ML20211G811
| ML20211G811 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 02/11/1987 |
| From: | DUQUESNE LIGHT CO. |
| To: | |
| Shared Package | |
| ML20211G790 | List: |
| References | |
| NUDOCS 8702250528 | |
| Download: ML20211G811 (7) | |
Text
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t Revise the Technical Specifications as follows:
Remove Page Insert Page 3/4 4-24 3/4 4-24 3/4 4-25 3/4 4-25 B 3/4 4-6 B 3/4 4-6 B 3/4 4-6a B 3/4 4-6a J
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MATERIAL PROPERTY BASIS CONTROLLING MATERIAL: WELD METAL COPPER CONTENT:
0.31 WT%
PHOSPHORUS CONTENT:
0.015 WT%
RT INITIAL:
O'F NOT RT AFTER 9.5 EFPY 1/4T,274*F NOT 3/4T,137'F CURVE APPLICABLE FOR HEATUP RATES UP TO 60'F/HR FOR THE SERVICE PERIOD UP TO 9.5 EFPY AND CONTAINS MARGINS OF 10*F AND 60 PSIG FOR POSSIBLE INSTRUMENT ERRORS 3000 i
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LIMITATIONS APPLICABLE FOR THE FIRST 9.5 EFPY 3/4 4-24 PROPOSED WORDING
MATERIAL PROPERTY BASIS CONTROLLING MATERIAL:
U.31 WT%
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RT INITIAL:
1/4T, 274*F NDT 3/4T,137'F CURVE APPLICABLE FOR C00LDOWN RATES UP TO 100*F/HR FOR THE SERVICE PERIOD UP TO 9.5 EFPY AND CONTAINS MARGINS OF 10*F AND 60 PSIG FOR POSSIBLE INSTRUMENT ERRORS 3000 i
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FIGURE 3.4-3 BEAVER VALLEY UNIT NO. 1 REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS APPLICABLE FOR THE FIRST 9.5 EFPY 3/4 4-25 PROPOSED WORDING t
REACTOR COOLANT SYSTEM BASES The heatup analysis also covers the determination of pressure-temperature limitations for the case in which the outer wall of the vessel becomes the controlling location.
The thermal gradients established during heatup produce tensile stresses at the outer wall of the vessel.
These stresses are additive to the pressure induced tensile stresses which are already present.
The thermal induced stresses at the outer wall of the vessel are tensile and are dependent on both the rate of heatup and the time along the heatup ramp; therefore, a lower bound curve similar to that described for the heatup of the inner wall cannot be defined.
Subsequently, i
for the cases in which the outer wall of the vessel becomes the i
stress controlling
- location, each heatup rate of interest must be analyzed on an individual basis.
The heatup limit curve, Figure 3.4-2, is a composite curve which was prepared by determining the most conservative case, with either the inside or outside wall controlling, for any heatup rate up to 60*F per hour.
The cooldown limit curves, Figure 3.4-3 are composite curves which were prepared based upon the same type analysis with the exception that the controlling location is always the inside wall where the cooldown thermal gradients tend to produce 3'
tensile stresses. while producing compressive stresses at the outside wall.
The heatup and cooldown curves were prepared based upon the l
most limiting value of the predicted adjusted reference temperature i
at the end of 9.5 EFPY.
l The reactor vessel materials have been tested to determine their initial RTgg; the results of these tests are shown in Table B 3/4.4-1.
Reactor operation and resultant fast neutron (E > 1 Mev) irradiation will cause an increase in the RTNTD.
Therefore, an adjusted reference temperature, based upon the fluence and copper content of the material in question, can be predicted using Figures B 3/4.4-1 and B 3/4.4-2.
The heatup and cooldown limit curves, Figures
-3.4-2 and 3.4-3 include predicted adjustments for this shift in RTNTD as well as adjustments for possible errors in the pressure and temperature sensing instruments.
l BEAVER VALLEY - UNIT 1 B 3/4 4-6 i
PROPOSED WORDING l
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5 10 15 20 25 30 35 SERVICE LIFE (EFPY)
FIGURE B 3/4.4-1 FAST NEUTRON FLUENCE (E>l MeV) AS A FUNCTION OF FULL POWER SERVICE LIFE (EFPY)
No Significant Hazard Determination Proposed Change Request No.
120 amends the Beaver Valley Power
- Station, Unit No. 1 (BVPS-1) Technical Specifications by revising the Reactor Coolant System (RCS) heatup and cooldown limit
- curves, (Figures 3.4-2 and 3.4-3),
and the bases for the limit curves (B l
3/4.4.9 and Figure B 3/4.4-1).
l Basis for proposed no significant hazards determination:
The proposed changes to Technical Specification and Bases entail incorporation of data taken from the summary report (WCAP 10867) on the second surveillance capsule evaluated in conjunction with the BVPS-1 Reactor Vessel Material Surveillance Program.
The surveillance program complies with 10CFR50, Appendix G and Appendix H,
which are intended to ensure that the reactor vessel has an adequate margin of safety with regard to material toughness throughout the service life of the facility.
Specifically, the program develops operating limits (RCS heatup and cooldown limit curves) to prevent non-ductile failure.
The operating limits are adjusted based on surveillance capsule analysis to account for the cumulative radiation effects on the reactor vessel material properties and to maintain an adequate margin of safety, i
Based on the three criteria in 10 CFR 50.92 for defining a significant hazards consideration, plant operation in accordance with the proposed amendment will not:
(1) Involve a
significant increase in the probability or consequences of an accident previously evaluated since all the reactor coolant system components are designed to withstand the effects of normal cycle loads due to temperature and pressure
- changes, as well as the loads associated with the i
postulated faulted conditions.
The fracture mechanics analysis performed to verify the fracture toughness of the reactor vessel is amended to account for the irradiation effects determined by the analysis of the surveillance
- capsules, and the operating limits of the plant are adjusted i
accordingly to offset any observed changes in the material.
Therefore, the likelihood of the previously analyzed accidents occurring is not increased by the incorporation of new heatup and cooldown limit curves.
The consequences of an accident previously evaluated in the FSAR will not be increased because the integrity of the reactor vessel is ensured by accounting for the irradiation effect.
(2) Create the possibility of a new or different kind of accident from any accident previously evaluated.
d The possibility of a new or different accident is not created because no physical changes in the reactor coolant system will result from the proposed Technical Specification Changes.
Pags 2 (3) Involve a
significant reduction in a margin cf safety since the incorporation of new limit curves ensures that the conditions under which the reactor vessel is operated are such that an adequate margin of safe is maintained.
The proposed changes to the reactor coolant system heatup and cooldown limit cures and the bases for these curves will ensure that the reactor vessel has an adequate margin of safety with regard to material toughness throughout the service life of the facility.
Therefore, the proposed changes will not increase the likelihood of a malfunction of safety-related equipment, increase the consequences of an accident previously
- analyzed, nor create the possibility of a malfunction different than previously evaluated.
Based on the above, it is proposed to characterize the change as not involving a
significant hazard.
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