ML20137K250
| ML20137K250 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 01/15/1986 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20137K241 | List: |
| References | |
| NUDOCS 8601230530 | |
| Download: ML20137K250 (4) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION n
-l WASHINGTON, D. C. 20555
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMEN 0 MENT NOS. 74 AND 60 TO FACILITY OPERATING LICENSE NOS. NPF-4 AND NPF-7 VIRGINIA ELECTRIC AND POWER COMPANY OLD DOMINION ELECTRIC COOPERATIVE NORTH ANNA POWER STATION, UNITS NO. 1 AND NO. 2 DOCKET NOS. 50-338 AND 50-339
==
Introduction:==
By letter dated April 30, 1985 (Serial No.85-290), the Virginia Electric and Power Company (licensee) requested a change to the Technical Specifications (TS) for the North Anna Power Station, Units No. 1 and No. 2.
Specifically, the proposed change would revise the reactor coolant pressure-temperature limits, which will be applicable through 10 Effective Full Power Years (EFPY).
Based on the revised pressure-temperature limits, accompanying changes have also been proposed for the reactor heatup limits and the low temperature overpressure protection (Power Operated Relief Valve) setpoints.
In support of the above proposed changes, the licensee submitted three Babcock
& Wilcox (BAW) reports.
They are: BAW-1638, " Analysis of Capsule V, Virginia Electric and Power Company, North Anna Unit No. 1, Reactor Vessel Materials Surveillance Program," May 1981; BAW-1794, " Analysis of Capsule V, Virginia Electric and Power Company, North Anna Unit No. 2, Reactor Vessel Materials Surveillance Program," October, 1983; and BAW-1872, " Reactor Vessel Pressure-Temperature Limit Curves for Virginia Pcwer Company North Anna Units 1 and 2,"
April 1985.
t Discussion:
Pressure-temperature limits must be calculated in accordance with the requirements of Appendix G,10 CFR Part 50, which became of fective on July 26, 1983.
Pressure-temperature limits that are calculated in accordance with the requirements of Appendix G, 10 CFR Part 50 are dependent upon the initial Reference Nil Ductility Transition Temperature (RT materialsinthebeltlineandclosureflangeregioNST)forthelimiting in the reactor vessel and the increase in RT resulting from neutron irradiation damage tothelimitingbeltlinemateYkII.
The NA-1&2 reactor vessels were procured to ASME Code requirements, which did not specify fracture toughness testing to determine the initial RT for each reactor vessel material.
TechnicalSpecificationTableB3/4.4-1$Ncifies initial RT f
NA-1&2vesI0Is.ormaterialsintheclosureflangeandbeltlineregionsofthe The initial RT values for the closure flange region materials weredeterminedusingthemetho$NecommendedbythestaffinStandardReview e60123053086gggg PDR ADOCK O PDR P
. -.. -. _.. ~.
. Plan, Section 5.3.2, Branch Technical Position MTEB 5-2 entitled, " Fracture Toughness Requirements." This method results in an initial RT for the limiting closure flange material of -22*F in each NAPS reactor gssel.
The N
initial RT testing of Nirradiated sample material.f r the beltline materials was determined by frac N
The unirradiated RT values for thelimitingbeltlinematerialswerereportedas38*FforUni5DIand56*Ffor Unit 2.
The limiting beltline material for each unit was the lower shell forgings.
The method recommended by the NRC staff for predicting the increase in RT resulting from neutron irradiation damage is documented in Regulatory Gui$T 1.99, Rev. 1, April 1977, " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials." This method of predicting neutron irradiation damage is dependent upon the predicted amount of neutron fluence and the amounts of copper and phosphorus in the beltline material.
The predicted amount of neutron fluence is dependent upon the neutron flux. The neutron flux is dependent upon the core design.
NA-1&2 began the transition from high to low leakage cores following the first fuel cycle.
The licensee plans to utilize low leakage cores for the remaining life of NA-1&2.
Using flux wire measurements and two dimensional discrete ordinate transport calculations, B & W Report BAW-1872 indicates that the gak ing/sec for NA-1 and 5 7 ide surface fast neutron flux (E<1MeV) during cycle 1 was 6.8 x 10 n/cm 10 2
n/cm /sec for NA-2.
The average peak inside surface fast neutron flux duringthetransitionperiodgomhig/secforeachunith flux leakage to low flux leaka was calculated to be 5.1 x 10 n/cm The equilibrium low leakage core was estimated as core cycle 5.
fast neutron flux during cycle 5 was calculated to be 4.2 x 10Thecalculatedpeg/cm/sec n
for each unit.
The fast neutron flux during the equilibrium core cycle was used in extrapolating the neutron fluence for the remainder of the reactor vessel's life, since the licensee indicates that none of the future power distributions (which might exist for an appreciable length of time) will be more severe than the equilibrium cycle.
In addition, the licensee used the calculated neutron fluence of the NA-1 reactor vessel for NA-2.
This is conservative for NA-2, since the neutron flux for NA-1 is greater than that for NA-2.
Thisneutrongluxagalysisresultsinapredictedneutronfluence y
for10EFgYofg.4x10 n/cm (E<1MeV) at the 1/4 T beltline location and i
- 2.3 x 10 n/cm (E<1MeV) at the 3/4 T beltline location.
f The licensee used the method recommended in Regulatory Guide 1.99 Rev.1 for predicting the increase in reference temperature resulting from neutron irradiation damage.
Tables I and II compare the increase in reference temperature predicted by Regulatory Guide 1.99 Rev. I to that measured from surveillance material in NA-1&2, respectively.
The measured increases in reference temperature for NA-1&2 surveillance materials are reported in B & W Reports BAW-1638 and BAW-1794, respectively.
T
Table I ComparisonofPredictedandggasuregIncreaseinReferenceTemperatureat Neutron Fluence of 2.49 x 10 n/cm for North Anna, Unit 1.
Material Increase in Reference Temperature (*F)
Measured from Predicted by Surveillance Tests Reg. Guide 1.99R.1 Base Metal Axially Oriented Spec.
21 87 Tangentially Oriented Spec.
39 87 Weld Metal 78 53 Heat Affected Zone 55 87 Table II ComparisonofPredictedandggasuregIncreaseinReferenceTemperatureat Neutron Fluence of 2.41 x 10 n/cm for North Anna, Unit 2.
Material Increase in Reference Temperature (*F)
Measured from Predicted by Surveillance Tests Reg. Guide 1.99R.1 Base Metal Axially Oriented Spec.
9 59 Tangentially Oriented Spec.
9 59 Weld Metal 2
46 Heat Affected Zone 10 59 This comparison indicates that, except for the weld metal in NA-1, the method recommended in Regulatory Guide 1.99, Rev. 1, conservatively predicts the amount of increase in reference temperature.
Since the weld metal will not be limiting during the first 10 EFPY of operation, the measured increase in reference temperature for the weld metal will not affect the pressure-temperature limits during this operating period.
Although the applicability of the pressure-temperature limit curves would be extended to 10 EFPY, the safety margins would be preserved by the proposed reductions in the pressure-temperature limit curves and in the associated low-temperature overpressure protection setpoints.
Evaluation:
The NRC staff has used the method of calculating pressure-temperature limits in USNRC Standard Review Plan 5.3.2, NUREG-0800, Rev. 1, July 1981 to evaluate J
the proposed pressure-temperature limits.
The amount of neutron irradiation damage to the limiting beltline material was estimated by using the method recommended in Regulatory Guide 1.99, Rev. 1.
Our conclusion is that the proposed pressure temperature limits meet the safety margins of Appendix G, 10 CFR Part 50 for 10 EFPY and may be incorporated in the NA-1&2 TS.
. j During the staff's review of the NA-1&2 surveillance material test results, it was determined that the future capsule withdrawals would not be in compliance -
with Appendix H, 10 CFR Part 50. The licensee has agreed with this conclusion and has indicated that they will submit revised schedules for capsules to be withdrawn in the future for staff review by April 15, 1986.
Since future capsule test results will not affect the operating limits proposed by the licensee for NA-l&2, a revised capsule withdrawal schedule does not change the i
staff's evaluation regarding the licensee's proposed pressure-temperature limits and associated low-temperature overpressure protection setpoints.
Therefore, based on all of the above, the staff finds the proposed changes to be acceptable for NA-1&2.
Environmental Consideration These amendments involve a change in the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20.
The staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure.
The Commission has previously published a proposed finding that these amendments involve no significant hazards consideration and there has been no public comment on such finding.
Accordingly, these amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 851.22(c)(9).
Pursuant to 10 CFR 551.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of these amendments.
Conclusion We have concluded, based on the considerations discussed above, that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations, and the issuance i
of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Date: January 15, 1986 Principal Contributors:
B. Elliot L. Engle
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