1CAN032301, License Amendment Request to Modify the Arkansas Nuclear One, Unit 1, Technical Specification 3.3.1, Reactor Protection System Instrumentation, Turbine Trip Function on Low Control Oil Pressure
| ML23089A261 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 03/30/2023 |
| From: | Couture P Entergy Operations |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| 1CAN032301 | |
| Download: ML23089A261 (1) | |
Text
- ) entergy Phil Couture Sr. Manager Fleet Regulatory Assurance - Licensing 601-368-5102
March 30, 2023
ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
Subject:
License Amendment Request to Modify the Arkansas Nuclear One, Unit 1, Technical Specification 3.3.1, "Reactor Protection System Instrumentation,"
Turbine Trip Function on Low Control Oil Pressure
Arkansas Nuclear One - Unit 1 NRC Docket No. 50-313 Renewed Facility Operating License No. DPR-51
In accordance with the provisions of Title 10 of the Code of Federal Regulations (10 CFR) 50.90, "Application for Amendment of License, Construction Permit, or Early Site Permit," Entergy Operations, Inc. (Entergy) is submitting a request for an amendment to the Technical Specifications (TSs) for Arkansas Nuclear One, Unit 1 (ANO-1 ).
The proposed amendment revises ANO-1 TS 3.3.1, Table 3.3.1-1, "Reactor Protection System Instrumentation," Function 9 "Main Turbine Trip (Oil Pressure)," for ANO-1 as follows:
- The allowable value setpoint is increased from ~ 40.5 psig to ~ 700 psig and the function name is modified to specify "Hydraulic Oil Pressure."
The proposed change to the TS is due to the replacement and relocation of the pressure switches from the low pressure auto-stop (AST) oil header that operates at approximately 100 psig to the high pressure turbine electrohydraulic control (EHC) oil header that operates at approximately 1800 psig to 1900 psig. The changes to the allowable value are needed due to the higher EHC system operating pressure. Relocation of the initiating pressure switches to the high pressure turbine EHC emergency trip header (ETH) is required to accommodate a modification to the EHC turbine control system while maintaining the function of transmitting the trip signal to the reactor protection system (RPS). This change does not adversely affect any RPS trip functions.
The enclosure provides a description of the proposed changes, technical evaluation of the proposed changes, regulatory evaluation, and a discussion of environmental considerations.
Attachment 1 to the enclosure provides the existing TS pages marked-up to show the proposed changes for ANO-1. Attachment 2 to the enclosure provides the existing ANO-1 TS pages retyped to show the proposed changes. Attachment 3 to the enclosure provides the existing ANO-1 TS Bases pages marked-up to show the proposed changes. Changes to the existing TS
Entergy Operations, Inc., 1340 Echelon Parkway, Jackson, MS 39213 1CAN032301 Page 2 of 2
Bases are provided for information only and will be implemented under the Technical Specification Bases Control Program.
There are no new commitments contained in this submittal.
Approval of the proposed amendment is requested by April 30, 2024. Once approved, the amendment shall be implemented prior to startup from refueling outage 1 R31 (spring 2024 ),
coincident with the aforementioned plant modifications to be performed in 1 R31.
If there are any questions or if additional information is needed, please contact Riley Keele, Manager, Regulatory Assurance, Arkansas Nuclear One, at 479-858-7826.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on March 30, 2023.
Sincerely, Philip Digitally signed by Philip Couture Date : 2023.03.30 Couture 07 :56:05 -05'00' PC/mar
Enclosure:
Evaluation of the Proposed Change
Attachments to the
Enclosure:
- 1. Technical Specification Page Markups
- 2. Retyped Technical Specification Pages
- 3. Technical Specification Bases Page Markups (Information Only)
- 4. CALC-22-E-0007-01 "Unit 1 Main Turbine Anticipatory Reactor Trip Setpoint Basis"
cc: NRC Region IV Regional Administrator NRC Senior Resident Inspector - Arkansas Nuclear One NRC Project Manager - Arkansas Nuclear One Designated Arkansas State Official Enclosure
Evaluation of the Proposed Change 1CAN032301 Enclosure Page 1 of 18
Table of Contents
1.0
SUMMARY
DESCRIPTION................................................................................................ 2 2.0 DETAILED DESCRIPTION................................................................................................ 2 2.1 Current Technical Specification Requirements........................................................ 2 2.2 Reason for the Proposed Change............................................................................ 2 2.3 Description of the Proposed Change....................................................................... 2
3.0 TECHNICAL EVALUATION
............................................................................................... 3 3.1 System Description.................................................................................................. 3 3.2 Existing Pressure Switch Configuration................................................................... 4 3.3 New Pressure Switch Configuration......................................................................... 4 3.4 Current Licensing Basis........................................................................................... 5 3.5 Failure Mode Analysis.............................................................................................. 8 3.6 Response Time of the Existing Versus New Pressure Switch Configurations......... 9 3.7 ARTS Pressure Switch Trip Setpoint and AV Determination................................. 10 3.8 Conclusion.............................................................................................................. 12
4.0 REGULATORY EVALUATION
......................................................................................... 12 4.1 Applicable Regulatory Requirements and Criteria................................................. 12 4.2 Precedent............................................................................................................... 1 4 4.3 No Significant Hazards Consideration.................................................................... 14 4.4 Conclusion.............................................................................................................. 16 5.0 ENVIRONMENTAL EVALUATION................................................................................... 16
6.0 REFERENCES
................................................................................................................. 17 7.0 ATTACHMENTS............................................................................................................... 17
1CAN032301 Enclosure Page 2 of 18
EVALUATION OF THE PROPOSED CHANGE
1.0
SUMMARY
DESCRIPTION
The proposed amendment would modify the Arkansas Nuclear One, Unit 1 (ANO-1), Technical Specification (TS) 3.3.1, Table 3.3.1-1, "Reactor Protection System Instrumentation,"
Function 9, "Main Turbine Trip (Oil Pressure)," allowable value (AV) and specify that it is for (Hydraulic Oil Pressure). The change is necessary to support replacement of the current Westinghouse electrohydraulic control system (EHC) which requires the replacement and relocation of the Reactor Protection System (R PS) Anticipatory Reactor Trip System (ARTS) pressure switches. The pressure switches act to cause a reactor trip upon loss of the Main Turbine, limiting the subsequent rise in Reactor Coolant System (RCS) pressure upon a loss of turbine load. Although this trip function is included in the ANO-1 TSs, it is not credited in the ANO-1 accident analyses.
2.0 DETAILED DESCRIPTION
2.1 Current Technical Specification Requirements
The current RPS trip on Main Turbine oil pressure is applicable in Mode 1 when 45% reactor power, having a low turbine auto-stop oil system pressure AV of 40.5 psig. If this requirement is not met, power must be reduced to < 45% within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, in accordance with TS 3.3.1, Condition F.
2.2 Reason for the Proposed Change
As part of installation of a new Ovation Turbine Control System (TCS), the current ARTS pressure switches installed on the low pressure auto-stop oil system which operates at approximately 100 psig are being replaced and new switches installed on the high pressure EHC system which operates at a pressure of approximately 1800 - 1900 psig. The low pressure auto-stop oil system will be removed. The EHC system supplies the turbine Emergency Trip Header (ETH) which acts to trip the turbine prior to losing controllability of the Main Turbine steam supply valves. While the new pressure switches perform the same trip function and in the same manner, the AV associated with the ARTS must be increased to accommodate the higher EHC oil pressure.
2.3 Description of the Proposed Change
Entergy Operations, Inc. (Entergy) requests ANO-1 TS 3.3.1, Table 3.3.1-1, Function 9, "Main Turbine Trip (Oil Pressure) be revised as follows:
- Changes the Function 9 description from "Main Turbine Trip (Control Oil)" to "Main Turbine Trip (Hydraulic Oil Pressure)"
to the enclosure provides the existing TS pages marked-up to show the proposed changes for ANO-1. Attachment 2 to the enclosure provides the existing ANO-1 TS pages retyped to show the proposed changes. Attachment 3 to the enclosure provides the existing ANO-1 TS Bases pages marked-up to show the proposed changes. Changes to the existing TS 1CAN032301 Enclosure Page 3 of 18
Bases are provided for information only and will be implemented under the Technical Specification Bases Control Program.
3.0 TECHNICAL EVALUATION
3.1 System Description
The Main Turbine oil pressure trip function anticipates the loss of heat removal capabilities of the secondary system following a turbine trip. This trip function acts to minimize the pressure and temperature transient on the reactor and RCS. A turbine trip from a power level below 45%
power will not directly initiate a reactor trip. Four pressure switches (one per each RPS channel) monitor the ETH pressure in the Main Turbine EHC system. A low pressure condition of 850 psig sensed by two-out-of-four pressure switches with reactor power 45% rated thermal power (RTP) will initiate a reactor trip. These pressure switches do not provide any input to the non-nuclear instruments or integrated control system.
The EHC system supplies hydraulic control oil fluid to the turbine throttle valves, governor valves, reheat stop valves, and reheat intercept valves. The existing turbine protection system consists of the auto-stop oil system and the EHC system. Auto-stop oil is composed of the same oil used for lubrication of the main turbine and provides a trip interface between critical turbine functions and the EHC system. On a turbine trip signal, the auto-stop oil system line is depressurized by the actuation of protective devi ces, solenoid trip valves, or an emergency trip valve when a turbine trip condition exists.
The EHC fluid is an incompressible fluid and when the solenoid/emergency trip valves are opened, the dump valves at each turbine governor valve, throttle valve, reheat stop valve, and reheat intercept valve are depressurized due to the high-pressure EHC fluid to these valves' actuators being released to drain back to the EHC fluid reservoir. The new Ovation TCS uses a separate set of three pressure transmitters to sense ETH pressure and initiate a turbine trip if EHC pressure decreases to 1000 psig. Two Testable Dump Manifolds (TDMs) replace the previously installed protective device trip interfaces. The trip signals will be sent to the TDMs which open to depressurize the EHC system. The governor, throttle, and reheat stop and intercept valves are spring actuated closed such that when the high pressure electrohydraulic (EH) fluid is removed from the valve actuators, the valves close instantaneously.
The EHC fluid is provided by skid-mounted hydraulic pumps that maintain operating pressure at approximately 1800 - 1900 psig. The EHC fluid pressure is sensed by the new ARTS pressure switches. When the decreasing ETH pressure reaches 850 psig as sensed by the pressure switches at 45% RTP, a reactor trip signal is initiated by at least two-out-of-four RPS channels.
The circuitry and logic associated with the ARTS pressure switches and RPS operates in the same manner as the current auto-stop oil pressure switches, and it remains independent of the Ovation TCS.
The reactor trip on turbine trip is an anticipatory trip input signal to the RPS. This trip is anticipatory in that it is not assumed to occur in any of the Safety Analysis Report (SAR)
Chapter 6 or Chapter 14 accident analyses. This trip meets the requirements of IEEE 279-1971, "Criteria for Nuclear Power Plant Protection Systems" (Reference 1), including separation, redundancy, single failure, and testability. The ARTS pressure switches are treated as safety-related and seismically qualified, although the location in the non-seismic Turbine Building does not permit full qualification.
1CAN032301 Enclosure Page 4 of 18
3.2 Existing Pressure Switch Configuration
The existing four ARTS pressure switches are located on the low pressure auto-stop oil system with output contacts serving as inputs to each of the four RPS instrument channels to indicate Main Turbine status to the RPS in a two-out-of-four channel trip logic (internal to the RPS). This signal initiates a reactor trip on a turbine trip if reactor power is 45% RTP on the power range nuclear instruments. The low pressure auto-stop oil header operates at a pressure of approximately 100 psig with the turbine latched. The current TS AV is 40.5 psig.
3.3 New Pressure Switch Configuration
The modification to the EHC system removes the low pressure auto-stop oil system, including the existing ARTS low auto-stop oil pressure switches. To support this modification, the RPS trip function will now be performed by four new pressure switches located on the Main Turbine ETH. Connected to the ETH is a Testable Dump Manifold (TDM) which depressurizes the ETH upon turbine trip. A separate TDM is installed on the Overspeed Protection Control (OPC) header which serves as a backup for turbine protection. As with the original pressure switches, each of the four new pressure switches has an output contact that provides an input to its corresponding RPS instrumentation channel via a contact buffer located in the RPS channel's cabinet. The RPS logic is not affected by the proposed change. The new ARTS pressure switches will continue to initiate a reactor trip on a turbine trip when reactor power is 45% RTP as sensed on the power range nuclear instruments (NIs).
The proposed change to the TSs is necessary due to the ARTS pressure switches being installed on the ETH, which is maintained at a much higher operating pressure than low pressure auto-stop oil. Latching and operation of the Main Turbine requires the control valves to be open (or capable of being opened/throttled), which is dependent on maintaining proper EHC supply header and ETH pressure.
As stated previously, the pressure switches are located in the ANO-1 Turbine Building, which is not seismically rated and functionally non-safety related; however, the switches are treated as safety-related and seismically mounted for conservatism and robustness, although used in a non-safety related application.
Each pressure switch will be calibrated to a setpoint of 850 psig with decreasing pressure. At the trip setpoint, contact outputs from the pressure switches will open to signal a Main Turbine trip to the RPS. Class-1E contact buffers in the RPS will sense the opening of the contact inputs and initiate an RPS trip. The trip response time for the downstream RPS logic remains unchanged at 150 milliseconds.
The new SOR Incorporated, Measurement and Control (Model 9N6-BB45-U1-C1A-JJTTNQ) pressure switches are rated for the type of EH fluid utilized by ANO-1 (Fyrquel 220 or equivalent) with a safe working pressure of 2500 psig. SOR pressure switches are rugged, field-mounted instruments. The pressure sensing element of the SOR pressure switch is a force-balance, piston-actuated assembly, and is sealed by a flexible diaphragm and a static O-ring. Each switch contains a dual pole, dual throw dry contact that is monitored by a contact buffer in its corresponding RPS cabinet.
1CAN032301 Enclosure Page 5 of 18
3.4 Current Licensing Basis
3.4.1 Safety Analysis Report (SAR)
The following SAR excerpts provide background and licensing basis information related to turbine/reactor control and the ANO-1 ARTS.
Section 4.1.1.2, "Transient Performance," states:
The RCS will follow step or ramp load changes under automatic control without relief valve or turbine bypass valve action as follows:
A. Step load changes - increasing load steps of 10 percent of full power in the range between approximately 22 percent and 90 percent full power and decreasing load steps of 10 percent of full power between 100 percent and approximately 22 percent full power.
B. Ramp load changes - increasing load ramps of 10 percent per minute in the range between approximately 22 percent and 90 percent full power are acceptable, or decreasing load ramps of 10 percent per minute from 100 percent to approximately 22 percent full power. From 90 percent to 100 percent full power, increasing ramp load changes of five percent per minute are acceptable.
The combined actions of the control system, the turbine bypass valves, and the main steam safety valves are designed to accept separation of the generator from the transmission system without reactor trip. However, this original design included a RPS high RCS pressure trip setpoint above the Pressurizer Electromatic Relief Valve (ERV) setpoint. As a result of the Three Mile Island, Unit 2 (TMI-2) event in 1979, the NRC ordered a reversal of these setpoints such that a high RCS pressure trip is reached before the ERV is challenged. This arrangement effectively eliminated the unit's ability to survive a load rejection from full power without a reactor trip unless operator action is taken.
Also as a result of the TMI-2 event, the NRC ordered the installation of the Anticipatory Reactor Trip System (ARTS). The ARTS will reduce the energy input to the RCS by a prompt trip upon loss of the OTSG [Once-Through Steam Generator] secondary heat sink.
This system trips the reactor upon either loss of main feedwater or turbine trip, actuating about eight seconds sooner than the RCS high pressure trip. The ARTS is redundant, safety grade, and meets the requirements of IE [Inspection and Enforcement] Bulletin 79-05B and NUREG-0578. The ARTS hydraulic pressure sensing switches, however, are located near the main turbine and main feedwater pumps in the non-seismic (non type-I) turbine building.
Section 7.1.2.2.3, "Description of Protection Channel Functions," describes the reactor trip on a turbine trip function as follows:
Each of the four protection channels receives input based on the status of the turbine (tripped/not tripped) and both main feedwater pumps (tripped/not tripped). The information is supplied by pressure switches which monitor the hydraulic control oil pressure for the main turbine and the main feed pump turbines. The main turbine or both main feed pump turbines must trip to initiate an ARTS trip.
1CAN032301 Enclosure Page 6 of 18
A contact buffer in each protection channel continuously monitors the state of the associated pressure switches. When the state of the pressure switch changes to a tripped condition, the contact buffer de-energizes the protection channel's trip relay.
Operating bypasses are provided to allow normal startup and shutdown of the plant.
Automatic removal of the bypasses occurs based on increasing flux level (greater than 45 percent full power for turbine trip and greater than 10 percent full power for loss of both main feedwater pumps).
The anticipatory trips were added to meet the requirements of the Three Mile Island, Unit 2 (TMI-2) Action Plan, NUREG-0737. The purpose of these trips is to provide a reactor trip signal in those cases where a loss of secondary heat sink would likely challenge the RPS (on some other parameter) or the Pressurizer Electromatic Relief Valve (ERV). The anticipatory trips limit the heat input to the system after a loss of heat sink, reducing the amount of heat that must be removed after the trip.
Section 7.2.3.3.4, "Loss of Load Considerations," states:
The nuclear unit is designed to accept a 10 percent step load rejection without safety valve action or turbine bypass valve action. The controls will prevent steam dump to the condenser when condenser vacuum is inadequate, in which case the safety valves may operate. The features that permit continued operation under load rejection conditions include:
A. Integrated Control System
During normal operation, the Integrated Control System (ICS) controls the unit load in response to load demand from the operator. During normal load changes and small frequency changes, turbine control is through the speed changer to maintain constant steam pressure. During large load and frequency upsets, the turbine governor takes control to regulate frequency.
B. 100 Percent Relief Capacity in the Steam System
This provision acts to reduce the effect of large load drops on the reactor system.
Consider, for example, a sudden load rejection greater than 10 percent. When the turbine generator starts accelerating, the governor valves and the intercept valves begin to close to maintain set frequency. In addition, the megawatt demand signal is reduced, which reduces the governor speed changer setting, feedwater flow demand, and reactor power level demand. As the governor valves close, the steam pressure rises and acts through the control system to reinforce the feedwater flow demand reduction already initiated by the reduced megawatt demand signal. In addition, when the load rejection is of sufficient magnitude, the turbine bypass valves open to reject excess steam to the condenser and the safety valves open to exhaust steam to the atmosphere. The rise in steam pressure and the reduction in feedwater flow cause the average reactor coolant temperature to rise which reinforces the reactor power level demand reduction, already est ablished by reduced megawatt demand, to restore reactor coolant temperature to the set value.
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As the turbine generator returns to set frequency, the turbine controls revert to steam pressure control rather than frequency control. This feature holds steam pressure within relatively narrow limits and prevents further large steam pressure changes.
Section 10.1, "Steam and Power Conversion System," describes the reactor trip on turbine trip function as follows:
The unit is designed to maintain station auxiliary load without a reactor trip upon loss of full load due to faults occurring on the system side of the 500 kV switchyard breakers. This loss of full load assumes the condition results in the opening of the main generator output breakers B5114 and B5118. This may be caused by protective features that result in a turbine/generator lockout or protective features in the switchyard that result in the opening of the generator output breakers. As described in SAR Sections 7.2.3.3.4, 'Loss-of-Load Considerations' and 14.1.2.8.1, 'Loss of Electric Power - Identification of Cause', [assuming plant power is less than the ARTS trip setpoint for Reactor Trip on Loss of Main Turbine] the ICS will attempt to maintain the unit on-line during a loss of load event which does not involve a turbine/generator lockout by initiating a runback to ~15% full power (note that a generator lockout also locks out the main turbine, defeating the need for a runback). However, the runback may not be successful in all cases and is not credited in the accident analyses. The steam bypass to the condenser and atmospheric dump valves will be utilized as necessary to achieve this load reduction.
Section 14.1.2.8.1, "Loss of Electric Power - Identification of Cause":
The unit is designed to withstand the effects of loss of electric load or electric power.
Emergency power systems are described in Chapter 8. Two types of power losses are considered:
A. A loss of load condition caused by separation of the unit from the transmission system.
B. A hypothetical condition resulting in a complete loss of all system and unit power except the unit batteries.
Section 14.1.2.8.3, "Results of Loss-of-Load Conditions Analysis," states:
The unit has been designed to accommodate a loss-of-load condition without a reactor or turbine trip.
Note that if generator output breakers B5114 and B5118 are opened by the primary tripping mechanism (286-G1-3 lockout relay), then the unit cannot maintain station auxiliary load without reactor trip because relay 286-G1-3 also trips the generator field breaker and generator lockout relays 286-G1-1 and 286-G1-2, which in turn trip the turbine lockout relay 286/T. In such an event, the plant will respond similar to the response following a turbine trip.
It is possible for the generator output breakers to trip based on backup protective schemes such as protective relaying in the switchyard, without directly causing a generator lockout.
Under such circumstances, a runback signal causes an automatic power reduction to 1CAN032301 Enclosure Page 8 of 18
15 percent reactor power. The runback may not be successful if the reactor high pressure setpoint is reached, at which time a reactor trip would occur.
The loss-of-load accident does not result in fuel damage or excessive pressures on the RCS.
There is no resultant radiological hazard to station operating personnel or to the public from this accident, since only secondary system steam is discharged to the atmosphere.
3.4.2 TS Bases
The discussion of the Main Turbine auto-stop oil pressure function is further described in the ANO-1 TS Bases B3.3.1.
Main Turbine Trip (Oil Pressure)
The Main Turbine Trip Function trips the reactor when the main turbine is tripped at high power levels. The Main Turbine Trip Function provides an early reactor trip in anticipation of the loss of heat sink associated with a turbine trip. The Main Turbine Trip Function was added to the Babcock and Wilcox (B&W) designed units in accordance with NUREG-0737 following the Three Mile Island Unit 2 accident. The trip lowers the probability of an RCS electromatic relief valve (ERV) actuation for turbine trip cases.
Each of the four turbine oil pressure switches feeds one of the four protection channels through a buffer that continuously monitors the status of the contacts. Therefore, failure of any pressure switch affects only one protection channel.
For the Main Turbine Trip (Oil Pressure) bistable, the Allowable Value of 40.5 psig is selected to provide a trip whenever main turbine oil pressure drops below the normal operating range. The reactor power bypass is designed to automatically remove the turbine oil pressure trip function from the bypassed condition at < 45% RTP. Alarms are available to alert operators when the bypass function is enabled. Should the automatic bypass removal function fail such that the channel remains in the bypassed state, the channel must be considered inoperable at power levels of 45% RTP and the appropriate condition is entered. Failure of the automatic bypass removal feature alone or the inability to place the channel in a bypassed state when < 45% RTP does not constitute channel inoperability. The automatic bypass removal feature is tested to ensure its continued availability during the monthly CHANNEL FUNCTIONAL TEST. The turbine trip is not required to protect against events that can create a harsh environment in the turbine building. Therefore, errors induced by harsh environments are not included in the determination of the setpoint Allowable Value.
The unit is designed to withstand a complete loss of load and not sustain core damage or challenge the RCS pressure limitations. Core protection is provided by the RPS RCS High Pressure Trip Function and RCS integrity is ensured by the Pressurizer code safety valves and operation of the ERV.
The LCO requires four channels of Main Turbine Trip (Oil Pressure) to be operable with reactor power 45% RTP. Below 45% RTP, a turbine trip does not actuate a reactor trip. In Modes 2, 3, 4, 5, or 6, there is no potential for a turbine trip, and the Main Turbine Trip (Oil Pressure) RPS trip function is not required to be operable.
1CAN032301 Enclosure Page 9 of 18
3.5 Failure Mode Analysis
The reactor trip on turbine trip function is not credited in the accident analysis. The new switches are designed for consistent, dependable operation at the higher ETH fluid oil pressure.
The tubing connecting the switches to the EHC header is stainless steel and is capable of withstanding the higher EHC system pressure. Postulated pipe breaks in the EHC header do not need to be considered in the design, as no safety-related equipment would be adversely impacted. A break in the piping would result in depressurization of the EHC system, closure of the associated turbine valves, and actuation of the ARTS pressure switches.
As stated, the pressure switches are designed with dry contacts. The Class-1E contact buffers within the RPS sense the opening of the respective pressure switch contact and are not modified by installation of the new ARTS pressure switches. The associated RPS logic is de-energize to actuate; therefore, a loss of power to a contact buffer or RPS channel will result in a trip of that channel. Inadvertent actuation is minimized since two of the four RPS channels must de-energize to initiate a reactor trip due to turbine trip (i.e., two RPS channels must lose power to initiate a trip). Likewise, only two of the four pressure switch contacts must open to initiate a reactor trip within the RPS; therefore, single failure criterion is met with respect to the design of the ARTS.
3.6 Response Time of the Existing Versus New Pressure Switch Configurations
Because the reactor trip on turbine trip function is not credited in the accident analysis, the ANO-1 SAR does not impose any response time requirements for the initiation of this trip (reference SAR Table 7-13, "Relationship Between Tested Trip Response Times and Trip Response Times Assumed in The Safety Analysis").
With the existing EHC configuration, the auto-stop oil system line is depressurized by the actuation of protective devices, solenoid trip valves, or an emergency trip valve on a turbine trip condition. The EHC fluid is an incompressible fluid and when the solenoid valves, emergency trip valves, or interface valve are opened during a turbine trip, the dump valves at each main steam governor and stop valve are depressurized and the high pressure EHC fluid to the Main Turbine governor valves, throttle valves, reheat stop valves, and reheat intercept valves actuators is released to drain.
With the new EHC TCS configuration, a solenoid valve trip block assembly is connected to the high pressure ETH. On a trip condition, the solenoid valves are de-energized and open to depressurize the dump valves that release the high pressure EH fluid to the Main Turbine throttle, governor, reheat stop, and reheat intercept valves actuators to drain back to the EHC reservoir (to approximately zero psig). By removing the low pressure auto-stop oil system and reconfiguring the ARTS pressure switches to directly sense the same EH fluid pressure used to maintain the steam valves open, the time required to sense a turbine trip will not be significantly affected. The nearly instantaneous depressurization of the EHC system when the EHC dump valves open, along with the 150 millisecond RPS response time, does not adversely affect the 8-second margin (described in SAR Section 4.1.1.2) between the ARTS reactor trip based on turbine trip and a reactor trip due to reaching the high RCS pressure trip setpoint due to the loss of turbine load.
Like the current monitoring system, the Ovation TCS receives several inputs related to the operation of secondary plant that indicate whether a trip of the Main Turbine should be initiated.
1CAN032301 Enclosure Page 10 of 18
If a trip condition is detected, the TCS will act to open the aforementioned EHC dump valves, rapidly depressurizing the EHC system such that the 850 psig setpoint of the ARTS pressure switches will be reached almost instantly. In turn, the RPS buffering contacts immediately detect activation of the ARTS pressure switches and initiate a reactor trip.
If a leak occurs on the EHC system such that EHC pressure decreases slowly, the TCS will perform the same response upon two of its three EHC pressure transmitters detecting a decreasing EHC pressure of 1000 psig. The most limiting turbine steam supply valve will begin to drift closed at 1024 psig (full closure at 480 psig) based on maintenance testing. With the EHC dump valves receiving a signal to open at 1 000 psig resulting in rapid depressurization of the EHC header, the 850 psig ARTS pressure switch setpoint is sufficient to ensure a reactor trip on turbine trip prior to losing controllability of the turbine steam supply valves even during a slow loss of EHC pressure.
In summary, the new response time to depressurize the ETH to less than the ARTS pressure switch actuation setpoint will be similar, if not faster than the existing configuration. Based on the above discussion, there will be no discernable difference in the time to initiate a reactor trip on a turbine trip (reactor power 45%) with the new TCS and the associated ARTS pressure switch configuration.
3.7 ARTS Pressure Switch Trip Setpoint and AV Determination
The setpoint of the new pressure switches is evaluated to be 850 psig (decreasing) based on the Ovation TCS's separate electronically sensed ETH setpoint of 1000 psig (representing a turbine trip). The total error associated with the ARTS pressure switches is conservatively determined to be +/- 107.00 psi; therefore, the TS 3.3.1, Table 3.3.1-1, Function 9 AV is established at 700 psig to allow for calibration tolerances, instrument uncertainties, instrument drift, and environmental errors. The difference in time between reaching a trip setpoint of 850 psig or the AV of 700 psig once the EHC dump valves open is evaluated as insignificant since depressurization occurs at near the speed of sound.
As stated previously, the ARTS is not credited in the accident analyses and, therefore, the related AV for the subject pressure switches does not protect a safety limit or act to prevent exceeding specified acceptable fuel design limits (SAFDLs) during design basis accidents (DBAs). The purpose of the switches is to actuate a reactor trip in response to a turbine trip event, not as a direct result of an accident such as a loss of coolant accident (LOCA) or a main steam line break (MSLB). Therefore, the control oil pressure setpoint is not a limiting setpoint used to protect a design or licensing basis limiting condition. The AV is derived from an evaluation of total loop uncertainties applied to a nominal setpoint value (850 psig). A summary description supporting the determination of the aforementioned pressure switch 850 psig setpoint and 700 psig AV follows.
In order to ensure that an instrument channel is capable of performing its specified function, ANO-1 performs testing of these instruments in accordance with current station procedures that govern the control of calibration requirements (including as-found and as-left tolerances), and the evaluation of out-of-tolerance instruments. Calibration accuracy is defined in the applicable instrument design output documents.
ANO-1 has not adopted TSTF-493, Revision 4, "Clarify Application of Setpoint Methodology for Limiting Safety System Settings (LSSS) Functions," and has no plans to adopt the TSTF in the future. The proposed change in the allowable value is based on the ANO-1 setpoint calculation methodology for the specific pressure switches and the hydraulic analysis performed as part of 1CAN032301 Enclosure Page 11 of 18
the TCS modification, which are consistent with the methods described in Regulatory Issue Summary (RIS) 2006-17, "NRC Staff Position on the Requirements of 10 CFR 50.36, 'Technical Specifications,' Regarding Limiting Safety System Settings During Periodic Testing and Calibration of Instrument Channels," (Reference 2) and Regulatory Guide (RG) 1.105, "Setpoints for Safety-Related Instrumentation" (Reference 7). Conformance with the issues described in RIS 2006-17 by ANO calculation CALC-22-E-007-01, "Unit 1 Main Turbine Anticipatory Reactor Trip Setpoint Basis" (Attachment 4) is addressed by adherence to Entergy procedure EN-DC-200, "I&C Uncertainties / Setpoint Calculations & Determinations" and ANO Design Guide IDG-001 "Instrument Loop Error Analysis and Setpoint Methodology Manual."
CALC-22-E-007-01 determined the following with respect to the Main Turbine ARTS setpoint:
The total error associated with the Emergency Trip Header Low Pressure Setpoint (TOT 1000) is as follows:
TOT1000 = +/- (Total Abnormal Transmitter Error2 + Total Abnormal Module Error2)0.5 TOT1000 = +/- 19.24 psi
The total error associated with the Main Turbine ART Setpoint (TOT 850) is as follows:
TOT850 = +/- (Total Abnormal Switch Error)
TOT850 = 46.00 psi or -106.64 psi TOT850 = +/- 107.00 psi
Note: TOT850 is conservatively rounded up to +/-107.00 psi to encompass the total error interval.
As can be seen in Figures 1 and 2 on the following page, there is no overlap between the lowest value associated with the TCS setpoint and the highest potential value of the ARTS setpoint.
The as-left tolerance (AL TOL) and as-found tolerance (AF TOL) of the pressure switch loop for the Main Turbine ARTS setpoint are determined to be as follows (Reference 8):
ALTOL = +/- 15.5 psi AFTOL = +/- 74.3 psi 1CAN032301 Enclosure Page 12 of 18
The as-left tolerance establishes the required accuracy band in which the switch must be calibrated and remain to avoid recalibration when periodically tested. The as-found tolerance, which includes the manufacturer drift specifcation for 18 months (+25% margin) or 22.5 months, establishes the limit of error the switch can be found to have during surveillance testing and still be considered within calibration and operable.
Based on the above, the ARTS pressure switch trip setpoint of 850 psig is sufficiently conservative with respect to the proposed AV of 700 psig whether applying the total pressure switch error or whether the pressure switch is operating at the most extreme as-found value.
Figure 1 EH System Pressure Figure 2 Summary Diagram 2000 psig 1000 psig TCS Turbine Trip Total Error
+/-19.24 psi
Normal 1800 psig Operating 950 psig Pressure Total Error
+/-107.00 psi
As-Found 900 psig Tolerance:
+/- 74.3 psi 1600 psig
TCS EH Supply Low 850 psig ARTS Reactor As-Left Major Trouble Alarm Trip Tolerance:
+/- 15.5 psi 1400 psig Auxiliary EH Pump Auto Start 800 psig
1200 psig TCS EH Trip Header Low Major Trouble Alarm 750 psig
1000 psig TCS Turbine Trip Total Error +/-19.24 psi 700 psig Tech Spec RPS Allowable Value Margin 23.76 psi 850 psig ARTS Reactor Trip See Fig. 2 Total Error +/-107.00 psi for detail
700 psig Tech Spec RPS Allowable Value 600 psig
3.8 Conclusion
The proposed change revises TS 3.3.1, Table 3.3.1-1, "Reactor Trip System Instrumentation,"
Function 9, "Main Turbine Trip (Oil Pressure)," to increase the AV. The proposed change to the TS AV is consistent with NUREG-1430, "Standard Technical Specifications Babcock and Wilcox Plants," Revision 5 (Reference 3). The proposed TS change does not impact or change any assumptions contained in the ANO-1 accident analyses. Therefore, the margin of safety related to the ability of the fission product barriers and accident mitigating systems to perform the respective specified safety functions during and following accident conditions is not affected.
The ARTS and related reactor trip on loss of turbine load is not credited in any accident 1CAN032301 Enclosure Page 13 of 18
analyses; therefore, the proposed change does not adversely impact maintenance of safety limits.
4.0 REGULATORY EVALUATION
4.1 Applicable Regulatory Requirements and Criteria
4.1.1 Regulations
10 CFR 50.36 sets forth the regulatory requirements for the content of the TSs. This regulation requires, in part, that the TS contain Surveillance Requirements (SRs). 10 CFR 50.36(c)(3),
states that SRs to be included in the TSs are those relating to test, calibration, or inspection, which assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the TS Limiting Condition for Operation (LCO) will be met.
10 CFR 50.65 sets forth the regulatory requirements for monitoring the effectiveness of maintenance at nuclear power plants. The existing turbine control system including the pressure switches that initiate the reactor trip on a turbine trip are within the scope of the Maintenance Rule and monitored at the plant level. The new Ovation TCS and associated pressure switches that provide a reactor scram function on a turbine trip will be monitored the same way. The RPS is also within the scope of the rule and monitored accordingly. The proposed TS change will have no effect on the monitoring of RPS-related features within the ANO-1 Maintenance Rule program.
4.1.2 General Design Criteria (GDC)
ANO-1 was not licensed to the 10 CFR 50, Appendix A, GDC. ANO-1 was originally designed to comply with the 70 "Proposed General Design Criteria for Nuclear Power Plant Construction Permits," published in July 1967. However, the ANO-1 SAR provides a comparison with the Atomic Energy Commission (AEC) GDC published as Appendix A to 10 CFR 50 in 1971. The applicable AEC GDC were compared to the 10 CFR 50, Appendix A, GDC as discussed in Section 1.4 of the ANO-1 SAR. The ANO-1 SAR contains these GDC followed by a discussion of the design features and procedures that meet the intent of the criteria. The relevant GDC are described below.
Criterion 20 - Protection System Functions
The protection system shall be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety.
Criterion 20 is applicable to this amendment request because the input into the RPS must ensure RPS actuation even if the input component fails. The normal operational state of the existing auto-stop control oil pressure switch is contacts closed. The contacts open when the control oil pressure drops below the setpoint. If the pressure switch fails, the contacts would open and therefore provide input to the associated RPS channel. The new low EHC trip header pressure switches are configured in the same manner as the existing auto-stop control oil pressure switches with contacts closed when EHC header pressure rises above the reset setpoint and contacts open when EHC header pressure drops below the trip setpoint. Pressure 1CAN032301 Enclosure Page 14 of 18
switch failure would result in contact opening and input provided to the associated RPS channel in the same manner as an EHC header pressure drop below the trip setpoint. Criterion 20 is met because the new EHC fluid oil header pressure switches are designed to fail into a safe state. Further conformance with GDC 20 is described in Section 1.4.16 of the ANO-1 SAR.
Criterion 22 - Protection System Independence
The protection system shall be designed to assure that the effects of natural phenomena, and of normal operating, maintenance, testing, and postulated accident conditions on redundant channels do not result in loss of the protection function, or shall be demonstrated to be acceptable on some other defined basis. Design techniques, such as functional diversity or diversity in component design and principles of operation, shall be used to the extent practical to prevent loss of the protection function.
Criterion 22 is applicable to this amendment request because the input into the RPS must ensure RPS channel separation is maintained to provide protection system independence. The low pressure auto-stop oil header pressure switches provide inputs to each of the four RPS protection instruments (two-out-of-four logic) to initiate a reactor trip on a turbine trip if reactor power is 45% as indicated by the power range neutron flux instruments. When a low EHC oil pressure condition is sensed that is below the setpoint by two-out-of-four pressure switches, the RPS initiates a reactor trip signal. The RPS trip function will now be performed by four new pressure switches located on the high pressure EHC trip header. As with the original pressure switches, the four new pressure switches have output contacts that provide redundant inputs to each of the four RPS protection instrument channels (two-out-of-four logic). The RPS logic is not affected by the change and the signal will still initiate a reactor trip on a turbine trip if reactor power is 45% as indicated on the power range neutron flux instruments. Separation between the pressure switches for each RPS channel and associated wiring is maintained in accordance with Institute and Electrical and Electronics Engineers (IEEE) 279-1971 (Reference 1) and ensures independence between the RPS channels. Further conformance with GDC 22 is described in Section 1.4.18 of the ANO-1 SAR.
Criterion 23 - Protection System Failure Modes
The protection system shall be designed to fail into a safe state or into a state demonstrated to be acceptable on some other defined basis if conditions such as disconnection of the system, loss of energy (e.g., electric power, instrument air), or postulated adverse environments (e.g.,
extreme heat or cold, fire, pressure, steam, water, and radiation) are experienced. Criterion 23 is applicable to this modification to the extent that inputs to the RPS are affected. The RPS and the turbine control systems are independent. A failure of the turbine control system does not affect the input to the RPS from the existing auto-stop oil low pressure switches. The low pressure auto-stop oil header pressure switches provide inputs to each of the four RPS protection instrument channels (two-out-of-four logic) to initiate a reactor trip on a turbine trip if reactor power is 45% RTP.
The RPS trip function will now be performed by four new pressure switches located on the high pressure EH trip header. The four new pressure switches have output contacts that provide inputs to each of the four RPS protection instrument channels (two of four logic) via contact buffers. The RPS logic is not affected by the change and the signal will still initiate a reactor trip on a turbine trip if reactor power is 45% RTP. The new switches were recommended by the turbine manufacturer (Westinghouse) for use in this application, and the supplier (SOR) 1CAN032301 Enclosure Page 15 of 18
provides many nuclear quality components to the industry and has a reliable operating history.
The new pressure switches do not provide any control inputs into the new Ovation TCS. The new control system trips the main turbine if EHC pressure drops to 1000 psig using separate instruments than the ARTS pressure switches. The control oil header pressure switches utilize the existing contact buffers to communicate with the RPS. The connection to the RPS from the contact buffers is not modified. Criterion 23 is met because the relocation and replacement of the pressure switches maintains system reliability, redundancy, and independence from the turbine control system. Further conformance with GDC 23 is described in Section 1.4.19 of the ANO-1 SAR.
4.2 Precedent
The following precedents are related to the proposed TS change in this submittal. These plants performed similar turbine control modifications that relocated the reactor trip on turbine trip pressure switches from a low pressure oil system to a high pressure oil system.
- The reactor trip on turbine trip setpoint was changed at Watts Bar Nuclear Plant, Unit 2, from 45 psig to 800 psig and the AV increased from 38.3 psig to 710 psig (Reference 4).
- The reactor trip on turbine trip setpoint was changed at Watts Bar Nuclear Plant, Unit 1, from 45 psig to 800 psig and the AV increased from 43 psig to 710 psig (Reference 5).
- The reactor trip on turbine trip setpoint was changed at H. B. Robinson Steam Electric Plant, Unit No. 2, from 45 psig to 800 psig and the AV increased from 40.87 psig to 769 psig (Reference 6).
4.3 No Significant Hazards Consideration
Entergy Operations, Inc. (Entergy) is proposing an amendment to revise the Arkansas Nuclear One, Unit 1 (ANO-1), Technical Specifications (TSs) to revise TS 3.3.1, Table 3.3.1-1, "Reactor Protection System Instrumentation," Function 9, "Main Turbine Trip - Oil Pressure." The proposed amendment revises the Allowable Value (AV) for this function from 40.5 psig to 700 psig. The proposed change is due to the replacement and relocation of the pressure switches from the auto-stop oil system that operates at approximately 100 psig to the high pressure turbine electrohydraulic control (EHC) trip header that operates at approximately 1800-1900 psig. The change to the AV is needed due to the higher EHC system operating pressure. Relocation of the initiating pressure switches to the high pressure turbine EHC trip header is necessary to accommodate the modification of the EHC Turbine Control System (TCS) while maintaining the function of transmitting the trip signal to the Reactor Protection System (RPS).
Entergy has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:
- 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
1CAN032301 Enclosure Page 16 of 18
Response: No
The proposed change reflects a design change to the TCS that results in the monitoring of a higher pressure oil system to detect a turbine trip, necessitating a change to the value at which a control oil pressure initiates a reactor trip on turbine trip. The control oil pressure is an input to the reactor trip instrumentation in response to a turbine trip event. The value at which the control oil initiates a reactor trip is not an accident initiator. A change in the nominal control oil pressure does not introduce any mechanisms that would increase the probability of an accident previously analyzed. The reactor trip on turbine trip function is initiated by the same protective signal as used for the existing low auto-stop oil pressure trip signal. There is no change in form or function of this signal and the probability or consequences of previously analyzed accidents are not impacted.
The existing test requirements for the low oil pressure TS instrument function related to those variables ensures that the instruments will function as required to initiate protective systems or actuate mitigating systems at the new setpoints derived in the setpoint calculation. Surveillance tests are not an initiator to any accident previously evaluated. As a result, the probability of any accident previously evaluated is not significantly increased.
The systems and components required by the control oil pressure TS instrument function for which surveillance tests are relevant are still required to be operable, meet the acceptance criteria for the surveillance requirements, and are capable of performing any mitigation function. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No
The EHC fluid oil pressure rapidly decreases in response to a turbine trip. The value at which the control oil pressure switches initiates a reactor trip is not an accident initiator.
The proposed TS change reflects the higher pressure that will be sensed after the pressure switches are relocated from the auto-stop oil system to the EHC emergency trip header.
Failure of the new switches would not result in a different outcome than is considered in the current design basis. Further, the change does not alter assumptions made in the safety analysis. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
- 3. Does the change involve a significant reduction in a margin of safety?
Response: No
The change involves a parameter that initiates an anticipatory reactor trip following a turbine trip. The accident analyses do not credit this anticipatory trip for protection of the fuel, reactor vessel, Reactor Coolant System pressure boundary, or the Containment Building. The original pressure switch configuration and the new pressure switch configuration both generate the same reactor trip signal. The difference is that the initiation of the trip will now be adjusted to a different system of higher pressure. This system function of sensing and transmitting a reactor trip signal on turbine trip remains the same.
The existing test requirements for the control oil pressure TS instrument function related to 1CAN032301 Enclosure Page 17 of 18
those variables ensures that the instruments will function as required to initiate protective systems at the new setpoints derived in the setpoint calculation. The as-left tolerance requirements of the calibration procedures are appropriately calculated to ensure that the tolerances will not have an adverse effect on equipment operability. The testing methods and acceptance criteria for systems, structures, and components, specified in applicable codes and standards (or alternatives approved for use by the NRC) will continue to be met as described in the plant licensing basis including the ANO-1 Safety Analysis Report.
There is no impact to safety analysis acceptance criteria as described in the plant licensing basis because no change is made to the accident analysis assumptions. Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Based on the above, Entergy concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92 (c), and accordingly, a finding of "no significant hazards consideration" is justified.
4.4 Conclusions
In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
5.0 ENVIRONMENTAL CONSIDERATION
A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration; (ii) a significant change in the types or significant increases in the amounts of any effluents that may be released offsite; or (iii) result in a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment needs to be prepared in connection with the issuance of the amendment.
6.0 REFERENCES
- 1. Institute of Electrical and Electronics Engineers (IEEE) 279-1971, "Criteria for Nuclear Power Plant Protection Systems," June 12, 1971
- 2. Regulatory Issue Summary (RIS) 2006-17, "NRC Staff Position on the Requirements of 10 CFR 50.36, 'Technical Specifications,' Regarding Limiting Safety System Settings During Periodic Testing and Calibration of Instrument Channels," August 24, 2006
- 3. NUREG-1430, "Standard Technical Specifications Babcock and Wilcox Plants," Revision 5, September 2021
1CAN032301 Enclosure Page 18 of 18
Issuance of Amendment to Modify Technical Specification Table 3.3.1-1, 'Reactor Trip System Instrumentation,' Turbine Trip Function on Low Fluid Oil Pressure (EPID L-2017-LLA-0357)," ML18255A156, dated October 30, 2018
Issuance of Amendment Regarding Reactor Protection System Instrumentation Turbine Trip Function (CAC No. MF9401; EPID L-2017-LLA-0189)," ML18052B347, dated March 28, 2018
- 6. NRC letter to Duke Energy Progress, Inc., "H. B. Robinson Steam Electric Plant, Unit No. 2
- Issuance of Amendment Regarding Technical Specification 3.3.1, Reactor Protection System Instrumentation Turbine Trip (TAC No. MF3463)," ML15040A073, dated September 22, 2015
- 7. Regulatory Guide (RG) 1.105, "Instrument Setpoint for Nuclear Safety-Related Instrumentation," Revision 3, ML993560062, December 1999
- 8. ANO Design Guide IDG-001-0, "Instrument Loop Error Analysis and Setpoint Methodology Manual"
7.0 ATTACHMENTS
- 1. Technical Specification Page Markups
- 2. Retyped Technical Specification Pages
- 3. Technical Specification Bases Page Markups (Information Only)
- 4. CALC-22-E-0007-01 "Unit 1 Main Turbine Anticipatory Reactor Trip Setpoint Basis"
Enclosure, Attachment 1
Technical Specification Page Markups RPS Instrumentation 3.3.1
Table 3.3.1-1 Reactor Protection System Instrumentation
APPLICABLE CONDITIONS MODES OR REFERENCED OTHER FROM SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS ACTION C.1 REQUIREMENTS VALUE
- 1. Nuclear Overpower -
- a. High Setpoint 1,2(a),3(d) D SR 3.3.1.1 104.9% RTP SR 3.3.1.2 SR 3.3.1.4 SR 3.3.1.5
- b. Low Setpoint 2(b),3(b) E SR 3.3.1.1 5% RTP 4(b),5(b) SR 3.3.1.4 SR 3.3.1.5
- 2. RCS High Outlet Temperature 1,2 D SR 3.3.1.1 618 °F SR 3.3.1.4 SR 3.3.1.5
- 3. RCS High Pressure 1,2(a),3(d) D SR 3.3.1.1 2355 psig SR 3.3.1.4 SR 3.3.1.5
- 4. RCS Low Pressure 1,2(a) D SR 3.3.1.1 1800 psig SR 3.3.1.4 SR 3.3.1.5
- 5. RCS Variable Low Pressure 1,2(a) D SR 3.3.1.1 As specified in the SR 3.3.1.4 COLR SR 3.3.1.5
- 6. Reactor Building High Pressure 1,2,3(c) D SR 3.3.1.1 18.7 psia SR 3.3.1.4 SR 3.3.1.5
- 7. Reactor Coolant Pump to 1,2(a) D SR 3.3.1.1 55% RTP with one Power SR 3.3.1.4 pump operating in each SR 3.3.1.5 loop.
- 8. Nuclear Overpower RCS Flow 1,2(a) D SR 3.3.1.1 As specified in the and Measured AXIAL SR 3.3.1.3 COLR POWER IMBALANCE SR 3.3.1.4 SR 3.3.1.5
- 9. Main Turbine Trip (Hydraulic Oil 45% RTP F SR 3.3.1.1 40.5700 psig Pressure) SR 3.3.1.4 SR 3.3.1.5
- 10. Loss of Main Feedwater Pumps 10% RTP G SR 3.3.1.1 55.5 psig (Control Oil Pressure) SR 3.3.1.4 SR 3.3.1.5
- 11. Shutdown Bypass RCS High 2(b),3(b) E SR 3.3.1.1 1720 psig Pressure 4(b),5(b) SR 3.3.1.4 SR 3.3.1.5
(a) When not in shutdown bypass operation.
(b) During shutdown bypass operation with any CRD trip br eaker in the closed position and the CRD System capable of rod withdrawal.
(c) With any CRD trip breaker in the closed position and the CRD System capable of rod withdrawal.
(d) With any CRD trip breaker in the closed position, t he CRD system capable of rod withdrawal, and not in shutdown bypass operation.
ANO-1 3.3.1-5 Amendment No. 215,264, Enclosure, Attachment 2
Retyped Technical Specification Pages RPS Instrumentation 3.3.1
Table 3.3.1-1 Reactor Protection System Instrumentation
APPLICABLE CONDITIONS MODES OR REFERENCED OTHER FROM SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS ACTION C.1 REQUIREMENTS VALUE
- 1. Nuclear Overpower -
- a. High Setpoint 1,2(a),3(d) D SR 3.3.1.1 104.9% RTP SR 3.3.1.2 SR 3.3.1.4 SR 3.3.1.5
- b. Low Setpoint 2(b),3(b) E SR 3.3.1.1 5% RTP 4(b),5(b) SR 3.3.1.4 SR 3.3.1.5
- 2. RCS High Outlet Temperature 1,2 D SR 3.3.1.1 618 °F SR 3.3.1.4 SR 3.3.1.5
- 3. RCS High Pressure 1,2(a),3(d) D SR 3.3.1.1 2355 psig SR 3.3.1.4 SR 3.3.1.5
- 4. RCS Low Pressure 1,2(a) D SR 3.3.1.1 1800 psig SR 3.3.1.4 SR 3.3.1.5
- 5. RCS Variable Low Pressure 1,2(a) D SR 3.3.1.1 As specified in the SR 3.3.1.4 COLR SR 3.3.1.5
- 6. Reactor Building High Pressure 1,2,3(c) D SR 3.3.1.1 18.7 psia SR 3.3.1.4 SR 3.3.1.5
- 7. Reactor Coolant Pump to 1,2(a) D SR 3.3.1.1 55% RTP with one Power SR 3.3.1.4 pump operating in each SR 3.3.1.5 loop.
- 8. Nuclear Overpower RCS Flow 1,2(a) D SR 3.3.1.1 As specified in the and Measured AXIAL SR 3.3.1.3 COLR POWER IMBALANCE SR 3.3.1.4 SR 3.3.1.5
- 9. Main Turbine Trip (Hydraulic Oil 45% RTP F SR 3.3.1.1 700 psig Pressure) SR 3.3.1.4 SR 3.3.1.5
- 10. Loss of Main Feedwater Pumps 10% RTP G SR 3.3.1.1 55.5 psig (Control Oil Pressure) SR 3.3.1.4 SR 3.3.1.5
- 11. Shutdown Bypass RCS High 2(b),3(b) E SR 3.3.1.1 1720 psig Pressure 4(b),5(b) SR 3.3.1.4 SR 3.3.1.5
(a) When not in shutdown bypass operation.
(b) During shutdown bypass operation with any CRD trip br eaker in the closed position and the CRD System capable of rod withdrawal.
(c) With any CRD trip breaker in the closed position and the CRD System capable of rod withdrawal.
(d) With any CRD trip breaker in the closed position, t he CRD system capable of rod withdrawal, and not in shutdown bypass operation.
ANO-1 3.3.1-5 Amendment No. 215,264, Enclosure, Attachment 3
Technical Specification Bases Page Markups (Information Only)
RPS Instrumentation B 3.3.1
APPLICABLE SAFETY ANALYSES (continued)
- 8. Nuclear Overpower RCS Flow and Measured AXIAL POWER IMBALANCE (continued)
This trip supplements the protection provided by the Reactor Coolant Pump to Power trip, through the power to flow ratio, for loss of reactor coolant flow events. The power to flow ratio provides direct protection for the limiting loss of flow transient which is the loss of two RCPs from four pump operation. The im balance portion of the trip is credited for steady state protection only.
The power to flow ratio of the Nuclear Overpower RCS Flow and Measured AXIAL POWER IMBALANCE trip also provides steady state protection to prevent reactor power from exceeding the allowable power when the primary system is operating with two or three pump flow. Thus, the power to flow ratio prevents overpower conditions similar to the Nuclear Overpower trip. This protection ensures that during reduced flow conditions the core power is maintained below that required to begin DNB.
The Allowable Value is selected to ensure that a trip occurs prior to core power, axial power peaking, and reactor coolant flow conditions reaching DNB or fuel centerline temperature limits. The Allowable Value for this Function is given in the unit COLR because the cycle specific core peaking changes affect the Allowable Value.
- 9. Main Turbine Trip (Hydraulic Oil Pressure)
The Main Turbine Trip Function trips the reactor when the main turbine is tripped at high power levels. The Main Turbine Trip Function provides an early reactor trip in anticipation of the loss of heat sink associated with a turbine trip. The Main Turbine Trip Function was added to the Babcock and Wilcox (B&W) designed units in accordance with NUREG-0737 (Ref. 4) following the Three Mile Island Unit 2 accident. The trip lowers the probability of an RCS electromatic relief valve (ERV) actuation for turbine trip cases.
Each of the four turbine oil pressure switches feeds one of the four protection channels through a buffer that continuously monitors the status of the contacts. Therefore, failure of any pressure switch affects only one protection channel.
For the Main Turbine Trip (Hydraulic Oil Pressure) bistable, the Allowable Value of 70040.5 psig is selected to provide a trip whenever main turbine oil pressure drops below the normal operating range. The reactor power bypass is designed to automatically remove the turbine oil pressure trip function from the bypassed condition at
< 45% RTP. Alarms are available to alert operators when the bypass function is enabled. Should the automatic bypass removal function fail such that the channel remains in the bypassed state, the channel must be considered inoperable at power levels of 45% RTP and the appropriate condition is entered. Failure of the automatic bypass removal feature alone or the inability to place the channel in a bypassed state when < 45% RTP does not constitute channel inoperability. The automatic bypass removal feature is tested to ensure its continued availability during the monthly CHANNEL FUNCTIONAL TEST. The turbine trip is not required to protect against events that can create a harsh environment in the turbine building. Therefore, errors induced by harsh environments are not included in the determination of the setpoint Allowable Value.
ANO-1 B 3.3.1-12 Amendment No. 215 Rev. 67, Enclosure, Attachment 4
CALC-22-E-0007-01 "Unit 1 Main Turbine Anticipatory Reactor Trip Setpoint Basis"
(Contains 33 pages)
[8]ANO 1 AN02 GGNS IP2 IP3 PLP
CALCULATION (1) EC# 89298 (2 l Page 1 of 15
COVER PAGE
(3 ) Design Basis Cale. I IBJ YES I NO I (4 ) IBJ CALCULATION EC Markup
(5) Calculation No: CALC-22-E-0007-01 (6) Revision: 000
(7)
Title:
UNIT 1 MAIN TURBINE ANTICIPATORY REACTOR (Bl Editorial TRIP SETPOINT BASIS DYES IBJ NO
(9) System(s): TCS, RPS (10) Review Org (Department): DESIC
(11) (12) Component/Equipment/Structure Safety Class: Type/Number:
PS-8584-1 PT-8550A
Safety I Quality Related PS-8584-2 PT-8550B
Augmented Quality Program IBJ Non-Safety Related PS-8584-3 PT-8550C PS-8584-4
( 13 l Document Type: CALC
( 14l Keywords (Description/Topical Codes):
Uncertainty, Setpoint, ARTS, TCS
REVIEWS
( 15 ) Name/Signature/Date ( 16) Name/Signature/Date ( 17) Name/Signature/Date
Digita lly signed by A la ina Difan i Digitall y signed by James C Kinectrics / See Page 2 ON : c n =A laina Oifan i, c =US, o =Project Barnett Engin eering, g c./:)~ ON: c n-=Jame s C Ba rnell, c =US,
emai l=adi fani@en le rgy.com ~ o =Ente rgy, ou =MF P A N O,
~~~ Reason : Prepa rer ema 1l=Jbarn 10@ en ter gy.com Locatio n: ANO Da le : 2023.02.1510 :39:04 -06'00 '
Date : 2 0 23.02.15 08 :2 1:00--06'00 '
Responsible Engineer Design Verifier Supervisor/ Approval
IBJ Reviewer Comments Attached Comments Attached
EN-DC-126 R009 Page 2 of 15 Calculation Cover Sheet
DOCUMENT TITLE: UNIT 1 MAIN TURBINE PROJECT NO.: ANO-490467 ANTICIPATORY REACTOR TRIP SETPOINT BASIS
DOCUMENT NO.: CALC-22-E-0007-01 CLIENT REV. NO.: 000
CLIENT: Entergy STATION/UNIT: ANO / Unit 1 STATUS (check if final): Final Unverified Assumptions or Preliminary Design Input YES NO QUALITY CLASS: SR AQ NSR used?
Prepared by: Reviewed by: Approved by:
Internal Revision Description Revision Affected Pages (Signature / Date) (Signature / Date) (For Final )
000.1 All Initial Issue
Mark W. Guy Sterrett Krzyzaniak
000.2 5, 7 - 14, Incorporated Client comments.
Attachment 2
Mark Krzyzaniak W. Guy Sterrett
000.3 5-14, Incorporated Client comments. Revised Attachment ARTS setpoint to 850 psig and increased 2 Turbine Bldg max temperature to 140F. Mark Krzyzaniak Revised pressure switch drift and W. Guy Sterrett temperature specifications for vented housing and sensor diaphragm, respectively.
Added As-Left and As-Found Tolerances.
Added Hydrostatic Pressure Effect as a bias.
000.4 3, 5-8, Incorporated Client comments. Relocated 10-15 figures of Att. 2 into Conclusion section.
Removed Att. 2. Added Relationships and Cross References on page 3. Mark Krzyzaniak W. Guy Sterrett
2/2/2023
Form 3.4-1 Rev. 1 Kinectrics AES Inc.
Page 3 of 12 Calculation Reference Sheet
CALCULATION CALCULATION NO: CALC-22-E-0007-01 REFERENCE SHEET REVISION: 000
I. EC Markups Incorporated (N/A to NP calculations)
- 1. None
II. Relationships Sht Rev Input Output Impact Tracking Doc Doc Y/N No.
- 1. CALC-ANO1-IC-22-00002 N/A 0 N
- 2. OP-1304.164 N/A 23 Y EC 89300
- 3. OP-1304.165 N/A 20 Y EC 89300
- 4. OP-1304.166 N/A 22 Y EC 89300
- 5. OP-1304.167 N/A 19 Y EC 89300
- 6. Tech Spec Table 3.3.1-1 N/A 278-B80 Y (LAR-2021-153)
- 7. Tech Spec Bases B 3.3.1 N/A 278-B80 Y (LAR-2021-153)
- 8. ULD-1-SYS-15 N/A 6 Y EC 89300
- 9.
- 10.
III. CROSS
REFERENCES:
- 1. IDG-001
IV. SOFTWARE USED:
Title:
________N/A_________ Version/Release: ___________ Disk/CD No.__________
V. DISK/CDS INCLUDED:
Title:
_________N/A________ Version/Release: ___________ Disk/CD No.__________
VI. OTHER CHANGES: None
EN-DC-126 R009 CALC-22-E-0007-01 Page 4 of 15 Record of Revision
Revision Record of Revision
Initial Issue.
000
EN-DC-126 R009 Page 5 of 15 Calculation Sheet
DOC. NO.: CALC-22-E-0007-01 CLIENT REV.: 000 INTERNAL REV.: 000.4
DOC. TITLE: UNIT 1 MAIN TURBINE ANTICIPATORY REACTOR TRIP SETPOINT BASIS
PREPARER: Mark Krzyzaniak REVIEWER: W. Guy Sterrett
Table of Contents 1.
Purpose:
........................................................................................................................................................... 6
- 2.
Conclusion:
...................................................................................................................................................... 6 2.1 Results 2.2 Evaluation
- 3. Input and Design Criteria................................................................................................................................ 8
- 4. Assumptions.................................................................................................................................................... 8
- 5. Method of Analysis......................................................................................................................................... 9
- 6. Calculation..................................................................................................................................................... 10 6.1 Emergency Trip Header Low Pressure Setpoint Error 6.2 Main Turbine Anticipatory Reactor Trip Setpoint Error
- 7. References...................................................................................................................................................... 15
Total Pages Attachment 1 - SOR Form 651 (01.20), Switches for the Nuclear Power Industry..................................... 17
Form 3.4-2 Rev.1 Kinectrics AES Inc.
Page 6 of 15 Calculation Sheet
DOC. NO.: CALC-22-E-0007-01 CLIENT REV.: 000 INTERNAL REV.: 000.4
DOC. TITLE: UNIT 1 MAIN TURBINE ANTICIPATORY REACTOR TRIP SETPOINT BASIS
PREPARER: Mark Krzyzaniak REVIEWER: W. Guy Sterrett
1.
Purpose:
The Unit 1 Turbine Control System (TCS) upgrade project EC 89300 (Ref. 7.1) replaces the original analog electrohydraulic single channel system with a new redundant digital electrohydraulic Ovation based system. The function of the existing Auto Stop oil system is replaced by a high-pressure emergency trip header of the Electro-Hydraulic Control (EHC) system. The Auto Stop Oil System provides a trip interface between critical turbine functions and the high pressure EHC fluid. If one of the protective devices detects a condition requiring a turbine trip, Auto Stop oil pressure is dumped (drained rapidly) which allows the EH trip interface valve to open causing high pressure EH fluid to also drain which closes the turbine steam admission valves resulting in a turbine trip. The Anticipatory Reactor Trip (ART) system was installed by DCP-79-1032 to implement the requirements of the Three Mile Island (TMI) Action Plan (NUREG-0737 Supplement No. 1). This anticipatory trip provides a reactor trip signal to the Reactor Protection System (RPS) in those ca ses where a loss of secondary heat sink would likely result in a reactor trip based on the primary response to the loss of heat removal capabilities. The anticipatory trip limits the heat input to the system after a loss of heat sink, reducing the amount of heat that must be removed after the trip.
Due to the removal of the Auto Stop oil system, four (4) new pressure switches (PS-8584-1, -2, -3 and -4), which monitor the high pressure Emergency Trip Header (ETH), replace the existing pressure switches sensing Auto Stop oil pressure and provide the same tu rbine trip status to the RPS.
Along with replacement of the pressure switches, the sensing line taps are also modified to now monitor pressure of the ETH rather than the Auto Stop oil system whose function is being replaced by the ETH. Due to this change, a new pressure switch setpoint of 850 psi has been selected (Ref. 7.1). This setpoint is separate from the ETH Low Pressure setpoint as sensed by pressure transmitters PT-8550A/B/C, which provide the TCS with a trip signal in conjunction with losing controllability of the steam admission valves. The purpose of the following calculation is to provide a basis for the new anticipatory reactor trip set point, taking into considerati on both the uncertainty of the new pressure switches and the need to not inadvertently initiate a reactor trip signal before a turbine trip is assured to have been initiated as sensed by pressure transmitters PT-8550A/B/C.
- 2.
Conclusion:
2.1 Results The total error associated with the Emergency Trip Header Low Pressure Setpoint (TOT 1000) is as follows:
TOT1000 = +/- (Total Abnormal Transmitter Error 2 + Total Abnormal Module Error 2)0.5 TOT1000 = +/- 19.24 psi
The total error associated with th e Main Turbine ART Setpoint (TOT 850) is as follows:
TOT850 = +/- (Total Abnormal Switch Error)
TOT850 = 46.00 psi or -106.64 psi TOT850 = +/- 107.00 psi Note that the TOT850 is conservatively rounded up to +/-107.00 psi to encompass the total error interval.
Form 3.4-2 Rev.1 Kinectrics AES Inc.
Page 7 of 15 Calculation Sheet
DOC. NO.: CALC-22-E-0007-01 CLIENT REV.: 000 INTERNAL REV.: 000.4
DOC. TITLE: UNIT 1 MAIN TURBINE ANTICIPATORY REACTOR TRIP SETPOINT BASIS
PREPARER: Mark Krzyzaniak REVIEWER: W. Guy Sterrett
Per Ref. 7.3, the As-Left Tolerance (ALTOL) and As-Found Tolerance (AFTOL) of the pressure switch loop for the Main Turbine ART Setpoint are determined to be:
ALTOL = +/- 15.5 psi AFTOL = +/- 74.3 psi The as-left tolerance establishes the required accuracy band in which the switch must be calibrated and remain to avoid recalibration when periodically tested. The as-found tolerance establishes the limit of error the switch can be found to have during surveillance testing and still be considered in calibration and operable. Refer to the Figures below for a summary of setpoints, total errors, and tolerances.
Form 3.4-2 Rev.1 Kinectrics AES Inc.
Page 8 of 15 Calculation Sheet
DOC. NO.: CALC-22-E-0007-01 CLIENT REV.: 000 INTERNAL REV.: 000.4
DOC. TITLE: UNIT 1 MAIN TURBINE ANTICIPATORY REACTOR TRIP SETPOINT BASIS
PREPARER: Mark Krzyzaniak REVIEWER: W. Guy Sterrett
2.2 Evaluation Considering the above uncertainty, the ART setpoint could be initiated at an actual maximum pressure of 957 psig while the ETH Low Pressure setpoint could be initiated at an actual minimum pressure of 980.76 psig. This results in a margin of 23.76 psi between the maximum ART setpoi nt pressure and the minimum ETH low pressure setpoint.
The ETH is a relatively static, high-pressure system operating at 1800 to 1900 psi (Ref. 7.1). In the original Auto Stop oil configuration, the lower oil system setpoint compared to process pressure, fluctuations were an important consideration. For the higher ETH process pressure, random fluctuations in pressure of the ETH down to the level of the setpoints being considered is not credible during normal operations due to the large delta between normal operating pressure and trip setpoint. Any depressurization ev ent of the ETH, such as during a trip or hydraulic line rupture, will cause fluid pressure to rapidly drop to near zero and both setpoints will be initiated. The 23.76 psi margin between these setpoints is sufficient to assure the anticipatory trip does not initiate a reactor trip signal before a turbine trip is assured to have been initiate d at or below the 1000 psi ETH Low Pressure Setpoint.
- 3. Input and Design Criteria Reference numbers used in the body of th e calculation are based on the following list:
3.1. Main Turbine ART Setpoint of 850 psi (Ref. 7.1) 3.2. Pressure switch uncertainty values (Ref. 7.2) 3.3. Pressure transmitter uncertainty values (Ref. 7.5) 3.4. Ovation electronics module uncertainty values (Ref. 7.7) 3.5. ETH Low Pressure Setpoint of 1000 psi (Ref. 7.1) 3.6. Pressure transmitter setting tolerance of 0.125% Span has been incorporated to aid in field calibration (Ref.
7.6)
- 4. Assumptions 4.1 The calibration error allowance for the switch is based on calibration using an input device having reference accuracy of one-half that of the switch and us ing the switch as the output/indicator (Ref. 7.3).
4.2 Since minimum and maximum temperatures of the Tu rbine Building are not as we ll documented as those for the Auxiliary and Reactor building, an assumed temperature at calibration of 60°F and a maximum temperature of 140°F are used for temperature effect calculations (Ref. 7.1 and 7.3). Calibration occurs during refueling outages which typically takes place during milder conditions in the spring and fall, so 60°F is a reasonable assumption.
Summer temperatures of 100°F are not uncommon. Based on th e proximity of the pressure switches to the turbine, and for potential steam leaks in the vicinity, for conservatism an estimated 140°F for the Turbine Building is adequate. This is also the assumed temperature used in the transmitter evaluations. Note that these temperatures are conservative and bound the maximum temperatur es recorded per WO-ANO-540018 taken in July 2021.
4.3 The calibration error allowance for the transmitter is based on calibration using a dead weight tester and a digital multimeter, each having a re ference accuracy of one-half that of the transmitter (Ref. 7.3).
Form 3.4-2 Rev.1 Kinectrics AES Inc.
Page 9 of 15 Calculation Sheet
DOC. NO.: CALC-22-E-0007-01 CLIENT REV.: 000 INTERNAL REV.: 000.4
DOC. TITLE: UNIT 1 MAIN TURBINE ANTICIPATORY REACTOR TRIP SETPOINT BASIS
PREPARER: Mark Krzyzaniak REVIEWER: W. Guy Sterrett
4.4 The transmitters measure the pressure of the Emergency Trip Header which has an operating pressure between 1800 and 1900 psi. For conservatism, the higher value of 1900 psi is assumed as the static line pressure within the trip header during normal operation (Refs. 7.1 and 7.11).
4.5 Annual drift specification is 2.5% span per Ref 7.2. The pressure switches are calibrated every refueling outage or 18 months (Ref. 7.4). Since ANO Tech Specs allow a 25% extension, 22.5 months is assumed for the maximum calibration interval. Therefore, per Ref. 7.3, a drift of 4.69% span (2.5%
- 1.5 years
- 1.25) is used.
4.6 Per Refs. 7.3 and 7.5, Static Pressure Zero and Span Effects are only applicable to differential pressure transmitters and not gauge pressure transmitters.
4.7 Since no calibration interval has been established for the pressure transmitters, the requested calibration frequency for the similar Unit 2 non-critical, essential TCS pressure transmitters of 4R or 72 months has been assumed. (Ref. 7.12) 4.8 Power Supply Effect is assumed to be negligible based on 1% load regulation of Mean Well HDR 30 Series power supply. (Refs. 7.9 and 7.10) 4.9 Electronics Module Reference Accuracy includes a term of +/-0.5
- Least Significant Bit (LSB). The module is a 14-bit unipolar device which results in an approximate 0.003% span which is a negligible contribution to its accuracy. (Ref. 7.7) 4.10 The elevation difference between the location of the pressure switches compared to the location of the pressure transmitters is assumed to be 20 feet (maximum) based on floor elevations of 386 (for pressure switches in cabinets C491 and C492, Ref. 7.8) and 368 (for pressure transmitters in rack R-8500, Ref. 7.9) and similar mounting heights above floors. This resu lts in a hydrostatic pressure (head) effect of -9.92 psig (-20 ft
- 0.434 psi/ft
- 1.145 (specific gravity of Fyrquel EH fluid, Ref. 7.1 Hydraulic Response Evaluation)). This represents a negative bias as the ARTS pressure switches sense an ETH pressure lower than the Main Turbine trip pressure transmitters.
- 5. Method of Analysis Calculations performed in support of this setpoint basis ar e accomplished using site and industry standard practices outlined in Reference 7.3. Random errors are combined using the Square Root Sum of Squares (SRSS) method while bias errors are algebraically combined. The total erro r for a given string is then the algebraic combination of the total random error and any bias terms.
Form 3.4-2 Rev.1 Kinectrics AES Inc.
Page 10 of 15 Calculation Sheet
DOC. NO.: CALC-22-E-0007-01 CLIENT REV.: 000 INTERNAL REV.: 000.4
DOC. TITLE: UNIT 1 MAIN TURBINE ANTICIPATORY REACTOR TRIP SETPOINT BASIS
PREPARER: Mark Krzyzaniak REVIEWER: W. Guy Sterrett
- 6. Calculation 6.1 Emergency Trip Header Low Pressure Setpoint Error The ETH has a maximum operating pressure of approximatel y 1900 psig (Ref. 7.11) and is responsible for tripping the Main Turbine by depressurizing wh ich results in closing of the steam admission valves, thereby removing steam from the turbine (Ref. 7.1). Low pressure protection for the ETH is provided electrically by three pressure transmitters (PT-8550A/B/C) in a 2-out-of-3 logic configuration. A mechanical means of protection also exists in that ETH pressure drops below the level at which the spring force of the steam admission valves overcomes the hydraulic fluid pressure and forces the valves closed. The electrical trip provided by the pressure transmitters has been selected by Westinghouse as 1000 psig such that the electrical trip is initiated in conjunction with a mechanical trip. Based on the EHC bleed down test performed to verify the minimum EH pressure necessary to maintain the main steam governor and throttle valves (GV/TVs) in a fu lly-open position (Ref. 7.13), the actual pressure at which the GVs begin to close occurs at 1024 psig (fully closed at 480 psig), whereas the TVs begin to close at 601 psig (fully closed at 243 psig). The automatic electrical trip at 1000 psig is deemed acceptable as Operator action to secure the turbine is to take place if EH pressure cannot be maintained before any potential loss of control of the valves due to spring force competing with hydraulic force (Ref. 7.1). This allows Operators a small window to initiate a manual trip rather than simply allowing an automatic trip to take pl ace. Figure 1 (in Section 2.1 Results) depicts the alarms and interlock (for Auxiliary EH Pump Auto Start) that result on a decreasing EH pressure scenario.
As part of providing the basis for the Main Turbine ART setpoint, the uncertainties for the ETH Low Pressure Setpoint are also developed to ensure adequate margin exists between the two setpoints. The instrument string responsible for the electrical TCS trip setpoint consists only of pressure transmitters and the Ovation modules (Ref.
7.9). The individual component uncertainties are developed below and combined in Section 2.1.
Form 3.4-2 Rev.1 Kinectrics AES Inc.
Page 11 of 15 Calculation Sheet
DOC. NO.: CALC-22-E-0007-01 CLIENT REV.: 000 INTERNAL REV.: 000.4
DOC. TITLE: UNIT 1 MAIN TURBINE ANTICIPATORY REACTOR TRIP SETPOINT BASIS
PREPARER: Mark Krzyzaniak REVIEWER: W. Guy Sterrett
Pressure Transmitter (TRX)
Pressure Transmitter Component IDs: PT-8550A, PT-8550B, PT-8550C (Ref. 7.1)
Manufacturer: Rosemount (Ref. 7.1)
Model: 3051S1TG4A2E11A1AHR7DA2C1Q4 (Ref. 7.1, 7.6)
Range Code: 4A (Ref. 7.1, 7.6)
Upper Range Limit (URL): 4000 psig (Ref. 7.5)
Calibrated Range: 0 to 3000 psig (Ref. 7.6)
Span: 3000 psig (Ref. 7.1)
Process/Environmental Conditions:
Calibration Temperature (AMB): 60°F (Assumption 4.2)
Abnormal Temperature (ABN) 140°F (Assumption 4.2)
Delta T (DT = ABN-AMB) 80°F Static Line Pressure 1900 psig (Assumption 4.4)
Error Summary:
Reference Accuracy (RA) +/- 0.025% Span (Ref. 7.5)
Setting Tolerance (ST) +/- 0.125% Span (Input 3.6)
Device Tolerance (DTOL) +/- (RA + ST) = +/-0.150% Span Calibration Error (CAL) +/- [(DTOL/2)2 + (DTOL/2)2]0.5 (Assumption 4.3)
CAL = +/-0.106% Span Drift (DR) +/- (0.20% URL for 10 yrs.) / 50°F / 1000 psi (Ref. 7.5, Assump. 4.7)
Drift (120 months) = +/-0.811% Span Drift (72 months) = +/-0.486% Span Temperature Effect (TE) +/- (0.009% URL + 0.025% Span) per 50°F (Ref. 7.5)
TE (for 80°F Delta T) = +/- (0.009% URL + 0.025% Span)
- DT / 50°F
= +/-0.059% Span Power Supply Effect (PE) < +/-0.005% Span / per voltage change (Ref. 7.5)
PE Negligible (Assumption 4.8)
Static Pressure Zero Effect (ZE) Not Applicable for Gauge Pressure Transmitter (Assumption 4.6)
ZE = 0 Static Pressure Span Effect (SE) Not Applicable for Gauge Pressure Transmitter (Assumption 4.6)
SE = 0
The errors of the Pressure Transm itter (TRX) are determined below:
Transmitter Error for Reference Conditions:
REF eTRX: = +/- (DTOL + CAL)
= +/- 0.256% Span
Form 3.4-2 Rev.1 Kinectrics AES Inc.
Page 12 of 15 Calculation Sheet
DOC. NO.: CALC-22-E-0007-01 CLIENT REV.: 000 INTERNAL REV.: 000.4
DOC. TITLE: UNIT 1 MAIN TURBINE ANTICIPATORY REACTOR TRIP SETPOINT BASIS
PREPARER: Mark Krzyzaniak REVIEWER: W. Guy Sterrett
Transmitter Error for Abnormal Conditions:
ABN eTRX: = +/- ((DTOL+CAL) 2 + DR2 + TE2 + ZE2 + SE2)0.5 ABN eTRX: = +/-0.553% Span
= +/-16.59 psi
Electronics Module (MOD)
Electronics Module Component IDs: C33A-1-A1-1.1.1, C33A-1-B8-1.2.8, C33A-2-C1-1.3.1 (Ref. 7.1)
Manufacturer: Westinghouse (Ovation) (Ref. 7.1)
Model: 5X00070G01 (Ref. 7.9)
Full Scale: 20 mA (Ref. 7.9)
Input Range: 4 to 20 mA (Ref. 7.9)
Input Span: 3000 psig (Ref. 7.1)
Error Summary:
Reference Accuracy (RA) +/- 0.10% Full Scale +/- 0.5 LSB (Ref. 7.7, Assump. 4.9)
RA = +/- (0.10%
- 20 mA / 16 mA) = +/- 0.125% Span Nonlinearity +/- 0.003% Full Scale (Ref. 7.7)
LIN = +/- (0.003%
- 20 mA / 16 mA) = +/- 0.004% Span Temperature Effect (TE) +/- 0.24% Full Scale from 0-60°C (32-140°F) (Ref. 7.7)
TE = +/- (0.24%
- 20 mA / 16 mA) = +/- 0.30% Span
The errors of the Ovation Electronics Module (MOD) are given below:
Ovation Electronics Module Error for Reference Conditions:
REF eMOD: +/- (RA2 + LIN2)0.5
= +/- 0.125% Span Ovation Electronics Module Error for Abnormal Conditions:
ABN eMOD = +/- (RA2 + LIN2 + TE2)0.5 ABN eMOD = +/- 0.325% Span
= +/- 9.75 psi
The errors of the Pressure Transmitter and Electronics Module are combined in the To tal Error for the ETH Low Pressure Setpoint (TE1000):
TOT1000 = +/- (ABN eTRX2 + ABN eMOD2)0.5 TOT1000 = +/- (16.592 + 9.752) 0.5 psi
TOT1000 = +/- 19.24 psi
Form 3.4-2 Rev.1 Kinectrics AES Inc.
Page 13 of 15 Calculation Sheet
DOC. NO.: CALC-22-E-0007-01 CLIENT REV.: 000 INTERNAL REV.: 000.4
DOC. TITLE: UNIT 1 MAIN TURBINE ANTICIPATORY REACTOR TRIP SETPOINT BASIS
PREPARER: Mark Krzyzaniak REVIEWER: W. Guy Sterrett
6.2 Main Turbine Anticipatory Reactor Trip Setpoint Error The Main Turbine ART setpoint of 850 psig is selected as d escribed below to not inadvert ently initiate a trip signal before a turbine trip is assured to have been initiated. This is accomplished by developing uncertainties for the instrument loop and showing that the ART setpoint, along with uncertainties, will not initiate prior to the ETH low pressure setpoint.
The instrument string responsible for supplying the Main Turbine anticipatory trip consists of the pressure switch and a contact buffer before ultimately reaching the RPS reactor trip assembly (Ref. 7.8). For purposes of this determination, only the uncertainty values associated with the new pressure switch will be considered. The contact buffer provides non-1E to 1E isolation and has no associated uncertainty values. The following calculation shows the development of the pressure switch un certainties and thus the total string e rror for the Main Turbine anticipatory trip.
Pressure Switch (SWTCH)
Pressure Switch Component IDs: PS-8584-1, PS-8584-2, PS-8584-3, PS-8584-4 (Ref. 7.1)
Manufacturer: SOR Measurement and Control (Ref. 7.1)
Model: 9N6-BB45-U1-C1A-JJTTNQ (Ref. 7.1)
Span: 1550 psi (1750max - 200min) (Ref. 7.2)
Calibration Interval: 18 Month (Ref. 7.4)
Setpoint (SP): 850 psig (Ref. 7.1)
Process/Environmental Conditions:
Calibration Temperature (AMB): 60°F (Assumption 4.2)
Abnormal Temperature (ABN) 140°F (Assumption 4.2)
Delta T (DT = ABN-AMB) 80°F
Form 3.4-2 Rev.1 Kinectrics AES Inc.
Page 14 of 15 Calculation Sheet
DOC. NO.: CALC-22-E-0007-01 CLIENT REV.: 000 INTERNAL REV.: 000.4
DOC. TITLE: UNIT 1 MAIN TURBINE ANTICIPATORY REACTOR TRIP SETPOINT BASIS
PREPARER: Mark Krzyzaniak REVIEWER: W. Guy Sterrett
Error Summary:
Repeatability (RA) +/- 1.0% Span = +/- 15.5 psi (Ref. 7.2)
Calibration Error (CAL) +/- [(RA/2)2 + (RA)2]0.5 (Assumption 4.1)
CAL = +/-1.1% Span = +/- 17.3 psi Drift (DR) (22.5 months) +/- 4.69% Span = +/- 72.7 psi (Ref. 7.2, Assump. 4.5)
Temperature Influence (TI) - (SP
- 0.0003)
- DT (Refs. 7.2 and 7.8)
TI (for Vented Housing) = -20.4 psi Hydrostatic Pressure Effect (HP) = -9.92 (Assumption 4.10)
The errors of the Pressure Switch (SWTCH) are determined below:
Switch Error for Reference Conditions:
REF eSWTCH: = +/- (RA2 + CAL2)0.5
= +/- 23.23 psi Switch Error for Abnormal Conditions:
ABN eSWTCH: = +/- (REF eSWTCH2 + DR2)0.5 + TI + HP
= +/- 76.32 psi + (-20.4 psi) + (-9.92 psi)
= 46.00 psi or -106.64 psi
= +/- 107.00 psi Therefore, the Total Error for th e Main Turbine ART Setpoint (TOT 850) is:
TOT850 = ABN eSWTCH
= +/- 107.00 psi
As-Left Tolerance:
The device repeatability (RA) is used to determine the as-left tolerance:
ALTOL: = RA (Ref. 7.3)
= +/- 1.0% Span = +/- 15.5 psi
As-Found Tolerance:
The SRSS (Square Root of the Sum of Squares) of the rep eatability (RA) and drift term (DR) is used to determine the as-found tolerance:
AFTOL: = +/- (RA2 + DR2)0.5 (Ref. 7.3)
= +/- (15.52 + 72.72)0.5 psi
= +/- 74.3 psi
Form 3.4-2 Rev.1 Kinectrics AES Inc.
Page 15 of 15 Calculation Sheet
DOC. NO.: CALC-22-E-0007-01 CLIENT REV.: 000 INTERNAL REV.: 000.4
DOC. TITLE: UNIT 1 MAIN TURBINE ANTICIPATORY REACTOR TRIP SETPOINT BASIS
PREPARER: Mark Krzyzaniak REVIEWER: W. Guy Sterrett
- 7.
References:
7.1 EC-89300, ANO Unit 1 Turbine Contro l System DEHC Upgrade Implementation 7.2 Attachment 1, SOR Form 651 (01.20), Switches for the Nuclear Power Industry 7.3 Design Guide IDG-001-0, Instrument Loop Erro r Analysis and Setpoi nt Methodology Manual 7.4 ANO-1 Procedures:
OP-1304.164, Unit 1 Reactor Protection System Channel A Refueling Calibration OP-1304.165, Unit 1 Reactor Protection System Channel B Refueling Calibration OP-1304.166, Unit 1 Reactor Protection System Channel C Refueling Calibration OP-1304.167, Unit 1 Reactor Protection System Channel D Refueling Calibration 7.5 TDR369 0670 Rev. 0, Rosemount 3051S Transmitter Technical Manual 7.6 CALC-ANO1-IC-22-00002 Rev. 0, ANO-1 Setpoint Documentation Package for Main Turbine Control System (TCS) 7.7 TDW120 6130 Rev. 0, Turbine Control System (TCS) Ovation I/O Reference Manual 7.8 ANO Drawings (for pressure switches):
E-260 Sh. 5 Rev. 3, Wiring Block Diagram Reactor Nuclear Instrumentation and Protection System M1Q-249 Sh. 1 Rev. 0, 885-NI/RPS Sch ematic Turbine Trip Subassembly A M1Q-251 Sh. 1 Rev. 1, 885-NI/RPS Sch ematic Turbine Trip Subassembly B M1Q-253 Sh. 1 Rev. 1, 885-NI/RPS Sch ematic Turbine Trip Subassembly C M1Q-255 Sh. 1 Rev. 1, 885-NI/RPS Sch ematic Turbine Trip Subassembly D M1Q-100 Sh. 1 Rev. 6, Schematic Diagram Bistable Trip String, Module Interlock, Test Trip Subassembly A M1Q-101 Sh. 1 Rev. 6, Schematic Diagram Bistable Trip String, Module Interlock, Test Trip Subassembly B M1Q-102 Sh. 1 Rev. 6, Schematic Diagram Bistable Trip String, Module Interlock, Test Trip Subassembly C M1Q-103 Sh. 1 Rev. 6, Schematic Diagram Bistable Trip String, Module Interlock, Test Trip Subassembly D M-504 Sh. 412 Rev. 0, Instrument Installation Tu rbine EH Trip Header RPS Pressure Switches E-649 Sh. 1 Rev. EC89300, Conduit & Tray Layout Turbine Generator Bldg. Area 1 Above El. 386-0 7.9 ANO Drawings (for pressure transmitters):
E-172 Sh. 4 Rev. 0, Schematic Diagram Turbin e Generator Main Turbine Control Cabinet E-777 Sh. 3 Rev. 1, Main Turbine Emergency Trip System Wiring E-777 Sh. 18 Rev. 1, Main Turbine Emergency Trip System Wiring E-777 Sh. 19 Rev. 1, Main Turbine Emergency Trip System Wiring E-773 Sh. 2 Rev. 1, Power Panel Assemblies Bill of Material E-765 Sh. 3 Rev. 1, C33A-1 & C33A-2 Turb ine Protection Cabinets Bill of Materials M-551 Sh. 1 Rev. EC89300, Instrument Location Area 1 Turbine Building Plan El. 361-0 to El. 386-0 7.10 TDW120 6610 Rev. 0 Turbine Control System (TCS) Ovation Electrical Parts 7.11 TDW120 3080 Rev. 19 Instruction Book 902,322 KW Steam Turbine Volume I Operation and Control Instruction Book 1250-C759 7.12 AR #21015372 PM Change Request for U2 TCS Pressure Transmitters 7.13 WO #567937-08
Form 3.4-2 Rev.1 Kinectrics AES Inc.
Attachment 1 Technical Manual SOR Switches for the Nuclear Power Industry SEE MORE AT SOR Inc.com
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Switches for the Nuclear Power Industr y
SOR Pressure, Vacuum and Temperature Switches are qualified by a combination of testing and analysis per IEEE-323-1974 &
1983 and IEEE-344-1975 & 1987. See SOR Test Repor t, 9058-102, 9058-105, and 8923-306 for qualification testing and explanations. ( Note: for nuclear qualified dif ferential pressure switches, see SOR catalog 1291. )
Qualification
- Thermal Aging testing
- Irradiation included
- Mechanical / Electrical Cycling 12 N 6
- Sine Beat
- Random Multifrequency
- LO C A
9RT 6 TA
SOR maintains a quality program committed to compliance with the applicable elements of 10CFR50, Appendix B, ANSI N45.2 and NQ A-1, including the repor ting requirements of 10CFR21. The products in this catalog are manufactured under this quality program which is audited by the Nuclear Procurement Issues C ommittee ( NUPIC ), and Nuclear Industr y Assessment Committee ( NIAC ).
SORInc.com l 913-888-2630 1/16 Switches for the Qualification Program Nuclear Power Industr y Summar y
Figure 1: AGING Figure 2 : SEISMIC 1,000,000 - 100 % RRS Cur ves at 1% Damping
500,000 - 10 0 -
50 -
30 -
100,000 - 25 -
16. 67 50,000 - 10 -
5 -
2 - SSE
10,000 - 1 - OBE
5000 -.5 -
.1 -
1000 -.5 1 2 3 4 5 10 20 30 50 1 00 100 1 20 1 40 1 60 1 80 200Frequency ( Hz)
Service Temperature ( F )
This is the RRS ( Required Response This graph is based on the Arrhenius Spectrum ) at 1% damping to which all equation and may be used as a general switches were seismically tested. All TRS guideline in determining the qualified life if ( Test Response Spectrum ) plots are contained the ser vice temperature is greater than or in test repor t 9058-102. Seismic damping less than 119.257 ° F. analysis to 0.5 %, 2 %, 3 %, 4 %, and 5 % is also available upon request.
Figure 3: LOCA
336 F/69 PSIG ¬ Cycle Test 276 F/61 PSIG 336 F/69 PSIG 340 316 F/39 PSIG 320 296 F/39 PSIG 300 246 F/24 PSIG 280 246 F/9 PSIG 250
200 Cycle Daily
¬ ¬ ¬ ¬ ¬¬¬ 140
Chemical Chemical Spray Spray This graph shows the combined environmental conditions to which cer tain switches were subjected at end-of-life conditions to simulate a LOCA ( Loss Of C oolant Accident).
The two thermodynamic transients were generated by injecting superheated steam into the autoclave in a controlled manner. The chemical spray consisted of 0.28 molar boric acid and 0.064 molar sodium thiosulfate buf fered to pH 10.5 with sodium hydroxide.
Note: Time values have been rounded. See test repor t 9058-102 for exact values.
2 /16 913-888-2630 l SORInc.com Switches for the Qualification Program Nuclear Power Industr y Summar y
Figure 4: H E LB 1 Figure 5 : HELB 2
223 F/6 PSIG ¬ Cycle Test346 F/39 PSIG¬ Cycle Test
230 210 F/0.5 PSIG 220 203 F/0 PSIG 350 271 F/29 PSIG 210 161 F 300 200 190 250 180 170 30-Sec 200 160 Ramp 150 150 140 Cycle Daily 130 20-Sec 100 120 Ramp 80 F/0 PSIGTo 110¬ ¬ ¬ ¬ ¬50 100 ¬ ¬ ¬Ambient
0 1 7 13 90.6 0 10 Sec 1.1 Hrs 2.8 Hrs 28 Hrs Time ( Hours)Time ( Hours)
This graph shows the combined environmental This graph shows the combined conditions to which cer tain switches were environmental conditions to which cer tain subjected at end-of-life conditions to simulate switches were subjected at end-of-life a HELB ( High Energy Line Break). The HELB 1 conditions to simulate a second more profile shown here was generated by injecting severe HELB ( High Energy Line Break).
superheated steam into the autoclave in a The HELB 2 profile shown here was controlled manner. generated by injecting superheated steam Note: Time values have been rounded. See test into the autoclave in a controlled manner.
repor t 9058-102 for exact values. Note: Time values have been rounded. See test repor t 9058-102 for exact values.
Qualification Program Explanation
- Thermal Aging to simulate a 20 -year life at a of the adjustable range. Temperature switch ser vice temperature of 119.257 ° F ( see Figure sensors were thermally cycled from 20° F 1). Switches were subjected to accelerated below set point to 20° F above set point. All thermal aging according to the Arrhenius cycling was conducted with full rated voltage model and based on the lowest activation and current applied to the switch contacts.
energy of all of the safety related, non-metallic materials of construction.
- Sine Beat testing at 1-50 Hz, 4.5g on line-mount temperature switches. This test was
- Radiation Aging to 31 or 186 megarads per formed to age the switch and determine minimum. Switches were subjected to various its response to these conditions. Only the amounts of gamma irradiation ( see test direct mount temperature switch was chosen repor t) to simulate that amount of radiation for this test as it is the only switch that may the switch might be exposed to during its be line mounted.
qualified life, plus the amount of radiation it might see during an accident plus margin.
- Random Multifrequency testing including five OBEs ( Operating Basis Ear thquake )
- Mechanical /Elec trical Cycling to 30,000 and one SSE ( safe shutdown ear thquake ) in cycles at full-scale pressure /temperature and each of four orientations ( see Figure 2). This rated electrical load. Pressure and vacuum test was per formed to age the switch and switches were cycled either pneumatically or determine its response to these conditions.
hydraulically from the low end to the high end
SORInc.com l 913-888-2630 3 /16 Switches for the Qualification Program Nuclear Power Industr y Explanation
- LO C A ( Loss Of C oolant Accident) testing
- HELB ( High Energy Line Break) testing to two on selected models ( see Figure 3 ). This test different profiles on selected models simulated LOCA conditions and established ( see Figures 4 and 5 ). This test simulated two the switchs response /condition before during dif ferent HELB conditions and established the and af ter the test. switchs response /condition before during and af ter the test.
The above testing brought the switches to end-of-life conditions as required by the IEEE standards and then subjected them to accident conditions. Please note that none of the qualification levels were established based on a specific application. Rather, they were chosen generically with the intent to be suitable for the majority of possible applications in nuclear power stations. It is the responsibility of the end user to establish if the qualification levels are suitable for the intended use.
Specifications Repeatabilit y of SOR Switches Pressure Switch +/-1% of Span Temperature Switch +/-1.5 % of Span Vacuum Switch +/-1% of Span (+/-1.5 % Post-LOCA)
Repeatability, as defined by ISA /ANSI S51.1, is the closeness of agreement among a number of consecutive measurements of the ouput ( set point) for the same value of the input under the same operating conditions, approaching from the same direction, for full range traverses.
Drif t Ma ximum Annual Drif t for all qualified models ( except # 9 & # 29 pistons with U8 diaphragm ) is 2.5 % of span. The Ma ximum Annual Drif t for # 9 & # 29 pistons with U8 diaphragms is 4.0 % of span.
Temperature Influence Formulas for Pressure and Vacuum Switches The formulas given below represent a general guideline for the expected influence of temperatures on the set points of the pressure and vacuum switches in this catalog.
Housing Sealed - SP = [ 0.027 ( psi / ° F ) - ( SP x 0.0003 / ° F )] x ( Tf - Ti )
Vented - SP = - ( SP x 0.0003 / ° F ) x ( Tf - Ti )
Where: SP = The change in the set point in ( psi ) from the intial value.
SP = The initial set point in ( psi ).
TI = The initial ambient temperature in ° F Tf = The final ambient temperature in ° F
Test Repor ts for SOR Pressure, Vacuum and Temperature Switches 9058-102 Qualification Test Repor t. 9058-120 R-series housing with M20x1.5 9058-103 DC rating on W sw itch element.* conduit connection.*
9058-105 U1 diaphrag m option for 8923-306 Switch without JJ conduit seal.
improved long-term drif t Af fects qualification levels.
and dead band. Contact SOR.
Af fects qualification levels. 8923-340 N6 housing.
Contact SOR.
9058-119 Terminal Block option in R-series housing.*
- Contact SOR for ordering information
4 /16 913-888-2630 l SORInc.com Switches for the Nuclear Power Industr y How to Order
Model Number System 12N6-B45-U8-C2A-JJTTN Q
Piston Housing DiaphragmSwitching Range Pressure P ort Accessories (Designator 1) (Designator 2) Element Spring (Designator 5) (Designator 6) (Designator 7)
(Designator 3) (Designator 4)
To specify aPressure Switch, begin with Step 1a. Use the sample model number above Vacuum Switch, begin with Step 1b. each table to position selected Temperature Switch, begin with Step 1c. designators within the model number.
Important Note: Some options may reduce the qualification level of a given model. The qualification of a switch is only as good as the weakest link. See SOR Qualification Test Repor ts 9058-102, 9058-105, and 8923-306 for fur ther details. Also, reference qualification levels given in Steps 2, 4, and 6 of this catalog.
Step 1a: Pressure Switch
Place designators in positions 1 and 4. 12N6-B45-U8-C2A-JJT TNQ
Piston Spring Adjustable Range Overrange1
(Designator 1) (Designator 4) psi bar [mbar] psi bar 12 4 0.5 to 6.0 [35 to 415]
12 5 0.75 to 12 [50 to 830]200 14 12 45 1 to 16 [70 to 1100]
4 4 2 to 25 0.14 to 1.7 4 5 3 to 50 0.2 to 3.5750 50 4 45 4 to 75 0.3 to 5 6 2 7 to 30 0.5 to 2 6 3 12 to 100 0.8 to 7 6 5 20 to 180 1.4 to 12 6 45 25 to 275 1.7 to 191500 100 5 3 25 to 240 1.7 to 16 5 5 35 to 375 2.4 to 26 5 45 45 to 550 3.1 to 38 29 4 80 to 400 5.5 to 28 29 45 150 to 1350 10 to 93 9 4 100 to 500 7 to 352500 170 9 5 200 to 1000 14 to 70 9 45 200 to 1750 14 to 120
- 1. The ma ximum input pressure /temperature that can be continuously applied to the switch without causing a permanent change of set point, leakage or material failure.
- 2. Dead Bands: Please contact SOR with model and increasing /decreasing set point value to obtain dead band information.
Step 1b : Vacuum Switch
Place designators in positions 1 and 4. 54N6-B11 8-M9-C2A-JJTTNQ
Piston Spring Adjustable Range (Vacuum Pressure) Overrange (Designator 1) (Designator 4) in. Hg mbar psi bar 54 118 30 1 1000 35 750 50
SORInc.com l 913-888-2630 5 /16 Switches for the Nuclear Power Industr y How to Order
Step 1c : Temperature Switch
Place designators in positions 1 and 4. 201N6-B12 5-U9-C7A-JJT TNQ
Probe Range Mounting Capillary Max Process Length Adjustable Range Overrange Temperature Pressure (Designator 1) (Designator 4) Type ft. m °F °C °F °C psi bar 201 125 Direct - -
203 125 Remote 6 1.8 205 125 Remote 10 3.040 to 225 5 to 107 360 182 2 07 125 Remote 15 4.5 209 125 Remote 20 6.0 2300 158 201 115 Direct - -
203 115 Remote 6 1.8 205 115 Remote 10 3.0150 to 66 to 37 5 190 520 270 2 07 115 Remote 15 4.5 209 115 Remote 20 6.0
Step 2: Select Housing 12N6-B45-U8-C2A-JJTTN Q Replace N6 in the sample model number with the appropriate housing designator.
Housing Specifications Qualification (Designator 2) Material Conduit DB E Radiation3 N6 Carbon Steel RT 316SS (CF8M)3/4 N PT (F) HELB 1 31 Mrad TA4 Ductile Iron HELB 2 & LO C A2 186 Mrad
- 1. Reference Page 3, Figures 4 & 5.
- 2. Reference Page 2, Figure 3.
- 3. The noted values represent the minimum irradiation aging dose applied during qualification testing.
- 4. Temperature switches in the TA housing are qualified for 31 Mrad and HELB only.
Step 3 : Select Switching Element 12 N 6 -B45-U8-C2A-JJT TNQ Replace B in the sample model number with the appropriate switching element designator.
Switch AC Rating DC RatingResistive Contact Form (Designator 3)
Volts Amps Volts Amps B 250 5 125 0.3 SPDT W* 250 5 - - SPDT BB 250 5 125 0.3 DPDT
- DC rating is optional. Contact SOR.
6 /16 913-888-2630 l SORInc.com Switches for the Nuclear Power Industr y How to Order
Step 4: Select Diaphragm System
Replace U8 in the sample model number with the 12 N 6 - B 4 5 -U8-C 2 A -J J T T N Q appropriate diaphragm system designator.
NOTE: If the designator 1 (chosen in step 1) does not appear under Compatible Designators, the line item is not available.
Diaphragm Diaphragm Diaphragm System Qualification Compatible (Designator 5) Material Designators Welded O -Ring DBE7 Radiation3 Cycles(Designator 1)
U14 HELB1 & 31 M r a d 5,000 9, 29 Standard ( 5, 6 Optional )
U85 31 Mrad or Ye s NoneLOCA24, 5, 6, 12 Standard 18 6 M r a d6 ( 9, 29 Optional )
U9 HELB1 31 M r a d316 S S T201, 203, 205, 30,000 207, 209 M4 Viton M9 EPR HELB1 & No54 LOCA2 31 Mrad or 18 6 M r a d6
- 1. Reference Page 3, Figures 4 & 5.
- 2. Reference Page 2, Figure 3.
- 3. The noted values represent the minimum irradiation aging dose applied during qualification testing.
- 4. The U1 is standard on the 9 & 29 Piston and it is optional on the 5 & 6 Piston. The U1 has significantly bet ter dead band and long term drif t for the 9 & 29 Piston. The U1 has marginally bet ter dead band and drif t on the 5 & 6 Piston.
Refer to Test Repor t 9058-105.
- 5. The U8 is standard on the 4, 5, 6 & 12 Piston and it is optional on the 9 & 29 Piston. If higher cycles or radiation levels are needed for the 9 & 29 the U8 can be specified, but this will increase the dead band and long term drif t.
6. The TA Housing selection is required for the 186 Mrad radiation level.
Step 5: Select Process Connection Replace C2A in the sample model number with the 12N6-B45-U8-C2A-J J T T N Q appropriate process connection designator.
NOTE: If the designator 1 (chosen in step 1) does not appear under Compatible Designators, the line item is not available.
Process Connection (Designator 6) Connection Material Connection Size/Type Compatible Designators (Designator 1)
C1A 1/4 N PT(F) 12, 4, 5, 6, 9, 29, 54 C2A 1/2 N PT(F)316SS C7A 1/2 N PT(M) 201, 203, 205, 207, 209
Step 6 : Select Accessories 12N6-B45-U8-C2A-JJT TNQ
Accessories (Designator 7) Description JJ Conduit seal with 17 ft. lead wire length. Optional.
This designator must be used for H E LB and LOCA applications.
RR Stainless steel tag attached with stainless steel wire to housing.
TT Oversized nameplate for tagging information. Required designator.
NQ Nuclear-qualified model. Required designator.
SORInc.com l 913-888-2630 7/16 Switches for the Nuclear Power Industr y How to Order
Approximate Weights
Housing Designator Piston/ Probe Weightlbs kg 12 5.63 2.55 4, 54 3.94 1.79 N6 5, 6, 9, 29 3.56 1.62 201 3.88 1.76 203, 205, 207, 209 3.75 1.70 12 7.13 3.23 4, 54 5.50 2.49 RT 5, 6, 9, 29 5.06 2.30 201 5.38 2.44 203, 205, 207, 209 5.25 2.38 12 8.81 4.00 4, 54 7.19 3.26 TA 5, 6, 9, 29 6.75 3.06 201 7. 0 6 3.20 203, 205, 207, 209 6.94 3.15
- 1. Includes weight of JJ conduit seal.
- 2. E xcludes weight of ex ternal wire leads.
- 3. E xcludes weight of armored capillary, bulb ex tension, and sensing bulb on remote mount temperature switches.
Dimensions SALES PAGE 1 MODEL # SALES ORDER # LINE ITEM # PURCHASE ORDER #
1 194.1 7.64 77.5 1 1.5 138.1 HOLE PATTERN FOR 3.05 0.06 5.44 MOUNTING WITH 1/4 NPT 69.1 (3) STEEL U-BOLTS 42.9 HEX PLUG 2.72 1/4-20 X 44.51.75 1.69 VENT INSIDE WIDTH 23.2 CONNECTION 25.4 & (6) NUTS 0.92 1.00 (NOT SUPPLIED)
50.8 2.00 101.6 TYP 4.00
179.8 7.08
APPROXIMATE SWITCH WEIGHT 110.7 (EXCLUDES EXTERNAL WIRE LEADS) 4.36 8 LB 13 OZ ELECTRICAL 14685 W 105TH ST LENEXA, KS 66215 USA 913-888-2630 CONNECTION ISO-9001 SORINC.COM 3/4 NPTF DRAWN BY PRODUCT CERTIFICATION DRAWING K MITCHELL CHECKED BY ALL DIMENSIONS ARE +/-1/16 IN S BOAL UNLESS OTHERWISE SPECIFIEDMM ENGINEER APPROVAL LINEAR = IN J MODIG DATE RIGID MOUNTING PROCESS 24 MAY 2011 CONNECTION THIS DRAWING IS THE EXCLUSIVE PROPERTY OF SOR.
SURFACE NO USE WHATSOEVER OF THE INFORMATION CONTAINED 94.5 Drawing 8923112MADE WITHOUT THE EXPRESS WRITTEN PERMISSION OF SOR.HEREON, NOR REPRODUCTION IN WHOLE OR PART MAY BE 3.72 TITLE NOTES: NUCLEAR DIMENSION DRAWING Designator: TA12, 124 TA W/JJ EO NUMBER: 5045 DRAWING NUMBER REV 1 1. DIMENSION APPROXIMATE AND BASED ON A FIVE THREAD ENGAGEMENT. Piston Number 12 8923112 3 SCALE: 0.71 SHEET 1 OF 1 DWG SIZE
DO NOT SCALE PRINT B Model Name: 8923112.ASSEM/3/1 Reset Form Dimensions in this catalog are for reference only. They may be changed without notice. C ontact the factor y for cer tified drawings for a par ticular model number.
8 /16 913-888-2630 l SORInc.com Switches for the Nuclear Power Industr y DimensionsSALES PAGE1 MODEL # SALES ORDER # LINE ITEM # PURCHASE ORDER #
67.1 174.1 6.86 1 2.64 28.6 106.8 26.2 1.13 4.20 105.9 26.2 1.03 4.17 1.03 1
GLASS TO METAL SEAL 1/8 NPT WITH EPOXY HEX PLUG WIRE SUPPORT VENT ELECTRICAL CONNECTION CONNECTION 3/4 NPT(F) RIGID MOUNTING SURFACE 172.1 7.1 6.77 40.5 40.5 0.28 1.59 1.59 1/4-20 MOUNTING HARDWARE (NOT SUPPLIED)
55.2 2.17 2X CLEARANCE FOR 6.40.25 HARDWARE
WITH 63.52.5 MIN TO 76.23.00 MAX C/C MOUNTING SLOT
33.3 LENEXA, KS 66215 USA PROCESS 1.31 913-888-2630SORINC.COM CONNECTION 17.0 69.9 ISO-9001
DRAWN BY 0.67 2.75 K MITCHELL 94.5 PRODUCT CERTIFICATION DRAWING CHECKED BY ALL DIMENSIONS ARE 1/16 IN S ROOS 3.72 UNLESS OTHERWISE SPECIFIED ENGINEER APPROVAL LINEAR = MMIN J MODIG Linear = mm/in. DATE 04 DEC 2019 THIS DRAWING IS THE EXCLUSIVE PROPERTY OF SOR, INC.
NO USE WHATSOEVER OF THE INFORMATION CONTAINED HEREON,NOR REPRODUCTION IN WHOLE OR PART MAY BE MADE
WITHOUT THE EXPRESS WRITTEN PERMISSION OF SOR, INC.
TITLE Drawing 8923118NUCLEAR DIMENSION DRAWING 12, 124 RT W/JJ NOTES: APPROXIMATE SWITCH WEIGHT EO NUMBER: 5433 DRAWING NUMBER REV (EXCLUDES EXTERNAL WIRE LEADS)Designator: RT8923118 4 1 1. DIMENSION APPROXIMATE AND BASED ON A FIVE THREAD ENGAGEMENT SCALE: 0.71 DWG SIZE 7 LB 2 OZ DO NOT SCALE PRINT SHEET 1 OF 1 B Piston Number 12 Model Name: 8923118.ASSEM\\4.1\\2019-DEC-04 11:18 Reset Form
SALES PAGE 1 MODEL # SALES ORDER # LINE ITEM # PURCHASE ORDER #
1 163.4 6.43 1/8 NPT 95.6 49.2 HEX PLUG 3.76 VENT 1 104.6 1.94 CONNECTION 4.12 48.5 19.5 GLASS TO 1.91 0.77 METAL SEAL WITH EPOXY WIRE SUPPORT
RIGID MOUNTING SURFACE
153.2 6.03 65.1 ELECTRICAL 2.56 27.0 CONNECTION 1/4-20 MOUNTING 1.06 3/4 NPTF HARDWARE (NOT SUPPLIED)
125.2 APPROXIMATE SWITCH WEIGHT 2X 7.1 4.93 (EXCLUDES EXTERNAL WIRE LEADS) 0.28 5 LB 10 OZ 60.1 MOUNTING 2.37 HOLES 14685 W 105TH ST LENEXA, KS 66215 USA913-888-2630
ISO-9001 SORINC.COM
DRAWN BY PRODUCT CERTIFICATION DRAWING K MITCHELL CHECKED BY ALL DIMENSIONS ARE +/-1/16 IN S BOAL UNLESS OTHERWISE SPECIFIED ENGINEER APPROVAL 27.7 PROCESS LINEAR = INMM J MODIG 1.09 CONNECTION 6.3 DATE 8.7 63.5 0.25 24 MAY 2011 0.34 2.50 NO USE WHATSOEVER OF THE INFORMATION CONTAINEDTHIS DRAWING IS THE EXCLUSIVE PROPERTY OF SOR.
94.5 HEREON, NOR REPRODUCTION IN WHOLE OR PART MAY BE 3.72 MADE WITHOUT THE EXPRESS WRITTEN PERMISSION OF SOR.
TITLE NUCLEAR DIMENSION DRAWING NOTES: 12, 124 N6 W/JJ EO NUMBER: 5045 DRAWING NUMBER REV 1 1. DIMENSION APPROXIMATE AND BASED ON A FIVE THREAD ENGAGEMENT. Drawing 8923115 8923115 5 SCALE: 0.88 DWG SIZE DO NOT SCALE PRINT SHEET 1 OF 1 B
Model Name: 8923115.ASSEM/5/0 Designator: N6 Reset Form Piston Number 12
Dimensions in this catalog are for reference only. They may be changed without notice. C ontact the factor y for cer tified drawings for a par ticular model number.
SORInc.com l 913-888-2630 9 /16 Switches for the
SALES PAGE 1 Nuclear Power Industr y DimensionsMODEL # SALES ORDER # LINE ITEM # PURCHASE ORDER #
1 194.1 7.64 77.5 1 1.5 138.1 HOLE PATTERN FOR 3.05 0.06 5.44 MOUNTING WITH 1/4 NPT 69.1 (3) STEEL U-BOLTS 42.9 HEX PLUG 2.72 1/4-20 X 44.51.75 1.69 VENT INSIDE WIDTH 23.2 CONNECTION 25.4 & (6) NUTS 0.92 1.00 (NOT SUPPLIED)
50.8 2.00 101.6 TYP 4.00
179.8 7.08
APPROXIMATE SWITCH WEIGHT 110.7 (EXCLUDES EXTERNAL WIRE LEADS) 4.36 7 LB 3 OZ ELECTRICAL 14685 W 105TH ST LENEXA, KS 66215 USA 913-888-2630 ISO-9001 SORINC.COM CONNECTION DRAWN BY 3/4 NPTF PRODUCT CERTIFICATION DRAWING K MITCHELL CHECKED BY Linear = mm/in. ALL DIMENSIONS ARE +/-1/16 IN S BOAL UNLESS OTHERWISE SPECIFIED ENGINEER APPROVAL LINEAR = INMM J MODIG
DATE RIGID MOUNTING PROCESS 24 MAY 2011 THIS DRAWING IS THE EXCLUSIVE PROPERTY OF SOR.
CONNECTION NO USE WHATSOEVER OF THE INFORMATION CONTAINED SURFACE Drawing 8923111 HEREON, NOR REPRODUCTION IN WHOLE OR PART MAY BE 54.8 MADE WITHOUT THE EXPRESS WRITTEN PERMISSION OF SOR.
2.16 TITLE NOTES: Designator: TA NUCLEAR DIMENSION DRAWING 4, 54 TA W/JJ EO NUMBER: 5045 DRAWING NUMBER REV Piston Numbers 4, 54 8923111 3 1 SCALE: 0.71 SHEET 1 OF 1 DWG SIZE
- 1. DIMENSION APPROXIMATE AND BASED ON A FIVE THREAD ENGAGEMENT. DO NOT SCALE PRINT B Model Name: 8923111.ASSEM/3/2 SALES PAGE 1 MODEL # SALES ORDER # LINE ITEM # PURCHASE ORDER #Reset Form
67.1 174.1 2.64 6.86 1 28.6 106.8 26.2 26.2 1.13 4.20 105.9 1.03 1.03 4.17 1
1/8 NPT GLASS TO HEX PLUG METAL SEAL VENT WITH EPOXY CONNECTION ELECTRICAL WIRE SUPPORT CONNECTION 3/4 NPT(F)
RIGID MOUNTING SURFACE 172.1 7.1 6.77 40.5 40.5 0.28 1.59 1.59 1/4-20 MOUNTING HARDWARE (NOT SUPPLIED)
55.2 2X CLEARANCE FOR 6.4 2.17 0.25 HARDWARE WITH 63.52.5 MIN TO 76.23.00 MAX C/C MOUNTING SLOT
33.3 LENEXA, KS 66215 USA PROCESS 1.31 913-888-2630 CONNECTION 17.0 69.9 ISO-9001 SORINC.COM
DRAWN BY 0.67 2.75 K MITCHELL 54.8 PRODUCT CERTIFICATION DRAWING CHECKED BY 2.16 ALL DIMENSIONS ARE 1/16 IN S ROOS UNLESS OTHERWISE SPECIFIED ENGINEER APPROVAL LINEAR = MMIN J MODIG
DATE 04 DEC 2019 NO USE WHATSOEVER OF THE INFORMATION CONTAINED HEREON,THIS DRAWING IS THE EXCLUSIVE PROPERTY OF SOR, INC.
WITHOUT THE EXPRESS WRITTEN PERMISSION OF SOR, INC.NOR REPRODUCTION IN WHOLE OR PART MAY BE MADE Drawing 8923117TITLE NUCLEAR DIMENSION DRAWING 4, 54 RT W/JJ NOTES:
APPROXIMATE SWITCH WEIGHTDesignator: EO NUMBER: 5433RT DRAWING NUMBER REV 1 1. DIMENSION APPROXIMATE AND BASED ON A FIVE THREAD ENGAGEMENT (EXCLUDES EXTERNAL WIRE LEADS) 8923117 4 SCALE: 0.71 DWG SIZE Piston Numbers 4, 545 LB 8 OZ DO NOT SCALE PRINT SHEET 1 OF 1 B
Model Name: 8923117.ASSEM\\4.1\\2019-DEC-04 11:16 Reset Form
Dimensions in this catalog are for reference only. They may be changed without notice. C ontact the factor y for cer tified drawings for a par ticular model number.
10 /16 913-888-2630 l SORInc.com Switches for the Nuclear Power Industr y Dimensions
SALES PAGE 1 MODEL # SALES ORDER # LINE ITEM # PURCHASE ORDER #
1 163.4 6.43 1/8 NPT 95.6 HEX PLUGVENT 3.76 49.2 CONNECTION 1 104.6 1.94 4.12 19.5 GLASS TO 48.5 0.77 METAL SEAL 1.91 WITH EPOXY WIRE SUPPORT
RIGID MOUNTING SURFACE
18.8 153.2 65.1 0.74 6.03 ELECTRICAL 2.56 CONNECTION3/4 NPTF 1/4-20 MOUNTING 27.0 25.7 HARDWARE 1.06 1.01 (NOT SUPPLIED)
APPROXIMATE SWITCH WEIGHT 125.2 (EXCLUDES EXTERNAL WIRE LEADS) 2X 7.1 4.93 3 LB 15 OZ 21.1 0.28 14685 W 105TH ST LENEXA, KS 66215 USA913-888-2630 0.83 MOUNTING SORINC.COM 60.1 HOLES ISO-9001 DRAWN BY 2.37 K MITCHELL PRODUCT CERTIFICATION DRAWING CHECKED BY ALL DIMENSIONS ARE +/-1/16 IN S ROOS UNLESS OTHERWISE SPECIFIED ENGINEER APPROVAL LINEAR = INMM J MODIG
DATE 20 SEP 2012 27.7 THIS DRAWING IS THE EXCLUSIVE PROPERTY OF SOR.
NO USE WHATSOEVER OF THE INFORMATION CONTAINED 1.09 PROCESS 6.3 MADE WITHOUT THE EXPRESS WRITTEN PERMISSION OF SOR.HEREON, NOR REPRODUCTION IN WHOLE OR PART MAY BE 63.5 CONNECTION 0.25 TITLE Linear = mm/in.
NOTES: 2.50 NUCLEAR DIMENSION DRAWING 54.8 4, 54 N6 W/JJ 2.16 EO NUMBER: 5166 DRAWING NUMBER REV 1 1. DIMENSION APPROXIMATE AND BASED ON A FIVE THREAD ENGAGEMENT. 8923114 7 SCALE: 1.00 SHEET 1 OF 1 DWG SIZE
DO NOT SCALE PRINT B
Model Name: 8923114.ASSEM/6/0+
Drawing 8923114 Reset Form Designator: TA Designator: N6 Piston Numbers 4, 54 Piston Numbers 4, 54
SALES PAGE 1 MODEL # SALES ORDER # LINE ITEM # PURCHASE ORDER #
1 DIMENSION APPROXIMATE AND BASED ON A FIVE THREAD ENGAGEMENT 1 194.1 SALES PAGE 1 MODEL # SALES ORDER # LINE ITEM # PURCHASE ORDER #7.64
77.5 1 1.5 138.1 HOLE PATTERN FOR 3.05 0.06 5.44 MOUNTING WITH 1/8 NPT 1/4 MNPT 69.1 (3) STEEL U-BOLTS HEX PLUG142.9 HEX PLUG 163.4 2.72 VENT1.69 VENT 6.43 1/4-20 X 44.51.75 CONNECTION23.2 CONNECTION95.6 25.4 49.2 INSIDE WIDTH 0.92 3.76 1.00 1.94 19.5 & (6) NUTS 1 104.6 GLASS TO (NOT SUPPLIED) 4.12 0.77 METAL SEAL 48.5 WITH EPOXY 1.91 WIRE SUPPORT
RIGID
MOUNTING
SURFACE 50.8 2.00 101.6 CG TYP 4.00 179.5 ELECTRICALCG 2.5 7.07 CONNECTION 18.8 153.2 0.10 10.2 3/4 NPTF 65.1 0.741/4-20 MOUNTING 6.03 0.40 2.56 HARDWARE APPROXIMATE SWITCH WEIGHT 27.0 110.4 (NOT SUPPLIED) (EXCLUDES EXTERNAL WIRE LEADS) 1.06 4.35 25.7 6 LB 12 OZ 1.01 ELECTRICAL APPROXIMATE SWITCH WEIGHT14685 W 105TH ST LENEXA, KS 66215 USA (EXCLUDES EXTERNAL WIRE LEADS)14685 W 105TH ST LENEXA, KS 66215 USA913-888-2630 125.2 CONNECTION ISO-9001 913-888-2630 SORINC.COM 4.93 3/4 FNPT ISO-9001 3 LB 15 OZSORINC.COM DRAWN BY 2X 7.1 DRAWN BY K MITCHELL 21.1 0.28 PRODUCT CERTIFICATION DRAWINGK MITCHELL CHECKED BY 0.83 MOUNTING PRODUCT CERTIFICATION DRAWINGALL DIMENSIONS ARE CHECKED BY1/16 INS BOAL 60.1 HOLES ALL DIMENSIONS ARE +/-1/16 INUNLESS OTHERWISE SPECIFIEDUNLESS OTHERWISE SPECIFIEDS ROOSENGINEER APPROVAL 2.37 MM LINEAR = MMENGINEER APPROVAL J MODIG 5.1 LINEAR = IN IN J MODIG DATE PROCESS CONNECTION 0.20 Linear = mm/in. DATE 13 JAN 2017 RIGID MOUNTING 20 SEP 2012 THIS DRAWING IS THE EXCLUSIVE PROPERTY OF SOR.THIS DRAWING IS THE EXCLUSIVE PROPERTY OF SOR.
NO USE WHATSOEVER OF THE INFORMATION CONTAINEDNO USE WHATSOEVER OF THE INFORMATION CONTAINED SURFACE PROCESS 6.3 MADE WITHOUT THE EXPRESS WRITTEN PERMISSION OF SOR.HEREON, NOR REPRODUCTION IN WHOLE OR PART MAY BEMADE WITHOUT THE EXPRESS WRITTEN PERMISSION OF SOR.HEREON, NOR REPRODUCTION IN WHOLE OR PART MAY BE 27.7 CONNECTION 0.25 TITLE TITLE NUCLEAR DIMENSION DRAWINGNUCLEAR DIMENSION DRAWING NOTES: 1.09 4, 54 N6 W/JJ 5, 6, 9, 29 TA W/JJ 63.5 Drawing 8923110EO NUMBER: 5166 DRAWING NUMBER REVDRAWING NUMBER REV 1 2.50 EO NUMBER: 53568923114 78923110 4
- 1. DIMENSION APPROXIMATE AND BASED ON A FIVE THREAD ENGAGEMENT. SCALE: 1.00 SCALE: 0.71 DWG SIZE 54.8 DO NOT SCALE PRINT SHEET 1 OF 1 SHEET 1 OF 1B DWG SIZE
Model Name: 8923114.ASSEM/6/0+ Designator: 2.16 TA DO NOT SCALE PRINT B Model Name: 8923110.ASSEM\\4.1\\2017-JAN-17 11:55 Reset Form Reset Form Piston Numbers 5, 6, 9, 29
Dimensions in this catalog are for reference only. They may be changed without notice. C ontact the factor y for cer tified drawings for a par ticular model number.
SORInc.com l 913-888-2630 11 / 1 6 Switches for the Nuclear Power Industr y Dimensions
SALES PAGE 1 MODEL # SALES ORDER # LINE ITEM # PURCHASE ORDER #
67.1 174.1 2.64 6.86 1 28.6 106.8 26.2 26.2 1.13 4.20 105.9 1.03 1.03 4.17 1
GLASS TO METAL SEAL 1/8 NPT WITH EPOXY WIRE SUPPORT HEX PLUG ELECTRICAL VENT CONNECTION CONNECTION 3/4 NPT(F) RIGID MOUNTING SURFACE
171.8 6.76 40.5 40.5 1/4-20 1.59 1.59 MOUNTING HARDWARE (NOT SUPPLIED)
54.9 2X CLEARANCE FOR 6.40.25 HARDWARE 2.16 WITH 63.52.5 MIN TO 76.23.00 MAX C/C MOUNTING SLOT
PROCESS 33.3 LENEXA, KS 66215 USA 7.1 CONNECTION 1.31 913-888-2630 0.28 17.0 69.9 ISO-9001 SORINC.COM 0.67 2.75 DRAWN BY PRODUCT CERTIFICATION DRAWING K MITCHELL CHECKED BY ALL DIMENSIONS ARE 1/16 IN S ROOS UNLESS OTHERWISE SPECIFIED ENGINEER APPROVAL Linear = mm/in. LINEAR = MMIN J MODIG
DATE 04 DEC 2019 THIS DRAWING IS THE EXCLUSIVE PROPERTY OF SOR, INC.
NO USE WHATSOEVER OF THE INFORMATION CONTAINED HEREON,NOR REPRODUCTION IN WHOLE OR PART MAY BE MADE WITHOUT THE EXPRESS WRITTEN PERMISSION OF SOR, INC.
Drawing 8923116TITLE NUCLEAR DIMENSION DRAWING 5, 6, 9, 29 RT W/JJ NOTES:
APPROXIMATE SWITCH WEIGHT EO NUMBER: 5433 DRAWING NUMBER REV 1 1. DIMENSION APPROXIMATE AND BASED ON A FIVE THREAD ENGAGEMENT Designator: (EXCLUDES EXTERNAL WIRE LEADS)RT 8923116 4 SCALE: 0.71 DWG SIZE 5 LB 1 OZ DO NOT SCALE PRINT SHEET 1 OF 1 B Piston Numbers 5, 6, 9, 29 Model Name: 8923116.ASSEM\\4.1\\2019-DEC-04 11:14 Reset Form
SALES PAGE 1 MODEL # SALES ORDER # LINE ITEM # PURCHASE ORDER #
1 163.4 6.43 1/8 NPT 95.6 49.2 HEX PLUG 3.76 1.94 VENT 1 104.6 CONNECTION 4.12 48.5 19.5 GLASS TO 1.91 0.77 METAL SEAL WITH EPOXY WIRE SUPPORT
RIGID MOUNTING SURFACE
152.9 6.02 65.1 ELECTRICAL 2.56 CONNECTION 1/4-20 MOUNTING 27.0 3/4 NPTF HARDWARE 1.06 (NOT SUPPLIED)
APPROXIMATE SWITCH WEIGHT 124.9 (EXCLUDES EXTERNAL WIRE LEADS) 4.92 3 LB 9 OZ 2X 7.1 14685 W 105TH ST LENEXA, KS 66215 USA 59.8 0.28 913-888-2630 2.35 MOUNTING ISO-9001 SORINC.COM HOLES DRAWN BY PRODUCT CERTIFICATION DRAWING K MITCHELL CHECKED BY ALL DIMENSIONS ARE +/-1/16 IN S BOAL UNLESS OTHERWISE SPECIFIED ENGINEER APPROVAL LINEAR = INMM J MODIG
DATE 24 MAY 2011 27.7 PROCESS NO USE WHATSOEVER OF THE INFORMATION CONTAINEDTHIS DRAWING IS THE EXCLUSIVE PROPERTY OF SOR.
1.09 CONNECTION HEREON, NOR REPRODUCTION IN WHOLE OR PART MAY BE 63.5 MADE WITHOUT THE EXPRESS WRITTEN PERMISSION OF SOR.Linear = mm/in.
2.50 6.3 TITLE NOTES: 0.25 NUCLEAR DIMENSION DRAWING 5, 6, 9, 29 N6 W/JJ EO NUMBER: 5045 DRAWING NUMBER REV 1 1. DIMENSION APPROXIMATE AND BASED ON A FIVE THREAD ENGAGEMENT. 8923113 6 SCALE: 1.00 DWG SIZE DO NOT SCALE PRINT SHEET 1 OF 1 B Drawing 8923113 Model Name: 8923113.ASSEM/6/0 Reset Form Designator: N6 Piston Numbers 5, 6, 9, 29
Dimensions in this catalog are for reference only. They may be changed without notice. Contact the factory for certified drawings for a particular model number.
12 /16 913-888-2630 l SORInc.com Switches for the Nuclear Power Industr y Dimensions SALES PAGE 1 MODEL # SALES ORDER # LINE ITEM # PURCHASE ORDER #
STANDARD MODELS 1 194.1 HOLE PATTERN FOR BULB
- APPROX LENGTH E 7.64 MOUNTING WITH MODEL # WEIGHT A-B C D 77.5 1 1.5 138.1 (3) STEEL U-BOLTS 3.05 0.06 5.44 1/4-20 X 44.5 203 1829 112.0 125.7 381.8 TO 1.75 72 4.41 4.95 15.03 23.2 1/4 NPT 69.1 INSIDE WIDTH 205 3048124.7138.4 394.5 TO 0.92 HEX PLUG 2.72 (NOT SUPPLIED) 6 LB 15 OZ 120 4.91 5.45 15.539.7 VENT 25.4 207 4572162.8176.5 432.6 TO 0.38 CONNECTION 1.00 23.2 180 6.41 6.95 17.03 0.92 194.6 208.3 464.3 209 60962407.668.20 18.28 TO
- WEIGHT EXCLUDES CAPILLARY, BULB AND EXTERNAL WIRE LEADS NON-STANDARD MODEL DIMENSIONS
50.8 101.6 2.00 4.00 TYP 6X 1/4-20 191.6 MOUNTING 7.54 NUTS (NOT SUPPLIED) 122.5 4.82 ELECTRICAL GLASS TO CONNECTION METAL SEAL 3/4 NPTF WITH EPOXY WIRE SUPPORT RIGID MOUNTING 52.3 SURFACE 2.06 POINT A 306.3 C 14685 W 105TH ST LENEXA, KS 66215 USA 12.06 913-888-2630 ISO-9001 SORINC.COM POINT B 22.4 TEMPERATURE DRAWN BY 0.88 HEX SENSING BULB PRODUCT CERTIFICATION DRAWING K MITCHELL CHECKED BY Drawing 8923124ALL DIMENSIONS ARE +/-1/16 IN S BOAL UNLESS OTHERWISE SPECIFIED ENGINEER APPROVAL LINEAR = INMM J MODIG ARMORED Designator: TADATE CAPILLARY BULB 1/2 NPTM E 24 MAY 2011 EXTENSION GLAND NUT NO USE WHATSOEVER OF THE INFORMATION CONTAINEDTHIS DRAWING IS THE EXCLUSIVE PROPERTY OF SOR.
D (TO WRENCH TIGHT) Remote MADE WITHOUT THE EXPRESS WRITTEN PERMISSION OF SOR.HEREON, NOR REPRODUCTION IN WHOLE OR PART MAY BE
TITLE Temperature NUCLEAR DIMENSION DRAWING NOTES: 203, 205, 207, 209 TA W/JJ SwitchEO NUMBER: 5045 DRAWING NUMBER REV 1 1. DIMENSION APPROXIMATE AND BASED ON A FIVE THREAD ENGAGEMENT. 8923124 3 SCALE: 0.47 DWG SIZE DO NOT SCALE PRINT SHEET 1 OF 1 B
Model Name: 8923124.ASSEM/3/2 Reset Form
SALES PAGE 1 MODEL # SALES ORDER # LINE ITEM # PURCHASE ORDER #
STANDARD MODELS 67.1 174.1 BULB
- APPROX LENGTH E 2.64 1 6.86 26.2 MODEL # WEIGHT A-B C D 28.6 1/8 NPT 106.8 GLASS TO 1.03 203 1829 112.0 125.7 1.13 HEX PLUG 4.20 METAL SEAL 72 4.41 4.95 TO 381.815.03 26.2 VENT 105.9 WITH EPOXY 205 3048124.7138.4 1.03 CONNECTION 4.17 1 WIRE SUPPORT 5 LB 120 4.91 5.45 TO 394.515.53 9.7 207 45724 OZ162.8176.50.38 180 6.41 6.95 TO 432.617.03 209 6096194.6208.3 240 7.66 8.20 TO 464.318.28
- WEIGHT EXCLUDES CAPILLARY, BULB AND EXTERNAL WIRE LEADS 183.8 NON-STANDARD MODEL DIMENSIONS 7.24
NOTES:
40.5 40.5 1. DIMENSION APPROXIMATE AND BASED ON A FIVE THREAD 1 1.59 1.59 ENGAGEMENT
ELECTRICAL 1/4-20 MOUNTING CONNECTION HARDWARE 66.9 3/4 NPT(F) (NOT SUPPLIED) 2.64
2X CLEARANCE FOR 6.40.25 HARDWARE RIGID MOUNTING SURFACE WITH 63.52.5 MIN TO 76.23.00 MAX C/C 52.3 17.0 MOUNTING SLOT LENEXA, KS 66215 USA 0.67 913-888-2630 2.06 Linear = mm/in.ISO-9001 SORINC.COM POINT A 33.3 TEMPERATURE DRAWN BY SENSING BULB PRODUCT CERTIFICATION DRAWING K MITCHELL 7.1 1.31 ALL DIMENSIONS ARE 1/16 IN ENGINEER CHECK 0.28 69.9 POINT B C306.3UNLESS OTHERWISE SPECIFIEDS ROOS ENGINEER APPROVAL 2.75 12.06 LINEAR = MMIN J MODIG
DATE Drawing 892312604 DEC 2019 NO USE WHATSOEVER OF THE INFORMATION CONTAINED HEREON,THIS DRAWING IS THE EXCLUSIVE PROPERTY OF SOR, INC.
WITHOUT THE EXPRESS WRITTEN PERMISSION OF SOR, INC.NOR REPRODUCTION IN WHOLE OR PART MAY BE MADE ARMORED BULB EXTENSION Designator: TITLE RT CAPILLARY 22.4 E NUCLEAR DIMENSION DRAWING 203, 205, 207, 209 RT 0.88 HEX Remote W/JJ DRAWING NUMBER REV EO NUMBER: 5433 1/2 NPT(M) GLAND NUT D (TO WRENCH TIGHT) 8923126 4 Temperature SCALE: 0.61 DWG SIZE DO NOT SCALE PRINT SHEET 1 OF 1 B
Model Name: 8923126.ASSEM\\4.1\\2019-DEC-04 11:31 Switch Reset Form
Dimensions in this catalog are for reference only. They may be changed without notice. Contact the factory for certified drawings for a particular model number.
SORInc.com l 913-888-2630 13 /16 Switches for the Nuclear Power Industr y Dimensions SALES PAGE 1 MODEL # SALES ORDER # LINE ITEM # PURCHASE ORDER #
STANDARD MODELS 163.4 BULB
- APPROX LENGTH 6.43 1 MODEL # WEIGHT A-B C DE 1/8 NPT 95.6 49.2 1829 112.0 125.7 HEX PLUG 3.77 1.94 203 72 4.41 4.95 TO 381.815.03 19.5 VENT 104.6 19.5 0.77 CONNECTION 4.12 1 0.77 205 3048124.7138.4 3 LB 120 4.91 5.45 TO 394.515.539.7 207 457212 OZ162.8176.50.38 180 6.41 6.95 TO 432.617.03 209 6096194.6208.3 240 7.66 8.20 TO 464.318.28 6.3 165.0
- WEIGHT EXCLUDES CAPILLARY, BULB AND EXTERNAL WIRE LEADS 0.25 6.49 65.1 NON-STANDARD MODEL DIMENSIONS ELECTRICAL 2.56 27.0 CONNECTION 1.06 3/4 NPT(F)
NOTES:
- 1. DIMENSION APPROXIMATE AND BASED ON A FIVE THREAD 1 71.9 2X 7.1 GLASS TO ENGAGEMENT 2.83 0.28 METAL SEAL MOUNTING WITH EPOXY HOLE WIRE SUPPORT 1/4-20 MOUNTING HARDWARE (NOT SUPPLIED)
RIGID MOUNTING 52.3 POINT A SURFACE 2.06
LENEXA, KS 66215 USA 913-888-2630 27.7 ISO-9001 SORINC.COM 1.09 DRAWN BY 8.7 63.5 TEMPERATURE SENSING BULB PRODUCT CERTIFICATION DRAWING K MITCHELL ENGINEER CHECK 0.34 2.50 Drawing 8923122ALL DIMENSIONS ARE UNLESS OTHERWISE SPECIFIED1/16 INS ROOS POINT B C306.3ENGINEER APPROVAL 12.06 LINEAR = MMIN J MODIG
DATE Designator: N604 DEC 2019 NO USE WHATSOEVER OF THE INFORMATION CONTAINED HEREON,THIS DRAWING IS THE EXCLUSIVE PROPERTY OF SOR, INC.
Remote WITHOUT THE EXPRESS WRITTEN PERMISSION OF SOR, INC.NOR REPRODUCTION IN WHOLE OR PART MAY BE MADE
TITLE ARMORED BULB 22.4 1/2 NPT(M) E Temperature NUCLEAR DIMENSION DRAWING 203, 205, 207, 209 N6 CAPILLARY EXTENSION 0.88 HEX GLAND NUT W/JJ NOTES: SwitchEO NUMBER: 5433 DRAWING NUMBER REV 8923122 6 1 1. DIMENSION APPROXIMATE AND BASED ON A FIVE THREAD ENGAGEMENT. D (TO WRENCH TIGHT) SCALE: 0.56 DWG SIZE DO NOT SCALE PRINT SHEET 1 OF 1 B
Model Name: 8923122.ASSEM\\6.1\\2019-DEC-04 11:23 Reset Form
SALES PAGE 1 MODEL # SALES ORDER # LINE ITEM # PURCHASE ORDER #
STANDARD MODELS 1 1BULB LENGTH BULB* APPROX 1.5 194.0 C E WEIGHT 1/4 NPT 0.06 7.64 1 105.7 9.7 HEX PLUG 125.0 4.16 0.38 7 LB 1 OZ VENT 4.92
- WEIGHT EXCLUDES EXTERNAL WIRE LEADS CONNECTION NON-STANDARD 138.1 HOLE PATTERN FOR MODEL DIMENSIONS 77.5 5.44 MOUNTING WITH 3.05 69.1 (3) STEEL U-BOLTS 42.9 2.72 1/4-20 X 44.51.75 1.69 23.2 25.4 INSIDE WIDTH 23.2 0.92 1.00 & NUTS 0.92 (NOT SUPPLIED)
GLASS TO 101.6 METAL SEAL 4.00 WITH EPOXY 50.8 WIRE SUPPORT 2.00 TYP RIGID MOUNTING SURFACE 206.9 8.14
137.8 ELECTRICAL 5.43 CONNECTION 31.8 3/4 NPTF 14685 W 105TH ST LENEXA, KS 66215 USA 1.25 HEX 913-888-2630 ISO-9001 SORINC.COM 1/2 NPTM Linear = mm/in. DRAWN BY PROCESS PRODUCT CERTIFICATION DRAWING K MITCHELL CONNECTION ALL DIMENSIONS ARE +/-1/16 IN CHECKED BY UNLESS OTHERWISE SPECIFIED S BOAL ENGINEER APPROVAL TEMPERATURE LINEAR = INMM J MODIG SENSING BULB Drawing 8923123DATE C E 24 MAY 2011 NO USE WHATSOEVER OF THE INFORMATION CONTAINEDTHIS DRAWING IS THE EXCLUSIVE PROPERTY OF SOR.
Designator: MADE WITHOUT THE EXPRESS WRITTEN PERMISSION OF SOR.HEREON, NOR REPRODUCTION IN WHOLE OR PART MAY BETA
TITLE Direct NUCLEAR DIMENSION DRAWING NOTES: 201 TA W/JJ EO NUMBER: 5045 DRAWING NUMBER REV 1 1. DIMENSION APPROXIMATE AND BASED ON A FIVE THREAD ENGAGEMENT. Temperature 8923123 3 SCALE: 0.60 DWG SIZE DO NOT SCALE PRINT SHEET 1 OF 1 Switch B Model Name: 8923123.ASSEM/3/0 Reset Form
Dimensions in this catalog are for reference only. They may be changed without notice. Contact the factory for certified drawings for a particular model number.
14 / 16 913-888-2630 l SORInc.com Switches for the Nuclear Power Industr y Dimensions SALES PAGE 1 MODEL # SALES ORDER # LINE ITEM # PURCHASE ORDER #
STANDARD MODELS BULB LENGTH BULB
- APPROXIMATE C E WEIGHT 67.1 174.1 105.7 9.7 2.64 6.86 1 4.16 0.38 5 LB 6 OZ 28.2 1/8 NPT 106.8 ELECTRICAL
- WEIGHT EXCLUDES EXTERNAL WIRE LEADS 1.11 HEX PLUG 4.20 CONNECTION NON-STANDARD MODEL DIMENSIONS 26.2 VENT 105.9 3/4 NPT(F) 26.2 1.03 CONNECTION 4.17 1 1.03
NOTES: 1
- 1. DIMENSION APPROXIMATE AND BASED ON A FIVE THREAD ENGAGEMENT
GLASS TO METAL SEAL WITH EPOXY WIRE SUPPORT
RIGID MOUNTING SURFACE
40.5 40.5 199.1 1.59 1.59 7.84 1/4-20 MOUNTING HARDWARE (NOT SUPPLIED)
7.1 2X CLEARANCE FOR 6.40.25 HARDWARE 0.28 82.2 3.24 WITH 63.52.5 MIN TO 76.23.00 MAX C/C LENEXA, KS 66215 USA 913-888-2630 MOUNTING SLOT ISO-9001 SORINC.COM
DRAWN BY PRODUCT CERTIFICATION DRAWING K MITCHELL 31.8 ALL DIMENSIONS ARE 1/16 IN ENGINEER CHECK 1.25 HEX UNLESS OTHERWISE SPECIFIED S ROOS ENGINEER APPROVAL 17.0 69.9 LINEAR = MMIN J MODIG 0.67 2.75 Drawing 8923125DATE C 1/2 NPT(M) 04 DEC 2019 E PROCESS THIS DRAWING IS THE EXCLUSIVE PROPERTY OF SOR, INC.
NO USE WHATSOEVER OF THE INFORMATION CONTAINED HEREON,NOR REPRODUCTION IN WHOLE OR PART MAY BE MADE CONNECTION Designator: WITHOUT THE EXPRESS WRITTEN PERMISSION OF SOR, INC.RT
TITLE TEMPERATURE Direct NUCLEAR DIMENSION DRAWING 201 RT W/JJ NOTES: SENSING BULB Temperature EO NUMBER: 5433 DRAWING NUMBER REV 1 1. DIMENSION APPROXIMATE AND BASED ON A FIVE THREAD ENGAGEMENT. SCALE: 0.70 8923125 4 DWG SIZE SHEET 1 OF 1 SwitchDO NOT SCALE PRINT B
Model Name: 8923125.ASSEM\\4.1\\2019-DEC-04 11:25 Reset Form
SALES PAGE 1 MODEL # SALES ORDER # LINE ITEM # PURCHASE ORDER #
STANDARD MODELS BULB LENGTH BULB
- APPROXIMATE 163.4 C E WEIGHT 6.43 1 105.7 9.7 49.2 1/8 NPT 95.6 4.16 0.38 3 LB 14 OZ 1.94 HEX PLUG 3.77
- WEIGHT EXCLUDES EXTERNAL WIRE LEADS 19.5 VENT 104.6 19.5 NON-STANDARD MODEL DIMENSIONS 0.77 CONNECTION 4.12 1 0.77
NOTES: 1
- 1. DIMENSION APPROXIMATE AND BASED ON A FIVE THREAD ENGAGEMENT
65.1 ELECTRICAL 2.56 GLASS TO METAL SEAL CONNECTION WITH EPOXY WIRE SUPPORT 27.0 3/4 NPT(F) 1.06 180.3 RIGID MOUNTING 7.10 SURFACE
1/4-20 MOUNTING HARDWARE 6.3 2X 7.1 (NOT SUPPLIED) 0.25 0.28 87.2 MOUNTING 3.43 HOLE 31.8 1.25 HEX LENEXA, KS 66215 USA 1/2 NPT(M) 913-888-2630 PROCESS Linear = mm/in.ISO-9001 SORINC.COM
DRAWN BY CONNECTION K MITCHELL PRODUCT CERTIFICATION DRAWING ENGINEER CHECK ALL DIMENSIONS ARE 1/16 IN S ROOS TEMPERATURE UNLESS OTHERWISE SPECIFIED ENGINEER APPROVAL C SENSING BULB Drawing 8923121LINEAR = MMIN J MODIG E DATE 05 DEC 2019 NO USE WHATSOEVER OF THE INFORMATION CONTAINED HEREON,THIS DRAWING IS THE EXCLUSIVE PROPERTY OF SOR, INC.
Designator: NOR REPRODUCTION IN WHOLE OR PART MAY BE MADEN6 WITHOUT THE EXPRESS WRITTEN PERMISSION OF SOR, INC.
27.7 Direct TITLE NOTES: 1.09 NUCLEAR DIMENSION DRAWING 201 N6 W/JJ 1 1. DIMENSION APPROXIMATE AND 8.7 63.5 TemperatureEO NUMBER: 5433 DRAWING NUMBER REV BASED ON A FIVE THREAD ENGAGEMENT. 0.34 2.50 8923121 6 SCALE: 0.75 DWG SIZE Switch SHEET 1 OF 1 DO NOT SCALE PRINT B
Model Name: 8923121.ASSEM\\6.1\\2019-DEC-04 11:21 Reset Form
Dimensions in this catalog are for reference only. They may be changed without notice. Contact the factory for certified drawings for a particular model number.
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16 /16 913-888-2630 l SORInc.com Page 1 of 1 Calculation/Evaluation Verification Form
DOCUMENT TITLE: Unit 1 Main Turbine Anticipatory Reactor Trip Setpoint Basis PROJECT NO.: ANO-490467
DOCUMENT NO.: CALC-22-E-0007-01 CLIENT REV.: 0 INTERNAL REV.: 0.004
CLIENT: Entergy UNIT: ANO 1 DELIVERABLE QUALITY CLASS: SR AQ NSR
VERIFICATION METHOD: DESIGN REVIEW ALT. CALC TESTING NOT REQUIRED
REVIEWER TO COMPLETE THE FOLLOWING ITEMS: YES NO N/A
- 1. Is the purpose/objective clearly stated?
- 2. Are all documents prepared in a clear, legible manner and in a proper format?
- 3. Were the design inputs/sources correctly selected (including Rev. / Edition) and incorporated into the design?
- 4. Have the applicable codes, standards and regulatory requirements been properly identified, applied and reflected?
- 5. Are the necessary design inputs for interfacing organizations specified in the design documents or in supporting procedures or instructions?
- 6. Are assumptions necessary to perform the design activity/activities? If yes:
A. Are the assumptions adequately described and reasonable?
B.Do any assumptions require later verification?
- 7. Are the acceptance criteria incorporated in the design/analysis documents sufficient to allow verification that design/analysis requirements have been satisfactorily accomplished?
- 8. Were appropriate design or analytical methods used?
- 9. Is the output reasonable compared to the inputs?
- 10. Was a computer program used? If yes, identify below:
A. Was the program name, version / rev., and the Computer ID on which the program was run identified in the design document? If yes, identify below:
Program name, Version/Rev:
Computer ID:
Program run date, File name:
B. Is the program listed on the Kinectrics ASL and has the SRN been reviewed for any program use?
C. Have existing user notices and/or error reports for the software/version been reviewed? If yes -
D. Were they applicable to the program run?
- 11. Was an alternate calculation performed to verify the design?
If yes, attach alternate calculation and/or identify alternate calculation number.
- 12. Was qualification testing performed to verify the design? If yes, does the testing demonstrate that the design or specific design feature:
A. Meets the intended performance under the most adverse design conditions?
B. Performs the intended function in all applicable operating modes?
- 13. If qualification testing is performed to verify the design, A. Are the test procedures and results adequately documented?
B. Is the test configuration clearly defined and documented?
Summary of Verification:
A detailed review of the calculation was performed to verify that the uncertainty addresses the applicable sources of uncertainty in the setpoint and that the resultant setpoint for the subject switches would be acceptable for performing the intended function. The setpoint was adjusted per client request. All comments were captured within the document and have been satisfactorily addressed.
Qualified Design Verifier Name/Signature: William G. Sterrett Date: 2/8/2023 (Attach Comment Control Form)
Form 3.6-1 Rev. 2 Kinectrics AES Inc.