Information Notice 2012-12, HVAC Design Control Issues Challenge Safety System Function: Difference between revisions

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{{#Wiki_filter:ML12115A012 UNITED STATES
{{#Wiki_filter:UNITED STATES


NUCLEAR REGULATORY COMMISSION
NUCLEAR REGULATORY COMMISSION
Line 22: Line 22:
OFFICE OF NEW REACTORS
OFFICE OF NEW REACTORS


WASHINGTON, DC
WASHINGTON, DC 20555-0001 July 24, 2012 NRC INFORMATION NOTICE 2012-12:                HVAC DESIGN CONTROL ISSUES CHALLENGE


20555-0001  July 24, 2012 NRC INFORMATION NOTICE 201
SAFETY SYSTEM FUNCTION
2-12: HVAC DESIGN CONTROL ISSUES CHALLENGE SAFETY SYSTEM FUNCTION


==ADDRESSEES==
==ADDRESSEES==
All holders of an operating license or construction permit for a nuclear power reactor or a non-power (research or test) reactor issued under Title 10 of the Code of Federal Regulations
All holders of an operating license or construction permit for a nuclear power reactor or a


(10 CFR) Part 50, "Domestic Licensing of Production and Utilization Facilities," except those who have permanently ceased operations and have certified that fuel has been permanently removed from the reactor vessel.
non-power (research or test) reactor issued under Title 10 of the Code of Federal Regulations


All holders of and applicants for
(10 CFR) Part 50, Domestic Licensing of Production and Utilization Facilities, except those


a power reactor early site permit, combined license
who have permanently ceased operations and have certified that fuel has been permanently


, standard design certification, standard design approval, or manufacturing license under 10 CFR Part 52, "Licenses, Certifications, and Approvals for Nuclear Power Plants."
removed from the reactor vessel.
 
All holders of and applicants for a power reactor early site permit, combined license, standard
 
design certification, standard design approval, or manufacturing license under 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants.


==PURPOSE==
==PURPOSE==
The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice (IN) to inform addressees about certain events involving heating, ventilation
The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice (IN) to inform
 
addressees about certain events involving heating, ventilation, and air conditioning (HVAC)
system design control issues that challenged, or potentially challenged, safety system functions.


, and air conditioning (HVAC) system design control issues that challenged
The NRC expects recipients to review the information contained within for applicability to their


, or potential
facilities and consider actions, as appropriate, to avoid similar occurrences. Suggestions


ly challenged
contained within this IN are not NRC requirements; therefore, no specific action or written


, safety system functions.  The NRC expects recipients to review the information contained within for applicability to their facilities and consider actions, as appropriate, to avoid similar occurrences.  Suggestions contained within this IN are not NRC requirements; therefore, no specific action or written response is required.
response is required.


==DESCRIPTION OF CIRCUMSTANCES==
==DESCRIPTION OF CIRCUMSTANCES==
Susquehanna Steam Electric Station (Susquehanna) HVAC Controller
Susquehanna Steam Electric Station (Susquehanna) HVAC Controller


On January
On January 3, 2011, PPL, the licensee for Susquehanna, identified a single-point vulnerability in


3, 2011, PPL, the licensee for
the reactor building HVAC system. The vulnerability was that a failure of a nonsafety-related


Susquehanna
temperature controller coincident with outside ambient air temperatures below 10 degrees


, identified a single-point vulnerability in the reactor building HVAC system.  The vulnerability was that a failure of a nonsafety
Fahrenheit (oF) could result in a spurious steam leak detection (SLD) system isolation on high


-related temperature controller coincident with outside ambient air temperatures below 10 degrees
differential temperature ( T), causing simultaneous isolation of main steam isolation valves


Fahrenheit (o F) could result in a spurious steam leak detection (SLD) system isolation
(MSlV), the high pressure coolant injection system, and the reactor core isolation cooling


on high differential temperature (T), causing simultaneous isolation of main steam isolation valves (MSlV), the high pressure coolant injection system, and the reactor core isolation cooling system. This vulnerability was common to both
system. This vulnerability was common to both Susquehanna Units 1 and 2 and had been in


Susquehanna Units
existence since the plants began licensed operations.


1 and 2 and had been in existence since the plants began licensed operations.
PPL initially reported the issue through an event notification (EN) (EN 46519) under


PPL initially reported the issue through a
10 CFR 50.72, Immediate Notification Requirements for Operating Nuclear Power Reactors, as an unanalyzed condition (10 CFR 50.72 (b)(3)(ii)(B)) and an accident mitigation concern


n event notification (EN) (EN 46519) under 10 CFR 50.72 , "Immediate Notification Requirements for Operating Nuclear Power Reactors,"
(10 CFR 50.72 (b)(3)(v)(D)). However, on February 28, 2011, PPL submitted an updated EN
as an unanalyzed condition (10
CFR 50.72 (b)(3)(ii)(B)) and an accident


mitigation concern (10 CFR 50.72 (b)(3)(v)(D))
that removed the accident mitigation consideration based on the low likelihood of a reactor
.  However, on February


28, 2011, PPL submitted an updated
ML12115A012 building temperature controller failure during a period when outside temperature was below


EN that removed the accident mitigation consideration based on the low likelihood of a reactor
10 oF (both conditions are required for the deficient SLD system isolation on high T to occur).


IN 201 2-12 building temperature controller failure during a period
PPL provided additional information pertaining to this issue in the form of a 10 CFR 50.73, License Event Report [LER] System, for an unanalyzed condition (LER 3872011001). The


when outside temperature was below
LER stated that the single-point vulnerability was discovered during the preparation of a


10 o F (both conditions are required for the deficient SLD system isolation
10 CFR 50.59, Changes, Tests and Experiments, determination for an engineering change to


on high T to occur).  PPL provided additional information pertaining to this issue in the form of a 10
remove the SLD high T isolation function to address obsolescence of the functions
CFR 50.73 , "License Event Report [LER] System," for an unanalyzed condition (LER 3872011001).  The LER stated that the single


-point vulnerability was discovered during the preparation of a
components. The licensee attributed the issue to a less than adequate single-failure analysis


10 CFR 50.5 9 , "Changes, Tests and Experiments,"
performed during the original plant design.
determination for an engineering change to remove the SLD high T isolation function to address obsolescence of the function's components. The licensee attributed the issue to a "less than adequate single


-failure analysis performed during the original plant design." 
The original single-failure analysis was performed consistent with accepted practices during the
  The original single


-failure analysis was performed consistent with
period of the initial plant design. In 2007, Susquehanna engineers received training on failure


accepted practices
modes and effects analysis (FMEA) techniques. This training updated the expectations for


during the period of the initial plant design. In 2007, Susquehanna engineers received training on failure modes and effects analysis (FMEA) techniques
FMEAs performed on nonsafety systems. Consequently, Susquehanna engineers used the


.  This training updated the expectations for FMEAs performed on nonsafety systems.  Consequently, Susquehanna engineers used the
new techniques when evaluating the impact of removing the SLD isolation function and, in the


new techniques when evaluating the impact of removing the SLD isolation function and , in the process , identified the single
process, identified the single-point vulnerability deficiency.


-point vulnerability deficiency.
The corrective actions for this issue included removing the isolation function of the SLD system


The corrective actions for this issue included removing the isolation function of the SLD system T instrumentation and performing a FMEA on all nonsafety systems that could cause an isolation of the emergency core cooling system
T instrumentation and performing a FMEA on all nonsafety systems that could cause an


or MSIV s as an extent of condition assessment.
isolation of the emergency core cooling system or MSIVs as an extent of condition assessment.


The report, "Susquehanna
The report, Susquehanna Steam Electric Station - NRC Integrated Inspection Report


Steam Electric Station
05000387/2011003 and 05000388/2011003 and Exercise of Enforcement Discretion, dated


- NRC Integrated Inspection Report 05000387/2011003 and 05000388
August 10, 2011 (Agencywide Documents Access and Management System (ADAMS)
/2011003 and Exercise of Enforcement Discretion
Accession No. ML112220409), provides the results of the NRC inspection related to this issue.


," dated August 10, 2011 (Agencywide
Diablo Canyon Power Plant Auxiliary Building Ventilation System Actuation Logic


Documents Access and
Diablo Canyon Nuclear Power Plant (DCNPP) completed modifications to its auxiliary building


Management
ventilation systems (ABVS) in November 2010. These modifications included replacement of


System (ADAMS)
relay-based actuation logic with a programmable logic controller (PLC). The licensee
Accession No.


ML112220409), provides the results of the NRC inspection related to this issue.
implemented the modification to address problems with reliability and availability (i.e.


Diablo Canyon Power Plant Auxiliary Building Ventilation System
obsolescence). The licensee reviewed the modification design to ensure applicable


Actuation Logic
single-failure criteria were met. Notwithstanding the licensees review, on January 10, 2011, during containment spray pump quarterly testing, a deficiency in the actuation logic of the


===Diablo Canyon Nuclear Power Plant===
recently installed PLC resulted in a complete loss of the Unit 2 ABVS when a damper failed to
(DC NPP) completed modification


s to its auxiliary building ventilation system
open as required because of leakage past a piston seal. This led one of the two ABVS exhaust


s (ABVS) in November 2010. These modification
fans to trip and prevented the other exhaust fan from starting; thus ABVS became inoperable.


s included replacement of relay-based actuation logic with a programmable logic controller (PLC).  The licensee implemented the modification to address problems with reliability and availability (i.e. obsolescence).  The licensee reviewed
The loss of the ABVS required the licensee to take action in accordance with Technical


the modification design to ensure applicable single-failure criteria were met. Notwithstanding the licensee's review, on January
Specification Limiting Condition for Operation 3.0.3 (i.e., action statement to reduce mode of


10, 2011, during containment spray pump quarterly testing, a deficiency in the actuation logic of the recently installed PLC resulted in a complete loss of the Unit 2 ABVS when a damper failed to open as required
plant operation) for approximately 20 minutes until operators restored the ABVS system through


because of
manual actions. The failure of the piston seal was attributed to using the seal beyond its


leakage past a piston seal.
defined service life, contrary to the requirements of the licensees preventive maintenance


This led one of the two ABVS
program for the seal.


exhaust fans to trip and prevented the other exhaust fan from starting; thus ABVS became inoperable.
DCNPP initially reported this event through a 10 CFR 50.72 EN (EN 46531) as an unanalyzed


The loss of the ABVS required the licensee to tak e action in accordance with Technical Specification Limiting Condition for
condition (10 CFR 50.72(b)(3)(ii)(B)) and an accident mitigation concern


Operation 3.0.3 (i.e.
(10 CFR 50.72(b)(3)(v)(D)). The licensee provided additional information in the form of a


, action statement to reduce mode of
10 CFR 50.73 LER for an unanalyzed condition and safety system functional failure


plant operation) for approximately 20
(LER 2752011002). In the LER, the licensee incorrectly attributed the cause of the loss of the ABVS to a nonconforming single-failure vulnerability in the ABVS system design that existed as
minutes until operators restored the ABVS system through manual actions.


The failure of the piston seal was attributed to using the seal beyond its defined service life
part of the original design for both DCNPP Units. It was later determined that the 2010
modifications to the ABVS control logic introduced a single-failure vulnerability, where ABVS


, contrary to the requirements of the licensee's preventive maintenance program for the seal.
exhaust fans tripped when a system damper was not fully opened.


===DC NPP initially reported this event through a 10===
The corrective actions for this issue consisted of modifying the design of both DCNPP units to
CFR 50.72 EN (EN 46531) as an unanalyzed condition (10
CFR 50.72(b)(3)(ii)(B)) and an accident mitigation concern (10 CFR 50.72(b)(3)(v)(D)).  The licensee provided additional information in the form of a


10 CFR 50.73 LER for an unanalyzed condition and safety system functional failure
satisfy the single-failure design criteria, revising the design change process to include a design


(LER 2752011002).  In the LER, the licensee incorrectly attributed the cause of the loss of the
evaluation of new and old failure modes based on the current licensing and design bases, and


IN 201 2-12 ABVS to a nonconforming single
revising the licensing basis.


-failure vulnerability in the ABVS system design that existed as part of the original design for both DC
The report, Diablo Canyon Power Plant - NRC Integrated Inspection Report


NPP Units.  It was later determined that the 2010 modifications to the
05000275/2011002 and 05000323/2011002, dated May 11, 2011 (ADAMS Accession


ABVS control logic introduced a single
No. ML111310608), provides the results of the NRC inspection related to this issue.


-failure vulnerability, where ABVS exhaust fans tripped when a system damper was not fully opened.
Point Beach Nuclear Plant (Point Beach) Control Room Emergency Filtration Fan Thermal


The corrective actions for this issue consisted of modifying the design of both DC
Overload


===NPP units to satisfy the single===
On February 3, 2007, Point Beach lost operability of the control room emergency filtration
-failure design criteria, revising the design change process to include a design evaluation of new and old failure modes


based on the current licensing and design bases, and revising the licensing basis.     The report, "
system (CREFS) because of an inadequately designed modification (LER 2662007001). In


===Diablo Canyon Power Plant===
October 2006, the licensee installed a modification (high efficiency CREFS fan motors) for the
- NRC Integrated Inspection Report 05000275/2011002 and 05000323/2011002
," dated May 11, 2011 (ADAMS Accession No. ML111310608), provides the results of the NRC inspection related to this issue.    Point Beach Nuclear Plant (Point Beach)
Control Room Emergency Filtration


Fan Thermal Overload  On February
purpose of increasing the low flow margin. During the design of this modification, an incorrect


3, 2007 , Point Beach lost operability of the control room emergency filtration system (CREFS) because of
assumption was made that outside temperature had a negligible effect on motor current draw, so no compensation for low temperature was included in the motor thermal overload design.


an inadequately designed modification (LER 2662007001).  In October 2006, the licensee installed a modification (high efficiency CREF
On February 3, 2007, with outside temperature at 6 oF, a CREFS fan tripped during a Technical


S fan motors) for the purpose of increasing the low flow margin. During the design of this modification, an incorrect assumption was made that outside temperature had a negligible effect on motor current draw
Specification surveillance test because of a thermal overload relay trip. After evaluating the


, so no compensation for low temperature was included in the motor thermal overload design.  On February
cause of the trip, the licensee declared both CREFS fans inoperable because the fan motors


3, 2007, with outside temperature at 6 o F , a CREF S fan tripped during a Technical Specification
had inadequately sized thermal overload heater elements.


surveillance test because of
The corrective actions for this issue included replacing the overload heater elements with


a thermal overload relay trip. After evaluating the cause of the trip, the licensee declared both CREF
elements having trip current setpoints adjusted to values that considered design requirements.


===S fans inoperable because the fan motor===
The report, Point Beach Nuclear Power Plant, Units 1 and 2, NRC Integrated Inspection Report
s had inadequately sized thermal overload heater elements.


The corrective actions for this issue included replacing the overload heater elements with elements having trip current setpoints adjusted
05000266/2007002 and 05000301/2007002, dated April 12, 2007 (ADAMS Accession No.


to values that considered design requirements.
ML071020081), provides the results of the NRC inspection related to this event.


The report, "Point Beach
==BACKGROUND==
 
Criterion III of Appendix B to 10 CFR Part 50 requires, in part, that licensees ensure that
Nuclear Power Plant, Units 1 a nd 2, NRC Integrated Inspection Report 05000266/2007002 and 05000301/2007002
," dated April
 
12, 2007 (ADAMS Accession No. ML071020081
), provides the results of the NRC inspection related to this event.


==BACKGROUND==
applicable regulatory requirements and design basis are correctly translated into specifications, drawings, procedures, and instructions. Furthermore, design changes, including field


===Criterion III of Appendix B===
changes, shall be subject to design control measures commensurate with those applied to the
to 10 CFR Part 50 requires, in part, that licensees ensure that applicable regulatory requirements and design basis are "correctly translated into specifications, drawings, procedures, and instructions."  Furthermore, "design changes, including field changes, shall be subject to design control measures commensurate with those applied to the original design..


." 
original design...
IN 201 2-12


==DISCUSSION==
==DISCUSSION==
In each event described in this
In each event described in this IN, a safety systems function was challenged or potentially
 
IN, a safety system's function was challenged or potentially challenged because of
 
design control issues.


In the first case, a long
challenged because of design control issues. In the first case, a long-standing design control


-standing design control issue was finally identified after the licensee adopted updated methods of analyzing
issue was finally identified after the licensee adopted updated methods of analyzing nonsafety


nonsafety system designs for single failures.
system designs for single failures. In the second and third cases, actual safety system


In the second and third cases, actual safety system functional failures occurred as a result of licensees implementing deficient modifications.
functional failures occurred as a result of licensees implementing deficient modifications. These


These events illustrate the importance of evaluating modifications rigorously to verify that design
events illustrate the importance of evaluating modifications rigorously to verify that design-basis


-basis requirements are satisfied.
requirements are satisfied.


==CONTACT==
==CONTACT==
This IN requires no specific action or written response.
This IN requires no specific action or written response. Please direct any questions about this


Please direct any questions about this matter to the technical contact
matter to the technical contacts listed below or to the appropriate Office of Nuclear Reactor


s listed below or to the appropriate Office of Nuclear Reactor Regulation or Office of New Reactors project manager.
Regulation or Office of New Reactors project manager.


/RA by JLuehman for/
/RA by JLuehman for/                                   /RA by SBahadur for/
    /RA by SBahadur for/
Laura A. Dudes, Director                              Timothy J. McGinty, Director
  Laura A. Dudes, Director


Timothy J. McGinty, Director
Division of Construction Inspection                    Division of Policy and Rulemaking


Division of Construction Inspection
and Operational Programs                            Office of Nuclear Reactor Regulation


Division of Policy and Rulemaking
===Office of New Reactors===
Technical Contacts:    Samir Darbali, NRR


and Operational Programs
301-415-3730
                        E-mail: Samir.Darbali@nrc.gov


Office of Nuclear Reactor Regulation
David Garmon, NRR
 
Office of New Reactors
 
Technical Contact
 
s: Samir Darbali, NRR


301-415-3730  E-mail: Samir.Darbali@nrc.gov
301-415-3512 E-mail: David.Garmon@nrc.gov


David Garmon, NRR
Note: NRC generic communications may be found on the NRC public Web site, http://www.nrc.gov, under NRC Library/Document Collections.


301-415-3512 E-mail:  David.Garmon@nrc.gov
ML12115A012              *via e-mail                   TAC ME7683 NRR/DIRS/IOE NRR/DE/EIC NRR/DPR/PRL


Note:  NRC generic communications may be found on the NRC public Web site, http://www.nrc.gov, under NRC Library/Document Collections
OFFICE NRR/DIRS/IOEB* Tech Editor* NRR/DE/EICB*
                                                                B/BC*          B/BC*          B/BC*
                                                              HChernoff      JThorp      JQuichocho


.
NAME      DGarmon            CHsu          SDarbali


ML12115A012            *via e-mail                                TAC ME 7683 OFFICE NRR/DIRS/IOEB
(EThomas for)
DATE      6/14/12          4/30/12        6/20/12          6/18/12        6/20/12        7/5/12 NRR/DPR/PG NRR/DPR/PG NRR/DPR/PGC NRR/DPR/P


* Tech Editor*
OFFICE    NRR/DE/D                                                                      NRO/DCIP/D      NRR/DPR/D
NRR/DE/EICB


* NRR/DIRS/IOEB/BC* NRR/DE/EICB/BC* NRR/DPR/PRLB/BC* NAME DGarmon CHsu SDarbali HChernoff
CB/LA*           CB/PM            B/BC         GCB/LA


(EThomas for) JThorp JQuichocho
PHiland                                                                          LDudes        TMcGinty


DATE 6/14/12 4/30/12 6/20/12 6/18/12 6/20/12 7/5/12 OFFICE NRR/DE/D NRR/DPR/PGCB/LA* NRR/DPR/PGCB/PM NRR/DPR/PGCB/BC NRR/DPR/PGCB/LA NRO/DCIP/D
NAME                          CHawes        ARussell          DPelton      CHawes


NRR/DPR/D NAME PHiland  (MCheok for)
(MCheok for)                                                                   (JLuehman for) (SBahadur for)
CHawes ARussell DPelton CHawes LDudes (JLuehman for)
DATE       6/28/12           7/10/12         7/12/12         7/18/12       7/18/12       7/24/12       7/24/12}}
TMcGinty (SBahadur for)
DATE 6/28/12 7/10/12 7/12/12 7/18/12 7/18/12 7/24/12 7/24/12}}


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

Latest revision as of 05:35, 12 November 2019

HVAC Design Control Issues Challenge Safety System Function
ML12115A012
Person / Time
Issue date: 07/24/2012
From: Laura Dudes, Mcginty T
Division of Construction Inspection and Operational Programs, Division of Policy and Rulemaking
To:
Garmon-Candelaria D
References
IN-12-012
Download: ML12115A012 (5)
v • d • e


UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

OFFICE OF NEW REACTORS

WASHINGTON, DC 20555-0001 July 24, 2012 NRC INFORMATION NOTICE 2012-12: HVAC DESIGN CONTROL ISSUES CHALLENGE

SAFETY SYSTEM FUNCTION

ADDRESSEES

All holders of an operating license or construction permit for a nuclear power reactor or a

non-power (research or test) reactor issued under Title 10 of the Code of Federal Regulations

(10 CFR) Part 50, Domestic Licensing of Production and Utilization Facilities, except those

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

removed from the reactor vessel.

All holders of and applicants for a power reactor early site permit, combined license, standard

design certification, standard design approval, or manufacturing license under 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants.

PURPOSE

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

addressees about certain events involving heating, ventilation, and air conditioning (HVAC)

system design control issues that challenged, or potentially challenged, safety system functions.

The NRC expects recipients to review the information contained within for applicability to their

facilities and consider actions, as appropriate, to avoid similar occurrences. Suggestions

contained within this IN are not NRC requirements; therefore, no specific action or written

response is required.

DESCRIPTION OF CIRCUMSTANCES

Susquehanna Steam Electric Station (Susquehanna) HVAC Controller

On January 3, 2011, PPL, the licensee for Susquehanna, identified a single-point vulnerability in

the reactor building HVAC system. The vulnerability was that a failure of a nonsafety-related

temperature controller coincident with outside ambient air temperatures below 10 degrees

Fahrenheit (oF) could result in a spurious steam leak detection (SLD) system isolation on high

differential temperature ( T), causing simultaneous isolation of main steam isolation valves

(MSlV), the high pressure coolant injection system, and the reactor core isolation cooling

system. This vulnerability was common to both Susquehanna Units 1 and 2 and had been in

existence since the plants began licensed operations.

PPL initially reported the issue through an event notification (EN) (EN 46519) under

10 CFR 50.72, Immediate Notification Requirements for Operating Nuclear Power Reactors, as an unanalyzed condition (10 CFR 50.72 (b)(3)(ii)(B)) and an accident mitigation concern

(10 CFR 50.72 (b)(3)(v)(D)). However, on February 28, 2011, PPL submitted an updated EN

that removed the accident mitigation consideration based on the low likelihood of a reactor

ML12115A012 building temperature controller failure during a period when outside temperature was below

10 oF (both conditions are required for the deficient SLD system isolation on high T to occur).

PPL provided additional information pertaining to this issue in the form of a 10 CFR 50.73, License Event Report [LER] System, for an unanalyzed condition (LER 3872011001). The

LER stated that the single-point vulnerability was discovered during the preparation of a

10 CFR 50.59, Changes, Tests and Experiments, determination for an engineering change to

remove the SLD high T isolation function to address obsolescence of the functions

components. The licensee attributed the issue to a less than adequate single-failure analysis

performed during the original plant design.

The original single-failure analysis was performed consistent with accepted practices during the

period of the initial plant design. In 2007, Susquehanna engineers received training on failure

modes and effects analysis (FMEA) techniques. This training updated the expectations for

FMEAs performed on nonsafety systems. Consequently, Susquehanna engineers used the

new techniques when evaluating the impact of removing the SLD isolation function and, in the

process, identified the single-point vulnerability deficiency.

The corrective actions for this issue included removing the isolation function of the SLD system

T instrumentation and performing a FMEA on all nonsafety systems that could cause an

isolation of the emergency core cooling system or MSIVs as an extent of condition assessment.

The report, Susquehanna Steam Electric Station - NRC Integrated Inspection Report 05000387/2011003 and 05000388/2011003 and Exercise of Enforcement Discretion, dated

August 10, 2011 (Agencywide Documents Access and Management System (ADAMS)

Accession No. ML112220409), provides the results of the NRC inspection related to this issue.

Diablo Canyon Power Plant Auxiliary Building Ventilation System Actuation Logic

Diablo Canyon Nuclear Power Plant (DCNPP) completed modifications to its auxiliary building

ventilation systems (ABVS) in November 2010. These modifications included replacement of

relay-based actuation logic with a programmable logic controller (PLC). The licensee

implemented the modification to address problems with reliability and availability (i.e.

obsolescence). The licensee reviewed the modification design to ensure applicable

single-failure criteria were met. Notwithstanding the licensees review, on January 10, 2011, during containment spray pump quarterly testing, a deficiency in the actuation logic of the

recently installed PLC resulted in a complete loss of the Unit 2 ABVS when a damper failed to

open as required because of leakage past a piston seal. This led one of the two ABVS exhaust

fans to trip and prevented the other exhaust fan from starting; thus ABVS became inoperable.

The loss of the ABVS required the licensee to take action in accordance with Technical

Specification Limiting Condition for Operation 3.0.3 (i.e., action statement to reduce mode of

plant operation) for approximately 20 minutes until operators restored the ABVS system through

manual actions. The failure of the piston seal was attributed to using the seal beyond its

defined service life, contrary to the requirements of the licensees preventive maintenance

program for the seal.

DCNPP initially reported this event through a 10 CFR 50.72 EN (EN 46531) as an unanalyzed

condition (10 CFR 50.72(b)(3)(ii)(B)) and an accident mitigation concern

(10 CFR 50.72(b)(3)(v)(D)). The licensee provided additional information in the form of a

10 CFR 50.73 LER for an unanalyzed condition and safety system functional failure

(LER 2752011002). In the LER, the licensee incorrectly attributed the cause of the loss of the ABVS to a nonconforming single-failure vulnerability in the ABVS system design that existed as

part of the original design for both DCNPP Units. It was later determined that the 2010

modifications to the ABVS control logic introduced a single-failure vulnerability, where ABVS

exhaust fans tripped when a system damper was not fully opened.

The corrective actions for this issue consisted of modifying the design of both DCNPP units to

satisfy the single-failure design criteria, revising the design change process to include a design

evaluation of new and old failure modes based on the current licensing and design bases, and

revising the licensing basis.

The report, Diablo Canyon Power Plant - NRC Integrated Inspection Report 05000275/2011002 and 05000323/2011002, dated May 11, 2011 (ADAMS Accession

No. ML111310608), provides the results of the NRC inspection related to this issue.

Point Beach Nuclear Plant (Point Beach) Control Room Emergency Filtration Fan Thermal

Overload

On February 3, 2007, Point Beach lost operability of the control room emergency filtration

system (CREFS) because of an inadequately designed modification (LER 2662007001). In

October 2006, the licensee installed a modification (high efficiency CREFS fan motors) for the

purpose of increasing the low flow margin. During the design of this modification, an incorrect

assumption was made that outside temperature had a negligible effect on motor current draw, so no compensation for low temperature was included in the motor thermal overload design.

On February 3, 2007, with outside temperature at 6 oF, a CREFS fan tripped during a Technical

Specification surveillance test because of a thermal overload relay trip. After evaluating the

cause of the trip, the licensee declared both CREFS fans inoperable because the fan motors

had inadequately sized thermal overload heater elements.

The corrective actions for this issue included replacing the overload heater elements with

elements having trip current setpoints adjusted to values that considered design requirements.

The report, Point Beach Nuclear Power Plant, Units 1 and 2, NRC Integrated Inspection Report 05000266/2007002 and 05000301/2007002, dated April 12, 2007 (ADAMS Accession No.

ML071020081), provides the results of the NRC inspection related to this event.

BACKGROUND

Criterion III of Appendix B to 10 CFR Part 50 requires, in part, that licensees ensure that

applicable regulatory requirements and design basis are correctly translated into specifications, drawings, procedures, and instructions. Furthermore, design changes, including field

changes, shall be subject to design control measures commensurate with those applied to the

original design...

DISCUSSION

In each event described in this IN, a safety systems function was challenged or potentially

challenged because of design control issues. In the first case, a long-standing design control

issue was finally identified after the licensee adopted updated methods of analyzing nonsafety

system designs for single failures. In the second and third cases, actual safety system

functional failures occurred as a result of licensees implementing deficient modifications. These

events illustrate the importance of evaluating modifications rigorously to verify that design-basis

requirements are satisfied.

CONTACT

This IN requires no specific action or written response. Please direct any questions about this

matter to the technical contacts listed below or to the appropriate Office of Nuclear Reactor

Regulation or Office of New Reactors project manager.

/RA by JLuehman for/ /RA by SBahadur for/

Laura A. Dudes, Director Timothy J. McGinty, Director

Division of Construction Inspection Division of Policy and Rulemaking

and Operational Programs Office of Nuclear Reactor Regulation

Office of New Reactors

Technical Contacts: Samir Darbali, NRR

301-415-3730

E-mail: Samir.Darbali@nrc.gov

David Garmon, NRR

301-415-3512 E-mail: David.Garmon@nrc.gov

Note: NRC generic communications may be found on the NRC public Web site, http://www.nrc.gov, under NRC Library/Document Collections.

ML12115A012 *via e-mail TAC ME7683 NRR/DIRS/IOE NRR/DE/EIC NRR/DPR/PRL

OFFICE NRR/DIRS/IOEB* Tech Editor* NRR/DE/EICB*

B/BC* B/BC* B/BC*

HChernoff JThorp JQuichocho

NAME DGarmon CHsu SDarbali

(EThomas for)

DATE 6/14/12 4/30/12 6/20/12 6/18/12 6/20/12 7/5/12 NRR/DPR/PG NRR/DPR/PG NRR/DPR/PGC NRR/DPR/P

OFFICE NRR/DE/D NRO/DCIP/D NRR/DPR/D

CB/LA* CB/PM B/BC GCB/LA

PHiland LDudes TMcGinty

NAME CHawes ARussell DPelton CHawes

(MCheok for) (JLuehman for) (SBahadur for)

DATE 6/28/12 7/10/12 7/12/12 7/18/12 7/18/12 7/24/12 7/24/12


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