Text

Proposed Tech Specs Re Cycle 5 Reload
	ML17156B021
Person / Time
Site:	Susquehanna
Issue date:	02/02/1989
From:	; PENNSYLVANIA POWER & LIGHT CO.
To:	;
Shared Package
ML17156B020	List: ML17156B019; ML17156B021; ML17156B022; PLA-3141, Forwards Application for Amend to License NPF-14,consisting of Proposed Amend 119,changing Tech Specs in Support of Ensuing Cycle 5 Reload;
References
	NUDOCS 8902100097
	Download: ML17156B021 (51)
	v • d • e

LIHITING COHDITIONS FOR OPERATIOH AHD SURVEILLAHCE RE U1REHEHTS 5ECT10N 3/4.0 APPLICABILITY, 3/4. 1 REACTIVITY CONTROL SYSTEHS 3/4. 1. 1 SHUTOOO HARGIN.................................,......

PAGE 3/4 0-1 3/4 1-1'/4.1.2 REACTIVITYNKNALIES...,........ ~.. ~....

~..

3/4 1-2 3/4. 1.3 COHTROL RODS 0

Control Rod Operability.....'.................

3/4 1-3 Control Rod kaximum Scram Insertion Times..............

3/4 1-6 Control Rod Average Scram Insertion Times..............

3/4 1-7 Four Control Rod Group Scram Insertion Times...........

3/4 1-8 Control Rod Scram Acc~lators.........................

Control Rod Drive Coupling.

3/4 1-9 3/4 1-11 Control Rod Position Indication....,............'.......

3/4 1-13 Rod Block Monitor.....................

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3/4. 1.5 STANDBY LIQUID CONTROL SYSTEH..........................

3/4. 2 PIER DISTRIBUTION LIHITS

" Control Rod Dr ive Housing Support......,...............

3/4.1.4 CONTROL ROD PROGRAH CONTROLS Rod Worth Ninimizer.t ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ t ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

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Rod Sequence Control System...........

3/4 1-15 3/4 1-16 3/4 1-17 3/4 1-1S 3/4 1-19 3/4,2.1 AVERAGE PLANAR LINEAR HEAT GEHERATIOH RATE.............

3/4 2-1 3/4 2a2 APO SETPOIHTS ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~ ~ ~ ~ oo ~ ~ ~ ~ ~ ~

'3/4 2-5 3/4.2.3 MINION CRITICAL PAAR RATIO...........................

3/4 2-6 3/4.2.4 LINEAR HEAT GENERATION RATE E448>> r ~ ~ ~ ~ ~ ~ ~ e ~ ~ 'r ~ ~ ~ ~ ~ o ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~

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AHF FUELo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ *'

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Amendment No 72 4CT og~

LIST OF FIGURES INDEX F1GVRR

3. 1. 5-1 3.1.5 2

3.2.1-f I 3,2.l-f0 3.2. 2-1 3.2. 3-1 3.2. 3-2 3.2.4

-1 3.2.4/2 3.4. i.

1-1'.4.6.1-1 B 3/4 3-1 B 3/4.4.6-1 5.1. 1-1 5.1.2-1 5.1. 3-la 5.1. 3-lb 6.2. 1-1 6.2. 2-1 SODIUN PEHTABORATE SOLUTION TENPERATURE/

CONCENTRATION REQUIRKNEHTS...............

~ ~ ~ ~ ~ ~.

SODIUN PEHTABORATK SOLUTIOH CONCEHTRATION

~ ~ ~ ~

~

~ ~

THIS PAGE INTENTIONALLY LEFT bLANK...... ~

~

~ ~ ~ ~ ~

4INRNJ~VKRAG&+bANA~HKA~~EHERAROH 4kTE"~PNG~~VERAG~tANA~POSURE, IEAXINJN AVERAGE PLANAR LINEAR HEAT GEHERATIOH RATE (MAPLHGR) VS AVERAGE BUNDLE EXPOSURE ~

ANF Bxb FUEL..;..................................

NAXINN AVKRAGE PLANAR LINEAR HEAT GENERATION RAT (NAPLHGR) VS.

AVERAGE BUNLE EXPOSURE, AHF 9x9 FUKLe ~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o, ~ r'o ~ i ~ ~ ~ ~

~ ~

LINEAR HEAT GENERATIQN RATE FOR APlN SKTPOIllTS VERSUS AVERAGE PLANAR EXPOSURE, ANF FUEL.........

'FL& DEPENDENT NCPR OPERATIHG LINIT..............

REDUCED PStER NCPR OPERATING LINIT...............

LINEAR HEAT GKHERATION RATE (LHGR) LIIIITVERSUS AVERAGE PLANAR EXPOSURE ANF Bx8 FUEL.............

LINEAR HEAT GEHERATIOH RATE (LHGR) LINIT VERSUS AVERAGK PLANAR EXPOSURE, ANF 9x9 FUEL............

THERNAL NEER LINITATIONS........................

NININN REACTOR VESSEL NETAL TENPERATURE VS.

REACTOR VESSEL PRESSURE.........................

REACTOR VESSEL WATER LEVEL......................

FAST NEUTRON FUJENCE (fPINeV) AT 1/4 T AS A FUNCTION OF SERVICE LIFE......... ~..............

EXCLUSION AREA

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~ ~ ~ ~ ~ ~ ~ ~ ~ ~

LN POPULATION ZONE

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ RE ~

NAP DEFINIHG UNRESTRICTED AREAS FOR RADIOACTIVK GASEOUS AN LIQUID EFFLUKNTS.........,....,...,.

'AP DEFINING UNRESTRICTED AREAS FOR RADIOACTIVE'ASKQUS AND LIQUID EFFNENTS...............'.......

OFFSITK ORGANIZATIQN......................,...,...

UNIT ORGANIZATION ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~ o ~ ~

PAGE 3/4 1-21 3/4 1 22 3/4 2-2 3/4 2-4 3/4 2 4a 3/4 2 7 3/4 2 9 3/4 2-9a 3/4 2-10b 3/4 2 10c 3/4 4-1b 3/4 4 18 B 3/4 3 8 8 3/4 4-7 52 53 5"4 55 6 4 SUSQUEHANNA - UNIT 1

~nant Ho. 72

TABLE 2.2. 1-1 REACTOR PROTECTION SYSTEH INSTRUHENTATION SETPOINTS c.. Heutron Flux-Upscale 3.

d.

Inoperative Reactor Vessel Steam Dome Pressure

- High Reactor Vessel Water Level - Low, Level 3

5.

Hain Steam Line Isolation Valve - Closure 6.

Hain Steam Line Radiation - High 7.

8.

Drywell Pressure - High Scram Discharge Volume Water Level - High a.

Level Transmitter b.

Float Switch FUNCTIONAL UNIT l.

Intermediate Range Honitor, Neutron Flux-High m

e 2;

Average Power Range Honitor:

a.

Neutron Flux-Upscale, Setdown I

b.

Flow Biased Simulated Thermal Power-Upscale 1)

Flow Biased 2)

High Flow Clamped TRIP SETPOINT

< 120/125 divisions of full scale

< 15K of RATED THERHAL POWER

< 0.58 M<59K, with a maximum of

< 113.5X of RATED THERHAL POWER

< llBX of RATED THERHAL POWER

< 1037 psig

.> 13.0 inches above instrument zero*

< lOX closed

< 7.0 x full power background

< 1.72 psig

< 88 gallons

< 88 gallons ALLOWABLE VALUES

< 122/125 d>v>sions of full scale

< 20K of RATED THERHAL POWER

0. 58 M>62K, with a maximum of

< 115.5X of RATED.

THERHAL POWER

< 120K of RATED THERHAL POWER

< 1057 psig

> 11.5 inches above instrument zero

< llew closed

< 8.4 x full power Background.

< 1.88 psig

< 88 gallons

< 7X closed fop 9.

10.

11.

12.

"See See Turbine Stop Valve - Closure Turbine Control Valve Fast Closure, Trip Oil Pressure - Low Reactor Hode Switch Shutdown Position Hanual Scram

< 5.5X closed

> 500 psig HA NA Bases Figure B 3/4 3-1.

Specification 3.4. 1. 1.2.a for single loop operation requirement.

I

> 460 psig HA

'2 s

2.

1 SAFETY LIMITS BASES

2. 0 INTROOUCTION The fuel cladding, reactor pressure vessel and primary system piping are the principal barriers to the release of radioactive materfals to the environs.

Safety Limits are established to protect the integrity of these barriers during normal plant operations and anticipated transients.

The fuel cladding ir;egrity Safety Limit is set such that no fuel damage is calculated to occur if t.-e limit is not violated.

Because fuel damage is not directly observable, a step-back approach fs used to establish a Safety Limit such that the MCPR is not less than the lfmit specified fn Specifications

2. 1.2 forO~

"e AFIF MCPR greater than the specified limit represents a conservative margin re ative to the conditions required to maintain fuel cladding integrity.

The fuel cladding is one of the physical barriers which separate the radioactive mate-rials from the environs.

The integrity of this cladding barrier is related to fts relative freedom from perforations or cracking.

Although some corrosion or use related cracking may occur during the life of.tKe cladding, fission product migration from this source is fncrementally coaulatfve and contfnuously me'asurable.

Fuel cladding perforations, howevera can result from thermal stresses which occur from reactor operatfon significantly above design condi-tions and the Limiting Safety System Settings.

While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration.

Therefore, the fuel cladding Safety Limit is defined with a margin to the conditions which would produce onset of transition boiling, MCPR of 1.0.

These conditions represent a significant departure from the con-dition intended by design for planned operation.

The MCPR fuel cladding inte-grity Safety Limft assures that during normal operation and during antic'.pated operational occurrences, at least 99.9X of the fuel rods in the core do not experience transition boiling (ref. XN-NF-524(A)).

2.1.1 THERHAL PRHER Log Pressure or Log Fleas~

The use of the XN-3 correlation fs valid for critical power calculations at pressures greater than 580 psig and bundle mass fluxes greater than 0.25 x 10e lbs/hr fta.

For operation it low pressures or low flows, the fuel cladding integrity Safety Limit fs established by a limiting conditfon on cote THERMAL POWER with the following basis:

Provided that the water level in the vessel downcomer is maintained above the top of the active fuel, natural circulation fs sufficient to assure a

minimum bundle flow for all fuel assemblies which have a relatively high powe~

and potentially can approach a critical heat flux condition.

For the ANF 9x9 fuel design, the mfnfaa bundle flow fs greater than 30,000 lbs/hr, For the ANF e~E Bx8 fuel, the mfnfaea bundle flow fs greater than 28,000 lbs/hr.

l For all designs, the coolant mfnfam flow and maxfam flow area fs such that the mass flux fs always greater than 0.25 x 10e lbs/hr fta.

Full scale cri-tical power tests taken at pressures down to 14.7 psfa indicate that the fuel assembly critical power at 0.25 x 10e lbs/hr fte is 3.35 Mwt or greater.

At SUS(UEHANNA UNIT 1 8 2-1 Amendment No. 72

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%VER OISTRIBUTIOH LIHIT5 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GEHERATIOH RATE LINITING CONOITIOH FOR OP ERATIOH 3.2.1 All AVERAGE PLANAR LINEAR HEAT GEHERATIOH RATES (APLHGRs) for tach type of fuel as a function of ~9tAQ~bAHAf~P05UR~o~Mu~n~VERAGE BUNDLE

. EXPOSURE for ANF fuel shall not exnean tha le ~lts sheen ln FIOures 3.2.1-1~ arrSSFj 3.2. l.2g,end-@4~1&

APPLICABILITY:

OPERATIOHAL COHOITIOH 1, vhen THERNAL PSfER is greater than or ACTION:

8th an APLHGR exceeding the lfaits of Figure 3. 2. 1-1$ 3. 2. 1-2, oWkAH initiate corrective action vithin 15 minutes and restore APLHGR to vfthin the riqufred liiits vithin 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERNAL PSIER to less than 25Z of RATED THERPQL ~ER vithin the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVE'ILLANCE RE UIRENEHTS

4. 2. 1 All APLHGRs shall be verified to be equal to or less than the lfeits determined froa Figures 3.2.1-l,p3.2.1-2)(sand-3Av4&t a.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.

Nthfn 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months afte~ c~letfon of a THENhL NMKR increase of at least 15X of RATEO THERNL KITER, and

c. " Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months vhen the reacto~ fs operating vfth a LINITIHG CONTROL ROO PATTERN for APufGR.

d.

The provisions of Specfffcatfon 4.0.4 are not applicable.

<<Sae Specfffcatfon 3.4.1.1.2.a for sfngle loop operation requfrments.

SUSQUEHANNA UHIT 1 3/4 2-1 Nee~nt No. 72 9 1987

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SAFETY LIMITS BASES

2. 1.2 THERMAL PMER Hi h Pressure and Hi h Flow Onset of transition boiling results in a 4ecrease in heat transfer from the clad and, therefore, elevated clad temperature and the possibility of clad fai lure.

However, the existence of critical power, or boiling transition, is not a directly observable parameter in an operating reactor.

Therefore, the rargin to boiling transition is calculated from plant operating parameters such a:. core power, core flow, feedwater temperature, and core power distribution.

The margin for each fuel assembly is characterized by the critical power ~atio CPR),

which is the ratio of the bundle power which would produce onset of transit on boiling divided by the actual bundle power.

The minimum value of this ratio for any bundle in the core is the minimum critical power ratio (MCPR).

The Safety Limit NCPR assures sufficient conservatism in the operating NCPR limit that in the event of an anticipated operational occurrence from the limiting condition for operation, at least 99.9% of the fuel ro4s in the core would be expected to avoid boiling transition.

The margin between calculated boiling transition (NCPR ~ 1.00) and the Safety Limit ICPR is based on a detai 1-ed statistical procedure which considers the uncertainties in monitoring the core operating state.

One specific unce tainty included

<n the safety limit is the uncertainty inherent in the XN-3 critical power correlation.

XN-NF-524 describes the methodology used in determining thi Safety Limit NCPR:

L(p) pe~)5t~j The XN-3 critical power correlation is based on a significant body of practical test data, providing a high degree of assurance that the critical power as evaluated by the correlation is ~ithin a small percentage of the actual critical power being estimated.

As long as the core pressure and flow are within the range of validity of the XN-3 correlation (refer to Sec-tion 8 2.1. 1), the assed reactor conditions used in defining the safety limit introduce conservatism into the limit because boun4ing high ra4ial power fa--

tors and bounding flat locil peaking distributions are used to estimate the neer of rods in boiling transition.

Still further conservatism is induced by the ten4ency of the XN-3 correlation to overpredict the number of rods in boiling transition.

These conservatisms and the inherent accuracy of the XN-3 correlation provide a reasonable 4egree of assurance that during sustained operation at the Safety Limit NCPR there woul4 be no transition boiling in the core.

If boiling transition were to occur, there is reason to believe that the integrity of the fuel would not necessarily be compromised.

Significant test 4ata ace~lated by the U.S. Nuclear Regulatory Commission and private organiza-tions indicate that the use of a boiling transition limitation to protect against cladding failure is a very conservative approach..

Nuch of the data in4icates that LlO fuel can survive for an extended period of time in an envirotaent of boiling transition.

SUS/UEHANNA - UNIT 1 B 2-2 isaendment No, 7>

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POWER DISTRIBUTION LIMlTS 3/4. 2. 2 APRM SETPOINTS LIMITING CONDITION FOR OPERATION 3.2.2 The APRM flow biased simulated thermal pow~r upscale scram trip setpoint (5) and flow biased neutron flux-upscale control rod block trip setpoint (SRE) shall be establfshed according to the following relationships:

Tri Set oint Allowable Value "SRS

< (O.MW + SOX)T

< (O.SN + 53X)T where:

S and SRg are in percent of RATED THERNL POWER,

> Loop recfrculatfon flow as a percentage of the loop frculatfon flow which produces a rated core flow of 1'00 eflifoh lbs/hr, T

< Lowest value of the ratio of FRACTION OF RATED THERMAL POWER divided by the NXIHUM FRACTION OF LIHITINQ POWER DENSITY.. %&em h~RAC+NN O~IMIRNG POWER<ENS~~R A~~aeM~~ct~NEARWEHWEHERA%0NN 4AT~~R~vfdeMy H~MpecA44cav nd The FLPD fOr ANF fuel fS the aCtual LHG" divided by the LINEAR HEAT GENERATION RATE from Ffgure 3.2.2-1.

T fs always less than or'qual to 1.0.

APPLICABILITv:

OPERATIONAL CONDITION 1, when THERNL POWER fs greater than or ACTION:

With the APRM flow biased simulated thermal power-upscale scram trip setpofnt ind/or the flow biased neutron flux-upscale control rod block trip setpoint less conservative than the value shown fn the Allowable Value column for 5 or SR<,

as above determined, initiate corrective action within 15 minutes and ad)ust 5 and/or SR< to be consistent with the Trip Setpofnt value 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 2SX of RATED THERNL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

greater than the FRTP during power ascensfon up to 90X of RATED THERNL POWER, rather than ad)ustfng the'APRM setpofnts, the APRM gain may be ad)usted such that APRM readings are greater than or equal to 100X times

MFLPD, provided that the ad)usted APRM reading does not exceed 100X of RATED TNERNL POWER, the required gain ad)ustment increment does not exceed 10X of RATED THERNL POWER, and a notice of the adjustment fs posted on the reactor control panel.

See Specification 3.4.1.1.2.a for single loop operation requfrements.

SUSQUEHANNA UNIT 1 3/4 2 5 Amendment No. 7>

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~f POvER OISTRIEU. ION Ll ITS 3/4.2.3 NIHIL. CRITICAL OSIER RATIO LIHITINC CONDITION FOR OPERATION 3.2.3 The HINIH'O'. CRITICAL POSER RAT10 (HCPR) shall be greater than or eoual to the greater oI Lne tvo values determined frow Figure 3.2.3 1 and F i QLI re 3.2. 3 2

APPLICABILITv.

OPERATIONAL CONOIT]ON 1, vhen THERHAL PSCR is greater than or ACTION:

.Nth HCPR less than the apylicable HCPR Iioit deteroined above,, initiate cor-rective action vithin 1S o<nutes and restore HCPR to Hth{n, the required licit

~ithin 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERHAL POCKR to less than AX of RATKO THERtNL PSCR uithin the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE RE UIREHEHTS 4.2.3.1 H PR shal>

be deterained to be greater than 'or equal to the applicable HCPR Iiait deteroinec froo Figure 3.2.3-1 and Figure 3.2.3 2:

I At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.

'Nthin 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after coapletion of a THERNL PoIER increase of at least 1$L of RATKO THEINAL POIKR, and c.

Initially and at least once per 12 hburs ~n the reactor is ope~ating v<tt a LIHITIH< CONTROL ROD PATTERN for HCPR.

d.

The yreviSiOnS Of SpeCifiCatien 4.0.4 are nOt appliCable.

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8.3 I,

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.10OOO COOOO 3OOOO 40000 Average Planar Exposure (MWD/MT)

'000 LlNEAR HEAT CiENERATlON RATE FOR APRM SETPOINTS VERSUS AVERAGE PLANAR EXPOSURE ANF FUEL RGURE M2 1 Susquahanna

- Un1t l.

3/4 2 Aaendaent No. 72 I

~

fa

~JI ill C.IS IS 1.7 CURVE A: EOC-BPT Inoperabfo, Main Turbine Bypass Operablo CURVE 8: Main Turbine Bypass Inoperablo; EOC-APT Operable CU E C: EOC-APT and Main Turbine pass Operable 1.39 1.3 1.31 1.2S 1.2 60 eo 70 80 Total Core Flow (% OF RATED) 90 100 FLOW DEPENDENT MCPR OPERATING LIMIT FIGURE 3.2.3-$

e

"~ 0 -"

~

t ik I

a

1.7 1.6 cn 1.5 C

I CL0 1A CL (3

(40,<.52)

(47.5,1.40)

(46.25, 1.42)

CURVE A: EOC-RPT Inoperable; Main Turbine Bypass Operable CURVE 8: Main Turbine Bypass Inoperable; EOC-RPT Operable CURVE C: EOC-RPT and Main Turbine Bypass Operable 142 1.40 (50,1.36) 52.7,1. 33) 1.33 1.2 40 50 60 70 80 Total Core Flow (% OF RATED) 90 100 FLOW DEPENDENT MCPR OPERATING LIMIT FIGURE 3.2.3-1

cn l7 0>>

l.o-CURVE A: EOC-RPT fnoperabfe:

Main Turbine Bypass Operable CURVE 8: Main Turbine Bypass Inoperable..

EOC-RPT Operable CURVE C: EOC-RPT and Main Turbine Bypass Operable l.6-w C

~

~0

~

OI: lA-CLO 4.3-O l.2-0 30 80 60 50 70 Core Power (% OF RATED)

REDUCED POWER MCPR OPERATING LIMIT Figure 3.2.3-2 90 100

4 g,i g~e

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1.7 (26,1.55)

(40,1.63)

CURVE A: EOC-RPT Inoperable:

Main Turbine Bypass Operable CURVE B: Main Turbine Bypass Inoperable; EOC-RPT Operable CURVE C: EOC-RPT and Main Turbine Bypass Operable

~~

1.5 CP L

(9 CL0 1.4 CLO (26.1.63)

(26,1.46)

(4O,1.61)

(40,1 43)

(66.<.60)

(65,1.48)

(86.1.4O)

(80,1.47)

(80,1.45) 1.42 1.40 (80,1 37)

(9O.1.33) 1.33 20 30 40 60 60 70 80 Core Power {% OF RATED)

REDUCED POWER MCPR OPERATING LIMIT Figure 3.2.3-2 90 100

P ER OISTRIBUTION LIHITS 3/4;.4 LINEAR HEAT GENERATION RATE GE'OE

'I~mNG NoirION POR OPERATION 3.2.4.L The NEAR HEAT GENERATION RATE (LHGR) f QK fuel sha11 not exceed 13.4 5vlft.

APPLICABILITY:

O QATIONAL CONDITION 1, when ERNL POWER fs greater than or lATRI IIIBINiICNH.

ACTION:

Mfth the LHGR'of any fue rod excaedfng e lfmft, fnftfata corractfve actfan wfthfn 15 mfnutas and res re the LHGR wfthfn the lfmft withfn 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL PSfSR to 1 ~ s than 2KC of RATN THEQQL RSN wfthfn the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE RE UIREHENTS 4.2.4.1 LHGRs for GE f 1 shall be rmfned to be equal to or less than the 1 fest:

a.

At 7east nca per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.

Nthfn hours after completfon o

a*THERMAL RSKR fncrease of at lees 15K of RATED THERE'LL POWER, c.

I tfa1ly and at least once per 12 ho when the reactor fs ratfng on a GING CONTROL ROD PA for LHGR.

d.

The provfsfons of Specfffcatfon 4.0.4 are t applfcable.

F SUSQUEHANNA UNIT 1 3l4 2 M Amendment No. 57

POWER 01 STRIBUTION LIMITS 3/4.2.4 LIHEAR HEAT GEHERATION RATE AHF FHE!

LIMITING CONOITION FOR OPERATION 3.2.4HLH Tha LINEAR HEAT GENERATION RATE (LHGRQ for ANF foal shall oot axoaao

, the LHGR limit determined from Figures 3.2./1 and 3.2.4

-2.

APoLICABILITY:

OPERATIONAL CONOITIOH 1, when THERMAL POWER is greater thar. or ACTION With the QGR of any fuel rod exceeding its applicable limit from Figure

..2,'r 3.2.4/5.2, initiate corrective action within 15 minutes and restore the LHGR to within the 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 thar 25X of RATEO THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE RE UIREMEHTS x

4.2.4/

LHGRs for ANF fuel shall be determined to be equal to or less than the I

limit~

b.

C.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a

THERMAL POWER increase of at least 15X of RATEO THERMAL POWER, and Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating on a LIMITIHG CONTROL ROD PATTERH for LHGR.

d.

The provisions of Specification 4.0.4 are not applicable.

SUSOUEHANHA UHIT 1 3/4 2 10a Amendment Ho.

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NOOO 20000 30000 40000 Average Ptanar Expaauro (hIWD/MTl UNEAR HEAT GENERATION RATE (LHGR) LIMIT VERSUS AVERAGE PLANAR EXPOSURE ANF SXB FUEL FIGURE 3.2.4.2-1

C

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REGION OF OPERATION 0

~

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~

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0 10000 20000 30000 40000 Average Planar Exposure (MWD/MT)

LINEAR HEAT GENERATION RATE (LHGRj LIMIT, VERSUS AVERAGE PLANAR EXPOSURE ANF 8X8 FUEL FIGURE 3.2A-1 60000

14 12 10 8

6

-- 0.0;:.

13.0 m

24,000;::

12.0

PERMISSABLE::

REGION OF OPERATION I

I 35,000; 9.5 48,000; 772 0

10000 20000 30000 40000 Average Planar Exposure (MWD/MT)

LINEAR HEAT GENERATION RATE (LHGR) LIMIT VERSUS AVERAGE PLANAR EXPOSURE ANF SXB FUEL FIGURE 3.2.4-2 60000

z n

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PEBMISSAB REGIO F

OPE N

0 10000 20000 30000 10000 Average Planar Exposure (MWD/MT) 0000 UMEAR HEAT GENERATION RATE {LHGR) LIMIT VERSUS AVERAGE PLANAR EXPOSURE ANF 8XS FUEL RGUBE 3.2.4.2-2

~ ~'t i ( y

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TRIP 2.

4.

5.

6.

TABLE 3.3.6-2 CONTROL ROD BLOCK IHSTRUHENTATION SETPOINTS F UNCTIOH TRIP SETPOINT ROD BLOCK HONITOR a.

Upscale b.

Inoperative c.

Downscal e APRH a.

Flow Biased Neutron Flux - Upscale O'W Inoperative Downscale Heutron Flux - Upscale Startup SOURCE RANGE HONITORS a.

Detector not full in HA b.

Upscale

<2x10 cps c.

Inoperative RA d.

Downscale

> 0.7 cps"~

INTERNEDIATE RANGE HONITORS a.

Detector not full in b.

Upscale c.

Inoperative d.

Downscale SCRAH DISCHARGE VOLUHE a.

Mater'evel - High

< 44 gallons REACTOR COOLANT SYSTEH RECIRCULATION FLOM ALLOMABLE VALUE

< 0.66 M+ 4W

< 0.66 M + 45K HA HA

> 5/125 divisions of full scale

> 3/125 of divisions full scale

< 0.58 M+ 50K" RA SX of RATED THERHAL POMER

< 0.5S M+

53X'A

> 3X of RATED THERHAL POMER b.

C.

d.

< 12K of RATED T}lERHAL POMER 14K of RATED THERHAL POMER HA<4x 10 cps RA

> 0.5 cps""

HA NA

< 108/125 divisions of full scale

< 110/125 divisions of full scale RA RA

> 5/125 divisions of full scale

> 3/125 divisions of full scale

. < 44 gallons a.

Upscale

< 10B/125 divisions of full scale

< ill/125 divisions of full scale

b. Inoperative'A HA c.

Comparator

< lOX flow deviation

< ill flow deviation

{M). The'trip setting of this function aust be maintained in accordance with Specificat.ion

3. 2. 2.

"Provided signal-to-noise ratio is >2.

Otherwise, 3cps as trip setpoint and 2.8cps for allowable value.

NSee Specification 3.4. l. 1.2.a for single loop operation requirements..

~f k)~

fgf ~

I f

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REACTOR COOLANT SYSTEN RECIRCULATION LIES SINGLE LOOP OPERATION LIHITING CONOITION FOR OPERATION 3.4.1.1.2 One reactor coolant recirculation looy shall be fn operation vfth the yuep speed c 80K of the rated puep syeed, ind a.

the following revised specification i{cits shall be followed:

I.

Speci ffcatfon 2.1.2:

the NCPR Safety Lfaft shall be fncreised to 1.07.

2.

Table 2.2.1-1:

the APN Flow Sfased Scrm Trfy Setyofnts, shall be as follows:

3.

Trf Se ofnt c

Specification 3.2.1:

The NPLHGR lfa{ts shall be is eHower Hl

+~hAkr

~~N~ue{~hMH syecfffed fn Figures 3.2.1-2 ind 3.2.1-3>

Syeciffcatfon 3.2.2:

the APN Setyofnts shall be as follows:

Tri Se oint Q%

Allowable Value 51/

c

+ $5QT c

i

)T 5

c (0.5N + 48K)T 5

c (0.5N + ~)T RS -

r~~/

RS-

~g. Table 3.3.5-2:

the RSH/APRH Control Rod block Setpo{nts shall be as foliowa:

a.

NN - Uyscal ~

Trf Se ofnt W'/o b.

APlN.Flow I{ised Tri Se oint Allowable Value c

~

Vrmllr

+5%

+8'/~

b.

APRN and LPRN>>>>

neutron flux noise levels shall be less than three tfaes their establ fshed basal inc level's when THEHNL )NB fs greater than the lfalt syec{ffed fn Ffgure 3/4.1.1.1-1.

c.

Total core flee shall be greater than or equal to 42 a{11{on lbs/hr when TEAL ONER fs greater than the i{aft specified {n Figure 3,4.1.1.1-1.

'APPLICASILITY:

OPERATIONAL CONITIONS 1>> and 2, except during two loop

~

operatf on. N SUSQUEHANNA UNIT 1 3/4 4-1c peendaent No. 72

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Specification 3.2.3:

The MINIMUM CRITICAL POWER RATIO (MCPR) shall be greater than or equal to the largest of the following values:

a.

1.42, b.

the MCPR determined from Figure 3.2.3-1 plus 0.01, and c.

the MCPR determined from Figure 3.2.3-2 plus 0.01.

i l

3/4. 2 POWER OI STRIBUTIOH LIHITS BASES The specifications of this section assure that the peak cladding tempera-turee following the postulated design basis loss-of-coolant accident will not exceed the 2200'F limit specified in 10 CFR 50.46.

3/4.2. 1 AVERAGE PLANAR LINEAR HEAT GENERATIOH RATE This specification assures, that the, peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limit specified in 10 CFR 50.46.

The peak cladding temperature (PCT) following a postulated loss-of. coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is dependent only secondarily on the rod to rod power distribution within an assembly. M~~~h~eak qu~~~sa-than-the-design-NGR-connoted-fo~enskfieat4on

.-7&~HGR-f4c+Hon-AVERAGE-PBN~EA~~ENHbRRN-RAT&+APH%+~r-GE-4uahd~& s LNGR-~he-h+gh~owered-ro~ided-b r-wn4eh-oesa+t5 han-220~.

The Technical Specification APLHGR for AHF fuel is specified to assure the PCT following a postulated LOCA will not exceed the 2200'F limit.

The limiting value for APLHGR is shown in Figures 3.2. 1-~3.2. 1-2~and-3 ~~

can/

The calculational proc!du used to establish the APLHGR sho~n on Figures 3.2. 1-1, and 3.2. 1-/Vs based on a lass-of-coolant accident analysis.

)

)[

The analysis wa performed using calculational models which are consistent with the requirements of Appendix K to 10 CFR 50..

These models are described in Reference 1'r XH-XF-BO-19, Volumes 2, 2A, 2B and 2C.

3/4. 2. 2 APRH SETPOIHTS The flow biased simulated thermal power upscale scram setting and flow biased simulated thermal power upscale control rod block functions of the APRM instru-ments limit plant operations to the region covered by the transient and accident analyses.

In addition, the APRH setpoints wst be adjusted to ensure that

>1% plastic strain and fuel centerline melting do not occur during the worst anticipated operational occurrence (AOO), including transients initiated from partial power operation.

For AHF fuel the T factor used to adjust the APRN setpoints is based on the FLPD calculated by dividing the actual LNGR by the LNGR obtained from Figure 3. 2.2-1.

The LNGR versus exposure curve in Figure 3.2.2-1 is based on AHF's Protection Against Fuel. Failure (PAFF) line shown in Figure 3.4 of XN-'NF-85-67(A), Revision 1.

Figure 3.2.2-1 corresponds to the ratio of.

PAFF/1.2 under which cladding and fuel integrity is protected during AOOs.

~

~~E-fuel-th~Mactor-used-to-id juste-APRN-setpo4ntM~~n-the-duel,~peci&cation-3-.2~

SUSQUEHANNA UNIT 1 B 3/4 2-1 Amendment Ho. 72

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PO'E'ER 0 ISTR18UTlON LIN1TP 6ASES 3/4.2.3 NlNlNUN CRIT1CAL POSER RAT10 Tht requirtd optrating i{aft NCPRs at sttady state operating conditions as SptC{fitd in SpeC{fiCat{On 3.2.3 art derived frem the tStabliShed fuel Claad>ng integrity Safety Lfeft HCPR, and an analysis of abnore41 aptrational transients.

FOr any abnarsal Operating tranSient analyS{S tValuatian with the initial Can.

dition of the rtactar bting at the steady state operating i{aft, it is required thatWe reSulting NCPR deeS nOt deartaSe belOw tht Safety L{aft NCPR at any tfat during the transient assm{ng fnstneerit trip sttting given in Specification 2.2.

To assure that the futl clad4{ng {nttgrfty Safety Lfa{t is not exceeded 4u<<ng any anticipated abnormal operational transient, the oost 1{aiting tran-sients have been analyzed to dettrefnt which result fn the largest rtduction in CRlTlCAL ~ER RAT10 (CPR).

The type of transients evaluated were loss of flaw, increase fn prtssurt and paver, positive reactivity insertian, and coolant temperature dtcrease.

The 1{a{ting transient yields tht largtst delta NCPR.

Vhtn added to tht Safety Licit NcpR ~ the required Sin{mum optrating liait NCPR of Specification 3.2.3.is obtained and presented in Figurts 3.2.3.1 and 3.2.3.2.

The evaluation of a given transient begins with the systta initial para~

!'t S

Snawn in tht CyCle SptC{f{C tranSitnt analySiS rtpOrt that art input ta~

RA 4+I ~A core dynas{c bthavior-transitnt computer prograa.

Tht outputs of this prograsL along <<ish the initial NcPR farm the input for further analysts of the

" -ally 1{aft{.g bundle.

The codes and aethodology t,". evaluate pressuriza

~{" 'nd non prtssur{zatfon events are described in XH.HF-?g-T1 and N-NF-84.105.

The principal result of this evaluatian is the reduction fn NcPR caused by the transient.

Figure 3. 2.3-1 defines core floe dependent HCPR operating i{sits w{ch o<<urt that.tht Safety L{a{t NCPR v{11 not be exceeded'during a flow incrtast transient resulting froa a abator generator speed cantrol failurt.

The t)av dtptn4ent NCPR iS Only CalCulattd fOr the manual flOw COntrOl aadt.

Thtrtfart, automatic flow control operation is tot pereftted.

Figure 3. 2.3.2 defines the paver 4eptndent NCPR Operating limit Wh{Ch ~SSurtS that the Safety Liait HCPR will not bt txcttded.in the event of a feedvater controller fai lure {nitiattd from a reduced pave~ condition.

Cycle specificanalyses are ptrtoried for tht est lie{ting local.an4 cart v{dt transients to detetafne theraal aargfn.

Additional analyses are perforetd ta determine tht NCPR operating 1{aft vfth tither the Nafn Turbine Sypass in.

OPerablt Or tht EOC-RPT fnaPtrable.

AnalySeS ta determine thermal sarg{n with both the EOC-RPT inoperable aqd Naia Turbine Sypass

{naptrable havt not been pttfomd.

Therefore, operation fn this condition fs not ptraftted.

At THERNAL NMN levels less thea or equal ta 25X of RATKO THERNAL PSrER.

tht reactor v{11 be operating at a{aiawa recirculation puep spee4

.an4 the aodtrator void content will be very saall.

For all 4tsignatt4 control rod patterns which aay be eeploytd at this point, operating plant experience

{nd{-

cates that tht rtsulting HcPR value fs in exctss of rtqu{rtetnts by a consiat~-

able margin.

Ouring {nit'ial start Mp ttstlng of the plant, a

MCPR evaluation SUS)UEHANNA - UNlT 1 8 3/a 2-2 Aeendetnt No.

fC

3/4.4 REACTOR COOLANT SYST'M BASES 3/4.4.1 RECIRCULATION SYSTEM Operation with one reactor recirculation loop inoperable has been evaluated and found acceptable, provided that the unit is operated in accordance with Specification 3,4,1.1.2.

P7

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~d-by-a-Acro

'HpHoation-factor-pedes-ex+ended-opere P2-

~n-w+ah-one-loo~uM~env4oe

$ 0/

For single loop operation, the RBM and APRM setpoints are adjusted by a~

decrease in recirculation drive flow to account for the ac.ive loop drive flow that bypasses the core and goes up through the inactive loop jet pumps.

Surveillance on the pump speed of the operating recirculation loop is imposed to exclude the possibility of excessive reactor vessel internals vibr a-tion.

Surveillance on differential temperatures below the threshold limits on THERMAL POWER or recirculation loop flow mitigates undue thermal stress on vessel nozzles, recirculation pumps and the vessel bottom head during extended operation in the single loop mode.

The threshold limits are those values which vill sweep up the cold vater from the vessel bottom head.

THERMAL POWER, core flow, and neutron flux noise level limitations are pre-scribed in accordance vith the recommendations of General E)ectric Service Information Letter Ho. 380, Revision 1, "BWR Core Thermal Hydraulic Stability,"

dated February 10, 1984, An inoperable jet pump is not, in itself, a sufficient reason to declare a

recirculation loop inoperable, but it does, in case of a design-basis. accident, increase the blowdown area and reduce the capability of reflooding the core;

thus, the requirement for shutdown of the facility with a jet pump inoperable.

Jet pump failure can be detected by monitoring jet pump performance on a prescribed schedule for significant degradation.

Recirculation pump speed mismatch limits are in compliance with the ECCS LOCA analysis design criteria for two loop operation.

The limits will ensure an adequate core flow coastdovn from either recirculation loop following a LOCA.

In the case vhere the mismatch limits cannot be maintained during the loop operation, continued operation is permitted in the single loop mode.

In order to prevent undue stress on the vessel nozzles and bottom head

region, the recirculation loop temperatures shall be within 504F of each other prior to startup of an idle loop.

The loop temperature must also be within 50eF of the reactor pressure vessel coolant temperature to prevent thermal shock to the recir'culation pump and recirculation nozzles.

Since the coolant in the bottom of the vessel is at a lover temperature than the coo?ant in the upper regions of the core, undue stress on the vessel would result ff the tem-perature difference vas greater than 145eF.

SUSQOEHAHNA NIT 1 8 3/4 4-1 Amendaent Ho.

72

+i op

~ ~~'I "a~

~

~ o 4

4

~

F I*

INSERT AZ For single loop operation, the MAPLHGR limits are multiplied by a factor of 1.0 for ANF fuel. This multiplication factor is derived from LOCA analyses initiated from single loop operation conditions.

The resulting MAPLHGR limits for single loop operation assure the peak cladding tem-O perature during a LOCA event remains below 2200 F.

The MINIMUMCRITICAL POWER RATIO (MCPR) limits for single loop op-eration assure that, the Safety Limit MCPR is not exceeded for any An-ticipated Operational Occurrence (AOO) and for the Recirculation Pump Seizure Accident.

I e

y

ML17156B021

Text

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