ML20248G990
| ML20248G990 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 04/03/1989 |
| From: | Moran D Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8904140024 | |
| Download: ML20248G990 (51) | |
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( o, UNITED STATES fL $ NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 j
/ l April 3, 1989 Docket No. 50-260 l
APPLICANT: Tennessee Valley Authority FACILITY: Browns Ferry Nuclear Plant, Units 1, 2 and 3
SUBJECT:
SUMMARY
OF MEETING ON MARCH 15, 1989 WITH TVA CONCERNING BROWNS FERRY SAFE SHUTDOWN TECHNICAL SPECIFICATIONS CHRONOLOGY i On March 15, 1989, the TVA Projects Division staff met with the Tennessee Valley Authority (TVA or the licensee) at Rockville, Maryland to discuss the Appendix R safe shutdown program implementation and appropriate Technical Specifications (TS). The attendance roster is included in Enclosure 1 and the viewgraphs used by TVA are found in Enclosures 2 through 4. TVA 1989 detailed the manner for generating in which the Browns theyshutdown Ferry safe are implementing (NRC TS Enclosures guidance 2 and 3). of March The licensee also discussed its Appendix R implementation methodology and outlined the specific safe shutdown program for Unit 2 (Enclosure 4). ! TVA stated that they will submit the Browns Ferry Appendix R safe shutdown program with the safe shutdown TS submittal. The licensee plans to provide the staff with this submittal by April 15, 1989. TVA also plans to place all necessary requirements in the TS and then reference the safe shutdown program document so that listing of all equipment, safe l shutdown instructions and references will be in one location. The staff indicated that the TVA approach appears to be generally acceptable; however, the staff stated that TVA did not completely follow the TS guidance in the staff's letter dated March 3,1989. Specific areas where TVA deviated from the staff's guidance are:
- 1. Failure to provide a list of Appendix R related safe shutdown equipment in the body of the Technical Specifications.
- 2. Failure to include the limiting conditions of operation as described in the letter in the body of the Technical Specifications.
8904140024 890403 PDR ADOCK 05000260 P PDC l
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1 i After discussions of the above issues, the staff closed the meeting by stating . that the TVA proposal appeared to be a viable alternative to staff's guidance ( as provided in its March 3, 1989 letter. The staff indicated that it would study i this proposal and provide a reply to TVA shortly. Original signed by David H. Moran, Project Manager TVA Projects Division Office of Nuclear Reactor Regulation
Enclosures:
- 1. Attendance List
- 2. Safe Shutdown Tech Specs Chronology
- 3. Technical Specification Change
- 4. Appendix R Implementation Methodology
- 5. Appendix R Safe Shutdown Program cc w/ enclosures:
See next page Distribution Docket File NRC PDR Local PDR Those on Attached List
/1 0FC :NRR:TVA/LA :NR gTVA/PM :TVA: P :TVA:AD/P l
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NAME :MSimms : 'n : a s :RPi n :SBlac9 DATE : 03/8/89 03/e;0 89: :03/4/89 : 0.f/y89 : : 0FFICIAL RECORD COPY
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General Counsel Regional Administrator, Region II Tennessee Valley Authority U.S. Nuclear Regulatory Comission 400 West Sumit Hill-Drive 101 Marietta Street, N.W. E11 B33- Atlanta, Georgia 30323 Knoxville, Tennessee 37902 Resident Inspector / Browns Ferry NP Mr. R. L. Gridley U.S. Nuclear Regulatory Comission Tennessee Valley Authority Route 12, Box 637 SN 1578 Lookout Place Athens, Alabama 35611 Chattanooga, Tennessee 37402-2801 Mr. C. Mason Dr. Henry Myers, Science Advisor Tennessee Valley Authority Comittee on Interior Browns Ferry Nuclear Plant and Insular Affairs P.O. Box 2000 . U. S. House of Representatives Decatur, Alabama 35602 Washington, D.C. 20515 Tennessee Valley Authority Mr. P. Carter. Rockville Office Tennessee. Valley Authority 11921 Rockville Pike Browns Ferry Nuclear Plant Suite 402 P.O. Box 2000 Rockville, Maryland 20852 Decatur, Alabama 35602 Mr. Oliver D. Kingsley, Jr. Mr. D. L. Williams Senior Vice President, Nuclear Power Tennessee Valley Authority Tennessee Valley Authority 400 West Sumit Hill Drive 6N 38A Lookout Place W10 885 1101 Market Street Knoxville, Tennessee 37902 Chattanooga, Tennessee 37402-2801 Chairman, Limestone County Comission P.O. Box 188' Athens, Alabama 35611 Claude Earl Fox, M.D. State Health Officer State Department of Public Health State Office Building Montgomery, Alabama 36130 l. l-l'p
= _ _ _ _ _ - - _ _
I t E,NCLOSURE 1 ATTENDANCE LIST BROWNS FERRY, UNIT 2 APPENDIX R Name Affiliation David H. Moran NRC Charlie Anderson TVA Joe Bynum TVA D. Zerinque TVA P. Carier TVA R. Wescott NRC John Lee Sparks NRC Tom Temple TVA James East TVA Bob Pierson NRC George Hubbard NRC l Joe McCarthy. .TVA Peter C. Hearn NRC
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- P. P. Carier, Acting Manager, of Site Li. censing, PAB C-BFN v :
( .(OM : Project Engineer Br'wns o Ferry Engineering Project, H.B. b Bounds,ineering,(EDBA10,'BFN.Lr I.u Lear Eng ' v
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SUBJECT:
BROWS FERRY NUCLEAR PLANT (BFN TEClulICAL SPECIFICATION a ti CHANGE
. )k APPENDIX"R SAFE SHUTDOWN P ,
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j This memorandum transmits.the proposed technical specification
- changes to incorporate the BFN Appendix R Safe Shutdown Program g into the BFN License. p,.j,,.jq.g g,; ,
II. JUSTIFICATION'J . .;. 4, . The attached proposed technical specification change package provides the necessary reference'to and administrative controls for the BFN Appendix R Safe Shutdown Program. This program is currently under development and the attached package is , consistent with it. The attached proposed changes are I administrative in nature and do not any equipment. These changes simply refe.in rence themselves, the program affect and . provide the necessary controls to ensure program revisions are . *
' in accordance with 10CFR50.59.' This is consistent with NRC
,, guidance in Ceneric Letter 86-10. 'The changes shown for , . . r. .r' J ,
technical specification section 6.0, " Administrative Controls," I' are based on the recently submitted (to.NRC).. upgrade of the f j ,., entireaection,6.0.jg.gjg. . . plg....I 3
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" Based on the nature of changes in the' attached package and the -
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discussion above no formal Nuclear Engineering ( E) safety anal D andEsiaisrefulred.1 ustificat to on) The attached enclosures 2 Description and 3 th(No Significant Hazards consideration) are adequate address e
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- - . . u.specification III. ACTION I
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g i . I Please provide the' attached technical specification change ' i package to Plant Operations. Review Commlttee (PORC) and fluclear
- Safety Review Board (NSRB) for their approval prior to I submitting to the NRC. If any additional information is i l
1 needed, please contact. Tom Temple (assigned co-sponsor from NE) I at j extension 5044. ,I ;
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IV. RESPONSE REOUEST . I 'o ' (1,;k.}y ., , l],,, L
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proposed License Condit lon vill reed e fo '
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- Browns Terry Nuclear. Plant, shall' laplement[li.c end memntain i in ofrect all
~~-~ provisions of"t,he approved AppendixiR Safe Shutdown Progrse as described -
in submit.t.a1 det,ed TOXOXG and es;epproved In.NEC SER det.ed YYYYYYYY . subject, t.o t,he following provision,:j.
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. . ,.4 p i n i, ,j l(i. s The !!censee any make changes t,o t,he approved Apperidix R Shutdown .
Progrant without prior approve!' or, t,h'e Commission only if t, hose changes .
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would not, adversely effect, t,he abilit,y,M['jshut,down meint.ain safein t'he event; of
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i '..-. J. 4.n 6 , c 1.0.NN tL y hiM % . . - . APPENDIX R SAFE SHUTDOLIN PROGEN; lf]t
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l BFN how developed en Ap.sendix R Safe Shutdown Progres. .
' This Progrea is to . j cneure that the equipment required by the . Appendix..R . Safe Shutdown Analysis is meintained and demonst, rated funct lonel as follows: h:' , ii.s .- .- i
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't Sete Shut .',.
,The functional requirement,s of the.. Program.a.down, systems ... ___ .end . equipment, "
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. . . .l . are out. lined in Section III.- of .
. . /+ the.M.l4,.d ' p Si.1.@lp .
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- 2. Test,ing and monitoring of t.he Appendix R 8ste Shutdown'syst.eas and -
'~ .l~~]%;l equipment are defined in Section VIlof the Progree ,
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~ ~ . .. , cannot, perform t, heir. Int. ended ' function p. appropriate. compen setory .a ,r-i. . ee .:.r ..h.L. .,
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,4. Changes sede t,o the BFN Appendix R Sefe Shutdown. Program vill be . . ., , s. . p- c.- . . . . ~. .p "
.., . processed inaccordence with. License ,Condi. ti.on. l.*_
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RESPONSIBIIITTTS
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6.5.1.6 The below. PORC shall be responsibleffor the activities listed a
\.',j The PORC may delegate 'the performance of reviews to
, subcommittees or qualified reviewers bue vill maintain jw
~,...-.. . -.. . : . . responsibility .T for these reviews.~.{
a.
- 1~."* '"'" ' "' *".I Y: .? ~ ' ' ' '
= = - - - - - . - - - - -
. Review to procedures, equipment, safety evaluations' fort :(1) procedures and (2) change
,,4.
. . a..r.:: t..,J. .: ,..-.s :6 systems or facilities to verify .
o qu,,gg,,,. that such actions ., do not, constitute, an ,unreviewed safety . i j. , ; g dl
. t. . h .l . . s . sl, ).' . /.i '1 ,* g
- b. ' ,
. . . . . Review of safety evaluations [for proposed tests and ,'
g g experimen:s tha. t aff.ect nuclear safety. o. l .. . o .. . .
,a
. n ,1,:, m ..'3.,, L- ,u, i
5[!.'.
-.- ;.'.;. _ ;f/f-.;" c. . Review of all proposed 7 changes! . tos
..~ ... . .-~,.,,,p.
- :. ~,:; .:L- .
.e .4 -
Technical S d -g 0
. . . . .. " " . .. a.- .. this 0yeratin5 LLcensa. %s .. . .A.,. .i..Kc; .
&: . a..: ..':.=~.~" .:.'=..:
* --. .specifications
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Y - 'EN ~? s . .mr u :w w.s"r-r- r .' -~ ~.,~ - .r . w i- , r . 3.. ..
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2.
. r.N. . r.,. .'.! ; ,; . -:.,Review : _ . d of . ' reports covering ' evaluation and recommendations - "~ to-a~ '~ .nr... . . " " :'. -=
.1 t" % . . ,.;..-l p.. . . .-Prevent . recurrence a'.m e of .all .N. 3. tRC . . viol..ati.ons. --- . . -- . .N. ;.f '.?,b. M e / . g*) P i
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4,. .. . ..u :.a. F.:.. . . . f Vf 'y
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hj t-T :.'Q e.;.).Reviev~of :all. reportable' avents and '; violations E.F *of Techn
' f., Midi'!. ::
. Specifications which' impact nuclear safety,m$2. - ' 7,:: . f 12., C., .
.; . ...:iGMW-f. ". ~ W.*--%.4i.i&&f.T . w lJh. w ; -.
. $.~3Ci.G./~/.f. .c. Review 3a .i;; f unitWe_operaticus .to . detect. po lear tential .nuc. . ._5:~ 2.z., f MG .!
m ---r w =..?::n:n .'n *vy l.~W a cety _q". .,/---f-"
.'..: ::t. ::.~4 - -:----:.w':n ~.- $. d.,. hazards. :.%'L'Q:
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r~.'.]e(:":- J. . L.f. . 2-b g ' ' . s+~
.'f 5 p. . .:.;. Performance e: f special.i reviews,#m. investigations ir analy'se's~
w E - r*hd...GE.d:.$ .:p.and reports thereon asi requested by. ther Planc Manager m Q- .or the .
.. .;;cet...~ .~.a:.-.u:...:... . :x .nC"k *?i;U..~*:
h.;- G.-Q. ---~. +yc;5.D 4. n..: m ,. m.. W 4".'.'. T.n...':.v ~2:4. Nuclea
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5.W.+4.,4'4-)- -: 3: 2eview of each unplanned.onsite effluent release of~":.? C'~,'.*:='.?W.
.. ~ - -
- ...----'W .T3..' c. . .&r '.x= ,'.-l".
~.
. -- -~
~'~ radioactive material to the. environs . including the * ",. -. .--. - - . - .
preparation and forwarding of reports covering evaluation, I
! recon:mendations and disposition of the corrective action to prevent 55 recurrence to. thel f# '
3,",n I
." 3.lh.SiteDirectorandtotheNuclear Safety Review Board.\p. ei (. ) .16,'j c
i-d
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Review proposed ' changes to ithe Radiological Effluent Manual i I Q.
. (REM)
+ .(j[,K'.['Q 7l
, AUTHORITY ! Ij l j .j j ,( r3 i
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- 6.5.1.'7-The PORC shall:. L NP i
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a. Recommend in writing to the Plant' Manager approval or
. disapproval of items .
. considered under 6.5.1.6 (c) above.
- i. : . ,l... . , .. .-- il .n
- b. .I ei ,
g Provide for a determination 'in writing with regard to whether or not g each item considered"under 6.5.1.6(a), and (b) above M constitutes an unreviewed' safety question. .. .
)
. .. ' . L q.. .; i o.. . . . , .A . . , P' lC f. c.
Provide written notification"vithin 24 hours to the site Director and the Nuclear. Safety Review Board of disagreement between the PORC and .the Plant Manager; however, the Plant
, Manager disagreements shall pursuant have responsibility to 6.1.1 above 'forj resolution of such
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- 6.5.2.s Audits of unit activitiesfshall be performed by the
.. appropriate audit group! under the ' cognizance of the NSRB.
Audits of these activities shall'ha conducted on a frequency- .
. commensurate with their safety significance and the results . 'd
- of their performance history. !.The audits will be conducted
-------- at the frequencies '11sted below unless a documented - _._ ;~ .
. assesuent based on past performance (e.g., performance based
- verification techniques', , trends, and .other performance
- indicators) is performed'to justify extension of their frequency. In no instance vill the frequency be extended by }
more g-than 12 months beyon.d'..the. li.sted frequency. 1 l . .. ,
" :/ * . . x...t ,: .. g ; .t .,:. ni ml
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I a. The conformance of, unit operation to ' provisions contained ./..
' ~within the Technical Specifications and applicable license . . .
~ :::f:7.-Nr. 'f ._.a : . . conditions..gae. lea;sn once per -12
. . . m v.s
. , . . . ....:. _.. .. n.s,.n.w .
- 4. months
.,2.._ i . u.1 c-i ,;,.-_2.g y,ig .
6-. .".~.. -
. 1. . . i .
. ,a ,
N.. b. 7'.e performa.nce, . training ?and qualifications of the entire ---
- 2. . a 4 , ...;,.J. facility oper.ations staff.i.at.l.e.ast..one per 12 months. :f.s
.. .. . . . . . . . . ...c .
. .. . _. . - . .. . ._;.._ _ _ ;,f.
j
- - W ey-Me.4,: S- W-W m.':-q.?n W m.we.w. , j.p h@M! . ..w.yL . . g l. ,44 ,,w,TJ.d.u.if.6';~, 1,, .
..i- 1
- l. M.M.'E6~~f. :C- G.The results ;of.' actions taken to. correct ; deficiencies ' c.pg q." 'L,e ' m" [,,,~~',[.',J ,
,. {
. te'i.:.;
, .prt[ q,,g-^ occurring -in unit equipmentdatructuresg systems or method of it.;4 :-glN,,fr,. -
j
,- .1.. :.--..:.. . . .;.., , ; operation that aff ect nuclear safety at leas'c once per 6 . .-- .-- - - ..d
- ~ . 7.:.tp. " - -- T mon ths C 3 "t ~ n-,';.. - ^-'--
.~
..~. 6- , .'.~;.
G.s.: ; .s.,cd k- . s . .dGr, ;2 .W. * ?-".'.r".. r ., r w .g;m ' e*w"*: 4'-m - - - U%' .h.f I
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j rw M.Mlr-id. v--The performance .o. f activities required by the Operational ~lp-$E'.r_W ---- * '.- . .. l
+$,'~-YW .? - ,- j.er IQuality Assurance Program to meet..the! criterih-of Appendix O.5'f.,5$~@h~;/ . '
7: P " 10 CyR 50, 'at 'lessr. once - I. ' h.'.l..d.Q,. 5%.i:$1,', 3.:. .}.1 &yh,i$y.y.h.per 24 months. i 3 -:r W ' W .i!.'S P.m5'
^
- ?. 6.*j f y:. ..JG.:.:r.-Q..!iz.'.e. '.";fThe Site Radiological Emergency Plan and implementing 7 x;;% ' .. m 7 f,---i
. . . . . _Ol'.h proeedures ae lease onee per 12 months. '.O .@". ,ifil"-'_'*-'W-t
.c ,_s.. ^ "'. a -W.- - ,
s . _ w _
- f. The Plant Physical Security Plan, the! Safeguards Contingency '
' l * '"
Plan,' and implementing procedures 'at least once per 12 months. . l i ;. ,1 ;.g : .. ;5. Apren \t. We. h%n %.ga % i s. The Facility Fire Protection Program ^and implementing 1 h l procedures l at least
! itonce .l i f pler.24 1.;lmont's. '
- h. An independent fire protection and loss prevention program
- inspection and audit shall,be performed once per 12 months
,j utilizing either qualified offsi*: licensco perr inel or an
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.7.';outside fire protection1j.v..ftppf firm.; b l . ,h'. ... I i -
- i. ' 'in ins'piction andl audit of;the fire protection and loss i
preventionprogramshallbeperformedjbfanoutsidequalified ; fire consuJeant at intervals no greater than 3 years. l
. I J. The radiological environmental monitoring program and the 1;t-l. 9.h l results thereof at least once per.24 months.
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- k. The program for hand 11/1g and controlling radiological vaste *
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- 1. The radiological effluent monitoringiprogram and the esults j thereof at least once'per 12, months.1 ; j i ; ; . ,-
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a b' PROCTDtJRES & PROC 2ANS ,
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- a. The applicable procedures recommended,in Appendix "A" of .
Regulatory..i Guide 1.33, Revision.2/mTebr o.uary 1978.
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- b. Limitations on the amount of , overtime worked by individuals " ' ~ "~~-~' ,
. performing safety-related . functions in accordance with NRC policy ,
statement on working hours (Generic Letter No. 82-12). j ,
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- c. Refueling operations. . ,3 G ;'$,$ '
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- d. Surveillance and test .activitiesiot.safetp.related equipment. ~
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- f. Site Radiological Emergency Plan implementation. : t.m . ..
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'2..,=,~M PROGRAM implementation-hy;-~y~M .:,. w . M-{ ufw . W.:,,.d.. ,, .y. 2l ;l.i.:.~-j.
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- j. ".. The emergency "o f NUREG-0737 operating' and proce.dures"isquir'ed Supplements 1 to 'NUREG-0737 s.....
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j I 6.8.2 Temporary changes to procedures El 6.8.1'above may be made ,
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- a. The intent of the original procedure is not altered.
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- b. For site controlled procedures,. the' change is approveri by two [
members of the plant' management staff,.at:least one of whom holds a Senior Reactor Operator's License on the unit affected. I
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'c. The.ch.ange is approved'in .accordance withl Specification 6.5.1.9 above within 14 days of implemen,tatio,n.h ,
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- 6.9 REPORTIlIG RE00TREMEffTS l,
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6.9 1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following identified reports , shall be submitted to the.' Director of the Regional Of fice of NRC,
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- . l' ':ENCLO9URE.2'i*
l Descript,lon and Justification L .l i - - t ' uni t - 2 ,+ ., { .. . ldi,A g,'.d:p..1 ' 1 . , , . lP.'- N b <le 0 t33 : REA80N FOR CHANGE .$j. i j
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oS . BFN Technical Specifications (TS) for ' unit,s 'U, 2,' and 3. ere being updated to - incorporate appropriate administret,1ve cont,rols for the BFN Appendix R 8efe , Bhutdown Program., This updat,e villiedd License Condition 2.C.5.(e), add Definition Section 1.0.NN, and revise, Chapter S.O Administrative Cont,rols.
- i
,These changes vil,1 ellow BFN to implement e NRC approved Appendix R 8efe
- 8 hut,down Progres i under the controls of the BFN T8, which is consistent, with
,the philosophy of NRC Generlo Letter 88-10. t'Chenges t,o this program vill be made under the guidelines of IOCFR50.59.'.. This process vill ensure that, the Progree vill be meintained without, e significant, reduction in the overall '
mergin of safety of the plant,, g.11 ; 1
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I DESCRIPTION AND JUSTIFICATION FOR CHANGE' . ; 'f; j
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ADDLICEN8ECONDITION2.C.5.(e)A8 FOLLOW 8:l- ! 2.C.5.( Al- Browns Ferry Nuclear Plant, shall leptement and maintain in effect all provisions of theispproved Arardix R 8efe Shutdown Program as . described in submittel dated XXXXXXXXX and as approved in NRC 8ER det,ed [ YYYYYYYY subject to the followins' Provisions: A 1 .l, t ', p *e .,- Brovns Ferry Nuclear Plant may make changes t,o the approved Appendix R 8efe Bhutdown Program wit,hout prior approval of the Commission only if i i those changes vould not, adversely effect' the ability t,o achieve end I maintain safe shutdown in the event,j of a. fire. '
' ' "I i lJU8TIFICATIONFORCHANGE .
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. I This change vill allow BFN to implement, e NRC e.pproved Appendix R 8efe :
I Shutdown Program. Adding this License Conditica vill allow BFN to revise specific features of the approved Program. .,vit,hout, prior NRC approval, provided that el such che'nges do.not' otherwise involve e change in a License Condit,lon or T8 or result, in entunreviewed safety quest,lon as defined by 10CFR50.59, and b) such changes do not result, in failure t,o complet.ef the Appendix R Befe Shutdown Program as approved by the Commission'. Consistent, with ot,her changes. implemented under 10CFR50.59 BFN vill maintain a current,' record of all such changes, including en
, onelysis of effects of t,he change on the Appendix R 8efe Shutdown i Program. These' changes are report,ed annually to NRC in t,he BFN ennual report,. y f f, l , : , :
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,y i Temporary changes t,o t,he Program may.be necessary in order t.o accomplish maintenance or modificat,lons. These ' changes vill be conduct.ed under
( .j BFN T8 8ection 6.0, "Administ, ret,1ve Controls". I ! , n.'! ;;!A ! 1 [ l.' .
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h,b ."rli' ;,l . . lADDTOTHEDEFINITIONSECTION1.0.'NNA8.FOLi0W8:
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l' !.0.NN- Appendix R Safe Shutdown Program ' ' BFN has developed en Appendix R 8efe Shutdown Progree. This Program is t,o ensure t,het, t,he equipment, required by t.he Appendix R Safe Shutdown [ i Analysis is maint,sined and demonst, rated Junctional as follows: l ! i . , p. ' A l. ;, l
- 1. The funct,lonel requirement,s of t,he Befe Shutdown systems and equipment, are outilned,in Section III of t,he Program. ' '
t '. s . .
- 2. Test.ing and monitoring of the Appendix' R 8efe Shutdown systems ,
and equipment, are defined, in Sect, lori VI of t,he Program. i
- 16. [ .Y J.q t ,
- 3. In t,he event, t,het, the Appendix ;R 8ei'e Shutdown syst. ems or i equipment cannot, perform their intended function, appropriate l compensatory measures will be implemented in accordance with !
8ection V of this Program. 1 [. .! , , ,
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- 4. Changes made t,o the BFN Appendix R defe Shutdown Program vill be processed in accordance with License Condition 2.C.5.(e ).
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,JU3TIFICATION FOR. CHANGE l . l ;, .s I ... * % j
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J'd i ' l The addition of this Sectionlls consistent with the process that was
% opproved by NRC for the A8ME;8ection XI test ing Frogram. This ellows BFN l to perform a more detailed test,ing and'aonitoring verification process t, hen under t,he regulet,lons of the TB. -lIn addit,lon, t,hese changes' piece administrative controls on e:NRC approved Appendix R 8efe Shutdown Program. Furthermore, any changes made to the Program would be done so under.the provisions of 10CFR50.59. This vill ensure that,any changes to t he Program, vill not, significant,1y. degrade .t,he overall safet,y of t,he p l a n t, . Adding items 1 t,hrough 4 specifically ident,1(y and reference 7
appropriate 8ect,lons of t,he Program and how BFN int, ends t,o implement,, , maint, sin, and amend the approved Program. ' l *,
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l ' ADD T8 6.5.1.6.J. TO READ A8 FOLLOW 8: , I Reviev'of the Appendix R 8efe Shut,down Program and implement,ina procedures. ,Bubmit all.PORC approved changes to NBRB. j 'ls li.'!- ,i;, t e . L J l : ,, I-l
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'JU871FICATION FOR CHANGE i 3
,,i Requiring PORC to reviev'(or be responsible for the review) t,he Program and implement,ing procedures ensures t,het, the Progree vill maintain a high '
level of Plant, management, evereness, as well es ensuring changes t.o t,he Program are reviewed by the effected plant, organizet,lons for overall ; safet,y implicat,lons. 8 I ! : 1,L j y,;,' , 'l i '. ..
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r. i-EXISTING T8 6.5.2.8.g CURRENTLY READO:,' f. ' k. :. .;il i.Hi.l.(L 5 ,i. { %' .- The Facility Fire Protection Progren und implementing procedures et least once per 24 months. ' ' ' - fj . ,
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CHANGE EX18 TING T8 6.5.2.8.s TO READ A8 FOLt0W81 The Facility Fire Protection Program and " Appendix R 8efe Bhutdown , Program" and laplementing procedures.;et .least once per 24 months. .
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JUSTIFICATION FOR CHANGE .
,.g.iy!! ~. .. . .. j ' P.p . ',I This is en ada'nistrative change that ens'ures'the N8R8 vill audit the Appendix R 8ere Shutd&n Progren'en a 24 month basis. This provides
, added assurance that the ' subject Progres is! adequately implemented et BFN. I ;- 't
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i lEXISTINGT86.8.1.sREAD8: . [ M[ ril,' , j Fire Protecilon Progree implementellon.(,f J '
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l CHANGE EXISTING 78 6.8.1.g TO READ:t ' p,g' b,M;! ;.L'l i... ,4 < ' M , Fire Protection Program "and Appendix R Bafe Shutdown Program" ' ' implementation. '
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JUSTIFICATION FOR CHANGE .. ' i, ! i Addingtherequirementforthe[AppendixR8efeShutdownProgramto
, Bection 6.8, " Procedures. end Programs", ensures that written procedures -
shell be established, implemented, and main eined for the subject
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- - program. This change is administrative and is required in order to ,
provide consistency with the' proposed.revis one to 8ections 6.5.1.6.j and .! l 6.5.2.8.g. , y, ij,' j l .:,' t ,'g [dl i : t f.e.1:1. i
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t f. . w.jg gig @a B22 '89 T02 28 0 61 g i l[. . '.%f tachment. To
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JOA RecordY ' NMN : ENCLO8URE 3 ' .iM
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\- i BROWNS FERRY NUCLEAR PLANT - .
l SIGNIFICANT HAZARDS CONSIDERATION ~ **? i ,' g 't s,7 '. y j. UNIT 2 c: y * . ... f * , Ui ' i h,;%l / 7.g. }l, '
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DESCRIPTION OF AMENDMENT REQUEST' '. i ' .
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- M'i k.r,,4'g '.I-add Definit. ion The proposed amendment, would add License'Conditf oril 2.C.S. (e ),8:ct,lon 1.0.NN, *
- and!
S.5.2.8, and 6.8.1 for BFN unit, 2. ' These changes are requested in order t.o '*
, implement, administret.ive controls for. the BFN Appendix R Safe Shutdown ^
~~ Program. ~l Implement,at, ion of these changes -la consistent vit.h NRC guidance in - -
0,:neric Lot,t,or 86-10. ,
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BA,818 FOR.PROPOBED,NO , . .. SIGNIFICANT HAZA,.RD8 CONSIDERATION DETERMINATION n..
.,.. g 7..g.:y NRC has provided stenderds for determining whether a significant, hazards '
consideret,lon exist,s as et,et,ed in 10CFR50.82(c).;;;This proposed amendment, t.o " ' ~"' 7" t.he BFN operet,ing J icense involves no'significenb :hezerde consideret,lons if
~ I operet,lon of t,he facility is conducted'in accordance with the propoced .~a, wendment. It, would not, (1) involve e significant . increase in t,he probability '
f consequences of an accident, from en 'eccident previously evoluste, or (2) f n {previously evaluet ed, or (3)reste involve the e .possibility significant. of a new or reduction in adifferent, mersin ofkind of accident free *
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' (1) The proposed amendment, does.not involve e. significant, increase in the probabilit,y or consequences of en accident, previously evaluated.
Implemented,lon of t hese technical specificat,lons (TS) provide ! consistency between t,he operet ing license, t,he Appendix R 8are Shutdown Program, and.t.he BFN Appendix R Safe Shutdown Analysis.
' These proposed changes are administrative in nature and do not alter the intended funct,lon or design basis of any sefat,y-related i equipment, es addressed in t,he BFN Final Safet,y Analysis Report,
. (F8AR). Implement,et,lon of t,hese changes Jensures t.het, t,he Appendix R i
/ Safe Shddown Program is maintalned in a manner consist,ent, vit.h NRC
- gulden'ce. By maint,sining this Program in accerdance with the i
' ' ' ~~~
proposed T8, BFN vill ensure t,he equipment, needed t,o perform safe shut.down in accordance with t,he.BFN Appendix R Safe Shutdown . Analysis is properly maint.eined ,and is functional to meet, t he ' requirements of 10CFR50 Appendix R. s ',
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, l (2) The proposed amendeont, does not, creet,e t,he possibilit,y of a n.v er *
' different, kind of accident, . from en accident- previously evaluet,ed.
This is an edelnist,rst.ive change which pieces controls on a previously approved NBC Progree. These changes will not, eliminet,e,, or modify any protect.lve funct,lons or equipment that, are required
' for BFN to achieve end maintain reactor shutdown cepebilit.les as -- --
required,by 10CFR50 Appendix l changes will not, result, in en,2.' ! !aplement,et. addit,lonell releaseion of these pathvey proposed to t,he .
! environment,. In developing the liFN Appendix R Sefe Shutdown Program en snelysis ves performed in order. to ident,ify t,het, equipment needed .
t,o comply wit,h 10CFR50 Appendix R. I Implement,ing t,his Progree and
~
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; the subject T8 does not : negate the other safety enelyses or accident ;
'I scenerlos in which BFN ves licensedk for. .
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. . . . _. ( 3 ) . The proposed change does nvolve
.. . :i . :.N ib..Y not,' i.Dh'kf'Ul0 .
e significant,
. . . l reduct,lon in the
,{
.. mergin of safety. This proposed:ssendment is adelnistrative and . i i pieces various adelnist.retive controls. on a NRC approved BFN l Appendix R Safe Shutdown Progree. Implementing these changes .
; enhances the overall sefety. of. BFN.4 The Pregram ident,ifles that
. . . - . equipment, required by t,he. Appendix R Safe Shutdevn Analysis to *
" ' , ' shut.down the reactor in accordance 'vit,h 10CFR50 Appendix R. In
- eddition, it, also provides, t,est,ing and monit,oring requirement s along {
' h. f.,..,.
, ^9':"I wit.h compenset.ory measures'if.the subject equipment, cannot, fulfill
' . ! l
. ~ f: . . / " ' i t,a funct,lon . .The proposed T8 ellow:BFN to change the Progree under
~ ~ "
." i
'p . e.' W.n,r..T.Ethe guidelines 'of '10CFR50.59(which is 'en approved industry practice
..'l~' 7~.~end consi stent, vit.h NRC Generlo, Let,t.cr '86-10. Any changes made #
~
! .~- -
J' ~ * , under *
, report. 10CFR50.59 This process ere will/ requiredito;be enable NRC submitted to' independent,1y to NRC keep in en abreast, ennual -" I
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uniuuvsuno s 3.6/4.6 PRIMARY SYSTEM BOUNDA9Y l LIMITING C0trDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS MOV
* ' 4 "o
'0 i200 3.6.C Coolant Leskane 4.6.C Co o lant- Leakane
- 2. Both the sump and air sampling 2. With the air sampling systems shall be OPERABLE during -
system INOPERABLE, grab REACTOR POWER OPERATION. From samples shall ha obtained and after the date that one of and analyzed at least once these systems is made or found every 24 hours. to be INOPERABLE for any reason, REACTOR POWER OPERATION is permissible only during the succeeding 24 hours for the sump system or 72 hours for the air sampling system. The air sampling system may be removed from service for a period of 4 hours for calibration, function testing, and maintenance without providing a temporary monitor.
- 3. If the condition in 1 or 2 above cannot be met, an orderly shutdown shall be initiated and the reactor shall be placed in the COLD SHUTDOWN CONDITION within 24 hours.
3.6.D. Relief Valves 4.6.D. Relief Valves
- 1. When more than one relief valves 1. Approximately one-half of are known to be failed, an all relief valves shall orderly shutdown shall be be bench-checked or initiated and the reactor -
replaced with a depressurized to less than 105 bench-checked valve psig within 24 hours. each operating cycle. All 13 valves will have been checked or replaced upon the completion of every second cycle. -
- 2. In accordance with .
Specification 1.0.MM, i each relief valve shall be manually opened until thermocouple and acoustic monitors downstream of the valve indicate steam is flowing from the valve. BFN - 3.6/4.6-10 AMENDMENT NO.15 5 Unit 2 " [ 9. I.iiMid6.i N.
.7
. ]
l DR __A__FT _ 4 i l 1 BROWNS FERRY NUCLEAR PLANT UNIT 2 APPENDIX R SAFE SHUTDOWN PROGRAM l 4 NOTE: t This document has been issued in draft form only. It is currently undergoing s l
-review by TVA personnel and its technical and administrative content is subject to change.
4 0726R Page 1
DRAFI s e- ,..
, ]
INDEX I i 1 SECTION I. PROGRAM DESCRIPTION ; SECTION-II - REFERENCES SECTION III REQUIRED SAFE SHUTDOWN EQUIPMENT - TABLE SECTION IV DRAWINGS SECTION V COMPENSATORY MEASURES - TABLE l
'.SECTION VI - TESTING AND MONITORING - TABLE SECTION VII SAFE SHUTDOWN INSTRUCTIONS INTERFACE SECTION VIII. TECHNICAL BASIS f
SECTION IX REVISION / CONTROL i 4-1 l l o l l
. DRAFT BFN APPENDIX R SAFE SHUTDOWN PROGRAM -
SECTION I PROGRAM AND DESCRIPTION 10 CFR 50 APPENDIX R app i 1' es to licensed nuclear reactors with respect to certain generic issues concerning fire protection features required to satisfy Criterion 3 to 10 CFR 50 Appendix A. When considering the effects of fire, those systems associated with achieving safe shutdown conditions assume major importance because damage to them can lead to core damage resulting from loss of coolant through boiloff. For this reason, a plant Appendix R evaluation (Reference 6) was performed for the Browns Ferry Nuclear Plant (BFN), to ensure that safe shutdown capability can be maintained during and after a fire in compliance with Section III.G, III.J. and III.L of Appendix R. The initial Appendix R evaluation identified the minimum systems required for postfire safe shutdown and the modifications that were necessary to ensure the operability of the minimum systems. The plant Appendix R evaluation was performed assuming concurrent operation of the three BFN units and did not factor in the unavailability of equipment based on the possible outage of a unit. A supplemental evaluation was performed for only Unit 2 operating. Based on this evaluation, Safe Shutdown Instructions (SSIs) were developed to provide the operators with the necessary actions to shutdown the reactor in the event of an Appendix R fire. To ensure the equipment necessary to support these actions is available and functional, BFN developed this Appendix R Safe Shutdown Program. This Program identifies those systems and components required to meet the intent of 10 CFR 50 Appendix R for safe shutdown of Unit 2. This Program also provides functional requirements, testing and monitoring requirements along with compensatory measures in the event the Appendix R Safe Shutdown equipment is not available to perform its function. Unit 2 safe shutdown capability relies upon portions of the Unit 1 and 3 auxiliary power systems, including the Unit 3 diesel generators, as well as certain Unit 2 specific and . common auxiliary power distribution boards. ?!uch of the equipment (eg., the backup control panel instruments for reactor water level, reactor vessel pressure, torus water level, and torus water temperature) or the mode of system component operation relied upon in the Appendix R evaluation is not included in the Unit 2 technical specifications. - l The basis for which systems are identified in Section III of this program are - j provided in Reference 6. The specific components, functional requirements (mode of operation) and fire areas / zones where the component is relied upon are provided in the calculations (References 1 through 5). f The BFN Appendix R Safe Shutdown Program will be maintained and amended in ) accordance with the BEN License Condition 2.C.5.(a) and appropriate technical , specifications in Section 6 " Administrative Controls". l l l l l
\
l 0726R Page 3 l
DRAFT SECTION II REFERENCES This section provides the references used in developing the BFN Appendix R Safe Shutdown Program.
- 1) ND-Q2999-880160 - Required Mechanical Equipment for a Unit 2 Safe Shutdown
- 2) ED-Q0000-88489 - Tabulation of Equipment Power Supplies and Detailing Criteria for the Auxiliary Power System
- 3) ND-Q2999-880161 - Appendix R Analysis -Unit 2 Manual Action Requirements
- 4) ND-Q2999-88008 - Appendix R Analysis-Unit 2 Manual Action Requirements for Areas Relaying on HPCI
- 5) ED-N2999-880681 - Appendix R-Separation Analysis
- 6) Engineering Evaluations in Support of 10 CFR 50 Appendix R Submittal, Vol. I, II, III 0726R Page 4
- : - 1
.- D
__ RAFT SECTION III - REQUIRED SAFE SHUTDOWN EQUIPMENT - is section provides a Table identifying the equipment required for BFN Appendix R Safe Shutdown. This Table lists specific components, the system containing the component, the function (s) the component provides, and identifies the fire areas / zones for which the equipment is required to safely shutdown Unit 2. REQUIRED SAFE SHUTDOWN EQUIPMENT l APPENDIX R EQUIPMENT DESCRIPTION UNIT FUNCTION AREAS /20NES AFFECTED MAIN STEAM SYSTEM PCV-1-4 MAIN STEAM LINE A 2 Operable from the 1, 2-1, 2-2, 3, 8, l RELIEF VALVE main control room 11, 12, 13, 14, 15, 17, 19, 21, . 22, 23, 24, 25 ! De-energized from 16 the backup control panel PCV-1-5 MAIN STEAM LINE A 2 Operable from the 2-2, 18 RELIEF VALVE-ADS main control room Operable from the 16 backup control panel PCV-1-18 MAIN STEAM LINE B 2 Operable from the 2-1, 2-2, 2-5, 2-6, RELIEF VALVE main control room 4, 5, 6, 7, 10, 20 De-energized from 16 - the backup control panel , 0726R Page 5 _ _ _
- Uaas1 REQUIRED. SAFE SHUTDOWN EQUIPMENT
. l APPENDIX R EQUIPMENT DESCRIPTION UNIT FUNCTION ' AREAS /20NES AFFECTED PCV-1-19 MAIN STEAM LINE.B 2 Operable from the 2-1, 2-2, 2-3, 2-4, 2-5, RELIEF VALVE-ADS main control room 2-6, 4, 5, 6, 7, 9, !
10, 18, 20 De-energized from 16 RMOV Board. PCV-1-22 MAIN STEAM LINE B 2 Operabic from the' 1, 2-1, 2-2, 3, 8, 11, RELIEF VALVE-ADS main control room 12, 13, 14, 15, 17, 19, 21, 22, 23, 24, 25 Operable from the 16 backup control panel PCV-1-23 MAIN STEAM LINE B 2 Operable from the 2-2 RELIEF VALVE main control room De-energized from 16 the backup control. panel PCV-1-30 MAIN STEAM LINE C 2 Operable from the 1, 2-1, 2-2, 3, 8,'ll, RELIEF VALVE-ADS main control room 12, 13,.14, 15, 17, 19, 20, 21, 22, 23, 24 25 , I' Operable from the 16 backup control panel PCV-1-31 MAIN STEAM LINE C 2 Operable from the 2-1, 2-2, 2-3, 2-4, 2-5, RELIEF VALVE-ADS main control room 2-6, 4, 5, 6, 7 - 9, 10, 18, 20 Deenergized from 16 , RMOV Board PCV-1-34 MAIN STEAM LINE C 2 Operable from the 2-2 RELIEF VALVE-ADS main control room
, Operable from the 16 backup control panel l
l 0726R Page 10
L - D s _RA__FI ( ,' REQUIRED SAFE' SHUTDOWN EQUIPMENT APPENDIX R
.cQUIPMENT DESCRIPTION UNIT FUNCTION AREAS / ZONES AFFECTED l
PCV-1-41 MAIN STEAM LINE D 2 Operable from the 1, 2-1, 2-2, 3, 8, 11. RELIEF VALVE main control room 12, 13, 14, 15, 17 l 19, 21, 22, 23, 24, 25 De-energized.from 16 the backup control panel i L PCV-1-42 Mi.IN STEAM LINE D 2 Operable from the 2-2 RELIEF VALVE main control room De-energized from 16 the backup control panel 1
- PCV-1-179 MAIN STEAM LINE A 2 Operable from the 2-l', 2-2, 2-3, 2-4, 2-5 i RELIEF VALVE main control room 2-6, 4, 5, 6, 7,
)
9, 10, 18, 20 l De-energized from 16 RMOV Board PCV-1-180 MAIN STEAM LINE D 2 Operable from the 2-2 RELIEF-VALVE main control room De-energized from 16 the backup control panel e 9 t
UKdf1
~
REQUIRED SAFE SHUIDOWN EQUIPMENT n . - .
. APPENDIX R lIPMENT DESCRIPTION UNIT- FUNCTION AREAS / ZONES AFFECTED e
RHR SERVICE WATER SYSTEM FCV-23-40 RHR HEAT EXCHANGER 1 Manually close 2-2, 2-5, 8, 11, 16, C SW OUTLET VLV with handwheel '18, 25
'FCV-23-40 RHR HEAT EXCHANGER 2 Manually open , 2-2, 2-5, 8, 11, C SW OUTLET VLV with handwheel 18, 25 Open with back- 16 up control switch l FCV-23-40 RHR HEAT EXCHANGER 3 Manually close 2-2, 2-5, 8, 11, 16 '
C SW 0UTLET VLV with handwheel 18, 25 l FCV-23-52 RHR HEAT EXCHANGER 1 Manually open 1, 2-1, 2-3, 2-4, 2-6 D COOL WATER OUTLET with handwheel 3, 4, 5, 6, 7, 9, VLV 10, 12, 13, 14, 15, 17, 19, 20, 21, 22 23, 24, 25 "r-23-52 RHR HEAT EXCHANGER 2 Manually open 1, 2-1, 2-3, 2-4, 2-6 D_ COOL WATER OUTLET with handwheel 3,4,5,6,7,9, ;
-VLV 10,.12, 13, 14, 15, i 17, 19, 20, 21, 22
.23, 24, 25 FCV-23-52 RHR HEAT EXCHANGER 3 Manually close 1, 2-1, 2-3,.2-4, 2-6 D COOL WATER CUTLET with handwheel 3,4,5,6, 7, 9, VLV' 10, 12, 13, 14, 15, 17, 19, 20, 21, 22 23, 24,~25 RHRSW PUMP A3 0 . Auto start 2-1, 2-2, 2-4, 2-5, 2-6 4, 5, 6.- 7, 8, 9 -
10, 11, 20 Manual breaker 1, 2-3, 25 operation 4 4 i
'0726R Page 12
- : eseLL s
REQUIRED SAFE SHUTDOWN EQUIPMENT APPENDIX R
-QUIPMENT DESCRIPTION WNIT FUNCTION AREAS /2ONES AFFECTED RFRSW PUMP B3 0 Auto start 3, 12, 13, 14, 15, 16, 17, 18, 19, 21, 22, 23, 24 Manual breaker 25 operation Start from backup 16 control switch RHRSW PUMP C1 0 start from backup 2-2, 2-5, 8, 11, 16, control switch 18, 25 RERSW PUMP C2 0 start from backup 2-2, 2-5, 8, 11, 16, control switch 18, 25 RERSW PUMP C3 0 Auto start 1, 2-1, 2-2, 2-3, 2-4, 2-5, 2-6, 4, 5, 6, 7, 8, 9, 10, 11, 20 Manual breaker 25 operation RHRSW PUMP D2 0 Start from backup 1, 2-1, 2-3, 2-4, 2-6, control switch 3, 4, 5, 6, 7, 9, 10, 12, 13, 14, 15, 17, 19, 20, 21, 22, 23, 24, 25 RERSW PUMP D3 0 Auto start 3, 12, 13, 14, 15, 16, 17, 18, 19, 21, 22, 23, 24 Manual breaker 25 operation ,
~
Start from backup 16 control switch A 0726R Page 13
DRAFI SECTION IV - D.RAWINGS - \ This section provides the plant Appendix R compartmentation drawings (47W216 series) identifying specific plant fire areas / zones. Tiese drawings are available in the program to ensure rapid identification of the physical plant areas affected by the loss of an Appendix R Safe Shutdown piece of equipment. This will minimize personnel error, ensure ccv.8istency in communications and help provide a rapid and correct response to an equipment loss. 9 e [ 0727R - - - - - - - - -Jage65
. DRAFT FIRE AREA DESIGNATIONS FOR BFN Fire Area No. Designation Location i UNIT 1 REACTOR BUILDING EL519 THRU 639 2-1 LWIT 2, REACTOR BUILDING EL519 THRU 639 West of Column Line R11 2-2 UNIT 2, REACTOR BUILDING EL519 THRU 565 East of Column Line R11 2-3 UNIT 2, REACTOR BUILDING EL593 North of Column Line r 2-4 UNIT 2, REACTOR BUILDING EL593 South of Column Line q and RHR Heat E:cchanger Reoms from El 565 and 593 2-5 UNIT 2, REACTOR BUILDING EL621 and El 639 North of Column Line r 2-6 UNIT 2, REACTOR BUILDING EL639 South of Column Line r 3 UNIT 3, REACTOR BUILDING EL519 THRU 639 4 UNIT 1, 4kV SHUTDOWN BOARD RM-B EL593 RB 5 UNIT 1, 4kV SHUTDOWN BOARD RM-A EL621 RB AND 250V BATTERY ROOM 6 UNIT 1, 480V SHUTDOWN BOARD RM-1A EL621 RB 7 UNIT 1, 480V SHUTDOWN BOARD RM-1B EL621 RB 8 UNIT 2, 4kV SHUTDOWN BOARD RM-D EL593 RB ,
9 UNIT 2, 4kV SHUTDOWN BOARD RM-C EL621 RB AND 250V BATTERY ROOM 10 UNIT 2, 480V SHUTDOWN BOARD RM-2A EL621 RB 11 UNIT 2, 480V SHUTDOWN BOARD RM-2B EL621 RB 12 UNIT 3, SHUTDOWN BOARD ROOM F EL593 RB 0727R Page 66
1 ,) .l
- DRAFT
(
- FIREAREADESIGNATIONS-FORBFN Fire Area No. Designation ' Location 13 UNIT 3 SHUTDOWN BOARD ROOM E EL521 RB 14 UNIT 3, 480V SHUTDOWN BOARD RM-3A EL621 RB 15 UNIT 3, 480V SHUTDOWN. BOARD RM-3B EL621 RB 16 CONTROL BUILDING 'FL593, 606 & 617 CB 17 UNIT 1, BATTERY AND BATTERY BD RM EL593 CB 18 UNIT 2, BATTERY AND BATTERY BD RM EL593 CB 19 UNIT 3, BATTERY AND BATTERY BD RM EL593 CB 20 UNIT 1&2 DIESEL GENERATOR BUILDING ALL ELEVATIONS DGB 21 UNIT 3 DIESEL GENERATOR BUILDING ALL ELEVATIONS DGB 22 UNIT 3, 4kV SHUTDOWN BD RM'S 3EA&3EB EL565 & 583, DGB 23 UNIT 3, 4kV SHUTDOWN BD RM'S 3EC&3ED EL565 s 583, DGB 24 UNIT 3, 4kV BUS TIE BOARD ROOM EL565 DGB 25 TURBINE BUILDING, CABLE TUNNEL, INTAKE ALL ELEVATIONS PUMPING STATION, AND RADWASTE BUILDING o
4 0727R Pare 67
D RAFT
.SECTION V - COMPENSATORY MEASURES l
This section provides a table which identifies the compensatory measures to be l taken in the event equipment required for appendix R Safe Shutdown is not available. For ease of use the format 'used is Section III for Equipment, Unit and Function is repeated here. Each equipment function has a letter designator next to it, this letter represents compensatory measure as is shown below. If the listed piece of equipment is not able to perform one of its functions the compensatory measure next to the unavailable function must be implemented. If the listed piece of equipment is not able to perform any of its functions, the compensatory measures for all functions of that equipment must be implemented. The listed compensatory measures may be removed when and if the required function is restored to service. In addition to restoring the function, the compensatory measures may be removed if an engineering analysis is performed, this program is changed and the Safe Shutdown Instructions are changed. A. Action must be taken ir. =ccordance with the referenced Technical Specification. B. Restore the equipment function in 7 days or provide equivalent shutdown capability by one of the following methods.
- 1) A temporary alteration in accordance with plant procedures that allows the equir:nent to perform its intended function.
- 2) A firewatch in the affected areas / zones as specified in Section III.
If the affected areas / zones are from the following set, the firewatch. must be continuous, otherwise it may be hourly (set later).
- 3) A temporary change to the SSI's which provide safe shutdown without the required function.
If equivalent shutdown capability is used, restore the equipment function in 60 days or provide an engineering evaluation and a change to this program that provides an alternate method to perform the Appendix R functivn, otherwise provide PORC review and Plant Manager approval of the equivalent shutdown capability to ensure its adequacy. This review shall - be conducted every 60 days until an alternate method is in place. 0728R Page 80
I r i D
' _ _R A_ _F _T COMPENSATORY MEASURES 9
APPENDIX R
. EQUIPMENT UNIT FUNCTION COMPENSATORY MEASURES PCV-1-4 2: Operable from the 'A(3.6.D) main control room De-energized from B the backup control
. panel PCV-1-5 2 ' Operable'from the A(3.6.D) main control room Operable from B the backup control NOTE: If the backup panel control function of valves PCV-1-22, 30, and 34 is available, no compensatory measure is required.
'PCV-1-18 2 Operable from the. A(3.6.D) cain control room De-energized from B the backup control panel PCV-1-19 2 Operable from the B main control room NOTE: 'Affected areas /
zones are 2-3, 2-4 and 9 only. De-energized from B RMOV Board - 0728R' Page 83
- 7- D
, _R__AI_I .c .
COMPENSATORY MEASURES
~.
APPENDIX R EQUIPMENT UNIT FUNCTION COMPENSATORY MEASURES PCV-1-22 2 Operable from the A(3.6.D) main control room Operable from back- B up control panel NOTE: If the backup control function-of valves PCV-1-5, 30, and 34 is available, no compensatory-measure is required. PCV-1-23 2 Operable from the A(3.6.D) main control room De-energized from back- B up control panel PCV-1-30 2 Operable from the A(3.6.D) nmin control room Operable from back- B up control panel PCV-t-31 2 Operable from B main control room NOTE: Affected areas / zones are 2-3, 2-4 and 9 only. De-energized from B RMOV Board PCV-1-34 2 Operable from the A(3.6.D) main control room ~ Operable from back- B - up control panel NOTE: If the backup _ control function of valves ) PCV-1-5, 22, and l 30 is available, ; no compensatory 1
, measure is required.
0728R Page 84
a DR __A__FT_ c COMPCaSATORY MEASURES APPENDIX R
. EQUIPMENT UNIT FUNCTION COMPENSATORY MEASURES PCV-1-41 2 Operable from the A(3.6.D) main control room De-energized from B the backup control panel PCV-1 2 Operable from the A(3.6.D) main control room
( De-energized from B the backup control panel PCV-1-179 2 Operable from B main control room ' NOTE: Affected areas / zones are 2-3, 2-4 and 9 only. De-energized from B RMOV Board PCV-1-180 2 Operable from the A(3.6.D) main control room De-energized from B the backup control panel 1 [. l l l 0728R Page 85 L
.. .c
' ~~
.D RAFT COMPENSATORY MEASURES
. APPENDIX R
, EQUIPMENT UNIT FUNCTION COMPENSATORY MEASURES RHR SERVICE WATER SYSTEM FCV-23-40 1 Manually close with handwheel '(Later)
FCV-23-40 2 Manually open with handwheel B Open with back- B up control switch FCV-23-40 3 Manually close with handwheel (Later) FCV-23-52 1 Manually close with handwheel B FCV-23-52 2 Manually open with B handwheel FCV-23-52 3 Manually close with (Later) handwheel RHRSW plt!P A3 0 Auto start B Manual breaker B operation. RHRSW PL?!P B3 0 Auto start B I Manual creaker B cperation Start from backup B control switch RHRSW Pl?!P C1 0 Start from backup B control switch NOTE: If the manual start from backup contr61 function is available on RHRSW C2, no f compensatory measure- _ is required. l 0728R Page 86 i C _ _______________________ _
- /
DRAFI s i COMPENSATORY MEASURES APPENDIX R EQUIPMENT UNIT FUNCTION COMPENSATORY MEASURES RHRSW PUMP C2 0 Start from backup B control switch NOTE: If the raanual start fr hackup control 1 an is available on RHRSW C1, no compensatory measure is required. RHRSW PUMP C3 0 Auto start B Manual breaker B operation RHRSW PUMP D2 0 Start from backup B control switch RHRSW PUMP D3 0 Auto start B Manual breaker B operation Start from backup B control switch i 4 0728R Page 87
e s DA __A_FT g ., _ _ SECTION VI - TESTING AND MONITORING This section provides a table which identifies the testing requirements, frequency and implementing procedures for Appendia R Safe Shutdown equipment. For ease of use the format for Equipment Unit ano Function is repeated here. Each testing requirement must be satisfied within the designated frequency in order for the equipment to be considered capable of performing its required safe shutdown function. Testing need not be performed when the reactor is in cold shutdown, however, reactor startup may not proceed unless all required testing has been satisfied within the required frequency. TESTING AND MONITORING TESTING EQUIPMENT UNIT FUNCTION REQUIREMENTS FREQ PROCEDURES PCV-1-4 2 Operable from Verify that the OC 2-SI-4.6.D.2 main control valve can be room opened from the main control room De-energized from Verify that the OC TI-73B the backup valve can be de-control panel energized from the backup control : panel ! PCV-1-5 2 Operable from Verify that valve OC 2-SI-4.6.D.2 l main control room can be opened , from the main ! control room Operable from the Verify that valve OC TI-73B backup control solenoid energizes panel when operated from the backup control panel ~ PCV-1-18 2 Operable from Verify that OC 2-SI-4.6.D.2 main control valve can be _ room opened from the main control room De-energized from Verify that the OC TI-73B the backup valve can be de-
, control panel energized from the backup control panel i
0729R Page 125
l ua_ a _I 1 1 l I l TESTING AND MONITORING TESTING UiPMENT UNIT FUNCTION REQUIREMENTS FREQ PROCEDURES 1 PCV-1-19 2 Operable from Verify that OC 2-SI-4.6.D.2 main control valve can be I room opened from the i main control room f
]
De-energized from No testing N/A N/A i RMOV Board required ) l PCV-1-22 2 Operable from Verify that OC 2-SI-4.6.D.2 main control valve can be j room opened from the ) main control room ! Operable from Verify that the OC TI-73B the backup valve solenoid control panel energizes when operated from the backup control panel PCV-1-23 2 Operable from Verify that the OC 2-SI-4.6.D.2 main control valve can be room opened from the main control room De-energized from Verify that the OC II-73B the backup valve can be de-control panel energized from the backup control panel ! PCV-1-30 2 Operable from Verify that valve LOC 2-SI-4.6.D.2 main control room can be opened from the main control room ' Operable from the Verify that valve OC TI-73B - backup control solenoid energizes . panel when operated from the backup control panel l PCV-1-31 2 Operable from Verify that the OC 2-SI-4.6.D.2 main control valve can be room opened from the main control room De-energized from No testing N/A N/A RMOV Board required i 0729R ( Page 126 '
...FT DRA -
.s TESTING AND MONITORING TESTING UIPMENT UNIT FUNCTION REQUIREMENTS FREQ PROCEDURES i
PCV-1-34 2 Operable from Verify that OC 2-SI-4.6.D-2 main control valve can ba room opened from the main control room Operable from Verify that the 00 TI-73B the backup valve solenoid control panel energizes when operated from the backup control panel PCV-1-41 2 Operable from Verify that the OC 2-SI-4.6.D.2 main control valve can be room opened from the main control room De-energized from Verify that the OC TI-73B the backup valve can be de-control panel energized from the backup control panel
<-1-42 2 Operable from Verify that the OC 2-SI-4.6.D.2 main control valve can be room opened from the main control room De-energized from Verify that the OC TI-73B the backup valve can be de-control panel energized from the '
backup control panel ! PCV-1-179 2 Operable from Verify that the OC 2-SI-4.6.D.2 - l main control valve can be i i rocm opened from the - main control room _ De-energized from No testing N/A N/A RMOV Board required I l I j i l l l l l i 0729R- Page 127 l l
q f / DR l __A__FT _
,- TESTING AND MONITORING TESTING 7UIPMENT. UNIT FUNCTION REQUIREMENTS -FREQ PROCEDURES b
l PCV-1-180 2 Operable from Verify that the. I main control 00 2-SI-4.6.D.2 valve can be room opened from the main control room
-De-energized from Verify that the OC .TI-73B the backup valve can'be de-i.*ontrol panel energized from the j
backup control ' panel RHR SERVICE WATER SYSTEM FCV-23-40 1 Manually close (later) with handwheel (later) FCV-23-40 2 Manually open with handwheel Verify that valve 18M (later) can be manually opened using local handwheel and clutch
~^V-23-40 3 Manually close (later) with handwheel (later)
FCV-23-52 1 Manua'11y close with handwheel Verify that valve 18M (later) can be manually opened using local handwheel and clutch FCU-23-52 2 Manually open with handwheel Verify that valve 18M (later) can be manually , opened using local handwheel and clutch FCV-23-52 3 Manually close (later) ~ with handwheel (later) ) i RRRSW PUMP A3 0 Auto start Verify that the pump 18M SI-4.9.A.3.a - automatically starts on a diesel generator running signal l Manual breaker Verify breaker can be (later)
, operation closed with manual pushbutton without control power l 0729R Page 128 (i _
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UIPMENT. ' UNIT FUNCTIOM;__ REQUIREMENTS FREQ PROCEDURES RHRSW PUMP B3 0 Auto start. Verify that the pump 18M SI-4.9.A.3.a 4
automatically starts on a diesel generator running signal Manual breaker Verify breaker can be (later) operation closed with manual pushbutton without control power 6 tart from backup Verify that breaker 18M (later) control switch can be closed using backup control switch RERSW PUMP C1 0 Start from backup Verify that breaker (later) l control cwitch can be closed using backup control switch I RHRSW PUMP C2 0 Start from backup Verify that breaker (later) ; control switch can be closed using l backup control switch 5 i "7SW PUMP C3 0 Auto start Verify that the pump 18M SI-4.9.A.3.a j automatically starts on a diesel generator running signal Manual breaker Verify breaker can be (later) operation closed with manual pushbutton without ; control power RHRSW PUMP D2 0 Start from backup Verify that breaker (later) control switch can be closed using backup control switch RHRSW PUMP D3 0 Auto start Verify that 18M SI-4.9.A.3.a the pump , automatically ' starts on a diesel generator running signal I Manual breaker Verify breaker (later)
, operation can he closed with munual pushbutton eithout control power 4
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~. SECTION VII - SAFE SHUTDOWN INSTRUCTIONS INTERFACE This.section is to ensure that close coordination is maintained between tne ,
3FN safe shutdown instructions and the Appendix R Safe Shutdown Program. Any. I changes to either of these documents should be reviewed to ensure consistency. If any changes'are needed to the BFN Appendix R Safe Shutdown , i Program, they will be don.e so in compliance with the BFN License Condition 2.C.5.(a) and technical specification section 6.0, " Administrative Control". Changes made to this Program which result in the addition or deletion of Appendix R equipment for a specific fire area / zone shall initiate changes to the corresponding BFN Appendix R Safe Shutdown Instructions (2-SSI-001 and/or 2-SSI-002). If the changes are needed immediately a temporary change to the appropriate safe shutdown instruction (s) can be used. This shall also require that the appropriate Nuclear Engineering Calculations (references 2.through 5) be evaluated and revised promptly. If these changes are not needed immediately, then the appropriate NE calculations shall be revised and issued prior to revising the safe shutdown instructions. 6 e J-0729R Page 187
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DRAFT SECTION VIII TECHNICAL BASIS > This section describes the bases for the compensatory measures listed in Table V and the monitoring and testing requirements for determining whether or not the componenta listed in Table 'III can perform their ' Appendix R Safe Shutdown function. t Each system is addressed separately. .However, certain bases for compensatory ) measures are generic to m.ost all the s'ystems. These are discussed below and will not be repeated for each system. Compensatory Measure A i Several components and systems required in this program are addressed by an existing Unit 2 Technical Specification (TS). 'If the TS operability requirements are equivalent or more stringent than the Appendix R requirements, then the limiting conditions for operation provide adequate compensatory measures The bases for this compensatory measure can be found in the TS and will not be repea ui' here. Compensatory Mearure B l For that equipment where Technical Specification LCO's do not provide adequate measures to meet the Appendix R analysis a separate. compensatory m ssure has been established. The specified measures are intended to assure safe s..utlown capability i is restored within 7 days by either restoring the failed equipment function or by taking temporary measures to assure equivalent shutdown capability exists. ' Equivalent shutdown capability is~ defined as 1) providing temporary equipment or. procedures which will ensure the out of service equipment function does not affect safe shutdown capability or 2) providing adequate fire watch capability to ensure fires are prevented and/or discovered in a time frame which will assure the out of service equipment is not needed to support reactor safe shutdown in case of fire. Dependence on temporary measures is acceptable for only a limited period of time. In order to prevent indefinite use of temporary measures a PORC review and Plant Manager approval of the adequacy of the temporary measures is required every 60 days until an alternative shutdown path is developed. i
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.,; SYSTEM'1 - MAIN STEAM
- a. MSRVs A minimum of.three MSRVs are required to be operable to support the Appendix R Safe Shutdown Analysis, Ref 6. In fire area 9.and fire zones.2-3 and 2-4, a postulated' fire could potentially disable all but three MSRVs. Since Technical Specifications allow'one relief valve to be out of service indefinitely and three specific valves are needed for fire-area 9 and fire zones 2-3 and 2-4, a separate compensatory measure must be implemented if one
.of these three valves is out of service. For a fire in any other fire areas / zones of the plant, at least four MSRVs would be available. Thus, even-if one MSRV is out of service, the required number of three MSRVs would remain available for safe shutdown and the action provided by TS.3.6.D is adequate.
For fire area CB 16, the MSRVs not required must be de-energized to prevent spurious operation. Thus, the MSRVs are tested for (1) MCR operability, (2) i I backup control panel operability, and (3) backup control panel de-energizing. ' Under the current TS 4.6.D.2, the MSRVs.are demonstrated operable once per i operating cycle from MCR by verifying steam flow downstream of the valve. The backup control operability and de-energizing shall be tested by verifying the valve solenoid energizes or is (e-energized. This is adequate since the valve i' operability is tested during t'.e TS surveillance. Also, TS 4.6.D.1 requires j all MSRVs to be bench-checke6 or replaced every 2 cycles. Based on this, Appendix R MSRV functions are adequately tested to provide assurance they will operate properly if' required. I 4 1 i 0729R Page 189 _ _ _ _ . - _ = - - _ _ _ _ _ _ . - _
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SYSTEM 23 - RHRSW ' l The plant Appendix R evaluation requires 'RHRSW pumps D2 and either C1 or C2 be ! operable to supply water to the RHR Heat Exchanger during an appendix R fire. If the-D2 or both C1 and C2 pump (s) are out of service, a compensatory measure
.will be established. "
These RHRSW pumps are regoired to be started from their backup control.
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station. The RHRSW pumps are routinely tested for operability. and flow rate.
- The flow rate of 4500 gpm for each pump, as required by TS 4.5.C.1.b. is adequate to remove decay heat for one unit during and Appendix R fire-(i.e.
support the analysis in Ref 6). The RERSW supply to the Unit 3 C & D RHR heat exchangers has been capped off and the discharge valves on the Unit 1 C & D heat exchangers will remain in their normally closed -position or will be manually closed. Hence, the-entire' flow from RHRSW pump D2 and either C1 or C2 will be delivered to the Unit 2 RHR heat exchangers. Therefore, only the verification that the breaker will close at the backup control station using the backup control switch is necessary (i.e. the pump is not required to be started) since the pumps are started and shown operable during their routine surveillance in the TS. The plant Appendix R evaluation requires RHRSW pumps A3, B3, C3, and D3 to be operable during Unit 2 reactor power operation to supply cooling water to the diesel generators. Compensatory measures will be taken when any of these pumps are out of service. For an Appendix R fire, these pumps are started from the 4KV Shutdown Board, the backup control station or they start automatically when any DG starts. The performance of the pumps (i.e. pump flow rates are within design limits) and the automatic start function are currently tested in the DG operability and EECW pump operability tests. The 4KV Shutdown Board start function (closing of the breaker) is also verified by the EECW pump operability test as the breaker must be operable for the pump to be operable. To verify the i backup control station start function is operable, the backup control switch ! must be-tested to ensure it will close the feeder breaker to the pump. The pump does not have to be started since pump operability and pump start are shown by the TS surveillance. Thus, the current testing for verifying operability of the EECW system by TS 3.5.C adequately verifies the system will perform'its required Appendix R function. The RERSW discharge valves on the C & D heat exchangers are required to be ' opened for Unit 2 and closed for Units 1 and 3. The RHRSW to the Unit 1 RHR heat exchsnger C and to Unit 3 RHR heat exchangers C & D have been capped , off. .Therefore, only the manual closing of FCV-1-23-52 and manual opening of . 2-FCV-23-40 and 2-FCV-23-52 are required to be verified every 18 months to show operability. i 0729R ,Page 190 '
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SECTION IX - REVISION / CONTROL OF THE PROGRAM This Section establishes the change process for tne BFN Appendix R Safe Shutdown Program. Any Employee any submit recommendations to improve this Program. Changes may include but is not limited'to the following:
- 1. Adding / deleting / revising compensatory measures or testing requirements
- 2. Updating the Program to comply with new regulations
- 3. Changes to improve the program
- 4. Providing alternate safe shutdown paths The control and updating of the BFN Appendix R Safe Shutdown Program will be
'in accordance with SDSP-2.2, " Controlled Manuals". This will ensure that this program is maintained in accordance with other such controlled programs at Browns Ferry Nuclear Plant. The additional restrictions provided in Attachment A shall also apply.
l i h l . I c- 0729R @mo g@@ _ i
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l DRAFT 4 ATTACHMENT A
, BFN Appendix R Safe Shutdown Program Revision and Control Process
- 1. A change request is generated by the employee (originator) detailing the basis and need for the proposed change (Form A), The originator obtains supervisors 2 concurrence with the. change. If the change request is for relief from a compensatory measure in effect, then the basis should state so and the respective Operations Supervisor must also concur with the proposed change.
Upon completion of the above, the change request is to be scnt to Systems Engineering.
- 2. System Engineering shall evaluate the proposed change and provide additional information if applicable. The proposed change shall be coordinated with Operations and the Nuclear Engineering (NE) - Nuclear Technology Branch (NTB) organization. NE-NTB will assign a change request number for tracking purposes. Upon approval of the appropriate System Engineering Supervisor, the change request is sent to NE-NTB.
- 3. A tracking number will be assigned by NE-NTB. The format will bE APPX R "XX-n YY" . Where XX is the year and YYYY is the assigned sequential nutaber.
- 4. NE-NTB will perform a design review of the change request and preliminary accept or reject the request on that basis. If NE-NTB rejects the change, appropriate justification will be provided on Form A which will be returned to Systems Engineering Section with a copy sent to the originating organization.
If NE-NTB preliminary accepts the change NTB will draft changes to the affected pages in accordance with technical specification 1.0.NN and perform a 10CFR50.59 safety evaluation in accordance with NEP 6.6. If the proposed change 3 does not involve an unreviewed safety question, the safety evaluation and l proposed change will be attached to Form A and copies will be forwarded to PORC members for review.
- 5. PORC members will review and comment on the change request, and resolve their comments with NTB. When all comments have been resolved, the change request will be submitted to PORC for approval in accordance with technical specification section 6.0, " Administrative Controls".
- 6. Upon PORC and Plant Manager approval, NE-NTB will establish an effective date ,
to incorporate the change into the Appendix R Safe Shutdown Program. This date ! shall be selected to ensure adequate time is available for applicable traihing, i procedure revision, safe shutdown instructions review / revision. '
- 7. If any changes are made to the change request af ter PORC approval, the change .
request must be resubmitted to System Engineering, NE-NTB, and PORC and their approvals obtained. 0729R Page 200
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RESPONSIBILITIES - i The following organizations have responsibilities in the BFN Appendix R Safe Shutdown' Program:
- 1. NE-NTB - Perform design review, prepare safety evaluation, submit and obtain PORC approval, and assign tracking number.
- 2. Operations - Responsible for verifying plant operations in accordance with the BFN Appendix R Safe Shutdown Program, any changes made to the program, and ensure that safe operation of the plant is not affected.
- 3. System Engineering - Ensures technical accuracy of the change and ensures safe plant operation is not affected.
1 l l 0729R Pame 201 >
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*J-FORM A APPENDIX R SAFE SHUTDOWN PROGRAM CHANGE REQUEST
.WMBER
- 1. DESCRIBE CHANGE (S)," PROVIDE PAGE(S) MARKUP, AND PROVIDE JUSTIFICATION FOR CHANGE.
f ORIGINATOR /DATE ORIG..SUP./DATE {
- 2. SYSTEM ENGINEERING REVIEW / RECOMMENDATIONS ACCEPT REJECT-I 1
SYST. ENG/DATE SUP./DATE
- 3. OPERATIONS REVIEW / RECOMMENDATIONS / COMMENTS ACCEPT REJECT ,
J REVIEWER /DATE SOS /DATE 0729R Page 202
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s; [* . FORM'A' (Continued)
- 4. NE-NTB REVIEW / RECOMMENDATIONS / COMMENTS-i_ ,
l l, ACCEPT j REJECT l SAFETY EVALUATION COPY OF AFFECTED PAGE ATTACHED f NE-NTB ENG./DATE i NE-NTB LNE./DATE i
- 5. PORC REVIEW / COMMENTS
_ ACCEPT REJECT l PORC CHAIRMAN -
'LAST PAGE i i
0729R Page 203
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DISTRIBUTION FOR'MEE' TING
SUMMARY
DATED: April.3,'1989 Faci.li ty: Browns Ferry Nuclear Plant,. Unit 2 41Dd d 6 File h NRC PDR
~ Local PDR-Projects Reading-ADSP Reading J. Sniezek D. Crutchfield B. D. Liaw.
S. Black-R. Pierson G. Gears D..Moran
.T. Daniels N. Markisohn M. Simms L'. Watson,-Region II OGC R. Wescott G.-Hubbard P.'Hearn ACRS(10)
GPA/CA (M. Callahan) (5) E. Jordan B. Grimes P. Gwynn J. Scarborough. i G. Marcus L. Norrholm , C. Ader. l J. Gray R. Borchardt BFN Rdg.. File i h e}}