ML20234D108
Text
Di. tibution cc:
Mr. C. C. Whelchel Pacific Gas & Electric Co.
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Doc. Rm.
Formal Docket No.
Suppl.
50-205 R&PRSB Files
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DIAR Files P. A. Morris - 2 E. G. Case t
R. H. Bryan I
tacific Gas & Electric Company F. N. Watson 245 Market street San Francisco 6, California Attention:
Mr. Richard E. Petersom General Counsel Gentlemen:
i As you know, the Reguistory staff of the commission is reviewing your i
application for a construction pet: sit for the proposed Bodega Bay Atomic Park Unit No. 1.
Transmitted herewith is a request for additional techni-cal information which the staff believes is necessary to complete its review of your application. Your reply abould be submitted as an amendment 1
to your application; three copies signed ander oath or afficantion and 47 additional copies should be provided.
We understand that you have research and development programs in progress which are expected to expend your knowledge in certain areas of the facility design. Where we have asked questions concerning design features included in these areas, you shemid, as provided by Part 50.35 of the commission's regulations, describe the reaserek and development program to be conducted te resolve the question er synestions. Also, there may be instances in which you are unable to answer gesstions at this time because details of the plant design have not been completed. la these instances, you should describe the criteria upon which the detailed design will be based.
l Please do not hesitate to contact us if you have any difficulty in under-standing the intent of the enclosed questions. The additional information '
j which you submit in reply to this letter will be evaluated by the staff and further discussions will be held with you cencerning any unresolved netters.
We will also advise you further concerning scheduling this project for discussions with the A M S.
Einevrely yours, Edson i Case, Assistant Director Facilities Licensing Division of Licensing and Regulation
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PACITIC GAS Alm ELECTRIC CCBtPANY BODEGA AttR$1C PARE - ONIT NO. 1-QUESTIONS RAISED BT TRS
_,g1 VISION OF 1.1CARSLNG.AgtJEguj4 TIM
'I RELATIVE TO CONSTRUCTION PERMIT AFFLICATION 1.
Page 1-1,*
As required by Paragraph 50.35. Title 10 of the Code tage 111-1 of Federai Regulations, (see revision effective
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Jaseary 29. 1963) describe ~the research and develop -
l' meet programs which will be conducted to resolve tMs4. design problems having a bearing on the safety of'the pisat.
2.
Page 1-1 Delineste the divistem of responsibility between Pacific Gas and Electric Company and General Electric I
company, in this project.
U 3.
1* age 11-1 ladicate whether a filter will be provided for the 1
air ejector discharge to the stack and the approximate delay time provided ahead of this filter. (Figure 11-1).
i 4.
Page 11-2 Justify, through reference to previous operacias.
experience and experiments, your fuel element design using stataloss steel casodias havlag a thickness of 11 mills.
5.
Pase 111-2 ladicate the height of water shielding which will be provided above a fuel element when it is being trans-ferred through the gate opening from the reactor to i
the storage pool. State the calculated radiation.
j dose rates at the pool surface under these conditions.
6.
tage 111-3 What precautions are being taken to ensure that a j
rupture of a time penetrating the containment wall will mot caese containment rupture?
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- Designates reference to section and pages of Exhibit C of the Bodega Say application for cemetraction permit.
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Page III-3 state the manner in which the biological shield will j
be cooled sad the design criteria to be fo11ewed in
'l dealgmiss this cooling system.
j 8.
Page III-4 ht are to be-the closing time specifications on the containment isolation valves and the turbine trip
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Fig. 111-2 State whether the drywell will be provided with an airlock entrance and Ladicate yewr intentieas relative to estering the drywell during operation. state the type of instrumostaties which will be provided to indicate recirculates pteep performance. ladicate what assurance there is that the equipment will operate satisfactorily under 150*F an61 eat conditions.
- 10. Page III-5 Indicate how the spent fast will be handled properatory to shipment off-site.
11.
Page III-6 How many scrans can be accesplished with the stored eastgy La the scram acciammalatorst-
- 12. Page III-7 State the core inlet sabacealing at rated conditions.
- 13. Page III-8 ht is the reactivity worth of starting a recircula-tion pump under various seresi and abnormal senditieast Are interlocks provided te (a) prevent starting a pump when the leep valve is open and (b) limit the rate of valve opening?
- 14. Page 111-9 h t conditions, such as low condesser vacuum, can prevent ose of the main feedwater pumpt
- 15. Fage III-9 What is-.theireactivity worth of tha. liquid poison systen sad what is its rate of reactivity additionf 1
What is the worth of the system when the refueling pool is connected to the reacter vesself
- 16. Page 111-10 What is the hest removal capacity of one tube bundle' of the emergency condenserf
- 17. Fig. 111-10 D~escribe the>research and developeast progree which I
has been or will be conducted to support the see of i
the internal steam separaters, and indicace the separation efficiesey expected.. Indicate the date l
which are presently available upon which the desiga will be base _d.:,
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State the desi a criteria to be followed in sising l
- 18. Fig.111-11 5
rage XY the piemus space in each fuel rod sad is preventing i
collapse from esternal pressure.
- 19. Page 111-10 Describe the techniques and equipment to be used for decay heat removal.during the first hour.after shot down from fati power under the assumption of failure
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of the entire feedwater system.
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- 20. Page 111-10 What is the heat removal capacity of the aestgency core spray system 7 Assume a mejor rupture of a recirealattom pipe line has caused the water level to fall below the core..
- 21. Page 111-10 Ubat are the consequences of a leak in a recirculation-loop that esoses the water level in the core to fall' in spite of maximum feedwater flow while the vessel pressure ramalas above 150.psit I
22.
Pese 111-15 Do say pipes or tubes._such as pneumatic er hydraulic lines, lead free the drywell area or the suppression poet tank-to the control room? If so, discuss the potential for the transmission of radioactivity through these conduits to the control room if a severe acoident occurs in the drywell.
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- 23. Page 111-16 Show that the start-up source will be capable of producing a satisfactory response in the nuclear instraumataties in view of the proposal that the amelaar instrumentation is to cover saly a range of nine decades.
24 Fage 111-18 Justify the emission of the followias as reactor scram signals a.
Seismic abock b.
Low control rod drive system hydraulic pressure c.
Low reactor pressure
- 25. rage 111-14 The resetor scram system includes a number of solenoid valves and other electrical devices. Is the reactor authentically misst down any time there is an interrup-ties of electrical power to these devices or to the
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control chain which activates these serse devices?
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- 26. Page 111-18 Explata how the two-channel reactor protection system l
is " fail safe" if both channels east be de-energized l
to produce a reactor scram. Is there any way, such as l
transistor shorting, in which a channel may fail and yet asistain sa energized outputt a
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- 27. Page 111-24 Describe the ekielding that will be pieced below the reactor vessel to permit maintenance operations in the "sub-pile" room.
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- 18. GENERAL If an eartt% or some.other occurrence should interrupt the supply of see water, what ausiliary sources of eesting water will always be available I
and hou loss could these weter supplies be used' to dissipate reacter decay best!
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- 29. GEWt.RAL ht reacter operations reem1 ting in reactivity chaeps will be permitted while the drywell vessel head is removed?
- 30. GassaAL ht water clean-up equiposat and off-gas removal equipment will be provided for the removal and control of fission products in the fuel storage pool?
- 31. GENERAL Explain the criteria to be followed in determining Page 111-4 whether both or saly one of a pair of isolation valves is to receive automatic elemure sigamis.
- 32. murmu.
h e is the magnitude of the pressure rise in the Refueling Centa4====t Building in the event of a feLiure of-a reacter auxiliary system (for example, a leak in the amargency condenser system er the high pressure deniasraitser)? h e is the desiSE Pressure rat.ing of the Refueling Containe nt Buildias?
- 33. Paa* IV-1 Justify the use of the proposed minimes burnous ratio criterien for the design of the core. h t burnout data correlation has been used in burnout analyses?
At what reactor power level is the burmont ratie of 4
1.5 expected to occurt At what reacter power level (steady state) would the burnout ratio reach 1.07 Discuss the asture of any conservatism used is estimating-the beat flumes used La burnout ratio calculations.
h e minimes burnout rette is reached as a consequence
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of a less of feedwater flow accident?
- 34. Page IV-3 State tdwre poissa curtains will be located ta the core, how they will be sepperted, and how they.will
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- 35. Page 17-4 state the average steam volmes fraction in the core at rated conditions and the average exit as well as the hot channel exit volmen fractions. Describe the research and development and/or analytical programs which* have been or will be conducted to determine the stability of the reactor at proposed conditions.
- 36. Page IV-6 State the design abutdown sesegin for the most reactive condition and at the most reactive time in cora life.
- 37. Page V-10 Describe the catch basins and how they will function.
2 Discuss whether they can overflow. Estimate maximas I
radioactivity concentrations to be allowed to accumwiste la a catch basia, both in solution or suspanaion and as deposited material.
- 38. Page VI-2 Provide sa analysis to demonstrate that the proposed design, operating procedures, and release limits for the circulating water and waste disposal systems in conjunction with the proposed monitoring equipment and procedures assure that no person will be esposed to concentrations of radioactivity in axeese of those permitted by Commission regulations if the marina l
environment is subjected to the effects of the operations contemplated. This analysis abould take into accouac re-concentration effects and other pertinent factors.
- 39. Page VI-2 What is expected to be the total acnumi discharge of Page VI-3 radioactive liquid and gaseous wastes from this plaut?
i 40.
Page VII-3 Delinesta the conditions under which (1) the asesrgency condenser will be in service while other containneat isolation valves are closed. (2) the meergency condenser toolation valves will be closed while other containment isolation valves are open and (3) all isolation valves are closed.
41; Page VII-9 What are the consequences of as accident in which the feedwater system fails and operator action or a reactor t
high pressure signal causes the isolation bleed valves to openf 422, Page VII-8 Describe the gas holdup system and state the system capacity.
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- 43. Faga VII-14 How far.from the nearest earthquake fault or branch fault is tho' reactor to be located? How far from the i
San Andreas fault is the reactor to be locatedt
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- 44. Page VII-15 Describe the research and development progree which will be eaaA=eted to support the belief that n' 2700 Mw-see nuclear excursion wf.11 met.reseit la damage to the reactor pressere vessei.. Indicate any design features which could reduce the likaltheed or magnitude of a red drop out accident.
- 45. Page VII-14 Diseues the possible effects on this plant should'an earthquake eause' displacements along minor faults under the plant.
44.
Fage VII-17 ht are the consequences of a major steen line break in the pipe tummel if the accident occurs during la-version eenditions? Treat both a esas in which the coatsimment isolation valve faections properly and a case in which the isolation valve fails to close.
- 47. Fage VII-18.
ht are the consequences of the MC04, taking no credit for fallout in the drywell and taking credit for plate out only as provided in' treatments of i
similar problems in TID 14844f
- 48. Page VII-23, 24 Submit samples of the calculations used in determiatag-the doses rates set forth in the Radiological Effects of Major Accidents.
- 49. GENERAL Describe the test program to be used to demonstrate the acceptability of the proposed cocetrol rod drivs system and indicate the acceptability criteria.
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