ML18040B289
| ML18040B289 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 03/12/1999 |
| From: | PENNSYLVANIA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML17164A990 | List: |
| References | |
| NUDOCS 9903290104 | |
| Download: ML18040B289 (40) | |
Text
ENCLOSURE C TO PLA-5040
., TECHNICALSPECIFICATION MARK-UPs.
9903290i04 9903%2 PDR ADQCK 05000387 P
~ 4 lr
- 2. 0 SAFETY LIHITS =(SLs)
'.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 785 psig or core flow(.10 million ibm/hr:
2.1.1.2 2.1.1.3 THERHAL POWER shall be. ~, 25K RTP.
With the reactor steam dome pressure a 785 psig and core flow w 10 million 1bm/hr:
- III, bp HCPR shall be >
for two recirculati,on loop operation or >
i Ill i
1ti l
0 p
ti 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 1.35 1.30 30,1.27,'
J J
0 P
~
J 0 0
,'ACCE, ABLE
,'CP VALUES',
~
0
- ~
~
~
-1.25-
- E,--
1.20 M
f.
1.15 O
1.10 1.05
~t
~
~
I 0
. - 40J1.24 50,1.22...
J t
0 - -
~ 01.17 60J1.19..
0
~
0 0 0 0
l0 0
~
0 0 UNACCEITABL NICPR VALU
~
0 ~
J J
0 0 0 0 J
~ 0 0 0
~0
.... 90,1.13 80,1.15
~ ~
J 0 ~
0 ~
h 108,1.1 100,1.1 h
t 1,00 30 40 0
60 70 80 90
, 100 110 Core Ftow (M Ihr)
Figure 2.1.'t.2-1*
MCPR Safety Limitvs 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 1,35 E.
1.30 m
1.25 CO o
1.20 o
1.15 0
1.10 F
1.05 1.00 30,1.28- - - - ~-
40,1.25 0
0 P
52,1.2 I
0 I
o ~
~
ACCEPTABLE MCPR VALUES ',
~
~
I I
PTABLE'LUES
/
e I
FAUNA MC I
I
'108,1.22 el o
o
~
o
~ 0
~
0 0 I
I 30 40 50, 0
70 80 90 100 110 Core Flow (Illb/hr)
Figure 2.1.1.2-2*
MCPR Safety Limitvs 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.
7
~
8.
9.
10.
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.
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.
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.
XN-NF-84-97 Revision 0, "LO
-Seismic Structural
Response
an ENC 9x9 Jet Pum Fuel Assembly,"
Exxon Nuclear ompany, Inc..
December 84.
PLA
- 728, "Response to NRC Question; eismic/LOCA A lysis of U2C2 Reload," Letter from W. Keiser PP8L) to E. Adensam (NRC), September 25, 1986.
I 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-.
XN-NF-80-19(A). Volume 1, and Volume 1 Supplements 1~
4-, e~
"Exxon Nuclear Methodology for Boiling Water Reactors:
Neutronic Methods for Design and Analysis;-"
'Ey1 yyyl 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; /~ih5'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 Reactor Reload F
n ineering ions.
November 1994..
ater 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.
25.
26.
27.
29.
30.
XN-NF )(A),
"RODEX 2A (BWR) Fuel d Thermal-Mechanical Res nse Evaluation Mode August 1986.
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 Dioxide/Gad nia Irradiation Examina 'n and Thermal Conducti '," November 1986. 082(P)(A) Revision 1 and Revision 1
upplement 1, "Application of ANF Design Methodolog for
'uel Assembly Reconstitution,"
May 1995.
ANF-91-048(P)(A),
"Advanced Nuclear Fuels Corporation Methodology for Boiling Water Reactors EXEM BWR Evaluation Model," January 1993.
-33(P)(A) Supplement 2,
"HUXY: A Generali Multiro Code with 10CRF50 Appen eatup Option," January 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.
- ~
el (1'
v o'@
~
i 6) d c
0 r t
e Q
tr~r.
~ gal C.
d.
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 Suppleme s 2. 3, and 4.
"COTRANSA2:.
A C uter Program for Boiling ater Reactor Transient An ses,"
August 1990.-
36.
XN-NF-84-105(P)(A),
olume 1 an olume 1 Supplements 1
and 2, "XCOBRA-T: A.C uter ode for BWR Transient Thermal-Hydraulic Core A sis," February 1987.
37.
XN-NF-84-105(P)(A),
ume 1
plement 4,. "XCORBRA-T:
A Computer Code f BWR Transien Thermal-Hydraulic Core
- Analysis, Void action Model Comp ison to Experimental Data,." June 8.
38.¹ EHF 0(P), Revision l. "Application o FB to ATRI
-10 f'r. Susquehanna Reloads."
March 1
39.¹ A-4595, "Response to NRC Request for. Additional Information on Siemens Report EHF-97-010; Revision 1," March 27.
1997.
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.
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
i'Reactor 'Core SLs B 2.1;1 BASES
'ACKGROUND (continued)
Operation above the boundary of the nucleate boiling regime could result in excessive cladding temperature because of the onset of transition boiling and the resultant sharp'eduction in heat transfer coefficient; Inside the steam 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
'ctivity to the reactor coolant.
APPLICABLE SAFETY ANALYSES 'he fuel cladding must not sustain damage as a result of normal operation and AOOs.
" The reactor. core SLs are 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.
(,9 eleve,~c.e 4) +~4' AQUA 9 -~~ (,Le~ace~c.e i) g~c O]AGQ a~h
~ O.OS~K~ Ybl4-A 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,
. 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.
W~ 4QF5 oah P Slh
'pSVo, 2.1.1.1 Fuel Claddin Inte rit
~ hNW+-b~
S.
The use of the ANFB~correlationis 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 established by a limiting condition on core THERMAL POWER.
with the following basis:
(continued)
SUSQUEHANNA - UNIT 1 B 2.0-2 Revision 0
Reactor Core SLs B 2.1.1 BASES APPLICABLE SAFETY ANALYSES 2.'1.1.1 Fuel Claddin Inte rit (continued)
'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 P 3.0, which is signif'icantly higher than the expected
- peaking. 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
"'ignificant body of practical test data.
As long as the 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
'nv nt of boiling transition.
a~i 't'X~'SK. 4YOi~~~o "~
PC~ ue is monitored using the ANFB Critical Power Corr iog The effects of 'channel bow on HCPR are explicitly included in the calculation of the AHFB-. HCPR SL.
C ~~~~~+~~
Explicit treatment of channel bow in the~ HCPR SL address 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.
is'equired to be above the top of the active. fuel to provide 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 irradiated 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 SAFETY ANALYSES 2.1. 1.3 Reactor Vessel Water Level (continued) 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 VIOLATIONS Exceeding an SL may cause fuel damage and create a potential 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.
E.tq7 -)9/7 (P)(Ãj 9m~ s o~ Q, 4@~@-ia c.
r c. i P er g~cceX<g;~~ "3~i ~zgP
~~a, r ChV-xWqa(P')(@) ~ qq'le~e,~h i Q.e~igicc O,'gqq, ia l cia'xc.~X Q~~e C.~ch~h~'-
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)
I
- 5. Le, B ram (PP8L) to NRC.
"Sus earn Electric Station est for
~Additional In on Siemen MF-97,-010',
'-4596.
Parch 27.
1997.
SUSQUEHANNA - UNIT 1 B 2.0-6 Revision 0
0
APLHGR B 3.2.1 BASES APPLICABLE SAFETY ANALYSES (continued)
Suppression Pool Cooling Mode, and Single Loop Operation (SLO)).
LOCA analyses were performed for the regions of the 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'-10fuel 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 n
contains four ABB lead use ass (LUAs).
The LUAs, a in 'g core regions per S ecification para mits f'r the ABB LU eveloped using the reference 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 REQUIREMENTS SR 3.2.1.1
. (continued) 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 5.
7.'.
'9.
10 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.
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. "
FSAR. Chapter 4 ~
FSAR, Chapter 6.
FSAR, Chapter 15 Final Policy Statement on Technical Specif'ications Improvements, July 22.
1993 (58 FR 39132).
CEN rence Safety Re
'g Water Reactor Rel 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; 1),'he critical power at which'oiling transi-tion is 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 SAFETY ANAI YSE The analytical methods and assumptions used in evaluating 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
'LUAs).
The L
s ed in non-1 ore regions 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.
APPLICABILITY 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
'5K RTP is-unnecessary-due to the large inherent 'margin that ensures that the HCPR SL is not exceeded even if a limiting transient occurs.
Studies of the variation of 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 (continued) 3.
PL-NF-87-001-A, "Qualification of Steady State core 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-.
5, 4Q-.
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.
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).
CE -,
erence Safety Report f'ater Reactor Reloa m ustion Engineering N
ons.
November 1994..
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@<>-Qp " 3~Xy N9'lf.
0
~ '
LHGR B 3.2.3 BASES APPLICABLE SAFETY ANALYSES (continued)
Protection Against Power Transients, (PAPT). defined. in references 5 and 6, provides the acceptance criteria for 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
'han'n two loop operation.
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 re LHGR multiplier for
'ingle recircula 'op operation.
tiplier ensures e
DBA LOCA will be less severe sn 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 (continued) 4.
XN-NF 85-67(P)(A), Revision 1, "Generic Mechanical 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.
, " 'nce Safety Re or Water Reactor Rel ustion Engineering a sons, November. 1994.
SUSQUEHANNA - UNIT 1 B 3.2-13 Revision 0
~-
APR in and Setpoints B 3.2.4
'ASES APPLICABLE SAFETY ANALYSES (continued)'MCPR),"
and LCO 3.2.3, "LINEAR HEAT GENERATION RATE (LHGR)," 1'imit the initial margins to these operating limits, 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,
'he margin degradation of either the LHGR or the MCPR during 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
e contains four ABB lead use assem (LUAs).
The LU s ded in nonli 're regions per Specification 4.2.1.
A s mechanical desi gn limit for the as been deve o
'he me rom Reference 7.
The APRM gain and setpoints satisfy Criteria 2 and 3 of the NRC Policy Statement (Ref. 6).
SUSQUEHANNA - UNIT 1 B 3.2-16 (continued)
Revision 0
0
~
~
~ l
APR in and Setpoints B 3.2.4 BASES SURVEILLANCE REQUIREMENTS SR 3.2.4.1 and SR 3.2.4.2 (continued) 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 and'ecognition of'he slowness of.changes in power distribution 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
'evels and because the MFLPD must be.calculated prior 'to exceeding 50K RTP unless performed in the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />...,-.'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.
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.
6.
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.
Final Policy Statement on Technical Specifications Improvements, July 22, 1993 (58 FR 39132).
7.
CENP ce Safety Re Water Reactor Rel a lons, November 1994.
ng stion Engineering SUSQUEHANNA -'UNIT 1 B 3.2-19 Revision 0
ENCLOSURE D TO PLA-5040
.,UNIT3,CYCLEQ2 C()RE.COMPOSITION....
Assembly Type SPC ATMUMM-10 SPC ATRIUIvPM 10 SPC 9x9-2 Previous Cycle Operational History Fresh Once-burned Twice-burned Number Of Assemblies 280 308 176
-~
I ENCLOSURE E TO PLA-5040
. CROSS REFKREKCK TO REACTIVITY,AND.POWER,..
DISTRIBUTIONLIMITSFOR TECHNICAL SPECIFICATIONS 5.
6.5 REFERENCES
ENCLOSURE E TO PLA-5040 Page 1 of2 The applicability of the NRC approved references to the specific COLR Limits for the Susquehanna Steam Electric Station Unit 1 is i
summarized below.
Technical Specification Technical Specification Reference Number Unit 1 Cycle 12 Proposed Technical Specifications Average Planar Linear Heat Generation Rate for Specification 3.2.1 Minimum Critical Power Ratio for Specification 3.2.2 Linear Heat Generation Rate for Specification 3.2.3 Average Power Range Monitor (APRM) Gain and Setpoints for Specification 3.2.4 Shutdown Margin for Specification 3.1.1 3
4 5
X X
X 7
8 9
X X
X X
"10 X
X 12 13 14 X
X 16 17 X
X SUSQUEHANNAUNIT 1
Cross Reference/Applicability for PP&L Methodology to Core Operating Limits Report Operating Limit PP&L Proposed Amendment No. 227 to License NPF-14 ENCLOSURE E TO PLA-5040 Page 1 of2 Unit 1 Reference Number 10 12 13 14 15 16 Applicability PP&L NRC approved topical report describing licensing analysis methods for BWR design and analysis SPC NRC approved topical report describing licensing analysis methods for BWR design and analysis 4
SPC NRC approved topical report describing mechanical design analysis for 9X9-2 SPC NRC approved topical report describing licensing analysis methods for BWR design and analysis SPC NRC approved topical report describing MCPR and MCPR Safety Limit methodology including channel bow impact SPC NRC approved topical report describing the ANFB CPR correlation for application to 9X9-2 SPC fuel GE NRC approved topical report describing the SAFER/GESTR-LOCA methodology. Analysis ofrecord for SPC 9X9-2 PP&L topical report describing PP&L methodology/licensing basis forpower uprate and increased core fiowand NRC SER.
PP&L NRC approved topical report describing methodology changes for the Loss ofFeedwater Heating analysis and implementation ofRETIE MOD 5.1 PP&L NRC approved topical report describing the licensing basis for extension ofSPC 9X9-2 discharge exposure PP&LNRC approved topical report describing methodology changes for implementation ofCASMO-3G and the ANFB CPR Correlation SPC NRC approved topical report describing generic mechanical design criteria which demonstrate acceptable results for SPC fuel designs (e. g., ATIUUM-10)
SPC NRC approved topical report describing LOCA methodology used for SPC ATRIUM-10fuel SPC NRC approved topical report describing LOCA methodology used for SPC ATIUUM-10fuel SPC NRC approved topical report describing transient methodology used to calculate ECPR for development ofMCPR Operating Limits SPC NRC approved topical report describing transient methodology used to calculate hCPR for development ofMCPR Operating Limits COLR Limit/Technical Sp'ecification Section
-Minimum Critical Power Ratio (MCPR)/3.2.2
-.Shutdown Margin/3.1.1
-'Minimum Critical Power Ratio (MCPR)/3.2.2
-Linear Heat Generation Rate (LHGR)/3.2.3
- LHGRforAPRM Setpoints/3.2.4
-Minimum Critical Power Ratio/3.2.2
- MCPR Safety Limit(two loop/single loop)/2.1.1
-'Minimum Critical Power Ratio (MCPR)/3.2.2
-Minimum Critical Power Ratio (MCPR)/3.2.2 i
-Average Planar Linear Heat Generation Rate (APLHGR)/3.2.1
-'Applies to all Core Operating Limits contained in the COLR.
-Minimum Critical Power Ratio (MCPR)/3.2.2
'inear Heat Generation Rate (LHGR)/3.2.3
-LHGR for APRM Setpoints/3.2.4
-Minimum Critical Power Ratio (MCPR)/3.2.2
-Shutdown Margin/3.1.1
-Linear Heat Generation Rate (LHGR)/3.2.3
-LHGRforAPRM Setpoints/3.2.4
-Average Planar Linear Heat Generation Rate (APLHGR)/3.2.1
-Pverage Planar Linear Heat Generation Rate (APLHGR)/3.2.1 Minimum Critical Power Ratio (MCPR)/3.2.
-MinimumCritical Power Ratio (MCPR)/3.2.2 17 SPC NRC approved topical report describing the ANFB-10 CPR Correlation
-Mmunum Critical Power Ratio t
(MCPR)/3.2.2 SUSQUEHANNAUNIT 1