ML20024G619

Proposed Tech Specs,Expanding Certain Specs to Govern Use of 8x8 Fuel Element Design

ML20024G619
Person / Time
Site:	Monticello
Issue date:	02/27/1974
From:	NORTHERN STATES POWER CO.
To:
Shared Package
ML20024G615	List: ML20024G614 ML20024G618 ML20024G619
References
NUDOCS 9102130498
Download: ML20024G619 (8)

Text

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? 2.0 3AFI:TY LIMI'In LDf1 TING SAFETY O TOTE!! SETTif00 2.1 FUEL CLADDING ILTEGRITY 2.3 FUEL CLADDIIF, ITTEGRITY Applicability Applicabli'ty:

Applies to the interrelated curiables Applies to trip settines of tac Instruments associated with fuel thermal behavior. and devices '.hich va p"ovido! te nrevent the reactor ..y.; ten re f :- A ir.i ts fran i being arceak d. l Objective: Ob,j ect ivo :

To establish limits below which the interrity To define the levr.1 of th_ I:ccess variables of the fuel cladding is preserved. at which nutcu2ic pratt eti ce cction is initiatt J tri prev ~1t tro sat s tv limits from i Leing excte'<d.

Specification: Speci fict. tion: i l A. When the reactor pressure is greater The limiting safety systm settings shall be l than 600 psig, the combination of reactor as specified below e

~ coolant core flow and reactor thermal i power transferred to the coolant sh111 A. Heutron Flux Scrs-not exceed the limit shown in Figure 2.1.1. The safety limit is exceeded when the 1. APRM -- The A.i M4 fl*tv r:r_r w trip setting reactor coolant. core flow and thermal sh tll be ur shown in Fi lmva 2.3.1 unle.

power transferred to the coolant results the cot.bination of pcwn:r and peak heat ,

in a point above or to the left of the l flux is above tt.e applicable curve in Figure limit line. 2.3.2. When the. ccabination of power and f.eak na is a e e curve in Fip re 2.3.2, 91021304 DR

ADOC c h PDR 263 2.1/2.3 6

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2.0 SATETY LIMITS LIMITING SAFETY SYSTEM SETTINGS

( S = 493,000 P (7x7 fuel)

~

l X l

f = 425,000 P (8x8 fuel)

X B. When the reactor pressure is less than Where:

600 psig or core flow is less than 5% P = percent of rated power of design, the reactor thermal power X = peak heat flux - (BTU / FIR /FT2) transferred to the coolant shall not shall be used.

exceed 300 MW.

2. IRM--Flux Scram setting shall be 25157 of rated neutron flux.

B. APRM Rod Block - The APPJi rod block setting shall be as shown in Figure 2.3.1 unless the combination of power and peak heat flux is l above the applicable curve in Figure 2.3.2.

C. 1. The neutron flux shall not exceed the When the combination of power and peak flux is scram setting established in Speci fica _ above the curve in Figure 2.3.2, a rod block trip tion 2.3.A for longer than 0.95 seconds setting (RB) as given by: ,

as indicated by the process computer.

RB = 437,400 F (7x7 fuel)

X RB = 382,400 p (8x8 fuel)

X where:

P = percent of rated power X = peak heat flux (BTU /liR/PI 2) shall be used.

2.1/2.3 C. Reactor Low Water Level Scram setting shall be Et 10'6 above the top of the active fuel.

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a Fases Continue.!:

2.1 The design basic critical heat f htx is bssed on an interrelationship of reactor t colant . low Tri t e'es quality. Stem quality is determin~l i.y rmu tor power, pru:nure, ,rei cool'nt n u +. c: '.n 1p:,. wh: in turn, is t function of feedwat.or temperature and tu a lesser Je.,r e react ( t ..x. levs '. E i- i-tion is based upon experir ent 21 data tnken over the pressure rrthr of internat i . s MJh, s.no .l. -

r-relation line was very con.,erv t'.i vely drawn L.elow all the av tilah te dat i. '; ; n( .* t% ( u r< l d ' n I'~

was drawn below the data. therr i r, a vr ry hich prob ibility th st crerati: n 9 t, wt .1 i .t ' s - f ~/ 1 '

would not result in a critical h"at flux occurrence. In addition, if a criti- .1 m' " h. , :er- to <- .e clad perforation would not necennari!y b., expected. Ciadline t<rper,tures v.cu'd irs- r-mitely 11rO'F which is beinw th+ rerroration temperiture af the e 1:.dd : n7 m t< r .1 TL: a b 13 'ri e i -

t- N "-

verified t y tests in the Genorrtl Electrie Test henctor (GEIH) where futi r. 3 : l e r i: C r i o t , I-; , u cperated ibove-the critical hea t flux for sirnificant periori cf tirre (P t i m' .c) ,i.mt cl .6 t- '

.n. (1)

Curves are presented for two different icer ures in Flou r e I' . 2 .1. Tn t. upe . t ar . " is based on a r.:- id operating pressure or 1000 nsir.. Th lower curve is based er . prm su: . of IUf' pr k. l a ne ~ - i reactor pressure ever e.cpected to eF. r .' 1290 psir, and therefore, t' e curn : r:!1 cc ca M1 0; . . .

conditions with interpointion. Ir rmeter precr tre shoc1d (ver e/.ceed IMO ; ' <ia t "v: ' cp 4' ~. a .

it would be assumed that the safety limit has l een viol ttod. For pr su - I ct. *. CO? i s ' y . E'

• i-the lowest pressure used in the critica' heat flux data, tuid 10Lo psig, tir r; - 's cus"e is ,puli w with increased margin.

The power shape assumed in the calculation of these curves was based on design limits and results in a

] total peaking factor of 3.08 for 7x7 fuel and 3.04 for 8x8 fuel. For any peaking of smaller magnitude, the curves are conservative. The actual power distribution in the core is established by specified I cantrol rod sequences and is monitored continuous 1y by the in-core local Power Range Monitor (1.PRit) System.

To maintain applicability of the safety limit curve, the safety limit will be lowered according to the j equation given in Figure 2.1.1 in the rare event of power operation with a total peaking factor in excess of the design value.

(1) T. Sorlie, et. al. " Experiences with Operating IMR Fuel Rods above the Critical Heat Flux" -

Nucleonics, Vol. 23, No. 4, April, 1965.

14 2.1 InSES REV

Bases Continued:

2.1 During transient operation, the heat flux would lag behind the neutron flux due to the inherent heat I transfer time constant of the fu e l . Also, the limiting safety systen scram settings are at values which will not allow the reactor to be operated above the safety limit during normal operation or during other plant operating situations wh ich have been analyzed in detail (4,5,6,7) . In addi. tion, control rod scrams are such that for normal operating transients the neutron flux transient is terminated before a significant increase in surface heat flux occurs. Scram times of each control rod are checked each refueling outage to assure the insertion times are adequate. Exceeding a neutron flux scram setting and a delay in the control rod action to reduce neutron flux to less than the scram setting within 0.95 seconds does not necessarily imply that fuel is damaged; however, for this specification a safety limit violation will be assumed anytime a neutron flux scram setting of the APRM's is exceeded for longer than 0.95 seconds.

Analysis within the nominal uncertainty range of all appropriate significant parameters, show that if the scram occurs such that the neutron flux dwell time above the '11miting safety system setting is less than 0.95 seconds, the safety limit will not be exceeded for normal turbine or generator trips, which are the most severe normal operating transients expected.

The computer provided with Monticello has a sequence annunciation program which will indicate the sequence in which scrams occur such as neutron flux, pressure, etc. This program also indicates when the scram set point is cleared. This will provide information on how long a scram condition exists and thus provide some measure of the energy added during a transient. Thus, computer information norna11y will be available for analyzing scrams; however, if the computer information should not be available for any scram analysis, Specification 2.1.C.2 will be relied on to determine if a safety limit has been violated.

During periods when the reactor is shut down, consideration must also be given to water level requirements due to the effect of decay heat. If reactor water level should drop below the top of the active fuel during this time, the ability to cool the core is reduced. This reduction in core cooling capability could lead to elevated cladding temperatures and clad perforation. The core will be cooled sufficiently to prevent clad melting should the water level be reduced to two-thirds the core height. Establishment of the safety limit at 12 inches above the top of the fuel provides adequate margin. This level will be continuously monitored whenever the recirculation pucps are not operating.

(4) FSAR Volume I, Section 111-2.2.3 (5) FSAE Volume III, Sections XIV-5 (6) Supplement on Transient Analyses submitted by NSP to the AEC February 13, 1973 (7) Letter f rom NSP to the AEC, " Planned Reactor Operation f rom 2,000 MWD /T to end of cycle 2", dated August 21, 1973 16 2.1 BASES REV

I Pases CcI1tinue1:

l 23 A. Heutron Flux Scram - Tt." riverve pownr ranc.e monitorine ( APF14) c:/ stem, c-n'

.sni-h i s calibratoi uning Since ficz:nn heat t tlance d sta taken durinc cteady state ccrrlitionn, reada i n m r :r.

l chambers. provide the insic i nput cirnals, the APR'! sy . tem renpen an d i r c . f:/

tve r ve neut.mn 11*m.

l ' br rd pawor)

During transients, the i nstant aneoun rate of he it t rans fe r f rom tho I'<1 (r- > .

is leca than th" instanttne;u- m ut m flux due to tho t:me cons' int J h fr a . Th"refor . durinc the transients induced by diuturb uicen an] rith an APE:1 ncrs cettir an .h swn r; Fioure P. 3.1 t

i thermal power of the fuci .:111 be lent tbtn that inii ited b:; th< rout?3n 'tr at tim scran re tint.

• ..'ne of Analycis reported in t h" FGra do. m onstr,t es th it, even with a fix-I JPOI .: r ; trip nettinc.,

the pontulated transienta rmult in violation of the fuel c irct:, limit uid ' m r. Mu cubatantial narcin from fuel damare. Y rofore, une of a flow-biiced cera_m provid m aj iltiv:ril urgin.

?. .l. wmll decruise the fin increace in the /JR?1 serr: wtt iv

• to reater than thV ch m. n in Fi m er margin present before the th-r W hydraulic caroty 11 it in rcach-d. Th. arP" c r e cottine was deterr:ined b:/ an analynic of rcirrinn rcquired to f rovile t rancmtL em - * - mancurarinc <iurin -

operation. A reduction in thi n cpcrating marrin wouLI increase the fr.qv not if riurieus cer c.s which have an alverne arrect on reactor naret., Leenuce or unnocennary thcr. .at ntre. whien it causes.

Thus, the fomer IPON APR*4 c-ttieq was celected b(cauce e prov Mc a 3+ q n' - 1r:n from th< thernal-hydraulic cafety limit, yet illows operating mrcin which niniminen ite ;,- _' lity or unnectncarv scramn. Therefore, it is in+onde ! t o ulti-ately rep 1nce (with prior AEC approval) the autoW ie flow referenced scram with a iixed 120 percent scran cetting, providing '!r1 ir~tial p m:r operation <nn-

+

firms the nuclear behavior characterictics ucri in th"ne tranni(nt an-nysis.

Die thermal hydraulic safety limit of Specification 2.1 was based on a total peaking factor of 3.08 A factor has been included on Figure 2.1.1 to adjust the safety I for 7x7 fuel and 3.04 for 8x8 fuel. limit in the event peaking factor exceeds the design value. Likewise, the scraml settin also be adjusted to assure MC'IFR does not become less than 1.0 in this degraded situation. This has been accomplished by use of Figure 2.3.2. If the combination of power and heat flux is greater I than shown by the curve; i.e., a peaking factor greater than the design value exists, the APRM scram setting is adjusted downward by the formula given in the specification. 'Ihe scramflux setting as given condition for by the equation will prevent MCIIFR from becoming less than 1.0 for the given heat the worst expected transients. If the APRM scram setting should require a change due to an abnormal peaking condition, it will be done by increasing the APRM gain and thus reducing the slope ard intercept point of the flow - biased scram curve by the reciprocal of the APRM gain change.

19 REV 2.3 BASES

1

• I' Bases Continurd:

2.3 the worst case MC!!FR during steady state operation is at 110% of rated pover. Peaking factors as specified in Section 3.2 of the FSAR were considered. The total peaking factor was 3.08 for I 7x7 fuel and 3.04 for 8x8 fuel. The actual power distribution in the core is established by specified contr61 rod sequences and is monitored continuously by the in-core LPRM system. As with the APRM scram setting, the APRM rod block setting is adjusted downward if peaking factors

• greater than the design value exist. Ihis assures a rod block will occur before MC11FR becomes 1ess than 1.0 even for this degraded case. The rod block setting is changed by increasing the APRM gain and thus reducing the slope and intercept point of the flow-biased rod block curve by the reciprocal of the APRM gain change.

The operator will set the Arm! rod block trip settings no greater than th,t shown in Ficure 2.3.1.

IIowever, the actual set point can t'e as much an 3'., greater than that chuwn on Firure 2.3.1 for re-circulation driving rinws loss than 50% of design and 2% greater than that shown for recircul tt ic.n driving flows greater than 5% of desirn due to the deviations discusr.m1 on Page 18.

C. Reactor Iow Water Irvel Scran - The reactor low water level scram in set ,t a point which will assure that the water level uned in the t ases f or the safety linit is niintained.

l The operator will set the low water level trip setting no lower than 10%" above the top of the i active fuel. Ilowever, the actual set point can be as much as 6 inches lower due to the deviations discussed on Page IP. s l

D. Reactor Low Low Water Level ECCS Initiation Trip Point - The emergency core cooling subsystems

( are designed to provide sul f icient cooling to the core to dissipate the enercy associated with

the loss of coolant accident anI to limit fuel clad temperature to well below the clad melting temperature to assure that core geometry remains intact and to limit any clad metal-water reaction to less than 1%. The desirn of the ECCS components to meet the above criterion wis dependent on three previously set p trameters

the maximum break size, the low water level scran set point, and the ECCS initiation set point. To lower the set point for initi ition of the ECCS could prevent the ECCS components from meeting their criterion. To raise the Ecrs initintion set point would be in a safe direction, but it would reduce the marc in est tblished to prevent actuation of the ECCS durinc nornal operation or during normlly expected transients.

2.3 BASE 21 ruy

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AEC DISTF LTION FOR PART 50 DOCKET MATER' 1676 (TDiPORARY FORM) CONTROL NO:

FILE:

FROM: DATE OF DOC DATE REC'D LTR MDiO RPT OTHER No72hern Staten Power Connany Minneapolis, Minn. 55401 L.O. Maver 2-27-74 3-4-74 N l

l To: ORIG CC OTHER SENT AEC PDR XXX J,r. O'Learv 3 signed CLASS UNCLASS PROP INFO INPUT NO CYS REC'D DOCKET NO:

XXX XXX 40 50-263 DESCRIPTION: ENCLOSURES:

l Lt tran9 the followinr,...channe to tech specs... Proposed change to tech specs, notarized i 2-27-74....

l DO NOT REMOVE s

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