ML17156B519
| ML17156B519 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 12/01/1989 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17156B518 | List: |
| References | |
| GL-88-11, NUDOCS 8912070337 | |
| Download: ML17156B519 (4) | |
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UNITED STATES NUCLEAR R EGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO GENERIC LETTER 88-11 PENNSYLVANIA POWER AND LIGHT COMPANY SUS UEHANNA STEAM ELECTRIC STATION UNITS 1
AND 2 DOCKET NOS.
50-387 AND -388
1.0 INTRODUCTION
In response to Generic Letter 88-11, "NRC Position on Radiation Embrittlement of Reactor Vessel Materials and Its Effect on Plant Operations,"
the Pennsylvania Power and Light Company (the licensee) requested permission to revise the pressure/temperature (P/T) limits in the Susquehanna Steam Electric Station, Units 1 and 2 Technical Specifications, Section 5.3.
The request was documented in a letter from the licensee dated December 12, 1988. This revision also changes the effectiveness of the P/T limits of 32 effective full power years (EFPY).
The proposed P/T limits were developed based on Section 1 of Regulatory Guide (RG) 1.99, Revision 2.
The proposed revision provides up-to-date P/T limits for the operation of the reactor coolant system during
- heatup, cooldown, criticality, and hydrotest.
To evaluate the P/T limits, the staff uses the following NRC regulations and guidance:
Appendices G and H of 10 CFR Part 50; the ASTM Standards and the ASME Code, which are referenced in Appendices G and H; 10 CFR 50.36(c)(2);
RG 1.99, Rev. 2; Standard Review Plant (SRP) Section 5.3.2; and Generic Letter 88-11.
Each licensee authorized to operate a nuclear power reactor is required by 10 CFR 50.36 to provide Technical Specifications for the operation of the plant.
In particular, 10 CFR 50.36(c)(2) requires that limiting conditions of operation be included in the Technical Specifications.
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 of 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 SRP Section 5.3.2.
Appendix G of 10 CFR Part 50 specifies fracture toughness and testing requirements for reactor vessel materials in accordance with the ASME Code and, in particular, that the beltline materials in the surveillance capsules be tested in accordance with Appendix H of 10 CFR Part 50.
Appendix H, in turn, refers to ASTM Standards.
891207033T 891201 PDR
- DOCK 05000387 P
PNU These 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 Char py upper shelf energy (USE).
Generic Letter 88-11 requested that licensees and permittees use the methods in RG 1.99, Rev. 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 of 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 ASTht 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 beltline.
2.0 EVALUATION The staff evaluated the effect of neutron irradiation embrittlement on each beltline material in the Susquehanna 1 and 2 reactor vessels.
The amount of irradiation embrittlement was calculated in accordance with RG 1.99, Rev. 2.
The staff has determined that the material with the highest ART at 32 EFPY for Unit 1 was the lower intermediate shell plate C2433-1 with 0.10% copper (Cu),
0.63% nickel (Ni), and an initial RT of 18 F.
The material with the highest ART at 32 EFPY for Unit 2 was the lola shell plate 6C1053-1-1 with 0.10%
Cu and 0.58% Ni, and an initial RTndt of 10 F.
The licensee has not removed any survei llance capsules from Susquehanna 1
and 2.
All surveillance capsules contained Charpy impact specimens and tensile specimens made from base metal, weld metal, and HAZ metal.
For the limiting beltline material in Unit 1, loper intermediate shell plate C2433-1, the staff calculated the ART to be 57.4 F at 1/4T (T = reactor vessel beltline thickness) for 32 EFPY.
For the limiting beltline material jn Unit 2, lower shell plate 6C1053-1-1, the staff calculated the ART to2be 49.3 F at 1/4T for 32 EFPY.
The staff used a neutron fluence of 5.3E17 n/cm at 1/4T for both units.
The licensee used the method in RG 1.99, Rev. 2, to calculate an ART of 57.6 F
at 32 EFPY at 1/4T for the same limiting lower intermediate shell plate in Unit 1 and 49.5 F for the same limitjng lower shell plate in Unit 2. The staff judoes that a difference of 0.2 F between the lic~nsee's ARTs and the staff's ARTs is acceptable.
Substituting the ART of 57.6 F into equations in SRP 5.3.2 for Unit 1 and 49.5 F for Unit 2, the staff verified that the proposed P/T limits for heatup,
- cooldown, and hydrotest meet the beltline material requirements in Appendix G of 10 CFR Part 50.
0 In addition to beltline materials, Appendix G of 10 CFR Part 50 also imposes P/T limits based on the reference temperature for the reactor vessel closure flange materials.
Section IV.2 of Appendix G states that when the pressure exceeds 20% of the preservice system hydrostatic test pressure, the temperature of the closure flanoe regions highly stressed by the bolt preload must exceed the reference temperature of t)e 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 10 F, the staff has determined that the proposed P/T limits satisfy Section IV.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.
Based on data presented by the licensee, the lowest USE for any beltline material in Susquehanna 1 was estimated at 81 ft-lb.
Using Figure 2 in RG 1.99, Rev. 2, it was estimated that the EOL USE would be 70.8 ft-lb.
This is greater than 50 ft-lb and, therefore, is acceptable.
Based on data presented by the licensee, the lowest USE for any beltline material in Susquehanna 2 was estimated at 61 ft-lb.
Using Figure 2 in RG 1.99, Rev. 2, it was estimated that the EOL USE would be 53 ft-lb.
This is greater than 50 ft-lb and, therefore, is acceptable.
3.0 CONCLUSION
The staff concludes that the proposed P/T limits for the reactor coolant system for heatup,
- cooldown, leak test, and criticality are valid through 32 EFPY because the limits conform to the requirements of Appendices G and H of 10 CFR Part 50.
The licensee's submittal also satisfies Generic Letter 88-11 because the licensee used the method in RG 1.99, Rev.
2 to calculate the ART.
- Hence, the proposed P/T limits may be incorporated into the Susquehanna Units 1
and 2 Technical Specifications.
4.0 REFERENCES
1.
Regulatory Guide 1.99, Radiation Embrittlement of Reactor Vessel Materials, Revision 2, May 1988 2.
NUREG-0800, Standard Review Plan, Section 5.3.2, Pressure-Temperature Limits
, 3.
Final Safety Analysis Report for Susquehanna Steam Electric Station, Units 1 and 2
4.
December 12, 1988, letter from H.
W. Keiser (PPSL) to W.
R. Butler (USNRC),,
subject:
Susquehanna Steam Electric Station,
Response
to Generic Letter 88-,11 5.
May 20, 1981, letter from N.
W. Curtis (PPSL) to B. J. 'Youngblood (USNRC),
subject:
Susquehanna Steam Electric Station, SER Issue No. 21, 22, 23, 24 6.
October 3, 1983, letter from N.
W. Curtis (PP&L) to A. Schwencer (USNRC),
subject:
Susquehanna Steam Electric Station, SER Issue No.
112 Dated:
December 1,
1989