ML20080F728
| ML20080F728 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 11/08/1994 |
| From: | CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | |
| Shared Package | |
| ML18064A586 | List: |
| References | |
| EA-RDS-94-02, EA-RDS-94-2, NUDOCS 9501300212 | |
| Download: ML20080F728 (78) | |
Text
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@possamur pesar PALISADES NUCLEAR PLANT EA-RDS-94-02 i
-am pumuusu ENGINEERING ANALYSIS COVER SHEET
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retat =er e+ sr. ts 11 Title t.mt oe e p.1,saces cue eet prs vee,,i m c , tee.. u,,-,- j l
INITIATION AND REVIEW l Calculation Status Preliminary Pending Final Superseded a o y .
Initiated (i.f t Review Method Technically Reviewed Reve Rm Appd Appd i
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- Description By Detail Qua l By appo By Date Alt Calc Review Test By Date ,
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9501300212 950123 PDR ADOCK 05000255 P PDR i
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- PALISADES NUCLEAR PLANT EA-RDS-94-02
- w resanns ANALYSIS CONTINUATION SHEET Sheet 2 Rev
- _1_
Table of Contents 1.0 Objective 3 2.0 Summary 3 3.0 Analysis Input 4 3.1 References 4 4.0 Assumptions 5 5.0 Analysis 5 5.1 Values of 'I' and 'M' 5 5.2 Values 5.2.1 of r'
' ART"sa' des 'CF' Values Pali 6
6 5.2.2 Palisades 'f' Values 8 5.3 Palisades PTS Screening Criteria Limits 9 6.0 Conclusions 11 Tables 5.1 Averages of Retired Steam Generator Weld Chemistries. 7 5.2 Best Estimate Cu and Ni Values for Palisades Axial Welds. 7 5.3 Palisades Fluence Values. 9 5.4 Possible Margin Gains. 10 Attachments l
i Attachment 1 Reference 3.1 Section 10 CFR 50.61 l Attachment 2 Reference 3.2 Pages 4.1 to 4.3 !
Attachment 3 Reference 3.3 Page 8-8 ;
Attachment 4 Reference 3.4 Attachment 1 page 8 l Attachment 5 Reference 3.5 All !
Attachment 6 Reference 3.6 Page 6-28 ;
Attachment 7 Reference 3.7 Summary table l l
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- PALISADES NUCLEAR PLANT EA-RDS-94-02
- w ress e ss ANALYSIS CONTINUATION SHEET l Sheet 3 Rev = j_
1.0 Objective l
1 This Engineering Analysis has been written to document calculations done to !
determine Palisades position with respect to the PTS screening criteria. These calculations incorporate the preliminary weld chemistry values obtained from the retired steam generators and the best available fluence data.
2.0 Sumary Calculations have been done to determine the Palisades reactor vessel material condition as it relates to the PTS screening criteria. Based upon the best available fluence values and axial weld chemistries which include the three preliminary copper and nickel weld values from the steam generators, the plant would exceed the 10 CFR 50.61 screening criteria after 210 EFPD's from 24:00 Hrs, October 31, 1994. This works out to a calendar date of May 29, 1995. If Palisades does not take credit for its inhouse fluence calculations, and instead uses cycle 9 fluence rates for cycle 11, the plant would exceed the 10 CFR 50.61 screening criteria after 115 EFPD's.
This gives a calendar date of February 23, 1995, assuming continuous full power operation.
The other part of the data to be collected from the retired steam generator welds is the initial RT,or. This data is not yet available. If the initial RT,or results are equal to or less than the generic value for Palisades axial weld of -
56 F, Palisades will recover a minimum of 10 F on its margin term. This gain would mean that the plant would exceed the 10 CFR 50.61 screening criteria in approximately 4.59 EFPY's.
Although the NRC rule on PTS is based on best estimate fluence and chemistry values, Palisades has not taken credit for the conservative bias of approximately 6%
in its current Westinghouse calculational methodology. Recently Palisades received a Technical Evaluation of its fluence metLdology from the NRC, reference 3.8. In this evaluation the NRC suggests that Palisades current fluence calculations are between 7% and 10% high. If Palisades is able to use the best estimate fluence v& lues submitted in its 6-5-92 submittal, the plant could run for another 735 EFPD's.
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- PALISADES NUCLEAR PLANT EA-RDS-94-02 mseners ressens ANALYSIS CONTINUATION SHEET Sheet 4 Rev a _L 44 3.0 Analysis Input References given in section 3.1 cover the data used in this Engineering Analysis.
3.1 References 3.1 10 CFR 50, current issue.
3.2 6-5-92 NRC Fluence Submittal, Docket 50-255 - Lic. DPR-20, 10CFR50.61 Pressurized Thermal Shock, Revised Projected Values of RT y , for Reactor Beltline Materials.
3.3 6-10-93 NRC Fluence Submittal, Docket 50-255 - Lic. DPR-20, 10CFR50.61 Pressurized Thermal Shock, Reactor Vessel Neutron Fluence, Additional Information.
3.4 2-23-94 NRC Fluence Submittal, Docket 50-255 - Lic. DPR-20, 10CFR50.61 Pressurized Thermal Shock, Revised Information.
3.5 Preliminary Chemistry Data from AEA for Palisades Retired Steam Generators.
3.6 6-21-94 NRC Fluence Submittal, Docket 50-255 - Lic. DPR-20, Palisades Plant, Reactor Vessel Material Surveillance Capsule Test '4eport.
3.1 EA-P-PTS-93-03, NI Detector Adjustment Factors for Cycle 11 Operations, Rev.1 3.8 NRC Fluence Evaluation, Docket 50-255, Palisades Plant, Transmittal of Technical Evaluation Report, 9-2-94.
All attachments relate directly to these references. The relevant pages from the separate references have been copied and included in the attachments so that all necessary information is readily available.
- . tenuuus pseer sausame PALISADES NUCLEAR PLANT EA-RDS-94-02 !
- m passmus ANALYSIS CONTINUATION SHEET Sheet 5 Rev 8 J_
4.0 Assumptions The calculations in this FA are based on the preliminary steam generator weld chemistry values provide by AEA, reference 3.5. All calculated values have been rounded off to three significant digits to be consistent with past submittals.
Projections of EFPD's ar.d EFPY's are based on inhouse fluence calculations for cycle 11 only. This inhouse model has been benchmarked against the Westinghouse fluence methodology and has been validated for use as a scoping tool. Westinghouse will be validating these calculations, however this data will not be available until the end of November. For dates that extend beyond cycle 11 it is important to note that the number of EFPD's or EFPY's may be changed by the fluence rates associated with the later cycles. The weld samples from the retired steam generator are only applicable to, and can only affect, Palisades axial weld chemistries. The 30 weld was and still is the limiting weld. This is the only weld addressed in this analysis. The welds removed from steam generator A contain W5214 weld material.
5.0 Analysis -
10 CFR 50.61 provides the foundation of the PTS screening criteria.
Calculations for the RTpy, are done using equation 1 from the rule.
RTm = I + M + b RTm Eq. 1 ARTm " Irradiation adjustment of RT '
I=RTm ( Inicial RT)
M = Margin term Each of the items in Equation I will be discussed with respect to Palisades current situation.
5.1 Values of 'I' and 'M' Palisades does not have an initial RT, value for its reactor vessel welds.
This forces the plant to use the generic value of -56 F for its axial welds, stated
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,,,,,,, FALISADES NUCLEAR PLANT EA-RDS-94-OR
--s rusases ANALYSIS CONTINUATION SHEET Sheet 6 Rev 8 _L in 10 CFR 50.61 for Linde 0091,1092 and 124 and ARCOS B-5 weld fluxes, reference 3.1. The initial RT,o, is one of the values that the plant intends to get from the l
retired steam generator welds, but has not yet received. l The value of M in Equation 1 is 66*F for welds when the generic value of I is used, and 56*F when a measured value of I is used. This is the 10 F margin term that ;
the plant hoped to recover by measuring a value of initial RT,or from the retired steam generator welds.
5.2 Valuts for ' ARTpt ,' l The value of ARTets is cal.ulated from two factors, CF and f, as shcwn in Equation 2 from 10 CFR 50.61.
ARTm = ( CF) f M ;* * ' M ** D Eq. 2 CF = Chemistry Factor f = Best estimate neutron fluence units of 10" n/cm -
5.2.1 Palisades 'CF' value.
The value of CF for Palisades comes from the table of generic weld CF's provided in a table in 10 CFR 50.61 for plants without credible surveillance data. This table relies on the copper and nickel content of the weld material to determine the CF.
Attachment 4 gives the copper and nickel contents for comparable heat No. W5214 welds other than the steam generator welds which are shown in Attachment 5. Table 5.1 shows the chemistry values for the three 'A' steam generator welds from Attachment 5 and their averages. The samples taken from A steam generator were tandem heat No.
W5214 welds, the B steam generator samples were from heat No. 348009; only the heat No. W5214 values are of interest in this EA, since welds fabricated using weld wire from this heat are limiting. The new data taken for heat No. 34B009 does not change the limiting weld for the Palisades reactor vessel. l
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Negr pasumus PALISADES NUCLEAR PLANT EA-RDS-94-02
--ans mammer ANALYSIS CONTINUATION SHEET Sheet 7 Rev 8 _1_
Weldment 'A' 'A/SG/A' 'A/SG/B' Sample Copper Nickel Copper Nickel Copper Nickel 1 0.341 1.093 0.367 1.154 0.353 1.203 2 0.310 1.003 0.291 1.156 0.233 1.149 3 0.266 1.090 0.278 1.059 0.237 1.024 Average 0.306 1.062 l 0.312 1.123 0.274 1.125 Table 5.1 Averages of Retired Steam Generater Weld Chemistries.
Table 5.2 uses the values from Table 5.1 and Attachment 4 to give all the weld sample values for copper and nickel. It also provides the averages of copper and nickel content for use in determining Palisades reactor vessel axial weld material CF from 10 CFR 50.61. Some of the copper values have been double counted because they were from tandem welds. This is the same averaging technique as used in Reference 3.4.
I.D. Copper I.D. Nickel 04463 IP2 0.20 04494 IP2 0.94
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0.20 04541 1.20 HBR2 Torus 0.159 04577 & 04604 1.00 0.159 04673 Mill IC 1.05 IP2 Sur 0.20 04674 IP2 1.12 IP3 Sur 0.16 04686 ML1 0.97 0.16 04687 IP21 0.92 IP3 Nozzle 0.15 04688 Pal 0.99 HBR2 Sur 0.34 D4690 1.13 0C1 Sur 0.285 HBR2 Torus 0.99 Pal Weldment A 0.306 IP2 Sur 1.03 0.306 IP3 Sur 1.12 PAL A/SG/A 0.312 IP3 Nozzle 1.09 0.312 HBR2 Sur 0.66 PAL A/SG/B 0.274 Pal Weldment A 1.062 0.274 Pal A/SG/A 1.123 Average 0.237 Pal A/SG/B 1.125 Average 1.03 Table 5.2 Best Estimate Cu and N1 Values for Palisades Axial Welds.
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- PALISADES NUCLEAR PLANT EA-RDS-94-02
- -sressmear ANALYSIS CONTINUATION SHEET Sheet 8 Rev v __1_
, The best estimate Cu value for Palisades axial welds is 0.237 and the Ni value is 1.03. These values can be used with Table 1 of 10 CFR 50.61, shown in Attachment 1, to determine a CF for use in calculating the Palisades PTS screening criteria fluence value. Using linear interpolation, as allowed by the rule, the CF -
242.36 F, which rounds to 242 F.
5.2.2 Palisades 'f' Values To date Palisades has only officially submitted fluence values for cycles 1 through 10, Reference 3.3 and 3.6; these values are restated in a more convenient format in Attachment 6. In order to calculate Palisades current accumulated fluence it is necessary to use cycle 10 fluence values from Reference 3.6, and cycle 11 fluence values from Reference 3.7.
Westinghouse analysis shows that the calculational methodology used to create the data shown in the references above has a conservative bias of approximately 6%.
Although the NRC rule on PTS is based on best estimate fluence and chemistry values, Palisades has not taken credit for the conservative bias in its current Westinghouse calculational methodology. Recently Palisades received a technical evaluation of its fluence methodology from the NRC, reference 3.8. In this evaluation the NRC evaluated the current Palisades fluence calculations as between 7% and 10% high. If necessary Palisades may choose in the future to recover this conservatism from its analysis. The best estimate fluence rates from Westinghouse for cycles 1 through 9 are shown in Attachment 3. For cycles 10 and 11 the best estimate fluence rates have been created by dividing the calculated fluence rates by 1.06.
Table 5.3 shows both the calculated and best estimate fluence rates, along with the cycle and cumulative fluence for both. The EFPD's for cycles 1 through 10 shown in Table 5.3 can be found in Attachment 6. For cycle 11 the EFPD's have been calculated as fol!ows. The current burn-up, 7222.1 MWD /MTV, times the MTU of the core, 81.202 MTV, divided by the rated power, 2530 MW, gives 231.8 EFPD's. Since the limiting welds are the welds at the 30* positions, only .luence .ates at these angles are used.
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- PALISADES NUCLEAR PLANT EA-RDS-94-02
-w ressassa ANALYSIS CONTINUATION SHEET Sheet 9 Rev = _1_
Cycle Cycle Fluence Rate Cycle Fluer.ce Cumulative Fluence Number EFPD's Calc's Best Est. Calc's Best Est. Calc's Best Est. I 1 379.4 4.70E10 4.43E10 1.54E18 1.45E18 1.54E18 1.45E18 2 449.1 4.70E10 4.43E10 1.82E18 1.72E18 3.36E18 3.17E18 3 349.5 4.70E10 4.43E10 1.42E18 1.34E18 4.78E18 4.51E18 l 4 327.6 4.70E10 4.43E10 1.33E18 1.25E18 6.llE18 5./6E18 1
i 5 394.6 4.70E10 4.43E10 1.60E18 1.51E18 7.71E18 7.27E18 6 333.4 4.79E10 4.52E10 1.38E18 1.30E18 9.09E18 8.58E18 7 369.9 4.79E10 4.52E10 1.53E18 1.44E18 1.06E19 1.00E19 8 373.6 2.34E10 2.21E10 7.55E17 7.13E17 1.14E19 1.07E19 9 298.5 2.00E10 1.89E10 5.16E17 4.87E17 1.19E19 1.12E19 l 10 356.9 1.94E10 1.83E10 5.98E17 5.64E17 1.25E19 1.18E19 11 231.8 1.66E10 l_57E10 3.33E17 3.14E17 1.28E19 1.21E19 Table 5.3 Palisades Fluence Values.
Using Palisades most up to date fluence calculations, f - 1.28. A best estimate value of, f = 1.21, could be used if the plant can recover the conservative bias in ~
its calculational methodology.
l 5.3 Palisades PTS Screening Criteria Limits Equations 1 and 2 from 10 CFR 50.61 can be solved for f, as shc,wn in Attachment 2, giving Equation 3 shown below.
o.za - go.ons - o.4 lo, ""';* '
Using this 270 F value for RT,rs, -56 F for I, 66 F for M, and 242 F for CF, in Equation 3, gives a screening criteria fluence value of 1.31*10" n/cm2 . This value and Palisades current fluence accumulation can be used to determine the number of EFPD's remaining before the plant reaches the PTS screening criteria. This is shown on the following page.
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- -s passa m PALISADES NUCLEAR PLANT EA-RDS-94-02 ANALYSIS CONTINUATION SHEET Sheet 10 Rev s _L Margin = 1. 31
- 101' n/ cm2 -1. 2 8
- 101' n/ cm2 = 3 . 0
- 1017 Fluence /EFPD = 1.66 =101 n/(cm 2-sec)
- 3600 sec/Hr=24 Hr/ Day F1uence/EFPD = 1.43=1015 n/cm1
- 0*A0 EF20's = = 210 EFPD's 1.43*1015 Table 5.4 shows the number of EFPD's/EFPY's left before the plant reaches the screen 5 criteria using different values of (RT,1, - I - M). This table includes values 4 EFPD's/EFPY's for both calculated and best estimate fluence data. The best estimate fluence rate for cycle 11 is 1.36*10'5 .
Value of Screening Criteria Margin Using Margin Using RTm -I-M Fluence Limit Calculated Fluence Best Est. G uence 260 1.31E19 210 EFPD's 735 EFPD's -
262 1.35E19 490 EFPD's 2.82 EFPY's 264 1.39E19 769 EFPD's 3.62 EFPY's 266 1.43E19 2.87 EFPY's 4.43 EFPY's 268 1.47E19 3.64 EFPY's 5.23 EFPY's 270 1.52E19 4.59 EFPY's 6.24 EFPY's 272 1.57E19 5.55 EFPY's 7.25 EFPY's 274 1.61E19 6.32 EFPY's 8.05 EFPY's 276 1.67E19 7.47 EFPY's 9.26 EFPY's 278 1.72E19 8.42 EFPY's 10.3 EFPY's 280 1.77E19 9.38 EFPY's 11.3 EFPY's Table 5.4 Possible Margin Gains.
I It is important to note two things about Table 5.4. First, the times stated do l not take into account any capacity factor deviation from 100%; outages would add to the number of days or years calculated. Second, the data assumes that all subsequent fluence will be accumulated at cycle 11 fluence rates.
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-- PALISADES NUCLEAR PLANT EA-RDS-94-02 manmemrr resmuss ANALYSIS CONTINUATION SHEET Sheet 11 Rev = __L If cycle 9 fluence rates are used to estimate the cycle 11 fluence, rather than using the inhouse calculations, it can be shown that Palisades has 115 EFPD's left in cycle 11 before reaching the screening criteria.
PTS screening criteria fluence = 1. 31+10 1' End of cycle 10 fluence = 1.25 *10 1' Cycle 9 fluence race = 2. 00 *1010
- 3 6 0 0 *2 4 = 1. 7 3
- 1015 1 1 EFPD's = 1.31*10 ' - 1.25*10 ' = 3 47 EFPD's 1.73*1015 Margin = 347 EFPD's - 232 EFPD's ( thru 10-31-94) = 115 EFPD's 6.0 Conclusion The objective of this EA has been met. Palisades PTS screening criteria mar ~ gin has been calculated using the preliminary and partial chemistry data received from testing done on the retired steam generator welds. The data provided shows that the Palisades reactor vessel weld material would not have reached its PTS screening criteria fluence value. A longer summary is available in section 2.0.
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I Attachment 1
t'Arti 30 0 04.ME5I4C LaCENSING OF PRODUCTION AND UTIUZATION FACILITIES
~
(cf Th? holder of a license author, m.7Le s regarcing core icadf 4 tzmg operation of a productson or utt. D*ms T?e samm .most at ce itzation factitty who destres (1) a ' Jpper and a.ckel contents and te char ge in technical specifications or "c.ence vakes used m the ca:r. 4 n
, s 2) to rnate a change m the factitty or f tr each beltane mater:41. If enese the procedures desertbed in the safety o.n nes differ from those sabm*ed .a
- analysis report or to conduct tests or respcnse to the engmal PTS rL;e n .d
- experiments not desenbed m the .mceY2d by the NRC. p.stificat:en must
- - safety analysts report. which mvolve Le prouded. If the value of RT , for any an unreviewed safety question or a m.tenal m the belthne is protected to change in technical specifications.
eveed the MS screening ct: tenon etoh h urs ant 0 90- m TW h MNN NMM M d 24 for protesten Agnanot peneewsoes Thermsg gerat.ng beerse or the propesed 7 se es Amospeanse erosste ser speem Sheet tvents w at:en date :f a change e te ..; ente prwernton meamwee ser egnesener nussear (a) Defmicons. For the purposes of o s ben eus'ed or te e :
power reestore see nereuel epm this section. ~ 'ne d 'e m .Ia req.est for -:c,d
.seaf a (a) Except se provided in paragraph (1)"ASME Code" means the *nca i has been swbr/..t:e1 ts Ib) of this section, all hghtwater nuclear Amencan Society of Mechanical S'5:5= men' must be s.bm:tted :,
power reactors must meet the fracture Engmurs. Boiler and Pronure Venel De.entter 11 :991. Oeerw.se n.s toughness and matenal serveillance U ' Section !!I. " Rules for the asessment must be si.bmated w.m ee program requirements for the seactor Lonstruction of Nuclear Power Plant rnt updaw of the presson.tempa.re coolant p essure boundary set forth in Components." edition and addenda as " " * ' "'*! N 8Ctef 5'5S'4
- Appendices G and H to this part. specified by I 50.58e. Codes and mammi seemance nput. w 5 us Standards eom tr e effectrse d.te of +,s v.,e (b) Proposed alternettves to the
- desenbed requirements m Appenances t', ..% y g g g g a w twer cumes first TFc,e e m.s Event" means an event of transient m st te updated whenever *
- cre .. .
= and H of this port er persons thevoof "q"n:f. cant change in proiected sa:.es 2f
- may be used where an exemption se pressurtred water reactors (PWRs) causms severe overcooling (thermal "", e up n a aquot fu e dange m granted by the Commission under me npiranon date fu operanon of the
- shock) concurrent with or followed by J 2 significant pressure m the reactor d e pressunsed cermal shd
'MS) screenmg cntenon :s 2"O*F Or (3) Reactor Vessel Beltline" steeris
- the region of the reactor vessel (shell F.tn. gegt and axial weld
' matenalincluding welds heat affected 8 nive4 s. w 300 F fu meumfee..i sones. and plates or forgings) that Q weld mater:als. "
For the purpose of directly surrounds the effective beight ofo e 'h*"'f C"',',",
,y e b "'
the active core and adjacent regions of g calculated as follows. escept as the reactor vessel that are predicted to provided in paragraph 15 (3) of e experience eufficient neutmn radiation ,,c.:on The calculation must be me:e damage to be considered in the selection for e ch weld and plate. or forge; n of the most limiting matenal wtth regard me reactor vessel belthne.
(4) I steens the reference I') l.T.n L RTns = Ireferen.e
. M . AR4, temperature for a reactor veneel ! means the mit:al matenal as defined in the ASME Code, ternperature (RT.n) of the cru r.c..:ed Peregroph N5-2331. RTeer means the ***l measured as defined .n !?.e reference temperatore as ediusted for ASME Code. paragraph NB-:JJ1 the effects of neutron radiathm for the Messured salues must be used if pettod of sum M Mm credible values are asatlable. if not ?e I 'I (5) "RT,,es" means the reference C*'"88'"'""'""**I"'i")'D' 0 ' '
temperehare calculated by the method d ds nt givon in paragraph (b)(2) of this section IJnds 0001.1092 and 124 and ARCOS 3-
,,,for use as a screen >ng entenen.
5 WJ h.
(ii)"M" rneens the margm to be added Pal Reywements. to cover uncertamt:es m the va'aes f
- 11) Far each pressunzad water nuclear initial RTen. copper and nickel power reactor for which an operatmg contents fluence and the cale;ht :r.:
Ivense has been is2ued. the beensee procedures. In Equation 1. M is 6*'F fr
.h..I subm t proiected values of RTm welds and 48"F for base me!alif aere c Gr enctor senet belthne maten4:s by salues of!ere used, and M .s 5e F far g mg salses for the time of submatal welds and 34*F for base metal .f g *e esmrai un d.te of the operstmg measured va!.es ofI are used g Lcense the pro,ected expiration date tf (m) ARTm is the mean veh.e of ie g 4 ch. age m the operatmg bcense has adiustment m reference temperat.re
- w. been requested. arid the projected caused by trradiation and shoid be
$ esp.rauen d te of a renewal term if a calc lated as follows:
req sest for 1. cense renewal has beeri ,
wmitted. The assessment must use the Equa tion l
<.A.Intive procedures given in ,q 2.g ARTm = lCF#",,y"g, n iregraph (b.(:) of this secaon. The ft. net on of copper and nickel centent
..wrer.: must spec fy the bases fi.>r CF is given in table 1 for welds an.* n i ter protection. :nciac':ng the table 2 for base metal (plates a...
50-47 June 30,1993 (reset) l 2
- l
$0.61tb) 50.6t(b) 4
- MWT O o DOWSTIC UCENSING OF PRODUCTION AND UTILIZATION FACILITIES {
forttr.gs).1.inest tr.terpoleurs is permitted. In Tables 1 and 2 "Wt-4 fastE 2 --CNEwSTay FaCToa rea bHn mod using b pmcedum of th;s paragraph are subtect to the opper" and "Wt-4 ruckel" are the best. *^L '# approval of the Director. Office of estimate values for the matenal, which
- '"- % clear Reactor Reguistion when used J will normally be the mesa of the Cooo* as provided m this section.
measured values for a plate er forgtrig or "
t o .oJo'o ac to oo:o ao
power reactor for which the value of wire heat number that matches the 0 - n 20 to ro ao RTm for any metenalin the belthne is entical venel weld. if these values are @ , ---j $ g' 2ci* 20 * *
- 80 proiected to exceed the PTS ecreenmg not available. the upper limiting values o os . 4 a m go. ,o entenon before the expiration date of 2g 3 ,o given m the matenal spectficat2ons to 0 04 .
which the vesul was built may be used. 0 05-
..i 2t asi ael as aer asi a the operstmg 12 cense, or the protected su U v; si. si esptranen data if a change m the license If not available. conservanve eenmates @ ' ;H- asi M j'st} 3J 37l 3[{] tes been equested, or the end of a (mean plus ene standard deviauon) o ne ... ] y 4.. sit se se so si "'*'I'0'" I' *9"'" I based on genene data 8 may be used d o os J rrt sat m sa: sai sai sa renewal has been subr.utted',IIC'"
the luetdication is provided. If none of theu O '0-
---H 44 m es, es, ori er er hcensee shall submit by March 16.1902.
alternatives are available. 0 35% copper ' ~ ~ - - asi aa' 72: 74, n! n n an analysis and schedule for and 104 nickel must be suumed.
(v)'T' means the best settmate l }$" ;""-]Is o,
0 94 . --.c 57 rsi ses soci tosi icei ios implementauon of such flux reduct:en programs as are reasonably practicab!e neutrun fluence. In units of108' n/cm8 " ' S- ei so sei nol us, nri or to avoid exceedmg the frIS ocreening
- 64j 104l ne' 522: its (E greater than 1 MeV). at the clad. base mstal mterface on the inside surface of lll" $ $ @ ,'fi j( l}[; itslll entenon set forth in paragraph (bJt:I of o is - _ _ this secuan. ne schedule for the vesul at the location where the o
--l rei er; tao, i : isoi isc is4 implementanon of flux reduction materialin questien receives the highest eo2 'c ea; 10 4 v25. use in 's4 ies measures may take into account me fluerce for the penod of service in quesuon. llE ." % l$ ',2s, , ;sp 2 y schedule for submittal and ant:e pa'ed o n _ _) esa n y , a , pg , , , ,,, ,, Commission approval of dataded plant.
o as. tooi trit tesi irrt isil iser aos specific analyses, submitted to TAsLE 1.--Cr4wsTny FACTost ros, c 2s soa nas 14e i t ee, 214 demonstrate acceptabie nok at salues of W u.os.*F !k ...
o as .. .. _.
,'Y!
no isa
$ ut h #h ja RTm above the screenirs hmit due to plant moddications. new mformanon or I
- % .g c re ... .. ire isa 'so isrt s'e,l 2ni ne te.4i isil 22
- a i: 24a new analysis techniques.
j ox .! tw vasti te7' 19: nsi 24er rs7 (5) For each pressurtzed water nuc! ear
.- IoloJotomotoao'caoIscoleao e si i4 isi tra see o 32 . power reactor for which the analysie l so! so;i ao sol ao i
o n__ . 13ei i ndi iseiise! teot aost irs;,asil usiaos; aeoi 2's assi me aas. asai a" g required by paragraph (b)(4) of this ao ao aol sol aoso N 0 3' 01... ..., ao ao ao - - - '** e4 iner acet am assi ano N section indicates that no reasonabiy as as 27 ism 'in 2n f" o0 03se.. . 1 22, es, 4tj arl 41. 2rl M tag avat anlaosg aos E practicable Oux from reduction presrsm will en 4i. at g j y y jaa
$ 0 34 .. ~... ; 24; 42. 54, sa' se s4 se ari 2r E {0 35---
oa_
ylhmeamammilva ies, ,
g $ prevent the velus of RTm exceeding the PTS screenmg criter:en -
o os .._ ae' 4e e9 es a se' es ;
c.ae .. _. its iss" assi arr anel assi sit before the expiration date of the '
o ! ~. . $ $ I N N 's!
as sai noi sos ice, ios. ice
' "- - "5 tes 307l 23 3e7l 3esj 38 operstmg license. or the projected e
c on os.. ..-._ , co se e4. i t s,122: 122! 122 esipiration date d a change m the
" (3) To venfy that the values of RT operating license has been requested. cr
! $.. e4 s h $ $$! N $s calculated se required by peregraph the end of a renewal term if a request l
o $2 -
o is
- sai n: ions
- so re, ios, vasi ist; 1724 tre im issi isil ici (b)(2) of this section are bounding values for hcanse renewal has ban submmed.
for the specific roectee veneet licensees the hcensee shall submit a safety o is - ei;, rei ice i42 isai iari ies shall consider plant.epecific informanon analysis to deterraans what. J any.
e4 o j - -._. nai.e. irs lsti aoo that could afect the level of modificenone to eqmpment. systems, oit T ~q gseroi erl EeI Eac a" embrittlement. This information includes and operation are necessary to prevent ois__ .
rei es, 4 i iu, vert assi aso but is not Ilmsted to the res . v vessel potenual failure of the reactor vessel as j o ie. est ioni im ist: istj nofis a moult of postulated m enots J operstmg temperature and e 4 ..eillance
- continued operation beyond the
$ ~ ~ --"-' $ $ @ $ results. Results from the plant.spectfle o af ~
on._. ; ici e nat ist! Ser a or isoi ies l h' ' $ 8*5 so, surveillance program shall be integrated screening entenon is allowed. In the analysia. the licensee may determme ses into the embnttlement estimate if.
c 24 -._ ., i tai eta asei see (i) The plant specific surveillance data reactor vessel matenals properties 144l i4
~
o n - _....; n ol tre 14ei aasl rra has been deemed credible es defined in based on available infonnanon.
o37_ _ j [(h o as _. .._ J iast im isoi ier ti asi
$Q Regulatory Guide 1.se Revtsion 2. and fill D' RTm value changes research neults. and plant purumance deta. and may use probabilistic ftactare o a - . ._ , iasl istt us, is4. istl 222 significantly.s ser as4l see mechanica techniques. This analys:s o ao ist ter. ,34 2ni ist 2so Any mformanon that is believed to eu must be submitted at least 3 > ears
..q.,im isei*r$$e.nel tool ris u iprove the accurecy of the RTm value before the value of RTm is projected to o!$ ] $$$$$$$
ou. ...4 i.e ui ies, aoe rasi aee aos significantly shall be reported to the Director. Office of Nuclear Reactor exceed the Ir!5 screening critenon or by one year after the effecuve cate of tt:s on .. ..; issi ies, ver rial sci tra' ses Regulation. Values of RTm that have acendment, whichever is later.
i o se .qise:itr'iscsiez.s;ars,su (6) After consideration of the J
Use
- n. ..
2 h e) $ l$ $# fe7 $
irit i n, aasi n rt a ui a s 3,,
' C'*asm i= RT '*l==' m coad*'.4 oedcanor a*= *= 'ahm d==nmaad
- heensee's snelysee (inchiding effec *s of proposed corrective actions. if any)
- c- - i trsj ine; aor 23sl uti ru,22o aw w w Srenessmanamun==." submitted in accordance with i f
..u acane nas a pareynen mas of mas ==sma w tieni ..im eue.d me sc=ams enwam Pao' paragraphs (b)(4) and (b)(5) uf this
+ one tran e.acin . . i. enc.i.4 i a. ^'"'""""'*"'"N"""**'"""8 " C h * "* ** " "*Y'"" * **
- wn,.i p.or.c u a me .a., .4 m. **'"'*d"""***'*'"*'*ai by-case basis. approve operation of the
. i .a e naa ens m 4. msw am pened , .. feetitty at values of RTm m ercese of exae en et smann. a.m.- the Ir!5 screenmg entanon. The ih m M M (f9904) MS 1
~'
meii su o Welt:
5BC UCENSING Of PRODUCTION AND UTluZATION FACluTIE
i Commiss.on wdl ccnsidir factors ! dIsigned to perform its functten m a (isl The reh bdity of the onsite significantly affectmg the potent.al for 4 rehable manner and be mdependent emergency ec power sources:
fa.lare of the reactor vessel m reachmg a (from the esistmg reactor tnp system) (ast) De expected frequericy of !oss of a decision. l from sensor output to trie final octuation offsite power and
(?)If the Commiseron concludr s. (device (iv) De probable time needed to pursu.nt to paragraph (b)(6) of this -
restore offeite power, section. that opersbon of the far.hty at 14' Each boihng water reaner tast Nave a standby liquid contrei s. stem (2) The reactor core and associated vah.es of RTm m saceos of the PTS coolant. control. and protection systems, screenhig crttenon cannot be approved ISI.CS) with the capabihty ui m:ecting intu the ra,irtur pressure sessel . includmg station bettenes and any other N on the besaa of the Lacensee's a4:alysee borated w, iter solution at .uch a finw necessary support systema. must '
subrnitted m accordance with provide e afficient capacity and N rate. lesci of boren conc ntrahon .ind capabihty to ensure that the cost is
$ paragraphs (bl(4) and !btf 5) of 'hissection. boren.10 the beenses isotepe shalland ennchtrent. request and cooled and apprognate containment g receise Commission approval pner to accounting for reactor pressure vessel ,
mtegnty is memtamed m the event of a l any operation beyond the critenoo. The v lume. that the resultmg reactivity station blackout for the specified !
request must be based upon c ntro is at least eqmaient to that duration. The capabihty for copir a with l mod.fications to equipment, s) stems. - resultmg fr m miection of 86 gailons per a station blackout of spec. fled duration and operation of the facdity to addition shall be determmed by an approgna'e to those previously proposed m the
%. minute of 13 weignt percent sodiumpentaborate decahydrate cnpmg analysis. Iltihties are especrea to solu, subrmited analyses that would reduce f natural boron.10 isotope abundance into have the basehne assumptions. j the potential for failare of the reactor a 251.mch inside diameter reactor analyses and related mformation used 3 pressure vessel for a given core design.
.eisel due to F'I.S events. or upon in their coping evaluanone availabie for !
urther anal) see based upon new The SLCS and its miection location must NRC review. I
_, Murmtion or impros ed metnodoing) be designed to perform its function in a (b) LJmtrotson of scope. Paragra ph fel rehable m nner. The SLCS mitiation of this secnon does not apply to those
!!c s2 aseverse ente for coeuenon of r%s must be automaisc and must be designed plants bcensed to operate pnor to /uly free = entic eet e treas arte wetmout s. rani to perform its function in a reinable 21.19st if the capabihty to withstand (Arwh events ect hgnt.neser.cootee manner for plants granted a construction station blackout was specifically nuvear pos,r anta, permit after July 26.1964, and for plants addressed m the operstmg hcense (a! Sphiebitry The requarneett f granted a construction permit pnor to roceedmg and was emphcitly approved
- n July 26,1964. that have already been y the NRC.
M(st.wster<oclad see"on app;ynuclest to til c:power imere:al designed and built to melude this (el lmplementation.--(1) Informorson ph.es
_. feature.
ird De'mt-n Far p :pe,ses of th s Submittal. For each hsht.watw cooled sact:or.. Armdpated Transient u cheut g- (5) Each bothng water reactor must nuctw poww plant hcensed to opnate have equipment to trip the reactor on or before /u!r 21.198R the licensee S.um (A fuS) means an ar tN.rtied coolant recirculaims pumpe shall submit the information defined eperational occurrence as defired n Appendia A of th.s part f-alcwert hy the [ automatically under conditions below to the Director of the Office of railure of the reactor trip port:en cf the Indicatsve of an ATWS. This equipment 3 Nuclear Reacw Regulation by Wlu '
pretert'on sytiem speci' lad m General I must be designed to perform its function j 198R For each hght water cooled Des gn Critenon 20 of Appendo A 'i bin a rehable manner. z nuclear power plant hcensed to operate this put '
(61Information sufficient to . efter the effective date of this
- amendment, the licensee shall submit (c) /fego
- remer'ts (1) Each pressurized demonstrate to the Commission the
== ater reactor must have equipment from the information defined below to the adequacy of items m paragraphs (c)(1) Director by 270 days after the date of sensor output to final actuation device, through !c)(5) of this section shall be d th license issustics.
submitted to the Commission as [i) A proposed station blackout k s>at stem.2 diverse from the to automatically reactor mitiate the trip specified in i 50.4. duration to be used in determming auxihary lor emergency) feedwster (d) Implementation. By 180 days after compliance with paragraph [s) of this
- system and mitiate a turbme tnp under the issuance of the QA guidance for secnon. including a lustification for the I conditions mdicative of an ATWS.This non-eafety related componenta. each selection based on the four factors equipment must be designed to perform 3 licensee shall develop and submit to the identified un parayaph (a) of this its function m a reliable manner and be 3 Commissioen. as spectfled la $ 30.4. a section:
independent tfrom sensor output to the , proposed schedule for meetag the (h) A desenption of the procedures I
imal actuation device) from the existmg . swquirements of paragraphs (c)(1) that will be implemented fo. station reactor tnp system. ; thro (c)($) of this section. Each shall blackout events for the duretion (2) Each pressurised water reactor incl an emplanation of the schedule determined in parayeph (c}(1)(i) of this reanufactured by Combustion along with a justification if the schedule section and for recovery therefrom: and Engmeenne or by Babcock and Wilcom calls for final implementation later than (hil A het of modifications to must heve a divesse serem system from the second refuehrig outage after July 28. equipment and associated procedi.res. if the sensor output to s'eterruption of 1304. or the date of issuance of a license any necessary to meet the requireraents power to the contpel rede. This scram authonsms operetion above 8 percent of of pareyeph (a) of this section. for :he system must be dostyled to perform its full power. A final schedule shall then specified station blackout duration function m a rehable manner and be be mutually agreed upon by the determined in paragraph (cW1)(i) of this mdependent from the existmg reactor ;
Commission and licensee, section and a proposed schedule for -
inp system (from sensor output to implementmg the stated modificaSons- 1 interruption of power to the control 9gMe Laosof esabsenseigmment rods! 88"*'*
8 The al'emate i tal Each boihng water reactor must ; (a) /taguiremonas. (ti tack light. ac'2) powerAhernate source (s), asoc socue:
defined n 150 2.
l will constitute acceptable capAihty to have an alternate rod miection ( ARl) ; water.coeied nuclear power plaat !
system that is diverse (from the reactor a licensed to operate must be able to withstand station blackout presided an inp system) from sensor output to the I withstand for a speafled duration and andysis is puformed which fmal actuation device.The ARI system demonstrates that the plant has this recover from a station blackout as must have redundant scram air header ; defined is 430.2. The specified etetten carabihty from onset of the station i enhaust valves.The AR1 must be blackout unni the attemate ac sourcef s1 blackout durative shall be based on the and required shutdown equipment are followmg factore:
started and hned up to operare The 'ime (t) The redundancy of the onsite emergency ac power sources: required for starrup and abgnment of the 50 40 June 30,1993 (retwt) 1
0 Attachment 2 l
l l
l l
j
]
4.0 PROJECTED RT,n I The following describes how the PTS reference temperatures are determined for each of the Palisades reactor vessel beltline I materials and includes projections for when each material will exceed the applicable screening criterion. The results are dependent on the best-estimate values for chemistry and fluence that have been addressed earlier in this report. Additionally this section provides response to NRC concerns as to how surveillance results from Palisades and other reactor vessels could affect the projected RT,,,
values.
4.1 Determination and Proiection of the PTS Reference Temeeratures The base equation for the PTS reference temperature from 10CFR50.61 is:
RTm =I*M+ ARTm (1)
"I" is defined as the initial reference temperature (RT ) of the unirradiated material. "I" values for the Palisades reactor vessel ~
beltline materials are:
Axial Weld I, - -56*F' Generic Value 10CFR50.61 (b)(2)(1) for Welds made Cire Weld I, = -56*F , with Linde 1092 and 124 Fluxes Plate I, = 0*F Value* reported in Reference 6. This represents the lietting plate.
- A less conservative value of I, = -10*F was measured by Battelle Columbus j
Laboratories in 1977 (Reference 39). A value of -5'F was used in C?Co's ,
1986 (Reference 16) and 1991 (Reference 1) PTS submittals. Confirmation of l 5'F could not be found by measurement or calculation.
4-1 1
1 i
"M" is defined as the margin term added to cover uncertainties as in the values of initial RT,,, (Cu and Ni content, fluence and the calculational procedures). Values of "M" for the Palisadr. essel beltline material are:
Axial Weld M, = 66
- F Value specified in 10CFR50.61(b)(2)(ii) for welds if generic values of "I" are used.
Cire Weld M, = 66
- F ,
Plate M, 34*F Value specified for base metal in 10CFR50.61 if measured value of "I" is used
" ART,.," is defined as:
ARTm = ( CF) f i " *
- 18 1" (2)
"CF", the chemistry factor, a function of Cu and Ni content, is derived from Tables 1 and 2 of 10CFR50.61.
In Section 2, the chemistry factors were determined-to be: ~
CF, - 217'F for the axial welds.
CF, = 228'F for the circumferential weld.
CF, = 165'F for the vessel plate material.
"f" is the best-estimate neutron fluence in units of 10" n/cma ,
(E > 1 MeV) at the clad-base metal interface of the vessel. j I
i 4-2 !
l l
4 The limiting fluence is determined by setting RT,,, equal to the screening criteria and solving for f. First, rearranging equations (1) and (2):
RTm = I
- M * ( CF) f ' * * * * *
- 10 Los n
( 0 . 2 8 - 0 .10 log f) log f = log RTm - I-M' s CF ,
0.10 (log f): - 0.28 log f + log m - I-M' =0 r CF ,
Using the quadratic equation to solve for log f:
0.28 (0.28)2 - 4 (0.10) log 8" ~'~
log i = > CF >
2 (0.10)
Because the positive root of the equation provides meaningless results, the equation may be simplified to:
0.28 - 0. 07 84 - 0.4 log "" ~'~
f = 10 exp CF ;
, 0.2 ,
The r.aximum allowed values of RT,,, is defined in 10CFR50.61(b)(2) for each of the Palisades beltline is:
Axial Weld RT,, , = 270'F Circumferential Weld RT,, , - 300*F Plate Material RT,, , = 270'F 4-3
a e
Attachment 3 i
1
)
i 1
l l
1
)
i l
1
l Table 8-4 (Continued)
Palisades Fast Neutron Fluence (E > l.0 MeV) Through Cycle 9 -
At the Reactor Vessel Clad Base Metal Interface Cycle Cycle Cycle Cumulative Length Flux Fluence Fluence Cycle (EFPD) (n/cm 2-s) (n/cm2 ) (n/cmh 30 Degrees ,
1 379.4 4.43E+ 10 1.45E+ 18 1.45E+18 2 449.1 4.43E+ 10 1.72E+18 3.17E+ 18 3 349.5 4.43E+ 10 1.34E+18 4.5 IE+ 18 4 327.6 4.43E+ 10 1.26E+18 5.77E+t8 5 394.6 4.43E+ 10 1.51E+18 7.28E+ 18 6 333.4 4.52E+10 1.30E+18 8.58E+18 7 369.9 4.52E+ 10 1.44E+18 1.00E+ 19 8 373.6 2.21E+10 7. I3E+17 1.07E+ 19 9 298.5 1.89E+10 4.87E+ 17 1.12E+ 19 -
45 Derrees 1 379.4 2.81E+10 9.22E+17 9.22E+ 17 2 449.1 2.81E+10 1.09E+18 2.0lE+ 18 3 349.5 2.81E+10 8.49E+ 17 2.86E+ 18 4 327.6 2.8 t E+ 10 7.96E+17 3.66E+ 18 5 394.6 2.81E+10 9.58E+17 4.62E+18 6 333.4 2.86E+10 8.23E+17 5.44E+ 18 7 369.9 2.86E+10 9.14E+ 17 6.35E+18 8 373.6 1.67E+10 5.39E+17 6.89E+ 18 9 298.5 1.09E+10 2.80E+17 7.17E+ 18 88
e
{
l l
l l Attachment 4
- _ - - - _ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - --a
o Attachment 1 Page 8 of 3 CALCULATION OF THE MEAN COPPER AND NICKEL CONTENT !
0F WELDS FABRICATED USING WELD WIRE FROM HEAT No. W5214 j The following identifications and copper content values are from Table I.l. '
- IP2-flange 1-0428 0.20 0.20 HBR2 - Torus Flange 0.159 l 0.159 IP2 - Surveillance 0.20 IP3
- Surveillance 0.16 0.16 l IP3 - Nozzle Cutout 0.15 HBR2 - Surveillance 0.34 OCl - Surveillance 0.285 i
Total 2.013 2.013 + 10 - 0.201 - Mean Copper Content -
- 2. NICKEL CONTENT Sample Identification Weight % Content 04494 - IP2 1-042 0.94 04541 1.20 Average of D4577 & 04604 1.00 04673 Millstone IC 1.05 04474 IP2 3-0428 1.12 044M 18.1 2-072A 0.97 04 W IP21-042A 0.92 044 5 PAL S/G 5-943 0.99 040N 1.13 H8Rt Torus Flange 0.99 IP2 - Surveillance 1.03 IP3 - Surveillance 1.12 IP3 - Nozzle Cutout 1.09 H8R2 - Surveillance 0.66 Total 14.21 i 14.21 + 14 = 1.015 = Mean Nickel Content l
9 Attachment 5 l
l l
l l
l
MPikon2
,, 'IF. STING OF W11DMIil ALS H)R CI'CO PHFI IMINARY ANALYSIS RESUI.TS ON SFCTIONS TAKliN FROM LARGli WI.I DMlWrS (*A* AND *ll*) AND THFPANS Mcwits of sinc chesnecal anlyses of weldnect.d seniples taken frone satiosas through the two 1.orge wckineenes and four Tecp.ms are giwa in the sabic helow. Altheingh these data are cosisedereal so he erwe anal manerate, final chetLs h.sve > sill so lic pcifossened and as wth the data should be escascal as beseg as a pseisensiasy : asiere inntil all the chaks have been niade.
Mn Cs Cia Mo Ni P Si S V Nsteun thseingh large
$- WcMenesee 'A'
~
Al/1/X l.157 0 0471 0 141 p.502 t ers t 0 010 0 264 00146 0u021 AllI/Y 1.249 0 0419 o llo 0507 I ou1 o olo o 26s o0174 o tus12 7 Al/l/Z l.176 ooll7 o 2M, p 4x7 ItrNI o 01I o 2xx 001x1 o(m2?
Nttien abroingh Large Weldment *ll*
lil/2/X 1249 o (N(N) o 245 o 546 1.21s o 012 o L74 oel59 o tu62x 3 Ill/2/Y I 404 091x1 o189 0517 1 olo 0 012 o lu. ooix2 osm2x 111/2/Z 1 26s n o tx9 o lw. o 540 l irex p ol? o ix? 0o177 in im/x Saison through Trepsin
- AXCdA*
A/ SCDA /21X l.117 otuoi o.47 o sox 1.154 0 011 0 14/ oolls o sm 21 A/SG/A/2/Y 1.120 o oi2x o 291 0 49x I Ise. oofI o 184 o017x o in Jo A/ SCDA /2/Z l.I14 o0111 0 27x 0 49x 1 os9 o p12 o?x4 o01x2 oem2i Socason thecingh Trepant
, *A/SG/It*
-. A/SG/U/3/X l 17 t He407 DEL 0 515 1.20 4 - 0 011 9 246 collo osm21 A/SCd11/3/Y I.102 0 91x9 0 2i t o.s23 1.I49 0 012 0 291 ooI5x o ini24 A/SG/Il/l/Z I . los ocull 0 217 0 519 1.024 0 011 o 10 2 00141 osm2s
, Saison shraingh Trepan
- ll/SG/A*
lt/SCJA/2/X '
l.292 00412 o 195 0.551 1.272 not7 o lat. ooixt o tmi t II/SG/A/2/Y I 214 o0179 0 19% o s50 1 14I o 016 o 202 00170 oim27 h/SCJA/2/Z l 246 o01xx o 21 6 o 544 i Ox o 016 0 2II o 01to otM127
- , Sation iluciagh Tsepan
- ll/SG/B'
.7 II/SCd11/2/X i 271 0 0 L97 o 162 0 541 1.826 o o16 o I9I oo177 o9029
.}', it/SG/II/2/Y I 217 o0402 0 2nx , 0.536 1.I16 o o16 o 104 oo165 00027 ivSGnv2rz i ixx o oi9i o 2ir, osu i wi7 . _-.m..-
o .o._ _i_e, u oois6 n in>2s s.s M
list October 1994 l
O e
Attachment 6 l
l l
1 l
l
t c
TABLE 613 (Cw==d)
CALCULA'IED FLUENCE (E>1.0 MeV) "IEROUGH CYCLE 10 AT THE )RESSURE VESSEL CLAD BASE METAL INTERFACE I Cycle Cycle Length Cycle Flux Cycle Fluence Cunnuative (EPPD) (Wem2 sec) (Wcm2) Fluence (Wcm2) i 30 Dearec 1 379.4 4.70E+10 1.54E+18 1.54E+18 2 449.1 4.70E+10 1.82E+18 3.36E+18 3 349.5 4.70E+10 1.42E+18 4.78E+18 4 '
327.6 4.70E+10 1.33E+18 6.11E+18 5 394.6 4.70E+10 1.60E+18 7.71E+18 6 333.4 4.79E+10 1.38E+18 9.09E+18 7 359.9 4.79E+10 1.53E+18 1.06E+19 8- 373.6 2.34E+10 7.55E+17 1.14E+19 9 298.5 2.00E+10 5.16E+17 1.19E+19 10 356.9 1.94E+10 5.98E+17 1.25E+19 I 45 Dearse 1 379.4 2.98E+10 9.78E+17 9.78E+17 2 449.1 2.98E+10 1.16E+18 2.13E+18 3 349.5 2.98E+10 9.00E+17 3.04E+18 ,
4 327.6 2.98E+10 8.44E+17 3.88E+18 r
5 394.6 2.98E+10 1.02E+18 4.90E+18 '
6 333.4 3.03E+10 8.73E+17 5.77E+18 7 369.9 3.03E+10 9.68E+17 6.74E+18 8 373.6 1.77E+10 5.71E+17 7.31E+18 9 298.5 1.15E+10 2.97E+17 7.61E+18 ,
10 356.9 1.32E+10 4.07E+17 8.0" S +18
!r u
6-28
4 f
Attachment 7 I
s
e ;
l Palisades Cycle Flux Values at Critical Locations I
t Cycle Flux E + 10 Cycle EFPD 0* 16' 30' 45*
1 379.4 4.59 03 4.70 2.98 2 449.1 4.59 o.03 4.70 2.98 3 349.5 4.59 6.03 4.70 2.98 4 327.6 4.59 6.03 4.70 2.98 5 394.6 4.59 6.03 4.70 2.98 6 333.4 4.87 6.25 4.79 3.03 7 369.9 4.87 6.25 4.79 3.03 8 373.6 2.16 4.89 2.34 1.77 9 298.5 2.08 3.06 2.00 1.15 10 356.9 1.51 2.40 1.94 1 mmmmmmmmmmmme mammmmmmmmmme summmmmmmmmmmemummmmmmmmmme ammmm. 32 mmmmmmma 11 422.0 1.42 2.21 1.66 1.09 Values for cycles 1 through 10 are from WCAP14014.
Values for cycle 11 are from Palisades in-house calculations. !'
c h
l i
I
)
1 ENCLOSURE 3 T0 10CFR50.61 PRESSURIZED THERMAL SH0CK - REVISED INFORMATION Consumers Power Company Palisades Plant Docket 50-255 A NON-PROPRIETARY VERSION OF THE CPC NOVEMBER 18, 1994 SUBMITTAL January 23, 1995
- j
A NON-PROPRIETARY VERSION OF THE CPC NOVEMBER 18, 1994 SUBMITTAL Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 DOCKET 50-255 - LICENSE DPR PALISADES FLANT 10CFR50.61 - PRESSURIZED THERMAL SH0CK - ADJITIONAL INFORMATION Consumers Power Comr.my (CPC) submittals dated February 23, 1994, November 8, 1994, ant 'hvember 10, 1994 described our plan to more accurately determine the chemicai and physical properties of the weld materials in the Palisades reactor vessel and the progress we have made. We have implemented a plan, the Palisades Reactor Vessel Integrity Project Plan (PRVIPP), to a point where we have performed chemistry and physical testing on weld material from our retired steam generators.
Our November 8, 1994 letter provided preliminary chemistry results from the testing of steam generator weld material. It also provided Revision 1 of Palisades engineering analysis EA-RDS-94-02 which postulated when the Palisades reactor vessel material would exceed the sr.reening criterion if the preliminary chemistry results were representative of three welds fabricated with weld wire from Heat No. 115214. Our November 10, 1994 letter informed the staff that we: (1) had received preliminary low temperature toughness 'ata from the physical testing material and were suspicious of its credibility for use in determining the initial RTuor of the weld material in the Palisades reactor vessel, and (2) were aware of preliminary information, in regard to the steam generator weld fabrication methodology, that indicated the data from the three welds from each steam generator should be treated as being representative of one weld. Our November 10, 1994 letter also stated that, on or before November 18, 1994, we would make a submittal containing: (1) our analysis of the steam generator weld test data and its effect on the operability of the Palisades reactor vessel, and (2) a description of the actions we plan to take in the near future as we continue to implement the PRVIPP.
This is that submittal.This letter transmits Revision 2 of Palisades engineering analysis EA-RDS-94-02 (Enclosure 1). Revision 2 incorporates steam generator weld material chemistry data into the industry database to estimate that the limiting Palisades reactor vessel material (welds fabricated using wire from Heat No. W5214) will not exceed the screening criterion until January 1999. In reaching this conclusion, the analysis continues to use the generic value of initial RTuor prescribed in 10CFR50.61 for the flux type used in the Palisades reactor vessel. While fracture toughness data obtained using ASTM test standard E 208 showed a higher than originally anticipated NDTT for
2 the steam generator material, subsequent testing and evaluation has lead to the conclusion that the steam generator material test results cannot be considered credible for use in establishing an initial RTum for as fabricated Palisades reactor vessei material. This is because of differences such as material thickness, number of weld passes, post-weld heat treatment, and the effects of thermal aging from having been exposed to a medium high temperature environment for a long period of time. These effects are addressed further in Attachment 9 to Enclosure 1.
The analysis presented in Enclosure 1 shows that the Plant may be operated for an additional 3.16 effective full power years (EFPY) or approximately four calendar years before the limiting Palisades reactor vessel material (Heat No.
W5214) exceeds the screening criterion (January 1999). It will, therefore, be necessary for the Company to submit, within approximately one calendar year, our plan to allow for operation through the end of licensed life. Our short-term actions, to be completed in 1995, will support the development of that plan. These short-term actions will include:
- 1. Independently analyze steam generator weld samples.
- 2. Evaluate performing microstructure analyses of the broken steam generator impact test samples.
- 3. Evaluate performing additional fracture toughness analyses using alternate methodology.
- 4. Before March 1, 1995, submit a request to use a site-specific surveillance plan using the available representative industry data on Heat No. W5214 welds. Preliminary analysis using this data, whicn is subject to staff approval, prnjects the time before the Palisades reactor vessel material will exceed the criterien to approximately seven EFPY, Additionally, we will evaluate heat treating the steam generator weld material samples and incorporating them in this plan.
Long term actions being considered are:
- 1. Using a plant-specific surveillance program.
- 2. Performing reactor vessel weld sampling.
- 3. Establishing methods to better define fluence.
- 4. Utilizing a lower leakage core.
- 5. Installing reactor vessel prestressed bands.
- 6. Performing a Regulatory Guide 1.154 analysis.
- 7. Performing a reactor vessel anneal.
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3 i CONCLUSION The latest calculations, using conservative fluence values, show that the Palisades reactor vessel can operate for 3.16 EFPY before exceeding the 10CFR50.61 screening criterion. This will allow operation at a 75 percent capacity factor uatil January 1999. When a site-specific integrated surveillance plan is approved, these values are expected to be increased by approximately four EFPY. Planned short-term actions, short term-actions being evaluated, and long term actions being considered may further increase the time before exceeding the criterion.
The version of Engineering Analysis EA-RDS-94-02 which is included as ;
Enclosure 1 to this letter contains the same information as that submitted November 18, 1994 and now requested to be withdrawn except that the information on Sheet 8 and in Attachment 4 is no longer considered >
proprietary.
SUMMARY
OF COMMITMENTS
- 1. Before March 1,1995, we will submit a site-specific integrated surveillance plan for staff approval.
- 2. Before March 1,1995 we will submit a plan to further evaluate the weld material from the retired steam generators.
1 Enclosure
ENCLOSURE 1 TO THE NONPROPRIETARY VERSION OF THE CPC NOVEMBER 18, 1994 SUBMITTAL Consumers Power Company Palisades Plant Docket 50-255 ENGINEERING ANALYSIS EA-RDS-94-02, REVISION 2 i
Cassin pysy m aassaars pasanser PALISADES NUCLEAR PLANT EA-RDS-94-02 ENGINEERING ANALYSIS COVER SHEET rotat m er e s eeti :o Title .v. t - -d e r se as e-t 2" c-ee-m :-ce ,9 r-INITIATION AND REVIEW Calculation Status Preliminary Pending Final Superseded O O O O Inttleted Inst Review Method Technically aevr R3v Appd Reviewed Appd CPCo
- Desertption By By Appo g
By Date Calc Review Test By Date Ecss Jim L 0 Original Issue snuggerua n-03-94 sta / eif<er ;;-:3-u 3::
Georgs a O Original issue / / Gerali a u- 3-94 a.3 2no a. .
Ross Jim L 1 Admin Revision snusservo n-07-94 GcP / eiffer u-07-94 3:
(,'c f GHGeek 2 Updated Revision Il-l~l h / g, n r/ fy /
Revision 1 Discussion This revision incorporates administrative coments made as a result of the PRC meeting. None of the calculations or results change in this revision.
Revision 2 Discussion TMs revision incorporates the final chemistry data and discussion of the results of the initial RT., test data. This has resulted in several changes from the two previous versions.
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- PALISADES NUCLEAR PLANT EA-RDS-94-02 mammars sammuur ANALYSIS CONTINUATION SHEET
, Sheet 2 Rev 1 Table of Contents 1.0 Objective 3
2.0 Summary 3.0 Analysis input 3 3
3.1 References 3 4.0 Assumptions 4 1
5.0 Analysis 5 1
5.1 Values of 'I' and 'M' 5 5.2 '
6 Values of 5.2.1 Pali' ART's'a' des 'CF' Values 6 5.2.2 Palisades 'f' Values B 5.3 Palisades PTS Screening Criteria Limits 9 6.0 Conclusions 9
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Tables 5.1 Averages of Retired Steam Generator Wald Chemistries. 7 5.2 Best Estimate Cu and Ni Values for Palisades Axial Welds. 8 Attachments Attachment 1 Reference 3.1 Section 10 CPR 50.61 Attachment 2 Reference 3.2 Pages 4.1 to 4.3 Attachment 3 Reference 3.3 Page 8-8 Attachment 4 Reference 3.4 Attachment 1 page 8 Attachment 5 Reference 3.5 All Attachment 8 Reference 3.6 Page 6-28 Attachment .7 Reference 3.7 Summary table Attachment.Se Reference 3.8 All Attachment 9~ Reference 3.9 All Attachment 10' Sample ID description
, . Cummet Negr
,,,,,,,,, PALISADES NUCLEAR PLANT EA-RDS-g4-02 a n m anns sumummer ANALYSIS CONTINUATION SHEET l Sheet 3 Rev = 1
, 1.0 Objective This Engineering Analysis is written to document calculations which determine l the compliance status of the Palisades reactor vessel weld material in respect to the PTS screening criteria. They incorporate the final weld chemistry values obtained from the retired steam generators and the best available fluence data.
2.0 Summary Calculations have been performed to determine the Palisades reactor vessel material condition as it relates to the PTS screening criteria. Based upon the best
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available fluence values and axial weld chemistries which include the averages of the eighteen copper and nickel weld samples from the steam generators, the plant would exceed the 10 CFR 50.61 screening criteria after 3.16 EFPY's from 24:00 Hrs, October 31, 1994. Assuming a 75% capacity factor this works out to a calendar date of mid January 1999.
- The other part of the data to be collected from the retired steam generator welds was the initial RTer. The data collected from these measurements suggests that the material has been affected by its use in the steam generators and cannot be er for this weld material.
used to provide the initial RT 3.0 Analysis Input References given in section 3.1 cover the data used in this Engineering Analysis.
3.1 References 3.1 10 CFR 50, current issue.
3.2 6-5-92 CPCo Submittal, Docket 50-255 - Lic. DPR-20, 10CFR50.61 Pressurized Thermal Shock, Revised Projected Values of RT,,, for Reactor Beltline Materials.
3.3 6-10-93 CPCo Submittal, Docket 50-255 - Lic. OPR-20,10CFR50.61 Pressurized Thermal Shock, Reactor Vessel Neutron Fluence, Additional Information.
M Muir
,,euens, PALISADES NUCLEAR PLANT mammes sessmer EA-R05-94-02 ANALYSIS CONTINUATION SHEET Sheet _4_ Rev = 1 3.4 2 23-94 CPCo Submittal, Docket 50-255 - Lic. OPR-20. 10CFR50.61 Pressurized Thermal Shock, Revised Information.
3.5 Testing of Weldmetals for CPCo Additional Chemical Analysis. Letter from Dr. G.
Gage, AEA, to John Kneeland, CPCo, November 14, 1994.
3.6 6-21-94 CPCo Submittal, Docket 50-255 - Lic. OPR-20, Palisades Plant. Reactor Vessel Material Surveillance Capsule Test Report.
3.7 EA-P-PTS-93-03, NI Detector Adjustment Factors for Cycle 11 Operations, Rev. 1 3.8 Palisades SG Upper Shell Long Seam Fabrication Technique Letter from Carl J.
1 Gimbrone, ABB, to John Kneeland, CPCo, November 15, 1994.
i 3.9 Server, W.L., Credibility of Using Steam Generator Welds as Surrogates for the '
Palisades Reactor Pressure Vessel Welds.
3.10 NRC Fluence Evaluation, Docket 50-255, Palisades Plant, Transmittal of Technical Evaluation Report. 9-2-94.
All attachments relate directly to these references. The relevant pages from the separate references have been copied and included in the attachments so that all necessary information is readily available.
4.0 Assumptions 4.1 The weld saeples from the retired steam generator are only applicable to, and can only. affect, Palisades axial weld chemistries, because only heat No. W5214 and 348009 weld materials were removed from the steam generators.
4.2 The 30* weld was and still is the limiting weld. This is ths only weld addressed in this analysis.
4.3 The welds removed from steam generator A contain the heat No. W5214 weld material. The welds removed from steam generator B contain the heat No. 348009
EMImm pesar sessums PALISADES NUCLEAR PLANT EA-RDS-94-02 mesmany puumuns ANALYSIS CONTINUATION SHEET Sheet 5 Rev = 1 weld material. The chemistry factor for heat No. 348009 weld material is still lower than the chemistry factor for heat No. W5214 weld material.
4.4 The calculations in this EA are based on integrating the averages of the eighteen steam generator weld chemistry values provide by AEA, reference 3.5.
with the previously available industry data. The samples taken from the steam generator constitute one weld and should be averaged into the industry data as one weld, as supported in reference 3.8.
4.5 The retired steam generator weld material is not capable of providing credible RT,,,, values for the Palisades axial welds, as supported in reference 3.9.
4.6 All calculated values have been rounded off to three significant digits to be consistent with past submittals. This is consistent with the accuracy of.
measured values and those values reported in the regulatory guidance.
5.0 Analysis 10 CFR 50.61 provides the foundation of the PTS screening criteria.
Calculations for the RT,,, are done using equation 1 from the rule.
RTm
M = Margin term
,if Each of the items in Equation I will be discussed with respect to palisades current situation.
5.1 Values of 'I' and 'N'
This means the plant must use the generic value of -56*F for its axial welds, stated in 10 CFR 50.61 for Linde 0091,1092 and 124 and ARCOS B-5 weld fluxes, reference
y 1
y o M Near
,,,,,,,, PALISADES NUCLEAR PLANT muummersausseau EA-R05-94-02 ANALYSIS CONTINUATION SHEET Sheet 6 Rev = 1 3.1. The initial RT,or was one of the values that the plant intended to get from the retired steam generator welds, but analysis of these welds showed that the material l had been affected by its use in the steam generators and could no longer be used to provide initial RT,er, reference 3.9.
The value of M in Equation 1 is specified in 10 CFR 50.61 as 66*F for welds, when the generic value of I is used.
5.2 Values for ' ART,n' The value of ART,n is calculated from two factors CF and f, as shown in Equation 2 from 10 CFR 50.61.
l ARTm * (CF)fG'" ******* Eq. 2 1
CF = Chemiscry Factor f = Best estimate neutron fluence uni ts of 10" n/cin*
5.2.1 Palisades 'CF' value.
The value of CF for Palisades comes from the table of generic weld CF's provided in a table in 10 CFR 50.61 for plants without credible surveillance data. This table ,
relies on the copper and nickel content of the weld material to determine the CF.
Attachment 4 gives the copper and nickel contents for comparable heat No. W5214 welds other than the steam generator welds which are shown in Attachment 5. Explanations of the weld designations are provided in attachment 10. Table 5.1 shows the chemistry va. lues for the three 'A' steam generator welds segments from Attachment 5 and their copper and nickel averages. The samples taken from A steam generator were tandem heat No. W5214 welds, the B steam generator samples were from heat No. 348009:
only the heat No. W5214 values are of interest in this EA, since welds fabricated using weld wire from this heat are limiting. The new data taken for heat No. 348009 does not change the limiting weld for the Palisades reactor vessel.
WWWWI
,,, PALISADES NUCLEAR PLANT EA-R05-94-02 memnury summuns ANALYSIS CONTINUATION SHEET Sheet 7 Rei a _1_
Weldment 'A' 'A/SG/A' 'A/SG/B' Sample Copper Nickel Copper Nickel Cooper Nickel 1 0.341 1.093 0.367 1.154 0.353 1.203 2 0.310 1.003 0.291 1.156 0.233 1.149 3 0.266 1.090 0.278 1.059 0.237 1.024 4 l 0.328 1.116 0.365 1.193 0.359 1.204 5 0.310 1.006 0.292 1.127 0.239 0.960 6 0.266 1.104 0.281 1.066 0.228 1.107 Average Cu 0.297 Average Ni 1.101 Table 5.1 Averages of Retired Steam Generator Weld Chemistries.
Table 5.2 uses the values from Table 5.1 and Attachment 4 to give all the weld sample values for copper and nickel. It also provides the averages of copper and nickel content for ese in determining Palisades reactor ve:sel axial weld material CF from 10 CFR 50.61. Some of the copper values have been double counted because' they were from tandem welds. This is the same averaging technique as'used in Reference
_, 3.4.
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, Cuesmal j Pruer possasus PALISADES NUCLEAR PLANT ma m mrvpassens EA-RDS-94-07.
ANALYSIS CONTINUATION SHEET Sheet 8 Res : 1 l
l 1.0. Copper I.D. Nickel l 04463 IP2 0.20 04494 IP2 0.94 l 0.20 04541 1.20 HBR2 Torus 0.159 04577 & 04604 1.00 0.159 04673 Mill IC 1.05 IP2 Sur 0.20 04674 IP2 1.12 IP2 Sur 0.16 04686 Mll 0.97 0.16 04687 IP21 0.92 IP3 Nozzle 0.15 04688 Pal 0.99 HBR2 Sur 0.34 04690 1.13 OCl Sur 0.285 HBR2 Torus 0.99 Palisades SG 0.297 IP2 Sur 1.03 0.297 IP3 Sur 1.12
~
Average 0.217 IP3 Nozzle 1.09 HBR2 Sur 0.66 Palisades SG 1.101 Average 1.02 Table 5.2 Best Estimate Cu and Ni Values for Palisades Axial Welds.
The best estimate Cu value for Palisades axial welds is 0.217 and the Ni value is 1.02. These values can be used with Table 1 of 10 CFR 50.61, shown in Attachment 1, to determine a CF for use in calculating the Palisades PTS screening criteria fluence value. Using linear interpolation, as allowed by the rule, the CF -
233.54*F, which rounds to 234*F.
5.2.2 Palisades 'f' Values To date, Palisades has only officially submitted fluence values for cycles 1 through 10, Reference 3.3 and 3.6; these values are restated in a more convenient format in Attachment 6. In order to calculate Palisades current accumulated fluence it is necessary to use cycle 10 fluence values from Reference 3.6, and apply cycle 9
{
fluence rates, reference 3.3, to cycle 11. Based on current core design the use of cycle 9 fluence rates for cycle 11 at the 30" weld is approximately 17% conservathe.
, tsuunners pygg ,
mammes paamuser PALISADES NUCLEAR PLANT EA-R05-94-02 ANALYSIS CONTINUATION SHEET
, Sheet 9 Rev =1 Attachment 6 shows that the EOC 10 accumulated fluence at the 30' weld location is 1.25*10 n/cma ,
5.3 Palisades PTS Screening Criteria Limits Equations 1 and 2 from 10 CFR 50.61 can be solved for f, as shown in Attachment 2, giving Equation 3 shown below, o . a - go . on - o . . to, "m y ~
- Eq. (3) f . to o.2 The maximum RTen allowed for Palisades axial welds is 270*F, reference 3.1. l Using this 270*F value for RT,n, -56*F for I, 66*F for M, and 234*F for CF, in Equation 3 gives a screening criteria fluence value of 1.49*10" n/cma . This value ;
and Palisades current fluence accumulation can be used to determine the number of !
EFPD's remaining before the plant reaches the PTS screening criteria. This is shown below, s PTS screening criteria fluence = 1. 49 *10" End of cycle 10 fluence = 1.25 *10" Cycle 9 fluence race = 2. 0 0 *101* *3 6 0 0 *2 4 = 1.7 3 *10" EFPD's = 1 ' #S *10 ~I' 5*10
= 1387 EFPD's 1.73*103 Margin = 1387 EFPD's - 232 EFPD's ( thru 10-31-94) = 1155 EFPD's i
i Using the 75% capacity factor and the 1155 EFPD's gives Palisades 4.21 years l before reaching the PTS screening criteria. This works out to a date sometime in mid January, 1999.
6.0 Conclucion The objective of this EA has been met. Palisades PTS screening criteria margin has been calculated using the chemistry data received from testing performed on the retired steam generator welds. The data provided shows that the Palisades reactor i
Dungul pysy
- PALISADES NUCLEAR PLANT EA-R05-94-02 ;
mammers semanns ANALYSIS CONTINUATION SHEET Sheet 10 Rev = i._
vessel weld material has not exceed the PTS screening criteria. l l
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. ting operatiott of a product.on or utt- : . " * ** s I
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change m tecr'.nical specificat.ons or .Me S a ~cs .sec .e
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$ the procedures seecribee ;n the safety . ..-* es d 'fer ' om '?.ese s.:- . .
! analytts report or to conduct tests or matase 'o tee cr 7.nai PTS . e , :
- experimen;a not descrtbed m the
. . F ed dv t5e NRC. c.st:f :a :- ...
- safety artal)sta re9ert, thach irnche .. ;rouced if tte vMue of RT,., L. , . ,
- an unrettewed safety Question or a
.e'eL41.n .he beltlarte is projecie: 3 change m techrucal specsfications.
ec,q .r.e FT5 screen .g c- *o me t of h c urs a .t 9-
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peser ,esente ser nesses eswam ?* % 'e m !a re;. utfJr u (a) Encept es provided m peregraph (1) "ASME Code' meene the ~f C* 4. *as oeta sut*. re t a lb) of this section, elllaghtweter musies, Amences Socicy of Mechemsel lsschmer.t must o's.b .re: :,
pe ,,seewee mest meet go g,,,,,,, Enginum. Soiler and Preneure Vessel Je'.e m er 3 m t Otter. se e mughneee esud musenal servedienee Code. Section RI. " Rules for the .
regwements fee the reemer Construction of Nuclear Power Plant .geatscimne andre- of si the de pren.:e ntm.".e. ;.. .- e3 coe at pressere beendary set fare in Components." edstion esel addende se - !s or 'Pe neat reactor ,ess..
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- Appendiese C and M to ese part, speesfied by I se.4as. Cadse end .! seeWence repon. as 5 . ., j (b) Propeesd alteenettves to the 7*m'"it ro e effecitwo date..,of "t.: j
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g ,, gg ni, **er camu first T* c e. . J - .i s d"*n m9 womeau m AppendicesC g,,,,,,,,,,,,,,,,,,,,,,,,,, ni,,t3, no,ied wher..n : e .. l
. and H of thw part w partene theseef ,,, ,,g ,g , , ,, ,, , , m y , gn.f. cant change m proiec:c: , a..es .f l
- may be esed when as sesernen m RTm. or upon a reque t for e :r4 4, n '
,,,,,,, ,,,,,, ,,,,,,,ggg, gg,,,,3 gmned by the Commmen edw "to npirshon date for speranen M ?.e l mis - shock) concervent wie er followed by
- ei Y-g ,gadicent
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pressartsed de al st . a
- (3)"Reester Wesel Bottline means DI ' '""'es entenon a 2 0 F '.r ptetu. 'forg nas and on.a. ne.d
- 6e regue W &e rncer voseel(ed j e * * "'I' 300"I I*' ' "* I " ' - i mWwml ,4 meedmg , wo,*lds. bem, g,Neced .e.d mater
- ale. For the purpose of {
- "dF = =* *o s E= d mdsh' d G.'."iw as f o e ae -
- he=*=mecerad **aa"'ec*e*d woesl e m em pm to aecceprovided m paragraph ib::3; af e empovWem h anges meanma demoge to be geno 6dend h te eclectes secton. The calculauon m.st :e 4:e for each weld and plate, sr !: ; : t !
. , of the meet lieuemg meterial with Ngard ,ne reactor vessel beltLne
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wmpwmurefw areseaw weseel til t' means the init.al tefe e 4 ;
WlaWeml se denned is ese ASE Code. ' err'perature (RT ,) of
- tie uru .: . .:
m maten.1 measured as cef:nea y e Puegraph W 335.RT m,,,,g g , eses g, de ASME Code. Paragraph NS-U4
,g,,,,,g,,,,,,, ;
g, ,g,,g g ,,,,,, ,,g,,,, g, g, Mc asured vskee must be useo I '
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kmpermem etnised by es need ya. and -54*F for welds tsee - - !
we la peregraph (bl(8) of tie secum Lade aset.toet and124 and *>O S- i
""""*"""""8"'""- s wohl fiumn.
(n)"M" rneens the mer; s e i . '
l di Ne%%8. erNt8. to Cover uncertairit:es in !? e ; i Dl Far eaen pressunzad weter nuclear e.itial RT n. copper erd ....'
power reactor for which en operating contents. (luence and 'he ca .. s leense has geen ssued. trie liten es procedures. In Equation 1. M e
- F - 3
.h..I sulam.: proiected values of RTm welde and et F for tase me ai i e" St enctor venen belthne rnatene.'e by 6alues ofIare used. and M e e :
- 4 mg sal es Nr 'Se time of subm.ttal, welds and 34*F for osu ne 4'. ! '
- espiratin date of the operatint g *L tense tre pro.ected empiration date if measured valwes ofI are ased I g (m) ARTm is the mean v. .e g a cheage in the operating hcense has 44ustment in reference te-;v . - i
. been rteuested. and the protected s,aused by rrradiation and s .J... -
$ ve>.ts:ica Jefe of a renewat term 5f a c41culated as foilows.
ret;mt for 1. cense renewal has been
@mit'ed The eseessment must use the Caustion 2. ARTm.:CF.f * * - *
. f..!.tne p ocedures given in (a) CF(*F)is the chemis"* . e
- ' er4; graph (L.i.*t of this section. The ft. net: ort of copper and n; cati ..* -
.. %.ter t must specify the bases for Cr ie given in table 1 fer we,:s v ,
!?e protection act.cing the table 2 for base metal';!a'es ** ' ;
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50 47 June 30. '993 eesti i
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PART 50 o DOedESTIC UCENSING CF PROOOCTION AND UTILIZATION FACluTIES
' forgir.gs). I;near mterpolacc$. TAstt 2 -OEws?av Fac ce :Cm D " " S0d M "'*3 'h' PNC'CW M Of ptrmitted. In Tablee 1 and 2 tNs peregraph are sub>ect to trie
"**t",,
cepper" ar.d "Wt.t ruckel" are the best. approval of the Director. Office of
- stimate values for de matenal. which N.ciear Fesctor Rep.ilenen when .see niu normally be the mean of the 'cooer ***'"'**
es prov ded m this section.
measured values for a plate er forpeg or
- o cao o .o o oc : no . 3e , M.. r41 F:r each :renunted water naclear for weld samples made with the weld power resetor for which the value of wtre heat nwnber that matches the 3 - 70 70 20 M M M M RTm for any metectal m the belthrie is 08' enucal vessel weld. lf tese values ere not available the upper hatur.g valaes
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Q Q g g g g to ;rciected ro euceed the m acmenmg entenen before the espiration date of o 23 ,
ao nann 3 n given in te matenal spec:ficatacns to o od .. tri m a agi w get a tr e Operet:rg beenee. or the proteced which the vessel was built may be used. o cs . - asi 3ti sie st si: st si eiet enen data if a char.ge m tes licea se if not available, conservanve estamttes N 3" 3" 1' 3' ras been equested or the end of a (mean plus one standard deviation)
.f,j
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'd '.'. 3]",37l lI ", "" renewal term if e equest for lacerse o se bened on genene data 5 may be used d c os -- . rr sat sei w w w se re .ewal has been submitted. the justdication to provided. If none of these o 'o. ~ 4i . w es. esi eri er er licensee snati subtrut by Mercr:16. :E.
etternatives are available. 0 38% copper 7' an analysis arid schedule for and 10% nickel must be soeumed. $
o is .. *- *-*--]1
$g368,3; 7
7,i egi'y yl [ y' , ,
,, esi treptementauon of euch flu reduct:en (v)"f" means the best estunate o**~ y st tsi sii teos rose ,os, ice progreme se are reesonably proct: cab!e routrun fluence.in uruta of tone n/cm8 4 ' S= j si so6 est not us4 nri nr to avoid exceedmg the m screening (E greater than 1 MeV). at the clad base l* , ", ( 'i,'j "'f l%.. ',N: ',a entenon set forth m paragraph (b){t :f metal mterface on the monde surface of a to.... _,. J ry sei i ,,, , w , . u.,o this section. The schedule for the vessel at the locanon where the e** i rei eri is isar isot tw is, implementanon of flum reducnon matenalin question receives the highest o ao . se ima6 '25. Se sw is in measures may take mto account the fluence for the pened of service in j'E7-o ss .,_ j ,een
- Q, , ,'g;l$ l
,,ri i3e, ,f,;
',s, schedule for submittel and ant:cipated Commise on oppmval of detailed plant, quescoa. ,,,,,,,,,,,,
o se - : inei veil iasi irst teil isei ame specific analyses. submitted to oM demonstrate acceptable nok at salues of Tsen.a 1. -Cnewistuv FACTom rom , 'ig i# tset its inee aseg *354 Waa.os. f' 'r*f . - vie $h o se . _..... sie ise see isrt s'el m,i r.ze a $ M* RTm bove a the screenirs limit due to plant moddications. new miortr.auon er
,,, m o n .... . . ias tes: inei isil asil aRe ae new enalysis techniques. '
e o 30 4 im ine8 sett sedi *n ** 2s7 pl For each pressurtzed water nuclear '
a loJoicesieseleseliselt as op / iu ,se' tr tesi am; assi me power mactor for wh2ch the ana'ysia i i o 32 .- iw issi i sett ass asse tra mquired by paragrepb (b)(4) of tie .
% o .. as act se as as se 8 88-- -- *' tesi isoi aser snel ano c on i' as ao se as se sol as 8 3' - - 1 * *** '**i aus. am assi ase section mdicates that no reasonsoly tar l ate J 3 as as as er af af ar isa ies a* att: aos f preenceble Gum rediscusa progrom wiil o os ~1 as se et ei es eil tr f assei g g ',#,,j -7 , r1,Ja m E 3 prevent the value of RTm from
.,,,,, o 04 . . _ . a4l 42.' sai sal se 54 54 ,3, ,,,, ,,g ,!e'r
,, aan a$ael ng sei va enceeding the m screenmg entenen e os . - . _ act es a ,an) asas arr sne6 ase6 ser before the empiratiott date of to ni og en ,,, t
. $ I es, $j ee' est$E$ ' ' ' - - - '#
'88l ser} mt asqasajaes operstmg license. or the protected E".~f e os CD-
- as m seiioni son,i m ios i ao se sei tis 6 saal is iar expiration date d a change in the operstmg itcense has been requested. cr D) To nnfy tW te wlun M RT j o$Y . .. "e# .h $ 3 $ ,) calculated as required b persymph e end of a mnewal term d a mquest for hesame renewal has been subnutte; o _ - not rai toet issa isei test sei (bM2) of thee escnon are values o ' *a. l se6 foitesisans isnt ital ire 1 for the speseSc rosesor sesent. licensees the hcensee shall subaut a safety analysis to determee what d any.
o 54. .% ed rol semi i4ai ieel inel ies shall consider plesegpeede saferseenos mod $ censes a gmyment eyewme, est ur' tesi irti esel me that could aSoct the leoni of Q "~~~ es and operousa are necessary to prevent o it7 rol e si i No $ embrtttlement. This Isisematles lacludes potential faalute d se mactor vusal as ae ret sel tant itse teri rise ase but is not Hested to the reacter vessel a result of postulated M evente J c it esi isnt tasi ist; 'ei asal see opere temperature and earwiuance o ao se( ioni im i rast see results. esults from the plant opectfle connnued opention beyond tha l li;;-i "$ ;"st ;"l :
on_ a to o ri i.ee 3=
mal see eurnin. ace ,m,.m abau be inio,eied mte the embnttlement estimate if, ocmenms =nien=a is ana-edan ce analysts. es bosee may determe oN snel as (i) The plaat.opectfic surveillance data reactor veneet menettais propertes
$ im tail i*ee i o as ._ __., noi iael i.es assi m has been deemed credib6e se defined m based on available iafonasuon.
- research resulta, and p;aat survedance
![to
~
as.._~
e s U] N.tant s'"' isoi asil ae ase insi i Regulatory Guide 1.se Revtsien 2. and 04 The RTmvehne changeo data. and may use probabilianc fractare on, in ters ten, toit aart anel aar significantly.e mecharuce techruques. This analysta 0 20 ist saa. ier ise: an. nri aos Ariy mformanon that is believed to must be submitted at least 3 years
.3 iss isii i r tie. r m teol its ou trnprove the accuracy of the RTm value before the value of RTm ta protected to lU 7]
o se . . .,teosina tw assi assi assi see d $ , $ $ signtficantly shall be reported to the Director. Office of Nuclear Reactor escoed the m screerung crttenen or b) one year aflat the effective date of tr..s Reguleoes. Values of RTm that how amendment, whichever is later.
o ss . ..-. .j isai isnt tort s'ri nel rrr men o as . tw irri .,i: aves a. , art sca (6) After consadorecon of the
$.T $
e 23..
i '$ Ih I,l$ l l[ $
i ry east assi part n. seet 3,,
i
' Chaas" = af .w m mi ame
- ea.fanns d aest ** 'da essermn=e a bcensee e snelysee (tachdme e!'ec s cf propeesd correcove actions,if any) o es ' trs{ ines ar; aaq ast; aes. aac eermea Mi 'h= "'""* ** ** *"* " subourted m accordance with
" 8'""*
. '** wn **'.'a'**'.8
. * *"'e*8,*%
e no en wa.a. er.".e Par 8FeP a and (bNS) of this i.+,no .eu. en .ncivame sectort. the r'a=='aa- may, on a case-
'D a * " * " " d * ***'*** *** . . a we .. .I em.uo si*' " e*a' by.csoe taois. approve operet:en of "e !
mai n.4 es.unuaien a o"n '".se a um fec"hty at values og g7m m eiceeo ef I
. . e ,s. e ,., e .. a. ;
no.ien.as e = es c the m ocesenmg crttarton. The I
.hme sei 1883(resen w-de !
i
rass w e ivnic,a.iw Acrainu wr renuscis0N AteO uTillZATION FACILITIES i
Co .m ss on mal consider 'acters .i des:gr.ed to perform its function .n a hi; The r,necihty;f rne enig, s,ga..rcanity affecting me potent.al fur
- rebaele rnanner and be independent emergency oc power scurces:
ta...re of trie reactor vessel a reser ..g
- ifrom the esishes reactor tr p system) (uil The expected freci ency of :t, .f a sec;s.on. <
. from sensor output to the fmal actaanon offsite power and (at !! 9e Coen:seien conc!. des. I,.d'nce livi ne probable time needed to
- urs nt to paragraph (b)(81 wi tras - restore offeite power.
sec:.on inat oparanen of the far.ht, at 41 Escr. St. . air w. er ea ce si (2) The reactor core er.d esecc:ew wan.es of RT,,,in excese of tne PTS " * * ' * '" D 4* 'd t w t i i s a.n coolant control and protechen eyatrats.
sc eecar g cr"eaon cannot be approved 'S!CS! a m .he capumiy un r:ect.e4 meludme station bettenee and any orrier I 'I* necessary support eystems. must N si on 2.e 04us of the latensee's analysee '"'"
borated *' '"' D"o"n w der soh.i 'a t o c n a IIn w provide sufficient especify and N pe.brrittee m eccordance
- sgrapas (5';4i ord $1!$1 of wiih to rate. letti of boren cent. eMation .tnd capabihty to ensure that the core is f sect.on. te hcenset srall request and bando isotope ennentrent. and cooled and appropnate contairimeest p eteae Comm:uton approval pr:or to 'CC8"ntir g for reactor pressure vessel integnty s maintained m ime event of a aa) operanon beyond the critenos. The ##' " t 9tstion blackout for the spec:fied con el s at . 4ast he resulting reactivity q isa;ent to that req.est must be based upnn duration The cepebihty for copreg ..t.i mod.ficatacos to equipment. systems. - resultirg from iniecuan of 44 gallons per e station blackout of spec.fied Orancai and operation of the facdity is addition l mmum U wettht percent sudium shell be deterrruned by en approgra ,
io thoes previsualy proposed in the pentaborate decahydrete solutiott at the coping snelysis. Utihties are especies .o submitted analyses that would reduce f naturalboron.10 isotope abundance mto have the beoehne assumptions.
tv pounnat for failure of me rucw 3 apressure 251.mch atteide diameter reactor vessel for a given core design.
enelyses, and related mformenon . sed m their copmg evaluanona evadecie 'cr
.viset das to f'!.S events, or upon
'urther anal)ses based upon new The SLCS and its iniection location must NRC review.
es designed to perform its function in e (b) L4metation of scope. peregrach ':t
,,, 4rmiion or improved meinesdoing) rehalele menner. The SLCS mitianon of this section does not appiy to 'hese 1 to 42 a sseuse seenes ser resussen et een rnust be automanc and raust be designed plante bcensed to operate pnor to /..y trem si.tes seeae tram erse esensus e. ram to pufonn its funcuan att a mheble 22. IsEE if the capetsihty to withstand ta rws; evenes %,, none.sese <eeeeee manner for plante granted a construction stenen blackout was specifically nucies, , , gn,,, permit after fuly 28.1984, and for plante addressed in the operettag beense tal t;/recFiery The requernents c,f granted a construction permit pnor to roceeding and wee emplicitly approved i Iuly 28.190e. that have already been y the NRC.
n.s sect on appiy to $11 c: ,mercial designed and built to melude this hr.t.wster.cocied nuclear power pl.i.cs (c) ImpletrentetsonH 11 In/or-etron
_ feature. Submitsel. For each hght.weeer cooled d,i fe'.nir+n For perp.ses of eb s nuclear power plaat hoensed to cperate sec't:,r.. An'icpated Transient Weh,ut g- (5) Each bothng weier reactor must have equipment to tnp the roecto, on or before fu!yit, tant the hcensee S.*et's i A f% 3) meene an er.ne.rceJ c ,g,,, ,,,,,,,i,,mg p,,,, shall submit the informenon def!ried eperanonal occurrence se defir'ei :n summatically under conditions below to the Director of the Office of Aprendia A of th:s part fr.J! owe <t by the E indiceit.e of an ATws. This equipment ; Nuclear Reacter Regulanon by W .'r faih.re of the reactor trip pert:en cf the t preteetmn syttom spect 8Ied in General I smet be dmened m perform m functm 2 !sAR nuclear For each poww p tIpea.weter heensed to cooled opereto Des gn Critenos 20 of Appendit A uf bin a reheble manner' . after the effecove date of this this part ' (6)Infonnation sufficient to ; amendment. the hcensee shall submit I (c) Aequirementa. (1) Each pronounsed demonstrate to the Comnussion the the safennenen denned below to 'ho t wster reactor must have equipment from adequacy of items is persysphs (cM1) Detector by 270 days after the date of I sensor output to final actuation device, throudi(CHI) of this section shall be license isomence.
d that is diverse from the reactor trip I submitted to the Commission es til A proposed etsuon blackout k system. to automatically intuate the speciAed in 9 30.4. duratics te be used m determming a sumihary (or emefgencyl feedwater (d)18tplementassen By 180 days shee compilance with pareysph lel of *e
- eystem and iniuste a turb ne in, under the leemance of the QA gendamse fee secuen. includmg a sueutication for 'he I condshone mdacauve of an ATWS.Thee non esfety related compemente, eneb selecnos beesd on the four factore equipment must be designed to perferen j 16 ceases shall develop and embed to the identined a pareysph (a) of thie ate funcnon in a rehable manner sad be g Commisseen. as spec Aest in i BE4 a secuam independent (from senser output to the g proposed schedule for Insetang the (n) Ai Jz of theprocedures final actuanon device) from the emasang ., regenrements of parepophe (cNI) that will be unplemehui . Menon reactor tnp system. 71 (cms) of this secuen. sech ohna blachest event .or the duration (2) Each pressustead weser reacter inc! an emplanaties of the schedele detenenned in paragraph (cH1)fil of this enenufactured by Combassem along with a tuotificanon if the schednis esenen and for recovery therefrom and Engmeenne er by these end Wiless calle for final implementanen later than lin) A liet of modincations to enuit have e devesesseem system tree the second refuelmg outage aAer July an, equipment and sesociated proceci. n :f tre sensor outpos es of 1994, or the date of issuance of a licones any, necessary to meet the require ems power to the ceased sah eerem authensang operetien ebeve S poroset of of parayeph (el of this seenon. Sr te ee stem must be dengued to perform its full power. A final schedule shou dien specified statoes blackout duration f.,nction m a reheble meaner and be be mutually agreed open by the determined in pareysph (cHtlli) of %e independens from the emissing reactoe , Commission and beenees. section. end a proposed scheduie for trip system (fror's sensor output to implemennt's the eteted modificatcas l interruption of pcmer to the control IOLS laae et as sensummig eunes '
8""
(2) Alternere oc source:Theeverae rodel ac power sourcetsi. as defined in 1 W L IJ) Each beihng wate, reactor must ; (el AeWesitement (11 Each lighb ,,11 conentute acceptable capatA'v o hewe an alternate red intection ( ARI) g weser. cooled nuclear power piesW withstand efetion blechout penmee an system that is diverse llrom the reactor a hconsed to operate meet be able to
- analysse is performed which
, inp sy stetrl from sensor output to the withstand fee a spessned dureties and demoneerstes that the plant has + e i final actuation device. The ARI system ; recover hoes a stenen bleekset as carabihty from onset of the statien I must have redundant scram air header defleed in i SS 3. The specified etetten '
blackout uani the alterviste ac swv' eahoust valves. The ARI must be l blechout duration shall be based on the and toquared shutdown equier-r s'e '
following factore: started and lined up to operste ** * *r (i) The redundancy of the onente required for startup and slig"er" 8 "e emergency ec power sources:
50-40 June M 1993 (reset) ;
r r -
8 4
e Attachment 2 W
5
~ -
4.0 PROJECTED RT,,,
The following describes how the PTS reference temperatures are determined for each of the Palisades reactor vessel beltline materials and includes projections for when each material will exceed the applicable screening criterion. The results are dependent on tae best estimate values for chemistry and fluence that have been addressed earlier in this report. Additionally this section prov ues response to NRC concerns as to how surveillance results from Palisades and other reactor vessels could affect the projected RT,,,
values.
4.1 Determination and Proinction of the PTS Reference Temneratures The base equation for the PTS reference temperature from 10CFR50.61 is:
RTm " I
- N
- ARTm (1)
"I" is defined as the initial reference temperature (RT,) of the
, unirradiated material. "I" values for the Palisades reactor vessel '
beltline materials are:
Axial Weld I, = -54
- F ' Generic Value 10CFR50.61 (b)(2)(1) for Welds made Cire Wald I, = - 56
- F , with Linde 1092 and 124 Fluxes Plate I, = 0*F Value* reported in Reference 6. This represents the limiting plate.
- A less conservative value of I, = -10*F was asasured by Battelle Columbus Laboratories in 1977 (Reference 39). A value of -5'F was used in CPCo's 1986 (Reference 16) and 1991 (Reference 1) PTS submittals. Confirmation of 5'F could not be found by measurement or calculation.
4-1
e.
"M",is defined as the margin term added to cover uncertainties as 'e the values of initial RT,,, (Cu and Ni content, fluence and the calculational procedures). Values of "M" for the Palisades vessel beltline material are:
Axial Weld M, 66' F Value specified in 10CFR50.61(b)(2)(ii) for el:s if generic values of "!" are r used. '
Cire Weld M, 66 ' F ,
Plate M, 34'F Value specified for base metal '
in 10CFR50.61 if measured value of "!" is used "ARin ," is defined as:
A R Tm * ( CE) f ' ' ' ' ' ' ' ' * * * * (2)
"CF", the chemistry factor. 4 function of Cu and Ni content, is derived from Tables 1 and 2 of 10CFR50.61.
In Section 2, the chemistry factors were determined to be:
CF, 217'F for the axial welds.
I CF, - 228'F for the circumferential weld.
CF, = 165'F for the vessel plate material.
1 "f" is the best-estimate neutron fluence in units of 10"' n/cm8 (E > 1 MeV) at the clad base metal interface of the vessel.
I l
42
I i
'. l l
The limiting fluence is determined by setting RT,,, equal to tne !
screening criteria and solving for f. First, rearranging equatt:ns l
(1) and (2):
l RTm = I - N - ( ct) f o . a s . 13 :., si l
l (0. 28 - 0.10 log f) log f = log Um ~ I'M I CF s i I
0.10 (log f)3 - 0.28 log f + log N ~ I" . o r CT 1 Using the quadratic equation to solve for log f:
- ~I~"
0.20 m
\ (0.20)8 - 4 (0.10) log 888 log f = ' CF >
~
2 (0.10)
- l 1
Because the positive root of the equation provides meaningless '
results, the equation may be simplified to:
0.28 - 0. 07 e 4 - 0. 4 log #*** ~ # ~ "
t = to exp $ ' C# 2 0.2 ,
The samisus allowed values of RTm is defined in 10CFR50.61(b)(2) for each of the Palisades beltline is:
Axial Weld RT,, 270'F Circumferential Weld RT,, , = 300'F Plate Material RT,, ,= 270*F i
43
l
)
l 1
l 1
i Attachment 3 1
I l
1 1
1 l
l I
l
Thie 5-4 (Continued 1 Palisades Fast Neutron Fluence (E > 1.0 MeV) Through Cycle 9 !
At the Reactor Vessel Clad Base Metal Interface Cycle Cycle Cycle Cumulative Length Flux Fluence Fluence M (EFPD) und;rd (n/cm2) In/cmh 30 Derrees i 379.4 4.43E+ 10 1.45E+ 18 1.45E+ 18 .
2 449.1 4.43E+ 10 1.72E+18 3.17E+ 18 3 349.5 4.43E+ 10 1.34E+18 4.5 IE + 18 4 327.6 4.43E+ 10 1.26E+18 5.77E+18 5 394.6 4.43E+10 1.51E+18 7.28E+18 6 333.4 4.52E+ 10 1.30E+18 8.58E+18 7 369.9 4.52E+10 1.44E+18 1.00E+19 s 373.6 2.21E+10 7.13E+17 1.07E+ 19 .
9 298.5 1.89E+10 4.87E+17 1.12E+ 19 45 Demes 1 379.4 2.8 IE+10 9.22E+17 9.22E+17 2 449.1 2.8IE+10 1.09E+18 2.01E+18 3 349.5 2.81E+10 8.49E+17 2.86E+18 4 327.6 2.8 IE+10 7.96E+17 3.66E+18 5 394.6 2.8 tE+10 9.58E+17 4.62E+18 6 333.4 2.86E+10 8.23E+17 5.44E+ 18 7 369.9 2.86E+10 9.14E+17 6.35E+18 8 373.6 1.67E+10 5.39E+17 6.89E+18 9 298.5 1.09E+10 2.80E+17 7.17E+ 18 88 I
. . l Attachment 4 e
I
Att4Chn967 I Page 8 of 3 CAI.CULATION OF THE MEAN COPPER AND NICKEL CONTENT OF WEtDS FABRICATED USING WELD WIRE FROM HEAT No. W5214 The following identifications and copper content values are from Table 1.1.
l I 1. COPPER CONTENT Sawle identification weight y. Copper 04463 - IP2 flange 10428 0.20
l
. a 0.20 0.159 HBRZ - Torus Flange '
0.159 l
l IPZ - Surveillance 0.20 IP3 Surveillance ,
0.16 l
,. 0.16 IP3 Nozzle Cutout . 0.15 H8RZ - Surveillance 0.34 OCl - Surve*!)ance 0.285 Total 2.013 2.013 + 10 = 0.201 - Mean Copper Content
~
- 2. NICKEL. CONTENT Sample identification Weight % Content 0.94 04494 - IPZ 1 042 1.20 04541 Average of D4577 & 04604 1.00 1.05 04673 Millstone IC 1.12 D4474 IPt 3-04t8 0.97 04606 ML1 2 07tA 0.92 D4087 IPfl-04tA 0.99 04400 PAL 5/G 5 943 1.13 04400 0.99 HWit Toru's Flange 1.03 1Pt - Surveillance 1.12 IP3 - Survalliance 1.09 IP3 - Nottle Cutout 0.66 H6At - Surveillance Total 14.21 14.21 + 14 = 1.015 Mean Nickel Content I
9 l
1 l
Attachment 5 1 1
1 m
.~_
l AEA Techselegy Facsimile Dese it- November 1994 To John Kneeland '
Company Consumers Power Conranny Focalmile number (9) 0101616 764 8196 From Dr. G. Gage i
Address Materials Performance Department: B388. Harwell Laboratory.
Didcot. Oxfordshire. OX11 ORA, Uruted Kingdom Telephone 235 434466 Feesimile number 235 432337 Page 1of3 Copies: Neil Irvine , AEA O'DONNELL: (9) 0101412 655 2928 -
MPD/082 e2474 .
Messegr. TESTING OF WELDMETALS FOR CPCO ADDITIONAL CRDUCAL ANALYSIS Attached is a table givir.g the values for the copper and nickel conesats of the repeat, second set of analyses. These values are the average vaines of the three ' _
t perfonned on each sample. They have also basa nonnalised appropnasely based as analyes results for the standards, however they have not been verified by the secnos manager.
I To aid compensoa l have also provided the data from the first set of samples alongade, 1 Also astsched is a copy of the far that I have received from TWI concenung visual cuaesmansa of the drasase surfaces ofdrop weight specimens; AQl, AQ2, AJ1 and BI2.
Untuesmessly I wiB be out of the ordse tornorrow. Tuesday 15/11/94, and hence not !
cousssante. As such t would propose makag a start on testing of the ressaamaag three besetsref Qasyy spennens (Weld 8; transverse, Weld A: longismhed and Weld 8:
loadsenmal) se Wednesday. I will try to conanct you before doing so is order to confirm that essh sesang is is keepirq with your wishes; Richard Miller indicated that you would be revwwing your reqarements today (Mondsy).
If MEREAREAW!n0EDG WittlTRA'GNI5!0N0f Till!TAWMilE!1AfERmG TQ Na ll!4 w
l resa enous nnem terrt>stree it 81 te . :
- t
_. , _ _ _ _ .._ -= -- - -- I
l-N 1 I
5
! ! ! I l
{ {
a I I f E
s I" I" lig I } lil l 11 I i ll ! ll !!! !iei! 1l llli11111 1llll I t n!! i ni in ni ni in ! !
l'
!e !!! !!! !!! !!! !!! !!! il i 'l%
d !!! EEE IEI
!.!.I. ... IE.E.
. !.!.!. ... ... g
~
li I i i ! ! ! ! !}
g 1
ili m li m li lli li lll li lll}illih[i 3
Ig I
i* gi a] o!
388 8 dnCWD 3WA13rWJ LtrEt>9EttwB e! et te. II<>t
- 4- 5 " -- -
. ;2:--
.,. , c. :2;; 23. 3, M TELEFAX TWI Abington He:1. Abington. Cambndge CB16AL '.!K Tempnone .a4 (01223 891182 Telex 81183 Teletax -44 (0)223 892668
. Or Gareth Gage Fax Room Ref: TF/ IWir To:
company: AEA Technoiogy From: Mr K Ben Meterate pettmance Dept Structural intagnty Oept:
Tem: Herwell. Didcot Date: 14 Novemoor 1994 Country: UK Dept Ref: KBAe/60.94 Fax No: 0236 432337 No of pages: 1 of 1 Pteese tauphone Faa Room on 0223 691162 Ext. 2220 if ooges are not received or are ancieer MESSAGE Pellini Testing of Steam Generater Wolds TWI Project $20741 Deer Gareth.
The four spec 6 mens (AEA refs AQ1. AQ2. AJ2. and St2, TWi refs W01-06. WQ109. WC104. sod %C2 04 respece.gey) have Deen heat bnted at 200*C for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. coced i. sing HQuad N . Droken open and enemined under a low powered tMnocular microscope (-x10).
All four seeomene ned been extenervely fractured dunng tne Pedini test (more than 80% of the frecture surfeos wee tinted)
None of the speamens snow any groes wouling defects which trugnt have influenced the results of the seese. One speamen (812) has e email pioner discononuity (- 3 x 0 Smm) on the frecture face 5mm sub-
- surfees Two specimeno (AJ2 and 882) showed some evidence of sheer lip formanen, up to imm w6de, on the top (tensen) surfeos of the speamen.
I am ar enging for all of the speelmene, broeien and untroken (escudog WQ2 01, which Richard Miller took wim hem efter tne day of teollng), to be retumed to you et AEA HarweL seet regerne.
Yours sancerely, senior Gnomeer.
s artgeIETY OtpARTEIElff NQ7d IYdCMddRN d M N N U E M M N
- 3 dwro Twic m accresnea ei et a . tt >i
e Attachment 6 I
l l
1 j
l 1
I
l TABI.E 613 (CW=d) o CALCULATED FLUENCE (E>l.0 MeV) THROUGH CYCLE 10 AT THE PRESSURE VESSEL CLAD BASE METAL INTERFACE 9
Cycle Cycle Len8th Cycle Flux Cyr . Fluence Cunnlative (EPPD) (Wem2 sec) (Wem2) Fluence .
(Wem2) 30 Deme 1 379.4 4.70E+10 1.54E+18 1.54E+18 2 449.1 4.70E+10 1.82E+18 3.36E+18 3 349.5 4.70E+10 1.42E+18 4.78E+18 4 327.6 4.70E+10 1.33E+18 6.1IE+18 5 394.6 4.70E+10 1.60E+18 7.71E+18 6 333.4 4.EE+10 1.3618 9.09E+18 7 369.9 4.79E+10 1.53E+18 1.06E+19 l 8 373.6 2.34E+10 7.5217 1.14E+19 1
9 298.5 2.00E+10 5.16E+17 1.1 m 19 10 356.9 1.94E+10 5.98E+17 1.25E+19 45 Desus 1 379.4 2.98E+10 9.785+17 9.78E+17 2 449.1 2.98E+10 1.165+18 2.13E+18 3 349.5 2.98E+10 9.0m17 3.04E+18 4 327.6 2.98E+10 8/.4E+17 3.88E+18 5 394.6 2.9ml0 1.02E+18 4.90E+18 6 333.4 3.03E+10 8.73E+17 5.77E+18 7 369.9 3.03E+10 9.685+17 6.74E+18 8 373.6 1.77E+10 - 5.71E+17 7.31E+18 9 298J 1.15E+10 2.97E+17 7.61E+18 10 356.9 1.32E+10 4.0m17 8.02E+18 6-28
0 Attachment 7 l
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t Palisades Cycle Flux Values at critical Locations ,
Cycle Flux E + 10 !
Cycle EFPD O' 16' 30' I
45' t 1 379.4 4.59 6.03 4.70 2.98 2 449.1 4.59 6.03 4.70 2.98 3 349.5 4.59 6.03 4.70 2.98 4 327.6 4.59 6.03 4.70 2.98 5 394.6 4.59 6.03 4.70 2.98 6 333.4 4.87 6.25 4.79 3.03 7 369.9 4.87 6.25 4.79 3.03 8 373.6 2.16 4.89 2.34 1.77 9 298.5 2.08 3.06 2.00 1.15 10 356.9 1.51 2.40 1.94 1 mummmmmmmmet mammmmmmmmmmmmmmm mmmmmmmmmmmmmmmmmmm mmmmmmm.32 mmmmm 11 422.0 1.42 2.21 1.66 1.09 Values for cycles 1 through 10 are from WCAP14014. l values for cycle 11 are from Palisades in house calculations.
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Attachment 8 W
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J%E9ED l l
i November 15, 1994 l RVG 94-089 Mr. John Kneele.nd Consumers Power Company Palisades Nuclear Plant 27780 Blue Star Memorial Highway Covert, MI 49043
Subject:
Palisades SG Upper Shell f.ong Scam Fabrication Technique <
Reference:
ABB Letter P PENG-94-022, Chemical Analysis of SG Weld Seam Samples and Materials Consultation (ABB CENO Proposal No. 1017-840-019-A) dated November 7,1994.
Dear Mr. Kneeland:
~
In support of your efforts to analyze and evaluate seseral steam generator (SG) long scam welds . _JB CE is providing material testing and consulting services on the Palisades RPV integrity. These services are described in the referenced letter. At a site meeting, ABB CE was requested to provide additional steam generator fabrication information.
As requested, ABP has reviewed the fabrication data for the original Palisades steam generators. This review was focused on the welding sequence of the long seam welds of the upper shell of the steam generators. From this effort we were able to determine that the original Palisades SGs' three long seam welds in the upper shell were made using a sequential weld method. Therefore these three seams should be considered as one weld with respect to ch-istry. ' Ibis meth.xi is discussed brie 0y below.
In order to minimim distortion and weld shrinkage, a process was used that increments!!y welded eachof the upper shelllong seams in sequence. These shell plates were machined with double U preps. TN upper shell was fabricated in a horizontal position (i.e., lying on its side). After alignment of these plates, weld deposit was then performed using an automated welding process on the OD weld. The automatic welling machine was run the entire length of the seam. Slag was manually chipped from the upper surface of the entire length of the weld bead. The automatic welding machine was repositioned at the far end of the weld seam and a second pass of weld material was deposited along the entire weld seam.
This process was continued imtil the initial OD weld deposit was approximately 1 1/2 inches thick. The shell was then rotated so that the second upper shell plate weld seam was in position. Weld deposit was then performed on the seam OD using the automatic welding process as described for the Grst seam. Upon completion of the second weld seam ABB Combustion Engineering Nuclear Power cm,%.-.,: % e n . ., m, ,- ,m m ,w
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procedure, the shell was rctated so that the third seam was in position. Weld deposit was performed on the third seam as described for seams one and two. The seam welding process then shifted to the ID of the weld joint. After backgrooving to sound weld metal and-performir:g a magnetic particle examination, the first weld pass was deposited on the ID of ;
the first weld joint using the automatic _ welding equipment. As with the OD welds, slag was removed and subsequent passes werc; performed until approximately one inch of weld was deposited on the ID of the first weld seam. The shell was then rotated so that the second weld seam was in position. This seam was also built up like the aforementioned ID weld.
The process was repeated for the third v eld seam. A second set of weld increments was I performed sequentially on each of the three ID weld seams. The second set of weld passes completed each of the ID weld seams. Upon completion of the ID weld seams, the welding process was shifted to the OD of the weld seams. Again the second increment of the OD ,
welding process was performed in a sequential fashion. The second increment of the OD j weld completed the OD weld. Weld wire was fed into the automatic welding machine from l
150 pound spools. Additional wire was added to the weld machine as necessary during the l
~ entire welding process. l i
Approximately six 150 pound spools of weld wire were required to fabricate the three upper ;
shell long seams. The welding sequence described above results in mixing of a portion of ;
each of the weld wire spools in each of the seams. Therefore effects of spool to spool :
variation in the weld deposits should be the same for each of the three seams. The chemical :
analysis results from the three welds should be averaged as a single datum point for a weld )
produced by this fabrication process. -
This welc!!ng sequence has been described in this letter to aid Consumers Power Company in its evaluation of the weld deposit chemistry. This description of the welding procedure and j the actual welding procedure are held by ABB Combustion Engineering as proprietary information. This document contains proprietary information and is not to be transmitted or reproduced without specific written approval from Combustion Engineering, Inc. Consistent j with the requirements of 10 CFR 2.790, transmittal of any proprietary information provided herein to the Nuclear Regulatory Commission must be accompanied by an affidavit from combustica Engineering, Inc.
If you have any questions regarding this letter, please call me at (203) 285-2567.
Sincerely.
COMBUSTION ENGINEERING, INC.
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Carl J. Gi rone Supervisor, Reactor Vessel Integrity
e Attachment 9 l
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3 CREDIBILITY OF USING STEAM GENERATOR WELDS AS SURROGATES FOR THE PALISADES REACTOR PRESSURE VESSEL WELDS An Independent Technical Opinion by W. L. Server, ATI Consulting j
In developing the proper materials to use as surrogates for the Palisades reactor pressure l vessel (RPV) beltline welds since no archive wcld materials exist (except for the Palisades l
surveillance weld), the welds in the retired steam generators were selected as candidates. The i pedigree of the welds in the steam generators was determined by ABB-CE, and some of the l welds in the steam generator wore found to be the same wcld wire heats as in the RPV.
Therefore, material was removed from the steam generators in order to further determine the l adequacy of the materials as surrogates for measuring copper nickel chemistry and use in a supplemental surveillance program. The two weld wires heats of concern were W5214 and 34B009. The following discussion provides details on the similarity and di1Terences between the RPV welds and the steam generator wolds as detennined comparing fabricanon information and measured chemistry and mechanical property data. Note that some of the comparison of chemistry and mechanical property data involves data from othat sister RPVs ..
that have the same weld wire heat in the beltline region and/or in their surveillance program: i H. B. Robinson Unit 2 (HBR), Indian Point 2 (IP2), Indian Point 3 (IP3), Salem Unit 1
- (SI), and two BWRa, Millstonc Unit 1 (MI) and Oyster Creek (OC).
ne fabrication Information for the steam generator welds is very similar to that of the Palisades RPV in that the same weld wire heats from the same manufacturer were used with the same flux type (Linde 1092), welding procedure, and approximate time frame in the same CE shop. (These items of similarity are generally truc for the sister RPV welds also.) ne dissimilarity issues come in nlative to difl'arences in the flux lot numbers, post weld heat treatment conditions, and number of wold arcs (related to vcssol wall thickness). The effect of different flux lot numbers should be inconsequential, but the effects of post-weld hest treatment conditions and thickness can be important. Table 1 illustrates these differences as compiled by ABB-CE. The differences appear to be minor, but there are differences.
Related to posMueld heat treatment is the question of extended time at service temperature and pressure. De steem generators were in service for approximately 8 EFPY at a nominal temperature of $007 under a pressure of about 700 psi. The effect of service agmg on these materials needs careful consideration in determining equivalency with the RPV welds. It is l known that thermal aging effects can arise at temperatures near or greater than 600'F for some femti': steels, but extended time at slightly lower temperatures also could play a j
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significant role. The results from mechanical testing should reveal any differences due to
.ve =&w 17.1PN 1 4
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., : ".1 inservice aging on the weld materials. It should be noted that Charpy V notch testing of the Palisades surveillance weld after thermal aging revealed a fairly signuicant difference reflecting time at temperature (5 EFPY at 5357) as shown m Figure 1. The Palisades survnllance weld was fabricated using a different weld wire heat (3277) but with equivalent flux type and welding procedures that closely matched or bounded the W5214 weld in the RPV; in fact, the surveillance weld baseline curve is very similar to that of the reported baseline curves for weld wire W5214 from IP2. IP3, and HBR as shown in Figure 2.
The pieces of steam generator welds were recently tested to determine copper and nickel chemistry, nil-ductility transition temperature (NDTT or NDT temperature), RTur, .tnd Charpy V notch transition cunes. There are no known NDTT measurements for the W5214 welds, and a value of 80T for the 34B009 has been reported. The measured values for the stcam generator materials are 20T for weld wire heat W5214 and -507 for weld wire heat 34B009. The 307 difference between the reported value for 34B009 and the measured value here suggests a potential aging effect. Other issues are also important relative to the drop weight NDT temperature determination: the brittle weld starting bead was fabricated using the latest ASTM E208 specification which was different from that used in the late 1960s through mid 1980s; for some materials, this different starting weld bend can result in difrerent measures of NDTT. Additionally, there appears to be a possible effect for the W5214 material relative to where the NDTT specimens were taken from the weld thickness which is somcwhat supported by looking at some of the higher energy Chupy specia:en resui:s (i.e., -
onc region of the weld with the highest copper level tends to provide higher icvels of toughness possibly indicative of a lower NDTT). The RT,1 value for the stcam generator W5214 weld is equivalent to the NDTT since the Charpy V-notch data supports greater than 50 ft lb and 35 mils lateral expansion at NDTT + 607.
The Charpy V-notch data also confirms a potential aging effect for the steam generator materials. The measured Charpy energy data for the steam generator W5214 weld metal is shown in Figurc 3 where a comparison is made with tbc combined baseline data from IF2, IP3, and I!BR in the same manner as shown in Figum 2. There is a definite shift in the 30 and 50 ft lb transition temperatures and a drop in upper shelf energy, ne shift difrerence is approximately 30T and the decrease in upper shelfis about 10 ft-lb. These differences strongly suggest a thermal aging effect for the steam generator W5214 material. This difference would also sugscst the inaccuracy of using the measured NDTT from the steam generator as appued to the Palisades RPV, even though the measured value of 207 is in the upper range ofItaown NDTT measurements for CE fabricated welds. Charpy V-notch testing of the similar steams generator weld Heat No. 34D009 is now underway, and a comparison of these results with the original (i.e., unitradiated and unagod) material will provide further evidence of this aging effect.
The aging phenomena have been observed in other vessel wcld materials as is described in a recent (unpublished) paper [Ref.1] containing data from the Doel I and II pressure vessels.
Charpy V-notch data from the Doel I reactor vessel wcld in the unaged and aged condition is shown in Figure 4 which clearly indicates the effects of aging on shift in the Charpy cune.
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The use of the ' steam generator matenal to add to the data base relative to copper nickel '
chemistry should be adequate since the bulk measureLents are unaffected by hett treatment or service aging considerations. The specific data from the chemistry measurements are discussed elsewhere, but the key results show good agreement between prior measurements for the 34B009 weld, but higher than expected copper values for the W5214 weld. These higher values for weld wire heat W5214 have been factor 4 into an average for all of the measurements for W5214 and applied to the Palisades RP weld.
In summary, the Charpy V-notch and NDTT values from the steam generator welds appear to exhibit a service aging phenomenon which invalidate their use directly as measures of the virgin mechanical properties for the subject welds. It may be possible to thermally anneal these steam generator materials to restore the properties back to their equivalent virgin condition, and this approach should bc pursued in combination with microstructural characterization work to better understand the embrittlement process. The success of achieving equivalent virgin mechanical properties will be used in further assessing the steam generator materials as appropriate surrogates for the supplemental surveillance program for the Palisades RPV.
Referease
- 1. Gerard, R., Fabry, A., Van de Velde, L, Puzzolants, J. L., Verstrepen, A., Van !
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Ransbecck, T., and Van Walle, E., "In-Service Embrittlement of the Pressure Vessel Welds at the Doel I and II Nuclear Power Plants," Effects of Radiation on MaWala 17th Tataenanianal Symoosium. ASTM STP 12XX, David S. Gelles, Randy ,
K. Nanstad, Arvind S. Kumar, and Edward E. Little, Editors, American Society for Testing and Materials, Philadelphia,1995 (currently under editorial review).
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l Table I l I
Palisades Vessel Beltune and Secum Generator Weld Fabrication Details Wald Seara 2112 A/C 3112 A/C 1-961 A/C 1951 AiC 14casion of Int. Shell lat. Shell Steam Gen. Steam Gen.
l Wald Long. Seuns Long. Seams No.1 No.2 Wold Wire Hess W5214 +Ni 200 W5214 +Ni200 W5214 +Ni200 34B009+Ni200 Nunsers 348009+Ni200 Max Type Linde 1092 1.inds 1092 Linde 1092 Ltade 1092 i flus 14e No.(s) 3617 3692,3617 3617 3708 Thickasus Ga.) 8.5 8.5 4.75 4.75 FWNT (Boers at 14.75 14 17.4 9.5 Temp > 1109 F) >
i
Palisades Surveillance Weld Data 7,.rt tic renees cor,. r<ttine e.wiin, versieri 2.c priet.e et 11:12:12 on 11 15 1994 Material: WELD SA302BM Capsules Heat No.:3277 orientation:TL Q Curve #1
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Material: Linde 1092 SAW Capsule: Heat No.:W5214 Crientation:
PCurve #1 IP2, IP3 ISR2 Data v Curve #2 Palisades S/G .
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. y Attachment 10 m
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, eams A and B A Steam Generator Seam
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