ML20247M921

Amend 153 to License NPF-1,permitting Use of Upgraded Fuel Assemblies Which Incorporate Features of Westinghouse VANTAGE5 Fuel Assemblies & Allowing Extended Fuel Burnup to 60 Gwd/Mt & Higher Nuclear Peaking Factors

ML20247M921
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	05/24/1989
From:	Knighton G Office of Nuclear Reactor Regulation
To:
Shared Package
ML20247M917	List: ML20247M912 ML20247M921 ML20247M933
References
NUDOCS 8906050194
Download: ML20247M921 (23)

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/ ^ UNITED STATES

..- II n NUCLEAR REGULATORY COMMISSION

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PORTLAND GENERAL ELECTRIC COMPANY THE CITY OF EUGENE, OREGON PACIFIC POWER AND LIGHT COMPANY DOCKET NO. 50-344-TROJAN NUCLEAR PLANT AMFNDMENT TO FACILITY OPERATING LICENSE Amendment No.153 License No. NPF-1

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for~ amendment by Portland General' Electric Company, et al., (the licensee) dated November 20, 1987, as supplemented Wayl 7, and August 12, 1988, complies with the standards and-requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I; B. The. facility will. operate in conformity with the application,

.the provisions of the Act, and the regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendr.ent is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

l 8906050194 890524 PDR -ADOCK 05000344 P PDC _

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2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-1 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.153, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifica-tions, except where otherwise' stated in specific license ,

conditions.

3. This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

. George W Knight

// $4 Director Projec Directorate V -

Division of Reactor Projects III, IV, Y and Special Projects

Attachment:

Changes to the Technical Specifications Date of Issuance: May 24, 1989 m

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I ATTACHMENT TO LICENSE AMENDMENT NO.153 TO FACILITY OPERATING LICENSE NO. NPF-1 DOCKET NO. 50-344 1..

! Revise Appendix A as follows:

Remove Pages Insert Pages

>. 2-2 2-2 2-5 2-5 2 -7 2-7 2-8 2-8 2-9 2-9 B 2-2 B 2-2 B 2-6 B 2-6 3/4 2-5 3/4 2-5 3/4 2-6a 3/4 2-6a 3/4 2-7 -

3/4 2-7 3/4 2-8 3/4 2-8 3/4 2-9 3/4 2-9 3/4 2-9a 3/4 2-9a B3/4 2-1 B3/4 2-1 -

B3/4 2-4 B3/4 2-4 83/4 2-5 B3/4 2-5 B3/4 2-6 B3/4 2-6 B3/4 4-1 B3/4 4-1 B3/4 5-2 B3/4 5-2 5-4 5-4 4

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SAFETY LIMITS BASES-N The curves are besed on an enthalpy hot channel' factor, F l $, of 1 and a reference cosine with a peak of 1.55,for axial power . An shah allowance is included for an increase in F AH at reduced power based on the expression:

F"y g = 1.56 [1 + 0.3 (1-P)) l where P is the fraction of RATED THERMAL POWER

- These limiting heat flux-conditions are higher than those calculated for the 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(AI) function of the Overtemperature AT trip. When the axial power imbalance is not within the tolerance, the axial power imbalance ef fect on the Overtemperature AT trips will reduce the set-points to provide protection consistent with core safety limits.

2.1. 2 REACTOR COOLANT SYSTEM PRESSURE The restriction of tis Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.

'The reactor pressure vessel' and pressurizer are designed to Section III of the ASME Code for Nuclear Power Plant which permits a maximum transient pressure of 110% (2735 psig) of design pressure. The Reactor Coolant System piping.* valves and fittings, are designed to ANSI B 31.7-1969, which permits a maximum transient pressure of 120%

(2985 psig) of component design pressure. The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated code requirements.

The entire Reactor Coolant System is hydrotested at 3107 psig to demonstrate integrity prior to initial operation.

9 TROJAN-UNIT 1 8 2-2 AmendmentNo.48,75,jdlk 153

__ ______._____m_._mm_..__.__ _ - m.____ _-_ .

l 0 klMITING SAFETY SYSTEM SETTINGS BASES 9

. loop flow. This latter trip will prevent the minimum value of the DNBR from going below the safety analysis DNBR limit during normal operational transients and anticipated transients when 3 loops are in operation and the Overtemperature AT trip setpoint is adjusted to the value specified for all loops in operation. With the Overtemperature AT trip setpoint adjusted to the value specified for 3 loop operation, the P-8 trip at 75% RATED THERMAL POWER will prevent the minimum value of the DNBR from going below the safety analysis DNBR limit during normal opera-tional transients and anticipated. transients with 3 loops in operation.

- Steam Generator Water Level The Steam Generator Water Level Low-Low trip provides core protec-tion by preventing operation with the steam generator water level below the minimum volume required for adequate heat removal capacity. The specified setpoint provides allowance that there will be sufficient water inventory in the steam generators at the time of trip to allow for starting delays of the auxiliary feedwater system.

Steam /Feedwater Flow Mismatch and Low Steam Generator Water Level .

The Steam /Feedwater Flow Mismatch il coincidence with a Steam Generator Low Water Level trip is not used in the transient and accident analyses but is included in Table 2.2-1 to ensure the functional capa-bility of the specified trip settings and thereby enhance the overall reliability of the Reactor Protection System. This trip provides added protection to the Steam Generator Water Level Low-Low trip. The Steam /

Feedwater Flow Mismatch portion of this trip gs activated when the steam flow exceeds the feedwater. flow by >1.51 x 10 lbs/ hour. The Steam Generator Low Water level portion of the trip is activated when the water level drops below 25 percent, as indicated by the narrow range instrument.

These trip values include sufficient allowance in excess of norwal operating values to preclude spurious trips but will initiate a reactor trip before the steam generators are dry. Therefore, the required capacity and starting time requirements of the auxiliary feedwater pumps are reduced and the resulting thermal transient on the Reactor Coolant System and steam generators is minimized.

Undervoltace and Underfrecuency - Reactor Coolant PumD Busses The Undervoltage and Underfrequency Reactor Coolant Pump bus trips provide reactor core protection when above the P-7 interlock setpoint against DNB as a result of loss of voltage (nominally 12.47 kV) or TROJAN-UNIT 1 B 2-6 Amendment No. 48, 737,)h%[

153

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POWER DISTRIBUTION LIMITS HEAT FlVX HOT CHANNEL FACTOR-F g {71 LIMITING CONDITION FOR OPERATION 3.2.2 F (Z) shall be limited by the following relationships:

n

' Fg (Z) 5 [2.50) [K(Z)] for P > 0.5 P

Fn (Z) 5 [(5.00)) [K(2)] for P S 0.5 ere P = THERMAL POWER RA1LD THERMAL POWER and K(Z) is the function obtained from Figure 3.2-2 for a given core height location.

APPLICABILITY: MODE 1-ACTION: ,

With Fg (Z) exceeding its limit:

a. Reduce THERMAL POWER at least 1% for each 1% F )

the limit within 15 minutes and similarly reduke(Z' the Power exceeds Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; subsequent STARTUP and POWER OPERATION may proceed provided the Overpower AT Trip Setpoints have been reduced at least 1% for each 1% F (2) exceeds the limit. The Overpower ATTripSetpointreductio0shallbeperformedwiththe reactor subcritical. ,,

b. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; THERMAL POWER may then be increased provided F (Z) is demonstrated through incore mapping tobewithinitslim9t.

l TROJAN-UNIT 1 3/4 2-5 AmendmentNo.5p@$[153 0

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POWER DISTRIBUTION' LIMITS SURVEILLANCE REQUIREMENTS (Continued) b) At least once par 31 EFPD, whichever occurs first.

2. When the F C is less than or equal'to the pRTP limit for theappropNatemeasuredcoreplane,additiEXalpower distribution maps shall be taken'and F G compared to F P and Ff at least once per 31 EFPDE

e. Changes in the F limits for RATED THERMAL POWER (FRTP) shal,1beprovideFforallcoreplanescontainingbanE*D' control rods and all,unrodded core planes in a Radial Peaking Factor Limit Report per Specification 6.9.1.7.

f. The F limits of e,' above, are not applicable in the' fol-lowinPcore plane regions as measured in percent of core height from the bottom of the fuel:

1. Lower core region from 0 to 15%, inclusive.

2. Upper core region from 86 to 100% inclusive.

3. ' Grid plane regions at 17.8 2%,32.112%,'47.4 2%,

60.6 !2% and 74.9 1 2%, inclusive.

4. Core plane regions within i 2% of core height (1 2.88 inches) about the bank demand position of the bank 'D' rods.

g. Evaluating the effects of F on F (2) to determine if F n(Z) is within its limit wheneveFFj e ceeds Fj.

. 4.2.2.3 F When F is measure 0(,I) shall be measured at least once per 31 EFPD.anoverallmeasure r distribution map and increased by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement i

uncertainty.

i TROJAN-UNIT 1 3/4:2-6: Amendment No. 5, 38, ##, 79, 78 M 153

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3/4 2-7 Amendment No.h 153

4 POWER DISTRIBUTION LIMITS RCS FLOWRATE AND F p LIMITING CONDITION FOR OPERATION 3.2.3 The combination of indicated Reactor Coolant System (RCS) total flow rate and F, shall be maintained within the region of allowable operation (abovB and to the left of the line) shown on Figures 3.2-3 and 3.2-4 for 4- and 3-loop operation, respectively.

Where:

N F

AH

a. FR " 1.56 (1.0 + 0.3 (1.0 - P)), and THERMAL POWER .

b.

P = RATED THERMAL POWER APPLICABILITY: MODE 1 ACTION:

With the combination of RCS total flow rate and F ofacceptableoperationshownonFi'gure3.2-3or$.outsidetheregion 2-4 (as applicable):

a. Within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:

1. Either restore the combination of RCS flow rate and F R -

to within the above limits, or l

2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High trip setpoint to ,y 55% of RATED THERMAL 90WER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, t

b. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of initially being outside the above limits, f verify through incore flux mapping and RCS total flow rate l comparison that the combination of Fratearerestoredtowithintheabov$andRCSto limits, or reduce THERMAL POWER to less than 55 of RATED THERMAL POWER within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

TROJAN-UNIT 1 3/4 2-8 Amendment No. 30, ##, dB, 76 M 153 ,

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l 4

POWER DISTRIBUTION LIMITS ACTION: (Continued)

c. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced THERMAL I POWER limit required by ACTION items a.2 and/or b above; subsequent POWER OPERATION may proceed provided that the combination of R and indicated RCS total flow rate are demonstrated, through incore flux mapping and RCS total flow )

rate comparison, to be within the region of acceptable opera- I tion shown on Figure 3.2-3 or 3.2-4 (as applicable) prior to i exceeding the following THERMAL POWER levels:

1. A nominal 50% of RATED THERMAL POWER,

2. A nominal 75% of RATED THERMAL POWER, and

3. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of attaining >_ 95% of RATED THERMAL POWER.

SURVEILLANCE RE001REMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable.

4.2.3.2 The combination of indicated RCS total flow rate and F be determined to be within the region of acceptable operation f o, shall Figure 3.2-3 and 3.2-4 (as applicable):

a. Prior to operation above 75% of RATED THERMAL POWER af ter each fuel loading, and ,

b. At least once per 31 Effective Full Power Days.

l Where:

F"6H FR " 1.56 (1.0 + 0.3 (1.0 - P) } , and N obtained by using the movable l F{H=MeasuredvaluesofFincore detectors to $$tain a power distribution map The measured values of FN shall be used to calculate F sinceFigures3.2-3aAH3.2-4inciudemeasurement c$1culationaluncertaintiesof3.5%forflowand 4% for incore measurement of F{H" 4.2.3.3 The RCS total flow rate indicators shall be subjected to a CHANNEL CALIBRATION at least once per 18 months.

4.2.3.4 The RCS total flow rate shall be determined by measurement at least once per 18 months.

TRO.1AN-UNIT 1 3/4 2-9 Amendment No. #8, 76, 153 l l

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TROJAN. UNIT 1 3/4 2 9a . Amendment No./A, Ts,g 153

. l 3/4.7 POWER DISTRIBUTION LIMITS BASES I

The specifications of this section provide assurance of fuel integ-rity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by: (a) maintaining the calculated DNBR in the core at or above design during nomal operation and in short term transients, and (b) limiting the fission gas release, fuel pellet temperature &

cladding mechanical properties to within assumed design criteria. j In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200*F is not exceeded.

The definitions of certain hot channel and peaking factors as used in these specifications are as follows:

F0 (2) Heat Flux Hot Channel Factor, is defined as the maximum local heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manu-facturing tolerances on fuel pellets and rods.

N F Nuclear Enthalpy Rise Hot Channel Facter, is defined as the aH ratio of the integral of linear power along the rod with the highest integrated power to the average rod power.

F,y(7) Radial Peaking Factor, is defined as the ratio of peak power density to average power density in the horizontal plane at core elevation 7. -

3/4.2.1 AXIAL FLUX DIFFERENCE (AFD)

The limits on AXIAL FLUX DIFFERENCE assure that the F boundenvelopeof2.50timesthenormalizedaxialpea' king ctor is 9a(2) not upper' exceeded during either normal operation or in the event of xenon redis-tribution following power changes.

Target flux dif ference is detemined at equilibrium xenon conditions.

The full length rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal position for steady state operation at high power levels. The value of the target flux difference obtained under these conditions divided by the f raction of RATED THERMAL POWER is the target flux dif fer-ence at RATED THERMAL POWER for the associated core burnup conditions.

Target flux differences for other THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value by the appropriate fractional THERMAL POWER level. The periodic updating of the target flux difference value is necessary to reflect core burnup considerations.

TRO.1AN-UNIT 1 B 3/4 2-1 Amendment No. 30, #8, 70 M 153

.,.A POWER DISTRIBUTION LIMITS BASES 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR. RCS FLOWRATE. AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR The limits on heat flux hot channel factor, RCS flowrate, and nuclear enthalpy rise hot channel. factor ensure that 1) the design limits on peak local power density and minimum DNBR are not exceeded and 2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200'F ECCS acceptance criteria limit.

Each of these is measurable but will normally cnly be detemined periodically as specified in Specifications 4.2.2 and 4.2.3. This periodic surveillance is suf ficient to insure that the limits are maintained provided:

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

b. Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.5.

c. The control rod insertion limits of Specification 3.1.3.5 '

are maintained.

l o. The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.

FN will be maintained within its limits provided conditions a.

throuh d. above are maintained. As noted on Figures 3.2-3 and 3.2-4, RCS flow rate and FN may be ' traded off' against one another (ie, a low measuredRCSf180rateisacceptableifthemeasuredF N is also low)

- to ensure that the calculated DNBR will not be below tN design DNBR value. This tradeoff is allowed up to a maximum FN of 1.56 (1+0.3(1-P))

whichisconsistentwiththeinitialconditionsasNmedfortheLOCAanalysis.l The relaxation of FN as a function of THERMAL POWER allows changes in the radial power shape Nr all permissible rod insertion limits.

When an gF measurement is taken, both experimental error and manuf actur-ing tolerance must be allowed for. 5% is the appropriate allowance for a full core map taken with the incore detector flux mapping system and 3% is the appropriate allowance for manufacturing tolerance. Application of these two penalties in a multiplication fashion is sufficient to provide a correction for the of fact of rod bow on F which has been conservatively estimated as5%inWCAP-8692,"FuelRodho, wing". The appropriate statistical combina-i tion of local power, manufacturing tolerance and rod bow uncertainties, results l in a penalty on F of 7.68%, whereas multiplying measured values of Fg by

! 1.03 x 1.05 resulks in a penalty of 8.15%.

TRO.1AN-UNIT 1 8 3/4 2-4 Amendment No. 30, 48, 79, 78 , M 15

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POWER DISTRIBUTION LIMITS BASES When RCS flowrate and F5H are measured, no additional allowances are i necessary prior to comparison with the limits of Figure 3.2-3 agd 3.2-4.

Measurement errors of 3.5% for RCS tota'l flow rate and 4% for FaH have been allowed for in determination of the design DNBR value.

The safety analysis DN8R values include a 14.4% margin for conservatism.

The effect of rod bow on DNBR has been determined to be a <1.5% penalty.

The available maggin more than offsets the effect of rod bow and no penalty is required on FAH or DNBR.

3/4.2.4 OUADRANT POWER TILT RATIO The quadrant power tilt ratio limit assures that the radial power distribution 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 limiting tilt of 1.025 can be tolerated before the margin for uncertainty in Fg is depleted. The limit of 1.02 was selected to provide an allowance for the uncertainty associated with the indicated power tilt.

. The 2-hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identificati6n and correc-tion of a dropped or misaligned rod. In the event such action does not correct the tilt, the margin for uncertainty on Fg is reinstated by reducing the power by 3 percent for each percent of tilt.in excess of 1.0.

TROJAN-UNIT 1 B 3/4 2-5 Amendment No. 30,##,48,79#,dht"153

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1 POWER DISTRIBUTION LIMITS  !

BASES 3/4.2.5 DNB PARAMETERS The limits on the DNB related parameters assure that each of the parameters are maintained within the normal steady state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR of greater than or equal to the safety analysis DNBR limit throughout each analyzed transient, The 12-hour periodic surveillance of these parameters thru instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation. The 18-month periodic measurement of the RCS total flow rate is adequate to detect flow degradation a'nd ensure correlation of the flow indication channels with measured flow such that the indicated percent flow will provide sufficient verification of flow rate on a 12-hour basis.

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TROJAN-UNIT 1 B 3/4 2-6 AmendmentNo.30,#8,jNy153

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3/4.4' REACTOR COOLANT SYSTEM BASES i

3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with all reactor coolant loops in operation, and maintain DNBR above the safety analysis DNBR limit during all nonnal operations and anticipated transients. Witn one reactor l coolant loop not in operation, THERMAL POWER is restricted to <38 percent-of RATED THERMAL POWER until the Overtemperature AT trip is reset..

Either action ensures that the DNBR will be maintained above the safety

. analysis DNBR limit. A loss of flow in two loops willcause a reactor trip if _ operating above P-7 (10 percent of RATED THERMALPOWER) while a loss of flow in one loop will cause a reactor trip if operating above P-8

.~ (39 percent of RATED THERMAL POWER).

In MODE 3, a single reactor coolant loop provides sufficient heat removal capability for removing decay heat; however, single failure con-siderations require that two loops be OPERABLE. Four loops in operation while control rod drive mechanisms are energized ensures that the DNB design basis can be met for a bank withdrawal from subcritical or low power accident.

In MODES 4 and 5, a single reactor coolant loop or RHR loop pro-vides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops be OPERABLE.

( Thus, if the reactor coolant loops are not OPERABLE, this specification requires two RHR loops-to be OPERABLE.

The operation of one Reactor' Coolant Pump or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System. 'The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control.

The restrictions on starting a Reactor Coolant Pump with one or more RCS cold legs less than or equal to 290*F are provided to prevent RCS pressure transients, caused by energy additions f rom the secondary system, which could exceed the limits of Appendix G to 10 CFR Part 50.

The RCS will be protected against overpressure transients and will not l

l exceed the limits of Appendix G by either (1) estricting the water vol-l ume in the pressurizer and thereby providing 6 volume for the primary i . coolant to expand into, or (2) by restricting starting of the RCPs to l

when the secondary water temperature of each steam generator is less than 50*F above each of the RCS cold leg temperatures, or (3) by restricting .

Starting of a RCP unless another RCP is running.

TROJAN-UNIT 1 B 3/4 4-1 Amendment No. If, 55,78,722,%

153 i

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'EMEREENCY CORE COOLING SYSTEMS BASES 3/4.5.3.2 ECCS SUBSYSTEMS The limitation for a maximum of one safety injection pump to be OPERABLE and the Surveillance Requirement that the other safety injection pump be-inoperable below 290*F provides assurance that a mass addition pressure l transient can be relieved by the operation of a single PORV.

3/4.5.4 DELETED -

3/4.5.5 REFUELING WATER STORAGE TMK (RWST)

The OPERABILITY of the RWST as part of the ECCS ensures that suf ficient negative reactivity is injected into the core to counteract any positive increase in reactivity caused by RCS system cooldown. RCS cooldown can be caused by inadvertent depressurization, a loss-of-coolant accident, or a steam line rupture.

The OPERABILITY of the RWST as part of the ECCS ensures that a sufficient supply of borated water is available for injection by the ECCS in the event of a LOCA. The limits on RWST minimum volume and boron concentra-

. tion ensure that 1) suf ficient water is available within containment to permit recirculation cooling flow to the core, 2) the reactor will remain subcritical in the cold condition fo1 Towing mixing of the RWS1 and the RCS -

l.

water volumes with all control rods inserted except for the n'.ost reactive control assembly, and 3) the reactor will remain suberitical in the cold condition following a large break LOCA (break flow area ?3.0 ft 2 for a cold leg break or >1.0 f te for a hot leg' break) assuming complete mixing of the RWST, RCS, ECCS water and other sources of water that may eventually reside in the sump, post-LOCA with all control rods assumed to be out.

These assumptions are consistent with the LOCA analyses.

TROJAN-UNIT 1 8 3/4 5-2 AmendmentNo.g2 78,793,M

. ,./,. ..

l.

, DESIGN FEATURES DESIGN PRESSURE AND TEMPERATURE 5.2.2 The reactor containment building is designed and shall be saintained for a maximum internal pressure of 60 psig and a temperature of 288'F.

PENETRATIONS 5.2.3 Penetrations through the reactor containment building are designed and shall be maintained in accordance with the original design provisions contained in Section 6.2.4 of the FSAR with allowance for, normal degradation pursuant to the applicable Surveillance Requirements.

5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 193 fuel assemblies with each fuel assembly containing 264 fuel rods clad with Zircaloy-4, except that limited '

substitutions of fuel rods by filler rods consisting of Zircaloy-4 or stainless steel, or by vacancies, may be made as justified by cycle-specific reload safety analysis. Each fuel rod shall have a nominal active fuel length of 144 inches. The initial core loading shall have a maximum

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enrichnent of 3.15 weight percent U-235. Reload fuel shall be similar in physical design to the initial core loading and of low enrichment.

CONTROL ROD ASSEMBLIES 5.3.2 ' The rea' c tor core shall contain 53 full length control rod assemblies.

'The full length control rod assemblies shall contain a nominal 142 inches of absorber noterial. The nominal values of absorber material shall- be 80 pe'rcent silver 15 percent indium and 5 percent cadmium. All control "

rods shall be clad with stainless steel tubing. Eight part length control rod assemblies originally installed in the core contained a nominal 36 inches of absorber material at their lower ends. The part length control rod I assemblies have been removed and are stored in the spent fuel pool.

5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained:

TROJAN-UNIT 1 5-4 Amendment No. 79. 7T1, 757,2Gk 153 i

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