ML17157A411

Amend 102 to License NPF-14,changing Tech Specs in Support of Cycle 6 Reload

ML17157A411
Person / Time
Site:	Susquehanna
Issue date:	11/02/1990
From:	Butler W Office of Nuclear Reactor Regulation
To:
Shared Package
ML17157A412	List: ML17157A411 ML17157A413
References
NUDOCS 9011080296
Download: ML17157A411 (28)

Text

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0 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 PENNSYLVANIA POWER 5 LIGHT COMPANY ALLEGHENY ELECTRIC COOPERATIVE INC.

DOCKET NO. 50-387 SUS UEHANNA STEAM ELECTRIC STATION UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 102 License No. NPF-14 1.

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

A.

The application for the amendment filed by the Pennsylvania Power 5 Light Company, dated July 2, 1990 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),

and the Commission's regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the regulations of the, Commission; C.

There is reasonable assurance:

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

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

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

2.

Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment and paragraph 2.C.(2) of the Facility Operating License No. NPF-14 is hereby amended to read as follows:

(2)

Technical S ecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 102and the EnviroIIIGIntal Protection Plan con-

=tained in Appendix B, are hereby incorpo%%ed'n the license.

PPSL shall operate the facigity in acco~danc~eith the Technical Specifica-tions and the'nvironme'ntaT Pr'o&erHon~%.

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This license amendment is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

Attachment:

Changes to the Technical Specifications Date of Issuance:

November 2, 1990

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Walter R. Butler, Director Project Directorate I-2'ivision of Reactor Projects - I/II LA en (p$

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This license amendment is effective as of its date of issuance.

FOR TNE NUCLEAR REGULATORY CGYÃISSION

Attachment:

Changes to the Technical Specifications Malter R. Butler, Director Project Directorate 1-2 Division of Reactor Projects - I/II Date of Issuance:

November 2, 1990

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ATTACHMENT TO LICENSE AMENDMENT NO. 102 FACILITY OPERATING LICENSE NO. NPF-14 DOCKET NO. 50-387 Replace the following pages of the Appendix A Technical Specifications with enclosed pages.

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

The overleaf pages are provided to maintain document completeness.*

REMOVE INSERT XX1 XX11 8 2-2 XX1 XX11*

8 2-2 3/4 2-1 3/4 2-2 3/4 2-9 3/4 2-9a 3/4 2-10a 3/4 2-10b 3/4 4-lb 3/4 4-lc 8 3/4 2-1 8 3/4 2-2 8 3/4 2-3 8 3/4 4-1 8 3/4 4-la 5-5 5-6 3/4 2-1*

3/4 2-2 3/4 2-9 3/4 2-9a 3/4 2-10a*

3/4 2-lob 3/4 4-lb 3/4 4-lc 8 3/4 2-1 8 3/4 2-2 8 3/4 2-3 8 3/4 4-1 8 3/4 4-la 5 5*

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LIST OF FIGURES INDEX FIGURE

3. l. 5-1

3. 1. 5-2

3. 2. 1-1

3. 2. 1-2

3. 2. 2-1

3. 2. 3-1 3.2.3 2

3. 2,4-1

3. 2. 4-2

3. 4. 1. l. 1-1

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

5. 1. 1-1

5. 1. 2-1

5. l. 3-la

5. l. 3-lb

6. 2. 1-1

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

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

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

AVERAGE BUNDLE EXPOSURE, ANF 8x8 FUEL........

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

AVERAGE BUNDLE EXPOSURE, ANF 9x9 F UEL

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

FLOW DEPENDENT MCPR OPERATING LIMIT...~.............

REDUCED POWER MCPR OPERATING LIMIT.

LINEAR HEAT GENERATION RATE (LHGR) LIMIT VERSUS AVERAGE PLANAR EXPOSURE ANF Bx8 FUEL..

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

THERMAL POWER RESTRICTIONS..

MINIMUM REACTOR VESSEL METAL TEMPERATURE VS.

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

REACTOR VESSEL WATER LEVEL......................

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

EXCLUSION AREA LOW POPULATION ZONE....

MAP DEFINING UNRESTRICTED AREAS FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS MAP DEFINING UNRESTRICTED AREAS FOR RADIOACTIVE GASEOUS AND LI(UID EFFLUENTS OFFSITE ORGANIZATION..

UNIT ORGANIZATION

%e PAGE 3/4 1-21 3/4 1"22 3/4 2-2 3/4 2-3 3/4 2-6 3/4 2-9 3/4 2-9a 3/4 2"10b:

3/4 2-10c 3/4 4-lb 3/4 4-18 B 3/4 3-8 B 3/4 4-7 5-2 5"3 5"5 6-3 SUSQUEHANNA - UNIT 1 xx1 Amendment No. 102

0 LIST OF TABLES TA5LE tlCEX lAQE 2.2

~ 1 1

3.3.1 1

3.3.1 2

1.3.1 1 1

3. 3. 2-1 3.3.2 2

3.3.2 3

l.3.2.1 1

SURVEILLANCE FREOUENCV NOTATION.

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0 ERA 10 L CON01 IONS....,.

RKACTOR lROTKCTION SYSTKII INSTRIPIKNTATIOH SET%?NTS.....,.,

REACTOR 1ROTECTION SYSTKN. INSTRNIKNTATIQN.....,,...

REACTOR PROTECTION SYSTBI RESPONSE TIIKS..........

RKACTOR PROTECTION SYSTEN INSTRLNKHTATION SURVK?LLANCK RKQUIRKNENTS...

ISOLATION ACTUATION INSTNOtENTATION........,......

ISOLATION ACTUATION INSTRLNENTATION SET@) INTS.....

2 I I

3/4 3.2 3/i 3o6 3/4 3

7 3/i 3 11 3/i 3 17 ISOLATION SYSTBI !NSTROIKNTATION RESPONSE'II%....

3/i 3 21 ISOLATION ACTUATION INSTRNIKHTATION SURVEILLANCE 3.3.3 1

3.3.3 2

3.3.3 3

i.3.3. 1-1 3.3.l.l 1 3.3.i.l 2 QJIRflIKNTS o ~

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RE EMERGENCY CORE COOL?NG SYSTBI ACTUAT?ON INSTNNKNTATION...................................

K%RGKHCY Coaf COOL?NG SYSTElI ACTUATION INSTNPlfiHTATION SETPOINTS o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

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EMERGENCY CORE COOLING SYSTQI RESPONSE TllIES'......

9%RGENCY CORf COOLIIC SYSTOI ACTUATION "

INSTRNIKNTATION SURVEILLANCE RE gIRQ%XTS......

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A%$ RECIRCULATION~ TR?t SYSTBI lNSTWATION

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NVS REC?RCVLATION POO TR!0 SYSTSI i ? ISTNKNTATION SETS INTS o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

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3/l 3 23 3/4 3 28 3/i 3 31 3/1 3 33 3/i 3 P 3/I 3 37 3/l 3 34 SUQVKHNWA UNIT 1

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leerent No.72 OCT c9t

0 SAFETY LIMITS BASES

2. 1.2 THERMAL POWER Hi h Pressure and Hi h Flow Onset of transition boiling results in a decrease in heat transfer from the clad and, therefore, elevated clad temperature and the possibility of clad failure.

However, the existence of critical power, or boiling transition, is not a directly observable parameter in an operating reactor.

Therefore, the margin to boiling transition is calculated from plant operating parameters such as core power, core flow, feedwater temperature, and core power distribution.

The margin for each fuel assembly is characterized by the critical power ratio (CPR),

which is the ratio of the bundle power which would produce onset of transition boiling divided by the actual bundle power.

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

The Safety Limit MCPR assures sufficient conservatism in the operating MCPR limit that in the event of an anticipated operational occurrence from the limiting condition for operation, at least 99.9X of the fuel rods in the core would be expected to avoid boiling transition.

The margin between calculated boiling transition (MCPR = 1.00) and the Safety Limit MCPR is based on a detail-ed statistical procedure which considers the uncertainties in monitoring the core operating state.

One specific uncertainty included in the safety limit is the uncertainty inherent in the XN-3 critical power correlation.

XN-NF-524 (A)

Revision 1 describes the methodology used in determining the Safety Limit MCPR.'he XN-3 critical power correlation is based on a significant body of practical test data, providing a high degree of assurance that the critical power as evaluated by the correlation is within a small percentage of the actual critical power being estimated.

As long as the core pressure and flow are within the range of validity of the XN-3 correlation (refer to Sec-tion B 2. 1. 1), the assumed reactor conditions used in defining the safety limit introduce conservatism into the limit because bounding high radial power fac-tors and bounding flat local peaking distributions are used to estimate the number of rods in boiling transition.

Still further conservatism is induced by the tendency of the XN-3 correlation to overpredict the number of rods in boiling transition.

These conservatisms and the inherent accuracy of the XN-3 correlation provide a reasonable degree of assurance that during sustained operation at the Safety Limit MCPR there would be no transition boiling in the core.

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

Significant test data accumulated by the U.S. Nuclear Regulatory Commission and private organiza-tions indicate that the use of a boiling transition limitation to protect against cladding failure is a very conservative approach.

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

ANF fuel is monitored using the XN-3 critical power correlation.

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

The~servatism has been evaluated by ANF to be greater than the maximum expected BCPL: P(A2) due to channel bow in C-lattice plants using channels fear os ~f~el bundle lifetime.

Since Susquehanna SES is a C-lattice plant and'ses c ihneTs for only one fuel bundle lifetime, monitoring of the MCPR limit with the XN-3 critical power correlation is conservative with respect to channel bow and addresses the concerns of NRC Bulletin No. 90-02 entitled "Loss of Thermal Margin Caused by Channel Box Bow."

SUSQUEHANNA - UNIT 1 8 2"2 Amendment No.

102

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2, RPTELT 3/4. 2 POWER OISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for each type of fuel shall not exceed the limits shown in Figures 3.2. 1-1 and 3.2. 1-2.

1 APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or

~21 21 1 RATER 1

ERPRL PR ER.

ACTION:

With an APLHGR exceeding the limits of Figure 3.2.1-1 or 3.2.1-2, initiate corrective action within 15 minutes and restore APLHGR to within the required limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 25X 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. 1 All APLHGRs shall be verified to be equal to or less than the limits determined from Figures 3.2. 1-1, and 3.2. 1"2.

a.

At least once per 24 hours2.777778e-4 days <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'ATED THERMAL POWER, and c.

Initially and at least o'ce 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 ROO PATTERN for APLHGR.

d.

The provisions of Specification 4.0.4 are not applicable.

SUSQUEHANNA - UNIT 1 3/4 2-1.

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6000 10000 16000 20000 26000 30000 36000 40000 Average Bundl'e Exposure (MWD/MT)

MAXIMUMAVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE BUNDLE EXPOSURE ANF 8X8 FUEL FIGURE 3.2.1-1 I

(30, 1.s2) 1.9 1.8" CURVE A: EOC-RPT Inoperable; Main Turbine Bypass Operable CURVE B: Main Turbine Bypass inoperable; EOC-RPT Operable CURVE C: EOC-RPT and Main Turbine Bypass Operable (40, 1.61) 1.4 (60, 1.43)

(62.6, 1.40)

(61.7, 1.41)

(67.6, 1.34)

A 1.41 1.40 1.34 1.2 30 40 60 BO '0 80 80 Total Core Flow (% OF RATED)

FLOW DEPENDENT MCPR OPERATING LIMIT FIGURE 3.2.3-1 100

1.7 CURVE A: EOC-RPT Inoperable:

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

1.4 (26, 1.44)

(26. 1.41)

(66. 1.44)

(80, 1.42)

A 1.41 B

1.40 l3 6, 1.34) 1.34 1.2 20 30 40 60 60 70 80 Core Power (% OF RATED)

REDUCED POWER MCPR OPERATING LIMIT Figure 3.2.3-2 90 100

~ QADI!f POWER DISTRIBUTION LIMITS 3/4. 2.4 LINEAR HEAT GENERATION RATE ANF FUEL LIMITING CONDITION FOR OPERATION 3.2.4 The LINEAR HEAT GENERATION RATE (LHGR) shall not exceed the LHGR limit determined from Figures 3.2.4-1 and 3.2.4-2.

APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is gr eater than orequal to 25K of RATED THERMAL POWER.

ACTION:

With the LHGR of any fuel rod exceeding its applicable limit from Figure 3.2.4-1 or 3 ~ 2.4-2, initiate corrective action within 15 minutes and restore the LHGR to within the limit within 2 hours2.314815e-5 days <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 LHGRs 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 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 on a LIMITING CONTROL ROD PATTERN for LHGR.

d.

The provisions of Specification 4,0.4 are not applicable.

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LINEAR HEAT GENERATION RATE (LHGR) LIMIT VERSUS AVERAGE PlANAR EXPOSURE ANF 8X8 FUEL FIGURE 3.2 4-1 60000

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3/4 4-1b Amendment No. 102

REACTOR COOLANT SYSTEM RECIRCULATION LOOPS - SINGLE LOOP OPERATION LIMITING CONOITION FOR OPERATION 3.4. 1. 1.2 One reactor coolant recirculation loop shall be in operation with the pump speed

< 80K of the rated pump speed and the reactor at a THERMAL POWER/core flow condition outside of Regions I and II of Figure 3.4.1.1.1-1, and a ~

the following revised specification limits shall be followed:

1.

Specification 2.1.2:

the MCPR Safety Limit shall be increased to 1.07.

2.

Table 2.2. 1-1:

the APRM Flow-Biased Scram Trip Setpoints shall be as follows:

Tri Set oint

< 0.58W + 54 Allowable Value

< 0.58W + 5 3.

4.

5.

Specification

3. 2. 2:

the APRM Setpoints shal.l be as follows:

T~

Allowable Value S

< (0.58W + 54K)T S

< (0.58W + 57X)T SRB (0.58W + 45K)T SRB (0.58W + 4SX)T Specification 3.2.3:

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

a.

l. 30, b.

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

the MCPR determined from Figure 3.2.3-2 plus 0.01.

Table 3.3.6-2:

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

a.

RBM - Upscale Tri Set oint

< 0.66W + 36 Allowable Value b.

APRM-Flow Biased Tri Set oint Al 1 owabl e Value

< 0.58W + 45 APPLICABILITY:

OPERATIONAL CONOITIONS 1" and 2*+, except during two loop operation.0 m%R SUS(UEHANNA - UNIT 1 3/4 4-lc Amendment No. 102

BASES The specifications of this section assure that the peak cladding tempera-ture following the postulated design basis loss-of-coolant accident will not exceed the 2200'F limit specified in 10 CFR 50.46.

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

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

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

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

H The calculational procedure used to establish the APLHGR shown on Figures

3. 2. 1-1, and 3. 2. 1-2 is based on a loss-of-coolant accident analysis.

The analysis was performed using calculational models which are consistent with the requirements of Appendix K to 10 CFR 50.

These models are described in XN-NF-80-19, Volumes 2, 2A, 2B and 2C.

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

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

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

For ANF,fuel the T factor used to adjust the APRM setpoints is based on the FLPO calculated by dividing the actual LHGR by the LHGR obtained from Figure 3. 2. 2-1.

The LHGR versus exposure curve in Figure 3. 2. 2-1 is based on'NF's Protection Against Fuel Failure (PAFF) line shown in Figure 3.4 of XN-. NF-85-67(A), Revision 1.

Figure 3. 2. 2-1 corresponds to the ratio of PAFF/1.2 under which cladding and fuel integrity is protected during AOOs.

SUSQUEHANNA - UNIT 1 B 3/4 2-1 Amendment No. 102

POWER DISTRIBUTION LIMITS BASES 3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRs at steady state operating conditions as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safety Limit MCPR, and an analysis of abnormal operational transients.

For any abnormal operating transient analysis evaluation with the initial con-dition of the reactor being at the steady state operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any-time during the transient assuming instrument trip setting given in Specification 2.2.

To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting tran-sients have been analyzed to determine which result in the largest reduction in CRITICAL POWER RATIO (CPR).

The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease.

The limiting transient yields the largest delta MCPR.

When added to the Safety Limit MCPR, the required minimum operating limit MCPR of Specification 3.2.3 is obtained and presented in Figures 3.2.3-1 and 3.2.3-2.

The evaluation of a given transient begins with the system initial param-eters shown in the cycle specific transient analysis report that are input to an ANF core dynamic behavior transient computer program.

The outputs of this program along with the initial MCPR form the input for further analyses of the thermally limiting bundle.

The codes and methodology to evaluate pressuriza-tion and non-pressurization events are described in XN-NF-79-71 and XN-NF-84-105.

The principal result of this evaluation is the reduction in MCPR caused by the

'ransient.

Figure 3.2.3-1 defines core flow dependent MCPR operating limits,which assure that the Safety Limit MCPR will not be exceeded during a flow increase transient resulting from a motor-generator speed control failure.

The flow dependent MCPR is only calculated for the manual flow control mode.

Therefore, automatic flow control operation is not permitted.

Figure 3.2.3-2 defines the power dependent MCPR operating limit which assures that the Safety Limit MCPR will not be exceeded in the event of a Feedwater Controller Failure, Rod With-drawal Error, or Load Reject without Main Turbine Bypass Operable initiated from a reduced power condition.

Cycle specific analyses are performed for the most limiting local and core wide transients to determine thermal margin.

Additional analyses are performed to determine the MCPR operating limit with either the Main Turbine Bypass in-operable or the 'EOC-RPT inoperable.

Analyses to determine thermal margin with both the EOC-RPT inoperable and Main Turbine Bypass inoperable have not been performed.

Therefore, operation in this condition is not permitted.

At THERMAL POWER levels less than or equal t 5X of RATED THERMAL POWER, the reactor'will be operating at minimum rec+rcul Jump speed and the moderator void content will be very small, EK all esignated control rod patterns which may be employed a% thA p~t g plant experience indi-cates that the resulting MCPR value is in excess of requirements by a consider-able margin.

During initial start-up testing of the plant, a

MCPR evaluation SUSQUEHANNA " UNIT 1 B 3/4 2"2 Amendment No. 102

BASES MINIMUM CRITICAL POWER RATIO (Continued) 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 25'f 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 1 imit.

3/4. 2.4 LINEAR HEAT GENERATION RATE This specification assures that the Linear Heat Generation Rate (LHGR) in any fuel rod is less than the design linear heat generation even if fuel pellet densification is postulated.

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B 3/4 2-3 Amendment No. IO2

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3/4. 4 REACTOR COOLANT SYSTEM BASES 3/4. 4. 1 RECIRCULATION SYSTEM Operation with one reactor recirculation loop inoperable has been evaluated and found acceptable, provided that the unit is operated in accordance with Spec ificati on 3. 4. 1. 1. 2.

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

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

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

The MINIMUM CRITICAL POWER RATIO (MCPR) limits for single loop operation assure that the Safety Limit MCPR is not exceeded for any Anticipated Opera-tional Occurrence (AOO).

In addition, the MCPR limits for single-loop opera-tion protect against the effects of the Recirculation Pump Seizure Accident.

That is, for operation in single-loop with an operating MCPR limit >1.30, the radiological consequences of a pump seizure accident from single-loop operating conditions are but a small fraction of 10 CFR 100 guidelines.

For single loop operation, the RBM and APRM setpoints are adjusted by a 8.5X decrease in recirculation drive flow to account for the active loop drive flow that bypasses the core and goes up through the inactive loop jet pumps.

Surveillance on the pump speed of the operating recirculation loop is im-posed to exclude the possibility of excessive reactor vessel internals vibration.

Surveillance on differential temperatures below the threshold limits on THER-MAL 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.

Specifications have been provided to prevent,

detect, and mitigate core thermal hydraulic instability events.

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

The boundaries of the regions in Figure 3.4. 1. 1. 1. 1-1 are determined using ANF decay ratio calculations and supported by Susquehanna SES stability testing.

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

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

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

The number and location of LPRM strings in each zone assure that with 50K or more of the associated LPRM upscale alarms OPERABLE sufficient monitoring capability is available-to detect core wide and regionM osc$ R5tions.

Operating plant in-stability data is used to determine the specific LPRM strings assigned to each lA fit in appropriate procedures.

SUS(UEHANNA " UNIT 1 B 3/4 4-1 Amendment No. 102

3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4. 1 RECIRCULATION SYSTEM (Continued)

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

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

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

Jet pump failure can be detected by monitoring jet pump performance on a

prescribed schedule for significant degradation:

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

The limits will ensure an adequate core flow coastdown from either recirculation loop following a LOCA.

In the case where the mismatch limits cannot be maintained during the loop operation, continued operation is permitted in the single loop mode.

In order to prevent undue stress on the vessel nozzles and bottom head

region, the recirculation loop temperatures shall be'ithin 50~F of each other prior to star tup of an idle loop.

The loop temperature must also be within 50'F of the reactor pressure vessel coolant temperature to prevent thermal shock to the recirculation pump and recircu1ation nozzles.

Since the coolant

,in the bottom of the vessel is at a lower temperature than the coolant in the upper regions of the core, undue stress on the vessel would result if the tem-

.'erature difference was greater than 145~F.

SUSQUEHANNA - UNIT 1 8 3/4 4"la Amendment No. >O2

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~ngment Ne. 29

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DESIGN FEATURES 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 764 'fuel assemblies with each fuel. as-sembly containing 62 or 79 fuel rods and two water rods clad with Zircaloy -2.

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

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

CONTROL ROD ASSEMBLIES r

5.3.2 The reactor core shall contain 185 control rod assemblies consisting of two different designs.

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

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

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, w'ith allowance for normal degradation pursuant to the applicable Surveillance Requirements, b.

For a pressure of:

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1250 psig on the suction side of the recirculation pumps.

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

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 j.

5-6 Amendment No. ]02