NRC Generic Letter 1985-10

NRC Generic Letter 1985-010: Technical Specifications for Generic Letter 1983-028, Items 4.3 and 4.4

ML031140409
Person / Time
Issue date:	05/23/1985
From:	Thompson H Office of Nuclear Reactor Regulation
To:
References
GL-85-010, NUDOCS 8505210131
Download: ML031140409 (16)

UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, D. C. 20555 May 23, 1985 TO ALL BABCOCK AND WILCOX PRESSURIZED WATER REACTOR LICENSEES AND APPLICANTS

Gentlemen:

SUBJECT: TECHNICAL SPECIFICATIONS FOR GENERIC LETTER 83-28, ITEMS 4.3 AND 4.4 (Generic Letter 85-10)

Item 4.3 of Generic Letter 83-28, "Required Actions Based on Generic Implications of Salem ATWS Events," established the requirement for the automatic actuation of the shunt trip attachment for B&W plants. Also, licensees are to submit any needed technical specification change requests as soon as practical following staff review and approval of the modified design. 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 control rods.

In the staff's evaluation of the B&W generic design modifications for auto- matic actuation of the shunt trip attachment, as described in individual letters to all B&W operating reactors, dated September 12, 1983, the staff concluded that technical specification changes should be proposed by licensees and that they would be reviewed on a plant specific basis. In the staff's

• review of plant specific responses to the generic letter, some licensees have indicated that changes to the technical specifications are not required. In such cases, the staff has found this to be unacceptable and has indicated that proposed technical specification changes should be submitted to reflect inde- pendent testing of the shunt and undervoltage trip attachments consistent with the design of the test features provided.

Therefore, licensees are requested to submit proposed technical specification changes which are responsive to the guidance noted in the enclosure. The enclosed guidance will be used to revise the Standard Technical Specifications for B&W plants, and it will be used by the staff as a basis to review changes to technical specifications submitted by licensees and for the review of proposed technical specifications for operating license applications.

For plants which have implemented the shunt trip modifications, a schedule for submittal of proposed technical specification changes should be established through discussions with the individual Project Manager for each facility. 6

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-2- In addition, discussions with the individual Project Managers should establish a schedule for plants which have not implemented the shunt trip modifications.

Proposed technical specifications should be submitted as soon as practical following staff review and approval of modified design. For operating license applicants, proposed technical specifications should include requirements which are responsive to the enclosed guidance.

This request for information was approved by the Office of Management and Budget under clearance number 3150-0011 which expires April 30, 1985. Should you have any questions, the staff contact is R. Karsch. Mr. Karsch can be reached on (301) 492-8563.

Tarigbal Signed by Hugh L Thompson, Jr Hugh L. Thompson, Jr., Director Division of Licensing Enclosure:

Sample Technical Specifications List of Generic Letters

• PREVIOUS CONCURRENCE SEE DATE

ORAB* SL:TSRG* SL:ORAB* C:ICSB* C:ORAB:DL* AD/SA:DL*

TAlexion:cl EButcher JHannon FRosa GHolahan DCrutchfield

03/4/85 03/4/85 03/6/85 03/11/85 03/7/85 03/18/85 D:DL -

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05/16 /85

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ENCLOSURE

TECHNICAL SPECIFICATION CHANGES

FOR REACTOR TRIP BREAKERS

(B&W PLANTS)

Background As a consequence of the Salem ATWS event, Item 4.3 of Generic Letter 83-28 established the requirements for the automatic actuation of the shunt trip

- attachment for reactor trip breakers. Further, licensees are to submit any needed technical specification change requests prior to declaring the modified system operable. A number of the responses from operating reactors have indicated that no technical specification changes are required for this modification.

The staff has reviewed the guidance provided in the Standard Technical Specifications (STS) for B&W Plants, NUREG-0103, and finds that additional clarification of both the limiting conditions of operation and surveillance requirements are appropriate as a result of the design modifications to include automatic actuation of the shunt trip attachments. In addition, Item

4.4 of the generic letter reauires that technical specification surveillance requirements be revised to include testing of the silicon controlled rectifiers (SCP). The STS for BW Plants will be revised to include the changes noted herein. Pending formal revision of the STS, this document provides guidance to licensees and operating license applicants on appropriate technical specifications in response to Items 4.3 and 4.4 of the Generic Letter.

K>1

-2- Discussion The operability requirements for the reactor trip breakers are specified in Table 3.3-1 of the STS (see Attachment 2). The specification states that both reactor trip breakers shall be operable in Modes 1 and 2, and when the breakers are in the closed position, the control rod drive system is capable of rod withdrawal, and fuel is in the reactor vessel. The action statement for an inoperable breaker requires that the breaker be placed in a tripped condition within one hour.

With the addition of the automatic actuation of the shunt trip attachment (STA), diverse features exist to effect a reactor trip for each breaker. If one of these diverse trip features is inoperable, a decision would have to be made with regard to the operability status of the reactor trip breaker. The definition of OPERABLE-OPERABILITY in Section 1.0 of the STS states that a component shall be operable or have operability when it is capable of performing its safety function. Since either trip feature being operable would initiate a breaker trip on demand, it would be overly conservative to treat a breaker as inoperable if one of these diverse trip features were inoperable. However, on the other hand the reliability of the reactor trip system would be reduced if each diverse trip feature is not maintained in an operable status.

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The reactor trip breaker surveillance test should independently verify the operability of the shunt and undervoltage trip features of the reactor trip breaker as part of a single sequential test procedure. Therefore, the surveillance test which identifies a failure of one diverse trip feature also confirms the operability of the other trip feature. As a consequence, there is a higher degree of confidence that this trip feature would be capable of initiating a reactor trip in the next 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. Accordingly, an additional action statement will be included in the STS for the reactor trip breakers as follows:

ACTION - With one of the diverse trip features (undervoltage or shunt trip attachment) inoperable, restore it to OPERABLE status in 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or place the breaker in trip in the next hour.

In accordance with the requirements of Item 4.4 of the Generic Letter, the SCR relays have also been included in the changes to Table 3.3-1 to define their operability requirements. The reactor trip system design for B&W

plants Includes two basic configurations; the Oconee design shown in Figure

3.4 and the Davis Besse design shown in Figure 3.5 (see Attachment 1). In the Oconee design the SCR relays for the regulating rods duplicate the trip function of the DC breakers for the safety rods. Therefore, for this design an inoperable channel should be placed in trip as required by action statement 7. However., in the Davis Besse design, the SCR relays provide a third means to insure that power is removed from all rods to Initiate a

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reactor trip. Therefore, placing inoperable channels of SCR relays in trip would only increase the potential for inadvertent reactor trips without significantly reducing the potential of an ATWS event, when considering the increased reliability of the reactor trip breaker afforded by the incor- poration of diverse trip features. For plants with the Davis Besse design, a new action statement will be included in Table 3.3-1 of the STS as follows:

ACTION - With one or both channels of SCR relays inoperable, restore the channels to OPERABLE status during the next COLD

SHUTDOWN exceeding 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The B&W generic design modification includes test features which permit independent testing to verify the operability of the shunt and undervoltage trip attachments. As noted above, operability as applied to the diverse trip features of breakers may have different degrees of safety significance. In order to be consistent with the intent ot the test features provided, the following notation will be included in the surveillance requirements specified in STS Table 4.3-1 for reactor trip breakers (see Attachment 2):

"The CHANNEL FUNCTIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip attachments of the Reactor Trip Breakers."

-5- Consistent with the requirements of Item 4.4 of the Generic Letter, the surveillance requirements for SCR relays will also be included in Table 4.3-1.

For plants with the Oconee design, the channel functional test of the SCR-

relays will be specified as monthly since the SCR relays used for the regu- lating rods duplicate the trip function of the DC breakers for the safety rods. For plants with the Davis Besse design, 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, safety function.

Attached are marked-up pages of the applicable STS tables with these changes. Proposed changes to plant specific technical specifications will be evaluated by the staff based on this guidance.

AtL IPOWER SOURCE8 FOR A PLANT PROCM

OIVEN ROD GROUP mS E INSTRUMENT CHANNflS

INTERRUPTED IN ORDER FOR ISENSORS AND

THE RODS IN THAT GROUP TANSMIllERS.

TO DROP INTO THE CORE SISTASLES. ETC.I AND

FIELD

CONTACT

S

TRIP MODULES

ILOGIC CHANNEW31I

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ECONTR ROD DRIVE CONTROL SV3TMl MANUAL TRIP --

RECTOR TRIP SysT

TRIP MODULE

OUTPUT TO THE CROCS

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4-SILICON

• CONTROLLED

GROUP 2 GROUP 2 RECTIFIER

GROUP? GROUPS ISCRI

RELAYS

SAFETY RODS REGULATING RODS F URIPAtTAGE

T~RI ATTCHMENT

Figure 3.4 Babcock 8 Wlleox Reactor Trip. System (Oconee. TMI, CR-3, ANO-1, Rancho Seco)

CR00 PLAN? PROCeS

NtRoL ROD OIVE CONiWOt sYSTEIM)

I TUMENT CHANNtEL f ISENSORS AN0

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BISTASLES. ETC., AND

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CONTACT

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4mVAC MVAC - I * I - l MAIN Bus SECONOART BUS

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BTRIP ATTACHMENT

Figure 3.5 Babcock EP Wilcox Reactor Trip System (Davis-Besse)

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I" TABLE 3.3-1

-t REACTOR PROTECTION SYSTEM INSTRUMENTATION

In SYSTEM INSTRUMENTATION

MINIMUM

TOTAL NO. CHANNELS CHANNELS APPLICABLE

FUNCTIONAL UNIT OF CHANNELS TO TRIP . OPERABLE MODES ACTION

1. Manual Reactor Trip 2 1 2 1, 2, and * 1

2. Nuclear Overpower 4 2 3 1, 2 2#

3. RCS Outlet Temperature--High 4 2 3 1, 2 3f

4. Nuclear Overpower Based on RCS

Flow and AXIAL POWER IMBALANCE 4 2(a)(b) 3 1, 2 22#

5. RCS Pressure--Low 4 2(a) 3 1, 2 3#

6. RCS Pressure--High 4 2 3 1, 2 3# (

7. Variable Low RCS Pressure 4 2(a) 3 1, 2 3#

8. Nuclear Overpower Based on Pump Monitor I'

4 2(a)(b) 3 1, 2 3# rn

9. Reactor Containment Pressure--High .E

4 2 3 1, 2 3#

10. Intermediate Range, Neutron Flux and Rate 2 0 2 1, 2, and * 4

11. Source Range, Neutron Flux and Rate A. Startup *2 2

0 2##, and * 5 B. Shutdown 2 0 1 3, 4 and 5 6 ('

12. Control Rod Drive Trip Breakers

13. Reactor Trip Module

2 per trip system

2 per trip

1 per trip system

1 per trip

2 per 1, 2, and *

trip system

2 per 1, 2, and *

7. 9 I

system system trip system 7

14. Shutdown Bypass RCS Pressure-High 4 2 3 1, 2 .8

15. SCR Relays 2 2 2 1, 2. and * 7 (Oconee Desinn)

.. 10 (Davis Besse Design)

TABLE 3.3-1 (Continued)

ACTION STATEMENTS (Continued)

ACTION 7 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. Within 1 hour: . S .

1. Place the inoperable channel in the tripped condition, or

2. Remove power supplied to the control rod trip device associated with the inoperable channel.

.

b. One additional channel may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing per Specification 4.3.1.1, and the inoperable channel above may be bypassed for up to

30 minutes in any 24-hour period when necessary to test the trip breaker associated with the logic of the channel being tested per Specification 4.3.1.1. The inoperable channel above shall not be bypassed to test the logic of a channel of the trip system associated with the inoperable channel.

ACTION 8 -

With less than the Minimum Number of Channels OPERABLE, declare the bypass inoperable and verify that all channels served by the bypass are OPERABLE, or satisfy the associated ACTION

requirements.

ACTION 9 - With one of the Reactor Trip Breaker diverse trin features (under- voltane or shunt trip attachment) inonerable, restore it to OPERABLE

status in 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or nlace the breaker in trin in the next hour.

ACTION 10 - With one or both channels of SCR Relays inoperable, restore the channels to OPERABLE status durinn the next COLD SHUTDOWN

exceeding 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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ELW-STS 3/4 3-5

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03 TABLE 4.3-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS

-- I

4n CHANNEL MODES FOR WHICH

CHANNEL CHANNEL FUNCTIONAL SURVEILANCE IS

FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED

1; -Manual Reactor Trip N.A. N.A. S/U(l) 1, 2, and *

2. Nuclear Overpower S D(2) and Q(6) M

H 1, 2

3. RCS Outlet Temprature--High S R M 1, 2

4. Nuclear Overpower Based on RCS

Flow and AXIAL POWER IMBALANCE S(4) M(3) and Q(6,7) M 1, 2

5. RCS Pressure--Low S R H 1, 2

6. RCS Pressure--High S R M 1, 2

.S 1, 2

7. Variable Low RCS Pressure S R M

8. Nuclear Overpower Based on Pump Monitor S R M I, 2

9. Reactor Containment Pressure--High S R M I, 2

10. Intermediate Range, Neutron Flux and Rate S R(rM S/U(I)(5) I, 7, alld *

11. Source Range, Neutron Flux and Rate S ftCfo) M and S/U(1)(5) 7. :,14, 'J,amVl

12. Control Rod Drive Trip Breaker N.A.

• f1.A. *Mand S/U(l)(10) I, 7, ,anl *

13. Reactor Trip Module N.A. N.A. I, 7, anm *

14. Shutdown Bypass RCS S* ':

S Ift(')) S/U(8) 1, 2 Pressure-High

15. SCR Relays H.A. Ni.A. M (Oconee Desinn) 1, 2, and *

R (Davis Besse Design)

I

TABLE 4.3-1 (Continued)

NOTATION

• - With any control rod drive trip breaker closed.

(1) - If not performed in previous 7 days.

(2) - Heat balance only, above 15% of RATED THERMAL POWER. Adjust channel if absolute difference is greater than or equal to (2) percent.

(3) - Compare incore to out-of-core measured AXIAL POWER IMBALANCE above

15% of RATED THERMAL POWER. Recalibrate if absolute difference is greater than or equal to (2) percent.

(4) - AXIAL POWER IMBALANCE and loop flow indications only.

(5) - Verify at least one decade overlap if not verified in previous

7 days.

(6) - Neutron detectors may be excluded from CHANNEL CALIBRATION.

(7) - Flow rate measurement sensors may be excluded from CHANNEL

CALIBRATION. However, each flow measurement sensor shall be calibrated at least once per 18 months.

(8) - Logic only, if not performed in previous 92 days.

(9) - The total bypass function shall be demonstrated OPERABLE during CHANNEL CALIBRATION testing of each channel affected by bypass operation.

(10) - The CHANNEL FUNCTIONAL TEST shall independentlv verify the OPERABILITY

of the undervoltane and shunt trip attachments of the Reactor Trip Breakers.

B&W- STS 3/4 3-8

I4 LIST OF RECENTLY ISSUED GENERIC LETTERS

GENERIC

LETTER NO. SUBJECT DATE

84-20 Scheduling Guidance for Licensee Submittals of Reloads that Involve Unreviewed Safety Questions 8/20/84

84-21 Long Term Low Power Operation in PWR's 10/16/84

84-22 Not used

84-23 Reactor Vessel Water Level Instrumentation in BWRs 10/26/84

84-24 Clarification of Compliance to 10 CFR 50.49 Environmental Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants 12/27/84

85-01 Fire Protection Policy Steering Committee Report 1/9/85

85-02 Staff Recommended Actions Stemming From NRC

Integrated Program for the Resolution of Unresolved Safety Issues Regarding Steam Generator Tube Integrity 4/15/85

85-03 Clarification of Equivalent Control Capacity 1/28/85 For Standby Liquid Control Systems

85-04 Operator Licensing Examinations 1/29/85

85-05 Inadvertent Boron Dilution Events 1/31/85

85-06 Quality Assurance Guidance for ATWS

Equipment that is not Safety-Related 4/16/85

85-07 Implementation of Integrated Schedules 5/02/85 for Plant Modifications

85-08 10 CFR 20.408 Termination Reports - Format 5/23/85

85-09 Technical Specifications for Generic Letter 83-28, Item 4.3 5/23/85

85-10 Technical Specifications for Generic Letter 83-28, Items 4.3 and 4.4 5/23/85

i'

-2- For plants which have not implemented the shunt trip modifications, proposed technical specifications should be submitted as soon as practical following staff review and approval of modified design. For operating license applicants, proposed technical specifications should include requirements which are responsive to the enclosed guidance.

This request for information was approved by the Office of Management and Budget under clearance number 3150-0011 which expires April 30, 1985.

Should you have any questions, s Ca 1eu7e u4 r A L AJ- 6e/0

Hugh L. Thompson, Director Division of Licensing Enclosure:

Sample Technical Specifications OSI s S : AB C:ICSB CSki AD/SA:DL

TAl l E her J annon FRosa Alahan DCrutchfield

03/.4/85 03/L /85 03/( /85 03//l/85 03/7 /85 03/ /85 D:DL

HThompson

03/ /85

- 2 -

For plants which have not implemented the shunt trip modifications, proposed technical specifications should be submitted as soon as practical following staff review and approval of modified design. For operating license applicants, proposed technical specifications should include requirements which are responsive to the enclosed guidance.

This request for information was approved by the Office of Management and Budget under clearance number 3150-0011 which expires April 30, 1985.

Should you have any questions, they should be directed to the NRC Project Manager for your facility.

Hugh L. Thompson, Jr., Director Division of Licensing Enclosure:

Sample Technical Specifications

• PREVIOUS CONCURRENCE SEE DATE

ORAB* SL:TSRG* SL:ORAB* C:ICSB* C:ORAB:DL*

TAlexion:cl EButcher JHannon FRosa GHolahan DCrt chfield

03/4/85 03/4/85 03/6/85 03/11/85 03/7/85 03/\ /85

.hD:DL

HThompson

03/ /85