ML19210A449
| ML19210A449 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 09/02/1971 |
| From: | Morris P US ATOMIC ENERGY COMMISSION (AEC) |
| To: | John Miller METROPOLITAN EDISON CO. |
| References | |
| NUDOCS 7910300460 | |
| Download: ML19210A449 (4) | |
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UNITED STATES
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September 2,1971 Docket No. 50-289 Metropolitan Edison Company Attn:
Mr. J. G. Miller Vice President & Chief Engineer P. O. Box 542 Reading, Pennsylvania 19603 Gentlemen:
In our continuing review of your application for an Operating License for the Three Mile Island Nuclear Station, Unit 1, we have concluded that supplemental information and possible modificationa are appropriate in relation to your design and proposed operating plans. Details of these items are provided in the enclosure.
We have previously discussed these matters with your representatives in meetings at your f acility and in our offices. However, if you desire further discussion or clarification of the material requested by this letter, please contact us.
Sincerely,
/
/
Peter A. Morris, Director Division of Reactor Licensing
Enclosure:
Additional Information Requested cc: w/ enclosure George F. Trowbridge, Esq.
Shaw, Pittman, Potts, Trowbridge, & Madden Suite 1017 - Barr Building 910 - 17th Street, N.W.
Washington, D. C.
20006 Jersey Central Powee & Light Co.
Attn: Robert H. Sims, Vice President John L. Thorpe, Manager of Safety & Licensing 260 Cherry Hill Road Parsipany, New Jersey 07054 1487 042 7910gog gQ
ADDITIONAL INFORMATION REQUESTED FOR THREE MILE ISLAND UNIT 1 1.
TENDON SURVEILLANCE PROGRAM The proposed surveillance program for the containment tendon system specifies that lift-off tests are to be performed on one vertical tendon oni,y; no lif t-off tests of dome or hoop tendons are specified.
As discussed with your r presentatives, the proposed program specifies less surveillance than the minimum that has been determined to be acceptable for other facilities using pre-stressed concrete containment structures.
Provide additional information (1) to modify your proposed surveillance program to conform to programs defined by the published Technical Specifications for other facilities with similar containment structures or (2) to rigorously demonstrate the basis for the acceptability of the proposed program.
2.
INDUSTRIAL SECURITY The additional information that we require with respect to the industrial security program for the Three Mile Island facility has been discussed with your representatives. This information should be submitted as soon as it is available and in the manner suggested during our prior discussions.
3.
METEOROLOGY We understand that you are planning to conduct on-site meteorology experiments at Tnree Mile Island.
Provide a description and the schedule for the experimental program.
4 NET POSITIVE SUCTION READ (NPSH) 0F ECCS AND SPRAY PUMPS Calculations on the NPSH margin for the ECCS pumps and reactor building spray pumps were presented in Amendment 20 to your application.
These calculations were based on the assumption that the reactor building pressure was in thermodynamic equilibrium with the ambient sump water temperature at the time that the pumps start to recirculate the sump water. This assumption led to a postulated reactor building pressure of about 3 psig.
This assumption is in contradiction to our published Safety Guide No. 1, which states that no increase in pre-accident pressure should be assumed, and is also in contradiction to your NPSH design basis as defined in the Preliminary Safety Analysis Report for this plant. We have concluded that the recomm(ndations of Safety Guide No. 1 are applicable to your design, and that your presently proposed provisions for assuring adequate NPSH are unaccept-able.
Provide additional information to assure that the necessary NPSH conditions will be ret including infcrmation with respect to any requirec changes in dasign er plant cperacien.
In addition, 1487 043
2-provide a description of the tests that viil be conducted to verify the assumptions made in the NPSH calculations.
5.
TESTS OF FIRE SUPPRESSION SYSTEMS IN THE AIR INTAKE TUNNEL We have concluded that appropriate startup and periodic testing is required for the fire suppression system in the air intake tunnel.
Provide a summary of the tests that will be made, and submit appropriate revisions to the proposed Technical Specifications.
6.
ROD EJECTION ANALYSIS We have concluded that you should revise your assumptions for the postulated rod ejection accident analysis.
In this context you should assume:
(i) a primary system to secondary system leak rate, prior to the accident, that is consistent with the proposed Technical Specification limit.
(ii) a realiatic assessment of primary system to secondary system leakage during the accident.
(iii) fission product releases from failed fuel rods consisting of 10% of the radioiodines and 207. of the noble gases.
(iv) fission product releases to the atmosphere as the result of (a) leakage directly from the containment, and (b) leakage f rom the primary system to the secondary system and then to the atmosphere.
Provide a reassessment of the course and consequences of the rod ejection accident incorporating these assumptions.
Further, in your analysis of the associated nuclear kinetics provide additional information to substantiate the acceptability of an ejected rod worth of 0.65% 4 k/k.
7.
IN-SERVICE SURVEILLANCE MONITORING FOR VIBRATIONS AND LOOSE PARTS Provide information with respect to the program you intend to follow on Three Mile Island Unit 1 to develop a practical system for the continuous in-service monitoring of the primary coolant system to detect the occurrence of unanticipated vibrations and the p resence of loose parts within the system.
Include a description of the initial efforts to be pursued in this development and the schedule established for its completion.
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8.
DECAY iiEAT REMOVAL SYSTEM ISOLATION VALVES The suction line of the relatively low pressure (N350 psi) decay heat removal system is connected to a hot leg in the primary coolant system (design pressare of 2500 psi). The systems are separated by three electric motor-operated valves, two located inside the containment structure and the other outside the structure. The valve nearest the primary coolant system is interlocked to prevent inadvertent overpressurization of the decay heat removal system.
We have concluded that your present design and operating procedures do not provide for an acceptably low probability for inadvertent valve opening.
In our opinion at least two of the valves should be designed to close automatically whenever the pressure in the primary coolant system exceeds the design pressure of the decay heat removal system, and these valves should be provided with safety-class interlocks, preferably of different principles, to prevent opening of the valves with the valves with the primary coolant system pressure above that of the decay heat removal system design pressure.
In addition, we believe that a high-pressure alarm signal for the decay heat removal system should be provided in order to notify the plant operator of an abnormal condition. The valve automatic cle:.re systems and the interlocks should conform to the intent of IfEE-279.
Provide (1) a description of the modifications you will make to the decay heat removal system to reduce the probability of an inadvertent opening of the isolation valves, or (2) a comprehensive assessment of the facility design and operating procedures to convincingly demonstrate that the present design and proposed procedures provide an acceptably low probability.
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