ML20011D738
| ML20011D738 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 07/01/1989 |
| From: | CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | |
| Shared Package | |
| ML20011D723 | List: |
| References | |
| NUDOCS 8912280341 | |
| Download: ML20011D738 (70) | |
Text
TABLE OF CONTENTS CHAPTER 4: REACTOR 4.1
SUMMARY
DESCRIPTION 4.2 FUEL SYSTEM DESIGN j
4.2.1 DESIGN BASIS 4.
2.2 DESCRIPTION
AND DESIGN DRAWINGS 4.2.3 DESIGN EVALUATION 4.2.4 TESTING AND INSPECTION PLAN 4.3 NUCLEAR DESIGN 4.3.1 DESIGN BASIS 4.3.2 NUCLEAR DESIGN DESCRIPTION 4.3.3 ANALYTICAL METHODS 4.4 TIERMAL AND HYDRAULIC DESIGN 4.4.1 CORE FLOW AND PRESSURE DROP 4.4.2 HYDRAULIC STABILITY 4.5 OPERATION WITH LESS THAN ALL LOOPS 4.6 PEACTIVITY CONTROL SYSTEMS 4.6.1 REACTIVITY CONTROL SYSTEMS OPERABILITY REQUIREMENTS 4.7 CONTROL ROD DRIVE SYSTEMS 4.7.1 CONTROL R0D BLADE ASSEMBLIES 4.7.2 CONTROL ROD DRIVES 4.7.3 CONTROL ROD DRIVE MATERIALS OF CONSTRUCTION 4.7.4 CONTROL ROD DRIVE HYDRAULIC CONTROL SYSTEM 4.7.5 CONTROL ROD SYSTEM INSTRUMENTATION AND CONTROL 4.7.6 CONTROL ROD SYSTEM PERFORMANCE REQUIREMENTS
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4.7.7 BWR SCRAM SYSTEM PIPE BREAKS 2
MIO588-0281A-BX01 gp PDh h CK 05oon353 41 891222 K
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4.7.8 CONTROL ROD DRIVE SUPPORT STRUCTURE 4.7.9 ACTIVITY OF COBALT-60 PRODUCED IN THE CONTROL RODS AND ACTIVATED CONTROL ROD TRANSFER ANALYSIS 4.7.10 CONTROL ROD DRIVE FAILURE MODES AND EFFECTS ANALYSIS s
4.7.11 CONTROL ROD DRIVE ANALYSIS OF OPERATOR ERRORS 4.8 LIQUID POISON SYSTEM (LPS) 4.8.1 LPS PURPOSE AND
SUMMARY
DESCRIPTION 4.8.2 LPS DESIGN CHARACTERISTICS 4.8.3 LPS INSTRUMENTS AND CONTROLS 4.8.4 LPS EFFECTIVENESS 4.8.5 LPS EQUIVALENT CONTROL CAPACITY f
4.8.6 INTEGRITY OF LPS PRESSURE ISOLATION VALVES 4.8.7 LPS INJECTION GAS PRESSURE t
4.8.8 LPS PERFORMANCE REQUIREMENTS 10 V
HIO588-0281A-BX01
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l \\
CHAPTER 4 4.0 REACTOR 4.1 St&&iARY DESCRIPTION l-The Big Rock Point Reactor is a commercial boiling water reactor rated at 240 MWt. The reactor core is small with an active region about six feet in height and six feet in diameter. Due to its size, it is very stable and has a relatively high leakage.
To compensate for the high leakage, K-infinity must be higher than for larger plants.
Eighty-four fuel assemblies and thirty-two control rods are used in the core. The ratio of interior control rods to interior assemblies is one to two.
1 The reactor has external recirculation loops with constant velocity pumps. Recirculation flow control is not employed and reactivity 4
maneuvering is done entirely with control rods.
Controls rods are l
manipulated individually and have half core symmetry, j
i A description of the Reactor Vessel and Internals Design is provided i
in Chapter 5 of this Updated FHSR.
In-core power distribution measurements are provided by the activation l
(q of eight fluxwires arranged in a half core synunetrical configuration.
l g
These are utilized to verify calculated axial power shapes.
The primary coolant system is pressurized to 1350 psia and maximum.
exit void fraction is about 55 percent.
I l
Reload cores for Big Rock Point are designed as outlined in the i
Physics Methodology Report,'Rev 3 of October 11, 1982. The reload design consists of selecting an assembly loading pattern and a control rod withdrawal sequence that meets the constraints of shutdown margin, rod worth, notch worth, maximum average planar linear heat generation rate, minimum critical power ratio, assembly power, and i
heat flux while allowing operation at the highest possible power l
level. The reload design also verifies that reactivity coefficients, j
delayed neutron fraction, liquid poison worth, and scram reactivity i
insertion are within assumptions used in the plant accident and-transient analysis.
1 The startup physics test program at the beginning of each fuel cycle consists of verification of shutdown margin, comparison of the zero power critical control rod density with prediction, and comparison of measured flux wire shapes with prediction, j
/
i 4.1-1 MI1287-1835A-BX01
The following is a list of acronyms used in Chapter 4:
ANF - Advanced Nuclear Fuels (formerly ENC)
BOC - Beginning of Cycle BOL - Beginning of Life BORAX - Boiling Reactor Experiment BWR - Boiling Water Reactor EBWR - Experimental Boiling Water Reactor EOC - End of Cycle EOL - End of Life ENC - Exxon Nuclear Company GWD - Gigawatt Days MAPLHGR - Maximum Arial Planar Linear Heat Generation Rate MCHFR - Minimum Critical Heat ' lux Ratio l
MCPR - Minimum Critical Power Ratio MT - Metric Tons MVT - Megawatt Thermal Re - Bare Rod Reynolds Number SPERT - Special Power Excursion Reactor Tests ST - Standard Tons VBWR - Vallecitos Boiling Water Reactor Ov 1
i f
OV 4.1-2 MI1287-1835A-BX01
s 4.2 FUEL SYSTEMS DESIGN 4.2.1 DESIGN BASIS The fuel used in e Big Rock reactor is designed such that it will operate without fuilure under all the different reactor conditions.
The fuel assemblics must be structurally compatible with the reactor core, including fuel channels and assembly supports. A detailed summary of the design criteria is presented in XN-NF-85-39(P),
" Summary of ENC Mechanical Design Criteria, Failure Mechanisms, and Material Properties for BWR Fuel Assemblies", January 1986. Following is an overview of the stress, strain, and various failure mechanisms addressed in the design basis.
(For detailed information refer to XN-NF-85-39(P)).
The fuel assembly is designed to withstand handling loads without deformation to the fuel bundle. Each tie rod upper end cap is e
designed to withstand axial loading.
Should excessive loads he applied to the assembly the bundle is designed to fail in a manner which assures no breaking of cladding will occur.
To prevent gaps from developing between fuel pellets, a plenum spring is placed in each fuel pin.
The plenum spring is designed to exert on the fuel column a force necessary to ensure the fuel remains
,-~g s mted through handling, shipping, loading and fuel densification.
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4 Spring relaxation as the result of radiation during the period' of
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fuel densification has been considered.
The fuel pins are prepressurized. The amount of prepressurization is l
designed so that thermal performance is acceptable at beginning of life and so pressure at end of life is not so large as to cause excessive stress.
Internal gas pressure is evaluated as a function of burnup and remains below system pressure.
The fuel must have the proper clearances te ensure adequate coolant flow along the fuel pins.
The design criteria for free space between rods and between rods and channels must be considered to avoid reductions in MCPR and is specified.
The failure mechanisms caused by vibration are addressed.in the design basis. Vibrations caused by flow can contribute to fretting corrosion failures.
Spacer grids are designed to prevent fretting as well as bow due to axial forces. The spring force exerted by the spacer grids must be sufficient to overcome flow induced vibration forces.
Ilydrogen absorption can contribute to early cladding failure by reducing ductility and by forming Zirconium Hydride. The hydrogen in the cladding is controlled through water chemistry and the initial concentration in the fuel and cladding. The design limit for hydrogen concentration has been addressed and is specified.
4.2-1 MI1287-1835A-BX01
1 i
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The corrosion mechanism can reduce bundle life under heavy crud conditions.
Heavy crud conditions are not expected since the limit for the maximum external oxide layer is specified as predicted by the RODEX2 code. Fuel temperatures are also determined by RODEX2 and are below the melting temperature of UO2.
The design limits for cladding temperatures on the inside surface of the cladding, for the average 7
temperature of the cladding, and on the outside surface are specified.
The maximum cladding ID teroperature is specified to protect against fuel-cladding chemical interaction.
This interaction between U02 and t
Zirealoy occurs around 1000*F.
The maximum cladding surface temperature is designed to minimize corrosion of the cladding. The limit on averags temperature is to maintain adequate material strength.
The design basis encompasses cladding stress and strain considerations.
The maximum steady-state strain at end of life has a specified design limit. The transient stresses induced during power changes expected under operating conditions and the transient strain have been specified.
The basis for cladding stress limits is that the fuel will not be i
damaged when incurring these stresbes. The methodology for calculating the stresses is presented in XN-NF-85-38(P) Rev 0.
Those stresses calculated include the following:
primary membrane stresses, bending stresses due to ovality, thermal stresses, stresses from mechanical bow, flow induced vibrational stresses, stresses from spacer springs, rs stresses on f t.el end caps, cladding stresses during power changes,
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and cladding cycle fatigue.
Fuel densification coupled with cladding creep can lead to increased localized strain and cladding-collapse.
The reduction in pellet to clad gap will not be greater than the initial cold gap up to a specified rod averaged exposure.
Various performance characteristics of the fuel rod are modeled by use of RODEX2 computer code. The use of RODEX2 is documented in XN4T-81-58(P) Supplement 1 & 2, Rev 2 "RODEX2 - Fuel Rod Thermal -
Mechanical Response Evaluation Code".
RODEX2 Models the following fuel chara aristics: gas release, radial thermal conduction and gap l
conductance, free rod volume and gas pressure calculations, pellet-cladding interaction, fuel swelling and densification, cracking and crack healing, cladding creep deformation and irradiation induced growth.
An evaluation of the design parameters for Big Rock Point fuel is
+
presented in Section 4.2.3.
4.
2.2 DESCRIPTION
AND DESIGN DRAWINGS 1
Dimensions, tolerances, and applicable standards are presented in Exxon Parts List XN-NF-PL-173 where drawings for the fuel bundles, fuel rods, poison rods, and a load map may be reviewed.
4.2-2 M11287-1835A-BX01
o The basic mechanical design of the 11 and I fuel types are essentially the same. !!-3 through the current I fuel types were modified from Il-1 and }I-2 by utilizing an eight tie rod upper tie plate locking design. The 11-1 and 11-2 fuel utilized twelve tie rods with a locking fork to secure the top tie plate. Additionally, insulator disks were removed and the ratio of pellet length to diameter was increased (XN-hT-82-63).
I fuel was further changed by increasing internal pressure and reducing pellet to clad gap to increase heat transfer characteristics. (XN-hT-85-38(P))
Each fuel bundle is made of 121 rods which are Zircaloy 2 tubes in a 11 x 11 array with the necessary handle, base, intermediate spacers, fasteners and other hardware. The fuel rods are positioned at the top and bottom of the fuel bundle by tie plates. The lower tie plate is machined stainless steel casting with a grid to accommodate the lower end of the zirealoy tubes.
The upper tie plate is a cast and machined grid plate with attached handle. An orientation lug is located on one side of the handle to permit orientation of the bundle.
A total of 8 rods are used as tie rods to secure the bottom and top tie plates.
(11-1 and 11-2 fuel types use a 12 tie rod locking fork mechanism.) The bottom of the tie rods are threaded into the lower tie plate. The upper ends of the tie rods have a locking device q
which holds the upper tie plate against the coil springs on the t'y remaining rods. The remaining rods are held securely between the tie plates. The coiled springs on the top of these rods support the upper tie plate, seat the fuel rods against the lower tie plate, and allow for relative expansion.
Three spacer grids maintain the array dimensions at intermediate levels of the fuel bundle. An inert spacer capture rod maintains the axial position of the spacer grids.
The fuel bundle contains 117 fuel rods, 4 of which contain a burnable poison. The fuel cladding is cold worked and stress relieved.
In addition, there are 4 inert rods in the array with the same outside diameter as the fuel pins. The dimensions of the fuel pellets and jackets are provided in XN-NF-85-38(P). All the fuel rods contain 70 inches of active fuel length, plus a plenum located at the top of the fuel pin. The fuel rods have an upper and lower plug-type end cap which are seal welded to the cladding. The fuel rods are pre-pressurized.
4.2 M11287-1835A-BX01
i f
4.2.3 DESIGN EVALUATION (XN-hT-85-38(P))
Structural testing of the fuel assembly was performed by applying a load between the upper tie plate bail and the lower tie plate. A specified load was applied and then returned to zero.
There was no l
evidence of permanent deformation. A load sufficient to cause i
failure was applied. Failure occurred at the tie rod end caps. No yielding of the cladding occurred.
To prevent fretting wear from flow induced vibration, a specified initial spring force in the grid spacers is required at the rod ends.
Positive contact is necessary at the central grid spacer. The spring force at the top and bottom grids is found to be a percentage of the initial force at a specified MWD /Mt.
The average sprir; force is as specified. Therefore the residual spring force is sufficient to prevent fretting.
Fuel assembly growth is determined by the amount of tie rod growth.
This growth must not be so large as to cause fuel rods to become l
disengaged from the upper tie plate. The maximum differential growth is shown to be as specified at EOL based upon actual measurements.
The nominal upper tie plate engagement is as specified. Therefore, adequate engagement in the upper tie plate is provided.
Tuel assembly liftoff due to hydraulic load on the assembly is evaluated. The minimum downward force exceeds the maximum lif ting force. Therefore, lift off will not occur.
Steady state strain due to cladding creep was analyzed using the RODEX 2 computer code through end of life and found to be within the design limit.
Hydrogen concentration was evaluated using the methodology referenced in XN-NF-85-38(P) Rev 0.
The total hydrogen concentration in the cladding was determined to be less than the design limits.
The maximum thickness of the oxide layer at the end of life exposure was determined to be well within the design limit to control cladding corrosion.
The stresses were analyzed and found to be within design limits.
Primary membrane stresses caused by coolant pressure and fill gas pressure were a maximum at BOL and EOL hot and were less than the design limit. Primary membrane plus primary bending due to ovality was less than the design limit. When the stresses from vibration, thermal gradient, mechanical and thermal bow, and spacer pressure are-included the maximum stress is less than the limit.
Stresses caused by changes in power coupled with corrosive fission products were evaluated by using the operating history of the fuel and applicable computer codes.
(XN-hT-85-38(P), Rev 0, P. 32) The results show the s
maximum cladding circumferential stress was below the design limit.
The same analysis evaluated the maximum transient stress which is A
4.2-4 M11287-1835A-BX01
O below the design limit. Cyclic stresses resulting from fuel shuffling and power maneuvering were evaluated.
The ratio of the expected to the allowable number of loading cycles must sum to less than as specified. The expected fractional fatigue life usage was determined and specified.
Power history was used to evaluate cladding creep collapse. The I
results are specified.
Maximum expected fuel pellet temperature is well below the melting temperature of U02. Cladding temperature was determined. All temperatures were below the design basis limits.
Internal fuel rod pressure is specified at end of life and is below the reactor system pressure.
Measurements of the spacing between fuel rods and rods and channels have determined the gap closure.
The initial spacing is sufficient to allow for this closure and maintain the design limit.
4.2.4 TESTING AND INSPECTION PLAN Various testing has been performed upon irradiated fuel at Big Rock Point. The Fuel Performance Improvement Program sponsored by the US Department of Energy with participation by L,onsumers Power Company g
and Exxon Nuclear Company was responsible for much of the evaluation of the fuel. Visual inspections, flaw detection with eddy current, fuel rod diameter measurement, rod lengths, and cladding external oxide thickness were some of the measurements made on both standard and experimental fuel rods. The results are published in various documents including XN-NF-85-121, DOE /ET/3521;-38, UC-73, " Fuel Performance Improvement Program Poolside Fuel Examination Big Rock Point, January 1985 (EOL 19)".
The US Department of Energy, Consumers Power Company, Exxon Nuclear Company, and GPU Nuclear were involved in an extended burnup demon-stration program from 1979 through 1985. Measurements of rod length, rod diameter, oxide thickness, eddy current testing, and visual examinatious were made as part of this program. The results are reported in various documents including XN-NF-85-102. DOE /ET-34006-18, UC-78, " Extended Burnup Demonstration Reactor Fuel Program Poolside Fuel Examination Big Rock Extenued Burnup Fuel", January 1985 (EOC 19)".
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4.2-5 Mll287-1835A-BX01
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FICURE 4.0 FUEL ASSEMBLY SKETCH
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water at 212'F, and is selected for its relatively high concentration of boren.
4.8.2.2 The discharge line from the tank is routed to alternate injection points into the bottom cf the reactor vessel and into the recirculating pump suction lines (See Drawina 0740G40121). A pressure equalizing I
line is provided between the steam drum and the tank, and a pressurization connection it provided from a bank of 2000 psig
)
nitrogen bottles (Drawing 0',40040107).
4.8.2.3 Normally, the tank is isolated from the nuclear steam supply system valves in both the discharge and the equalizing lines. Also, the line from the nitrogen bottles is isolated so that the tank is near atmospheric pressure. The squib valves providing the isolation (equalization, pressurization, and injection) are designed for very i
high reliability and leak tightness. Each valve consists of a sealed inlet fitting in which flow is normally blocked by a precision machined shear plug, and a trigger assembly in which a ram is forced out by an explosive charge to shear off the plug.
The explosive charge is provided with two electrically fired squibs per valve, each of whirA is energized from a different d-c circuit.
Two parallel circuits supply all of the seven explosive valves so that an external failure or internal breaking of a circuit due to staggered firing will not prevent all valves from being energized.
Also, each individual line to the fourteen charges is fused so that a short circuit in an element during firing will not disturb the remainder. In the normal position, all elemeats are connected in series including the control relays and each circuit is continuously monitored, a failure in any part of a circuit causing an alans to be given in the main control room.
This type of valve, but with only a single explosive squib, has a demonstrated reliability of 99.96%. The pressurization line and equalizing line each have two full capacity valves in parallel and the injection line has three one inch valves in parallel so that the mathematical probability of valve failure is infinitesimal. The explosive, or squib, life is a function of environmental temperature.
The squibs provided in the poison system were replaced by low temperatute primers via Facility Change FC-180 and are suited for the operating range of -60*F to +160*F; the maximum espected ambient temperature in this zone is 120*F.-
Vendor documentation (storage life versus temperature curve) shows O
that expected lifetime can exceed ten years when ambient temperatures are kept at 120*F or below.
As degradation is primarily temperature 4.8-2 MIO688-0302A-BX01
i s
i dependent, appropriate action will be taken to ensure that primers that approach the tise temperature effective storage life limit do l
not remain in the poison system.
The explosive valves used are manufactured by Conax Corporation of
}
Buffalo, New York.
These have been used predominately by the missile industry. The primary modification for application at Big Rock Point i
is in the use of all stainless steel pressure parts rather than aluminum.
4.8.2.4 As the solution will start to precipitate at 90'F, the tank temperature J
is maintained at 150'T by a dual element electric immersion heater, with automatic temperature control.
The poison tank is insulated, so that upon loss of power to the heaters, the solution will not cool to i
j saturation temperature for approximately 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />.
J 4.8.2.5 Upon initiation of the poison valves admitting full 2000 psig nitrogen pressure to the poison tank, poison is forced into the reactor within
]
a few seconds; however, if the primary system is at full operating l
pressure, the nitrogen volume will be insufficient for forcing out more than a few gallons of solution.
The driving force for the i
remaining volume is achieved from the static head due to the elevated position of the tank and the head across the recirculating pumps.
If both recirculating pumps are down, a valve on the injection line into the pump suction is closed by an interlock on the pump motor breakers, i
and poison it forced directly into the reactor through a check valve.
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The primary purpose of nitrogen pressurization is to insure positive
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i displacement of poisoa solution when the reactor recirculation system is static, such as during refueling, when there is no initial driving head to establish a siphon through the discharge dip tube in the poison tank.
As sufficient pressurized nitrogen is available to start flow into the reactor with all drum safety valves blowing (1870 i
psia), all other operating conditions are considered less stringent.
For example, sufficient pressurized nitrogen is available to displace the entire poison solution from the tank without benefit of siphon for system operating pressures up to 350 pais. At any operating pressure with natural or pumped recirculation flow, a minimum of 9 feet friction loss through the reactor will establish siphon through the injection line and empty the poison tank even without benefit of l
nitrogen pressurization.
t 4.8.2.6 The top outlet arrangement of the poison injection connection was-selected to avoid precipitation of sodium pentaborate in contact with external piping and valves. The resulting siphon required is 3 feet I
when the tank is full and 9 feet when it is empty.
Integrity of this I
siphon, once established, is assured by the following (a) small size (3") of injection line, (b) in most cases internal pressure prevents air leakage into the line,-(c)-air in-leakage is improbable under any circumstances as the associated piping and flanges are designed for 2000 psig pressure.
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4.8-3 MIO688-0301A-BX01
i
)
4.8.2.7 Two spring-return switches are provided on the control console to initiate poison injection and both of these must be held closed to fire the explosive valves, thus insuring that operation of the system is a deliberate action.
In order to allow the operator the option of interrupting poison injection, once started, a remote controlled air operated valve is provided downstream of the three explosive injection valves. This
)
valve will open on air or power failure.
If the valve is closed, it can be reopened, and poison injection will continue. Operability requirements for this valve are addressed in Section 4.8.8 below.
4.8.2.4 Within the first five minutes, sufficient solution is introduced into 2
the primary system to produce a boron concentration of 1300 ppm.
This is equivalent to -16% Ak/k, which is more than a i quate to J
reduce the power level to zero by offsetting the effect of decreasing voids. Injection of the remaining solution raises the boron level to 2000 ppe, which produces -25% Ak/k, and is sufficient to hold the core suberitical even after it has cooled completely and with all j
control rods removed.
4 4.8.2.9 All components of this system can be tested for proper operation except that actual opening of the electrically actuated explosive valves would allow poison solution to enter the reactor. Thus, the
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primer-triggers are tested using a special test firing circuit. The t
primer-trigger is installed in a spare valve identical to those used in the poison system to insure the trigger assembly develops sufficient force to shear the integral inlet cap.
An isolation valve was added via Facility Change FC-481 to provide isolation between the Reactor Vessel poison nozzle and the poison injection inlet check valve.
This valve is normally locked open
,i during operation and may be closed to enable check valve internal inspection during periods when the LPS is not required to be operable.
In response to General Electric Service Information Letter (SIL #186) the LPS squib valve control circuit was modified via FC-398 to include a method of verifying the current limiting resistors integrity.
I A single coil solenoid valve was installed via Facility Change FC-149 to electrically fail close the Control Valve for poison to recirculating pump on loss of power. The valve is to be open when thet recirculating pumps run and closes on loss of electrical power or loss of air.
4.8.3 LPS INSTRUMENT AND CONTROLS Emergency injection of liquid poison is initiated manually from the control console.
In. order-to insure against accidental initiation, two control switches are provided in series, each of which must be i
closed individually to energize the injection valves. A separate
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manual control is provided to close tne injection line isolation.
1 valve and stop the flow of liquid poison, if desired.
This valve is i
1 4.8-4 MIO688-0302A-BX01 1
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O
\\V a normally open air operated valve (See Drawina 0740040107).
Indication of poison tank pressure is provided to assure that the tank has become pressurized; and poison tank level switches indicate that the tank is emptying.
Temperatures of the poison tank and squib primers are recorded. High temperature on the squib primers is annunciated on a local panel with remote annunciation of a trouble alarm to the control room. The recorder was installed via Facility Chan8e FC-265 in response to a CPCo December 7,1973 commitment to monitor squib primer temperatures.
4,8.4 LPS EFFECTIVENESS (REFERENCE'1).
As part of the Probatilistic Risk Assessment results relating to the efficacy of a recirculation Pump Trip, CPCo concluded that the Liquid Poison System will act much more rapidly than described in Section 4.8.2.8 above. The basis for the improved performance characteristics and LPS Evaluation are provided in Chapter 15, Section 15.8.6 of this l
Updated FHSR.
4.8.4.1 Fault Tree Analysis (Reference 1)
A fault tree analysis was developed to analyze the ability of the LPS to inject poison (borated water) into the reactor vessel following a situation in which the control rods have failed to insert into the
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core (such as during an Anticipated Transient Without Scram (ATWS) event). This analysis is provided in Attachment 4 of Reference 1.
4.8.5 LPS EQUIVALENT CONTROL CAPACITY (Reference 2)
Pevised rule 10 CFR 50.62(c)(4), Requirements For Reduction of Risk From Anticipated Transients Without Scram (ATWS) included Equivalent Control Capacity requirements for standby liquid control (BRP LPS) l which was defined by the NRC in a letter dated January 21, 1981 (Generic Letter 85-03). The equivalent capacity of 86 gpa of 13 weight 1 sodium pentaborate was the Equivalent Capacity requirement.
The Big Rock Point liquid poison system exceeds the requirements of this rule. Smaller vessels require proportionally less sodium pentaborate flow to meet the rate of poison injection requirement of 10 CFR 50.62(c)(4) than does the 251 inch diameter vessel on which the rule was based. The Big Rock Point reactor vessel has a diameter of only 106 inches and sodium pentaborate solution is on the order of 19 weight percent.
Injection rate with the reactor at power is 132 gpe.
It should be noted that because of unique operating features associated with plant design that the injection rates established by the rule are probably not appropriate for Big Rock Point.
It is useful to get f-a sufficient amount of poison to the core to shut the reactor down
'e quickly during a Big Rock ATWS.
The effectiveness of the poison system in providing a relatively quick shutdown of the reactor and a 4.8-5 MIO688-0302A-BX01
i l
detailed description of the poison system itself was provided as a part of our February 26, 1981 Automatic Recirculating Pump Trip risk evaluation. The system operates on naturel circulation and requires no charging pumps. A high pressure nitrogen accumulator is available to provide initial force to establish poison flow to the reactor.
With the react.or at power, sufficient sodium pentaborate can be injected to the primary coolant to result in reactor shutdown on the order of one minute following actuation. Five minutes of natural circulation injection provides a sufficient amount of poison to permit suberiticality even at cold conditions.
The Big Rock Point poison system has been designed to requirements more restrictive than this portion of the rule and therefore no l
further action is required.
4.8.6 INTEGRITY OF LPS PRESSURE ISOLATION VALVES (Reference 3) t Generic Letter 87-06, Periodic Verification of Leak Tight Integrity of Pressure Isolation Valves, dated March 13, 1987, requests that all Licensees review methods in place for assuring the leak-tight integrity of pressure isolation valves as independent barriers against abnormal leakage, rapidly propagating failure, and gross rupture of the reactor coolant pressure boundary.
Pressure isolation valves (P1Vs) are defined for each interface as any two valves in_ series within the i
reactor coolant pressure boundary which separate the high pressure Reactor Coolant System (RCS) from an attached low pressure system and are normally closed during power operation. The following addresses the LPS interface with the RCS and responds to the Generic Letter requirements in relation to the Poison System.
The LPS contains five explosive squib valves which isolate the RCS from the normally depressurized liquid poison tank.
(Refer to Drawing 0740G40107.) Although squib valvss are inherently leak tight and cannot be routinely stroked, leakage to the LPS from the RCS is monitored by a low range poison tank pressure gauge installed in the main control room. This parameter is checked and recorded in the control room log once a shift and monitored during hourly readings.
Any Reactor Coolant System pressure boundary leakage would cause a detectable pressure increase.
Consumers Power Company feels that these requirements combined with
the system design features as discussed above are adequate to detect-Reactor Coolant System pressure boundary leakage and initiate prompt mitigative actions if required.
4.8.7 LIQUID POISON SYSTEM INJECTION GAS PRESSURE The setpoint of the steam drum safety valves affects both the over-pressure protection for the primary system and the operation of the standby liquid poison system. Within this Updated FHSR,- Sections -
O 5.4.2.7 (Steam Drum Relief Valves) and 4.8.8 (Liquid Poison System-Operability) below requires that the maximum setting'of the first 4.8-6 MIO688-0302A-BX01
safety valve be 1700 psis with subsequent pressure increment settings -
of 10 poi so that the allowable pressure of 1870 psia (1700 plus 10%)
in the nuclear steam supply system is not exceeded. This is also supported by Section 4.8.2.5 of this Updated FRSR which assumes a primary system pressure of 1870 psia when describing the liquid poison system capability.
LPS Operability Requirements provides a relationship between safety valve setroint and required liquid poison systen nitrogen pressure.
4.8.8 LIQUID POISON SYSTEM PERFORMANCE REQUIREMENTS (Reference Technical Specifications) 4.8.8.1 LPS Tabulated Requirements Absorber Material-
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Sodium Pentaborate Solution Poison Tank Capacity, 850 Gallons J
Sodium Pentaborate in Solution, Wt Percent:
Minimum Concentration 19 Maximum Concentration 30 Initial Injection Oas Pressure,
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Psia Condition Pressure Head Off 500 Head on - Relief Valve Settings:-
1250 1470 1500 1800 l
1750 2080 l
System Actuation Remote Manual Type of Injection Valves Explosiv', Electrically e
Operated, Supplied by Conax Corporation
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4.8.8.2 LPS Availability, Concentration, and Surveillance i
The liquid poison system will be available for operation at all times during refueling and power operation.
The reactor will be shutdown in any situation where the poison solution tank level drops below an equivalent of 850 gallons 19 weight percent sodium pentaborate or where the poison solution storage temperature drops to_less than 5'F v
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above saturation temperature. The maximum allowable concentration l
shall be 30 weight percent of sodium pentaborate. The minimum worth i
of the liquid poison system (based on normal water level) shall be 251 Ak Components of the system will be checked at one to two monthIkke.rvals for proper operation except for actuation of the i
injection valves. The liquid poison systes shall be used at any time
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when suberiticality cannot be assured by the normal shutdown mechanism, i
Injection will be continued until a minimum shutdown margin of 0.01 nolde/kIfkrated after poison has been injected until the boron' Ak is assured in the most reactive core.
The reactor will concentration in the reactor water has been reduced to 10) ppe or
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1ess. One squib primer'and trigger 'a'is'embly from the' equalising ~1ine~
i will be removed and test-fired at least every 18 months. The'se will
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l be tested on an alternate basis ensuring valve replacement every 36 months. One squib primer and trigger assembly from the remaining five units will be removed and test-fired at least every 18 months.
These will be tested on an alternate basis.
In no case shall a squib primer and trigger assembly remain in service longer than five years.
The tests will consist of monitoring of the input firing current and j
shearing of the integral inlet cap.
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Section 4.4 keferences Reference 1.
CPCo letter dated February 26, 1981, Probabilistic Risk Assessment 3
Results Relating to the Efficacy of a Recirculation Pump Trip.
(Liquid Poison System Effectiveness) 2.
CPCo letter dated October 14, 1985, Implementation Schedule For 10 CFR 50.62, Requirements For Reduction of Risk From Anticipated i
Transients Without Scram (ATWS),
(Liquid Poison System capacity) 3.
CPCo letter dated May 15, 1987, Response to Generic Letter 87-06, Periodic Verification of Leak Tight Integrity of Pressure Isolation Valves (LPS interface with RCS)
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