ML20077K278

Proposed Tech Specs Re Removal of Some of Thimble Plugging Devices During Refuel 4 & Use of COLR

ML20077K278
Person / Time
Site:	Callaway
Issue date:	07/30/1991
From:	UNION ELECTRIC CO.
To:
Shared Package
ML20077K262	List: ML20077K260 ML20077K278
References
ULNRC-2450, NUDOCS 9108070174
Download: ML20077K278 (9)

Text

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4 ULNRC- 24 50 TECHNICAL SPECIFICATION BASES CHANGES Bases 2.1.1 Pages D 2-1 B 2-2 Bases 3/4.2.2 and Pages B 3/4 2-4 3/4.2.3 D 3/4 2-5 INSERT A Bases 3/4.3.3.6 Page B 3/4 3-4 Bases 3/4'.G.1.4 Pago B 3/4 6-2 Bases 3/4.7.1.1 Page B 3/4 7-1 l.

9108070174 910730

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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 result in excessive cladding temperatures because of the onset of departure from nacleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNS 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 the location of DNB for axially uniform and non-uniform heat flux distributions.

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

The DNB design basis is as follows: there must be at least 3 95 percent e probability that the minimum DNBR of the limiting rod during Condition I and II events is greater than or equal to the DNBR limit of the DNB correlation beieg used-(th "Il l correlati;n for Optimized fuci (OPA) = d the WRB-2 correlation for VANTAGE 5 fuel in th's application). The correlation DNBR limit is estab-lished based on the entire applicable experimental data set such that there is a 95 percent probability with 95 percent confidence that DNB will not occur when the minimum ONBR is at the DNBR limit (1.17 for-beth-the =C-i ad-WRB-2 correlatio ).

! In meeting this design basis, uncertainties in plant operating parameters, nuclear and thermal parameters, and fuel fabrication parameters are considered statistically such that there is at least a 95% probability with 95% confidence level that the minimum DNBR fer the limiting rod is greater than or equal to the DNBR limit. The uncertainties in the above plant parameters are used to determine the plant DNBR uncertainty. This DNBR uncertainty, combined with the correlation DNBR limit, establishes a design DNBR value which must be met in plant safety analyses using values of input parameters without uncertainties.

For Callaway, the design DNBR values are-1.32 and 1.35 for thimble ad typial nih, rapatively, fec OIA, end 1.33 and 1.34 for thimble and typical cells, respectively, for VANTAGE 5 fuel. In addition, margin has been maintained-4tr--

l b:th f=1 dai;a by meeting safety analysis DNBR limits of 1.42 ad 1.45 f;r L -thir,tle end typicel cells, resputively, for CIA, cad 1.61 and 1.69 for thimble L and typical cells, respectively, for VANTAGE 5 fuel.

The curves of Figure 2.1-1 show the loci of points of THERMAL POWER,

)/ Reactor Coolant System pressure and average temperature below which the calculated DNBR is no less than the design DNBR value or the average enthalpy

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at the vessel exit is less than the enthalpy of satursted liquid.

CALLAWAY - UNIT 1 B 2-1 Amendment No. L5,49,44

SAFETY LIMITS  :

BASES I 2.1.1 REACTOR CORE r as 7pec(CoQti tlie Ued)in & Core Q*rding limth Aeted (COLA')

Tl.e cur /es are based on a measured nuclear enthalpy rise hot _ channel factor, F H' O ' I* OI I U ' and a reference cosine with a peak of 1.55 for axial power shape. An allowance is includeo for an increase in F H at reduced power based on theg = prn :ica:: i-etudion 3 tren in & COL A.

r,g - 1. e C D 0. : (br):fcrOrA,cr.d-rN H ^

1*= 1+ c' (iF} Or v s s f=1 whcrc r i: the fracticr cf RATCO TllCr"AL POWCR.

Ther,e limiting heat flux conditions are higher than those calculated for -

- tne range of all-control rods fully withdrawn to the maximum allowable control-rod insertion assuming. the axial power. imbalance is within the limits of the f i(t!) function of the Overtemperature trip. When the axial power imbalance-is not within the tolerance, the axial power imbalance effect on the Overtem-perature AT-trips will reduce the setpoints to provide protection consistent with core safety limits.

- 2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this- safety limit protects the integrity of the Reactor Coolant System (RCS) from overpressurization and thereoy prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.

The reactor vessel,- pressurizer, and the RCS- piping and valves are designed to Section III of the ASME Code for Nuclear Power Plaic 3 which pemits a maximum transient pressure of 110% (2735 psig) of des %n pressure.

The-Safety Limit of 2735 psig is the efore consistent with the design criteria and associated Code require ents.

The entire RCS is hydrotested at greater than or equal to 125% (3110'psig) of design pressure to demonstrate integrity prior to initial-operation.

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CALLAWAY - UNIT 1 B 2-2 Amendment No. J5, 44

POWER' DISTRIBUTION LIMITS 1 - -

(BASES 3/4.2.2-'and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR AND NUCLEAR'ENTHALPY RISE HOT CHANNEL FACT,O_R, (Continued)

Each of these is measurable but will nonnally only be determined period-ically as specified in Specifications 4.2.2 and 4.2.3. This periodic surveil-lance is sufficient to ensure that the limits are maintained provided:

a. Control rods in a single group move together with no individual rod insertion differing by more than i 12 steps, indicated, from the-group demand position,

b. Control rod banks are sequenced with overlapping groups as described in Specification 3.1.3.6.

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

! -maintained.

d. The axial power distribution, expressed in tems of AXIAL FLUX DIFFERENCE, is maintained within the limits.

3g FN will be maintained within its limits provided conditions a. through d.

above are maintained. The relaxation of F N as a function of THERMAL POWER allows-_ changes in the radial power shape fN all pennissible rod insertion limits.

When an Fq measurement is taken, an allowance for both experimental error and manufacturing tolerance must be made. An allowance of 5% is appropriate for a-full-core map taken with the incore detector flux mapping system and a 3%

allowance is appropriate for manufacturing tolerance.

N

, When F 3g is measured (i.e., inferred), no additional allowances are

, -necessary prior to comparison with the limits of Section 3.2.3. An error allow-l ance of 4% has been included in the limits of Section 3.2.3.

Specifications 3.2.2 and 3.2.3 contain the F q nd a F-delta-H l_imits appli-cable to VANTAGE 5 fuel. The OFA f;:1 i: :n:ly :d t; icwer limits-sinc; it wili h v; cap;rienced burnup, thereby redu;ing the attainable OFA-specific het

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_ cf the fecters OFA fuel such will thet be much the expected less then peak pcwer icvels that necessary and peekth; tc approach radi;l 0FAp;w F r g

L  :.nd F-delta P. n:ly;i: limit:.

Margin between the safety analysis DNBR limits (1.12 :nd 1.5 for the

- Optim.ized fuel thimble and typicci ccil:, rc:pc;tively, :nd 1.61 and 1.69 for the VANTAGE 5 thimble and typical cells) and the design DNBR limits (1. :nd 1.35-f;r thc Optimized fuel tWti; a6 typic;l :cil; and 1.33 and 1.34 for the VANTAGE 5 thimble and typical cells, respectively) is maintained. A fraction of this margin is utilized to accomodate the tran;ition 0;rc 0.TR p;n:lty l J l CALLAWAY - UNIT 1 B 3/4 2-4 Amendment No. J5,28, 44 l

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1 POWER DISTRIBUTION LIMITS BASES 3/4.2.2_and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR AND NUCLEAR ENTHALPY R!S

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HOT CHfNNEL FACTOR (Continued) c ZNS6KrA (1 !A fvr "ANTA E 5 tuel) ;nd th

• appropriate fuel rod bow DNBR penalty (+en--/3#4

- & n 1.5L per WCAP 8691, Rev. 1)g e The margin between design and safety analysis DNBR limits of 0.3% for Optimizco fusi end- 17.4% for VANTAGE 5 fuel includes greater than-3hmargin for bcth Opti-ized fuel and-V"' TACE 5 fac1 for plant designflexibil{ty.

9 'lo The hot channel factor F (z) is measured periodically and increased by a cycle and he1ght dependent power factor appropriate to either Nomal Operation or RESTRICTED AFD OPERATION, W(z)NO or Wz)RAFD0, to provide assurance that the limit on the hot channel factor, F (z), is met.

9 W(z)NO accounts for the effects of normal operatioa transients and was detemined from expected power control maneuvers over the full range of burnup conditions in the core.

W(z)RAFD0 accounts for the more restrictive operating li.iits required by RESTRICTED AFD OPERATION which result in less severe transient values.The W(z) functions are specifiea in the Core Operating Limits Report per Specification 6.9.1.9.

" Provisions to account for the possibility of decreases in margin to the FQ(z) limit during intervals between surveillances are provided. Any decrease in the minimum margin to the F0 (z) limit compared to the minimum margin detennined from the previous flux map is determined by comparing the ratio of:

maximum Fh(z) over z

( g(7) taken from the current map to the same ratio from the previous map. The ratios to be compared from the two flux maps do not need to be calculated at identical z locations. Increases in this ratio indicate that the minimum margin to the i

FQ (z) limit has decreased and that additional penalties must be applied to the measured QF (z) to acccunt for further decreases in margin that could occur before the next surveillance. More frequent surveillances may also be substi-tuted for the additional penalty.

3/4.2.4 QUADRANT POWER TILT RATIO The QUADRANT PCWER TILT RATIO limit assures that the radial power distri-bution satisfies the design values used in the power capability analysis.

Radial power distribution measurements are made during STARTUP testing and periodically during power operation.

The limit of 1.02, at which corrective action is required, provides DNB and linear heat generation rate protection with x-y plane power tilts. A CALLAWAY - UNIT 1 8 3/4 2-5 Amendment No U,28,44 j/ , 5c

F INSERT A

.the flow anomaly penalty (3.3%), and the thimble plug removal penalty (3.1%).

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• q,'.';g2Q INSTRUMENTATION

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REVISIO$ 1

') BASES jg i -

y&j.. c 3/4.3.3.3 SLISMIC INSTRUMENTATION The OPERABILITY of the seismic instrumentation ensures that sufficient '

capability is available to promptly determine the magnitude cf a seismic event and evaluate the response of those features important to safety. This capability is required to permit comparison of the measured response to that used in the j design basis for the facility to determine if plant shutdown is required j pursuant to Appendix A of 10 CFR Part 100. The instrumentation is consistent 1 with the recommendations of Regulatory Guide 1.12, " Instrumentation for )

Earthquakes," April 1974 i i

I 3/4. '! . 3. 4 METEOROLOGICAL INSTRUMENTATION .

i The OPERABILITY of the meteorological instrumentation ensures that suf ficient meteorological data are available for estimating potential radiation doses to the public as a result of routine or accidental release of radioactive materials to the atmosphere. This capability is required to evaluate the need ..

lor initiating protective measures to protect the health and safety of the public and is consistent with the recommendations of Regulatory Guide 1.23, "Onsite Meteorological Programs," February 1972.

3/4.3.3.5 REMOTE SHUTDOWN INSTRUMENTATION jg the OPERABILITY of the Remote Shutdown System ensures that sufficient W,.w '

capability is available to permit shutdown and maintenance of HOT SHUTDOWN of the facility.from locations outside of the controi room and that a fire will not preclude achieving safe shutdown. The Remote Shutdown System transfer switches, power circuits, and control circuits are independent of areas where a fire could damage systems normally used to shutdown the reactor.

This capability is required in the event control room habitability is lost and is consistent with General Design Criteria 3 and 19 and Appendix R of 10 CFR Part 50.

1 t/4.3.3.6. ACCID [NT MONITORING INSTRUMENTATION l

lhe GPLRABILITY of the accident monitoring instrumentation ensures that

' sulIir.ient information is atailable on selected plant parameters to monitor and assess these variables following an accident. This capability is consistent with the recommendations of Regulatory Guide 1.97, Revision 2, " Instrumentation f or Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," December 1980, and NURW-0737, " Clarification of IMI Action Plan Requirements," November 1980, A.r chr ///e/ in /~.f"M g A en//x7f, f

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CONTAINMENT SYSTEgm REVistoy ,7

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._ 3/4.6.1.4 INTERNAL PRESSURE i.

The limitations on containment internal pressure ensure that: (1) the *L containment structure is prevented from exceeding its design negative pressure differential with respect to the outside atmosphere of 3.0 psig, and (2) the -

containment peak pressure does not exceed the design pressure of 60 psig during steam line breek conditions. ,

e#1 The marimum peak pressure expected to be obtained from a steam line break event is 48"psig. The limit of 1.5 psig for initial positive containment pressure  !

will limit the total pressure to psig, which is less than design pressure and is consistent with the safety a yses. ,

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3/4.6.1.5 AIR TEMPERATURE 1I 3

The limitations on containment average air temperature ensure that the . ..

overall containment average air temperature does not exceed the initial .

temperature condition assumed in the safety analysis for a steam line break  :

accident. Measurements shall be made at all listed locations, whether by '

3 fixed or portable instruments, prior to determining the average air temperature. -

} + 3/4.6.1.6 CONT AINMENT VESSEL STRUCTURAL INTEGRITY g' This limitation ensures that the structural integrity of the containment vessel will be maintained in accordance with safety analysis requirements for the life of *.he facility. Structural integrity is required to ensure that the containment will withstand the maximum pressure of 50 psig in the event of a steam line break accident. The measurement of containment tendon lift-off force, the tensile tests of the tendon wires or strands, the visual examination of tendons, anchorages and exposed interior and exterior surfaces of the

. containment, and the Type A leakage test are sufficient to demonstrate this capability.

The Surveillance Requirements for demonstrating the containment's structural integrity are in compliance with the recommendations of proposed Regulatory Guide 1.35, " Inservice Surveillance of Ungrouted Tendons in Prestressed Concrete Containment Structures," April 1979, and proposed Regulatory Guide 1.35.1,

" Determining Prestressing Forces for Inspection of Prestressed Concrete Con-tainments," April 1979.

The required Special Reports from any engineering evaluation of containment abnormalties shall include a description of the tendon condition, the condition of the concrete (especially at tendon anchorages), the inspection procedure, the tolerances on cracking, the results of the engineering evaluation and the corrective actions taken.

CAILAWAY - UNif i B 3/4 6-2 '

_ _ _ . ._ _ _ _ _ _ .._ ~- _._ _ . - .. _ _ _ _ .- _ _ _ _ ._ __ __ . . _ .

i , W h-fN M I Q ~*'[.~ E *. N O I b E 7 7 W Q W *3 N V 7,'";*' % : Ms F. g. 3 , g .V ",T .;4.J _ y<,g 3/4.1 PLANT SYSTEMS YlSfQg y ,

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3 *i' t. BASE 5

- m6 fj, 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line Code safety valves ensures that the 3econdary Coolant System pressure will be limited to within 110% (1320 psia) of its desiga pressure of 1200 psia during the most severe anticipated system operational transient. The maximum relieving capacity is associated with a '

Turbine trip from 102% RATED THERMAL POWER coincident with an assumed loss of no steam bypass to the condenser).

condenser heat' sink (i.e., /R p 3 .

The specified Val lift settings and relieving capacities are in accordance with the ra luirements of Section III of the ASME Boiler and Pressure Code, (1971 Edition). The total relieving capacity for all valves on all of the steam lines is (M M x 10s) lbs/h wCch is 115% of the total secondary steam flow af Gre-x 10' lbs/h at RATED THERMAL POWE A minimum of-two OPERACLE s 'ety valves per steam ge eratoi- ensures tha sufficient relieving capa ity is available for th allowable THERMAL OWER restrict on in ..

Table 3.7 /8. f 2 /M#/* W//A /4 *

• yenere or erplujyed.

STARIUP and/or POWER OPERATION is allowable with safety valves inoperable '

l within the limitations of the ACTION requirements on the basis of the reouction in Secondary Coolant System _ steam flow and THERMAL POWER required by the reduced .

• Reactor Trip Settings of the Power Range Neutron Flux channels. The Reactor -

,. dj Trip Setpoint reductions are derived on the following bases:

ForfouS1'oopoperation: .-

i SP =

{ x (109) ,

i Where:.

4 SP = Reduced Reactor Trip Setpoint in percent of RATED THERMAL POWER, v' = Maximum number of inoperable safety valves per steam line, l

CALLAWAY - UNIT 1. 8 3/4 7-1

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