ML20034C790
| ML20034C790 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 04/26/1990 |
| From: | PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | |
| Shared Package | |
| ML20034C376 | List: |
| References | |
| NUDOCS 9005110101 | |
| Download: ML20034C790 (9) | |
Text
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Unit 2 PBAPS
=
J NOTES FOR TABLE 3.2.C, i
1.
For the starte and run positions of the Reactor Mode Selector Switch, there
{
shs11 be two operable or tripped trip systems for each function.
The SRM and 1
IRM blocks need not be operable in "Run" mode, and the APRM and RBM rod blocks need not be operable in "Startup" mode.
If the first column cannot be met for one of the two trip systems, this condition may exist for up to seven days i
provided that during that time the operable system is functionally tested Immediately and daily tnereafter; if this condition lasts longer than seven days, the system shall be tripped.
If the first column cannot be met for both trip systems, the systems shall be tripped.
2.
The equation for Trip Level Setting will be used in the event of operation with l
[
a maximum fraction of limiting power density (MFLPD) greater than the fraction of rated power (FRP) where:
FRP = fraction of rated thermal power (3293 HWt)
MFLPD = maximum fraction of limiting power density where the limiting power density is the value specified in the CORE OPERATING LIMITS REPORT.
The ratio of FRP to MFLPD shall be set equal to 1.0 unless the actual operating value is less than the design value of 1.0, in which case the actual operating value will be used.
W = Loop Recirculation flow in percent of design. W is 100 for core flow of 102.5 million lb/hr or greater.
Trip level setting is in percent of rated power (3293 MWt).
4W is the difference between two loop and single' loop effective recirculation drive flow rate at the same core flow. _ During single loop operation, the re%ction in trip setting is accomplished by correcting the flow input of the f W biased rod block to preserve the original (two loop) relationship between the rod block setpoint and recirculation drive flow, or by adjusting the roo block setting. AW = 0 for two loop operation.
3.
IRM downscale is bypassed when it is on its lowest range.
4.
This function is bypassed when the count rate is > 100 cps.
5.
One of the four SRM inputs may be bypassed.
6.
This SRM function is bypassed when the IRi4 range switches are on range 8 or above.
7.
The trip is bypassed when the reactor power is < 30%.
8.
This function is bypassed when the mode switch is placed in Run. 9005110101 900426 ADOCK050g7 PDR P
Unit 2 PBAPS 3.5. BASES (Cont'd) l J.
Local LHGR This specification assures that the linear heat generation rate in any 8XB fuel rod is less than the design linear heat generation.
The maximum LHGR shall be checked daily during reactor operation at > 25%
power to determine if fuel burnup, or control rod movement has caused changes in power distribution.
For LHGR to be at the design LHGR below 25% rated thermal power, the peak local LHGR must be a factor of cpproximately ten (10) greater than the average LHGR which is precluded by a considerable margin when employing any permissible control rod pattern.
K.
Minimum Critical Power Ratio (MCPR)
Operating Limit MCPR The required operating limit MCPR's at steady state operating conditiona are derived from the established fuel cladding integrity Safety Limit MCPR and analyses of the abnormal operational transients presented in Supplemental Reload Licensing Analysis and Reference 7.
For any abnormal operating transient analysis evaluation with the initial condition of the reactor being at the steady state operating limit it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip setting given in Specification 2.1.
To assure that the fuel cladding integrity Safety Limit is not violated during any anticipated abnormal operational transient, the most limiting t2cnsients have been analyzed to determine which result in the largest reduction in critical power ratio (CPR).
The transients evaluated are as described in reference 7.
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Unit 2 PBAPS 3.5.K. BASES (Cont'd)
The largest reduction in critical power ratio is then added to the fuel cladding integrity safety 1.imit MCPR to establish the MCPR operating Limit for each fuel type.
Analysis of the abnormal operational transients is presented in Reference 7.
Input data and operating conditions used in this cnalysis are shown in Reference 7 and in the Supplemental Reload i
Licensing Analysis, i
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3.5.L.
Average Planar LHGR (APLHGR), Local LHGR and Minimum 5
Critical Power Ratio (MCPR)
In the event that the calculated value of APLHGR, LHGR or MCPR exceeds-its limiting value, a determination is made to ascertain the cause and initiate corrective action to restore the value to within prescribed limits.
The status of all indicated limiting fuel bundles is reviewed cs well as input data associated with the limiting values such as power distribution, instrumentation data (Traversing In-Core Probe -
l TIP, Local Power Range Monitor - LPRM, and teactor heat balance instrumentation), control rod configuration, etc., in order to determine whether the calculated values are valid.
In the event that the review indicates that the calculated'value oxceeding limite is valid, corrective action is immediately undertaken to restore the value to within prescribed limits.
Following corrective action, which may involve alterations to the control rod configuration and consequently changes to the core power distribution, revised instrumentation data, including changes to the relative nautron flux distribution, for up to 43 in-core locations is obtained and the power distribution, APLHGR, LHGR and MCPR calculated.
Corrective action is initiated within one hour of an indicated value oxceeding limits and verification that the indicated value is within prescribed limits is obtained within five hours of the initial indicaticn.
In the event that the calculated value of APLHGR, LHGR or MCPR 1
oxceeding its limiting value is not valid,-i.e., due to an erroneous instrumentation indication, etc., corrective action is initiated-within one hour of an indicated value exceeding limits.
Verification i
that the indicated value is within prescribed limits is obtained within five hours of the initial indication.
Such an invalid indication would not be a violation of the limiting condition for operation and therefore would not constitute a reportable occurrence.
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- t PBAPS' '
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6.9.1 RoutineReports(Cont'd) c.
Annual Safety / Relief Valve Report Describe all challenges "a the primary coolant system safety and relief-valves. Challenges are dt.iined as the automatic opening of the primary coolant safety or relief valves in response to high reactor pressure.
i d.
Monthly Operating R,,eport.
1 Routine reports of operating statistics and shutdown experience,and a tiarrative summary of the operating experience shall be submitted on a monthly basis to the Office of Management and Program Analysis (or its -
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successor), U.S. Nuclear Regulatory Commission, Washington, DC 20555, with a copy to the appropriate Regional Office, to he submitted no later'than the 15th of the month following the calendar month covered by the report.
e.
Core Opefating Limits Report (1) Core operating limits shall be-established and shall be documented in the CORE OPERATING LIMITS REPORT prior to each Operating Cycle, or prior to any remaining portion of an Operating Cycle, for the following a.
The APLEGR for Specification 3.5.1, b.
The MCPR for Specification 3.5.K.
The Kf core flow adjustment factor for $pecification 3.:i.K.
c.
I d.
The LHGR for Specification 3.5.J.
The upscale flow biased Rod Block Monitor setpoint and the upscale e.
high flow clamped Rod Block monitor setpoint ot' Specification 3.2.C.
].
i (2)
The analytical methods used to determine the core operating limits shall be i
those previously reviewed and approved by the NRC, specifically those I
described in the following documents as amended and approved:
NEDE-24011-P-A, " General Electric Standard Application for Reactor a.
Fuel" (latest approved version) b.
Philadelohia Electric Company Methodologies as described in:
(1) PECo-FMS-0001-A, " Steady-State Thermal Hydraulic Analysis of Peach Bottom Units 2 and 3 using FIBWR Computer Code" (2) PECo-FMS-0002-A, " Method for Calculating Transient Criticel Power RatiosforBoilingWaterReactors(RETRAN-TCPPECo)"
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Unit 2 PBAPS 6.9.1 Routine Reports (Cont'd)
(3) PECo-FMS-0003-A, " Steady-State Fuel performance Methods Report" (4) PECo-FMS-0004-A, " Methods for Performing BWR Systems Transient Analysis" (5) PECo-FMS-0005-A, " Methods for Performing BWR Steady-State Reactor Physics Analysis" (3) The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical i
limits, core thermal-hydraulic limits, ECCS limits, noc11ar limits such as shutdown margin, transient analysis limits,andaccidentanalysislimits)ofthesafety analysis are met.
(4) The CORE OPERATlHG LIMITS REPORT, including any mid-cycle revisions or supplements, shall be submitted upon issuance for each Operating Cycle to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.
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i Unit 3 1
PBAPS i
l NOTES FOR TABLE 3.2.C 1.
For the startup and run positions of the Reactor Mode Selector Switch, there shall be two operable or tripped trip systems for each function..The SRM and IRM blocks need not be operable in "Run" mode, and the APRM and RBM rod blocks need not be operable in "Startup" mode.
If the first column cannot be met for e
one of the two trip systems, this condition may exist for up to seven days j
provided that during that time the operable system is functionally tested immediately and daily thereaftert if this condition lasts longer than seven i
days, the system shall be tripped.
If the first column cannot be met for both trip systems, the systems shall be tripped.
2.
The equation for Trip Level Setting will be used in the event of operation with-l a maximum fraction of limiting power density (MFLPD) greater than the fraction j
of rated power (FRP) where i
FRP=fractionofratedthermalpower(3293MWt)
MFLPD = maximum fraction of limiting power density where the limiting power density is the value specified in the CORE OPERATING LIMITS REPORT.
The ratio of FRP to MFLPD shall be set equal to 1.0 unless the actual operating v; Je is less than the design value of 1.0, in which case the actual operating value will be used.
i W = Loop Recirculation flow in percent of design. W is 100 for core flow of 102.5 million lb/hr or greater.
Trip level setting is in percent of rated power (3293 MWt).
AW is the difference between two 1000 and single loop effective recirculation drive flow rate at the same core flow. During single loop operation, the reduction in trip setting is accomplished by correcting the flow input of the flow biased rod block to preserve the original (two loop) relationship between the rod block setpoint and recirculation drive flow, or by l
l adjusting the rod block setting.
AW = 0 for two loop operation.
~
3.
IRM downscale is bypassed when it is on its lowest range.
l 4.
This function is bypassed when the count rate is 3,100 cps.
l 5.
One of the four SRM inputs may be bypassed.
i 6.
This SRM function is bypassed when the IRM range switches are on range 8 or above.
7.
The trip is bypassed when the reactor power is < 30%.
8.
This-function is bypassed when the mode switch is placed in Run.
l ^
l
Unit 3 PBAPS 3.5 BASES (Cont'd) l J.
Local LHGR This specification assures that the linear heat generation rate in any 8X8 fuel rod is less than the design linear heat generation.
The maximum LHGR shall be checked daily during reactor operation at > 25%
power to determine if fuel burnup, or control rod movement has caused changes in power distribution.
For LHGR to be at the design LHGR below 25% rated thermal power, the peak local LHGR must be a factor of epproximately ten (10) greater than the average LHGR which is precluded by a considerable margin when employing any permissible control rod pattern.
K.
Minimum Critical Power Ratio (MCPR)
Operating Limit MCPR The required operating limit MCPR's at steady state operating conditions are derived from the established fuel cladding integrity Safety Limit MCPR and analyses of the abnormal operational transients presented in Supplemental Reload Licensing Analysis and Reference 7.
For any abnormal operating transient analysis evaluation with the initial condition of the reactor being at the steady state operating limit it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip setting given in Specification 2.1.
To assure that the fuel cladding integrity Safety Limit is not l
violated during any anticipated abnormal operational transient, the t
most limiting transients have been analyzed to determine which result in the largest reduction in critical power ratio (CPR).
The transients evaluated are as described in reference 7.
F
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/
'e Unit 3 PBAPS 6.9.1 Ro,utine Reports (Cont'd) c.
Annual Safety / Relief Valve Report Describe all challenges to the primary coolant system safety and relief valves. Challenges are defined as the automatic opening of the primary coolant safety or relief valves in response to high reactor pressure.
d.
. Monthly Operating Report Routine reports of operating statistics and shutdown experience,and a narrative summary of the operating experience shall be submitted on a monthly basis to the Office of Management and Program Analysis-(or its successor), U.S. Nuclear Regulatory Comission, Washington, DC 20555, with a copy to the appropriate Regional Office, to be submitted no later than the 15th of the month following the calendar month covered by the report.
e.
Core Operating Limits Report (1) Core operating limits sna11 be established and shall be documented in the CORE OPERATING LIMITS REPORT prior to each Operating Cycle, or prior to any remaining portion of an Operating Cycle, for the following:
a.
The APLHGR for Specification 3.5.1, b.
The MCPR for Specification 3.5.K.
The Kf core flow adjustment factor for Specification 3.5.K.
c.
d.
The LiiGR for Specification 3.5.J.
The upscale flow biased Rod Block Monitor setpoint and the upscale e.
high flow clamped Rod Block monitor setpoint of' Specification 3.2.C.
(2)
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 as amended and approved:
NEDE-24011-P-A, " General Electric Standard Application for Reactor a.
Fuel" (latest approved version) b.
Philadelphia Electric Company Methodologies as described in:
(1) PECo-FMS-0001-A, " Steady-State Thermal Hydraulic Analysis of Peach Bottom Units 2 and 3 using FIBWR Computer Code" (2) PECo-FMS-0002-A, " Method for Calculating Transient Critical Power i
Ratios for Boiling Water Reactors (RETRAN-TCPPECo)"
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i Unit 3 P8APS
.6.9.1 Routine Reports (Cont'd)
(3)
PECo-FMS-0003-A, " Steady-State fuel Performance Methods Report" (4)
PECo-FMS-0004-A, " Methods for Performing BWR Systems Transient Analysis" (5)
PECo-FMS-0005-A,
- Methods for Performing BWR Steady-State Reactor Physics Analysis" i
(3) The core operating limits ?*all be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits ECCS limits, nuclear limits such as shutdown margin,. transient analysis limits. and accident analysis l
limits) of the safety analysis are met.
(4) The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements, shall be submitted upon issuance for each Operating Cycle to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.
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