Text

Amend 70 to License NPF-1,deleting Requirements Re part-length Control Rod Assemblies,Allowing Removal from Reactor Core
	ML20049J007
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	03/03/1982
From:	Clark R; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20049J004	List: ML20049J003; ML20049J007; ML20049J009;
References
	NUDOCS 8203110163
	Download: ML20049J007 (19)
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PORTLAND GENERAL ELECTRIC COMPANY THE CITY OF EUGENE, OREGON PACIFIC POWER AND LIGHT COMPANY

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DOCKET.NO. 50-344 TROJAN NUCLEAR PLANT I

AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 70 i

License No. NPF-1 l

l 1.

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

A.

The application for amendment by Portland General Electric Company, the City of Eugene, Oregon, and Pacific Power and Light Company (the licensee) dated February 10, 1982 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and 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 rules and 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 r

E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

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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. 70, are hereby incorporated in the license. The licensee shall operate the facility in accordance with.the Technical

~ Specifications, except as noted in paragraph 2.C.(11)

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below.

3.

This license amendment is effective upon entry into Operational Mode 2 (Startup) for Fuel cycle 5.

i FOR THE NUCLEAR REGULATORY COMMISSION

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Jh.fM i as Robert A. Clark, Chief Operating Reactors Branch #3 Division of licensin'g

Attachment:

Changes to the Technical Specifications Date of Issuance: March 3, 1982

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ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO. 70 TO FACILITY OPERATING LICENSE NO. NPF-1 l

- DOCKET NO. 50-344 4

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Revi,se Appendix A as follows:

Remove Pages Insert Pages i

1-3 1-3 i.

3/4 1-18 3/4 1-18

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3/4 1-19

~3/4.1-19 3/4 1-20 3/4 1-20 3/4 1-26 3/4 1-25 4 i 3/4 2-3 3/42-3 1'

3/4 2-6a 3/4 2-6a j i 3/4 2-10 3/4 2-10 3/4 2-11 3/4 2-11 5

3/4 10-1 3/4 10-1 3/4 10-2 3/4 10-2 3/4 10-5 3/4 10-5 B 3/4 1-4 B 3/4 1-4 B 3/4 2-1 B 3/4 2-1 B 3/4 2-4 B 3/4 2-4 5-4 4 9

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DEFINITIONS i

CHANNEL FL'NCTIONAL TEST 1.11 A CHANNEL FUNCTIONAL TEST shall be the. injection of a simulated signal into the channel as close to the primary sensor as practicable to verify OPERABILITY including alarm and/or trip functions.

CORE ALTERATION 1.12 CORE ALTERATION shall be the movement or manipulation of any l

component within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of.C0RE ALTERATION shall not preclude l

completion of movement of a component to a safe conservative position.

SHUTDOWN MARGIN 1.13 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is or would be subcritical from its present condi-assuming all full lenth rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.

IDENTIFIED LEAKAGE 1.14 IDENTIFIED LEAKAGE shall be:

Leakage (except CONTROLLED LEAKAGE) into closed systems, such a.

as 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 sources that are b'oth -specifically located and known either not to. interfere with the operation of leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE, or Reactor coolant system leakage through a steam generator c.

to the secondary system.

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UNIDENTIFIED LEAKAGE 1.15 UNIDENTIFIED LEAKAGE shall be all leakage'which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE.

TROJAN - UNIT l 1-3 Amendment No. 7 0 4

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REACTIVITY CONTROL SYSTEMS 3/4.1.3 MOVABLE CONTROL ASSEMBLIES GROUP HEIGHT LIMITING CONDITION FOR OPERATION J

3.1. 3.1 All full length (shutdown and control) rods shall be OPERABLE

&nd positioned within + 12 steps (indicated position) of their group step counter demand position.

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APPLICABILITY: MODES 1* and 2*

i ACTION:

With one or,more full length rods inoperable due to being a.

immovable as a result of excessive friction or mechanical I

interference or known to be untrippable, determine that the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

With more than one full length rod inoperable or misaligned from any other rod in its group by more than + 12 steps (indicated position), be in HOT STANDBY within 6 h'ours.

With one full length rod inoperable or misaligned from c.

its group step counter demand height by more tian +'12 steps (indicated position), POWER OPERATION may continue provided that within one hour either:

1.

The rod is restored to OPERABLE status within the above alignment requirements, or 2.

The rod is declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied.

POWER OPERATION may then continue provided that:

a)

An analysis of the potential ejected rod worth is performed within 3 days and the rod worth is deter-mined to be < 0.98% ak at zero power and < 0.21%

ak at RATED THERMAL POWER for the remainder of the fuel cycle, and b)

The SHUTDOWN MARGIN requir,ement of Specif1:at.lon 3.1.1.1 is determined at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , and

See Special Test Exceptions 3.10.2 and 3.10.4.

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l l LIMITING CONDITION FOR OPERATION (Continued) c)

The THERMAL POWER level is reduced to < 75% of RATED THERMAL POWER within one hour and withTn the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the high neutron flux trip setpoint is reduced to < 85% of RATED THERMAL POWER,*6r 1

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The remainder of the rods in the group with the inoperable rod are aligned to within + 12 steps of the inoperable rod within-one hour while maintaining the rod sequence and insertion limits of Figures 3.1-1 and 3.1-2; the THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.5 during subsequent operation.

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SURVEILLANCE REQUIREMENTS I

j 4.1. 3.1.1 The position of each full length rod shall.be determined to be within the group demand limit by verifying the individual rod positions at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time ir.tervals when the Rod' Position Deviation Monitor is inoperable, then verify the group positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4.1. 3.1. 2 Each full length rod not fully inserted shall be determined to be OPERABLE by movement of at least 10 steps in any one direction at least once per 31 days.

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t REACTIVITY CONTROL SYSTEMS POSITION INDICATOR CHANNELS i

LIMITING CONDITION FOR OPERATION 3.1.3.2 Control rod position indicator channels for control and shutdown rods and the demand position indication system shall*be OPERABLE and j

capable of determining the control rod positions within + 12 steps.

i APPLICABILITY: MODES 1 and 2.

ACTION:

With a maximum of one rod position indicator channel per group in-a.

i operable either:

I 1.

Determine the position of the non-indicating rod (s) indirectly by the movable incore detectors at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and i

immediately after any motion of the non-indicating rod which i

exceeds 24 steps in one direction since the last determination l

of the rod's position, or 2.

Reduce THERMAL POWER TO < 50% of RATED THERMAL POWER within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

b.

With a maximum of one demand position indicator per bank inoperable either:

1.

Verify that all rod position indicators for the affected bank are OPERABLE and that the most withdrawn rod and the least withdrawn rod of the bank are within a maximum of 12 steps of each other at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , or l

2.

Reduce THERMAL POWER to < 50% of RATED THERMAL POWER within i

8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

SURVEILLANCE REQUIREMENTS 4.1. 3. 2 Each rod position indicator channel shall be determined to be OPERABLE by verifying the demand position indication system and the rod position indicator channels agree within 10, steps at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the Rod Position Deviation Monitor is inoperable, then compare the demand position indication system and the rod position indicator channels at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

TROJAN-UNIT 1 3/4 1-20 Amendment No. 7 0 l

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l ii PART LENGTH ROD INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.6 This Specification has been deleted due to tt}e ren. oval of part length rods from the reactor.

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SURVEILLANCE REQUIREMENTS 4.1.3.6 This Specification has been deleted due to the removal of part length rods from the reactor.

TROJAN-UNIT 1 3/4 1-26 Amendment No. 70 3_

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t POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) 4.2.1.3 The target flux difference of each OPERABLE excore channel shall be determined by measurement at least once per 92.. Effective' Full Power Days.

j 4.2.1.4 The target flux difference shall be updated at least once per 31 Effective Full Power Days by either determining the target flux l '

difference pursuant to 4.2.1.3 above or by linea'r interpolation between i.

the most recently measured value and 0 percent at the end of the cycle life.

TROJAN-UNIT l 3/4 2-3 Amendment No. 70.'

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POWERDISTRi'BUTIONLIMITS SURVEILLANCE REQUIREMENTS (Continued) b)

At least once per 31 EFPD, whichever occurs first.

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When the Ff is less than or equal to the FyTP 2.

limit for y

the appropriate measured core plane, additional power distribution maps shall be taken and Fj compared to F}TP and F y at least once per 31 EFPD.

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The F limits for RATED THERMAL POWER within specific core xy planes shall be:

1.

FRTP 11.71 for all core planes containing bank y

"D" control rods.

2.

F P 11.65 for unrodded core planes.

f.

The F limits of e, above, are not applicable in ~the fol-lowing core 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 85 to 100% inclusive.

3 Grid plane regions at 17.8 + 2%, 32.1 + 2%, 47. 4 + 2%,

60.6 + 2% and 74.9 + 2%, inclusive.

4.

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

g.-. Evaluating the effects of F on F (Z) to determine if F (Z) xy n

q iswithinitslimitwheneverFjexceedsF

4. 2. 2. 3 F (Z) shall be measured at least once per 31 EFPD. When F (Z) is 9

n measured, an overall measured value shall be obtained from a power distribution map and increased by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty.

TROJAN-UNIT 1 3/4 2-6a Amendment No. 6, 30, ff,7 0

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POWER DISTRIBUTION LIMITS QUADRANT POWER TILT RATIO LIMITING CONDITION FOR OPERATION 3.2.4 THE QUADRANT POWER TILT RATIO sh'all not exceed 1.02.

APPLICABILITY: MODE 1 AB0VE 50% OF RATED THERMAL POWER

ACTION:

a.

With the QUADRANT POWER TILT RATIO determined to exceed 1.02 but < 1.09:

1.

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

a)

Either reduce the QUADRANT POWER TILT RATIO to within its limit, or b)

Reduce THERMAL POWER at least 3% for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1.0 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 .

2.

Verify that the QUADRANT POWER" TILT RATIO is within its limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding the limit. or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the. Power Range Neutron Flux-High Trip setpoints to < 55% of RATED THERMAL POWER within the next 4 houri'.

3.

Identify and correct the cause of the out of limit con-dition prior to increasing THERMAL POWER; subsequent POWER OPERATION abcVe 50% of RATED THERMAL power may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour until verified acceptable at 95% or greater RATED THERMAL POWER.

j b.

With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to misalignment of either a control or shutdown rod:

1.

Reduce THERMAL POWER at least 3% for each 1% of indi-cated QUADRANT POWER TILT RATIO in excess of 1.0, within l

30 minutes.

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See Special Test Exception 3.10'.2.

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f LIMITING CONDITION'FOR OPERATION (Continued) 2.

Verify that the QUADRANT POWER TILT RATIO is within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High trip Setpoints to 155% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

l 3.

Identify and correct the cause of the out of limit con-dition prior to increasing THERMAL POWER; subsequent i

l POWER OPERATION above 50% cf RATED THERMAL POWER may l

proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour until verified acceptable at 95% or greater RATED THERMAL POWER.

c.

With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to causes other than the misalignment of either a control or' shutdown rod:

1.

Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Trip Setpoints to 155% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

2.

Identify and correct the cause of the 'out of limit con-dition prior to increasing THERMAL POWER:

subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour until verified at 95% or greater RATED THERMAL POWER.

SURVEILLANCE REQUIREMENTS 4.2.4 The QUADRANT POWER TILT RATIO shall.be determined to be within the limit above 50% of RATED THERMAL POWER by:

Calcu'lating the ratio at least once per 7 days when the alarm a.

is OPERABLE.

b.

Calculating the ratio at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during steady state operation when the alarm is inoperable.

c.

Using the movable incore detectors to determine the QUADRANT POWER TILT RATIO at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when one Power Range Channel is inoperable & THERMAL POWER is > 75 percent of RATED THERMAL POWER.

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3/4.10 SPECIAL TEST EXCEPTIONS SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.10.1 The SHUTDOWN MARGIN reouirement of Specification 3.1.1.1 may be suspended for measurement of c.ontrol rod worth and shutdown margin provided:

Reactivity equivalent to at least the highest estimated control a.

rod worth is available for trip insertion from OPERABLE control rod (s).

APPLICABILITY: MODE 2.

ACTION:

a.

With the reactor critical (Keff > 1.0) and with less than the above reactivity equivalent available Tor trip insertion, immediately

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initiate and continue boration at > 30 gpm of 7000 ppm boron or. its equivalent until the SHUTDOWN MARGTN required by Specification 3.1.1.1 is restored.

b.m. With the reactor suberitical (Keff < 1.0) by less than the above reactivity equivalent, immediately initiate and continue boration at

> 30 gpm of 7000 ppm boron or its equivalent until the SHUTDOWN RARGIN required by Sp cification 3.1.1.1 is restered.

SURVEILLANCE REQUIREMENTS 4.10.1.1 The positior of each full length rod either partially or fully withdrawn shall be determined at least once per 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

4.10.1.2 Each full length rod not fully inserted shall be demonstrated OPERABLE by verifying its rod drop time to be < 2.2 seconds within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to reducing the SHUTDOWN MARGIN to less than the -limits of Specification 3.1.1.1.

4.10.1.3 This Specification has been deleted due to the removal of part length rods from the reactor.

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GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION 3.10.2 The group height, insertion and power distribution limits'of Specifications 3.1. 3.1, 3.1. 3. 4, 3.1. 3.5, 3.2.1 and 3.' 2. 4 may be l

suspended during the performance of PHYSICS TESTS provided:

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a.

The THERMAL POWER is maintained < 85%# of RATED THERMAL POWER,-

and b.

The limits of Specifications 3.2.2 and 3.2.3 are maintained and determined at the frequencies specified in Specification 4.10.2.2 below.

APPLIC, ABILITY.: MODE 1 ACTION:

With any of the limits of Specifications 3.2.2 or 3.2.3 being exceeded while the requirements of Specifications 3.1. 3.1, 3.1. 3. 4, 3.1. 3.5,.

3.2.1 and 3.2.4 are suspended, either:

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a.

-Reduce THERMAL POWER sufficient to satisfy the ACTION.

requirements of Specif1 cations 3.2.2 and 3.2 3, or b.

Be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.10.2.1 The THERMAL POWER shall be determined to be < 85% of RATED THERMAL POWER at least once per hour during PHYSICS TESTS.-

4.10.2.2. The Surveillance Requirements of Specifications 4.2.2 and 4.2.3 shall be performed at the following frequencies during '

,J PHYSICS TESTS:

a.

Specification 4.2.2 - At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

F b.

Specification 4.2.3 - At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

A THERMAL POWER'l.imit of 90% of RATED THERMAL POWER is permissible during tests performed as part of the Augmented Startup Test Program.'

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4 3.10.4 The limitatio'ns of Specifications 3.1.1. 4, 3.1. 3.1, 3.1. 3. 4 a nd

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.';o 3.1".3.5 may b'e suspended during the performance of PfiYSICS TESTS provioed:

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'The THERMAL POWER does not exceed 5% of RATED THERMAL POWER,

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b.,The reactor trip setpoints on the OPERABLE Intermediate and d

Power Range Nuclear Channels are set at < 25%~ of RATED s

THERMAL POWER.

, APPLICABILITY: MODE 2.

ACTION:

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With the THERMAL POWER > 5% of RATED THERMAL POWER, immediately open the reactor trip breakers.

.s SURVEILLANCE REQUIREMENTS

^

4.10.4.1 The THERMAL POWER shall be determined to be < 5% of RATED THERMAL POWER at least once per hour during PHYSICS TESTS.

4.10.4.2 Each Intermediate and Power Range Channel shall be subjected to a CHANNEL FUNCTIONAL TEST within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months prior to initiating PHYSICS TESTS.

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REACTIVITY CONTROL SYSTEMS t

BASES 3/4.1. 3 MOYABLE CONTROL ASSEMBLIES (Continued)

Control rod positions and OPERABILITY of the rocl., position indicators are requir.ed to be ' verified on a nominal basis of once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months with more frequent verifications required if an automatic monitoring channel is inoperable. These verification frequencies are adequate for assuring that the applicable LC0's are satisfied.

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TROJAN-UNIT 1 B 3/41-4 Amendment No. 58, 7 0

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3/4.2 POWER DISTRIBUTION LIMITS I,

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, and (b) limiting the fission gas release, fuel pellet temperature &

1 cladding mechanical properties to within assumed design criteria.

Ir. addition, limiting the peak linear power density during Condition I t

events provides assurance that the initial conditions assumed for the i

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:

,F(Z)

Heat Flux Hot Channel Factor, is defined as the maximum local g

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.

F)H Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of linear power along t'he rod with the' highest integrated power to the average rod power.

xY(Z)

Radial Peaking Factor, is def'ined as the rati.o of peak F

power density to average oower density in the horizontal plane at core elevation Z.

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3/4.2.1 AXIAL FLUX DIFFERENCE (AFD)

The limits on AXIAL FLUX DIFFERENCE assure that the F (Z) upper n

bound envelope of 2.32 times the normalized axial peaking ' factor is not exceeded durinp either normal operation or in the event of -xenon redis-tribution following power changes.

Target flux difference 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 inse.-ted 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 fraction of RATED THERMAL POWER is the target flux differ-ence at RATED THERMAL POWER for the associated core burnup conditions.

Target flux differences for other THERMAL POWER levels are obtained l-by multiplying the RATED THERMAL POWER value.by the appropriate fractional l

THERMAL POWER level.

The periodic updating of the target flux 'differe*nce value is necessary to reflect core. burnup considerations.

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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 dbsign limits on j

peak local power density and minimum DNBR are not exceeded and 2) in the 2

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 sufficient to insure that the limits are maintained provided:

I a.- Control rods in a single group move together with no individual j

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.

3 The control rod insertion limits of Specification 3.1. 3.5 c.

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are maintained.

i d.

The axial pcwer distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limi.ts.

t FfHwillbemaintainedwithinitslimitsprovidedconditionsa...

through d. above are maintained. As noted on Figures 3.2-3 and 3.2-4, RCS flow rate and Fyg may be " traded off" against one another (i.e., a low measured RCS flow rate is acceptable if the measured F{g is also low) to ensure that the calculated DNBR will not be below tne design DNBR value.

This tradeoff is allowed up to a maximum F g of 1.49 (1+0.2(1-P))

whichisconsistentwiththeinitialconditionsas]sumedfo'rtheLOCAanalysis.

N The relaxation of F4g as a function of THERMAL POWER allows changes in the radial. power shape Yor all permissible rod insertion limits.

When an Fg measurement is taken, both experimental ' error and manufactur-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 effect of rod bow on Fg, which has been conservatively estimated as 5% in WCAP-8692, " Fuel Rod Bowing". The appropriate statistical combina-tion of local power, manufacturing tolerance and rod bow uncertainties, results in a penalty on Fg of 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, H 7 0 l

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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 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. Each fuel rod shall have a nominal active fuel length of 144 inches and contain a maximum total weight of 1760 grams uranium. 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 enrichment of 3.5 weight percent U-235.

CONTROL ROD ASSEMBLIES 5.3.2 The reactor core shall contain 53 full length control rod assemblies.

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:

TROJAN-UNIT 1 5-4 Amendment No. 7 0 l

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