Information Notice 1997-81, Deficiencies in Failure Modes and Effects Analyses for Instrumentation and Control Systems
\\
I
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C. 20555-001
November 24, 1997
NRC INFORMATION NOTICE 97-81: DEFICIENCIES IN FAILURE MODES AND EFFECTS
ANALYSES FOR INSTRUMENTATION AND CONTROL
SYSTEMS
Addressees
All holders of operating licenses for nuclear power reactors except those who have 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 design inadequacies in safety-related instrumentation and control systems in
which the original failure modes and effects analysis (FMEA) failed to identify the inability to
perform intended function(s) in the presence of a single failure, which defeats the system
independence and redundancy. It is expected that recipients will review the information for
applicability to their facilities. However, suggestions contained in this information notice are not
NRC requirements; therefore, no specific action or written response is required.
Description of Circumstances
Five events have been identified related to the design deficiency noted above:
1.
On July 1, 1997, while Crystal River Unit 3 was in mode 5, the licensee discovered that in
a postulated design-basis loss-of-coolant accident (LOCA), concurrent with a loss of
offsite power (LOOP) and a single failure (such as a loss of the train A dc bus), the
train B engineered safeguards system actuation (ESSA) signal could not be bypassed.
The inability to bypass the train B ESSA signal removes the operators ability to restore
control complex chillers and causes certain train B high pressure injection (HPI) valves to
remain open. These valves cannot be closed remotely when the LOCA is caused by a
break on one of the HPI lines. This condition could lead to a potential for inadequate core
cooling. In addition, the operators may not be able to cope with the event, because the
station emergency operating procedures (EOPs) did not contain adequate guidance
(LER 97-21, Accession No: 9708290125).
2.
On July 25, 1997, while Three Mile Island, Unit 1 (TMI-1) was at 100-percent power, the
TMI licensee discovered that in a postulated condition of a large-break LOCA, concurrent
with LOOP and a single failure (such as a loss of the train A dc bus), the engineered
safeguards (ES) system components of train B would be actuated. However, because of
, rf as F aurlde q1-8D III I X
7111901
-IN 97-81 November 24, 1997 the loss of power to ESSA system channels A and C (due to loss of the train A dc bus),
the operators would not be able to bypass the train B ESSA signal. The plant's design for
accident conditions with loss of dc bus "A", did not consider the need for throttling low- pressure injection and building spray pump flow to ensure adequate net positive suction
head when the suction is transferred to the reactor building sump from the borated
water storage tank. Maintaining the minimum required flow ensures that these pumps
remain operable, so that the cooling requirements of the reactor core and reactor building
are met. However, the plant design upon loss of train A dc power, prevents the operators
from taking manual control of the valves required to transfer low-pressure injection and
reactor building spray pump suction from the refueling water storage tank (RWST) to the
reactor building sump upon indication of a low-water-level condition in the RWST
(LER 97-009, Accession No: 9709020257).
3.
On May 15, 1997, while Waterford Unit 3 was at 100-percent power, the licensee
discovered that in the postulated conditions of a LOCA with one RWST-level monitoring
channel placed in a tripped state [as allowed by the Technical Specifications(TS)], if a
single failure, such as a failure of another RWST-level channel occurs, a potential for
premature initiation of the recirculation mode exists. In another situation, with one
channel of steam generator (SG) differential pressure (DP) instrumentation associated
with the emergency feedwater actuation signal (EFAS) placed in a tripped state, an event
such as a main steam line break or a feedwater line break concurrent with a single failure
such as loss of another SG DP instrument channel, results in a potential for not isolating
the faulted SG from the emergency feedwater supply line (LER 97-16, Accession
No: 9706180379).
4.
On September 16, 1997, while Arkansas Nuclear One, Unit 2 (ANO-2) was at 100-
percent power, the licensee discovered that a potential for premature actuation of the
recirculation mode exists in case of a LOCA concurrent with one of the RWST-level
instrumentation channels in a tripped state and a failure in another RWST-level
monitoring channel. In another situation, the automatic Isolation of a faulted SG during
certain main steam line or feedwater line breaks will not occur If one instrumentation
channel monitoring SG DP is in a tripped state and another SG DP channel has failed
(LER 97-03, Accession No: 9706190213).
5.
On October 30, 1996, while ANO-2 was at 100-percent power, the licensee discovered
that while one plant protective system (PPS) channel is in bypass, a scenario consisting
of a LOOP concurrent with a single failure, such as a loss of the train A dc bus, would
result in a failure of certain engineered safeguard function (ESF) systems to actuate
automatically. ESF systems affected are the containment Isolation system (CIS),
containment spray system (CSS), and emergency feedwater system (EFWS). The
consequence of a dc bus failure alone could lead to the same failures with loss of off-site
power and loss of on-site power in the affected train (LER 96-04-01, Accession
No: 9702120360).
Discussion
The first two events described above are examples in which a single failure in a train (loss of dc
bus A) not only prevented the train A safety system from performing the intended design
function(s), but also prevented equipment in train B safety systems from performing their
IN 97-81 November24, 1997 intended design function(s). The basic logic design of the ES system at both Crystal River
Unit 3 and Three Mile Island Unit 1 consists of three primary analog monitoring instrumentation
channels, powered from two redundant battery-backed dc buses and provides an ESSA signal
to train A and train B equipment in a two-out-of-three logic scheme. Two of the three
monitoring channels are powered from the train A dc bus and the third channel is powered from
the train B dc bus. Since the logic arrangement is based on a two-out-of-three configuration
and the train A dc bus feeds two channels, if dc bus A is lost, the ESSA signal cannot be
bypassed because the bypass circuitry is powered from the same source as the signal initiation
channels. This problem could exist at any other plant at which the ESSA signal is generated
with a two-out-of-three logic configuration with the three monitoring channels fed from only two
power sources.
As described in events 3 and 4 above, a failure in a single instrument channel with one other
instrument channel of the same function placed in the "tripped" state as permitted by the TS,
created a unique control-logic configuration that could prevent certain safety system(s) from
performing intended design function(s) for mitigating a LOCA or steam/feedwater line break
event(s). The ANO-2 vital power design for the plant protective system (PPS) consists of one
emergency diesel generator (EDG), one battery/dc distribution system and two inverters for
each power division. The ESF system actuation logic is based on a two-out-of four
configuration. The design configuration for some measurement channels (steam generator
pressure, pressurizer pressure, containment pressure) is such that upon loss of power, these
channels do not fail to their safe state and, therefore, during a loss-of-power condition, are
unable to automatically actuate the associated ESF systems if needed.
As described in event 5 above, during a postulated LOOP event concurrent with a loss of one
dc bus, the CIS, CSS, and EFWS will fail to actuate if one PPS channel fed from the operable
dc bus is in a bypassed state as permitted by the TS. The result of this existing condition is that
during the period when one PPS channel is bypassed, the plant could have been operated
outside its original design basis since many required automatic safety functions would have
been unavailable. If a channel is not in bypass, all PPS functions except the function that
controls feedwater flow/SG level to prevent SG overfill, will be available even if one power
division is lost concurrent with a LOOP. The SG overfill prevention feature Is not a required
safety function for the postulated loss-of-power scenario.
The previous ANO-2 TS allowed one PPS channel to be in bypass for up to 48-hours to perform
maintenance or testing based upon the low probability of a fault such as loss of a power division
affecting more than one channel during the 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> interval. Thus, the failure of the automatic
actuation capability of some ESFAS functions was considered sufficiently unlikely for the limited
time one channel was permitted to be in bypass. Amendment 159 to the ANO-2 TS issued on
April 3, 1995, increased the allowed time In bypass for one PPS channel during plant operation
at full power from "48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />" to "until the next cold shutdown." Reviews to support this
amendment conciuded that bypassing of a specific protective channel combined with a single
' Ji
IN 97-81 November 24, 1997 failure would not prevent required protective actions. The discovery of the potential
unavailability of required protective actions under the conditions described above indicated
that the FMEA providing an acceptable basis for Amendment 159, was in error.
The FMEA did indicate that any of those PPS channels that generate a trip on a decreasing
value of the process signal will actuate on loss of power (because the logic will perceive the
loss of power as a decrease in the value of the process signal). However, the FMEA failed to
recognize that those PPS channels that generate a trip on an increasing value of the process
signal will never actuate if power Is lost to the measurement channel. Thus, the FMEA was in
error In its conclusion that "the vital ac power system did not have a single failure mechanism
that could cause failure of two vital AC power channel inputs" (a minimum of two channels is
required for a two-out-of four logic).
To alleviate the problem described in events I and 2 above, the Crystal River licensee decided
to revise the EOPs to provide procedural guidance on a loss of ES train bypass capability upon
loss of dc bus A. The revised EOPs will provide guidance to recognize the failure and an
alternative method to bypass the B train ESSA signal. This will allow the operators to regain
control of the necessary ES equipment to assure adequate high-pressure injection flow to the
reactor. It will also allow operators to throttle HPI flow when the subcooling margin is restored
to maintain RCS pressure below pressurized thermal shock limits. The TMI licensee is also
revising the EOPs to (1) proceduralize methods to diagnose an inability to bypass the ESSA
signal, (2) provide alternate methods to energize the train A dc bus so that the ESSA signals
can be bypassed, and (3) identify required operator actions necessary to achieve minimum flow
requirements for reactor core and reactor building cooling.
For the concerns described in events 3 and 4 above, relating to events (LOCA, main steam line
break or feedwater line break) with one instrument channel in a tripped state while another
channel has failed, the ANO-2 licensee provided the needed temporary administrative controls
until a permanent fix involving a TS change is implemented. The Waterford licensee has also
evaluated the conditions related to channel trip and provided temporary administrative controls
until a permanent fix involving a TS change is implemented.
To address the problem described in event 5 above, the ANO-2 licensee evaluated the
condition and concluded that the PPS will continue to perform its design function if a channel is
bypassed for no longer than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. As a short term corrective action, the licensee
established administrative controls which prevent a PPS channel from remaining in bypass for
more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. A "Night Order" was issued to the ANO-2 operations personnel to remind
them of the guidance contained in the EOPs regarding actions to mitigate potential SG overfill
events. The licensee has indicated that its long term corrective action will be to install hardware
modifications which will allow a single channel to remain in bypass indefinitely with no loss of
safety function. The licensee will install these modifications by the end of refueling outage
2R13 in 1999, and incorporate changes to the Safety Analysis Report (including the FMEA for
dc bus failures) to reflect those modifications and resolve issues discovered during the root
cause evaluation of the condition following that outage.
The preceding examples describe inadequacies in the "design-process," in which the original
FMEA failed to identify and correct potential failure modes In the plant design, and the
'J GIN 97-81 November 24, 1997 independent design verification process failed to detect the FMEA inadequacies. This concern
is addressed in Title 10, Code of Federal Regulations, Part 50, Appendix B, Criterion IlIl,
"Design Control," which stipulates, "Measures shall be established to assure that applicable
regulatory requirements and the design basis, as defined in 50.2 and as specified In the license
application, for those structures, systems, and components to which this appendix applies are
correctly translated into specifications, drawings, procedures, ... components. The design
control measures shall provide for verifying or checking the adequacy of design ....
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.
a k W. Roe, Acting Director
(Diviion of Reactor Program Management
ice of Nuclear Reactor Regulation
Technical contacts: S.V. Athavale, NRR
301-415-2974 E-mail: svalnrc.gov
Thomas Koshy, NRR
301-415-1176 E-mail: txk~nrc.gov
Attachment: List of Recently Issued NRC Information Notices
v-/ Attachment
November 24, 1997 LIST OF RECENTLY ISSUED
NRC INFORMATION NOTICES
Information
Date of
Notice No.
Subject
Issuance
Issued to
97-80
97-79
Licensee Technical
Specifications
Interpretations
Potential Inconsistency
in the Assessment of
the Radiological Conse- quences of a Main Steam
Line Break Associated
with the Implementation
of Steam Generator Tube
Voltage-Based Repair
Criteria
11/21/97
11/20/97
All holders of OLs for
nuclear power reactors
All holders of OLs for
pressurized-water reactors
implementing a steam
generator tube voltage- based repair criteria in
accordance with the
guidance presented in
'Voltage-Based Repair
Criteria for Westinghouse
Steam Generator Tubes
Affected by Outside
Diameter Stress Corrosion
Cracking,' issued August 3,
1995
97-78
Crediting of Operator
Actions in Place of
Automatic Actions and
Modifications of Operator
Actions, Including Response
Times
10/23/97
All holders of OLs for
nuclear power reactors
except those who have
permanently ceased
operations and have
certified that fuel has
been permanently removed
from the reactor vessel
OL = Operating License
CP = Construction Permit
\\lIN
97-81 November24, 1997 independent design verification process failed to detect the FMEA inadequacies. This concern
is addressed in Title 10, Code of Federal Regulations, Part 50, Appendix B, Criterion IlIl,
"Design Control," which stipulates, "Measures shall be established to assure that applicable
regulatory requirements and the design basis, as defined in 50.2 and as specified in the license
application, for those structures, systems, and components to which this appendix applies are
correctly translated into specifications, drawings, procedures, ... components. The design
control measures shall provide for verifying or checking the adequacy of design ...."
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.
original signed by
Jack W. Roe, Acting Director
Division of Reactor Program Management
Office of Nuclear Reactor Regulation
Technical contacts: S.V. Athavale, NRR
301-415-2974 E-mail: sval
nrc.gov
Thomas Koshy, NRR
301 415-1176 E-mail: txk@nrc.gov
Attachment: List of Recently Issued NRC Information Notices
Tech Editor has reviewed and concurred on 10/1/97
- SEE PREVIOUS CONCURRENCES
DOCUMENT NAME: 97-81.IN
E -
P...
.kAI
- .maMnln.ra*
-
Min .nn
To receiv, a copY of Uils document, mimc96 M tne DOX:
iA. - Copy WIUOU
81 UILUU
IU1 u
re1IUM.U8US
e
-
-
..
...
..
-
-..
-
OFFICE
HICB:DRCH
I
SC:HICB
BC:HICB
I
PECB:DRPM
I
SC:PECB L
NAME
SVAthavale*
JMauck*
JSWermiel*
TKoshy*
RDennig*
DATE
09/24/97
09/24/97
09/24/97
11/17/97
10/09/97 OFFICE
C:PECB/DRPM
[
(A)Q:DRPM
I_
I
_I___
_NAME
_SRichards*
L
lDATE l10/14/97 .
11A4/97
OFFICIAL RECORD COPY
K>
IN 97-xx
October xx, 1997 application, for those structures, systems, and components to which this appendix applies are
correctly translated into specifications, drawings, procedures, ... components. The design
control measures shall provide for verifying or checking the adequacy of design ...."
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: S.V. Athavale, NRR
301-415-2974 E-mail: sval@nrc.gov
Thomas Koshy, NRR
301-415-1176 E-mail: txk@nrc.gov.
Attachment: List of Recently Issued NRC Information Notices
- SEE PREVIOUS CONCURRENCES
DOCUMENT NAME: INFORMATION NOTICE (97-103)
T. m,,.,. a
- nV af thif t
nument- hintlata hI the hax: C
-
ConY without attachment/enclosure
SEw -
Copy with attachment/enclosure
'N'
-
No copy
OFFICE
HICB:DRCH
I
SC:HICB
I
BC:HICB
l PECB:DRPM
SC:PECB L
NAME
SVAthavale*
JMauck*
JSWermiel*
TKoshy*
RDennig*
DATE
09/24/97
09/24/97
09/24/97
10/08/97 v
,7 10/09/97 OFFICE
C:PECB/DRPM
I
(A)D:DRPM
L
_ _
NAME
SRichards
JWRoe
l
DATE
09/ /97
10/ /97
OFFICIAL RECORD COPY
IN 97-XX
November , 1997
Page5of 5 independent design verification process failed to detect the FMEA inadequacies. This concern
is addressed in Title 10, Code of Federal Regulations, Part 50, Appendix B, Criterion Ill, "Design
Control," which stipulates, "Measures shall be established to assure that applicable regulatory
requirements and the design basis, as defined in 50.2 and as specified in the license
application, for those structures, systems, and components to which this appendix applies are
correctly translated into specifications, drawings, procedures, ... components. The design
control measures shall provide for verifying or checking the adequacy of design .....
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: S.V. Athavale, NRR
301-415-2974 E-mail: sval@nrc.gov
Thomas Koshy, NRR
301-415-1176 E-mail: txkenrc.gov
Attachment Ust of Recently Issued NRC Information Notices
Tech Editor has reviewed and concurred on 10/1/97
- SEE PREVIOUS CONCURRENCES
DOCUMENT NAME: INFORMATION NOTICE (97-103)
To r--
a *
..
fa this dnorntnt h3nsta In h
bharox 'C
- ConY without sttachmantenclosure
'El -
Copy with ottachment/enclosure
N -
No copy
OFFICE
fHICB:DRCH
I
SC:HICB
I
BC:HICB
H
PECB:DRPM
I
SC:PECB
NAME
SVAthavale*
1JMauck*
JSWermiel*
TKoshy*
RDennig*
DATE
09/24/97
09/24/97
09/24/97
10/08/97
10/09/97 OFFICE
C:PECB/DRPM
(A)D:DRPM
[___
NAME
SRichards*
JWRoe
DATE
10/14/97
11/
/97
OFFICIAL RECORD COPY
IN 97-xx
October xx, 1997
Page 5of 5 independent design verification process failed to detect the FMEA inadequacies. This concern is
addressed in Title 10, Code of Federal Regulations, Part 50, Appendix B, Criterion lii, "Design
Control," which stipulates, "Measures shall be established to assure that applicable regulatory
requirements and the design basis, as defined in 50.2 and as specified in the license
application, for those structures, systems, and components to which this appendix applies are
correctly translated into specifications, drawings, procedures, ... components. The design control
measures shall provide for verifying or checking the adequacy of design ...."
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: S.V. Athavale, NRR
301-415-2974 E-mail: sval
nrc.gov
Thomas Koshy, NRR
301-415-1176 E-mail: txk@nrc.gov
Attachment: List of Recently Issued NRC Information Notices
- SEE PREVIOUS CONCURRENCES
DOCUMENT NAME: INFORMATION NOTICE (97-103)
-
B--.L.
s -.-
.
L. 6ke
.i.n.- *l.^ -
_
..
k. .
.k.rhr
aln..
- E- _
Inv
with .ntschnmntthnnn.,r.
-
Nn rnnw
OFFICE
HICB:DRCH
SC:HICB
BCH
L--LB PECB:DRPM
SC:PECB L
NAME
SVAthavale*
JMauck*
JSWermiel*
TKoshy*
RDennig*
DATE
09/24/97
09/24/97
09/24/97
10/08/97
10/09/97 OFFICE
C:PECB/DRPM
U(A)DDRPM
NAME
I~chards
Mo_
DATE
/A /97 l10/
/97 A
I V
OFFICIAL RECORD COPY
IN 97-xx
October xx, 1997 independent design verification process failed to detect the FMEA inadequacies. This concern
is addressed in Title 10, Code of Federal Regulations, Part 50, Appendix B, Criterion IlIl, "Design
Control," which stipulates, "Measures shall be established to assure that applicable regulatory
requirements and the design basis, as defined in 50.2 and as specified in the license
application, for those structures, systems, and components to which this appendix applies are
correctly translated into specifications, drawings, procedures, ... components. The design
control measures shall provide for verifying or checking the adequacy of design ...."
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: S.V. Athavale, NRR
301-415-2974 E-mail: sval nrc.gov
Thomas Koshy, NRR
301-415-1176 E-mail: txk@nrc.gov
Attachment: List of Recently Issued NRC Information Notices
- SEE PREVIOUS CONCURRENCES
DOCUMENT NAME: INFORMATION NOTICE (97-103)
T- .lv
l -
.
-.
ta
ns
nLtP
In th
.
b-
=C r nny, without attechmentenclosure
,e sCGIv S..t
w-
-.
-
-unn ,
-
-X
n
Ad
l-- VA
-
-M -.
'E'
-
CoDy with attachment/enclosure
WN -
No copy
OFFICE
HICB:DRCH
[I
S ICB
I
BC:HICB
H
PECB:DRPM
H
C:PECB
NAME
SVAthavale*
jMauck*
JSWermiel*
TKoshy*
RDennig*
DATE
09/24/97
09/24/97
09/24/97
10/08/97
10/09/97 OFFICE
C:PECB/DRPM
(A)D:DRPM
I
_
I
NAME
SRichards
JWRoe
DATE
09/ /97
10/ /97
-
d nMr-T fT
AlIr rnnnn
onfrv
`44j /V/1/17 u01-bIUAL KLUUKU burl
IK>
IN97-xx
October xx, 1997 independent design verification process failed to detect the FMEA inadequacies. This concern
is addressed in Title 10, Code of Federal Regulations, Part 50, Appendix B, Criterion l1l,
"Design Control," which stipulates, "Measures shall be established to assure that applicable
regulatory requirements and the design basis, as defined in 50.2 and as specified in the license
application, for those structures, systems, and components to which this appendix applies are
correctly translated into specifications, drawings, procedures, ... components. The design
control measures shall provide for verifying or checking the adequacy of design ...
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: S.V. Athavale, NRR
(301)415-2974 E-mail: sval@nrc.com
Thomas Koshy, NRR
(301)415-1176 txk@nrc.com
Attachments: 1. List of Recently Issued NRC Information Notices
DOCUMENT NAME: INFORMATION NOTICE (97-103)
.
.A_..
.-
s K.
in
K'w.
- E-
-
r-
'F,
-
r
V with attnrhmantlenclosure
'N' -
No coov
TO mecelve a Copy
%1r
.....
....
t1 Re s
x
vpywsvta~al~a~nwuw
-~-
t- OFFICE
HICB:DRCH
S
C:ZIC
BC:HICB
PECB:DRPM
I0
A
S
NAME
SVAthavale*
JMauck*
JSWermiel*
TKoshy
& Z
IRb&iNly l
DATE
09/24/97
09/24/97
09/24/97
10/
97
10/
/97 OFFICE
C:PECB/DRPM
l
AD:DRPM
NAME
SRichards
JWRoe
DATE
09/
/97
10/
/97
OFFICIAL RECORD COPY
IN97-xx
October xx, 1997 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: S.V. Athavale, NRR
(301)415-2974 E-mail: sval@nrc.com
Thomas Koshy, NRR
(301)415-1176 txk@nrc.com
Attachments: 1. List of Recently Issued NRC Information Notices
DOCUMENT NAME: INFORMATION NOTICE (97-103)
To receive a copy of this document. Indicate In the box:
Ca -
Coov without attachment/enclosure
'E'
-
Copy with attachmentlenclosure
'N'- No copy
[OFFICE
HICB:DRCH
SC:HICB
LBC HICB
I
PECB:DRPM
I
SC:PECB
I
gNAME
SVAthavale*
JMauck*
JSWermiel*
TKoshy 60tc' _RDennig
IDATE
09/24/97
09/24/97
09/24/97
10/ o/97
10/ /97 lOFFICE 1C:PECB/DRPM
AD:DRPM
L
IL
NAME
[SRichards
lJWRoe
l
DATE
09/ /97
10/
/97
'1
OFFICIAL RECORD COPY
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:
S.V. Athavale, NRR
(301)415-2974
Thomas Koshi, NRR
(301) 415-1176 Attachments:
1. List of Recently Issued NRC Information Notices
DOCUMENT NAME:
INFORMATION.NOTICE
(97-103)
9/
To edve a CODY of this documant. kullexte ItheX box. 'C'
a-o
wiYWthouit sttachmentenc nclsur
E' a GDYv with mftachrnenthk naum
- 111 - No COY v
OFFICE
HICB:DRCH l
l
$ICBI
I BC:HICB4 E
I PECB:DRPM
W
DDRC2/-
NAME
SVAthaval
-
JSWermilen
TKoshi
RLSpSard
DATE
09/.A-/97
09/1 97
09/2. /97
09/ /97
09l 2S'/97 OFFICE
ABC:PECB/DRPM
[AD:DRPM
U-UI
NAME
EFGoodwin
JWRoe
DATE
09/ /97
09/ /97 OFFICIAL RECORD COPY