ML20205S024

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Proposed Tech Specs,Supporting Plant Uprating to 3,565 Mwt Core Power Level.Supporting Documentation Encl
ML20205S024
Person / Time
Site: Callaway Ameren icon.png
Issue date: 03/31/1987
From:
UNION ELECTRIC CO.
To:
Shared Package
ML20205R988 List:
References
NUDOCS 8704060505
Download: ML20205S024 (21)
v • d • e


Text

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.FA c.rtzrf: OMMzNG L.ccENre NW-30 l

, M (4) UE, pursuant to the Act and 10-CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source of special nuclear material without. restriction to chemical or phys-ical fonn, for sample analysis or . instrument calibration or asso-ciated with radioactive apparatus or components; and (5) UE, pursuant to the Act and 10 CFR Parts'30,' 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. This license shall be deemed to contain and is subject to the conditions specified in the Comission's regulations set forth in 10 CFR Chapter .I .

and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Comission now or hereafter in effect; and is subject' to the additional conditions'specified or incorporated below:

4 (1) Maximum Power Level

( 3.4"45 UE is authorized to operate pe facility 'at reactor core power levels not in excess of G1huegawatts thermal (100% power) in i

accordance with the conditions specified herein and in Attach-ment 1 to this license. The preoperational tests, startup tests 4 and other items identified in Attachment 1 to this license shall be completed as specified. Attachment 1 is hereby incorporated

- into this license. .

! (2) Technical Soecifications and Environmental Protection Plan TheTechnicalSpecificationscontainedinAppendixAa$dtheEnviron-

  • mental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this ifcense. UE shall operate the facility in accordance with the Technical Spec-3

- ifications and the Environmental Protectian Plan; i '3) Environmental Oualification (Section 3.11, SSER #3)*

Nove ***

(a) Prior to :'~. .mfgo-30,85,

.. . . ,v l9 UE shall environmentally qualify all electrical equipment accordingly to the provisions of 10 CFR 50.49. ,

, (b) Prior to restart following the first refueling outage, UE shall have qualified the reactor vessel-level instrumentation system high volumb sensor, t

  • The parenthetical notation following the' title of many license conditions denotes the section of the Safety Evaluation ~ Report and/or its supplements wherein the license condition is discussed. .

kt fla affrevsk. amendmark.r.

      • Saa Amenolma k sk. S.

8704060505 870331 i PDR ADOCK 05000483

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INDEX DEFINITIONS SECTION . PAGE 1.0 DEFINITIONS 1.1 ACTI0N............................................................. 1-1 1.2 ACTUATION LOGIC TE5T............................................... 1-1 1.3 ANALOG CHANNEL OPERATIONAL TE5T.................................... 1-1 1.4 AXIAL FLUX DIFFERENCE.............................................. 1-1 1.5 CH AN N E L CA L I B RATI O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 1.6 CHANNEL CHECK...................................................... 1 - 1.7 CO NTA I NMENT I NTEG RITY. . . . . . . . . . . . . . . . . . . . . . . . . . . '. . . . . . . . . . . . . . . . . . . 1-2

1. 8 CONTROLLED LEAKAGE................................................. 1-2
1. 9 CORE ALTERATION.................................................... 1-2 3,40 AE512k TYJdlMWPO4V. ./. . ,(. . . /. . ./ . . .j(. . . /. . /. . ./. . ./. . / f) 1.x /co05 E EQUIVA LENT I-131. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-2 1-3 1.X//l-AVERAGEDISINTEGRATIONENERGY....................................

n' 1. M/2 ENGINEERED SAFETY FEATURES RESPONSE TIME. . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 1-3

1. J4/J FREQUENCY N0TATION. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

1.JyfDENTIFIEDLEAKAGE................................................. 1-3 1.)yfMASTERRELAYTE5T................................................. 1-3 1.32/hMEMBER(5) 0F THE PUBLIC............................................ 1-3 1-4

1. w/@FFSITE 005E CALCULATION MANUAL. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

1-4 1.)f//0PERABLE-0PERASILITY.............................................

1- 4 1 36/(O P ERATI ONA L MOD E - MOD E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

1-4

1. j%2C PHY S I C S T E ST 5. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

1-4 1.J2)tlPRE55URE BOUNDARY LEAXAGE..........................................

1-4

1. P 22f RO C E5 5 CO NTRO L P R0G RAM. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

1-4 1.f4MPURGE-PURGING.................................................... 1-5 1.)$2%UADRANTPOWERTILTRATI0..........................................

1-5 1.)6),f RAT ED THE RMA L P0WE R. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

1-5 l 1.J 1)f REACTOR TRIP SYSTEM RESPONSE TIME. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

1-5

1. p$ E P O RTAB L E EY' INT. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

1-5

1. p)f5 HUTDOWN MA RG I N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

I Amendment No. 15 CALLAWAY - UNIT 1

l

.. 1 1

INDEX DEFINITIONS SECTION

- PAGE DEFINITIONS (Continued) 1.)02.15 I TE 8 0 UND A R Y. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 1.)i3 4 LAVE RELAY TE57.................................................. 1-5

. 1.aryi SOLIDIFICATION......._............................................. 1-5

1. g 2,50VRCE y CHECK...................................................... 1-5
1. My STAGG ERED TEST 8AS IS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-6

. - 175JpERMALP0WER..................................................... 1 1.)(gfTRIP ACTUATING DEVICE OPERATIONAL TE5T. . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-6

1. p VNIDENTIFIED LEAKAGE.............................................. 1-6
1. mn UNRESTRICTED AREA................................................. 1-6 1.)9fpVENTILATION EXHAUST TREATMENT SY5 TEM.............................. 1-6
1. 40 piVENT I N G . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-6 1./1 A ASTE GAS HOLDUP 5YSTEM........................................... 1-7 TABLE 1.1 FREQUENCY N0TATION........................................... 1-8 TABLE 1.2 OPERATIONAL M0 DES............................................ 1-9 I

I 1

1 CALLAWAY - UNIT 1 II Amendment No. 15

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- .. 1,

l DEFINITIONS CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when:

a. All penetrations reauired to be closed during accident conditions are either: ,
1) Capable of being closed by an OPERABLE containment automatic isolation valve system, or ,

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2) Closed by manual valves, blind flanges, or deactivated automatic l valves secured in their closed positions, except as provided in Table 3.6-1 of Specification 3.6.3.
b. All equipment hatches are closed and sealed,

~

- c. Each air lock is in compliance with the requirements of Specification

_ 3. 6.1. 3, , ,

d. The containment leakage rates are within the limits of Specification 3.6.1.2, and .
e. The sealing mechanism associated with each penetration (e.g., welds, bellows, or 0-rings) is OPERABLE.

CONTROLLED LEAKAGE

1. 8 CONTROLLED LEAKAGE shall be that seal water flow from the reactor coolant pump seals.

CORE ALTERATION

1. 9 CORE ALTERATION shall be the movement or manipulation of any component within the reactor vessel with the vessel head removed and fuel in the vessel.

Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.

DES)GN THEPML POWEV 10 D IGN THE L POWER hall be design to 1 reac r core eat tr sfe ate t the rea tor coola of 3565 t.

DOSE EQUIVALENT I-131

[0 '

1 32 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie / gram) l which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in I Table III of TID-14844, " Calculation of Distance Factors for Power and Test  ;

Reactor Sites."

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, CALLAWAY - UNIT 1 1-2 Amendment No. 15

,-r - , . , - - -

DEFINITIONS Y - AVERAGE DISINTEGRATION ENERGY

1. I shall be the average (we'ighted in proportion to the concentration of l each radionuclide in the reactor coolant at the time of sampling) of the sua of the average beta and gamma energies per disintegration (in MeV) for ifotopes, other than todines, with half-lives greater than 15 minutes, making up at least 95% of the total noniodine activity in the coolant.

ENGINEERED SAFETY FEATURES RESPONSE TIME 2

1. The ENGINEERED SAFETY FEATURES (ESF) RESPONSE TIME shall be that time j i terval from when the monitored parameter exceeds'its ESF Actuation Setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e. , the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include

. diesel generator starting and sequence loading delays where applicable. .

FREQUENCY NOTATION

~

I3

1. 4 Re)quirementsshallcorrespondtotheintervalsdefinedinTable1.1.The FREQUENC l IDENTIFIED LEAKAGE
1. IDENTIFIED LEAKAGE shall be:

l

a. Leakage (except CONTROLLED LEAKAGE) into closed syste.ms, such as ,

i pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or

b. Leakage into the containment atmosphere from sourcas that are both specifically located and known either not to interfere with the operation of Leakage Detection Systems or not to ce PRESSURE BOUNDARY LEAKAGE, or 1
c. Reactor Coolant System leakage through a steam generator to the l Secondary Coolant System.

MASTER RELAY TEST l

1. A MASTER RELAY TEST shall be the energization of each master relay and l verification of OPERABILITY of each relay. The MASTER RELAY TEST shall include i a continuity check of each associated slave relay. l MEMBER (5) 0F THE PUBLIC
1. MEMEER(S) 0F THE PUBLIC shall include all persons who are not occupa- l l tionally associated with the plant. This category does not include employees l of the licensee, its contractors or vendors. Also excluded from this category '

are persons who enter the site to service equipment or to make deliveries.

This category does include persons who use portions of the site for recrea-tional, occupational, or other purposes not associated with the plant.

CALLAWAY - UNIT 1 1-3 Amendment No. 15 l

1

l DEFINITIONS l

OFFSITE DOSE CALCULATION MANUAL

1. The 0FFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology l and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip Setpoints, and in the conduct of the Environmental Radiological Monitoring Program.

OPERABLE - OPERABILITY

1. A system, subsystem, train, component or device shall be OPERABLE or }

have OPERABILITY when it is capable of performing its specified function (s),

and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its

-function (s) are also capable of performing their related support function (s). -

. OPERATIONAL MODE - MODE

1. An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive l combination of core reactivity condition, power level, and average reactor coolant temperature specified in Table 1.2.

PHYSICS TESTS

1. PHYSICS TESTS shall be those tests performed to measure the fundamental l 1, nuclear characteristics of the core and related instrumentation: (1) described
in Chapter 14.0 of the FSAR, or (2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission.

PRESSURE BOUNDARY LEAKAGE

1. PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube l 1(akage) through a nonisolable fault in a Reactor Coolant System.cceponen' body, pipe wall, or vessel wall.

PROCESS CONTROL PROGRAM

1. The PROCESS CONTROL PROGRAM shall contain the current formula, sampling, [

analyses, tests, and determinations to be made to ensure that the processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulatea wet solid wastes will be accomplished in such a way as to assure. compliance with 10 CFR Part 20, 10 CFR Part 71 and Federal and State regulations, burial ground requirements, and other requirements governing the disposal of the radioactive waste.

PURGE - PURGING l 1.kPURGEorPURGINGshallbeanycontrolledprocessofdischargingairor l gis from a confinement to maintain temperature, pressure, humidity, concentra-tion or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

1-4 Amendment No.15 CALLAWAY - UNIT 1

DEFINITIONS - -

QUADRANT POWER TILT RATIO

1. QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore l detector calibrated output to the average of the upper excore detector cali-brated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. With one excore detector inoperable, the remaining three detectors shall be used for computing the average.

RATED THERMAL p0WER 1 RATED THERMAL POWER shall be a total core heat transfer rate to the [

reactor cooiant of L. .

REACTOR TRIP SYSTEM RESPONSE TIME

~ ~

1. The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from

- ]

~

when the monitored parameter exceeds its Trip Setpoint at the channel sensor until loss of stationary gripper coil voltage.

REPORTABLE EVENT

1. A REPORTABLE EVENT shall be any of those conditions s'pecified in l Section 50.73 to 10 CFR Part 50.

SHUTDOWN MARGIN

1. SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which l the reactor is suberitical or would be suberitical from its . resent condition assuming all full-length red cluster assemblies (shutdown anc control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.

SITE 20UNDARY 1 The SITE BOUNDARY shall be that line beyond which the land is neither l owned, nor leased, nor otherwise controlled by the licensee.

SLAVE RELAY TEST

1. A SLAVE RELAY TEST shall be the energization of each slave relay and [

verification of OPERA 3ILITY of each relay. The SLAVE RELAY TEST shall include a continuity check, as a minimum, of associated testable actuation devices.

SOLIDIFICATION

1. ! SOLIDIFICATION shall be the conversion of wet wastes into a form that l meets shipping and burial ground requirements.

SOURCE CHECK -

1. A SOURCE CHECK shall be the qualitative assessment of channel response l when the channel sensor is exposed to a source of increased radioactivity.

CALLAWAY - UNIT 1 1-5 Amendment No.15 i

'EFINITIONS D .

(

1 STAGGERED TEST BASIS 1.)( A STAGGERED TEST BASIS shall consist of: -

N a. A test schedule for n systems, subsystems, trains, or other designated components obtained by dividing the specified test interval into n equal subintervals, and

b. The testing of one system, subsystem, train, or other designated

{

component at the beginning of each subinterval.

THERMAL POWER 1.MHERMAL POWER shall be the total core heat transfer rate to the reactor coolant. l

_ TRIP ACTUATING DEVICE OPERATIONAL TEST . -

1.

_ Tr[ip Actuating Device and verifying OPERABILITY of alarm, l int trip functions. The TRIP ACTUATING DEVICE OPERATIONAL TEST shall include adjustment, as necessary, of the Trip Actuating Device such that it actuates at the required Setpoint within the res ifc red accuracy.

UNIDENTIFIED LEAKAGE 1.M UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE. l UNRESTRICTED AREA

1. An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY l

access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials, or any area within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes.

VENTRATION EXHAUST TREATMENT SYSTEM 1.I1l A VENTILATION EXHAUST TREATMENT SYSTEM shall be any system designed l andt installed to reduce gaseous radiciodine or radioactive material in particulate '

form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or partic-ulates from the gaseous exhaust stream prior to the release to the environment.

Such a system is not considered to have any effect on noble gas effluents.

Engineered Safety Features (ESF) Atmospheric Cleanup Systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

VENTING

1. k VENTING shall be any controlled process of discharging air or gas from a l confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not pro-vided or required during VENTING. Vent, used in system names, does not imply a VENTING process.

1 CALLAWAY - UNIT 1 1-6 Amendment No. 15 l l

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~

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DEFINITIONS .

WA$TE GAS HOLDUP SYSTEM

1. A WASTE GAS HOLOUP SYSTEM shall be any system designed and installed to l reduce radioactive gaseous effluents by collecting Reactor Coolant System of.'-

gases from the Reactor Coolant System and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

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..g.- j 580 .

__a 0.4 0.6 0.8 1.0 1.2 0 0.2 FRACTION OF RATED THERMAL POWER i

FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION CALLAWAY - UNIT 1 2-2

_ - . _-. -._- _ _ -_ _- _ _ _ _ _ = . - - -

^

R 'I C.

.)

j I

4 n

TABLE 2.2-1 (Continued)

  1. TABLE NOTATIONS l* C i E NOTE 1: OVERTEMPERATURE AT l -< -

y AT h [ (3 f ,,3) $ AT, (K -K h [T (3 f ,,5) - T'] + Ka(P - P') - f (AI)]

z  :

4  !

H Where: =

AT Measured AT by RTD Manifold Instrumentation; I * =

Lead-lag compensator on measured AT; 4 .

,  ! 13,.t = Time constants utlitzed in lead-lig compensator for AT, is = 8 s,

! v = 3 s; 3 f gg = Lag compensator on measured AT; -

'? = Time constant utilfred in the las compensator for AT, ta = 0 s; w

ta .

= F (Referenced AT at .d.::;; THERMAL POWER); I AT, .

Ki = 1.15; K = 0.0251/*F; l I

f [' =

The function generated by the lead-lag compensator for T,,,

dynamic compensation;

14. Is ' '= Time constants utfifred in the lead-lag compensator for T,,,, t, = 28 s, ts = 4 s; .

T = Average temperature. *F; l

[

a 1

g 3 , ,,3

=

Lag compensator on measured T,,,;

E y te =

Time constant utilized in the measured T,,, lag compensator, te = 0 s;

! t;' .

e s

v' s.) '

g TABLE 2.2-1(Continyjg h TABLE NOTATIONS (Cg(!1nued) i N -

, NOTE 1: (Continued)

T' <

508.4*f (Referenced T,,, a THERMAL POWER);

e K3 = 0.00116;

  • P = Pressurizer pressure, psig; P' = 2235 p'sig (Nominal RCS operating pressure);  !

5 = Laplace transform operator, s 8;  ;

and f (AI) is a function of the indicated difference between top and bottom detectors of the power-range neutron ion chambers; with gains to be selected based on measured instrument j

y response during plant STARIUP tests such that: r -

(1) For qt b

% etween -35% and + 6%, f (AI) = 0, where qt and g are percenh THERMAL l

POWER in the top and bottom halves of the core respectively, and qt

  • i% s total ME N i POWER in percent of A g THERMAL POWER; !_ ,

(11) For each percent that the magnitude of. qt ~ % exceeds -35%, the AT Trip setpoint shall be automatically reduced by 1.91% of its value at TifERMAL POWER; and l (iii) For each percent that the magnitude '6f qt -

% exceeds +6%, the AT Trip Setpoint shall be automatically reduced by 1.89% of its value at MT MEN NR. -

gkM 4

i *

[ NOTE 2: The channel's maximum Trip setpoint shall not exceed its computed Trip setpoint by more than

  • zee %

of AT span. 23 E

  • r e

i .

3 m "I p, TABLE 2.2-1 (Continued) -

h TABLE NOTATIONS (Continued) 5 i 5

-< NOTE 3: OVERPOWER AT -

t g

AT (1 f 13

5) $ AT, (K4 - Ks IYty5 1 I rs e T - K. [T (3 f ,,5) - M - f (at))

t Where: AT = Measured AT by RTD Manifold Instrumentation; .

I 3

= lead-lag compensator on measured AT; t,t i

= Time constants utilfred in lead-lag compensator for AT, ti, = 8 s., ta = 3 s; ,

d, Ifrs a

= Lag compensator on measured AT; 13 = Time constant utilized in the la compens'ator for AT, is = 0 s; AT, . = *F (Referenced AT at THERMAL POWER);

K, = 1.000; -

1 Ks = 0.02/*F for increasing average temperature and 0 for decreasing average temperature; I

3j'S 3 = 'The function generated by the rate-lag compensator for T,,, dynamic ,

compensation; I

g t, = Time constant utilized in the rate-lag compensator for T,,,, t, = 10 s; I

g 1+ 5 = . Lag compensator on measured T,,g;

! a - -

t g 1 =

Time constant utlitzed in the measured T,,, lag compensator, to = 0 s; l V.

f 9 0

l 1 e O

~

l '

TA8tE 2.2-1 (Continued) '

9 j {E TABLE NOTATIONS (Continued)

NOTE 3: (Continued) b Ka = 0.0065/*F for T > T" and Ka = 0 for T < T"; l

-e

" T = Average Temperature. *F; j T" =

Indicated T,,,at THERMAL POWER (Calibration temperature for AT f

instrumentation, < 588.4*F);

S = Laplace transform operator, s s; and .

I.

f (AI) = 0 for all AI. ,

j.

m NOTE 4: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than

+t1% of AT span. l O

33 O

E E, -

g .

t

'O e

.Attschmsnt 3 ULNRC-1471 March 31, 1987 CALLAWAY UPRATING SUBMITTAL

, ATTACHMENT 3 ENVIRONMENTAL EVALUATION Union Electric has performed an environmental evaluation to 4 - support uprating the Callaway Plant from the currently licensed power level to the core ~ thermal power of 3565 MWt which corresponds to a NSSS power level of 3579 MWt. This change does not provide a significant-adverse environmental impact, nor a significant increase ~in effluent- nor an adverse land or cultural resource altering act' >

,. The original licensing evaluations for the plant, including the NRC Environmental Evaluations (Ref. NUREG-75/Oll, 3/75, Section 1.1) , were based on a NSSS thermal power level of 3579 MWt. Therefore, the. proposed uprating remains within the bounds of the original environmental analyses.

I l

hh-

Attachment.4 ULNRC-1471 March ~31, 1987 i

CALLAWAY UPRATING SUBMITTAL ATTACHMENT 4 SIGNIFICANT HAZARDS EVALUATION t

1 i

i

)

l 5

l l

4 L

r Attachment-4 1.

-ULNRC-1471 March 31, 1987

SIGNIFICANT HAZARDS EVALUATION

! FOR CALLAWAY' PLANT UPRATING'TO.3579 MWt i

i 2

i -This evaluation supports Union Electric Company's license i amendment request to increase the thermal output of the Callaway Plant. At present,-Callaway is licensed to operate at.a reactor core thermal power level of 3411 MWt, which corresponds to a Nuclear Steam Supply System-(NSSS) thermal power level-of 3425 MWt.- This amendment requests the necessary License and. Technical Specification changes to operate the Callaway Plant at a reactor

  • core thermal power level of 3565 MWt, which corresponds to an NSSS thermal power level of 3579 MWt. This represents an 1

increase of 4.5% and is equivalent to the Engineered Safety

Features Design Rating of the Callaway Plant.- Union Electric has
submitted a license amendment request (ULNRC-1470, dated March 1

31, 1987) to utilize Westinghouse 17x17 VANTAGE 5 fuel beginning

, with Callaway Cycle 3 operation. The' requested uprating-is scheduled to begin concurrently with Cycle 3 operation.

In order to implement the uprating, Callaway Plant, Facility I Operating License NPF-30, Section 2.C.(1) requires revision to i indicate operation is authorized at reactor core power levels not in excess of 3565 megawatts thermal. Technical Specification i changes include revision of the definitions of Design Thermal Power (Section 1.10) and Rated Thermal Power (Section l.26),

rescaling of Figure 2.1-1 (Core Safety Limits), and revision of the notes associated with OT Delta T and OP Delta T (Table 2.2-1).

~

! In support of the requested uprating, the NSSS system and component designs were analyzed to verify continued compliance with licensing criteria and standards currently required by the Callaway operating license and the functional requirements specified in the FSAR for operation at 3579 MWt. The. review was

. conducted in accordance with the methodology documented in Westinghouse topical report WCAP-10263, "A Review Plan for-i Uprating'the Licensed Power of a Pressurized Water Reactor Power Plant". This methodology has been referenced in connection with the approval of similar license amendments for!other facilities. ~

WCAP-lO263 methodologies were also utilized in the review of the i interfaces between the NSSS and th'e' Balance of Plant-(BOP) systems'and components. Reviews of the BOP systems and components were conducted'and verify their capability to' meet the requirements for operation at 3579 MWt. . Review'of the turbine >

generator is based on the Valves Wide Open heat balance which-represents an NSSS power of 3562 MWt. The ability of the turbine.

generator to operate typ to 3579 MWt will be determined by careful-i i-f

. .__.1__..,-, _ ___ . _ . __m...a__,_,...-,_,.., ,, ,, u m._

monitoring of plant performance parameters. These reviews demonstrate that the Callaway Plant is capable, in the present design configuration, of operating at the uprated conditions without violating any of the design or safety limits specified in the Callaway FSAR and Facility Operating License NPF-30.

In conjunction with the use of Optimized Fuel Assemblies (OFA) during Callaway Cycle 2 operation, accident analyses were performed at the uprated power conditions. These were documented in ULNRC-1207, dated 11/15/85 and ULNRC-1247, dated 1/28/86. In conjunction with the change to VANTAGE 5 fuel for Cycle 3, certain of the accident analyses were redone. These analyses were also based on the uprated power condition and are documented in ULNRC-1470, dated March 31, 1987. The analyses that were not redone are those that are still bounded by the OFA analyses performed for Cycle 2. The results of the accident analyses demonstrate that all safety limits are met at the uprated power conditions.

Radiological dose consequences pertaining to the uprated power conditions have been completed and documented as part of the license amendment in support of Callaway Cycle 3 for the utilization of VANTAGE 5 fuel, which are documented in ULNRC-1470 dated March 31, 1987. The evaluation is also applicable to this amendment request. The environmental impact of operating Callaway Plant at the uprated condition was previously evaluated in the Callaway Environmental Report as a bounding assumption.

This has been approved by NRC in NUREG-75/011, Section 1.1 dated 3/75.

a) This change does not involve a significant increase in the probability or consequences of an accident previously evaluated. This is based on the environmental evaluation being previously approved at the uprated conditions and the radiological evaluation which is presented in the Callaway Cycle 3 amendment request for the utilization of VANTAGE 5 fuel.

b) This change does not create the possibility of a new or different kind of accident from any accident previously evaluated. This is based on system and component reviews which verified their capability to operate at the uprated conditions.

c) This change does not involve a significant reduction in a margin of safety. Accident analyses were performed at the uprated conditions and demonstrate the DNB design basis remains unchanged, the RCS pressure limit of 2700 psig is not exceeded, and the LOCA results remain well below the regulatory limits given in 10CFR50.46.

Based on the above discussions, the amendment request does not involve a significant increase in the probability or

consequences of an accident previously evaluated; does not create the possibility of a new or different kind of accident from any 4

accident previously evaluated; does not involve a reduction in the required margin of safety. Based on the foregoing, the 4

requested amendment does not present a significant hazard.

e 4

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4

+

Attcchment 5, Appendix A-ULNRC-1471'-

March-31, 1987 4

6 l'

t 4

i t

t 1 e i CALLAWAY UPRATING SUBMITTAL i

l ATTACHMENT 5: APPENDICES i

i APPENDIX A f

NSSS UPRATING LICENSING REPORT t

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4 t

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f

. . - - _ . . . . ._ . . . _ - - - - . , ~ . . _ . - _ . . . . , _ _ -

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