ML20217J589

Safety Evaluation Related to Reactor Pressure Vessel Pressurized Thermal Shock for Beaver Valley Power Station, Unit 1

ML20217J589
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Site:	Beaver Valley
Issue date:	10/07/1997
From:	NRC (Affiliation Not Assigned)
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NUDOCS 9710210188
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, , *Uov y* t UNITED STATES g

2 j NUCLEAR REGULATORY COMMISSION WASHINoTON. D.C. 30636 0001 4

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION I i

RELATED TO THE REACTOR PRESSURE VESSEL PRESSURIZED THERMAL SHOCK W MESNE LIGHT COMPANY BEAVER VALLEY POWER STATION. UNIT NO. 1 j DOCKET NO. 50-334

1.0 BACKGROUND

Title 10 of the Code of Federal Reoulations, Part 50.61 (10 CFR 50.61),

" Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Event:," requires licensees to evaluate their reactor pressure vessel (RPV) materials for susceptibility to failure under pressurized thermal shock (PTS) conditions. This evaluation establishes a value for each vessel

), which is based on a material, combination its of PTSthe reference material's initial temperature propert (RT,,,ies and the effect of irradiation-induced changes on the reference temperature. Since these

irradiation-induced changes cause each material's RT value to increase continuously,10CFR50.61requiresthatcompliancew,Ytbtherulemustbe demonstrated based on projections to the end of the facility's operating license (E0t). Each material's RT value is assessed against the screening criteria temperature appropriate f,o,r,that product form (270 'F for axial

' welds, plates, and forgings, and 300 'F for circumferential welds). The vessel material which has a RT value closest to or exceeding its screening temperatureatEOListhevesseY'slimitingmaterial. Finally,10 CFR 50.61 requires that a licensee's PTS assessment be updated and submitted to the NRC whenever there it a significant change in projected RT,,, values.

The RT,,, value is defined as the sum of:

reference temperature (RT et b) the m(a) ean the unirradiated value of the shiftr.il-ductility in nil-ure ) due to irradiation, and (c) a margin ductility reference tempera [ui), ((ART t,he v term to account for uncertainties in nickel content of the material, the EOL fluence, and t calculat< onal the copser and 3rocedures. The mean value ART is the product of a chemistry factor (CF, sasedonamaterial'scopperan7,nickelcontent,orderivedempiricallyfrom surveillance program tests) and a fluence factor. The margin value is i

(generic or material-dependentuponthemethodusedforobtainingRT*Nc'torcanbedeterminedfrom specific data credible surve)illance data.andAdditional whether or not the chemistry discussion of this evaluational method

. can be found in 10 CFR 50.61 and NRC Regulatory Guide 1.99, Revision 2,

" Radiation Embrittlement of Reactor Vessel Materials" (RG 1.99, Rev. 2).

ENCLOSURE 1

l 9710210188 971007 i PDR ADOCK 05000334 l P PM

i

To support the evaluation of RPV materials, licensees are also requi md by -

IC CFR Part 50, Appendix H, to implement an RPV material embrittlement surveillance program, and the motiodology for evaluation of the surveillance data is cutlined in 10 CFR 50.61. The 'nformation from this program provides a direct comparison between surveillance material embrittlement (as characterized by a measured shift in RT This may allow for the direct determination of a "surveill.,) and fluence.

ance-based" CF and obviate the need to use the generic CF methodology given in 10 CFR 50.61(c)(1)(iv)(A). Further 10 CFR 50.61(c)(2 data to detemine)whether the values given by use of the generic methodology are bounding values. If the surveillance data demonstrate a shift in RT whichexceedsthatproposedbythegenericmethod,thenthesurveillanceTata should be used to produce a bounding material CF value, in previous submittals 1.2) Du the Beaver Valley Power [ Station,quesne Light Company (the licensee) evaluated Unit No.1 (BVPS-1) RPV materials under the most recent revisions of 10 CFR 50.61. Based on these evaluations, the licensee determined that the limiting material in the BVPS-1 vessel (lower shell plate B6903-1 iicense (EOL, Janua)ywould r 29,2016). exceed Theits PTS screening licensee criteria took action before beginning in end of operational cycle 10 as required by 10 CFR 50.6)(b)(3) to implement flux reduction methods and avoid causing the limiting material's RT,,, value to exceed the screening criteria prior to EOL.

In the licensee's current submittal and responses to the NRC staff's requests for additional information (3,4,5), a reevaluation of the BVPS-1 RPV with regard to PTS was performed. This reevaluation was based on: (1 accounting for new information based on the implemented flux reduction progra)m. -(2) recalculation of surveillance capsule and vesssi fluences using updated neutron cross-section data (ENDF/B-VI) and state-of-the-art techniques, (3) a proposed revision of the limiting material's margin term based on a reassessment of the available data on RPV entrittlement, and (4) reanalysis of the surveillance materials' Charpy impact curves using a systematic hyperoolic tangent curve-fitting routine 6). The technical discussion of these changes was submitted in WCAP-14543, " Evaluation of Pressurized Thermal Shock for the Beaver Valley Unit 1 Reactor Vessel".and WCAP-14554, " Beaver Valley Unit 1 Radiation Analysis and Neutron Dosimetry Evaluations" as attachments to references-3 and 4, respectively. While some of these items will affect the RT,f*e,ty sa evaluation report will focus on addressing plate B6903-1 since theevalu RT[t'eB6903-1 assessment.

plI assessments Additional changes fortoall other the PTSvessel materials-are evaluation of other bounded until EOL BVPS-1 RPV materials are discussed in sect' on 4.0.

2.0 DISCUSSION OF THE LICENSEE'S EVALUATION The reassessment of plate B6903-1 submitted by the licensee proposed several These changes can be divided independent changes to the calculation on RT,he surveillance capsule into four main concerns: (1) reassessment of t fluences; (2) reassessment of the peak vessel inside diameter (ID) EOL

fluence; (3) reanalysis of the unirradiated (reference) and irradiated Charpy curves for the surveillance program materials and; (4) an adjustment to the R!m margin term to use " credible" surveillance data. The nature and effect of each of these changes is addressed in the sections which follow, i

2.1 Surveillance Capsule and Reactor Vessel Fluence Reassessment in WCAP-14554, the licensee reevaluated the neutron flux environment in the BVPS-1 vessel. This reevaluation used the updated cross-sectional libraries of the ENOF/8-VI database. In addition, the licensee also accounted for other changes in the opei ional history of the facility which were not included in previous fluence ssments. These changes in the neutron transport code model modified the licensee's previous best-estimate fluence values for surveillance capsules V, U, and W which had been removed from the vessel in 1979 (1.02 ), 1984 (3.58 EFPY (5.89 EFPY), Effective res Full Power The changes Years in(EFPY)he t ca>sules,fluence),

as proposed and 1988 by the licensee,pectively.

are compared in the second and tiird columns of Table 1. The effect of changing the fluences associated with surveillance capsules V, V, and W was to modify the CF derived from the surveillance data. This change in the CF is discussed in more detail in section 2.4.

The licensee's new fluence evaluation also modified the projected vessel ID fluences at EOL. The aforementioned changes in the neutron transport code model to account for ENDF/B-VI cross-sections and operational history data effected these changes in the RPV E0L fluence. Of significance, this new evaluation included changes to the EOL fluence as a result of the flux reduction methods implemented by the licensee beginning in operational cycle

10. These changes resulted in the EOL fluence vessel 10 decreasing from a value of 2.88 x 10,for the limiting material at n/cm (asrecordedintheNRC staff's Reactor Vessel Integrity Databgse (RyID), based on information provided in reference 7) to 2.818 x 10 n/cm.

2.2 Reanalysis of Charpy Impact Curves To assess the embrittlement of the surveillance progrui materials with fluence, it is necessary to examine the shift in the Cliarpy impact curve 30 ft-1b temperature between the material's unirradiated and irradiated states.

Previously, the licensee had relied on hand-drawn curves through the

, unirradiated and irradiated material data for assessing this shift. These I procedures result in the shift values given in column 4 of Table 1. In thie I analysis, the licensee developed new Charpy curves for the unirradiated anu l irradiated tests of the surveillance material by using the curve-fitting program CVGRAPH. CVGRAPH provides hyperbolic tangent curve fits to the Charpy l test data which can then be compared to establish the shift in the 30 ft-lb temperature. The new values determined by the licensee for the shift in the 30 ft-lb temperature for each capsule / specimen orientation combination are given in column 5 of Table 1.

The effect of the change in the shift values was to modify the CF derived from the surveillance data. This CF change is discussed in more detail in section 2.4.

2.3 Adjustment of the Margin Ters to Account for " Credible

• Surveillance Data In the following sections, the rationale proposed by the licensee for outermining a margin tors for their PTS analysis is compared to the margin term as determined by the methods of 10 CFR 50.61.

2.3.1 Current Regulatory Framework The methodolcgy for t.alculating a material's RT m value, which is established in 10 CFR 50.61(c), defines the purpose of the margin term in the calculation and notes how it is to be datermined. The margin term is "to be added to

' account for uncertainties in the valles of RT cup contents, fluence, and the calculational procIclures. copper "

Theand nickel margin term is calculated as:

Margin = 2 /(0,2 , ,,a) where 'u I for DRT isthestandarddeviationforRT,7Ya'teB6903-1is0'FsinceRT and as is the standarsi deviation  !

i The accepted value for o for p l isbasd3on material-spectfic data. uGeneric values of as for plate or forg$3ng en i (17 'F) and weld These values were(developed prior to 1988 from the database which was(used to28 'F establish 1.99, Rev.the 2. embrittlement terrelations contained in 10 CFR 50.61 and RG These values wre based on the la scatter band of the actual embrittlement data available at that time (from power reactor surveillance I programs) from the embrittlement correlations' predictions. These generic values for aa are to be used in a material's PTS evaluation if credible survelliance data do.5 not exist for the material.

If credible surveillance data does exist, as determined by the credibility criteria given in 10 CFR 50.61 and RG 1.99 Rev. 2, the aa for the material is one-half of the generic aa for that material type. One of these credibility criteria requires that when each surveillance datapoint for a material is compared to the best-fit embrittlement correlation to that data (as described by the surveillance-based CF), all datapoints must be within the generic value af aa for that material type for the surveillance data to be credible. For plate B6903-1, this would require that for a plot of the observed shifts in RT vs. fluence, the datapoints from individual surveille.uce capsule must noU7be greater that 17 *F away from the best fit curve through the data.

2.3.2 Licensee's Proporad Analysis In its PTS submittal, the licensee proposed to cetermine whether the BVPS-1 surveillance plate data was " credible" based on criteria not specified in the rul2. The licensee proposed that the surveillance data be considered credible if the data was within 29.1 *F of the best fit curve. The licensee's proposed

&;teria is based on its analysis of an updated set of embrittlement data from power reactor surveillance prograins. According to the licens % this dataset would now include over 200 datapoints for plate and forging ma%-ials. The licensee noted that Lacause additional scatter was present in the datapoints

which have been added since RG 1.99 Rev. 2 was developed, the value of the "lo" scatter band had increased with respect to the RG 1.99 Rev.2 correlation.

Therefore, without developing a new embrittlement correlation, the licensee determined that the "lo" value would be 29.1 'F when all of these data)oints are considered against the RG 1.99 Rev. 2 correlation. The licensee tien used 29.1 *F as their " generic o " with which to evaluate the " credibility" of the surveillancc. data for the plate B6903-1 material. Table 2 has been included in this report to demonstrate the method. Using this information, the licensee determined the data to be " credible" and proposed a value of 14.55 'F for the plate B6903-1 aa , one-half of their generic value of 29,1 'F.

Therefore, by using this value and oa = 0 'F, the licensee proposed that the margin term in their PTS evaluation should be 29.1 'F. The data contained in Table 2 and reported by the licensee indicates that the scatter of the BVPS-1 plate data around the embrittlement function defined by the best-estimate CF is less than 29.1 'F. '

2.4 Change in the Chemistry Factor for the Plate Material B6903-1 By combining the updated surveillance capsule fluence assessments and the revised values for the shift in the surveillance material's 30 ft-lb transition temperature, the licensee determined a new CF for plate B6903-3.

The relationship for using surveillance data to calculate a material's CF is given by Equation 5 of 10 CFR 50.61. Using this method, the licensee determined that the CF for pli.te B6903-1 should be reduced from 167.4 *F (as previously recorded in the staff's RVID) to 163.4 'F.

2.5 Conclusions of the Licensee's Analysis Per the discussion above, the licensee used the following information to determine the EOL RT value for their limiting pl T fluence for plate B6N3-1 was taken as 2.818 x 10" ate. n/cma,he theE0L CF (27.1 basedEFPY) on the l

" credible" surveillance data was taken as 163.4 'F, the RT for the plate I was 27 'F, and the margin term was assumed to be 29.1 'F. E 3ng this data, the licensee determined the E0L RT value for plate B6903-1 to be 264.5 *F.

Therefore,thelicenseeconcluded5.Sattheplate'sRT value would not exceed the screening criteria of 270 *F given in 10 CN 50.61 through EOL.

3.0 NRC STAFF'S EVALUATION The NRC staff has examined the licensee's proposed analysis and does not concur with the fluence values and the i argin term used in the licensee's analysis. The NRC staff does agree that the RT wonu value for plate B6903-1 is 27 *F and that t~ne value of ou - 0 *F is appropr$ ate. The remainder of the NRC staff's evaluation is addressed below.

3.1 Surveillance Capsule and Reactor Vessel Fluence Reassessment The NRC staff has recently examined the Westinghouse methodology for performing reactor vessel dosimetry assessments in detail in conjunction with its review of the updated Palisades PTS assessment. Based on the information

~

supplied by the licensee-(WCAP-14554) to support the fluence values submitted for BVPS-1, the NRC staff has concluded that the method applied for BVPS-1 and for Palisades were the same. Therefore, the NRC staff has arrived at the following assessment based in part on the NRC staff's findings with regard to the Palisades issue, as (scumented in an NRC staff safety evaluation (SE) of December 20,1996[8).

In general, the fluence values submitted for the BVPS-1 vessel and the surveillance capsules in WCAP-14554 were determined by performing neutron transport model calculations and then ad,iusting the results of these calculations based on dosimetry wire analysis and application of Westinghouse's FERRET com> uter code. The FERRET code is a log-normal least-squares fitting routine w11ch is used to adjust the calculated group fluxes in order to obtain a "best fit" to the dosimetry wire data. This application determines a multiplicative factor, the measured-to-calculated fluence ratio (M/C) by which the results of the neutron traasport calculation are modified.

For BVP5-1 in the neutron energy spectrum range of interest (E > 1.0 Mev), the values of M/C for each surveillance capsule were determined as 3,928 for l Capsule V,1.003 for Capsule U, and 0.866 for Capsule W. The capule adjustments were then averaged to determine a value of M/C for 0.932 for vessel plate B6903-1. Therefoie, the calculated fluence values for the ve;sel i

were effectively reduced by approximately 7 percent due to the application of l FERRET and the dosimetry data, i

In the case of Palisades, the NRC staff noted that validity of the modifications to the neutron group-fluxes as a result of the FERRET Iw-normal least-squares fitting routine had not-yet been confirmed by the NRC staff. In fact it was noted that, "A major concern with the application of the FERRET adjustment is that, while the adjustment does provide a- best fit :f the messured data, the dosimeter cross sections, measured reaction u tes and calculated spectrum _ adjustments are made without any physical basis." [8] The NRC staff has not reviewed the FERRET code and does not approve-its use in calculating fluence values at this time. In addition, as was also the case l

with the data from Palisades, a su'oset of the dosimetry data (camely, the copper and uranium dosimetry wires) relevant to measuring the fluence at high neutron energies demonstrated an M/C ratio-closer to unity. Therefore, the-NRC staff has determined that the M/C multiplier of 0.932 cannot be included in BVPS-1 fluence assessment without additional justification and that no adjustment to the calculated fluence values (i.e. a M/C ratio - 1.000) will be used-in the NRC staff's assessment.

Additionally, tn further support the NRC staff's conclusions, a sensitivity study on the-effect of varying the M/C ratio was conducted. The NRC staff examined several M/C values for the vessel plate location ranging between 0.932 and 1.000. The M/C ratior for each individual surveillance capsule (based on each capsule's irradiation history) were proportionally varied in the same way to maintain consistency. Table 3 shows the fluence values which would be attributed to the vessel plate and each of the surveillance capsules as the M/C ratio is varied from 0.932 to 1.000. Each column of the table represents a " consistent" set of capsule and vessel . fluence values.

Additional results are presented in Table 4 and are addressed in Section 3.4.

3.2 Reanalysis of Charpy Impact Curves The NRC staff then reviewed the changes in the RT , shifts which the licensee submitted as a result of applying the CVGRAPH program to unirradiated and irradiated surveillance data. The NRC staff has independently examined the data submitted by the licensee using a version of CVGRAPH provided by ATI Consulting [6). Based on the it's review, the NRC staff accepted the values '

submitted by the licensee. The:e values for each surveillance capsule and specimen orientation are recorded in column 5 of Table 1.

3.3 Assessment of the Margin Term While the NRC staff accepts the fac+ that the scatter in the power reactor I surveillance material database has likely increased as new data has been added, the NRC staff's view is that use of the embrittlement models, credibility criteria, and margin assessments must be considered as a whole.

In fact, in the letter dated April 30, 1993 [9] from G. E. Edison, NRC, to i J. D. Sieber, DLC, in which the NRC staff recommended the use of the 34 'F l margin tt n, the NRC staff noted that this value should be used until a' review j of the updated surveillance database is complete. Until all parts of the '

analysis procedure can be revised, the NRC staff finds that it is prudent to i remain consistent with the basis on which this analysis procedure was constructed.

The NRC staff has examined the licensee's argument and does not agree that the margin term for the PTS analysis of plate B6903-1 should be reduced from the value recommended by the NRC staff, 34 'F, to 29.1 *F. The NRC staff had previously recommended a margin value of 34 'F because the surveillance data war not credible under the criteria of 10 CFR 50.61(c)(2)(1)(C). The NRC staff's assessment of the credibility of the plate B6903-1 surveillance data is demonstrated in Table 5 which shows that half of the surveillance data exceed the criteria and the dataset is therefore, not credible for margin reduction.

3.4 Change in the Chemistry Factor for the Plate Mahrial B6903-1 If surveillance data is not available or not deemed to be credible then 10 CFR 50.61(c)(1)(iv)(A) would require the material's CF to be determined based upon the material's chemical composition (weight percent copper and nickel) per Tables 1 and 2 of 10 CFR 50.51. These tabulated values would be expected to provide an adequately conservai %e Assessment of the material's embrittlement trend. The composition of plats N O3-1 (0.20% copper, 0.54% nickel) would provide for a table CF of 141.8 'F. However, as will be shown in.this section, the "noncredible" surveillance data for plate B6903-1 demonstrates that the rate of embrittlement is significantly greater than that which would be Expected for a material with a CF of 141.8 'F. Therefore, the NRC staff has determined that the use of the table CF for this material would not provide a bounding assessment of the material's embrittlement behavior and

q. ,. .

I would not provide an adequate level of assurance against pressurized ' thermal shock, even when a margin of 34 *F is assessed. Since per 10 CFR 50.61(c)(2),

sic, should be RT ist each vessel beltline material "To verify that is a bounding value... RT,oy [ licensees shall co,,n,s)ider plant-specific information th could affect the level of embrittlement" and since the NRC staff has determined that the "noncredible" surveillance data is, per 10 CFR 50.61(c)(3) l " believed to improve the accuracy of the RT will be used in the NRC staff's evaluation.,,, value significantly," this data Using an M/C value of 1.000, the NRC staff's fluence values as recorded in Table 3, and the a:cepted values for the observed shifts in RT recorded in Table 1, column 5, the NRC staff determined a CF for plate B698-1 from the existing "noncredible" surveillance data. The NRC staff applied Equation 5 of 10 CFR 50.61 for this determination and calculated a value of 159.9 'F for the CF. This is lower than the value determined by the licensee (163.4 *F), as would be expected since the NRC staff's evaluation has assigned slightly higher fluence values to the capsule data while accepting the same value for the shift in RT,o,.

To better understand the impact of the M/C ratio on the BVPS-1 PTS analysis, the NRC staff also performed a sensitivity study between vessel M/C ratios of 0.932 ard 1.000. The fluence values for the limiting B6903-1 plate and each of the surveillance capsules (as calculated by the NRC staff by stepping the M/C ratios in this range toward unity) are presented in Table 3. Table 4 then shows the effect of changing the M/C ratio on the surveillance-based CF, t!e fluence factor, and, ultimately, the value of RT for plate B6903-1. For each value of M/C shown in Table 4, the NRC staffs evaluation demonstrates that the criteria of 270 resulting

• F at EOL. value Ad of RT"dStionally, column I was added to Table 4 tofor demonstrate the overall effect of using the licensee's proposed margin value of 29.1 *F instead of the NRC staff's value of 34 *F.

3.5 Conclusions of the NRC staff's Analysis The NRC staff has concluded that the appropriate values ,for uge in evalaating BVPS-1 plate B6903-1 at EOL are a fluence of 3.024 x 10' n/cm , a CF based on the surveillance data of 159.9 'F, an RT value of 27 *F, and a margin term of 34 *F. WithrespecttothemargEcS3erm, while the NRC staff accepts the fact that the scatter in the )ower reactor surveillance material database has likely increased as new data las been added, the NRC staff's view is that use of the embrittlement models, credibility criteria, and margin assessments must be considered as a whole. Until all parts of the analysis procedure can be revised, the NRC staff find; that it is prudent to remain consistent with the basis on which this analysis procedure was constructed. Therefore, the NRC staff's analysis uses a a of 17 *F, finds the plate B6903-1 surveillance data to not be credible, and , assesses a margin term of 34 *F in the PTS analysis. iissed on this data, the NRC staff has concluded that a limiting E0L RT,,, value for the BVPS-1 vessel of 267.8 'F.

Therefore, altbugh the NRC staff does not concur with the licensee's fluence

methodology or their proposed margin term, the NRC staff does concur with the licensee's finding that the RT 'eria of 10 CFR 50.61 untti remainsbelowthescreeningcrit EOL,value Further, the for the BVPS-1 NRC staff's sensitivity study has shown that the variation of the assumed fluence values between those calculated by the NPC staff and those proposed by the licensee does not affect this conclusion.

4.0 EVALUATION OF OTHER BVPS-1 RPV MATERIALS As noted in section 1.0, the NRC staff's SE has focused on the evaluation of the BVPS-1 limiting material. However, to confirm that plate B6903-1 remained the vessal's limiting material under this evaluation and to document additional information for inclu: ion in the NRC staff's RVID, the NRC staff also evaluated the other BVPS-1 materials. These evaluations were conducted in a manner consistent- with the method used by the NRC staff for plate B6903-1. The pertinent data and results of these evaluations are given in Table 6.

5.0 CONCLUSION

S The NRC staff has concluded that, based on the information supplied by the licensee, plate B6903-1 is the limiting BVPS-1 RPV material for its PTS evaluation and that plate B6903-1 will remain below the PTS screening criteria of 10 CFR 50.61 through the end of the facility's operating license.

Principal Contributor: M. Mitchell Date: October 7, 1997

TABLE 1 +

CHANGES TO BVPS 1 hlRVEILLANCE CAPSULE DATA AS PROPOSED BY THE LICENSEE Capsule - Specimen . Old Fluence New Fluence Old RT. New RT, Orientation (x 10) (x10) Shift (*F) Shift (*F)

' V - Transverse 0.255 .O.316 140.0 137.8 V - Longitudinal 0.255 0.316 130.0 128.1 U - Transverse 0.654 0.690 135.0 131.8 U - Longitudinal 0.654 0.690 120.0 118.9 W - Transverse 0.949 0.915 185.0 179.9 W - Longitudinal 0.949 0.915 -150.0 147.7 TABLE 2 LICENSEE'S DETERMINATION OF PLATE B6903-1 SURVEILLANCE DATA CREDIBILITY BASED ON A BEST-FIT CHEMISTRY FACTOR OF 163.4 'F Capsule - Measured Pradicted Measured - Credible Specimen Shift in RT., Shift in RT., Predicted l Mes. Pred. l Orientation (* F) (0F)

< 29.1 'F V - Transverse 137.8 111.7 26.1 YES V Longitudinal 128.1 111.7 16.4 YES U - Transverse 131.9 146.4 -14.5 YES U - Longitudinal 118.9 146.4 - 27.1 YES W - Transverse 179.9 159.3 20.6 YES W - Longitudinal 147.7 159.3 -11.6 YES

i TABLE 3

• Neutron Fluences (E > 1.0 MeV, in units of 10 n/cm 8) .

as a Function of the Assumed Measured to Calculated Fluence Ratio for Plate B69031 Fluence - Licensee's (M/C) = (M/C) = NRC Staff's item Values 0.95 0.975 Values Based on a Based on a Plate B69031 Plate B69031-(M/C) = 0.932 (M/C) = 1.000 i- Plate B69031 2.818 2.872 2.948 3.024

@ EOL Capsule V O.316 0.322 0.331 0.340

@ 1.02 EFPY Capsule U O.691 0.690 0.68P 0.688 9 3.58 EFPY Capsule W 0.915 '0.954 1.006 1.058

@ 5.89 EFPY -

u II

TABLE 4 -

ANALYSIS OF BVPS-1 PLATE B6903-1 AS A FUNCTION OF THE ASSUMED M/C RATIO 1

(M/C) = (M/C) = (M/C) = (M/C) = NRC staff's 0.932 0.932 0.95 0.975 Values

i. with l (M/C) =  ;

l i

Licensee's 1.00  !

Margin Plate B69031 Fluence @ EOL 2.818 2.818 2.872 2.948 3.024 (x 10" n/cm')

Fluence Factor (ff) 1.276 1.276 1.280 1.287 1.293 Chernistry Factor based on the Surveillance Data 163.4 163.4 162.4 161.1 159.9 fr. .') Table 3 (CF in *F)

ART.,

(CF x ff) 208.5 208.5 207.9 207.3 206.8 (in *F) .

RTmTu (Material Specific) 27 27 27 27 27 (in *F)

Margin (in *F) 29.1 .34 34 34 34 Plate B6903-1 RTn, Value @ EOL 264.5 269.4 268.9 268.3 267.8 (in *F)

RTn. = RTeru + ARTer + Margin 4

TABLE 5

• NRC STAFF'S DETERMINATION OF PLATE B69031 SURVEILLANCE DATA CREDIBILITY BASED ON A BEST FIT CHEMISTRY FACTOR OF 159.9 'F Capsule - . Measured Predicteo Measured - Credible Specimen Shift in RT,ey Shift in RT,.y Predicted Orientation l Mes. - Pred. l

('F) (* F) < 17 'F V - Transverse 137.8 112.4 25.4 NO j V Longitudinal- 128.1 112.4 -15.7 YES U Transverse 131.9. 143.1 - 11.2 YES U - Longitudinal 118.9 143.1 -24.2 NO W Transverse 179.9 162.4 17.5 NO W - Longitudinal 147.7 162.4 -14.7 YES

i- I l

TABLE 6 VALUES OF PTS RELATED DATA FOR ALL BVPS 1 RPV MATERIALS (Limiting material in bold)

Material Chemistry Fluence - Fluence RT,em Margin ART,er RTm Factor Factor Plate 100.5 3.132 1.301 43 34 130.8 207.8-B66071 Plate 100.5 3.132 1.301 73 34 130.8 237.8 B6607 2 Plate- 159.9 3.024 1.293 27 34 206.d 267.B B69031 Plate -98.7 3.024 1.293 20 65.5 127.6 213.1 B7203 2

. Axial Welds 192.4 0.721 0.908 -56 44 174.7- 162.7 19-714 A/B Axial Watcs - 209.1 0.721 0.908 -56 65.5 189.9 199.4 20 714 A/B Circ. Wald 127.0 3.024 1.292- -56 65.5 164.1 173.6 11 714

5 6.0- REFERENCES .

1. Letter from J.-D. Sieber. DLC, to United States Nuclear Regulatory m ission (U.S. NRC), " Beaver Valley Power Station, Unit No.1. Docket Com' No. 50-334, License No.- DPR-66,10 CFR 50.61(b)(1) RT-PTS Submittal,"

December 16, 1991.

2. - Letter from J. D. Sieber, DLC,' to U.S. NRC, " Beaver Valley Power Station, Unit No. 1,-Docket No. 50-334, License No. DPR-66,-10 CFR 50.61(b)(1)

LRT-PTS Submittal," March 16, 1992.

3. Letter from S. C. Jain, DLC, to U.S. NRC, " Beaver Valley Power Station, Unit No. 1, Docket No.-50-334, License No. DPR-66, 10 CFR 50.61(b)(1)'

RT-PTS Submittal," August 2,1996.

! 4. Letter from S. C. Jain, DLC to U.S. NPO, " Beaver Valley Power Station, Unit No. 1, Docket No. 50-334, License No. DPR-66, Response to Request for

- Additional Information Concerning Reactor Vessel-PTS Assessment,"

Harch 14, 1997.

5. x Letter from S. C. Jain, DLC, to U.S. NRC, "Be:ver Valley Power Station,"

Unit No.-1, Docket No. 50-334, license No. DPR-66, Response to second s Request for Additional Information Concerning Reactor Vessel PTS

• l; Assessment," June 5, 1997.

L

( 6. CVGRAPH, Charpy-V-notch Curve-Fitting Routine, Version 4.0, developed by-ATI Consulting, March 1995.

7. Letter from G. S. Thomas, . DLC,- to U.S. NRC, " Beaver Valley Power-Station, Unit'No.:1, Docket-No.150-334, License No. DPR-66, 10 CFR 50.61(b);

Pressurized Thermal Shock,": September 6,1994.

8. Letter from J'. N.! Hannon, U.S. NRC, to T. C.-Bordine, Consumerc Power,

" Palisades: Evaluation of Updated Reactor Pressure Vessel-Fluence Values,"

December 20, 1996.

9. Letter from G. E.-Edison, U.S. NRC, to J. D. Sieber, DLC, " Beaver >

Valle/ Unit'l --Fracture Toughness Requirements for Reactor Vessel,"

April 20, 1993.

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