ML18040B289

Proposed Tech Specs Including Rev 0 to EMF-1997(P)(A) Into TS Section 5.6.5 & Including Revised MCPR Safety Limits in TS Section 2.1.1.2

ML18040B289
Person / Time
Site:	Susquehanna
Issue date:	03/12/1999
From:	PENNSYLVANIA POWER & LIGHT CO.
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ML17164A990	List: ML17164A989 ML18040A969 ML18040A970 ML18040B289
References
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Download: ML18040B289 (40)

Text

ENCLOSURE C TO PLA-5040

., TECHNICAL SPECIFICATION MARK-UPs.

9903290i04 9903%2 PDR ADQCK 05000387 P PDR

~ 4 lr

2. 0 SAFETY LIHITS =(SLs)

'.1 SLs 2.1.1 Reactor Core SLs

< or core II, 2.1.1.1 With the reactor steam dome pressure 785 psig flow ( .10 million ibm/hr:

THERHAL POWER shall be. ~, 25K RTP.

2.1.1.2 With the reactor steam dome pressure a 785 psig and core flow w 10 million 1bm/hr:

I for bp HCPR shall be >

two recirculati,on loop operation or >

i Ill i 1ti l 0 p ti 2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel.

2.1.2 Reactor Coolant S stem Pressure SL Reactor steam dome pressure shall be s 1325 psig.

2.2 SL Violations With any SL violation, the following actions shall be completed within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:

2.2.1 Restore compliance with all SLs; and 2.2.2 Insert all insertable control rods.

SUSQUEHANNA - UNIT 1 2.0-1 Amendment 178

. SLs 2.0 1.40 J

• ~

J J

1.35 0 P ~ 0 0 ~ ~

'ACCE,

, ABLE

,'CP VALUES',

1.30 30,1.27,' ~ 0

' E,--

~ ~

-1.25- ~ t

~

I

~

0 50,1.22... 0 ~

J 0 ~

J J J J

1.20 . - 40J1.24 t 0 -- ~

01.17 0 0 0 0 ~ 0 0 0 h

t h M

f. 1.15 60J1.19 .. .... 90,1.13 O 108,1.1 80,1.15 1.10 0 0 0 l0 0 ~ 0 0 0 UNACCEITABL NICPR VALU 100,1.1 1.05 ~ 0 ~ ~ 0 ~0 1,00 30 40 0 60 70 80 90 , 100 110 Core F tow (M Ihr)

Figure 2.1.'t.2-1*

MCPR Safety Limit vs Core Flow Two Loop Operation

• Note: Operation to this figure is only approved for Unit 1 Cycle 11 SUSQUEHANNA - UNIT 1 2.0-2 Amendment 178

SLs 2.0 1,40 I

~ 0 ~ 0 0 1,35 0 0 P I

I I

E. ACCEPTABLE 1.30 30,1.28- - - - ~-

MCPR VALUES ',

m 1.25 '108,1.22 CO FAUNA 1.20 40,1.25 o 52,1.2 o ~ ~ ~ ~

I I I el o o 1.15 0 0

o o 1.10 PTABLE'LUES ~

MC F /

e 1.05 I I

1.00 30 40 50, 0 70 80 90 100 110 Core Flow (Illb/hr)

Figure 2.1.1.2-2*

MCPR Safety Limit vs Core Flow.

Single Loop Operation

• ote: Operation to this figure is only approved for Unit 1 Cycle 11 SUSQUEHANNA - UNIT 1 2.0-3 Amendment 178

I i>

'fl ~ J.

• eporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.4 Monthl 0 eratin Re orts Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the main steam safety/relief valves, shall be submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the report.

5.6.5 CORE OPERATING LIMITS REPORT COLR

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, 'and shall'e documented in the COLR for'he following:

1. The Average Planar. Linear Heat Generation Rate for Specification 3.2.1:

2. The Minimum Critical Power Ratio for Specification 3'.2.2; .

3. The Linear Heat Generation Rate for Specification 3.2.3;

4. The Average Power Range Monitor (APRM) Gain and Setpoints for Specification 3.2.4: and

5. The Shutdown Margin for Specification 3.1.1

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1. - F-87-001-A, "Qualification of Steady State C Physic s for BWR Design and An uly, 1988.

2. PL-N - - , "Qualification of Tra nalysis e ods for BWR Desi n and Anal sis," Jul 1 PL-NF-90-001-A, "Application of Reactor Analysis Methods for BWR Design and Analysis," July. 1992.

(continued)

SUSQUEHANNA - UNIT 1 5.0-21 Amendment 178

~ l Reporting Requirements 5.6 5.6 Reporting Requirements 5 6 5. COLR (continued)

XN-NF-80-19(P)(A), Volume 4. Revision 1. "Exxon *Nuclear.

Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads," Exxon Nuclear Company, Inc. June 1986.

XN-NF-85-67(P)(A), Revision 1. "Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel, "Exxon Nuclear Company, Inc.. September 1986.

PLA-3407, "Proposed Amendment 132 to License No.

NPF-14: Unit 1.Cycle 6 Reload, "Letter from H. W.

Keiser'(PPSL) to W..R. Butler (NRC); July 2. 199 , .as .

approved by letter from Mohan C. Thadani (NRC) o H. W Keiser (PP8L), "Cycle 6 Reload, Susquehanna earn Electric Station, Unit 1 (TAC No. 77165),"

ovember 2, 1990.

7 ~ Let r from Elinor G. Adansam (NRC) H. W. Keiser (PPEL "Issuance of Amendment No. to Facility Operati License No. NPF S quehanna Steam Electric ation, Unit.2," Oct er .3, 1986.

8. PLA-3533, Rev ed Proposed endment 67. to License No.

NPF-22: Unit 2 cle 5 Re ad, "Letter from H. W.

Keiser (PPIEL) to R. tier (NRC), March 7'. 1991, as approved by letter James J. Raleigh (NRC) to H. W Keiser (PPSL), "Sus anna Steam Electric Station, Unit 2. Cycle 5 oad TAC No. 79140)," April 22, 1991.

9. XN-NF-84-97 Revision 0, "LO -Seismic Structural Response an ENC 9x9 Jet Pum Fuel Assembly," Exxon Nuclear ompany, Inc.. December 84.

10. PLA 728, "Response to NRC Question; eismic/LOCA A lysis of U2C2 Reload," Letter from W. Keiser PP8L) to E. Adensam (NRC), September I

25, 1986.

XN-NF-82-06(P)(A), Supplement 1, Revision 2, "Qualification of Exxon Nuclear Fuel for Exten d Burnup Supplement 1 Extended Burnup Qualif'icatio of ENC 9x9 Fuel," May 1988.

(continued)

SUSQUEHANNA - UNIT 1 5 '-22 Amendment 178

Reporting Requirements 5.6 5.6 . Reporting Requirements 5.6.5 COLR (continued) $ ~ggX~a 5 $ (0h~<~bar i'LVo) t +2-.

4-, e~

XN-NF-80-19(A). Volume 1, and Volume 1 Supplements 1~

"Exxon Nuclear Methodology for Boiling Water

'Ey1 yyyl Reactors: Neutronic Methods for Design and Analysis;-"

ANF-524(P)(A); Revision 2 and Supplement 1, Revision 2.

"Advanced Nuclear Fuels Corporation Critical Power:

Methodology for Boiling-Water Reactors.",

November 1990.

ANF-1125(P)(A) and ANF-1125(P)(A), Supplement 1, "ANFB Critical Power Correlation", April 1990.

NEDC-32071P, "SAFER/GESTR-LOCA Loss of Coolant Accident Analysis," GE Nuclear Energy, Hay 1992.

NE-092-001A, Revision 1, "Licensing Topical Report for Power Uprate With Increased Core Flow," Pennsylvania Power 5 Light Company, December 1992 ~c 4 NQ c. Sq.g .

( Quw~4ac; /~i h5't3)

PL-NF-90-001. Supplement 1-A,. "Application of Reactor Analysis Methods for BWR Design and Analysis: Loss of Feedwater Heating Changes and Use of RETRAN HOD 5.1,"

August 1995.

PL-NF-94-005-P-A. "Technical Basis for SPC 9x9-2 Extended Fuel Exposure at Susquehanna SES", January, 1995.

CENPD- - , " ce Safety Re or ater Reactor Reload F n ineering ions. November 1994..

PL-NF-90-001, Supplement 2-A, "Application of Reactor Analysis Methods for BWR Design and Analysis:

CASMO-3G Code and ANFB Critical Power Correlation",

July 1996.

(continued)'USQUEHANNA

- UNIT 1 5. 0-23 Amendment 178

eporting 'equi rcments 5.6

5. 6 Reporting Requirements 5.6.5 COLR (continued)

ANF .89 98(p)(A) Revision I, and Revision 1 Supplement 1, "Generic Mechanical Design Criteria for BMR Fuel Designs," Advanced Nuclear Fuels Corporation.

May 1995.

XN-NF-81-58(P)(A) Supplements 1 and 2 Revision 2, "RO Fuel Rod Thermal-Mechanical Response Evaluation Mo May 1986.

24. XN-NF )(A), "RODEX 2A (BWR) Fuel d Thermal-Mechanical Res nse Evaluation Mode August 1986.

25. XN-NF-82-06(P)(A) an uppl'e s 2,,4 and 5 Revision 1. "Qualificati f Exxon Nuclear Fuel for Extended Burnup," Octo XN-NF-85-92(P), "Exxon Nuclear nium nia Irradiation Examina 'n 26.

Dioxide/Gad Conducti '," November 1986.

and Thermal

27. 082(P)(A) Revision 1 and Revision 1 upplement 1, "Application of ANF Design Methodolog for Assembly Reconstitution," May 1995. 'uel ANF-91-048(P)(A), "Advanced Nuclear Fuels Corporation Methodology for Boiling Water Reactors EXEM BWR Evaluation Model," January 1993.

29. -33(P)(A) Supplement 2, "HUXY: A Generali Multi ro Code with 10CRF50 Appen eatup Option," January

30. XN-CC-33(P)(A ion 1, "HU . eneralized Multirod Heat with 10CFR50 Appendix K Hea tion Users nua ." November 1975.

XN-NF-80-19(P)(A), Volumes 2, 2A, 2B, and'2C "Exxon Nuclear Methodology for Boiling Water Reactors: EXEM BWR ECCS Evaluation Model," September 1982.

35. ~ XN-NF-80-19(P)(A), Volumes 3 Revision 2 "Exxon Nuclear Methodology for Boiling Water Reactors Thermex: Thermal Limits Methodology Summary Description," January 1987.

(continued)

SUSQUEHANNA - UNIT 1 5. 0-24 Amendment 178

Reporting Requirements 5'

5. 6 Reporting Requirements COLR (cdntinued)

~ ~ i XN-NF-79-71(P)(A) Revision 2, Supplements 1, 2, and 3, "Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors," March 1986.

(1' 3 'NF-1358(P)(A), Revision 1, "The Loss of Fe'edwater Heating Transient in Boiling Water Reactors." .Septe r .

2.

35. ANF-9 P)(A) Volume 1 Revision 1 and Volume v Suppleme s 2. 3, and 4. "COTRANSA2: . A C uter Program for Boiling ater Reactor Transient An ses," August 1990.-

<> o'@

36. XN-NF-84-105(P)(A), olume 1 an and 2, "XCOBRA-T: A .C olume 1 Supplements uter ode for BWR Transient 1

~ Thermal-Hydraulic Core A sis," February 1987.

XN-NF-84-105(P)(A), plement 4,. "XCORBRA-T:

i 6) 37.

A Computer Code f BWR ume 1 Transien Thermal-Hydraulic Core d c 0 r Analysis, Void Data,." June 8.

action Model Comp ison to Experimental

• ~ 38.¹ EHF 0(P), Revision l. "Application o FB to ATRI -10 f'r. Susquehanna Reloads." March 1 el 39.¹ A-4595, "Response to NRC Request for. Additional t e Q Information on Siemens Report EHF-97-010; tr Revision 1," March 27. 1997.

~r.

~ gal C. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency .Core Cooling Systems (ECCS) limits, nuclear limits such as SDH, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the-NRC.

(continued)

SUSQUEHANNA - UNIT 1 5.0-25 Amendment 178

BASES i 'Reactor 'Core SLs B 2.1;1

'ACKGROUND Operation above the boundary of the nucleate boiling regime (continued) could result in excessive cladding temperature because of the onset of transition boiling and the resultant in heat transfer coefficient; Inside the steamsharp'eduction film, high cladding temperatures are reached, and a cladding water (zirconium water) reaction may take place. This chemical reaction results in oxidation of- the fuel cladding to a structurally weaker form. This weaker form may lose its integrity, resulting in an uncontrolled release of to the reactor coolant. 'ctivity APPLICABLE SAFETY ANALYSES

'he fuel cladding must not normal operation and AOOs.

sustain damage as a result The reactor. core SLs are of established to preclude violation of the fuel design criterion that an MCPR limit is to be established, such that at least 99.9X of the fuel rods in the core would not be expected to experience the onset of transition boiling.

. The Reactor Protection System setjoints (LCO 3.3.1.1.

"Reactor Protection System (RPS) Instrumentation" ), in combination with the other LCOs, are designed to prevent any anticipated combination of transient conditions for Reactor Coolant System water level. pressure. and THERMAL POWER level that would result in reaching the MCPR limit.

(,9 eleve,~c.e 4) +~4' 2.1.1.1 Fuel Claddin Inte rit ~

W~ 4QF5 oah P Slh hNW+-b~

'pSVo, AQUA 9 -~~ (,Le~ace~c.e i) The use of the ANFB~correlationis S. valid f'r critical power calculations at pressures > 600 psia an bundle mass fluxes

) 0.1 x.10'b/hr-ft'ef-.~.. For peration at low pressures or ow ows, the fuel cladding integrity SL is g~c O]AGQ a~h established by a limiting condition on core THERMAL POWER.

~ O.OS~K~ Ybl4-A with the following basis:

No Provided that the water level in the vessel downcomer is maintained above the top of the active fuel, natural circulation is sufficient to ensure a minimum bundle flow for all fuel assemblies that have a relatively high power and potentially can a'pproach a critical heat flux condition. For the SPC 9x9 fuel design, the minimum bundle flow is approximately 30 x 10'b/hr. For the SPC ATRIUM-10 design, (continued)

SUSQUEHANNA - UNIT 1 B 2.0-2 Revision 0

Reactor Core SLs B 2.1.1 BASES APPLICABLE 2.'1.1.1 Fuel Claddin Inte rit (continued)

SAFETY ANALYSES

'he minimum bundle flow is > 28 x 10'b/hr.'or both the SPC 9x9-2 and ATRIUM-10 fuel designs, the coolant ...

mini'mum bundle flow and maximum area are such that the mass flux is always > .25 x 10'b/hr-ft'. Full scale critical power test data taken from various SPC and GE f'uel designs at pressures from 14.7 psia to 1400 psia indicate the fuel assembly critical .power at 0.25 x 10'b/hr-ft's approximately 3.35 HWt. At 25K RTP, a bundle power of approximately 3.35 HWt ~

corresponds to a bundle radial peaking factor of 3.0, which is signif'icantly higher than the expected

- Ppeaking. factor'.. Thus., a THERMAL,.POWER limit of 25K RTP for reactor pressuies ( 785 psig is conservative.

2.'I'. 1. 2 MCPR The HCPR SL ensures sufficient conservatism in the operating HCPR limit that, in the event of an AOO from the limiting condition ot 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 (i.e.,

HCPR = 1.00) and the HCPR SL is based on a detailed

" statistical procedure that considers the uncertainties in monitoring the core operating state. One specific uncertainty included in the SL is the uncertainty. in the ANFB critical power correlation. Referencep'>~d-5 describesthe 1 used in determ'ng the HCPR SL.

,* o,4 AV 5-$ <

~

The ANFBcrs >ca power correlationm ased 'on a body of practical test data. As long as the

"'ignificant core pressure and flow are within the .range of validity of the 4MFB correlation<(refer to Section B.2.1.1.1), the assumed reactor conditions used in def'ining the SL introduce conservatism into the limit because bounding high radial power factors and bounding flat local peaking distributions are used to estimate the number of'ods in boiling transitio These conservatisms and the inherent accuracy o e ANFB~correlation~provide a reasonable degree of assurance that during sustained operation at the HCPR SL there would be no transition boiling in the core. If boiling transition were to occur, there is reason to believe that the integrity of the fuel would not be compromised.

(continued)

SUSQUEHANNA - UNIT 1 B 2.0-3 Revision 0

Reactor Core SLs B 2.1.1 BASES APPL ICABLE 2. 1. 1. 2 MCPR (continued)

SAFETY ANALYSES

'Significant test data accumulated by the NRC and'private organizations indicate that the use of a boiling transition limitation to protect against cladding failure is a,very conservative approach. Much of the data indicate that BWR fuel can survive for an extended period of time in an nt of boiling transition. 'nv a~i 't'X~'SK. 4YOi~~~o "~

PC~ ue is monitored using the ANFB Critical Power

, C ~~~~~+~~

Corr address

~

iog The effects of 'channel bow on HCPR are explicitly included in the calculation of the AHFB-. HCPR SL.

Explicit treatment of channel bow in the HCPR SL the concerns.af NRC Bulletin No., 90-02 entitled..

"Loss of Thermal Margin Caused by Channel Box Bow."

Monitoring requi red for compliance with the HCPR SL. is specified in LCO, 3.2.2, Minimum Critical Power Ratio.

2.1.1.3 Reactor Vessel Water Level During MODES '1 and. 2 the reactor vessel water level.

to be above the top of the active. fuel to provide is'equired core cooling capability. With fuel in the reactor vessel during 'periods when the reactor is shut down, consideration must be given to water level requirements due to the effect of decay heat. If the water level should drop below the top of the active i rradiated fuel during this period, the ability to remove decay heat is reduced. This reduction in cooling capability could lead to elevated cladding temperatures and clad perforation in the event that the water level becomes ( 2/3 of. the core height. The reactor vessel water level SL has been established at the top of the active irradiated fuel to provide a point that can be (continued)

SUSQUEHANNA - UNIT 1 B 2.0-4 Revision 0

eactor Core SLs B 2.1.1 BASES

..APPLICABLE 2.1. 1.3 Reactor Vessel Water Level (continued)

SAFETY ANALYSES monitored and to also provide a'dequ'at'e margin for. effective action.

SAFETY LIMITS The reactorcore SLs are established to protect the integrity of the fuel clad barrier to the release of radioactive materials to the environs. SL 2.1.1.1 and SL 2.1. 1-.2'nsure that the core operates within the fuel design criteria. SL 2. 1.1.3 ensures that the reactor vessel water level is greater than the top of the active irradiated fuel in order to prevent el'evated clad temperatures and, resultant clad perforations.

APPLICABILITY SLs 2.1.1.1, 2.1.1.2, and 2.1.1.3 are applicable in all MODES.

SAFETY LIMIT Exceeding an SL may cause fuel damage and create a potential VIOLATIONS for radioactive releases in excess of 10 CFR 100,. "Reactor Site Criteria," limits (Ref. 3). Theref'ore; it is required to insert all insertable control rods and restore compliance with the SLs within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time ensures that the operators take prompt remedial action and also ensures that the probability ot an accident occurring during this period is minimal.

REFERENCES 10 CFR 50," Appendix A, GDC 10.

ANF 524 {P){A), Revision 2, "Critical Power Methodology for Boiling Water Reactors," Supplement 1 Revision 2 and Supplement 2, November 1990.

3. 10 CFR 100.

E.tq7 -)9/7 (P)(Ãj 9m~ s o~ Q, 4@~@-ia c. r c. iP er g~cceX<g;~~ "3~i ~zgP ~~a, r ChV-xWqa(P')(@) ~ qq'le~e,~h i Q.e~igicc O,'gqq, ia cia'xc.~X Q~~e C.~ch~h~'-

l ir~gQ Cocci

~Q,% ki~g Kcs~h jr 5> 3 AXy (continued)

SUSQUEHANNA - UNIT 1 B 2.0-5 Revision 0

eactor Core SLs 8 2.1.1 BASES REFERENCES (continued)

5. Le, .

Electric Station B ram (PP8L) to NRC. "Sus est for earn I ~

Additional In . on Siemen MF-97,-010',

Parch 27. 1997. '-4596.

SUSQUEHANNA - UNIT 1 B 2.0-6 Revision 0

0 APLHGR B 3.2.1 BASES APPLICABLE Suppression Pool Cooling Mode, and Single Loop Operation SAFETY ANALYSES (SLO)). LOCA analyses were performed for the regions of the (continued) power/flow map bounded by the IOOX rod line 'and 'the APRMod block line (i.e., the ELLA region). The ELLA region is analyzed to determine whether an APLHGR multiplier as a function of'ore flow is required. The results of the analysis demonstrate the PCTs are within the 10 CFR 50.46 limit, and that APLHGR multipliers as a function of core flow are not required.

The GE and SPC LOCA analyses consider the delay in Low Pressure Coolant Injection (LPCI) availability when the unit is operating in the Suppression Pool Cooling Mode. The delay in LPCI availability is due to the time required to

'ealign valves from the Suppression Pool Cooling Mode to the LPCI mode. The results of the- analyses demonstrate that the PCTs are within the 10'CFR 50.46 limit.

Finally, the GE and SPC LOCA analyses were performed for Single-Loop Operation. The results, of the SPC analysis f'r ATRIUM'-10 fuel shows that an APLHGR limit which is 0.8 times the two-loop APLHGR limit meets the 10 CFR 50.46

. acceptance criteria, and that the PCT is. less than the limiting two-loop PCT. The results of the GE analysis shows that the two loop APLHGR limit for 9x9-2 fuel is acceptable in SLO.

The (LUAs).

per LU S

n ecification contains four The LUAs, a ABB in para

'g lead use ass eveloped using the reference core regions mits f'r the ABB The APLHGR satisfies Criterion 2 of'he NRC Policy Statement (Ref. 10).

LCO The APLHGR limits specified in the COLR are the result of the DBA analyses.

APPLICABILITY The APLHGR limits are primarily derived from LOCA analyses that are assumed to occur at high power levels. Oesign calculations and operating experience have shown that as power is reduced, the margin to. the required APLHGR limits (continued)

SUSQUEHANNA - UNIT 1 8 3.2-2 Revision 0

APLHGR B 3.2.1 BASES SURVEILLANCE SR 3.2.1.1 . (continued)

REQUIREMENTS given the large inherent margin to operating limits at low...

power levels and because;the APLHGRs must be calculated prior to exceeding 50K RTP.

REFERENCES NEDC-32071 (P), "Susquehanna Steam Electric Station Units 1 and 2: SAFER/GESTR Loss of Coolant Accident Analysis," May 1992.

Letter from C. 0. Thomas (NRC) to J. F. Quirk (GE),

"Acceptance for referencing of Licensing Topical Report NEDE-23785, Revision 1, Volume. III(P),"= 'The GESTR-LOCA and SAFER Models for the Evaluation of Loss of Coolant Accident,'une 1984.

ANF-91-048(P)(A), "Advanced Nuclear Fuels Corporation Methodology for Boiling Water Reactors EXEM BWR-Evaluation Model," January 1993.

ANF-CC-33(P)(A) Supplement 2. "HUXY: A Generalized Multirod Heatup Code with 10CFR50 Appendix K Heatup Option," January 1991.

5. XN-CC-33(P)(A) Revision 1, "HUXY: A Generalized Multirod Heatup Code'with 10CFR50 Appendix K Heatup Option Users Manual." November 1975..

XN-NF-80-19(P)(A), Volumes 2, 2A, .ZB. and 2C "Exxon Nuclear Methodology f'r Boiling Water Reactors: EXEM BWR ECCS Evaluation Model," September 1982. "

7.'.

FSAR. Chapter 4 ~

FSAR, Chapter 6.

'9. FSAR, Chapter 15 10 Final Policy Statement on Technical Specif'ications Improvements, July 22. 1993 (58 FR 39132).

CEN - - ,

Water Reactor Rel

" rence Safety Re 'g busti on Engineering ions, November 1994.

SUSQUEHANNA - UNIT 1 B 3.2-4 Revision 0

HCPR B 3.2.2, B 3.2 POWER DISTRIBUTION LIMITS B 3.2.2 MINIMUM CRITICAL POWER RATIO (HCPR)

BASES BACKGROUND MCPR is a ratio of the fuel assembly power that would result in the onset of boiling transition to the actual fuel assembly power. The HCPR Safety Limit (SL) is set such that 99.9% of the fuel rods avoid boiling transition if the'limit is not violated (refer to the Bases for SL 2.1.1.2). The operating limit MCPR is established to ensure that no fuel damage results during anticipated operational occurrences (AOOs). Although fuel damage does not necessarily occur if a fuel rod actually experienced boiling transition (Ref; critical power at which'oiling transi-tion is 1),'he calculated'o occur has been adopted, as a fuel design. criterion.,

I The onset of transition boiling is a phenomenon that is readily detected during the testing of various fuel bundle designs. 'ased on these experimental .data, correlations have been developed to predict critical bundle power (i.e.,

the bundle power level at the onset of transition boiling) for a given set of plant parameters (e.g., reactor vessel-- '.

pressure, flow, and subcooling). Because plant operating ,

conditions and bundle power levels. are monitored and determined relatively easily, monitoring the MCPR is a convenient way of'nsuring that fuel failures due to .

inadequate cooling do not occur.

... APPLICABLE The analytical methods and assumptions used in evaluating SAFETY ANAI YSE the AOOs =to establish the operating "limit:MCPR"arepresen'te'd'n References 2 through . To ensure that the HCPR SL is not exceeded during any transient event that occurs with moderate frequency, limiting transients have been analyzed to determine the largest reduction in critical power ratio (CPR). The types of transients evaluated are loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease. The limiting transient yields the largest change in CPR (hCPR). When the largest hCPR is added to the MCPR SL, the required operating limit MCPR is obtained.

The MCPR operating limits derived from the transient analysi s are dependent on the operating core flow and power (continued)

SUSQUEHANNA -" UNIT 1 B 3.2-5 Revision 0

Oy ~

HCPR B 3.2.2 BASES APPLICABLE state to ensure adherence to fuel design limits during SAFETY ANALYSES the worst transient that. occurs with moderate frequency.

(continued)

• These analyses may also consider other combinations of plant conditions (i.e., control.rod scram speed, bypass valve performance, EOC-RPT, cycle exposure. etc.). Flow dependent HCPR limits are determined by analysis of slow flow runout transients using the methodology of Reference 2.

The e contains four ABB lead use assemb The L s ed in non-1 ore regions

'LUAs).

per Specification 4.2.1. n limits for the ABB LUAs have be e using t e metho s 2 and 12.

The'HCPR satisfies Criterion 2 of the NRC Policy Statement .

(Ref. M}.

LCO The HCPR operating limits specified in the COLR are the.

result of the Design Basis A'ccident (DBA) and transient analysis. The operating limit HCPR is determined by the larger of the, flow dependent HCPR and power dependent HCPR limits.

APPLI CABI LITY The HCPR operating limits are primarily derived from transient analyses that are assumed to occur at high power levels. Below 25K RTP, the reactor is operating at a minimum recirculation pump speed and the moderator void ratio i.s small., Surveillance of, thermal limits below RTP is- unnecessary- due to the large inherent 'margin that

'5K ensures that the HCPR SL is not exceeded even transient occurs. Studies of the variation of limiting if a limiting transient behavior have been performed over. the range of power and flow conditions. These studies encompass the .

range of key actual plant parameter values important to typically limiting transients. The results of these studies demonstrate that a margin is expected between performance and the HCPR requirements. and that margins increase as power is reduced to 25K RTP. This trend is expected to (continued)

SUSGUEHANNA,- UNIT I B 3.2-6 Revision 0

t ~ N 4

HCPR

. B 3.2.2 BASES l

REFERENCES 3. PL-NF-87-001-A, "Qualification of Steady State core (continued) Physics Methods for BWR Design and Analysis."

April 28; 1988.

PL-NF-89-005-A, Qualification of Transient Analysis Hethods for BWR Design and Analysis," July 1992, including Supplements 1 and 2.

XN-NF-80-19 (P)(A), Volume 4, Revision 1, "Exxon Nuclear Methodology for Boiling Water Reactors:

Application of the ENC Methodology to BWR Reloads,"

Exxon Nuclear Company, June 1986.

NE-092-001, Revision 1, "Susquehanna Steam Electric Station Units 1 &.2: Licensing Topicaleport for ."

Power Uprate with Increased Core flow," December 1992, and NRC Approval Letter: Letter from T. E. Hurley (NRC) to R. G. Byram (PP&L), "Licensing Topical Report for Power Uprate With Increased Core Flow, Revision 0, Susquehann'a Steam Electric Station, Units 1 and 2 (PLA-3788) (TAC,Nos. M83426 and H83427),", November 30, 1993.

TM 7 4-. PL - ,

" se to NRC Request f Information on Sie .010, Revision 1" (Only Applicab e XN-NF-79-71(P)(A) Revision 2, Supplements 1, 2. and 3, "Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors," Harch 1986.

5, 4Q-. XN-NF-84-105(P)(A), Volume 1 and Volume 1 Supplements 1 and 2, "XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis," February 1987.

Final Policy Statement on Technical Specifications Improvements, July 22, 1993 (58 FR 39132).

" erence Safety Report CE Reactor Reloa " m ustion Engineering f'ater N ons. November 1994..

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LHGR B 3.2.3 BASES APPLICABLE Protection Against Power Transients, (PAPT) . defined. in SAFETY ANALYSES references 5 and 6, provides the acceptance criteria for (continued) LHGRs calculated in evaluation of the AOOs.,

For SPC 9x9-2 fuel, there is a LHGR,multiplier defined in the Core Operating Limits Report (COLR) for, single recirculation loop operation. This multiplier ensures that the DBA LOCA will be less severe in single loop operation two loop operation. 'han'n t 1 core contains four ABB lead use assemblies '

(LUAs). As are loaded in nonlimiting core per Specification 4. . . eparate LHGR limit he ABB LUAs have been developed using e ogy- from Reference 8.

Similar to'SPC 9x9-2 fue ensures recircula 'op will e DBA LOCA re operation.

LHGR multiplier for be less severe sn tiplier 'ingle le loop ion than in two loop operation.

The LHGR satisfies'riterion 2 .of the NRC 'Policy. Statement (Ref.

7).'CO The LHGR is a basic assumption in the fuel design analysis.

The fuel has been designed to operate at rated core power with sufficient design margin to the LHGR calculated to, cause a 1X fuel cladding plastic strain. The operating limit to accomplish this objective is specified in the COLR.

APPLICABILITY The LHGR.limits are derived from fuel design analysis" that is limiting at high power level conditions. At core thermal power levels ( 25K RTP, the reactor is operating with a substantial margin to the LHGR limits and..therefore, the Specification is only required when the reactor is operating at a 25K RTP.

ACTIONS A.1

'I If any. LHGR exceeds its required 'limit, an assumption regarding an initial condition of the fuel design analysis is not met. Therefore, prompt action should be taken to (continued)

SUSQUEHANNA -,UNIT 1 B 3.2-11 Revision 0.

LHGR .

B 3.2.3 BASES REFERENCES 4. XN-NF 85-67(P)(A), Revision 1, "Generic Mechanical (continued) Oesign for Exxon Nuclear Jet Pump BWR Reload Fuel,"

Exxon Nuclear Company, Inc., September 1986..'.

PL-NF-94-005-P-A. "Technical Basis for 9X9-2 Extended Fuel Exposure at Susquehanna SES," January 1995 ANF-89-98(P)(A) Revision 1 and Revision 1 Supplement 1, "Generic Mechanical Design Criteria for BWR Fuel Designs," Advanced Nuclear Fuels Corporation, Nay 1995.

7. Final Policy Statement on Technical Specifications Improvements. July 22. 1993 (58 FR 39132).

8. CE - - ,

" 'nce Safety Re or ustion Engineering Water Reactor Rel a sons, November. 1994.

SUSQUEHANNA - UNIT 1 B 3.2-13 Revision 0

~-

APR in and Setpoints B 3.2.4

'ASES APPLICABLE and LCO 3.2.3, "LINEAR HEAT GENERATION RATE .

SAFETY ANALYSES (LHGR)," 1'imit the initial margins to these operating limits, (continued)'MCPR),"

at rated conditions so that spec'defied acceptable fuel design limits are met during transients initiated from rated conditions. At initial power-levels less than r'ated levels, margin degradation of either the LHGR or the MCPR during

'he a transient can be greater than at the rated condition event. This greater margin degradation during the transient is primarily offset by the larger initial margin to limits at the lower than rated power levels. However, power distributions can be hypothesized that would result in reduced margins to the pre-transient operating limit. When combined'ith the increased severity of certain transients at other than rated conditions, the SLs could be approached.

At substantially reduced power levels, highly peaked power distributions could be obtained that could reduce thermal margins to the minimum levels required for transient events.

• To prevent or mitigate such situations. the MCPR margin .

degradation at reduced power and flow is factored into the power and flow dependent MCPR limits (LCO 3.2.2). .For LHGR (Ref. 4 and 5), either the,APRM gain is adjusted upward by the ratio of the core limiting MFLPD to the FRTP,. or the flow biased APRM scram level is reduced by the ratio of FRTP to the core limiting MFLPD. The adjustment in the APRM gain can be performed provided it is during power ascension up to 90K of RATED THERMAL POWER. that the adjusted APRM reading does not exceed 100K of RATED THERMA 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. Either of these adjustments effectively counters the increased severity of some events at other than rated conditions by proportionally increasing the APRM gain or proportionally lowering the flow biased APRM scram setpoints, dependent on the-.increased peaking that may be encountered.

The n (LUAs).

e The LU s contains four ABB ded in nonli 're lead use assem regions per Specification 4.2.1.

limit for me the A s rom Reference as been deve o 7.

'he mechanical desi gn The APRM gain and setpoints satisfy Criteria 2 and 3 of the NRC Policy Statement (Ref. 6).

(continued)

SUSQUEHANNA - UNIT 1 B 3.2-16 Revision 0

0

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APR in and Setpoints B 3.2.4 BASES SURVEILLANCE SR 3.2.4.1 and SR 3.2.4.2 (continued)

REQUIREMENTS is operating within the assumptions of the safety analysis..

These SRs are only required to determine the MFLPD and, assuming MFLPD is greater than FRTP, the appropriate gain or setpoint, and is not intended to be a CHANNEL FUNCTIONAL .,

TEST for the APRM gain or flow biased neutron flux scram circuitry. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency of SR 3.2.4.1 is chosen to coincide with the determination of other thermal..limits, specifically those for the APLHGR (LCO 3.2.1). The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency 'is based on both engineering judgment of'he slowness of .changes in power distribution and'ecognition during normal operation. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> allowance after THERMAL POWER > 25K RTP is achieved .is acceptable given the large inherent margin to operating. limits at low power and because the MFLPD must be .calculated prior 'to 'evels exceeding 50K RTP unless performed in the previous 24 hours...,-.'hen MFLPD is greater than FRTP, SR 3.2.4.2 must be performed. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency of'R 3.2.4.2 requires a more frequent verification when MFLPD is greater -than'he fraction of rated thermal power.(FRTP) because more rapid changes in power distribution are typically expected.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 10. GDC 13, GDC 20, and GDC 23.

2. FSAR, Section 4.

3. FSAR, Section 15 4; 'PL-NF-94-005-P-A, "Technical Basis" f'r 9X9-2'xtended Fuel Exposure at Susquehanna SES ~ January 1995.

ANF-89-98(P)(A) Revision 1 and Revision 1 Supplement 1. "Generic Mechanical Design Criteria f'r BWR Fuel Designs." Advanced Nuclear Fuels Corporation, May 1995.

6. Final Policy Statement on Technical Specifications Improvements, July 22, 1993 (58 FR 39132).

7. CENP - . ce Safety Re ng Water Reactor Rel stion Engineering a lons, November 1994.

SUSQUEHANNA -'UNIT 1 B 3.2-19 Revision 0

ENCLOSURE D TO PLA-5040

.,UNIT3,CYCLEQ2 C()RE.COMPOSITION....

Assembly Type Previous Cycle Number Of Operational History Assemblies SPC ATMUMM-10 Fresh 280 SPC ATRIUIvPM 10 Once-burned 308 SPC 9x9-2 Twice-burned 176

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I ENCLOSURE E TO PLA-5040

. CROSS REFKREKCK TO REACTIVITY,AND.POWER,..

DISTRIBUTION LIMITSFOR TECHNICAL SPECIFICATIONS 5.

6.5 REFERENCES

ENCLOSURE E TO PLA-5040 Page 1 of2 i

The applicability of the NRC approved references to the specific COLR Limits for the Susquehanna Steam Electric Station Unit 1 is summarized below.

Technical Specification Technical Specification Reference Number Unit 1 Cycle 12 Proposed Technical Specifications 3 4 5 7 8 9 "10 12 13 14 16 17 Average Planar Linear Heat X X X Generation Rate for Specification 3.2.1 Minimum Critical Power X X X X X X X Ratio for Specification 3.2.2 Linear Heat Generation Rate for Specification 3.2.3 Average Power Range X X X Monitor (APRM) Gain and Setpoints for Specification 3.2.4 Shutdown Margin for Specification 3.1.1 SUSQUEHANNA UNIT 1

Cross Reference/Applicability for PP&L Methodology to Core Operating Limits Report Operating Limit ENCLOSURE E TO PLA-5040 PP&L Proposed Amendment No. 227 to License NPF-14 Page 1 of 2 Unit 1 Reference Number Applicability COLR Limit/Technical Sp'ecification Section PP&L NRC approved topical report describing licensing analysis methods for -Minimum Critical Power Ratio (MCPR)/3.2.2 BWR design and analysis -.Shutdown Margin/3.1.1 SPC NRC approved topical report describing licensing analysis methods for -'Minimum Critical Power Ratio (MCPR)/3.2.2 BWR design and analysis 4 SPC NRC approved topical report describing mechanical design analysis for -Linear Heat Generation Rate (LHGR)/3.2.3 9X9-2  ;LHGR for APRM Setpoints/3.2.4 SPC NRC approved topical report describing licensing analysis methods for -Minimum Critical Power Ratio/3.2.2 BWR design and analysis SPC NRC approved topical report describing MCPR and MCPR Safety Limit  ;MCPR Safety Limit (two loop/single loop)/2.1.1 methodology including channel bow impact -'Minimum Critical Power Ratio (MCPR)/3.2.2 SPC NRC approved topical report describing the ANFB CPR correlation for -Minimum Critical Power Ratio (MCPR)/3.2.2 i

application to 9X9-2 SPC fuel GE NRC approved topical report describing the SAFER/GESTR-LOCA -Average Planar Linear Heat Generation Rate methodology. Analysis of record for SPC 9X9-2 (APLHGR)/3.2.1 PP&L topical report describing PP&L methodology/licensing basis for power -'Applies to all Core Operating Limits contained in the uprate and increased core fiow and NRC SER. COLR.

PP&L NRC approved topical report describing methodology changes for the -Minimum Critical Power Ratio (MCPR)/3.2.2 Loss of Feedwater Heating analysis and implementation of RETIE MOD 5.1 PP&L NRC approved topical report describing the licensing basis for extension 'inear Heat Generation Rate (LHGR)/3.2.3 10 of SPC 9X9-2 discharge exposure -LHGR for APRM Setpoints/3.2.4 PP&L NRC approved topical report describing methodology changes for -Minimum Critical Power Ratio (MCPR)/3.2.2 implementation of CASMO-3G and the ANFB CPR Correlation -Shutdown Margin/3.1.1 SPC NRC approved topical report describing generic mechanical design criteria -Linear Heat Generation Rate (LHGR)/3.2.3 12 which demonstrate acceptable results for SPC fuel designs (e. g., ATIUUM-10) -LHGR for APRM Setpoints/3.2.4 SPC NRC approved topical report describing LOCA methodology used for -Average Planar Linear Heat Generation Rate 13 SPC ATRIUM-10 fuel (APLHGR)/3.2.1 SPC NRC approved topical report describing LOCA methodology used for -Pverage Planar Linear Heat Generation Rate 14 SPC ATIUUM-10fuel (APLHGR)/3.2.1 SPC NRC approved topical report describing transient methodology used to Minimum Critical Power Ratio (MCPR)/3.2.

15 calculate ECPR for development of MCPR Operating Limits SPC NRC approved topical report describing transient methodology used to -Minimum Critical Power Ratio (MCPR)/3.2.2 16 calculate hCPR for development of MCPR Operating Limits SPC NRC approved topical report describing the ANFB-10 CPR Correlation -Mmunum Critical Power Ratio (MCPR)/3.2.2 17 t SUSQUEHANNA UNIT 1