ML17146A416

Proposed Tech Specs Re Cycle 2 Reload.Nshc Encl

ML17146A416
Person / Time
Site:	Susquehanna
Issue date:	06/19/1986
From:	PENNSYLVANIA POWER & LIGHT CO.
To:
Shared Package
ML17146A415	List: ML17146A414 ML17146A416 ML17146A417 ML17146A419 ML17146A421 ML17146A422 ML17146A423 ML18040B150
References
NUDOCS 8606240307
Download: ML17146A416 (82)

Text

INDEX DEFINITIONS SECTION

l. 0 DEFINITIONS PAGE ACTIONo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1-1 1.2 AVERAGE QbiiNA EXPOSURE...................................

1.3 AVERAGE PLANAR LINEAR HEAT GENERATION RATE................

1.4 CHANNEL CALIBRATION.............'..........................

1.5 CHANNEL CHECKS ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

1.6 CHANNEl. FUNCTIONAL TEST...................................

1.7 CORE ALTERATION........................................... 1-2 1.8 CRITICAL POWER RATIO.. 1-2 1.9 DOSE EQUIVALENT I-131..................................... 1-2

l. 10 Z-AVERAGE DISINTEGRATION ENERGY........................... 1-2 1.11 EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME......... 1-2 1.12 END-OF"CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME.. 1"2

l. 13 FRACTION OF LIMITING POWER DENSITY................

l. 14 FRACTION OF RATED THERMAL POWER.................... 1" 3

1. 15 FRE(UENCY NOTATION....................,........... 1-3

l. 16 GASEOUS RAOWASTE TREATMENT SYSTEM............ ~... ~ 1-3

1. 17 IDENTIFIED LEAKAGE........................ 1-3

1. 18 ISOLATION SYSTEM RESPONSE TIME.................... 1-3

1. 19 LIMITING CONTROL ROD PATTERN...................... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

1~%

1. 20 LINEAR HEAT GENERATION RATE....................... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

1~%

l. 21 LOGIC SYSTEM FUNCTIONAL TEST.....................- 1-4 1.22 MAXIMUM FRACTION OF LIMITING POWER DENSITY................. 1"4

l. 23 MEMBER(S) OF THE PUBLIC........................... 1-4

1. 24 MINIMUM CRITICAL POWER RATIO......... 1-4

1. 25 OFFSITE DOSE CALCULATIQN MANUAL........... 1" 4 SUS)UEHANNA - UNIT 2 8606240307 860619 PDR ADOCK o5o00388 P PDR

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION PAGE 3/4.0 APPLICABILITY............................................. 3/4 0"1 3/4. 1 REACTIVITY CONTROL SYSTEMS 3/4.1. 1 SMUTDOWN MARGIN........................................ 3/4 1-1 3/4.1.2 REACTIVITY ANOMALIES................................... 3/4 1-2 3/4. 1. 3 CONTROL RODS Control Rod Operability...................... 3/4 1-3 Control Rod Maximum Scram Insertion Times.... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 1-6 Control Rod Average Scram Insertion Times.............. 3/4 1-7.

Four Control Rod Group Scram Insertion Times........... 3/4 1-8 Control Rod Scram Accumulators..............'........... 3/4 1-9 Control Rod Drive Coupling.............................. 3/4 1-11 Control Rod Position Indication........:.......... 3/4 1-13

- Control Rod Drive Housing Support...... 3/4 1" 15 3/4.1.4 CONTROL ROD PROGRAM CONTROLS Rod Worth Minimizer.................................... 3/4 1-16 Rod Sequence Control System................ 3/4 1-17 Monitor...............'.......................

V Rod Block 3/4 1-18 3/4.1. 5 STANDBY L'I(UID CONTROL SYSTEM.............. 3/4 1-19 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE............. 3/4 2-1 3/4 2.2 APRH SETPOINTSo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 2-5 3/4.2.3 MINIMUM CRITICAL POWER RATIO....................... 3/4 2 6 3/4.2.4 LINEAR HEAT GENERATION RATE............................ 3/4 2-10 Cof FoEL, 3/e 2-FMC. f=oEC. 3/q 2-SUS)UEHANNA - UNIT 2 iv

T INDEX LIST OF FIGURES PAGE

3. 1. 5-1 SODIUM PEHTABORATE SOLUTION TEMPERATURE/

CONCEHTRATION REQUIREMENTS ........................ 3/4 1-21 3.1. 5" 2 SOOIUM PEHTABORATE SOLUTION CONCENTRATION ......,,. 3/4 1-22

3. 2. 1-1 MAXIMUM AVERAGE PLAHAR LINEAR HEAT GENERATION RATE (MAPLHGR) VS. AVERAGE PLANAR EXPOSURE, G6 PIIEL IPPE 8CR383 ( . 8, ) .... 3/I 2-2

3. 2,1-2 MAXIMUM'AVERAGE PLANAR LINEAR HEAT GEHERATION RATE (MAPLHGR) VS. AVERAGE PLANAR EXPOSURE, GE PI/EL IPPE ECR233 (*. 'R * ') . 3/I 2-3 3.2. l-3 fh Rwl'%VIA A~E'RA4K PL,&tJA~ LINEAR HEAT CSEhJEPATlOhl 'RATE (lAAft'h>R) VSa AVE'RAOE l3t3V~LK Pl&)STREP F)0(M 8 3-2 2-.I

- '3.2.$ -la.

3 Z.K-lb FLOWbePeuOEN~

P~

~

L lhlEAQ HFAT cu&IEQAT10N RATS FoR APRhh ~I="TI oI<T~ UER$ cs AvERRcuE PLANAR WPOOVRK i E~N 9xg FoGl aPEemNG LlrnlT (RON Sar~< O.aroW +HO'h)

'5EMIJOELIT hhCPR ~WATlt4h LlfAIT (RSIA CGT < 0.&QN+ QZ9bg RK~WEO FbwER hhCPR OPGRATI M Co LIM IT 28 3.Z.4.2-l

~ ~ LINEAR AE'AT 46NERA71~ RAl'E. (QgPR.) UNII'T VEQSLIS gQ5RAQK KAhlAR E'IIP~~~~ > EMIN,'ON fXg PVK(

3. 4. 1. 1-1 THERMAL POWER LIMITATIONS.........................:

I 3/4 4-1b

3. 4. 6. 1-1 ~ MINIMUMREACTOR VESSEL METAL TEMPERATURE VS.

REACTOR VESSEL PRESSURE .............:............. 3/4 4-18

4. 7. 4-1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIOHAL TEST ..'...... 3/4 7-15 B 3/4 3-1 REACTOR VESSEL MATER LEVEL ........................ 8 3/4 3-8 B 3/4.4.6"1 FAST NEUTRON FLUENCE (E>iMeV) AT 1/4 T AS OF SERVICE LIFE .... ..

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

A'UNCTION

~ -- ~ ~ " ~ ~ B 3/4 4-7

5. l. 1-'1 . EXCLUSION AREA . 5-2

5. l. 2-1 LS POPULATION ZONE ............................... 5"3 5.1.3-1a . MAP DEFINING UNRESTRICTED AREAS FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS ..........:...........

5. l. 3-1b MAP OEFIHING UNRESTRICTED AREAS FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUEHTS ....................... 5-5 ~

6. 2. 1-1 '-

OFF SITE ORGANIZATION 6-3

6. 2. 2-1 UNIT ORGANIZATION ............ -.'.. 6-4 SUSQUEHANNA'- UNIT 2 xxii Amendment Ho. 2

INDEX LIST OF TABLES TABLE PAGE SURVEILLANCE FRE(UENCY NOTATION ...................

1.2 OPERATIONAL CONDITIONS ............................ 3-10 2.2. 1-1 REACTOR PROTECTION SYSTEH INSTRUMENTATION I

S ETPO NTS ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ i ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 2-4

3. 3. 1" 1 REACTOR PROTECTION SYSTEM INSTRUMENTATION ......... 3/4 3-2

3. 3. 1-2 REACTOR PROTECTION SYSTEM RESPONSE TIMES .......... 3/4 3-6

4. 3. 1. 1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE RE(UIREMENTS ......;... 3/4 3-7

3. 3. 2-1 ISOLATION ACTUATION INSTRUMENTATION ............... 3/4 3-11 3.302 2 ISOLATION ACTUATION INSTRUMENTATION.,SETPOINTS ... ~ . 3/4 3-17 3.3 ~ 2 3 ISOLATION: SYSTEM INSTRUMENTATION RESPONSE TIME .... 3/4'-21

4. 3. 2. 1-1 ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE RE(UIREMENTS ........."............;................ 3/4 3-23

3. 3. 3-1 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION ................................... 3/4 3-28 3.3. 3-2 EMERGENCY CORE COOLING SYSTEH ACTUATION INSTRUMENTATION SETPOINTS ......................... 3/4 3"31 3.3. 3-3 EHERGENCY CORE COOLING SYSTEH RESPONSE TIMES ...... 3/4 3-33

4. 3. 3. 1-1 EMERGENCY CORE COOLING SYSTEH ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS ..;...... 3/4 3-34

3. 3.4. 1-1 ASS RECIRCULATION PUMP TRIP SYSTEH INSTRUMENTATION 3/4 3-37

3. 3. 4. 1"2 ATWS RECIRCULATION PUHP TRIP SYSTEM INSTRUMENTATION SETPOINTS ....... 3/4 3-38 SUSQUEHANNA - UNIT 2 XXl 11

'INDEX LIST OF TABLES Continued TABLE PAGE 4.8.1. 1.2-1 DIESEL GENERATOR TEST SCHEDULE .................... 3/4 8-7 4.8.1.1.2-2 UNIT 1 ANO UNIT 2 DIESEL GENERATOR LOADING TIMERS.. 3/4 8-8 4.8.2.1"1 BATTERY SURVEILLANCE REQUIREMENTS ................. 3/4 8-15 3.8.4.1-1 PRIMARY CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES ...."................ 3/4 8-26 3.8.4.2"1 MOTOR OPERATED VALVES THERMAL OVERLOAD PROTECTION ........... 3/4 8-31 3.11.1.1-1 MAXIMUM PERMISSIBLE CONCENTRATION OF DISSOLVED OR ENTRAINED NOBLE GASES RELEASED FROM THE SITE TO UNRESTRICTED AREAS IN LIQUID WASTE ................ 3/4 11-2 4, 11.1. 1. 1-1 RADIOACTIVE LIQUID WASTE SAMPLING ANO ANALYSIS P RO GRAM ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 11-3 4.11.2.1.2-1 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS P ROGRAM ..... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 11-10

3. 12. 1-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ..... 3/4 12-3

3. 12. 1-2 REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES .. 3/4 12"9

4. 12. 1-1 DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS .. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 12-10 B3/4.4.6-1 REACTOR VESSEL TOUGHNESS ........................;. 8 3/4 4-6 5.7.1-1 COMPONENT CYCLIC OR TRANSIENT LIMITS .............. 5-8 6.2.2-1 MINIMUM SHIFT CREW, COMPOSITION .................... 6"5 SUSQUEHANNA - UNIT 2 XXV1

1. 0 OEFINITIONS

'he following terms are defined so that uniform interpretation of these specifications may be achieved. The defined terms appear in capitalized type and shall be applicable throughout these Technical Specifications.

ACTION 1.1 ACTION shall be that part of a Specificatioo which prescribes remedial measures required under designated conditions.

AVERAGE l~t EXPOSURE bundle

l. t The AVERAGE BUNOLE EXPOSURE shall be equal to the suci of the axially averaged exposure of all the fuel rods in the specified bundle divided by the nl aoer of fuel rods in the fuel .

The AVERAGE PLANAR EXPOSURE shall be applicable 'to a specific'lanar height and is equal to the suia of the exposure of all the fuel rods in the specified oundle at the specified height divided by the nuihber of fuel rods in the fuel buncle.

AVERAGE PLANAR LINEAR HEAT GENERATION RATE 1.3 The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) shall be applicable to a specific planar height and is equal to the sum of the LINEAR HEAT GENERATION RATES for all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle.

CHANNEL CALIBRATION le4 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors.

The CHANNEL CALIBRATION shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST. The CHANNEL CALIBRATION may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is calibrated.

CHANNEL CHECK le5 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observati'on. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.

CHANNEL FUNCTIONAL TEST le6 A CHANNEL FUNCTIONAL TEST shall be:

a.. Analog channels - the injection of a'simulated signal into the channel as cl'ose to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions and channel failure trips.

b. Bistable channels - the injection of a simulated signal into the sensor to verify OPERABILITY including alarm and/or trip functions.

The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is tested.

SUS)UEHANNA - UNIT 2

DEFINITIONS FRACTION OF LIMITING POWER DENSITY 1.13 The FRACTION OF LIMITING POWER DENSITY (FLPD) shall be the LHGR existing at a given location divided by the or that bundle type.

• LHQR. sycei6rd in S'@~on 3.Z.Z FRACTION OF RATED THERMAL POWER 1.14 The FRACTION OF RATED THERMAL POWER (FRTP) shall be the measured DERMAL POWER divided by the RATED THERMAL POWER.

FRE UENCY NOTATION 1.15 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table l. l.

GASEOUS RADWASTE TREATMENT SYSTEM

1. 16 A GASEOUS RADWASTE.TREATMENT SYSTEM shall be any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of r educing the total radioactivity prior to release to the environment.

IDENTIFIED LEAKAGE 1.17 IDENTIFIED LEAKAGE shall be:

a. Leakage into collection systems, such as pump seal or valve packing leaks, that is captured and conducted to a collecting tank, or

. b. Leakage'nto the containment atmosphere from sources that are both specifically located and known either not to interfere with the opera-tion of the leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE.

ISOLATION SYSTEM RESPONSE TIME 1.18 The ISOLATION SYSTEM RESPONSE TIME shall be, that time interval'rom when the monitored parameter exceeds its isolation actuation setpoint at the channel sensor until the isolation valves travel to their required positions. Times shall include diesel generator starting and sequence loading delays where applicable. The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

LIMITING CONTROL ROD PATTERN 1.19 A LIMITING CONTROL ROD PATTERN shall be a pattern which results in the

'core being on a thermal hydraulic limit, i.e., operating on a limiting value for APLHGR, LHGR, or MCPR.

LINEAR HEAT GENERATION RATE" 1.20 LINEAR HEAT GENERATION RATE (LHGR) shall be the heat generation per unit length of fuel'od. It is the integral of the heat flux over the heat transfer area associated with the unit length.

SUSQUEHANNA - UNIT 2 1-3

2.1 SAFETY LIMITS BASES

2. 0 INTRODUCTION The fuel cladding, reactor pressure vessel and primary system piping QJ are the principal barriers to the release of radioactive materials to the 5 environs. Safety Limits are astahlished to protect the integrity or these barriers during normal plant operations .and,anticipated transients. The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated. Because fuel damage is not directly observable, a step-back approach is used to establish a Safety Limit such that the MCPR is not less than the limit specified in Specification 2.1.2. MC greater than the specified limit represents a conservative margin relative to the conditions required to maintain fuel cladding integrity. The fuel cladding is one of the physical barriers which separate the radioactive materials from the environs. The integrity of this cladding barrie is related to its relative freedom from perforations or cracking. Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses which occut from reactor operation significantly above design conditions and the Limiting Safety System Settings. While fission product migration from cladding perfor a-tion is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater.

thermal stresses may cause gross rather than incremental cladding deterioration.

Therefore, the fuel cladding Safety Limit- is defined with a margin to the con-ditions which would produce onset of transition boiling; MCPR of 1.0. These conditions represent a significant departure from the condition intended by design for planned operation. Thc /ACPR fuel cloddy'na tnfgriA Gaf~ ti'~g't'ssvres Mcd dv~nn normal ojherattost and during.antlcspatw operatyshnal occvrtkaes, ctt feast'9.vPo ofwd. <typal rcds The use of the ~

Wxl-3 correlation is not valid for all critical power calculations at pressures below 785 psig or core flows less than 10% of rated flow. Therefore, the fuel cladding integrity Safety Limit is established by other means. This is done by establishing a limiting condition on core THERMAL POWER with the following basis. 'ince the pressure drop in the bypass region is essentially all elevtation head, the core pressure drop at low power and flows will always be greater than 4.5 psi. Analyses show that with a bundle. flow of 28 x 10'bs/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow with a 4.5 psi driving head will be greater than 28 x 10'bs/hr. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow fs approximately 3.35 MWt. With the design peaking factors, this corresponds to a THERMAL POWER of more than 50K of RATED THERMAL POWER. Thus, a THERMAL POWER limit of 25K of RATED THERMAL POWER for reactor pressure below 785 psig is conservative.

SVS(UEHANNA - UNIT 2 B'-1 Amendment No. 26

SAFETY LIMITS gpZ~ ~ir~ ~~> S~roe p ~.g BASES

2. 2 THERMAL POWER Hi h Pressure and Hi h Flow e fuel cladding integrity Safety Limit is set such that no me anistic fuel da ge is calculated to occur if the limit is not violated. nce the parameter which result in. fuel damage are not directly observabl during reactor ope ation, the thermal and hydraulic conditions resultin in a departure fr nucleate boiling have been used to mark the beg'ing of the region where 1 damage could occur. Although it is recogni ed that a departure from cleate boiling would not necessarily resul in "damage to BWR fuel rods, the cr ical power at which boiling transition 's calculated to occur has been adop ed as a convenient limit. Ho~ever, e uncertainties in monitoring the core crating state and in the procedur 'sed to calculate the critical power res lt in an uncertainty in the val e of the critical power. Therefore, the el cladding integrity Safet Limit is defined as the CPR in the limiting fuel sembly for which more th 99.9X of the fuel rods in the core are expected t avoid boiling transiti n considering the power distribution within the core nd all uncertainti The Safety Limit MCPR is d ermined usin the General Electric Thermal Analysis Basis, GETAB , which is statistic model that combines all of the uncertainties in operating paramet s and e procedures used to calculate critical power. The probability of e o urrence of boiling transition is determined using the General Electric r tical guality (X) Boiling Length (L),

GEXL, correlation. The GEXL t.orrelati is valid over the range of conditions used in the tests of the d ta ed to develop the correlation.

The required input to the sta ~stical m el are the uncertainties listed in Bases Table B2. l. 2-1 and the minal values of the core parameters listed in Bases Table 82. 1.2-2.

The bases for the uncer inties in the core pa ameters are given in NEDO-20340 and the basis r the uncertainty in the XL correlation is given in NEDO-10958-A . The p er distribution is based on a ypical 764 assembly core in which the rod p ttern was arbitrarily chosen to p oduce a'skewed power distribution having t greatest number of assemblies at tlat highest power levels. The worst d stribution during any fuel cycle would hot be as severe as the distributio used in the analysis.

"General Ele ric BWR Thermal Analysis Bases (GETAB) Data, Correla 'on and Design appl ation,'" NEDO-10958"A.

General ectric "Process Computer Performance Evaluation Accuracy" NEDO-2 40 and Amendment 1, NEDO-20340-1 dated June 1974 and December 197 respe tively.

SUS)UEHANNA - UNIT 2 B 2-2

SAFETY LIMITS h/sd 5~i'0& 9, /. g BASES

2. 1.2 THERMAL POWDER High Pressure and High Flow Onset of transition boiling results in a decrease in heat transfer from the clad and, therefore, elevated clad temperature and the possibility of clad failure. However, the existence of critical po~er, or boiling transition, is not a directly observable parameter in an operating reactor. Therefore, the margin to boiling transition is calculated from plant operating parameters such as core power, core flow, feedwater temperature, and core power distribution. The margin for each fuel assembly is characterized by the critical power ratio (CPR),

which is the ratio of the bundle power which would produce onset of transition boiling divided by the actual bundle power. The minimum value of this ratio for any bundle in the core is the minimum critical power ratio (MCPR).

The Safety Limit MCPR assures sufficient conservatism in the operating MCPR limit that in the event of an anticipated operational occurrence from the limiting condition for operation, at least 99.9Ã of the fuel rods in the core would be expected,to avoid boiling transition. The margin between calculated boiling transition (MCPR = l.'00) and the Safety Limit MCPR is based on a detail-ed statistical procedure. which considers the uncertainties in monitoring the core operating state. One specific uncertainty included in the safety limit is the uncertainty inherent in the XN-3 critical po~er correlation. XN-NF-524 describes the methodology used in determining the Safety Limit MCPR..

I The XX-3 critical power correlation is based on a significant body of practical test data, providing a high degree of assurance that the critical power as evaluated by the correlation is within a small percentage of the actual critical power being estimated. The assumed reactor conditions used in defining the safety limit introduce conservatism into the limit because bounding high radial power factors and bounding flat local peaking distribu-tions are used to estimate the number of rods in boiling transition. Still further conservatism is induced by the tendency of the XN-3 correlation to overpredict the number of rods in boiling transition. These conservatisms and the inherent accuracy of the XN-3 correlation provide a reasonable degree of assurance that'uring sustained operation at the Safety Limit MCPR there would be no transition boiling in the core. Lf boiling transition were to occur, there is reason to believe that the integrity of the fuel would not necessarily be compromised. Significant test data accumulated by the U.S: Nuclear Regulatory Commission and private organizations indicate that the use of a boiling transi-tion limitation to protect against cladding failure is a very conservative approach.

Much of the data indicates that LMR fuel can survive ,or an extended period of time in an environment of boiling transition.

Bases Table B2.1.2-2 NOMINAL VALUES OF PARAMETERS USED IN THE STAT TICAL ANALYSIS OF FUEL CLADDING INTEGRITY SAFETY MIT THERMAL P ER 3323 MW Core Flow 108.5 Mlb/hr Dome Pressure 1010.4 psig Channel Flow Area 0..1089 i' R-Factor High enri ment - 1.043 Medium e ichment - 1.039 Low enr chment - T.030 SUS/UEHANNA - UNIT 2 B 2-4

REACTIVITY CONTROL SYSTEMS 3/4. 1. 2 REACTIVITY ANOMALIES LIMITING CONDITION. FOR OPERATION dilfcrcnce tehrW>~ mon>+or~ Core keg rrnd~c prrdrcfcd <ore, keg 3.1.2 The reactivity shall not exceed 1X delta k/k.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION:

digaexc greX&r With the reactivity than K delta .k/k:

a. Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> perform an analysis to determine and explain the cause of the reactivity difference; operation may continue if the difference is explained and corrected.

b. Otherwise, be in at least HOT SHUTOOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

SURVEILLANCE REOUIREMENTS

.12

+EH&FR-and

• i<<

the predicted shall be verified to be thonlkorcd less than or clorc 4<@.

equal to A delta k/k: ~c e4f

a. Ouring the first'tartup following CORE ALTERATIONS, and 7~ MWD ~~ o4 core evpceorc.

b. At least once per during POWER OP ERATION.

SUS(UEHANNA -UNIT 2 3/4 1-2

3/4. 2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION

@ QE4oel and AveR4aE S~ou exseev~e 4r e~

3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for each type of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limits shown in Figures 3.2.1-1, 3.2. 1-2, and 3.2.1-3."

APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or TEE EE E 0 EE.

ACTION:

With an APLHGR exceeding the limits of Figure 3.2.1-1, 3.2.1-2, or 3.2.1-3, initiate corrective action within 15'inutes and restore APLHGR to within the required limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 25K of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

T SURVEILLANCE RE UIREMENTS 4.2.1 All APLHGRs shall be verified to be equal to or less than the limits determined from Figures 3.2. 1-1,. 3.2.1-2, and 3.2.1"3:

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

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

c. Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with a LIMITING CONTROL R00 PATTERN for APLHGR.

d. The provisions of Specification 4.0.4 are not applicable.

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

SUSQUEHANNA - UNIT 2 3/4 2-1 Amendment No. 26

m'3 Cll C

g 4 12 1000; 12.2 I

eooo; 12.6 10.000:

12.8 16,000;:

12.9 ='0.000; 12.6 0;

200; 11.7 12.0 0.

LLI

~

4I-11 Q LLI gZ 30,M 0:

Ke D+

10 PERMISSIBLE REGION OF OPERATI 9

6,000 10.000 16.000 20.000 26.000 30.000 AVERAGE PLANAR EXPOSURE IMWdlt)

FIGURE 3.2.1-1 HAXIHUH AVERAGE PLANAR LINEAR HEAT GENERATION RATE (HAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE INlTIAL CORE FUEL TYPE 8CR183 (LOW ENRICHHENT) .

13 1102; 22,046;:

12.2 023.'...:. 16,535;:..: ..:... 12.6 ~

~ ~ ~

~

11 I

~ 5512;

.:::: 12.8 g)

I 12 ..:...... 12.6 220; 27,558; 11.7

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MAXIMUMAVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE GE FUEL TYPES SCR183 (1.83/o ENRICHED)

FIGURE 3.2.1-1

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FIGURE 3.2.1-2 HAXIHUH AVERAGE PLANAR LINEAR HEAT GENERATION RATE (HAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE INITIAL CORE FUEL TYPE 8CR233 (HEDIUH ENRICHHENT)

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AVERAGE PLANAR LINEAR HEAT 'AXIMUM GENERATION RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE GE FUEL TYPES 8CR233 {2.33% ENRICHED)

FIGURE 3.2.1-2

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MAXIMUMAVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE BUNDLE EXPOSURE EXXON 9X9 FUEL FIGURE 3.2.1-3

'4 POWER DISTRIBUTION LIMITS 3/4.2. 2 APRM SETPOINTS LIMITING CONDITION FOR OPERATION 3.2.2 The APRM flow biased simulated thermal power-upscale scram trip setpoint (S) and flow biased neutron flux-upscale control rod block trip setpoint (SRB) shall be established .according to the following relationships:

Tri Set oint Allowable Valu'e S < 0. 8W + 59K)T S < 0. BW + 6ZX)T

< (0.58W + 50K)T < (0.58W + 53K)T SRB SRB where: S and S are in percent of RATED THERMAL POWER, W = LooIIB recircu1ation flow as a percentage of the loop recirculation flow which produces a rated core flow of 100 million lbs/hr, T ~ Lowest value of the ratio of FRACTION OF RATED THERMAL POWER divided by the, MAXIMUM FRACTION OF LIMITING POWER DENSITY.

FR4CTION CIP LINIITIIhIIP POWER 'bEN5ITQ (FLPD) %r 66 WLI+l IS Mherc'.,'The.

adLAal LulEAR HEATGEPIPFATIOAl 'FREE (I.Hcow') dividf.'d IcbtII <3. +

b Peg Thc Spec'14C&O n 3.2. 9.

FLPD4br Fecoo I,

fuel C nd I's Wc. ac&al LH&'R BI'sided bg AE'4T Q&IKRATIoA4 'RPTR 5om Fi'pic Wd'INEAR s.2.2-l.

T l5 aliAIayt less "lhasa oI duct.( <cp [. o APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or f IATEII YhEIN 0 ER.

ACTION:

With the APRM flow biased simulated thermal power upscale scram trip setpoint and/or the flow biased neutron flux-upscale control rod block, trip setpoint less conservative than the value shown in the Allowable Value column for S ar S>>, as above determined, initiate corrective action within 15 minutes and aU)ust S and/or SR to be consistent with the Trip Setpoint value", within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce IIHERMAL POWER to less than 25K of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

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

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

b. Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at 1east 15K of RATED THERMAL POWER,'and

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

d. The provisions of Specification 4.0.4 are not applicable.

'With MFLPD greater than the FRTP during power ascension up to 90K of RATED THERMAL POWER, rather than adjusting the APRM setpoints, the APRM gain may be adjusted such that APRM readings are greater than or equal to 100K times MFLPD, provided that the" adjusted APRM reading does not exceed 100K of RATED THERMAL POWER, the required gain adjustment increment does not exceed 10K of RATED THERMAL POWER, and a notice of the adjustment is posted on the reactor control panel.

SUS(UEHANNA UNIT 2 '/4 See Specification 3.4.1.1.2.a for single loop operation requirements.

- 2.-5 Amendment No. 26

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LINEAR HEAT GENERATION RATE FOR APRIVI SETPOINTS VERSUS AVERAGE PLANAR EXPOSURE EXXON SXS FUEL FIGURE 3.2.2-1

POWER DISTRIBUTION LIHITS 3/4.2.3 MINIHUH CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION

.2.3 The HINIMUM CRITICAL POWER RATIO (MCPR) shall be equa o or greater n the RCPR limit determined fram Figure 3.2.3-1a or Fig 3.2.3-lh, as icable, times the Kf shown in Figure 3.2.3-2, provide hat the end-of-cycle re culation pump trip (EOC-RPT) system is OPERABLE per pecification 3.3.4.2 and turbine bypass system is OPERABLE per Specific on 3.7.8,.with:

ave - B A B

= 0.86 seconds, control average scram insertion time limit to notch 39 pe ecification 3. 1.3.3, 688 + 1.65 { )~(0.052),

N n

E i=1 ave n N.

where:

n number surveil lance tests p ormed to date in cycle, Ni = numb of, active control rods mea ed in the i surveillance tests, measured in the i.

-th ti = av s

age scram time eillance test, to notch and 39 of al ods N> tal number of acti've rods measured in cification 4.1.3.2.a.

APPL BILITY:

OP TIONAL CONDITION 1, when THERMAL POWER is greater or equal to 25X of ED THERMAL POWER.

SUS)UEHANNA - UNIT 2 3/4 2-6 Amendmen o. 10

POWER OISTRIBUTION LIMITS LIMITING CONOITION FOR OPERATION Continued ACTION:

a. With the end-of-cycle recirculation pump trip system i able per Specification 3.3.4.2, operation may continue and the visions of Specification 3. 0.4 are not applicable provided tha ithin 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, MCPR is determined to be greater than or equal to e MCPR limit as a function of average scram time as shown in Figu .2.3-.1a or .

Figure 3.'2".3-1b, as applicable, EOC-RPT inope e curve, times the Kf sh~n in Figure 3-2.3-2

b. ith the turbine bypass system inoperab er Specification 3.7.8, o ~ation may continue and the provisi of Specification 3.0.4 are no pplicable provided that within ur, MCPR is determined to be gre than or equal to the MCPR t as a function of average scram time shown in .Figure 3.2.3-1a Figure 3.2.3-1b, as applicable, turbin pass inoperable curv imes the Kf sho~n in Figure 3.2.3-2.

c. With MCPR s than the app able MCPR limit determined from Figure 3.2. a or Figure 2.3-lb, as applicable, and Figure 3.2.3-2, initiate cor ive act'ithin 15 minutes and restore MCPR to within the req ed li within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to

,A~

~ less than 25K of T THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

~

SURVEILLANCE RE UIREMEHTS 4.2.3 MCPR, with a t = 1.0 for the pr'o e performanc in accordance f the initial scram time measurements Specification 4.1.3.2, or

b. x as d ned in Specification 3. used to determine the limit withi 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of the conclusion each scram time surveillance tes equired by Specification 4.1.

c T provisions of Specification 4.0.4 a ot applicable.

shall be termined to be equal to or greater than th pplicable MCPR limit determi d from Figure 3.2.3-1a or Figure 3.2.3-lb; as licable, and Figure 3.2. 3-At.,least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,

b. Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER rease of at

'm least 15% of RATEO THERMAL POWER, and Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reacto operating with a LIMITING CONTROL ROO PATTERN FOR MCPR.

The pAvisions of Specification 4.0.4 are not applicable.

SUSQUEHANNA - UNIT 2 3/4 2-7 Amendment No.10

POWER DISTRIBUTION LIMITS 3/4. 2.3 MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION 3.2.3 The MIHIMUM CRITICAL POWER RATIO (MCPR) shall bey

~c ~c ~al~s def~rmif,& fro~ Fibre 3 K3 of F (porc. 3.Z. 3>>lb, ~ c asap)(faye ~ F ~ ~

APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than T EIUS ER.

of'ATlB

~hei ACTION:

4t7WC )

With less than the applicable MCPR limit determined MCPR

• I I .i I la ~ \

, to within the required limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or'reduce THERMAL POWER to less than 25" of RATED THERMAL POWER wi.hin the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE REOUIREMEHTS 4.2.3. 1 MCPR shall be determined to be greater than or to the applicable MCPR limit determined

a. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,'qual from Figurc 8,P.s-la. '3.2.s -)b >And FIgorc,3.2.3-Z:

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

c. Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with a LIMITIHG CONTROL ROD PATTERN for MCPR.

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1.7 E 16 C 1.6 Figures to be supplied later 1.4 tX:

O 1.3 1.2 40 60 60 70 80 90 100 Total Core Flow (% OF. RATED)

FLOW DEPENDENT MCPR OPERATING LIMIT (RBM SET AT < 0.66W + 40%)

FIGURE 3.2.3-1a 1.7

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FIGURE 3.2.3- tb

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1.3 1.2 20 30 40 60 60 70 80 80 100 Core Power {/o OF RATED)

REDUCED POWER MCPR OPERATING LIMIT Figure-3.2.3-2

POWER DISTRIBUTION LIMITS 3/4.2.4 LINEAR HEAT GENERATION RATE MITING CONDITION FOR OPERATION 3.2.4 The LINEAR HEAT GENERATION RATE (LHGR) shall not exceed 13. kl/ft.

APPLICAB ITY: OPERATIONAL CONDITION 1, when THERMAL POWER is g ater than or equal to of RATED THERMAL POWER.

ACTION:

With the LHGR o any fuel rod exceeding the'limit, initiat corrective action within 15 minutes nd restore the LHGR to within the limi within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.,or reduce THERMAL POW to less than 25K of RATED THERMAL P ER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE RE UIREMENTS 4.2.4 LHGRs shall be determin to be equal or less than the limit:

a. At least once per 24 ho s,

b. Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after comp tio of a THERMAL POWER increase of at least 15K of RATED THERMAL P , and

c. Initially and at least once e 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> when the reactor is operating on a LIMITING CONTROL ROD TTER for LHGR.

d. The provisions of Specif cation 4.0. are not applicable.

g4 gA

,SUSQUEHANNA - UNIT 2 3/4 2-10

PE-trJIPJTH POWER DISTRIBUTION LIMITS P~( e-Z 3/4.2.4 LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.4.1 The LINEAR HEAT GENERATION RATE (LHGR) for GE fuel shall not exceed

'13.4 kw/ft.

APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or f IATKD IIIEINL K .

ACTION:

With the LHGR of any fuel rod exceeding the limit, initiate corrective action within 15 minutes and restore the LHGR to within the limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 25K of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

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

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

b. Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at least 15'f RATED THERMAL POWER, and C. Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating on a LIMITING CONTROL R00 PATTERN for LHGR.

d. The provisions of Specification 4.0.4 are not applicable.

3/4 2-10

POWER DISTRIBUTION LIMITS

~

3/4.2.4

~ ~ LINEAR HEAT GENERATION RATE ENC FUEL LIMITING CONDITION FOR OPERATION 3.2.4.2 The LINEAR HEAT GENERATION RATE (LHGR) for ENC fuel shall 'not exceed the LHGR limit determined from Figure 3.2.4.2-1.

APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL is greater than or I 'I POWER IIETEE TEEEIE I Ell.

ACTION' With the LHGR of any fuel rod exceeding the limit, initiate corrective action within 15 minutes and restore the LHGR to within the limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 25K of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE RE UIREMENTS 4.2.4.2 LHGRs for ENC fuel shall be determined to be equal to or less than the limit:

At 1 east once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion'of a THERMAL POWER increase of. at least 15K of RATED THERMAL POWER, and Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating on a LIMITING CONTROL ROD PATTERN for LHGR.

d. The provisions of Specification 4.0.4 are not applicable.

3/4 2-10a

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

INSTRUMENTATION END-OF"CYCLE RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.4.2 The end-of"cycle recirculation pump trip (EOC-RPT} system instrumentation channels shown in Table 3.3.4.2-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.4.2-2 and with the END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEH RESPONSE TIME as shown in Table 3.3.4.2-3.

APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or I IIEET IIIIIII "Ill ACTION:

With an end-of-cycle recirculation pump trip system instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.4.2-2, declare the channel inoperable until the channel is restored to OPERABLE status with the channel setpoint adjusted consistent with the Trip Setpoint value.

With the number of OPERABLE channels one less than required by the Minimum OPERABLE Channels per Trip System requirement for one or both trip systems, place the inoperable channel(s} in the tripped condition within one hour.

co With the number of OPERABLE channels two or more less than required by the Minimum OPERABLE Channels per Trip System requirement for one trip system and:

1. If. the inoperable channels consist of one turbine control valve channel and one turbine stop valve channel, place both inoperable channels in the tripped condition within one hour.

2. If the inoperable channels include two turbine control valve channels or two turbine stop valve channels, declare the trip system inoperable.

d. With one trip system inoperable, restore the inoperable trip system to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or evaluate MCPR to be equal to or greater than the applicable MCPR limit without EOC-RPT within 1 hour" or take the ACTION required by Specification 3.2.3.

e. With both trip systems inoperable, restore at least one trip system to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or evaluate MCPR to be equal to or greater than the applicable MCPR limit without EOC-RPT within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />" or take the ACTION required by Specification 3.2.3.

"If MCPR 1s evaluated to be equal to or greater than the applicable MCPR limit without EOC-RPT within 1 hour, operation may continue and the pro-visions 'of Specification 3.0.4 are not applicable.

SUSQUEHANNA - UNIT 2 3/4 3-40

REACTOR COOLANT SYSTEM RECIRCULATION LOOPS -.SINGLE LOOP OPERATION LIMITING CONDITION FOR OPERATION 3.4.1.1.2 One reactor coolant recirculation loop shall be in operation with the pump speed < 90K of the rated pump speed, and aO the following revised specification limits shall be followed:

1. Specification 2. 1.2: the MCPR Safety Limit shall be increased to 1.07.

2. 'able 2.2. 1-1: the APRM Flow-Biased Scram Trip Setpoints shall be as follows:

Tri Set oint Allowable Value

< 0.58W + 5 '5R.

3. Specification 3.2. 1: The MAPLHGR limits shall be the limits specified in Figures 3.2. 1-1, 3.2. 1-2, and 3.2. 1-3, multiplied bye e. 0-Specification 3.2.2: the APRM Setpoints shall be as follows:

Tri SRB Set oint

.5 55%lT

< (0.58W + 46K)T

~NtT Allowable Value SRB

< (0.58W + 49%)T

5. Table 3.3.6-2: the RBM/APRM Control Rod Block Setpoints shall be as follows:

a. RBM - Upscale Trio Set oint Allowable Value

l. , < 0.66W + 35

2. < 0.66W + 37K < 0.66W + 40K 5.a. 1 and 5.a.2 shall be used in conjunction with the MCPR limits specified in Figures 3.2.3-1a and 3.2.3-Ib, respectively.

b. APRM-Flow Biased Tri Set oint Allowable Value

<0.5 +46

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

c. Total core flow shall be greater than or equal to 42 million lbs/hr when THERMAL POWER is greater than the limit specified in Figure 3.4. 1. l.l-l.

APPLICABILITY: OPERATIONAL CONDITIONS 1" and 2", except during two loop operation.8 ACTION:

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

SUSQUEHANNA - UNIT 2 3/4 4-lc Amendment No. 26

0 PLANT SYSTEHS 3I4.7.8 HAIN TURBINE BYPASS SYSTEH LIMITING CONDITION FOR OPERATION 3.7.8 The main turbine bypass system shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITION l.

e UPE.UI <+

ACTION: With the main turbine bypass systempnoperab1e, restore tha system to opERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or sbmme Nm NCPR to be sepal to or greater than the applicable HCPR limit without bypass within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or take the ACTION required by Specification 3.2.3.

SURVEILLANCE REQUIREMENTS 4.7.8 The main turbine bypass system shall be demonstrated OPERABLE at least once per:

a. 7 days by cycling each turbine 'bypass valve through at least one complete cycle of full travel, and

b. 18 months by:

1. Performing a system functional test which includes simulated automatic actuation and ver'ifying that each automatic valve actuates to its correct'position.

2. Demonstrating TURBINE BYPASS SYSTEM RESPONSE TIME to be less than or equal to 0.30 second.

+ }p phC.~ is evalvatrd icB be. egvc( Q or grex+er than whw MO~

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SUS(UEHANNA - UNIT 2 3/4 7-30

C 4

3/4. 1 REACTIVITY CONTROL SYSTEMS BASES WV l

3/4. 1. 1 SHUTOOWN MARGIN fv s 3L A sufficient SHUTOOWN MARGIN ensures that 1) the reactor can be made 3 cf subcritical from all operating conditions, 2) the reactivity transients '5, g 5

associated with postulated accident conditions are controllable within acceptable limits, and 3),the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition. 0 Since core reactivity values will vary through core life as a function of c/f. l-2 EAcTIUIT 0

- Since the'HUTOOHN MARGIN requirement is small, a careful check on actual reactor conditions compared to the predicted conditions is necessa~. Any changes in reactivity from that of the predicted (predicted cyre k ff) can be determined from the core monitoring system (monitored core k ff). In the absence of any deviation in plant operating conditions or reactivity anomaly, these values should be essentially equal since the calcuIational methodologies are consistent.

The predicted core k eff is calculated by a 30 core simulation code as a function ff of cycle exposure. This is performed for projected or anticipated reactor operat-ing states/conditions throughout the cycle and is usually done prior to cycle operation.- The monitored core keff ff is the kefff as calculated by the core monitor ing system for actual plant conditions.

Since the comparisons are easily done, frequent checks are not an imposition on normal operation. A lX deviation in reactivity from that of the predicted is larger than expected for normal operation, and therefore should be throughly evaluated. A deviation as large as 3Z would not exceed the design conditions of the reactor.

REACTIVITY CONTROL SYSTEMS BASES 3/4. l. 3 CONTROL ROOS The specification of this section ensure that (1) the minimum SHUTOOWN MARGIN is maintained, (2) the control rod insertion times are consistent with those used in the accident analysis, and-(3) limit the potential effects of the rod drop accident. The ACTION statements permit variations from the basic re-quirements but at the same time impose more restrictive criteria for continued operation. A limitation on inoperable rods is set such that the resultant effect on total rod worth and scram shape will be kept to a minimum. The re-quirements for the various scram time measurements ensure that any indication of systematic problems with rod drives will be investigated on a timely basis.

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

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

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

The control rod system is designed to bring the reactor subcritical at a rate fast enough to prevent the MCPR from becoming less than the 1'imit speci-fied in Specification 2. 1.2 during the co~~ ~id'. transient analyzed in +he.

cycle silicic. traneen< ana4sis io ~ This analysis shows .that the negative reactivity rates resuming From the scram with the average response of a11 the drives as given in the specifications, provide the required protection and MCPR remains greater than the limit specified in Specification-2.1.2. The occurrence of scram times longer then those specified should be viewed as an indication 'of a systematic problem with the rod drives and therefore the surveillance interval is reduced in order to prevent operation of the reactor for long periods of time with a potentially serious problem.

The scram discharge volume is required to be OPERABLE so that it will be available when needed to .accept discharge water from the control rods during a reactor scram and will isolate the reactor coolant system from the containment when required.

Control rods with inoperable accumulators. are declared inoperable and Spe-cification 3. 1.3. 1 then applies. This prevents a pattern of inoperable accumu-lators that would result in less reactivity insertion on a scram than has been analyzed even though control rods 'with inoperable accumulators may still be in-serted with normal drive water pressure. Operability of the accumulator most ensures that there is a means available to insert the control rods even under the unfavorable depressurization of the reactor.

SUS(UEHANNA - UNIT 2 B 3/4 1-2 Amendment No. 26

ll REACTIVITY CONTROL SYSTEMS BASES CONTROL RODS (Continued)

Control rod coupling integrity is required to ensure compliance with the analysis of the rod drop accident in the FSAR. The overtravel position feature provides the only positive means of determining that a rod is properly coupled and therefore this check must be performed prior to achieving criticality after completing CORE ALTERATIONS that could have affected the control rod coupling integrity. The subsequent check is performed as a backup to the initial demonstration.

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

The control rod housing support restricts the outward movement of a control rod to less than 3 inches in the event of a housing failure. The amount of rod reactivity which could be added by this small amount of rod withdrawal is less than a normal withdrawal increment and will not contribute to any damage to the primary coolant system. The support is not required when there is no pressure to act as a driving force to rapidly eject a drive housing.

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

3/4. 1.4 CONTROL ROD PROGRAM CONTROLS Control rod withdrawal and insertion sequences are established to assure that the maximum insequence individual control rod or control rod segments which are withdrawn at any time during the fuel cycle could not be worth enough to .

result in a peak fuel enthalpy greater than 280 cal/gm in the event of a control rod drop accident. The specified sequences are characterized by homogeneous, scattered patterns of control rod withdrawal. When THERMAL POWER is greater than 2P~ of RATED THERMAL POWER, there is no possible rod worth which, if dropped at the design rate of the velocity limiter, could result in a peak enthalpy of 280 cal/gm. Thus requiring the RSCS and RWM to be OPERABLE when THElUQL POWER is less than or equal to 20K of RATED THERMAL POWER provides adequate control.

The RSCS and RWM provide automatic supervision to assure that out-of-sequence. rods will not be withdrawn or inserted. pgRP>~P l~ f0 The ana s accident is presen of the FSAR and the techniques of th d in a topical report, Refere pp ements, References '2 and 3.

The RBM is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power operation. Two channels are provided. Tripping one of the channels will block erroneous rod withdrawal soon enough to prevent fuel damage. This system backs up the written sequence used by the operator for withdrawal of control rods.

SUSQUEHANNA - UNIT 2 B 3/4 1-3

Parametric Control Rod Orop Accident analyses have sho~n that for a wide range of keg reactor parameters (which envelope the operating ranges of these variables), the fuel enthalpy rise during a postulated control rod drop acci-dent remains considerably lower than the 280 cal/gm limit. For each operating cycle, cycle-specific parameters such as maximum control rod worth, Ooppler coefficient, effective delayed neutron fraction, and maximum four bundle local peaking factor are compared with the inputs to the parametric analyses to deter mine. the peak fuel rod enthalpy rise. This value is then compared against the 280 cal/gm design limit to demonstrate compliance for each operating cycle. If cycle-specific values of the above parameters are outside the range assumed in the parametric analyses, an extension of the analysis or a cycle-specific analysis may be required. Conservatism present in the analysis, results of the parametric studies, and a detailed description of the methodology for per-forming the Control Rod Orop Accident analysis are provided in NN-NF-80-19 Volume l.

0 REACTIVITY CONTROL SYSTEMS BASES 3/4. 1.5 STANDBY LI UID CONTROL SYSTEM The standby liquid control system provides a backup capability fear.

bringing the reactor from full power to a cold, Xenon-free shutdown, assuming that none of the withdrawn control rods can be inserted. To meet this objective it is necessary to inject a quantity of boron which produces a concentration of 660 ppm in the reactor core in approximately 90 to 120 minutes. A minimum quantity of 4587 gallons of sodium pentaborate solution containing a minimum of 5500 lbs. of sodium pentaborate is required to meet this shutdown require-ment. There is an additional allowance of 165 ppm in the reactor core to account for imperfect mixing. The time requirement was selected to override the reactivity insertion rate due to cooldown following the Xenon poison peak and the required pumping rate is 41. 2 gpm. The minimum storage volume of the solution is established to allow for the portion below the pump suction that cannot be inserted and the filling of other piping systems connected to the reactor vessel. The temperature requitement for the sodium penetrate solution is necessary to ensure that the sodium penetaborate remains in solution.

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

Surveillance requirements are established on a frequency that assures a high reliability of .the system. Once the solution is established; boron con-centration will not vary unless more boron or water is added, thus a check on the temperature and volume once each 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> assumes that the solution is available for use.

Replacement of the explosive charges in the valves at regular intervals will assure that these valves will not fail because of deterioration of the charges.

R. C. Stirn and 'J. A. Woolley, "Rod Drop Accide ysis for Large BWR s, Topical Report NEDO-10527 2 C. J. Paone, R. C.

1972 Stirn and . o, upplement 1 to NED0-10527, July J. M. Haun, C. one and R. C. Stirn, Addendum 2,. ed Cores,"

Sup 1 to NED0-10527, January 1973 SUSQUEHANNA -, UNIT 2 B 3/4 1-4

3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section assure that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the 2200'F limit specified in 10 CFR 50.46.

3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE This specification assures that the peak cladding temperature following the postulated design basis lass-of-coolant accident will not exceed the limit specified in 10 CFR 50.46.

The peak cladding temperature (PCT) following a postulated loss-of-coolan a ident is primarily a function of the average heat generation rate of all the ds of a fuel assembly at any axial location and is dependent only second ily on the rod to rod power distribution within an assembly. T peak clad temp ature is calculated assuming a LHGR for the highest power rod which is equal to r less than the design LHGR corrected for densificati . This LHGR times 1. is used in the heatup code along with the expos dependent steady state gap onductance and rod-to"rod local peaking fac r. The Technical Specification AVER PLANAR LINEAR HEAT GENERATION RATE ( HGR) is this LHGR of the highest powere rod divided by its local peaking ctor. The limiting value for APLHGR is sho in Figures 3.2. 1-1. 3.2. 1-2 d 3.2. 1"3.'he calculational proce re used to establis he APLHGR shown on Figures

3. 2. 1"1, 3. 2. 1-2 and 3.2. 1-3 i ased. on a loss f-coolant accident analysis.

The analysis was performed using neral Ele ic (GE) calculational models which are consistent with the requi ents f Appendix K to 10 CFR 50. A

'omplete discussion of each code emplo in the analysis is presented in Reference 1. Differences in this ana s compared to previous analyses can be broken down as follows.

In ut Chan es

1. . Corrected Vapor tion Calculation - Coeff ients in the vaporization correlation u d in the REFLOOD code were co cted.

2. Incorpor d more accurate bypass areas - The byp areas in the top gu' were recalculated using a more accurate te nique.

3. C rected guide tube thermal resistance.

4. Correct heat capacity of reactor internals heat nodes.

~ REPLACE mi~H

- 3/4"2-1 SEcAo w 3/g SUSQUEHANNA UNIT 2 8

The peak cladding temperature (PCT) following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is dependent only secondarily on the rod to rod power distribution within an assembly. For GE fuel, the peak clad temperature is calculated assuming a LHGR for the highest powered rod which is equal to or less than the design LHGR corrected for densification. This LHGR times 1.02 is used in the heatup code along with the exposure dependent steady state gap conductance and rod-to-rod local peaking factor. The Technical Speci-fication AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) for GE fuel is this LHGR of the highest powered rod divided by its local peaking factor which results in a calculated LOCA PCT much less than 22004F. The Technical Specification APLHGR for Exxon fuel is specified to assure the PCT following a postulated LOCA will not exceed the 22004F limit. The limiting value for APLHGR is shown in Figures 3.2.1-1,~ 3.2.1-2>and 3.'Z. l-3.

The calculational procedure used to establish the APLHGR shown on Figures 9-z.l-l 9.z.)-p,a< s.z.l3is based on a loss-of-coolant accident analysis. The analysis was performed using calculational models which are consistent with the requirements of Appendix K to 10 CFR 50. These models are described in Reference 1 or XN-NF-80-19, Volumes 2, 2A,. 2B and 2C.

POWER DISTRIBUTION LIMITS'ASES AVERAGE PLANAR LINEAR HEAT GENERATION RATE (Continued)

1. Core CCFL pressure differential - 1 psi - Incorporate assumption t flow from the bypass to lower plenum must over e a 1 psi pre re drop in core.

2. Incorporat RC pressure transfer assumpt' The assumption used in the SAFE-REFL pressure transfer whe he pressure is increasing was changed.

A few of the changes affect dent calculation irrespective of CCFL.

These changes are listed below.

iver,

.. 1. Break Areas he DBA break area was ca lated more accurately.

b. Model Cha

1. mproved Radiation and Conduction Calculation - Inco ration of CHASTE 05 for heatup calculation.

A list of the significant plant input parameters to the loss-of-coo t accident analysis is presented in Bases Table 8 3.2. i-l.

3/4. 2. 2 APRM SETPOINTS The fuel cladding integrity Safety Limits of Specification 2 ased on distribution which would yield the design LHG D THERMAL POWER.

The flow bias lated thermal power-upscale etting and flow biased simulated thermal powe ale control ock functions of the APfN instru-ments must be adjusted to ensu he MCPR does not become less than 1.06 or that > 2X plastic str es not in the degraded situation. The scram settings and rod settings are adjuste 'cordance with the formula in this speci on when the combination of THERMAL and MFLPD indicates a hi aked power distribution to ensure that an LHGR tr 'ould not be increased in the degraded condition.

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SUS(UEHANNA - UNIT 2 B 3/4 2-2

The flow biased simulated thermal power-upscale scram setting and flow biased simulated thermal power-upscale control rod block functions of the APRH instru-ments limit plant operations to the region covered by the transient and accident analyses. 'In addftfon, the APRH setpoints must be adjusted to ensure that

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

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Sases Table S 3.2.1-1 SIGNIFICANT INPUT PARAMETERS TO THE LOSS-OF-COOLANT ACCIOENT ANALYSIS Plant Parameter .

Core THERMAL HER ...................... 3439 Mwt" w ch corresponds to 105K of ated steam flow Vessel Steam Outpu ...............'..... 6

14. 15 x 0 ibm/hr which cor-respon to 105K of rated steam flow Vessel Steam Oome Pressu .............. 10 psia Oesign Basis Recirculation L Area for:

ne'reak

a. Large Breaks 4.153 ft N

b. Small Breaks 1.0 ft 0.02 ft Fuel Parameters:

PEAK CHNICAL, INITIAL SPEC -ICATIQN OESIGN MINIMUM.

L EAR HEAT XIAL CRITICAL FUEL BVNOLE G ERATION RATE P KING POMER FUEL TYPE GEOMETRY (kw/ft) FA TOR RATIO Initial Core 8 x 8 13. 4 1.4 1. 18 A more detailed listing f input of'ach model and its sourc is presented in Section II of Refer nce 1 and Section 6.3 of the FSAR.

"This power level m ts the Appendix requirement of 102 . The co e heatup calculatio assumes a bundle power consistent with operati of the highest pow HEAT GENERATIO ed RATE rod at 1 imi t.

10'f its Technical Specification LIN SUSQUEHANNA UNIT 2 B 3/4 2-3

POWER OISTRIBUTION LIMITS BASES 3/4. 2. 3 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRs at steady state operating conditions as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safety Limit MCPR, and an analysis of abnormal operational transients.

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

tion 2.2.

To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting transients have been analyzed to determine which result in the largest reduction in CRITICAL POWER RATIO (CPR). The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion,, and coolant tem-perature decrease. The limiting transient yields the largest delta MCPR. When added to the Safety Limit MCPR, the required minimum operating limit MCPR of Specification 3.2.3 is obtained and presented in Figures3.2.3-la~aa4-3.2.3-lb,and 3 23-2.

When the less operationally limiting Rod Block Monitoring trip setpoint

(.66W + 42K from Table 3.3.6-2) fs used, the more limiting MCPR curve Figure 3.2.3"1b is applicable due to a larger delta MCPR from the~A%~ Rod With-drawal Error (RWE) transient. Figure 3.2.3-la is applicable when the Rod Block Monitor trip setpoint (.66W + 40K from Table 3.3.6-2) is used.

e evaluation of a given transient begins with the system initial parameters sh in FSAR Table 15.0-2 that are input to a GE-core dynamic behavior tra sient mputer program. The code used to. evaluate pressurization events 15 described NEOO-24154 (3) and the program used in nonpressurizatio vents is described in 00-10802 (2) . The outputs of this program alo with the initial MCPR form input for further analyses of th'e the ly limiting bundle with the singl hannel transient thermal hydrauI TASC code described in NEOE-25149 (4) . The pri al result of this eva tion is the reduction in HCPR caused by the transie The purpose of the Kf factor of Figu 3. -2 is to define operating limits at other than rated core flow conditi s. t less than 100K of rated flow the required MCPR is the product of t CPR an e K factor. The K factors assure that the Safety Limit R will not be Vio ted during a flow increase transient resulting from otor-generator speed con 1 failure. The Kf factors may be applied to bo manual and automatic flow control des.

The K factor ues shown Tn Figure 3.2.3-2 were developed gen cally and are a pile e to all BWR/2, BWR/3 and BWR/4 reactors. The K fac s were derived Tng the flow control line corresponding to RATBO THERMAL PO at rate core flow..

(.@p~ ~/ SEcY(0% 3//+. 2-'!NW<7 SUSQUEHANNA - UNIT 2 B 3/4 2-4 Amendment No. 26

The evaluation of a given transient begins with the system initial param-eters shown in the cycle specific transient analysis report that are input to a Exxon-core dynamic behavior transient computer program. The outputs of this program along with the initial HCl'R form the input for further analyses of the thermally limiting bundle. The codes and methodology to evaluate re riza-tion and non-pressurization events are-described in XN-HF-79-71. The princspal result of this evaluation is the reduction in HCPR caused. by the transient.

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POWER OISTRIBUTION LIMITS BASES MINIMUM CRITICAL POWER RATIO (Continued) or the manual flow control mode, the Kf factors were calculate ch that for the imum flow rate, as limited by the pump scoop tube se >nt and the correspon THERMAL POWER along the rated flow contr >ne, the limiting bundle's relative er was adjusted until the MCPR c es with different core flows. The ratio the MCPR calculated at 1ven point of core flow, divided by the operating lim'PR, determi the Kf.

For operation in. the automati control mode, the same procedure was employed except the initial distribu was established such that the MCPR was equal to the op ing limit MCPR at THERMAL POWER and rated flow.

The Kf fa s shown in Figure 3.2.3-2 are conservator oi the General Electri ant operation because the operating limit MCPRs of Spe ti .2.3 are greater than the original 1.20 operating limit MCPR use r

'a-he generic derivation of Kf.

At THERMAL POWER levels less than or equal to 25K of RATED THERMAL POWER, the reactor will be operating at minimum recirculation. pump speed and the moderator void content will be very small. For all designated control rod patterns which may be employed at this point, operating plant experience indi-cates that the resulting MCPR value is in excess of requirements by a considerable margin. During initial start-up testing of the plant, a MCPR evaluation will be made at 25K of RATED THERMAL POWER level with minimum recirculation pump speed. The MCPR margin will thus be demonstrated such that future MCPR evaluation below this power level will be shown to be unnecessary.

The daily requirement for calculating MCPR when THERMAL POWER is greater than or equal to 25K of RATED THERMAL POWER is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes. The~requirement for calculating MCPR when a limiting control rod pattern is approached ensures that MCPR will be known following a change in THERMAL POWER or power shape, regardless of magnitude, that could place operation at a thermal limit.

3/4.2.4 LINEAR HEAT GENERATION RATE This specification assures that the Linear Heat Generation Rate (LHGR) in any rod is less than the design linear heat generation even if fuel pellet densification is postulated.

References:

1. Gener'al. Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K, NEOE-20566, November 1975.

SUS(UEHANNA - UNIT 2 8 3/4 2-5 Amendment No. 10

3/4. 4 REACTOR COOLANT SYSTEM BASES 3/4. 4. 1 RECIRCULATION SYSTEM Operatfon with one reactor recirculation loop inoperable has been evaluated and found acceptable, provided that the unit fs operated fn accordance with Specification 3.4.1.1.2.

For single loop operation, the MAPLHGR limits 'are multiplied by a factor of O.ck

~1k'.

. This mulb plicatron pre'ctvctes ectcndH operation cviTh one loop au+ o4 service ~

For single loop operation, the RBM and APRM setpoints are adjusted by a 7X decrease in recirculation drive flow to account for the active loop drive flow that bypasses the core and goes up through the inactive loop jet pumps.

Surveillance on the pump speed of the operating recirculation loop is imposed to exclude the possibility of excessive reactar vessel internals vibration.

Surveillance on differential temperatures below the threshold limits of THERMAL POWER or recirculation loop flow mitigates undue thermal stress on vessel nozzles, recirculation pumps and the vessel bottom head during extended opera-tion in the single loop mode. The threshold limits are those values which will sweep up the cold water from the vessel bottom head.

THERMAL POWER, core flow, and neutron flux noise level limitations are prescribed in accordance with the recommendations of Genera1 Electric Service Information Letter No. 380, Revision 1, "BWR Core Thermal Hydraulic Stability," dated Febru-ary 10, 1984.

An inoperable jet pump is not, in itself, a sufficient reason.to declare a re-circulation loop inoperable, but it does, in case of a design basis accident, increase the blowdown area and reduce the capabflity of ref loodfng the core; thus, the requirement for shutdown of the facility with a jet, pump inoperable.

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

4 Recirculation pump speed mismatch limits are fn compliance with the ECCS LOCA analysis design criteria for two loop operation. The limits will ensure an adequate core flow coastdown from either recirculation loop following a LOCA.

In the. case where the mfsmatch limits cannot be mafntafned during the loop operation, continued operation fs permitted fn the single loop mode.

In order to prevent undue stress on the vessel nozzles and bottom head region, the recirculation loop temperatures shall be within 50 F of each other prior to startup of an idle loop. The loop temperature must also be within 50oF of the reactor pressure vessel coolant temperature to prevent thermal shock to the recirculation pump and recirculation nozzles. Since the coolant fn the bottom of the vessel fs at a lower temperature than the coolant fn the upper regions of the core, undue stress on the vessel would result ff the temperature differ ence was greater than 145'F.

SUSQUEHANNA - UNIT 2 B 3/4 4-1 Amendment No. 26

1 PLANT SYSTEMS BASES 3/4 7.6 FIRE SUPPRESSION SYSTEMS The OPERABILITY of the fire suppression systems ensures that adequate fire suppression capability is available to confine and extinguish fires occurring in any portion of the facility where safety related equipment is located. The fire suppression system consists of the water system, spray and/or sprinklers, CO systems, Halon systems and fire hose stations. The collective ca)ability of the fire suppression systems is adequate to minimize potential damage to safety related equipment and is a major element in the facility fire protection program.

In the event that portions of the fire suppression systems are inoperable, alternate backup fire fighting equipment is required to be made available in the affected areas until the inoperable equipment is restored to service. When the inoperable fire fighting equipment is intended for use as a backup means of fire suppression, a longer period of time is allowed to provide an alternate means of fire fighting than if the inoperable equipment is the primary means of fire suppression.

The surveillance requirements provide assurances that the minimum OPERABILITY requirements of the fire suppression systems are met. An allowance is made for ensuring a sufficient volume of Halon in the Halon storage tanks by verifying the weight and pressure of the tanks.

In the event the fire suppression water system becomes inoperable, immediate corrective measures must be taken since this system provides the major fire

'suppression capability of the plant. The requirement for a twenty"four hour report to the Commission provides for prompt evaluation of the acceptability of the corrective measures to provide adequate fire suppression capability for the continued protection of the nuclear plant.

3/4. 7. 7 FIRE RATED ASSEMBLIES The OPERABILITY of the fire barriers and barrier penetrations ensuie that fire damage will be limited. These design features minimize the possibility of a single fire involving more than one fire area prior to detection and extinguishment. The fire barriers, fire barrier penetrations for conduits, cable trays and piping, fire windows, fire dampers, and fire doors are periodically inspected to verify their OPERABILITY.

3/4.7e8 MAIN TURBINE BYPASS SYSTEM The, required OPERABILITY of the main turbine bypass system is consistent with the assumptions of the feedwater controller failure analysis in Nyct spy@'c frees>en'nalysis.

SUSQUEHANNA - UNIT 2 B 3/4 7-4

0

~,

DESIGN FEATURES 5.3 REACTOR CORE FUEL ASSEMBLIES or7V 5.3.1 The reactor cor shall contain 764 fuel assemblies with each fuel assembly containing 62 fuel rods and two water rods clad with Zircaloy -2.

Each fuel rod shall have a nominal active fuel length of 150 inches. The initial core loading shall have a maximum average enrichment of 1.90 weight percent U-235. Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum average enrichment of~weight percent U-235. +.o CONTROL ROD ASSEMBLIES 5.3.2 The reactor core shall contain 185 control rod assemblies, each consisting of a cruciform array of stainless steel tubes containing 143 inches of boron carbide, B4C, powder surrounded by a cruciform shaped stainless steel sheath.

5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4. 1 The reactor coolant system is designed and shall be maintained:

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

b. For a pressure of:

l. 1250 psig on the suction side of the recirculation pumps.

2. 1500 psig from the recirculation pump discharge'to the get pumpso

c. , For a temperature of 575 F.

'VOLUME 5.4.2 The total water and steam volume of the reactor vessel and recirculation system is approximately 22,400 cubic feet at a nominal T of 528'F.

SUSQUEHANNA - UNIT 2 5-6

NO SIGNIFICANT HAZARDS CONSIDERATIONS The following three questions are addressed for each of the proposed changes:

I. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

II. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

III. Does the proposed change involve a significant reduction in a margin of safety?

o Definition 1.2, Avera e Ex osure I. No. This change reflects the addition of the average exposure definition appropriate for Exxon Nuclear Company (ENC) fuel. The ENC POWERPLEX core monitoring system determines Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) based on average bundle exposure rather than average planar exposure, which is the related term for General Electric (GE) fuel. This additional definition is therefore administrative in nature.

II. No. See I above.

III. No. See I above.

o Definition 1.13,. Fraction of Limitin Power Densit I. No. This chan e is administrative in nature' the definition was altered to reflect the appropriate Linear Heat Generation Rate to be used in determining FLPD, since a Linear Heat Generation Rate curve specifically for determination of APRM setpoints has been provided in this analysis. This is justified under Specification 3/4.2.2, APRM setpoints.

II. No. See I above.

III. No. See I above.

o S ecification 3/4.1.2, Reactivit Anomalies I. No. This change is administrative in nature in that it reflects how POWERPLEX detects reactivity anomalies; POWERPLEX monitors Keff, which is a more direct measurement of reactivity than rod density.

This better monitoring method has no analytical ramifications.

II. No. See I above.

III. No. See I above.

o S ecification 3/4.2.1, Avera e Planar Linear Heat Generation Rate The changes to this specification reflect the use of the average bundle exposure definition discussed above, changes to remaining GE MAPLHGR figures to reflect a change in exposure units, the removal of all GE .711/

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enriched fuel, and the addition of appropriate limits for all Cycle 2 ENC 9X9 (XN-1) fuel.

No. Except for the new XN-1 fuel limits, each of the above changes are administrative in nature. For the average bundle exposure definition, this was previously discussed (see Definition 1.2).

Figures 3.2.1-1 and 3.2.1-2 for GE fuel have simply been altered due to the conversion of the abscissa units from MWD/t to MWD/MT for consistency with the new Figure 3.2.1-3. The current Figure 3.2.1-3 has been deleted since the .711% enriched fuel it applied to will not be part of the Cycle 2 core.

New Figure 3.2.1-3 illustrates the MAPLHGR limits for XN-1 fuel.

These limits are based upon an ENC analysis of t'e Loss of Coolant Accident (LOCA) as described in XN-NF-86-60 (attached). Based on this analysis, operation within the proposed MAPLHGR limits will ensure that the Peak Cladding Temperature (PCT) remains below 2200'F, local Zr-H2 0 reaction remains below 17%, and core-wide hydrogen production remains below 1% for the limiting LOCA as required by (

10CFR50.

With respect to GE fuel, the attached Reload Summary Report shows that the XN-1 fuel is hydraulically and neutronically compatible with GE fuel. Therefore the existing MAPLHGR limits, based on the GE LOCA analysis provided in the FSAR, remain applicable for Unit 2 Cycle 2 operation with GE fuel.

No. Although there are'bvious physical differences between the XN-1 fuel and the GE P8X8R fuel, there are no significant differences in their operating characteristics. Therefore, the addition of ZN-1 fuel in the Cycle 2 core does not create the possibility of a new or different kind'of accident from any accident previous evaluated.

No. The analyses performed were done in accordance with 10CFR50 Appendix K and did not predict a significant reduction in any safety margin. The methodology used to perform the Cycle 2 safety analyses contain similar inherent conservatisms to those which supported the initial core.

o S ecification 3/4.2.2, APRM Set pints This specification has been changed to explicitly define T for GE fuel and for ENC fuel. Since for ENC fuel T is dependent on a transient-based LHGR, a new figure, 3.2.2-1, has been provided.

No. For GE fuel, the method of calculating T has not changed.

Clarification was simply provided to ensure that T was properly determined. This part of the change is therefore editorial.

For ENC fuel, the T factor is modified by an exposure-dependent LHGR which is based on Exxon's "Protection Against Fuel Failure" (PAFF) line shown on Figure 3.4 of ZN-NF-85-67, Revision 1. This LHGR is

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provided in new Figure 3.2.2-1, which corresponds to the ratio of PAFF/1.2. Under this limit, cladding and fuel integrity are protected during anticipated operational, occurrences (AOO's),

including an overpower condition for transients initiated from partial power. Therefore, this change will ensure fuel design limits are not violated.

No. Again, for GE fuel the change is editorial. For ENC fuel, as stated in I above, the new LHGR limit provides assurance that cladding and fuel integrity are protected during AOO's. Since the ENC fuel is hydraulically and neutronically compatible with GE fuel, no new events were postulated to occur.

No. For GE fuel, no change has occurred. Since no limit previously existed for ENC fuel, the only comparison that can be made is with the method of calculating T for GE fuel. Since both methods are shown to provide appropriate protection against 1X clad strain and fuel centerline melting, no significant reduction in safety margin has occurred.

o S ecification 3/4.2.3, Minimum Critical Power Ratio No.. Some editorial chang'es to this specification consistent with the methodologies being utilized to determine MCPR operating limits have been provided (see the attached marked-up Technical Specification changes). As detailed in the Susquehanna SES Unit 2 Cycle 2 Reload Summary Report, QCPR results for local transients have been completed based on approved methods (see Summary Report Reference 13). The methodology for determining QCPRs for core-wide transients is reviewed here, but actual Technical Specification operating limits are not supplied because the calculations for the Feedwater Controller Failure (FWCF) and Load Re)ection Without Bypass (LRWOB) transients have yet to be completed. When these hCPRs are known, operating limits will be submitted based on these and the local transient results.

The plant transient model used to evaluate the system affects of the FWCF and LRWOB transients is ENC's COTRANSA code (See Summary Report Reference 16). This output will be utilized by the XCOBRA-T methodology (see Summary Report Reference 23) to determine b CPRs.

The COTRANSA code has been used in previous approved licensing submittals. The ZCOBRA-T code is appropriate for use in this application because it provides a more realistic treatment of transient phenomena than previously utilized methods and has been benchmarked against transient critical heat flux tests as reported in the above mentioned reference.

All core-wide transients will be analyzed deterministically (i.e.,

using bounding values of input parameters).

Based on the above, the method used to develop operating limit MCPRs for the Technical Specifications does not involve a significant

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of an accident previously II. No. The methodology described can only be evaluated for its affect on the consequences of analyzed events; it cannot create new ones.

The consequences of analyzed events were evaluated in I above.

III No. As stated in I above and in greater detail in the Summary Report, the methodology used to evaluate core-wide transients is consistent or more realistic than previously approved methods and meets all pertinent regulatory requirements for use in this application. Therefore, its use will not result in a significant decrease in any margin of safety.

o S ecification 3/4.2.4, Linear Heat Generation Rate

.This specification has been changed to provide appropriate limits for ENC fuel. The GE limit of 13.4 kw/ft has not changed.

I. No. New specification 3/4.2.4.2 and Figure 3.2.4.2-1 reflect appropriate LHGR limits for ENC fuel under steady-state conditions.

The figure is based on information provided in the fuel mechanical design analysis (XN-NF-85-67, Rev. 1) and assures margin to design limits for the life of the fuel.

II. No. This change reflects an additional control which has been previously accepted for GE fuel. Addition of this control to ENC fuel will not create the possibility of a new or different accident.

III. No. This new control has been shown to ensure compliance with all relevant fuel mechanical design criteria and therefore ensures appropriate safety margin.

o S ecification 3/4.3.4.2, End-of-C cle Recirculation Pum Tri S stem Instrumentation I. No. New action statements have been provided to ensure compliance with appropriate MCPR limits when EOC-RPT is inoperable. The requirements are consistent with those in the current MCPR Specification; therefore this change is administrative in nature.

II. No. See I above.

III. No. See I above.

o S ecification 3/4.4.1.1.2, Recirculation Loo s Sin le Loo 0 eration I. No. This specification has been changed to preclude extended operation with one recirculation loop out-of-service. Since this specification previously allowed such operation, this change constitutes an additional restriction which is much more conservative than the current provisions. Therefore, it will not increase the probability or consequences of any previous evaluation.

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II. No. See I above.

III. No. See I above.

o S ecification 3/4.7.8, Main Turbine B ass S stem I. No. This change is similar to that proposed for specification 3/4.3.4.2 and is proposed to make this specification consistent with the changes to 3/4.2.3, Minimum Critical Power Ratio. Since this change is consistent with the requirements in the current MCPR specification, no change in level of control has occurred.

Therefore, this change is administrative in nature.

II. No. See I above.

III. No. See I above.

o S ecification 5.3.1, Fuel Assemblies I. No. As written, this specification provides GE P8XSR general core design information. The proposed changes provide the same information for the ENC fuel being introduced in Cycle 2. This general information was part of a much more elaborate set of inputs used to generate the attached analyses and the Technical Specification limits discussed above. Since the Technical Specifications and associated analyses have been shown not to increase the probability or consequences of any previous evaluation, the proposed change to this section is primarily editorial and therefore will not degrade the current level of safety at Susquehanna SES Unit 2.

II. No. See I above.

III. No. See I above.

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