ML19210A586
| ML19210A586 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 09/21/1967 |
| From: | Morris P US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Neidig R METROPOLITAN EDISON CO. |
| References | |
| NUDOCS 7910300626 | |
| Download: ML19210A586 (4) | |
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?.EC Cocunant Room For--a l Suppl.
!;2L ',en in; 222-3 '.c2 ding Cris:
- 3. Crimes
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- 2. S. Soyd SEP 211967 Izc 2eadins Dockst 50-289 N. Dube (5)
L. Kornblich (2)
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J. 2. Buchanan, 02NL 22 Dranch Chiefs Metropolitan Edison Company 2'r Branch Chiefe P.O. Box 542 g,'tevine Reading, Pennsylvania 19603
- 2. Shevholt Attention:
Mr.1. E. Neidig Vice President Gentlemsn:
This is a request for supplemental information to your application for a construction permit and operating license for the Three Mile Island Nuclear Station to be located in Danphin County, Pennsylvania.
As indicated in our ici.ter of August 25, 1967, additionet informa-tion on the subject of charmal shock to the reactor vessei during emergency coolant injection is required. The attsched question list treats this subject and lists other information which we have con-cluded will be required to complets our review.
In order to facilitate our technical review, we urge that you provide full and complete answers to the attached cuestions so that further questions covering the same material will not be required. We will be available to amplify the meaning of any of the questions.
Sincerely yours, Origina! Signed by M. M. Mann F. A. Morris, Director Division of Reactor 1.icensing
Enclosure:
Requestad Additional Information ec:
Mr. coorse 1. Treridge, z.,q.
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. 11.2 L'ith respect to the brittic fracture mode of failure provide the following additional information:
11.2.1 The assuned distribution of the initial NDT temperature through the plate thickness. State also the experimental basis for this assumption and the degree of conservatism involved.
11.2.2 Thg assumed time-integrated neutron flux (nyt) at thc reactor vessel inner diameter.
11.2.3 The pr-file of the NDT temperature shif t through the thick-ness of the plate.
11.3 An estimate of the effect of an initial vessel temperature higher than that assumed in the analysis on the extent of yielding and defor-c.ation o f the vessel.
11.4 An estimate of the maximum allowable pressure stress, when combined u!th other stress present in the vessel, whi:5 could be tolerated without failure.
11.5 An estimate of the maximum neutron flux exposure (nvt) of the vessel that could be' tolerated without vessel failure.
11.6 The effect of potential local penetrations present in the vessel cladding, exposing the base metal to the coolant, on the results of the analysis.
11.7 The nu:aber of thermal shock cycles, induced by ECCS operation, that tr.- vessel could withstand at the end of its fatigue life.
11.8 Experimental data on the thermal shock effects in thick plates under stress, tested below the NDT temperature.
11.9 An evaluation of the capability of the safety injection nos:les and accumulator piping to withstand the transient.
11.10 An evaluation of the effects of thi', transient on the core barrel and other internals with regard to assuring that distortion would not restrict the flow path of the emergency core coolant.
1492 02B
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3-12.0 rI:37N, ?RODUCT
- TILE.ist raoM pugn 12.1
?rovide the details of the method of calculating the primary cool-ant activity Icvols for the one percent failed fuel case, including purification cycling of primary systen, fission product release acsumptions from the failed fucis o f fcces of burnup and fuel temper-ature on fission product release from fuel, etc.
Provide all formulae, assumptions, and justifications for same. Justify the cicanup system reduction factors stated in the PSAR.
12.2 Provide a plot of fuel temperature versus the volumetric fraction of the total fuel at that temperature in the core at end-of-life conditions. Describe the cthod of calculation, state all assump-tions, and provide typi.:a1 radial pin profiles and the gross peaking factors u,,d.
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