ML18040B194

Forwards Application for Proposed Amend 58 to License NPF-22,changing Tech Specs to Support Cycle 3 Reload.Unit Scheduled to Shutdown on 880305 & Restart on 880503.Reload Summary & Transient Analysis Repts Encl.Fee Paid

ML18040B194
Person / Time
Site:	Susquehanna
Issue date:	12/23/1987
From:	Kenyon B PENNSYLVANIA POWER & LIGHT CO.
To:	Butler W Office of Nuclear Reactor Regulation
Shared Package
ML17146B090	List: ML17146B089 ML17146B092 ML17146B094 ML17146B095 ML17146B096 ML17146B102 ML18040B194
References
PLA-2953, NUDOCS 8712310143
Download: ML18040B194 (62)

Text

'(.'CELEMTED DJ i IBUTJON DEMONS ~TION SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:8712310143 DOC.DATE: 87/12/23 NOTARIZED: YES DOCKET FACIL:50-388 Susquehanna Steam Electric Station, Unit 2, Pennsylva 05000388 AUTH. NAME AUTHOR AFFILIATION KENYON,B.D. Pennsylvania Power & Light Co.

RECIP.NAME RECIPIENT AFFILIATION BUTLER,W.R. Project Directorate I-2 p ~g I

SUBJECT:

Forwards application for Proposed Amend 58 to License NPF-22,changing Tech Specs to support Cycle 3 reload.

DISTRIBUTION CODE: AOOID TITLE: OR COPIES RECEIVED: LTR Submittal: General Distribution ENCL i SIZE: +  !,

NOTES:1cy NMSS/FCAF/PM. LPDR 2cys Transcripts. 05000388 RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD1-2 LA 1 0 PD1-2 PD 5 5 THADANI,M 1 1 A

INTERNAL: ACRS 6 6 ARM/DAF/LFMB 1 0 NRR/DE ST/ADS 1 '1 NRR/DEST/CEB 1 1 NRR/DEST/MTB 1 1 NRR/DEST/RSB 1 I NRR/DOEA/TSB 1 1 NRR 8/ILRB 1 1 'D OGC/HDS2 1 0 EG FI 01 1 1 RES/DE/EIB 1 1 8

EXTERNAL: LPDR 2 2 NRC PDR 1 1

., NSIC 1 1 NOTES: -. 3 3 R

8 A

'D 8

TOTAL NUMBER OF COPIES REQUIRED: LTTR 30 ENCL 27

a lt II

~ ~

Pennsylvania Power 8 Light Company

, Two North Ninth Street,~ Allentown, PA 18101 ~ 215/7706151 I

OEG 23 1987 Bruce D. Kenyon Senior Vice President-Nuclear 21 5/770-41 94 Director of Nuclear Reactor Regulation Attention: Dr. W. R. Butler, Project Director Project Directorate I-2 Division of Reactor Projects U.S. Nuclear Regulatory Commission Washington, D.C. 20555 SUSQUEHANNA STEAM ELECTRIC STATION PROPOSED AMENDMENT 58 TO LICENSE NO. NPF-22:

UNIT 2 CYCLE 3 RELOAD SUBMITTAL PLA-2953 FILES R41>>2, A17-2, A7-8C Docket No. 50-388

Dear Dr. Butler:

The purpose of this letter is to propose changes to the Susquehanna SES Unit 2 Technical Specifications in support of the ensuing Cycle 3 reload. Changes to the following Technical Specifications are requested:

Index 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 3/4.3.6 Control Rod Block Instrumentation 3/4.4.1 Recirculation System B 2.1 Safety Limits B 3/4.2.1 Average Planar Linear Heat Generation Rate B 3/4.2.2 APRM'etpoints

~

B 3/4.2.3 Minimum Critical Power Ratio B 3/4.4.1 Recirculation System The following attachments to this letter are provided to illustrate and technically support each of the changes:

Marked-up Technical Specification Changes No Significant Hazards Considerations PL-NF<<87-007 "Susquehanna SES Unit 2 Cycle 3 Reload Summary Report",

December 1987 Susquehanna SES Unit 2 Cycle 3 Proposed Startup Physics Tests Summary Description, November 1987 ANF-87-125, Revision 1, "Susquehanna Unit 2 Cycle 3 Plant Transient Analysis", November 1987 ANF-87-126, Revision 1, "Susquehanna Unit 2 Cycle 3 Reload goo I Analysis", November 1987 t 8712310i43 871223 PDR ADOCK 0500038)

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DEC 23 l98i, 2 FILES R41-2, A17-2, A7-8C PLA-2953 Dr. W. R. Butler Susquehanna SES Unit 2 is currently scheduled to be shutdown for refueling and inspection on March 5, 1988 and to restart as early as May 3, 1988. We request that your approval be conditioned to become effective upon startup after this outage, and we will keep you informed of any schedule changes.

Any questions with respect to this proposed amendment should be directed to Mr. R. Sgarro at (215) 770-7916. Pursuant to 10CFR170, the appropriate fee is enclosed.

Very truly yours, B. D. Kenyon Sr. Vice President-Nuclear Attachments cc:i NRC Document Control Desk (original) g NRC Region I Mr. J. Stair, NRC Resident Inspector-SSES Mr. M. C. Thadani, NRC Project Manager-Bethesda Mr. T. M. Gerusky, Pennsylvania DER

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8712310143l INDEX I

BASES SECTION PAGE 3/4. 0 I APP LICAB L ITY. B 3/4 0-1 3/4. 1 REACTIVITY CONTROL SYSTEMS 3/4. l. 1 SHUTDOMN MARGIN... B 3/4 1-1 3/4. 1. 2 REACTIVITY ANOMALIES...., .. B 3/4 1-1 3/4. l. 3 CONTROL RODS. 8.3/4 1"2

. 3/4.1.4 CONTROL ROD PROGRAM CONTROLS........ ~.......,... B 3/4 1-3 3/4. 1.5 STAHDBY LIQUID CONTROL SYSTEM.....,............. B 3/4 1"4 7.

3/4. 2 POWER DISTRIBUTION LIMITS 3/4.2. 1 AVERAGE PLANAR LINEAR HEAT GENERATION R ATE ~ ~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ B 3/4 2-1 3/4.2.2 APPM SETPOINTS ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ B 3/4 2"2 I 3/4. 2.,3 MIHIMUM CRITICAL POWER RATIO. B 3/4 2"A ~

3/4.2.4 LINEAR HEAT GENERATION RATE .........,... B 3/4 2-JS B 3/4. 3 INSTRUMENTATION 3/4. 3. 1 REACTOR PROTECTION SYSTEM INSTRUMENTATION... B 3/4 3-1 3/4. 3. 2 ISOLATION ACTUATION INSTRUMENTATION....,.... B 3/4 3"2 3/4. 3~ 3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION.... B 3/4 3-2 3/4. 3. 4 RECIRCULATION PUtIP TRIP ACTUATION INSTRUMENTATION. 8 3/4 3-3 3/4. 3. 5 REACTOR CORE ISOLATION COOLINGSYSTEM ACTUATION INSTRUMEHTATIOH. B 3/4 3-4 3/4. 3. 6 CONTROL ROD BLOCK INSTRUMENTATION. B 3/4 3-4

' SUSQUEHANNA " UNIT 2 '11

l I 0

0

INDEX LIST OF FIGURES FIGURE PAGE

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

CONCENTRATION REQUIREMENTS .. 3/4 1-21

3. l.'5" 2 SODIUM PENTABORATE SOLUTION CONCENTRATION 3/4 1-22

3. 2. 1-1 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE'(MAPLHGR) VS. AVERAGE PLANAR EXPOSURE, GE FUEL TYPE 8CR183 (1.83K ENRICHED) 3/4 2"2

3. 2. 1-2 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VS. AVERAGE PLANAR EXPOSURE, GE FUEL TYPE 8CR233 (2.33K ENRICHED) ................ 3/4 2-3

3. 2. 1-3 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION

~it RATE (MAPLHGR) VS. AVERAGE BUNDLE EXPOSURE, REF 9x9 FUEL .............. ..... ... . 3/4 2-4 3.2. 2-1 LINEAR HEAT GENERATION RATE FOR APRM SETPOINTS VERSUS AVERAGE PLANAR EXPOSURE, ~~C-RlVF FvE.L

...,..... 3/4 2-6a 3~ 2. 3" 1 FLOW DEPENDENT MCPR OPERATING LIMIT.. 3/4 2"8 3.2.3 2 REDUCED POWER MCPR OPERATING LIMIT.............. 3/4 2-9

3. 2. 4. 2" 1 LINEAR HEAT GENERATION RATE (LHGR) L:MIT VERSUS AVERAGE PLANAR EXPOSURE, &HNN'x9 FUEL 4~~ .......... 3/4 2-10b

/coze t=L.oW

3. 4. 1. 1" 1 3i 4.1mZ THERMAL POWERALIMITATIONS

>>+"~ ~aP O4'8Rw'i m

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TH&RPl~l Po~6R u~~TATious 3 4 4 lb

3. 4. 6. 1" 1 MINIMUM REACTOR VESSEL MIFTAL TEMPERATURE VS.

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

4. 7. 4" 1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST .... 3/4 ?"15 B 3/4 3"1 REACTOR VESSEL WATER LEVEL ..... B 3/4 3-8 B 3/4.4.6"1 FAST NEUTRON FLUENCE (E>1MeV) AT 1/4 T AS A FUNCTION OF SERVICE LIFE B 3/4 4"7

5. 1. 1-1 EXCLUSION AREA ....... ~ 5" 2

5. l. 2-1 LOW POPULATION ZONE . 5-3

5. l. 3-la MAP DEFINING UNRESTRICTED AREAS FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS

5. 1. 3" lb MAP DEFINING UNRESTRICTED AREAS FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS 5-5 SUSQUEHANNA - 'UNIT 2 xx11 Amendment No. 3)

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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 MCP is not less than the limit specified in Specification 2. l. 2 for both GE and n.. fuel. MCPR greater than the specified limit represents a conser-vative 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. Al-though 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 pro" duct migration from cladding 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 incre-mental cladding deterioration. 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 antici-pated operational occurrences, at least 99.9X of the fuel rods in the core do not experience transition boiling (ref. XN-NF-524(A)).

2.'l. l THERMAL POWER Low Pressure or Low Flow ~<'P~~C ~l+h ENSE' calcu 's The use of the XN-3 correlation is not valid for all critical powe at pressures below 785 psig or core flows less than 1 of rated flow. Ther e, the fuel cladding integrity Safety Limit i ablished by other means. Ths 's done by establishing a limiting c tion on core THERMAL POWER with the llowing basis. Since the essure drop in the bypass region is essentially all e tion head, th re pressure drop at low power and flows will always be greater n 4 si. Analyses show that with a bundle flow of 28 x 10~ lbs/hr, bu essure drop is nearly independent of bundle power and has a value .5 psi. the bundle flow with a 4. 5 psi driving head will be gre than 28 x 10'bs/ . Full scale ATLAS test data taken at pressures m 14.7 psia to 800 psia indicat at the fuel assembly critical powe this flow is approximately 3.35 MWt. Wi e design peaking ors, this corresponds to a THERMAL POWER of more tha X of RATED THE POWER. Thus, a THERMAL PO'WER limit of 25K of RATED THERMAL for eactor ressure below 785 si is conservative.

SUSQUEHANNA " UNIT 2 B 2-1 Amendment No. 31

7 he use.of PAe A'rer 3 corre,la~i'on t's va. lid for crier'ce.l power calcaladi'ons at pressures gree,A<

+Aan 5'$'0 psi'~ artd 4nn die tttass f luxes

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o/louring basis lrovr Peel +ha~ tAe wat'el. level i'n t'Ae vessel

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+ lux c on t'ic'7ti'o n, Car +h et RN< '9X 9 foe I elespr>>

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+he mt'rtirnutvt bundle 5 lit ni t's g rea, ter tea vt

~P> ooo /hs/Pr r or tx ll ctvs /gxtz v +Ae cc o/trank Flow an d xrtaxlrnuN f/ow o.ree is sue,g

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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 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 transition (MCPR = 1. 00) and the Safety Limit MCPR is based on a de-tailed statistical procedure which considers the uncertainties in monitoring the core ooerating state. One specific uncertainty included in the safety limit is the uncertainty inherent in the XN-3 critical power correlation. XN-NF-524 describes the methodology used in determining the Safety Limit MCPR.

The

.X,HsE.RV S.

XN"3'critical power cor lation is based on a significant body of prac-tical test data, providing a 'gh degree of assuranCe that the critical power as evaluated by the correlat'on is within a small percentage of the actual criti-cal power being estimated. e assumed reactor conditions used in defining the safety limit introduce conservatism into the limit because bounding high radial power factors and bounding flat local peaking 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 sus-tained operation at the Safety Limit MCPR there would be no transition boiling in the core. If boiling transition were to occur, here 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 or-ganizations indicate that the use of a boiling transition limitation to protect against cladding failure is a very c'onservative approach. Much of the data in-dicates that LMR fuel can survive for an extended period of time in an environ-ment of boiling transition.

SUSQUEHANNA - UNIT 2 8 2-2 Amendment No. 31

As long as WAe core pressure and + lons o,re.

will'n +de ra,nate of ya. Ii Wy I'l 'of VAe XS-3 cor r8 la1 ~~n (reFer to Seci~/o n 0 4 I./)>

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION Rowan (RPI R6R~Q 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION ~&4e~&AVBQ~ttNBtC" PLANAR EXPOSURE

~NHI'VERAGE shall not exceed the limits shown in Figures 3.2. 1-1, 3.2.1-2, 3.2.1-3." +4>~ BCS<et awd ~QE~6E. Q,QNhl P. pgpyS+gg and 6'AlF +Me/

APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or

• ACTION:

With an APLHGR exceeding the limits of Figure 3. 2. 1-1, 3. 2. 1-2, or 3. 2. 1-3, initiate corrective action within 15 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 25K of RATED THERMAL POWER within the next 4 hours.

SURVEILLANCE RE UIREMENTS 4.2. 1 All APLHGRs shall be verified to be equal to or less than the limits determined from Figures 3. 2. 1-1, 3. 2. 1-2, and 3.2. 1-3:

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

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

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

d. The provisions of Specification 4.0.4 are not applicable.

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

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

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0 5000 10000 15000 20000 25000 30000 35000 40000 Average Bundle Exposure {MWD/MT)

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

A/

I.

I P

t

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-+ 59K)T

< (0.58W + 50K)T SRB SRB 0'58W + 53 T where: S and S are in percent of RATED THERMAL POWER, B

W = too/recirculation 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. %here:

a~ The FRACTION OF LIMITING POWER DENSITY (FLPD) for GE fuel is the actual LINEAR HEAT GENERATION RATE'(LHGR) divided by 13:4 per Specification 3.2.4.1, and RNP

b. The FLPD for~m 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 E E 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 as above determined, initiate corrective action within 15 minutes and adjust 3 B,and/or SRB to be consistent with the Trip Setpolnt 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 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 4.2.2 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

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

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

SUSQUEHANNA - UNIT 2 3/4 2-5 Amendment No. 3l

I

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~ ~ ~ ~ ~ ~ 8.3 00" 20000 30000 40000 60000 PIanar Exposure {MWD T) 'verage I

LI R HEAT GENERATION RATE FOR APRM S POINTS VERSUS AVERAGE PLANAR EXPOSURE EXXON FUEL FIGURE 3.2.2-1

~cp(KgeJ vent.g Ivzup p,cyan, p z SUSQUEHANNA - UNIT 2 3/4 2-6e Amendment Np 3]

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8.3 10000 20000 30000 40000 50000 Average Planar Exposure (M WD/MT)

LINEAR HEAT GENERATION RATE FOR APRM SETPOINTS VERSUS AVERAGE PLANAR EXPOSURE ANF FUEL FIGURE 3.2.2-1

1.7 CURVE A: EOC-RPT tnoperabfe; Mafn Turbine Bypa:ss Operable CURVE 8: Main Turbine Bypass fnoperable; EOC-RPT Operable RVE C: EOC-RPT and Main Turbine By ass Operable Ul 1.5 C

CL O A 1.4 CL I

(3 CO 1.31 1.3 1.30 C

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FLOW DEPENDENT MCPR OPERATING LIMIT FIGURE 3.2.3-1

1.7 AD m

CURVE A: EOC-RPT Inoperable:

Main Turbine Bypass Operable CURVE 8: Main Turbine Bypass Inoperab e; EOC-RPT Operable CURVE C: EOC-RPT and Main Turb e Bypass Operable 1.6

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CURVE A: EOC-RPT Inoperable:

Main Turbine Bypass Operable CURVE 8: EOC-RPT Operable: Main 1.6 Turbine Bypass Inoperable CURVE C: EOC-RPT and Main Turbine Bypass Operable (25,1.52)

{40,1.50) g) 1.5 (65,1.47)

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CL (25,1.44)

(40,1.42)

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(s,. )

(65,1.34) 1.34 1.32 (75,1.32) 1.2 20 30 40 50 60 70 80 90 100 Core Power (% OF RATED)

REDUCED POWER MCPR OPERATING LIMIT Figure 3.2.3-2

POWER DISTRIBUTION LIMITS 3/4.2.4 LINEAR HEAT GENERATION RATE VNF FUEL LIMITING CONDITION FOR OPERATION puP 3.2.4.2 The LINEAR HEAT GENERATION RATE (LHGR) for MC.fuel shall not exceed the LHGR limit determined from Figure 3. 2. 4. 2-1.

APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or 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 />.

SURVEILLA'NCE RE UIREMENTS A~F 4.2.4.2 LHGRs forM& fuel shall be determined to be equal to or less than the 1 imi t:

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 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 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 2-10a Amendment No.31

r

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.. 0.0, 13.0

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. PERMlSS E:.

REGl OF CQ OP ATlON 8 ~ ~ ~ ~ ~ ~ ~ ~

4Q ~ ~ 48,000; C 7.72 10000 20000 30000 400 60000 Average Planar Exposure {MID/MT)

L(NEAR HEAT GENERATlON RATE (LHGR) LlMlT VERSUS AYERAGE PLANAR EXPOSURE EXXON &X9 FUEL FlGURE 3.2.4.2-]

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6) P ERMISSABLE (9 h REGION OF ~ ~

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6 0 10000 20000 30000 40000 50000 Average Planar Exposure (MWD/MT}

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

I' TABLE 3.3.6-2 CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE ROD BLOCK MONITOR

a. Upscal ett 0.66 W+ 42X < 0.66 W + 45K

b. Inoperative NA NA

c. Downs cal e > 5/125 divisions of full scale > 3/125 of divisions full scale

2. APRH

a. Flow Biased Neutron Flux Upscale'~ < 0.58 W + 50K* < 0.58 W + 53K~

b. Inoperative NA NA

c. Downscale > SX of RATED THERMAL POWER > 3X of RATED THERMAL POWER

d. Neutron Flux - Upscale Startup < 12K of RATED THERMAL POWER < 14K of RATED THERHAL POWER

3. SOURCE RANGE MONITORS

a. Detector not full in NA NA

b. Upscale < 2 x 10 cps <4xlO cps C. Inoperative NA NA

d. Downsca1e ) 0 7 cps')k > 0.5 cps*"

4. INTERMEDIATE RANGE MONITORS ao Detector not full in NA NA

b. Upscale < 108/125 divisions of full scale < 110/125 divisions of full scale C. Inoperative NA NA

d. Downscale > 5/125 divisions of full scale > 3/125 divisions of full scale

5. SCRAM DISCHARGE VOLUME

a. Water Level High < 44 gallons < 44 gallons

6. REACTOR COOLANT SYSTEM RECIRCULATION FLOW

a. Upscal e < 108/125 divisions of full scale < ill/125 divisions of full scale

b. Inoperative NA NA

c. Comparator < lOX flow deviaticn < llX flow deviation The Average Power Range Monitor rod block function is varied as a function of recirculation loop flow (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, 3 cps as trip setpoint and 2.8 cps for allowable value.

HSee Specification 3.4.1.1.2.a for single loop operation requirements.

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 operationy and:

the, +Cat.+

a..

v t Total core flow shall be greater than or equal to aa sF 0+C, &La~ Ccsadi+ic3u

~ million lbs/hr, or RMAL POWER~ less than or equal to the limit specified in Figure 3. 4. 1.1. 1. 1" APPLICABILITY: OPERATIONAL CONDITIONS 1" and 2", except during single loop operation.4 ACTION:

a. With one reactor coolant system recirculation loop not in operation, comply with the requirements of Specification 3. 4. 1. 1. 2, or take the associated ACTION.

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. l. 1. 1-1, and initiate l measures to place the unit in at least STARTUP within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTOOWN within the ne'xt 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. ghee We,chi:+oY 0+ o

c. With two reactor coolant sys recirculation loo in operation and total core flow less than million lbs/hr and HERMAL POWER greater than the limit specified in Figure 3.4.1.1.1-1:

e s+tsVC +be C4.'43.C+n+ +o ct.

4 /covC Slo~

Cash d>t 43>

e

}

1. less than or equal to the limit specified in Figure 3.4.1.1.1-1, or F <<s'e e~'~ <<Mti+low I

2. Increase core flow to greater than 4 million lbs/hr, or

3. Determine the APRM and LPRM""" neutron flux noise levels within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, 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 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and i'. 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 <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> by increasing core flow to greater than ml >on s r,

~c4~~W i~> <v e and/or by uc4Ae~rf-TitERM~CMN

+4 a. 'YH+g~AL ~~</co<e less than or equal to the limit specified in bloc c.a~d- i+ Fi gure 3.4. l. 1. 1-1. I "See Special Test Exception 3.10.4.

"""Detectors 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.

OSee Specification 3.4.1.1.2 for single loop operation requirements.

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

Figure 3.4.1.1.1-1 THERMAL POWER LIMITATIONS 80 REGION GR TER THAN UMIT 70 C} 0 4 LU ~ I

~~ eo 60

( \ p r ~ ~

'0 40

)

h E 30 "REGION LESS THAN MIT 20 L

0 J O I ~

10

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0 20 30 40 60 80 70 80 Core Row (% RAYED)

SUSIlUEHAHHA " UNIT 2 3/4 4-1b Amendment H0..26

'

-4

Eigure 3'.4..1.1.1 1 THERMAL POWER/CORE FLOW LIMITATIONS 80

---

.- REGION GREATER .:-"--..: .

C5 70 THAN LIMIT

'<

j:

>o rp o 40 CD 30 I L

f- 20 REGION LESS THAN LIMIT L

10 0

20 30 40 60 60 70 80 Core Flow (% RATED)

0 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 < 40K of the rated pump speed, and Bo&~

a. the following revised specification limits shall oe followed:

l. 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

< 0.58W + 55 < 0.58W + 58 .

3.

4.

Specification 3.2.

o~d Specification 3.2.2:

~ 1: The HAPLHGR the APRH limits shall be the limits specified Fl uvc.

4-.I 8.2.1-3

.Ri Setpoints shall be as follows: 'mvltiq6ed 4y INSERT S.

S <

SRB (0.58W + 55K)T

< (0.58W + 46K)T SRB

<

'l)

Allowable Value (0.58W + 49K)T

'L

.0 ~

4x Table 3.3.6-2:

fo 1

'1 ows:

the RBM/APRM Control Rod Block Setpoints shall be as a., RBM - Upscale Tri Set oint Allowable Value

< 0.66W + 3 < 0.66W + 40 k-.a;-1-and~~?-shaR-be-used ie-eonjunet+o~~4e-M

b. APRM-Flow Biased Tri Set oint Allowable Value

< 0.58W + 46

b. APRM and LPRM""" neutron flux noise levels shall be less than three times their established baseline levels when THERMAL POWER is greater than the limit specified in Figure 3/4. l. 1.2-1.

2

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.Z-1.

z APPLICABILITY: OPERATIONAL CONDITIONS 1" and 2", except during two loop oper ation. 0 ACTION:

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

SUS(UEHANNA " UNIT 2 3/4 4-lc Amendment No. 31

0 C

I

Speci&ico+'aaa ~.2.>: T48 PIINI&UM CRI'TICAL PowFRI RIA~ID (Wc~IRI sI a.ll 4e cgeoaew +I o.~

oe eqao.( ao <Ne law's+ as wl e salia~lugaolaes:

o., h . 3 1

)

b +he 8C'Pki de+e>yniNed Svo~ Figure

~

pIus a.al ~a.Zd C. <4,l %CYAN> d,eke>mi~ed &&0~ ~ iqwwe. E.Z.z-2.

@~AS o. 0 h ~

4 I

t k

a

REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION Continued

b. With any of the limits s'pecified in 3/4.1. 1.2a not satisfied:

C

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

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

C. With the APRM or LPRM""" neutron flux noise levels greater than or equal to three times their established baseline levels when THERMAL POWER is greater than the limit specified in Fig" ure 3, 4.1. 1. -1, immediately initiate corrective action and res ore e noi'se levels to within the requi'red limits within

.

2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> by initiating an orderly reduction of THERMAL POWER to~

+

less than or equal to the limit specified in Figure 3.4. 1. 1.< l.

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

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

e. With total core flow less than 42 million lbs/hr when THERMAL POWER is greater thorn the limit specified in Figure 3~4.1.1.<l, immediately initiate corrective action by either:

1. Reducing THERMAL POWER to less than or equal to the limit specified in Figure 3.4.1.1.W1 within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />,'r p

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

SURVEILLANCE RE UIREMENTS 4.4. l. 1.'2. 1 'pon entering single loop operation and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter, verify that the pump speed in the operating loop is < Sf% of the rated pump speed.

8O'Po 4.4. l. l. 2. 2 With THERMA OWER greater than the limit specified in Fig-ure 3.4.1.1. -1, determine the APRM and LPRM""" neutron flux noise levels within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. - Continue to determine the noise levels at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and within 30 minutes after the completion of the THERMAL POWER increase ) 5X 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 30K""*" differential temperature.

requirements are met if THERMAL POWER is <

in the of RATED operating THERMAL POWER or the recirculation loop fTow recirculation loop is < 50K"""" of rated loop flow:

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

0

Figure 3.4.1.1.2-1 SINGLE LOOP OPERATION THERMAL POWER LIMITATIONS 80

.--. REGION GREATER 70 THAN LIMIT I

~ 60

~O 50 0 40 P 30 20 REGION LESS THAN LIMIT O 10 20 30 40 50 60 70 80 Core Flow {% RATED)

3/4. 2 POMER DISTRIBUTION LIMITS BASES The specifications of this section assure that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the 2200 F limit specified in 10 CFR 50.46.

3/4.2. 1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE This specification assures that the peak cladding temperature following the postulated design basis 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. for GE fuel, the peak clad temperature is calculated assuming a LHGR for the highest powered rod which is equal to 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 peaki.ng factor. The Technical Specification 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 2200 F. The Technical Specifi-cation A or fuel is specified to assure the PCT following a postu-lated 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 Fig-ures 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 con-sistent 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 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 >1% plastic strain and fuel centerline melting do not occur during the worst anticipated operational occurrence (AOO), including transients initiated from partial power operation.

Ruf For d~~ fuel the T factor used to adjust the APRM setpoints is based on the FLPD 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 PNF SExxen-'s Protection Against Fuel Failure (PAFF) line shown in Figure 3.4 of XN-NF-85-67 Revision 1. Figure 3. 2. 2-1 corresponds to the ratio of PAFF/1. 2 un er w hach cladding and fuel integrity is protected during AOO's.

SUS(UEHANNA " UNIT 2 B 3/4 2-1 Amendment No. 31

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POWER OISTRIBUTION LIMITS BASES APRH SETPOINTS (Continued)

For GE fuel the T factor used to adjust the APRH setpoints is based on the FLPD calculated by dividing the actual LHGR by the LHGR limit specified for GE fueh in Specification 3.2.4.1.

3/4. 2. 3 HINIMUM 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 HCPR, 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 Specifica-tion 2.2.

To assur e that the, fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting transients have been analyzed to determine which result in the largest reduction in CRITICAL POWER RATIO (CPR). The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease. The limiting transient yields the largest delta MCPR. When added to the Safety Limit MCPR, the required minimum operating limit HCPR of Specification 3.2.3 is obtained and presented in Figure 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 e-i~em 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 HCPR caused by the transient.

Figure 3.2.3-1 defines core flow dependent MCPR operating limits which assure that the Safety Limit HCPR will not be violated during a flow increase tran-sient resulting from a motor-generator speed control failure, The flow depend-ent HCPR 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 HCPR operating limit which assures that the Safety limit HCPR will not be violated in the event of a feedwater controller failure 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 Hain Turbine Bypass inoperable have not been performed. Therefore, operation in this condition is not permitted.

SUSQUEHANNA - UNIT 2 B 3/4 2"2 Amendment No. 31

I 41

3/4.4 REACTOR COOLANT SYSTEM BASES 3/4. 4. 1 REC I RCULAT ION SYSTEM Operation with one reactor recirculation loop inoperable has been evaluated and found acceptable, provided that the unit is operated in accordance wi th Specification 3.4.1 1.2. ~

des-exuded-operatien~~n~o~s~

~+68 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 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 Febru-ary 10, 1984.

An inoperable jet pump is not, in itself, a sufficient reason to declare a re-circulation loop inoperable, but it does,'in case of a design basis accident, increase the blowdown area and reduce the 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 differ-ence was greater than 145'F.

SUSQUEHANNA - UNIT 2 B 3/4 4-1 Amendment No. 31

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Attachment to PLA-2953 Page 1 of 4 NO SIGNIFICANT HAZARDS CONSIDERATIONS The following three questions are addressed for each of the proposed Technical Specification changes:

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

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

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

S ecificatioa 3/4.2.1, Avera e Planar Linear Heat Generation Rate The changes to this specification reflect editorial changes to correct misarranged wording that was issued with Amendment 31, and the replacement of references to "Exxon" with "ANF". A change to increase the allowed exposure for GE 2.33X enriched fuel to 40,675 MWD/MTU is also proposed.

No. The editorial changes to correct misarranged wording and the vendor reference are wholly editorial in nature and therefore have no impact on any safety analysis.

The change to the GE limit is based on a GE LOCA analysis. This new curve was previously approved by the NRC in Amendment 64 to the Unit 1 Operating License, it is a fuel-dependent limit, and is being applied to the same type of GE fuel in this Unit 2 proposal. As stated in the staff safety evaluation for Amendment 64, "The resulting peak cladding temperature (PCT) limit and local oxidation fraction were calculated by GE based on the same plant conditions and systems analysis used to derive the current MAPLHGR limits defined in the SSES FSAR. The calculated values are well within the 10CFR50.46 Appendix K limits."

These conclusions still apply.

No. The editorial changes cannot create new concerns; based on the methods and results of the GE analysis discussed above, no new events are postulated due to the extended burn-up limit.

No. The editorial changes have no safety impact. The previously approved methods and results of the GE analysis ensure that the margin of safety is not reduced due to the change in the GE fuel MAPLHGR limit.

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Attachment to PLA-2953 Page 2 of 4 S ecification 3/4.2.2, APRM Set pints All proposed changes to this specification are editorial.

No. The proposed changes correct the vendor reference from "Exxon" to "ANF". This has no impact on safety analyses since administrative in nature.

it is entirely II. No. See I above.

III. No. See I above.

S ecification 3/4.2.3, Minimum Critical Power Ratio The changes to this specification reflect the results of the cycle-specific transient analyses.

No. Limiting core-wide transients were evaluated with ANF's COTRANSA code (see Summary Report Reference 18) and this output was utilized by the XCOBRA-T methodology (see Summary Report Reference 19) to determine delta CPRs. Both COTRANSA and XCOBRA-T have been approved by the NRC in previous license amendments. All core-wide transients were analyzed deterministically (i.e., using bounding values as input parameters).

Two load events, Rod Withdrawal Error and Fuel Loading Error, were analyzed in accordance with the methods described in XN-NF-80-19 (A)

Vol. 1" (see "Summary Report 'Reference 15) . This methodology has been approved'by the NRC.

Based on the above, the methodology used to develop the new operating limit MCPRs for the Technical Specifications does not involve a significant increase in the probability or consequences of an accident previously evaluated.

No. The methodology described can only be evaluated for its affect on the consequences of analyzed events; it cannot create new ones. The consequences of analyzed events were evaluated in I above.

No. As stated in I above, and in greater detail in the attached Summary Report, the methodology used to evaluate core<<wide and local transients is consistent with previously approved methods and meets all pertinent regulatory criteria for use in this application. Therefore, its use will not result in a significant decrease in any margin of safety.

S ecification 3/4.2.4, Linear Heat Generation Rate All proposed changes to this specification are editorial.

No. The propose'd changes correct the vendor reference from "Exxon" to "ANF". This has no impact on safety since in nature.

it is entirely administrative

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III. No. See I above.

S ecification 3/4.3.6, Control Rod Block Instrumentation The proposed change to this specification is editorial and was previously submitted to the NRC via proposed amendment 52, dated June 30, 1987.

No. The proposed change restores footnote "////" to Trip Function 2a.

This footnote was always meant to apply in this location. This change has no impact on safety since it is entirely editorial in nature.

II. No. See I above.

III. No. See I above.

S ecification 3/4.4.1, Recirculation S stem

a. Two Loop Operation: The changes to these requirements are due to the cycle specific stability analysis. The new analysis resulted in a varying "detect and suppress" region flow boundary, which in turn resulted in the need for the editorial changes to the action statements.

No. COTRAN core stability calculations performed for U2C3 predict stable reactor operation outside of the detect and suppress region of operation in SSES Unit 2. The detect and suppress region is defined by the area above and to the left of the 80% Rod Block line, the 45X constant flow line, and the line connecting the 66X Power/45X Flow, 69%

Power/47X Flow points extrapolated to the APRM Rod Block line.

Operation outside or on the boundary of this region is supported by COTRAN calculations which result in decay ratios of less than or equal to 0.75 as required by the NRC SER on COTRAN (see Summary Report Reference 14). This region is slightly larger than the region previously specified for SSES Unit 2. The results of this analysis are presented in Summary Report Reference 4.

PP&L has performed a stability startup test in SSES Unit 2 during initial startup of Cycle 2 to demonstrate stable reactor operation with ANF 9x9 fuel. The test results (see Summary Report Reference 7) show very low decay ratios with a core containing 324 ANF 9x9 fuel assemblies.

Based on the above, operation within the limits specified by the proposed Technical Specifications will not significantly increase the probability or consequences of unstable operation.

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Attachment to PLA-2953 Page 4 of 4 II. No. The methodology described above can only be evaluated for its affect on the consequences of unstable operation; it cannot create new I

events. The consequences were evaluated in above.

III. No. The methodology used to determine the regions of potentially unstable operation and stable operation were based on the guidance provided in the NRC SER for COTRAN. Also, SSES Unit 2 Technical Specifications have implemented surveillances for detecting and suppressing power oscillations. This along with the tests and analyses described in I above assures SSES Unit 2 complies with General Design Criteria 12, Suppression of Reactor Power Oscillations. Therefore, the proposed change will not result in a significant decrease in safety margin.

b. Single Loop Operation: The proposed changes reflect the changes submitted in support of Cycle 2 operation (reference proposed amendment 52 to License No. NPF-22, dated June 30, 1987), which is still pending with the NRC. The only change not explicitly evaluated in that submittal was the cycle-specific single loop MCPR

' I. No.

limit, and an administrative change to the Single Loop Operation (SLO) figure on Thermal Power Limitations.

The new MCPR limit is a result of the SLO analysis discussed in the attached ANF report, ANF-87-125. The 0.01 MCPR penalty during SLO is still proposed. The change to the figure number is entirely editorial

.

in nature and therefore has no impact on safety.

II. No. See I above.

III. No. See I above.

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