ML17345B222

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Proposed Tech Specs for Reactor Vessel Flux Reduction Program
ML17345B222
Person / Time
Site: Turkey Point  NextEra Energy icon.png
Issue date: 08/19/1983
From:
FLORIDA POWER & LIGHT CO.
To:
Shared Package
ML17345B221 List:
References
NUDOCS 8308230353
Download: ML17345B222 (13)
v • d • e


Text

ATTACHE".ENT A P."-G."OSED TEC'r'fbi COL SPEC IFiCAT s~h AI'i"-tl0i'=t/TS 8308230353 8308i9 PDR ADQCK 05000250 P

PDR

LIST OF FIGURES

~Fi ure 2.1-1 2.1-la

2. 1-lb 2.1-2 3.l-la 3.1-1b 3.1-1c 3.1-1d 3.1-2 3.1-2c 3.1-2d 3.2-1 3.2-1a 3.2-1b 3.2-lc 3.2-2 3\\2 3 3.2-3a 3.2-'0 0.12-1 6.2-1 6.2-2 Title Reactor Core Thermal and Hydraulic Safety Limits, Three Loop Operation Deleted Deleted Reactor Core Thermal and Hydraulic Safety Limits, Two Loop Operation Reactor Coolant System Heatup and Cooldown Pressure Limits Reactor Coolant System Heatup and Cooldown Pressure Limits Reactor Coolant System Heatup and Cooldown Pressure Limits Reactor Coolant System Heatup and Cooldown Pressure Limits

- Radiation Induced Increase in Transition Temperature for A302-B Steel Radiation Induced Increase in Transition Temperature for A302-B Steel Radiation Induced Increase in Transition Temperature for A302-B Steel Control Group Insertiori Limits for Unit 0, Three Loop Operation Control Group Insertion Limits for Unit 0, Two Loop Operation Control Group Insertion Limits for Unit 3, Three Loop Operation Control Group Insertion Limits for Unit 3, Two Loop Operation Required Shutdown Margin K(Z) vs. Core Height Deleted Maximum Allowable Local KW/FT Sampling Locations Offsite Organization Chart Plant Organization Chart B3.1-1 B3.1-2 B3.2-1 B3.2-2 Effect of Fluence and Copper Content on Shift of RTNDT for Reactor Vessel Steels Exposed to 550 F Temperature Fast Neutron Fluence (E>1 MEV) as a function of Effective Full Power Years Target Band on Indicated Flux Difference as a Function of Operating Power Level Permissible Operating Band on Indicated Flux Difference as a Function of Burnup (Typical)

655 Z400 Ps )'g 648 6'=5

~ 625 0

~ 628 5

~000 Ps)'~

6P5-Psq.z

<<c5

<<CQ

<<65 5BB c75

.I '2

,5

.4

.5

.6

.7

.8

.R I.

).I l.~

PQVER (fraction of noninel)

Figure 2.l-l Reactor Core Thermal and Hyoraulic Safety Limits, Three Loop Operation

Thi s Figure intentional lg~eleted.

Figure 2.1-la

This Fi gure intentional i~el eted.

Figure 2. 1-lb

R:-AC OR COOL~HT T:-~PZ~~TURK C"ertc=oerature

$ T < 3To (K1 Q. Ql07 (T-574 )

~~ Q. 000453 (p-2235}

f (i q) )

c? t&0'c rat ec po~'er~

hve a~e te=er ture, Pres u i"er p es u e, psIS a

unction of the ina'catec'if e

e ce bet"een top zn6 bot oa detectors of the po" r-ra 8e uclear Ron cheers; arith ~aiw to be se ec" e zse6 on measure i sarment response Burin~ start~

tests such that:

or qt

.b)

"~ thin. + 10 percent and -14 pe cent ~~ere qt an8 qb are the pezce t po~er in the too anc. bottom halves of the core respective y, an6 qt +

ab is total core po.-er in perce"" of rated pot-er, f (Lc) ~ 0-7o each pe cent that the aapnitv e of (qt qb) ence c's.'0 percent the Delta-T trip setpoi"- shall be automatically "educed by 3.5 perce" t of its value at inter~a po -er.

-or each pe:cent that the by 2 parcae=

of its valve Klaxon'u e.ot se tpo i""=-sha at 'n e

J (g

qb)

QKceecs -i 4 3

be auto-.aticall"=ecucM

(:nree Loop Cperzt'on) 1.095

(:vo loco Cperation) 0.88 2

~ 3 2

0 Ove.o~er5T-T1 09

-Y1 dT g2 (T

J

)

f ($ Q) 0 lndicated T at atec po"er, c

T eve zge e pcIu e

P lnd ca.tH.,averzge te~erat re at no-.,,inal coed-ons and rated poser, Kl 0 fo-decrezsing average temperature; 0.2 sec-/7 for incraasi"g average to=7erzture 0.00068 or T equal to or -ore thzn T; 0 for T less than T Pate o

cna"ge of tmerzt re, c/sec dT'

~

f (ne)~As do:i e" abave.

Pressuri"er Lo~ Press -irer pressure eeuz~

to or greater tha" 1835 psg-Hign

~

~ ~ ~

I than 2385 ps-g-to or less than 9?~

o= full Pressur'er J

c

~ ~ 1eyv

~

w pv 1y, ~

1 s

la@

atua pressure eo ='-co or ess sc zle-Pe ac to r Coolant Plo-Lo= reactor coolant flow e uz'o or greater than 90Z o

co=a indicate" flow-Low reacto cool'znt p~ notor f ecuency equal to or greats thzn 56.1-Ez-

~ Undervol age on reactor coolant pu~ notor bus equa'o o"

i grezter thzn 60X of nor zl voltzge-S t can Can erato rs V

Low-los stean generator pater level equal to or greater than 15Z of nar o~ range instru=eat scale.

2

~ 3 3

6.

DlB PARfNETERS ~.

The ollowing DtiB related parameter limits shall be maintained during power operation:

a.

Reactor Coolant Systan Tavg ( 578.2'F b.

Press uri zer Press ure> 2220 psi a*

c.

Reactor Coolant Flow

> 268,500 gpm With any of the above pa ram ters exceedi ng its 1 imiit, restore the parameter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce thermal power to less than 5~ of rated thermal power using normal shutdown procedures.

Compliance with a.

and b; is demonstrated by verifying that each of the parameters is within its limits at least once each 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Compliance with c. is demonstrated by veri;y'ng within its limits after each refueling cycle.

"ilat ihe parcae er l 5 "Limit not applicable during either a

TH":." iAL PO'.".R ramp increase in excess oi (5")

RATED THERN~~

POWER per minute or a

THERt~iAL POWER step increase in excess of (10")

RATED THERMAL POWER.

3. 1-7

reactivity in rtion upon ejection greater th 0.3" tk/k at rated power.

Inoperable rod worth shall be determined within 4 weeks.

b.

A control rod shall be considered iroperable -if (a) the rod cannot be moved by CRO:", or (b) the rod is misaligned fran its bank by more than 15 incnes, or (c) the rod drop time is not met.

c.

Is a control rod cannot be moved by the drive mechanism, shutdown margin shall be increased by boron addizion to compensate for the withdrawn worth of the inoperable rod.

CGHTROL ROD POSITIOH IHO ICATION I

either the power range channel deviation alarm or the rod deviation monitor alarm is not operable, rod positions snail oe logged once per shift and after a load change areater than 10> of rated power.

If botn ala rms are inoperable for, two hours or more, the nuclear overpower zri p shall be reset to 935 of rated power.

POlhiiR 0 ISTR IBUTIOH L INITS a.

Hot channel factors:

(1)

F~ Limit The hot channel factors (defined ~ Bases) must olloving limits at all t'mes exmpz during lo>>

tests:

Fq (Z)

< (EFq]L/P) x K(Z),

or P >.0.5 Fq (2)

< (2 x [Fq]L) x K(2),

or P

< 0.5 F~~< 1.62 I.1.0

+ 0.3 (1 - P)]

1'here P is the fraction of razed power at which operating; K(2) is the function given in Figure the core height location oi F~.

meet zhe

~s povr r phys ic s zhe core is 3.2-3; l is Plugging 1 evel LFq]L

< 5g

2. 32 Figure Humber for K(Z) 3.2-3 (2)

Augmented Surveil lance (PIIOS)

If [Fn]p, as predicted by approved physics cal col arions; exceeds L'F ]t then the pow r wiii be limited to a turnon pow r practica T, equal to the ratio of [Fq]t divided by LFq]p, or, for operation at pow r levels above pT aupmen<<d surveillance of hot channel factors shall oe implemented, except in Base 'Load 3.2-3

I I I I I I

I II IIII I

II I

I I l I

I I

II I

I I I I I I /II I I I

I

> I III I

I I

Thi s Figure -intent.ional lg-.deleted.

Figure 3.2-3a

The curves of Figure 2.1-1 POl~'ER, Reactor Coolant System press calculated DllBR is no less than the

-at the vessel exi t i s 1 ess than the show the loci of points of THE%i~i ure and average temperature for which tne design DhBR value or the average enthalpy enthalpy of saturated liquid.

The curves are conservative for an enthalpy hot channel

factor, F$ H, of..62 and a reference cosine with a peak of 1.55 or axial power shape.

An allowance is included for an increase in F",H at reduced power based on the expression':

FgH

< 1.62 I.l + 0-3 (1-P) j where P is the fraction of RATED THERHAL PO>>ER.

These limiting heat flux conditions are higher than those calculated for tne range of all control rods fully withdrawn to.he maximum allowable control rod insertion limit assuming the axial power imbalance is within the limits of tne f(Lq) function of the Overzemperature aT tri p.

linen the axial pow r imbalance is not '"i thin the tolerance, the axi al gower imoal ance effect 0Tl the Over

=.: erature

< 7 trips will reduce the setpoin.=~

to provide pro ection; consistent with core safety limits.

Fuel rod bowing reduces the values of DiiB ratio (DISR).

The amount of the DllBR reduction is 4.7i.'or LOPAR fuel with the L-grid DtiB correlation and B.b'or the OFA fuel with the i'RB-1

.correlation.

The penal.ies are calculated pursuant to "Fuel Rod Bqw Evaluation,"

HCAP-8691-P-A, Rev.

1 (proprietary) and QCAP-8692 Rev.

1 (non-proprietary).

The restrictions of the Core Thermal Hydra'ulic Safety Limits assure that an amount of ONER margin greater tnan or equal to the above penalties is retained to offset the rod bow DAR penalty.

B2. 1-2

An up-.er bound envelope as def;ned by nor..al i:ed peaking factor axial de-ercence of Fic re 3.2 3,

has been Cetermired to be consistent

~:.". tne tec;n'.cal s-e 'cat'.'ons on power distri ution control as given in e'er 1on 3

T :

esu>ts oi ha loss of coolant accidert analyses based on this uper bcund envelope indicate a

peak clad temperature could theoretically exceed tne 2200'F limits.

To ensure the cr>teria are not violated, DIOS will be used to provide.a i-ore exact indication of Fo.

Note that NAOS and a penalty on

~

are only required above P

i.o meet the aCceptance criter a

as iuStified in th ana!yses.

Below PT, the nuclear analyses of credible power shapes consistent witn these speci =icationS hav Si,awn ti",at the

) imiit of L"~~L/P times;igure 3.2-3 is not exceeded provided the limits of Figure 3.2-3 are applied.

h'hen an FO measurement is taken, both exper',mental error and manufac.uring t0 1 ef ance must be al 1 owed or.

Fi ve percent i s the appropri ate al 1 owanc

'or a 'ul 1

cor~

map taken with tne movable incore detector i lux mapping syst m and three percent is the appropr',ate allowance for manufacturing toleranc These uncertainties only apply if tr e map. is taken for purposes other than the determination of P~L and PR~.

<n tne speci iud limit oi F,.., there is an 8 percent allowance for uricertainti>s which means thct normal oper".ioo of the core is expect " to esui

~

in F",><1. 6?Jl.OB.

The log'.c be!;;nd the~iarger uncertaiinty iin =iis cas~

hat (a

Aorm 1 "er-.urbations in -he rad', a!-~power si ape (e. o., roc-misai icn.-. n:)

a-,

e - i,, in most cas=s witrot~ necessar'.

at,'e ting -;<,

(b) 1 t ough the operator has a

Ci rect i ni1 uenc~- on,=< throuon r ovum iL oi r"cs, and can 1 imit it to the desired value, lg has no direct control over F<<

arc(c) an error in the prediction ior radial p6'~el

shape, which may be detected durinr, startup physics tests can be compenSated

-'or in F~ by tignt=r ax'.al contro1, bt c"mpensation

<for F,H is less readily available.

Khan a

.e=sur=ment of F'.. is taken, experiment=1 ef rot.-ust be al wed fo.

and

<" is a

ropria'e a~llcwance

,or a iull core map taken with the m"vab

~ l incore de'actor flux mapping systefli ~

l!easurements of the hot channel factors are requ;red as part of start-up physics tests, at least once each full power month of operation, ard when"v r abnormal'o'wer distribution conditions require a reduction of core power to a

level based on measured hot cl annel factors.

The incore map taken following

~ initial loadino provioes confirmation of the basic nuclear B3.2-4

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