ML20211P463
| ML20211P463 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 10/08/1997 |
| From: | Berkow H NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20211P467 | List: |
| References | |
| NUDOCS 9710200201 | |
| Download: ML20211P463 (16) | |
Text
_ _ _ _ _ _ _ _ _ _ _ _.
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t UNITED STATES NUCLEAR REGULATORY COMMISSION o
I WASHINGTON. D.C. 30l#4001 SOUTHERN NUCLEAR OPERATING COMPANY. INC.
GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON. GEORGIA DOCKET NO. 50 321 EDWIN 1. HATCH NUCLEAR PLANT. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.210 License No. DPR 57 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment to the Edwin 1. Hatch Nuclear Plant, Unit 1 (the f acility) Facility Operating License No, DPR 57 filed by Southern Nuclear Operating Company, Inc. (Southern Nuclear), acting for itself, Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia (the licensess), dated May 9,1997, as supplemented September 3,1997, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B.
The fac3lity will operate in conformity with the application, the provisions of the Act, and the rules and 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 will not 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, cp (o.w o Y 9710200201 971008 PDR ADOCK 05000321 P
2-2.
Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR 57 is hereby amended to read as follows:
(2)
Technical Soecifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 210, are hereby incorporated in the license. Southern Nuclear 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 shall be implemented prior to Unit 1 startup from the fall 1997 refueling outage.
i FOR THE NUCLEAR REGULATORY COMMISSION 0f) 1
/
Her rt N. Berkow, Director Project Directorate ll 2 Division of Reactor Projects l/11 Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date of issuance: October 8, 1997 l
=.
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ATTACHMENT TO LICENSE AMENDMENT NO. 210 FACILITY OPERATING LICENSE NO. DPR 57
_ ppg,MQ,50 321 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain verticallines indicating the areas of change.
Remove inag1 3.3 19 3.3 19 5.019 5.019
(
l Control Rod Block Instrumentation 3.3.2.1 febte 3.3.2.1 1 (pese 1 of 1) j Control and Stock Instrumentation APPLICA8LE NC*)EE OR CTMER SPECIFIED REQUlkED SURVEILLANCE ALLOWABLE FUNCT10N CONDill0NS CHANNELS REQUltEMENTS VALUE 1.
Red Block Monitor s.
Low Power aanee -upscale (a) 2 SR 3.3.2.1.1 s 115.5/125 sa 3.3.2.1.4 divlstens of sa 3.3.2.1.7 futt scate b.
Intenaediate Power (b) 2 sa 3.3.2.1.1 s 109.7/125 teree = Upscale SR 3.3.2.1.4 divlelone of SR 3.3.2.1.7 fwlt scale c.
Nish Power sense-upscote (c) 2 SR 3.3.2.1.1 s 105.9/125 I
st 3.3.2.1.4 divisions of SR 3.3.2.1.7 futI seate W.
Insp (d) 2 SR 3.3.2.1.1 mA I
e.
Downscete (d) 2 sa 3.3.2.1.1 t 93/125 1
sa 3.3.2.1.7 divisions of futt scale I '),2(')
1 st 3.3.2.1.2 mA i
I
- 2. and Worth Mintelser sa 3.3.2.1.3 sa 3.3.2.1.5 sa 3.3.2.1.8 3.
asector mode Switch-shutdown (f) 2 sa 3.3.2.1.6 NA 1
Posttion s
(a) THERMAL POWER R 291 and
- M1 RTP.
(b) THERMAL POWER t M1 and
- M1 ATP.
(c) TMERMAL POWR R M1 (d) THERMAL POWER t 291.
(e) With TMERMAL POWER < 101 RTP.
(f) Reactor mode awltch in the shutoown position.
HATCH UNIT 1 3.3-19 Amendment No. 210
i Reporting Requirements 5.6 1
5.6 Reporting Requirements (continued).
5.6.5 f0RE OPERATING LIMITS REPORT (COLR) a.
Core operating limits shall be established prior to each reload cycle, or prior to any remaining l>ortion of a reload cycle, and shall be documented in the CO.R 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.
b.
The analytical methods used to determine the core operating limits ; all be those previously reviewed and' approved by the NRC, specifically those described in the following documentst 1)
NEDE44011-P-A, " General Electric Standard Application for hactor Fuel," (applicable amendment specified in the COLR).
2)
" Safety tvaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No. 157 to Facility Operating License DPR-57," dated September 12, 1988.
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 mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.
(continued)
HATCH UNIT 1 5.0-19 Amendment No. 210
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Control Rod Block Instrumentation B 3.3.2.1-B 3.3 -INSTRUMENTATION B 3.3.2.1 Control Rod Block-Instrumentation BASES BACKGROUND Control rods provide the primary means for control of reactivity changes.
Control rod block instrumentation includes channel sensors, logic circuitry, switches, and relays that are designed to ensure that the fuel cladding integrity safety limit, and specified fuel design limits are not violated during postulated transients and accidents.
During high power operation, the rod block monitor (RBM) provides protection for control rod withdrawal error events.
During low power operations, control rod blocks from the rod worth minimizer (RWM) enforce specific control rod sequences designed to miti ate the consequences of the control rod t
drop accident (C WA). During shutdown conditions, control rod blocks from the Reactor Mode Switch - Shutdown Position Function ensure that all control rods remain-inserted to prevent inadvertent criticalities.
The purpose of the RBM is to limit control rod withdrawal if localized neutron flux exceeds a predetermined setpoint during control rod manipulations.
It is assumed to function to block further control rod withdrawal to preclude a violation of the MCPR Safety Limit (SL)The RBM sup>1ies a or a specified acceptable fuel design limit (SAFDL).
L trip signal to the Reactor Manual Control System (tMCS) to appropriately inhibit control rod withdrawal during power operation above the low power range setpoint. The RBM has two channels, either of which can initiate a control rod-block when the channel output exceeds the control rod block setpoint. One RBM channel inputs into one RMCS rod block circuit and the other RBM channel inputs into the second RMCS rod block circuit.
l The RBM channel signal is generated by averaging a set of local power range monitor.
heights surrounding the con (LPRM) signals at various core trol rod being withdrawn. A signal from one of the four redundant average power range monitor (APRM) channels supplies a reference: signal for one of the RBM channels, and a signal from another of the APRM channels supplies the reference signal to the second RBM channel. This reference signal.is used to determine which RBM range setpoint (low, intermediate, or high) is enabled.
If the APRM is indicating less than the low power range -
setpoint, the RBM is automatically bypassed. The RBM
~
(continued)
HATCH UNIT 1 B 3.3-42 Amendment No. 210
Control Rod Block Instrumentation B 3.3.2.1 BASES BACKGROUND is also automatically bypassed if a peripheral control (continued) rod is selected (Ref. 1). A rod block signal is also generated if an RBM Downscale trip or an Inoperable trip occurs. The Downscale trip will occur if the RBM channel signal decreases below the Downscale trip setpoint after the RBM signal has been normalized. The Inoperable trip will occur during the nullino (normalization)' sequence, if: the RBM channel fails to null, too few LPRM inputs are available, a module is not plugged in, or the icnction switch is moved to any position other than " Operate."
The purpose of the RWM is to control rod patterns during startup and shutdown, such that only specified control rod sequences and relative positions are allowed over the operating range from all control rods inserted to 10% RTP.
The sequences effectively limit the potential amount and rate of reactivity increase during a CRDA.
Prescribed control rod sequences are stored in the RWM, which will initiate control rod withdrawal and insert blocks when the actual sequence deviates beyond allowances from the stored sequence. The RWM determines the actual sequence based position indication for each control rod. The RWM also uses feedwater flow and steam flow signals to determine when the reactor power is above the preset power level at which the RWM is automatically bypassed (Ref. 2).
The RWM is a single channel system that provides input into both RMCS rod block circuits.
'With the reactor mode switch in the shutdown position, a control rod withdrawal block is applied to all control rods to ensure that the shutdown condition is maintained. This Function prevents inadvertent criticality as the result of a control rod withdrawal during MODE 3 or 4, or during MODE 5 when the reactor mode switch is recuired to be in the shutdown channels, position. The reactor moce switch has two each inputting into a separate RMCS rod block circuit. A rod block in either RMCS circuit will provide a control rod block to all control rods.
(continued)
HATCH UNIT 1 B 3.3-43 Amendment No. 210
7..
Control Rod Block Instrumentation B 3.3.2.1 BASES
-APPLICABLE 1.
Rod Block Monitor (continued)
SAFETY ANALYSES, LCO, and effects (for channels that must function in harsh APPLICABILITY environments as defined by 10 CFR 50.49) are accounted for.
The RBM is assumed to mitigate the consequences of an RWE l
event when operating k 29% RTP.
Below this power level, the consequences of an RWE event will not violate the MCPR SL or i
the 1% plastic strain design limit; therefore, the RBM is not required to be OPERABL- (Ref. 3).
I 2.
Rod Worth Minimirer i
The RWM enforces the banked position withdrawal sequence (BPWS) to ensure that the initial conditions of the CRDA analysis are not violated. The analytical methods and assumptions used in evaluating the CRDA are summarized in i
References 4, 5, 6, and 7.
In addition, the Reference 6 i
analysis (Generic BPWS analysis) may be modified by plant specific evaluations. The BPWS requires that control rods be moved in groups, with all control rods a; signed to a l
specific group required to be within specified banked positions. Requirements that the control rod sequence is in compliance with the BPWS are specified in LCO 3.1.6, " Rod Pattern Control."
5 The RWM Function satisfies Criterion 3 of the NRC Policy Statement (Raf. 10).
Since the RWM is a system designed to act as a tackup to 1
operator control of the rod sequences, only one channel of the RWM is available and required to be OPERABLE (Ref. 7).
j Special circumstances provided for in the Required Action of LCO 3.1.3, " Control Rod OPERABILITY," and LCO 3.1.6 may l
necessitate bypassing the RWM to allow continued operation with inoperable control rods, or to allow correction of a control rod pattern not in compliance with the BPWS. The
!r (continued) 1 HATCH UNIT 1 B 3.3-45 Amendment No. 210 l
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. k UNITED STATES e
3 NUCLEAR RECULATORY COMMISSION War,eilNGToN, D.C. SpeeMcM SOUTHERN NUCLEAR OPERATING COMPANY. INC.
GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF OALTON. GEORGIA DOCKET NO. 50 366 EDWIN 1. HATCH NUCLEAR PLANT. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.151 License No. NPF 5
-1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment to the Edwin 1. Hatch Nuclear Plant, Unit 2 (the facility) Facility Operating License No. NPF 5 filed by Southern Nuclear Operating Company, Inc. (Southern Nuclear), acting for itself, Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia (the licensees), dated May 9,1997, as supplemented September 3,1997, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I;
. B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and 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 will not 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 Ulth 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2-2.
Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF 5 is hereby amended to read as follows:
(2)
Technical Snecifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No.151 are hereby incorporated in the license.
Southern Nuclear 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 uf issuance and shall be implemented within 30 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION He ert N. Berkow, Director Pr ect Directorate ll 2 Division of Reactor Projects - 1/II Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes
' Date of issuance:
October 8,1997
ATTACHMENT TO LICENSE AMENDMENT NO.151 FACILITY OPERATING LICENSE NO. NPF 5 DOCKET NO. 50 366 Replece the following pages of the Appendix *A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain verticallines indicating the areas of change.
Remove ADaan 3.3 20 3.3 20 5.0 15 5.019 k
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Control Rod Block Instrumentation 3.3.2.1 febte 3.3.2.1 1 (pepe 1 of 1)
C(1 trol Rod Block Instrwentation APPLICARLt MODE 8 CR CTER j
SPECIFit0 REQUIRED SURVEILLANCE ALL(WABLE FUNCTION CONDIT10NS CNAhWELS REgu1REMEN18 VALUE l
1.
Red tieck monitor s.
tow Power Rence-upscale (e) 2 st 3.3.2.1.1 s 115.5/125 SR 3.3.2.1.4 divialens of SR 3.3.2.1.7 futt scale b.
Intermediate Power (b) 2 sa 3.3.2.1.1 5 109.7/125 Renee a upscele SA 3.3.2.1.4 divistens of sa 3.3.2.1.7 futi seete c.
Nish Power Renee -upscote (c) 2 sa 3.3.2.1.1 s 105.9/125 SR 3.3.2.1.4 divialens of I
84 3.3.2.1.7 futi scale d.
Insp (d) 2 SA 3.3.2.1'.1 mA l
e.
Downscete (d) 2 84 3.3.2.1.1 t 93/125 SR 3.3.2.1.7 divistens of l
j futt scale 2.
Red Worth Minleiter 1('),2(')
1 sa 3.3.2.1.2 NA sa 3.3.2.1.3 g
st 3.3.2.1.5 j
SA 3.3.2.1.5 i
3.
Reacter Nede switch-shutdown (f) 2 sa 3.3.2.1.6 MA Positten
't e) T M RMAL POWER R 295 and 4 M 1 RTP.
(b) TMRMAL POWER R M1 and
- 841 RTP.
(c) THERMAL PCWER g $41.
i (d) 1MRMAL POWER R 291.
(e) With TERMAL PCWER e 101 RTP.
(f) Asoctor mode switch in the shutdown positten.
1
)
HATCH UNIT 2 3.3-20 Amendment No. 151 i
i
Reporting Requirements 5.6 5.6 Reporting Requirements (continued)-
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, tnd shall be documented in the CO.R for the followir.g:
1)
The Average Planar Linear Heat Generation Rate for Specification 3.2.1.
2)
The Minimum Critical Power Ratio for Specification 3.2.2.
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)
NEDE-24011-P-A, " General Electric Standard Application for Reactor Fuel," (applicable amendment specified in the COLR).
2)
" Safety Evaluation by the Office of Nuclear Reactor Regulation. Supporting Amendment Nos. 151 and 89 to Facility Operating Licenses DPR-57 and NPF-5," dated l
January 22, 1988.
c.
The core operating limits shall be detertained 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 SDN, transient analysis limits and accident analysis limits-) of the safety analysis are met, d.
The COLR, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.
(continued)
HATCH UNIT 2 5.0-19 Amendment No.151
B 3.3.2.1 B 3.3 INSTRUMENTATION B 3.3.2.1 Control Rod Block Instrumentation BASES BACKGROUND Control rods provide the primary means for control of reactivity changes.
Control rod block instrumentation includes channel sensors logic circuitry, switches, and relays that are desianed,to ensure that the fuel cladding integrity safety limit limits are not violated (SL)ing postulated transients andand the specified fuel dur i
accidents. During high power operation, the rod block monitor (RBM) provides protection for control rod withdrawal error events. During low power operations, control rod blocks from the rod worth minimizer (R ) enforce specific control rod sequences designed to miti te the consequences of the control rod drop accident (CRDA.
During shutdown conditions, control rod blocks from th Reactor Mode Switch - Shutdown Position Function ensure that all control rods remain inserted to prevent inadvertent critica11 ties.
The purpose of the RBM is to limit control rod withdrawal if localized neutron flux exceeds a during control rod manipulations. predetermined setpoint
-It is assumed to function to block further control rod withdrawal to preclude a violation of the MCPR SL or a specified acceptable fuel l
design limit CSAFDL). The RBM supplies a trip signal to the I
Reactor Manual Control System (durin)g power operation above RMCS to appropriately inhibit control rod withdrawal the low power range setpoint.
The RBM has two channels, either of which can initiate a control rod block when the i
channel output exceeds the control rod block setpoint. One RBM channel inputs into one RMCS rod block circuit and the other RBM channel inputs into the second RMCS rod block circuit.
The RBM channel signal is generated by averaging a set of local power rante monitor (LPRM) signals at various core heights surrounding the control rod being withdrawn. A signal from one of the four redundant average power range monitor (APRM) channels supplies a reference signal for one of'the RBM channels, and a signal from another of the RBM channels su) plies the reference signal to the second RBM channel T11s reference signal is used to determine which RBM range setpoint (low, intermediate, or high) is enabled.
If the APRM is indicating less than the low setpoint, the RBM is automatically bypassed. power rangeThe RBM is also automatically bypassed if a peripheral control rod is (continued)
HATCH UNIT 2 B 3.3-42 Amendment No. 151
Control Rod Block Instrumentation B 3.3.2.1 BASES i
j BACKGROUND selected (Ref.1). A rod block signal is also gene' rated if (continued) an RBM Downscale trip or an Inoperable trip occurs.
The i
Downscale trip willoccur if the RBM channel signal decreases below the Downscale trip setpoint after the RBM signal has i
been normalized. The Inoperable trip will occur during the nulling (normalization) sequence, if:
the RBM channel fails to null, too few LPRM inputs are available, a module is not plugged in, or the function switch is moved to any position other than " Operate."
2 The purpose of the RWM it to control rod patterns during startup and shutdown, such that only specified control rod sequences and relative positions are allowed over the operating range from all control rods inserted to 10% RTP, The sequences effectively limit the potential amount and rate of reactivity increase during a CRDA.
Prescribed l
control rod sequences are stored in the RWM, which will initiate control rod withdrawal and insert blocks when the actual sequence deviates beyond allowances from the stored i
sequence. The RWM determines the actual sequence based position indication for each control rod. The RWM also uses feedwater flow and steam flow signals to determine when the reactor power is above the preset power level at which the 4
RWM is automatically bypassed (Ref. 2). The RWM is a single channel system that provides input into both RMCS rod block circuits.
)
With the reactor mode switch in the shutdown position, a control rod withdrawal block is applied to all control rods to enwre that the shutdown condition is maintained. This Function prevents inadvertent criticality as the result of a i
control rod withdrawal during MODE 3 or 4, or during MODE 5 l
when the reactor mode switch is required to be in the l
shutdown position. The reactor mode switch has two channels, each inputting into a separate RMCS rod block circuit. A rod block in eithar RMCS circuit will provide a
}
control rod block to all control rods.
3 L
i j
(continued) f HATCH UNIT 2 B 3.3-43 Amendment-No. 151 l
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Control Rod Block Instrumentation B 3.3e201 i
BASES APPLICABLE L
Rod Block Monitor (continued)
SAFETY ANALYSES, LCO, and effects (for channels that must function in harsh APPLICABILITY environments as defined by 10 CFR 50.49) are accounted for.
The RBM is assumed to mitigate thw consequences of an RWE event when operating a: 29% RTP.
Below this power level, the i
consequences of an RWE event will not violate the MCPR SL or 4
the 1% plastic strain design limit; ?.herefore, the RBM is notrequiredtobeOPERABL-(Ref.3).
2.
Rod Vorth Minimiter The RWM enforces the banked position withdrawal sequence (BPWS to ensure that the initial conditions of the CRDA l
analys)is are not violated.
4 The analytical methods and assumptions used in evaluating the CRDA are summarized in References 4, 5, 6, and 7.
In addition, the Reference 6 analysis (Generic BPWS analysis) may be modified by plant specific evaluations. The BPWS requires that control rods be moved in groups, with all control rods assigned to a specific group required to be within specified banked positions. Requirements that the control rod sequence is in compliance with the BPWS are specified in LCO 3.1.6, " Rod Pattern Control.
The RWM Function satisfies Criterion 3 of the NRC Policy Statement (Ref. 10).
Since the RWM is a system designed to act as a backup to o>erator control of the rod sequences, only one channel of tle RWM is available and required to be OPERABLE (Ref. 7).
Special circumstances provided for in the Required Action of LCO 3.3.3, " Control Rod OPERABILITY," and LCO 3.1.6 may necessitate bypassing the RWM to allow continu0d operation with inoperable control rods, or to allow cmeetion of a control rod ptttern not in compliance with the BPWS. The (continued)
HATCH UNIT 2 B 3.3-45 Amendment No. 151
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