ML20210E009
| ML20210E009 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 01/02/1987 |
| From: | Imbro E, Parkhill R, Sinkule M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II), NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| To: | |
| Shared Package | |
| ML20210C958 | List: |
| References | |
| 50-424-86-128, NUDOCS 8702100211 | |
| Download: ML20210E009 (37) | |
See also: IR 05000424/1986128
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Report No.: 50-424/86-128 .
Licensee: Georgia Power Company
P.O.; Box 4545 %
Atlanta, GA 30302
-Docket No.: 50-424- Construction Permit No.: CPPR-108
Facility: Vogtle, Unit 1 Module: Module No. 22, liidependent
Design Review
Reviews Conducted: -April 15-19, 1985; July 1-2, 1985, July 24-Aug.2, 1985;
August 11-15, 1986; and October 6-9, 1986
On-Site Inspections Conducted: Weeks of April 16, 1985 and August 15, 1986 ,
NRC Offices Participating in Inspections / Reviews: ;
Office of Inspection and Enforcement (IE, Bethesda, MD)
Reviewers: E. Imbro, QA Branch, IE
T. McLellan, Reactor Construction Branch,-IE ,
T. DelGaizo, Consultant (WESTEC Services)
G. Morris, Consultant (WESTEC Services)
J. Blackman, Consultant (WESTEC Services)
J. Kaucher, Consultant (WESTEC Services)
J. Leivo, Consultant (Leivo Associates)
G. Harstead, Consultant (Harstead Engineering)
E. Willhaus, Consultant (Harstead Engineering)
Prepared by: /[m J / P/'Ae [h
R. Parkhill, Team Leader
_ N/#4 /Pt.
Date Sf'gned
Inspection Specialist, IE j
Approved by: 8[ w # A2 /M'/
E. Imbro, Section Chief .Date Signed
Licensing Section, QA Branch, IE
M. Sinkule, Chief
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Date " Signed' -
Projects Section 3C
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Division of Reactor Projects, Region II
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V0GTLE ELECTRIC GENERATING PLANT UNIT 1 I
READINESS REVIEW PROGRAM
MODULE N0. 22
INDEPENDENT DESIGN REVIEW
_ CONTENTS
Topic Page
S u mma ry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
Purpose and Scope of Review.............................. 4
Methodology.............................................. 5
N RC S ta f f Eval ua ti on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 j
- General.................................... 7
Structural................................. 8
Hazards.................................... 9
Mechanical Analysis........................ 11
Control Systems............................ 15
Electrical................................. 22
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System Design.............................. 26
Generic Concern............................ 30
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! Statement of Module Acceptability (Conclusions).. .... . . . . 34
Li s t of Pe rsons Contac ted. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 35
Acronyms................................................. 36
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V0GTLE ELECTRIC GENERATING PLANT UNIT 1
READINESS REVIEW PROGRAM
MODULE NO. 22
INDEPENDENT DESIGN REVIEW
SUMMARY
In Module No.22 of Georgia Power Company's Readiness Review Program for
the Vogtle Electric Generating Plant, the Stone & Webster Engineering
Corporation conducted an Independent Design Review of the design of the
auxiliary feedwater system by the Bechtel Engineering Corporation. In
areas where the auxiliary feedwater system did not contain the iiecessary
design attributes to be reviewed, other plant systems / designs were selected
for review by SWEC. The NRC monitored the conduct of this module including
an inspection of the review plan preparation, inspection of review plan :
implementation, inspection of the resolution of observations, and an
inspection of implementation of corrective and preventive actions. In
this report, the NRC accepts Module 22, requiring no further evaluation j
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or followup activities in the context of this module. The resolution of
electrical power observation 22F-11 remains an open licensing issue. Based
on its review, inspection, and monitoring of the Modules, the NRC also con-
cludes that a technically adequate design process has been implemented for I
the Vogtle Plant and that substantial additional assurance of design adequacy j
has been provided.
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V0GTLE ELECTRIC GENERATING PLANT UNIT 1
READINESS REVIEW PROGRAM
MODULE NO. 22
INDEPENDENT DESIGN REVIEW
1. Purpose and Scope of Review
The purpose of this evaluation is to determine if the results of the
Independent Design Review, performed by the Stone & Webster Engineering
Corporation (SWEC) and presented in the report of Module No. 22, are an
effective and accurate assessment of (1) the state of the design of the
Vogtle Electric Generating Plant, Unit 1 (VEGP) and (2) the implementa-
tion of the design process for VEGP by the architect-engineer firm,
Bechtel Engineering Corporation (BEC). Further, the evaluation tested
the technical justification of the SWEC conclusions in Module No. 22 with
regard to the adequacy of design and effectiveness of the design process,
the appropriateness and effectiveness of corrective or preventive actions
arising from the SWEC report, and the resolution and disposition of
potential generic concerns raised by SWEC.
tileIndependentDesignReview,wasundertakenbyGeorgia
Module No. 22,(GPC) as a result of NRC concerns that the Readiness Review
Power Company
Program, as structured prior to April 1985, took a fragmented approach to
the matter of design process implementation such that intradiscipline
interface, intraorganization interface, and other common design aspects
(e.g., high-energy-line break protection, seismic /nonseismic interactions,
etc.) might not be adequately evaluated. The integrated evaluation of a
specific sample of the design (i.e. the auxiliary feedwater system aug-
mented by other ple,. design attributes) as performed in Module No. 22
resolved the NRC's concerns and provided a single document for evaluation
of the state of VEGP design, with exception of the civil / structural
discioline. Module 1 (Reinforced Concrete Structures), Module 8
(Structural Steel) and Module 13 (Foundations and Backfill, Coatings,
and Post Tensioning) address the status of the design in the civil / structural
area, except for structural interfaces with the AFW system which were
evaluated in Module 22. The design aspects of Module 4 (Mechanical
Equipment and Piping), Module 6 (Electrical Equipment), and Module 16
(NSSS Interface), were integrated into Module 22.
In view of the above, the NRC's review of Module 22 constitutes an overall
review of the design and design process implementation in the mechanical
systems, mechanical components, electrical, and the instrumentation and
controls disciplines, along with common design areas such as high- and
moderate-energy-line break, seismic II/I, flooding, internally generated
missiles, fire protection, and environmental qualification. While this
report does not address the results of the civil / structural modules
(Modules 1, 8, and 13), which have been reported separately, the conclusions
of Section 4 of this report do consider the results of these modules in
drawing overall conclusions as to the adequacy of the design of '/EGP and
the effectiveness of implementation of the design process for the plant.
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2. Methodology
The NRC approached the review and evaluation of Module No. 22 in a manner
similar to monitoring of Independent Design Verification Programs (IDVPs)
at other NTOL facilities. Specifically, the NRC's inspection program
involved three phases: (1) inspection of design review checklists, (2)
inspection of implementation of the IDR plan (including an assessment of
the technical depth of the review), (3) inspection of the technical docu-
mentation supporting the IDR conclusions including justification of
observation resolutions between the IDR reviewer (SWEC) and the architect-
engineer (BEC), and inspection of the effectiveness of implemented
- corrective or preventive actions (subsequent to closecut of the items by
GPC).
As can be seen in Figure 1, the NRC inspections were performed as follows:
Inspection Location Dates
1. Review Plans SWEC, Boston 7/1-7/2/85
2. Implementation BEC, Los Angeles 7/29-8/2/85
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3. Resolution of VEGP Site 8/11-8/15/86
i Observations and BEC, Los Angeles 10/6-10/9/86
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Corrective Action
Figure 1 contains information on the civil / structural modules for purposes
of showing the complete scope of the design review activities of the
, Readiness Review Program and since some NRC inspections involved Module 22
review along with review of one or more of the civil / structural modules.
This report addresses only Module 22 (and those portions of Modules 4, 6,
and 16 which were incorporated in Module 22), and therefore the civil /
structural information is provided for information only to give a total
overview of the Vogtle design process.
In addition, the scope and depth of the design reviews being performed
in conjunction with Modules 4 and 16 were evaluated at the VEGP Site from
April 15 through April 19, 1985. Since these observations were eventually
merged with the Module 22 findings, the resolution of observations and
effectiveness of corrective action associated with these items were
reviewed by the NRC during the appropriate Module 22 inspections cited
above.
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In each of the above inspections NRC inspectors, with contractor support,
reviewed the design disciplines of mechanical systems, mechanical components,
electrical power, and instrumentation and control. In addition, at each
inspection, the action taken to resolve NRC concerns from previous inspec-
tions was reviewed and evaluated. A report was prepared for each inspection
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l and prior NRC inspection concerns were closed out in subsequent reports,
provided appropriate action to close the concern had been observed.
Otherwise, these items were carried as open into the next NRC inspection.
The NRC inspection reports associated with the Module 22 review are as
follows:
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V0GTLE READINESS
REVIEW PROGRAM
DESIGN VERIFICATION
CIVIL / STRUCTURAL MECH. EQUIPMENT / SYSTEM-BASED
REVIEW PIPING REVIEW TECHNICAL ASSESSMENT
if If if
- , . MODULES 1,8 , MODULES 4,6 SWEC IDR
AND 13 AND 16 (CH.7) MODULE 22
(AUX. FEED REVIEW)
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APRIL, 1ses II
yj,gseecism. REVIEW PLAN
Jutv. ises PREPARATION
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1MPLEMENTATION
OF PLAN
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BPC/5WEC ACTION- d$rllS$ii['"
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BPC/SWEC ACTION-
ITEM RESOLUTIONS
ITEM RESOLUTIONS
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BEC CORRECTIVE e
ACTION
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READINESS REVIEW *
VERIFICATION OF
FIGURE ] CORRECTIVE ACTION
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Inspection Date Report No.
7/01/85-7/02/85 50-424/85-34R
7/29/85-8/02/85 50-424/85-34R
8/11/86-8/15/86 50-424/86-105R
10/6/86-10/9/86 50-424/86-128R
3. NRC Staff Evaluations
An evaluation of each Section of the SWEC IDR report (Module 22) is
provided in this section, using section numbers corresponding to the
IDR report. Included are a brief description of the section, what was
reviewed, the basis for acceptance, and a statement of any required
follow-up action or follow-up evaluation.
a. Section 1 - Introduction
(1) IDR Report
This section is a general introductory section including background
information on the program and general information as to the content
of the report.
(2) Inspection Results
This section was reviewed for background information only. No followup
action or evaluation of this section is required.
b. Section 2 - Scope
(1) IDR Report
This section provides the basis for selection of the Auxiliary
Feedwater System and other plant design aspects as being a repre-
sentative sample for an independent design review and further
describes the scope of the review.
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(2) NRC Inspection Results
The NRC staff agrees with the rationale for selection of the
Auxiliary Feedwater System and other plant design aspects as being
a representative sample of the design. Further, the staff concurred
with the scope of the review at the time of inspection of the initial
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design review plan and checklists during the inspection at SWEC
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Headquarters (Boston) in July 1985. Accordingly, the staff finds
this section of the report acceptable and no followup action or
evaluttion is required.
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c. Section 3 - Results and General Conclusions
This section of the IDR report contains the SWEC conclusions for
each of the major sections of the report as follows:
Section Subject
3.1 General
3.2 Structural
3.3 Hazards
3.4 Mechanical Analysis
3.5 Control Systems
3.6 Electrical Design
3.7 System Design
3.8 HVAC System Design
3.9 HVAC Duct Design and Analysis
3.10 Evaluation of Generic /
Programmatic Concerns
This section is the heart of the report. Each of these sections
is assessed by the NRC staff separately in the following material.
Since these subsections cannot be assessed without reference to
the appropriate subsets of section 4 (Review Description) and
section 5 (Review Observations, Responses, Assessments), the
respective subsets of sections 4 and 5 are considered with the
section 3 material in this report.
c.1 Section 3.1 - General
(1) IDR Report
In this subsection, SWEC draws the following general conclusion:
"In conclusion, certain areas of weakness were identified.
These areas were primarily with respect to the adequacy of
the project documentation and program implementation. In
a few instances this led to isolated design deficiencies
and in some cases hardware changes. Due to the judicious
application of design margin based on past engineering
experience during design development, the extent of these
changes was limited. The current programs and corrective
actions, as indicated in Sections 3.10 and 5 of this report
and in Section 7 of each of the civil / structural module
reports, provide reasonable assurance that the final plant
design will be technically adequate and in compliance with
licensing commitments."
(2) NRC Inspection Results
Overall NRC staff conclusions are presented in Section 4 of
this report, Statement of Module Acceptability. Since overall
staff conclusions are presented in detail in Section 4, they
are not repeated in this section. For the reasons given in
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Section 4, the staff finds this subsection of the report to be
acceptable. No followup action or evaluation of ini subsection
is required.
c.2 Section 3.2 - Structural (SBTA)
Section 4.2 - Structural
Section 5.A - Structural Observations
(1) IDR Report
Since other Readiness Review Modules (e.g. Modules 1,8, and 13)
addressed VEGP structural design, Module 22 focused on structural
interfaces with the AFW system, design of raceways and supports,
and structural interfaces with pipe supports, raceway supports,
and equipment supports. The IDR team concluded that this portion
of the design was adequate and met licensing commitments and
regulatory requirements.
(2) NRC Inspection Results
The structural assessment by SWEC consisted of the examination
of two design criteria, two specifications, fourteen calculations
and various other design documents. The NRC's assessment of the
observations resulting from this review are presented below:
Observation 22-A6 indicated that for FCRs relating to conduit
support systems span changes, only the effect of the change on
the support was addressed and not the other components such as
the conduit, fittings and clamps. As a result of this observation,
BEC performed a review of past FCRs associated with increased
conduit spans and determined that although the technical justi-
fication documentation for the changes was not complete, structural
evaluation of the changes indicated that the modified arrangements
were structurally adequate. The twelve FCRs identified by the IDR
were revised to include all required technical justification.
Instructions were issued to clarify the requirements of providing
technical justification for concurring with FCRs. In addition,
training of personnel was conducted as part of the implementation
effort. The NRC has reviewed the corrective actions taken,
performed a review of a sampling of FCR written af ter imple-
mentation of the corrective action program and concludes that
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l the matter has been satisfactorily resolved. This item is closed.
The remaining observations in the structural area are not
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individually discussed because they were of lesser technical
significance and were satisfactorily resolved. These observations
were reviewed by the staff and all dealt with cases where
engineering judgment in lieu of documenting the details of var-
ious designs had occurred. The staff reviewed the corroborating
details and concluded that no governing design criteria or any
project licensing commitments had been violated. All of the
observations were resolved by BEC by supplying the missing
corroborating information. The staff therefore finds this section
I of the report acceptable. No followup action or evaluation is
required.
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. c.3 Section 3.3 - Hazards
Section 4.3 - Hazards
4 Section 5.B - Hazards'
(1) IDR Report
This subsection of the report concludes that the hazards area
- was found to be technically acceptable. SWEC stated that
project actions initiated in response to hazards observations
' were assessed as adequately addressing resolution of .the-
concerns. SWEC further stated that project commitments to
revise or generate appropriate procedural. requirements,
evaluation sheets, and calculations.were adequate to ensure
safety system functional integrity during postulated in-plant
hazards.
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(2) NRC Inspection Results
The hazards assessment by SWEC included an examination of
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approximately: 23 design criteria and project technical
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guides; 4 topical reports; 24 calculations; 33 drawings; 13
specifications; 31 intradiscipline interface documents; and 5
vendor interface documents. The NRC's assessment of the
.! observations resulting from this review are presented below:
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Observation 22-B01. identified an incorrect implementation of
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high-energy-line break criteria, resulting in hardware modi-
fications to maintain safe shutdown capability. Specifically,
longitudinal breaks had not been postulated at intermediate
high stress points for high energy ASME III Class 2 and 3 piping.
Subsequent postulation of longitudinal breaks and evaluation of
consequences on essential system and component targets resulted
in design changes to assure safe shutdown capability. As a
result of this observation, the Licensee reported the event in
accordance with 10 CFR 50.55e. The NRC staff reviewed the
resolution of the observation and concurs that appropriate
corrective action has been taken and considers the specific
technical issue to be closed. With regard to the generic
- - implications of a project procedure which failed to correctly
implement design criteria, this matter is addressed by the IDR
in Section 3.10 and is assessed by the staff in the appropriate
section of this report.
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Observation 22-B02 indicated that due to the unfinished status
of the design with regard to documentation of protection against
internally generated missiles, the IDR reviewer was unable to
adequately assess project activities in this area. Specifically,
at the time of the IDR, missile postulation calculations were
undergoing a general revision. During the NRC's inspection at
BEC in October 1986, the staff reviewed the Summary Report of
the VEGP Internally Generated Missile Analysis and calculation
X6CXD-25 (Missile Analysis-Rotating Equipment). The staff
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concluded that a technically adequate analysis for protecticn
against internally generated missiles is in progress (nearly
complete) and is being sufficiently documented. Confirmation
by the Licensee that the analysis was completed is documented
in P. D. Rice's letter to Region II dated November 20, 1986.
This observation is closed.
Observation 22-B03 indicated that maintenance of safe shutdown
capability was not adequately demonstrated in the documentation
of analysis of internal flooding. Specifically, the observation
identified inadequate flooding effects analysis regarding' the
consideration of flooding sources and effects beyond the volume
under evaluation, the procedures and tools used to identify
potential flood flow paths, and consideration of electrical
consequences beyond the flooded volume. The project subsequently
revised the hazards walkdown instructions to assure proper con-
sideration for other plant areas. During the NRC's inspection
at BEC in October,1986, the staff reviewed the hazards walkdown
documentation as well as two Auxiliary Building flooding calcu-
lations (X6CXC-26 and X6CXC-30). The staff concluded that IDR
concerns are being fully addressed and that internal flooding
effects are being sufficiently analyzed. Confirmation by the
Licensee that the internal flooding portion of the hazards
finalization program is to be completed prior to fuel load is
documented in P. D. Rice's letter to Region II dated November 20,
1986. This observation is closed.
Observation 22-804 identified various instances where project
documentation failed to identify the source of design data or
present technical justification for assumptions. The project
subsequently demonstrated that technical deficiencies did not
result from the examples cited in this observation with excep-
tion of the longitudinal breaks discussed in Observation 22-B01
(above). The staff reviewed a sample of these items and con-
siders this observation closed. The generic issue of hazards
documentation is discussed in Section 3.c.8 of this report.
Observation 22-B05 concerned an unjustified assumption that
- non-nuclear safety-related items supplied to standard com-
mercial practice provided adequate design margin to withstand
seismic loadings of an SSE without structural collapse. As an
example of this assumption, the IDR identified room space heaters
which could impact safety-related equipment should they collapse
in a seismic event. The project committed to upgrade the support
of these heaters and comitted to adequately consider seismic
capability of commercial grade equipment as part of the hazards
program. Subsequent to upgrading the supports of all wall-
mounted heaters having potentially unacceptable interactions
with safety grade equipment, BEC performed a calculation which
concluded that the heaters would not have collapsed in a seismic
event. The staff considers this observation closed for the
following reasons: (1) the supports for the heaters in question
have been upgraded, (2) the hazards walkdown is considering
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potential. failure of category 2 equipment-(verified by the
staff during thez l0/06/86 inspection at BEC), and (3) BEC has
. concluded that the heaters would not have collapsed even prior
to the upgrade.
Observation 4-92 identified that the. evaluation of internally
. generated missiles did not evaluate a fan blade missile escaping
through the expansion joint, such as the event that occurred at
the Palo Verde Nuclear Station. The project consnitted to an
evaluation of this missile. During the inspection at BEC the
week-of October 6,1986, the staff reviewed documentation of
the walkdown of all Joy fans in.the plant and the analysis of
potential missiles, particularly those potentially escaping
through flexible expansion joints. The evaluation concluded
that no safe-shutdown equipment would be impacted by potential
missiles. The staff considers this evaluation sufficient.to
address the concern and considers this observation closed.
The remaining observations in the hazards area are.not indi-
vidually discussed because they were of lesser technical
significance and were satisfactorily resolved.
In view of the technical items discussed above, the NRC staff
concurs with.the IDR conclusion that design of the plant in
the area of hazards protection is technically adequate. With
regard to the_ hardware changes made as result of the possible
longitudinal pipe cracks or potential failure of commercial
. grade equipment in a seismic event, these features are very
-conservative design measures imposed on nuclear plants to
provide protection against highly ~unlikely occurrences. Al-
though these measures must be incorporated, and as a result'
of the IDR they have been _ incorporated at -VEGP, the identified
weaknesses in the design process are not considered to be sub-
stantial. With regard to the shortcomings in project documentation
in tha hazards area, these are discussed in subsection c.8 of
this report. The staff finds this subsection of the report
acceptable. No followup action or evaluation is required.
c.4 Section 3.4 - Mechanical Analysis
Section 4.4 - Mechanical Analysis
Section 5.C - Mechanical Analysis - Pipe Stress
Section 5.0 - Mechanical Analysis - Pipe / Duct Supports
' Section 5.M4 Mechanical Equipment and Piping
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(1) IDR Report
b This subsection of the report concludes that, with exception of
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-identified observations, piping analysis, pipe support designs,
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and duct support designs met technical requirements of project
licensing commitments and applicable codes and standards.
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Further, corrective and preventive measures were judged to'be
adequate.
(2) NRC Inspection Results
The pipe stress, pipe support, and duct support assessmen'ts-
by SWEC included the examinatian of-approximately: 25 design
criteria, 57 calculations, 28 specifications, 78 drawings and
69 other miscellaneous documents. The NpC's assessment of the
observations resulting from this review are presented below:
Observation 22-C5 identified the incorrect enveloping of response
spectrum in three piping analysis problems associateo with the
auxiliary feedwater system. In the first case, two separate sets
of enveloped response spectra were used; however, a multiple
response spectra analysis was not performed. In the second case,
the selected-response spectra did not-envelop all potential
responses. While in the third case, the project. performed-a
multiple response spectrum analysis which is inconsistent with
Design Criteria DC-1005 and FSAR commitments. As a result of-
this observation, BEC reviewed all high energy, Class 1 and
other analyses within its scope and corrected five analyses.
In addition, DC 1005 and the FSAR were revised to reflect the.
changes discussed above. .The NRC staff reviewed the resolution
of the observation and concurs that the appropriate corrective
actions have been taken. 'In view of'the limited number of
discrepancies found, it does not appear that any systematic .
breakdown in control has occurred. This observation is closed.
Observation 22-C10 identified a situation where a weldolet
formula contained in a Bonney Forge Handbook was used to account
for the stress intensification present at an extruded outlet tee
intersection. As a result of this observation, the project
developed justification for the use of the formula based upon
similarity to geometrics defined in figure NC-3673.2(b)-1 of
the ASME code of record for Vogtle as well as comparison to
test result performed by Bonney Forge and the Pressure Vessel
Research Council (PVRC). In addition, a broadness review was
performed to determine other cases where stress intensification
factors may have been used without strict compliance to pro.iect
documents. A very limited number of situations were found where
nonstandard stress intensification factors were used. Evaluation
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of these situations by the project indicated stress levels which
I can easily accommodate much larger stress intensification factors
than were used. The staff reviewed the relevant documentation
and concluded that the matter had been addressed adequately. In
addition, there has not been a breakdown in the design process
nor implication of safety impact. This observation is closed.
I Observation 22-C11 concerned neglecting the weight contribution
- of pipe support attachments in the piping model for stress
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analysis. The project stated that for Class 1 lines, this
factor is considered in.the analysis, while for Class 2 and
3 lines the consideration is of no practical concern. The
majority of supports used in Class 2 and 3 systems are rigid
frames or other designs which do not contribute additional
weight to the piping system. Further most systems are not
subjected to significant dynamic loadings to which such
factors are important. The project agreed, however, to
require evaluation of such effects if.the added weight
4 exceeded 10% of the piping weight for future analyses and
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as part of the'as built reconciliation program. The staff
reviewed the justification for the project's position and the
implementation of its commitments and concluded that the
i matter wds satisfactorily resolved. .This observation is closed.
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. 0bservation 22-C12 indicated that the static anchor movement
loading used in ,the RCL loop break is not consistent with
project criteria (DC-1018) and that static equivalent
analyses were performed to evaluate the effects of a LOCA
which may not be conservative. The project stated that the -
anchor movements used were an approximation and that since
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the leak before break criterion is now being used on the
!~ > project, appropriate LOCA displacements would be.used when
made available by Westinghouse. In addition it was decided
to confirm the acceptability of the static equivalent method
by performing dynamic LOCA analyses for the affected lines.
The required analyses have been performed and were reviewed
l by the staff. The results indicate that the original piping
design is adequate, therefore the matter is considered
- closed..
i- In observation 22-001, a review of all voided pipe whip
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restraints was conducted by BEC and 11 voided restraints
were identified as supporting structure for 12 pipe supports.
- - All of the pipe supports in question have been updated and
< additional personnel training in the use of action tracking
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logs was conducted by BEC. The staff reviewed the response
- assessment and 18 pipe support drawings and associated rupture
, restraint drawings to verify concurrence with deletions
identified by BEC. No discrepancies were identified. The
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staff concludes that the item is closed and the generic
l concerns are answered.
In observation 22-D02, BEC conducted a full review of all
pipe support design loads for all lines in the BEC scope and
identified 60 pipe supports with significant load deviations
out of 2,142 supports on 208 isometrics reviewed. Of these
60 supports identified with load deviations, only one required
j physical change. BEC has issued an instructional memorandum
and conducted personnel training on its provisions. The staff
,
reviewed the response and assessment of observation 22-002.
, In addition, the staff reviewed documentation for completeness
e
1 13
1
4
9,y., _ . , - 9 , , - - - -, . ,-,.,_,w-,.7-._, ,.,..y_ _._,e+ ,_.,m r_,.,,.., , _ _ . _g,, ., - . ,, . , . . . - . . - ~ . , _ . - . _
._ . .-_ - ~_ . __ _ , _ - - __ . .
.
and detail design of 10 pipe supports from SWEC's sample in.
Appendix E, " Corrective / Preventive Action Commitment" of the
. I DR .- No discrepancies were identified during.the review.of
pipe support documentation. The staff concludes from the
.
review of the response, assessment and documentation that this
observation is closed, and the generic concerns are answered.
,
Observations 22-D04 and-011 refer to undocumented use pipe
1 support design guides in calculations and undocumented use:
of computer programs in calculations, respectively. In the
'
case .of 22-D04, BEC has revised DC-1017- revision 5 to assure
that proper documentation is prepared during the reconciliation
f program. In observation 22-011. BEC reviewed 30 calculations
j for use of the correct computer program name and omissions of
f the program name and no discrepancies were identified. The
i staff reviewed the response end assessment of observation 22-D04
and 011. In addition, the NRC inspection reviewed the calcu- :
'
l
1ations for 10 pipe supports from SWEC's sample in Appendix E
i- of the IDR and did not identify any discrepancies. The staff
i concludes' from the review of. the response, assessment and the
revising of DC 1017 by Bechtel, that the observations are
closed and the generic concerns are answered.
f Observation 22-D06 identified an incorrect movement used for a
spring hanger design. BEC reviewed all 300 spring hangers in
i
, their scope for errors in thermal movements and identified four
,
cases in which errors were made. The identified errors in -
thermal movement had insignificant effect on spring size -
- selection and hot load settings. BEC revised the computerized
- load and movement sumary sheets in 1984 to include all required-
!
data in one section of the' computer printout to eliminate the
i
potential for error. The staff reviewed the response, assessment
! and the calculation for pipe support VI-1301-012-H026 and did not
j identify any discrepancies. Based on BEC's review of all the
- summary sheets for 300 springs in their scope and the revision
of the computerized summary sheets, the staff concludes that
,
j
the observation is closed and the generic concerns are answered.
.
Observation 22-D12 concerned undocumented calculations for
~
-
!
,
snubber settings. BEC stated that the undocumented snubber
L setting in pipe support calculation V1-1201-053-H005 was an .
! isolated error. To verify this was an isolated error, a
j
review of 71 snubber calculations that were_ done by both
' the originator _ and the checker cited in the original finding
was performed and an additional sample of 60 snubber
l calculations by other designers were reviewed for correct
l
.
snubber settings. BEC identified 1 out of the 131 snubbers
reviewed to have a thermal growth discrepancy and the pre-
l
liminary review of the estimated thermal stresses indicated
the displacement discrepancy would not exceed code allowables.
The staff reviewed Pipe Support Design Manual, Section 4.8.1
and 5 snubbers from the sample of 71 for correct settings and
! 14
1
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!
. ., - - . - - - - - - _ - - - . -- . - - . - - - -
.
no discrepancies were identified. The staff also reviewed
PSDM Section 4.8.1 and concluded it was acceptable for
calculating snubber settings. Based on this review, the
staff finds the resolution of this item acceptable and this
item is closed.
Observation 22-D19 concerned inconsistencies between the stress
criteria for HVAC duct supports and FSAR commitments. The IDR
noted that the stress allowables in calculation X2003 and
Section 3.4 of Design Manual DC-2167, Revision 4, were not
consistent with the commitment in FSAR Sections 1.9.52 and
1.9.140. In response to this observation, BEC did a review
of all civil / structural design criteria for consistency with
FSAR commitments, emphasizing HVAC duct supports and electrical
raceway supports. In its review, BEC concluded that the incon-
sistency with regard to the stress allowables was isolated and
did not cause any hardware changes. The staff reviewed the
documentation associated with the response and identified a
minor administrative error with regard to FSAR Table 3.2.2-2.
The Licensee subsequently issued a SAR change notice to correct
the error. Based on the staff's review of the documentation and
the issuance of the SAR change notice, the staff concluded that
this observation had been resolved adequately. This item is
closed.
The remaining observations in the mechanical analysis area
are not individually discussed because they were of lesser
technical significance and were satisfactorily resolved.
Based upon the review of all observations in the mechanical
analysis area, the staff concurs with the EA conclusion that
piping analysis and pipe support designs meet technical
requirements of project licensing commitments and applicable
codes and standards. The staff finds this section of the
report acceptable and no followup or additional evaluation
activities are required.
c.5 Section 3.5 - Control System
Section 4.5 - Control System
Section 5.E - Control System
,
l (1) IDR Report
l
'
This subsection of the report concludes that the controls
area was found to be technically acceptable. SWEC stated
that project actions initiated in response to the controls
observation adequately resolve the concerns.
!
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. . . . _
__ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ . - _ - _ _ _ _ _ _ _ _ .
. .
.
l
(2) NRC Inspection Results
There were 19 IDR observations in the instrumentation and
controls discipline. Of these, 10 had been designated Level
IV (invalid observation) by the IDR team following subsequent
clarification or review of additional information supplied by
the project. The staff reviewed the proiect responses to these
10 observations, interviewed the IDR team, and agreed that with
the clarifications as documented in the IDR, that these obser-
vations had no safety significance.
The following staff comments apply to the observations which
the IDR had designated either Level I or II, and which in some
cases had potential for hardware changes, FSAR revision, or the
appearance of safety significance. The staff reviewed the
project responses to these observations, interviewed the IDR
team, and examined available supporting documentation.
Observation 22-E03 noted an apparent inconsistency between FSAR
commitments regarding AFW " actuated devices" described in the
FSAR and the devices as shown on P& ids. The IDR reviewer had
interpreted " actuated devices" to exclusively signify automatic
actuation but this term was used in the FSAR in a more general
sense (e.g., an actuated device may have only remote manual
control). In response to this observation, BEC explained the
FSAR interpretation, and for additional assurance reviewed
the P& ids, logic diagrams, loop diagrams and other AFWs
documents to assure consistency of design documentation; no
discrepancies were reported in the actuation logic as a
result of that review. As a further result of that review,
BEC identified additional AFW actuated devices that would be
added to the FSAR list for consistency in reporting. The
staff concludes that this observation resulted from differing
interpretation of terminology; there was no technical discrepancy
in actuation logic; and the proposed FSAR revision would improve
reporting consistency. This observation is closed.
Observation 22-E04 identified an inconsistency between the
logic diagram and the elementary diagram for the turbine
driven AFW pump room intr.ke damper. The logic diagram
correctly showed the damper opening on an AFW start signal,
but the elementary showed the damper closing on the signal.
Investigation by BEC verified that the elementary diagram was
incorrect and that the error was missed in the drawing check.
BEC reported that the basic error was selection of a normally
open auxiliary relay contact rather than normally closed
(since this damper closes when its solenoid is energized).
16
__ _. _ l
-
. .
BEC also explained that the contact states required for this
circuit represent about 10% of the damper applications,
implying this may have contributed to the error. BEC checked
the remaining similar cases (28) and found no errors; they
further explained that BEC had independently discovered the
error, and in any case correct operation would be confirmed
during preoperational testing. The staff concludes that this
appears to be a typical isolated error, moreover that it
would be detected in preoperational testing. This observation
is closed.
Observation 22-E06 identified a concern that a vendor has not
adequately addressed orifice cavitation in his sizing criteria,
and that the project may have approved technically unacceptable
criteria. If cavitation is not properly considered, the orifice
would be damaged and its performance affected. BEC had not
specified cavitation criteria to the vendor. In response to
this observation, the project has implemented a program to
evaluate flow elements and orifices for cavitation, and has
identified orifices requiring resizing or replacement with
multistage orifices. At the time of inspection, calculations
had not been issued. The staff agreed that the corrective
action is appropriate, and at BEC, reviewed calculation X4C1202V22
Rev. 3, " Orifice Cavitation Verification Calculation". Revision
2 had been previously reviewed and accepted by the IDR team.
Revision 3 resolved the findings of the IDR team by performing
additional calculations for orifices experiencing cavitation,
and accounting for additional information acquired during
preop testing. The staff spot checked the safety injection
pump miniflow orifice to determine the potential for cavitation.
Based on design information provided by Westinghouse during the
inspection (Westinghouse letter MED-PVE-4635 of 10/8/86), the
staff concluded that there is adequate assurance that severe
cavitation would not be expected for any operating conditions of
that system. This observation is closed.
Observation 22-E08 identified incorrect specification of
orifice and flow transmitter ranges at 0 - 1000 gpm. A BEC
calculation established the low end of the required measure-
ment range to be 146 gpm. The orifice and transmitter range
,
!
l
specified would preclude meeting channel accuracy requirements
l at 146 gpm. BEC has committed to change the ranges of this
I instrumentation to 0 - 600 gpm, stating that this will meet R.G.
1.97 requirements for 0 - 110% of normal design flow while
.
meeting the accuracy requirements. BEC also states that with
l the new range, the reading could be off scale high when
measuring flow monitoring to a faulted steam generator but
that quantitative flow monitoring to a faulted steam generator
.
'
is not within the R.G. 1.97 basis; they also state that steam
generator levels are available and the off scale reading might
- help identify the break. The staff agrees with the interpre-
tation of R.G.1.97, and concludes that the corrective action
l
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I 17
I
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. .
for AFW flow monitoring is appropriate. BEC stated that the
error. was due to incorrect specification by a mechanical
engineer of the flow range based on the high flow when , feeding
the break, and failure of the controls engineer to recognize
the accuracy problem. A broadness review by BEC indicated that
two setpoint determinations donc by the same mechanical engineer
were incorrect. BEC has concluded.that the problem is limited
to these two enginears, but has committed to further strengthening
its loop verification program to identify any similar design
deficiencies. This connitment includes a review of all orifice
plates and corresponding accuracy requirements. The staff agreed
with the corrective action program proposed, and performed a
followup review of BEC's loop consistency review program. Con-
sistency reviews for approximately 487 loops had been completed,
.
and corrective actions such as changing meter scales had been
identified. The staff discussed with BEC the method of estab-
lishing loop requirements from mechanical / process discipline
inputs, methods for documenting the bases, and methods for
tracking change packages issued against the loop. Several
examples were sampled and found to be thorough and consistent.
This observation is closed.
Observation 22-E09 stated that BEC had not specified over range
protection for differential pressure transmitters, nor had
vendors provided documentation of over range protection. Over
range protection would be provided to assure integrity of the
instrument. BEC indicated that for the specification in question
(X5AD04), over pressure protection had been provided as a
standard feature even though not specified. BEC demonstrated
to the IDR team that adequate over pressure protection had been
provided for a typical pressure transmitter and differential
pressure transmitter in the AFW, and indicated that these require-
,
ments were assured by review of data sheets during bidding and
design. For instruments such as differential pressure switches /
- gauges where such protection may not be standard, BEC specified
j
over range protection equal to the rating of the device body;
, body rating is specified by considering the maximum process
pressure. The staff believes omission of over range protection
requirements from specifications is a significant documentation
,
!
e
deficiency, but agrees with the IDR assessment that adequate
'
over range protection appears to have been provided at Vogtle.
This observation is closed.
Observation 22-E10 identified deficiencies in controlling and
I
documenting the basis and calculation of set points and set
! point tolerances. In response, BEC has issued Desk Instruction
l
X5B006 detailing the scope, methodology, documentation require-
l ments, and responsibilities for the setpoint program so as to
assure compliance with R.G. 1.105. BEC has agreed that all
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. .
safety related B0P instrument loops used to initiate safety-
related actions or required for technical specification
verification will be within the scope of that Desk Instruction.
BEC also indicated that GPC .is preparing calculations .for NSSS
setpoints. 'The: staff believes that ~the corrective action is
appropriate, and performed a followup review of the component
cooling water (CCW) . tank . level setpoint determination and of
the setpoint program in general. BEC calculation X4C1203T02
established the hi/lo control setpoints for maintaining tank
level (via makeup and drain) and the hi/lo alann levels for
operator response to leaks in the system; this~ calculation
determined design basis thermal surge, leak rates, and NPSH
requirements (10-10 level CCW pump trip). Dedicated level
switches positioned according to a BEC level . setting diagram
(LSD) were used to implement control and alarm action. During
the inspection at BEC, it was determined that the installed level
switch elevations did not conform to all of the settings in the
calculation; certain installed settings conformed with LSD
1X5DT0025 Rev. O, but not with Rev.1. Revision 1 of the LSD was
consistent with the approved calculation. Moreover, Revision 2
of the LSD was the current revision, and had changed some of the
calculated settings back.to the installed settings '(i.e.,
installed e'ievations), apparently to avoid rebuilding the level
switch bridle. The staff requested that BEC reconcile the fol-
. lowing discrepancies.in the level settings and determined any
effects on .the system operation:
Function Installed Elev. Calculated Elev.
Hi control 254' -6 1/4 " 253' -9 "
(close valve)*
Lo control 253' -1 1/4 " 253' -2 "
(open valve)
Lo-lo CCW 252' -7 1/4 " 252' -9 "
pump trip
- This discrepancy was of most interest, since the installed
setpoint is only 11 3/4" from the top of the circular cross-
,
l section tank and is within 1 3/4" of the hi alarm setpoint.
!
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BEC addressed these discrepancies as follows:
1) BEC determined that the installed level settings had not
been revised to. agree with the calculation and that this
probably resulted from inadvertently omitting cognizant
reviewers from review of the DCN and LSD revisions.
2) BEC reevaluated calculation X4C1203T02 and determined that
thermal surges using the higher installed setpoints would
still not result in overflow of the surge tank into its
drain tank system during anticipated operating conditions
of the CCW. BEC also determined that there was ample HPSH
margin for the slightly lower "lo-lo" setting.
3) Even if the CCW surge tank were to overflow, the connecting
drain system and drain tank is designed to process and
contain the overflow in an orderly fashion and in accordance
with CCW design requirements.
To determine any generic implications of this discrepancy,
the staff requested a broadness review of all similar surge
tank instrumentation having potential safety significance, ,
to determine if other discrepancies might exist between
installed and calculated settings or elevations. The
auxiliary CCW system and the ESF chilled water systems were
reviewed; the mechanical calculations and the level setting
diagrams were reported to be in complete agreement for these
systems and no field discrepancies were reported. The staff
concluded that the CCW surge tank level setting discrepancies
were of no safety significance for the reasons given above, and
that there did not appear to be any significant generic implica-
tions resulting from this discrepancy.
To support the review of the balance of the setpoint program,
BEC provided a summary of virtually all of the setpoint
calculations within BEC scope. The staff reviewed the
summary, interviewed BEC staff regarding the calculation
bases, and spot checked some calculations. The staff
identified seven variables having setpoints important to
safety that were explicitly listed and that are often in 80P
scope; BEC was asked to identify responsibility for those
setpoints and to provide the calculation or equivalent basis
for any setpoints in BEC scope. The following was determined
for those variables:
20
--
. .
Variable Respons. Remarks
RWST level }[ None
Condensate storage GPC No control setpoint
tank level
Turbine impulse GPC Permissive for RPS
pressure
Turbine autostop BEC BEC calc. X5CP6161
oil pressure
Containment pressure W None
Main FW pump tripped GE BEC doc. CX5DT1101-26C
(hydro. press)
Turbine driven AFW Terry BEC calc. X4C1302V03
pump overspeed
The staff reviewed the BEC calculations and documents above,
and interviewed BEC personnel regarding the methods and as-
sumptions for establishing setpoint tolerances. The staff
concluded that a sound basis had been established by BEC for
setpoint tolerance calculations, input data was well con-
trolled, results were well documented in auditable form,
and BEC personnel appeared to have a good understanding
of the process.
To address the B0P setpoint calculations in GPC scope, the
staff requested the following information from GPC:
1) GPC's engineering procedure for performing setpoint
tolerance calculations.
2) GPC's setpoint tolerance calculation for turbine impulse
pressure.
3) Documentation supporting " engineering judgment".
4) Documentation supporting consideration of environmental
effects on excore flux detector cable and the effect on
channel accuracy.
The staff interviewed cognizant GPC personnel, reviewed the
information submitted by GPC and concluded that GPC's method-
ology appeared to be consistent with WCAP 11269 previously
approved by the staff for VEGP. The documentation submitted
l-
appeared to be complete and consistent. Based on the foregoing
'
review of both BEC's and GPC's setpoint tolerance calculation
l
program, the staff concludes that observation 22-E10 is closed.
l
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.. .. .. - - ____ - ____ _
. .
Observation 22-E16 identified inconsistencies in FSAR Table
10.4.9.5 regarding CWST level alarms; alarms are shown on the
shutdown panels in the. table, which does not agree with design
documentation (there are no alarms on the shutdown panels).
BEC reviewed the design documentation which correctly and
consistently shows these alarms on main control room panels,
and determined that the FSAR was in error; some additional
similar errors were found in the table regarding location and
depiction of various instruments. In no cases were discrepancies
in design information. found. An independent review by the staff
concluded that.the design was consistent and that correction of
these errors in FSAR Section 9.2.6.6 and Table 10.4.9.5 as
proposed by BEC will resolve this observation. Consequently,
this observation is closed..
The remainirg observations in the controls area are not indi-
vidually discussed in this report because they are of lesser
technical significance and wre satisfactorily resolved. In
view of the NRC's review of all observations in the controls
area, the staff concurs with the EA conclusion that the controls
area is technically acceptable. The staff finds this subsection
of the report to be acceptable and no further followup or
evaluation activities are required.
c.6 Section 3.6 - Electrical Design
Section 4.6 - Electrical Design
Section 4.10 - Equipment Qualification
Section 5.F - Electrical Design
Section 5.M6 Module 6, Electrical Equipment
(1) IDR Report
This subsection of the report concludes that the electrical design
area was found to be technically acceptable. Corrective and
preventive actions instituted in response to observations were
adequate to. resolve specific technical concerns. Evaluation of
potential generic or programmatic concerns are given in Section 3.10.
and Appendix F of the IDR Report.
(2) NRC Inspection Results
The IDR electrical design assessment included an examination
by SWEC of approximately: 225 drawings, 30 calculations, 30
specifications, 30 design criteria and over 100 other documents
and correspondence. The IDR team identified 71 observations,
of which 20 were eventually determined invalid. The following
!!RC staff assessment of the observations is provided:
Observation 22-F52 identified that the correct lubricants
were not used for pulling Okonite cables and some Rockbestos
22
.
cables. Subsequent to installation of the cables, approval
from Rockbestos to use Polywater J lubricant was received.
Approval for the use of Polywater J was also received from
Okonite; however, Okonite expressed some concern over the
lcng term effects of the lubricant on the cable insulation.
The Licensee has committed to an evaluation of the long term
effects of Polywater J on the cable insulation as part of the
environmental qualification report. The staff, during the
inspection at BEC in October 1986, reviewed the Okonite Company
letter dated November 27, 1985. This letter informs BEC that
the Okonite Company has review the test data supplied by .
American Polywater, and based on the test data, Okonite j
approves use of "Polywater J" lubricant on class 1E cables. !
The Team also reviewed American Polywater Co. letter dated
'
June 2,1986 and the Kerite Co. letter dated August 24, 1984. '
These letters conclude that Polywater J lubricant does not
have any deleterious effects on the cables. The staff finds
this resolution acceptable and this observation is closed.
Observation 22-F11 identified that conditions exist wherein
safety-related cable installations in vertical trays and
-
conduits are not adequately supported. This condition can
result in the cables putting excessive loading on terminal
end connections, with the potential for failure _of the
.
'
connections giving rise to safety concerns. The staff was ..
informed during the week of October 20, 1986 that GPC plans
to justify by analysis their current vertical cable support
configuration rather than install additional cable supports.
Review of preliminary supporting documentation provided by
- Bechtel led IE to defer this issue to NRR for evaluation.
This observation is considered closed with reg'ard to Module 22, -
but remains an open licensing issue pending evaluation by NRR. ~
Observation 22-F6 identified that some of the DC power cable
sizing calculations do not take the actual installed configu .
ration (such as: junction boxes, cable length, conduit seal -
assemblies etc.) into account. However. the project has a
cable sizing program to ensure that power cables are reviewed
by the respor.sible engineer for proper sizing after construction
has reported the installed length. Additionally, the project
has committed to the following:
l a. A review of all significant Class 1E DC loads for cable
size consistency between the calculation and the one lines.
b. A review of Train A power cables for consistency between
the one line and EE580 system.
I Correct calculations should include actual lengths, installed
cable size, and an analysis of voltage drop acceptability. The
project has recognized this need and is committed to update
calculation X3CK08 evcry six months. The above planned cor-
rective actions should resolve this action item. To verify
the implementation of these actions, the staff reviewed
23
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v-~ ws. . _ _. _ __ _ _ _ _ _ _ _ _ . _,_ _ . , _ _ , _ ,_,
- ______ ___
,.
.
calculation X3Cr.08 Revision 5 as a sample and noted that the
calculation.now uses installed length of cable in place of
designed length. The installed length is taken from the
EE-580 cable schedule which identifies installed cables by
the letters "cc" (construction completed). The staff verified
this fact from EE-580 dated 5/13/86. As an example, two
cables, LADLCA and LADLCB with design length of 43' and 51'
have "cc" value of.36'and 54', respectively. The staff veri-
fied that the 54' value was used for calculation of the
voltage drop. Cable sizing and voltage drop calculations
for motor feeder cables for valves HV-5120 and HV-5122 were
reviewed by the staff. The staff noted that the calculation
now accounts for length.and wire sizes, for electric conduit
seal assemblies, and for junction boxes. The staff found all
of the above actions satisfactory. This observation is closed.
Observation 22-F17 concerned whether analyses exist to support
the requirements of CMEB 9.5-1 for separation of different
safe-shutdown Trains of Electrical Cable within the same fire
area. The staff found that BEC had completed the Appendix 'R'
analysis and this analysis resulted in the following:
a. Approximately 33 cases of raceways require application of
3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> _ fire wrap. Three out of 33 raceways carry power
feed and the remaining 30 are utilized for control or
instrumentation applications. The power feed raceways
identified were 1BE311RM156,1BE311TLAM and 1BE304RS162.
The team noted that BEC performed an analysis for derating
of power feed cables for the fire wrap. (Ref: BB-51481,
file No. X3BE01). The derating factor used was taken from
the vendor's test report No. 82-5-355 F dated July 1982.
b. In one instance, a concrete wall was constructed to
redefine and separate fire zones 14C and 14D on
elevation 'C' as shown on drawing #AX4DJ8011.
The above actions by BEC are acceptable to the staff and
this observation is closed.
Observation 22-F34 concerned the validity of the 480 volt
cable sizing calculations. The existing calculation
apparently:
a. Contains an inconsistency between the conductor
temperature used and the design crit:ria.
b. Does not account for additional cable length due to
twisting.
c. Does not take accident ambient temperature into account.
24
._
.
The staff reviewed calculation X3CK06 Revision 6 for high
conductor temperature analysis. BEC used 81 C as a maximum
temperature in 2-3 such cases. Using the correction factor
for 81 C results in an additional voltage drop of 1.019%.
This calculation also addresses the effects of high ambient
temperature during LOCA conditions on voltage drop. Calculation
X3CA03-1 shows pre-LOCA worst case voltage drop on bus 1ABA as
0.9428 PU and during LOCA as 0.8653 PU, respectively. The NRC
team noted from calculation X3CK05 Rev.7, and calculation X3CK06
Rev.6, that terminal voltages using worst case (high conductor
temperature, LOCA ambient temperature and degraded bus voltages)
at motor terminals of motors located inside the containment
(containment cooler unit and purge unit fans) were within the
design limits. The staff also noted from calculation X3CK06
that an allowance has been added for twisted conductor cables.
Calculation of this allowance was done in accordance with the
vendor's method described in letter AX3AJ01-16-1 dated
November 12, 1986. All of the above actions by BEC are
acceptable to the NRC team and resolves this concern. This
observation is closed.
Observation 22-F35 indicates that design criteria used for
ampacity calculations 60 not address the deratings required
for punched bottom trays vs cable in air, covers or solid
bottom tray fittings. The project committed to revise design
documents so that maintained spaced cables in solid bottom
trays will be sized on the basis of the more conservative
standard, ICEA-P-54-440; to provide 25% uprating in ampacity
across the board for all the cables; to provide an engineering
evaluation for each instance where nonventilated covers are
used to evaluate if the 25% margin has been encroached due
to factors such as degraded voltage; and to revise the FSAR
design criteria and calculations to indicate the planned case-
by-case reduction from the committed generic allowance of 25%.
The staff noted that BEC has added the 25% uprating for ampacity
requirements of all the cables and an additional 10% derating
used for fire stops for trays without covers. For trays which
are covered, a derating factor of 27% is being used. This factor
resulted from testing done at San Onofre station by BEC. The
staff noted this has been referenced as note #10 on page 6 of
calculation X3CK01, reference 4, BEC IOM-E 2.6.4/LS 301 dated
10/22/84. The staff found that for solid bottom trays with
covers, ICEA-P54-440 values of ampacity are being used and
design criteria DC-1809 ampacity tables have been revised to
this effect. BEC has conducted analysis of cases where
encroachment of the specified generic margin has occurred to
verify that the affected calculations are revised. In view
of the above, this item is closed.
Observation 22-F37 concerned the adequacy of the DC system
design due to numerous findings related to first minute /last
minute voltages at the battery system and cable lengths that
25
. .
did not consider actual installed lengths, junction boxes,
and electric conduit seal assemblies. The team noted that i
J
BEC now uses the installed lengths.of cable and accounts for
junction boxes and electric conduit seal assemblies in
voltage drop calculations (see the resolution to observation
F06). The staff reviewed voltage drop calculations, battery
vendor's test data, and equipment qualification reports for one
DC operated valve and for a control' panel of the auxiliary
feedwater turbine. The team noted that the first minute voltages.
at batteries A,B,C were 116.23, 116.23 and 115.05 volts. If
these values were used for voltage drop calculations, the voltage
at DC valve HV-5120 will be 99.5 V, which is lower than the 100 V
for which this valve has been qualified. BEC stated that since
the vendor tests were conducted at 55 F and normal ambient in
the battery room will be at 70*F, the adjusted voltage values
for batteries A,B,and C will be 117.41, 117.41 and 115.93 V.
Using this voltage, starting voltage at valve HV-5120 will be
100.39 V (0.39 V above 100 V). The team found the last minute
voltages at batteries A,B were 108.78 V, 108.78 V and voltage
at battery panel C was 107.0V. This reflects 90.8 Y (qualified
for 90 V) at the PORV and 105.08 V (minimum required voltage is
105 V) at the AFW turbine driven pump control panel. These values
are marginal but since there is no specific requirement of a
minimum margin, these values are accepted by the staff and this
observation is closed.
The remaining observations in the electrical area are not indi-
vidually discussed in this report since they were of less
technical significance and were satisfactorily resolved. In
view of the review of all observations, the staff concurs with
the EA conclusion that the electrical design is technically
acceptable. This subsection of the report is acceptable and
no further followup or evaluation is required.
c.7 Section 3.7 - System Design
Section 3.8 - HVAC System Design
Section 3.9 - HVAC Duct Design
Section 4.7 - System Design
Section 4.8 - HVAC Sy: tem Design
Section 4.9 - HVAC Duct Design
Section 5.G - System Design & HVAC System Design
Section 5.M4 Module 16, NSSS
'
(1) IDR Report
i
In these subsections, SWEC concluded that the plant design
was technically acceptable and was in accordance with
licensing commitments and code requirements, with exception
of the specific observations noted.
l
26
l
!
.
(2) NRC Inspection Results
The IDR reviews included examination of approximately 85
drawings, 70 calculations, 24 specifications, 27 design
criteria, and over 100 other documents such as computer
reports, correspondence, procedures, vendor documents,
and change documents. As a result of this review, 27
valid observations were identified in the systems design
area. In addition, 5 valid observations were identified
in HVAC system design and 4 valid observations in HVAC' duct
design. The following NRC staff assessment of the observations
is provided:
Observations 4-19 and 16-3 identified that protection had not
been provided on the low design pressure AFW and SIS pump
suction piping from potential overpressure due to back
leakage through the discharge check valve of a nonoperating
pump. Review of other systems with a potential for similar
over pressure through back leakage revealed no other
instar:ces where protection was required. The potential for
overpressure of the AFW and SIS systems was corrected by the
installation of suction piping relief valves. The staff
reviewed the documentation associated with these observations
(including the change control packages for installation of
the relief valves) and concluded that appropriate action had
been taken. Further, the staff noted that the conditions
necessary to cause the overpressure do not exist when these
systems are relied upon to perform their safety functions,
since all system pumps would be operating at that time. These
observations are closed.
Observations 4-24 and 4-25 dealt with the sizing of the
Component Cooling Water (CCW) surge tank. The project's
initial responses were that this surge tank was considerably
oversized and that substantial margin existed in spite of the
concerns identified. Subsequently, it was discovered that
the complete surge tank was not higher than all of the system
piping, rendering a portion of the tank unavailable as a
surge volume. With this further reduction in tank margin,
the staff became concerned with certain assumptions made in
the sizing calculations. During the inspecticn at BEC
headquarters the week of October 6, the staff evaluated all
aspects of the design of this tank. Calculations X4C1203S02
and X4C1203T02 were reviewed and discussed with BEC. As a
result of this review, the staff concluded that the design
assumptions were sufficiently conservative and that the tank
design was in accordance with commitments and regulatory
requirements. This observation is closed.
Observations 16-1 and 16-2 identified instances where the
safety injection / residual heat removal systems could be
isolated from their over pressure protection paths should
27
. . . . -- _. - - . . - _. . . . .
- -
, .
. ;~
.
certain manual valves be shut. The design was modiMed to
'.' provide unblocked relief paths. . Review of relief paths in
'
other' safety-related systemsLrevealed no similar instances
where relief paths.could be blocked. The staff reviewed
,
-
documentation associated with these observaticas and deter-
mined that appropriate action had been taken. .These observa-
tions-are closed.
,
Observation 22-G02 dealt with the calculation of AFW turbine
driven pump room ventilation by natural convection. BEC
agreed that the present duct work arrangement differed from
that of the calculation but asserted that the overall effect
on the calculation would be ir.significanti BEC agreed to .
revise the calculation following the test to measure flows,
temperatures, and heat loads committed to in the answer to
,
- NRC question 410.54. During the NRC's inspection at BEC in
October 1986, the staff reviewed the test results and the -
!
subsequently revised calculation (X4C21E9V01). The test
l
shoved that the natural convection air flow was considerably
.
.
less than originally predicted (1200 CFM vice 3800 CFM) but -
that the resulting peak temperature (extrapolated to peak
outside design temperature of 98*F) exceeded-the limit for
the space by 2 F (122'F vice 120 F). Initial reviews C
'
indicated that the 2 F increase in peak temperature had no '
-
'
impact on installed equipment. Further, the test conditions '
were quite conservative in that certain steam leaks during
the test added considerable heat load to the room. Since >
- modeling of the natural circulation flow is complex and since
the confirmatory test was committed to prior to the IDR, the
'
staff considers this matter to be part of the: normal design
process. Confirmation by the Applicant inia letter from
D. P. Rice to Region 11 dated November 20, 1986 verified that
the equipment was qualified for a 2 F increase in peak
I
temperature and that appropriate documentation, including
appropriate FSAR sections, have been updated to reflect
extrapolated test results. This observation is closed.
i
Observation 22-GC5 involved conditions where spot cooling was
relied upon for the self cooling-of certain HVAC cooler
motors but that duct layout drawings did not direct adequate
air flow. As a result of this observation, a review of all
spot cooling applications was performed. Two instances were
identified where spot cooling applicatiens d'd not meet the
design intent. Design changes were initiated to adequately
l
account for spot t.ooling requirements in these cases. The 2
l staff confirmed the design changes by reviewing Des)gn Change -
~
,
Notices H-346-M and H-347-M during t.he inspection of BEC in s
i
L October 1986. Stbsequent calculatiohs by BEC indicated that g i s
28
e
,,.,._m---
---
,-,-- y p _,y .. . , _. .,, 7 ,%,_y._, .. ., , , , _ , , , , ,.,,,,,,_,m_r
h '
W
.
l'
.
. ^' h I
- design' temperatures would not have been exceeded, even with-
l 13 - .
out the design changes. This observation is closed.
- d -
-
M 3
Lu
,#' ~'f*
!
~ L .3 Observation 22-G14. stated-that.it was not evident that the
N" calculation con' firming AFW process design (X4C1302S08) had
y performed single failure) analyses for accidents other than
L
.the main feedline rupture. The observation questioned.whether
!. A single failure andysts had been performed for other potential
MF
~
accidents such as hain' steam line rupture and station blackout
'T, '
, ' with loss of normal feedwater. The project responded that
'
single failures had been properly. considered for the conditions
- analyzed in.the cited calculation and that system perforirance
had been evaluated for the worst single failures which could
- maximize or minimize system flow.- Subsequent review by the
IDR determined this to be a case of undocumented engineering
Judpent. After reviewirig the bases for the judgments, the
'IDR concurred that single failure had been considered for
- % worst"casciondi tions. Similar review of single failure
1:
'
u considerations by s the staff also resultad in' a concurrenct
U ,
~ y' Ctha(d.:: close ingle failure criteria had been met. This observation is
'
y
.
,
c
,L$- -
-' '? Otse'rvations 22-G171alon'g with 4-14 and 4-17, were related
5,' @ to concerns" associated with AFW system piping design pressure.
g. N, - Specifically, questions were raised'as _ to whether or not the
i '(
'
piping on the discharge side of the turbine driven AFW pump
."
could be pressurized in'ekces's of system design pressure under
'
a turbine over speed con _dition. As a result of these bbser-
vations, the design pressure of the piping between the turbine
V 1
driven AFW pump discharde and the restriction orifices was
'Mu@ . ,
increased from 1800 ps'ig tc 1975 psig. The problem wasdimited
-
- ,
4 to the AFW system which.is the only safety-related system with a
,
'
? ' turbine pump which;is subject to an over speed condition. With
T['- regard to the piping between the restriction orifices and the
stop check isolation valves, BEC asserted that this piping was
( ,~ [
--
,
not subjected tc#e over pressure condition since the isolation
t va'.ves are locked open.3 Further the VEGp commitment that normally
locbd open valves will not be manually c70 sad during system
- teiting of the AFW system was proposed 'in- FSAR Amendment 13 -
,
-(questioc'0420.32) and accepted by the' NRC in Section 7.3.3.7
'
'
cf.:the NE's SER of June 1986. In view of'the above, the staff
! codsiders the action taken to resolve this observation to be
! O[- 9ppropriate and this observation is closed.,,
- qy , , ,
- 1
The remaining observations in the system design, HVAC design,
l h, and HVAC duct design areas are not individually discussed
j '- N because they were of lesser technical significance and were
d.. b satisfactorily resolved and closed out.
-
i
m
In view of the technical discussicas @1ove, the NRC staff
I
^
concurswiththeIDRconclusions9thatithedesignoftheplant
'
is technically adequate. With r'egard tog hardware changes,
' '
! \
!
d 4 -
29 N
j )- %
1! ,
J
1
>!
2 - -- a : ~ _- . - - --. --
.
such as addition of suction side relief valves to prevent
over pressure or the increase in AFW turbine driven discharge
piping design pressure, the implications of these changes
relative to the overall design has been discussed in the
individual observations. With regard to other potential
generic concerns, these items are discussed in subsection
c.8 of this report. The staff finds this subsection of the
report to be acceptable. No followup action or evaluation
is required.
c.8 Section 3.10 - Evaluation of Generic Concerns
Appendix F - Evaluation of Generic Concerns
(1) IDR Report
In order to assess the collective significance of IDR
' observations, two separate analyses were performed. The
' first involved categorization of SBTA observations into
specific areas of weakness while the second included a
cumulative evaluation of IDR and other Readiness Review
observations related to the scope of the SBTA. From these
analyses, eight generic concerns were identified and
analyzed. The eight concerns are listed below:
Item Potential Concern
Finalization Programs Existing documentation did not clearly
indicate that the programs would correct
and prevent deficiencies of the type
noted in individual observations.
Calculations This concern was related to 1) errors
existing in project calculations
2) completeness and implementation of
guidelines for performing and checking
calculations 3) formalization of
'
.
calculation update requirements 4)
documentation of cumulative use of
design margins and 5) the degree of
implementation of ANSI N45.2.11
requirements.
Use of NON-VEGP Data Use of NON-VEGP specific technical
data may not have been properly
evaluated and documented for
applicability.
Piping Design Concerns were identified relative to
1) the use of stress intensification
factors 2) technical and procedural
noncompliance 3) documentation of
design and operating modes and
4) control and completion of outstanding
technical issues.
30
- -
- - . - -- -- .
-
. .
Pipe Support Design Concerns were identified relative to
1) technical and administrative adequacy
of support calculations and 2) control
and completion of outstanding technical
issues such as load changes and support
location changes.
Hazards Program Concerns were identified relative to
1) the adequacy of documentation
2) adequacy of intradiscipline interface
and design change control 3) adequacy
of project procedures and 4) use of
inappropriate plant layout information.
Electrical Design Concerns were identified relative to
1) inconsistencies between electrical
design documents 2) the adequacy and
accuracy of the electrical design
criteria and 3) the adequacy of DC
system design.
BEC/ Westinghouse Concerns were identified relative to
Interface 1) over pressure protection design
2) piping system desigr. load
verification and 3) handling of NSSS
design information.
Following a detailed assessment of each of these items in
Section 3.10 and Appendix F, the IDR concluded:
" Based on the detailed assessments above, the IDR team
concludes that with comple"on of the required corrective
actions, and program enhc :1ments being implemented by the
project, all identified potential generic / programmatic
concerns have been adequately addressed."
(2) NRC Inspection Results
Based upon the large number of staff inspections of the VEGP
Readiness Review Program associated with plant design and the
design process, as detailed in this report, particularly the
inspection of corrective and preventive actions at BEC
1
headquarters from October 6-9, 1986 (which directly evaluated
i
the generic concerns), the NRC staff concurs with the IDR
l
assessment that the generic concerns have been adequately
'
addressed. The staff finds this subsection of the report to
be acceptable. No followup activities or further evaluation
is required. Specific staff comments associated with each of
the potential generic concerns are provided below:
!
31
,
'
- - -
_.
.
Item Comment
Finalization Programs Enhancements to the Finalization Programs
resulting from IDR observations have
resolved potential concerns that these
programs might not adequately address
the observations of the IDR. When
completed, the Finalization Programs
will provide additional assurance
that the design of VEGP is in accordance with
commitments.
Calculations During the course of several . inspections,
the staff has independently reviewed a
number of calculations in all technical
disciplines. The number and magnitude
of problems identified by the IDR and
the NRC are not unusual for a design
project of this magnitude. The staff
has-confidence that the calculations
adequately support the plant design.
Use of NON-VEGP Use of NON-VEGP data was principally
Data applied as corroborative evidence in
support of engineering judgments. In
some cases, additional information was
required to substantiate the use of
this data for the VEGP. In no case
did hardware changes result solely
from the application of NON-VEGP data.
Finally, a broadness review indicated
that NON-VEGP data was not extensively
used.
Piping Design During the course of the inspections,
design criteria documents and a large
number of calculations were reviewed by
the staff. Based on these reviews, the
staff concurs that these generic concerns
have been adequately addressed.
Pipe Support Design During the various NRC 1nspections, a
large number of pipe support calculations
were reviewed. With minor exceptions
the calculations were ad;teate and
outstanding technical issues were
adequately controlled.
Hazards Program At the time of the IDR, the VEGP Hazards
Program, both from the standpoint of
work accomplished and documentation
available, was further advanced than
typical hazards programs at a similar
stage of design. The enhancements
32
-
. - . - - . . -- _. -- .- . . .- ..
. .
1 made as a result of the IDR, as
subsequently verified by the staff, 1
substantially improved the-program in
- . specific areas. The final result is
a comprehensive, thorough, and well
,
documented program which achieves
design objectives. The staff concurs
!- with the IDR conclusion that a generic
program breakdown did not occur. With
.
regard to the failure to postulate
longitudinal breaks (observation 22-B01),
'
the hardware changes resulting from '
this observation (relocation of some
,
electrical conduit) provide an additional
measure of conservatism to the design of-
the plant. The safety implications of
plant operation without having relocated
.
'
.this conduit are not significant.
Electrical Design The staff's review of the electrical
- area indicated specific weaknesses
which were corrected, but not a
,- programmatic breakdown of the design
'
process. Although there was generally
little margin in the DC system design,
the basic requirements were met and.
i the overall electrical design is
i considered adequate.
[ BEC/ Westinghouse Root causes of these discrepancies
Interface involved interpretation of code
requirements, inattention to detail
when transferring data, and designs-
2 based on anticipated criteria changes,
i Specific corrective actions were
undertaken to remedy these root causes.
The staff concurs.that no programmatic
i breakdown occurred and that a proper.
- and controlled interface exists.
'
d. Section 4 - Review Description
. (1) IDR Report
The technical content of this section has been discussed in
the appropriate subsection of paragraph 3.c, above.
~
(2) NRC Inspection Results
!
Inspection results are given in paragraph 3.c,
- above.
.
1
33
i
i
1 ,
._. r - , -. , _ . - . . . . , . . , , , . , , . .. .. - _ _ _ - , _ _ . . _ _ - _ _ . _ , . _ . . - _ - - - . - _ _ , . -__-
.
.
e. Section 5 - Review Observations, Responses, Assessments
(1) IDR Report
The technical content of this section has been discussed
in the appropriate subsection of paragraph 3.c, above.
(2) NRC Inspection Results
Inspection results are given in paragraph 3.c, above.
f. Appendices
(1) IDR Report
The report appendices provide either (1) background
information (e.g. the review plan, personnel contacted),
(2) supplementary information (e.g. documents reviewed,
action commitment summary), or (3) further review
analysis (e.g. Appendix F - Generic / Programmatic Project
Assessment).
(2) NRC Inspection Results
Information in the appendices was reviewed either for
background information or with the corresponding
subsection of Section 3 (discussed in paragraph 3.c,
above). No followup activities or evaluation of these
appendices are required.
4. Statement of Module Acceptability
In paragraph 3 of this report (NRC Staff Evaluations), each section of
the IDR Report (Readiness Review Module 22) was technically evaluated.
In each instance, the specific sections of the IDR reoort were found
acceptable to the NRC staff, for reasons enumerated in the individual
subparagraphs. Taking each of these subparagraphs collectively, the NRC
finds Module 22 to be acceptable. No followup activities or evaluations
are required.
In view of the results of Module 22 and the independent assessments of
the structural design, (Modules 1, 8, and 13), both the analysis of
individual observations and the evaluation of potential generic concerns,
the NRC concludes that a technically adequate design process has been
successfully implemented at VEGP. The Applicant's Readiness Review
Program has provided additional assurances that the design of VEGP is in
accordance with licensing commitments and regulatory requirements.
Further, the PRC's direct monitoring of the IDR provides additional
credibility to the validity of the program results. No further pre-
license design review activities are considered necessary.
34
/
. .
List of Persons Contacted
Throughout the course of the five inspections conducted by the NRC in monitoring
Module 22, a large number of personnel were contacted from each of four organi-
zations: Georgia Power Company, Stone & Webster Engineering Corporation, Bechtel
Engineering Corporation, and Southern Company Services. Individuals contacted
during a specific inspection are identified in the report of that inspection.
The report numbers are identified in Section 2 of this report.
The following is a summary of key individuals of each organization contacted by
the NRC:
Georgia Power Co./ Southern Company Services
Batum, 0. Deputy:to V.P. Engineer
Foster, D. V.P. Project Support
Hayes, C. Vogtle QA Manager '
McManus, R. Assistant Project Const. Manager
Ramsey, W. Manager, Readiness Review
Read, D. General Manager, QA
Rice, P. V.P. Project Engineering
Bechtel Engineering Corp.
Beer, M. Westinghouse Technical Assistant
Blasingame, J. Manager of Engineering
Capito, D. Piping Group Supervisor
Carpa, P. Chief Engineer, Westec Power Div.
Cereghino, S. Nuclear Group Supervisor
Freid, S. Assistant Chief Mech / Nuclear
Johnson, W. Control Systems Group Supervisor
'
Malcolm, M. Deputy Project Engineering Manager
Marsh, F. Project Engineering Manager
McDaniel, P. Mechanical Deputy Group Supervisor
Morrow, D. Electrical Group Supervisor
Sanders, A. Assistant Project Engineer
Smith, R. Electrical RR Coordinator
Strohman, D. Project QA Engineer
Walvekar, K. Mechanical Group Supervisor
Stone & Webster Engineering Corp.
Allen, J. Project Manager
Bushnell, G. Hazards Reviewer, IDR
Eifert, W. Chief Engineer, EA
Frank, S. Mechanical Systems Reviewer
>
Gardel, W. Electrical Reviewer
Genosi, A. Engineering Assurance / Module 4
King, D. Engineering Manager
Lockaby, J. Duct & Pipe Supports Reviewer
Mathur, R. Structural Reviewer
Mullen, C. Controls Reviewer
Stamm, S. Tech. Manager - IDR
Tsai, C. Pipe Stress Reviewer
35
1
.
.The following acronyms are used throughout this report:
AFW Auxiliary Feedwater System
ANSI American National Standards Institute
ASME American Society of Mechanical Engineers
BEC Bechtel Engineering Corporation
B0P- Balance of Plant
CCW Component Cooling Water
CWST Condensate Water Storage Tank
DC Direct Current
EA Engineering Assurance
FCR Field Change Request
FSAR Final Safety Analysis Report
GPC Georgia Power Company
HVAC Heating,-Ventilating & Air Conditioning
IDR Independent Design Review
IDVP Independent Design Verification Programs
IE NRC Office of. Inspection and Enforcement
LOCA Loss of Coolant Accident
NRC Nuclear Regulatory Commission
NRR NRC Office of Nuclear Reactor Regulation
NSSS Nuclear Steam Supply System
P&ID Piping & Instrument Diagram
PORV Power Operated Relief Valve
PSDM Pipe Support Design Manual
PVRC Pressure Vessel Research Council
RCL Reactor Coolant Loop
RR Readiness Review
SAR Safety Analysis Report
SBTA System Based Technical Assessnent
SER Safety Evaluation Report
SIS Safety Injection System
SWEC Stone & Webster Engineering Corp.
VEGP Vogtle Electric Generating Plant
1
36