Text

Amend 58 to License NPF-22,revising Tech Specs in Support of Fuel Reload for Cycle 4 Operation
	ML17156B445
Person / Time
Site:	Susquehanna
Issue date:	11/03/1989
From:	Butler W; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML17156B446	List: ML17156B445; ML17156B447;
References
	NUDOCS 8911130025
	Download: ML17156B445 (38)
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'a*<<+'NITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O. C. 20555 PENNSYLVANIA POMER 5 LIGHT COMPANY ALLEGHENY ELECTRIC COOPERAT!VF.,

INC.

ROCKET NO. 50-388 SUS UEHANNA STEAM ELECTRIC STATION UNIT 2 AMENnII'ENT TO FACILITY OPERATING LICENSE Amendment No. 58 License No.

NPF-22 The Nuclear Regulatory Commission (the Commission or the NRC) has found that:

A.

The application for the amendment filed by the Pennsylvania Power 5

Light Company, dated June 16, 1989 as clarified October 6,

1989, complies with the standards and requirements o+ the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I; 8.

The ~acility will ooera+e in conformity with the application, the Drovisions of the Act, and the regulations o< the Commission; C.

There is reasonable assurance:

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

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

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations ard 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 paraqraph P.C.(2) of the Facility tiperatinq License No. NPF-V is hereby amended to read as follows:

(2>

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 58 and the Environmental Protection Plan con-tained in Appendix 8, are hereby incorpo~ated in the license.

PP8L shall operate the facility in accordance with the Technical Soecifica-tions and the Environmental Protection Plan.

85'ii130025 89ii03 PDR ADOCK 0 000388 P

PDC

3.

This 1icense amendment is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

Attachment:

Changes to the Technical Specifications Date of Issuance:

November 3, 1989 Mohan C. Thadani for Walter R. Butler, Director Project Directorate I-2 Division of Reactor Projects I/II PDI-2/LA O'rien

)(/) /89 PDI-2/PM MThadani

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

FOR THE NUCLEAR REGULATORY COMMISSION C

t Palter R. Butler, Director Pro'ect Directorate I-2 Division of Reactor Pro'ects I/?I

Attachment:

Chanaes to the Technical Specifications Date o~ '.ssuance:

November 3, Ig8g

ATTACHMENT TO LICENSE AMFNOMENT NO. 58 FACIlITY OPERATINt~ LICENSE NO.

NPF-22 DOCVFT NO. 50-388 Replace the following pages of the Aopendix A Technical Specifications with enclosed pages.

The revised paoes are identified by Amendment number and contain vertica> lines indicating the area n+ change.

The over1eaf pages are provided to maintain document completeness.*

RENOVE 111 1V xxi xxii B 2-1 B 2-2 3/4 2-1 3/n 3/4 2-3 3/4 2-4 3/4 2-5 3/4 2-6 3/4 2-Ga 3/4 2-7 3/4 2-8 3/4 2-9 3/4 2-10 3/4 2-10a 3/4 2-10b 3/4 4>>l 3/4 n-la 3/4 4-jb 3/4 4-lc 3/4 4-j(I 3/4 4-le INSERT 1 1 1*

1V xxi*

XX11 B 2-1 8 2-2 3/4 2-l 3/4 2-2 "3/n 2-3 3/4 2-4 3/4 2-5 3/4 2-6 3/4 2-7 3/4 2-8 3/4 2-9 3/4 2-10 3/4 4-1 3/4 4-la*

3/4 4-jb 3/4 4-jc

?/4 4-1d 3/4 4 le*

I I

4

ATTACHMENT TO LICENSE AMENDMENT NO. 58 FACILITY OPERATING LICENSE NO.

NPF-22 OOCKET NO. 50-388 RE??OVE 3/4 4-lf 3/4 4-lg 8 3/4 2-1 8 3/4 2-2 8 3/4 2-3 8 3/4 4-1 B 3/4 4-2 5-5 5-6

?MSERT 3/4 4-1f*

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

5*

5-6

~ I INDEX SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION PAGE 2.1 SAFETY LIMITS THERMAL POWER, Low Pressure or Low Flow...................

2-1 V

THERMAL POWER, High Pressure and High Flow................

2-1 Reactor Coolant System Pressure.........,..........,......

2-1 Reactor Vessel Water Level.........,......;,............

2-2 2."

LIM'..:NG SAFETY SYSTEM SETTINGS Reactor Protection System Instrumentation Setpoints.......

2-3 BASES

2. I SAFETY LIMI75 THERMAL'OWER, Low Pressure or Low Flow...................

B 2-1 THERMA'OWER, High Pressure and High Flow................

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

8 2-5 Reactor Vessel Water Level..

8 2-5

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

B 2-6 SUSQUEHANNA - UNIT 2

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

3/4. 1 REACTIVITY CONTROL SYSTEMS 3/4. 1. 1 SHUTDOWN MARGIN............,.........

3/4. 1. 2 REACTIVITY ANOMALIES.................

3/4. 1. 3 CONTROL RODS Contr ol Rod Operability..............

Control Rod Maximum Scram Insertion Times.

Control Rod Average Scram Insertion Times PAGE 3/4 0" 1 3/4 1-1 3/4 1-2 3/4 1-3 3/4 1-6 3/4 1-7 Four Control Rod Group Scram Insertion Times...........

3/4 1-8 Control Rod Scram Accumulators Control Rod Drive Coupling..

~Control Rod Position Indication.

Control Rod Drive Housing Support.

3/4. 1.4 CONTROL ROD PROGRAM CONTROLS Rod Worth Minimizer.

'Rod Sequence Control System.

Rod Block Monitor.

3/4.1.5 STANDBY LIQUID CONTROL SYSTEM.

3/4. 2 POWER DISTRIBUTION LIMITS 3/4. 2. 1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE...........

3/4 2.2 APRM SETPOINTS 3/4.2.3 MINIMUM CRITICAL POWER RATIO.

3/4.2.4 LINEAR HEAT GENERATION RATE 3/4 1-9 3/4 1-11 3/4 1-13 3/4 1-15 3/4 1-16 3/4 1-17 3/4 1-18 3/4 1-19 3/4 2-1 3/4 2-3 3/4 2-6 3/4 2-9

~

SUSQUEHANNA - UNIT 2 iv Amendment No. 58

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INDEX ADMINISTRATIVE CONTROLS

6. 13 PROCESS CONTROL PROGRAM.

6-23 6.14 OFFSITE DOSE CALCULATION MANUAL....,......,..............

6-24 6.15 MAJOR CHANGES TO RADIOACTIVE HASTE TREATMENT SYSTEMS.....

6-24 SUSQUEHANNA - UNIT 2 XX1

LIST OF FIGURES FIGURE INDEX PAGE

3. 1. 5-1

3. l. 5-2

3. 2. 1-1 SODIUM PENTABORATE SOLUTION TEMPERATURE/

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

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

AVERAGE PLANAR EXPOSURE, ANF 9 X 9 FUEL 3/4 1-21 3/4 1-22 I

I 3/4 2-2 I

3.2.2 1

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

3/4 2-5

3. 2. 3-1 3 ~ 2. 3-2 3.2.4 -1 FLOW DEPENDENT MCPR OPERATING LIMIT..............

~.

REDUCED POWER MCPR OPERATING LIMIT...........

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

3/4 2-7 3/4 2 8 I

t 3/4 2-10 I

3.4. 1. l. 1-1

3. 2. 6. 1-1

4. 7. 4"1 B 3/4 3-1 REACTOR VESSEL WATER LEVEL..

~.

THERMAL POWER/CORE FLOW LIMITATIONS................

MINIMUM REACTOR VESSEL METAL TEMPERATURE VS.

REACTOR VESSEL PRESSURE....

SAMPLE PLAN 2)

FOR SNUBBER FUNCTIONAL TEST.........

3/4 4-jb J

I 3/4 4-18 3/4 7-15 B 3/4 3-8 B 3/4.4.6-1 FAST NEUTRON FLUENCE (E)lMeV) AT 1/4 T AS A FUNCTION OF SERVICE LIFE.......

B 3/4 4-7

5. l. 1-1

5. 1. 2-1

5. l.3-la EXCLUSION AREA..

5-2 MAP DEFINING UNRESTRICTED AREAS FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS......;..............

5-4 LOW POPULATION ZONE 5-3

5. 1. 3-lb MAP DEFINING UNRESTRICTED AREAS FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS.....

5-5 SUSQUEHANNA " UNIT 2 XX11 Amendment No. 58

~ 2.1" SAFETY LIMITS BASES

2. 0 INTRODUCTION The fuel cladding, reactor pressure vessel and primary system piping are the principal barriers to the release of radioactive materials to the environs.

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

The fuel cladding integrity Safety 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 NCPR is not less than the limit specified in Specification

2. 1.2 for ANF.

fuel.

MCPR greater than the specified limit represents a conservative margin relative to the conditions required to maintain fuel cladding integrity.

The fuel cladding is one of the physical barriers which separate the radioactive materials from the environs.

The integrity of this cladding barrier is related to its relative freedom from perforations or cracking.

Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable.

Fuel cladding perforations,

however, can result from thermal stresses which occur from reactor operation significantly above design conditions and the Limiting Safety System Settings.

While fission product migration from clad-ding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding deteri-oration.

Therefore, the fuel cladding Safety Limit is defined with a margin to the conditions which would produce onset of transition boiling, MCPR of 1.0.

These conditions represent a significant departure from the condition intended by design for planned operation.

The MCPR fuel cladding integrity Safety limit assures that during normal operation and during anticipated operational occurrences, at least 99.9% of the fuel rods in the core do not experience transition boiling (ref.

XN-NF-524(A) Revision 1).

2. 1. 1 THERMAL POWER Low Pressure or Low Flow The use of the XN-3 correlation is valid for critical power calculations at pressures greater than 580 psig and bundle mass fluxes greater than 0.25 x 10 lbs/hr-ft For operation at low pressures or low flows, the fuel cladding integrity Safety Limit is established by a limiting condition on core THERMAL POWER with the following basis:

Provided that the water level in the vessel downcomer is maintained above the top of the active fuel, natural circulation is sufficient to assure a mini-mum bundle flow for all fuel assemblies which have a relatively high power and potentially can approach a critical heat flux condition.

For the ANF 9 x 9

fuel design, the minimum bundle flow is greater than 30,000 lbs/hr.

For this

design, the coolant minimum flow and maximum flow area is such that the mass flux is always greater than 0.25 x 10 1bs/hr-ft Full scale critical power tests taken at pressures down to 14.7 psia indicate that the fuel assembly critical power at 0.25 x 10 lbs/hr-ft is 3.35 Mwt or greater.

At 25% thermal power a bundle power of 3.35 Mwt corresponds to a bundle radial peaking factor of greater than 3.0 which is significantly higher than the expected peaking

Thus, a

THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressures below 785 psig is conservative.

SUSQUEHANNA - UNIT 2 B 2-1 Amendment No. 58

SAFETY LIMITS BASES

2. l. 2 THERMAL POWER Hi h Pressure and Hi h Flow Onset of transition boiling results in a deci ease 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 parametei s 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 tran-sition 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 transi tion (MCPR = 1. 00) and the Safety Limit MCPR is based on a de-tailed statistical procedure which considers the uncertainties in monitoring the core operating state.

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

XN-NF-524 (A), Revision 1 describes the methodology used in determining the Safety Limit MCPR.

The XN-3 critical power correlation is based on a,significant body of prac-tical 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 criti-cal power being estimated.

As long as the core pressure and flow are within the range of validity of the XN-3 correlation (refer to Section B 2. 1. 1), the assumed reactor conditions used in defining the safety limit introduce conser-vatism into the limit because bounding high radial power factors 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 boi ling transi-tion 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 transition limitation to protect against cladding failure is a

very conservative approach.

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

SUS(UEHANNA - UNIT 2 B 2-2 Amendment No. 5B

3/4. 2 POWER DISTRIBUTION LIMITS 3/4.2. 1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2. 1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for all fuel shall not exceed the limit shown in Figure 3.2. 1-1.

APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or ATE HE "E

ACTION; With an APLHGR exceeding the limit of Figure 3.2. 1-1, initiate corrective action within 15 minutes and restore APLHGR to within the required limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 25K of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE RE UIREMENTS 4.2. 1 All APLHGRs shall be verified to be equal to or less than the limit determined from Figure 3.2. 1-1:

a.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.

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

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

d.

The provisions of Specification 4.0.4 are not applicable.

SUS(UEHANNA - UNIT 2 3/4 2-1 Amendment No.5B

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O MAXIMUMAVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE BUNDLE EXPOSURE ANF SXB FUEL FIGURE 3.2.1-1

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 Value S

0.58W + 59%)T SRB

< (0.58W + 50%)T SRB

< (0.58W + 53%)T where:

S and S

B are in percent of RATED THERMAL POWER, W = LooPrecirculation flow as a percentage of the loop recirculation flow which produces a rated core flow of 100 million lbs/hr, T = Lowest value of the ratio of FRACTION OF RATED THERMAL POWER divided

'by the MAXIMUM FRACTION OF LIMITING POWER DENSITY.

The FRACTION I

OF LIMITING POWER DENSITY (FLPD) for ANF fuel is the actual LHGR divided by the LINEAR HEAT GENERATION RATE from Figure 3.2.2-1.

T is always less than or equal to 1.0.

APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or O )Or IIATEE ThERMAL POWER.

ACTION:

With the APRM flow biased simulated thermal power-upscale scram trip setpoint and/or the flow biased neutron flux-upscale control rod block trip setpoint less conservative than the value shown in the Allowable Value column for S or S B, as above determined, initiate corrective action within 15 minutes and adjust 3 and/or SRB to be consistent with the Trip Setpoint value" within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE RE UIREMENTS 4.2.2 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, as required:

"With MFLPD greater than the FRTP during power ascension up to 90% 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 100% times MFLPD, provided that the adjusted APRM reading does not exceed 100% of RATED THERMAL POWER, the required gain adjustment increment does not exceed 10% of RATED THERMAL POWER, and a notice of the adjustment is posted on the reactor control panel.

See Specification 3.4. 1. 1.2.a for single loop operation requirements.

SUSQUEHANNA - UNIT 2 3/4 2-3 Amendment No. 58

POWER DISTRIBUTION LIMITS 3/4. 2. 2 APRM SETPOINTS LIMITING CONDITION FOR OPERATION 4.2.2 (Continued) a.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.

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

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

d.

The provisions of Specification 4.0.4 are not applicable.

1 SUSQUEHANNA - UNIT 2 3/4 2-4 Amendment No. 58

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10000 20000 30000 40000 Average Planar Exposure (MWD/MT) 50000 LINEAR HEAT GENERATION RATE FOR APRH SETPOINTS VERSUS AVERAGE PLANAR EXPOSURE ANF FUEL FIGURE 3.2.2-1

POWER DISTRIBUTION LIMITS 3/4. 2. 3 MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION 3.2.3 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be greater than or equal to the greater of the two values determined from Figure 3.2.3-1 and Figure

3. 2. 3-2.

APPLICABILITY:

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

.qqq.f tATED TIIEIIMAL P ER.

ACTION:

With MCPR less than the applicable MCPR limit determined above, initiate correc-tive action within 15 minutes and restore MCPR to within the required limit with-in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 25K of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE RE UIREMENTS 4.2.3. 1 MCPR shall be determined to be greater than or equal to the applicable MCPR limit determined from Figure 3.2.3-1 and Figure 3.2.3-2:

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

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

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

d.

The provisions of Specification 4.0.4 are not applicable.

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

5B

AD ITI 2.0 9

(30,1.8'I) 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 B

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(60.77, 1.4 (68.93,1.33 B

1.41 1.40 1.33 O

1.2 30 40 60 60 70 80 Total Core Flow (% OF RATED)

FLOW DEPENDENT MGPR OPERATING LIMIT FIGURE 3.2.3-1 90 100

1.7 (26.1.64)

(26.1.62)

(40.1.62)

CURVE A: EOC-RPT Inoperable:

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Main Turbine Bypass Operable CURVE B: Main Turbine Bypass Inoperable; EOC-RPT Operable CURVE C: EOC-RPT and Main Turbine Bypass Operable (66.1.49)

(40,1 60)

(66,1.47)

(80,1.47)

(26,1.44)

(40,1.42)

(86.1.39)

(80.1.46)

(80.1.37)

(90 9,141) 8 1.41 1.40 1.33 (90.9.1.33) 20 30 80 40 60 60 70 Core Power (% OF RATED)

REDUCED PONER MCPR OPERATING lIMIT FIGURE 3.2.3-2 90

- 100

0

" POWER DISTRIBUTION LIMITS 3/4.2.4 LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.4 The LINEAR HEAT GENERATION RATE (LHGR) shall not exceed the LHGR limit determined from Figure 3.2.4-1.

APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or qqq.

f RATEO THE fl L E

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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 25K of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

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

a.

At least once per 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months s, l

b.

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

Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating on a LIMITING CONTROL ROD PATTERN for LHGR.

d.

The provisions of Specification 4.0.4 are not applicable.

SUS(UEHANNA - UNIT 2 3/4 2-9 Amendment No. 58

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EJl CO 0

10000 20000 30000 40000 60000 Average Planar Exposure (MID/MT)

LINEAR HEAT GENERATION RATE (LHGR) LIMIT VERSUS AVERAGE PLANAR EXPOSURE ANF 9X9 FUEL FIGURE.3.2.4-I

0 3/4.4 REACTOR COOLANT SYSTEM 3/4. 4. 1 RECIRCULATION SYSTEM RECIRCULATION LOOPS -

TWO LOOP OPERATION LIMITING CONDITION FOR OPERATION

3. 4. 1. 1. 1 Two reactor coolant system recirculation loops shall be in operation and the reactor shall be at a

THERMAL POWER/core flow condition less than or equal to the limit specified in Figure 3.4. l. 1. 1-1.

APPLICABILITY:

OPERATIONAL CONDITIONS l~ and 2", except during single loop operation.0 ACTION:

a.

With one reactor coolant system recirculation loop not in operation, comply with the requirements of Specification 3.4. l. 1.2, or take the associated ACTION.

b.

With no reactor coolant system recirculation loops in operation, immediately initiate an orderly reduction of THERMAL POWER to less than or equal to the limit specified in Figure 3.4. 1. 1. 1-1, and initiate measures to place the unit in at least STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c.

With two reactor coolant system recirculation loops in operation and the reactor at a

THERMAL POWER/core flow condition greater than the limit specified in Figure 3.4. l. l. 1-1:

Restore the reactor to a

THERMAL POWER/core flow condition less than or equal to the limit specified in Figure 3. 4. l.l. 1-1, or 2.

Determine the APRM and LPRM*** neutron flux noise levels within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and:

a)

IF the APRM and LPRM"*~ neutron flux noise levels are less than three times their established baseline

levels, continue to determine the noise levels at least once per 8

hours and within 30 minutes after the completion of a THERMAL POWER increase of at least 5X of RATED THERMAL POWER, or b)

If the APRM or LPRM""* neutron flux noise levels are greater than or equal to three times their established baseline

levels, immediately initiate corrective action and restore the noise levels to within the required limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months by returning the reactor to a THERMAL POWER/core flow condition less than or equal to the limit specified in Figure 3.4. l.l. 1-1.

See Special Test Exception

3. 10.4.

- - Detectors A and C or one LPRM string per'ore octant plus detectors A and C

of one LPRM string in the center of the core should be monitored.

¹See Specification 3.4. 1. 1.2 for single loop operation requirements.

SUSQUEHANNA - UNIT 2 3/4 4-1 Amendment No. 58

REACTOR COOLANT SYSTEM SURVEILLANCE RE UIREMENTS 4.4. l. l. l. 1 Each pump discharge valve and bypass valve shall be demonstrated OPERABLE by cycling each valve through at least one complete cycle of full travel during each startup"" prior to THERMAL POWER exceeding 25K of RATED THERMAL POWER.

4.4. 1. 1. 1.2 Each pump discharge bypass valve, if not OPERABLE, shall be verified to be closed at least once per 31 days.

4.4. 1. 1. 1.3 Each pump MG set scoop tube electrical and mechanical stop shall be demonstrated OPERABLE with overspeed setpoints less than or equal to 102.5 and 105~, respectively, of rated core flow, at least once per 18 months.

4.4. 1. 1. 1.4 Establish a baseline APRM and LPRM neutron flux noise value at a point within 5~

RATED THERMAL POWER of the 100K rated rod line with total core flow between 35K and 50K of rated total core flow during startup testing following each refueling outage.

""If not performed within the previous 31 days.

SUSQUEHANNA - UNIT 2 3/4 4-la

100 Figure 3.4.1.1.1-1 THERMAL POWER/CORE FLOW LIMITATIONS 90 C5 so 7O

<o 60 0

50 CL 4o A

30 I-0 20 10

R EGI ON G R EATER THAN LIMIT

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"" REGION LESS THAN LIMIT::::

0 30 36 40 46 50 56 BO Core Row (% RATED)

B6 70 Figure 3.4.1.1.1-1 THERMAL POWER/CORE FLOW LIMITATIONS SUSQUEHANNA - UNIT 2 3/4 4-1b Amendment No. 58

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

< 80K of the rated pump speed and a.

the following revised specification limits shall be followed:

1.

Specification

2. 1.2:

the MCPR Safet;y 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 000 3.

5.

I Specification 3.2.2:

the APRM Setpoints shall be as follows:

I Tri Set oint Al 1 owabl e Value 5 M)T

~R)T SRB

< (0.58W + 45K)T SRB

< (0.58W + 48K)T I

Specification 3.2.3:

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

a.

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

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 + 36o Allowable Value 000 b.

C.

b.

APRM-Flow Biased Tri Set oint Al 1 owabl e Va'lue

< 0.58W + 45o APRM and LPRM""" neutron flux noise levels shall be less than three times their established baseline levels when THERMAL POWER is greater than the limit specified in Figure 3.4. 1. 1. 1-1 Total core flow shall be greater than or equal to 42 million lbs/hr when THERMAL POWER is greater than the limit specified Figure 3.4. 1. 1. 1-1.

APPLICABILITY:

OPERATIONAL CONDITIONS 1* and 2~, except during two loop operation.¹ ACTION:

a.

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

SUS(UEHANNA - UNIT 2 3/4 4-lc Amendment No.58

g

'I REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION Continued) b.

C.

d.

With any of the limits specified in 3/4.1.1.2a not satisfied:

l.

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

2.

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

With the APRM or LPRM*"* neutron flux noise levels greater than or equal to three times thei r established baseline levels when THERMAL POWER is greater than the limit specified in Figure 3.4. l. 1. 1-1, immediately initiate corrective action and restore the noise levels to within the required limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months by returning the reactor to a THERMAL POWER/core flow condi tion less than or equal to the limit specified in Figure 3.4. 1. 1. 1-1.

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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

With total core flow less than 42 million lbs/hr when THERMAL POWER is greater than the limit specified in Figure 3.4. 1. 1.2-1, immediately initiate corrective action by either:

1.

Returning the rector to a THERMAL POWER/core flow condition less than or equal to the limit specified in Figure 3.4. 1. 1. 1-1 within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , or 2.

Increasing total core flow to greater than or equal to 42 million lbs/hr within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE RE UIREMENTS

4. 4. 1. 1. 2. 1

4. 4. l. 1. 2. 2 Upon entering single loop operation and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter, verify that the pump speed in the operating loop is

< 80% of the rated pump speed.

With THERMAL POWER greater than the limit specified in Figure 3.4. 1. 1. 1-1, determine the APRM and LPRM~~" neutron flux noise levels within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

Continue to determine the noise levels at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and within 30 minutes after the completion of the THERMAL POWER increase 5% of RATED THERMAL POWER.

4.4.1.1.2.3 Within 15 minutes prior to either THERMAL POWER increase resulting from a control rod withdrawal or recirculation loop flow increase, verify that the following differential temperature requirements are met if THERMAL POWER is

< 30%~*** of RATED THERMAL POWER or the recirculation loop flow in the operating recirculation loop is

< 50%*"*" of rated loop flow:

SUS(UEHANNA - UNIT 2 3/4 4-1d Amendment No. 58

~

REACTOR COOLANT SYSTEM SURVEILLANCE RE UIREHENTS Continued 4.4. 1. 1. 2. 4

4. 4. 1. 1. 2.

5'.4.

l. 1. 2. 6

4. 4. 1. 1. 2. 7

4. 4. 1. l. 2. 8 4.4. 1. 1. 2. 9 a.

< 1454F between reactor vessel steam space coolant and bottom head drain line coolant, b.& < 504F between the reacto~ coolant within the loop not in operation and the coolant in the reactor pressure

vessel, and c.fP

< 504F between the reactor coolant within the loop not in operation and operating loop.

a.

Establish a baseline APRH and LPRH neutron flux noise value at a point within 5X RATED THERMAL POWER of the 100K rated rod line with total core flow between 35K and 50K of rated total core flow during startup testing following each refueling outage, or b.

In lieu of establishing a single loop operation baseline value, utilize the value established pursuant to Specification 4.4.1. 1. 1.4 if a baseline value is needed to meet the requirements of Specification 3.4. l. 1.2.

The pump 'discharge valve and bypass valve in both loops shall be demonstrated OPERABLE by cycling each valve through at least one complete cycle of full travel during each star tup"" prior to THERMAL POWER exceeding 25K of RATED THERMAL POWER.

The pump discharge bypass valve in the OPERABLE loop, if not OPERABLE, shall be verified to be closed at least once per 31 days.

The pump HG set scoop tube electrical and mechanical stop shall be demonstrated OPERABLE with overspeed setpoints less than or equal to 102.5X and 105K, respectively, of, rated core flow, at least once per 18 months.

The pump discharge valve and bypass valve in the inoperable loop, if not OPERABLE, shall be verified to be closed at least once per 31 days.

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

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

b.

The indicated total core flow differs by more than lOX from the established total core flow value from single recirculation loop flow measurements.

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

REACTOR COOLANT SYSTEM SURVEILLANCE RE UIREMENTS Continued c.

The indicated diffusar

-to-lower plenum differential pressure of any individual jet pump differs from estab-lished single recirculation loop patterns by more than 10K.

.4.4.1.1.2.10 The SURVEILLANCE REQUIREMENTS associated with the specifications referenced in 3.4.1.1.2a shall be followed.

See Special Test Exception

3. 10.4.

If not performed within the previous 31 days.

'I Oetectors A and C of one LPRM string per core octant plus detectors A and C of one LPRM string in the center of the core should be monitored.

Initial value.

Final value to be determined based on startup testing.

Any required change to this value shall be submitted to the Commission within 90 days of test completion.

See Specification 3.4. l. 1. 1 for two loop operation requirements.

This requi~ement does not apply when the loop not in operation is isolated from the reactor pressure vessel.

Ouring startup testing following each refueling outage, data shall be recorded for the parameters listed to provide a basis for establishing the specified relationships.

Comparisons of the actual data in accordance with the criteria listed shall commence upon the performance of subsequent required survei llances.

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

26

I

~

'/4. 2 POWER OISTRIBUTION 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 loss-of-coolant accident will not exceed the limit spec-ified in 10 CFR 50 46.

The peak cladding temperature (PCT) following a postulated loss-of-coolant accident is primarily a function of the average he=t 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 Specification 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 Figure 3.2.1-1.

The calculational procedure used to establish the APLHGR shown on Figure 3

~ 2.1-1 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 KN-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 instruments limit plant operations to the region covered by the transient and accident analyses.

In addition, the APRM setpoints must be adjusted to ensure that >V'lastic strain and fuel centerline melting do not occur during the worst anticipated operational occurrence (AOO), including transients initiated from partial power operation.

For ANF fuel the T factor used to adjust the APRM setpoints is based on the 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 ANF's Protection Against Fuel Failure (PAFF) line shown in Figure 3.4 of XN-NF-85-67(A), Revision 1.

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

SUS(UEHANNA - UNIT 2 B 3/4 2-1 Amendment No. 58

POWER DISTRIBUTION LIMITS BASES 3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRs at steady state operating conditions as speci-ified 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 requi red that the resulting MCPR does not decrease below the Safety Limit MCPR at any 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 tr ansient, 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 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 parameters 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 pressurization and non-pressurization events are described in XN-NF-79-71 and XN-NF-84-105..

The princi-pal result of this evaluation is the reduction in MCPR caused by the transient.

Figure 3.2.3-1 defines core flow dependent MCPR operating limits which assure that the Safety Limit MCPR wi 11 not be violated during a flow increase tran-sient resulting from a motor-generator speed control fai.lure.

The flow depend-ent 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 wi 11 not be violated in the event of a Feedwater Controller Failure, Rod Withdrawal 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 core wide tran-sients to determine thermal margin.

Additional analyses are performed to determine the MCPR operating limit with either the Main Turbine Bypass inoperable 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.

SUS(UEHANNA - UNIT 2 B 3j4 2-2 Amendment No. 58

POWER DISTRIBUTION LIMITS BASES MINIMUM CRITICAL POWER RATIO (Continued)

At THERMAL POWER levels less than or equal to 25% of RATED THERMAL POWER, the reactor wi 11 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 indicates that the re-sulting 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 wi 11 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 lin'ear heat generation even if fuel pellet densification is postulated.

SUSQUEHANNA - UNIT 2 B 3/4 2"3 Amendment No. 5B

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 Specification 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 conditions.

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's not exceeded for any Anticipated Operational Occurrence (AOO).

For single loop operation, the RBM and APRM setpoints are adjusted by a 8.5%

I decrease in recirculation drive flow to account for the active loop drive flow that bypasses the core and goes.

up through the inactive loop jet pumps.

Surveillance on the pump speed of the operating recirculation loop is imposed to exclude the possibility of excessive reactor vessel internals 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 General Electric Service Information Letter No. 380, Revision 1, "BWR Core Thermal Hydraulic Stability," dated February 10, 1984.

An inoperable jet pump is not, in itself, a sufficient reason to declare a re-circulation loop inoperable, but.it does, in case of a design basis accident, increase the blowdown area and reduce the capability of reflooding the core;

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

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

prescribed schedule for significant degradation.

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

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

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

In order to prevent undue stress on the vessel nozzles and bottom head region, the recirculation loop temperatures shall be within 50'F of each other prior to startup of an idle loop.

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

Since the coolant in the bottom of the vessel is at a lower temperature than the coolant in the upper regions of the core, undue stress on the vessel would result if the temperature difference was greater than 145'F.

SUS(UEHANNA - UNIT 2 B 3/4 4-1 Amendment No. 58

REACTOR COOLANT SYSTEH BASES 3/4.4.2 SAFETY/RELIEF VALVES The safety valve function of the safety/relief valves operate to prevent the reactor coolant system from being pressurized above the Safety Limit of 1325 psig in accordance with the ASHE Code.

A total of 10 OPERABLE safety/relief valves is required to limit reactor pressure to within ASHE III allowable values for the worst case upset transient.

Demonstration of the safety/relief valve liftsettings will occur only during shutdown and will be performed in accordance with the provisions of Specifica-tion 4.0.5.

3/4.4.3 REACTOR COOLANT SYSTEH LEAKAGE 3/4. 4. 3. 1 LEAKAGE DETECTION SYSTEHS The RCS leakage. detection systems required by this specification are provided to monitor and detect leakage from the reactor coolant pressure boundary.

3/4. 4. 3. 2 OPERATIONAL LEAKAGE The allowable lcekage rates from the reactor coolant system have been based on the predicted and experimentally observed behavior of cracks in pipes.

The normally expected background leakage due to equipment design and the detection capability of the instrumentation for determining system leakage was also considered.

The evidence obtained from experiments suggests that for leakage somewhat greater than that specified for UNIDENTIFIED LEAKAGE the probability is small that thc imperfection or crack associated with such leakage would grow rapidly.

However, in all cases, if the leakage rates exceed the values specified or the leakage is located and known to be PRESSURE BOUNDARY LEAKAGE, the reactor will be shutdown to allow further investigation and corrective action.

The Surveillance Requirements for RCS pressure isolation valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystcm LOCA.

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

Chloride limits are specified to prevent stress corrosion cracking of the stainless steel.

The effect of chloride is'not as great when thc oxygerr cmeentration in thc coolant is low, thus the 0.2 ppm limit on chlorides is ~ittcd during NEER OPERATION.

During shutdown and refueling operations, ~ temperature necessary for stress corrosion to occur is not present so a 0.5 ppe conccntrition of chlorides is not considered harmful during these periods.

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

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/+a ~ ~, ~g FIGURE 5.1.3-1b MAP DEFINING UNRESTRICTED AREAS FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS SUSQUEHANNA - UNIT 2 5-5

OESIGN FEATURES

5. 3 REACTOR CORE FUEL ASSEMBLIES 5..3. 1 The reactor core shall, contain 764 fuel assemblies with each fuel assembly containing 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 ROO ASSEMBLIES 5.3.2 The reactor core shall contain 185 control rod assemblies, each consisting of a cruciform ar ray of stainless steel tubes containing 143 inches of boron carbide,

84C, powder surrounded by a cruciform shaped stainless steel sheath.

5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE ANO 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.

C.

For a pressure of:

1.

1250 psig on the suction side of the recirculation pumps.

2.

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

For a temperature of 575'F.

VOLUME 5.4.2 The total water and steam volume of the reactor vessel and recirculation system is appr.oximately 22,400 cubic feet at a nominal T

of 528'F.

ave SUS(UEHANNA - UNIT 2 5-6 Amendment No. 58

ML17156B445

Text

Attachment:

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