ML20215N282
| ML20215N282 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 10/21/1986 |
| From: | Architzel R, Imbro E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| To: | |
| Shared Package | |
| ML20215N277 | List: |
| References | |
| 50-327-86-45, 50-328-86-45, NUDOCS 8611050169 | |
| Download: ML20215N282 (23) | |
See also: IR 05000327/1986045
Text
,
U.S. NUCLEAR REGULATORY COMMISSION
OFFICE OF INSFECTION AND ENFORCEMENT
Division of Quality Assurance, Vendor, and Technical
Training Center Programs
Report Nos.:
50-327/86-45, 50-328/86-45
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:
Knoxville, TN
Inspection Conducted:
July 21-25, 1986
Inspection Team Members:
Team Leader:
R. E. Architzel, Senior Inspection Specialist, IE
Mechanical Systems:
F., Mollerus, Consultant, Mollerus Engineering Inc.
Mechanical Components:
~A. V. duBouchet, Consultant
Civil / Structural:
A. Unsal, Consultant, Harstead Engineering
Electrical Power:
S. V. Athavale, Inspection Specialist, IE
Instrumentation &
Control:
L. Stanley, Consultant, Zytor Inc.
Nuclear Systems:
J. M. Leivo, Consultant
,
Operations:
P. E. Harmon, Resident Inspector, SQN
Licensing:
J. J. Holonich, Project Manager, NRR
F. Rinaldi, Structural Engineer, NRR
0. Terao, Mechanical Engineer, NRR
h
W }/-[b
'
Ralph E. Architzel u
Date
Team Leader
W Y Wm b.
Icl2//)(
Eugene V. Imbro
' Date
Section Chief
g10$$b$$IbO
7
G
, _- -
.
.
LIST OF ABBREVIATIONS
Atomic Energy Commission
.
ANSI
American National Standards Institute
Anchor Point Movement
CFR
Code of Federal Regulations
Combustible Gas Control System
C/R
Commitment / Requirement
Design Basis Document
DBVP
Design Baseline and Verification Program
Division.<of Nuclear Engineering
Engineering Assurance
Engineering Change Notice
Engineered Safety Features
Environmental Qualification
Final Safety Analysis Report
GDC
10CFR50, Appendix A, General Design Criteria
Heating, Ventilation and Air Conditioning
i
INP0
Institute of Nuclear Power Operation
IEEE
Institute of Electrical and Electronics Engineers
Loss of Coolant Accident
NEB
Nuclear Engineering Branch
NMS
Neutron Monitoring System
Non-Conformance Report
'
NRC
Nuclear Regulatory Commission
Nuclear Steam Supply. System
Piping and. Instrumentation Diagram
RDBD
Restart Design Basis Document
Significant Condition Report
SQEP
Sequoyah Engineering Procedure
'
Sequoyah Nuclear Plant
Tennessee Valley Authority
'
i
.--
__
- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
.
SEQUOYAH NUCLEAR POWER PLANT
Design Baseline and Verification Program v
-
Inspection Report 50-327/86-45 & 50-328/86-45
July 21-25, 1986
1.
INTRODUCTION AND BACKGROUND
,
The design baseline and verification program (DBVP) was developed by the
Division of Nuclear Engineering (DNE) to resolve design control issues
described in several TVA sponsored evaluations and audits and NRC inspections.
The Sequoyah Nuclear Plant (SQN) Design Baseline and Verification Program will
be used by TVA to provide the required level of confidence that the modifica-
tions to selected plant systems, implemented since receipt of the operating
license, have not resulted in any violation of the plant's licensing basis. The
program is described in the " Program Plan for the Engineering Assurance
Independent Oversight Review for the Sequoyah Nuclear Plant Design Baseline and
Verification Program;" ' dated May 9, 1986 and forwarded to the NRC as an
enclosure to Mr. R. L. Gridley's letter dated June 27, 1986.
2.
PURPOSE
NRC inspection activities related to the TVA's DBVP and associated Engineering
Assurance (EA) independent technical oversight of Sequoyah Nuclear plant are
planned to be conducted in several phases:
(1)
Inspection of program preparation and initial implementation (EA Review
plans and procedures, DBVP procedures, walkdown results).
(2)
Inspection of program implementation, including design criteria
preparation, Engineering Change Notice (ECN) and system evaluations.
(3)
Inspection of DBVP and EA oversight results and corrective actions.
.
This inspection focused on the development and updating of the design criteria.
The NRC previously conducted an inspection (Report Nos. 50-327/86-38 and
50-328/86-38) of the DBVP. This previous inspection focused on overall
DBVP plan and scope, implementing procedures, and the conduct and results of
walkdowns.
The purpose of this inspection was to (1) review TVA's program for design criteria
preparation, (2) review a sample of the revised and newly issued design
criteria, and (3) overview the efforts of TVA's Engineering Assurance (EA)
group for independent review of the design criteria. An integral part of the
design criteria preparation was the generation and use of a (new) commit-
ment / requirement (C/R) data base.
The purpose of the Sequoyah design criteria reconstitution program is to
generate revised design criteria documents which address system and general
functional design requirements governing the design of structures, systems and
components. At TVA, these design criteria documents include current licensing
commitments and regulatory requirements, as well as design criteria that are
.not commitments but TVA self-imposed standards of " good engineering practice."
The revised design criteria will be used by TVA as the basis to review all
plant modifications made to those systems or portions of plant systems within
1
s
_
9
-
.
.
.
'
4.
SUMMARY OF FINDINGS
.
4.1 Program for Design Criteria Preparation
,
Team review of SQEP-18, " Procedure for Identifying Commitments and Requirements
as Source Information for Sequoyah Design Criteria Development," showed that a
program has been established to identify the licensing commitments and other
design requirements. The team also reviewed SQEP-29, " Procedure for Preparing
the Design Basis Document," which addresses the methods to be used to capture
the C/Rs and other design input in an upper tier, commitment driven, compila-
tion of design and design documentation requirements.
During a prior inspection of the DBVP (Inspection Report 86-38), the team had
performed a preliminary review of the computerized list of licensing commit-
ments and design requirements called the " Commitment / Requirement Data Base."
The team found that this list, developed by TVA/Impell, was not independently
verified. Although Quality Assurance controls were not applied to the
information retrieval process, the Engineering Assurance group has conducted
reviews in this area on a sampling basis. The NRC team does not consider that
these reviews were done in sufficient technical depth or were of sufficient
scope to allow meaningful conclusions to be drawn regarding the completeness of
the C/R data base. This data base forms a portion of the basis for the newly
issued design criteria. d,ocuments.
.
The team discussed future plans the project was considering to verify the
accuracy of the C/R data base. Consideration was being given to independently
verifying incorporation of all C/Rs in design output documents, for example.
Prior to restart, TVA considers that the independent verification provided
during preparation of new and revised design criteria provides the required
verification of inc,o'rporation of C/Rs. The team remains concerned regarding
the identification of C/Rs. Therefore, TVA should clearly define their basis
for concluding that the C/R data base has accurately captured all licensing
commitments, regulatory requirements and other pertinent design information as
applicable, e.g. NSSS vendor interface requirements (Inspection Report 86-38,
Observation No. 5.4).
4.2 Revised and New Issue Design Criteria
The team identified a concern regarding the incorporation of proprietary
information in the design criteria due to a lack of availability (within
TVA) of proprietary source documents containing C/Rs (Observation No. 3.4)
The team identified several cases where C/Rs applicable to selected design
criteria were not captured (Observation Nos. 4.4, 5.5, 6.6, and 6.7).
Environmental Qualification requirements were found to be incorporated in an
inconsistent manner among various design criteria (Observation Nos 4.5 and
6.5). The team also identified missing requirements in the design criteria for
'
the neutron monitoring system, apparently at the interface between the NSSS
and TVA. These findings, collectively, raise a concern regarding the compre-
hensiveness of the design criteria.
Several technical observations were identified relating to the coordination
(between design criteria) of overcurrent protection and cable sizing for
medium voltage motors (Observation No. 5.6); the collection of varied single
l
failure definitions and commitments into a single document (Observation
!
l
3
-
-
-
-
<
'
.
-
.
the scope of the .DBVP since operating license issuance to provide assurance
'
that SQN is in conformance with its licensing basis.
As a part of the Design Baseline and Verification Program, TVA is preparing a
.
"
Design Basis Document (DBD) and a Restart Design Basis Document (RDBD). The
DBD defines, establishes, and maintains the design requirements for the
Sequoyah Nuclear Plant. Although a design basis currently exists for the
Sequoyah Nuclear Plant, the design basis documents were not always readily
retrievable in a verified form. Thus, TVA identified a need for a verified,
controlled design basis document to be maintained throughout plant life. The
DBD is intended tc, be used to evaluate and control design changes, to respond
to abnormal operations and events, to evaluate limiting conditions for
operation, to perform safety reviews, to assess conditions adverse to quality,
to assess operating experience reports',' end'to provide an interface with
outside organizations. The RDBD will be the initial issue of the DBD and will
cover those safety-related systems identified by TVA calculation SQN-0SG7-048
which are required to support hot shutdown and mitigate postulated accidents
described in FSAR Chapter 15.
The DBD contains general design criteria for site, plant, structures, and sys-
tems which establish the plant-specific design input requirements. The 080
also contains certain detailed design criteria, system descriptions, design
input drawings, engine'er.ing decisions, analysis results, and engineering para-
meters for detailed de' sign. The design commitments and requirements to be used
as a basis for developing the design criteria were identified using TVA
procedure SQEP-18.
~
3.
INSPECTION ACTIVITIES
The following activities were generally performed by all team members.
Evaluation of applicable DBVP procedures for generation of the
commitment / requirement data base and for updating the design basis
document.
Review of selected design criteria generated or updated as a result of
the DBVP.
Review a sample of the commitment / requirements associated with the
selected design criteria to verify their incorporation.
Examination of the results to date of the independent oversight review of
C/Rs and design criteria.
In addition, the HRR team members conducted a walkdown at the Sequoyah Nuclear
Plant to physically examine hardware which is the subject of substantive
technical issues. Observations from the walkdown in the small bore piping and
HVAC duct and support disciplines are also discussed in this report.
1
2
.
.
.
No. 6.8); an'd implementation of 10 CFR 50 General Design Criteria requirements
for containment penetrations (Observation No. 8.2)
Several concerns were identified by team members from the NRC Office of
Nuclear Reactor Regulation who are reviewing interim acceptance criteria being
applied for SQN piping, pipe supports, and cable tray designs (Observation
Nos. 8.1-8.4).
These findings are documented in this report for completeness,
but will be resolved independently by NRR during their review of these issues.
During this inspection, the team reviewed several system functional ' design
criteria, as well as two recently issued Civil Engineering Branch general
design criteria that supersede Watts Bar design criteria which TVA had used to
design pipe anchors and pipe supports at Sequoyah Nuclear Power Plant. The
team identified two design requirements specified in the Watts Bar design cri-
teria for the design and modification of pipe supports that TVA did not specify
in the Sequoyah design criteria issued on June 23, 1986. The new design
criteria does not require that the local stresses generated in the piping by
stiff pipe clamps with large preload be checked and allows pipe axial loading
of floor or wall sleeves that have not been designed for such pipe loads
(Observation No. 3.4).
The Sequoyah Project has not found any items which they determined necessitated
the issuance of a new desian criteria or any revision to existing design
criteria (prior to restart) in the Civil / Structural discipline during review of
the licensing commitments and design requirements. Several examples of
situations which the team' considers should have resulted in such criteria
changes were identified (Observation Nos. 7.2-7.3).
4.3 EA Independent Oversight Review Program Plan
The team expressed certain reservations with the review plan being implemented
for the instrumentation and control discipline to assess the technical adequacy
of design changes. The team's concerns included:
(1) establishing an appropriate balance between important technical
issues and quality assurance issues as described in the checklists;
(2)
the representativeness of the four systems selected, and
(3) the relatively small sample sizes chosen for. review of design
criteria, engineering change notices, and field change notices.
The instrumentation and control discipline specific action plan for EA review
of the C/R data base states that approximately ten representative commitments /
requirements would be selected from a variety of source documents including
letters and memorandums. These choices will then be reviewed for their inclu-
sion in the Sequoyah commitments / requirements (C/R) data base. Review attributes.
described in the implementing checklist include the identification of source
documents, training of personnel for this activity, kreparation of the C/R data
sheet used for input to the Sequoyah commitment / requirement data base, and dis-
tribution of the data base output to involved us~ers.
i
While a sample size of ten may be sufficient to validate the processing of
identified commitments and requirements through the Sequoyah C/R data base, the
t
4
,
--
-
_ _ _ _ _ _ __
proposed action plan is unlikely to be sufficiently comprehensive to vabdate
the method used to initially identify individual commitments and requirements
from the designated source documents. Since the overall effectiveness of the
C/R data base-is dependent upon the accuracy of the determination of specific
commitments and requirements, greater emphasis appears necessary to validate
the method and procedure used for identification of commitments and
requirements.
Although no specific observations were identified, team members reviewing in
.Other disciplines also noted a lack of technical depth and breadth in the
samples selected for oversight by EA.
Limited oversight products were
available in the mechanical systems and nuclear systems areas.
In the electric
power area, the team found that the EA oversight was thorough and identified
findings (Action Items) similar to those identified by the team.
In the Civil / Structural. area the team found that the plan of action and the
attributes shown in the'EA oversight review plan indicate that an adequate plan
has been established to review the Sequoyah project work.
5.
SPECIFIC COMMENTS
Specific comments of individual NRC discipline inspectors are categorized as
observations. The observations and a description of the activities performed
by each discipline of:the NRC team are provided in Attachment A of this report.
TVA actions relating '.c individual observations will be reviewed by the NRC
during future inspections. These observations elaborate on the general
comments stated in this report and in some cases provide additional coments
not considered to be of a general nature.
6.
MEETING SUMMARIES - REFERENCES
A summary of the meetings. held relating to the DBVP inspection and a list of
references are provided in Attachment B.
5
.
.
.
- - .
--
.
- . .
. .
.
-
- - .
-- ..
.
Attachment A - Inspsction Activities and Obsarvations
NOTE: The observation numbers used in this report are a continuation of the
~
.
numbers used for the previous DBVP inspection.
(ReportNos. 50-327/86-38 &
,
i
50-328/86-38). The references are listed in Attachment B.
.
i
-
1.0 OPERATIONS
!
!
In the operations area the team examined selected design criteria, EA action
items from the previous inspection, incorporation of various commitments /re-
quirements, and coordinated with other team disciplines in the conduct of the
inspection. No observations were identified in the operations area.
,
2.
MECHANICAL SYSTEMS
In the mechanical systems discipline, the team reviewed the.following design
criteria generated for the baseline restart effort.
4
'
Responsible
<
Design Criteria No.
Ti tle
TVA Branch
SQN-DC-V-3.1.1
Steam Generator Blowdown System
MEB
SQN-DC-V-4.1.1
Main Steam System
MEB
>
SQN-DC-V-4.2
Feedwater System
MEB
SQN-DC-V-7.4
Essential Raw Cooling Water System
MEB
SQN-DC-V-13.9.3
. Auxiliary Building Ventilation & Cooling
MEB
SQN-DC-V-13.9.8
Auxiliary Feedwater System
MEB
<
SQN-DC-V-13.9.9
Component Cooling Water
MEB
SQN-DC-V-16.0
Auxiliary Contro1 Air System
MEB
SQN-DC-V-27.6
Residual Heat Removal System
MEB
!
The team reviewed the Engineering Assurance Oversight Review
!
Plan - Mechanical (Reference 9). This plan is organized by the following
activities.
- Walkdowns
!
Licensing Commitments
!
Design Basis
l
Change Control Board
"
Evaluation of Change Documents
i
!
Comparison of Design Documents with Walkdown Results
System Evaluations
Modifications to Control Room Drawings
l,
Unreviewed Safety Question Determination (USQD)
l
Each activity description includes a plan of action, description of sample
j
size, and an attributes check list. The latter is basically a check list to
j
evaluate the acceptability of the activity being monitored.
1
One observation was identified concerning TVA's access to and incorporation of
j
proprietary information in the design criteria (Observation No. 2.3).
,.
)
Observation No. 2.3 - Status of NSSS Vendor Proprietary Information
i
Several of the commitment / requirements (C/Rs) listed in the analysis report for
l
l
the component cooling water system are NSSS vendor reports and memoranda that
!
}
A-1
b.
.,,
.,,,-,--,--..---.,,,-,,,n---.
- - .
, , - - , , , , , , - - , ,
- , , , , - - . - - - . - - - . - . - - - , - . , , . , ~ . , , . -
.,----,v,
,,, . - - + . .
-
Attachment A - Inspection Activities and Observations
are considered by the NSSS vendor to be proprietary and unavailable to outside
-
The NSSS vendor also stated.that the C/Rs contained in
organizations.
.
proprietary memoranda are also contained in formal documents that have been made
available to TVA. The inspection team was informed that the formal documents
are being used as a source of C/Rs for the design criteria; the NSSS vendor
proprietary correspondence is not. TVA's basis of the acceptability of_this
approach was the NSSS vendor's statement that all C/Rs are contained in the
formal documents. The titles of the proprietary correspondence will be kept in
the C/R tracking system until restart. At that time, TVA Branch Chiefs will
evaluate all remaining C/Rs not addressed in the design criteria.
The inspection team will continue to follow developments in this NSSS vendar
interface area and TVA's efforts to assure that all NSSS vendor C/Rs are
incorporated in the design basis documents.
3.
MECHANICAL COMPONENTS
In the mechanical components discipline, the team reviewed Sequoyah Design
Criteria SQN-DC-V-2. 14, Piping System Anchors installed in Category I
Structures, which was issued June 30, 1986 and supersedes Watts Bar design
Criteria WB-DC-40.31.15 for new pipe anchor designs and modifications to
existing pipe anchor designs.
-
'
The team also reviewed Seguoyah Design Criteria SQN-DC-V-24.1, Location and
Design of Piping Supports and Supplemental Steel in Category I Structures,
which was issued June 23,.1986 and supersedes Watts Bar Design Criteria
WB-DC-40.31.9 for new pipe support designs and modifications to existing pipe
support designs. Observation No. 3.4 was identified relating to elimination
of requirements from a design criteria document.
<
Observation No. 3.4 - Pipe Support Design Criteria
Sequoyah Design Criteria SQN-DC-V- 24.1, which TVA issued on June 23, 1986 for
the design of new pipe supports and modifications to existing pipe support
designs, supersedes Watts Bar Design Criteria WB-DC-40.31.9, but does not carry
forward the following two design requiremer.ts which are contained in the Watts
Bar design criteria document.
(1) Section 8.2.12 of the Watts Bar Design Criteria, Stiff Pipe Clamps, which
notes that stiff pipe clamps with extremely large pre-load may induce sig-
nificant localized piping stresses, and requires that piping stresses be
investigated for stiff clamps identified in NRC Information Notice 83-80.
(2) Section 8.3.1.2 of the Watts Bar Design Criteria, which prohibits pipe
axial loading of a floor or wall sleeve unless the sleeve has been designed
for a pipe load.
l
This issue needs to be further examined by TVA.
4.
NUCLEAR SYSTEMS
In the nuclear systems area the team reviewed TVA's Engineering Assurance (EA)
oversight work products. At the time of the inspection, the Nuclear Engineer-
ing Branch (NEB) EA products available for NRC review in this area were:
A-2
_ _ _
..
,
Attachment A - Insptcticn Activities and Observations
'
.
(1) A review of the design criteria for the combustible gas control system
(Reference'4) in accordance with SQEP-29, " Procedure for Preparing the
.
"
Design Basis Document for Sequoyah Nuclear Plant."
_
(2) A review of commitment / requirement (C/R) identification for the auxiliary
feedwater and essential raw cooling water systems in accordance with
SQEP-18, " Procedure for Identifying Commitments and Requirements as Source
Information for Sequoyah Design Criteria Development."
The team also reviewed a . sample of DBVP Project generated design criteria. At
the time of the inspection, approximately. half of the NEB design criteria had
been issued; none of the NSSS design criteria had been issued. The team
reviewed the design criteria issued for the combustible gas control system
(CGCS), containment isolation, single failure, and remote shutdown criteria
from locations outside control room (Refs. 4 through 7, respectively).
In
addition, the team reviewed the design criteria for the neutron monitoring
system that had recently been transmitted by _the NSSS vendor, but had not yet
been issued by TVA.
Three observations were identified relating to incorporation of commitments,
consistency of the design criteria, and completeness of design criteria
(Observations Nos. 4.4 - 4.6).
Observation No. 4.4 - Spray Shield Commitment / Requirement for Certain Hydrogen
Igniters in Upper Compartment
.
A spot check of CGCS C/Rs; identified by TVA for commitment changes (and
apparently excluded from the analysis report used in preparing design criteria)
indicated that TVA's C/R to provide enlarged spray shields for upper compart-
ment igniters (Reference 8) has not been included in the CGCS design criteria
(Reference 4). Thi,s' missing requirement does not. appear to have been
identified during the EA review of the design criteria. The team recognizes
that the igniters are provided for mitigation of accidents beyond the scope of
FSAR Chapter 15 events; however, since TVA appears to take credit for the
enlarged spray shields in assuring functionality of the igniters following a
degraded core loss of coolant accident, the team believes the requirement for
' enlarged shields should be included in the CGCS/ hydrogen mitigation system
design criteria.
Observation No. 4.5 - Clear and Consistent EQ Requirements in Design Criteria
The team observed that some of the environmental qualification (EQ) require-
ments as stated in'the few design criteria sampled appear to be somewhat
inconsistent and might be misinterpreted by a user of the criteria who is not
particularly knowledgeable of EQ requirements and procedures.
For example, the
CGCS design criteria do not clearly state what EQ requirements apply to
existing equipment.
Observation No. 4.6 - Missing Criteria for Neutron Monitoring System
While the Neutron Monitoring System (NMS) design criteria had not yet been
issued by TVA, the team reviewed the design criteria provided recently by the
NSSSvendor(Reference 3). Based on the NMS' boundary definition provided in
References 1 and 2, the design criteria appear to be incomplete in the
~
following areas:
l
A-3
l
'
, . - - . - , - -
- - - - - -
.
-
- - . - -
-
- - - - - -
- - - . - . .
Attachment A - Inspection Activities and Observations
.
(1) No requirements are stated or referenced for providing indication of
source range flux outside the control room as required to accomplish safe
-
shutdown from outside the control room; for example, these requirements
are not presented in paragraphs 3.5.1.n or Tables 3.1-1, 3.1-4 of the
document.
(2) No requirements are stated or referenced for electrical penetrations that
are required for NMS functionality.
An indirect reference to IEEE-317
~
provided in Section 3.5.2.8 of the document is not considered sufficient -
to cover the special penetration requirements for these high impedance
.
instruments.
This observation also reflects the concern expressed in Observation No. 4.2
(Inspection Report 86-38) that system functional boundaries should be expli-
citly defined for this system.
For example, it appears that the NSSS vendor
has limited the scope of the NMS design criteria to the system and equipment in
its scope of supply, rather than addressing the functionality of the integrated
system for safe restarts TVA should give sufficient attention to the NSSS/Bal-
ance-of-Plant interfaces, as well as to overall system functionality, so that
no significant items are inadvertently omitted from the review scope.
5.
ELECTRIC POWER
In the electric power di,scipline, the team reviewed portions of design criteria
SQN-DC-V-11.4.1 (Revis' ion 2), for auxiliary power; SQN-DC-V-11.3 for cable
applications; and TVA':s DBVP procedures SQEP-16 (Revision 0), SQEP-29
'
.(Revision 1), and SQEP-18.(Revision 1). The team also reviewed the action
items identified during EA's independent oversight review of design criteria
SON-DC-V.11.2, SQN-DC-V.11.6, SQN-DC-V-11.8 and DBVP procedure SQEP-16
~
(Revision 0). The team noted that of the eight design criteria documents
required in this discipline prior to restart, the EA oversight team had
completed reviews for six design criteria documents. This review resulted in
the identification of 15 Action Items, similar in nature to the observations
identified by the team.
Two observations were identified relating to commitment / requirement incorpora-
tion and cable sizing (Observation Nos. 5.5 and 5.6).
Observation No. 5.5 - Commitment / Requirement Inclusion in Design Criteria for
the Auxiliary Power System
The team reviewed portions of Design Criteria SQN-DC-V-11.4.1 (Revision 2), and
noted the following weak areas.
(1) Many commitment / requirements have been identified on the commitment /
requirement evaluation sheet but were not captured in the design criteria.
Examples include the following.
SON EEB DRW 1072 - TVA's commitment pursuant to IE Bulletin 79-25, which
prohibits the use of Westinghouse type BFD and NBFD relays.
SQN EEB DRW 1073 - TVA's commitment pursuant to IE Bulletin 79-11, which
prohibits the use of Westinghouse type DB-50 and DB-75 circuit breakers
with overcurrent devices.
A-4
. .__
_
_-
'
~ Attachment A - Insp;ction Activities and Observations
SQN EEB DRW 1077 - This.C/R requires mandatory fusing of control ~ circuits.
.
,
SON EEB DRW 1078 - This C/R deals with problems of spurious operation of
diesel generator breaker differential relaying.
(2) Many commitment / requirements which are applicable to the auxiliary power
system were not shown on the commitment / requirement evaluation sheet for
this system. Examples includes the following.
SQN EEB DRW 1092 - This C/R addresses the increased' fault interrupting duty
of 6.9 kv breakers for 3-phase faults during periods when the turbine
generator is operated at higher voltages (25.2 kv) to meet the demands of
the distribution grid.
SON EEB DRW 1017, 1018, 1019 - These C/Rs address the topic of class 1E
station battery capacity, and were not included in the auxiliary power C/R
evaluation.
Observation No. 5.6 - Cable Sizing for Overload Currents
Section 8.5.3d of Design Criteria SQN-DC-V-11.4.1 (Revision 2), stipulates that
for medium voltage motors the overload signal will activate an alarm, only.
In
such situations, since the overload signal does not trip the breaker, the motor
feeder cable could carry an overload current from the time of alarm actuation
until the breaker is tripped manually.
Sizing of the cables should address
this situation.
The team found that design criteria SON-0C-V-11.3 does not address overload
current condition but-stipulates the use of 125% of full load current as the
basis for sizing the, feeder cables.
In situations where overload setting 'is
more than 125% of the full load current, the cable sizing may not be adequate.
This margin of 25% may not be sufficient since the margin may be required for
low voltage operation of the motor and by the motor's service factor (capability
of supplying greater than rated load).
In situations where a motor is required
to operate below its normal rated voltage and/or the motor service factor is
more than 1.0, the motor draws more tha.: its normal full load current. The team
believes that for cable sizing, the c'riteria should address evaluation of each
load on a case by case basis and require sizing the cable accordingly, instead
of directing cable sizing using a blanket rule of 125% of full load current.
6.
INSTRUMENTATION AND CONTROL
The inspection team reviewed TVA procedures SQEP-18 for identification of
commitments and requirements and SQEP-29 for preparation of the design basis
document. A detailed review was made of TVA Design Criteria 26.2 for
environmental qualification, 32.0 for the auxiliary control air system, and
2.16 for single failure criteria. Portions of Westinghouse Design
Criteria 27.6 for the residual heat removal system, 27.8 for the neutron
monitoring system, and 27.9 for the reactor protection system were reviewed for
their instrumentation and control aspects.
In the course of this NRC team
review, several system and technical subject output sorts from the TVA data
base for commitments and requirements were used as well as a TVA computerized
document data base (RIMS) output sorted by the reactor protection system
keyword. A number of commitment / requirement record files in the electrical,
A-5
.
.
.
.
--
_
-
-
_ _ - -
Attachment A - Inspection Activitics and Obstrvations
-
mechanical, and nuclear disciplines were examined, as well as some source
documents such as the FSAR, project letters, and plant drawings.
The NRC team also examined instrumentation and control action items identified
by the TVA EA oversight team during their review of TVA procedures SQEP-18 and
-29 and during their review of the steam generator blowdown and accident
monitoring design criteria. The EA team indicated that one or two more design
criteria would be reviewed; however, the team was concerned with the lack of
depth of technical review being demonstrated by the instrumentation and control
discipline.
Four observations were identified in the Instrumentation and Control discipline
relating to specifications for replacement parts, inclusion of C/Rs in design
criteria, and definition of the single failure criterion (Observation
Nos. 6.5-6.8).
Observation No. 6.5 - Replacement Part and Equipment Qualification
TVA Design Criteria SQN-DC-V-32.0 for the safety-related auxiliary control air
system has inconsistent criteria for the qualification of Class IE replacement
parts and equipment in that section 3.3.1.f specifies the use of IEEE Trial Use
Standard 323-1971. For example, a number of containment isolation valve
solenoids for this system are located in a narsh environment, and must now
conform to IEEE 323-1974 to satisfy the requirements of 10 CFR 50.49.
Although replacement equipment can be exempted from these requirements, sound
reasons must be provided as described in Regulatory Guide 1.89.
Consequently,
the design criteria should be changed to specifically describe those conditions
where use of the IEEE 323-1971 standard might be permitted.
Observation No. 6.6 - Auxiliary Control Air System Design Criteria
A comparison was made of the design basis commitments and requirements data
base output with the design criteria document for the auxiliary control air
system (SQN-DC-V-32.0).
In three instances, the team noted that applicable
commitments or requirements had not been explicitly converted into design
criteria, as follows.
C/R SQN MEB LWB 1208 stated that an air compressor time delay relay had
been replaced with a different relay to achieve a time delay requirement
of 3 seconds. This time delay is used to initiate rapid loading of the
after the air header pressure has
auxiliary control air compressor shortly (which had a range of 20 to 200
dropped to 80 psig. The original timer
seconds) did not assure that a minimum 70 psig air header pressure
requirement would be maintained. The design criteria document did not
reflect this particular time delay requirement.
C/R SQN NED FAK 1229 identified a licensee event report (LER) that
contained a statement that TVA would upgrade the air supply to
safety-related for containment radiation monitoring containment isolation
valves. The design criteria does identify various valves connected to the
auxiliary control air system, but does not list the radiation monitor
isolation valves. The team found that upgrading the air supply to
containment radiation monitor containment isolation valves was not
addressed during revision of the design criteria document.
A-6
Attachment A - Inspection Activities and Observations
'
C/R SQN MEB LWB 1060 identified physical separation requirements for air
-
headers inside cbntainment in response to NUREG 0737. A comitment was
made that there was no adverse interaction due to a design basis event
,
within' containment that would cause a pressure transient that could-fail
equipment needed to mitigate the transient. The design criteria document
did not identify this commitment or the resultant implementation of
physical separation requirements for this system.
Observation No. 6.7 - Oil Free Compressed Air Requirement
TVA commitment / requirement SQN WES RMM 1110 identified a Westinghouse technical
bulletin for solenoid pilot valves in response to NRC IE Bulletin 80-11. In
section 4-9 of the Westinghouse SIP /10-1 manual, a requirement was stated in
the Westinghouse document that the compressed air system must be oil free.
This requirement was omitted from the design criteria document because a
satisfactory definition of " oil free" could not be developed in response to an
internal TVA review comment. No technical justification was provided that use
of a 5 micron filter would suffice to satisfy the oil free requirement.
Observation No. 6.8
, Single Failure Design Criteria
The team reviewed the single failure criteria document, SQN-DC-V-2.16,
dated 7/14/86, with respect to the commitment / requirement data base output.
Three types of concerns . ere noted by the team; (1) the collection of single
w
failure criteria from various sources has led to a loss of clarity regarding
certain constraints for application to both fluid and electric systems; (2) the
~
resolution process used during review of the document prior to issuance has
produced some definitions that are no longer sufficient, and (3) some specific
single failure criteria have not been captured in design documents from the
commitment / requirement data base output.
Thesinglefailurebesigncriteriadocumentdidnotaddresseachoftheinitial
assumptions of: (1) a loss of offsite power, (2) a design basis event, and (3)
the consequential failures that are a direct result of that event prior to the
application of an individual single failure.
The design criteria identified the possibility of passive component failure in
electric systems as stated by the definition section in 10 CFR 50 Appendix A.
However, the concept of passive failure has been applied only to mechanical
fluid systems where certain failures are not postulated during the short-term
period.
Industry practice has been that postulated failures in electric
systems make no distinction between short-term and long-term. periods, and are
generally identified as active failures to minimize confusion.
When an undetectable failure exists that cannot be detected by periodic
surveillance tests, IEEE Standard 379-1977, endorsed by Regulatory Guide 1.53,
states.that the preferred course of action is to redesign the system or the
test scheme to make the failure detectable. The preferred course of action
outlined in the TVA design criteria was to revise the test scheme to eliminate
the undetectable condition, and if this could not be achieved, then a redesign
may be considered. The team considered that TVA design criteria is too
permissive since it appears to encourage the retention of undetectable failures
in the system design.
The TVA design criteria's definition of. independence is "t'he freedom from
effect of one train of equipment on another train." This definition does not
A-7-
,
-
_ - -
.
.
.
-
'
Attachment A - Inspection Activities and Observations
C/R SQN MEB LWB 1060 identified physical separation requirements for air
headers inside containment in response to NUREG 0737. A commitment was
made that there was no adverse interaction due to a design basis event
within containment that would cause a pressure transient that could fail
equipment needed to mitigate the transient. The design criteria document
did not identify this commitment or the resultant implementation of
physical separation requirements for this system.
b
Observation No. 6.7 - Oil Free Compressed Air Requirement
TVA commitment / requirement SQN WES RMM 1110 identified a Westinghouse technical
bulletin for solenoid pilot valves in response to NRC IE Bulletin 80-11. In
section 4-9 of the Westinghouse SIP /10-1 manual, a requirement was stated in
the Westinghouse document that the compressed air system must be oil free.
This requirement was omitted from the design criteria document because a
satisfactory definition of " oil free" could not be developed in response to an
internal TVA review comment.
No technical justification was provided that use
of a 5 micron filter would suffice to satisfy the oil free requirement.
Observation No. 6.8 - Single Failure Design Criteria
The team reviewed the single failure criteria document, SQN-DC-V-2.16,
dated 7/14/86, with respect to the commitment / requirement data base output.
Three types of concerns were noted by the team; (1) the collection of single
failure criteria from v'arious sources has led to a loss of clarity regarding
certain constraints for application to both fluid and electric systems; (2) the
resolution process used during review of the document prior to issuance has
produced some definitions ,that are no longer sufficient, and (3) some specific
single failure criteria have not been captured in design documents from the
commitment / requirement data base output.
The single failure design criteria document did not address each of the initial
assumptions of: (1) a loss of offsite power, (2) a design basis event, and (3)
the consequential failures that are a direct result of that event prior to the
application of an individual single failure.
The design criteria identified the possibility of passive component failure in
electric systems as stated by the definition section in 10 CFR 50 Appendix A.
However, the concept of passive failure has been applied only to mechanical
fluid systems where certain failures are not postulated during the short-term
period.
Industry practice has been that postulated failures in electric
systems make no distinction between short-term and long-term periods, and are
generally identified as active failures to minimize confusion.
When an undetectable failure exists that cannot be detected by periodic
surveillance tests, IEEE Standard 379-1977, endorsed by Regulatory Guide 1.53,
states that the preferred course of action is to redesign the system or the
test scheme to make the failure detectable. The preferred course of action
outlined in the TVA design criteria was to revise the test scheme to eliminate
the undetectable condition, and if this could not be achieved, then a redesign
may be considered. The team considered that TVA design criteria is too
permissive since it appears to encourage the retention of undetectable failures
in the system design.
The TVA design criteria's definition of independence is "the freedom from
effect of one train of equipment on another train." This definition does not
A-s
Atitachment A - Inspection Activities and Obssrvations
~
~
~~
.
)
. Observation No. 7.2 - Comitments/ Requirements Related to Drilled in Anchors
TVA engineers reviewed the computer listing of the comitments/ requirement data
base related to all systems.
In this review they determine whether a
comitment/ requirement should be included in the restart phase of the D8VP.
TVA excluded most of the comitments/ requirements which relate to drilled-in
anchors and placing supports on block walls. TVA's basis for not revising the
design criteria prior to restart was that corrective actions to check adequacy
in this area were being independently pursued by a significant condition
J
report. The NRC team believes that these comitments/ requirements should be
,
re-evaluated to determine whether they should be included in the design
4
criteria for the restart phase because they relate to TVA design standard
DS-C1.7.1 which was recently revised to cover interim criteria.
Observation No. 7.3 - Revision of Design Criteria for Restart
i
Team review of comitment no. SQNCEB-CG1170 from an "all systems" comitment/
requirement data base computer listing shows that this item was included in the
restart phase and that a revision to design criteria SQN-DC-V-1.1.1 ~was
i
necessary. The comitment addresses attachments to reinforced masonry block
walls, which are only to be made with through wall bolts.
No revision to the
design crite'rf a has been made and there was no schedule to revise this design
criteria before restart to incorporate the C/R. TVA should reexamine the
incorporation of C/Rs prior to restart in the Civil Engineering Branch area.
>
.
8.
LICENSING
.
,
.
In the licensing area, theiteam focused on the review of TVA newly developed
i
interim design criteria and related engineering design evaluation, principally
in the Civil Engineering area (Civil / Structural and Mechanical Components
disciplines in this report). The acceptability of TVA's interim acceptance
criteria is still under review by the NRC's Office of Nuclear Reactor Regula-
tion. One day was spent at the Sequoyah plant site, where a walkdown of cable
tray and piping systems in containment was conducted.
On July 23, 1986, the NRR team members conducted a walkdown at the Sequoyah
Nuclear Plant. The team was aware of extensive reanalysis of small bore piping
systems currently underway at Sequoyah due to several substantive technical
issues identified by TVA in their SCR/NCR restart activities. Accordingly, the
j
purpose of the walkdown was to gain a better understanding of the technical
issues by viewing specific examples of those designs where the concerns are
evident. The team also observed HVAC support designs and typical duct span
lengths. The team walkdcwn primarily included areas inside reactor containment
"
i
and in the auxiliary building.
,
Four observations were identified in the licensing area relating to anchor
,
point movement loads, incorporation of 10 CFR 50 General Design Criteria (GDC)
requirements in TVA's design criteria and the results of the site walkdown of
>
cable tray, piping, and HVAC systems (Observations Nos. 8.1, 8.2, 8.3 and 8.4).
'
All observations stated below constitute open licensing issues that will be
,
independently resolved by NRR and the licensee. Consequently, these are
considered closed for the purpose of this inspection.
,
,
4
A-9
i
-n-
.
w..,-
- , ,, ,,
, , , , - - ~ ~
n,,----
- - , , . - - - - - - - - - - . - - . -
--,-,--,--..,,w.~,-
- , - - - - - -
- - - -
- - _ .
._
=_
.
.
-.
..
_
'
AttacNmentA-InspectionActivitiesandObservations
,
.
Observation No. 8.1 - Anchor Point Movement Loads
The team found that anchor point movement (APM) loads associated with a design
basis accident (double-ended guillotine break in the reactor coolant loop) were
!
included in the Watts Bar Design Criteria WB-DC-40.31.9 (previously used for,
the design of Sequoyah supports) but were not included in the new Sequoyah
Design Criteria SQN-DC-V-24.1. The licensee's response to the team's
observation was that DBA anchor point movement loads were not a licensing issue
!
at Sequoyah. Primary system movements at branch lines resulting from the~ DBA
loading were found to be less than 1/4 inch.
Because the APM results in a
secondary stress in the attached branch lines in a plant faulted condition, the
licensee concluded that the ASME Code does not require its evaluation.
4
'
According to the licensee, this was a Westinghouse-NRC negotiated position.
Furthermore, according to the licensee, this load case does not occur under the
Leak-Before-Break (LBB) position recently approved by the NRC.
Although secondary stresses in the piping due to plant faulted conditions (e.g.
SSE loading) are generally neglected because they are not expected to cause
'
gross structural failure due to local yielding and minor distortions in the '
,
pipe, the -licensee should provide the documentation of the basis for' concluding
that the anchor point movements associated with a DBA event are sufficiently
t
small to produce secondary, self-limiting type stresses. Under the Leak-
Before-Break position recently established by the NRC, the issue of anchor
point movement associated with a DBA becomes moot. However, the licensee must
ensure that the associated requirements established in the LBB position are
adequately implomented in its plant. The NRC office of Nuclear Reactor
Regulation sent a letter to TVA (S.A. White from J. Youngblood, dated September
29,1986), requesting additional information relating to the treatment of
anchor point movement loads for Sequoyah.
i
1
Observation No. 8.2 - Conformance to GDC for Containment Isolation
Appendix A to Design Criteria SQN-DC-V-2-15, " Containment Isolation System,"
,
'
states that the Sequoyah design was recently questioned by an NRC inspector who
cited the utility for noncompliance with the General Design Criteria (GDC).
Because the basis for the original design criteria for containment isolation
was AEC criterion 53, TVA referenced this in SQN-DC-V-2.15 and intends .to
remove the GDC reference in the Final Safety Analysis Report and replace it
j
with a reference to AEC Criterion 53. Since the Sequoyah operating license
application date (December 9,1973) falls more than six months after the
effective date of the GDC (May 21,1971), the plant must meet the GDC;
!
therefore, the licensee cannot change the references, but rather must demon-
l
strate compliance with the GDC.
Observation No. 8.3 - Cable Tray Systems
!
The following concerns were identified by the team relating to cable tray
support systems.
(1) Review of cross sections of cable trays identified cable masses outside
,
l
the trays.
Consideration should be given to securing the cables and to
-
evaluating resulting eccentricities. An example of this condition was
found in the plant at Elevation 714' and Coordinates A12S.
'
(2) A few locations in the cable tray systems appeared to contain large unsup-
ported spans. These locations were identified to technical personnel
A-10
- --- -
- -
. .
-
- - - -
- -
..-- - - - - ,
. _ - -
.
.
._
Attachment A - Inspection Activities and Observations
-
involved in the evaluation of these systems accompanying the team.
,
This situation was noted at elevation 714 and coordinates A1Q and A2Q,
at elevation 734 at unspecified coordinates, and at elevation 685 in the
auxiliary relay room.
(3) Cables were observed dropping from conduits and cable trays to other cable
trays in large bundles without significant support along the vertical drop
span. This was observed particularly in the cable spreading room, but
also at other locations'. TVA indicated this situation was being evaluated.
Results of the analytical and testing programs will be reviewed by the
NRC.
(4) Other concerns identified during the site visit were identified by TVA
personnel as currently being evaluated. These concerns include the
following:
Supports located near free edges and connections near edges of support
plates (Action Item 86-07).
Punch-shear problems between large vertical support tubes and lateral
support tubing.
Correction of weld connections between main and secondary structural
members.
.
(5) Various team discussions with TVA technical staff did not clarify how
TVA would assure that all employee concerns applicable to the interim
criteria would be evaluated.
l
The NRC Office of Nuc, lear Reactor Regulation is evaluating cable tray support
systems at Sequoyah.f A request for additional information was sent to TVA
(Mr. S.A. White from J. Youngblood, dated August 28,1986) on this subject.
Observation No. 8.4 - Piping and HVAC Systems
For small bore piping inside containment, the NRC observed several large,
extended motor-operators on valves which were unrestrained. These large
motor-operated valves were found to be extremely flexible and sensitive to
dynamic excitations. This generic issue was previously identified by TVA in a
recent SCR (SCRSQNCEB8614). TVA found that the effect of torsion due to
seismic acceleration of large eccentric masses (e.g., valve operators) was at
times neglected. The safety implications were that excessive pipe stresses and
pipe support loads could result. Additionally,~ excessive displacements could
potentially impair the operability of the valve and cause damage to adjacent,
sensitive equipment.
In 1979, a program for review of unsupported valve
operators resulted in the correction of several deficiencies in this area.
However, as-documented in the above noted SCR, some problems may still exist.
Motor-operated and pneumatic valves will be reviewed in the alternate program
and deficiencies will be evaluated and corrected.
The second issue observed by the NRC relates to non-seismic piping inside
containment. The team observed a large portion of a small bore piping system
which was not supported in the horizontal direction.
The licensee believed
the line to be non-seismic and, thus, should be addressed in the seismic (II/I)
interaction program. There currently appear to be at least two programs in
place at TVA which may address this potential issue.
A-11
,
.
- - _
__
_ _
,
Attachment A - Insp:ction Activities and Observations
Alternately analyzed Category I (L) type non-ANS safety class line (those
'
performing a secondary safety function) are being evaluated by the TVA
mechanical discipline through a contract with EQE Inc., which performs
walkdowns to assess the seismic integrity of the piping system.
Secondly, the staff is aware of a Sequoyah probabilistic seismicity
calculation (Analysis No. SQNNAL5-003) and an evaluation by Stevenson &
Associates which calculated the probability of a seismic event inducing a
failure in a non-seismically designed pipe.
This observation was identified to document and track resolution of these
~
issues. The NRC Office of Nuclear Reactor Regulation is evaluating TVA's
alternate analysis programs and interim acceptance criteria and will follow
resolution of these issues.
1
i
f
6
A-12
--.
.-_
__ ___
. _ _ , _ -
.- _ _ , _
1
,
Attichm:nt B - M:etings and References
.
.
1.
Meetings
~
Inspection activities were conducted at the DNE offices in Knoxville,
Entrance and exit meetings were held to discuss the inspection
plans and findings, respectively.
The following describes the general purpose
of these meetings.
Table B.1 is provided as a matrix of meeting attendance and
principal persons contacted. Other licensee personnel were also contacted.
Meeting 1:
On July 21, 1986, an entrance meeting was held at the DNE offices
in~ Knoxville.
The NRC explained the plans for the assessment of TVA's program
for design criteria preparation and the associated Engineering Assurance
oversight.
Meeting 2: On July 25, 1986 an exit meeting was held at the DNE offices in
Knoxville. The scope and findings of the inspection were discussed.
The team
members presented the more significant findings within each discipline.
Table A.1 MEETINGS
Name
Organization
Title
Meeting Attended
1
2
REArchitzel
USNRC-IE
Team Leader
x
x
SVAthavale
USNRC-IE
NRC-Electric Power
x
x
PEHarmon
USNRC-RII
Resident Insp.,5QN
x
ADuBouchet
NRC-Consultant
NRC-Mech. Components
x
x
FJMollerus
NRC-Consultant
NRC-Mech. Systems
x
x
AIUnsal
NRC-Consultant
NRC-Civil / Structural
x
x
JMLeivo
NRC-Consultant
NRC-Nuclear System
x
x
LStanley
NRC-Consultant
NRC-Instr./ Controls
x
x
WCDrotleff
<TVA-DNE
Dir. DNE
x
x
JEHuston
TVA-DNQA
Dep. Dir. Nuc. QA
JFWeinhold
TVA-DNE
EA Manager
x
APCappozzi
TVA/S&W
Consultant - EA
x
x
MP8erardi
TVA-EA
x
x
AWLatti
TVA-DNE
Manager DBVP
x
x
RPSvarney
TVA-EA
Civil / Structural Engr.
x
x
APagano
TVA-EEB
EEB Asst. BC
x
JPLittle
TVA-MEB
Supervisor
x
JPDurnhan
Impell
Consultant
x
JFCox
TVA-DNE
Ast. PE SQEP-K
x
x
MJScruggs
TVA-DNE
Elec. Engineer SQEP-K
x
x
CFBowman
TVA-DNE
DPB Mgr.
x
JJSas
TVA-DNE
Dep. Director DNE
x
JARaulston
TVA-DNE
Chief Nuc. Engr.
x
FAKoontz
TVA-NEB
Gr. Head T/H & Plant Sup.x
CWParker
TVA-NEB
Nuc. Engr.
x
JJWilder
TVA-NEB
Nuc. Engr.
x
BHall
Licensing - Sequoyah
x
x
GRReed
TVA-DNE
Elec. Engr. - DBTF
x
RL0lberding
TVA-DNE
Mech. Engr.
x
RCWilliams
Reg. Rep.
x
DLWilliams
Nuc. Engr.
x
'
,
B-1
--
1
.
Atttchment B - Meetings and References
.
Table A.1 MEETINGS (cont.)
Name
Organization
Title
Meeting Attended
1
2
JWilliams
Staff Specialist
x
FRinaldi
NRC-NRR
Structural Engr.
x
x
DTerao
NRC-NRR
Mechanical Engr.
x
x
JHolonich
NRC-NRR
Project Mgr.
x
2.
REFERENCES
1.
TVA Calculation SQN-0567-048, Revision 1 dated 5/15/86,
" Identification of Systems Required for Restart."
" Design Ba~eline and Verification Program, Sequoyah Nuclear Plant,"
2.
s
Revision 0 dated 5/1/86.
3.
TVA Design Criteria SQN-DC-V-27.8 and " Status of C/R Review" for
" Neutron Monitoring System," transmitted by Westinghouse letter dated
7/16/86 to TVA for information and use; not yet issued by TVA.
4.
TVA Design Criteria SQN-DC-V-26.1 Revision 0 dated 7/11/86,
,
" Combustible Gas Control System."
5.
TVA Design Criteria SQN-DC-V-2.15 Revision 0, " Containment Isolation
System."
6.
TVA Design Criteria SQN-DC-2.16 Revision 0, " Single Failure."
7.
TVA Design Criteria SQN-DC-2.17 Revision 0, " Remote Shutdown Criteria
from Locations Outside the Main Control Room."
8.
TVA letter 11/2/84, Mills to NRR, topic:
Hydrogen Igniter Spray
Shield Design.
9.
Review Plan No. 4100R2, Tennessee Valley Authority, Sequoyah Nuclear
Plant, Engineering Assurance Oversight Review Plan, Mechanical,
July 23, 1986.
'
B-2
l
l
_
,
. _ _ _ .
..
_.
r-
_-
-
-
,
%%g
Mr. C. C. Mason
-4-
ggyg
i
Distribution (w/ encl)
'
LPDR
DQAVT Reading
QAB Reading
RArchitzel, IE
LSpessard, IE
EVImbro,-IE
HJMiller, IE
i
BKGrimes, IE
!
JMTaylor, IE
.
'
RWStarostecki, IE
,
HRDenton, NRR
.
GZech, RII
l
KBarr, RII
BBHayes, 01
BDebbs, RII
SRConnelly,-OIA
ELJordan, IE
i
JYoungblood, NRR
HThompson, NRR
,
l
DMuller, NRR
'
Inspection Team (8)
JHolonich, NRR
i
Resident Inspector
?
j
Regional Administrator, RII
1
CRStahle, NRR
-
TMNovak, NRR
!
i
ELD
I
i
i
!
I
j
D
S
Wh
l
IE:0QAVT:QAB
IE:0QAVT:QAB
IE:DQAVT:,QAB:C'
IE:
- DD
j
REArchitzel
EVImbro
HMiller
ler
10/)//86,
10/q//86
10/11*/'/86
0/ ./86
,
j
Q
IE:h
$(.
IE:DD
I
1
s
'LSpsssard
RWStarostecki
JM aylor
j
Op/86
10/gf/86
10/9/86
b
10/) /86
,
i
i
I
_ - _ . _ . . _ .
. _ _ _ _ ~ . , ,
.. .__ _ _ _
.-_._ .._ .,___ _ . _ _,_, _ _ ,_._ . _ _.-_ ___- _ -._
.
Mr. C. C. Mason
-4-
g
g
Distribution (w/ encl)
,
LPDR
DQAVT Reading
QAB Reading
RArchitzel, IE
LSpessard, IE
EVImbro, IE
HJMiller, IE
BKGrimes, IE
JMTaylor, IE
RWStarostecki, IE
HRDenton, NRR
l
GZech, RII
KBarr, RII
l
BBHayes, 0I
___
(
BDebbs, RII
SRConnelly, 0IA
ELJordan, IE
JYoungblood, NRR
l
HThompson, NRR
!
DMuller, NRR
-'
,
Inspection Team (8)
-
.
-
JHolonich, NRR
Resident Inspector
-
Regional Administrator, RII
CRStahle, NRR
TMNovak, NRR
ELD
.-
WL -
?,t
IE:DQAVT:QAB
IE:DQAVT:QAB
IE:DQAVT:'QAB:C
IE:D
- DD
REArchitzel
EVImbro
HMiller
H ' ler
lof//86
10A'q '/86
10/J/86
0/ /86
.
QAJ .D
IE:lig:D
$gb
g
IE:DD
I
'
s
LSpfssard
RWStarostecki
JM a lor
Ohg/86
10(gf/86
10/F/86
10/){86
__