ML20085N099

From kanterella
Jump to navigation Jump to search
Proposed Tech Specs Supporting Implementation of BWR Thermal Hydraulic Stability Solution
ML20085N099
Person / Time
Site: Monticello Xcel Energy icon.png
Issue date: 06/22/1995
From:
NORTHERN STATES POWER CO.
To:
Shared Package
ML19330G080 List:
References
NUDOCS 9506300047
Download: ML20085N099 (18)
v • d • e


Text

i

.M.

o l

'f Exhibit B [,

' Monticello Nuclear Generating Plant l License Amendment Recuest dated June 22.1995 Proposed Changes Marked Up on Existing  :

Technical Specification Pages  ;

Exhibit B consists of the existing Technical Specification pages with the proposed changes _ l marked up on those pages. Existing pages affected by this change are listed below: 5 U l 15 l 107 (Insert text for page 107 provided on  ;

separate page) l 108  :

109 114 (Insert text for page 114 provided on separate page) .

211 249b 4

9506300047 950622

~

DR ADOCK 05000263 PDR

- m

.a

- f.

Bases Continued:

For analyses of the thermal consequences of the transients, the Operating MCPR Limit (T.S.3.11.C) is conservatively assumed to exist prior to initiation of the transients.

This choice of using conservative values of controlling parameters and initiating transients at the design power level, produces more pessimistic answers than would result by using expected values of control-parameters and analyzing at higher power levels.

Deviations from as-left settings of setpoints are expected due to inherent instrument error, operator setting error, drift of the setpoint, etc. Allowable deviations are assigned to the limiting safety system settings for this reason. The effect of settings being at their allowable deviation extreme is minimal with respect to that of the conservatisms discussed above. Although the operator will set the setpoints within the trip settings specified, the actual values of the various setpoints can vary from the specified trip setting by the.

allowable deviation.

A violation of this specification is assumed to occur only when a device is knowingly set outside of the limiting trip setting or when a sufficient number of devices have been affected by any means such that the automatic function is incapable of preventing a safety limit from being exceeded while in a reactor mode in which the specified function must be operable. Sections 3.1 and 3.2 list the reactor modes in which the functions listed above are required.

A. Neutron Flux Scram The average power range monitoring (APRM) system, which is calibrated using heat balance data taken during steady state conditions, reads in percent of rated thermal' power (1670 MWt). Because fission chambers provide the basic input signals, the APRM system responds directly to-average neutron flux. During transients, the instantaneous rate of heat transfer from the fuel (reactor thermal power) . is less than the instantaneous neutron flux due to the time constant of the fuel. Therefore, during abnormal operational transients, the thermal power of the fuel will be less-than that indicated by the neutron flux at the scram setting. Analyses demonstrate that, with a 1204 scram trip setting, none of the abnormal operational. transients analyzed violate the . fuel Safety Limit and there is a substantial margin from fuel damage. 5:r:fere, th: :: f f1= rsfer;.. d

= trip pr: cide: : re 2dditie- 21 : rgi- A .g. Q 4g _ W e g qd, 4 Sc u .

( Spcd fi cdem 'L . 3. A . l) covides p ro b,cA ion +o 9 h ,, ,g g 15 6 c- k'#i i" -N c. g { ; kc} c V C-P h o h ' A REV-135 1/2/03 br\c.c WN d bf d/d A I C- i WsbM h ky, 1

i 1

1

__-m____.a_._____ - . _ . _ _ . -_ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ - . _ . - _______-_m____en , - - . _ e* -- w--,w-- *.+--,,.-e.~.m.,. ,,e. v ,v w wmm..,.we..rin-r,.,i.iue w er - --w w.w-w w -+v-,,..-.we.,m-ww.ww-. m,-w-=>--

.w l

3.0 LIMITING CONDITION FOR OPERATION 4.0 . SURVEILLANCE REQUIREMENTS I F. Recirculation System F. Recirculation System I 3/. The reactor may be started and operated. or 1. See Specification 4.6.G operation may continue with only one g ](- recirculation loop in operation provided that:  ?. % f-11--ing b : lin: n:12: 1;.;1: ril? 5:

ch::in;d p.-i;r te y e . -il... - m;. .ol, . . ..

a. The following changes to setpoints and recircu12:isn pup in ;pu;;i: 2: ; ::::

l '3, T. I.I safety limit settings will be made within .

thern:1 p;r:r g;; t: th:x th t p: if * '

i 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after initiating operation with 3 Figur; 3. 5.1 :: rith ; :::: fler gre- rr

' .pc-! 3 6.b'1 only one recirculation loop in operation. than = pro.id J ;.L i L-..lb.. 1- . L- -

not L; .. ._; Lii h d i..;; ti; 12:t-::::

o c. 1. The Operating Limit MCPR (MCPR) will refueling. h ;elin; r:12 : rill 5: ' r'- r p be changed per Specification 3.11.C. with-only e recire"' -'-* p -"-- t =g .

p 2. The Maximum Average Planar Linear lleat 2 E tchli:5 : S:::l'n: : rr pl

Generation Rate (MAPIJICR) will be le-1 changed 2: neted in Table 3.11.1.

FC / Cec.e-i- k r A m 3 IIM . L. L;. Lli;L i;;:lin: .""' ---' ' DD M

3. The APRM Neutron Flux Scram and APRM ...uimou flua ...... l...l.

Rod Block setpoints will be changed as noted in Specification 2,3.A and Table 3. "ith :nly ::: recire21st!-- 1-ar in 3.2.3. ;per;;ica ;; ; ;;r; ti;r 2 ;l p: :: gr :t:

th... th_: ;p ;ifi:d in Tigrrr

  • 1 -r ='th .
b. Total : : fler uill be  !.tcined greater  ; ._ ; rc. fi:- grr-*-- ^ -- '"- -

-:--'- the than 39 bwhenwore-thermal-power-14-above 4el-lowing-noist--1.. ls-ee-leastwc per S the limit. specified in Eigur 1 5.1. he"r p-ri-i -- ' "ithin-30-minutes--aliter-a TedaicJ < fcca 4(c. Abo A '3 . J~. #, I cu J cc :- th:=:1 F ;r i .cr;;;; cf sr ater-than M os/6 3 . f.F. 'A afc re.t 5' "' r-t-d thermal-Power.

  • N N eo (c a. d o r (,,,pol M i "

eaP rc.c; rc Aba i =es. i^ o

.. c. qi y s ; M Tec L M v e 7' ' 9 ,o 4[ 'c 5.

Cor; F l oti I" LeiSo liiaI5-

=-end-LeRM wt -- .~ 6 1 1 .

gf 3.S'. C.I c+4 '3.S. F. A by I

  • S* 'S O c.o d ral rods u s Oc^ c- ostl y "

o pecd wh o, s 3. 6. A. 2. ud 3. 5, f. 3 3.5/4.5 4o< op( mb o^ v e% o n l y oM 107 , , , , _

REV u,_, ,nA v c.c5 c c.d Ak s > 1 oog of io * <c.u h c s 6.0 de. 96.cd

, L HA E4h sh a bvs.

y

+

.t insert Text for Page 107 F. Recirculation System

1. Intentional entry into the stability exclusion region of the power-flow map defined in the Core Operating Limits Report (COLR) is prohibited. If entry into the stability exclusion region does occur, immediately perform one or more of the following until the stability exclusion region has been exited:
a. Insert control rods,
b. Increase the speed of an operating recirculation pump.
2. Entry into the stability buffer region of the power-flow map as defined in the COLR is prohibited unless the power distribution controls as defined in the COLR are in effect. If the power distribution controls are not in effect and entry into the stability buffer region does occur, immediately perform one or more of the following until the stability buffer region has been exited:
a. Insert control rods,
b. Increase the speed of an operating recirculation pump.

. . . - _ - . _ - . _ - - . - . - . . . . . . . - - . - . . . . - - . .-. . . , . . - . . - . . . . . . - - . . - . . . = - . - - . . - . . - . . . - - - . - . . - . . . .

r "

3.0 LIMITING CONDITION FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS

c. "r'cr :: ::ntirr:d :; retien ith enly ena recirculetter p" p ir eperation,
1. the urvettience requirement: ef specification A _S F.2 rhell be- met-within.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or,
2. entier ch:11 he taken-to+.
a. reducc total car; fic; te les than-45%

and ;-

b. -reduce cere thermal-power-to-less-than-the 1-imit ; ccified ir. Pigure-3.5.1.

d _ If-the-core-plate AP-noise-level-Ls-f4und-to-be.

-greater-than-1,0-pei end-2-t-imee-Lee-established-

.b as e l i ne -du ri ng-the-pe r formenc e--ef--Spec i fication W F & immediately-initiate corrective. action.

-d recter: the ncir levele te -ithir the r49utred li-it- =ith!- 2 be re by A-creasing-core -

Elow-and/er initiating en erderly reduction-of

. core thermal pe"er- by insertin; control- rods.

c. If th: " P"J' and/cr LP"" n;etree-f-lux-noire levels

-are-found-te be greater-than-three-times- their

-established her line valuer during the-performance of-Specificatier ^.5,F,3, irmoediately initiate

--eerreet-tv: :: tier te restore-the-noise-levels-to withir.-th; rcquired limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> by increasing core-flow and/or-initiating-an-orderly reduetten cf cere thermal-power--by-inserting-control-rods..

3.5/4.5 - 108

, ,o,.

.REV ...

n s. 4 wy sg s L

___---_____.-__m.-____ _ _ _ . _ _ _ _ _ _ _ _ _ . . _ _ - _ _ _ _ _ _ _ _ _ -_ - _ _

m-_.e-.-..--._-m--- - - - . - - , - . - . . . ~ - - .~.. _ ....-~-.-- - -- ._ --

-.-4 _ . - - - , __ m_._ ._ _ - . . _ _ _ _ _ _ _ _ m

. i I

i 0, l e.-c w ,w .

- n.o. \ m. c. m  :

s %b.

- t. v ct. w <

80 r

\ i b

I I

i

<,,3 9 % F J. ,i j ,i l 70- ~ -- ---

L ~ ~ - ~ ~ - ~ l ~ ~ ~ -~ ~ -- ~ E - - ~ ----: - L ~ -- - ~ ~~~ i i

.  !  : j i  ;

L Opera n  !  !

6 0'-..

NOT Alle d

\.... .,i e-(0.sa7)F+1s.2 .............J'.................

e

'O i P= Core Thermal  !

8 .  ! Power (% of  :

! Rated) i ce  :  ;

1 gg  :

~ - - ~

F= Indicated Total ~ ~ ~ ~ ~ ~ - - ~ ~ ~ - -

  • ~

o 50~

~ -- -

i .~---- Core Flow {% of  :* .

Mated) p w

l-i  :  :

a3 ,

40 . . . .

C3  :

Q.,

a . .

E  :

- Operation Allowed .

E a0- .........! . - - - -

if TS 3.5.F.1.c and TS 4 5.F.3 Surveillance f: l l

Is' Complete Also See o .

C NO Restriction i TS 4.5.F.1.a U on Operatio ns , ........ .........i... ............ .. ..................;.. ................

2o. ...

.  :  :  : g i .  :  !

,0_

7.....\..x.........l...............

\

! .' i \  !

< l45% Fl .  :

T 0 . .' > >

.\ 80 30 40 50 60 70 20 Indicated Core Flow (% of Rated) .  !

e e o f- 2, Q,,,.wey[o y a m . 3. - -fman .~ y A n-r--.

m n 4.f e. .n.

- 4 - . . - . . - . . - -

Figta.-io v.u i

-.. _n- -,.. r,. ,,.

F V TT U 3 J 6 e %# vu vwe vv I

109 3.5/4.5 V , *..!

,. n ttofe, s # r # =

4-Bases 3.5/4.5 Continued:

F. Recirculation System paci f f < = tia=_ ?_5.F.1 is based-upon-preriding assurance-thee-neutron-flux li=it cycle. ;;cille;.ie..e " " d. .

ha small probability of occurring in the high power / low flow corner of the operating domai

,[' 6 M detecte d suppressed. Under certain high power / low flow conditions that could occur during a Mg recirculatio'li'pugp trip and subsequent Single Loop Operation (SID) where reverse floy edisInrs in inactive jet Wg pumps, a hydraulichactor kinetic feedback mechanism can be enhanced such that sus'tained limit cycle oscillations of flow no't!r with peak to peak levels several times normal y es are exhibited. Although b .4 large margins to safety limit e maintained when these limit cycle 11ations occur, they are to be ea monitored for, and suppressed when us noise exceeds the threy tiliie baseline value by inserting rods and/or p+ vincreasing coolant flow. The line in F'llure 3.5.1 is based-on the 80% rod line below which the probability

a. of limit cycle oscillations occurring is negligible.

./

APRM and/or LPRM oscillations in excess foth'ose specifie -in Specification 3.5.F.1.e could be an indication that a condition of thermal' hydraulic instability exists and that appropriate remedial action should be taken. By restricting-core flow to greater than or equal N 39% of rated, which corresponds to the core flow at the 80% rpd'line with 2 recirculation pumps running at R4 mum speed, the region of the power / flow map where Ifs'e oscillations are most likely to occur is avoided (Rb 1).

Above 45% f ated core flow in Single Loop Operation there is the potential to set up h flow-indue noise in the core. Thus, surveillance of core plate AP noise is required in this reg f the r zer/flez =ap-t-- rier u he epercter: te t eke --epprepele te-reme#el-ac t i:-- if : :P  ::nditi:n :xie Specification 3.6. A.2 governs the restart of the pump in an idle recirculation loop. Adherence to this specification limits the probability of excessive flux transients and/or thermal stresses.

Rekr-enec c : 1r-General Electric-Service Information-Let4er Nov480,- R 1,- F bruary-101 -1984-NEXT PAGE IS 121 3.5/4.5 Bases 114 REV 429--4/9/91-

'< g

t INSERT TEXT FOR PAGE 114 i

The reactor is designed such that thermal hydraulic oscillations are prevented or can be readily detected and suppressed without

, exceeding specified fuel design limits. To m,nimize i the likelihood of a thermal-hydraulic instability, a power-flow exclusion region, to -

3 be avoided during normal operation, is calculated using the approved methodol as stated in specification 6.7.A.7. Since the .

, exclusion region may change each fuel cycle the limits are contained in the Core og Limits Report. Specific directions are '

provided to avoid operation in this region and to immediately exit upon an entry. tries into the exclusion region are not part of normal operation. An entry m,ay occur as the result of an abnormal event such as a single recirculation pump tnp. In these events, .

operation in the exclusion region may be needed to prevent equipment damage, but actual time spent inside the exclusion region is minimized. Though operator action can prevent the occurrence and protect the reactor from an instabikty, the APRM flow biased scram function will suppress oscillations prior to exceeding the fuel safety limit. ,.

~ Power distribution controls are established to ensure the reactor is operated within the bounds of the stabihty analysis. With these controls in place, there is confidence that an oscillation will not occur outside of the stability exclusion region. Without these -

controls, it is theoretically possible to operate the reactor in such a manner as to cause an oscillation outside of the exclusion .

region. A nominal 5% power-flow buffer region outside of the exclusion region is provided to estathh a stabihty margen to the -

analytically defined exclusion region. The buffer region may be entered only when the power distnbution controls are in place, i

Continuous operation with one recirculation loop was analyzed and the adjustments specified in specification 3.5.F.2 were determined by NEDO-24271, June 1980,?Monticello Nuclear Generating Plant Single Loop Operation".

[

D

,,....r. - - -- . , . - , , - ,--..,,,,,..........m. . ,,,.- ._e .-me,.,.,-e-,,,.%,..,.,-. .. ,v,,,.m.w., _ . , - . . . , , , , ,.m..-..,,..s .w-,,%<

3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMEMTS 3.11 REACTOR FUEL ASSEMBLIES 4.11 REACTOR FUEL ASSEMBLJ.E.E Applicability Aeolicability The Limiting Conditions for Operation The Surveillance Requirements apply to associated with the fuel rods apply to those the parameters whicti monitor the fuel parameters which monitor the fuel rod rod operating conditions.

operating conditions.

Obiective Obiective The objective of the surveillance The obiective of the Limiting conditions for Requirements is to specify the type and Operation is to assure the performance of frequency of surveillance to be applied the fuel rods, to the fuel rods.

Specifications Specifications A. Averace Planar Linear Heat Generatino A. Avermoe Planar Linear Heat Ratio (APLHGR) Generation Rate (APLHGR)

During two recirculatiot loop operation, The APLHGR for each type of fuel the APLHGR for each typ3 of fuel as a as a function of average planar function of average planar exposure exposure shall be determined dail shall not exceed the applicable ilmiting during reactor operation at 125% y values specified in the Core Operating rated thermal power.

Limits Report. When hand calculations are required, the APLUGt for each type of fuel as a function of average planar exposure shall not exceed the limiting value for the most limitin lattice (excluding natural uranium provided in the Core Operating Limits eport, Durib n re'ci/cu t Sn doop p%wer' e Q u,c b W reAAA* ^ ' I O "j ' ^ #

opeta or) the PL GR 1 '# '*Q* u -(o r o bc rd ICA 'fw y itEnhondition<4 -

for o ret idn r a shhl not eke d h a v vafuee' c

uel/ '

llggMNK I ' '~' ' od 4 '"'] h:4.1 @rM ' d4 FCCL*1 O L *p mpit plfed/by . 5. 4[fc Il,p 4 . g' . o 4 ,.

time durin # "(4;(,( ;,_1 L2I O. % o r' If itatisanfermined de thak h wer o b Rration, e APL / 6. 3 b c d otM- @(M limitingconditionforokerationis action s all be b , gg g(3 ovc. g/cslu cS p g lY.[Ilc1 hd beincy initiatedexceeded with [n 15 minutes to restore operation to within the prescribed -Me s o- M powee limits. Surveillance and corresponding aN ton e d C-action shall continue until reactor C orecx4non bgo rs f ro g;je d 8

operation is within the prescribed limits. If the APLHGR is not returned i

deIudeM to 4be- C 4 /C- c d-;aO, f L As VY. a r d .

to within the prescribed limits within two thanhours,ithin 25% w the nextreduce thermal fourpower hours. to less 211 3.11/4.11 REV 132 S/20/?' --

'4

7. Core Ooeratina Limits Report
a. Core operating limits shall be established and documented in the Core Operating Limits Report before each reload cycle or any remaining part of a reload cycle for the following:

l l Rod Block Monitor Operability Requirements (Specification 3.2.C.2a) l Rod Block Monitor Upscale Trip Settings i 5 ys e ?ov>cc 4' p.( " Mg Wbi l ik7 I-8-$'O*b b Y g.(i d e 3,5,p) t i

Item 4 a)

Ee,c(Table 3.2.3,, s et m\ <Aw Maximum Average Planar Linear Heat Generation Rate Limits (Specification 3.ll.A)

Linear Heat Generation Ratio Limits (Specification 3.ll.B)

Minimum Critical Power Ratio Limits (Specification 3.11.C)

Power to Flow Map (Bases 2.3.A)

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, mpecifically those described in the following documents:

NEDE-240ll-P-A, " General Electric Standard Application for Reactor Fuel" (the approved version at the time the reload analyses are performed)

NSPNAD-8608-A, " Reload Safety Evaluation Methods for Application to the Monticello Nuclear Generating Plant" (the approved version at the time the reload analyses are performed)

NSPNAD-8609-A, " Qualification of Reactor Physics Methods for Application to Monticello" (the approved version at the time the reload analyses are performed)

ANF-91-048[(P)(A), " Advanced Nuclear Fuels Corporation Methodology for Boiling Water Reactors-EXEM I BWR Evaluation Model," Siemens Power Corporation (the approved version at the time the reload analyses are performed) d The core operatin limits shall be determined such that all applicable limits (e.g , fuel thermal-mechanica limits, core thermal-hydraulic limits, ECCS limit s, nuclear limits such as shutdown margin, transient analysis limits and accident analysis limits) of the safety analysis are met.

d. The Core Operating Limits Report, including any mid-cycle revisions or supplements, shall be supplied
  • upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

6.7 hdDO- 3N60, C 6W  % " $' b "P N .5 4c" M'I A7 W$b uhc m w a mg y, n ,,, a w ,,,. . s n . 2 U - +- lu.- u. to _a m .i p c. cs . pc < G. m c 0 f +d E 4116 D C> 0, Syp - te.guch- d., G V R Omm.' Guy [c.y Tea a %6;\Ay b A IN io 'L L ; cucing

+. .w ke_ McAt4ea o

velm ;s l o g y\.nc.., " % c ,k, i. n e_ cyos vu.. r at + he

( c. ,, 7 .

.c e

, cp

,,a

t __.C , : ea fW."

,[e t.

i r i

.j 1

-1 1

' Exhibit C . i f

Monticello Nuclear Generating Plant - l License Amendment Reauest dated June 22.1995 Revised Technical Specification Pages l Exhibit C consists of the Technical Specification pages with the proposed enanges 'i l incorporated. Existing pages affected by this change are listed below:

L M

15 ,

107 108 109 114 211 249b

0-4 Bases Continued:

For analyses of the thermal consequences of the transients, the Operating MCPR Limit (T.S.3.11.C) is conservatively assumed to exist prior to initiation of the transients.

This choice of using conservative values of controlling parameters and initiating transients at the design power level, produces more pessimistic answers than would result by using expected values of control parameters and analyzing at higher power levels.

Deviations from as-left settings of setpoints are expected due to inherent instrument error, operator setting error, drift of the setpoint, etc. Allowable deviations are assigned to the limiting safety system settings for this reason. The effect of settings being at their allowable deviation extreme is minimal with respect to that of the conservatisms discussed above. Although the operator will set the setpoints within the trip settings specified, the actual values of the various setpoints can vary from the specified trip setting by the allowable deviation.

A violation of this specification is assumed to occur only when a device is knowingly set outside of the limiting trip setting or when a sufficient number of devices have been affected by any means such that the automatic function is incapable of preventing a safety limit from being exceeded while in a reactor mode in which the specified function must be operable. Sections 3.1 and 3.2 list the reactor modes in which the functions listed above are required.

A. Neutron Flux Scram The average power range monitoring (APRM) system, which is calibrated using heat balance data taken during steady state conditions, reads in percent of rated thermal power (1670 MWt). Because fission chambers provide the basic input signals, the APRM system responds directly to average neutron flux. During transients, the instantaneous rate of heat transfer from the fuel (reactor thermal power) is less than the instantaneous neutron flux due to the time constant of the fuel. Therefore, during abnormal operational transients, the thermal power of the fuel will be less than that indicated by the neutron flux at the scram setting. Analyses demonstrate that, with a 120% scram trip setting, none of the - abnormal operational transients analyzed violate the fuel Safety Limit and there is a substantial margin form fuel damage. Also, the flow biased neutron flux scram (specification 2.3.A.1) provides protection to the fuel safety limit in the unlikely event of a thermal-hydraulic instability.

2.3 BASES 15 REV

o e

.t--

3.0 LIMITING CONDITION FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS F. Recirculation System F. Recirculation system

1. Intentional entry into the stability 1. See Specification 4.6.G exclusion region of the power-flow map defined in the Core Operating Limits Report (COLR) is prohibited. If entry into the stability exclusion region does occur, immediately perform one or more of the following until the stability exclusion region has been exited:
a. Insert control rods,
b. Increase the speed of an operating recirculation pump.
2. Entry into the stability buffer region of the power-flow map as defined in the COLR is prohibited unless the power distribution controls as defined in the COIR are in effect. If the power distribution controis-are not in effect and entry into the stability buffer region does occur, immediately perform one or more of the following until the stability region has been exited:
a. Insert control rods,
b. Increase the speed of an operating recirculation pump.

3.5/4.5 107 REV

y.

soz .

.o,

g:

3.0 LIMITING CONDITION FOR OPERATION 4.0 SURVEILIANCE REQUIREMENTS F. Recirculation System

3. The reactor may be started and operated, or l operation may continue with only one recirculation loop in operation provided that:
a. The following changes to setpoints and safety limit settings will be made within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after initiating operation with only one recirculation loop in operation.
1. The Operating Limit MCPR (MCPR) will

-be changed per Specification 3.11.C.

2. The Maximum Average Planar Linear Heat Generation Rate-(MAPLRGR) will.be l changed per Specification 3.ll.A.
3. The APRM Neutron Flux Scram and APRM Rod Block setpoints will be changed as noted in Specification 2.3.A and Table

, 3.2.3.

b. Technical Specifications 3.5.F.1 and i 3.5.F.2 are met.
4. With no reactor coolant system recirculation loops in operation:
a. Comply with Technical Specifications 3.5.F.l'and 3.5.F.2 by inserting control rods and then comply with specifications 3.6.A.2 and 3.5.F.3 for operation with only one recirculation loop in operation,-

OR

~

b. The reactor shall be placed in hot shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

108 3.5/4.5 REV i

_ _ _ _ . _ . _ _ _ _ _._.____..m.m_ . - . . ________m_ m.,- , _ _ _ _ _ _ _ - _ _ . s~ a. _.-_--.-w... . . .. - . . - - , . . , _ . ,c . ,,, + ... - . .- . ... . , . - .._. . .. ~ . , .. . . . , . . ,,., , , . . . ~ . . , .

c.: )

.s- I 1..  ;

O ,

M  !

i t

i 1

d dO M

B s-N in.

O

.a

4

(-

Bases 3.5/4.5 continued:

F. Recirculation System The reactor is designed such that thermal hydraulic oscillations are prevented or can be readily detected and suppressed without exceeding specified fuel design limits. To minimize the likelihood of a thermal-hydraulic instability, a power-flow exclusion region, to be avoided during normal operation, is calculated using the approved methodology as stated in specification 6.7.A.7. Since the exclusion region may change each fuel cycle the limits are contained in the Core Operating Limits Report. Specific directions are provided to avoid operation in this region and to immediately exit upon an entry. Entries into the exclusion region are not part of normal operation. An entry may occur as the result of an abnormal event such as a single recirculation pump trip. In these events, operation in the exclusion region may be needed to prevent equipment damage, but actual time spent inside the exclusion region is minimized. Though operator action can prevent the occurrence and protect the reactor from an instability, the APRM flow biased scram function will suppress oscillations prior to exceeding the fuel safety limit.

Power distribution controls are established to ensure the reactor is operated within the bounds of the stability analysis. With these controls in place, there is confidence that an oscillation will not occur outside of the stability exclusion region. Without these controls, it is theoretically possible to operate the reactor in such a manner as to cause an oscillation outside of the exclusion region. A nominal 5%

power-flow buffer region outside of the exclusion region is provided to establish a stability margin to the analytically defined exclusion region. The buffer region may be entered only when the power distribution controls are in place.

Continuous operation with one recirculation loop was analyzed and the adjustments specified in specification 3.5.F.2 were determined by NEDO-24271, June 1980, "Monticello Nuclear Generating Plant Single Loop operation". Specification 3.6.A.2 governs the restart of a recirculation pump in an idle recirculation loop. Adherence to this specification limits the probability of excessive flux transients and/or thermal stresses.

NEXT PAGE IS 121 3.5/4.5 '114 REV

- 1
D 4, I e ,

3.0 LIMITING CONDITION FOR OPERATION 4.0 SURVEILIANCE REQUIREMENTS  !

3.11 REACTOR RJEL ASSEMBLIES 4.11 REACTOR FUEL ASSEMBLIES I Aeolicability Aeolicability Conditions for eration associated The Surveillance Requirements apply to the TheLimitinhlrodsapplyto with the iu ose parameters which Parameters which monitor the fuel rod Perating conditions.

monitor the fuel rod operating conditions.

Obiective Obiective The obiective of the Limiting Conditions for The objective of the Surveillance ro ra Opgs. tion is to assure the performance of the fuel ggreynt g is oggeQet

{

fy be h l$ to the fuel rods.

Specifications Specifications A. g Planar Linear Heat Generatine Ratio A. Averare Planar Linear Heat Generation Kate (Al'UWR) gig gi ationgo w r operation, The AP111GR for each tvne of fuel as a function of avera function of average p nar exposure shall not bedetermineddaigepTanarexposureshall reactor exceedtheabplicablelimitinkeport.

in the Core perating Limits values dsen ecified hand OPerationat225%yduringtermalpower.

rated t  ;

calculations are required, the AP111GR for each type of fuel as a function of average planar exposure shall not exceed the limiting value for the most limiting lattice natural uranium) provided in th(excludinge Core Operating  ;

1 Limits Report.

During one recirculation loop power operation, the AFLHGR limiting condition for operation for each type of fuel 's hall not exceed the most l limiting of:

a. The above values multiplied by 0.85, or
b. The above values multiplied by the appropriate flow and power dependent correction factors provided in the Core Operating Limits Report.

If at any time during power operation it is determined that the APillGR limiting co,ndition i for operation is being exceeded, action shall i be initiated within 15 minutes to restore t operation to within the prescribed limits.

Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits. If the APLHGR is not returned to within the prescribed limits within two hours 25% within, the reduce nextthermal power to less than four hours.

3.11/4.11 Q

.o V ,

7. Core Operatine Limits Report
a. Core operating limits shall be established and documented in the Core Operating Limits Report before each reload cycle or any remaining part of a reload cycle for the following:

Rod Block Monitor Operability Requirements (Specification 3.2.C.2a)

Rod Block Monitor Upscale Trip Settings (Table 3.2.3, Item 4.a)

Recirculation System Power to Flow Map Stability Regions (Specification 3.5.F) l Maximum Average Planar Linear Heat Generation Rate Limits (Specification 3.ll.A)

Linear Heat Generation Ratio Limits (Specification 3.11.B)

Minimum Critical Power Ratio Limits (Specification 3.11.C)

Power to Flow Map (Bases 2.3.A)

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

NEDE-24011-P-A, " General Electric Standard Application for Reactor Fuel" (the approved version at the time the reload analyses are performed)

NSPNAD-8608-A, " Reload Safety Evaluation Methods for Application to the Monticello Nuclear Generating Plant" (the approved version at the time the reload analyses are performed)

NSPNAD-8609-A, " Qualification of Reactor Physics Methods for Application to Monticello" (the approved version at the time the reload analyses are performed)

ANF-91-048(P)(A), " Advanced Nuclear Fuels Corporation Methodology for Boiling Water Reactors-EXEM BWR l Evaluation Model," Siemens Power Corporation (the approved version at the time the reload analyses are performed)

NEDO-31960, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology," June 1991 (the approved version at the time the reload analyses are performed)

NEDO-31960, Supplement 1, "BWR Owners' Group Long-Tana Stability Solutions Licensing Methodology," March 1992 (the approved version at the time the reload analyses are performed)

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, transient analysis limits and accident analysis limits) of the safety analysis are met.
d. The Core Operating Limits Report, including any mid-cycle revisions or supplements, shall be supplied upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional-Administrator and Resident Inspector.

6.7 249b REV

_ _ , _ _ ._. __ - - . - ,. - . . - -. -- - .. .

Retrieved from "https://kanterella.com/w/index.php?title=ML20085N099&oldid=2738210"