ML20056G486
| ML20056G486 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 08/27/1993 |
| From: | Hunger G PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML19310D669 | List: |
| References | |
| NUDOCS 9309030178 | |
| Download: ML20056G486 (20) | |
Text
PHIIADELPHIA ELECTRIC COMPANY NUCLEAR GROUP HEADQUARTERS 955-65 CHESTERBROOK BLVD.
WAYNE. PA 19087-5691 10CFR50.90 cis>caoa m 10CFR50.46(a)(3)(ii)
STATION SUPPORT DEPARTMENT August 27, 1993 Docket Nos. 50-352 50-353 l License Nos. NPF-39 l NPF-85 U. S. Nuclear Regulatory Commission Attn: Document Control Desk ,
Washington, DC 20555 j l l l
SUBJECT:
Limerick Generating Station, Units 1 and 2 i Technical Specifications Change Request and l Notification of Change in the Calculated Peak l Fuel Cladding Temperature in Accordance with 10CFR50.46(a)(3)(ii)
Gentlemen:
Philadelphia Electric Company (PECo) is submitting Technical Specifications (TS) Change Request No. 92-08-0, in accordance with 10 CFR 50.90, requesting an amendment to the TS (i.e., Appendix A) of Operating License Nos. NPF-39 and NPF-85 for Limerick Generating Station (LGS), Units 1 and 2, respectively. The proposed TS changes reflect an expanded operating domain for LGS, Units 1 and 2, resulting from the proposed implementation of the Average Power Range Monitor -
l Rod Block Monitor Technical Specifications / Maximum Extended Load Line Limit Analysis (ARTS /MELLLA). The NRC has approved expanded operating domains at other Boiling Water Reactor (BWR) type plants, such as Brunswick Steam Electric Plant, Units 1 and 2, by Safety Evaluation (SE) dated October 23, 1990 and October 12, 1989, respectively, and the Fermi Plant, Unit 2, by SE dated May 15, 1991.
l Attachment I to this letter describes the proposed changes, and contains t information supporting a finding that the proposed changes do not involve a i Significant Hazards Consideration and information supporting an Environmental l Assessment. Attachment 2 contains the TS pages showing the proposed changes to l the LGS, Units 1 and 2 TS. Attachment 3, " Maximum Extended Load Line Limit and l ARTS Improvement Program Analysis for Limerick Generating Station Units 1 and 2,"
NEDC-32193P, Revision 1, dated July 1993, contains the safety analysis prepared by General Electric (GE) to support this change request and the implementation (
of ARTS /FELLLA at LGS, Units 1 and 2.
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U. S. Nuclear Regulatory Commission August 27, 1993 Document Control Desk Page 2 ;
! This letter also satisfies the requirements of 10 CFR 50.46(a)(3)(ii), )
reporting a change to the evaluation model utilized to analyze the Loss-of- i Coolant Accident (LOCA) for LGS, Units 1 and 2. PEco has adopted the GE l SAFER /GESTR model for evaluation of the LOCA. This model has been approved by j the NRC as discussed in a letter from C. O. 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 the loss-of-Coolant Accident," dated June 1,1984, and is currently being utilized at the majority of operating Boiling Water Reactors (BWRs).
i Attachmant 4, " Limerick Generating Station Units 1 and 2 SAFER /GESTR-LOCA !
! Loss-of-Coolant Accident Analysis," NEDC-32170P, Revision 1, dated June 1993, i contains the plant specific safety analysis prepared by GE in support of this !
change to the LOCA evaluation model. As a result of the use of the SAFER /GESTR l model, the limiting Licensing Basis fuel rod peak cladding temperature for the postulated Licensing Basis LOCA at the licensed power limit of 3293 MWt at LGS,
, Units 1 and 2, changes from 2,195 *F to 1,310 *F. Since this change in the peak cladding temperature is greater than 50 *F, Attachment 4 is being submitted in accordance with 10CFR.50.46(a)(3)(ii). In anticipation of implementing the ARTS /MELLLA at LGS, Units 1 and 2, the SAFER /GESTR model has been used to calculate the fuel rod peak cladding temperature during a LOCA with the l ARTS /MELLLA improvements. The SAFER /GESTR model calculates that the peak l cladding temperature will increase less than 20 *F with the ARTS /MELLLA improvements based on 10 CFR 50, Appendix K assumptions.
Attachments 3 and 4 contain information proprietary to GE. GE requests that the documents be withheld from public disclosure in accordance with 10 CFR 2.790(a)(4). The affidavits supporting this request in accordance with 10 CFR 2.790(b)(1), are provided with Attachments 3 and 4, respectively.
The improvements associated with the Maximum Extended Load Line Limit (MELLL) mode of operation and the ARTS orogram are a prerequisite for the Power Rerate Program implementation at LGS, ' , ~
- I and 2.
We request that, if approved, the amendments be effective by January 19, 1994, for LGS Unit 1, and February 18, 1995, for LGS Unit 2. These dates support the planned modification schedules. Specifically, the modifications associated with implementation of the ARTS /MELLA clinges are planned to be made during the fifth refueling outage for Unit 1, currently scheduled to begin on January 29, 1994, and the third refueling outage for Unit 2, currently scheduled to begin on i March 4, 1995.
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U. S. Nuclear Regulatory Commission Arpst 27,1993 Documer t Control Desk Page 3 i
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If you have any questions, please do not hesitate to contact us. .
I Very truly yours, G.A..L:1.Hunger,Jp..fa .
! Director l Licensing Section Attachments l cc: T. T. Martin, Administrator, Region I, USNRC w/ attachments N. S. Perry, USNRC Senior Resident Inspector, LGS "
W. P. Dornsife, Director, PA Bureau of Radiological Protection w/ attachments
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i COMMONWEALTH OF PENNSYLVANIA:
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COUNTY OF CHESTER :
i G. R. Rainey, being first duly sworn, deposes and says:
That he is Vice President of Philadelphia Electric Company; the Applicant I herein; that he has read the foregoing Application for Amendment of Facility Operating License Nos. NPF-39 and NPF-85 (Technical Specifications Change Request l No. 92-08-0) to reflect the implementation of the Average Power Range Monitor -
l Rod Block Monitor Technical Specifications / Maximum Extended Load Line Limit Analysis (ARTS /MELLLA) at Limerick Generating Station Units 1 and 2, and knows the contents thereof; and that the statements and matters set forth therein are true and correct to the best of his knowledge, information and belief.
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_/i /Oi Vice President l ,
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Subscribed and sworn to befo e me th so day (of/ u d N V- ] 1993.
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l v jj-Notary Public l thanzsw i Eva A Samon.f6ary P.m !
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l ATTACHMENT 1 LIMERICK GENERATING STATION ,
UNITS 1 AND 2 ,
l Docket Nos. 50-352 l 50-353 i
l License Nos. NPF-39 NPF-85 i l
TECHNICA*. SPECIFICATIONS CHANGE REQUEST No. 92-08-0 ]
" Average Power Range Monitor - Rod Block Monitor Technical Specifications / Maximum Extended Load Line Limit Analysis (ARTS /MELLLA)" ,
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Supporting Information for Changes - 15 pages j
i U. S. Nuclear Regulatory Commission August 27, 1993 Document Control Desk Page 1 !
Philadelphia Electric Company (PECo), Licensee under Facility Operating License Nos. NPF-39 and NPF-85 for Limerick Generating Station (LGS), Units 1 and 2, respectively, requests that the Technical Specifications (TS) contained in Appendix A to the Operating License be amended as proposed herein, to reflect an ,
expanded operating domain for LGS, Units 1 and 2. This results from the !
implementation of the Average Power Range Monitor - Rod Block Monitor Technical Specifications / Maximum Extended Load Line Limit Analysis (ARTS /MELLLA). i Attachment 3, " Maximum Extended Load Line Limit and ARTS Improvement Program l Analysis for Limerick Generating Station, Units 1 and 2, "NEDC-32193P, Revision 1, dated July 1993, contains the safety analysis prepared by General Electric (GE) to support this Change Request and the implementation of ARTS /MELLLA at LGS, Units I and 2. These changes include references to the SAFER /GESTR methodology ,
used for the analysis of the spectrum of loss of Coolant Accidents (LOCAs) for LGS, Units 1 and 2. The safety analyses prepared by GE to support this change ,
to the LOCA evaluation model is provided in Attachment 4, " Limerick Generating Station Units 1 and 2 SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis," NEDC-32170P, Revision 1, dated June 1993. The proposed changes are indicated by the '
i vertical bars in the margin of the TS pages listed in Attachment 2. All affected TS pages are included in Attachment 2.
We request that, if approved, the amendments be effective by January 19, i 1994, for LGS Unit 1, and February 18, 1995, for LGS Unit 2. These dates support the planned modification schedules. Specifically, the modifications associated with implementation of the ARTS /MELLA changes are planned to be made
- during the fifth refueling outage for Unit 1, currently scheduled to begin on i January 29, 1994, and the third refueling outage for Unit 2, currently scheduled I to begin on March 4, 1995.
This TS Change Request provides a discussion and description of the proposed TS changes, a safety assessment of the proposed TS changes, information supporting a finding of No Significant Hazards Consideration, and information i supporting an Environmental Assessment.
Discussion and Descriotion of the Proposed Chanaes The Technical Specifications (TS) changes associated with the implementation of Average Power Range Monitor - Rod Block Monitor Technical Specifications / Maximum Extended Load Line Limit Analysis (ARTS /MELLLA) at Limerick Generating Station (LGS), Units 1 and 2 are as follows. .
A) Add the following terms to TS Section 1.0, " Definitions."
i) Subsection 1.9a - Downscale Trip Setpoint (DTSP), TS Page 1-2.
ii) Subsection 1.15a - High Trip Setpoint (HTSP), TS Page 1-3.
iii) Subsection 1.16a - Intermediate Trip Setpoint (ITSP), TS Page 1-3.
iv) Subsection 1.20a - Low Trip Setpoint (LTSP), TS Page 1-4.
v) Subsection 1.22a - Flow Dependent Maximum Average Planar Linear Heat Generation (MAPFAC(F)), TS Page 1-4.
U. S. Nuclear Regulatory Commission August 27, 1993 '
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vi) Subsection 1.22b - Power Dependent Maximum Average Planar '
Linear Heat Generation (MAPFAC(P)), TS Page 1-4.
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B) Revise the following term in TS Section 1.0, " Definitions." l i) Subsection 1.23 - Flow and Power Dependent Minimum Critical i Power Ratio (MCPR(F)) and (MCPR(P)), respectively, TS Page 1-4.
C) Delete the following terms from TS Section 1.0, " Definitions."
i) Subsection 1.13 - Fraction of Limiting Power Density, TS Page 1-3.
I ii) Subsection 1.14 - Fraction of Rated Thermal Power, TS Page 1-3.
l iii) Subsection 1.21 - Maximum Fraction of Limiting Power Density, I
TS Page 1-4.
D) Replace the current flow-referenced Rod Block Monitor (RBM) setpoint TS with the RBM power-referenced setpoints as described below.
- 1) Revise Limiting Condition for Operation (LCO) " Rod Block ,
! Monitor," TS Section 3.1.4.3, Page 3/4 1-18.
ii) Revise " Control Rod Withdrawal Block Instrumentation," TS Table 3.3.6-1, Page 3/4 3-59.
l iii) Revise " Control Rod Block Instrumentation Setpoints," TS Table 3.3.6-2, Pages 3/4 3-60 and 3/4 3-60a.
iv) Revise " Control Rod Block Instrumentation Surveillance Requirements," TS Table 4.3.6-1, Pages 3/4 3-61 and 3/4 3-62.
v) Revise " Control Rod Program Controls," TS Bases 3/4.1.4, Page B 3/4 1-3.
E) Change the following TS to reflect Implementation of power-dependent
! and flow-dependent fuel thermal limits in order to eliminate Average Power Range Monitor (APRM) trip setdown requirements and to support ,
the power-dependent RBM trips. (Note that power and flow-dependent l fuel thermal limits are developed for GE fuel designs and are also applicable to the Asea Brown Boveri (ABB) and Siemens Nuclear Power (SNP) qualification fuel bundles.)
i) Revise "APRM Setpoints," TS Table 2.2.1-1, Page 2-4. I ii) Delete the flow referenced trip setpoint discussion from
" Reactor Protection System Instrumentation Setpoints for the Average Power Range Monitor," TS Bases, Page B 2-7. l l
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iii) Revise LC0 " Average Planar Linear Heat Generation Rate," TS l Section 3/4.2.1, Page 3/4 2-1. !
iv) Delete LC0 and Surveillance Requirements "APRM Setpoints" TS !
Section 3/4.2.2, Page 3/4 2-7.. !
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v) Revise LCO and Surveillance Requirements for " Minimum Critical l Power Ratio," TS Section 3/4.2.3, Pages 3/4 2-8 and 3/4 2-9. l vi) Revise the Bases discussion of warage Planar Linear Heat Generation Rate (APLHGR), TS Bases 3/4 2.1, Page B 3/4 2-1. ;
vii) Delete the Bases discussion of APRM Setpoints, TS Bases 3/4 2.2, Page B 3/4 2-2.
viii) Revise the Bases discussion of Minimum' Critical Power Ratio .
(MCPR) TS Bases 3/4 2.3, Pages B 3/4 2-4 and B.3/4 2-5.
ix) Revise Reactor Protection System Instrumentation Surveillance Requirements, TS Table 4.3.1.1-1, Page 3/4 3-8.
x) Revise Administrative Controls " Core Operating Limits Report,"
TS Section 6.9.1.9, Page 6-18a. :
F) Reactor Recirculation System !
i) Delete LC0 ACTION a.l.c, TS Section 3.4.1.1, Page 3/4 4-la.
l ii) Revise LC0 ACTION a.2, TS Section 3.4.1.1, Page 3/4 4-la.
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iii) Revise TS Bases 3/4 4.1, Page B 3/4 4-1. I i
l G) Standby Liquid Control System ;
i) Revise and add to Surveillance Requirements, TS Section 4.1.5, Pages 3/4 1-19 and 3/4 1-20.
, ii) Revise TS Bases 3/4 1.5, Pages B 3/4 1-4 and B 3/4 1-5 l
H) Depressurization System j i) Revise TS Bases 3/4 6.2, Page B 3/4 6-3 I) References i) Revise References cited in TS Bases-3/4.2.4, Page B 3/4 2-5.
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U. S. Nuclear Regulatory Commission August 27, 1993 Document Control Desk Page 4 This proposed TS Change Request reflects the operation of LGS, Units I and 2 in the region above the rated rod line. The current operating envelope is modified to include the extended operating region bounded by the rod line which passes through the 100% power /75% core flow point (i.e., approximately the 121%
rod line), the rated power line, and the rated load line, as shown in Figure 2.1 of Attachment 3, " Maximum Extended Load Line Limit and ARTS Improvement Program Analysis for Limerick Generating Station Units 1 and 2,"NEDC-32193P, Revision 1, dated July 1993. The technical analysis is referred to as the Maximum Extended Load Line Limit Analysis (MELLLA) and the shaded area in Figure 2.1 is referred to as the Maximum Extended Load .Line Limit (MELLL) region. The evaluation utilizes LGS, Unit 1 Cycle 5 as the reference cycle for the safety analyses. The applicability of MELLL operation to subsequent LGS cycles will be supported by analyses for plant / cycle specific operating limits associated with operation in the MELLL domain, to be performed as - part of the cycle specific reload calculations. In addition, the implementation of the Average Power Range Monitor, Rod Block Monitor, and Technical Specifications Improvement Program will-increase plant operating efficiency by updating the thermal limits requirements and improving plant instrumentation responses and accuracy.
The improvements associated with the MELLL mode of operation and the Improvement Program reflected in the proposed TS Change Request are described bel ow.
- (1) A power-dependent MCPR thermal limit similar to that used by Boiling j
! Water Reactor (BWR) type 6 plants is to be implemented to complement =
l the new power-biased RBM system. This will update reactor thermal j limit administration. 1 i
(2) The APRM trip setdown requirement is to be replaced by more !
meaningful power-and flow-dependent thermal limits to reduce the i need for manual setpoint adjustments and to allow more direct ,
thermal limits administration. This will improve the human / machine interface, update thermal limits administration, increase ;
reliability, and provide more accurate protection of the plant safety. l (3) The flow-biased RBM trips will be replaced with power-dependent i trips. The Low Power Range Monitor (LPRM) inputs to the RBM system will be reassigned to:
I l (a) improve the response characteristics of the system;-
(b) improve the response predictability; and I (c) reduce the frequency of nonessential alarms.
This will improve plant instrumentation responses, accuracy, and reliability, as well as the human / machine interface.
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U. S. Nuclear Regulatory Commission August 27, 1993 Document Control Desk Page 5 i
(4) The Rod Withdrawal Error analysis was performed in a manner that more accurately reflects actual plant operating conditions and is consistent with the system changes. This updates the thermal limits administration.
(5) The Standby Liquid Control System was evaluated and the Baron - 10 enrichment percentage will be revised to accommodate operation in the MELLL region. ;
Safety Assessment The proposed Technical Specifications (TS) changes will permit Limerick Generating Station (LGS), Units I and 2, to operate in an expanded operatir.9 domain. Operation in the expanded operating domain is based on Maximum Extendeo load Line Limit Analysis (MELLLA) performed by General Electric (GE) using methods described in Attachment 3, " Maximum Extended Load Line Limit and ARTS' Improvement Program Analyses for Limerick Generating Station Units 1 and 2," ;
NEDC-32193P, Revision 1, dated July 1993. The current operating envelope is i modified to include the extended operating region bounded by the rod line which ;
passes through the 100% power /75% core flow point (i.e., approximately the 121%
rod line), the rated power line, and the rated load line, as shown in Figure 2.1 of Attachment 3, " Maximum Extended Load Line Limit and ARTS Improvement Program i Analysis for Limerick Generating Station Units 1 and 2,"NEDC-32193P, Revision 1, dated July 1993. Plant operating efficiency is increased by the Average Power Range Monitor-Rod Block Monitor Technical Specifications (ARTS) program which updates thermal limits requirements and improves plant instrumentation responses
, and accuracy. This requires modification to the Average Power Range Monitor l (APRM) and Rod Block Monitor (RBM) systems. >
This safety assessment summarizes the information provided in Attachment 3, " Maximum Extended Load Line Limit and ARTS Improvement Program Analysis for Limerick Generating Station Units 1 and 2,"NEDC-32193P, Revision 1, dated July 1993 prepared by GE. NEDC-32193P provides the analysis to support the operation of LGS, Units I and 2, in the expanded operating domain, justifying the proposed TS changes as follows.
- 1. Delete the flow-biased Average Power Range Monitor (APRM) scram and rod block trip setpoint setdown requirements, delete reference to the "k" flow adjustment factor, k, y introduce power and flow dependent adjustments to the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) and Minimum Critical Power Ratio (MCPR) limits, revise the documentation requirements of the Core Operating Limits Report (COLR), and delete the definitions of the Fraction of Rated Thermal Power (FRTP), Fraction of Limiting Power Density (FLPD), and the Maximum Fraction of Limiting Power Density (MFLPD).
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U. S. Nuclear Regulatory Commission August 27, 1993 j Document Control Desk Page 6 l l
The above proposed TS changes eliminate the requirement for setdown !
of the flow-biased APRM scram and rod block trip setpoints when the ;
MFLPD is greater than the FRTP and substitute adjustments to the !
MCPR and MAPLHGR operating limits that are flow and power dependent. .
An alternative method of assuring that the safety limit MCPR and :
fuel thermal-mechanical design bases are not violated is to define :
the operating limit MCPR and MAPLHGR limits such that no postulated ;
transient event, if initiated from other than rated power or flow -
conditions, could result in violation of either the safety limit MCPR or the fuel thermal-mechanical design bases. Evaluations to l determine the flow and power dependent requirements on the MCPR and !
MAPLHGR limits were performed and are described in detail in- !
Attachment 3, " Maximum Extended Load Line Limit and ARTS ;
Improvement Program Analysis for Limerick Generating station Units l 1 and 2," NEDC-32193P, Revision 1, dated July 1993. The results of i these evaluations were used to determine the following set of flow l and power dependent parameters for the fuel designs present in LGS, ;
Units I and 2. (
I a) A new power dependent MCPR limit adjustment factor, MCPR(P), i b) A new power dependent MAPLHGR limit adjustment factor, l MAPFAC(P),
c) A new flow dependent MCPR limit, MCPR(F), which replaces the current k,, MCPR multiplier, and d) A new flow dependent MAPLHGR limit adjustment factor, )
MAPFAC(F).
Since the MCPR(F) limit replaces the k, factor, all references to k, will be deleted. Consistent with the approach of the elimination of cycle specific parameters, such as k,, from the TS the parameters which are used to determine thermal operating limits are reported in the COLR for each cycle. The governing MCPR and MAPLHGR limits for any power and flow condition are the more limiting of the values as adjusted by the core power and flow dependent factors.
The power dependent requirements, MCPR(P) and MAPFAC(P), were derived from analyses of the limiting anticipated operational occurrences which are the main generator load rejection with no bypass, turbine trip with no bypass, and feedwater controller l failure maximum demand event. The MCPR(P) bounds the initial MCPR needed to assure that the fuel safety limit will not be violated for each transient event. The power-dependent MCPR requirements are also consistent with the definition of the Rod Block Monitor (RBM) setpoints and the Rod Withdrawal Error (RWE) analyses. The MAPFAC(P) requirement was determined from 'the same transient evaluations used to determine the MCPR(P) and -demonstrates compliance with the fuel thermal-mechanical design basis.-
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U. S. Nuclear Regulatory Commission August'27, 1993 Document Control Desk Page-7 i
The flow dependent requirements, MCPR(F) and MAPFAC(F), were derived from results of analyses of the slow-flow (i.e., recirculation flow) ;
runout events. The MCPR(F) requirement serves the same purpose as j the current k, multiplier which it replaces. The MAPFAC(F) !
requirements are specified in terms of a MAPLHGR multiplier, which i is applied to the full power and flow fuel type and exposure i dependent-MAPLHGR limits. This multiplier has been derived such i that the peak transient MAPLHGR during the slow-flow runout ;
transients will not increase above the fuel . thermal-mechanical design bases values.
The ficte dependent MAPLHGR requirements also bound the MAPLHGR [
requirements which are needed to demonstrate compliance to 10 CFR ;
50.46 and 10 CFR SD, Appendix K. [
The proposed elimination of the APRM setpoint setdown and '
substitution of core flow and power dependent adjustment factors ,
provides more direct administration of fuel thermal limits compared l with the current practice of modifying the APRM gain adjustment t factors. j The definitions of MFLPD and FRP are used only in the determination j of the required setdown of the flow-biased APRM scram and rod block j setpoints. Since we propose to delete the setdown requirements, :
these definitions are also being deleted in the proposed TS changes, j i
- 2. Modify the flow-biased APRM scram and rod block trip equations to i accommodate an expanded operating domain.
The purpose of the flow-biased APRM rod block trip setpoint is to block control rod withdrawal when core power exceeds design bases and approaches ' the scram level. Should operation continue in a manner such that the core power / flow condition ; exceeds that specified by the APRM rod block setpoint, the flow-biased APRM scram trip setpoint would initiate a scram. As such, the flow-biased APRM rod block provides a buffer in power and flow conditions from the flow-biased APRM scram function. Both the current and proposed formulation of the flow-biased APRM scr:m equations are clamped such that a maximum value of the trip setpoint is less than or equal to the trip setpoint of the fixed neutron flux scram. The proposed formulation does, however, reach this maximum value at a lower flow condition. The APRM scram trip setpoint will continue to initiate a scram if the increasing core power / flow condition continues beyond the APRM rod block setpoint.
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U. S. Nuclear Regulatory Commission August 27, 1993 Document Control Desk Page 8
- 3. Replace the RBM flow-biased trip equation with power-dependent setpoint definitions.
The RBM system is explicitly designed to mitigate the consequences of the RWE event and is not assumed to be available to mitigate any other anticipated operational occurrence. The proposed modification to the RBM system configuration is described in Attachment 3," Maximum Extended Load Line Limit and ARTS Improvement Pr., gram Analysis for Limerick Generating Station Units 1 and 2", NEDC-32193P, Revision 1, dated July 1993. The modified RBM system uses advances in electronics to enhance instrumentation accuracy and to improve the signal to thermal margin correlation. Modification of the RBM trip logic provides a system response which more accurately reflects the actual margin to the safety limit at various power conditions. The more accurate response of the RBM system will reduce the number of rod blocks which are not associated with reduced thermal limit margins. The operator will be better able to predict system response which improves the human / machine interface and enhances safety. Coincident with the analyses of the modified RBM system, a generic RWE approach was taken such that neither the safety limit MCPR or the fuel thermal-mechanical design basis is jeopardized.
This approach included determining appropriate MCPR requirements and corresponding RBM power dependent setpoints for the modified RBM system for current fuel designs. By an appropriate selection of the setpoints, the RWE will not be the limiting event and will not ,
determine the operating limit MCPR. In this respect, the RBM setpoints are dependent upon the operating limit MCPR values which depend on the cycle-specific conditions. For these reasons, the proposed TS changes incorporate a reference to the COLR.
- 4. Revise the enrichment of Boron-10 used in the Standby Liquid Control System (SLCS) sodium pentaborate solution to accommodate operation in the Maximum Extended Load Line Limit (MELLL) region.
The SLCS operation or safety function (i.e., mitigation of postulated Anticipated Transients Without Scram (ATWS) events), is not impacted as a result of changing the Boron-10 enrichment to 29%.
This is required in order to maintain the suppression pool tcmperature below 190* F due to operation in the MELLL region in case of an ATWS event, and ensures that the SLCS has the capability to mitigate such an event, and shutdown the reactor as originally designed.
U. S. Nuclear Regulatory Commission August 27, 1993 Document Control Desk Page 9 Additional discussions of various systems, structures, and components that -
have been evaluated for operation of LGS, Units 1 and 2 in the extended operating domain, are provided in Attachment 3, " Maximum Extended Load Line Limit and ARTS Improvement Program Analysis for Limerick Generating Station Units 1 and 2,"NEDC-32193P, Revision 1, dated July 1993.
Information Succortina A Findina of No Sionificant Hazards Consideration We have concluded that the proposed changes to the Limerick Generating Station (LGS), Units 1 and 2 Technical Specifications (TS), which reflect the --
implementation of the Average Power Range Monitor - Rod Block Monitor Technical Specifications / Maximum Extended Load Line Limit Analysis (ARTS /MELLLA), do not involve a Significant Hazards Consideration. In support of this determination, -- --
an evaluation of each of the three (3) standards set forth in 10 CFR 50.92 is provided below.
- 1. The oroposed Technical Specifications (TS) chanaes do not involve a sianificant increase in the probability or ConseouenceS of an accident previously evaluated.
The consequences of any accident previously evaluated will not be increased due to operation of Limerick Generating Station (LGS),
Units 1 and 2 in the expanded operating domain, as discussed below.
a) Anticipated Operational Occurrences (A00)
The core-wide A00 included in the LGS, Unit 1 Cycle 5 reload analyses were re-examined for operation in the Maximum Extended Load Line Limit (MELLL) region. Re-examination for operation in the MELLL region for LGS, Unit I will be done as part of its next cycle reload analyses. The analytical methods as well as the input assumptions are consistent with the bases for the LGS, Unit 1 Cycle 5 Core Operation Limits Report (COLR). No design or safety limits will be exceeded.
b) Reactor Vessel Overpressure Protection The Main Steamline Isolation Valve (MSIV) closure with a neutron flux scram event was analyzed at the 102% power /75%
flow point using the nuclear parameters resulting from the End .,
of Cycle (E0C) 5 target exposure shape. The results show the peak reactor vessel pressure (i.e.,1,264 psig) is below the 1,375 psig limit.
U. S. Nuclear Regulatory Commission August 27, 1993 Document Control Desk Page 10 e
c) Loss-of-Coolant Accident (LOCA)
The current Emergency Core Cooling System (ECCS) LOCA analysis s documented in Attachment 3, " Maximum Extended Load Line cimit and ARTS Improvement Program Analysis for Limerick Generating Station Units 1 and 2,"NEDC-32193P, Revision 1, dated July 1993, and already includes the proposed Average Power Range Monitor -
Rod Block Monitor Technical Specifications (ARTS) and MELLL Analysis (MELLLA) application.
The LOCA evaluation utilized the General Electric (GE)
SAFER /GESTR methodology. This mechodology was approved by the NRC as discussed in a letter from C. O. Thomas (NRC) to J. F.
Quirk (GE) " Acceptance for Referencir.9 of Licensing Topical Report NEDE-23785, Revision 1, Volume III(P), The GESTR-LOCA and SAFER Models for the Evaluation of the Loss-of-Coolant Accident,' dated June 1, 1984. The results of this bounding evaluation show for the application of ARTS /MELLLA- to LGS Units 1 and 2, the Peak Cladding Temperature (PCT) for a design basis LOCA at 102% core power /75% core flow is judged to be less than the 20* F increase compared to the rated core flow case with 10CFR50, Appendix K assumptions. With the introduction of ARTS, the setdown factor on the fl ow-referenced APRM rod block system is replaced with a set of power- and flow-dependent Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) and Minimum Critical Power Ratio (MCPR) adjustment factors. One of the criteria for the ARTS Improvement Program has been to ensure that the criteria in 10CFR50.46 are met through the application of the flow-dependent MAPLHGR multipliers. This analysis also establishes a new licensing basis peak cladding temperature of 1,310 *F at the current thermal power liinit of 3,293 MWt.
Bounding short-term containment response analysis of the design basis LOCA- event (i.e., a double-ended guillotine break of a recirculation line) was performed. The results show the peak containment drywell pressure is bounded by the LGS Updated Final Safety Analysis Report (UFSAR) analysis values and remain well below the design value of 55 psig. The bounding event for the containment drywell temperature response is a main steamline break. Under MELLL operation, the increased reactor vessel coolant subcooling has no impact on the steam break flow. Tharefora, the peak containment drywell temperature for MELLL operation is bounded by that presented for the main steamline break in the UFSAR Table 6.2-1 and is below the design temperature of 340 *F. The peak values of drywell pressure and temperature are relatively insensitive to reactor operating conditions.
U. S. Nuclear Regulatory Commission August 27, 1993 Document Control Desk Page 11 The containment dynamic loads analysis for a LOCA is based on the short-term LOCA analysis described above. The loads considered for MELLLA include pool swell, condensation oscillation, and chugging. The analysis results show the comparisons are bounded by the corresponding design basis load definition in the UFSAR. The peak containment wetwell airspace pressure during a suppression pool swell period is calculated to be 38.0 psig which is within the design limit of 55 psig.
In addition, the radiological analysis for a LOCA is not affected by implementation of the ARTS /MELLLA changes.
d) Results of the Anticipited Transients Witnu ' ram (ATWS) analysis conducted for operation in the MELLu .anain showed 4
that the maximum values of the key performance parameters (i.e., fuel cladding temperature and reactor vessel bottom pressure) were within the generic limits reported in GE's
" Assessment of BWR Mitigation ATWS," Vol. II, NED0-24222, dated February 1981. The other key performance parameter, suppression pool temperature, shows an increase above 190 *F due to operation in the MELLL region. As a result, injection of a sodium pentaborate solution with a higher enrichment of Boron-10 from the Standby Liquid Control Systen. is required to reduce reactivity sooner. Evaluation of the required Boron -
10 enrichment of 29% has been determined to be sufficient to satisfy this requirement. Thus, operation in the MELLL domain will have no adverse impact on the capability to mitigate postulated ATWS events in the expanded operating region.
e) Introduction of a statistically based Rod Withdrawal Error (RWE).
The proposed new RBM systerr. with power-dependent setpoints requires new RWE analyses be performed to determine the MCPR requirements and corresponding setpoints. The generic analysis and its effect on the MCPR safety limits and Critical Power Ratio (CPR) correlations are discussed in Attachment 3,
" Maximum Extended Load Line Limit and ARTS Improvement Program Analysis for Limerick Generating Station Units 1 and 2," NEDC-32193P, Revision 1, dated July 1993, Subsection 10.3.1. The new RWE analysis for LGS, Units 1 and 2, is valid for all GE fuel types including GE 11 and is also applicable to Asea Brown Boveri (ABB) and Siemens Nuclear Power (SNP) qualification fuel bundles. This analysis method will be applied to future core reload analyses to maintain the MCPR safety limit.
U. S. Nuclear Regulatory Commission August 27, 1993 Document Control Desk Page 12 f) Control Rod Drop Accident LGS, Units 1 and 2 employs banked position withdrawal sequencer (BPWS) for control rod movement. The Control Rod Drop Accident (CRDA) for BPWS plants have been generically analyzed for GE fuel designs (i.e., "GE Standard Application for Reactor Fuel, GESTAR II," NEDE-240ll-P-A-10 and GESTAR II United States Supplement, NEDE-24011-P-A-10-US, dated February 1991) and are applicable to ABB and SNP qualification fuel bundles. A CRDA event is a startup accident evaluated at hot and standby conditions which are unaffected by operation in the MELLL domain. There is no change to the CRDA analysis basis or results as presented in the NEDE reports cited above and therefore the conclusions for CRDA are applicable for operation in the MELLL domain and with the ARTS Improvement Program.
g) Fuel Loading Error Power operation does not impact the analysis of the fuel loading error accidents. Thus, the bases for the LGS, Unit I cycle 5 COLR are applicable to the proposed implementation of the ARTS /MELLLA.
h) Recirculation pump runout The results in Attachment 3, , " Maximum Extended Load Line Limit and ARTS Improvement Program Analysis for Limerick Generating Station Units 1 and 2,"NEDC-32193P Revision 1, dated July 1993 Table 8-1, show the Reactor Internal Pressure Differences are bounded for MELLL operation by the design basis results for recirculation pump runout along the rated rod line and thus, the reactor internals have adequate design margin for operation in the MELLL region.
i) Recirculation pump runback The recirculation pumps runback intermediate speed will be reset at 42%. The resulting power level when operating on the MELLL rod line is now sufficiently low enough to be within the nominal capacity of the two feedwater pumps in the event of a feedwater pump trip.
U. S. Nuclear Regulatory Commission August 27, 1993 Document Control Desk Page 13 j) Reactor Internals Vibration To support the operation of LGS, Units 1 and 2 in the MELLL region, the vibration measurements at Browns Ferry Unit I were analyzed to determine if there would be any detrimental effects to the reactor internals due to this mode of operation. ine results show the maximum steam flow that will be generated in the MELLL region will be no more than from rated power / flow condition. Therefore, the reactor internals inside the reactor vessel shroud and in the upper region of the reactor vessel will not be affected by operation in the HELLL region.
Vibration of the reactor internal components in the annulus region (i.e., outside the reactor vessel shroud) is expected to increase slightly due to the increase in the recirculation drive flow. However, this drive flow will never exceed that at rated core power and rated core flow. Therefore, the flow induced vibration effect on the reactor internal components outside the shroud during MELLL operation would not exceed the acceptance criteria currently established. Thus, operation in the MELLL region will not have any detrimental effects on the reactor internals due to flow induced vibration.
k) Single Loop Operation (SLO)
To support the additional operating domain above the rated line, three key issues were addressed in this study. These were the MCPR fuel cladding integrity safety limit, the MCPR operating limit, and the LOCA analysis for the SLO mode. The studies show the results for one-loop operation cases would not be more limiting than thermal and overpressure consequences of a two pump operation. This conclusica is applicable to SLO conditions within the LGS, Units 1 and 2 expanded operating domain.
The updated SAFER /GESTR-LOCA analysis for SLO was performed using the conservative input assumptions and a 0.9 MAPLHGR multiplier. The required MAPLHGR multipliers at the SLO power-flow condition are more restrictive than the value asr;med in the SAFER /GESTR-LOCA analysis for SLO. Therefore, this analysis for SLO is conservative and bounding for LGS, Units 1 and 2.
U. S. Nuclear Regulatory Commission August 27, 1993 Document Control Desk Page 14 No new component and/or system interactions that could lead to an accident are created by the proposed changes. No new challenges to equipment are involved with implementation of ARTS /MELLLA changes.
The probability of any accident is not increased by operating in the expanded operating domain because formulation of the flow-biased Average Power Range Monitor (APRM) rod block trip equation, including a new maximum value for the APRM rod block has been established to maintain margin between the APRM rod block setpoint and the APRM scram setpoint. Additionally, the proposed changes will have no effect on any accident initiating mechanisms. No equipment that is assumed to fail in an accident is affected by implementation of the ARTS /MELLLA changes. Equipment environment, operating conditions, and equipment interactions are not adversely affected by the proposed changes.
The radiological consequences of all analyzed events are unchanged and in addition, the consequences of all the transients will not cause the MCPR safety limit to be exceeded.
Therefore, the proposed TS changes do not involve an increase in the probability or consequences of an accident previously evaluated.
- 2. The proposed TS chances do not create the possibility of a new or different kind of accident from any accident previously evaluated.
Implementation of the ARTS /MELLLA changes does not create any new failure mode or sequence of events that can lead to an accident of a different type than any previously evaluated. The APRM rod block trip setpoint will continue to block control rod withdrawal when core power significantly exceeds normal limits and approaches the scram level. The APRM scram trip setpoint will continue to initiate a scram if the increasing core power / flow condition continues beyond the APRM rod block setpoint. The proposed changes to the RBM system have been designed to enhance the reliability and accuracy of the RBM system without impacting the degree of isolation of the RBM system from other plant systems. The function of the RBM system will not change. Implementation of the ARTS /MELLLA changes does not increase challenges or create any new challenge to safety-related systems or equipment, or other equipment whose failure could cause an accident. The SLCS retains the capability to shutdown the reactor as originally designed. Also, Implementation of the ARTS /MELLLA changes does not involve any new challenge to a fission product barrier.
Therefore, the proposed TS changes do not create the possibility of a new or different kind of accident from any previously evaluated.
l V. S. Nuclear Regulatory Commission August 27, 1993 !
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- 3. The proposed TS chanaes do not involve a sianificant reduction in a l marain of safety.
i The margin of safety as defined in the TS is not reduced because the results of all the safety analysis are within allowables values.
The margin of safety for the MCPR safety limit and the limits associated with a LOCA will be maintained. The peak analyzed containment pressure does not change and thus, the margin of safety for the containment does not change. The SLCS retains the capability to bring the reactor to a cold shutdown condition from full power steady state operating conditions, as originally designed.
Therefore, the proposed TS changes do not involve a reduction in a margin of safety.
Information Succortina an Environmental Assessment An environmental assessment is not required for the changes proposed by this Technical Specifications (TS) Change Request because the requested changes to the Limerick Generating Station (LGS), Units 1 and 2 TS conform to the criteria for " actions eligible for categorical exclusion," as specified in 10CFR51.22(c)(9). The requested changes will have no impact on the environment.
The proposed changes do not involve a significant hazards consideration as discussed in the preceding section. The proposed changes do not involve a significant change in the types or significant increase in the amounts of any effluents that may be released offsite. In addition, the proposed changes do not involve a significant increase in individual or cumulative occupational radiation exposure.
Conclusion i
The Plant Operations Review Committee and the Nuclear Review Board have reviewed these proposed changes to the Limerick Generating Station (LGS), Units l 1 and 2 Technical Specifications (TS) and have concluded that they do not involve l an unreviewed safety question, and will not endanger the health and safety of the l
public.
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