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ATOMIC ENERGY COMMISSION C
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EI3 R. S. Boyd, Assistant Director for Boiling Water Reactors, L
GR.U;D GULF NUCLEAR STATION QUESTIONS PLANT MAME: Grand Gulf LICE':SI::G STAGE: CP DCCKET " '5ER : 50-416 RIS?0"SIBLE B:LCCH:
PWR01 REQUSSTED COMPLETION DATE:
February 9, 1973 APPLICA':TS RESPONSE.DATE NECESSARY FOR NEXT ACTION PLANNED ON PROJECT:
N/A DESCRIPTIO:: OF RESPONSE:
N/A REVIEW STATUS:
N/A Attached are the Accident Analysis Branch initial questions on the Grand Culf Nuclear Station PSAR.
Areas questioned include the site, i
enb neered safeguards for iodine removal, and design bases accident consequences.
During a recent site visit, Accident Analysis Branch personnel learned that significant design changes were bein', made in the Grand Gulf conuire.:n sys.tyms., and that an amendncat describ-in; these changes was in preparation des'ign basis accident Preliminary calculations of consequences yielded results exceeding iart 100 guidelines.
We anticipate major revisions to our calculational models as a result of the expected changes in containment design. This major change, if net submitted in the next few weeks, pay. affect the overall revie" cchedule.
- w Harold R. Denton, Assistant Director for Site Safety
,j a q y a Directorate of Licensing Enclosure (As Stated) cc: w/o enclosure A. Giambusso W. Mcdonald cc: w/ enclosure S. Hanauer J. Hendrie W. Butler G. Lainas G. Cetley W. Nischan E. Ad..non K. Murphy
". Zav doski G806160075 080606 PDR FOIA CONNOR88-91 PDR Q
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Clarify the specified exclusion radius distance of 780 moters.
Figure 2.1-2 of the SAR indicates a distance of 2460 feet (750 neters) as the distance from the SGTS release point to the nearest site boundary.
The SAR states that MP&L does not own all land within the specified exclusion radius, but that it does hold options to buy all land within that area not presently owned Indicate when those options will be exercised.
Define the plant restricted area, as discussed in Section 2.1.2.2 of the Standard Format.
Provide a map showing the restricted area boundary lines and discuss your. access control measures:
Proeide details on the bases used by you to conclude that expl'osions or the 1
release of hazardous materials shipped on nearby transportation routes or s.tcred in tae plant vicinity will'not advers,ely affect pla,nt operations.
Include' the f ollbw'ing deta'ils :
(a)
Ine' type's and lot qhantit' des oE hazardous materials
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8 stored or shipped, and the. number of shipments per l
unit time (day, month, year).
(b) The calculational methods or models used to predict the consequences of explosion or hazardous materials release on the plant.
(c)
Specific values of parameters used.a the calculations, e.g. distances, pressures, cor.contrations.
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GRAND GULF NUCLEAR STATION ACCIDENT.\\NALYSIS BRANCH QUESfl0N LIST (Q-1)
In section 6.4.2, evaluate the radioiodine loading of the filters (and resulting heat generation) based upon all the assuruptions stated in Safety Guide #3, including the fission product source term.
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Regarding Section 9.4.1.2, provide evaluation and location of the dual fresh air inlets to the control room ventilation system.
Provide the flow rates during normal and emergency operation.
Calculate the cfm of un-filtered air leaking into the control room from ducts, doors, and other openings when the control room is isolated (no f resh air uake-up).
Use the following assumptions in the analysis:
1/8 inch water gauge differential fo,r openings.
which may be effec'ted by external win'ds.
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1/16 inch water gauge differential for openings protected from direct wind ef.fects.
maximum design pressure differential for closed dampers on suction side of supply fans.
s 100 cu'bic feet air exchange par door open[ng, 'i.e.
ingress or egress, event.,
ihe assumption that no inlea.ka,ge,occu~rs'while the control room is isola,ted nay be.
op timis t ic.
An. analysis should,be performe'd to d'etermine the thyroid, whole body gamma and beta skin doses assuming the inleakage based on resulta from the above analysis.
With regard to Section 15.1.X.2(2) of the Standard Format, and your SAR Section 15.4.1.2.2.2.1, the radiological consequences of a LOCA must a
a be based on both /best estimato and / conservative estimate of fission product release.
Thus an additional analysis is required using conser-vative source terms.
(See Safety Guide No. 3).
I The fission product source term used to estimate consequen-es of a i
postulated loss of coolant accident as discussed in Section 15.4.1.2.2 of the PSAR is not acceptabic for the purpose of comparison with siting criteria.
Preliminary cel ulations for your containment design using an AEC Safety Guide 3 source term yield consequences over Part 100 guidelines.
Describe the changes in plant design you propose to meet the siting guidelines of Part 100.
Provide a com-parabic analysis using the source teret as described in AEC Safety -
' Guide #3.
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Provide the dose consequences at the site boundary and the low population zone boundary and in the control room following the design basis loss of coolant accident concurrent with the SSE due to ' leakage through,the main steamline isolation valves. Assume loss o[ all non-Category I (seismic). components, and assume main'steamline isolation valve leakage to ba at Technical Sp,ecification limits.
Deednstrate'that ledkage via the MSLIV's iti conjuaction yith primary contain=ent leakage following the I
pcstulated LCCh wiIl not resolt,in consp4uences exceeding.thS guideli'ne of 10 CFR'i) art 100 and of General Design Criterion 19.
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Preliminary calculations of the consequences of a refueling accident in the primary containment, using assumptions in Safety Guide No. 25, yield unacceptable consequences if the primary containment is not Describe the Technical Specifications which will be proposed secure.
regarding primary containment integrity during fuel handling operations within the primary containment.
Analyze the fisbion product movement in the spent fuel' air handling system (Pa3e 9.1-6) due to a refueling accident in the auxiliary building. Discuss instrument, response times, air flow, and exhaust duct isolation times. Demonstrate that the auxiliary building ventilation system will be automatically iso' lated bef' ore any significant.
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. release of radioact'iv'i,ty occurs as a result of the. refuel!.n?, accident'
' within the auxiliary building. 'Sec tion.15.4.2 contains an analysis of a f uel handlin,g accident' withih th.e primary containabSt only..
Provide an analy. sis of the off-site' consequenc.es of a befue-ling accident witiin
. the auxiliary buildi6g..
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Identify toxic material, such as chlorine. that may be stored en i
or in the vicinity of the sit y, which, assuming a container rupture, may interfere with control room operation.
Identify the location of storage on-site and lise the distances between the location of any such material and the air intake to the control roen.
Provide an analysis of the severity of such accidents, and discuss the steps to mitigate their consequences.
The description of the analysis should clearly list all. assumptions.
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Grand Gulf Nuclear Station Raview Schedule Dates for Containment System Branch Present Dates Proposed Dates (1973)
(1973) 1st Q to PM 4-6 5-4 9.
i 1st Q to MP&L 4-13 5-11 MP&L Response 6-8 6-22 Position &/or 2nd Q to PM 8-24 9-7 Position &/or 2nd Q to MP&L 8-31 9-14 r
MP&L Response 10-23 10-26 SER Input to PM 12-14 12-14 SER out 12-28 12-28 1
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.t uni Attached is an ANC letter which requests
$14,500 to provide technical assistance R. L. Tedesco an in revising CONTZXPT for Mark III application.
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.t.u u We should have this capability as soon as cc: H. Menzel possible so that we can use CONTEMPT during
. Cudlin un J. Glynn l
l the testing program.
I reco=end that we T o (8. -..ae.a.o are muuo try to get this additional funding.
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