Information Notice 1997-07, Problems Identified During Generic Letter 89-10 Closeout Inspections
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UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C. 20555-0001 March 6, 1997 NRC INFORMATION NOTICE 97-07: PROBLEMS IDENTIFIED DURING GENERIC
LETTER 89-10 CLOSEOUT INSPECTIONS
Addressees
All holders of operating licenses or construction permits for nuclear power reactors.
Purpose
The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice to alert
addressees to the general conclusions derived from NRC inspections of programs developed
at nuclear power plants in response to Generic Letter (GL) 89-10, "Safety-Related Motor- Operated Valve Testing and Surveillance." 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 response to operating events, research results, and the findings in NRC Bulletin 85-03,
"Motor-Operated Valve Common Mode Failures During Plant Transients due to Improper
Switch Settings," the NRC staff requested in GL 89-10 that holders of nuclear power plant
operating licenses and construction permits ensure the design-basis capability of their safety- related motor-operated valves (MOVs) by periodically reviewing MOV design bases, verifying
MOV switch settings, testing MOVs under design-basis conditions where practicable, improving evaluations and corrective actions associated with MOV failures, and determining
trends of MOV problems. The NRC staff issued seven supplements to GL 89-10 to provide
further guidance to the industry on implementation of the generic letter.
On September 18, 1996, the NRC staff issued GL 96-05, "Periodic Verification of Design- Basis Capability of Safety-Related Motor-Operated Valves." GL 96-05 contains detailed
guidance on the development of long-term programs to ensure the design-basis capability of
safety-related MOVs. It also includes updated information on long-term MOV performance.
In the area of MOV periodic verification, the recommendations of GL 96-05 supersede those
of GL 89-10.
Over a number of years, industry and NRC activities associated with GL 89-10 have
increased, reflecting both the evolution of technological development and experience gained
over time and the rising expectations of both the industry and the NRC staff. Activities have
included generic communications, workshops, MOV Users' Group meetings, symposia on
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IN 97-07 March 6, 1997 pumps and valves, and a massive MOV testing and analysis effort by the Electric Power
Research Institute (EPRI). As a result, information on MOV performance has been widely
disseminated over the past few years.
Description of Circumstances
Most nuclear power plant licensees have notified the NRC that they consider their programs
to verify the design-basis capability of safety-related MOVs in response to GL 89-10 to be
complete. The NRC staff has been conducting inspections of the development, implementation, and completion of these programs. In performing the inspections, the NRC
staff has followed Temporary Instruction (TI) 2515/109, "Inspection Requirements for Generic
Letter 89-10, Safety-Related Motor-Operated Valve Testing and Surveillance." The NRC staff
recently updated this TI to provide guidance on GL 89-10 closeout inspections and on the
scope of GL 89-10 programs. The NRC staff plans to complete its review of the GL 89-10
programs at most nuclear plants in 1997.
Through MOV testing, analyses, and operational events over the past few years, the nuclear
industry and the NRC staff have identified weaknesses in the original design, manufacture, maintenance, and testing of safety-related MOVs. During inspections to review completion of
GL 89-10 programs, the NRC staff has found that some licensees have not fully verified the
design-basis capability of their safety-related MOVs. For example, the NRC staff has found
that little testing bases existed in support of original assumptions by some licensees (and
actuator and valve manufacturers) for friction coefficients and efficiencies affecting thrust and
torque requirements and actuator output when sizing and setting MOVs. As a result, licensees have had unexpected difficulty in demonstrating to the staff that they have
adequately completed their GL 89-10 programs.
When reviewing the development and implementation phases of the GL 89-10 programs, the
NRC inspectors identified specific items and concerns that needed attention before
completion of the programs. These items and concerns are discussed in the inspection
reports prepared by the NRC staff. During inspections to evaluate completion of the
GL 89-10 programs, the NRC staff found that some licensees had not resolved the items and
concerns identified in the previous inspection reports. In addition, some licensees had not
recognized that the MOV program has to be kept up to date on the basis of new information
on MOV performance.
In GL 89-10, the NRC staff recommended that MOVs within the scope of the generic letter
be tested under design-basis conditions where practicable. In Supplement 6 to GL 89-10,
the NRC staff provided guidance for licensees on grouping MOVs that were not practicable to
test dynamically. Some licensees have also chosen to group MOVs to minimize the amount
of dynamic testing under their GL 89-10 programs. The MOV grouping guidelines
recommend that dynamic test data be obtained on a reasonable sample of MOVs and that
the resulting information be applied to the remaining MOVs in the group.
During GL 89-10 closeout inspections, the NRC staff found that some licensees provided
weak justification for the design-basis capability of MOVs that have not been dynamically
IN 97-07 March 6, 1997 tested. As stated in Supplement 6 to GL 89-10, the NRC staff considers plant-specific test
data to be the best source of information when attempting to justify the design-basis
capability of MOVs. The plant-specific test data would be obtained from the specific MOV
being evaluated or, if testing was not practicable, from other similar MOVs under similar fluid
conditions at the plant.
In developing the justification for the design-basis capability of MOVs that are not dynamically
tested, it is important to consider the extent and reliability of the information being applied to
the MOV under evaluation. For example, MOVs of similar manufacture and fluid conditions
have been found to have a range of performance characteristics. Therefore, reliance on data
from a few MOVs tested under industry programs or at other plants might be insufficient to
justify the design-basis capability of similar MOVs at a specific plant. Plant-specific testing
needs to be repeatable or at least validated through the performance of statistically valid
testing.
If MOV-specific data and plant-specific data for similar MOVs are not available, other sources
of information appropriate for the plant's MOVs must be found. In the search for this
information, the range of performance under similar fluid conditions needs to be considered.
For example, EPRI made significant efforts to predict bounding thrust requirements through
its program of separate effects tests, flow loop testing, and analytical methodology. In a
safety evaluation (SE) dated March 15, 1996 (Accession number 9608070280), the NRC staff
approved the EPRI MOV Performance Prediction Methodology (PPM) when used in
accordance with certain conditions and limitations. Selective application of the EPRI test data
or methodology might not be reliable without full consideration of the NRC staff SE on the
EPRI PPM. Further, the NRC staff has determined that it is difficult to select the specific
point of flow isolation of tested valves and to apply flow isolation data from one valve to
another.
Key parameters to be addressed in verifying the design-basis capability of MOVs are valve
friction coefficients (i.e., valve factor), stem friction coefficients, and load sensitive behavior
(i.e., rate-of-loading effects). During GL 89-10 closeout inspections, the NRC staff found that
some licensees were using qualitative arguments to justify assumptions for these quantitative
parameters. As discussed previously, MOVs that have not been dynamically tested need to
have adequate justification for their design-basis capability. The most reliable source of
information on valve friction coefficients, stem friction coefficients, and load sensitive behavior
is the specific licensee's plant. Licensees can best demonstrate the validity of their
assumptions for these parameters by ensuring that sufficient test data are available for their
specific plants and by analyzing the data for the plant- and valve-specific parameters.
Pressure locking and thermal binding of gate valves were particular MOV performance
problems identified in GL 89-10. To some extent, the NRC staff has addressed licensee
responses to this issue in GL 89-10 inspections. The NRC staff issued GL 95-07, "Pressure
Locking and Thermal Binding of Safety-Related Power-Operated Gate Valves," to provide
specific recommendations to licensees and to request their responses in regard to pressure
locking and thermal binding of gate valves. GL 95-07 also requested that licensees submit
IN 97-07 March 6, 1997 their responses to this issue separate from their submittals on their GL 89-10 programs.
Nevertheless, the NRC staff may request information from licensees during GL 89-10
inspections regarding the operability of specific MOVs found to be susceptible to pressure
locking or thermal binding.
On February 28, 1992, the NRC staff issued NRC Information Notice (IN) 92-18, "Potential for
Loss of Remote Shutdown Capability During a Control Room Fire." In that IN, the NRC staff
alerted licensees to conditions (sometimes referred to as "hot shorts") found at several plants
that could result in the loss of capability to maintain the reactor in a safe shutdown condition
in the unlikely event that a control room fire forced reactor operators to evacuate the control
room. During NRC inspections of MOV programs and other licensee activities, the NRC staff
has identified weaknesses in the responses of some licensees to potential short-circuiting of
MOV control circuitry in the event of a plant fire.
Attachment 1 to this information notice contains examples of licensee problems in supporting
specific aspects of their bases for stating GL 89-10 actions have been completed. Attach- ment 2 contains a list of recently issued NRC information notices.
Related Generic Communications
BL 85-03 "Motor-Operated Valve Common Mode Failures During Plant Transients Due to
Improper Switch Settings," dated November 15, 1985 GL 89-10 "Safety-Related Motor-Operated Valve Testing and Surveillance," dated
June 28, 1989 GL 89-10 "Safety-Related Motor-Operated Valve Testing and Surveillance," dated
Sup. 1 June 13, 1990
GL 89-10 "Safety-Related Motor-Operated Valve Testing and Surveillance," dated
Sup. 2 August 3, 1990
GL 89-10 "Safety-Related Motor-Operated Valve Testing and Surveillance," dated
Sup. 3 October 25, 1990
GL 89-10 "Safety-Related Motor-Operated Valve Testing and Surveillance," dated
Sup. 4 February 12, 1992
GL 89-10 "Safety-Related Motor-Operated Valve Testing and Surveillance," dated
Sup. 5 June 28, 1993 GL 89-10 "Safety-Related Motor-Operated Valve Testing and Surveillance," dated
Sup. 6 March 8, 1994 GL 89-10 "Safety-Related Motor-Operated Valve Testing and Surveillance," dated
Sup. 7 January 24, 1996
IN 97-07 March 6, 1997 GL 95-07 "Pressure Locking and Thermal Binding of Safety-Related Power-Operated
Gate Valves," dated August 17, 1995 GL 96-05 "Periodic Verification of Design-Basis Capability of Safety-Related Motor- Operated Valves," dated September 18, 1996 IN 92-17 "NRC Inspections of Programs Being Developed at Nuclear Power Plants in
Response to Generic Letter 89-10," dated February 26, 1992 IN 92-18 "Potential for Loss of Remote Shutdown Capability during a Control Room
Fire," dated February 28, 1992 IN 96-48 "Motor-Operated Valve Performance Issues," dated August 21, 1996 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.
Thomas T. Martin, Director
Division of Reactor Program Management
Office of Nuclear Reactor Regulation
Technical contacts: Thomas Scarbrough, NRR
(301) 415-2794 E-mail: tgs@nrc.gov
(301) 415-2853 E-mail: wfb@nrc.gov
Attachments:
1. Examples of Problems Identified During GL 89-10
Closeout Inspections
2. List of Recently Issued NRC Information Notices
. P *;
Attachment 1 IN 97-07 March 6, 1997 EXAMPLES OF PROBLEMS IDENTIFIED DURING GL 89-10 CLOSEOUT INSPECTIONS
Thrust and Torque Requirements for Non-Dynamically Tested Motor-Operated Valves
(MOVs)
Some licensees had made general assumptions regarding the reduction in valve factor that
were based on increasing valve size, differential pressure, or fluid temperature without
sufficient test data to justify these assumptions quantitatively. In addition, some licensees
have had difficulty in justifying the capability of certain MOVs that have been sized and set on
the basis of unsupported assumptions for thrust or torque requirements. Licensees typically
predict the thrust required to operate gate and globe valves from the sum of (1) the product
of a valve factor, differential pressure across the valve, and the area of the valve disk; (2) the
product of the system pressure and the stem cross-sectional area; and (3) the drag of the
valve packing material on the valve stem. Some licensees assumed a generic valve factor of
0.5 (or less) in predicting the thrust required to operate non-dynamically tested gate valves
on the basis of their assumption that the selected valve factor was conservative. However, industry and plant-specific gate valve testing has revealed thrust requirements can exceed
that predicted by a 0.5 valve factor. Similarly, industry and plant-specific globe valve testing
has revealed that a valve factor of 1.1 to predict the thrust requirements might not be
adequate for all globe valves. With respect to butterfly valves, industry and plant-specific
testing has revealed that vendor calculations might not adequately predict the torque required
to operate some butterfly valves. On the basis of industry testing and analyses, the Electric
Power Research Institute (EPRI) is revising its application guide for predicting MOV thrust
and torque requirements.
Use of Industry Valve Information
Some licensees have found that testing of certain MOVs under dynamic conditions is
impracticable and that sufficient test information on similar MOVs at their plants is not
available. Consequently, these licensees have obtained MOV performance information from
other licensee or industry test programs and the MOV Performance Prediction Methodology
(PPM) developed by EPRI. In comparing test data from other sources, it is important to
understand the similarity of the valves; test conditions of differential pressure, temperature, and flow; diagnostic equipment and uncertainty; evaluation of the data and any anomalies
(such as high static seating loads); and calculation of valve factor (including flow area
assumptions). In addition, sufficient data need to be obtained to account for the variability in
thrust requirements for similar valves under applicable conditions. EPRI tested a sample of
valves of varying manufacture, type and size to validate a bounding methodology for
predicting thrust requirements for a wide variety of valves. The NRC staff identified concerns
regarding certain specific MOV tests by EPRI during its review of the methodology. These
concerns were resolved with respect to the bounding nature of the EPRI methodology in
developing the NRC staff safety evaluation.
Some licensees were not addressing the results of the EPRI methodology that predicted
potential valve damage and unpredictable thrust requirements for specific valves, and some
Attachment 1 IN 97-07 March 6, 1997 licensees did not address the limitations on the applicability of the EPRI methodology (such
as limitations due to the specific valve manufacturer).
Justification for Stem Friction Coefficient and Load Sensitive Behavior Assumptions
The efficiency of the conversion of actuator output torque to stem thrust is a function of the
stem friction coefficient and the dimensions of the valve stem and its thread. Load sensitive
behavior relates to the change in this efficiency when different thrust levels are exerted
through the stem. Typically, as the thrust level increases, the stem friction coefficient
increases and the thrust delivered at the torque switch trip decreases (referred to as a "rate- of-loading" effect). Some licensees initially assumed a stem friction coefficient of 0.15 (or
less) or rate-of-loading effect of 15 percent (or less) and planned to justify these assumptions
as part of their dynamic testing under GL 89-10. However, in some cases, insufficient data
or higher-than-expected values obtained during the MOV testing caused the staff to question
the licensee's initial assumptions when the data were evaluated in a statistically valid
manner. For example, one licensee may have to revise the initial assumption for rate-of- loading effects up to 25 percent. Stem friction coefficient and rate-of-loading effects may
vary between MOVs because of factors such as stem lubricant, lubrication frequency, environmental conditions, and manufacturing tolerances of the stem and stem nut.
Therefore, i is difficult to apply information on stem friction coefficient and rate-of-loading
effects from sources other than the licensee's testing program. EPRI developed bounding
values for load sensitive behavior associated with gate valves as part of its MOV PPM. The
NRC staff discusses conditions and limitations of the EPRI methodology in a safety
evaluation dated March 15, 1996. Also, some licensees have improperly considered load
sensitive behavior (or rate-of-loading effects) to be a random uncertainty, rather than a bias
error or a bias/random combination error.
GrouDing of MOVs
In GL 89-10, the NRC staff recommended that licensees test their safety-related MOVs under
design-basis conditions where practicable. In Supplement 6 to GL 89-10, the NRC staff
reiterated that recommendation but provided information on grouping MOVs in situations
where a licensee either is not able to test some MOVs under design-basis conditions or
chooses not to dynamically test some MOVs. For example, the NRC staff considered it
important to (1) assess, when grouping MOVs, such similarities as valve manufacturer, model
and size, valve flow, temperature, pressure, installation configuration, valve materials and
condition, seatlguide stresses, and performance during testing; (2) test a representative
sample of MOVs in each group (nominally 30 percent and no less than 2 MOVs); (3)test
each MOV in a group with diagnostics under static conditions; and (4) evaluate any adverse
information from individual MOV testing and determine its applicability to the entire group.
Some licensees have used approaches for grouping and testing MOVs other than those
described in Supplement 6 to GL 89-10. The NRC staff has found that some licensees have
not adequately justified testing only one MOV in a group, or a very small sample of MOVs in
the group. Also, some licensees have selected a valve factor based on a sample of tests
that does not accommodate reasonable variation in the valve factor for other MOVs in the
Attachment 1 IN 97-07 March 6, 1997 group (for example, the bounds on the valve factor for a group of valves was not always
appropriate for the scatter observed in the data). Although some licensees have grouped
MOVs in ways that could not be justified, some other licensees have established such a large
number of groups (as many as 50) that it is difficult to have sufficient test data for each
group. Some licensees have adequately justified including MOVs with small variations in size
into the same group in order to minimize the number of groups and allow sufficient data to be
obtained for each group.
De-raded Voltage Calculations
The NRC staff discussed in Supplements 1 and 6 to GL 89-10 determination of the voltage
assumed at MOVs for design-basis conditions. Various methods are used by licensees to
determine the reduction in voltage from the grid lo the MOV being evaluated. During
GL 89-10 closeout inspections, the NRC staff found that some licensees had not fully justified
their assumptions for the grid voltage assumed in their MOV calculations. For example, some licensees assumed full grid voltage as the starting point for calculations, rather than the
degraded grid relay setpoint.
Justification for Weak Link Analyses
In Information Notice 96-48, "Motor-Operated Valve Performance Issues," the NRC staff
discussed recent failures of MOV keys. Some licensees have also identified cracks in motor
shafts for some MOVs. Further, missing bolts or incorrect bolting material has been found in
some MOVs. These problems could be related to inadequate justification of the weak link
components in MOV analyses. For example, replacement of a motor pinion key with a key of
stronger material could cause the weak link to shift to another internal part, such as the
motor shaft.
Analytical Evaluation of Potential Pressure Lockinq of Gate Valves
In Supplement 6 to GL 89-10, the NRC staff provided one acceptable approach for
addressing potential pressure locking and thermal binding of MOVs. In GL 95-07, "Pressure
Locking and Thermal Binding of Safety-Related Power-Operated Gate Valves," the NRC staff
gave more detailed information and recommendations to address potential pressure locking
and thermal binding of gate valves.
During recent GL 89-10 inspections, the NRC staff identified weaknesses in some
approaches used by licensees to evaluate the effects of pressure locking of MOVs. Some
licensees are relying on analytical approaches (without test-based justification) to provide
confidence that the motor actuator can overcome the thrust resulting from pressure locking of
its valve. The NRC staff found that some licensees assumed overly optimistic actuator
efficiencies in predicting the thrust delivered by the motor actuator under pressure locking
conditions. In addition, the staff found that some licensees have insufficient justification for
assumptions of significant leakage from the valve bonnet over a short period, and of a very
low increase in bonnet pressure with rising temperature.
Attachment 1 IN 97-07 March 7, 1997 Evaluation of Test Data
Some licensees have not thoroughly evaluated test data to ensure that the results are
reliable. For example, an abnormally low thrust requirement or a back-calculated valve factor
might indicate that the design-basis differential pressure and flow were not achieved during
the test. Further, anomalies in the data traces could reveal valve or actuator damage. Some
licensees have not justified extrapolation of test data based on percentage of design-basis
differential pressure and absolute value of differential pressure as discussed in the EPRI
MOV program.
Trackina and Trendinq of MOV Problems
Tracking and trending are important aspects of a licensee's periodic verification program.
The NRC staff provided comments on MOV tracking and trending methods in initial reports of
GL 89-10 inspections. It also identified weaknesses in the development of MOV tracking and
trending methods at some nuclear plants. During GL 89-10 closeout inspections, the NRC
staff found that some licensees have not fulfilled their plans to develop MOV tracking and
trending methods and that some licensees have highly informal methods without specific
guidelines or schedules.
K)
Attachment 2 IN 97-07 March 6, 1997 LIST OF RECENTLY ISSUED
NRC INFORMATION NOTICES
Information Date of
Notice No. Subject Issuance Issued to
97-06 Weaknesses in Plant- 03/04/97 All holders of OLs
Specific Emergency or CPs for nuclear
Operating Procedures power reactors with
for Refilling the with once-through
Secondary Side of Dry steam generators
Once-Through Steam
Generators
91-85, Potential Failures of 02/27/97 All holders of OLs
Rev. 1 Thermostatic Control or CPs for nuclear
Valves or Diesel power reactors
Generator Jacket
Cooling Water
97-05 Offsite Notification 02/27/97 All holders of OLs
Capabilities or CPs for nuclear
power reactors and
test and research
reactors
97-04 Implementation of a New 02/24/97 All materials, fuel
Constraint on Radioactive cycle, and non-power
Air Effluents reactor licensees
97-03 Defacing of Labels to 02/20/97 All material licensees
Comply with 10 CFR involved with disposal
20.1904(b) of medical waste
OL = Operating License
CP = Construction Permit
IN 97-07 March 6, 1997 GL 95-07 "Pressure Locking and Thermal Binding of Safety-Related Power-Operated
Gate Valves," dated August 17, 1995 GL 96-05 "Periodic Verification of Design-Basis Capability of Safety-Related Motor- Operated Valves," dated September 18, 1996 IN 92-17 "NRC Inspections of Programs Being Developed at Nuclear Power Plants in
Response to Generic Letter 89-10," dated February 26, 1992 IN 92-18 "Potential for Loss of Remote Shutdown Capability during a Control Room
Fire," dated February 28, 1992 IN 96-48 "Motor-Operated Valve Performance Issues," dated August 21, 1996 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
Thomas T. Martin, Director
Division of Reactor Program Management
Office of Nuclear Reactor Regulation
Technical contacts: Thomas Scarbrough, NRR
(301) 415-2794 E-mail: tgs@nrc.gov
(301) 415-2853 E-mail: wfb@nrc.gov
Attachments: Tech Editor has reviewed and concurred on 11/27/96
1. Examples of Problems Identified During GL 89-10
Closeout Inspections
2. List of Recently Issued NRC Information Notices
DOCUMENT NAME: G:\WFB\IN_89_10 *SEE PREVIOUS CONCURRENCE
To receive a copy of this document, Indicate In the box cbopy wlo attachmentvendosure E=Copy with attachmentlenclosure N =No copy =
OFFICE Reviewers BCEMEB BCXPECB lIjjj-j
NAME TScarbrough*with RWessman* AChaffee* Tfartin
WBurton:jkd*comments wlcomments
DATE 12/03/96 12/06/96 01/03/97 02d097
, _ OFFICIAL RECORD COPY
- t February 11, 1 97 GL 95-07 "Pressure Locking and Thermal Binding of Safety-Related Power perated
Gate Valves," dated August 17, 1995 GL 96-05 "Periodic Verification of Design-Basis Capability of Safety elated Motor- Operated Valves," dated September 18, 1996 IN 92-17 "NRC Inspections of Programs Being Developed a Nuclear Power Plants in
Response to Generic Letter 89-10," dated Febrdry 26, 1992 IN 92-18 "Potential for Loss of Remote Sutdown Capability during a Control Room
Fire," dated February 28, 1992 IN 96-48 "Motor-Operated Valve Perforpiance Issues," dated August 21, 1996 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.
Thomas T. Martin, Director
Division of Reactor Program Management
Office of Nuclear Reactor Regulation
Technical contacts: Thomas Scarbrough, NRR
(301) 415-2794 E-mail: tgsenrc.gov
(301) 415-2853 E-mail: wfb@nrc.gov
Attachments: Tech Editor has reviewed and concurred on 11/27/96
1. Examples of Problems Identified During GL 89-10
Closeout Inspections
2. List of Recently Issued NRC Information Notices
DOCUMENT NAME: G:\WFBUIN_89 10 *SEE PREVIOUS CONCURRENCE
To receive a copy of this document, Indicate Inthe box C=Copy w/o attachment/enclosure E=_opy Wth attachment/enclosure N No py
OFFICE Reviewers l BC\EMEB I BC\PECB M ID
. NAME TScarbrough*with RWessman* AChaffee* Toirartin
WBurton:jkd*comments w/comments
DATE 12/03/96 12/06/96 01/03/97 02A497 OFFICIAL RECORD COPY
Is
bi. IN 96-xx
K) December xx, 1996 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.
Thomas T. Martin, Director
Division of Reactor Program Management
Office of Nuclear Reactor Regulation
Technical Contacts: Thomas Scarbrough, NRR William Burton, NRR
(301) 415-2794 (301) 415-2853 E-mail: tgsenrc.gov E-mail: wfb@nrc.gov
Attachments:
1. List of Recently Issued NRC Information Notices
2. Examples of Problems Identified During GL 89-10
Closeout Inspections
DOCUMENT NAME: G:\WFBUN_89j10 *SEE PREVIOUS CONCURRENCE l ij4 I
To receive a cowy of this document Indicate in the box C=Coov wio attachmentlenclosure E=Coov with attachmentlenclosure N = COWv
OFFICE Reviewers I BC\EMEB _ BC\PE7f A) I D\DRPM I
NAME TScarbrough*wifth RWessman* AChake\, TMartin
WBurton:jkd*comments w/comments ) I_ _ _
DATE 12/03196 12/06/96 Dl @ V 12/ /96 OFFICIAL RECORD COPY <
( (vII-/f 1
- .IN 96-xx
- .iDecember xx, 1996 This information notice requires no specific action or written response. Ihave any
questions about the information in this notice, please contact one 9We technical contacts
listed below or the appropriate Office of Nuclear Reactor Re ion (NRR) project manager.
Tho T. Martin, Director
ision of Reactor Program Management
Office of Nuclear Reactor Regulation
Technical Contacts: Thomas S rbrough, NRR William Burton, NRR
(301) -2794 (301) 415-2853 E- l: tgs@nrc.gov E-mail: wfbenrc.gov
Attachments:
1. List of Re tly Issued NRC Information Notices
2. Examp of Problems Identified During GL 89-10
seout Inspections
DOC ENT NAME: G:AWFB\IN 89 10
To receive a co of this document. Indicate In the box C=copy wlo attachment/enclosure E=Copy Wtth attachmentlenclosure N = No copy
OFFI Reviewers l BC\EMEB BC\PECB l I DUDRPM
NAME TScarbrou3WP it{Wessm ) ACh47/ -g0 TMartin
WBurton jkd _f__ __m
DATE 123/96 61J96 1 /15/77 Il1 /96 OFFICIAL RECORD COPY