ML20059N256

Responds to NRC Re Reactor Pressure Vessel Fluence Assessment.Util Believes Startup & Operation Justified for Listed Reasons

ML20059N256
Person / Time
Site:	Yankee Rowe
Issue date:	09/28/1990
From:	Haseltine J YANKEE ATOMIC ELECTRIC CO.
To:	Russell W Office of Nuclear Reactor Regulation
References
BYR-90-128, NUDOCS 9010110105
Download: ML20059N256 (9)

Text

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YANKEEATOMICELECTRIC COMPANY "C"*,?*g*g)ll*,"l" yN W

580 Main Street, Bolton, Massachusetts 01740-1398

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l September 28, 1990 BYR 90-128

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i United States Nuclear Regulatory Commission Document Control Desk 4

Washington, DC 20555 Attention:

Mr. William Russell i

Associate Director for Inspection and Technical Assessment Office of Nuclear Reactor Regulation

?

References.

(a) License No. DPR-3 (Docket No. 50-29) l (b) Letter, NRC to Yankee Atomic Electric Company, dated l

August 31, 1990 (c) Letter, Yankee Atomic Electric Company to NRC, dated i

July 5, 1990 i

Subject:

Reactor Pressure Vessel Fluence Assessment

Dear Sir:

As noted in the NRC Safety Assessment of the Yankee reactor pressure l

vessel (Reference (b)), Yankee committed to preparing and submitting an updated fluence analysis by October 1, 1990.

In a telecon with Dr. Thomas Murley, NRC, on September 25, 1990, Mr. John DeVincentis of Yankee f

Atomic Electric Company (Yankee) reported that preliminary results of the.

updated fluence analysis, which is still under review, indicated a change from the fluence referenced in previous analyses sent to the NRC. The. preliminary results indicate a higher peak fluence-(i.e., 2.6 x 1019 versus 2.3 x 1019 2

n/cm ) and a higher azimuthal variation of fluence (i.e.,

i 1.2 x 1019 versus 3.7 x 1018 n/cm2 at 45').

Mr. DeVincentis reported that Yankee assessed the effeet of the changes f

and determined that the resulting PTS analysis was still within the bounds of i

the NRC Safety Assessment and that startup of the Yankee plant from its refueling outage was, therefore, justified.

Dr. Murley conc,urred with that-judgment.

Mr. DeVincentis also informed Dr. Mur3ey that Yankee would not be able to meet the October 1, 1990 submittal date for.the final updated fluence' analysis.

u because further work had to be performed in order to verify the preliminary results and assess their effects.

In a subsequent telecon between NRC and Yankee on September 26, 1990, it was agreed that the safety assessment performed by Yankee on the preliminary fluence values be submitted on September 28, 1990 and that the final updated fluence analysis would be 0l\\

submitted within 60 days of the telecon.

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United States Nuclear Regulatory Commission September 28, 1990 Attention: Mr. William Russell Page 3 BYR 90-12 8

-The fracture mechanics were run with the limiting thermal hydraulic parameters from all three accident categories (small break LOCA, main steam line break, and transient). The small break LOCA was dominant because it had the highest combination of conditional failure probability and event frequency. As provided in Table 3, the results indicate a conditional failure probability of 2.75 x 10-2 The limiting material is the upper axial weld.

The results are still within the bounds assumed by the NRC in its Safety Assessment, i.e., 10-1 to 10-2 for conditional failure probability. There are some conservatisms used in the calculation for the axial weld which should be taken into account in assessing the results. First, one flaw is assumed in r

each axial weld for fracture mechanics.

The volume of the weld material is very small, and one flaw in the upper _ axial weld is equivalent to approximately 50 flaws per cubic meter.

Second, the axial welds are located between the cold leg nozzles which see higher temperatures than below the nozzles. The beltline area, including the axial welds, was assumed to see the same cold temperatures for the analysis.

Third, the thermal hydraulic conditions assumed were stagnated flows during the entire event which results in the coldest temperatures for the longest times.

Fourth, the axial weld reference temperatures are based upon bounding copper and nickel contents of 0.35 wt.% copper and 0.70 wt.% nickel. A sensitivity study was performed to assess the change in fracture mechanics results if the copper and nickel content were 0.30 wt.% and 0.70 wt.%, respectively. The results show that the conditional failure probability decreases to 7.6 x 10-3 Therefore, the results are highly sensitive to the copper and nickel assumptions.

Based on the results described above Yankee concludes that startup and operation of the plant are justified for the following reasons:

1.

It is very unlikely that Yankee would experience a PTS event.

The design features on which this conclusion is based are not affected by the change in fluence.

2.

The PTS conditional failure probability with the new fluence distribution is within the bounds assumed by the NRC in its Safety Assessment.

3.

The upper axial weld is the limiting material for PTS.

Its reference l

temperature is based upon bounding copper and nickel content. The results are sensitive to the chemistry content, and if lower copper contents are identified, better results are obtained.

4.

The conservatisms associated with the flaw density within the upper axial weld, the thermal hydraulic conditions assumed for the small break LOCA, and the location of the axial welds between the cold leg nozzles would i

provide margin for any remaining uncertainties.

1 Based on the above, Yankee concludes'that reasonable assurance of the public health and safety continues to be provided.

l l

United States Nuclear Regulatory Commission September 28, 1990 Attention:

Mr. William Russell Page 2 BYR 90-128 Safety Assessment A safety assessment was performed to determine the effects of the preliminary updated fluence values on previous PTS analyses. The assessment first determined the 1990 reference temperatures for the beltline materials.

Then, using these reference temperatures, PTS fracture mechanics analyses were performed for each material to obtain their conditional f ailure probabilities.

The reference temperatures were determined using the updated fluence values. The method for calculating the fluences in the axial and azimuthal directions was the same as described in Yankee's Reactor Vessel Evaluation Report (Reference (c)).

The resulting fluence distribution is shown in Table 1.

The reference temperature values for the beltline materials were calculated using the metheds eontained in the NRC Safety Assessment.

I 1.

Upper Plate The reference temperature estimated by the NRC for the upper plate is 19 n/cm and the evaluation of based on a peak fluence of 2.3 x 10 G. R. Odette. Odette estimated the shift in reference temperature to be 245'F.

To be consistent with this approach, the trend curve for Yankee surveillance data has been increased by 60'F as shown in Figure 1 to reflect a 245'F shift at a 2.3 x 1019 n/cm fluence. Application of the new fluence distribution from Table 1 and use of the revised trend curve, Figure 1, results in the reference temperature distribution as shown in Table 2 for the upper plate.

2.

Lower Plate In the NRC Safety Assessment, the NRC estimated that the increased nickel content in the lower plate would contribute an additional 80'F above the-revised trend curve Figure 1, for the upper plate. Application of the new fluence uistribution for the lower plate from Table 1, use of the revised trend curve, and addition of 80'F results in the reference temperature distribution for the lower plate as shown in Table 2.

3.

Circumferential and Axial Welds The NRC estimated the reference temperatures for the circumferential and axial welds using Regulatory Guide 1.99, Revision 2 methodology; bounding values for copper and nickel of 0.35 wt.% and 0.70 wt %, respectively; q

and 50'F for the irradiation temperature effect.

Application of the updated fluence distribution for the welds from Table 1, use of Regulatory Guide 1.99, Revision 2 methodology, and addition of 50*F.

results in the reference temperature distributions for the welds as shown.

in Table 2.

The reference temperatures for each material were-input into the VISA II Code as before. The distribution of input parameters, flaw density using the Marshall distribution, flaw length, assumption of one flaw per beltline material, and number of simulations were identical to previous submittals to the NRC.

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i United States Nuclear Regulatory Commission

-September 28, 1990.-

l Attention:

Mr. William Russell Page 4 L

BYR.90-12 8 r

1 Additional Information f

i

'As requested by the NRC, volume and inside surface areas of the axial welds, circumfert:ntial weld, upper and lower plates are contained in Table:4.

Within 60 days we will submit final fluence values and will revise the PTS l

analysis to account for the final values.

i Sincerely, cb John D. Haseltine

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t Director

. Yankee Project JDH/gjt/WPP77/186 Attachments cc P. Sears (NRC, NRR)

R. Wessman (NRC, NRR)

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Fluence Olstritution for 8ettline Materials Peak fluence et 21.44 EFPY 2.6 e19 tVcm2 A2IMUTHAL YARIAT!ON Axlet Welds 0 to 5 5 to 10 10 to 1515 to 20 20 to 25 25 to 30 30 to 35 35 to 40 40 to 45 40 to 45 l

Upper Plate 10 to 20 0.360 0.376 0.390 0.375 0.332 0.280 0.235 0.202 0.187 0.187 4

20 to 30 1.236 1.292 1.339 1.287 1.141 0.960 0.806 0.695 0.641 0.641 i

30 to 40 1.949 2.037 2.111 2.029 1.799 1.514 1.271 1.096 1.011 1.011 40 to 50 2.222 2.323 2.408 2.314 2.051 1.726 1.449 1.250 1.153 1.153

% cf Height 50 to 60 2.335 2.441 2.530 2.431 2.155 1.814 1.523 1.313 1.212 1.212 60 to 70 2.335 2.441 2.530 2.431 2.155 1.814 1.523 1.313 1.212 1.212 70 to 80 2.400 2.509 2.600 2.499

'2.215 1.864 1.565

'1.349 1.245 1.245 80 to 90 2.357 2.464 2.553 2.454 2.175 1.831 1.537 1.325 1.223 1.223 90 to 100 2.335 2.441 2.530 2.431 2.155 1.814 1.523 1.313

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I cire Weld 2.136 2.233 2.314 2.224 1.972 1.659 1.393 1.201 1.108 j

Lower Plate 35 to 40 0 to 10 2.136 2.233 2.314 2.224 1.972 1.659 1.393 1.201 1.108 1.201

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% cf Height 10 to 20 1.680 1.756 1.820 1.749 1.551 1.305 1.096 0.945 0.872 0.945 20 to 30 0.970 1.014 1.050 1.009 0.895 0.753 0.632 0.545 0.503 0.545 30 to 40 0.168 0.176 0.182 0.175 0.155 0.130 0.110 0.094 0.087 0.094 I

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Mean Reference Temperatures Based on New Fluence calculation i

Peak Fluence at 21.44 EFPY =

2.6 e19 n/cm2 A 2 1 M U T h.* L VARIATlON Axlal Welds O to 5 5 to 10 10 to 1515 to 20 ?0 to 25 25 to 30 30 to 35 35 to 40 40 to.5 40 to 45 Upper Plate 10 to 20 115 118 122 118 108 99

<96 -

<96 175 20 to 30 203 205 208 205 197 185 172 163 157 24C 30 to 40 235 237 240 237 228 210 205 195 188 277 40 to 50

^243 245 250 245 236 225 213 203 197 285

% of Heloht 50 to 60 248 250 252 250 240 230 217 205 200 289 60 to 70 248 250 252 250 240 230 217 205 200 289 70 to 80 250 252 255 252 243 231 218 208 204 290 80 to 90 247 250 253 250 241 230 217 206-202 289 90 to 100 245 248 251 248 240 230 216 205 201 289 i

circ Weld 323 326 328 326 319 308 297 288 283 1

i Lower Plate 35 to 40 1

0 to 10 320 323 325 323 315 303 288 280 275 288

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% of Heleht 10 to 20 304 306 310 305 298 285 275 265 258 273 I

20 to 30 265 257 272 267 260 248 235=

225 220 238 30 to 40

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<176 142 Notes to determine the reference teoperature, en initial temperature of 30F for plates and 10F for welds sust be added to these mean reference tenperatures.

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1 Table 3 l

1 PTS Fracture Mechanics Results 1990 Peak Fluence Peak Reference Conditional Failure

( x 1E+19 n/cm-2)

Temperature -F Probability e

Upper Plate

2. 6' 285 2.8 E-03 Lower Plate 2.31 355 6.6 E-05 i

Circ. Weld 2.31 338 6.8 E-04 Upper Axial Weld 1.24 300 2.4 E-02 Lower Axial Weld 1.2 298 2 E-05 2.75 E-02 F

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Volume and Inside Surface Area of Beltline Materials I

I Volume (ft.-3)

Inside Surface Area (in.-2) 4 Upper Plate 124 250T Lower Plate 46 9470 circ. Weld 3

300 3

Upper _ Axial Weld 0.63 66' Lower Axial Weld 0.24 25 j

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