Page 3/5
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