ML20005F649
| ML20005F649 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 11/20/1989 |
| From: | DETROIT EDISON CO. |
| To: | |
| References | |
| NUDOCS 9001170094 | |
| Download: ML20005F649 (81) | |
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- l# M I ) Significant Hazards Consideration E'y @
[ ] Enviromental Impact - Categorical Exclusion " DIST. //~.a t M - @ [ ] Environmental Evaluation m }; ' e9 I ) Other ""- -- c4 D) is UFSAR change required? [, LCR No. g > { ) Yes ( ]No p,
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- 8) Section(s), Table (s), Figure (s), etc. Affected (Attach marked-up pages) ,
Figure 9.1-3 Sheet 1 c.,M Dt. o m .1 l C) Reason for Change The UFSAR Figure g.1-3, Sheet 1, is being changed by EDP-10230 which adds a 7'-0* high wire mesh enclosure for tool storage to the 5th Floor of the Reactor Bulloing between Columns 12-A and 13-A and between the columns and the decontamination pit. D) Reference end $succe Documents (loontity) - EDP-10230 Tech Spec b,,D f PDC-10230 Procedure
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DER PE (Attached) Yes McF F Ed Test Ay,MS gggggD j 1 Effectiveness Review (Attached) [ ]Yes [x]N DTCr7eteAP Awt h +FF Other Drawings. Design Calculations, Correspondence, etc. Nt F9 //5- Ars we el p .... . . . . . . . . * "" * * * * * " * **
- P A R T 2 : OPERATING LICENSE CHANGES [x] N A ***""* *" m ** *"*
A) Document [ ] Operating License [ ] Tech Specs [ ] Environmental Protection Plan [ ] Tech Spec Clarification B) Section(s), Table (s), Figure (s), etc. Affected (Attach marked-up pages) O*:) [(Reference
} Significant Hazards Consideration
) Enviromental Impact and Source
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( )Other D) is UFSAR change required? l [ ] Yes [ ] No LCR No. I STATUS ' ASB a NRC approval required by (date): , 9/n/v4 An [ ] Emergency [ ] Exigent condition will occur if not approved by: rjlST. (State date): _ _ . Explanation ncv. _. i F) Implementation 1 DER No. - p....uu..uno............ P ART 3 : A P P R O V A L S* """"""" * """* """"" * * "" * """* "
- l A) Originator R. Y. fut4 /J Date d:6 L>r M
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Date bbN8} C) Nuclear Generation Unit Head Date /7-T/6 D) General Director, Nuclear Engineering [ ]NA Date ~7 E) Plant Manager & Date OD F) Other , Date G) Director, Nuclear Licensing % Date Y//# M H) OSRO Approval (Tech Spec Amendments) { ]NA Date ll) NSRG Approval (Operating License Amendments) ( l NA Date l Form FIP-RA2-01 Att 1 P1/1030189 DTC 5 TCLCR for UFSAR File: 1735 DTC t ; tdt.S lor other
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EDP lotof Test 6 ate.tM Mes ms . t*r.u nt .ot k ti.9tt. ct PDC i o t e s- Tech Spec 1.h.s. t Maus us utut u mQ Procedure ute .ituim ABN w/A ! DER sS-eis % SE (Attached) a& cowl. Rev.i 4 S&oNs- > 1 Effectiveness Review (Attached) Other Drawings, Design Calculations, Correspondence, etc. pn..u.n..uu......*""*
- PART 2: OPERATING LICENSE CHANGES [>0NA **"*"*"*"E"*""** l A) Document i
[ ] Operating License [ } Tech Specs [ ] Environmental Protection Plan
] Tech Spec Clarification i B) ILection(s), Table (s), Figure (s), etc. Affected (Attach marked-up pages)
C) Reference and Source Documents Attached STATUS , O [[ ]] Significant EnviromentalHazards Consideration impact - Categorical [ ] Environmental Evaluation Exclusion ASB-o w 'Jln I'A 4 DIST. [ ] Other D)is UFSAR change required? RW. [ ] Yes [ ] No LCR No , E) Priority > NRC approval required by (date): An ( ) Emergency [ ) Exigent condition will occur if not approved by: yng mntgn (State date): ,
,g Explanation DSR P9-/A/. t475 Akr!(
l MM ** :%;;Q 1 m M-W _ F) Implementation DER No. g% p,.u.no.....u....... ....
- PART 3: A PP R O V A L S* "" " " " "" * "* "" **""*""""*"" *" " * "" l A) Originator ?.Trk,% ELL & -r ,,4 3 r d Date 6 9. s 3 e c. m m u. tr o , .
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Date GVFW C) Nuclear Organization Unit Head I.bM7eb Date 4 *M'IT D) Director, Nuclear Engineering [ ] M[ Levt Date / J/O E) Plant Manager / Nb Date 74/d F) Director Nuclear Licensing Date k7/
- 0) OSRO Approval bd NA Date ,_
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. "The not volume of the i
secondfloorsoftheturbinebuildin!3 secondary containment is 2.8 x 106f The reactor building is a Category I structure designed and con-
structed in accordance with all applicable local and state build- .
- ing code requirements.
Substructures and exterior walls of the building up to the i refueling floor consist of poured-in-place, reinforced concrete. ! The building structure above the refueling floor is a steel frame j covered with insulated oetal siding and is sealed against leakage. The building is designed for an esternal pressure of
! 0 295 psig and for low inleakage and outleakage (depending on I
wind condition's) during reactor operation. . I 6.2.1.2.2.1 Rosetor Buildine Penetrations . mw-onnel andxequipse are eq pped w th -
- A se opent for pe went r-strip-t seal for airtihhtness. Personne entra ces ions
- '
tothe\econdary e ntain areat}efell ing loc %php a. The reactor core isolation cooling system / core spray %,ga pump room at Elevation 551 ft 0 in. , DI b. The auxiliary building basement from the CRD pump room at Elevation 551 ft 0 in. l i c. Between the turbine and auxiliary building at Elevation / 564 ft 0 in. .i
- d. en to the reactor building at 583 ft 6 in.
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- e. g:pentry to the reactor building at 583 ft 6 in.
- f. Between the reactor building and the auxiliary building at Elevation 613 ft 6 in.
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- g. Between the reactor building refueling floor and the auxiliary building at Elevation 684 ft 6 in.
f h. Between the reactor building refueling floor and the
a miliary buildin at Elevat n 701 ft 0 i . 1 l Al of these entr es have a vos bule with d ble doors t j ami ain secondary ontainment in egrity. The double doors are i 1 eithe interlocked t prevent the ening of on door until e
- other or is closed one of the rs is keyl ked closed.
. The keys or the locke closed doors re administ tively contro11e the nuclea shift superv or. Penetrations f'or piping and ducts are designed for leakage
. characteristics consistent with containment requirements for the l 1O i
v entire building. Electrical cables and instrument leads pass .'h through ducts sealed into the building wall. J p\ 6.2-14 REV 1 3/88 s
;
i CONTINUATION SHEET
- LCR l8 19 l- 11 11 14 l- lU lF lS l TSC l l l- l l l l Revision 0 PageJ of (,
A) Document : i i
- 1. Access openings for personnel and equipment are equipped with weather-strip-type seals, except for ;
the railroad bay entry, for air tightness to meet secondary containment negative building pressure requirements. The railroad bay entry doors have inflatable seals which provide the air tightness requirements as well as site flood protection. Personnel entrances to the secondary containment are d at the following locations.
- 2. All of these entries have a vestibule with double doors to maintain secondary containment integrity. .
The double doors are administratively controlled to prevent both doors from being open at the same i time, thus maintaining secondary containment integrity. Additionally, as an administrative aid, the - doors have either interlocks to prever.t the opening of one door und the other door is closed or one ! cf the doors is key locked closed. The interlock feature is not considered QAl safety-releted. Failure ; Cf these interlock circuits would not cause the doors to open on their own accord. Keys for the i Iceked closed doors are administratively controlled by the Nuclear Shift Supervisor, in the case of the railroad bay airlock, the doors have inflatable seals which are considered active components, i' Therefore, to meet single failure criteria and maintain secondary containment integrity, the inner door seal is supplied from Division I of non-interruptible control air and the outer door seal is supplied from Division 11 of non-interruptible control air. The railroad bay airlock doors also have low seal - ! pressure alarms which are monitored in the main control room. 1
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9.3 PROCESS AUXILIARIES t 9.3.1 comeressed Air system ; i 9.3.1.1 Deslan Bases ! The Fermi 2 station and control air system provides the plant with a reliable source of clean, dry, oil-free compressed air for plant operation. The control air system is designed t'o provide oil and dirt-free air with a dowpoint of -40'F (at pressure). l
;
The control air compressors, aftercoolers, dryers, and receiver ' tanks are provided to supply air to some of the engineered l sa*ety feature (ESF? equipment in the plant when the normal sup- : ply of control air hs not available. Because the noninter-L ruptible portion of the control air system provides control air i to EST equipment, it is classified as a safety-related system. t I f ! The station air and interruptible control air systems are con-structed in compliance with standards for Quality Group D compo-t nents. The criteria are met by designing the systems to ASME section VIII and ANSI B31.1.0 code requirements. These systems i are nonseismic. The noninterruptible control air system is constructed in compli- l ance with upgraded standards for Quality Group D components. These criteria are met by designing this system to ASME j section III, class 3 requirements. The system is Category I. O ; i l 9.3.1.2 System Descriotion I The air system is composed of two subsystems. The first is the r l supply and distribution of station air and the second is the j supply and distribution of interruptible and noninterruptible control air. The station air and interruptible control air 2 ! supply equipment is located in the turbine building. The non-interruptible control air system is located in the auxiliary i building. The station and control air systems are the source of ! compressed air for use in routine maintenance operations, in equipment process cycles such as domineraliser backwashing, and l as an instrument and control media. The compreesed air system is
;
shown in Figure 9.3-1. , i The station air system consists of three one-half-capacity l 1225 sefm, two-stage nonlubricated reciprocating compressors i equipped with inlet filter-silencers, and intercoolers and l aftercoolers. Two 150-ft3-capacity air receivers and the i ) station air distribution piping, valves, and fittings complete e the station air equipment. j. In operation, ambient air from the turbine building is drawn This into the station air compressors via the inlet filter-silencers. air is compressed, cooled, and discharged into the station air ! O receivers. Normal practice is to have two compressors in operation on automatic control. These two compressors maintain i
;
i 9.3-1 REV 2 3/89
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*r=*- i O * #* =r-(1aa i)i*a ir i natic regulation of each of the compressors through successive one compressor remains on standby to provide ,
unloading steps, assurance against loss of capacity due to a single compressor outage. l From the station air receiver, the station air is distributed I l throughout the plant via the station air header / riser system. The station air system is sized to minimite the pressure' loss of ! l air at the point of use. i The noninterruptible control air portion of the system consists i of two 100 percent-capacity 100 scia, single-stage nonlubricated reciprocating air compressors two 100 percent-capacity parallel l strings of oil filters, air dryers, and afterfilters; two. control ; air receivers and associated piping, fittings, and valves. l l I ? l During normal plant operation, the source of noninterruptible and i interruptible control air is through interconnections between the i station and control air systems. Compressed air from the station i air system is supplied through one of these interconnections to ! the Division I and II noninterruptible control air compressor - discharge headers. The air then flows from each header through i l its divisional 100 percent-capacity filter and. dryer. It is i cleaned of all particles of dirt 10.5pm (nominal), 10.9pm abso- : 2 lutec and then dried by a regenerative desiccant-type dryer to a l'
-40'F dowpoint (at line pressure). After leaving the filter /
dryer, the noninterruptible control air flows to its divisional control air receiver from which it eventually flows to its point i of use through its divisional noninterruptible control air i distribution system. l Another station air connection supplies the interruptible control . stem through the interruptible control air system's supply l air filtersy/ dryer, which functions to meet the same air quality cri- l teria as the noninterruptible filter / dryers. The interruptible i control air flows to the interruptible control air receiver, j l which supplies the interruptible control air distribution i system. The station and control air compressors, air receivers, ; filters, and dryers are designed to operate in an ambient temper- ! ature range of 60'F to 100'F, a range of 20 percent to 100 per- l cent relative humidity, and a radiation field of 1 mR/hr. l The control air distribution system is divided into two distinct j parts: interruptible and noninterruptible. Moninterruptible l control air (NIAS) supplies, through two separate distribution ; systems (Divisions I and II), equipment in the following systems: d 2
- a. Standby gas treatment system (SGTS) I i
- b. Control center air conditioning system (CCACS) ;
l 61 l 9.3-2 REV 2 3/89 J l
s . M M* 11% u pts rsnusaursAn j j "4 % 6 4 A i
- c. Main steam isolation valve leakage control system j j
.O (MSIVLCS) . .
c j d. Primary containment atmosphere monitoring system (PCAMS) j i l; e. Emergency equipment cooling water system (EECWS) j f. Primary containment pneumatic supply system . l w.L Torus to_reactogbtijding vacuum relief system. _h. .d he re . In raaasion, Ru te.ivisiol u wI\esk NIASdeeIdes.A prov r . con rol air for the ; i following: - i a. Primary containment isolation of drywell equipment and floor drain sump pump discharge lines 2 l l . b. Back-up supply for Division I (N2) pneumatic supply to , i the primary containment. -
- Division II NIAS supplies, in addition, air operated valves in j the-following systems
I a. Righ pressure coolant injection system (IPCI) ' i b. Reactor core isolation cooling system (RCIC). , l[ All other control air users are connected to the interruptible
) control air distribution system. Interruptible control air (IAS) l 1s supplied through its own set of filters, dryer, and receiver 4
tank, which is fed from the station air system. The station air compressors and aftercoolers are cooled by the turbine building closed cooling water system (TBCCWS). The control air compressors and aftercoolers are cooled by the reactor building closed cooling water system (RBCCWS) or the EECWS. During normal operation, any one of three installed station air comgressors will be in operation. One of the other two will be in auto" and the third compressor will be in the "off" position. Normal operating pressure from the station air compressors is 100 psig. If the station air header pressure drops below 95 psig, the compressor in " auto" will automatically start. If the . pressure drops to 90 psig, an alarm in the control room will be initiated and the third compressor may be manually started from 2 . the control room panel. If the station air header pressure continues to decrease, at 45 psig the station air supply header , isolates and only supplies the IAS and NIAS. An alarm as , initiated in the control room. Should station air supply pressure to either division of NIAS decrease to 85 psig, its division's control air compressor -O automatically starts. If the pressure continues to decrease, at 75 psig the station air supply isolates from the NIAS and alarms. l 9.3-3 REV 2 3/89
l' ,) . LICENSING WANE 500EST i,. . CONTROLLED Len 18191 - 1111181 - IUlrisi
# Revision I Pane 1 of f 5" I m m m m p FANT 1: W5&R. PLAN. OR runwunAN sw.vI5ICEI [ JEA ;;; ;;;#l j
1 A) Document g N j UFSAR i 5) Bection(s), Table (s), Figure (s), etc. Affected ( Attach marked-up pages) ' i UFSAR Fig. 9.1-23 and Section 9.1.3.4 & Section 9 1.3.2 I ' l C) Reason for thanae' Removed FPCCU vacuus breaker valves G410vrv67 A & B l and 04100F065A h B. ~ Lem esm qg z}l ) g Q t I D) Referenos an6 h ts (Identify) r{s$ r) . i EDP 10109. Re i Test n/A l , i l PDC 10109. Rev. 7 Tech 5pec ~ 3.9.9 ( ABN N/A Procedure 0/A , DER N/A SE (Attached)'59-0041. Rev. # w C I Effectiveness Review (Attached) , Other LCR 89-123-151 b-Drawings, Design Calculations, Correspondence, etc. I h$ jeeeeeeeeeeeeeeeeestePART 2: OPERATING LICENSE CBANGES [ I]NA e########l p' E t A) Document ' ' ' ) ( ) Operating License ( ) Tech Specs [ ] Environmental Protection l Plan ' ( Tech Spec Clarification w B) SectLon(s), Table (s), Figure (s), etc. Affected (Attach marked-up pages) ,. $ l *' .-{l C) Reference and Source Coc eents Attached ' n ( ) Significant Har.ards Consideration STATUS U ( ) Environental Impact - Categorical Exclusion " l" ' [ ] Environmental Evaluation ASB Y ( ) Other ruer
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D) Is UFSAR change required? ; ( ) Yes ( ) No LCR No. REV. E) Priority NRC approval required by (date): An ( ) Emergency ( ) Exigent condition will occur -d un .w vm py; (State date): . Explanation l F) Implementation DER No. TW5TITUTTIT5TTUTFeteenPART,3: APPROT A13emmmmmmmmmmem l MXxvecap Q1' a a Half Y-Eb M A) Originator M. A. Majumder A/,b d - /2. /M.[e- h/S&W Date 4- 2#- ry B) Technical Expert
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. Hi NS E Appr0 val k'NA Date I furs FJP-RA2-01 Att 'l : 1,1 092185
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r L.cR 89-uc-uF3,Rayh Fksa 4 es 5-FI N N 8A" SC 39 6041, Rav G UPaAA 544. 81 .l. 5 . 4 () - Table 9.1-4 presents the major parameters used in the calcula tions. Results indicate that the dose from this postulated accident would not exceed a fraction of 10 CFR 100 limits. i Among the sequence of events that would occur prior to the postu-late f ailure of the f uel pool cooling system is the cooling of the fuel in the pool for 21 days af ter its removal from the reactor. Prior to and during this period of time, the fuel clad-ding temperature would have been reduced to a (maximum) value of slightly above 125'F. Should the fuel pool water temperature increase to the boiling point, the (maximum) fuel cladding ! temperature would be slightly above 212'F. This fuel cladding temperature is a fraction of the fuel cladding temperature during normal plant operation. i The physical characteristics of the fuel and the integrity of the fuel cladding would not experience i changes that could cause an activity concentration in the fuel j i pool water in excess of the _ activity in the reactor water during full-power operation. Dose calculations performed by Edison, ' based on the above, indicate that the design criteria applied to the spent fuel pool cooling system are adequate to provide reasonable assurance that the plant can be operated without undue risk to the health and safety of the public, consistent with the 4 requirements of Criterion 2 of Appendix A to 10 CFR 50. l ( In summary, the spent fuel storage pool cooling -system's design,
siphon-breaking piping arrangement, redundant transfer gates, emergency makeup water supply from the RNR service water system, l and RHR backup capability provide a completely reliable system j for the storage and cooling of spent fuel. j 9.1.3.4 Testino and Insoection i Prior to power operation following a refueling outage, a deter-mination will be made that the heat generation rate in the fuel pool $s within the current capacity of the FPCCS to maintain the .!LETA pool temperature at 125'r or less. The feu
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No special tests are required for instrumentation on the FPCCS. The instrumentation will be subjected to routine testing. The FPCCS Preoperational Test program is discussed in Chapter 14. 9.1.4 Puel Handline System 9.1.4.1 Desien Bases I O The fuel handling system provides a safe and ef fective means for transporting and handling fuel from the time it reaches the plant until the time it leaves the plant af ter postirradiation cocling. 9.1-27
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unas s ac. 8b I O 2-normal makeup water source for the system is provided from the : condensate storage tank to the skimmer surge tanks. Backup cooling is provided to the fuel pool by means of a perma- , nontly piped cross tie to the RNR system. In this mode of opera-i tion, one RHR pump and the corresponding RHR division heat exchanger will provide the means to cool the fuel pool . This , l cooling circuit is established by opening cross-tie valves ,
- V8-3264 and V8-3029 and closing FPCCS valves V8-3006 and V8-3253 (see Figure 9.1-23). For the designed piping configuration, the I RNR pump will deliver approximately 5400 gpg, and the RRR heat I i exchanger will remove approximately 35 x 10' Stu/hr at 125*F
- pool water temperature (see Table 9.1-1). To ensure the avail- i ability of backup cooling via the RNR system, the cross-tie pip- ,
l ing, the FPCCS piping from the skimmer tanks to the first anchor l downstream of valve V8-3006, and the FPCCS piping from the first anchor upstream of valve V8-3253 to the fuel pool diffusers are l Category I. J i toth FPCCS heat exchangers operating in parallel are designed to *
- remove the maximum heat load produced by various combinations of
- spent fuel discharged from the equilibrium fuel cycle at the time the RHR system is isolated from the pool, plus the heat being i released by batches discharged at previous refueling (see Subsec-tion 9.1.3.1). The FPCCS is designed to maintain the fuel stor-age pool water temperature below 125'F while removing the maximum
- 8 normal heat load from the pool with the RSCCW temperature at its j'
maximum. The pool operating temperature is permitted to rise approximately 25'F above the normal operating temperature of 125'F when larger than normal batches of fuel are stored or when the FPCCS becomes ! incapacitated. In this case, either of the heat exchangers in l , the RHR system can be used in conjunction with the FPCCS to supplement pool cooling. Table 9.1-1 also lists the characteris-tics of an RHR subsystem in the fuel pool cooling mode. The design of the fuel storage pool is such that the top of the stored fuel is at a lower elevation than the bottom of the pool gate between the reactor well and fuel storage pool. There are no connections to the fuel storage pool that could drain the pool below the elevation of the bottom of the pool gate when the gate ! is removed for refueling, or below the normal pool level when the ! gate is in place. To prevent water from being siphoned out of l the _ pool ,_ the pipina_ saterino the fuel Etorag Doo_1_ is _ fitted /pg,grt %LM wit E RE d el y: M r:: = 5:::h:::RM J 551 01:r:ti 5
~
l r A level indicator, mounted at the valve rack, monntors reactor d well water during refueling. A high rate of leakage through the refueling bellows assembly, drywell to reactor seal, or the fuel storage pool gates is indicated on the operating floor instrument racks. , Fuel storage pool water is continuously recirculated. The O. circulation patterns within the reactor well and storage pool are ( l Wo&A LLy Sv64dfacb VC.W.5 9.1-21 pgy pp kWocd w/L.L. $fQ sere r4 we##WEDm wm wa. nk meena U 44.5Y. .
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- P ART 1 : UFSAR, PLAN, OR PROGRAM REVISION [ ]NA *""*"***""*"****"I I
A) Document UFSAR ; 4 S) Section(s) Table (s), F6gure(s), etc. Affected (Attach marked-up pages) ! UFSAR Figure 9.1-3 Sht.1 (6A721-2003-1)
, C) Reason for Change: The referenced UFSAR Figure is revised per EDP-9734. The impacting EDP ;
(Index item No.10, Pg.1 of 8) page is included in the attachments. ,
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-J EDP-9734 Tech Spec PDC Procedure . ,4f.
ABN SE (Attached) 89-0065 "' ' r _ 3,, _ j DER PE (Attached) N" Test Effectiveness Review (Attached) [ ]Yes [ JNo Other Drawings. Design Calculations, Correspondence, etc. p... .n.u.u. .u... ....u .*
- P ART 2 : OPERATING LICENSE CHANGES PCNA *"*"""""**"**"*l 1
A) Document [ ) Operating License [ ] Tech Specs [ ] Environmental Protection Plan ) [ ) Tech Spec Clarification S) Section(s), Table (s), Figure (s), etc. Affected (Attach marked-up pages) i
- C) Reference and Source Documents Attached l [ ] Significant Hazards Consideration [ ] Environmental Evaluation i [ ] Enviromental Impact / Categorical Exclusion [ } Other D) is UFSAR change required?
I [ ]Yes ()No LCR No. STATUS E) Priority NRC approval required by (date): ASO Y I["j" , An [ ) Emergency [ ) Exigent condition will occur if not approved by DIST. (State date): MtX Explanation F) Implementation DER No. p...u..u...o.u.u n...." PART 3: A P PA O V A L S* " " " " " " " " * " " * " " " " " " " * * " * * " * " ""
- l A) Originator H. Sahiner Date d[N/87 S) Technical Expert .4 , Date 6[88k C) Nuclear Generation Unit Head Date 7b/81 D) General Director. Nuclear Engineering [ )NA /J Date 7 N E) Plant Manager b Date 7 D F) Other Date O) Director, Nuclear Licensing Date f//MM H) OSRO Approval (Tech Spec Amendments) []NA Date 1
l ll) NSRG Approval (Operating License Amendments) [ } NA Date l L rs . e.s t t n D P. *) A i Aee 1 D9 et A*2A1DD F% T o M TPi eB dam t af f A B f Aim .1821
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i B) Esotion(s), Table (s), Figure (s), etc. Affected (Attach marked-up pages) i ! Flauros 7 3-9 sheets 3. 9 and 10. s r n - % . s cr u i.ob m .io. cr u l eco.it 4 i; l C) Reason for thange To revise Ur5AR FLaures 7.3.9 sheets 3. 9 and '10 to yg g h l incorporate the sodifhostions per ED,-I l576 Rev. 5. g y I by 3
- D) Reference and source Dooments (Identify) h I EDP 45"6 Rev. # Test N/A e
. PDC 1 15"6 Rev. F Tech spec N/A h g ABN l I/h Procedure N/A g S2 i DER 55-0074 SE (Attached) 59-9966 Gl% oo Effec",1veness Review (Attached) N/A e4 i Other 30ER 87-002 a i Drawin , Design Calculations, Correspondence, etc. h g., l******************gPARTtt OPERATING LIG NSE GRAN 35 [><)RA *********l ! A) Document ( ) Operating License [ ] Tech Specs [ ] Environmental Protection
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[ ] Tech Spec Clarification l B) Section(s), Table (s), Figure (s), etc. Affected (Attach marked-up pages) ,[ l C) Ileforence and Source h iemaats Attached WIbYN l [ ] Significant Hazards Consideration gggpa q /m /gq [ ] Environental Impact - Categorical Exclusion
- [ ] Environmental Evaluation OlST
l [ Other .o w-l D) Is R 5AR change required? ( ) Yes ( ) No LCR No. ! E) Priority NRC approval required by (date): l An i JEnergency ( ) Exigent condition will occur if not approved by: (State date): i Explanation - F) Implementation
! DER No.
4 i....................,A., 3, ,,,,o m .,............. ,,,. , . , . ,.3 A) Originator R. Guamaraju bm ! S& S//7 89 B) Technical Expert ,f/_, . - Date6/2f[8$ C) Nuclear Oraanization Unit Head b6 N', A[ Date If D) Director. Nuclear Entineering ( INA!) I. Date4/f/df E) Plant Manager ab Date u /7 F) Director. Nuclear Licensing X Date [/[ G') OSRO Approval ( ) NA Date IH) MSRC Anoroval I I WA Z Date W
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. D5N: /VM4.WS (el d Revisionh Page / cf /7 p........ ............rAMT u ve aaR. PLAN. OR PROGRAM REVillON [ }NA ""--""**""""*l A) Document
//F5/94 CONTROLLED B) Section(s). Table (s), Figure (s), etc. Affected (Attach marked-up pages) %
X- / 3 I.T. S. L iB. J.r. 2. L 2 s2 fa r i 71.t. 2.1% 2.1.2.1. Am n s..n.1L t-n . sn io, o. r C) Reason for Change m& few# w em'- < hu r-Mews:' vs-
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- D) Reference and Source Documents (identity) . ...
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MDP Tech Spec PDC Procedure i ABN SE (Attached) M w(.O DER PE (Attached)- Test Effectiveness Review (Attachec) DdYes [ JNo 1 Other L C. I.~ & ; . '),e* t) F f Dr_awings. Design Calculations. Correspondence. etc. < p...........................'PART2: OPERATING LICENSE CHANGES I ]NA '""""**""'""*l A) Document [ ] Operating License [ } Tech Specs [ ] Environmental Protection Plan j [ ] Tech Spec Clarification B) Section(s), Table (s), Figure (s), etc. Af f ected (Attach marked-up pages) C) Reference and Source Documents Attached [ ] Significant Hazards Consideration [ ] Environ,v. ental Evaluation j l ) Enviremental Impact / Categorical Exclusion ( ) Other
- D) is UFSAR change required?
l [ ] Yes ( ) No LCR No
- E) Priority
+
NRC approval required by (date):
An [ ] Emergency [ ) Exigent condition will occur if not approved by:
- (State date)
1 ! Explanation F) lmplementation DER No. . { p.. ..u n n...... .... . .... . . . P A RT 3 : A P P R O V A L S* " " " " " " " "" " " " ""** " """ **"" * ""' " l A) Originator M
.. Date 4/[$Ml9 B) Technical Expe t i u Date -
C) Nuclear Generation U t Head Date Yof/d D) General Director, Nuclear Engine 3ripg NA - Date 4[-Y[7 E) Plant Manager J/ Date F) Other Date
' G) Director, Nuclear Licensing s/ - _
Date E//MP7 H) OSRO Approval (Tech Spec Amendments) [ ] NA Dete
- 1) NSRG Approval (Operating License Amendments) ) NA Date l torm FIP-RA2 01 Att 1 P1/1030189 DTC:IN TCLCR for UFSAR File: 1735 DTC: ! ) TDLCR for other
l
. l EFFECTIVENESS REVitW ;
I u I F I .51 I
- Reference Lcm 1819 l - I/I212l-Revision 4 Page 2.of 17 . _ _
r..........................................PART1: UF S A R I ]N A *" *" * " * """*"* "" * * * """
- l A) Quality Assurance Program I JYes M No Does the change (s) cease to satisfy the criteria of 10CFR50. Appendix B and the UFSAR program commitments previously acr.epted by the NRC?
Provide the basis for each change on Attachment 2. Page 2. _3) Fire Protection Program ( )Yes ( ) No Does the chenpe(s) significantly decrease the level of fire protection in the plant? I IYes [ ] No Does the change (s) result in fallure to complete h Protection Program i opproved by the NRC prior to license issue? Provide the basis for each chance on Attachment 2, Pane 2 P"" PART 2: R ADIOLOGICAL EMERGENCY RESPONSE PREPAREDNES PLAN DONA """;;;;*"" l A)I ) Yes I , No Does the change (s) decrease the effectiveness of the RERP Plan?
! )Yes ! )No Does the RERP Plan, as changed, cogse to meet the standards of 10CFR50,47(b) and 10CFR50 Appendix E?
Provide the basis for each chen;e on Attachment 2. Page 2.
- po..........................PARTN S E C U RIT Y PLAN S DO N A """"** * **"""* *"""*"""""l
- A) Document l
~ A) I J Yes I J No Does the change (s) decrease the effectiveness of the Physical Security f Plan or Security Personnel Training and Qualification Plan prepared l
pursuant to 10CFR50.34(c) or 10CFR737
! ) Yes ( ) No Does the change (s) decrease the effectiveness of the first four categories of Informational Background, Generic Planning Base, Licensee Planning i
Base, and/or responsiblity matrix of the Safeguards Contingency Plan l prepared pursuant to 10CFR50.34(d) or 10CFR737 I Provide the basis for each change on Attachment 2, Peg e 2. - p u . . . . .. . . . . . . . . . . . . . . . . . . . . P AR T 4 : PROCESS CONTROL PROGRAM DdN A*"""*;"""*""* l A) [ ] Yes [ }No Does the change (s) reduce the overall conformance of the solidified l waste product to existing criteria for solid wastes in accordance with I l Technical Specification 6.13? Provide the bests for each change on Attachment 2 Pa-}e 2. l pu..........n........................ PARTS: ODCM LX N A * * * * " *""" " """" : : "" * * * * * * " l A) I ) Yes [ ] No Does the change (s) reduce the accuracy or reliability of the dose l calculations or setpoint determinations in in accordance with Technical l Specification 6.147 Provide the basis for each chan ge on Attachment 2, Page 2. r . . . . . . . . . . . . . . . . . . . . . . . . . . . . , A R T m A , , R D m . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . ... . l A) Originator d Date MM
@ B) Technical Expert / _
Date U/ ' y
/ r
/j C) Quality Assurance (For Security Plans only) Date l
- 10) OSRO (Not required for UFSAR Changes) Date l rorm FIP-RA2-01 Att 21/2 030189 DTC: [ ] TCLCR for UFSAR File: 1735
] DTC: [ ] TDLCR for other
- - . - -m, _ . . - , -
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. EFFECTIVENESS REVIEW DOCUMENTAWON .
I . Reference LCR 1819 l- l /l 2.12.l- lul#151 ) I Revision O Pene3 of 17 1 _ D;cument i
- t/F68R Listed below is each change by section and page; the reason for the thenge; end the. basis for etncluding that the revised plan or proprem continues to satisfy the criteria for that plon or program.
Section/Page Change Besis . i l
/ 2 2./. 2 y I'22-/ ADD 7"?M /9stis[edf To 7ML.,,, TWW- i-xWier is n >
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i, . ( FERMl 2 UFSAR L2 6%IL2. 0P: P. 4- os 17
. S&l ia ,
l- The operating conditions are the following: Mode Switch Average Reactor j condition Position Coolant Temnarature
- 1. Power operation Run Any temperature
- 2. Startup Startup/ hot Any temperature
- standby ,
j 3. Not shutdown Shutdown Greater than 200*F
- 4. Cold shutdown Shutdown Less than or equal to 200'F Less than or oo.ual to 140'F
- 5. Refueling Shutdown or l refuel ;
} The Shift Technical Adviscrs are required to have a Bachelor's ! ! Degree in engineering or equivalent from an accredited institu-i tion and 2 years'of related nuclear experience. They complete an 2 , INPO-accredited Shift Technical Advisor training program. This : program contains all of the SRO training materials and practical i skills training that has been identified on a job task listing : ! that is Termi 2 specific. The Shift Technical Advisors are ' normally assigned to a shift, and participate in the training cycle for licensed operator requalification. l2 > The Nuclear Shift Supervisors, Nuclear Assistant Shift Super-l visors, Nuclear Supervising Operators, Nuclear Power Plant Oper-l ators, and Nuclear Assistant Power Plant Operators are trained in i the following areas of radiation protections i !~ a. Use of portable radiation detectors ! b. Limits of exposure rates and accumulated doses
- c. Use of protective barriers and signs
- d. Use of protective clothing and breathing apparatus
- e. Limiting contamination
- f. Pertinent plant and federal regulations.
shift personnel, however, do not perform technical quantitative functions in chemistry, radiochemistry, and other areas. Tech- , nicians who are qualified to meet the requirements of ANSI N18.1
=
perform these functions; during off hours they are called in, when needed, by the Nuclear Shift Supervisor. 13.1.2.5 Minimum oualification Recuirements for Nuclear Plant 2 Personnel During the selection of personnel and the scheduling of training : assignments for the plant staff, the requirements of Regulatory Guide 1.8, Personnel Selection and Training, will be met. The combined experience and education and an assigned training program are such that the criteria in Section 4 of ANSI N18.1 (1971), Selection and Training of Nuclear Personnel, are met. For these determinations, the following plant staff positions are identified with tne classifications contained in Section 4 of - ANSI N18.1: , 13.1-15 REV 2 3/89 l
. FinMi t uFSaR b 5 05 0 . Fermi 2 Staff Position ANSI N15.1 Classification .
General Director - Nuclear Engineer in Charge (Section ) Engineering 4.6.1) i Plant Manager - Kuclear Plant Manager (Section 4.2.1) ! Production Superintendent - Operations Plant Manager (Section 4.2.1) i I Maintenance Manager (Section ! (Superintendent-Maintenanceadtre
' 4.2.3) M:f101;;;i: / m - M d 6<
- 5 )
Quality control Inspectors ANSI N45.2.6 , Nuclear Assistant Power Plant Operators not requiring NRC operator - license (Section 4.5.1) 2 C05?rel ";;; vi;;r " Suposdeur Let regsiring-WRC-
'- <2/utk o US?M? U;;ti.fL '*3'It t M'O Pk,; w
$ fi M aman" f %v,s %n ~ &tA2..,.m. . . .. . m. 2. .,.
c..... . nance- .. .u ,., _ a_nC f/e<7tu .11canas. (section 4.3.2) s -n vu~ por . . . ~ lifed %D $(M.S.sM4sn4/4 ye r e .. _ 3133% r 5 j,,,g,5 .
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g;;;;y . , ,,,,,pt,g,,_,,e 2l inunn m .u.. (: tica . 3.:). General Maintenance Journeyman Repairman (Section 4.5.3) Superintendent - Technical Technical Manager 2 Engineering (Section 4.2.4) General Supervisor - Instrumentation and Control Paintenance Instrument and (section 4.3.2) Cont rol ;".1 --" I- - A s p ,) , a z u apu lso+ Moi t'st"w% WtUc" & Oenessi Furwm.n Macle;r :sr:rvleer not._ requiring-NRC(g;aW6{.t.) lastrumsat m 8 t ntrol 1-icense---(Section .4. 3.-2)- i 1 j Fe es.u - 14achar--Insttement Supervisor--not-requiring ! and rM t:ci anc Lwnsi- tiectioni.3.t ! Nuclear Instrument Repairman Technician (Section 4.5.2) Engineering Technicians Technicians (Section 4.5.2) 13.1.2.6 Oualifications of Plant Personnel The resumes of the initial appointees to the managerial and supervising technical positions for Fermi 2 were included in the , original FSAR. The qualification summaries of personnel cur-rently filling these or equivalent positions are maintained onsite and are available for review by the NRC. - O t 13.1-16 REV 2 3/89 l
' LW 6i In. vr-S 2cv p i-FKMl 2 UFSAn P. (c er p
13.5.4 Oeerstino Procedures j The preparation of the Operating Procedures is under the direc-tion of the Operations Support Engineer. The implementation of 1 these procedures mainly falls to the operating group and is performed under the direction of a licensed senior reactor i operator or reactor operator. Procedures that are prepared ) include, but are not limited to, those in Regulatory Guide 1.33, i Appendix A. J j! 13.5.4.1 General Ooeratino Procedures j The General Operating Procedures provide the necessary instruc- , tions for the integrated operation of all plant systems. Sign- ' offs are provided to ensure that necessary operat;,ng instruc-
tions, tests, and calibrations have been completed and are also t used for confirming the completion of major steps in the proper sequence. General Operating Procedures are prepared under the
- direction of the operations Engineer and implemented by Oper- !
- ations personnel.
13.5.4.2 system operatino Procedures ! System operating Procedures provide the necessary se uence of
steps to properly operate a particular system, inclu ing the
- following, as necessary: ,
- a. Normal operation
! b. Startup operation
- c. Shutdown operation
- d. Standby operation
- e. Automatic initiation
- f. Manual initiation.
13.5.4.3 Alarm Response Procedures Alarm Response Procedures provide guidance on actions to be taken l by the control room operator when the alarm annunciators actu-ate. Each procedure contains the followings
- a. The title of the annunciator
- b. The actuating device
- c. The setpoint(s) of the actuating device l
- d. The possible causes of the actuation
- e. The immediate action to be taken by the operator and those actions which occur automatically
- f. The subsequent action to be taken to return the system to its normal mode of operation, if necessary. -
13.5-3
. FECMituPSAR P.1ov n j
I 13.5.4.4 Abnormal Operatins Procedures
- 1. .
l Abnormal Operating Procedures provide operator guidance for l l stabilizing the plant or for restoring normal operating uendi- ) tions following a perturbation. l I 13.5.4.5 Emercency Operatino Procedures l Emergency Operating Procedures provide operator guidance to miti-l gate, reduce, or eliminate the consequence of an accident or i i potentially hazardous condition that has already occurred, to implement the emergency plan, or to prepare for possible haz-
- ardous natural occurrences.
13.5.5 Maintenance Procedures Maintenance activities that affect the performance of safety-1 related equipment are preplanned and performed in accordance
; with written procedures, documented instructions, and drawings appropriate to the activity. Procedures for performing various categories of maintenance are prepared following the guide- '
lines contained in Regulator Q The l Superintendent - Maintenance these procedures.
- _[g A.MaintenanceFr' c4Tv5Ddm y { Appendix _.. is responsible for 4 Nuclear Shift Supervisor before performing maintenancefon# plant equipment. This ensures that the operability of redundant safety-related systems is maintained as required by the Technical Specifications.
O 13.5.6 Technical Procedures
)
Technical Procedures include the procedures necessary to provide periodic maintenance, calibration, and testing of plant instru-mentation and components. These procedures have provisions for meeting surveillance schedules and for ensuring that measurement accuracies are adequate to keep parameters within operational and i safety limits. Procedures for these tests and the control of measuring and test equipment used in conducting these' tests are prepared in accordance with Regulatory Guide 1.33, Appendix A. 2l The Superintendent - Technical Engineering and the Superintendent
- Maintenance and Modification are responsible for these procedures.
13.5.7 Reactor Enoineerino Procedures Reactor Engineering Procedures provide guidance for activities ; associated with fuel and core management and nuclear performance 2 evaluation. The Principal Engineer - Reactor is responsible for these procedures. 13.5.8 Radiation Protection Procedures I Radiation Protection Procedures describe the methods for person-nel exposure control and monitoring; area radiation surveys; O portable radiation surveys; portable radiation-monitoring )i! 13.5-4 REV 2 3/89
cm ca.i.e. m <?ev; ' FERM12 UFSAR R O OF 17 ' 17.2 OUALITY ASSURANCE PROGRAM POR PLANT OPERATION The Detroit Edison Company (Edison) operational quality assurance (QA) program is based on American National Standards Institute (ANSI) Standard N18.7-1976, " Administrative Controls and Quality Assurance for the operational Phase of Nuclear Power Plants," as modified by Regulatory Guide 1.33 as addressed in Appendia A of the UFSAR. The program is structured and implemented in accord-ance with the guidance of the ANSI standards referenced therein and the associated regulatory guides that endorse them. Compli-ance with this guidance ensures a comprehensive QA program and an effective implementation of that program for compliance with the requirements of Appendix B to 10 CFR 50. 17.2.1 Orsanization The organizational structure, responsibilities, authorities, and functions of the nuclear organisation (Nuclear Generation) are l2 described in this subsection. Those' corporate organisational units that support the operation and maintenance of the plant and perform activities subject to the requirements of the QA program are also described. Those organisational units
! include Purchasing and Inspection, and are discussed in Sub-section 17.2.1.5.
l The Edison corporate organization is described in Subsection
- 13.1.1. That portion of the corporate organisation that is involved with activities subject to the QA program is shown in .
I Figure 17.2-1. 17.2.1.1 Senior Vice President - Nuclear Generation The Senior Vice President - Nuclear Generation has the ultimate management authority for establishing QA policy. He reports , directly to the Chairman of the Board and Chief Executive ! Officer. The authority and responsibilities of the Senior Vice President - Nuclear Generation are discussed in Subsec- 2 tion 13.1.1. He has the overall responsibility for the implementation of the QA program by Nuclear Generation. He is assisted by the Vice President - Nuclear Operations; the Vice ! President - Nuclear Engineering and Services; the Director.- Nuclear Quality Assurance and Plant Safety; and the Nuclear l2 Safety Review Group (NSRG) Chairman. 17.2.1.2 Vice President - Nuclear Operations The duties and responsibilities of the Vice President - Nuclear Operations are described in more detail in Subsection 13.1.2. The Vice President - Nuclear Operations is supported by the Plant I l Mana rod etion ene the Director - Nuclear Training. tut:/ge.' -Add Nuclear. .d ' d /2 MN*a- apWMA .
- O 1
17.2-1 REV 2 3/89 l
il Si- Ipp. pq p
', 8 9 oF l '7 17.2.1.2.1 Plant Manaaer - Nuclear Production
I The Plant Manager - Nuclear Production (Plant Manager) is - I responsible for the operation, maintenance, modification, outage j management, radiological protection, security, and plant g, r,# " l 4 administration of Fermi 2 and for the _ implementation oL. ' M7x " t quality-related procedures.'FA detailed'Bescrip M n 6f the plant 4 i organisation, and qualifications includingfor all responsibilities, key staff positions, authorities, is given induties,Fp #/ # $j Subsection 13.1.2. The Plant Manager is the chairman of the . g. onsite Review organisation (OSRO). In matters involving v, , l implementation of the QA program for the operation of the plant, 1l he is supported by the Supervisor - Production Quality Assurance and his staff. f 17.2.1.2.2 Director - Nuclear Trainina 1 The Director - Nuclear Training is responsible for developing and implementing training programs in support of the safe and ' efficient operation of the plant. f The training program is described in Section 13.2. 17.2.1.3 Vice President - Nuclear Enaineerine and Services The Vice President - Nuclear Engineering and Services is responsible for plant engineering, nuclear technology, O radiological emergency response preparedness, nuclear fuel, computer services, information systems, nuclear procurement, administrative services, and licensing in support of the plant
-)
operation. Reporting to the Vice President - Nuclear Engineering and Services are the Director - Nuclear Licensing; General Director - Nuclear Engineering; the Director - Nuclear Services; 2 the Gen?ral Supervisor - Nuclear Fuel; Director - Nuclear Materia.Ls Management; and the Supervisor - Radiological Emergency Response Preparedness. 17.2.1.3.1 Director - Nuclear Licensina The Director - Nuclear Licensing reports to the Vice President - 2 Engineering and Services and is responsible for_ nuclear licensing activities and for ensuring compliance with regulatory require-ments. The Director or the operating authority is responsible for communications with the NRC Regional Office on reportable deficiencies for activities covered by the Nuclear QA program. 17.2.1.3.2 General Director - Nuclear Encineerine 1 l The General Director - Nuclear Engineering reports to the Vice President - Engineering and Services and is responsible for design engineering and nuclear engineering including plant engineering, nuclear technology, and qualifications engineering l ! in support of plant operations. Nuclear Quality Assurance 17.2-2 REV 2 3/89
.-y . - , . . - . . , . . - . . , - . _ , , , . ~ _ , , ~ . . .m.m-.,__m.-., , , - __ - _ . . _ - _ . . _ . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -
ME 61-INL. prs bp I J. i FERUI 2 UFSAR P, io op f 1
- ,
{. 13.1.2.2.4 Director - Nuclear Materials Manacement The Director - Nuclear Materials Management provides rocurement
- services for all materials, supplies, tools, and serv ces; i reviews documentation to ensure proper information for safety I related or important to safety applicatione is specified on the procurement documents; and receives, stores, and distributes 2 materials and supplies. Reporting to the Director are the
- General Supervisor - Nuclear Materials, Supervisor - Material l Engineering, and Supervisor - Nuclear Procurement. ;
! 13.1.2.2.5 General sueervisor - Nuclear Fuel
- l 1
l The General Supervisor - Nuclear Fuel is responsible for fuel l cycle analysis, uranium and enrichment programming, core '
- analysis, reactor dynamics, fuel design, fuel fabrication contract administration, fuel assurance, fuel storage and l shipping, fuel performance, fuel accountability, control rod
) analysis, and fuel burnup. l
;
sunervisor - Radiolonical Emereenev Reanonne
~
13.1.2.2.6 Prenaredness l 2 ^
The Supervisor - Radiological Emergency Response Preparedness j is responsible for the onsite emergency response planning and
- coordination with offsite emergency response organisations.
13.1.2.3 Vice President - Nuclear Operations ' ! The Vice President - Nuclear Operations is responsible for the '; operation, maintenance, modification, training, outage manage-ment, security, and plant administration of nuclear power , plants. Reporting to the Vice President - Nuclear Operations are the Plant Manager - Nuclear Production, the Director - Nuclear Training, and the Assistant to the Vice President - Nuclear 2 Operations. 13.1.2.3.1 Director - Nuclear Trainino , l The Director - Nuclear Training is responsible for developing and i implementing training programs in support of safe and efficient l operation of the plant. j The training program is described in Section 13.2. ; I 13.1.2.3.2 Plant Manaaer - Nuclear Production The Plant Manager - Nuclear Production is responsible for the , operation, maintenance, modification, outage management, radio- : logical protection, security, and plant administration of Fermi 2. The Plant Manager shall delegate in writing the succes- ' i sion t his responsibility during any absence. C Off!W Wh/tScotfims &// be- Ib' Mgdir%%s ycoup wMed Rspe/ka&ed b't a rs n N r MstsM
& Tu s4cc Azeuckni; Alack.u Mrurion '
13.1-7 REV 2 3/89 4 ,
(42 64 - IGG- UF6 kV @
- .
- FERM12 UFSAR P, il op 37 '
. Reporting to the Plant Manager - Nuclear Production are the Superintendent - Operations, the Superintendent - Maintenance
(~') (_ and Modifications, the Director - Nuclear Security, the Superintendent - Radiation Protection, the Superintendent -
-)
Technical Engineering, the Supervisor - Planning and Scheduling, 2 and the Assistant to the Plant Manager. The Radiation Protection ! Manager reports directly to the Plant Manager regarding radiological control (see Subsection 13.1.2.3.2.5). , In addition to Fermi 2 responsibilities, the Plant Manager l will provide support for the decommissioned Fermi 1 facility, including the reactor, the fuel and repair building, and the 2 sodium storage and treatment building. This facility should l take no more than 1 percent of the Plant Manager's time. The Superintendent - Radiation Protection will perform the surveil-l lance requirements of the Fermi 1 Technical Specifications. On the average, these will not require more than 100 work-hours of effort each year. The plant staff is organized into sections under the direction of the Plant Manager - Nuclear Production. Figures 13.1-3 and ' 13.1-4 are the Nuclear Production organisation charts; each 2 classification in the figures represents a job position for one or more individuals. The functior . responsibilities, and authorities of plant personnel 4 - Jescribed below. 13.1.2.3.2.1 Suoerintendent c. cations The Operations Section is ress- ole for the operation of plant } equipment and systems inclu? idwaste and plant chemistry. The Superintendent - Operat. As responsible for the activities of this section. The Superintendent - Operations exercises overall managerial and supervisory responsibility for the startup testing and safe, reliable, and efficient operation of the plant and all associated 2l equipment. It is the Superintendent's responsibility to have a staff of trained and properly licensed personnel to accomplish the various plant responsibilities and to ensure that qualified-personnel are available to fill the plant complement positions. Prior to operation, the Superintendent - Operations was respon-2 sible for the Fermi 2 startup and testing activities. The Operations Section is shown in Figure 13.1-4. All operations, testing, or maintenance work must be approved either by the Superintendent or by an assigned delegate, as 2l established in procedures in the Plant Operations Manual. In the absence of the Superintendent, the succession to the Superintendent's responsibilities is specified by procedures in 2 the Plant Operations Manual. Reporting to the Superintendent are the Operations Engineer, the Os Operations Support Engineer, the General Supervisor - Radwaste, J 13.1-8 REV 2 3/89
lA.A 6 % % 4EV D FERMI 2 UFSAR P, q g g and coordination of refueling. The Operations Support Engineer 4
) '
l shall have an SRO license. The General Supervisor - Radwaste is responsible for the supervi-sion, both technical and operational, of plant activities involv-ing radioactive waste processing. The Radwaste Supervisor - 1 2 Operations reports to the General Supervisor - Radwaste. i The Startup Engineer - Test Phase was functionally responsible for the checkout and initial operation and preoperational testing and the ascension-to-power testing programs. 2l The General Supervisor - Chemistry is responsible for directing the sampling of plant fluid systems, for the chemical laboratory, for prescribing the procedures to be followed for sample preparation and analysis and result reports, and for environ-mental regulatory compliance. The General Supervisor - Chemistry 2 evaluates results, reports, and laboratory techniques. The ' i Chemical Engineer, the Chemist, and an engineer report to the
- General Supervisor - Chemistry. -
l The Chemical Engineer is responsible for the operation, mainte-
=
nance, and calibration of the plant chemical processing and water treatment equipment. 2l . The Chemist is responsible for maintaining the radiological and
- chemical parameters of the plant within the requirements of the ,)
l Technical Specifications and the Radiological Effluent Technical " 4 Specifications. l 2 13.1.2.3.2.2 Superintendent - Maintenance and Modifications 2 The Superintendent - Maintenance and Modifications is respon-sible for the maintenance of the plant and all associated equipment and the modification to plant structures, systems, and equipment. Reporting to the Superintendent - Maintenance ervisor - Mod &f4eet4ene, MC""" and Modifications the General are the Supervisor General Sup/ Instrument and Control,t**^' h 's
- Meintenen;;
-the Super-visor - Maintenance-Gupper+r-tM 0;;;r:1 T;r;=; W#
m intunanuw, th: Ocner41-Foseman - Sh:p:, nd th; Tr;; ntive NW/fCA,M 2 M/4 Malatenance Cecidinstcr.- O'^'M" " The Superintendent - Maintenance and Modification is responsible for the maintenance of plant structures, systems, and equipment. In this capacity, compliance with the Technical Specifications related to maintenance, written procedures, and work practices is ensured; and duties include the lubrication program, the preven-tive maintenance program, instrument spare parts, routine cali-l bration, instrument and control troubleshooting, and the standards calibration program. ,'
/et,a '^, M & 2 .a i u , ', :" . d( l- Tey..Liew K 0:nstal-Gupervisor b Medific:tichgrevides n.p..;; sib \
and direcW= erh of maint-ence L. _-r 'includi-- ' J r^ T ' ' ~ Tw a't r0 0 tO r ) si; ;UN;r OQC ,,tO th0 pl*"t-s Th8 \
^#
5 ~ -rm
-,,9,,d ij i~g2 t i_ -,q x -
( 13.1-10 REV 2 3/89
lCC 8% DZ2- VFS QGV $ FERMl 2 UFSAR P,1$ op 37 the Startup Engineer - Test Phase, and the General Supervisor - () Chemistry. The Operations Engineer is responsible for the overall operations l of the plant equipment and systems. As designated in procedures in the Flant Operations Manual, the Operations Engineer or a j delegate approves written work requests for equipment operation,. maintenance, or tests. During the Operations En ineer's absence, , the Operations Support Engineer may assume the d ties and respon-i sibilities of that position. The operations Engineer shall have !
; a senior. reactor operator's (SRO) license. Reporting to the
- Operations Engineer is the Nuclear Shift Supervisor and shift i organization. I 1 l i' l The Nuclear Shift Supervisor (s) is responsible for and exercises
- supervisory control over the operating personnel on shift. ,
During the absence of the Superintendent - Operations, the Opera- i
- tions Engineer, and the Operations Support Engineer, the Nuclear ;
Shift Supervisor is responsible for the overall operation of the l
- plant. !
Each Nuclear Shift Supervisor shall.have an SRO license. Report-
ing to the Nuclear Shift Supervisor is the Nuclear Assistant Shift Supervisor. l l The Nuclear Assistant Shift Supervisor assists the Nuclear Shift 4 l Supervisor in duties as directed. Each Nuclear Assistant Shift 4 Supervisor shall have an SRO license. Reporting to the Nuclear l i Assistant Shift Supervisor are the Nuclear Supervising Operators. The Nuclear Supervising Operators manipulate the reactor controls I and other controls and direct the activities of the Nuclear Pnwer < Plant Operators and Nuclear Assistant Power Plant @ rators. Normally, two Nuclear Supervising Operators are assigned to each ' operating shift. Each Nuclear Su reactor operator's (RO) license. pervising Operator-shall have a The Nuclear Power Plant Operators are responsible, under the direction of the Nuclear Shift Supervisor,-the Nuclear Assistant Shift Supervisor, or the Nuclear Supervising Operators, for oper-ating auxiliary systems and.for assisting in the refueling of the ! plant as directed. Among their regular duties are the opration l of such plant equipment as pumps, turbine generator auxiliaries, blowers, radwaste systems, compressors, and auxiliary service
!. equipment. Additional-dutiss include radiation monitoring, ; . recordkeeping, and general housekeeping.
The Nuclear Assistant Power Plant Operators. assist the Nuclear j Power Plant Operators in duties outside the main control room. , The Operations Support Engineer is responsible for providing sup ~ ; i port to the operating organization, including procedure-prepara- '
/ ) tion and revision, in-plant training, corrective action support, ;
t i - 13.1-9 REV 2 3/89l. e
3 l <>u m SPJ v~w w-am pt pu7 % dic Mt7tf,ccypm N AWaA M sfc .:r.: ire /- T 'p FEnus a ursAR ite on.az.m 8%c $2acar.cs, 4 80'4 P. i + on 17 li
- s. m w a ig n g eneral-supervisor - s ein s;ti:n; le 4 *o responsible for 5, M)ceesdineting::iniunencetraini..s. P pertin; t; th; C: Oral .
y hpervisnr M^difi : tion; ce; th; A::ict;nt C;n;r:1 For:::a ~ 2 b Modificetiene, ihe : int;;;nce ;:ntracter, the N;;l;;r " tiding , 4
)4'(y ";;;ialiet, end the ri;1d :sperint: dente. 1 g W i The General Supervisor --Maintens ;;/ Instrument and Control is Wresponsible for repair, maintenance, and calibration of the S g Ninstrument and control devices located in the plantgf Reporting u t to the General Supervisor - Maint :::::/ Instrument and Control 4e 2 w
s f the C;;wrela renwsoes
;;.;n - Nuclear Instrument and Control.
wm (fTheGeneralWrikaF4aintenanceisresponsibleforensuring M fy s k that-elect @ el-and mechanical maintenance is performed in an expeditious and efficient nnel ace g g to procedures. Reporting tg.t h Ggn r 1 . 8ErY- htihance are the doe 4etant 2 i GO Oral-F8teman fdMa n e Ence, Meint;nenc; Per? ?: - ?lt:t . h E1::trie:1 : d P.e henical, P. int:nc : Te;;;;n - Plant Shops, , k For:ran - Muclear-het-rement end Cenisel M;trele;y, and the (' s'g . \) Nuclear Tools Supervisor p 74 wt2//9 5ymeme ja.w/M Gced 3vms/n=% a I wnr .~nr;- h General. Foreman - Puclear Instguqent and Cont 91 is responsi- l1 i bleN for planning and dixocting the day,-to-day activi, ties pf'the instrument repairmen. These activities, include issuln orders to the work crew and ensuring that the work isgdone workin accordance with approved procedures. Responsibilit'ies also include keeping maintenance records on the' equipment assigped
/the Maintenance Instrument and' Control group'. Reporting to.,the Deneral Foreman'- Nuclear Instrument and C'ontrol is the Foreman -
Nuhear Instrument'and Control. N / 4 'N N,/ N ,/ _ The Foreman - Nuclear Instrumentrand Control is responsible' for'directin Repairmen. \g the day-to-day mork of the Nuclear Instrument
/ y -
N ,/ - The Supervisor 5 Maintenance support is responsible for provid-s - N l2 ing technical service and operating support for all phases of s\
' maintenance N ,e activity y e-N
's
, s x y'/ N The Preventive Maintenance Coordinator is responsidle for the administration and controlxof the spare parts, preventive 2 maintenanh , Nuclear Plant Reliability Data System (NPRDS), and epvi'ronment4 gualificationdrqqrams in .the maintenance area of efesponsibility,to help ensnre plant sa.fety. \ \
13.1.2.3.2.3 Director - Nuclear Security The Director - Nuclear Security is responsible for the physical- ! security of Edison nuclear power plants and the facilities, material, equipment, and construction associated with them. The physical security responsibility includes developing and imple-menting the Physical Security Plan, the Safeguards Contingency O Plan, the Security Personnel Training and Qualifications Plan, the Safeguards Information Protection Program, and security 13.1-11 REV 2 3/89
I LM BR-I n vFs
~ "
FERMI 2 UFSAR policy; conducting personnel screening for all personnel requir- ! ing unescorted access into the protected area; and-implementing ) i the access authorization and fitness for duty programs. The Director - Nuclear Security also conducts investigations or initiates investigations of attempts to breach nuclear security, whether committed by a person employed at Fermi or a member of the public.
! 13.1.2.3.2.4 Suoerintendent - Radiation Protection The Superintendent - Radiation Protection is responsible for l
the admini'stration and supervision of the Radiation Protection Section. This group is responsible for radiological engineering, l
health physics, radiation protection, radiological effluents, Radiation
- and ALARA programs.' Reporting to the Superintendent
Protection is the Radiation Protection Manager. i ! 13.1.2.3.2.5 Radiation Protection Manacer .
-t 1 .
The Radiation Protection Manager has the responsibility and j authority to formulate and administer plant programs and proce-dures which ensure radiation protection for plant personnel, ' members of the public, and the environment. This position receives delegated authority from the Plant Manager in the area of radiological control, which includes radiation protection, qp, radioactive effluents, radioactive waste disposal, and radio-logical health. The Radiation Protection Staff reports to the i l Qgy Radiation Protection Manager. ') The Radiation Protection Staff is responsible for all operational health physics activities, such as the placement of barrier ropes and signs; dosimetry; and the health physics instrumentation program, including calibration, repair,.and routine maintenance. The Staff is also responsible for (1) interpreting regulations and commitments related to radiation protection and radioactive effluents control and for developing and maintaining program documents to ensure that applicable requirements are satisfied and (2) for assessing, developing, and implementing methods and technology to maintain radiation exposure and radioactive effluents as low as reasonably achievable. 13.1.2.3.2.6 Suoerintendent - Technical Encineerina The Superintendent - Technical Engineering is responsible for the onsite engineering related to the operation of the plant. In this capacity, duties include the evaluation and reporting of equipment performance, performance testing, surveillance testing, technical and administrative pr'ocedures, and computer maintenance interfacing. Reporting to the Superintendent - ! Technical Engineering are the Principal Engineer - Process Nuclear Computer System, the Program Manager - Inservice l{} Inspection and Performance Evaluation, the Principal Engineer - Plant Systems, and the Principal Engineer - Reactor. ); L 13.1-12 REV 2 3/89
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- ( ]Yes [ ] No LCR No
) E) Priority STATUS l NRC approval required by (date): a . , _ < An [ ] Emergency I lExiaent condition will occur if not approved by: Abtl LLucz.v j (State date): ARMS INFoRMATioN myntrut: ru ev NM M - OTC ..,,,hg,-f s,. . ,, o y,,,j,g. jgg,qfg3 REV Explanation PAGt.. .REV , p;s F) Implementation SEP 2 6198947Jgj i p............................PART3:
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- 82073 J CHAPTER 13: CONDUCT OF OPERATIONS This chapter describes the framework within which Fermi 2 will be operated. It summarises the organisational structsre, the training program, review and audit procedures, plant procedures,
~
plant records,.and industrial security. 13.1- ORGANIZATIONAL STRUCTURE . .: l The Detroit Edison Company (Edison) is providing the operational 'l l organization for operating Fermi 2. This chapter describes the ' Edison organisation as it pertains to testing and operation of' l the plant. < l incorporated and ! Edison is an investor-owned public utility,d sale of electrical engaged in the generation, transmission, an I , energy in the State of Michigan. l Edison has had considerable experience in designing, construct-ing, and operating fossil-fueled facilities-for generating elec-tricity. Normally, the design engineering effort and the con-struction management for such facilities have been performed by. Edison personnel with-the assistance of design.and construction
! contractors. Such contractors-are guided and directed in their work by the responsible divisions that report through the Edison Corporate organization. The corporate' functions, responsibil-ities, and authorities related to Fermi 2 are described in Sub-
sections 13.1.1, 13.1.2, and 13.1.6. Figures 13.1-1 through 13.1-4 show the Corperate organisation; the Nuclear Generation 2 organizations and the Nuclear Production (plant) organization, including the operations. organization. , l - 13.1.1 Corporate Oraanization
~
Chairman of the Board and Chief Executive Officer The Chairman of the Board.and Chief Executive Officer, subject to the control of the Board of Directors, has general charge of Edison's usiness and affairs. g:y:x M... gtJ nd Chief Operating officer e Vice President _r_ .. ..nr e mad the Senier 2 l V4ce-f sident % ci::: C..setin report to the Chairman. See g
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j' -) I L 1 . 1 . 1- President and Chief Operatina Officer M
/3 . / . / . O, The President and Chief Operating Officer, subject to-the control i of the Board of Directors and the Chairman of the Board, is in-
' charge of Edison's mennee6eer engineering and operational affairs. The individuals reporting to the President who have responsibilities ralgt to Fermi 2 are9the Senior.Vice.Presi-dent - Power Supply the, Senior Vice President-(fiivisions and 2 Administration d/h M h.,4-WMW #"-[17 g !( '"' " D 1 l- 13.1-1 REV 2 3/89 i
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. Fanui UFSAR REY o R S ov e 1
/6. /. / . J. / [
-lbl . l . l .1- Senior Vice President - Power Sucolv The Senior Vice President - Power Supply is responsible for effectivelyandcompetitivelyprovidingfullpwersufplyto large users and for buying, marketing, and selling bu k power.
Be is also responsible for operation and maintenance of fossil fuel power generation facilities, including fossil fuel supply. Reporting to the Senior Vice President are the Vice President - Power Generation, Director - Environmental Protection, Manager - Power Supply Transactions, and the Director - Business Planning.
/8././ A././ p A
-13.1.1.1,424- Vice Presidenf - Power Generatfo_n The Vice President - Power Generation is responsible for ensuring l that nonnuclear power generation facilities are available to operate as required and to operate efficiently. He is respon-sible for ensuring that the operation and maintenance is accom-plished in a safe manner within legal, corporate, and contractual guidelines. Reporting to the Vice President - Power Generation 2 are the Director - Engineering Research and;the Director - Design l
Engineering. (Refer to Subsection 13.1.1.f for a description of the responsibilities of these two individuals.) -
/S./ /.d./.O- M 4414,1.1<1. Manacer - Power Sucolv Transactions The Manager - Power Supply Transactions is responsible for '
O the planning, marketing, and operations associated with bulk power transactions. Reporting to the Manager - Power Supply Transactions is the Director - Operations.
- 13. t . /. 9. :2. /
j 3.1.1r172- Senior Vice President- fDivisions and AdministrationM
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i - Reporting to the Senior Vice President-(Divisions and Administra-
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tion) is the Vice President - Administration.
/3in.D.D./
E4.1.1.1.3 Vice President - Administration [ The Vice President - Administration has the overall responsibil-l ity for Edison's administrative services and materials procure- . ment and management. Reporting to the Vice President - Admin-istration are the Director - General Purchasing and the General Superintendent - Stores and Transportation. (Refer to Subsection 13.1.1.38for a description of the responsibilities of these individuals.) IS. t . t . t 3ri rir2
&kw md Vice President - 5:10>e Relatione
/Ja w w<a.uw 0 l
h The Vice President - Employe4 elations has the overall responsi-i de T bility for corporate policies and practices regarding human
/>4 ) resources (including the employment, development, compensation, g' and for related deal-
/ f./(;andutilizationofemployesatalllevels) ings with unions and government agencies. Reporting to the Vice l llI ,
13.1-2 REV 2 3/89 i
, _ , . _ _ . . _ _ . , . , , - , , , , . _ _ , , , . , ..,,_g,, , ,%,. . . , . , , ,_..%. p.,.9g-.-,y_,,,w
LU2 69-14,-URS ' 4 FERMI 2 UFSAR President uw & :l 7 5dictic:h are the-Manager - Organisation
-O Planning and Development and Manager - Personnel Services. (For .
a description of the responsibilities of the Manager - Organi-
- f. sation, Planning and Development, and the Manager - P'ersonnel
/S'~' Services, refer to subsection 13.1.1.4.)
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-13.1.1.3 Senior Vice President - Nuclear Generation 2 i The Senior Vice President - Nuclear Generation is responsible ,
for the overall plant safety, including operation, maintenance, modification, quality assurance, training, security, outage f management, engineering, and administration of Edison nuclear
- power plants. Refer to subsection 13.1.2 for the detailed description of the Nuclear Generation Organisation.
/S. /. ) < 3
-13.1.1.4 Coroorat'e Oraanization - Nuclear Generation Suonort Assisting Nuclear Generation are several service grou's p that provide technical and services support as requested by Nuclear Generation. The responsibilities of these groups and their interrelationships with Nuclear Generation are described in the following paragraphs. These reporting chains are shown in Figure 13.1-1.
, / 3. f . I . S . I M ! -13 .1.-l .-4 . Director - Desian Encineerino l The Director - Design Engineering reports to the Vice President - Power Generation and is responsible for providing quality engi- 2-1 neering designs and the project management required to build or modify systems and structures for Power Generation and other-i organizational units as requested in a cost effective and timely' , l manner.
- /2 1 I . b . R
-13.1.1.4.4 Director - Encineerino Research 4 The Director - Engineering Research reports to the Vice Presi-dent - Power Generation and is responsible for providing laboratory, analytical, and technical services and some 2
experimental capability to ensure the safe and economical operation of. power plants and parts of the electrical system. '
/ 6, / . t . .p . B ,
-13.1.1.473 Director - Environmental Protection k i The Director - Environmental Protection reports to the Senior Vice President - Power Supply and is responsible to favorably influence the development and enactment of environmental laws and regulations; factor regulatory, technical, and managerial 2 considerations into cost effective compliance programs; and ensure that company facilities are operating iri compliance with applicable laws and regulations.
O 13.1-3 REV 2 3/89 i
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43.1.1.6.6 Director - General Purchasino K O t The Director - General Purchasing reports to the Vice President - W Administration and is responsible for Edison procurement of equipment, materials, and contract services. The Inspection 2 Division, reporting to the Director - General Purchasing, is responsible for determining that the quality of purchased materials and equipment meets the applicable purchase specifica-tions. These determinations normally are made from information obtained by inspection of the vendor's products and facilities. The work of the Inspection Division is coordinated through 2l Nuclear Quality Assurance by the Supervisor - Nuclear Procurement Quality Assurance.
/B.I.t.8.o~ g 2l ;;.l.1.4.5 General Suoerintendent - Stores and Transoortation The General Superintendent - Stores and Transportation reports to-the Vice President - Administration and is responsible for Edison control of inventory of stock materials, receiving, warehousing, and disbursing, reclamation and salvage of all materials and equipment, and maintenance of the motor vehicle fleet. >
/ S. / . # . 3,6
;'
l -10.1.1. 4. 6- Manacer - Personnel Services g Man l _-d $ agpfeead , w W and is responsible for Edison policies on allP rsonnel Services ld repor employment functions, employe and recreational services, person-i l l
\ nel data systems, medical and safety services, and affirmative action programs. Reporting to the Manager - Personnel Services g
i is the Medical Staff - Director. The Medical Staff - Director is responsible for providing or i securing first-aid and medical treatment for occupational and ) nonoccupational injuries or illnesses and for conducting pre-l- employment and periodic physical examinations in accordance with ANSI N546-1976. Medical services are coordinated through the l Director - Nuclear Services. (
/3. t.1. 3 . 9 2 -1&-lvlv4. 7 - Manace r - Orca niza tion , Plannino, and Development [
The Manager - Organizatiegy: g ,y g g ji Development reports to the Vice President ..r. the Edison policies on education and management development,
.___...... and is responsible for N psychological services, policy and benefits, wage and salary i
services, and employe training (limited Fermi 2). Reporting to the Manager - Organization, Planning, and Development is the ',
Director - Psychological Services. The Director - Psychological Services is responsible for perform- , ing psychological screening processes for Fermi 2 personnel including testing and programs on behavior reliability. , i 13.1-4 REV 2 3/89 i T
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.' ' P, 7 o p 8 17.2 00ALITY ASSURANCE PRN"-9A4 POR Pgu_ T OPERATION The Detroit Edison Company (Edison) oprational quali.ty assurance p (QA) program is based on American National Standards Institute
' (ANSI) Standard N18.7-1976, ' Administrative Controls and Quality
Assurance for-the Operational Phase of Nuclear Power Plants," as 1 modified by Regulatory Guide 1.33 as addressed in-Appendix A of the UFSAR. The program is structured and implemented in accord-ance with the guidance of the ANSI standards referenced therein l and the associated regulatory guides that endorse them. Compli-ance with this guidance ensures a comprehensive QA program and an effective implementation of that program for compliance with the requirements of Appendix B to 10 CFR 50. l 17.2.1 oraanization , The organizational structure,' responsibilities, authorities, and i functions of the nuclear organization (Nuclear Generation) are l2 j described in this subsection. Those corporate organisational units that support the operation and maintenance of the plant and perform activities subject to the requirements of the QA program are also described. Those organizational units
l include Purchasing and Inspection, and-are discussed in Sub-l section 17.2.1.5. i ! The Edison corporate organisation is described in Subsection l 13.1.1. That portion of the corporate organization that is ! involved with activities subject to the QA program is shown in Figure 17.2-1. 17.2.1.1 Senior Vice President - Nuclear Generation- 2 The Senior Vice Presidtnt - Nuclear Generation hasHethe ultimate repor s management authority g g. g-y).shing gpolicy. Chief *---"--(.4tbade di.;;;1y to the "'-'----
---- A officer. The authority and responsibilities of the Senior : <
Vice President - Nuclear Generation are discussed in Subsec- 2 tion 13.1.1. He has the overall responsibility for the implementation of the QA program by Nuclear Generation. He is assisted by the Vice President - Nuclear Operations; the Vice President - Nuclear Engineering and Services;'the Director - Nuclear Quality Assurance and Plant Safety; and the Nuclear l2 Safety Review Group (NSRG) Chairman. r 17.2.1.2 Vice President - Nuclear Operations The duties-and responsibilities of the Vice President - Nuclear operations are described in more detail in Subsection 13.1.2. The Vice President - Nuclear Operations is supported by the Plant l Manager - Nuclear Production and the Director - Nuclear Training. 1 I t-1 i 17.2-1 REV 2 3/89
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***** PART 2: OPEAATING LICENSE CHANGES I X]NA '
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- 00) NSRO Approval [ ] NA crm FlP-RA2-01 Att 1 P1/1 Og2188 Date l DTC: gg File: /73g-
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l(_ A) DIcument r U%Me(s), F6gure(s), etc. Affected (Attach marked-up pages) P) 5:stion(s) Tab (,. 3 . 4
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- PART 2: +
, A) Oscument [ ] Operating License [ ] Tech Specs [ ] Environmental Protection Plan 1 Tech Spec Clarification B) li:ction(s), Table (s), Figure (s), etc. Affected (Attach marked-up pages) STATUS A btf N W I C) Reference and Source Documents Attached p [ ] Significant Hazards Consideration ggy , Q [[ ]] Enviromental impact - Categorical Exclusion vironmental Evaluation s . REV. , D)is UFSAR change required? gg g HQM LU# l LCR No.
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!- (State date): ARMS.lNFoRMATioN SYSTEMS DTc -721 crk. .. Osw Ef/- M.89-t//6 Explanation FAGE REV. .@ pgg --
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4 n2 . , Date I C) Nuclear Organlaation Unit Head D) Director, Nuclear Engineering [ INA / Date fbk Date .bM f ! ^ E) Plant Manager.[ aMMe" - t/ iu _ rs.4.n
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i i/g,#Z u4 J K =~ J4:ArfM " Date swjr/s , l F) DTrector, Nuclear Licensing Date ' 0) OSRO Approval B NA i I Date l ' lH) NSRG Approval E NA DTC: fg File: 77 _ . 1 form FlP-RA2-01 Att 1 P1/1092188 l _ _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ - . _ _ _ _ _ _ . . . . _ _ _ , _ _ . . . . . , , , . , , , , , , . _ _ _ , . . . - , . . , ,, ,.%..m,. ,
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() If an initiation signal' occurs during the test, the core spray system is signaled to start and the system returns to the automatic startup mode and is. ready to deliver water to the reactor. Similarly, LPCI pumps and valves are tested periodically during , reactor operations. With the injection valves closed and the I return line'open to the suppression pool, full-flow pumping capa- . bility is demonstrated. The injection valves are tested, and the l testable check valves are operated,'as described previously for the core spray valves. The system test conditions during reactor-shutdown are.shown on the RHR/LPCI system process diagram, Figures 6.3-14 through 6.3-16. The portion of the LPCI outside the drywell is inspected for leaks during tests.- Controls and instrumentation are tested as described in'Section 7.3. l l On receipt of an LPCI initiation signal during tests, the valves - L in the test bypass lines and in the shutdown cooling system are ' closed automatically to ensure that'the LPCI~ pump discharge is r routed properly to the reactor vessel. l Detailed specifications for ECCS component testing are contained l in Chapter 14 and the Technical specifications. l l The valves performing an isolation function between high-pressure
- and low-pressure portions of systems connected to the RCS.are
( tested in accordance with the Technical' Specifications. t Table 6.3-9 lists-the valves-that perform an isolation function ! between high-pressure and low-pressure portions of systems 4
- connected to the RCS. -These pressure isolation valves meet the i
.- requirements of the ASME Code Section XI, Pump and Valve Testing Program, and are categorized as A or AC. The testing program for ,- the valves, which is referenced in the Technical Specifications, clpipff-consists of the following methods: ,ptq
- a. Exercise the valve and verify.the valve position during *"N
i refueling and - af ter maintenance before the - return to_ ""' '"~'jd l service in accordance with IWV-3300 orL%IWV-MM$b4 ,,Q ~;522 l b. Exercise the valve (full stroke).for operability during - the cold-shutdown mode as time permits, but not more
frequently than once every 3 months f c. Measure the full-stroke time (not for check valves) l d. Leak test the valve seat before reaching power operation following refueling and after valve maintenance before the return to service. l These valves will not lue routinely exercised every 3 months i lt('T
'V during plant operation as required by IWV-3410 because of the following.
I ) y 6.3-45
-- - - - - . , . - - , - - - . , - , , - - , _ , , - _ . ~ , . - , . ~ ~ - , , , - , .,,,..,_,,,,w,,, ,ar -w--a-.-
LICBISING GRANGE REQUEST
*' IER 18191 - 1113101 - IUlFISI ,
Revision II Page 1 of MC' l
,/ ]NA meme l I m m ee m m PART 1: UFSAR. PLAN, OR ri==3mLM REVISION [
A) E-:-r:3t EDP-10166. Rev. F. HPCI Drainpot Level Switch lI
- B) Section(s), Table (s), Figure (s), etc. Affected (Attach marked-up pages) l UFSAR Fig. 7 31. Sheet 1 (See page 2 of this LCR). uut-Me i
C) Ranson for (hange EDP-10166 relocated the lower tap. E41L442 for i level switch. E41N014 from 1" drain line to the 10" drainpot. This change i needs to be reflected on UFSAR Fig. 7 3-1. Sheet 1. D) Reference and Source Doousents (Identify) 4 EDP 10166,Rev.d Test N/A i
- PDC 10166 Tech Spec N/A ABN N/A Procedure N/A ,
+
i DER N/A SE ( Attactied) 59-0073 i Effectiveness Review (Attached) N/A l Other ' ] Drawings, Design Calculations, Correspondence, etc. j leeeeeeeeeeeeeeeeeeeeFART 2: OPERATING LICENSE CHANGES [I ]NA #########l ! A) Document i [ ] Operating License [ ] Tech Specs [ ] Environmental Protection i Plan [ ] Tech Spec Clarification B) Section(s), Table (s), Figure (s), etc. Affected (Attach marked-un namen) CT ATI I4
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E) M ority NRC approval required by ( . g g ggggg,@ An [ ] Emergency [ ] Exigent condition will occur if not approved by: A (State date): Explanation t h3 Y i r hs F) Implementation , DER No. j ($' ,e j
--.m.---PART 3: APPROVALS **.. - = = = = s**===== =r = ====l 8p- l; KJ4K MMW .1Y*fC7 j j!
[ I A) Originator A. G. Reyes /S&W Date /j ID/8 j l i B) Technical Expert ; d Date E/s/s j i C) Nuclear Organization Unit Head &7 [[b b Date fd f [ " l ' D) Director, Nuclear Engineering [ lNA [. M Date [ 2 E) Plant Manager Date G .F F) Director.NuclearLicensing! _
~
Date h /4 , G) OSRO Approval [ ] NA Date { l lH) NSRG Approval [ Date NA l Forn FIP-RA2-01 Att 1 P1/1 092188 DTC: g g File: 173g- ,
l,; ;
.l
;
- LCR No. 89. Iso-UFs '
Page Z_ y % /;O
- .
a dev$ apac 4 APPEC.TED DOCUMENT :- . i
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LICENSING CHANHE RECUEST LCR 18 l 9 l- 11 l 3l 41 lU lFl Sl s' L#M de */-#f Revision 0 page7of 2
, p u..u....* *"* " * *
- PART 1 : UFSAR. PLAN. OR PROGRAM REVISION [ ]NA " * *""""""*"""*l A) Document l UFSAR
- 5) Section(s) Table (s), Figure (s), etc. Affected (Attach marked-up pages)
FIGURE 9.2-2 fmot.syg - l C) Reason for Change Removal of signal isolator per EDP-9752
- io A8 D) Reference and Source Documents (identify) l' '
EDP-9752 Tech Spec PDC-9752 Procedure . . ASN SE (MttimetN 89-0075 6 decrA > J N s.4 gy ,' g% g DER PE (Attached) po i Test M Effectiveness Review (Attached) ( JYes >l3No ?0
- Other h
g Drawings. Design Calculations. Correspondence etc. i n....u.u............* *"
- PA R T 2 : OPERATING LICENSE CHANGES I X]NA "*""*"""""""*l
,y N A) Document , g [ ] Operating License [ ] Tech Specs [ ] Environmental Protection Plan .J . e ' 1 Tech Spec Clarification ly
- 3) Doction(s), Table (s), Figure (s), etc. Affected (Attach marked-up pages) l1 l
% C) Reference and Source Documents Attached
[ ] Signific. ant Hazards Consideration [ ] Environmental Evaluation [ ] Enviromental impact / Categorical Exclusion [ ] Other ! D)is UFSAR change required? STATUS
' 1Yes [ }No LCR No 9themW AOilE @ , ,
E) l'riority ie ie .< y ASU " Nw NRC approval required by (date)d' 'e i 6 d 0; u evi} _m[ a'h ..- An [ ) Emergency [ ] Exigent coWdifi~on will occur if not approved by: * * " (State date): RI:V Explanation F) implementation , DER No. ,
,............................, ART.A,,ADm..................................................,
A) Originstor J.H.Tanianmoell [LJ ~ me Date YJS/ frf
- 5) Nnfc'aIENert os.:r[['f46 " . Date 6-7-89 )
C) Nuclear Ger eration Unit Head Y/ / Date S '-/ O D) General Director, Nuclear Engine n [ )NA !/ 4 Date 4 // E) Plant Manager Date 4 F) Other Date
#M
- 0) Director, Nuclear Licensing Date H) OSRO Approval (Tech Spec Amendments) bd NA Date {
MM V-/ 97 i ll) NSRO Approval (Operating License Amendments) N NA Date ' l l Form FIP-RA2-01 Att 1 P1/1030189 DTC: M TCLCR for UFSAR File: 1735 E _ _ ___. J6Rdul5MMUk?Annt _ _ -__. _ . . _ _ . . - . - . . ~
LCR- 134-UF5 REV O q PAGE 2 OF 2
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- LCR 18101 - 11 l 117 l -
1 Ul r l sf Revision n Page 1 of ~ ' ' " p::::::::::nm..u.PART 1: UFSAR. PLAN, OR PROGRAM REVISION [ JNA
. A) Document
"*******"*J '
. EDP - 9811 -
B) Section(s), TeWe(s), Figure (s), etc. Affected (Attach marked-up pages)
~ h ur 7.6-11
.., o C) Reason for Change Modification of Fuel Pool Water Hioh Level alam circuit reouires 3o its input removed frtxn Fue' Pool Coolina Trouble annunciator .inout.
l Table must be revised to show M9ah Water Level as havino its own(3 in. above ilN nomall setpoint, rather than beina one of several inouts to the Common alarm D) Reference and liource Documents (identify) h m EDP 9811 PDC 9811 Test Tech Spec gg g ABN DER Procedure ,)6 SE (Attached) 89-0081 l4 Effectiveness Review (Attached) Other NRC-87-0257 HED 784 .4 R Dr
..............awings, Design Calculations. Correspondence, etc.
S
^
2, ..u .* * "* * * * *
- PART 2: OPERATING LICENSE CHANGES lx ]NA """*""""*"*"l A) Document
'9% g g*g [ ] Operating License [ ] Tech Specs [ ] Environmental Protection Plan I Qw ] Tech Spec Clarification ]S y. 8) Hection(s), Table (s), Figure (s), etc. Affected (Attach marksd-up pages) 9 Q T J C) Reference and Source Documents Attached STATU5 I ..I [ ] Significant Hazards Consideration [ ] Enviromental impact - Categorical Exclusion ASB ~7QN , I [ ] Environmental Evaluation 1 [ ] Other OlST. D) is UFSAR change required? p'g
._rn a rm q.I qj pgi Mt:V.
] Yes [ ] No E) Oriority LCR No 4 (t@iI.k 1 Eo ;3 a Ud
, "'ghuhwW-l NRC approval required by (det : An [ ] Emergency [ ) Exigent condition will occur if not approved by: { (State date): l Explanation ' i F) Implementation DER No.
,............................,A,,.A,,,D<S.....................................................l A) Originator Bruce R. Johnson /Impell a mm Date S #9
- 8) Technical Expert m ,
/ .b_ ,
Date (o/9/sw
) Nuclear Organl ation Unit Head bd -
. Date i1// 7 D) Director, Nuclear Engineering [ ]NA / /
. Date 7f [#,)
E) Plant Manager MM // Date 4 -1M _F) Director, Nuclear Licensing Date f/0/fr'$f
- 0) OSRO Approval [ ] NA Date H) NSRO Approval [ ] NA Date 1 Corm F4P-RA2-01 Att 1 P1/1 Og2188 DTC: File: ; -a,
'~
.. ' FERMitursAR M E "9 * ' "~ N RE"V O N
P. 2. o s= 2
- .. TABLE 7.6-11 BPENT FUEL POOL LFAKAGE A1 ARMS IN CONTROL ROOM t.oi.e Condition .
(nominal) Fuel pool system temperature high- 130*F Fuel pool water level low 4 in. below normal level Fuel pool system trouble Any alara contacts from surge tank high or low, gate , leakage, refueliaa h11ows leakagefprX-!XMnor i - pump A or pump s assenarg oressure' low. Fuel pool water level h gh 3in,abovenormallev] ! DELETE. , ADD i . . i iO l - 1 ( l l l 1 1
~,
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1 1 x 7.6-127 i
.- +-.- ..---- - -- - - -- __-..__ - - . . , . , -e,-- - . ~ .
LI(ENSING CRANGE REQUEST LCR 18191 - 1114l01 - IUlFISI e- Revision 6 Pane 1 of 20
. I. m . m .PART 1 UFSAR. PLAN. OR FIDGRAN REVISION [ JNA m.me l g' A) Document O- gDP-9942. Rev.'8
- 5) Section(s). Table (s), Figure (s), etc. Affected (Attach ,sarked-up pages)
?
t I (See Pate 2 of this LCR) 'i C) Esason for Change EDP-9942. Rev. 5 modified the Fersi 2 ATNS ( ARI/RPT) initiation circuits and impacts the UFSAR descriptions and figures. jl r
;
D) Reference and som os Docuses;ts (Identify) EDP-9942. Rev. 6 Test U/A y, i PDC 9942 Tech Spec Not affected g ABN N/A Procedure NPP 44.040.009 & .010 l'! g$ . g : DER 58-2096 SE (Attached) 59-0075 g , g j Effectiveness Review.(Attached) N/A B Other Drawings, Design Calculatiors, Correspondence, etc.
) g jeeeeeeeeeeeeeeeeeeeePART 2: OPERATING LICENSE CRANGES [I ]NA ********#l h' A) Dooment <
[ ] Operating License [ ] Tech Specs [ ] Environmental Protection , Plan l i [ ] Tech Spec Clarification ag l B) Section(s), Table (s), Figure (n), etc. Affected ( Attach marked-up pages) 5 n. E, C) Reference and Source Documents Attached [ ] Significant Hazards Consideration STATUS : [ ] Enviromental Impact - Categorical Exclusion O .' ' ' ' , [ ] Environmental Evaluation
~
l l [ Other neTN/I" J/)487' 1 D) Is U13AR change required?
- [ ] Yes- [ ] No LCR No. RW l B) Priority L NRC approval required by (date):
l An [ ] Emergency [ ) Exigent condition will occur u noc approvec Dy i (State date): Explanation - s=aunrmrA A'il l I" R , ENi to 6 is
- d44iPGE Ra i nibi L4 la utLaibhEtanus F) Implementation * * " " ~
DER No. i = = ...= = m.,m 3 : A . ,Au,...................................i A) Originator A. G. Reyes d_ b /S&W Date 7N9 B) Technical Expert I M Date 7 20 99 _C u lea Organization Unit Head ke' Date 7!F k/f D) Director. Nuclear Engineering [ )NA,M [ [h Date M//M E) Plant Manager s Date 1-4->4 F) Director. Nuclear Licensing - A ~ Date M / h l C) OSRO Approval [ ] NA Date l lH) NSRG Approval [ NA Date l ! Fora FIP-RA2-01 Att 1 F1/1092188, DTC g g File: n g L l
e (INTINUATI(N REET u:n 18191 - 1114101 - lUlFl51 O '=c.L;i-iiii i l Revision # Page 2 of 20 l A) Document l EDP-9942. Rev. A I -j j Part 1. B) Section(s). Table (s). Figures (s), etc. Affected Sections LCR Page No. , l 5.2.2.3 3 7.6.1.18.2 4 thru 7 7.6.2.18 8
- 7.7.1.2.3.1 9 & 10 i
15.0.2 . h 11
15.8 12 & 13 Figure 7.5-1 Sht. 1 14 CI7tal .wa - l ' c ,mi - a ce7 3 Figure 7.5-1 Sht. 2 15 Figure 7.5-1 Sht. 3 16 er; , m . ,c o m . .o Figure 7.5-1 Sht. 4 17 c.T1 st. .ncol - D Figure 7 5-1 Sht. 6 18 c,m i.ny, s-Figure 7 7-3 sht. 1 19 erlaL Li n.I l Figure 7.7-3 sht. 4 A0 f. r,ni . Dion 'l i 1 l l l 0 l l Form FIP-RA2-01 Att 5 P1/1 092158 DTC: File:
L4 Ho . 89-140-LA FS FERMI 2 UFSAR kM $ l
- 3
_ 4 to Each accumu-O nitrogen to be supplied as an additional backup.lator has adequate storage capacity to an SRV at a drywell pressure of 62 psig. This provides adequate pneumatic storage to cover interruptions if the pneumatic sup-plies are switched from the normal to the emergency backup
- sources. .There is M so a separate, fully qualified pneumatic
! supply to eight non-ADS SRVs that allows these SRVs to be oper -
ated as a backup to the automatic depressurization system (ADS) I l for reactor pressure relief. This separate system is similar to the ADS-SRV pneumatic supply, except it does not include NIAS as one of its supply sources. Consideration of leakage from the accumulator and valve operator was included in sizing the , accumulator. However, normal leakage from this system is so
- small, 0.016 scfm or 0.08 percent of supply per minute, _ that it had essentially no influence on the size of the accumulator.
- , Refer to Figure 5 2-1 for a description of the prikary contain-l ment pneumatic supply system.
The NPRS automatically depressurises the NSSS suf ficiently to
- permit the LPCI and core spray systems to operate. Depressuriza-
- tion occurs when five of the SRVs are opened automatically (ADS).
i Descriptions of the operation of the automatic .depressurization j feature are found in Subsections 6. 3. 2 and 7. 3.1. The NSSS can be depressurized manually if the main condenser is , The SRVs T l h not available as a heat sink after reactor shutdown./ are operated by remote manual controls from the main control room. Controls for two of the relief valves are located on the remote control panel, and can thus be operated outside the main control room. 5.2.2.3 Report on Overpressure Protection The report entitled "Enrico Fermi Unit 2r Summar{ Technical Report of Reactor Vessel Overpressure Protection (GE Document Number 22A4070) provides sufficient information and documenta-tion to show compliance with all requirements of Article 9 of ASME B&PV Code Section III, 1968, Nuclear vessels, in the area of the vessel overpressure protection design of the Fermi 2 reactor pressure vessel. Included are the design basis for h sizing of the SRVs; the overpressure Erotection analysisr*4 e analysis of the system avm41mbilitvera 6 x e C 1 ed Asdn#Uestr witho1W sErar ..de a [ [uldidn7 r un _r p. Tse otracts in the vesseJ. pressure transients or vaAv. pacity are also shown. The overpressure protection analysis used the actual Fermi 2 scram characteristics (e.g., for BWR/4 scram and c'ontrol rod drive (CRD) systems). This report is included as Appendix 5A. 5.2.2.4 Main steam Safety / Relief Valves l The Fermi 2 valves are the Target Rock Corporation, two-stage, pilot-operated SRVs. The two-stage, pilot-operated SRVs are 5.2-14
l .) i
.m maa LCR Wo. 89 140 0F3
' w.
l Paje 4 y" ! 7.6.1 17.6 poerational Considerations l f.4.1.17.6 1 M i i i Under moraal operating conditions, control air is supplied by the
- ~ station air system and only the sentrol air dryers are in opera- !
tion. In the event of low control air headst pressure, the non- t i interruptible sontrol air system divisions are isolated from the i i interruptible air users. me control air osopressors will etart automatically prior to this isolation. .
- 7.6.1.17.4 2 Omertier 1Aferat11mn '
i meadoue. instruments are providod in the main eentrol room to dis-
play and record the Division I and II control air pressures. I Recorders register the automatic initiation of the sontrol air i I system compressors. g 7.6.1 18 Alternate Rod Insertion I 7.6.1.18 1 Boulosent Identification
- e, l
The alternate rod insertion (ARI) components of the CRD system i are designed to sitigate the potential consequences of an i anticipated transient without scram (AWS) event. The ARI components are redundant to the RPS. Sd'E 7.6.1 18.2 Eeuiosent Desien M YYNON ' [ /NS W
; , - -
w - ! l( The ARI . ogic roc vos r etor does ese and wa -level sig
> nals on the n lear iler syst as e in Fig e 7.7-3.
i he ic wil cause e immedia energ tion offme ARI vai es wh a ther e rea or high- esure t <,p setpo t or low- I ter-lev 2 set nt is re ed. 2 energi ion of the ! I alves e ressur es the se air he der inde logic d vent adent of thaf Ives of RPS. ditional ediate resIponse j - l to t initia on signals. neludes he recir lation y / motor ' gen stor fi d breaker ip (see beectio 7 7 1.2.3. . a signa to insert f ato div$sions (two- ut-of-t e conte rode is enerated in wo espa-ogic in I given divis on) and 1 results/in the eneygination ight C1 es 13 de val es. Two o ' these str /F160A of theand F 1)0A and 3, vent the,ser ekup air supply lineust down-scramvalves. Se er to F1 e 7.6-36. These ARI alves al act to blo9ai the supply of ' to the se an header.f Oneck val es F161A an 3 p!' ovide-r-flow path around the F160 valv s in the av t one or e of ( /them fails / ?tv,T addi lonal ARI alves, F162 , B, C, and , vent
/ the A and 9 scram he er to the taosphere. the het r '
depress 1ses, the ram valve , will e ing ope amming th yrods. at eachTwohyd u41c contr 1 unit [ I valves, 163A a l 3, we the ser ir header o the scram ischarge lume or n I i 7.6-75 a
LCR No. 99 140-UF.S fiW.c)
'O Pnge L. a{
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p%A MV/$N q 1 - an vent ves, osing . es iso r I ocha _ vol . All ght va e at ly one , 7.6 1.19 Safety / Relief Valves 7 4.1.19.1 Ryste_m.3dentifis_a M on The suelear pressure relief system is designed to prevent over-pressurisation of the muelear system that sould lead to the tail-ure of the reactor soolant pressure boundary. l
.. 7.6.1.19.2 Safety /Rel lief Valve Seuissent Desien Safety / relief valves (SRVs) are dual-functioning types e auto-antic self-actuating and solenoid operated.' ghe valves are self-actuated when reactor pressure exceeds spring set pressures that ,
are edjustable in range. Se R Vs are divided into three spring-set-pressure groups. The first group consists of five valves set to open een vessel pressure eseeeds 1110 psig, the second group consists of five valves set to open when vessel pressure escoeds ' l 1120 psig, and the third group consists of five valves set to I open when vessel pressure escoeds 1130 pois. The solenoid-l operated air pilot valves parait remote manual or automatic open-l Ang. The pilot valve controls the pneumatic pressure applied to ! an air cylinder operator that controle valve opening and clos-ing. Each valve associated with ADS has an secumulator to store ( pneumatic energy with sufficient espacity for several relief valve operations. The valves are espable of remote manual open-ing at any pressure above 100 psig and staying open, once opened,
) g i
, untti pressure decreases to 50 pelg. Five of the SRVs are used -
for ADS (Subsection 7.3.1.2.2). Two ERVs are used for low-low , setpoint relief (Subsection 7.6.1.19.g) . 7.61193 Initiatine circuits Reactor pressure eseeeding the setpoint actuates the SRV. The SRV can also be manually actuated (by remote manual switch) or automatically by the ADS and low-low setpoint relief logic. 7.6.1.19 4 Loeic and Seeuencine no automatie logie is involved in the overpressure safety func-
- tion of the RVt. (See Subsection 7.6.1.19 9 for low-low set-point relief logic and subsection 7 312.3 for the ADS logic.)
7.6 1.19.5 Broasses and Interlocks , Bypasses are not used in the moraal ERV funetton. An arming cir-O- eult is used as an interlock to prevent the low-low setpoint valves free prematurely actuating during normal plant operation. h interlock is required because the reopening setpoint of the low valve is near the normal reactor operating range. -
. 7.6 -76 u-____________.________._.___.__________
,y .
LCR Wo. 89 140- UF5 p5ERT A psy, p 7.5 1.18.2 a miment nesian Page.f__af 20 gnitiatina circuits There are three initiating signals used for the ARI logics, namely, o haetor done high pressure o practor low toter level 2 o Manual initistion in the Main Centrol room Anyone of the above signals een initiate the divisional ARI logics as shown in Figure 7.7-3, aht. 4. Additional immediate response to the initiation signals includes the recirculation punp motor generator field breaker trap (see subsection 7.7.1.2.3.1) . Emic two divisional ARI logic systems are prwided: Division I, consisting of' logic &annels A and C, and Division II for logic eennels B and D. Se signal to insert the control rods is generated in two separate divisions on two out ef-two logic &annels in a given division. The ARI :ogic receiws reactor dame pressure and water level signals from O. the nucleer bailer system. She logic causes autcastic energistion of the ARI solenoid valves when either the reactor high pressure trip setpoint or loMater-1wel 2 setpoint is reased. She ARI logic can also be initiated manually from the main control room. Each ARI logic channel is prwaded with a 'dinarmed/ armed' pustbutton switch. Both pustbutton switches in a given division must be depressed to energise the ARI logic and initiate control rod insertion. She ARI initiation signals are designed to sea 3-in
, the initiation logic to assure conpletion of the ARI function until it is reset manually. A reset pus) button per division is prwided in the main control room to clear the ARI logie. A tiner is uses in each of the ARI gM logic sannels to inhibit the reset function pr f38 secones after the I initiation signal is received. A 30 second time delay is selected to ensure cmpletion of the ARI function before the logic can be reset. ,t% ,
the initiation of the two separate ARI logics results in the energisation of eight Class 1E DC solenoid valves (four per division). two of these, F168h and B, vent the scram air supply line just downstream of the F118A and B backup scram valves. Refer to Figure 7.6-36. These ARI valves also act to block the supply of air to the scram header. Check valves F161A and a pewide an air-flow path around the F168 valves in the event one or sore ' of them fails. Four additional ARI valves, F1624, B, C, and D, vent the A and B scram header to the atmosphere. As the header depressurises, the scram valves at each hydraulic control unit will spring open scramming the rods. Two ARI valves: P163A and B, vent the scram air heaaer to the scram discharge volume dram and vent valves, closing these valves and isolating the scram dis &arge voltsne. All eight ARI valves are normally deenergized.
~ '
, P g DFiM5ERT h h,f 89-140-$5 prmunciatjan and Indication pagg 7 q Jg,,,
he mamm1 initiation pushbotton switch in the main control rom activates - O- an annunciator window Wenever it is placed in ' armed' position. A separate annunciator window is activated upon initiation of the ARI logic circuits. We 'open' and 'close' position of the ARI aclenoid valws are also indicated in the main control room.
- I Etatability Pour (4) separate ARI initiation logic dennels are provided to permit maintenance repair, test or calibration of all circuit devacee (at power) up to, but not including the final trip devices (ARI aclanoid valves).
Each ARI logic channel is provided with a tact $ack and indicating lights to verify logic setivation in any given division. e 4 O e e 9
L C R 2 . g aj. i 4 o . g 3 Psaastaupsan Rav.p Page.0 af "O 7.6.2.16 Rod Beauence Control 8vstes . , 7.6.2.16.1 ponformance To General Functional Reautrements The RBC8 protects against a rod drop accident that assumes the drop of the highest-worth rod that can be developed at any time by one inadvertent error by the operator. The ROM and the RSC8 are compatible and redundant to each other. 7.6.2 16 2 Sonfarmance To Specific Resulatory Reeutremertts There are no specific regulatory requirement for the 28c8. The - Fermi 2 R8C8has been designed to seat the eriteria accepted by the Regulatory Statf in their review of the Browns Ferry RSC8
-. design.
7.6.2.17 Control Air.8ystes Conformance To General Functional Reauirements The instrumentation and control of the control air system is designed to parait reliable operation and testing of sash divi-sional loop of the control air system. The control air system is designed to fulfill the safety and power generation design bases stated in subsection 7.1.2 1.32. 7.6.2.18 Alternate Rod Insertion
, Conformance To Ge_neral Functional Reautrements The sensors, transmitters, trip units, associated logic, and ARI valves are Class 1E, redundant to and diverse from the reactor protection system, are seismically and environmentally qualified to meet IEEE 323-1974 and IEEE 344-1975, and are, supplied with 1 Class 1E de power.
The ARI equipment is physically separated into two redundant divisions. Either division will be automatically energised to actuate and scram the reactor upon receipt of high reactor pres-sure or vessel low-water-level 2 signals. 7.6.2.19 Safety / Relief Valves Analysis hh _
,7.6.2.19.1 Conformance To General Functional Reautrements The SRVs furnished meet requirements of the RSME Boiler and Pressure vessel Code Section III, Article 9. The valves are operable in two modest self-actuated or power-actuated solenoid pressure relieving mode. The automatic mode is independent of the power-actuated mode. Failure of the power-actuated mode does not affect the self-actuated mode. -
_ _ _j _ n e JRZ Ayc Soy REo be. S K^ # MANf h m . ( d e e P & S h 7. Go. /. / 8 . 3 h / M A # ?; ) N - 7.6-93'~
/' panm appaAn - LCR Ho.88).140-prs fled. 9 ePART _ OF __ SECT)OQ
~ _ 7. 7.1. 2. 5.1 Pese ' 4 zo the speed setting of the speed controller for each motor-generator set converter. d The master controller signal adjusts eagh actor-generator set variable speed converter. The master controller signal is .
compared with the actual speed of the generator by the speed con-troller. The speed controller signal causes adjustment of the speed converter, resultig in a change of the cenerator speed - until the feedback from the generator equals tF.e master con- S. troller signal. A prov ton a been nelud in t Fera 2 des n to t p th recir lati pump otor-g erator field reake on re ipt eac r hi pres ce or actor ow-wa r-lev 2 in der mit ate e off ta of n anti pated ransi@e t ett ut oc am .
.( s) e ent. e tri signal is app ied to ual-b aker rip ils each enerat fiel breake . Th recir lati p rip nctio sin e-act e-fai re pr f and eo se of
. Clas it eq- pne t p to t fina break ,s. -
7.7.1.2.3.2 Motor-GeneratoI_gg,gg S MM7 MM M Bach of the two actor-generator sets and its controls are identicair therefore, only one description is given of the - actor-generator set. Figure 5.5-2 shows the general arrangement and rating of the actor-generator set. The actor-generator set can continuously supply power to the pump motor at any speed between approximately 19 percent and 96 percent of the drive motor speed. The actor-generator set is capable of starting the-
]
pump and accelerating it from standstill to the desired operating speed when the pump motor thrust bearing is fully loaded by reac-ter pressure acting on the pump shaft. The main components of the actor-generator set are
- a. Drive actor - The drive motor is an ac induction actor
; hat drives the input shaf t of the variable speed * '
converter
- b. Generator - The variable-frequency generator is driven ay the output shaft of the variable-speed converter.
During normal operation, the generator exciter is powered by the drive motor. The excitation of the generator is provided from an auxiliary source during pump startup
- c. Varuebhe-speed converter and actuation device - The var:,abhe-speed converter transfers power from the drive motor to the generator. The variable-speed converter
, actuator automatically adjusts the slip between the con-verter input shaft and output shaft as a function of the signal from the speed controller. If the speed control-O- ler signal is lost, the actuator causes the speed con- J g
verter slip to remain "as is." Manual reset of the 7.7-14
;
.... _1
;~ .
LCR No. 09-140-UFS l Rw.0 wREVI510& _ _ - To MCTIOA) 7 7.1 2. 3.] _ m_ _
, ,, j + - -- .. .. s _ _.-........ A .. # N
) pm'yi.sim be.r been inc4Wded s Nf E 'b 5
.,s axp, a .g p '~%
a f fh* y A7gs #U*k ,4cpsn4. The Peral t RPF design employs two trip oolls in each poeirculation system motor generator set t I
. field treaker. This design provides for redundant trips of both ',
actor generator sets following the transient and failure-to-soras. To -
;f aintaine the possibility of field treakers being tripped laadvertently, the automatic trip signals are arranged in two-out-of-two legte.
\, The field breaker automatic trip signal is a combined &RI/RPT logic. That g is, a low reactor vessel water level (level 8) or high reactor vessel pres-aure signal will anitiate the trip of both field breakers. (#efer te ,
fs y 7.7-3 ,isAat4.). l (
\ The RPT mag Magnug1ly in lated by the SaBe two pushbuttonsI(on a AUG Uf m ,
l t divisional tasis) as ARIj I one division will trip tioth fielddifference breakers. being that initiation of (' I i The RPT logie delays recirculation pump trip on low reactor vessel l i unter level for 9 seconds. This time delay was provided to account for L the differeneo in the pump osastdown time if the field breaker is l tripped rather than t,he 800 set drive actor, as was assumed in the LOCA analysis.The annual reset of the generator field breaker trip seal.in circuit dess seEAasa.any time delay due to the rapid operation of th's circuit breaker.
..... .. A O
u
- LCR Wo. 69-140.UES L .
penamaursan Sav. 9 PART OF 3Et7A00_--/S.O--2 fage. .l.l. N 20 ,O -
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l 15,8 ANTIC!pAft0TRAN$1ENTSWITHOUTSCRAM(ATWS) f> 16.8.1 ATWS Rule 10CFR60.62 Anticipatedtransientsaretransientsexpectedtooccurduring)the life of the plant. Anticipated transients without scram
.. those entremely low probability events in which an anticip(ATWS ated are transient occurs and is not followed by en automatic reactor shutdown (scram)whenrequired. The postuistion of the
- normal scram
- failure tn ATWs can only be deduced if more than one *aingle failure criteria *
(p is essemed. The NRC has since established the requirements to further reduce the risk to the public from such a postulated event. These requirements are specified in 10CFR50.62, ' Requirements for Reduction of Risk from Anticipated Transients Without Scram (ATWS) Events for Light Water-Cooled Nuclear Power plants" For the SWR, 10CFR50.62 re vires an k alternaterodinsertion(AR!hsystem,emanualStandbyL1 id Control System ($LCS). and an automafic recirculation pump trip ( T) function, j l The BWR Owners troup had prepared a topical report on the sub. ject of ATW$ and discussed the details of the design options how 10CFR$0.62 is )l satisfied. Reference 6 is the NRC approved topical report. Deco is a f
, member of the BWR Owners Stoup and Fermi t's design is consistent with those discussed in Reference 1. . Details of the ARI can be found l-inSubsection4.5.yand7.6.1.10. Details of the SLCS and enriched
% boron can be found in Subsection 4.5.2.4. Details of ATWSRPT can be
#2 found in Subsection 7.7.1.2.3.1.
x 15.8.2 References L. B. claasspi[and ,5. C. Bokert, 8 es et SWR D(sie[s for,-
- 1. '
Mitig,ation4f Ant l cipated'Transierfts without scram /wwo-2 26, ober U 4 y to Amene nts 'and 2. 1 Ele pany, usinera lectrit A Re t i 2 Gen
.1 rie eoo,p>.<o,uet.q.7 7p, 15.8-1 1 3/88
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! 3. s. p.'sheng,. svaluatton 4 Anticipateefransient wit'hout i'l ! Scras for A,he wanticello Nuclear Generatine plant', IE550-I [ 3910, pte A e. f
\ A eene at slecer oo. y, A. W ent of sw( wie eati l
\. Arws. Volume J , _
-2422,;t[ pecepnerj v79./
- 5. gatter from s. p. sylvia ,ln trait asinon, to Irac, Anticipated Transle.nt Without Scram (ATWS)J dated una i5,1989 (NRC-89 0140p 1--_ _ _
. G:, NEDE 31096-P-A ' Anticipated Transients without scram 3
. response to NRC ATWS rule 10CFR50.62" February 1987. ,
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- 5. THE TRIP (
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- 6. DELETEC
- 7. FORCONFIN AND INDICA PURCHASE S
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. ....:::en.nnon. PART 1: UFSAR PLAN. OR PRDOPAM REVISION I INA """""*""""*]
*- A) Document UFSAR
- 5) Bettion(s). TeWo(sL. F61ure(s), etc. Affected (Attach marked-up pages) rSGUME 9.2-12 sheet 2 t.'. m 1 m t ms .t C) Rossen for Change Removal of auto isolator N30Kt07 per EDP-9753 l
D) Reference and Source Documents (identify) EDP -9753 Tech Spec , J PDC v753 Procedure ABN SE Attacheg) AQ.nnen DER PE((Attached)' Test
- Effectiveness Review (Attached) { JYes ( Pio Other
' Drawings. Design Calculations. Correspondence. etc.
**"""""";;;;;;'s p . . . . *. . . .u . . . . . . . . . . . . . . . . P ART 2 : OPERATING LICENSE CHANGE 5 f r1NA A) Document
( ) Operating License ( ) Tech Specs ( ) Environmental Protection Plan 1 Tech Spec Clarification _ i B) Hoct6en(s). Table (s). Figure (s), etc. Affected (Attach marked-up pages) . ! C) Reference and Source Documents Attached [ ) Significant Hazards Consideration ( ) Environmental Evaluation l ( ) Enviromental Impact / Categorical Exclusion ( )Other l D) is UFSAR change required? '
' ]Yes ( ) No LCR No n A B P"N O O U U II#I31 ,
I E) f'riority 1 b r fib IV ! h et+q b, g $aayyj 5TATUS NRC approval required by (date): ._ _ _ (' ', _"_ An ( ) Emergency ( ) Exigent conilition will occur if not approved BF ' " " ' ' "
"/
ntcT (State date) Explanation REV. F) implementation . DER No. ,
,............................,AR,,. A,,,Dm,........................=.......................,
A) Originator / .s. /b tws /,r, ped Oste //st/t1 Ye ca Empert M (AAf B. d. "JA y Date 7D V C) Nuclear Generation Unit Head /M - 44 .::rw - Date 8h///) D) Generat Director. Nuclear Engineering [ NA M Date /b E) Mont Manager 1. Date p1M Date F) Other A) Director. Nuclear Licensing A Date 9//zvft H) OSRD Approval (Tech Spec Amendments) ( lNA Date
- 11) NSRO Approval (Operating License Amendments) NA Date l DTC:IB TCLCR for UFSAR File: 1735 term FIP-RA2-01 Att 1 P1/1030189
=, . - - . . - _ _ _ - _. -_ pr~.__s _t ym r~O on osa, __ __, _ ,_
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> Ltm 18 19 'l - 11 14 16 l- IU IF lS l
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) 5:ument ,
g , UFSAR 3.5.1.3 Tomado Generated Missiles W e B) Sect 6on(s), Table (s), F6gure(s), etc. Affected (Attech marked-up pages) % ? UFSAR Section 3.5.1.3 Tornado Generated Missiles
;
k i C) Reason for Change g, UFSAR sections were revised to reflect tornado missile hazard analysis performed by Sergent Si Lundy. , l D) Reference end Source Documents (identity) } g'
- y Tech Spec ;
EDP O PDC Procedure 1, J l ABN SE (Attached) 89-0094 .? K.). E v g g, DER 88-1528 PE (Attached) g) Test Effectiveness Review (Attached) [ ]Yes [ ]No * ' '
- Other Sergent Si Lundy Report No. SAD-524
._ _( _ _ Drawings, Design Calculations, Corres >ondence, etc.
_I
--g i-- - - - - - - - - - - - - - - - - - - - - - PART 2 : OPERATiftG UCENSE CHANGES [X]NA -_-_------------ -_____
4 A) Document g3 g 0" I [ ] Operating License [ ] Tech Specs [ ] Environmental Protection Plan -
] Tech Spec Clarification i
B) Mction(s), Table (s), F6gure(s), etc. Affected (Attech marked-up pages) C) Reference and Source Documents Attached , [ ] Significant Hazards Consideration [ ] Environmental Evaluation [ ] Enviromental Impact / Categorical Exclusion [ ] Other _ D) is UFSAR change required? T'" [ ] Yes ( ) No LCR No. $m ri .: is .. l E) Priority Q G g g g g ggy b JabV MTATUS NRC approval required by (date): _ , , An ( ) Emergency [ ) Exigent condition will occur if not approved by: N " ^' ? l (Statedate): mT o/I J/ /4M I Explanation ilEV. . l . cf** wu, an F) linplementation DER No. r -- -- -------------- --- ~ r xART 3 : APPROV4LE --- " "---- "-------- l A) Originator: D. G. Nick d Date MM B) Technical Expert: A.1.Hassoun I Cf nv Date 27hf C) Nuclear Generation Unit Head Date ik8i D) General Director, Nuclear Engineering [ )NA Date f f E) Plant Manager Date /o L-d _ F) Other Date
- 0) Director Nuclear Licensing 4 Date /0///M *I H) OSRO Approval (Tech Spec Amendments) [ } NA Date II) NSRO Approval (Operating License Amendments) { l NA Date l Corm FIP-RA2-01 Att 1 P1/1030189 DTC W TCtCR for UFSAR File: 1735 D1C i 11 ( R tese other
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V i 1 CONT 6NUAT60N SHEET LCR 18 19 l- 11 14 16 l- IU IF is l l TSC l l 1- 1 l l l Revision 0 Page 2 of 2 A) Document 3.5.1.3.2 Additional Analyses The missile barriers listed in Subsection 3.5.3 provide protection against tornado generated missiles; hswever, three areas received additional analysis to ensure resistance to tornado generated missiles. They are the spent fuel pool, the fan blades of the cooling towers in the residual heat r:moval(RHR) complex and the miscellaneous penetrations and openings in the exterior walls of the R3 actor / Auxiliary Building. 3.5.1.3.2.3 Exterior Wells - Reactor / Auxiliary Buildig An analysis of tornado misille hazard due to certain vulnerable areas on the exterior walls of the Reactor /Auxillery Building was performed to assess the probability of any design-basis missile penetrating the building. The analysis resulted in a calculated cumulativeper silegenetratl year. The calculated probab prebability of 1.16x10' per Year, the level interpreted by experts working in the design of nuclear power plants as acceptable for nat considering protective measures against external events. The exterior walls of the Reactor / Auxiliary Building in Fermi 2 have been designed to resist the impact of tornado-generated missiles. The missle protection adequacy of certain small penetrations and some doors 6nd HVAC Intake pciosures on the exterior walls was evaluated. The vulnerable areas consist of 51 small penetrations located on the south, west and north walls; a security door on he south wall; en HVAC Intake enclosure on the south wall; a removable precast panel on the south wall; and the equipment access door on the south wall. The analysis (Reference 16) was performed to evaluate the probability of missile perforation due to i the vulnerable areas. The tornado missile hazard associated with these penetrations and openings is limited to the penetration of missiles through the cited areas which may then impact and damage safety related items inside the buildings. For the gquipment Access Aullding door, perforation of the door is considered in the analysis. .po,w/y1 The Individual calculated probabilities are smaller thary1x10'7 per year. Their aggregate as an indicator of total plant damage probability is 1.1Bx10' . About 50% of this probability is due to a precast removable concrete panel which had been conservatively treated as an open penetration in this report. In addition, some of the penetrations are protected with steel blind flanges, however, these flanges were conservatively ignored in the analysis. Given the conservative approach f the analysis, further refinement of the calculated probability would yield a value less than 1x10' Since the conservatlyly calculated missile damage probability of 1.15x10'7 per year is very i ccmparable to 1x10' per year for considering design-basis events as described above, no further analysis is required. The occurrence of tornado missile perforation due to cited vulnerable areas is n:t considered a design-basis event. 3.5 MISSILE PROTECTION References 1
- 16. "Probabilistic Analysis of Tornado Missile Hazard Due to Some Penetration and Openings in Reactor / Auxiliary Building Walls," Sargent 81. undy Report SAD-524, February 23,19P9.
Form FIP-RA2-01 Att 5 P1/1030189 DTC: [ ] TCLCR for UFSAR File: 1735 DTC: [ ] TDLCR for other _ _ _ - _ _ _ _ _ _ _ _ _ - _ _ .--__ -_ _}}