Text

Proposed Tech Specs Adding Unit & Cycle Specific Footnotes Relative to Positive Moderator Temp Coefficient & Reduced Thermal Design Flow
	ML20071K973
Person / Time
Site:	Byron, Braidwood
Issue date:	07/26/1994
From:	; COMMONWEALTH EDISON CO.
To:	;
Shared Package
ML20071K967	List: ML20071K966; ML20071K973;
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	NUDOCS 9408020129
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ATTACHMENT 2 MARKED UP PAGES FOR PROPOSED CHANGES TO APPENDIX A TECHNICAL SPECIFICATIONS OF FACILITY OPERATING LICENSES NPF-37, NPF-66, NPF-72, AND NPF-77 BYRON STATION UNITS 1 & 2 BRAIDWOOD STATION UNITS 1 & 2 REVISED PAGES: REVISED PAGES:

lil ill 2-1 2-1 2-2 2-2 2-2a 2-2a B 2-1 B 2-1 2-5 2-5 3/42-8 3/4 2-8 83/42-4 B 3/4 2-4 3/4 9-1 B 3/4 9-1 l

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SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS z;,_ ~

SECTION _. PAGE 2.1 SAFETY LIMITS . , _ _ _ .

2.1.1 REACTOR C0RE................................................ 2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE............................. 2-1 FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION.. 2-2 1%u eE Z. 8- la PGAc,c2 &CG SA% LI M aT - %c tc6;b s a oTWAiid 2 -Z 2 2.2 LIMITING SAFETY SYSTEM SETTINGS ,

2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0INTS............... 2-3 TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS.... 2-4 BASES SECTION , ._

PAGE 2.1 SAFETY LIMITS _ . . _ , _

2.1.1 REACTOR C0RE................................................ B 2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE............................. B 2-2 2.2 LIMITING SAFETY SYSTEM SETTIN35 ,, , _ , _ _ _ _ _ _

2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0!NTS............... B 2-3 BYRON - UNITS 1 & 2 III

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2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating i op coolant temperature (T,yg) shall not exceed the limits shown in Figure 2.1-1 for fo cor ope ati n 4

APPLICABILITY: MODE ACTION:

Whenever the point defined by the ccmbination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pres-surizer pressure line, be in HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and comply with the requirements of Specification 6.7.1.

REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig.

APPLICABILITJ: MODES 1, 2, 3, 4, and 5.

ACTION:

H0 DES 1 and 2:

Whenever the Reactor Coolant System prusure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within this limit within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and comply with the requirements of Specification 6.7.1.

MODES 3, 4 and,5:

Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within this limit witdin 5 minutes, 'and comply with the requirements of Specification 6.7.1.

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BYRON - UNITS 1 & 2 2-1

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FIGURE 2.1-1 a i

REACTOR CCRE SAFETY LIMIT - FOUR LOOPS IN OPERATION M Epplicable 40 041i. Aepl?.ble k vuM 2 u4I ocxiplE BYRON - UNITS 1 & 2 ct e9eh b2a

2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could I result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB. This relation has been developed to predict the DNB flux and T the location of DNB for axially uniform and nonuniform heat flux distributions.

The local DNB heat flux ratio (DNBR) is defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, and is indicative of the margin to DNB.

'Dl3W g j -The DMS de:ign bc:i: i: :: f 11:w:: there mu:t be at le::t 05 per nt

$r:babi'ity that the miefmum DNBR cf the limiting red during Cenditien I ond II events i; greatee-tha&oe-equal--te-the-DNBft-14st-+f-th: ONS correlation being used (the-W"B-1 :Orretation for Optimized fuel Assembly (OFA) fuel :nd the 2 ,

h00-2 correktien-for "ANTAO: 5 fuel in th r*s applicatica). The :str:Mtir t >

0"SR 'ie4t i estabMshed beseo-on-the-entire epplicable exp:riment:1 d:te-eet cue that there-is-a-95-pereent-probabiMty-w+th 05 percent cef4 dent: t hn-GHB e44-not-occur when-the-minimum-DNBR-is- at-the-correktion-DNE4 limit (1.17 f r both the URB-1 and upg_3 cc77e7;t$3n ), };

In :: ting thi design basis, uncertaintic; in plant oper:t-ing-par:r ters, euclear and thermal-par:reten, and-fuel fabricatier per ete"E e"e read de"ad-statisticalTy-+uck-t-hat there is at least s 05 confidence that th minimum 0""".

r the 'imi-ting-rods-is-greater-than-or equal to th ONSR linit. The uncer-

-tcintic+-in-the-above-plent--parameter; are used to det mine the pl:nt DMBR encertainty. Th4: DNBR uncertainty, combined with th: correl: tier DMSP "-it, s tablishes a design DNBR value which-must~be-met-in tientw afety an:ly:is teng-vakes-of input ~p s-w h om, uom usiipe .

i.zs G 4u +pht ud 4 h.-w e<lls The design DNBR valueg are%L.34 and 1. u for a typical cell and a thimble cell, thimble respectively for 0FA fuel, cell for the VANTAGE 5 fusl qp)d 1.33 formargin In addition, a typical pas beencell and 1.32 for a maintained inbothdesignsbymeetingsafetyanalysisDNBRlimitsogl.49foratypical cell and 1.47 for a thimble cell for 0FA fuel, and 1.67 a n .65 for a typical cell and a thimble cell, respecti in perf_o safety analyses, Q AGE 5 f p,4,, z. t - t a )

u

,$ f,.N g;c, l y The curves of Figure 2.1-1"show the loci of points o R

'l Reactor Coolant System pressure and average temperature for which the minimum design DNBR is no less than the design DNBR value, or the average enthalpy at the vessel exit is less than the enthalpy of saturated liquid.

LL4 Appbeable le ual 1. Appl; cal, 4e U44 2 umI cempida- of c'Jela I BYRON - UNITS 1 & 2 B 2-1 AMENDMENT N0.536-4 o p H.Jis.J G I h M . b G

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INSERT A The DNBR thermal design criterion is that the probability that DNB will not occur on f the most limiting rod is at least 95% (at a 95% confidence level) for any Condition I or 11 event.

In meeting this design basis, uncertainties in plant operating parameters, nuclear and thermal parameters, and fuel fabrication parameters are considered. As described in the UFSAR, the effects of these uncertainties have been statistically combined with the correlation uncertainty. Design limit DNBR values have been determined that satisfy the DNB design criterion.

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_ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ - _ _ _ _ _ ____A

TABLE 2.2-1 (Continued)

$ REACTOR TRIP SYSTEM INSTRUNENTATION TRIP SETPOINTS E

[ FUNCTIONAL UNIT TRIP SETPOINTS ALLOWABLE VALUE

= -

3 12. Reactor Coolant Flow-Low 290% of loop mini- 189.3% of loop alpi-

_ mum measured flow mum measured flow

13. Steam Generator Water

[

Level Low-Low

a. Unit 1 233.0% of narrow 231.0% of narrow range instrument range instrument '

span span

b. Unit 2 236.3% of narrow 234.8% of narrow '

range instrument range instrument ' -

span span ,

m 14. Undervoltage - Reactor 25268 volts - 24920 volts - -

E, Coolant Pumps each bus each bus 3

15. Underfrequency - Reactor 257.0 Hz 256.08 Hz 3 Coolant Pumps .

')

16. Turbine Trip <

3 .

a. Emergency Trip Header 21000 psig 1815 psig -

Pressure ._)

b. Turbine Throttle Valve 21% open 21% open ~

Closure __3 k= 17. Safety Inje-tion Input N.A. N.A. '7 from ESF -

3 -

$ 18. Reactor Coolant Pump M.A. N.A.

, Breaker Position Trip >

Minimummeasuredflowf97,600gpm) 92,85ccypm 4% Llol Ql'ieable 4e u.ar i , gl,cabla +odnit 2 d !OWl*2N oE c9"I# 8-n n- ... u.. o. ,...w n e ,.._.,,,, ,. ,,,,;y z ,c w c ,,, g_

POWER DISTxIBUTION LIMITS 3/4.2.3 RCS FLOW' RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR LIMITING CONDITION FOR OPERATION i I

3.2.3 Indicated Reactor Coolant System (RCS) total flow rate and F q shall be mainty ne g g gljo g f,og {ou lyo o eration.

-a.EllCS Total Flowrate > 390,400 gpm, and

b. F H $ 1.55 [1.0 + 0.3 (1.0-P)] for OFA fuel ((

F H i 1.65 [1.0 + 0.3 (1.0-P)] for VANTAGE 5 fuel 5!

where:

Measured values of F H are obtained by using the movable incore detectors. An appropriate uncertainty of 4% (nominal) or greater shall then be applied to the measured value of F before it is H

compared to the requirements, and THERMAL POWER P _ RATED THERMAL POWER APPLICABILITY: MODE 1.

ACTION:

With RCS total flow rate or F g outside the region of acceptable operation:

a. Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either:

1. Restore RCS total flow rate and AF"H to within the above limits, or

2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux-High Trip Setpoint to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

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BYRON - UNITS 1 & 2 3/4 2-8 AMENDMENT NO. 36

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l POWER DISTRIBUTION LIMITS BASES i

HEAT FLUX HOT CHANNEL FACTOR and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continuea)

c. The control rod insertion limits of Specification 3.1.3.6 are maintained, and

d. The axial power distribution, expressed in terms of AXI DIFFERENCE, is maintained within the limits. [571, doo3F=,

F g will be maintained within its limits provided the Conditions a. thro V

d. above are maintained.N The combination of_t e RCS flow requirement (390,400 g and the requirement 0on F " guarantee thatdhe (DNBPttsetrTfrthe rafetyanaiy will be met.

)]:

( [I.So fev ne dypice /aad-/bble cells, Margin between the safety analysis limit DNBRs (1.49 and 1.47 for the OFA fuel typical and thimble cells, r tively and 1.67 and 1.65 for the VANTAGE 5 typical and thimble cel d the design limit DNBRs4(1.34 and 1.32 for the OFA fuel typical and thimb ells, and 1.33 1.32 for the VANTAGE 5 fuel typical and thimble cells, re ct(v y)]'s mainL -

1 A fraction of this margin is utilized Q.zh 4G p;wlt &%L>le ommoda s<lls anett4ca-c6re%

DNBR penalty (maximum of 12.5%) and the appropria rod bow DNBR penalty 5 (less than 1.5% per WCAP-8691, Revision 1). The rest of the margin between <E design and safety analysis DNBR li ' C can be W s i tr - 2 l g6%'pa flexibility. P,ds,d Thc%lDcoig end e The RCSaflow requirement is n msaurs ow rate ,

of97,600gpm)whichir i dinthfmprovedThermalDesignProceduredxcribed g in i n R 4.4.1 and 15.'.

and is used to calibra A precision heat balonce is perforeco once each cycle he RCS flow rate indicators. Potential fouling of the feedwater venturi, which r,aght not be detected, could bias the results from the precision heat balance in a non-conservative manner. Therefore, a penalty of g gf _ 0.1% is accacced f g potential feedwater venturi fouling. A maximum measurement uncertaintyoT2.2%)hasbeenincludedintheloopminimummeasuredflowrateto l account for potential undetected feedwater venturi fouling and the use of the RCS flow indicators for flow rate verification. Any fouling which might bias ;

the RCS flow rate measurement greater than 0.1% can be detected by monitoring [

and trending various plant performance parameters. If detected, action shall be :

taken, before performing subsequent precision heat balance measurements, i.e.,

either the effect of fouling shall be quantified and compensated for in the RCS flow rate measurement, or the venturi shall be cleaned to eliminate the fouling.

Surveillance Requirement 4.2.3.4 provides adequate monitoring to detect i possible flow reductions due to any rapid core crud buildup.

Surveillance Requirement 4.2.3.5 specifies that the measurement instrumen- ;

tation shall be calibrated within seven days prior to the performance of the '

calorimetric flow measurement. This requirement is due to the fact that the drift effects of this instrumer,tation are not included in the flow measurement uncertainty analysis. This requirement does not apply for the instrumentation !

whose drift effects have been included in the uncertainty analysis. ,

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BYRON - UNITS 1 & 2 B 3/4 2-4 AMENDMENT NO. 36

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3/4.9 REFUELING OPERATIONS _ ,, _ . _ , ,

3/4.9.1 BORON CONCENTRATION _ _ _ . . _ _ . . . . .

,,__ _.. LIMITING CONDITION FOR OPERATION _

3.9.1 The boron concentration of all filled portions of the Reactor Coolant System and the refueling canal shall be maintained uniform and sufficient to ensure that the more restrictive of the following reactivity conditions is met:

a. A K,ff of 0.95 or less, or b.8Aboronconcentrationofgreaterthanorequalto2000 ppm.

4* et i tdre ce+vAG- of <)re4u 4% or ut"al 49 45e pe*.

I APPLICABILITY: MODE 6*. -

i ACTION: -

1 With the requirements of the above specification not satisfied, immediately i

suspend all operations involving CORE ALTERATIONS or positive reactivity I changes and initiate and continue boration at greater than or equal to 30 gpm )

of a solution containing greater than or equal to 7000 ppm boron or it:, equiv-alent until X,ff isreducedtolessthanorequalto0.95orgeboron concer)tration is restored to greater than or equal to 2000 ppg whichever is l the more restrictive. T q CZ5e %)

. SURVEILLANCE REOUIREMENTS . . - - .

4. 9.1.1 The more restrictive of the above two reactivity conditions shall be determined prior to:

a. Removing or unbolting the reactor vessel head, and

b. Withdrawal of any full-length control rod in excess of 57 steps (approximately 3 feet) from its fully inserted position within the reactor vessel.

4.9.1.2 The boron concentration of the Reactor Coolant System and the refueling canal shall be determined by chemical analysis at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

M e :>

4.9.1.3 Valves CV111B, CV8428, CV8441, CV8435, and CV8439 shall be verifiec closed and secured in position by mechanical stops or by removal of air or $

g [

electrical power at least once per 31 days. ;

e. . _

The reactor snall De maintained in MODE 6 whenever fuel is in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.

& lof Applicable lo una i.. APPlleabl,r lo um!4 I_ M I &c ocwpM. o{ ofb L

^1*

u.4pplie.bic 4e u nl+ 1. Llo4 W 'd' ' # A "# ' -

BYRON - UNITS 1 & 2 3/4 9-1

O 3/4.9 REFUEL 1NG OPERATIONS BASES 3/4.9.1 BORON CONCENTRATION The limitations on reactivity conditions during REFUELING ensure that:

(1) the reactor will remain suberitical during CORE ALTERATIONS, and (2) a uniform boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel. The limitation on Keff of no greater than 0.95 is sufficient to prevent reactor criticality during A 25cp refuelin'g uncertainties. operations and includes a 1% Ak/k conservative al}(nwancedorSimi includes h conservative uncertainty allowance of 50 ppm. These limitations are consistent with the initial conditions assumed for the boron dilution incident in the safety analyses. The locking closed of thi required valves during refueling operations precludes the possibility of uncontrolled boron dilution of the filled portions of the RCS. This action prevents flow to the RCS of unborated water by closing flow paths from scurces of unborated water.

3/4.9.2 INSTRUMENTATION The OPERABILITY of the Source Range Neutron Flux Monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core.

3/4.9.3 DECAY TIME ,

The minimum requirement for reactor suberiticality prior to movement of irradiated fuel assemblies in the reactor vessel ensures that sufficient time has elapsed to allow the radioactive decay of the short-lived fission products.

This decay time is consistent with the assumptions used in the Jafety analyses.

3/4.9.4 CONTAINMENT BUILDING PENETRATIONS The requirements on containment building penetracion closure and OPERABILITY ensure that a release of radioactive material within containment wi.11 be restricted from leakage to the environment. The OPERABILITY and closure restrictions are sufficient to restrict radioactive material release from a fuel element rupture based upon the lack of containment pressurization potential while in the REFUELING MODE.

The Byron Station is designed such that the containment opens into the fuel building through the personnel hatch or equipment hatc'h. In the event of a fuel drop accident in the containment, any g.tseous radioactivity escaping from the containment building will be filtered through the Fuel Handling Building Exhaust Ventilation System.

3/4.9.5 COMMUNICATIONS The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity conditions during CORE ALTERATIONS. ,

$ $ NY R { .h A h b BYRON - UNITS 1 & 2 B 3/4 9-1

O 6

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BRAIDWOOD AFFECTED PAGES l

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SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS PAGE SECTION 2.1 SAFETY LIMITS 2.1.1 REACTOR C0RE................................................ 2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE............................. 2-1 l FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPE:'.ATION.. 2-2 FICdRE 2..t-w VEACTA CsE. ShfA utMT,ptygugh na C. pig 2pty 7-A 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0!NTS............... 2-3 TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS.... 2-4 BASES P PAGE SECTION 2.1 SAFETY LIMITS 2.1.1 REACTOR C0RE................................................

8 2-1 <

2.1.2 REACTOR COOLANT SYSTEM PRESSURE............................. 8 2-2 l

2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0INTS...............

8 2-3 i

BRAIDWOOD - UNITS 1 & 2 III 1

MAadbMEWT Me.

1

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2.0' SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS f

2.1 SAFETY LIMITS !

REACTOR CORE 1

2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest l operating loop coolant temperature (T,yg) shall nnt exceed the limits shown in :

Figure 2.1-(for four ( loop . operation.r2. \ - \ c[g) !

MODES 1 M( 2. F p r-t, APPLICABILITY: _, ._ __> l ACTION:

Whenever the point defined by the combination of the highest operating loop i average temperature and THERMAL POWER has exceeded the appropriate pres- c surizer pressure line, be in HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ,-and comply with the requirements of Specification 6.7.1.

REACTOR COOLANT SYSTEM PRESSURE I 2.1.2 The Reactor Coolant Sve. tem nraccora ch=11 ant .- .ad 97?c gg f APPLICABILITY: MODES 1, 2, 3, 4, and 5.

ACTION: y MODES 1 and 2: I Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be {

in HOT STANDBY with the Reactor Coolant System pressure within this limit ;

within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and comply with the requirements of Specification 6.7.1. ,

MODES 3, 4 and 5: I Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within this limit within j 5 minutes, and comply with the requirements of Specification 6.7.1. j t

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l

'O 40 60 80 100 120 POWER (PERCENT)

FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION f\ q' \ s c.cdA \e 'Um4 L OW1 bd ~~l ov&t( togkMM l o cv o G, I BRAIDWOOD - UNITS 1 & 2 2-2 l b w e ,r W,

680 . .. . . . . . . . . . . . . . l l

. .. . . . . .. . . . . . . . . . i 670 '

2411. psL nc . . . .- . ..

660 . .

'2251Y palas*r**

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g ... ... . . ...

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% . .. . . . .. . . . . . . . . I g . .. . . . .. . . . . . . .

s. 640

: g: N.. '...

2 2.6.0.6.

m.

6 ..

's. l.a.. :... :. . :. .

cp .. . .. . .. . . . . . . . .

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E 630 .

. .. .w .

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CD . . .. \. .. . . . . .

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620 .

. . x.

. . . J. . L . .....

. . . ... . . J. . .t .. '.

A.

610

. .s... . .....

600 . . . : . . .. . . . , . . . .

590 580 O 0.2 0.4 0.6 0.8 1 1.2 Power (Fraction of Nominal)

I Figure 2.1-10,.

]'

Reactor Core Safety Limit - Four Loops in Operation fNwbc.c6.\E To At i_ cmA Ow A D .3>hrh n(_N dM\ t 6 6, 1-n C,nn e.uOb -Owcr% 1f1 6. un . Mi M _ 7 NO . l l

2.1 $AFE7Y L2MITS ,

BASE 5 _

~

2.1.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation tamperature.

Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly asasurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been This relation has been developed to predict the DN8 flux e ,

related to DNB.

and the location of DNB for axially uniform and nonunifore heat flux distri- 6 -

butions. The local DNB heat flux ratio (DNBR) is defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat g y f,1ux, and is indicative of the margin to DN8.

he M TAe_0$8. design _ basis is as fo1Ws- thars -'et be et least a 95 penent l

m^ ^ probability that the minimum DNBR of the liatting-rod during Condition-I-end-11 41stion bein evente-to-greater-than-en-equal-to-the-CMSR 14stt-of the Ok8-cor: 4 weed-7 i fuel-in-this-app 14 cation).-The-correlation-DNBR-limit-is established 5:50d en '

Ahe ~ entire applicable-experimental data set such that- there-is- a 95 percent probabillty-with-95-percent-conf 4 dance-that-DNS-wi4not-occur-when th: :!nM't _

ONRA-le-et-the-corcelqtio,n4pBy f att-fir 17-for$th 7 the;#A8,1 :n? ".5.-b$ p,/

oorrelationb jp c g_ g g eh ivd -plent lopetett i p:r:.::t+es, e,qk,g k-meet 4ne-th4senen%5iTG4ncertsintlesAn%%g nttchte-end-thermal-parameters, and fuel fabrication parameters-ere-considered statistica14y-such-that-there-is et 1 ::t :-95-confidence-that-the-stei :: OF""~

fee-the-l4miting rods-is-greater-than or equal to the-0NBR-limit. The en r-f teinties-in-the above plant parameters are used to deterstne the.plantJN84-.

l

un: :rt letyr-This-ONBA-encertaintyr-comb 4eed-wMh-th: :crr:1;ti:n 0""" li:M.

l us4ag-values-of-input-parametera-without-uncertakties2 R The-o f des values are 1.34 and 1.32 for a typical cell and a thimble cell, respectively for OF#tfuel, and 1.33 for a typical cell and 1.32 for a thimble cell for the (,

VANTAGE 5 fue1D In addition, margin has been maintained in both designs by meeting safety analysis DNBR limits of 1,49 for a typical cell and 1.47 for a k thimble es11 for 0FA fuel, and 1.67 and 1.65 for a typical cell and a thimble ('

cell, actival < fora ha VANTAGE 5 fue1 3 1n performing safety analyses. .

.m. 2 - wW :

ctreves-o figure'2.1-1*show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum '

design DNBR is no less than the design DNBP. value, or the average 6nthalpy at '

thevesselexitislessthan,theentj.a)py4:

s Qu, rat _ed iquid. -

y w-

%w gcLb Cwb ~ .

'-%uh\9i w.

, okL\\hff a

0ptimized_ Fuel Assemblies

. - eqC waCo %. um u. 2 % J a

8 2-1 Amendment No.

BRAIDWOOD - UNITS 1 & 2

l

. . q INSERT A The DNBR thermal design criterion is that the probability that DNB will not occur on the most limiting rod is at least 95% (at a 95% confidence level) for any Condition I or il event.

In meeting this design basis, uncertainties in plant operating parameters, nuclear and thermal parameters, and fuel fabrication parameters are considered. As described in the UFSAR, the effects of these uncertainties have been statistically combined with the correlation uncertainty. Design limit DNBR values have been determined that satisfy the DNB design criterion.

l

. t l

t 2

to

-TABLE 2.2-1 (Continued) 3 o

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS ~

6 a

TRIP SETPOINT ALLOWABLE VALUE 7FUNCTIONALUNIT

>90% of loop mini- >89.3% of loop #

I

$ 12. Reactor Coolant Flow-Low mum measured flow

minimum measured g flow

f w

Steam Generator Water

)

o- 13.

to Level Low-Low ( l

a. Unit 1 >33.0% of narrow >31.0% of narrow N range instrument range instrument /

span span [ i

>17% (Cycle 3); >16.3% (Cycle 3); )

b. Unit 2 [36.3%(Cycle 4 [34.8%(Cycle 4and{ l and after) of after) of narrow ;

narrow .?nge range instrument (

ro span instrument span c u' <

>5268 volts - >4920 volts -

14. Undervoltage - Reactor each bus each bus Coolant Pumps

>57.0 Hz >56.08 Hz

15. Underfrequency - Reactor Coolant Pumps

16. Turbine Trip

>1000 psig >815 psig

/ l

a. Emergency Trip Header }

Pressure '

>1% open >1% open I b. Turbine Throttle Valve _

R Closure j rn N.A. N.A. L 5 17. Safety Injection Input i M from ESF N.A.

N.A.

E

18. Reactor Coolant Pump Breaker Position Trip /

CMinimum measured flow = 97,600 gpm (cuy50pf

=w % um t w wuv a une o g s m oge9E T.

- w# N % E % w a i o,ek h w A she w g w aw e g eg,

e .

I POWER DISTRIBtJTION_ LINITS 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR LIMITING CONDITION FOR OPERATION 3.2.3 Indicated Reactor Coolant System (ACS) total flow rate and Ffg shall be maintained as follows for four loop operation.

4

a. 0 RCS Total Flowrote > 390,400 gpm, and Y>fuel b.$ NgF i 1,55 [1.0 + 0.3 (1.0-P)] for 0FA ,

Fh 51.65 [1.0 +0..' (1.0-P)) for VANTAGE 5 fuel where:

MeasuredvaluesofFhareobtainedbyusing.themovableincore detectors. An appropriate uncertainty of 4% (nominal) or greater I N

shall then be applied to the measured value of Ffg before it is compared to the requirements, and THERMAL POWER p , RATED THERMAL POWER APPLICABILITY: MODE 1.

ACTION:

dith RCS total flow rate or F outside the region of acceptable operation:

a. Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either:

1. Restore RCS total flow rate and Fh to within the above limits, or

2. Redu:e THERMAL POWER to less than 50% of RATED THERMAL POWER and rath:c6 thc Power Range Neutron Flux-High Trip Setpoint to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , m hp\ scrddQ_ -\t L\v\sk b CM Lk A& 2 LAv\4t \

cewm es cype n.

w @Qi>\rnb\L % LWt L mui da A Q ShdV\

I em tupt G, BRAIDWOOD - UNITS 1 & 2 3/4 2-8 AmendmentNo.[

' - + "

i ,

l POWER DISTRIBUTION LIMITS l IASO. i i

HEAT FLUX HOT CHANNEL FACTOR, and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE i HDT CHANNEL FACTOR (Continued)

c. The control rod insertion limits of Specification 3.1.3.6 are -

i ,

t maintained, and The axial power distribution, expressed in terms of AXIAL FLUX -

d.

DIFFERENCE, is maintained within the limits. ((pgte~ O*j ~

F g will be maintained within it6 limits provided the Conditions a.~through~ .

d. above are maintained.N The combination of the ACS flow requirement and the requirement ong guarantee that the DNBR analysis used-in-th de4 {

trill be met. (gc3, Margin between the sa fety~analysi's limit DNBRs7,ccy 49 and 1.47 for omo the OFA 4\g((\ygp ;

fuel typical and thimble cells, res ectively and 1. and 1.65 r the !

i VANTAGE 5 ty~pical and thimble cell nd the design limit DNBRs 1.34 and 1.32 i for the 0FA fuel tPjifHTand thimbi cells, and 1.33 anti,R _the i

respectivelypis maintained. A i i

VANTAGE 5 fuel typicp1 C W.; M rand h_thimblexalls.,in

@ e o.M AA\g?te \\@j ( '

A fraction of thitmirgiiris1tilized to' accommodate ~the trahsition core DNBR penalty (maximum of 12.5%) and the appropriate fuel rod now DNBR penalty (less than 1.5% per WCAP-8691, Revision 1). The rest of the mar in betwocn l design and safety, analysis DNBRy linjts-cap be gn flexibilhy. i ,

Un v C oyq Ogmd ~yerm,used for_ plant de\ &ncy keewrc.W 3

Mhe: RCSiflow requirer.ent-is based ~en tKa lo6p minimba me'asured fIowhate] ,

l of 97,600 gpm&hich is used in the Improved Thermal Design ProcedureVdescribed

? g> a w in 4&AR 4.4.1 and 15.0.3. A precision heat balance is performed once each cycle and is used to calibrate the RCS flow rate indicators. Potential fouling of the <

feedwater venturi, which might not be detected, could bias the results from the precision heat balance in a non-conservative manner. Therefore, a penalty of ,

. UE 14 assessed for potential feedwater venturi fouling. A maximum measurement

'.[y ancartainty of 2722%has been included in the loop minimum measured flow Ma'ccount for potential undetected feedwater venturi fouling and the use of the RCS flow indicators for flow rate verification. Any fouling which might bias the RCS flow rate measurement greater than 0.1% can be detected by monitoring and trending various plant perfonnance parameters. If detected, action shall be taken, before performing subsequent precision heat balance measurements, i.e. ,

either the effect of fouling shall be quantified and compensated for in the RCS flow rate measurement, or the venturi shall be cleaned to eliminate th6 fouling, j Surveillance Requirement 4.2.3.4 provides adequate monitoring to detect possible flow reductions due to any rapid core crud buildup.

Surveillance Requirement 4.2.3.5 specifies that the measurement instrumen-tation shall be calibrated within seven days prior to the performance of the ,

calorimetric flow measurement. This requirement is due to the fact that the drift effects of this instrumsntation are not included in the flow measurement uncertainty analysis. This rN.11 resent does not apply for the instrumentation those drift effects have been included in the uncertainty analysis.

NdN

% ACMM L M ON (W.d QQ uvA % e g _L. (b ,

BRAIDWOOD - UNITS 1 & 2 B 3/4 2-4 Amendment No.

ATTACHMENT 3 EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION Cornmonwealth Edison (CECO) has evaluated the proposed amendment and determined that it involves no significant hazards consideration. According to 10 CFR 50.92(c), a proposed amendment to an operating license involves no significant hazards if operation of the facility in accordance with the proposed amendment would not:

(1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.

The proposed changes would modify the Techr..al Specifications c mceming (1) the moderator temperature coefficient (MTC), (2) the boron concentration necessary to meet shutdown margin (SDM) requirements, and (3) the thermal design flowrate.

The MTC change would allow a slightly positive MTC (PMTC) below 100 percent of rated full power. The principal benefit of this change is that it would facilitate the design of future reload fuel cycles to yield significant fuel cost savings. Technical Specification changes are aiso required to meet SDM requirements to accommodate the positive MTC and the potential of lengthened reload fuel cycles due to increased energy requirements. To assure suberiticality requirements are met following a postulated loss-of-coolant accident (LOCA), the boron concentration is increased in the refueling water s! mage tank (RWST) and the accumulators. The safety analyses for the Byron and Brn md Updated Final Safety Analysis Report (UFSAR) transients have been p,6 >usly based on a maximum MTC being less than or equal to O pcm/*F at all times when the reactor is critical. The proposed change to the Technical Specification would allow a +7 pcm/*F MTC for power levels up to 70 percent with a linear ramp to O pcm/*F at 100 percent power. CECO has reviewed the revised USFAR safety analyses. These analyses conservatively bound the positive MTC and increase in boron concentration, incorporates the revised thermal design flows, and addresses increased tube plugging levels. The acceptable results of the revised analyses are provided in WCAP 13964 " Commonwealth Edison Company Byron and Braidwood Units 1 and 2 increased SGTP/ Reduced TDF/PMTC i Analysis Program Engineering / Licensing Report".

The thermal design flow (TDF) is a minimum RCS flow value assumed in the accident analyses and reactor core thermal / hydraulic design calculations. These calculations demonstrate that the necessary heat is removed from the core to meet various transient acceptance criteria. The minimum measured flow (MMF) currently used for the licensing basis is a total core flow of 390,400 gpm for Byron /Braidwood. I This value is reflected in Technical Specification Table 2.2-1 (Functional Unit 12) as !

l l

I

,.' l

)

a footnote of 97,600 ppm per loop for the reactor coolant flow-low reactor trip. The i MMF value must be serified in accordance with Technical Specification 3/4.2.3.

1 A reduction in TDF has been factored into the accident analyses that rely on RCS flowrate. This msults in a reduction in the limiting condition for operation (LCO) value for RCS flow reflected in the Technical Specifications. The reduced flow requirement continues to provide a margin to account for future steam generator tube plugging (SGTP). The revised TDF value corresponds to a MMF value of 371,400 gpm, which assumes a 3.5 percent flow measurement allowance. This value is reflected in the footnote to Technical Specification Table 2.2-1 as 92,850 gpm, minimum measured loop flow for the reactor coolant flow-low reactor trip. The revised LCO flow value will be incorporated in Technical Specification 3/4.2.3.

The proposed changes also include an administrative change to correct the wording in the MTC Technical Specification LCO 3.1.1.3a to clarify that both LCO 3.1.1.3a and b must be met over the fuel cycle.

WCAP 13964 utilized a NRC approved safety analysis methodology. Based on Commonwealth Edison's review and approval of WCAP 13964, it has been determined that the changes associated with the analyses do not involve a significant hazard. Specifically:

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

(1) The reduced thermal design flow and positive moderator temperature coefficient program includes corresponding increases to the RWST and accumulator required boron concentration. The analysis program and associated boron concentration changes will not affect the operability and integrity of plant systems and components. The analysis program also does not result in a condition that would challenge the design, material, and construction standards of the plant systems and components.

Additionally, the safety functions of the evaluated systems and components remain unchanged. The safety analyses necessary to support the reduced TDF and PMTC program were performed (WCAP 13964) and found to be acceptable and consistent with the Byron and Braidwood original safety analysis bases. All Departure from Nucleate -

Boiling (DNB) Ratio (DNBR) design limits were determined such that

there was a 95 percent probability at a 95 percent confidence level that a DNBR value of 1.25 for a typical and thimble cell were verified to have been met. The present Technical Specification limit for Nuclear Enthalpy Rise Hot Channel Factor, F" n, of less than 1.65 ensures that the limiting DNB ratio during normal operations and operational transients (Condition i I and Condition ll events) is greater than or equal to the DNBR limit of the correlation being applied tnes rus! integrity will not be challenged. ,

3-2

The accidents which are found to be sensitive to PMTC were analyzed as part of this effort and the results were found to be acceptable. On a cycle-by-cycle basis, the impact of PMTC on Anticipated Trip Without Scram (ATWS) risk will be addressed by determining the Unfavorable Exposure Time (UET) per established Westinghouse Owners Group methodology, with corrective actions to be taken as appropriate to assure acceptable risk. The increase in the RWST and accumulator boron concentration will have no adverse impact on the previously evaluated accidents. The SGTP/TDF/PMTC program does not affect the integrity of the safety related systems and components such that their function to control radiological consequences is affected and all fission barriers will remain intact. The effects on offsite doses have been considered. The incorporation of a PMTC, a reduction in TDF and increased tube plugging levels will result in a small increase in offsite doses, however, the total doses remain a small fraction of the 10 CFR 100 limits. As such, the accident analysis acceptance criteria continue to be satisfied.

Therefore, the probability or consequences of an accident previously analyzed in the UFSAR is not increased by the SGTP/TDF/PMTC program.

B. The proposed changes do not create the possibility of a new or different type of accident from any accident previously evaluated.

(2) The methodology and manner of plant operation as a result of the proposed changes is unchanged. The increased SGTP, reduced TDF, and PMTC program, which includes changes to the RWST and accumulator boron concentration, does not impact the safe operation of the reactor provided that the existing and proposed Limiting Conditior's for Operation (LCOs) and the associated action requirements are satisfied.

The reactor response to normal temperature fluctuations will be different due to PMTC, however, the normal reactor control systems, as designed, will continue to maintain a stable primary system temperature and reliable power production. The assumptions do not create any new failure modes that could adversely impact safety related equipment. The related Safety Limits and LCOs in the plant Technical Specifications will be evaluated and satisfied for each future reload core design via the 10 CFR 50.59 process. All DNBR Limits have been satisfied. The typical and thimble fuel cells were verified to maintain a DNBR value of 1.25 at a 95 percent probability and 95 percent conference level. Other than the analysis for tube plugging, the proposed changes do not involve any equipment additions or modifications at the stations. Current;y installed equipment will not be operated in a manner different than previously designed. Changes will be mede to technical data within the existing 3-3

f, station procedures, however, the analytical methods used to determine the data also remain unchanged. All aspects of the SGTP/TDF/PMTC program have been evaluated, and no new or different accidents or i failure modes have been identified for any system or component important to safety. No new credible limiting single failure has been created.

Because the SGTP/TDF/PMTC program does not adversely affect the j integrity of the steam generator or any other equipment, it is determined '

that the proposed changes do not create the possibility of a new or different type of accident from any accident previously evaluated.

C. The proposed changes do not !nvolve a significant reduction in a margin of safety.

(3) The performance and integrity of the evaluated safety-related systems and components are not affected by the proposed changes. The radiological consequences of all previously analyzed accidents remain unchanged. The reduced TDF and PMTC program, which includes changes to the RWST and accumulator boron concentration, will have no effect on the availability, operability, or performance of the evaluated safety-related systems or components. The reactor response to normal temperature fluctuations will be different due to PMTC, however, the normal reactor control systems, as designed, will continue to maintain a stable primary system temperature and reliable power production. The margin of safety associated with the licensing basis safety analysis is not significantly reduced by the proposed changes. All acceptance criteria for the specific UFSAR Chapter 15 safety analyses (Non-LOCA and l LOCA) have been satisfactorily evaluated and verified using NRC approved methodologies. Therefore, there is no significant reduction in the margin of safety as defined in the bases of any affected Technical Specification.

Based on the above evaluation, Commonwealth Edison has concluded that implementation of a PMTC, revised RWST and accumulator boron concentrations, and reduced RCS thermal design flow does not involve a significant hazards consideration with respect to the provisions of 10CFR50.92.

1 i

3-4

.

ML20071K973

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