Information Notice 1997-90, Use of Nonconservative Acceptance Criteria in Safety-Related Pump Surveillance Tests
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C. 20555-0001
December 30, 1997
NRC INFORMATION NOTICE 97-90: USE OF NONCONSERVATIVE ACCEPTANCE
CRITERIA IN SAFETY-RELATED PUMP
SURVEILLANCE TESTS
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 potential problems associated with safety-related pump surveillance testing. 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
Several recent inspections in the area of safety-related pump performance have resulted in the
issuance of Notices of Violation of Appendix B, Criterion Xi, "Test Control," of Part 50 of Title 10
of the Code of Federal Regulations (10 CFR Part 50) because licensees have concentrated on
inservice testing (IST) requirements without ensuring that design requirements were met.
There are two applicable primary requirements for safety-related pump testing: (1) to ensure
that Criterion Xl is met in that each safety-related pump achieves its minimum design-required
performance and (2) to ensure that each safety-related pump meets the requirements of
Section Xl, "Inservice Testing,' of the American Society of Mechanical Engineers Boiler and
Pressure Vessel Code (ASME Code). Criterion Xi of Appendix B to 10 CFR Part 50 requires
that a test program be established with written test procedures that incorporate the
requirements and acceptance limits contained in applicable design documents. Although
licensees have established IST acceptance criteria that meet the requirements specified in the
ASME Code, the criteria at some plants allowed safety-related pumps to degrade below the
performance assumed in the accident analyses.
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December 30, 1997
Description of Circumstances
In June 1996, the NRC staff found that the licensee for the Point Beach Nuclear Plant had failed
to ensure that the safety-related pump tests for several pumps had acceptance criteria that
incorporated the acceptance limits from applicable design documents (Inspection Report
50-266/96-06 and 50-301/96-06, dated September 5, 1996 [Accession 9609170044]). These
tests were used to ensure operability of the pumps, but the acceptance criteria were based only
on the ASME Code requirements. The NRC staff found that the average performance of the
service water pumps documented as acceptable by IST tests was 12.4 percent below the
average performance assumed in the large-break, loss of coolant accident (LOCA) analysis.
The NRC staff also noted that the IST-based acceptance criterion of 10-percent degradation
would have allowed the safety injection (SI) pumps to degrade more than the 5 percent
specified in the final safety analysis report (FSAR) small-break LOCA analysis. The NRC staff
found that other pumps would also have been allowed to degrade below minimum
requirements.
In December 1996, a system operational performance inspection (SOPI) at the Donald C. Cook
Nuclear Power Plant revealed a similar situation (Inspection Report 50-315/96-13 and
50-316/96-13, dated February 4, 1997 [Accession 9702070476]). NRC inspectors reviewed the
ability of the centrifugal charging pumps (CCPs) to satisfy design requirements. One of these
requirements, as specified in the plants technical specifications, was the ability to provide
adequate reactor coolant system boration from the refueling water storage tank (RWST).
During this inspection, the licensee performed a charging system flow calculation and
determined that the allowable CCP degradation to ensure adequate reactor coolant system
boration from the RWST was 8.5 percent for Unit 1 and 2.5 percent for Unit 2, rather than the
10-percent degradation limit specified by the ASME Code. The ASME Code acceptance limits
would have allowed the CCP performance to degrade to less than the technical specification
requirements.
In January 1997, a SOPI at the Kewaunee Nuclear Power Plant identified inadequate auxiliary
feedwater (AFW) and residual heat removal (RHR) pump surveillance testing (Inspection
Report 50-305/97-02, dated March 28, 1997 [Accession 9704040065]). In the AFW instance, the licensee was using an IST acceptance limit of 160 gallons per minute (gpm) delivered to the
steam generator for each of the pumps, although the license-basis requirement was 200 gpm
(not including 40 gpm of recirculation flow). The surveillance procedure used the original
vendor's pump performance curve with an acceptance criterion of 10-percent degradation from
December 30, 1997 these curves. When this criterion was applied to the AFW pumps, it was determined that all
AFW pumps were performing below license-basis requirements. This situation resulted in the
plant being outside of its accident analysis and requiring an operability determination for the
AFW pumps. For the RHR pumps, the IST lower band acceptance limit was 1330 gpm, which
was below the accident analysis required flow of 1400 gpm. In this case, the pumps were
shown to be operating right at the accident analysis limit.
In March 1997, an inspection at Grand Gulf Nuclear Station determined that the surveillance
test procedure acceptance criterion for one of the standby service water (SSW) pumps and the
high-pressure core-spray (HPCS) service water pump did not ensure that they were capable of
meeting their accident analysis criterion (Inspection Report 50-416/97-05, dated May 29. 1997
[Accession 9706030042]). The reference values for the surveillance test procedure acceptance
criterion for the SSW pumps were based on initial preoperational testing because this testing
demonstrated better performance than predicted by the vendor curves [124 pound per square
inch differential (psid) at 10,500 gpm]. The accident analysis was based on the vendor curves.
The SSW pumps were rebuilt in April 1996. Post-maintenance testing of the "A" pump revealed
hydraulic performance of 131.9 psid at 10,500 gpm, which was 5 percent below the previous
reference value. The licensee established this performance point as the new reference value.
With the ASME Code allowable degradation of 10 percent, the new minimum ASME Code
acceptance criterion at 10,500 gpm was 118.8 psid, which was below the value on the vendor's
curve. The accident analysis criterion for the HPCS service water pump of 83.8 psid at
700 gpm was based on the vendor's pump curves. The surveillance test procedure acceptance
criterion for this pump considered a developed head of 81.9 psid acceptable. Therefore, both
surveillance procedures would have considered their respective pumps to be operable, even
though their performance did not satisfy the accident analysis criterion.
In April 1997, an inspection at Fort Calhoun Station determined that the surveillance test
procedure acceptance criteria for pumps in five safety-related systems did not ensure that they
were capable of meeting their accident analysis performance criteria (Inspection Report
50-285/97-06, dated June 27,1997 [Accession 9707010011]). These pumps were the AFW
pumps, the high- and low-pressure Si pumps, the containment spray pumps, and the raw water
pumps. In particular, the motor-driven AFW pump's minimum performance criterion for delivery
to the steam generators was 1033 psid at 200 gpm. Since this analysis did not account for the
errors due to main steam safety valve set point tolerance and accumulation, the errors changed
the minimum performance requirements to 1079 psid at 200 gpm. However, using the
degradation of 10 percent allowed by the ASME Code, the surveillance test procedure
acceptance criterion considered the pump to be operable with 990 psid at 200 gpm.
Subsequent to these inspection findings, the licensee initiated a review of all safety-related
pumps in its IST program. Four other pumps were found to have surveillance procedure
minimum performance acceptance criteria lower than the criteria specified in their respective
accident analyses.
December 30, 1997 In May 1997, a SOPI at the Prairie Island Nuclear Generating Plant found that the lower
acceptance limits for the IST for the AFW pumps were below the most limiting accident analysis
value specified in the updated FSAR (Inspection Report 50-282/97-08 and 50-306/97-08, dated
July 16, 1997 [Accession 9707220149]). The maximum degradation allowed by design from
the reference pump curve was about 4 percent, whereas the IST acceptance criteria allowed
10-percent degradation. Because the AFW pumps appeared to be operating close to the
accident analysis limit (200.8 gpm vs. 200 gpm), a specific evaluation had to be performed to
ensure operability.
Discussion
These examples identify inadequacies in surveillance test procedure acceptance criteria that
had the potential for, and in some cases did result in, pumps not meeting their accident analysis
acceptance criteria. IST is intended to monitor degradation of components. The ASME Code
does not require that pumps be tested at design-basis conditions. Many licensees use the
ASME test to verify compliance with the ASME Code and the pump design requirements
contained in plant design-basis documentation such as the FSAR. The ASME Code allows a
specific percentage of degradation of pump hydraulic performance from an established
reference value before action must be taken. If the minimum design performance as specified
in the plant design documentation is more stringent than the ASME Code acceptance criteria, then the test acceptance criteria must be adjusted to avoid the actual pump performance being
allowed to degrade below the minimum acceptable design performance. In addition, licensees
need to ensure that original plant design-basis calculations, or revisions to these calculations, are properly integrated into surveillance test procedure acceptance criteria. The NRC
published guidance on this issue in NUREG-1482, "Guidelines for Inservice Testing at Nuclear
Power Plants." Section 5.6, "Operability Limits of Pumps," states that operability limits must
always meet, or be consistent with, licensing-basis assumptions in a plant's safety analysis.
Related guidance can also be found in ASME OM-SG-1997, Part 15, "Standard for
Performance Testing of Emergency Core Cooling Systems in Pressurized Water Reactor Plants
(OM Part 15)," which will be published in late 1997. A similar standard for boiling-water
reactors (OM Part 20) is scheduled to be published in 1998.
December 30, 1997 This information notice requires no specific action or written response. However, recipients are
reminded that they are required to consider industry-wide operating experience (including NRC
information notices) where practical, when setting goals and performing periodic evaluations
under Section 50.65, "Requirement for monitoring the effectiveness of maintenance at nuclear
power plants," to Part 50 of Title 10 of the Code of Federal Regulations. 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.
W. T-
/Jkck W. Roe, Acting Director
ivision of Reactor Program Management
Office of Nuclear Reactor Regulation
Technical contacts:
Eric R. Duncan, Rill
Robert M. Lerch, Rill
(630) 829-9739
(630) 829-9759 E-mail: erd@nrc.gov
E-mail: rml5@nrc.gov
Gerard F. O'Dwyer, Rill
Thomas F. Stetka, RIV
(630) 829-9624
(817) 860-8247 E-mail: gfo~nrc.gov
E-mail: tfs@nrc.gov
Joseph Colaccino, NRR
301-415-2753 E-mail: jxcl@nrc.gov
Attachment: List of Recently Issued NRC Info
ces
K-,--,
Attachment
December 30, 1997
Page 1 of I
LIST OF RECENTLY ISSUED
NRC INFORMATION NOTICES
Information
Date of
Notice No.
Subject
Issuance
Issued to
97-89
97-88
97-87
97-86
97-85
Distribution of Sources and
Devices Without Authorization
Experiences During Recent
Steam Generator Inspections
Second Retrofit to
Industrial Nuclear Company
IR 100 Radiography Camera, to Correct Inconsistency in
10 CFR Part 34 Compatibility
Additional Controls for
Transport of the Amersham
Model No. 660 Series
Radiographic Exposure Devices
Effects of Crud Buildup
and Boron Deposition on
Power Distribution and
Rupture in Extraction
Steam Piping as a
Result of Flow-Accelerated
Corrosion
Seismic Adequacy of
Thermo-Lag Panels
12/29/97
12116/97
12/12197
12/12/97
12111/97
12/11/97
12/10/97
All sealed source and device
manufacturers and distributors
All holders of OLs for pressurized- water reactors except those who
have permanently ceased
operations and have certified that
fuel has been permanently
removed from the reactor
All industrial radiography
licensees
Registered users of the Model
No. 660 series packages, and
Nuclear Regulatory Commission
industrial radiography licensees
All holders of OLs for pressurized- water reactors, except those
licensees who have permanently
ceased operations and have
certified that the fuel has been
permanently removed from the
reactor vessel
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
All holders of OLs for nuclear
power reactors
97-84
95-49, Sup. 1 OL = Operating License
CP = Construction Permit
December 30, 1997 This information notice requires no specific action or written response. However, recipients are
reminded that they are required to consider industry-wide operating experience (including NRC
information notices) where practical, when setting goals and performing periodic evaluations
under Section 50.65, "Requirement for monitoring the effectiveness of maintenance at nuclear
power plants," to Part 50 of Title 10 of the Code of Federal Regulations. 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:
Eric R. Duncan, Rill
(630)829-9739 E-mail: erd@nrc.gov
Robert M. Lerch, Rill
(630) 829-9759 E-mail: rml5@nrc.gov
Gerard F. O'Dwyer, Rill
(630) 829-9624 E-mail: gfo@nrc.gov
Thomas F. Stetka, RIV
(817) 860-8247 E-mail: ffsenrc.gov
Joseph Colaccino, NRR
301-415-2753 E-mail: jxcl@nrc.gov
Attachment: List of Recently Issued NRC Information Notices
OFC
PECB:DRPM*
TECH ED*
RII/DRS*
RIII/DRS
l
NAME
T. Greene
R. Sander
P. Pelke
M. Ring
DATE
12/11
/97
111/25/97 j 11/26 /97
12/ 1 /97 OFC
I EMEB:DE*
SC/EMEB:DE*
C/EMEB:DE*
RIV/DRS*
NAME
J. Colaccino
D. Terao
R. Wessman
T. Stetka
12/ 10 /97
12/ 12 /97
12/ 15 / 97
11/ 26 /97
- 1F.
i
OFC
SC/PECB:DRPM
C/PECB:DRPM
D/DRPM
NAME
R. Dennig*
S. Richards*
J. Roe*
DATE
12/
16 /97
12 / 22 /97
112/ 22 /97
- - See previous concurrence
[OFFICIAL RECORD COPY]
DOCUMENT NAME: 97-90.IN
IN 97-XX
December XX, 1997 This information notice requires no specific action or written response. However, recipients are
reminded that they are required to consider industry-wide operating experience (including NRC
information notices) where practical, when setting goals and performing periodic evaluations
under Section 50.65, "Requirement for monitoring the effectiveness of maintenance at nuclear
power plants," to Part 50 of Title 10 of the Code of Federal Regulations. 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:
Eric R. Duncan, Rill
(630)829-9739 E-mail: erd@nrc.gov
Robert M. Lerch, Rill
(630) 829-9759 E-mail: rml5@nrc.gov
Gerard F. O'Dwyer, Rill
(630) 829-9624 E-mail: gfo~nrc.gov
Thomas F. Stetka, RIV
(817) 860-8247 E-mail: tfs@nrc.gov
Joseph Colaccino, NRR
301-415-2753 E-mail: jxcl@nrc.gov
Attachment: List of Recently Issued NRC Information Notices
OFC
PECB:DRPM*
TECH ED*
RIII/DRS*
RIll/DRS*
NAME
T. Greene
R. Sander
P. Pelke
M. Ring
DATE
12/11
/97
11/25/97
11/26 /97
12/ 1 /97 OFC
EMEB:DE*
SC/EMEB:DE*
C/EMEB:DE*
RIV/DRS*
NAME
J. Colaccino
D. Terao
R. Wessman
T. Stetka
DATE
12/ 10
/97
12/ 12
/97
12/ 15 / 97
11/
26 /97 OFC
SC/PECB:DRPM
C/PECB:DRPM
D/DRPM
NAME
R. Dennig*
S. Richards
J.
oe
DATE
112/
16
/ 97
1 2-/
9
4V7 t
1 l-
/ 97
-
ee pevius
cncurenc
- - See previous concurrence
'JtS.
/qI'/? f 7
[OFFICIAL RECORD COPY]
DOCUMENT NAME: G:\\TAG\\INIST
2IN 97-XX
December XX, 1997 This information notice requires no specific action or written response. However, recipients
are reminded that they are required to consider industry-wide operating experience (including
NRC information notices) where practical, when setting goals and performing periodic
evaluations under Section 50.65, "Requirement for monitoring the effectiveness of
maintenance at nuclear power plants," to Part 50 of Title 10 of the Code of Federal
Regulations. 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:
Eric R. Duncan, Rill
630-829-9739 E-mail: erd@nrc.gov
Robert M. Lerch, Rill
630-829-9759 E-mail: rml5@nrc.gov
Gerard F. O'Dwyer, RilI
630-829-9624 E-mail: gfo@nrc.gov
Thomas F. Stetka, RIV
817-860-8247 E-mail: ffs@nrc.gov
Joseph Colaccino, NRR
301-415-2753 E-mail: jxcl@nrc.gov
Issued NRC Information Notices VI4k
Attachment: List of Recently
-
OFC
PECB:DRPM*
TECH ED*
RIII/DRS*
RIII/DRS*
NA ME
T. Greene
R. Sander
P. Pelke
M. Ring
DATE
[ 12/11
/97
11/25/97
11/26 /97
12/ 1 /97 l
OFC
l EMEB:DE*
SC/EMEB:DE*
C/EMEB:DE*
RIV/DRS*
NAME
J. Colaccino
D. Terao
R. Wessman
T. Stetka
DATE
l 12/ 10
/97 l 12/ 12
/97
12/ 15 / 97 l 11 /26 /97
-
..
II
OFC
SC)~ggowM
C/PECB:DRPM
D/DRPM
NAME
R.ti4endig
S. Richards
J. Roe
DATE
I --/ (v / 97 I
/ 97 I
/97 I - bee previous concurrence
[OFFICIAL RECORD COPY]
DOCUMENT NAME: G:\\TAG\\INIST
K-i- S~ 97-XX
December 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:
Eric R. Duncan, Rill
630-829-9739 E-mail: erdenrc.gov
Robert M. Lerch, Rill
630-829-9759 E-mail: rml5@nrc.gov
Gerard F. O'Dwyer, Rill
630-829-9624 E-mail: gfo@nrc.gov
Thomas F. Stetka, RIV
817-860-8247 E-mail: ffs@nrc.gov
Joseph Colaccino, NRR
301-415-2753 E-mail: jxclnrc.gov
Attachment: List of Recently Issued NRC Information Notices
[OFC
PECB:DRPM
TECH ED*
RIII/DRS*
RIII/DRS*
NAME
T. Greene (Q)
R. Sander
P. Pelke
M. Ring
l DATE
12/11
/97
111/25/97
11/26 /97 J 12/
_
/ 97 l OFC
EMEB:DE
SC/EMEB:DE
C/EMEB:DE)
RIV/DRS*
NAME
J. Colaccino
7i
D. Terao
an J
R. W
T. Stetka
lDATE
_la_
_
_ /97 fN
Z-f 97 l A/
1s / 97 l 11 / 26 /97 OFC
SC/PECB:DRPM
C/PECB:DRPM
DIDRPM
NAME
R. Dennig
S. Richards
J. Roe
DATE
I_____
I
/97
1 /
/97 J
- - See previous concurrence
[OFFICIAL RECORD COPY]
DOCUMENT NAME: G:kTAG\\INIST
(301) 41'
3 E-mail: jxcl@nrc.gov
Attachment: List of Recently Issued NRC Information Notices
OFC
PECB:DRPM
TECH EDITOR
RIII/DRS
RIII/DRS
NAME
T. Greene
M!svW~
M. Ring
P. Pelke
DATE
I
197
/(/1 197 I
/97 I
/97 OFC
IRillIDRS
EMEB:DE
SLIEMEB:DE
CIEMEB:DE
NAME jJ. Colaccino
T. Derao
JR. Dennig
R. Wessman
DAE
j
197 I
/97 i
/971 / /97 OFC
C/PECB:DRPM
D/DRPM
NAME
l S. Richards
J. Roe
DATE
l
/97 I
/97 l
[OFFICIAL RECORD COPY]
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11/26/97 09:57am
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Thomas Stetka
WND2.WNP4(TAG)
11/26/97 1:44pm
IN REVIEW (TAC M99024) -Reply
I reviewed the proposed IN and concur with no comments.