Text

Amend 90 to License NPF-14,changing Tech Specs to Support Cycle 5 Operations W/Advanced Nuclear Fuel Corp 9x9 Reload Fuel
	ML17156B163
Person / Time
Site:	Susquehanna
Issue date:	05/15/1989
From:	Butler W; Office of Nuclear Reactor Regulation
To:	;
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ML17156B164	List: ML17156B163; ML17156B165;
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	NUDOCS 8905250434
	Download: ML17156B163 (37)
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UNITED STATES UCLEAR REGULATORY COMMISSIO WASHINGTON, D. C. 20555 PENNSYLVANIA POWER 8 LIGHT COMPANY ALLEGHENY ELECTRIC COOPERATIVE INC.

DOCKET NO. 50-387 SUSQUEHANNA STEPS ELECTRIC STATION UNIT I AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.

90 License No. NPF-14 1.

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

A.

The application for the amendment filed by the Pennsylvania Power 5

Light Company, dated February 2, 1989 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 ComIission; 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 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-14 is hereby amended to read as follows:

I (2)

Technical S ecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 90 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.

8905250434 890515 PDR ADOCK 05000387 P

PDC 3.

This license amendment is effective as of its date of issuance to be implemented upon startup for Cycle 5 operations currently scheduled for June 2, 1989.

FOR THE NUCLEAR REGULATORY COMMISSION

Attachment:

Changes to the Technical Specifications Date of Issuance:

May 15, 1989

/s/

Walter R. Butler, Director Project Directorate I-2 Division of Reactor Projects I/II 89 DI-2/PM MThadani:mr g /9-/89 OG, PDI-2/0 utl r P/$ /89 P/j g/89

3.

This license amendment is effective as of its date of issuance to be implemented upon startup for Cycle 5 operations currently scheduled for June 2, 1989.

FOR THE NUCLEAR REGULATORY COMMISSION Halter R. Butler, Director Project Directorate I-2 Division of Reactor Projects I/II

Attachment:

Changes to the Technical Specifications Date of Issuance:

May 15, 1989

ATTACHMENT TO LICENSE AMENDMENT NO. 90 FACILITY OPERATING LICENSE NO. NPF-14 DOCKET NO. 50-387 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.*

RENOVE 111 1V INSERT iii*

1V XX1 xxii 8 2-1 8 2-la 8 2-2 XX1 xxii*

2-3*

2-4 8 2-1 8 2-la*

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 2-5a 3/4 2-6 3/4 2-7 3/4 2-9 3/4 2-9a 3/4 2-10 3/4 2-10a 3/4 2-10b 3/4 2-10c 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 2-9

~

3/4 2-9a 3/4 2-10 3/4 2-10a 3/4 2-10b 3/4 2-10c ATTACHMENT TO LICENSE AMENDMENT NO.

FACILITY OPERATING LICENSE NO. NPF-14 DOCKET NO. 50-387 3/4 3-53 3/4 3-54 3/4 4-1b 3/4 4-1c 8 3/4 2-1 8 3/4 2-2 8 3/4 4-1 8 3/4 4-2 3/4 3-53*

3/4 3-54 3/4 4-1b*

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

IN0EX SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION

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

THERMAL POWER, High Pressure and High Flow.

Reactor Coolant System Pressure.

Reactor Vessel Water Level...

~

PAGE 2-1 2-1 2-1

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

2-3 BASES

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

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

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

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

B 2-3 B 2-3 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Protection System Instrumentation Setpoints........

B 2-4 SUSQUEHANNA - UNIT 1 Amendment No.

B2 AUG 8 0 1988

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

3/4. 1 REACTIVITY CONTROL SYSTEMS 3/4. l. 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..............

Four Control Rod Group Scram Insertion Times...........

4 3/4 1-7 3/4 1-8 Control Rod Scram Accumulators..

~..

3/4 1-9 Control Rod Drive Coupling.............................

Control Rod Position Indication..............

Control Rod Drive Housing Support......................

3/4. 1.4 CONTROL ROD PROGRAM CONTROLS 3/4 1-11 3/4 1-13 3/4 1"15 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 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 2-5 3/4. 2. 3 MINIMUM CRITICAL POWER RATIO................

3/4 2-6 3/4.2. 4 LINEAR HEAT GENERATION RATE ANF FUEL.......................

3/4 2-10a SUSQUEHANNA - UNIT 1 iv Amendment No. 90

LIST OF FIGURES INDEX FIGURE

3. l. 5-1

3. l. 5-2

3. 2. 1-1 3

~ 2. 1-2

3. 2. 2-1

3. 2. 3-1

3. 2. 3"2

3. 2. 4"1

3. 2. 4-2

3. 4. l. 1-1

3. 4. 6. 1-1 B 3/4 3-1 B 3/4.4.6"1 5.1. 1-1

5. 1. 2"1

5. 1. 3-la 5.1. 3"lb

6. 2. 1-1 6.2. 2-1 SODIUM PENTABORATE SOLUTION TEMPERATURE/

CONCENTRATION REQUIREMENTS SODIUM PENTABORATE SOLUTION CONCENTRATION.........

THIS PAGE INTENTIONALLY LEFT BLANK..

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

AVERAGE BUNDLE EXPOSURE, ANF Bx8 FUEL MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VS.

AVERAGE BUNDLE EXPOSURE, ANF 9x9 FUEL

~

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

FLOW DEPENDENT MCPR OPERATING LIMIT.

REDUCED POWER MCPR OPERATING LIMIT LINEAR HEAT GENERATION RATE (LHGR) LIMIT VERSUS AVERAGE PLANAR EXPOSURE ANF 8x8 FUEL..

LINEAR HEAT GENERATION RATE (LHGR) LIMIT VERSUS AVERAGE PLANAR EXPOSURE, ANF 9x9 FUEL...............

THERMAL POWER LIMITATIONS MINIMUM REACTOR VESSEL METAL TEMPERATURE VS.

REACTOR VESSEL PRESSURE REACTOR VESSEL WATER LEVEL.

FAST NEUTRON FLUENCE (E)lMeV) AT 1/4 T AS A FUNCTION OF SERVICE LIFE 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....

OFFSITE ORGANIZATION.. ~....

UNIT ORGANIZATION.

PAGE 3/4 1-21 3/4 1-22 3/4 2-2 I

l 3/4 2-4 3/4 2-4a 3/4 2"7 3/4 2-9 3/4 2-9a 3/4 2-10b I

3/4 2-10c 3/4 4-lb 3/4 4-18 B 3/4 3-8 B 3/4 4-7 5-2 5-3 5"5 6-3 6"4 SUSQUEHANNA - UNIT 1 XX1 Amendment No.

9O

lg LIST OF TABLES TASLE 1.2 INDEX SURVEILLANCE FREQUENCY NOTATION OPERATIONAL CONDITIONS PAGE 1-9 1-10 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 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOIHTS o ~

~

~ ~

~ ~ ~ ~

~

~ ~

~

~ ~

~

REACTOR PROTECTION SYSTEM IHSTRUMEHTATIOH.........

2-4 l

3/4 3-2 REACTOR PRDTECTIOH SYSTEM RESPONSE TIMES,.........

3/4 3-6 REACTOR PROTECTION.SYSTEM IHSTRUMENTAT1ON SURVEILLANCE REQUIREMENTS..

3/4 3-7 ISOLATIOH ACTUATION IHSTRUMEHTATION SURVEILLANCE REQUIREMENTS 3/4 3 23 ISOLATIOH ACTUATIOH INSTRUMEHTATION...............

3/4 3-11 ISOLATION ACTUATIOH INSTRUMENTATION SETPOINTS....,

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

3/4 3-21 3,3.3 1

3 ~ 3 3 2 EMERGENCY CORE COOLING SYSTEM ACTUATIOH INSTRUMEHTATION ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

EMERGENCY CORE COOLING SYSTEM ACTUATION IHSTRUMEHTATIOH SETPOIHTS.........................

3/4 3 28 3/4 3-31 3.3.3 3

4. 3.3. 1-1 3.3.4. 1-1 EMERGENCY CORE COOLING SYSTEM ACTUATION IHSTRNENTATION SURVEI LLAHCE REQUIREMENTS.........

ATi5 RECIRCULATION PtNP TRIP SYSTEM 3/4 3-34 EMERGENCY CORE COOLING'YSTEM RESPOHSE TIMES

~" ~ ~ ~

3/4 3-33 3.3.4.1 2 NSTRSWATION

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

I AAS RECIRCULATION PLNP TRIP SYSTEM IHSTRQKHTATION SETPOINTS

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 0 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

3/4 3-37 3/4 3-38 SUSQUENANHA UNIT 1 Agen~ne Ho 72 7

C95N

SRFETY LIMITS ANQ ITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY. SYSTEM SETTINGS

~REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS 2.2. 1 The reactor protection system instrumentation setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2. 1-1.

APPLICABILITY:

As shown in Table 3.3.1-1.

ACTION:

With a reactor protection system instrumentation setpoint less conservative than the value shown in the Allowable Values column of Table 2.2. 1-1, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3. 1 until the channel is restored to OPERABLE status with its setpoint ad'justed consistent with the Trip Setpoint value.

SUSgUEHANN UNIT 1

TABLE 2.2.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS 3.

c.

Neutron Flux-Upscale d.

Inoperative Reactor Vessel Steam Dome Pressure

- High Reactor Vessel Water Level - Low, Level 3

5.

Main Steam Line Isolation Valve - Closure 6.

Main Steam Line Radiation - High 7.

8.

Drywell Pressure

- High Scram Discharge Volume Water Level - High a.

Level Transmitter b.

Float Switch FUNCTIONAL UNIT l.

Intermediate Range Monitor, Neutron Flux-High m

2.

Average Power Range Monitor:

I a.

Neutr on Flux-Upscale, Setdown C

b.

Flow Biased Simulated Thermal Power-Upscale (1)

Flow Biased (2)

High Flow Clamped TRIP SETPOINT

< 120/125 divisions of full scale

< 15K of RATED THERMAL POWER

< 0.58 W+59X, with a maximum of

< 113.5X of RATED THERMAL POWER

< 118K of RATED THERMAL POWER NA

< 1037 psig

> 13.0 inches above instrument zero*

< lOX closed

< 7.0 x full power

Background

< 1.72 psig

< 88 gallons

< 88 gallons ALLOWABLE VALUES

< 122/125 divisions of full scale

< 20K of RATED THERMAL POWER

< 0.58 W+62X, wi a maximum of

< 115 5X of RATED THERMAL POWER

< 120K of RATED THERMAL POWER NA

< 1057 psig

> ll 5 inches above instrument zero

< illclosed

< 8.4 x full power background

< 1.88 psig

< 88 gallons

< 88 gallons O

10.

12.

Manual Scram Turbine Stop Valve - Closure Turbine Control Valve Fast Closure, Trip Oil Pressure

- Low ll.

Reactor Mode Switch Shutdown Position

< 5.5X closed

> 500 psig NA

< 7X closed

> 460 psig NA ED

See Bases Figure 8 3/4 3-1.

See Specification 3.4.1.1.2.a for single loop operation requirement.

BASES

2. 0 INTROOUCTION 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 Specifications

2. 1.2 for ANF 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 99.9X of the fuel rods in the core do not experience transition boiling (ref. XN-NF-524(A) Revision 1).

I

2. 1. 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 106 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 9x9 fuel design, the minimum bundle flow is greater than 30,000 lbs/hr.

For the ANF 8x8 fuel, the minimum bundle flow is greater than 28,000 lbs/hr.

I 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 cri-tical 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 SUSQUEHANNA - UNIT 1 B 2-1 Amendment No.

90

2.1 SAFETY LINITS BASES

2. l. 1 THERHAL POWER Low Pressure or Low Flow (Continued) 25X 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

THERHAL POWER limit of 25X of RATED THERHAL POWER for reactor pressures below 785 psig is conservative.

SUSQUEHANNA - UNIT l B 2-la Aaendient No. 72 OCT 9 ggy

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 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 99.9X 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 detail-ed 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 describes the methodology 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 Sec-tion B 2. l. 1), the assumed reactor conditions used in defining the safety limit introduce conservatism into the limit because bounding high radial power fac-tors 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 organiza-tions 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.

SUSQUEHANNA - UNIT 1 B 2-2 Amendment No.

9O

0 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 each type of fuel as a function of AVERAGE BUNDLE EXPOSURE for ANF fuel shall not exceed the limits shown in Figures 3.2. 1-1 and 3.2. 1-2.

APPLICABILITY:

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

EETER ER.

ACTION:

With an APLKGR exceeding the limits of Figure 3.2.1-1 or 3.2.1-2, 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, and 3.2. 1-2.

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 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 1 Amendment No.

90

14 13 12 11 10 a

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13.0:.::.:: :':.:::-::13.0

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.:....:... 35,000;..

104 nn 10.4

- '- '-::PERMISSABLE- '" '- '-- '- '" '" " '" '- '" "

REGION OF OPERATION

~

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0

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0 6000 10000 16000 20000 26000 Average Bundle Exposure (MWD/MT) 30000 36000 MAXIMUMAVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE BUNDLE'EXPOSURE ANF 8X8 FUEL FIGURE 3.2.1-1

AD 12

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PERMISSABLE::::

REGION OF OPERATION

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20,000; 10.2 26,000;.:

ae 30,000;:

8.8

.. 40,000; 7.6 36,000;,:...:.....

8.2 0

6000 10000 15000 20000 26000 30000 35000 40000 Average Bundle Exposure (MWD/MT)

O MAXIMUMAVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE BUNDLE EXPOSURE ANF 9X9 FIDEL RGURE 3.2."l-2

LEFT INTENTIONALLYBLANK SUSQUEHANNA - UNIT 1 3/4 2-4 Amendment No.

gp

POWER DISTRIBUTION LIMITS 3/4.2.2 APRM SETPOINTS

'LIMITING CONDITION FOR OPERATION 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 Set oint¹ Allowable Value S

8W + 59K)T Rl SRB

< (0.58W + 50K)T SRB

< (0.58W + 53K)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 lbs/hr, T

= Lowest value of the ratio of FRACTION OF RATED THERMAL POWER divided by the MAXIMUM FRACTION OF LIMITING POWER DENSITY.

The FLPD for ANF fuel is the actual LHGR divided by the LINEAR HEAT GENERATION RATE from Figure 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 RATE TIIE MAI PRIIER.

ACTION:

With the APRM flow biased simulated thermal power-upscale scram trip setpoint and/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 above determined, initiate corrective action within 15 minutes and adjust S and/

or SRB to be cons i stent with the Tr ipSetpoi nt val ue" within2 hours 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 .

>th greater than the FRTP during power ascension up to 90K 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 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 10K 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 operation requirements.

SUSQUEHANNA - UNIT 1 3/4 2-5 Amendment No.

90

POWER DISTRIBUTION LIMITS 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 siaulated 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:

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 coepletion 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 MFLPD greater than. or equal to FRTP.

d.

The provisions of Specification 4.0,4 are not applicable.

SUSQUEHANNA - UNIT 1 3/4 2-Sa

~ndaent No. 72

18 a~

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16.0 26,400;:.

14.0

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'......':... 43,200; 8.0 48,000;.....:

8.3 10000 20000

$0000 40000 Average Planar Exposure (MWD/MT) 50006 LINEAR HEAT GENERATION RATE FOR APRH SETPOINTS VERSUS AVERAGE PLANAR EXPOSURE ANF FUEL FIGURE 3.2.2-1 SUSQUEHANNA - UNIT 1 3/4 2-6 Amendment No.

9O

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 THERMAL POWER is greater than or E

E T

ERRA R.

ACTION'ith MCPR less than the applicable MCPR limit determined above, initiate cor-rective 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 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 'URVEILLANCE 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 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 operat-ing with a LIMITING CONTROL ROD PATTERN for MCPR.

d.

The provisions of Specification 4. 0.4 are not applicable.

SUS(UEHANNA - UNIT 1 3/4 2-7 Amendment No.

90

1.7 (40,1.52) cs 1.6

~~

Q.0 tK (47.5,1A0) 0-O (46.25, 1.42)

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

ass Operable A

8 1.42 1.40 1.3 (50.1.36) 52.7.1.33)

C 1.33 1.2 40 50 eo 70

~

80 Total Core Flow (% OF RATED) 90 100 FLOW DEPENDENT MGPR OPERATING LIMIT FIGURE 3.2.3-1

CA C/l AD rn 1.7 1.8 (26,1.66)

(40.1.63)

CURVE A: EOC-RPT Inoperable:

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

~~

1.6 C

~~

CQ GJ 0 1.4 O

(26,1.63)

(26,1A6)

(40,1 43)

(40,1.61)

(86,1.60)

(66.1.48)

(86.1.40)

(80.1.47)

(80,1.46) 1.42 1.40 1.3 (80,1.37)

C (96,1.33) 1.33 O

1.2 20 30 80 60 60 70 Core Power (% OF RATED)

REDUCED POWER MCPR OPERATING LIMIT Figure 3.2.3-2 90 100

THIS PAGE IS DELETED.

SUSQUEHANNA - UNIT 1 3/4 2-10 Amendment No.

g0

~ (

I

POWER DISTRIBUTION LIMITS 3/4.2.4 LINEAR HEAT GENERATION RATE

'ANF FUEL LIMITING CONDITION FOR OPERATION 3.2.4 The LINEAR HEAT GENERATION RATE (LHGR) for ANF fuel shall not exceed the LHGR limit determined from Figures 3.2.4-1 and 3.2.4-2.

APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than X fRTE E

OE.

ACTION:

With the LHGR of any fuel rod exceeding its applicable limit from Figure 3.2.4-1 or 3.2.4-2, initiate corrective action within 15 minutes and restore the LHGR to within the limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less 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 LHGRs for ANF fuel shall be determined to be equal to or less than the 1 imit:

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 on a LIMITING CONTROL ROD PATTERN for LHGR.

d.

The provisions of Specification 4.0.4 are not applicable.

SUSQUEHANNA - UNIT 1 3/4 2-10a Amendment No.

9O

C/l Cl)

AD m

~~

E lUK C0 C

C9 COIx L

CQO C:

18 16 14 12 10

.0.0; 1B.O 25,400;:

14.1

PERMISSABLE REGION OF OPERATION 42,000; S.3 O

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

LINEAR HEAT GENERATION RATE (LHGR) LIMIT VERSUS AVERAGE PLANAR EXPOSURE ANF 8X8 FUEL FIGURE 3.2.4-1 60000

C/l C/l AD I

C)

O

~~

E lO LK C0 C

C9 CQ LtgO C

18 14 12 10 B

d 0.0;.:

13.0 I

I

24,000;.:

12.0

PERMISSABLE::

REGION OF OPERATION 35,000; 9 {)

48,000; 7.72 O

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

LINEAR HEAT GENERATION RATE (LHGR) LIMIT VERSUS AVERAGE PLANAR EXPOSURE ANF 9XB FUEL FIGURE 3.2.4-2 60000

TABLE 3.3.6-1 (Continued)

CONTROL ROO BLOCK INSTRUMENTATION ACTION ACTION 60 Oeclare the RBH inoperable and take the ACTION required by Specification 3.1.4.3.

ACTION 61 With the number of OPERABLE Channels:

a.

One less than required by the Hinimum 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 Hinimum 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 Hinimum OPERABLE Channels per Tr.p 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 > 3'f RATEO THERHAL POWER.

With'ore 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 or fewer fuel assemblies (adjacent to the SRHs) are in the core.

The RBH shall be automatically bypassed when a peripheral control rod is selected or the reference APRH channel indicates less than 30" of RATEO THERHAL POWER.

b.

This function shall be automatically bypassed if detector count rate is

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

Ce d.

This function is autoeatically bypassed when the associated IRH channels are on range 8 or higher.

This function is automatically bypassed when the IRH channels are on range

'3 or higher.

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

SuSquaauue - UNIT X 3/4 3-53 Amendment No. 43

TABLE 3.3.6-2 CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS c=

TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE QJ 3

ColI 4.

5.

ROD BLOCK MONITOR a.

Upscale ¹¹ b.

Inoperative c.

Downscale APRH a.

Flow Biased Neutr~y Flux - Upscale b.

Inoperative c.

Downscale d.

Neutron Flux Upscale Startup SOURCE RANGE MONITORS a.

Detector not full in b.

Upscale c.

Inoperative d.

Downscale INTERMEDIATE RANGE HONITORS a.

Detector not full in b.

Upscale c.

Inoperative d.

Downscale SCRAH DISCHARGE VOLUME

< 0.66 W+ 42K

< 0.66 W + 45K HA NA

> 5/125 divisions of full scale

> 3/125 of divisions full scale

< 0.58 W + 50K" HA

> 5X of RATED THERHAL POWER

< 12K of RATED THERMAL POWER

< 0.58 W + 53K*

NA

> 3X of RATED THERMAL POWER

< 14K of RATED THERMAL POWER NA

< 2x10 cps HA

> 0.7 cps"*

NA<4x10 cps HA

> 0 5 cps"*

NA NA

< 108/125 divisions of full scale

< 110/125 division of full scale RA NA

> 5/125 divisions of full scale

> 3/125 divisions of full scale O

a.

Water Level - High

< 44 gallons

< 44 gallons CD 6.

REACTOR COOLANT SYSTEM RECIRCULATION FLOW a.

Upscale

< 108/125 divisions of full scale

< ill/125 divisions of full scale b.

Inoperative NA NA c.

Comparator

< lOX flow deviation

< 11K flow deviation he verage ower Range Honitor rod block function is varied as a function of recirculation loop flow (W).

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

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

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

¹¹See Specification 3.4.1. 1.2.a for single loop operation requirements.

e' Figure 3.4,1.1.1-1 THERMAL POWER LNITATIONS IO 70 lD 40 60 g

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SUSQUEHANNA UNIT 1 3/i 4-$

AII~nt Xo.

S6

REACTOR COOLANT SYSTEM RECIRCULATION LOOPS - SINGLE LOOP OPERATION LIMITING CONDITION FOR OPERATION

~

4 (p

3. 4. 1. l. 2 One reactor coolant recirculation loop shall be in operation with the pump speed

< 80K of the rated pump speed, and 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:

Tri Set oint

+

3.

Specification 3.2. 1:

The MAPLHGR limits Figures 3.2.1-2 and 3.2.1-3

~

4.

Specification 3.2.2:

the APRM Setpoints Allowable Value I

shall be as specified in I

shall be as follows:

Tri Set oint Allowable Value

'45)T ~)T SRB

< (0.58W + 45K)T SRB

< (0.58W + 48K)T 5.

Specification 3.2.3:

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

a.

1. 42, b.

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

the MCPR determined from Figure 3.2.3-2 plus 0.01.

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

+ 3 Allowable Value b.

APRM-Flow Biased Tri Set oint Allowable Value

+

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. l. 1. 1-1.

APPLICABILITY:

OPERATIONAL CONDITIONS 1" and 2", except during two loop operation.0 SUSQUEHANNA - UNIT 1 3/4 4-lc Amendment No.

gO 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. 1-1.

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

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

The peak cladding temperature (PCT) following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is dependent only secondarily on the rod to rod power distribution within an assembly.

The Technical Specifi-I ation APLHGR for ANF fuel is specified to assure the PCT following a postulated LOCA will not exceed the 2200'F limit.

The limiting value for APLHGR is shown in Figures 3.2.1-1 and 3.2.1-2.

I The calculational procedure used to establish the APLHGR shown on Figures 3.2.1-1, and 3.2.1-2 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-80-19, Volumes 2, 2A, 2B and 2C.

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

In addition, the APRM setpoints must be adjusted to ensure that

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

For 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 (PAFF) line shown in Figure 3.4 of XN-NF-85-67(A), Revision 1.

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

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

90

POWER DISTRIBUTION LIMITS BASES 3/4.2.3 MINIMUM CRITICAL POWER RATIO 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 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 Specification 2.2.

To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting tran-sients 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 and presented in Figures 3.2.3-1 and 3.2.3-2.

The evaluation of a given transient begins with the system initial param-eters 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 pressuriza-tion and non-pressurization events are described in XN-NF-79-71 and XN-NF-84-105.

The principal 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 exceeded 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.

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

At THERMAL POWER levels less than or equal to 25K of RATED THERMAL POWER, the reactor will be operating at minimum recirculation pump speed and the moderator void content will be very small.

For all designated control rod patterns which may be employed at this point, operating plant experience indi-cates that the resulting MCPR value is in excess of requirements by a consider-able margin.

During initial start-up testing of the plant, a

MCPR evaluation SUSQUEHANNA - UNIT 1 8 3/4 2"2 Amendment No.

9O

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, provi,ded 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 1.0 for ANF fuel.

This multiplication factor is derived from LOCA analyses initiated from single loop operation conditions.

The resulting MAPLHGR limits for single loop operation assure the peak cladding 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 8.5X 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 vibra-tion.

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

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

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

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

380, Revision 1, "BWR Core Thermal Hydraulic Stability,"

dated February 10, 1984.

An inoperable jet pump is not, in itself, a sufficient reason to declare a

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

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

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

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

The limits will ensure an adequate core flow 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 tem-perature difference was greater than 145'F.

SUS(UEHANNA - UNIT 1 8 3/4 4-1 Amendment No. 90

'r 3/4. 4 REACTOR COOLANT SYSTEM BASES Continued 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 ASHE Code.

A total of 10 OPERABLE safety-relief valves is required to limit reactor pressure to within ASHE III ~llow-able 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 Specification 4.0.5.

3/4.4.3 REACTOR COOLANT SYSTEH LEAKAGE 3/4.4.3.

1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are provided to monitnr 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 deteriin)ng 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

~dded assurance of valve integrity thereby reducing the probability of gross valve failure and consequent

)ntersystea LOCA.

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

Chloride li~its are specified to prevent stress corrosion cracking of the stainless steel.

The effect of chlor ide is not as great when the oxygen concentration in the coolant is low, thus the 0.2 ppa li~it on chlorides is permitted during REER OPERATION. 'uring shutdown and refueling operations, the temperature necessary for stress corrosion to occur is not present so

~ 0.5 ppe concentration of chlorides is not considered harmful during these periods SUSQUEHANNA UNIT 1 B 3/4 4 2 Amendment No. 56

ML17156B163

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