ML17158A230

Amend 103 to License NPF-22 Re Changes to TS Consisting of Raising Authorized Power Level from 3,293 Mwt to New Limit of 3,441 Mwt

ML17158A230
Person / Time
Site:	Susquehanna
Issue date:	04/11/1994
From:	Russell W Office of Nuclear Reactor Regulation
To:
Shared Package
ML17158A231	List: ML17158A230 ML17158A234
References
NUDOCS 9404250031
Download: ML17158A230 (92)

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 PENNSYLVANIA POWER 8E LIGHT COMPANY ALLEGHENY ELECTRIC COOPERATIVE INC.

DOCKET NO. 50-388 SUS UEHANNA STEAM ELECTRIC STATION UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.

103 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 It Light Company, dated November 24,

1993, as supplemented by letters dated January 7 and February 14,

1994, 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'onformity with the application, the provisions of the Act, and the regulations of the Commission; C.

There is reasonable assurance:

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

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

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

940425003i 94041i PDR ADOCK 05000388 P

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2.

3.

Accordingly, Facility Operating License No. NPF-22 paragraph 2.C. (1) is hereby amended to read as follows:

(1) Maximum Power Level P6nnsyl vanf a Power and Light Company (PP8 L) i s authori zed to operate the facility at reactor core power levels not in excess of 3441 mega-watts thermal: ('100K power) in accordance,.with the conditions specified herein and in Attachment 1 to.this"license.

The, preoperational

tests, startup tests and other items identified in Attachment 1 to this license"shall'e completed as specified.

Attachment 1 is hereby incorporated into this license.

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Further, the license is amended by changes to the Technical Specifications as indicated

-in the attachment to this license, amendment and paragraph, 2.C.(2) of the Facility Operating License No. NPF-22 is hereby amended to read as follows:

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(2) Technical S ecifications and Environmental Protection Plan The Technical Specifications contained in. Appendix> A, ay,revised...

through Amendment No.

103 and the Environmental Protection,Pl,an contained in Appendix B, are hereby incorporated in the license.

PP8L shall operate the facility in accordance with the Technical Specifications-and the Environmental-Protection Plan...,,,,

4.

This license amendment is effective as of its.dateiof issuance.and is to be implemented prior to startup in Cycle 7, currently scheduled for May 21, 1994. i%i'OR THE NUCLEAR REGULATORY COMMISSION

~ lt Attachments:

1.

Page 3 of License 2.

Changes to the Technical Specifications Date of Issuance:

April 11, 1994

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William T. Russell;- Director

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Office of Nuclear Reactor Regulation

• Page 3 is attached, for convenience, for the composite license. to reflect this change.

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(3)

PPSL, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to

receive, possess, and use at any time any byp'roduct'source and special nuclear material as" sealel'eutron sources"'for'eactor

startup, sealed neutron sources for reactor instrumentation and r'adiation monitoring equipment calibration, and as fission detectors in amounts as required; (4)

PPSL, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to

receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or ins'trument calibration or associated with radioactive apparatus or components; and (5)

PP&L, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to

possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C.

This license shall be deemed to contain and is subject to the conditions-.

specifie'd 'in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules,-

regulations and order s of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1)

Naximum Power Level Pennsylvania Power 8,Light. Company (PP8L) is authorized to operate the fa'cility at *reactor core power levels not in excess of 3441 megawatts thermal (100X power) in accordance with the conditions specified herein and in Attachment 1 to this license.

The preoperational

tests, startup tests and other items identified in Attachment 1 to this license shall be completed as specified.

Att'achment 1 is hereby incorporated into this license.

(2)

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

103, and the Environmental Protection

.Plan contained in Appendix 8, are hereby incorporated in the license.

PPKL shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

Amendment No. 4, 8, 103

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'I ATTACHMENT TO LICENSE AMENDMENT NO.

103 FACILITY OPERATING LICENSE NO. NPF-22 DOCKET NO. 50-388

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Replace the fol-lowing pages of 'the-Ap'pendix'tA Technical Specifications with the 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.*

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ATTACHMENT TO LICENSE AMENDMENT NO.

103 FACILITY OPERATING LICENSE NO. NPF-22 DOCKET NO. 50-388 Replace the following pages of the Appendix A Technical Specifications with enclosed pages.

The revised pages are identified by Amendment number and contain vertical lines indicating the area of change;, The..-overleaf:

pages are provided to maintain document completeness.*-

REMOVE 3/4 3-25 3/4 3-26 3/4 3-41 3/4 3-42 3/4 3-53 3/4 3-54 3/4 4-1 3/4 4-la 3/4 4-1b 3/4 4-lc 3/4 4-1d 3/4 4-le 3/4 4-1f 3/4 4-3 3/4 4-4 3/4 4-5 3/4 4-6 3/4 4-7 3/4 4-8 3/4 4-19 3/4 4-20 3/4 5-3 3/4 5-4

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ATTACHMENT TO LICENSE"AMENDMENT NO.

103 FACILITY OPERATING LICENSE NO. NPF-22 DOCKET NO. 50-388 Replace the following pages of the Appendix A Technical Specifications with enclosed pages.

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

The overleaf pages are provided to maintain document completeness.*

REMOVE B 3/4 3-3 B 3/4 3-4 B 3/4 4-5 B 3/4 4-6 B 3/4 4-7 B 3/4 5-1 8 3/4 5-2 B 3/4 6-3 B 3/4 6-4 5-5 5-6 5-7 5-8 6-20a 6-20b INSERT B 3/4 3-3 B 3/4 3-4*

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DEFINITIONS PHYSICS TESTS 1.28 PHYSICS-TESTS shall be those tests performed to measure the fundamental nuclear characteristics of-the reactor core and related instrumentation and 1) described in Chapter 14 of the FSAR, 2) -authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Comission.

PRESSURE BOUNDARY LEAKAGE 1.29 PRESSURE BOUNDARY LEAKAGE shall be leakage through a non-isolable fault in a reactor coolant systeII component bogy, pipe wall or vessel wall.

PRIMARY CONTAINMENT INTEGRITY 1.30 PRIMARY CONTAINMENT INTEGRITY shall exist when:

aO b.

C.

d.

All primary containment penetrations required to be closed during accident conditions are either:--

1.

Capable of being:closed by an OPERABLE primary'ontainment automatic isolation system, or II 2.

Closed by at least one'manual'valve, blind flingi~yr. deactivated automatic valve secured in"its c'fosed position, excijt-as provided in Table 3.6.3-1'of Specification 3.&.3.

All primary con'tainment equipment hatches ar'e closed and sealed.

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Each primary containment air lock is OPERABLE pursuanC'to Specification 3.6.1.3.

The primary. containment leakage rates are within the limits of Speci fication 3.6.1.2.

e.

The suppression chamber is OPERABlE pursuanf;,to,Specification 3.6.2.1.

f.

The sealing mechanism associated with each primary containment penetration; e.g., welds, bellows or O-rings, is OPERABLE.

PROCESS CONTROL PROGRAM

1. 31 The PROCESS CONTROL PROGRAM (PCP) shall contain the sampling, analysis, and formulation determination by which SOLIOIFICATION of radioactive wastes frea liquid systems is assured.

PURGE - PURGING 1.32 PURGE or PURGING shall be the controlled process of dis"barging air or gas from a confinement to maintain temperature;

pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

SUSQUEHANNA - UNIT 2 1-5

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RATED THERMAL POWER 1.33 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3441 MWT.

REACTOR PROTECTION SYSTEM RESPONSE TIIVI 1.34 REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time Interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until deenergization of the scram pilot valve solenoids.

The response time may be measured by any series of sequential, overlapping or.total steps such that the entire response time is measured, REPORTABLE EVENT 1.35 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50.

ROD DENSITY

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1.36 ROD DENSITY shall be the number of control rod notches inserted as a fraction of

'he total number of control rod notches.

Allrods fullyinserted ls"equivalent to 100%

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SECONDARY CONTAINIVIENTINTEGRI g

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1.37 SECONDARY CONTAINMENTINTEGRITY sliall exist when:

All secondary containment penetrations required to be closed during accident conditions are either:

1.

Capable of being closed by an OPERABLE secondary containment automatic isolation system, or 2.

Closed by at least one manual valve, blind flange, or deactivated automatic damper secured in its closed position, except as provided in Table 3.6.5.2-1 of Specification 3.6.3.2.

b. Allsecondary containment hatches and blowout panels are closed and sealed.

c.

The standby gas treatment system is OPERABLE pursuant to Specification 3.6.5.3.

d. At least one door in each access to the secondary containment is closed.

e.

The sealing mechanism associated with each secondary containment penetration, e.g.,

welds, bellows, resilient material

seals, or O-rings, is OPERABLE.

f.

The pressure within the secondary containment is less than or equal to the value required by Specification 4.6,5.1a.

SUSQUEHANNA - UNIT 2

.~<VAIL 1-8 Amendment No.

I,03

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2.0 SAFETY LIMITSAND LIMITINGSAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS THERMALPOWER Low Pressure or Low Flow 2.1.1 THERMALPOWER shall not exceed 25% of RATED THERMALPOWER with the reactor vessel steam dome pressure less than 785 psig or core flow less than 10 millionIbm/hr.

APPLICABILITY: OPERATIONAL CONDITIONS 1 AND 2..

ACTION:

With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel steam dome pressure less than 785 psig=or core flow less than 10.millionJbm/hr:, be. in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.

THERMALPOWER Hl h Pressure and Hl h Flow td

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2.1.2 The MINIMUMCRITICALPOWER RATIO {MCPR) shall'not be less than 1.06 with the reactor vessel steam dome pressure greater than 785 pslg and core floe greater than 10 million Ibm/hr.

APPLICABILITY: OPERATIONALCONDITIONS 1 AND 2.

ACT1ON:

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With MCPR less than 1.06 and the reactor vessel steam dome pressure greater than 785 psig and core flowgreater than 10 millionIbm/hr., be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.

REACTOR COOLANT SYSTEM PRESSURE e ~

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-.ted ee 2.1.3 The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed 1325 psig.*>>

ACTION:

With the reactor coolatnt'system'pressure, as measured in the reactor vessel steam dome, above 1325 psig, be in at least HOT SHUTDOWN with reactor coolant system pressure less than or equal to 1325 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.

See Specification 3.4.1.1.2.a for single loop operation requirement.

SUSQUEHANNA - UNIT 2 2-1 amendment tte.

t03 i

SAFETY LIHITS AND LIHITING SAFETY SYSTEH SETTINGS SAFETY LIHITS (Continued)

REACTOR VESSEL 'WATER LEVEL

2. 1.4 The reactor vessel water level shall be above the top of the active irradiated fuel.

Al APPLICABILITY:

OPERATIONAL CONOITIONS 3, 4 and 5

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ACTION:

With the reactor vessel water level at or below the top of the active irradiated fuel, manually initi'Ate the ECCS to restore the watdP level, after depres-surizing the reactor vessel, if required.

Comply with the requirements of Speci fication 6.7.1.

SUSQUEHANNA - UNIT 2 2-2

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS

2. 2 LIMITING.SAFETY SYSTEM SETTINGS REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The reactor protection systea 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 instreaentation setpoint less conservative than the value sham in the A11owable Values coleman of Table 2.2.1-1, declare the channel inoperable and apply the applicable ACTIN statement requireoent of Specification 3.3.1 until the channel is restored to OPERABLE status with its setpoint ad)usted consistent with the Trip Setpoint value.

SUSQUEHANNA - UNIT 2 "2-3

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TRIP. MTPOINT,.'",. ',

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Intermediate Range Monitor, Neutron Flux-High 6 120/125 divisions of fWI scale

"'": 'ALLOWABLEVALUES 6 122/126 divisions of full scale 2.

Average Power Range Monitor:

a.

Neutron Flux-Upscale, Setdown C 15% of RATED THERMALPOWER 6 20% of RATED THERMALPOWER b.

Flow Biased Simulated Thermal Power-Upscale 1)

Flow Biased 2)

High Flow Clamped C 0.58 W+69%¹,

with a maximum of 113.5% of RATED THERMAL POWER 6 0.68 W+62%¹,

with a maximum of 6 115.5% of RATED THERMAL POWER TABlE2ega)~$

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Neutron Flux-Upscale d.

Inoperative s 118% of RATED THERMALPOWER NA S 120% of RATEP THERMALPOWER NA 3.

Reactor Vessel Steam Dome Pressure I igh C 1087 psig 6 1093 psig 3

CL3 Z0 4.

Reactor Vessel Water Level - Low, Level 3 5.

Main Steam Une Isolation Valve - Closure 6.

Main Steam Line Radiation - High 7.

Drywall Pressure - High 8.

Scram Discharge Volume Water Level - High a.

Level Transmitter b.

Float Switch 9.

Turbine Stop Valve - Closure

10. Turbine Control Valve Fast Closwe, Trip Oil Presswe - Low

11. Reactor Mode Switch Shutdown Position 12.

Manual Scram a 13.0 inches above instrument zero C 10% closed 6 7.0 x full power background s 1.72 psig C 88 gallons 6 88 gallons S 6.6% closed R 600 psig a 11.6 inches above instrument zero 6 11% closed c 8.4 x full power background J

6 1.88 psig c 88.gallons g 88 gallons g 7% dosed 2 460 psig NA NA See Bases Figure 8 3/4 3-1.

O'ee Specification 3.4.1.1.2.a for single loop operation requirements, r+pg~KM'cf~

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2.1 SAFETY LIMITS )"

'i BASES

2.0 INTRODUCTION

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

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

The fuel cladding integrity Safety Liqjit,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 Limitsuch that the MCPR is not less than the limitspecified 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 stillgreater 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 norrrial operation and duiing anticipated operational occurrences, at least 99.9% of the fuel ro'ds in the core do not experience transition boiling (ref. XN-NF-524(A) Revision 1).

2.1.1 THERMALPOWER Low Pressure or Low o

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The use of the XH-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 Limitis established by a limiting condition on core THERMALPOWER 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~ is 3.35 Mwt or greater.

At 25%

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

Thus, a THERMAL POWER limit of 2596 of RATED THERMAL POWER for reactor pressures below 785 psig is conservative.

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SUSQUEHANNA - UNIT 2 B 2-1 Amendment No.

~i~, 103

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

5 The Safety Umit MCPR assures sufficient conservatism in the operating MCPR limitthat 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 Umit MCPR 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 ls the uncertainty inherent in the XN-3 critical power correlation.

XN-NF-624 (A), Revision 1 and PL-NF-90401 describe the methodologies used in determining the Safety Umit MCPR.

The XN-3 critics) power correlation is based on a significant body of prac5cal test data, providing e high degree of assurance that the critical power as evaluated by the correlation

's within e 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 limitIntroduce conservatism into the limit because bounding'high radial power factors and bounding flat local peaking distributions are used to estimate the number of rods ln boiling tranation.

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'e reasonable d(Igree of assurance that during sustained operation at the Safety Umit MCPR there wove be no transition boNng in the. core.

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

Significant test data accumulated by the U.S. Nuclear Regulatory Commission end private orginizations indicate that the use of a boNng transition limitation to protect against cladding failure Is a very conservative approach.

Much of the date intlcates that LWR fuel can survive for an extended period ot 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 4CPR (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.

ANNA - UNIT 2 B 2-2 Amendment No.9J

LIMITINGSAFETY SYSTEM SETTING BASES REACTOR PROTECTION SYSTEM INSTRUMENTATIONSETPOINTS (Continued)

Turbine S o Valve-Closure 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% of valve 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.

This function is not required when THERMALPOWER is below 30% of RATED THERMAL POWER.

The Turbine Bypass System is sufficient at this low power to accommodate a

turbine stop valve closure without the necessity of shutting down the reactor. This function is automatically bypassed at turbine first stage pressures less than the analytical limit of 147.7 psig, equivalent to THERMAL POWER of. about 30% RATED THERMAL POWER.

Turbine first stage pressure of 147,7 psig is equivalent to 22% of rated turbine load.

10. Turbine Control Valve Fas Closure Tri Oil Pressure-Low 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.

This function is not required when THERMALPOWER is below 30% of RATED THERMAL POWER.

The Turbine Bypass System is sufficient at this low power to accommodate a

turbine control valve closure without the necessity of shutting down the reactor.

This function is automatically bypassed at turbine first stage pressures less than the analytical limit of 147.7 psig, equivalent to THERMAL POWER of about 30% RATED THERMAL POWER. Turbine first stage pressure of 147.7 psig is equivalent to 22% of rated turbine load.

React r Mode Swi ch Shutdown 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.

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

SUSQUEHANNA - UNIT 2 8 2-7 Amendment No. 97, 103

1 I

REACTIVITY CONTROL SYSTEMS 3/4.1.5 STANDBY LI UID CONTROL SYSTEM LIMITING CONDITION FOR OPERATION

3. 1.5 The standby liquid control system shall be OPERABLE.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2 and 5*.

ACTION:

a.

b.

In OPERATIONAL CONDITION 1 or 2:

1.

With one pump-and/or one explosive valve inoperable, =restore the inoperable pump and/or explosive valve to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

2.

With the standby liquid control system otherwise inoperable, restore, the'ystem to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be in at least-,HOT'SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

In OPERATIONAL CONDITION 5*:

l.

With one pump and/or one explosive valve inoperable, restore-the inoperable pump and/or explosive valve to OPERABLE status within 30 days or insert all insertable control rods within the next hour.

2.

With the standby liquid control system otherwise i'noperable, insert all insertable control rods within one hour.

SURVEILLANCE RE UIREMENTS

4. 1.5 The standby liquid control system shall be demonstrated OPERABLE:

a ~

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying that; L

1.

The temperature of the sodium pentaborate solution is within the limits of Figure 3.1.5-1.

2.

The available volume of sodium pentaborate solution is within the limits of Figure 3.1.5-2.

3.

The heat tracing circuit is OPERABLE by actuating the test feature and determining that the power available light on the local heat tracing panel energizes.

th any contro rod withdrawn.

Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

SUSQUEHANNA - UNIT 2 3/4 1-19

REACTIVITYCONTROL SYSTEMS SURVEILLANCEREQUIREMENTS Continued

b. At least once per 31 days by;

1. Verifying the continuity of the explosive charge.

, 2. Determining that the available weight of sodium pentaborate is greater than or equal to 5500 Ibs and the concentration of boron in solution is within the limits of Figure 3.1.5-2 by chemical analysis.

3. Verifying that each valve, manual, power operated or automatic, in the flow path that is not locked, sealed, or otherwise secured in position, is In its correct position.

c. Demonstrating that, when tested pursuant to Specification 4.0.5, the minimum flow requirement of 41.2 gpm at a pressure of greater than or equal to 1224 psig is met.

S

d. At least once per 18 months during shutdown by;

1. Initiating one of the standby liquid control system loops, including an explosive valve, and verifying that a flow path from the pumps to the reactor pressure vessel is-available by pumping demineralized water into the reactor vessel.

The replacement charge for the explosive valve shall be from the same manufactured batch as the one-fired or. from another batch which has been certified by having one of that batch

=

successfully fired. Both injection loops shall be tested in 3B months.

2.

Demonstrating that all heat traced piping is unblocked by pumping from the storage tank to the test tank and then draining and flushing the discharge piping and test tank with demineralized water.

r 3.

Demonstrating that the storage tank heaters are OPERABLE by verifying the expected temperature rise for the sodium pentaborate solution in the storage tank after the heaters are energized.

This test shall also be performed anytime water or boron is added to the solution or when the solution temperature drops below the limitof Figure 3.1.5-1.

This test shall also be performed whenever both heat tracing circuits have been found to be inoperable and may be performed by any series of sequential, overlapping or total flow path steps such that the entire flow path is included.

SUSQUEHANNA - UNIT 2 3/4 1-20 Amendment No. 103

I

~

~

POWER DISTRIBUTION LIMITS 3 4.2.2 APRM SETPOINTS LIMITINGCONDITION FOR OPERATION 3.2.2 The APRM flow biased simulated thermal power-upscale scram trip setpoint (SI and fiow biased neutron flux-upscale control rod block trip setpoint (S<a) shall be established according to the following relationships:

TRIP SETPOINT ¹"-

ALLOWABLE::VALUE;0!:.

S 6 (0.58W + 59%) T SRS 6 (0.58W + 50%) T S M (0.58W + 62%) T SRS 6 (0.58W + 53%) T where:

S and SR> are in percent of RATED THERMALPOWER, W

=

Loop recirculation flow as a percentage of the loop recirculation flowwhich produces a 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 (FLPD) for SNP fuel is the actual LHGR divided by the applicable LINEAR HEAT GENERATION RATE for APRM Setpoints limit specified in the CORE OPERATING LIMITSREPORT.

T is always less than or equal to 1.0.

APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMALPOWER.

C 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 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 <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMALPOWER to less than 25% of RATED THERMALPOWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCEREQUIREMENTS 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 flowbiased 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 THERMALPOWER, 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 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 2 3/4 2-2 Amendment No. ~

103

MlTIN 4;2.2 (Continued)

P TtN a.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b.

Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at least 15% of RATED THERMALPOWER, and c.

Initiallyand at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the rector, is operating with MFLPD greater than or equal to FRTP.

d.

The provisions of Specification 4,0.4 are not applicable.

SUSOUEHANNA - UNIT2 3/4 2-3 haedment No. fS, 9>

AUG 4 ig93

C/l C

C/l C:

TABLE 3.3.1-1 (Continued)

REACTOR PROTECTION SYSTEH INSTRUHENTATION M

FUNCTIONAL UNIT 7.

Drywell Pressure - High 8.

Scram Discharge Volume Mater Level - High a.

Level Transmitter b.

Float Switch 9.

Turbine Stop Valve - Closure APPLICABLE OPERATIONAL CONDITIONS 1, 2(-)

5 1

2(-)

5 1(i)

HINIHUH OPERABLE CHANNELS PER TRIP SYSTEH a

2 2

2 2

4(k)

ACTION 1

3 1

3 6

CA>I CQ 10.

Turbine Control Valve Fast Closure, (i)

Valve Trip System Oil Pressure - Low 1 ~

ll.

Reactor Hode Switch Shutdown Position

~ (

~

t.

6 12.

Hanual Scram 1,

2 3, 4 5

1, 2

3, 4 5

~

~ ~

1 1

~1 2

2 1

7

/t 3

1 8

9

TABLE 3.3.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION ACTION STATEMENTS ACTION 1 ACTION 2 ACTION 3 ACTION 4 ACTION 5 ACTION 6 ACTION 7 ACTION 8 ACTION 9 Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Verify all insertable control rods to be inserted in the core and lock the reactor mode switch in the Shutdown position within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

Suspend all operations involving CORE ALTERATIONS and insert all insertable control rods within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

Be in at least STARTUP within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />...;r,i r.

Be in STARTUP with the main steam line isolation valves closed within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Initiate a reduction in THERMALPOWER within 15 minutes, and reduce THERMALPOWER to less than 3096 of RATED THERMALPOWER within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Verify all insertable control rods to be inserted within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

Lock the reactor mode switch in the Shutdown position within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

Suspend all operations involving CORE ALTERATIONS, and insert all insertable control rods and lock the reactor mode switch in the SHUTDOWN position within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

SUSQUEHANNA - UNIT 2 3/4 3%

Amendment No. 103

TABLE 3.3.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION TABLE NOTATIONS (a)

(b)

(c)

(d)

(e)

A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter.

Upon determination that a trip setpoint cannot be restored to within its specified value during performance of the CHANNEL CALIBRATION,the appropriate ACTION, 3.3.1a or 3.3.1b, shall be followed.

This function is automatically bypassed when the reactor mode switch is in the Run position.

IApl The "shorting links" shall be removed from the RPS circuitry prior to and during the time any control rod is withdrawn and shutdown margin demonstrations performed per Specification 3,10.3,

~ ~

The non-coincident NMS reactor trip function logic is such that all channels go to both trip systems.

Therefore, when the "shorting links" are removed, the Minimum OPERABLE Channels Per Trip System is 4 APRMS and 6 IRMA'<<"

~ 4K I An APRM channel is inoperable if there are fess, than 2 LPRM inputs" per level or less than 14 LPRM inputs to an APRM channel.

This function is not required to be OPERABLE when the reactor pressure vessel head is unbolted or removed per Specification 3.10.1.

, ~

(g)

(h)

This function is automatically bypassed when the reactor mode switch is not in the Run position.

This function is not required to be OPERABLE when PRIMARY CONTAINMENTINTEGRITY is not required.

With any control rod withdrawn.

This function sha8 not be automatically bypassed when turbine. first stage pressure is greater than an allowable value of,. 136 psig.

Also actuates the EOC-RPT system.

Not required for control rods removed per Specification 3,9.10.1 or 3.9.10.2.

SUSQUEHANNA - UNIT 2 3/4 3-5 Amendment No. ~" 0 ~0~ f

TISLE 3.3. 1-2 REhCTOR PRPTECTION SYSTEN RESPONSE TINES cn

.+V

~ \\

~

FlNHIONhL NIT lntenedfate Raaiy lfsaftors:

a.

tteutroa Fl~ ~ gg

~.

laoperetfve Average teer Iaaee Nsafto&:

a.

Neuttea Flux - ~le, Setdae b.

Flm Ifased SfmlaM Tleraal t~r - +scale c.

Ffxe4 tteutrea Flea - Upscale d.

Inoperetf ve RESPONSE Tile Seconds Nh g 0 QUA

< 0.0$

Nh 5.

6.

7.

8.

Reactor Vessel Stem Dose tressure - High Reactor Vessel %ter Level Los, Love) 3 Nafn Stem Line Iso)atfea Valve - Closure Nafn Stem Line Radiatiea - High 0~1) tressure - Hfyl Scrm Ofscterge Vol~ Mater Level - High a.

Level Transaftter b.

Float Sssftcb Turbfae Stop Valve - Closure Jurbfae Ceatrel Valve Fast Closure, Yrfy Oil tresswe - Lee Ioacter Nwto Ssrftcb StwMae tesftion Noawal Scrm

< 0.55 c 1.05

.< 0.06 NA Nh

< 0.06

< P.ON Via Nh eweutrea detectors are ~t fna response tfm testing.

Resp+me tfae sM11 be aeasured fthm tie detector wtyut ee'ry tte iaput ef the first electronic c~onent fn the chenae1.

%et facludfag sfmlaied ttisaal pear tfm ccostaat.

INoasured fna actuation of fast-acting solenoid.

0 K

zz I

Cz IQ 1.

PRIMARY CONTAINMENTISOLATION a.

b.

co d.

e.

Reactor Vessel Water Level 1)

Low, Level 3 2)

Low Low, Level 2 3)

Low Low Low, Level 1 Drywall Pressure

- High Manual Initiation SGTS Exhaust Radiation - High Main Steam Line Radiation - High h 13.0 inches h -38.0 inches0 a -129 inches 6 1.72 psig NA 23.0 mR/hr 6 7.0 x full power background 2 11.5 inches R -45.0 inches h -136 inches 6 1.88 psig NA 5 31.0 mR/hr 6 8.4 x full power background 2.

SECONDARY CONTAINMENTISOLATION

. ISOLATIONACTUATIQNINSTRUhhENTATlQN 8ETPOINTS

'"== -'"-'- TRIP R)AUCTION;":P:."""'."-".-'=~..-".-" '.".':::-':-::', TRiP SETPOINT '-='""'"

'. ": ".ALLOWABLEVAIUE a

Ca) a.

b.

C.

d.

e.

f.

Reactor Vessel Water Level - Low Low, Level 2 Drywall Pressure

- High Refuel Floor High Exhaust Duct Radiation-High Railroad Access Shaft Exhaust Duct Radiation - High Refuel Floor Wall Exhaust Duct Radiation-High Manual Initiation a -38.0 inches 6 1.72 psig s 2.5 mR/hr 6 2.5 mR/hr S 2.5 mR/hr NA

~ <5.0 inches 6 1.88 psig 6 4.0 mR/hr 6 4.0 mR/hr 6 4.0 mR/hr NA 3.

MAINSTEAM ONE ISOLATION.

2 z0 a.

b.

ce d.

Reactor Vessel Water Level - Low Low Low, Level 1

Main Steam Line Radiation - High Main Steam Une Pressure - Low Main Steam Une Flow - High h -129 inches 6 7.0 x full power background h 861 psig c 113 psid h -136 inches s 8.4 x full power background h 841 psig s 121 psid

-."-.:; '::-::;:'::i-':-'-::;.;.":;:; iSaiAi )OiAvion'iO'j')iij'iuiaiV'AiiCiai'SH~OIjiS.'".':.."- -"".".:.-... '

C zz Cz e.

Condenser Vacuum - Low f.

Reactor Building Main Steam Line Tunnel Temperature - High

~ 9.0 inches Hg vacuum 6 1774F

~ 'YC:: ".'". "":: ~

" '"" --- ':..:.'::."".::'TRIP,,RINCTION'::,.':":::-'.":""-;:.'-::"'.: "~-"""""'"': ",'.":,: "'::.".:;TRIP,'SETPQINT."'.:":;.';::",';-":

AIN S IN LINE ISO IContfnued)

'."ALLOWABLEVALUE-'

8.8 inches Hg vacuum E 1844F Ca)

D Cal co g.

Reactor Building Main Steam Line Tunnel h Temperature - High h.

Manual Initiation i.

Turbine Building Main Steam Line Tunnel Temperature - High 4.

REACTOR WATER CLEANUP SYSTEM ISOLATION a.

RWCU 4 Flow - High b.

RWCU Area Temperature

- High c.

RWCU/Area Ventilation h Temperatwe: High d.

SLCS Initiation e.

Reactor Vessel Water Level - Low Low, Level 2 f1.

RWCU Flow - High f2.

Non-Regenerative Heat Exchanger Dittcharge Temperatwe-High g.

Manual Initiation 6.

REACTOR CORE ISOLATION COOUNG SYSTEM ISOLATION

~ 994F NA C 1974F

~ 60 gpm C 1474 F or 1314F¹

~ 694F or 40.S4F¹ NA R -38 inches

-:-6-462 gpm 6 1444F NA i

6 108'F NA 5 200 F

~ 80gpm 6 1544F or 137'F¹ 4i 4 6 724F or 43.54F¹ NA R %5 inches 6 472 gpm 6 1504F CO

~0z0 a.

RCIC Steam Line h Pressure - High b.

RCIC Steam Supply Pressure

- Lovif c.

RCIC Turbine Exhaust Diaphragm Pressure

- High

- s.138" H,O a 60psig 6 10.0 psig 6143 HO h 53psig 6 20.0 psig These trip functions need not be OPERABLE from October 19, 1989 to Januaty49,1990

0 m

ZZ

)

Z

. "'::TRIP, RINCT30N "':; "-':.'"=" -'-'.'

REACTOR CORE ISOLATIO TEM ISOLATIO IContinued) d.

RCIC Equipment Room Temperature

- High s 1674F e.

RCIC Equipment Room h Temperature

- High f.

RCIC Pipe Routing Area Tem ture - High g.

RCIC Pipe Routing Area h Te ratwe - High h.

RCIC Emergency Area Cooler Qmp. - High

')'as'4F k

" "-":-"'-"""-"'"-'-'-"'-'-"'-'-::"'-'-TABLE3.3.2-2: ICOI1gIIIIad)':.:I '-':"'"'::

.',,:" ISOL'ATIONACTUAT)QNINSTRU5JIENTATIQQ SETPOINTS'"-;

s 1744F S 984Fo S 1744F¹¹ 984F¹¹ ~

s 1744F Glo Ca) rD a.

) tP HPCI Steam Line Flow - High -+ 4 b.

HPCI Steam Supply Pressure -'.44 HPCI Twbine Exhaust Diaphragpgressge

"'i h

C.

g d.

HPCI Equipment Room Tempereture - High e.

t HPCI Equipment Room h Temparatwe -

igh f.

HPCI Emergency Area Cooler Anp. - High HPCI Pipe Routing Area Tem ture - High "

i.

Manual Initiation Drywall Pressure

- High 6.

HIGH PRESSURE COOLANT INJECTION'STEM ISOLATION

',4A' 47'sig '~': -"-';~'.-';,<<!

S.387 Inches H,O

~

Ot 2

1 psig s fggsig s 167%F 48 I'

16.

F

~

t

~ <<

s 1674F¹¹ NA s 1.88 psig S 399 inches H20 2 90psig s 20 psig s 1744F s 984F s 1744F s 1744F¹¹ o 2 Cl Z0 l~

t h.

HPCI Pipe Routing Area h Tengprature - Higg s 894F¹¹ L

Manual Initiation NA j.

Drywall Pressure

- High s '1.72 psig

'hese trip functions need not be OPERABLE rom October 39, 1989 to January'.:l9; ]990;"-

984F¹¹o NA s 1.88 psig

DC z

I

=I

'. ';=.--'

;::;;:,-,',"."

',"-."-':-:~.;,"-.,;;;.;,'=;-'j,"."::..,',:'..'.'.'.;."."::,":-":.,:l80LATIONACTUATION;lNSTRUIIIIEQTATIONSETPQINTS--'"-;- ".':

.'-'..:." '-.;- ".: ';.:-.,:~'-:;"'i,.;:";.TjgP,'RlgQTION',i'j'::",g:".:,i'i;.'.";<'...'.'.$-,;,",'":."":; 5,""; -".:TRIP.SETPQINT'.,:-".: '"'.I'-

'".'-"'-. -ALLQWABLEVANE"'.='"

7.

RHR SYSTEM SHUTDOWN COOUNG/HEAD SPRAY NIODE ISOLATION Ca>

~D Ca) lbO e.

Reactor Vessel Water Level - Low, Level 3 b.

Reactor Vessel (RHR Cut-in Permissive) Pressure - High c.

RHR Flow - High d.

Manual Initiation e.

Drywall Pressure - High 2 13.0 inches 6:98 psig S 2S,OOO gpm NA S 1.72 psig R 11.5 inches 5 108 psig 5 26,000 gpm S 1.88 psig

.Oy,.-,Laweysetpolnta.for TSHt633;2N609$$ F and.,TDSH-633..2N802.;E,.F...)

q~;.;.-.;I>)l,,,,II,<~,k,z;,;,~v'lgy~g3-.;;:,-.':,;,,-,,:,...:,, -.,-..

Cm TNLE h.3.c. I-I (Cont inued)

ISOlAlION ACIUAIIOH IHSIRlSEHZAIION SURVEILLANCE RE IRL'HEHZS CINNNEL CIIANNEL fUNCTIONAL CHARNEL 1RIP F UNCT lON CHECX TEST CALlBRAZION 5.

REACTOR CORE ISOLATIOH COOLING SYSIEH ISOLATION OPERAIIORAL CONDITIONS FOR QIICH SURVE ILLANCE RE IREO 6.

C.

e.

i.

HIGH RCIC Steam Line d Pressure - High RCIC Steae Supply Pressure,-

Lm RCIC Turbine Exhaust Oiaphrage Pressure - High RCIC Equipaent Roaa Teaperature - High RCIC Equipment Rooa d Temperature - High RCIC Pipe Routing brea Teaperature

- High RCIC Pipe Routing Area d Temperature - High RCIC Eaergency Area Cooler Teaperature - High Hanual-Initiation Orwell Pressure - High -.

PRESSURE COOLANT INECZION NSIE

'A HA NA NA H:ISOLATION H'

N NA R- ',"

I, 2, 3

I, 2, 3

I, 2, 3

I, 2, 3

1, 2, 3*

1,2,3 S, Z, 3',

2, 3

I, 2, 3.

I, 2, 3

ct QO Z o 0 0 ~C Dl CQ

~ 0

~

lO r

C" CJ.

HPCI Steaa line d Pressure

- High HPCI Steam Supply Pressure - Le>>

g gal% P $ 5 C.

IIPCI Turbine Exhaust Diaphraya Pressure

- lligi>

"lliese lrip l>>>>etio>>s>>eed>>ot be Dl'LRAIIlk from Oclobet l9, 1989 lo Jan>>ary l9, 1990.

I, 2, 3

I, 2, 3

l, 2, 3

TABLE4.3.2.1-1 {Continued)

ISOLATIONACTUATIONINSTRUMENTATIONSURVEILLANCEREQUIREMENTS C

copC zR C

Tlup FUNCTION HIGH PRESSU E CO S

0 (Continued) d.

HPCI Equipment Room Temperature - High e.

HPCI Equipment Room h Temperature - High f.

HPCI Emergency Area Cooler Temperatwe - High g.

HPCI Pipe Routing Area Temperature -.High h.

HPCI Pipe Routing Area h Temperature - High CHANNEL CHECK NA NA NA NA CHANNEL FUNCTIONAL TEST M

CHANNEL CAUBRATION Q

a Q

Q OPERA'PONAL CONDITIONS FOR WHICH SURVEILLANCEREQUIRED 1".2, 3 1;2,3 n 2.3 1 2,3

~ ~ ~ e 1,2,3 Cal

~O Cs) l4 i.

Manual Initiation j.

Drywell Pressure - High 7.

RHR SYSTEM SHUTDOWN COOLINGJHEAD SPRAY MODE ISOLATION a.

Reactor Vessel Water Level - Low, Level 3 b.

Reactor Vessel (RHR Cut-in Permissive)

Presswe - High M

.=M:

NA

1. 2. 3 1,2,3 1,2,3 1,2,3 c.

RHR Flow - High d.

Manual Initiation e.

Drywall Presswe - High S

NA W

R.

r> =.M'A 1,2,3

1. 2, 3 1,2,3 o3 9

R O

4 ~

OO ~

tee ~

When handling irradiated fuel in the secondary containment and during CORE ALTERATIONS vessel.

When any turbine stop valve is open.

When VENTING or PURGING the drywall per Specification 3.$ 1.2.8.

This trip function need not be OPERABLE from October 19, 1989 to January 19, 1990.

and operations with a potential for draining the reactor

~ ~

t INSTRUMENTATION SURVEILLANCE RE UIREHENTS 4.3.4.2.

1 Each end-of-cycle recirculation pump trip system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3.4.2.1-,1.

4.3.4.2.2.

LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least= once per 18 months.

4.3.4.2. 3 The instrument r esponse time portion of the END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME of each trip function shown in Table 3.3.4.2-3 shall be measured at least once per 18 months..

Each test shall include at least the logic of one type of channel input, turbine control valve fast closure or turbine stop valve closure, such that both types of channel inputs are tested at least once per 36 months.

The measured time shall be added to the most recent breaker arc suppression time and the resulting END-OF-CYCLE-RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME shall be verified to.be within its limit.

4.3.4.2.4 The time interval necessary for breaker arc suppression from energization of the recirculation pump circuit breaker trip *coil shall be measured at least once per 60 months.

't

~ ~

~ l ~

tj SUSQUEHANNA " UNIT 2 3/4 3-41

TABLE 3.3.4.2-1 END-OF-CYCLE RECIRCULATIONPUMP TRIP SYSTEM INSTRUMENTATION TRIP FUNCTION MINIIHUMOPERABLE CHANNELS PER TRI SVSTE<<<<l <<sl 1.

Turbine Stop Valve - Closure 2 <<bl 2.

Turbine Control Valve - Fast Closure 2 <<b)

(a)

.A trip system may be placed in an inoperable status for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for required surveillance provided that the other trip system is OPERABLE.

(b)

This function shall not be automatically bypassed when turbine first stage pressure is greater than an allowable value of 136 psig.

SUSQUEHANNA - UNIT 2 3/4 3%2 Amendment No. 103

TABLE 3. 3. 6-1 (Continued)

CONTROL ROD BLOCK INSTRUMENTATION ACTION c"

r

~

ACTION 60 Declare the

-RBM inoperable and take the ACTION required by Specification 3.1.4.3.

ACTION 61> -

With the number of OPERABLE Channels:

~ A a.

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

ACTION 62 b.

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

I ~

I

'i ~ ~ s

~

l* sl eccl

~

s<

~ ~ re ~ c s

s rid,'ith the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, place the inoperable channel in the tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

. -;:E....,t~ lst; <~i~~ltfil:l~.p NOTES r

s ~

r ~ s C

I ~

~

'ith THERMAL POWER ) 30K of RATEO THERMAL POWER.

With more than one control rod withdrawn.

Not applicable to control rods removed per Specification 3.9.10. 1 or 3.9. 10.2.

• • • Not required when eight or fewer fuel assemblies (adjacent to the SRNs );

are in the core.

(a)

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

(b)

This function shall be automatically bypassed if detector count rate is

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

(c)

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

(d)

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

(e)

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

SUS(UEHANNA - UNIT 2 3/4 3-53 Amendment No.

50 AUG 30 tsss

" '-:-':"-.";,',"",:,",-'.-;:;.,""CONTROL.ROD.BLOCK INSTRUIIIINTATION8ETPOINTS,'.

toC CrCmx zz Cz

=I N

a.

Upscale¹¹ b.

Inoperative c.

Downscale 2.

APRM 6 0.63 W + 41%

NA h 6/126 divisions of full scale

..'"i'","'

".""'",; ","'",. : TIQP,:FUNCTION,;,;:::i;,:"':-",'.;.i : -".",:;'.

"',s.".>.";

1.

ROD BLOCK MONI 0 S 0.63 W + 43%

NA h 3/126 of divisions full scale a.

Flow Biased Neutron Flux Upscale¹¹ 1)

Flow Biased

) r'r 2)

High Flow Clamped b.

Inoperative c.

Downscale d.

Neutron Flux - Upscale Startup 3.

SOURCE RA GE MON 0 S

s 0.68W+ 60%

~ 108% of RATED THERMALPOWER NA h 5% of RATED THERMALPOWER 6 12% of RATED THERMALPOWER 6 0.68 W + 53%

6 111% of RATED THERMALPOWER NA h 3% of RATED THERMALPOWER s 14% of RATED THERMALPOWER Ca)

D co QiD a.

Detector not full in b.

Upscale c.

Inoperative d.

Downscale 4.

I EMEDIAE A

G 0 ITOS NA s 2 x 10'ps NA 2 0.7 cps NA S 4 x 10s cps NA 2 0.6 cps a.

Detector not full in b.

Upscale c.

Inoperative d.

Downscale NA C 108/126 division of full scale NA 2 6/126 division of full scale NA 6 110/126 division of full scale NA

~ 3/126 divisions of full scale 6.

SC M

CHA G VOL a.

Water Level - High 6.

EA 0

CO LA S

C CU TIONF 0 6 44 gallons 6 44 gallons O

~+z0 a.

Upscale b.

Inoperative c.

Comparator 6 114/126 divisions of full scale NA 6 10% flow deviation 6 117/126 divisions of full scale NA 6 11% flow deviation The Average power Range Monitor rod block function is viried as e function'of realrculatIjy lorop',flow (w), The trip settirIg of this function must c<+ yM's be maintained in accordance with specification 3.2,?. '-':

'?,:-,;&',ri+:7,";"~:,",'0;:,::.,"'-'."".'

" " Provided signal to-noise ratio is R 2, Otherwise,.3 cps as trip setpaint and 2,8:,cps,far allowable,,valuer

¹¹ See Specificatian 3.4.1;1.2.i for single,'loop,operation requirements.',

~ 9:-'.;,':':r:.,'..','-.~>'.-".>.',-'j;:~;,.;'::.,".:":.'+",:,j.'.

3/4.4 REACTOR COOLANT SYSTEM 3/4. 4. 1 RECIRCULATION SYSTEM RECIRCULATION LOOPS -

TWO LOOP OPERATION LIMITING CONDITION FOR OPERATION 3.4.l.l. 1 Two reactor coolant system recirculation loops shall be in operation with the reactor at a

THERMAL POWER/core flow condition outside of Regions I and II of Figure 3. 4. l. l.l-l.

APPLICABILITY:

OPERATIONAL CONDITIONS 1" and 2"+, except during single loop I

I ACTION:

a.

In OPERATIONAL CONDITION 1:

l.

With:

wl >>'T>>"~

• 1 a) c)

No reactor coolant system recirculation loops in operation, or Region I of Figure 3. 4. l.l. 1-1 entered, ot

>> )>)>>>>

~

Region II of Figure 3. 4. 1. l..i-l..entered and"core therma) hydraulic instability occurring as evidenced by:

1),,Two or more APRM readings oscillating with at least

" '-"j-'one oscillating greater than or equal to 10K of RATED THERMAL POWER peak-to-peak, or 2)

Two or more LPRM upscale alar'ms activating and deactivating with 'a 1 to 5 second period, or 3)

Observation of a sustained LPRM oscillation of greater than 10 w/cm peak-to-peak with a 1 to 5

second period, or d)

Region II of Figure 3.4. l. 1. 1-1 entered and less than 50%

of the required LPRM upscale alarms

OPERABLE, immediately place the reactor mode switch in the shutdown position.

• See Special Test Exception

3. 10.4.

PSee Speci fication 3. 4. 1. 1. 2 for single loop operation requirements.

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

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

60 NOV 22 tg89

%4 REACTOR COOLANT SYSTEM SURVEILLANCEREQUIREMENTS ACTION: (Continued)

2. If Region II of Figure 3A.1.1.1-1 is entered and greater than or equal to 50% of the required LPRM upscale alarms OPERABLE, immediately exit the region by:

a) inserting a predetermined set of high worth control rods, or b) increasing core flow.

~ ~

~ I

>>s,)

3. With less than 50% of the required ARM;upsc'aie alarms: OPERABLE, follow ACTION a.1.d upon entry into Region II of Figure 3.4;1.1.1-1.

b.

In OPERATIONALCONDITION2 with no reactor coolant system recirculation loops in operation, return at least one reactor coolant system recirculation

'oop to operation, or be in HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

c. With any pump discharge valve'not OPERABL'h remove the assocINed loop from operation, close the valve*-and,comply witljthi requireitients of Specification 3A.1.1.2.

4 4 ~

g

d. With any pump discharge bypass valv'e not"OPERABLE'close the'valve and verify closed at least once per.31 days. -;:~

1

~

Qtf q

v l/

4.4.1.1.1.1 Each pump discharge valve and bypass valve shN'be demonstrated OPERABLE by cycling each valve throug5fat least one'complete cycle of full travel during each startup prior to~THERMAL POWER exceeding 25% of RATED THERMAL.POWER:

~ ~-

4.4.1.1.1.2 Each pump MG set scoop tube electrical and mechanical stop shall be demonstrated OPERABLE with overspeed setpoints less'than or equal to a core flow'jf.109.5 million Ibm/hr and 110.5 million Ibm/hr respectively, at least'once"per 18 months.

4.4.1.1.1.3 At least 50% of the required LPRM'upscale alarms shall be determined OPERABLE by performance of the following on each LPRIVI upscale alarm:

1)

CHANNEL FUNCTIONALTEST at least once per 92 days, and 2)

CHANNELCALIBRATIONat least once per 184 days.

If not performed within the previous 31 days.

SUSQUEHANNA - UNIT 2 3/4 4-1e Amendment No. 88, 103 J

REACTOR COOLANT SYSTEM RECIRCULATION LOOPS - SINGLE LOOP OPERATION LIMITINGCONDITION 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:

Tr(p Setpo(nt

'-" '"-"'" ~"""-"'-'--'-"N4wableValue:

+

s 0.58W + 54%

s b.58W+ 57%

3.

Specification 3.2.2:

the APRM Setpoints shall be as follows:

TItjSetpol Q VSI4&<"""'""

t' s (0.58W + 54%) T sR s (0.5 W., +, 5%) T' S s (0.58W + 57%) 7 SRa k (0.58W

+'48%)'T'.

Specification 3.2.3,'he MINIMUMCRITICALPOWER RATIO (MCPR) shall be grieater thy'r equal to,the apgJi~j,gin ale Loop Ifperation WPR limit as specified in the CORE OPERATING LIMITSREPORT.

5.

Specification 3.2.4: The LINEARHEAT GENERATION RATE (LHGR) shall be less than or equal ta the applicable Single Loop Operation LHGR limit as specified in the CORE OPERATING LIMITS REPORT.

6, Table 3.3.6-2:

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

a. RBM-Upscale b.

APRM - Flow Biased Tr(p 8etpo(nt::-:.;=: '

0.83W+ 35%

Trip Sitpo(nt:

0.58W+ 45%

-,'9

-'. AIIowabie,Va(ua; s 0.63W + 37%

AIIcAveb)a'.Vefoe '

0.58W + 48%

APPLICABILITY: OPERATIONAL CONDITIONS 1'nd 2 +,

except during two loop operation.¹ ACTION:

a.

In OPERATIONALCONDITION 1:

1. With a) no reactor coolant system recirculation loops in operation, or b)

Region I of Figure 3.4.1.1.1-1 entered, or c)

Region II of Figure 3.4.1.1.1-1 entered and core thermal hydraulic instability occurring as evidenced by:

N

-U IT2 3/4 4-1c Amendment No. ~i ~

REACTOR COOLANT SYSTEM LIMITING CQNQITIQN FQR OPERATION Continued ACTEQN:

(Contfnuad) 1)

Two or more APRM readings oscfllat'ing with at least one oscillating greater than or equal to 10% of RATEQ THERMAL POWER peak-to-peak, or 2)

Two or more LPRM upscale alarms activating and deactivating with a 1 to 5 second period, or 3)

Observation of a sustained LPRM oscillation of greater than 1Q w/cm~ peak-to-peak with a 1 to 5 second period, or d)

Region II of Figure 3.4, 1. 1. 1-1 entered and less than 50%

of the required LPRM upscale alarms

QPERALBE, immediately place the re~ctor mode s~itch in the shutdown position.

2.

If Region II of Figure 3.4. l.l. 1-1 is entered and greater thao or equal to 5QX of the required LPRM upscale alarms are OPERABLE, immediately exit the region by:

a) inserting a predetermined set of high worth control rods.

or b.

C.

d.

b) increasing core flow by increasing tt.e speed of the opera%i ng;rec~ycul atione,puep...~:

~!:;,->q 3.

With less than 50% of the required LPRM upscale alarms OPERABLE, follow ACTION a. l.d upon entry into Region II of Figure 3.4.1.1.1-1.

In OPERABLE CONOITION 2 with no reactor coolant system recirculation loops in operation, return at least one reactor coolant system recirculation loop to operation, or be in HOT SHUTOOQI within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

With any of the limits specified in 3/4.1. 1.2a not satisfied:

l.

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

2.

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

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

With any pump discharge valve not OPERABLE remove the associated loop from operation, close the valve and verify closed at least once per 31 days.

SUS(UEHANNA - UNIT 2 3/4 4-1d Amendment No. 60 NOV 28 l989

REACTOR COOLANT SYSTEM LIIVIITINGCONDITION FOR OPERATION Continued

"- '-'.'f;-Qfifhwnypump-discharge bypass valve not OPERABLE close the valve and verify closed at least once per 31 days.

SU VEIL CE REQUIREMENTS 4.4.1.1.2.1 Upon entering single loop operation and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter, verify that the pump speed in the operating loop is s 80% of the rated pump speed.

tr0(

4 4.1.1.2.2 At least 50% of the required LPRM upscale alarms shall be determined OPERABLE by performance of the following on each LPRM upscale alarm.

,1L,QHAQJf4El,guNCTIONALTEST at least once per 92 days, and 2V 'CHANNEL'ALIBRATIONat least once per 184 days.

4 4.1.1.2.3 Within 15 minutes prior to either THERMALPOWER increase resulting from a control rod withdrawal or recirculation loop flow increase, verifythat the following differential temperature requirements are met if THERMAL POWER is M 30%""" of RATED THERMALPOWER or the recirculation loop flow in the operating recirculation loop is C 50%'""" of rated loop flow:

a.

S 145 F between reactor vessel steam space coolant and bottom head drain line coolant, b.¹¹ 6 504F between the reactor coolant withinthe loop not in operation and the coolant in the reactor pressure vessel, and c.¹¹ S 50 F between the reactor coolant withinthe loop not in operation and operating loop.

4.4.1.1.2A The pump discharge valve and bypass valve in both loops shall be demonstrated OPERABLE by cycling each valve through at least one complete cycle of full travel during each startup"" prior to THERMAL POWER exceeding 25% of RATED THERMALPOWER.

4.4.1.1.2.5 The pump MG set scoop tube electrical and mechanical stops shall be demonstrated OPERABLE with overspeed setpoints less than or equal to a core flow of 109.5 million Ibm/hr and 110.5 million Ibm/hr respectively, at least once per 18 months.

4.4.1.1,2.6 During single recirculation loop operation, all jet pumps, including those in the inoperable loop, shall be demonstrated OPERABLE at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying that no two of the following conditions occur:¹¹¹ a.

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

SUSQUEHANNA - UNIT 2 3/4 4-1e Amendment No. 7I

~

103

REACTOR COOLANT SYSTEM SURVEILLANCEREQUIREMENTS Continued b.

The indicated total core flow differs by more than 1096 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.

Initial value.

Final value to be determined based on Power Uprate startup testing.

Any required change to this value shall be submitted to the.

Commission within 90 days of Power Uprate startup test program completion.

L See Specification 3A.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. ~i

~ 103

4 J1

'$h'l 44$

4

~

~ -, ~ $ 4$ ~ ~ ~, '4

'.+ t~'44 ~

~f "tTlh1> I~

'44'+

Ao

'4

'1 \\-

yl($tvAP

~

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REACTOR COOLANT SYSTEIVI RECIRCULATION PUIVIPS LIMITINGCONDITION FOR OPERATION

)'A

~

p'A.1.3 Recirculation pump speed mismatch shall be maintained within:-

a. 5'f each other with core flow greater than or equal to 75 million Ibm/hr.

b.

10% of each other with core flow less than 75 million lbm/hr.

APPLICABILITY:

OPERATIONALCONDITIONS 1 AND 2" when both recirculation loops are in operation.

t ACTION:

l I

tl I

With the recirculation pump speeds different by more. than the specified; limits, either:

\\

a.

Restore the recirculation pump speeds to within the specified limitwithin 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or b.

Declare the recirculation loop of the pump..with the. slower speed not in operation and take the ACTION required by Specification.3,4;1.1.1.

tl t t

SURVEILLANCEREQUIREIVIENTS 4.4.1.3 Recirculation pump speed mismatch shall be'verified to be within the limits at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

r r

See Special Test Exception 3.10.4.

SUSQUEHANNA - UNIT 2 3/4 4-3 I

Amendment No. ~9) 103

~

~

~

~

REACTOR COOLANT SYSTEM IDLE RECIRCULATION LOOP STARTUP LIMITING CONOITION FOR OPERATION

,i ~'jl'a'0 I.i' Ak

'('.4.1.4 An idle recirculation loop shall not be started unless the temperature differential between the reactor pressure vessel steam space coolant and the bottom head drain line coolant is less than, or equal to 145F; and:

~ OT.aD 4'.

When both loops have been idle, unless the temperature differential between the reactor coolant within the idle loop to be started up and the. coolant in the reactor pressure vessels is less than or equal to 50~F, or, b.

When only one, loop has-been idle, unless the temperature differential between the reactor coolant within the idle and operating recirculation loops is less than or equal to 504F, the operating loop flow rate is less than or equal to 50% of rated loop flo'w, and the reactor is operating at a THERMAL POWER/core flow condition below the 80% Rod Line shown in Figure 3.4. l. 1. 1-1, ji

~

i ~

.aa APPLICABILITY:

OPER(TIONAL CONDITIONS 1, 2, 3 and 4.

ACTION:

4 ~

~ ~

With temperature differences and/or flow rates exceeding the above limits, suspend startup of any idle recirculation loo'p.

Jt SURVEILLANCE RE UIREMENTS 4.4.1.4 The temperature differentials and flow rate shall be determined to be within the-liaits within 15 minutes prio~ to startup of an idle recirculation loop.

SUSQUEHANNA - UNIT 2 3/4 4-4 Amendment No 60 NOV 22 t98g

~

~

I I

REACTOR COOLANT SYSTEM 3 4.4.2 SAFETY RELIEF VALVES LIMITINGCONDITION FOR OPERATION 3.4.2 The safety valve function of at least 12 of the following reactor coolant system safety/relief.valves shall be OPERABLE with the specified code safety valve function liftsettings:

2 safety-relief valves 5 1175 psig a1%

6 safety-relief valves 5 1195 psig ~1%

8 safety-relief valves I 1205 psig %1%

APPLICABILITY: OPERATIONALCONDITIONS 1,, 2, AND 3.

~ACTIO

'i f)ilk3

a. '

WIth the safety valve %net)on of oneo<. mpre of the above required'safety/relief

".'.". valves inoperable, be in at least HOT SHOWDOWN withih 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. Wit)) one or morel, pfifety/rl)lief valves stuck open, provided that suppression

'po61 average water'temperature Is less than 105'F, close the stuck open relief valve(s); if unable to close the open valveis) within 2 minutes qr ifsuppression pool water temperature is 105oF or greater, place the reactor mode switch in the Shutdown position.

c.

With one or more safety/relief,, valve pcoustic

'monitors inoperable, restore the inoperable monitor(s) to O~BL'g stattvUwithjn 7 days orbe irrat least, HOT SHUTDOWN, within'the neitt'12 hotirs and,it)~).D SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURVEIL NC EQUIR MENT 4.4,2>>>>>> The acoustic monitor, for each. safety/relief valve shall be demonstrated, PERABLE with'the setpoint verified to be 0.25 of thaguU ppqn noise Ieye)>>,py.pqrfq kflce of a:

, a..

CAANNEL,FLINCTIONALTEST at least once per 31 days'; and 5 b.

Calibration in accordance with procedures prepared in conjunction with its manufacturer's recommendations at least,pnce per 1,8,,months.">>

~ 1&

'I

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>> ~

~I,,Q gI~IsAh I

~ ~

I The lift setting pressure shall correspond to ambient conditione 6f the'valves.at nominal operating temperatures and pressures.

OO Up to 2 inoperable valves may be replaced with spare OPERABLE valves with lower setpoints until the next refueling.

Initial setting shall be in accordance with the manufacturer's recommendation.

Adjustment to the valve full open noise level shall be accomplished during the startup test program.

The provisions of Specification 4.0A are not applicable provided the surveillance is performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure is adequate to perform the test.

Compliance with these requirements for the "S" SRV acoustic monitor is not required for the period beginning January 21, 1994, until the next unit shutdown of sufficiant duration to allow for containment entry, not to exceed the sixth refueling and inspection outage.

SUSQUEHANNA - UNIT 2 3/4 4-6 Amendment No. 7%,

103

~

~

I REACTOR COOLANT SYSTEM 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEHS LIMITING CONOITION FOR OPERATION 3.4.3.1 At least the following reactor coolant system leakage detection systems shall be OPERABLE:

b.

Two drywell floor drain sump level channels, and One primary containment atmosphere gaseous radfoactivity monitoring systetw channel and one containment atmosphere particulate radioactivity monitoring systes channel aligned to the drywell.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

a.

b.

With one or both channels of the drywell floor drain sump level monitoring system inoperable, be fn at least HOT SHUTMWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUT058 within the following 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> s.

With both channels of the gaseous radioactivity monitoring systea inoperable or with both channels of the particulate radioactivity monitoring systea inoperable, operation may continue for up to 30 days provided grab samples of the containment atmosphere are obtained and analyzed at least once'er 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

If at least one channel of the affected monitoring systea cannot be returned to OPERABLE status and aligned to the drywell within 30 days, or the grab samples are not obtained and analyzed as required, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and fn COLD SHUTOOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURYEILLANCE RE UIREMENTS 4.4.3.1 The reactor coolant systta leakage detection systems shall be desonstrated OPERABLE by:

a.

Prfaary contafnaent atmosphere particulate and gaseous monitoring system-performance of a CHANNEL CHECK at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, a

CHANNEL FUNCTIONAL TEST at least once'per 31 days and a CHANNEL CALIBRATION at least once per 18 months.

b.

Orywell floor drain suep level monitoring systea.perforeance of a CHANNEL FUNCTIONAL TEST at least once per 31 days and a CHANNEL CALIBRATION at least once per 18 months.

SUSQUEHANNA - UNIT 2 3/4 4-6 Amendaant No+

63 48$

REACTOR COOLANT"'SYSTEM OPERATIONAL LEAKAGE LIMITINGCONDITION FOR OPERATION 3.4.3.2 Reactor coolant system leakage shall be limited to:

a@

b.

C.

d.

e.

No PRESSURE BOUNDARY LEAKAGE:

5 gpm UNIDENTIFIED LEAKAGE.

25 gpm total leakage, averaged over any 24-hour period.

1 gpm leakage at a reactor coolant system pressure of 1035 a 10 psig from any reactor coolant system pressure isolation valve, specified in Table 3.4.3.2-1.

2 gpm increase in UNIDENTIFIED LEAKAGEwithin any 4-hour period.

ACTION:

aO With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24

'ours.

b.

With any reactor coolant system leakage greater thanthe limits inb;and/or c., above, reduce the leakage rate to within the limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

C.

With any reactor coolant system pressurajsolation valve leakage greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one closed manual or deactivated automatic valve, or be in at least HOT SHUTDOWN withinthe next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN withinthe following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

d.

With one or more of the high/low pressure interface valve leakage pressure monitors shown in Table 3.4.3.2-1 inoperable, restore the inoperable monitor(s) to OPERABLE status within 7 days or verify the pressure to be less than the alarm pressure at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; restore the inoperable monitor(s) to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

e.

With any reactor coolant system UNIDENTIFIEDLEAKAGEincrease greater than 2 gpm within any 4-hour period, identify the source of leakage increase as not service sensitive Type 304 or 316 austenitic stainless steel within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SUSQUEHANNA - UNIT 2 3/4 4-7 Amendment No.

$ O3

REACTOR COOLANT SYSTEM SURVEILLANCE RE UIREMEMTS

4. 4. 3. 2. 1 The reactor coolant system leakage shall be demonstrated to be within each of the above limits by:

a.

Monitoring the primary containment atmospheric particulate and gaseous radioactivity at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, and b.

Monitoring the drywell floor drain sump level at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

c.

Determining the total IDENTIFIED LEAKAGE at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

4.4.3.2.2 Each reactor coolant system pressure isolation valve specified

'n Table 3.4.3.2-1 shall be demonstrated OPERABLE by leak testing pursuant to Specification 4. 0. 5 and verifying the leakage of each valve to be within specified limit:

a.

At least once per 18 months, and b.

Prior to returning the valve to service following maintenance, repair or replacement work on the valve which could affect its leakage rate.

The provisions of Specification 4.0.4 are not applicable for. entry into OPERATIONAL CONDITION 3.

4;4.3..2.3 The high/low pressure interface valve leakage pressure monitors shall be demonstrated OPERABLE with the alara setpoints per Table 3.4.3.2-1 by performance of a:

a.

CHANNEL FUNCTIONAL TEST at least once per 31 days, and b.

CHANNEL CALIBRATION at least once per 18 months.

SUSQUEHANNA - UNIT 2 3/4 4-8 Amendment No. S3 NY 4 158

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RE CTOR COOLANT SYSTEM REACTOR STEAM DOIVIE LIMITINGCONDITION FO OPERATION 3.4.6.2 The pressure in the reactor steam dome shall be less than 1050 psig.

APPLICABILITY:

OPERATIONAL CONDITIONS 1 and 2"'.

ACTION:

With the reactor steam dome pressure exceeding 1060 psig, reduce t6e pressure to less than 1050 psig within 15 minutes or be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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SURVEILLANCEREQUIREMENTS

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'I 4.4.6.2 The reactor steam dome pressure shall be verified to be less than T050 psig at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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Not applicable during anticipated transients.

SUSQUEHANNA - UNIT 2 3/4 4-20 Amendment No. 103

EMERGENCY CORE COOLING SYSTEMS LIMITING CONDITION FOR OPERATION Continued ACTION:

(Continued) d.

For the ADS:

2.

With one of the above required.ADS valves inoperable,.provided the HPCI'system, the CSS and. the LPCI system are

OPERABLE, restore the inoperable ADS valve.to OPERABLE status within 14 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and reduce reactor steam dome pressure to < 100 psig within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

With two or more of the above'equired: ADS'alves inoperable, be in at least HOT SHUTDOWN"within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and reduce-reactor steam dome pressure to < 100 psig within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

With a CSS header'P instrumentation-channel inoperable,. restore the inoperable channel to. OPERABLE. status with 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or determine the ECCS header 4P locally..at,-least once per, 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />;,,otherwise, declare the CSS inoperable.

In the event an ECCS system is actuated and'njects water into the reactor coolant system, a Special Report shall, be prepared and sub-mitted to the Commission'ursuant to Specifi'cation 6.9.2 wi'thin 90 days describing the circumstances of the =actuation and the total accumulated actuation cycles to date.

The current value of the usage factor for each affected injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.

g.

With-the condensate transfer pump: discharge low, pressure alarm instrumentation: inoperable,'onitor the CSS LPCI, and HPCI pressure locally at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

'O I

SUSQUEHANNA " UNIT 2 3/4 5-3

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCEREQUIREMENTS 4.5.1 The emergency core cooling system shall be demonstrated OPERABLE by:

a, " <At)east'once'pen34 days:

1.

For the CSS, the LPCI system, and the HPCI system:

a)

Verifying that the system piping from the pump discharge valve to the system isolation valve is filled with water by:

1.

Venting at the high point vents 2.

Performing a'CHANNEL FUNCTIONAL'fEST of the condensate transfer pump discharge low pressure alarm Instrumentation a 2.

3.

4, b.

Ve 2.

3.

b)

Verifyingthat each valve, manual, power-operated, or automatic, in the flow path that is not locked, sealed, or-othqqeise secured in position, is in its.

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For the CSS, performance of a CHANNEL FUNCTIONAL.TEST of the core spray>

header hP inqtrumentation:

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For the LPCI system, verifying that at least one LPCI system subsystem cross-tie valve is closed with power removed from the valve operator.

For the HPCI system, verify!n'g that:the'pump.,flcniiR cantro)ler Is'in;ge correct position.

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'r'ach LPCI pump in-each subsystem develops'a-Aowibf at'east 12;200 gpm against a test line pressure of h 222 psig, corresponding to a reactor vessel to primary containment differential pressure ~ 20 psid.

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The HPCI pump develops a flowofat least 5000 gpm against a test line pressure of h 1140 psig when steam is being supplied to the turbine at 920)c,+140, -20 pslge c.

At least once per 18 months:

1.

For the CSS, the LPCI system, and=the HPCI system, performing a system functional test which includes simulated automatic actuation of the system throughout its emergency operating sequence and verifying that each automatic valve in the flow path actuates to its correct position. Actual injection of coolant into the reactor vessel may be excluded from this test.

The provisions of Specification 4.0.4 are not applicable provided the surveillance is performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure is adequate to perform the test.

Except that an automatic valve capable of automatic return to its ECCS position when an ECCS signal is present may be in position for another mode of operation.

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INSTRUIVIENTATION BASES 4

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er e 3 4.3 4 RECIRCULATION PUIVIP TRIP ACTUATION INSTRUMENTATIO

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'he anticipated transient without scram (ATWS)rrecirculation pump-,ti'tjstem-.pgpvtdes a -:

means of limiting the consequences of the uniikelp'ccurrence'Woa,faffur8-fo'sciarn"during an anticipated transient.

The response of the pla'nt'to this'postulatedrevent falls. within the envelope of study events in General Electric Co76psÃy To'pical"Report'gbO>10349, dated March 1971 and NEDO-24222, dated December 1979..'he end-of-cycle recirculation pump trip (EOC-RPQT,.sysSr'a-".is'~a+t~i.-the'vReactor protection, System.-gnd.'IS=aii=essential safety supplemeig.td'.Pe reactor trgp~The purpose

. of-.thefOC"-RpNIPto (ecover the loss of thermal'margin whicl1%ccurs,'at'.thgAf-of-cycle.

The Phgliomal. PhenOmeriOn inVOIVed iS that the: yOidtq'r'eua@Wityf',feeggaCIC.,'due tO a.

pressurization transient can add positive rea'ctivity.to'the reactto< syst'et't'a:faster rate.

than'he control. rods add negative scram "reactivity.

- EyN'.~EOC-RP~yste~ips-.,both,,

recirculation pumps~aredpcing,coolant, flow-in order.te.'ague'e';tea"voidwoJ)apsqJathe'core during two of:the moselimiting'pressurization eveqtsgf~The two,'.eyeiit~r-which the Eoc-RPT protective..feature:,wfiiI",function are clos'uyre of the turbine yfop'valves and fast closure.

of the<turbIne control v'elves",~"'"

',a ~ho.

Wfv 4 fastclos@e sens'or from'ea'jh.of.two turbine,control valves prjyj)jis~pt ty.the EOC-RPT

'srysteyyy; a,fast~co ure 'sensor fromsefich of thii other tvy$Lturbihficrfyttqfr~lvea Provides input tothe s'e'cond EOC-RPT system.

Similarly, a positioa Owitjtl,for,'eadj~of two turbine->>'

,atffP'V'alVea'PrOVideS inPut tc One EQC-.RPT~SStern; a

POSftjtrim'Svr'rtei hgrn ea'oh Ofthe Other

. two stop>elves provides input to:ELrv'oth'eg EOC-RPT. system'r.-.%Lsfrch.'EQYfglf'systeni, the.<ens jrf relqy<poptacts~Aare, arranged.to form a<-.out.of< logic'foFthe"fast'clo(rure of

..turbine googol. va!ves-and a.2 out-of%'togic.forthe'turb'je stop vaMeO~The-.operation of

,eitheijfpgic.wilf..you'ate the EOC-RPT system, and trjpAmth reclrcu~gjpur6py ~

js=functio07is not'reqgged'->vyjfqrt THERMA'L'"PGM/EH is-below 30% of ~EQ,yTHEFtMAL

" POWER.

The.Turbine.Bypassgystem iy.aufficient, at this low. poWer"to:accommodate a

turbine-',titop, valve"or'coritrooI-'valise"cfos1Ire vlittlout the necessity.'of trippnfp thy'reactorr recjrculationvpumps. -*Thisluhction~Is aytomatiIHHty'ybQypassed agtlrbine;flis'tstag~y pressUres less than tlie'Tanaryntfdat.l it:rmffion47 7ps'ig, e.q.uTvalent.to"THERllrlAL"'pOytgo~io~0%:::

RATED THERMAL'pOWER!"furpiftefirst stage pres'sure of 147jpp~sl e'qblvalkhfgo 22%

of rated. turbine..load;E";,,"." j "";-,j.."..'..--:-~'.".-.+..w~ A'. "~~~.='~~~'.

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be 'manually:-,'-)bypassed-'byus'e~of.

a keyswitch which is administratively controlled.

The manual bypasses "and the automatic Operating Bypass at less than 30% of RATED THERIVIALPOWER are annunciated in the control room.

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~ prre et The EOC-RPT response time is the time assumed in the analysis between initiation of valve motion and complete suppression of the electric arc, i,e175 ms, 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.

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

103

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INSTRUMENTATION.

BASES 3/4.3.5 REACTOR'CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION The reactor core isolation cooling system actuation instrumentation is provided to initiate actions to assure adequate core cooling in the event of reactor isolation from its primary heat sink and the loss of"feedwater flow to the reactor vessel without providing actuation of any of the emergency core cooling equipment.

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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.'

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~'L The control rod block functions are provided consistent with the requirements of the specifications in Section 3/4. 1.4, Control Rod Program Controls, and Section 3/4.2 Power Distribution Limits.

The trip logic is arranged so that a tFip 'hn any one of the inputs will result in a control rod bl ock.

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.i/]wilt';~yt Operation with..a trip set. less conservative than. its Trip Setpoint but within its specified Allowable Value is acceptable on.the basjs,tpat the difference between each Trip Setpoint and, tPe Al].owab1e Value.,is, equal to or less than the drift allowance assumed for. each trip in the safety analyses.

h The Rod Block Monitor (RBM) portion of the contrpl rod block instrumentation contains'dltiplexi'n'$ Bbcuitry'which interfaces with the reactor manual control system.

The RBM is a redundant system which includes two channels of information which must ag'reebefore-'rod'it(6'N'on is permitted.

-Each of these redundant chan-nels has a self-test feature which is implicitly tested during the performance of surveillance pursuant to this specification as well as the control rod operability specification (3/4.1.3.1).

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3/4.3. 7 MONITORING INSTRUMENTATION 3/4.3.7. 1 RADIATION MONITORING INSTRUMENTATION.

The OPERABILITY of the radiation monitoring instrumentation ensures that; (1) the radiation levels are continually measured in the areas served by the individual channels; (2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded; and (3) sufficient information is available on selected plant parameters to monitor and assess these variables following an accident.

This capability is consistent with the recommendations of NUREG-0737, "Clarification of THI Action Plan Requirements,"

November, 1980.

3.4.3.7. 2 SEISMIC MONITORING INSTRUMENTATION The OPERABILITY of the seismic monitoring instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety.

This capability is required to permit comparison of the measured response to that used in the design basis for the unit.

This instrumentation is consis-tent with the recommendations of Regulatory Guide 1.12 "Instrumentation for Earthquakes",

April 1974.

SUSQUEHANNA " UNIT 2 B 3/4 3-4

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REACTOR COOLANT SYSTEM BASES PRESSURE/TEMPERATURE LIMITS (Continued)

The pressure-terperature lfrftlines shown fn Ffgure 3.4.6.1-1, curves C

and A, for reactor criticality and for fnservice"leak and hydrostatic'testing have been provided to assure corplfance with the rinfae terperature requfre-rents of Appendix G to 10 CFR Part 50 for reactor criticality and for inservice leak and hydrostatic testing.

3/4.4.7 MAIN STEAM LINE ISOLATION VALVES Double isolation valves are provided on each,,of,.the:rafa,>Mar.lines to.

minimize the potential leakage, paths. from the contafnrent fn case of a line break.

Only one valve fn each line fs required to rafntafn the integrity of the contafnrent.

The surveillance requfrerents are based-on the;.operating history of this type valve.

The maxfmur closure tire has been selected to contain ffssfon products and'o ensure'he cod% fs"hot-'ncovered'fdlTeMng 1fne breaks.

3/4.4.8 'TRUCTURAL INTEGRITY The fnspectfon progrars for-ASME Code Class 1-," 2'and 3'"'coeponents'ensure that the structural"fntegrfty of these corponents:id11 b4 rafntafned at an acceptable level throughout the life of the plant.

Corponents of the reactor coolant syster were designed to provide access to perrft fnservfce fnspectfons'"fn accordance'with-'Sectton XX-of'he'ASME Boile~

and Pressure Vessel Code X971 Edftfin-and'Addenda through"X972."

The inservic>> fnspectfon prbgrar for"ASME Cede'Class~1, 2 and 3 components will be perforred fn accorda'nce with Sictfon"XI of'he ASMK"Boiler and Pressure Vessel Codi and applicable addenda as required by 10 CFR 50.55a(g) except where specfffc."written relief has been granted by the NRC pursuant to 10 CFR 50.55a(g)(6)(f).

3/4.4. 9 RESIDUAL HEAT REMOVAL A single shutdown cooling mode loop provides sufficient heat reroval capability for removing core decay heat and mixing to assure accurate teapera-ture indication, however, single failure considerations require that two loops be OPERABLE or that alternate methods capable of decay heat reaoval be demonstrated and that an alternate method of coolant mixing be fn operation.

BASES TABLE B 3/4.4.6-1 REACTOR VESSEL TOUGHNESS CM0Cm zz Cz

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Weld Seuj I.D:,-,-or lI/fit':!,";:Tyje.";:".'-"

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0.13 Ni(%)

0.68

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56.7 N/P 46.7 MIn, Upper

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Weld NIA 624263/

E204A27A 0.06 0.89

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NOTE: 'hese values are given only for the benefit of calculating the 32 EFPY RT per R.G. 1.99 Rev. 2.

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3 4.5 EMERGENCY CORE COOLING SYSTEM BASES 3 4.5.1 and 3 4.5.2 ECCS - OPERATING AND SHUTDOWN The core spray system (CSS) is provided to assure that the core is adequately cooled following a,loss-of-coolant

accident, and together with the LPCI mode of the RHR system, provides adequate core cooling capacity for all break sizes up to and including the double-ended reactor recirculation line break, and for smaller breaks following depressurization by the automatic depressurization system (ADS).

The CSS is a primary source of emergency core cooling after the reactor vessel is depressurized and a source for flooding of the core in case of accidental draining.

The surveillance requirements provide adequate assurance that the CSS will be OPERABLE when required.

Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a

complete functional test requires reactor shutdown.

The pump discharge piping is maintained full to prevent water hammer damage to piping and to start cooling at the earliest moment.

The low pressure coolant injection (LPCI) mode of the RHR system is provided to assure that the core is adequately cooled following a loss-of-coolant accident.

Two subsystems, each with two pumps, provide adequate'core flooding for all break sizes up to and including the double-ended reactor recirculation line break, and for small breaks following depressurization by the ADS.

The surveillance requirements provide adequate assurance that the LPCI system willbe OPERABLE when required. Although all active components are testable and fullflowcan be demonstrated by recirculation through a test loop during reactor operation, a

complete functional test requires reactor shutdown.

The pump discharge piping is maintained full to prevent water hammer damage to piping and to start cooling at the earliest moment.

The high pressure coolant injection (HPCI) system is provided to assure that the reactor core is adequately cooled to limitfuel clad temperature in the event of a small break in the reactor coolant system and loss of coolant which does not result in rapid depressurization of the reactor vessel.

The HPCI system permits the reactor to be shut down while maintaining sufficient reactor vessel water level inventory until the vessel is depressurized.

The HPCI system continues to operate until reactor vessel pressure is below the pressure at which CS system operation or'LPCI mode of the RHR system operation maintains core cooling.

'he capacity of the system is selected to provide the required core cooling. The HPCI pump is designed to deliver greater than or equal to 5000 gpm at reactor pressures between 1187 and 150 psig. Initially,water from the condensate storage tank is used instead of injecting water from the suppression pool into the reactor, but no credit is taken in the safety analyses for the condensate storage tank water.

SUSQUEHANNA - UNIT 2 B 3!4 5-1 Amendment No. 103

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EMERGENCY CORE COOLING SYSTEM BASES ECCS-OPERATING and SHUTDOWN (Continued.',

With the HPCI system inoperable, adequate core cooling-is assured by the OPERABILITY of the redundant and diversified automatic depressurization system and both the CS and LPCI systems.

In addition, the reactor core isolation cooling (RCIC) system, a system for which no credit is taken in the safety analysis,,will automatically provide makeup at reactor operating pressures on a reactor low water level condition.

The HPCI out-of-service period of 14 days is. based on the demonstrated OPERABILITY of redundant and diversified low pressure core cooling systems and the RCIC system.

The surveillance requirements provide adequate assurance that the HPCI system will be OPERABLE when required.

Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test with reactor vessel injection requires reactor shutdown.

The pump discharge piping is maintained full to prevent'water haaeer damage and to provide cooling at the earliest moment.

Upon failure of the HPCI system to function properly after a small break loss-of-coolant accident, the automatic depressurization system (ADS) automa-tically causes selected safety-relief valves to open, depressurizing the reactor so that flow from the low pressure core cooling systems can enter the core in time to limit fuel cladding temperature to less than 22004F.

ADS is conserva-tively required to be OPERABLE whenever reactor vessel pressure exceeds 100 psig.

This pressure is substantially below that for which the low pressure core cooling systems can provide adequate core cooling for events requiring ADS.

ADS automatically controls six 'selected safety-relief valves although the safety analysis only takes credit for five valves.

It is therefore appropriate to permit one valve to be out-of-service for up to 14 days without materially reducing system reliability.

3/4.5. 3 SUPPRESSION CHAMBER The suppression chamber is required to be OPERABLE as part of the ECCS to ensure that a sufficient supply of water is available to the HPCI,=CSS and LPCI systems in the event of a LOCA.

This limit on suppression chamber minimum water volume ensures that sufficient water is available to permit recirculation cooling flow to the core.

The OPERABILITY of the suppression chamber in OPERATIONAL CONDITIONS 1, 2 or 3 is also required by Specification 3.6.2.1.

Repair work might require making the suppression chamber inoperable.

This specification will permit those repairs to be made and at the same time give assurance that the irradiated fuel has an adequate cooling water supply when the suppression chamber must be made inoperable, including draining, in OPERATIONAL CONDITION 4 or 5.

In OPERATIONAL CONDITION 4 and 5 the suppression chamber minimum required water volume is reduced because the reactor coolant is maintained at or below 2004F.

Since pressure suppression is not required below 2124F, the minimum water volume is based on NPSH, recirculation volume, vortex prevention plus a safety margin for conservatism.

SUSQUEHANNA - UNIT 2 B 3/4 5-2

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CONTAINMENTSYSTEMS BASES 3 4 6.2 DEPRESSURIZATON SYSTEMS The specifications of this section ensure that the primary containment pressure willnot exceed the design pressure of 53 psig during primary system blowdown from full operating pressure.

The suppression chamber water provides the heat sink for the reactor coolant system energy release following a postulated rupture of the system.

The suppression chamber water volume must absorb the associated decay and structural sensible heat released during reactor coolant system blowdown from 1053 psia.

Since all of the gases in the drywell are purged into the suppression chamber air space during a loss of coolant accident, the pressure of the liquid must not exceed 53 psig, the suppression chamber maximum pressure.

The design volume of the suppression chamber, water and air, was obtained by considering that the total volume of reactor coolant and to be considered is discharged to the suppression chamber and that the drywell volume is purged to the suppression chamber.

Using the minimum or maximum water volumes given in this specification, containment pressure during the design basis accident is approximately 45.0 psig which is below the design pressure of 53 psig.

Maximum water volume of 133,540 ft results in a downcomer submergence of 12 feet and the minimum volume of 122,410 ft'esults in a submergence approximately 24 inches less. The majority of the Bodega tests were run with a submerged length of 4 feet and with complete condensation.

Thus, with respect to the downcomer submergence, this specification is adequate.

The maximum temperature at the end of the blowdown tested during the Humboldt Bay and Bodega Bay tests was 1704F and this is conservatively taken to be the limit for complete condensation of the reactor

coolant, although condensation would occur for temperatures above 170'F.

Should it be necessary to make the suppression chamber inoperable, this shall only be done as specified in Specification 3.5.3.

Under full power operating conditions, blowdown from an initial suppression chamber water temperature of 904F results in a water temperature of approximately 128'F immediately following blowdown which is below the 1704F used for complete condensation via quencher devices.

At this temperature and atmospheric pressure, the available NPSH exceeds that required by both the RHR and core spray pumps, thus there is no dependency on containment overpressure during the accident injection phase.

If both RHR loops are used for containment cooling there is no dependency on containment overpressure for post-LOCA operations.

Experimental data indicate that excessive steam condensing loads can be avoided ifthe peak local temperature of the suppression pool is maintained below 200 F during any period of relief valve operation.

Specifications have been placed on the 'envelope of reactor operating conditions so that the reactor can be depressurized in a timely manner to avoid the, regime of potentially high suppression chamber loadings.

SUSQUEHANNA - UNIT 2 8 3/4 6-3 Amendment No.

103

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8ecause of the large volume and thermal capacity of the suppression pool, the volume and temperature normally changes very slowly and monitoring these parameters daily is sufficient to establish any temperature trends.

By requiring the suppression pool temperature to be frequently recorded during periods of significant heat addition; the temperature trends will be closely followed so that appropriate action can be taken. The requirement for an external visual examination followingany event where potentially high loadings could occur provides assurance that no significant damage was encountered.

Particular attention should be focused on structural discontinuities in the vicinity of the relief valve discharge since these are expected to be the points of highest stress.

In addition to the limits on temperature of the suppression chamber pool water, operating procedures deffne the action to be taken in the event a safety-relief valve lnadvertentty opens or sticks open.

As a minimum this action shall include:

(1) use ot all available means to close the valve, I2) initiate suppression pool water cooling, I3) initiate reactor shutdown, and (4) if other safety-relief valves are used to depressurize the reactor, their discharge shall be separated,.

from that of the stuckwpen safety relief valve to assure mixing and uniformity ot energy'nsertion to the pool.

Ourfng a LOCA. potential leak paths. between the drywall and suppression chamber airspace could result in excessive containment pressures, since the steam flow into the airspace would bypass the heat sink capabllltfes ot the pool.

Potential sources of bypass leakage are the suppression chamber-drywall vacuum breakers IVBs), penetratlons In the diaphragm floor, and cracks in the diaphragm floor/liner plate and downcomers.located in the suppression chamber afrspace.

The containment pressure response to the postulated bypass leakage can be mitigated by manually actuating the suppression chamber sprays. An analysis was performed for a design bypass leakage area of A/(k)'~equal to 0.0636 ft'o verifythat the operator has sufficient time to initiate the sprays prior to exceeding the containment design pressure of 53 psig.

The limit of 10% of the design value of 0.0535 ft'nsures that the design basis for the steam bypass analysis is met.

The drywall-to-suppression chamber bypass test at a differential pressure of at least 4.3 psi verifies the overall bypass leakage area for simulated LOCA conditions ls less than the specified limit. For those outages where the drywall-to-suppression chamber bypass leakage teat is not conducted, the VB leakage test verifies that the VB leakage area ls less than the bypass limit.

with a 70% margin to the bypass limitto accommodate the remaining potential leakage area through the passive structural components.

Previous drywall-~uppressfon chamber bypass test data indicates that the bypass leakage through the passive structural components will be much fess than the 70% margin.

The VB leakage limit, combined with the negligible passive structural leakage area, ensures that the drywall-to-suppression chamber bypass leakage limit is mat for those outages for which the drywall-to.suppression chamber bypass teat is not scheduled.

The OPERABIUTY of the primary contalnmant isolatfon valves ensures that the containment atmosphere willbe isolated from the outside environment in the <<vent of a release of radioactive material to the containment atmosphere or pressurization ot the containment and la consistent with the requirements of GOC 54 through 57 of Appendix A to 10CFR 50.

Containment isofatfon within the time limits specified forthose isolation valves designed to close automatically

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DESIGN FEATURES 5.3 REAC R COR FUEL ASSE BLIES 5.3.1 The reactor core shall contain 764 fuel assemblies.

Each assembly consists of a matrix of Zircaloy clad fuel rods with an initial composition of non-enriched or slightly enriched uranium dioxide as fuel material and water rods.

Limited substitutions of Zirconium alloy filler rods for fuel rods, in accordance with NRC-approved applications of fuel rod configurations, may be used.

Fuel assemblies shall be limited to those fuel designs that have been anatyzed with applicable NRC staff-approved codes and methods, and shown by test or analyses to comply with all fuel safety design bases.

A limited number of lead use assemblies that have not completed representative testing may.be placed in non-limiting core regions.

Each fuel rod shall have a nominal active fuel length of 150 inches.

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

CONTROL R D ASSEMBLIES 5.3.2 The reactor core shall contain 185 cruciform shaped control rod assemblies.

The control material shall be boron carbide powder (BiC), and/or Hafnium metal.,

The control rod shall have a nominal axial absorber length of 143 inches.

Control rod assemblies shall be limited to those control rod designs approved by the NRC for use in BWRs.

5.4 REACTOR COOLANT SYSTE DESIGN PRESSURE A D TE E

UR 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, eC b.

For a pressure of:

'I; 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 5754F.

VOLUM 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 T,, of 532'F.

SUSQUEHANNA - UNIT 2 5-B amendment Me. 9 ~'03

DESIGN FEATURES

5. 5 HETEOROLOGICAI TOMER LOCATION

5. 5. 1 The meteoro1ogical tower shall be located as shown on Figure 5. 1. 1-1.

5.6 FUEL STORAGE CRITICALITY 5.6.1. 1 The spent fuel storage racks are designed and shall be maintainea with:

a.

A k ff equivalent to less than or equal to 0.95 when flooded with unborated water, which includes all calculational biases and uncertainties as described in Section 9.1.2 of the FSAR.

b.

A nominal 6.625 inch center to-center distance between fuel assemblies placed in the storage racks.

5.6.1.2 The k ff for new fuel for the first core loading stored dry in the

'pent fuel storage racks shall not exceed 0.98 when aqueous foaw moderation is assumed.

DRAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 816'9".

CAPACITY 5.6.3.1 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 2840 fuel assemblies.

5. 6..3.2 A multi-purpose storage rack may be used to store up to IO sound and/or defective fuel assemblies and/or other reactor internals.

5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.7.1 The components identified in Table 5.7.1-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7.1-1.

SUSQUEHANNA - UNIT 2 5-7

TABLE 5.7.1-1 COMPONENT CYCLIC OR TRANSIENT LIMITS COMPONENT CYCLIC OR TRANSIENT LIMIT DESIGN CYCLE OR TRANSIENT Reactor 120 heatup and cooldown:

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cycles 80 step change cycles 180 reactor trip cycles 130 hydrostatic pressure and leak tests

'0'F to 551 F to 70'F Loss of feedwater heaters 100% to 0% of RATED THERMALPOWER Pressurized to R 930 psig and 6 1250 psig.

SUSQUEHANNA -,UNIT 2 5-8 Amendment No. ip3

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I (Continued) 6.9.3.2 The analytical methods used to determine the core operating limits shall be those topical reports and those revisions andlor supplements of the topical report pfeviougy reviewed and approved by the NRC. which describe the methodology applicable to the current cycle.

For Susquehanna SES the topical reports are:

PL-NF-87-001-A, Qualification of Steady State Core Physics Methods for BWR Design and Analysis, July, 1988.

2.

3.

4.

7.

PL-NF-89-005-A, "Qualification of Transient Analysis Methods for SWR Design and Analysis, July, 1992.

PL-NF-90-001-A, Application of Reactor Analysis Methods for BWR Design and Analysis." July. 1992.

XN-NF-80-19(P)(A), Volume 4, Revision 1, "Exxon Nuclear Methodology for Soiling Water Reactors: Application of the ENC Methodology to BWR Reloads.

Exxon Nuclear Company, inc.. June 1986.

XN-NF-85-67(P)(A), Revision 1.

Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel, Exxon Nuclear Company, inc., September 1986.

PLA-3407, "Proposed Amendment 132 to License No. NPF-14: Unit 1 Cycle 6

Reload, Letter from H. W. Kaiser (PPLL) to W. R. Butler (NRC), July 2, 1990.

Letter from Elinor G. Adensam (NRC) to.H. W. Kaiser (PPSL),

Issuance of Amendment No. 31 to Facility Operating Ltcense No.. NPF Susquehanna Steam Electric Station, Unit 2.

October 3, 1986.

PLA-3533, Revised Proposed Amendment 67 to License No. NPF-22: Unit 2 Cycle 5 Reload,'etter from H. W. Kaiser (PPLL) to W. R. Butler (NRC), March 7, 1991.

9.

XN-NF-84-97, Revision 0, LOCA-Seismic Structural Response of an ENC 9x9 Jet Pump Fuel Assembly, Exxon Nuclear Company, Inc., December 1984.

10. PLA-2728.

Response to NRC Question: SeismiclLOCA Analysis of U2C2 Reload, Later from H. W. Kaiser (PP5L) to E. Adenstam (NRC), September 25, 1986.

11. XN.NF-8246(P)(A). Supplement 1, Revision 2, Qualification of Exxon Nuclear Fuel for Extended Bumup Supplement 1 Extended Bumup Qualification of ENC 9x9

Fuel, May 1988.

12. XN-NF-80-19(A), Yolume 1, and Yolume 1 Supplements 1 and 2, Exxon Nuclear Methodology for Boiling Water Reactors:

Neutronic Methods for Oesign and AnaiysN.

Exxon Nuclear Company, Inc.. March 1683.

13. XN-NF-524(A), Revision 1, Exxon Nuclear Crftlcal Power Methodology for Boiling Water Reactors, Exxon Nuclear Company, lnc., November 1983.

SUSQUEHANNA - UNIT 2 6-20a Amendment No.~

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ADIVIINISTRATIVECONTROLS CORE OPERATING LIMITS REPORT (Continued)

14. XN-NF-512-P-A, Revision 1 and Supplement 1, Revision 1, "XN-3 Critical Power. Correlation," October, 1982.

15. NEDC-32071P,"SAFER/GESTR-LOCALossofCoolantAccidentAnalysis,"

GE Nuclear Energy, May 1992.

16. NE-092-001A, Revision 1, "Licensing Topical Report for Power Uprate With Increased Core Flow," Pennsylvania Power & Light Company, December 1992.

1 17.

NRC SER on PP&L Power Uprate LTR (November 30, 1993).

6.9.3.3 The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, transient analysis limits and accident analysis limits) of the safety analysis are met.

6.10 RECORD RETENTION ln addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated.

6.10.1 The following records shall be retained for at least 5 years:

a.

Records and logs of unit operation covering time interval at each power level.

SUSQUEHANNA - UNIT 2 B-20b Amendment No. ~),

103

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-Om1 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT N0.103TO FACILITY OPERATING LICENSE NO. NPF-22 PENNSYLVANIA POWER

& LIGHT COMPANY ALLEGHENY ELECTRIC COOPERATIVE INC.

SUS UEHANNA STEAM ELECTRIC STATION UNIT 2 DOCKET NO. 50-388

1.0 INTRODUCTION

By letter dated November 24,

1993, as supplemented by letters of January 7 and February 14, 1994, the Pennsylvania Power and Light Company (PP8L or the licensee) submitted a request for a revision of Facility Operating License No.

NPF-22 for Susquehanna Steam Electric Station (SSES),

Unit 2, to uprate the current licensed power level from 3293 MWt to a new limit of 3441 MWt.

The amendment application also submitted a number of changes to the Technical Specifications (TSs) to implement uprated power operation.

The supplemental letter of January 7,

1994, cor rected a single typographical error.

The February 14, 1994, letter transmitted an affidavit.

The supplemental letters did not affect the application or the staff's initial proposed no significant hazards consideration determination.

2.0 EVALUATION PP&L's letter of June 15,

1992, submitted "Licensing Topical Report NE-092-001, Revision 0, for Power Uprate With Increased Core Flow," for Susquehanna Steam Electric Station (SSES),

Units 1

and 2.

The report was submitted to support future proposed amendments to the Units 1 and 2 licenses to permit a

4.5-percent increase in reactor thermal power and an 8-percent increase in core flow for each unit.

The initial submittal was revised and supplemented by letters of July 24, September 17, and December 18,

1992, and January 8,

January 25, April 2, August 5, August 12, and September 29, 1993.

On November 30,

1993, the Director, Office of Nuclear Reactor Regulation, issued a letter, supported by an enclosed safety evaluation, which informed PP&L that the revised licensing topical report adequately supported the proposed power uprate for SSES.

Therein the staff concluded that SSES could operate safely with the proposed 8-percent increase in core flow, the proposed 4.5-percent increase in reactor thermal power, the corresponding 5-percent

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increase in main turbine inlet steam flow, and the corresponding increases in flows, temperatures, pressures, and capacities required in supporting systems and components at these uprated conditions, but that authorization for any increase in reactor thermal power would be based on a review of the TS changes submitted with the amendment application.

The safety evaluation and letter are attached.

As stated in the conclusion section of the November 30, 1993, safety evaluation, there were four open items that PP&L was to address in the proposed license amendment application.

These four items were (1) the startup test plan, (2) the anticipated transient without scram (ATWS) analysis, (3) the pipe whip and jet impingement evaluation, and (4)"the program to upgrade the emergency operating procedures.

PP&L addressed each of these items in the subject amendment application.

2.1 Post-Power Uprate Startup Test Program PP&L plans to perform a post-power uprate startup test program similar in denature to the original Susquehanna startup test program described in Chapter 14 of the Final Safety Analysis Report (FSAR), but with the scope of testing limited to those tests or portions of tests affected by power uprate or increased core flow.

The test program will be conducted in four separate segments or test plateaus.

Each test plateau will contain one or more test conditions which defines'uprate power levels and core flows at which the tests are to be performed.

The test plateaus and test conditions were described in the application.

The current 100-percent power level (3293 HWt) represents about 95.7-percent power of the proposed maximum.uprated power level (i.e.,

100X power equals 3441 HWt).

One of the test plateaus will bracket this point (i.e.,95-96X of the uprate power level) with varying core flow.

The last two plateaus are at 97-98-percent and 99-100-percent of the proposed uprated power level.

Generally, all tests scheduled to be performed in one test condition are to be completed before proceeding to the next higher test condition.

After all testing in each plateau is completed, the test results for all tests will be reviewed by the Plant Operations Review Committee (PORC) before operations authorization is given to proceed to the next test plateau.

The requirements for power-uprate startup testing come from a review of Chapter 14 of the FSAR, the General Electric (GE)

Power Uprate Startup Test Specification, the proposed TSs for power uprate and the Susquehanna Licensing Topical Report NE-092-001 described previously.

The tests which will be performed for the power-uprate startup test program were described in five tables in Attachment 1 to PP&L's application.

The staff has reviewed the proposed test program and finds it acceptable.

It is recognized that changes to the test program may occur as it is executed.

2.2 High Energy Line Breaks In Section 3.9. 1 of the November 30, 1993, safety evaluation, the staff reported that the licensee was still evaluating the calculations supporting the disposition of potential targets of pipe whip and jet impingement from

postulated high energy line breaks (HELBs) to determine the effects of the power uprate.

The staff also stated that the licensee expected the evaluation to confirm the adequacy of the existing design under power-uprate conditions.

Because the licensee had not completed these calculations, the staff could not reach any conclusion regarding the impact of the uprated power level operation on HELBs.

In the November 24, 1993, letter, the licensee submitted information to indicate that these calculations were complete.

The results of the licensee's analysis showed that the effects of power uprate on HELBs were proportional to the increase in reactor vessel pressure which resulted in higher loads,

stresses, and displacements on the piping, supports, and whip restraints.

However, the increases were relatively smal.l

and, as expected, the original design-basis HELB commitments in the FSAR were still satisfied.

The staff has reviewed the results of the licensee's analysis and concurs with the licensee's conclusions that, for the power uprate, no further action is required regarding protection against the effects of pipe whip and jet impingement due to HELBs.

The resul.ts of the analysis are consistent with the results of analyses performed at other plants during similar power uprates.

The staff, therefore, concludes that protection against the effects of postulated breaks in HELBs will remain acceptable after the power uprate.

2.3 Anticipated Transient Without Scram Analysis (ATWS)

PP8L had not addressed the Susquehanna ATWS analysis for power-uprate conditions in the Susquehanna Licensing Topical Report NE-092-001 because the licensee had not completed the calculations and analyses when the topical report was submitted.

Although GE has performed generic bounding ATWS

analyses, these analyses cannot be used for Susquehanna because the licensee:

(I) uses non-GE fuel and (2) has taken exceptions to Revision 4 of the emergency procedure guidelines (EPGs) for responding to ATWS, which are assumed in the GE generic analyses.

The results of the ATWS analysis for SSES, Unit 2, for power-uprate conditions were sent with the November 24, 1993, submittal.

Seven limiting events were analyzed.

All events were initiated at the extended load line limit, 100 percent of uprated power (3441 HWt) and 87 percent of rated core flow (87 HLb/hr)..

The licensee's ATWS analysis predicts that the most limiting transient is rapid closure of the main steam isolation valves (HSIVs).

In this pressurization transient, the computer analyses predict that the peak reactor pressure vessel pressure could reach 1317 psig, the peak suppression pool temperature could rise to 178.9 'F and the peak fuel cladding temperature could be 1463 'F.

The staff has reviewed the licensee's ATWS analysis for Unit 2 for power-uprate conditions and has determined that the results are acceptable.

2.4 Emergency Operating Procedures Emergency operating procedures (EOPs) to support uprated power operation are under development with implementation, to include operator training, scheduled to take place before startup in Fuel Cycle No. 7.

Presently, the plant-specific technical guidance has been revised and verified.

All EOPs to support power uprate have been revised and reviewed by shift management and training personnel.

Comments are being resolved, and five of the six affected EOPs have been completed and are being verified.

The sixth EOP is in the comment resolution stage.

The final revised EOPs for power uprate will be reviewed in the same manner as other changes to the EOPs are being reviewed during the normal inspection programs.

2.5 Proposed TS Changes Operation with a 4.5-percent increase in reactor thermal power and an 8-percent increase in core flow results in a 5-percent increase in main turbine inlet steam flow, approximately a 30 psig increase in design reactor pressure and other changes in system pressures, temperatures and flows.

To implement "

the power uprate, the licensee submitted a number of changes to the TSs to revise such parameters as the authorized power level, core flow, reactor

pressure, steam pressures and flows, turbine first-stage pressure setpoints, average power range monitor (APRN) setpoints for two-loop and single-loop operation, changes in some reactor protection system (RPS) setpoints (such as the turbine pressure that initiates the recirculation pump trip system),

high pressure coolant injection (HPCI) steamline flow and pump discharge

pressure, thermal power stability restrictions, and resetting the safety/relief valve setpoints.

The specific TS changes are as follows:

I.

Change Oefinition 1.33 to redefine rated thermal power as 3441 megawatts thermal.

The staff's safety evaluation of November 30,

1993, evaluated all aspects of operation of the Susquehanna units at an increased thermal power of 3441 megawatts including:

the reactor thermohydraulic and neutronic performance, thermal-hydraulic stability, the ability of the control rod drive system to control core reactivity at the increased reactor pressure, the structural integrity of the reactor coolant and connected

systems, overpressure protection with the new safety-relief valve settings, the effect of revised LOCA loads on the reactor

system, containment systems and emergency core cooling system performance, the effect of increased core flow on reactor internals and

pumps, the performance of the steam, feedwater and auxiliary systems, the capability of the High Pressure Coolant Injection, Reactor Core Isolation Cooling, Residual Heat Removal and Core Spray Systems, the impact of the increased thermal power on containment system and standby gas treatment system performance, the changes to the plants'nstrumentation and control

systems, the functioning of all safety-related service water systems, the capability of the non-safety-related cooling systems, the impact of the increased thermal power on the heating, ventilating and air conditioning

systems, the impact on the radwaste

systems, the impact of the increased thermal power on postulated design basis accidents, the environmental

qualification of mechanical and electrical equipment under the incr eased pressures, temperatures and humidity and the effect of the increased power on generic issues.

The staff also issued an environmental assessment, dated March 11, 1994, that evaluated the potential impact of operation at the increased thermal power with respect to potential radiological and non-radiological effects on the environment.

As part of the power uprate

program, the licensee conducted an extensive design-basis reconstitution and design basis upgrade program.

The NRC staff in effect performed a

licensing review of all systems that would be effected by operation at increased thermal power and the associated increased

core, feedwater and steam system flows and pressures.

As a result of the extensive evaluation, the staff concluded that the Susquehanna units can operate safely with a 4.5 percent increase in reactor thermal

power, an 8 percent increase in core flow, the corresponding 5 percent increase in steam flow and the corresponding increases in flows, temperatures, pressures and capacities required in supporting systems and components.

The proposed increase in thermal power from 3293 MWt to 3441 MWt is acceptable.

2.

In Section

2. 1. 1 and 2. 1.2, replace the reference to 10 percent of'ated'=.'ore flow with a reference to the actual core flow of 10 million lbs/hr

'nder power uprate conditions.

The references to "rated core flow" in TS 2. 1. 1 and 2. 1.2 have been deleted to avoid confusion since allowable core flow is being increased by 8 percent.

As discussed in the Bases for TS 2. 1. 1, boiling transition will not occur in fuel bundles if core power is less than 25 percent of rated thermal power, regardless of pressure or core flow.

Specifying a specific minimum core flow before exceeding 25 percent power is more precise than specifying a percentage of maximum core flow and is acceptable.

3.

In Table 2.2. 1-1, Reactor Protection System Instrumentation Setpoints, Item 3, change the trip setpoint and allowable value for Reactor Vessel Steam Oome Pressure-High to g 1087 psig and g 1093 psig, respectively, to reflect the higher reactor pressure with power uprate.

This scram function is designed to terminate a pressure increase transient not terminated by direct scram or high flux scram.

The nominal trip setpoint is maintained above the reactor vessel maximum operating pressure.

The allowable value is set below the analytical limit used in the transient analyses.

For the uprated transient

analyses, the licensee used 1105 psig.

The results of the overpressure protection analyses using this revised analytical limit showed that the peak pressure remained below the 1375 psig American Society of Mechanical Engineers (ASME) limit and met all licensing requirements.

The 36 psig increase in the allowable value to < 1093 psig is acceptable as well as the new increased trip setpoint.

4.

In the Bases for Section

2. 1. 1 on Thermal

Power, change the value on fuel bundle radical peaking factor at 25 percent thermal power from "greater than 3.0" to "approximately 3.0" because of the higher thermal power with power uprate.

This is still si'gnificantly higher than the expected peaking factor and is acceptable.

5.

In the Bases for the Reactor Protection System Instrumentation Setpoints, add a paragraph to 2.2. 1.9 on Turbine Stop Valve - Closure and 2.2. 1. 10 on Turbine Control Valve Fast Closure to clarify that the anticipating scram function is not required when Thermal power is below 30 percent, since the turbine bypass valves can bypass up to 30 percent of the steam flow directly to the condenser to alleviate a potential pressurization transient.

The added Bases also notes that the new analytical limit, used in the transient

analyses, is 147.7 psig, which is equivalent to 30 percent rated thermal power under uprated power conditions.

The added paragraphs are clarifications rather than changes to the present Bases and are acceptable.

6.

Revise specification 4. 1.5.C to require the Standby Liquid Control pumps to develop a discharge pressure greater than or equal to 1224 psig versus the current requirement of 1190 psi.

The increased discharge pressure acceptance criteria is based on the increased reactor pressure with power uprate and takes into account that operating with increased core flow will result in additional friction losses through the core and a slightly larger core differential pressure (approximately 4 psi).

The 34 psig increase in Standby Liquid Control pump test discharge pressure acceptance criteria ensures that the pumps will inject sufficient sodium pentaborate into the core at the approximately 30 psig increased reactor pressure to bring the reactor subcritical.

The increased acceptance criteria is acceptable.

7.

TS 3.2.2 on Average Power Range Monitor (APRM) Setpoints contains the definition of "W" for the flow biased APRM scram equation.

The word "rated" is being deleted from the definition of "W" since rated core flow is being increased.

The definition of "W" is not altered.

The change is being made for editorial purposes and is acceptable.

8.

Action 6 in Table 3.3. 1-1 on Reactor Protection System Instrumentation is being revised to clarify the current requirements.

The revision does not change the intent.

Action 6 currently reads:

"Initiate a reduction in THERMAL POWER within 15 minutes and reduce turbine first-stage pressure until the function is automatically bypassed. within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> s."

As noted in Item 5 above, the turbine bypass valves can bypass up to 30 percent of the steam flow directly to the condenser.

The licensing basis analysis for the Minimum Critical Power Ratio (MCPR) operating limit (for the Generator Load Rejection Without Bypass) transient takes credit for operation of the anticipating scram on control valve fast closure at greater than 30 percent of rated thermal power.

The revision to Action 6 clarifies that the action only applies when the Reactor Protection System (RPS) scram functions and End-of-Cycle Recirculation Pump Trip (EOC-RPT) on turbine main stop valve closure or control valve fast closure are not automatically bypassed.

The revised Action 6 reads:

"Initiate a

reduction in THERMAL POWER within 15 minutes and reduce THERMAL POWER to less than 30 percent of rated THERMAL POWER within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />."

The revisions to the action statement clarify the current requirements; they do not change their intent and are acceptable.

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9.

Note (j) in Table 3.3. 1-1 on Reactor Protection System Instrumentation is being revised to increase the scram bypass limit to 136 psig from 108 psig to reflect the higher steam pressure with power uprate.

The setpoint change is related to Item 8 above.

Setting the value of first-stage turbine pressure at 136 psig ensures that the analytical limit of 147.7 psig, which represents 30 percent rated thermal power, is not exceeded.

The proposed revision to Table 3.3. 1-1, Note (j), and Table 3.3.4.2-1, Note (b), does not change the operation of the RPS and EOC-RPT bypasses on turbine stop valve closure and control valve fast closure below 30 percent power.

The turbine first stage pressure switches will still be calibrated in the same

manner, and, by procedure, the reactor operator will not exceed 30 percent power if the trip bypass annunciator does not clear.

The setpoints for the RPS and EOC-RPT bypass functions were selected to allow sufficient operating margin to avoid scrams during low power turbine generator trips.

This small absolute setpoint increase maintains the safety basis for the setpoint and is acceptable.

10.

In Table 3.3.2-2, the main steam line flow high differential pressure setpoint is being changed from g 107 psid to S

113 psid and the allowable value is being changed from < 110 psid to g 121 psi'd to reflect the hi'gher steam line pressure with power uprate.

Footnote "**" is being added to Table 3.3.2-2 to indicate that these values will be confirmed during the power uprate startup testing.

If revisions to the setpoint and allowable value are required, they will be forwarded to the NRC for approval within 90 days of the completion of the test program.

The main steam line flow high differential pressure setpoint changes reflect the redefinition of rated main steam line flow that occurs with power uprate.

The allowable value is maintained at the same percentage of rated steam flow as the differential pressure changes due to the increased uprate steam flow.

The analytical limit of 140 percent of uprated steam flow is maintained for the uprated analyses.

The relationship between the allowable value and the analytical limit was retained to ensure that a

trip avoidance margin is maintained for the normal plant testing of HSIV's and turbine stop valves.

The increase in the absolute value of the trip setpoint still provides a high assurance of isolation protection for a main steam"line break accident which satisfies the original intent of the design.

The proposed main steam line flow high differential pressure setpoint changes are acceptable.

11. In Table 3.3.2-2, the Reactor Water Cleanup (RWCU) system flow-high isolation trip setpoint is being changed from 426 gpm to 462 gpm and the allowable value is being changed from 436 gpm to 472 gpm.

RWCU flow is being increased by 10 percent to maintain reactor coolant water chemistry at the higher power level and increased core flow.

The basis for the RWCU flow-high isolation is to ensure a

RWCU System isolation in case of a pipe break.

The high flow setpoint is set high

12.

13.

14.

15.

enough to avoid spurious trips from normal operating transients but low enough to ensure an isolation during a pipe break.

The proposed TS limits will result in a negligible reduction in the margin between the RWCU isolation setpoint and the 4350 gpm flow postulated during an RWCU line break and will avoid spurious isolations.

The proposed change in the trip setpoint maintains the original design intent with the 10 percent increase in the purification rate and is acceptable.

In Table 3.3.2-2, on Isolation Actuation Instrumentation Setpoints, the High Pressure Coolant Injection.(HPCI) and the Reactor Core Isolation (RCIC) steam line flow-high are being changed to account for changes in steam conditions and flows that result from operation at uprated conditions.

For the RCIC system, the trip setpoint and allowable value for the high delta pressure in the steam line are being increased to less than or equal to 138" HzO and less than or equal to 143" H 0, respectively.

The trip setpoint and allowable value for the HPCI steam line flow-high are being increased to less than or equal to 387" H 0 and less than or equal to 399" H<0, respectively.

The setpoint and allowable value are set so that isolation occurs at greater than 272X normal steam flow and less than 300X steam flow.

Setting the isolation at less than or equal to 300X of normal flow ensures that the isolation will occur if a steam line were to rupture.d The original setpoints were calculated using information obtained during the Susquehanna startup program.

The revised setpoints and allowable values were calculated using the same startup data and adjusted for uprate conditions.

The revised setpoints maintain the current design intent and are acceptable.

In Table 4.3.2. 1-1, on Isolation Actuation Instrumentation Surveillance Requirements, footnote "**" is being revised to delete reference to reactor pressure.

The original purpose of this footnote was to describe the functioning of the permissive circuitry that allowed the main steam isolation valves (HSIV) low condenser pressure isolation to be bypassed.

In the startup phase of the Susquehanna

units, GE deleted the reactor pressure setpoint input to the bypass circuitry.

This change is being made to have the footnote conform to the installed configuration.

This change is editorial in nature and is acceptable.

In Table 3.3.4.2-1, on End-of-Cycle Recirculation Pump Trip System Instrumentation, note "(b)" is being revised to specify that the EOC-RPT shall not be automatically bypassed when turbine first-stage pressure is greater than an allowable value of 136 psig for the reason stated in item 9, above.

This setpoint provides adequate margin between the analytical limit of 147.7 psig, which represents 30 percent rated thermal power (under power-uprate conditions) to ensure that the trip is not bypassed above

'30 percent power.

This maintains the current design requirement under uprate conditions and is acceptable.

In Table 3.3.6-2 (Page 3/4 3-54) on Control Rod Block Instrumentation

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Setpoints, and Specification 3.4. 1. 1.2.a.6.a (Page 3/4 4-lc),

on Single Loop Operation, the rod block monitor (RBH) flow biased rod blocks are being changed.

In the table, item l.a is being revised to change the trip setpoint and allowable value to less than or equal to 0.63 W + 41X and less than or equal to 0.63 W + 43X, respectively.

In the new specification 3.4. 1. 1.2.a.6.a, the trip setpoint and allowable values will be less than or equal to 0.63 W + 35X and less than or equal to 0.63 W +

37X, respectively.

The downward rescaling is made necessary by the re-definition of rated thermal power.

These TS changes do not represent a

change from current limits.

The RBH flow biased rod blocks are used in the Rod Withdrawal Error (RWE) analysis.

In order to maintain Critical Power Ratio (CPR) margins similar to previous Susquehanna

cycles, the flow biased rod blocks were changed in terms of megawatts thermal but the change was not appreciable.

The rescaling of the RBH flow biased rod block to reflect the re-definition of rated thermal power maintains the same level of protection as previously provided.

The proposed change to the RBH trip setpoints and allowable value maintain the current level of protection and are acceptable.

16.

In Table 3.3.6-2, Control Rod Block Instrumentation Setpoints, item 2.a.;

the Average Power Range Honitor (APRH) rod block upscale value has been changed to add a high flow clamp setpoint at 108X of rated thermal power with a high flow clamped allowable value at 111X.

The addition of the high flow clamp to the flow biased APRH rod block function maintains the normal margins between the rod block and the scram power levels in the increased core flow (ICF) regions.

When the reactor core flow is greater than 100 million ibm/hr, the APRH clamp provides an alarm to help the operator avoid scrams while operating in the ICF region.

The additional APRH trip provides an additional margin of safety in the ICF regions and is acceptable.

17. In Table 3.3.6-2, Control Rod Block Instrumentation Setpoints, item 6.a.,

the reactor coolant system recirculation flow upscale rod block trip setpoint and allowable value are being increased to 114/125 divisions of full scale and 117/125 divisions of full scale, respectively.

The upscale rod block setpoint and allowable value are being increased to allow operation. in the ICF region.

The purpose of the Reactor Coolant System recirculation flow upscale rod block is to prevent rod movement when an abnormally high increase in reactor recirculation flow causes an increase in neutron flux that results in an increase in reactor power.

However, this increase in neutron flux is monitored by the Neutron Honitoring System that can provide a rod block.

No design basis accident or transient analysis takes credit for rod block signals initiated by the Reactor Coolant Recirculation System.

The increase in the upscale trip setpoint from 108/125 divisions to 114/125 divisions of full scale and the increase in the allowable value from 111/125 divisions to 117/125 divisions is necessary to operate with increased core flow and is acceptable.

18. Surveillance Requirements 4.4.1. 1. 1.2 and 4.4.1.1.2.5 on the Reactor Coolant System are being revised to allow core flows in the ICF region of up to 108 million ibm/hr.

The reactor recirculation pump motor generator set scoop tube electrical and mechanical overspeed stop setpoints are being increased to a core flow of 109.5 million ibm/hr.

and 110.'5 million ibm/hr., respectively.

The electrical stop is maintained above the maximum operating core flow and below the mechanical'stop.

The 109.5 million ibm/hr. electrical stop setpoint, specified by General

Electric, is based on BWR operating history.

The electrical stop is a system design feature and is not used in any safety analysis.

The 110.5 million ibm/hr.

mechanical stop setpoint is used in transient analysis to limit core flow during a recirculation pump controller failure.

The 110.5 million ibm/hr.

mechanical stop setpoint, specified by General Electric, is also based on BWR operating history.

The cycle specific analyses, performed for power

uprate, used the 110.5 million ibm/hr. mechanical stop setpoint.

The 110.5 million ibm/hr setpoint was used by the licensee in the Unit 2, Cycle 7, reload analysis and is acceptable.

19. Figure 3.4. l.l. 1-1 on Thermal Power Stability Restrictions has been redrawn to reflect the new definition of Rated Thermal Power to retain the same stability operating restrictions in terms of megawatts thermal as currently prescribed by this graph.

The core thermal hydraulic stability curve and associated bases are maintained at the current rod lines and power levels.

Those values are redefined to reflect the redefinition of rated thermal power.

Since the current operating restrictions are maintained, power uprate has no detrimental effect on the level of protection provided by the TSs.

The revised figure pr ecludes operation in the region of potential thermal-hydraulic instability and is acceptable.

20.

A new specification, 3.4. 1. 1.2.5 is being added to the Limiting Condition for Operation (LCO) on the Reactor Coolant System, Recirculation Loops Single Loop Operation, to specify that a 0.70 Linear Heat Generation Rate (LHGR) multiplier has been added to Specification 3.2.4 when in single recirculation loop operation.

Operation with one recirculation loop out of service is allowed, but is not considered a normal mode of operation.

Single loop operation (SLO) is a special operational condition when only one of the two recirculation loops is operable.

In this operating condition, the reactor power will be limited to less than 80 percent of rated by the maximum achievable core flow, which is typically less than 60 percent of rated core flow.

A postulated LOCA (Loss of Coolant Accident) occurring in the active recirculation loop during SLO would cause a more rapid coastdown of the recirculation flow than would occur in two loop operation, where one active loop would remain intact.

This rapid coastdown causes an earlier boil'ing transition and deeper penetration of boiling transition into the bundle, which tends to increase the calculated PCT (Peak Clad Temperature).

However, the PCT effects of early boiling transition are substantially offset by the mitigating effect of the lower power level achievable at the start of such an event.

An LHGR reduction (multiplier) of 0.70 will be imposed when the plant is in SLO.

The SLO results are less limiting (i.e., lower PCT's) than the results for the two

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~ loop DBA LOCA.

Thus, the licensing PCT is based appropriately on two loop operation rather than SLO.

As discussed in Section 3.3.3 of the staff's safety evaluation of November 30, 1993, the. licensee used the staff-approved SAFER/GESTR (S/G) methodology to assess the Emergency Core Cooling System (ECCS) capability for meeting the 10 CFR 50.46 criteria.

The addition of an LHGR reduction of 0.70 when the plant is in SLO provides an additional margin of safety and is acceptable.

21. Footnote ****to Specification 4.4. 1. 1.2.7 on the Reactor Coolant System is being changed to reference the power uprate startup test program as distinguished from the initial startup test program when the unit was first licensed.

This footnote provided a mechanism for changing the power limits specified if the results of the initial startup test program determined that it was necessary.

The footnote is being modified to allow operation at uprated power with the present p'ower limits: 'Should the power uprate startup test program determine a need to'hange the power limits, they will be submitted to the Commission within 90 days as required by the revised footnote.

This is consistent with the original BWR startup test program requirement and is acceptable.

O

22. Specifications 4.4. 1. 1. 1.2, 4. 1. 1.2.5, 3.4. 1.3; and Figure 3:4. 1.1. 1-1, specify performance requirements and limits for the Reactor. Recirculation System.

These specifications are referenced to 102;5 percent and 105 percent of the current rated core flow.

The references to -"rated core flow" are being replaced with actual equivalent core flows; As discussed in item 18 above, the electrical and mechanical stops will be set at 109.5 million ibm/hr.

and 110.5 million ibm/hr., respectively.

The specifications are equivalent and unchanged..

This change is:-being made for editorial purposes to avoid confusion since rated core-flow is being increased.

These changes are also consistent with the changes made in Section

2. 1.

As discussed in the staff's safety evaluation of November 30, 1993, the staff evaluated operation of the Susquehanna units at increased core flows of up to 110.5 million ibm/hr and determined that the new mechanical and electrical setpoints were acceptable.

23. Specification 3.4.2, Reactor Coolant System,,Safety Relief Valves (SRV) is being changed to reduce the number of setpoint groups from 5 to 3.

Two valves will be set at 1175 psig plus or minus 1 percent, 6 will be set at 1195 psig plus or minus 1 percent and 8 will be set at 1205 psig plus or minus 1 percent.

Also, the number of Operable safety valves are being increased from 10 to 12.

The staff's assessment of the licensee's reactor overpressure protection analysis was discussed in Section 3.2.2 of the November 30, 1993, safety evaluation.

The licensee's analysis showed that for the most limiting pressurization transient, Main Steam Isolation Valve (MSIV) closure with failure of the valve position scram, the peak pressure remained below the 1375 psig ASME limit and met all licensing, requirements.

The margin'etween peak allowable pressure and the maximum safety setpoints (1205 psig

+

1 percent) is unchanged.

The difference is that in

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the present

TSs, 3 of the 16 SRVs are set at 1205 psig, whereas with the power uprate

TSs, 8 of the 16 SRVs will be set at 1205. psig.

.Thelicensee performed analysis on the effects of the setpoint changes for-the design conditions and the emergency and faulted conditions."'The increased RPV dome pressure does not affect the design condition-and, therefore, stresses remain unchanged.

With the changed setpoints, there will be reduction in the simmer margin which will be compensated: for by more stringent leak test requirements during valve refurbishment".

The Crosby SRVs used at Susquehanna have. not had the-problems of "weeping" associated with the Target Rock SRVs used at some other BWRs.

Since the licensee's analysis demonstrates that reactor pressure will be limited to within ASME Section III allowable values for the worst-case upset"transient; the revised SRV lift settings are acceptable.

24. Specification" 3.4.3.2.d, Reactor Coolant System; Operational--Leakage, is being revised to indicate that. the 1 gpm leakage rate'imit currently-"-.

applicable applies at the uprate'd maximum allowable pressure of 1035 psig.

The steam dome pressure for leakage is being increased by 35 psig (reactor design pressure).

This pressure is chosen on the basis of steam line pressure drop characteristics and excess-steam flow capability of the turbine observed-'during plant operation up to'the'<currentrrated:~power i

level.

Increasing the leakage rate pressure'to"1035"psig':is consistent with the expected-uprated-operating pressure.

'ncreasing the reactor steam dome pressure has been analyzed and found to'be within allowable limits.

Keeping the current' gpm leakage rate limit at the increased reactor system pressure is conservative and is acceptable.

25.

In Specification 3.4;6-2 and 4.4.6.2, Reactor Coolant System, Reactor Steam Dome, the reactor steam dome pressure limits have been increased from'040 psig to 1050 psig.

Operating pressure for uprated power is increased by a minimum amount necessary to assure that satisfactory reactor pressure control is maintained.

The operating pressure was chosen on the basis of steam'ine pressure drop characteristics and excess steam flow capability of the turbine observed during plant operation up to the current rated power level.

Satisfactory reactor pressure control requires an adequate flow margin between'the upr ated operating condition and the steam flow capability of the turbine control valves at their maximum stroke.

An operating dome pressure of 1032 psig is expected and is being assumed in the transient analysis.

The 1050 psig limit was chosen to maintain an adequate level of operating flexibilitywhile maintaining an adequate distance from the high pressure scram for trip avoidance.

This limit is the initial pressure value used in the overpressure protection safety analysis for power uprate, for which all licensing criteria have been met.

The 10 psig increase in the steam dome pressure limit was discussed in the staff's safety evaluation of November 30,

1993, and is acceptable.

26. Specification 4.5. I.b.3, Emergency Core Cooling Systems, has been revised to specify a test line pressure for the flow surveillance of the HPCI system of greater than or equal to 1140 psig at nominal reactor operating

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The staff's assessment of the,HPCI, system under power uprate conditions was discussed in Section 3.3.2 of th'e November"'30, 1993 safety evaluation.

As noted in.item 25 above, the. steam<,dome. pressure at the uprated power is expected to be 1032 psig.

Thpper"'pressilre limit is being set at 1050 psig.

The. licensee has proposed~that HPCI test acceptance pressure be set at 1140 psig, approxim'ately 100 psig above the expected steam dome, pressure.

The...staff, concludes that this test criteria will assure that the HPCI system will be able'o i'nject th'e required 5000 gpm at the higher reactor operating pressures, associated with power uprate.

The proposed HPCI pump 'test criteria is 'acceptable.

27.

In Bases Table B 3/4 4.6-1, the characteristics of the limiting plate material were revised per R.G. 1.99, Revision 2.

The change is in accordance with Generic Letter 92-01 and is acceptable.

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28. Specification 5.4.2 on Design, Features, Reactor Coolant System,

Volume, was revised to show that the nomi,nal.T,, is being changed'from 528 F to 532 'F.

This change is being made to reflect,.the higher average saturation temperature that results from a 35 psi in'crease in reactor design pressure.,

The staff's assessmeqt.

of the effect of 4 'F i'nc'rease in average primary coolant temperature on stresses and'fatigue 'bsage"factors was discussed in Sections 3.2.3, 3.2.4, and 3.2.'5oV the November 30, 1993, safety evaluation.

The. effects of power uprate have"been evaluated to ensure that the increase in system temperatures c'auses minor"increases in thermal loadings on pipe supports, equipment nozzles, and in-line components.

The results of the analyses scrog that 'at"uprated conditions, all ASHE components will satisfy design specification requirements'nd code limits when evaluated to the rules of Subsection NB-3600 of the ASHE Boiler and-Pressure Vessel Code Section III.

The effects of thermal expansion as a result of power uprate were found to be insignificant.

The slight increase in average coolant temperature is a consequence of the increase in reactor.,pressure.'he increase.,ih temperature. results in no significant increase in thermal. stresses and is acceptable.

'9.

In Table 5.7. 1-1, Component Cyclic or Transient 'L'imit's', the design cycle or transient limit far, the reactor was changed to ra'ise the upper limit for a heat cycle from 546 'F to 551 'F.

This'change is being made to reflect the higher average saturation temperature that results from a 30 psi increase in reactor design pressure.

The purpose of this specification is to limit the number of heatup and cooldown cycles.

The effects of power uprate have been evaluated to ensure that the reactor vessel components continue to comply with the existing structural requirements of the ASHE Boiler and Pressure Vessel Code.

The analyses were performed for the design, n'ormal, upset, emergency, and faulted conditions.

The increase in the temperature limitation is not significant with respect to the affect it has upon the RPV and associated components.

The staff's assessment of stresses and fatigue usage factor for the reactor vessel were discussed in Section 3.2.3 of the November 30, 1993 safety evaluation.

The 5 'F increase in the upper transient limit was determined to not be significant and is acceptable.

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~ 30. Administrative Control Section 6.9.3.2 describes and lists topical reports that are used to determine core operating limits.

Topical reports..15 through 19 are LOCA methodology reports and are being deleted.

These reports describe Siemens LOCA methodology.-

The GE SAFER/GESTR LOCA methodology is being used for this uprated cycle.

In addition, other minor methodology changes were made for power uprate transient analysis.

GE topical report NEDC-32071P, PPEL topical report NE-092-001, and the NRC Safety Evaluation Report on PP&L power uprate licensing topical. are added as Topical Reports No.

15, 16, 1'7, respectively..

<<The referenced reports.

and safety evaluations have been previously approved by the.

NRC staff-and are an acceptable basis for the Core Operating Limit's'eport~

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The li'censee's application was submitted on November 24;. 1993.

The'-

Commission's safety evaluation was issued a week later on November 30,- 1993-.

One of the TS changes proposed by th'e licensee was a revision.'sto the list'f topical reports on TS Page 6-20b approved by the NRC.'and'which are the"basis for the "Core Operating Limits Report."

The licensee's proposed wording for -.

Reference 17 was:

"NRC SER on PP&L Power Uprate Ltr (1'ater).:"

The NRC's safety evaluation was issued on November 30, 1993'.

-The:- staff substituted "November 30, 1993," in place of "(later)." --This change updates the TS submittal and is acceptable."

f f In the original license, NPF-22, issued on March 23, 1984, there was a

typographical error in the first line of paragraph 2.C.(1) in that the "L" was omitted from PP&L.

This error was corrected by this amendment with the licensee's concurrence and-did not 'change the original no 'significant hazards consideration determination.

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3.0 STATE CONSULTATION

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/ rs In accordance with the Commission's regulations, the Pennsylvania State official was notified of the proposed issuance of the amendment.

The State official had no comments.

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4. 0 ENVIRONMENTAL CONSIDERATION Pursuant to 10 CFR 51.21, 51.32, and 51.35, an environmental assessment and finding of no significant'mpact have been prepared and published-in the Federal

~Re ster on Narcb 18, 1994 (59 FR 12990).

Accordingly, based upon the environmental assessment, the staff has determined that the issuance of this amendment will not have a significant effect on the quality of the human environment.

5.0 CONCLUSION

The Commission has concluded, based on the considerations discussed

above, that:

(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed

manner, (2) such

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Attachment:

Letter dated November 30, 1993, to R.

G.

Byr am, PPLL, from T. Murley,

NRC, transmitting safety evaluation.

Principal Contributors:

W. Lefave M. Razzaque H. Garg R.

Goel W.

Long C.

Wu K. Eccleston R. Stransky R. Clark Date:

April ll, 1994

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<<**+>>

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20566 0001 November 30.

1993 Docket Nos. 50-387 and 50-388 Rr. Robert G.

Byram Senior Vice President-Nucleai Pennsylvania Power and Light Company 2 North Ninth Street Allentown, Pennsylvania 18101

SUBJECT:

LICENSING TOPICAL REPORT FOR POMER UPRATE MITH INCREASED CORE

FLOM, REVISION 0, SUS(UEHANNA STEAN ELECTRIC SJAT/ON, UNITS 1 AND 2 (PLA-3788)

(TAC NOS.

H83426 AND.%3427)

Your letter of June 15, 1992, submitted "Licensing.Topical Report NE-092-001, Revision 0, for Power Upr ate Mith Increased Core Flow,,", for Susquehanna SteawI Electric Station (SSES),

Units 1 and 2; The report was submitted to support future proposed

'amendments to the Units 1 and 2 licenses to permit a 4.5-percent increase in reactor thermal power and an 8-percent,increase in core flow for each unit.

Your initial submittal was revised and supplemented by letters of July 24, September 17, and December 18, 1992 and January 8,

January 25, April 2, August 5, August 12, ~d September 29, 1993.

As discussed in the enclosed safety evaluation,

.we have. concluded that the revised (Revision 2) licensing topical report adequately,.supports your proposed power uprate.

Me have also concluded that

SSES, Units 1 and 2, can operate safely with the proposed 8-percent increase in core. flow, the proposed 4.5-percent increase in reactor thermal power, the..corresponding 5-percent increase in main turbine inlet steam fl,ow, and the corresponding increases in flows; temperatures, pressures, and capacities required in supporting systems and components at these uprated conditions.

However, authorization for any increase in reactor thermal power will be based on our review of, the..technical specifications you will submit when you submit the amendment application.

Sincerely,

Enclosure:

Safety Evaluation cc w/enclosure:

See next page homas E

urley, irector Office of Nuclear Reactor Regulation

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Mr.-Robert G.

Byram Pennsylvania Power 5 Light Company Susquehanna Steam Electric Station, Units 1 5 2 CC:

Jay Silberg, Esq.

Shaw, Pittman, Potts 8 Trowbridge 2300 N Street N.W.

Washington, D.C.

20037 Bryan A. Snapp, Esq.

Assistant Corporate Counsel Pennsylvania Power i Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 Mr. J.

M. Kenny Licensing Group Supervisor Pennsylvania Power 5 Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 Mr. Scott Barber

'enior Resident Inspector U. S. Nuclear Regulatory Commission P.O.

Box 35 Berwick, Pennsylvania 18603-0035 Mr. William P. Dornsi fe, Director Bureau of Radiation Protection Pennsylvania Department of Environmental Resources P. 0.

Box 8469 Harrisburg, Pennsylvania 17105-8469 Mr. Jesse C. Tilton, III Allegheny Elec. Cooperative, Inc.

212 Locust Street P.O.

Box 1266 Harrisburg, Pennsylvania 17108-1266 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, Pennsylvania 19406 Mr. Harold G. Stanley Vice President-Nuclear Operations Susquehanna Steam Electric Station Pennsylvania Power and Light Company Box 467 Berwick, Pennsylvania 18603 Mr. Herbert D. Woodeshjck Special Office of the President Pennsylvania Power and Light Company Rural Route 1, Box 1797 Berwick, Pennsylvania 18603 George T. Jones Vice President-Nuclear Engineering Pennsylvania Power and Light Company 2 North Ninth Street Allentown, Pennsyl vani a 18101