ML19210A460
| ML19210A460 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 11/22/1967 |
| From: | Morris P US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Neidig R METROPOLITAN EDISON CO. |
| References | |
| NUDOCS 7910300483 | |
| Download: ML19210A460 (5) | |
Text
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UNITED STATES i
\\k ATOMIC ENERGY COMMISSION Distribution:
F. @ C
AEC Document Room WASHINGTON. D.C.
20545 Su 4
REG Reading DRL Reading IN REPLY PEFER TO:
Docket No. 50-289 NOV 2 21S67 RPB-3 Reading Orig: BGrimes M. M. Mann R. S. Boyd b
L. Kornblith, CO (2) g F. W. Karas L. L. Kintner D. F. Ross bec:
J. R. Buchanan, ORNL Metropolitan Edison Company P. O. Box 542 Reading, Pennsylvania 19603 Attention:
Mr. R. E. Neidig Vice President Gentlemen:
At our meetings on October 17-19, 1967, we indicated that additional information was needed to complete your applica-tion for a construction permit for the Three Mile Island reactor. We generally discussed the type of information that would be required. Much of this information is required because the answers to previous questions were insufficient.
We have prepared the attached list of questions to illustrate the kind of information that is required. We will need to defer completion of our review until the additional informa-tion is provided.
Sincerely jours, Jrignal Siped by Petsr A. Morris Peter A. Morris, Director Division of Reactor Licensing
Enclosure:
Requested Additional Information cc: George F. Irowbridge, Esq.
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13.0 GENERAL 13.1 Provide an outline of the emergency plans to be followed in case of a major accident at the facility.
13.2 Provide a parametric study of various masses, energies and Lmpact areas which might result from a breakup at overspeed of the last stage wheel of the turbine to show that the wheel section chosen is indeed the worst missile.
13.3 Provide a brief description on vendor qualification on the waste evaporators that will assure a 104 decontamination factor.
14.0 SITE 14.1 Discuss the design of the dikes and intake structure for protection against erosion during flood conditions. Can " ice jams" occur on the Susquehanna, and if so what forces are considered in the design of the intake structure?
14.2 The flow model of the river from which the flood protection levels were established is not acceptable. The rationale for using such a program is that if a profile of a known flood can be matched by a computed profile over a significant reach of the river the parameters used in the calculation can be used in the computation of higher floods. Since the computed profile grossly conflicts with available data, the extension of these computations to a larger flood is not applicable. A recomputation in which a longer reach (perhaps 5 to 7 miles) is taken into consideration might produce acceptable results.
Additional river cross sections might be required to justify the final results.
15.0 STRUCTURAL DESIGN 15.1 Clarify the statement made in response to question 7.2 which indicates that horizontal and vertical " frequencies" wf.11 be added.
Indicate that the appropriate stressss will be added directly and linearly in all cases.
15.2 We understand that calculations of thr. response of the primary system internals to simultaneous earthquake and accident loads are being performed in response to our previous question 3.3.
Outline the scope of the calculations and the technique to be used. Provide a schedule on which the preliminary and detailed calculations will be performed and the date when the stress and deformation limits and loading combinations which you consider appropriate will be deter-mined. Provide any results which you have obtained to date.
15.3 Clarify your response to question 3.2 to indicate the provisions which will be made to insure that cranes cannot be displaced from the track.
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,.? i Vith regard to the locacion of vital scructures on dissimilar foundation it:erials provide (1) a detailed description of the transient analysis to be performed to determine the stresses and deflections of critical piping between beildings and specify the criteria for stresses and deflections, (2) discuss provisions to be made to prevent physical interaction between buildings due to different seismic response or settlement and tilting, and (3) the criterion for allowable settlement or tilting of the turbine building and the method used to calculate this phenomenon.
15,5 Clarify the response to question 3.2 to indicate the load combinations to be applied in the piping design.
C'_arify the limits to be imposed on (1) net principal tensile stress excluding
.5 bending or flexural stress and (2) net principal tensile stress when 1ceal 5tading arising from thermal loads is included.
~3' atate whether the equipment in the turbine and auxiliary building whien is necessary to a safe shutdown is included in the tornado protection criteria.
O REA0 TOR DESIGN
- n recent discussions you indicated that in-core thermocouples would be installed in the first reactor of this type to provide confirmatory information on the thermal characteristics of the core. In this regard, since the Three Mile Island reactor will be one of the first of this type on the line, please provide the information requested in Question No. 4.7 transmitted by our letter dated Au gus t 25, 1967.
Flease respond to the concern expressed in the ACRS letter on the Oconee Station dated July 11, 1967, that further evidence should be obtained to show that fuel-rod failure in loss-of-coolant accidents will not affect significantly the Ability of tha ECCS to prevenc clad melting. Please index your PSAR and supple-2*tts vtth respect to the other concerns expressed in the letter.
4 ovide a detailed cutline of the research program required to ~srify the analysis 1* hads on therztl shock effects in thick plates under stress below the ND"
- erparaturs.
Idantify any other area related to the pressure vessel and piping
- hermal shock problem tnat requires a research and development program for c reof-of-principle, and outline the required program.
We relieve that research and development above that which you have indicated v.11 be required to justify the use of core barrel check valves as a solution
- the ste4m bubble problem. Further consideration should be given to testing
( vibration effects on the valves (caused by core barrel vibrations) and (2) 5 04 charseteristics in the reactor after loss of a valve. We believe that if
- t a Icar of a valve is not detectable, the DN3 ratio at the overpover condition 3
titer 1:ss of a valve must be not less than 1.3 (based on the W-3 correlatien.
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. 17.0 ENGINEERED SAFETY FEATURES 17.1 We require assurance that the design of the heat exchangers for the fan coolers will be fully capable of removing the specified heat load under the steam-air accident conditions. We understand that rather than submit a preliminary design, you prefer to specify that proof tests by vendors will be run under simulated accident conditions. Specify the type of tests to be required and indicate how degradation of performance over the lifetime of the plant will be taken into account.
17.2 Provide further detail on the fluid block system including the size of the fluid reserve and the consequences if one of the isolation valves on a line did not close. List the lines penetrating containment which will be affected by this system.
17.3 We believe that the ability to take samples from the recirculated water after a loss-of-coolant accident should be maintained to allow determination of the boron content and chemical composition of the liquid. This is particularly important whan a chemical iodine removal agent is relied on.
Indicr*.e (1) the method by which sampling will be accomplished, (2) how soon after an acci-dent the sample could be taken, (3) whether adequate equipment and qualified personnel will be available on-site or if the sample would have to be trans-ported, and (4) precautions to be taken in sampling to avoid major radioactivity releases.
17.4 Provide a complete, detailed list of experiments which must be performed to substantiate the use of a chemical spray system which is relied on to meet Part 100 site guidelines.
You may indicate those experiments which you believe will be carried out by other programs but you should indicate clearly your responsi-bility for obtaining the necessary data in case p.ograms other than your own do not produce conclusive results. The following list of areas constitutes a necessary, bot not necessarily sufficient, list of items which must be resolved by the research and development program effort.
(1) Verification that Griffith's model is applicable to an air film around a reactive spray drop over the temperature and pressure range of interest.
(2) Determination of the effect of simultaneous water condensation on the iodine removal rate.
(3) Verification that Griffith's model is valid over a wide range of initial iodine concentrations and forms.
(4) Stability of the solution in the pressure-temperature-radiation environ-ment over long time periods.
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. (5) Verification of scaling assumptions.
(6) Projection of efforts if (1) Griffith's model and/or (2) the sodium thiosulfate reagent must be abandoned.
(7) Simulation of injection system.
17.5 In view of the large reduction factor which must be supplied by the proposed chemical spray in order to meet Part 100 and in view of the uncertainty in the fraction of iodine in the containment which is in the organic or particu-late forms, we believe that all practical measures should be taken in other areas to reduce the potential off-site dose after an accident. This should include reducing the leak rate to the lowest practical value.
17,6 Discuss the consequences of opening a valve from the containment sump during injection after a loss-of-coolant accident if the sump is empty and the corresponding case for the borated water storage tank.
17s7 Discuss the consequences to containment integrity aad doses to the public if the design b7 sis loss-of-coolant accident were to occur after the plant had been operatrag with steam generator tube leakage (at least 10 gpm) and secondary safety valve leakage. Discuss the implications of the calculation with respect to technical specification limits on operation of the plant with generator tube leakage and secondary safety valve leakage.
_.0 CONTROI. AND INSTRUMENTATION f.1 In response to a previous staff question (reference Supplement No. 1, Question 4
No. 9.2) relating to the performance of equipment located within containment, you stated that the fan and valve motors will have a system of insulation and enclosure which has demonstrated capability to perform under the post-accident envircnment. Please confirm that this capability will be determined by proto-type environmental tests conducted under Metropolitan Edison's supervision that will illustrate the ability of the equipment, including insulation and lubricating systems, to function in the accident environment.
Indicate the
- ceditions and length of time for which the prototype will be tested. Sections 3.7 and 4.4 of the IEEE Standard, Nuclear Power Plant Protection Systems (Revision 9) should be addressed.
Assuming that it becomes necessary to abandon the control room during full power
,e operation, please discuss the procedures that would have to be performed exter-ral to the control room which would ensure a safe shutdown of the plant for an indefinita time.
Include, in your discussion (1) the instrumentation that would h required to monitor vital plent parameters, (2) components that would require manual actuation, and (3) the available times to take the actions necessary.
Ocincident loss of off-site power should be conside:ed. How much of the above
- apability will be incorporated in your design?
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,1 km oc o 18.3 Analyze the con;equences of an accidental phase reversal at an emergency bus under accident conditions. The single failure criterion should not be used as a basis for the ans.lysia in those cases where machinery rotating in reverse has an adverse effect on its redundant counterparts.
id.i Discuss, in detail, your criteria relating to the physical separation of the installed instrument and logic channels which initiate protective and emergency safety feature action. This discussion should include:
(a) separation between redundant instruments, (b) separation between redundant relays and breakers,
( c) routing of redundant wiring, (d) permanently installed test equipment which may be common to redundant instrument channels.
List the emergency equipment which is powered by the engineered safeguards busses. List the expected loads and luading sequence and discuss your philos-
- phy of loading the diesel above its nameplate: rating. Under what conditions will the engineered Safeguards busses be tied together? Will the tie be auto-netic or manual?
.5 On page 9-41 of the PSAR you state that there is manual provision for switching to full recirculation for post-accident control room ventilation. Please dis-cuss your justification for the absence of automatic switching in response to a signal indicative of an accident conditicn.
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Vtat are the expected ranges of the atmospheric and liquid monitcring systems?
Indicate the relauionship between your design basis accident analysis and the ranges, sensitivities and detector locations of the radiation monitoring system.
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