Text

Amend 45 to License NPF-22,changing Facility Tech Specs to Support Fuel Reload & Cycle 3 Operation
	ML18040B209
Person / Time
Site:	Susquehanna
Issue date:	04/25/1988
From:	Butler W; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML18040A879	List: ML18040A878; ML18040B209;
References
	NUDOCS 8805030488
	Download: ML18040B209 (42)
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UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O. C. 20555 PENNSYLVANIA POWER

& LIGHT COMPANY ALLEGHENY ELECTRIC COOPERATIVE, INC.

DOCKET NO. 50-388 SUSQUEHANNA STEAM ELECTRIC STATION, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 45 License No.

NPF-22 1.

The Nuclear Regulatory Commission (the Commission or the NRC) having found that:

A.

B.

The application for the amendment filed by the Pennsylvania Power Light Company, dated December 23, 1987 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),

and the Commission's regulations set forth in 10 CFR Chapter I; The facility will operate in conformity with the application, the provisions of the Act, and the regulations of the Commission; C.

There is reasonable assurance:

(i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

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

2.

Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment and paragraph 2.C.(2) of the Facility Operating License No.

NPF-22 is hereby amended to read as follows:

(2)

Technical S ecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 45 and the Environmental Protection Plan con-tained in Appendix 8, are hereby incorporated in the license.

PP&L shall operate the facility in accordance with the Technical Specifica-tions and the Environmental Protection Plan.

8805030488 880425 PDR ADOCK 05000388'

3.

This license amendment is effective prior to startup for Cycle 3

operation.

FOR THE NUCLEAR REGULATORY COMMISSION

/s/

Attachment:

Changes to the Technical Specifications Date of Issuance:

April 251 1988 Walter R. Butler, Director Project Directorate I-2 Division of'eactor Projects I/II c>

N)"hr'en I/>>$ 88 PDI-2/PM NTh@dani:mr V/4 /88 fGC),/

/88 PDI-2/D WButler

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This license amendment is effective prior to startup for Cycle 3

operation.

FOR TME NUCLEAR REGULATORY COMMISSION

Attachment:

Changes to the Technical Specifications Date of Issuance:

April 25, 1988 Walter R. Butler, Director Project Directorate I-2 Division of Reactor Projects I/II

ATTACHMENT TO LICENSE AMENDMENT NO. 45 FACILITY OPERATING'LICENSE NO. NPF-22 DOCKET NO. 50-388 Replace the following pages of the Appendix A Technical Specifications with enclosed pages.

The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.

The overleaf pages are provided to maintain document completeness.*

REMOVE ill 1V Xl Xli Xxl Xxl 1 8 2-1 8 2-2 3/4 2-1 3/4 2-2 3/4 2-3 3/4 2-4 3/4 2-5 3/4 2-6 3/4 2-6a 3/4 2-7 3/4 2-8 3/4 2-9 3/4 2-10 3/4 2-10a 3/4 2-10b 3/4 3-53 3/4 3-54 3/4 4-1 3/4 4-la 3/4 4-1b 3/4 4-1c INSERT iii*

lV X1*

Xii XX1*

Xxll 8 2-1 8 2-2 3/4 2-1 3/4 2-2*

3/4 2-3 3/4 2-4 3/4 2-5 3/4 2-6*

3/4 2-6a 3/4 2-7*

3/4 2-8 3/4 2-9 3/4 2-10*

3/4 2-10a 3/4 2-10b 3/4 3-53*

3/4 3-54 3/4 4-1 3/4 4-la*

3/4 4-1b 3/4 4-lc

FACILITY OPERATING LICENSE NO.

NPF-22 DOCKET NO. 50-388 REMOVE 3/4 4-ld 3/4 4-le 3/4 4-1$

3/4 4-2 INSERT 3/4 4-ld 3/4 4-le*

3/4 4-H'*

3/4 4-lg 3/4 4-2*

B 3/4 2-1 B 3/4 2-2 B 3/4 4-1 B 3/4 4-2 B 3/4 2-1 B 3/4 2-2 B 3/4 4-1 B 3/4 4-2*

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INDEX SAFETY'LIMITS ANO LIMITING SAFETY SYSTEM SETTINGS SECTION PAGE

2. 1 SAFETY LIMITS THERMAL POWER, Low Pressure or Low Flow...................

2-1 THERMAL POWER, High Pressure and High Flow................

2-1 Reactor Coolant System Pressure...

~.......

Reactor Vessel Water Level..............

~

2 1

~

2 2 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Protection System Instrumentation Setpoints.......

2-3 t

BASES

2. 1 SAFETY LIMITS THERMAL POWER, Low Pressure or Low Flow...................

8 2-1 THERMAL POWER, High Pressure and High Flow.......

B 2-2 Reactor Coolant System Pressure........................

B 2-5 Reactor Vessel. Water Level.....

B 2-5 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Protection System Instrumentation Setpoints........

B 2-6 SUSQUEHANNA - UNIT 2

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION 3/4. 0 APPLICABILITY.

3/4.1 REACTIVITY CONTROL SYSTEMS 3/4. l. 1 SHUTDOWN MARGIN.

PAGE

~

3/4 O-l 3/4 1-1 3/4. 1. 2 REACTIVITY ANOMALIES...... ~...

3/4. 1.3 CONTROL RODS Control Rod Oper ability..........

3/4 1-2 3/4 1"3 Control Rod Maximum Scram Insertion Times..............

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

Four Control Rod Group Scram Insertion Times...

Control Rod Scram Accumulators..

Control Rod Drive Coupling.....,

Control Rod Position Indication.....

Control Rod Drive Housing Support......

3/4. 1.4 CONTROL ROD PROGRAM CONTROLS Rod Worth Minimizer.

Rod Sequence Control System.......

Rod Block Monitor.

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

3/4. 2 POWER DISTRIBUTION LIMITS 3/4 1-7 3/4 1-8 3/4 1"9 3/4 l-ll 3/4 1-13 3/4 1-15 3/4 1-16 3/4 1-17 3/4 1-18 3/4 1"19 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE~............

3/4 2-1 3/4 2.2 APRM SETPOINTS 3/4. 2. 3 MINIMUM CRITICAL POWER RATIO.

3/4 2-5 3/4 2-6 3/4. 2. 4 LINEAR HEAT GENERATION RATE.......

GE FUELe

~

ANF FUEL...........

3/4 2-10 3/4 2-10 3/4 2-10a SUSQUEHANNA - UNIT 2 iv Amendment No. 45 f

INDEX LIMITING CONDITIONS FOR OPERAT'ION AND SURVEILLANCE RE UIREMENTS SECTION 3/4. 11 RADIOACTIVE EFFLUENTS 3/4. 11. 1 LIQUID EFFLUENTS Concentration....

ose..............,..

0 Liquid Waste Treatment System.

Liquid Holdup Tanks.............

3/4. 11.2 GASEOUS EFFLUENTS Dose Rate Dose-Noble Gases.

Dose Iodine-131, Tritium and Radionucl Particulate Form....,....

ides in Gaseous Radwaste Treatment System....................

Ventilation Exhaust Treatment System.......

Explosive Gas Mixture............

Main Condenser...........

Venting or Purging......

~........

3/4. 11.3 SOLID RADWASTE SYSTEM......

3/4.11.4 TOTAL DOSE.........,............

3/4. 12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4. 12. 1 MONITORING'ROGRAM...........

3/4. 12.2 LAND USE CENSUS.....................

PAGE 3/4 11-1 3/4 11-6 3/4 11-7 3/4 11-8 3/4 11-9 3/4 11-13 3/4 11-14 3/4 11-15 3/4 11-16 3/4 11-17 3/4 11-18 3/4 11-19 3/4 11-20 3/4 11-22 3/4 12-1 3/4 12-13 3/4. 12. 3 INTERLABORATORY COMPARISON PROGRAM..............,.....

3/4 12"14 SUSQUEHANNA - UNIT 2 xi

BASES INDEX SECTION 3/4. 0 APPLICABILITY 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1. 1 SHUTDOWN MARGIN.

3/4.1.2 REACTIVITY ANOMALIES.

3/4. 1.3 CONTROL RODS PAGE B 3/4 0"1 B 3/4 1-1 B 3/4 1-1 B 3/4 1-2 3/4. 1.4 CONTROL ROD PROGRAM CONTROLS B 3/4 1-3 3/4. 1.5 STANDBY LIQUID CONTROL SYSTEM.....,.............

B 3/4 1"4 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE 3/4.2.2 APRM SETPOINTS...............

3/4. 2. 3 MINIMUM CRITICAL POWER RATIO.

3/4.2.4 LINEAR HEAT GENERATION RATE..

3/4.3 INSTRUMENTATION B 3/4 2-1 8 3/4 2-1 8 3/4 2-2 B 3/4 2-3 3/4.3. 1 REACTOR PROTECTION SYSTEM INSTRUMENTATION........

B 3/4 3"1 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION..............

B 3/4 3-2 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION...

B 3/4 3-2 3/4. 3. 4 3/4.3.5 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION......

REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION................

B 3/4 3"3 B 3/4 3-4 3/4.3'.6 CONTROL.ROD BLOCK INSTRUMENTATION...............

B 3/4 3-4'SUSQUEHANNA

" UNIT 2 X11 Amendment No. 45

INDEX ADMINISTRATIVE CONTROLS 6.13 PROCESS CONTROL PROGRAM..........

~.......

6-23

6. 14 OFFSITE DOSE CALCULATIONMANUAL..'........................

6-24 6.15 MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS.....

6-24 SUSQUEHANNA - UNIT 2 XX1

LIST OF FIGURES INDEX FIGURE

3. l. 5-1 SODIUM PENTABORATE SOLUTION TEMPERATURE/

CONCENTRATION REQUIREMENTS PAGE 3/4 1-21

3. l. 5-2 3.2.1 1

MAXIMUMAVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VS.

AVERAGE PLANAR EXPOSURE, GE FUEL TYPE 8CR183 (1.83K ENRICHED) 3/4 2-2 SODIUM PENTABORATE SOLUTION CONCENTRATION.........

3/4 1-22

3. 2. 1-2 3 ~ 2 1 3 3.2. 2-1 3 ~ 2 ~ 3 1

3. 2. 3-2 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VS.

AVERAGE PLANAR EXPOSURE, GE FUEL TYPE 8CR233 (2.33X ENRICHED)

MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VS.

AVERAGE PLANAR EXPOSURE, ANF 9 X 9 FUEL LINEAR HEAT GENERATION RATE FOR APRM SETPOINTS VERSUS AVERAGE PLANAR EXPOSURE, ANF FUEL...........

FLOW DEPENDENT MCPR OPERATING LIMIT REDUCED POWER MCPR OPERATING LIMIT...

3/4 2-3 3/4 2-4 3/a 2-6a 3/4 2-8 3/4 2"9 3.2.4.2-1 LINEAR HEAT GENERATION RATE (LHGR) LIMIT VERSUS AVERAGE PLANAR EXPOSURE, ANF 9 X 9 FUEL............

3.4. l.l. 1-1 THERMAL POWER/CORE FLOW LIMITATIONS

3. 4. 1. 1. 2-1 SINGLE LOOP OPERATION THERMAL POWER LIMITATIONS....

3/6 2-10b i 3/4 4"lb 3/6 a-1g

3. 2. 6. 1-1 MINIMUM REACTOR VESSEL METAL TEMPERATURE VS.

REACTOR VESSEL PRESSURE 3/4 4-18

4. 7. 4"1 B 3/4 3-1 B 3/4.4.6"1 REACTOR VESSEL WATER LEVEL.

~

FAST NEUTRON FLUENCE (E>lMeV) AT 1/4 T AS A

FUNCTION OF SERVICE LIFE,..................

B 3/4 3"8 B 3/4 4-7 SAMPLE PLAN 2)

FOR SNUBBER FUNCTIONAL TEST.........

3/4 7-15

5. l. 1-1

5. 1. 2-1

5. 1. 3-la 5.1.3-1b SUSQUEHANNA EXCLUSION AREA................

LOW POPULATION ZONE MAP DEFINING UNRESTRICTED AREAS FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS.

MAP DEFINING UNRESTRICTED AREAS FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS

- UNIT 2 XX11 5-2 5"3 5-4 5-5 Amendment No.

45

2. 1 SAFETY 'LIMITS BASES

2. 0 INTRODUCTION The fuel cladding, reactor pressure vessel and primary system piping are the principal barriers to the release of radioactive materials to the environs'afety Limits are established to protect the integrity of these barriers during normal plant operations and anticipated transients.

The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated.

Because fuel damage is not directly observable, a step-back approach is used to establish a Safety Limit such that the MCPR is not less than the limit specified in Specification

2. 1.2 for both GE and ANF fuel.

MCPR greater than the specified limit represents a conser-vative margin relative to the conditions required to maintain fuel cladding integrity.

The fuel cladding is one of the physical barriers which separate the radioactive materials from the environs.

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

Al-though some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable.

Fuel cladding perforations,

however, can result from thermal stresses which occur from reactor operation significantly above design conditions and the Limiting Safety System Settings.

While fission pro-duct 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 incre-mental 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 condition intended by design for planned operation.

The MCPR fuel cladding integrity Safety limit assures that during normal operation and during antici-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. l. 1 THERMAL POWER Low Pressure or Low Flow The use of the XN-3 correlation is valid for critical power calculations at pressures greater than 580 psig and bundle mass fluxes greater than 0.25 x 10 lbs/hr-ft For operation at low pressures or low flows, the fuel cladding integrity Safety Limit is established by a limiting condition on core 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 is sufficient to assure a

minimum bundle flow for all fuel assemblies which have a relatively high power and potentially can approach a critical heat flux condition.

For the ANF 9 x 9 fuel design, the minimum bundle flow is greater than 30,000 lbs/hr.

For the ANF and GE 8 x 8 fuel, the minimum bundle flow is greater than 28,000 lbs/hr.

For all designs, the coolant minimum flow and maximum flow area is such that the mass flux is always greater than 0.25 x 10 lbs/hr-ft Full scale critical power tests taken at pressures down to 14.7 psia indicate that the fuel assembly critical power at 0.25 x 10 lbs/hr-ft is 3.35 Mwt or greater.

At 25K thermal power a bundle power of 3.35 Mwt corresponds to a bundle radial peaking factor of greater than 3.0 which is significantly higher than the expected peaking factor.

Thus, a THERMAL POWER limit of 25K of RATED THERMAL POWER for reactor pressures below 785 psig is conservative.

SUS(UEHANNA - UNIT 2 B 2-1 Amendment No.

45

SAFETY LIMITS BASES

2. 1.2 THERMAL POWER Hi h Pressure and Hi h Flow Onset of transition boiling results in a decrease in heat transfer from the clad and, therefore, elevated clad temperature and the possibility. of clad failure.

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

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

The margin for each fuel assembly is characterized by the critical power ratio (CPR), which is the ratio of the bundle power which would produce onset of tran-sition boi ling 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 MCPR assures sufficient conservatism in the operating MCPR limit that in the event of an anticipated operational occurrence from the limiting condition for operation, at least 99.9X of the fuel rods in the core would be expected to avoid boiling transition.

The margin between calculated boiling transition (MCPR = 1.00) and the Safety Limit MCPR is based on a de-tailed statistical procedure which considers the uncertainties in monitoring the core operating state.

One specific uncertainty included in the safety limit is the uncertainty inherent in the XN-3 critical power correlation.

XN-NF-524 describes the methodology used in determining the Safety Limit MCPR.

The XN-3 critical power correlation is based on a significant body of prac-tical test data, providing a high degree of assurance that the critical power as evaluated by the correlation is within a small percentage of the actual criti-cal power being estimated.

As long as the core pressure and flow are within the range of validity of the XN-3 correlation (refer to Section B 2. 1. 1), the assumed reactor conditions used in defining the safety limit introduce conser-vatism into the limit because bounding high radial power factors and bounding flat local peaking distributions are used to estimate the number of rods in boiling transition.

Still further conservatism is induced by the tendency 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 degree of assurance that during sustained operation at the Safety Limit MCPR there would be no transition boiling in the core.

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

Significant test data accumulated by the U.S.

Nuclear Regulatory 'Commission and private organizations indicate that the use of a boiling transition limitation to protect against cladding fai lure is a very conservative approach.

Much of the data indicates that LWR fuel can sur-vive for an extended period of time in an environment of boiling transition.

SUSQUEHANNA - UNIT 2 B 2-2 Amendment No.

45

3/4. 2 POWER DISTRIBUTION LIMITS

, I aI)I 3/4. 2. 1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2. 1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for each type of fuel as a function of AVERAGE PLANAR EXPOSURE for GE fuel and AVERAGE BUNDLE EXPOSURE for ANF fuel shall not exceed the limits shown in Figures 3.2. 1-1, 3.2.1-2, and 3.2.1-3."

APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or DADE I RATED TIIERIIAI I ER.

ACTION:

With an APLHGR exceeding the limits of Figure 3.2.1-1, 3.2.1-2, or 3.2.1-3, initiate corrective action within 15 minutes and restore APLHGR to within the required limits 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 25K of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE RE UIREMENTS 4.2. 1 All APLHGRs shall be verified to be equal to or less than the limits determined from Figures 3.2. 1-1, 3.2. 1-2, and 3.2. 1-.3:

a.

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

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 15K of RATED THERMAL POWER, and c.

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 with a LIMITING CONTROL ROD PATTERN for APLHGR.

d.

The provisions of Specification 4.0.4 are not applicable.

"See Specification 3.4.1.1.2.a for single loop operation requirements.

SUSQUEHANNA - UNIT 2 3/4 2-1 Amendment No. 4~

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PERMISSABLE REGION OF OPERATION O

60OO

. 10000 16000 20OOO 26DOO 300OO 36000 Average Planar Exposure.(MWD/MT)

MAXIMUMAVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE GE FUEL TYPES 8CR183 ('l.83% ENRICHEO)

FIGURE 3.2.1-1

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33,069; 11.2 9-0 5000 10000 15000 20000 25000 30000 35000 40000 45000 Average Planar Exposure (MWD/MT)

MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (HAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE GE FUEL TYPES 8CR233 (2.33K ENRICHED)

FIGURE 3.2.1-2

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O MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (HAPLHGR) VERSUS AVERAGE BUNDLE EXPOSURE ANF 9X9 FUEL FIGURE 3.2.1-3

POWER DISTRIBUTION LIMITS 3/4.2.2 APRM SETPOINTS LIMITING CONDITION FOR OPERATION 3'.2 The APRM flow biased simulated thermal power-upscale scram trip setpoint (S) and flow biased neutron flux-upscale control rod block trip setpoint (SRB) shall be established according to the following relationships:

Tri Set oint

~ ¹ Allowable Value S

W + 59K)T RIT SRB

< (0.58W + 50K)T SRB

< (0.58W + 53K)T where:

S and S

B are in percent of RATED THERMAL POWER, W = LooPrecirculation flow as a percentage of the loop recirculation flow which produces a rated core flow of 100 million lbs/hr, T = Lowest value of the ratio of FRACTION OF RATED THERMAL POWER divided by the MAXIMUM FRACTION OF LIMITING POWER DENSITY.

Where:

b.

The FRACTION OF LIMITING POWER DENSITY (FLPD) for GE fuel is the actual LINEAR HEAT GENERATION RATE (LHGR) divided by 13.4 per Specification 3.2.4.1, and The FLPD for ANF fuel is the actual LHGR divided by the LINEAR HEAT GENERATION RATE from Figur e 3.2. 2-1.

T is always less than or equal to 1,0.

APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or f RATER TIEERIIAI R

ER.

ACTION:

With the APRM flow biased simulated thermal power-upscale scram trip setpoint and/or the flow oiased neutron flux-upscale control rod block trip setpoint less conservative than the value shown in the Allowable Value column for S or S B, as above determined, initiate corrective action within 15 minutes and adjust 3 and/or SRB to be consistent with the Trip Setpoint 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 25K of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE RE UIREMENTS 4.2.2 The FRTP and the MFLPD shall be determined, the value of T calculated, and the most recent actual APRM flow biased simulated thermal power-upscale scram and flow biased neutron flux-upscale control rod block trip setpoints verified to be within the above limits or adjusted, as required:

"With MFLPD greater than the FRTP during power ascension up to 90X of RATED THERMAL POWER, rather than adjusting the APRM setpoints, the APRM gain may be adjusted such that'PRM readings are greater than or equal to 100K times MFLPD, provided that the adjusted APRM reading does not exceed 100K of RATED THERMAL POWER, the required gain adjustment increment does not exceed lOX of RATED THERMAL POWER, and a notice of the adjustment is posted on the reactor control panel.

See Specification 3.4. 1. 1.2.a for single loop oper ation requirements.

SUS(UEHANNA - UNIT 2 3/4 2-5 Amendment No. 45

POWER OISTRIBUTION LIMITS 3/4.2.2 APRM SETPOINTS LIMITING CONDITION FOR OPERATION 4.2.2 aO b.

C.

d.

(Continued)

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 15K of RATED 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 with HFLPD greater than or equal to FRTP.

The provisions of Specification 4.0.4 are not applicable.

SUSQUEHANNA - UNIT 2 3/4 2-6 Amendment No.

31 cj) trodi vQ.<~>

+cUXi<uulp.~ Lu& ol

AD 18

~

0.0;.------::"-::-.

16.0 O

IZg CM 0

CO g) 14

~'o

0) ~

C ~

(3 u 12 eW K CC CL co Q 6) 10

.5 o U LL 0

25,400;:

14.0

.. 43,200; S.O 48,000;:;,.

0 50000 10000 20000 30000 40000 Average Planar Exposure (MWD/MT)

LINEAR HEAT GENERATION RATE FOR APRH SETPOINTS VERSUS AVERAGE PLANAR EXPOSURE ANF FUEL FIGURE 3.2.2-1

POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION 3.2.3 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be greater than or equal to the greater of the two values determined from Figure 3.2.3-1 and Figure

3. 2. 3-2.

APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMA POWER is greater than or T

THE EE ACTION:

With MCPR less than the applicable MCPR limit determined above, initiate correc-tive action within 15 minutes and restore MCPR to within the required limit with-in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 25X of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE RE UIREMENTS 4.2.3.1 MCPR shall be determined to be greater than or equal to the applicable MCPR limit determined from Figure 3.2.3-1 and Figure 3.2.3-2:

a.

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

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 15'f RATED THERMAL POWER, and c.

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 operatin'g with a LIMITING CONTROL ROD PATTERN for MCPR.

d.

The provisions of Specification 4.0.4 are not applicable.

SUSQUEHANNA - UNIT 2 3/4 2-7 Amendment No. 3l C~ L~u'~> '~~'I

~ECLLG.CLCrig': CA( ILIA P~

I<$ 'll

1.7 1.6 (40,1.B1)

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

~~

CQ CL 1.4 CC 0O 1.3 (60,1.44) +

(60.77.1.43)

(67 B9,1.34)

(69.23,1.32)

A B

C 1.43 1.34 1.32 1.2 40 60 60 70 80 Total Core Flow (/ OF RATED)

FLOW DEPENDENT MCPR OPERATING LIMIT FIGURE 3.2.3;1 90 100

1.7 (26,1.62)

(26,1.44)

(40,1.60)

(40,1.42)

CURVE A: EOC-RPT Inoperable:

Main Turbine Bypass Operable CURVE B: EOC-RPT Operable: Main Turbine Bypass Inoperable CURVE C: EOC-RPT and Main Turbine Bypass Operable (66,1.47)

(80,1.44)

(66,1. 39) 1.42 (25,1.39) 40,1. 37

(

)

(65,1.34)

(80,1 36)

(76,1.32)

C 1.32 20 30 40 60 60 70 Core Power (/o OF RATED)

REOUCED POMER HCPR OPERATING LIHIT Figure 3. 2. 3;2 80 90 100

POWER DISTRIBUTION LIMIT 3/4.2.4 LINEAR HEAT GENERATION RATE GE FUEL LIMITING CONDITION FOR OPERATION

~

3.2.4.1 The LINEAR HEAT GENERATION RATE (LHGR) for GE fuel shall not exceed 13.4 kw/ft.

APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or.

f RAT TIIEINL Pil" ACTION:

With the LHGR of any fuel rod exceeding the limit, 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 than 25X of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE RE UIREMENTS 4.2.4. 1 LHGRs for GE fuel shall be determined to be equal to or less than the limit:

a.

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

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 RATED THERMAL POWER, and c.

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 LIMITING CONTROL ROD PATTERN for LHGR.

d.

The provisions of Specification 4.0.4 are not applicable.

SUSQUEHANNA - UNIT 2 3/4 2-10 Amendment No. 31

~r) N~

~Mcg-

POWER DISTRIBUTION LIMITS 3/4.2.4 LINEAR HEAT GENERATION RATE ANF FUEL LIMITING CONDITION FOR OPERATION 3.2.4.2 The LINEAR HEAT GENERATION RATE (LHGR) for ANF fuel shall not exceed the LHGR limit determined from Figure 3.2.4.2-1.

APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or ATED E MAL Ell.

ACTION:

With the LHGR of any fuel rod exceeding the limit, 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 than 25K of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE RE UIREMENTS

4. 2. 4. 2 LHGRs for ANF fuel shall be determined to be equal to or less than the limit:

a.

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 15K of RATED 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 LIMITING CONTROL ROD PATTERN for LHGR.

d.

The provisions of Specification 4.0.4 are not applicable.

SUSQUEHANNA " UNIT 2 3/4 2-10a Amendment No.

45

CA AD m

E lg lK 0

CQ 1B 14 10 C

C9 X

L CQ tD

.. 0.0;.;

13.0

. ~ ~ ~ ~

~

~.'

~

~ -

i

..:..:....:.... 24,000;.

12.0

~ ~ ~

35,000;

~

1

~

9.5

. PERMISSABLE:.

REGION OF OPERATION 48,000; 772

~

0 10000 20000 30000 40000 Average Planar Exposure (MWD/MT) 50000 O

LINEAR HEAT GENERATION RATE (LHGR) LIMIT VERSUS AVERAGE PLANAR EXPOSURE ANF 9X9 FUEL FIGURE 3.2.4.2"1

I

~ ~

TABLE 3. 3.6-1 (Continued)

CONTROL ROD BLOCK INSTRUMENTATION ACTION ACTION 60 Declare the RBM inoperable and take the ACTION required by Specification 3.1.4.3.

ACTION 61 With the number of OPERABLE Channels:

aO One less than required by the Minimum OPERABLE Channels per Trip Function requirement, restore the inoperable channel to OPERABLE status within 7 days or place the inoperable channel in the tripped condition within the next hour.

b.

Two or more less than required by the Minimum OPERABLE Channels per Trip Function requirement, place at least one inoperable channel in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 62 With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, 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 .

NOTES With THERMAL POWER

> 30K of RATED THERMAL POWER.

With more than one control rod withdrawn.

Not applicable to control rods removed per Specification 3.9. 10. 1 or 3.9.10.2.

Not required when eight of fewer fuel assemblies (adjacent to the SRMs )

are in the core.

(a)

The RBM shall be automatically bypassed when a peripheral control rod is selected or the reference APRM channel indicates less than 30K of RATED THERMAL POWER.

(b)

This function shall be automatically bypassed if detector count rate is

> 100 cps or the IRM channels are on range 3 or higher.

(c)

This function is automatically bypassed when the associated IRM channels are on range 8 or higher.

(d)

This function is automatically bypassed when the IRM channels are on range 3 or higher.

(e)

This function is automatically bypassed when the IRM channels are on range l.

SUSQUEHANNA - UNIT 2 3/4 3-53 Amendment No.16

g TRIP FUNCTION TABLE 3.3.6-2 CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS TRIP SETPOINT ALLOWABLE VALUE m

I 3.

ROD BLOCK MONITOR a.

Upscale¹¹ b.

Inoperative c.

Downscale APRM a.

Flow Biased Neutron Flux - Upscale¹¹ b.

Inoperative'.

Downscale d.

Neutron Flux - Upscale Startup SOURCE RANGE MONITORS 0.66 M + 42X NA

> 5/125 divisions of full scale

< 0.58 M+ 50K*

NA

> 5X of RATED THERMAL POWER

< 12X of RATED THERMAL POWER

< 0.66 M + 45K NA

> 3/125 of divisions full scale 1

< 0.58 M+ 53K*

NA

> 3X of RATED THERMAL POWER

< 14X of RATED THERMAL POWER 4J I

4.

a.

Detector not full in b.

Upscale c.

Inoperative d.

Downscale INTERHEDIATE RANGE MONITORS NA<2x10 cps HA

> 0.7 cps""

NA

< 4 x 10 cps NA) 0 5 cps%*

a.

b.

C.

d.

Detector not full in Upscale Inoperative Downscale NA NA

< 108/125 divisions of full scale

< 110/125 divisions of full scale NA NA

> 5/125 divisions of full scale

> 3/125 divisions of full scale 5.

SCRAM DISCHARGE VOLUME a.

Mater Level - High

< 44 gallons REACTOR COOLANT SYSTEM RECIRCULATION FLOM

< 44 gallons O

a.

Upscale

< 108/125 divisions of full scale

< ill/125 divisions of full scale b.

Inoperative HA NA c.

Comparator

< 10X flow deviation

< 11K flow deviation k

k k<<k kk (M).

The trip setting of this function must be maintained in accordance with Specification 3.2.2.

"*Provided signal-to-noise ratio is

> 2.-

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

¹¹See Specification

3. 4. 1. 1. 2. a for single loop operation requirements.

I 3/4.4 REACT'OR COOLANT SYSTEM 3/4. 4. 1 RECIRCULATION SYSTEM RECIRCULATION LOOPS -

TWO LOOP OPERATION LIMITING CONDITION FOR OPERATION 3.4. 1.1.1 Two reactor coolant system recirculation loops shall be in operation and:

a.

Total core flow shall be greater than or equal to 55 million lbs/hr, or b.

The reactor is at a THERMAL POWER/core flow condition less than or equal to the limit specified in Figure 3.4. 1. 1. 1-1.

APPLICABILITY:

OPERATIONAL CONDITIONS 1" and 2", except during single loop operati on. ¹ ACTION:

a ~

b.

C.

With one reactor coolant system recirculation loop not in operation, comply with the requirements of Specification 3.4. 1. 1.2, or take the associated ACTION.

With no reactor coolant system recirculation loops in operation, immediately initiate an orderly reduction of THERMAL POWER to less than or equal to the limit specified in Figure 3.4. 1. 1. 1-1, and initiate.

measures to place the unit in at least STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

With two reactor coolant system recirculation loops in operation and total core flow less than 55 million lbs/hr and the reactor at a THERMAL POWER/core flow condition greater than the limit specified in Figure 3.4.1.1.1-1:

1.

Restore the reactor to a THERMAL POWER/core flow condition less than or equal to the limit specified in Figure 3.4. 1. 1. 1-1, or 2.

Increase core flow to greater than 55 million lbs/hr, or 3.

Determine the APRM and LPRM""" neutron flux noise levels within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and:

a)

If the APRM and LPRM""* neutron flux noise levels are less than three times their established baseline levels, continue to determine the noise levels at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and within 30 minutes after the completion of a THERMAL POWER increase of at least 5X of RATED THERMAL POWER, or b)

If the APRM or LPRM""" neutron flux noise levels are greater than or equal to three times their established baseline

levels, immediately initiate corrective action and restore the noise levels to within the required limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months by increasing core flow to greater than 55 million lbs/hr, and/or by returning the reactor to a THERMAL POWER/core flow condition less than or. equal to the limit specified in Figure 3.4.1.1.1-1.

"See Special Test Exception

3. 10.4.

""~Detectors A and C of one LPRM string per core octant plus detectors A and C

of one LPRM string in the center of the core should be monitored.

¹See Specification 3.4. 1. 1.2 for single loop operation requirements.

SUSQUEHANNA - UNIT 2 3/4 4-1 Amendment No. 45

REACTOR COOLANT SYSTEM SURVEILLANCE RE UIREMENTS I

~

4.4. 1. 1. l. 1 Each pump discharge valve and bypass valve shall be demonstrated OPERABLE by cycling each valve through at least one complete cycle of full travel during each startup"* prior to THERMAL POWER exceeding 25K of RATED THERMAL POWER.

4.4. l. 1. 1.2 Each pump discharge bypass valve, if not OPERABLE, shall be verified to be closed at least once per 31 days.

4.4. 1.1.1.3 Each pump MG set scoop tube electrical and mechanical stop shall be demonstrated OPERABLE with overspeed setpoints less than or equal to 102.5 and 105K, respectively, of rated core flow, at least once per 18 months.

4.4. 1. 1. 1.4 Establish a baseline APRM and LPRM neutron flux noise value at a point within 5X RATED THERMAL POWER of the 100K rated rod line with total core flow between 35K and 50K of rated total core flow during startup testing following each refueling outage.

""Ifnot performed within the previous 31 days.

SUSQUEHANNA - UNIT 2 3/4 4-la

80 70 5 col g

60 u

40 q""

30 20 10

~ 1

~

~ t ~

II

~

0

\\

REGION LESS THAN LIMIT

. REGION GREATER -.;"""""

THAN-LIMIT

'I I

4

~ 8 1 ~

~ ~

0 20 30 40 60 60 Core Flow (% RATED) 70 80 Figure 3.4.1.1. 1-1 THERMAL POWER/CORE FLOW LIMITATIONS SUSQUEHANNA - UNIT 2 3/4, 4-1b Amendment No. 4g

REACTOR COOLANT SYSTEM RECIRCULATION LOOPS " SINGLE LOOP OPERATION LIMITING CONDITION FOR OPERATION 3.4. 1. 1.2 One reactor coolant recirculation loop shall be in operation with the pump speed

< 80K of the rated pump speed, and a.

the following revised specification limits shall be followed:

1.

Specification

2. 1.2:

the MCPR Safety Limit shall be increased to 1.07.

2.

Table 2.2. 1-1:

the APRM Flow-Biased Scram Trip Setpoints shall be as follows:

3.

5.

Tri Set oint Allowable Value

+

Specification 3.2. 1:

The MAPLHGR limits shall be the limits specified in Figures 3.2. 1-1 and 3.2. 1-2 multiplied by 0.81 and Figure 3.2. 1-3 multiplied by 1.0.

Specification 3.2.2:

the APRM Setpoints shall be as follows:

Tri Set oint Allowabl e Value 555)T

~55)T SRB

< (0.58W + 46K)T SRB

< (0.58W + 49X)T Specification 3.2.3:

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

a 137 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.

6.

Table 3.3

~ 6-2:

the RBM/APRM Control Rod Block Setpoints shall be as fol 1 ows:

a.

RBM - Upscale Tri Set oint

+

Allowable Value 6W+

b.

APRM-Flow Biased Tri Set oint Al 1 owabl e Value

+ 46 b.

APRM and LPRM""" neutron flux noise levels shall be less than three times their established baseline levels when THERMAL POWER is greater than the limit specified in Figure 3/4. 1. 1.2-1.

c.

Total core flow shall be greater than or equal to 42 million lbs/hr when THERMAL POWER is greater than the limit specified in Figure 3.4.1. 1.2-1.

APPLICABILITY:

OPERATIONAL CONDITIONS 1* and 2", except during two loop operation.0 ACTION:

a.

With no reactor coolant system recirculation loops in operation, take the ACTION required by Specification 3.4. 1. 1. 1.

SUSQUEHANNA - UNIT 2 3/4 4-1c Amendment No. 45

REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION Continued b.

C.

d.

e.

With any of the limits specified in 3/4.1.1.2a not satisfied:

l.

Upon entering single loop operation, comply with the new limits within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

2.

If the provisions of ACTION b. 1 do not apply, take the ACTION(s) required by the referenced Specification(s).

With the APRM or LPRM""" neutron flux noise levels greater than or equal to three times their established baseline levels when THERMAL POWER is greater than the limit specified in Fig-ure 3.4. 1. 1.2-1, immediately initiate corrective action and restore the noise levels to within the, required limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months by initiating an orderly reduction of THERMA POWER to less than or equal to the limit specified in Figure 3.4.1.1.2-1.

)

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

With one or more jet pumps inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

With total core flow less than 42 million lbs/hr when THERMAL POWER is greater than the limit specified in Figure 3.4. 1. 1.2-1, immediately initiate corrective action by either:

1.

Reducing THERMAL POWER to less than or equal to the limit specified in Figure 3.4.1. 1.2-1 within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , or 2.

Increasing total core flow to greater than or equal to 42 million lbs/hr within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE RE UIREMENTS

4. 4. l. 1. 2. 1

4. 4. 1. 1.2. 2

, 4. 4. 1. 1. 2. 3, Upon entering single loop operation and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter, verify that the pump speed in the operating loop is

< 80K of the rated pump speed.

l With THERMAL POWER greater than the limit specified in Fig-ure 3.4.1.1.2-1, determine the APRFi, and LPRN*** neutron flux noise levels within 1hour.

Continue to determine the noise levels at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and within 30 minutes after the completion of the THERMAL POWER increase

> 5X of RATED THERMAL POWER.

Within 15 minutes prior to either THERMAL POWER increase resulting from a control rod withdrawal or recirculation loop flow increase, verify that the following differential temperature requirements are met if THERMAL POWER is

< 30K""*"of RATED THERMAL POWER or the recirculation loop fTow in the operating r'ecirculation loop is

< 50K**","of rated loop flow:

SUSQUEHANNA UNIT 2 3/4 4-1d Amendment No.45

REACTOR COOLANT SYSTEM SURVEILLANCE RE UIREMENTS Continued 4.4.1.1.2.4 4.4. 1. 1. 2. 5 4.4. 1. 1. 2. 6

4. 4. 1. 1. 2. 7

4. 4.1. 1. 2. 8 4.4. 1.1. 2. 9 a.

< 145'F between reactor vessel steam space coolant and bottom head drain line coolant, b.¹¹

< 504F between the reactor coolant within the loop not in operation and the coolant in the reactor pressure

vessel, and c.¹¹

< 504F between the reactor coolant within the loop not in operation and operating loop.

a.

Establish a baseline APRM and LPRM neutron flux noise value at a point within 5X RATED THERMAL POWER of the 100K rated rod line with total core flow between 35'nd 50K of rated total core flow during startup testing following each refueling outage, or b.

In lieu of establishing a single loop operation baseline value, utilize. the value established pursuant to Specification 4.4.1.1. 1.4 if a baseline value is needed to meet the requirements of Specification 3.4. 1. 1.2.

The pump discharge valve and bypass valve in both loops shall be demonstrated OPERABLE by cycling each valve through at least one complete cycle of full travel during each startup"" prior to THERMAL POWER exceeding 25K of RATED THERMAL POWER.

The pump discharge bypass valve in the OPERABLE loop, if not OPERABLE, shall be verified to be closed at least once per 31 days.

The pump MG set scoop tube electrical and mechanical stop shall be demonstrated OPERABLE with overspeed setpoints less than or equal to 102.5X and 105K, respectively, of. rated core flow, at least once per 18 months.

The pump discharge valve and bypass valve in the inoperable loop, if not OPERABLE, shall be verified to be closed at least once per 31 days.

During single recirculation loop operation, all jet pumps, including those in the. inoperable loop, shall be demonstrated OPERABLE at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying that no two of the following conditions occur:¹¹¹ a.

The indicated recirculation loop flow in the operating loop differs by more than 10K from the established single recirculation pump speed-loop flow characteristics.

b.

The indicated total core flow differs by more than 10K from the established total core flow value from single recirculation loop flow measurements.

SUS)UEHANNA - UNIT 2 3/4 4-le Amendment No. 26

REACTOR COOLANT SYSTEM SURVEILLANCE RE UIREMENTS Continued c.

The indicated diffuses

-to-lower plenum differential pressure of any individual jet pump differs from estab-

, lished single recirculation loop patterns by more than 10K.

4.4. 1. 1.2. 10 The SURVEILLANCE REQUIREMENTS associated with the specifications referenced in 3. 4. 1. 1. 2a shall be followed.

See Special Test Exception 3. 10.4.

If not performed within the previous 31 days.

Detectors A and C of one LPRM string per core octant plus detectors A and C of one LPRM string in the center of the core should be monitored.

Initial value.

Final value to be determined based on startup testing.

Any required change to this value shall be submitted to the Commission within 90 days of test completion.

See Specification 3.4. 1. 1. 1 for two loop operation requirements.

¹¹ This requirement does not apply when the loop not in operation is isolated from the reactor pressure vessel.

¹¹¹ During startup testing following each refueling outage, data shall be recorded for the parameters listed to provide a basis for establishing the specified relationships.

Comparisons of the actual data in accordance with the criteria listed shall commence upon the performance of subsequent required surveillances.

SUS(UEHANNA - UNIT 2 3/4 4-1f Amendment No.

26

80 70 LLl BO o<

50 0

40 0

30 lao 20 O

10

-";."REGlQN GREATER THAN LlMIT

. REG(ON LESS THAN LIMIT 20 30 40 60 Bo Core Flow (% RATED) 70 80 Figure 3 4 1 1 2-1 SINGLE LOOP OPERATION THERNL POWER LIMITATIONS SUS(UEHANNA - UNIT 2 3/4 4-lg Amendment No. 45

REACTOR COOLANT SYSTEM JET PUMPS LIMITING CONDITION FOR OPERATION 3.4. 1.2 All jet pumps shall be OPERABLE.

APPLICABILITY:

OPERATIONAL CONDITIONS 1 and 2 when both recirculation 1oops are in operation.

ACTION:

With one or more jet pumps inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE RE UIREMENTS 4.4. 1.2""

Each of the above required jet pumps shall be demonstrated OPERABLE l'rior to THERMAL POWER exceeding 25K of RATED THERMAL POWER and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months " by determining recirculation loop flow, total core flow and diffuser to-lower plenum differential pressure for each jet pump and verifying that no two of the following conditions occur when the recirculation pumps are operating at the same speed:

a.

The indicated recirculation loop flow differs by more than 10K from the established pump speed-loop flow characteristics.

P b.

The indicated total core flow differs by more than 10K from the established total core flow value derived from recirculation loop

. flow measurements.

c.

The indicated diffuser-to-lower plenum differential pressure of any individual jet pump differs from established patterns by more than 10K.

"During the startup test program, data 'shall be recorded for the parameters listed to'rovide a basis for establishing the specified relationships.

Comparisons of the actual data in accordance with the criteria listed shall commence upon the conclusion of the startup test program.

""See Specification 4.4. 1. 1.2. 9 for single loop operation requirements.

SUSQUEHANNA - UNIT 2 3/4 4-2 Amendment No.

26

3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section assure that the peak cladding temperature 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. l AVERAGE PLANAR LINEAR HEAT GENERATION RATE This specification assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limit spec-ified 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.

For GE fuel, the peak clad tem-perature is calculated assuming a

LHGR for the highest powered rod which is equal to less than the design LHGR corrected for densification.

This LHGR times 1.02 is used in the heatup code along with the exposure dependent steady state gap conductance and rod-to-rod local peaking factor.

The Technical Specification AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) for GE fuel is this LHGR of the highest powered rod divided by its local peaking factor which results in a calcu-lated LOCA PCT much less than 2200'F.

The Technical Specification APLHGR for ANF fuel is specified to assure the PCT following a postulated LOCA will not exceed the 22000F limit.

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

The calculational procedure used to establish the APLHGR shown on Figures 3.2. 1-1, 3.2. 1-2, and 3.2. 1-3 is based on a loss-of-coolant accident analysis.

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

These models are described in Reference 1

or XN-NF-B0-19, Volumes 2, 2A, 2B and 2C.

3/4.2.2 'PRM SETPOINTS The flow biased simulated thermal power-upscale scram setting and flow biased simulated thermal power-upscale control rod block functions of the APRM instruments limit plant operations to the region covered by the transient and accident analyses.

In addition, the APRM sdtpoints must be adjusted to ensure that >1X plastic strain and fuel centerline melting do not occur during the worst anticipated operational occurrence (A00), including transients initiated from partial power operation.

For ANF fuel the T factor used to adjust the APRM setpoints is based on the FLPD calculated by dividing the actual LHGR by the LHGR obtained from Figure 3.2.2-1.

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

Figure 3.2.2-1 corresponds to the ratio of PAFF/1.2 under which cladding and fuel integrity is protected during AOO's.

SUSQUEHANNA - UNIT 2 B 3/4 2-1 Amendment No.

45

POWER DISTRIBUTION LIMITS BASES APRM SETPOINTS (Continued)

For GE fuel the T factor used to adjust the APRM setpoints is based on the FLPD calculated by. dividing the actual LHGR by the LHGR limit specified for GE fuel in Specification 3.2.4. 1 ~

3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRs at steady state operating conditions as speci-ified in Specification 3.2.3 are derived from the established fuel cladding integrity Safety Limit MCPR, and an analysis of abnormal operational transients.

For any abnormal operating transient analysis evaluation with the initial con-dition of the reactor being at the steady state operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip setting given in Specifica-tion 2.2.

To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting transients have been analyzed to determine which result in the largest reduction in CRITICAL POWER RATIO (CPR).

he type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease.

The limiting transient yields the largest delta MCPR.

When added to the Safety Limit MCPR, the required minimum operating limit MCPR of Specification 3.2.3 is obtained and presented in Figure 3.2.3-1 and 3.2.3-2.

The evaluation of a given transient begins with the system initial parameters shown in the cycle specific transient analysis report that are input to an ANF core dynamic behavior transient computer program.

The outputs of this program along with the initial MCPR form the input for further analyses of the thermally limiting bundle.

The codes and methodology to evaluate pressurization and non-pressurization events are described in XN-NF-79-71 and XN-NF-84-105.

The princi-pal result of this evaluation is the reduction in MCPR caused by the transient.

Figure 3.2.3-1 defines core flow dependent MCPR operating limits which assure that the Safety Limit MCPR will not be violated during a flow increase tran-sient resulting from a motor-generator speed control failure.

The flow depend-ent MCPR is only calculated for the manual flow control

mode, Therefore, automatic flow control operation is not permitted.

Figure 3.2.3-2 defines the power dependent MCPR operating limit which assures that the Safety limit MCPR will not be violated in the event of a feedwater controller failure initiated from a reduced power condition.

Cycle specific analyses are performed for the most limiting local core wide tran-sients to determine thermal margin.

Additional analyses are performed to determine the MCPR operating limit with either the Main Turbine Bypass inoperable or the EOC-RPT inoperable.

Analyses to determine thermal margin with both the EOC-RPT inoperable and Main Turbine Bypass inoperable have not been performed.

Therefore, operation in this condition is not permitted.

SUSQUEHANNA - UNIT 2 B 3/4 2-2 Amendment No. '5

0 3/4.4 REACTOR COOLANT SYSTEM 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.

For single loop operation, the MAPLHGR limits are multiplied by a factor of 0.81 for GE fuel and 1.0 for the ANF fuel.

These multiplication factors are derived from LOCA analyses initiated from single loop operation conditions.

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

The MINIMUM CRITICAL POWER RATIO (MCPR) limits for single loop operation assure that the Safety Limit MCPR is not exceeded for any Anticipated Operational Occurrence (AOO) and for the Recirculation Pump Seizure Accident.

For single loop operation, the RBM and APRM setpoints are adjusted by a 7X deer'ease in recirculation drive flow to account for the active 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 vibration.

Surveillance on differential temperatures below the threshold limits of THERMAL POWER or recirculation loop flow mitigates undue thermal stress on vessel

nozzles, recirculation pumps and the vessel bottom head during extended opera-tion in the single loop mode.

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

THERMAL POWER, core flow, and neutron flux noise level limitations are prescribed in accordance with the recommendations of General Electric Service Information Letter No.

380, Revision 1, "BWR Core Thermal Hydraulic Stability," dated Febru-ary 10, 1984.

An inoperable jet pump is not, in itself, a sufficient reason to declare a re-circulation 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 coastdown from either recirculation loop following a LOCA.

In the case where 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 50'F of each other prior to startup of an idle loop.

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

Since the coolant in the bottom of the vessel is at a lower temperature than the coolant in the upper regions of the core, undue stress on the vessel would result if the temperature differ-ence was greater than 145'F.

SUS(UEHANNA - UNIT 2 B 3/4 4"1 Amendment No.

45

REACTOR COOLANT SYSTEM BASES 3/4.4.2 SAFETY/RELIEF VALVES The safety valve function of the safety/relief valves operate to prevent the reactor coolant system from being pressurized above the Safety Limit of 1325 psig in accordance with the ASME Code.

A total of 10 OPERABLE safety/relief valves is required to limit reactor pressure to within ASME III allowable values for the worst case upset transient.

Demonstration of the safety/relief valve liftsettings will occur only during shutdown and will be performed in accordance with the provisions of Specifica-tion 4.0.5.

3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.3.

1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the reactor coolant pressure boundary.

3/4.4.3.2 OPERATIONAL LEAKAGE The allowable leakage rates from the reactor coolant system have been based on the predicted and experimentally observed behavior of cracks in pipes.

The normally expected background leakage due to equipment design and the detection capability of the instrumentation for determining system leakage was also considered.

The evidence obtained from experiments suggests that for leakage somewhat greater than that specified for UNIDENTIFIED LEAKAGE the probability is small that the imperfection or crack associated with such leakage would grow rapidly.

However, in all cases, if the leakage rates exceed the values specified or the leakage is located and known to be PRESSURE BOUNDARY LEAKAGE, the reactor will be shutdown to allow further investigation and corrective action.

The Surveillance Requirements for RCS pressure isolation valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA.

3/4.4. 4 CHEMISTRY The water chemistry'imits of the reactor coolant system are established to prevent damage to the reactor materials in contact with the coolant.

Chloride limits are specified to prevent stress corrosion cracking of the stainless steel.

The effect of chloride is not as great when the oxygen cmeentration in the coolant is low, thus the 0.2 ppm limit on chlorides is ~itted during POWER OPERATION.

During shutdown and refueling operations, tha temperature necessary for stress corrosion to occur is not present so a 0.5 ppe concentration of chlorides is not considered harmful during these periods.

SUS/UEHANNA - UNIT 2 B 3/4 4-2 Amendment No. Rim

ML18040B209

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