ML20155F001
| ML20155F001 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf, 05000000 |
| Issue date: | 01/12/1974 |
| From: | Butler W US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Fraley R Advisory Committee on Reactor Safeguards |
| Shared Package | |
| ML20155E140 | List:
|
| References | |
| FOIA-88-91 NUDOCS 8806160269 | |
| Download: ML20155F001 (9) | |
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Q '<'v JAN 12 574 Occ;nt ::oc. 50-416 nad Si?-4L7 d
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'!ote to !'ayr: cod Fr.: ley, Txecutive Secretary h!vbory Co.citteo ca deictor Safe;:nards i
A'iS !ERS TO CLRTAIN I!NCIRIt'S 0:1 THC GnA?'O GULF N1'CLPAR UT'l:CIATI!!C STATIO i 7.e /L"P. 5.'nco.uittee on the Granet Gulf.uclear Station rc.:ur.sted at it,
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- c r ':1-22, l'.'73 neatinr. t! at the J.caulatory ctcff provide critten
- t..u'ic ra to certain inanirloo.
The enclored amn: crc are tcInc furnished for uce i.y t..e N ci'r.ittee i rior to i ts January 17-13, 1 7.'4 tecting to fern t:te basis for furthor discussiona with tl.c tientilatory staf f, l'or 'f our inf or:entien, the Re;ulatory staff will be represented at the Jnnuary 17-18, 1974 rmetiac.1.y technical ur:eciallots in the creas of alta scim:icity, containc.hnt cyster:a, reactor cysteras, and structurcl anclycos.
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- Jalter R. Lutlar, Chief
.l.irbt Jarer hvoctora Projects Trcnei. 2-2 Directorato oC Litonsing t,
Cnelosure:
As ut.ited i
2 Distribution:
Docket Files Lh*R l-2. File GF0wsley RP' Asst. Dirs.
RP Branch Chiefs TR Asst Dirs.
TR Branch Chiefs RCDcYoung MMaigret G806160269 080606 PDR FOIA PDR' CONNOR OB-91 W W--l-5,, :
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s ENCLOSURE t
RESPONSES TO INQUIRIES 1.
INOUIRY:
What is the basis for the time vs. quantity relationship of, hydrogen generation which t :e staff assumes in establishing design criteria for Grand Gulf and for other plants?
(pg. 28 of the transcript).
Related questions posed by the Subcommittee (in the discussion on pp. 15-28) should be addressed in the response, particularly Dr. Bender's question on pg. 16:
"Recognizing that the hydrogen release mechanism is arbitrary, is it reasonable to press so hard for getting the initiation of the fans and mixing in a period like 10 minutes?"
RESPONSE
4 The response to this question is provided in Section 6.2.5 of tha
'S E R.
2.
INQUIRY:
What mechanism or phencrena for failure might lead to ECCS degradation beyond that considered in the interim ac.ceptance criteria? Has the staff considered what such phenomena, mechanisms, and failures might i
be?
(Tr. pg. 28; related discussion starts on pg.15 and.also on pp. 61-63).
RESPONSE
The present Interim Acceptance Criteria and the recently issued Rule gcVerning "Acceptance Criteria for Diergency Core Cooling Systems for Light-Water-Cooled-Nuclear Power Reactors" requires that the ECC systems be designed to accommodate the effects of a single active
o 2-failure.
With this assumption, both sets of criteria set forth a minimal deg ee of acceptance performance standards.
Obviously, any additional failure or degradation beyond those already assumed would cause the calculated clad temperature to increase and could' exceed the allowable value (2200'F or 2300*F); that is, if a pump were to produce less flow than assumed in the analysis the effect would be to cause an increase in clad temperature and it might be characterized as a degradation of performance in the ECCS.
However, this.does not suggest that the core would not be adequately cooled.
Quite to the contrary, it has -been testified at length during the recent rulemaking hearing on ECCS that clad temperatures considerably higher than 2300*F could occur and a coolable core geometry maintained. However, the extent of permissible ECCS degradation can no,t be quantified at this time.
Thus, it is believed the clad temperature, may approach clad melting, and in some cases, perhaps some clad melting may occur and can still be tolerated.
The specific failure modes that can cause various degrees of degradation or consequences in arriving,at the present single failure assumption used to satisfy the criteria are disc,ussed above. Additional
- single failures beyond this to cause any single degradation have been considered as discussed in response to interrogatory 5.
In summary, the ECCS is designed to accommodate any single active failure and those consequences or degradations in systema resulting from it.
However, present systems have not been designed to accept two unrelated single failures and still satisfy the criteria discussed above.
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3.
INQUIRY:
What is the basis for the staff's conclusion that reliability of the hydrogen mixing system is acceptable? This includes both its reliability to function when called upon and reliability not to fail and to permit inadvertent steam bypass (Tr. pg. 39).
RESPONSE
The response to this question is provided in Sections 6.2.1.6 and 6.2.5 of the SER.
4.
INQUIRY:
What is the staff's opinion of the applicant's description of the capability of the safety related features and systems to withstand a hydrogen burning event?
(TR. pg. 60; related discussion starts on pg. 48).
RESPONSE
General Design Criterion 41 requires that systems to control hydrogen, oxygen and other substances which may be released into the reactor containment be provided as necessary to control their concentrations following postulated accidents.
We have' reviewed the systems for the control of combustible gas concentrations proposed for the Grand Gulf facility using the guidelines of Regulatory Guide 1.7.
We have concluded the systems are acceptable as indicated in our Safety Evaluation Report, Section 6.2.5 provided that the system be auto-l matically actuated.
The acceptability of the system is based on l
limiting the hydrogen concentration within the containment to below H
4 v/o which is conservatively considered to be the lower flammability 1
limit of hydrogen in air.
As' has'been our practice, our review is
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based on the system's capability to limit concentrations to this level and thereby preclude burning; we have not evaluated the cap-ability of the containment to withstand the effects of burning of flammable hydrogen mixtures.
5.
INoMTRV!
What is the effectiveness of the' ECCS consisting of 1 LPCI and of 2 LPCI's, in providing adequate cooling in the event of a LOCA?
(Tr. pg. 63) Can the staff evaluate the applicant's statements on this subject (Tr. pg. 61-63).
RESPONSE
j The staff has not made independent calculations of the peak clad temperature or metal-water reaction with core cooling by only one or two LPCI pumps. Although 'the staff could estimate the 4
flooding time with one or two LPCI pumps and calculate the tempera-ture up to the time of flooding, the accuracy of the peak clad tempera-ture estimates for the transient following flooding would be uncertain because of the uncertainty in the quantity of residual water remaining in the lower plenum and the effectiveness of the heat transfer at low flooding rates has not been specifically evaluated for a BWR.
6, INQUIRY:
Is there a means for ascertaining during the life of the plant whether core spray nozzles are blocked? (Tr. pg. 64).
RESPONSE
Blockage of individual or a few core spray nozzles cannot be detected unicss the water level in the vessel is lowered below the level of the core spray sparger and the system spray observed while the cort
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While this is done prior to initial l
startup, it is not performed during plant operation since all fuel' would have to be removed from the vessel.
Blockage of a large a
number of nozzles would be detected by the full flow core spray system test performed during refueling outages.
7, JNQUIRY:
What is an acceptable method for determining the representative set of stresses for a single direction, when considering more than one mode shape?
(Tr. pg. 74; related discussion starting on pg. 66).
RESPONSE
The acceptable procedure for deterr.ining the representative stresses for a single direction is to take the square root of the sum of the stresses obtained from each mode. For closely spaced modes, absolute sum of the stresses for the closely apaced modes is first obtained and then the square root of the sum of the squares approach is used co obtain the resultant stresses from all modes.
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