ML18064A585

Forwards Rev 1 to Calculation EA-RDS-94-02, Evaluation of Palisades Current PTS Screening Criteria Margin

ML18064A585
Person / Time
Site:	Palisades
Issue date:	01/23/1995
From:	Haas K CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
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ML18064A586	List: ML18064A585 ML20080F728
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NUDOCS 9501300210
Download: ML18064A585 (86)

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{{#Wiki_filter:consumers Power POWERING MICHIGAN'S PIUlliRE55 Palisades Nuclear Plant: 27780 Blue Star Memorial Highway, Covert, Ml 49043 January 23, 1995 Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 DOCKET 50-255 - LICENSE DPR PALISADES PLANT 10CFR50.61 PRESSURIZED THERMAL SHOCK - REVISED INFORMATION KurtM. Haas Plant Safety and Licensing Director This submittal of revised information is being made in response to the NRC letter dated December 22, 1994. That letter notified Consumers Power Company (CPC) that the Combustion Engineering (CE) proprietary information in our November 8, 1994 and November 18, 1994 submittals would be released to the public unless, within 30 days, we requested withdrawal of that information and the request was approved. In a subsequent telephone conference call between the NRC, CE and CPC it was agreed to implement a sequence of action in which: (1) ABB/CE would release the proprietary information contained in our November 8, 1994 and November 18, 1994 submittals to the public; (2) CPC would revise those submittals and submit them in a non-proprietary format containing the same information as the originals but without any indications that some of the information had been proprietary; and (3) CPC would request that the original versions of those submittals be withdrawn from the docket. CPC, by this letter, and in conformance with the release of information as described in Combustion Engineering letter P-PENG-95-003, dated January 11, 1994, from Peter Leombruni to Kurt M. Haas (Enclosure 1), hereby submits versions of our November 8, 1994 and November 18, 1994 submittals in a non-proprietary format (Enclosures 2 and 3) and requests withdrawal of our original November 8, 1994 and November 18, 1994 submittals from the docket.

. ~ Revisions 1 and 2 of Palisades Engineering Analysis EA-RDS-94-02 which were submitted in our November 8, 1994 and November 18, 1994 submittals have since been superseded by Revision 3 of EA-RDS-94-02 which was submitted on January 13, 1995. They are here submitted to support completeness of the publicly available record on the docket after the originals are withdrawn.

SUMMARY

OF COMMITMENTS This letter contains no new commitments. However, commitments Nos. 1 and 2 listed below which were made in our November 18, 1994 submittal will remain active.

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. Kurt M. Haas Plant Safety and Licensing Director CC Administrator, Region III, USNRC NRC Resident Inspector - Palisades Enclosures \\_ l 2

ENCLOSURE 1 TO 10CFR50.61 PRESSURIZED THERMAL SHOCK - REVISED INFORMATION Consumers Power Company Palisades Plant Docket 50-255 ABB/CE LETTER P-PENG-95-003 January 23, 1995

Mr. Kurt M. Haas Plant Safety and Licensing Director Consumers Power Company Palisades Plant 27780 Blue star Memorial Highway Covert, MI __ 49043 January 11, 1994 P-PENG-95-003

SUBJECT:

Request to Withdraw Proprietary Information

Reference:

Letter, W.T. Russell (NRC) to K.M. Haas (CPCo), dated December 22, 1994

Dear Mr. Haas:

In response to the Reference, ABB Combustion Engineering Nuclear Operations (ABB CENO) requests that Consumers Power Company withdraw the proprietary information contained in

• Attachment 1 to the November 8, 1994, application and Enclosure 1 to the November 18, 1994, application referred to in the Reference. As discussed in the January 10, 1994,

. telephone call between the NRC, Consumers Power Company, and ABB CENO, we are. providing the enclosed replacement documents. These replacements ~re identical to the original submittals except the proprietary markings are absent. The NRC may destroy the originally submitted documents rather than return them to us.

• ABB CENO has decided to release the particular information pertinent to the Palisades reactor vessel to fully support, in a timely fashion, Consumers Power Company's efforts to substantiate the structural integrity of the vessel. In your transmittal to the NRC, we request that Consumers Power Company notify the NRC that ABB CENO continues to assert that our weld chemistry and fabrication process information and copyrighted data base is proprietary to Combustion Engineering and should be protected from public disclosure by the NRC. We will be requesting a meeting with the NRC outside of your docket to discuss a means to meet legitimate NRC requirements while providing for proper protection of our proprietary documents. If you have any questions regarding this matter; please contact me.

Sincerely, COMBUSTION ENGINEERING, INC. w~~r-Peter Leombruni Project Director

Enclosure:

As Stated ABB Combustion* Engineering Nuclear Power* Combustion Engineering, Inc. P.O. Box 500 1000 Prospect Hill Road Windsor. Connecticut 06095-0500 Telephone (203) 688* 1911 Fax (203) 285-9512 Telex 99297 COMBEN WSOR

ENCLOSURE 2 TO 10CFR50.61 PRESSURIZED THERMAL SHOCK - REVISED INFORMATION Consumers Power Company Palisades Plant Docket 50-255 A NON-PROPRIETARY VERSION OF THE CPC NOVEMBER 8, 1994 SUBMITTAL January 23, 1995

A NON-PROPRIETARY VERSION OF THE CPC NOVEMBER 8, 1994 SUBMITTAL Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 DOCKET 50-255 - LICENSE DPR PALISADES PLANT 10CFR50.61 PRESSURIZED THERMAL SHOCK - PRELIMINARY INFORMATION This letter is being submitted to make the staff aware there is a possibility that the Palisades Reactor Vessel material will exceed the 10CFR50.61 screening criterion earlier than the currently estimated date in the year 2004 (Reference the Palisades February 23, 1994 submittal and the NRC SER dated July 12, 1994). That possibility is based on preliminary and partial test data recently obtained from weld samples removed from the retired Palisades steam generators. The steam generator welds were made using the same weld materials (wire from Heat No. W5214 with Linde 1092 flux and wire from Heat No. 34B009 with Linde 124 flux) and procedures in the same shop in the same time period as welds made in the Palisades Reactor Vessel. The recently acquired preliminary and partial test data shows higher copper and nickel content than the existing best estimate copper and nickel content for welds fabricated using weld wire from Heat No. W5214. If these preliminary chemistry results were assumed to be credible and were added to the industry data base for welds fabricated using weld wire from Heat No. W5214 (which is more limiting than Heat No. 34B009), they would raise the best estimate copper content of Heat No. W5214 welds to 0.237 weight percent copper. The existing best estimate 0.201 weight percent copper was used to derive the year 2004 estimate. Using that postulated value of 0.237 weight percent copper as the best estimate copper content for Heat No. W5214 and current fluence (current fluence is obtained using methodology similar to that evaluated as conservative by the NRC), Palisades Engineering Analysis EA-RDS-94-02 (Attachment 1) shows that the Palisades Reactor Vessel would exceed the 10CFR50.61 screening criterion 115 effective full power days (EFPD) after October 31, 1994. As described by the Palisades February 23, 1994 submittal, the Palisades reactor vessel project plan is to use the chemistry results in conjunction with the measured initial RTNoT (IRT-NDT) values that are being obtained from drop weight and Charpy tests on the steam generator weld material. Both chemistry and IRT-NDT data are being obtained as part of the ongoing test program from which the preliminary copper and nickel content results were obtained. It is expected that the resulting measured IRT-NDT values wi 11 be lower than the 10CFR50.61 prescribed generic value of -56°F. A measured value of IRT-NDT equal to or lower than -56°F, in combination with the reduced margin prescribed by 10CFR50. 61 for use with a measured I RT-Non wi 11 offset the postulated reactor life reduction caused if the preliminary higher copper and

2 nickel content values become final. Postulating a measured IRT-NDT equal to or less than minus 56°F adds at least 3.64 effective full power years (EFPY) after October 31, 1994 before the screening criterion would be exceeded. Thus the presently postulated worst case prediction using only the preliminary chemistry results may be greatly improved by the IR~NDT results from tests scheduled to be performed November 9, 10 and 11, 1994. Even without consideration of the effect of measured IRT-NDT values, the time remaining before the Palisades Reactor Vessel material will exceed the 10CFR50.61 criterion is increased from 115 EFPD to 210 EFPD after October 31, 1994 if the fluence for cycle 11 (current full cycle) is used. This is achieved by using in-house fluence calculation methodology rather than using the cycle 9 fluence addition rate (which was used to calculate cycle 11 fluence value that resulted in 115 EFPD). It should be recognized that Palisades has made a transition to a low-leakage core design which has resulted in successively lower fluence rates since Cycle 8. Use of the Cycle 11 core design and in-house fluence calculation methodology is being independently reviewed by Westinghouse. The results are expected by November 30, 1994. In view of the above, the Palisades Reactor Vessel material would not exceed the screening criterion even if the preliminary chemistry data from the steam generator welds are postulated as final and added to the data base to derive best estimate chemistry for Heat No. W5214 welds. This gives reasonable assurance of safety for at least 115 EFPD of operation after October 31, 1994. The screening criterion would not be exceeded for at least 210 days after October 31, 1994 if the preliminary chemistry values are postulated as final and we also use current fluence as determined by Palisades in-house fluence calculation methodology. Palisades engineering analysis EA-RDS-94-02 has been evaluated under the requirements of 10CFR50.59 and that evaluation has been reviewed by the Palisades Plant Review Committee (PRC). The version of EA-RDS-94-02 included as Attachment 1 contains the same information as that submitted November 8, 1994 and now requested to be withdrawn except that the information on Sheet 7 and in Attachment 4 is no longer considered proprietary.

SUMMARY

OF COMMITMENTS This letter contains no commitments. Attachment

ATTACHMENT I TO THE NON-PROPRIETARY VERSION OF THE CPC NOVEMBER 8, 1994 SUBMITTAL Consumers Power Company Pali sades Pl ant Docket 50-255 PALISADES ENGINEERING ANALYSIS EA-RDS-94-02, Revision I

PALISADES NUCLEAR PLANT ENGINEERING ANALYSIS COVER SHEET EA-RDS-94-02 Total Ml.IOOer _*of Sheets _l_l_ Title Evaluation of Palisades Current PTS Screening Criteria Margin INITIATION AND REVIEW Calculation Status Preliminary Pending Final Superseded 0 0 l( 0 Initiated In it Review Method Technically Reviewed Revr Rev Appd Appd CPCo I Description -By Detail Qual By Appd Bv Date Alt Cale Review Test Bv Date ~1' > l\\*f>1

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J.L* e::rfv II I 0 3/ q 'I lbt-f 0 Original Issue x

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1 -IL,, ;?- le... I~ L.,. s J 9501300212 950123 PDR ADOCK 05000255 P PDR

M*!Mtln' dd!lilllSS PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET. EA-RDS-94-02

• Sheet _2_ Rev # _l_

Table of Contents 1.0 Objective 2.0 Summary 3.0 Analysis Input 3.1 References 4.0 Assumptiol)s 5.0 Analysis 5.1 Values of 'I' and 'M' 5.2 Values of 'ARTPrs' 5.2.1 Palisades 'CF' Values 5.2.2 Palisades 'f' Values 5.3 Palisades PTS Screening Criteria Limits 6.0 Conclusions 5.1 5.2 5.3 5.4 Attachment Attachment Attachment Attachment Attachment Attachment Attachment Tables Av~rages of Retired Steam Generator Weld Chemistries. Best Estimate Cu and Ni Values for Palisades Axial Welds. Palisades Fluence Values. Possible Margin Gains. Attachments 1 Reference 3.1 Section 10 CFR 50.61 2 Reference 3.2 Pages 4.1 to 4.3 3 Reference 3.3 Page 8-8 4 Reference 3.4 Attachment 1 page 8 5 Reference 3.5 All 6 Reference 3.6 Page 6-28 7 Reference 3.7 Summary table 3 3 4 4 5 5 5 6 6 8 9 11 7 7 9 10

1.0 Objective . PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET EA-RDS-94-02

• Sheet _3_ Rev # _1_

This Engineering Analysis has been written to document calculations done to determine Palisades position with ~espect 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 Summary 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 cycre 11, the plant would exceed the 10 CFR 50.61 screening criteria after 115 EFPD's. r' 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 RTNor _This data is _not yet available.' If the initial RTNDT 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 methodology 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 values submitted in its 6-5-92 submittal1-the plant could run for another 735 EFPD's.

MCllE'VS I HSllUS 3.0 Analysis Input PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET EA-RDS-94-02 Sheet _4_ Rev # _l_ 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 ~ Lie. DPR-20, 10CFR50.61 Pressurized Thermal Shock, Revised Projected Values of RTPrs for Reactor Beltline Materials. 3.3 *6-10-93 NRC Fluence Submittal, Docket 50-255 - Lie. DPR-20, 10CFR50.61 Pressurized Thermal Shock, Reactor Vessel Neutron Fluence, Additional Information. 3.4 2-23-94 NRC Fluence Submittal, Docket 50-255 - Lie. 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 - Lie.. DPR-20, Pali.sades 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 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 inforniation is readily available..

..... I/I'S,,_ 4.0 Assumptions PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET EA-RDS-94-02 Sheet _5_ Rev # _l_ The calculations in this EA are based on the preliminary steam generator weid 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 and 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, hriwever this data *will not be available until the end of November. For dat~s that extend beyond cycle If 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 r~tired 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 analysi~. 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 RTPTs are done using equation 1 from the rule. RTPTS =*I + M + 4 RTPTS Eq. 1 4RTPTs = Irradiation adjustment of RT I = RT NDT ( Initial RT) M = Margin term Each of the items in Equation 1 will be discussed with respect to Palisades current situation. 5.1 Values of 'I' and 'M~ Palisades does not have an initial RTNDT value for its reactor vessel welds. This forces the plant to use the generic value of -56°F for its axial welds, stated

PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET EA-RDS-94*02 Sheet _6_ Rev # _I_ in 10 CFR 50.61 for Linde 0091, 1092 and 124 and ARCOS 8-5 weld fluxes, reference 3.1. The initial RTNDT is one of the values that the plant intends to get from the retired steam generator welds, but has not yet received.- 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 RTNDT from the retired steam generator welds. 5.2 Values for 'ARTPrs' The value of ARTPrs is calculated from two factors, CF and f, as shown in Equation 2 from 10 CFR 50.61. 5.2.1 &RTPTS = (CF) f (0.28 - O.lOlog fl Eq. 2 CF = Chemistry Factor f = Best estimate neutron fluence uni ts of 1019 n/ cm 2 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 credibl~ surveillance data. This table relies on the copper and nickel content of the weld material to determine the CF. gives the copper and nickel contents for comparable heat No. W5214 welds other than thee-steam generator welds which are shown in Attachment 5.

Table 5.1 shows the chemJstry 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 8 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.

c:J 1 2 3 I Ave.rage II Table 5.1 Weldment Copper 0.341 0.310 0.266 0.30.6 I PALISADES NUCLEAR PLANT ANALYSIS CONTlt:fUATION SHEET , A,.. 'A/SG/A' . Nickel Copper Nickel 1.093 0.367 1.154 1.003 0.291 1.156 l.090 0.278 1.059 1.062 I 0.312 .1.123 EA-RDS-94-02 Sheet _7 _ Rev # _l_ 'A/SG/B' Copper Nickel 0.353 1.203 0.233 1.149 0.237

1. 024 0.274 1.125 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 ~nd ni~kel. It also provides the averages of copper and nickel cont~nt for use in determining Palisades reactor vessel axial weld material CF from JO *CFR 50.. 61. Some of the copper values have be_en double counted because they were from tandem welds. This is the same averaging technique as used in Reference

• 3.4.

I I. D. I

• . CoEEer II I.D.

I Nickel I 04463 IP2 0.20 04494 IP2 0.94 0.20 04541

1. 20 '

HBR2 Torus 0.159 04577 & 04604 1.00 0.159 04673 Mi 11 lC 1.05 IP2 Sur 0.20 04674 IP2 1.12 IP3 Sur

0. i6 04686 Mll 0.97 II II 0.16 04687 IP21 0.92 IP3 Nozzle 0.15 04688 Pal 0.99 HBR2 Sur.

0.34 04690 1.13 oc 1 Sur:.';c~,,; * ' 0.285 HBR2 Torus 0.99. Pal We l dmeo~?lr

• 0.306 IP2 Sur 1.03 n

0.306 IP3 Sur 1.12 PAL A/SG/A 0.312 IP3 Nozzle 1.09 II II 0.312 HBR2 Sur 0.66 PAL A/SG/B. 0.274 Pal Weldment A 1.062 II II 0.274 Pal A/SG/A 1.123 Averaoe 0.237 Pal A/SG/B 1.125 Average 1.03 Table 5.2 Best Estimate Cu and Ni Values for Palisades Axial Welds.

EA-RDS-94-02 M'IT'#S I JlllllillaS PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET Sheet _8_ Rev # _l_ The best ~stimate 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 Westinghcruse 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 analysis. are shown Palisades may choose in the future to recover this conservatism from its The best estimate fluence rates from Westinghouse for cycles 1 through 9 in Attachment 3. For cycle~ 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 tumulative flu~nce 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 follows. The current burn-up, 7222.1 MWD/MTU, times the MTU of the core, 81.202 MTU, divided by the rated power, 2530 MW, gives 231.8 EFPD's. Since the limiting welds are the welds at the 30° positions, only fluence rates at these angles are used.

• • -un; a:aas Cycle Cycle Fluence Number EFPD' s Cale's 1

379.4 4.70El0 2 449.l

4. 70El0 3

349.5 4.70El0 4 327.6

4. 70EIO 5

394.6

4. 70E10 6

333.4

4. 79El0 7

369.9

4. 79EIO 8

373.6 2.34El0 9 298.5 2.00ElO 10 356.9 I. 94El0 11 231.8

1. 66El0 PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET Rate Cycle Fluence Best Est.

Ca le' s Best Est.. 4.43El0

1. 54El8 l.45El8 4.43El0 l.82El8 I. 72El8

4. 43El0.

I. 42El8

1. 34El8 4.43El0

1. 33El8 I. 25El8 4.43El0 I. 60El8 l.51El8

4. 52EIO I. 38El8 l.30El8 4.52El0 l.53El8 l.44El8 2.21El0 7.55El7 7.13El7 I. 89El0 5.16El7 4.87El7 I. 83El0 5.98El7 5.64El7 I. 57El0 3.33El7 3.14El7 Table 5.3 Palisades Fluence Values.

EA-RDS-94-02 Sheet _9_ Rev # _l_ Cumulative Fluence Ca le' s Best Est. I. 54El8 I. 45El8 3.36El8 3.17El8

4. 78El8 4.51El8 6.11El8 5.76El8 7.71El8 7.27El8 9.09El8 8.58El8 1.06El9 l.OOE19 I.14El9 1.07El9 l.. 19El9 l.12El9 I.25El9 l.18El9 L28El9 l.21El9 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.

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. 0.28 - Jo.0784 - 0.4 log f = 10 °* 2 (RTns - t - Ml CF Eq. (3) The maximum RTPrs allowed for Palisades axial welds is 270°F, reference 3.1. Using this 270°F value for RTPTs' -56°F for I, 66°F for M, and 242°F for CF, in Equation 3, gives a screening criteria fluence value of l.31*1019 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.

PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET EA-RDS-94-02 Sheet 10 Rev # _l_ Margin= 1.31*1019 n/cm 2 -l.28*10 19 n/cm 2 = 3.0*1017 Fluence/EFPD = l.66*10 10 n/ (cm2-sec)

• 3600 sec/Hr*24 Hr/Day Fluence/EFPD = 1.43*1015 n/cm 2 EFPD 1s = 3. O*l0 17 = 210 EFPD 1s

1. 43 *10 15 Table 5.4 shows the number of EFPD's/EFPY's left before the plant reaches the screening criteria using different values of (RTPrs - I - M).

This table includes values of EFPD's/EFPY's for both calculated and best estimate fluence data. The best estimate fluence rate for cycle 11 is l.36*1015

• Value of Screening Criteria Margin Using Margin Using RT OTC I - M Fluence Limit Calculated Fluence Best Est. Fluence 260 1.31El9 210 EFPD's 735 EFPD's -

262

l. 35El9 490 EFPD' s 2.82 EFPY' s 264 1.39El9 769 EFPD's 3.62 EFPY' s 266

1. 43El 9 2.87 EFPY's 4.43 EFPY's 268 1.47El9

3. 64 EFPY' s 5.23 EFPY Is 270 l.52El9 4.59 EFPY's 6.24 'EFPY's 272

1. 57E19 5.55 EFPY's 7.25 EFPY's 274 1.61El9 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 1.77El9 9.38 EFPY's 11.3 EFPY's Table 5.4 Possible Margin Gains.

It is* important to note two things about Table 5.4. First, the times stated do 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.

PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET EA-RDS-94-02 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 19 End of cycle 10 fluence = 1. 25*10 19 Cycle 9 fluence rate= 2.00*10 10 *3600*24 = 1.73*10 15 EFPD's = 1.31*1019 - 1.25*1019 = 347 EFPD's

1. 7 3 *1015 Margin= 347 EFPD 1s - 232 EFPD 1s (thru 10-31-94) = 115 EFPD 1s 6.0 Conclusion The objective of this EA has been met.

Palisades PTS screening criteria margin 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.

PAit r :)0. uWMESTIC LICENSING OF PROOU~TION ANO UTILIZATION FACILITIES I < c > The holder of a license author-izing. operauo~ of a. production or uti* lizat1on facility who desires cl l a change in technical specifications or .,. c 2 l to make a chan1e in the facility or

the procedures described in the safety

analysis report or to conduct tests or a: experiments not described in the

~ safety analysis l'ePGrt. which involve ,.. an unrev1ewed safety question or a. lJ change in technical specifications. shall submit an application for a.mend* ment of hts license pursuant to § 50.90. t 50.IO A a a 1ptm1e1 crtt9'tll far 1'9cture p!Hefttlon W9 for llytilWllllli nudlir (*)Except 11 provided in p1r*F*Ph lb) of thi11ection. 111 lightw1ter nucle1r power reecton mu1t meet the fnicture tou11hne11 1nd material 1urvei111nce Prosr*m requirement.I for the reactor cool1nt pressure boundary 1et forth in ~ Appendicet G and H to thi1 part. ~ (bl Propo1ed altem1tives to the a: described requirement* in Appendices G ~ and H of thi1 1>9rt ar portion* thereof

• may be uled when an exemption is L

granted by the Commiuion under f 50.12 I 50.11 Fnicv1 Touttv-A1-I '"*It* for 'nltldon AplMt flr9uurtnd nwm.i 9hoctl E venta. (a) Definitiof16. For the purpoaes of this 1ection: (1) "ASME Code" meana the American Society of Mechanical Engineera. Boiler and Pre11ure Vessel Code. Section m. "RuJe1 for the Con1truction of Nuclear Power Plant Components," edition ant:! addenda ae specified by I 50.SSI. Codes ind Standard1. (2) "Pre11urized Thermal Shock Event" mean* an event or transient in preuurized water re1ctors (PWR1) causin11 1evere O\\'el"Coolill8 (thermal

1hock) conCWTent with or followed by

"' 1isnificant pressure in the reactor ~ vessel. =: (3) "Reactor Ve111el Beltline" means o the region of the reactor vesael (shell .n material ineludina weld1. heat affected zones. and platea or forgtnssl that directly 11UTOW1d1 the effective beisht of the active eon and adjacent region1 of the reactor ve111el that are pnidicted to experience sufftcient neutron radiation damage to be considered in the 1election of the mo1t Umitins material with regard to radiation dama11e. (4) "Initial RT'""'" meant tbe reference temperature for a reactor ve11el material H defined in the ASME Code. Pal'lsnph N&-2331. RT..., meant the reference temperatani 11 adjusted for the effect* of neutron l'ldiation for the period of Hl'Yice in qufttion. (5) "RT""" mean1 ttie reference temperaturw celc:ulated by the method given in 1>9nsraph_ (b)(Z~ of.tbi1 11eetion for use 11 a 1Creen1111 critenon.. ~ f1] Rec;w:-emP.nts." [l) f:Jr each pressurizad water nucli:ar power reactor for which an operating lir:ense has been issued. the lir.ensee >h<.!:I submit projected values of RT,,, f,x_~eactor ves&el belthne materials by 1,p*:*:g vabes for !he time of submittal. ~ rr:e e'<:pirat1on d.,te of lhe operating N !1cense. the projected expiration date if ~ d chC1:1ge in the operating license has u.. been r<:?qucsted. and the* projected ~ ex;i;ra:1on date of a renewal term if a rcc;ut!st for license renewal has been

• rn!imn:ed. The assessment must *use the 1.uk~l<1!1ve procedures given in p*1ra;;iraph (ui(2) of this section. The

.r~sr'ssmer.i must speedy the bases for tta! pro1ection. including the 50-47 <l~sJi::;:it:or:3 re~ardi:-.g sore ioad* -, pa::P.:-ns. T!:e subm:::a1 must ::31 *;~"e rnpper and 111ckel contents. and :!:e ~~cr.ce values used in the ca>::.i:d::*J:-i ,*Jr each bcltane mater1dl. If these

• ~ *.,;:-i:1::es differ from those subm:~'.ed ::i.

~sponse to the on;pnal PTS rule ~nd

,tcce;ited by the NRC. juslificat:on must be provided. If the value of RT,.,., for ar..v material :n the beltHne is projected to e'<:-eed the PTS screenin11 craenon b1!fore the expiration date of the

• Jpera t:r.g license or the proposed

~9:ra!1on date if a change tr.. :!:e '..;e::;e r..c1s bel!n :e"uested. or :he end of a ~!1ew=,! te~ J a ~equest for !:.:c:-:se

r>r..e'-'.J: ~as bec:i. subr,::::ed. :his dSSchmer.: must. be s\\.:brr::tted by December 16. !991. Otl:erw:se. :his uSSCSSmen! m'.lSt be suom::!ed w;:h :::e

,ext ;i;

ida '.e of the pressure*!err.;i" :*,: _:e

.1m::s. or :he next reactor **essei mater1.. il surveillance report. or 5 "*:u.:s from :r.e effective date of this :ui~. "hir.hP.ver cu mes first: ThcsE' *.. ::r-.. *:..,.:s musl be upddtcd wheni::ver :::.,re,s d significanl change in projected vaL.;es 'Jf RT l'!'S* or upon a reque11t for a chd::ge :n the expiration date for operat:on of the facility. (2] The pressurized thermal sh..::cic. {PTSJ screening criterion is Z70"F :cir plates. forgings. and axial weld Sil ntdtemils. or J00°F for circwnfere'.'!:iai ilS weld materials. For the purpose of ~ companson wilh this criterion. the va1~e u.. of RT..,., for the reactor vessel must tie rs calculated as follows. except as provided in paragraph lb](3) of :h:s sec:!on. The calculation must be-:nade for each weld and plate. or forg!:-:~. :n the reactor vessel beltline. E'q*.id:ion 1. RTP'T'S=l+\\1-.lRT,..,.5 (ij **1" means the initlal referen~e temperature (RT~DTJ of the unirrc.d:c>t~d material measured as defined 1n the ASME Code. Paragraph :'-'l>-Z3J1. ~feasured values must be used 1f credible values are available: tf :lo!. :~~ f illowir~ goneric mean values m'.l~t be used: o*F for welds made with L~:-:de ~o flux, and -S6°F for welds made with Unde 0091. 1092 and 124 and ARCOS 3-5 welc.J fluxes. (ii) "M" means the margin to be a.iced to cover uncertainties tn the va:*Jes of initial RT,.DT* copper ar.d*nic:kel contents. f1uence and the calc:i!a!:*J:-,.,: procedures. In Equauon 1. ~I is 66'F fur welds and 48F for base metal if ger.e::c values of I dre used. and M iS 56'F f0r welds and J4"F for base me:al if measured v<1lues of I are used. (iii) ART PTs ia the mean value of !!:e adjustment in reference tem;iera::::e cauaP.d by im:diation and should be calculated as follows: Equa !ion Z: ART,,.9 = ( CF]r"' :*"' '" '"'- n (iv) CF ("Fl is lhe chemist:y fawi~. B function of copper and mckel car.tent. CF is given in table 1 for welds and ::i table Z for base metal (plates and June 30. 1993 (reset)

50.6l(b) PART 50 *DOMESTIC LICENSING OF PRODUCTION ANO UTILIZATION FACILITIES 50.6l(b) ., forgings). Linear interpolation i1 permitted. In Table. 1 and 2 "Wt-"

opper" and "Wt-9' nickel" are the best-timate values for the material. which iU normally be the mean of tbt measured values for a plate or forgin1 or for weld aamplea made with the weld wire heat number that matcheJ the critical vessel weld. If these values are not available. the upper limiting values given in the material specifications to which the veuel was built may be used.

If not available. conservative e1timates (mean plus one standard deviation) based on generic data 1 may be used if justification is provided. If none of these alternatives are available. 0.359' copper and 1.~ nickel mu1t be auumed. (v) "r' means the best estimate neutron fluence. in unit1 oflO" n/cm 1 (E greater than 1 MeV), at the clad-base* metal interface on the inside surface of the vessel at the location where the material in que1tion receives the highest fluence for the period of 1ervice in question. TA.BU: 1.-CHEMISTRY FACTOA F"OA WELDS, 'F TABLE 2.-CHEMISTRY FACTOR FOR METAL, °F Coppt'I'. Wt-% lliickef. WI*"'* o :o.20io..cilo1SOl0 ao 1oo*1 20 0...................i 20i 20i 20201 20: 20: 20* 20 0.01.. *-*---1 20f 201 20i 20i 201 20 0.02................. i 201 201 20j 201 201 201 2D 0.03................,1 201 20/ 201 20i 20. 20: 20 o.°'***************** 221 291 261 2111* 28[ 211 216 0 05................. \\ 251 311 311 l1 31

31.

31 o oe................. 28 37, 37 37 37

111 *37 0.07........... -....

1 31 ! 43!.. 441 441 ooe................ 34 481 s11 s11 5,i s11 51 o ot*-*---~ 'r1 531 sei sal sai sa1 se

0. 10 *... _.

41 5tf *t 851 671 57, 67 0.11._.. ____ ,5 621 72, 74 771 77: 77 0.12..,.......... _ 1 49 871 :!J n 86 et1I 89 0.13.......... *-**** 53 71 91 99 991 99 0.14........ _ s11 1s1 g,' 1001 1ost 1oei 1oe 0.15.. -.. **--* 61 80~ 98 110 115 1171 117 O.HI... _. __ 65 a.I 1°' 1191 123 12si 125 0.17............. -.. 88/ aej TTO 1211 1 132113~! 135 0.19................ 73f 92. 1'5 13' 141 14 14, IJ.19......... - 711 ~j 1201 1421 1501 15': 15' 0.20..... --******1 821 1~ 125l 1491 1591 1641 165 o.2i......... __ 1 eel 1011 129; 1551 1s1I 112: 11' 0.22............. _.., 9Tt 112, 1341 1811 176! 1811 18' 0.23............ -... 1 as/ 117! 1381 1~ uwl 190! HM ~:~~:~;::*--***l* 1~~ 1 !s' ~~- }~ ~ 1 ' ffii m 0.211.................,, 138 1~[ 1~1 219 2331 239 Nickll Wt*'!I. 029................. 12' 142 1e..1 1sq 2<11 241, 243 0.30.............. J 1291 1481 187! 1941 2251 2491 257 o o.ao o.~10.eolo.ao 1.ooluo o 31..............

• 134; 1511 1

112; 1ael 228 1 1 2ss1* zee SI! 032................ 1311 155 1751 202/ 231 290 274 20 20 20 20 20 20 20 llS o.33*-*******-*** 1ui 1eoi 1eol 205 23' :zMi 2t12

• • • • • ' 201 201 20 20 20 201 20 ~ OJ'*************-* 1~ 151 1MI 209 2391218j 290 2

22 1 28 27 27 271 27f 27 ff 0 :JL........... 153 19 117 212 2'1 272l 291 85 41 4T' 41 411 4T :8 0.38................ 151 17'3 191 218 245 275 303 °'***********-****j' 241 ,3[ s.i s..I s.' 54* 5, g-~**-*********** ~Ml ~n ~ = 241 =-79 ~ o 05................. zel 491 157' eei 1515! e91 ti8 1 IJOS....*...... -... f 29! 52j nl 152! 82! s21 82 o.>>*-***-***-* 171 1 227 216 317 0.07. _______ -J 321 561 851 951 95, 951 glj 0.40.............. 175 189 231 257 218 320 1J.oe. __ **--*i ~ se1 001 1oe1 1oe1 1oe1 1oe 0-08****-**-1 40 81 ' !Mi 115 1* 122! 122! 122 (3) To verify that th* values of RTPTS

0. 10......... -......) "' 851 1111122 1331 1351 135

o. 11............. -.. f '8 ee 101 130 1uj "'j 141 calculated 11 required by paragraph 0.12_..... _ _j 521 721 1031 1351 1531 1e1 1~1 (b)(Z} of this 1ection aN boundins values o.1L.. ---j sej 76j '°'j 139J 1e2J 1121 176 for the specific reactor ve111J. liceDHft o."*****-*--t 11/ 791 109 1.. 2 188, 18211111 shallcon1iderplant-1pedftcinformation O. 'L--~ *I Ooi "~,.., "' "'* 200 that could effect tlw lnol of 0.1t*-*******--

701 11111 115j 1'9f 17151 199! 211 0.11... *-***- ~ 12 11 1511 1.. 1 2011 ~ embrittlement. Thi8 information includea

o. 11*-*-**--

7'I

• 1122 15ot1 191 2,.1* ZJCJ but it not limited to the reactor vestel 0.11.. ___

13 100 12111 1s11 111 220 m operatins temperature and IUl'Veillance 0 20

• • • • -**-*~

1111 10.I 129 1eo '" 223 245 results. Results f?Om the plant-specific 0.21......... ___ 82 1081 133 1M, § 252 o Z?........ __ 9~ n2 137 1 m surveillance pri>gram shall be integrated 0.23........ __ 101 1111 1..c> m into the embrittlement estimate if, 0.2*.*-*-*-1 106 121f 14' 173 211 (i) The plant-specific 1urveillance data 025-*******-******j 110 1ze. 1" 1 : 2111 243 272 ha1 been deemed credible H defined in 0.28.............. 1131 1301 151 1eo -11 2481 279 R_..*lato,.., Guide 1.91 Revtlion 2. and 0.27................ 1 1191 13'1 15!1 1 :11 2481 290

• 1 0:29**-***-*--l 1~11381 1eol 1a1 211 2111294 (ii) The RT"' valu.. changes*

o ~*********---j 1211J 1.ca1 1&&; 111 I 222 2S& 211 significantly.' o J0............ ___ 1 131 1461 157* 19'1 22~ 257I' 290 Any infonnation that is believed to 0 31 *************** 13111 1s1! n2! 1981 221 290. m irnprovct the accuracy of the RT""' value 0.32.................,1 1"° 1551 175i 202; 231/ 293! 299 iignificantly shall be reported to the 0 33................. 14' 1601180/ 20Sj 23'12981 2911 o.34...... **********i 149. uw 11141 20912:18 269j 302 Director. Office of Nuclear Reactor o.35................. 1S3, 1eel 197 212 241 212 305 Regulation. Valuea of RTm that have 0.3e................. 158 172 191! 219 2451 275j le.a 0 J7.......... - 162 1n1 19451 2201 2481 2781 311 'l.38........... -.. 1eer 182. 200j 22:Jn 2501 291: 3~, 'Chanpe ta RT,.,. nlun ani conwidenicl 39........ *-***** 171. 185i 203! 22 ~/ 2951 317 111nific:an1ifaitlaardlel'U.dalanllilwdin

• • '°***-

175f 1881 2071 231 257J 288i.* 320 pvalfeph (bNa) ol tbill IRlion m IM al1arna111 .-1llu* de~ illl ~h. (bl(.1) al thia MC1icn. or boch vah1n. eiu:aed th* ICl'ftllina criterian. prior 0*1* Crom reactor v-i. C*brica1*d 10 Ille MIM teri*I specificauon in lhe tame lhop 11 Ille vneel in question *nd In the *me nma period i1 an 1xam,io1 of '".-naric da!L

• June 30, 1113 (met) 10 1hc e~prralion of the OJ191'9Unf lh:8n*. includina

""~ renewcc term. 1f 1pplic:able. for che pl*nl. been modified using the procedures of tltis paragraph an! subject to the approval of the Director, Office of Nuclear Reactor Resulation when use<l as provided in this section. ( ~) For each oressurized water nuclear power reactor for which the value of RT,,.,, for any material in the beltline is projected to e)(ceed the PTS screening criterion before the e)(piration date of the operating license. or the projected expiration data if a change in the license has been :-equested. or the end of a renewal term if a request for license renewal has been submitted. the

• licensee shall submit by March 16. 1992.

an analysis and schedule for implementation of such flux reduction programs as are reasonably practicable to avoid e)(ceeding the PTS 1C?"eening criterion set forth in paragraph (b)(2) of this section. The schedule for implementation of nux reduction measures may take into account t~e schedule for 1ubmittal and antic: pated Commission approval of detailE'd plant* specific analy.es. 1ubmitted to demo111trate acceptabie risk at values of RTP't"I above the screening limit due to plant modificahoru. new infol'IY'..ation or new analysis techniques. (5) For each pressurized water nuclear power reactor for which the analysia ~ required by paragraph (b)(4J of this ~ section indicates that no reasonably ff practicable flwc i-eduction program will

8 prevent the value of RT PT'S from exceeding the PTS 1Creening criterion before the e)(piration date of the operating license. or the projected e)(piration date if a change 1n the operating license has been requested. or the end of a renewal term if a request far license renewal has been submitted.

the licensee shall 1ubmit a safety analysis to determine what. if any. modificationt to equipment. systems. and operation are necessary to prevent potential failure of the reactor vessel as a result of po1tulated PTS events if continued operation beyond the screening criterion it allowed. In the analysis, the Ucenaee may determine reactor vessel maleriala properties baaed on available information. M'!search results. and pl&llt 1urveillance data, and may use probabilistic fracture mechanics techniques. This analysi.3 must be submitted at least 3 years before the value of RT P"TS is projected to exceed the PT'S screening criterion or by one year after the effective date of this. amendment, whichever is later. (6) After consideration of the licensee's an11lyse1 finch1ding effects of propoted corrective actions. if any) submitted in accordance with paragraphs (b)(4) and (b}(5) of this section. the Commi11ion may, on a caae-by-ca1e basis. approve operation of the facility at value1 of RT PT'S in e)(ce11 of the PT'S 1cree~ criterion. The

.-Art f ~y

• IJUMc~TIC L.JCENSlNG OF PROOUCTION ANO UTILJZATION FACILITIES Commission will consider factors Significantly affecting the potent.Jal for fail~re of the reactor vessel in reaching a decision.

('.") If the Comm1S1ion conc!udr.s. pursu..snt to paragraph (bJ(SJ <Jf this s1:ct1on. that operation of the facility at valul'!s of RT P'T'!I in 11xce11 of the PTS screenmg criterion cannot be approved on the basis of the licenaee's aJJalyses submitted in accordance with !:::I paragraphs [bj(-1) and (b)(S} of this ~ section. the licensee shall requr:st and

1l receive Cummissi.on approval prior to any operation beyond the criterion. The request must be bated upon modifications to equipment. systems.

and operation of the facility ill addition to those previously proposed in the submitted analyses that would reduce thP potential for failure of the reactor "~~s.,l due to PTS events. or upon

• urther analyses based upon new

.. 1:orm11tion or improved mctn0Jol0gy. ~ !0.!2 Atqulr~enta for rWductfon of rl'Jll lror.- 111.ttcipP!-C trati~:er.ta *llt'lout t.;r~m (A TWS) eveni. ic.r lignt-water-<coi.ct nuc;le*r po*er.~lc1:it1. (a).4;;;/ir:cl:::i:ry. The reqai!ment~ uf ~ii:s ~ection apply to ~II c:-1merc1.<1l l!:;;;t-wc:teP-r.ocled nuclear power pl,1il:s tr,j De'i!IU'"n.. F.1r p*~:p_;~es of th:s sr.ctior.....,r.ti.::!pated Tran.~ie:lt \\\\'d'!uut S*.rarn" [A'l'\\'*5) means an ar.tk:p:ed operational occurrence as dr.fi!"!ed in pi;-cndix A of this part fo!lowerl hy the failure of the reactor trip po;:t:on cf the prcter.tion sy~tem spccH°if!d in G~neral Oes:sn Criterion 20 of Appendix A *if this part. (cl Requirements. (1) Each pressurized ~ater reactor must have equipment from sensor output to final actuation device.

that is diverse from the reactor trip

~ system. to automatically initiate the a: auxiliary (or eme?1Jency) feedwater u.. system and initiate a turbine trip under

condil1on1 indicative of an A TWS. This equipment must be desiened *to perfonn its function in a reliable manner and be independent (from sensor output to the final actuation device) from the exi1tin9 reactor trip system.

(2) Each pre11uriud water reactor manufactured by Combustion Engineerins or by *bccck and Wilcox must have a diverte ICl'&Jft 1y1tem from the sensor output to interruption of power to the control rads. This ecram system must be desiped to perfonn its function in a reliable manner and be independent from the existin11 reactor trip system (from sensor output to interruption of power to the control rods). (3) Each boiling water reactor must have an alternate rod injection (ARI) system that is diverse (from the reactor trip 1ystem) from sensor output to the final actuation device. The ARJ system must have redundant scram air header exhaust valves. The ARI must be

designed to perform its function in a

~ reliable manner and be independent a: (from the existing reactor trip system) ~ from sensor output to the final actuation L device.

14) Each boding water reu<"tor must hnve d stunc..lby l1qu1d conrrul s-,stc:n (Sl.CSJ w11h the capubd1ty ui 1n.1r.cting into the rr.;ir.tur pressure ve~sel d borated water solution iii ~uch il now rate, levci of boron conc,.ntratiun and boron-10 isotope enrichment. and accounting for reactor pressure vessel volume. that the resulting reactivity control is at least equivalent to that

_ resulting from in1ect1on of 86 gailons per l5! minute of 13 weight percent sodium <9S pentaborate decahydrate solution at the ~ natural boron-10 isotope abundance into a 251-inch inside diameter reactor i pressure vessel for a given core design. The SLCS and its injection location must be designed to perform its function in a reliaLle manner. The SLCS initiation must be automatic and must be des111ned to perform its function in a reliable manner for plants granted a construction permit after July 26. 1984. and for plants granted a construction pennit prior to July 26. 1984. that have 11lready been desisned and built to include this feature. r (SJ Each boilin11 water reactor must ! have equipment lo trip the reactor ~ coolant recirculating pumps ~ eutom11tically under conditions

11. indicative of an ATWS. This equipment

~ must be designed to perform its function Lin a reliable manner. (6) Information sufficient to demonstrate to the Commission the adequacy of items in paragraphs (c)(t) through (c)(S) of this section shall be submitted to the CommiHion al specified in I 50.4. (d) Implementation. By 180 days after the i11uance of the QA guidance for non-eafety related components. eacb 8 licensee 1hall develop and submit to the

Commission. H specified in I 50.4. *

• propoled 1chedule for meetins the

~ requirements of parasraph* (c)(l)

through (c)(5) or this section. Each 1ball include an explanation or the 1chedule alons with a justification if the ecbedule call* for rmal implementation later than the second refuelins outase after July 28.

19&1. or the date of iHuance of a license authorizing operation above 5 percent of full power. A final schedule 1hall then be mutually agreed upon by the _ Commi11ion and licensee. r11o.11 1..-................. I,..... 8 (a)&quiNmenr.. (1) Each lisht* ~ water-cooled nuclear power plant "' licenled to operate muat be able to ~ with*tand for a 1pecifiad duration and .., recover from a station blackout u "'1 defined in I 50.Z. The 1pecifltd st1tion blackout duration 1hall be baled on the folluwins factors: (i) The redundancy of the omit* emef'Rency ac power soura1: (ti) The rehabality of the on11te eme?IJency IC power 1ources: (iii) The expected frequency of !au of offsite power: ind (iv) The probable time needed to restore offsite power. (2) The reactor core and 111ociated coolant. control. and protection systems. includins 1tation batteries and any other necessary support system1. mu1t provide sufficient capacity and capability to ensun!! that the core is cooled and appropnate containment integrity is marntained in the event of a statron blackout for the specified duration. The capability for copins w11h a station blackout of specified duration shall be detenruned by an appropria re coping analysia. Ullhties are expected ro have the baseline assumptions. analyses. and related information used in their copi"8 evaluations availebie for NRC review. (b J limitation of scope. Parasra p h [ c) of this section does nol apply to those plants licensed to operate pnor to fuiy ZI. 1988. if the capability to withstand atation blackout was specifically addn!11ed in the operating license proceedins and wH explicitly approved by the NRC. (cl lmp/emeritation.-{1) /nformar1on Submittal. For each lijht-water-cooled nuclear power plant licenaed to operate on or before fuly 21, :988. the licensee shall submit the infonnation defined below to the Director of 1he Office of

Nuclear Reactor Regulation by.4.pril 17.

~ 1989. For each lisht-weter*cooled ~ nuclear power plant licensed to o_perate ... after the effective dale of !hi1 ~ amendment. the licensee 9hall submit the infonnation defined below to the Director by 270 day1 after the date of license iuuance. (ii A proposed stalion blackoul duration to be used in determining compliance with paragraph [a) of this 1ection. includins a justification for the selection bated on lhe four factors identified in paragraph (a) of thi1 section: (ii) A deec:ription of the proceduree that will be implemented fo. station blackout event* for the duration determined In parqraph (c)(l)(i) of this eection and for recovery therefrom: end (iii) A li1t of modificatfon110 equipment and as1ociated procedures. 1f any. nece11ary to meet the requirement* of paragraph (a) of thi1 section, for :he specified 1tation blackeut duration determined in paragraph (c)(1J(iJ of this section. and a proposed schedule for implementins the 1tated modifica~ions.

• (2) AlttJrnattJ ac sourre: The alternate ac power sourcefs). as defined in § 50.Z.

will constitute acceptable capalHlity to withstand station blackout proviJed an analysi1 is performed which demon1trates that the plant has this carability from onset of the station blackout until the alternate ac sourc*efsl and required shutdown equipment are 1tarted and lined up to operate. The *1me required for 1tartup and alignment of the June 30, 1993 (reset)

4.0 PROJECTED RTers 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 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?rs values. 4.1 Determination and Projection of the PTS Reference Temoeratures The base equation for the PTS reference temperature from 10CFRS0.61

s

RTPTS = I + M + ARTP'n ( 1 )

• r* is defined as the initial reference temperature (RT..,1 ) of the unirradiated material.
• 1* values for the Palisades reactor vessel beltline materials are:

Axial Weld Circ Weld Plate I 1

• -ss*F } Generic Value 10CFRS0.61 (b)(2)(i) for Welds made re * -ss*F with Linde 1092 and 124 Fluxes Value* reported in Reference 6. This represents the limiting plate.
• A less conservit1ve vilue of I, * -1o*F was measured by Battelle Columbus Laboratories in 1977 (Reference 39). A value of -s*f was used in CPCo's 1986 (Reference 16) and 1991 (Reference 1) PTS subntittals. Confirmation of

-s*F could not be found by measurement or calculation. 4-1

"H* is defined ~s the margin term added to cover unc~rtainties as in th~ values of initial RTPrs (Cu and Ni content, fluence and the calculational procedures). Values of "M* for the Palisades vessel beltline material are: Axial Weld I Value specified in 10CFRS0.6l(b)(2)(ii) if generic values of used. for welds "I" are Circ Weld Plate Value specified for base metal in 10CFRS0.61 if measured value of "I* is used "11RT Prs" is defined as: 4RTPTS = (CF) f(o.2e - 0.10 109 £l "CF*, the chemistry factor, a function of Cu and Ni content, is derived from Tables 1 and 2 of 10CFRS0.61. In Section 2, the chem;stry factors *were determined to be: CF.

• 211*F for the axial welds.

CFc

• 22a*F for the circumferential weld.
• CFP
• l&S*f for the vessel plate material.
• f* is the best-est1*ate neutron fluence in units of 1019 n/cm2 (E > 1 MeV) at the clad-base metal interface of the vessel.

4-2 (2)

The limiting fluence is determined by setting RTPrs equal to the ) screening criteria and solving for f. *First, rearrangirig equations ( 1) and ( 2) : RTPTS = I + M + (CF) f (0.28 -0.10 log fl ( RT - I-M) ( o. 2 8 - o. 1 o log f) log f,. log PTSCF 0.10 (log f) 2 - 0.28 log f +log (RTPTS - I-M) = 0 CF Using the quadratic equation to solve for log f: Because the positive root of-the equation provides meaningless results, the equation may be simplified to: o

• 2 a -

~ o. 0184 - o. 4 log

• CF

. [ ~ ( RT,.. -I -H) l f

• 10 exp 0. 2 The max11111 allowed values of RTPts is defined in 10CFRS0.6l(b)(2) for*11cfl of th* Palisades beltline is:

Axial Weld Circumferential Weld Plate Material* . RTprsa

• 21o*F RTpr1c
• 3oo*F RTPTSp
• 21o*F 4-3

Cycle 2 3 4 5 6 7 8 9 2 3 4 s 6 7 8 9 Table 8-4 (Continued) Palisades Fast Neutron Fluence (E > 1.0 MeV) Tilrough Cycle 9 At the Reactor Vessel Clad-Base Metal lnterface Cycle .Cycle Cycle Length Flux Flue nee CEFPD) (n/cm2-s) (Q/cm2) 30 Degrees 379.4 4.43E+l0 1.45E+18 449.1 4.43E+IO

l. 72E+ 18 349.5 4.43E+l0 l.34E+l8 327.6 4.43E+l0 l.26E+l8 394.6 4.43E+l0 l.51E+l8 333.4 4.52E+l0 l.30E+ 18 369.9 4.52E+l0 l.44E+l8 373.6 2.21E+l0 7.13E+l7 298.5 l.89E+l0 4.87E+l7 45 Degrees 379.4 2.81E+IO 9.22E+l7 449.l 2.81E+l0 l.09E+l8 349.5 2.81E+l0 8.49E+l7 327.6 2.8 lE+lO 7.96E+l7 394.6 2.81E+l0 9.58E+l7 333.4 2.86E+l0 8.23E+l7 369.9 2.86E+l0 9.14E+l7 373.6 l.67E+l0 5.39E+l7 298.S l.09E+l0 2.80E+l7 8-8 Cumulative Flue nee

<n/crn;) 1.45E+ 18 3.17E+l8 4.5IE+18 5.77E+ 18 7.28E+l8 8.58E+l8 l.()()E+ 19 l.07E+l9 l.12E+ 19 9.22E+i7 2.0lE+ 18 2.86E+l8 3.66E+l8 4.62E+l8 S.44E+l8 6.3SE+l8 6.89E+l8 7.17E+l8

Attachment l Page 8 of 3 -CALCULATION OF THE MEAN COPPER ANO NICKEL CONTENT OF WELDS FABRICATED USING WELD WIRE FROM HEAT No. W5214 Th~ following identifications and copper ~ontent values are from Table I.l.

1. COPPER CONTENT Sample Identification 04463 - IP2-flange 1-0428 HBR2 - Torus Flange IP2 - Surveillance IP3 - Surveillance IPJ - Nozzle Cutout HBR2 - Surveillance OCl - Surveillance Weight r. Copper 0.20 0.20 o.159 0.159

0. 20 -

o.16 0.16 0.15 0.34 0.285 Total 2.013 2.013 + 10

• 0.201 *Mean Copper Content

2. NICKEL CONTENT Sample Identification 04494 - IP2 1-042 04541 Average of 04577 l 04604 04673 Millstone IC 04674 IP2 3-0421 04UI, Ml.I 2-972A.

0**1:: 1P21-042A 04_fll;';PAt* S/G 5-943 04aaA*c ; ~ HBR2 Torus Flange 1P2 - Surveillance IP3 - Surveillance 1P3 - Nozzle Cutout HBR2 - Surveillance ). Weight i Content 0.94 1.20 1.00 1.05 1.12 0.97 0.92 0.99 1.13 0.99 1.03 1.12 1.09 0.66 Total

14. 21 14.21 + 14

• 1.015 *Mean Nickel Content
• I I
• I
• I
• I G. Gage 3 I SI Oclobcr l ~)4 MPD/OKl TESTING OF WELDMETALS FOR CPCO PltF.l.IMINARY ANALYSIS RESULTS ON SF.CTIONS TAKEN FROM LARGE WEU>MENTS ('A' AND '11') AND TREPANS H.csults of the chcnucal anlysis of >VCldmc&al s;anaplcs &aken f'rom sections throu~h &he IW\\l l;argc wddmcnts and four Trcpans arc given m &he table bclQW. Although these data are considered 10 be &rue anJ au..*1m1lc, final chcxks have s111l 10 be pcrfonut'd ;uuJ as such 1hc data should be 1rea1_al as bcmat~:~ ~laaina.1}' na1ure m1ul all IJ1c checks have ln:111ti:wic.

..'~.:?~*~.~;:

.~t~1*,

. '-

~' Mn Cr Cu Mo Na I' S1 s v Sl."t:llOO lhrougll l..arge Wcldment 'A' Al/llX 1.157 00:171 0.341 0.502 l 0'}1 O.OIO 0 2<.4 (),()14(, 0 0021 Al/IN. 1.249 0.0:09 0.310 0.507 I.OU.> 0.0 IO 0.2{,"i 0.0174 0.0022 Al/l/Z l.176 IUH17 o.:u><. 0.4M7 I 01)(> 0011 0 2XX O.OIKI ooon Sauon through Lctrf.I! Wddmcn1 'B' Bl/2JX 1.249 0.0400 0.215 0.546 UIS 0012 0 l7

• OOIW 0 (Ml:>X Bl/2JY 1.304 ocn1n

0. I <\\9 0.517 l.OIO 0 012 0.16(,

OOIK:l. OIMW". Ulf2/Z 1.2(1'\\ 0.CHK') O. l 1>f1 0.540 I. ll')l( 0 01'.' o.1ie 00177 0 IMl!X Scc1ion through Trcpiln 'MSG/A' NSG/NllX 1.D7 0_0401 o.:*67 O.iOK I. I :".J 0011 0.247 0.0115 Cl.OW I NSCJNllY 1.120 o.o.rn~ 0.29 I 0.4'JK I.I :'it> 0.011 () 284 00178 o.oo:w AJSG/A/2/Z 1.114 o.o:-n 0.278 0.4*)1( I.WW 0012 O.'.:'X4 0.0IK2 ll.ll021 -Scc11on 1hrough Trcp.1111 'AJSGll AJSG/U/3/X 1.17.\\ O.cM07 CU.5\\ 0.515 1.201 0011 ll HI OOllO 0 11021 AJSG/UfJtY

l. I02 O.O.l89 0.2H 0.52J 1.149 0.012 o.2'JI 0 Ol)K -

0 002.J AJSGIU!VL. I. I05 0 114 l I orn 0.5 l'J 1.024 0.01)

0. H>.!

0 0141 0 ()(12) ~-------- Su:uou tluoogh Tre1wi '8/SG/A' -- ~----------- *-----*- --------~ U/S('J AJ2/X 1.21)2 ().()412 O. l'J:'i 0.551 1.272 0.017 o 1xc. <l.O I KI 00011 U/SG/A/2/Y I 214 o Ol?<J () 195

0. 'Vi()

1..n1 C).016 0.202_ 0 Cll70 o.oon HISGIMJ.f.l 1.24<> O.Ol!fK 0.2(){, fl. '\\44 I llH 0016 O l 11 OOIM 0 0027 Seel.Ion through Trepan 'B/SG/B' B/SG/Dll.JX l.271 0.0.197 (I lto2 0'.541 I. r.u1 0.016 0 I<> I 0.0177 U.0029 H/SG/Bfl/Y l.237 O.C>402 0.208 U.536 1.1 l6 0.0 I(, () 704 () () 16) o.oon I o 01'.'16 A/SG/B/2f/. l.IKK 0.CH<Jl 0.20') 0.5l2 l."107 0.01<1 ()]I'!. 0_001~

---

L--:------ ------

TABLE 6-13 (Continued) CALCULATED FLUENCE (E>l.0 MeV) lHROUGH CYCLE 10 AT nm PRF.SSURE VESSEL CLAD-BASE METAL INTERFACE Cycle Cycle Length Cycle Flux Cycle Fluence Cumulative (EFPD) (D/cm2-sec) (n/cm2) Fluence (n/cm2) 30 Degree 1 379.4 4.70E+10 1.54E+18 1.54E+18 2 449.1 4.70E+10 1.82E+l8 3.36E+18 3 349.5 . 4.70E+10 1.42E+18 4.78E+18 4 327.6 4.70E+l0 1.33E+l8 6.11E+l8 5 394.6 4.70E+10 1.60E+18 7.71E+l8 6 333.4 4.79E+10 1.38E+18 9.09E+l8 7 369.9 4.79E+10 1.53E+18 1.06E+19 8 373.6 2.34E+10 7.55E+17 l.14E+19 9 298.5 2.00E+lO 5.16E+17 1.19E+19 10 356.9 1.94E+10 5.98E+17 l.25E+19 45 Degree 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 5 394.6 2.98E+10 l.02E+18 4.90E+18

6.

333.4 3.03E+10 - 8.73E+l7 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.S 1.15E+10 2.97E+17 7.61E+18 10 356.9 1.32E+10 . 4.07E+l7 8.02E+18 6-28

Cycle I 1 2 3 4 5 6 7 8 9 10 11 EFPD 379.4 449.1 349.5 327.6 394.6 333.4 369.9 373.6 298.5 356.9 422.0 Palisades Cycle Flux Values at Critical Locations I Cycle Flux oo I 16° I 4.59 6.03 4.59 6.03 4.59 6.03 4.59 6.03 4.59 6.03 4.87 6.25 4.87 6.25 2.16 4.89 2.08 3.06

1. 51 2.40 1.42 2.21 E + 10 30° 4.70 4.70 4.70 4.70 4.70 4.79 4.79 2.34 2.00

1. 94 1.66 Values for cycles 1 through 10 are from WCAP14014.

I Values for cycle 11 are from Palisades in-house calculations. 45° I 2.98 2.98 2.98 2.98 2.98 3.03 3.03

1. 77 1.15

1. 32

1. 09

ENCLOSURE 3 TO 10CFR50. 61 PRESSURIZED THERMAL SHOCK - REVISED INFORMATION Consumers Power Company Palisades Plant Docket 50-255 A NON-PROPRIETARY VERSION OF THE CPC NOVEMBER 18, 1994 SUBMITTAL January 23, 1995

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 PLANT 10CFR50.61 - PRESSURIZED THERMAL SHOCK - ADDITIONAL INFORMATION Consumers Power Company (CPC) submittals dated February 23, 1994, November 8, 1994, and November 10, 1994 described our plan to more accurately determine the chemical 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 screening criterion if the preliminary chemistry results were representative of three welds fabricated with weld wire from Heat No. W5214. Our November 10, 1994 letter informed the staff that we: (1) had received preliminary low temperature toughness data from the physical testing material and were suspicious of its credibility for use in determining the initial RTNor 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 RTNor 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 NOTT 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 RTNoT for as fabricated Palisades reactor vessel 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 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, which is subject to staff approval, projects the time before the Palisades reactor vessel material will exceed the criterion 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.

3 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 until 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 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. 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

Rev 0 0 1 2 Calculation Status

• Description Original Issue Original Issue 2nd Review Admin Revision Updated Revision*

PALISADES NUCLEAR PLANT ENGINEERING ANALYSIS COVER SHEET INITIATION AND REVIEW Preliminary 0 Initiated I nit By Ross Snuggerud Ross Snuggerud Appd By Date 11-03-94. GCP ll-07-94 GCP Pending Final 0 0 Review Method Alt Detail Qual Cale* Review Test I I I / Revision I Df~cussion EA-RDS-94-02 Total NUl!tler of Sheets _J_Q_ Superseded 0 Technically Revr Reviewed Appd By BY Date Jim L Biffer ll-03-9.1 ..J1...:"'. George H Goralski il-03-94 ~~G Jim L

• Biffer GHGer.Jtt.

ll-07-94 GCP CPCo Appd 111,_11*17*1r lit,l/ (777 This revision incorporates administrative comments made as a result of the PRC meeting. None of.'the calcul~tions or results cha~ge in this revision. Revision 2 Discuss~O~' Tttts&~*tt!vision* incorporates-the final chemhtry data and discussion of the results of the initial RTNDT test data. This *has resulted in several changes from the tw~ previous ver~ions.

PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET EA-RDS-94-02 Sheet _2 _ Rev = _2_ Table of Contents 1.0 Objective 2.0 Summary 3.0 Analysis Input 3.1 References 4.0 Assumptions 5.0 Analysis 5.1 Values of 'I' and 'M' 5.2 Values of '11RT Prs'

5. 2.1

• Palisades 'CF' Values 5.2.2 Palisades 'f' Values 5.3 Palisades PTS Screening Criteria Limits 6.0 Conclusions Tables*

5.1 5.2 Averages of Retired Steam Generator Weld Chemistries. Best Estimate Cu and Ni Values for Palisades Axial Welds. Attachment 4' ;,,. Attachment Attachment :. ft****': Attachment;. '::.8\\.~{tL:,., Attachment~*:rf;:Ji,.: Attachment lOi' :.

• Attachments Reference 3~1 Section 10 CFR 50.61 Reference 3.2 Pages 4.1 to 4.3 Reference 3.3 Page 8-8 Reference 3.4 Attachment 1 page 8 Reference 3.5 All Refe~ence 3.6 Page 6-28 Reference 3.7 Summary table Reference 3. 8 A 11 Reference 3.9 All Sample ID description 3

3 3 3 4 5 5 6 6 8 9 9 7 8

1.0 Objective PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET EA-RDS-94-02 Sheet _3_ Rev = 2 This Engineering Analysis is written to document calculations which determine the compliance status of the Palisades reactor veisel weld material in respect to the PTS screening criteria. They incorporate the final weld chemistry v~lues obtained from the retired steam generators and the best availabie fluence data. 2.0 Su*ary Calculations have been performed 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 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% tapacity 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 RTNor* The data collected from these me.asurements suggests that the material has been affected by its use in the steam generators and cannot be used to provide the initial RTNDT for this weld material. 3.0 Analysis Input References given in section 3.1 cover the data used in this Engineering Analysis.

3. 1 Referene-~s-3.1 10 CFR 50, current issue.

3.2 6-5-92 CPCo Submittal, Docket 50-255 - Lie. DPR-20, IOCFRS0.61 Pressurized Thermal Shock, Revised Projected Values of RTPrs for Reactor Beltline Materials. 3.3 6-10-~3 CPCo Submittal, Docket 50-255 - Lie. DPR-20, 10CFR50.61 Pressurized Thermal Shock, Reactor Vessel Neutron Fluence, Additional Information.

PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET EA-RDS-94-02 Sheet _4_ Rev = _2_ 3.4 2-23-94 CPCo Submittal, Docket 50-255 - Lie. DPR-20, 10CFR50.61 Pressurized Thermal Shock, Revised Information. 3.5 Testing of Weldmetals for CPCo Additional Chemical Analysis, Letter from Or. G. Gage, AEA, to John Kneeland, CPCo, November 14, 1994. 3.6 6-21-94 CPCo Submittal, Docket 50-255 - Lie. DPR-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. Gimbrone, ABB, to John Kneeland, CPCo, November 15, 1994. 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

• Asswnpt 1 o_ns 4.1 The wel4:;:$111Ples from the retired steam generator are only applicable to, and can onh{:affect, Palisades axial weld chemistries, because only heat No. W5214 and 348009weld materials were removed from the steam generators.

4.2 The 30° weld was and still is the limiting weld. This is the 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

PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET EA-RDS-94-02 Sheet _s_ Rev = _l_ 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 RTNor 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 reporte.d in the regulatory guidance. 5.0 Analysis 10 CFR 50.61 provides the foundation of the PTS screening criteria. Calculations for the RTPrs are done using equation 1 from the rule. ,*,M RTns = I + M + fl RTns Eq. 1 llRTns = Irradiation adjustment of RT I = RTND'l' ( Initial RT) M = Margin term Each of the items in Equation 1 will be discussed with respect to Palisades current situation. 5.1 Values of 'I' and 'M' Palisades does not have an initial RT~r value for its reactor vessel welds. 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 8-5 weld fluxes, reference

PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET EA-RDS-94-02 Sheet _*_5_ Rev = __ 2_ 3.1.. The initial RT~r 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 had been affected by its u~e in the steam generators and ~ould no longer be used to provide initial RT~r* reference 3r9. 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 '6RTPrs' The value of 6RTPrs is calculated from two factors, CF and f, as shown. in Equation 2 from 10 CFR 50.61.

ARTns = (CF) f (o.2e - 0.10109 fl Eq. 2 CF = Chemistry Factor f = Best estimate neutron fluence

• uni ts of 10 19 n/ cm 2 5.2.l 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 th~ c6pper and nickel content of the weld mat~rial to determi~e the CF. gives the copper and nickel contents for comparable heat No. W5214 welds other than tha.steam.gener~tor welds which are-sho~n in Attachment 5. Explanations of the weld:d.s1gnations are provided i"n attachment IO. *Table 5.1 shows the chemistry v.~J:~*¥~1.for the* three 'A' steam generator welds segments from Attachment s -*: *~;(* :~:;_.. ~ and their copper-**and nickel averages. The samples taken from A steain generator were* iandem heat Nor W5214 welds, the B steam generator samples were from heat No. 348009: only the heat No. ws214* ~alues are of interest in this EA, since welds fabricated using weld wire from this heat are limiting. The new data taken for heat ~o. 348009 does not change the limiting weld for the Palisades reactor vessel.

• We l dment Sample Copper 1

0.341 2 0.310 3 0.266 4 0.328 5 0.310 6 0.266 Average Cu I A' PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET I A/SG/A' Ni eke 1 Copper Nickel 1.093 0.367 1.154 1.003 0.291

1. 156 1.090 0.278 1.059 1.116 0.365 1.193 1.006 0.292 1.127 1.104 0.281 1.066 0.297 Average Ni EA-RDS-94-02 Sheet _7_ Rev =

? I A/SG/B I Copper Nickel 0.353 1.203 0.233 1.149 0.237 1.024 0.359

1. 204 0.239 0.960 0.228 1.107 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 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 tande~ welds. This is the same averaging technique as used in Reference 3.4.

_.._.,n ; a illm I I.O. I 04463 IP2 ti HBR2 Torus II U IP2 Sur IP3 Sur IP3 Nozzle HBR2 Sur OCl Sur Palisades SG II II Averaqe Table 5.2 PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET caeeer II I. 0. 0.20 04494 IP2 0.20 04541 0.159 04577 & 04604 0.159 04673 Mi 11 IC 0.20 04674 IP2 0.16 04686 MLI 0.16 04687 IP21 0.15 04688 Pal. 0.34 04690 0.285 HBR2 Torus 0.2.97 IP2 Sur 0.297 IP3 Sur 0.217 IP3 Nozzle HBR2 Sur Palisades SG Averaae EA-RDS-94-02 Sheet _8_ Rev = _1__ I Nickel I 0.94

1. 20

1. 00

1. 05 1.12 0.97 0.92 0.99 1.13 0.99 1.03 1.12 1.09 0.66 1.101 I. 02 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 i.s 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 173 conservative.

~!:!: .,-,n a a11111 PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET EA-RDS-94-02 Sheet _9_ Rev = ? shows that the EOC 10 accumulated fluence at the 30° weld location is l.25*10 19 n/cm2

• 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.

0.20 - Jo.0784 - 0.4 log f = 10 °* 2 tRTns HJ CF Eq. (3) The maximum RTPrs allowed for Palisades axial welds is 270°F, r~ferente 3.1. Using this 270°F value for RTPrs* -56°F for I, 66°F for M, and 234°F for CF, -in Equation 3, gives a screening criteria fluence value of l.49*1019 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 below. PTS screening criteria fluence = l. 49 *10 19 End of cycle 10 fluence = 1.25*1019 Cycle 9 fluence rate= 2.00*10 10 *3600*24 = 1.73*10 15 EFPD 1s = 1. 49 *1019 - 1. 25 *1019 = 13 87 EFPD 1s

1. 7 3 *1015 Margin = 1387 EFPD 1s - 232 EFPD 1s ( thru 10-31-94) = 1155 EFPD's

. '~ :,_- -' Using*th~*-751 capacity factor and the 1155 EFPD's gives Palisades 4.21 years before reaching the PTS screening criteria. This works out to a date sometime i~ mid January, 1999. 6.0 Conclusion 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 reacto~

-.-*ws o a ms PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET vessel weld material has not exceed the PTS screening criteria. EA-RDS-94-02 Sheet 10 Rev = 2

.r'~ ::...,

r. c

• The holder of a iicen.se author*

izmi operation of a. product;on or uti* ~iza.tion !a.c1lity ;vho ::les1res . i 1 a cha.nge in ~echn1cal spec1!1cat11:ins or

2. ta make a chan1e m the !ac1l1ty or

~ the procedures descnbed :n the safety

analy~is report or to conduct tests or

experiments not. described in the

safety llnalys1s regort. which invol~*e

- an unre111ewed s&!ety c:iuest.1on or & L chanu in technical s'pec1tica.t1ons. sha.ll si.bm1t an application !or a.mend* ment of hl.5 licerue pursuant ~o § 50.90. t 50.IO *: n***...,.. far flslln ,..,... w.......... far~*-......... ,.................... ~.... (*)Except 11 providld in para.,.ph lb) of th11 1ection. 1ll liptw1ter nud11r power re1cton muat meet the hc:twl tou1hnn1 end m1tm1l nrvlill1nce Protr*m requimnenll for the r11ctor coolant prn1ure bound1ry Mt forth in ~ Appendicn G ud H to this pu1. ~ (bJ Propoe.ed 1ltnn*t1Yn to the

dncnbeci rwquimn1nt1 in Appendic:e1 G

ind Hof thi1 pert* pertiaft1 therwof

• m1y be uMd wt.en an nemptiOft i9 L

sr-ntld by the Comminion 1111w I 50.tZ

>~

_:~ ~--

5 e*:~~;i.~~.:~.~;;:~J:.~~ .<~;-~~::~-:~.~~:~~~~:~;/:a'.::;.~;

• 1r ~ac~ ~eit.:ne ma:e~:di. ::-..... ~

4 3 ~ -.. a:-:':'.:! d:~fer ~om ~::o!e,... ~"':":... 4 -~~;icr.se ro the or.:;mai PTS ~ *. e, - ~ * !*:::e:i"~d ~y r~e.\\'RC. ;:.;s1:f:~ 0 :.;:-: -*.. ~!! ;;rov1ded. !f rhe Vilue of RT.....,:,;~'.-" -:-:d ~C!r:al :n :he be!tline 19 proiec:ec :J

• e'~~c:J :he PTS 1creen:ng c::'.e:".J!'1 I I0.11 FtllCtUN,...,.._'It* f..,...,..

1~:~=-~ '.~e eX.;Jira~:on Ca!e of :~.e

o:o::::. Atllnlt,._,...... '"*"'-'

>;i~:a:::-:g :::e:-:~e or :he ;irop-:s~~ ~, J:~a~:c~ :a:e J a i::-:a:-:e&?.:-. *:--.~. --=- ..1 (a) De.finitiom. For the purpo.es of ~..... ~ '.::e*~~. :e-;:.;es:ed. ;ir :he e::: ;:* J U-.i(~,.~~~ Code" mHna the

~~~::*~: ~:~:c~:;~;:.~~};,.:*-;~;~

AmericanSocietyofMeclianiCll .,sscs.:ner.: ":"lu3t be,_~:;..::e: :. Et\\tineere. Boiler and Pre111Ue Vn11l L.Jece::1ber :5. :991. O:~.e:*.,.;e.: ; Code. Section m. "IW.lee-for the .. >SC!,91'nent ~:.;st be S\\.~~:::e:i... - -~ Conatructton o!Nuclur Power Pl111t

• ex.t J;Jda:<? of r!':e ;iress-.:e**~"'.':~*:*. -.

Components:* edition anti addenda u

~: :s. or :~e !'lex.I reactor "es 3 ~:*

1p1d(ied by I 50.SS.. Codn lftd ~.. ter:...l su1*edlince re;ior:. J: 5.*.-, Stand1rd1.

rom

l'.e effective diite oi *~.s :* *. ~

(2) "Pre11Urized Tberm1l Slloc:k "h1r !~l"vcr i:umea first T~c'" <. * -*.. ~ ; Event" m1an1 Ill event or tr1n1i1nt in rn.,~t Ot! upddll?d wht?r.e:' er * ::~ *e,,, preuW'ized w1ter re1ctors (PWRiJ ,;~n:Cicant chin&* 1n pro1ec:c:: "a.* ~~ ;f e1u11n1 uvere overcoolins (thumil RT,.,.,. or upon a requht ior a ::-:a~.~".:i

1hock) concmrent with or !ollowld by the ex.piration date for ope

a1:.::'l ;i( ::-.~

~ !~~t preuure lft the re1ctor . f4f~\\ithe pressurized :!':e:-:-r.al s.*... -~ (JJ "Re1ctor Veuel Btltline* mHna 'PTSJ sc:eer.ing criter:on.s z.*o*F :;~ = the reaion of the rwact l ( h U pl 4tes. forg::'lgs. and ax... : weid .., m11.,;11 inductt111 wef:'a:':.': iff~ !ii ITld t~r1.ils. or JOO"F for i:.rcamf~:~ ~ *.,. ionn. llld pllte1 ar f01'liftpl U..t M weid mater:al1. For the ;:>urpose Jf directly mmrand1 the elfectiYI beilht of. ~ cnmp..inson wah thi1 er.tenon. ::-.e. i. -e the 1c:tiY1CDnind1dj1Cllftt f'llion1 of "" of RT.,, for the reactor,*esse! :::~s: :e tha reactor *nnl thit.,. pl'ldictld to ~ Cilculillted as follows. e~cept -as '1qlerilftce nfftc:ient neutrva rtdiition provided in paragraph 1 '!(J; <Jf :.*. i d1mqe to be coniidlNd ill the Mlection ~ec::on. The Cilculition ~u3! '.:!e -, :e

• of the molt limitifta materiil with reprd

.or eiich we!d ind plate. or:.:~-.*-:.., to r1di1tion dlm1p. the reac:or veuel beltbe. (41 "Initial RT..;," m111111 tile merenct Fqud::on l. RT.,, *I+ \\t -.lR7 _ 1 t1mper1ture for 1 Nlctor YftMl Ii) "!" means the init:al refe:~:-:.~ m1t1riel 11 deftned in the ASME Code. tcrr.;:>erature (RT ~DTl of :he uru~~,.:,..: hr1.,.,tl ~Z331.1T..,me11111 the marendl measured 11 def:r.ed ~ *-~ rt!erence tempentm'I 11 ad;u.tld for ASME Code. Paragrap~ :\\=B-:Jj~ the 1ffec:ta of nntrun radlitian far the Meuured uluet mu!l be t:sed r period of llniCll in qu9ttion. credible values are 1vailab:e: :f ~.*~ * (5) "RT,.." 11111n1 tt11 Nflf'lftee f 1Howir:i scnenc meiin vai'JI!! r"." ** * ~~ ter.lJ)tt'llVI mlc:Wlttd bJ the t111thod used: O"F for wel'1s m-tde w*:~ L. - ;- **' liVID ia,.,.....ph (bl(ZJ of t.bia MCtion nux. and -56'F for weld1 rr:ace "* * - for 1111 11 1 ICl"lllliftl critlrian. Linde 0091. 1092 and 124 and.~.? *:: s 3-5 welJfluxes.

Jj Rec;

• 11:*em~nt8.

[l] F:Jr each ;:ireuurind water nuclt:ar power reac:or for 1A*h1c!'l 1n operating l1r:ense has been iuued. the lir:ensee ~h~ ii !ulJm: t proiected values of RT,., '*Jr :eactor *~s&el belthne materiiils by >J'\\':"q val.Jes :*ur !!'IC! nme of subm:ttal. ~ *r.e !!'rmanun date of the operittn; "' !icense. t/':e.projected ex.p1r111tion date if ~ d cha:1se 1r. the operil1n3 licen91 has "" been r':lquested. and the projected

fl 1:.,;-i;ra:1on J4te of i renewal term if a

ct;ut!~: for l:cense renewil bas been*

J! 1mu:ed. T!'le usessment mull use t!-.e ... :~_!drive p~ocedures 91ven in

r

• -il~rdph {Li*(~J of th1s 1ection. T!'le

.,~,'*>srr.er.: mi.st sriec:fy the bises for 1:-:1* prorec11or.. :r.cl-.dins the 50-47 (ii) "M" mean1 the mar;~:-. :J ~e,.. to cover unceMiin:1e1 i:t t=:e *.,. *, '

r.1liil RT~oT* copo;ier ar:d :-.:.:..-!

contents. fluence and !he ca.~.*,. -. procedures. In Equauon 1. ~! :~.;., F '

• welds and 4a"F for base me*a1

\\'ilues of l 111re ustd. and M ~ 'il'l : welds ind ~*F for base :ne:d:.~ musuntd vali.;es of rare '.lsed (iii) ~RT"" 11 the ~ean,.d.-~.: aJjustment in refor:mce te~.~c:, *. ** c..au1P.d by irrl!diauon and s:-..i _. ~

• Ciilci.olated as foilows:

EqulliOn Z: ~RT..n,..tCF;r*.*. " * (i\\*) CF ("F) i1 lhe chcm1w* ',,

• fonct:on of copper and r.:cJ.ei ~~~ ** *
• CF i111iven 1n table 1 ~or we.~i r:

table Z for base metdl :;:!ate1,- : June 30. 1993, *uen

PART SO* DOMESTIC LICENSING OF PROOUCTION ANO UTIL.IZATION FACILJTlES ~orgm.gll). Llnea.r interpolation i1 itted. In Tablea 1 and Z "'Wt-' per" and "'Wt-" ruckal" a.-. U,t be5t* umat1 valuea for tht matetial. which N1il normally bt the mean of the me31ur9d valuet for a plate or forgui1 or for weld 1&mple1 made with th8 weld wi~ heat number that matchn the c:itical vesHI weld. If these values &rl! not available. the upper limi:i"i valuH given in :.he matenal specifications to which :.he veuel wu bwlt ir.ay be used. !! :1ot available. conservative eatimatea (mean plus one 1tandard devfauon) bued on generic dat* 1 may bt uaed if ju1tification is provided. I! none of these aitem1tivea are available. 0.3!1" copper and 1.0" n1ckll mu1t be aa1umed. (v) ""r' meant th* beet 11tilnat* r.eutrun nuence. in unit. 0£10 1* n/cm* (E greater than 1 MeV), at the clad-baH* metal interlace on th1 in1id* 1urface of the ve11el at the loc.etion where tht material in qutttion NCeivta the hiahest fluence for tht period of 1trvice in quntion. TA&L.E 1.-CHEMISTRY FACT°"'°" WELDS. "F 0 0..20 0.40 0.111 O.IO 1.00 '-20 TABl.E 2.~HE!wliS"T'l=IV FAC"T'CR ~OR METAi.. °F

.:iooer.

"'-c*.i 'M *, ...,.~ ----- 0. 0.20

• 0 o60
• 0 60
• C QO ' :JO 1 20 o

20 1 20: 20< zo* 20 20 20 0 01 20I 201 201 201 201 201 20 0 02.............. 201 201 201 201 201 201 2D 0 03............, 201 201 20; 2t), 20: 20* 20 0 °'................ 221 291 211 I 2111 ~' 291 211 005............... ; 2S1 J11 311 111 J1:

31.

31 o oe................. 291 J71 J1 J7. J71

i1*

J7 0.01 ********-****I J1 'l!.. 1 '"....... Q 08 *............ ! ]41 '81 51 I 5 !

• 51' 51" 5 !

0 °'*----' 771' 531 511 5ai 5al 511. 51

0. 10.... --~ ~! 51!

95)* SS! 571 57 67 0 11*-**-**---i..... 62! 721 7,1 77! 77: 77 0.12.......... __ 1, 53 ,11 111 7'91 13 "I ll(li O 13.......... *-***( 711 !1!11 91 96 Ml M 0 "********-1 57,. 751. 111 1001 1051 !Oii 1ot

0. 15.* -....

j !! .,~ *I 1101 11§1,, 7! I I 7 0.11... s.&1 10.; 118! rz3! 125; 125 0 17 ********-*-./ Al *1 TTOI 1271 132! 1J!I! 135 0.11... -****--***~ '131 N 11!1i 13*1 1'1 I 1'4' 1 ** ,, 19......... _.j 711 i11 1301 1'2! 1501 15"1 15" 0.20... ___ 1. 12[ '~ 125! "'" 1se1 16",,65 0.21......... -- M 1011 1211 1!1!11 167j 112: 17* 0 22.........*..* -.. IT. IT2. 1341 T!ITI 1781 11911 HM 0.23...... ---***' Ml 1171 1381 1~ 1.. 1 190i 196 i~~~;;~---***l: !?~I ;~:I ~~ j~7!1 ~l: ~ ru 0.21......... --*** 11 I ,:1 18 219 ;;;j 2"Jt o 21................. 12*,.2 16"1 191 2* q 2*;1 2'8 0.30........... **! 1291 !'81 tf7t 1961 2251 2*1' 257 0 31............. 13*. 151 1 1721 1 1911 22912551 291 ~ 032............... 1311 1ul "' 2021 231 zeot 27* 20 201 20 20 ZQ 20 20 ~ 0.21-**-**-*- IUI 1 teoi 205I 23.a Jiii 292 ...!.**_****.**.*_**.*--.*.**! 201 2015-1 20 20 za 20 20 0.3"...... *****-* 1411 1.. 1a . ~ !!!! 2IO 2221 27 27 27 'D 27 ~ 0.35--*-*-** 1~ 1 117 21 241 27Zl - ~ o.Go&.......... *-***i 2*1 oll 5'j !Ioli 5"! s** 5* oo*?!**-*****-**- !...... ~ z !!!,a 0.03............. J .., *1

• 1 611., i 0..31...............

151 17'3 191 211 245 ~ 303 o o5 -1 **... nl eel... 1,

• • 1*

-*,::1 i:>.oe:::::::::~::::::i ;, !12j... 12, iii u. ii 0.31- "1l 1 221

111 0.07 -*~ §32 511 161 151 951 Ml 911 UO.---*-***

175 I Zll 21 ".l.OI.____ 511 IOI 1~ 1oa1 IQll IOI 0.08._._. 11 Nl 115 122 122i 122 ( ) T fy th th l f RT

0. !0........ ---1 el ~7 1, 1331 1351 1;n 3

0 ven It e V. UH 0 l'T'I ~*:i=:.. **-*j ~~ ~ t :: ~: 1 i :~r' :: ~~~.~.=:-?::=...

o. 14**-*---t., ;;i '

1..al 1u 112,. lhall comict.r pI.nt....,..;ftc infarmatton

0. 15.~-

.. I '12~ '.., I 7!.!II 200 tUl caalct aJf9ct the lftll of 0.1*-****--- ~ *I,,!II "" 1111 1 z,, 0.17.... __ Ill "" 151! 1.. 1 2071.. tmbrttdemnt. nm IDfarmattoa indudn

o.,.__

llf 1221 '"'! 1111 2=* m but it not limited to the rnctor venel 0-1*- a 1001 'S 1s1 1* 1 220 Zll operatiftl temperatvl and lm"leillance 0 20**** 1°' 1 I 2*5 .. multi. Results rrom the plant---"11c 0.21...... ___ i H*I ' 1 ' § 251 "l'W"... 0 2:?........ 171 1i21,311 1 217 IUrYtillance ProtJr'UD mall be lntqrated 023**-****--4 101! 1111 1.eoi 1 into the embrittlement estimate if. 02**-*-1 10&l 1211 '.. I 1 (i) Tht plant-specific 1U:Yeillance data o 2s ***-*--****t i 1ol 12' 1"' 1 * " m ha1 btea deemed CNdiblt at dtfintd in o.29 *** *** ********1 ' 131 1301 1911 1 "" 2* 2,., R-*lato-Guide UM ReYt9ion 2. *nd 0.J7 *****-*-*****~ lttl 1)61 155! I 14t1 -

• ct-

"I o.a..-*--*--1 12211311 1eol 1 2'I 2111 294 (ii) Tht RTm valuecha"lff a 2'1........ --; 121 1..,zt 1sc. 1111 m 25i&l 211 S18nilicantly'.' o JO... ********-*i 131 14'1 1u: 1NI 22:ii 2571 290 Any informarion that i1 believed to o J1.. *******-**1 139 1511 1121 199; 22111 290 m irnpl'O'll'* the iccunc:y of the RT"" value 032... :........... ]1 icol iss1 175; 202* 2311 29:11 291 iienific*ntly inall bt reported to the o 33............... 1... 1 1eo1 11111 20Si n.i1 261i m o Jt...... *****-**i "' 1M1 1MI 2001 mi* m 1 :J02 Director. Offtce of Nuclear Reactor o 39.. *******-****! 1531 1111 1111 21212*1 2721

• Resuletioa. Valun of RT"" th*t havt

~*~*:::_:::~********i :~51 :~1 ;~;~I~:::~~::~ 0.31......... _ 1 192 2001 ml 2501 21 3!6 I Cha11p910IT,,..**lunarweo1111dl11a 0 31......... __ 1 71 11151 2031 2211 ZSill 2151 317 111niLC8111 If lilMr lbuU. dllt-Md in .a.. __ 175 Ill!. 207! 231 j 2S7'1' za1.* l20 p111qrapll (bllZI ol tti. -- ot IM alt'"'81e I

• • 1... deWSillold...,.,..,.,a lb~I ol !&a MCllCft.

or boU1 val-1iu:Md Illa llC1ftainc cnllnOtl. pnor ' Cara frOlll rwactor *-la !*llncalld 10 Ula.,.,_ m*tmal si-c1ncauoa 111 tlla.. - lilop 11 Ule

• nae!.a qunuon and 111 llM.. -

!!ma pat!Qd i1.an "'""'"* o1 **.....-:.da1a. - 10 1hct "ll'N"°" of th*Olll9f'letll 1"2n*. 1ncludi"1 on~ ren1wcc ""* 111pphut1l1. for 1h1 planl been modififi<i Ull"i tl!e procedurn of th.is p1ragrapl'I are 1ubiect 10 the approval of the Director. Office of ~;;ciear Reac:or R!11Jlation when 'J~ u provided 1n th11 aection. (; l F'or eac~ :irnsurized war er n:.iclear power reactor for whrch the value of RT,.,, for any material in the beltline is pro1ected to exceed the P"TS 1creen1ns cntenon before the expirarion date of the operat:ng !icense. or the proj~ted e~p1:-anon data 1f 1 char:~ in the lice!'lse r.u been :-equested. or the end of a renewal term if a :-equest for licer.se re!':ewal has been submitted. the licensee shall subnut by March 16. !99Z. an analy111 and schedule for implementation of such flux reduct:on programs 11 ue re11onably pract:cab!e to avoid exceedinl the PTS 1creen1:'lg criterion set forth in paragraph (b)(z: of rhrs section. The 1chedule far implementation of flux reduction meuures may take into account t~e schedule for 1ubmittal and antic: pated Commission 1pprov1I of deta1lf'd plant* specific analylft. 1ubmitted to demoaatratt accep~ble nak 1t values of RT"' above the acreening limn due.10 plant mcdificahoru. new inforrr~uon or new 1nalysi1 techniques. (SJ For each preuurized.water nuclear ~ power l'llctor for whicll tha aralyli1

• required by paragraph (b)(4) o! this - -

section indicates that no reHonably It prectic*blt nu Nduction program w1il

8 prevent the valut of RT,,. from excttdins tht PT'S ec:reenins crit&non before the expiration dat1 of the operating license; or th1 projected upirauon date if 1 chana* in the operatin1 licenn h11 been ~quested. or the end of a renew.I term if 1 request for liceDN renewal bu been submitted.

the licenaH 1hall 1ubmit a safety an*ly1il to determiu what i! any. modiftc:.ttom to equipment. systems. and operaUoa are neceuary to prevent potential failure of tha reactor ve11el aa a l'ltult of postW.ted PTS event* il continued operation beyoad tha screenina criterion la allowed. In the analysia. the llcemea may determine reactor veael m.alariala propertiu butd on available information. re1ean:h re1ulta. and pla.nt 1W"Veillance d11ta. and may uM probabilatic !ractu:e mechanic:a teclutique1. This a:ialysis must be submitted at le:ist 3 years before th1 value of RT rn is projected to exceed them ICZ't!ertina critenon orb] one year an.: the effective date of li::s 1mendment; whichever i1 later. (6) After consideration of the licen1ee's 1nelysn (incl~ding etfects of propoaed c:on'9Cttve actions. if any) submitted in accordance wtth paragraphs (b)(4) and (b){S) uf this sect:on. the Commj11jon may. on 1 cue-by-c11e basi1. approve oper111on of '.r.e fllcility 1t valuea of RT~ in exce11 of the PI'S *c::ftDU11 critenoc. n.

C.:im:r.:ss:on will cons1dtr 'ac!crs s1sn:ficar.tly affeciin!I !~e ;iotent.Jal for fa.. "re oi the reilctor vesael "'" l"l!ac~:ns a c~c1s1on. (:-1 !~!!le Co~:niS11on conc::~d~s.

1..

s~nt to paragraph (bJ(Sj..ii tr.is -

st:c::ion. that operallon of the far..litv at valul!s of RT~ 1n exc1111 of the PTS sc:-eenmg cr:tel"'!on cannot b1 a:;iprove<.I on :.'le bu11 of the Ucenae1*1 aJ1aiysl!I submmed an 11ccordanc:11 wuh Rl pa~ag:apns ib![~i ar.d *b)fSJ of '.!':is

f section. t!-.e licensee shall requt'!sl and i!."!!Le1ve C;.irnmiuion appro11al prior to a~y operauon beyond the criterioa. Tb.e request must b1 bated upon mod1ficatioa1 to equipment. s~*stems.

and operatum of tha facility ill addi t1on to thoae prevLOualy pro110Nd in the 9ubr:ittted analyses tha1 would reduce thl! potcn11al for failure of tht! reactor 1 t::~s.. I d11t! !ll P'l'S even IS. or. upon

• urther analys111 based upon new 1:orm11uon or improved mct:wJol.,gy

~ !O.U Atq~ems fOf rWductJon of r1'i11 Ira,.. 11.tleitwfecl tr..it*tr.a wrt,,out 1..;r*m (A TWS) 1Venta rc.r ligtlt*....,<Oalecl nuclHr ~

IM:"ltS.

(a).J.;;:lkc!::i:ty. The reqai*'!ment! r.Jf

n:s :1ect!on a~*piy to ~u c~-,meirctal 1!:;'11-w<:ti!P*r.oc.IPd nuclear power ;Jl,1a:s t~*i De'!nir'"n. F*ir p*~=l=-'!:H of th!9 Sf'C'.to~.. ".\\rHi.:!pat1d Tran*ie:tt \\\\'d!.,,ut s.. **ijr!t [Ar\\'.3) muns an ar.11.;:i:-?:e:i cperat1onal occurnnc1 H dP.fir!ed !n 1\\pi:icndix A of this part fo!loweri lly th1 failure of the rea!=tor trip portion cf the prcter.ti1'n sy~tem specifiitd in C.:neral Outsn Criterio:t zo of App1ndi~ A '!r this part.

(cl R~qui~m~nts. (ti Each pre11urized ~aler reactor must h1111 equipment from sensor output to final actuation device.

that is divel"M from the reactor trip

f. system. 10 automatically initiate the

auxiliary (or 1merpncy) feedwater

. "' 1ystem and initiate a turbine trip under ~ condil1on1 indie11tiv1 of an A TWS. 1bi1 equipment mutt be detaped to pnfonD ill function in a r.liabl1 manner and be independent (from 11nsor output to the final actuation device) from the *aialinl reactor trip 1ystem. (ZJ Each pNllllriled water r.actor

• manufactured by c.butiaa-Engineenna or "7 ll'n ~ ud Wilcox must ha111 1 div.,_,....,stem from the sensor outii"'ltlflillrruplioaof power to !hi cona9t:~*Tlsil acram s~stem mua_t be dftillled to perform ill f1onc11on in a reliable manner and be*

independent from th* 1x11t1nt reactor tnp system [from sensor output to interruptton of power to the control rods). (:I) Each boilin1 water !"lactor must have an altemat1 rod injection (ARIJ system that ii div1r11 (from th1 reactor trtp 1ystem) from sentor output to th1 fin~I actuation device. The ARI system must have redundant 1cram 1ir header uhaust valves. The ARI mull be 4

designed to perform its function in a

• ~ reliable manner and be independent

(from the l!J11st1n@ reactor tnp !yStl!m)

~ from sensor output to the final act"Jallon 't_dev1ce. '41 Eac~ ~c.!::-:lj.....-d*ler ?*J~~or :"...;st h,;ve d ;:<.1nr.iby,,qu1J c;onr~u1 s.,:~m !SI.CS) w1!h the ca;iui.rn1ty ui.n*:,..c::n~ into the r~;ir.lur pressure vc~~el d borated w.ilr.r 9olu11on ;it **1ch J now rare. \\evc1 of boron conc.. n1ra11on.ind boron*lO isotope enrichment. and accounting for reactor preuure vessel volumt!. that the resulting reactivity control ;sat least equivalent to that resulting from 1n1ec11on of 86 gallons per !i minute of lJ weight percent sodium !:! pentaborate decahydrate solution at the ff natural boron*lO isotope abundance 1n10 I ZSl*1ncb inside diameter 1'91C!Or Z pre!Sure ves!t!l for a given core design. The SLCS and 111 injection location must be designed to perform its function 1n a reliaLle manner. The SLCS 1n1tiii11on must -be automauc and must be des111ned to perform its function in a reliable manner for plan!I granted a con1truc!lon permit afteP July ZS. 1984. and for plants sranted a construction permit pnor to July ze. 1984. that have 1tlready been designed and built to include thi1 feature. r (!!) Each boilin1 water reactor muat ! have equipment to trip the rHctor I coolant.recirculatin1 pumps 1utom111ically under condition* 1£ indicative of an ATWS. Thia equipment mutt ba dnianed to perform ill function Lin a reliable manner. (&J Information sufficient to demon1trat1 to th1 Cornmi11ion the .adequacy or item* in para1nph1 (c)(l) throusb (c)(5J of thia section 1haU be submitted to the Commi11ion u 1pitcified in t 50.4. (d) lm;Jl*mentation. By 11D dayt after the iHUIDc:e o( the QA pidance for non... feti l"llated compoaeata. **cb a llCllllft 1haU dtlv1lop and 1ubmit to tbe a Commiuioa. H specified iD I 5CH. a

• propoeed achedule for mnUnt the

~ Nq11irement1 or paraan11taa (cM11

lhrouab (c)(5) ol thia MCUon. Each tball indude an explanation of the 1ehedul9 alona with a juttificatioa if th* achedllle calla ror ranal impl1m1ntatioa laler thu the HCOnd refu1lin9 outa1* after f uJJ 21.

1914. or the date or i11uance of a lic:eue authonzinS ope,..tion above S percent of full power. A final schedule 1hail thea be munaally a.,eld upon by th* _ Commi11ion and licen.... -r........_........ -...... I,...* (a) lfMtuirettMtt'-- (ti Each lisht* !:; water-cooled nud1u power plant "' liceued to 01'9f9tl m111t be able to ' ~ with1tand ror a apecif\\ed duratioa and .., l"ICOver from a 1tation bleckout u ..., defined in I 50.2. The 1pectfted 1tat1on blackout duration 1hall be baled on the folluwin9 factors: (ii The redundancy of the on1it1 1mefll1ncy ac power so~e: (11) The M!h1bal1ty..:if rhe cn,ne eme111ency ac. power 1oun:n: (111) The expected frequency *Jf JH -:f off!1te ?CJWe~ and [1v) The probable time needed :o M!store ofC.111 power. IZJ The reactor core and 111oc1ated coolant. control. and protecuon svs1ems. including 1tat1on b11t1nn end 11ny other ,necesHry support 1y1tem1. mutt provide 1ufficient capacity and cap1b1l11y to ensure that the core is cooled and 11ppropn1te containment 1ntegr11y !S ma1nta1ned 1n the event of a s11111on blackout for the spe!::fied durauon. The capability for cop1r.~ w>1h a 1tation blackout of spec:fied dJra11on shall be detemuned by an appropr.i'e cnpinl analys11. U11li11e1 are ex;iec:eo *o have 1b1 baseline 1uwnp11on1. 1n1ly111. and related 1nform111on ~'ed in their cop111f ev1l:Jatlon1 1vaiiab1e '.r:r NRC review. (bl Limitation of scope. Para@raph 'cl of this section does not apply to !hoH plant1 licenlld to operate pnor to fuiy

21. l.9M. 1l the capability to wtthsrand 1tation blackout WH 1pecifically addnued in the operating license proceedina and wH explu:11ly approvl!tl by the NRC.

(c) lmp/t1mt1ntation.-4,1J lnformat1or1 Submittal. For each lilftt*wallr~aoll!d nuclear power plant licenHd to operate on or befoN /uly 21. 19111. the licen1N shall 1ubnut the information defined below to the Director of th* Office of

Nuclear Reactor Resulat1on by.~pr

l 17.

I 19111. For Heft light-wat1r-cooled _

"' nuclear power plant licen1ed to operate

after the 1ffectiv1 date of thi*

1mendlll1nL th1 licen111 shall submit the information defined below to the Dil"ICtor by Z70 day1 1fter th1 date of licellM iNUanc

e.

(i) A propolld 1tation blackout duiation to be ulld in de11rm1n1n1 compliance with parasr1ph (*l of th11 11etion. indudin1 a 1u1tifte1!ion for the 1electioa baled on the four f1ctor1 identified in parasrapb (a) of thi1 llCtiOG: (ii) A dncnption of the procedurl!I that will be u11pl1mented ro. station blackout natl for Ille duration determined In parqnph (c)(1 l(iJ of tlm nction and for recovery therefrom: and (iii) A li1t of modiflcallon1 to equipment and 111ocia11d proced1..~!s. :f any. necessary to mHt th1 requ1~~-!nt1 or para.. ph (aJ or thil MCllOn. f:ir :he 1peafild 1tation blackout dura11on d1t1111uned in plHIJ'IPh (cJl1)(1J of !!'111 section. end a propo1ed schedule '.or 1mpl1ment1n1th*111u~d modificJ ".o:-:s. (2) Alt11mat11 ac source: The di *e r:ia

• e ac powlt' 1ourcel9). 111 defined 1n i 5.J :.

will constitute acceprable capJlid11y :.:i_ with1tand 1t1tion bl1ckou1 prn~1Jed an analysi1 11 performed which d1monatnt11 that th1 plant has 1h1s c1~.1bility from onaet of the st;it1on blackout until rhe 1ltemate ac rn*Jr:e* s' and Nqull'td 1hutdown equ1pmer.! a~e 1tuted and lined up re opera ti! T'-e * -:- e requiNd for111tn1p and alignment *i **e Jun. 30. 1993 (reset)

4.0 PROJECTED RTns The.following describes how the PTS reference temperatures are determined for each of the Palisades reactor vessel beltl ine materials and includes projections for when each material will eice~d the applicable screening crHerion. The results are dependent on ~~e best-estimate ~alues 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~*s va 1 ues. 4.1 Determination and Proiection of the PIS Reference Temceratures Tht base equation for the PTS reference tt111p1r1tur1 from 10CFRS0.61 is: ( 1 )

• i-is defined u the initii.l reference t111p1ratur1 (RT.T) of the unirrad11t1d material. *1* values for the Palisades reactor*vessel b11t1in1 materials are:

Axhl Weld Cfrc Weld Platt 11

• -s6*F } *Generic Value 10CFRS0.61 (b)(2)(1) for Welds made le * -s6*F. with Lindt 1092 and 124 Fluxes Value* r1porttd tn Rlf1renc1 6. This represents th* l i*1ttng phtt.
• A less conservative value of I, * -1o*F was m.asurtd by Battellt Columbus Laboratorits tn 1977 (R1f1renc1 39). A value of -s*F was used in CPC~'s 1986 (R1f1renc1 16) and 1991 (Reference 1) PTS submtttals. Confirmation of

-S'F could not be found by measurement or calculition. 4

• l

"M*,is defined as the margin term added to cover uncertainties as., the vilues of initial RT11rs (Cu and Ni content, fluence and the calculitional procedures). Values of "M* for the Palisades vessel beltline material are: Axial Weld Circ Weld Plate M,

• 66° F -1 Mc "' 66° F Ml'
• 34"F Value specified in 10CFRS0.6l(b)(Z)(ii) for ~elcs if generic values of "I" are used.

Value specified for-base metal in 10CFRS0.61 if measured value of

• I
• i s used "ARTPn* is defined as:

ARTl"t'S

• _ (CF) f (o. 21 - o.10 l09 t1 "CF*, the chemistry factor, a function* of Cu and Ni content, is derived fro11 Tab 1 ts 1 and 2 of 10CFRSO. 61.

In Section 2, the ch.. 1stry factors were d1t1rmintd to be: CF.

• 211*F for the axial welds.

CFc

• 22a*F for the circu*f1rent111 weld.

CF 111

• l&S*F for the v~ssel plate 111t1rh1.

.,. is the blst-1st1.at1 neutron flu1nc1 1n units of 1019 n/c"'" (E > l MIV) at the clad-base metal int1rfac1 of th* vessel. I 4-2 ( 2)

The 1 imiting f1uence is determined by setting RTPrs equal to the screening criteria and solving for f. First, rearranging equations ( 1 ) and ( 2) : R:' P"!'S = I

• M * (CF) f *o. 211 -J. i::i :09 !l

( RT - I-M) ( o. 28 ".' o. :.o log f) log f

• log P'!'SCF o.10 (log f) :z - o. 28 log f
• log ( RTP':'Sc; I-M)
• a Using the quadratic equation to solve for log f:

0.21 * ~ (0.21>2 - ' (0.lO) log (RT,,. ;PI -M) log f * ----.:.-----2-(-0-.-1-0-) __....__.......;;.;... _ ___., Because the positiv1 root of the equation provides 1111n;ngless results, the equation may be simplified to: [ 0

• 28 -

~ 0. 07 u f

• 10 exp 0.2 f"!... 1.... allowed values of RT,T1 is defined in 10CFRS0.6l(b)(2)

'o*.*acb of the Palisades b11t11n1 is: Ax11l Weld Circumferential Weld Pl ate M1t1rhl RT11TI*

• 27Q*f RT"1c
• 300*f RT11TSp
• 27Q*f _

4-3

Cyde 2 3 4 s 6 .7 8 9 2 3 4 s 6 7 8 9 T.ll:ile ~~ <Cominuedl -*

• P:ilisades Fast Neutron Fluence <E > l.O ~feV) Through Cyde 9
• _At the Reactor Vessel Clad-Base ~etal lnterf;u;e Cycle Cycle Cycle CumulJLive Length Flux Fluence Fluem;e

<EFPD) (nkm2-s) (ry'cm2) 1ry1.:m*i 30 Degrees 379.4 4.43E+IO L45E+l8 1.4.SE+ 18 449.1

• 4.43E+l0 L72E+l8 l l7E+l8 349.S 4.43E+l0 l.34E+l8 4.5 lE+l8 327.6-4.43E+l0 l.26E+l8 5.77E+l8 394.6 4.43E+l0 l.51E+l8 7.:!8E+l8 333.4 4.52E+l0 l.30E+ 18 8.58E+l8 369.9 4.52E+l0 l.44E+l8 l.OOE+l9 373.6 2.21E+l0

7. l3E+l7 l.07E+l9 298..5 l.89E+l0 4.87E+l7 l.l2E+l9 45 Degrees

.379.4 2.81E+l0 9.22E+l7 9.22E+l7 449.l 2.81E+l0 l.09E+l8 2.0IE+l8 349.-' 2.81E+l0. 8.49E+l7 2.86E+l8 327.6 2.81E+l0

• 7.96E.f.17 3.66E+l8.

. 394.6 2.81E+l0 9.58E+l7 4.62E+l8

• 333.4:

2.86E+l0 8.23E+l7 5.44E+l8 ~...,.. 369.9 2.86E+l0 9.14E+l7 6.35E+l8

373.6 l.67E+l0 5.39E+l7 6.89E+l8 298.-'*

l.09E+l0 2.80E+l7 7.17E+l8 8-8

Attachm~nt. \\ Page 8 of '3 CALCULATION OF THE HEAN COPPER ANO NICKEL CONTENT OF WELDS FABRICATED USING WELD WIRE FROM HEAT No. W5214 The followtn9 tdenttftcat1ons and copper content values are from Table I. 1.

1. COPPER CONTENT Sample ldenttficat1on 04463

• IP2-flange 1-0428 HBR2 - Torus Flange IPZ - Survetllance IPJ - Surveillance lPl - Nozzle Cutout HBR2 - Surveillance OCI - Survefllance Weight r. Copper 0.20 0.20 o.1 S9 o.1 sg

0. 20

0. 16

0. l6

0. 15

0. 34 0.285 Tot.al 2.013 2.013 + lO

• 0.201
• Hean Copper Content

2. NICKEL CONTENT Sample Identtftcatton 04494

• I PZ 1-042_

04541 Average of 04577 l 04604 04673 Mtllstone IC 04674 IPZ 3-0421 04... Ml.I Z*072A Q4U1, IPZ1*042A .04... PAL S/G 5*943 04*90- .;_;;t 'Hla2 Torus Flange lPZ. - Survetlhnce IPl - Surveillance 1P3 - Nozzle Cutout HBRi - Surveillance Weight % Content 0.94

1. zo 1.00 1.05 1.12 0.97 0.92 0.99 1.13 0.99 1.03 l.12 1.09 0.66 Tohl 14.21 14.21 + 14

• 1.015 *Hean Ntckel Content

Facsimile Da.. To Company A.EA Tec.baoiOIY I 4th November 1994 John Kneeland Consumers Power Company Facsimile number (9) 0101 616 764 8196 From Address Dr. G. Gage Materials PerformaDce Department: 8388, Harwell L.aboratoey, Didcot. Oxfordshire, OXl I ORA, United Kingdom Telephone 235 434466 Fac1imile number 23.S 432337 Paae Copies: Maa1e: I of3 Neil Irvine, AEA O'DONNELL: (9) 010 l 412 655 2928 TESTING OF WELDMETALS FOR. CPCO ADDITIONAL CHEMICAL ANALYSIS MPD/082

1. 2474 Attlched is a table giving the values for the copper and nickel contents of the repeat, second set of analyses. These values are the averqe values of the three determinations performed on each sample. They have also been nonnalised appropriately based on analylia results for the standards. however they have nor been verified by the section manqei'.

To a.id comparison I have also provided the data from the first set ol samples alonpide. Also lllChcd is I copy of the fax thar I have received from rwr concemina visual namjutian oftbe fnc:ture surfaces of drop weiabt specimens; AQl, AQ2. AJl and 812. Ur4artaawly I will be out of the office tomorrow, Tuesday 15/l l/94, and hence not cO'*C"hh; As such I would propose makin1 a start on testing of the mn1inin1 three ba=WcilCwpy specimens (Weld 8; transverse. Weld A: longitwlin&l and Weld B: louli"Mi*I) ca. Wednesday. I will try to contact you before doin1 so in order to confirm that lllda ta1ing is in keeping with your wishes; Richard Miller indicated thll you would be reviewiq your requirements today (Monday). IF 111EU ARE ANY PROBLEMS W1Tll TRANSMISSION OF T111S f ACS/MILE Pl.EASE RING TEL Na lJ.f 4J4JJl

co co '1" j 1..*. COMPARISON OP CXlfl'£R AND NICK.EL OOtflENTS OF SAMPLES ANAL YSEI> IN SECOND SET Will{ ll{QSE OBTAINED ON TIIE ORJGINAL SAMPLES Sl!COND SET Ni Cu Cu Ni FIRSTSHT Saclioa.... Larp: W 11111 l'A' Wddmau'A' AallX 1.116 0.121 0.141 1.091 Al/llX A&l/Y 1.006 0.110 0.110 1.001 Al/lfY Alll/Z 1.104 0.266 0.266 1.090 Al/l/Z lluiaalbraupl.MF Section through Large

• • • *'B' Wddmc111 '8' Bal/X 1.092 0.191 0.21.S 1.21.S BlnJX MllY 0.906 O.ltl 0.119 l.OIO 0112/Y
• llZ l.O.S1 0.196 0.196 1.098 Blfl/Z Seam........ TRpM Scaion th.-ougb Trqian

'AJSG/A' 'A/SG/A' A/SG/AIJIX l.l9l 0.16.S 0167 l.U4 AJSG/A/2/X A/SG/A/]fY 1.121 0.292 0.291 llS6 A/SG/Arl/Y A/S£1/AIJIZ l.066 0.211 0.278 1.059 AJSGIW11Z SclCliaa danlu&b TRfUI Sca.ioa throup Trqiant 'AISGIB' 'A/SGIB' AIS£ilfflJJX 1.204 0.1'9 0.1.B 1.201 A/SG/8111X AJSG/B/J/Y 0.960 0.219 0.211 I 149 A/SG/BllfY A/SO/B/VZ 1.107 0.221 0.217 I 024 A/SG/B/l/Z ~ duoup Tsqien ScQicNI through T1q)1111 'BISGIA' 'B/SG/A' BISKJIA/JIX 1.256 0.191 0.19.S 1.272 BISGIA/2/X BISKJ/A/JfY 1.292 0.192 0.19.S l.lll BISGIA12/Y B/SGIA/l/Z 0.998 0.204 0.206 I Ill UISGIAJllZ *--- -*----*-- Sa.1im duauP Trqian Seaioli through TlqJan 'BISGIB' 'B/SG/B' BISGIBIJIX 1.111 0.16.S 0.162 1.126 8ISGWl/X BISG'BIJIY I.Oii 0.201 o.:zoa l.ll6 BISG'Dll/Y BISG9fllZ 1.292 0.209 0.209 1.107 B/SG'8/'llZ The *- d.u Clllll1eSpOtlli IO the ~ vlluai ol lk tbnc clclcrmiftllioo performed 0o cada 1W11plc. All rcmlu lww bcca lllbjcacd to the oonnalialioo prooaiwc bwd on calibrMioa wilb........ llp"'imnn, IMNever-.. rar dM: __.. _ ol--.*' 1aaw,a &o be vaificd by..., IDClioo maoaga. G.Gasr; 14th NO\\'Clllba 1994

~ .. c.; ~*.-.-.:~:-:.-

22.

~;: TWI WI TELEFAX T'v\\11. Abington Hall. -"b1ngton. Cambridge C81 6AL. ~K Telephone: *44 (O)i23 891162 Telex.81183 Telefax *44 (0)223 892588 To; Or Gareth Gage Fax Room R1t. TF/ tl:i(<i;° Company-: AEA Tecilnology From: Mr K Bell Dept: Material* Performance Oept: Stn.icturtl Integrity Town: Harwell, Dldeot Date: 14 November 1 ~ country: UK Dept Ref: KB/Ktl/60. 94 Fu No: 0235 ~ 432337 No of pa;H: 1 of 1 PleaH r.i.phone Fax Room on 0223 891162 Ext. 2220 if pages are not receivea or,,. unclear 'MESSAGE

Dear Gareth,

P1lllnl T11tlng of Stum Generator Welda TW1Project120711 The four1~mena (ASA refsAQ1. AQ2, AJ2. and Bl2, TWI refw W01-08. W01.0Q, W01..04, and W02.0. respectiYely) have bMn heat tinted at 2ao*c for 2 l'loura. cooled 1..&lng liquid N2, Crok1n open and examined under a low powered binocular micl'Oacopt (-x10). All four apecimans l'lad oaen 1xten1iV1ly fractured during the Pellini test (more than 80% of tl'11 frac:turt IUrfll~ WU tfnt.d). None of the 1P9cimen1 atiow any gro11 welding d*f9ctl Which migl'lt have inftuenced th* rMUl'ta of the teats. One apecimen (812) hu a amall planar dilcontinuity (- 3 x 0.5mm) on the fractura faoe, Smm 1ub-surf1ce. Two 1peelmen1 (AJ2 and 912) lhowecl some 1vi~nce of 1he1r ~P formation, up to 1 mm wide, on the top (tension) 1urflle1 of th* specimen. I am arranging tor all of the apedmena. broken and unbraic.n (excludi119 W02-01. which Ric:hard Miller took with tiim after t"'* day of titting), to be ratumld to you 1t AEA HatWell. BHt re;arda. YOl.lrl tin~l"liy, Normal 'PM conditions of contract agp~ as app~e

TABLE 6-13 (O>nrinued) CALCULATED FI.VENCE (E>l.0 MeV) 'IHROUGH CYQ.E 10 -* AT 11iE PRESSURE VESSEL CI..Al).BASE METAL INTERFACE Cycle Cycle Length Cycle Flux Cycle Flucnce Cumulative (EPPD) (n/cm2-sec) (n/cm2) Fluence (n/cm2) 30 Degree 1 379.4 4.70E+10 l.54E+l8 l.54E+l8 2 449.1 4.70E+l0 l.82E+18 3.36E+l8 3 349.S 4.70E+l0 L42E+l8 4.78E+l8 4 327.6 4.70E+l0 l.33E+l8 6.11E+l8 s 394.6 4.70E+l0 l.60E+l8 7.71E+l8 6 333.4 4.79E+l0 l.38E+l8 9.09E+18 7 369.9 4.79E+10 l.S3E+l8 l.06E+l9 8 373.6 2.34E+10 7.SSE+l7 l.14E+l9 9 298.S 2.00E+lO S.16E+l7 l.19E+l9-10 3.56.9 l.94E+l0 S.98E+l7 l.2SE+l9 4S Degree 1 379.4 2.98E+l0 9.78E+l7 9.78E+l7 2 449.1 2.98E+l0 l.16E+l8 2.13E+l8 3 349.S 2.98E+l0 9.00E+l7 3.04E+l8 .4 327.6 2.98E+l0 8.44E+l7 3.88E+l8 s 394.6 2.98E+l0 l.02E+18 4.90E+l8

• 6 333.4 3.03E+l0 8.73E+l7 5.77E+l8 7

369.9 3.03E+l0 9.68E+l7 6.74E+l8 8 . 373.6 l.77E+l0 S.71E+17 7.31E+18 9 298-' l.lSE+lO 2.97E+l7 7.61E+l8 10 3.56.9 l.32E+l0 4.07E+l7 8.02E+l8 6-28

Cycle l 2 3 4 5 6 7 8 9 10 11 EFPO 379.4_ 449. l ,349,5 327.6 394.6 333.4 369.9 373.. 6 298.5 356.9 422.0 Palisades Cycle Flux Values at Critical Locations Cycle Flux oo 16° 4.59 6.03 - 4.59 6.03 4.59 6.03 4.59 6.03 4.59 6.03 4.87 6.25 4.87 6.25 2.16 4.89 2.08 3.06

1. 51 2.40 1.42 2.21 E + 10 30° 4.70 4..70 4.70 4.70 4.70 4.79 4.79 2.34 2.00

1. 94 1.66 Values for cycles 1 through 10 are from WCAP14014.

45° . 2. 98 2.98 2.98 2.98 2.98 3.03 3.03

l. 77 1.15

l. 32 1.09 Values far cycle 11 are from Palisades in-house calculations.

Mr. John Kneeland Consumers Power Company Palisades Nuclear Plant 27780 Blue Star Memorial Highway. Covert, Ml 49043 Jl* 1111 _,.1\\191 November 15, 1994 R YG-94-089

Subject:

Palisades SG Upper Shell Long Scam Fabrication Technique;'

Reference:

ABB Letter P-PENG-94-022, Chemical Analysis of SO Weld Seam Samples and Materials Consultation (ABB CENO Proposal No. 1017-840-019-A) dated N ovembcr 7, 1994.

Dear Mr. Kneeland:

In support of your efforts to analyze and evaluate several steam generator (SG) long seam welds, ABB CE is providing material testing and consulting services on the Palisades RPV integrity. 1b.ese services are described in the referenced letter. At a site meeting; ABB CE was requested to provide additional steam generator fabrication information. As requested, ABB has reviewed the fabrication data for the original Palisades steam generators. Titls 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 scam 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 chemisny. This method is discussed briefly below. In order to minimize distortion and weld shrinkage, a process was used that incrementally welded each. of the upper shell long seams in sequence. These shell plates were machined with doubl~ U preps. 'The 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 welding 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 wetd seam.

This process was continued until 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 scam was in position. Weld deposit was then pcrfonned on the seam OD using the automatic welding pra<;:ess as described for the first seam. Upon completion of the second weld seam ABB Combustion Engineering Nuclear Power r o R<.. ~'Jo

procedure, the shell was rotated so lhat lhe third seam was in posit*ion. Weld deposil was perfonned 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 mecal and* perfonning a magnetic particle examination, the first weld pass was deposited on the ID of the first weld joint using the automatic welding equipment As wich the OD welds, slag was removed and_su.bsequent passes were performed unlil 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 pn;x:ess was repeated for the third weld seam. A second set of weld increments was 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 secon'd increment of the oo* weldfog process was performed in a sequential fashion. The second increment of the OD weld completed the OD weld. Weld wire was fed into the automatic welding machine from 150 pound spools. Additional wire was added to the weld machine as necessary during the

• entire welding process.

Approximately six 150 pound spools of weld wire were required to fabricate the three upper shell long scams. 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 welding sequence has been des.cribed in this letter to aid Consumers Power Company in its evaluation of the weld deposit chemistry; This description of the welding procedure and the. actual welding procedure are held by ABB Combustion Engineering as proprietary information. This document contains proprietary infonnation and is not to be transmitted or reproduced without specific written approval from Combustion Engineering, Inc. Consiste.nc with the requirements of 10 CFR 2. 790, transmittal of any proprie~ry information provided herein to the Nuclear Regulatory Commission must be accompanied by an affidavit from Combustion Engineering, II¥:. If you have any questions regarding this letter, please call me at (203) 285-2567.

• Sincerely, COMBUSTION ENGINEERING, INC.

~~~ Carl J. Gimbrone Supervisor, Reactor Vessel Integrity

C~DIBILITY OF USING STEA)'I GE~ERATOR WELDS AS St:RROGATES FOR THE PALISADES REACTOR PRESSL'"RE VESSEL WELDS An Independent T edutlcal Opinion by W. L. Server, ATI Consulting ill dcv1?loping Lhe proper materials to use as-surrogates for the Palisades rea~tor pressure vessel (RPV) bcltline welds since no archive weld materials exist (except for the Palisac.i~s surveillance weld), the welds in the retired steam generators wc=re selected as candidates. Th~ pedigree of the welds in the steam jcnerators was dlitermined by ABB-CE, and some of the welds in the steam generator w1.?re fowid to be the same weld wire heats as in the RPV. -Therefore, material was removed from -the steam ienerators in order to further determine the adequacy of the materials as surrogates for measurina copper-nickel chemistry and llse in a supplemental surveillance program. The-Lwo weld wires heats of concern wen~ W5214 and 34B009. The following discussion provides details on the similarity and differences between lhe RPV welds and the steam generator welds as detennined comparing fabrication* information and measured chemistry and mechanical property data. Note that some of the comparison of chemistry and mechanical property data involves data from other sister RPVs __ that have the same weld wire heat in the beltline region and/or in their surveillance program: H. B. Robinson Unit 2 (HBR), Indian Point 2 (TP2), Indian Point 3 (IP3), Salem Unit 1 (Sl), and two BWRs, Millstone Unit 1 (Ml) and Oyster Creek (OC). The fabrication infonnation 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), weldins procedure, and_ approximate time fraine in the same CE shop. (These items of similarity arc aencrally true for the sister RPV welds also.) *1be dissimilarity issues come in relative to differences in the flux lot numbers, post-weld heat treatment_ conditions, and number of weld arcs (related to vessel wall thickness), The effect of different flux lot numbers should be inconsequential, but the effects of post-weld heat 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 ~weld heat treatment is the question of extended time at service temperature and pressure~ Tho steam aeneraton were in service for approximately 8 EFPY at a nominal tcmpcratura of SO<>°F under a pressure of abuut 700 psi. The effect of service aging on these materials needs careful consideration in determining cquivalency with the RPV welds. Tt is kc.own that thermal aging effect,- can arise at temperatures near or greater th..'\\ll 600QF for some ferritic steels, but extended time at sli~tly lower tcml)Craturcs also could play a significant role. The results from mechanical testing should reveal any differences due to No~111b1Jf 17. 199' l NOV 1 7 '94 1 4: 38

inscrvice aging on the weld materials. It should be noted that Charpy V -notch testing of the Palisades surveillance weld a:ft~r thermal aiing revealed a fairly significant difference reflecting time at temperature (5 EfPY at 535°F) as shown in Figure l. The Palisades surveillance weld was fabricated usini a different weld wire heat (3277) but with equivalent flux type and weldini procedures that closely match~d or bounded the WS2 l 4 weld in the RPV; in fact, the surveillanc~ weld baseline curve is very similar to thar of the reported baseline curves for weld wire W5214 from IP2, IP3, and HBR as shoYtn in Fiwure 2. The pieces of steam generator welds were recently tested to determine copper and nickel chemistry, nil-ductility trElilSition tcmix.:.rature (NDTT or NDT temperature), RT Nor> and Charpy V -notch transition curves. There are no known NDTT measw-ements for the WS214 welds, and a value of -80°F for the 343009 has been reported. The measured values for the steam gener~tor materials are -20°F for weld wire heat W5214 and -50°F for weld Vvire heat 34B009. The 30°F difference between the reported value for 34B009 a.od the measured value h1:rc suggests a potential aging effect. Other issues are also important relative to the drop weight NOT temperature determination: the brittle weld startinw bead was fabricated using the latest ASTM E208 specification which was different from that used in the late 1960s through mic.l 1980s; for some materials, this different starting weld bead can result in different measures of NDTT. Additionally, there appears to be a possible effect for the W5214 material relative to where the NOTT specimens were taken from the weld thickness which is somewhat supported by looking at some of the hiiber energy Charpy !i-pecimen results (i.e., -- one region of the weld with the highest copper level tends to provide hijher levels of touihness possibly indicative of a lower NOTT). The RT NDT value for the steam generator WS214 weld is equivalenno the NDIT since the Charpy V-notch data supports greater than SO ft-lb mid 35 mils lateral expansion at NDTT + 60(\\F. 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 shoVvn in Figure 3 where a comparison is made with the combined baseline data from IP2, IP3, and IIBR ill the same manner as shovim in FiiW'C 2. There is a definite shift in the 30 aod 50 ft-lb transition temperatures and a drop in upper shelf energy. The shift dilTerence is upproximately 300F and the decrease in upper shelf is about 10 ft*lb. These differences strongly suggest a thennal aging effect for the steam 8el'.lerator WS214 material. This difference would also maacst the inaccuracy of using the measured NOTT from the steam generator as applied to lhe Palisades RPV, even though the measured value of *20°F is in the upper range: Qf mcwn NOTT measurements for CE-fabricated welds. Charpy V -notch testing of the similar *lteam generator weld Heat No. 34B009 is now underway, and a e<>mparison of these results \\Vith the original (i.e., Wlirradiated and mi.aged) material will provide further evidence of this aging effect. The aging phenomena have been observed in other vessel weld materials as is described in a rccc:nt (unpublished) paper [Ref. l] containing data from the Deel I and TI pressure vessels. Cb.arpy V -notch data from the Doe! I reactor vessel weld in the unaged and BKed condition is sho\\\\on in Figure 4 which clearly indicates the effects of aging on shift in the Charpy curve. Nwrm/Hr J'?. !9fU 2

The USC of the steam iener:itor malerial to add to the data ba:se relative to copper-nickel chemistry shauld be adequate since the bulk measurements are unaffected by heat treatment or service aging considerations. The specific data from the chcmisliy measurements are discussed elsewhere, but the key results show good agreement between prior measurements for the 34BOCJ9 weld, but higher than expc:cted copper values for the W5214 weld. These higher values for weld wire heat W5214 have been factored into an average for all of the measurements for W 5 214 and applied to the Palisades RPV weld. In summary, the Charpy V-nolch and ~1)TT values from the steam generator welds appear to exhibit a service aging phenomenon which invalidaLe their use directly as mea:;ures of the virgin n1echanical properties for the subject welds. It may be possible to thermally anneal these steam generator materials to restore the properties back to Lheir equivalent virgin condition. and this approach should be pursued in combination with micro~;tructural charactcrizaLion work to better-understand I.he embrittlement process. The success of achieving equivalent virgin mechanical properties will be used in further assessing the steam geucrator materials as appropriate surrogates for the supplemental surveillance program for the Palisades RPV. Reference

1.

Gerard, R., Fabry, A., Van de Velde, J., Puzzolante, J. L., Verstrepe11, A., Van Ransbccck, T., and Van Walle, E., "ln-Sen1ce Embrittlcmcnt of the Pressure VeHel Welds at the Doel I and D Nuclear Power Plants," Effects of Ra,diation . on Ma~rws: 17th International 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. 199S (currently under editorial review). No~.:111/Jef' I 7, J 994 3

Weld Seam Location of Weld Weld Wire Hem Numbers F1ux Type flux Lot No.M Thiclmem (In.) PWHT (lloun II& Temp > 1100 F) Table 1 Palisades Vessel Beltline and Steam Generator W "Id F11 brication D-etails 2-112 A/C 3-Ul A/C 1-9!1 A/C Im. Shell Int. Shell Steam Oen. LoIJi. Se<lllll Long. Seams No. 1 W5214+Ni 200 W5214+Ni200 WS214+Ni2.00 34B009+Ni200 Linde 1092 Linde 1092 Llndc 1092* 3617 3692, 3617 3617 8.5 8., 4.75 14.75 14 17.4 1*951 A.JC Steam Gen. No. 2 34B009+Ni200 Linde 1092 3708 4.15 9.5

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