ML20211A148: Difference between revisions

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k TENNESSEE VALLEY AUTHORITY CHATTANOOGA. TENNESSEE 374o1 g h '.
TENNESSEE VALLEY AUTHORITY CHATTANOOGA. TENNESSEE 374o1 g h '. 157B Lookout Place
157B Lookout Place SEP 301986
                    %%                    SEP 301986
- U.S. Nuclear Regulatory Conaission j
      - U.S. Nuclear Regulatory Conaission                                                 j Region II-Attn:   Dr. J. Nelson Grace, Regional Administrator 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323
Region II-Attn:
Dr. J. Nelson Grace, Regional Administrator 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323


==Dear Dr. Grace:==
==Dear Dr. Grace:==
OFFICE OF INSPECTION AND ENFORCEMENT BULLETIN 85 MOTOR-OPERATED VALVE COMMON MODE FAILURES DURING PLANT TRANSIENTS DUE TO IMPROPER SWITCH SETTINGS -
OFFICE OF INSPECTION AND ENFORCEMENT BULLETIN 85 MOTOR-OPERATED VALVE COMMON MODE FAILURES DURING PLANT TRANSIENTS DUE TO IMPROPER SWITCH SETTINGS -
BROWNS FERRY NUCLEAR PLANT By {{letter dated|date=May 13, 1986|text=letter dated May 13, 1986}}, an interim response to IE Bulletin 85-03 was provided for Browns Ferry. Enclosure 1 contains the additional information required to complete item a of the bulletin. The information is based on the generic methodology of General Electric Company NEDC 31322 which was submitted to NRC by letter to J. M. Taylor from T. A. Pickens of the BWR Owners's Group dated September 2, 1986.
BROWNS FERRY NUCLEAR PLANT By {{letter dated|date=May 13, 1986|text=letter dated May 13, 1986}}, an interim response to IE Bulletin 85-03 was provided for Browns Ferry. Enclosure 1 contains the additional information required to complete item a of the bulletin. The information is based on the generic methodology of General Electric Company NEDC 31322 which was submitted to NRC by letter to J. M. Taylor from T. A. Pickens of the BWR Owners's Group dated September 2, 1986. provides an updated list of commitment milestones for completion of the remaining items of the bulletin.
Enclosure 2 provides an updated list of commitment milestones for completion of the remaining items of the bulletin. If there are any questions, please get in touch with J. D. Wolcott at (205) 729-3604.
If there are any questions, please get in touch with J. D. Wolcott at (205) 729-3604.
To the best of my knowledge, I declare the statements contained herein are complete and true.
To the best of my knowledge, I declare the statements contained herein are complete and true.
Very truly yours, TENNESSEE VALLEY AUTHORITY R. G idley,   rector Nuclear Safety and Licensing Enclosures cc: See page 2 8610150002 860930 PDR ADOCK 05000259 G                 PDR                                                   p/
Very truly yours, TENNESSEE VALLEY AUTHORITY R. G idley, rector Nuclear Safety and Licensing Enclosures cc: See page 2 8610150002 860930 PDR ADOCK 05000259 G
f
PDR p/
_ _ ~ , . ,
f f,
f,
_ _ ~,.,


Dr. J. Nelson Grace                       SEP 30 986 cc (Enclosures):
. Dr. J. Nelson Grace SEP 30 986 cc (Enclosures):
Mr. James Taylor, Director Office of Inspection and Enforcement U.S. Nuclear Regulatory Connission Washington, D.C. 20555 U.S. Nuclear Regulatory Consission Document Control Desk Washington, D.C. 20555 Mr. G. G. Zech Director, TVA Projects U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 i
Mr. James Taylor, Director Office of Inspection and Enforcement U.S. Nuclear Regulatory Connission Washington, D.C. 20555 U.S. Nuclear Regulatory Consission Document Control Desk Washington, D.C.
20555 Mr. G. G. Zech Director, TVA Projects U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 i
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r ENCLOSURE 1 BROWNS FERRY NUCLEAR PLANT (BFN)
r ENCLOSURE 1 BROWNS FERRY NUCLEAR PLANT (BFN)
RESPOMSE TO IE Bulletin 85-03
RESPOMSE TO IE Bulletin 85-03
            . Item a
. Item a
            . Review and document the design basis for the operation of each valve. This       -
. Review and document the design basis for the operation of each valve. This documentation should include the maximum differential pressure expected during both opening and closing the valve for both normal and abnormal events to the extent that the valve operations and events are included in the existing, approved design basis, (i.e.,-the design basis documented in pertinent licensee submittals such as FSAR analyses and fully approved operating and emergency procedures, etc.).
documentation should include the maximum differential pressure expected during both opening and closing the valve for both normal and abnormal events to the extent that the valve operations and events are included in the existing, approved design basis, (i.e.,-the design basis documented in pertinent licensee submittals such as FSAR analyses and fully approved operating and emergency procedures, etc.). When determining the maximum differential pressure, those single equipment failures and inadvertent equipment operations (such as inadvertent valve closures or openings) that are within the plant design basis should be assumed.
When determining the maximum differential pressure, those single equipment failures and inadvertent equipment operations (such as inadvertent valve closures or openings) that are within the plant design basis should be assumed.
Additional Response to Item a The required review and documentation of the design basis for operation of each valve identified to be within the bulletin scope has been completed. The additional information required to complete table 1 of our May 13, 1986 response, documentation of the maximum expected differential pressures for each of the subject valves, is provided in table 1.
Additional Response to Item a The required review and documentation of the design basis for operation of each valve identified to be within the bulletin scope has been completed. The additional information required to complete table 1 of our May 13, 1986 response, documentation of the maximum expected differential pressures for each of the subject valves, is provided in table 1.
TVA participated with the Boiling Water Reactor Owners's G.*oup (BWROG) in producing NEDC-31322, "BWR Owners Group Report on the Operational Design Basis of Selected Safety-Related Motor Operated' Valves." The information provided in table 1 is worst case differential pressures for each safety-related valve function as determined by application of the NEDC-31322 methodology to BFN.
TVA participated with the Boiling Water Reactor Owners's G.*oup (BWROG) in producing NEDC-31322, "BWR Owners Group Report on the Operational Design Basis of Selected Safety-Related Motor Operated' Valves." The information provided in table 1 is worst case differential pressures for each safety-related valve function as determined by application of the NEDC-31322 methodology to BFN.
The worst case differential pressures in table 1 have been compared to the original valve purchase specifications with favorable results. In no case did the worst case differential pressures exceed the design specified value.
The worst case differential pressures in table 1 have been compared to the original valve purchase specifications with favorable results.
In no case did the worst case differential pressures exceed the design specified value.
S
S
            ' Response Update for Items b   c. d. and f The schedule for completion of the requirements of bulletin items b, e and d is provided in enclosure 2. Items 2 and 3 of the schedule are being extended from the original commitment dates of October 31, 1986 and January 7, 1987 respectively, due to a delay in obtaining vendor services.     (Note: Attachment 2 of our {{letter dated|date=May 13, 1986|text=May 13, 1986 letter}} contained a typographical error in the completion date for Item 2). The Browns Ferry Units will not be operating during the extended timeframe and thus, there is no adverse impact on nuclear safety.
' Response Update for Items b
: c. d. and f The schedule for completion of the requirements of bulletin items b, e and d is provided in enclosure 2.
Items 2 and 3 of the schedule are being extended from the original commitment dates of October 31, 1986 and January 7, 1987 respectively, due to a delay in obtaining vendor services.
(Note: Attachment 2 of our {{letter dated|date=May 13, 1986|text=May 13, 1986 letter}} contained a typographical error in the completion date for Item 2).
The Browns Ferry Units will not be operating during the extended timeframe and thus, there is no adverse impact on nuclear safety.
The final planned response is the written report required by item f of the bulletin, which is due 60 days af ter completion of the program.
The final planned response is the written report required by item f of the bulletin, which is due 60 days af ter completion of the program.
Y
Y


Table 1                                         .
Table 1 Max. Differential Valve Valve Nonnal Safety Function Pressure (Ib/in2d)
Valve Max. Differential Valve                               Nonnal       Safety Function             Pressure (Ib/in2d)
Unlaue No.
Unlaue No.         Description       Position     Description                 Openine Closine 71-2       RCIC stem           Open         Valve must close and           -
Description Position Description Openine Closine 71-2 RCIC stem Open Valve must close and 1105 isolation valve isolate containment on RCIC stem line break.
1105 isolation valve                   isolate containment on           -
71-3 RCIC stem Open Valve must close and 1105 isolation valve isolate containment on RCIC stem line break.
RCIC stem line break.
71-8 RCIC stem supply Closed Valve must open on RCIC 1105 valve initiation to supply reactor stem to the turbine.
71-3       RCIC stem           Open         Valve must close and           -
71-9 RCIC turbine Open Valve must spring close on stop valve RCIC isolation signal to stop the turbine.
1105 isolation valve                   isolate containment on RCIC stem line break.
71-17 RCIC suction valve Closed Valve must close if opened 39.4 to the suppression and containment isolation pool is required.
71-8       RCIC stem supply     Closed       Valve must open on RCIC     1105         -
71-18 RCIC suction valve Closed Valve must close if opened 39.4 to the suppression and containment isolation pool is required.
valve                             initiation to supply reactor stem to the turbine.
71-19 ROIC suction valve Open Valve must close when RCIC 37.5 to the condensate is aligned to take suction storage tank from the suppression pool.
71-9       RCIC turbine         Open         Valve must spring close on     *
71-25 RCIC cooling water Closed Valve must open on RCIC 1342.9 supply valve initiation to ensure an adequate cooling supply.
* stop valve                       RCIC isolation signal to stop the turbine.
71-34 RCIC mininun Closed Must both open and close to 1335.8 1469.7 recirculation flow provide adequate punp valve recirculation flow and adequate system flow to the vessel, respectively.
71-17       RCIC suction valve   Closed       Valve must close if opened     -
71-38 RCIC test return Closed Valve must close on an 1200**
39.4 to the suppression               and containment isolation pool                               is required.
to condensate initiation signal while in the test return mode to ensure adequate flow to the reactor vessel.
71-18       RCIC suction valve   Closed       Valve must close if opened     -
71-39 RCIC discharge Closed Valve must open on RCIC 1129.3 injection valve initiationtoinjectwater to the reactor vessel through the feedwater system.
39.4 to the suppression               and containment isolation pool                               is required.
i
71-19       ROIC suction valve   Open         Valve must close when RCIC     -
37.5 to the condensate                 is aligned to take suction storage tank                       from the suppression pool.
71-25       RCIC cooling water   Closed       Valve must open on RCIC       1342.9         -
supply valve                       initiation to ensure an adequate cooling supply.
71-34       RCIC mininun         Closed       Must both open and close to   1335.8     1469.7 recirculation flow               provide adequate punp valve                             recirculation flow and adequate system flow to the vessel, respectively.
71-38       RCIC test return     Closed       Valve must close on an         -
1200**
to condensate                     initiation signal while in the test return mode to ensure adequate flow to the reactor vessel.
71-39       RCIC discharge       Closed       Valve must open on RCIC       1129.3         -
injection valve                   initiationtoinjectwater to the reactor vessel through the feedwater system.
i a
* Valve is included for conpleteness; it has no motor-operated safety function.
* Valve is included for conpleteness; it has no motor-operated safety function.
                                  ** Based on actual operating experience in test return mode.
a
** Based on actual operating experience in test return mode.
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Y
Y
  -_ _ .~__._ -_._,_ _ _ _ _
-_ _.~__._ -_._,_ _ _ _ _


Table 1                                           -
Table 1 (Continued)
(Continued)
Max. Differential Valve Valve Nonnal Safety Function Pressure (Ib/in2d)
Max. Differential Valve           Valve             Nonnal       Safety Function             Pressure (Ib/in2d)
Unique No.
Unique No.     Description         Position     Description                 Openino Closino 73-2       HPCI steam           Open           Valve must close to isolate     -        1105 isolation valve                     containment in case of a HPCI stem line break.
Description Position Description Openino Closino 73-2 HPCI steam Open Valve must close to isolate 1105 isolation valve containment in case of a HPCI stem line break.
73-3       HPCI steam           Open           Valve must close to isolate     -        1105 isolation valve                     containment in case of a HPCI steam line break.
73-3 HPCI steam Open Valve must close to isolate 1105 isolation valve containment in case of a HPCI steam line break.
73-16     HPCI steam supply     Closed       Valve must open on HPCI       1105         -
73-16 HPCI steam supply Closed Valve must open on HPCI 1105 valve initiation to supply reactor steam to the turbine.
valve                               initiation to supply reactor steam to the turbine.
73-26 HPCI suction valve Closed Must open to allow the HPCI 50.3 40.6 to the suppression punp to take suction from pool the torus and must close if isolation of primary containment is required.
73-26     HPCI suction valve   Closed       Must open to allow the HPCI   50.3     40.6 to the suppression                 punp to take suction from pool                               the torus and must close if isolation of primary containment is required.
73-27 HPCI suction valve Closed Must open to allow the HPCI 50.3 41.5 to the suppression pwp to take suction from,f the torus and must close i pool isolation of primary containment is required.
73-27     HPCI suction valve   Closed       Must open to allow the HPCI   50.3     41.5 to the suppression                 pwp   to take pool                                the torus  andsuction from,f must close i isolation of primary containment is required.
73-30 HPCI minimum Closed Must both open and close to 1361.8 1439.1 recirculation flow provide adequate pupp valve recirculation flow and adequate cooling flow to reactor vessel respectively.
73-30     HPCI minimum         Closed       Must both open and close to   1361.8   1439.1 recirculation flow                 provide adequate pupp valve                               recirculation flow and adequate cooling flow to reactor vessel respectively.
73-35 HPCI test return Closed Must close on HPCI 1250**
73-35     HPCI test return     Closed       Must close on HPCI               -      1250**
to condensate initiation to ensure adequate flow to the reactor vessel.
to condensate initiation to ensure adequate flow to the reactor vessel.
73-40     HPCI suction valve   Open         valve must close when HPCI       -
73-40 HPCI suction valve Open valve must close when HPCI 95.2 to the condensate is aligned to take suction storage tank from the suppression pool.
95.2 to the condensate                   is aligned to take suction storage tank                       from the suppression pool.
73-44 HPCI discharge Closed Must open on !?CI initiation 1129.3 injectionvalve to inject water to the reactor vessel through the feedwater system.
73-44     HPCI discharge       Closed       Must open on !?CI initiation 1129.3         -
i 73-81 HPCI steam supply Open Must close to isolate 1105 l
injectionvalve                     to inject water to the reactor vessel through the feedwater system.
bypass valve primary containment on a HPCI Steam line break.
i             73-81     HPCI steam supply     Open         Must close to isolate           -
** Based on actual operating experience in test return mode.
1105 bypass valve                       primary containment on a l                                                            HPCI Steam line break.
              ** Based on actual operating experience in test return mode.
P-
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          .                                                                                                                                            t
t j
    .  .                                                                                                                                              j ENCLOSURE 2 BROWNS FERRY NUCLEAR PLANT (BFN)
ENCLOSURE 2 BROWNS FERRY NUCLEAR PLANT (BFN)
REVISED COMMITMENT LISTING FOR IE BULLETIN 85-03
REVISED COMMITMENT LISTING FOR IE BULLETIN 85-03 1.
: 1. Provide the remaining information for item a as requested in the bulletin by October 1, 1986.           (Complete)
Provide the remaining information for item a as requested in the bulletin by October 1, 1986.
: 2. BFN will review and establish the correct valve switch settings by February 15, 1987.
(Complete) 2.
: 3. BFN will implement any required switch settings by April 15, 1987.
BFN will review and establish the correct valve switch settings by February 15, 1987.
: 4. BFN will perform differential pressure testing to verify the established switch settings are correct or provide justification of the acceptability of those which cannot be tested by June 4,1987.
3.
: 5. BFN will implement any required procedures and/or procedure revisions necessary to ensure the correct switch setting can be maintained throughout the life of the plant by September 3,1987.
BFN will implement any required switch settings by April 15, 1987.
: 6. Provide a written report in accordance with item f of the bulletin 60 days after completion of the program, i
4.
BFN will perform differential pressure testing to verify the established switch settings are correct or provide justification of the acceptability of those which cannot be tested by June 4,1987.
5.
BFN will implement any required procedures and/or procedure revisions necessary to ensure the correct switch setting can be maintained throughout the life of the plant by September 3,1987.
6.
Provide a written report in accordance with item f of the bulletin 60 days after completion of the program, i
0293c s
0293c s
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          ., . - . . .        . - ~ _ _ - . - -         - . - - - _ , . . - . . . _ _ . - . _ . , _ _ . .    - - . . _ . , . -.. - . _ . . . - . - ~}}
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Latest revision as of 00:40, 6 December 2024

Forwards Addl Info in Response to IE Bulletin 85-003 Re motor-operated Valve Common Mode Failures.Required Review & Documentation of Design Basis for Operation of Each Valve within Scope of Bulletin Completed
ML20211A148
Person / Time
Site: Browns Ferry  Tennessee Valley Authority icon.png
Issue date: 09/30/1986
From: Gridley R
TENNESSEE VALLEY AUTHORITY
To: Grace J
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
References
IEB-85-003, IEB-85-3, NUDOCS 8610150002
Download: ML20211A148 (6)
v • d • e


Text

-_

k TENNESSEE VALLEY AUTHORITY CHATTANOOGA. TENNESSEE 374o1 g h '.

157B Lookout Place SEP 301986

- U.S. Nuclear Regulatory Conaission j

Region II-Attn:

Dr. J. Nelson Grace, Regional Administrator 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323

Dear Dr. Grace:

OFFICE OF INSPECTION AND ENFORCEMENT BULLETIN 85 MOTOR-OPERATED VALVE COMMON MODE FAILURES DURING PLANT TRANSIENTS DUE TO IMPROPER SWITCH SETTINGS -

BROWNS FERRY NUCLEAR PLANT By letter dated May 13, 1986, an interim response to IE Bulletin 85-03 was provided for Browns Ferry. Enclosure 1 contains the additional information required to complete item a of the bulletin. The information is based on the generic methodology of General Electric Company NEDC 31322 which was submitted to NRC by letter to J. M. Taylor from T. A. Pickens of the BWR Owners's Group dated September 2, 1986. provides an updated list of commitment milestones for completion of the remaining items of the bulletin.

If there are any questions, please get in touch with J. D. Wolcott at (205) 729-3604.

To the best of my knowledge, I declare the statements contained herein are complete and true.

Very truly yours, TENNESSEE VALLEY AUTHORITY R. G idley, rector Nuclear Safety and Licensing Enclosures cc: See page 2 8610150002 860930 PDR ADOCK 05000259 G

PDR p/

f f,

_ _ ~,.,

. Dr. J. Nelson Grace SEP 30 986 cc (Enclosures):

Mr. James Taylor, Director Office of Inspection and Enforcement U.S. Nuclear Regulatory Connission Washington, D.C. 20555 U.S. Nuclear Regulatory Consission Document Control Desk Washington, D.C.

20555 Mr. G. G. Zech Director, TVA Projects U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 i

I i

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_/

r ENCLOSURE 1 BROWNS FERRY NUCLEAR PLANT (BFN)

RESPOMSE TO IE Bulletin 85-03

. Item a

. Review and document the design basis for the operation of each valve. This documentation should include the maximum differential pressure expected during both opening and closing the valve for both normal and abnormal events to the extent that the valve operations and events are included in the existing, approved design basis, (i.e.,-the design basis documented in pertinent licensee submittals such as FSAR analyses and fully approved operating and emergency procedures, etc.).

When determining the maximum differential pressure, those single equipment failures and inadvertent equipment operations (such as inadvertent valve closures or openings) that are within the plant design basis should be assumed.

Additional Response to Item a The required review and documentation of the design basis for operation of each valve identified to be within the bulletin scope has been completed. The additional information required to complete table 1 of our May 13, 1986 response, documentation of the maximum expected differential pressures for each of the subject valves, is provided in table 1.

TVA participated with the Boiling Water Reactor Owners's G.*oup (BWROG) in producing NEDC-31322, "BWR Owners Group Report on the Operational Design Basis of Selected Safety-Related Motor Operated' Valves." The information provided in table 1 is worst case differential pressures for each safety-related valve function as determined by application of the NEDC-31322 methodology to BFN.

The worst case differential pressures in table 1 have been compared to the original valve purchase specifications with favorable results.

In no case did the worst case differential pressures exceed the design specified value.

S

' Response Update for Items b

c. d. and f The schedule for completion of the requirements of bulletin items b, e and d is provided in enclosure 2.

Items 2 and 3 of the schedule are being extended from the original commitment dates of October 31, 1986 and January 7, 1987 respectively, due to a delay in obtaining vendor services.

(Note: Attachment 2 of our May 13, 1986 letter contained a typographical error in the completion date for Item 2).

The Browns Ferry Units will not be operating during the extended timeframe and thus, there is no adverse impact on nuclear safety.

The final planned response is the written report required by item f of the bulletin, which is due 60 days af ter completion of the program.

Y

Table 1 Max. Differential Valve Valve Nonnal Safety Function Pressure (Ib/in2d)

Unlaue No.

Description Position Description Openine Closine 71-2 RCIC stem Open Valve must close and 1105 isolation valve isolate containment on RCIC stem line break.

71-3 RCIC stem Open Valve must close and 1105 isolation valve isolate containment on RCIC stem line break.

71-8 RCIC stem supply Closed Valve must open on RCIC 1105 valve initiation to supply reactor stem to the turbine.

71-9 RCIC turbine Open Valve must spring close on stop valve RCIC isolation signal to stop the turbine.

71-17 RCIC suction valve Closed Valve must close if opened 39.4 to the suppression and containment isolation pool is required.

71-18 RCIC suction valve Closed Valve must close if opened 39.4 to the suppression and containment isolation pool is required.

71-19 ROIC suction valve Open Valve must close when RCIC 37.5 to the condensate is aligned to take suction storage tank from the suppression pool.

71-25 RCIC cooling water Closed Valve must open on RCIC 1342.9 supply valve initiation to ensure an adequate cooling supply.

71-34 RCIC mininun Closed Must both open and close to 1335.8 1469.7 recirculation flow provide adequate punp valve recirculation flow and adequate system flow to the vessel, respectively.

71-38 RCIC test return Closed Valve must close on an 1200**

to condensate initiation signal while in the test return mode to ensure adequate flow to the reactor vessel.

71-39 RCIC discharge Closed Valve must open on RCIC 1129.3 injection valve initiationtoinjectwater to the reactor vessel through the feedwater system.

i

  • Valve is included for conpleteness; it has no motor-operated safety function.

a

    • Based on actual operating experience in test return mode.

4 i

Y

-_ _.~__._ -_._,_ _ _ _ _

Table 1 (Continued)

Max. Differential Valve Valve Nonnal Safety Function Pressure (Ib/in2d)

Unique No.

Description Position Description Openino Closino 73-2 HPCI steam Open Valve must close to isolate 1105 isolation valve containment in case of a HPCI stem line break.

73-3 HPCI steam Open Valve must close to isolate 1105 isolation valve containment in case of a HPCI steam line break.

73-16 HPCI steam supply Closed Valve must open on HPCI 1105 valve initiation to supply reactor steam to the turbine.

73-26 HPCI suction valve Closed Must open to allow the HPCI 50.3 40.6 to the suppression punp to take suction from pool the torus and must close if isolation of primary containment is required.

73-27 HPCI suction valve Closed Must open to allow the HPCI 50.3 41.5 to the suppression pwp to take suction from,f the torus and must close i pool isolation of primary containment is required.

73-30 HPCI minimum Closed Must both open and close to 1361.8 1439.1 recirculation flow provide adequate pupp valve recirculation flow and adequate cooling flow to reactor vessel respectively.

73-35 HPCI test return Closed Must close on HPCI 1250**

to condensate initiation to ensure adequate flow to the reactor vessel.

73-40 HPCI suction valve Open valve must close when HPCI 95.2 to the condensate is aligned to take suction storage tank from the suppression pool.

73-44 HPCI discharge Closed Must open on !?CI initiation 1129.3 injectionvalve to inject water to the reactor vessel through the feedwater system.

i 73-81 HPCI steam supply Open Must close to isolate 1105 l

bypass valve primary containment on a HPCI Steam line break.

    • Based on actual operating experience in test return mode.

P-

t j

ENCLOSURE 2 BROWNS FERRY NUCLEAR PLANT (BFN)

REVISED COMMITMENT LISTING FOR IE BULLETIN 85-03 1.

Provide the remaining information for item a as requested in the bulletin by October 1, 1986.

(Complete) 2.

BFN will review and establish the correct valve switch settings by February 15, 1987.

3.

BFN will implement any required switch settings by April 15, 1987.

4.

BFN will perform differential pressure testing to verify the established switch settings are correct or provide justification of the acceptability of those which cannot be tested by June 4,1987.

5.

BFN will implement any required procedures and/or procedure revisions necessary to ensure the correct switch setting can be maintained throughout the life of the plant by September 3,1987.

6.

Provide a written report in accordance with item f of the bulletin 60 days after completion of the program, i

0293c s

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