ML20055C514
| ML20055C514 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 05/09/1990 |
| From: | Hebdon F Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20055C515 | List: |
| References | |
| GL-88-16, NUDOCS 9005240247 | |
| Download: ML20055C514 (47) | |
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GULF STATES UTILITIES COMPANY D,0CKE,T NO._50-458 0CK RIVER BEND STATION, UNIT 1 i
AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 42 i
License No. HPF-47 l
1.
The. Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Gulf States Utilities Company (the licensee) dated November 17,1989,. as modified March 30, and 1
April 16, 1990, complies with the standards and requirements of the-Atomic Energy Act of 1954, as amended (the Act), and the Comission's i
rules and regulations set forth in 10 CFR Chapter I; j
1 B.
The facility will operate in conformity with the application, as l
amended, the provisions of the Act, and the rules and regulations i
of the Comission; C.
There is reasonable assurance:
(1) 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 Comission's regulations; D.
The issuance of this license amendment will not-be inimical to the comon 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 Comission's regulations and all applicable requirements have been satisfied.
9005240247'h$$$9 458 PDR ADOCK PDC P
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.i 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 Facility Operating License No. NPF-47 is hereby amended to read as follows:
(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained-in Appendix A, as revised through Amendment No. 42 and the Environmental Protection Plan contained in Appendix D, are hereby incorporated in the license. GSU shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
The license amendment is effective 60 days from its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Frederick J. Het/ don, Director Project Directorate IV Division of Reactor Projects - III, IV, Y and Special Projects Office of Nuclear Reactor Regulation
Attachment:
Charges to the Technical r
Specifications Date of Issuance:
May 9, 1990 l
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ATTACHMENT TO LICENSE AMENDMENT NO. 42 FACILQYOPgATINGLICENSENO.'NPF-47 A
DOCKET-NO. 50-458-Replace the following pages of the Appendix "A"' Technical Specifications with the enclosed pages. The revised pages are-identified by Amendment number ~and contain vertical lines indicating the area of change. The overleaf-pages-are provided to maintain docunent completeness.-
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REMOVE PAGES INSERT PAGES i
i ii 11 xxii xxii 1-2~
1-2 B 2-7 B 2-7 3/4 2-1 3/4 2-1 3/4 2 -
3/4 2-3 3/4 2-4 3/4'2-5 3/4 2-6 3/4 2-6A
'3/4 2-6B 3/4'2-6C 3/4 2-8 3/4 2-8L 3/4 2-9 3/4 2.-10 3/4 2-11 3/4 2-11 3/4 3-6 3/4 3-6 3/4 3-9 3/4 3-9 B 3/4'2-1 B 3/4 2-1 B 3/4 2-2 B 3/4 2-2 l
B 3/4 2-4 8 3/4 2-4 5-5 5-5 L-6-19 6-19 l.
[.
e s
t DEFINITIONS SECTION F
'1.0 DEFINITIONS PAGE 1.1-ACTI0N.........................................................
.1-l '
1.2 AVERAGE PLANAR. EXPOSURE........
1-1 1.3 AVERAGE PLANAR LINEAR HEAT GENERATION RATE.....................
1-1 1.4 CHANNEL CALIBRATION............................................
1-1
^t e
'1. 5 CHANNEL CHECK..................................................
1-1
'i 1.6. CHANNEL' FUNCTIONAL TEST........................................
1-1 1.7' CORE ALTERATION................................................
1-2 i
1.8 CORE MAXIMUM FRACTION OF LIMITING POWER DENSITY................
1-2 1.9 CORE OPERATING LIMITS REPORT (C0LR)............................
1-2 l
1.10 CRITICAL POWER RAT10...........................................
1-2 1.11 DOSE EQUIVALENT I-131..........................................
1,.
1.12.DRYWELL INTEGRITY..............................................
1-2 11.13lE-AVERAGE DISINTEGRATION ENERGY................................
1 ;
1.14 EMERGENCY.00RE'C00 LING SYSTEM (ECCS) RESPONSE. TIME.............
1-3 e
11.15 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME......
1-3 4
- 1.16 FRACTION OF LIMITING POWER DENSITY.............................
1-3 1.17 FRACTION OF RATED THERMAL P0WER................................
1-3 y
1.18 FREQUENCY N0TATION.............................................
1-4 l
1.19 GASE0US~RADWASTE TREATMENT (OFFGAS) SYSTEM.....................
1-4
'1.'20 IDENTIFIED LEAKAGE.............................................
1-4 o
1.21' ISOLATION SYSTEM RESPONSE TIME.................................
1-4
'1.22 LIMITING CONTROL R00 PATTERN...................................
1-4 1.23-LINEAR HEAT GENERATION RATE....................................
1-4 t
_1.24 LOGIC SYSTEM FUNCTIONAL TEST...................................
1-4 L
RIVER BEND - UNIT 1 i
Amendment No. 42 E
WI INDEX DEFINITIONS SECTION DEFINITIONS (Continued)
PAGE 1.25 MEMBER (S) 0F THE PUBLIC..........................................-
1-5 1.26 MINIMUM CRITICAL POWER RATI0....................................'.
1-5 1.27 0FFSITE DOSE CALCULATION MANUAL..................................
1-5
.1.28 OPERABLE - OPERABILITY...........................................
1-5 1.29 OPERATIONAL CONDITION - C0NDITION................................
1-5 1.30 PHYSICS TESTS....................................................
1-5 1.31 PRESSURE BOUNDARY LEAKAGE........................................
1-6
- 1.32 PRIMARY CONTAINMENT INTEGRITY - FUEL HANDLING....................
1-6 1.33 PRIMARY CONTAINMENT INTEGRITY - 0PERATING........................
1-6 1.34 PROCESS CONTROL PR0 GRAM..........................................
1 1.35 RATED THERMAL P0WER..............................................
1 1.36 REACTOR PROTECTION SYSTEM RESPONSE TIME..........................
1-7 1.37 REPORTABLE EVENT.................................................
1-7 1.38 ROD DENSITY......................................................
1 1~.39 SECONDARY CONTAINMENT INTEGRITY - FUEL BUILDING..................
1-7 1.40 SECONDARY CONTAINMENT INTEGRITY - 0PERATING......................
1-7 1.41 SHUTDOWN MARGIN..................................................
1-8 1.42 SITE B0VNDARY....................................................
1-8 1.43 SOLIDIFICATION...................................................
1-8 1.44 SOURCE CHECK.....................................................
1-8 1.45 STAGGERED TEST BASIS.............................................
1-9 1.46 THERMAL-P0WER....................................................
1-9 1.47 TURBINE BYPASS SYSTEM RESPONSE TIME..............................
1-9 i
RIVER BEND - UNIT 1 11 Amendment No. 42
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r INDEX~
DEFINITIONS SECTION
.PAGE G
1 DEFINITIONS (Continued)
- 1. 48 UNI D ENTI F I ED LE A KAG E.............................................
1-9
.f 1,49 UNRESTRICTED AREA................................................
1-9 1.50 VENTILATION EXHAUST TREATMENT SYSTEM.............................
1-9 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS THERMAL POWER, Low Pressure or Low F1ow..........................
2-1 THERMAL POWER, High Pressure and High F1ow.......................
2-1 Reactor Coolant System Pressure..................................
2-1 Reactor Vessel Water-Level.......................................
2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Protection System Instrumentation Setpoints..............
2 L BASES L
1 2.1 SAFETY LIMITS i
1 THERMAL POWER, Low Pressure or Low F1ow..........................
B 2-1 THERMAL POWER, High Pressure and High F1ow.......................
B 2-2 Reactor. Coolant System Pressure..................................
B 2-5 Reactor Vessel Water Leve1.......................................
B 2-5.
2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Protection System Instrumentation Setpoints..............
B 2-6 t
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RIVER BEND - UNIT 1 iii Amendment No. 42 1
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s INDEX
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LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE' REQUIREMENT i
SECTION
_ PACE 3/4.0 APPLICABILITY.~............................................
3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1-SHUTDOWN MARGIN.'.......................................
i 3/4 1-1.
3/4.1.2 REACTIVITY AN0MALIES...................................
3/4 1-2 3/4.1.3 CONTROL RODS Control Rod Operability................................
3/4 1-3 J
Control Rod Maximum Scram Insertion Times..............
3/4 1-6 i
Control Rod Scram Accumulators.........................
3/4 1-9 Control Rod Drive Coupling.............................
3/4 1,
Control Rod Position Indication........................
.3/<4 1-13 Control Rod Drive Housing Support.....................
3/4 1-15 i
3/4.1.4 CONTROL R0D PROGRAM CONTROLS Control Rod Withdrawa1..................................
3/4 1-16
' Rod Pattern Control System.......................
3/4 1-17 4
' 3/4.1.5
' STANDBY LIQUID CONTROL SYSTEM..........................
3/4 1-19 i
3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE.............
3/4 2-1 3/4.2.2 APRM SETP0lNTS.........................................
3/4 2-7 3/4.2.3 MINIMUM CRITICAL POWER RATI0...........................
3/4 2-8 3/4.2.4 LINEAR HEAT GENERATION RATE............................
3/4 2-11 RIVER BEND - UNIT 1 iv
g r _7 y.
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]NDEX ADMINISTRATIVE CONTROLS SECi10N i
PAGE 6.13 PROCESS CONTROL PROGRAM (PCP)................................
6 22
~6.14 0FFSITE DOSE CALCULA7!ON MANUAL (0DCM).......................
4-23 6.15 MAJOR CHANGES 70 RAD 10 ACTIVE LIQUID GASEOUS AND $0L10 WASTE TREATMENT SYSTEM5............................................
6-23 t
RIVER BEND - UNIT 1 ni
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INDEX LIST OF' FIGURES FIGURE TITLE P_ AGE AE 3.4.1.1-1 Thermal Power versus Core Flow.....................
3/4 4-3 3.4.6.1-1 Minimum Temperature-Required Versus Reactor Pressure.....-......................................
3/4 4-24 4.7.4-1 Sample' Plan for Snubber Functional Test............
3/4 7-15 B 3/4 2.3-1 Power Flow Operating Map...........................
B 3/4 2-6 B 3/4.3-1
-Reactor Vessel Water Leve1.........................
B 3/4 3-B B 3/4.4.6-1 Fast Neutron Fluence (E>l MeV) at 1/4 T as a Function of Service Life...........................
B 3/4 4-B 5.1.1-1 Exclusion Area.....................................
5-2 5.1.2-1 Low Population Zone................................
5-3 5.1.3-1 Map Defining Unrestricted Areas and Site Boundary for Radioactive Gaseous and Liquid Effluents.......
5-4 l
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RIVER BEND - UNIT 1 xxii Amendment No. Z3, 27,42 l
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- 1. O DEFINITIONS-
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The following tems are defined so that unifom interpretation of these i
specifications may be achieved. The defined teres appear in capitalized i
type and shall be appitcable throughout these Technical Specifications.
ACTION t
1.1 ACTION shall be that part of a specification which prescribes remedial sensures required under designated conditions.
AVERAGE PLANAR EXPOSURE The AVERAGE ' PLANAR EXPOSURE shall be applicable to a specific planar heigh 1.2 and is equal to the sum of the exposure of all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle.
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AVERAGE PLANAR LINEAR HEAT GENERATION RATE 1.3 The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) shall be applicable to a specific planar height and is equal to the sum of the LINEAR HEAT GENERATION i
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 i
1.4 A CHANNEL CALIBRATION shall be the adjustment as necessary, of the channel output such that it responds with the necessary ran,ge and accuracy to known values of.the parameter which the channel monitors.
The CHANNEL CALIBRATION shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST. The CHANNEL CALI-BRATION may be performed by any series of sequential, overlapping or total
)
l channel steps such that the entire channel is calibrated, l
CHANNEL CHECK l
1.5 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where pos-sible, comparison of the channel indication and/or status with other indications
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and/or status delived frosi independent instru.T,ent channels sieasuring the saa o
parameter.
CHANNEL FUNCTIONAL TEST L
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- 1. 6 A CHANNEL FUNCTIONAL TEST shall be:
a.
Analog channels - the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY including alarm 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.
RIVER BEND - UNIT 1 1-1
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i DEFINITIONS The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is tested.
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 removed and fuel in the vessel.
Normal movement of the SRMs. IRMs, LPRMs, TIPS or special movable detectors is not considered a CORE ALTERATION.
Suspension of CORE ALTERATIONS shall not pre-clude 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 (CMFLPD) shall be the i
highest value of the FLPD which exists in the core.
CORE OPERATING LIMITS REPORT
- 1. 9 The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document that provides core operating limits for the current reload cycle.
These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specifications 6.9.3.1, 6.9.3.2, 6.9.3.3 and 6.9.3.4.
Plant operation within these core operating 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 l
assembly which is calculated by application of the GEXL 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 l
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."
DRYWELL INTEGRITY 1.12 DRYWELL INTEGRITY shall exist when:
l All drywell penetrations required to be closed during accident a.
conditions are either:
1.
Capable of being closed by an OPERABLE drywell automatic isola-tion system, or I
RIVER BEND - UNIT 1 1-2 Amendment No. 29,42
6
,, p-DEFINITIONS I
- 2. -
Closed by.at least one manual valve, blind flange, or deactivated automatic valve secured in its closed position, except as pro '
vided in Specification 3.6.4.
b.
All drywell equipment hatches'are closed and sealed, The drywell airlock is in compliance with the requirements of c.
Specification 3.6.2.3.
d.
The drywell leakage rates are within the limits of Specification 3.6.2.2.
The suppression pool is in compliance with the requirements of e.
. Specification 3.6.3.1.
f.
The sealing mechanism associated with each drywell penetration; e.g., welds, bellows or 0-rings, is OPERABLE.
E-AVERAGE DISINTEGRATION ENERGY 1.13 I shall be the average, weighted in proportion to the concentration of-l each radionuclide in the reactor coolant at the time of sampling, of the 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 95% of the total non-iodine activity in the coolant.
EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME 1.14 The EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME shall be that time l
interval from when the monitored parameter exceeds its ECCS actuation setpoint 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, overlapping or l
total steps such that the entire response time is measured.
END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME 1.15 The END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME shall be l'
that time interval to complete suppression of the electric are between the fully open contacts of the recirculation pump circuit breaker from initial movement of the associated:
i a.-
Turbine stop valves, and b.
Turbine control valves.
The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.
FRACTION OF LIMITING POWER DENSITY 1.16 The FRACTION OF LIMITING POWER DENSITY (FLPD) shall be the LHGR existing
(
'at a given location divided by the specified LHGR limit for that bundle type.
FRACTION OF RATED THERMAL POWER 1.17 The FRACTION OF RATED THERMAL POWER (FRTP) shall be the measured THERMAL l
POWER divided by the RATED THERMAL POWER.
I RIVER BEND - UNIT 1 1-3 Amendment No.42 l
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S.
f DEFINITIONS' FREQUENCY NOTATION 1.18 The FREQUENCY NOTATION specified for the performance of Surveillance l
Requirements shall correspond to the intervals defined in Table'1.1.
GASEOUS RADWASTE TREATMENT (OFFGAS) SYSTEM 1.19 The GASEOUS RADWASTE TREATMENT (OFFGAS) SYSTEM 1.:the system designed i
and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.
IDENTIFIED LEAKAGE 1.20 IDENTIFIED LEAKAGE shall be:
l l
Laakage into collection systems, such as pump seal or valve packing a.
i leaks, that is captured and conducted to a sump or collecting tank, or b.
Leakage into the drywell atmosphere from sources that are both specifically located and known either not to interfere with the operation of the leakage detection systems or-not to be PRESSURE BOUNDARY LEAKAGE.
ISOLATION SYSTEM RESPONSE TIME 1.21 The ISOLATION SYSTEM RESPONSE TIME shall be that time interval from when.
l the monitored parameter exceeds its isolation _ actuation setpoint at the channel sensor until the isolation valves travel to their required positions.
Times shall include diesel generator starting and sequence loading delays _where appli-cable.
The response time may'be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.
LIMITING CONTROL ROD PATTERN 1.22 A LIMITING CONTROL ROD PATTERN shall be a pattern which results in the l
core being on a thermal hydraulic limit, i.e., operating on a limiting value for APLHGR, LHGR, or MCPR.
i LINEAR HEAT GENERATION RATE 1.23 LINEAR HEAT GENERATION RATE (LHGR) shall be the heat generation per unit l
length of fuel rod.
It is the integral of the heat flux over the heat transfer area associated with the unit length.
l LOGIC SYSTEM FUNCTIONAL TEST l
1.24 A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all logic components, l
l 1.e., all relays and contacts, all trip units, solid state logic elements, etc, l
RIVER BEND - UNIT 1 1-4 Amendment No. 42
DEFINITIONS of a logic circuit, from sensor through and including the actuated device, to-verify OPERABILITY.
The LOGIC SYSTEM FUNCTIONAL TEST may be performed by any series of sequential, overlapping or total system steps such that the entire l
logic system is tested.
MEMBER (S) 0F THE PUBLIC
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1.25. MEMBER (S) 0F THE PUBLIC shall include all persuns who are not occupa-l l
tionally associated with the plant.
This category does not include employees of the utility, its contractors or vendors.
Also excluded from this category are persons who enter the site to service equipment or to make deliveries.
This category does include persons who use portions of the site for recreational, occupational or other purposes not associated with the plant.
MINIMUM CRITICAL POWER RATIO 1.26 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be the smallest CPR which l
exists in the core.
OFFSITE DOSE CALCULATION MANUAL (0DCM) 1.27 The OFFSITE DOSE CALCULATION MANUAL shall contain the methodology and l
parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents and in the calculation of gaseous and liquid effluent monitoring alarm / trip setpoints.
It shall also contain a table and figure defining current radiological environmental monitoring sample locations.
OPERABLE - OPERABILITY 1.28 A system,' subsystem, train, component or device shall be OPERABLE or I
have OPERABILITY when-it is capable of performing its specified function (s) and when all necessary attendant instrumentation, controls, electrical power, L
cooling or seal water, lubrication or other auxiliary equipment that are l
required for the system, subsystem, train, component or device to perform its function (s) are also capable of performing their related support function (s).
OPERATIONAL CONDITION - CONDITION 1.29 An OPERATIONAL CONDITION, i.e., CONDITION, shall be any one inclusive l
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combination of mode switch position and average reactor coolant temperature as specified in Table 1.2.
PHYSICS TESTS 1.30 PHYSICS TESTS shall be those tests performed to measure the fundamental l
nuclear characteristics of the reactor core and related instrumentation and 1) described in Chapter 14 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission.
L RIVER BEND - UNIT 1 1-5 Amendment No. 42
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? DEFINITIONS 1
-1.31 PRESSURE BOUNDARY LEAKAGE shall be leakage through a non-isolable fault' I
in a reactor coolant' system component body, pipe wall orLvessel wall.
FRIMARY CONTAINMENT INTEGRITY - FUEL HANDLING 1.32 PRIMARY CONTAINMENT INTEGRITY - FUEL HANDLING shall exist when:
l All containment penetrations required to be closed during accident a.
conditions are closed by at least one manual valve, blind flange, j
or deactivated automatic valve secured in its closed position.
J Up to twelve vent and drain line pathways may be opened under-administrative control for the purposes of surveillance testing j
provided the total calculated flow rate through the open vent and
. drain line pathways is less than or equal to 70.2 cfm.
b.
All containment hatches are closed, Each containment air lock is in compliance with the requirements of i
c.
Specification 3.6.1.4.
PRIMARY CONTAINMENT INTEGRITY - OPERATING 1.33 PRIMARY CONTAINMENT INTEGRITY - OPERATING shall exist when:
'l All containment' penetrations required to be closed during accident a.
L conditions are-either:
1.
Capable of being closed by an OPERABLE containment automatic isolation system, or 2.
Closed by at least one manual valve, blind flange, or deacti-vated automatic ~ valve secured in its closed position, except as provided in Specification 3.6.4.
b.
All containment equipment hatches are closed and sealed, Each containment air lock is in compliance with the requirements of c.
L Specification 3.6.1.4.
f 1
d.
The' containment leakage rates are within the limits of Specification 3.6.1.3.
l e.
The suppression pool is in compliance with the requirements of Speci-fication 3.6.3.1.
f.
The sealing mechanism associated with each primary containment penetra-tion; e.g., welds, bellows or 0-rings, is OPERABLE.
PROCESS CONTROL PROGRAM (PCP) 1.34 The. PROCESS CONTROL PROGRAM shall contain the current formula, sampling, l
L analyses, tests, and determinations to be made to ensure that the processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Part 20, 10 CFR Part 61, 10 CFR Part 71 and
'44 1
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RIVER ~ BEND
-UNIT 1 1-6 Amendment No. 35,42
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DEFINITIONS
-Federal.and State regulations and other requirements governing the disposal of the radioactive waste.
i RATED THERMAL POWER 1
1.35 RATED THERMAL POWER shall be a total reactor core heat transfer rate to l
the reactor coolant of 2B94 MWT.
1, REACTOR PROTECTION SYSTEM RESPONSE TIME l-1.36 REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time interval from j
when the monitored parameter exceeds its trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids.
The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.
REPORTABLE EVENT 1.37 -A REPORTABLE EVENT shall be any of those conditions specified in 10 CFR 50.73..l l.
R0D DENSITY 1.38 R0D DENSITY shall be the number of control rod notches inserted as a l
fraction of the total number of control rod notches.
All rods fully inserted l
l-is equivalent to 100%. ROD DENSITY.
SECONDARY CONTAINMENT INTEGRITY - F0EL BUILDING 1.39 SECONDARY CONTAINMENT INTEGRITY - FUEL BUILDING shall exist when:
l-All Fuel Building penetrations required to be closed during acci-a.
dent conditions are closed by valves, blind flanges, or dampers secured in pcsition, b.
All Fuel Building equipment hatch covers are installed.
The Fuel Building Charcoal Filtration System is in compliance with
.c.
the requirements of Specification 3/4.6.5.6.
d.
At least one door in each access to the Fuel Building is closed, except for routine entry and exit of personnel and equipment.
The pressure within the Fuel Building is maintained in compliance e.
with the requirements of Specification 4.6.5.1.a.
SECONDARY CONTAINMENT INTEGRITY - OPERATING 1.40 SECONDARY CONTAINMENT INTEGRITY - OPEi ING shall exist when:
l All Auxiliary Building penetrations, Fuel Building penetrations and a.
Shield Building annulus penetrations required to be closed during accident conditions are either:
RIVER BEND - UNIT 1 1-7 Amendment No. 42
-3 DEFINITIONS' i
1.
Capable of being closed by an OPERABLE secondary containment automatic isolation signal, or.
a 2.
Closed by at least one manual valve, blind flange, or deacti-vated automatic valve or. damper, as applicable, secured in its closed position, except as provided in Specification 3.6.5.2.
b.
All Auxiliary Building, Fuel Building and Shield Building Annulus equipment hatches are closed and sealed.
The Standby Gas Treatment System is in compliance with the require-c.
ments of Specification-3.6.5.4.
d.
The Fuel Building Charcoal Filtration System is in compliance with the requirements of Specification 3.6.5.6.
At'least one door in each access to the Auxiliary Building, Fuel-e.
Building and Shield Building Annulus is closed, except for routine entry and exit of personnel and equipment.
i f.
The sealing mechanism associated with each Auxiliary Building, Fuel Building and Shield Building Annulus penetration, e.g., welds, bellows or 0-rings, is OPERABLE.
g.
The pressure within the secondary containment is less than or equal-to the value required by Specification 4.6.5.1.a.
SHUTDOW; MARGIN 1.41 SbJTDOWN MARGIN shall be the amount of reactivity by which the reactor l
is subtritical or would be subcritical assuming all-control rods are fully inserted except for the single control rod of highest react'ivity worth which is-assunEl to be fully withdrawn and the reactor is in the shutdown condition;
}
cold, i.e. 68'F; and xenon free.
SITE BOUNDARY 1.42.,The SITE BOUNDARY shall be that line beyond which the land is not owned,
~l leased, or otherwise controlled by the licensee.
1 4
SOLIDIFICATION 1.43 SOLIDIFICATION shall be the conversion of wet wastes into a form that l'
meets shipping and burial ground requirements.
SOURCE CHECK 1.44 A SOURCE CHECK shall be the qualitative assessment of channel response l
when the channel sensor is exposed to a source of increased radioactivity.
RIVER BEND - UNIT 1 1-8 Amendment No.42
t - 9;
=*.1 c..
DEFINITIONS STAGGERED TEST BASIS 1.45 A STAGGERED TEST BASIS shall consist of:
A test schedule for n systems, subsystems, trains or other designated a.
components obtained by dividing the specified test interval into equal subintervals.
~b.
The testing of one system, subsystem, train or other designated compo-
[
nent at the beginning of each subinterval.
THERMAL POWER 1.46 _ THERMAL POWER shall be the total reactor core heat transfer rate to the l
TURBINE BYPASS SYSTEM RESPONSE TIME 1,47 The TURBINE BYPASS SYSTEM RESPONSE TIME consists of two components:
l (a) the time from initial movement of the main turbine stop valve or control valve until'80% of turbine bypass capacity is established, and (b) the time from initial movement of the main turbine stop valve or control valve until
-initial movement of.the turbine bypass valve.
The response times may be measured by any series of sequential, overlapping or total steps such that both entire response time components are measured.
.l 1.48 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE.
l UNRESTRICTED AREA 1.49 An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY l
access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials, or any area within the site boundary used for residential quarters or for industrial.
I commercial, institutional, and/or recreational purposes.
VENTILATION EXHAUST TREATMENT SYSTEM 1.50 A VENTILATION EXHAUST TREATMENT SYSTEM _is any system designed and installed l
to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing fodines or particulates from
.the gaseous exhaust stream prior to the release to the environment (such a system is not considered to have any effect on noble gas effluents).
Engineered Eafety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.
(
RIVE.K BEND - UNIT 1 1-9 Amendment No. 42 l
- , a
~
l t
_ TABLE 1.1-
.i SURVEILLANCE' FREQUENCY NOTATION 1
. NOTATION' s
FREQUENCY t
S At least once'per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.:
D At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
W 1
At least'once per 7 days.
M At least once per 31 days.
Q-At least once per 92 days.
SA At least once per 184 days, i
A At least once per 366 days, R
1 At least once per 18 months-(550 days)..
~S/U Prior to each reactor startup.
P Prior to each radioactive release.
t N.A.
Not applicable.
k v
1 i
7 RIVER BEND - UNIT 1 1-10
7 s
LIMITING SAFETY SYSTEM SETTINGS' BASES REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued)'
Average Power Ranae Monitor (Continued) i t
The APRM trip system is ' calibrated using heat balance data taken during steady state conditions.
Fission chambers provide the basic input to the system and.therefore the monitors respond directly and quickly to changes due to transient operation for the case of the Neutron Flux-High setpoint; i.e.,
for a power increase, the THERMAL POWER.of the fuel will be less than.that indicated by the neutron flux due to the time constants of the heat transfer associated with the fuel.. For the Flow Biased Simulated Thermal Power-High setpoint, a time constant is introduced into the flow biased APRM in order to l
simulate the fuel thermal transient characteristics.
A more conservative maximum value is used for the flow biased setpoint as shown in Table 2.2.1-1.
The APRM setpoints were selected to provide adequate margin for the Safety Limits and yet allow operating margin that reduces the possibility of unneces-sary shutdown.
The flow referenced trip setpoint must be adjusted by the specified formula in Specification 3.2.2 in order to maintain these margins when CMFLPD is > to FRTP.
3.
Reactor Vessel Steam Dome Pressure-High High pressure in the nuclear system could cause a rupture to the nuclear system process barrier resulting in the release of fission products. A pressure increase while operating will also tend to increase the power of the reactor by compressing voids thus adding reactivity.
The trip will quickly reduce the neutron flux, counteracting the pressure increase.
The trip setting is slightly' higher than the operating pressure to permit normal operation without spurious trips.
The setting provides for a wide margin to the maximum allowable design pressure and takes into account the location of the pressure measurement compared to the highest pressure that occurs in the system during a transient.
This trip setpoint is effective at low power / flow conditions when the turbine control valve fast closure and turbine stop valve closure trips are bypassed.
For a load rejection or turbine trip under these conditions, the transient analysis indicated an adequate margin to the thermal hydraulic limit.
4.
Reactor Vessel Water Level-Low The reactor vessel water level trip setpoint has been used in transient analyses dealing with coolant inventory decrease.
The scram setting was chosen far enough below the normal operating level to avoid spurious trips but, high enough above the fuel to assure that there is adequate protection for the fuel and pressure limits.
l l
RIVER BEND - UNIT 1 B 2-7 Amendment No. 42
i 1
LIMITING SAFETY SYSTEM SETTINGS BASES REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued) 5, Reactor Vessel Water Level-Hiah A reactor scram from high reactor water level, approximately two feet above j
-normal operating level, is intended to offset the addition of reactivity effect-associated with the introduction of a significant amount of relatively cold feedwater.
An excess of feedwater entering the vessel would be detected by the level increase in a timely manner.
This scram feature is only effective when the reactor mode switch is in the Run position because at THERMAL POWER levels below 10% to 15% of RATED THERMAL POWER, the approximate range of_ power level for changing to the Run position, the safety margins are more than
=i i
adequate without a reactor scram.
6.
Main Steam Line Isolation Valve-Closure The main steam line isolation valve closure trip was provided to limit the amount of fission product release for certain postulated events.
The MSIV's are closed eutomatically'from measured parameters such as high steam flow, high steam line radiation, low reactor water level, high steam tunnel temperature and low steam line pressure.
The MSIV's c)csure scram anticipates the pressure and flux transients which could follow MSIV closure and thereby-protects reactor vessel pressure and fuel thermal / hydraulic Safety Limits.
7.
Main Steam Line Radiation-Hiah The main steam line radiation detectors are provided to detect a gross failure of the fuel cladding. When the high radiation is detected, a trip is j
initiated to reduce the continued failure of fuel cladding.
At the same time the main steam line isolation valves are closed to limit the release of-fission products.
The trip setting is high enough above background radiation levels to prevent spurious trips yet low enough to promptly detect gross failures in q
the fuel cladding.
8.
Drywell Pressure-High High pressure in the drywell could indicate a break in the primary pressure boundary systems or a loss of-drywell cooling. The reactor is tripped in order to minimize the possibility of fuel damage and reduce the amount of energy being added to the coolant and to the primary containment.
The trip setting was j
selected as low as possible without causing spurious trips.
9.
Scram Discharge Volume Water Level-High The scram discharge volume receives the water displaced by the motion of I
the control rod drive pistons during a reactor scram.
Should this volume fill up to a point where there is insufficient volume to accept the displaced water, control rod insertion would be hindered.
The reactor is therefore tripped when
g-- --
3 4
.g.
3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE 1
LIMITING CONDITION FOR OPERATION L
3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for each type of fuel as a function of AVERAGE PLANAR EXPOSURE shall.not exceed the limits specified in the COLR.
The limits specified in the COLR shall be reduced to a j
i i
value of 0.84 times the two recirculation loop operation limit when-in single loop operation.
l J
APPLICABILITY:
OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or
{
equal to 25% of RATED THERMAL POWER.
l 1
ACTION:
With an APLHGR exceeding the limits specified in the COLR, initiate corrective l
i action within 15 minutes and restore APLHGR to within the required limits
(,
within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER-
- within the next'4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
I l
1 l.
SURVEILLANCE REQUIREMENTS 4.2.1 All APLHGRs shall be verified to be equal to or less than the limits-determined specified in the COLR.
l 4
t a.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, b.
Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and i
Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating c.
with a LIMITING CONTROL ROD PATTERN for APLHGR.
d.
The provisions of Specification 4.0.4 are not applicable.
U i
i RIVER BEND - UNIT 1 3/4 2-1 Amendment No. 12, 3I, 33,42 s
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i RIVER BEND - UNIT 1 3/4 2-5 Amendment No. 22,42 l
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RIVER BEND - UNIT 1 3/4 2-6 Amendment No.12,42
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t RIVER BEND - UNIT 1 3/4 2-6a Amendment No. 12,42.
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i Figure 3.2.1-7 has been deleted.
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RIVER BEND - UNIT 1 3/4 2-6b Amendment No. 33,42
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5 Figure 3.2.1-8 has been deleted.
RIVER BEND - UNIT 1 3/4 2-6c Amendment No. 33,42
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O POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION 3.2.3 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be equal to or greater than both MCPRf and MCPR limits at indicated core flow and THERMAL POWER as p
specified in the COLR.
l APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25f of RATED THERMAL POWER.
ACTION:
With MCPR less than the applicable MCPR limit, initiate corrective action l
within 15 minutes and restore MCPR to within the required limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.2.3 MCPR shall be determined to be equal to or greater than the MCPR limit specified in the COLR:
l a.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, b.
Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating c.
with a LIMITING CONTROL ROD PATTERN for MCPR.
d.
The provisions of Specification 4.0.4 are not applicable.
RIVER BEND - UNIT 1 3/4 2-B Amendment No.42
4 p--
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Figure 3.2.3-1 has been deleted,
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RIVER BEND - UNIT 1 3/4 2-9 Amendment No. 33,42
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i Figure 3.2.3-2 has been deleted, t
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RIVER BEND - UNIT 1.
3/4 2-10 Amendment No.42 i
s a
POWER DISTRIBUTION LIMITS 3/4.2.4 LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.4 The LINEAR HEAT GENERATION RATE (LHGR) shall not exceed the limits specified in the COLR.
APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.
ACTION:
With the LHGR of any fuel rod exceeding the limit, initiate corrective action within 15 minutes and restore the LHGR to within the limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.2.4 LHGR's shall be determined to be equal to or less than the limit:
a.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, b.
Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> af ter completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is c.
operating on a LIMITING CONTROL ROD PATTERN for LHGR.
d.
The provisions of Specification 4.0.4 are not applicable.
RIVER BEND - UNIT 1 3/4 2-11 Amendment No. 33,42
(
v 1
TABLE 3.3.1-1 (Continued)
REACTOR PROTECTION SYSTEM INSTRUMENTATION TABLE NOTATIONS (a)
A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter.
(b)
Unless adequate shutdown margin has been demonstrated per Specification to and during the time any control rod is withdrawn *.3.1.1, t f
t (c)
An APRM channel is inoperable if there are less than 2 LPRM inputs per level or less than 11 LPRM inputs to an APRM channel.
(d)
This function is not required to be OPERABLE when the reactor pressure vessel head is removed per Specification 3.10.1.
i-(e)
This function shall be automatically bypassed when the reactor mode switch l
1s not in the Run position.
\\
4 (f)
This function is not required to be OPERABLE when DRYWELL INTEGRITY is not required.
(g) With any control rod withdrawn.
l per. Specification 3.9.10.1 or 3.9.10.2.Not applicable to control rods removed (h)
This function shall be automatically bypassed when turbine first stage pressure is < 187 psig,** equivalent to THERMAL POWER less than 40% of RATED THERMAI POWER.
l l
l l
l i
"Not required for control rods removed
- To allow for instrumentation accuracy,per Specification 3.9.10.1 or 3.9.10.2.
< 177 psig turbine first stage pressure shall be used. calibration and drift, a setpoint of I
RIVER BEND - UNIT 1 3/4 3-5
- o TABLE 3.3.1-2' 2
E REACTOR PROTECTION SYSTEM RESPONSE TIMES
- 5 RESPONSE TIME FUNCTIONAL UNIT E
(Seconds)
Z 1.
s a.
Neutron Flux - High NA b.
Inoperative NA 2.
Average Power Range Monitor *:
a.
Neutron Flux - High, Setdown M4 b.
Flow Biased Simulated Thermal Power - High
<0.09**
c.
Neutron Flux - High 70.09 d.
Inoperative NA R
3.
Reactor Vessel Ste a Dome Pressure - High
<0.35 4.
Reactor Vessel Water Level - Low, level 3 71.05 Y
5.
Reactor Vessel Water Level - High, level 8 E1.05 6.
Main Steam Line Isolation Valve - Closure 70.09 7.
Main Steam Line Radiation - High NA 8.
Drywell Pressure - High NA 9.
Scram Discharge Volume Water Level - High a.
Level Transmitter NA i
b.
Float Switches MA
- 10. Turbine Stop Valve - Closure l
- 11. Turbine Control Valve Fast Closure, Valve Trip System
-<0.06 Dil Pressure - Low
<0.07f
> 12.
Reactor Mode Switch Shutdown Position NA 2
- 13. Manual Scram M4 cf' 2
5
- Neutron detectors are exempt from response time testing. Response time shall be measured from the detector output or from the input of the first electronic component in the channel.
z
- Not including simulated themal power time con; tant specified in the COLR.
l
. g
- Measured from start of turbine control valve fast closcrc
?
2 m
x
.s
~
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-s
t TABLE 4.3.1.1-1 (Continued)
~
5 l
y REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIRDENTS A
z (f) The LPRMs shall be calibrated at least once per 1000 effective full power hours (EFPH) using the TIP system.
E (g) Calibrate Rosemount trip unit setpoint at least once per 31 days.
Z (h) Verify measured drive flow to be less than or equal to established drive flow at the existing flow i
control valve position.
(i) This calibration shall consist of verifying the simulated thermal power time constant is within the limits specified in the COLR.
(j) This function is not required to be OPERABLE when the reactor pressure vessel head is removed per Specification 3.10.1.
(k) With any control rod withdrawn.
Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
(1) This function is not required to be OPERABLE when DRYWELL INTEGRITY is not required per Specifica-tion 3.10.1 R (m) Verify the Turbine Bypass Valves are closed when THERMAL POWER is greater than or equal to 40Y, RATED THERMAL POWER.
Y (n) The CHANNEL RECTIONAL TEST and CHANNEL CALIBRATION shall include the turbine first stage pressure instruments.
(o) The CHANNEL CALIBRATION shall exclude the flow reference transmitters; these transmitters shall be calibrated at least once per 18 months.
(p) This period may be extended to the first refueling outage, not to exceed 9-15-87.
N
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8
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5 J
l-l I
L IN51tuMENTAT10N 3/4.3.2
!$0LAT!0N ACTUATION INSTRUM[NTATION LIMITING CONDITION FOR OP[tATION 3.3.2 The isolation actuation instrumentation channels shown in Table 3.3.11 i
shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip setpoint column of Table 3.3.2 2 and with 150LAT10N SYSTEM RESPON$t 1
i TIME as shown in Table 3.3.2 3.
APPLICABILITY:
As shown in Table 3.3.2 1.
ACTION:
L a.
With an isolation actuation instrumentation channel trip setpoint I
less conservative than the value shown in the Allowable Values column of Table 3.3.2-2, declare the channel inoperable until the channel l
1s restored to OPERABLE status with its trip setpoint adjusted l
consistent with the Trip $ttpoint value.
l b.
With the nueer of OPERABLE channels less than required by the Minimum CPERABLE Channels er Trip Systes requirement for one trip sys.tes place the inoperab e channel (s) and/or that trip systen in the tripped condition" within one hour. The provisions of Specification 3.0.4 are not applicable, With.the number of OPkRABLE channels less than required by the Minimun c.
OPERABLE Channels per Trip Systes requirement for both trip systems, l
place at least one trip system ** in the tripped condition within one 4
hour and take the ACT10N required by Table 3.3.2-1.
l l
1
- An inoperable channel need not be placed in the tripped condition where this would cause the Trip Function to occur.
In these cases, the inoperable channel i
shall be restored to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or the ACTION required by 1
Table 3.3.21 for that Trip Function shall be taken.
C*The trip system need not be placed in the tripped condition if this would cause the Trip Function to occur.
When a trip system can be placed in the tripped condition without causing the Trip Function to occur, place the trip system with the most inoperable channels in the tripped condition; if both systens have the same number of inoperable channels, place either trip systen in the tripped condition.
i RIVER BEND - UNIT 1 3/4 3-10
, s...
i 3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section assure that the peak cladding temper-ature following the postulated design basis loss-of-coolant accident will not exceed the 2200'F limit specified in 10 CFR 50.46.
4 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE The peak cladding temperature (PCT) following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is dependent only second-arily on the rod to rod power distribution within an assembly. The peak clad temperature is calculated assuming a LHGR for the highest powered rod which is equal to or less than the design LHGR corrected for densification.
This LHGR times 1.02 is used in the heatup code along with the exposure-dependent steady state gap conductance and rod-to-rod local peaking factor.
The Technical Specification AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) is this LHGR of the highest powered rod divided by its local peaking factor.
The limiting value for APLHGR is specified in the CORE OPERATING LIMITS REPORT (COLR).
l The daily requirement for calculating APLHGR when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER is sufficient since power distribu-tion shifts are very slow when there have not been significant power or control rod changes.
The requirement to calculate APLHGR within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the com-pletion of a THERMAL POWER increase of at least 15% of RATED-THERMAL POWER ensures thermal limits are met after power distribution shifts while still allotting time for the power distribution to stabilize.
The requirement for calculating APLHGR af ter initially determining a LIMITING CONTROL R00 PATTERN exists ensures that APLHGR will be known following a change in THERMAL POWER or power shape that could place operation into a condition exceeding a thermal limit.
The calculational procedure used to establish the APLHGR specified in the COLR is based on a loss-of-coolant accident analysis.
The analysis was per-formed using General Electric (GE) calculational models which are consistent I
with the requirements of Appendix K to 10 CFR 50.
A complete discussion of each code employed in the analysis is presented in NEDE-20566(1)
Differences in this analysis compared to previous analyses can be broken down as follows.
a.
Input Chances l
1.
Corrected Vaporization Calculation - Coefficients in the vaporization correlation used in the REFLOOD code were corrected.
2.
Incorporated more accurate bypass areas - The bypass areas in the top guide were recalculated using a more accurate technique.
RIVER BEND - UNIT 1 B 3/4 2-1 Amendment No. 12,33,42 l-l
\\
POWER DISTRIBUTION LIMITS BASES AVERAGE PLANAR LINEAR HEAT GENERATION RATE (Continued) 3.
Corrected guide tube thermal resistance.
4.
Correct heat capacity of reactor internals heat nodes, b.
Model Chance 1.
Core CCFL pressure differential - 1 psi - Incorporate the assumption that flow from the bypass to lower plenum must overcome a 1 psi pressure drop in core.
2.
Incoporate NRC pressure transfer assumption - The assumption used in the SAFE-REFLOOD pressure transfer when the pressure is increasing was changed.
A few of the changes affect the accident calculation irrespective of CCFL.
These changes are listed below.
a.
Input Change 1.
Break Areas - The DBA break area was calculated more accurately.
b.
Model Chance 1.
Improved Radiation and Conduction Calculation - Incorporation of CHASTE-05 for heatup calculation.
A list of the significant plant input parameters to the loss-of-coolant accident analysis is presented in Bases Table B 3.2.1-1.
For plant operation with a single recirculation loop, the MAPLHGR limits are multiplied by 0.84.
The constant factor 0.84 is derived from LOCA analyses l
initiated from single recirculation loop operation to account for earlier boil-ing transition at the limiting fuel mode compared to the standard LOCA evaluations.
F4.2.2 APRM SETPOINTS The fuel cladding integrity Safety Limits of Specification 2.1 were based on a power distribution which would yield the design LHGR at RATED THERMAL POWER.
The flow biased simulated thermal power-high scram trip setpoint and the flow biased neutron flux-upscale control rod block trip setpoints of the APRM instru-ments must be adjusted for both two recirculation loop operation and single recirculation loop operation to ensure that MCPR does not become less than the fuel cladding safety limit or that > 1% plastic strain does not occur in the degraded situation.
The scram settTngs and rod block settings are adjusted in accordance with the formula in this specification, when the combination of THERMAL POWER and CMFLPD indicates a peak power distribution, to ensure that an LHGR transient would not be increased in degraded conditions.
RIVER BEND - UNIT 1 B 3/4 2-2 Amendment No. I2, 3I, 33,42
y POWER 01STRIBUT!0N LIMITS 1
1 1
\\
i Bases Table B 3.2.1-1 1
i SIGNIFICANT INPUT PARAMETERS TO THE LOSS OF-COOLANT ACCIDENT ANALYSIS i
Plant Parameters; Core THERMAL POWER.................... 3015 Mwt* which corresponds to 105% of rated steam flow Ve s sel Ste am Output................... 13. 08 x 106 lbm/hr which corresponds to 105% of rated steam flow Ves sel Steam Dome Pressure............. 1060 psia i-Design Basis Recirculation Line Break Area for:
t 2
a.
Large Breaks 2.2 ft,
2 b.
Small Breaks 0.09 ft,
Fuel Parameters:
PEAK TECHNICAL INITIAL SPECIFICAT]DN DESIGN MINIMUM LINEAR HEAT AKIAL CRITICAL FUEL ASSEMBLY GENERATION RATE PEAKING POWER FUEL TYPE GEDMETRV' (kw/ft)
FACTOR RATIO Initial Core 8x8 13.4
- 1. 4 1.17**
l A more detailed listing of input of each model and its source is presented in Section 11 of NEDE 20566(1) and subsection 6.3.3 of the FSAR.
- This power level meets the Appendix K requirement of 102%.
The core heatup calculation assumes an assembly power consistent with operation of the highest powered rod at 102% of its Technical Specification LINEAR HEAT GENERATION RATE limit.
RIVER BEND - UNIT 1 B 3/4 2-3 Amendment No. 33 a
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4 POWER DISTRIBUTION LIMITS BASES 3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRs at steady state operating conditions specified in the COLR are derived from the established fuel cladding integrity i
Safety Limit MCPR specified in Specification 2.1.2 and an analysis of abnormal operational transients.
For any abnormal operating transient analysis, 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 settings 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 loss of flow, increase in pressure an)d power, positive reactivity insertion.
nd coolant temperature decrease.
The limiting transient yields the largest delta MCPR.
When added to the Safety Limit MCPR, the required minimum operating limit MCPR specified in the COLR is obtained.
Analysis of transients occurring during single recirculation loop operation indicates that the maximum operating limit MCPR will be bounded by the limits specified in the COLR.
The power-flow map l
of Figure B 3/4 2.3-1 shows typical regions of plant operation, The evaluation of a given transient begins with the system initial param-eters identified in Reference 2 that are input to a GE core dynamic behavior transient computer program.
The codes used to evaluate transients are described in Reference 2.
The principal result of this evaluation is the reduction in MCPR caused by transient.
The purpose of the MCPR and MCPR specified in the COLR define operating l
f p
limits at other than rated core flow and power conditions.
At less than 100%
of rated flow and power the required MCPR is the larger value of the MCPR and f
MCPR at the existing core flow and power state.
The MCPR s are established p
f l-to protect the core from inadvertent core flow increases such that the 99.9%
MCPR limit requirement can be assured.
The MCPR,s were calculated such that, for the maximum core flow rate and the. corresponding THERMAL POWER along the 105%-of-rated steam 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 MCPRs were calcu-lated at different points along the 105%-of rated steam flow control line corresponding to different core flows.
The calculated MCPR at a given point of core flow is defined as MCPR.
f RIVER BEND - UNIT 1 B 3/4 2-4 Amendment No, If, 3Z,42
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O DESIGN FEATURES 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor shall contain 624 fuel assemblies.
Each assembly shall t
consists of a matrix of Zircalloy clad fuel rods with an initial composition i
of slightly enriched uranium dioxide (00 ) as fuel material.
Fuel assemblies 2
shall be limited to those fuel designs approved by the NRC staff for use in BWR's.
CONTROL ROD ASSEMBLIES 5.3.2 The reactor core shall contain 145 cruciform shaped control rod assemblies.
The control material shall be boron carbide powder (B.C) and/or hafnium metal.
The control rod assemblies shall be full length.
5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained:
a.
In accordance with thf. code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to tht applicable Surveillance Requirements, b.
For a pressure of:
1.
1250 psig on the suction side of the recirculation pump.
2.
1650 psig from the recirculation pump discharge to the outlet side of the discharge shutoff valve.
3.
1550 psig from the discharge shutoff valve to the jet pumps.
c.
For a temperature of 575'F.
VOLUME 5.4.2 The total water and steam volume of the reactor vessel and recirculation system is approximately 16,000 cubic feet.
RIVER BEND - UNIT 1 5-5 Amendment No. 23,42
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i
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.u DE52GN FEATURES l
4 4
5.5 METEOROLOGICAL TDWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1.1-1, 5.5 FUEL STORAGE CRITICALITY 5.6.1.1 The spent fuel storage vacks are designed and shall be maintained with:
A k,gg less than or equal to 0.95, when flooded with unborated water, a.
includin all calculational uncertainties and biases as described in Section
.1 of the FSAR.
b.
A fuel assembly center-to center storage spacing of 7 in, within rows and 12.25 in, between rows in the Low Density 5torage Racks in the upper containment pool.
A fuel assembly center-to center storage spacing of 5.28 in., with a e
neutron poison material between storage spaces, in the High Density Storage Racks in the spent fuel storage facility in the Fuel Building,
, The stor,sge of spent fuel in the upper containment fuel storage pool is pronibited during OPERATIONAL CONDITIONS 1 and 2.
5.6.1.2 For the first core loading, the K,gg for new fuel stored dry in the spent fuel storage racks shall be administratively controlled to not exceed O.98 when optimum soderation (foam, spray, fogging, or small droplets) is assumed.
s 5.6.1.3 provisions shall be taken to avoid the entry of sources of optimum moderation (foam, spray, fogging, or small droplets) to preclude that K,gg for new fuel, stored in the new fuel storage facility, could exceed 0.98.
i DRAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent
. inadvertent draining of the pool below elevation 95'.
l CAPACITY r
5.6.3 The spent fuel storage pool in the fuel building is designed and shall be maintained with a storage capacity limited to no more than 2680 fuel i
assemblies. Only fuel manufactured by General Electric may be stored in the spent fuel pool.
i 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT l
i 5.7.1 The components identified in Table 5.7.1-1 are designed and shall be i
maintained within the cyclic or transient limits of Table 5.7.1-1.
e l
l RIVER BEND - UNIT 1 56
ADMINISTRATIVE CONTROLS l
l' SEMIANNUAL EFFLUENT RELEASE REPORT (Continued) l to MEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY l
(Figure 5.1.3) during the report period.
All assumptions used in making these assessments (i.e., specific activity, exposure time and location) shall be in-i l
cluded in these reports.
The assessment of radiation doses shall be performed in accordance with the methodology and parameters of the OFFSITE DOSE CALCULA-T!ON MANUAL (ODCM).
The Semiannual Radioactive Effluent Release Report to be submitted 60 days after i
January 1 of each year shall also include an assessment of radiation doses to 1
the likely most-exposed MEMBER OF THE PUBLIC from reactor releases and other l
nearby uranium fuel cycle sources (including doses from primary effluent pathways and direct radiation) for the previous calendar year to show conformance with i
40 CFR Part 190 Environmental Radiation Protection Standards for Nuclear Power Operation.
Acceptable methods for calculating the dose contribution from liquid and gaseous offluents are given in Regulatory Guide 1.109, Rev.1, October 1977.
The Semiannual Radioactive Effluent Release Reports shall include a list and description of unplanned releases from the site to UNRESTRICTED AREAS of radio-active materials in gaseous and liquid effluents made during the reporting period.
The Semiannual Radioactive Effluent Release Reports shall include any changes made during the reporting period to the PROCESS CONTROL PROGRAM (PCP) and to the ODCM, as well as a listing of new locations for dose calculations and/or environmental monitoring identified by the land use census pursuant to Specifi-cation 3.12.2 SPECIAL REPORTS 6.9.2 Special reports shall be submitted in the following manner:
Special reports shall be submitted to the U.S. Nuclear Regulatory a.
Commission, Document Control De@., W shington, DC 20555, with a copy to the Regional Office of the NRC and a copy to the NRC Resi-dent Inspector, within the time period specified for each report.
l b.
Special reports in regard to Corbicula will be submitted to the NRC
(
within 30 days of identification of infestation.
In accordance with the settlement agreement dated October 10, 1984, these reports shall describe the level of infestation, affected systems and measures taken to prevent further infestation.
CORE OPERATING LIMITS REPORT l
l 6.9.3.1 Core operating limits shall be established prior to startup from each reload cycle, or prior to any remaining portion of a reload cycle, for the following:
The AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGR) for a.
Specification 3.2.1.
RIVER BEND - UNIT 1 6-19 Amendment No.17,42
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f ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Continued) b.
The MINIMUM CRITICAL POWER RATIO (MCPR) for Specification 3.2.3.
The LINEAR HEAT GENERATION RATE (LHGR) of Specification 3.2.4.
c.
d.
The REACTOR PROTECTION SYSTEM (RPS) response time for APRM thermal l
time constant for Specification 3.3.1.
\\
and shall be docume.ited in the CORE OPERATING LIMITS REPORT (COLR).
t I
6.9.3.2 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 NEDE 24011-P-A, " General Electric Standard Application for Reactor Fuel" (latest approved version).
6.9.3.3 The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal-h margins, ydraulic limits, ECCS limits and nuclear limits such as shutdown and transient and accident analysis limits) of the safety analysis are met.
6.9.3.4 The CORE OPERATING LIMITC REPORT, including any mid-cycle revision 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 Inspector.
6.10 RECORD RETENTION 6.10.1 In addition to the appliu ble record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated.
6.10.2 The following records shall be retained for at least 5 years:
Records and logs of unit operation covering time interval at each a.
I power level.
b.
Records and logs of principal maintenance activities, inspections, ropair, and replacement of principal items of equipment related to nuclear safety.
{
RIVER BEND - UNIT 1 6-19a Amendment No. 42 1
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)
I ADMINISTRATIVE CONTROLS RECORD RETENTION (Continued)
I i
c.
All REPORTABLE EVENTS 1
d.
Records of surveillance activities, inspections, and calibrations required by these Technical Specifications.
Records of changes made to the procedures required by Specification e.
6.8.1.
f.
Records of radioactive shipments.
g.
Records of sealed source and fission detector leak tests and results.
h.
Records of annual physical inventory of all sealed source material
(
of record.
i.
Records of emergency drills and exercises.
6.10.3 The following records shall be retained for the duration of the unit Operating License:
Records and drawing changes reflecting unit design modifications made a.
to systems.and equipment described in the Final Safety Analysis Report, b.
Records of new and irradiated fuel inventory, fuel transfers, and assembly burnup histories, Records of radiation exposure for all individuals entering radiation c.
control areas, d.
Records of gaseous and liquid radioactive material released to the environs.
Records of transient or operational cycles for those unit components i
e.
identified in Table 5.7.1-1.
i l
f.
Records of reactor tests and experiments.
g.-
Records of training and qualification for current members of the unit
- staff, h.
Records of inservice inspections performed pursuant to these Technical Specifications.
1.
Records of quality assurance activities required by the Operational Quality Assurance Manual that are not listed in Specification 6.10.1.
l l
l RIVER BEND - UNIT 1 6-20 Amendment No. 42 l'
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