ML20205P942

Application for Amend to License NPF-3,revising Tech Spec Tables 3.3-1 & 4.3-1 in Response to Generic Ltrs 83-28 & 85-10 Re Silicon Controlled Rectifier Relays & Reactor Protection Sys Instrumentation

ML20205P942
Person / Time
Site:	Davis Besse
Issue date:	03/27/1987
From:	Myers T, Shelton D TOLEDO EDISON CO.
To:
Shared Package
ML20205P917	List: ML20205P912 ML20205P942 ML20205P954
References
GL-83-28, GL-85-10, TAC-55352, TAC-59932, NUDOCS 8704030410
Download: ML20205P942 (12)

Text

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Dockr,t No. 50-346' i License No. NPF-3 Serial No.- 1312 Attachment APPLICATION FOR AMENDMENT.

TO FACILITY OPERATING LICENSE NO. NPF-3 FOR DAVIS-BESSE NUCLEAR POWER STATION UNIT NO. 1 ,

i Enclosed are requested changes to the Davis-Besse Nuclear Power Station Unit No. 1. Facility Operating License No. NPF-3. Also included are the Safety Evaluation and Significant Hazards Consideration.

The proposed changes (submitted under cover letter Serial No.1312) concern:

Section 3/4.3.1, Reactor Protection System Instrumentation. Table 3.3-1; Section 3/4.3.1, Reactor Protection System Instrumentation Surveillance l Requirements Table 4.3-1 By T. J y he Nuclar/rs Licensing Director i For D. C. Shelton j

Vice President, Nuclear Sworn and subscribed before me this 27th day of March, 1 87.

l $xtwoch Notary Public, State of Ohio

. b;cfu My commission expires d /f /

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8704030410 870327 PDR I

P ADOCK 05000346 l PDR

Dock t Ns. 50-346 License No. NPF-1 Serial No. 1312 Attachment The following information is provided to support issuance of the requested changes to the Davis-Besse Nuclear Power Station, Unit No.-1 Operating License No. NPF-3, Appendix A, Technical Specifications, Table 3.3-1 and Table 4.3-1.

A. Time Required to Implement: This change is to be effective thirty (30) days af ter issuance of the License Amendment.

B. Reason for Change (Facility Change Request 85-0064, Revision C):

Item 4.4 of Generic Letter 83-28 requires that the appropriate surveillance and test sections of the Technical Specifications be revised to include testing of the silicon controlled rectifiers used to interrupt power to the control rods.

Appendix A, Technical Specifications, Tables 3.3-1 and 4.3-1, do not contain silicon controlled rectifier relays as a functional unit.

C. Safety Evaluation: See attached Safety Evaluation.

D. Significant Hazards Consideration: See attached Significant Hazards Consideration.

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Docket No. 50-346 License No. NPF-3

, Serial No. 1312

. Attachment 1 Page 1 SAFETY EVALUATION DESCRIPTION OF THE PROPOSED ACTIVITIES This License Amendment Request is as a result of a directive from the NRC in their safety evaluation report on our addition of the shunt trip device in each of four Reactor Trip Breakers and as a result of Generic Letter 85-10. The sole purpose of this request is to meet the NRC directive to incorporate the testing requirements into the Technical Specifications.

SYSTEM AFFECTED Control Rod Drive Control System DOCUMENTS AFFECTED NRC Supplemental Safety Evaluation Report (Log No. 1683)

B&W Report on " Review of On-Line Test Intervals for the Reactor Trip System "

FCR 84-0026 (Revision A), Addition of the Shunt Trip Device to each Reactor Trip Breaker Technical Specifications for Generic Letter 83-28, Items 4.3 and 4.4 (Generic Letter 85-10) Log No. 1756 l -

B&W " Acceptance Criteria for Reactor Trip Breaker Response Time Testing" Document Identifier 51-1151182-00 SAFETY FUNCTION OF SYSTEM AFFECTED The safety function of the reactor trip breakers is to remove the power from the Control Rod Drive Mechanisms allowing the safety and regulating rods to drop into the core. This brings the reactor to a suberitical condition whenever the reactor trip breaker receives an automatic trip i command signal (undervoltage conditions) from the Reactor Protection System (RPS), Anticipatory Reactor Trip System (ARTS), or a manual trip

command signal from the operator. This safety function is accomplished by l the undervoltage device (UVD) on loss of voltage due to interruption of l the circuit by an RPS, ARTS, or manual trip signal.

l The aafety function of the shunt trip device is to provide additional l assurance that the reactor trip breakers will open when an automatic trip command signal from the RPS, ARTS, or a manual trip command signal from the operator is received.

The function of the Silicon Controlled Rectifier (SCR) relays is to interrupt the gate control signals to the SCRs in each of the nine CRDM motor power supplies and the motor return power surply. When the gate signals are interrupted, the SCRs will revert to their open state on the negative half-cycle of the applied voltage, thus removing power at the output of the motor power supplies. There is no safety function accociated with the SCR relays.

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D;ck;t N . 50-346 License No. NPF-3 Serial No. 1312 Attachment 1 Page 2 EFFECTS ON SAFETY / PROPOSED CHANGES There are three alternate means of interrupting power to the Control Rod Drive Mechanism as follows:

1. Interruption of power to the undervoltage device (UVD) for the Control Rod Drive Mechanism breaker. This releases spring stored energy to rotate the breaker trip shaft, opening the main breaker contacts.

2. Interruption of power to an undervoltage relay in parallel with the undervoltage device (UVD) permitting the undervoltage relay contacts to energize the shunt trip device powered from 125 VDC Essential Buses. This shunt trip device independently provides additional torque to rotate the reactor trip breaker trip shaft thereby opening the main breaker contacts.

3. Interruption of the gate control signals to the SCRs in each of the nine Control Rod Drive Mechanisn (CRDM) motor supplies, and the motor return power supply. The trip c'evices in this case are ten relays connected with their coils in ptrallel. Contacts of these relays interrupt the gate control signals to the SCRs in each power supply.

When the gate signals are interrupted the SCRs will revert to their open state on the next negative half-cycle of the applied AC voltage, thus removing all power at the outputs of the motor power supplies.

Because the power supplies have redundant halves, two sets of ten relays each are provided. Reactor Protection System (RPS) channel 3 energizes one set of trip relays and RPS channel 4 energizes the other set through auxiliary relays (in parallel with the undervoltage device and undervoltage relay for the shunt trip device) in CRD breakers C and D. The trip relays can remain in their non-tripped state only if the associated RPS chamiel is energized. When an RPS i channel trips, the associated trip relays de-energize, interrupting l the SCR gate control signals. I The maximum allowable trip command time delay (difference between time trip input is received by the reactor trip breaker from the RPS and time power is interrupted to the control rod drive mechanisms) within the control rod drive control system is 100 milliseconds. Testing currently measures this trip command time delay (or reactor trip breaker opening time) upon loss of voltage at the UVD due to interruption by an RPS trip signal. Testing shall be modified to ensure the' trip command time delay of 100 milliseconds is not exceeded upon actuation of the shunt trip device.

Docket,No. 50-346

License No.,NPF-3 1 Serial No. 1312

, Attachment 1 Page 3.

The: proposed amendment request involves the-folicwing changes:

, 1. The-addition of silicon controlled rectifier relays'to Tables /3.3-1 and 4.3-1.

2.- The addition of Action Statements 9 and 10.to Table 3.3-1.

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3.- The addition of Notation 9 to Table 4.3-1..

The addition of the silicon controlled rectifier relays to: Table 3.3-1,- _

-Reactor Protection System Instrumentation, and Table 4.3-1,' Reactor Protection System Instrumentation Surveillance Requirements, will provide a surveillance schedule.for testing these devices and assure the operabil -

ity requirements of the reactor trip system are met. It will also ensure that a second diverse and independent trip is initiated when the. gating g

signals to the silicon controlled rectifiers (SCRs) are: interrupted.

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The addition to the(Davis-Besse Unit No. 1 Technical Specification should

. g' q not result in additional testing of the reactor trip. breakers as proce-3 dures have previously been modified to incorporate the requirement.to independently confirm the shunt and undervoltage trip devices. This was accomplished following the addition of the shunt trip device. Testing of the SCR relays is currently performed. This testing does not trip.the Reactor Trip Breakers. The frequency of the channel functional test of the SCR relays will be specified as at least once per 18 months. The less frequent testing (for the SCR relays in the Davis-Besse design is due to

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their less critical, i.e., duplicative function. That is, the non-1E, non-seismic SCR relays are a backup tripping means co the Reactor Trip Breakers. This is consistent with the NRC guidance provided in Generic Letter 85-10.

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The frequency of testing as a result of the independent verification was ,

addressed by the B&W Owners Group in their report entitled " Review of ~l On-Line Test Intervals for the Reactor Trip Systems."

This report concluded:

"The wearout evaluation indicated that the Reactor Trip Systen (RTS)  :

components are.not susceptible to wearout caused by testing. The i breakers are the major components affected by test cycling and the CE AK-2 breaker has a design cycle objective of 12,500 cycles.' Aging of the trip shaft bearing lubricant is virtually eliminated as a concern when the Mobil 28 lubricant is installed" (this is completed -

j at Davis-Besse). "Therefore, for the breakers, common mode failure due to wearout is not a significant source of RTS unavailability. j Other components do not exhibit histories that indicate the wearout l 1s a concern."

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' Dock::t No. 50-346 3i y

-License No. NPF-3 Serial No. 1312- 1-Attachment 1 Page 4  %

v Surveillance.testingnowrequiresI3'tripspermonthor36peryear. The design number of cycles 12,500 over the-40 year plant life allows for an

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average of 312 per year. This margin is more thanfsufficient to allow for

.r reacter trip and bench testing (the'12 month schedule for: maintenance of +

the breakers). 'X f-p

' Action' Statements 9 and 10 in Table 3.3-1 will not be subject.to the'!

provision of Appendix _ A, Technical Specification 3.0.4,: Limiting Condition for Operation. Exceptions to this. Specification are provided because) changing modes with inoperable equipment would not affect plant safety}

For the control rod drive trip. breakersi (Action 9) with one trip device -

inoperable, the other trip device would still function to ensure reactor- ..

shutdown upon an automatic trip command signal from the RPS, ARTS, or al . h/ , s manual trip command signal from the operator. For the SCR relays (Action 10) .g with one or both channels inoperable, both the undervoltage and shunt . trip devices would function to open their respective trip breaker upon an

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automatic trip command signal from the RPS, ARTS, or a ma'nual trip command ,

signal from the operator. Thus redundant tripping means would exist to ensure plant safety. ' '

Proposed Action Statement 9 wording to Table 3.3-1 'is consistenti with that provided by NRC Generic Letter 85 '10 (Log' No.1756) with andtexception. "'L C term " devices" was retained instead of " attachment". The tern'" attach-ment" applies to a Westinghouse breaker; 3rhereas the term '" devices" ' - '

applies to our GE breakers. ' '

Proposed Notation 9 wording to Table 4.3-1;is consistent with that provid . b

l The~ word edbyNRCGenericLetter85-10(LogNo.g;756)with'oneaddition.

"both" has been added to clarify that the undervoltage trip device and shunt trip device are tested independently of each other.

UNREVIEWED SAFETY QUESTION EVALUATION -

The implementation of these proposed changes.would not:

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1. Increase the probability of occurrence of an, accident previously /

evaluated in the Safety Analysis Report because these changes are. .

increasing the surveillance requirements -(10CFR50.59(a)(2)(1)).-

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2. Increasetheconsequenceofanaccidentprevhous,1yevaluatedinthe. I safety analysis report becarce these changes at'e increasing the -I surveillance requirements (10CFR50.59(a)(2)(1))'.'

-3. Increase the probability of.a malfunction of equipment important to t safety because the increased surveillance requiremer:ts reduce the probability .of reactor trip breaker malfunction (10CFR50.59(a)(2)(i)). '

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4. Increase the consequences of a malfunction of equipment important to safety because these iicreased surveillance requirements ensure that M

\s an alternate means for safe reactor shutdown is operable (10CFR'50.59(a) (2)(i)) . ,

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• Dock;t'Na. 50-346 License No. NPF-3 Serial No. 1312 Attachment 1 g-,, Page 5 1  ;

Mg 5. Create a possibility for an accident of a different type than any evaluated previously in the safety analysir, report because no station 4

equipment is-modified by these changes and the control rod drive control system will continue to be tested to ensure its operability (10CFR50.59 (a) (2) (ii)) .

6. Create a possibility for a malfunction of equipment of a different type than any evaluated previously in the safety analysis report because no station equipment is modified by these changes and the control rod' drive control. system will continue to be tested to ensure its operability (10CFR50.59(a)(2)(ii)).

7. Reduce any margin of safety as defined in the basis for these Techni-cal Specifications because reactor trip breaker surveillance testing will measure reactor trip breaker opening time for UVD and shunt trip device actuation. This testing ensures that the RPS action function associated with each channel is completed within the time limit assumed in the safety analysis (10CFR50.59(a)(2)(iii)).

CONCLUSION Based on the above, an unreviewed safety question is not involved, s

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D;ckat N]. 50-346 License No. NPF-3 Serial No. 1312 Attachment 2 Page 1 SIGNIFICANT HAZARDS CONSIDERATION DESCRIPTION OF THE PROPOSED ACTIVITIES This License Amendment Request has been prepared as a result of a directive from the NRC in their safety evaluation report on our addition of the shunt trip device to each of four Reactor Trip Breakers and as a result of Generic Letter 85-10. The sole purpose of this request is to meet the NRC directive to incorporate the testing requirements into the Technical Specifications.

SYSTEM AFFECTED Control Rod Drive Control System DOCUMENTS REFERENCED NRC Supplemental Safety Evaluation Report (Log No. 1683)

B&W Report on " Review of On-Line Test Intervals for the Reactor Trip System" FCR 84-0026 (Revision A) Addition of the shunt trip device to each Reactor Trip Breaker Technical Specifications for Generic Letter 83-28, Items 4.3 and 4.4 (Generic Letter 85-10) Log. No. 1756 B&W " Acceptance Criteria for Reactor Trip Breaker Response Time Testing" Document Identifier 51-1151182-00.

SAFETY FUNCTION OF SYSTEM AFFECTED The safety function of the reactor trip breakers is to remove the power from the Control Rod Drive Mechanisms (CRDMs) allowing the safety and regulating rods to drop into the core. This brings the reactor to a suberitical condition whenever the reactor trip breaker receives an automatic trip command signal (undervoltage conditions) from the Reactor Protection System (RPS), Anticipatory Reactor Trip System (ARTS), or a manual trip command signal from the operator. This safety function is accomplished by the undervoltage device (UVD) on loss of voltage due to interruption of the circuit by an RPS, ARTS, or manual trip signal.

The safety function of the shunt trip device is to provide additional assurance that the reactor trip breakers will open when an automatic trip command signal from the RPS, ARTS, or a manual trip command signal from the operator is received.

The function of the Silicon Controlled Rectifier (SCR) relays is to interrupt the gate control signals to the SCRs in each of the nine CRDM motor power supplies and the motor return power supply. When the gate signals are interrupted, the SCRs will revert to their open state on the negative half-cycle of the applied voltage, thus removing power at the output of the motor power supplies. There is no safety function associated with the SCR relays.

Dick 0t No.!50-346 License No. NPF-3 i Serial No.-1312-Attachment-2 Page 2 4

EFFECTS ON SAFETY / PROF 0 SED CHANGES 8

~ There are three . alternate means of interrupting power to the Control Rod -

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c Drive Mechanism as follows:

. 1. Interruption of power.to the.undervoltage device (UVD) for the Control Rod Drive Mechanism breaker. This releases spring stored energy 1to rotate the breaker trip shaft,' opening the main breaker contacts.

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2. Interruption of power to an undervoltage -relay in parallel with the undervoltage device (UVD) permitting the undervoltage relay contacts to energize the shunt trip device powered from 125 VDC Essential Buses. This shunt trip device independently provides additional torque to rotate the reactor trip breaker trip shaft.thereby opening the main breaker contacts.

3. Interruption of the gate control signals to the SCRs in each of the nine Control Rod Drive Mechanism (CRDM) motor supplies, and the motor j return power supply. The trip devices in this case are ten relays connected with their coils in parallel. Contacts of these relays interrupt the gate control signals to the SCRs in each power supply.

. When the gate signals are interrupted the SCRs will revert to their open state on the next negative half-cycle of the applied AC voltage, thua removing all power at the outputs of the motor power supplies.

Because the power supplies have redundant halves, two sets of ten relays each are provided. Reactor Protection System (RPS) channel 3 .

energizes one set of trip relays.and RPS channel 4 energizes the other set through auxiliary relays (in parallel with the undervoltage

device and undervoltage relay for the shunt trip device) in CRD .

i breakers C and D. The trip relays can remain in their non-tripped state only if the associated RPS channel is energized.- When an RPS

, channel trips, the associated trip relays de-energize, interrupting the SCR gate-control signals.

4 The maximum allowable trip command time delay (difference between time trip input is received by the reactor trip-breaker from the RPS and time-power is interrupted to the control rod drive mechanisms) within the

, control rod drive control system is 100 milliseconds. Testing currently.

measures this trip command time delay (or reactor trip breaker opening time) upon loss of voltage at the UVD due to interruption by an RPS trip

, signal. Testing shall be modified to ensure the trip command time delay of 100 milliseconds is not exceeded upon actuation of the shunt trip device.

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Dockst No.~50-346

. License No. NPF-3

-- Serial No. 1312 Attachment 2' Page 3 The proposed amendment request involves the.following changes:

' 1. The addition of silicon controlled rectifier relays to Tables 3.3-1 and 4.3-1.

2 .' The addition of Action Statements 9 and 10 to Table 3.3-1.

- The addition of Notation 9 to Table 4.3-1.

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The addition of the. silicon controlled rectifier relays to Table 3.3-1, Reactor Protection System Instrumentation, and Table 4.3-1, Reactor E . Protection System Instrumentation Surveillance Requirements, will provide ~

a surveillance schedule for testing these devices and assure the operabil--

ity requirements of the reactor trip. system are met. It will also ensure that a second diverse and independent trip is initiated when the gating signals to the silicon controlled rectifiers (SCRs) are interrupted.

The additions to the Davis-Besse Unit No. 1 Technical Specifications should not result-in. additional testing of the reactor trip breakers as.proce-dures have previously been modified to. incorporate the requirement-to independently confirm the shunt and undervoltage trip devices. This was

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accomplished following the addition of the shunt. trip device. . Testing of

the SCR relays is currently performed. This testing does not trip the Reactor Trip Breakers. The frequency of the channel functional test of the SCR relays will be specified as at least once.per 18 months..'The less frequent testing for the SCR relays in the Davis-Besse design is due to.

their less critical, i.e., duplicative function. ' That is, the non-1E,-

l' non-seismic SCR relays are a backup tripping means to the. Reactor Trip Breakers. This is consistent with the NRC guidance provided.in Generic 1

Letter 85-10.

j

- The frequency of testing as a result of the independent: verification was addressed by the B&W Owners Group in their report entitled " Review'of On-Line Test Intervals for the Reactor Trip Systems."

This report concluded:

"The wearout evaluation indicated that the Reactor Trip System (RTS) components are not susceptible to wearout caused by testing. The breakers are the major components affected by test cycling and the g GE AK-2 breaker has .a design cycle objective of 12,500 cycles. Aging '

of the trip shaft bearing lubricant is virtually eliminated as a

{ concern when the Mobil 28 lubricant is installed" (this is completed i at Davis-Besse). "Therefore, for the breakers, common mode failure due to wearout is not a significant source of RTS unavailability.

Other components do not exhibit histories that indicate the wearout 4

is a concern."

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I D:ckit No. 50-346 License No. NPF-3 Serial No. 1312 Attachment 2 Page 4

' Surveillance testing now requires 3 trips per month or 36 per year.- The design number of: cycles 12,500 over the 40 year plant life allows.for an average of 312 per year. This margin is more than sufficient to allow for reactor trip and bench testing (the 12 month schedule for maintenance of the breakers).

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Action Statements 9 and 10 in_ Table 3.3-1 will not be subject to the e- provision of Appendix A,_ Technical Specification 3.0.4,-Limiting Condition for Operation. Exceptions to this Specification are provided because changing modes with inoperable equipment would not affect plant safety.

For the control rod drive trip ~ breakers-(Action 9) with one trip device inoperable.-the other trip device would still function to ensure reactor shutdown upon an automatic trip command signal'from the RPS, ARTS, or'a-manual trip command signal from.the operator. For the SCR relays (Action 10) .

with one or both channels inoperable, both the undervoltage and shunt trip devices would function'to open their respective trip breaker upon an

, automatic trip command signal from the RPS, ARTS, or.a manual trip command i signal from the operator. Thus redundant tripping means would exist to ensure plant safety.

i Proposed Action Statement 9 wording to Table 3.3-1 is consistent with that provided by NRC Generic Letter 85-10 (Log No. 1756) with one exception. The term " devices" was retained instead of " attachment". The term " attach-ment" applies to a Westinghouse breaker, whereas the term " devices"-

applies to our GE breakers.

l Proposed Notation 9 wording to Table 4.3-1 is consistent with that'provid-ed by NRC Generic Letter 85-10 (Log No. 1756) with'_one addition. The word "both" has been added to clarify that the undervoltage trip device and shunt trip device are tested independently of each other.

l SIGNIFICANT HAZARDS CONSIDERATION The proposed changes do not involve a significant hazards consideration because the operation of the Davis-Besse Nuclear Power Station, Unit No. 1, in accordance with these changes would not:

1. Involve a significant increase in the probability or consequences of  !

an accident previously evaluated because these changes.are  !

increasing the surveillance requirements (10CFR50.92(c)(1)).

2. Create the possibility of_a new or different kind of accident from at.y accident previously evaluated because no station equipment is i

modified by these changes and the control rod drive control system-

] will continue to be tested to ensure its operability (10CFR50.92(c)(2)).

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Docket No. 50-346 License No. NPF-3 Serial No. .1312 Attachment 2 Page 5

3. . Involve a significant reduction in a margin of safety because reactor trip breaker surveillance testing will measure reactor trip breaker opening time for UVD.and shunt trip device actuation. This.

testing ensures.that the RPS action function associated with each channel is completed within the time limit assumed in the' safety analysis (10CFR50.92(c)(3)).

CONCLUSION On the basis of the above, Toledo Edison has determined that the Amend-ment Request does not involve a significant hazards consideration.

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