ML20049H389
| ML20049H389 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 02/05/1982 |
| From: | Randall P, Tourigny E Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20049H387 | List: |
| References | |
| NUDOCS 8203020272 | |
| Download: ML20049H389 (3) | |
Text
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...f SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 64 TO FACILITY' OPERATING LICENSE NO. DPR-40 OMAHA PUBLIC POWER DISTRICT FORT CALHOUN STATION, UNIT NO. 1 DOCKET NO. 50-285
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Introduction:==
By application dated January 15, 1982, as supplemented by letter dated January 27, 1982, Omaha Public Power District (OPPD or the licensee) requested an amendment to the Technical Specifications (TS) which changes the reactor coolant system pressure-temperature limits for operation through 6.1 effective full power years (EFPY's). Tha licensee is presently-in fuel cycle 7.
This TS change will permit the licensee to complete fuel cycle 7.
The end of 6.1 EFPY's of operation will occur early in fuel cycle 8.
- Thus, another similar amendment request will have to be submitted and assessed before the. licensee is substantially into fuel cycle eight. This safety
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evaluation addresses operations through 6.1 EFPY's.
Discussion:
10 CFR Part 50, Appendix G, " Fracture Toughness Requirements", requires that pressure-temperature limits be established for reactor coolant system heatup and cooldown operations, inservice leak and hydrostatic tests, and reactor core operation. These limits are required to ensure that the stresses in the reactor vessel remain within acceptable limits. They are intended to provide adequate marg'ns of safety during any condition of normal operation, including anticipated operational occurrences.
10 CFR Part 50, Appendix H, " Reactor Vessel Material Sur' eillance Program v
Requirements", requires a material surveillance program to monitor changes l
in the fracture toughness properties of ferritic materials in the reactor vessel beltline region of water cooled power reactors resulting from their exposure to neutron irradiation and the thermal environment. Under this program, fracture toughness test data are obtained from material specimens withdrawn periodically from the reactor vessel. These 6ata permit the determination of the conditions under which the vessel can be operated with adequate margins of safety against fracture throughout its service life.
Thus, the results of this program help to establish the pressure-temperature limits addressed in 10 CFR Part 50, Appendix G.
f-B203020272 820205 PDR ADOCK 05000285 P
One principal effect of neutron irradiation upon a reactor pressure vessel-is that it causes the vessel material nil-ductility temperature to increase with time. The pressure-temperature operability limits must therefore be modified periodically to account for this. By periodically revising the pressure-temperature limits to account for radiation damage, the stresses and stress intensities'in the reactor vessel are maintained within acc'eptable limits.
Possible changes to other systems must be evaluated as a result of changing the pressure-temperature limits..The only other systems changes insofar as this application is concerned areTS changes regarding operation of safety injection systems. One would not desire an actuation of a safety injection system to damage the reactor pressure vessel.
Safety Evaluation:
The first part of the staff safety evaluation addresses the proposed pressure-temperature limit curves. The staff accepts the licensee's prediction that the peak fluence at the inner diameter wall of the pressure vessel will be 8.4 x 1018 n/cm2 (E>l Mev) at the end of 6.1 EFPY's. Based upon this value, the licensee calculates a nil-ductility transition temperature shift to be equal to 2380F for the t position. The staff believes this value to be conservative and therefore accepts it.
The staff has performed independent calculations to check the pressure-temperature limits proposed in Figures 2-1A, 2-1B, 2-2A, and 2-28. These calculations assumed 2380F for the adjustment of reference temperature, an initial nil-ductility transition reference temperature of OoF, and a
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vessel wall thickness of 7.125 inches. The figures proposed by the licensee were found to be acceptable by the staff.
The socond part of the s'taff safety evaluaticn considers an administrative temperature change. The licensee desires to modify a temperature value as obtained from the curve shown in Figure 2-3 of t.he TS. Currently, the TS states on page 2-5 that "the predicted TNDI shift for an integrated fast neutron (E>l Mev) exposure of 4.4 x 1019 n/cmz is 3500F, the value obtaine~d from the ciirve shown in Figure 2-3".
The licensee proposed to use 3440F i
'instead of 3500F. Since figure 2-3's abscissa and ordinate are not linear, it is difficult to find highly accurate temperature values as a function of Therefore, OPPD's proposal of 3440F versus the 350oF presently fluence.
in the TS is considered an administrative temperature change and is acceptable.
The third part of the staff safety evaluation addresses safety injection system changes as a result of the pressure-temperature limit changes.
l Low temperature overpressurization can be induced by. operation of safety injection systems. The licensee proposes that whenever the reactor coolant system cold leg temperature is belcw certain temperature values, certain As an example, high pressure safety injection pumps shall be dis'abled.
inadvertent actuation of three (3) HPSI pumps and three (3) charging. pumps, coincident with the opening of one of the two power operated relief valves 4
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\\ J (PORV's), would result in a peak primary system pressure of 1190 psia.'
l A minimum permissible temperature of 3370F corresponds with 1190 psia on i
the proposed Figure 2-1B (1000F/hr cooldown rate curve). Thus, at least one HPSI pump is disabled at 3370F. The s.taff finds such changes acceptable.
The licensee also proposes that whenever the reactor coolant system cold r
leg temperature is below 1530F, the cooldown rate of Figure 2-1B shall be.
limited to a maximum rate of 200F/hr. This is to ensure that the safety system can safely function. The staff agree-Environmental Consideration We have determined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power. level and will not result in any significant environmental impact. Having made this determination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR $51.5(d)(4), that an environmental impact statement or negative declaration and environ-mental impact appraisal need not be prepared in connection with the issuance of this amendment.
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Conclusion We.have concluded, based on the considerations discussed above, th.at:
(1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered i
and does not involve a significant decrease in a safety margin, the amendment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
j Date: February 5, 1982
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Principal Contributors:
E. G. Tourigny, Project Manager P. N. Randall, Materials Engineer t
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