Information Notice 2008-02, Findings Identified During NRC Component Design Bases Inspections
| ML073450262 | |
| Person / Time | |
|---|---|
| Issue date: | 03/19/2008 |
| From: | Michael Case NRC/NRR/ADRA/DPR |
| To: | |
| References | |
| SECY-05-0118 IN-08-002 | |
| Download: ML073450262 (8) | |
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, DC 20555-0001 March 19, 2008 NRC INFORMATION NOTICE 2008-02: FINDINGS IDENTIFIED DURING COMPONENT
DESIGN BASES INSPECTIONS
ADDRESSES
All holders of operating licenses for nuclear power reactors, except those who have
permanently ceased operations and have certified that fuel has been permanently removed
from the reactor vessel.
PURPOSE
The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice (IN) to inform
addressees of issues identified during recent component design bases inspections (CDBIs)
regarding the capability of selected components to perform their design bases safety functions.
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 IN are not NRC requirements; therefore, no specific action or written response is required.
BACKGROUND
Starting in January 2006, CDBIs replaced safety system design and performance capability
(SSDPC) inspections in the baseline inspection program of the Reactor Oversight Process.
Unlike SSDPC inspections, CDBIs are not limited to one or two systems. Instead, CDBIs verify
that design bases have been correctly implemented for selected risk-significant and low-margin
components and that operating procedures and operator actions are consistent with design and
licensing bases. This is to ensure that selected components are capable of performing their
intended safety functions. Inspectors also review the initial design and subsequent
modifications to verify that important design features have not been altered or disabled during a
modification. The basis for the change from SSDPC inspections to CDBIs is described in
SECY-05-0118, Results of the Pilot Program to Improve the Effectiveness of Nuclear
Regulatory Commission Inspections of Engineering and Design Issues, dated July 1, 2005, available through the Agencywide Documents Access and Management System (ADAMS)
Accession No. ML051390465.
DESCRIPTION OF CIRCUMSTANCES
Following is a summary of the findings identified by CDBIs from their initiation in January 2006, through September 2007. This summary also includes several findings that were identified
during the pilot CDBI inspection performed in 2004. The areas of concern were identified in
SECY-05-0118. Enclosure 1 lists the referenced NRC inspection reports. The specific CDBI
findings can be accessed under ADAMS Accession No. ML080670419.
(1) Potential Air Entrainment and Vortexing of Safety-Related Fluid Systems
NRC inspectors identified instances where design deficiencies could have led to air entering
safety-related water systems, potentially resulting in damage to pumps and a degradation or
loss of system function. The majority of these design issues were related to inappropriate
setpoint calculations for refueling water storage tank and condensate storage tank (CST) level
instruments. These level instruments, and associated alarms, are used either to stop
safety-related pumps or to realign their suctions to alternate water supplies. The failure to stop
operating pumps or to transfer their suctions can result in the ingestion of gas from these tanks
which could cause degraded system performance and would eventually cause damage to the
pumps. In some cases, both trains of a redundant safety-related system could be affected by
this design deficiency. The potentially affected systems for pressurized-water reactors included
auxiliary feedwater, containment spray, high pressure safety injection, and low pressure safety
injection/residual heat removal (RHR). The potentially affected systems for boiling-water
reactors included high pressure core spray and reactor core isolation cooling.
The level setpoint issues included: (a) the use of inappropriate methodology to predict the
onset of vortexing in tanks, (b) the failure to allow for the time required for the transfer of the
pump suctions to be completed (operator response times and valve stroke times), and (c) the
failure to correctly translate analysis results into operating procedures. One case involved the
failure to account for instrument uncertainty, resulting in potentially inadequate vortexing margin
for the RHR pumps during reactor coolant system mid-loop operation. Another case involved
failure to account for the potential effect of air entrainment on the level instrument sensing lines.
Some deficiencies involved the existence of air in the suction piping associated with safety- related pumps. This stemmed from the failure to ensure that the piping was full of water
following testing or maintenance activities. One case involved a system design inadequacy that
created the potential for air intrusion into operating auxiliary feedwater (AFW) pumps, potentially
resulting in a common mode failure of the AFW system. The AFW pump discharge pressure
trip design was not in compliance with the plant licensing basis because it could not be shown
that the AFW pumps would trip prior to pump damage, due to air ingestion if the CSTs or
suction piping failed during a seismic or tornado event.
(2) Emergency Diesel Generators
NRC inspectors identified instances where the emergency diesel generators (EDGs) loading
calculations failed to account for the increased electrical load resulting from EDG operation at
the maximum frequency allowed by technical specifications. The operation of rotating
equipment at higher speeds (resulting from higher electrical frequencies) would result in
increased EDG loads under accident conditions. In some cases, the EDG capacity margin was
found to be small compared to the potential load increase.
Other EDG deficiencies included: (a) failure to verify that fuel oil testing results were within
specified limits, (b) failure to correctly determine tube plugging limits for associated heat
exchangers, (c) failure to account for all potential electrical loads in analyses, (d) failure to verify
fuel transfer equipment was rated for the required temperature, and (e) inadequate procedures
related to ground faults and tornado depressurization. (3) Testing
NRC inspectors identified instances where test acceptance criteria failed to ensure the
capability of the equipment to perform its function under the most limiting conditions. Examples
included the acceptance criteria for valve and pump surveillance tests as well as design
requirements and safety-related battery tests that did not include appropriate minimum voltage
acceptance criteria. NRC inspectors identified deficiencies in test programs for molded case
circuit breakers. Other test deficiencies included: (a) failure to account for EDG under- frequency in pump test acceptance criteria, (b) failure to appropriately account for instrument
uncertainties, (c) failure to ensure adequate test equipment, (d) failure to account for valve
pressure locking effects, and (e) failure to verify the minimum containment cooling coil fouling
factor assumed in analyses.
(4) Cooling Water Systems
NRC inspectors identified instances where the licensee did not demonstrate that systems were
capable of performing their intended safety function under the most limiting conditions. One
finding involved the failure to appropriately account for service water strainer plugging in the
flow calculation. Another finding was that system design calculations did not reflect the
maximum strainer pressure drop allowed by the plant operating procedures, resulting in
potentially non-conservative analyses.
Other issues included: (a) failure to evaluate the potential loss of a non-safety-related service
water valve positioner, (b) failure to adequately evaluate the potential effects on system
performance of a closed valve that was leaking significantly, (c) failure to address maximum
component cooling water piping temperatures in the pump room heat up calculation, (d)
modification to remove four fans from the safety-related screenhouse ventilation system without
adequately verifying the adequacy of remaining fans, and (e) potential plugging of emergency
(auxiliary) feedwater flow control valves when aligned to the service water system, which served
as a backup water supply.
Procedure deficiencies included: (a) failure to verify by testing that adequate flow was provided
to an essential service water cooling tower, (b) inadequate operating procedures to perform time
critical operations after the loss of component cooling water, and (c) plant procedures that would
have allowed a plugged strainer to be bypassed without an evaluation of the potential impact on
downstream equipment.
(5) Station Blackout
NRC inspectors identified deficiencies related to plant operation after a postulated loss of all
alternating current (AC) power. This scenario may require equipment to operate in off-normal
conditions and may require the use of non-safety-related equipment to maintain the plant in a
safe condition for the design basis coping duration. The identified deficiencies included: (a)
failure to perform analyses or have adequate procedures to ensure that building temperatures
would be acceptable for the operation of equipment credited for coping with a station blackout, (b) failure to perform coping analyses that covered the entire coping duration, (c) failure to
address the degraded reliability of a non-safety-related gas turbine, and (d) failure to perform an
evaluation to support a procedure change that substituted manual actions for automatic actions during a station blackout event in accordance with Title 10 of the Code of Federal Regulations
(10 CFR), Section 50.59, Changes, Tests and Experiments.
(6) Motor Operated Valves
NRC inspectors identified where licensees failed to identify the maximum differential pressures
across containment sump isolation valves, and analyze the ability of the valves to open with
these pressures. The postulated scenario involves the pressurization of the piping between
these valves and the suction of the safety injection/RHR pumps after an accident where the low
pressure pumps are not injecting flow. The failure of these valves to open could result in a loss
of long term emergency core cooling. In addition, inspectors identified where a licensee failed
to use adequate methodology to determine if a motor-operated valve (MOV) could be
susceptible to pressure locking phenomenon.
NRC Inspectors identified deficiencies including non-conservative inputs to MOV voltage
calculations, and incorrect control logic for containment isolation valves. NRC inspectors also
identified deficiencies in MOV overload protection, as described in section (9) of this IN.
(7) Operability Evaluations
NRC inspectors identified deficiencies including the failure to identify degraded conditions and
perform operability evaluations, the failure to identify past operability issues for potential
reportability, and the failure to perform adequate operability evaluations for degraded conditions.
One licensee failed to promptly identify and correct a long-standing condition involving an
inadequate safety analysis dose calculation, and failed to maintain previously imposed
compensatory measures that were credited in plant analyses.
(8) Standby Batteries and Direct Current Electrical Distribution Systems
NRC inspectors identified deficiencies in the 125 volts direct current (Vdc) system voltage drop
calculations including: (a) failure to include acceptance criteria for end use equipment such as
circuit breakers, (b) failure to analyze an alternate supply for the 125 Vdc system, (c) improper
methodology for determining first minute voltage, (d) failure to consider effect of accident
temperatures on cable resistance, and (e) failure to use conservative design inputs for battery
inter-cell resistance and battery terminal voltage. Deficiencies were identified in battery sizing
calculations involving failure to correctly model battery loads and charger restoration
procedures.
NRC inspectors identified instances of improper battery and battery charger maintenance
including failure to follow procedures for cleaning battery terminals, inadequate procedures for
cleaning spilled electrolyte, use of a non-safety-related battery charger without proper isolation, failure to periodically energize a spare battery charger, improper torque values for battery
terminals, and ineffective corrective action for high battery inter-cell resistance. NRC inspectors
also identified instances of inadequate battery testing listed under section (3) of this IN. (9) Alternating Current Auxiliary Power Systems
NRC inspectors identified instances of deficiencies in voltage calculations for AC auxiliary power
systems. These included: (a) lack of motor starting studies, (b) use of incorrect brake
horsepower, (c) using incorrect assumption for voltage drop in cables, and (d) errors in control
circuit analyses for motor control centers. NRC inspectors identified instances where
calculations for offsite power availability were inadequate including failure to analyze all offsite
sources, and failure to properly analyze a modified transformer.
NRC inspectors also identified problems with motor overload protection including failure to
provide overload protection for a deep well pump, continuously bypassing overload devices for
several safety-related motors including MOVs, and failure to verify the adequate sizing of fan
motor overload devices.
(10) Circuit Breakers
NRC inspectors identified instances of improper maintenance of circuit breakers involving failure
to follow vendor maintenance recommendations. Other findings related to circuit breakers
included failure to use the correct short circuit rating in a design calculation, and numerous
testing deficiencies described under the Testing topic of this IN.
DISCUSSION
This IN provides examples in which the design basis was not correctly implemented for certain
systems and components as required by 10 CFR Part 50, Appendix B, Criterion III, Design
Control. The issues often involved the capability of the plant system or component to perform
its intended safety function under the most limiting conditions (e.g., pressure, temperature)
assumed in the plant design and licensing basis. In some cases, the issues were original
design deficiencies that had not been adequately addressed. In other cases, these design
deficiencies were introduced as a result of plant design changes, calculation revisions, and/or
changes to operating and test procedures. Some of the design basis discrepancies resulted in
technical specification required systems and components being declared inoperable. There
were also instances where the licensee did not incorporate the requirements and acceptance
limits contained in applicable design documents into test procedures, as required by 10 CFR
Part 50, Appendix B, Criterion XI, Test Control. The issues discussed in this IN often involved
plant systems and components with low design margin. Systems and components are
designed and operated, as described in the current licensing basis, with design margins and
engineering margins of safety to ensure, among other things, that some loss of quality does not
mean immediate failure to meet a specified function. The current licensing basis includes
commitments to specific codes and standards, design criteria, and some regulations that also
dictate margins. Many licensees add conservatism so that a partial loss of quality does not
affect their commitments for design and operational margin. The loss of conservatism that is
not credited in the current licensing basis does not affect operability or functionality. The issues
discussed in this IN often involved the failure to account for factors, or the use of improper
assumptions (i.e., not using the most limiting conditions) that could reduce or eliminate the
available design margin.
The NRC and its licensees rely on calculations and analyses that predict the performance of the
facility under various accident sequences. Design deficiencies have the potential of affecting redundant systems and components and introducing common mode failures that were not
considered in accident or transient analyses. The plant risk assessment model assumes the
capability of safety systems and components to perform their intended safety function
successfully. Therefore, the accuracy of these design basis calculations and analyses has
become increasingly important as the industry and NRC implement risk-informed regulatory
initiatives.
CONTACT
This IN requires no specific action or written response. Please direct any questions about this
matter to the technical contact listed below.
/RA/
Michael J. Case, Director
Division of Policy and Rulemaking
Office of Nuclear Reactor Regulation
Technical Contact:
Donald Norkin, NRR/DIRS
301-415-2954 E-mail: dpn@nrc.gov
Enclosure: NRC Inspection Report References redundant systems and components and introducing common mode failures that were not
considered in accident or transient analyses. The plant risk assessment model assumes the
capability of safety systems and components to perform their intended safety function
successfully. Therefore, the accuracy of these design basis calculations and analyses has
become increasingly important as the industry and NRC implement risk-informed regulatory
initiatives.
CONTACT
This IN requires no specific action or written response. Please direct any questions about this
matter to the technical contact listed below.
/RA/
Michael J. Case, Director
Division of Policy and Rulemaking
Office of Nuclear Reactor Regulation
Technical Contact:
Donald Norkin, NRR/DIRS
301-415-2954 E-mail: dpn@nrc.gov
Enclosure: NRC Inspection Report References
Distribution: IN Reading File
ADAMS Accession Number: ML073450262 OFFICE IRIB:DIRS TECH EDITOR BC:IRIB:DIRS D:DIRS
NAME DNorkin HChang TKobetz FBrown
DATE 02/28/2008 12/19/2007 3/4/2008 3/7/2008 OFFICE LA:PGCB PGCB:DPR BC:PGCB:DPR D:PGCB:DPR
NAME CHawes DBeaulieu MMurphy MCase
DATE 03/12/2008 3/11/2008 03/19/2008 03/19/2008 OFFICIAL RECORD COPY
NRC INSPECTION REPORT REFERENCES
[Note: A brief description of the specific CDBI findings from the below inspection reports is
available at ADAMS Accession No. ML080670419.]
Plant NRC Inspection Report (IR) Number ADAMS No.
Byron Units 1 and 2 IR 05000454/2006009; 05000455/2006009 ML070120571 Callaway IR 05000483/2006009 ML070560002 Catawba Units 1 and 2 IR 05000413/2007006; 05000414/2007006 ML071490122 Donald C. Cook Units 1 and 2 IR 05000315/2007002; 05000316/2007002 ML071060236 Comanche Peak Units 1 and 2 IR 05000445/2006009; 05000446/2006009 ML070360606 Dresden Units 2 and 3 IR 05000237/2007006; 05000249/2007006 ML071830531 Duane Arnold IR 05000331/2006007 ML061580073 Edwin I. Hatch Units 1 and 2 IR 05000321/2006007; 05000366/2006007 ML062370129 Fermi IR 05000341/2007003 ML072540412 Ginna IR 05000244/2007006 ML073060346 Grand Gulf IR 05000416/2006008 ML061070259 Hope Creek IR 05000354/2006015 ML070190243 Indian Point Units 2 and 3 IR 05000247/2007007; ML070890270
IR 05000286/2007006 ML080320244 James A. Fitzpatrick IR 05000333/2007006 ML072430509 Kewaunee IR 05000305/2005002 ML050950237 IR 05000305/2007006 ML071550470
McGuire Units 1 and 2 IR 05000369/2006007; 05000370/2006007 ML061740013 Monticello IR 05000263/2006009 ML062190468 Nine Mile Point, Units 1 and 2 IR 05000220/2006008; 05000410/2006008 ML063350016 Oyster Creek IR 05000219/2007006 ML071870171 Oconee Units 1, 2, and 3 IR 05000269, 05000270, 05000287/2006006 ML061180004 Palisades IR 05000255/2006009 ML070440439 Peach Bottom Units 2 and 3 IR 05000277/2006009; 05000278/2006009 ML061520381 Pilgrim IR 05000293/2006006 ML061800215 Point Beach Units 1 and 2 IR 05000266/2006006; 05000301/2006006 ML063200093 Prairie Island Units 1 and 2 IR 05000282/2007007; 05000306/2007007 ML072180551 Quad Cities Units 1 and 2 IR 05000254/2006003; 05000265/2006003 ML063330597 Robinson IR 05000261/2007006 ML071100274 Saint Lucie IR 05000335/2007006; 05000389/2007006 ML073090538 Salem Units 1 and 2 IR 05000272/2006006; 05000311/2006006 ML060930500
San Onofre Units 2 and 3 IR 05000361/2006009; 05000362/2006009 ML063420342 Seabrook Station IR 05000443/2007006 ML071590071 Shearon Harris IR 05000400/2006007 ML063400335 Three Mile Island Unit 1 IR 05000289/2007006 ML070920411 Virgil C. Summer IR 05000395/2004009 ML050060203 IR 05000395/2006008 ML062080036 Vogtle Units 1 and 2 IR 05000424/2007006; 05000425/2007006 ML072050572 Vermont Yankee IR 05000271/2004008 ML043340269 IR 05000271/2006007 ML062720159 Waterford IR 050003822007007 ML071940133 Watts Bar IR 05000390/2007006 ML071100271 Wolf Creek IR 05000482/2007006 ML072880678 Enclosure