ML20073J480
| ML20073J480 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 04/24/1991 |
| From: | Butler W Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20073J486 | List: |
| References | |
| NUDOCS 9105080209 | |
| Download: ML20073J480 (11) | |
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7, UNITED STATES
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i NUCLEAR REGULATORY COMMISSION WASHINGTON, D C. 20666 j
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PUBLIC SERVICE ELECTRIC & GAS COMPANY l
ATLANTIC CITY ELECTRIC COMPANY Y
DOCKET NO. 50-354 l
H0oE CPEEK GENERATING STATION i
AMEllDMENT TO FACILITY OPERATING LICENSE Amendment No. 42 License No. NPF-57 1.
The Nuclear Regulatory Commission (the Comission or the NRC) has found 4
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that:
i A.
The application for amendment filed by the Public Service Electric &
Gas Company (PSE&G) dated December 28, 1990 complies with the-i standards and requirements of-the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set j
forth in 10 CFR Chapter I;
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C.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; i
l 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 j
Chapter I; j
D.
The issuance of this amendment will not be inimical to the comon i
defens 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 Comission's regulations and all applicable-requirements have been i
satisfied.
2.
Accordingly, the license is amended by changes.to the Technical Specifica-tions as indicated in the attachment to this license amendment, and-paragraph 2.C.(2) of Facility-Operating License No. NPF-57 is-hereby, i
amended to read as follows:
(2) Technical Specifications and-Environmental Protection Plan l
The Technical Specifications contained in Appendix A,:as revised through Amendment No. 42, and the Environmental Protection Plan contained in-Appendix B, are-hereby incorporated in the. license.
PSE&G shall operate the-facility in accordance with the Technical Specifications and the L
Environmental Protection Plan.
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9105080209 910424 PDR ADOCK 05000354
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This license amendment is effective as of its date of issuance and shall be implemented within 60 days of its date of issuance.
FOP Tile NUCLEAR REGULATORY COMMISSION
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Walter R. Butler, Director Project Directorate I-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
April 24,1991
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ATTACHMENT TO LICENSE AMENDMENT NO. 42 FACILITY OPERATING LICENSE NO. NPF-57 DOCKET NO. 50-354 Replace the following pages of the Appendix "A" Technical Specifications with the attached pages.
The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
Overleaf pages provided to maintain document completeness.*
Remove Insert 1-1 1-1*
1-2 1-2 B 2-1 B 2-1 B 2-2 B 2-2 B 2-3 B 2-3*
B 2-4 B 2-4 B 3/4 2-3 B 3/4 2-3 8 3/4 2-4 8 3/4-2-4 5
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- 1. 0 DEFINITIONS i
l The following terms are defined so that unMore interpretation of these specifications may be achieved.
and shall be applicable throughout these Technical Specifications.The def1h ACTION 1.1 ACTION shall be that part of a Specification which prescribes remedial measures required under designated conditions.
AVERAGE PLANAR-EXPOSURE
- 1. 2 The AVTRAGE PLANAR EXPOSURE shall be applicable to a specific pisnar height and is equal to the sus of the exposure of all the fuel rods in the spe fled bundle at the specified height divided by the number of fuel
" ?.he fuel bundle.
rod AVERAGE ^ C4AR LINEAR HEAT GENERATION RATE
- 1. 3 The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) shall be applicabl to a specific planar height and is equal to the sum of the LINEAR HEAT GENERATION RATES for all the fuel rods in the specified bundle at the:
specified height divided by the number of fuel rods in the fuel bundle.
CHANNEL CALIBRATION
- 1. 4 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors.
The CHANNEL CALIBRATION shall encompass the entire channel includi the sensor and alare and/or trip functions, and shall include the L FUNCTIONAL TEST.
The CHANNEL CALIBRATION may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is calibrated.
CHANNEL CHECK i
- 1. 5 A CHANNEL CHECK shall be the-qualitative assessment of channel behavior during operation by observation.
This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independsat instrument channels measuring the sank parameter.
CHANNEL F W IONAL TEST
- 1. 6 A CHA15EL. FUNCTIONAL TEST shall be:
$nalog channels - the injection of a simulated signal into the channel a.
as close to the sensor as practicable to verify 0PERASILITY including alare and/or trip functions and channel failure trips.
b.
Bistable channels - the injection of a simulated signal into the sensor to verify OPERABILITY including alarm and/or trip = functions.
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The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is tested.
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HOPE CREEK 1-1
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1 DEFINITIONS CORE ALTERATION
- 1. 7 CORE ALTERATION shall be the addition, removal, relocation or movement of fuel, sources, incore instruments or reactivity controls within the reactor pressure vessel with the vessel head renioved and fuel in the vessel.
Normal movement of the SRMs, IRMs TIPS, or special movable detectors is not considered a CORE ALTERATION.
Suspension of CORE ALTERATIONS shall not preclude completion of the movement of a component to a safe conservative position.
CORE MAXIMUM FRACTION OF LIMITING POWER DENSITY 1.8 The CORE MAXIMUM FRACTION OF LIMITING POWER DENSITY (CHFLPD) shall be highest value of the FLPD which exists in the core.
CORE OPERATING LIMITS REPORT
- 1. 9 The CORE OPERATING LIMITS REPORT is the unit-specific 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 6.9.1.9.
Plant operation within these limits is addressed in individual specifications.
CRITICAL POWER RATIO 1.10 The CRITICAL POWER RATIO (CPR) shall be the ratio of that power in the assembly which is calculated by application of the applicable NRC-approved critical power correlation to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.
DOSE EQUIVALENT I-131 1.11 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-131, I-132, I-133, I-134, and I-135 actually present.
The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites."
l E-AVERAGE DISINTEGRATION ENERGY 1.12 E shall be the average, weighted in proportion to the concentration of j
each radionuclide in the reactor coolant at the time of sampling, of the J
sum of the average beta and gamma energies per disintegration, in MeV, for isotopes, with half lives greater than 15 minutes, making up at least i
l 95% of the total non-iodine activity in the coolant.
l EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME 1.13 The EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ECCS actuation set-point at the channel sensor until the ECCS equipment-is capable of performing its safety function, i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.
Times shall include diesel generator starting and sequence loading delays where applicable.
The response time may be measured by any series of sequential, l
overlapping or total steps such that the entire response time is measured.
I MOPE CREEK 1-2 Amendment No. 42 l
2.1 SAFETY LIMITS BASES
- 2. 0 INTRODUCTION The fuel cladding, reactor pressure vessel and primary system piping are the principal barriers to the release of radioactive materials to the environs.
Safety Limits are established to protect the integrity of these barriers during normal plant operations and anticipated transients. The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated.
Because fuel damage is not directly observable, a step-back approach is used to establish a Safety Limit such that the MCPR is not less than 1.07 for two recirculation loop operation and 1.08 for single recirculation loop operation.
MCPR greater than 1.07 for two re-circulation loop operation and 1.08 for single recirculation loop operation represents a conservative margin relative to the conditions required to maintain fuel cladding integrity.
The fuel cladding is one of the physical barriers which separate the radioactive materials from the environs.
The integrity of this cladding barrier is related to its relative freedom from perforations or cracking.
Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incre-mentally cumulative and continuously measurable.
Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation significantly above design conditions and the Limiting Safety System Settings.
While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration.
Therefore, the fuel cladding Safety Limit is defined with a margin to the conditions which would produce onset of transition boiling, MCPR of 1.0.
These conditions represent a signi-ficant departure from the condition intended by design for planned operation.
2.1.1 THERMAL POWER, Low Pressure or Low Flow The use of the applicable NRC-approved critical power correlation is not valid for all critical power calculations performed at reduced pressures below 785 psig or core flows less than 10% of rated flow.
Therefore, the fuel cladding integrity Safety Limit is established by other means.
This is done by estab-lishing a limiting condition on core THERMAL POWER with the following basis.
Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flows will always be greater than 4.5 psi.
Analyses show that with a bundle flow of 28 x 103 lbs/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi.
- Thus, the bundle flow with a 4.5 psi driving head will be greater than 28 x 103 lbs/hr.
Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt.
With the design peaking factors, this corresponds to a THERMAL POWER of more than 50% of RATED THERMAL POWER.
Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative.
l HOPE CREEK B 2-1 Amendment No. 42 l
-4 SAFETY LIMITS BASES 2.1.2 THERMAL POWER, High Pressure and High Flow The fuel cladding integrity Safety Limit is set such that no fuel damage.
is calculated to occur if the limit is not violated.
Since the parameters which result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions resulting in a departure from nucleate boiling have been used to mark the beginning of the region where fuel damage could occur.
Although it is recognized that a departure from nucleate boiling would not necessarily result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit.
However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical power.
Therefore, the fuel cladding integrity Safety Limit is defined as the CPR in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transition considering the power distribution within the core and all uncertainties, Tne Safety Limit MCPR is determined using a statistical model that combines j
all of the uncertainties in operating parameters and in the procedures used to calculate critical power.
Calculation of the Safety Limit MCPR is defined in
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Reference 1.
The requi ed inputs to the statistical model are the uncertalaties listed in Bases Table 62.1.2-1.
Reference:
1.
General Electric Standard Application for Reactor Fuel, NEDE-24011-P-A (The approved revision at the time the reload analyses are performed.
The approved revision number shall.be identified in the CORE OPERATING.
LIMITS REPORT.)
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HOPE CREEK B 2-2 Amendment No. 42 l
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Bases Table 82.1.2-1 UNCERTAINTIES USED IN THE DETERMINATION OF THE FUEL CLA00!NG SAFETY LIMIT
- Standard Deviation Quantity
(% of Point)
Feedwater Flow
.1.76 Feedwater Temperature 0.76 Reactor Pressure 0.5 Core Inlet Temperature 0.2 Core Total Flow Two Recirculation Loop Operation 2.5 Single Recirculation Loop Operation 6.0 Channel Flow Area 3.0 Friction Factor Multiplier 10.0 Channel friction Factor Muttiplier 5.0 TIP 8.eadings Two Recirculation Loop Operation 8.7 Single Recirculation Loop Operation 9.1 l
l R Factor
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Critical Power 3.6 i
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- The uncertainty analysis used to establish the core wide Safety Limit MCPR is based on the assumption of quadrant power symmetry for the reactor core. The values herein apply to both two recirculation loep operation anri single recirculation loop operation, except-as noted.
HOPE CREEK B 2-3 Amendment No.15 NE 15 logg
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HOPE CREEK 8 2-4 Amendment No. 42
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POWERDISTRIBUTIONgMJJTS BASES L
3/4.2.3 MINIMUM CR1TICAL POWER RATIO The required operating limit MCPRs at steady state operating conditions as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safety Limit MCPR, and an analysis of abnormal operational transients.
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 trcnsient assuming instrument trip setting given in Specification 2.2.
To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting tran-sients have been analyzed to determine which result in the largest reduction in CRITICAL POWER RATIO (CPR).
The type of transients evaluated were less of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease.
The limiting transient yields the largest delta MCPR.
When added to the Safety Limit MCPR, the required minimum operating limit MCPR of Specification 3.2.3 is obtained.
The evaluation of a given transient begins with the system initial para-meters shown in FSAR Table 15.0-3 that are input to a GE-core dynamic behavior transient computer program.
The codes used to evaluate transients are discussed in Reference 2.
The purpose of the K, factor specified in the CORE OPERATING LIMITS REPORT is to define operating lihits at other than rated core flow conditions.
At less than 100% of rated flow the reouired MCPR is the product of the operating limit MCPR and the K, during a flow,ncrease transient resulting from a m factor.
The K factors assure-that-the Safety Limit MCPR will not be violated i
speed control failure.
The K factors may be applied to both manual and auto-f matic flow control modes l
The K, factors values specified in the CORE-0PERATING LIMITS REPORT were developed benerically and are applicable to all BWR/2, BWR/3 and BWR/4 reactors.
The K, factors were derived using the flow control line corresponding to RATED THERML POWER at rated core flow.
The K, factors are determined in the following manner:
The change in CPR is deterhined as a function of core flow along the rated power flow control line.
Then, for a given scoop tube setpoint in the manual flow control operating mode, the MCPR at reduced flow is established that would give the Safety Limit MCPR if the care flow was increased to the scoop tube setpoint.
The ratio of the MCPR at reduced flow to the operating limit MCPR is the K factor at that reduced flow.
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HOPE CREEK B 3/4 2-3 Amendment No. 4?
POWER DISTRIBUTION LIMITS BASES MINIMUM CRITICAL POWER RATIO (Continued)
For operation in the automatic flow r.ontrol mode, the same procedure is employed except the MCPR at low flow is established such that the MCPR is equal to the operating limit MCPR at RATED THERMAL POWER and rated core flow.
The K factors specified in the CORE OPERATING LIMITS REPORT are conserva-f tive because the operating limit MCPRs of Specification 3.2.3 are equal to or greater than the original 1.20 operating limit MCPR used for the generic deriva-tion of K.
f At THERMAL POWER levels less than or equal to 25% of RATED THERMAL POWER, the reactor will be operating at minimum recirculation pump-speed and the modera-tor void content will be very small.
For all designated control rod patterns which may be employed at this point, operating plant experience indicates that the resulting MCPR value is in excess of requirements by a considerable margin.
During initial start-up testing of the plant, a MCPR evaluation will be made at 25% of RATED THERMAL POWER level with minimum recirculation pump speed.
The MCPR margin will thus be demonstrated such that future MCPR evaluation below this 4
power level will be shown to be unnecessary.
The daily requirement for calculat-ing MCPR when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes.
The requirement for calculat-1 ing MCPR when a limiting control rod pattern is approached ensures that MCPR will be known following a change in THERMAL POWER or power shape, regardless of magnitude, that could place operation at a thermal limit.
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3/4.2.4 LINEAR HEAT GENERATION RATE i
This specification assures that the Linear Heat Generation Rate:(LHGR) in j
any rod is less than the design linear heat generation even if fuel pellet F
densification is postulated.
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References:
1.
General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K, NEDE-20566, November 1975.
2.
General Electric Standard Application for Reactor Fuel, NEDE-24011-P-A (The approved revision at the time the reload analyses are performed, The approved revision number shall be identified in the CORE OPERATING i
LIMITS PEPORT.)
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1 HOPE CREEK 8 3/4 2-4 Amendment No. 42 c.-.,
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