ML19317F440
| ML19317F440 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 06/01/1973 |
| From: | US ATOMIC ENERGY COMMISSION (AEC) |
| To: | |
| Shared Package | |
| ML19317F429 | List: |
| References | |
| NUDOCS 8001140678 | |
| Download: ML19317F440 (2) | |
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m SUPPLEMENT TO SAFETY EVALUATION h
OCONEE UNITS 2,AND 3 (CONTAINMENT SYSTEMS)
DOCKET NOS. 50-270, 287 1.
Primary Containment The applicant has analyzed the ability of the containment to withstand the peak pressure which might be encountered during a To ensure that the most severe break loss-of-coolant accident.
size and location were selected, the applicant performed analyses The breaks for a spectrum of breaks at various locations.
producing the highest containment pressure were a 7 ft split of l
the pump suction piping and the double-ended hot leg break which resulted in a containment pressure of about 54 psig.
The design pressure is 59 psig. We have reviewed the assumptions used by the applicant in this analysis and performed confirmatory analysis for the 7 ft split break.
The applicant calculated the mass and energy release rates to the containment using the CRAFT code for both the blowdown and reflood periods. -We have reviewed the assumptions used in this code and found them to be reasonably conservative.
In our analysis of containment pressure, we used the mass and energy release rates calculated'by the applicant for the blowdown period.
During the reflooding period, however, we used the FLOOD-2 code to predict We calculated containment the mass and energies to the containment.
pressure using the CONTEMPT code and calculated the same containment T
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We, therefore, conclude that the design pressure of e
59 psig is acceptable.
2.
Subcompartment Analysis The applicant has presented the results of calculations of pressure differentials across the walls of compartments inside the containment. We have performed similar calculations which result in values higher than that calculated by the applicant for the steam generator compartments.
For the east compartment, i
the applicant calculates 9.7 psi for a double-ended hot leg break.
The applicant stated that this structure was designed for 11.3. psi.
Our analysis of the same postulated break results in a peak calculated differential pressure of approximately 18 psi. Due to the difference between the applicant's and our analysit, we conclude that the steam generator compartment design pressure has not been adequately justified.
1 The applicant has also presented the calculated pressure differential response in the reactor. cavity for two size breaks of a reactor coolant pipe of 3.0 and 8.0 square feet.
The results, as presented by the applicant, indicate that the 8.0 square foot break would result in a 195 psi peak pressure differential.
The stated design value 'a 205 psi. We have performed a similar analysis which reasonably agrees with the applicant.
We, therefore, conclude that the design pressure of 205 psi is acceptable.
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