ML20079A600
| ML20079A600 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 09/27/1990 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20079A599 | List: |
| References | |
| NUDOCS 9010150220 | |
| Download: ML20079A600 (8) | |
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 40 TO FACILITY OPERATING LICENSE NO. NPF-42 WOLF CREEK NUCLEAR OPERATING CORPORATION WOLF CREEK GENERATING STATION DOCKET NO. 50-482
1.0 INTRODUCTION
By application dated June 20, 1988, and as supplemented by letters dated May 22, June 8, and August 1, 1990, the Wolf Creek Nuclear Operating Corporation (the licensee) proposed changes to the Technical Specifications (TS) (Appendix A to Facility Operating License No NPF-42) for the Wolf Creek Generating l
Station regarding heatup, cooldown and cold overpressure mitigation system i
power-operated relief valve'(PORV) setpoint pressure / temperature (P/T) limits i
as required by 10 CFR Part 50, Appendix H.
The current P/T. limits were developed during the initial licensing process using results of the theoretical neutron transport analysis.
The proposed P/T limits were developed based on the data obtained from the actual surveillance capsule which was withdrawn from the Wolf
~ Creek Generating Station (WCGS) reactor during the first refueling outage.
Based on the revised P/T limits, new PORV setpoints were derived for the cold i
overpressure mitigation system.
The new data for both the pressure / temperature L
limits and the PORV setpoints are applicat,le for seven effective full power years (EFPY).
As of Septembe"
, 1990, the Wolf Creek facility had accumulated 3.86 EFPY of operation.
2.0 DISCUSSION 2.1 Pressure / Temperature (P/T) Limits (a) Background 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 ASTM Standards and ASME Code, which are referenced in Appendices G and H; 10 CFR 50.36(c)(2);
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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 parti-cular, 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 to 10 CFR Part 50 describe specific requirements for f'acture toughness and reactor vessel material surveillance.
These must l
r be considered in setting P/T limits.
An acceptable method in constructing the P/T limits is described in SRP Section 5.3.2.
Appendix G to 10 CFR Part 50 specifies fracture toughness and testing requirements for reactor vessel materials in accordance with the ASME Code and, in particular, to test the beltline materials in the surveillance capsules in accordance with Appendix H to 10 CFR Part.50.
Appendix H, in turn, refers to the ASTM Standards.
These tests define the condition of vessel embrittlement.at the time.of capsule withdrawal in terms of the increase in the reference temperature (RTNDT).
Appendix G also requires the licensee to predict the effects of neutron irradiation on vessel embrittlement by calculating the adjusted RT and upper shelf energy NDT (USE).
A method that is acceptable to the NRC staff is described in Regulatory Guide 1.99, Rev. 2.
Appendix H to 10 CFR Part 50 requires the licensee to establish a sur-L veillance program to periodically withdraw surveillance capsules from the reactor vessel.
Appendix H refers to the ASTM Standards which, in turn, requires that the capsules be installed in the vessel before startup_and that they contain test specimens that-are made from plate, weld, and heat-affected-zone materials of the reactor beltline.
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(b) Evaluation The licensee removed the first surveillance capsule, U, from the reactor vessel after 1.08 EFPY of operation.
The surveillance specimens in capsule U were tested and analyzed by Westinghouse Corporation.
The results were published in the Westinghouse report, " Analysis of Capsule U From the stolf Creek Nuclear Operating Corporation - Wolf Creek Reactor Vessel Radiation Surveillance Program," WCAP-11553.
In accordance with Appendix H to
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10 CFR Part 50, the licensee submitted the report on November 4,1987, a l
year after the capsule withdrawal.
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Capsule U contained Charpy impact specimens and tensile opecimens that were made from base metal, weld metal, and heat affected zone (HAZ) metal.
The base metal specimens were made from the lower shell plate, R2508-3; the weld metal specimens were made from the girth weld located between the intermediate shell plate and lower shell plate.
The weld was made by the i
submerged arc weld method using 3/16-inch Mil 8-4 filler wire, heat number I
90146 and the weld flux was Linde 124, lot number 108.
The'HAZ specimens were made from the HAZ metal of plate R2508-3.
The specimens were tested in accordance with Appendices G and H to 10 CFR Part 50 and ASTM E185-82.
Specifically, the Charpy impact. tests were performed per ASTM E23-82 and l
tension tests _were performed per ASTM E8-83 and E21-79.
The Charpy impact I
tests of the R2508-3 plate showed that as a result of neutron fluence 3.39E18 2
n/cm, the increase in RT was 30'F and the decrease in the USE was H0T 3 ft-1b.
The tests showed that the R2508-3 plate has the highest increase in RTNOT; therefore, it was designated as the limiting (controlling) material.
The licensee used the method in Regulatory Guide 1.99, Rev. 2, to calculate an adjusted RT of 108'T for the R2508-3 plate at 7 EFPY and 1/4T location NDT (T is the reactor vessel thickness).
The staff performed a similar calcu-lation and verified the licensee's RT value to be correct (see Table 1).
NOT Substituting the RT of 108'F into equations in SRP 5.3.2, the staff NOT verified that the proposed P/T limits for heatup, cooldown, criticality, and hydrotest are acceptable.
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t TABLE I The NRC Staff Calculated Adjusted Reference Temperature For The Limiting Reactor Beltline Material at Wolf Creek Unit 1.
Material:
lower shell plate A553B, Class 1 Code No.:
R2508-3 Copper Content:
0.07%
Nickel Content:
0.62%
Initial RTNDT:
40*F i
2 Neutron Fluence n/cm at 7 EFPY 6.825E18 at 32 EFPY 3.12E19 Increase in RT 33'F NDT at 7 EFPY Adjusted RT DT at 7 EFPY 107'F (Licensee Calculated 108*F)
In addition to beltline materials, Appendix G to 10 CFR Part 50 also imposes 1
P/T limits on the react)r vessel closure flanges.
Section IV.2 of Appendix G states that when pressure exceeds 20 percent of the preservice system I
hydrostatic test pressure, the temperature of the closure flange regions that are highly strested by the bolt preload must exceed the RT of the NDT 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 RT of 20'F, the staff has determined that the proposed P/T limits NDT satisfy Section IV.2 of Appendix G.
Section IV.B of Appendix G requires the predicted USE at end-of-life to be above 50 ft-lb.
At 1.08 EFPY, the measured USE is 92 f t-lb for the inter-mediate to low shell weld metal.
This is an 8 percent reduction from the
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unirradiated value of 100 ft-lb and is limiting.
Using the method in Regulatory Guide 1.99, Rev. 2, the staf f predicted that the USE of the weld metal at end-of-life will still be above 50 f t-lb and thus the projected USE of the limiting metal at end-of-life satisfiesSection IV.B of Appendix G.
The staff concludes that the proposed WCGS P/T limits on the reactor coolant system for heatup, cooldown, leak test, and criticality are valid through 7 EFPY because the limits conform to requirements of Appendices G and H to 10 CFR Part 50.
The licensee also conforms to Generic Letter l
88-11 because it used the method in Regulatory Guide 1.99, Rev. 2, to calculate the reference temperature, RT The proposed P/T limits may NDT.
be incorporated into the WCGS TS.
2.2 Low Temperature Overpressure Protection (a) Background Low temperature overpressure protection (LTOP) is provided by the PORVs on the pressurizer.
These PORVs are set at pressures low enough to prevent violation of the 10 CFR Part 50 Appendix G heatup and cooldown curves should a RCS pressure transient occur during low temperature operations.
The licensee, in its June 20, 1988, submittal, identified the most limiting overpressure transients analyzed to determine the PORV setpoints for LTOP.
The PORV setpoint limits have been set by two design criteria.
These are i
t'he limiting transients for mass addition and heat addition.
In response to an NRC request for additional information (letter dated August 21, 1989, from D. V. Pickett, NRC to B. D. Withers, WCNOC), the licensee provided additional description of the methodology used for determining the PORV setpoints and revised Figures 3.4-2, 3.4-3, and 3.4-4 for the Wolf Creek TS.
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(b) Evaluation l
1 The present WCGS TSs 3.1.2.3 and 3.5.3 require one charging pump to be operable for boron injection or recirculation cooling in Modes 4, 5, and 6.
For LTOP considerations, the most limiting mass addition transient l
was analyzed assuming letdown isolation with one centrifugal charging pump in operation with maximum charging flow.
This 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 heat input transient was analyzed assuming a 50'F temperature difference-between the hotter secondary side of a steam generator and the RCS cold leg.
A reactor coolant pump startup in one loop was assumed 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 pressure-tempera-ture curves for WCGS are not exceeded.
By letter dated May 22, 1990, the licensee provided supplementary informa-tion related to the thermal-hydraulic analyses used to support the proposed TS changes.
The revised LTOP setpoints identified in the amendment request were derived using the same methodology employed in.the development of the original LTOP setpoints currently described in the plant's Updated Safety Analysis Report.
The original methodology is documented in the Westinghouse l
Owners' Group (WOG) report dated July 1977 and its supplement of September 1977.
The WOG methodology u ed the LOFTRAN computer code to generate the PORV setpoint overshoot for a bounding envelope of mass addition rates.
I The plant specific PORV setpoints and overshoot were then determined with 1
i plant specific parameters and updated algorithms applicable to WCGS.
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!.. Based on the results of the most limiting LTOP transient, the licensee proposed TS PORV setpoints that are no higher than 1910 psig when the RCS I
average temperature is equal to 377'F.
The PORV setpoints are progres-sively reduced at lower RCS temperatures to maintain the margin of safety required by Appendix G considerations.
The combined effect of the new pressure-temperature limits is reflected in a modification to Figure 3.4 !
in the WCGS TS.
The new figure 3.4-4 presents the maximum allowed PORV setpoints as a function of RCS temperature and includes margin for possible instrument error.
The modifications are based on analyses using approved methodology and are acceptable.
A proposed change to TS 3.4.9.1 will administrative 1y restrict the RCS heatup rate to less than or equal to 60*F/hr for an indicated RCS average temperature less than or. equal to 200'F.
This change will maintain the margin of safety associated with 10 CFR Part 50, Appendix G requirements and is acceptable.
The licensee's proposed changes in indicated temperature and PORV setpoints in the TSs are consistent with the above discussed COMS alignment temperatures L
and the heatup and cooldown rates identified by the updated Figures 3.a-2, 3.4-3, and 3.4-4 in TS 3.4.9.
The staff finds that they are reasonably conservative and are acceptable.
The staff review and approval applies to l
the revised Figures and Bases pages provided in the licensee's May 22, 1990, and June 6, 1990, submittals.
f-Based on the evaluation above, the staff concludes that the licensee's proposed TS and their associated bases are acceptable to support the updated pressure-temperature limits applicable for a period up to seven EFPY.
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3.0 ENVIRONMENTAL CONSIDERATION
Pursuant to 10 CFR 51.21, 51.32 and 51.35, an environmental assessment and finding of no significant impact was published in the Federal Register on September 14, 1990 (55 FR 37989).
Accordingly, based upon the environmental assessment, the Commission has determined that issuance of this amendment will not have a significant effect on the quality of the human environment.
4.0 CONCLUSION
The staff has 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.
Date:
September 27, 1990 Principal contributors:
John C. Tsao EMTB/NRR Michael A. McCoy, SRXB/NRR Douglas V. Pickett, PDIV-2/NRR
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