ML20107F895

Forwards Responses to Conclusions on Page 11 of Eg&G Interim Rept Re Reg Guide 1.97,Rev 2,covering Reactor Bldg or Secondary Containment Radiation & Radiation Exposure Rate

ML20107F895
Person / Time
Site:	LaSalle
Issue date:	02/22/1985
From:	George Alexander COMMONWEALTH EDISON CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
RTR-REGGD-01.097, RTR-REGGD-1.097 9797N, NUDOCS 8502260224
Download: ML20107F895 (4)

Text

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Commonwealth Edison O One First Nitiorni Plaza. Chicsgo. Illinois Address Reply to: Post Office Box 767 Chicago, Illinois 60690 February 22, 1985 f

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Mr. Harold R. Denton, Director '

Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

LaSalle County Station Units 1 and 2 Regulatory Guide 1.97 Revision 2 Response to Iterim Report by EG & G NRC Docket Nos. 50-373 & 50-374 Reference (a) C. W. Schroeder to A. Schwencer letter dated June 29, 1982 ,

(b) T. R. Tramm to H. R. Denton letter j dated December 6, 1983 (c) A. Schwencer to D. L. Farrar letter dated December 13, 1984 ,

Dear Mr. Denton:

Reference (c) contained the interim report by EG & G Idaho I on Commonwealth Edison's response to Regulatory Guide 1.97 Revision

2. Included in the report was a request for additional information which was discussed with Mr. Bournia and Mr. Joyce of the NRC on February 20, 1985.

Attached are CECO's responses to the conclusions on_page 11-of the interim. report. One_ signed original and forty copies are provided for your use.

Sincerely, 8502260224 850222 G. L. Alexander PDR ADOCK 05000373 Nuclear Licensing Administrator

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F PDR bs-encls. O' t N cc: A. Bournia NRC Resident Inspector - LSCS g 9797N'

ATTACHMENT

' Conclusion 1 The licensee.should. provide the information identified in Section 2 of thistreport to document.their commitment on conformance to

' Generic Letter 82-33 (Section 3.1).

! Response This questica requests 8 items of information for each of the 69

-variables listed in our Reference (a)' submittal. Pages 2 and 3 of that submittal' addressed two'of the iteos-in summary form. The two L

relevant statements wereL" Seismic qualification ofELaSalle equipment

-was' completed to the IEEE~ 344-1975 Standards under the SQRT. program" and " Edison will complyEwith.the quality assurance' requirements Lusingeits' approved. quality assurance program, as described.in Topical; Report CE-1 as revised.

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The other.six items requested can nctTbe answered at this' time.

Reference-(b) contains the approved LaSalle: schedule for Detailed ~

! Control Room Design Review (DCRDR)'and for Regulatory? Guide 1.97

. . Revision 2. - The DCRDR-report lis-scheduled for submittal to.the NRC~

.by_11-01-85. The' review,will1 encompass the adequacy of~

instrumentation display,'information, arrangement,'and task ..

c . analysis. If deficiencies are' disclosed,. instrumentation additions, deletions, relocations,.or replacements will be required. It a >

, ;follows that withoutra. final determination-of; instrumentation:

e l acceptability. Land arrangement,ethat submission of= details-such-as instrumentTrange environmental qualification,_ redundance-'and sensor:

, Llocation,1 power:s,upply,flocation.of display, and schedule-of installation ortupgradetwould1beipremature.-. I believe this? fact;was controlling when the schedule-for submittingiaiReg.JGuide 1.97 final-

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report-(including.scheduleefor installation).wasl established and '

approved;as 08-01-86.:

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1After the:NRC reviews our1DCRDR report we will-beEableitoireevaluate' Ax 4thefschsdule7for submittingEthe: requested <information. -

L Conclusion 2'~~ (Review'er :does z. noti agree)) ' ER

Neut'ron flux--th's flicensee's

present instrument'ation ais -acceptables

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on an interimibasisiunti19 Category?l: Li nstrumentation:is developedz E

and!insta11edf(Section-3.3.1). .

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{ Ceco is1 pursuing: two coursesJof action.; LCurrently:we are : reviewing i ,

whether; recently. ' developed Lequipmentt meets . Reg.. ~ Guide 11.97 iRev. f 2.'

" Also' in : conjunction 1with :the .DCRDRiwel will ev'aluate :whither: this; W ~

parameterlis5 required.. Lor.can:be?clas'siffed as a: Categoryj3 variable.

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Conclusion 3 Radiation exposure rate--the licensee should show that the ranges supplied for-this variable encompass the radiation level at the

-instrumentLlocation (Section 3.3.4). '

Response-Revised 02-11-85 ISSUE'll. -VARIABLE E2' LE2:-; Reactor Building or Secondary Containment Radiation Issue Definition Regulatory Guide [1597. specifies that'" Reactor building or secondary containment area radiation" (variable E2):should'be monitored over the range lof 10-1-to 104 R/h-for. Mark I-and II-containments, and-over.the range of--1'to 107 R/hr for Mark III containments. The'

. classification for. Mark Is and IIcis Category 2; for Mark III,,the-classification is. Category I.. .

Discus'sion ' r~

As di.scussedsin the varidble CIA-position statement)(Issue 6),.

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-Secondary! Containment:AreaRadiation.isaninappropriateparame(er; t j ito uselto. detect'or: assess. primary containment leakage.'. <

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, ~ Conclusion.,

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ItLis, Edison'spositionthatthe/instrumentationTfor-this$ariab'e-

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-islnot1neededias theJplantsnoble gasieffluent.monitorsJare'more

' usefu14and :pract'icallin (detecting the primary ' containa'ent (leakgge c The'rangeLof~theinoble gas ~' effluent monitorsHat LaSalle01s110-' '

luci/ccJto;10.5 uci/cc which;meetsf the' RG J1~.97. requirements.-

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Conclusion'4: j ,3',, .

~ t ' { Status; of4tandbyi'poweriandf other energy ' sources-_-theslicens'ee1 , D'

should eshow4 that =the Tstatus :is" monitored. fortall'. recommended, powert

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%g 180thi on2sitecan'dioff-site power sources $y.areimonitored-fosEstatusiper/-

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' the J requirements-J of .;i Reg . : Guidefl . 97. Rev.~2. 5

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'k Conclusion 5 Reactor-building or secondary containment area radiation--the

~1icensee should supply additional justification for this deviation (Section 3.3.8).

Response

Revised 02-11-85 E3: Radiation Exposure Rate Issue-Definition RegulatoryJGuide 1.97 specifies in Table 1, variable E3, that

-radiation exposure rate (inside buildings or areas where access is

! required,to serd ce equipment important to safety)_be monitored over

- the ' range of 10 - to 104 R/hr for; detection of-significant releases,:for release assessment,_and forilong term surveillance.

Discussion Inigeneral,; access.is'not required to_any-areas ~of the secondary

' containment'to service equipmentfimportantito safety in a '.

po st-ac ciden t , situ'a tion .- If and when accessibility is reestablished

'in the'long term, it will_be_done.by a combination of portable-

> radiation survey instrument and post-accident sampling of?the,,

. 'secondaryicontainment-atmosphere.- The radiation exposure rate-g; ;monitorslinsi'de secondary (containment;at LaSalle= County Station;are' Je sfor: normal operation and1are not ~ intended ito continually Leonitort gross'~1nt'rusions off,the. reactor _'.s -fission products into . secondary .

containment. Theselmonitors areTinstalled inflow-or moderate tradiationiareastto identify abnormaluoccurrences-which-would; produce' Eradiation1 environments l ranging from a fewimilliradsiper' houre tol-several: rads . per.Thour :as : shown jin L FSAR LTable '12.3-13.f ? Abnormal 1 occurrencesTare any,incidentsLthat;do not propagate intofanytof'the' f , / accidents-discussedJin ChapterL 15:of.the1FSAR.- ,

[rTheNfollowing~listof-FSAR~informationcontains"thedataused;to b 3establishithe rangesEof theLa' nitors o Lin L FSARETable:12.3 13,o"Arsai ,

'. mRadiation Monitors.":

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,- ' s { 1..? 1(SAR7 Table 112.3 -3, !" Reactor < Building Design Data"1 a

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During' Full PoweriOperation".

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Conclusion It is Edison's position that the specified radiation exposure rate monitors inside the Reactor Building are used only to measure tt e radiation' dose rates during normal operation and to detect abnor.nal occurrences which do not constitute an accident while the reactor is critical. One area monitor is designed to detect and monitor fuel handling' incidents (including fuel handling accidents) and has an adequate range up to 103 rads /hr. Seven Reactor Building monitors have a range up to 10 rad /hr. The remaining twenty monitors have an upper range of 1 rad /hr or 0.1 rad /hr.

.Because the LaSalle design does not require access to a harsh

-environment area to service safety-related equipment during an accident, this dose rate variable is only used to determine abnormal occurrences and is provided from existing area radiation monitors.

This parameter is reclassified as Category 3 and the monitors furnished for this variable have ranges that encompass the expected radiation levels at their locations.

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