ML17223A821
| ML17223A821 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 06/11/1990 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17223A820 | List: |
| References | |
| NUDOCS 9006210031 | |
| Download: ML17223A821 (7) | |
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UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF tIUCLEAR REACTOR REGULATION RELATED TO AMENDMENT N0.104 TO FACILITY OPERATING LICENSE NO. DPR-67 FLNIDA POMER 5 LIGHT COMPANY ST.
LUCIE PLANT UNIT NO.
1 DOCKET NO. 50-335 INTRODUCTION By application dated December 5, 1989, the Florida Power and Light Company (the licensee) requested an amendment which would incorporate revised pressure/temperature (P/T) limits and the results of a revised low temperature overpressure protection (LTOP) analysis into the Technical Specifications (TS) for St. Lucie Unit 1.
The current St. Lucie Unit 1 TS for P/T and LTOP are applicable to 10 effective full power years (EFPY).
Accordingly, the St.
Lucie 1 TS require revision prior to the plant reaching 10 EFPY.
Below is the staff's evaluation of the proposed changes.
PRESSURE/TEHPERATURE LIHITS In response to Generic Letter 88-11, "NRC Position on Radiation Embrittlement of Reactor Vessel Haterials and Its Effect on Plant Operations,"
the licensee requested permission to revise the P/T limits in the St. Lucie 1 TS, Section 3.4.
This revision also changes the effectiveness of the P/T limits from 10 to 15 EFPY.
The proposed P/T limits were developed using 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 ASTH Standards and the ASHE Code, which are referenced in Appendices G and H; 10 CFR 50.36(c)(2);
RG 1.99, Rev. 2; Standard Review Plan (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 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" TS 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.
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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 tn ASTM Standards.
These tests define the extent of vessel embri+tlement 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).
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 ASTM 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.
Evaluation The staff evaluated the effect of neutron irradiation embrittlemer t on each beltline material in the St.'Lucie 1 reactor vessel.
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 15 EFPY for St. Lucfe 1 at 1/4T (T vessel beltline thickness) was the lower shell longitudinal weld seams (3-203A, 8, and C) with 0.30% copper (Cu), 0.64%
nickel (Hi), and an initial RT t of -56'F.
At 3/4T, the material with the highest ART at 15 EFPY was lowIII shell plate C-5935-1 with 0.15% Cu, 0.57% Ni, and an initial RTdt of 20'F.
The licensee has removed one surveillance capsule from St. Lucie 1.
The results from capsule W-97 were published in Combustion Engineering Report TR-F-MCM-004.
A11 surveillance capsules contained Charpy impact specimens and tensile specimens made from base metal, weld metal, and HAZ metal.
For the limiting beltline material at I/4T, the lower shell longitudinal weld
For the limiting beltline material at 3/4T, lower shell plate C-5935-1, the staff calculated the ART to be 137.2'F.
The staff used a neutron fluence of 1.22E19 n/cm~ at 1/4T and 4.34E18 n/cm't 3/4T.
The ART was determined using Section 1
of RG 1.99, Pev.
2 because only one surveillance capsule has been removed from the St. Lucie 1 reactor vessel.
The licensee used the method in RG 1.99, Rev. 2, to calculate an ART of 191'F at 1/4T and 137'F at 3/4T.
The licensee's ART of 191'F is more conservative than the staff's ART of 181.4'F; therefore it is acceptable.
Substituting the ART of 191'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 of 10 CFR Part 50.
In addition to beltline materials, Appendix G nf 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 flange regions highly stressed by the bolt preload must exceed the reference temperature of the material in those regions by at least 1?O'F for normal operation and by 90'F for hydrostatic pressure tests and leak tests.
Based on the flange reference temperature of 30'F, the staff has determined that the proposed P/T limits satisfy Section IV.2 cf Appendix G.
Section IV.B of Appendix G requires that the predicted Charpy USE at end of life be above 50 ft-lb.
'The licensee has USE data for all plate materials and one weld in the beltline area, but does not have any USE data for the intermediate shell longitudinal weld seams (2-203A, B, and C) and tho. lower shell longitudinal weld seams (3-?03A, B, and C).
The staff will obtain tne USE of these two welds
',n the near future.
Presently, the staff has determined that all the materials for which USE data are available will meet the requirement
.hat the Charpy USE at end of life be above 50 ft-lb.
Technical Findin The staff concludes that the proposed P/T limits for the RCS for heatup,
- cooldown, leak test, and criticality are valid through ]5 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 St. Lucie 1 TS.
LOl( TEMPERATURE OVERPRESSURE PROTECTION The Reactor Coolant System (RCS) P/T limits during plant heatup and cooldown are specified in Technical Specification Figures 3.4-2a and 3.4-2b for St. Lucie Unit 1.
The P/T curves in the current TS are based on an assumed design basis neutron fluence through 10 EFPY.
By letter dated December 5, 1989, the licensee provided its updated P/T curves in proposed TS Figures 3.4-2a (for heatup and core critical) and 3.4-2b (for cooldown and inservice testing),
changes in the values of the RCS cold leg temperature at which LTOP should be enabled, and the justification for the changes.
The new P/T curves are based on the irradiation damage prediction methods of RG 1.99, "Radiation Embrittlement of Reactor Vessel Materials,"
Revision 2, U.S. Nuclear Regulatory Commission, May 1988, and are applicable for a period up to 15 EFPY.
New cooldown rates as a function of indicated reactor coolant temperature are also proposed in an updated Figure 3.4-3.
LTOP is provided by the PORVs on the pressurizer.
These PORVs are set at pressures 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, in'its December 5, 1989 submittal, identified the most limiting overpressure transients in determining the PORV setpoints for LTOP.
The FPRV setpoint limits have been previously set by analysis of the limiting transients for mass addition and energy addition.
TS 3.4.13 maintains the same pressure setpoints and revises the values of the applicable temperatures for LTOP.
A modification to TS 3.4.14, "Reactor Coolant Pump-Starting," identifies the new applicable temperature for LTOP in Mode 4.
4-The most limiting mass addition transient was analyzed assuming a spectrum of inadvertent safety injection actuation assumptions.
The transient analysis is typically performed to determine the pressure overshoot past the LTOP setpoint such that the Appendix G curves are not exceeded during the transient.
The energy input transient was analyzed assuming a 30"F temperature difference between the steam generator and the RCS.
A reactor coolant pump startup in one loop was assumed in order to maximize the heat transfer effect.
As was the case for the mass addition transient, the pressure overshoot is calculated such that the Appendix G P/T curves for Unit 1 are not exceeded.
The present TS 3.5.3, "ECCS Subsystems
- T 325'F," allows only one high pressure safety injection (HPSI) pump to b3"operable prior to decreasing the RCS temperature below ?53'F and requires disabling of all HPSI pumps prior to decreasing the RCS temperature below 220'F.
Jn the proposed revision, the RCS temperature in LCO 3.5.3b and Action b woula change from 253"F to 270'F, and the RCS temperature in LCO 3.5.3c would change from 220'F to 236'F.
These changes are necessary as a result of a reanalysis of inadvertent safety injection actuation in a LTOP condition.
During heatup, a HPSI pump may be returned to service at 236'F.
To provide a reasonable operational margin for returning a HPSI pump to service, the applicability of 3.5.3 in Hode 4 would be revised from a RCS cold leg temperature above 235'F to a PCS cold leg temperature above 250'F.
This change is made to a footnote in TS 3.5.3.
The licensee's analyses were performed using the same methodology as the prior application for 10 EFPY with some changes in the analysis assumptions.
For the revised analysis, the LTOP enable temperatures were determined by following the guidance that for LTOP, the enable temperature is the water temperature corresponding to a metal temperature of at least RT >
+ 90"F at the vessel beltline, which was calculated by the licensee to bÃ04'F during heatup and 28] 'F during cooldown.
The results indicated that a change in the present PORV setpoints of 350 psia and 530 psia is not required.
The Oefinitions Section 1.16, "Low Temperature RCS Overpressure Protection Range,"
is revised to identify the cold leg temperature for the LTOP range as less than or equal to 304'F during heatup or less than or equal to 281'F during cooldown.
The licensee-proposed changes in TS 1.16, 3.4.13, 3.4.14, and 3.5.3, and the associated Bases sections reflect the above discussed LTOP alignment temperatures and the heatup and cooldown rates identified by the updated Figures 3.4-2a, 3.4-2b, and 3.4-3 in TS 3.4.9.
The staff finds that they are reasonably conservative and acceptable.
Technical Findin Based on the above evaluation,,the staff concludes that the proposed TS 1.16, 3.4.13, 3.4.14, 3.5.3, and their associated Bases are acceptable to support the updated P/T limits identified in TS 3.4.9.1 applicable for a period up to 15 EFPY.
ENVIPCNMENTAL CONSIDERATION This amendment involves a change to a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20.
We have determined that this 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 exclu ion 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 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 of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
REFERENCES 1.
Regulatory Guide 1.99, "Radiation Embrittlement o
Reactor Vessel Materials," Revision 2, Hay 1988.
2.
NUREG-0800, Standard Review Plan, Section 5.3.2, Pressure-Temperature Limits.
3.
Letter from J.
H. Goldberg (FPL) to USNRC Document Control Desk,
Subject:
"St. Lucie 1 Proposed License Amendment P-T Limits and LTOP Analysis,"
December 9, 1989.
4.
S. T. Byrne, "Florida Power and Light Company St. Lucie Unit No. 1, Post-Irradiation Evaluation of Reactor Vessel Surveillance Capsule W-97,"
Combustion Engineering Report No. TR-F-MCM-004, December 1983.
Date:
June 1I, 1990 Princi al Contributors:
sao H. McCoy
'OAEEO:
AMENDMENT NO.
"O'TO FACILITY OPERATING LICENSE NO. DPR ST. LUCIE, UNIT 1
'WtbjC'f,geQ NRC 8 Local PDRs PDII-2 Reading S. Yarga, 14/E/4 G. Lainas, 14/H/3 H. Berkow D. Hiller J. Norris OGC-WF D. Hagan, 3302 HNBB E. Jordan, 3302 HNBB B. Grimes, 9/A/2 G. Hill (4), P-137 Wanda Jones, P,-130A J. Calvo, 11/F/23 J.
Tsao H.
McCoy ACRS (10)
GPA/PA OC/LFHB PD Plant-specific file [Gray File M. Sinkule, R-II Others as required cc:
Plant Service list
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