ML20247L242
| ML20247L242 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 07/19/1989 |
| From: | Conlon T, Hunt M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20247L168 | List: |
| References | |
| 50-324-89-11, 50-325-89-11, IEIN-85-009, IEIN-85-9, NUDOCS 8908010190 | |
| Download: ML20247L242 (33) | |
See also: IR 05000324/1989011
Text
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UNITED STATES
~ '
[jcpcro ug
. .. ,,og NUCLEAR REGULATORY COMMISslON
.[ n REGIOid il
g , g: 101 MARIETTA STREET,N.W.
'8' e AT LANTA. GEORGI A 30323
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- Report Hos.: 50-325/89-11.and 50-324/89-11
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Licensee: : Carolina Power and Light Company
P. O. Box 1551
Raleigh, NC' 27602
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. Docket Nos.: 50-325 and 50-324 License Nos.: DPR-71 and DPR-62~
' c. ' , . Facility Name: Bruntwick I and 2
,
11 Inspection Conducted: May 22-26 (onsite); May 29 - June 9, (NRC
Region'II Office) and June 20-22,1989(onsite)
Inspector:
M. D. Hunt
N4 Mb ,
///I!b
/ fate Signed ,
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' Accompanying Personnel: J.~R. Harris
W. Levis
P. A. Taylor '
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A. B. Ruff
D. C. Wa rd
G. R. Wiseman
Approved b pfM ~7 /9 C9
ffate Signe'd
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T. E. Conlon, Chief ,
Plant Systems Section i
Engineering Branch 1
Division of Reactor Safety ;
SUMMARY
Scope:
This special announced inspection was conducted in two site visits. During the
initial visit emphasis was placed on examination of the licensee's methods of
meeting the requirements of 10 CFR 50, Appendix R, Sections III.G. III.J. and ;
III.L., for safe shutdown capabilities, associated circuits of concern and
alternative shutdown capabilities. A review of information for the closure of
! a fire protection related unresolved item was conducted. A second visit was
made to review the licensee's proposed corrective action to an identified
violation.
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89080101h
PDR ADC fM hPDC 24 J
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Results:
During this inspection, the NRC inspectors reviewed the results of a corporate
QA audit which identified three areas in Unit 2 that were not in compliance {
with Appendix R. The licensee performed an engineering evaluation which {
outlined the corrective action for each finding and the compensatory actions 4
required until Unit 2 can be brought into compliance. The unit was thought to {
be in compliance as required on April 1988. These findings were determined to j
be a Licensee Identified Violation [ Paragraph 2]. During inspection of Unit 1,
the NRC inspectors found that an area in the Southwest Corner of the Reactor
Building did not meet the separation requirements of Appendix R for redundant
shutdown trains of equipment. The licensee promptly placed a firewatch and '
removed stored combustibles from the area. During the second visit to the j
site, the licensee agreed to take corrective actions that would bring the area !
into compliance with Appendix R. These findings were identified as a violation {
and are further discussed in paragraph 2. An overall assessment of the j
licensee's performance, based on findings during this inspection, revealed both j
strengths and weaknesses as summarized below:
Strengths
The licensee's QA audit identified the several items that were not in
compliance with Appendix R on Unit 2.
The licensee took prompt compensatory actions upon determining that Unit 1
was not in compliance.
Once the licensee was convinced that the SW corner of Unit I reactor
building was not in compliance, they committed to take the necessary
actions to bring the area into full compliance.
Weaknesses
Initially the licensee failed to recognize the significance of the
discrepancies found in the Unit I reactor building SW corner.
The licensee took credit for the ADS /LPCI mode of operation for backup to
the loss of the HPCI/RICI systems but did not have an analysis to provide
assurance that three safety relief valves would provide adequate reactor
pressure control to protect the core.
The licensee waited a considerable length of time to conduct an audit to
determine if 'Jnit 2 was in compliance. This type audit should have been
completed tr.M earlier to provide assurance that all modifications were
complete and etisfactory.
In various cases, it was acted that the as-built conditions did not match
drawings as noted in the Unit I reactor building SW corner, and on some
electrical drawings minor differences were noted relating to fuse sizing.
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~ Maintenance activities were not coordinated with the operating procedures
. to insure that shutdown. activities would not be impacted, as noted by the
sealing of the reactor building access doors at elevation 50'.
The number.of operating procedures employed in the shutdown modes appeared
to be rather large, that is 32 for each unit. This could have been
. brought about by the fact that the plant is divided .into 34 Unit 1 and
Unit 2-fire areas.
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REPORT DETAILS
1. Persons Contacted
Licensee Employees
D. Barnett, Senior Specialist, Electrical
- C. Blackman, Manager, Operations
J. M. Brown, Engineering
- T. Caldwell, Appendix R Coordinator
.
- S. Callis, Licensing
- W. Caraway, Nuclear Engineering Departments
- A. Cheatham, Manager, Environmental and Radiological Control
- W. J. Dorman, Supervisor, Quality Assurance
- K. Elliot, Nuclear Safety
- K. E. Enzor, Director, Regulatory Compliance
- L. Fincher, Nuclear Engineering Department
- S. Hardy, Nuclear Engineering Department
- J. Harness, General Manager
- J. E. Harrell, Director, Outages i
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H. !iarrelson, Operations
- T. Harris, Regulatory Compliance
- W. R. Hatcher, Security Supervisor
- A. S. Hegler, Operations
- R. Helme, Technical Support
B. Hinds,' Project Engineer, Nuclear Engineering Department
- M. A. Jones, On-site-Nuclear Safety
A. Lane, Senior Engineer, Electrical
- D. Lichty. Operations
- T. Mull, Operations
J. O' Conner, Technical Support
- R. Oates, Licensing
- B. Parks, Technical Support
- R. M. Poulk, Project Specialist
L. Rothman, Nuclear Engineering Department
- J. Royal, Nuclear Engineering Department
- W. Simpson, Site Planning and Control
- S. Tabor, Technical Support
- R. L. Warden, Maintenance Manager .
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iD. Warrne, Technical Support
4. Wyllie, Technical Support
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Other licensee employees contacted during this inspection included
engineers, operators, security force members, and administrative l
personnel. l
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Other Organizations
Bob Sergy, Impel
NRC Resident Inspectors
- W. H. Ruland, Senior Resident Inspector
- W._ Levis, Resident Inspector
- Attended exit interview on May 26, 1989
- Attended exit interview on June 22, 1989
Acronyms and initialisms used throughout this report are listed in the
last paragraph.
2. Compliance with 10 CFR 50, Appendix R, Sections III.G. and III.L. (Module
64100)-
An inspection was conducted to determine if the fire protection features,
provided for structures, systems and components important to safe shutdown
at Brunswick Unit I and Unit 2, were in compliance with' 10 CFR 50,
Appendix R, Sections III.G. and III.L. The scope of'this inspection was
to determine if the fire protection features provided for reactivity
control, reactor coolant makeup, reactor pressure control, reactor heat
reuoval, process monitoring function and safe shutdown system support
' functions were capable of limiting potential fire damage so that one train
of safe shutdown systems essential for achieving and maintaining plant
shutdown from either the control room or remote shutdown stations is free
of fire damage.
a. Safe Shutdown Capabilities
The. licensee performed an analysis to determine the shutdown trains
which will be used to shut down the plant in the event of a fire in
any plant area. This postfire safe shutdown analysis is described in
the licensee's ASCA. This document was submitted to the NRC and the
postfire safe shutdown capability described within the ASCA was
approved by the NRC by letter dated December 30, 1986. The systems
required to meet each of the performance goals outlined in 10 CFR 50,
Appendix R, are established in the ASCA. These systems are:
Perf ormance Goal Systems
Reactivity Control None (deenergize scram solenoid
valves)
Reactor Coolant Makeup Train A: HPCI
Train B: RCIC
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Reactor Pressure Control Train A: HPCI and One (1) ADS SRV
Train B: RCIC and Three (3) ADS
SRV.
Reactor Heat Removal Train A: RHR in TC and SDC mode
Train B: RHR aligned to TC mode
and SDC mode
Processing Monitoring Train A: Instruments for reactor
Instrumentation pressure, reactor level,
suppression pool level,
'and suppression pool
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Train B: Instruments for reactor
pressure, reactor level,
suppression pool level,
and suppression pool
temperature
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Safe Shutdown System
Support Functions
Emergency Power System Train A: Emergency AC and DC
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(EPS) Power Distribution
Power Distribution
Diesel Generator Cooling Train A: SW
Water (DGCW) Train B: SW
Train B: SW
The components necessary for postfire safe shutdown associated with
each system above were identified and plant fire areas and fire zones
were defined to provide separation of Train A and Train B shutdown
systems. For each plant area postfire safe shutdown, Train A or B
has been verified available. In areas where separation could not be
achieved between Train A and Train B as required by 10 CFR 50,
Appendix R,Section III.G., the licensee has provided an alternate
shutdown capability independent of the area in accordance with
Section III.L. of Appendix R, or an exemption from the requirements
of Section III.G. was granted by the NRC.
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Brunswick is divided into 34 Unit I and Unit 2 fire areas. Many of
the fire areas are common to both units. Table 6.2.1. of the ASCA
describes the alternate shutdown method for each fire area. Of the
34 fire areas, 28 require the licensee to take manual actions to
mitigate the consequences of a fire and/or require alternate shutdown
operations from designated alternate control stations remote from the
control room.
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The licensee ~ has developed 32 ASSD Procedures for each unit to
implement the -postfire safe shutdown capability established in the
ASCA. ASSD-01 for each unit includes a table which establishes the-
safe shutdown train to be used for a fire in each ASSD Fire Area.
ASSD-2 for each unit is the procedure developed for a control
building fire which could result in a control room evacuation and ;
alternative shutdown using Train B. The remaining ASSD procedures i
would not require a complete control room evacuation but could result
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in the necessity to take actions at remote alternative centrol
stations.
In order to ensure safe shutdown capabilities, where cables. or
equipment of redundant trains of systems necessary to achieve and
maintain hot shutdown conditions are located within the same fire
area outside the primary containment, 10 CFR 50, Appendix R,
Section III.G.2 requires that one train of hot shutdown systems be
maintained free of fire damage oy providing fire protection features
which meet the requirements of either !II.G.2.a., III.G 2.b. , or
.III.G.2.c.
On the brsis of the above Appendix R criteria, the inspectors
reviewed the separation provided for a. sample of safe shutdown
equipment cables required to implement the performance goals
described above.
(1) Fire Protection for Safe Shutdown Systems / Components
An inspection was made to determine if redundant cabling for the
Units 1 and 2 safe shutdown system,' required to achieve and
maintain hot and cold shutdown conditions have been provided
with adequate separation or protected in accordance with
Appendix R,Section III.G.2.
Included in the review was an evaluation of the fire protection
features (fire barriers, raceway barriers, cable coatings,
spacial separation, fire detection and fire suppression)
installed to comply with Appendix R,Section III.G.2. and NRC
approved exemptions.
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The following cabling and compcnents were reviewed.
(a) Reactor Coolant Makeup
Train A - HPCI (Unit.1)
Equipment Cables Cable Fire Zone Location
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1-E41-F0.01 JE7-VF2 RB1-01-H, 0, G, N; RB1-02
(steam inlet) JE7-X49 RB1-01-G, N; RB1-04
1-E41-F004 B14-VE8/1 RB1-01-G, 0, N; RB1-04 'l
(suction CST) JE7-VF1 RB1-01-H, 0, G, N; RB1-02 {
1-E41-F006 B17-KG6/3 RB1-01-G
(discharge B17-KG6/4 RB1-01-G
to vessel) B17-514/1
1-E41-F042 B22-VF2 RB1-01-G, 0, N; RB1-02
(suction SP) G22-VF2/1 RB1-01-G, 0, N; RB1-02
JE7-VF1 RB1-01-H, 0, G, N; RB1-02
Train A - HPCI (Unit 2)
Equipment Cables Cable Fire Area Location
(valves)
2-E41-F001 B21-VE9/1 RB2-01-G, N, 0; RB2-02
(steam inlet) 821-VE9/2 RB2-01-G, N, 0; RB2-02
B21-VE9/3 RB2-01-G, N; RB2-02
B21-JF1 RB2-01-G
2-E41-F004 B14-VE8/1 RB2-01-G
(suction CST) G14-JFI RB2-01-G
2-E41-F006 KG6-S1H RB2-08
(discharge KG6-S1H/1 RB2-08
to vessel) KG6-S1H/2 RB2-08
B17-S1H RB2-01-G
B17-S1H/1 RB2-01-G
2-E41-F042 B22-VF2 RB2-01-G, N, 0; RB2-02
(suctionSP) B22-VF2/1 RB2-01-G, N, 0; RB2-02
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Train B - RCIC (Unit 1)
Equipment Cables ' Cable Fire Area LocationL
(valves)
1-E51-F007 DS4-QB3/1 RB1-01-G, 0
(inboard DS4-QB3/3 RB1-01-G, 0
isolation DS4-QB1/4 RB1-01-G, 0
to vessel)
1-E51-F0101 B38-VE1 RB1-01-D, G.
(suction CST)
1-E51-F013 AQ7-IG7 RB1-01-H; RB1-10
(discharge
to vessel) B41-JH1 RB1-01-G
B41-KS5 RB1-01-G
B41-KS5/1 RB1-01-G
1-E51-F029 B46-VE3 RB1-01-D, G
(suctionSP) B46-VE3/1 RB1-01-D, G
1-E51-F045 B44-JH1 RB1-01-G
(steam inlet) 844-RS4 RBI-01-G
B44-VE7 RB1-01-D, G, N
B44-VE7/1 RB1-01-D, G, N
IG7-RS4/3 RB1-01-H, G
I4H-JH1 RB1-01-D, G, H
Train B - RCIC (Unit 2)
Equipment Cables Cable Fire Area Location
(valves)
2-E51-F010 B38-VE1 RB2-01-D, G
(suction CST) 838-VEI/1 RB2-01-D, G
JH1-VE3 RB2-01-D, G; RB2-04
2-E51-F013 B41-KS5 RB2-01-G
(discharge B41-KS5/1 RB2-01-G
to vessel)
2-E51-F029 B46-VE3 RB2-01-D, G 4
RB2-01-D, G !
(suction SP) B46-VE3/1
JH1-VE3 RB2-01-D, G; RB2-04
2-E51-F045 B44-VE7 RB2-01 D, G
(steam inlet) B44-VE7/1 RB2-0'.-D, G
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(b) Reactor Pressure Control
The separation of the HPCI (Train A) and RCIC (Train B)
components necessary to control reactor pressure described
in paragraph 2.a.(1)(a) were reviewed. In addition, the
.following ADS Safety / Relief valve separation was reviewed.
Train' A ADS Safety / Relief (Unit 1)
Equipment Cable Cable Fire Area Location
1-B21-F013F JG8-QC3* RB1-01-G
Train A ADS Safety / Relief (Unit 2)
Equipment Cable Cable Fire Area Location
2-821-F013F JG8-QC3* RB2-01-G; RB2-06
Train B ADS Safety / Relief (Unit 1)
Equipment Cables Cable Fire Area Location
1-821-F013B 'RS4-WI3 RB1-01-G
RS4-XN0 RB1-01-D, G
QC1-WI3/1 RB1-01-G, 0
1-B21-F013E QCO-WI3 RB1-01-G, 0; RB1-03-A
1-821-F013G QC8-WI3/1 RB1-01-G
Train B ADS Safety / Relief (Unit 2)
Equipment Cables Cable Fire Area Location
2-B21-F013G QCI-WI3/1 RB2-01-G, 0; RB2-03-A
RS4-WI3 RB2-01-G
2-B21-F013E QCO-WI3 RB2-01-G, 0; RB2-03-A
2-B21-F013G QC8-Wi3/1 RB2-01-G
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(c) Reactor Heat-Removal
Train A RHR Aligned to TC Mode (Unit 1) ,
Equipment Cables Cable Fire Area Location
1-E11-C002A- JF1-XN1/4 RB1-01-G, N; RB1-04
(Pump 1A) NC6-YE7 RB1-01-C
1-E11-C002C NC8-YE8 RB1-01-C
(PumpIC)
1-E11-F020A CQ8-DG8 RB1-01-C, D, G N, 0
(suctionSP) CQ8-VA6 RB1-01-C
CQ8-VA8 RB1-01-C
DG0-DG8 RB1-01-D, G-
DG8-JF1 RB1-01-D, G; RB1-04
DG8-VD2 RB1-01-C, D, G. N, 0
DG8-VD2/1' RB1- 01-C, D, G, N, 0
1-E11-F028A CQ8-DG0 RB1-01-C, G, N, 0
(containment CQ8-VA6/1 RB1-01-C
spray) CQ8-VA8/1 RB1-01-C
DGO-DI2 RB1-01-G
DG0-JF1 RB1-01-G; RB1-04
DGO-VC1 RB1-01-C, G N, 0
DG0-VC1/1 RBI-01-C, G, N, O
JE6-VC3 RB1-01-C, G, H, N, 0
Train A RHR Aligned to TC Mode (Unit 2)
Equipment Cables Cable Fire Area Location
2-E11-C002A NC6-YE7 RB2-01-C
(Pump 1A)
2-E11-C002C NC8-YE8 RB2-01-C
(Pump 10)
2-E11-F020A CQ8-DG8 RB2-01-C, G, N, 0
(suctionSP) CQ8-VA6 RB2-01-C
CQ8-VA8 RB2-01-C
DGO-DG8 RB2-01-G
DG8-JF1 RB2-01-G
DG8-VD2 RB2-01-C, G, N, 0
DG8-VD2/1 RB2-01-C, G, N, 0
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'2-E11-F028A CQ8-DG0. RB2-01-C, G. N, 0
-(containment .CQ8-VA6/1 RB2-01-C
spray) CQ8-VA8/1 RB2-0.-C
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DGO-JF1 RB2-01-G
DGO-VC1 RB2-01-C, G. N, 0
DG0-VC1/1 RB2-01-C, G. N, O
JE6-VC3 RB2-01-C, G. N, 0
Train B RHR Aligned to TC Mode (Unit 1)
Equipment Cables Cable Fire Area Location;
1-E11-C002B B28-D00 .RB1-01-G, 0
(Pump 18) B28-D11 RB1-01-G
D11-XN0/1 ' RB1 01-D, G
NC7-YD7 RB1-01-D
1-E11-C002D D11-XN0 RB1-01-D, G'
(PumpID) NC9-YD9 RB1-01-D
1-E11-F020B DMS-DN6 RB1-01-G'
(suction SP) DMS-VC2/3 RB1-01-D, G
DN6-DQ2 RB1-01-D, G
DN6-VD3 RB1-01-D, G
DN6-VD3/1 RB1-01-D, G
DN6-VD3/3 RB1-01-D, G
DQ2-VA7 RB1-01-D
DQ2-VA9 RB1-01-D
1-E11-F0038 DK8-VA5 RB1-01-D, G ,
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(HXoutlet) DK8-VA5/1 RB1-01-D, G
TrainBRHRAlignedtoTCMode(Unit 2J
E_quipment Cables Cable Fire Area Location
2-E11-C002B NC7-YD7 RB2-01-D
(PumpIB)
2-E11-C002D NC9-YD9 RB2-01-D
(Pump ID)
2-E11-F020B DMS-ON6 RB2-01-G
(suctiv. SP) DMS-VC2/3 RB2-01-G
DN6-DQ2 RB2-01-0, G
DN6-JF8 RB2-01-G: RB2-04
DN6-VD3 RB2-01-D, G
DN6-VD3/1 RB2-01-D, G
DN6-VD3/3 RB2-01-D, G
DQ2-VA7 RB2-01-D
DQ2-VA9 RB2-01-D
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2-E11-F0038 DK8-VA5' .
RB2-01-D G
(HX outlet) DK8-VA5/1 RB2-01-D, G
Train A RHR Aligned to SDC Mode (Unit 1)
Equipment Cables Cable Fire Area Location
1-E11-F006A DE9-J F1 RB1-01-G, RB1-04
(SDC suction) DE9-VA6/1 RB1-01-C, G, N, 0
'l-E11-F047A DG1-JF1 RB1 .-G; RB1-04
(HX inlet) DGI-VB9/1 RB1-91-C, G. N, 0
1-E11-F008 G50-JH2 RB1-01-G
(RHR outboard
isolation)
Train A RHR Aligned to SDC Mode (Unit 2)
Equipment Cables Cable Fire Area Location
2-E11-F006A DE9-JF1 RB2-01-G
(SDCsuction) DE9-VA6/1 RB2-01-C, G. N, 0
2-E11-F006C DFO-JF1 RB2-01-G
(SDC suction) DF0-VA8/1 RB2-01-C, G, N, 0
2-E11-F008 B50-JH2 RB2-01-G; RB2-04
(RHR outboard
isolation)
2-E11-F047A DGI-JF1 RB2-01-G
(HX inlet) DG1-VB9/1 RB2-01-C, G, N, 0
Train B RHR Aligned to SDC Mode (Unit 1)
Equipment Cables Cable Fire Aren Location
l 1-E11-F006B DL1-VA7 RB1-01-D, G
(SDC suction) DL1-VA7/1 RB1-01-D, G
DMS-DN6 RB1-01-G
DM5-VC2/3 RB1-01-D, G
DN6-902 RB1-01-D, G
DN6-JF8 RB1-01-G
DN6-VD3 RB1-01-D, G
DN6-VD3/1 RB1-01-D, G
DN6-VD3/3 RB1-01-D, G
DQ2-VA7 RB1-01-0
DQ2-VA9 RB1-01-D
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1-E11-F006D . DL2-VA9 RB1-01-D, G
1-E11-F008 B26-L6C RB1-01-0; RB1-06
(RHR outboard B26-L6C/1 RB1-01-G; RB1-06
isolation) 'B50-JH2 RB1-01-G
B50-KM3/2 RB1-01-G, 0; RB1-06
B50-L6C 'RB1-01-G
KM3-L6C RB1-01-G; RB1-06
KM3-L6C/1 RB1-01-G; RB1-06
Train B RHR Aligned to SDC Mode (Unit 2)
Equipment Cables Cable Fire Area Location
2-E11-F006B DL1-VA7 RB2-01-D, G
(SDC sucticn) DL1-VA7/1 RB2-01-D, G
OMS-DNG RB2-01-G
DMb-VC2/3 RB2-01-D, G
DN6-DQ2 RB2-01-D, G
DN6-JF8 RB2-01-G; RB2-04
DN6-VD3 RB2-01-0, G
DN6-VD3/1 RB2-01-D, G
DN6-VD3/3 RB2-01-D, G
DQ2-VA7 RB2-01-0
DQ2-VA9 RB2-01-D
2-E11-F006D DL2-VA9 RB2-01-D, G
(SDC suction) DL2-VA9/1 RB2-01-D, G
2-E11-F008 B26-LIF RB2-01-G
(RHR outboard B26-LIF/1 RB2-01-G
isolation) B2A-LIF RB2-01-G
l B50-JH2 RB2-01-G; RB2-04
L B50-KM3/2 RB2-01-G; RB2-06
l
B50-L6C RB2-01-G, H
B50-L6C/1 RB2-01-G, H
KM3-LIF RB2-01-G, 0; RB2-06
KM3-L6C RB2-01-G; RB2-06
KM3-L6C/1 RB2-01-G; RB2-06
LIF-L6C RB2-01-G, O
LIF-L6C/1 RB2-01-G, 0
__ __ _ __ - ___-__- - -__ ______-___ _ _ - _ _ _ _ _
_ - _ - -- - - - _ _ _ _ . - - _ . - . _ _ . . _ _ _ - _ - _ _ _
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(d) Process Monitoring Instrumentation
The cables associated with Train B instrumentation on the
remote shutdown panel for the following instruments were.
reviewed:
Train B Instrumentation (Unit 1)
Equipment Cables' Cable Fire' Area location
E51-FI-3340 B53-RS4 RB1-01-G
(RCIC flow)
E51-FIC-3325 IG7-RS4 RB1-01-G, H
(RCIC flow)
821-LI-R604BX IJ7-RS4 RB1-01-D, G
(remote inst.
panel)
CAC-LI-3342 IJ7-RS$/1 RB1-01-D, G
- .(suppression N06-WOS RB1-01-D, G
poollevel() RS4-WOS RB1-01-D, G
Train B Instrumentation (Unit 2)
Equipment Cables Cable Fire Area Location
E51-FI-3340 B53-RS4 RB2-01-G
,
(RCIC flow)
i
E51-FIC-3325 IG7-RS4/2 RB2-01-G, H
(RCIC flow)
1 B21-LI-R604BX IJ7-RS4 RB2-01-D, G
(remoteinst.
panel)
CAC-LI-3342 IJ7-RS4/1 RB2-01-D, G
(suppression WOS-WOV/1 RB2-01-D
pool level) RS4-WOS RB2-01-D, G
N08-WOS RB2-01-D, G
N - - __ _ _ _ _ ___ _ _ _____ _ __ _ _ - _ ____ _ __
____ _ _- - - - - - _--- . - - - _ - - _ - - - - _ _ _ - _ _ _ _ - - - . , . - _ - - _ - - - - - - - _ - . -_ . . - - - - . - . - - - - - - -- . - - - _ - - - - - - _ _ _
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(e) Safe. Shutdown System Support Functions
The cable' separation for following SW components which - 4
provides. the DGCW and RHR Cooling Water support functions i
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was reviewed:
Train A SW (Unit 1)
Equipment Cables Cable Fire Area Location
1A-NSWP NA7-SIF SW-1A
(Pump 1A) 51F-YF8 RB1-01-D, G; SW-1B
1-E11-F002A DE5-VQO. . RB1-01-E, G, 0
(RHRHX DE5-VQ0/1 RB1-01-E, G, 0
outlet) DE5-JF1 RB1-01-G; RB1-04
1-SW-V105 DM1-V!4 RB1-01-G, F
(inlet to DM1-V14/1 RB1-01-G, li
l
Train A SW (Unit 2)
Equipment Cables Cable Fire ' Area Location
2-CWSP2A NA4-SIG SW-1A, IB
(Pump 2A)
2-E11-F002A DE5-VQ0 RB2-01-E, 0, G
(RHRHX DES-VQO/1 RB2-01-E, O. G
outlet)
2-SW-V105 DM1-VI4 RB2-01-G, H
, (inlet to DMI-VI4/1 RB2-01-G, H
2-SW-V117 DP2-KF9 RB2-01-D, G
(inlet to RHR DP2-KF9/1 RB2-01-D, G
l'
room cooler)
I
1
TrainBSW(Unit 11
Equipment Cables Cable Fire Zone Location
IB-NSWP NAB-YF4 SW-IA, IB
(PumpIB)
IC-CSWP NA6-51X SW-1A, IB
(Pump IC)
I
--_--.~_--___-----.--_-_-__.--____-__--_.---___---__-_--.-----_L.. -------.--___--._____Q
- , - . -_ -- _ _ - _ ._ __ - _ _ _ _ - _ - - _ _
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' Train B SW (Unit 2)
' Equipment Cables Cable Fire Zone Location
7
"
2B-NSWP. NAB-SIX .SW-1A, IB
,
(Pump 2B)
}
'2-E11-F002B DN9-VQ1 RB2-01-F, G
1 i(RHRLHX< DN9-VQ1/1 .RB2-01-F, G
,L l outlet)'
y
The separation' between the sample of redundant cables / equipment-
4 listed above was' reviewed against the requirements of.10 CFR 50,
.L ' Appendix'.R.Section III.G. and the approved exemption requests i
'E contained in the' NRC's December 30, 1986, Safety Evaluation
p Report (SER).- Based on this review, the routing of the above
p+ sample of Units 1 and 2 safe shutdown cables, and the available 4
-y l' fire protection features, it appears that these ' cables' have
f adequate separation to maintain one safe shutdown train free of
fire damage, except the separation provided between the Unit I-
- i .RCIC (Train B) discharge isolation valve to the . vessel,
i 1-E51-F013, and redundant HPCI (Train A) cables. in the southwest-
corner of the Reactor Building (RB) 20' ele'vation.-
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By: letter dated April 24, .1984, as supplemented by letters of
f1 December 21, 1984, and October 28, 1985, CP&L requested a number
of exemptions from the requirements of 10 CFR 50, Appendix R.
Included in this submittal was an exemption request,
p Section 7.2.1., from the- separation requirements from
L, Section III.G. of Appendix R for the southwest corner of the
b Unit 1 RB 20' elevation. At this location, Train A HPCI cables
from the north RB cross over to the south side and enter the.
>
Control Building and come in close proximity to Train B RCIC
cables. Train A cables and Train B equipment (1-E51-F013) and
cables B41-KSS, B41-KS5/1, DK1-QD4, and DK1-QD4/1 (cables
serving valves 1-E51-F013 and 1-B32-F023B, respectively) are
'
located within 20 feet inside this zone.
Valve 1-E51-F013 provides the RCIC pump discharge isolation
function. Valve B32-F023B is utilized for the Shutdown Cooling
mode of RHR system' operation, which is required to achieve cold
r.hutdown. This valve is the suction valve of Reactor
Recirculation Pump IB, and is normally open. For the Shutdown
Cooling mode of RHR system operation, the valve is required to
be closed to prevent loss of core cooling. For the Suppression
Pool . Cooling or Low-Pressure Injection modes of RHR system
operation, the position of this valve has no adverse effect on
system function either closed or open.
b _ _ _ __ _ _ __ __ _ _ _____ _ _ _ _____ __ _
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'The NRC . SER in section 2.3.1. approves the lack of cable
separation meeting the requirements of Appendix R,
Section III.G. based on the licensee providing equivalent
protection of (1) establish 20 foot wide separation zones free
of significant quantities of intervening combustibles between .
the redundant safe shutdown trains on the 20 foot elevation, (2)
"
reroute exposed electrical cables in the separation zones out of
the zone, place the cables in conduit, enclose the cables in
noncombustible enclosures s or wrap t ie cables in 1-hour fire
rated barriers, and (3) install close'y spaced closed sprinklers
and draft stops across each separation zone to serve as water
curtains. However, the licensee has only provided a water
curtain at this location, as described in Section 5.2.(3) of the
ASCA Report. Based on the May 1989, NRC inspection walkdown it
was identified that the Train B RCIC system cables were not
routed as shown on the sketch, Figure 7.2-2, sheet 3, Revision
1, that accompanied the licensee's submittal. The cables for
Train B RCIC system at this location were actually routed in a
tray which is approximately four feet from a Train A tray
containing HPCI cables. During the June 1989 NRC Followup
inspection, the acceptability of this configuration wts reviewed
in depth. The review identified that the licensee had not O
fulfilled the requirement; of the NRC's December 30, 1986, SER .
and is in violation of Scction III.G. of Appendix R since one
train necessary for safe shutdown is not maintained free of fire
damage. This is identified as Violation Item 50-325/89-11-01,.
Failure to Provide Appendix R Separation Between HPCI (Train A)
and RCIC (Train B) Cables in the Southwest Corner, 20' Elevation
of Unit 1.
A walkdown of the zone was performed on June 21, 1989 by CP&L
and NRC' personnel. The resalt of this walkdown ideatified the
need for several changes to the existing configuration. The
licensee committed to the following actions:
1. Cables B41-KS5, B41-KS5/1, DK1-QD4 and DK1-QD4/1 will be
rerouted and/or wrapped with a 3-hour rated fire barrier
within the separation zone.
2. Tio additional sprinkler heads will be provided between
valve 1-E51-F013 (Train B) and cable tray 31L/CA (Train A).
3. Fire stops will be installed in trays 39A/BB, 39A/CB, and
39A/DB at the east and west ends of the separation zone.
4. The additions identified above will be implemented prior to
start-i.p from the next Unit I refueling outage.
The inspectors considered the proposed modifications and
schedule acceptable. Following the completion of these
modifications identified above, the fire protection features for
_ _ _ _ _ _ - _ - _ i
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the southwest corner 20' Elevation of Unit reactor building -
separation zone will conform to the requirements of Appendix
R III.G and approved exemptions to provide'a level of assurance
l
that a single fire will not adversely impact the redundant safe
'
shutdown trains.
The inspectors discussed with the licensee representatives the
fact that a fire in the SW corner of Unit I could result in the
loss of both trains A and B of the shutdown systems. The
licensee advised that the ASSD shutdown procedures were used in
conjunction with the emergency operating procedures and these
would instruct the operator to use the ADS /LPIC mode of
shutdown. However a fire in the SW corner of the reactor
building would require the operation of the ADS /LPIC system from
an auxiliary shutdown panel which had controls for only three
safety relief valves [SRV]. The NRC inspectors inquired 2@t
the analysis that was performed to insure that the time th mye
was uncovered was within an acceptable time frame. The 11cnntes
advised that there was no analysis for the use of three SRVs to
depressurize the reactor vessel to enable low pressure
injection. However, the licensee had an analysis performed
after the inquiry by the NSSS that stated that after the initial
pressurization due to reactor isolation, the reactor pressure
would be reduced by cycling of the SRVs. At approximately 38
minutes the water level is at the top of the core and cycling of 3
the SRVs causes the water level to swell at each operation until
at approximately 40 minutes with three SRVs open the LPCI pumps
can pump coolant to the core.
It was therefore concluded that even though a fire in the SW
corner of Unit I reactor building disabled both designated
shutdown trains, the unit could have been brought to hot standby ,
without exceeding Peak Clad Temperatures. This information -
taken along with the licensee's immediate corrective actions by ,
'i
removing stored combustibles, posting a firewatch and the
comitment to make the needed modifications to bring the area '
into compliance with Appendix R, III.G were considered in
determining the severity of the violation, 325/89-11-01.
The licensee was required to be in compliance with Appendix R in
4:ril 1989 for Unit I and April 1988 for Unit 2. Following
these compliance dates, the licensee conducted an indepth audit
of the fire protection program at Brunswick to verify the
commitments made in the ASCA were satisfied. This audit,
documented in CQAD 88-2245, resulted in a number of findings,
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I concerns, comments and audit follow-up items. Of particular
significance are the three Unit 2 audit findings which identify
Appendix R noncompliance. These findings were:
- Transfer Contactor Design: . The automatic transfer switches
for valves 2-B21-F016, 2-E11-F008, 2-E11-F009, 2-E41-F002,
2-E41-F079, 2-E51-F007 and 2-E51-F062 required to transfer
power to an alternate source have a failure mode which
would prevent this transfer.
l-
- RCIC Steam Supply Isolation Valve Power - Valve 2-E51-F007
l is required to be operable for Train B shutdown. The
licensee identified that, with a loss of offsite power,
fire damage to the autostart capability of the Train B
diesel generator and a spurious closure of 2-E51-F007, RCIC
will be inoperable. This is due to the need to manually
start the diesel generator which takes approximately 30
minutes. Valve 2-E51-F007 is presently supplied with AC
,
power only and RCIC injection is required within 20 minutes
according to the ASCA. Therefore, safe shutdown could not
be achieved as described in the ASCA.
- SRV Pneumatics - The Unit 2 SRVs required for postfire safe
shutdown (2-B21-F0138, 2-B21- F01 E , 2-B21-F013F and
2-B21-F013G) are supplied with air from the accumulators.
The licensee identified that, due to known leakage in the
accumulator system, the air volume necessary to allow three
i to five actuations of the SRVs over a period of 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />
is not available. Therefore, safe shutdown could not be
achieved using the methodology outlined in pl Ant procedures
0-ASSD-02, 2-ASSD-05 or 2-ASSD-06.
Following the discovery of these nonconformances, the licensee
prepared an Engineering Evaluation EER 89-0054, which provided
an evaluation / disposition and outlined corrective action for
each nonconformance. These corrective actions included
establishing continuous firewatches fi. the areas of concern and
stating modifications would be complete to correct these issues
by the end of the next Unit 2 outage. These items are
identified as a non-cited Violation (NCV) 324/89-11-01,
Appendix R Nonconformances Identified by the Licensee in CQAD
88-2245. This licensee identified violation is not being cited
because criteria specified in Section V.G of the NRC Enforcement
Policy were satisfied.
I
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b. Associated Circuits of Concern
The separation and protection requirements of 10 CFR 50, Appendix R,
apply not only to safe shutdown circuits but also to associated
circuits that could prevent operation or cause the undesired
operation of safety shutdown systems and equipment. The
identification of these associated circuits of concern was performed
for Brunswick Nuclear Plant (BNP) in accordance with NRC Generic
Letter 81-12 and subsequent NRC clarifications. Associated circuits
of concern are defined as those circuits that have a physical
separation less than that required by Section III.G.2 of Appendix R,
and have one of the following:
A common power source (common bus) where the shutdown equipment
and the power source is not electrically protected from the
circuit of concern by coordinated breakers, fuses, or similar
devices; or j
A connection to circuits of equipment whose spurious operation
(spurious signal) would adversely affect the shutdown
capability; or a common enclosure with the shutdown cables, and
Type (1) are not electrically protected by circuit
breakers, fuses or similar devices, or
Type (2) will allow propagation of the fire into the
enclosure.
(1) Associated Circuits by Common Power Supply (C mmon Bus)
Circuits a,d c6bles associated by comman power supply are simply
nonsafe shutdown cables whose fire-in:fuced failure will cause
the loss of power source '(bus, distribution panel, or MCC) tjiat
is necessary to support safe shutdown. This problem could efist
for power, congrol or instrumentation circuits. The problenl of
associatec circuits of concern by common power supply is
resolved by ensuring adequate electrical coordination between
the safe shutdown pcwer source supply breaker and the component
feeder brsakers or fuses. Such coordination ensures that the
protective device nearest to the fault operates prior to the
operation of any " upstream" devices, and limits interruption of
electrical service to a minimum amount of equipment.
l
The examination of the breaker / fuse coordination was performed
on a sample selection of circuits involving the power
distribution boards and equipment listed below:
t
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Coordination Calculation
Number and/or Drawing No. Description Breaker
9527-300-3-ED00-03-P Bus IE Supply Breaker
Drawing No. 4K/-5-ES from Bus ID AE6
Drawing No. 4KV-5-ES Breaker from Bus 1E to
Unit Substation
4160/480 V Xfrmr AF8
Drawing No. 4KV-5-ES From Unit Substation
Xfrmr to Bus SE AU9
82125-E-134-F DG Output Breaker
Feeding Bus IE AE9
82125-E-134-F Cony Service Water
Pump AF6
82125-E-134-F_ Reactor Core Spray
Pump AF2
9527-001-3-ED00-15-F Feeder Breaker to MCC
Drawing 480V-3-ES IXA from Bus E5 AU4
Not Recorded Feeder Breaker to DG
MCC AY8
Supply Breaker from
Drawing No. 1-DCC-8 Feeder Breaker to
125/250 V DC Distri-
bution Panel IB from
Battery GM-3
Drawing No. 1-DCC-17 Feeder Breaker to
power source for 125/
250 V DC Dist Panel IB)
from DC Dist Panel IB
Drawing No. SK82125-2-7416 Feeder Breaker to 125/- GM-3
250 V Distribution
Panel IB from Battery
Feeder Breaker to 125/ GM-1
'
250 V DC Distribution
'
Panel IB from Battery
Charger
Drawing No. SK82125-Z-7460 Feeder Breaker to 480- C69
and 7462 120/240 Volt Xfrmr from
MCC ICB to 50KVA Standby
UPS IB
,__ _ _ - _ - _ - - - _ _ - _ _ _ _ _ _ _ _ __. - _ _ _ _ _ _ _ __ _ _ -_ _ _ __- _ _ _ _ _ _ _ - _ _ _ - _ _ _ _ _ _ _ - _
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Drawing No. SK82125-Z-7460 Circuit No.1 off of Fuse:
and 7462 UPS Dist Panel 1A Gould-
Shawmut
Class
RK-1
Catalog
No.
'
A2K150R
The licensee controls Appendix R fuses in accordance with their
Drawings LL-09100 and 091000. Since fuse sizes are a vital
element in electrical circuit coordination / protection, a
walkdown inspection was performed on a selected number of fuses
to ensure that they agreed with the type and size listed on the
drawings. The following is a list of fuses examined: l
Fuse Designation
Location Node / Address Description
Bus E3 4KV SWGR AJS and AJ6/AY-35A Fu Gould-Shawmut
Breaker Motor, Breaker Amp-Trap /A2K35R ;
Close and Trip Circuits i
Bus E4 4KV SWGR AK0/AX-15A Fu Under- Gould-Shawmut
voltage Ckt One-Time 15
AL4 and AL5/AX-35A Fu Gould-Shawmut i
Breaker Motor,. Breaker Amp-trap A2K354 !
close and trip ,
Panel H12-P601 JF9/DD-F4 Gould-Shawmut !
Amp-trap A25210
,
Panel H12-P601 JF9/DD-F5 Gould-Shawmut
Amp-trap A25Z3
Panel H12-P612 JD/BB-F2 and F3 Gould-Shawmut
Amp-trap A25Z3
UPS Distribution HG4/ Circuit 6 Gould-Shawmut
Panel 2A Amp-trap A25X30
'
Remote Shutdown RS4/DB-line Bussman FNM 3.5
Panel
l'
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Remate Shutdown RS4/DA(-) Bussman FNM 15
Panel
Battery Charger GB6/TB5 Fu 1 Bussman FNM 3.2 )
2B-1 l
1
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Battery Charger GB6/TB4 Fu 2 Bussman FNM 3.2
MCC 2XB DN1/F1,.FN Bussman FNM 1.6
MCC 2XDB B43/F4 Bussman FNM 6
MCC 2XDB B52/F1, F3, F4 Bussman FNM 6
' The fuses listed above were observed in the plant and were in
accordance with the plant Appendix R Fuse Schedule Drawings.
The licensee acknowledged a basic problem with the
administrative control of fuses, especially non Appendix R
fuses. This is documented in their Nonconformance Report (NCR)
A-88-024 and LER 1-89-001 Supplement 1. . The corrective action
for the NCR is to develop a fuse size / type index which will
provide guidance to personnel. The Appendix R Fuse Schedule
Drawings I L-09100 and LL-091000 will be used as a pattern for
this index. The initial implementation should be completed by
the summer of 1989. The licensee also has a program underway in
which Appendix R fuse identification label plates (red
background with white lettering) are installed on MCCs, panels
and switchgear cubicles. This program should be completed by
the end of the year.
IE Information Notice 85-09, Isolation Transfer Switches and
Postfire Shutdown Capability was issued January 31, 1985. This
Notice identifies a potential . problem concerning fuses in
control circuits that are common for operation of equipment from
the Control Room and remote shutdown area. A fire in the
Control Room could cause these common fuses to blow before
transfer is made to the Remote Shutdown Area. If the control
circuit is needed at the Remote Shutdown Area to energize a
piece of equipment and if the fuse (s) blew before transfer,
equipment would not be operable without replacing the blown
fuse (s). Brunswick Nuclear Plant review for Associated Circuits
recognized the above potential and modified circuits as shown in
typical circuit modifications in Figure 4.1-1, 2, and 6 of
section 4 in their Alternative Shutdown Capability Assessment
(ASCA) Report.
(2) Associated Circuits Causing Spurious Operation (Spurious
Signals)
Circuits associated because of spurious operation are those that
can, by fire-induced failures cause safe shutdown equipment or
nonsafe shutdown equipment to operate or not to operate in a way
that defeats the function of safe shutdown systems or equipment.
Examples include uncontrolled opening or closing of valves or
circuit breakers due to fire-induced damage to nonsafe shutdown
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instrument and control circuits that affect the control circuit
interlock of the safe shutdown components.
The analysis of spurious operations considered equipment and
placed them into one of the following two categories:
(a) Spurious operation of equipment which could affect proper
safe shutdown system operation; and
(b') Spurious operation of equipment which could cause an
uncontrolled loss of primary coolant.
The equipment that falls' into the first category was addressed
by including them on the safe shutdown list for the affected
safe shutdown system and analyzing them as a safe shutdown
component.
The . equipment that falls into the second category was analyzed
on a case-by-case basis. For all potential spurious actuations
of equipment that could cause a loss of primary coolant a
resolution was provided. These resolutions fell into one or a
combination of the following types:
(a) Pre-fire action (e.g., maintain a breaker open during
normal operations)
(b) Plant modification which has been accomplished (e.g., ;
replace single-pole circuit breaker with a new two-pole
circuit breaker)
(c) Post-fire operator action (e.g., open a breaker)
Post-fire actions will take place shortly following the
identification of a substantial fire at BNP. The results
of the above analysis are summarized in BNP's Alternative
Shutdown Capability Assessment (ASCA) Report tables 3.5-49
and 3.5-50. BNP has made many plant modifications and the
protracted time span of the various submittais makes
portions of these tables and other information in the ASCA
out-of-date and it should be updated. This concern was
identified as item #9 in BNP's Audit Report QAA/0021-88-06.
The changes and modification with regard to Associated ,
Circuit of Concern were discussed with the NRC inspector i
and they are considered as meeting Appendix R requirements. 1
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.g
('3)) Associated Circuits by Common Enclosure
The common enclosure concern is found when redundant trains.are
routed together with a nonsafety circuit which crosses from one- ~
raceway or enclosure to another.and the nonsafety circuit'is not
electrically protected, or fire can destroy both redundant -
trains .due to fire ' propagating 'into enclosures containing
redundant safe shutdown circuits.
The licensee advised that their electrical coordination, the use
of IEEE-383 cable insulation, cable tray . covers, conduits an_d
rated fire seals at fire barriers, and the other actions taken
~
with ' regard to Associated Circuits of Concern provides. this
c
protection. In addition, cable wrapping / sprinklers have been
' installed in areas that present ? challenge to the concept of
common enclosure separation. This - is discussed in another
section of this report. It is considered that the licensee's'
action with regard to Appendix R Common Enclosure concerns are .
satisfactory.
c. Alternative Shutdown Capabilities
The licensee program for providing alternative shutdown capability is
addressed in their Alternative Shutdown Capability Assessment Report
(ASCA) issued to the NRC April _ 24, 1984, with subsequent supplements.
Safety Evaluation Report dated . December 30, 1986 approved the
licensee program for providing alternative shutdown capability for
the fire areas identified in the ASCA report for Unit I and Unit 2.
The inspectors reviewed operating personnel training, shift staffing
and the licensee use of alternative safe shutdown procedures (ASSD),
as these activities relate the safe shutdown of the plant during _a
given fire scenario. These activities were reviewed to determine if
the requirements of Appendix R.Section III.G.3 and III.L for
obtaining hot shutdown conditions and subsequent cold shutdown are
being met.
(1) Operator Training and Shift Staffing-
'
The inspector reviewed the licensee program for providing
training to licensed and non-iicensed operators who are required
to perform the functions of the Alternative Safa Shutdown
procedures (ASSD). It was noted that before an of erator can be
assigned .to an ASSD function on-the-job training .ind classroom
training are required to be completed. The inspeytors verified
that a comprehensive ASSD training progra was provide by the
review of lesson plans, selected. examinations, epd completed
training reports. The initial training to the 09erators also
provided "real time' walk through ASSD drill, whyre both Units
are require to be shutdown from outside the main icontrol room.
All of this initial training has been provided bylthe Operations
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Real Time Training Organization. The "on going" training
functions are. currently being turned. over' to the . Brunswick
Training Unit. Study Guides and lesson plans are being prepared
so that this transition will be completed smoothly. ASSD
training 'will be incorporated into hot license training and
licensed operator requalification training. !n addition an
L
annual ASSD drill is being planned which will . exercise the ASSD
';
shutdown functions from outside the main control room.
The licensee normal shift- staffing was reviewed to . verify that
sufficient manpower is available to operate the equipment and
L
systems described in the ASSD procedures. Fire Protection-
"
Procedure FFP 031, Fire Brigade and ASSD Staffing Roster is used
to ensure that operating personnel selected are qualified ASSD
staff members. This determination is made at the beginning of
the shift from a weekly A*SD staffing qualification roster.
Each shift ASSD staffing member is notified that they are
assigned ASSD duties and what their assignments are. Once
signed by the ASSD members and the Shift Foreman for, each unit
,
the ASSD staffing roster is posted on the control room work-
schedule bulletin board. The inspectors examined the ASSD
staffing roster posted for the day shift and noted operating
personnel assigned were on shift.- The roster consist of
(2) SRO, (3) R0, (5) A0 and (1) shift operations technician.
FFP-031 also identified the fire . brigade members and it was
noted that these member are primarily Radwaste operator with the
Shift Fire Comander being SR0 qualified. It was determined
that fire brigade members are separate from the' shift ASSD
staff.
Adequate shift staffing was further demonstrated during a
simulated walk-through of ASSD-02, Control Building Fire. This
procedure requires the use of 10 operators to take the plant
from hot shutdown to cold shutdown and is the most demanding of
the ASSD procedures with respect to manpower. The simulated
walk-through of ASSD-02 was conducted using an off-shift crew.
. The walk-through began in the main control room followed by
manning the remote shutdown panels (both units), and other
stations in the plant. The walk-through of ASSD-02 ended when
RHR shutdown cooling was established.
(2) Review of Alternative Safe Shutdown Procedures
The licensee ASCA identified fire areas within the Service Water
Building, Diesel Generator Building, Turbine Building, Reactor
Building, Control Building, and East Yard where a loss of
equipment due to fire an alternative shutdown approach is
requi red. The identified fire areas within these building
results in the issuance of 32 ASSD procedures for each Unit to
manage the shutdown process.
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The " entry" into the alternative safe shutdown procedures is
determined once the fire area is identified using Prefire Plan-
PFP-013, General' Fire Plan. If it is determined that an ASSD
fire area is involved, operational personnel " enter" ASSD-01,
Alternative Safe Shutdown Procedure Inden to assess severity of
the fire, effects on ASSD equipment, control room habitability,
etc.- If it is determine that alternative shutdown actions are
required then manual scram of the reactor can be taken if not
already done. The applicable ASSD procedure for a given fire
area is selected and initiated. For each ASSD fire area, the
safe shutdown train to be used (i.e., train A or B) is also
identified.
The inspectors selected several ASSD procedures to review as
listed below to verify that Appendix R Section III.L performance
goals have been incorporated into the procedures. The equipment
and systems to accomplish subcritical reactivity control,
reactor pressure, level, and decay 2 eat controls, establish hot
shutdown conditions and subsequent cold shutdown conditions were
determined to be provided and no discrepancies were identified.
0-PFP-013, Rev. 5, General Fire Plan
0-ASSD-00, Rev. 1, User's Guide
1-ASSD-01, Rev.1, Alternative Safe Shutdown Procedure Index
0-ASSD-02, Rev. 1, Control Building
1/2 - ASSD-05, Rev. 1, Reactor Building North
1/2 - ASSD-06, Rev. 1, Reactor Building South
1-ASSD-09, Rev. 1, Service Water Building
1-ASSD-14, Rev. 1, Diesel Generator Building 23' Elevation, DG
cell 2
In assessing the licensee's procedures, the inspectors noted
that the licensee's ASSD procedures in many cases require
reentering a fire area to take a manual action at a piece of
safe shutdown equipment. These actions are all in response to
establishing SW flow to the diesel generators, RHR room coolers
and RHR heat exchanger. The licensee justifies these actions
based on the fact that all actions are not required until at
least one hour following a fire. The inspectors found this
position to be acceptable and is clearly stated in the
licensee's ASCA. However, for a fire in the southwest corner of
the Unit I reactor building, the licensee's procedures require
the operator to immediately enter the building via the 50'
elevation doors in the northwest corner to reach the remote
shutdown panel. This access path is in the same fire area as .
the southwest 20' elevation. Threfore, the licensee prepared
an Engineering Evaluation, EER-OF .., to justify reentering the
.
fire area. This evaluation was reviewed and found acceptable by
the inspectors.
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In addition' to the review of ASSD procedures a walk-through of
ASSD-02, Control Building was conducted. The fire area for
initiating this procedure requires abandonment of the control
room and the use of 10 operators to accomplish hot shutdown and
subsequent cold shutJown. Train B equipment and systems are
specified for use by ASSC-02 procedure. The purpose of
walking-through ASS 0-02 was to verify that:
- Communications between various stations is adequate and
-
Identification plates installed on valves and
instrumentation agree with that called for in the procedure
steps.
- Lighting at stations, access and egress paths is adequate.
- Equipment and valves to be operated can be reached and are
not obstructed.
- Sound power phone headsets and procedures to be used are
available and contain the latest revision. '
- Steps of procedures are clear and can be accomplished.
-
Instrumentation identified in IEN 84-09 is available to
monitor system process variables.
The review of the above procedures and walk-through of ASSD-02
resulted in the following concerns.
(a) 0-ASSD-02, Control Building and 2-ASSD-05, Reactor Building
North, in Section A, Attachment I list conditions for which
the operators will initiate rapid depressurization of the
reactor vessel (i.e. manually open the three available
safety-relief valves (SRV's) at the remote shutdown panels)
followed by flooding the core with train B RHR system in
the low pressure core injection mode (LPCI). The inspector
concerns with using this contingency method is:
- The licensee ASCA report doesn't address using the
above contingency method, which if used will uncover
the core. Appendix R,Section III.L.2.b requires that
the reactor coolant level be maintained above the top
of the core.
An analysis was provided after the inspection that
supported the use of 3 SRV's without exceeding the
peak cladding temperature of the fuel rods.
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L~ '(b). During the inspections of. alternative shutdown equipment,
i
MCC, and ~ remote shutdown panels on; May-23,1989 . the ,
iinspectors . noted .tnat the reactor. building 50' elevation
r, - doors 302 'and 304. for Unit I were. sealed shut 'with a
L silicone sealant. A . review ' of. work authorizations j
indicated that these -doors were sealed on. November.21,
1988. This action was necessary to: stop air. inleakage to
the reactor building, which was affecting the satisfactory
~ completion of the secondary containment integrity test in
.
progress at the time.
The inspectors question the ability to gain access through j '
these doors and what ASSD procedures required this access.
-
It was determine after further review that certain fire
scenarios could block access to the reactor building
through the normal 20' elevation. It is -necessary to get-
to the 20' elevation' as the remote shutdown panels, MCC,
etc. are at this elevation for both . train A and B '
alternative shutdown equipment and need to be operated to
accomplish safe shutdown of the Units.
- Two Unit 1 ASSD procedures were-identified as requiring the
need to gain access to the reactor building via the sealed
doors. ASSD-05, Reactor Building North provides for access
to the reactor building 20' elevation through inner door
303 and outer door 304 (sealed shut). ASSD-06, Reactor
Building South provides for access to the reactor building
20' elevation through inner door 301 and outer door 302
(sealed shut). The licensee could not assure access
through the sealed doors 302 and 304 therefore immediately-
took the following compensatory measures:
- Station a fire watch in Unit i reactor building 20'
eleva:. ion.
- Stationed a tool box containing appropriate tools ;j
outside each sealed door to gain access if required. '
- Work request was initiated to remove silicone sealant
after the inner doors 301 and 303 repairs are
completed.
The licensee completed repairs to inner door 301 on June 1,
1989. On June 2,1989 an ASSD drill of ASSD-06 was
conducted to determine if access could be gained through
sealed door 302. The resident inspectors witness the
actions of MCC operator during the drill of ASSD-06. The
licensee timed the drill duration from drill initiation
until alternate power was established to valve E41-F002, q
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HPSI steam supply and the time to open the valve. The
inspectors noted that the silicone sealant did impede the
operator, but he was able to open the door within three
minutes using only the tools that are normally in his ASSD
equipment bag which is obtained at the ASSD equipment l
cabinet. The total time-for the operator to accomplish his
required actions was well within HPCI initiation and water
injection time lines.
The inspectors assessed ASSD-05 operator actions and
concluded that gaining access through door 304 that similar
difficulties would be experien::ed, but the required time
lines could be met. It appears that sealing the doors is a
minor safety significant issue. The licensee was informed
that a weakness exist in controlling and maintaining ASSD
equipment. The licensee stated that problems of this kind
should be avoided in the future when Plant Program
Procedure PLP-01.5, Alternative Shutdown Capability
Controls, is fully implemented. The procedure list in
table 1 all ASSD equipment including access / egress paths,
lighting, and communications. The inspectors detennined
that when fully implemented, this procedure should prevent
similar occurrences in the future.
Within the areas inspected no violations or deviations were
identified.
3. Compliance with 10 CFR 50, Appendix R, Section III.J., Emergency Lighting
Section III.J. requires emergency lighting units with at least an 8-hour
battery power supply to be provided in all areas needed for operation of
safe shutdown equipment and in access and egress routes thereto.
During the walkthrough of ASSD-02 discussed in paragraph 2.c. of this
report, the inspectors verified that the required emergency lights were
provided at each local control station and along access and egress paths.
Based on this sample, it appears the licensee has provided emergency
lighting in accordance with Section III.J.
4. Licensee Actions on Previous Items
(Closed) Unresolved Item 325/88-39-01 and 324/88-39-01, BTP 9.5-1,
Apendix A and 10 CFR 50, Appendix R Fire Program Implementation.
The licensee has incorporated the passive fire protection features into
the periodic test (PT)/ surveillance program initiated on May 16, 1980, to
ensure cable and conduit fire barrier wrap features required to satisfy
commitments to Appendix R and BTP 9.5-1, Appendix A are functional. The
scope of the program includes 18 month visual inspections of plant fire
barrier features i.e.; Kaowool wraps, Thermo-Lag 330-1, Interam ES0A,
Flamemastic 77 or Flame safe S100, Pyrocrete , marinite board,
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Flexi-Blanket _330-660, within safe shutdown areas of the plant. These
fire . protection features are installed within areas of the control
building, Unit 1 and Unit 2 reactor buildings, diesel . generator building
and service water intake structure.
As part of thir program, the licensee. on November 10,. 1988,' issued
approved drawings which identify boundaries of application in the above
structures.
The inspectors verified the development of Revision 0 of periodic tests
34.15.9.3 through 34.15.9.7 all of which were issued in March 1989. The
test acceptance criteria appears adequate to determine if fire barrier-
features inspected are functional. The procedures require that identified
deficiencies be . tracked by the impairment log; and correction af any
malfunctional fire _ barrier. be initiated by WR/J0. The procedures also
establish criteria for compensatory measures. upon discovery of
nonfunctional' barriers. This appears adequate. 'The licensee stated that
performance of ~ these procedures 1 has- not been implemented but are
incorporated into the next fuel cycle schedule. During this inspection a
detailed review of- specific previously identified NRC concerns in this
area. was not ' conducted but will be conducted during ' subsequent NRC
inspections. These items are closed.
5.- Exit Interview
The inspection scope and results were summarized on May 26,1989 and
June 22, 1989, with those persons indicated in Paragraph 1. The
inspectors described the areas inspected and discussed in detail the
inspection results which included the following violations:
50-325/89-11-01, Failure to Provide Appendix R Separation Between
HPCI (Train A) and RCIC (Train B) Cables in the Southwest Corner, 20'
Elevation of Unit 1.
Non-Cited Violation 50-324/89-11-01,- Appendix R Nonconformances
Identified by the Licensee in CQAD 88-2245.
During the first exit meeting the licensee committed to get an engineering
evaluation to verify the use of ADS /LPCI as a backup for the loss of the
shutdown trains due to a fire in the SW corner of Unit I reactor building.
'
During the second exit meeting, the licensee committed to take actions
required to bring the fire area in the SW corner of Unit 1 into compliance
with Appendix R,Section III G.
The licensee did not identify as proprietary any information reviewed by
the inspectors. No dissenting comment were received from the licensee.
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1 1
6. Acronyms and Initialisms
ADS Automatic Depressurization System
A0 Auxiliary Operator
ASCA Alternative Shutdown Capability Assessment
ASSD Alternative Safe Shutdown Procedures
AUX Auxiliary i
'
BNP Brunswick Nuclear Plant
CFR Code of Federal Regulations
CKT Circuit
CON 7 Conventional
CST Condensate Storage Tank
DIST Distribution
DG Diesel Generator
F Fuse
FU Fuse.
.HPCI High Pressure Coolant Injection
HX Heat Exchanger
KVA Kilovolt Amperes
LPCI Low Pressure Core Injection
MCC Motor Control Center
RCIC Reactor Core Isolation Cooling
R0 Reactor Operator
SP Suppression Pool
SR0 Senior Reactor Operator
TC' Torus Cooling
UPS Uninterruptible Power Source i
'
V Volt
XFRMR Transformer
i
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