Text

Proposed Tech Specs Re thermal-hydraulic Stability,Idle Recirculation Loop Startup & Post Accident Monitoring
	ML20199H937
Person / Time
Site:	Fermi
Issue date:	01/28/1998
From:	; DETROIT EDISON CO.
To:	;
Shared Package
ML20199H924	List: ML20199H937; NRC-98-0003, Application for Amend to License NPF-43,revising Tech Specs to Improve Administrative Controls Re thermal-hydraulic Stability & Idle Recirculation Loop Startup & Eliminating Neutron Flux Indication from TS;
References
	NUDOCS 9802050158
	Download: ML20199H937 (26)
	v • d • e

. . . ._

Enclosure 3 to -

NRC-98-0003 Page1 ENCLOSURE 3 FERMI 2 NRC DOCKET NO. 50-341

. OPERATING LICENSE NPF-43 REQUEST TO REVISE TECilNICAL SPECIFICATIONS

.- TECilNICALS]'ECIFICATION Cil ANGE Attached is a mark up of the existing Technical Specifications (TSs), indicating the proposed changes (Part 1) and a typed version of the TSs incorporating the proposed changes with a list ofincluded pages (Part 2).

+

9802050158 900128 PDR ADOCK 05000341 P PDR to ,

i NRC-98-0003 Page 2 l

I ENCL,OSURE 3 - PART I PROPOSED TECIINICAL SPECIFICATION PAGE MARK-UP

l E I.5S DELETED l

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JE l

1 LIST 0 FIG RES '- I 1

UfdE E 3.1.5 1 S0DIUM PENTA 80 RATE VOLUME /

i CONCENTRATION REQUIREMENT 3..................... 3/4 1-21 l 3.4.1.4 1 "!*".' i .*a" " ": . !^"I TLW

_ m M .................i 3/4 4 6a 2

4 3.4.6.1 1 MINIMUM REACTOR PRES 5URE VESSEL METAL TEMPERATURE l

V5. REACTOR VESSEL PRE 55URE.................... 3/4 4-11 3.4.10 1 THERMAL POWER V5. CORE FL0W. . . . . . . . . . . . . . . . . . . . 3/4 4 31 4.7.5 1 5 AMPLE PLAN 2) FOR $ NUB 8ER FUNCTIONAL TEST..... 3/4 7-21 i

8 3/4 3 1 REACTOR VESSEL WATER LEVEL..................... B 3/4 3 7 B3/4.6.2-1 I LOCAL P00L TEMPERATURE LIMIT...,............... B 3/4 6 5 f

83/4.7.3-1 ARRANGEMENT OF SH0RE BARRIER SURVEY P0!NT3..... B 3/4 7-6 i

5.1.1-1 EXCLUS10H AREA................... ............. 5-2 .

5.1.2-1 LOW POPULATION 20NE............................ 53 1

5.1.3 1 MAP DEFINING UNRESTRICTED AREAS AND SITE SOUNDARY FOR RAD 10 ACTIVE GASEOUS AND LIQUID

EFFLUENT 5...................................... 54 t

i.

FERMI - UNIT 2 xxi Amendment No. 75,f1,37.54. 77

. .. . .. . - . . - -. - - . ~ . . _ - - - . - . -

l l

l .

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TABLE 3.3.7.5-1 (Continued) f i

i E ECIDEN"151111TORillG IIISTMllENTATION

~MINilRIM APPLICABLE L e- CHA191ELS OPutATIONAL i

5 REQUIRED NUMBER 0F CHANNELS OPERABLE ColelT10h5 EllE

[ INSTRUMDif

~13. Standby Gas Treatment System Radiation Monitors SGTS-NobleGas(Low-range)I l/0PERABLE 1/UPDtABLE' I, 2, 3~ 81

a. SGTS subsystem SGTS subsystem i

-SGTS - Noble Gas (Mid-range) 1/0PERABLF. 1/0PutABLE - 1, 2, 3 81

b. SGTS subsystem SGTS subsystem SGTS - AIM-Noble Gas (Mid-range) l/0PERABLE 1/0PutABLE 1, 2, 3 81

! c. SGTS subsystem SGTS subsystem l w 2 SGTS - AXM-Noble Gas (High-ra.,ge} 1/0PERABLE 1/0PULABLE 1, 2, 3 81

'u d. SCTS subsystem SGTS subsystem

4. -

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14. Ileutesa-Fl= 2 + .

-DELETED ^ -

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15. Deleted 1/ valve i 2,3 82 F 16. Primary containment solation Valve Position 1/ valve i M OTN a..

D .= '

a.

l 2 8Also includeO in the OFFSITE DOSE CALCULATielt MANUAL.

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1ARLE 4.3.7.5-1 (Continued) .;

4 ACCIDENT MONITORING INSillUMENTAil0ll SWVtitLANCE_IltWillDEllT$

..- a m icAstr E- CIInfliiEL CluteIEL ePEnnisonnt i 14- LN3JilUMini - ttKCK CAllipl4110ll CORDIlleftS.

w

13. - Standby Gas Treatment System Radiatten Monitors

'a. SGT5 - Noble Gas (Low-range) M R I 2. 3

- b. SGIS - Noble Gas (Mid-range) _ M R I, 2, 3 !

c. SGTS - AIM-Iloble Gas (Mid-range) M R I, 2, 3 i

d. 3GIS . AIM-Hoble Gas (High-range) M R 1, 2, 3

_-w ;

14. 49esteen-fhnt- --#- -R-- .2)

DELETEb > - s

15. Deleted ) ;

16. Primary Centalament isolatt Valve Position M R 1, 2. 3 e:

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l REACTOR COOLANT SYSTEM '

10LI RICIRCULATION LOOP STARTUP D L6Ib 3 j

LIMITING CONDITION FOR OPEILATION .

l 4.1.4 An idle recirculation, loop shall_ met be_ n<arted unless WlEmiAL a0WE j

.1SSW rr:==;r;;- ;gy;2 :.; t.M dthe t 1 RfifeEDufHieeslFthe reactor prigssd' vessel 8Tlas space coolant and the bottom head drain line nelant is less than er equal.te 145Y. ands

a. When both loops have been idle, unless the temperature differential between the reactor coolant within the idle loop to be started up

~ and the coolant in the teactor pressure vessel is less than or equal i

$ to 50'F. or d b. When only one loop has been idle unless the tesperature differential 1d1L and opera _tingJac Q between the reactor coolant w'. thin the,F _..". .t.t ;;;t:tS; tien looosJs less than stoqual to ggO

,hi ht:]:]jd t!:t, n ;p t U . . O' 7;ted h ;;;- --f 4- .;. -=

!;K;cula-l r

APPL 1tAllLITY: OPERAT10kAL C0FJNilDNS 1, 2. 3 and 4.

L f AtT10N:

f bl Fib.

Withb"J5 hath temperature differenceshE5 "EMesceeding the

[DbNb '

I above limits, suspend startup cf any idle rectrculation loop.

i l '

sutvritt Antt ' REQUIREMENTS![

b /"

4.4.1.4 ThdNi"IL5".R temperature differentials f D shall be

~

detersined to ice annin e limits within 15 minutes pr or so startup of an idle retirculation loop.

4 FIRMI - UNIT 2 3/44-6 Amendernt No. 53 .

i

-9f7/94

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Amendment No. A1, 87 FERMI - UNIT 2 3/44-6a '

9/7/94

i SENlS$ y . - ,

l 1 ExHt 5lT CO2E TWEfN1RL- ;

- HVDEAULIC. .TAlSTA DIU TY DE BE l

' REACTOR COOLANT SYSTEM OPEFt1TEB INT 46 Safp1 DE s l 3/4.4.10 CORE THERMAL HYDRAULIC STABIL17Y E)('lT EE6)DMS AS SPGDF/G.D LIMITING CONDITION FOR OPERATION /// 10 f 160EE 5.4.10-1 ,

.. _3.4.10_ _The Reactor ccre shall no .. :;;r:ted-6: *^;f - A r * ;!::

- ---- - : Of /

! Q';g: 3 . 4 : " '. . -- _

, JPPLICABILITY: OPERATIONAL CONDITION 1 THE SCVWil FE6 tow AC

C I h- SPEElFIEb IM Fl6l LEE 5A,lB-l, I

a. With the Reactor operating in - ' ii_A ef_.fi_;;r: 3.4.10 -1 I sition.

QdbC '

mediatelWla:a.

THE EX the_RanctarEE6

._SwitchAthe SHUTD0 ION A 5 SPEalFIEb IN FIGUEE 34.Ib l,I I

b. LTth the Reactor operating ~tn4ogier, ! cf Ef;;r: 3.0.10 ' I by inserting control imediately igg to .::ttT-hBb rods C COLE FLD the--21 5;dr=li: insteht11ty

c If I

! /g : : occurs-as-evidence-by r-whHe-n ith"J~..:;ien 9, APRM-readinas eset!!ating by ;r::t:r th: er

.S E E equal-to-1000f-RATED-THERMAL-POWER-peak-te pe k er LPRM-eead4 ate IM.EPT.1 oscillating-greater-than-or-equal-to-30-watts /c::' ;; k : ;;:n,

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FD2 E6Vl5Eh T&YT -

! SURVEILLANCE REQUIREMENTS i

4.4.10.1 The provisions of Specification 4.0.4 are not applicable.

,4.4<10.2 With-THER MMAer-than-4000f-RAYED dE"%^.bP0b ::d :r:

Flow- ess-than-500+f-Rated-core -Flx, ::rify th:t th: 7::::: ::r: i: ,;;;

operating-in-Region-A-or-Region B ef Figer: 3.4.10-1 :: 1:::t :7,:7 :.;:q g

--hours I

\ .5~EE IN5E VT 2. GDi2. TEUISEb I c-XT' i

%bb FDDTNDIT. - s EE INSEET.5 FD2 ABBED TE7m x .

9/7/94 3/44-30 Amendment No. 53.

_ _ _ . [ERM _- UNIT 2

4 IN3ERT 1

(Replace the text in TS Section 3.4.10.c, Page 3/4 4-30 with the text below) p '% ,_ -

c. If core thermal hydraulic instability occurs as evidenced by a sustained increas

- APRM or LPRM peak-to-peak noise level reaching 2 or more times its initial level, and occurring with a characteristic period ofless than 3 seconds, immediately place the Reactor Mode Switch in the Shutdown position.

'~~----% M x J INSERT 2 (Replace the text in TS Section 4.4.10.2, Page 3/4 4-30 with the text below) l - -- ? W. , y 4.4.10.2 When operating within the Stability Awareness Region as specified in Figure 3.4.10-1. verify that the reactor core is not exhibiting core thermal hydraulic instabi!ity by monitoring APRM and LPRM signals immediately and at least once every hour.

INSERT 3 (Add the below text to the bottom of Page 3/4 4-30) v v

'* Restarting an Idle Recirculation Loop or resetting a Recirculation Flow Limiter are not acceptable methods ofimmediately increasing core flow to leave the Exit Region.

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( vw s)vamodimw , g Amendment No. JJ,87 FERMI - UNIT 2 3/4 4-31 9/7/94

85 85 M; !/ /////M k\\\ !e0 7

Stability Reg. ion Desen.p tions i

75 75:

3 d 5 >9S% Rod Line <40% Core Flow i :remmer Z :

Exit Regiort i

M in Reg *m 60 Y

$ 60- .

- and -

55 M7% RM N. <M Cue % l

$ SSU : >77% Rod Line. <45% Core Flow f 50

>103% Rod Line <50% Core Flow i q l Q- 50 -

St M A e sRy m:

-- 4 5 w 45 V : Win Scram or Exit Regions l

, q) : ~ '

40 i 40U

>62% Rod Line <45% Core Flow 35 35 ' '

' >98% Rod Line. <55% Core Flow 33 30 2

V i 25 ,,,, ,, ,, , ,,, ,,,, ,,,, - 25 30 35 40 45 50 55 60 ,

Core Flow (% Rated)

THERMAL PfMER VERSUS CORE FLOW THIS "N6V4" F/60Pd TO FIGURE 3.4.10-1 EEYl.Ac6 EXtSTu46 FIGGll& ,

1 3/4 '.~ 4 REACTOR ~ COOLANT SYSTEM BASES 'N l

. )/4.4.1 RECIRCULATION SYSTEM (Continued)

Sudden equalization of a temperature difference greater than 145'F bet' ween- i the reactor vessel bottom head coolant and the coolant in the upper region of !

the reactor vessel by increasing core flow rate would cause undue stress in the -

l

.. reactor vessel bottom head. ----

1:eir: :7.t; ;r; ig:::d t: er;hibit idi . h:; :t:rt ; :i:;; th; 7M l

- 1in; 0: =inini:: th: ;:t::t!:1 f:r 'ittistit; := t5:=1 Sydr::'^:

u . u u s u .. -

3/4.4.2 - SAFETY / RELIEF VALVES The safety valve function of the safety / relief valves operate to prevent

- the reactor coolant system from being pressurized above the Safety Limit of 1325 psig ia accordance with the ASME Ccde. A totti of 11 OPERABLE safety / relief valves is required to limit reactor pressure to within ASME III allowable values -i for the worst case upset translent.

Demonstration of the safety / relief valve lift settings will occur only

- during shutdown and will be performed in accordance with the provisions of .

Specification 4.0.5.

l . The low-low set system ensures that a potentially high thrust load 4

(designated as load case C.3.3) on the SRV discharge lines is eliminated during ,

subsequent actuations. This is achieved by automatically lowering the closing setpoint of two valves and lowering the opening setpoint of tw valves following the initial ~ opening. 50fficient redundancy is provided for ths low-low set' system such that failure of any one valve to open or close at its reduced setpoint does not vloiste the design basis. 4 r

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FERMI- .: UNIT 2 B 3/4 4-la Amendment No. Si 87

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_ /5CEAA _----

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LGME THE"2G&l6M BEFbCE OSCILLATlbMS .DELIELbW I

g VI Mn REACH .hN L)MhCCEPTA8 LE MMA0lTUDE. M' l'

i REACTOR COOLANT SYSTEM Q)l56b~I ibTTom c.f t-0C)279

%bM /5 i M BASES NMMb ~ _

,l3~ .i 3/4.4.10 CORE THERMAL HYDRAULIC STABILITY XENbN ebN CENTPAT16N ur noise in a stable

' ) bqj J BWR l modecores which is typically due to randonoperate with boiling and the flowpresence noise. As t'h of global r flow -

gy i conditions are changed, along with other systes parameters ,

3Q fa subcooling, sower distribution, etc.) the thermal hydraulic /rhtte kinetic l feedback mechanism can be enhanced such that random perturbations may result

in sustained limit cycle or divergent oscillations in power and flow.

/}Twomajormodesofoscillationshavebeenobservedin8W

\ICdD EW

}& the fundamental or core wide oscillation mode in which the entire come oscillates in phase in a given axial plane. The second mode involves regional The first mode is

, v1 g with oscillation in which one half of the core oscillates 180 degrees out of phase gg the other half. Studies have indicated that adequate margin to the i

z j (w

afglimit Minimum Critical Po_wer Ratio (SLMCPR) may not exial ring

.... . oscillations.

m TJ/6 SCPAh1AND EXIT EE610Nf a/)N kCC;V

..Mer.

" a. => a :f d=r; . 4Mre reseMast stable conditions of the Sant (high power / low flow)h.ee - ees " are usually entered as the result 4 of a plant transient (for exa.mple, recircu ation pump triys) and therefore ara 5

g soera11y not considered part of the normal operating domin. Sincegil ility ennts (including test experience) have occurrei in eitherq..;;= 4 t.y h

)S9

lCt sc lat y e ngtons are avoided to minimize the possibility of encountering s and not t 1 challenging the SLMCPR. Therefore, intentional W'

i operation n * -; N A er " s not allowed. It is recognized that during 2 @ $

v1 certain abnormal con . ons within the plant, it may become necessary to enter

o. Wa A e,

,) y btx4 cou d m ac plant for the safety purpose ofpurpose or for the protecting equipment of protecting a safetywhich, were or fuel it to fai b lu operating limit. In these cases, the appropriate actions for the region entered would be performed as required. Q Qgg EEshs.t m Most oscillations that have occur ed during test gandoperattoihave occurred at or above the 96% rod I w rore/ ow near natural circulation. l N This behavior is constitent with anal se;s whic 4redict reduced stability Vi ma g f1' . As core flow is increased or

\ pq^;rgin with increasiW e p . the_orobability poweroscilla ens or occurring decreas will decrease.

b fUstsobservedinTE(wKs..,

~

L i;.. m a,;re 4. .J17' boundy the maj ity of the stability events and

% B . Since A.;; -fepresents the least stable region AlJ of the power / flow operating domain,TheTotential to rapidly encounter la d magnitude core thermal hydraulic oscillations is increased. Durinc +pe 4 tr*"5 s..... !.'. atS ' th' .. '..'.*.r*

r

"*t

.....h. "_5"f fi ci'atS...*_* t'E?=M....j;"i..w.e

"..*.,Y . . . .. ....

eu m 7-

,E p h.the prompt action of manuall _.scransLing_.tfle f p' 23t ith'eh

.. A T; =t;i : iT nwincte enn ;=t=ti= f tt.: Suica". glg V, 2 ' THC SCilkk'1 EE&totd IS GNT525b O2 \NW6M Q5EILLATIOl1.L

.M26 DETdCTED 11 E6&OR6D To EMIdea vec,e cm,om

4) h +

jg - OF TRE SLM CP gmg, FERMI - UNIT 2 8 3/4 4-8 Amendment No. A1,87

e l

+-

INSERT.1 j (Add the following sentence to the beginning of the third paragraph of-~

Bases Section 3/4.4.10, Page B 3/4 4-8) j v V e _

- Figure 3.4.10-1 specifies the Scram, Exit, and Stability Awareness Regions by providing :

a description of these region boundaries based on Rod Line, Core Flow and Thermal

- Power, -

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4 l

REACTOR C00LAKT SYSTEM s BASES 3/4.4.10CORETHERMALHYDRAULICSTABILITY(Continued) -

" {Esed-on-test hydroutic-esci11ations-is- =d :;:ntig =;4a4e,4en ess- 5 the 5 5;5 A.:rt re,t=y t M =th:-erefr: r cy ef-cors. thermal expected-end-;=5:r dieted-te fue r=;; k t=r thr 5n;h.

thu =;A.t=

A u s:i=;

, ; n;i= 2 ;;==

M = rgir, t; t :-

h. i 1

lower-;;=r SLMCPR M 11-typk: end-hiI1; h hq: S t-i= 5 the 5 5;t= A. "ith == {

.margia-te SLTPP. W : t=r ;nhhtlity ef =:ilhth=, nitir.; t;;;r. by contret -red- h=rtin i: j=tifud, t = = r, if = r.1 ht ur.; er; et;;r=d- '1 whOe-exiting-Region-8,-th =:ter :(115 r=11y :;r:md. '

~

> 4he-tetentiet-for-core-therne14;d=:1u :=tlutt=: : ==r utw; ;f i Regh= A and-8-is-vwM 11 M thenfere :;=hl = 24*ements-st: =t necessary eet:tde-ef tf n: n;i= :.

k o .

Based on test and operating exserience, the frequency of core thermal hydraulic oscillations is less in t1e Exit Region than in the Sc'am Region.

Core decay ratios are expected and predicted to be lower in this region since the Exit Region covers a lower power and higher flow range than the Scram

/ Re ion. Also, the margin to the SLMCPR wil typically be larger in the Exit Re ion than in the Scram Region. With more margin to SLMCPR and a lower probability of oscillations, leavin the Exit Region by immediate control rod i insertion or core flow increase is ustified. Restarting an Idle Recirculation Loop or resetting a Recirculation Flow limiter are not acceptable methods of increasing core flow to leave the Exit P.egion because j these actions would not support timely completion of this immediate action. if I oscillations are observed at any time, the reactor will be manually scrammed. '

. The potential for core thermal hydraulic oscillations to occur outside of the Scram or Exit Regions is small, but could occur. Therefore, frequent monitoring of .PRM and LPRM signal., is appropriate when operating in the ..

g Stability Awareness Region. /

/

Q ,. '

/

-- !.5 5 ..

i 9/7/94 FED!! UNIT 2 8 3/4 4 9 Amendment No. 53'

Enclosure 3 to NRC 98 0003 l' age 3

/

ENCLOSURE 3-PART2 PROPOSED TECilNICAL SPECIFICATION REVISED PAGES INCLUDED PAGES:

xxl 3/4 3 61a 3/4 3 63a 3/44-6 3/4 4-6a 3/4 4-30 3/4 4-31 113/44 12 113/448 B 3/4 4 9

_ . - _ - . _ _ _ _ - _ . . _ . _ - _ _ _ . - _ . _ _ _ . _ . - _ _ _ _ _ _ _ i umu uuimunimuse

LIST OF FIGURES ._

FIGURE PJ_GI 3.1.5 1 SODIUM PENTABORATE VOLUME /

CONCENTRATION REQUIREMENTS................... 3/4 1-21 3.4.1.4-1 DELETED...................................... 3/4 4 6a l 3.4.6.1-1 MINIMUM REACTOR PRESSURE VESSEL HETAL TEMPERATURE VS. REACTOR VESSEL PRESSURE.................. 3/4 4-21 3.4.10 1 THERMAL POWER VS. CORE FL0W.................. 3/4 4 31 4.7.5 1 SAMPLEPLAN2)F0RSNUBBERFUNCTIONALTEST... 3/4 7-21 B 3/4.3-1 REACTOR VESSEL WATER LEVEL................... B 3/4 3-7 83/4.6.2-1 LOCAL POOL TEMPERATURE LIMIT................. B 3/4 6-5 B 3/4.7.3 1 ARRANGEMENT OF SHORE BARRIER SURVEY POINTS... B 3/4 7-6 5.1.1-1 EXCLUSION AREA............................... 5-2 5.1.2-1 LOW POPULATION Z0NE.......................... 5-3 5.1.3-1 MAP DEFINING UNRESTRICTED AREAS AND SITE B0UNDARY FOR RADIDACTIVE GASE0US AND LIQUID EFFLUENTS.................................... 5-4 FERMI - UNIT 2 xxi Amendment No Jp,f),E),E/,JJ,

I ,. TABLE 3.3.7.5-1 (Continued)

r

-g ACCIDENT MONITORING INSTRUMENTATION MINIMUM APPLICA8tE

'E REQUIRED NUMBER CHAfG8ELS OPERATIONAL Z' INSTRUMENT- 0F CHANNELS OPERABLE CONDITIONS ACTION m-

13. . Standby Gas Treatment S'; stem Radiation Monitors

a. SGTS - Noble Gas (Low-range)I l/0PERABLE l/0PERABLE 1, 2, 3 81 SGTS subsystem SGTS subsystem

b. SGTS - Noble Gas (Mid-range) 1/0PERABLE l/0PERABLE 1, 2, 3 81 SGTS subsystem SGTS subsystem

c. SGTS - AXM-Noble Gas (Mid-range) l/0PERABLE l/0PERABLE 1, 2, 3 81 SGTS subsystem SGTS subsystem M* d. SGTS - AXM-Noble _ Gas (High-range) 1/0PERABLE l/0PERABLE 1, 2, 3 81 y SGTS subsystem SGTS subsystem e

5~ 14. Deleted '

.15. Deleted

16. Primary Containment Isolation Valve Position 1/ valve 1/ valve 1, 2, 3 82 a

e g.

w Also included in the OFFSITE DOSE CALCULATION MANUAL.

1 4

r, ,. +m ,--r, , _r <---c , -_- . ,-+

TABLE 4.3.7.5-1 (Continued)

E ~ ACCIDENT MONITORING INSTRtMENTATION SURVEILLANCE REQUIRimnTS APPLICA8tE E- . CHANNEL CHANNEL OPERATIONAL INSTRUMENT'. CHECK CALIBRATION COMITIONS

13. . Standby Gas Treatment System Radiation Monitors a.- SGTS - Noble Gas (Low-range) M R - 1, 2, 3

b. SGTS - Noble Gas _(Mid-range) M R- 1, 2, 3 c.; : SGTS - AXM-Noble Gas (Mid-range) M R 1, 2, 3

d. SGIS - AXM-Noble Gas (High-range) M R 1, _ 3

14. Deleted j

15. Deleted w

16. Primary Containment Isolation Valve Position M R 1, 2, 3

.a 4

4

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. _ . _ - . . . . _ . . _ . . . _ . . . _ , . - . _ . , . - . - . _ _ _ . - - . . _ - . . . . _ _ . - - . _ - . . - ~ . . . . .

REACTOR COOLANT SYSTEM IDLE RECIRCULATION LOOP STARTUP M ONDITION FOR OPERATION 3.4.1.4 An idle recirculation loop shall not be started unless the temperature l

differential between the reactor pressure vessel steam space coolant and the j bottom head drain line coolant is less than or equal to 145'f, and: j

a. When both loops have been idle, unless the temperature differential !

between the reactor coolant within the idle loop to be started up l and the coolant in the reactor pressure vessel is less than or equal-l' _to 50'F, or ,

e

b. When only one loop has been idle, unless the temperature differential between the reactor coolant within the idle and

-operating recirculation loops is less than or equal to 50'F. t AFPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 and 4.

]

ACTION: '

i With temperature differences exceeding the above limits, suspend startup of any l idle recirculation loop.

i SURVEILLANCE RE0VIREMENTS 4.4.1.4 The temperature differentials shall be determined to be within the l limits within 15 minutes prior to startup of an idle recirculation loop.

t.

f i

l i

l-

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FERMI - UNIT 2 3/4 4-6 Amendment No.JJ L -

FIGURE 3.4.1.4 DELETED l

1 1

FERMI - UNIT 2 ?/4 4-6a Amendment-No. U ,E/,

l REACTOR COOLANT SYSTEM r

=1/4.4.10 CORE THERMAL HYDRAULIC STABILITY LIMITING CONDITION FOR OPERATION 3.4.10 The Reactor core shL11 not exhibit core thermal hydraulic instability or be operated in the Scram or Exit Regions as specified in Figure 3.4.10-1.

APPLICABILITY: OPERATIONAL CONDITION 1 ACTION: ;

a. With_the Reactor operating in the Scram Region as specified in figure 3.4.10 1, immediately place the Reactor Mode Switch in the Shutdown position,

b. With the Reactor _ operating in the Exit Region as specified in Figure 3.4.10-1, immediately initiate action to leave the Exit Region by inserting control rods or by increasing core flow.* i

c. If core thermal hydraulic instability occurs as evidenced by a sustained increase in APRM or LPRM peak-to peak noise level reaching 2 or more times its initial level, and occurring with a characteristic period of less than 3 seconds, immediately place the Reactor Mode Switch in the Shutdown position.

SURVEILLANCE RE0VIREMEN15 4.4.10.1 The provisions of Specification 4.0.4 are not applicable.

4.4.10.2 When operating within the Stability Awareness Region as specified in Figure 3.4.10 1, verify that the reactor core is not exhibiting core thermal hydraulic instability by monitoring APRM and LPRM signals immediately and at least once every hour.

Restarting an Idle Recirculation loop or resetting a Recirculation Flow limiter are not acceptable methods of immediately increasing core flow to leave the Exit Region, o FERMI - UNIT 2~ 3/4 4-30 Amendment No. J),

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/ I" Stability Region Descriptions

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3/4.4 REACTOR COOLANT SYSTEM BASES :

I 3/4.4.1 RECIRCULATION SYSTEM (Continued)

Sudden equalization of a temperature difference greater than 145'F 1 between the reactor vessel bottom head coolant and the coolant in the upper i region of the reactor vessel by increasing core flow rate would cause undue ;

stress in the reactor vessel bottom head.

3/4.4.2 SAFETY / RELIEF VALVES The safety valve function of the safety / relief valves operate to prevent the reactor coolant system from being pressurized above the Safety Limit of 1325 psig in accordance with the ASME Code. A total of 11 OPERABLE safety / relief valves is required to limit reactor pressure to within ASME III allowable values for the worst case upset transient.

Demonstration of the safety / relief valve lift settings will occur only during shutdown and will be performed in accordance with the provisions of i Specification 4.0.5.

The low low set system ensures that a potentially high thrust load (designated as load case C.3.3) on the SRV discharge lines is eliminated during subsequent actuations. This is achieved by automatically lowering the closing set)oint of two valves and lowering the opening setpoint of two valves following tie initial opening. Sufficient redundancy is provided for the low-low set system such that failure of any one valve to open or close at its reduced-setpoint does not violate the design basis.

FERMI - UN!_T 2 B-3/4-4 la - Amendment No. 53, 87,- _

l l

REACTOR COOLANT SYSTEM

]

BASES 3/4.4.10 CORE THERMAL HYDRAULIC STABILITY l

BWR cores typically operate with the presence of global flux noise in a stable mode which is due to random boiling and flow noise. As the power / flow conditions are changed al power distribution, etc. the thermal concentration,subcoollng,ongwithotherreactorparameters(x perturbations may result in sustained limit cycle or divergent oscillations in '

power and flow.

Two major modes of-oscillations have been observed in BWRs. The first mode is the fundamental or core wide oscillation mode in which the entire core oscillates in phase in a given axial plane. The second mode involves regional oscillation in which one half of the core oscillates 130 degrees out of phase with the other half. Studies have indicated that adequato margin to the Safety Limit Minimum Critical Power Ratio (SLMCPR) may not exist during oscillations. l Figure 3.4.10 1 specifies the Scram, Exit, and Stability Awareness Re i aroviding a description of these region boundaries based on Rod Line,gions Core by Flow and Thermal Power. The Scram and Exit Re conditions of the plant (high power / low flow).gions The Scram represent and Exitthe least are Regions stable usually entered as the result of a plant transient (for example, recirculation pump trips) and therefore are generally not considered part of the normal operating domain. Since all stability events (including test experience) have occurred in either the Scram or Exit Regions, these regions are avoided to j minimize the )ossibility of encountering oscillations and potentially challenging tle SLMCPR. Therefore, intentional operation in the Scram or Exit Regions is not allowed. It is recognized that during certain abnormal conditions within the plant, it may become necessary to enter the Scram or Exit Regions for.the purpose of protecting equipment which, were it to fail

could impact plant safety or for the purpose of protecting a safety or fuel, operating limit. In these cases, the a entered would be performed as required.ppropriate actions for the region Most oscillations that have occurred during testing and operation have occurred at.or above the 96% rod line with core flow near natural circulation.

This behavior is consistent with analyses which margin with incre r.ing power or decreasing flow. As predictcore reduced stability or 'l flow is increased power decreased, the probability of oscillations occurring *will decrease. The Scram and Exit Regions bound the majority of the stability events and tests '

observed in GE BWRs. Since the Scram Region represents the least stable region of the power / flow operating domain, the potential to rapidl

,large magnitude core thermal hydraulic oscillations is increased. y encounter During transients,.the operator may not have sufficient time to leave the Scram '

region before oscillations develop and reach an unacceptable magnitude.

Therefore, the prompt action of manually scramming the plant when the Scram '

Region is entered or when oscillations are detected is required to ensure protection of the SLMCPR.

FERMI: : UNIT 2 B 3/4 4-8 Amendment No.- A3, W ,

. - .. L

REACTOR COOLANT SYSTEM BASES _,_

3/4.4.10 CORE THERMAL HYDRAVLIC STABILITY (Continued)

Based on test and operating exserience, the frequency of core thermal hydraulic oscillations is less in tie Exit Region than in the Scram Region.

Core decay ratios are expected and predicted to be lower in this region since the Exit Region covers a lower power and higher flow range than the Scram Region. Also, the margin to the SLMCPR will typically be larger in the Exit Region than in the Scram Region. With more margin to SLMCPR and a lower probability of oscillations, leaving the Exit Region by immediate control rod insertion or core flow increase is ;ustified. Restarting an Idle Recirculation Loop or resetting a Recirculation Flow limiter are not acceptable methods of increasing core flow to leave the Exit Region because these actions would not support timely completion of this immediate action. If oscillations are observed at any time, the reactor will be manually scrammed.

The potential for core thermal hydraulic oscillations to occur outside of the Scram or Exit Regions is small, but could occur. Therefore, frer,uent monitoring of APRM and LPRM signals is appropriate when operating in the Stability Awareness Region.

FERMI - UNIT 2 B 3/4 4-9 Amendment No. E ,

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Detroit Misen

! j January 28,1998 NRC 98 0003 '

i

U. S. Nuclear Regulatory Commission

Attention: Document Control Desk 4

Washington, D C 20555 l j

References:

1) Fennl2 Docket No. 50 341 i- NRC License No. NPF-43

2) NRC Generic Letter 94 02,"Long Term Solutions and ,

Upgrade ofInterim Operating Recommendations for Thermal- ,

Hydraulic Instabilities in Boiling Water Reactors", dated July 11.1994 ,

3) Detroit Edison letter to NRC, NRC 94 0089," Detroit Edison i

- Response to NRC Generic Letter 94 02", dated September 8,

i 1994

4) Detroit Edison letter to NRC, NRC 97 0105," Proposed Technical Specification Change (License Amendment)- -+

Neutron Monitoring System", dated December 10,1997

5) BWROG Letter 94078,"BWR Owners' Group Guidelines for . r Stability Interim Corrective Action", dated June 6,1994

6) NRC Letter Boger to Tully,"NRC Evaluation of BWR-Owners' Group Topical Report NEDO-31558 Pesition on NRC Regulatory Guide 1.97, Revision 3, Requirements for

, Post Accident Monitoring System", dated January 13,1993 7)- NRC Letter Colburn to Gipson," Regulatory Guide 1.97 -

Boiling Water Reactor Neutron Flux Monitoring - Fermi 2",

dated February 17,1994 ;

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USNRC NRC 98 0003 Page 2

8) NRC Letter Colburn to Gipson," Regulatory Guide 1.97 -

lloiling Water Reactor Neutron Flux Monitoring MPA A 17",

dated May 10,1993

9) GE Nuclear Energy Letter, Lim to Thorson," Fermi 2 Recirculation Operating Loop f> rive Flow Limit Prior to Idle Pump Restan, Mill 96009, dated January 29,1996

10) NUIUiG 1433," Standard Technical Specifications, General Electric Plants,llWR/4", Revision 1, April 1995.

I1) NEDO 31960 A,"IlWR Owners' Group Long Term Stability Solutions Licensing Methodology," dated November 1995

Subject:

Proposed Technical Specification Change (License Amendment)-

Thermal llydraulic Stability, idle Recirculation Loop Startup, and Post Accident Monitorine _

Pursuant to 10CFR$0.90, The Detroit Edison Company (Deiroit Edison) hereby files an application to amend the Femii 2 Technical Specifications (TS). The proposal revises the Technical Specificatims to improve the administrative controls related to thermal hydraulle stability and idle recirculation loop startup, and eliminates neutron flux indication from the TS post accident monitoring parameters.

Enclosure 1 provides a description and evaluation of the proposed changes.

Enclosure 2 provides an analysis of the significant hazards consideration assessment using the standards in 10CFR50.92.

Enclosure 3 provides marked up pages of the existing Technical Specifications to show the proposed changes and a typed version of the affected Technical Specification pages with the proposed changes incorporated.

As described in Reference 4 Detroit Edison plans to install the new Power Range Neutron Monitoring (PRNM) System during the sixth refueling outage (RF06). As also discussed in Reference 4, the Stability Option til automatic trip will be installed during RFO6 in the " indicate only" mode. Assuming successful operation of the system in the following operating cycle, the automatic trip would be activated during the seventh refueling outage (RF07). The proposed Technical Specification changes related to the PRNM hardware changes are provided in the Reference 4 request. Part of the enclosed TS changes address the administrative controls for the interim period

- prior to the activation of the Option ill trip. Technical Specification changes for the activation of the Option til trip will be submitted separately in time for the scheduled activation during RF07, currently planned for the spring of the year 2000.

The enclosed thermal hydraulic stability changes are being proposed separately from the Reference 4 Power Range Neutron Monitoring (PRNM) System TS changes

,c USNRC -

NRC 98 0003 '

Page 3 ,

because the changes are not directly related to the FRNM hardware change or the i t

associated General Electric Licensing Topical Report. In addition, it is the intent that the enclosed changes be implemented prior to the PRNM hardware changes to ,

enhance thermal hydraulic stability controls in the interim period until the automatic !

trip function is installed in RF07. Because the proposed license amendment :

represents enhanced administrative controls related to thermal hydraulic stability, approval is requested as soon as practical. The changes will be implemented within 60 days of approval.

Detroit Edison has evaluated the proposed Technical Specifications against the criteria of 10CFR$0.92 and determined that no significant hazards consideration is !

involved. The Fermi 2 Onsite Review Organization has reviewed and recommended :

5 approval of the proposed TS changes. The Nuclear Safety Review Group has reviewed the proposed changes and concurs with the enclosed determinations. In i accordance with 10CFR50.91, Detroit Edison is providing a copy of this letter to the State of Michigan.

If you have any questions or comments, please contact Mr. Joseph Conen of my stafT .

at (734) 5861960.

1 Sincerely, ,

/

Enclosures ec: A.B. Beach B. L. Burgess O. A. liarris A. J. Kugler l Supervisor, Electric Operators.

Michigan Public Service Commission i

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i 4-i USNRC I NRC 98 0003 Page 4 4

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' l Douglas R. Gipson do hereby amrm that the foregoing statements are based on !

facts and circumstances that are true and accurate to the best of my knowledge and l 3-L belief. !

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1 Douglas R.Dipson i i Senior Vice President :

Nuclear Generation i

oG71 v day of [A WM fe/1998 before me ~

personallylf On this

, appeared Douglas R. Gipson being first duly swompnd says that he executed the foregoing as his free act and deed. t

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4 Enclosure 1 to NRC 98 0003 Page1 ENCLOSURE 1 FERMI 2 NRC DOCKET NO 50 341 OPERATING LICENSE NPF-43 REQUEST TO REVISE TECilNICAL SPECIFICATIONS:

TilERMAL ilYDRAULIC STAHILITY, IDLE RECIRCULATION LOOP STARTUP AND POST-ACCIDENT MONITORING DESERIPTION AND EVAL,UATION Of Tile PROPOSED CIIANGES

6 Enclosure I to NRC 98 0003 l' Page 2 i

BACKGHOUND

~

As discussed in Reference 4, Detroit Edison is planning to replace the power range ;

monitor portion of the Neutron Monitor'ag System with a GE digital NUMAC Power Range Neutron Monitoring (PRNM) retrofit system. The new equipment will ultimately include capability for an automatic Oscillation Power Range Monitor (OPRM) trip to detect and suppress possible thermal hydraulic instabilities in the plant. The new OPRM trip function, when enabled, uMi implement the Boiling Water Reactor Owners Group '

(BWROO) defined " Stability 0-l, tion 111" alternative (Reference 11). Ilowever, the OPRM trip function will not be enabled during the first cycle of operation with the new equipment. This proposed Technical Specification (TS) change request is being submitted to provide improved administrative controls for thermal hydraulic stability during the '

phased implementation of(PRNM) System changes.

Detroit Edison is proposing the follow'ng three specific TS changes associated with thermal hydraulic stability and post accident monitoring: ;

A. Changes to prohibited and restricted operating regions, Limiting Condition for '

Operation (LCO). Actions, Surveillance Requirements and Bases in the Core Thermal liydraulic Stability Specification consistent with NRC Generic Letter 94 02 (TS 3.4.10, Figure 3.4.101, and B 3/4.4.10).

B. Stability related improvements in the operating restrictions in the Idle Recirculation Loop Startup Specification (TS 3.4.1.4, Figure 3.4.1.4 1, and B3/4.4.1).

C. Elimination of Neutron Flux from parameters required to be monitored by Post Accident Monitoring TS, as allowed by References 6 and 7 (Tables 3.3.7.51 and 4.3.7.5 1)

Each of these changes are individually evaluated in the Evaluation section of this submittal.

EVALUITION A. Changes it, prohibited and restricted operating regions, LCO, Actions, Surveillance Requirements, and Bases in the Core Thermal Hydraulic Stability Specifiention consistent with NRC Generic Letter 94-02 (TS 3.4.10, Figure 3.4.101, and H 3/4.4.10).

Chances to prohibited and Restricted Oneratine Recions:

The proposed change modifies the current restrictions on operating at power levels and core flow levels that are designed to avoid core thermal hydraulle instabilities. Associated bases are also being revised. The current TSs are based upon the guidance of GE Service information Letter (SIL) 380. Revision 1 and NRC Bulletin 88-07, Supplement 1. The proposal updates the restrictions to include the additional interim corrective actions Detroit Edh has taken in response to NRC Generic Letter (GL) 94-02 (Reference 2).

.t Enclosure 1 to NRC 98 0003 Page 3 In response to GL 94 02, the Bolling Water Reactor (BWR) Owners' Group (BWROG) issued Reference 5 to provide improved guidelines for stability interim coructive actions.

Reference 3 provided Detroit Edison's response to GL 94 02. Detroit Edison ecumitted to modifylag operating procedures and updating training to be consistent with 6e BWROO guidance. Implementation of the guidance in this proposed change is conskient with the actions already taken.

Chances to Ficure 3.4.101:

TS Figure 3.4.10 1, "Thennal Po,ver versus Core Flow" has been modified to show new Scram, Exit, and Stability Awareness Regions in terms of percent rod line, core flow and thermal power. The new regionc are intended to replace references to Regions A UJ B, as well as the previous region of survei!!ance applicability (Greater than 30% of RATtID TilERMAL POWER and Core Flow less than 50% of Rated Core Flow). Since thesc Stability Region boundaries are based on Rod Lines (lines of constant control rod configuration on a power /Dow map), the applicable Rod Line representations have been re computed for this Figure. The re computed Rod Line representations are consistent ,

with empirical data taken along the 100% Rod Line during both recent and historical startup testing.

The change to the nomenclature of stability regions from Region A and B to the Scram and Exit Regions respectively, is consistent with the latest BWROO Guidance. For the proposed revision to Figure 3.4.101, the Scram Region boundaries remain the same as Region A, but the Exit Region Boundaries are larger than Region B, providing additional stability margin. The Ogure was also revised to show core Dow starting at 30 percent rated core Dow as opposed to the present 20 percent rated core flow. This change was made to improve the readability of the ugure since core flows less than 30 percent are not possible over the rod lines ofinterest.

Chances to Limitine Condition for Ooeration (I CQ);

The new Limiting Condition for Operation statement that the reactor core shall not exhibit core thermal hydraulic instability is to ensure that the LCO is broad enough so that the guidance of ACTION 3.4.10.c to scram the reactcr when an instability occurs remains applicable when in any region and not only v...cn in the Exit Region.

Chances to ACTION:

Changes to ACTION 3.4.10.a are intended to reflect the new nomenclature adopted for Region A as discussed above.

Changes to ACTION 3.4.10.b are intended to reflect the new nomenclature adopted for Region B as discussed above Leaving the Exit Region by a core flow increase is permitted in addition to leaving the region by control rod insertion, A new footnote states that Restarting an Idle Recirculation Lesp or resetting a Recirculation Flow Limiter are not acceptable means ofimmediately increasing core Dow to leave the Exit Region. This footnote is intended to prevent delays in leaving the Exit region that might typicahy be

.s l Enclosure 1 to NRC-98 0003 l

, page 4 i associated with these activities, because execution of these activities could not be completed in a short enough time to be considered immediate. A core flow increase will result in establishing a more stable power /Dow relationship with respect to thermal-hydraulic instability, and thus is an acceptable means to leave the Exit Region.

Changes to ACTION 3.4.10.c have been made to eliminate a tie between completing a manual scram in the event an instability is detected and operation in Region B (now the Exit Region). The new ACTION 3.4.10.c now clearly requires a muual scram whenever instability is detected, regardless of what region the reactor is operating within. In addition, the definition of what constitutes an instability has been revised to re0cct the Irtest BWROO guidance, and is intended to be able to describe reactor / nuclear instrumentation behavior at the onset ofinstability. This contrasts with the existing TS guidance, which is more representative of conditions where a fully developed instability is in progress. The improvements to ACTION 3.4.10.c will result in appropriate manual protective actions being applicable over a larger operating region and at a more appropriate time.

ChangmS_urveillance Reauirements The change to the Surveillance Requirements (SR) re0cet a more appropriate action to be taken when operating near Region B (now the Exit Region). The rcvised Surveillance requirement would require an immediate and periodic (once every hour) assessment of reactor stability by monitoring the nuclear instrumentation for signs ofinstability. This guidance is app opriate for operation near the Exit Region, resulting in early detection of any instability that might occur, and provides superior protection against instability compared to the existing TS guidance to check on a 4 hourly basis that the reactor core is not operating in Region A or B. The current SR does not account for the pctential for instability when operating outside Region B (now the Exit Region), while the new guidance requires that the reactor be assessed for stability when operating near the Exit Region. The change to the region of applicability for the Suneillance Requirement is in accordance with the latest BWROG guidance and bounds the Exit Region by 5% in core now and Rod Line. The shape of this new Stability Awareness Region is intended to resemble the shape of the Scram and Exit Regions in that all three region boundaries are loosely based on predicted lines of constant Core Decay Ratio (a measure of stability of an oscillating system). The new Stability Awareness Region does not bound the current region of surveillance applicability in all areas since the current region was based on simplicity not on predicted lines of constant Core Decay Ratio, While the proposed Stability Awareness Region is smaller than the existing surveillanr region, it does provide an extension to the current region of surveillance applicability at high Rod Lines up to 55% Core Dow to support the concept of a line of constant Core Decay Ratio.

a l Enclosure I to !

NRC 98 000) !

Page5

}

i Changes to llASES Changes to the DASES section support the changes in nomenclature from Regions A and H to the Scram, Exit and Stability Awareness Regions and to reflect that core flow increases, within appropriate limitations, are permissible for leaving the Exit Region.

Changes were made to reDect knowledge gained from instability events that occurred outside Regions A and D. Specifically, while rare, the potential exists for instabilities to occur near the Exit Region boundary, and low Xenon concentration can be a factor in creating conditions where instability can occur. Reactor pressure is also a contributing .

, factor, but has a second order effect when compared to Xenon concentration, subcooling and power distribution.

Changes were also made to reflect knowledge gained in the development of hardware based solution trip algorithms as to the impact of even small instabilities on the Safety Limit Minimum Critical Power Ratio (SLMCPR). This includes the need to scram for any instability, whether core wide or regional, in order to protect the SLMCPR.

In summary, these proposed changes are consistent with the industry developed guidelines for avoiding instabilities and are in agreement with the requirements in OL 94 02,

11. Stability related improvements in the operating restrictions in the Idle Recirculation Loop Startup Specification (TS 3.4.1.4, Figure 3.4.1.41, and H3/4.4.t).

The proposal climinates redundant provisions in TS Section 3.4.1.4 related to power and core flow operating regions where an lie Recirculation Loop Startup is permitted because these restrictions are functionally redundant to those specified in TS Section 3.4.10.

i Associated bases are also being changed. -This includes deleting text in TS 3.4.1.4 and deleting TS Figure 3.4.1.4 1. In addition, the TS Section 3.4.1.4 requirement to reduce the operating loop now to less than or equal to 50% of rated loop flow prior to starting an idle loop is being eliminated. This change is proposed because compliance with this specification requires operation at low core Cows where core thermal hydraulie stability can be a concem and the original generic bases for this specification an. no longer appropriate. The original generic bases were related to scram avoidance and jet pump riser brace vibration (Reference 9). The following provides a discussion of these bases. ;

Feram Avoidance 11ases: The scram avoidance bases were related to the neutron Dux spike seen during idle loop startup and reduced margin to the APRM flow biased neutron -

. Aux scram from single pump operation due to lower measured drive flow. Detroit Edison

does not use an APRM flow biased neutron flux scram but instead employs an APRM flow biased simulated thermal power scram utilizing a filter circuit with a time constant of approximately 6 seconds to filter the neutron flux signal.- In addition, Detroit Edison also

. employs a jog circuit on the recirculation pump discharge valves that results in a less severe APRM response to the idle loop restart transient. The combination of the less

- severe idle loop restart APRM response and APRM scram setpoints that are based on a

.t Enclosure 1 to NRC 98 0003 page 6

filtered neutron flux signal eliminates any need to reduce core thermal hydraulle stability i margin to gain margin for APRhi scram avoidance orjet pump riser brace vibration by lowering operating loop flow to less than or equal to 50% of rated loop flow prior to an idle loop startup (Reference 9).

Jet pumn Riser Ilrace Vibration Ilases: This bases is related from an experience with an earlier llWR plant where restrictions were imposed due to a concem about jet pump riser brace vibration. The riser brace loads increased with the amount of flow mismatch between the two recirculation loops that occurred following the restart of the idle pump. '

When returning to two loop operation from single loop, the transition would have been through the mismatch speed zone if the operating loop was allowed to be at greater than 50% speed. To avoid this, the operating loop flow had to be lowered to 50% before .

starting the idle loop. Riser braces were modified for later plants, including Fermi 2t therefore, these restrictions are unnecessary at Fermi 2. Furthermore, the flow mismatch resulting from restart of the idle pump is adequately addressed by requirements in the Fermi 2 Technical Specifications 3.4.1.3 (Reference 9).

These changes are also consistent with the IlWR Improved Standard Technical Specification, NUREG 1433, Rev.1 (Reference 10).

C. Elimination of neutron flux from parameters required to be monitored by Post !

4 Accident Monitoring Specification as allowed by Reference 6 (TS Tables 3.3.7.5-1 and 4.3.7.5-1).

The proposed change eliminates Neutron Flux instrumentation from the list of required post accident monitoring instrumentation contained in TS Tables 3.3.7.5 1 and 4.3.7.5 1.

, The basis for this change is that Neutron Flux instrumentation was determined not to be Category I instrumentation for IlWRs (as defined in NRC Regulatory Guide 1.97). Since the purpose of the post accident monitoring TS is to assure that instrumentation critical to post accident operations (Category I instrumentation) is available, the Neutron Flux instrumentation does not need to be included in the scope of this TS.

The NRC,in a letter dated hiay 10,1993. (Reference 8) requested Fermi to review their neutron flux monitoring instrumentation against the criteria of NEDO 31558 and provide a letter to the NRC documenting results of the review. This letter also informed Detroit Edison that neutron flux monitoring instrumentation for IlWRs is no longer considered Category 1 instruraentation and that Femti could request removal of the neutron flux monitoring instrumentation from the post accident monitoring technical specifications if the system met the attemate criteria proposed in NEDO 31558. Detroit Edison responded by submitting results of the review to the NRC in a letter dated September 28,1993.

From this, the NRC concluded that: 1) the deviations from NEDO 31558 are acceptable and 2) the post accident neutron flux monitoring instrumentation at Femti 2 is an acceptable alternative to the guidance of RO 1.97 Category I criteria. This conclusion is documented in a letter dated February 17,1994 (Reference 7).

Ilased upon the above, the proposed change is acceptable.

] j. . !

I i

Enclosure I to '

NRC 98 0003

Page 7 l

-i SIGNIFICANT HAZARDS CONSIDERATION i in accordance with 10CFR50.92, Detroit Edison has made a determination that the !

i proposed amendment iv>olves no significant hazards considerations. To make this i

determination, Detroit Edison must establish that operation in accordance with the j proposed amendment would not: (1) involve a significant increase in the probability or j consequences of an accident previously evaluated; or (2) create the possibility of a new or i different kind of accident from any accident previously evaluated; or (3) involve a j significant reduction in a margin of safety. The significant hazards consideration !

' assessment is presented in Enclosure 2. - ;

ENVIRONMENTAL IMPACT f Detroit Edison has reviewed the proposed Technical Specification changes against the [

[ criteria of 10CFR51.22 for environmental considerations. The proposed change does not i

- involve a significant hazards consideration, nor significantly change the types or -!

significantly increase the amounts of eilluents that may be released offsite. In addition the ,

proposed changes do not involve a significant increase in individual or cumulative 1 occupational radiation exposures.11ased on the foregoing, Detroit Edison concludes that -I the proposed TS meet the criteria given in 10CFR51.22(c)(9) for a categorical exclusion i from the requirements for an Environmental Impact Statement. ;

CONCLUSION liased on the evaluation above: 1) there is reasonable assurance that the health and safety ,

of the public will not be endangered by operation in the proposed manner, and 2) such l

- activities will be conducted in compliance with the Commission's regulations, and proposed amendments will not be inimical to the common defense and security or to the health and safety of the public.

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1 Enclosure 2 ta l I

NRC 98 0003 PageI ENCLOSURE 2 FERMI 2 NRC DOCKET 50 341 OPERATING LICENSE NPF-43 REQUEST TO REVISE TECilNICAL SPECIFICATIONS:

10CFR50.92 EVALUATION llASIS FOR SIGNIFICANT liAZARDS DETERMINATION:

The proposed Technical Speci0 cation changes described in Enclosure 1 do not involve a significant hazards consideration. This determination was reached by evaluating: 1) the changes to TS 3.4.1.4 and 3.4.10 associated with thermal hydraulic stability and idle recirculation loop startup, and 2) the changes to TS 3/4.3.7.5 associated with post-accident monitoring. These evahtations are provided below:

r Thermal livdraulie Stability and Idle Recirculation Looo Startup:

1. The prop < . A d changes do not involve a signincant increase in the probability or consequences of an accident prev lously evaluated.

These changes act to prohibit operations which have been found to carry a signincant potential for the formation of core thermal hydraulic instabilities and climinates inappropriate technical specifications for maintaining <50%

recirculation loop flow before starting the idle recirculation pump. As such, operation in compliance with the proposed provisions does not affect any initiating mechanism for previously evaluated accidents or the response of the plant to a previously evaluated accident. The actions tsken lead to placing the plant in a safe condition and are not themselves associated with an initiator for a previously evaluated accident. Therefore, the change does not represent a significant increase in the probability or consequences of any previously evaluated accident.

Enclosure 2 to NRC 98 0003 Page 2

2. The proposed TS changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

As discussed above, the change acts to restrict operations previously allowed. The change also provides remedial actions that act to place the plant in a safe condition.

The actions specified are within the analyzed domain of plant operations. Unless an inst < bility event is in progress, the new allowance to use a core flow increase to leas . the Exit Region is no different than normal plant maneuvering. If an instability event is in progress, the new ACTION 3.4.10.c to scram the reactor takes precedence. The allowance to start an idle loop with the active loop flow >$0% of rated flow has been shown to have no adverse affect on scram avoidance orjet pump riser brace vibration. Therefore, the proposed changes do not create a new or different type of accident.

3. The proposed TS changes do not involve a significant reduction in a margin of safety.

Consistent with the lat,;st 13WROG guidance, the changes act to expand the Exit region compared to the current TS for core thermal hydraulic instability and provide improved remedial actions which promptly terminate the potential for instability. These changes therefore do not involve a significant reduction in a margin of safety.

Post Accident Monitoring:

1. The' proposed TS changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed change does not involve a change in plant design or a change in the manner in which the plant is operated. The long term post accident design requirements of the Neutron Monitoring System (NMS) are not based on operator use for transients with scram, accidents with scram, and other occurrences without scram (Reference 6). For lesser events such as transients without scram, the NMS enhances the operator actions, since successful verification that power is below approximately 3% power can avoid n-n routine operator actions (Reference 6).

These lesser events establish design requirements for the NMS. The failure of this instrumentation during post accident conditions will not prevent the operator from determining reactor power levels. Altemate parameter status will be available from which reactor power may be inferred. Ilased on the multiple inputs available to the operator, sufficient infonnation will be available upon which to base operational decisions and to conclude that reactivity control has been accomplished. This change will therefore not represent a significant increase in the probability or consequences of an accident previously evaluated.

Enclosure 2 to NRC-98 0003 Page 3

(

2, The proposed TS changes do not create the possibility of a new or difTerent kind of '

accident from any accident previously evaluated.

The proposed change does not introduce a new mode of plant operation and does .

not involve the installation of any new equipment or modifications to the plant.

Therefore, it does not create the possibility of a new or different khd of accident .

from any accident previously evaluated.

3. The proposed TS changes do not involve a significant reduction in a margin of safety.

The proposed change climinates a TS listing of a function to tellect the actual safety significance. As such it has no efrect on actual plant operation and thus no impact on any margin of safety.

Based on the above, Detroit Edison has determined that the proposed amendment does not involve a significant hazards consideration.

l l r l

Enclosure 3 to l NRC-98 0003 Page1 ,

ENCLOSURE 3 FERMI 2 NRC DOCKET NO. 50 341 OPERATING LICENSE NPF-43 REQUEST TO REVISE TECIINICAL SPECIFICATIONS TECIINICAL SPECIFICATION CilANGE ,

Attached is a mark-up of the existing Technical Specifications (TSs), indicating the proposed changes (Part 1) and a typed version of the TSs incorporating the proposed changes with a list ofincluded pages (Part 2).

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Enclosure 3 to NRC.98 0003 Page 2 ENCLOSURE 3 - PART I PROPOSED TECIINICAL SPECIFICATION PAGE MARK UP

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11ST 0 FIG RES -

EIG1LRE EMI 3.1.5 1 SODIUM PENTA 80 RATE VOLUME /

CONCENTRATION REQUIREMENTS..................... 3/4 1-21 3.4.1.4-1 D

\ ". .i.""E 'WEA-C

.M . 000E FLOW. . . .3/4. .4.Ga ...........

3.4.6.1-1 MINIMUM REACTOR PRESSURE VESSEL METAL TEMPERATURE VS. REACTOR VESSEL PRESSURE.................... 3/4 4 21 3.4.10 1 THERMAL POWER VS. CORE FL0W.................... 3/4 4 31 4.7.5 1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST..... 3/4 7-11 B 3/4 3 1 REACTOR VESSEL WATER LEVEL..................... 83/437 8 3/4.6.2 1 I LOCAL POOL TEMPERATURE LIMIT................... B 3/4 6 5 8 3/4.7.3-1 ARRANGEMENT OF SHORE BARRIER SURVEY POINTS..... 8 3/4 7-6 5.1.1-1 EXCLUSION AREA................................. 52 .

5.1.2 1 LOW POPULATION 20NE............................ 53 5.1.3-1 MAP DEFINING UNRESTRICTED AREAS AND SITE B0UNDARY FOR RAD 10 ACTIVE EASEQUS AND LIQUID EFFLUEN1$...................................... 5-4 I

FERXI - UNIT 2 xxi Amendment No. pp,/2,37.54,77

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TAntt 4.3.7.5-1 (Continued)

E ACCIDENT MONITORING INSTRUMENTATION StRVEllLANCE Rt90lRt3ENTS

- a m iCAstt CimMMEL CManNtt ortRATIonAL si .

Call 8AAT104 M INSTRUMENT

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13. Standby Cas Treatment System Radiation Monitors

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SCTS - Moble Gas (Low-range) M R I, 2, 3

a. I, 2, 3

b. SGTS - NcSle cas (Mid-range) M R SGTS - AIM-Moble cas (Mid-range) M R 1, 2, 3

c. I, 2, 3

d. SGIS-AIM-MobleGas(High-range) M R "15.CDeleted W **** x caterea - J

16. Primary Containment Isolatio Valve Position h R 1, 2, 3 b DELETS

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REACTOR COOLANT SYSTEM 10tt Rit1RCULAT10W LOOP STARTUP DELE'i&3 11MITING CONDITION FOR OPinATION e 1 n.3.,4.1.4 An idle retirculation c100Mhall m , . m. .. . . .= . . . ..

mot

...-_ be,st,arted salass "ff55Elsf Detue5E ihe'MitirTr' t

4 5Isdie5sei'sW" 3gy, g as space coolant and the bottom head drain line coolant is less than er equal to 1457 ends unless the temperature differential

a. When both loops have been idle,hin the idle loop to be started up between the reactor coolant wit

~ and the coolant in the reactor pressure vessel is less than er equal .

to 50*F, or O b. When only one loop has been idle, unless the troperature differential cula-between the reacter coolant within the id11',_and _o_Perating Y

tion __ loons __is less than gt :equal toJ.J0V 5 . Of 7:t;d.,...

h;p.. fig

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A7PL1CABILITY: OPERAT10kAt C0hD1T10N51, 2, 3 and 4.

Wb. b ACTION:

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WithkWW5M.;th temperature differences @h:'r fYEhenceeding the above limits, suspend startup cf ary idle recircuution loop.

SURvt1!t ANtt 'Rt001REMIN1s _

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[D6/ itT /* N / 6 [6 -

4.4.1.4 TheQ[d"iUOL'A temperature differentials Ord f5 -@ shall be determined to be witnin the limits within 15 minutes prior to startup of an idle retirculation loop.

FERMI . UNIT 2 3/4 4-6 Amendmnt No. 53 9/7/94

i+w.*e4.adm=_ma--eJA--..em.cm.4),_M.-- Je--du . men..w...-m..-ah*-*=6ma.M. he- AAEdM* dea _A# eih.Am4b4A. a s Er A _ A-e_Aa.5m.-ma.Ae. J a A .,a _A4J_mA. lim.msa 4. .am E4 4 _. d m.d a m e a Md-e.a. .ha-e4_JLeM__G_A j

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Amendment No. A1, 87 FERMI - UNIT _2 3/44-6a -

9/7/94

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EXHt SIT Co2E T+1Efu1RL-N'lDEAULIC TA15D151UTY OE EE REACIDR_f00LANT SYSTEM OW4TEL 1MTA6 MAM DE 3/4.4.10 CORE THERMAL HYDRAULIC STABILITY E)(IT EE6)DNS A5 SF6CJF/C.b LINITING CONDITION FOR OPERATION /// 10 Pl60EE 5.4.10-)

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_3.4.10_ _The Reactor core shall no /p:;r:ted-in *:;'-- ..

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i APPLICABILITY: OPERATIONAL CONDITION 1 THE SCNM EE6'oA K M' pdj%C -

SPEt\FIEh IM FIGM 3A.lb-1, l

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a. With the Reactor operating in Ede Switch--ier

the- ^' "'-"-- 'SHUTD0 I osition.

hkC ' mediate THE EX lace. tha Reacter EEGION A 5 SPECWI With the Reactor operatingTQegier

Of it; r: 3.'.10-1.

U hi BA.Ib l, I I

' b.

imediately initt to .:tti=--M)by inserting control i kb rods W OE COPG FLD W

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l_6 5/6 T9S m-EXWh/^ l I

o Ifr-wh44 t4nf..egten-B, ::re the--:1 hydr:eli: in:tability *

[eoccur4-as-evidence-by-APRM-readtea! es:441 sting by gr :tr th: er equal-to-10%-of-RATED-THERMAL-P0WER-peak te pe:E er L**" re: ding +

SEE Ig,EPT 1 oscillat4ng-greater-than-or-4 qual-to-30-watts /c:' p::k t; p;;t, imediately place-the-Reactor-Mode-Switch-in-the-SHUID0'."' pesitier.. g FD9 EdVISEh ^

TiXT -

i SURVEILLANCE REOUIREMENTS i

4.4.10.1 The provisions of Specification 4.0.4 are not applicable. ~

^"ra '- " >' "e---

.4 10.2 Wit er-than-405-of-RATED--THE"'"~"2 l'"I"1 F ow- ess-than-50%-of-Rated-Core Flee f v- i' +'-* *'- -- L l

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9/7/94 FERMI 3 Amendment No. 53

___- __ UNIT 2-._ _ _ _ . ____/4 4-30_ _ _ _

m

[ INSERT 1 l (Replace the text in TS Section 3.4.10.c, Page 3/4 4 30 with the text below) ]

l

- =

3 , 4

c. If core thermal hydraulic instability occur as evidenced by a sustained increase in APRM or LPRM peak-to-peak noise level reaching 2 or more times its initial level,

. and occurring with a characteristic period ofless than 3 seconds, immediately place -

. the Reactor Mode Switch in the Shutdown position.

< M x J INSERT 2 (Replace the text in TS Section 4.4.10.2, Page 3/4 4-30 with the text below) 4.4.10.2 When operating within the Stability Awareness Region as specified in Figure 3.4.10-1, verify that the reactor core is not exhibiting core thermal hydraulic

instability by monitoring APRM and LPRM signals immediately and at least ,

.. once every hour, k -

! INSERT 3 (Add the below text to the bottom of Page 3/4 4-30)-

y e

' Restarting an Idle Recirculation Loop or resetting a Recirculation Flow Limiter are not acceptable methods ofimmediately increasing core flow to leave the Exit Region.

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i 3/4 4-?' Amendment No. JJ,87 F N I - UNIT 2 9/7/94

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l 80 Stability Region Descriptions Scram Regiort i } ~d

>96% Rod Une. <40% Core Flow e 70-~ ' 7n 65 h 65 Exit Regiort

} 4 Not in Scram Region 1 60 y ( 60-:- ~

- and -

._ I >67% Rod Une. <40% Core Flow 55-h O

>77% Rod Une. <45% Core Flow

>103% Rod Une, <50% Core Flow ,

50

) _ 50 f :

, (U : 2 Stability Awareness Region:

-45 de b@

45h y ^

Notin Scrarn or ExR Regions ;

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>62% Rod Une. <45% Core Flow 1

2 >72% RM N. <M Core h 35J 35

>98% Rod Une. <55% Core Flow I :

30 y .

2 25 2

,,,, , ,, ,,,, ,,,, ,,,, 25 l 35 40 45 50 55 60 30 Core Flow (% Rated)

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THERMAL POWER VERSUS CORE FLOW THIS "N6VJ" F/tTOPd TO FIGURE 3.4.10-1 gx tcTudh P'

i 3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 RECIRCULATION SYSTEM (Continued)

Sudden equalization of a temperature difference greater than 145'F between the reactor vessel bottom head coolant and the coolant in the upper region of the reactor vessel by increasing core flow rate would cause undue stress in the -

.. reactor vessel bottom head. - _ _

/. -

acquir;;;r.t: cr i ;;;;d t: pr;hibit idi: 1::; :t:rt ; i:^;; th: 77% red j lir.: t: mi.;ici:: th: ; t:nti:1 fer '-iti: tin; : r: th:r=:1 Sydr: ltc 1 ...unia.m -

[~ ~ ' ' ' y ' ~ x n -

3/4.4.2 SAFETY / RELIEF VALVES 7s { j j -{ gg The safety valve function of the safety / relief valves operate to prevent the reactor coolant system from being pressurized above the Safety Limit of 1325 psig in accordance with the ASME Code. A total of 11 OPERABLE safety / relief valves is required to limit reactor pressure to within ASME !!! allowable values for the worst case upset transient.

Demonstration of the safety / relief valve lift settings will occur only during shutdown and will be performed in accordance with the provis' ions of .

Specification 4.0.5.

The low-low set system ensures that a potentially high thrust load (designated as load case C.3.3) on the SRV discharge lines is eliminated during subsequent actuations. This is achieved by automatically lowering the closing setpoint of two valves and lowering the opening setpoint of two valves following the initial opening. Sufficient redundancy is provided for the low-low set system such that failure of any one valve to open or close at its reduced setpoint does not violate the design basis. .

e e

4 FERMI - UNIT 2 B 3/4 4-la Amendment No. EJ 87

_ J 5CBRI5

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LGME THFC'2EGlom SEFDEE O.SclLLATIONS .DEtlELOf

~5 VI AND EG ACM .AN UM ACCEPT.AE LE MA6MlTODE. -j '

MJ 1 REACTOR COOLANT SYSTEM h f m ~J'y 4r M M IhT h% ci NJ279 M;b'\

4 Basts MN M ~A 8[ .

XENDN f N CEMTPkTlDN 3/4.4.10 CORE THERMAL HYDRAULIC STABILITY, ---

ux noise in a stable

)bqj J BWR modecores which istypically operate due to random with boiling andthe flowpresence noise. As Ch of global ow xy < conditions are changed, along with other systen arameters' hydraulic /r'4Tet r %e;; kinetic #; 1 ,

i sQ I i

subcooling, power distribution, etc.) the therma fa feedback mechanism can be enhanced such that randos perturbations may result

] in sustained limit cycle or divergent oscillations in power and flow.

EW f Two major modes of oscillations have been observed in BWRs. The first mode is Cb the fundamental or core wide oscillation mode in which the entire core DD oscillates in phase in a given axial plane. The second mode involves regional

, v1 g ' with the other half. oscillation Studiesinhave which one halfthat indicated of the core oscillates adequate margin to180 the degrees out of phase Ug j Saf aty Limit Minimum Critical Power _Batio (SLMCPR) may not eR ring

} oscillations. #e SCPW)m EM EE61oN_C t-\I/ M P.e ten a :nd a :f Fj;;r:T.? e rese7EiiTeast stable conditions of the fi; ant (high power / low flow). h on n sad " are usually entered as the result of a plant transient (for example, recirculation pump trips) and therefore ar.

Al y C kscerallynotconsideredpartofthenormaloperatingdomain.i.1,itySincegal events (includingb. test experi

S9 y ese regions are avoided to minimize the possibility of encoun ing W Ct osci lat s t t 11 challenging the SLMCPR. Therefore, intentional

@y v1 operation in ' ie.s A er " s not allowed.

certain_ abnormal con It is recognized that during ons within the plant, it may become r.ecessary to enter jgb d'T a A er rx4 cou d m ac for the plant purpose safety or for theofpurpose protecting equipment of protecting a safety which, or fuelwere it, g. to fail, h F Lu operating limit. In these cases, the appropriate actions for the region entered would be performed as required.

7- ggg gg Ed\)lst m Most oscillations that have occurred during test g and operation have N occurred at or above the 96% rod ikwi h core / ow near natural circulation. j This behavior is consistent with analysts whic / predict reduced stability 4 As core flow is increased or

\ margin with ppwar_.gecreased.

increasing power er decreas oscilla ng fl' w.

ons occurring will decrease.

b "--i ca A ~c f Ti d21._the_ probability qt,the ity

4. E bound majof the stability events and X I'N sts observed D E B s. Since"d

ifin epresents the least stable region

'[Ul of the power / flow operating domain, e otential to rapidly encounter la d magnitude core thermal hydraulic oscillations is increased. Durino --

pg nj ;(c~ransients, th_e_olerator may not have sufficient7p time y 7- toMu:ll t

, ontrol req ^S

%m_itigate-ths oscilliticGfge theynbs ungcceptible mnited:./ ThereT5re, the prompt action of manuallv scramrning th_e plant wh'eiiV

~

, b_

A T3-eater:M r: quired to en:'.'re protecti:n f th; SE:MC-PR. g

' v; z ' THC SCR4tn EE& SON I5 ENTEPEb O2 \NGER 050.ILLATIoti 5 e) 9 i A26 DETdCTED 1126CO R.6D To E4_TdE6 PEole CTION j

'f OF TRE SLN) CP - - = ~ -

FERMI - UNIT 2 B 3/4 4-8 Amendment No. EJ, 87

INSERT 4

_ (Add the follcwing sentence to the beginning of the third paragraph of

, Bases Section 3/4,4,10, Page B 3/4 4 8) v v Figure 3.4.10-1 specifies the Scram, Exit, and Stability Awareness Regions by providing

a description of these region boundaries based on Rod Line, Core Flow and Thermal

. Power.

A --

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4 1

5 1

5 4

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- . . . . . - . - . - . - - _ _ - .. ~. - .- - -. . . - - .-.

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REAtTOR COOLANT SYSTEM l

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BASES i i

3/4.4.10 CORE THERMAL HYDRAULIC STABILITY (Continued) ;

Based-on-tcet rd i;:r:ti.q; er;r!E:c, the fr;:rry ef cors-thermal N !

hydraul-ic-oscillations-is-ess in .".gi= 5 the 5 5;t= A. 5 = y 7:ti : =; ,

espected rd-;;ndt:hd te h hur 5 thh n;t= :hr 5;i= " :==: :

inx ;== rd hi-tr fh: nr.;; thx t;in A. A h;, th = rgir. t; t h .

SEMGPR will typi :Ily h l=;n u S;i= " the 5 5;i= A. "ith = = g i marg!r. te SLTPR rd a !rr ;;:id!!ity Of = ilhti=:, citir.; S;hr. 2 by centrel- red unrtin i: jntift:d. Sxx=, if ;niluti=; ere eteeried- 1 dih exiting-Region 4, th n:t= will 5 22=11y :;md.

Th pet =tial-forcere-thermal hydr =1h :=ilhti=: t: ;== nt:i5 ef ,!

-S;i=:' A =d4 h " cry ill rd thr fere :;rbl n; f rrrt: r: =t

= :::try eettid: ef thn: n;i=:.

i

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l l

Based on test and operating experience, the frequency of core thermal l hydraulic oscillations is less in the Exit Region than in the Scram Region.

Core decay ratios are expected and predicted to be lower in this region since -

the Exit Region covers a lower power and higher flow range than the Scram

/ Region. Also, the margin to the SLMCPR will typically be larger in the Exit Region than in the Scram Region. With more margin to SLMCPR and a lower probability of oscillations, leaving the Exit Region by immediate control rod

. insertion or core flow increase is justified. Restarting an Idle i Recirculation Loop or resetting a Recirculation Flow Limiter are not acceptabie methods of increasing core flow to leave the Exit Region because these actions would not support timely completion of this immediate action. I If /l oscillations are observed at any time, the reactor will be manually scrammed. 'l l I.

The potentihl for core thermal hydraulic oscillations to occur outside of j 4

\ .

the Scram or Exit Regions is small, but could occur. Therefore, frequent i monitoring of APRM and LPRM signals is appropriate when operating in the j Stability Awareness Region. /

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I 9/7/94 FERMI - UNIT 2 B 3/4 4 9 Amendment No. 53'

Enclosure 3 to NRC-98 0003 ,

Page 3 --

ENCLOSURE 3-PART2 i PROPOSED TECilNICAL SPECIFICATION REVISED PAGES INCL,UDED PAGES:

xxi 3/4 3-61a 3/4 3-63a 3/44-6 3/44-6a 3/4 4-30 3/4 4-31 B 3/4 4-1a B 3/4 4-8 B 3/4 4-9 0

LIST OF FIGURES FIGURE PAq[

3.1.5-1 SODIUM PENTABORATE VOLUME /

CONCENTRATION REQUIREMENTS................... 3/4 1-21 3.4.1.4-1 DELETED...................................... 3/4 4-6a l 3.4.6.1-1 MINIMUM REACTOR PRESSURE VESSEL METAL TEMPERATURE VS. REACTOR VESSEL PRESSURE.................. 3/4 4-21 3.4.10-1 THERMAL POWER VS. CORE FL0W.................. 3/4 4-31 4.7.S-1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST... 3/4 7-21 B 3/4.3-1 REACTOR VESSEL WATER LEVEL................... B 3/4 3-7 8 3/4.6.2-1 LOCAL P0OL TEMPERATURE LIMIT................. B 3/4 6-5 B 3/4.7.3-1 ARRANGEMENT OF SHORE BARRIER SURVEY POINTS... B 3/4 7-6 5.1.1-1 EXCLUSION AREA............................... 5-2 5.1.2-1 LOW POPULATION Z0NE.......................... 5-3 5.1.3-1 MAP DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADI0 ACTIVE GASE0US AND LIQUID EFFLUENTS.................................... 5-4 1

FERMI - UNIT 2 xxi Amendment No. Ep,fJ,EJ,E/,77,

,, TABLE 3.3.7.5-1 (Continued)

' hh - ACCIDENT MONITORING' INSTRUMENTATION MINIMUM APPLICABLE..

EE RE' 'E0 NUMBER CHANNELS OPERATIONAL INSTRUMENT or iHANNELE OPERABLE CONDITIONS ACTION

13. Standby Gas-Treatment' System Radiation Monitors

a. SGTS - Noble Gas (Low-range)# 1/0PERABLE l/0PERABLE 1, 2, 3 81 SGTS subsystem SGTS subsystem

b. SGTS - Noble Gas (Mid-range) 1/0PERABLE 1/0PERABLE 1, 2, 3 81-SGTS subsystem SGTS subsystem-

c. SGTS - AXM-Noble Gas (Mid-range)- 1/0PERABLE 1/0PERABLE 1, 2, 3 81 SGTS subsystem SGTS subsystem R**

d. SGTS - AXM-Noble Gas (High-range) I/0PErABLE 1/0PERABLE 1, 2, 3 81 SGTS su:2ystem

?> SGTS subsystem a

EI 14. Deleted '

15. Deleted '

16. Primary Containment I:olation Valve Position 1/ valve 1/ valve 1, 2, 3 82 3I '

.i r+

F o #

Also included in the OFFSITE DOSE CALCULATION MANUAL.

t I

, . - ~

TABLE 4.3.7.5-1 (Continued) f 7

jE ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS

APPLICABLE CHANNEL CHANNEL OPERATIONAL EE ~

Z CHECK CALIBRATION CONDITIONS I

INSTRUMER n> \

13. Standby Gas Treatment System Radiation Monitors SGIS - Noble Gas (Low-range) M R 1, 2, 3 a.

SGIS - Noble Gas (Mid-range) M R 1, 2, 3 b.

SGTS - AXM-Noble Gas (Mid-range) M R 1, 2, 3 c.

SGIS - AXM-Noble Gas (High-range) M R 1, 2, 3 d.

14. Deleted l

15. Deleted M I, 2, 3

16. Primary Containment Isolation Valve Position M R Y

0

=

E I

.af

l

? ,.

j Rt?tTOR COOLANT SYSTEM IDLE RECIRCULATION LOOP STARTVP LlHITING CONDITION FOR OPERATION 3.4.1.4 An idle recirculation loop shall not be started unless the temperature j differential between the reactor pressure vessel steam space coolant and the bottom head drain line coolant is less than or equal to 145'F, and:

a. When both loops have been idle, unless the temperature differential between the reactor coolant within the idle loop to be started up and the coolant in the reactor pressure vessel is less than or equal to 50'F, or

b. When only one loop has been idle, unless the temperature differential between the reactor coolant within the idle and operating recirculation loops is less than or equal to 50*F.

I APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 and 4.

AC.TLQH:

With temperature differences exceeding the above limits, suspend startup of any j idle recirculation loop.

SURVEILLANCE RE0VIREMENTS 4.4.1.4 The temperature differentials shall be determined to be within the j limits within 15 minutes prior to startup of an idle recirculation loop.

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FERMI - UNIT 2 3/4 4-6 Amendment No.EJ

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FIGURE 3.4.1.4 1 - DELETED l

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i FERMI - UNIT 2 3/4 4-6a Amendment No. E3,E7,

9 REACTOR COOLANT SYSTEM n

3/4.4.10- CORE THERMAL HYDRAUL1C STABILITY LIMITING CONDITION FOR OPERATICN s 3.4.10 The Reactor core shall not exhibit core thermal hydraulic instability or be operated in the Sc am or Exit Regions as specified in Figure-3.4.10-1.

APPLICABILITY: OPERATIONAL CONDITION 1 ACTION:

a. With the Reactor operating in the Scram Region as specified in Figure 3.4.101, immediately place the Reactor Mode Switch in the Shutdown position. [

With the Reactor operating in the Exit Region as specified in Figure

b. ,

3.4.10-1, immediately initiate action to leave the Exit Region by '

inserting control rods or by increasing core flow.*

c. If core thermal hydraulic instability occurs as evidenced by a sustained increase in APRM or LPRM peak-to-peak noise level reaching 2 or more times its initial level, and occurring with a characteristic period of less than 3 seconds, immediately place the Reactor Mode Switch in the Shutdown position.

SURVEILLANCE RE0VIREMENTS 4.4.10.1 The provisions of Specification 4.0.4 are not applicable.

4.4.10.2 When operating within the Stability Awareness Region as specified in j Figure 3.4.10-1, verify that the reactor core is not exhibiting core thermal hydraulic instability by monitoring APRM and LPRM signals I immediately and at least once every hour. !

Restarting an Idle Recirculation loop or resetting a Recirculation '

Flow limiter are not acceptable methods of immediately increasing core I flow to leave the Exit Region. !

FERMI - UNIT 2 3/4 4-30 Amandment No. El,

I *

'en 85 85-L 5

8-i V/////) &\kN i 1 80 Stability Region Descriptions

_4 -75 N '75 Scram Region:

f

>96% Rod Line. <40% Core Flow

\ 170 m 70-65-x' h :

65 Exit Region:

Not in Scram Region 60

$ 60- / .

% \- - and -

[ b 55

>67% Rod Line. <40% Core Flow 55- ' . .

>77% Rod Line <45% Core Flow

/ >103*A Rod Line. <50% Core Flow O \

\ 50 s O- 50-Stability Awareness Region:

45 Not in Scram or Exit Regions O

45- ~

N _40

- a4

}- 40- % _

>62% Roc Line <45% Core Flow

>72% Rod Line. <50% Core Flow

-35 352 k >98% Rod Line <55% Core Flow g 25 ! .... ....,...............25 60 35 40 45 50 55 g 30 ita Core Flow (% Rated)

E THERMAL POWER VERSUS C9RE FLOW FIGURE 3.4.10-1 w.

3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 RECIRCULATIOh SYSTEM (Continued)

Sudden equalization of a temperature difference greater than 145'F between the reactor vessel bottom head coolant and the coolant in the upper region of the reactor vessel by increasing core flow rate would cause undue stress in the reactor vessel bottom head.

3/4.4.2 SAFETY / RELIEF VALVES The safety valve function of the safety / relief valves operate to prevent the reactor coolant system from being pressurized abeve the Safety Limit of 1325 psig in accordance with the ASME Code. A total of 11 OPERABLE safety / relief valves is required to limit reactor pressure to within ASME III allowable values for the worst case upset transient.

Demonstration of the safety / relief valve lift settings will occur only during shutdown and will be performed in accordance with the provisions of Specification 4.0.5.

The low-low set system ensures that a potentially high thrust load (designated as load case C.3.3) on the SRV discharge lines is eliminated during subsequent actuations. This is achieved by automatically lowering the closing setpoint of two valves and lowering the opening setpoint of two valves following the initial opening. Sufficient redundancy is provided for the low-low set system such that failure of any one valve to open or close at its reduced setpoint does not violate the design basis, e

I FERMI UNIT 2 B 3/4_4-la Am@ndern A E _lIL

.o REACTOR COOLANT SYSTEM BASES 3/4.4.10 CORE THERMAL HYDRAULIC STABILITY BWR cores typically operate with the presence of global flux noise in a stable mode which is due to random boiling and flow noise. As the power / flow conditions are changed, along with other reactor parameters (xenon concentration, subcooling, power distribution, etc.) the thermal hydraulic / reactor kinetic feedback mechanism can be enhanced such that random perturbations may result in sustained limit cycle or divergent oscillations in power and flow.

Two major modes of oscillations have been observed in BWRs. The first mode is the fundamental or core-wide oscillation mode in which the entire core oscillates in phase in a given axial plane. The second mode involves regional oscillation in which one half of the core oscilhtes 180 degrees out of phase with the other half. Studies have indicated that adequate margin to the Safety Limit Minimum Critical Power Ratio (5LMCPR) may not exist during oscillations. j Figure 3.4.10-1 specifies the Scram Exit. and Stability Awareness Regions by .j providing h description of these region boundaries based on Rod Line, Core ;

Flow and Thermal Power. The Scram and Exit Regions represent the least stable conditions of the plant (high power / low flow). The Scram and Exit Regions are !

usually entered as the result of a plant transient (for example, recirculation pump trips) and therefore are generally not considered part of the normal operating domain. Since all stability events (including test experience) have occurred in either the Scram or Exit Regions, these regions are avoided to l minimize the )ossibility of encountering oscillations and potentially challenging tie SLMCPR. Therefore, intentional operation in the Scram or Exit '

Regions is net allowed. It is recognized that during certain abnormal conditions within the plant, it may become necessary to enter the Scram or Exit Regions for the purpose of protecting equipment which, were it tu fail, ,

could impact plant safety or for the purpose of protecting a safety or fuel operating limit. In these cases, the appropriate actions for the regien entered would be performed as required.

Most oscillations that have occurred during testing and operation have occurred at or above the 96% rod line with core flow near natural circulation.

This behavior is consistent with analyses which predict reduced stability l margin with increasing power or decreasing flow. As core flow is increased or power decreased, the probability of oscillations occurring will decrease. The '

Scram and Exit Re io I observed in GE BWks.ns Since bound the Scram the majority of represents Region the stabilitythe events least and tests stable !

region of the power / flow operating domain, the potential to rapidly encounter large magnitude core thermal hydraulic oscillations is increased. During transients, the operator may not have sufficient time to leave the Scram '

region before oscillations develop and reach an unacceptable magnitude.

Therefore, the prompt action of manually scramming the plant when the Scram l Region is entered or when oscillations are detected is required to ensure ,

protection of the SLMCPR.

l FERMI - UNIT 2 B 3/4 4-8 Amendment No. 5 , R7,

REACTOR COOLANT SYSTEM BASES 3/4.4.10 CORE THERMAL HYDRAVLIC STABILITY (Continued)

Based on test and operating ex)erience, the frequency of core thermal hydraulic oscillations is less in t1e Exit Region than in the Scram Region.

Core dtcay ratios are expected and predicted to be lower in this region since the Exit Region covers a lower power and higher flow range than the Scram Region. Also, the margin to the SLMCPR will typically be larger in the Exit Region than in the Scram Region. With more margin to SLMCPR and a lower probability of oscillations, leaving the Exit Region by immediate control rod insertion or core flow increase is ;p.,tified. Restarting an Idle Recirculation loop or resetting a Recirculation Flow limiter are not acceptable methods of increasing core flow to leave the Exit Region because these actions would not support timely completion of this immediate action. If oscillations are observed at any time, the reactor will be manually scrammed. ,

The potential for core thermal hydraulic oscillations to occur outside of the Scram or Exit Regions is small, but could occur. Therefore, frequent monitoring of APRM and LPRM signals is appropriate when operating in the Stability Awareness Region.

l FERMI - UNIT 2 B 3/4 4-9 Amendment No. JJ,

_ ___ _ _ _ _ _ _ _ _ _ _________

ML20199H937

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