ML19256C738

Forwards Addl Info Re Upgrading of safety-grade Anticipatory Reactor Trip Sys (Arts),In Response to 790907 Request.One Oversize Drawing Available in Central Files

ML19256C738
Person / Time
Site:	Davis Besse
Issue date:	10/03/1979
From:	Roe L TOLEDO EDISON CO.
To:	Reid R Office of Nuclear Reactor Regulation
References
540, TAC-45184, NUDOCS 7910120235
Download: ML19256C738 (19)

Text

TOLECO Docket No. 50-346 EDISON License No. NPF-3 LCwEtt E. RCE v e. ..n e.rt Serial No. 540 ac~t a : +=~i 1419) 259 5242 October 3, 1979 Director of Nuclear Reactor Regulation Attention: Mr. Robert W. Reid, Chief Operating Reactors Branch No. 4 Division of Operating Reactors United States Nuclear Regulatory Commission Washington, D. C. 20555

Dear Mr. Reid:

This letter is in response to your September 7, 1979 request for addi-tional information on the upgrading of the anticipatory reactor trip system (ARTS) at the Davis-Besse Nuclear Power Station, Unit 1 (D3-1).

Toledo Edison is designing and purchasing an ARTS which will be separate from the Babcock and Wilcox reactor protecticn system. The original de-sign was submitted on May 21, 1979 (TECo letter Serial No. 1-71). Since that time the design has been modified to incorporate recent reactor trip experience. Details are provided in the attached response to item 6 with the enclosed revision 1 to Bechtel Co. Drawing No. SK-E-410. The modifications provide improved testability and remove a low reactor power block signal. The revised design only blocks the turbine trip signal below 207. power. Loss of feedwater signals remain effective at all power levels.

In order to expedite as much as possible the installation of the safety grade trip system, Toledo Edison has already issued the design for vendor bids. If there are no changes in the proposed system as a result of the NRC design review, the schedule may be able to meet your requested im-proved installation date of six months after NRC approval. At this time such an improvement is tenuous and cannot be coc=itted to. In lieu of that assurance, attachment A provides an identification and discussion of a modification to the interim ARTS system that could be provided within six months of your concurrence. y b \ .

Very truly yours, 4

• W hp LER/TJM ,, _ ._

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Enclosure pp a/18 7 91012 c 235 pog#g THE TCLECC ECISCN COMPA'JY ECISCN PLAZA 300 MACISCN AVENLE TCLECC. CKO A3552 ' #

Docket No.50-346 License No. NPF-3 Serial No. 540 October 3, 1979 ATTACHMENT A Interim (Control Grade)

Anticipatory Resctor Trip System (I-ARTS)

In response to the NRC letter of September 7, 1979 Toledo Edison has re-viewed both the final (safety grade) and interim (control grade) ARTS designs. This attachment discusses Toledo Edison's proposed modifica-tion that would improve the I-ARIS in case the safety grade system sche-dule could not be acceptably expedited.

The Davis-Besse Nuclear Power Statier., Unit 1 (DB-1) 1-ARTS features were discussed with your staff during a meeting at the NRC offices on May 8,1979 and formalized by letter (Serial No. 501) on May 11,1979.

Davis-Besse Drawing No. SK-E-409, Revision 1 was transmitted in that letter.

Two of the reverse delta pressure switches conn?cted across the main feedwater check valves which are presently used to sense a loss of main feedwater flow for the Steam and Feedwater Rupture Control System (SFRCS) are also used as inputs to I-ARTS. These pressure switches are tested monthly. The only portion of I-ARTS not testable at power is ths main turbine trip input signal.

Toledo Edison is proposing to add redundancy to the main turbine trip input signal as an upgrade to the I-ARTS. This upgrade would provide a parallel contact to XKT1182 (B-5 on Drawing No. SK-E-409) with parallel wiring to sense a turbine trip from a loss of hydraulic control oil pressure. This provides redundancy in the only portion of the I-ARTS not testable at power.

The added redundancy in the turbine trip circuitry and the current testability of the I-ARTS will provide an anticipatory reactor trip system acceptable for an extended period until the safety grade system can be delivered and installed.

This improvement can be installed prior to start up after the spring outage scheduled for March 1980 unless the safety grade ARTS schedule can be expedited to approximately six months after NRC design approval.

pp a/19 i687 255 A-1

Docket No. 50-346 License No. NPF-3 Serial No. 540 October 3, 1979 Attachment B Response to Request for Additional Information on the Safety Grade Anticipitory Reactor Trip System (ARTS)

Item 1 For your proposed design, state the degree of conformance with the acceptance criteria listed in Column 7.2 of Table 7-1 (" ACCEPTANCE CRITERIA FOR CONTROLS")

of the Standard Review Plan. Justify any non-conformance.

Response

The conformance of the DB-1 proposed safety grade anticipatory reactor trip system (ARTS) with criteria lieted in column 7.2 of Table 7-1 of the standard review plan. In general, the ARTS meets or exceeds all the criteria for which Davis-Besse Nuclear Power Station, Unit I was reviewed for in the issuance of its operating license. The numbers below corresponds to those of Table 7-1.

Criteria 1-10CFRPart 50 a) 10CFR50.34 - the ARTS complies with this criteria b) 10CFR50. 36 - proposed technical specifications will be submitted upon system installation c) 10CFR50.55a- the ARTS complies with this criteria. Specifics of paragraph h are described in the response to item 2 below.

Criteria 2-Ihe ARTS conforms to the General Design Criteria (GDC) Appendix A to 10CFRPart 50 as described in the DB-1 Final Safety Analysis Report (FSAR),

Appendix 3D. Additional information is provided below:

1) To meet GDC 1, Quality Standards and Records, the requirements of the Toledo Edison Company Quality Assurance Program Specification for Operation Phase Suppliers / Contractors are complied with.

2) To meet GDC 3, Fire Protection, the requirements of the FSAR and the Fire Hazard Analysis Report are complied with.

3) The response to Question 2 discusses in further detail the design basis of the system which conforms to the requirements of: GDC 20, Protection System Functions, GDC 21, Protection System Reliability and Testability; GDC 22, Protection System Indepen fence; GDC 23, Protection System Failure Modes; 3DC 24, Separation of Protection snd Centrol Systems.

i687 256 B-1

Docket No.50-346 License No. NPF-3 Serial No. 540 October 3,1979 Response to Item 1 (continued)

4) GDC 35, Emergency Core Cooling, and GDC 37. Testing of Emergency Core Cooling System, are not applicable to the Safety Grade Anticipatory Reactor Trip System.

Criteria 3-IEEE Standards - The Safety Grade Anticipatory Reactor Trip System conforms to the IEEE Standards of interest as follows:

1) The degree of conformance with IEEE 279-1971 is discussed in response to Question 2.

2) IEEE 317 is not applicable.

3) The System complies with IEEE 336-1971 as discussed in response to FSAR Question 7.1.1.

4) Provisions are made to permit testing in accordance with Section 5 of IEEE 338-1971.

5) The equipment complies with the requirements of IEEE 344-1975 and as further explained under item 4 when addressing Regulatory Guide 1.100.

6) For discussion of IEEE 379 compliance, refer to the response to Item 2 of this attachment where confor=ance to IEEE 279-1971 is discussed.

7) The IEEE 384 separation requirements are met by the separation of criteria discussed in response to FSAR Question 8.1.2 which addresses Regulatory Guide 1.75.

Criteria 4-Regulatory Guides Regulatory Guides - The Safety Grade Anticipatory Reactor Trip System conforms to the regulatory guides of interest as follows:

1) Regulatory Guides 1.11, 1.70 and 1.120 are not applicable.

2) Regulatory Guides 1.22,1.29 and 1.30 are complied with as stated in FS AR Appendix 3D.

3) The intent of Regulatory Guide 1.47 is met as discussed in response to FSAR Question 7.5.2.

4) Regulatory Guide 1.53 was not specifically committed to in the licensed design; however, the application of the single failure criterion is discussed in response to Item 2 of this attachment where conformance to IEEE 279-1971 is discussed.

5) Regulatory Guide 1.62 is complied with as discussed in response to Item 2 of this attachment where IEEE 279-1971 is discussed.

6) Regulatory Guide 1.63 is not applicable.

7) Regulatory Guide 1.68 will be discussed af ter equipment f abrication .

8) The extent of compliance with Regulatory Guide 1.75 is discussed in response to FSAR Question 3.1.2

9) Regulatory Guide 1.89 is not a licensing basis for design; however the requirements of IEEE 323-1971 are met.

i687 257 3-2

Docket No. 50-346 License No. NPF-3 Serial No. 540 October 3, 1979 Response to Item 1 (continued)

10) Regulatory Guide 1.100 is not a licensing basis for design; however, the requirements of IEEE 344-1975 are met as supplemented by the following (applicable sections of IEEE 344-1975 are referenced):

Section 5.3 - Static Coefficient Analysis: This section states that a static coef ficient of 1.5 is used for equipment analysis to take into account the effects of both multi-f requency excitation and multimode response. The use of 1.5 as a static coefficient is not acceptable unless specifically justified by analysis. Similarly, any alternate static co-ef ficient the SUPPLIER may choose to utilize shall also be justified by analysis.

Section 6.6.2.1 - Derivation of Test Input Motion: This section states that for equipment with more than one predominate frequency, the shake table motion should produce a test response spectrum (IRS) acceleration at the test frequencies equal to 1.5 times the acceleration given by the specified required response spectrum (RRS) or less if justified. The section also states that the TRS need not envelop the RRS provided the factor of 1.5 and the concept that the TRS not envelop the RRS is not acceptable unless specifi-cally justified.

Section 6.6.2.5 - Sine Sweep Test: Those portions of this section which indicate that, for qualifying equipment using the sine sweep test input , the TRS must envelop the RRS according to the criteria in Section 6.6.2 and 6.6.2.1 are not acceptable in the absence of specific justification.

11) Regulatory Guide 1.105 is not a licensing basis for design; however, the design meets this guidance with the exception of the requirement of securing devices and the design verification of the instrument qualifi-cation program reco= mended in Regulatory Guide 1.89. As noted above, the design meets IEEE 323-1971.

12) Regulatory Guide 1.118 is not a licensing basis for design; however the requirements of Section 5 of IEEE 338-1971 are met as noted in response to Criteria 3 (4) above.

Criteria 5-Branch Technical Positions The Safety Grade Anticipatory Reactor Trip System conforms to the BTPs as dis-cussed below:

1) The design meets the requirements of S!P ICSB 1,9, 21 and 22.

2) BIP ICSB 5 is not applicable since the scope of the ARTS is upstream of the scram circuitry.

3) BTP ICSB 12 and 14 are not applicable.

4) The design meets the requirements of BTP ICSB 26 except for the turbine trip pressure switches which are located in a non-seismic category I building.

1687 2?5B B-3

Docket No. 50-346 License No. NPF-3 Serial No. 540 October 3, 1979 Item 2 Provide a discussion of the following:

a. design basis information required by Section 3 of IEEE-279-1971, and

b. conformance with the design requirements of Section 4 of IEEE-279-1971.

Response

A. The design basis information of the Safety Grade Anticipatory Reactor Trip System as required by Section 3 of IEEE Sta. 279-1971 are as follows:

1. The generating station conditions which require protective action are:

a. Loss of Main Feedwater Indicated by reverse main feedwater check valve delta-pressure

b. Turbine - Generator trip

c. Main Feedpump trip

d. Low Steam Generator level

2. The generating station variables that are required to be monitored in order to provide protective actions are:

a. Turbine - Generator Status

b. Main Feedpump Turbine Status

c. Steam and Feedwater Rupture Centrol System (SFRCS) Status

3. The number and location of sensors provided to monitor for protective function purposes, those variables that have spatial dependence.

Turbine - Generator Status -

Four pressure switches mounted on the fast acting solenoid for the turbine - generator control valves.

Main Feedpump Turbine Status -

Four pressure switches mounted on each pump monitoring the oil pressure of the control valves.

SFRCS Status -

Four signals, one from each SFRCS logic channels, actuated on a low Steam Generator level, a loss of main feedwater, or low steam line pressure.

4 Prudent operational limits for each variable in eac'1 reactor operation mode:

Turbine - Generator Status -

The normal operational limits of the pressure in the oil system is 1600 psig.

1687 259 B-4

. Docket No. 50-346 License No. NPF-3 Serial No. 540 October 3, 1979 Main Feedpump Status -

The normal operational limits of the pressure in the oil system is 200 psig SFRCS Status -

Refer to FSAR Section 7.4.1.3.10, Paragraph 4

5. Margin between operational limit and level marking onset of unsafe conditions:

Turbine - Generator Status -

The minimum operational pressure in the oil system is 1300 psig. This pressure provides a margin of 700 psig between the minimum operation pressure and the pressure requiring protective action.

Main Feedpump Turbine Status -

The minimum operational pressure in the oil system is 130 psig. This pressure provides a margin of 55 psig between the minimum operation pressure and the pressure requiring protective action.

SFRCS Status -

Refer to FSAR Section 7/4/1/3/10, Paragraph 5

6. The level that, when reached, will require protective action:

Turbine - Generator Status -

The Reactor will be automically tripped when the oil pressure decreased to 600 psig.

Main Feedpump Status -

The Reactor will be automatically tripped when the oil pressure decreased to 75 psig.

SFRCS Status -

Refer to FSAR Section 7.4.1.4.10, Paragraph 6.

7. Range of transient and stfady-state conditicas of the energy supply and the environment during normal, abnormal and accident circumstances throughout which the system must pe rf o rm.

The essential power supply is discussed in Chapter 8 of the FSAR.

The SFRCS and the Safety Grade ARTS are located in the control room. All other sensors are located in the turbine building where the environmental conditions are 40 to 100 percent relative humidity, atmospheric pressure and temperatures to 120 F.

1687 2s60 3-5

Docket No. 50-346 License No. NPF-3 Serial No. 540 October 3, 1979

8. The malfunctions, accidents or other unusual events which could physically damage protection system components and for which provisions must be incorporated to retain necessary protective action:

The Safety Grade ARTS is designed to withstand physical damage or loss of function caused by earthquakes. It is also located in a building area designed to protect the equipment f rom flood, lightning, wind and missiles.

Pressure switches used as sensors for the system will conform to IEEE 279-1971 and be environmentally qualified.

However, seismic criteria are not included in qualification regarding mounting and location for that portion of the trip system located within the nonseismic Category I turbine building.

9. The Safety Grade ARTS is digital and therefore, the system response time is less than 1 second.

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. .i.. ----i-

, Docket No. 50-346 License No. NPF-3 Serial No. 540 October 3, 1979 B. The following discussion is keyed to Section A of IEEE Std. 279-1971 and demonstrates a compliance with the above mentioned s t andard :

1. General Functional Requirement - The Safety Grade ARTS will, with precision, and reliability, automatically perform its protective function, whenever the station conditions monitored by the system reach a preset level, under the design condition listed in the discussion of Section 3 of IEEE Std. 279-1971.

2. Single Failure Criterion - No single f ailure can prevent the system from performing its protective function.

3. Quality of Components and Modules - The system consists of high quality components and modules with minimum maintenance requirements and low failure rates. Quality control procedures will be used during f abrication and testing to verify compliance with requirements specified for the particular equipment.

4. Equipment Qualification - Type test data will be available to verify that the system equipment meets, on a continuing basis, the performance requirements determined to be necessary for achieving the system requirements.

5. Channel Integrity - Each channel of the system will be designed, manufactured, and located so that the channel integrity is maintained under the design conditions listed in the discussion of Section 7 of IEEE Std. 279-1971.

6. Channel Independence - Each system channel is located in its own cabinet. The cabinets act as a barrier against fire and mechanical damage from external so mces.

The cabinets are in a room which of fers environmental and missile protections.

7. Control and Protection System Interaction

a. Classification of Equipment - Equipment that is used for protection and control function is classified as part of the protection system and meets the requirements of IEEE Std. 279-1971.

b. Isolation Devices - Output signals from the system will be through isolation devices which will be classified as part of the system and meet all the requirements of IEEE Std. 279-1971.

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Docket No. 50-346 License No. NPF-3 Serial No. 540 October 3, 1979

c. Single Random Failure - A single random f ailure resulting in a control system action si=ultaneously causing a channel failure and a station condition requiring protective action is incredible.

d. Multiple Failures Resulting from a Credible Single Event -

No control system action can result in a condition requiring protective action and can concurrently prevent the protective action of any Safety Grade ARTS channel.

8. Derivation of System Inouts - The system inputs are digital inputs that are direct measures of the station parameters.

9. Capability for Sensor Checks - All inputs are digital. Input sensor indicating lights will be provided. Test pushbuttons for these Laputs signals will also be provided at each cabinet.

10. Capability for Test and Calibration -

a. Manual testing will be provided for each input signal to the system to simulate sensor operation.

b. Msaual calibration capability is provided for pressure switches from the Turbine - Generator and Main Feedpump Turbines. These can be independently isolated and simulated process parameters applied to check calibration. For manual calibration capability of the SFRCS inputs, refer to FSAR Section 7.4.2.3.1 Paragraph 10b.

11. Channel Bypass or Removal from Operation - Each Safety Grade ARTS sensing and logic channel is provided with one key operated rotary test trip bypass switch. Thi. switch enables the operator to change the two-out-of-f our coincidence matrices into a two-out-of-three mode for one given variable. The channel bypass permits .

the testing, calibratien and maintenance of a particular generating station variable of a single channel during power operation. With the bypass in ef fect the three remaining channels of that station variable provide the necessary protection.

Since enly two channels of a variable need exceed the trip setpoint to cause a trip, a single failure will not prevent the station variable logic from fullfilling its protective function.

12. Operating Bypasses - The operating bypasses consist of blocking the Turbine - Generator Status input when the reactor power is below 20 percent. The bypass is automatically initiated and automatically removed when reactor pcwer is above 20 percent.

The bypass signals are part of the protective system and originate from the Reactor Protection System.

B-8 1687 263

Docket No. 50-346 License No. NPF-3 Serial No. 540 October 3,1979

13. Indication of Bypasses - Initiation of the channel bypass will be continuously indicated at the system cabinets and the station computer and annunciator. Initiation of the operating bypass will be continuously indicated at the system cabinet, at the RPS cabinet and at the station computer and annunciator.

14. Access to Means for Bypassing - The activation of the Safety Grade ARTS channel bypass is accomplished by using key switches, which are under administrative control. Also, to initiate a bypass, a corresponding cabinet door must be opened. The cabinet door keys are also under administrative control.

15. Multiple Setpoints - The Safety Grade ARTS does not use multiple se tpo int for any staticn parameters.

16. Completion of Protective Action Once It Is Initiated - The reactor, once tripped by the system, cannot be restarted unti3 the operator deliberately resets the individual cabinets when station parameters return to normal.

17. Manual Initiation - Manual Laitiation of the system is accomplished by means of the reactor trip buttons located on the main control board.

18. Access to Setpoint Adjustments, Calibrations, and Test Points -

Setpoint adj us tment and calibration of the station parameter sensing switches are under administrative controls. The test points will be accessible only when the system cabinet doors are open.

The door keys are under administrative control. Open doors will be alar =ed by the statien computer and annunciator.

19. Identification of Protective Actions - Protective action will be initiated whenever the generating station parameters sensed exceed the setpoint. These parameters are alarmed on the station annunciator and/or station computer. Each trip will also be indicated by the logic system in the system logic cabinets. Failures of power supplies are also monitored at the station computer and annunciator and indicated at the logic cabinets.

20. Information Readout - This is a digital system but analog signals f rom the station parameters are monitored by the station computer.

21. System Repair - The periodic testing can locate failure in a logic system. The modular design of the system will allow for quick repair of =alfunctions.

22. Identification - The identification of the equipment, including cabinets, trays and cables of the system redundant portions, is accomplished by color coding and numbering as described in Chapter 8 of the FSAR.

i687 264 B-9

Docket No. 50-346 License No. NPF-3 Serial No. 540 October 3, 1979 Item 3 Provide a description of any changes to and/or interfaces with the existing protection system. Include diagrams (block, location, functional and/or elementary wiring), as necessary, to clearly depict the changes and/or inter-faces. In addition, provide an analysis which demonstrates that these changes and/or interfaces will not degrade the existing protection system.

Response

The Safety Grade ARTS interfaces with two protectica syste=s. These include the Reactor Protection System (RPS) and the Steam and Feedwater Rupture Centrol System (SFRCS) . All interf aces are shown on SK-E-410, attached.

The output frem the SFRCS consist of an isolated output from each of the 601, 2, 3 & 4 output logic (refer to FSAR Figure 7- 28 ). These output signals are generated whenever there is low steam generation level, low steam pressure- or high reverse differential pressure across the main feedwater check valve on either steam g+nerator.

An isolated output signal from the RPS is used to block the rurbine Trip input signal when reactor power is less than 20%. This signal will be generated and reset automatically from an isolated bistable in each of the four RPS channels.

The output from the ARTS will be terminated in the RPS cabinet as part of the reactor trip circuit. The signals will be isolated and independent.

No single failure will create an adverse effect on plant safety. All interfaces between protection systems and the ARTS are isolated. The isolation devices will be qualified to withstand any adverse condition which could degrade the operation of the protection system.

Item 4 Identify equipment which is identical to equipment utilized in existing safety-grade systems. For the equipment which is not identical, briefly describe the difference.

s Resconse The logic portion of the Safety Grade ARTS is designed to be similar to the Safety Features Actuation System (SFAS) logic. The components of the ARTS supplied may be either similar or identical to the SFAS, depending on the vendor selected for the system. All cabling for the system will be identical to cable installed at the plant. All compcnents of the system shall be required to be of equal quality to components presently installed at the plant. Any substantial differences in components for this system from those used .nthe plant will not be fully determined until a vendor has been chosen and has submitted his design.

1687 265 B-10

Docket No. 50-346 License No. NPF-3 Serial No. 540 October 3, 1979 Item 5 For all critical equipment , provide an expected delivery date.

Response

Currently without vendor bids received, there is no delivery schedule.

It could be expected if there is no alteraticn of the bid design as a result of NRC review that delivery may be expedited to six months af ter conceptual approval. If there are changes affecting design, additional delivery time would be expected.

Item 6

?

In general, the equipment shall be seismically and environmentally qualified.

Therefore, provide the follow descriptive information for the qualification test program:

a. equipment design specification requirements

b. test plan

c. test setup

d. . test procedures

e. acceptablilty goals and requirements If the above information is not available at this time, provide a schedule for its submittal.

? -

Response

A. The equipment shall be designed to f unction continuously at ambient temperatures ranging from 40 F to 110 F and at a relative humidity up tc 80 percent. The ARTS cabinets sh'all be seismically tested to meet the qualification for Class lE electrical equipment for the plant.

Enclosed are five of the latest drawings of ARTS (Bcchtel Co. Drawing

,No. SK-E-410, Revision 1) that include the design criteria and syste=s description. This design has been modified f rom the original sub=ittal, of Revision 0 (TECo letter 1-71, dated May 21, 1979) to incorporate reactor trip experience since May of this year. The low reactor power block signal that re=oved ARTS protection below 20% reactor power has been removed for all trip signals except the =ain turbine trip.

B-E This information is dependent on vendor selection and is not available at this time. Toledo Edison will provide this information three months prior to equipment delivery. .

B-11 1687 266

Dock at No. 50-346 License No. NPF-3 Serial No. 540 October 3, 1979 Item 7 Identify the portion (s) of the design which are within the scope of supply of B&W and/or other contractors.

Response

The low power block bistable and associated hardware located in the reactor protection system cabinets will be supplied by B&W. All other components will be supplied by other vendors who will be chosen later.

Item ,8 Provide the criteria for the overall reactor protection system installation testing which will demonstrate that the new trip has been installed properly.

If this information is not available at this time, provide a schedule for its submittal.

Response

This criteria will be provided with the response to Item 6, B-E above.

It is noted that this design has minimal interface (low block) with the reactor protection system.

Item 9 The safety evaluations for the anticipatory trips are either missing or are incomplete. Submit supporting analysis to justify the proposed trip signals by addressing the following items:

a. Provide an analysis that quantifies the improvement in the time-to-reactor-trip for both the turbine trip and the loss of main feedwater signals;

b. Address the need to bypass these trips at 20% power versus bypass at a lower power (approximately 10%);

c. Discuss the adequacy of the proposed trip signals for loss of main feedwater for a variety of f ailure scanarios (such as feedwater valve closures), i.e.

see the Ocenee 1 transient of 6/11/79; and

d. Provide an evaluation of why a reactor trip on low steam generator level is not a viable anticipatory trip signal when the other signals are bypassed i.e. , see the Crystal River 3 transient of 9/02/79.

B-12 1687 267

. Docket No. 50-346 License No. NPF-3 Serial No. 540 October 3, 1979 Response to Question 9 9a. The primary purpose of anticipatory reactor trip system (ARTS) is to reduce the probability of lif ting the PORV for turbine trip / loss of main feedwater type events. For a reactor high pressure trip setpoint of 2300 psig, it was shown in References 1 and 2 that the PORV would not lift with a setpoint of 72400 psig. The margin to the PORV setpoint can be increased, however, by use of ARTS. Figure 9a-1.shows the pressure increase from nominal operating pressure as a function of time to trip for the loss of main feedwater event.

From this figure, it can be seen that an ARTS that detects and trips the plant at 4 seconds results in a peak pressure increase of 60 psi; whereas the high R.C. pressure trip which would occur at 8 seconds results in a peak pressure increase of 184 psi. The anticipatory trip signals which have been selected will initiate a reactor trip in less than one second af ter a total loss of all feedwater flow. As seen on Figure 9a-1, a one second time to trip results in a 12 psi pressure increase, compared to a 184 psi pressure Increase for the high pressure trip at 8 seconds.

The analyses presented above are for a loss of main feedwater event which pro-duces higher peak pressures than turbine trips produce. A reactor trip after a turbine trip f rom full power without ARTS would not occur if the ICS were adjusted properly. With ARTS, the reactor trip will occur within one second of the turbine trip.

9b. The only trip that is bypassed in the ARTS at power levels <t20% is the sain turbine trip. Davis-Besse Unit I has a 25% turbine bypass capacity. For any turbine trips at power levels 4 25% there are no pressure transients on the steam generator (SG' secondary side because the turbine bypass system has the capacity to control the SG pressure at the nocmal 370 psig. Following the turbine trip, ICS will run the reactor power back to 15% at a rate of 20% per minute.

9c. The ARTS at Davis-Besse provides a reactor trip on the loss of both main feed-water pumps (MFP) or a steam feedwater line rupture control system (SFRCS) initiation . SFRCS in turn is initiated on loss of four reac~ tor coolant pumps.

reverse delta pressure across either main feedwater line check valve, low main steam line pressure and a low steam generator water level.

The Oconee 1 transient of 6/11/79 was a reactor startup situation with 1 main feedwater pump (MFP) reset and not operating. When the operating feedwater pump tripped, the reactor did not automatically trip on loss of feedwater because the low discharge pressure trip on the reset MEP was not reached prior to the operator manually tripping the plant. The Davis-Besse ARTS reactor trip based on a total loss of feedwater flow to either SG, low SO level in either 50 or a trip of both MFP would have detected this event.

It should be noted that the purpose of .'".T3 is to decrease the probability of PORV actuation on turbine trip / loss of main feedwater type events. Since it has been demonstrated (1,2) that with a reactor trip of 2300 psig and PORV se tpo in t 2400 psig, no lif ting of the PORV will occur, the addition of ARTS only increases the margin to PORV setpoint pressure.

1687 268 3-13

Docket No. 50-346 License No. NPF-3 Serial No. 540 October 3, 1979 Response to Question 9 (continued) 9d. The Davis-Besse ARTS only bypasses the turbine trip signal at low power levels.

All loss of feedwater trips, including low level Lt either steam generator, are functional throughout the power range. This provides adequate coverage for the Crystal River transient of August 2, 1979.

REFERENCES:

1. BSW Report to the NRC, May 7,1979, " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plant."

2. Toledo Edison Report to the NRC, May 16,19 79, " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plant, Volume 3."

3_1, 1687 269

CHANCE IN REACTOR COOLANT SYSTEM PRESSURE VS TIME TO TRIP FOR A LOSS OF MAIN FEEDWATER FROM 100% POWER 240,.

l 200 .

q 160, ,,

c.

c., ,.

<7 g -

a v

g 120-M

=A E

c 9 80--

G 4m 40,,

0 ,

0 2 4 6 8 10 l'2 TIME TO TRIP , SEC.

Figure 9a-1 1687 270

TABLE 9b-1 POWER LEVEL SEiSITIVITY TDE TO REACH TIME TO FILL POWER LEVEL PORV SETPOINT PRESSURIZER 100% 3 min. 10 min.

75% 6 min. 11 min.

50% 12. 3 min. 13 min.

25% 15 min. 16.6 min.

NOTE: RESULTS ARE FOR THE CASE OF NO AUXILIARY FEEDWATER INITIATION WHICH RESULTS IN THE SHORTEST ACTUATION TIMES. REACTOR TRIPS ON HIGH RC PRESSURE TRIP (2300 psig) .

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