ML17164A959
| ML17164A959 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 02/17/1999 |
| From: | Bajwa S NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML17164A960 | List: |
| References | |
| NUDOCS 9902230098 | |
| Download: ML17164A959 (24) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 PP&L INC.
ALLEGHENYELECTRIC COOPERATIVE INC.
DOCKET NO. 50-388 SUS UEHANNASTEAM ELECTRIC STATION UNIT2 AMENDMENTTO FACILITYOPERATING LICENSE Amendment No. 1m License No. NPF-22 1.
The Nuclear Regulatory Commission (the Commission or the NRC) having found that:
A.
The application for the amendment filed by PPBL, Inc., dated August 4, 1998, as supplemented by letters dated December 16, 1998, January 12, 1999, and January 28, 1999, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I; B.
The facilitywilloperate in conformity with the application, the provisions of the Act, and the regulations of the Commission; C.
'There is reasonable assurance:
(i) that the activities authorized by this amendment can be'conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D.
The issuance of this amendment willnot be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
9902230098 9902i7 PDR ADQCK 05000388 P
PDR 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of the Facility Operating License No. NPF-22 is hereby amended to read as follows:
(2)
Technical S ecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through A'mendment No. 1@'nd the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license.
PPLL shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of its date of issuance and is to be implemented within 30 days after its date of issuance.
FOR THE NUCLEAR REGULATORYCOMMISSION S. Singh Bajwa, Director Project Directorate I-1 Division of Licensing Project Management Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
I'ebruary 17, 1999
ATTACHMENTTO LICENSE AMENDMENTNO.
154 FACILITYOPERATING LICENSE NO. NPF-22 DOCKET NO. 50-388 Replace the following pages of the Appendix A Technical Specifications with enclosed pages.
The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
'EMOVE 2.0- 1 2.0- 2 2.0- 3 5.0- 21 5.0 - 22 5.0- 23 5.0- 24 5.0- 25 8 2.0-2 8 2.0- 3 8 2.0-4 8 2.0- 5""
8 3.2-5 8 3.2-6 8 3.2-9 INSERT 2.0- 1 2.0-2 2.0- 3 5.0- 21 5.0 - 22 5.0- 23 5.0- 24 5.0- 25 8 2.0- 2 8 2.0- 3 8 2.0-4 82.0-5 8 3.2 - 5 8 3.2-6 8 3.2 - 9
SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs
- 2. l.l. 1 With the reactor steam dome pressure
< 785 psig or core flow < 10 million ibm/hr:
THERMAL POWER shall be s 25K RTP.
- 2. 1. 1.2 With the reactor steam dome pressure a 785 psig and core flow a 10 million ibm/hr:
MCPR shall be a 1.11 for two recirculation loop operation or a 1.12 for single recirculation loop operation.
- 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' TS / 2.0-1 Amendment e. 1~
SLs 2.0 (Figure 2.1.1.2-1)
THIS PAGE INTENTIONALLYLEFT BLANK SUSQUEHANNA - UNIT 2 TS / 2.0-2 Amendment te. 1W
SLs 2.0 (Figure 2.1.1.2-2)
THIS PAGE INTENTIONALLYLEFT BLANK SUSQUEHANNA - UNIT 2 TS / 2.0-3 Amendment z. 1~
4 k
epor ting Requirements 5.6 5.6 Re ortin Re uirements (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 be documented in the COLR for the 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.
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.
PL-NF-90-001-A, "Application of Reactor Analysis Methods for BWR Design and Analysis," July, 1992.
2.
XN-NF-80-19(P)(A), Volume 4. Revision 1, "Exxon Nuclear Methodology f'r Boiling Water Reactors:
Application of the ENC Methodology to BWR Reloads,"
Exxon Nuclear
- Company, Inc. June 1986.
(continued)
SUSQUEHANNA - UNIT 2 TS / 5.0-21 Amendment tb. 1."4
eporting Requirements 5.6
- 5. 6 Reporting Requirements 6.
7.
10 12.
13.
XN-NF-85-67(P)(A), Revision 1, "Gener ic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel, "Exxon Nuclear Company, Inc., September 1986.
r XN-NF-80-19(A), Volume 1, and Volume 1 Supplements 1,
2, and 3, "Exxon Nuclear Methodology for Boiling Water Reactors:
Neutronic Methods for Design and Analysis,"
Exxon Nuclear Company, Inc..
March 1983.
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.
"SAFER/GESTR-LOCA Loss of Coolant Accident Analysis,"
GE Nuclear
- Energy, May 1992.
NE-092-001A, Revision 1, "Licensing Topical Report for Power Uprate With Increased Core Flow," Pennsylvania Power
& Light Company, December 1992.
NRC SER on PP&L Power Uprate LTR (November 30, 1993)
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 MOD 5. 1."
August 1995.
PL-NF-94-005-P-A, "Technical Basis for SPC 9x9-2 Extended Fuel Exposure at Susquehanna SES". January.
1995.
NEDE-24011-P-A-10, "General Electric Standard Application For Reactor Fuel, February, 1991.
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)
SUSQUEHANNA - UNIT 2 TS / 5.0-22 Amendment e. isa
0
epor ting Requirements 5.6 5.6 Reporting Requirements 5.6.5 COLR (continued) 14.
ANF-89-98(P)(A) Revision 1 and Revision 1
I Supplement 1, "Generic Mechanical Design Criteria for BWR Fuel Designs,"
Advanced Nuclear Fuels Corporation, Hay 1995.
'5.
ANF-91-048(P)(A),
"Advanced Nuclear Fuels Corporation Methodology for Boiling Water Reactors EXEH BWR Evaluation Model," January '1993.
16.
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.
17.
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.
18.
XN-NF-79-71(P)(A) Revision 2, Supplements 1, 2, and 3.
"Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors,"
March 1986.
19.
EMF-1997 (P)(A) Revision 0.
"ANFB-10 Critical Power Correlation," July 1998.
and EMF-1997 (P)(A)
Supplement 1 Revision 0, "ANFB-10 Critical Power Correlation
- High Local Peaking Results." July 1998.
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 SDM, 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 f'r each reload cycle to the NRC.
(continued)
SUSQUEHANNA - UNIT 2 TS / 5.0-23 Amendment e. 1w
Reporting Requirements 5 '
5.6 Reporting Requirements 5.6.5 COLR (continued)
THIS PAGE INTENTIONALLYLEFT BLANK (continued)
SUSQUEHANNA - UNIT 2 TS / 5.0-24 Amendment hh. 1m
eporting Requirements 5.6 7
5.6'eporting Requirements 5.6.5 COLR (continued)
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SUSQUEHANNA - UNIT 2 TS / 5.0-25 Amendment e. 1~
Reactor Core SLs B 2.1.1 BASES BACKGROUND (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 reduction 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 activity to the reactor coolant.
APPLICABLE The fuel cladding must not sustain damage as a result of SAFETY ANALYSES normal operation and AOOs.
The reactor core SLs are established to preclude violation of the fuel. design criterion that an NCPR 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 boi ling.
The Reactor Protection System setpoints (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 THERHAL POWER level that would result in reaching the HCPR limit.
- 2. 1. 1. 1 Fuel Claddin Inte rit The use of the ANFB (Reference
- 2) and ANFB-10 (Reference 4) correlations are valid for critical power calculations at ressures
> 600 psia for ANFB'and > 571 psia for ANFB-10 and undle mass fluxes > 0.1 x 10'b/hr-ft'or ANFB and
> 0.115 x 10'b/hr-ft'or ANFB-10.
For operation at low pressures or low flows, the fuel cladding integrity SL is established by a limiting condition on core THERHAI POWER.
with the following basis:
Provided that the water level in the vessel downcomer is maintained above the top of'he 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 approach 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 2 TS /
B 2.0-2 Revision 1
Ane'anent bb. 1'
Reactor Core SLs B 2.1.1 BASES APPLICABLE SAFETY ANALYSES
- 2. 1. 1. 1 Fuel Claddin Inte rit (continued) the minimum bundle flow is > 28 x 10'b/hr.
For both the SPC 9x9-2 and Atrium-10 fuel designs, the coolant minimum 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 fuel designs a't 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 MWt.
At 25K RTP, a bundle power of'pproximately 3.35 MWt corresponds to a bundle radial peaking factor of'pproximately 3.0, which is signif'icantly higher than the expected peaking factor.
Thus.
a THERMAL POWER limit of 25K RTP for reactor pressures
< 785 psig is conservative.
2.1.1.2 MCPR The HCPR SL ensures sufficient conservatism in the operating HCPR limit that, in the event of an AOO from the limiting condition of operation, at least 99.9X of the fuel rods in the core would be expected to avoid boi ling transition.
The margin between calculated boiling transition (i.e.,
MCPR = 1.00) and the MCPR 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.
Reference 2 describes the methodology used in determining the HCPR SL.
The ANFB and ANFB-10 critical power correlations are based on a significant body of practical test data.
As long as the core pressure and flow are within the range'of validity of the correlation (refer to Section B 2. 1. 1.1.). the assumed reactor conditions used in defining 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 rods in boiling transition.
These conservatisms and the inherent accuracy of'he ANFB and ANFB-10 correlations 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 2 TS /
B 2.0-3 Revision 1
Arerxbent No. 1."4
Reactor Core SLs B 2.1.1 BASES APPLICABLE SAFETY ANALYSES 2.1.1. 2 HCPR (continued)
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 environment of boiling transition.
SPC 9x9-2 fuel is monitored using the ANFB critical power correlation, and the SPC ATRIUM-10 fuel is monitored using the ANFB-10 Critical Power Correlation.
The effects of channel bow on HCPR are explicitly included in the calculation of the MCPR SL.
Explicit treatment of channel bow in the HCPR SL addresses the concerns of 'the NRC Bulletin No. 90-02 entitled "Loss of Thermal Margin Caused by Channel Box Bow."
The Unit 2 core contains four GE lead use assemblies (LUAs).
The LUAs are loaded in nonlimiting core regions per Specification 4.2. 1.
The HCPR SL generated using Reference 2 is acceptable for the GE LUAs.
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 required 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'he 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 monitored and to also provide adequate margin for effective action.
(continued)
SUSQUEHANNA - UNIT 2 TS /
B 2.0-4 Revision 1
Amenchent No. 154
Reactor Core SLs 8 2.1.1 BCES (continued)
SAFETY LIMITS The reactor core 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 ensure 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 elevated 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).
Therefore, 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 of an accident occurring during this period is minimal.
REFERENCES 10 CFR 50, Appendix A, GDC 10.
ANFB 524 (P)(A). Revision 2, "Critical Power Methodology f'r Boiling Water Reactors,"
Supplement 1
Revision 2 and Supplement 2,
November 1990.
3.
4.
EMF-1997, Revision 0 (October 1997) and Supplement 1,
Revision 0 (January 1998), -"ANFB-10 Critical Power Correlation,"
and associated NRC SER dated 7/17/98.
SUSQUEHANNA - UNIT 2 TS / 8 2.0-5 Revision 1
kendnent hb.
3Rd
MCPR 8 3.2.2 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.2 HINIHUM CRITICAL POWER RATIO (MCPR)
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 99.9K 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 HCPR 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),
the critical power at which boiling transition is calculated to occur has been adopted as a fuel design criterion.
The onset of transition boiling is a phenomenon that is readily detected during the testing of various fuel bundle designs.
Based 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 ensuring that fuel failures due to inadequate cooling do not occur.
APPLICABLE SAFETY ANALYSES The analytical methods and assumptions used in evaluating the AOOs to establish the operating limit HCPR are presented in References 2 through 9.
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 HCPR SL. the required operating limit HCPR is obtained.
The MCPR operating limits derived from the transient analysis are dependent on the operating core flow and power SUSQUEHANNA - UNIT 2 TS /
B 3.2-5 (continued)
Revision 1
kerxbent No. 1~
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)
'hese analyses may also consider other combinations of plant conditions (i.e., control rod scram speed, bypass valve
~
~
erformance.
EOC-RPT. cycle exposure, etc.).
Flow dependent CPR limits are determined by analysis of slow flow runout transients using the methodology of Reference 2.
The Unit 2 core contains four GE lead use assemblies (LUAs).
The LUAs are loaded in nonlimiting core regions per specification 4.2. 1.
HCRR operating limits for the GE LUAs have been developed using the methodology from References 2
and 11.
The HCPR satisfies Criterion 2 of the NRC Policy Statement (Ref. 10).
LCO The HCPR operating limits specified in the COLR are the result of the Design Basis Accident (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 MCPR operating limits are prima ily 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 is small.
Surveillance of thermal limits below 25K 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 (continued)
SUSQUEHANNA - UNIT 2 TS /
B 3.2-6 Revision 1
Anendmet lb. 154
HCPR B 3.2.2 BASES REFERENCES (continued)
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 Methods f'r 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
8 2:
Licensing Topical Report for Power Uprate with Increased Core Flow," December
- 1992, and NRC Approval Letter:
Letter from T.
E. Hurley (NRC) to R.
G.
Byram (PP8L),
"Licensing Topical Report for Power Uprate With Increased Core Flow, Revision 0, Susquehanna Steam Electric Station, Units 1 and 2
(PLA-3788)
(TAC Nos.
M83426 and M83427),"
November 30, 1993.
EHF-1997, Revision 0 (October 1997) and Supplement 1.
Revision 0 (January 1998),
"ANFB-10 Critical Power Correlation,"
and associated NRC SER dated 7/17/98.
XN-NF-79-71(P)(A) Revision 2, Supplements 1, 2, and 3.
"Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors,"
March 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.
10.
Final Policy Statement on Technical Specifications Improvements.
July 22, 1993 (58 FR 39132).
11.
NEDE-24011-P-A-10, "General Electric Standard Application for Reactor Fuel," February 1991.
SUSQUEHANNA - UNIT 2 TS /
B 3.2-9 Revision 1
4andmt bb. 1~