Information Notice 1998-22, Deficiencies Identified During NRC Design Inspections
UNITES STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C. 20555-0001
June 17, 1998
NRC INFORMATION NOTICE 98-22: DEFICIENCIES IDENTIFIED DURING NRC DESIGN
INSPECTIONS
Addressees
All holders of operating licenses for nuclear power reactors, except those licensees who have
permanently ceased operations and have certified that fuel has been permanently removed
from the vessel.
Purpose
The U.S. Nuclear Regulatory Commission (NRC) is issuing this Information notice to alert
addressees to Issues identified during recent design team inspections regarding the capability
of selected systems to perform their design bases safety functions. It is expected 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.
Background
In October 1996, as a result of concerns that some plant configurations and operations were
inconsistent with their design and licensing bases, the NRC formed three NRC-led teams of five
contracted engineers from architect-engineer firms to perform design-focused, inspections of
risk-significant safety systems. These Inspections were implemented to evaluate the capability
of the selected systems to perform their safety functions, the adherence of the selected
systems to their design and licensing bases, and the consistency of the as-built configuration
and system operation against the final safety analysis report (FSAR). The inspections were
conducted in accordance with NRC Inspection Procedure 93801, eSafety System Functional
Inspection,* and focused on the engineering design and configuration control sections of the
procedure. As of May 1, 1998, 16 Inspections have been completed.
Description of Circumstances
The following summarizes the most significant technical and programmatic issues that have
been identified by the 16 Inspections completed to date. In some Instances, there erg ongoing
NRC staff reviews concerning specific regulatory and technical aspects of the issues.
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IN 98-22 June 17,1998 Modifications or Evaluations That Resulted In Operation of the Plant Outside the
Licensing Bases
At some plants, there were Issues concerning operation outside the licensing bases
stated In the FSAR. At some plants, licensee 10 CFR 50.59 evaluations failed to Identify
unreviewed safety questions, and licensees made significant changes to plant
operations or equipment without NRC's approval. In some Instances, this was due to
revisions to calculations that were not subject to 10 CFR 50.59 screening programs. At
Farley and Perry, Inspectors Identified Issues involving the lack of protection from
tornado missiles. At Perry, the suppression pool cleanup (SPCU) system was
frequently operated at power, which was Inconsistent with the updated safety analysis
report (USAR) description. Due to the system Interface between the SPCU and the high
pressure core spray (HPCS) system, whenever the SPCU system Is In operation, the
HPCS system must be aligned to the suppression pool as opposed to its preferred
source, the condensate storage tank. At Diablo Canyon, the team identified a single- failure vulnerability with the component cooling water (CCW), auxiliary saltwater, and
residual heat removal (RHR) systems and an Inability to use the containment spray
system during the containment recirculation mode of the RHR system.
At Cooper, the licensee made a modification to the reactor equipment cooling system
which resulted in a leakage of 200 gallons per day from sampling valves that were
Inadvertently left In the open position. During the modification process, the need for
controlling the position of these valves was not recognized. Therefore, in accident
conditions, this leakage could have depleted the available water to the system, resulting
in an Inability of the system to support Its long-term post-accident cooling functions.
Errors In Analyses for Emergency Core Cooling System (ECCS) Pump Suction Swap- over from Refueling Water Storage Tank (RWST)/Borated Water Storage Tank (BWST)
to the Reactor Sump During a Loss of Coolant Accident (LOCA)
At five plants, licensees made errors in the calculations that were performed to ensure
adequate coolant would be available to support operation of low-pressure pumps during
and after swap-over from the RWST/BWST to the reactor building sump during a
postulated LOCA. These errors resulted from the use of non-conservative reactor
building pressures, valve stroke times, and operator response times; failing to account
for limiting system configurations; and Instrument uncertainties. The errors affected the
calculations for setting the RWST/BWST level Instrument alarms, the emergency
operating procedures (EOPs), and technical specifications. At Three Mile Island (TMII),
the licensee declared both the decay heat removal system and the reactor building
spray pumps Inoperable as a result of these concerns.
At D.C. Cook, the licensee did not sufficiently evaluate the Instrument uncertainties and
flow biases that would cause the RWST level Instrumentation to Indicate lower than
IN 98-22 June 17, 1998 actual level. In addition, the licensee was unable to demonstrate the adequacy of
drainage paths from the Inactive to the active containment recirculation sumps to
support operation of the ECCS pumps with suction from the ECCS sump. On the basis
of these concerns, the licensee declared both trains of the ECCS and the containment
spray system inoperable and Initiated a dual-unit shutdown.
At H. B. Robinson, a design modification by the licensee allowing for single safety
Injection (SI) pump operation resulted In insufficient net positive suction head (NPSH) for
two of three SI pumps. At Ginna, the licensee implemented procedure changes after a
slight negative NPSH was calculated for the RHR pump A.
At Wolf Creek, the licensee's RWST Instrument loop uncertainty calculations did not
consider density variations resulting from temperature changes and boron
concentration, which affected the RWST alarm and swap-over setpoints and the
accuracy of the RWST level Indication.
Inadequate Testing of Safety-Related Components
Inspection teams Identified numerous examples of Inadequate testing of safety related
components, including the lack of testing for certain molded-case circuit breakers at
St. Lucie, Arkansas Nuclear One (ANO), and TMI; leak and functional testing of valves
(including check valves) at TMI, St. Lucie, Ginna, Farley, D.C. Cook, Palisades, and
Davis-Besse; post-modification testing of safety injection pumps at H.B Robinson;
testing of the safety Injection lock-out relay at Indian Point 2; and testing of sections of
the auxiliary service water supply path and pumps at Diablo Canyon. In some
instances, the licensees did not perform periodic tests, while In other cases, the testing
was inadequate to demonstrate the operability of all safety functions. For example, the
licensee tested certain check valves In the forward flow but not the reverse flow
direction. At Vermont Yankee, the inspectors determined that testing of the RHR heat
exchangers was Inadequate because of invalid test instrument uncertainty assumptions.
Issues Concerning Implementation of Computer Evaluation Models Used for Analyzing
ECCS Response to Design Basis Accidents
In three inspections, the team Identified Issues concerning the computer evaluation
models used for analyzing the ECCS response to postulated design-bases accidents.
At H.B. Robinson, the computer model Indicated the existence of a second peak in fuel
clad temperature that was significantly higher than the peak temperature reported by the
licensee. At Ginna, the team Identified errors In the analysis report that called Into
question the level of review and the validity of some Inputs. At Indian Point 2, the
licensee had not established procedural controls to ensure that Input data assumptions
used In the model would not be Invalidated by plant modifications. The team also
Identified the lack of formal design control procedures between the licensee and the
vendor for supplying and verifying the validity of Input data and assumptions.
IN 98-22 June 17, 1998 System Operation at a Temperature In Excess of the Design Basis
At three plants, the inspection team found that the plant had been operated while
system ambient temperatures were in excess of the design or licensing bases. At
Vermont Yankee, the licensee aflowed the suppression pool temperature to exceed the
design-basis temperature used in the analyses for the standby cooling system pump
NPSH, containment pressure, piping stress, and equipment qualification. At D.C. Cook, the licensee operated the plant with the essential service water temperature (ultimate
heat sink) In excess of the design-basis temperature, which could affect the qualified life
of equipment In the control room and reduce the rate of heat removal from the spent fuel
pool.
At Palisades, LOCA analyses concluded that the post-accident CCW temperatures
could exceed the design-basis temperature; however, the licensee had not evaluated
system performance at the higher temperatures.
Errors Made in Evaluating Post-accident Temperatures for Safety Related Pump Rooms
At four plants, licensees made errors in calculating the maximum pump room
temperatures that would be expected during post-accident conditions. At Palisades, Ginna, and Indian Point 2, the calculations for the auxiliary feedwater pump rooms
incorrectly used nominal rather than the maximum expected ambient temperature
conditions before the accident. At Cooper, the licensee calculated the heat load for the
RHR pump room using only one of the two pumps. Also at Cooper, operating
procedures were Inadequate to ensure that the maximum calculated RHR service water
booster pump room temperature would not be exceeded.
Lack of Controls or Specified Outage Times for Limiting System Line-ups That Couild
Challenge Design-Basis Considerations
Inspectors identified issues pertaining to the lack of controls or specified outage times
(either within technical specifications or administrative procedures) for ensuring systems
are maintained In a configuration that would support all design basis considerations. At
Cooper, the licensee had no controls in place to limit the time the RHR system is
operated In the suppression pool cooling mode. In this mode of operation, the system
would not be capable of automatically realigning Into the Injection mode given certain
single-failure assumptions.
At Indian Point 2, Inspectors found that there were no controls for taking Instrument
busses off their invertors and supplying them from alternate power sources. Under
accident conditions, certain Instrumentation would be lost because of the shedding of
loads from the alternate sources. Also at Indian Point 2, Inspectors found that there
were no controls for limiting the time the Si system could be used to fill accumulator
tanks. During this process, If a safety systems actuation occurred, a portion of the Si
flow credited In the accident analysis would be diverted.
IN 98-22 June 17, 1998 Other Significant Issues Identified During the Course of the Inspections:
At Vermont Yankee, the licensee operated the RHR pumps at minimum flow values that
were significantly less than those recommended by the pump vendor.
At Robinson, control cables for all three Si pumps were routed in the same conduit In
violation of single failure and separation criteria.
At ANO, vendor-specified flow limits for the steam generators were not incorporated Into
plant procedures. As a result, operators were unaware that flow limits were exceeded
during a plant transient.
At D.C. Cook and Cooper, failures of instrument air regulators could result in the
Inoperability of redundant safety trains as a result of the over pressurization of air
operated valves.
Discussion
The majority of the issues identified have resulted from errors In the original design or design
modifications, calculational errors, inadequate corrective action, inadequate testing, and
documentation discrepancies. Many of the original design, design modifications, and
calculational errors can be attributed to the Inadequate specification and control of system and
discipline Interfaces, inadequately verified calculational assumptions, or the use of superseded
calculations. Licensees failed to evaluate the impact of calculational revisions on other
calculations, operating and test procedures. Changes to operating and test procedures were
not always reviewed against the existing calculations to ensure calculational assumptions were
still bounding. Also, the lack of a controlled, easily retrievable design basis has, in some
instances, hindered the ability of licensee engineers to Identify all design basis safety functions
of a system or component
Inadequate corrective actions have often resulted from weaknesses In root cause analyses or
from failing to assign ownership to engineering Issues. In addition, the depth of Internal self
assessments has not always been sufficient to Identify configuration management
weaknesses.
Additional details regarding the specific issues Identified during the NRC design Inspections can
be found In the following NRC Inspection Reports:
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IN 98-22 June 17, 1998 PLANT
INSPECTION REPORT#
ACCESSION #
Ginna
50-244/97-201
9710060295 H.B. Robinson 2
50-261/97-201
9708280104 Vermont Yankee
50-271/97-201
9709020247 Diablo Canyon
50-275197-202
9712030153 Three Mile Island
50-289196-201
9704210100
Arkansas Nuclear 1
50-313197-201
9803120197 Donald C. Cook I & 2
50-315/97-201
9712030232 St. Lucle 1 & 2
50-335196-201
9703280271
9703280234 Davis-Besse
50-346197-201
9709180174 Joseph M. Farley I & 2
50-348/97-201
9705230286 Washington Nuclear 2
50-397/96-201
9704250204 Indian Point 2
50-247/98-201
9804020083 Palisades
50-255/97-201
9801130395 Wolf Creek
50-482/97-201
9803030253 Cooper
50-298/97-201
9802260235 Perry 1
50-440/97-201
9706130253 This Information notice requires no specific action or written response. If you have any
questions about the Information In this notice, please contact one of the technical contacts lIsted
below or the appropriate Office of Nuclear Reactor Regulation (NRR) project manager.
Acting Director
Division of Reactor Program Management
Office of Nuclear Reactor Regulation
Technical contacts: Jeffrey Jacobson, NRR
301-415-2977 E-mall: jbj@nrc.gov
Attachment: Ust of Recently Issued NRC Information Notices 74iP6-e
% fiAY
DOCUMENT NAME: S:MDRPM SEC\\98-22.lN
Tech Ed. concurred on 5/15/98 To receive a copy of this document, Indicate In the box C-Copy wbo ettachment/enclosure ECopy wfth attachmentlenciosure N v No copy
lOFFICE
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DATE
06102/98
06/08/98
06/08198
06/09/98
06/ 0198
- See previous concurrenceOFFICIALRECORDCOPY
Attachment
June 17, 1998
Page 1 of I
LIST OF RECENTLY ISSUED
NRC INFORMATION NOTICES
Information
Date of
Notice No.
Subject
Issuance
Issued to
98-21 Potential Deficiency of
6/4198
All holders of operating licenses
Electrical Cable/Connection
Systems
for nuclear power reactors, except
those licensees who have
permanently ceased opertions
and have certified that fuel has
been permanently removed from
the reactor vessel.
98-20
98-19
98-18
98-17
98-16
98-15
Problems with Emergency
preparedness Respiratory
Protection Programs
Shaft Binding In General Electric
Type SBM Control Switches
Recent Contamination Incidences
Resulting from Failure to Perform
Adequate Surveys
Federal Bureau of Investigations
(FBI) Awareness of National
Security Issues and Responses
(ANSIR) Program
Inadequate Operational Checks
of Alarm Ratemeters
Integrity of Operator Licensing
Examinations
613/98
6/3/98
5/13/98
517/98
4130/98
4120/98
All holders of operating licenses
for nuclear power reactor;
non-power reactors; all
fuel cycle and material licensees
require to have an NRC approval
All holders of operating licenses
for nuclear power reactors
Part 35 Medical Ucensees
All U.S. Nuclear Regulatory
Commission fuel cycle and power
and non-power reactor licensees
All Industrial Radiography
Licensees
All holder of operating licenses
for nuclear power reactors except
those that have permanently
ceased operations and have
certified that fuel has been
permanently removed from the
reactor vessel
OL = Operating License
CP = Construction Permit
IN 98-22 June 17, 1998 PLANT
INSPECTION REPORT #
ACCESSIONA
Ginna
50-244/97-201
9710060295 H.B. Robinson 2
50-261/97-201
9708280104 Vermont Yankee
50-271/97-201
9709020247 Diablo Canyon
50-275/97-202
9712030153 Three Mile Island
50-289/96-201
9704210100
Arkansas Nuclear 1
50-313/97-201
9803120197 Donald C. Cook 1 & 2
50-315/97-201
9712030232 St. Lucie 1 & 2
50-335/96-201
9703280271
9703280234F
Davis-Besse
50-346/97-201
9709180174 Joseph M. Farley 1 & 2
50-348197-201
9705230286 Washington Nuclear 2
50-397/96-201
970425020n4 Indian Point 2
50-247/98-201
9804020083 Palisades
50-255/97-201
9801130395 Wolf Creek
50-482/97-201
9803030253 Cooper
50-298/97-201
9802260235 Perry 1
50-440/97-201
9706130253 This information notice requires no specific action or written response. If you have any
questions about the information in this notice, please contact one of the technical contacts listed
below or the appropriate Office of Nuclear Reactor Regulation (NRR) project manager.
JT
. IXoeoyActing Director
Division of Reactor Program Management
Office of Nuclear Reactor Regulation
Technical contacts: Jeffrey Jacobson, NRR
301-415-2977 E-mail: jbj~nrc.gov
Attachment: List of Recently Issued NRC Information Notices
DOCUMENT NAME: S:%DRPMSEC\\98-22.IN
Tech Ed. concurred on 5/15198 To1
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06/10/98
OFFICIAL RECORD COPY
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IN 98-XX
June xx, 1998 PLANT
INSPECTION REPORT#
ACCESSION##
.Ginna
50-244/97-201
9710060295 H.B. Robinson 2
50-261/97-201
9708280104 Vermont Yankee
50-271/97-201
9709020247 Diablo Canyon
50-275/97-202
9712030153 Three Mile Island
.50-289/96-201
9704210100
Arkansas Nuclear 1
50-313/97-201
9803120197 Donald C. Cook 1 & 2
50-315/97-201
9712030232 St. Lucie 1 & 2
50-335/96-201
9703280271
9703280234 Davis-Besse
50-346/97-201
9709180174 Joseph M. Farley 1 & 2
50-348/97-201
9705230286 Washington Nuclear 2
50-397/96-201
9704250204 Indian Point 2
50-247/98-201
9804020083 Palisades
50-255/97-201
9801130395 Wolf Creek
50-482/97-201
9803030253 Cooper
50-298/97-201
9802260235 Perry 1
50-440/97-201
9706130253 This information notice requires no specific action or written response. If you have any
questions about the information in this notice, please contact one of the technical contacts listed
below or the appropriate Office of Nuclear Reactor Regulation (NRR) project manager.
Jack W. Roe, Acting Director
Division of Reactor Program Management
Office of Nuclear Reactor Regulation
Technical contacts: Jeffrey Jacobson, NRR
301-415-2977 E-mail: jbj~nrc.gov
Attachment: List of Recently Issued NRC Information Notices
DOCUMENT NAME: G:\\TXK\\AEINSPI.
To receive a co
of this document, indicate in the box C=Copy wo attachmentlenclosure E=Copy with attachmentlenclosure N = No copy
OFFICE
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