Text

Amends 184 & 158 to Licenses NPF-14 & NPF-22,respectively, Incorporating long-term Power Stability Solution Instrumentation Into SSES Units 1 & 2 TS
	ML17146B161
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Site:	Susquehanna
Issue date:	07/30/1999
From:	Bajwa S; NRC (Affiliation Not Assigned)
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UNITED STAT =S NUCLEAR REGULATORS COMMISSION wAsHINGTON, D.c. 2055&0001 PPRL INC.

ALLEGHENYELECTRIC COOPERATIVE INC.

DOCKET NO. 50-387 SUSQUEHANNA STEAM ELECTRIC STATION UNIT 1 AMENDMENTTO FACILITYOPERATING LICENSE Amendment No. 184 License No. NPF-14 1.

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

A.

The application for the amendment filed by PP8L; Inc., dated June 19, 1998, and supplemented by letters dated November 23, 1998, and June 23, 1999, 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 facilitywilloperate in conformity with the application, the provisions of the Act, and the regulations of the Commission; C.

There is reasonable assurance:

(i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities willbe 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.

9908040%89 990730 PDR I5IDQCK 05000387 P

PDR 2.

Accordingly, 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-14 is hereby amended to read as follows:

(2) Technical S ecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.184 and the Environmental Protection Plan contained in Appendix 8, are hereby incorporated in the license. PP8L shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of its date of issuance and is to be implemented within 30 days following startup from the Unit 1, eleventh Refueling Inspection Outage, currently scheduled for spring 2000.

FOR THE NUCLEAR REGULATORY COMMISSION

Attachment:

Changes to the Technicai Specifications S. Singh Bajwa, Director Project Directorate l-1 Division of Licensing Project Management Office of Nuclear Reactor Regulation Date of Issuance:

Ouly 30, 1999

ATTACHMENTTO LICENSEE AMENDMENTNO. 184 FACILITYOPERATING LICENSE NO. NPF-14 DOCKET NO. 50-387 Replace the following pages of the Appendix A Technical Specifications with enclosed pages.

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

REMOVE III Iv V

Vl 3.4-1 3.4-2 3.4-3 3.4-4 3.4-5 8 3.3-44 8 3.3-45 8 3.3-46 8 3.3-47 8 3.3-48 8 3.3-49 8 3.3-50 8 3.3-51 8 3.4-1 8 3.4-2 8 3.4-3 8 3.4-4 8 3.4-5 8 3.4-6 8 3.4-7 8 3.4-8 8 3.4-9 INSERT I

II III IV 3.3-15a 3.3-15b 3.4-1 3.4-2 3.4-3 3.4-4 3.4-5 8 3.3-43a 8 3.3-43b 8 3.3-43c 8 3.3-43d 8 3.3-43e 8 3.3-43f 8 3.3-43g 8 3.3-43h 8 3.3-43i 8 3.4-1 8 3.4-2 8 3.4-3 8 3.4-4 8 3.4-5 8 3.4-6 8 3.4-7 8 3.4-8 8 3.4-9

TABLE OF CONTENTS (TECHNICAL SPECIFICATIONS) 1 0

1.1 1.2 1.3 1.4 I

2.0 2.1 2.2 3.0 3.0 3.1 3.1.1 3.1.2 3.1.3 3.1.4 3.1.5 3.1.6 3.1.7 3.1.8 3.2 3.2.1 3.2.2 3.2.3 3.2.4 3.3 3.3.1.1 3.3.1.2 3.3.1.3 3.3.2.1 3.3.2.2 3.3.3.1 33,32 3.3.4.1 3.3.4.2 3.3.5.1 3.3.5.2 3.3.6.1 3.3.6.2 3.3.7.1 USE AND APPLICATION Definitions Logica 1 Connectors Completion Times Frequency SAFETY LIMITS (SLs)

SLs SL Violations LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY SURVEILLANCE REQUIREMENT (SR) APPLICABILITY REACTIVITY CONTROL SYSTEMS Shutdown Margin (SDM)

Reactivity Anomalies Control Rod OPERABILITY Control Rod Scram Times Control Rod Scram Accumulators Rod Pattern Control Standby Liquid Control (SLC) System Scram Discharge Volume (SDV) Vent and Drain Valves POWER DISTRIBUTION LIMITS Average Planar Linear Heat Generation Rate (APLHGR)

Minimum Critical Power Ratio (MCPR)

Linear Heat Generation Rate (LHGR)

Average Power Range Honitor (APRM) Gain and Setpoints INSTRUMENTATION Reactor Protection System (RPS) Instrumentation Source Range Monitor (SRH) Instrumentation Oscillation Power Range Monitoring (OPRH)

Instrumentation Control Rod Block Instrumentation Feedwater

- Main Turbine High Water Level Trip Instrumentation Post Accident Monitoring (PAH) Instrumentation Remote Shutdown System End of Cycle Recirculation Pump Trip (EOC-RPT)

Instrumentation Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation Emergency Core Cooling System (ECCS) Instrumentation Reactor Core Isolation Cooling (RCIC) System Instrumentation Primary Containment Isolation Instrumentation Secondary Containment Isolation Instrumentation Control Room Emergency Outside Air Supply (CREOAS)

System Instrumentation 1.1-1 1.1-1 1.2-1 1.3-1 1.4-1 2.0-1 2.0-1 2.0-1 3.0-1 3.0-4 3.1-1 3.1-1 3.1-5 3.1-7 3.1-12 3.1-15 3.1-18 3.1-20 3.1-25 3.2-1 3.2-1 3.2-3 3.2-5 3.2-7 3.3-1 3.3-1 3.3-10 3.3-15a 3.3-16 3.3-21 3.3-23 3.3-26 3.3-29 3.3-33 3.3-36 3.3-48 3.3-52 3.3-63 3.3-67 SUSQUEHANNA - UNIT 1 Amendment No. <8)

TABLE OF CONTENTS (TECHNICAL SPECIFICATIONS) 3.3 3.3.8.1 3.3.8.2 3.4 3.4.1 3.4.2 3.4.3 3.4.4 3.4.5 3.4.6 3.4.7 3.4.8 3.4.9 3.4.10 3.4.11 3.5 3.5.1 3.5.2 3.5.3 3.6 3.6.1.1 3.6.1.2 3.6.1.3 3.6.1.4 3.6.1.5 3.6.1.6 3.6.2.1 3.6.2.2 3.6.2.3 3.6.2.4 3.6.3.1 3.6.3.2 3.6.3.3 3.6.4.1 3.6.4.2 3.6.4.3 INSTRUMENTATION (continued)

Loss of Power (LOP) Instrumentation Reactor Protection System (RPS) Electric Power Monitoring REACTOR COOLANT SYSTEM (RCS)

Recirculation Loops Operating Jet Pumps Safety/Relief 'Valves (S/RVs)

RCS Operational LEAKAGE RCS Pressure Isolation Valve (PIV) Leakage RCS Leakage Detection Instrumentation RCS Specific Activity Residual Heat Removal (RHR) Shutdown Cooling System-Hot Shutdown Residual Heat Removal (RHR) Shutdown Cooling System-Cold Shutdown RCS Pressure and Temperature (P/T) Limits Reactor Steam Dome Pressure EMERGENCY CORE COOLING SYSTEMS (ECCS)

AND REACTOR CORE ISOLATION COOLING (RCIC)

SYSTEM ECCS-Oper ating ECCS-Shutdown RCIC System CONTAINMENT SYSTEMS Primary Containment Primary Containment Air Lock Primary Containment Isolation Valves (PCIVs)

Containment Pressure Drywell Air Temperature Suppression-Chamber-to-Drywell Vacuum Breakers Suppression Pool Average Temperature Suppression Pool Water Level Residual Heat Removal (RHR) Suppression Pool Cooling Residual Heat Removal (RHR) Suppression Pool Spray Primary Containment Hydrogen Recombiners Drywell Air Flow System Primary Containment Oxygen Concentration Secondary Containment Secondary Containment Isolation Valves (SCIVs)

Standby Gas Treatment (SGT) System 3.3-72 3.3-75 3.4-1 3.4-1 3.4-6 3.4-8 3.4-10 3.4-12 3.4-14 3.4-17 3.4-19 3.4-22 3.4-24 3.4-31 3.5-1 3.5-1 3.5-8 3.5-12 3.6-1 3.6-1 3.6-4 3.6-8 3.6-17 3.6-18 3.6-19 3.6-22 3.6-25 3.6-26 3.6-28 3.6-30 3.6-32 3.6-34 3.6-35 3.6-38 3.6-42 SUSQUEHANNA - UNIT 1 AmendmentNo.

<ss

Ie

TABLE OF CONTENTS (TECHNICAL SPECIFICATIONS)

I 3.7 3.7.1 3.7.2 3.7.3 3.7.4 3.7.5 3.7.6 3.7.7 3.8 3.8.1 3.8.2 3.8.3 3.8.4 3.8.5 3.8.6 3.8.7 3.8.8 3.9 3.9.1 3.9.2 3.9.3 3.9.4 3.9.5 3.9.6 3.9.7 3.9.8 3.10 3.10.1 3.10.2 3.10.3 3.10.4 3.10.5 3.10.6 3.10.7 3.10.8 4.0 4.1 4.2 4.3 PLANT SYSTEHS Residual Heat Removal Service Water (RHRSW) System and the Ultimate Heat Sink (UHS)

Emergency Service Water (ESW) System Control Room Emergency Outside Air Supply (CREOAS) System Control Room Floor Cooling System Main Condenser Offgas Hain Turbine Bypass System Spent Fuel Storage Pool Water Level ELECTRICAL POWER SYSTEMS AC Sources-Operating AC Sources-Shutdown Diesel Fuel Oil, Lube Oil, and Starting Air DC Sources-Operating DC Sources-Shutdown '.............

Battery Cell Parameters Distribution Systems-Operating Distribution Systems-Shutdown REFUELING OPERATIONS

. Refueling Equipment Interlocks.........

Ref'uel Position One-Rod-Out Inter lock Control Rod Position..............

Control Rod Position Indication Control Rod OPERABILITY-Refueling...,...

Reactor Pressure Vessel (RPV) Water Level Residual Heat Removal (RHR)-High Water Level Residual Heat Removal (RHR)-Low Mater Level SPECIAL OPERATIONS Inservice Leak and Hydrostatic Testing Operation Reactor Node Switch Interlock Testing Single Control Rod Withdrawal -Hot Shutdown Single Control Rod Withdrawal -Cold Shutdown Single Control Rod Drive (CRD) Removal -Refueling Multiple Control Rod Withdrawal Refueling Control Rod Testing-Operating SHUTDOWN MARGIN (SDH) Test-Refueling DESIGN FEATURES Site Location Reactor Core Fuel Storage 3.7-1 3.7-1 3.7-4 3.7-6 3.7-10 3.7-13 3.7-15 3.7-17 3.8-1 3.8-1 3.8-17 3.8-20 3.8-23 3.8-29 3.8-32 3.8-37 3.8-41 3.9-1 3.9-1 3.9-3 3.9-5 3.9-6 3.9-8 3.9-9 3.9-10 3.9-13 3.10-1 3.10-1 3.10-4 3.10-6 3.10-9 3.10-13 3.10-16 3.10-18 3.10-20 4.0-1 4.0-1 4.0-1 4.0-1 SUSQUEHANNA - UNIT 1 Amendment No.i84[

C 4

I ~

TABLE OF CONTENTS (TECHNICAL SPECIFICATIONS) 5.0 5.1 5.2 5.3 5.4 5.5 5.6 5.7 ADHINISTRATIVE CONTROLS Responsibility......

Organization.......

Unit Staff Qualifications Procedures

~

Programs and Hanuals Reporting Requirements High Radiation Area 5.0-1 5.

0-1'.

0-2 5.0-5 5.0-6 5.0-7 5.0-19 5.0-27 SUSQUEHANNA - UNIT 1 1V Amendment No.

184

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OPRM Instrumentation 3.3.1.3 3.3 INSTRUMENTATION 3.3. 1.3 Oscillation Power Range Monitor (OPRH) Instrumentation LCO 3.3. 1.3 Four channels of the OPRH instrumentation shall be OPERABLE.

Each OPRM channel Period Based Algorithm (Sp) Allowable Value shall be less than or equal to 1.09 at a confirmation count permissive (Np) of 10.

APPLICABILITY:

Thermal Power ) 25K RTP.

ACTIONS:

-NOTE-Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A.

One or more required channels inoperable.

A.1 OR Place channel in trip.

30 days A.2 OR A.3 Place associated RPS trip system in trip.

Initiate alternate method to detect and suppress thermal hydr aulic instability oscillations.

30 days 30 days B.

OPRM trip capability not maintained.

B.1 AND Initiate alternate method to detect and suppress thermal hydraullic instability oscillations.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months B.2 C.

Required Action and C.1 associ,ated Completion Time not met.

Restore OPRM trip capabi1 ity Reduce THERMAL POWER

( 25K RTP.

120 days 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months SUSQUEHANNA - UNIT 1 TS I 3.3-15a Amendment No.18)

OPRM Instrumentation 3.3.1.3

'SURVEILLANCE REQUIREMENTS

NOTE-When a channel is placed in an inoperable status solely for performance of requi red Surveillances.

entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the OPRM System maintains trip capability.

SURVEILLANCE FREQUENCY SR 3.3.1.3.1 Perform CHANNEL FUNCTIONAL TEST.

184 days SR 3.3. 1.3.2 Calibrate the local power range monitors.

1000 MWD/MT average core exposure SR 3.3.1.3.3

-NOTE Neutron detectors are excluded.

Perform CHANNEL CALIBRATION.

24 months SR 3.3. 1.3.4 Perform LOGIC SYSTEM FUNCTIONAL TEST 24 months SR 3.3.1.3.5 Verify OPRM is not bypassed when THERMAL POWER is ) 30K RTP and core flow ~ 60 MLb/Hr.

24 months SR 3.3.1.3.6

-NOTE---------- -------

Neutron detectors are excluded.

Verify the RPS

RESPONSE

TIME is within limits.

24 months on a

STAGGERED TEST BASIS

)

SUSQUEHANNA - UNIT 1 TS / 3.3-15b Amendment No. i84

0 t~

Recirc ation Loops Operating 3.4 '

'3.4 REACTOR COOLANT SYSTEM (RCS) 3.4. 1 Recirculation Loops Operating LCO 3.4.1 Two recirculation loops with matched flows shall be in operation.

OR One recirculation loop may be in operation provided the following limits are applied when the associated LCO is applicable:

a.

LCO 3.2. 1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)." single loop operation limits specified in the COLR; b.

LCO 3.2.2, "MINIMUMCRITICAL POWER RATIO (MCPR)," single loop operation limits specified in the COLR; c.

LCO 3.2.3, "LINEAR HEAT GENERATION RATE (LHGR)," single loop operation limits specified in the COLR, and d.

LCO 3.3. 1. 1, "Reactor Protection System (RPS)

Instrumentation,"

Function 2.b (Average Power Range Monitors Flow Biased Simulated Thermal Power-High),

Allowable Value of Table 3.3.1. 1-1 is reset for single loop operation.

e.

Recirculation pump speed is ~ 80X.

Note----

Required limit and setpoint resets for single recirculation loop operation may be delayed for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after transition from two recirculation loop operation to single recirculation loop operation.

APPLICABILITY:

MODES 1 and 2.

SUSQUEHANNA - UNIT 1 TS' 3.4-1 Amendment No.

4 ZQ, 184

Recirc ation Loops Operating 3.4.1

'CTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

No recirculation loops operating while in MODE 1 A. 1 Place reactor mode switch in the shutdown position.

Immedi ately B.

Recirculation loop flow mismatch not within limits.

B.1 Declare the recirculation loop with lower flow to be "not in operation."

2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months

)

C.

No recirculation loops in operation while in MODE 2 OR Single Recirculation Loop required limits and setpoints not

'stablished within required time.

C.l Be in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SUSQUEHANNA - UNIT 1 TS / 3.4-2 Amendment No.

$78, 184

SUSQUEHANNA - UNIT 1 TS / 3.4-3 AmendmentND.

478, 184

Recir ation Loops Operating 3.4.1 SURVEILLANCE REQUIRBIENTS SURVEILLANCE FREQUENCY SR 3.4.1.1

NOTE Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after both recirculation loops are in operation.

Verify recirculation loop jet pump flow mismatch with both recirculation loops in operation is:

a.

c 10 million ibm/hr when operating at

< 75 million ibm/hr total core flow; and b.

s 5 million ibm/hr when operating at

~ 75 million ibm/hr total core flow.

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months SR 3.4.1.2

NOTE-Only requi red to be met during single loop operations.

Verify recirculation pump speed is within the limit specified in the LCO.

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months SUSQUEHANNA - UNIT I TS / 3.4-4 Amendment No.

478, f84

Recirc tion Loops Operating 3.4.1 (Figure 3.4.1-1) l IHIIIS PAGE IIINTEMTIIONALLYLEIFl IHlLAIMK SUSQUEHANNA - UNIT 1 TS / 3.4-5 Atttendgent Np. $78, 184

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OPRH Instrumentation B 3.3.1.3

,8 3.3 INSTRUMENTATION 8 3.3.1.3 Oscillation Power Range Monitor (OPRH)

BASES BACKGROUND General Design Criterion 10 (GDC 10) requires the reactor core and associated coolant, control, and protection system to be designed with appropriate margin to assure that acceptable fuel design limits are not exceeded during any condition of normal operation including the effects of anticipated'operational occurrences.

Additionally, GDC 12 requires the reactor core and associated coolant control and protection systems to be designed to assure that power oscillations which can result in conditions exceeding acceptable fuel design limits are eithei not possible or can be reliably and readily detected and suppressed.

The OPRN System provides compliance with GDC 10 and GDC 12, thereby roviding protection from exceeding the fuel MCPR safety imit.

References 1, 2.

and 3 describe three separate algorithms for detecting stability related oscillations:

the period based detection algorithm, the amplitude based algorithm, and the growth rate algorithm.

The OPRN System hardware implements these algorithms in microprocessor based modules.

These modules execute the algorithms based on LPRH inputs and generate alarms and trips based on these calculations.

These trips result in tripping the Reactor Protection System (RPS) when the appropriate RPS trip logic is satisfied, as described in the Bases for LCO 3.3. 1. 1, "RPS Instrumentation."

Only the period based detection algorithm i s used in the safety analysis (Ref.

1, 2.

6 IC 7).

The remaining algorithms provide defense-in-depth and additional protection against unanticipated oscillations.

The period based detection algorithm detects a stabi lity-related oscillation based on the occurrence of a fixed number of consecutive LPRH signal period confi rmations followed by the LPRH signal amplitude exceeding a specified

'setpoint.

Upon detection of a stability related oscillation, a trip is generated for that OPRN channel.

The OPRH System consists of four OPRH trip channels, each channel consisting of two OPRH modules.

Each OPRN module receives input from LPRNs.

Each OPRH module also receives (continued)

SUSQUEHANNA - UNIT 1 TS /

B 3.3-43a Revision 0

Amendment No.

184

I

OPRH Instrumentation B 3.3.1.3

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'ASES BACKGROUND (continued) input from the Neutron Monitoring System (NHS) average power range monitor (APRH) power and flow signals to automatically enable the trip function of the OPRH modu'le.

Each OPRH module is continuously tested by a self-test function.

On detection of any OPRH module fai lure, either a

trouble alarm or INOP alarm is activated.

The OPRH module provides an INOP alarm when the self-test feature indicates that the OPRM module may not be capable of meeting its functional requi rements.

APPLICABLE SAFETY ANALYS It has been shown that BWR cores may exhibit thermal-hydraulic reactor instabilities in high power and low flow portions of the core power to flow operating domain.

GDC 10 requires the reactor core and associated coolant control and protection systems to be designed with appropriate margin to assure that acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.

GDC 12 requires assurance that power osci llations which can result in conditions exceeding acceptable fuel design limits are either not possible or can be reliably and readily detected and suppressed.

The OPRH System provides compliance with GDC 10 and GDC 12 by detecting the onset of osci llations and suppressing them by initiating a reactor scram.

This assures that the HCPR safety limit will not be violated for anticipated osci llations.

The OPRH Instrumentation satisfies Criterion 3 of the NRC Policy Statement.

LCO Four channels of the OPRH System are required to be OPERABLE to ensure that stability related osci llations are detected and suppressed prior to exceeding the MCPR safety limit.

Only one of the two OPRH modules'eriod based detection algorithm is required for OPRH channel OPERABILITY.

The minimum number of LPRHs required OPERABLE to maintain an OPRH channel OPERABLE is consistent with the minimum number of LPRHs requi red to maintain the APRM system OPERABLE per LCO 3.3.1.1.

(continued)

SUSQUEHANNA - UNIT 1 TS I B 3.3-43b Revision 0

Amendment No.

184

OPRM Instrumentation B 3.3.1.3

~ BASES LCO (continued)

The Allowable Value for the OPRM Period Based Algorithm setpoint (Sp) is derived from the Analytic Limit corrected for instrument and calibration errors (Ref.

9 8 10).

APPLICABILITY The OPRM instrumentation is required to be OPERABLE in order to detect and suppress neutrnn flux osci llations in the event of thermal-hydraulic instability.

As described in References

1. 2, and 3, the power/core flow region protected against anticipated oscillations is defined by THERMAL POWER

) 30K RTP and core flow s 60 Mlb/Hr.

The OPRM trip is required to be enabled in this region, and the OPRM must be capable of enabling the trip function as a result of anticipated transients that place the core in that power/flow condition.

Therefore, the OPRM js required to be OPERABLE with THERMAL POWER a 25K RTP and at all core flows while above that THERMAL POWER.

It is not necessary for the OPRM to be operable with THERMAL POWER < 25K RTP because transients from below this THERMAL POWER are not anticipated to result in power that exceeds 30K RTP.

ACTIONS A Note has been provided to modify the ACTIONS related to the OPRM instrumentation channels.

Section 1.3, Completion Times, specifies that once a Condition has been entered subsequent divisions, subsystems, components, or variables expressed in the Condition discovered to be inoperable or not within limits will not result in separate entry into the Condition.

Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional f'ailure with Completion Times based on initial entry into the Condition.

'However, the Required Actions for inoperable OPRM instrumentation channels provide appropriate compensatory measures for separate inoperable channels.

As

such, a Note has been provided that al'Iows separate Condition entry for each inoperable OPRM instrumentation channel.

A1 A2 and A3 Because of the reliability and on-line self-testing of the OPRM instrumentation and the redundancy of the RPS design, an allowable out of service time of 30 days has been shown SUSQUEHANNA - UNIT 1 TS / 8 3.3-43c Revision 0

Amendment No.

184

OPRH Instrumentation B 3.3.1.3

. BASES ACTIONS Al A2 and A3 (continued) to be acceptable (Ref.

7) to permit restoration of any inoperable channel to OPERABLE status.

However, this out of'ervice time is only acceptable provided the OPRH instrumentation still maintains OPRM trip capability (refer to Required Actions B. 1 and B.2).

The remaining OPERABLE OPRH channels continue to provide trip capability (see Condition B) and provide operatoi information relative to stability activity.

The remaining OPRH modules have high reliability.

With this high reliability, there is a low probability of a subsequent channel fai lure within the allowable out of service time.

In addition, the OPRH modulqs continue to perform on-line self-testing and alert the operator if any further system degradation occurs.

If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the OPRM channel or associated RPS trip system must be placed in the tripped condition per Required Actions A. 1 and A.2.

Placing the inoperable OPRH channel in trip (or the associated RPS trip system in trip) would conservatively compensate for the inoperability, provide the capability to accommodate a single failure and allow operation to continue.

Alternately, if it is not desi red to place the OPRH channel (or RPS trip system) in trip (e.g.,

as in the case where placing the inoperable channel in trip would result in a full scram),

the alternate method of detecting and suppression thermal hydraulic instability osci llations is required (Required Action A.3).

This,alternate method is described in Reference 5.

It consists of increased operator awareness and monitoring for neutron flux osci llations when operating in the region. where osci llations are possible.

If indications of oscillation, as described in Reference 5 are observed by the operator, the operator will take the actions described by procedures which include initiating a manual scram of the reactor.

B.l and B.2 Required action B.l is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped OPRH channels within the same RPS trip system result in not maintaining OPRH trip capability.

OPRH trip capability is (continued)

SUSQUEHANNA - UNIT 1 TS I B 3.3-43d Revision 0

Amendment No.

184

OPRN Instrumentation B 3.3.1.3

'ASES ACTIONS B. 1 and B.2 (continued) considered to be maintained when suffici nt OPRM channels are OPERABLE or in trip (or the associated RPS trip system is in trip), such that a valid OPRM signal'ill generate a

trip signal in both RPS trip systems (this would requi re both RPS trip systems to have at least one OPRM channel OPERABLE or the associated RPS trip system in trip).

Because of the low probability of the occurrence of an instability, 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is an acceptable time to initiate the alternate method of detecting and suppression thermal hydraulic instability osci llations described in Action A.3 above.

The alternate method of detecting and suppressing thermal hydraulic instabi lity osci llations would adequately address detection and mitigation in the event of instability osci llations.

-Based on industry operating experience with actual instability osci llations, the operator would be able to recognize instabilities during this time and take action to suppress them through a manual scram.

In addition, the OPRM System may still be available to provide alarms to the oper ator if the onset of'sci llations were to occur Since plant operation is minimized in areas where osci llations may occur, operation for 120 days without OPRM trip capabi lity is considered acceptable with implementation of the alternate method of detecting and suppression thermal hydraulic instability osci llations.

C.1 With any Required Action and associated Completion Time not

met, THERMAL POWER must be reduced to < 2M RTP within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Reducing THERMAL POWER to < 25K RTP places the plant in a condition where instabilities are not l.ikely to occur.

The 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is reasonable, based on operating experience, to reduce THERMAL POWER < 25K RTP from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE REQUIREMENTS SR 3.3.1.3.1 CHANNEL FUNCTIONAL TEST is performed to ensure that the entire channel will perform the intended function.

A Frequency of 184 days provides an acceptable level of system (continued)

SUSQUEHANNA - UNIT 1 TS /

B 3.3-43e Revision 0

Amendment No.

184

OPRM Instrumentation B 3.3.1.3

'ASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.3.1.3.1 average availability over the Frequency and is based on the reliability of the channel.

(Ref.

7)

SR 3.3.1.3.2 LPRH gain settings are determined from the local flux profiles measured by the Traversing Incore Probe (TIP)

System.

This establishes the relative local flux profile for appropriate representative input to the. OPRH System.

The 1000 HWD/HT Frequency is based on operating experience with LPRH sensitivity changes.

SR 3.3.1.3.3 A CHANNEL CALIBRATION verifies that the channel responds to the measured parameter within the necessary range and accuracy.

CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations.

Calibration of the channel provides a check of the internal reference voltage and the internal processor clock frequency.

It also compares the desired trip setpoints with those in processor memory.

Since the OPRH is a digital system, the internal reference voltage and processor clock frequency are, in turn, used to automatically calibrate the internal analog to digital convertors.

As noted, neutron detectors are excluded from CHANNEL CALIBRATION because they are passive

devices, with minimal drift, and because of the difficulty of simulating a

meaningful signal.

Changes in neutron detector sensitivity are compensated for by performing the 1000 HWD/HT LPRH calibration using the TIPs (SR 3.3. 1.3.2).

The Frequency of 24 months is based upon the assumption of the magnitude of equipment drift provided by the equipment supplier.

(Ref. 7)

SR 3.3.1.3.4 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific channel.

The functional testing of control rods, in LCO (continued)

I SUSQUEHANNA - UNIT 1 TS /

B 3.3-43f Revision 0

Amendment No.

184

~ )t,

OPRM Instrumentation B 3.3.1.3

'BASES SURVEILLANCE REQUIREMENTS SR 3.3. 1.3.4 (continued)

3. 1.3, "Control Rod OPERABILITY", and scram discharge volume (SDV) vent and drain valves, in LCO 3. 1.8.

"Scram Discharge Volume (SDV) Vent and Drain Valves". overlaps this Surveillance to provide complete testing of the assumed safety function.

The OPRM self-test function may be utilized to perform this testing for those components that it is designed to monitor.

The 24 month Frequency is based on engineering judgement, reliability of the components and operating experience.

SR 3.3.1.3.5 This SR ensure that trips initiated from the OPRM System will not be inadvertently bypassed when THERMAL POWER is

> 30K RTP and core flow is ~ 60 MLb/Hr. This normally involves calibration of the bypass channels.

Adequate margins for the instrument setpoint methodology are incorporated into the actual setpoints (Reference 7).

If any bypass channel setpoint is nonconservative (i.e., the OPRM module is bypassed at ~ 30K RTP and core 'flow is c 60 MLB/Hr), then the affected OPRM module is considered inoperable.

Alternatively, the bypassed channel can be manually placed in the conservative position (Manual Enable).

If placed in the manual enable condition, this SR is met and the module is considered OPERABLE.

The 24 month Frequency is based on engineering judgment and reliability of'he components.

SR 3.3.1.3.6 This SR ensure that the individual channel response times are less than or equal to the maximum values assumed in the safety analysis (Ref.

6)

The OPRM selt-test function may be utilized to perform this testing for those components it is designed to monitor.

The LPRM amplifier cards inputting to the OPRM are excluded from the OPRM response time testing.

The RPS

RESPONSE

TIME acceptance criteria are included in Reference 8.

(continued)

SUSQUEHANNA - UNIT 1 TS /

B 3.3-43g Revision 0

Amendment No. i84

'I l

OPRM Instrumentation B 3.3.1.3

~

'ASES SURVEILLANCE REQUIREMENTS SR 3.3.1.3.6 (continued)

As noted, neutron detectors are excluded from RPS

RESPONSE

TIME testing because the principles of detector operation virtually ensure an instantaneous response time.

RPS

RESPONSE

TIME tests are conducted on an 24 month STAGGERED TEST BASIS.

This Frequency is based upon operating experience, which shows that random fai lures of instrumentation components causing serious response time degradation, but not channel failure, are infrequency occurrences.

REFERENCES 1.

NED0-31960-A, "BWR Owners Group Long-Term Stability Solutions Licensing Methodology",

November 1995.

2.

NEDO 31960-A.

Supplement 1,

"BWR Owners Group Long-Term Stability Solutions Licensing Methodology",

November 1995.

3.

NRC Letter, A. Thadani to L. A. England, "Acceptance for Referencing of Topical Reports NED0-31960, Supplement 1,

'BWR Owners Group Long-Term Stability Solutions Licensing Methodology', July 12.

1994.

4.

Generic Letter 94-02.

"Long-Term Solutions and Upgrade of Interim Operating Recommendations for Thermal-Hydraulic Instabilities in Boiling Water Reactors",

July 11, 1994.

5.

BWROG Letter BWROG-9479, "Guidelines for Stabi lity Interim Corrective Action", June 6.

1994.

6.

NED0-32465-A, "BWR Owners Group Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology and Reload Applications". August 1996.

7.

CENPD-400-P-A, Rev.

01, "Generic Topical Report for the ABB Option III Oscillation Power Range Honitor (OPRH)",

Hay 1995.

8.

FSAR, Table 7.3-28.

I SUSQUEHANNA - UNIT 1 TS /

B 3.3-43h (continued)

Revision 0

Amendment No.

184

OPRM Instrumentation 8 3.3.1.3

~

'BASES REFERENCES (continued) 9.

NFE-1-11-003.

"Unit 1 Cycle 11 Stability Option III Analyses" 10.

EC-078-1010, "Oscillation Power Range Monitor (OPRM)

Period Based Algorithm (Sp) Technical Specification Limit Value".

(continued)

SUSQUEHANNA - UNIT 1 TS,/

B 3.3-43i Revision 0

Amendment No.

184

Recirculation Lcops Operating B 3.4.1 3.4 REACTOR COOLANT SYSTEH (RCS)

B 3.4. 1 Recirculation Loops Operating BASES BACKGROUND The Reactor Coolant Recirculation System is designed to rovide a forced coolant flow through the core to remove eat from the fuel.

The forced coolant flow removes more heat from the fuel than would be possible with just natural circulation.

The forced flow, therefore, allows operation at significantly higher power than would otherwise be possible.

The recirculation system also controls reactivity over a wide span of reactor power by varying the recirculation flow rate to control the void content of the moderator.

The Reactor Coolant Recirculation System consists of two recirculation pump loops external to the reactor vessel.

These loops provide the piping path for the driving flow of water to the reactor vessel jet pumps.

Each external loop contains one variable speed motor driven recirculation

pump, a motor generator (HG) set to control pump speed and associated piping, jet pumps,

valves, and instrumentation.

The recirculation pump, piping, and valves are part of the reactor coolant pressure boundary and are located inside the drywell structure.

The jet pumps are rea0tor vessel internals.

The recirculated coolant consists of satut ated water from the steam separators and dryers that has been subcooled by incoming feedwater.

This water passes down the annulus between the reactor vessel wall and the core shroud.

A portion of the coolant flows from the vessel, through the two external recirculation loops, and becomes the driving flow for the jet pumps.

Each of the two external recirculation loops discharges high pressure flow into an external manifold, from which individual recirculation inlet lines are routed to the jet pump risers within the reactor vessel.

The remaining portion of the coolant mixture in the annulus becomes the suction flow for the jet pumps.

This flow enters the jet pump at suction inlets and is accelerated by the driving flow.

The drive flow and suction flow are mixed in the jet pump throat section.

The total flow then passes through the jet pump diffuser section into the area below the core (lower plenum), gaining sufficient head in the process to drive the required flow upward through the core.

The subcooled water enters the bottom of the fuel channels and contacts the fuel cladding, where heat (continued)

SUSQUEHANNA - UNIT 1 TS /

B 3.4-1 Revision 1

Amendment No.

184

Recirculation Loops Operating B 3.4.1 BASES BACKGROUND (continued) is transferred to the coolant.

As it rises.

the coolant begins to boi 1, creating steam voids within the fuel channel that continue until the coolant exits the core.

Because of reduced moderation, the steam voiding introduces negative reactivity that must be compensated for to maintain or to increase reactor power.

The recirculation flow control allows operators to increase recirculation flow and sweep some of the voids from the fuel channel, overcoming the negative reactivity void effect.

Thus, the reason for having variable recirculation flow is to compensate for reactivity effects of boi ling over a wide range of power generation without having to move control rods and disturb desirable flux patterns.

Each recirculation loop is manually started from the control room.

The HG set provides regulation of individual recirculation loop drive flows.

The flow in each loop is manually controlled.

APPLICABLE SAFETY ANALYSES The operation of the Reactor Coolant Recirculation System is an initial condition assumed in the design basis loss of coolant accident (LOCA) (Ref. 1).

During a LOCA caused by a recirculation loop pipe break, the intact loop is assumed to provide coolant flow during the first few seconds of the accident.

The 'initial core flow decrease is rapid because the recirculation pump in the broken loop ceases to pump reactor coolant to the vessel almost immediately.

The pump in the intact loop coasts down relatively slowly.

This pump coastdown governs the core flow response for the next several seconds unti 1 the jet pump suction is uncovered (Ref. 1).

The analyses assume that both loops are operating at the same flow prior to the accident.

However, the LOCA analysis was reviewed for the case with a flow mismatch between the two loops, with the pipe break assumed to be in the loop with the higher flow.

While the flow coastdown and core response are potentially more severe in this assumed case (since the intact loop starts at a lower flow rate and the core response is the same as if both loops were operating at a lower flow rate),

a small mismatch has been determined to be acceptable based on engineering judgement.

The recirculation system is also assumed to have sufficient flow coastdown characteristics to maintain fuel thermal margins during abnormal operational transients (Ref. 2),

which are analyzed in Chapter 15 of the FSAR.

(continued)

SUSQUEHANNA UNIT 1 TS /

B 3.4-2 Revision 1

Amendment No.

184

Recir ation Loops Operating B 3.4.1

~

'ASES APPLICABLE SAFETY ANALYSES (continued)

Plant specific LOCA analyses have been performed assuming only one operating recirculation loop.

These analyses have demonstrated that, in the event of a LOCA caused by a pipe break in the operating recirculation loop, the Emergency Core Cooling System response will provide adequate core cooling. provided that the LHGR limit for SPC 9x9-2 f'uel and GE lead use assemblies and the APLHGR limit for SPC ATRIUM'-10fuel is modified.

The transient analyses of Chapter 15 of the FSAR have also been performed for single recirculation loop operation-and demonstrate sufficient flow coastdown characteristics to maintain fuel thermal margins during the abnormal operational transients analyzed provided the MCPR requi rements are modified.

During-single recirculation loop operation, modification to the Reactor Protection System (RPS) average power range monitor (APRM) instrument setpoints is also required to account for the different relationships between recirculation drive flow and reactor core flow.

The APLHGR, LHGR, and MCPR setpoints for single loop operation are specified in the COLR.

The APRM flow biased simulat'ed THERMAL POWER setpoint is in LCO 3.3. 1. 1, "Reactor Protection System (RPS) Instrumentation."

In

addition, a restriction on recirculation pump speed is incorporated to address reactor vessel internals vibration concerns and assumptions in the event analysis.

Recirculation loops operating satisfies Criterion 2 of the NRC Policy Statement (Ref. 5).

LCO Two recirculation loops are required to be in operation with their flows matched within the limits specified in SR 3.4. 1. 1 to ensure that during a

LOCA caused by a break of the piping of one recirculation loop the assumptions of the LOCA analysis are satisfied.

With the limits specified in SR 3.4. 1. 1 not met, the recirculation loop with the lower flow must be considered not in operation.

With only one recirculation loop in operation, modifications to the requi red APLGHR limits (LCO 3.2. 1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE"),

LHGR limits (LCO 3.2.3, "LINEAR HEAT GENERATION RATE (LHGR)"), MCPR limits (LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)"), and APRM Flow Biased Simulated Thermal Power High setpoint (LCO 3.3. 1. 1) may be applied to allow continued operation consistent with the (continued)

SUSQUEHANNA - UNIT 1 TS /

B 3.4-3 Revision 1

Amendment No. f84

Recirculation Loops Operating B 3.4.1 BASES LCO (continued) safety analysis assumptions.

Furthermore, restrictions are placed on recirculation pump speed to ensure the initial assumption of the event analysis are maintained.

The LCO is modified by a Note that allows up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to establish the requi red limits and setpoints after a change from two recirculation loops operation to single recirculation loop operation.

If the limits and setpoints are not in compliance with the applicable requirements at the end of the this period, the ACTIONS requi red by the applicable specifications must be implemented.

This time is rovided to stabilize operation with one recirculation loop y:

limiting flow in the operating loop, limiting total THERMAL POWER; and, fully implementing and confirming the required limit and setpoint modifications.

APPLICABILITY In MODES 1 and 2, requirements for operation of the Reactor Coolant Recirculation System are necessary since there is considerable energy in the reactor core and the limiting design basis transients and accidents are assumed to occur.

In MODES 3. 4, and 5, the consequences of an accident are reduced and the coastdown characteristics of the recirculation loops are not important.

ACTIONS A.1 When operating with no recirculation loops operating in MODE 1, the potential for thermal-hydraulic osci llations is greatly increased.

Although this transient is protected for expected modes of oscillation by the OPRM system, when operable per LCO 3.3.1.3 (References 3 and 4), the prudent response to the natural circulation condition is to preclude potential thermal-hydraulic osci llations by immediately placing the mode switch in the shutdown position.

B-1 Recirculation loop flow must match within required limits when both recirculation loops are in operation.

If flow mismatch is not within requi red limits, matched flow must be

'estored within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

If matched flows are not restored, (continued)

SUSQUEHANNA - UNIT 1 TS /

B 3.4-4 Revision 1

Amendment No.

184

Recirculation Loops Operating B 3.4.1 BASES ACTIONS B-1 (continued) the recirculation loop with lower flow must be declared "not in operation."

Should a

LOCA occur with recirculation loop flow not matched, the core flow coastdown and resultant core response may not be bounded by the LOCA analyses.

Therefore.

only a limited time is allowed prior to imposing restrictions associated with single loop operation.

Operation with only one recirculation loop satisfies the requirements of the LCO and the initial conditions of the accident sequence.

The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time is based on the low probability of an accident occurring during this time period, providing a reasonable time to complete the Required Action, and considering that frequent core monitoring by operators allows abrupt changes in core flow conditions to be quickly detected.

These Required Actions do not requi re tripping the recirculation pump in the lowest flow loop when the mismatch between total jet pump flows of the two loops is greater than the requi red limits.

However, in cases where large flow mismatches

occur, low flow or reverse flow can occur in the low flow loop jet pumps, causing vibration of the jet pumps.

If zero., or reverse flow is detected, the condition should be alleviated by changing recirculation pump speed to re-establish forward flow or by tripping the pump.

C.1 With no recirculation loops in operation while in NODE 2 or if after going to single loop operations the required limits and setpoints cannot be established, the plant must be brought to MODE 3, where the LCO does not apply within

.12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

In this condition, the recirculation loops are not required to be operating because of the reduced severity of DBAs and minimal dependence on the recirculation loop coastdown characteristics.

The allowed Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is reasonable to reach MODE 3 from full power conditions in an orderly manner without challenging plant systems.

(continued)

SUSQUEHANNA - UNIT 1 TS / 8 3.4-5 Revision 1

Amendment No.

184

Reci r ation Loops Operating B 3.4.1

'ASES SURVEILLANCE REQUIREMENTS SR 3.4.1.1 This SR ensures the recirculation loops are wi,thin the allowable limits for mismatch.

At low core flow (i.e.,

< 75 million ibm/hr), the MCPR requi rements provide larger margins to the fuel cladding integrity Safety Limit such that the potential adverse effect of early boiling transition during a

LOCA is reduced.

A larger flow mismatch can therefore be allowed when core flow is < 75 million ibm/hr.

The recirculation loop jet pump flow, as used in this Surveillance, is the summation of the flows from all of the jet pumps associated with a single recirculation loop.

The mismatch is measured in terms of core flow. If the flow mismatch exceeds the specified limits, the loop with the lower flow is considered inoperable.

The SR is not requi red when both loops are not in operation since the mismatch limits are meaningless during single loop or natural'irculation operation.

The Surveillance must be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after both loops are in operation.

The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency is consistent with the Surveillance Frequency f'r jet pump OPERABILITY verification and has been shown by operating experience to be adequate to detect off normal jet pump loop flows in a timely manner.

SR 3.4.1.2 As noted, this SR is only applicable when in single loop operation.

This SR ensures the recirculation pump limit is maintained.

The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency is based on operating experience and the operators inherent knowledge of the current reactor status.

(continued)

SUSQUEHANNA - UNIT 1 TS /

B 3.4-6 Revision 1

Amendment No.

184

Recircu ation Loops Operating B 3.4.1

'ASES REFERENCES 1.

FSAR, Section 6.3.3.7.

2.

FSAR, Section 5.4.1.4.

3.

NEDO-31960-A "BWROG Long Term Stability Solutions Licensing Hethodology,"

November, 1995.

4.

NEDO-31960-A "BWROG Long Term Stability Solutsions Licensing Hethodology, Supplement 1," November 1995.

5.

Final Policy Statement on Technical Specifications Improvements, July 22, 1993 (58 FR 39132).

SUSQUEHANNA - UNIT 1 TS /

B 3.4-7 Revision 1

Amendment No.

184

)

SUSQUEHANNA - UNIT 1 TS / 8 3.4-8 Revision 1

Amendment No.

184

SUSQUEHANNA - UNIT 1 TS I 8 3.4-9 Revision 1

Amendment Mo.

184

I 0

~ ~a

~0

++*<<+

UNITED STATLS NUCLEAR REGULATORY.COMMISSIGiI WASHINGTON, D.C. 20555-0001 PPIIL INC.

ALLEGHENYELECTRIC COOPERATIVE INC.

DOCKET NO. 50-388 SUS UEHANNASTEAM ELECTRIC STATION UNIT 2 AMENDMENTTO FACILITYOPERATING LICENSE Amendment No. 158 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 PP8L, Inc., dated August 5, 1998, and supplemented by letters dated November 23, 1998, and June 23, 1999, 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 facilitywilloperate in conformity with the application, the provisions of the Act, and the regulations of the Commission; C.

There is reasonable assurance:

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

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

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

2.

Accordingly, 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:

(2)

Technical S ecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.158 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. PP&L shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of its date of issuance and is to be implemented within 30 days following startup from the Unit 1, eleventh Refueling Inspection Outage, currently scheduled for spring 2000.

FOR THE NUCLEAR REGULATORY COMMISSION S. Singh Bajwa, Director Project Directorate I-1 Division of Licensing Project Management Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

Ju'ly 30 f999

ATTACHMENTTO LICENSE AMENDMENTNO. 158 FACILITYOPERATING LICENSE NO. NPF-22 DOCKET NO. 50-388 Replace the following pages of the Appendix A Technical SpeciTications with enclosed pages.

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

REMOVE INSERT III IV V

Vl 3.4-1 3.4-2 3.4-3 3.4-4 3.4-5 8 3.3-44 8 3.3-45 8 3.3-46 8 3.3-47 8 3.3-48 8 3.3-49 8 3.3-50 8 3.3-51 8 3.4-1 8 3.4-2 8 3.4-3 8 3.4-4 8 3.4-5 8 3.4-6 8 3.4-7 8 3.4-8 8 3.4-9 I

II III IV 3.3-15a 3.3-15b 3.4-1 3.4-2 3.4-3 3.4P 3.4-5 8 3.3-43a 8 3.3-43b 8 3.3-43c 8 3.3-43d 8 3.3<3e 8 3.3-43f 8 3.3-43g 8 3.3-43h 8 3.4-1 8 3.4-2 8 3.4-3 8 3.4-4 8 3.4-5 8 3.4-6 8 3.4-7 8 3.4-8 8 3.4-9

I I

NTS (TE ICAL SPECIFICATIONS) 2.0 2.1 2.2 3.0 3.0 3.1 3.1.1 3.1.2 3.1.3 3.1.4 3.1.5 3.1.6 3.1.7 3.1.8 3.2 3.2.1 3.2.2 3.2.3 3.2.4 3.3 3.3.1.1 3.3.1.2 3.3.1.3 3.3.2.1 3.3.2.2 3.3.3.1 3.3.3.2 3.3.4.1 3.3.4.2 3.3.5.1 3.3.5.2 3.3.6.1 3.3.6.2 3.3.7.1

.ND APPLICATION tefinitions

.ogical Connectors

ompletion Times

.'requency SAFETY LIMITS (SLs)

SLs SL Violations LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY.

SURVEILLANCE REQUIREMENT (SR) APPLICABILITY REACTIVITY CONTROL SYSTEMS Shutdown Margin (SDM)

Reactivity Anomalies Control Rod OPERABILITY Control Rod Scram Times Control Rod Scram Accumulators Rod Pattern Control Standby Liquid Control (SLC) System Scram Discharge Volume (SDV) Vent and Drain Valves POWER DISTRIBUTION LIMITS Average Planar Linear Heat Generation Rate (APLHGR)

Minimum Critical Power Ratio (MCPR)

Linear Heat Generation Rate (LHGR)

Average Power Range Monitor (APRM) Gain and Setpoints INSTRUMENTATION Reactor Protection System (RPS) Instrumentation Source Range Monitor (SRM) Instrumentation Oscillation Power Range Monitor (OPRM)

Instrumentation Control Rod Block Instrumentation Feedwater

- Main Turbine High Water Level Trip Instrumentation Post Accident Monitoring (PAM) Instrumentation Remote Shutdown System End of Cycle Recirculation Pump Trip (EOC-RPT)

Instrumentation Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation Emergency Core Cooling System (ECCS) Instrumentation Reactor Core Isolation Cooling (RCIC) System Instrumentation Primary Containment Isolation Instrumentation Secondary Containment Isolation Instrumentation Control Room Emergency Outside Air Supply (CREOAS)

System Instrumentation 1.1-1 1.1-1 1.2-1 1.3-1 1.4-1 2.0-1 2.0-1 2.0-1 3.0-1 3.0-4 3.1-1 3.1-1 3.1-5 3.1-7 3.1-12 3.1-15 3.1-18 3.1-20 3.1-25 3.2-1 3.2-1 3:2-3 3.2-5 3.2-7 3.3-1 3.3-1 3.3-10 3.3-15a 3.3-16 3.3-20 3.3-22 3.3-26 3.3-29 3.3-33 3.3-36 3.3-48 3.3-52 3.3-63 3.3-67 SUSQUEHANNA - UNIT 2 Amendment

~o.

464.

158

TABLE OF CONTENTS (TECHNICAL SPECIFICATIONS) 3.3 3.3.8.1 3.3.8.2 3.4 3.4.1 3.4.2 3.4.3 3.4.4 3.4.5 3.4.6 3.4.7 3.4.8 3.4.9 3.4.10 3.4.11 3.5 3.5.1 3.5.2 3.5.3 3.6 3.6.1.1 3.6.1.2 3.6.1.3 3.6.1.4 3.6.1.5 3.6.1.6 3.6.2.1 3.6.2.2 3.6.2.3 3.6.2.4 3.6.3.1

, 3.6.3.2 3.6.3.3 3.6.4.1 3.6.4.2 3.6.4.3 INSTRUMENTATION (continued)

Loss of Power (LOP) Instrumentation Reactor Protection System (RPS) Electric Power Monitoring REACTOR COOLANT SYSTEM (RCS)

Recirculation Loops Operating Jet Pumps Safety/Relief Valves (S/RVs)

RCS Operational LEAKAGE RCS Pressure Isolation Valve (PIV) Leakage RCS Leakage Detection Instrumentation RCS Specific Activity Residual Heat Removal (RHR) Shutdown Cooling System-Hot Shutdown Residual Heat Removal (RHR) Shutdown Cooling System-Cold Shutdown RCS Pressure and Temperature (P/T) Limits Reactor Steam Dome Pressure EMERGENCY CORE COOLING SYSTEMS (ECCS)

AND REACTOR CORE ISOLATION COOLING (RCIC)

SYSTEM ECCS-Operating ECCS-Shutdown RCIC System CONTAINMENT SYSTEMS Primary Containment Primary Containment Air Lock............

Primary Containment Isolation Valves (PCIVs)

Containment Pressure Drywell Air Temperature Suppression-Chamber-to-Drywell Vacuum Breakers Suppression Pool Average Temperature Suppression Pool Water Level Residual Heat Removal (RHR) Suppression Pool Cooling Residual Heat Removal (RHR) Suppression Pool Spray Primary Containment Hydrogen Recombiners Drywell Air Flow System Primary Containment Oxygen Concentration Secondary Containment Secondary Containment Isolation Valves (SCIVs)

Standby Gas Treatment (SGT) System 3.3-73 3.3-76 3.4-1 3.4-1 3.4-6 3.4-8 3.4-10 3.4-12 3.4-14 3.4-17 3.4-19 3.4-22 3.4-24 3.4-31 3.5-1 3.5-1 3.5-8 3.5-12 3.6-1 3.6-1 3.6-4 3.6-8 3.6-17 3.6-18 3.6-19 3.6-22 3.6-25 3.6-26 3.6-28 3.6-30 3.6-32 3.6-34 3.6-35 3.6-38 3.6-42 SUSQUEHANNA - UNIT 2 Amendment No.

$64, 158

TABLE OF CONTENTS (TECHNICAL SPECIFICATIONS) 3.7 3.7.1 3.7.2 3.7.3 3.7.4 3.7.5 3.7.6 3.7.7 3.8 3.8.1 3.8.2 3.8.3

'.8.4

'.8.5 3.8.6 3.8.7 3.8.8 3.9 3.9.1 3.9.2 3.9.3 3.9.4 3.9.5 3.9.6 3.9.7 3.9.8 3.10 3.10.1 3.10.2 3.10.3 3.10.4 3.10.5 3.10.6 3.10.7 3.10.8 4.0 4.1 4.2 4.3 PLANT SYSTEMS Residual Heat Removal Service Water (RHRSW) System and the Ultimate Heat Sink (UHS)

Emergency Service Water (ESW) System Control Room Emergency Outside Air Supply (CREOAS) System Control Room Floor Cooling System Main Condenser Offgas Main Turbine Bypass System Spent Fuel Storage Pool Water Level ELECTRICAL POWER SYSTEMS AC Sources-Operating AC Sources-Shutdown Diesel Fuel Oil, Lube Oil, and Starting Air DC Sources-Operating DC Sources-Shutdown Battery Cell Parameters Distribution Systems-Operating

, Distribution Systems-Shutdown REFUELING OPERATIONS Refueling Equipment Interlocks........

Refuel Position One-Rod-Out Interlock Control Rod Position Control Rod Position Indication Control Rod OPERABILITY-Refueling..

Reactor Pressure Vessel (RPV) Water Level Residual Heat Removal (RHR) High Water Level Residual Heat Removal (RHR)-Low Water Level SPECIAL OPERATIONS Inservice Leak and Hydrostatic Testing Operation Reactor Mode Switch Interlock Testing Single Control Rod Withdrawal -Hot Shutdown Single Control Rod Withdrawal -Cold Shutdown Single Control Rod Drive (CRD) Removal Refueling Multiple Control Rod Withdrawal -Refueling Control Rod Testing-Operating SHUTDOWN MARGIN (SDM) Test-Refueling DESIGN FEATURES Site Location Reactor Core Fuel Storage 3.7-1 3.7-1 3.7-4 3.7-6 3.7-10 3.7-13 3.7-15 3.7-17 3.8-1 3.8-1 3.8-18 3.8-22 3.8-25 3.8-32 3.8-37 3.8-42 3.8-48 3.9-1 3.9-1 3.9-3 3.9-5 3.9-6 3.9-8 3.9-9 3.9-10 3.9-13 3.10-1 3.10-1 3.10-4 3.10-6 3.10-9 3.10-13 3.10-16 3.10-18 3.10-20 4.0-1 4.0-1 4.0-1 4.0-1 SUSQUEHANNA - UNIT 2 Amendment No.

$ 54, 1S8

r TABLE OF CONTENTS (TECHNICAL SPECIFICATIONS) 5.0 5.1 5.2 5.3 5.4 5.5 5.6 5.7 ADMINISTRATIVE CONTROLS Responsib'ility......

Organization.......

Unit Staff Qualifications Procedures Programs and Manuals Reporting Requirements High Radiation Area 5.0-1 5.0-1 5.0-2 5.0-5 5.0-6 5.0-7 5.0-19 5.0-25 SUSQUEHANNA - UNIT 2 iv Amendment No.454, 158

~ I

OPRH Instrumentation 3.3.1.3 3.3 INSTRUMENTATION 3.3. 1.3 Oscillation Power Range Monitor (OPRM) Instrumentation LCO 3.3. 1. 3 APPLICABILITY Four channels of the OPRH instrumentation shall be OPERABLE.

Each OPRH channel Period Based Algorithm (Sp) Allowable Value shall be less than or equal to 1.10 at a confi rmation count permissive (Np) of 10.

Thermal Power o 25K RTP.

ACTIONS:

NOTE-Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A.

One or more required channels inoperable.

A.1 OR Place channel in trip.

30 days A.2 Place associated RPS trip system in trip.

OR 30 days A.3 Initiate alternate.

method to detect and suppress thermal hydraulic instability osci llations.

30 days B.

OPRM trip capability not maintained.

C.

Required Action and associated Completion Time not met.

B. 1 Initiate alternate method to detect and suppress thermal hydraullic instability osci llations.

AND 8.2

. Restore OPRH trip capabi 1 ity C.l Reduce THERMAL POWER

< 25K RTP.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 120 days 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months SUSQUEHANNA - UNIT 2 TS I 3.3-15a Amendment No.158 I

t OPRM Instrumentation 3.3.1.3 SURVEILLANCE REQUIREMENTS NOTE----------------------------------

When a channel is placed in an inoperable status solely for performance of requi red Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the OPRM System maintains trip capabi 1 ity.

SURVEILLANCE FREQUENCY SR 3.3.1.3.1 Per form CHANNEL FUNCTIONAL TEST.

184 days SR 3.3.1.3.2 Calibrate the local power range monitors.

1000 MWD/MT average core exposure SR 3.3.1.3.3 NOTE Neutron detectors are excluded.

Perform CHANNEL CALIBRATION.

24 months SR 3.3.1.3.4 Perform LOGIC SYSTEM FUNCTIONAL TEST 24 months SR 3.3. 1.3.5 Verify OPRM is not bypassed when THERMAL POWER is ) 30K RTP and core flow( 60 MLb/Hr.

24 months SR 3.3.1.3.6

NOTE Neutron detectors are excluded.

Verify the RPS

RESPONSE

TIME is within limlts.

24 months on a

STAGGERED TEST BASIS I

SUSQUEHANNA - UNIT 2 TS / 3.3-15b Amendment No.

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Recircu ation Loops Operating 3.4.1

'.4 REACTOR COOLANT SYSTEM (RCS) 3.4. 1 Recirculation Loops Operating LCO 3.4.1 Two recirculation loops with matched flows shall be in operation.

OR One recirculation loop may be in operation provided the following limits are applied when the associated LCO is applicable:

a.

LCO 3.2. 1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," single loop operation limits specified in the COLR; b.

LCO 3.2.2, "MINIMUMCRITICAL POWER RATIO (MCPR)," single loop operation limits specified in the COLR; LCO 3.2.3, "LINEAR HEAT GENERATION RATE (LHGR)," single loop operation limits specified in the

COLR, and LCO 3.3. 1. 1, "Reactor Protection System (RPS)

Instrumentation,"

Function 2.b (Average Power Range Monitors Flow Biased Simulated Thermal Power-High),

Allowable Value of Table 3.3.1. 1-1 is reset for single loop operation.

e.

Recirculation pump speed is ~ 80K

-Note-Required limit and setpoint resets for single recirculation loop operation may be delayed f'r up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after transition from two recirculation loop operation to single recirculation loop operation.

APPLICABILITY:

MODES 1 and 2.

SUSQUEHANNA - UNIT 2 TS / 3.4-1 Amendmentno.

$54, 158

Recirculation Loops Operating 3.4.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME

)

A.

No recirculation loops operating while in MODE 1

A.1 Place reactor mode switch in the shutdown position.

Immediately B.

Recirculation loop flow mismatch not within limits.

B.1 Declare the recirculation loop with lower flow to be "not in operation."

2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months C.

No recirculation loops in operation while in MODE 2 OR Single Recirculation Loop required limits and setpoints not established within required time.

C.1 Be in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SUSQUEHANNA - UNIT 2 TS / 3.4-2 Amendment No.

$5$,

158

THIS PAGE INTENTIONALLYLEFT BLANK SUSQUEHANNA - UNIT 2 TS / 3.4-3 Amendment No.

454, 158

Recirculation Loops Operating 3.4.1 SURVEILLANCE REQUIREHENTS SURVEILLANCE FREQUENCY SR 3.4.1.1

-NOTE-Not requi red to be performed unti 1 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after both recirculation loops are in operation.

Verify recirculation loop jet pump flow mismatch with both recirculation loops in operation is:

a.

( 10 million ibm/hr when operating at

< 75 million ibm/hr total core flow; and b.

s 5 million ibm/hr when operating at

) 75 million ibm/hr total core flow.

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months I

SR 3.4.1.2

--NOTE Only required to be met during single loop operations.

Verify recirculation pump speed is within the limits specified in the LCO.

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months SUSQUEHANNA - UNIT 2 TS / 3.4-4 Amendment No.

454, 158

C

Recirculation Loops Operating 3.4.1 IC 5'Figure 3.4.1-1)

THIS PAGE INlENl IONALLYLEIFT BLANK SUSQUEHANNA - UNIT 2 TS / 3.4-5 Amendment No.

454,

$ 58

OPRH Instrumentation 8 3.3.1.3 3.3 INSTRUMENTATION B 3.3. 1.3 Oscillation Power Range Monitor (OPRM)

BASES BACKGROUND General Design Criterion 10 (GDC 10) requires the reactor core and associated coolant, control, and protection system to be designed with appropriate margin to assure that acceptable fuel design limits are not exceeded during any condition of normal operation including the effects of anticipated operational occurrences.

Additionally, GDC 12 requires the reactor core and associated coolant control and protection systems to be designed to assure that power osci llations which can result in conditions exceeding acceptable fuel design limits are either not possible or can be reliably and readily detected and suppressed.

The OPRH System provides compliance with GDC 10 and GDC 12, thereby roviding protection from exceeding the fuel MCPR safety imlt.

References 1, 2, and 3 describe three separate algorithms for detecting stability related oscillations:

the period based detection algorithm, the amplitude based algorithm.

and the growth rate algorithm.

The OPRH System hardware implements these algorithms in microprocessor based modules.

These modules execute the algorithms based on LPRH inputs and generate alarms and trips based on these calculations.

These trips result in tripping the Reactor Protection System (RPS) when the appropriate RPS trip logic is satisfied, as described in the Bases for LCO 3.3. 1. 1, "RPS Instrumentation."

Only the period based detection algorithm is used in the safety analysis (Ref.

1, 2, 6

8 7).

The remaining algorithms provide defense-in-depth and additional protection against unanticipated oscillations.

The period based detection algorithm detects a stabi lity-related oscillation based on the occurrence of a fixed number of consecutive LPRH signal period confi rmations followed by the LPRH signal amplitude exceeding a specified setpoint.

Upon detection of a stability related oscillation, a trip is generated for that OPRH channel.

The OPRM System consists of four OPRH trip channels, each channel consisting of two OPRH modules.

Each OPRH module receives input from LPRHs.

Each OPRM module also receives (continued)

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l OPRH Instrumentation B 3.3.1.3

'ASES BACKGROUND (continued) input from the Neutron Monitoring System (NHS) average power range monitor (APRM) power and flow signals to automatically enable the trip function of the OPRH module.

Each OPRM module is continuously tested by a self-test function.

On detection of any OPRH module failure, either a

trouble alarm or INOP alarm is activated.

The OPRM module provides an INOP alarm when the self-test feature indicates that the OPRM module may not be capable of meeting its functional requirements.

APPLICABLE SAFETY ANALYSES It has been shown that BWR cores may exhibit thermal-hydraulic reactor instabilities in high power and low flow portions of the core power to flow operating domain.

GDC 10 requires the reactor core and associated coolant control and protection systems to be designed with appropriate margin to assure that acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.

GDC 12 requires assurance that power osci llations which can result in conditions exceeding acceptable fuel design limits are either not possible or can

,be reliably and readily detected and suppressed.

The OPRH System provides compliance with GDC 10 and GDC 12 by detecting the onset of osci llations and suppressing them by initiating a reactor scram.

This assures that the MCPR safety limit will not be violated for anticipated osci llations.

The OPRH Instrumentation satisfies Criterion 3 of the NRC Policy Statement.

LCO Four channels of the OPRH System are required to be OPERABLE to ensure that stability related osci llations are detected and suppressed prior to exceeding the HCPR safety limit.

Only one of the two OPRH modules'eriod based detection algorithm is required for OPRH channel OPERABILITY.

The minimum number of'PRHs required OPERABLE to maintain an OPRM channel OPERABLE is consistent with the minimum number of LPRHs required to maintain the APRH system OPERABLE per LCO 3.3.1.1.

(continued)

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P

OPRM Instrumentation B 3.3.1.3

'ASES LCO (continued)

The Allowable Value for the OPRM Period Based Algorithm setpoint (Sp) is derived from the Analytic Limit corrected for instrument and calibration errors.

APPLICABILITY The OPRM instrumentation is required to be OPERABLE in order to detect and suppress neutron flux osci llations in the event of thermal-hydraulic instability.

As described in References

1. 2, and 3. the power/core flow region protected against anticipated osci llations is defined by THERMAL POWER a 30K RTP and core flow c 60 Mlb/Hr.

The OPRM trip is required to be enabled 'in this region, and the OPRM must be capable of enabling the trip function as a result of anticipated transients that place the core in that power/flow condition.

Theref'ore, the OPRM is requi red to be OPERABLE with THERMAL POWER ) 25K RTP and at all core flows while above that THERMAL POWER.

It is not necessary for the OPRM to be operable with THERMAL POWER < 25K RTP because transients from below this THERMAL POWER are not anticipated to result in power that exceeds 30K RTP.

ACTIONS A Note has been provided to modify the ACTIONS related to the OPRM instrumentation channels.

Section 1.3, Completion Times, specifies that once a Condition has been entered subsequent divisions, subsystems, components.

or variables expressed in the Condition discovered to be inoperable or not within limits will not result in separate entry into the Condition.

Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional

, failure with Completion Times based on initial entry into the Condition.

However, the Required Actions for inoperable OPRM instrumentation channels provide appropriate compensatory measures for separate inoperable channels.

As

such, a Note has been provided that allows separate Condition entry for each inoperable OPRM instrumentation channel.

Al A2 and A3 Because of the reliability and on-line self-testing of the OPRM instrumentation and the redundancy of the RPS design, an allowable out of service time of 30 days has been shown (continued)

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OPRH Instrumentation B 3.3.1.3 BASES ACTIONS Al A2 and A3 (continued) to be acceptable (Ref.

7) to permit restoration of any inoperable channel to OPERABLE status.

However, this out of service time is only acceptable provided the OPRM instrumentation still maintains OPRM trip capability (refer to Required Actions B. 1 and B.2).

The remaining OPERABLE OPRH channels continue to provide trip capability (see Condition 8) and provide operator information relative to stability activity.

The remaining OPRH modules have high reliability.

With this high reliabi lity, there is a low probabi lity of a subsequent channel fai lure within the allowable out of service time.

In additions the OPRH modules continue to perform on-line self-testing and alert the operator if any further system degradation occurs.

If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the OPRH channel or associated RPS trip system must be placed in the tripped condition per Required Actions A. 1 and A.2.

Placing the inoperable OPRH channel in trip (or the associated RPS trip system in trip) would conservatively compensate for the inoperability, provide the capabi'lity to accommodate a single failure and allow operation to continue.

Alternately, if it is not desired to place the OPRH channel (or RPS trip system) in trip (e.g.,

as in the case where placing the inoperable channel in trip would result in a fu11 scram),

the alternate method of detecting and suppression thermal hydraulic instability osci llations is required (Required Action A.3).

This alternate method is described in Reference 5.

It consists of increased operator awareness and monitoring for neutron flux osci llations when operating in the region where osci llations are possible.

If indications of oscillation, as described in Reference 5 are observed by the operator, the operator will take the actions described by procedures which include initiating a manual scram of the reactor.

B.l and 8.2 Required action B. 1 is intended to ensure that appropriate actions are taken if multiple, inoper able, untripped OPRH channels within the same RPS trip system result in not maintaining OPRH trip capability.

OPRH trip capability is (continued)

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t OPRM Instrumentation 8 3.3.1.3 BASES ACTIONS B. 1 and 8.2 (continued) considered to be maintained when sufficient OPRM channels are OPERABLE or in trip (or the associated RPS trip system is in trip). such that a valid OPRM signal will generate a

trip signal in both RPS trip systems (this would requi re

'oth RPS trip systems to have at least one OPRM channel OPERABLE or the associated RPS trip system in trip).

Because of the low probability of the occurrence of an instability.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is an acceptable time to initiate the alternate method of detecting and suppression thermal hydraulic instability oscillations described in Action A.3 above.

The alternate method of detecting and suppressing thermal hydraulic instability osci llations would adequately address detection and mitigation in the event of instability osci llations.

Based on industry operating experience with actual instability osci llations, the operator would be able to recognize instabilities during this time and take action to suppress them through a manual scram.

In addition, the OPRM System may still be available to provide alarms to the operator if the onset of osci llations were to occur.

Since plant operation is minimized in areas where osci llations may occur, operation for 120 days without OPRM trip capability is considered acceptable with implementation of the alternate method of detecting and suppression thermal hydraulic instability oscillations.

C.1 With any Required Action and associated Completion Time not

met, THERMAL POWER must be reduced to < 25K RTP within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Reducing THERMAL POWER to < 25K RTP places the plant in a condition where instabilities are not likely to occur.

The 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is reasonable, based on operating experience.

to reduce THERMAL POWER < 25K RTP from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE REQUIREMENTS SR 3.3.1.3.1 A CHANNEL FUNCTIONAL TEST is performed to ensure that the entire channel will perform the intended f'unction.

A Frequency of 184 days provides an acceptable level of system (continued)

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Amendment No.

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OPRM Instrumentation B 3.3.1.3 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.3.1.3.1 average availability over the Frequency and is based on the reliability of the channel.

(Ref. 7)

SR 3.3.1.3.2 LPRH gain settings are determined from the local flux profiles measured by the Traversing Incore Probe (TIP)

System.

This establishes the relative local flux profile for appropriate representative input to the OPRH System.

The 1000 HWD/MT Frequency is based on operating experience with LPRM sensitivity changes.

SR 3.3.1.3.3 A CHANNEL CALIBRATION verifies that the channel responds to the measured parameter within the necessary range and accuracy.

CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations.

Calibration of the channel provides a check of the internal reference voltage and the internal processor clock f'requency.

It also compares the desi red trip setpoints with those in processor memory.

Since the OPRH is a digital system, the internal reference voltage and processor clock frequency are, in turn, used to automatically calibrate the internal analog to digital convertors.

As noted, neutron detectors are excluded from CHANNEL CALIBRATION because they are passive

devices, with minimal drift, and because of the difficulty of simulating a

meaningful signal.

Changes in neutron detector sensitivity are compensated for by performing the 1000 HWD/MT LPRM calibration using the TIPs (SR 3.3. 1.3.2).

The Frequency of 24 months is based upon the assumption of the magnitude of equipment drift provided by the equipment supplier.

(Ref. 7)

SR 3.3.1.3.4 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific channel.

The functional testing of control rods, in LCO (continued)

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'V

OPRM Instrumentation B 3.3.1.3 BASES SURVEILLANCE REQUIREMENTS SR 3.3. 1.3.4 (continued)

3. 1.3.

"Control Rod OPERABILITY". and scram discharge volume (SDV) vent and drain valves, in LCO 3. 1.8, "Scram Discharge Volume (SDV) Vent and Drain Valves", overlaps this Surveillance to provide complete testing of the assumed safety function.

The OPRM self-test function may be utilized to perform this testing for those components that it is designed to monitor.

The 24 month Frequency is based on engineering judgement, reliability of the components and operating experience.

SR 3.3.1.3.5 This SR ensure that trips initiated from the OPRM System will not be inadvertently bypassed when THERMAL POWER is

~ 30%

RTP and core flow is c 60 MLb/Hr. This normally involves calibration of the bypass channels.

Adequate margins for the instrument setpoint methodology are incorporated into the actual setpoints (Reference 7).

If any bypass channel setpoint is nonconservative (i.e.. the OPRM module is bypassed at a 30K RTP and core flow is

~ 60 MLB/Hr), then the affected OPRM module is considered inoperable.

Alternatively, the bypassed channel can be manually placed in the conservative position (Manual Enable).

If placed in the manual enable conditions this SR is met and the module is considered OPERABLE.

The 24 month Frequency is based on engineering judgment and reliability of the components.

SR 3.3.1.3.6 This SR ensure that the individual channel response times are less than or equal to the maximum values assumed in the safety analysis (Ref. 6)

The OPRM self-test function may be utilized to perform this testing for those components it is designed to'onitor.

The LPRM amplifier cards inputting to the OPRM are excluded from the OPRM response time testing.

The RPS

RESPONSE

TIME acceptance criteria are included in Reference 8.

(continued)

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OPRM Instrumentation B 3.3.1.3

'ASES SURVEILLANCE REQUIREMENTS SR 3.3. 1.3.6 (continued)

As noted, neutron 'detectors are excluded from RPS

RESPONSE

TIME testing because the principles of detector operation virtually ensure an instantaneous response time.

RPS

RESPONSE

TIME tests are conducted on an 24 month STAGGERED TEST BASIS.

This Frequency is based upon operating experience, which shows that random failures of instrumentation components causing serious response time degradation, but not channel fai lure, are infrequency occurrences.

REFERENCES 1.

NED0-31960-A, "BWR Owners Group Long-Term Stability Solutions Licensing Methodology",

November 1995.

2.

NEDO 31960-A, Supplement 1,

"BWR Owners Group Long-Term Stability Solutions Licensing Methodology",

November 1995.

3.

NRC Letter, A. Thadani to L. A. England, "Acceptance for Referencing of Topical Reports NED0-31960, Supplement 1,

'BWR Owners Group Long-Term Stability Solutions Licensing Methodology', July 12, 1994.

4.

Generic Letter 94-02, "Long-Term Solutions and Upgrade of Interim Operating Recommendations for Thermal-Hydraulic Instabilities in Boiling Water Reactors",

July ll, 1994.

5.

BWROG Letter BWROG-9479, "Guidelines for Stability Interim Corrective Action", June 6.

1994.

6.

NED0-32465-A, "BWR Owners Group Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology and Reload Applications", August 1996.

7.

CENPD-400-P-A, Rev.

01, "Generic Topical Report for the ABB Option III Oscillation Power Range Monitor (OPRM)",

May 1995.

8.

FSAR, Table 7.3-28.

(continued)

I SUSQUEHANNA - UNIT 2 TS I B 3.3-43h Revision 0

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Recirc ation Loops Operating B 3.4.1 B 3.4 REACTOR COOLANT SYSTEH (RCS) 8 3.4. 1 Recirculation Loops Operating BASES BACKGROUNO The Reactor Coolant Recirculation System is designed to provide a forced coolant flow through the core to remove heat from the fuel.

The forced coolant flow removes more heat from the fuel than would be possible with just natural circulation.

The forced flow, therefore, allows operation at significantly higher power than would otherwise be possible.

The recirculation system also controls reactivity over a wide span of reactor power by varying the recirculation flow rate to control the void content of the moderator.

The Reactor Coolant Recirculation System consists of two recirculation pump loops external to the reactor vessel.

These loops provide the piping path for the driving flow of water to the reactor vessel jet pumps.

Each external loop contains one variable speed motor driven recirculation

pump, a motor generator (HG) set to control pump speed and associated piping, jet pumps, valves.

and instrumentation.

The recirculation pump, piping, and valves are part of the reactor coolant pressure boundary and are located inside the drywell structure.

The jet pumps are reactor vessel internals.

The recirculated coolant consists of saturated water from the steam separators and dryers that has been subcooled by incoming feedwater.

This water passes down the annulus'etween the reactor vessel wall and the core shroud.

A por tion of the coolant flows from the vessel, through the two external recirculation

loops, and becomes the driving flow for the jet pumps.

Each of the two external recirculation loops discharges high pressure flow into an external manifold. from which individual recirculation inlet lines are routed to the jet pump risers within the reactor vessel.

The remaining portion of the coolant mixture in the annulus becomes the suction flow for the jet pumps.

" This flow enters the jet pump at suction inlets and is accelerated by the driving flow.

The drive flow and suction flow are mixed in the jet pump throat section.

The total flow then passes through the jet pump diffuser section into the area below the core (lower plenum), gaining sufficient head in the process to drive the required flow upward through the core.

The subcooled water enters the bottom of the fuel channels and contacts the fuel cladding, where heat (continued)

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1

Recircu ation Loops Operating B 3.4.1 BASES BACKGROUND (continued) is transferred to the coolant.

As it rises, the coolant begins to boi 1, creating steam voids within the f'uel channel that continue until the coolant 'exits the core.

Because of reduced moderation, the steam voiding introduces negative reactivity that must be compensated for to maintain or to increase reactor power.

The recirculation flow control allows operators to increase recirculation flow and sweep some of the voids from the fuel channel, overcoming the negative reactivity void effect.

Thus, the reason for having variable recirculation flow is to compensate for reactivity effects of boiling over a wide range of power generation without having to move control rods and disturb desirable flux patterns.

Each recirculation loop is manually started from the control room.

The MG set provides regulation of individual recirculation loop drive flows.

The flow in each loop is manually controlled.

APPLICABLE SAFETY ANALYSES The operation of the Reactor Coolant Recirculation System is an initial condition assumed in the design basis loss of coolant accident (LOCA) (Ref. 1).

During a LOCA caused by a recirculation loop pipe break, the intact loop is assumed to provide coolant flow during the first few seconds of the accident.

The initial core flow decrease is rapid because the recirculation pump in the broken loop ceases to pump reactor coolant to the vessel almost immediately.

The pump in the intact loop coasts down relatively slowly.

This pump coastdown governs the core flow response for the next several seconds until the jet pump suction is uncovered (Ref. 1).

The analyses assume that both loops are operating at the same flow prior to the accident.

However, the LOCA analysis was reviewed for the case with a flow mismatch between the two loops, with the pipe break assumed to be in the loop with the higher flow.

While the flow coastdown and core response are potentially more severe in this assumed case (since the intact loop starts at a lower flow rate and the core response is.the same as if both loops were operating at a lower flow rate),

a small mismatch has been determined to be acceptable based on engineering judgement.

The recirculation system is also assumed to have sufficient flow coastdown characteristics to maintain fuel thermal margins during abnormal operational transients (Ref. 2),

which are analyzed in Chapter 15 of the FSAR.

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E

Recircu ation Loops Operating 8 3.4.1 BASES APPLICABLE SAFETY ANALYSES (continued)

Plant specific LOCA analyses have been performed assuming only one operating recirculation loop.

These analyses have demonstrated that, in the event of a LOCA caused by a pipe break in the operating recirculation loop, the Emergency Core Cooling System response wi 11 provide adequate core cooling, provided that the LHGR limit for SPC 9x9-2 fuel and GE lead use assemblies and the APLHGR limit for SPC ATRIUM'-10fuel is modified.

The transient analyses of Chapter 15 oi the FSAR have also been performed for single recirculation loop operation and demonstrate sufficient flow coastdown characteristics to maintain fuel thermal margins during the abnormal operational transients analyzed provided the MCPR requi rements are, modified.

During single recirculation loop operation, modification to the Reactor Protection System (RPS) average power range monitor (APRM) instrument setpoints is also required to account for the different relationships between recirculation drive flow and reactor core flow.

The APLHGR, LHGR and MCPR setpoints for single loop operation are specified in the COLR.

The APRM flow biased simulated THERMAL POWER setpoint is in LCO 3.3. 1.1, "Reactor Protection System (RPS) Instrumentation."

In

addition, a restriction on recirculation pump speed is incorporated to address Reactor Vessel Internals vibration concerns and assumptions in the event analysis.

Recirculation loops operating satisfies Criterion 2 of the NRC Policy Statement (Ref. 5).

LCO Two recirculation loops are required to be in oper ation with their flows matched within the limits specified in SR 3.4. 1. 1 to ensure that during a

LOCA caused by a break of the piping of one recirculation loop the assumptions of the LOCA analysis are satisfied.

With the limits specified in SR 3.4. 1. 1 not met. the recirculation loop with the lower flow must be considered not in operation.

With only one recirculation loop in operation, modifications to the required APLHGR limits (LCO 3.2. 1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE"), LHGR limits (LCO 3.2.3, "LINEAR HEAT GENERATION RATE (LHGR)"), MCPR limits (LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)"), and APRM Flow Biased Simulated Thermal Power-High setpoint (LCO 3.3.1.1) may be applied to allow continued operation consistent with the safety analysis assumptions.

Furthermore, restrictions are (continued)

SUSQUEHANNA - UNIT 2 TS I B 3.4-3 Revision 1

Amendment No. i58

Recircu ation Loops Operating B 3.4.1.

BASES LCO (continued)

I placed on recirculation pump speed to assure the initial assumptions of the event analysis are maintained.

The LCO is modified by a Note that allows up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to establish the required limits and setpoints after a change from two recirculati'on loops operation to single recirculation loop operation.

If the limits and setpoints are not in compliance with the applicable requi rements at the end of this period, the ACTIONS requi red by the applicable specifications must be implemented.

This time is rovided to stabilize operation with one recirculation loop y:

limiting flow in the operating loop, limiting total THERMAL POWER; and, fully implementing and confirming the required limit and setpoint modifications.

APPLICABILITY In MODES 1 and 2, requi rements for operation of the Reactor Coolant Recirculation System are necessary since there is considerable energy in the reactor core and the limiting design basis transients and accidents are assumed to occur.

In MODES 3, 4.

and 5, the consequences of an accident are reduced and the coastdown characteristics of'he recirculation loops are not important.

ACTIONS A.l When operating with no recirculation loops operating in MODE 1, the potential for thermal-hydraulic osci llations is greatly increased.

Although this transient is protected for expected modes of oscillation by the OPRM system, when operable per LCO 3.3. 1.3 (References 3 and 4), the prudent response to the natural circulation condition is to preclude potential thermal-hydraulic osci llations by immediately placing the mode switch in the shutdown position.

ACTIONS (continued)

B. 1 Recirculation loop flow must match within required limits when both recirculation loops are in operation.

If flow mismatch is not within required limits, matched'low must be restored within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

If matched flows are not restored, the recirculation loop with lower flow must be declared "not SUSQUEHANNA - UNIT 2 TS /

B 3.4-4 (continued)

Revision 1

Amendment No.

158

Recirculation Loops Operating B 3.4.1 BASES in operation."

Should a

LOCA occur with recirculation loop flow not matched, the core flow coastdown and resultant core response may not be bounded by the LOCA analyses.

Therefore.

only a limited time is allowed prior to imposing restrictions associated with single loop operation.

Operation with only one recirculation loop satisfies the requirements of the LCO and the initial conditions of the accident sequence.

The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time is based on the low probability of an accident occurring during this time period, providing a reasonable time to complete the Required Action, and considering that frequent core monitoring by operators allows abrupt changes in core flow conditions to be quickly detected.

These Required Actions do not requi re tripping the recirculation pump in the lowest flow loop when the mismatch between total jet pump flows of the two loops is greater than the requi red limits.

However, in cases where large flow mismatches

occur, low flow or reverse flow can occur in the low flow loop jet pumps, causing vibration of the jet pumps.

If zero or reverse flow is detected, the condition should be alleviated by changing recirculation pump speed to re-establish forward flow or by tripping the pump.

C.1 With no recirculation loops in operation while in MODE 2 or if after going to single loop operations the requi red limits and setpoints cannot be established.

the plant must be brought to NODE 3, where the LCO does not apply within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

In this condition, the recirculation loops are not requi red to be operating because of the reduced severity of DBAs and minimal dependence on the recirculation loop coastdown characteristics.

The allowed Completion Time of

~ 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is reasonable to reach MODE 3 from full power conditions in an orderly manner without challenging plant systems.

(continued)

SUSQUEHANNA - UNIT 2 TS /

B 3.4-5 Revision 1

Amendment No.

158

Recirculation Loops Operating B 3.4.1 BASES SURVEILLANCE REQUIREMENTS SR 3.4.1.1 This SR ensures the recirculation loops are within the allowable limits for mismatch.

At low core flow (i.e.,

< 75 million ibm/hr), the MCPR requirements provide larger margins to the fuel cladding integrity Safety Limit such that the potential adverse effect of early boi ling transition during a

LOCA is reduced.

A larger flow mismatch can therefore be allowed when core flow is < 75 million ibm/hr.

The recirculation loop jet pump flow, as used in this Surveillance, is the summation of the flows from all of the jet pumps associated with a single recirculation loop.

The mismatch is measured in terms of core flow. If the flow mismatch exceeds the specified limits, the loop with the lower flow is considered inoperable.

The SR is not requi red when both loops are not in operation since the mismatch limits are meaningless during single loop or natural circulation operation.

The Surveillance must be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after both loops are in operation.

The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency is consistent with the Surveillance Frequency for jet pump OPERABILITY verification and has been shown by operating experience to be adequate to detect off normal jet pump loop flows in a timely manner.

SR 3.4.1.2 As noted. this SR is only applicable when in single loop operation.

This SR ensures the recirculation pump limit is maintained.

The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency is based on operating experience and the operators inherent knowledge of the current reactor status.

(continued)

I SUSQUEHANNA - UNIT 2 TS /

B 3.4-6 Revision 1

Amendment No.

158

Recircu ation Loops Operating B 3.4.1 BASES REFERENCES 1.

FSAR, Section 6.3.3.7.

2.

FSAR, Section 5.4. 1.4.

3.

NEOO-31960-A "BWROG Long Term Stability Solutions Licensing Hethodology,"

November, 1995.

4.

NEOO-31960-A "BWROG Long Term Stability Solutions Licensing Hethodology, Supplement 1," November 1995.

5.

Final Policy Statement on Technical Specifications Improvements, July 22, 1993 (58 FR 39132).

(continued)

SUSQUEHANNA - UNIT 2 TS /

B 3.4-7 Revision 1

Amendment No.

158

0

Recircu ation Loops Operating B 3.4.1 BASES THIS PAGE INTENTIONALLYLEFT BLANK (continued)

I SUSQUEHANNA - UNIT 2 TS / 8 3.4-8 Revision 1

Amendment No.

158

Recir tion Loops Operating B 3.4.1 BASES (continued)

SUSQUEHANNA - UNIT 2 TS /

B 3.4-9 Revision 1

Amendment No.

158

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