ML20235W537

Forwards Design Summary Re plant-specific Info for ATWS Implemetation (10CFR50.62).Encl Provides Evaluation That Demonstrates That for All Loss of Offsite Power Scenarios, Control Rods Will Be Released Due to Loss of Voltage

ML20235W537
Person / Time
Site:	Davis Besse
Issue date:	02/28/1989
From:	Shelton D TOLEDO EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
1638, TAC-59086, NUDOCS 8903130081
Download: ML20235W537 (19)

Text

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1 TOLEDO EDISON A Centenor Energy Company DONAU] C. SHELTON Voe PrerchbNuch (4

• 24S "

Docket Number 50-346 License Number NPF-3 Serial Number 1638 February 28, 1989 United Stctes Nuclear Regulatory Commission Document Control Desk Vashington, D.C. 20555 Subj ect : Plant-Specific Submittal for ATUS Implementation (10 CFR 50.62)

(TAC 59086)

Gentlemen:

The attached " Design Summary" provides the plant-specific information requested by Nuclear Regulatory Commission (NRC) letter dated August 10, 1988 (Log Number 2664: "NRC Evaluation of BV0G Generic Report - Design Requirements for DSS and AMSAC") for the Davis-Besse Nuclear Power Station.

The equipment to protect against Anticipated Transients Without Scram at Davis-Besse vill be designed based on the generic design described in Babcock and Vilcox (B&W) Document 47-1159091-00, " Design Requirements for Diverse Scram System (DSS) and ATVS Mitigation System Actuation Circuitry (AMSAC)" and the additional guidance provided by the NRC Safety Evaluation Report (SER) transmitted with the August 10, 1988 letter.

Diverse power supply requirements described in SER Sections 5.6 and 6.1 vere further clarified in an August 17, 1988 NRC/ Babcock and Wilcox Owner's Group (BV0G) meeting. In this meeting and in a subsequent NRC letter [G. Holahan (NRC) to L. C. Stalter (BV0G) dated September 7, 1988] the NRC described three acceptable options that licensees could use to resolve the power supply independence issue. Toledo Edison has selected Option 2, which requires that a power source for the DSS be provided via a non-battery backed 480-volt bus that is independent [i.e., not associated with the Reactor Trip System (RTS)]

and non-class 1E. The attachment to this letter provides an evaluation that demonstrates that for all loss of offsite power scenarios the control rods i

vill be released due to loss of voltage to the 480-volt supply to the control rod holding mechanisms. Part i of the attached " Design Summary" addresses DSS design requirements.

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5 8903130081 890228 PDR ADOCK 05000346 P PDC THE TOLEDO EDISON COMPANY EDISON PLAZA 300 MADISON AVENUE TOLEDO. OHIO 43652

Docket Numb:r 50-346 License Number NPF-3

. Sefial Number 1638 Page 2 To meet the requirements of the AMSAC, Toledo Edison vill utilize the existing Steam and Feedvater Rupture Control System (SFRCS). Part 2 of the attachment to this letter provides the basis for using the SFRCS to fulfill the AMSAC )

function since SFRCS performs the same. function. Part 3 of the attachment j includes an evaluation to demonstrate that common mode failures vill not  !

propagate through the power supplies and disable both SFRCS and the RTS since SFRCS is partially powered through the same essential 120 VAC busses as the RTS. I Toledo Edison plans to design and install the ATUS systems by the end of the 1 Sixth Refueling Outage, which is currently scheduled to begin in February 1990. Toledo Edison is prepared to further discuss the attached design details with the NRC at a mutually convenient time. If you have any questions concerning this matter, please contact Mr. R. W. Schrauder, Nuclear Licensing Manager, at (419) 249-2366.

Very trul ours, 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 l

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Docket Numb 2r 50-346 Licensa Numbar NPP-3 Serial Number 1638

. Attachment Page 1 DESIGN

SUMMARY

Toledo Edison's conceptual design for equipment required to address the Anticipated Transient Without 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 ATVS Mitigation System Actuation Circuitry (AMSAC) function. Part 3 provides an evaluation to demonstrate that common mode failures vill not propagate through the power supplies and disable both SFRCS and the .8.

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 contacts will 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. Beth 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 alarm.

In addition, all " rod bottom" lights will be lit on the Position Indication Panel in the Control Room.

The system vill be designed to be testable with the reactor on-line. While one channel is being tested, a DSS bypass will be provided 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.

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 convenienet, 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 Wilcox Owners Group (BV0G) " Design Requirements for DSS and AMSAC" document. Since Toledo Edison has not completed the final design for the DSS, design details, such as component location and specific vendors, have not yet been finalized.

Dockat Numb:r 50-346  ;

Licznsa Numb r NPF-3 )

Serial Number 1638 1

, Attachment l Page 2 1 5.1 DIVERSITY FROM THE EXISTING REACTOR TRIP SYSTEM The diversity of the DSS Equipment from the existing Reactor Trip System (RTS) vill include all signal conditioners, bistables, logic channels, logic power supplies, and the relays used to de-gate the silicon controlled rectifiers.

Sensors: The sensors that will 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 0 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 ATVS rule l (10CFR50.62) specifically excludes diverse design or diverse manufacturer requirements for the sensors.

i Signal Conditioners: The transmitter output signal (4-20 milliamps) is provided directly to the bistable. No signal conditioners are required. ,

Bistables: The bistables used for the DSS vill be from a manufacturer other than Bailey Controls Company (manufacturer of the l bistables used in the RTS).

Logic Channels: The DSS logic channels vill not use equipment manufactured by Bailey Controls Company (manufacturer of the logic ]

j channels used in the RTS).

Logic Power Supplies: The only DSS logic is the bistable used to compare the input signal representative of the RCS pressure to the internal setpoint to provide a contact closure to initiate a channel trip. Any power supply used for this bistable vill be of a different manufacturer than Bailey Controls Company (manufacturer of the power supplies used in the RTS). ,

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. The relays used for the DSS will be different from those used by the RTS.

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, will be separate from the existing safety-related RTS circuits and components.

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Docket Numbrr 50-346 Liesnsa Number NPF-3 Serial Number 1638

' Attachment Page 3 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 the additional information requested by Appendix A of the SER. (See 5.6 for description of the power supply for the DSS.)

The Davis-Besse RTS utilizes four safety grade reactor trip breakers to provide the trip function, as shown in Figure 3. In addition, two of the reactor trip breakers contain " electronic trip relays" which trip the 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.

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 0UALITY ASSURANCE FOR TEST, MAINTENANCE, AND SURVEILLANCE The DSS vill 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 syatems. 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 ATVS Equipment That Is Not Safety Related," as set forth in Generic Letter 85-06.

5.6 SAFETY RELATED (1E) 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 bistables (logic) vill be via a separate transformer directly from off-site power that is not backed up by an emergency diesel generator. (See Figure 2, " Power Arrangement DSS Simplified Diagram")

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'Dockat Numbar 50-346 Licznsa Numbar NPF-3 Serial Number 1638

, Attachment Page 4 The DSS function 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 CRDHs vould 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. The two SCRs normally hold the roller nuts engaged. One energized 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 power to the DSS the reactor is tripped without relying on the RTS. With loss of power to one channel of the DSS, the remaining channel of the DSS can trip the reactor without relying on the RTS.

5.7 TESTABILITY AT POVER The DSS will be testable with the reactor at power from the sensor output up to and including the interruption of power to the programmer lamp in the CRDCS. Testing vill de-gate either the "A" or "B" channel of power to the CRDMs.

Testing vill be performed by placing the opposite channel of DSS into the bypass condition, which will prevent an inadvertent trip, then introducing a test signal into the DSS channel 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.

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 to readily determine sensor functionality.

5.8 INADVERTENT ACTUATION Inadvertent actuation of the DSS will 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.

Dockat Numb 2r 50-346 License Number NPF-3 Serial Number 1638

. Att'achment Page 5

)

5.9 MAINTENANCE BYPASSES l 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 vill be accomplished by the use of the test switch, as shown in Figure 1, under key lock control. When the switch is placed in the test position, an indication in the Control Room will annunciate that the DSS is disabled. Access to the test 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 bypasses, since there are no normal operating conditions where the RCS pressure is expected to exceed the trip setpoint.

5.11 INDICATION OF BYPASSES The test switch for the DSS will have an annunciator output to the control room annunciator panel to continuously indicate that the DSS has been disabled. When the test switch is returned to the normal position, the annunciator condition vill be cleared.

5.12 MEANS FOR BYPASSING See Section 5.9.

5.13 COMPLETION OF PROTECTIVE ACTION The DSS trip will result in a " fault" condition in the CRDCS which seals in per the CRDCS logic design. Reset of this trip condition vill require deliberate operator action at the Diamond Control Panel, and vill not be possible until the programmer lamp current is restored. The Diamond Control Panel is mounted on the Main Control Panel in the Control Room.

5.14 INFORMATION READ 0UT The DSS design vill provide the operator complete, accurate, and timely information concerning system status. If any channel is in the tripped condition, the status will be provided to the operator via the plant computer.

In addition, the operator vill receive a CRDCS system fault annunciator alarm along with a programmer lamp fault light on the Diamond Control Panel in the main control room. The high pressure condition vill also be indicated on the Post Accident Monitoring Panel in the main control room.

The alarmed conditions will clear when the tripped condition is reset by the operator. Indications of test, and maintenance are as previously described in Section 5.9.

i Docket Numb 2r 50-346 Licensa Number NPF-3 Serial Number 1638

. Attachment i Page 6 ]

l 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.

i 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 SER Section 5.5.

Dockst Numb r 50-346

'Licensa Numbzr NPF-3 Serial Number 1638

. Attachment Page 7 l-Part 2 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-1E system which actuates auxiliary feedvater (AFW) and initiates a turbine trip for the following conditions:

Lov steam generator level High steam generator level Lov 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 flo. is provided below.

Following a loss of main feedvater event, a lov 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 tc th- 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 lov steam generator level occurs approximately 27 seconds following the loss of feedvater event. Auxiliary 1 Feedvater flow is established, in either case, approximately 30 seconds following SFRCS initiation due to auxiliary feedpump turbine acceleration time. The resultant peak RCS pressure for this case is 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 B&W 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&W evaluated (Reference 5) and found acceptable for an ATVS event, peak pressures up to 4300 psia at Davis >Besse.

Docket Numbar 50-346

' License Numbar NPF-3 Serial Number 1638

, Attachment Page 8 Therefore, the existing SFRCS design satisfies the requirements of the ATVS rule for AMSAC. Since SFRCS is partially powered from the same inverters as.

the RTS, an analysis is required to show that faults will not propagate

'through the power supplies and disable both SFRCS and the RTS. This andvsis has been completed (See Part 3). The analysis concludes that no credible failures can propagate through the common power supplies and disable both SFRCS and the RTS.

Part 3 of this attachment addresses the design requirements associated with i

diversity from existing RPS and safety-related (1E) power supplies. The l remaining design requirements specified in Section 5 of the SER are those l

required for a Class-1E System. As noted above, SFRCS is a Class-1E system.

l- Therefore, an item by item comparison for SFRCS is not provided.

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I Docket Number 50-346 License Numbsr NPF-3 Serial Number 1638

, Attachment Page 9 Part 3 - SFRCS POVER SUPPLY CONFIGURATION AND OPERATIONAL REQUIREMENTS SFRCS Power Supply Configura" on 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 powered 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 Specifications The SFRCS system voltage and frequency specifications for vitrl (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%)

48 VDC: 39.78 VAC 1 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 vill result in the system reverting to a one-out-of-one trip configuration for the actuation channel affected. Failure of both 1E vital buses on a single actuation channel vill result in a turbine trip and initiation of respective 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 Vith 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 opto-isolators is approximately 12 VDC. Based upon a 28 VDC normal output from the power supplies, the devices will 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 capacitors

Docket Number 50-346 License Number NPF-3 Serial Number 1638 Atfachment Page 10 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.

A failure of the capacitors will depend on the amount of time and percentage above these vendor approved voltages. Even with a capacitor failure, since the SFRCS is a de-energize to trip system, the power supply failure vill only revert the SFRCS to a one-out-of-one trip for the affected actuation channel.

5FRCS Operation With An Overfrequency Or Underfrequency Condition There is no effect on the SFRCS from an overfrequency or underfrequency conditions. The only result will 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 load required by the SFRCS components.

Vital Bus Power Supplies - System Operation l Inverter Output Specifications Output Voltage 118 VAC i 1%

Output Frequency 60 Hz i 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 i 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 i 1/2%. Thus, loss of either the station battery or the 480 volt bus will not cause the loss of the inverter output. j Conclusions j

- Loss of Offsite Power or loss of the AC power supplies to the SFRCS has no effeet 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 change required for an undervoltage is >50% and as noted above, system functionality vill 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.

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I Docket Number 50-346 License Number NPF-3 Serial Number 1638 Attachment Page 11

- The failure of one channel in either SFRCS or the RTS cannot prevent either system from performing its design function.

l[ i 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 based on the following:

- Manufacturing Processes: The Bailey 880 RTS equipment and the SFRCS equipment are manufactured by two different companies (Bailey Metering Company and Consolidated Controls Corporation), 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.

' Docket Number 50-346 License Number NPF-3 Serial Number 1638 Attachment Page 12 References

[1]. Letter to Mr. Donald C. Shelton (Toledo Edison) from Albert V. DeAgazio (NRC) dated August 10, 1988; NRC Evaluation of BWOG Generic Report -

" Design Requirements for DSS and AMSAC".

[2]. Letter to Mr. L. C. Stalter (Chairman of the BV0G ATVS 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&V Document 47-1159091-00) " Design Requirements for DSS (Diverse Scram System) and AMSAC (ATVS Mitigation System Actuation Circuitry)"

[5]. B&W Report 12-1174341-00 "DB-1 ATVS Justification", dated February 9, 1989.

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DETAILED CONTROL ROD GROUP l

POWER ARRANGEMENT FIGURE 3 i

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