ML20210K602
| ML20210K602 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 04/19/1986 |
| From: | Architzel R, Imbro E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| To: | |
| Shared Package | |
| ML20210K600 | List: |
| References | |
| 50-327-86-27, 50-328-86-27, NUDOCS 8604280400 | |
| Download: ML20210K602 (77) | |
See also: IR 05000327/1986027
Text
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U.S. NUCLEAR REGULATORY COMMISSION
OFFICE OF INSPECTION AND ENFORCEMENT
Division of Quality Assurance, Vendor,
and Technical Training Center Programs
Report Nos.: 50-327/86-27;50-328/86-27
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Docket Nos.: 50-327;50-328
Licensee: Tennessee Valley Authority
6N, 38A Lookout Place
1101 Market St.
Chattanooga, TN 37402-2801
Facility Name: Sequoyah Nuclear Plant, Units 1 & 2
Inspection At: Gilbert / Commonwealth, Reading PA
TVA-Knoxville TN
TVA-Soddy Daisy TN
Inspection Conducted: February 12-14, 18-21 and March 3-7, 1986
Inspection Team Members:
Team Leader: R. E. Architzel, Senior Inspection Specialist, IE
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Mechanical Systems: F. Mollerus Consultant, Mollerus Engineering Inc.
R. Parkhill, Inspection Specialist, IE*
Mechanical Components: A. V. du Bouchet, Consulting Engineer
Civil / Structural: A. Unsal, Harstead Engineering
Electrical Power: S. V. Athavale, Inspection Specialist, IE
, Instrumentation &
Control: L. Stanley, Consultant, Zytor Inc.
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Ralph'E. Architzel
da'/d /IlbII& 9
Team Leader J Date
Approved by: [ d'Date
f!/'[
Eugene V. Imbro
Section Chief
Quality Assurance Branch
- Part time
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8604290400 860422
PDR ADOCK 05000327
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1. INTRODUCTION AND SUPMARY
1.1 INTRODUCTION
The following subparagraphs provide an introduction to.the objectives, format and
focus of the Sequoyah design control inspection.
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1.1.1 BACKGROUND AND OBJECTIVES
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Several recent developments at TVA have raised uncertainties regarding the
l adequacy of the design control process which has been and is being implemented
' for the Sequoyah Nuclear Power Plant. Some of these issues were identified in
an NRC letter dated January 15, 1986, requesting additional information regard-
'
ing the following:
An explanation of why the design control problems identified at Browns e
! Ferry Nuclear Plant and identified via employee concerns at Watts Bar
!
do not apply at Sequoyah.
I An explanation as to how all of the many issues arising from employee-
j concerns, NSRS concerns, QA audits, etc., will be reviewed, and how !
j the scope of the reviews will bound the issues.
A more complete description of the basis for TVA's conclusion that
Sequoyah design controls were adequate prior to June 1985. -
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An explanation of how any design changes initiated at Watts Bar to
correct a potential safety problem would be addressed with respect to
Sequoyah, assuming the plant design feature in question is identical
j for both plants.
A description of the electrical design calculations review program,
l currently in progress for the Sequoyah facility. This should include _
i a description of how this program is being incorporated in.the overall
l design review effort.
- Although TVA had not yet responded to this request,-the NRC.was aware of several
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design control reviews which were being conducted.
One of these reviews, the Gilbert / Commonwealth (G/C) review of changes to the main
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and auxiliary feedwater system, was being conducted as a follow-on to a previous-
Sequoyah Design Control Survey conducted by G/C. NRC Region II inspected the
i results of this survey (Inspection Report Nos. 50-327/85-48 and 50-328/85-48) and
i also inspected the later review in process (Inspection Report Mos. 50-327/86-07
1
and 50-328/86-07).- The survey ~ addressed design control program adequacy in
light of implementation of a new set of Office of Engineering procedures in June
- 1985.
,
TVA asked Gilbert / Commonwealth to review the technical adequacy of design control
for Sequoyah Nuclear Plant Units 1 and 2 for the period from issuance of the
j operating license until June, 1985, when TVA issued revised design control proce-
- dures. Gilbert / Commonwealth (G/C) selected the main and auxiliary feedwater
i system for review. G/C Report No. 2614, issued on March 3, 1986, documents the
j results of G/C's review of the main and auxiliary feedwater system.
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In a December 12, 1985 meeting with the NRC, TVA committed to review the
current configuration of Sequoyah Nuclear Plant Units 1 and 2 for suitability
for operation. This review was to include confirmation that unreviewed safety
questions did not exist. Questions had been raised as to whether out of sequence
implementation of ECNs, at the discretion of the Office of Nuclear Power could
result in an unreviewed safety question. Additionally, the large backlog of
unimplemented and partially implemented ECNS was of concern. The scope of TVA's
review was limited to the essential raw cooling water (ERCW) system for Units 1
and 2, Unit I reactor coolant system (RCS) and Unit 2 chemical and volume control
system (CVCS). The results of TVA's 3 system review were provided to the team
in a TVA report entitled " Technical Review of Unimplemented and Partially Imple-
mented Engineering Change Notices", which TVA issued during the last week of the
inspection on March 3, 1986.
The objectives of this NRC inspection were to:
Assess the adequacy and conclusions of the Gilbert / Commonwealth (G/C)
engineering review of modifications made to the main and auxiliary
feedwater system since the plant was licensed.
Review the TVA Office of Engineering (0E) e.aluation of unimplemented/
partially implemented changes in three systems (RCS, CVCS, ERCW) to
determine whether the partial implementation or failure to implement
a modification constitutes an unreviewed safety question (USQ).
Assess the adequacy of the Sequoyah design control process in light
of inspection results from items 1 and 2 and an additional independent
NRC review.
The purpose of the independent review portion of the inspection was to examine,
on a sampling basis, detailed design and engineering which was performed to
support the modifications. This approach was used to allow the NRC team to
understand the depth of the technical reviews performed by G/C and TVA and to
enhance the confidence which the team had regarding the other reviews' con-
clusions and recommendations.
1.1.2 REPORT FORMAT AND DEFINITIONS
The areas examined during this inspection are addressed by discipline in the
following sections. Each section consists of three sub parts which address the
Gilbert / Commonwealth review, the TVA "3 system review," and independent NRC
review, respectively. Section 7 summarizes the meetings held during the inspec-
tion and lists the persons contacted.
Deficiencies, unresolved items, and observations are defined below and are
included in an appendix to this report.
Deficiencies
Errors, inconsistencies or procedure violations with regard to a
specific licensing commitment, specification, procedure, code or
regulation are described as deficiencies. Follow up action by the
licensee is required.
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Unresolved Items
Unresolved items are potential deficiencies which require more
-information to reach a conclusion. Follow-up action by the
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license is required.
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Observations
Observations are presented where it is considered appropriate to
call attention to matters that are not deficiencies or unresolved-
items, but which merit licensee consideration. No l',censee response
is required since there is no specific regulatory requirement for
i such matters.
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1.1.3 INSPECTION EFFORT
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The inspection was an NRC effort conducted with contractor assistance. Team
members were selected to provide technical expertise and design experience in
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the disciplines listed. Most of the team members had previous experience as
employees of architect-engineering firms or reactor manufacturers working on
large commercial nuclear power plants. The others had related design experience-
- on commercial nuclear facilities, test reactors, or naval reactors.
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Beginning on February 12-14, 1986, the team inspected the G/C review at their
offices in Reading, Pennsylvania. TVA personnel were present during this
- phase, and provided information regarding the TVA design control process and
j the scope and findings of the G/C review as of that date. The inspection con-
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tinued on February 18-19, 1986 at TVA's Office of Engineering in Knoxville,
t Tennessee, where the principal thrust of the inspection was an overview of the
$
i OE "3 system" review. The remainder of the inspection was conducted at
the Sequoyah Nuclear Plant on February 20-21 and March 3-7, 1986. This
! portion of the inspection included reviews of both the G/C and OE work and
independent assessment of design control by the NRC. An exit meeting was held
at the Sequoyah plant site on March 14, 1986.
The inspection team-reviewed the organizations' staffing and procedures and
j interviewed personnel to determine the responsibilities of and the relation-
j ships among the entities involved in both the reviews and the design process.
j Primary emphasis was placed on reviewing the adequacy of design details (or
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products) as a means of measuring how well the design process had functioned
j in the selected sampling area. In reviewing the design details, the team
- focused on the following items:
!4 Validity of design inputs and assumptions. 1
i Validity of design specifications.
Validity of analyses.
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Identification of system interface requirements.
Potential indirect effects of changes,
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j Proper component classification.
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- Revision control,
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Application of design information transferred between organizations.
Design verification methods.
The team inspected five engineering disciplines within the project. The five
disciplines were mechanical systems (Section 2), mechanical components
(Section 3), civil / structural (Section 4), electric power (Section 5), and
instrumentation and controls (Section 6).
1.2 SUMMARY - MECHANICAL SYSTEMS
The team reviewed the mechanical systems related changes made to the main
feedwater/ auxiliary feedwater system, unimplemented and partially implemented
ECNs for the essential raw cooling water system, the reactor coolant system and
the chemical and volume control system, and selected completed ECNs and Work
Plan packages for the essential raw cooling water system.
The team found that a facility change which was accomplished via a TempoVa'ryH
Alteration Control Form remained in effect for several years without completion
of the Work Plan, closeout of the ECN and final pre-operational testing. Tem .
porary Alteration procedures were used, in effect, for permanent design and
management controls and, in at least one instance, did not provide timely engineer-
ing review and closure, including appropriate post-modification testing
(Deficircy D3.2-1).
The team also questioned the engineering analysis associated with ECN L6317.
This ECN implements the injection of Furmanite to stop a steam leak from a
valve body gasket in an ASME Section III, Class B system. The procedure
requires drilling into the valve bonnet, and the potential for drilling into
- closure bolting and redistribution of stress. The team was unable to find
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stress analysis or other documentation showing that the stress remains within
allowable values following application of the procedure used for ECN L6317
(Deficiency D2.1-1).
The team performed an overview of the G/C review of changes made to the
main / auxiliary feedwater system and TVA review of unimplemented ECNs. Except
for the deficiency and unresolved item noted above, the team concurs with the
results and conclusions of these reviews in the area of mechanical systems.
1. 3 SUMMARY -
MECHANICAL COMPONENTS
The team reviewed the three technical issues and three observations which G/C
identified in the piping discipline. The team also reviewed the equipment
seismic qualification portions of four of the nine technical issues and two of
the nine observations which G/C identified in the instrumentation and controls
discipline. The team considers the G/C review of the main and auxiliary feed-
water system to be adequate. However, in one instance, the team identified an
ECN which may have been implemented without a valid unreviewed safety question
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determination (USQD) (Deficiency D3.1-1).
TVA reviewed a total of 52 partially implemented and unimplemented ECNs for the
"3 system review" in order to determine if an unreviewed safety question (USQ)
existed. The team reviewed 16 ECNs relating to mechanical components which
involved modification or replacement of piping, pipe supports or equipment for -
potential USQs. The team concurs with TVA's evaluation of the 16 ECNs. However,
the team identified a second instance in which an ECN may have been implemented
without a valid USQD (Deficiency 03.2-2). ,
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The team also reviewed the design basis for the main and auxiliary feedwater
system at Sequoyah, and identified several instances of TVA's failure to
implement piping code requirements or FSAR commitments.
One example is that friction is not considered in the design of pipe supports,
although the piping code of record (B31.1) for Sequoyah Nuclear Plant requires
consideration of friction forces in pipe support design. If pipe supports were
reevaluated considering friction forces in conjunction with original operating
loads, some installed supports may not meet required factors of safety
(Deficiency D3.3-1).
The team found that pipe supports for piping, field routed to generic qualifi-
cation criteria, have not been systematically evaluated -for additional reactions
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due to thermal loads. The team was concerned with the adequacy of corrective
action when informed that the TVA Civil Engineering Branch (CEB) issued two
nonconfermance reports (NCRs) in 1982 which indicated broad concerns, including
evaluation of appropriate loads such as thermal, with the field routed piping
program at Sequoyah Nuclear Plant. TVA CEB has not yet closed out these NCRs
(Deficiency D3.3-4).
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The turbine driven auxiliary feedwater pumps at Sequoyah Nuclear Plant have
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fundamental frequencies in the 13-20 hertz range, well below the FSAR commit-
ment to procure pumps with a minimum fundamental frequency of 33 hertz, Moreover,
the pump nozzles were modeled as rigid anchors in the piping analysis, since TVA
Civil Engineering Branch has no requirements to model flexible equipment in piping
analyses. The team is concerned that piping subsystems which contain flexible
equipment could be subjected to amplified pipe stresses, pipe support reactions
and equipment nozzle loads during a seismic event. Piping subsystems at Sequoyah
Nuclear Plant which contain flexible equipment therefore may not be adequately
qualified by analysis (Deficiency 03.3-5).
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1.4 SUMMARY - CIVIL / STRUCTURAL
The team reviewed all the technical issues and the observations, as they relate
to the civil / structural area, that were raised by Gilbert / Commonwealth. The
team determined that G/C performed a thorough review in determining the root
cause and the technical justification of each modification.
The team reviewed all the fifty-two ECNs from the TVA three system review and
determined that only four were related to Civil / Structural. The team agrees
with the TVA conclusion that these four ECNs in their current configuration
would not involve an Unreviewed Safety Question.
The team also reviewed the technical adequacy of the ECNs from the TVA three
system review, original cable tray support calculations and the seismic
analysis of the refueling water storage tank.
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The team noted that all three ECNs reviewed lacked some aspects of technical
justification for the change performed. In two cases, reinforcing bars were cut
without evaluation of the structural adequacy of the elements in question (e.g.,
in the pressurizer cavity wall) (Deficiency D4.3-1). In the other case, platforms
were added to the lower steam generator supports without verifying that original
design requirements were not violated (Deficiency D4.3-3).
The original calculations for the cable tray supports showed that there were
deviations from the design criteria. The wrong acceleration response spectra
curve was used in design (Deficiency D4.3-4). Also, approximately 10% of the
cable tray supports reviewed used an unconservative design load (Deficiency D4.3-5).
The team noted that torsional shear stresses were not considered in the weld
design of asymmetrical cable tray supports (Deficiency D4.3-6).
1. 5 SUMMARY - ELECTRIC POWER ;
The team reviewed three technical issues which G/C identified in the electrical
power area. The team also reviewed 6 ECNs which were reviewed by G/C for
auxiliary feedwater system modifications. The team considers the G/C review of
ECNs for auxiliary feedwater system to be adequate in the electrical discipline.
The team reviewed 5 ECNs related to the TVA "3 system review" in the electrical
area and determined that the methodology used by the Office of Engineering for
i this review was acceptable. The team agrees with the TVA review's conclusions
i and recommendations. In addition, the team reviewed TVA's design control program
used for plant modifications and selected electrical power calculations.
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The team findings raise questions as to the adequacy of the system to review and
update electrical calculations which verify that the existing configuration of
the plant systems support all modes of plant operation. The team identified that
some calculations (such as the station battery sizing calculation, battery charger
sizing calculations and the 120V Vital ac inverter sizing calculations) were not
performed while other electrical calculations (such as those for 125V dc voltage
drop, emergency diesel generator load analysis, ac auxiliary power supply voltage
and loading analysis) were performed recently and the current versions are not
finalized. The team also found that TVA did not perform an analysis of the trip
settings of the thermal overloads of motor operated valves in safety-related
, systems.
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The team noted that TVA has generated detailed engineering and administrative
procedures, but that in some cases these procedures either were not implemented
or were implemented improperly.
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1.6 SUP9tARY - INSTRUMENTATION AND CONTROL
During the Sequoyah design control inspection, the team reviewed approximately
eighty design modifications involving the instrumentation and control
discipline. Results of the Gilbert / Commonwealth review of the main and
' auxiliary feedwater system were reviewed, and proposed actions in response to
specific technical issues were evaluated. The team reviewed results from the
Tennessee Valley Authority three system review, and conducted an independent
review of other design modifications, technical issues, completed temporary
alteration control forms, work plans, and surveillance test procedures.
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As discussed in section 6.1, the team determined that the Gilbert / Commonwealth
review was comprehensive for this technical discipline. The team expanded upon
the Gilbert / Commonwealth review by examining the effect that several design
modifications had on the original plant design basis, and determined that TVA
had not naintained control of the original or interim plant design basis for
either procurement of replacement instruments or setpoint calculation changes
(Deficiencies D6.1-1 and D6.1-3). In addition, a solenoid valve replacement
design modification did not satisfy a feedwater isolation requirement and
invalidated the unreviewed safety question determination (Deficiency D6.1-2).
The team determined that the TVA three system review, which assessed the safety
impact of unimplemented and partially implemented engineering change notices on
subsequent safe operation of the plant, was satisfactory for the instrumentation
and control discipline. Again, the team expanded on the scope of the three system
review by examining the adequacy of the proposed technical approach described in
a number of engineering change notices. The validity of a TVA Nuclear Engineering
Branch analysis that would have authorized downgrading in the environmental quali-
fication of reactor coolant system resistance temperature detectors used to
generate the over-power and over-Temperature delta-T reactor trips violated IEEE-
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279-1971 (Deficiency D6.2-1). '
A systematic weakness in the specification of vendor hydrostatic pressure tests
to demonstrate instrument pressure boundary integrity after seismic qualification
testing was identified by the team (Deficiency D6.3-1). Evidence of a possible
systematic weakness was also found in the determination of quality assurance and
seismic analysis requirements to be applied to individual design modifications
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(Unresolved Item U6.3-2).
In two specific instances involving the auxiliary feedwater pump discharge
pressure switch and the feedwater bypass control valve solenoid operator, the
team was informed that interim component replacements had been accomplished
without an authorizing engineering change notice (Deficiencies D6.1-1 and D6.1-
2).
1.7 G/C REVIEW CONCLUSIONS AND RECOMMENDATIONS
G/C evaluated about 700 documents, including 109 engineering change notices
(ECNs). The G/C review team identified 19 technical issues and 18 observations.
G/C defined apparent failures to meet design or licensing criteria as technical
issues. Issues which G/C considered less significant, such as documentation
discrepancies, were identified as observations. G/C prepared recomended actions
for each technical issue. The G/C report also details the action plans which TVA
drafted in response to G/C's recommended actions. TVA did not respond to issues
which G/C categorized as observations.
G/C concluded that, assuming completion of the TVA action plans and resolution
of open items, the modifications made to the main and auxiliary feedwater system ,
will have maintained the technical adequacy of this system since the operating
license was issued. Due to the root causes of some of the technical issues, this
conclusion could not be extrapolated to other systems without further evaluation
of additional modifications.
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These root causes identified by G/C are the lack of:
Engineering Evaluation.
Documentation of Evaluation.
Detail in Design Output.
Identification of Design Input and Design Bases.
Additionally, the G/C team concluded that plant configuration control and Un-
reviewed Safety Question Determination activities require further attention.
Resolution activities were conducted during the G/C review. The final report
contained Technical Issue Data Sheets which in addition to describing the
technical issue involved, reflect descriptions of actions taken or planned
by TVA. The status of the resolution activities is as follows:
One Technical Issue Data Sheet (No. 3) is closed, based upon
recognition of an existing program.
Most of the Technical Issue Data Sheets have acceptable resolution
activities planned and, in some cases, accomplished.
The team reviewed the proposed corrective actions by TVA for the specific
technical issues identified by G/C during the inspection. No significant
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problems were identified regarding these actions insofar as they were defined.
TVA actions to address root causes identified by G/C and to address G/C observa-
tions had not been identified at the time of the inspection.
1.8 TVA THREE SYSTEM REVIEW CONCLUSIONS AND RECOM ENDATIONS
The TVA Office of Engineering evaluated a total of 52 unimplemented and partially
implemented engineering change r.otices for the three systems reviewed, and
concluded that 14 of the 52 ECNs reviewed involved a USQ. However, 12 of the
ECNs were either scheduled for completion prior to restart, or scheduled for
review to confirm that partial implementation at time of restart would not
involve a USQ. TVA OE concluded that two ECNs had not been scheduled for
completion prior to restart, and that failure to implement these ECNs prior to
restart would have resulted in a USQ. The NRC team noted that the TVA "3 system" :
review was not intended to be a detailed engineering evaluation of the proposed I
changes and in general concentrated only on the need to complete changes based
upon NRC commitments and safety analysis considerations.
3 The TVA review team also identified weaknesses in the configuration control area j
which related to these ECNs. One weakness was that TVA lacked an adequate system ,
to track unimplemented and partially implemented ECNs. This was evidenced by the ,
need to drop 17 ECNs and add 16 ECNs to achieve the final 3 system sample of 52 ,
unimplemented ECNs from the original status list. The "as-configured" drawings i
were also noted to have numerous mistakes. In addition, the TVA review team ,
considered the large number of unimplemented ECNs to represent a potential problem. i
The TVA review team recommended that:
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, A review be initiated for all safety-related systems to assess ECNs not
! scheduled for completion prior to restart. In addition existing procedures
should be strengthened to ensure that a USQD is performed for all partially
, implemented ECNs.
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i A review be initiated to check all "as-configured" drawings located in the'
control room to verify their accuracy. In addition existing procedures
should be strengthened to ensure that these drawings are accurately main-
- tained in accordance with accepted quality assurance practices.
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Procedures should be implemented to review all ECNs over two years old
which have not been implemented to determine if the modification is still
necessary. If the ECN cannot be justified, the ECN should be cancelled
and the "as-designed" drawings should be revised to "back-out" the
, modification.
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Administrative Instruction-19 should be revised to require OE concurrence
i on all USQDs prepared by the Office of Nuclear Power for partially imple-
- are not irnpacted.
The ECN form should be revised to specify if implementation of the ECN is
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required to resolve a Condition Adverse to Quality (CAQ) or to comply with-
a licensing commitment.
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i Criteria should be established to define and/or limit the scope of ECNs.
- Several ECNs were so general that the scope could not be identified.
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Several of the configuration control issues identified during the 3 system review
were outside the scope of this design control inspection and were not examined
by the team.
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j 1.9 OVERALL CONCLUSIONS
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The NRC team considers that the G/C and the TVA "3 system" reviews were adequately
- implemented to achieve their objectives. Both reviews resulted in recommendations
i and conclusions which relate to the adequacy of the overall design control
i process. In general, the NRC design control inspection findings validate these
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findings and recommendations. i
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In addition to the four " root causes" identified by Gilbert / Commonwealth, the !
4 following concerns were identified by the NRC team during the inspection.
In several cases standard industry codes and practices were not followed
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in the samples of original design work examined by the NRC staff in
conjunction with the review of the Gilbert / Commonwealth effort'.
l There are a lack of available calculations supporting the original design
in some disciplines. For example, calculations do not exist to support-
j the sizing of the station batteries, vital inverters, and battery chargers.
'; This is also evidenced by the auxiliary power system reviews being moni-
tored by the NRC Office of Nuclear Reactor Regulation and by lack or-
j unavailability of EDS performed pipe support calculations.
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Temporary alteration procedures have been used for permanent design; ,
management controls did not provide timely engineering review and closure.
Seismic requirements were not being adequately handled as design input in
some instances nor were they translated into design output documer.ts, as '
< evidenced by lack of specification of these requirements in purchase documents
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and vendor data sheets for vendor equipment.
Final design implementation and details do.not always agree with assumptiens ,
.and statements in Unreviewed Safety Question Determinations.
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The more significant specific NRC team findings which support the concerns
identified above included:
The failure to include friction in the design of supports for large bore pipe. l
) The lack of systematic evaluation of thermal loads in pipe supports for
i field routed pipe.
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The failure to model flexible equipment in piping analyses, i
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The failure to consider the effects of cutting rebar (for example in the
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i The use of incorrect loading assumptions in about 10% of the cable tray :
j calculations examined. ~;
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, The lack of control of setpoint calculations in the instrumentation and ';
- control area. -
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} A final NRC determination regarding the overall. adequacy of design control for
l Sequoyah could not be made, considering the-need for significant corrective 3
! actions to address G/C and TVA reviews and the.NRC team findings. The NRC will !
continue to monitor these actions. ,
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Positive observations by the NRC team during the inspection included controlled ;
! and comprehensive design guides and specification in some disciplines and a com- ;
puterized data base for design document (e.g.-calc'11ation) retrieval.
l 2.0 MECHANICAL SYSTEMS ;
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I.. This portion of the inspection evaluated the mechanical systems design aspects
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of actual and proposed plant modifications in several systems with emphasis on
technical adequacy. The principal areas of investigation were: .
An overview of the Gilbert / Commonwealth technical review of changes made
to the main feedwater/ auxiliary feedwater system since the time Sequoyah i
received its operating license. '
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An assessment of TVA's review of unimplemented ECNs for selected systems. ;
) Review of additional ECNs for the essential raw cooling water and support i
j systems for the main / auxiliary feedwater system, such as heating, !
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ventilation and air conditioning. i
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2.1 OVERVIEW 0F THE GILBERT / COMMONWEALTH TECHNICAL REVIEW 0F MAIN FEEDWATER/
The Gilbert / Commonwealth review team reviewed a -total of fifteen (15) ECNs and
twenty-two (22) Design Change Requests that modified the main and auxiliary
feedwater system. The Gilbert / Commonwealth review identified one technical
issue in the area of mechanical systems. Technical issue 1 is related to ECN
L6112.
This ECN permits the installation of a globe valve in the feedwater isolation
valve stem leakoff lines. This would isolate the leakage path from the inner
valve packing to the leakage collection system resulting in the outer valve
packing becoming a part of the pressure boundary. Gilbert / Commonwealth observes,
however, that closing leakoff lines would permit a potential pressure buildup
on the outer layers of stem packing leading to the following concerns:
i Leakage through the outer stem packing, which could spray the valve
motor operator and render the isolation valve inoperable.
Creation of a pressure / temperature in the bonnet area which could prevent
or restrict valve operation.
Potential for wire drawing along the stem in the upper layers of packing.
The TVA response to this finding is that, due to the service conditions of
these valves, the likelihood of leakage through the outer packing is very
j minimal and the possibility of the bonnet pressure / temperature affecting valve
operability is not a serious concern. However, TVA has reviewed its practice
of capping valve leakoff lines and has decided to remove previously installed
caps on the basis that a cap is a permanent modification rather than the intended
- temporary measure. On leakoff lines where globe valves have been installed in-
stead of caps, the globe valves will be kept closed until the valve is repacked
or adjusted. TVA will periodically inspect feedwater isolation valves with closed
globe valves in leakoff lines for leakage through the outer packing. The
inspection team finds this to be an acceptable resolution of this technical issue.
The review of G/C obcervation No. 2 resulted in an additional technical concern
,
of the NRC inspection team. ECN L6317 was used to plug a steam leak in the bonnet
flange of check valve L-VLV-3891. Plugging is accomplished by injecting a compound
known by the trade name "Furmanite" into the area outside of the bonnet to body
gasket. The G/C observation addressed an apparent drawing error depicting
,
several valves as gate, versus globe valves.
The NRC team's concern with ECN L6317 is the lack of consideration of the change
in stresses in the closure flanges and closure bolting that could result from
,
drilling and tapping to inject Furmanite. The Furmanite technique has been used
frequently at TVA to temporarily stop valve leaks. Two other ECNs, L6169 and
L6157, used to repair leaking valves in the auxiliary feedwater, were reviewed
by the team. Stress analyses were included with both of these ECNs, thus indi-
cating that changes in stress were a consideration with other Furmaniting
operations (Deficiency D2.1-1).
s
, 11
,
. -- - - - - . . - - . . . - . . . - - . . - -- . - - . . . - .. -
'# . .
.
!
b
l 2.2 ASSESSMENT OF THE REVIEW 0F UNIMPLEMENTED AND PARTIALLY IMPLEMENTED ECNs
The TVA review of the essential raw cooling water system unimplemanted and i
l that were determined to involve unreviewed safety questions.
1
!
resolve problems of reduced flow to components cooled by essential raw cooling
water that have been caused by corrosion buildup in small diameter carbon steel :
piping. TVA noted that the unimplemented parts of the ECN will be resolved in
.. conjunction with ECN L6534. The latter ECN, scheduled to-be implemented prior
j to restart, will revise essential raw cooling water piping drawings to reflect
- the "As-configured" plant, i.e., remove stainless steel piping from drawings
where it has not been installed. TVA will then analyze the configuration to
- . resolve any Unreviewed Safety Questions prior to restart. Those carbon steel-
j lines not involving an Unreviewed Safety Question will be changed to stainless
- steel after startup and will be implemented by new ECNs based on on-going plant
{ performance testing.
1 ECN L6014 was issued to modify essential- raw cooling water pump motor' bearing
f cooling lines. The present lines are 1-inch diameter and are exposed to freezing
temperatures. The heat tracing on these lines is non-IE and not annunciated in
the control room. During freezing weather there is a potential that the lines
could freeze and prevent cooling of the pump motor bearings. The unimplemented
ECN is to make the lines self-draining. The ECN will not be implemented before
i startup. Instead, a special requirement will be put in place to either inspect
l- the heat tracing to verify operability or all the pumps will be operated during.
freezing conditions.
The inspection team finds the TVA resolution of the above unimplemented/
i partially implemented ECNs to be satisfactory, ,
!
j In addition to the above essential raw cooling water mechanical system ECNs,
- the inspection team identified ECN L5500 as needing clarification and
! resolution. The ECN requires locking devices to be installed on sixteen valves
i to assure proper system alignment. The ECN status reported by TVA was that 8
i out of sixteen valves identified in the ECN are locked. A walk-down by the
1 inspection team identified that an additional 4 of the 16 identified in the ECN
i (182-67-718 A&B),arealsolocked. This leaves the locking of valves 1&2-67-
j 507, A&B, in the open position as the unimplemented portion of ECN L5500. These
- 4 valves are in the essential raw cooling water supply lines to the emergency
j diesel generators. The inspection team observed that the four 67-507 valves and
j cross connect piping provide a configuration that will not prevent flow of cooling
, water to any of the emergency diesel generators if one of the 67-507 valves is
mistakenly closed. This is because the valves and-piping configuration form
+
redundant flow paths to the diesel generators. However, Form E56021.01, "Unimple- '
j mented or Partially Implemented Design Item Review" for ECN L5500, can be inter-
'
preted as implying there is a remaining Unreviewed Safety Question. TVA agreed
j to clarify the determination.
The TVA review of the reactor coolant system unimplemented and partially imple-
mented ECNs resulted in two mechanical engineering related ECNs that were
determined to' involve USQs. The two ECNs implemented involve modifications ,
i to the pressurizer safety valve loop seal lines to both prevent excessive steam '
leakage and reduce loading on the safety valve discharge piping and supports. l
l Both were scheduled for completion before restart. ;
i 12
l
.
mum-uy- - - -
,..fw-gs-.-w-em.,-my- p p g. -y y y ma m &-* m.-aw. g g;y--gi.m--w* e., .rw--,gy3++ met --py'.-- a-m gq=-y.,p,w .s a 94ie y-yemam7n--ga.a.y-%vv'-e* er ' fr--T-N-eT1 m n a ' m W't*
. - - - _ _ _ _ _ _ _ _ _ _ _ _
. .
.
ECN L6401 A was to remove the loop seal drain and cap the line. The drain, from
the bottom of loop seal back to the pressurizer, prevented accumulation of con-
densate in the loop seal. Removing and cooping the lice allows the loop seals
to fill the pipe with liquid condensed from the-pressurizer steam space. This
condensate prevents steam from leaking through the safety valves. The outstand-
ing work, to leak test the capped drain line, was scheduled for completion prior
to restart.
The inspection team considers the modifications to the loop seals to be adequate
provided the leak test is completed.
A review of the unimplemented and partially implemented ECNs for the Chemical
and Volume Control System did not identify any mechanical system related ECNs
with unreviewed safety questions.
2.3 ADDITIONAL REVIEW BY THE NRC
TVA has implemented a significant system change to the auxiliary feedwater system
by ECN L5842. This ECN replaces unreliable auxiliary feedwater motor driven
pump discharge valves PCV-3-122 and PCV-3-132 with cavitating venturis. The
function of the flow control valves was to limit pump flow and prevent pump
cavitation during design basis events, such as a main steam line break when the
,
steam generators depressurize and subject the auxiliary feedwater pumps to lcw
'
discharge pressure runout conditions. The reason for replacing the flow control
valves was excessive maintenance of the hydraulic control system used to operate
the flow control valves.
The venturis installed in the pump discharge lines are passive devices designed
with the intent to limit motor driven auxiliary feed pump runout ficw to 600
gpm per pump by cavitating in the throat of the venturi. Conversely, the
design must allow 440 gpm flow per pump under normal conditions when the steam
generators are at pressure. The venturis have been installed and undergone
post modification testing (PMT). Tne inspection team reviewed calculations,
purchase orders for the venture and the results of the post modification tests.
The inspection team observed that performar.ce deviations found during testing
are in the process of being resolved by TVA.
The inspection team reviewed several essential raw cooling water ECNs that had
a status of " complete" in the files of TVA Office of Engineering. This
involved the review of the ECN and the contents of related Work Plan files.
This review resulted in a concern that Temporary Alteration Control Forms are
- being used as a method to permanently modify a system without completion and
closeout of ECNs.
The inspection team has found the change from flow control valves to cavitating
venturis to be satisfactorily implemented subject to resolution of the test
deviations.
ECN L5320, dated 10/23/80, was issued to replace essential raw cooling water
pump motor switchgear overcurrent relays with larger capacity relays. The ,
change was initiated to avoid sticking or fusing of the relays that may prevent
pump restart. The condition was determined to exist during Preoperational
Tests of larger essential raw cooling water pumps installed under ECNs issued
prior to the Sequoyah operating license. The problem was reported in Preoperational
13
.
=__ , . . . . , . - - . .--.- - .- .- -
- . .
3
.-
1-
! Test Deficiency PT-566. The Office of Engineering status of the ECN is "com-
- plete," but it has not been closed out by written notification of completion. i
'
by the Office of Nuclear Power. The inspection team was advised that the Work
= Plan, WP8957, has been cancelled and the field change incorporated per
! Temporary Alteration Control Form 80-734-67, dated 11/14/80. The Temporary-
'
Alteration Control Form appears to be still in effect.
'
The concern with the above situation is that the Temporary Alteration Control
Form has continued to remain in effect since 1980 without completion of a Work.
Plan and closecut of the ECN. This is contrary to TVA's policy concerning
limited use of Temporary Alteration Control Foms. In addition, without a work
' plan, tests showing the new relays are adequate have not been performed and PT-566
'
has not been closed out (Deficiency D2.3-1).
f in other cases, the inspection team reviewed completed ECNs and their work plan
- packages where testing to verify adequacy of ~ the modification was successfully.
4
accomplished and documented. These included:
f ECN L5369 - Modify Logic Function of FCV-67-152 (WP 10117).
i
}' ECN L5651 - Modify Essential Raw Cooling Water Supply Valve to SI Pump
OpeningLogic(WP10467).
3. MECHANICAL COMPONENTS
I This portion of the inspection evaluated the mechanical component aspects of
Gilbert / Commonwealth's review of the main and auxiliary feedwater system for
'
] Sequoyah and TVA's review of unimplemented and partially implemented engineering l
<
j change notices for the RCS, CVCS and ERCW, systems. The team also reviewed, on
a sampling basis, the design basis for piping, pipe supports and equipment. l
3.1 OVERVIEW 0F THE GILBERT / COMMONWEALTH TECHNICAL REVIEW 0F MAIN FEEDWATER/
I
! The team reviewed the three technical issues and three observatior.s which G/C
'
identified in the piping discipline. The team also reviewed the equipment
seismic qualification portions of four of the nine technical issues and two of
the nine observations which G/C identified in the instrumentation and controls .l
(!&C) discipline. The following summarizes the technical issues reviewed by ,
the team and its conclusions relating to the G/C review.
G/C technical issue No. 12 (piping) identified a TVA piping analysis procedure
,
for combining pipe support loads which failed to identify the largest magnitude ;
i of load reversal. G/C's recommended actions included correction or replacement '
! of the TVA procedure and identification and reevaluation of all affected unsym-
! metrical pipe support designs. TVA will assess the potential impact on unsym-
t metrical pipe support designs. TVA also noted that the TVA computer program
- which contained this error has been replaced.
!
i' G/C technical issue 17 (piping) was based upon G/C's review 'of two piping
{ subsystems which TVA reanalyzed for decreased operating temperature. G/C
- identified inconsistencies in TVA's handling of the increased pipe support loads ,
i l
t :
'
l
4
'
14
i
-- .. - - - - - - . =- . .
.- i
i
which resulted from the reanalysis. G/C recommended that TVA reconcile the
j. latest issues of the pipe support load summary sheets and the pipe support
4
design sheets to the latest revisions of the piping analyses. TVA will quantify-
'
the requirements for pipe support reanalysis and will apply these criteria to
all pipe supports which do not reflect revised pipe support loads.
! G/C technical issue 18 (piping) identified TVA's apparent failure to reanalyze
,
a piping subsystem using the increased stress intensification factors (SIFs)
!
idantified in a TVA report. G/C recommended that the piping subsystem be
4
' evaluated for the increased SIFs and that TVA use the increased SIFs in future i
piping reanalysis. TVA noted that the corrective actions planned to address G/C
technical issue 17 would insure the consistent implementation of increased
j SIFs for piping reanalysis.
i
i
'
G/C technical issue 4 (I&C) identified the installation of quick exhausters
on four control valves with a greater offset distance from the pipe centerline
!
than the valve centers of gravity. G/C noted that the exhausters could exper-
fence seismic accelerations greater than the limiting seismic accelerations
1 monitored at the valve centers of gravity. G/C recommended that TVA seismic
,
qualification of components added to existing equipment be performed with
, respect to the installed configurations. TVA's action plan includes a revision .
J of the governing design interface document, and a sampling of past design
i' changes at Sequoyah to determine possible hardware impact due to inadequate
seismic qualification documentation.
! G/C technical issue 7 (I&C) identified replacement solenoid valves which I
! lacked seismic qualification documentation. G/C also questioned the seismic
adequacy of the solenoid valve mounting details. G/C recommended that TVA
,
provide assurance that the seismic qualification of the solenoid valves'is
appropriatefortheirinstallation,andthattheexistingmountingdetailsto
j be modified conform with recommended installation practices. TVA s action plan '
- indicates that the solenoid valves do not require seismic qualification because
i of the unique function and location (non-seismic turbine building structure).
However, TVA will detail solenoid mounting requirements to limit seismic responses.
j G/C technical issue 9 (I&C) identified the installation of replacement solenoid
,
valves on level control valves subject to possible seismic accelerations greater
i than the limiting accelerations monitored at the level control valve centers of
! gravity. G/C initially identified this concern in technical issue no. 4. G/C's !
.
] recommended actions and TVA's commitments to resolve this technical issue are
j discussed in technical issue 4. j
i
!
s
G/C technical issue 19 (I&C) identified the improper remounting of a resistur ;
i box from within the turbine-driven auxiliary feedwater pump speed controller.
i The Unit I resistor box was not mounted on the wall as required by the ECN, but
-
was instead supported from a conduit attached to the speed controller box. G/C -
' recommended that the installation of the Unit I resistor box be modified to be t
a rigid wall mount. TVA has committed to both generic and specific corrective
i action to address this technical issue.
I
i
The team also reviewed five G/C observation:, relating to the mechanical components !
area. TVA did not develop action plans to address observations as part of the ~
G/C report, so corrective actions were not developed for all the observation. ;
The team identified one finding relating to these observations. ,
. +
1 i
'
15 t
- t
I-
f
l
'
, , - . - -.
.- . . . . ~ -_,w- - . , - . . ~ . . - - , _ , _
_ _ -- .~. _ _ _ - - . _ . . _ _ . _ __. ____. _ _
. .
.
i
. G/C observation 6 (I&C) is related to technical issue 4, which concerned
i the installation of quick exhausters on four control valves. G/C noted a lack
'
'
of qualification documentation for the quick exhausters installed on the control
valves. The team noted that the exhausters were installed "at-risk" without docu-
1
mented evaluation with respect to the vendor's seismic qualification report.
Such an evaluation was specified in the unreviewed safety question determination
4
(USQD) which TVA originally prepared for the ECN which authorized the installation
, of the exhausters. TVA's failure to track and implement the USQD requirement to
1
evaluate the quick exhausters and instrument tubing may have voided the USQD
(Deficiency 03.1-1).
} The team considers the G/C review of the main and auxiliary feedwater system to
, be adequate, with the exception noted. The team also notes that G/C did not-
! generally specify the TVA procedures governing the scope of work which G/C
, reviewed, in order to establish a detailed basis for the technical issues and
- observations which G/C identified. In addition, the Gilbert scope of work in
j the mechanical components discipline did not appear to include a detailed review
of FSAR commitments. The team considers the TVA action plans developed for the
'
G/C technical issues adequate to correct the issues, if properly implemented.
3.2 ASSESSMENT OF THE REVIEW 0F UNIMPLEMENTED AND PARTIALLY IMPLEMENTED ECNs
)
"
In the mechanical components area, the team reviewed 16 of the 52 ECNs which TVA
OE reviewed, broken down by system as follows: 10 of 29 ECNs for the essential
i raw cooling water system, three of which involved a USQ; four of 17 ECNs for the
i neither of which involved a USQ. These ECNs involved modification or replacement
j of piping, pipe supports and equipment.
i
l The team reviewed the adequacy of the USQ determination which TVA OE made for
! re qualify modified or replacement piping, pipe supports and equipment. Seven
i of the 16 ECNs which the team reviewed in the mechanical components discipline ,.
a are summarized below.
'
i
ECN L5009 was written in 1981 to change carbon steel piping and valves in the
essential raw cooling water system to stainless steel piping and valves because
of corrosion problems. The changeout included the majority of 2 in, and smaller
lines. The TVA OE review noted that the majority of the work within the scope
i
t
of the ECN had not been implemented in the field. TVA therefore addressed the-
original scope of ECN L5009 in the following manner: (1) the TVA Offices of
,
'
Nuclear Power and Engineering established a level of work within the scope of :
the ECN for scheduled completion prior to restart; (2)'ECNs L6534 and L6560 were
written to void other portions of the original scope of work to reconcile the
j
as-designed and as-confiqured piping drawings, and (3) remaining items not com-
l plate or not scheduled for completion by restart would be implemented, as needed j
j by a future ECN. The TVA DE review team concluded that a USQ existed, pending -(
j analysis of the anticipated system configuration prior to restart, j
f ECN L5017 was written in 1980 to add a motor operator and associated controls to !
a manual valve in the essential raw cooling water system. The motor operator and ,
controls had been installed, with the exception of power connection to a valve I
heater. However, ECN L6258 was subsequently issued to remove power from the l
, valve, and to lock the valve open. The TVA DE review team therefore concluded
r
that a USQ did not exist. The team noted a documentation problem in that the
16
s
1
.. . = , . . . . --.---.-. .- - - . - - - - - - . - - -.
. e
'
.
^
.
j
i valve summary sheet prepared for the piping reanalysis which qualified the added
- motor operator did not tabulate the weight of the valve operator or the offset
f
dimension of the valve operator center of gravity. However, the piping isometric
j drawing and the computer analysis did contain the correct data (Observation 03.2-1).
i
i
4
ECN L5235 was written in 1980 to resolve deficiencies in essential raw cooling
water flow to the electric board room and main control room air conditioners.
4_ Stainless steel piping was installed but not connected to existing carbon steel
! piping. In 1985 ECN L6491 was written to void ECN L5235, as reevaluation had- '
j reduced the required flow rates to the equipment. The TVA DE. review team deter- '
- mined that the installed stainless steel piping was being analyzed for position ,
l. integrity under seismic loads to confirm that adjacent piping and equipment 4
L would not be impacted during a seismic event. The analysis was scheduled
J a USQ would not exist. The NRC team agreed with the TVA conclusion. '
l ECN L5500 was written in 1982 to install valve locking devices in the essential
i raw cooling water system. The TVA DE review team determined that eight of 16
valves had been locked, and noted that no open work plan existed to complete the
ECN. However, the TVA DE review team concluded that no USQ existed, since system
!
function had not been degraded. The team found the status of this ECN to be
I different than that stated by TVA OE, as previously discussed in section 2.2.
1- L5500 also authorized installation of extension operators for two of the valves.
1 The USQD prepared for the ECN specified a seismic analysis to confirm that the
!* new valve stem operators would not invalidate the existing seismic qualifications
t for the valve, piping or associated components. However, the TVA Civil Engineer -
) ing Branch was not able to obtain the new valve seismic documentation for review,
i (Deficiency 03.2-2). ECN L6534, as previously noted, was written in 1985 to void
} portions of the scope of work originally authorized by ECN L5009. The TVA DE
,
-
review team noted that the ECN was partially implemented but was scheduled for
completion prior to restart, and that failure to implement the ECN would involve
, a USQ.
ECN L5737, written in 1982, added a check valve upstream of a primary water valve :
.
in the CVCS system. The TVA OE review team determined that the check valve had
i
not been installed, and would not be installed. However, the TVA DE review team
i
concluded that a USQ did not exist, since the primary water system was not neces-
l sary to mitigate a design basis event. The team noted that the revised piping =
i
flow diagram which implemented the ECN design change failed to show a required
i piping class break in the line containing the new valve (Deficiency 03.2-3).
i ECN L6462, written in 1985, authorized the installation of a support for a 3/4
) inch sample connection coming off the downstream piping of the letdown heat
exchanger. The TVA DE review team concluded that no USQ was involved,'since the
i
support had been installed and was scheduled for post-modification testing prior
! to restart. However, the team noted a technical problem with the change In that
the support clamp used 3/4 inch bolts instead of the 7/8 inch bolts specified on
the standard support detail (Deficiency 03.2-4). ,
The team concurs with TVA OE's evaluation of the 16 ECNs which the team reviewed
for potential USQs. However, the team documented four ovficiencies during the
!
course of this review. The team considers Deficiency 03.2-2, which documents
l TVA's failure to track and implement a USQO commitment, to have potential generic
j significance (see also Deficiency 03.1-1).
!
i
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- 17
i
{
i
I .
!
.. _ . - _ __._ ._. . . .. - _ .- ___. ._ _
. e
,
i
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3.3 ADDITIONAL REVIEW BY THE NRC
l During the evaluation of the Gilbert and TVA reports which are detailed in
,
Sections 3.1 and 3.2, the team also reviewed the mechanical components design
l basis for the main and auxiliary feedwater system at Sequoyah. Deficiencies
which the team identified during the course of this sampling review are sum-
,
'
marized below.
! Friction is not considered in the design of pipe supports for Sequoyah. This
i is a formal design policy for Sequoyah Nuclear Plant which is detailed in Section ,
i 7.19 of-the TVA pipe support design manual: "There is no requirement to use fric-
tion in the design of pipe supports at SQN." The team notes that it is an in-
y dustry standard to account for frictional forces associated with piping thermal
.
displacements of 1/16 in, or greater. Moreover, USAS B31.1-1967, the piping
code of record for TVA safety class B, C and D piping at Sequoyah Nuclear Plant,
requires consideration of frictional forces due to piping themal expansion
l (Deficiency D3.3-1). Pipe supports reevaluated for applied frictional forces
in conjunction with original operating loads may not meet required AISC code '
factors of safety.
j
'
FSAR Section 3.9.2.5.2 specifies maximum seismic valve accelerations of 2g
vertical and 3g horizontal for the safe shutdown earthquake condition.
However, some valves installed at Sequoyah Nuclear Plant are being qualified on
-
a case by case basis to computed acceleration levels which exceed these
j
'
allowables (Deficiency 03.3-2). There is no current TVA procedure for Sequoyah
Nuclear Plant which details the qualification of valves to accelerations higher
than 2g vertical and 3g horizontal.
! FSAR Tables 3.9.2-1 and 3.9.2-3 specify a minimum fundamental valve frequency
i of 33 hertz. However, the TVA specification used to procure valves for Sequoyah l
, Nuclear Plant prior to 1975 specified a minimum fundamental frequency of 25 hertz l
(Deficiency D3.3-3). It appears that little amplification of the design floor
,
response spectra occurs for the frequency shift from 33 hertz to 25 hertz, so that
valves procured tn this specification do not appear to be subject to higher
loads than originally considered. This preliminary conclusion is based upon
cursory team review of the design floor response spectra, and is subject to
more detailed review by TVA technical personnel.
1
FSAR Sections 3.9.2.5 and 3.9.2.6 specify piping which may be field routed to
generic qualification criteria es a function of TVA piping safety class, dia-
'
meter, material, temperature and pressure. Typical seismic supports for process j
pipe 2 inch diameter and smaller are detailed on TVA 47053-series drawings. Sheet j
.
'
1A of the series details the general definitions, requirements and guidelines to i
be followed when using the typical seismic support drawings to field route pipe. '
Note 11 of sheet 1A specifically indicates that "the above guidelines do not
consider thermal expansion or anchor movements." The team requested docu-
mentation to' confirm that pipe supports for field routed pipe subject to thermal
. expansion were being systematically evaluated for additional support reactions
, due to themal loads. However, the TVA Civil Engineering Branch was not able to
3
demonstrate that such loads had been systemmatically evaluated (Deficiency D3.3-4)
Moreover, TVA Civil Engineering Branen provided the team with two-
nonconformance reports written in 1982 which indicated in-house concerns with
the field routed piping program, including correct design loads, at Sequoyah
4 Nuclear Plant. The team was concerned with TVA's lack of timely corrective ~
4
18
1 .
!
k
_
. e
action regarding the structural adequacy of the pipe supports used to field
route pipe at Sequoyah Nuclear Plant.
FSAR Tables 3.9.2-1 and 3.9.2-3 specify a minimum fundamental pump frequency of
33 hertz. The team reviewed the seismic report prepared for the turbine driven
auxiliary feedwater pumps which are installed at Watts Bar Nuclear Plant. The
seismic report was used to qualify nozzle loads for the turbine driven auxiliary
feedwater pumps installed at Sequoyah Nuclear Plant. The free vibration analysis
performed for the pump yields lateral fundamental frequencies of 13 and 16 hertz,
and a vertical frequency of 20 hertz. The pump procurement specification does
not specify a minimum fundamental frequency. The pump nozzles were modeled as
rigid anchors in the piping analysis. TVA has no requirements to model flexible
equipment in piping analyses (Deficiency D3.3-5). The team is concerned that
piping subsystems which contain flexible equipment could be subjected to amplified
pipe stress, pipe support reactions and equipment nozzle loads during a seismic
event. The team does not consider piping subsystems at Sequoyah Nuclear Plant
which contain flexible equipment to be adequately qualified by analysis.
4.0 CIVIL / STRUCTURAL
This portion of the inspection evaluated the structural aspects of the Gilbert /
Commonwealth study of the main and auxiliary feedwater system modifications and
the TVA review of the unimplemented and partially implemented ECNs for the essen-
tial raw cooling water, reactor coolant and chemical and volume control systems.
The team also inspected limited areas of the original design to determine techni-
cal adequacy and adherence to design criteria.
4.1 OVERVIEW 0F THE GILBERT / COMMONWEALTH TECHNICAL REVIEW OF MAIN FEEDWATER
,
The following summarizes those issues which are directly related to the civil / i
l
structural area. l
l
G/C technical issue 2 identified the possibility of overloading the cable tray
supports due to abandoned cables. Due to a lack of tracking or identifying
abandoned cables, original load requirements might have been exceeded in all ,
j
' Category I buildings. The team agreed with this Gilbert / Commonwealth finding.
The TVA action plan states that Electrical Engineering Branch will identify
the areas where the cable tray loads exceed the above design limits. After
such identification the Civil Engineering Branch will perform analysis to deter-
,
mine the structural adequacy of the affected cable tray supports.
G/C technical issue 8 involves the failure of the embedded plates for Unit 1
main feedwater deadweight supports 1-H4-282 and 1-H4-322. Although these
supports were adequately redesigned by the issuance of ECN L6263, TVA failed to
identify the root cause and the generic aspects of the problem. Gilbert /
{
Commonwealth, in their review, identified that the original design of these
i embedded plates did not consider capacity reduction due to lack of adequate edge
distance.
'
The team agrees with this Gilbert / Commonwealth finding. In response
to this finding, TVA initiated Significant Condition Report (SCR) Civil Engineering
Branch-8607 to identify all the embedded plates with possible edge distance
19
_ _ _
.. .
-
. .
I
problems and then try to show by analysis that they meet the current design stan-
dards. This SCR states that TVA will perform systematic walkdowns for the
auxiliary, control and diesel generator buildings to identify all the affected
embedded plates. A review of the reactor building will be based on a drawing
review with inspection of those plates determined to be affected by a free edge.
G/C technical issue 11 is a seismic system interaction which was found during
the Gilbert / Commonwealth walkdown of the plant. G/C noted that a 2" 0.D. pipe
was lucated near the tubing supplying air to auxiliary feedwater bypass valve
2-LCV-3-148A. The lateral displacement of this pipe during a seismic event could
impair the safety function of the air supply. The TVA action plan for this item
states that a walkdown will be performed for all seismic category I(L) systems
and all violations will be documented. TVA will also develop and implement
dispositions for each of the identified violations.
G/C technical issue 14 addresses the lack of seismic analysis for various
modified brackets for the auxiliary feedwater system control valves. These s
brackets were supplied by the vendor, but due to interference some were field
modified. Although Design Change Request (DCR) 814 was initiated to address
the modified brackets for valve 1-LCV-3-148, it failed to include other valve
brackets. Gilbert / Commonwealth identified other valves which have modified
brackets. The NRC team agrees with this finding. The TVA action plan states
that they have identified all the modified braces. The TVA Civil Engineering
Branch will document that these modified seismic brackets are sufficient to
support the valves during a seismic event. 3
Since the Gilbert / Commonwealth teview was restricted to the main and auxiliary
feedwater system, the number of Engineering Change Notices (ECN) reviewed were
limited. For these ECNs, Gilbert / Commonwealth performed a thorough job in
reviewing the root cause and the technical justification. The team agrees
with the Gilbert / Commonwealth conclusion that the results of their review could
not be extrapolated to all structural design modifications.
4.2 ASSESSMENT OF THE REVIEW OF UNIMPLEMENTED AND PARTIALLY IMPLEMENTED ECNs
The team reviewed all the fifty two (52) ECNs evaluated during the TVA review
of unimplemented and partially implemented ECNs to determine those that were
related to Civil / Structural. Of these, only four (4) were related to the struc-
tural area. ECN L5887 was unimplemented, ECNs L5202, L5037 and L6495 were
partially implemented. The team reviewed the contents of these ECNs, as well
as the associated Unreviewed Safety Question Determination (USQD). The team
agrees with the TVA conclusion that in the Civil / Structural area these four
ECNs in their current configuration would not involve an Unreviewed Safety
Question.
The team reviewed in detail the technical contents and the back-up documentation
for ECNs L6495, L5202, and L5307. The team also reviewed the original design
calculations to determine the technical adequacy of the changes generated by
these ECNs.
Due to changes initiated by ECN L6495, a 4 inch sleeve was installed through the
pressurizer cavity wall which resulted in the cutting of various reinforcing bars.
,
The USQD for this ECN states that Civil Engineering Branch would evaluate the
installation of this new sleeve to ensure that the structural integrity of the
wall is not degraded. Contrary to this statement, the team could not locate any
20
- - . -
_
. .
.
calculations which evaluated the effects of cutting these rebars. A similar
problem was noted during the review of ECN L5202. In conjunction with this ECN
various reinforcing bars were cut from the Diesel Generator Brilding Floor.
Again, the team did not find evidence that the structural adequacy of the
floor was investigated for these cuts (Deficiency D4.3-1).
ECN L5034 added platforms for access to the steam generator ports. These
permanent platforms are built on the lower steam generator girders. Although
calculations were performed for the design of these platforms, the team could
not locate any calculations where the structural adequacy of the steam generator
lower support girders was checked (Deficiency 04.3-3).
4.3 ADDITIONAL REVIEW BY THE NRC
Besides the two reviews performed by Gilbert / Commonwealth and TVA, the team
independently reviewed the original design of the auxiliary building cable
tray supports at Elevation 714.0' and the seismic analysis of the refueling
water storage tank.
In the design of the cable tray supports, acceleration response curves for 1%
damping given a N-S earthquake were used. The response spectra curves for the
same amount of damping have higher accelerations for the E-W earthquake. These
latter accelerations should have been used in the design (Deficiency D4.3-4).
TVA design criteria require that for an 18 inch cable tray, the loads on the
cable tray support should be 75 pounds / foot for the top tray and 45 pounds / foot
for the lower trays. Contrary to this requirement, 10% of the cable tray support
designs reviewed by the team showed that only 45 pounds / foot was used for the top.
tray (Deficiency D4.3-5).
The team also reviewed the weld design at the attachment of the vertical
member of the cable tray support to the embedded plate at the ceiling. Due to
unsymmetrical configuration of certain cable tray supports, torsional stresses
will be induced on the welds between the vertical member and the embedded
plate. The team's review showed that such additional stresses were not
considered in weld design (Deficiency D4.3-6).
During the review of TVA drawing 48N1333, the team noticed that a surface base
plate with threaded bolt anchors was used for certain cable tray supports.
The calculations for this base plate used the rigid base plate approach and was
performed prior to the issuance of current TVA design standard 05-C1.7.1, which
requires the consideration of plate flexibility. Recently, a nonconformance
report was issued stating that TVA engineers were still using the rigid base
plate theory in their design of the pipe support base plates. The team believes ;
l
that this might also be true for the design of the cable tray supports (Unresolved 1
Item U4.3-7).
The team further noted that the TVA pipe support design manual still states that
base plates are to be analyzed as rigid, whereas the current civil design standard
DS-C1.7.1 requires the consideration of plate flexibility (Observation 04.3-8).
'21
o .
The team also reviewed the TVA design criteria for seismically qualifying tanks.
This design criteria states that the natural frequency of a tank when considered
to be full of fluid should not be less than 33 hertz. The team did not believe
that large tanks were routinely engineered to achieve such rigidity, and there-
fore reviewed calculations for the seismic analysis of the refueling water
storage tank. The seismic analysis of the tank shovs its frequency to be about
6.5 hertz. Although the team's review did not raise question regarding the
structural adequacy of this tank, the team believes that the TVA design criteria
should be revised to reflect the actual methodology used for seismic design.
(Observation 04.3-9).
5.0 ELECTRIC POWER
, This portion of the inspection evaluated the electric power aspects of the
following three areas:
Inspection of Gilbert / Commonwealth's review of the ECNs completed since
operating license issuance on the, auxiliary and main feedwater system.
Evaluation of findings of TVA's three system review of unimplemented and
partially implemented ECNs.
Review of TVA's design control process for the subject design changes,
including selected aspects of the original plant design. A brief tour of
the plant was conducted to assess the implementation of several modifications.
5.1 OVERVIEW OF THE GILBERT / COMMONWEALTH TECHNICAL REVIEW OF MAIN FEEDWATER/
Gilbert / Commonwealth reviewed 12 ECNs in the electrical area. The team reviewed
,
TVA's action plan for resolution of the technical concerns identified by G/C.
1
The team found G/C's methodology and review program acceptable. The team noted
that the scope of G/C's review was limited to the design control process as it
related to the ECNs, and did not examine original plant design if this did not
relate to the ECNs. G/C's review identified the following technical concerns
in the area of electrical power.
G/C technical issue 2 noted that TVA does not keep track of abandoned cables
which have been removed from electrical circuits but not physically removed from
the trays. Un-energized cables lying in the trays do not have a significant
impact on the derating factor for the energized cables in their proximity.
G/C technical issue 3 identified that TVA could not provide documentation to
confirm that all unqualified cables in the main steam valve room have been
replaced with qualified cables. ,
{
l
G/C technical issue 6 concerned the installation of cavitating venturies and !
removal of existing flow control valves in the AFW pump discharge, which caused !
the design condition horsepower demanded by the pump motors to increase. TVA ;
did not review the ratings of the emergency diesel generator, switchgear, power :
feed cable for the AFW pump motor, and the existing protection circuits to verify
that the installed equipment is adequate to meet the increased horsepower demand.
22
i
l
._ _ _ _ . . =_. ._. _ . _ . . . -
_ _. .
4
.
The NRC team agrees with the G/C technical issues.' TVA action plans reviewed-
by the team for these issues appear adequate to correct the technical issues, if
j adequately implemented.
5.2 ASSESSMENT OF THE REVIEW OF UNIMPLEMENTED AND PARTIALLY IMPLEMENTED ECNs ,
The team reviewed five ECNs related to electric power to assess the methodology
, used by TVA for the.unreviewed safety question determinations. The team found
I the review results and the methodology acceptable.
1
The ECN packages reviewed by the team also contained the USQDs which supported
~
the original concept of the change. These USQD's are performed by the OE Nuclear
Engineering Branch. Although the TVA_"3 system review" was not intended to be- ,
'
_ an evaluation of these USQD's they were reviewed by the NRC team during the -
course of the inspection. The team noted that some unreviewed safety question -
1
determination analyses were very brief. For example, the USQD analysis for
l
ECN-L-5057 did not consider the increased time response of the instrument nor ,
was a failure mode and effects analysis performed, although the modified circuit
has more components and power supplies. This is an apparent weakness in the
, USQD development process, perhaps caused by the development of USQDs before the ,
design details are developed. This aspect was also noted by G/C. -The team also 1
felt that the USQD review would envelope a wider spectrum of attributes if the
reviewers were not limited to one discipline. .
5.3 ADDITIONAL REVIEW BY THE NRC
The team reviewed the design control process used for the ECNs and selected '
j aspects of the original design. This review included the applicable engineer-
! ing and administrative procedures. In addition, the team reviewed the original
! calculations and the related drawings for the 6.9 kv, 125v dc and 120v vital
] ac power systems, associated with the ECNs for the AFW system.
l_
'
The team reviewed the procedures for the flow of information as.it relates to
the issuance and implementation of ECNs and TACFs. The progression of the ECN
process generates two controlled sets of drawings one "as-designed" and the other
"as-configured." TVA has committed to implement a drawing control system which I
will provide engineering a current, as-constructed status. Such a system was not l
developed at the time of the inspection. The team noted that the "3 system"
review identified numerous errors and inconsistencies between the "as-designed
and "as-configured" drawings.
The team reviewed the Temporary Alteration Control Form procedure, AI-9 and TACFs
'
85-2009, 84-0115, 84-0107, 84-0081, 82-0242, 82-0214, 81-0458, and 80-0625. The
team noted that the TACF program lacks tracking of open TACFs to ensure followup
actionisinitiatedasrequiredbyprocedures,lacksengineeringevaluationbyOE
l of the temporary design, and lacks independent verification of as configured "
.
changes. The team also found that TVA uses TACFs for permanent changes (Deficiency
D5.3-1).
The team reviewed the sizing calculation which was performed to purchase the
station batteries. The calculation was never revised, although the plant systems-
have undergone many changes resulting in dc loading and duty cycle changes for - I
the station batteries. The team noted that the methodology used for this cal-'
culation did not consider the correction factors for ambient temperature and :
aging. The team concluded that the adequacy of the existing capacity of the
i battery system to support safety system operation during a total loss of ac power,
l as described in FSAR Section 8, has not been established.
+
.
23
1
_ _ _ . - . _ , ., . -, ,, -
-_- - - - - -_ ,- ,,.
_ _
. .
The team also found that sizing calculations for the battery chargers and for
the 120 v vital ac inverters do not exist, or if they were performed, could not
be located during the inspection (Unresolved Item US.3-2).
The team reviewed the control circuits of motor operated valves which must
function following an accident. At Sequoyah, the thermal overload trip function
for these valves has not been bypassed by the ESF signal. The trip settings
for the motor operated valves were set in accordance with procedure #SNP-INSTR
- 17, which allows the adjustment of the trip settings to be between 16 to 30
seconds for locked rotor current. The team found that some ESF valves may take
up to 60 seconds to complete their travel during accident conditions when the
- ystem voltage is degraded. In such a situation, the motor will experience a
running current for an extended time. Although this current may be less than
that for normal voltage operation, the time duration is long and therefore the
team is concerned that this current / time characteristic may be enough to operate
the thermal overload trip and defeat the safety function of the valve. TVA did
not perform an analysis on a case by case basis to verify that a spurious thermal
overload trip during valve travel will not prevent the valves from completing
their intended safety function (Unresolved Item U5.3-3).
TVA issued ECN-L5842 to replace AFW pressure control valves PCV-3122 and PCV-3-
132 with cavitating venturies. This revised system configuration permits a higher
flow resulting in an increased horsepower requirement (from 486 to 540 hp) for
the AFW pump motor. The team reviewed the emergency diesel generator load
analysis for the blackout and safety injection mode, and noted the following
(Unresolved Item U5.3-4):
The loading analysis was carried out using a revised load of 540 hp (for-
the AFW pump) lumped on the 25 second step of the load sequencer. In
reality, the load gradually increases from 486 hp to 540 hp in seven seconds.
The team is concerned because the seven second ramp overlaps the 30 second
step, which was identified by TVA as a critical step in the calculation. !
The loading sequence for the 25 and 30 second steps had not been analyzed j
to verify that the drop and recovery of the output voltage and frequency are <
within design limits. l
The team also noted several incorrect assumptions for the diesel generator
loading analysis. The analysis assumes all the pressurizer heaters connected
to the IE bus will be turned off; however, the loading table shows that the
pressurizer heaters remain connected to the bus. Another assumption of the
analysis states that the transformer name plate rating was used in the
sequencing table; however, only the transformer's connected loads were used.
The total transformer load which was analyzed to remain connected to the
6.9 kv bus consists of two 1500 kva transformers and one 3G0 kva transformer.
'
However, the team found that three 1500 kva transformers and one 300 kva
transformer remain connected. TVA did not consider the effect of the
magnetizing in-rush current of the third 1500 kva transformer for the voltage
and the frequency analysis of the sequencer zero block loading.
The team also noted that the calculated loading of the diesel generator
is 4184.9 kw but the rated capacity of the diesel generator is 4000 kw.
TVA indicated that although the rated capacity of the diesel generator is
lower than the imposed demand, an overload capacity of 10% (400 kw) will
24
, .
be available for two hours. TVA calculations indicate that the load will
be within the diesel's capacity within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
The team reviewed the control circuit drawings for the 6.9 kv switchgear and
circuit breaker for the AFW pump motor feeder breaker. The team found that the
breaker control circuit does not have a provision to detect and annunciate the
loss of control power. Such a loss could be caused by routine maintenance,
blown fuses, or an open circuit, making the motor driven AFW train inoperable.
The team acknowledged the licensee's statements regarding indication of control
power availability via the breaker's status lights; however, it is not clear
that the commitment to Regulatory Guide 1.47, which requires system level indi-
cation of bypassed or inoperable status, is satisfied without a more specific
annunciation (Unresolved Item US.3-5).
The team examined 125V dc voltage drop calculations to verify that the input
terminal voltages at the breaker control circuits for the motor driven AFW pump,
the AFW steam throttle valve and the vent fan for the steam driven AFW pump are
adequate. The team noted that these calculations have some errors and have un-
verified assumptions. The calculation for the vent fan and the throttle valve
voltage drop was a rough draft which was provided to the team at the end of the
inspection (Observation 05.3-6).
6.0 INSTRUMENTATION AND CONTROL
The team used a three step process for its review of particular instrumentation
and control design modifications at Sequoyah. First, the team examined the
recent review performed for Tennessee Valley Authority by Gilbert / Commonwealth
of completed design modifications for the main and auxiliary feedwater system.
Each of the eighteen instrumentation and control technical issues and observa-
tions identified by Gilbert / Commonwealth was reviewed by the team. Next, an
examination was made of an internal Tennessee Valley ,'uthority review for the
potential safety impact of unimplemented and partially implemented engineering
change notices in the essential raw cooling water, reactor coolant, and chemical
volume and control systems. A majority of the instrumentation and control
engineering change notices included in the Tennessee Valley Authority review
were examined by the team during this evaluation. In the third segment, the
team selected other instrumentation and control design modifications to provide
an independent evaluation of the effectiveness of the Tennessee Valley Authority
design control process. In this latter review, the technical approach chosen
to address a number of specific design issues was examined by the team.
6.1 OVERVIEW 0F THE GILBERT / COMMONWEALTH TECHNICAL REVIEW OF MAIN FEEDWATER/
Approximately half of the nineteen technical issues and eighteen observations
developed by Gilbert / Commonwealth involved the instrumentation and control
discipline. The team reviewed the technical content of each of these topics
with both Gilbert / Commonwealth and Tennessee Valley Authority personnel, and
made an assessment of the proposed Tennessee Valley Authority actions to
resolve the Technical Issue concerns.
25
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--
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9 4-
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w
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.
,-
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- - - - - - _ -
. - -- . -.- _ _ - _ _ _ - - _
>
l .
m y
'
i
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-
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.,
l The team also found that sizing calculations for the battery chargers and for
- the 120 v vital ac inverters do not exist, or if they were' performed, could not-
be located during the inspection (Unresolved Item U5.3-2). .
The team reviewed the control circuits of motor operated valves which must '
function following an accident. ~ At Sequoyah, the thermal overload trip function
- for these valves has not been bypassed by the ESF signal. fThe trip settings
for the motor operated valves were set in accordance with procedure #SNP-INSTR
,
- 17, which allows the adjustment of the trip settings to be between 16 to 30
'
s6conds for locked rotor current. The team found that some ESF. valves may take
up to 60 seconds to complete;their travel during accident conditions when the
system voltage is degraded. -In such a, situation, the motor will experience a
running current for an extended time. Although this curt)nt may be less than
i that for normal voltage operation, the time duration isJong and therefore the
i team is concerned that this current / time characteristic may be enough to operate
the thermal overload trip and defeat the safety function of the valve. TVA did
not perform an analysis on a case by case basis to verify that a spurious thermal
overload trip during valve travel will not prevent the valves from completing
their intended safety function (Unresolved Item US.3-3).
i TVA issued ECN-L5842 to replace AFW pressure control valves PCV-3122 and PCV-3-
j 132 with cavitating venturies. This revised system configuration permits a higher
!
flow resulting in an increased horsepower requirement (from 486 to 540 hp) for
"; the AFW pump motor. The team reviewed the emergency diesel generator load
analysis for the blackout and safety. injection mode, and noted the follo.*ing
(Unresolved Item U5.3-4): '4
i The loading analysis was carried out using a revised load of 540 hp (for
the AFW pump) lumped on the 25 second step of the load sequencer. In
i reality, the load gradually increases from 486 bp to 540 hp.in.seven seconds.
- The team is concerned because the seven second ramp overlaps the 30 second
step, which was identified by TVA as a critical step in the calculation.
The loading sequence for the 25 and 30 second steps had not been analyzed
to verify that the drop and recovery of the output voltage and frequency are r
,
within design limits.
a The team also noted several incorrect assumptions for the diesel generator
loading analysis. The analysis assumes all the pressurizer heaters connected
tn the IE bus will be turned off; however, the loading table shows that the
- essurizer heaters remain connected to the ~ bas. Another assumption of the
analysis states that the transformer name plate rating was used in the
, sequencing table; however, only the transformer's connected loads were used.
i
The total transformer load which was analyzed to remain connected to the
6.9 kv bus consists of two 1500 kva transformers and one 300 kva transfoiser.
,
However, the team found that three 1500 kva transformers and one 300 kva
j
- ?
transformer remain connected. TVA did not consider the effect of the
1 _ - magnetizing in-rush current of the^ third 1500 kva transformer for the voltage
{ and the frequency, analysis of.the sequencer zero block loading.
3
'
.
7 %
The team also noted that the calculated loading of the diesel generator
4
is 4184.9 kw but the rated capacity of the diesel generator is 4000 kw.
TVA indicated that although-the rated capacity of the diesel generator is
lower than the imposed demand, an overload capacity of 10% (400 kw) will
-
24 -
-
,
.=
'
e- -
!
. . _ . _ . - . . _ . - . - . - ~ ~ .-- - -- - - . . - . - - - . - .-.-_...,.: - - - - , , - - - , ,u-.
. ._ .- - - . .-. .-. -. - .
.. .
.
.
4
be available for two hours. TVA' calculations indicate tha't the load will
j be within the diesel's capacity within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
The team reviewed the control circuit drawings for the 6.9 kv.switchgear end
circuit breaker for the AFW pump motor feeder breaker. The team found that the
breaker control circuit does not have a provision to detect and annunciate the
'.
< loss of control power. Such a loss could be caused by routine maintenance,
blown fuses, or an open circuit, making the motor driven AFW train inoperable. ,
The team acknowledged the licensee's statements regarding indication of control-
power availability via the' breaker's status lights; however, it is not clear
that the commitment to Regulatory Guide 1.47, which requires system level indi-
cation of bypassed or inoperable status, is satisfied without a more specific
annunciation (Unresolved Item US.3-5).
The team examined 125V dc voltage drop calculations to verify that the input
terminal voltages at the breaker control circuits for the motor driven AFW pump,
the AFW steam throttle valve and the vent fan for the steam driven AFW pump are.
4
adequate. The team noted that these calculations have some errors and have un-
l verified assumptions. The calculation for the vent fan and the throttle valve
'
voltage drop was a rough draft which was provided to the team at the end of the
inspection (Observation 05.3-6).
t
6.0 INSTRUMENTATION AND CONTROL
The team used a three step process for its review of particular instrumentation
,
'
and control design modifications at Sequoyah. First, the team examined the
recent review performed for Tennessee Valley Authority by Gilbert / Commonwealth
l of completed design modifications for the main and auxiliary feedwater system.
} Each of the eighteen instrumentation and control technical issues and observa-
tions identified by Gilbert / Commonwealth was reviewed by the team. Next,.an
examination was made of an internal Tennessee Valley Authority review for the
potential safety impact of unimplemented and partially implemented engineering
4
change notices in the essential raw cooling water, reactor coolant, and chemical
volume and control systems. A majority of the instrumentation and control
- engineering change notices included in the Tennessee Valley Authority review
>
were examined by the team during this evaluation. In the third segment, the
team selected other instrumentation and control design modifications to provide
an independent evaluation of the effectiveness of the Tennessee Valley Authority
.
!
- design control process. In this latter review, the technical approach chosen j
to address a number of specific design issues was examined by the team.
i
6.1 OVERVIEW OF THE GILBERT /CDMMONWEALTH TECHNICAL-REVIEW 0F MAIN FEEDWATER/ 1
,
i Approximately half of the nineteen technical issues and eighteen observations
developed by Gilbert / Commonwealth involved the instrumentation and control
discipline. The team reviewed the technical content of each of these topics
i
with both Gilbert / Commonwealth and Tennessee Valley Authority personnel, and
made an assessment of the proposed Tennessee Valley Authority actions to-
resolve the Technical Issue concerns.
L
a
25
l
h
l
i
'
- -.._ ,__- --- ,- . - , ~ . . . , - . _ , . _ , . . . . . _ . , _ , , . . . - .u . _. -- , - _ _
-- _ . - . . . -- -- -. ._ .
-.- . . .- . .-
-n-
.. .
!
,
l
To determine the dep+h of Gilbert / Commonwealth's review, the team selected five
4 of these topics 'or on in-depth inquiry regarding. specific details of the
design modifications.
,
G/C technical issue 4 described inadequacies in' environmental and seismic qualifi-
cation of the auxiliary feedwater control valve mechanical exhauster added to
!
achieve valve operating times less than 60 seconds. The team reviewed post--
modification test data which indicated control valve response times between 12.1
.
and 12.6 seconds, and noted that subsequent periodic testing for response time *
{.
characteristics has been required at 92 day intervals by TVA surveillance
! procedures. The team concurs with the TVA commitment to review seismic and-
environmental qualification data for safety-related non-electrical devices, and .
.
agrees with the TVA plan to either qualify or eliminate the mechanical exhauster
l
from the design.
l
G/C technical issue 5 described auxiliary feedwater pump discharge pressure switch '
ratings that were not consistent with s'ystem operating pressures. The team
j extended this review by examining the original instrument data sheet, an
j interim replacement data sheet, and a subsequent replacement data sheet. The
i
tean roted that Tennessee Valley Authority did not use the previous design
}
basis or engineering work for these pressure switch replacements. Specifically,
- design basis changes for proof pressure, maximum operating pressure, contact
'
operating voltage, span adjustment range, contact operation, and setpoint values
did not have a documented basis (Deficiency D6.1-1). Actions proposed by Tennessee
Valley Authority to change the operating pressure to-1315 psig, overrange pressure ,
to 1500 psig, and proof pressure to 2500 psig are considered acceptable. The team
is not convinced that this is an isolated situation, as stated by Tennessee Valley
Authority, because of the absence of design basis control observed for these
instrument replacements.
G/C technical issue 7 described documentation inconsistencies.in safety classi-- i
! fication and an inappropriate seismic mounting arrangement of replacement solenoid
valves used to control the main feedwater bypass control valves. The team learned
that an interim nonsafety-related solenoid valve replacement had also been made
] without an engineering change notice authorizing the change. .The team explored
a
!
the feedwater isolation safety function required of these solenoid valves with
- Tennessee Valley Authority personnel, and determined that the-use of non-Class
IE replacement solenoid valves invalidated the unreviewed safety question deter-
l mination. Engineering change notice ECN 5717 stated that quality assurance was
required, but did not require seismic analysis. The unreviewed safety question
determination stated that a Class 1E solenoid was provided; however, this
,
requirement was not satisfied. Reasons presented by Tennessee Valley Authority.
for using a non-Class IE solenoid were based on the fact that the turbine
building is not a seismic category I structure,'the avoidance of Class IE cables
.in this location, and the fail-safe characteristics of the solenoid. - However -
this analysis failed to address the need to satisfy the functional requirement
- to isolate the main feedwater bypass control = valves (Deficiency D6.1-2). The
proposed Tennessee Valley Authority commitment to provide a Class IE solenoid
valve, not qualified for harsh environmental conditions'or seismic events,but;
.
mounted to limit. seismic responses, is considered satisfactory.
G/C technical issue 13 identified that response time' acceptance criteria had not
, been specified for the main feedwater bypass control valves and associated
components. The proposed commitment to document response time requirements forJ
! the main feedwater bypass valves, feedwater control valves,'and feedwater isolation
l
!
I
26
- .: . .- ... .-.. .-- - . - . .. -
_ , .. :, _ _. -. .' , , , ,
e 4 -- Aua ve,. ~ e .+w--m. . -
,
. .
,
- '
j' valves in an FSAR change, to provide-appropriate interva. to operations
f personnel for surveillance testing, and to reiterate to the lineers the need -
for specifying acceptance test criteria in ECNs, is considereu satisfactory,
f
G/C technical issue 15 identified that setpoint.and time delay values for
- auxiliary feedwater pump suction pressure switches were,not adequately documented.
- These pressure switches initiate automatic transfer of pump suction to the safety-
4
related essential raw cooling water system. The team noted that numeric changes
made in input values for a 1979 setpoint calculation had not been carried through-
.
to-calculational results, and that this calculation had not'been identified for -
update when setpoint changes were made in 1981 and 1984 (Deficiency D6.1-3). The
proposed actions to update the design input calculation and prepare safety-related
setpoint calculations are considered satisfactory.
.
.
The team concluded that the Gilbert / Commonwealth instrumentation and control
technical review was comprehensive and thorough within the' selected review
i scope. While examination of the maintenance of continuity in the plant's' design
basis as design modifications were processed was not part of the Gilbert / Common-
wealth review, the team has independently examined original and interim design
, basis documents for certain modifications. . Specific design basis inconsistencies '
identified by the team suggest that a review of this area.would be appropriate.
6.2 ASSESSMENT OF THE REVIEW OF UNIMPLEMENTED AND PARTIALLY IMPLEMENTED ECNs
Approximately forty implemented or partially implemented engineering change
notices included in the Tennessee Valley Authority three system review were
'
examined by the team for their instrumentation and control aspects. During this
review, the team placed its major emphasis on those engineering change notices i
involving the essential raw cooling water and reactor coolant systems.
The TVA review identified two engineering change notices that would have resulted !
in an unreviewed safety question based on planned restart of Sequoyah prior to ,
i
their implementation. The team reviewed the Tennessee Valley Authority evaluation
'
for four I&C related engineering change notices which would have resulted in a
USQ if these were not implemented prior to restart, and agreed with the USQD in~ -
each instance. The reviewed engineering change notices were: '
- ECN-L-6527, which would disconnect certain essential raw l
I
l cooling water, reactor coolant, saf ety injection, ;
and residual heat removal system local handswitches '
I
from their respective electrical circuits since the ;
) switches are not qualified for harsh environment- ;
{ conditions (replacement planned). ~i
!
ECN-L-5773, which would replace unqualified air-operated relief .!
valves (PORVS) with environmentally qualified j
solenoid-operated valves (replacement planned). '
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ECN-L-6439, which would relocate pressarizer level and
. . .
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i pressure transmitters outside the crane wall. -
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Functional testing and calibration following .!
these modifications (already planned) were '
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determined to be needed to preclude an USQ. l
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ECN-L-6495, which involved the replacement of valves and
relocation of the condensate pot for pressurizer
level transmitters. An unreviewed safety question
would have existed if the transmitters weren't
calibrated. Such calibration was planned prior
to restart.
In the three system review program, Tennessee Valley Authority attempted to
determine whether implemented and partially implemented engineering change
notices would have an adverse impact on existing unreviewed safety question
determinations or on subsequent safe operation of the plant. The team was
informed that the purpose of the three system review was not to assess the
technical adequacy of design modification changes described in individual
engineering change notices, but to determine if the status of the ECN involved
an USQ. During the inspection, the team identified the following instance
where the proposed technical solution was not adequate.
A Nuclear Engineering Branch (NEB) analysis for ECN-L-6449 incorrectly author-
ized the downgrading of reactor coolant system resistance temperature detectors
used as inputs to the reactor protection system from environmental qualification
category A to C (Deficiency D6.2-1).
The engineering change notice required complete qualification of these detectors,
but this aspect would have been rescinded by acceptance of a previous NEB analysis.
During the inspection, Tennessee Valley Authority initiated corrective action
through a significant condition report to void the analysis and restore the
environmental qualification to category A.
6.3 ADDITIONAL REVIEW BY THE NRC
The team examined portions of the Sequoyah design modification program beyond
the Gilbert / Commonwealth main and auxiliary feedwater system and the Tennessee
Valley Authority three system reviews.
Tennessee Valley Authority has not routinely specified the performance of a
hydrostatic pressure test following seismic qualification tests for instruments
connected to the reactor coolant or safety system pressure boundary. Adequate
guidance has been provided in TVA procedures regarding this requirement. The
team reviewed two instrument procurements where one vendor had successfully
demonstrated acceptable test results and one vendor had not. The latter
instruments were connected directly to the reactor coolant pressure boundary
(Deficiency 06.3-1).
The team noted that approximately 20 percent of the 80 ECNs reviewed during
the inspection were inconsistent in the specification of quality assurance and
seismic requirements for safety-related design modifications. Seven ECNs speci-
fied Quality Assurance without requiring seismic analysis; the team believes
this is incorrect under TVA procedures. Eight other ECNs had the initial deter-
mination changed to a correct designation by a reviewer during the approval
process. The team reviewed applicable procedures for these determinations, and
believes that more clarity is needed to achieve consistency in this process
(Unresolved Item U6.3-2).
28
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Removal of wiring for a differential pressure ~ switch used for the essential raw
cooling water travelling screen was accomplished by Temporary Alteration Control
Form 82-0258 in late 1982. The temporary alteration justified deletion of this
safety-related function based upon the availability of an automatic cycling
4
timer. Installation of a bubbler type instrument, which will restore safety-
related level. indication across the screens as an initiator of the screen wash
system, is under development by ECN-L-5758, initiated in 1982. The team con-
siders this an excessive period of time to restore the safety function
(Observation 06.3-3).
During the inspection, the team experienced difficulty with retrieval of instru-
mentation and control design basis data and Sequoyah licensing positions relative
to industry standards and regulatory guides. The team concludes that this diffi-
culty of retrieval of key information can affect the maintenance of consistency
throughout the design process.
Tennessee Valley Authority design criteria for electrical separation of equipment, ,
,
cables and instrument tubing appeared to be comprehensive and technically sound.
!
'
The team also reviewed examples of completed work plans used to implement
engineering change notices, post-modification test results, and surveillance test :
procedures. The team had no further questions in these areas.
3
7.0 BACKGROUND
7.1 MEETINGS
,
! Inspection activities were conducted at G/C Offices in Reading, PA, OE Offices
in Knoxville, Tennessee, and at OE Offices at the Sequoyah site. Entrance and
exit meetings were held to discuss the inspection plans and preliminary findings,
respectively. A final exit meeting was held at the Sequoyah site on March 14, 1986.
The following describes the general purpose of the various meetings. Table 7.1 is
provided as a matrix of meeting attendance.
l Meeting #1: This meeting was held at the officer of Gilbert / Commonwealth at
Reading, PA, on February 12, 1986. The NRC inspection team explained the plans ,
for the assessment of the efforts of Gilbert / Commonwealth for the review of the '
historical modifications to the AFW system. G/C presented the methodology, scope, l
j and status of the review. TVA presented historical and current aspects of the :
engineering design change process, including organization and procedure '
changes.
Meeting #2: This meeting was held at the offices of Gilbert / Commonwealth on
,
February 14, 1986. The NRC team explained the preliminary findings of the review
i
at Gilbert / Commonwealth. A tentative schedule for inspection activities at
Knoxville and'the Sequoyah site was presented.
Meeting #3: This meeting was held at the Office of Engineering in Knoxville,
4
Tennessee on February 18, 1986. TVA presented organization charts and changes-
in the engineering and administrative procedures. TVA also discussed the status,
,
preliminary findings and methodology of the "three system review", which was ' .!
being conducted by the Office of Engineering.
Meeting #4: This meeting was held at the Office.of Engineering offices at the
Sequoyah site on February 21, 1986. The team discussed preliminary findings-to l'
date. Plans for. a one week inspection 'of TVA's design change process were also
discussed.
f 29 I
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_ - ._. _ __
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Meeting #5: On March 3, 1986, a meeting was held with TVA personnel to discuss
the preliminary findings of the inspection.
Meeting #6: On March 7, 1986, a meeting was_ held with TVA personnel to discuss
the preliminary findings of the inspection.
Meeting #7: On March 14,-1986, a final exit meeting was held at the Sequoyah
si te .- NRC management discussed the scope and findings of.the inspection.
'
Each team member presented the more significant findings within his discipline.
,
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TABLE 7.1 MEETINGS
Name Organization Title Meeting Attended
1 2 3 4 5 6 7
BKGrimes USNRC-IE Director,DQAVT/01E x
HJMIller USNRC-IE Dep. Director DQAVT/0IE x x
EVImbro USNRC-IE Section Chief QAB/0IE x x x x-x
REArchitzel USNRC-IE Team Leader x x x x x x x
RParkhill USNRC-IE NRC-Mechanical Systems x x x x
SVAthavale USNRC-IE NRC-Electric Power x x x x x x
JJBlake USNRC-RII Section Chief x
SPWeise USNRC-RII Section Chief x
KMJenison USNRC-RII Sr. Resident Inspector x x
WCLiu USNRC-RII Reactor Inspector x
ADuBouchet NRC-Consultant NRC-Mechanical Components x x x x x x
FJMollerus NRC-Consultant NRC-Mechanical Systems x x x x x x
AIUnsal NRC-Consultant NRC-Civil / Structural x x x x x x
LStanley NRC-Consultant NRC-Instrument / Controls x x x x x x
HLAbercrombie TVA Site Director x
PRWallace TVA Plant Manager x x
JPVineyard TVA SQEP Project Manager x x x x
HBRankin TVA SQN DES. SVCS Manager x x x x x x
JDHutson TVA Elec. Issues Manager x x x x
RW01 son TVA Modification Manager x x
WEAndrews TVA Site quality Manager x
WCDrotleff TVA/S&W Manager of Engineering x
LMNobles TVA Supdt.. (0&E) x
CNJohnson TVA Civil Project Manager x x x
JCKey TVA Mech. Project Engineer x x x x
GTHall TVA Elec. Project Engineer x x x
APCapozzi TVA/S&W Ass. Chief. Engineer x
MRSealacik TVA Elec. Mod. Supv. x x
GBKirk TVA Compliance Supv. x x
JEstaub TVA I&C Supv. x x x x x x x
RDaniels TVA Mechanical Supv. x
LAlexzender TVA Mech. Mod. Supv. x
RMMooney TVA Systems Engr. Supv. x
JRochelle TVA Comp. Analysis Engineer x
DLCowart TVA Quality Surveillance Supv. x
FEDenny TVA Quality Assurance Engineer x x x
RCBirchell TVA Mech. Engineer x x
VABianco TVA Nuclear Engineer x x x x x
MRCooper TVA Compliance Engineer x
MACooper TVA Mech. Engineer- -x x
WBWest TVA Engineer x
31
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TABLE 7.1 MEETINGS (cont.)
Name Organization Title Meeting Attended
1 2 3 4 5 6 7
CConstantine TVA Engineer x
GRBell TVA Elec. Engineer x
DHatcher TVA Civil Engineer x
EWhitaker TVA Licensing Engineer x
R0'Donnel TVA Staff Engineer x x
JDSmith- TVA x x
WJLeininger G/C Team Leader x x x
RCronk G/C Electrical lead x x x
MAkins G/C Mech. lead x x v
CWhitehead G/C Piping / structural lead x x a
HAManning G/C Q/A Lead x x
CPaschall G/C Program Lead x x
GSanders G/C x x x
JFTortora G/C I&C Engineer x x x
RKGramling G/C Piping Engineer x x
7. 2 PERSONS CONTACTEC - MECHANICAL SYSTEMS
Name Title Organization
E. L. Booker Mechanical Maintenance TVA-ONP
5. S. Long Systems Engineering TVA-ONP
D. Hafley Project Engineer TVA-0NP
M. Cooper Compliance Engineer TVA-ONP
D. Reed Work Plan Files TVA-ONP
R. W. Olson Modification Branch Manager TVA-0E
R. E. Daniels Mechanical Section Supervisor TVA-0E
D. A. Burch Mechanical Engineer TVA-0E
J. C. Key Mechanical Project Engineer TVA-0E
H. B. Rankin Design Services Manger TVA-ONP
,
32
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7.3 PERSONS CONTACTED - MECHANICAL COMPONENTS
'
Name Title Organization
K. Mogg Section Supervisor, SQEP-CEB TVA
D. Hatcher Civil Engineer, SQEP-CEB TVA
S. Yeh Mechanical Engineer, SQEP-CEB TVA
R. Gish Mechani al Engineer, SQEP-CEB TVA
J. Rochelle Senior Mechanical Engineer, CEB TVA
C. Winton Mechanical Engineer-Modification, SQEP TVA
R. Field Civil Engineer-0E, SQEP TVA
S. Jackson Mechanical Engineer-0E, SQEP-MEB TVA
T J. Means Mechanical Associate-0E, SQEP-CEB TVA
T.B. Bucy Civil Section #1 Supervisor-0E, SQEP-CEB TVA
R.E. Daniels Mech Section #2 Supervisor-0E, SQEP-MEB TVA
D.A. Burch Mechanical Engineer, SQEP-MEB TVA
L.L. Edmondson Electrical #1-0E, SQEP-EEB TVA
L.M. Chacon Civil Engineer, SQEP-CEB TVA
J.E. Staub I&C Section Supervisor-0E, SQEP-EEB TVA
R.L. Kelly Systems Engineer-NUC PR, SQN-SE TVA
A. Smith I&C Engineer-0E, SQEP-EEB TVA
,
'
D. Widner Mechanical Engineer-Modifications TVA
R.W. Olson Modification Manager, SQN TVA
J.A. Southers Civil Engineer-0E, SQEP-C2 TVA
j J.L. Hamilton QE/QC Supervisor-QA Staff, SQN-QA TVA
J.E. Maddox OE-Project Manager Staff, SQEP TVA
7.4 PERSONS CONTACTED -
CIVIL / STRUCTURAL
'
Name Title Organization
,
'
C.W. Whitehead Piping Supervisor Gilbert
G.B. Sanders Civil Supervisor Gilbert
A.M.Patel Civil Engineer Gilbert
M. Cones Technical Supervisor TVA OE
T.B.Bucy Supervising Civil Engineer TVA DE
5.J.Patel Civil Engineer TVA OE
A.Rather Civil Engineer TVA OE <
L.M.Chacon Civil Engineer TVA OE !
C.Constantine Mechanical Engineer TVA OE
S. Taylor {
N. Black Supervising Electrical eng TVA OE
J.Rochelle Section Supervisor, CEB TVA OE
)
'
R.Gish Mechanical Engineer TVA DE !
C. Johnson Civil Project Engineer TVA DE
l
7.5 PERSONS CONTACTED - ELECTRICAL
!
Name Title Organization
G.T. Hall Electrical Project Engineer OE/DETS/EEB/TVA
J.D. Hutson Electrical Issues Manager OE/DETS/EEB/TVA
N. Black Section Supervisor OE/DETS/EEB/TVA
s
!
R.P. Cronk Section Manager Electrical Gilbert / Commonwealth l
33
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7.6 PERSONS CONTACTED I&C
Name Title Organization
A.P. Smith I&C Project Engineer Gilbert / Commonwealth
J.F. Tortora I&C Engineer Gilbert / Commonwealth
H.A. Manning Mgr., Corp. QA Programs Gilbert / Commonwealth
C.C. Pascha11 Mgr. , Quality Management Gilbert / Commonwealth
B.K. Williams Nuclear Engineer, NEB TVA, Knoxville
M. Belew Principal Elec. Engineer TVA, Knoxville
A.F. Pagano Head, I&C Engineering TVA, Knoxville
J. Hudson Mgr., SQN Electrical Issues TVA, Knoxville
J.E. Staub Supervisor, Elec. Section TVA, Sequoyah
G.R. Bell Electrical Engineer TVA, Sequoyah
L.M. Begley Electrical Engineer TVA, Sequoyah
B.H. Brown Supervisor, Mech. Section TVA, Sequoyah
B. Buchanan Electrical Engineer TVA, Sequoyah
L.M. Chacon Civil Engineer TVA, Sequoyah
L.L. Edmondson Engineering Associate TVA, Sequoyah
R.D. Gish Civil Engineer TVA, Sequoyah
J.W. Hatcher Electrical Engineer TVA, Sequoyah
R.C. Jenkins Supervisor, Procurement Sect. TVA, Sequoyah
J.L. Wood Procurement Section TVA, Sequoyah
A.L. Smith Electrical Engineer TVA, Sequoyah
K.R. Spino Engineering Associate TVA, Sequoyah
J.W. Webb Electrical Engineer TVA, Sequoyah
V.A. Bianco Nuclear Engineer TVA, Sequoyah
D.A. Burch Mechanical Engineer TVA, Sequoyah
C. Constantine,Jr. Mechanical Engineer TVA, Sequoyah
E.H. Turner Electrical Engineer TVA, Sequoyah
L. Nobles TVA, Sequoyah
F. Denny Electrical QA Engineer TVA, Sequoyah
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4
LIST OF DEFICIENCIES, UNRESOLVED ITEMS, AND OBSERVATIONS
,
Item Title
D2.1-1 Deficiency Furmanite Leaking Va'lve Bonnet
02.3-1 Deficiency Long Term Unincorporated ECNs
D3.1-1 Deficiency Exhauster Installation
03.2-1 Observation Valve Operator
D3.2-2 Deficiency USQD Requirement ;
D3.2-3 Deficiency Piping Flow Diagram .
D3.2-4 Deficiency Sample Connection Support
03.3-1 Deficiency Pipe Support Friction Design
D3.3-2 Deficiency Valve Accelerations
D3.3-3 Deficiency Valve Fundamental Frequency
D3.3-4 Deficiency Alternate Pipe Support Criteria
D3.3-5 Deficiency Pump Fundamental Frequency
D4.3-1 Deficiency Evaluation of Structures for Reinforcing Bar Cuts
4.3-2 N/A Number not used
D4.3-3 Deficiency Steam Generator Access Platform Design
D4.3-4 Deficiency Cable Tray Support Response Spectra
D4.3-5 Deficiency Loads on Cable Tray Supports
'
D4.3-6 Deficiency Torsional Shear Stress Effects on Weld Design
U4.3-7 Unresolved Item Cable Tray Support Base Plate Analysis
04.3-8 Observation Base Plate Design Criteria
04.3-9 Observation Design Criteria for Tanks
D5.3-1 Deficiency Temporary Alterations using TACFs
US.3-2 Unresolved Item Sizing Calculations
US.3-3 Unresolved Item Motor Operated Valves Thermal Overload Trip Setting
US.3-4 Unresolved Item Diesel Generator Loading Calculations
A-1
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Item Title
.U5.3-5 Unresolved Item Loss of Control Power Annunciation.
.
05.3-6 Observation Voltage Drop Calculations
- D6.1-1 Deficiency AFW Pump Discharge Pressure Switch Ratings
'
D6.1-2 Deficiency Feedwater Bypass Control Valve Solenoid Replacement
4
D6.1-3 Deficiency AFW Pump Suction Pressure Switch Setpoint Calculation
D6.2-1 Deficiency Reactor Coolant System Narrow Range Resistance Tem-
.
perature Detector Qualification Category Change.
'
D6.3-1 Deficiency Specification of Hydrostatic Test.to Demonstrate
Instrument Pressure Boundary Integrity after Seismic
Qualification Testing.
,
U6.3-2 Unresolved Item Engineering Change Notice Quality Assurance and
Seismic Analysis Designations.
06.3-3 Observation Essential Raw Cooling Water Screen Wash Pump Control
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D2.1-1 (Deficiency) Furmanite Leaking Valve Bonnet l
DESCRIPTION: ECN L6317 authorized the use of Furmanite to stop steam leakage
of check valve 1-VLV-3-891 in one of the steam supply lines to the Unit I
turbine driven auxiliary feedwater pump. Furmanite is a commercially available
material that is injected into the vicinity of the gasket space to stop
excessive leakage until the valve can be repaired or replaced in the next ,
outage. The "Furmaniting" procedure can be done without isolating the valve.
.
The Furmanite procedure, N-84533, requires drilling into the bonnet flange to
inject Furmanite into the space between the flanges outside of the gasket and
inside a dam formed by inserting wire into, or peening, the seam between the
flanges. In the case of valve 1-VLV-3-891, the injection of Furmanite was accom-
plished by drilling radially inward from the outside of the bonnet flange to a
point that is slightly outside the flange stud holes. The holes are then tapped
for a 3/8-inch adapter. Using the adapter as a guide, a 1/8-inch drill is used
to drill into the bolt clearance hole to form a path for injecting Furmanite.
The procedure requires drilling into points near the center of the bonnet
flange, to inject Furmanite into the clearance annuli between each closure stud
and its hole. Drilling and Furmanite injection was made into each closure stud
clearance. Drilling into studs is avoided through the skill of the mechanics
doing the work.
The safety concern is primarily for the pressure integrity of the closure after
performance of the Furmaniting procedure, namely:
The valve is an ASME III Class B component as classified in the work plan.
The drilling procedure involved metal removal from the bennet flange and
possibly from the closure studs. Stresses can increase due to metal removal
and stress concentration factors.
If Furmaniting is effective, there could be a shift in gasket loading and
resultant stresses in the components of the valve.
Neither the ECN, Unreviewed safety Question Determination or the work plan address
the need for a stress analysis or otherwise dispose of this issue. Other ECNs
authorizing Furmaniting have included a stress analysis (References 4 & 5).
BASIS: TVA has committed to apply ASME code requirements to piping and
components used in safety-related systems (Reference 5) and ANSI-N45.2.11 (Reference
6) to the design process for plant modifications.
The ASME Code addresses stresses in Class 2 valves by providing rules for minimum
thickness (Reference 7) including the standard design rules of ANSI B16.34 (Ref-
erence 8). The Code does not appear to anticipate that valve closures may be
modified after installation in order to facilitate continued operation. Some
guidance is provided by ASME Section XI (Reference 9) that is applicable to in-
service performance of repairs and replacement of components. Article IWC-4000
of section XI provides rules for weld repair of cavities created by removal of
defects in order to restore the component to its original strength and config-
uration. Weld repair of the holes created by Furmaniting is impractical in most
situation. However, an engineering evaluation of stress increases should be
included for Furmaniting ECNs applied to components covered by the ASME Code
under Section III and XI.
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REFERENCES
1. ECN L6317, Aux Feedwater System 03, Furmanite and Peen the Bonnet on Valve
1-VLV-3-891, 10/14/84.
2. Furmanite Procedure N-84533,12/18/84
3. ECN L6157, 5/21/84.
4. ECN L6169, 6/4/84.
5. TVA Design Criteria No. SQN-DC-V-3.0, 12/12/75.
6. ANSI N45.2.11-1974, " Quality Assurance Requirements for the Design of
Nuclear Power Plants."
7. ASME Boiler and Pressure Vessel Code,Section III, Nuclear.
Power Plant Components, Subsection NC, Class 2 Components.
8. ANSI Standard Specification B16.34, Fittings, Flanges and
Valves,
e 9. ASME Boiler and Pressure Vessel Code,Section XI, Rules for
In-service Inspection of Nuclear Power Plant Components.
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D2.3-1 (Deficiency) long Term Unincorporated ECNs j
DESCRIPTION: ECN L 5320, dated 10/23/80, authorizes replacement of overcurrent
relays for the essential raw cooling water pump. motors with larger capacity
overcurrent relays. The original relays were provided for 600-hp essential raw
cooling water pumps.
A new essential raw cooling water pumping station was constructed using essential
raw cooling water pump motors rated at 700-hp. Preoperational testing of the
new pumps using the original relays in the 6.9 KV shutdown board identified
that the overcurrent relays are set too low. This resulted in relay chattering
and relay hang up. The relays had to be cleared manually. The Preoperational
Test Deficiency Report (Reference 2) identified that if the relay hung up for an
inactive pump that was needed for auto-start, the pump could not be started until
the relay was manually cleared.
ECN L5320 was issued in response to Deficiency Report PT-566 and Design Change
Request SQ-DCR-741. The latter is a generic DCR requesting issuance of ECNs
based on Pre-op Test Deficiencies (pts)". The generic DCR was issued to allow
timely resolution of pts. The Unreviewed Safety Question Determination Form
accompanying ECN 5320 indicated that no Unreviewed Safety Question was intro-
duced by the change.
Team review of field documentation showed the following:
The Office of Engineering is carrying the status of ECN L 5320 as " complete."
This is interpreted as complete insofar as remaining work by the Office of
Engineering is concerned, but not necessarily field implementation. The
Office of Engineering changes the status of the ECN to closed upon written
notification by the Office of Nuclear Power that the field work and testing
is complete.
Work Plan WP 8937, associated with the ECN L5320 has been cancelled.
Temporary Alteration Control Form 80-734-67 was issued and approved for
emergency action on 10/4/80.
BASIS: Current TVA policy in effect since July 1985 requires that a Design
Change Request be issued within 60 days of a Temporary Alteration Control Form
.I if the temporary alteration is not returned to normal. The present condition
'
is a situation where a change has been made without the change process associated
- with ECN preparation and Work Plan development, including testing. ANSI N45.2.11
(Reference 4) requires a process of design analysis and verification (e.g., test)
which appear to be circumvented by the modifications implemented using only a
Temporary Alteration Control Form.
REFERENCES
1. ECN L5320, Replace Essential Raw Cooling Water Pump Overcurrent Relays,
- 10/23/80.
2. Preoperational Test Deficiency Report PT-566.
3. Temporary Alteration Control Form 80-734-67.
4. ANSI N45.2.11-1974, " Quality Assurance Requirements for the Design of
Nuclear Power Plants."
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D3.1-1 (Deficiency) Exhauster Installati]n
DESCRIPTION: TVA ECN L5267 added quick exhausters to the controls for eight
valves installed in Sequoyah Nuclear Plant. The USQD indicates that:
"Masoneilan is evaluating the installed configuration of the exhauster on the
control valve and will be sending documentation of the seismic qualification
later (with the possible but not expected addition of new support requirements
for air tubing)". However, the TVA Electrical Engineering Branch could not
obtain a seismic qualification document. The exhausters may therefore have
been installed without a valid USQD. TVA confirmed that a total of eight
exhausters have been installed to TVA in-house alternate support spacing criteria.
The team reviewed Detail J120 (Reference 1), which provides a schematic of the
quick exhauster and associated 1/2 in. tubing.. Tubing spans for two of the
eight installations (control valves 2-LCV-3-148 and -171) appear to exceed the
maximum allowable TVA support spacing for spans with concentrated weights
(Reference 2). The team noted that none of <the eight installed exhausters appear
to have been evaluated with respect to a governing Masoneilan seismic qualification
document. Moreover, two of' the eight exhausters do not even appear to be installed
in accordance with the TVA alternate support spacing criteria.
BASIS: Section 4.1.2.8 of TVA engineering procedure EN DES-EP 4.52 (Reference
3) requires, in part, that a USQD be prepared for each L-ECN or TVA-approved
'
design change against a nuclear power plant with an issued operating license
before physical work is authorized. Section 4.5 specifies that completion of
design work for an L-ECN involves, in part, a completeness review of the
associated USQD to be certified by memo. TVA engineering procedure EN DES-EP
2.03 (Reference 4), Section 3.4, notes that failure to adhere to requirements
specifically identified in the USQD evaluation nullifies the USQD evaluation.
Finally, two exhausters were not installed in accordance with the TVA span
allowables tabulated in Reference 2.
REFERENCES
1. TVA Drawing No. 47W600-120, Mechanical Instruments and Controls,
Revision 11, dated April 22, 1980.
2. TVA Table No. 47A053-15B, Mechanical Seismic Support Process Pipe 2 in.
,
Diameter & Less, Revision 0, dated September 15, 1977.
3. TVA Engineering Procedure EN DES-EP 4.52, Engineering Change Notices
(ECNs) After Licensing-Handling, Revision 1, dated April 24, 1984.
4. TVA Engineering Procedure EN DES-EP 2.03, Unreviewed Safety Question
Determination-Handling and Preparation, Revision 6, dated April 24, 1984.
,
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03.2-1- (Observation) Valve Operator
DESCRIPTION: TVA ECN L5017 added a motor operator and associated controls to
manual valve 0-67-546A installed in the essential raw water cooling system.
The valve summary sheet for the calculation package prepar+' for the piping
analysis (Reference 1) did not tabulate the weight of the valve operator or the
offset dimension of the valve operator center of gravity. However, the. piping
isometric and the computer analysis contained the correct data. The piping
input data form for recording all design basis input data used for analysis
considerations should be corrected.
REFERENCES
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1. TVA Piping Analysis N2-67-1A, Revision 4, dated January 24, 1986.
2. TVA Design Criteria No. SQN-DC-V-13.3, Detailed Analysis of Class I Piping
Systems, Pevision 3, dated August 13, 1984.
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03.2-2 (Deficiency) USQD Requirement
DESCRIPTION: TVA ECN L5500, in oart, adds extension operators and covers to
Units 1 and 2 valves 67-507A installed in the essential raw cooling water system.
The ECN calls for a seismic analysis to seismically re-qualify the existing
valves. .The USQD for the ECN specifies that: "a seismic analysis will be done
to show that the new valve stem extensions do not invalidate the existing seismic
qualifications for the valve, piping or associated components". However, the
TVA Civil Engineering Branch was not able to obtain new seismic qualification
documentation for the valves.
BASIS: Section 4.1.2.8 of TVA engineering procedure EN DES-EP 4.52 (Reference
1) requires, in part, that a USQD be prepared for each L-ECN or TVA-approved
design change against a nuclear power plant with an issued operating license
before physical work is authorized. Section 4.5 specifies that completion of
design work for an L-ECN involves, in part, a completeness review of the
associated USQD to be certified by memo. TVA engineering procedure EN DES-EP
2.03 (Reference 2), Section 3.4, notes that failure to adhere to requirements
specifically identified in the USQD evaluation nullifies the USQD evaluation.
REFERENCES
1. TVA Engineering Procedure EN DES-EP 4.52, Engineering Change Notices
(ECNs) After Licensing-Handling, Revision 1, dated April 24, 1984.
2. TVA Engineering Procedure EN DES-EP 2.03, Unreviewed Safety Question
Determination-Handling and Preparation, Revision 6, dated April 24, 1984.
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D3.2-3 (Deficiency) Piping Flow Diagram
DESCRIPTION: TVA ECN L5737 added a check valve to the primary water piping to
the CVCS spent resin header. The primary water piping is TVA Class G while the
CVCS piping is. pressure boundary piping qualified to TVA Class D. The check
valve was added to the flow diagram (Reference 1) but the change of class from
D to G is not shown on the revised flow diagram.
BASIS: TVA design criteria no. SQN-DC-V-3.0 (Reference 2), Section 3.5,
requires in part that piping drawings be clearly marked to indicate the TVA
pipe classification of all piping represented, and that all interface boundaries
of higher and lower class piping be pinpointed exactly.
REFERENCES
1. TVA SQN Flow Diagram No. 47W809-4, CVCS/ Chemical Control, Revision 7,
dated March 14, 1983.
2. TVA Design Criteria No. SON-DC-V-5.0, The Classification of Piping, Pumps,
Valves and-Vessels, Revision 1, dated June 28, 1985.
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D3.2-4 (Deficiency) Sample Connection Support
DESCRIPTION: TVA ECN L6462 authorized the installation of a support for a 3/4
inch sample connection coming off the downstream piping of the CVCS letdown heat
exchanger. The support, designed to seismic Ca'tegory I criteria, is being added
to reduce vibration in the 3/4 inch sampling line connected to the 3 inch process
line. The support was designed to a minimum frequency of 50 hertz. The details
for the clamped connection of the support to the 3 inch process line are derived
from a standard support detail sheet (Reference 1) and are attached to the ECN.
The support detail is based on TVA Civil Engineering Branch Report No. 77-42
(Reference 2), which originally specified two heavy Bergen-Patterson clamps with
four 7/8 inch diameter bolts torqued to 100 ft-lbs. However, the standard support
detail now specifies Basic Engineers heavy duty pipe clamp BE-122, which uses 3/4
inch. bolts instead of 7/8 inch bolts for the 3 inch clamps. In addition, split
washers or lock nuts were not specified for the installation in order to maintain
the bolt design torque under vibratory loads. The team notes that the support
detail specified in the TVA Civil Engineering Branch report is based on static
and not dynamic loads.
BASIS: The basis for this deficiency is the use of smaller diameter bolts than
called for in the standard support detail. In addition, the 100 ft-lbs
originally specified for the 7/8 inch bolts and specified in the ECN to be used
l for the 3/4 inch bolts may overstress the 3/4 inch bolts.
REFERENCES
1. TVA SQN Drawing No. 47A406-2-4, Mechanical-Unit 2/ Category Support for
Support Detail 2-4, Revision 1, dated August 2, 1985.
2. TVA Civil Engineering Branch Report No. 77-42, Static Pipe Support Tests
and Development-Sequoyah Nuclear Plants 1 & 2, dated October 25, 1977.
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D3.3-1 (Deficiency) Pipe Support Friction Design-
DESCRIPTION: Section 7.19 of the TVA pipe support design manual (Reference 1)
indicates that friction forces are not considered for pipe support design at
Sequoyah Nuclear Plant. The team noted that forces due to friction for thermal
displacements greater than 1/16 inch are generally taken into account when com-
puting pipe support reactions due to dead and thermal loads. Moreover, USAS
B31.1-1967, the piping code of record for TVA safety class B, C and D piping at
Sequoyah Nuclear Plant, requires consideration of frictional forces due to
piping thermal expansion.
A pipe support which cannot. accommodate piping thermal movement without
resistance is subject to friction force in addition to bearing force. The
magnitude of the applied friction force is about one-third of the piping dead
load and operating thermal force bearing:on the support. The direction of the
applied friction force is the same as the direction of the piping thermal move-
ment and is perpendicular to the direction of the pipe support bearing force.
Friction force is potentially significant for a pipe support which is weaker in
resisting friction force than in resisting bearing force. In order to cualify
a pipe support subject to friction and bearing force by analysis, the stresses
due to friction and bearing forde must be separately computed, and combined in
accordance with AISC code requirements.
BASIS: The basis for this deficiency is TVA's failure to analyze pipe supports
for friction forces due to thermal displacements, as required by piping code
B31.1. Section 120.2.3, Anchors or Guides, requires that: "where anchors or
guides are provided to restrain, direct, or absorb piping movements, their
design shall take into account the forces and moments at these elements caused
by internal pressure and thermal expansion". Section 121.2.1, Anchors and Guides,
paragraph (e), specifies taat: " Brackets shall be designed to withstand forces
and moments induced by sliding friction in addition to other loads."
REFERENCES
1. TVA SQN Pipe Support Design Manual, Vol. 3, Revision 0, dated April 2?,
1983.
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D3.3-2 (Deficiency) Valve Accelerations
DESCRIPTION: FSAR Section 3.9.2.5.2 stipulates that: " seismic valve accelerations
are maintained below 29 vertical, and 39 horizontal for the SSE condition".
However, the teams found that valves are being qualified on a case by case basis
to acceleration levels which exceed the FSAR comitment. As an example, the team
reviewed piping analysis N2-78-6A (Reference 1), which reanalyzes the 4 inch spent
fuel pit cooling line from containment penetration X-83 to a three-way rigid restraint.
The system was analyzed for static, normal thermal, design basis accident thermal
and pressure, design basis accident inertia, operating basis earthquake seismic, and
safe shutdown earthquake seismic conditions. The piping subsystem contains one 4
inch diaphragm valve and two 3/4 inch globe valves (Reference 2). The 4 inch
Grinnell valve was qualified to computed accelerations of 2.66g vertical and 7.679
horizontal; the 3/4 inch Hancock valve was qualified to computed accelerations of
1.199 vertical and 19.58g horizontal, and the 3/4 inch Kerotest valve was qualified
to computed accelerations of 2.929 vertical and 7.339 horizontal. TVA technical
personnel have indicated that Appendix F of the Sequoyah Nuclear Plant quality
assurance manual (Reference 3) formed the technical basis for the procurement of
valves prior to 1975. Section 6.1.1 of that document requires qualification of
valves and other components supported by piping systems to 3.0 g horizontal
acceleration and 2.0 g vertical acceleration in accordance with the FSAR commit-
ment. FSAR Table 3.9.E-3 also notes that: " pumps and valves are supported to
assure each component is not seismically loaded in excess of the "g" loading
specified in the design specification".
BASIS: Sore valves are being qualified to computed acceleration levels which
exceed the FSAR comitments and the valve procurement qualification criteria.
REFERENCES
1. TVA SQN Piping Analysis N2-78-5A, Revision 1, dated January 11, 1985.
2. TVA SQN Drawing No. 47K454-58, Reactor Building Unit 1/ System N2-78-6A/
Isometric of Static, Thermal, and Dynamic Analysis of Spent Fuel Pit Cooling
Piping, Revision 1, dated June 19, 1980.
3. TVA Appendix F of SQN Quality Assurance Manual, Design Criteria for Quali-
fication of Seismic Class I and Seismic Class 11 Mechanical and Electrical
Equipment, Revision 1, dated February 10, 1972.
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D3.3-3 (Deficiency) Valve Fundamental Frequency
DESCRIPTION: FSAR Tables 3.9.2-1 and 3.9.2-3 indicate a minimun fundamental
frequency for pumps and valves of 33 hertz. However, Section 3.1.3 of the
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specification used to procure mechanical and electrical equipment for Sequoyah
Nuclear Plant prior to 1975 (Reference 1) notes that equipment exhibiting a
fundamental frequency of 25 hertz can be considered rigid, and can be qualified
by static rather than dynamic analysis. This is a less conservative requirement
than the FSAR commitment.
BASIS: The technical specification used to procure pumps and valves for Sequoyah
Nuclear Plant prior to 1975 specifies a minimum fundamental frequency which is
unconservative with respect to the FSAR commitment.
REFERENCES
1. TVA Appendix F of the Sequoyan Nuclear Plant Quality Assurance Manual,
Design Criteria for Qualification of Seismic Class I and Seismic Class II
Mechanical and Electrical Eouipment, Revision 1, dated February 10, 1972.
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D3.3-4 (Deficiency) Alternate Pipe Support Criteria
DESCRIPTION: FSAR Sections 3.9.2.5 and 3.9.2.6 define the extent of piping to l
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be explicitly analyzed and piping which will be supported to generic qualifi.
cation criteria for Sequoyah Nuclear Plant. TVA Class B, C, D, G. K and M
process piping and instrument lines that do not require complete analysis may
be field routed. Piping 3/8 inch through 1 inch and subject to a maximum
temperature of 650 degrees F may be field routed. Piping 1-1/4 inch through 4
inch and subject to a maximum temperature of 200 degrees F may also be field
routed. Mechanical seismic supports for process pipe 2 inch and less are
detailed on a series of standard detail sheets (Reference 1). General defini-
tions, requireinents and guidelines are tabulated on sheet 1A. Note 11 indicates
that: "The above guidelines do not consider thermal expansion or anchor movements".
In light of this note, the team asked for documentation to confirm that field
routed pipe subject to thermal loads was evaluated to calculate the additional
support reactions due to thermal loads. The TVA Civil Engineering Branch was
not able to obtain documentation which could confirm the systematic evaluation
of thermal loads on supports.
Moreover, TVA Civil Engineering Branch produced two nonconformance reports
written in 1982 which indicated in-house concern with the field routing program
at Sequoyah Nuclear Plant. Nonconformance report SQNSWP8215, dated September 21,
1982, notes, under Description of Condition, that:
A joint Civil Engineering Branch-SWP review on SQN alternate analysis
has shown generic technical and documentational deficiencies on the analysis
criteria used (Civil Engineering Branch-74-2, Civil Engineering Branch-80-5,
Civil Engineering Branch-76-5, and Civil Engineering Branch-75-9) and their
application on SQN. These deficiencies are in the following areas:
1. Incomplete documentation of analysis assumptions and criteria
exceptions.
2. Improper documentation of the report criteria.
3. Inadequate compliance with applicable analysis criteria during
implementation of alternate piping analysis; particularly for
determination of support locations and design loads.
Nonconformance report SQNSWP8222, dated December 21, 1982, notes, under Des-
cription of Condition, that:
EN DES has failed to develop a procedurally controlled system to ensure
that all piping, defined in the Sequoyah Design Criteria SQN-DC-V-13.7
" Alternate Piping Analysis and Support Criteria for Category I Piping
Systems" (last revision 11/8/74), has been supported according to the appro-
priately specified criteria. As a result, there exists the possibility that
piping segments may not have been analyzed nor their dependent pipe supports
released to CONST.
TVA corrective action to address these nonconformance reports should have been
prompt and included systematic identification and evaluation of pipe supports
for-field routed piping subject to thermal loads.
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TVA Civil Engineering Branch indicated that efforts are currently underway to
address these nonconformance reports.
BASIS: The basis for this deficiency is TVA's failure to systematically
address pipe support thermal loads for field routed pipe. TVA's failure to
address nonconformance reports SQNSWP8215 and -8222 in timely fashion
represents inadequate corrective action. Criterion XVI, Corrective Action,
of 10 CFR 50, Appendix B requires in part that measures be established to
assure that conditions adverse to quality such as failures, malfunctions,
deficiencies, deviations, defective material and equipment, and nonconformance
are promptly identified and corrected.
REFERENCES
1. TVA Mechanical Seismic Support, Process Pipe 2" Diameter and Less, Sequoyah
Nuclear Plant, Drawing Series 47A053, Rev. 5, February 14, 1981.
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D3.3-5 (Deficiency) Pump Fundamental Frequency
DESCRIPTION: The team reviewed piping analysis N2-3-3A, -4A, -5A (Reference 1)
for the auxiliary feedwater system from penetrations X-40A and X-40B to the
discharge of auxiliary feedwater turbine pump 1A-S and motor pumps 1A-A and IB-B.
The Upset nozzle forces due to pressure, dead weight, thermal and operating
basis earthquake loads for the discharge nozzle of turbine driven auxiliary
feedwater pump 1A-S are tabulated on page 251. of the analysis along with the
maximum allowable nozzle forces and moments provided by Ingersoll-Rand, the pump
vendor, on page 25C of the analysis. As noted on page 105B of the calculation,
the nozzle loads for the turbine driven auxiliary feedwater pump were qualified
by similarity to the Watts Bar pump based on a review of its stress report
(Reference 2).
The free vibration analysis performed for the pump in the stress report yields
fundamental frequencies of 13 and 16 hertz in the lateral directions and 20 hertz
in the vertical direction. Sequoyah Nuclear Plant FSAR Tables 3.9.2-1 and 3.9.2-3
commit to pump fundamental frequencies equal to or greater than 33 hertz. The
team reviewed the procurement specification for the turbine driven and motor
driven pumps (Reference 3). Section 5 Seismic Requirements, of the procurement
document does not specify a minimum fundamental pump frequency. The pump nozzles
were modeled as rigid anchors in the piping analysis. There appear to be no
requirements to model flexible equipment in piping analysis.
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The team is concerned that piping subsystems which contain flexible equipment
could be subjected to amplified pipe stresses, pipe support reactions and equip-
ment nozzle loads during a seismic event. The team does not consider piping
subsystens at Sequoyah Nuclear Plant which contain flexible equipment to be
adequately qualified by analysis.
BASIS: The procurement document did not specify a minimum pump fundamental
frequency of 33 hertz in accordance with the FSAR commitment. As a consequence,
piping and supports downstream of the turbine driven auxiliary feedwater pumps
are subject to possible amplification of pipe stresses and support reactions
during a seismic event.
REFERENCES
1. TVA SQN Piping Analysis N2-3-3A, 4A, SA, Revision 5, dated September 27,
1985.
2. Mcdonald Engineering Report No. ME-161, Seismic-Stress Analysis of Auxiliary
Feedwater Pumps /TVA Specification 1547/ Watts Bar Nuclear Plant Units 1 and
2/ Manufactured by*Ingersoll-Rand Company, dated October, 1974.
3. TVA Specification 9955 for Steam-Turbine-Driven and Electric-Motor-Driven
Auxiliary Feed Pumping Units for Sequoyah Nuclear Plant Units 1 and 2,
Contract Date July 9, 1971.
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D4.3-1 (Deficiency) Evaluation of Structures for Reinforcing Bar Cuts
i DESCRIPTION: ECN L6495 (Reference 1) replaced the angle valve (1-68-446A)
with a straight through gate valve. In addition, the condensate pot had to be
lowered, necessitating the installation of a new sleeve through the pressurizer
cavity wall. Due to the installation of the sleeve, two horizontal and two
vertical reinforcing bars were cut inside the pressurizer cavity wall. These
were shown on drawing 41N730-1 (Reference 2) related to Workplan 11847.
The USQD for ECN L6495 states that Civil Engineering Branch would evaluate the
installation of the new sleeve to ensure that the structural integrity of tne
wall is not degraded. Contrary to this statement, the original calculations for
this wall (Reference 3) were not revised to evaluate the effects of cutting these
reinforcing bars.
ECN L5202 (Reference 4) deals with the interface of conduit, cabling and piping
between the existing diesel generator building, the powerhouse and the additional
diesel generator building. Due to interference of certain reinforcing bars, Field
Change Request 1476 was issued to cut them as shown on drawing 10N321-2 (Reference 5).
Although the location of the reinforcing bar cuts are shown on this drawing, a
review of calculation 10N320 (Reference 6) showed that the effects of cutting
- reinforcing bars on the structural adequacy of the slab were not evaluated.
BASIS: Section 4.1.2.8 of TVA engineering procedure EN DES-EP 4.52 (Reference
7) requires, in part, that a USQD be prepared for each L-ECN or TVA-approved
design change against a nuclear power plant with an issued operation license
before physical work is authorized. Section 4.5 specifies that completion of
design work for an L-ECN involves, in part, a completeness review of the
associated USQD to be certified by memo. TVA engineering procedure EN DES-EP
2.03 (Reference 8), Section 3.4, notes that failure to adhere to requirements
specifically identified in the USQD evaluation nullifies the USQD evaluation.
REFERENCES
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2. TVA Calculation 41N730-1. Concrete Steam Generator and Pressurizer Enclosure-
Reinf, Rev. 1, 6/29/79.
3. TVA Calculation 41N730-1, Reactor Building Pressurizer Compartment, Final
Design, Rev. 2, 8/9/82.
5. TVA Drawing 10N321-2, Concrete Floors and Walls Reinforcement - Sheet 2,
Rev. 2, 7/22/83.
6. TVA Calculation 10N320, Diesel Generator Building Superstructure and Slabs,
Rev. 1, 10/27/80.
7. TVA Engineering Procedure EN DES-EP 4.52, Engineering Change Notices (ECNs)
After Licensing-Handling, Revision 1, dated April 24, -1984.
8. TVA Engineering Procedure EN CES-EP 2.03, Unreviewed Srfety Question Deter-
mination-Handling and Preparate n, Revision 6, dated April 24, 1984.
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D4.3-3 (Deficiency) Steam Generator Access Pl'_ iorm Design
DESCRIPTION: ECN L5034 (Reference 1) added platforms for access to the steam
generator ports. These permanent platforms were built on the lower steam gene-
rator girders. Although calculations were performed for the design of these plat-
forms, the effect of the platforms on the lower steam generator girders was not
evaluated.
The USQD for ECN L5034 states that the additional loads are transmitted to the
lower steam generator supports and do not exceed the design basis load. A review
of the calculations for the platforms (Reference 2) did not show any consideration
related to the structural adequacy of the girders for the additional loads from
the platforms.
BASIS: Section 4.1.2.8 of TVA engineering procedure EN DES-EP 4.52 (Reference
3) requires, in part, that a USQD be prepared for each L-ECN or TVA-approved
design change against a nuclear power plant with an issued operating license
before physical work is authorized. Section 4.5 specifies that completion of
design work for an L-ECN involves, in part, a completeness review of the
associated USQD to be certified by memo. TVA engineering procedure EN DES-EP
2.03 (Reference 4), Section 3.4, notes that failure to adhere to requirements
specifically identified in the USQD evaluation nullifies the USQD evaluation.
REFERENCES
2. TVA Calculation 48N908-1, Reactor Building Steam Generator Access Platform,
Rev. 2, 2/20/86
3. TVA Engineering Procedure EN DES-EP 4.52, Engineering Change Notices (ECNs)
After Licensing-Handling, Revision 1, dated April 24, 1984.
4. TVA Engineering Procedure EN DES-EP 2.03, Unreviewed Safety Question
Determination-Handling and Preparation, Revision 6, dated 4/24/86.
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D4.3-4 (Deficiency) Cable Tray Support Response Spectra
DESCRIPTION: Response spectra curves for 1% damping given a North-South earth-
quake were used in the design of the cable tray supports for Elevation 714.0'
(Reference 1). The response spectra curves for'the same amount damping have
higher accelerations in the East-West earthquake (Reference 2). Although the
present design criteria for cable tray supports allows the use of a less con-
servative 2% response spectra curve, the existing design may be unconservative
(Reference 3) when the competing effects of the accelerations are considered in
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combination with the damping values. That is, the North-South earthquake with
1% damping maybe less conservative than the East-West earthquake (which has
higher accelerations) using 2% damping.-
BASIS: TVA design criteria no. SQN-DC-V-1.3.4 (Reference 3), Section 4.0,
requires that seismic loads be computed and added to the static loads. The
design criteria also states that these seismic loads are determined from the
peak accelerations of floor spectra. In this particular deficiency, the TVA
engineer failed to use the maximum peak acceleration of the floor spectra for
the two horizonal earthquake. -
REFERENCES
1. TVA Calculation 48N1330,48N1332,48N1333, Auxiliary Building Cable Tray
Supports El. 714.0 ', Rev 2, 2/6/80.
2. TVA Dynamic Earthquake Analysis of the Auxiliary Control Building and
Response Spectra for Attached Equipment, Rev 1,
3. TVA Design Criteria for Category I Cable Tray Support Systems, SQN-DC-V-1.3.4,
Rev 0, 8/20/75.
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D4.3-5 (Deficiency) Loads on Cable Tray Supports
DESCRIPTION: The TVA Design Criteria for Category I Cable Tray Support Systems
(Reference 1) states that for an 18 inch tray, the loads on cable tray supports
should be 75 pounds per linrar foot for the top' tray and 45 pounds per linear
foot for the additional trays. TVA calculations for the cable tray supports
MK 268, MK 42, MK 18A and MK 18B show that for the-top trays only 45 pounds per
linear foot were taken as the loading in the support design (Reference 2). These
represent about 10% of the cable ~ tray support calculations reviewed by the team.
The rest of the support calculations adhered to the loading requirements of the
design criteria. Since a loading lower than required by the criteria was used
in the design, the as-built cable tray supports might be overloaded.
BASIS: TVA design criteria SQN-DC-V-1.3.4 (Reference 1), Section 4.0,
requires that for an 18 inch tray the static maximum loading of the top tray in
a tier should be 75 pounds per linear foot.
REFERENCES
1. TVA Design Criteria for Category I Cable Tray Support Systems, SQN-DC-V-1.3.4,
Rev 0, 8/20/75. i
2. TVA Calculation 48N1330,34,35,74, Auxiliary Building Cable Tray Support Below
El. 734.0 ', 2/2/79.
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D4.3-6 (Deficiency) Torsional Shear Stress Effects on Weld Design
DESCRIPTION: Certain cable tray supports will be effected by torsional shear
stresses during an earthquake due to their asymmetrical geometry. TVA drawing
48N1334 (Reference 1) shows that cable tray supports MK4 through MK4G are
loaded on one side of the support. This configuration will lead to the twisting
of the vertical structural member, inducing torsional stresses into the weld
between this member and the embedded plate.
Team review of TVA calculation (Reference 2) revealed that the additional stresses
due to torsion on the welds were not considered in the cable tray support design.
BASIS: An incomplete analysis was performed for the design of the welds by not
considering the torsional shear stresses. Such consideration is required by the
AISC specification that TVA invokes in section 3.0 of design criteria SQN-DC-
V-1.3.4 (Reference 3).
REFERENCES *
1. TVA Drawing 48N1334, Miscellaneous Steel Cable Tray Supports
El. 714.0 '- Sh 6. Rev 15, 4/13/77.
2. TVA Calculation 48N1330,48N1332,48N1333, Auxiliary Bldg.
Cable Tray Supports El. 714.0 ', Rev 2, 2/6/80.
3. TVA Design Criteria for Category I Cable Tray Support Systems,
SQN-DC-V-1.3.4, Rev. O, 8/20/75.
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U4.3-7 (Unresolved Item) Cable Tray Support Base Plate Analysis
DESCRIPTION: TVA drawing 48N1333 (Reference 1) shows that a surface base plate
with threaded bolt anchors was used for certain cable tray supports. TVA cal-
culation (Reference 2) shows that the design of the base plate and the anchor
bolts used the rigid base plate analysis. TVA design standard (Reference 3)
requires that plate flexibility be considered to determine the anchor tensile
loads.
BASIS: Although the design of this particular base plate was perforred before
the issuance of the design standard, there is a possibility that cable tray
support base plates designed recently might still be using the rigid base plate
approach. A nonconformance report (Reference 4) written on the base plate design
for pipe supports states that the requirements of the design standard have not
been followed since the issuance of the standard.
REFERENCES
1. TVA Drawing 48N1333, Miscellaneous Steel Cable Tray Supports El. 714.0 -
Sh 5, Rev 5, 7/16/75. .
2. TVA Calculation 46N1330, Auxiliary Building Cable Tray Support Below El.
734.0 ', 2/23/79.
3. TVA Design Standard DS-C1.7.1, General Anchorage to Concrete, Rev 3,
11/16/84.
4. TVA Nonconformance Report SQN Civil Engineering Branch 8404, 5/10/84
,
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04.3-8 (Observation) Base Plate Design Criteria
DESCRIPTION: TVA pipe support design manual (Reference 1) section 7.18.2
states that base plates are analyzed as rigid plates for Sequoyah. This is in
contradiction to TVA civil design standard (Reference 2) where it states that
flexible plate analysis will be performed to determine the anchor tensile loads.
The team determined that TVA engineers currently use the flexible plate analysis
for base plate design. The pipe support manual should be revised to reflect the
actual methodology used and eliminate inconsistencies in the design guidance.
REFERENCES
1. TVA Pipe Support Design Manual, Rev, 4/22/83.
2. TVA Civil Design Standard DS-C1.7.1, General Anchorage to Concrete, Rev 3,
11/16/84.
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04.3-9 (Observation) Design Criteria for Tanks
DESCRIPTION: The design criteria for seismically qualifying tanks (Reference
1) states in paragraph 2.3 that the natural frequency of a tank when considered
to be full of fluid should not be less than 33 hertz. A review of the seismic
analysis of the refueling water storage tank (Reference 2) showed that the funda-
mental frequency of this tank is about 6.5 hertz.
The seismic analysis performed showed that the 33 hertz criteria could not be
met. The seismic loads were calculated for a flexible tank and the tank was
designed to withstand such loads. Although the team does not question the struc-
tural adequacy of the tank, the design criteria does not include analysis methods
for flexible tanks and should be so revised.
REFERENCES
i
1. TVA Design Criteria SQN-DC-V-13.6, Design Criteria for Seismically Qualifying
'
Tanks and Reservoirs and Their Supports, 2/23/73.
2. TVA Calculation Modifications to Correct Design Deficiency and Increase Capacity
Refueling Water Storage Tank, 8/22/75.
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.3-1 (Deficiency) Temporary Alterations using TACFs.
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DESCRIPTION: During the review of the ECNs, the team noted that some ECNs were
generated as followup documentation for the TACFs.- Therefore, the team reviewed ,
'
- the Sequoyah plant administrative procedure for' temporary plant modifications
using Temporary Alteration Control Forms (TACF) (Reference-1). The team reviewed *
,
t
selected TACFs issued during a period from 1980 through 1985 (Reference 2).
TACF procedure section 3.4 gives the plant manager an option to require the shift ,
t technical adviser to perform an independent verification of the controlled copies
of the drawings. These drawings, located in~the control room, are marked up for .
'
the "as-built" configuration by the individual completing work under the TACF. '
Failure to require an independent verification of "as-built" drawings also applies
i to the drawing changes made as a result of ECNs. We noted that G/C's review and
the TVA "three system" review both identified drawing errors in the main control
~, room drawings. The team believes that such errors would be minimized if the control
! room "as-built" drawings received independent verification when changes were entered.
t
j The team reviewed eight TACFs (Reference 2) and noted the following:
i
TACF-84-81-62 - Issued and installed on 7/2/84 but found no ECN issued.
.
,
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TACF-80-625-68 - Issued and installed on 7/31/80, PORC review for USQD was ,
done six months later on 2/23/81. Signature block date on
the USQD was missing,
j TACF-82-214-82 - Issued and installed on 10/6/82 and the ECN was issued two
years later in December 1984. ECN is still open. ;
l
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TACF-82-242-82 - Issued and installed on 9/22/62, ECN 6217 was issued two :
} years later in December 1984. ECN is still open.
4
{ TACF-84-81-62 - Issued and installed on 7/2/84 ECN is not yet issued. ,
- ,
, TACF-85-2009-62 - Issued and installed on 7/8/85, no further related documents
j were found.
All of the above TACFs were found open. The team reviewed the computer printout
list of the TACFs. TVA informed us that this list is published in the third
.
quarter of each month. The team reviewed this list and noted that the list does
'
not indicate any requirements for issuance of the followup documents required to
close the TACFs. The team concluded that TVA does not have an effective tracking ;
system for open TACFs which require an ECN or other followup documents to complete .!
closure. The above referenced computer list does not provide any dates by which l
'
TACFs must be closed and/or followup documents issued. -The team found that of
i TACFs could remain open for an indefinite period.
l The team observed that TACFs have been used routinely for permanent changes in
! the plant. This approach is apparently used when the plant believes that OE
i cannot support a change in a timely fashion. The TACFs do not require review
< and/or approval from the Office of Engineering. The team feels that in the
i
absence of a proper tracking system, many TACFs used for the permanent changes
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will remain open and unanalyzed by the responsible design organization for a
long time, resulting in the possible degradation of the safety systems. In
addition, the team is concerned that the Office of Nuclear Power may not have
sufficient controls in place to be considered a. responsible design organization
for original engineering.
The team also noted that the TACF procedure section 5.2 directs the use of a TACF
for a long term change and a Maintenance Request (MR) or the Plant Instructions
for a short term change. However, the procedure does not define the time duration
for a short term or long term change.
BASIS: ANSI-45-2.11 (Reference-3) Section 6 " Design Verification" requires that
the original design of the safety systems or components, must receive an inde-
pendent verification. Section 8. " Design Change Control" of ANSI-45.2.11 states
that " Changes shall be justified and subjected to design control measures
'
commensurate with those applied to the original design."
REFERENCES
1. TVA Procedure #AI-9-Rev. 20 for Temporary Modifications using the TACF Form.
2. TVA TACFS #80-0625, 81-0458, 82-0214, 82-0242, 84-0081, 84-0107, 84-0115,
and 85-2009.
3. American National Standards Institution ANSI-N45-2.11.
4 TVA'S Topical Report #TVA-TR-75-1A Rev.8.
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US.3-2 (Unresolved Item) Sizing Calculations
DESCRIPTION: The team reviewed the operation of the steam driven AFW system
during a loss of ac power. The steam throttle valve and the vent fan for this
system operate continuously during operation of the system and are supplied from
the 125 V de station battery system. The team reviewed the battery sizing cal-
culations to verify that the battery system's capacity is adequate to meet the
<
system demand.
The team determined that TVA does not have proper calculations for the sizing of
the station batteries. The existing calculations (Reference 3) do not address
the correction factors for the operating ambient temperature and for aging. The
calculations were performed before initial operation of the plant, and have never
been reviewed or revised; although, the loading profile of the de
system has undergone changes.
In the absence of analysis and/or calculations the team could not verify that
the installed equipment has adequate capacity to meet the design demands.
Although the battery calculation was performed before the issuance of IEEE-485
- (Reference 2), it is necessary to use temperature correction and aging factors
~
for assessment of the battery's performance. Changes to the loading must be
evaluated to prove that the battery system will have a sufficient capacity to
meet the design commitments per FSAR Section 8.3.2.1.
The team examined the system to determine if a similar problem exists with the
sizing calculations for the battery charger and the 120 V vital ac inverter.
TVA informed us that sizing calculations for these components do not exist.
These calculations were performed before procurement of these components but
were not documented.
1
BASIS: TVA has committed to implement the guidance of ANSI N45.2.11. Section 3
of this standard states that:
I "3.1 General
Applicable design inputs, such as design bases, regulatory requirements,
codes and standards , shall be identified, documented and their selection reviewed
and approved. Changes from specified design inputs, including the reasons for the
,
changes, shall be identified, approved, documented and controlled.
The design input shall be specified on a timely basis and to the level of
detail necessary to permit the design activity to be carried out in a correct
manner and to provide a consistent basis for making design decisions, accom-
plishing design verification measures, and evaluating design changes.
3.2 Requirements
The design input shall include but is not limited to the following, where *
applicable:
1. Basic functions of each structure, system and component.
2. Performance requirements such as capacity, rating, system output. !
3. Codes, standards, and regulatory requirements including the applicable
issue and/or addenda.
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4. Design conditions such as pressure, temperature fluid chemistry and voltage....
6. Environmental conditions anticipated during storage, construction and
operation such as pressure temperature, humidity, corrosiveness, site
elevation, wind direction, nuclear radiation, electromagnetic radiation
and duration of exposure...."-
REFERENCES
1. Sequoyah Unit 1 FSAR Section 8.3.2.1.1.
2. 1EEE-Standard 485-1978-lEEE Recommended Practice for Sizing large lead
Storage Batteries for Generating Stations and Substations. .
3. Battery contract #73C8-83800 Calculation for Battery Sizing.
1
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US.3-3 (Unresolved Item) Motor Operated Valve Thermal Overload Trip Setting
DESCRIPTION: The team reviewed elementary diagrams (Reference 4) for motor
operated valves and noted that the thermal overload trip for the ESF motor operated
valves is not bypassed by an accident signal. The team noticed that the over-
load trip settings of these thermal overload relays were set by the TVA con-
struction staff, in accordance with the Procedure SNP-INSP-INSTR #17 (Reference
1). This procedure directs the technician to set the relays based on a range
of 16-30 seconds of locked rotor current. The team found that some motor operated
valves take up to 60 seconds to complete their travel under the the degraded
voltage conditions; therefore, the arbitrary setting of 16-30 seconds may result
in a trip by the overloads during valvc travel. The team found that the setting
duration o 16-30 was transmitted to the #315-LB-K(2) dated May 8,1974 (Reference 2).
However, the Office of Engineering did not perfonn analyses on a case-by-case basis
to verify that a spurious trip of the thermal overload during the travel will not
prevent the valve from completing its intended safety function.
BASIS: Incomplete travel of the motor operated valves may defeat the engineered
safety system's purpose of safe shutdown by preventing the safety systems to
initiate or complete the recuired safety functions on demand.
REFERENCES
1. SNP-INSP. INSTR. #17-Overload Relay Heater Inspection.
2. Memo-315-LB-K(2), 5/8/74 - Selecting and Testing of MCC Overload Elements.
Stations.
3. 45N779-SH.1 through 16 - Wiring diagrams, 480 V Shutdown Auxiliary Power
Schematic Diagrams.
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US.3-4 (Unresolved Item) Diesel Generator Loading Calculations
DESCRIPTION: The team reviewed the diesel
and 6.9 kv one-line drawings (Reference 2) generator
and noticed loaditems:
the following analysis (Reference
The diesel generator loading analysis was carried out using a 540 hp load lumped
on the 25 second step of the sequencer for the AFW pump motor. However, in
reality the load gradually increases from 486 hp to 540 hp in seven seconds.
This seven second ramp overlaps the 30 second step, which is the critical step for
diesel loading. TVA did not perform an analysis to examine the effects of this-
situation on the voltage and frequency response and recovery limits to verify
that the response is within the values given in Regulatory Guide 1.9 (Reference 3).
The diesel generator loading analysis assumes that all the pressurizer heaters
are turned "off" by the accident signal; however, the loading table correctly
shows heaters which are energized. The 6.9 kv bus-one line drawing (Reference 2)
has a drafting error in note No. 6 in which tripping of the pressurizer heaters
was omitted. These are considered documentation items in that the calculation
used the correct configuration. . ,
One assumption of the analysis states that the transformer nameplate rating was
used for the load analysis; however, the loading table indicates that the actual
connected loads ratings were used. The load table shows that the transformer
load on the 6.9 kv bus consists of two-1500 kva and one-300 kva transformers. How-
ever, design drawings (Reference 2) show that there are three 1500 kva and one
300 kva transformers. The team found that TVA did not analyze the effect of
the third 1500 kva transformer, which remains connected to the 6.9 kv bus during
the zero block loading along with the other two 1500 kva and one 300 kva
transformers. This will affect the frequency and voltage recovery of the diesel
generator in the two second interval between closing of the diesel breaker and
application of the first block load.
BASIS: TVA has committed to implement guidance of Regulatory Guide 1.9 (Reference
3). Section C-4 of this guide states that "The diesel generator unit design
should be such that at no time during the loading sequence should the frequency
and voltage decrease to less than 95% of nominal and 75% of nominal respectively
(a larger decrease in voltage and frequency may be justified for a diesel
generator unit that carries only one large connected load). Frequency should
be restored to within 2% of nominal and voltage should be restored to within 10%
of nominal within 60% of each load sequence time interval." ANSI-N45.2.11
(Reference 4) Section 4 stipulates use of correct design inputs for the design
analysis.
REFERENCES
1. Diesel Generator Load Analysis #B25-86-0204-300.
2. TVA Drawing #45N-724-1,2,3,4 - 6.9 kv One Line Diagram TVA Drawing #45N-765
Sh.1 through Sh.18, and 6.9 kv Shutdown Aux Power Schematic Diagram.
3. U.S. NRC-RG-1.9, Rev.2, - Selection, Design and Qualification of Diesel-
Generator Units used as Standby (on site) Electric Power Systems at Nucle?r
Power Plants.
4. ANSI-N45.2.11, 1975 - Quality Assurance Requirements for the Design of Nuclear
Power Plants.
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A-30
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U5.3-5 (Unresolved Item) Loss of Control Power Annunciation.
DESCRIPTION: The team reviewed TVA drawings (References 1 and 2) for the
6.9kv feeder breaker control circuit for the AFW pump motor and noticed that
the breaker control circuit does not have a provision to detect and annununciate
the loss of control power. In the event of loss of control power, the circuit
breaker will not be able to close when required. This will prevent automatic
operation of the AFW pump, a required function important to the safety of the
plant. TVA informed the team that the control room operators monitor the breaker
status indicator lights. The "off" status of these lights (neither open nor
closed indication) can be taken as an indication of the loss of control power.
TVA further informed the team that at the end of each shift, a documented record
is prepared by the operator for those lights which have changed their status
(from "0N" to "0FF" or from "0FF" to "0N"). The team acknowledged these comments;
however, noted that it is possible that a change in status of these lights could
go unnoticed by the operators for some time.
Regulatory Guide 1.47 states that, "A practical indicating system covering
a wide range of commonly expected conditions, however could be designed if it
included provisions for automatic indication of each bypass or deliberately
induced inoperable condition that meets all three of the following guidelines.
1. The bypass or 6.uperatie condition affects a system that is designed to
perform auP...atically c function that is important to the safety of the
public,
2. The bypass will be utilized t,j plant personnel or the inoperable condition
can reasonably be expected to occur more frequently than once per year, and
3. The bypass or inoperable condition is expected to occur when the affected
system is normally required to be operable."
The team feels that AFW system meets the three conditions stated above. The AFW
systen is important to the safety of the public; plant operators use removal of
control power to naintain equipment; and the inoperable condition is expected to
occur when the AFW system is required such as during accident conditions. In
addition it is possible that loss of control power may occur due to blown fuses,
short circuits, and open circuits.
RG 1.47 further states that " Bypass indication should aid the operator in
recognizing the effects on plant safety of seemingly unrelated or insignificant
events. Therefore, the indication of bypass conditions should be at the system
level, whether or not it is also at the component or channel level. For example,
in a design which utilizes a de power system to control circuit breakers, de-
energizing during maintenance should result in an indication for each safety
system whose operation is dependent on that power system that the safety system
,
is inoperable." The team feels that the above guidance also applies when de-
energizing of de control power occurs automatically due to system fault.
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BASIS: TVA has committed to implement the guidance of Regulatory Guide 1.47,
Bypass and Inoperable Status Indication (Reference 3). Loss of control power
for the breaker control circuit will prevent the auxiliary feedwater pump from
being able to respond to system demand, yet this condition is not indicated as
an inoperability at the system level.
REFERENCES
1. TVA Drawing #45N-724-1,2,384-6.9kv One Line Diagram.
>
2. TVA Drawing #45N-765 SH.1 through 18, 6.9ky Shutdown Aux Power Schematic
Diagram.
3. US NRC RG 1.47 - Bypassed and Inoperable Status Indication for Nuclear Power
Plant Safety Systems.
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05.3-6 (Observation) Voltage Drop Calculations
DESCRIPTION: The team examined calculation #B43-86-0210-924 (Reference 1) and a
draft calculation (Reference 2) to verify that the input terminal voltages at
the AFW pump motor feed breaker control circuit ~, at the steam throttle valve and
at the vent fan of the steam driven AFW pump system are adequate.
The team noted that calculation B43-86-0210-924 (Reference 1) has many unverified
assumptions. Since the validity of the results of this calculation depend on
the correctness of these assumptions, the team feels that assumptions 3.5 and
3.9, explained below, should be verified before restart. Assumption 3.5 states
that drawings used are of the latest revision and all equipment has been
installed. The calculation indicates that this assumption is unverified. The
team feels that this should be verified because the installed length of cables
and wires can be different and thus can change the value of the actual voltage
drop. Similarly, the latest revision of the drawings may show changes in the
loading of the circuit, which in turn will change the calculated values of the
voltage drop. Assumption 3.9 states that the minimum pickup voltage for the
Westinghouse AR Series relay is approximately 85 V dc. The calculation indicates
that this value is unverified. The team feels that in the absence of the
correct value of pickup voltage, adequacy of the available voltage (after
voltage drop) at the input terminals of the control circuit cannot be verified.
The draft calculation (Reference 2) for the voltage drop for the 125 V de valve
and the vent fan circuit was noted to have the following calculation errors. !
The cable length for cable 2SG223 was taken as equal to the distance between
junction box JB-3044 and the motor starter. The correct length should be twice
this length since the actual circuit run is from the junction box to the motor
starter and back. The team noticed that the temperatures used for cable resistance
correction was not consistant between the two calculations. One calculation uses
90*C, and the other calculation uses 40*C. These errors and inconsistencies
should be corrected and were provided to TVA.
REFERENCES
- 1. Calculation #B43-86-0210-924 - 125V dc Vital Instrument Power System Voltage
Drop Study.
2. Calculation #(not assigned) Rough Draft-Voltage Drop Study for 125 VDC Steam
Throttle Valve and Vent Fan for Steam Driven AFW Pump System. !
3. ANSI-N45.2.11-1974 - Quality Assurance Requirements for the Design of Nuclear
Power Plants.
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D6.1-1 (Deficiency) AFW Pump Discharge Pressure Switch Ratings.
DESCRIPTION: Auxiliary feedwater pump discharge pressure switches 1-PS-3-148,
-156, -164, and -171 provide a safety-related interlock for positioning of
bypass control valves. Gilbert / Commonwealth reviewed two engineering change
notices where existing pressure switches were replaced with environmentally
!
qualified devices (References 1 through 4). The team noted that Gilbert
Commonwealth had not compared the technical requirements for the replacement
instruments with the original procurement instrument data sheet to assure that
design basis requirements remained satisfied. G/C stated that such a comparison
was not in their assigned scope of review. The instrument data sheet is used
specify technical requirements for procurement of the pressure switches from
i equipment vendors. Consequently, the team performed this design basis compari-
son and determined that an intermediate replacement had also been made for these
pressure switches. Results of this comparison are provided below:
Technical Original Interim Current
Characteristic (Ref. 5) (Ref. 6) (Ref. 7)
l
l Proof Pressure A500 psi *2000 psig or 2000 psig or
150% des.pr. 150% des.pr.
Maximum Pressure 1200 psig 1650 psig *1085 psig
1 Process Connection 0.5 inch 0.25 inch 0.25 inch
Contact Rating 0.4 ampere 0.5 ampere 0.5 ampere
,
j Contact Voltage 140 VDC *125 VDC 140 VDC
- Contact Action Close-decr. *0 pen-decr. Open-decr.
Trip Setpoint 500 psig *485.3 psig *400 psig
Adjustment Range 285 to 660 psi *5 to 200 psig 45 to 550 psig
'
Manufacturer Custom Comp. Asco Static-0-Ring
f
The team found no indication that those changes denoted by an asterisk (*) to
the original design basis for the interim or current replacements had been
technically documented. TVA stated that existing plant documentation was not
revised when those modifications were initiated; rather, a new instrument data
j
sheet was prepared in each instance,
j For the interim modification, the voltage specification of 125 volts de was in
4 error since it did not accommodate a battery recharging condition. The trip
'
setpoint change to 485.3 psig was not supported by a calculation, and implied
an unrealistic setpoint accuracy for this instrument.
The current modification has a design basis impact for maximum pressure and
trip setpoint characteristics. The 1085 psig maximum design pressure did not
provide for additional margin above the maximum system operating pressure, and
1 the trip setpoint change to 400 psig was not supported by an appropriate
calculation.
l
The team noted that Gilbert / Commonwealth's review had not identified that the
TVA instrument data sheets and Static-0-Ring vendor drawing were not labelled i
as safety-related for the end user. A minor catalog number transposition error l
between the vendor drawing (5N6-B45-NX-CIA-JJTTX6) and the TVA instrument data
sheet (5N6-B45NX-CIA-TTJJX6) was noted by the team.
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BASIS: For the design modifications involving both interim and current replace-
ment pressure switches, a number of changes were made in the design basis with-
out a documented engineering justification when the modification was prepared,
approved, and implemented. The team did not find evidence that the reduction
in proof pressure and changes in maximum operating pressure values were satis-
factory from a system perspective as required by IEEE Std. 279-1971 section 3(7).
Setpoint changes for these safety-related instruments were not supported by
calculations as required by sections 3(4), 3(5), and 3(9) of IEEE Std. 279-1971.
One change in the direct current voltage rating of the switch contacts did not-
conform with IEEE Std. 279-1971 section 3(7).
Instrument data sheets and vendor drawings were not labelled as safety-related,
even though other TVA drawings have been marked in accordance with IEEE Std. 494-1974. This aspect was not identified in the Gilbert / Commonwealth review.
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REFERENCES
1. ECN-L-5823, AFW Pump Disch. Press. Sw. Replacement, Rev. O, 10/5/83.
2. ECN-L-5883, AFW Pump Disch. Press. Sw. Replacement, Rev. O, 10/20/83.
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3. Gilbert / Commonwealth Technical Issue Data Sheet 5, Rev. O, 1/24/86.
4. Gilbert / Commonwealth Technical Issue Data Sheet 16 Rev. O, 1/24/86.
5. TVA Instrument Data Sheet Specification 1596, Rev. O, 6/11/75.
6. TVA Instrument Data Sheet PR-W-3098 (Watts Bar), Rev. 2, 8/3/82.
7. TVA Instrument Data Sheet PR. SE-0307, Rev. O, 10/1/84.
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I D6.1-2 (Deficiency) Feedwater Bypass Control Valve Solenoid Replacement.
! ~
! DESCRIPTION: Replacement solenoid valves 1-FSV-3-35A.-48A,-90A, and -103A were .
installed to improve the response time of -the main feedwater bypass control
valves (Reference 1). Similar replacement solenoid valves for Sequoyah Unit 2
.
were designated as non-quality assurance material (Reference 2). A subsequent
unreviewed safety question determination for this modification stated that.
Class IE solenoid valves were provided; however, this requirement was not
,
satisfied (Reference 3).
Gilbert / Commonwealth's review of this modification identified documentation
- inconsistencies in the safety-related versus non-safety-related designation for
l-
these replacement solenoid valves, and recommended that the solenoid valve and -
its electrical circuits be made safety-related to provide redundancy for main
,
feedwater isolation from the steam generators (References 4 through 6). During
[
the Gilbert / Commonwealth plant walkdown, the non-Category I seismic mounting of
t
the replacement solenoid valve was identified as a deficiency.
The team held a number of discussions with TVA personnel regarding the feedwater
l
j isolation safety function required of these solenoid valves. TVA's reasons for-
l using non-Class IE replacement solenoid valves were based on the valve's~1oca-
! tien in the non-Category I turbine building, the desire to avoid use of Class 1E
'
cables in this building, and the fail-safe characteristics of the solenoid.
,
However, this analysis failed to address the need to satisfy the isolation safety
j function requirement. TVA should have recognized this safety function requirement
j when the solenoid valves were replaced, and should have upgraded the original
,
non-Class IE solenoid valves at that time, consistent with the USQD.
.
1 In response to the recent Gilbert / Commonwealth review of completed design modi-
! fications, TVA has indicated that a Class 1E solenoid qualified for service
i
conditions that exclude 10CFR50.49 environmental considerations will be speci-
! fied and that detailed solenoid mounting requirements will be developed to limit
seismic responses.
BASIS: A feedwater isolation safety function has been required of the solenoid l
{ valves associated with the feedwater bypass control valves. Replacement solenoid
] valves did not meet the Class 1E requirements needed to ensure accomplishment of
this safety function. TVA's reasons for providing non-Class 1E solenoid valves
'
did not adequately address the need to satisfy the feedwater isolation safety ,
j function. The installation of non-Class IE-solenoid valve violates the unreviewed i
j safety question determination. j
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,
REFERENCES
1. ECN-L-5717. FW Bypass Control Valve Solenoid Change, Rev. O, 5/14/80.
4
2. TVA Memorandum, SWP 801016 022 Transfer of Solenoid Valves from l
'
- Watts Bar Nuclear Plant to Sequoyah Nuclear Plant, 10/15/80.
i 3. TVA Unreviewed Safety Question Detemination for ECN-t-5717,
1 SWP 830217 802, 2/17/83.
i 4. Gilbert /Comonwealth Technical Issue Data Sheet 7. Rev. 0,1/24/86. !
! 5. Gilbert / Commonwealth Technical Issue Data Sheet 13. Rev.1,1/28/86.
6. Gilbert / Commonwealth Observation Sheet, Rev. O. 1/24/86.
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D6.1-3 (Deficiency) AFW Pump _ Suction Pressure Switch Setpoint Calculation
DESCRIPTION: Automatic transfer of auxiliary feedwater pump suction from the
condensate storage tank to the essential raw cooling water system is accomplished
by safety-related instruments monitoring for auxiliary feedwater pump low suction
pressure. Gilbert / Commonwealth could not determine whether setpoint values and time
delay requirements were adequately reevaluated as required after system testing
and stated that existing calculations did not take into consideration the
Technical Specification limiting safety setting requirement.
Gilbert / Commonwealth recommended that a new calculation for these setpoints be
performed, but did not identify that the existing calculation of record
(Reference 5) should have been updated or superseded when the pressure switch
mndifications were made. The team noted that this calculation had not been
referenced, updated, or superseded as a result of setpoint changes listed in a
1981 memorandum (Reference 6) and three subsequent engineering change notices
(References 1, 2, and 3).
- In their review Gilbert / Commonwealth did not state that this calculation had
not been marked as a safety-related calculation, and that numeric changes made
in input values were not carried through to calculational results.
,
BASIS: The adequacy and control of existing design basis documentation was not
'
addressed in that the original setpoint calculation should have been referenced
1 in subsequent TVA design documents and then either corrected or superseded.
j Such controls are required by ANSI N45.2.11 section 4.2 Design Analyses, and
- section 8, Design Change Control.
1
- REFERENCES
1. ECN-L-5721, AFW Pump Suction Setpoints, Time Delays, Rev. O, 4/3/84.
2. ECN-L-6124, AFW Pump Suction Press. Sw. Setpoints, Rev. O, 4/25/84
3. ECN-L-6254, AFW Pump Suction Press. Sw. Setpoints, Rev. O, 11/19/84.
4. Gilbert / Commonwealth Technical Issue Data Sheet 15 Rev. 1, 1/28/86.
5. TVA Calculation, SQN-CA-0053, AFW Setpoints, Rev. O, 4/6/79.
6. TVA Memorandum, MEB-810519-022, AFW Time Delays, Rev. O, 5/19/81.
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D6.2-1 (Deficiency) Reactor Coolant System Narrow Range Resistance
Temperature Detector Qualification Category Change.
DESCRIPTION: Reactor coolant system temperature detectors are used in the
reactor protection system for the determination of the reactor coolant system
average temperature which is used to compute reactor trip parawters such as
Overpower and Overtemperature Delta T. These detectors were originally desig-
nated as TVA qualification category A (References I and 2), but were changed
to category C. Category A components are those that are subject to harsh
environmental conditions of design basis accidents for which they must function
to mitigate the consequences. Category C components are those that are subject
to harsh environmental conditions of design basis accidents but are not required
for the mitigation of that accident and whose failure in any mode would not be
detrimental to plant safety. The stated basis for this change was their use as
back-up rather than primary trip signals as described in FSAR transient and
accident analyses (Reference 3). Westinghouse had provided a similar basis for
the elimination of environmental and seismic qualification for ex-core neutron
detectors in late 1983 (Reference 4).
The team did not agree with this change in qualification category. The
instrument sensors connected to the reactor protection system must be environ-
mentally qualified for their intended service conditions. During the inspection,
the team was advised that the Office of Engineering had initiated a revision to
the engineering change notice to restore these sensors to qualification category A.
BASIS: The change from qualification category A to C violated a requirement
that reactor protection system sensors be qualified for their intended service
conditions as stated by section 4.4 of IEEE Std. 279-1971. All reactor trips
should be designed to meet the requirements of IEEE Std. 279 in order to prevent
a possible degradation of the reactor protection system (Reference 5).
REFERENCES
1. ECN-L-6449, Narrow Range RCS Class IE RTD's, Rev. O, 7/24/85.,
2. TVA Unreviewed Safety Question Determination for ECN-L-6449, B25 850918 509,
Rev. 1, 9/18/85.
3. TVA Quality Information Release, B45 851231 268, 10CFR50.49 Category and
Operating Times Calculation Change for Reactor Coolant System Resistance
Temperature Detectors.
4. Westinghouse Letter, WAT-D-5709, NEB 830930 637 Seismic'and Environmental
Qualification of Ex-Core Neutron Detectors, 9/22/83.
5. NUREG 0800, Branch Technical Position ICSB 26, Requirements for Reactor
Protection System Anticipatory Trips, pg. 7A-18, Rev. 2, 7/81.
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D6.3-1 (Deficiency) Specification of Hydrostatic Test to Demonstrate
Instrument Pressure Boundary Integrity After Seismic
Qualification Testing.
DESCRIPTION: Process instruments connected directly into safety class piping
must conform with seismic category I requirements and maintain the pressure
boundary integrity of safety class piping. The demonstration of system pressure
boundary integrity is ordinarily achieved by separate hydrostatic pressure tests c
performed immediately before and after a seismic qualification test.
<
During the team's review of specific process instruments used at Sequoyah, it
was determined that procedural guidance existed for the specification of
hydrostatic test requirements. For example, TVA procedure OEP-09, which has
been applicable to instrument procurement since June 1985, stated that tests
and acceptance criteria for hydrostatic pressure tests may be included in
procurement specifications where applicable (Reference 3). In addition, the
Sequoyah Office of Engineering Project Manual specifically required that
component test requirements include a consideration of hydrostatic pressure
tests (Reference 4).
However, the team determined that TVA had not specified a design performance
test for hydrostatic pressure integrity following the seismic qualification test
for instruments purchased for recent plant modifications (References 1 and 2).
For one procurement contract, the instrument vendor successfully demonstrated ,=
hydrostatic pressure integrity before and after the seismic qualification test
(Reference 5). However, for e second proct.rement contract, the vendor did not
perform a hydrostatic test after the seismic qualification test (Reference 6).
PASIS: TVA procedural requirements with respect to the specification of a
hydrc .;atic pressure test after seismic qualification have not been satisfied.
The r ssure boundary integrity of one set of instruments connected to the =
reactor coolant system has not been demonstrated after the seismic qualification
test.
REFERENCES
1. ECN-L-6380, RCP Bypass Line dP Switch Replacement, Rev. O, 4/29/85.
2. ECN-L-5620, AFW Turbine Discharge Pressure Transmitter, Rev. O,
3/14/83.
3. TVA Procedure OEP-09, Attachment 9 General Content and Format
Requirements for Procurement Specifications, section 8.2.2
4. TVA OE Sequoyah Project Manual,Section VII, Expansion to OEP-06,
item 4.4, Test and Inspection Requirements, 1/10/86.
5. Foxboro N-E110M Differential Pressure Transmitter Qualification
Report, B70 851125 528, Rev. O, 1/28/86.
6. Static-0-Rirg 103AS-BB803-NX-JJTTX6, Differential Pressure
Indicating Switch, Actcn Environmental Test Corp. Reports
18878-84N-1, Rev. 1, 8/30/84 and 18878-84N-3, Rev. 1, 9/25/84.
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U6.3-2 (Unresolved Item) Engineering Change Notice Quality Assurance
and Seismic Analysis Designations
DESCRIPTION: During the preparation, review, and approval of an engineering
change notice, the application of quality assurance and seismic analysis
requirements must be designated by yes or no entries on the form (References 1
through3).
The team reviewed eighty (80) individual engineering change notices for the
1980 through 1985 period, and noted an approximate 9 percent error rate and a
10 percent reversal rate for the designation of quality assurance and seismic
analysis requirements. Several variations in these designations were noted by
the team; namely, the application of one requirement without the other, the
application of neither requirement for safety-related equipment modifications,
and the reversal of an initial designation for one or both of these
requirements.
The team believes that the final designation of the following engineering
change notices were in error by specifying the application of quality assurance
without requiring seismic analysis. Each modification involved one or more
class IE components which are required to meet both the quality assurance
requirements of 10CFR50 Appendix B and the seismic requirements of IEEE Std. 344-1975. A "no" entry for seismic analysis on the engineering change notice
would not provide confirmation of seismic adequacy for these class 1E components:
ECN-L-5057, Reactor Coolant Pump UV and UF PPS Sensors.
ECN-L-5092, AFW Turbine Resistor Box Moved to Wall Mount.
ECN-L-5314, Pressure Switch Moved Outside Crane Wall.
ECN-L-5339, AFW Flow Control Valve Replacement.
ECN-L-5490, AFW Speed Control Moved to Wall Mount.
ECN-L-5717, AFW Control Valve Solenoid Replaced.
ECN-L-5758, Travelling Screen Bubbler dP Instrument Added.
The team noted that the following engineering change notices had a reversal of
the initial determination for one or both of these requirements:
ECN-L-5057, Reactor Coolant Pump UV and UF PPS Sensors,
QA changed from no to yes.
ECN-L-5620, AFW Turbine Pump Surveillance Point Added,
QA and seismic changed from no to yes. '
ECN-L-5717, AFW Control Valve Solenoid Replaced, QA changed from no to yes.
ECN-L-5726, Instrument Line Insulation and Re-Routing,
QA and seismic changed from nc to yes.
ECN-L-5760, Venturi Flow Restrictors Added,
QA and seismic changed from no to yes.
ECN-L-5789, Main Feedwater Solenoid Valve Leakage,
QA and seismic changed from no to yes.
ECN-L-5884, AFW Flow Transmitter Changed, seismic changed from no to yes.
ECN-L-6109, Reactor Coolant Pump 011 Reservoir Level Monitor, seismic changed
from no to yes.
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Since approximately 20 percent of the initial determinations for engineering
change notices reviewed by the team were in error, the team's opinion is that
individual engineers have had obvious difficulty in understanding how the written
criteria were to be applied to a given design modification situation. This view
appears to be supported by the additional ECNs identified by the team that
remained in error following review and approval steps. The team's assessment
is that while the criteria were technically correct they lacked sufficient
clarity necessary for a more uniform application.
BASIS: Criteria for making determinations regarding quality assurance and
seismic analysis were provided in superseded and current TVA design change
procedures (References 1 and 3). Section 4.3.1 of TVA procedure OEP-09 states
that nuclear safety related work includes the specification of quality assurance
requirements and applicable industry codes. The seven engineering change notices
identified by the team where quality assurance aspects were applied without cor-
responding seismic analysis requirements did not conform with these TVA procedures
or provide a justification for the omission of seismic analysis.
REFERENCES
1. TVA Procedure EN DES EP 4.52, ECN's After Licensing, Rev. 1,4/24/84.
2. TVA Procedure OEP-11, Change Control, Rev. O, 4/26/85.
3. TVA Procedure OEP-09, Procurement, Rev. O, 4/26/85.
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06.3-3 (Observation) Essential Raw Cooling Water Screen Wash Pump Control
DESCRIPTION: Redundant differential pressure switches connected across the ERCW
travelling screens had been used to initiate operation of four screen wash
pumps to remove debris. During preoperational tests, it was determined that
pressure drop across the strainers exceeded design values and were causing
improper operation of the backwash and backflush subsystems (Reference 1).
The pump motor circuit wiring for each switch was disconnected on a temporary
basis because switch unreliability had caused constant operation of the screen
wash pumps (Reference 2). The TACF identified this as an alteration to safety-
related equipment, and stated that screen backwashing would be by means of an
automatic timer or by operator manual action until such time as new sensors
were installed. This design modification has been implemented on the basis of
the TACF which has not been superseded by an authorizing ECN.
The engineering change notice (Reference 3) to implement a safety-related bubbler
type differential pressure sensing measurement was initiated in 1982, yet remains
unimplemented. The team considers the . period of time during which this safety
function has been disabled by a temporary modification to be excessive, and that
the design change process would be enhanced if corrective actions were complete'1
in a more timely fashion in such instances.
REFERENCES
1. ECN-L-5512, ERCW Strainer Pre-Op Deficiencies, Rev. O, 9/30/82.
2. TACF-82-258-67, Disconnection of dP Sensor Wiring, Rev. O, 10/7/82.
3. ECN-L-5758, ERCW Screen Wash DP Sensor Change, Rev. O, 12/8/82.
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