Text

Amend 91 to License NPF-22,changing TS in Support of Cycle Six Reload
	ML17157C054
Person / Time
Site:	Susquehanna
Issue date:	10/28/1992
From:	Chris Miller; Office of Nuclear Reactor Regulation
To:	;
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 PENNSYLVANIA POWER 5 LIGHT COMPANY ALLEGHENY ELECTRIC COOP RATIVE INC.

DOCKET NO. 50-388 SUS UEHANNA STEAM ELECTRIC STATION UNIT 2 AMENDMENT TO FACILITY OPERATING LICENS Amendment No. 9j License No. NPF-22 1.

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

A.

The application for the amendment filed by the Pennsylvania Power 8

Light Company, dated June 30,

1992, and its supplements dated July 21, 1992 and October 19,

1992, 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; B.

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 regulati'ons 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 applicable 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:

92ii060i55 921028,-

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(2) Technical S ecifications and Environmental Protection Plan The Technical, Specifications contained in Appendix A, as revised through Amendment No 91 and the Environmental Protection Plan con-tained in Appendix B, are hereby incorporated in the license.

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

3.

This license amendment is effective as of its date of issuance and is to be implemented within 30 days of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

Attachment:

Changes to the Technical Specifications Date of Issuance:

october 28, 1992 Charles L. Miller, Director Project Directorate I-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

ATTACHMENT TO LICENSE AMENOMENT NO.

91 FACILITY OPERATING LIC NSE NO. NPF-22

~TN 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 Xxl XXi1 8 2-1 8 2-2 8 2-3 8 2-4 8 2-5 8 2-6 8 2-7 8 2-8 8 2-9 INSERT ill iv*

xxi*

xxii xxiii xxiv*

8 2-1 8 2-2 8 2-3 2-4 8 2-5 8 2-6 8 2-7 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 2-7 3/4 2-8 3/4 2-9 3/4 2-10 3/4 2-1*

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

3/4 2-5 3/4 2-5a 3/4 2-6 3/4 2-7 3/4 2-8 3/4 2-8a

'I l

i

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.*

3/4 4-1b 3/4 4-lc 3/4 4-1f 3/4 2-Sb 3/4 2-9 3/4 2-10 3/4 2-10a 3/4 4-1b 3/4 4-1c 3/4 4-lf 8 3/4 1-1 8 3/4 1-2 8 3/4 1-3 8 3/4 1-4 8 3/4 2-1 8 3/4 2-2 8 3/4 4-1 8 3/4 4-2 5-5 5-6 8 3/4 1-1*

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

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

5 5*

5-6

INDEX SAFETY LIMITS AND LIMITINGSAFETY SYSTEM SETTINGS

~SE Tl N

PAGE 2.1 AF TY LIMIT THERMAL POWER, Low Pressure or Low Flow THERMAL POWER, High Pressure and High Flow Reactor Coolant System Pressure Reactor Vessel Water Level 2.2 LIMITIN AFETY Y TEM ETTIN Reactor Protection System Instrumentation Setpoints 2-1 2-1 2-1 2-2

~

2 3 BASES 2.1 AFETY LIMIT THERMAL POWER, Low Pressure or Low Flow THERMAL POWER, High Pressure and High Flow R.actor Coolant System Pressure Reactor Vessel Water Level B 2-1 B 2-2 8 2-3 B 2-3 2.

IMITIN AFETY Y TEM TTIN Reactor Protection System Instrumentation Setpoints B 2-4 SUSQUEHANNA - UNIT 2 Amendment No.91

INOEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS 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 Control Rod Operability..

PAGE 3/4 0-1 3/4 1"1 3/4 1-2 3/4 1-3 Control Rod Maximum 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 Accumulators..........

Control Rod Drive Coupling..

Control Rod Pos ition Indicati on..

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 1-9 3/4 1"11 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 3/4. 2. 3 3/4. 2. 4 MINIMUM CRITICAL POWER RATIO..

LINEAR HEAT GENERATION RATE 3/4 2-6 3/4 2-9 SUSQUEHANNA - UNIT 2 1V Amendment No. 58 1989

INDEX AOHINISTRATIVE CONTROLS

6. 13 PROCESS CONTROL PROGRAM..

~

6.14 OFFSITE OOSE CALCULATION MANUAL......

~

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

6-24 SUS(UEMANNA - UNIT 2 XX)

INDEX LIST F FIG RE

~FI I~R 3.1.5-1 3.1.5-2 3.2.1-1 SODIUM PENTABORATE SOLUTION TEMPERATURE/

CONCENTRATION REQUIREMENTS...................

3/4 1-21 SODIUM PENTABORATE SOLUTION CONCENTRATION..... 3/4 1-22 MAXIMUMAVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VS. AVERAGE PLANAR EXPOSURE, SNP 9 X 9 FUEL.................... ~........... 3/4 2-2 3.2.2-1 3.2.2-2 3.2.3-1 3.2.3-2 3.2.3-3 3.2.3-4 3.2.4-1 3.2.4-2 LINEAR HEAT GENERATION RATE FOR APRM SETPOINTS VERSUS AVERAGE PLANAR EXPOSURE, SNP (EXCLUDING XN-1) FUEL................. ~..... 3/4 2-5 LINEAR HEAT GENERATION RATE FOR APRM SETPOINTS VERSUS AVERAGE PLANAR EXPOSURE, SNP XN-1 FUEL.... 3/4 2-5a FLOW DEPENDENT MCPR OPERATING LIMIT............

3/4 2-7 POWER DEPENDENT MCPR OPERATING LIMIT (EOC-RPT AND MAINTURBINE BYPASS OPERABLE)......

~ 3/4 2-8 POWER DEPENDENT MCPR OPERATING LIMIT (EOC-RPT OPERABLE AND MAINTURBINE BYPASS INOPERABLE)............... ~.......... ~....... 3/4 2-Sa POWER DEPENDENT MCPR OPERATING LIMIT (EOC-RPT INOPERABLE AND MAIN TURBINE BYPASS OPERABLE)............................

~... ~... 3/4 2-Sb LINEAR HEAT GENERATION RATE (LHGR) LIMITVERSUS AVERAGE PLANAR EXPOSURE, SNP 9X9 (EXCLUDING XN-1) FUEL.................. ~....... 3/4 2-10 LINER HEAT GENERATION RATE (LHGR) LIMITVERSUS AVERAGE PLANAR EXPOSURE, SNP XN-1 9X9 FUEL...... 3/4 2-10a 3.4.1.1.1-1 THERMAL POWER RESTRICTIONS.. ~... ~.............'3/4 4-1b 3.4.6.1-1 MINIMUMREACTOR VESSEL METALTEMPERATURE VS.

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

3/4 4-18 4.7.4-1 B 3/4 3-1 B 3/4.4.6-1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONALTEST.....

~

~ 3/4 7-15 REACTOR VESSEL WATER LEVEL

. ~...... ~...........

B 3/4 3-8 FAST NEUTRON FLUENCE (E) 1MeV) AT 1/4 T AS A FUNCTION OF SERVICE LIFE.......,........,.......

B 3/4 4-7 SUSQUEHANNA - UNIT 2 XXII Amendment No.9>

'IST OF FIG RES ContInUed INDEX

~FI I~R PAGE 5.1.1-1 EXCLUSION AREA...........

~

~ 5 2 5.1.2-1 5.1.3-1a 5.1.3-1b LOW POPULATION ZONE

~

~.. ~...

~ ~................

5-3 MAP DEFINING UNRESTRICTED AREAS FOR RADIOACTIVE GASEOUS AND LIQUIDEFFLUENTS...................

5-4 MAP DEFINING UNRESTRICTED AREAS FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS.... ~..............

5-5 SUSQUEHANNA - UNIT 2 XXIII Amendment No.~

LIST OF TABLES INOEX

~

/

TABLE 1.2 2.2. 1-1 3.3. 1-1 3.3. 1-2 4.3.1. 1-1 3.3. 2-1

3. 3. 2-2 3.3. 2-3 4.3.2.1-1

.PAGE SURVEILLANCE FREQUENCY NOTATION...................

1-9 OPERATIONAL CONDITIONS............................

1-10 REACTOR PROTECTION SYSTEM INSTRUMENTATION ETPOINTS.........................................

S REACTOR PROTECTION SYSTEM INSTRUMENTATION.........

E-4 3/4 3-2 REACTOR PROTECTION SYSTEM RESPONSE TIMES..........

3/4 3-6 REACTOR PROTECTION. SYSTEM INSTRUMENTATION SURVEILLANCE RE(UIREMENTS.........................

3/4 3-7 ISOLATION ACTUATION INSTRUMENTATION...............

3/4 3-11 t

ISOLATION ACTUATION INSTRUMENTATION SETPOINTS.....

3/4 3-17 ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME....

3/4 3-21 ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS......................................

3/4 3-23 3.3.3-1 EMERGENCY CORE. COOLING SYSTEM ACTUATION INSTRUMENTATION..........,......................

3/4 3-28 3.3.3"2 3.3. 3-3 4.3.3. 1"1 3.3.4. 1-1 3.3.4.1-2 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS......................

EMERGENCY CORE COOLING SYSTEM RESPONSE TIMES '..

~

3/4 3-31 3/4 3-33 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE RE(UIREMENTS.........

3/4 3-34 ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION...................................

3/4 3-37 ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION SETPOINTS.........................

3/4 3-38 4.3.4. 1-1

3. 3.4. 2-1 ATWS RECIRCULATION.PUMP TRIP ACTUATION INSTRUMENTATION SURVEILLANCE RE(UIREHENTS......

END-OF"CYCLE RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION................................

3/4 3-39 3/4 3-42

3. 3.4. 2-2 3.3.4. 2-3 END-OF-CYCLE RECIRCULATION PUMP TRIP SETPOINTS....

3/4 3"43 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM

RESPONSE

TIME.....................................

3/4 3-44 SUS(UEHANNA UNIT 2

~~.tllP XX1 V q >/2 iL~ 8 Amendment No.

3l

'LLCVL4'l.c'.""

Act.~

2.1 AFETY LIMIT

~ BASE 2 0 INTR D

TI N The fuel cladding, reactor pressure vessel and primary system piping are the principal barriers to the release of radioactive materials to the environs.

Safety 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 SNP fuel. MCPR greater than the specified limit represents a conservative 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.

Although 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 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 condition intended by design for planned operation.

The MCPR fuel cladding integrity Safety Limit assures that during normal operation and during anticipated operational occurrences, at least 99.9% of the fuel rods in the core do not experience transition boiling (ref. XN-NF-524(A) Revision 1).

2.1.1 THERMAL POWER Low Pressure or Low Flow The use of the XN-3 correlation is valid for critical power calculations at pressure greater than 580 psig and bundle mass fluxes greater than 0.25 x 10'bs./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 SNP 9 x 9 fuel design, the minimum bundle flow is greater than 30,000 Ibs/hr. For the SNP 9 x 9 design, the coolant minimum flow and maximum flow area is such that the mass flux is always greater than 0.25 x 10'bs/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'bs/hr-ft's 3.35 Mwt or greater.

At 25% 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 THERMALPOWER limit of 25% of RATED THERMAL POWER for reactor pressures below 785 psig is conservative.

SUSQUEHANNA - UNIT 2 B 2-1 Amendment No.~

AF TY LIMIT BASE 2.1.2 THERMAL P WER Hi h Pre r

an Hi h Fl w 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 transition 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 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 99.9% 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 LimitMCPR is based on a detailed 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 (A), Revision 1 and PL-NF-90-001 describe the methodologies used in determining the Safety Limit MCPR.

~ 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 within 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 Section B 2.1.1), the assumed reactor conditions used in defining the safety limit introduce conservatism 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 transition 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 failure is a very conservative approach.

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

SNP fuel is monitored using the XN-3 Critical Power Correlation.

SNP has determined that this correlation provides sufficient conservatism to preclude the need for any penalty due to channel bow.

The conservatism has been evaluated by SNP to be greater than the maximum expected hCPR (0.02) due to channel bow in C-lattice plants using channels for only one fuel bundle lifetime. Since Susquehanna SES Unit 2 is a C-lattice plant and uses channels for only one fuel bundle lifetime, monitoring of the MCPR limit with the XN-3 Critical Power Correlation is conservative with respect to channel bow and addresses the concerns of NRC Bulletin No. 90-02 entitled "Loss of Thermal Margin Caused by Channel Box Bow."

SUSQUEHANNA - UNIT 2 B 2-2 Amendment No,9j

AFE LIIVIIT

'ASES 2.1.

REA R

LANT Y TEN PRESSURE The Safety Limit for the reactor coolant system pressure has been selected such that it is at a pressure below which it can be shown that the integrity of the system is not endangered.

The reactor pressure vessel is designed to Section III of the ASME Boiler and Pressure Vessel

Code, 1968 Edition, including Addenda through Summer 1970, which permits a maximum pressure transient of 110%, 1375 psig, of design pressure, 1250 psig.

The Safety Limit of 1325 psig, as measured by the reactor vessel steam dome pressure indicator, is equivalent to 1375 psig at the lowest elevation of the reactor coolant system.

The reactor coolant system is designed to the USAS Piping Code, Section B31

~ 1, which permits a maximum pressure transient of 120%, 1375 psig, of design pressure, 1150 psig for suction piping and 1500 psig for discharge piping. The pressure Safety Limit is selected to be the lowest transient overpressure allowed by the applicable codes.

2.1.4 RE CT R VE SEL WATER LEVEL With fuel in the reactor vessel during periods when the reactor is shutdown, consideration must be given to water level requirements due to the effect of decay heat.

Ifthe water level should drop below the top of the active irradiated fuel during this period, the ability to remove decay heat is reduced.

This reduction in cooling capability could lead to elevated cladding temperatures and clad perforation in the event that the water level became less than two-thirds of the core height.

The Safety Limit has been established at the top of the active irradiated fuel to provide a point which can be monitored and also provide adequate margin for effective action.

SUSQUEHANNA - UNIT 2 B 2-3 Amendment No. o"'

2.2 LIIVIITIN AFETY YSTEIVI SETTING BASE 2.2.1 REA T R

R TE Tl N

Y TEN IN TR NIENTATI N ETP INT The Reactor Protection System instrumentation setpoints specified in Table 2.2.1-1'are the values at which the reactor trips are set for each parameter.

The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their Safety Limits during normal operation and design basis anticipated operational occurrences and to assist in mitigating the consequences of accidents.

Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.

1.

In r

n M n' N

r n Fl x - Hi h The IRM system consists of 8 chambers, 4 in each of the reactor trip systems.

The IRM is a 5 decade 10 range instrument.

The trip setpoint of 120 divisions of scale is active in each of the 10 ranges.

Thus as the IRM is ranged up to accommodate the increase in power level, the trip setpoint is also ranged up.

The IRM instruments provide for overlap with both the APRM and SRM systems.

The most significant source of reactivity changes during the power increase is due to control rod withdrawal.

In order to ensure that the IRM provides the required protection, a range of rod withdrawal accidents have been analyzed.

The results of these analyses are in Section 15.4 of the FSAR.

The most severe case involves an initial condition in which THERMAL POWER is at approximately 1% of RATED THERMALPOWER. Additional conservatism was taken in this analysis by assuming the IRM channel closest to the control rod being withdrawn is bypassed.

The results of this analysis show that the reactor is shutdown and power peak is limited to 21% of RATED THERMAL POWER with the peak fuel enthalpy well below the fuel failure threshold of 170 cal/gm.

Based on this analysis, the IRM provides protection against local control rod errors and continuous withdrawal of control rods in sequence and provides backup protection for the APRM.

2.

Avr P wrRn Mni r

For, operation at low pressure and low flow during STARTUP, the APRM scram setting of 15% of RATED THERMAL POWER provides adequate thermal margin between the setpoint and the Safety Limits. The margin accommodates the anticipated maneuvers associated with power plant startup.

Effects of increasing pressure at zero or low void content are minor and cold water from sources available during startup is not much colder than that already in the system.

Temperature coefficients are small and control rod patterns are constrained by the RSCS and RWM. Of all the possible sources of reactivity input, uniform SUSQUEHANNA - UNIT 2 B 2-4 Amendment No. 91

LIMIT TEM ETTIN

~ BASES REACT E

I N

YSTEM IN TRUMENTATI N ETPOINT (Continued)

Av P w R

M ni or (Continued) control rod withdrawal is the most probable cause of significant power increase.

Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks and because several rods must be moved to change power by a significant amount, the rate of power rise is very slow.

Generally the heat flux is in near equilibrium with the fission rate.

In an assumed uniform rod withdrawal approach to the trip level, the rate of power rise is not more than 5% of RATED THERMAL POWER per minute and the APRM system would be more than adequate to assure shutdown before the power could exceed the Safety Limit. The 15% neutron flux trip remains active until the mode switch is placed in the Run position.

The APRM trip system is calibrated using heat balance data taken during steady state conditions.

Fission chambers provide the basic input to the system and therefore the monitors respond directly and quickly to changes due to transient operation for the case of the Fixed Neutron Flux-Upscale 118% setpoint; i.e. for a power increase, the THERMAL POWER of the fuel will be less than that indicated by the neutron flux due to the time constants of the heat transfer associated with the fuel. For the Flow Biased Simulated Thermal Power-Upscale

setpoint, a time constant of 6 a 1 seconds is introduced into the flow biased APRM in order to simulate the fuel thermal transient characteristics.

A more conservative maximum value is used for the flow biased setpoint as shown in Table 2.2.1-1.

The APRM setpoints were selected to provide adequate margin for the Safety Limits and yet allow operating margin that reduces the possibility of unnecessary shutdown.

The flow referenced trip setpoint must be adjusted by the specified formula in Specification 3.2.2 in order to maintain these margins when MFLPD is greater than or equal to FRTP.

3.

R r V I

m 0 m Pr ur -Hi h High pressure in the nuclear system could cause a rupture to the nuclear system process barrier resulting in the release of fission products.

A pressure increase while operating will also tend to increase the power of the reactor by compressing voids thus adding reactivity.

The trip will quickly reduce the neutron flux, counteracting the pressure increase.

The trip setting is slightly higher than the operating pressure to permit normal operation without spurious trips. The setting provides for a wide margin to the maximum allowable design pressure and takes into account the location of the pressure measurement compared to the highest pressure that occurs in the system during a transient.

This trip setpoint is effective at low power/flow conditions when the turbine stop valve closure trip is bypassed.

For a turbine trip under these conditions, the transient analysis indicated an adequate margin to the thermal hydraulic limit.

SUSQUEHANNA - UNIT 2 B 2-5 Amendment No. ~>

LIMITIN E

Y TEIVI E

IN REA N

S EM IN TR ENT Tl ETP INT (Continued) 4, R

V IW rLvlL The reactor vessel water level trip setpoint was chosen far enough below the normal operating level to avoid spurious trips but high enough above the fuel to assure that there is adequate protection for the fuel.

5.

Mi mLin I

I inVlv-I r

The main steam line isolation valve closure trip was provided to limit the amount of fission product release for certain postulated events.

The MSIV's are closed automatically from measured parameters such as high steam flow, high steam line radiation, low reactor water level, high steam tunnel temperature and low steam line pressure.

The MSIV's closure scram anticipates the pressure and flux transients which could'follow MSIV closure and thereby protects reactor vessel pressure and thermal/hydraulic Safety Limits.

6.

Min mLi R

i in-Hih The main steam line radiation detectors are provided to detect a gross failure of the fuel cladding. When the high radiation is detected, a trip is initiated to reduce the continued failure of fuel cladding.

At the same time the main steam line isolation valves are closed to limit the release of fission products.

The trip setting is high enough above background radiation levels to prevent spurious trips yet low enough to promptly detect gross failures in the fuel cladding.

No credit was taken for operation of this trip in the accident analyses; however, its functional capability at the specified trip setting is required by this specification to enhance the overall reliability of the Reactor Protection System.

7.

0 w

II Pre r -Hi High pressure in the drywell could indicate a break in the primary pressure boundary systems.

The reactor is tripped in order to minimize the possibility of fuel damage and reduce the amount of energy being added to the coolant.

The trip setting was selected as low as possible without causing spurious trips.

8.

rmDi har Vlm W rLvIHih The scram discharge volume receives the water displaced by the motion of the control rod drive pistons during a reactor scram.

Should this volume fill up to a point where there is insufficient volume to accept the displaced water at pressures below 65 psig, control rod insertion would be hindered.

The reactor is therefore tripped when the water level has reached a point high enough to indicate that it is indeed filling up, but the volume is still great enough to accommodate the water from the movement of the rods at pressures below 65 psig when they are tripped.

SUSQUEHANNA - Unit 2 B 2-6 Amendment No. 91

LIMITIN A ETY Y TEIVI ETTI

'BASE REA T P

TE Tl N

Y TEM INSTR MENTATI N ETPOINT (Continued)

9. Trin Vlv-I ur The turbine stop valve closure trip anticipates the pressure, neutron flux, and heat flux increases that would result from closure of the stop valves. With a trip setting of 5.5%

'fvalve closure from full open, the resultant increase in heat flux is such that adequate thermal margins are maintained during the worst case transient assuming the turbine bypass valves operate.

10. Trin nrlVIv F I

r Tri ilPr r-w The turbine control valve fast closure trip anticipates the pressure, neutron flux, and heat flux increase that could result from fast closure of the turbine control valves due to load rejection coincident with failure of the turbine bypass valves.

The Reactor Protection System initiates a trip when fast closure of the control valves is initiated by the fast acting solenoid valves and in less than 30 milliseconds after the start of control valve fast closure.

This is achieved by the action of the fast acting solenoid valves in rapidly reducing hydraulic trip oil pressure at the main turbine control valve actuator disc dump valves.

This loss of pressure is sensed by pressure switches whose contacts form the one-out-of-two-twice logic input to the Reactor Protection System.

This trip

setting, a faster closure time, and a different valve characteristic from that of the turbine stop valve, combine to produce transients which are very similar to that for the stop valve.

Relevant transient analyses are discussed in Section 15.2 of the Final Safety Analysis Report.

11.

Re or Mo wi h

hu d wn Posi ion The reactor mode switch Shutdown position is a redundant channel to the automatic protective instrumentation channels and provides additional manual reactor trip capability.

The Manual Scram is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability.

SUSQUEHANNA - UNIT 2 B 2-7 Amendment No. ~

I.

/

)

~

t A

~

3/4. 2 POWER DISTRIBUTION LIMITS 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 all fuel shall not exceed the limit shown in Figure 3.2. 1-1.

APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or AAA f RATER TIIERMAL PRLIER.

ACTION:

With an APLHGR exceeding the limit of Figure 3.2.1-1, initiate corrective action within 15 minutes and restore APLHGR to within the required 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 4.2.1 All APLHGRs shall be verified to be equal to or less than the limit determined from Figure 3.2. 1-1:

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.

SUSQUEHANNA - UNIT 2 3/4 2-1 Amendment No.58 NQV 3 tg-~

12 C0~

C O CO v g

8 ~

0)

LQ m E v lO X

C 11 10 8

8

.: :'PERMISSABLE:::'::.::::.::.:::

REGION OF OPERATION

. 40,000;.:

. 48,000;.:

0 I

~

~ ::::::.:.:::.:.::.:.:: 36,000;::

'2 0

7

~

0.0;::::::::::.: 20,000;

]A 6 1A 0

~.

~ ~ :

~ ~.

~.

~ ~.

~ ~ :

~ ~

~ ~.

~.

~ ~.

~ ~

= ~

~ ~

%aloC...,..

~

...:: ~:.:::.::. '":.:.:: ':::.:30,000;:

8.9 7

0 6000 10000 16000 20000 25000 30000 35000 10000 46000 60000 Averape Bundle Exposure (MWD/MT)

MAXIMUMAVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE BUNDLE EXPOSURE SNP 9X9 FUEL FIGURE 3.2.1-1

7 g

~

0 WE D

TRIB Tl N LIMIT 4.2.2 A

IVI ETP INT LIMITIN NDITION FOR OPERATION

t 3.2.2 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 oin ¹ All w I Vl S ~ (0.58W + 59%) T SRS ~ (0 58W + 50%) T S N (0.58W + 62%) T SRB < (0.58W + 53%) T where:

S and SRB are in percent of RATED THERMAL POWER, W =

Loop recirculation flow as a percentage of the loop recirculation flow which produces a rated core flow of 100 million Ibs/hr, T

=

Lowest value of the ratio of FRACTION OF RATED THERMAL POWER (FRTP) divided by the MAXIMUM FRACTION OF LIMITING POWER DENSITY. The FRACTION OF LIMITINGPOWER DENSITY (FLOP) for SNP fuel is the actual LHGR divided by the applicable LINEAR HEAT GENERATION RATE from Figure 3.2.2-1 or 3.2.2-2.

T is always less than or equal to 1.0 equal to 25% of RATED THERMAL POWER.

~A',Tl(~N:

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

SURVEILLANCE REQUIREIVIENTS 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 90% of RATED THERMAL POWER, rather than adjusting the APRM setpoints, the APRM gain may be adjusted such that APRM readings are greater than or equal to 100% times MFLPD, provided that the adjusted APRM reading does not exceed 100% of RATED THERMAL POWER, the required gain adjustment increment does not exceed 10% of RATED THERMALPOWER, and a notice of the adjustment is posted on the reactor control panel.

See Specification 3.4.1.1.2.a for single loop operation requirements.

SUSQUEHANNA - UNIT 2 3/4 2-3 Amendment No. 91

POWER DISTRIBUTION LIMITS 3/4. 2. 2 APRM SETPOINTS LIMITING CONDITION FOR OPERATION 4.2 ~ 2 a.

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 15X 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 MFLPO greater than or equal to FRTP.

The pr'ovisions of Specification 4.0.4 are not applicable.

SUSQUEHANNA - UNIT 2 3/4 2-4 Amendment No. 58

>QV 3 1989

18 I~

m~

16 0

o~

C g0 Ue

~M eg 12 XK

~Q

.CO 10 UU 0.0; 16.0

~ ~

.:25,400; 14.0

43,200; 9.0 48,000; 8.3 0

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

LINEAR HEAT GENERATION RATE FOR APRM SETPOINTS VERSUS AVERAGE PLANAR EXPOSURE SNP (EXCLUDING XN-1) FUEL FIGURE 3.2.2-1 60000

~

Sw 14 0

~~ co

~cg t~

C g0

+

12 Uo

~M gZK

~A

'4+

10 C ()

8 ~

U L.

0.0;::

16,500; 14.76::14.75

~

~ I

~

67,300; 8.67

~

0

~

0 10000 20000 30000 40000 60000 Average Planar Exposure (MWD/MT) 80000 LINEAR HEAT GENERATION RATE FOR APRM SETPOINTS VERSUS AVERAGE PLANAR EXPOSURE SNP XN-I FUEL FIGURE 3.2.2-2

POWER Dl TRIBUTION LIIVIIT 4.2.

MINIM IVI RITI AL P WER RATI LIMITINGCONDITION FOR OPERATION 3.2.3 The MINIMUMCRITICALPOWER RATIO (MCPR) shall be greater than or equal to the greater of:

a)

The Flow-Dependent MCPR value determine from Figure 3.2.3-1, and b)

The Power-Dependent MCPR value determined from one of the following figures, as appropriate:

Figure 3.2.3-2:

EOC-RPT and Main Turbine Bypass Operable Figure 3.2.3-3:

Main Turbine Bypass Inoperable Figure 3.2.3-4:

EOC-RPT Inoperable using a linear interpolation between Curve A and Curve 8 of the appropriate figure, based on the results of each scram time surveillance test required by Specification 4.1.3.3.

~*':

equal to 25% of RATED THERMAL POWER.

~AT~IN:

With MCPR less than the applicable MCPR limit determined above, initiate corrective action

'within 15 minutes and restore MCPR to within the required limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMALPOWER to less than 25% of RATED THERMALPOWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 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, 3.2.3-3 and 3.2.3-4 as

, appropriate:

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% 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 LIMITINGCONTROL ROD PATTERN for MCPR.

d.

The provisions of Specification 4.0.4 are not applicable.

SUSQUEHANNA' UNIT 2 3/4 2-6 Amendment No. 91

(30,1.90)

(37,1.76)

(46,1.63)

(60,1.47)

(76.8,1.34)

(100,1.34) 40 60 60 70 80 ao Total Core Flow (% QF RATED)

FLOW DEPENDENT MCPR OPERATING LIMIT FIGURE 3.2.3-1 100

~4 2.0 1.9 1.8 Legend CURVE A: MINIMUMALLOWABLESCRAM SPEED FROM T.S. 3.1.3.3 CURVE B: AVERAGE SCRAM SPEED 4.4 ft/sec E

l.7 Ol C

~~

lP l.8 CL0 lK 1.6 (26,1.66)

(26,1.64)

(4O,1.64)

(40,1.6s)

~ (66,1.41) l.4 1.3 (66,1.40)

(81.1,1.36)

(80.2,1.34)

A 8

(100,1.36)

(100,1.34) 1.2 20 30 80 40 60 60 70 Core Power (% RATED)

POWER DEPENDENT MCPR OPERATING LIMIT EOC-RPT AND MAINTURBINE BYPASS OPERABLE FIGURE 3.2.3-2 ao 100

2.0 1.8 (25,1.80)

Legend CURVE A: MINIMUMALLOWABLESCRAM SPEED FROM T.S. 3.1.3.3 CURVE B: AVERAGE SCRAM SPEED 4.4 ft/sec E

1.7 Ch C

~~

1.8 CL0 IZ 1.6 1.4 (25,1.79)

(40,1.70)

(40,1.es)

(es,1.s7)

(es,1.se)

(84,1.S1)

~ A ~~a (84,1.49)

(100,1.4e)

(100,1.44) 1.3 1.2 20 30 80 80 40 60 60 70 Core Power (% RATED)

POWER DEPENDENT MCPR OPERATING LIMIT EOC-RPT OPERABLE AND MAINTURBINE BYPASS INOPERABLE FIGURE 3.2.3-3 100

2.0 1.9 1.8 Legend CURVE A: MINIMUMALLOWABLESCRAM SPEED FROM T.S. 3.1.3.3 CURVE B: AVERAGE SCRAM SPEED 4.4 ft/sec E

1.7 0)

C 1.6 CL0 0:

~O 1'6 1.4 (26,1.6S)

(26.1.65)

(40,1.57)

(40,1.66)

(61.5,1.45)

(66,1.43)

(66 5 1 42)

(1OO,1.46)

(100,1.42) 1.3 1.2 20 40 80 80 60 60 70 Core Power (% RATED)

POWER DEPENDENT MCPR OPERATING LIMIT EOC-RPT INOPERABLE AND MAINTURBINE BYPASS OPERABLE FIGURE 3.2.3-4 100

P WER Dl TRIB Tl N LIMIT 4.2.4 LINE R

EAT ENERATI N RATE LIMITIN NDITION FOR OPERATION 3.2.4 The LINEAR HEAT GENERATION RATE (LHGR) shall not exceed the LHGR limit determined from either Figure 3.2.4-1 or 3.2.4-2, as appropriate.

~*'"':

equal to 25% of RATED THERMAL POWER.

~AT~IN:

With the LHGR of any fuel rod exceeding its applicable limitfrom Figure 3.2.4-1 or 3.2.4-2, t

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 THERMALPOWER to less than 25% of RATED THERMALPOWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 4.2.4 LHGRs 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 15% 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 LIMITINGCONTROL ROD PATTERN for LHGR.

d.

The provisions of Specification 4.0.4 are not applicable.

SUSQUEHANNA - UNIT 2 3/4 2-9 Amendment No,9-

E I

R lK c0

~~

IcQ (9

Ix L

QC 18 14 12 10 8

8 0

-. 0.0;..

-'....:..'.-....::.....'....':.....'...-'........: '24,000;:

13.0::

12.0 PERMISSABLE REGION OF OPERATION

35,000; --":

9.5 48,000; 7.72 10000 20000 30000 40000 Average Planar Exposure (MWD/MT)

LINEAR HEAT GENERATION RATE (LHGR) LIMIT VERSUS AVERAGE PLANAR EXPOSURE SNP 9X9 (EXCLUDING XN-1) FUEL FIGURE 3.2.4-1 60000

E I

K c0

~~

I CQ C9 x

I C

18 14 12 10 8

8 0

".: 0.0;:.":.. 15,600;

.:131.:::131

~

r

'::. PERMISSABLE REGION OF OPERATION r

67,300; 7.7

~

10000 20000 30000 40000 60000 Average Planar Exposure (MWD/MT)

LINEAR HEAT GENERATION RATE (LHGR) LIMIT VERSUS AVERAGE PLANAR EXPOSURE SNP XN-1 9X9 FUEL FIGURE 3.2A-2 80000

~

e ~

REA T LA Y TEM RE IR I

L P

IN LEL PERATI N

LIMITIN NDITION FOR OPERATION 3.4.1.1.2 One reactor coolant recirculation loop shall be in operation with the pump speed s 80% of the rated pump speed and the reactor at a THERMALPOWER/core flow condition outside of Regions I and II of Figure 3.4.1.1.1-1, 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:

in All w I Vl s 0.58W + 54%

s 0.58W + 57%

3.

Specification 3.2.2:

the APRM Setpoints shall be as follows:

In S s (0.58W + 54%) T SRB s (0,58W + 45%) T All w I

V S s (0.58W + 57%) T SRB s (0.58W + 48%) T 4.

Specification 3.2.3:

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

a.

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

the MCPR determined from Figure 3.2.3-2; Figure 3.2.3-3 or Figure 3.2.3-4 as appropriate, plus 0.01.

5.

Table 3.3.6-2:

the RBM/APRM Control Rod Block Setpoints shall be as follows:

a.

RBM - Upscale s 0.66W + 36%

s 0.66W + 39%

b.

APRM-Flow Biased

~*'"':

operation.4

~AT(

NN:

a.

In OPERATIONAL CONDITION 1:

s 0.58W + 45%

s 0.58W + 48%

1.

With a) b)

c) no reactor coolant system recirculation loops in operation, or Region I of Figure 3.4.1.1.1-1 entered, or Region II of Figure 3.4.1

~ 1.1-1 entered and core thermal hydraulic instability occurring as evidenced by:

SUSQUEHANNA - UNIT 2 3/4 4-1c Amendment 4 ~>

REA T R

OOLANT SYSTEM SURVEILLANCE REQUIREMENTS Continued b.

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

c.

The indicated diffuser-to-lower plenum differential pressure of any individual jet pump differs from established single recirculation loop

. patterns by more than 10%.

4.4.1.1.2.7 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.

1 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.

¹ 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 At least once per 18 months (555 days), 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 required surveillances.

+

The LPRM upscale alarms are not required to be OPERABLE to meet this specification in OPERATIONAL CONDITION 2.

SUSQUEHANNA - UNIT 2 3/4 4-1f Amendment No. 91

3/4. 1 REACTIVITY CONTROL SYSTEMS BASES 3/4. 1. 1 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made

'subcritical from all operating conditions,

2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and 3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

Since core reactivity values will vary through core life as a function of fuel depletion and poison burnup, the demonstration of SHUTDOWN MARGIN will be performed in the cold, xenon-free condition and shall show the core to be subcritical by at least R + 0.38K delta k/k or R + 0.28K delta k/k, as appro-priate.

The value of R in units of I delta k/k is the difference between the beginning of cycle shutdown margin minus the minimum shutdown margin in the cycle, where shutdown margin is a positive number.

The value of R must be positive or zero and must be determined for each fuel loading cycle.

Two different values are supplied in the Limiting Condition for Operation to provide for the different methods of demonstration of the SHUTDOWN MARGIN.

The highest worth rod may be determined analytically or by test.

The SHUTDOWN MARGIN is demonstrated by control rod withdrawal at the beginning of life fuel cycle conditions, and, if necessary, at any future time in the cycle if= the first demonstration indicates that the required margin could be reduced as a function of exposure.

Observation of subcriticality in this condition assures subcriticality with the most reactive control rod fully withdrawn.

This reactivity characteristic has been a basic assumption in the analysis of plant performance and can be best demonstrated at the time of fuel loading, but the margin must also be determined anytime a control rod is incapable of insertion.

3/4. 1. 2 REACTIVITY ANOMALIES Since the SHUTDOWN MARGIN requirement is small, a careful check on actual reactor conditions compared to the predicted conditions is necessary.

Any changes in reactivity from that of the predicted (predicted core k ff) can be eff determined from the core monitoring system (monitored core k ff).

In the eff

'bsence of any deviation in plant operating conditions or reactivity anomaly, these values should be essentially equal since the calculational methodologies are consistent.

The predicted core k

is calculated by a 3D core simulation eff code as a function of cycle exposure.

This is performed for projected or anticipated reactor operating states/conditions throughout the cycle and is usually done prior to cycle o'peration.

The monitored core k

is the k ff as eff calculated by the core monitoring system for actual plant conditions.

C

,SUSQUEHANNA - UNIT 2 B 3/4 1"1 (glLL4

( '.,~,')1 u L '.).w Amendment No. 31 p

JQLM~ LING)~

i.( '..

~l Q:T~ajjl.

REA TIV TR L

Y TEM BASE REA TIVI AN IVIALIE (Continued)

Since the comparisons are easily done, frequent checks are not an imposition on normal operation.

A 1% deviation in reactivity from that of the predicted is larger than expected for normal operation, and therefore should be thoroughly evaluated.

A deviation as large as 1% would not exceed the design conditions of the reactor.

4.1.

NTR LR D

The specification of this section ensure that (1) the minimum SHUTDOWN MARGIN is maintained, (2) the control rod insertion times are consistent with those used in the accident

analysis, and (3) limit the potential effects of the rod drop accident.

The ACTION statements permit variations from the basic requirements but at the same time impose more restrictive criteria for continued operation.

A limitation on inoperable rods is set such that the resultant effect on total rod worth and scram shape will be kept to a minimum.

The requirements for the various scram time measurements ensure that any indication of systematic problems with rod drives will be investigated on a timely basis.

Damage within the control rod drive mechanism could be a generic problem, therefore with a control rod immovable because of excessive friction or mechanical interference, operation of the reactor is limited to a time period which is reasonable to determine the cause of the inoperability and at the same time prevent operation with a large number of inoperable control rods.

Control rods that are inoperable for other reasons are permitted to be taken out of service provided that those in the nonfully-inserted position are consistent with the SHUTDOWN MARGIN requirements.

The number of control rods permitted to be inoperable could be more than the eight allowed by the specification, but the occurrence of eight inoperable rods could be indicative of a generic problem and the reactor must be shutdown for investigation and resolution of the problem.

The control rod system is designed to bring the reactor subcritical at a rate fast enough to prevent the MCPR from becoming less than the limit specified in Specification 2.1.2 during the core wide transient analyzed in the cycle specific transient analysis report.

The MCPR operating limits in Specification 3.2.3 are a function of average scram speed.

Therefore, the results of the required scram time testing (Specification 4.1.3.3) are used to adjust the MCPR operating limits to assure the validity of the cycle specific transient analyses.

This ultimately assures that MCPR remains greater than the limitspecified in Specification 2.1.2.

The occurrence of scram times longer then those specified should be viewed as an indication of a systematic problem with the rod drives and therefore the surveillance interval is reduced in order to prevent operation of the reactor for long periods of time with a potentially serious problem.

The scram discharge volume is required to be OPERABLE so that it will be available when needed to accept discharge water from the control rods during a SUSQUEHANNA - UNIT 2 B 3/4 1-2 Amendment No. 91'

~

. REACTIVITY CDNTROL SYSTEMS BASES CONTROL RODS (Continued) reactor scram and will isolate the reactor coolant system from the containment when required.

Control rods with inoperable accumulators are declared inoperable and Spe-cification 3.1.3. 1 then applies.

This prevents a pattern of inoperable accumu-lators that would result in less reactivity insertion on a scram than has been analyzed even though control rods with inoperable accumulators may still be in-seWed with normal drive water pressure.

Operability of the accumulator ensures that there is a means. available to insert the control rods even under the most unfavorable depressurization of the reactor.

Control rod coupling integrity is required to ensure compliance with the analysis of the rod drop accident in the FSAR.

The overtravel position feature provides the only positive means of determining that a rod is properly coupled and therefore this check must be performed prior to achieving criticality after completing CORE ALTERATIONS that could have affected the control rod coupling integrity.

The subsequent check is performed as a backup to the initial demonstration.

In order to ensu~e that the control rod patterns can be followed and therefore that other parameters are within their limits, the control rod position indication system must be OPERABLE.

The control rod housing support restricts the outward movement of a control rod to less than 3 inches in the event of a housing failure.

The amount of rod reactivity which could be added by this small amount of rod withdrawal is less than a normal withdrawal increment and will not contribute to any damage to the primary coolant system.

The support is not required when there is no pressure to act as a driving force to rapidly eject a drive housing.

The required surveillance intervals are adequate to determine that the rods are OPERABLE and not so frequent as to cause excessive wear on the system components.

3/4. 1.4 CONTROL ROD PROGRAM CONTROLS Control rod withdrawal and insertion sequences are established to assure that the maximum insequence individual control rod or control rod segments which are withdrawn at any time during the fuel cycle could not be worth enough to result in a peak fuel enthalpy greater than 280 cal/gm in the event of a control rod drop accident.

The specified sequences are characterized by homogeneous, scattered patterns of control rod withdrawal." When THERMAL POWER is greater than 20K of RATED THERMAL POWER, there is no possible rod worth which, if dropped at the design rate of the velocity limiter, could result in a peak enthalpy of 280 cal/gm.

This requiring the RSCS and RWM to be OPERABLE when THERMAL POWER is less than or equal to 20K of RATED THERMAL POWER provides adequate control.

SUSQUEHANNA - UNIT 2 B 3/4 1-3 C'mendment No. 31

~J I BD Ilk

REA TIVI TR L SYSTEMS 0

BASES NTR L

D R

RAM CONTROLS (Continued)

The RSCS and RWM logic automatically initiates at the low power setpoint (20% of RATED THERMALPOWER) to provide automatic supervision to assure that out-of-sequence rods will not be withdrawn or inserted.

Parametric Control Rod Drop Accident analyses have shown that for a wide range of key reactor parameters (which envelope the operating ranges of these variables),

the fuel enthalpy rise during a postulated control rod drop accident remains considerably lower than the 280 cal/gm limit. For each operating cycle, cycle-specific parameters such as maximum control rod worth, Doppler-coefficient, effective delayed neutron fraction, and maximum four-bundle local peaking factor are compared with the inputs to the parametric analyses to determine the peak fuel rod enthalpy rise.

This value is then compared against the 280 cal/gm design limit to demonstrate compliance for each operating cycle.

If cycle-specific values of the above parameters are outside the range assumed in the parametric analyses, an extension of the analysis or a cycle-specific analysis may be required.

Conservatism present in the analysis, results of the parametric studies, and a detailed description of the methodology for performing the Control Rod Drop Accident analysis are provided in PL-NF-90-001, and XN-NF-80-19 Volume 1.

The RBM is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power operation. Two channels are provided. Tripping one of the channels willblock erroneous rod withdrawal soon enough to prevent fuel damage.

This system backs up the written sequence used by the operator for withdrawal of control rods.

~

4.1 TANDBY LI ID NTR L SY TEM The standby liquid control system provides a backup capability for bringing the reactor from

'ull power to a cold, Xenon-free shutdown, assuming that none of the withdrawn control rods can be inserted.

To meet this objective it is necessary to inject'a quantity of boron which produces a concentration of 660 ppm in the reactor core in approximately 90 to 120 minutes.

A minimum quantity of 4587 gallons of sodium pentaborate solution containing a minimum of 5500 lbs. of sodium pentaborate is required to meet this shutdown requirement.

There is an additional allowance of 165 ppm in the reactor core to account for imperfect mixing. The time requirement was selected to override the reactivity insertion rate due to cooldown following the Xenon poison peak and the required pumping rate is 41.2 gpm.

The minimum storage volume of the solution is established to allow for the portion below the pump suction that cannot be inserted and the filling of other piping systems connected to the reactor vessel.

The temperature requirement for the sodium pentaborate solution is necessary to ensure that the sodium pentaborate remains in solution.

With redundant pumps and explosive injection valves and with a highly reliable control rod scram system, operation of the reactor is permitted to continue for short periods of time with the system inoperable or for longer periods of time with one of the redundant components inoperable.

SUSQUEHANNA - UNIT 2 B 3/4 1-4 Amendment No. 91

4.2 E

Dl IBUTI N LIMIT BASE 4,2 VERA E

LANAR LINEAR HEAT GENERATION RATE This specification assures that the peak cladding temperature followingthe 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.

The Technical Specification APLHGR for SNP 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 Figure 3.2.1-1.

The calculational procedure used to establish the APLHGR shown on Figure 3.2.1-1 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 XN-NF-80-19, Volumes 2, 2A, 28 and 2C.

4 2 2 APRM TP INT 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 setpoints must be adjusted to ensure that a1% plastic strain and fuel centerline melting do not occur during the worst anticipated operational occurrence (AOO),

including-transients initiated from partial power operation.

For SNP 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 or Figure 3.2.2-2 as applicable.

The LHGR versus exposure curve in Figures 3.2.2-1 and 3.2.2-2 are based on SNP's Protection Against Fuel Failure (PAFF) lines shown in Figure 3.4 of XN-NF-85-67(A),

Revision 1 and Figure A.2 of PL-NF-92-001.

Figures 3.2.2-1 and 3.2.2-2 correspond 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. $1

P WE DI TRIB T N LIMIT BASE 4.2.

MIN M RITI AL P WER RATI The required operating limit MCPRs at steady state operating conditions as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safety Limit MCPR, and analyses of abnormal operational transients.

For any abnormal operational transient analysis with the initial condition 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 settings given in Specification 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).

The 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.

The required MCPR Operating Limits as a function of core power, core flow and plant equipment available condition are presented in Figures 3.2.3-1 through 3.2.3-4.

The transient analyses to determine the MCPR operating limits are performed using NRC approved methods described in PL-NF-90-001 and corresponding supplements.

The pressurization events were analyzed based on more realistic scram speed as well as on the Technical Specification 3.1.3.3 limits., The MCPR operating limits are specified as a function of scram speed.

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 transient resulting from a motor-generator speed control failure. The flow dependent MCPR is only calculated for the manual flow control mode.

Therefore, automatic flow control operation is not permitted.

Figures 3.2.3-2, 3.2.3-3 and 3.2.3-4 define the power dependent MCPR operating limits which assure that the Safety limit MCPR will not be violated in the event of a Feedwater Controller Failure, rod Withdrawal Error, or Load Reject Without Main Turbine Bypass operable initiated from a full power or reduced power condition.

Cycle specific analyses are performed for the most limiting local and core wide transients 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. 91

0 4.4 REA T R

LANT Y TEM BASE 4 4.1 RE IR LATION YSTEM 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.

LOCA analyses for two loop operating conditions, which result in Peak Cladding Temperatures (PCTs) below 2200'F, bound single loop operating conditions.

Single loop operation LOCA analyses using two-loop MAPLHGR limits result in lower PCTs.

Therefore, the use of two-loop MAPLHGR limits during single loop operation assures that the PCT during a LOCA event remains below 2200'F.

The MINIMUMCRITICALPOWER RATIO (MCPR) limits for single loop operation assure that the Safety Limit MCPR is not exceeded for any Anticipated Operational Occurrence (AOO).

In addition, the MCPR limits for single-loop operation protect against the effects of the Recirculation Pump Seizure Accident. That is, for operation in single-loop with an operating MCPR limit a 1.30, the radiological consequences of a pump seizure accident from single-loop operating conditions are but a small fraction of 10CFR100 guidelines.

For single loop operation, the RBM and APRM setpoints are adjusted by a 8.5% decrease 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 operation in the single loop mode.

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

Specifications have been provided to prevent, detect, and mitigate core thermal hydraulic instability events.

These specifications are prescribed in accordance with NRC Bulletin 88-07, Supplement 1, "Power Oscillations in Boiling Water Reactors (BWRs)," dated December 30, 1988.

LPRM upscale alarms are required to detect reactor core thermal hydraulic instability events.

The criteria for determining which LPRM upscale alarms are required is based on assignment of these alarm to designated core zones.

These core zones consist of the level A, B and C alarms in 4 or 5 adjacent LPRM strings.

The number and location of LPRM strings in each zone assure that with 50% or more of the associated LPRM upscale alarms OPERABLE sufficient monitoring capability is available to detect core wide and regional oscillations.

Operating plant instability data is used to determine the specific LPRM strings assigned to each zone.

The core zones and required LPRM upscale alarms in each zone are specified in appropriate procedures.

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 refloading 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.

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

REACTOR COOLANT SYSTEM BASES Recirculation pump speed mismatch limits are in compliance with the ECCS LOCA analysis design cr iteria 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 1454F.

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 valveh is required to limit reactor pressure to within ASME III allowable values for-the worst case upset transient.

Oemonstration of the safety/relief valve lift settings 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 OETECTION SYSTEMS The RCS leakage detection systems required by this specification are, provided to monitor and detect leakage from the reacto~ 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 UNIOENTIFIEO 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 BOUNOARY LEAKAGE, the reactor will be shutdown to allow fur ther 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 L'OCA.

SUSQUEHANNA - UNIT 2 B 3/4 4-2 Amendment No 6O NUV 22 l989

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r FIGURE 5.1.3-1b NP DEFINING UNRESTRICTED AREAS t'OR RADIOACTIVE GASEOUS AND L'IgUID EFFLUENTS SUSgUEHI8.'N - UNIT 2 5-5

0'ESI N FEAT RES F

EL A EMBLIE 5.3.1 The reactor core shall contain 764 fuel assemblies.

Allfuel assemblies shall contain 79 fuel rods and two Zircaloy-2 water rods.

Each fuel rod shall be clad with Zircaloy-2 and have a nominal active fuel length of 150 inches.

Reload fuel shall have a maximum average enrichment of 4.0 weight percent U-235.

NTR L R D A EIVIBL E 5.3.2 The reactor core shall contain 185 control rod assemblies consisting of two different designs.

The "original equipment" design consists of a cruciform array of stainless steel tubes containing 143 inches of boron carbide (B4C) powder surrounded by a stainless steal sheath.

The "replacement" control blade design consists of a cruciform array of stainless steal tubes containing 143 inches of boron carbide (B4C) powder near the center of the cruciform, and 143 inch long solid hafnium rods at the edges of the cruciform, all surrounded by a stainless steel sheath.

.4 EA T R

LANT Y TEM DE I

N PRE RE A D TEMPERAT RE 5.4.1 The reactor coolant system is designed and shall be maintained:

a.

In accordance with the code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements, b.

For a pressure of:

1.

1250 psig on the suction side of the recirculation pumps.

2.

1500 psig from the recirculation pump discharge to the jet pumps.

c.

For a temperature of 575'F.

~VL~QME 5.4.2 The total water and steam volume of the reactor vessel and recirculation system is approximately 22,400 cubic feet at a nominal Ta<<of 528'F.

SUSQUEHANNA - UNIT 2 5-6 Amendment No. 91

ML17157C054

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RESPONSE

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