ML18152A449
| ML18152A449 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 06/04/1993 |
| From: | Branch M, Matt Thomas NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18152A450 | List: |
| References | |
| 50-280-93-12, 50-281-93-12, NUDOCS 9306160176 | |
| Download: ML18152A449 (14) | |
See also: IR 05000280/1993012
Text
Report Nos.:
Licensee:
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTA STREET, N.W.
ATLANTA, GEORGIA 30323
50-280/93-12 and 50-281/93-12
Virginia Electric And Power Company
Glen Allen, VA 23060
Docket Nos.:
50-280 and 50-281
Facility Name:
Surry 1 and 2
Inspection Conducted:
May 3-7, 1993
License Nos.: DPR-32 and DRP-37
Inspector: mt~ ~
M. Thomas
Accompanying Inspectors: H. Whitener
M. Miller
Approved by:
C -~~J~,,.
k_
M.Bran~ ~
Test Programs Section
Engineering Branch
Division of Reactor Safety
SUMMARY
Scope:
,-~-7~
Date Signed
c/lf/93
Date Signed
This routine, announced inspection was conducted in the areas of design
changes and modifications and engineering support activities.
Results:
In the areas inspected one violation and one inspectorfollowup item {IFI)
were identified.
Violation 50-280,281/93-12-0l, for failure to follow procedures in
updating design drawings within the required time period following
the implementation of two design change packages {DCP) {paragraph
3).
IFI 50-280,281/93-12-02, Labeling of test valves installed in the
emergency diesel generator air start system {paragraph 2.b.) *
9306160176 6gggg~80
~DR
ADOCK
2
The various engineering groups worked well together to resolve
complex problems that could potentially affect plant operations.
Timely and effective engineering support was provided to resolve
the pressurizer safety relief valve issue.
Engineering has provided timely support in resolving deviation*
reports (DRs) and requests for engineering assistance (REA).
A weakness was noted in the documentation of the 10 CFR 50.59
safety evaluation for DCP 92-49 (paragraph 2.b.).
The licensee's program for reducing the DCP and engineering work
request (EWR) modification backlog provided adequate justification
for cancelling the DCPs reviewed.
The licensee's self assessment efforts continue to demonstrate the
licensee's commitment to improving the quality and effectiveness
of engineering support provided to the plant .
..
1.
Persons Contacted
Licensee Employees
REPORT DETAILS
- W. Benthall, Manager, Nuclear Licensing
- R. Bilyeu, Licensing Engineer
D. Blake, Design Engineer, Station Engineering
- R. Blount, Superintendent, Station Engineering
- A. Fletcher, Assistant Superintendent, Station Engineering
- B. Foster, Mechanical Engineering Supervisor, Station Engineering
R. Green, Systems Engineering Supervisor, Station Engineering
- D. Hart, Supervisor, Quality Assurance
- M. Kansler, Station Manager
- R. MacManus, Systems Engineering Supervisor, Station Engineering
- J. Price, Assistant Station Manager, Nuclear Safety and Licensing
- R. Saunders, Assistant Vice President Nuclear Operations
- J. Swientoniewski, Supervisor, Station Nuclear Safety
E. Watts, Electrical Engineering Supervisor, Station Engineering
Other licensee employees contacted during this inspection included
engineers, operators, craftsmen, and administrative personnel.
NRC Resident Inspectors
J. York, Senior Resident Inspector
- S. Tingen, Resident Inspector
- Attended exit meeting
Acronyms and initialisms used throughout this report are listed in the
last paragraph.
2.
Design Changes and Plant Modifications (37700)
a.
Plant Modifications to Improve Reactor Safety
The inspectors reviewed the initiatives taken by the licensee to
identify and implement plant modifications to improve reactor
safety. This included reviewing the licensee's efforts to reduce
The licensee implemented a Level I Project Modification Package
Backlog Reduction program designed to reduce the EWRs and DCPs
from 333 in August 1992 to 150 by June 1993.
The inspectors
reviewed trend reports and monthly status reports and determined
that the EWR/DCP reduction was on schedule as of April 1993 with
172 packages remaining in the backlog population.
The packages in
the backlog population were in various stages of completion.
In
some cases the engineering design work, modification
implementation, and technical review were complete and final
closeout awaited only sign off of the documentation.
At the other
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2
extreme, some DCP numbers were assigned, but were not funded and
no engineering work had been performed.
Closeout methods varied
from closure with the completion of documentation, closure with
partial implementation, closure with no implementation, closure by
cancellation, and closure of assigned numbers which were not
approved for funding.
This process was consistent with the VEPCO
General Nuclear Standard (STD-GN-0001, Revision 10) which provided
interim instructions for changes in the design control process.
Subsequent to September 16, 1991, all plant modifications (major
and minor) are to be performed using the DCP process and EWRs are
to be used for engineering technical evaluations.
initiated prior to this date may be completed in the old format,
canceled, or converted to the new DCP format as appropriate.
STD-
GN-0001, Revision 11 and VPAP 0301, Revision 1, Design Change
Process, specify that cancellation of EWRs/DCPs will be
accomplished through a field change (FC) which identifies the
reason for cancellation and if there was any 'impact on systems or
projects resulting from not implementing the design change.
A sample of FCs for canceled EWRs/DCPs was reviewed to evaluate
any impact on system reliability and safe operation of the plant.
FCs reviewed and evaluated included the following:
DCP 90-013-1, Revision 8 - 2/20/92
This was a corporate issued DCP specifying replacement of
the motor operators for six valves.
Replacement was
completed for two valves, two valves did not require
replacement, and two valves required maintenance.
Subsequent to the operational readiness review, the DCP was
canceled for the four valves not requiring operator
replacement.
While cancellation was justified, the FC was weak in that
the reason for not replacing the operators on four of the
six valves had to be determined from other sources.
EWR 89-467, Revision A - 9/25/92
This EWR was an evaluation to perform a commercial grade
dedication for replacement relief valves in the component
cooling water system in order to upgrade the valves to
safety related status. Based on the current design change
standards, EWRs are not the proper vehicle for performing
commercial grade dedication of equipment.
The EWR was
canceled and the new valves will be procured and qualified
under the Virginia Power Safety Related Dedication
Procedures and Standards program.
The FC provided adequate justification for the cancellation .
3
EWR 88-012, Revision J - 9/11/92
This EWR provided instructions for replacing existing
auxiliary feedwater system check valves. According to a
systems engineering evaluation, replacement was not
necessary.
The existing valves were performing
satisfactorily.
Due to valve performance, cancellation was considered
justified. Should valve replacement be necessary in the
future, it would be addressed under the current design
change standard with a DCP.
EWR 89-730, Revision B - 8/31/92
This EWR was issued for engineering authorization to replace
a leaking one inch Pacific gate valve (no longer available}
with a one inch Velan gate valve. Discussions with System
Engineering indicated that in a recent leak rate test the
Pacific valve was not leaking.
The licensee concluded that
it was not necessary to replace the Pacific valve.
Cancellation of this EWR was considered justified based on
valve performance .
DCP 92-48-1
Engineering reviewed a request for engineering assistance
(REA} to develop a DCP to install new ventilation flow
elements. Although the new elements were slightly different
from the original elements, the differences were minor and
considered within the scope of the Item Equivalency Program.
Procurement Engineering performed an evaluation to address
the differences.
The new elements are to be installed by
work orders 123458 and 123459.
Design Engineering will
assist by providing a package of electrical wiring diagrams,
equipment support drawings, and the general installation
sequence identifying special implementation requirements.
Cancellation of the DCP was considered justified in that the
flow elements will be installed under an approved, standing
program.
DCP 90-16-1, Revision 6 - 11/19/92
This DCP was issued to modify the reactor cavity seal.
Prior to performing this modification a walkdown revealed
dimensional discrepancies between the actual cavity seal
area and the new seal design. Therefore, the DCP was not
implemented.
Instead, the original method of cavity seal
was accomplished via EWR 90-328. This method is performing
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4
satisfactorily. Consequently, DCP 90-16-1, which would
require major redesign, will not be implemented.
In that the current cavity seal method is performing
satisfactorily, cancellation of the redesigned seal is
considered justified.
EWR 90-158, Revision A - 10/20/92
This EWR was issued to relocate an electrical receptacle
located near a battery in the emergency service water pump
house to reduce an explosion hazard.
An engineering
walkdown prior to implementation found that the receptacle
had been removed, the wires taped and the opening capped.
In that there were sufficient receptacles in the pump house,
this specific receptacle was eliminated rather that
relocated. Cancellation of the EWR was considered
justified.
EWR 90-237, Revision A - 9/4/92
This EWR was issued in response to valve thermal
binding/bonnet pressurization problems identified in INPO
Significant Operating Experience Report 84-07.
Engineering
reviewed all air operated {AOV) and motor operated {MOV)
safety related gate valves for Surry Units 1 and 2 and
issued a Type 1 Report on the investigattve methods and
findings.
The review consisted of an initial screening to
identify the AOVs and MOVs susceptible to the thermal
binding/bonnet pressurization phenomena.
A detailed
analysis was performed on valves identified by the screening
process.
The licensee concluded that none of the valves
have a significant possibility of experiencing thermal
binding/bonnet pressurization leaking.
In response to the Operating Experience Review Group
concerns, engineering developed calculations showing that,
for the MOVs of concern, the motor operators have the
capacity to open the valves against the binding forces.
Based on the valve operation histories, the analysis of
operating conditions, and force calculations, cancellation
of this DCP was considered justified.
Review of the above sample of FCs issued to cancel EWRs/DCPs
indicated that safe operation of the plant or reliability of
systems has not been compromised.
Based on discussions with
licensee engineering personnel, and review of the above
documentation, the inspectors concluded that the licensee has a
..
b.
5
satisfactory process for identifying and implementing plant design
changes to improve reactor safety and to reduce the EWR/DCP
backlog.
Planning, Development and Implementation of Plant Modifications
The inspectors reviewed the DCPs listed below to: (1) determin~
the adequacy of the 10 CFR 50.59 safety evaluations performed; (2)
verify that the DCPs were reviewed and approved in accordance with
TS and applicable administrative controls; (3) verify the subject
modifications were installed (for those that could be physically
inspected) in accordance with the DCP package; (4) verify that
applicable plant operating and design documents (drawings, plant
procedures, FSAR, TS, etc.) were revised to reflect the subject
modifications; (5) verify that the modifications were reviewed and
incorporated into the operations training program as applicable;
and (6) verify that post modification test requirements were
specified and that adequate testing was performed.
The following
DCPs were reviewed:
DCP 88-32, Addition of Diesel Generator Sequencing, Unit 2
This modification was initially dated December 11, 1988, to
add an emergency diesel generator (EDG) load sequencing
scheme that would be initiated by a loss of offsite power
(LOOP).
The purpose of this scheme was to ensure that the
maximum EDG load capabilities would not be exceeded under
the worst case load applications, and therefore resolve NRC
concerns described in IE Information Notice 85-91.
The
second part of this DCP was dated February 26, 1991, to
modify the control circuits for the auxiliary feedwater
pumps.
The control circuits were modified by eliminating
the latching relays and replacing them with auxiliary type
relays. Four auxiliary relays were added to control
circuits.
This DCP included both the engineering design change
packages and the installation work plans.
The inspectors
conducted a detailed walkdown inspection to verify that
components were installed and the drawings reflected the as-
built plant condition.
The inspectors determined that this
modification was implemented in a satisfactory manner.
DCP 91-12, RSHX Service Water Flow Element Modifications,
Units 1 and 2
This DCP was implemented to replace existing pitot venturi
flow elements (l-SW-FE-105A and -105B, 2-SW-FE-205A and
-205B) that monitor SW flow in the supply headers to the
RSHXs; and venturi flow elements (l-SW-FE-106A,B,C,D, 2-SW-
FE-206A,B,C,D) that monitor SW outlet flow from each of the
6
RSHXs, with V-Cone flo~ elements.
The V-Cone flow element
provided stable and accurate flow indication during the RSHX
flow test that was performed on April 6, 1991.
The inspectors reviewed the DCP in accordance with the
criteria specified above and performed a field inspection to
verify that the components were installed and the applicable
drawings reflected the as-built plant configuration. The
inspectors determined that this DCP was satisfactorily
implemented.
DCP 92-49, Removal of Motor Operators From Ol/02-RH-MOV-
100/200
This modification required the removal of motor operators
from the RHR containment isolation valves 01-RH-MOV-100 and
02-RH-MOV-200.
In the DCP's "Statement of the Problem", the
reasons for the removal of these valves' motor operators
were the initial installation had an improper design, there
were significant difficulties with proper operation, and the
valves repeatedly failed the "Type C" leak testing. These
problems were caused by the design configuration originated
for the initial installation of the motor operators by DCP 74-001 performed in 1975.
The previous manual valves had
motor operators remotely installed above them using a long
drive shaft. Since this arrangement was not effective the
motors were later disconnected and the valves were manually
operated.
The purpose of DCP 74-001 was to install an overflow line
from the RWST to the safeguards valve pit to prevent the
uncontrolled release of radioactive water.
The basis for
this modification was that pumps were installed in the pit
to pump the water to the liquid waste system.
The motor
operators were installed on the valves to automatically
close upon a high level alarm on the RWST to prevent over
filling.
(The valves are in the lines feeding the RWST and
are only used during refueling.)
In 1984 another modification was performed in this area.
EWR 84-089 was initiated to prevent the release of
unmonitored gaseous effluents from the RWST.
This
modification required the capping of the vent on top of the
RWST and extended .the overflow line down further in the
valve pit.
EWR 84-089 discussed the basis of DCP 74-001
where the manual valves had motor operators installed to
provide the automatic shutoff to prevent RWST overfill.
During the review of this modification the inspectors
determine that the licensee's basis for removing the motor
operators was acceptable: However, several concerns were
identified during the review of the safety evaluation for
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7
the removal of the motor operators which indicated an
inattention to details. The licensee's safety evaluation
form required a response and an explanation for each
question.
The explanation provided as justification for
several of the questions was that the RWST overflow
occurrence had been previously analyzed in DCP 74-001 and
EWR 84-089.
The inspectors determined that DCP 74-001 and
EWR 84-089 analyses clearly stated that the valves had motor
operators added to provide automatic shutoff to prevent the
RWST from being overfilled. Therefore, using the analyses
from DCP 74-001 and EWR 84~089 for removing the motor
operators was inappropriate.
In the "Programs Review
Checklist
II the question for ALARA, asks if the work in the
DCP affects systems, facilities, and/or equipment which
process or contain radioactive materials, fluids or gases?
The answer checked was no.
The purpose of the earlier
modifications, DCP 74-001 and EWR 84-089, was to control the
release of radioactive effluents. The inspectors identified
these responses in the safety evaluation as concerns that
appeared to be due to inattention to details.
The inspectors concluded that the licensee's basis for
removing the motor operators was justified and there was no
safety concern.
However, the responses to several of.the
questions indicated an inattention to detail .
DCP 92-72, EG Check Valve Testing Modification, Units I
and 2
This DCP was implemented to install a test valve in each EDG
air start system between the compressor and the safety-
related air receiver check valve, in order to test for back
leakage by the check valve.
Leak testing of the check valve
is performed to meet an ASME Section XI Inservice Testing
commitment.
The inspectors reviewed this DCP to the criteria specified
above and performed a field inspection to verify proper
installation. During the field inspection, the inspectors
noted that the drawings had been updated to reflect the
modification, but only one of the six test valves installed
by this DCP was labeled. The inspectors discussed this
discrepancy with licensee personnel who indicated that the
valves were being labeled as part of the licensee's upgraded
labeling program, which was still ongoing.
The licensee
further indicated, and the inspectors verified, that the
valves were only for test purposes and performed no safety
function.
The inspectors will verify labeling of the test
valves during a future inspection. This item will be
identified as IFI 50-280, 281/93-12-02, Labeling of test
valves in emergency diesel generator air start system .
C
3.
8
The inspectors noted that, except for the discrepancies discussed
above, the DCPs were satisfactorily implemented.
None of the
discrepancies noted had a safety impact. Violations or deviations
were not identified in the areas inspected.
Drawing Control
The inspectors reviewed the licensee's program and procedures that were
developed and implemented to maintain configuration control of the
applicable drawings after DCP implementation.
The program was examined
to ensure that design control was maintained and that the drawings
affected by DCPs were updated in a timely manner to reflect the as-built
plant. The procedures reviewed for the configuration drawing control
program included the following documents:
VPAP-0301, Virginia Power Administrative Procedure, Design Change
Process
SUADM-ADM-11, Surry Power Station Administrative Procedure,
Station Drawing Revision and Distribution
VPAP-0301 was the detailed procedure that established the process for
managing the preparation and implementation of design changes.
It also
established interfaces among the various organizations and defined the
controls necessary to assure safe implementation of station design
changes.
In addition, the VPAP discusses the requirements for updating
the design drawings in accordance with SUADM-ADM-11.
The purpose of
SUADM-ADM-11 was to prescribe the method for revision of controlled
drawings, for annotation of drawings to reflect design changes in
progress, and to provide guidelines for proper distribution and
maintenance of station controlled drawings.
The inspectors selected
drawings from completed DCPs to determine if the drawings were updated
in a timely manner as required by the licensee's procedures.
In
addition, the drawings in the Control Room were examined to ensure that
they were also updated.
The drawings for the following DCPs were
examined:
DCP 93-20, EPH 34.5 KV BUS NO. 5
DCP 92-64, CHARGING PUMP LOGIC
DCP 89-09, POWER SUPPLY-ATWS MIGITATION SYSTEM
DCP 87-26, ATWS MITIGATION SYSTEM
DCP 88-32, ADDITION OF DIESEL GENERATOR SEQUENCING
DCP 92-63, LP HEATER DRAIN PUMP RECIRCULATION LINE
DCP 91-12, RSHX SERVICE WATER FLOW ELEMENT MODIFICATIONS
DCP 92-72, EG CHECK VALVE TESTING MODIFICATIONS
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9
During the drawing review, the inspectors identified on May 6, 1993,
that the priority drawings for DCP 93-20 and DCP 92-64, which required
revision within 15 days of the operational readiness review (ORR)
completion date, were not completed.
The ORR completion date for DCP
93-20 was April 8, 1993, and the drawing not updated was 11448-FE-lA.
The ORR completion date for DCP 92-64 was April 9, 1993 and the drawings
not updated were 11548-ESK-SP, -SQ, -SR, and -5S.
The licensee's
failure to update these 15-day priority drawings as required by SUADM-
ADM-11 is identified as Violation 50-280, 281/93-12-01, Priority
Drawings Not Updated.
The licensee took immediate corrective action by
issuing deviation reports for DCP 92-64 and DCP 93-20, identifying the
discrepancies and requiring that the drawings be updated.
Both DRs were
dated May 6, 1993.
The inspectors verified that the licensee updated
the overdue drawings on May 6, 1993.
One violation was identified in the areas inspected.
4.
Engineering and Technical Support Activities
The inspectors reviewed activities performed by Station Engineering in
an effort to assess the effectiveness of the support being provided to
the plant operations and maintenance staffs for day-to-day activities.
These activities included responding to DRs, REAs, Systems Engineering
activities, and self assessments .
The inspectors concluded that the various engineering groups worked well
together to resolve complex problems and, in general, provided timely
and effective engineering support to the plant.
a.
Deviation Reports and REAs
The inspectors reviewed Engineering's involvement in resolving
DRs, which included reviewing DR trend reports over the last year.
The inspectors noted that from January 1992 to May 1993, a total
of 654 DRs were assigned to Station Engineering (458 DRs for 1992
and 196 DRs for 1993).
Engineering was only overdue in responding
to 16 of the DRs (exceeding 30 days) in 1992.
There have been no
late DR responses for 1993. There were 13 DRs currently open for
1993 and there were no open 1992 DRs.
The inspectors also reviewed Engineering's involvement in
resolving REAs.
Since June 1992, 390 REAs were received by
Engineering.
A total of 338 REAs have been closed by Engineering,
11 REAs have been reviewed by Engineering and were awaiting review
by the MMRT, and 41 REAs were under review by Engineering.*
The inspectors concluded that Station Engineering provided timely
responses for assigned DRs and REAs .
b.
10
Systems Engineering
The licensee's engineering organization is diversified in
specialty areas including corporate, maintenance, design, system,
material, modification implementation, and testing. The
inspectors reviewed engineering activities to ascertain
involvement in plant operations and maintenance.
The Surry Station Engineering Services (SSES) 3.01, Revision 2,
Controlling Procedure for System Engineering, identifies
Maintenance Engineering (ME) as the component engineering experts
and describes the System Engineer (SE) as the system expert
responsible for system management and oversight. Duties included
ensuring system performance per design basis; maintaining
cognizance of system conditions; coordinating diverse group
efforts to resolve problems; and serving as the expert for system
design, regulatory, testing and operational questions.
In this
regard the SE writes and reviews procedures, performs system
tests, reviews test results and trends system performance.
Additionally, the SE reviews all work orders on assigned systems
prior to work.
When multiple work orders are involved the SE
generates a post maintenance testing (PMT) package which details
the sequence of testing, a flow chart, procedures, and procedure
changes to accomplish a meaningful PMT.
This package is
distributed to operations, management, maintenance and engineering
for review and comment.
The inspectors reviewed DRs and REAs and determined that
engineering's response in support of plant operations and
maintenance was adequate and prompt.
The inspectors also reviewed
portions of the Station Engineering Weekly Events Reports for
February 1993 - May 1993 and the System Engineering Quarterly
Report for the fourth quarter of 1992.
These reports present the
historical problems of the system, the problems experienced during
the report quarter, actions taken to resolve problems and
recommendations to upgrade and enhance system reliability. The
quarterly report also trends system parameters.
The. inspectors
reviewed portions of these reports for the EDG and instrument air
systems and concluded that engineering was responsive in
identifying and resolving problems.
The inspectors also followed a real time event which occurred
during this inspection. During startup, the pressurizer code
safety valves were simmering.
The licensee obtained a waiver of
compliance to permit gagging the relief valves and performing the
system hydrostatic test. Subsequently, system pressure was
reduced to about 1900 psig to allow the code valves to seat and
stabilize. System pressure was then brought up to 2135 psig,
where the plant will be operated. This is 100 ps~ below normal
operating pressure and required NRC approval and an emergency TS
change.
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c.
11
Engineering was active in obtaining approval for operation at
reduced pressure. Corporate engineering performed calculations to
show adequate departure from nucleate boiling safety margin,
design engineering walked down the system and developed alternate
plans of action in case leakage did not stop at reduced pressure,
System Engineering was involved in obtaining the waiver of
compliance for the hydrostatic test and Maintenance Engineering
was involved in details of gagging the valves and communicating
with the vendor, and a management oversight committee followed the
events closely. The inspectors considered this a good example of
engineering groups working together to support the plant.
Based on review of records and observations of the licensee's
response to the above event the inspectors concluded that
engineering activity supports plant operations and maintenance.
Self Assessment
The inspectors reviewed certain aspects of the licensee's self
assessment program.
One assessment reviewed was the appraisal of
the engineering programs.
The licensee developed a rating system
and performed a survey of 24 engineering programs to evaluate the
programs' effectiveness. A score of 5 indicates average program
effectiveness.
The survey results showed that four of the
programs (Appendix R, Setpoint Coordination, TS Compliance
Coordination, and Computer Software Control) were rated as average
to less than average. Corrective action plans were developed and
entered in the licensee's commitment tracking system to enhance
and improve the programs.
The inspectors reviewed the action
plans for three of the programs.
The inspectors also reviewed selected QA audits of engineering
activities. Audits reviewed included activities such as the
vendor information program, drawing update, and configuration
control.
The audits were detailed and identified areas of
weakness and strength.
The inspectors determined that the licensee's self assessment
efforts continue to demonstrate the licensee's commitment to
improving the quality and effectiveness of engineering support
provided to the plant.
Violations or deviations were not identified in the areas inspected .
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5.
12
Exit Interview
The inspection scope and results were summarized on May 7, 1993, with
those persons indicated in paragraph 1.
The inspectors described the
areas inspected and discussed in detail the inspection results listed
below.
Proprietary information is not contained in this report.
Dissenting comments were not received from the licensee.
The following
findings were discussed:
Violation 50-280, 281/93-12-01, Failure to follow procedures in
updating design drawings within the required time period.
(paragraph 3)
IFI 50-280, 281/93-12-02, Labeling of test valves installed in the
emergency diesel generator air start system. (paragraph 2.b.)
6.
Acronyms and Initialisms
ADV
DR
EOG
FC
IFI
kV
MDV
NRC
ORR
psi
psig
REA
RH
TS
As Low As Reasonably Achievable
Air Operated Valve
American Society of Mechanical Engineers
Anticipated Transient Without Scram
Design Change Package
Deficiency Report
Engineering Work Request
Field Change
Final Safety Analysis Report
Inspector Followup Item
Institute of Nuclear Power Operations
Kilovolts
Low Pressure
Motor Operated Valve
Nuclear Regulatory Commission
Operational Readiness Review
Pounds Per Square Inch
Pounds Per Square Inch Gauge
Quality Assurance
Request for Engineering Assistance
Residual Heat
Residual Heat Removal System
Recirculation Spray Heat Exchanger
Refueling Water Storage Tank
Technical Specification