ML20066K534
| ML20066K534 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 02/20/1991 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | GPU Nuclear Corp, Jersey Central Power & Light Co |
| Shared Package | |
| ML20066K539 | List: |
| References | |
| DPR-16-A-147 NUDOCS 9102270242 | |
| Download: ML20066K534 (14) | |
Text
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NUCLE AR REGULATORY COMMISSION f
.I W ASHING T ON, D. C, 20566
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5, GPU NUCLEAR COPPORATION AND JERSEY CENTRAL POWER & LIGHT COMPANY DOCKET NO. 50-219 OYSTER CREEK NUCLEAR GENERATING STATION AMENDMENT TO PROVISIONAL _0PERATING LICENSE Arendment No.147 License No. DPR-16 1.
The Nuclear Regulatory Comission (the Conunission) has found that:
A.
The application for amendment by GPU Nuclear Corporation, et al.,
(the licensee), dated May 7 1990 as supplemented September 14, 1990 and December 13, 1990, complieswIththestandardsandrequirements oftheAtomicEnergyActof1954,asamended(theAct),andthe Commission's rules and regulations set forth in 10 CFR Chaptor I; B.
The facility will operate in conf nnity 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 amendnent can be conducted without endangerin9 the health and sdfety of ihe public, and (ii) that such activities will be canduc',ed in compliance with the Comission's regulations; 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.
9102270242 910220 PDR ADOCK 05000219 P
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2 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 Provisional Operating License No. DPR-16 is hereby amended to read as follows:
(2) Technical-Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.147, are hereby incorporated in the license.
GPU Nuclear Corporation shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of issuance, to be implemented within 30 days of issuance.
FOR THE NUCLEAR REGULATORY COMISSION
- /0 f
'I'W John F. Stolz, Director Project Directorate 1-4 Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications
- Date of Issuance:
February 20. 1991
ATTACHMENT TO LICENSE AMENDMENT NO.147 i
PROV1_SIONAL OPERATING LICENSE NO. DPR-16 DOCKET NO. 50-219 i
Replace the following pages of the Appendix A Technical Specifications with the enclosed pages as indicated. The revised pages are identified by amendment number and contair, vertical lines indicating the areas of change.
- Remove, insert Page i Page i Page 1.0 4 Page 1.0 4 Page 1.0-7 Page 1.0 7 Page 1.0-8 Page 3.10-1 Page 3.10-1 Page 3.10 2 Page 3.10-2 Page 3.10 3 Page 3.10 3 Page 3.10 4 Page 3.10-4 Page 3.10-5 l
Page 3.10-6 I
Page 3.10-6a Page 3.10-7 Page 3.10-8 Page 3.10-9 Page 3.10-10 Page 3.10-11 Page 3.10-12 Page 6.12 Page 6.12 Page 6.12a Page 6.12b I
i 1
4
TAnte or coxTggIA section 1 Definitions Page 1.1 Oterable - Operability 1.0-1 1.2 operating 1.0 1 1.3 Power Operation 1.0-1 1.4 startup Mode 1.0-1 1.6 Run Mode 1.0-1 1.6 shutdown Conditicn 1.0-1 1.7 Cold shutdown 1.0-2 1.8 Place in shutdown condition 1.0-2 1.9 Place in Cold shutdown condition 1.0-2 1.10 Place in teolated condition 1.0-2 2.11 Refuel Mode 1.0-2 1.12 Refueling Outage 1.0-2 1.13 Primary Containment Integrity 1.0-2 1.14 secondary containment Integrity 1.0-3 1.15 Deleted 1.0-3 1.16 Rated riux 1.0-3 1.17 Reactor Thermal Power-To-Water 1.0-3 I
1.18 Protective Instrumentation Logic Definitions 1.0 3 1.19 Instrumentation surveillance Definitions 1.0-4 1.20 TDSAR 1.0-4
)
1.21 Core Alteration 1.0-4 1.22-Minimum Critical Power Ratio 1.0-4 1.23 staggered Test Basis 1.0-4 1.24 surveillance Requirements 1.0-5 1.25 Fire suppression Water system 1.0-b 1.26-Traction of Limiting Power Density (FLPD) 1.0-5 1.27 Maximum Traction of Limiting Power Density (MrLPD)1.0-5 1.28 Fraction of Rated Power (TRP) 1.0-6 1.29 Top of Active ruel (TAT) 1.0-6 1.30 Reiertable Event 1.0-6 1.31 Identified Leakage 1.0-6 1.32 Unidentified Leaksge 1.0-6 1.33 Process Control Plan 1.0-6 1.34 Augmented of fgas system (AOC) 1.0-6 1.35 Member of the Public-1.0-6 1.36 offsite Dose Calculation Manual 1.0-6 1.37 Purge 1.0-7 1.38 Exclusion Area 1.0-7 1.39 Reactor Vessel Pressure Testing 1.0-7 1.40 substantive Changes
-1.0-7 1.41 Dose Equivalent 1-131 1.0-7 1.42 Average Planar Linear Heat Generation Pate 1.0-7 1.43 Core operating Limits Report 1.0-8 1.44 Local Linear Heat Generation Rate 1.0-8 section 2 safety Limits and Limiting safety system settings Etat 2.1 safety Limit - Fuel Cladding Integrity 2.1-1 2.2_
safety Limit - Reactor Coolant syetem Pressure 2.2-1 2.3 Limiting safety system settings 2.3-1 OYSTER CREEK-1 Amendment No.: 64, 84, 97, 108. 147
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1.19 IN!?p"MrNTAT!cN st'pvEIttANCc rEr1NITIONS A.
Channel check A qualitative determination of acceptable operability by observation of channel behavior during operation.
This determination shall include, where possible, comparison of the channea with other independent channels measuring the same variable.
B.
Channel Test Injection of a simulated signal into the channel to verify its proper response including, where applicable, alarm and/cr trip initiating action.
C.
Channel Calibration Adjustment of channel output such that it responds, with acceptable range and accuracy, to known values of the parameter which the channel measures.
Calibration chall encompass the entire channel, including equipment actuation, alarm or trip.
D.
Source Check A SOURCE CHECK is the qualitative assessment of channel response when the channel senser is exposed to a source of radioactivity.
1.20 EplhB Oyster Creek Unit No. 1 racility Description and Safety Analysis Report as amended by revised pages and figure changes contained in Amendments 14, 31 and 45.*
1.21 COPE ALTERA?!ON A core alteration is the addition, removal, relocation or other manual movement of fuel or controls in the reactor core.
Control rod movement with the control rod drive hydraulic system is not defined as a core alteration.
1.22 EFITICAL POWER RATIO The critical power ratio is the ratio of that power in a fuel assembly which is calculated, by application of an NRC approved CPR ccrrelation, to cause some point in that assembly to experience boiling transition divided by the actual assembly operating power.
1.23 STAGGERED TEST BASIS A Staggered Test Basis shall consist oft A.
A test schedule for n systems, subsystems, trains or other designated components obtained by dividing the specified t**t interval into n equal subintervals.
'Per Errata dtd. 4-9-69 OYSTER CREEK 1.0-4 Amendment No. 14. 108, 147 TSCR-180
_ ~ _ _. -. _ _ _ _ _ _.... _ _
i 1.37 PURGE PURGE OR PUD 0!NO is the controlled process of discharging air or gas i
from.a confinement and replacing it with air or gas.
1.38 EXCLUSION AREA L
EXCLUs!ON AREA is defined in 10 CFR part 100.3(2). As used in these technical specification, the Exclusion Area boundary is the perimeter line around the OCNGS beyond which the land is neither owned, leased, nor otherwise subject to control by CPU (ref. ODCM Figure 1-1). The area outside the Exclusion Area is ternied OFFSITE.
1.39 EEACTOR VESSEL PRESSURE TESTING system pressure testing required by AsME Code section XI, Article IWA-5000, including system leakage and hydrostatic test, with reactor vessel completely water solid, core not critical and section 3.2.A satisfied.
1.40 SUBSTANTIVE CHANGES SUBSTANTIVE CRANGES are those which affect the activities associated with a document or the document's meaning or intent.
Examples of non-substantive changes ares (1) correcting spelling, (2) adding (but not
- deleting) sign-off spaces, (3) blocking in notes, cautions, etc., (4) changes in corporate and personnel titles which do not reassign responsibilities and which are not' referenced in the Appendix A Technical specifications, and (5) changes in nomenclature or editorial changes which clearly do not change function, meaning or intent.
1.41 DOSE EOUIVALENT I-131 Dose EQUIVALENT I-131 shall be that concentration of I-131 microcuries per gram which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-231, I-132, 2-133, I-134, and I-135 actually
- present. The thyroid dose conversion factors used for this calculation shall be those listed in Tablu E-7 or Regulatory Guide 1.109, ' Calculation of Annual Doses to Man from Routine Releases of Reactor Ef fluences for the Purpose of
- Evaluating Compliance with 10 CFR Part 40 Appendix I".
1.42 AVERACE PLANAR LINEAR HEAT GENERATION RATE
~The-AVERAGE PLANAR LINEAR HEAT CENERATION RATE (APLHCR) shall be applicable to a specific planar height and is equal to the sum of the-heat
~
generation rate per unit length of fuel rod for all the fuel rods in the
- specified bundle at the specified height divided by the number of fuel rods in the fuel bundle at that height.
OYSTER CREEK 1.0-7 Amendment No.: 108, 120, 125, 126, 138, 147 vo ca.
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a.43 CORE OPrRAT2NG (TM7TS REPORT The Oyster Creek CORE OPERATING LIMITS REPORT (COLR) is the document that provides core operating limits for the current operating reload cycle.
These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.f.
Plant operatAon within these operating limits is addressed in individual specifications.
3.44 LOCAL LINEAR MEAT CENERATION PL?g The LOCAL LINEAR HEAT GENERATION RATE (LLHGR) shall be applicable to a specific planar height and is equal to the AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHCR) at the specified height multiplied by the local peaking factor at that height.
OYSTER CREEK 1.0-0 Amendment No.: 147
r_____
d 3.10 cop.E LIMITS Aeolleability: Applies to core conditions required to meet the Final Acceptance Criteria for Emergency Core Cooling ?arformaneo.
Obieetives-To assure conformance to the peak clad temperature limitations during a postulated loss-of-coolant accident as specified in 10 CTR 50.46 (January 4, 1974) and to assure conformance to the operating limits for local linear heat generation rate and i
minimum critical power ratio.
Soeeification:
A.
Average Planar LNGR During power operation the maximum AVERAGE PLANAR LINEAR HEAT CENERATION RATE (APLHGR) for each fuel type as a function of exposure shall not exceed the limits specified in the CORE OPERATING LIMITS REPORT (COLR).
If at any time during power operation it is determined by normal surveillance that the limiting value for APLHGR ik being exceeded, action shall be initiated to restore operation to j
within the prescribed limits.
If thh APLHGR is not returned to i
within the prescribed limits within two (2) hours, action shall be initiated to bring the reactor to the cold shutdown condition within J6 hours. During this period surveillance and' l
. corresponding action shall continue until reactor operetion is j
within the prescribed limits at which time power operation may be continued.
B.
Local LHGR During power operation, the Local LINEAR HEAT GENERATION RATE (LHGR) of any rod in.any fuel assembly, at any axial location shall not exceed the maximum allowable LHCR limits specified in-
-the COLR.
If at any time during operation it,is determined by normal-i surveillance that the limiting value of LHGR is being exceeded,
{
action shall be inJtiated to restore operation to within the prescribed limits.
If the LHGR is not returned to within the prescribed limits within two (2) houre, action shall be initiated to bring the reactor to the cold shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. During this period, surveillance and corresponding action shall continue until reactor operation is within the prescribed limits at which time power operation may be continued.
1 OYSTER CREEK ~
3.10-1 Amendment No. 48, 75.-129, 147 L
TSCR 180 E
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Minimum Critical Pow >r Ratio (MCPR)
During steady state power operation the MINIMUM CRITICAL POWER RATIO (MCPR) shall be equal to or greater than the MCPR linit as specified in the COLR.
When APRM status changes due to instrument failure (AFEM or LPRM input failure), the MCPR requirement for the degraded condition shall be met within a time interval of eight (B) hours, provided that the control rod block is placed in operation during this interval.
For core flows other than rated, the nominal value fer MCFR shall be increased by a factor of k, where kg is specified g
in the COLR.
If at any time during power operation it is determined by normal surveillance that the limiting value for HCPR is being exceeded for reasons other than instrument failure, action shall be initiated to restore operation to within the prescribed limits.
If the steady state MCPR is not returned to within the prescribed limits within two (2) hours, action shall be initiated to bring the reactor to the cold shutdown conditier. within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
During this period, surveillance and ectresponding action shall continue until reactor operation is within the prescribed limit at which time power operation may be continued.
Barest The Specification for average planar LHGR assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the 2200'T limit s pec4 fied in 10 CTR 50.4 0.
The analytical methods and assunptions used in evaluating the fuel design limits are presented in TSAR Chapter 4.
LOCA analyses are performed for each fuel design at selected exposure points to determined APLHCR limits that meet the PCT and maximum oxidation limits of 10 CTR 50.46.
The analysis is performed using GE celcul0tional models which are consistent with the requirements of 3 0 CTR $0, Appendix K.
The PCT following a postulated LOCA is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is not strongly influenced by the rod to rod power distribution within an assembly.
Since expected location variations in power distribution within a fuel assembly af fect the calculated peak clad temperature by less than 1 20'T relative to the peak temperature for a typical fuel design, the limit on OYSTER CREEK 3.10-2 Amendment No.: 48, 75, 111. 129. 14
i
'th) avereg) linJar h st 93n ratien rate is suf ficient to assuro that calculated temp 3ratures are b>10w the limit s specified in 10 CTR 60.46.
The maximum everage planar LHGR limits for the various f uel types currently being used are specified in the COLR. The MAPLHCR limits f or both five-loop and f our-loop operation
[
with the idle loep unisolated are shown.
Four-loop operation with the idle loop isolated (suction,' discharge and discharge bypass valves closof t requires that a KAPLHCR multiplier of 0.98 be applied to all fuel types.
Additional requirenents for isolat ed loop operation are given in specification 3.3.r.2.
Fuel design evaluations are ;*rf ormed te demonstrate that the cladding in plastle strain and other fuel design limits are not excocded during anticipated operational occurrences for operation with LHGRs up to the operating limit LHGR.
The analytical methods and assumptions used in evaluating the anticipated operational occurrences to establish the.
operating limit HCPR are presented in the FSAR, Chapters 4, l
- 6 and 16 and in Technical specification 6.9.1.f.
To assure that_the safety Limit HCPR is not exceeded during any moderate frequency transient event, limiting transients have been analyzed to determine the largest reduction in Critical Fower Ratio (CPR).
The types of transients evaluated are pressurisation, positive reactivity insertion and coolant temperature decrease. The operational MCPR limit is selected to provide margin to accommodate transients and uncertal.nties in monitorirg the core operating state, manuineturing, and in the critical power correlation itself. This limit is derived by addition of a
the CPR for the most limiting transient to the safety limit MCPR designated in Specification 2.1.
The APRM response :ls used to predict when the rod block occurs in the analysis of the rod withdrawal error transient.
The transient rod position at the rod block and corresponding MCPR c'an be determined. The MCPR has been evaluated for different APRM responses which would result from changes in the APRM status as a consequence of bypassed APRH channel and/or failed / bypassed LPRH inputs.
The steady state HCPk required to protect the minimum transient CPR for the worst case APRM status condition (APRM Status 1) is determined in the rod withdrawal error transient-analysis. The steady state HCPR values for APRM status conditions le 2, and 3 will be evaluated each cycle.
For those cycles where the rod withdrawal error transient is not the most severe transient the MCPR Value for APRH status conditions 1, 2, and 3 will be the same and be equal to the limiting transient MCPR value.
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I OYSTER CREEK 3.10-3 Amendment No. 75, 129, 147 f
Th3 time int 0rval of Eight (8) houro to cdjust tha stoody stCt0 Cf HCPR to cec unt for a d:grOdttien in th3 APRM
!I stctus 10 justificd en th3 b2sio of instituting o control j i rod block which precludes the possibility of experiencing a i
rod withdrawal error transient since rod withdrawal is physically prevented. This time interval is cdequate to allow the operator to either increase the HCPR to the appropriate value or to upgrade the status of the APRM system while in a condition which prevents the possibility of this transient occurring.
Transients analyzed each fuel cycle will be evaluated with respect to the operational MCPR limit specified in the COLR.
The purpose of the kg factor is to define operating limits at other than rated flow conditions.
At less than 100% flow the required MCPR is the product of the operating limit HCPR and the k factor.
Specifically, the k g
g f actor provides the required thermal margin to protect against a flow increase transient.
The k factor curves, as shown in the COLR, were g
developed generically using the flow control line corresponding to rated thermal power at rated core flow.
For the manual flow control mode, the k factors were g
calculated sten that at the maximum flow state (as limited oy the punp scoop tube set point) and the corresponding core power (along the rated flow control line), the limiting bundle's relative power was adjusted until the MCPR was slightly above the safety Limit.
Using this relative bundle power, the MCPR's were calculated at different points along the rated flow control line corresponding to different core flows.
The ratio of the McPR calculated at a given point of core flow, divided by the operating limit MCPR determines the value of k.
g OYSTER CREEK 3.'4-4 Amendment No.: 75, 129, 140, 147 TSCR-180 l
(4) o cummary of metcorological dsto calicetcd during tho yocr j
ch:11 be includ:d in tho report cubmitted within 60 d yo efter i
January 1 of each year.
Alternatively, summary meteocological data may be retained by GPU Nuclear and made available to the
.7C upon request.
e.
Annual Radiolecit al Environment al Report A report of radiological environmental surveillance activities during each year shall be i
submitted before May 1 of the following year.
Each report shall include the following information required in Specification 4.16 for radiological environmental surveillances (1) a summary description of the radiological environmental monitoring program, (2) a map and a table of distances and directions (compass azimuth) of locations of sampling stations from the reactor, (3) results of analyses of samples and of radiation measurements, (In the event some results are not available, the reasons shall be explained in the report.
In the event the missing results are obtained, they shall be reported to the NRC as soon as is reasonable.)
(4) deviation (s) from the environmental sampling schedule in Table 4.16.1.
(5) identification of environmental samples analyzed when instrumentation was not capable of meeting detection capability in Table 4.16.2.
(6) a summary of the results of the land use survey.
(7) a summary of Lhe results of licensee participation in an NRC approved inter-laboratory crosscheck program for environmental samples.
(8) results of dose evaluations to demonstrate compliance with 40 CFR Part 190.10a.
f.
CORE OPERATING LIMITS REPORT (COLR) 1.
Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle for the followings
- a. The AVERACE PLA:iAR LINEAR HEAT GENERATION RATE ( APLHGR) for Specification 3.10.A
- b. The Kg core flow adjustment factor for Specification 3.10.C.
- c. The HINIMUM CRITICAL POWER RATIO (HCPR) for Specification 3.10.C OYSTER CREEK 6-12 Amendment No.: fr% J&ff, %e 147
- d. The LOCAL L3NEAR HEAT GENERAT3ON RATE (LLHCR) for SF*eification 3.10. B.
and shall be documented in the COLR.
2.
The analytical methods used to determine the core operating limits shall be those previously reviewed and approyed by the NRC, specifically those described in the ic110 wing documents,
- a. GPU Nuclear (CPUN) Topleal Report (TR) 020, Methods for the Analysis of Boiling Water Rosetors Lattice Physics, (The approved revision at the time reload analyses are performed shall be identified in ths COLR.)
- b. CPUN TR 021, Methods for the Analysis of Boiling Water Reactors Steady State Physics, (The approved revision at the time reload analyses are performed shall be identified in the COLR.)
- c. CPUN TR 033, Methods for the Generation of Core Kinetics Data for RETRAN-02, (The approved revision at the time reload analyses are performed shall be identified in the COLR.)
- d. CPUN TR 040, steady-state and Quasi-Steady-State Methods Used in the Analysis of Accidents and Transients, (The approved revision at the time reload analyses are performed shall be identified in the COLR.)
- e. CPUN TR 045, BWR-2 Transient Analysis Model Using the Rctran code, (The approved revision at the time reload analyses are performed shall be identified in the COLR. )
- f. NEDE-31462P and NEDE-31462, Oyster Creek Nuclear Generating station SAFER /CORECOOL/CESTR-LOCA Loss 4of-Coolant Accident Analysis, (The approved revision at the time reload analyses are perf ormed shall be io6ntified in the COLR. )
- g. NE0E-24011, General Electric standard App 11ection for Reactor ruel, (The approved revision at the time reload analyses are performed shall be identified in the COLR.)
- h. NEDE 24195, General Electric Reload ruel Application for Oyster Creek, (The approved revision at the time reload analyses are performed shall be identified in the COLR.)
- 1. XN-75-55-(A); XN-75-55, supplement 1-(A); XN-75-55, supplement 2-(A), Revision 2, " Exxon Nuclear Company WREM-Based NJP-BWR ECCS Evaluation Model and Application to the oyster Creek Plant,' April 1977 OYSTER CREEK 6-12a Amendment No.147 l
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- j. XM-76 36(NP)*(A); KN-76 36(NP), supplenont 1-(A),
spray Cooling H3at TransfCr Phas) Test R3sults, ENC - 8x& BWR Tuc1 60 and 63 Active Rods, Interim Report,' October 1976 4
3.
The core operating Ilmits shall be determined such that all i
applicable limits (e.g., f uel thermal-mechanical l'imits, core thermal-hydraulle limits, ECCS limits, nuclear 1&n4ts such as shutdown margin, transient analysis limits, and accident analysis limits) of the safety annlysis are met.
4.
The CORT OPERATING LIMITS k! PORT, including any mid*eyele revisions or supplements shall be provided, upon issuance for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident 2nspector.
Basis 6.9.1.e An annual report of radiological environmental surveillance activities includes factual data summarising results of activities required by the surveillance program.
In order to aid interpretation of the data, CPUN may choose to submit analysis of trends and comparative non regional radiological environmental data.
In addition, the licensee may choose to discuss previous radiological environmental data as well as the observed radiological environmental impacts of station operation (if any) on the environment.
f 6.9.2 REPORTAstr tyrNTS The submittal of Licensee Event Reports shall be accomplished in accordance with the requirements set forth in 10 CPR 60.73.
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OYSTER CREEK 6-12b Amendment No.: 84, 100, 134, 147
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