ML022190261

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Attachment: Dominion'S Additional Information on Reactor Vessel Neutron EMBRITTLEMENT-NORTH Anna
ML022190261
Person / Time
Site: North Anna  Dominion icon.png
Issue date: 08/05/2002
From:
NRC/NRR/DRIP/RLEP
To:
Virginia Electric & Power Co (VEPCO)
Tabatabai O, NRR/DRIP/RLEP, 415-3738
References
Download: ML022190261 (10)
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Text

Reactor Vessel Neutron Embrittlement - North Anna The following information concerning Reactor Vessel Beltline Neutron Fluence, Pressurized Thermal Shock, Charpy Upper Shelf Energy, and Limits for Heatup and Cooldown was prepared in support of North Anna license renewal application. This information demonstrates an ability to comply with applicable regulations governing reactor vessel integrity including 10 CFR 50 Appendix G, 10 CFR 50 Appendix H, and 10 CFR 50.61 during a postulated 20-year license renewal period.

1. Calculated Beltline Fluence The reactor vessel beltline neutron fluence values applicable to a postulated 20 year license renewal period were calculated using the Virginia Power Reactor Vessel Fluence Methodology Topical Report [Ref. 1]. The methodology described in that report was developed in accordance with Draft Regulatory Guide DG 1053 [Ref. 2]. The reactor vessel fluence calculational methodology was benchmarked using a combination of Virginia Power surveillance capsules, pressure vessel simulator measurements, and Surry Unit 1 ex-vessel cavity dosimetry measurements.

The underlying requirement of DG-1053 is that the fluence determination should be made on a plant-specific, best-estimate basis rather than on a generic conservative basis. The methodology used to determine the best-estimate fluence must be demonstrated to have an associated uncertainty of +/-20 percent at the 1-sigma level. This level of uncertainty is consistent with the assumptions made in the development of the Pressurized Thermal Shock (PTS) screening criteria for vessel welds and plates.

Table 1 Calculated Fluence Values North Anna Unit Number 1 2 EFPY at BOLRP 32.3 34.3 EFPY at EOLRP 50.3 52.3 Fluence* at Clad/Base Metal Interface BOLRP 3.920 3.960 EOLRP 5.900 5.910 Fluence* at 1/4 of wall thickness BOLRP 2.446 2.471 EOLRP 3.681 3.687 Fluence* at 3/4 of wall thickness BOLRP 0.952 0.962 EOLRP 1.433 1.435

  • Note: All fluence values are in units of 1019n/cm2 (E > 1.0 Mev) 08/01/02 PAGE 1 OF 10

Reactor Vessel Neutron Embrittlement - North Anna

2. Pressurized Thermal Shock The following values were calculated in accordance with 10 CFR 50.61.

Table 2 North Anna Unit 1 Values of RTPTS at 50.3 EFPY Inter. Shell Lower Shell Circ.

Limiting Materials 990311/ 990400/ Weld 298244 292332 25531 Initial Ref. NDT Temp. (ºF) 17 38 19 Copper Content (%) 0.12 0.16 0.11 Nickel Content (%) 0.82 0.83 0.13 Table* Chemistry Factor (ºF) 86.0 Table* Margin (ºF) 34 Table* Ref. PTS Temp. (ºF) 174.3 S/C** Chemistry Factor (ºF) 88.9 93.1 S/C** Margin (ºF) 17 28 S/C** Ref. PTS Temp. (ºF) 182.5 180.4 Table 3 North Anna Unit 2 Values of RTPTS at 52.3 EFPY Inter. Shell Lower Shell Circ.

Limiting Materials 990496/ 990533/ Weld 292424 297355 716126 Initial Ref. NDT Temp. (ºF) 75 56 -48 Copper Content (%) 0.10 0.13 0.07 Nickel Content (%) 0.85 0.83 0.05 Table* Chemistry Factor (ºF) 67.0 96.0 Table* Margin (ºF) 34 34 Table* Ref. PTS Temp. (ºF) 205.1 227.7 S/C** Chemistry Factor (ºF) 10.4 S/C** Margin (ºF) 14.9 S/C** Ref. PTS Temp. (ºF) -18.2

    • Note: Chemistry factor determined using credible surveillance capsule (S/C) data.
3. Upper Shelf Energy The requirements on upper shelf energy are included in 10 CFR 50, Appendix G. 10 CFR 50, Appendix G requires utilities to submit an analysis at least 3 years prior to the time that the upper shelf energy of any of the RPV material is predicted to drop below 50 ft-lb, as measured by Charpy V-notch specimen testing.

08/01/02 PAGE 2 OF 10

Reactor Vessel Neutron Embrittlement - North Anna There are two methods that can be used to estimate the change in upper shelf energy (USE) with irradiation, depending on the availability of credible surveillance capsule data as defined in Revision 2 of Regulatory Guide 1.99. For vessel beltline materials that are not in the surveillance program or not credible, the Charpy upper shelf energy is assumed to decrease as a function of fluence and copper content, as indicated in Regulatory Guide 1.99, Revision 2 [Ref. 3].

When two or more credible surveillance data sets become available from the reactor, they may be used to determine the Charpy USE of the surveillance material. The surveillance data are then used in conjunction with the Regulatory Guide data to predict the change in USE of the RPV due to irradiation.

Using the 1/4 thickness fluence of Section 1, the values of upper shelf energy (USE) in Tables 4 and 5 were calculated for the North Anna Unit 1 and Unit 2 reactor pressure vessels at the end of the license renewal period being evaluated.

Table 4 North Anna Unit 1 USE Values at 50.3 EFPY Inter. Shell Lower Shell Circ.

Limiting Materials 990311/ 990400/ Weld 298244 292332 25531 Initial USE Value (ft-lbs)* 92 85 102 Decrease (%) 28 34 34 USE Value (ft-lbs) 65.9 56.5 67.8 Table 5 North Anna Unit 2 USE Values at 52.3 EFPY Inter. Shell Lower Shell Circ.

Limiting Materials 990496/ 990533/ Weld 292424 297355 716126 Initial USE Value (ft-lbs)* 74 80 107 Decrease (%) 26 30 28 USE Value (ft-lbs) 54.8 56.3 76.6

  • Note: Initial values are measured.

As shown by these results, the upper shelf energy (USE) values at the end of the license renewal period are greater than the NRC (10CFR50) Appendix G requirement of 50 foot-pounds for the limiting materials.

4. Limits for Heatup and Cooldown Figure 1 presents the heatup curves, without margin for instrumentation errors, for a maximum rate of 60ºF/hour for the limiting material in the North Anna Units 1 and 2 reactor pressure vessel beltline. Note on these curves, that moderator temperature is Reactor Coolant system water temperature. Likewise, Figure 2 presents the cooldown curves, without margin for instrumentation errors, for a maximum rate of 100ºF/hour for the limiting material in the North Anna Units 1 and 2 reactor pressure vessel beltline. The heatup curves of Figure 1 and the 08/01/02 PAGE 3 OF 10

Reactor Vessel Neutron Embrittlement - North Anna cooldown curves of Figure 2 are based upon the limiting adjusted reference temperature (ART) values from Tables 6 and 7; and are valid for up to 50.3 EFPY in Unit 1 and for up to 52.3 EFPY in Unit 2. Since these curves provide sufficient margin on the operating window relative to the pump seal requirements, no additional actions are required for the license renewal periods of North Anna Unit 1 and Unit 2.

Maximum allowable low temperature over-pressure protection system (LTOPS) power operated relief valve (PORV) setpoints have been developed which bound both North Anna Units 1 and 2.

They were developed based on end of license renewal heatup and cooldown curves using the current Westinghouse methodology [Ref. 5]. The setpoints conservatively account for instrument uncertainties and the pressure difference between the wide range pressure transmitter and the reactor vessel limiting beltline region.

The following PORV setpoints, which depend upon the reactor coolant system (RCS) temperature, will provide adequate margin to the North Anna Units 1 and 2 Appendix G limits throughout a 20 year license renewal period with no restrictions on the number of RCPs running:

RCS Temperature PORV Setpoint TRCS < 130ºF 395 psig 130ºF < TRCS < 305ºF 450 psig Table 6 North Anna Unit 1 ART Values at 50.3 EFPY Beltline Materials ART at 1/4 T ART at 3/4 T Intermed. Shell Forging 990311/298244 Table* Chemistry Factor 166.1 ºF 145.6 ºF Lower Shell Forging 990400/292332 S/C** Chemistry Factor 174.0 ºF 152.8 ºF Circumferential Weld 25531 S/C** Chemistry Factor 171.5 ºF 149.4 ºF

    • Note: Chemistry factor determined using credible surveillance capsule (S/C) data.

Table 7 North Anna Unit 2 ART Values at 52.3 EFPY Beltline Materials ART at 1/4 T ART at 3/4 T Intermed. Shell Forging 990496/292424 Table* Chemistry Factor 198.7 ºF 182.7 ºF Lower Shell Forging 990533/297355 S/C** Chemistry Factor 218.5 ºF 195.6 ºF Circumferential Weld 716126 S/C** Chemistry Factor -19.2 ºF -21.7 ºF

    • Note: Chemistry factor determined using credible surveillance capsule (S/C) data.

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Reactor Vessel Neutron Embrittlement - North Anna

5. Reactor Vessel Surveillance Program The revised North Anna Unit 1 and 2 surveillance capsule withdrawal schedules [Ref. 4], which include provisions for license renewal, in the form of footnotes, are provided in Tables 8 and 9, respectively. Dominion anticipates implementation of the recommendation of GALL report for the withdrawal of the final plant-specific surveillance capsules.

==

Conclusion:==

The aforementioned information demonstrates an ability to comply with applicable regulations during a postulated 20-year license renewal period. Required analysis will be performed and implemented in accordance with the requirements of the applicable regulations, and in anticipation of the expiration of affected plant Technical Specifications.

08/01/02 PAGE 5 OF 10

Reactor Vessel Neutron Embrittlement - North Anna Table 8 SURVEILLANCE CAPSULE WITHDRAWAL SCHEDULE1 FOR NORTH ANNA UNIT 1 Capsule Capsule Lead Capsule Status4 Estimated Insert Est. Capsule Ident. Location2 Factor3 Withdrawal EFPY/Year Fluence (x1019)5 EFPY/Year V 165° 1.6 Active 1.1/1979 NA 0.30 U 65° 1.0 Active 5.90/1987 NA 0.88 W 245° 1.03 Active 14.7/1998 NA 2.04 Z 305° 0.69 Active6 16.1/2000 NA 1.48 Z 165° 1.6 NA 16.1/2000 1.48 Z 165° 1.6 EOL/2018 NA 4.64 T 55° 0.69 Standby7 16.1/2000 NA 1.48 T 245° 1.03 NA 16.1/2000 1.48 T 245° 1.03 NA NA 3.52 (EOL)

Y 295° 1.03 Standby7 NA NA 4.24 (EOL)

S 45° 0.55 Standby7 NA NA 2.27 (EOL)

X 285° 1.6 Standby7 EOL/2018 NA 6.59 1

Withdrawal schedule meets requirements of ASTM E-185-82, Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, dated July 1, 1982.

2 See North Anna UFSAR Figure 5.4-4 for original capsule installation locations.

3 Lead Factor is defined in ASTM E-185-82 as the ratio of the neutron flux density at the location of the specimens in a surveillance capsule to the neutron flux density at the reactor pressure vessel inside surface at the peak fluence location.

4 Capsules required to satisfy the requirements of ASTM E-185-82 during the current license period are designated Active. Capsules not required by ASTM E-185-82, but which are maintained for contingencies, are designated Standby.

5 Surveillance capsule neutron fluence estimates based on fluence analysis presented in WCAP-11777, Analysis of Capsule U from the Virginia Electric and Power Company North Anna Unit 1 Reactor Vessel Radiation Surveillance Program, dated February 1988.

6 Capsule X may be withdrawn at EOL in lieu of Capsule Z to satisfy ASTM E-185-82 fourth capsule requirement for the current license period.

7 Capsules T, Y, S, and X are available to satisfy potential fluence monitoring requirements during a postulated 20 year license renewal period.

08/01/02 PAGE 6 OF 10

Reactor Vessel Neutron Embrittlement - North Anna Table 9 SURVEILLANCE CAPSULE WITHDRAWAL SCHEDULE1 FOR NORTH ANNA UNIT 2 Capsule Capsule Lead Capsule Status4 Estimated Insert Est. Capsule Ident. Location2 Factor3 Withdrawal EFPY/Year Fluence (x1019)5 EFPY/Year V 165° 1.66 Active 1.0/1982 NA 0.25 U 65° 1.19 Active 6.3/1989 NA 1.07 W 245° 1.19 Active 15.3/1999 NA 2.58 Z 305° 0.81 Standby6 15.3/1999 NA 1.76 Z 165° 1.66 NA 15.3/1999 1.76 Z 165° 1.66 NA NA 5.82 (EOL)

T 55° 0.81 Standby6 15.3/1999 NA 1.76 T 65° 1.19 NA 15.3/1999 1.76 T 65° 1.19 NA NA 4.67 (EOL)

Y 295° 1.19 Standby6 NA NA 5.50 (EOL)

S 45° 0.65 Standby6 NA NA 3.00 (EOL)

X 285° 1.72 Active7 EOL/2020 NA 7.95 1

Withdrawal schedule meets requirements of ASTM E-185-82, Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, dated July 1, 1982.

2 See North Anna UFSAR Figure 5.4-4 for original capsule installation locations.

3 Lead Factor is defined in ASTM E-185-82 as the ratio of the neutron flux density at the location of the specimens in a surveillance capsule to the neutron flux density at the reactor pressure vessel inside surface at the peak fluence location.

4 Capsules required to satisfy the requirements of ASTM E-185-82 during the current license period are designated Active. Capsules not required by ASTM E-185-82, but which are maintained for contingencies, are designated Standby.

5 Surveillance capsule neutron fluence estimates based on fluence analysis presented in WCAP-12497, Analysis of Capsule U from the Virginia Electric and Power Company North Anna Unit 2 Reactor Vessel Radiation Surveillance Program, dated January 1990.

6 Capsules Z, T, Y, and S are available to satisfy potential fluence monitoring requirements during a postulated 20 year license renewal period. Capsule Y may be withdrawn in lieu of capsule X to satisfy ASTM E-185-82 fourth capsule requirement for the current license period.

7 Withdrawal of Capsule X at EOL satisfies ASTM E-185-82 requirement for EOL capsule, and provide material properties data at a fluence which exceeds that expected to be achieved at the end of a postulated 20 year license renewal period.

08/01/02 PAGE 7 OF 10

Reactor Vessel Neutron Embrittlement - North Anna Figure - 1 North Anna Units 1 and 2 Reactor Coolant System Heatup Limitations (Heatup Rates up to 60ºF/hour With Margins of 0 ºF and 0 psi for Instrumentation Errors)

Applicable to the End of License Renewal Period 08/01/02 PAGE 8 OF 10

Reactor Vessel Neutron Embrittlement - North Anna Figure - 2 North Anna Units 1 and 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100ºF/hour With Margins of 0 ºF and 0 psi for Instrumentation Errors) Applicable to the End of License Renewal Period 08/01/02 PAGE 9 OF 10

Reactor Vessel Neutron Embrittlement - North Anna

Reference:

1. Virginia Power Topical Report VEP-NAF-3A, "Reactor Vessel Fluence Analysis Methodology," dated November, 1997.
2. Draft Regulatory Guide DG-1053, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," June 1996 previous draft was DG-1025, September 1993.
3. NRC Reg. Guide 1.99 Revision 2, Radiation Embrittlement of Reactor Vessel Materials, May, 1988.
4. Letter from J. P. OHanlon to USNRC, Virginia Electric and Power Company, North Anna Power Station Units 1 and 2, Reactor Vessel Surveillance Capsule Withdrawal Schedules, Serial No.98-646, dated December 17, 1998.
5. WCAP-14040-NP-A, Rev. 2, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, January 1996.

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