ML17146A796

Amend 64 to License NPF-14,revising Tech Spec to Change MAPLHGR & Min Critical Power Ratio Limits,Preclude Single Loop Operation & Change Affected Tech Spec Bases

ML17146A796
Person / Time
Site:	Susquehanna  (NPF-14-A-064, NPF-14-A-64)
Issue date:	05/07/1987
From:	Butler W Office of Nuclear Reactor Regulation
To:
Shared Package
ML17146A797	List: ML17146A796 ML17146A798
References
NUDOCS 8705150398
Download: ML17146A796 (38)

Text

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UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 PENNSYLVANIA POPPER

& LIGHT COMPANY ALLEGHENY ELECTRIC COOPERATIVE, INC.

DOCKET NO. 50-387 SUS UEHANNA STEAM ELECTRIC STATION, UNIT I AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.

64 License No. NPF-14 1.

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

A.

The application for the amendment filed by the Pennsylvania Power Light Company (PP&L), dated December 12, 1986, 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 2.

E.

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

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

(2)

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

64, and the Environmental Protection Plan con-tained in Appendix 8 are hereby incorporated in the license.

PP&L shall operate the facility in accordance with the Technical Specifica-tions and the Environmental Protection Plan.

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3.

This amendment is effective as of the date of issuance.

FOR THE NUCLFAR REGULATORY COMMISSION

Enclosure:

Changes to the Technical Specifications Date of Issuance:

May 7, 1987

/S/

Walter R. Butler, Director Project Directorate I-2 Division of Reactor Projects Previously concurred*:

LA:BWD-3:DBL*

BWD-3:DBL EHylton/hmc MThadani 03/26/87 03/26/87 OGC*

MYoung 03/31/87 D:PDI-2:DRP WButler 05/

/87

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3.

This amendment is effective as of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

Enclosure:

Changes to the Technical

.Specifications Date of Issuance:

tiay 7, 1987 Malter R. Butler, Director Project Directorate I-2 Division of Reactor Projects I/IT

ENCLOSURE TO LICENSE AMENDMENT NO.'4 FACILITY OPERATING LICENSE NO. NPF-14 DOCKET NO. 50-387 Replace the following pages of the Appendix A Technical Specifications with enclnsed pages.

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

The corresponding overleaf, pages are also provided to maintain document completeness.

REMOVE INSERT XX1 XX11 3/4 2-3 3/4 2-4 3/4 2-5 3/4 2-6 3/4 2-8 Deleted 3/4 2-9 3/4 2-10 3/4 3-53 3/4 3-54 3/4 4-1b 3/4

. 4-1c B 3/4 2-1 B 3/4 2-2 B 3/4 4-1 B 3/4 4-2 B 3/4 7-3 B 3/4 7-4 XX1 xxii (over leaf) 3/4 2-3 3/4 2-4 3/4 2-5 (overleaf) 3/4 2-6 3/4 2-9 3/4 2-9a 3/4 2-10,"(overl ea f) 3/4 3-53 (overleaf>

3/4 3-54

-'3/4 4-1b (overleaf) 3/4 4-Ic B 3/4 2-1 (overleaf)

B 3/4 2-2 B 3/4 4-1 B 3/4 4-2 (over leaf)

B 3/4 7-3 (overleaf)

B 3/4

.7-4

LIST OF.FIGURES INDEX FIGURE

3. 1. 5"1 SODIUM PENTABORATE SOLUTION TEMPERATURE/

CONCENTRATION REQUIREMENTS.....................

PAGE 3/4 1-21

3. l. 5-2 SODIUM PENTABORATE SOLUTION CONCENTRATION.........

3/4 1-22 3.2. 1-1

3. 2. 1-2 7

3. 2. 3"1

3. 2. 3-2

3. 2.4. 2-1 3.4. 1. 1-1 3.4. 6. 1-1 B 3/4 3-1 B 3/4. 4. 6-1

5. l. 1-1 5.l. 2-1

5. 1. 3-la

5. 1. 3-lb

6. 2. 1-1

6. 2. 2-1 THIS PAGE INTENTIONALLY LEFT BLANK..

MAXIMUMAVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VS.

AVERAGE PLANAR EXPOSURE, GE FUEL TYPE 8CR233 (2. 33K ENRICHED).....

MAXIMUMAVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VS.

AVERAGE BUNDLE EXPOSURE, EXXON 8x8 FUEL.

FLOW DEPENDENT MCPR OPERATING LIMIT...

REDUCED POWER MCPR OPERATING LIMIT.,

LINEAR HEAT GENERATION RATE (LHGR) LIMIT VERSUS AVERAGE PLANAR EXPOSURE EXXON 8x8 FUEL..............

THERMAL POWER LIMITATIONS MINIMUM REACTOR VESSEL METAL TEMPERATURE VS.

REACTOR VESSEL PRESSURE...

REACTOR VESSEL MATER 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 LIQUID EFFLUENTS...

OFFSITE ORGANIZATION UNIT ORGANIZATION 3/4 2-2 3/4 2-3 3/4 2-4 3/4 2-lob 3/4 4-lb 3/4 4-18 B 3/4 3"8 B 3/4 4-7 5"2 5-3 5-5 SUSQUEHANNA - UNIT 1 Xxl Amendment No.

64

I I

I'I

LIST OF TABLES TABLE INDEX SURVEILLANCE FREQUENCY NOTATION

~ $

PAGE 1-9 1.2 2.2. 1-1 B2. 1. 2-1 OPERATIONAL CONDITIONS....................

REACTOR PROTECTION SYSTEM INSTRUMENTATION S ETPO INTS

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UNCERTAINTIES USED IN THE DETERMINATION OF 2-4 B2. 1. 2"2

3. 3. 1-1

3. 3. 1-2

4. 3. l. 1"1

3. 3. 2-1 3.3.2 2

3. 3. 2"3

4. 3. 2. 1"1 3 ~ 3 ~ 3 1 3e3o3 2

3 ~ 3 ~ 3 3

4. 3.3. 1-1 THE FUEL CLADDING SAFETY LIMIT....................

HOHINAL VALUES OF PARAMETERS USED IN THE STATISTICAL ANALYSIS OF FUEL CLADDING INTEGRITY S

e AFETY LIMIT......................................

REACTOR PROTECTION SYSTEM INSTRUMENTATION.........

B 2-3 B 2-4 3/4 3-2 REACTOR PROTECTION SYSTEM -INSTRUMENTATION SURVEILLANCE REQUIREMENTS.......;.................

ISOLATION ACTUATION INSTRUMENTATION................

ISOLATION ACTUATIOH INSTRUMENTATION SETPOINTS.....

ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME....

ISOLATION ACTUATIOH INSTRUMENTATION SURVEILLANCE REQUIREMENTS o ~ o ~ oo ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ eo ~ ~ ~ ~ ~ ~

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EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUHEATATION o ~ o ~ ~ ~ ~ ~ ~

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EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS...................

EHERGENCY CORE COOLING SYSTEM RESPONSE TIMES.

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMQlTS....

3/4 3-7 3/4 3"11 3/4 3"17 3/4 3-21 3/4 3-23 3/4 3-28 3/4 3-31 3/4 3-33 3/4 3"34 REACTOR PROTECTION SYSTEM RESPONSE TIMES..........

3/4 3"6 3.3.4. 1"1

3. 3.4. 1-2 A'MS RECIRCULAT!OH PUMP TRIP SYSTEM INSTRUMENTATION o e o ~ o ~ ~ ~ ~

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ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATIOH SETPOINTS....................

3/4 3"37 3/4 3-38 SUSQUEHANNA - UNIT 1 XXH Amendment Ho.

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PERIVIISSABLE REGION OF OPERATION:.'

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0 6000 10000 16000 20000 26000 30000 36000 40000 46000 Average 'Planar Exposure (MWD/MT)

MAXIMUIVIAVERAGE PLANAR LINEAR HEAT GENERATION RATE {MAPLHGR)VERSUS AVERAGE PLANAR EXPOSURE GE.FUEL TYPES 8CR233 {2.33% ENRICHED)

FIGURE 3.2.1-1

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".': PERIV!ISSABLE

--: REGION OF.:...:.:.:.::::::::::::::::

10.4 OPERATION 9

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0 30000 36000 6000

10000, 16000 20000 26000 Average Bundle'xposure (IVIWD/MT)

MAXIMUMAVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR). VERSUS AVERAGE BUNDLE EXPOSURE EXXON SXS FUEL FIGURE 3.2.1-2

h1

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:

~T Al 1 owabl e Value S

< (0.58W + 59X)T S < (0.58W + 62X)T SRB

< (0.58W + 50X)T SRB < (0.58W + 53X)T where:

S and SRB are in percent of RATED THERMAL POWER, W

= Loop recirculation flow as a percentage of the'loop recirculation flow which produces.

a rated core flow of 100 million lbs/hr, T (GE fuel)

= Lowest value of the ratio of FRACTION OF RATED THERMAL POWER divided by the MAXIMUM FRACTION OF LIMITING POWER DENSITY.

T is always less than or equal to 1.0.

T (Exxon fuel) = 1.0 APPLICABILITY:

OPERATIONAL CONDITION 1, 'when THERMAL POWER is greater than or equal to 25X of RATED 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 SRB, as above determined, initiate corrective action within 15 minutes and adjust S and/or

/

5 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 THERMAL PIER 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 4.2.2 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:

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'east 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 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 90X 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 100X times

MFLPD, provided that the adjusted APRM reading does not exceed 100X of RATED THERMAL POWER, the required gain adjustment increment does not exceed 10X of RATED THERMAL POWER, and a notice of the adjustment is posted on the reactor control panel.

'V See Specification 3.4.1.1.2.a for single loop operation requirements.

SUSQUEHANNA " UNIT 1 3/4 2"5 Amendment No.

56

POWER DISTRIBUTION LIMITS 3/4. 2. 3.

MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION 3.2.3 The MINIMUM'CRITICALPOWER 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 I

R TER THERMA EERIER.

ACTION:

With MCPR less than the applicable MCPR limit determined above, initiate cor-rective action within 15 minutes and restore MCPR 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 25K of RATED THERMAL POWER w'ithin 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 4.2.3. 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 <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 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 with a LIMITING CONTROL ROD PATTERN for MCPR.

d.

The provisions of Specification 4.O.4 are not POWER increase of at the reactor is operating MT applicable.

SUS(UEHANNA " UNIT 1 3/4 2"6 Amendment No.

64

THIS PAGE IS DELETED 5 ~

SUSQUEHANNA - UNIT 1 3/4 2-8 Amendment No. 64

CA 1.7 1.B CURVE A: EOC-RPT, inoperable; Main Turbine Bypass Operable CURVE B: EOC-RPT Operable; Main Turbine Bypass Operable or Inoperable U) 1.5

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B 1.33 1.29

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1.2 40 BQ '0 80 Total Core Flow (% OF RATED) 90 100 FLOW DEPENDENT MCPR OPERATING LIMIT FIGURE 3.2.3-1

1.7 CURVE A: EOC-RPT Inoperable:

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

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REDUCED POWER MCPR OPERATING LIIVIIT Figure 3.2.3-2 90 100

'OWER DISTRIBUTION LIMITS 3/4. 2.4 LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.4 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

~i Iliiii i.

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 "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 />, C.

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.

SUS(UEHANNA - UNIT 1 3/4 2"10 Amendment No. 45

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TABLE 3.3.6-1 (Continued)

~ y CONTROL ROO BLOCK INSTRUMENTATION ACTION ACTION 60 -

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

'ACTION 61 With the number of OPERABLE Channels:

a.

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

b.

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

ACTION 62, With the number of OPERABLE channels less than requir'ed by the Minimum OPERABLE Channels per Trip Function requirement, place the inoperable channel in the tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

aO b.

c d.

e.

NOTES With THERMAL POWER > 3'f RATED THERMAL POWER.

With more than one control rod withdrawn.

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

Not required when eight or fewer fuel assemblies (adjacent to the SRHs) are in the core.

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

This function shall be automatically bypassed if detector count rate is

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

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

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

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

SVS(UEHANNA " UNIT 1 3/4 3"53 Amendment No. 43

TABLE 3.3.6-2 CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS TRIP FUNCTION TRIP SETPOINT ALLOWABLE O'ALUE 2.

3.

ROD BLOCK MONITOR a.

Upscale b.

Inoperative c.

Downscale APRM a.

Flow Biased Neutron Flux - Upscale b.

Inoperative c.

Downscale d.

Neutron Flux - Upscale Startup SOURCE RANGE MONITORS a.

Detector not full in b.

Upscal e c.

Inoperative d.

Downscale INTERMEDIATE RANGE MONITORS

< 0.58 W+ 50K" NA

> 5X of RATED THERMAL POWER

< 12'f RATED THERMAL POWER

< 0.58 W+ 53'"

NA

> 3X of RATED THERMAL POWER

< 14K of RATED THERMAL POWER NA

< 2x10 cps HA

> 0.7 cps*",

NA

< 4 x 10 cps RA

> 0.5 cps~"

< 0.66 W+ 42K

< 0.66 W + 45M NA HA

> 5/125 divisions of full scale

> 3/125 of divisions full scale a 4 b.

C.

d.

Detector not full in Upscale Inoperative Downscale NA

< 108/125 divisions of full scale HA

> 5/125 divisions of full scale NA

< 110/125 divisions of full scale HA

> 3/125 divisions of full scale 5.

6.

SCRAM DISCHARGE VOLUME a.

Water Level - High

< 44. gallons REACTOR COOLANT SYSTEM RECIRCULATION FLOW

< 44 gallons a.

b.

C.

Upscale Inoperative Comparator

< 108/125 divisions of full scale

< ill/125 divisions of full scale HA HA

< 10X flow deviation

< 11K flow deviation d

R il M

'1 dk1 kf f

1 1

p (W).

The trip setting of this function must be maintained in accordance with Specification 3.2.2.

• "Provided signal-to-noise ratio is >2.

Otherwise, 3cps as trip setpoint and 2.8cps for allowable value.

¹¹See Specification 3.4. 1.1.2.a for single loop operation requirements.

. Figure 3.4.1.1.1-1 THERMALPOWER LIMITATIONS 80

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SUS(UEMANNA " UNIT 1 3/4 4-1b Amendment No.

56

I C

3

5 1

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 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 Allowable Value

.58W +

3.

Specification 3.2. 1:

The MAPLHGR limits shall be as follows:

a.

GE fuel:

the limits specified in Figure 3.2.1-1 multiplied by 0.81.

b.

Exxon fuel:

the limits specified in Figure 3.2.1-2 multiplied by 0.0.

4.

Specification 3.2.2:

the APRM Setpoints shall be as follows:

5.

Table 3.3.6-2:

follows:

Tri Set oint Allowable Value 555)T ~5 5)T SRB

< (0.58W + 46K)T

~

SRB

< (0.58W + 49K)T the RBM/APRM Control Rod Block Setpoints shall be as a.

RBM - Upscale Tri Set 'oint.

< 0.66W + 3 b.

APRM-Flow Biased Tri Set oint

< 0.58+ 46 Al 1 owabl e Value

.6

+4 Allowable Value il

. 5 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 Figurq 3/4. l. 1. 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. 1. 1-1.

APPLICABILITY:

OPERATIONAL CONDITIONS 1" and 2~, except during two loop operation.8 SUS(UEHANNA - UNIT 1 3/4 4"lc 5

"'mendment No.

.64

4

3/4.2 POWER OISTRIBUTION LIMITS BASES T>e 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 lass-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 Dn 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 ihe 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.

3. 2. 1-2 and 3 2 1-3.

The calculational procedure used to establish the APLHGR shown on Figures 3.2.1"1, 3.2.1-2 and 3:2.1-3 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 Reference 1 or 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 kPRM instru-ments limit p'lant operations to the region covered by'he transient and accident analyses.

In addition, for GE fuel, the APRM setpoints must be adjusted to en-sure that > lX plastic strain does not occur in the degraded situation.

The scram settings and rod block settings are adjusted in accordance with the formula in this specification when the combination of THERMAL POWER and MFLPD indicates a higher peaked power distribution to ensure that an LHGR transient ~ould not be increased in the degraded condition.

For the Exxon fuel, no adjustment is required since operation within the MCPR and MAPLHGR operating limits assures that fuel mechanical design criteria are not violated.

SUS(UEHANNA - UNIT 1 B 3/4 2-1 Amendment No. 45

I C

I II

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-%he 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'n 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 a Exxon-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 4n MCPR caused by the transient.

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 ass'ures that the Safety Limit MCPR will not be exceeded.in the event of a feedwater'controller failure 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 aqd Main Turbine Bypass inoperable have not been performed.

Therefore, operation in this condition is not permitted.

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 consider-able margin.

During initial start-up testing of the plant, a

MCPR evaluation SUSQUEHANNA - UNIT j.

B 3/4 2-2

-" Amendment No.

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.

For single loop operation, the MAPLHGR limits for Exxon fuel are multi-plied by a factor of 0.0.

This multiplication factor precludes extended opera-tion with one loop out of 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 reactor vessel internals vibra-tion.

Surveillance on differential-temperatures below the threshold limits on THERMAL POWER or recirculation loop flow mitigates undue thermal stress on vessel nozzles, recirculation pumps and the vessel bottom head during extended operation 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 pre-scribed in accordance with the recommendations of General Electric Service Information Letter No. 380, Revision 1, "BMR Core Thermal Hydraulic Stability,."

dated February 10, 1984.

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 inoperab1e.

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 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'tress 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 tem-perature difference was greater than 145 F.

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

" Amendment No. 64

3/4.4 REACTOR COOLANT SYSTEM BASES Continued 3/4. 4. 2 SAFETY/RELIEf VALVES The safety valve function of the safety/relief valves operate to prevent the rea'ctor coolant system from being pressurized above the Safety Limit of 1325 psig in accordance with the ASME Code.

A total of 10 OPERABLE safety-relief valves is required to limit reactor pressure to within ASME III allow-able values for the worst case upset transient.

Demonstration of the safety/relief valve lift settings will occur only during shutdown and will be performed in accordance with the provisions of Specification 4.0.5.

3/4'.4.3 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.3.

1 LEAKAGE DETECTION SYSTEMS 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 leakage 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 smally that the 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 arid..known to be PRESSURE BOUNDARY LEAKAGE, the reactor will be shutdown to allow further invest'igation 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 intersystem LOCA.

3/4.4. 4 CHEMISTRY The water chemistry limits of the reactor coolant system are established to prevent damage to'he 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 the oxygen concentration in the

'oolant is low, thus the 0.2 ppm limit on chlorides is permitted during POWER OPERATION. 'uring shutdown and refueling operations, the temperature necessary for stress corrosion to occur is not present so a 0.5 ppm concentration of chlorides is not considered harmful during these periods.

SUSQUEHANNA " UNIT 1 B 3/4 4-2 Amendment No. 56

l l

~ I 1

~

PLANT SYSTEMS BASES 3/4 7.4 SNUBBERS (continued)

The service life of a snubber is evaluated via manufacturer input and information through consideration of the snubber service conditions and asso-ciated -installation and maintenance records (newly installed snubber, seal

replaced, spring replaced, in high radiation area, in high temperature

area, etc...).

The requirement to monitor the snubber service life is included to ensure that the snubbers periodically undergo a performance evaluation in view of their age and operating conditions.

These records will provide statistical bases for future consideration of snubber service life.

The requirements for the maintenance of records and the snubber service life review are not intended to affect plant operation.

';3/4 7. 5 SEALED SOURCE CONTAMINATION The limitations on removable contamination for sources required leak testing, including alpha emitters, is based on 10 CFR 70.39(c) limits for plutonium.

this limitation will ensure that leakage from byproduct,

source, and special nuclear material sources will not exceed allowable intake values.

Sealed sources are classified into three groups according to their use, with surveillance requirements commensurate with the probability of damage to a source in that group.

Those sources which are frequently handled are required to be tested more often than those which are not.

Sealed sources which are continuously enclosed within a shielded mechanism, i.e., sealed sources within radiation monitoring or boron measuring

devices, are considered to be stored and need not be tested unless they are removed from the shielded mechanism.

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, C02 systems, Halon systems and fire hose stations.

The collective capability 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.

Mhen 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.

SUSQUEHANNA - UNIT 1 B 3/4 7-3 Amendment No.36

PLANT SYSTEMS BASES 3/4 7.6 FIRE SUPPRESSION SYSTEMS (continued)

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 RATEO ASSEMBLIES The OPERABILITY of the fire barriers and barrier penetrations ensure 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 barr ier penetrations for conduits, cable trays and piping, fire windows, fire dampers, and fire doors are periodically inspected to verify their OPERABILITY.

3/4 7.8 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 the cycle specific transient analysis.

SUS(UEHANNA - UNIT 1 B 3/4 7-4

'Amendment No.

64