ML101690164
ML101690164 | |
Person / Time | |
---|---|
Site: | River Bend |
Issue date: | 06/17/2010 |
From: | O'Keefe N NRC/RGN-IV/DRS/EB-2 |
To: | Mike Perito Entergy Operations |
References | |
EA-10-095 | |
Download: ML101690164 (49) | |
See also: IR 05000458/2010006
Text
UNITED STATES
NUCLEAR REGULATORY COMMI SSI ON
R E G I ON I V
612 EAST LAMAR BLVD, SUITE 400
ARLINGTON, TEXAS 76011-4125
June 17 2010
Michael Perito
Vice President, Operations
Entergy Operations, Inc.
River Bend Station
5485 US Highway 61N
St. Francisville, LA 70775
SUBJECT: RIVER BEND STATION - NRC TRIENNIAL FIRE PROTECTION INSPECTION
REPORT 05000458/2010006 AND NOTICE OF VIOLATION
Dear Mr. Perito:
On June 2, 2010, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at
River Bend Station facility. The enclosed inspection report documents the inspection results,
which were discussed on April 23, 2010, with Mr. Eric Olson, General Manager of Plant
Operations, and in a telephonic exit meeting on June 2, 2010 with Mr. Jerry Roberts and other
members of your staff.
The inspection examined activities conducted under your license as they relate to safety and
compliance with the Commissions rules and regulations and with the conditions of your license.
The team reviewed selected procedures and records, observed activities, and interviewed
personnel.
This report documents four NRC-identified violations. One violation is cited in the enclosed
Notice of Violation and the circumstances surrounding it are described in detail in the subject
inspection report. The violation is being cited in the Notice because of your failure to correct a
significant non-compliance with your License Condition 2.C.(10), Fire Protection, within a
reasonable time as described in the NRC Enforcement Manual. The NRC has also identified
three other issues that were evaluated under the risk significance determination process as
having very low safety significance (Green). The NRC also determined that violations are
associated with these issues. These violations are being treated as Noncited Violations
(NCVs), consistent with Section VI.A of the Enforcement Policy. These NCVs are described in
the subject inspection report.
You are required to respond to this letter and should follow the instructions specified in the
enclosed Notice of Violation when preparing your response. The NRC will use your response,
in part, to determine whether further enforcement action is necessary to ensure compliance with
regulatory requirements.
Entergy Operations, Inc. -2-
If you contest the noncited violations or their significance, you should provide a response within
30 days of the date of this inspection report, with the basis for your denial, to the Nuclear
Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with
copies to: (1) the Regional Administrator, Region IV, 612 East Lamar Blvd., Arlington, TX
76011-4125; (2) the Director, Office of Enforcement, United States Nuclear Regulatory
Commission, Washington, DC 20555-0001; and (3) NRC Resident Inspector at River Bend
Station facility. The information you provide will be considered in accordance with Inspection
Manual Chapter 0305.
In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter, its
enclosures, and your response (if any) will be available electronically for public inspection in the
NRC Public Document Room or from the Publicly Available Records (PARS) component of the
NRCs document system (ADAMS). ADAMS is accessible from the NRC Website at
http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room). To the extent
possible, your response should not include any personal privacy, proprietary, or safeguards
information so that it can be made available to the Public without redaction.
Sincerely,
/RA/
Neil OKeefe, Chief
Engineering Branch 2
Division of Reactor Safety
Docket No. 50-458
License No. NPF-47
Enclosure: Inspection Report No. 05000458/2010006
w/Attachments:
1 - Notice of Violation
2 - Supplemental Information
cc w/Enclosure:
Senior Vice President and COO
Entergy Operations, Inc
P. O. Box 31995
Jackson, MS 39286-1995
Vice President, Oversight
Entergy Operations, Inc.
P. O. Box 31995
Jackson, MS 39286-1995
Senior Manager, Nuclear Safety & Licensing
Entergy Nuclear Operations
P. O. Box 31995
Jackson, MS 39286-1995
Entergy Operations, Inc. -3-
Manager, Licensing
Entergy Operations, Inc.
5485 US Highway 61N
St. Francisville, LA 70775
Attorney General
State of Louisiana
P. O. Box 94005
Baton Rouge, LA 70804-9005
Ms. H. Anne Plettinger
3456 Villa Rose Drive
Baton Rouge, LA 70806
President of West Feliciana
Police Jury
P. O. Box 1921
St. Francisville, LA 70775
Mr. Brian Almon
Public Utility Commission
William B. Travis Building
P. O. Box 13326
Austin, TX 78701-3326
Mr. Jim Calloway
Public Utility
Commission of Texas
1701 N. Congress Avenue
Austin, TX 78711-3326
Louisiana Department of Environmental Quality
Radiological Emergency Planning and
Response Division
P. O. Box 4312
Baton Rouge, LA 70821-4312
Joseph A. Aluise
Associate General Counsel - Nuclear
Entergy Services, Inc.
639 Loyola Avenue
New Orleans, LA 70113
Chief, Technological Hazards
Branch
FEMA Region VI
800 N. Loop 288
Denton, TX 76209-3606
Entergy Operations, Inc. -4-
Electronic distribution by RIV:
Regional Administrator (Elmo.Collins@nrc.gov)
Deputy Regional Administrator (Chuck.Casto@nrc.gov)
DRP Director (Dwight.Chamberlain@nrc.gov)
DRP Deputy Director (Anton.Vegel@nrc.gov)
DRS Director (Roy.Caniano@nrc.gov)
DRS Deputy Director (Troy.Pruett@nrc.gov)
Senior Resident Inspector (Grant.Larkin@nrc.gov)
Resident Inspector (Charles.Norton@nrc.gov)
Branch Chief, DRP/C (Vincent.Gaddy@nrc.gov)
RBS Administrative Assistant (Lisa.Day@nrc.gov)
Senior Project Engineer, DRP/C (Bob.Hagar@nrc.gov)
Project Engineer, DRP/C (Rayomand.Kumana@nrc.gov)
Public Affairs Officer (Victor.Dricks@nrc.gov)
Public Affairs Officer (Lara.Uselding@nrc.gov)
Project Manager (Alan.Wang@nrc.gov)
Branch Chief, DRS/TSB (Michael.Hay@nrc.gov)
RITS Coordinator (Marisa.Herrera@nrc.gov)
Regional Counsel (Karla.Fuller@nrc.gov)
Congressional Affairs Officer (Jenny.Weil@nrc.gov)
Senior Enforcement Specialist Ray.Kellar@nrc.gov
OEMail Resource
OEDO RIV Coordinator (Leigh.Trocine@nrc.gov)
File located: S:\DRS\REPORTS\(final) RBS2010006 rpt-STG ADAMS ML
SUNSI Rev Compl. ; Yes No ADAMS ; Yes No Reviewer Initials NFO
Publicly Avail ;Yes No Sensitive Yes ; No Sens. Type Initials NFO
SRI:DRS/EB2 RI:DRS/EB2 RI:DRS/EB2 RI:DRS/EB2 SRA:DRS
SGraves SAlferink BCorrell NOkonkwo MRunyun
/RA/ /RA/ /RA/ /RA/ /RA/
6/9/10 6/9/10 6/9/10 6/9/10 6/9/10
SES:ACES C: DRP/PBC C:DRS/EB2
RKellar VGaddy NFOKeefe
/RA/ /RA/ /RA/
6/15/10 6/17/10 6/17/10
OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax
NOTICE OF VIOLATION
Entergy Operations, Inc. Docket No. 50-458
River Bend Station License No. NPF-47
During an NRC inspection completed on June 2, 2010, a violation of NRC requirements was
identified. In accordance with the NRC Enforcement Policy, the violation is listed below:
License Condition 2.C.(10), Fire Protection, requires that the licensee comply with the
requirements of their fire protection program as specified in Attachment 4. Attachment
4, Fire Protection Program Requirements, states, in part, that the licensee shall
implement and maintain in effect all provisions of the approved fire protection program
as described in the Final Safety Analysis Report for the facility. The fire protection
program requirements are described in section 9.5.1 and appendices 9A and 9B.
Section 9B.4.7 specifies, in part, Fire protection features shall be capable of limiting fire
damage so that one train of systems necessary to achieve and maintain hot shutdown
conditions from either the control room or emergency control station(s) is free of fire
damage.
Contrary to this requirement, in May 2007, the licensee determined that they failed to
ensure that one train of systems necessary to achieve and maintain hot shutdown
conditions from either the control room or emergency control station(s) was free of fire
damage. Specifically, the Division 1 standby service water support system to the
Division 1 emergency diesel generator, which was required to achieve safe shutdown,
was not protected such that it remained free from fire damage under all conditions.
The non-emergency high temperature trips for the emergency diesel generator would be
disabled by design when automatically started in emergency mode due to loss of offsite
power. Since standby service water could be lost due to fire damage during a control
room fire, the emergency diesel generator would continue to run without cooling, and
potentially fail prior to operators restoring standby service water at the remote shutdown
panel. The licensee failed to promptly restore compliance in the three years since
identifying the non-conforming condition, during which time the licensee has completed
two refueling outages, six unplanned outages, and a planned system outage of sufficient
duration. This condition was entered into the licensees corrective action program as CR-
This violation is associated with Green significance determination process finding 05000458/2010006-01.
Pursuant to the provisions of 10 CFR 2.201, Entergy Operations, Inc. is hereby required to
submit a written statement or explanation to the U.S. Nuclear Regulatory Commission,
ATTN: Document Control Desk, Washington, DC 20555-0001 with a copy to the Regional
Administrator, Region IV, 612 East Lamar Blvd., Arlington, TX 76011-4125, and a copy to the
NRC Resident Inspector at River Bend Station within 30 days of the date of the letter
transmitting this Notice of Violation (Notice). This reply should be clearly marked as a Reply to
a Notice of Violation: EA-10-095 and should include for each violation: (1) the reason for the
violation, or, if contested, the basis for disputing the violation, (2) the corrective steps that have
been taken and the results achieved, (3) the corrective steps that will be taken, and (4) the date
when full compliance will be achieved. In your response, please provide a description of the
-1- Enclosure
process(es) used and your assessment of the appropriateness of the decisions to extend the
completion of necessary plant modifications beyond the November 2009 refueling outage. Your
response may reference or include previous docketed correspondence, if the correspondence
adequately addresses the required response. If an adequate reply is not received within the
time specified in this Notice, an order or a Demand for Information may be issued as to why the
license should not be modified, suspended, or revoked, or why such other action as may be
proper should not be taken. Where good cause is shown, consideration will be given to
extending the response time.
If you contest this enforcement action, you should also provide a copy of your response, with
the basis for your denial, to the Director, Office of Enforcement, United States Nuclear
Regulatory Commission, Washington DC 20555-0001.
Because your response will be made available electronically for public inspection in the NRC
Public Document Room or from the NRCs document system (ADAMS), accessible from the
NRCs website at www.nrc.gov/reading-rm/pdr.html or www.nrc.gov/reading-rm/adams.html, to
the extent possible, it should not include any personal privacy, proprietary, or safeguards
information so that it can be made available to the public without redaction. If personal privacy
or proprietary information is necessary to provide an acceptable response, then please provide
a bracketed copy of your response that identifies the information that should be protected and a
redacted copy of your response that deletes such information. If you request withholding of
such material, you must specifically identify the portions of your response that you seek to have
withheld and provide in detail the bases for your claim of withholding (e.g., explain why the
disclosure of information will create an unwarranted invasion of personal privacy or provide the
information required by 10 CFR 2.390(b) to support a request for withholding confidential
commercial or financial information). If safeguards information is necessary to provide an
acceptable response, please provide the level of protection described in 10 CFR 73.21.
Dated this 17th day of June 2010.
-2- Enclosure
ENCLOSURE
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Docket: 50-458
License: NPF-47
Report No.: 05000458/2010006
Licensee: Entergy Operations, Inc.
Facility: River Bend Station
Location: 5485 U.S. Highway 61
St. Francisville, LA
Dates: April 5 through June 2, 2010
Team S. Graves, Senior Reactor Inspector
Leader: Engineering Branch 2
Division of Reactor Safety
Inspectors: S. Alferink, Reactor Inspector
Engineering Branch 2
Division of Reactor Safety
B. Correll, Reactor Inspector
Engineering Branch 2
Division of Reactor Safety
N. Okonkwo, Reactor Inspector
Engineering Branch 2
Division of Reactor Safety
Approved Neil OKeefe, Branch Chief
By: Engineering Branch 2
Division of Reactor Safety
-3- Enclosure
SUMMARY OF FINDINGS
IR 05000458/2010006; 4/5/10 - 6/2/10; Entergy Operations, Inc.; River Bend Station; Fire
Protection (Triennial)
The report covered a two week triennial fire protection team inspection by specialist inspectors
from Region IV. Four Green findings were identified and categorized as one cited violation
(NOV) and three noncited violations (NCVs). The significance of most findings is indicated by
their color (Green, White, Yellow, Red) using Inspection Manual Chapter 0609, Significance
Determination Process. The crosscutting aspects were determined using Inspection Manual
Chapter 0310, Components within the Cross-Cutting Areas. Findings for which the
significance determination process (SDP) does not apply may be Green or be assigned a
severity level after NRC management review. The NRCs program for overseeing the safe
operation of commercial nuclear power reactors is described in NUREG-1649, Reactor
Oversight Process, Revision 4, dated December 2006.
A. NRC-Identified and Self-Revealing Findings
Cornerstone: Mitigating Systems
- Green. The team identified a cited violation of License Condition 2.C.(10), Fire
Protection, for failing to ensure that the Division 1 standby service water support
system to the Division 1 emergency diesel generator, which was required to achieve
safe shutdown, was protected such that it remained free from fire damage under all
conditions. This condition was identified by the licensee in May 2007, and entered
into their corrective action program as a significant non-conforming condition in CR-
RBS-2007-02102. The licensee subsequently initiated compensatory measures in
the form of manual actions to protect the Division 1 emergency diesel generator.
This issue was documented as a licensee-identified noncited violation in Inspection
Report 2009002. River Bend has subsequently completed two refueling outages, six
forced outages, and one emergency diesel generator work window of sufficient
duration since identification of this condition and failed to correct the non-
conformance. The team determined that schedule changes resulted in a new
completion date of January 2011.
The failure to ensure that one train of systems necessary to achieve and maintain
hot shutdown conditions from either the control room or emergency control station(s)
was free of fire damage and to correct this significant non-conforming condition in a
timely manner is a performance deficiency. This performance deficiency was more
than minor because it was associated with the protection against external factors
(fire) attribute of the Mitigating Systems Cornerstone and adversely affected the
cornerstone objective of ensuring the availability, reliability, and capability of systems
that respond to initiating events in order to prevent undesirable consequences. The
team evaluated this deficiency using Inspection Manual Chapter 0609, Appendix F,
Fire Protection Significance Determination Process, because it affected fire
protection defense-in-depth strategies involving post fire safe shutdown systems with
plant-wide consequences. A Phase 3 SDP risk assessment was performed by a
senior reactor analyst. The bounding change in conditional core damage frequency
for a 1-year exposure is the Fire Mitigation Frequency (4.30E-08/year) multiplied by
the change in conditional core damage probability (0.9) for a value of 3.87E-08/year.
This value indicates the finding has very low safety significance (Green). Because
-4- Enclosure
the licensee failed to correct this violation, this violation is being treated as a cited
violation, consistent with the NRC Enforcement Policy. This finding had a
crosscutting aspect in the Work Control component of the Human Performance area
because the licensee did not appropriately plan work activities to support long-term
equipment reliability by limiting temporary modifications, operator workarounds,
safety systems unavailability, and reliance on manual actions H.3(b). (Section
1R05.01)
- Green. The team identified a noncited violation of Technical Specification 5.4.1.d,
Fire Protection Program Implementation. Specifically, Procedure AOP-0031
Shutdown from Outside the Main Control Room, Revision 307, had steps that could
not be implemented as written. Two steps were to be performed before the
necessary ac power was available, and two steps required diagnostic assessment
without the availability of instrumentation.
The failure to ensure that Procedure AOP-0031, Revision 307 could be implemented
as written is a performance deficiency. The performance deficiency was more than
minor because it was associated with the procedure quality attribute of the Mitigating
Systems Cornerstone and it adversely affected the cornerstone objective of ensuring
the availability, reliability, and capability of systems that respond to initiating events
to prevent undesirable consequences. Using Attachment 2 to Appendix F, Fire
Protection Significance Determination Process, this issue was determined to be a
safe shutdown finding, and was assigned a degradation rating of Low because the
examples involved procedural deficiencies that could be compensated for by
operator experience. Since this finding was assigned a low degradation rating, the
safety significance screened as very low (Green). This finding was entered into the
licensees corrective action program as CR-RBS-2010-01592, CR-RBS-2010-01831,
CR-RBS-2010-01775, CR-RBS-2010-01821, and CR-RBS-2010-1846. This finding
had a crosscutting aspect in the Resources component of the Human Performance
area, in that the licensee did not ensure that procedures were complete, accurate,
and up to date to assure nuclear safety [H.2.(c)]. (Section 1R05.05.b.1)
- Green. The team identified a noncited violation of License Condition 2.C.(10), Fire
Protection, for the failure to implement and maintain in effect all provisions of the
approved fire protection program. Specifically, the team identified, during a timed
walkdown of the procedure that it took operators over 6 minutes to isolate feedwater,
but the simulator showed that the steam lines could be flooded in 2 minutes.
Overfilling the reactor pressure vessel and flooding the main steam lines could make
reactor core isolation cooling unavailable. Reactor core isolation cooling was
credited for decay heat removal and inventory control in the event of a fire.
The failure to ensure that feedwater would be isolated prior to overfilling the reactor
pressure vessel and flooding the main steam lines making reactor core isolation
cooling unavailable is a performance deficiency. The performance deficiency was
more than minor because it was associated with the protection against external
events (fire) attribute of the Mitigating Systems Cornerstone and it adversely affected
the cornerstone objective of ensuring the availability, reliability, and capability of
systems that respond to initiating events to prevent undesirable consequences. The
team evaluated this finding using Inspection Manual Chapter 0609, Appendix F, Fire
Protection Significance Determination Process, because it affected fire protection
defense-in-depth strategies involving post fire safe shutdown systems with plant-
-5- Enclosure
wide consequences. A senior reactor analyst performed a Phase 3 evaluation to
determine the risk significance of this finding since it involved a control room fire that
led to control room abandonment. The Phase 3 evaluation determined that the
finding had very low safety significance because a fire in only one of 109 electrical
cabinets in the control room could result in this overfill event. The finding was
entered into the licensees corrective action program as CR-RBS-2010-01808. The
finding did not have a crosscutting aspect since it was not indicative of current
performance, in that the licensee had established the incorrect response time more
than three years prior to this finding. (Section 1R05.05.b.2)
- Green. The team identified a noncited violation of License Condition 2.C.(10), Fire
Protection, related to the licensee's failure to implement and maintain in effect all
provisions of the approved fire protection program. Specifically, during testing
required by the approved fire protection program the licensee failed to adequately
test the remote shutdown emergency transfer switch functions used to assure
isolation of safe shutdown equipment from the control room in the event of a control
room evacuation due to fire. The switch functions had not been adequately tested
since 1997.
The failure to ensure isolation from the control room for safe shutdown equipment
controlled from the remote shutdown panel during surveillance testing of emergency
transfer switches is a performance deficiency. The finding was more than minor
because it was associated with the procedure quality attribute of the Mitigating
Systems Cornerstone in that it adversely affected the cornerstone objective of
ensuring the availability, reliability, and capability of systems that respond to initiating
events to prevent undesirable consequences. The team evaluated the finding using
Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance
Determination Process, because it affected fire protection defense-in-depth
strategies involving post fire safe shutdown. Using Appendix F, Attachment 2,
Degradation Rating Guidance Specific to Various Fire Protection Program
Elements, the team determined that the finding constituted a low degradation of the
safe shutdown area since the control room isolation feature was expected to display
nearly the same level of effectiveness and reliability as it would had the degradation
not been present. This finding screened as having very low safety significance
(Green). This violation was entered into the licensees corrective action program as
CR-RBS-2010-01783. Because the emergency transfer switch surveillance
procedures had been in effect since 1997, there was no crosscutting aspect
associated with the violation, in that it is not indicative of current licensee
performance. (Section 1R05.05.b.3)
B. Licensee-Identified Violations
None.
-6- Enclosure
REPORT DETAILS
1. REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R05 Fire Protection (71111.05T)
This report presents the results of a triennial fire protection inspection conducted in
accordance with NRC Inspection Procedure 71111.05T, Fire Protection (Triennial), at
the River Bend Station. The inspection team evaluated the implementation of the
approved fire protection program in selected risk significant areas, with an emphasis on
the procedures, equipment, fire barriers, and systems that ensure the post fire capability
to safely shut down the plant.
Inspection Procedure 71111.05T requires the selection of three to five fire areas for
review. The inspection team used the fire hazards analysis section of the River Bend
Station Individual Plant Examination of External Events to select the following five risk
significant fire areas (inspection samples) for review:
C-15 Division I Standby Switchgear Room
C-17 Control Room Ventilation Room (El. 116)
C-25 Control Room
AB-2/Z-1 and Z-2 High Pressure Core Spray and High Pressure Core Spray
Hatch Area
PT-1 Piping Tunnel
The inspection team evaluated the licensees fire protection program using the
applicable requirements, which included plant Technical Specifications, Operating
License Condition 2.C.(10), NRC safety evaluations, NRC supplemental safety
evaluations, 10 CFR 50.48, and Branch Technical Position 9.5-1. The team also
reviewed related documents that included the Final Safety Analysis Report (FSAR),
Section 9.5.1; Technical Requirements Manual; the fire hazards analysis; and the post
fire safe shutdown analysis.
Specific documents reviewed by the team are listed in the attachment. Five inspection
samples were completed.
.01 Protection of Safe Shutdown Capabilities
a. Inspection Scope
The team reviewed piping and instrumentation diagrams, safe shutdown equipment list,
safe shutdown design basis documents, and the post fire safe shutdown analysis to
verify that the safe shutdown methodology had properly identified the components and
systems necessary to achieve and maintain safe shutdown conditions for equipment in
the selected fire areas. The team also reviewed and observed walkdowns of the
procedures for achieving and maintaining safe shutdown in the event of a fire to verify
that the licensee properly implemented the safe shutdown analysis provisions.
-7- Enclosure
For each of the selected fire areas, the team reviewed the separation of redundant safe
shutdown cables, equipment, and components located within the same fire area. The
team also reviewed the licensees method for meeting the requirements of 10 CFR
50.48; Branch Technical Position 9.5-1, Appendix A; and 10 CFR Part 50, Appendix R,
Sections III.G. Specifically, the team evaluated whether at least one post fire safe
shutdown success path remained free of fire damage in the event of a fire. In addition,
the team verified that the licensee met applicable license commitments.
b. Findings
Introduction. The team identified a Green, cited violation of License Condition 2.C.(10)
Fire Protection, for failing to ensure that one train of systems necessary to achieve and
maintain hot shutdown conditions from either the control room or emergency control
station(s) is free of fire damage and failing to promptly correct this non-conforming
condition.
Description. On May 21, 2007, during a review of industry operating experience, the
licensee determined that the Division 1 emergency diesel generator could be disabled
during a main control room fire due to fire damage to a required support system.
Specifically, the non-emergency high temperature trips for the emergency diesel
generator would be disabled by design when the engine is automatically started in
emergency mode due to loss of offsite power. Since standby service water could be lost
due to fire damage during a control room fire, the emergency diesel generator would
continue to run without cooling and potentially fail prior to operators restoring standby
service water at the remote shutdown panel. The Division 1 emergency diesel generator
is the credited source of ac power used to safely shut down the reactor in the event of a
fire requiring evacuation of the main control room with concurrent loss of offsite power.
The licensee documented this non-conformance in Condition Report
CR-RBS-2007-02102 as a significant non-conforming condition and implemented
compensatory measures in the form of operator manual actions. The manual actions
were added to Procedure AOP-0031, Shutdown from Outside the Main Control Room,
Revision 307, to immediately trip the emergency diesel generator after an emergency
mode start and transfer control to the remote shutdown panel prior to control room
evacuation. Once transferred, operators would ensure the availability of standby service
water and perform a manual normal-mode start of the emergency diesel generator, in
which the high temperature trips would remain functional.
This non-conforming condition was reported to the NRC as an unanalyzed condition that
significantly degrades plant safety, in accordance with 10 CFR 50.72(b)(3)(ii)(B) and
subsequently in July 2007, in Licensee Event Report (LER) 05000458/07-003-00.
The team was concerned that the licensee had not been timely in restoring compliance.
In late 2008, the NRC concluded that this non-conforming condition constituted a
licensee-identified Green noncited violation. At that time, the licensee had scheduled
corrective action for this condition for November 2009. The team learned that this was
later rescheduled because the modification package was not completed in time and
parts were not available to support the scheduled date. While the licensee had
concluded that the work could be done online, the modification was not ready so it was
rescheduled for the next refueling outage in January 2011.
-8- Enclosure
The team noted that the licensee had concluded that multiple spurious operations had to
occur for the condition to impact safe shutdown in the event of a fire. Further
discussions with the licensee resulted in the team concluding that the loss of offsite
power also was inappropriately considered as a fire-induced spurious actuation in the
control room fire scenario, and because the standby service water system could be
subject to maloperation due to fire-damage, The licensee classified this scenario as an
event requiring multiple fire induced spurious actuations in order to occur. This incorrect
conclusion contributed to licensee decisions to delay completion of corrective actions.
The team pointed out that demonstrating the ability to safely shutdown in the event of a
fire in the control room is a deterministic design requirement, not a spurious operation.
Similarly, the postulated loss of standby service water is the result of fire damage, not a
spurious operation.
The Onsite Safety Review Committee evaluated the core damage frequency and
concluded that the risk of rescheduling the modification was very low. However, the
team noted that this condition was classified by the licensee as being operable but a
significant non-conforming condition. Regulatory Issue Summary 2005-20 references
Inspection Manual Part 9900, Revision to Guidance Formerly Contained in NRC Generic Letter 91-18, Information to Licensees Regarding Two NRC Inspection Manual
Sections on Resolution of Degraded and Non-conforming Conditions and on
Operability, which states, in part, that degraded or non-conforming conditions must be
corrected in a timely manner, commensurate with the safety significance. Also, for
technical specification systems, structures, or components, the NRC expects that issues
requiring compensatory measures and issues involving manual actions in lieu of
automatic system response would indicate conditions that should be fixed expeditiously.
While the licensee used this guidance in their decision making process, the team was
concerned that the licensee did not appropriately consider this guidance before delaying
implementation of the modification. Further, at the time of this inspection, the plant had
conducted two refueling outages, six unplanned outages, and a planned system outage
of sufficient duration since identifying the condition. The team concluded that the total
time to restore compliance did not reflect timely corrective action, and rescheduling to
the January 2011 refueling outage rather than adjusting online maintenance schedules
did not reflect a work control process that was focused on scheduling work activities so
as to minimize reliance on manual actions.
Section 7.2 of Inspection Manual Part 9900 states, in part, that "In determining whether
the licensee is making reasonable efforts to complete corrective actions promptly, the
NRC will consider safety significance, the effects on operability, the significance of the
degradation, and what is necessary to implement the corrective action. The NRC may
also consider the time needed for design, review, approval, or procurement of the repair
or modification; the availability of specialized equipment to perform the repair or
modification; and whether the plant must be in hot or cold shutdown to implement the
actions. If the licensee does not resolve the degraded or nonconforming condition at the
first available opportunity or does not appropriately justify a longer completion schedule,
the staff would conclude that corrective action has not been timely and would consider
taking enforcement action."
-9- Enclosure
In applying this guidance to this issue, the staff concluded that:
- The systems affected by the non-conforming condition and the compensatory
measures are systems required to be operable by technical specifications. These
systems are also required to be operable to meet License Condition 2.C.(10) and the
safe shutdown requirements of the approved fire protection program.
- The non-conforming condition was more significant based on the reliance upon
manual actions in lieu of automatic functioning, and because compensatory actions
were necessary to ensure the operability of the affected systems.
- Scheduling the modification for completion in the second refueling outage following
identification of the issue was justified based on the proximity of the first outage to
the date of identification and the time needed for design and procurement activities.
- Delay of the modification to the third refueling outage, rather than scheduling a work
window sooner, did not appear to have adequately considered the factors described
in Part 9900. Further, delays in design and procurement appeared to be the result of
factors within the control of the licensee, given proper priority.
Based on the above, the staff has concluded that corrective action for this non-
conforming condition was not timely commensurate with the safety significance of the
condition.
Analysis. The failure to ensure that at least one train of equipment necessary to achieve
hot shutdown from either the control room or emergency control station(s) is maintained
free of fire damage as required by the licensees fire protection program, and to correct
this significant non-conforming condition in a timely manner is a performance deficiency.
This performance deficiency was more than minor because it was associated with the
protection against external factors (fire) attribute of the Mitigating Systems Cornerstone
and adversely affected the cornerstone objective of ensuring the availability, reliability,
and capability of systems that respond to initiating events in order to prevent undesirable
consequences. The team evaluated this deficiency using Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, because it
affected fire protection defense-in-depth strategies involving post fire safe shutdown
systems with plant-wide consequences. A Phase 3 SDP risk assessment was
performed by a senior reactor analyst.
Because the River Bend control room included the plant instrumentation and relay
cabinets, the senior reactor analyst added a generic fire ignition frequency for a relay
room to the control room fire ignition frequency listed in the Individual Plant Examination
for External Events. The analyst multiplied an appropriate severity factor (SF) by the
sum of the control room fire initiation frequency (CRFIF) and the instrument room fire
initiation frequency (IRFIF) and multiplied this result by a nonsuppression probability
(NPCRE) to account for the likelihood that operators failed to extinguish the fire within 20
minutes, assuming that it would take operators 2 minutes to detect the fire. The
resulting fire would require a control room evacuation with a control room evacuation
frequency determined as follows:
- 10 - Enclosure
Control Room Evacuation Frequency = (CRFIF + IRFIF) * SF * NPCRE
= (9.5E-03/year + 1.42E-03/year) * 0.2 * 1.30E-02
= 2.84E-05/year
As described in the Individual Plant Examination for External Events, the control room
had 109 panels. Because multiple failure combinations could result in a start of the
Division 1 diesel generator without service water supplied, the senior reactor analyst
determined the combined partial fraction for all possible scenarios. The analyst
determined partial fraction for each loss of electrical scenario by dividing the number of
affected cabinets by the total number of cabinets:
Scenario Number Fraction (number/109)
Cabinets with Diesel Generator 1 controls 4 FDG1 = 3.67E-02
Cabinets with Division 1 power 1 FDiv1 = 9.17E-03
Cabinets with power from both divisions 1 FBDIV = 9.17E-03
Cabinets with service water 3 FSW = 2.75E-02
A fire could result in the inadvertent start of a diesel generator either directly, by affecting
the diesel control circuits, or indirectly, by affecting the power to the associated vital bus.
Therefore, the probability that a fire could result in the start of the Division 1 emergency
diesel generator (PDGStart) was calculated as follows:
PDGStart = FDG1 + FDiv1 + FBDiv
= 3.67E-02 + 9.17E-03 + 9.17E-03
= 5.50E-02
To determine the probability that a main control room fire would fail the service water
system at the same time as starting the Division 1 emergency diesel generator (PFailure),
the analyst performed the following calculation:
PFailure = PDGStart * FSW
= 5.50E-02 * 2.75E-02
= 1.52E-03
The resulting Fire Mitigation Frequency is the Control Room Evacuation Frequency
(2.84E-05/year) multiplied by the combined failure probabilities (1.52E-03) for a value of
4.30E-08/year.
The analyst determined the change in conditional core damage probability by subtracting
the base case conditional core damage probability given abandonment of the control
room (0.1) from the assumed conditional core damage probability given the performance
deficiency (1.0) for a value of (0.9). The bounding change in conditional core damage
frequency for a 1-year exposure is the Fire Mitigation Frequency (4.30E-08/year)
- 11 - Enclosure
multiplied by the change in conditional core damage probability (0.9) for a value of
3.87E-08/year. This value indicates the finding has very low safety significance (Green).
This finding had a crosscutting aspect in the Work Control component of the Human
Performance area because the licensee did not appropriately coordinate work activities
to support long-term equipment reliability by limiting temporary modifications, operator
workarounds, safety systems unavailability, and reliance on manual actions H.3(b).
Enforcement. License Condition 2.C.(10) Fire Protection, requires that the licensee
comply with the requirements of their fire protection program as specified in Attachment
4. Attachment 4, Fire Protection Program Requirements, states, in part, that the
licensee shall implement and maintain in effect all provisions of the approved fire
protection program as described in the Final Safety Analysis Report for the facility. The
fire protection program requirements are described in section 9.5.1 and appendices 9A
and 9B of the Final Safety Analysis Report. Section 9B.4.7, specifies, in part, Fire
protection features shall be capable of limiting fire damage so that one train of systems
necessary to achieve and maintain hot shutdown conditions from either the control room
or emergency control station(s) is free of fire damage.
Contrary to this requirement, in May 2007 the licensee determined that they failed to
ensure that the one train of systems necessary to achieve and maintain hot shutdown
conditions from either the control room or emergency control station(s) would be free of
fire damage. Specifically, the Division 1 standby service water support system to the
Division 1 emergency diesel generator, which was required to achieve safe shutdown,
was not protected such that it remained free from fire damage under all conditions.
Because the licensee failed to correct this violation, this violation is being treated as a
cited violation, consistent with the NRC Enforcement Policy,Section VI.A.1, which
states, in part, that a cited violation requiring a formal written response from a licensee
will be considered if the licensee failed to restore compliance within a reasonable time
after a violation was identified. The NRC Enforcement Manual further explains that the
purpose of this criterion is to emphasize the need to take appropriate action to restore
compliance in a reasonable period of time once a licensee becomes aware of the
violation, and take compensatory measures until compliance is restored when
compliance cannot be reasonably restored within a reasonable period of time.
The licensee had compensatory measures in place; however compliance had not been
restored.
This violation is identified as VIO 05000458/2010006-01, Failure to Ensure at Least One
Train of Equipment Necessary to Achieve Hot Shutdown Conditions is Free of Fire
Damage.
.02 Passive Fire Protection
a. Inspection Scope
The team walked down accessible portions of the selected fire areas to observe the
material condition and configuration of the installed fire area boundaries (including walls,
fire doors, and fire dampers) and verify that the electrical raceway fire barriers were
appropriate for the fire hazards in the area. The team compared the installed
- 12 - Enclosure
configurations to the approved construction details, supporting fire tests, and applicable
license commitments.
The team reviewed installation, repair, and qualification records for a sample of
penetration seals to ensure the fill material possessed an appropriate fire rating and that
the installation met the engineering design.
b. Findings
No findings.
.03 Active Fire Protection
a. Inspection Scope
The team reviewed the design, maintenance, testing, and operation of the fire detection
and suppression systems in the selected fire areas. The team verified the manual and
automatic detection and suppression systems were installed, tested, and maintained in
accordance with the National Fire Protection Association code of record or approved
deviations, and that each suppression system was appropriate for the hazards in the
selected fire areas.
The team performed a walkdown of accessible portions of the detection and suppression
systems in the selected fire areas. The team also performed a walkdown of major
system support equipment in other areas (e.g., fire pumps, and Halon supply systems)
to assess the material condition of these systems and components. The team reviewed
the electric and diesel fire pump flow and pressure tests to verify that the pumps met
their design requirements.
The team assessed the fire brigade capabilities by reviewing training, qualification, and
drill critique records. The team also reviewed pre-fire plans and smoke removal plans
for the selected fire areas to determine if appropriate information was provided to fire
brigade members and plant operators to identify safe shutdown equipment and
instrumentation, and to facilitate suppression of a fire that could impact post fire safe
shutdown capability. The team inspected fire brigade equipment to determine
operational readiness for fire fighting.
The team observed an unannounced fire drill on April 13, 2010, and the subsequent drill
critique using the guidance contained in Inspection Procedure 71111.05AQ, Fire
Protection Annual/Quarterly. The team observed fire brigade members fight a
simulated fire in Fire Area C-14, Standby Switchgear 1B Room, located in the Control
Building. The team verified that the licensee identified problems, openly discussed them
in a self-critical manner at the drill debrief, and identified appropriate corrective actions.
Specific attributes evaluated were: (1) proper wearing of turnout gear and self-contained
breathing apparatus; (2) proper use and layout of fire hoses; (3) employment of
appropriate fire fighting techniques; (4) sufficient firefighting equipment was brought to
the scene; (5) effectiveness of fire brigade leader communications, command, and
control; (6) search for victims and propagation of the fire into other areas; (7) smoke
removal operations; (8) utilization of pre-planned strategies; (9) adherence to the pre-
planned drill scenario; and (10) drill objectives.
- 13 - Enclosure
b. Findings
No findings.
.04 Protection from Damage from Fire Suppression Activities
a. Inspection Scope
The team performed plant walkdowns and document reviews to verify that redundant
trains of systems required for hot shutdown, which are located in the same fire area,
would not be subject to damage from fire suppression activities or from the rupture or
inadvertent operation of fire suppression systems. Specifically, the team verified that:
- A fire in one of the selected fire areas would not directly, through production of
smoke, heat, or hot gases, cause activation of suppression systems that could
potentially damage all redundant safe shutdown trains.
- A fire in one of the selected fire areas or the inadvertent actuation or rupture of a
fire suppression system would not directly cause damage to all redundant trains
(e.g., sprinkler-caused flooding of other than the locally affected train).
- Adequate drainage is provided in areas protected by water suppression systems.
The team reviewed the separation of safe shutdown cables, equipment, and
components within the same fire areas, and reviewed the methodology for meeting the
requirements of 10 CFR 50.48, Appendix A to Branch Technical Position 9.5-1 and
10 CFR Part 50, Appendix R, Section III.G. Specifically, this was to determine whether
at least one post fire safe shutdown success path was free of fire damage in the event of
a fire in the selected areas.
b. Findings
No findings.
.05 Alternative Shutdown Capability
a. Inspection Scope
Review of Methodology
The team reviewed the safe shutdown analysis, fire hazards analysis, operating
procedures, piping and instrumentation drawings, electrical drawings, the Final Safety
Analysis Report, and other supporting documents to verify that hot and cold shutdown
could be achieved and maintained for fires in areas where the licensees post fire safe
shutdown strategy relied on manipulating shutdown equipment from outside the control
room. The team verified that hot and cold shutdown could be achieved and maintained
with or without offsite power available.
The team conducted plant walkdowns to verify that the plant configuration was
consistent with the description contained in the safe shutdown and fire hazards
analyses. The team focused on ensuring the adequacy of systems selected for
- 14 - Enclosure
reactivity control, reactor coolant makeup, reactor decay heat removal, process
monitoring instrumentation, and support systems functions.
The team also verified that the systems and components credited for safe shutdown
would remain free from fire damage, with the exceptions discussed in this report.
Finally, the team verified that the transfer of control from the control room to the
alternative shutdown location would not be affected by fire-induced circuit faults (e.g., by
the provision of separate fuses and power supplies for alternative shutdown control
circuits), with the exceptions discussed below.
Review of Operational Implementation
The team verified that licensed and non-licensed operators received training on
alternative shutdown procedures. The team also verified that a sufficient number of
personnel, exclusive of those assigned as fire brigade members, were trained and
available onsite at all times to perform an alternative shutdown.
The team reviewed the adequacy of the procedures utilized for alternative shutdown and
performed an independent walkthrough of the procedure to ensure their implementation
and human factors adequacy. The team also verified that the operators could be
reasonably expected to perform specific time critical actions within the time required to
maintain plant parameters within specified limits, with the exceptions discussed below.
Some of the time critical actions verified included the restoration of alternating current
electrical power, establishing control at the remote shutdown and local shutdown panels,
establishing reactor coolant makeup, and establishing decay heat removal.
The team reviewed periodic surveillance testing of the alternative shutdown transfer
capability, including transfer and isolation of instrumentation and control functions, to
verify that the tests were adequate to demonstrate the functionality of the alternative
shutdown capability. The team also reviewed a sample of wiring diagrams, vendor
manuals, connection drawings, and circuit diagrams for the remote transfer circuits,
control circuits, and the remote shutdown panel to verify the field configurations matched
the design documents.
b. Findings
b.1 Introduction. The team identified a Green noncited violation of Technical Specification 5.4.1.d, Fire Protection Program Implementation, for failing to ensure that the
alternative shutdown procedure, AOP-0031 Shutdown from Outside the Main Control
Room, Revision 307, could be implemented as written, with three examples.
Description. Procedure AOP-0031 Shutdown from Outside the Main Control Room,
Revision 307, was used in the event of a fire in the control room which required control
room evacuation. This procedure contained the necessary steps to safely shut down the
reactor with or without offsite power available. During a walkdown of the procedure, the
team identified three examples where this procedure could not be performed as written.
Example 1: Step 5.10.5 required the operators to verify at least one of three breakers
(ACB04, ACB06, or ACB07) was closed to supply power to the Division I
vital switchgear. The team determined that operators would not able to
perform the step as written during a control room fire scenario with a loss of
- 15 - Enclosure
offsite power since these three breakers would be open and locked out.
Breakers ACB04 and ACB06 would open by design upon the loss of offsite
power. The Division I diesel generator output breaker, ACB07, would be
open because the operators performed an emergency stop of the diesel
generator in the control room as a manual action to prevent damage to the
diesel generator. Further, a caution note before step 5.10.5 informed the
operator not to close these breakers without specific instruction from the
Control Room Supervisor. The team also noted that Procedure AOP-0031
did not require the diesel generator to be started again until step 5.14.2.
Example 2: Step 5.13 required the Reactor Building Operator to start 1LSV*C3A,
Penetration Valve Leakage Control Air Compressor. This compressor
provides air pressure to maintain the safety relief valves open during
sustained operation of the residual heat removal system in the alternate
shutdown cooling mode, if required. During a loss of offsite power, this
compressor would not have ac power available until after the Division 1
emergency diesel generator was started. As noted above, Procedure
AOP-0031 did not require the diesel generator to be started until step
5.14.2. Step 5.14.1 directed the Control Room Supervisor to verify that
steps 5.10.5 and 5.13 were completed. This step occurred before
establishing electrical power in step 5.14.2. During interviews with the
operators, the team concluded that the Control Room Supervisor would
direct an operator to start the diesel generator upon realization that ac
power was required to perform steps 5.10.5 and 5.13.
Example 3: Steps 5.14.5.3 and 5.15.3 required the operators to perform a diagnostic
evaluation for fire damage to cables and motor-operated valves in the form
of IF fire-induced cable [valve] damage has occurred to the following,
THEN perform the following The procedure did not provide guidance or
identify protected instrumentation for assessing whether this fire damage
occurred. The post fire safe shutdown analysis credited the actions
specified in steps 5.14.5.3 and 5.15.3 for the plant to reach and maintain
hot shutdown. The team was concerned that it might not be practical to
identify specific cable damage within the time constraints.
Analysis. The failure to ensure that Procedure AOP-0031, Revision 307, could be
implemented as written is a performance deficiency. The performance deficiency was
more than minor because it was associated with the procedure quality attribute of the
Mitigating Systems Cornerstone and it adversely affected the cornerstone objective of
ensuring the availability, reliability, and capability of systems that respond to initiating
events to prevent undesirable consequences. The team evaluated the finding using
Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance
Determination Process, because it affected fire protection defense-in-depth strategies
involving post fire safe shutdown systems with plant-wide consequences. Using
Appendix F, Attachment 2, Degradation Rating Guidance Specific to Various Fire
Protection Program Elements, the team determined that the finding constituted a low
degradation of the safe shutdown area since the procedural deficiencies could be
compensated by operator experience and familiarity. This finding screened as having
very low safety significance (Green).
- 16 - Enclosure
This finding had a crosscutting aspect in the Resources component of the Human
Performance area because the licensee did not ensure that procedures used to assure
nuclear safety could be implemented [H.2.(c)].
Enforcement. Technical Specification 5.4.1.d states, in part, that written procedures
shall be established, implemented, and maintained covering fire protection program
implementation. Contrary to this requirement, prior to June 2, 2010, the licensee failed
to implement and maintain a required fire protection program procedure. Specifically,
the licensee failed to ensure that Procedure AOP-0031, Shutdown from Outside the
Main Control Room, Revision 307, could be implemented as written.
Because this violation was of very low safety significance and it was entered into the
licensees corrective action program as CR-RBS-2010-01592, CR-RBS-2010-01831,
CR-RBS-2010-01775, CR-RBS-2010-01821, and CR-RBS-2010-1846, this violation is
being treated as an NCV, consistent with the Enforcement Policy and is identified as
NCV 05000458/2010006-02, Failure to Ensure Alternative Shutdown Procedure could
be Implemented as Written.
b.2 Introduction. The team identified a Green noncited violation of License
Condition 2.C.(10), Fire Protection, for the failure to implement and maintain in effect
all provisions of the approved fire protection program. Specifically, during a timed
walkdown of the procedure the team identified that it took operators over 6 minutes to
isolate feedwater, but the simulator showed that the steam lines could be flooded in 2
minutes. Overfilling the reactor pressure vessel and flooding the main steam lines could
make reactor core isolation cooling unavailable. Reactor core isolation cooling was
credited for decay heat removal and inventory control in the event of a fire.
Description. Design Criterion 240.201A, Post-Fire Safe Shutdown Analysis, Revision
4, contained a listing of the equipment and their function relied upon for post fire safe
shutdown in the approved fire protection program. This analysis credited the use of the
reactor core isolation cooling system and safety relief valves during a control room fire
scenario which forces evacuation. Procedure AOP-0031, Shutdown from Outside the
Main Control Room, Revision 307, was used to shut down the reactor in the event of a
fire that required evacuation of the control room. This procedure contained the steps to
safely shut down the reactor with or without offsite power available. Step 5.10.1 of
Attachment 13 to AOP-0031 provided instructions for opening the circuit breakers for the
motor-driven feedwater pumps and removing the control power fuses within 5 minutes of
evacuating the main control room. Without prompt isolation of the feedwater system,
feedwater could continue to inject and overfill the reactor vessel up to the steam lines.
Flooding the reactor vessel up to the level of the steam lines could disable the reactor
core isolation cooling system and damage the steam lines. The reactor core isolation
cooling system was relied upon in this scenario to restore and maintain reactor vessel
level and control pressure. Overfilling the reactor vessel could also damage the safety
relief valves since they were not analyzed to pass high pressure water. The safety relief
valves are located on the main steam lines upstream of the inboard main steam isolation
valves and are required to open to vent steam to the suppression pool and prevent
reactor vessel overpressure.
Calculation G13.18.12.2-27, 10 CFR 50 Appendix R Manual Action Time Frame,
Revision 1, provided best estimate times for the performance of manual actions to
prevent placing the reactor in an unrecoverable condition. This calculation identified that
- 17 - Enclosure
operators must isolate feedwater with a high priority upon leaving the control room.
The post fire safe shutdown analysis concluded that a time limit of 5 minutes met the
intent of high priority as stated in the calculation.
During a timed walkdown of Procedure AOP-0031, Revision 307, the team noted that it
took 6 minutes 45 seconds for the operators to isolate feedwater injection outside of the
main control room. During subsequent discussions, licensee staff was unable to provide
a technical basis to support why the 5-minute time limit to isolate feedwater was
acceptable. To improve understanding of the issue and to obtain an estimate of the time
available to isolate feedwater, the team observed a simulator scenario with the high
reactor level (Level 8) feedwater trip disabled due to fire damage, and the feedwater
pumps continuing to inject. The level 8 trip is an automatic initiation, which during a fire
scenario was not verified to be free of fire damage and functional. In this scenario, the
inspectors observed that it took approximately 2 minutes for the reactor water level to
reach the level of the main steam lines. From this scenario, the inspectors determined
that the 5-minute time limit appeared nonconservative, in that the licensee could not
demonstrate that it would be sufficient to ensure the availability of all equipment relied
upon for post fire safe shutdown, specifically the reactor core isolation cooling system
would not be available if operators were not able to prevent filling the steam lines with
water.
Analysis. The failure to ensure that feedwater would be isolated prior to overfilling the
reactor pressure vessel and flooding the main steam lines making reactor core isolation
cooling unavailable was a performance deficiency.
The performance deficiency was more than minor because it was associated with the
protection against external events (fire) attribute of the Mitigating Systems Cornerstone
and it adversely affected the cornerstone objective of ensuring the availability, reliability,
and capability of systems that respond to initiating events to prevent undesirable
consequences. The team evaluated this finding using Inspection Manual Chapter 0609,
Appendix F, Fire Protection Significance Determination Process, because it affected
fire protection defense-in-depth strategies involving post fire safe shutdown systems with
plant-wide consequences. A senior reactor analyst performed a Phase 3 evaluation to
determine the risk significance of this finding since it involved a control room fire that led
to control room evacuation.
Since the River Bend Station control room included the plant instrumentation and relay
cabinets, the senior reactor analyst added a generic fire ignition frequency for the relay
room (FIFIR) to the control room fire ignition frequency (FIFCR) listed in the Individual
Plant Examination for External Events. The analyst multiplied the combined fire ignition
frequency by a severity factor (SF) and a non-suppression probability indicating that
operators failed to extinguish the fire within 20 minutes assuming a 2 minute detection
that required a control room evacuation (NPCRE). The resulting control room evacuation
frequency (FCR-EVAC) was:
FCR-EVAC = (FIFCR+FIFIR) * SF * NPCRE
= (9.50E-3/yr + 1.42E-3/yr) * 0.2 * 1.30E-2
= 2.84E-5/yr
- 18 - Enclosure
The control room had a total of 109 cabinets. The analyst determined that a single fire in
only one of these cabinets could lead to the spurious operation and loss of control
function for the feedwater system which could result in overfilling the reactor vessel to
the main steam lines or above. The analyst calculated a bounding change in core
damage frequency for the finding (CDFFIRE-MFW) by multiplying the combined fire ignition
frequency by the fraction of panels containing the affected circuits.
CDFFIRE-MFW = FCR-EVAC * 1 / 109
= 2.84E-5/yr * 0.0092
= 2.61E-7/yr
This frequency was considered to be bounding since it assumed:
1) Fire damage in the applicable cabinet would create circuit faults such that the
feedwater pumps continued to operate and the level 8 trip would be disabled,
resulting in overfilling the reactor vessel above the main steam lines and,
2) The conditional core damage probability given a control room fire with evacuation
and the spurious operation of the feedwater system was equal to one, and
3) The performance deficiency accounted for the entire change in core damage
frequency (i.e., the baseline core damage frequency for this event was zero).
In accordance with the guidance in Manual Chapter 0609, Appendix H, Containment
Integrity Significance Determination Process, the senior risk analyst screened the
finding for its potential risk contribution to large early release frequency (LERF) since the
bounding change in core damage frequency provided a risk significance estimate
greater than 1E-7.
The issue represented a Type A finding, based on the guidance in Appendix H, because
the finding influenced the likelihood of accidents leading to core damage. As
documented in Appendix H, Table 5.1, accident sequences that lead to large early
release frequency for boiling water reactors with Mark III containment include high
pressure transient events.
The analyst determined that most of the sequences involving control room evacuation
with spurious operation of the feedwater system resulted in the reactor coolant system
being at high pressure at the time of vessel breach. Using Table 5.2, Phase 2
Assessment Factors - Type A Findings at Full Power, the analyst selected a large early
release frequency factor of 0.2 for these sequences. The sum of the large early release
frequency score as stated in Step 3.2, LERF Significance Evaluation, was then
quantified. The change in large early release frequency was estimated to be 5.22E-08.
This value agrees with the result of the change in core damage frequency evaluation
that the finding was of very low safety significance (Green).
The finding did not have a crosscutting aspect since it was not indicative of current
performance, in that the licensee had established the incorrect response time more than
three years prior to this finding.
- 19 - Enclosure
Enforcement. License Condition 2.C.(10), Fire Protection, requires that the licensee
comply with the requirements of their fire protection program as specified in Attachment
4. Attachment 4, Fire Protection Program Requirements, states, in part, that the
licensee shall implement and maintain in effect all provisions of the approved fire
protection program as described in the Final Safety Analysis Report for the facility. The
fire protection program requirements are described in section 9.5.1 and appendices 9A
and 9B. Appendix 9A references Design Criterion 240.201A.
Design Criterion 240.201A, Post-Fire Safe Shutdown Analysis, Revision 4, contained a
listing of the equipment and their function relied upon for post fire safe shutdown in the
approved fire protection program. This analysis credited the use of the reactor core
isolation cooling system during a control room fire scenario.
Contrary to this requirement, prior to June 2, 2010, the licensee failed to implement and
maintain in effect all provisions of the approved fire protection program. Specifically, the
licensee failed to ensure that the reactor core isolation cooling system would be
available for post fire safe shutdown during a control room fire scenario. Because this
violation was of very low safety significance and it was entered into the licensees
corrective action program as CR-RBS-2010-01808, this violation is being treated as an
NCV, consistent with the Enforcement Policy and is identified as NCV 05000458/2010006-03, Failure to Implement and Maintain in Effect all Provisions of the
Approved Fire Protection Program.
b.3 Introduction. The team identified a Green noncited violation of License Condition
2.C.(10), Fire Protection, related to the licensee's failure to implement and maintain in
effect all provisions of the approved fire protection program. Specifically, the licensee
failed to adequately test the remote shutdown emergency transfer switch functions used
to assure isolation of safe shutdown equipment from the control room in the event of a
control room evacuation due to fire.
Description. License Condition 2.C.(10), Fire Protection, requires that the licensee
comply with the requirements of their fire protection program as specified in Attachment
4. Attachment 4, Fire Protection Program Requirements, states, in part, that the
licensee shall implement and maintain in effect all provisions of the approved fire
protection program as described in the Final Safety Analysis Report for the facility. The
fire protection program requirements are described in section 9.5.1 and appendices 9A
and 9B. Section 9A.3.4.5, Test and Test Control, requires in part, that a test program
be established and implemented to assure that testing is performed and verified by
inspection to demonstrate conformance with the design and system readiness
requirements. For a fire in the control room requiring control room evacuation, the
functions of the emergency transfer switches are: 1) transfer control of selected
equipment to the remote shutdown panel and other local control stations, and 2) isolate
the applicable fire area circuits to prevent fire damage from disabling or causing
maloperation of equipment. The remote shutdown panel emergency transfer switches
are required to be operated during control room evacuation events per procedure
AOP-0031, Shutdown from Outside the Main Control Room, Revision 307.
Alignment for remote operation is accomplished via a series of transfer switches and
multiplying relays. The River Bend Station design uses General Electric type SB-9 and
Electro Switch type 20KB switches, in conjunction with General Electric model CR120BC
and Gould model J11A relays. During review, the team identified that the testing
- 20 - Enclosure
methodology in the surveillance procedures did not appear adequate to ensure isolation
of power, control and instrumentation circuits from the control room, in that the licensees
surveillance procedures did not ensure that all contacts on the transfer switches used for
isolation of the associated fire area performed their intended function as required. If a
contact used for control room isolation failed to reposition when the emergency transfer
switch was taken to the Emergency position, the surveillance procedures, as written,
would not identify the failed contact. The licensee's surveillance test procedures verified
that the control function was transferred from the main control room to the remote
shutdown panel by operating the equipment from the remote panel. For the isolation
function however, the procedures only checked that control room indicating lights
extinguished on the main control panels as the method of verifying control room circuit
paths were isolated. Using electrical schematic and wiring diagrams, the team was able
to identify examples where control room indicating lights might be extinguished without
ensuring that the control room portion of the circuit was isolated from the emergency
control circuit. The surveillance procedures did not verify that all other parallel control
circuit paths in the associated fire area were isolated. In the event that one or more
contacts used for control room isolation failed to reposition, a fire induced circuit failure
could cause the control power fuses to open or cause maloperation, and result in a loss
of equipment or system required to function to achieve and maintain safe shutdown
conditions in the event of a control room fire. A review of licensee documents indicated
that the isolation function of the emergency transfer switches had not been adequately
tested since 1997.
The licensee performed internal reviews of maintenance and corrective action
documents searching for failures of the emergency transfer switches and multiplying
relays. The licensee also performed reviews of past operability and surveillance tests for
equipment operated by the transfer switch circuitry, and reviewed industry operating
experience for documented failures of the switch and relay types used at River Bend
Station. The industry operating experience review revealed one documented failure of
the SB-9 type switch, but was determined to be due to a switch configuration not
applicable to River Bend Station. The licensee documented their basis for having
reasonable assurance of operability of the emergency transfer switches and relays,
which justified continued operation until their next refueling outage scheduled for
January 2011, at which time validation testing and analysis of the transfer and isolation
circuitry will be performed. The team reviewed a licensee document detailing remote
shutdown panel transfer switch reliability, Corrective Action 1 to LO-LAR-2010-00120,
and held internal discussions with a regional senior reactor analyst to review the
licensees continued operability conclusions and agreed that reasonable assurance of
operability existed.
Analysis. The failure to ensure isolation from the control room during surveillance
testing of emergency transfer switches for safe shutdown equipment controlled from the
remote shutdown panel is a performance deficiency. The performance deficiency was
reviewed against Inspection Manual Chapter 0612, Appendix B "Issue Screening" to
determine whether the performance deficiency was of minor or more-than-minor
significance. The performance deficiency was determined to be sufficiently similar to
Example 4.L of Inspection Manual Chapter 0612, Appendix E, "Examples of Minor
Issues" to reasonably conclude that it satisfied at least one of the minor screening
questions. The finding was more than minor because it was associated with the
procedure quality attribute of the Mitigating Systems Cornerstone in that it adversely
- 21 - Enclosure
affected the cornerstone objective of ensuring the availability, reliability, and capability of
systems that respond to initiating events to prevent undesirable consequences.
The team evaluated the finding using Inspection Manual Chapter 0609, Appendix F,
Fire Protection Significance Determination Process, because it affected fire protection
defense-in-depth strategies involving post fire safe shutdown. Using Appendix F,
Attachment 2, Degradation Rating Guidance Specific to Various Fire Protection
Program Elements, the team determined that the finding constituted a low degradation
of the safe shutdown area since the control room isolation feature is expected to display
nearly the same level of effectiveness and reliability as it would had the degradation not
been present. This finding screened as having very low safety significance (Green).
Because the emergency transfer switch surveillance procedures had been in effect since
1997, there was no crosscutting aspect associated with the violation, in that it is not
indicative of current licensee performance.
Enforcement. License Condition 2.C.(10), Fire Protection, requires that the licensee
comply with the requirements of their fire protection program as specified in Attachment
4. Attachment 4, Fire Protection Program Requirements, states, in part, that the
licensee shall implement and maintain in effect all provisions of the approved fire
protection program as described in the Final Safety Analysis Report for the facility. The
fire protection program requirements are described in section 9.5.1 and appendices 9A
and 9B. Section 9A.3.4.5, Test and Test Control, requires in part, that a test program
be established and implemented to assure that testing is performed and verified by
inspection to demonstrate conformance with the design and system readiness
requirements. Contrary to these requirements, the licensee failed to implement and
maintain in effect all provisions of the approved fire protection program as described in
the Final Safety Analysis Report for the facility, in that the transfer switch testing
program did not verify that each required emergency transfer switch was capable of
performing the required isolation function in accordance with their approved fire
protection program.
Because this violation was of very low safety significance and it was entered into the
licensees corrective action program as CR-RBS-2010-01783, this violation is being
treated as an NCV, consistent with the Enforcement Policy and is identified as NCV 05000458/2010006-04, Failure to Implement and Maintain in Effect all Provisions of the
Approved Fire Protection Program.
.06 Circuit Analysis
a. Inspection Scope
The team reviewed the post fire safe shutdown analysis to verify that the licensee
identified circuits that could impact the ability to achieve and maintain safe shutdown.
The team verified, on a sample basis, that the licensee properly identified cables and
equipment required to achieve and maintain hot shutdown conditions in the event of a
fire in the selected fire areas. The team verified that cables associated with safe
shutdown-related equipment were protected from the adverse effects of fire damage or
were analyzed to show that fire induced cable faults (e.g., hot shorts, open circuits, and
shorts to ground) would not prevent safe shutdown.
- 22 - Enclosure
The team evaluated cables for selected components from the reactor core isolation
cooling and residual heat removal systems. For the sample of components selected, the
team reviewed process and instrumentation diagrams, electrical schematics, and wiring
diagrams to identify power, control, and instrumentation cables necessary to support
safe shutdown equipment operation. In addition, the team reviewed cable routing
information to verify that fire protection features were in place to satisfy the separation
requirements specified in the fire protection license basis.
Since the licensee utilized thermoset cables for most applications, the team reviewed the
following cable failure modes for selected required and associated circuits:
$ Spurious actuations resulting from any combination of conductors within a single
multiconductor cable;
$ A maximum of two cables considered where multiple individual cables may be
damaged by the same fire;
$ The vulnerability of three phase power cables resulting from three phase proper
polarity hot shorts for decay heat removal system isolation valves at high-
pressure to low-pressure interfaces.
In addition, on a sample basis, the adequacy of circuit protective coordination for safe
shutdown power sources was evaluated. Also, on a sample basis, the adequacy of
electrical protection provided for non-essential cables that share a common enclosure
with cables for required safe shutdown equipment was reviewed to ensure that the
non-essential cables are adequately protected to preclude common enclosure concerns.
Specific components reviewed by the team are listed in the attachment.
b. Findings
No findings.
.07 Communications
a. Inspection Scope
The team reviewed the adequacy of the communication systems to support plant
personnel in the performance of alternative post fire safe shutdown functions and fire
brigade duties. The review verified that the licensee established and maintained in
working order the credited primary and backup communication systems. The review
also verified that problems with communication equipment necessary for alternative safe
shutdown support were properly categorized in the corrective action program and
received the appropriate priority. The team evaluated the environmental impacts such
as ambient noise levels, coverage patterns, and clarity of reception. The team verified
that the design and location of communications equipment such as repeaters, private
branch exchanges, and transmitters would not cause a loss of communications during a
fire.
The team verified the contents of designated storage lockers and reviewed the
alternative shutdown procedure to verify that portable radio communications and fixed
- 23 - Enclosure
emergency communications systems were available, operable, and adequate for the
performance of designated activities.
b. Findings
No findings.
a. Inspection Scope
The team reviewed emergency lighting system required for alternative shutdown to verify
that it was adequate to support the performance of manual actions required to achieve
and maintain safe shutdown conditions, and to illuminate access and egress routes to
the areas where manual actions would be required. The locations and positioning of
emergency lights were observed during a walkthrough of Procedure AOP-0031,
Shutdown from Outside the Main Control Room, Revision 307, and during review of
manual actions implemented for the fire areas other than the control room.
The team verified the licensee installed emergency lights with an 8-hour capacity,
maintained the emergency light batteries in both fixed and portable configurations in
accordance with manufacturer recommendations, and tested and performed
maintenance in accordance with plant procedures and industry practices.
b. Findings
No findings.
.09 Cold Shutdown Repairs
a. Inspection Scope
The team verified that the licensee identified repairs needed to reach and maintain cold
shutdown and had dedicated repair procedures, equipment, and materials to accomplish
these repairs. The only repair credited by the licensee was the use of electrical jumpers
for temporary Division I 480 Vac power to Residual Heat Removal (RHR) shutdown
cooling inboard isolation valve E12-MOV-F009, in the event of a main control room fire
and the loss of Division II 480 Vac electrical power.
Using Attachment 6, Jumper Procedure for E12-F009 to Procedure AOP-0031,
Revision 307, the team evaluated whether these repairs could be accomplished as
written to bring the plant to cold shutdown within the time frames specified in their design
and licensing bases. The team verified that the repair equipment, components, tools,
and materials needed for the repairs were available and accessible on site. For
equipment that was not pre-staged, the team verified that the equipment could be
procured and installed within the time frames specified in their design and licensing
basis.
b. Findings
No findings.
- 24 - Enclosure
.10 Compensatory Measures
a. Inspection Scope
The team verified that compensatory measures were implemented for out-of-service,
degraded or inoperable fire protection and post fire safe shutdown equipment, systems,
or features (e.g., detection and suppression systems and equipment; passive fire
barriers; and pumps, valves, or electrical devices providing safe shutdown functions or
capabilities). The team verified that the short-term compensatory measures
compensated for the degraded function or feature until appropriate corrective action
could be taken, and that the licensee was effective in returning the equipment to service
in a reasonable period of time, with the exception described in section 0.1 of this report.
The team reviewed licensee manual actions used to mitigate the effects of fire in order to
assess their feasibility and reliability. The team reviewed the manual actions against the
items listed in NUREG-1852, Demonstrating the Feasibility and Reliability of Operator
Manual Actions in Response to Fire, dated October 2007. The manual actions were
found to be in accordance with the guidance.
b. Findings
No findings.
.11 B.5.b Inspection Activities
a. Inspection Scope
The team reviewed the licensees implementation of guidance and strategies intended to
maintain or restore core cooling, containment, and spent fuel pool cooling capabilities
under the circumstances associated with loss of large areas of the plant due to
explosions or fire as required by Section B.5.b of the Interim Compensatory Measures
Order, EA-02-026, dated February 25, 2002 and 10 CFR 50.54(hh)(2).
The team reviewed licensees strategies to verify that they continued to maintain and
implement procedures, maintain and test equipment necessary to properly implement
the strategies, and ensure station personnel are knowledgeable and capable of
implementing the procedures. The team performed a visual inspection of portable
equipment used to implement the strategy to ensure availability and material readiness
of the equipment, including the adequacy of portable pump trailer hitch attachments, and
verify the availability of on-site vehicles capable of towing the portable pump. The team
assessed the off-site ability to obtain fuel for the portable pump, and foam used for
firefighting efforts. The strategies and procedures selected for this inspection sample
included:
- Spent Fuel Pool Makeup/Spray Strategies, OSP-0066, Extensive Damage
Mitigation Procedure, Revision 003, Attachment 13, Spent Fuel Pool
Emergency Makeup/Spray Strategies.
- 25 - Enclosure
Procedure, Revision 003, Attachment 8, RCIC Operation with a Loss of AC and
DC Power.
b. Findings
No findings.
4. OTHER ACTIVITIES [OA]
4OA2 Identification and Resolution of Problems
Corrective Actions for Fire Protection Deficiencies
a. Inspection Scope
The team selected a sample of condition reports associated with the licensees fire
protection program to verify that the licensee had an appropriate threshold for identifying
deficiencies. The team reviewed the corrective actions proposed and implemented to
verify that they were effective in correcting identified deficiencies. The team evaluated
the quality of recent engineering evaluations through a review of condition reports,
calculations, and other documents during the inspection.
b. Findings
No findings.
4OA6 Meetings, Including Exit
Exit Meeting Summary
On April 23, 2010, a preliminary exit meeting was held in which the team presented the
preliminary inspection results to Mr. Eric Olson and other members of the licensee staff.
On June 2, 2010, an additional exit meeting was held telephonically, and the inspection
results were presented to Mr. Jerry Roberts and other members of the licensee staff.
The licensee acknowledged the findings presented. The team asked the licensee
whether any of the material examined during the inspection should be considered
proprietary. No proprietary information was identified.
4OA7 Licensee-Identified Violations
None
ATTACHMENT: SUPPLEMENTAL INFORMATION
- 26 - Enclosure
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
C. Forpahl Manager, Programs and Components
D. LaBorde Ops Procedures
D. Lorfing Manager, Licensing
E. Olson General Manager, Plant Operations
G. Krause Assistant Ops Manager
H. Goodman Engineering Director
J. Roberts Director, Nuclear Safety Assurance
K. Huffstatler Senior Licensing Specialist
L. Woods Manager, Quality Assurance
M. Chase Manager, Training
R. Kerar Senior Engineer - Fire Protection
NRC Personnel
G. Larkin, Senior Resident Inspector
C. Norton, Resident Inspector
M. Runyun, Senior Reactor Analyst
K. Bucholtz, Technical Specifications Branch, Office of Nuclear Reactor Regulation
R. Elliott, Technical Specifications Branch, Office of Nuclear Reactor Regulation
C. Schulten, Technical Specifications Branch, Office of Nuclear Reactor Regulation
R. Telson, Reactor Inspection Branch, Office of Nuclear Reactor Regulation
-1- Attachment
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
05000458/2010006-01 VIO Failure to Ensure at Least One Train of Equipment
Necessary to Achieve Hot Shutdown Conditions is
Free of Fire Damage (Section 1R05.01)
Opened and Closed
05000458/2010006-02 NCV Failure to Ensure Alternative Shutdown Procedure
could be Implemented as Written (Section
1R05.05.b.1)05000458/2010006-03 NCV Failure to Implement and Maintain in Effect all
Provisions of the Approved Fire Protection Program
(Section 1R05.05.b.2)05000458/2010006-04 NCV Failure to Implement and Maintain in Effect all
Provisions of the Approved Fire Protection Program
(Section 1R05.05.b.3)
Discussed None
Updated None
-2- Attachment
LIST OF DOCUMENTS REVIEWED
CALCULATIONS
Number Title Revision
12210-E-137 Electrical 480 Volts Continuous Load Cable Ampacity 0
Calculation
12210-E-169 Electrical Cable Sizing 0
E-200, Att. 3 4160 VAC Protective Device Coordination 1
G13.18.12.2-027 10 CFR 50 Appendix R Manual Action Time Frame 1
G13.18.12.2-106 Evaluation of Ability to Secure Reactor Feedwater During a
0
Main Control Room Fire
G13.18.12.4 RCIC Room Heatup Analysis 26
G13.18.12.4 RCIC Room Heatup with the Room Door Held Open 29
G13.18.13.2*84 Condenser Pressure During Loss of Circulating Water 0
G13.18.14.0*016 Number of SRV Cycles Expected for Isolation Event 1
G13.18.14.0*029 Reactor Level Response to a Fire in the Control Room 1
G13.18.2.6*034 Number of SRV Actuations from LSV Air Receiver Tanks 2
G13.18.3.6.07 Coordination Study of Appendix R and Class 1E Low Voltage 1
Protection Devices
G13.18.3.6.07 Safe Shutdown Common Enclosure Associated Circuit 1
Analysis
G13.18.3.6.12 10 CFR 50 Appendix R Analysis of Fire Area PT-1 0
DRAWINGS
Number Title Revision
0214.200-034-047 Schematic Diagram of Series DCF & DCM Controller For 8
Cummings Engine, Sht 1 of 2
0214.200-034-047 Schematic Diagram Of Series DCF & DCM Controller 8
For Cummings Engine, Sht 2 of 2
0242.562-082-319 Schematic and Wiring Diagram for FVR Starter G
0242.562-082-341 Composite Diagram for 1EHS-MCC2L F
0244.514-552-009 Schematic 40KVA Manual Transfer Switch 120VAC 1 A
phase 60HZ
-3- Attachment
Number Title Revision
12210-EB-45N-9 Ventilation & Cooling, Sections SH-13, Auxiliary Building 9
12210-EB-48A-7 Fire Protection & Plumbing Auxiliary Building EL 70-0 7
SH-1
12210-EB-82A-7 Fire Protection & Plumbing Control Building 7
12210-EE-18G-4 Wiring Diagram Fire and Smoke Detection Control 4
Building EL. 115-0 &116-0
12210-EE-34B Cable Tray Arrangement SH-6 6
12210-EE-34CJ Cable Tray Identification SH-4 4
12210-EE-34CL Cable Tray Identification SH-1 5
12210-EE-34DD-3 Cable Tray Identification, Turbine Bldg 3
12210-EE-34DD-4 Cable Tray Identification, Turbine Bldg 4
12210-EE-34EB-5 Cable Tray Identification Reactor Building 5
12210-EE-34FC Cable Tray Identification SH-1 5
12210-EE-34FF-4 Cable Tray Identification Reactor Building 4
12210-EE-34JG-4 Cable Tray Identification, Elect Tunnels & Norm SWGR 4
BLDG
12210-EE-34JK Cable Tray Identification SH-3 3
12210-EE-36BT-5 Wiring Diagram Elect. Pen. Terminal Cab., 1RCP*TCR 5
14A and 1RCP*TCA14
12210-EE-420M Seismic Conduit Inst. Plan El. 115-0 - 116-0 11
12210-EE-490J Seismic Conduit Inst. Plan El. 95-9 3
12210-EE-490Q Seismic Conduit Inst. Plan El. 95-9 6
12210-EE-80W-8 Communications Plan Standby Switchgear Area Control 8
Building
12210-EE-9BZ-5 Wiring Diagram Engine Driven Fire Pumps, Fire Pump 5
House
12210-ESK 6FPW02 Elementary Diagram, 480 V Control CKT Fire Protection 9
System Auxiliaries, RBS - Unit 1
12210-ESK 7FPW02 Elementary Diagram, 120 V Control CKT Engine Driven 11
Fire Pump Control , RBS - Unit 1
12210-ESK-3X Control Switch Contact Diagram 2
-4- Attachment
Number Title Revision
12210-ESK-7FPW03 Elementary Diagram, 120 V Control CKT Engine Driven 11
Fire Pump Control, RBS - Unit 1
828E239AA, Sht. 1 Elementary Diagram, Remote Shutdown System 20
84-51380-23 Sht. 3 Composite Diagram For 1EHS-MCC-2K A
84-51380-23 Sht. 6 Composite Diagram For 1EHS-MCC-2K A
84-51380-23-C97 Schematic and Wiring Diagram for FVR Starter O
851E225AA, Sh. 13 G.E. Elementary Diagram, Automatic Depressurization
System
944E115 SH-32 Connection Diagram Remote Shutdown VB 2
944E115 SH-34 Connection Diagram Remote Shutdown VB 2
944E115 SH-36 Connection Diagram Remote Shutdown VB 2
944E115 SH-37 Connection Diagram Remote Shutdown VB 8
944E115 SH-38 Connection Diagram Remote Shutdown VB 2
944E115 SH-39 Connection Diagram Remote Shutdown VB 2
944E115 SH-45 Connection Diagram Remote Shutdown VB 13
944E115 SH-46 Connection Diagram Remote Shutdown VB 10
CDB-VBN01A1, SH. 1 Power Distribution Panel Board Schedule Control Room 11
Appendix R Safe Shutdown Analysis Emergency
CE-001A, Sheet 1 4
Lighting, Control Building El. 98-0
Appendix R Safe Shutdown Analysis Emergency
CE-001B 6
Lighting, Control Building El. 116-0
Appendix R Safe Shutdown Analysis Emergency
CE-001C 4
Lighting, Control Building El. 136-0
Appendix R Safe Shutdown Analysis Emergency
CE-001F 6
Lighting, Diesel Generator Building El. 98-0
Appendix R Safe Shutdown Analysis Emergency
CE-001H, Sheet 1 1
Lighting, Auxiliary Building El. 95-0
Appendix R Safe Shutdown Analysis Emergency
CE-001J 5
Lighting, Auxiliary Building El. 114-0
Appendix R Safe Shutdown Analysis Emergency
CE-001K, Sheet 1 5
Lighting, Auxiliary Building El. 141-0
Appendix R Safe Shutdown Analysis Emergency
CE-001Q 3
Lighting, Standby Cooling Tower El. 118-0
-5- Attachment
Number Title Revision
Appendix R Safe Shutdown Analysis Emergency
CE-001U 2
Lighting, Turbine Building El. 67-6
Appendix R Safe Shutdown Analysis Emergency
CE-001V 2
Lighting, T-Tunnel El. 123-6
Appendix R Safe Shutdown Analysis Emergency
CE-001W 4
Lighting, Switchgear Building El. 98-0
DD-5617-I Fire Damper Schedule U
DD-5617-J Fire Damper, Vertical Mound and Horizontal Mount (CAT V
I)
EB-003AB Fire Area Boundaries Plant Plan View - Elevations 65- 5
0 to 90-0
EB-003AC Fire Area Boundaries Plant Plan View - Elevations 83- 6
0 to 106-0
EB-003AD Fire Area Boundaries Plant Plan View - Elevations 109- 9
0 to 148-0
EB-003AE Fire Area Boundaries Plant Plan View - Elevations 113- 4
0 to 186-3
EB-003BB Fire Protection Features Plant Plan View - Elevations 4
65-0 to 90-0
EB-003BC Fire Protection Features Plant Plan View - Elevations 5
83-0 to 106-0
EB-003BD Fire Protection Features Plant Plan View - Elevations 5
109-9 to 148-0
EB-003BE Fire Protection Features Plant Plan View - Elevations 5
113-0 to 186-3
EB-003M Fire Protection Arrangement SH-12 6
EB-003N Fire Protection Arrangement SH-13 9
EB-003P Fire Protection Arrangement SH-14 7
EB-045D Ventilation and Cooling, Plan El 95-9 SH 4, Auxiliary 10
Building
EB-082B Fire Protection & Plumbing Control Building 7
EB-048B Fire Protection & Plumbing Aux. Bldg El 95-9 & 114-0 7
SH-2
EE-001AA 480 V One Line Diagram, Standby Bus 1EJS*LDC 1A & 16
2A
-6- Attachment
Number Title Revision
EE-001AB 480 V One Line Diagram, Standby Bus 1EJS*LDC 1B & 17
2B
EE-001AC Start Up Electrical Distribution Chart 43
EE-001TA 480 V One Line Diagram, EHS-MCC2A & 2L, Auxiliary 19
Building
EE-001TE 480 V One Line Diagram, EHS-MCC2JA & 2K, Auxiliary 20
Building
EE-001ZD 125 VDC One Line Diagram ENB-MCC1 Auxiliary BLDG 6
EE-003KW Wiring Diagram, 1C61*PNLP001 Bay D, Control Building 7
EE-003LX Wiring Diagram, 1C61*PNLP001 Bay C, Control Building 7
EE-003LY Wiring Diagram, 1C61*PNLP001 Bay A and B, Control 14
Building
EE-007AT External Connection Diag. PGCC Termination Cabinet 8
1H13*P745 Bay B
EE-007D External Connection Diag. PGCC Termination Cabinet 10
1H13*P730 Bay E
EE-007DE External Connection Diagram PGCC Terminal Cabinet 10
H13*P710 Bay B
EE-007DQ External Connection Diagram PGCC Terminal Cabinet 10
H13*P713 Bay B
EE-007EB External Connection Diagram PGCC Terminal Cabinet 8
H13-P715 Bay B
EE-008BJ 4160V Wiring Diagram, Bus NNS-SWG2A 9
EE-009NB 480 V Wiring Diagram, 1EHS-MCC2B, Auxiliary Building 7
EE-009PA 480 V Wiring Diagram, 1EHS-MCC2J, Auxiliary Building 5
EE-009PE 480 V Wiring Diagram, 1EHS*MCC2KL, Auxiliary 7
Building
EE-009PG 480 V Wiring Diagram 1EHS*MCC2K Auxiliary Building 9
EE-009PU 480 V Wiring Diagram 1EHS*MCC14A Standby 12
Switchgear ROOM 1A
EE-009PUC Wiring Diagram Uninterrupted Power Supply ENB 302
EE-009SY 480 V Wiring Diagram, 1EHS*MCC2L, Auxiliary Building 11
EE-009SZ 480V Misc Wiring Diagram, 1EHS*MCC2L Auxiliary 17
Building
-7- Attachment
Number Title Revision
EE-009W 480 V Wiring Diagram, MISC Wiring Details Fire Pump 14
House
EE-018AE Wiring Diagram Fire and Smoke Detection Sys. 8
Auxiliary Building
EE-018F Wiring Diagram Fire and Smoke Detection Control 5
Building EL. 98-0
EE-018H Wiring Diagram Fire and Smoke Detection Control 8
Building EL. 136-1 5/8
EE-018Z Wiring Diagram Fire and Smoke Detection Control 3
Building EL. 136-1 5/8
EE-027A Arrangement Main Control Room 15
EE-80 Communication Plan Normal Switchgear Area & General 9
Notes
EE-80B-3 Communication Plan Normal Switchgear Building, Elev 3
123-6
EE-10C-5 125 VDC Wiring Diagram STBY 1ENB*MCC1 5
EE-27C-7 Arrangement Control BLDG Standby Switchgear Area 7
EE-32A Arrangement Duct line Plan & Details 10
EE-34FD Cable Tray Identification Auxiliary Building
EE-34KC Cable Tray identification, Aux Boiler & Water Treatment 3
Building
EE-36BD-5 Wiring Diagram Elect Pen. Termin CAB. 1RCP*TCR12A 5
- 1RCP*TCA12
EE-36BW Wiring Diagram Elect. Pen. Terminal Cabinet, 5
1RCP*TCR 15A and 1RCP*TCA15
EE-37 T-9 Arrangement, Sleeves, Inserts & Openings, Aux. 9
Building EL 114-0 & 141-0
EE-460AF Seismic Conduit Installation, Drywell Plan EL 141-0 8
Reactor Building
EE-460F Seismic Conduit Installation, Drywell Plan EL 95-9 10
Reactor Building
EE-490X Seismic Conduit Installation, Drywell Plan EL 114-0 9
Auxiliary Building
EE-55C Conduit Plan & Details, Fire Protection Pump House 7
-8- Attachment
Number Title Revision
EE-80AJ-5 Communication Plan Normal Switchgear Building & 5
Elect Tunnel Elev. 67-6
EE-80AK Communications Plan Tunnels Sh. 1 3
EE-80AL Communications Plan Tunnels Sh. 2 4
EE-80D Communications Plan Aux. BLDG Elev 70-0 & 95-9 5
EE-80U Communications Plan Main Control Room 6
EE-80V Communications Plan HVAC & Battery Rooms Control 5
Building
EE-8AZ 4160V Wiring Diagram, Standby Bus 1ENS-SWG1B 10
EE-9BJ 480 V Wiring Diagram, 1EJS-LDC2B, Auxiliary Building 8
EE-9MX 480 V Wiring Diagram, 1EHS-MCC2C, Auxiliary Building 9
EE-9RV 480V Misc Wiring Diagram, 1EHS*MCC16A &16B 6
Standby Cooling Tower Area
ESK-05SWP04 Elementary Diagram 4.16 kV SWGR Standby Service 27
Water Pump P2A, SH-1
ESK-06CCP09 Elementary Diagram, 480 V CONT CKT Reac. Plant 14
CMPNT. CLG WTR ISOL VALVE
ESK-06DTM25 Elementary Diagram, 480 V CONT CKT MNST LINE DR 11
ISOL MOVS
ESK-06EJS02 Elementary Diagram, 480V DC Switchgear Standby Bus 13
1B & 2B Supply ACB
ESK-06FPW01 Elementary Diagram, 480 V Control CKT Motor Driven 10
Fire Pump Control
ESK-06RHS06, Sh. 1 Elementary Diagram, 480 V Control CKT Residual Heat 12
Removal System
ESK-06RHS22 Elementary Diagram, 480V Control CKT, Residual Heat 11
Removal System
ESK-06RHS22, Sh. 1 Elementary Diagram, 480 V Control CKT Residual Heat 11
Removal System
ESK-07HVC25 Elementary Diagram, 120 V Control Circuit Remote 9
Shutdown Transfer Relays
ESK-11EJS02, Sh. 1 Elementary Diagram, 480V SWGR Standby Bus UNDV 11
TRIP RELAYS
-9- Attachment
Number Title Revision
ESK-11ICS06 Sh. 1 Elementary Diagram 125 VDC Control Circuit RCIC 7
Turbine Exhaust to Suppr Pool V
ESK-7HVN07, Sh. 1 Elementary Diagram, 120 V Control Circuit Remote 4
Shutdown Transfer Relays
GE-828E445AA, Elementary Diagram, Nuclear Steam Supply Shutoff
28
Sheet 13 System
GE-828E445AA, Elementary Diagram, Nuclear Steam Supply Shutoff
28
Sheet 14 System
GE-828E445AA, Elementary Diagram, Nuclear Steam Supply Shutoff
34
Sheet 7 System
Elementary Diagram, Reactor Protection System Motor
GE-944E981, Sheet 1 9
Generator Control System
PID-15-01A Engineering P&I Diagram, System 251, Fire Protection- 18
Water & Engine Pumps
PID-15-01B Engineering P&I Diagram, System 251, Fire Protection- 13
Water & Engine Pumps
PID-15-01C Engineering P&I Diagram, System 251, Fire Protection- 13
Water & Engine Pumps
PID-15-01D Engineering P&I Diagram, System 251, Fire Protection- 7
Water & Engine Pump
PID-15-01E Engineering P&I Diagram, System 251, Fire Protection- 11
Water & Engine Pump
PID-22-01E Engineering P&I Diagram, System 409, HVAC - 15
Auxiliary Building
PID-27-06A System 209 Reactor Core Isolation Cooling 43
PID-27-07A Engineering P&I Diagram, System 204, Residual Heat 36
Removal - LPCI
PID-27-07B Engineering P&I Diagram, System 204, Residual Heat 41
Removal - LPCI
PID-27-07C Engineering P&I Diagram, System 204, Residual Heat 25
Removal - LPCI
TLD-FWP-015 Test Loop Diagram, Motor Fire Water Pump Discharge 0
FWP-PS115
- 10 - Attachment
ENGINEERING REPORTS (ER)
Number Title Revision
98-0296 Determine the Appropriate Battery Replacement
0
Frequency for the Appendix R Emergency Lights
RB-2001-0136-000 Document the Basis for the Scope and Frequency of 0
Fire Protection Testing
RB-2003-0711-001 Revising Post fire Safe Shutdown Operator Manual
0
Action Evaluations Following Release of RIS 2006-10
RB-2004-0140-000 Evaluate the Impact on the Post Fire Safe Shutdown 0
Analysis if Automatic Functions are NOT Lost Due to a
Fire
RB-2004-0275-000 Summarize all RBS NFPA Code Deviations 0
SD171 SD112 SD97 SD82 SD86
WORK ORDERS
Number Title Revision/Date
51642307 FPW-Batt1A Replace Bank 6/2/2008
00192017 FPW-Batt1B Replace Bank 6/25/2009
51522151 Diesel Fire Pump Battery 18 month Surveillance 1/26/2009
52226058 Diesel Fire Pump Battery Quarterly Surveillance 3/09/2010
52249598 Diesel Fire Pump Battery Quarterly Surveillance 3/31/2010
00218207 RBS EP Remote Radio: Perform Annual Maintenance 2/01/2010
00130765 EHS-MCC2J Breaker 1CB AOP-0031 Attachment 6 1
Needs To Be Verified
160308 FPW-P4 Annual Maintenance [3 Year] 0
- 11 - Attachment
ENGINEERING CHANGES
Number Title Revision/Date
EC12206 Child to EC-8684 Modify Div 1DG Controls, Not Bypass 12/1/2009
Trips, LOP-Only Start Ref. CR-RBS-2007-2102 LT-
ACE, Reportable Regulatory Issue Non Control Room
Work
EC1933 Install Transfer Switches that Allow Division I to Supply 10/16/2009
Motive Power and Control Power to Valve E51-
MOVF063 following evacuation of the Main Control
Room due to a fire
EC21964 Restore Breaker EHS-MCC2J-1CB to Original 0
Configuration
EC2570 Engineering Change Provide An Alternate Power 1/5/2010
Source for E51-MOVF063 During a main Control Room
Fire Div 1 & Non-Safety Pre Outage Phase
EC2571 Provide An Alternate Power Source for E51-MOVF063 10/15/2009
During a main Control Room Fire Div II Outage Phase
EC8684 Modify Div 1-2 DG Controls, Not Bypass Trips, LOP- 12/10/2009
Only Start; Ref. CR-RBS-2007-2102 LT-ACE,
Reportable Regulatory Issue
ECR1784 Engineering Change Request - Revise Division 1-2 DG 8/1/2007
Controls to Leave Overheat Trips Active After LOP-Only
Auto-Start
ECR6274 Engineering Change Request - Revise Division 1-2 DG 11/18/2008
Controls to Leave Overheat Trips Active After LOP-Only
Auto-Start
- 12 - Attachment
CONDITION REPORTS (CR)
RBS-2001-00613 RBS-2010-01410 RBS-2010-01578* RBS-2010-01825*
RBS-2006-03776 RBS-2010-01529* RBS-2010-01589* RBS-2010-01828*
RBS-2008-03475 RBS-2010-01537* RBS-2010-01592* RBS-2010-01831*
RBS-2009-05823 RBS-2010-01538* RBS-2010-01594* RBS-2010-01846*
RBS-2009-05843 RBS-2010-01540* RBS-2010-01599* RBS-2010-01851*
RBS-2009-05882 RBS-2010-01546* RBS-2010-01750* RBS-2010-01955
RBS-2010-00697 RBS-2010-01552* RBS-2010-01766* LAR-2010-00022*
RBS-2010-01087 RBS-2010-01557* RBS-2010-01775* LO-NOE-2009-00516
RBS-2010-01192* RBS-2010-01559* RBS-2010-01783* LO-LAR-2010-00120
RBS-2010-01234* RBS-2010-01566* RBS-2010-01808*
RBS-2010-01405 RBS-2010-01567* RBS-2010-01821*
- Issued as a result of inspection activities.
PREVENTIVE MAINTENANCE TASKS
WM-105-00 PMRQ 19005-01 PMRQ 19005-04
WM-105-04 PMRQ 19005-03 PMRQ 19005-05
- 13 - Attachment
PROCEDURES
Number Title Revision/Date
AB-095-506 Pre-Fire Strategies - HPCS Pump Room, Fire Area 4
AB-2/Z-1
AB-095-517 Pre-Fire Strategies - HPCS Piping Area, Fire Area 4
AB-2/Z-2
AOP-0031 Shutdown From Outside the Main Control Room 307
AOP-0052 Fire Outside the Main Control Room in Areas 18
Containing Safety Related Equipment
CB-116-127 Pre-Fire Strategies - HVAC Room Fire Area C-17 3
CB-136-138 Pre-Fire Strategies - Control Room Fire Area C-25 4
CB-98-117 Pre-Fire Strategies - Standby Switchgear 1B Room 2
Fire Area C-14
CB-98-118 Pre-Fire Strategies - Standby Switchgear 1A Room 2
Fire Area C-15
Preparation of Fire Protection Engineering
Evaluations
EN-DC-330 Fire Protection Program 0
EN-LI-102 Corrective Action Process 14
EN-OP-104 Operability Determination Process 4
EN-TQ-125, Fire Brigade Drills Scenario 0
Attachment 9.1
FPP-0010 Fire Fighting Procedure 12
FPP-0015 Post Fire Ventilation/Smoke Management 0
FPP-0070 Duties of Fire Watch 11
FPP-0100 Fire Protection System Impairment 10
FPP-0101 Fire Suppression System Inspection 11
OSP-0601 Remote Shutdown System Control Circuit Operability 1
Test (Switches 43-1EGAN05, 43-1EJSA01,
43-1ENSC04, 43A-1ENSA01, 43B-1ENSA03,
43C-1ENSA09, 43D-1ENSC04, 43E-1ENSC01,
43F-1ENSA01, and 43G-1ENSA03)
OSP-0602 Remote Shutdown System Control Circuit Operability 0
Test (Switches 43-1HVCN30, 43-1HVCN31,
43-1HVCN32, 43-1HVKA01)
- 14 - Attachment
Number Title Revision/Date
PT-070-427 Pre-Fire Strategies- E-Tunnel West and F-Tunnel 3
Fire Area PT-1
PT-070-428 Pre-Fire Strategies- F-Tunnel Electrical Fire Area 3
PT-1
PT-070-429 Pre-Fire Strategies- G-Tunnel Fire Area PT-1 3
RBNP-038 Site Fire Protection Program 6B
SOP-0027 Remote Shutdown System (#200) 302
SOP-0027, Control Board Lineup - Remote Shutdown (Standby) 302
Attachment 2
STP-200-0605 Remote Shutdown System Control Circuit Operability 303
Test (Switches S1, S6, S7, S8, S9, and S12)
STP-200-0606 Remote Shutdown System Control Circuit Operability 303
Test (Switches S1, S2, S3, S4, S5, and S11)
STP-200-0607 Division I remote Shutdown System Control Circuit 302
Operability Test (Switch S10)
STP-200-0613 Remote Shutdown System Control Circuit Operability 1
Test (Switches 43-1SWPA45, 43-1SWPA46)
STP-251-3201 Fire Hose Station Visual Inspection 11
STP-251-3300 Surveillance Test Procedure for Diesel Fire Pump 14
Battery Quarterly Surveillance
TPP-7-021 Fire Protection Training and Qualifications 11
B.5.b COMMITMENTS
P-16812 P-16818 P-16820
P-16821 A-16837 P-16881
COMPONENTS REVIEWED DURING CIRCUIT ANALYSIS
Component ID Description
1CCP*MOV15B Containment Return Inboard Isolation Valve
1B21*F0501D Safety Relief Valve
1B21*MOVF016 Main Steam Line DR Inboard Isolation Valve
1B21*MOVF019 Main Steam Line DR Inboard Isolation Valve
- 15 - Attachment
Component ID Description
1B21*PTN068A Reactor Vessel Pressure Transmitter
1B21*PTN068B Reactor Vessel Pressure Transmitter
1B21*PTN068E Reactor Vessel Pressure Transmitter
1B21*PTN068F Reactor Vessel Pressure Transmitter
1E12*FTN052B RHR B Discharge Flow Transmitter
1E12*MOVF004B RHR Pump B Suppression Pool Suction Valve
1E12*MOVF006B RHR B Shutdown Cooling Suction
1E12*MOVF006A RHR A Shutdown Cooling Suction
1E12*MOVF009 RHR Shutdown Cooling Inboard Isolation Valve
1E12*MOVF008 RHR Shutdown Cooling Outboard Isolation Valve
1E12*MOVF011B RHR B Discharge to Suppression Pool
1E12*MOVF024B RHR B Test Return/HX Discharge to Suppression Pool
1E12*MOVF040 RHR Discharge to Radwaste Inboard Isolation valve
1E12*MOVF042B RHR B Injection Valve
1E12*MOVF064B RHR B Min Flow Line Isolation Valve
1E12*VF082 RHR B/C Discharge Line Fill Pump Suction
1E12*PC003 RHR B/C Line Fill Pump
1SWP*P2B Standby Service Water Pump
1SWP*MOV40B Standby Service Water Pump 2b Discharge
1SWP*MOV505A Standby Service Water Division I / Division II Crossover Valve
1SWP*MOV027A Control Building Chilled Water pump SWP*P3A
1SWP*P2D Standby Service Water Pump motor
1EHS*MCC2J 480 Volts Auxiliary Building Motor Control Center
1EHS*MCC2K 480 Volts Auxiliary Building Motor Control Center
1SWP*MOV73B 1HVR*UC5 Service Water Supply Valve
- 16 - Attachment
MISCELLANEOUS DOCUMENTS
Number Title Revision/Date
Fire Area C-15 Summary Table, Division I
Standby Switchgear Room (EL. 98)
Fire Area C-17 Summary Table, Control
Room Ventilation
Fire Area AB-2 Summary Table, HPCS &
Fire Area PT-1 Summary Table, Piping
Tunnel
Snapshot Assessment on B.5.b Strategy 3/31/2010
Implementation
PDMS Cable Routing Sheets for:
1E51*MOVF068
1ICSNRC016
1ICSNRC017
1ICSNRC022
1ICSNCK618
1ICSNCK619
1ICSNRK620
Addendum 2 to 229.180 Specification for Floor and Wall Sleeve 2
Seals
Branch Technical Guidelines for Fire Protection for Nuclear 8/23/1976
Position (BTP) APCSB Power Plants, docketed prior to July 1,
9.5-1 & Appendix A 1976
Design Change Notice Change Cable Designation from 12/1/1995
95-1100 1RHSNRC517 to 1RHSNRC527.
Design Criterion No. Specification for Procurement and Storage 1
228.412 of Thermo-Lag Fire Barrier Materials
Design Criterion No. Specification for Floor and Wall Sleeve 2
229.180 Seals
Design Criterion No. Post Fire Safe Shutdown Analysis 4
240.201
Design Criterion No. 10CFR50 APPENDIX R, Post fire Safe 4
240.201A, Appendix C Shutdown Equipment List and Logic
Diagram
Design Criterion No. Circuit Analysis for RBS 10CFR50 Appendix 4
240.201A, Appendix E R Safe Shutdown Equipment List
Components
- 17 - Attachment
Number Title Revision/Date
EDCR C-24501 Engineering Design and Coordination
Report Communication Equipment Hold
Down
EDS-EE-006 Installation, Modification and Maintenance of 3
Thermo-Lag Fire Barrier Systems
EEAR-93-E0059 Communication Cat. I, II & III Engineering 11/11/1993
Evaluation and Assistance Request
Final Safety Analysis Fire Hazards Analysis 10
Report, Appendix 9A
Final Safety Analysis Fire Protection Program Comparison With 15
Report, Appendix 9B Appendix R to 10 CFR 50
Letter Response Providing Information Regarding
Implementation Details for the Phase 2 and 1/11/2007
3 Mitigation Strategies
Letter Supplementary Response Regarding
Implementation Details for the Phase 2 and 5/14/2007
3 Mitigation Strategies
LER 07-003-00 Licensee Event Report - Unanalyzed
Condition of Emergency Diesel Generator in 7/19/2007
Post-Fire Safe Shutdown Scenario
NUREG-0800 Standard Review Plan, Section 9.5.1, Fire
1981
Protection Program
Procedure Action
AOP-0031R305PR-306
Request
Procedure Action
AOP-00301R307CN-A
Request
Regulatory Guide 1.68.2 Initial Startup Test Program to Demonstrate 2
Remote Shutdown Capability for
Water-Cooled Nuclear Power Plants
Specification No. Specification for Standby Diesel Generator 3
244.700 Systems
System Training Manual
Remote Shutdown System 2/2/2009
R-STM-0200.04
System Training Manual Fire Protection & Detection 6
R-STM-0250
System Training Manual Reactor Core Isolation Cooling (RCIC) 6
R-STM-209 System
- 18 - Attachment
Number Title Revision/Date
System Training Manual Standby Diesel Generators 8
R-STM-309S
Technical Requirements Fire Detection Instrumentation 5
Manual Section 3.3.7.4
Technical Requirements Fire Suppression Systems 122
Manual Section 3.7.9.1
Technical Requirements Spray and/or Sprinkler Systems 5
Manual Section 3.7.9.2
Technical Requirements Halon Systems 5
Manual Section 3.7.9.3
Technical Requirements Hose Stations 5
Manual Section 3.7.9.4
Technical Requirements Fire-Rated Assemblies 5
Manual Section 3.7.9.6
VTD-C742-0112 Cummins Service Bulletin For Battery and 0
Cable Specification (Pub. #3379024-011)
VTD-G080-1264 General Electric Control and Instrument 0
Switches
VTD-G080-1476 General Electric Type SB-9 Control 0
Switches Renewal Parts
Vendor Technical Manual for Exide
VTM-E355-0002 07/09/1997
Corrective Action 1 to White Paper - Remote Shutdown Panel
LO-LAR-2010-00120 Transfer Switch Reliability
- 19 - Attachment