ML18009A766
| ML18009A766 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 12/26/1990 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML18009A765 | List: |
| References | |
| NUDOCS 9101020379 | |
| Download: ML18009A766 (6) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 e
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO.
TO FACILITY OPERATING LICENSE NO.
NPF-63 CAROLINA POWER 5 LIGHT COMPANY SHEARON HARRIS NUCLEAR POWER PLANT UNIT I DOCKET NO. 50-400
1.0 INTRODUCTION
By letter dated September 10, 1990, as supplemented November 20, 1990, Carolina Power and Light Company (CPAL) submitted a request for changes to the pressure-temperature (P-T) limits in the Shearon Harris Nuclear Power Plant, Unit I (Harris), Technical Specifications (TS), Section 3.4.
The November 20, 1990, letter provided clarifying information that did not change the initial determination of no significant hazards consideration as published in the Federal Re ister (55 FR 40462) on October 3, 1990.
This revision changesatse p-T
>msts from 3 to 5 effective full power years (EFPY).
The proposed P-T limits were developed based on Regulatory Guide (RG) 1.99, Revision 2, and they provide limits for the operation of the reactor coolant system during heatup, cooldown, criticality, and hydrotest.
Each licensee authorized to operate a nuclear power reactor is'required by 10 CFR 50.36 to provide TS for the operation of the plant.
In particular, 10 CFR 50.36(c)(2) requires that limiting conditions of operation be included in the TS.
The P-T limits are among the limiting conditions of operation in the Technical Specifications for all commercial nuclear plants in the U.S.
Appendices G and H to 10 CFR Part 50 describe specific requirements for fracture toughness and reactor vessel material surveillance that must be considered in setting P-T limits.
An acceptable method for constructing the P-T limits is described in Standard Review Plan (SRP) Section 5.3.2.
To evaluate the P-T limits, the staff uses the following NRC regulations and guidance:
Appendices G and H to 10 CFR Part 50; the American Society of Testing Materials (ASTM) Standards and the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), which are referenced in Appendices G and H; 10 CFR 50.36(c)(2);
RG 1.99, Revision 2; SRP Section 5.3.2; and Generic Letter 88-11.
Appendix G to 10 CFR Part 50 specifies fracture toughness and testing requirements for reactor vessel materials in accordance with the ASME Code.
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In particular, Appendix G specifies that the beltline materials in the surveillance capsules be tested in accordance with Appendix H to 10 CFR Part 50.
Appendix H, in turn, refers to ASTM Standards for surveillance testing requirements.
These surveillance tests define the extent of vessel embrittlement at the time of capsule withdrawal in terms of the increase in reference temperature.
Appendix G also requires the licensee to predict the effects of neutron irradiation on vessel embrittlement by calculating -the adjusted reference temperature (ART) and Charpy.upper shelf energy (USE).
Gener ic Letter 88-11 requested that licensees and permittees use the methods in RG 1.99, Revision 2, to predict the effect of neutron irradiation on reactor vessel materials.
This guide defines the ART as the sum of unirradiated reference temperature, the increase in reference temperature resulting from neutron irradiation, and a margin to account for uncertainties in the prediction method.
Appendix H to 10 CFR Part 50 requires the licensee to establish a
surveillance program to periodically withdraw surveillance capsules from the reactor vessel.
Appendix H refers to the ASTN Standards which, in turn, require that the capsules be installed in the vessel before startup and that they contain test specimens made from plate, weld, and heat-affected-zone (HAZ) materials of the reactor vessel beltline.
2.0 EVALUATION To review the licensee's
- request, the staff evaluated the effect of neutron irradiation embrittlement on each beltline material in the Harris reactor vessel.
The amount of irradiation embrittlement was calculated in accordance with RG 1.99, Revision 2.
The staff determined that the material with the highest ART at 5
EFPY at 1/4T and 3/4T (T = reactor vessel beltline thickness) was intermediate shell plate 84197-2 with 0.1%
copper (Cu), 0.5% nickel (Ni), and an initial RT dt of 86'F.
ndt For the limiting beltline material, plate 84197-2, the staff calculated the ART to be 174.1'F at 1/4T and 8.3'F at 3/4T.
The staff used a
neutron fluence of 5.495j8 (E18=10
) neutrons/square centimeter (n/cm
)
at 1/4T and 2.17E18 n/cm at 3/4T.
The licensee used the method in RG 1.99, Revision 2, to calculate an ART of 175'F at 1/4T and 159'F at 3/4T for the same limiting metal.
Substituting the ART of 175'F into equations in SRP 5.3.2, the staff verified that the proposed P-T limits for heatup,
- cooldown, and hydrotest meet the beltline material requirements in Appendix G to 10 CFR Part 50.
The licensee has not removed any survei 1lance capsules from the Harris reactor vessel.
According to the Harris Final Safety Analysis Report (FSAR), the first surveillance capsule will be removed at about 3 EFPY, which will be the next refueling outage.
The staff has determined that all surveillance capsules contained Charpy impact specimens and tensile specimens made from base metal, weld metal, and HAZ metal.
In addition to beltline materials, Appendix G to 10 CFR Part 50 also imposes P-T limits based on the reference temperature for the reactor vessel closure flange materials.
Section IV.A.2 of Appendix G states that when the pressure exceeds 20'5 of the pre-service system hydrostatic test
- pressure, the temperature of the closure flange regions highly stressed by the bolt preload must exceed the reference temperature of the material in those regions by at, least 120'F for normal operation and by 90'F for hydrostatic pressure tests and leak tests.
Based on the flange reference temperature of O', the staff has determined that the proposed P-T limits satisfy Section IV.A.2 of Appendix G.
Section IV.B of Appendix G requires that the predicted Charpy USE at end of life be above 50 ft-lb.
The material with the lowest initial USE is plate B4197-2 with an initial USE of 74 ft-lb.
Using the method in RG 1.99, Revision 2, the predicted Charpy USE of the plate at the end of life will be greater than 50 ft-lb and, therefore, is acceptable.
The staff concludes that the proposed P-T limits for the reactor coolant system for heatup,
- cooldown, leak test, and criticality are valid through 5
EFPY because the limits conform to the requirements of Appendices G and H to 10 CFR Part 50.
The licensee's submittal also satisfies Generic Letter 88-11 because the licensee used the method in RG 1.99, Revison 2, to calculate the ART.
- Hence, the proposed P-T limits may be incorporated-into the Harris TS.
Low temperature overpressure protection (LTOP is provided by the pressurizer overpressure relief valves (PORVs These PORVs are set to open at a pressure low enough to prevent violation of the Appendix G
heatup and cooldown curves should an RCS pressure transient occur during low temperature operations.
The licensee identified the most limiting overpressure transients analyzed to determine the PORV setpoints for LTOP in Amendment 19.
The same setpoints are utilized for the five EFPY.
An analysis was performed by the licensee to ensure that the pressure over shoot beyond the LTOP setpoint is such that the new Appendix G pressure-temperature curves are not exceeded during the transient.
The licensee stated that the new curves, in conjunction with the associated TS changes in the heatup and cooldown ranges and the existing LTOP system setpoints, provide the required assurance that the reactor pressure vessel is protected from brittle fracture up to five EFPY of operation.
In addition, the proposed change will add an effective lower temperature limit to TS Figure 3.4-4 for use of the LTOP system to maintain protection of the RCS.
This modification would extend the low and high PORVs setpoint lines from 100'F down to 90'F.
The limit was added to ensure that LTOP setpoints are used only in a region where the system can provide the necessary protection.
Below 90'F,. overpressuy'ization protection is provided by administrative procedure that will implement TS 3.4.9.4(a) in which the reactor coolant system is depressurized with a vent area greater than 2.9 square inches.
All other LTOP setpoint design bases remain the same as those used previously for the three EFPY 4.0 setpoint determination.
The staff reviewed the current and the extended PORV's setpoints to the new derived pressure-temperature curves and concluded that the setpoints for the PORVs are within the allowable range and are, therefore, acceptable.
ENVIRONMENTAL CONSIDERATION 5.0 This amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes the surveillance requirements.
The staff has determined that the amendment involves 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 issued a proposed finding that this amendment involves no significant hazards consideration, and there has been no public comment on such finding.
Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.
CONCLUSION 6.0 The Commission made a proposed determination that this amendment involves no significant hazards consideration which was published in the Federal Re ister (55 FR 40462) on October 3, 1990, and consulted with thee.i~ac o
Nort Carolina.
No public comments or requests for hearing were
- received, and the State of North Carolina did not have any comments.
The staff has concluded, based on the considerations discussed
- above, that: (I) there is reasonable assurance that the health and safety of the public wi 11 not be endangered by operation in the proposed
- manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
REFERENCES 2.
3.
4 ~
Regulatory Guide 1.99, Radiation Embrittlement of Reactor Vessel Materials, Revision 2, May 1988 NUREG-0800, Standard Review Plan, Section 5.3.2, Pressure-Temperature Limits Shearon Harris, Unit I, Final Safety Analysis Report Shearon Harris, Unit I, Technical Specifications, January 1987
r 5.
Letter from A.B. Cutter, CP&L to USNRC Document Control Desk,
Subject:
Request for License Amendment Reactor Coolant System Pressure-Temperature Limits, September 10, 1990 6.
Generic Letter 88-11, NRC Position on Radiation Embrittlement of Reactor Vessel Materials and Its Impact on Plant Operations, USNRC, July 12, 1988 Dated:
December 26, 1990
-Princi al Contributors:
J. Tsao L. Tran R. Becker