Text

Proposed Tech Specs Including Effects of Nuclear Fuel Design Changes
	ML20236W761
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	11/20/1987
From:	; PORTLAND GENERAL ELECTRIC CO.
To:	;
Shared Package
ML20236W729	List: ML20236W727; ML20236W739; ML20236W761; ML20236W768;
References
	NUDOCS 8712080215
	Download: ML20236W761 (20)
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-_-___-_ .______ _ ____ _ _ _ _ _

LCA 161 Attachment Page 1 of 20 680 . _ _ .

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540 0 .2 .4 .6 .8 1.0 1.2 l FRACTION OF RATED THERMAL POWER FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION l TROJAN UNIT 1 8712080215 871120 PDR ADOCK 05000344 P PDR TROJAN-UNIT 1 2-2 Amendment No. 48, 76 i

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LCA 161 j Attachment 4 Page 6 of 20 j l SAFETY LIMITS BASES The curves are based on an enthalpy hot channel factor, F N and a reference cosine with a peak of 1.55 g for axial power shah,.An of 1.56 allowance is included for an increase in F aH at reduced power based on the expression-F"H a = 1.56 D + 0.3 N )]

where P is the fraction of RATED THERMAL POWER l 1 These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the !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(al) 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 this Safety Limit protects the integrity of the Reactor Coolant System f rom 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.

i TROJAN-UNIT 1 B 2-2 Amendment No. 48, 76 L ___-_ - _ _ _

l LCA 161 i Attachment i Page 7 of 20 l I

LIMITING SAFETY SYSTEM SETTINGS BASES 1 loop flow. This latter trip will prevent the minimum value of the DNBR from going below the safety analysis ONBR 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 w'.ll 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 f the minimum volume required for adequate heat removal capacity. The j specified setpoint provides allowance that there will be sufficient water ;

inventory in the steam generators at the time of trip to allow for i starting delays of the auxiliary feedwater system.

Steam /Feedwater Flow Mismatch and Low Steam Generator Water Level The Steam /Feedwater Flow Mismatch in 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 normal operating values tu preclude spurious trips but will iritiate 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.

Undervoltage and Underfrecuency - Reactor Coolant Pump 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 ?-6 Amendment No. 48, 737 I

u-.----. --

LCA 161 i Attachment..

Page 8 of 20 POWER DISTRIBUTION LIMITS HEAT FLUX' HOT CHANNEL FACTOR-Fg[2_)

LIMITING CONDITION FOR OPERATION

~

3.2.2- Fg (Z) shall be limited by the following relationships:

Fn (Z) 5 [2.50] [K(Z)] for P > 0.5 :l P

Fn (Z) $ [(5.00)] [K(Z)] for P 5 0.5 where P = THERMAL POWER RATED 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% nF (Z) exceeds the limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; POWER OPERATION may proceed for up to a total.of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; 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 AT Trip Setpoint reductick shall be performed with the reactor subcritical. l

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

1 l

l TROJAN-UNIT 1 3/4 2-5 Amendment No. 6

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i L - -- 1

.p LCA 161 Attachment Page 9 of 20' POWER DISTRI'BUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) b) . At least once per 31 EFPD, whichever occurs first. J

2. When the F C is less than or equal to the F RTP limit for theappropNatemeasuredcoreplane,additiUXalpower-distribution maps shall be taken and F 0 compared to FRTP and'F at least once per 31 EFPDN x

e. Changes in the F limits for RATED THERMAL POWER (FRTP) shall be provide 8YforallcoreplanescontainingbanE"D" control rods and all unrodded core planes in a Radial Peaking Factor Limit Report per Specification 6.9.1.7.

f. limits of e, above, are not applicable in the fol-TheFlYcore lowin 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 f rom 86 to 100% inclusive.

3. Grid plane regions at 17.8 1 2%, 32.1 1 2%, 47.4 1 2%, j 60.6 12% and 74.9 1 2%, inclusive.

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

g. Evaluating the effects of F on F n

iswithinitslimitwheneveFFjeRc(2)todetermineifF(Z) x eedsFj.

4.2.2.3 F (2) sh611 be measured at least once per 31 EFPD. When Fn (2)-

.is measure 0, an overall measured value shall be obtained from a power distribution map and increased by 3% to account f or manuf acturing tolerances and further increased by 5% to account for measurement uncertainty.

I 1ROJAN-UNIT 1 3/4 2-6a Amendment No. 5, 30, ##, 70, 75 L___ ___

LCA 161 Attachment Page 10 of 20 k

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LCA 161 Attachment Page,11 of 20 POWER DISTRIBUTION LIMITS.

RCS FLOWRATE AND'F p LIMITING CONDITION FOR OPERATION

-1 3.2.3 'The combination of indicated Reactor Coolant System (RCS) total flow rate-and Fofshall be maintained within the region of allowable operation (abovD 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

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

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P = RATED THERMAL POWER APPLICABILITY: MODE 1 ACTION:

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With the combination of RCS total flow rate and F, outside the region of acceptable operation'shown on Figure 3.2-3 or 3.2-4 (as applicable):

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

1. Either restore the combination of RCS flow rate and F R 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 < 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

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

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TROJAN-UNIT 1 3/4 2-8 Amendment No. 30, ##, #B, 75

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! l LCA 161 )

Attachment .

j Page 12 of 20 l POWER DISTRIBUTION LIMITS ACTION: (Continued) J

c. Identify and correct the cause of the out-of-limit condition l prior to increasing THERMAL POWER above the reduced THERMAL 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-tion shown on Figure 3.2-3 or 3.2-4 (as applicable) prior to exceeding the following THERMAL POWER levels:

1. A nomitial 50% of RATED THERMAL POWER,

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

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

SURVEILLANCE REQUIREMENTS 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 g shall be determined to be within the region of acceptable operation et Figure 3.2-3 and 3.2-4 (as applicable):

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

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

Where:

N F

AH F

p = 1.56 {1.0 + 0.3 (1.0 - P)}, and N obtained by using the movable F{H=MeasuredvaluesofFincore uctectors to 8$tain a power distribution ma The measured values of FN F sinceFigures3.2-3ab$shallbeusedtocalculate 3.2-4 include measurement c01culationaluncertaintiesof3.5%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.

l TROJAN-UNIT 1 3/42-9 Amendment No. 48, 76

LCA 161 Attachment Page 13 of 20 al i .

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TROJAN-UNIT 1 3/4 2-9a Amendment No./p, 76

l i

LCA 161 Attachment Page 14 of 20 3/4.2 POWER DISTRIBUTION LIMITS ;

BASES 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 normal operation and in short term transients, -1 and (b) limiting the fission gas release, fuel pellet temperature &

cladding mechanical properties to Mthin assumed design criteria. j In addition, limiting the peak linear power density during Condition I i events pruides 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:

Fn(Z) 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-f acturing tolerances on fuel pellets and rods.

Nuclear Enthalpy Rise Hot Channel Factor, is defined as the F"H a

ratio of the integral of linear power alcng the rod with the i highest integrated power to the average rod power.

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

3/4.2.1 AXIAL FLUX DIFFERENCE (AFD)

The limits on AXIAL FLUX DIFFERENCE assure that the Fg (7) upper bound envelope of 2.50 times the normalized axial peaking ractor is not I

exceeded during either normal operation or in the event of xenon redis-tribution following power changes.

Target flux dif ference is determined 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 dif ference 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 dif ference value is necessary to reflect core burnup considerations.

TROJAN-UNIT 1 B 3/4 2-1 Amendment No. 30, 48, 70 l L __. . - _ .

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LCA 161 Attachment Page 15 of 20 ]

l POWER DISTRIBUTION LIMITS i l

BASES 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR. RCS FLOWRATE. AND f NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR The limits on heat flux hot channel factor, RCS flowrate, and nuclear j enthalpy rise hot channel factor ensure that 1) the design limits on j peak local power density and minimum DNBR are not exceeded and 2) in the i 1

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 only be determined 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 dif fering by more than i 12 steps f rom 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.

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

FN will be maintained within its limits provided conditions a. 1 throuhNd.abovearemaintained. 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 i l

measuredRCSflbbrateisacceptableifthemeasuredFN is also low) toensurethatthecalculatedDNBRwillnotbebelowth$designDNBR value. This tradeof f 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 f$r all permissible rod insertion limits..

When an F measurement is taken, both experimental error and manufactur-g ing tolerance must be allowed for. 5% is the appropriate allowance for a full f core map taken with the incore detector flux mapping system and 3% is the l appropriate allowance for manufacturing tolerance. Application of these two penalties in a multiplication fashion is sufficient to provide a correction for the effect of rod bow on F which has been conservatively estimated as5%inWCAP-8692,"FuelRodbo, wing". The appropriate statistical combina-tion of local power, manuf acturing tolerance and rod bow uncertainties, results l in a penalty on F gof 7.68%, whereas multiplying measured values of Fg by 1.03 x 1.05 results in a penalty of 8.15%.

TROJAN-UNIT 1 B 3/4 2-4 Amendment No. 30, #8, 70, 76 1

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a LCA 161 Attachment Page 16 of 20 I

POWER DISTRIBUTION LIMITS BASES When RCS flowrate and F$H are measured, no additional allowances are .

j necessary prior to comparison with the _ limits of Figure 3.2-3 agd 3.2-4. '

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

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

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

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

3/4.2.4 OVADRANT POWER TILT RATIO The quadrant power tilt ratio limit assures that the radial power distribution satisfies the design values used in the .oower capability.

analysis. Radial power distribution measurements are made during startup i testing and periodically during power operation. 1 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 F0 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 i than 1.02 but less-than 1.09 is provided to allow identification and correc- i tion of a dropped or misaligned rod. In the event such action does not l '

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.

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TROJAN-UNIT 1 B 3/4 2-5 Amendment No. 30,##,#5,704 f 1

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LCA 161 Attachment L

Page 17 of 20 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 and 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.

9 TROJAN-UNIT 1 B 3/4 2-6 Amendment No. 30, #8

LCA 161 Attachment Page 18 of 20 3/4.4 REACTOR COOLANT SYSTEM BASES ,

I 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION

'l The plant is' designed to operate with all reactor coolant loops in ]

operation, and maintain DNBR above the safety analysis DNBR limit during l all ~ normal operations. and. anticipated transients. - With one reactor 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 tnat 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 i 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 f ailure cortsiderations 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 j 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 from 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 exceed the limits of Appendix G by either (1) restricting the water vol-ume in the pressurizer and thereby providing a volume for the primary coolant to expand into, or (2) by restricting starting of the RCPs to 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.

l TROJAN-UNIT 1 B 3/4 4-1 Amendment No. 5#, 56, 78, 722

f LCA 161 Attachment Page 19 of 20 EMERGENCY CORE COOLING SYSTEMS BASES 3/4.5.3.2 ECCS SUBSYSTEMS The limitation for a maximum of one safety injection pump to be OP'ERABLE and the Surveillance Requirement that the other safety injection pump be inoperable below 290*F provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV.

3/4.5.4 DELETED 3/4.5.5 REFUELING WATER STORAGE TANK (RWST)

The OPERABILITY of the RWST as part of the ECCS ensures that sufficient 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 l subcritical in the cold condition following mixing of the RWS1 and the RCS water' volumes with all control rods inserted except for the most reactive control assembly, and 3) the reactor will remain subcritical in the cold condition'following a large break LOCA (break flow area >3.0 ft2 for a cold leg break or >1.0 ft2 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 Amendment No. 32, 78, 103

LCA 161 Attachment Page 20 of 20 DESIGN FEATURES DESIGN PRESSURE AND TEMPERATURE 5.2.2 The reactor containment building is designed and shall be maintained 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 amintained 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 enrichment of 3.15 weight percent U-235. Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrich-ment of 3.5 weight percent U-235.

CONTROL ROD ASSEMBLIES

5.3.2 The reactor core shall contain 53 full length control rod assemblies. l The full length control rod assemblies shall contain a nominal 142 inches of '

absorber material. The nominal values of absorber material shall be 80 percent 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 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:

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TROJAN-UNIT 1 5-4 Amendment No. 70, 776 I

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