Information Notice 2007-07, Potential Failure of All Control Rod Groups to Insert in a Boiling Water Reactor (BWR) Due to a Fire: Difference between revisions

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{{#Wiki_filter:UNITED STATES
{{#Wiki_filter:UNITED STATES


NUCLEAR REGULATORY COMMISSION
===NUCLEAR REGULATORY COMMISSION===
OFFICE OF NUCLEAR REACTOR REGULATION


OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C. 20555-0001


WASHINGTON, D.C. 20555-0001 February 15, 2007 NRC INFORMATION NOTICE 2007-07:                 POTENTIAL FAILURE OF ALL CONTROL ROD
===February 15, 2007===
NRC INFORMATION NOTICE 2007-07:


===POTENTIAL FAILURE OF ALL CONTROL ROD===
GROUPS TO INSERT IN A BOILING WATER
GROUPS TO INSERT IN A BOILING WATER


REACTOR DUE TO A FIRE
===REACTOR DUE TO A FIRE===


==ADDRESSEES==
==ADDRESSEES==
Line 40: Line 43:
possibility of fire-induced hot shorts preventing all control rod groups from inserting when
possibility of fire-induced hot shorts preventing all control rod groups from inserting when


required due to a postulated fire. The NRC expects that recipients will review the information
required due to a postulated fire. The NRC expects that recipients will review the information


for applicability to their facilities and consider actions, as appropriate, to avoid similar problems.
for applicability to their facilities and consider actions, as appropriate, to avoid similar problems.
Line 53: Line 56:
from inserting when the operator places the reactor mode selector switch in the SHUTDOWN
from inserting when the operator places the reactor mode selector switch in the SHUTDOWN


position in the control room. The inspection results are summarized in Inspection Report
position in the control room. The inspection results are summarized in Inspection Report


05000397/200608, dated August 18, 2006 (Agencywide Documents Access Management
05000397/200608, dated August 18, 2006 (Agencywide Documents Access Management
Line 63: Line 66:
shutdown systems necessary to accomplish the reactivity control shutdown function and are
shutdown systems necessary to accomplish the reactivity control shutdown function and are


credited in the post-fire safe shutdown procedures developed by this licensee. However, the
credited in the post-fire safe shutdown procedures developed by this licensee. However, the


potential for fire to cause a loss of this required shutdown function has not been evaluated. The
potential for fire to cause a loss of this required shutdown function has not been evaluated. The


Final Safety Analysis Report states Fail safe circuits (electrical divisions 4, 5, 6, and 7) are
Final Safety Analysis Report states Fail safe circuits (electrical divisions 4, 5, 6, and 7) are


designed to fail in a safe manner if subjected to fire damage. For example, reactor
designed to fail in a safe manner if subjected to fire damage. For example, reactor


scram, once initiated, cannot be overridden as a consequence of fire. The licensees analysis
scram, once initiated, cannot be overridden as a consequence of fire. The licensees analysis


was based on the assumption that the operator would initiate and confirm shutdown before
was based on the assumption that the operator would initiate and confirm shutdown before
Line 81: Line 84:
Fires can potentially damage circuits prior to the decision to initiate a plant shutdown using
Fires can potentially damage circuits prior to the decision to initiate a plant shutdown using


alternative shutdown capability and control room evacuation. At CGS, a reactor shutdown
alternative shutdown capability and control room evacuation. At CGS, a reactor shutdown


would not be initiated unless:
would not be initiated unless:
 
* there are indications that the fire threatens safe operation of the plant,  
* there are indications that the fire threatens safe operation of the plant,
 
* the operator observes degraded equipment performance, or
* the operator observes degraded equipment performance, or


Line 99: Line 100:
At CGS, the licensee revised plant procedures to ensure that the RPS would be de-energized
At CGS, the licensee revised plant procedures to ensure that the RPS would be de-energized


prior to initiating depressurization. This action would ensure that all control rods insert into the
prior to initiating depressurization. This action would ensure that all control rods insert into the


reactor prior to opening the safety/relief valves (SRVs) and starting low pressure injection. Due
reactor prior to opening the safety/relief valves (SRVs) and starting low pressure injection. Due


to various power supplies that may be in the cabinet, the response by CGS may not be effective
to various power supplies that may be in the cabinet, the response by CGS may not be effective
Line 112: Line 113:
nuclear power plant has a fire protection plan that satisfies General Design Criterion (GDC) 3 of
nuclear power plant has a fire protection plan that satisfies General Design Criterion (GDC) 3 of


Appendix A of Part 50. Criterion 3 specifies that Structures, systems, and components
Appendix A of Part 50. Criterion 3 specifies that Structures, systems, and components


important to safety shall be designed and located to minimize, consistent with other safety
important to safety shall be designed and located to minimize, consistent with other safety
Line 122: Line 123:
with 10 CFR 50.48(a) and imposes a backfit requirement for plants licensed to operate prior to
with 10 CFR 50.48(a) and imposes a backfit requirement for plants licensed to operate prior to


January 1, 1979, to comply with section III.G of Appendix R. For plants licensed to operate
January 1, 1979, to comply with section III.G of Appendix R. For plants licensed to operate


after January 1, 1979, similar requirements were incorporated into NUREG-0800, Standard
after January 1, 1979, similar requirements were incorporated into NUREG-0800, Standard
Line 128: Line 129:
Review Plan, Section 9-5.1, Fire Protection Program, and were incorporated into the licensee
Review Plan, Section 9-5.1, Fire Protection Program, and were incorporated into the licensee


programs during the licensing process for these plants. These licensees then had planned
programs during the licensing process for these plants. These licensees then had planned


implementation as a condition of the Operating License for the facility. To satisfy GDC 3, each
implementation as a condition of the Operating License for the facility. To satisfy GDC 3, each


licensee must have fire protection features that are capable of limiting fire damage so that:
licensee must have fire protection features that are capable of limiting fire damage so that:
Line 142: Line 143:


==DISCUSSION==
==DISCUSSION==
The control rods are divided into four control rod groups. The scram of each control rod group
The control rods are divided into four control rod groups. The scram of each control rod group


is controlled by separate circuits within the RPS system. The system design has two trip logic
is controlled by separate circuits within the RPS system. The system design has two trip logic


channels functioning in a 1-out-of-2 twice arrangement. This design requires that one trip logic
channels functioning in a 1-out-of-2 twice arrangement. This design requires that one trip logic


be satisfied in both trip logic channels before the control rod group will scram. The RPS circuits
be satisfied in both trip logic channels before the control rod group will scram. The RPS circuits


are a fail-safe design in that the circuits are normally energized, and the loss of power will
are a fail-safe design in that the circuits are normally energized, and the loss of power will


initiate a scram. Also, the RPS scram circuits are routed separately from other circuits to
initiate a scram. Also, the RPS scram circuits are routed separately from other circuits to


prevent any possibility for interaction.
prevent any possibility for interaction.
Line 160: Line 161:
mode switch in the control room is placed in the SHUTDOWN position and circuit damage does
mode switch in the control room is placed in the SHUTDOWN position and circuit damage does


not prevent the scram. For fires in the control room, a hot short between conductors to the
not prevent the scram. For fires in the control room, a hot short between conductors to the


mode switch could keep the associated trip channel logic energized. Two hot shorts without the
mode switch could keep the associated trip channel logic energized. Two hot shorts without the


occurrence of an open circuit or short to ground have the potential of affecting the scram
occurrence of an open circuit or short to ground have the potential of affecting the scram


function. A hot short as described above would have to be present in both of the trip channel
function. A hot short as described above would have to be present in both of the trip channel


logic circuits associated with the same trip channel. This would keep the trip channel energized
logic circuits associated with the same trip channel. This would keep the trip channel energized


so that half of the 1-out-of-2 twice logic would not be satisfied. The result would be that the
so that half of the 1-out-of-2 twice logic would not be satisfied. The result would be that the


associated rod group would not scram. The other three rod groups would not be affected and
associated rod group would not scram. The other three rod groups would not be affected and


would scram as expected.
would scram as expected.


CGS is a BWR-5 design. This plant design has the Power Generation Control Complex which
CGS is a BWR-5 design. This plant design has the Power Generation Control Complex which


provides cable separation in the control room subflooring when the wires exit the control
provides cable separation in the control room subflooring when the wires exit the control


cabinets. In part, due to this cable separation, the NRC determined this issue at CGS to be of
cabinets. In part, due to this cable separation, the NRC determined this issue at CGS to be of


very low safety significance.
very low safety significance.
Line 186: Line 187:
To accomplish alternative safe shutdown, for a fire in the control room, most BWRs rely upon
To accomplish alternative safe shutdown, for a fire in the control room, most BWRs rely upon


using three to six SRVs to depressurize the reactor vessel. The vessel is then reflooded with a
using three to six SRVs to depressurize the reactor vessel. The vessel is then reflooded with a


low pressure coolant injection system using a residual heat removal pump in the low pressure
low pressure coolant injection system using a residual heat removal pump in the low pressure


coolant injection mode. By design, the negative reactivity, added by all four rod groups during a
coolant injection mode. By design, the negative reactivity, added by all four rod groups during a


scram, provides adequate shutdown margin to offset the positive void and temperature
scram, provides adequate shutdown margin to offset the positive void and temperature


reactivity would have been added to the vessel. A typical BWR reactor has about 180 control
reactivity would have been added to the vessel. A typical BWR reactor has about 180 control


rods. One of the four rod groups remaining in the fully out position would place the reactor
rods. One of the four rod groups remaining in the fully out position would place the reactor


outside of the design basis.
outside of the design basis.
Line 210: Line 211:
S
S


This information notice requires no specific action or written response. Please direct any
This information notice requires no specific action or written response. Please direct any


questions about this matter to the technical contacts listed below or the appropriate Office of
questions about this matter to the technical contacts listed below or the appropriate Office of
Line 217: Line 218:


/RA/
/RA/
                                                    Michael Case, Director


===Michael Case, Director===
Division of Policy and Rulemaking
Division of Policy and Rulemaking


Office of Nuclear Reactor Regulation
===Office of Nuclear Reactor Regulation===
Technical Contacts: John M. Mateychick, R IV
 
===Edward McCann, NRR===
817-276-6560
301-415-1218 E-mail:  JMM3@nrc.gov
 
E-mail:  EVM@nrc.gov
 
===Phillip M. Qualls, NRR===
301-415-1849 E-mail:  PMQ@nrc.gov
 
ML063540138 RIV/SRI
 
AFPB/FPE
 
AFPB/FPE


Technical Contacts:  John M. Mateychick, R IV                Edward McCann, NRR
AFPB/BC


817-276-6560                            301-415-1218 E-mail: JMM3@nrc.gov                    E-mail: EVM@nrc.gov
TECH EDITOR


Phillip M. Qualls, NRR
PGCB/PM


301-415-1849 E-mail: PMQ@nrc.gov
PGCB/LA


ML063540138 RIV/SRI      AFPB/FPE  AFPB/FPE    AFPB/BC      TECH EDITOR  PGCB/PM    PGCB/LA      PGCB/BC     DPR/DD
PGCB/BC


JMateychick  PQualls  EMcCann      SWeerakkody  CClark        QNguyen    CHawes -    CJackson    MCase
DPR/DD


JMateychick
PQualls
EMcCann
SWeerakkody
CClark
QNguyen
CHawes  -
CMH
CMH


12/20/06     2/05/2007 2/02/2007   2/08/2007   02/12/2007   2/01/2007 2/14/2007     02/15/2007 02/15/2007}}
CJackson
 
MCase
 
12/20/06
2/05/2007
2/02/2007
2/08/2007
02/12/2007
2/01/2007
2/14/2007
02/15/2007
02/15/2007}}


{{Information notice-Nav}}
{{Information notice-Nav}}

Latest revision as of 04:40, 15 January 2025

Potential Failure of All Control Rod Groups to Insert in a Boiling Water Reactor (BWR) Due to a Fire
ML063540138
Person / Time
Issue date: 02/15/2007
From: Michael Case
NRC/NRR/ADRA/DPR
To:
References
IN-07-007
Download: ML063540138 (5)
v • d • e


UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON, D.C. 20555-0001

February 15, 2007

NRC INFORMATION NOTICE 2007-07:

POTENTIAL FAILURE OF ALL CONTROL ROD

GROUPS TO INSERT IN A BOILING WATER

REACTOR DUE TO A FIRE

ADDRESSEES

All holders of operating licenses for boiling water reactors (BWRs), except those who have

permanently ceased operations and have certified that fuel has been permanently removed

from the reactor.

PURPOSE

The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice to inform

addressees of an issue discovered at Columbia Generating Station (CGS) regarding the

possibility of fire-induced hot shorts preventing all control rod groups from inserting when

required due to a postulated fire. The NRC expects that recipients will review the information

for applicability to their facilities and consider actions, as appropriate, to avoid similar problems.

However, suggestions contained in this information notice are not NRC requirements; therefore, no specific action or written response is required.

DESCRIPTION OF CIRCUMSTANCES

During an inspection at CGS completed on July 13, 2006, the NRC inspectors identified that

two hot shorts, caused by a postulated fire, could prevent one of the four groups of control rods

from inserting when the operator places the reactor mode selector switch in the SHUTDOWN

position in the control room. The inspection results are summarized in Inspection Report 05000397/200608, dated August 18, 2006 (Agencywide Documents Access Management

System, Accession No. ML062300334).

The reactor protection and control rod drive systems are identified as part of the minimum safe

shutdown systems necessary to accomplish the reactivity control shutdown function and are

credited in the post-fire safe shutdown procedures developed by this licensee. However, the

potential for fire to cause a loss of this required shutdown function has not been evaluated. The

Final Safety Analysis Report states Fail safe circuits (electrical divisions 4, 5, 6, and 7) are

designed to fail in a safe manner if subjected to fire damage. For example, reactor

scram, once initiated, cannot be overridden as a consequence of fire. The licensees analysis

was based on the assumption that the operator would initiate and confirm shutdown before

control circuiting is damaged; therefore, evaluation of the effects of fire damage to the reactor

protection (RPS) and control rod drive systems was not performed.

Fires can potentially damage circuits prior to the decision to initiate a plant shutdown using

alternative shutdown capability and control room evacuation. At CGS, a reactor shutdown

would not be initiated unless:

  • there are indications that the fire threatens safe operation of the plant,
  • the operator observes degraded equipment performance, or
  • there is visible damage to vital plant equipment or cabling.

Since a reactor scram would possibly not be initiated until the fire had damaged vital plant

equipment, the NRC staff noted the possibility that hot shorts could occur prior to making the

decision to scram the unit.

At CGS, the licensee revised plant procedures to ensure that the RPS would be de-energized

prior to initiating depressurization. This action would ensure that all control rods insert into the

reactor prior to opening the safety/relief valves (SRVs) and starting low pressure injection. Due

to various power supplies that may be in the cabinet, the response by CGS may not be effective

for other BWR designs.

BACKGROUND

Title 10 of the Code of Federal Regulations (CFR) Part 50.48(a) requires that each operating

nuclear power plant has a fire protection plan that satisfies General Design Criterion (GDC) 3 of

Appendix A of Part 50. Criterion 3 specifies that Structures, systems, and components

important to safety shall be designed and located to minimize, consistent with other safety

requirements, the probability and effect of fires and explosions.

Part 58.48(b) of Title 10 of the Code of Federal Regulations identifies some methods to comply

with 10 CFR 50.48(a) and imposes a backfit requirement for plants licensed to operate prior to

January 1, 1979, to comply with section III.G of Appendix R. For plants licensed to operate

after January 1, 1979, similar requirements were incorporated into NUREG-0800, Standard

Review Plan, Section 9-5.1, Fire Protection Program, and were incorporated into the licensee

programs during the licensing process for these plants. These licensees then had planned

implementation as a condition of the Operating License for the facility. To satisfy GDC 3, each

licensee must have fire protection features that are capable of limiting fire damage so that:

a. 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

b. Systems necessary to achieve and maintain cold shutdown from either the control room or

emergency control station(s) can be repaired within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

DISCUSSION

The control rods are divided into four control rod groups. The scram of each control rod group

is controlled by separate circuits within the RPS system. The system design has two trip logic

channels functioning in a 1-out-of-2 twice arrangement. This design requires that one trip logic

be satisfied in both trip logic channels before the control rod group will scram. The RPS circuits

are a fail-safe design in that the circuits are normally energized, and the loss of power will

initiate a scram. Also, the RPS scram circuits are routed separately from other circuits to

prevent any possibility for interaction.

For CGS, in all fires other than a control room fire, the circuits will be de-energized when the

mode switch in the control room is placed in the SHUTDOWN position and circuit damage does

not prevent the scram. For fires in the control room, a hot short between conductors to the

mode switch could keep the associated trip channel logic energized. Two hot shorts without the

occurrence of an open circuit or short to ground have the potential of affecting the scram

function. A hot short as described above would have to be present in both of the trip channel

logic circuits associated with the same trip channel. This would keep the trip channel energized

so that half of the 1-out-of-2 twice logic would not be satisfied. The result would be that the

associated rod group would not scram. The other three rod groups would not be affected and

would scram as expected.

CGS is a BWR-5 design. This plant design has the Power Generation Control Complex which

provides cable separation in the control room subflooring when the wires exit the control

cabinets. In part, due to this cable separation, the NRC determined this issue at CGS to be of

very low safety significance.

To accomplish alternative safe shutdown, for a fire in the control room, most BWRs rely upon

using three to six SRVs to depressurize the reactor vessel. The vessel is then reflooded with a

low pressure coolant injection system using a residual heat removal pump in the low pressure

coolant injection mode. By design, the negative reactivity, added by all four rod groups during a

scram, provides adequate shutdown margin to offset the positive void and temperature

reactivity would have been added to the vessel. A typical BWR reactor has about 180 control

rods. One of the four rod groups remaining in the fully out position would place the reactor

outside of the design basis.

RELEVANT GENERIC COMMUNICATIONS

Regulatory Issue Summary 2004-03, Revision 1, Risk-Informed Approach for Post-Fire

Safe-Shutdown Circuit Inspections discusses the probability of two hot shorts occurring for a

postulated fire.

CONTACT

S

This information notice requires no specific action or written response. Please direct any

questions about this matter to the technical contacts listed below or the appropriate Office of

Nuclear Reactor Regulation (NRR) project manager.

/RA/

Michael Case, Director

Division of Policy and Rulemaking

Office of Nuclear Reactor Regulation

Technical Contacts: John M. Mateychick, R IV

Edward McCann, NRR

817-276-6560

301-415-1218 E-mail: JMM3@nrc.gov

E-mail: EVM@nrc.gov

Phillip M. Qualls, NRR

301-415-1849 E-mail: PMQ@nrc.gov

ML063540138 RIV/SRI

AFPB/FPE

AFPB/FPE

AFPB/BC

TECH EDITOR

PGCB/PM

PGCB/LA

PGCB/BC

DPR/DD

JMateychick

PQualls

EMcCann

SWeerakkody

CClark

QNguyen

CHawes -

CMH

CJackson

MCase

12/20/06

2/05/2007

2/02/2007

2/08/2007

02/12/2007

2/01/2007

2/14/2007

02/15/2007

02/15/2007


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