ML20213E463

Forwards Branch Review of Proposed Tech Specs Re Min Critical Power Ratio,Eccs Response Times & Main Turbine Bypass Sys,Per 830701 Request

ML20213E463
Person / Time
Site:	Columbia
Issue date:	07/11/1983
From:	Sheron B Office of Nuclear Reactor Regulation
To:	Thomas C Office of Nuclear Reactor Regulation
References
CON-WNP-0596, CON-WNP-596 NUDOCS 8307270051
Download: ML20213E463 (9)

Text

led 0mnm0:Ennaca M

RS3 R/F RSB 'P/ F: WNP-2 TCollins R/F d,l *, l Sl3 BSheron WHodges TCollins gy. 39 ;'

HEMORATIDUM FOR:

Cecil Thomas, Chief Standardization & Special Projects Branch Division of Licensing FROM:

Brian W. Sheron, Chief Reactor Systems Branch Division of Systems Integration

SUBJECT:

REVIEW OF WNP-2 TEQiNICAL SPECIFICATI0iG Per your request of July 1,1983 Reactor Systems Branch has reviewed the proposed technical specifications for WNP-2.

The sections covered by our review are: 1.0, 2.0, 3/4.2.2-4, 3/4.3.2-5, 3/4.3.9, 3/4.4.1-2, 3/4.4.9,3/4.5,3/4.7.3,3/4.7.9,3/4.9.11,3/4.10.4, and Bases 2.0, 3.0, 4.0.

Our cor.ments are enclosed.

OrigIH3ItipP3 $

Brian W. Sterca Brian W. Sheron, Chief Reactor Systems Branch Division of Systems Integration cc:

R. Houston R. Mattson R. Auluck D. Hoffman RSB Section B Members CCf4 TACT:

T. Collins X24478 0307270051 830V1'i 77/

p ADOCK 0500039

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CFFICI AL RECCRO CCPY

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1.

Minimum Critica'l Power Ratio (3/?.2.3 and Basesi The MCPR specification and Bases must be ccerected tc reflect use of ODYN option 5 for determination of the operating limi: MCPR.

Attacned is a copy of the LaSalle specifica icn.

This can be used as e model for WNP-2.

2.

ECCS Rescorse Times (Table 3.3.3-3)

The response time given in the specificatiens fcr LPCS and LPCI is 5 46 seconds.

The LOCA analyses assume a response time of 40 seconds.

The specification is therefere non-conservative and shculd be changed to i 40 seconds.

es 3.

Main Turbine Eypass System (Bases 3/4.7.9)

The last sentence of the Bases needs to be corrected to read "The main turbine bypass system provides pressure relief during the feedwater i

centroller failure event so that the safety limit MCPR is not violated."

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?Ok!ER DISTRIBUTICN LIMITS l

3/4.2.3 MINIMUM CRITICAL F0WER RATIC i

LIMITING CONDITION FOR OPERATICN

3. 2. 3 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be equal to or greater than the MC?R limit shown in Figure 3.2.3-1 -ices the K, shewn in Figure 3.2.3-2.

n.-rLI, r.LiiY:

OP:nn 10hnL c.On.-

0.i

, w.u.en inc.o..

D 1 4 i s n.

PCnen is greater : nan er

.~:

e:ual to 25% of RATED THERMAL ?OWER.

ACTICN:

With MCPR less than the MCPR limit determined _from Figure 3.2.3-1 times K, I

determined from Figure 3.2.3-2, initiate corrective action within 1E minutes and restere MCPR to within the required limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> er recuce THERMAL

? WER to less than 25% of RATED THERMAL POWER within the next a hours.

.e n.. Ar e ; Li.n"C :

e Ln....v._e h 5 n s r

n

4. 2. 3 MCPR shall be determined to be equal to cr greater than the a:plicable limit datermined fr:m Figures 3.2.3-1 and 3.2.3-2:

a.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, b.

Within 12 hcurs after completion of a THERMAL POWER increase of at least IE% of RATED THERMAL POWER, and c.

Initially 'and at least once per 12 hcurs when the reacter is operating with a LIMITING CONTROL' ROD PATTERN for MCPR.

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c. a t.,..,..... t., L

..., c : i n u...

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,,.,..,..y 7.a u.v. i C. 7,l. i,,s l t' e n..g, n A..,ilu Ki 7 The required operating limit M^ prs at steady st=te operating ccnditions

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.::w cperating limit, it is required that the resulting MCFR does not decrease teles the Safety Limit MCPR at any time during the transient assu. Fag instre-cent trip setting given in Specificatien 2.2.

To assure that the fuel cladding integrity Safety Limit is not exceeded durinc. an.y anticie ted abnormel operational transient, the mest limiting a

transients have been analvzed to determine which result in the largest reduc-tion in CRITICAL PO.WER RA'TIO (CPR).The type. of transier,ts evaluated were less or i,tew, increase in pressure and pcwer, positive reactivity insertion, any e

coolant terperature decrease.

The limiting transient yibids the largest delta MCPR.

When added to the Safety Limit MCPR of 1.05, the recuired minicu i cperating limit MC?R of Specification 3.2.3 is cbtained and presented in Ficure 2.2.3-1.

The evaluatica of a given transient bec. ins with the s.ystem initial aarameters e

shcen in :5AR Table 15.0-1 that are input to a GE core dynamic behavior trarsient cc.mputer procram.

The ccde used to evaluate pressuri ation events is cescribed o : r.s s. :..,, (.P) a

. s su

-i n n...,, e s s u r :.... : c o.

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d M."N"M 3 CP.IIICAL POWER EATIO (C ntinued)

I described further in F.eference 5, produced generic Statistical Acjustment Factors which have been applied to plant and cycle specific C:YN results to yield cperating limits which provide a 55% prcbability with 95% conficence that the limiting pressurization e,ent will not cause M~PR to fall belcw the

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e.

u.

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..: c.,..,...

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.:.,. ese.:.,

. =

.y 4.

m, s., :,

....e.,

o...

esceed a 5% significance level cc pared to the distributien assu.ed in the

.c.

O' 'N s.atistical analyses, the M:PR limit must be increased linearly, as a f t nction of the mean 20% scram time, to a core conservative value which re'lects an NRC determined uncertainty penalty of 4.4%.

This penalty is 4

.-..g

'.o

'.'.a. - l.= n

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v

;..

. :. a.

e. :i 7,.,i e. c.1,. u,. e..

e.

,...... cm..

. u.,.

' '-i-'.; p: int in the cycle.

It is not applied in full until the = tan of ali c.

i rent cycle 20% scram times reaches the 0.86 seconds value of 5;.ecifica-

.icn 3.1.3.3.

In aractice, however, the recuirements of 3.1.3.3 would. cst e

'.'.keiy be reached, i.e., individual data set average > 0.85 secs, and the

... e uired actions taken well before the running averace exceeds 0.85 secs.

res The 5% significance level is defined in Reference 4 as:

I

(

n n = p - 1.55 (N / 2 N.)1/2 I

1 c

1_,

1 where p

=

cean value for statistical scram time distributien i

to 20% inserted =.688 standard deviation of above distribution =.052 o

=

N

=

number of rods tested at BOC, i.e., all cperable 1

rods t

i n

r I N. =

total number of operable rods tested in the 1

.i=1 current c.vc.ie e.

.: c

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=.=...~..'.s

...4 c.. i s e.

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3 4.

u,

=

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t

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u..

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Ze

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.=..-... -

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. 6:

r'.l n.

4 C..e

v. *.

• y. i... V. (;; ;.as;

.a,-.m roa r, r,.a 70 (u,,.:....a)

.n T'at is, a cycle length was assumed which is ic ger than any past or c: rte plated rafualing interval and the number cf reds tested was t=ximized c -- ' :.. *. ~. c i..,., i :. #. y :..A C

e e TV c-s.1 s.a.1.)..e 4.

c.

-.:s..:

su.

e. i.,.

s. u. a.e - -. s :o e n.

.u

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..,.. c c :. ),

.. n. e. c, e. r.

a.

y c. s.,..

:,... g.2.. 9..:. 4 :e.,

a, a. z. ;. e.

a.

p s. ;., g i...

c.w.

v.

.y i

t.:

a. s.... g,.

e. 1.,. a.

s e.

. s.

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S. n.- v

.. e..

s.

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s.

g

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n.

i...

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..u. 3 a

p y s,,,. s. c,

.w3 p,. p p s

. s u

cn. s..,.

K,,

c. c.. o r,

.h=. K, a.-

s.,.i. u t

.w.

t s.

i i

<=.-.. e c.e..... t. u = s +uc c.2. e +.y L 1..1.

v....:.: w.4.1 1 n,,.

u,. s. tole.s,d.

t e

h Kf f c.,.. r s u-

.o.

u

..o ware d=ri,ed usino THERMAL PCWER and core fic.< corresponding to 105% of

. s. :., <..

...a.

g. f.,... n e

r u=

=*.'

a....... e c a c.. i.: s. e. a e,

u.

s. u.,.... x.:,...

a. <ic...:.,

s.....e.

. o..

i

-c.

.o a.-

.'e correspending THERMAL PCWER aloitg the 105% cf rated steam flow

..s. cl ':. e, s.u.e 11 :s.. i ui ng u,,, s l e s re i c. : s.e.

t

,,,.... r... s

c. a. u z. s c. s t,n.1 1 s h e Me ~ ~t.w.--.

s-v...

s.

,...e

-e s

.r yas siightly abeve the Safety Limit.

Using this relative bundle power, the MCPRs were calculated at different points along the 105% of rated steam flow centrol line corresponding to cifferent core ficas.

The ratio of the M:PR calculated at a given point of core flow, divided by the cpirating limit M"PR, n.s.

,, 1.,.es.he A.

s A*. THERMAL POWER levels less than or ecual to 25% of RATED THERMAL PCWER, the resctor will be operating at minicum recirculation p6p s' peed and the a.

.s, s

.... :. c... s.. : a c n,.=. n. wi l l u.,e s.e.v s.. l 1.

ror c.il a n. e n o... s. a c,n., o s n i red vie o

... c

....: s: u

..v patterns which msy be empicyed at this point, cpErating plant experience

'.ndicates that the resulting MCPR value is in excess of requirements by a c:.s#dirable margin.

During initial start-up testing of the plant, a MCPR

.:n c.

c.e

-e c. c.. :. r..,. ri i

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on

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c.

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P0k'ER D'ISTRI$UTION SYSTEMt 5

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Refere,ces:

1 1.

General Electric Company Analytical Model for Less-of-Ceciant 1

i Analysis in Accordance with 10 CFR 50, Ap.:endix K, NE3E-20E56,

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l 2.

R. E. Linford, Analytical Methods of Plant Transient Evaluations f:r

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n 2.

Qual.ification of the One Dicensional Core Transient Model For 4

E iling v'ater Reactors, NECD-24154, October 1978.

j 4

TASC 01-A Ccaputer Program For the Transiant Analysis of a Single Channel, Technical Description, NECE-25119, January 1950.

5.

"qualificaticn cf the One-Dimensional Core Transient M del f:r Ecilin; k'ater Reactors" General Electric Co. Licensing Tepical 1

1

.s p o r t.,., uu, 2,, : 4.,ols. I ane a II and.i n; t,,_:,- \\,ol. 1.1 as sup-7 v

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plemented by letter dated September 5, 19507 frc: F. H. Euchholz (GE) to P. 5. Check (NRC).

3/4.2.4 LINEAR HEAT GENERATION RATE i'

The specification assures that the LINEAR HEAT GENERATION RATE (LHGR) in any red is less than the design linear heat generation even if fuel peliet censificatien is postulated.

The power spike penalty specinied is based on the analysis preser.ted in Section 3.2.1 of the GE topical report NECM-10735 Supplemen 5, and assumes a linearly increasing variation in axial caos 1

bet-aen core botto: and top and assures with a 95% confidence that no'nore t'.an ene fuel red exceeds the design LINEAR HEAT GENERATION RATE due te p:wer 3 : 1 k i e' 3.

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