ML17157A244

Proposed Tech Specs to Support Cycle 6 Reload

ML17157A244
Person / Time
Site:	Susquehanna
Issue date:	07/02/1990
From:	PENNSYLVANIA POWER & LIGHT CO.
To:
Shared Package
ML17157A243	List: ML17157A242 ML17157A244 ML17157A245 ML18026A235
References
IEB-90-002, IEB-90-2, NUDOCS 9007120261
Download: ML17157A244 (38)

Text

SUSQUEHANNA SES UNIT 1 CYCLE 6 TECHNICAL SPECIFICATION CHANGES e

June 1990 PENNSYLVANIA POWER 5 LIGHT COMPANY

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INOEX LIST OF FIGURES FIGURE PAGE

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

CONCENTRATION REQUIREMENTS ........... 3/4 1-21

3. 1. 5-2 SOO IUM P EN TABORAT E SOLUTION CONCENTRATION 3/4 1-22

3. 2. 1-1 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (HAPLHGR) VS. AVERAGE BUNOLE EXPOSURE, ANF BxB FUEL 3/4 2-IE

3. 2. 1-2 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (HAPLMGR) VS. AVERAGE BUNOLE EXPOSURE, ANF 9x9 FUEL .. 3/4 2-Q

3. 2. 2-1 LINEAR HEAT GENFRATION RATE FOR APRM SETPOINTS VERSUS AVERAGE PLANAR EXPOSURE, ANF FUEL............ 3/4 2-7

3. 2. 3-1 FLOW OEPENOENT MCPR OPERATING LIMIT.... 3/4 2-9

3. 2. 3-2 REOUCEO POWER MCPR OPERATING LIMIT 3/4 2-9a

3. 2. 4-1 LINEAR HEAT GENERATION RATE (LHGR) LIHIT VERSUS AVERAGE PLANAR EXPOSURE ANF Bx8 FUEL .. 3/4 2-10b

3. 2. 4-2 LINEAR HEAT GENERATION RATE (LHGR) LIMIT VERSUS AVERAGE PLANAR EXPOSURE, ANF 9x9 FUEL. ~ 3/4 2-10c

3. 4. l. 1~1- 1 THERMAL POWER RESTRICTIONS. 3/4 4-jb

3. 4. 6. 1-1 MINIMUM REACTOR VESSEL METAL TEHPERATURE VS.

REACTOR VESSEL PRESSURE .. 3/4 4" 18 8 3/4 3-1 REACTOR VESSEL WATER LEVEL 8 3/4 3-8 8 3/4.4.6-1 FAST NEUTRON FLUENCE (E>lHeV) AT 1/4 T AS A FUNCTION OF SERVICE LIFE ~........,.............. ~ . 8 3/4 4-7

5. 1. 1-1 EXCLUSION AREA . 5-2

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

5. 1. 3-la HAP OEFINING UNRESTRICTEO AREAS FOR RAOIOACTIVE GASEOUS ANO LI(UIO EFFLUENTS . 5-4

5. 1. 3-lb HAP OEFINING UNRESTRICTEO AREAS FOR RADIOACTIVE GASEOUS ANO LI(UIO EFFLUENTS . 5-5

6. 2. 1-1 OFFSITE ORGANIZATION 6-3

6. 2. 2-1 UNIT ORGANIZATION .......... 6-4 SUS(UEHANNA - UNIT 1- xxl Amendment No. 93 I

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0 cooo 3pppQ Npoo willful]uc,>>>>n sxeo zcooo Qoooo AVeiege Quails EXPPWrO (MINDlMTI

'Re@lac>>. 4llowl<g p>>g MAXQALNAAVEAhCiE ~NAA UHEAB HEAT GEMERATIOH RATE ) YEASUS AVERAGE BUhlOll= EXPOSURE ANF SXS FOEL QGUf 3.2.1-1

14 0.0; 19,000; 13.0 13.0 13 CO~

C CD CO vCO tx: 12 CO 0 ~ ~ I 25,000; 11.5 11 35,000; 6 10.4 C 30,000; E tD 10.4 CO 10 ' PERMISSABLE 37,000; X S REGION OF 10.4 CO g OPERATION CO C

6000 10000 16000 20000 25000 30000 35000 40000 Average Bundle Exposure (MWD/MT)

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

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CURVE A: EOC-APT (noperable.

Main Turbine Bypass OperPala 1.7. EOC-APT Operable CURVE C: EQC-APT and

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FLOW DEPENDENT MGPR OPERATING LIMIT aeuRE 3.2.3->

Vei'irreo ~ii-h figurc. on Poi'iomrno pn~r r:

CURVE A: EOC-RPT Inoperable; (30, 1.92) 1.9 Main Turbine Bypass Operable CURVE B: Main Turbine Bypass Inoperable; EOC-RPT Operable 1.8 CURVE C: EOC-RPT and Main Turbine Bypass Operable E

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(61.7, 1.41) A 1.41 1.4 B 1.40 (62.6, 1.40) 1.34 (67.6, 1.3 4) 1.3 1.2 30 40 60 60 70 80 90 100 Total Core Flow (% OF RATED)

FLOW DEPENDENT MCPR OPERATING LIMIT FIGURE 3.2.3-1

1.7 CURVE A:-EOC-APT inoperable:

Main Turbine Bypass Operabl CURVE 8: Mairl Turbine Bypass lnop ble; 1.B EOC-APT Operable CURVg C: EOC-APT and Main rbine (26.1.66)

Bypass Operable (66.1.60) us.1.631 H0.1.61)

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REDUCED POWER MGPR OPEBATlNG LlMlT Figure 3.2.3-2 ggq3IQC P. 43 it~1 plg QVQ (3~ QC)(lC i .>t>"le 4 0<( P

CURVE A: EOC-RPT Inoperable:

Main Turbine Bypass Operable CURVE B: Main Turbine Bypass Inoperable; 1.6 EOC-RPT Operable CURVE C: EOC-RPT and Main Turbine Bypass Operable 1.5 O)

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REDUCED POWER MCPR OPERATING LIMIT Figure 3.2.3-2

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0 $ 0000 20000:QOlNM) 40000 Average Phn Exposure (MlND/MT)

LQ4l&R HEAT GENERATION BATE {LHGA) LIMIT VERSUS AVERAGE PLANAR EXPOSURE'NF SXB FUEl FIGURE 3.W.O->

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CO 0) 43,200; C 9.0 8

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

LINEAR HEAT GENERATION RATE (LHGR) LIMIT VERSUS AVERAGE PLANAR EXPOSURE ANF 8X8 FUEL FIGURE 3.2.4-1

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THERMAL POWER RESTRICTIONS Figure 3.4.1.1.1-1

REACTOR COOLANT SYSTEM RECIRCULATION LOOPS - SINGLE LOOP OPERATION LIMITING CONDITION FOR OPERATION

3. 4. 1. I. 2 One reactor coolant recirculation loop shall be in ooeration with the oumo soeed < 80% of the rated pump speed and the . eac:or at a

.DERMAL POWER/core flow condition outside of Regions I and II of Figure 3.4.1.1. l.-'nd he following revised soeci fication > imits shall oe fol!owed:

Speci ficat on 2.'.2: the ~CPR Safety 'mit shall oe '.ncreaszc:- '.;i

2. able 2.2. 1-1: the APRM Flow-Biased Scram Trip Setpoints snal' as fol>ows:

Trio Setpoint Allowable Value

< 0.>8W i 4

3. Specification 3. 2. 2: the APRM Setpoints shall be as follows:

Trio Set oint Allowable Value

~ TT SRB

< (O.SN + 45Z)T SRB

< (0.58W + 48Z)T Specification 3,2.3: The MINIMUM CRITICAL POWER RATIO (MCPR) shall be greater than or equal to the largest of the following values:

l,SO the MCPR determined from Figure 3.2.3-1 plus 0.01, and

c. the MCPR determined from Figure 3.2.3-2 plus 0.01.

5. Table 3.3.6-2: the RBM/APRN Control Rod Block Setpoints shall be as l fo1 1ows:

a. RQf - Upscale Tri Set aint Allowable Value

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~. APR}f.Flow Biased Tri Set oint Allowable Value

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4 APPLI CAB IL ." OPBATIONAL CONOITIONS 1 and P+, except dur ing two loop operation.l SUSQUEHANNA - UNIT 1 3/4 4-1c Amendment Ho 96 L

SAFETV LIMITS 8ASES 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. The~efo~e, the margin to boiling transition is calculated from plant operating parameters such as core power, core flow, feedwater temperature, and core power distribution. The margin for each fuel assembly is characterized. by the critical power ratio (CPR),

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

The Safety Limit MCPR assures sufficient conservatfsa in the operating MCPR limit that in the event of an anticipated operational occurrence from the limiting condition for operation, at least 99.9% of the fuel rods in the core would be expected to avoid boiling transition. The aargin between calculated boiling transition (MCPR = 1.00) and the Safety Lfaft MCPR is based on a detail-ed statistical procedure which considers the uncertainties in monitorfng the core operating state. One specific uncertainty included in the safety lfaft is the uncertainty inherent in the XN-3 critical power correlation. NN-NF-524 (A)

Revision 1 describes the methodology used in deterlining the Safety Lilit MCPR.

The XN-3 critical power correlation is based on a significant body of practical test data, providing a high degree of assurance that the critical power as evaluated by the correlation is within a small percentage of the actual critical power being estimated. As long as the core pressure and flow are within the range of validity of the XN-3 correlation (refer to Sec-tion 8 2. 1. 1), the assumed reactor conditions used in defining the safety limit introduce conservatisa into the lieft because boundfng high radial power fac-tors and bounding flat local peaking distributions are used to estimate the number of rods in boiling transition. Still further conservatfsa is induced by the tendency of the XN-3 correlation to overpredfct the number of rods in boiling transition. These conservatfsas and the inherent accuracy of the NN-3 cor~elation provide a reasonable degree of assurance that during sustained operation at the Safety Lfaft MCPR there would be no transition boiling in the core. If boiling transftfon were to occur, there is reason to believe that the integrity of the fuel would not necessarily be coeproefsed. Significant test data accumulated by the U.S. Nuclear Regulatory Comfssion and private organiza-tions indfcate that the use of a boiling transition lfaftatfon to protect against cladding failure fs a very conservative approach. Much of the clata indicates that LQR fuel can survive. for an extended period of tfee fn an environment of boiling transition.

yaNF guet is iaoo>M absig Ite wg a'.iifie ppivea. ~diabAvi'-

ti t this ooitek'hab ]wiahei sutViaieot- h~ooetm to paeoiooie the I

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>P ~ M epcW hW4C h4% li~'~a ~,om /el bW4 Ir+A~) iWq~ef Sc (~)due.% Si~ Nob ~~ $ 6 Iat4.'Pg iS e I d-l~HiCe, Jb4$ ~ Jipvf w]'5 the, ~9 eri3i<<l a<<eo - aoroek%ia ie caoaeovoWVe ui ih Ieoreat'* oivoooel bow ~nla44 the eo  ! 2-2 Aaaehent No. >0 SUS)UEHANNA yt gQ UNIT 1 8ailetie h'o 'iao-e1 eotiti.4 "~ 8 of thor+el ieoar> caoaot by chwei Goe ibov." 3/4.2 PAAR QISTRIBUTIQN LIMITS BASES The specifications of this section assure that the peak cladding tempera-ture following the postulated design basis loss-of-coolant accident will not, exceed the 22004F licit specified in 10 CFR 50.46. 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE This specification assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limit specified in 10 CFR 50. 46. The peak cladding temperature (PCT) following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is dependent only secondarily on the rod to rod power distribution within an assembly. The Technical Specifi- I ation APLHGR for ANF fuel is specified to assure the PCT following a postulated LOCA wi ll not exceed the 2200 F litiit. The liiiiting value for APLHGR is shown in Figures 3.2.1-1 and 3.2.1-2. I The calculational procedure used to establish the APLHGR shown on Figures 3.2.1-1, and 3.2.1-2 is based on a loss-of-coolant accident analysis. I The analysis was performed using calculational models which are consistent with the requi repents of Appendix R to '10 CFR 50. These eodels are described in XN-NF-80-19, Volumes 2, ZA, ZB and ZC. 3/4.2. 2 APRN SETPOINTS The flow biased simulated thermal power-upscale scram setting and flow biased simulated thermal power-upscale control rod block functions of the APRM instru-ments 11lIit plant operations to the region covered by the transient and accident analyses. In addition, the APRM setpoints aust be adjusted to ensure that >IX plastic strain and fuel centerline melting do not occur during the worst anticipated operational occurrence (AOO), includin'g transients initiated from partial power operation. For ANF fuel the T factor used to adjust the APN setpoints is based on the FLPO calculated by dividing the actual LHGR by the LHGR obtained froa 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 NN-NF-85-67(A), Revision 1. Figure 3.2.2-1 corresponds to the ratio of PAFF/1.2 under which cladding and fuel integrity is protected during AOOs. SUSQUEHANNA - UNIT 1 B 3/4,2-1 Aaendment No. 90 0 / ISSUER DISTRIBUTION LIMITS BASKS 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 integri ty Safety Limit MCPR, and an analysis of abnormal operational transients. For any abnormal operating transient analysis evaluation with the initial con-dition of the reacto~ being at the steady state operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip setting. given in Specification 2.2. To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient,, the most limiting tran-sients have been analyzed to determine which result in the largest reduction in CRITICAL POWER RATIO (CPR). The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease. The limiting transient yields the largest delta MCPR. When added to the Safety Limit MCPR, the required efnfam 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 begfns with the system initial param-eters shown in the cycle specific transient analysfs report that are input to an ANF core dynalic behavior transient computer program. The outputs of this program along with the initial HCPR form the input for further analyses of the thermally limiting bundle. The codes and methodology to evaluate pressuriza-tion and non-pressurization events are described in XN-NF-79-71 and XN-NF-84-105. The principal result of this evaluation is the reduction in MCPR caused by the transient. Figure 3.2.3-1 defines core flow dependent MCPR operating limits which assure that the Safety Lfaft HCPR will not be exceeded during a flow increase transient resulting froe a motor-generator speed control failure. The flow dependent MCPR is only calculated for the manual flow control mode. Therefo~e, automatic flow control operation fs not permitted. Ffgure 3.2.3-2 defines the power dependent HCPR operating lfaft whi h assures that the S fety Liwit MCPR will not be exceeded in the even f edwater ntroller ilu initfate froa a reduced power condition. R &<or ~h Tiykihc 0 Cycle specific analyses are per forme or ~ most a ting local and core wide transfents to determine thermal margin. Addftfonal analyses are performed to determine the HCPR operating lfaft with either the Hain Turbine Bypass in-operable or the EOC-RPT fnoperable. Analyses to determine thersal margin with both the EOC-RPT inoperable and Hain Turbine Bypass inoperable have not been performed. Therefore, operation in thfs condition is not perwftted. At THERMAL POWER levels less than or equal to 25X of RATKO THERMAL POWER, the reactor wi 11 be operatfng at mfnimul recirculation pump speed and the moderato void content wf]1 be very small. For all designated control rod patterns which may be employed at this point, operating plant experience indi-cates that the resulting HCPR value is fn excess of requfreaents by a consider able margin. Ouring initial start-up testing of the plant, a HCPR evaluation SUSQUEHANNA - UNIT 1 B 3/4 R-2 Aaandeent No. 9O POWER OISTRIBUTION LIMITS SASES MINIMUM CRITICAL POWER 'RATIO (Cont'f nuad) will be made at 2N of RATED THERMAL POWER level with minimum recirculation pump speed. The MCPR margin will thus be demonstrated such that future MCPR evaluation below this po~er level will be sho~n to be unnecessary. The daily requirement, for calculating MCPR when THERMAL POWER is greater than or equal to 2W of RATEO THERMAI. POWER is sufficient since power distribution snif.s are very slow when there have not been significant power or control rod cnangas. The requirement for calculating MCPR when a limiting control rod gattarn is approached ensures that NCPR will be known following a cnange in ;HERMAL POWER or power shape, regardless of magnitude, that cauld place operation at a tnarmai 1 imi t. 3/4.2.4 L NEAR HEAT GENERATION RAT< This specification assures that the Linear Heat Gener at.'an Rata (':-:GRi 'n any fuel rod is less than the aes gn linear heat generation even '.f fuel w'el pellet densification is postulated. Re a

1. General E1ectrsc for Loss-of-Coolant Ana'lysi s in Accordance x K, NEQE-20566, Novemoer 1975.

SUSQUEHANNA UNIT 1 a 3/4 2-a Amendment 9o. 57 lI ~ 'g ll~( s, IDcp Opal Qi lcR p ~t bl - addi ivlC 7 . ~ec,7z o'~ 7eclr'culprit'loo ~uprp Glrive Qel2urc p0-lde~j. is ) "',cr -pr~~~ jg aqle-loop ~itn'n ~svatjnq MC'~ lim'it ) f.pQ the cadio lactic;.I eo sequent'es of d p p gael q~e 8<<ld~o~ P ~ >ln lQ foQ7. Car~d<ctonS 0<< ~q: q, ~.I -.-.~+;-~ o K'~~1Q .-I~ l,neS. 3/4.4 REACTOR COOLANT SYSTEM aASES 3/4.4. 1 RECIRCULATION SYSTEM OperatiOn with One reaCtOr reCirCulatian lOOp inaperable haS been eValuated and fOund aCCeptable, prOVided that tne unit iS Operated in aCCOrdanCe with Specification 3. 4. 1. 1. 2. LOCA analyses for two loop operating conditians, wnich result in ~eak Cladding Temperatures (PCTs) below 2200 F, bound single loop aoerating conditions. Single loop ooeration LOCA analyses using two-loop HAPLHGR wi:s "esult fn !ower PCTs. Therefore, ne use of twa-laoo HAPLHGR !~mits "ur ~g >i~gle 'aoo aoeration assures that the PC during a LOCA event remai"s ."e::w )2~0oc -xe urVINUH CRI '.GAL , OWER RATIO (MCPR) limits for single laao .era SSSure that :ne Safety '.mit HCPR is ~ot exceeded for any Antic!bated

3. =X "ecrease in recirculation drive flow to account far the active laoo ".rive "aw that bypasses the care and goes up through the inactive loop ;et :umos.

Surveillance an the sump speed af the operating recirculation 'ooo 's 'moosed :o exc'uae'he possibility af excessive reactor vessel internals i'bra-

ion. Surveillance on differential temperatures below the threshold limi s on THERMAL POWER or recirculation loop flow mitigates undue thermal stress an veSSel nOZZ!eS, reCirCulatiOn pumpS and the VeSSel bettOm head during eXtended operation in the single loop mode. The threshold limits are those values wnich

~I11 sweep up the cold water from the vessel bottom head. Specificstions have been provided to prevent, detect,. and mitigate core thermal hydraulic instabi ligy events. These specifications are prescribed in aCCardanCe with NRC Sulletin 88-07, Supplement 1, POWer OSCi llatianS in 8ailing water Reactors (BWRs)," dated Oecember 30, 1988. The boundaries of the regions n Figure 3. 4. 1. l. 1. 1-1 are determined using ANF decay ~atio calculations and SuppOrted by SuSquehanna 565 Stability teSting. LPRM upscale alarms are required to detect reactor core thermal hydraulic instability events. The criteria for determining which LPRM upscale alarms are required iS baSed On aSSignment Of theSe alarmS tO deSignated COre ZaneS. These core zones consist of the level A, 8 and C alarms in 4 or 5 adjacent 'RM strings. The number and location of LPRM strings in each zone assure that with 50% or more af the associated LPRM upscale alarms OPERABLE sufficient monitoring capability is available to detect core wide and regional ascillations. Operating plant instability data is ueed to determine the specific LPRM strings assigned to each zone. The core zones and required LPRM upscale alarms- in each zone are specified in appropriate procedures. An inoperable jet pump is not, in itself', a sufficient gleason to declare a reCirCulatian lOOp inaperable, but it dOeS, in CaSe Of a deSign-baSiS-aCCident, increase the blowdown area and reduce the capability of reflooding the core; thuS, the requi~ement fOr ShutdOWn Of the faCility with a jet pump inaperable. -'et pump failure can be detected by monitoring jet pump performance on a prescribed schedule for significant degradation, SUSQUEHANNA - UNIT 1 8 3/4 4-1 Amendment Vo. 96 DESIGN FEATURES S. 3 REACTOR CORE FUEL ASSEutlLIES 5.3.1 The reactor core shall contain 764 fuel asseebltes with tach fuel ~ asseeoly containing 62 or 79 fuel rods and two water rods clad with Drcaloy -2. Each fuel rod shall have a noatnal ac ive fuel length of 150 inches. Reload fuel percent U-235. CONTROL ROO ASSBSLIES L 5.3.2 The reactor core shall contain MS control rod assaebltes, earth consisting of a cructfora array of stainless steel tubes containing 143 inches ( of boron carbide, 84C, powder surrounded by a crucifona snaoeo stainless steel sneath.

5. 4 REAC'OR ".VOLANT SySTEH DESIGN PRESSURE AHO TBIPERATURE 5.4. 1 The reactor coolant systole is designed and sha)1 ae eatntained:

a. In accordance with the code requt~nts specified in Secti'on 5. 2 of the FSAR, with allowance for noraal degradation pursuant to tne applicable Surveillance Requtreaents,

b. For a pressure of:

l. 1250 psig on the suction side of the recirculation pus!ps.

2. 1500 pstg face the recirCulation puap discharge to the jet p~si

c. For a tassperature of 575'F.

5.4.2 The total water and stem vol~ of the reactor vessel and recirculation system is approxtaately 22,400 cubic i'eet at a eatnal T of SN~F. SUS)UEHANNA UNIT 1 .5 6 Aa1gndrt!ent No. 72 Insert A CONTROL ROD ASSEMBLIES 5.3.2 The reactor core shall contain 185 control rod assemblies consisting of two different designs. The "original equipment" design consists of- a cruciform array of stainless steel tubes containing 143 inches of boron carbide (B4C) powder surrounded by a stainless steel sheath. The "replacement" control blade design consists of a cruciform array of stainless steel tubes containing 143 inches of boron carbide (B4C) powder near the center of the cruciform, and 143 inch long solid hafnium rods at the edges of the cruciform, all surrounded by a stainless steel sheath. NO SIGNIFICANT HAZARDS CONSIDERATIONS The following three questions are addressed for each of the proposed Technical Specification changes:

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

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

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

S ecification 3 4.2.1 Avera e Planar Linear Heat Generation Rate The changes to this specification are solely to Figure 3.2. 1-1, which provides appropriate MAPLHGR limits to bound the exposure that the ANF SxS fuel will experience during Cycle 6 operation. No. The increased allowed exposure is based on an additional MAPLHGR evaluation performed by ANF (See Summary Report Reference 3). This evaluation is consistent with previously approved methods, and ensures that the peak cladding temperature for the ANF SxS fuel remains below 2200'F, local Zr-HzO reaction remains below 178, and core-wide hydrogen producti on remains below 18 for the limiting LOCA event as required by 10 CFR 50.46. Therefore, the additional MAPLHGR limits do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. No. The analysis described above can only be evaluated for its effect on the consequences of analyzed events; it cannot create new ones. The consequences of analyzed"events were evaluated in 1. above.

3. No. As discussed in l..above, the analysis to support the MAPLHGR limits at higher exposures is consistent with previously approved methods and meets all pertinent regulatory criteria for use in this application. Therefore, the proposed change will not result in a significant decrease in any margin of safety.

S ecification 3 4.2.3 Minimum Critical Power Ratio The changes to this specification provide new operating limit MCPR curves based on cycle-specific transient analyses. No. Limiting core-wide transients were evaluated with ANF's COTRANSA code (see Summary Report Reference 29) and this output was utilized by the XCOBRA-T methodology (see Summary Report Reference 30) 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 local 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 6). This methodology has been approved by the NRC. Based on the above, the methodology used to develop the new operating limit HCPRs for the Technical Specifications does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. No. The methodology described can only be evaluated for its effect on the consequences of analyzed events; it cannot create new ones. The consequences of analyzed events were evaluated in 1. above.

3. No. As stated in l. 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.

Based on the above, the use of the methodology'tilized to produce the U1C6 MCPR operating limits will not result in a significant decrease in any margin of safety. S ecification 3 4.2.4 Linear Heat Generation Rate Proposed changes to this specification provide appropriate limits at extended burnups for ANF 8x8 fuel. No. ANF-90-018(P), Revision 1 (see Summary Report Reference 5) supports the new maximum 8X8 discharge exposure. This report demonstrat'es that margin to 8x8 fuel mechanical design limits is assured for all anticipated operational occurrences throughout the life of the fuel provided that the fuel rod power history remains within the power histories assumed in the a'nalyses. Based on the above, the U1C6 LKGR operating 1'imits do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. No. This change reflects appropriate limits which ensure compliance with all relevant fuel mechanical design criteria. Application of these limits will not create the possibility of a new or different event.

3. No. As described in 1. above, ANF-90-018(P) Revision 1 demonstrates appropriate safety margin to fuel mechanical design limits for all anticipated operational occurrences throughout the life of the fuel.

S ecification 3 4.4.1 Recirculation S stem Two Loo 0 eration The changes to this specification (i.e., Figure 3.4. 1. 1. 1-1) reflect cycle-specific stability analysis. No. COTRAN core stability calculations were performed for Unit 1 Cycle 6 to determine the decay ratios at predetermined power/flow conditions. The resulting decay ratios (See Summary Report, Reference 3) were used to define operating regions which comply with the interim requirements of NRC Bulletin No. 88-07, Supplement 1 "Power Osci llations in Boiling Water Reactors," (See Summary Report, Reference 19). As in the previous cycle, Regions B and C of the NRC Bulletin have been combined into a single region (i.e., Region II), and'egion A of the NRC Bulletin corresponds to Region I. Region I has been defined such that the decay ratio for all allowable power/flow conditions outside of the region is less than 0.90. To mitigate or prevent the consequences of instability, entry into this region requires a manual reactor scram. Region I for Unit 1 Cycle 6 has been calculated to be slightly larger than Region I for the previous cycle. Region II has been defined such that the decay ratio for all allowable power/flow conditions outside of the region (excluding Region I) is less than 0.75. For Unit 1 Cycle 6, Region II must be immediately exited if it is inadvertently entered. Similar to Region I, Region II is slightly larger than in the previous cycle. In addition to the region definitions, PP&L has performed stability tests in SSES Unit 2 during initial startup of Cycles 2 and 3 to demonstrate stable reactor operation with ANF 9x9 fuel. The test results for U2C2 (See Summary Report, Reference 20) show very low decay ratios with a core containing 324 ANF 9x9 fuel assemblies. Figure 3/4. 1. 1. 1-1 is also referenced by Specificati'on 3/4.4. 1. 1.2, which governs Single Loop Operation (SLO). The evaluation above applies under SLO conditions as well. Based on the above, operation within the limits specified by the proposed changes,will ensure that the probability,and consequences of unstable operation will not significantly increase.

2. No. The methodology described above can only be evaluated for its effect on the consequences of unstable operation; it cannot create new events. The consequences were evaluated in 1. above.

3. No. PP&L believes that the use of Technical Specifications that comply with NRC Bulletin 88-07 Supplement 1, and the tests and analyses described above, will provide assurance that SSES Unit 1 Cycle 6 will comply with General Design Criteria 12, Suppression of Reactor Power Oscillations. This approach is consistent with the SSES Unit 1 Cycle 5 method for addressing core stability (See Summary Report, References 22 and 23).

S ecification 3 4.4.1 Recirculation S stem Sin le Loo 0 eration The changes to this specification include a revised HCPR limit and correction of a typographical r error.

1. No. The revised HCPR limit reflects the result of ANF's analysis of a recirculation pump seizure accident on a generic basis for the

Susquehanna units (See Summary Report Reference 4). Past analyses of this accident utilized ANF's transient methodology to establish a delta CPR which would preclude fuel failures due to overheating or clad strain. The generic analysis performed for U1C6 and future SSES cycles used Safety Limit HCPR methodology to determine the extent of rods which might experience boiling transition should HCPR reach 0.90. This accident methodology results in increased consequences (less than 2~ of the fuel rods were calculated to experience boiling transition at the 95~ confidence level, and significantly fewer rods would be expected to fail, as opposed to none using the transient methods). This result, however, is not a significant increase in consequences when compared to LOCA results. Furthermore, it meets the regulatory acceptance criteria for radiological consequences since they are but a small fraction of 10 CFR 100 guidelines, even with the conservative assumption that all rods which experience boiling transition fail. The typographical error is an inadvertent omission of the "W" in the APRH flow biased trip setpoint. This is an editorial correction to a previously approved amendment; no technical change is being proposed. Based on the above, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. No. The analysis which supports the SLO HCPR limit revision can only be evaluated for its effect on the consequences of analyzed events; cannot create new ones. The consequences of analyzed events were it evaluated in l. above. The typographical- correction is purely administrative in nature.

3. No. See l. above. The analysis used to determine the revised SLO HCPR limit meets all pertinent regulatory requirements for use in this application, and concluded that the consequences were but a small fraction of 10 CFR 100 guidelines.

The typographical correction is purely administrative in nature. Based on the above, the proposed changes will not result in a significant decrease in any margin of safety. S ecification 5.3.1 Fuel Assemblies The proposed changes to this section delete unnecessary references to the initial core loading. No. References to the initial core loading, which has been completely discharged, are unnecessary and proposed to be deleted. The ANF-5 9x9 fuel has similar thermal hydraulic and nuclear operating characteristics to the ANF-4 9x9 design which has been previously approved by the NRC (See Summary Report Reference 7) for coresidence with the ANF 8x8 fuel that wi 11 remain in the core. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. 0

2. No. 'As described above, the ANF 9x9 fuel has been previously evaluated for coresidence with ANF Bx8 fuel in the core. No new events have been determined to result from this change.

3. No. Based on its similar operating characteristics, previously approved analyses, and the analyses and limits which are proposed in this application, the U1C6 reload fuel will not result in a significant decrease in any margin of safety.

S ecification 5.3.2 Control Rod Assemblies The changes to this specification are provided in order to recognize the replacement blade design being introduced in UlC6. No. The main differences between the replacement Ouralife 160C control blades and the original equipment control blades are:

a. the Duralife 160C control blades utilize improved B~C tube material (i.e. high purity stainless steel vs. commercial purity stainless steel) to eliminate cracking during the lifetime of the control blade; b.. the Ouralife at each edge 160C control blades utilize three solid hafnium rods of the cruciform which replace the three B4C rods that are most susceptible to cracking to increase control blade life; C. the Ouralife 160C control blades contain additional B4C tubes in place of the stiffeners, have an increased sheath thickness, utilize a full length weld to attach the handle and velocity limiter, and contain additional coolant holes at the top and bottom of the sheath which result in a crevice-free structure;

d. the Ouralife 160C control blades utilize low cobalt-bearing pin and roller materials in place of stellite which was previously utilized;

e. the Ouralife 160C control blades are longer by approximately 3.1 inches in order to facilitate fuel moves within the reactor vessel during refueling outages at Susquehanna SES; and

f. the Ouralife 160C control blades are approximately 16 pounds heavier as a result of the design changes described above..

The Duralife control blade has been evaluated to assure it has adequate structural margin under loading due to handling, and normal, emergency, and faulted operating modes. The loads evaluated include those due to normal operating transients (scram and jogging), pressure differentials, thermal gradients, seismic deflection, irradiation growth, and any other lateral and vertical loads expected for each condition. The Ouralife 160C control blade stresses, strains, and cumulative fatigue have been evaluated and result in an acceptable margin to safety. The control blade insertion,capability has been evaluated and it has been determined to be capable of insertion into the core during all modes of plant operation within the limits of plant analyses. The Duralife 160C control blade coupling mechanism is equivalent to the original equipment coupling 'mechanism and is fully compatible with the existing control rod drives'n the plant. In addition, material selected is compatible with the reactor environment. The impact of the increased weight of the control blades on the seismic and hydrodynamic load evaluation of the reactor vessel and internals has been reviewed and found to have a negligible effect on existing analyses. With the exception of the crevice-free structure and the extended handle, the Duralife 160C control blades are equivalent to the NRC approved Hybrid I Control Blade Assembly (Summary Report Reference 9). The mechanical aspects of the crevice-free structure were approved by the NRC for all control blade designs in Summary Report Reference 10. A neutronics evaluation of the crevice-free structure for the Duralife 160C design was performed by GE using the same methodology as was used for the Hybrid I control blades in Reference 9. These calculations were performed for the original equipment control blades and the Duralife 160C control blades described above assuming an array of ANF 9x9 fuel, The Duralife 160C control blade has a slightly higher worth than the original equipment design, but the increase in worth is within the criterion for nuclear interchangeability. The increase in blade worth has been taken into account in the appropriate U1C6 analyses. However, as stated in Summary Report Reference 9, the current practice in the lattice physics methods is to model the original equipment all 8 C control blade as non-depleted. The effects of control blade dep/etion on core neutronics during a cycle are small and are inherently taken into account by the generation of a target k-effective for each cycle. As discussed above, the neutronics calculations of the crevice-free structure show that the non-depleted Ouralife 160C control blade has direct nuclear interchangeability with the non-depleted original equipment all 84C design. The Duralife 160C also has the same end-of-life reactivity worth reduction limit as the all B4C design. Therefore, the Duralife 160C can be used without changing the current lattice-physics models as previously approved for the Hybrid I control blades (Summary Report Reference 9) . The extended handle and the crevice-free structure features of the Duralife 160C control blades result in a one pound increase in the control blade weight over that of the Hybrid I blades, and a sixteen pound increase over the Susquehanna SES original equipment control blades. In Summary Report Reference 9, the NRC approved the Hybrid I control blade which weighs less (by more than one pound) than the 0 lattice control blade. The basis of the Control Rod Drop Accident analysis continues to be conservative with respect to control rod drop speed since the Ouralife 160C control blade weighs less than the D lattice control blade, and the heavier 0 lattice control blade speed is used in the analysis. In addition, GE performed scram time analyses and determined that the Duralife 160C control blade scram times are not significantly different than the original equipment control blade scram times. The current Susquehanna SES measured scram times also have considerable margin to the Technical Specification limits. Since the increase in weight of the Ouralife 160C control blades does not significantly increase the measured scram speeds and the safety analyses which involve reactor scrams utilize the Technical Specification limit scram times, the safety analyses are not affected. Since the Duralife 160C control blades contain solid hafnium rods in locations where the 84C tubes have failed, and the remaining B4C rods are manufactured with an improved tubing material (high purity stainless steel vs. commercial purity stainless steel), boron loss due to cracking is not expected. PPEL plans to track the depletion of each control blade and discharge any control blade prior to a ten percent loss in reactivity worth. Therefore, the requirements of IE Bulletin 79-26, Revision 1 do not apply to the Duralife 160C control blades. Based on the discussion above, the new control blades proposed to be utilized in U1C6 do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. No. The replacement blades can only be evaluated for their effectiveness as part of the overall reactivity control system, which is evaluated in terms of analytical consequences in l. above. Since they do not cause any significant change in system operation or function, no new events are created.

3. No. The analyses described in 1. above indicate that the replacement blades meet all pertinent regulatory criteria for use in this application, and are expected to eliminate the boron loss concerns expressed in IE Bulletin 79-26, Revision 1. Therefore, the proposed change does not result in a significant decrease in any margin of safety.

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