ML20246D223
| ML20246D223 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 06/30/1989 |
| From: | Shelton D TOLEDO EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| TAC-59086, NUDOCS 8907110262 | |
| Download: ML20246D223 (35) | |
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TOLEDO i Docket Number 50-346 EDISON
< License Number NPF-3 ^ '* * ' '"'
- C""* "Y Serial Number 1682 DONALD C. SHELTON Vce Presutent-Nuctor (419]249 2300 June 30, 1989 D
United States Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555
Subject:
Revision to Plant-Specific Submittal for ATVS Implementation (10CFR50.62) (TAC Number 59086)
Gentlemen:
On February 28, 1989 (Serial Number 1638), Toledo Edison submitted a l plant-specific " Design Summary" which provided Davis-Besse's conceptual design
. requirements for meeting the Anticipated Trar.sients Without Scram (ATVS) rule (10CFR50.62). By letter dated May 3, 1989 (Log Number 2916), the Nuclear Regulatory Commission (NRC) requested Toledo Edison to provide' additional information to allow completion of the Davis-Besse's ATVS implementation review.
The matrix provided in Attachment I correlates the additional information requested by the NRC to the revised section of the Design Summary which provides the information. The revised Design Summary provided in Attachment 2
' incorporates the additional information requested and supersedes the Design Summary provided in Toledo Edison's previous submittal. Part 3 has not changed but has been included for completeness.
Toledo Edison plans to design and install the ATVS systems by the end of the Sixth Refueling Outage, which is currently scheduled to begin in February 1990. If you have any questions concerning this matter, please contact Mr. R.
V. Schrauder, Nuclear Licensing Manager, at (419) 249-2366.
Very truly yours,
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EBS/dlm Attachments cc: P. M. Byron, DB-1 NRC Resident Inspector A. B. Davis, Regional Administrator, NRC Region III T. V. Vambach, DB-1 NRC Senior Project Manager 8907110262 890630 6 p[.$
PDR ADOCK 0500(
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,I THE TOLEOO EDISON COMPANY ED! SON PLAZA 300 MADISON AVENUE TOLEDO, OHIO 436
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i-Attachment 1 REQUEST FOR ADDITIONAL INFORMATION 1
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. '.Dockat Numbar'50-346
'~- Licen32 Numbtr NPF-3 Serial Number 1682 Attachment 1 Page 1 TABLE 1 Request for Additional Information NRC Ouestion TE Response
- 1. Diversity from Existing RPS Toledo Edison must further describe Further. detail is provided in Part how diversity is to be achieved 1, Section 5.1. The NRC advised between the DSS and RTS equipment... Toledo Edison during a telephone Adequate diversity is to be achieved conference call held on June 13, by the use of components from 1989, that this was not a concern, different manufacturers / and that the NRC had no problem with manufacturing processes, the use of -Toledo Edison using existing mechanical versus electronic devices electromechanical devices of the that uses different principles of same model number.
operation.... This information should be included in the Toledo Edison final, plant-specific Davis-Besse submittal in order for the staff to make a determination of compliance with the ATWS Rule.
- 2. Electrical Independence from the Existing RPS Toledo Edison must ensure that the Reference Part 1, Sections 5.2 and DSS /AMSAC applications are bounded 5.6 for DSS. As discussed in the by the previously-documented testing June 13, 1989, telephone conference and so state in the final call, this question does not apply Davis-Besse plant-specific to AMSAC since AMSAC will be a Class submittal. IE system and there are no 1E to non-1E interfaces.
- 3. Physical Separation from Existing RPS To allow the staff to determine if Reference Part I and 2, Section 5.3 this part of the Davis-Besse design complies with the current approved plant design requirements, specific details on component location and physical separation should be j supplied in the plant specific 1 submittal.
. ' , Dockst LicinsaNuxbtr Nu;btr50-346' NPF-3 '
Serial Number 1682 Attachment 1 Page 2 TABLE 1 Request for Additional Information NRC Ouestion TE Response
- 4. Environmental Qualification (EO) and Quality Assurance (0A) for Testing, Maintenance, and Surveillance In its plant-specific submittal, Reference Part 1 and 2, Section 5.5 Toledo Edison should provide a description of the measures /
programs implemented for Davis-Besse to assure that the equipment diversity provided in accordance with the ATVS Rule vill be maintained during component repair, replacement, and modifications and/or design changes, etc.
throughout the life of the plant.
- 5. Safety-Related (lE) Power Supplies For the staff to make a final Reference Part 2 Sections 5.1, 5.2 determination of acceptance, the and 5.6 concerns noted in the diversity and independence sections with respect to the use of SFRCS equipment in ATVS designs should be addressed.
- 6. Testability at Power
... the plant specific submittal Reference Parts 1 and 2 Sections 5.7 should also address the time limits associated with channel testing, disabling of channels, actions to be taken if one channel fails, etc.
- 7. Maintenance Bypasses, Operating Bypasses, Indication of Bypasses, j and Means for Bypassing i The Davis-Besse " conceptual design" Reference Part 2, Sections 5.9 and 5.10 docs not provide information describing the use of operational bypasses, the means of bypassing, or any indications of bypass conditions for the AMSAC. Therefore, Toledo Edison should include this information in the plant-specific design submittal for Davis-Besse.
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',',Dodb.atNurbur50-346 Licinsa.Nulb2r NPF-3'
. Serial Number 1682
-Attachment 1 TABLE 1 Page 3 Request for Additional Information L NRC Ouestion TE Response
- 8. Completion of Protective Action l Toledo Edison should provide Reference Parts 1 and 2, Section
): specific information which confirms 5.13 that'both the DSS and AMSAC at Davis-Besse are designed.such that, upon receipt'of a trip signal, the protective action goes to completion and deliberate operator action is required Lto reset the systems in order to comply with the ATVS Rule.
In addition to the specific information on the system's design, Toledo Edison should include a discussion of any required operator actions.
- 9. Information Readout Therefore, the Toledo Edison Reference Parts 1 and 2, Section plant-specific submittal for 5.14 Davis-Besse should provide more detailed information relating to how the operator'is provided.vith accurate, complete, and timely information (i.e., what actuates or deactuates alarms, annunciators, lights, and what functions are performed by specific switches, etc.) pertinent to the DSS /AMSAC system status. In addition, Toledo Edison should provide a discussion of how human-factors engineering practices are incorporated into the
' design of ATVS prevention / mitigation system components located in the control room. The coordination of displays used to provide the status of ATVS systems / equipment to the-operator with existing displays should be addressed specifically.
- 10. Safety-Related Interfaces These concerns (i.e., the sharing of Reference Part 2. Section 5.15 power supplies via SFRCS/AMSAC and the adequacy of isolation devices) should be adequately addressed in the plant-specific submittal in order for the staff to evaluate the applicability of the devices for use in the DSS /AMSAC systems.
t-4 Attachment 2 DESIGN
SUMMARY
L , ' Dock 2t Numb r 50-346 Liccnsa Numb:r NPF-3 Serial Number 1682
!. Attachment 2 Page 1 DESIGN
SUMMARY
(Revision 1)
Toledo Edison's conceptual design for equipment required to address the Anticipated Transient Vithout Scram (ATVS) rule relative to the design requirements specified in Section 5 of the NRC's Safety' Evaluation Report (as transmitted by Reference 1 dated August 10, 1988) is provided in Parts 1, 2, and 3 below. Part 1 provides a description of the Diverse Scram System (DSS).
Part 2 includes the basis and justification for using the existing Steam and Feedvater Rupture Control System (SFRCS) to satisfy the ATUS Mitigation System Actuation Circuitry (AMSAC) function. Part 2 also includes the design requirements for AMSAC. Part 3 provides an evaluation to demonstrate that common mode failures vill not propagate through the power supplies and disable both SFRCS and the RTS.
Part 1 DESCRIPTION OF THE DIVERSE SCRAM SYSTEM _(DSS)
A conceptual functional diagram of Davis-Besse's proposed Diverse Scram System (DSS) is provided in Figure 1. The DSS will consist of two channels of instrumentation, each having a reactor coolant pressure input to a bistable with'a trip setpoint of approximately 2450 psig. The bistable output will be a contact closure that energizes new DSS relays in the Control Rod Drive Control System (CRDCS) cabinets. The DSS relay output vill open programmer lamp circuits causing de-gating of one group of Silicon Controlled Rectifiers (SCR). A coincident second DSS channel actuation vill de-gate a second group of SCRs, thus removing the power from the Control Rod Drive Mechanisms and allowing the control rods to drop into the reactor core. Both CRDCS groups' (channels "A" and "B") SCRs must de-energize to release the control rods.
Control Room indication of a trip initiated by DSS will be through an annunciator alarm entitled " DSS TRIP" on the main control board. In addition, all " rod bottom" lights vill be lit on the Control Rod Position Indication Panel in the Control Room, and the "A" & "B" lamp fault lights vill illuminate on the " Diamond Operator's Control Panel (See Figure 4b).
The system vill be designed to be testable with the reactor on-line. While one channel is being tested, a DSS bypass switch vill be actuated to prevent an inadvertent reactor trip. A means to alert the operators that a DSS channel is in a bypass or tripped condition vill be provided (See Figure la).
DESIGN REQUIREMENTS FOR THE DSS This section presents the specific design requirements Toledo Edison plans to use to fulfill the design and implementation criteria for the Diverse Scram System. For convenience, the paragraphs are numbered to coincide with comparable paragraphs of the NRC's Safety Evaluation Report. The generic design requirements are also addressed. Most of the generic design requirements have been addressed, at least in part, by the Babcock and Vilcox Owners Group (BWOG) " Design Requirements for DSS and AMSAC" document. Since Toledo Edison has not completed the final design for t!.e DSS, design details, such as specific vendors, have not been finalized. Component location has been established (See Figures 4b and 4e).
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'Dockat Nu;btr 50-346
. Licenza Number NPF-3 Serial Number 1682 Attachment 2 Page 2 5.1 DIVERSITY FROM THE EXISTING REACTOR TRIP SYSTEM l
The diversity of the DSS Equipment from the existing Reactor Trip System (RTS) vill include all signal conditioners, bistables, logic channels, i logic power supplies, and the relays used to de-gate the silicon l controlled rectifiers.
I Sensors: The sensors that vill be used for the DSS are independent of the existing sensors that input to the Reactor Protection System (RPS) equipment presently used at Davis-Besse Unit 1 to provide the reactor trip. The sensors that vill be used to provide the input to the DSS are Reactor Coolant Pressure Transmitters PT6365B and PT6365A (Rosemount Model 1154 transmitters designed and installed to be qualified to meet the post accident conditions for Davis-Besse Unit 1). The range of each transmitter is O to 3000 psig.
In addition to being independent of the Reactor Trip System (RTS),
the transmitters being utilized for the DSS are a different model than those used by the RTS. Rosemount Model 1152 transmitters are used in the RTS to measure reactor coolant pressure. The range of each transmitter is 0 to 2500 psig. The ATVS rule (10CFR50.62) specifically excludes diverse design or diverse manufacturer requirements for the sensors.
Signal Conditioners: The DSS utilizes a 4 to 20 milliamp signal (proportional to O to 3000 psig) from the respective vide range pressure transmitters through isolation provided by Foxboro VAI modules directly to the bistable. The bistable modules utilize an internal setpoint only. No signal conditioners are required.
The RPS utilizes a buffer amplifier to provide signal conditioning and isolation. The buffer amplifier and bistable modules are supplied from Bailey Controls. The output of the buffer amplifier is a 0 to 10 VDC signal to the bistable input. The bistable modules are capable of either an internal or external setpoint.
Bistables: The bistables used for the DSS vill be of a different manufacturer Tentatively Science Applications International Corporation (SAIC)] than the RTS bistables, and vill utilize analog operational amplifiers to determine the comparison of input signals with the established setpoint. They operate on a 125 VAC power input, and have their own internal DC power supply for the logic.
The bistables used by the RTS are manufactured by Bailey Controls Company, and utilize analog operational amplifiers to determine the comparison of input signals with the established setpoint. The Bailey bistables require a 15 VDC power source, which is supplied by the RPS cabinet power supply. The cabinet power supply provides DC power to all modules in each RPS channel.
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. pockst Numb 2r 50-346
. Lipcnst Numbsr NPF-3 r Serial Number'1682, Attachment 2 i Logic Channels: The DSS' utilizes two out of.tvo logic with an energize to trip output. .When the bistable setpoint is reached a contact closure vill energize a circuit at the CRDCS, interrupting the power to the programmer lamp.-
The RPS utilizes a two out of four logic vith a deenergize to trip output. When the bistable setpoint is reached the signal to the corresponding trip breaker is interrupted tripping the breaker..
Logic Power Supplies: .The DSS bistable modules utilize an internal logic power supply, with an input of 120 VAC and output voltages of.
+/- 10 to +/- 32 VDC. This power supply is packaged as an integral part of the bistable. module.
The RPS utilizes a separate system 15 VDC power supply externally located from the bistable module. This power supply provides the power source to all RPS modules within the associated RPS channel.
Relays: The DSS relays used to de-gate the SCRs vill provide a mechanism for removing power from the Control Rod Drive Mechanisms that is diverse from the four reactor trip breakers used by the RTS to initiate the reactor trip. The RTS also de-gates the SCRs using
'125 VAC electromechanical relays. Two methods of implementing diverse relays are currently under evaluation. One is the use of solid state relays in the circuit that interrupts the programmer lamps, and the second is the use of electromechanical 24 VDC relays with mechanical contacts. Both of these alternatives will utilize equipment from a different manufacturer than those used by the RTS to electronically trip the SCRs. The best alternative vill be selected (both are diverse from the RTS relays). ,
5.2 ELECTRICAL INDEPENDENCE FROM THE EXISTING RTS The DSS will be electrically independent from the sensor output up to and including the relays that de-gate the SCRs in the CRDCS. The DSS will be installed as a non-safety related system and, as such, vill be separate from the existing safety related RTS circuits and components.
l Isolation of the sensor signal vill be provided through a Foxboro 2AO-VAI CUSTOM (ECEP 9206) STYLE A CS-N/SRC VOLTAGE-TO-CURRENT CONVERTER. The Foxboro Company Corporate Quality Assurance Laboratory Type Test Report (00AAB44 Rev A), which is proprietary to Foxboro, is available at Davis-Besse for NRC review. The report provides the validation of this isolation capability of the voltage to current converter, and includes l the additional information requested by Appendix A of the the NRC's l Safety Evaluations Report (SER). the application used at Davis-Besse for l the DSS is bounded by the Foxboro report. (See Part 1, Section 5.6 for a description of the DSS power supply.)
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The Davis-Besse RTS utilizes four safety grade reactor trip breakers to l
provide the trip function, as shown in Figure 3. In addition, two of the reactor trip breakers contain " electronic trip relays" which degate the
Dock 3t Numb;r-50-346 Lic:nca Numb:r NPF-3 Serial Number 1682 Attachment 2 Page 4 CRDCS "A" and "B" SCRs. The 1E to non-1E isolation is provided by the coil to contact isolation of the electronic trip relay as shown in Figure 4.
The DSS trip contacts are in the non-1E portion of the electronic trip circuit in the CRDCS. Therefore, requirements for Class 1E isolation of this portion of the DSS do not apply.
5.3 PHYSICAL SEPARATION FROM EXISTING RTS The DSS at Davis-Besse vill meet the BV0G requirements for physical separation from the RTS.
The DSS sensing channels and input signal isolation, vill be located in the control room cabinet room. The signal isolation vill be located in cabinets C5763A and C5755G for DSS channels 1 and 2 respectively (Refer to Figure 4e).
The bistables vill be located in cabinet C5760A, which is Class IE. The bistables will be mounted in a standard 19" instrument rack. Physical separation vill be provided for the DSS hardvare via distance and barriers where required.
The actuation channel switching devices vill be located in the CRDCS Power Supply Interface Cabinet located in Electrical Penetration Room Number 1.
- 5. 4 ENVIRONMENTAL QUALIFICATION The equipment vill be purchased and installed to meet the requirements for the environmental conditions expected in the locations selected for installation of DSS components.
5.5 QUALITY ASSURANCE FOR TEST, MAINTENANCE, AND SURVEILLANCE The DSS will be controlled in accordance with the general requirements of the Toledo Edison Quality Assurance Program in a manner similar to that currently used for other non-safety related systems. Testing, maintenance, and any specified surveillance vill be conducted and controlled in accordance with approved procedures. Collectively, the controls applied to the DSS vill meet or exceed the " Quality Assurance Guidance for ATUS Equipment That Is Not Safety Related," as set forth in Generic Letter 85-06.
The maintenance procedures in place at Davis-Besse require maintenance personnel to replace any components that have failed with identical components. If those components are not available, then a change to different components may be made provided that a plant modification request is processed. Plant modifications for the DSS will be handled in l a manner similar to a "0" system which requires a safety review be l prepared in accordance with 10CFR50.59.
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The above measures / programs, currently in place at Davis-Besse, assure that the equipment diversity provided in accordance with the ATUS Rule vill be maintained during component repair, replacement, and modifications and/or design changes, throughout the life of the plant.
5.6 SAFETY RELATED (IE) POWER SUPPLIES The power source for the DSS is not associated with the power sources .
(i.e., batteries) that are used by the RTS. The power source for the DSS l bistables (logic) vill be via a separate transformer directly from l off-site power that is not backed up by an emergency diesel generator. !
(See Figure 2 ' Power Arrangement DSS Simplified Diagram")
The DSS fun m or vill be fulfilled upon loss of off-site power as follows:
Coincident with the loss of the station main turbine generator, busses E2 and F2 vill be de-energized, resulting in interruption of power to each of the control rod drive mechanism groups. This feed is via bus E2 and F2 through the reactor trip breakers to the control rod drive control system. Therefore, a loss of power from both E2 and F2 vould result in release of all control rods (See' Figure 3, " Detailed Control Rod Group Power Arrangement").
With this arrangement, should either bus E2 or F2 be
~de-energized, either CRDCS "A" or "B" channel SCRs for each of the CRDMs would lose power. Each of the regulating and safety rod groups have two power sources. Each power source supplies one of the two CRDCS channels needed to hold the CRDM engaged.
T.% two SCRs normally hold the roller nuts engaged. One energited SCR is sufficient to maintain engagement preventing a trip of the rod. If either E2 or F2 was energized providing power to one SCR, there would be power available for the associated DSS bistable to perform its intended function.
In summary, with total loss of offsite power to the DSS the reactor is tripped without relying on the RTS. With loss of offsite power to one channel of the DSS, the remaining channel of the DSS can trip the reactor without relying on the ETS.
5.7 TESTABILITY AT POWER The DSS will be testable with the reactor at power from the sensor output up to and includj t the interruption of power to the programmer lamp in the CRDCS. Testin.,;; vill de-gate either the "A" or "B" channel of power ;
to the CRDMs. !
The DSS vill be tested by a periodic functional test, with a test frequency of once every six months. Testing vill be performed by placing !
the opposite channel of DSS into the bypass condition, which will prevent l an inadvertent trip, then introducing a test signal into the DSS channel
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-Se' rial Number 1682 Attachment 2 Page 6-
'being tested (Figure 1). Adjusting the test trip input to the trip setpoint vill-turn off the associated programmer lamps and generate a programmer fault indication on the CRDCS panel in the control room (Refer to Figures 4a and c).
i I The sensors that vill be used for input to the DSS will also'be used for indication in the Control Room. The Control Room Operators will be able j to raadily determine' sensor functionality.
In addition to the channel functional tests, the total system vill be checked for calibration, and adjusted if necessary at a frequency of once per refueling outage cycle.
If at any time a DSS channel is deemed to be inoperable, a maintenance work order vill be prepared to initiate corrective actions to make the system operable. During power operations, the DSS channels vill normally be maintained in an sperable condition. The DSS will not be required to be operable for any shutdown condition.
5.8 INADVERTENT ACTUATION Inadvertent actuation of the DSS vill be prevented by using a two-out-of-two channel logic to initiate a reactor trip. This meets the B&W generic design criteria as noted in the SER.
5.9 MAINTENANCE BYPASSES The DSS design vill permit bypassing to allow maintenance, repair, test, or calibration during power operation in order to preclude inadvertent actuation of protective actions at the system level. This will be accomplished by the use of the bypass switch, as shown in Figures 1 and 4d. 'When the switch is placed in the bypass position, an indication in the Control Room vill annunciate that the DSS is in a not-normal condition (See alarm logic, Figure la). The bypass switch will be located in a locked cabinet. Access to the bypass switch vill be administrative 1y controlled by the Shift Supervisor.
5.10 OPERATING BYPASSES The DSS will trip the reactor on high RCS pressure. The design of the DSS does not include any operating bypassea, since there are no normal operating conditions where the RCS pressure is expected to exceed the trip setpoint.
5.11 INDICATION OF BYPASSES The bypass switch for the DSS will have an annunciator output (See Figure la) to the control room annunciator panel to continuously indicate that the DSS has been disabled. When the bypass switch is returned to the normal position, the annunciator condition vill be cleared.
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- Li,ccnae Nurb:r.NPF-3 Serial Number 1682 Attachment 2 Page 7 j I
5.12 MEANS FOR BYPASSING See Section 5.9.
I' 5.13 COMPLETION OF PROTECTIVE ACTION When the RCS pressure has exceeded the trip setpoint, both DSS bistables trip and initiate an interrupt to t-he programmer lamps in both the "A" and "B" channels in the CRDCS. The bistable output seals in and can only be reset'by depressing the reset button provided on the DSS panel ,
installed next to each bistable (Refer to Figure 4d).
The bistable trip also results in the illumination of programmer lamp fault lights on the Diamond Operators Fanel (Refer to Figure 4c), and initiates a fault condition in the CRDCS. The CRDCS fault condition can be cleared only when the programmer lamps are restored.
The operator clears the fault condition by depressing the " Fault Reset" button on the Diamond Operators Panel. The sensing logic for the programmer lamp circuit is shown in Figure 4a.
As a result of the reactor trip, the operator vill perform other actions delineated in the reactor trip procedure.
5.14 INFORMATION READOUT The Davis-Besse plant modification process utilizes a project team for conceptual and detailed design review. Human Factors is one of the disciplines included in this review.
The indications for RCS pressure are located on the Post Accident Monitoring Panel in the main control room (Refer to Figure 4e).
For the following, refer to Figure la for the logic, and Figure 4b for the location. The indication that a DSS channel is in a "NOT NORMAL" condition, or " DSS TRIP" condition is provided by an annunciator vindow on the main control board, and a Computer alarm on the CRT located on the operator's desk. The " DSS TRIP" also provides a computer input in the sequence of events monitor in the plant computer. The annunciator provides both an audible alert, and a visual indication of the condition.
The programmer lamp fault lights are on the Diamond Operator's Panel, and illuminate on a DSS channel trip.
The operators are trained on the use of the annunciators. The arrangement of the annunciators was included in an overall annunciator study as part of the Detailed Control Room Design Review Group, and was integrated into the general annunciator upgrade planned for implementation at the same time as the DSS implementation.
Administrative controls for conduct of testing and key controls for access vill provide assurance that the operators are made aware of the status of the system, and that testing is in progress.
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. Serial Number 1682 Attachment 2-Page 8
-5.15 SAFETY RELATED INTERFACES As shown in Figure 3, the RPS input is via the'undervoltage device (UVD) relays in the CRD trip breakers. The function of these breakers is to interrupt power to the CRDCS transformers. This removes power from the CRDCS SCRs, and the.CRDMs. As shown in Figure 4, the DSS contacts, which de-gate the SCRs, are twice removed from the reactor trip breaker UVD.
Thus, there vill be no direct interface between DSS and RPS. In addition, there vill be no interfaces between DSS and the Safety Features Actuation System (SFAS). Therefore, the existing RPS and SFAS continue to meet all applicable safety criteria.
5.16 TECHNICAL SPECIFICATIONS The technical specifications requirements for surveillance and testing of the DSS will be addressed by the Technical Specification Improvement Program (TSIP). The NRC previously acknowledged that this was a reasonable position during its August 17, 1988, meeting with the B&W Owners Group. Toledo Edison vill test, maintain, and perform surveillance as described in our response to the SER Section 5.5.
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Serial Number 1682 Attachment 2 Page 9 s
Part 21 SFRCS-COMPARISON TO AMSAC DESIGN REQUIREMENTS The AMSAC design requirements in 10CFR 50.62 " Requirements for Reduction of Risk From ATVS Events for Light-Vater-Cooled Nuclear Power Plants" are specified as follows:
"Each pressurized water reactor must have equipment from sensor output to final actuation device that is diverse from the reactor trip system to automatically initiate the auxiliary feedvater system and initiate a turbine trip under conditions indicative of an ATVS. This equipment must be designed to perform its function in a reliable manner and be independent (from sensor output to the final actuation device) from the existing reactor trip system."
The' existing Steam and Feedvater Rupture Control System (SFRCS) installed at Davis-Besse satisfies this requirement. SFRCS is a Class lE system which actuates auxiliary feedvater (AFV) and initiates a turbine trip for the following conditions:
Low steam generator level High' steam generator level Low steam generator pressure High steam generator to main feedvater differential pressure Loss of four reactor coolant pumps As requested by Reference 1, a discussion of the method of detecting total loss of feedvater flow is provided below.
Following a loss of main feedvater event, a low steam generator level (10 inches collapsed liquid level above the lover tube sheet) vill initiate SFRCS which vill actuate the AFV system and trip the main turbine. A Davis-Besse specific ATVS analysis was recently performed by B&W (Reference 3) to determine the peak RCS pressure for an SFRCS initiation using the existing lov steam generator level signal. This new analysis was performed to account for the numerous modifications to the SFRCS and AFV systems which have been implemented since the Davis-Besse June 9, 1985 event which affect the previous BV0G ATVS analysis. The analysis was performed to demonstrate the acceptability of using the existing SFRCS to fulfill the AMSAC function. The analysis shows that SFRCS initiation on low steam generator level establishes sufficient Auxiliar; Feedvater flow in appropriate time to limit the peak RCS pressure (4103 psia).
The results of the analysis demonstrate that the peak pressure falls within the range (3621 psia to 4190 psia) previously identified for the BEV plants in the earlier BV0G ATVS analysis. This range of pressures has been previously evaluated by the NRC (Reference 4) and found acceptable. Additionally, B&V evaluated (Reference 5) and found acceptable for an ATVS ever.t, peak pressures up to 4300 psia at Davis-Besse.
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Serial Number 1682 Attachment 2 Page 10 1 Therefore, the existing SFRCS design satisfies the requirements of the ATUS i rule for AMSAC. Since SFRCS is partially povered from the same inverters as I the RTS, an analysis is required to show that faults vill not propagate l through the power supplies and disable both SFRCS and the RTS. This analysis I has been completed (See Part 3). The analysis concludes that no credible j failures can propagate through the common power supplies and disable both l i SFRCS and the PTS. a l I Part 3 of this attachment addresses the design requirements associated with )
diversity from existing RPS and safety related (IE) power supplies. The j remaining design requirements specified in Section 5 of the SER are those j required for a Class 1E System. As noted above, SFRCS is a Class IE system. 1 The following addresses the design requirements, as identified in the Generic SER, for the Davis-Besse AMSAC/SFRCS. For convenience, the paragraphs are numbered to coincide with comparable paragraphs of the NRC's Safety Evaluation Report. )
DESIGN REQUIREMENTS FOR AMSAC/SFRCS 5.1 DIVERSITY FROM THE EXISTING REACTOR TRIP SYSTEM (RTS)
Diversity from the existing RTS is assured by:
- Manufacturing Processes: The Bailey 880 RTS equipment and the SFRCS equipment are manufactured by two different companies (Bailey 4 Metering Company and Consolidated Controls Corporation respectively), >
at two different manufacturing facilities utilizing independent manufacturing procedures.
- Principle of Operations: The SFRCS is primarily digital in operation while the RTS is primarily an analog system.
- System Interfaces: The SFRCS uses primarily optical isolation technology for its interfaces while the RTS systems uses relay contacts and operational amplifiers.
5.2 ELECTRICAL INDEPENDENCE FROM THE EXISTING RTS The SFRCS is a Class 1E system. There is no 1E to non-1E isolation required between the RTS and SFRCS. The power supply inputs enter the SFRCS cabinets via isolation transformers, but these are not for 1E to non-1E isolation.
5.3 PHYSICAL SEPARATION FROM THE EEISTING RTS The Class 1E separation for individual channels is provided for the RPS, and the SFRCS. Figure 4e shows the location of the cabinets in the cabinet room. Only one SFRCS cabinet is in the same rov as an RPS cabinet, and it is the same Class IE channel.
pocket Numb:r 50-346 Lic:nsa Numbar NPF-3 Serial Number 1682 Attachment 2 Page 11 5.4 ENVIRONMENTAL QUALIFICATION The SFRCS is a Class 1E system, and as such the system, its inputs, and its ouputs actuated equipment are purchased, installed, and maintained to meet the appropriate environmental normal operating and accident conditions.
5.5 QUALITY ASSURANCE FOR TEST, MAINTENANCE, AND SURVEILLANCE The SFRCS is classified as a "0" system, and as such was purchased and installed, and is maintained and tested, in accordance with the requirements of the Toledo Edison Quality Assurance Program. The requirements for the SFRCS that relate to 10CFR50.62 vill be documented in the SFRCS system description.
The maintenance procedures in place at Davis-Besse require the maintenance personnel to replace any components that have failed with identical components. If those components are not available, then a change to different components may be made provided that a plant modification request is processed. The plant modification for such a system requires a safety evaluation be prepared in accordance with 10CFR50.59.
The above are the measures / programs presently implemented at Davis-Besse that assure the equipment diversity provided in accordance with the ATVS Rule vill be maintained during component repair, replacement, and modifications and/or design changes, throughout the life of the plant.
5.6 SAFETY RELATED (IE) POWER SUPPLIES The SFRCS is a Class 1E system. There is no 1E to non-1E isolation required between the same channels of RPS and SFRCS. The power supply inputs enter the SFRCS cabinet via an isolation transformer but this is not for 1E to non-lE isolation. (See Figure 5) 5.7 TESTABILITY AT POVER The SFRCS system surveillance testing and actions requirements are addressed by Technical Specification sections 3.3.2.2/4.3.2.2. The testing requirement of monthly at-power (functional) testing and complete system (calibration) testing every eighteen months satisfies the AMSAC testing requirements of at-power testing every six months and complete system testing every refueling.
5.8 INADVERTENT ACTUATION The SFRCS is a Class 1E system, and as such two channel trips are required to initiate actuation of the desired functions.
___ _______. ____ _ _ _a
, 'DoSkst Nu2btr 50-346
. Licensa Numb 2r NPF-3
+
~ Serial Number 1682 L Attachment 2 Page 12 5.9 MAINTENANCE BYPASSES No maintenance bypasses exist in the SFRCS. There are parameter input bypasses which exist and are used only while in Modes 4, 5 or.6.' These input bypasses are to facilitate testing and are not to be used during normal operation. The use of the input bypass is annunciated to the operator; 5.10 OPERATING BYPASSES'
.The operating bypass for the SFRCS is based upon steam line pressure. As the plant cools down, the Low Pressure and High Steam Generator Level trip is bypassed and reset as follows:
- a. The bypass pressure setpoint is reached with two out of two low pressure block pressure switches providing an enable and annunciator alarm.
- b. The Operator depresses the SFRCS low pressure block push buttons.
- c. The block seals in and is annunciated that the block permissive is gone and the block is established,
- d. .The block is automatically removed when the steam generator is repressurized and steam pressure is high enough that an inadvertent lov pressure actuation vill not occur.
5.11 INDICATION OF BYPASSES See Section 5.10 5.12 MEANS FOR BYPASSING See Section 5.10 5.13 COMPLETION OF PROTECTIVE ACTION The SFRCS is an automatic reset system. The system does satisfy IEEE 279 Section 4.16 requirements for " Completion of Protective Action Once It Is Initiated". The system satisfies this requirement by having component level seal ins which ensure that the components' protective action goes to completion. To ensure the component level seal ins are engaged, a two second SFRCS logic level seal in provides adequate titte for the component logic to seal in. . Component level seal ins must be reset by manual operator action.
5.14 INFORMATION READOUT The operator is provided information for the SFRCS system by twelve annunciator vindows and 2 steam generator level indicators, one from each steam generator, in the center console. If any input to the SFRCS
'Dockst Numb 2r 50 346
. - Licinsh Nuzbar NPF-3
- " Se' rial Number 1682-
. Attachment 2 Page 13 l
changes to the tripped condition, a "SFRCS Trouble" annunciator vindow illuminates. This vindow also annunciated a loss of power supply. The remaining vindows are as follows:
- 1. Two vindows for a low pressure trip annunciation
- 2. Two vindows for a high steam generator level trip and a high stram generator to feedvater pressure differential trip-
- 3. Two vindows for a low level trip and a loss of all four reactor 1 coolant pumps trip
- 4. Two vindows for operating bypass permissive indication
- 5. Two vindows for operating bypass block established. .
- 6. - One vindow for SFRCS Actuation channel initiation.
Human factors for SFRCS vere addressed during the Davis-Besse Detailed Control Room Design Review (DCRDR). Due to numerous problems identified for SFRCS, a special study team was assigned to address all aspects of.
operator interface with the SFRCS. As a result, during the fifth refueling outage an extensive regrouping of controls and displays were implemented along with changes to annunciators. Some new displays (e.g.,
Steam Generator level and pressure) were added to a mimic layout of the Steam Generator, Feedvater and Auxiliary Feedwater Control on the center console.
5.15 SAFETY RELATED INTERFACES Refer to Section 5.6 5.16 TECHNICAL SPECIFICATIONS The SFRCS is currently addressed by Technical Specifications.
. 'Dbeket Numbar 50-346
.Licensa Humbar NPF-3 Serial Number ~1682
Attachment 2 Page 14 Part 3'- SFRCS POWER SUPPLY CONFIGURATION AND OPERATIONAL REQUIREMENTS SFRCS Power Supply Configuration Each SFRCS Logic Channel is powered from a separate 120 VAC 1E vital bus
-(Figure 5). Two Logic Channels (one in each Actuation Channel) are povered from battery backed inverters which also provide power to two RPS channels.
The remaining Logic Channels are powered from Emergency Diesel Generator backed sources which are independent of the RPS power supplies.
SFRCS System Power Specification'ns The SFRCS system voltage and frequency specifications for vital (logic) AC power are:
Voltage 120 VAC i 10%
Frequency 60 Hz 1 2%
The incoming 120 VAC is converted to 28 or 48 VDC to power the SFRCS logic components by internal power supplies. The internal DC Logic power supplies have a transformed input power specification of:
28 VDC: 22.55 VAC 1 2.82 VAC (12.5%) @ 60 Hz 1 3 Hz (5%)
l 48 VDC: 39.78 VAC 3 4.8 VAC (12%) @ 60 Hz 1 3 Hz (5%)
Loss Of Power Effects On SFRCS The SFRCS is. designed as a Class 1E system. Failure of one of-the 1E vital busses to zero volts will result in the system reverting to a one-out-of-one l trip configuration for the actuation channel affected. Failure of both IE l vital buses on a~ single actuation channel vill result in a turbine trip and initiation of respcetive Auxiliary Feedvater train due to the de-energize-to-trip design of the SFRCS.
On a Loss of Offsite Power, power to all four Reactor Coolant Pumps is lost.
This results in actuation of SFRCS which trips the turbine and initiates Auxiliary.Feedvater.
SFRCS Operation With An Overvoltage Or Undervoltage Condition The SFRCS power supplies are non-regulated and simply rectify and filter the incoming AC. Because the power supply DC output voltage follows the incoming AC voltages, the power supplies vill not limit an overvoltage or undervoltage condition. The limiting undervoltage components are the opto-isolators used in the field buffers and the relay drivers. The voltage requirement for the i
opto-isolators is approximately 12 VDC. Based upon a 28 VDC normal output H from the power supplies, the devices vill still provide their intended function with a 57% decrease in voltage. The limiting overvoltage components are the capacitors in the 48 VDC and 28 VDC power supplies. The capaciters have a vendor approved operating voltage which is approximately 12% above the actual power supply normal operating voltage. The capacitors have a maximum surge rating of approximately 75% above power supply normal operating voltage.
LDocket Numbar 50-346 Li,ctnsa Nunb2r NPF-3
' Serial Number 1682 Attachment Page'.15
'A failure of the capacitors vill depend'on the amount of time and-percentage aboveLthese vendor. approved voltages. Even with a capacitor failure, since the SFRCS is a de-energize to trip system, the power supply failure vill only L revert the SFRCS to a one-out-of-one trip for the affected actuation channel.
SFRCS Operation'With-An Overfrequency Or Underfrequency Condition There is no effect on the SFRCS from an overfrequency or.underfrequency conditions. 'The only result vill be an increase or decrease in input impedance to;the-transformers. This has no effect due to the sizing of the transformers and the small normal'1oad required by the SFRCS components.
Vital Bus Power Supplies - System Operation Inverter.0utput Specifications Output Voltage 118 VAC i 1%
Output Frequency 60 Hz + 1%
Each inverter is supplied with 125 VDC from a station battery and 3-phase rectifier povered from a 480 VAC diesel backed bus. The 480 VAC'is rectified and then diode auctioneered with the 125 VDC from the station battery. This DC output from the diode auctioneering circuit is the input to the inverters.
During normal operation the inverter frequency is synchronized to an AC source. On loss of the AC reference voltage, the inverter frequency is controlled by an internal oscillator which has a specified frequency of 60 Hz i 1/2%. Thus, loss of either the station battery or the 480 volt bus will not cause the loss of'the inverter output.
Conclusions
- Loss of Offsite Power or loss of the AC power supplies to the SFRCS has no effect since the SFRCS is a fail-safe system (de-energize to trip). The loss of power results in a turbine trip and initiation of auxiliary feedvater.
Undervoltage is not considered a credible failure since the voltage che.nge required for an undervoltage is >50% and as noted above, system functionality will still be maintained for a 57% decrease in voltage.
- Overvoltage has no detrimental effect upon the system until a power supply capacitor failure occurs. At that time, the power supply fails and the SFRCS is reverted to a one-out-of-one trip requirement for the affected actuation channel.
- Overfrequency and underfrequency have no effect on the operation of the system due to the sizing of the transformers and small normal load required by the SFRCS components.
The failure of one channel in either SFRCS or the RTS cannot prevent either' system from performing its design function.
h '
D:cht Nu b:r 50-346 Li.c naa Nurb:r NPF
Serial Number 1682L A'ttachment 21
-Page:16' It is also TE's' position.that SFRCS is not part'of'the Reactor Trip System,
.that the SFRCS equipment is diverse from the RTS equipment, and that there is:
no common failure mechanism which can prevent both systems,from performing
!their intended' function.
-s.
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' Dock t Numbar 50-346 Li,c nsa Numbar NPF-3 Serial Number 1682 Attachment 2 Page 17 References (J J, Let ter to Mr, Donald C, Shelton (Toledo Edison) from Albert V, D*6;r. win. . _ . . _!
(NRC) dated August 10, 1988; NRC Evaluation of BV0G Generic Report -
" Design Requirements for DSS and AMSAC".
[2]. Letter to Mr. L. C. Stalter (Chairman of the BV0G ATUS Committee) from Gary Holahan (NRC Staff), dated September 7, 1988; August 17, 1988, B&V/NRC ATVS Meeting
[3]. B&W Calculation 32-117357-00 "DB-1 LOFV ATVS analysis", dated February 28, 1989.
[4]. Reference 4 of the NRC SER dated February 1988, Safety Evaluation of Topical Report (B&W Document 47-1159091-00) " Design Requirements for DSS (Diverse Scram System) and AMSAC (ATVS Hitigation System Actuation Circuitry)"
[5]. B&V Report 12-1174341-00 "DB-1 ATVS Justification", dated February 9, 1989.
i I
LIST OF FIGURES t-i irev1 DSS FUNCTIONAL DIAGRAM la DSS ALARM LOGIC 2 POWER ARRANGEMENT DSS SIMPUFIED DIAGRAM 3 DETAILED CONTROL ROD GROUP POWER ARRANGEMENT 4 rev 1 DSS INTERFACE WITH RPS ELECTRONIC TRIP 4a CONTROL ROD DRIVE CONTROL SYSTEM ALARM LOGIC 4b DSS CONTROL ROOM INTERFACE 4c DIAMOND OPERATOR'S PANEL 4d DSS RACK LAYOUT 4e EQUIPMENT LOCATION 5 POWER ARRANGEMENT SFRCS SIMPLIFIED DIAGRAM 4
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C s -
' }s
' An V o 2
0Yi 1
Il il i l , L