Text

Amend 46 to License NPF-1,approving Use of Westinghouse Improved Thermal Design Procedure & WRB-1 Critical Heat Flux Correlation
	ML19330C206
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	07/25/1980
From:	Clark R; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML19330C137	List: ML19330C136; ML19330C138; ML19330C140; ML19330C206;
References
	NUDOCS 8008080069
	Download: ML19330C206 (42)
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UN[TED STATES '

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NUCLEAR REGULATORY COMMISSION t

j WASHINGTON. D. C. 20655

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PORTLAND GENERAL ELECTRIC COMPANY THE CITY OF EUGENE, OREGON C-PACIFIC POWER AND LIGHT COMPANY DOCKET NO. 50-344 TROJAN NUCLEAR PLANT AMENDMENT TO FACILITY OPERATING LICENSt:

Amendment No.46 License No. NPF-1 1.

The Nuclear Regulatory Comission (the Comission) 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 licensees) dated January 5,1979, as supplemented February 22'and November 5,1979, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chaoter I; B.

The facility will operate in conformity with the application, the pro-visions of the Act, and the rules and regulations of the Comission; 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 Comission's regulations; D.

The issuance of.this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and 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 satis-fied.

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:

goosoneo6q

_ _.. ~.

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 46, are hereby incorporated in the license..The licensee shall operate the facility in accordance with the. Technical Specifications, except as noted in paragraphs 2.C.(10)-and 2.c.(11) below.

3.

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

FOR THE NUCLEAR REGULATORY COMISSION Robert A. Clark, Chief

~

Operating Reactors Branch #3 Division of Licensing

Attachment:

Changes to the Technical-Specifications Date of Issuance: July 25, 1980-

+

v

a ATTACHMENT TO LICENSE AMENDMENT NO. 46 FACILITY ODERATING LICENSE NO. NPF-1 DOCKET NO. 50-344 Replace the following oages of the Apoendix "A" Technical Specifications with the enclosed paaes. The revised paaes are identified by Amendment The number and contain vertical lines indicating the areas of change.

correspondina overleaf pages are also provided to maintain document comoleteness.

Pace IV IX 2-2 2-3 2-5 2-7 2-8 2-9 B 2-1 B 2-2 3 2 B 2-6

-3/4 2-8 3/4 2-9 3/4 2-9a 3/4 2-9b 3/4 2-12 3/4 2-13 3/4 3-8 B 3/4 2-1 B 3/4.2-4 B 3/4 2-5 B 3/4 2-6 B 3/4 4-1 1

TROJAN - UNIT 1 i

n

INDEX LIMITING CONDITIONS FOR OPERAT' ION AND SURVEILLANCE REQUIREMENTS SECTION Page 3/4.0 APPLICABILITY.....'........................................

3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 B0 RATION CONTROL Shutdown Margin - T,yg > 200*F..........................

3/4 1-1

~

Shutdown Margin - T,yg i 200'F..........................

3/4 1-3 Boron Dilution.........................................

3/4 1-4 Moderator Temperature Coefficient......................

3/4 1-5 Minimum Tempera ture for Cri tical i ty....................

3/4 1-6 3/4.1.2 BORATION SYSTEMS Flow Paths - Shutdewa..................................

3/4 1-7 Flow Paths - Operating.................................

3/4 1-9 Charging Pump - Shutdown...............................

3/4 1-11 Charging Pumps - Operating.............................

3/4 1-12

' Boric Acid Transfer Pumps - Shutdown...................

3/4 1-13 Boric Acid Transfer Pumps - Operating..................

3/4 1-14 Borated Water Sources - Shutdown.......................

3/4 1-15 Borated Water Sources - Opera ting......................

3/4 1-16 3/4.1.3 MOVABLE CONTROL ASSEMBLIES Group Height...........................................

3/4 1-18 Position Indicator Channels............................

3/4 1-20 Rod Drop Time..........................................

3/4 1-21 Shutdown Rod Insertion Limit...........................

3/4 1-22 Control Rod Ir.s erti on Limi ts...........................

3/4 1-23 Part Length Rod Insertion Limits.......................

3/4 1-26 TROJAN-UNIT 1 III

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION Pace 3/4.2 POWER OISTRIBUTION LIMITS 3/4.2.1 Ax i a l Fl ux Di ffe rence......................'............

3/4 2-1

~

3/4.2.2 Hea t Fl ux Hot Channel Facto r...........................

3/4 2-5 3/4.2.3 RCS Flowrate and Fq....,..............................

3/4 2-8 3/4.2.4 Oua d ran t Powe r Ti l t R a t i o..............................

3/4 2-10 3/4.2.5 DNB Parameters.........................................

3/4 2-12 3/A.3 INSTRU51ENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION....................

3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION......................................

3/4 3-14 3/4.3.3 MONITORINC INSTRUMENTATION Radiation Moni toring Instrumentation...................

3/a 3-33 Movable Incore Detectors...............................

3/4 3-37 Seismic Instrumentation................................

3/4 3-38 Meteoroloaical Instrumentation.........................

3/4 3-41 Remote Shutdown Instrumentation........................

3/4 3-44 Chl orine Detection Systems.............................

3/4 3-47 Fire Detection Instrumentation.........................

3/4 3-48 3/4.4 REACTOP COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS No rmal 0 o e ra ti o n................ :......................

3/4 4-1 TROJAN-UNIT 1 IV Amendment No. JS, 46 Q

.~

na a--

~

INDEX BASES SECTION Pace 3/4.0 APPLICABILITY.............................................

B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS

~

3/4.1.1 BORATION CONTR0L........................................

B 3/4 1-1 3/4.1.2 BORATION SYSTEMS........................................

B 3/4 1-2 3/4.1.3 MOVABLE C0fRROL ASSEMBLIES..............................

B 3/4 1-3 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE...................................

B 3/4 2-1 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR, RCS FLOWRATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR............

B 3/4 2-4 3/4.2.4 OUADRANT POWER TILT RATI0...............................

B 3/4 2-6 3/4.2.5 DNB PARAMETERS..........................................

B 3/4 2-6 TROJAN-UNIT 1 IX Amendment No. 30,46 i

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-e-e+

INDEX BASES SECTION PAGE 3/4.3 INSTRUMENTATION 3/4.3.1 PROTECTIVE INSTRUMENTATION..............................

B 3/4 3-1 3/4.3.2 ENGINEEREDSAFETYFEATUREINSTRd'ENTATION...............

B 3/4 3-1 M

3/4.3.3 MONITORING INSTRUMENTATION..............................

B 3/4 3-1 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT L00PS...................................

B 3/4 4-1 3/4.4.2 and 3/4.4.3 SAFETY VALVES...............................

B 3/4 4-1 3/4.4.4 PRESSURIZER.............................................

B 3/4 4-2 3/4.4.5 STEAM GENERATORS........................................

B 3/4 4-2 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE.......................... B 3/4 4-2 3/4.4.7 CHEMISTRY...............................................

B 3/4 4-3 3/4.4.8 SPECIFIC ACTIVITY......................................

B 3/4 4-4 3/4.4.9 PRESSURE / TEMPERATURE LIMITS.............................

B 3/4 4-4 3/4.4.10 STRUCTURAL INTEGRITY...................................

B 3/4 4-9 TROJAN-UNIT 1 X

e 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (T

) shall not exceed the limits shown in Figures 2.1-1 and 2.1-2 for 4 and ggloop operation, a

respectively.

APPLICABILITY: MODES 1 and 2.

ACTION:

Whenever the point defined by the combination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate' pressurizer pressure line, be in HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig.

APPLICABILITY: MODES 1, 2, 3, 4 and 5.

ACTION:

MODES 1 and 2

.Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

MODES 3, 4 and 5 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes.

TROJAN-UNIT 1 2-1

_s w,

+

=

e

==

ew a=

-~w

e i::idiNiN 5

ji' qu::

.I.

'tUNACCEPTABLE

'~

~2 660

j 00ERATION

i:

+4' M 2400 PSIA l 1

l' II I

l

-- l-

1....

MO 2250 PSIA 1

- l j

l I.._..

l.Al l-i i

l I

I i

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~ 2000 PSIA

.j l

j 620 1

C E'

1775 PSIA a

i

~

j "E

600 l

l M

N k,

i 580

\\

ACCEPTABLE-r

' OPERATION i

e i

.4 560 t

I

....2-......-

e j.. _...;.

......L.....

540 0

0.2 0.4 0.6 0.8 1.0 1.2 FRACTION OF RATED THERMAL POWER Figure 2.1-1 Reactor Core Safety Limit-Four Loops in Operation.

TROJAN-UNIT I 2-2 Amendment NO. 46 8

~

l :

i l

.__._.....t 4

UNACCEPTABLE i 660

- - - l l

OPERATION 4

2400 PSIA l l

.. g.._.

...4.

l

~

i i

I I

.. j..;. y..

t 640

.l' I

t.

l l

2250 PSIA :

i 2000 PSIA.

620 f

l

\\

1850 PSIA

-- ! ~ ~ ~ '

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u

.P U

600 1

c:

4 580

\\; \\

\\

ACCEPTABLE _,.

OPERATION 560

-~-

i T

i i

1 1

._......._...+

540 0

0.2 0.4 0.6 C.8 1.0 1.2 l

i FRACTION OF RATED THERMAL POWER Figure 2.12 Reactor Core Safety Limit-Three Loops in Operation.

i TROJAN-UNIT 1 2-3 Amendment t'O. 46 l

e.

SAFETY LI"MITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The reactor trip system instrumentation setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1.

APPLICABILITY: As shown for each channel in Table 3.3-1.

ACTION:

With a reactor trip system instrumentation setpoint less conservative than the value shown in the Allowable Values column of Table 2.2-1, declare the channel inoperable and apply the applicable ACTION state-ment requirement of Specification 3.3.1.1 until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.

1' I

l TROJAN-UNIT 1 2-4 l

TABLE 2.2-1 REACTOR TRIP SYSTEl1 INSTRUMENTATION TRIP SETPOINTS h-FUNCTIONAL UNIT-

. TRIP SETPOINT ALL04ABLE VALUES

1. Manual Reactor Trip-tiot Applicable Not Applicable

2. Power Range, Neutron Flux Low hetpoint - 1 25% of RATED Low Setpoint - 5 26% of RATED TilERilAL POWER TilERMAL POWER I

liigh Setpoint - 5 109% of RATED High Setpoint - 5 110% of RATED TilERt-iAL P04ER

-THERMAL POWER

)

3. Power Range, Neutron Flux,

< 5% of RATED TilERMAL P0llER with

< 5.5% of RATED THERMAL POWER 4

Hiqh Positive Rate a tik.e constant > 2 seconds

)

With a time constant 1 2 seconds

4. Power Range, Neutron Flux,

< 5% of RATED TilERMAL POWER with

< 5.5% of RATED THERMAL POWER High Negative Rate a time constant 1 2 seconds Eith a time constant 1 2 seconds i

'?

5. Intermediate Range, Neutron 5 25% of RATED THERMAL POWER 5 30% of RATED TliERNAL POWER m

Flux 5

6. Source Range, Neutron Flux

< 10 ' counts per second 1 1.3 x 10 counts per second

7. Overtemperature AT See flote 1 See Note 3

8. Overpower AT See flote 2 See Note 3 R

9. Pressurizer Pressure--Low 3

1 1865 psig

> 1855 psig

10. Pressurizer Pressure--High 5 2385 psig 5 2395 psig

11. Pressurizer Water Level--liigh

_927 of instrument span

_93% of instrument span

12. Loss of Flow 1 90% of design flow oer loop

1 89% of design flow per loop

E

Design Flow is 92,925 gom per loop.

1-TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS

!E FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES E

q

13. Steam Generator Water 3.5% of narrow range instrument 2; 4% of narrow range instrument' Level--Low-Low span-each steam generator span-each' steam generator

14. Steam /Feedwater Flow

< 40% of full steam flow at

< 42.5% of full steam flow at R'TED THERMAL POWER coincident R' TED TilERMAL POWER coincident A

A Mismatch and Low Steam Generator Water Level with steam generator water level with steam generator water level 3,25% of narrow range instrument

3. 24% of narrow range instrument span--each steam generator span--each steam generator

15. Undervoltage-Reactor

> 68% each bus

> 67% each bus Coolant Pumps

16. Underfrequency-Reactor 3.57.5 llz - each bus 3.57.4 Hz - each bus Q2 Coolant Pumpt O)

17. Turbine Trip A.

Low Trip System 3,800 psig 2;700 psig Pressure B.

Turbine Stop Valve

> 1% open

> 1% open Closure

18. Safety Injection Input Not Applicable Not Applicable from ESF

19. Reactor Coolant Pump Not Applicable Not Applicable Breaker Position Trip

.A 8

>*.g.

TABLE 2.P-1 (Continued) j zZ

_ REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOTATION 1

(T-T)*K (P-P')-f (AI)]

NOTE 1:

Overteperature AT 1 aTg [K -K2 j

3 j

2 Indicated AT at RATED IllERMAL POWER (set to 100% of RATED THERMAL POWER when where: AT

=

AT = 61.45"F)

Average temperature, F T

=

T' 1 584.8 *F (indicated T,yg at RATED THERMAL POWER)

Pressurizer pressure, psig P

=

2235 psig (indicated 1:CS nominal operating pres,sure)

P'

=

1+1 S j

The function generateil by the lead-lag controller for T dynamic compensation

=

j 3

avg tj 2

Time constants utilized in the lead-lag controller for T,yg j = 30 secs,

&1

=

i v2 = 4 secs.

Laplace transfonn operator S

=

y Operation with 4 Loops Operation with 3 Loops g

=,

g K

= 1.32 K

= 1.17 j

j 5

K

= 0.02109 per *F K

= 0.02109 per *F 2

2 E

K

= 0.00100 per psi K

= 0.00100 per psi 3

3 1-

TABLE 2.2-1 (Continued) 2

-2 REACTOR TRIP SYSTEN INSTRUMENTATION TRIP SETPOINTS R

l-NOTATION (Continued)

Z of thd p(owe)ris a function of the indicated difference between top and bottom detectors and f al range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:'

(1) for qb q-qNndqbetween - 45 percent and + 3 percent, fj (AI) = 0 (wher are percent RATED TilERMAL P0l!ER in the top and bottom halvesofthecbrerespectivel,andq is tal RHAL NR in t*Ob

~

percent of RATED TilERMAL POWER (ii) for each percent that the magnitude of (q

-q exceeds - 45 percent, the AT trip setpoint shall be automaticalfy renu)ced by 2.20 percent of

'?

its value at RATED TilERMAL POWER.

~

(iii) for each percent that the magnitude of (q. - q exceeds + 3 percent, the AT trip setpoint shall be automaticalty rebu)ced by 2.66 percent of its value at RATED TilERMAL P04FR.

g-Note 2:

Overpower AT"5 ATg [K -K5' 4S T-K6(T-T")-f(AI)]

4 2

3-Where:

AT Indicated AT at rated power

=

g T

Average temperature, F

=

T" =

Indicated T at RATED TilERMAL POWER 1 584.8 *F avg k

K a

4=

1.08 K5=

0.02/*F for increasing average temperature K6=

0.00137/*F for T > T"; K6 = 0 for T 1 T" l

TABLE P.2-1 (Continued)

_4$

REACTOR TRIP SYSTEtt 14STRUMENTATION TRIP SETPOINTS f

Ei

'10 TAT 10N (Continued)

M

'3S=

The. function generated ley the rate laq controller for Tavo 1+'3S dynamic compensation j

'3 = Time constant utilized in the rate lag controller for Tavg 3 = 10 secs.

2

- 1 S=

Laplace transform operator f (AI) = 0 for all AI 2

NOTE 3: The channel's maximum trip point shall not exceed its computed trip point by more than 2 percent.

e

2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temoerature.

Operation above the upper boundary of the nucleate boiling reoime could result in excessive cladding temperatures because of the onset of deoarture from nucleate boilino (DNB) and the resultant sharp reduction in heat transfer coefficient.

DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temper-ature and Pressure have been related to DNB.'

This relation has been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions.

The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the ~3rgin to DNB.

The DNB design basis is that there must be at '. east a 95 percent probability with 95 percent confidence that DNB will not occur when the minimum DNBR is at the DNBR limit.

In meeting this design basis, uncertainties in plant operating l

parameters, nuclear and thermal parameters, and fuel fabrication param-eters are considered statistically such that there is at least a 95 percent confidence that the minimum DNBR for the limiting rod is greater i

than or equal to the DNBR limit. The uncertainties in the above plant parameters are used to determine the plant DNBR uncertainty. This DNBR uncertainty, combined with the correlation DNBR limit, establishes a design DNBR value which must be met in plant safety analyses using l

values of input _ parameters _ without uncertainties.

l The curves of Figures 2.1-1 and 2.1-2 show the loci of points of THERMAL POWER, Reactor Coolant System pressure, and averaoe temperature for which the calculated DNBR is no less than the desig-zlR value or the average enthalpy at the vessel exit is less than thu m ;halpy of saturated liquid.

TROJAN-UNIT 1 B 2-1 Amendment No. 46 e

e me

- ye I

9 SAFETY LIMITS BASES N

The curves are based on an enthalpy hot channel factor, F and a reference cosine with a peak of 1.55 f r axial power shaN., of 1,49 An N

allowance is included for an increase in F at reduced power based on 3g the expression:

F"H = 1.49 [1 + 0.2 (1-P)1 a

wherePisthefractionofRAkEDTHERMALPOWER 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 effect 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 intecrity of the Reactor Ccolant 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 osig) of component design pressure. The Safety Limit of 273S 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.

TP.1JAN-UNIT 1 B 2-2 Amendment No. 46

~

2.2 LIMITINGSAFEiYSYSTEMSETTINGS BASES 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS The Reactor Trip Setpoint Limits specified in Table 2.2-1 are the values at which the Reactor Trips are set for each parameter. The Trip 4

Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits. Opera-tion with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that each Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.

4 Manual Reactor Trip

-The Manual Reactor Trip is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip 4

capability.

i Power' Range, Neutron Flux i

The Power Range, Neutron Flux channel high setcoint crovides react:-

core protection.against re&ctivity excursions wnich are too rapid to be 4

protected by temperature and pressure protective circuitry.

The low set point provides redundant protection in the pcwer range for a power excursion beginning from low power. The trip associated with the low l

setpoint may be manually bypassed when P-10 is active (two of the four power range channels indicate a power level of above approximately 10 percent of RATED THERMAL POWER) and is automatically reinstated when P-i 10 becomes inactive (three of the four channels indicate a power level i

below approximately 10 percent of RATED THERMAL POWER).

Pcwer Range, Neutron Flux, High Rates The Power Range Positive Rate trip provides protection against

' rapid flux. increases which are characteristic of rod ejection events from any power level. Specifically, this trip complements the Power Range Neutron Flux High and Low trips to ensure that the criteria are met for rod ejection from partial power.

TROJAN-UNIT 1 8 2-3 1

.r

..,~, -

LIMITING SAFETY SYSTEM ~ SETTINGS BASES The Power Range Negative Rate trip provides protection to ensure that the minimum DNBR is maintained above the design DNBR value for multiple control rod drop accidents. The analysis of a single control rod drop accident indicates a return to full power may be initiated by the auto-matic control system in response to a continued full power turbine load demand or by the negative moderator temperature feedback. This transient will not result in a calculated DNBR of jess than the design DNBR value, therefore single rod drop protection is not required.

Intermediate and Source Ranoe, Nuclear Flux The Intermediate and Source Range, Nuclear Flux trips provide reactor core protection during reactor startuo. These trips provide redundant protection to the low setpoint trip of the Power Range, Neutron Flux channels. The Source Range Channels will initiate a reactor trip at about 10+,' counts Der second unless manually blocked when P-6 becomes active. The Intermediate Range Channels will initiate a reactor trip at a current level procertional to aperoximately 25 percent of RATED THERMAL POWER unless manually blocked when P-10 becomes active.

No credit was taken for operation of the trios associated with either the Intermediate or Scurce Range Channels in tne accident analyses; however, their functional capability at the specified trip settings is required by this specification to enhance the overall reliability of the Reactor Drotection System.

Overtemperature AT The Overtamperature AT trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power distribution, provided that the transient is slow with respect to piping transit delays from the core to the temperature detectors (about 4 seconds), and pressure is within the range between the High and Low Pressure reactor trips. This setpoint includes corrections for changes in density and heat capacity of water with temperature and dynamic com-pensation for piping delays from the core to the loop temperature detectors. With normal axial power distribution, this reactor trip limit is always below the core safety limit as shown in Figure 2.1-1.

If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the reactor trip is automatically reduced according to the notations in Table 2.2-1.

TROJAN-UNIT 1 B 2-4 Amendment No. 46

LIMITING SAFETY SYSTEM SETTINGS' BASES Operation with a reactor coolant loop out of service below the 4 loop P-8 set point does no't require reactor prote: tion system set point modification because the P-8 set point and associated trip will prevent DNB during 3 loop operation exclusive of the Overtemperature AT set point.. Three loop operation above the 4 loop P-8 set point is permis-sible after resetting the Kl. K2 and K3 inputs to the Overtemperature aT channels and raising the P-8 set point _to its 3 loop value.

In this mede of operation, the P-8 interlock and trip functions as a High 4

Neutron Flux trip at the reduced power level.

Overpower AT The Overpower aT reactor trip provides assurance of fuel integrity, e.g., no melting, under all possible overpower conditions, limits the required range for Overtemperature.aT protection, and provides a backup to the High Neutron-Flux trip. The setpoint includes corrections for changes in density and heat capacity of water with temperature, and dynamic compensation for piping delays from the core to the loop temperature detectors. No credit was taken for operation of.this trip in the accident analyses; however, its functional capability at the specified trip set-ing-is required by thi: : a:f ficati:n to enhance the Overall reliability of the Reactor Protection System.

l Pressurizer Pressure The~ Pressurizer-High and. Low Pressure trips are provided to limit the pressure range in which reactor operation is permitted. The High Pressure tripLis backed up by the pressurizer code safety valves for RCS: overpressure protection,-and is therefore set lower than the set pressure for.these valves (2485 psig). The Low Pressure trip provides protection by tripping the reactor in the event of a loss of reactor coolant pressure.

Pressurizer Water Level The Pressurizer High Water Level, trip ensures protection against Reactor Coolant System overpressurization by limiting the water level to a volume sufficient to retain'a steam bubble and prevent water relief h

TROJAN-UNIT 1-

's 2-5 L

V-1

-[

r LIMITING SAFETY SYSTEM SETTINGS BASES through the pressurizer safety valves. No credit was taken for operation of this trip in the accident analyses; however, its functional capability at the specified trip setting is required by this specification to enhance the overall reliability of the Reactor Protection System.

Loss of Flow The loss of Flow trips provide. core protection to prevent DNB in the event of a loss of one cr more reactor coolant pumos.

Above 10 percent of RATED THERMAL POWER, an automatic reactor trip will occur if the flow in any two loops droos below 90% of nominal full loop flow. Above 38% (P-8) of RATED THERMAL POWER, an automatic reactor trip will occur'if the flow in any single loop drops below 90% of nominal full loop flow. This latter trip will prevent the minimum value of the DNBR from going below 1.73 during normal operational transients and anticipated transients when 3 loops are in operation and the Over-temperature aT trip setooint 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 ONBR from coing below 1.73 during normal operational 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 in coincidence with a Steam Generator low Water Level trio 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 tne overall TROJAN-UNIT 1 8 2-6 Amendment No. 46

e p

/

e ::g-d = -

-..e--

__3-~

y_

gy w_

. - - + - _,.

4.

-.-.:=

=

--4_.

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+ - - 2 7 :*. T

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m-

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- O. J _ a _---

__---.-.9 N

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=

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_f-...__.

-~ --~* K g.

y "1

y O

o-a o

=

D --

g._

-._---.m___

a-

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=~_ = y r. 4. _.

3

=

g

... -= -._

5 10. _, ~ y~

4

=.: -

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_...:=---.3-

_-.-.4-

>=

4 u.

g N.

N

__gg

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g 48 O

= =_= - * : r;==r g-E g

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==

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=.

47 w

zW

=

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e

~ ~ t*

_ _ _ -. -. _i

.;!' = * ; 1.. *.. * * = :.' *. :1: **

N o

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ai o

o (zPd anznVWWON -(ZW l

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i.

TROJAH-UNIT 1 3/4'2-7

- ~ -

~--

POWER DISTRIBUTION LIMITS RCS FLOWRATE AND FR 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 q

operation (above 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:

~

AH

, and a.

Fg = 1.49 {l.0 + 0.2-(1.0 - P)}

THERMAL POWER b*

P = RATED THERMAL POWER APPLICABILITY: ' MODE 1 i

ACTION:

With the combination of RCS total flow rate and F outside the reofon of acceptable operation shown on Figure 3.2-3 or $.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 to within the above limits, or R

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 2a hours of initially being outside the above limits, verify through incore flux mapping and RCS total flow rate comparison that the combination of F and RCS total flow q

rate are restored to within the above limits, or reduce THERMAL P0uER to less than 5% of RATED THERMAL POWER within the rext 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

b TROJAN-UNIT 1 3/a 2-8 Amendment No. 30, 44, 46 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 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 reoion 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 nominal 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 attainino > 95% of RATED THERMAL

, POWER.

SURVEILLANCE REOUIREMENTS 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.3 shall be de: ermined :o be within the regica of acceptacie operation of Figure 3.2-3 or 3.2-4-(as applicable):

a.

Prior to operation above 75% of RATED THERMAL POWER after each fuel loading, and b.

At least once per 3] Effective Full Power Days.

Where:

N 4

p aH

, and FR

1.49 {1.0 + 0.2 (1.0 - P)}

N N

F H = Measured values of F obtained by using the movable incore detectors to b$tajn a power distribution map.

1 The measured values of F sinceFigures3.2-3a$$shallbeusedtocalculate F

3.2-4 include measurement q

calculational' uncertainties of 3.5% for flow and N

4% for incore measurement of FaH' i

4.2.3.3 The RCS total flow rate indicators shall be subiected to a s

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.

TROJAN-UNIT 1 3/4 2-9 Amendment No.46

.W 7y 4

C

+6

r o_

O

, og

1

~

s I

i i

48 "j-'

' l:"

. E: *..."

.g-

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. Measurement uncertainties of 3.5% ~

for flow and 4% for incore N

!!!!N! !!

2!!

id5; measurement of F[H are included -:;'

.jli ll~ h."l-

.ll

.:llj;;.

q

- in this figure.

.g i

g:.q k.

ii 25!!!N j,

. ?!:

f!'1[i_.

t

- [i _

L )g DNB LIMIT!*

f 1 -.

.. _.. _ /..

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[p LOCA LIMIT T h (1.074,43.5)

N y

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s

~

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w l

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l 42 E

ACCEPTABLE

~

[r j

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!!ll

/

o

/

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a u.

.f J

i 40 W

/

N m

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i

' UNACCEPTABLE

>a 38

/-

3

~

j

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(1.0,37.17) ' l

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i i-i

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36 h '.... p'.... :.

i

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(0.9'S2. '35.5)

.._....i.

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2.... !.

34 0.90 0.94 0.98 1.02 1.06 1.10 1.14 F[g /1.49 (1 + 0.2 ('1.P))

Figure 3.2-3 Flow vs. F[g Limit for 4 Loops in Operation.

1 i

l TROJAN-UNIT l 3/4 2-9a Amendment No. 46

, q..

.g.

, p.

.;p 38 Measurement uncertainties of 3.5% g::k--[

.j:.

.{

4...,..p..

j b

.b for flow and 4% for incore

~~~

M'!'

'T N

measurement of F are included AH in this figure.

g 36 s.

d

.. +..

h __g

__.. p_..

/.

. y.

[7..}.._.

..F.)IM DNB LIMIT

!.'. ' l 7..

f: !

N I

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g.

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i i

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/p LOCA LIMIT

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c:

ACCEPTABLE Q(1.074,31.03) _

g

/

t 3

s a

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0

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1:

24 0.90 0.94 0.98 1.02 1.06 1.10 1.14 F,Ng /1.49 (1 + 0.2 (1.P))

N Figure 3.2 4 Flow vs. F Lim t for 3 Loops in Operation.

gH 1

i TROJAN-UNIT 1 3/4 2-9b Amendment No. 46 o

1

POWER DISTRIBUTION LIMITS QUADRANT POWER TILT RATIO 1

LIMITING CONDITION FOR OPERATION 3.2.4' THE QUADRANT POWER TILT RATIO shall not exceed 1.02.

APPLICABILITY: MODE 1 ABOVE 50% OF RATED THERMAL POWER

ACTION:

a.

With the QUADRANT POWER TILT RATIO determined to exceed 1.02 but < l.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 1 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

3.

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 acceptable at 95% or greater RATED THERMAL POWER.

I b.

With the QUADRANT POWER TILT RATIO datermined to exceed 1.09 due to misalignment of either a shutdown, control or part length rod:

1.

Reduce THERMAL POWER at least 3% for each 1% of indi-cated QUADRANT POWER TILT RATIO in excess of 1.0, within 30 minutes.

See Special Test Exception 3.10.2.

TROJAN-UNIT 1 3/4 2-10 I

q,

.g._

q.

==_

-5

.y

POWER DISTRIBUTION LIMITING CONDITION FOR OPERATION (Continued)

- - 2.

Verify that'the QUADRANT POWER TILT RATIO is within its limit within 2 hour2.314815e-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 1 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

3.-

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 7

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 shut-down, control or part length 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 1 55% 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:

a.

Calculating the ratio at least once per 7 days when the alarm I

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.

TROJAN-UNIT 1 3/4 2-11

\\

a m

POWER DISTRIBUTION LIMITS DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB.related parameters shall be maintained within the limits shown on Table 3.2.1:

Reactor Coolant System T,yg.

a.

b'.

Pressurizer Pressure

~

c.

Reactor Coolant System Total F1ow Rate APPLICABILITY:

MODE 1 ACTION:

With any of-the'above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 4.2.5 Each of the parameters of Table 3.2-1 shall be veified to be within their limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

TROJAN-UNIT 1 3/4 2-12 Amendment No. 46

,j TABLE 3.2-1 c.

if DNB PARAMETERS c-

((

LIMITS

~"

4 Loops In 3 Loops In PARAMETER Operation Operation Reactor Coolant-System T 1 589 F

,5 580.4*F avg.

Pressurizer Pressure 3,2220 psia

3,2220. psia

Reactor Coolant System Figure 3.2-3 Figure 3.2-4 Total Flow Rate R

a Y

C

Limit not applicable during either a TilERMAL POWER ramp indrease in excess of 5% RATED THERMAL-POWER per minute or a THERMAL POWER step increase in excess of ]0% RATED TilERMAL POWER.

n 3

a o.

is

TABLE 3.3-1 (Continued) a.

Less than or equal to 5% of RATED TH'ERMAL POWER, place.

the inoperable channel in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and restore the inoperable channel to operable status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after increasing. THERMAL POWER above 5% of RATED THERMAL POWER; otherwise, reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

Above 5% of FATED THERMAL POWER, place the inoperable channel in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ; operation may continue until performance of the next required CHANNEL FUNCTIONAL TEST.

ACTION 8 - With the number of OPERABLE channels one less than the Total Numbers of Channels and with the THERMAL POWER level above P-7, place the inoperable channel in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ; operation may continue until performance of the next required CHANNEL FUNCTIONAL TEST.

ACTICN 9 - With a channel associated with an operating loop inoperable, restore the inoperable channel to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel associated with an operating loop may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance: testing per Specification 4.3.1.1.1.

l, ACTION 10 - With one channel inoperable, restore the inoperable channel to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to below P-8 within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

Operation below P-8 may continue pursLaa; to ACT:3 li.

ACTION 11 - With less than the Minimum Number of Channels OPERABLE, operation may continue provided the inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 12 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and/or open the reactor trip breakers.

REACTOR TRIP SYSTEM INTERLOCKS DESIGNATION CONDITION AND SETPOINT FUNCTION i

r P-6 With 2 of 2 Intermediate Range Prevents or defeats Neutron Flux Channels < 6 x 10-II the manual block of

amps, source range reactor trip.

TROJAN-UNIT 1 3/4 3-7 Amendment No. 33

TABLE 3.3-1 (Continued)

DESIGNATION

' CONDITION AND SETPOINT FUNCTION P-7 Hith 2 of 4 Power Range Neutron Prevents or defeats Flux Channels > 11% of RATED the automatic block of THERMAL POWER or 1 of 2 Turbine reactor trip on: Low impulse chamber pressure channels flow in more than one

> 66 psia.

primary coolant loop, reactor coolant pump under-voltage and under-frequency, turbine trip, pressurizer low pressure, and pressurizer hioh level.

P-8 With 2 of 4 Power Range Neutron Prevents or defeats THERMAL POWER.-

the automatic. block of Flux channels > 39% of RATED reactor trip on low coolant flow in a single loop.

P-10 With 3 of 4 Power range neutron Prevents or defeats flux channels < 9% of RATED the manual block of: P&ver THERMAL POWED.

range low setpoint reactor trip, Intermediate

. range reactor trip, and intermediate range rod stops.

Provides input to P-7.

' TROJAN-UNIT 1-3/4 3-8 Amendment No. 33,46 l

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

cladding mechanical properties to within assumed design criteria.

In addition, limiting the peak linea'r 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:

g(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-facturing tolerances on fuel pellets and rods.

N F

Nuclear Enthalpy Rise Hot Channel Factor, is defined as the g

ratio of the integral of linear power along the rod with the hi;hast integratad ;ower t: the aver:;a r:d ::ler.

xY(Z)

Radial Peaking Factor, is defined as the ratio of peak F

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 F (Z) upper n

bound envelope of 2.32 times the normalized axial peaking Yactor is not exceeded during either nonnal operation or in the event of xenon redis-tribution following power changes.

I Target flux difference is determined at equilibrium xenon conditions with the pari: length control rods withdrawn from the core. 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 fraction of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER for the associated core burnup conditions.

Target flux differences for othe' THERMAL POWER levels are obtained r

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.

l TROJAN-UNIT.1 B 3/4 2-1 Amendment No. 30,46 i

l

~

L l

'l POWER DISTRIBUTION LIMITS BASES Although it-is intended that the plant will be operated with the AXIAL FLUX DIFFERENCE within the +5% target band about the target flux difference, during rapid plant THERMAL POWER reductions, control rod motion will cause the AFD to deviate outside of the target band at re-duced THERMAL POWER levels.- This deviation will not affect the xenon redistribution sufficiently to change the envelope of peaking factors which may be reached on a subsequent return to RATED THERMAL POWER (with the-AFD within the target band) provided,the time duration of the devi-ation is limited.- Accordingly, a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months penalty deviation limit cumu-lative during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is provided for operation outside of the target band but within the limits of Figure 3.2-1 while at THERMAL POWER levels between 50% & 90% of RATED THERMAL POWER.

For THERMAL POWER levels between 15% & 50% of rated THERMAL POWER, deviations of' the AFD outside of the target band are less significant. The penalty of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months actual time reflects this reduced significance.

Provisions for monitoring the AFD on an autanatic basis are derived from the plant process computer through the AFD Monitor Alarm. The computer determines the one minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for at least 3 of 4 or 2 of 3 OPERABLE excore channels are outside the target band & the THERMAL POWER is greater than 90% of RATED THERMAL POWER.

During operation at THERMAL POWER levels between 50% & 90% & 15% & 50%

RATED THERMAL POWER, the computer outputs an alarm message when the penalty deviation accumulates beyond the limits of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months & 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , respectively.

Figure B 3/4 2-1 shows a typical monthly target band near the begin-i ning of core life.

1 i

TROJAN-UNIT-1 B3/4 2-2 r

,--e n.

--n-

Percent of Roted P'"

5%

100%

-~ O..

- --+- -- - --

y.==..,

_g--

90%

-:=q _ _.

^

J._ :-.._-

--._===g.

..; c=3 u _ ---'.

3.

E'm.

g

, _-=a

-,~-

t_.'_-.=_

g

. - ~, _

.~e

.m _

a g

Target Flum Difference 7T 50%

/-

l I

I k

-i -

m i

30%

i 20%

i 10%

i

-30%

20%

10%

0

+10%

+20%

+30%

)

INDICATED AXIAL FLUX OlFFERENCE Fipre B 3/4 21 TYPICAL INDICATED AXIAL FLUX DtFFERENCE VERSUS THERMAL POWER AT BOL TROJAN-UNIT 1 B 3/4 2-3 t

i

POWER DISTRIBUTION LIMITS BASES ~

~~

3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR, RCS FLOWRATE, AND 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 desian 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.

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' 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:

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 Specifications 3.1.3.5 and 3.1.3.6 are maintained.

d.

The axial power distr'bution, expressed in tenns of AXIAL FLUX DIFFERENCE, is maintained within the limits.

N F

will be maintained within its limits provided conditions a.

througgHd.aboveagemaintained.'AsnotedonFigures3.2-3and3.2-4, RCS flowrate and F may be ". traded off" against one anotger (i.e., a lowmeasuredRCSfiHowrate is acceptable if the measured F is also low) to ensure that the calculated DNBR will not be belgw the $sion DNBR value.

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

which is consistent with the iqitial conditions asSbd for the LOCA analysis. The relaxation of F' as a function of THERMAL POWER allows changes in the radial power shkhe for all permissible rod insertion limits.

1 When an F measurement is taken, both experimental error and manu-facturing tolehance must be allowed for.

Five percent 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 F, which has been conservatively estimated as 5% in NCAP-8692, " Fuel Rod Bowing." The aporopriate statistical combination of local power, manufacturing tolerance, and rod bow uncertainties results in a penalty on Fg of 7.68%, whereas i

multiolying measured values of F by 1.03 x 1.05 results in a penalty of 8.15%.

TROJAN-UNIT l-B 3/4 2-4 Amendment No. 30,46

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POWER DISTRIBUTION LIMITS

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BASES N

When RCS flowrate and F are measured, no additional allowances are necessary prior to compahYson with the limits of Figures 3.2-3 and 3 2-4.

Measurement errors of 3.5% for RCS total flow rate and 4% for 5

F'aH have been allowed for in determination of the design DNBR value.

The design DNBR values include a 20.2% margin for conservatism.

The effect of rod bow on DNBR has been determined to be a 10.6% Denalty.

Therefore, the availcble margin more thqm offsets the effect of rod bow, and no penalty is required on DNBR or F' g.

3/4.2.1 00ADRANT 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 F is depleted. The limit of 1.02 was selected to provide an allowance i

n for the uncertainty associated with the indicated cower tilt.

1 The 2-hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correction of a dropped or misaligned rod.

In the event such action does not correct the tilt, the margin for uncertainty on F is g

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, ff,46 am

o POWER DISTRIBUTION LIMITS I

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 naintain a minimum DNBR of 1.73 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 adequace to detect flow depradation 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.

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TROJAN-UNIT 1 B 3/4 2-6 Amendment No. 30,46 L

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3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS The plant is designed to operate with all reactor coolant loops in operation, and maintain DNBR above 1.73 during 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 that the ONBR will be maintained above 1.73.

A loss of flow in two loops will cause a reactor trip if operating above P-7-(10 percent of RATED THERMAL POWER) while a loss of flow in one loop will cause a reactor trip if operating above P-8 (38 percent of RATED THERMAL POWER).

A sinole reactor coolant loop provides sufficient heat removal capability for removing core decay heat while in HOT STANDBY:

however, single failure '.ansiderations require placing a RHR loop into operation in the shutdown coe'ing mode if component repairs and/or corrective actions cannot be made within the allowable out-of-service time.

3/4.?.2 and 3/4.4.3 SAFETY VALVES The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 psig.

Each safety valve is designeq to relieve 420,000 lbs per hour of saturated steam at 110% of the valve's I

setpoint. The relief capacity of a single safety valve is adequate to relieve any overpressure. condition which could occur during shutdown.

In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization.

During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2735 psig. The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from a complete loss of load assuming no reactor trip until the first Reactor Protective System trip setpoint is reached (i.e., no credit i-s taken for a direct reactor trip on the loss of load) and also assuming no operation of the power-l operated relief valves or steam dump valves.

Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the provisions i

of Section XI of the ASME Boiler and Pressure Code,1974 Edition.

TROJAN-UNIT 1 B 3/4 4-1 Amendment No.46

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REACTOR COOLANT SYSTEM l

BASES 3/4.4.4 PRESSURIZER A steam bubble in the pressurizer ensures that the RCS is not a hydraulically solid system and is capable of accommodating pressure surges during operation. The steam bubble also protects the pressurizer code safety valves and power operated relief valves agains water relief.

The power operated relief valves and steam bubble function to relieve RCS pressure during all design transients up to and including the design step load decrease with steam dump. Ope. ration of the power operated relief valve minimizes the undersirable opening of the spring-loaded pressurizer code safety valves.

3/4.4.5 STEAM GENERATORS One OPERABLE steam generator provides sufficient heat removal capability to remove decay heat after a reactor shutdown. The requirement for two OPERABLE steam generators, combined'with other requirements of the Limiting Conditions for Operation ensures adequate decay heat removal capabilities for RCS temperatures greater than 350 F if one steam ge erator becomes inoperable due to single failure considerations.

Below 35C.,

decay heat is removed by the RHR system.

The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained. The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.

Inservice inspection of steam generator tubing is essential in o'rder to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion.

Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.

The plant is-expected to be operated in a manner such that the secondary coolant will be maintained within those parameter limits found i

to result in negligible corrcsion of the steam generator tubes.

If the secondary coolant chemistry is not maintained within these parameter limits, localized corrosion may likely result in stress corrosion cracking.

The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the primary coolant l

Change No. 3 TROJAN-UNIT 1 B 3/4 4-2 December 19, 1975

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