ML20196L357
| ML20196L357 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 06/24/1988 |
| From: | Cheng T, Fair J, Hermann R NRC OFFICE OF SPECIAL PROJECTS |
| To: | |
| Shared Package | |
| ML20196L356 | List: |
| References | |
| 50-327-88-12, 50-328-88-12, NUDOCS 8807070392 | |
| Download: ML20196L357 (66) | |
See also: IR 05000327/1988012
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UNITED STATES NUCLEAR REGULATORY COMMISSION
QF,FICEOFSPECIALP(0JECTS
TVA PROJECTS DIVISION
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Report No.:
50-327/88-12 and 50-328/88-12
Docket No.:
50-327/328
Licensee:
Tennessee Valley Authority
6N, 38A Lookout Place
1101 Market Street
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Chattanooga, TN 37402-2801
Facility Narne:
Sequoyah Nuclear Plant, Unit 2
Inspection At:
Knoxville, TN
Inspection Conducted:
February 15-19, 1988
Inspectors:
/, AI
6/a.N / W
L ohn R. Fair, Team Leader
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Thornas M. 01eng
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Consultants:
A. V. <tuBouchet
R. E. Serb, A. I. Unsal, G. Harstead,
T. Tsai, O. Mallon
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Approved By:
- L Robert A. Hermann
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Chief. Engineering Branch
TVA Projects Division
Office of Special Projects
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8807070392 880624
ADOCK 05000327
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SEQUOYAH NUCLEAR POWER PLANT
DESIGN CALCULATION REVIEW PROGRAM
INSPECTION REPORT 50-327/88-12 AND 50-328/88-12
FEBRUARY 15-19, 1988
1.
INTRODUCTION AND BACKGROUND
The design calculation review program was developed by the Division of Nuclear
Engineering (DNE) because past audit findings ard other reviews have shown that
the design basis for TVA's nuclear power plants have not been adequately docu-
mented by supporting calculations or that such calculations, if performed, nay
no longer be retrievable. This program augmented the Sequoyah Nuclear Plant
(SQN) design baseline and verification program (OBVP) by including a technical
adequacy review of supporting calculations, a feature not included in the DBVP.
The design calculation review plan was initially described in an enclosure
to TVA letter from R. L. Gridley dated January 20, 1987 and Revision 2 to
Section III.4 of Sequoyah's Nuclear Performance Plan dated March 27, 1987
The design calculation review plan was subsequently updated in enclosure to
TVA letters from R. L. Gridley dated July 31, 1987 and August 21, 1987. The
design calculation review plan addressed the essential calculations required
to support the SQN design basis in the four technical branches of DNE.
The NRC conducted three previous inspections of the design calculation review
program and documented the results of these inspections in reports 50-327,
328/87-06, 50-327, 328/87-27 and 50-327, 328-87-64 (References 10,18,23).
TVA has responded to the observations identified in these reports (R.'ferences
4,20,21,24,~25).
In addition to the inspections on the cesign calculation
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review program, the NRC has conducted an Integrated Design Inspection (IDI)
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of the Essential Raw Cooling Water (ERCW) system.
The results of these
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inspections are documented in reports 50-327,328/87-48,50-327,328/87-74
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and 50-327, 328/88-13 (References 26,28,30). TVA has responded to the
observations identified in these reports (References 27, 29).
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2.
PURPOSE
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The purpose of this inspection was to review TVA's corrective actions associ-
ated with the civil engineering portion of the calculation review program.
The inspection scope included a review of rigorous piping analyses, regenerated
piping support calculations and the followup of TVA corrective actions
associated with NRC observations in the civil engineering area documented
in previous NRC design control inspection reports including the IDI.
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3.
RESULTS OF NRC INSPECTION
The following paragraphs characterize the inspection findings and conclusions
in each area of ~the civil engineering calculations review effort. The results
of the followup review of findings from previous design calculation, DBVP and
IDI items are provided in Appendices A, B and C respectively. Appendix 0
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contains a listing of open post-restart items resulting from this inspection.
.
3.1 Rigorous Piping Analysis Review
3.1.1
Scope
The Civil Engineering Branch (CEB) calculation review program in the rigorous
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piping analysis area was originally based on the recommendations for corrective
action contained in CEB summary report, "Evaluation of Programs Establishing
Technical Adeq acy of the Civil Calculations," dated January 30, 1987 (RIMS No.
,
B41 870130 013 .
To assess the adequacy of Sequoyah Unit 2 (SQN 2) rigorous piping analyses TVA
contracted Gilbert /Coninonwealth (G/C) to perform a technical review of five
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sample rigorous analyses. The review was completed in May 1987 and documented
in Gilbert /Conmonwealth. Inc. Report No. 2689 (RIMS B41870519 250).
.
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The five rigorous analysis piping problems reviewei by G/C were selected from
those SQN 2 problems for which engineering change notices had been executed
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subsequent to operating license issuance. They were selected to assure the
following eight attributes identified by TVA would be overviewed:
Equipment nozzle qualification.
Valve with operator acceleration qualifi,
. . .
High energy lines.
Multiple seismic zones.
Primary loop displacements.
DBA load cases.
SCV penetrations.
ECNs.
The problems selected also represent a variety)of piping systems, classifica-
tions and physical locations (i.e., structures . TVA explained during the
inspection that excepting provisions for SCV connected piping reanalysis, pre-
and post-operating license piping analysis criteria are essentially the same.
Therefore, review and conclusion based on post operating license analysis
review are applicable to all analyses irrespective of the current revision
date (i.e., pre or post operating license). The review scope as described is
considered representative of SQN 2 rigorous piping analyses.
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3.1.2
TVA Disposition of Review Findings
TVA responded to the descrepancies identified by G/C review in a report, "TVA
Review Program in Response to Gilbert / Commonwealth Report No. 2689" (RIMS B41
870617 250). Each disposition includes TVA determination of validity, signifi-
cance, and required corrective action. TVA disposition of fifteen of the
-ninety-four G/C identified discrepancies associated with two of the five sample
problems were reviewed during the inspection. An additional review of the G/C
Report No. 2689 was performed by Robert L. Cloud Associates (RLCA), (RIMS 841
870521250). The RLCA report provided an additional review of the adequacy of
Sequoyah's rigorous piping analyses. The following paragraphs sumarize
observations resulting from the review of these rigorous piping analysis reports.
The corrective action in the TVA report for each of four review discrepancies
(Items 15, 18, 33 and 34) was to include an entry to the SQN Analysis Open
!tems Log. This would assure identification of each discrepancy for correction
dering the next reanalysis for the affected problem.
In three cases the TVA
repb t . indicated that the entry to the log had already been made. However,
these open items had not been included in the log as of February 18, 1988. The
log was updated on February 19, 1988 to include all open items from the TVA
raport.
The TVA report incorrectly dispositioned an analysis review discrepancy
(Item 46) as invalid. The discrepancy identified two valves of Problem
N2-74-5A for which the analysis specified weight was double the correct
weight.
In addition, the RLCA assessment of the G/C review also found the
valve weights had been doubled in the analysis.
However, TVA in their report
concluded the weights had been correctly specified and the discrepancy was not
valid.
The review of this discrepancy during the inspection confirmed that
it was valid as originally identified in the G/C report. TVA assessed the
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technical significance of the discrepant condition and concluded it does not
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significantly affect design. The assessment was based on the analyses of the
problem that had been performed by both G/C and RLCA which used correct valve
weights. The results of these analyses did not identify a design problem.
This inspection identified five deficiencies with TVA's resolution of the G/C
findings.
Four of these deficiencies involved TVA's failure to enter correc-
tive actions in the analysis open items log as required by the TVA dispositions.
The fifth deficiency, the incorrect conclusion that a review discrepancy was
not valid, was not regarded as technically serious.
In summary, although some
deficiencies with TVA's resolution of the G/C findings were identified during
the inspection, the identified deficiencies were not significant and the
deficiencies did not require additional analyses or plant modific?tions for
restart resolution.
As part of the inspection the SQN rigorous analysis handbook, "Sequoyah Nuclear
Plant Rigorous Analysis Handbook Class 2 and 3 Analysis," December 9,1987, was
reviewed. The rigorous analysis handbook addresses design requirements and
analysis procedures including detailed modeling procedures, qualification and
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verif1 cation procedures which invoke extensive design and analysis procedure
checklists, and documentation requirements. No concerns regarding the handbook
were identified during the inspection.
3.1. 3,
TVAGenericConcernIdentificationandDisposition(RCLDisplacement_sl
TVA produced a final closure report for the rig rous piping analysis review
effort, "Closure Report for Technical Review of Rigorous Piping Stress
Analysis," January 1988 (RIMS B41 880129 007). The TVA review closure report
was reviewed during the inspection.
In that report TVA reviewed the generic
implication of the G/C review results.
The TVA review of the G/C identified
discrepancies concluded that three concerns potentially affected other SQN 2
rigorous analyses by generic implication. These concerns were valve / flange
weights, reactor coolant loop (RCL)/ steel containment vessel (SCV) movements,
and equipment nozzle movements / allowable loads. TVA's report identified that
a review of all post-OL reanalyzed stress analysis calculations reveale.d one
discrepancy. This discrepancy was that equipment nozzle movements / allowable
loads for the revised Westinghouse RCL analysis (TVA Stress Report No. SD 105,
dated May 7,1974) had not been incorporated into analyses for loop attached
piping. To resolve this discrepancy TVA contracted G/C to perform an evalua-
tion of the revised loop analysis resultant displacement effects.
The seismic
anchor motion (SAM) at twenty-four of the fifty piping attachments to the RCL
were found to change by less than one-sixteenth of an inch. The revised
displacements at the other twenty-six attachments were evaluated by a combin-
ation of computer and hand calculations. These evaluations concluded design
criteria were met with the exception that five nozzle loads exceeded currently
specified Westinghouse allowables.
Based on further review G/C concluded
nozzle equivalent stresses were acceptable (Summary of Activities for CAQR
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761161 dated February 18,1988).
Based on the TVA evaluation, the final
disposition of this concern by incorporation of the revised SAM displacements
into problem reanalyses post restart is acceptable.
3.1.4
EA Review
SQN 2 rigorous analysis adequacy was also addressed by TVA Engineering
Assurance (EA) Audit 87-09, Concern Observation C-1.
The concern remained
open in three EA corrective action status reports to EA 87-09 (December 2,
1987; February 2, 1988 and February 12,1988). The TVA Closure Report for
Technical Review of Rigorous Piping Stress Analysis concludes that the EA
concern is closed except as it relates to an equipment "Q"
list.
In the EA
memorandum dated February 12, 1988 EA agrees with the closure report except
that it notes that the need remains to receive and roview information from
Bechtel regarding design basis accident (DBA) zero period acceleration (ZPA)
effects. The issue of DBA ZPA effects is discussed in Section 3.7 of this
Inspection Report.
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3.2 TVA's Calculation Regeneration Program for Rigorously Analyzed
Pipe Supports in SQN Unit 2 and Comon
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3.2.1
Procedures and Cr,iteria
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TVA's program plan and implementing instructions to regenerate the calculations
which EDS and Basic Engineers originally prepared for the rigorously analyzed
pipe supports in SQN Unit 2 and common are detailed in the following CEB
procedures:
.1.
CEB-C121.80, "Program Plan for Calculation Regeneration of Pipe Supports
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on Rigorously Analyzed Category 1 Piping - Sequoyah 2," Revision 1, dated
August 28, 1987.
2.
CEB-D121.81, "Generation and Control of Rigorous Analysis Problem
Connectivity Diagrams for Category 1 Piping:
Sequoyah 2," Revision 1,
dated August 28, 1987.
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CEB-D121.83, "Functional Verification of Supports for Rigorously Analyzed
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3.
Category 1 Piping:
Sequoyah 2," Revision 3, dated December 14, 1987.
4.
CEB-C121.84, "Control of Correspondence and Transmission of Design
Documents between TVA and Engineering Servicos Contractors," Revision 1,
dated August 28, 1987.
5.
CEB-D121.85, "Generation of Pipe Support Design Data:
Sequoyah 2,"
Revision 2, dated November 19, 1987.
6.
CEB-Ci 21.88, "Control of Input and Output from the Sqn Hanger Tracking
Subprogram of CCRIS," Revision l', dated October 19, 1987.
7.
CEB-C1 21.89, "Modification Priorities for Pipe Supports on Rigorously
Analyzed Category 1 Piping - Sequoyah Unit 2," Revision 2, dated
December 18, 1987.
8.
CEB-C1 21.90, "Gang Hanser and Terminal Procedure," Revision 0, dated
August 31, 1987.
9.
CEB-C121.91, "Handling of Pipe Support Calculation Review / Regeneration
Results - Sequoyah 2," Revision 0, dated December 18, 1987.
10. CEB-C1 21.92, "Red Lining of Pipe Support Drawings," Revision 0, dated
December 14, 1987.
On July 17, 1987, TVA issued design criteria SQN-DC-V-24.2, "Supports for
1 Piping," to specify the design criteria which
Rigorously Analyzed Category (SWEC) used to regenerate the pipe support
Bechtel and Stone & Webster
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calculations.
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TVA reviewed pipe supports which did not meet the long-term design criteria of
SQN-DC-V-24.2 to the interim design criteria of CEB-Cl 21.89.
Pipe supports
which meet the interim design criteria will be modified post-restart.
3.2.2,
Review Results
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On February 15, 1988, TVA ind'.cated that 4,984 of the 5,612 pipe support
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calculations in the calculaCion regeneration program met the long-term design
criteria of SQN-DC-V-24.2. The remaining 628 pipe supports were evaluated to
the interim design criteria of CEB-CI 21.89. 447 pipe supports met the interim
criteria and are scheduled for modification post-restart. TVA has already
modified the remaining 181 pipe supports.
According to TVA, the relative high proportion of pipe supports requiring
modification is due in part to the requirements that friction forces be
considered in the design of pipe supports, and that the maximum allowable
stresses in built-up pipe supports be limited to nine-tenths of the material
yield stress.
In order to assess the adequacy of TVA's pipe support calculation regeneration
program, a sample of 23 pipe support calculations which meet the long-term
design criteria of SQN-DC-V-24.2 was reviewed. These pipe supports are
installed in the essential raw cooling water (ERCW), component cooling water
(CCH) and main steam (MS) piping systems:
1.
ERCW system:
1-ERCWH-56(Bechtelcalculation
1-ERCWH-61 (Bechtel calculation
1-ERCWH-62 (Bechtel calculation
1-ERCWH-44 & -100 (Bechtel calculation; common hanger)
1-ERCWH-132 (Bechtel calculation)
47A450-21-216 (SWEC calculation)
2.
CCH system:
2-CCH-62 (Bechtel calculation)
2-CCH-80 & -81 (Bechtel calculations; common hanger)
1-CCH-677 (SWEC calculhtion)
1-CCH-685 & -688 (SWEC calculation; common hanger)
1-CCH-699 (SWEC calculation)
1-CCH-736 & -737 (SWEC calculation; common hanger)
3.
MS system:
2-MSH-302(Bechtelcalculation
2-MSH-304 (Bechtel calculation
2-MSh-310 (Bechtel calculation
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2-MSH-346 & -347:(Bechtel calculation; common hanger)
2-MSH-384'(Bechtelcalculation)
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2-MSH-432 (Bechtel. calculation)
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Three deficiencies were identified during the course of this review.
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The first deficiency was that:Bechtel did not perform a check for beam web
crippling due to pipe bearing in -the sample of pipe support calculations
reviewed during the inspection. To address this concern, TVA sampled an
additional 50 pipe support calculations and confirmed that 49 of the 50 beams ~
satisfied the web crippling check specified in the edition of the AISC code
which design criteria SQN-DC-V-24.2 references. TVA confirmed that the
remaining beam would meet the comparable criterion specified in the 1986
edition of the AISC code.
TVA stated they would revise the design criteria
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SQN-DC-V-24.2 to specify that a beam web crippling check for pipe bearing be
performed in accordance with the 1986 edition of the AISC code. This
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deficiency is closed based upon the results of TVA's generic review and
evaluations which demonstrated that AISC structural criteria was met.
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The second deficiency was that SWEC had incorrectly coded a 2,263 lb. thermal
load into the SANDUL computer run used to analyze pipe support 1-CCH-677. The
thermal load actually_ used to analyze the pipe support was 263 lbs. The load
was not coded into the-correct input field of the program, and the program
input echo did not flag the truncation of the first digit. To address this
deficiency, TVA retrieved an additional 25 pipe support calculations with
~ SANDUL computer runs, and confirmed that the input loads for these runs were
correctly coded. TVA will also revise the calculation for pipe support
1-CCH-677. This deficiency is closed based upon the results of TVA's generic.
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review and TVA's decision to revise pipe support calculation 1-CCH-677.
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The third' deficiency was that Bechtel's calculations 2-MSH-346 & -347 for a
combined spring hanger / snubber pipe support failed to qualify the pipe support.
Calculation 2-MSH-346 for-the spring hanger portion of the pipe support
concluded that the spring hanger would bottom out when subjected to piping
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thermal displacement. The spring hanger would therefore be subject to the
snubber design seismic loads in addition to the spring hanger design dead load.
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However, the calculation did not identify the need to modify the spring hanger
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support to eliminate this unacceptable condition.
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To address this deficiency, TVA inspected the spring hanger and confirmed
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that adequate spring hanger travel exists to acconinodate ciping themal
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displacement. TVA will revise the calculation for pipe support 2-MSH-346
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to incorporate revised field measurements which confirm the adequate travel
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capacity of the spring hanger under piping thermal movement. This deficiency
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is considered to be isolated, and is closed based upon TVA's inspection of the
spring hanger and TVA's decision to revise the calculation.
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A sample of 10 pipe support calculations for pipe supports which TVA has
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scheduled for post-restart nodification was also reviewed:
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1.
ERCW system:
1-ERCWH-9 (Bechtel calculation)
1-ERCWH-21 (SWEC calculation)
1-ERCWH-51(Bechtelcalculation)
1-ERCWH-174 (Bechtel calculation)
1-ERCWH-226 (Bechtel calculation)
47A450-21-228 (SWEC calculation)
2.
CCH system:
1-CCH-198(SWECcalculation
1-CCH-205 (SWEC calculation
3.
MS system:
2-H1-301 (Bechtel calculation)
2-MSH-313 (Bechtel calculation)
The staff concurred with TVA's decision to modify these pipe supports post-
restart.
Based upon staff review of TVA's program plans and design criteria, and review
of a sample of the. pipe support calculations which Bechtel and SWEC prepared,
the staff concludes that TVA's pipe support calculations were generally
prepared in accordance with program technical and quality assurance criteria.
3.2.3
EA Rev'y
TVA's engineering assurance (EA) group has audited Bechtel and SWEC to confirm
that the technical and quality assurance provisions of their contracts with TVA
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were being adequately implemented. TVA has documented these audits in the
following TVA reports:
1.
Sequoyah Nuclear Plant - Tasks Performed Under Personal Services Contract
TV-72104A - Procured Services Audit 87P-51, dated October 14, 1987 (an
audit of Bechtel in San Francisco during the periods August 17-21 and
August 31-September 3,1987),
2.
Sequoyah Nuclear Plant (SQN) - Watts Bar Special Projects - Personal
Services Contract TV-72102A - Engineering Assurance - Procured Services
Staff Audit 87P-53 (an audit of SWEC in Boston during the periods August
25-28 and October 5-9, 1987, and in Knoxville during the period August
4-7,1987).
3.
Division of Nuclear Engineering (DNE) Engineering Assurance (EA) Audit
87-09 (Technical) - DNE Calculations Review Effort, dated December 2, 1987
(a followup to EA audit 87-09(T); includes EA review of 26 pipe support
calculations).
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4.
Division of Nuclear Engineering (DNE) - Engineering Assurance (EA) Audit
87-09 (Technical) - DNE Calculations Review Effort - Civil Discipline,
dated February 1,1988-(an additional followup ? audit).
TVA EA additionally reviewed 54 Bechtel and SWEC pipe support calculations,
and identified 20 deficiencies.
Bechtel and SWEC have provided TVA EA with
acceptable resolutions to these deficiencies.
Under the direction of TVA's lead pipe support engineer, TVA engineers also
performed a three-tier review of the regenerated pipe support calculations in
accordance with CEB-D121.87, which specifies that each pipe support calcula-
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tion be checked against procedure NEP 3.1, calculations, that ten percent of
the pipe _ support calculations be reviewed programatically, and that 100 pipe
support calculations be reviewed line-by-line. TVA has indicated that, in
addition to screening each pipe support calculation to the NEP 3.1 requirements.
TVA programatically reviewed approximately one-third of the pipe support
calculations, and reviewed 80 pipe support calculations line-by-line. TVA was
planning to issue a report on or about May 1, 1988 to summarize this review.
3.3 Thermal Monitoring of Supports
As part of the pipe support calculation regeneration effort, TVA developed a
set of restart criteria, CEG-CI 21.89 (Reference 50). The staff accepted this
criteria subject to restrictions (Reference 51). TVA revised this restart
criteria based on additional discussions with the NRC staff (Reference 52).
One of the revisions to the restart criteria allowed TVA to monitor snubber
swing angles during plant heatup to verify that thermal binding would not
These measurements were to be used instead of using the calculated
oci,cr.
piping th?rmal movement for computing the angular swing for comparison with
allowable tolerances. TVA identified 13 supports to be monitored during the
he6iup (TVA memorandum Hosmer to Abercrombie dated December 18, 1987 RIMS No.
B25 871218 020). The staff review identified that four supports 2-H63-2,
2-H63-3, 2-H63-4 and 2-H63-5 were being monitored by strain gage to obtain
thermal loads. These strain gage reasurments were not part of the agreed upon
criteria (Reference 52). TVA responded (Reference 41) that the four supports
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met the allowable stress criteria in CEB-Cl-21.00. TVA also proposed using
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the measured loads to qualify the supports to the long term design criteria
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SQN-DC-V-24.2 The staff does not accept this long term solution unless the
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entire piping analysis problem is reanalyzed to determine a new load
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distribution of all supports.
(Unresolved Item URI 88-12-01.)
3.4 U-bolt Allowable Loads
TVA's design criteria for piping supports, SQN-DC-VC-24.2, Figure I-7 has a
table of load ratings for U-bolts.
Based on review of the pipe support
criteria in September 1987, the staff questioned the basis for the allowable
loads used for U-bolts. TVA provided the basis for the U-bolt allowable loads
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as a Browns Ferry test report, CEB-85-06.
According to CEB-85-06 the load
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ratings were developed based on the winter addenda to the 1983 ASME Code. The
staff review of CEB-85-06 questioned whether the allowable loads have been
appropriately derived using the ASME Code criteria.
TVA has not responded to
this question. TVA presented an additional basis for the allowable U-bolt
loads. The additional basis included a comparison of the allowables with a
load rating procedure using the factor of safety quoted in industry standard
MSS SP-58. This standard is referenced in ANSI P31.1 - 1967. The load rating
calculations also included a check of deflection criteria in SQN-DC-V-24.2.
.The staff did not agree with the appropriateness of the deflection criteria
used for the lateral load test. TVA had used an average value from tests of
cinched and uncinched U-bolts.
Based on further staff questions, TVA
responded that they did not have cinched U-bolts in field installations. TVA
typical drawing 17 W586-3, Revision 23 showed a gapped U-bolt configuration.
A TVA check at the field during the inspection also identified that a gapped
U-bolt configuration was typical. TVA then demonstrated that the U-bolts
could meet a reduced allowable based on test data using the uncinched U-bolt
tests. This was considered acceptable by the staff for restart. The staff
will have further followup reviews of TVA's development of standard component
support allowable loads after restart.
(UnresolvedItemURI 88-12-02.)
3.5 Seismic Load Combination for Piping
During the staff's IDI review during the week of February 1, 1988, it was
identified that TVA's criteria for evaluating spatial earthquake responses for
piping analysis used a two directional square root of the sum of the squares
(SRSS) procedure.
The method of combining these responses originally stated in
FSAR Section 3.7 was a vectorial combination. TVA in Amendment I to the UFSAR
(1984) clarified that the spatial combination was SRSS. However, the staff SER
(1979) stated that the combination was absolute sum. Since the change to the
FSAR was evaluated by TVA under a 10 CFR 50.59, there was no evidence that the
NRC staff had reviewed and approved this spatial combination method. TVA's
method of spatial combination is less conservative than the method required by
current licensing criteria in Regulatory Guide 1.92.
In order to assess the
significance of this issue, the staff requested that TVA evaluate a sample of
five piping systems using absolute sum and SRSS for the two directional
The results were documented in a letter from TVA (Reference 42).
response.
This study showed a difference of approximately 10% between the two combination
methods.
In addition, a recent review of another nuclear facility revealed
that a two directional SRSS of spatial responses had been reviewed and approved
by the NRC staff.
Since this method had been previously reviewed and accepted
by the NRC staff and TVA had provided a clarification of the method in an FSAR
update the staff accepted the results of TVA's sample study as sufficient to
resolve this issue. Therefore, this issue is considered resolved for Sequoyah.
However, since the two directional SRSS does not meet current licensing
criteria, TVA should upgrade this analysis method for any future criteria
changes that deviate from the original FSAR design criteria.
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3.6 Employee Concerns Element Report 220.11
The staff evaluation of Element Report 220.11(B) summarizes the status of
TVA's corrective actions to address the issue of thermal expansion of
restrained pipe support structural members at SQN. The evaluation of the
element report specified an open issue that the calculations of restrained
thermal growth pipe supports 2-HIM-101, -102 -103 and 2-H36-111 should be
reviewed when these calculations were complete.
,
On November 30, 1987, TVA revised subsection 6.3.13. Environmental Thermal
Effects, of design criteria SQN-DC-V-24.2 to permit critical buckling stresses
which exceed 0.9 of the material yield stress to be handled on a case-by-case
basis. This revision to subsection 6.3.13 represents a relaxation of the
criterion which the staff originally reviewed and accepted as part of TVA's
calculation regeneration program. The staff raised a concern with TVA's
modification of the design criteria to eliminate the buckling allowable
limits. TVA responded that buckling lim'.s had not been exceeded and that
the actual criteria that had been used was an inelastic evaluation of the
end connections (Reference 39). This type of evaluation is consistent with
,
criteria used for structural evaluation of environmental thermal effects and
is acceptable. TVA also comitted to revise design criteria SQ"-0C-V-24.2 to
allow inelastic analysis on a case-by-case basis and eliminate the statement
on exceeding buckling allowables.
3.7 Employee Concerns Element Report 221.2
The staff review of Element Report 221.2(B) identified that TVA had not
followed the recomendations in civil engineering report CEB 80-58 for
evaluating the zero period acceleration (ZPA) effects for the containment
design basis accident (OBA) loads,
in response to this concern, TVA evaluated
a sample of five piping systems attached to the steel containment vessel.
During the inspection, TVA presented the results of an evaluation of the
containment penetrations for this load case. This evaluation demonstrated
that the penetrations were adequate for the increased loads due to the DBA ZPA
effects. The staff requested the results of the rest of tt.e piping analyses
including the supports. TVA stated that due to the low level of deflection
caused by the ZPA loads the supports would not be loaded due to the support
construction gaps. The staff disagreed with TVA's reasoning on this issue.
TVA was attempting to use two contradictory sets uf assumptions for the
analysis,
in determining that the piping had a rigid response, TVA assumed the
supports were active. Then TVA assumed the supports were not active for the
loads generated assuming the piping response was rigid.
In response to the
staff concern, TVA completed an evaluation of the supports where loads
increased by more than.10% on the five sample piping analyses (Reference 40).
The results of this evaluation demonstrated that the supports met either the
interim or long term criteria. This sample evaluation was acceptable for
restart. TVA should complete the evaluation of DBA ZPA effects for the
remaining piping systems as a post restart effort.
(Unresolved item URI
88-12-03.)
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- 3.8 Design Basis Accident Response Spectrum Generation for Steel
Coritainnent Vessel
In its safety) evaluation on the use of ASME Code Case N-411 damping (Refer
damping (for evaluation of the piping systems attached to the steel containmen
vessel
SCV) under the load caused by the containment vibratory motions
associated with a design basis accident (DBA) is acceptable provided the DBA
response spectra at various locations on the SCV have been generated by con-
servative analysis techniques. TVA's generation of the DBA response spectra
for the SCV was documented in Report No. CEB-86-20-C, R0, entitled, "Sequoyah
Nuclear PlantgDesign Basis Accident Non- Axisymmetric Pressure Loading Dynamic
& Static Analysis of the Steel Containment Vessel and Response Spectra for
Attached Equipment."
The staff reviewed the TVA report during the inspection. The staff assessment
of the adequacy of the DBA response spectrum generation is discussed below:
1.
Analysis Model and Input - The containment was represented by a fixed base
axisymmetric model containing shell elements. The structural damping was
1% of critical for the DBA dyncmic analysis. The analysis model and
damping value are acceptable.
The DBA pressure transients were generated by Westinghouse assuming any
one of the six hot or cold legs could break in a guillotine manner, and
hence the spatial distribution of the pressure transients was non-
axisynnetric. All pressure transients reached the steady-state pressure
of about 12 psi at no late.r than about 0.9 seconds after the initiation
of the accident. Based on its review, the staff found the input pressure
transients acceptable for the dynamic analysis and generation of DBA
response spectrum.
2.
Structural Analysis and Generation of Response Spectrum - For the analysis
of the axisynnetric containment model under pressure transient TVA used
the SUPERSHELL computer code that was originally developed by Ghosh and
Wilson at the University of California, Berkeley.
It was specifically
developed for the analysis of an axisymmetric model subjected to non-
axisynnetric loadings. The staff reviewed the verification of computer
code and found that the code is acceptable for the DBA dynamic analysis.
SUPERSHELL outputs displacement response time histories at any specified
structural node. However, it only outputs acceleration response time
histories at those nodes located at the 0-degree aximuth.
To generate
the response spectrum at the nodal points of interest, TVA first double-
differentiated the displacement time history at the specified node
to convert it into an acceleration time history, and then computed the
response spectrum from the obtained acceleration time history. Both the
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numerical double differentiation and spectrum computation were done by a
TVA in-house computer code NUPRPOS.
In 1986, TVA verified the double
differentiation algorithm in SUPRPOS by applying it to one full cycle of
harmonic motion that was digitized into twenty equal time steps. The
output was in good agreement with the analytical solution. The staff
considered this verification inadequate with respect to the DBA analysis
because of the uncertainties involved when the double differentiation was
applied.
For the purpose of restart, TVA was requested to repeat their
1986 verification except with the harmonic motion re-digitized in ten
and eight time steps, respectively. The output from both runs compared
closely with the analytical solution, and the staff considered TVA's
revised verification sufficient for restart. Because the structural
response in the case of the DBA analysis is transient in nature and
contains more than one cycle of oscillation, the staff requested a verifi-
cation of SUPRPOS after the restart of Unit 2 by comparing the spectrum
generated directly from the SUPERSHELL acceleration time history output
at the 0-degree azimuth nodes to the corresponding spectrum generated
indirectly from the displacement time history output.
The staff also reviewed the response spectra generated at the various
locations on the containment, and expressed a concern with the adequacy of
the 0.9-second duration which TVA adopted in the analysis. The concern
was that at some locations the response might not have reached its real
maximum yet when the analysis was cut off at the end of 0.9 seconds
because the containment was very lightly damped (1% damping).
The staff
accepted the existing DBA response spectra for restart, but requested TVA
to verify on a long tenn basis that the existing DBA response spectra did
not miss their real maximum due to the 0.9-second cutoff during the
spectrum generation.
,
In conclusion, the staff concluded that the existing DBA response spectra are
acceptable for restart. TVA was requested to take two post-restart actions:
(1) verify the adequacy of the double differentiation technique adopted by
SUPRPOS by comparing the response spectrum directly generated from the
SUPERSHELL acceleration time history at the 0-degree azimuth nodes with the
corresponding response spectrum generated from the SUPERSHELL displacement
time history, and (b) verify that the existing OBA response spectra did not
miss the real maximum response due to the analysis being cut off at the end
of 0.9 seconds. TVA comitted to perform ttsse post-restart actions in a
letter dated March 2, 1988 (Reference 41).
(Unresolved Item URI 88-12-04.)
3.9 Deficiency Evaluation of Voids in ERCW Pumphouse Support Cells
The ERCW pumping station is supported on two overlapping concrete filled sheet
pipe cells founded on rock.
The concrete in the cells was placed using tremie
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concrete methods. The sheet piling was originally only considered as a
formwork for placing concrete and was not considered as part of a Category I
structure. After completion for the concrete placement, the concrete was
cored and geophysical logs of the concrete were developed. The cores
indicated cavities and areas of sof t con: rete. One of the cores indicates
that approximately 7 feet of void exits between the two northern most intake
liners. The liners are about 10 feet apart at this location.
In order to justify the strength of the concrete in the ERCW act.ess cells, TVA
submitted the data to the staff. The data was submitted in a report titled,
"Rock and Concrete Investigation Report" dated January 1978. The report
indicates that a total of eight holes were drilled. Two holes were core
drilled and six were percussion drilled.
Between elevation 620 feet and
elevation 630 feet (the bottom of the cells), six of the cores indicated
cavities. Sonic cross hole measurements were taken on the two northern and the
two southern holes. These measurements indicated that the cavities were not
continuous at these two locations. Thif, is not surprising since one boring at
each location did not indicate a cavity. Also these holes are not in the area
of concern.
An analysis was performed assuming areas of sound concrete, soft concrete and
voids or gravel pockets.
The analysis assumed the following:
'
1.
Concrete does not take tension.
2.
Areas assumed to be voids or gravel pockets do not take any load.
3.
Areas assumed to be soft concrete would have a reduced modulus and a
reduced allowable stress.
The results of the analysis indicate that the stress levels are acceptable when
subjected to SSE loads.
This analysis gives reasonable assurance that the ERCW
pumping station will not fail or be subject to excessive deflections when
subjected to the postulated SSE event. However, the staff requested that TVA
confirm the extent and size of the cavities or gravel pockets as a
post-restart action.
TVA has agreed to perform additional post-restart evaluations of the ERCW
pumping station concrete to address the staff concern (Reference 36). The
following post-restart actions were agreed to by TVA:
1.
An evaluation program will be submitted to the staff for review and
approval.
Special emphasis will be placed on determining the extent and
size of the cavities or gravel pockets.
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Once the as-built condition is determined TVA will:
2.
Review the seismic cualification of ERCW equipment,
a.
Re-evaluate effect of ERCW pumping station deflections on ERCW
b.
piping,
Confirm that the design requirements of the OBE concurrent with a
c.
water level at elevation 704 feet are satisfied.
This item is no longer considered a re-start issue for Sequoyah Unit 2 but
remains open until the above comitments are completed.
(Unresolved Item
URI 88-12-05.)
4.
OBSERVATIONS FROM CALCULATION REVIEW EFFORT
The inspection effort focused on the resolution of open restart items from
previous design calculation, OBVP and IDI inspections in the civil engineering
Although the restart open items were resolved during the inspection,
These open items are listed
area.
several post restart open items were identified.
in Appendix D of this report.
5.
REVIEW 0F PREVIOUS INSPECTION FINDINGS
The inspection reviewed TVA's responses and corrective actions documented the
following previous NRC inspections associated with the design calculation, DBVP
and IDI reviews.
50-327/86-27 and 50-328/86-27
50-327/86-38 and 50-328/86-38
50-327/86-45 and 50-328/86-45
50-327/86-55 and 50-328/86-55
50-327/87-06 and 50-328/87-06
50-327/87-14 and 50-328/87-14
50-327/87-27 and 50-328/87-27
50-327/87-31 and 50-328/87-31
50-327/87-48 and 50-328/87-48
50-327/87-64 and 50-328/87-64
50-327/87-74 and 50-328/87-74
50-327/88-13 and 50-328/88-13
The review of the open items from these inspection reports are documented in
Appendices A, B and C.
Although only open items in the civil engineering area
were reviewed during the inspection, this report documents the restart closure
Some of the items
of all open issues from the previous inspection reports.
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addressed in these appendices were closed in previous inspection reports based
on ongoing licensing reviews at the time of tre inspection report. A
discussion of these items are included in the sppendices to document the basis
for restart closure. Tables A.1 and B.1 contain a complete listing of all
' items identified for followup review during the design baseline verification
program and calculation program inspections conducted by the NRC and identifies
the inspection reports the items were discussed, closed or transferred for
licensing review.
Inspection Report 86-27 listed five items as observations
that did not require a response from TVA. These items were not included in
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Table B.1.
6.
MEETING SUMMARIES AND REFERENCES
A summary of attendeos at the entrance and exit meetings and a list of
references are provided in Appendix E.
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APPENDIX A
LICENSEE ACTION FOR PREVIOUS CALCULATION
REVIEW PROGRAM INSPECTION FINDINGS
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Inspection Report No. 87-06
(Closed) Observation GEN-2, Q Q Operability Determinations
Inspection Report 87-06 identified the concern that the new corporate QA
procedure for corrective actions NQAM, Part I, Section 2.16 required that
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component operability be determined by its technical specification
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safety-related function rather than its design related function. The
inspection report also identified that Nuclear Engineering Procedure (NEP) 9.1
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was in the process of beirg) revised to agree with the new corporate procedu
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TVA's response (Reference 4 stated that NQAM, Part 1, Section 2.16
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"Corrective Action," Attachment 5, was revised to address this concern.
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Inspection Report 87-27 noted that TVA withdrew the recent revision to NEP 9.1.
4
The resolution of this item was further discussed with the OSP staff
'
(Reference 32) and based on those discussions TVA agreed to revise NQAM 2.16
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and Attachment 5, "Guidelines for Potential Operability Determinations" to
Sequoyah Instruction AI-12. These revisions were initiated by a memorandum
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from Capozzi to Kazanas dated January 12,1988(RIMSB05880112002). TVA
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subsequently revised NEP 9.1 on 3/31/88 to address the staff's concern. The
P
revision assures that component operability assessments are performed for cases
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where design criteria are not met. This observation is closed.
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.(ClosedRestart)
Observation MEB-3, Waterhammer
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Inspection Report 87-06 identified that TVA had performed an analysis of a
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feedwater waterhusner due to a postulated pipe break upstream of the feedwater
check velve but had not formally issued the analysis. TVA's cesponse
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(Reference 4) stated that the check valves'and piping had been designed to
withstand the pressure associated with the waterhamer and that further
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analysis of this event was not justified.
Inspection Report 87-27 provided
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the detailed chronology of TVA's internal correspondence on the feedwater
"
waterhammer issue and requested TVA's justification for not issuing an analysis
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of the piping system when it had been identified by TVA documentation as a
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licensing comitment.
Inspection Report 87-64 stated the issue was still under
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review by CEB and TVA would provide a revised response to the observation.
TVA's revised response (Reference 33) still contends that the original evalua-
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tion of the check valve and piping for the waterhamer pressure met TVA's
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licensing comitments and was adequate. However, TVA performed an analysis of
!
the feedwater piping using forcing functions developed for the waterhamer
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transient. The piping was analyzed using a three-dimensional inelastic finite
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element analysis and these results were compared to criteria contained in
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- Appendix:F of1the ASME Code. The results of this analysis indicated that
- piping' supports may fail ~or defortiduring.the postulated transient, however,
the results of the analysis'also demonstrated piping system integrity would be
--
maintained.(ASME Code Appendix F limits were met). The-results of this
analysis are considered: acceptable by the staff for Sequoyah restart. This
- observation is closed for restart. The staff will be performing additional
review of the details of this analysis as a post-restart effort.
(Unresolved
Item URI 88-12-06.)
(Closed) Observation MEB-6, Component Cooling Water System Desian Pressure
Inspection Report 87-06 identified a concern with TVA's calculation of system
design pressure for the component cooling water system. TVA's response
(Reference 4) stated a calculation of s revised system operating pressure had
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been completed and the operating pressure remained below the design pressure.
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- Inspection Report 87-27 identified a number of open issues with TVA's revised
calculation. Additional review of this item documented in Inspection Report
87-64.
Inspection. Report 87-64 requested the following items to be addressed
in a revised calculation as a confirmatory item.
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1.
A design pressure calculation based on:
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(a)
'a static head produced by the surge tank water leve, it the high~ end
of the normal level control range,
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(b) the lowest pump flow (highest total dynamic head) that can occur for
any normal operating mode of the CCS, and
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. (c) the lowest expected operating coolant temperature.
2. .
TVA should show by calculation that CCS pressure variations meet the
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requirements of Paragraph 102.2.4 of the Power Piping Code B31.1.0 - 1967.
The team considered that events such as closure of the normal surge tank
vent and increase in surge tank pressure to its relief valve setpoint plus
accumulation can be considered pressure variations provided the event
n ets the spirit of the phrase "occasional periods of noeration for short
periods" contained in B31.1.0 - 1967 and is not permined to be a normal
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mode of operation.
3.
TVA should conduct a review to determine if all components meet the
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calculated design pressure.
Inspection Report 87-64 stated that TVA should revise the FSAR and notify the
NRC if the calculation of design pressure calculation for the component cooling
,
water system used assumptions other than pump shutoff head.
!
TVA's response _ (Reference 24) states the revised calculation shows that no
portion of the system exceeds design pressure under normal operating corditions
A-2
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,,,#-
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and a separate calculation shows that CCS pressure variations do not exceed
the allowances of ANSI B31.1.0 - 1967.
In addition, TVA comitted to revise
FSAR Section 9.2.1.2 in the April 1939 update. The staff considers these
actions sufficient to close this observation.
(Closed) Observation EEB-2, Breaker Coordination
Inspection Report 87-06 identified an error with the corrective action taken by
TVA to resolve breaker coordination problems for the 480V diesel generator and
TVA's response (Reference 4) stated
essential raw cooling water system boards.
that the correct corrective action for this observation would have been obtained
by following the ECN procedure.
Inspection Report 87-64 found TVA's corrective
action described in ECN6883 acceptable. The inspection report also agreed with
the post-restart classification for the completion of the corrective action.
The inspection report held the observation open pending TVA's CCTS commitment
to complete the corrective action. TVA's response (Reference 24) stated that
the corrective action (ECN 6883) was complete. This observation is closed.
(Closed) Observation CEB-1, Rigorous Piping Analysis N2-67-8A
Inspection Report 87-06 identified three issues associated with this piping
analysis. TVA's response (Reference 4) agreed with the observation findings
Report 87-27 closed the first
and provided the proposed corrective actions. The inspection report transferred
two items based on TVA's corrective actions.
the third item to licensing for review.
The third item of Observation CEB-1 stated that procurement documents for a 1
by 2 inch valve exempted the valve from seismic qualification requirements which
are included in the SQN FSAR and TVA design criteria.
In response to the
observation TVA issued CEB Report 87-10C and Condition Adverse to Quality
Report (CAQR) SQF870070. Report 87-10C documents the rationale for generic
qualification of small bore hand operated globe and gate valves.
The CAQR
documents the deficiency and identifies the following corrective actions:
1.
Verify and document seismic qualification of the specific valve
identified.
2.
Review other valve specifications that could have resulted in similar
deficiencies.
3.
Verify and document seismic qualification of valves identified by action
item 2.
Corrective action item 1 has been completed. A copy of the qualification
calculation was reviewed during the inspection. At the time of the inspection
correction action item 2 was nearly complete and few, if any, additional valves
were expected to be identified. Completion of the documentation for this
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action has been designated as a post-restart effort. Corrective action item 3
has proceeded based on the unofficial list generated by action-item 2.
The
majority of valves and all "seismically sensitive" valves (i.e., valves with
extended operators such as air or motor operated valves) idantified by item 2
have been evaluated. Completion of the documentation for this action has also
been designated as a post-restart effort. TVA's completed corrective actions
are considered adequate to address the safety concern with the seismic
qualification of valves. The staff considers completion of the documentation
as a post-restart effort acceptable. This item is closed.
(Closed) Observation CEB-2, Structural Steel Sizing Calculations
(Closed) Observation CEB-3, Structural Steel Details
(Closed) Observation CEB-4, Platform Steel Calculations and Drawings
(Closed) Observation CEB-5, Revisions to Steel Platform Calculations
(Closed) Observation CEB-6, Seismic Loads for Steel Platforms
Inspection Report 87-06 raised several concerns with the structural design
adequacy of steel platforms located in different safety-related buildings.
L
These were contained in observations CEB-2 through CEB-6. TVA's response
(Reference 4), stated that Significant Condition Report (SCR) SQN 8711 was
1
revised to address the concerns and provided a discussion of TVA's corrective
actions.
Inspection Report 87-27 noted that TVA's corrective actions had not
been completed.
Inspection Report 87-64 discussed TVA's corrective actions to demonstrate the
TVA selected six platforms for
structural design adequacy of the platforms.
reanalysis.
Three of these platforms are located in the auxiliary building and
the other three were selected from the reactor building.
All platforms were
I
walked down by T' \\ engineers to obtain as-built information, which was later
used in tha reanalysis. The inspection report stated that TVA Calcolation B25
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870926 805 which contained the reanalysis of the auxiliary building platform at
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elevation 724'-3" w reviewed. The inspection report also concluded chat
TVA's approach for d uonstrating structural design adequacy of the steel
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platforms was acceptable, however, the inspection report identified three
concerns that were left as open confirmatory items.
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During this inspection, TVA's resolution for these three concerns was reviewed.
The results of the review of the three concerns is discussed below:
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1.
TVA-used 0.0 psf live load in the reanalysis of the steel platforms when
combined with seismic loads. However, TVA has issued an administrative
control program (B25 871127 009) to restrict the amount of live load that
can be imposed on safety related platfoms during plant operation. The
restriction states:
"When maintenance or repair activity is required on
safety-related platforms which impose loads of more than 1000 pounds for
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> duration or more within any ten feet square area (100 square
feet) or less, prior evaluation must be made by Division of Nuclear
Engineering (DNE) to determine the effect of that loading for all
applicable design criteria loading combination.
If this evaluation
concludes that the requirements of the loading combination can not be met,
temporary supports or modifications will be used during the maintenance or
repair activity to support the tempcrary loads.
In addition, platforms
shall not be utilized for long-tenn storace of materials or equipment."
TVA qualified certain connections by torsional tests performed at
2.
Singleton Materia'.; Engineering Laboratory as shown in TVA document B46
870904 001. Because these tests are not standard tests and are not
covered in the AISC code, an independent review of the tests was
performed by Dr. Edwin G. Burdette (TVA Consultant) of the University
of Tennessee. He has confirmed the adequacy of the testing procedure
and the applicability of the test results.
3.
TVA concluded that the bending stresses in the weak axis of one beam
exceeded the FSAR stress limits for OBE load case. However, the maximum
weak axis bending stress for the SSE load condition is below the yield
TVA will modify this beam for long term operation of the plant.
stress.
The results of this evaluation of the six steel platforms also showed that
some self-drilling anchors did not meet the long term safety factors, however,
the safety factors obtained were within the NRC approved safety factors for
interim operation. These results are included in a TVA letter to the NRC
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(Reference 24). As stated by TVA, modifications will be installed after
,
restart for all cases which do not meet the long term criteria.
The staff found the actions taken by TVA to resolve these observations to be
adequate for restart. Therefore, this observation closed.
(Closed) Observation CEB-12, Use of Variable Damping for Conduits
Inspection Report 87-06 identified that TVA was using a variable damping value
for qualifying conduit supports instead of the value shown in Table 3.7.2.4 of
the Sequoyah FSAR. TVA's response (Reference 4) stated test data existed to
justify the damping values used in the evaluations and that the FSAR would be
revised to add new damping values for metal conduit.
Inspection Report 67-27
closed the observation and transferred the issue to OSP for review.
This
issue was addressed in a staff safety evaluation report on variable damping
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(Reference 55). Based on the staff safety evaluation ~ report, this observation
is closed.
Inspection Report & . 87-27
(Closed) Observation MEB-10, Loss of Station AC Power Calculation
Inspectioa Report 87-E7 raised a concern with the lack of a calculation or
other basis which substantiated the adequacy of HVAC to maintain adequate
ambient temcerature for essential equipment during a loss of station ac power.
TVA's rescor,se (Reference 20) stated that this concern was beyond the design
basis of the plant. However, TVA provided additional information to address
the concern.
Inspection Report 87-64 stated that TVA's original response was
inadequate. The inspection report also found TVA's subsequent acknowledgement
cf a comitment to maintain hot shutdown following a loss of station ac power
for a two hour period adeauate to resolve the issue pending TYA's submittal
of a revised response. TVA provided a follow-up response to this observation
(Reference 24) documenting the basis for meeting the two hour comitment.
Therefore, this
TVA's response is considered adequate to resolve this issue.
observation is closed.
(Closed) Observation EEB-7, HVAC Temperature and Flow Process Safety Limits
Inspection Report 87-27 identified that the MEB 480 volt board air handling
unit temperature switch setpoint calculation did not establish process safety
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limits'for a large number of safety-related HVAC temperature and flow
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measurements. TVA's response (Reference 20) stated the issue had been
addressed by a new calculation.
Inspection Report 87-64 identified an
additional issue with TVA's corrective action. This issue involved a conflict
i
between the lower process safety limit temperature used for the fifth vital
l-
battery room heater control and another calculation for battery operability.
l
The inspection report recommended that the lower process safety limit
l
'
calculation be revised to show the correct minimum value.
TVA's response
'
(Reference 24) comitted to reevaluate the fifth vital battery room temperature
30, 1989. This resolution is considered
as a post-restart effort by June
acceptable since the calculation revision will not affect the plant hardware.
i
'
Therefore, this observation is closed.
(Closed) Observation EEB-9, Containment Electrical Overcurrent Protection
Inspection Report 87-27 identified a concern with the overcurrent trip settings
TVA's
used to protect the circuits of penetration assemblies Nos. 52 and 53.
response (Reference 20) stated that no corrective action was required based on
Inspection Report 87-64 provided
manufacturers test data for the penetrations.
a review of calculations TVA had performed using vendor's test data to
demonstrate the adequacy of the penetrations. The inspection report identified
A-6
.
.
..,
,
s
.
an error in the ambient air temperature used in TVA's calculations and
recomended that TVA revise the calculation. This was not considered a restart
issue based on calculations performed by the inspection team but was held open
TVA's response
p(ending TVA's comitment to revise the calculation. Reference 24) to the' insp
calculation by August 1, 1988.
Based on this comitment this observation is
closed.
(Closed) Observation EEB-10, Pump Start Time Delay Relay Setpoint Calculations
Inspection Report 87-27 identified that no time delay setpoint calculations had
been prepared by TVA for both the 15 to 25 second and 0.5 second time delay
relays used in pump start circuits for the ERCW, CCS, and AFW systems. TVA's
response (Reference 20) stated the issue was being addressed. . Inspection
Report 87-64 identified that TVA subsequently revised procedure PM 86-02,
"Method for Electrical Calculations," to specifically list a time delay relay
category in the set of required calculations. TVA identified 38 specific time
delay relay applications reouiring setpoint calculations, and designated 12 of
these as post-restart.
Inspection Report 87-64 concurred with TVA's
designation of the 12 post-restart items and held the observation open pending
TVA's
TVA's correspondence confirming entry of the calculations in the CCTS.
response (Reference 24) to the inspection report comitted to complete the
30, 1989. TVA's letter to the NRC
p(ost-restart calculations by JuneReference 43) contained this item on the CCTS.
observation is closed.
(Closed) Observation EEB-11, Component Cooling System Setpoint Calculations
Inspection Report 87-27 identified that CCS flow alarm accuracy values had been
discussed between EEB and MEB, but justifications for selecting particular
values were not documented in an MEB calculation (RIMS No. B44 870602 001).
TVA's response (Reference 20) stated that the flow alarm setpoints were not
needed for the hot shutdown of the plant but they were considered desirable.
TVA's response stated that demonstrated accuracy calculations for these alarm
setpoints were planned as a post-restart item.
Inspection Report 87-64
'
concurred with TVA's post-restart designation and held the observation open
TVA's response
p(ending TVA confirmation of entry of the item in the CC1S. Reference 24)
TVA's letter to the NRC (Reference 43) contained this item on the CCTS.
Based on this commitment this observation is closed.
(Closed) Observation CEB-13, Regenerated CEB Pipe Support Calculations
Inspection Report 87-27 identified that CEB's calculation for pipe support
H10-635 demonstrated that the pipe support failed when friction forces were
considered, and that CEB did not document this deficiency on the calculation
A-7
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cover sheet or in the CAQR which CEB subsequently prepared.
In addition, the
ei
CEB calculation for pipe support H10-1219 did not include a thermal check of
the pipe support and CEB did not note this as an unverified assumption on-the
calculation cover sheet or on CEB's pipe support calculation log.
TVA's
response (Reference 20) stated the supports were being reviewed as part pipe
support calculation effort.
Inspection Report 87-64 identified that these
. calculations were being regenerated as part of the pipe support calculation
effort and held the observation open pending)TVA's confirmation the actions
were complete. TVA's response (Reference 24 stated the calculations had been
completed and that pipe support H10-635 required a modification that was
ccmpleted to meet the design criteria.
Based on TVA's corrective actions this
observation is closed.
(Closed) Observation CEB-15, Technical Adequacy of Miscellaneous Structural
'
Steel _
Inspection Report 86-27 stated that TVA reviewed 54 randomly selected features
to determine the design adequacy of miscellaneous structural steel at Sequoyah
Nuclear Plant Unit 2.
The inspection report raised a concern that this initial
sample size was not large enough to represent the total population of miscel-
igneous structural steel. TVA's response (Reference 20) stated the sample
evaluation had identified five CAQR's that were being evaluated for operability
.
requirements and that miscellaneous steel calculatior.s would be reviewed and
revised as necessary after restart.
Inspection Report 87-64 identified that
TVA hhd increased their sample size to review 38 ado 1tional calculations and
also planned to select 60 equipment support calculations for review to deter-
mine whether the appropriate vendor loads were used in the design. TVA stated
this effort was scheduled for completion by November 30, 1987.
The inspection
report found the sample size which TVA selected to determine the design
adequacy of miscellaneous structural steel acceptable. The inspection report
identified that interim criteria were being prepared by TVA specifically for
the miscellaneous steel members which deviated from the FSAR requireaents.
The inspection report left the observation open pending TVA's confimation
that the evaluation was complete.
The staff reviewed TVA's criteria for the evaluation of miscellaneous structural
steel. TVA used design criteria SQN-DC-V-1.3.3.1, which is ccmpatible with the
FSAR loading combinations and allowable stresses.
TVA's review of the design
records showed that there were differences between the design loads and vendor
loads used for the design of equipment supports.
TVA issued CAQRs SQP870188,
SQP870209 and SQP8702109 to resolve these discrepancies. The resolution for
CAQR SQP870188 required a modification to the containment spray heat exchanger
support as docuWnted in TVA calculation B25 880131357. This calculation was
reviewed by the staff and found to be acceptable.
A-8
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.
As for generic evaluation related to inconsistencies in the use of vendor
loads, TVA selected 60 equipment support designs and evaluated them to
determine whether the appropriate vendor loads have been utilized in the design
of equipment supports. This evaluation was performed in conjunction with the
TVA response to NRC IDI Ceficiency 04.6-1.
The staff's review of the 60
equipment support evaluations are covered under this 10I deficiency.
The staff found the actions taken by TVA to resolve t;e concerns with the
adequacy of miscellaneous structural steel adequate and therefore considers
this observation closed.
(Closed) Observation CEB-16, Conduit and HVAC Duct Support Calculations
Inspection Report 87-27 idontified that CEB's review of recently regenerated
conduit and HVAC duct support calculations identified numerous discrepancies
-
between the calculations and the design criteria. TVA's response (Reference 20
) provided the details of TVA's corrective actions.
Inspection Report 87-64
noted that TVA had not completed the corrective actions and held the
observation open for further review.
TVA's response (Reference 24) provided
the results of the completed corrective acticns. These results are discussed
in detail below.
HVAC Ducts and Duct Supports
In order to resolve CAQR SQT870843, TVA has selected five worst case duct
systems which were qualified by computer analyses. Gilbert / Commonwealth (TVA
Consultant) was contracted to perform analyses on the five duct samples. The
results of this evaluation is contained in Gilbert / Commonwealth (G/C) report
for Task R0006.
The staff review of this report found that TVA used the 7% damped amplified
response spectra (ARS) to determine the seismic loads for both OBE and SSE in
four of the five duct samples, namely 1, 3, 4 and 5.
This is in violation of
the FSAR in which the use of 2% and 5% damped ARS was required for steel
structures with bolted connections under OBE and SSE loading cases.
During the
inspection, TVA performed additional calculations using the loads calculated
from the 5% damped ARS to show that the HVAC ducts and duct supports meet the
restart criteria.
5% damping for SSE loading condition is acceptable to the
staff for the HVAC duct evaluation. The staff reviewed these preliminary
calculations and concluded that the sampled HVAC systems met the restart
criteria requirements.
However, the staff noted that these preliminary
calculations should be finalized prior to restart. TVA submitted the final
calculations (B25 880224) for staff's review (Reference 35). The staff
reviewed these calculations and found them acceptable for restart.
TVA should
qualify these four duct samples plus additional samples from other duct systems
to long term criteria post-restart.
A-9
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!The staff reviewed the support calculations (RIMS Nos. 823 871120 450, B25
871120 453 through B25 871120 455) related to sample 2 and also reviewed the
preliminary G/C calculations for. sample 2 in which overstresses in connection
welds and drilled-in . anchors were identifed. The staff review of the G/C
calculations found that the overstressed welded connections are adequate to
transfer the axial loads and would act as pinned connections rather than fixed
connections, as modelled in the computer analyses. The staff requested that
these preliminary calculations for qualifying the 'overstressed welds be revised
'
to reflect the pinned end connections prior to restart. TVA' submitted the
finalized G/C calculations (RIMS Nos. B25 880224 308 through 825880224313)
which considered the welded connections as pinned connections (Reference 35).
The staff reviewed these calculations and found the results acceptable. Also,
the staff accepted that the drilled-in anchors for. sample 2 met a short term
safety factor of 2 0.
However, for post-restart, TVA needs to qu111fy all
i
these anchors to the long term criteria and the~ G/C calculations should be
revised to ccincide with the evaluation results as shown in Tables 1 through 6
of their report for Task R0006. The. documentation received from TVA
(Reference 35) showed that TVA has comitted to qualify the drilled-in anchors
used in HVAC supports for compliance with the long term criteria and to revise
the G/C calculations for sample 2 to reflect Table 2 of Report R0006, after
restart.
,
.
In conclusion, the staff found that' the five worst case duct samples meet the
I
restart criteria. TVA should compute the evaluation of these five duct samples
' to the long-tenn criteria and select additional samples from other duct systems
l
to evaluate to the long-term criteria.
(UnresolvedItemURI 88-12-07.)
l
!
Conduit and Conduit Supports
l
Sequoyah CAQR SQT870626, Revision 1, identified several issues regarding the
adequacy of the seismic design of the conduits and conduit supports. The major
issue was the compliance of the existing seismic design with design
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L
requirements specified in design criteria-SQN-0C-V-13.10.
In addition, the
(x
following six specific issues were identified:
'
,
,
Adequacy cf conduit runs containing one-hole finger clamps;
1.
2.
Adequacy of conduit runs containing cast iron parts;
3.
Adequacy of conduit systems with rigid and flexible supports intermixed;
s
l
4.
Adequacy of conduit supports for axial loads;
l
5.
Adequacy of conduits supported on structures and/or other equipment;
l.
s
6.
Effect of differential seismic movements between structures.
To resolve the major issue, i.e., compliance with the design criteria
requirements, TVA performed walkdowns of the plant and evaluated a
A-10
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py.
.
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<
representative worst case sample of 60 supports and the related conduits based
on the restart criteria.
For the restart evaluation of these 60 sample support
cases under the SSE condition, TVA used (4) 2% damped amplified response
spectra.(ARS) for the conduit system, (b) an absolute-sum combination of the
loads from all three earthquake components, and (c) the allowables for the OBS
condition. The criteria used by TVA for this sample is more conservative than
requires (a)y accepted Icng-term criteria for the SSE loading condition
the currentl
loads from the vertical and one horizontal earthquake components; and (c) the
SSE condition allowables. According to TVA's evaluation, all sampled cases
(conduits and supports) were found to meet the restart criteria with three
exceptions. The three exceptions are (a) both the aluminum conduit and
Unistrut clamp at support AB25 were overstressed, (b) the aluminum conduit at
support AB45 was overstressed, and (c) the Unistrut clamps at support 2AE3 were
overstressed. TVA re-evaluated these three support cases using the long term
criteria for the earthquake component combination and allowables and the
existing seismic loads calculated from the 2% damped ARS. The re-evaluation
results showed that all three supports met the long term critaria, however, the
'
aluminum conduit stresses et supports AB25 and AB45 still exceeded the long
term allowables by approximately 30% and 9%, respectively.
The staff reviewed
TVh's evaluation results of supports AB25 (RIMS Nos. B41-880205-020 and
B25-880209-800), AB45 (RIMS Nos. 841-871106 and SCG2S-88-008) and 2AE3 (RIMS
Nos.B25-880209-802andB41-871106-086) and concurred with the TVA judgre.nt
that the conduit stresses would be within long term allowables if the currently
accepted 5% damped ARS were used for the input motion. Therefore, the staff
concludes that TVA's evaluation for the 60 representative sample supports and
related conduits is an ecceptable restart resolution for the major, issue
regarding the compliance of the existing conduits and supports with the design
criteria requirements. As a post-restart corrective action item, TVA will
re-evaluate the conduits at these two supports and additional samples to be
selected from the remaining conduit systems using seismic loads based on a 5%
damped ARS.
To resolve the remaining six issues, TVA applied the earthquake experience data
developed by its consultant, EQE, Inc., and concluded that the existing condi-
tions are adequate for restart. This evaluation was summarized in the EQE
report, "Seismic Evaluation of Specific Issues for Conduit Systems at Sequoyah
Nuclear Plant, Unit 2," Revision 0, August 28, 1987 (B25 871106 018). The
staff did not find the EQE earthquake experience data a sufficient basis and
j
TVA was required to base their evaluation on plant-specific analyses and/or
test data.
In response to the staff concern, TVA provided new information for
review. The staff evaluation of this information is discussed in the
following:
Issue (1) - The staff had a concern that no test data existed to confirm the
capacity of one-hole finger clamps to resist axial loads due to clamping
fri cti or. , TVA performend preliminary calculations which demonstrated that all
conduit sizes except the 4 and 5 inch diameter met restart criteria without
l
using the clamping friction. The staff concurred with this preliminary
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evaluation and requested TVA'to provide a more conclusive pre-restart
resolution for those 4" and 5" conduit runs containing one-hole clamps.
Subsequent to the inspection, TVA performed a four-activity program to
demonstrate the adequacy of 4" diameter and 5" diameter conduits at SQN that
contain one-hole clamps.
In activity (1), TVA re-confirmed that for all
conduit sizes except for 4" diameter and 5" diameter, the axial leads can be
resisted by the first one-hole clamp beyond a 90 elbow with sufficient margin.
The allowables adopted in this evaluation were 0.7 tirres the ultimate lateral
load capacities established previous 1) from testing conducted by TVA.
In
activity (2), TVA performed a 100% walkdown of the auxiiiary building floor at
elevation 690'-0" which was believed by TVA to contain the greatest
concentration of large size conduits, i.e., 4" diameter and 5" diameter
conduits. Of the 122 4" diameter cc'iduit runs identified from the walkdown,
only eigtt runs had tro or more one-h71e clamps in series on straight runs.
This walkdown confirmed TVA's assumptien used for the test that the 50' of
straight run of the 4" diameter conduit with six one-hole c'. amps was a bounding
configuration.
In activity (3), TVA evalated the three worst cases out of the
eight 4" diameter conduit runs with two or w re one-hole clamps in series
assuming that all clamps provided three directional restraints. The evaluation
results showed that all three worst cases met 'he accepted restart criteria.
In activity (4), TVA established axial load capeity of one-hole clamps for 4"
diameter conduits based on seven static tests of a 50' straight conduit run
containing six in-line clamps at 10' spacing.
All tests stopped at an applied
axial load of 2500 lbs. except for the first run in shich the test was stopped
,
'
prematurely at 1800 lbs.
The tests did not result in failure of the clamps.
l
The tests demonstrated that the one-hole clamps, when installed to the standard
TVA requirements (finger-tight plus 1/8 to 1/4 turn), can develop sufficient
I
i
axial load capacity and can be relied upon as a three directional support on
i
conduit runs.
l
The staff reviewed the information on the results of the four activity program
l
and concludes that it is a sufficient restart resolution for the issue
concerning the adequacy of conduit runs contain one-hole clamps.
l
l
Issue (2) - TVA provided test information on conduits containing straight run
l
and 90* elbow couplers made of cast iron. The tests were performed for
application to Bellefonte Nuclear Plant. The test data gave an average
ultimate strength of 30.5 ksi for the cast iron couplers, which represented a
minimum margin of safety about 1.75 with respect to the maximum conduit
stresses within the 60 worst case sample of conduit supports.
The staff
accepted this information as a sufficient resolution for the issue of cast iron
conduit bodies.
Issue (3) - The staff was informed that TVA's previous walkdown of the plant
confirmed that all Category I conduit supports, except two, are rigid supports,
and that the two flexible supports were of the rod hanger type and had already
been replaced. TVA documents SCR SQN MEB8610-R1, ECN L5599, FCR 4636 and Work
Plan 12186 verified that the replecement of the two rod hanger supports were
completed. Thus, TVA concluded that the only remaining concern of this issue
A-12
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was the seismic' interaction between Category II conduit and siesmic Category I
conducit (two o'ter one 'or II/I). TVA presented sample analyses showing that
-
"Category 11 conduits supported by rod hangers would not fail and become
free-falling missiles. Regarding the' potential for the Category II conduits to
become swinging missiles, TVA indicated that this issue was resolved viactheir
resolution of Employee Concern Element 22600, SQN-02 and -03, Interaction Item
TPW/734/005. The staff reviewed the action items and found that TVA's
resolution.for-Issue (3)isadequate.
Issue (4) - The staff was informed that only Unistrut hardware have been used
for Category I conduit supports at SQN, as stated in TVA-document'NCR W-387-P.
In addition, during its walkdown of the plant TVA verified that for some
supports which were not originally designed for resisting axial loads. there
was no washtr installed between clamps and Unistrut sections. Because of this
construction error, these supports have to resist 3-directional loadings. This
was documented in NCR SQNCEB8502, Revision 3, and SCR SQNCEB8612, Revision 1.
The axial capacity of one-hole clamps was addressed by TVA during its
resolution for Issue (1). The axial capacity of 2-hole Unistrut clamps has'
been accepted by the staff when reviewing TVA's investigations for Employee
' Concern Element 228.0-SON. Since TVA's evaluation of the 60 worst case conduit
,
supports was based on the supports being 3-directional, the staff believes that
TVA's resolution for Issue (4) is adequate because in rasolving Issue (1), TVA
had demonstrated that the one-hole clamps _also provide sufficient 3-directional
support with respect to the restart criteria.
Issue (5) - TVA provided sample analyses to show that Category I conduits would
4
not be overstressed when subjected to the loading due to the relative
,
displacement between structures and/or equipment on which the conduits were
!
rigidly supported. Only one non-Category I conduit, 1" in diameter, was
identified as the worst case in which the conduit was overstressed. The staff
found_the analyses acceptable and hence Issue (5) is adequately resolved.
_ Issue (6) - TVA identified the worst case differential movement to be about
l
m-
0.072" for a 3" diameter conduit running between the auxiliary and shield
1
l
building at Elevation 705.25'. TVA evaluated the conduit taking into account
l
the effects of both differential movement and inertia loads, and found the
conduit stress to be within the allowable. The staff found the evaluation
acceptable and hence concludes that TVA's resolution of Issue (6) is
acceptable.
'
The staff considers the issue related to conduits and supports closed.
(Closed) CEB-17, CEB Corrective Action Program Description
Inspection Report 87-27 opened this item to track the TVA's response to the
concern with the adequacy of the TVA civil engineering calculation program.
TVA's response to this concern (Reference 31) provided an update to the status
of the issues identified in the inspection report.
In addition, TVA provided
A-13
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updatedinformationonthecalc01ationprogram(Reference 19). The adequacy of
the civil program plan was addressed in the staff's safety evaluation of the
'
Sequoyah Nuclear Performance Plan.-
Based on the staff's safety evaluation
report, this observation is closed.
,
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. TABLE A.1
,;
DESIGN CALCULATION ISSUES
'
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,
Inspection Report Number
'
a
I
1
4
Observation
87-06
87-27
87-64
.
,
y
GEN-1
0
0
C
GEN-2
0
0
MEB-1
0
C
MEB-2
0
C
'
MEB-3
0
0
0
MEB-4
0
C
0
C
i
ME3 5
s
MEB-6
0
0
0
'
0
C
'
'-
MEB-7
MEB-8
0
0
C
'
j-
MEB-9
0
0
C
NEB-1
'
0
0:
C
NEB-2
0
0
C
f
NEB-3
O
C
l
- EEB-1
0
0
C
l
EEB-2
0
0
0
EEB-3
0
0
C
,
EEB-4
0
C
EEB-5
0
C
i
'
CEB-1
0
T
'CEB-2
0
0
0
CEB-3
0
0
0
t
!1
A-15
1
<,
.
.
t-
4
'
'-
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.,:
,
,
'
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,
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(
l
)
%
Inspection Report. Number
Observation
87-06
87-27
87-64
,
,
.'
.
CEB-4
0
0
0
CEB-5
0
0
0
'
CEB-6
0
0
0
CEB-7
0
C
>CEB-8
0
C
e
CEB-9-
0
C
-CEB-10
0
C
CEB-11
0
0
C
CEB-12
0
T
.
GEN-3
O
C
'
MEB-10
0
0
-
'
EEB-6
0
C
EEB-7
0
0
l-
EEB-8
0
C
EEB-9
0
0
EEB-10
0
0
[
EEB-11
0
0
CEB-13
0
0
CEB-14
0
C
CEB-15
0
0
CEB-16
0
0
CEB-17
0
0 - Item is discussed as open issue in tho Inspection Peport
C - Item is closed in the' Inspection Report
T - Item is closed in the Inspection Report based on licensing review
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APPENDIX B
LICENSEE ACTION FOR PREVIOUS DBVP INSPECTION FINDINGS
i
INSPECTION REPORT NO. 86-27
(Closed) Deficiency 03.2-2, US00 Requirement
Inspection Report 86-27 stated that ECN L5500 added extension operators and
covers to Sequoyah Unit 1 and 2 valves 67-507A installed in the essential raw
'
cooling water' system. The inspection report identified that TVA did not have
seismic qualification documentation for these new valve stem extensions.
TVA's
response (Reference 5) to the inspection report stated that approximately
300 valves with remote operators were affected by this deficiency and that SCR
SQNCEB8621 was written to address the issue.
Inspection Report 86-55
identified that remote valve stem operators and associated piping would be
t
seismically qualified for rigorously analyzed piping prior to restart and that
alternately analyzed piping would be evaluated for the additional concentrated
~
weight effects after restart.
Inspection Report 87-14 addressed the review of
TVA's corrective actions for SCR SQNCEB8621 and closed this deficiency.
Although Inspection Report 87-14 closed this deficiency, the inspection report
noted that TVA's decision to evaluate the added mass of extended valve
operators for alternately analyzed piping after restart had been previously
submitted to the NRC. The evaluation of TVA's alternately analyzed piping
program is contained in Section 2.4 of the staff's safety evaluation of TVA's
Sequoyah Nuclear Performance Plan (Reference 44).
For alternately analyzed
piping systems, the torsional effects of large motor operated and
pneumatic-operated values were evaluated as pre-restart items whereas TVA's
'
evaluation of the effects of large concentrated weights as a post-restart
effort was considered acceptable. Based on the staff's safety evaluation
report this deficiency is closed.
(Closed) Deficiency 03.3-1, Pipe Support Friction
-
Deficiency D3.3-1 identified that TVA had not considered friction forces for
pipe suppoit designs as required by the USAS 831.1.0
1967 Code. TVA stated
that the effects of friction due to thermal loads had not been considered at
-
Sequoyah and p(roposed to evaluate the effects of friction on pipe supports on
sample basis References 5, 6) after restart.
was discussed in Inspection Report No. 50-327/86-55. TVA's response
(Reference 8) to the inspection report provided additional information on the
scope of the evaluation.
Inspection Report 87-14 left the deficiency open
pending TVA's completion of the evaluation.
Subsequent to this inspection
report, TVA initiated a program to regenerate nissing pipe support calculations
for rigorous piping analyses (Reference 45). The criteria used for these
support evaluations, SQN-DC-V-24.2, included the effects of friction. The
staff review of regenerated pipe supports verified that friction was being
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considered in the pipe support calculations.
In addition, for alternately
analyzed piping, TVA has committed to perform a study of the effects of'fric-
tion forces as part of the post-restart program. This study is discussed in
the staff safety evaluation of Sequoyah's Nuclear Performance Plan
-
(Reference 44).
Based on review of the regenerated pipe support calculations
and TVA's conunitment for the post-restart alternate analysis program this
deficiency is closed.
,
(Closed) Deficiency 03.3-4, Alternate Pipe Support Criteria
inspection Report 86-27 identified that supports for field routed piping may
not have been properly evaluated for the reactions due to piping system thermal
loads. The inspection report also identified two nonconformance reports
prepared in 1982 dealing with alternately analyzed piping.
Field routed piping
is generally two inches and smaller in diameter and uses typical supports.
Alternately analyzed piping includes larger piping sizes and uniquely.
engineered supports.
Field routed pipe was included as part of TVA's
alternately analyzed piping program. TVA's response (Reference 5) to the
inspection report referenced the alternately analyzed piping program as
addressing the specific issue of thermal expansion flexibility. TVA's response
also stated that resolution of the thermal expansion issue was not a restart
item. Inspection Report 86-55 closed Deficiency D3.3-4 based on a separate
licensing review of TVA's alternately analyzed piping program. The evaluation
of TVA's alternately analyzed piping program is contained in Secticn 2.4 of the
staff's safety evaluation report of TVA's: Sequoyah Nucle 6.- Performance Plan
(Reference 44). Alternately analyzed piping systems with operating
temperatures greater than 200 F were evaluated prior to restart.
Based on the
staff's safety evaluation of TVA's alternately analyzed piping program this
deficiency is closed.
(Closed) Deficiency 04.3-3, Steam Generator Access Platform Design
Inspection Report 86-27 identified that the steam generator lower supports were
TVA's response (Reference 5)y attached platform loads arided by ECN
not evaluated for permanentl
stated the calculations would be revised after
restart. TVA's revised response (Reference 6) to the inspection report
icentified an additional concern with the documentation of the loads for
attachments to the steam generator lower supports.
Inspection Report 86-55
stated that a TVA walkdown of these supports identified additional pip 6
supports that were attached to these steam generator supports which were not
accounted for in the original uesign.
Inspection Report 87-14 stated that Westinghouse completed a re-evaluation of
these supports using the TVA walkdown information and thn re-evaluation showed
861219 601)pports are structurally adequate to carry the additional loa
that the su
The inspection report stated TVA was still evaluating the
.
attachment of the supports to the concrete.
Inspection Report 87-64 stated TVA
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- completed the analysis of. the attachments using the load information obtained
from the Westinghouse evaluaticin. The inspection report stated that the
calculations by TVA, B25 8711120 452, showed thah the attachment stresses are
within FSAR requirements.
In addition, the inspection report identified that
TVA had evaluated the crane w&ll for the additional loads obtained from the
Westinghouse analysis. This calculation, B25 870903 454, showed that the crane
wall is adequate to carry these additional loads.
The inspection report identified a concern that the walkdowns performed on the
steam generator support were not in accordance with the TVA QA requirements.
!
TVA comitted to perform walkdowns in accordance with their QA requirements to
obtain the as-built information as a post-restart effort. The inspection
report held the deficiency open pending TVA's entering this comitment on the
'
CCTS. TVA's response (Reference 25) stated that this walkdown would be
performed by Unit 2 cycle 4 refueling outage. During this inspection it was
verified that this commitment was on CCTS control number NC0-88-0008-001. This
commitment is acceptable to the staff and therefore this deficiency is closed.
(Closed) Unresolved Item U5.3-2, Sizing Calculations
Inspection Report 86-27 identified that TVA did not have adequate calculations
for the sizing of the 125v station batteries, the battery charger and the 120v
vital ac inverter. TVA's response (Reference 5) stated that the sizing of
'these components had been reviewed as part of the program to upgrade the
electrical calculations and that the components bad been determined to have
adequate capacity for the existing loads.
Inspection Report 86-55 closed this
item based on a licensing review that was addressing the adequacy of TVA's
electrical calculations. The adequacy of the 125v de vital instrunent power
system voltage calculations was addressed in Section 2.3.3.2.2 of the staff's
safety evaluation of the Sequoyah Nuclear Performance Plan (Reference 44).
Based on the staff's safety evaluation report this item is closed.
(Closed) Unresolved Item US.3-4, Diesel Generator Loading Calculations
Inspection Report 86-27 identified several errors with the assumptions used by
TVA in the diesel generator loading analysis. TVA's response (Reference 5)
stated that the diesel generator analysis had been revised to correct the
,
l
concerns and that a new procedure for the preparation of the diesel generator
loading calculation would be prepared as part of the electrical calculation
'
program.
Inspection Report 86-55 closed this item based on a licensing review
that was addrest.ing the adequacy of TVA's electrical calculations. The
adequacy of the diesel generator loading calculations was addressed in Section
!
2.3.3.2.1 of the staff's safety evaluation of the Sequoyah Nuclear Performance
Plan (Reference 44).
Based on the staff's safety evaluation report this item
is closed,
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(Closed) Unresolved Item U5.3-5, Loss of Control Power Annunciation
'
Inspection Report 86-27 identified that there was a lack of control room
- annunciation for 'the loss of control power tc the auxil,iary feedwater pump.
The inspection report cited the requirements provided in Regulatory Guide 1.47,
Bypass and Inopt:rable Status Indication for Nuclear Power Plant Safety Systems,
TVA's response,(Reference 5) stated they were preparing a design concept for
n
the implementation of Regulatory Guide 1.47 requirements.
Inspection Report
86-55 closed thir, item based on a licensing review that was addressing TVA's
"
. implementation of Regulatory Guide 1.47. The implementation of Regulatory
'
Guide 1.47 was addressed in a separate safety evaluation report previously
transmitted to TVA (Reference 46). Based on the staff's safety evaluation
,
report this item,is closed.
(Closed) DeficiencyD6.3-1,SpecificationofHydrostaticTesttoDemonstrith,
,
.
Instrument Pressure Boundary Integrity After
Seismic Qualification Testing
Inspection Report 86-27 identified that TVA had not specified a design
performance test for hydrostatic pressure integrity following the seismic
l
qualification test for instruments purchased for recent plant modifications.
.
The inspection report referenced the requirements of TVA's procedures OEP-06
and OEP-09. TVA's stated position in its response (Reference 5) was that the
requirements for seismic testing and hydrostatic testing are totally separate
and independent of each other and that separate hydrostatic test following the
component seismic qualification was not required. However, TVA did comit to
'.
additional actions to test the onsite pressure switches to the rated overrange
of the units.
Inspection Report 86-55 accepted TVA's corrective actions for
the onsite pressure switches as demonstrating pressure integrity based on a
review of the test data but still identified a concern with the future
procurement specifications for other instroments. TVA submitted a revised
response to this issue (Reference 11). TVA's response stated that a review of
environmental cualification binders had revealed that the other instruments had
been pressure tested.
In addi* ion, TVA stated that standard procurement
'
specifications require hydrostatic tests to r:cet the requirements of ANSI
B31.1Property "ANSI code" (as page type) with input value "ANSI</br></br>B31.1" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process.. TVA still did not agree with the position that separate hydrostatic
tests were required after the seismic qualification.
Inspection Report 87-31
'r
closed this issue and referred it to licensing for review. The current staff
position on the seismic qualification of mechanical and electrical equipment is
contained in Section 3.10 of the Standard Review Plan (NUREG-0800).
This
position requires tests and analyses to confirm operability during and after a
seismic event including loads from normal and accident conditions. These
i
normal loads include the system operating pressure.
However, the staff
position does not specify a hydrostatic test be performed during the
qualification. Therefore, the staff accepts TVA's position and previous
,
corrective actions as adequate.
This deficiency is closed.
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INSPECTION REPORT NO. 86-38
(Closed) Observation 3.2, Margins of Safety for Restart
Inspection Report 86-38 identified that TVA was using interim criteria to
determine whether hardware modifications had to be made prior to the restart of
Sequoyah. These interim criteria involved piping systems, cable tray systems,
pipe supports and concrete anchorages.
Inspection Report 86-55 closed this
item based on a licensing review that was addressing TVA's restart design
criteria.
The review of these criteria are contained in Sections 2.3.2, 2.4
and 2.5 of the staff's safety evaluation of the Sequoyah Nuclear Performance
Plan (Reference 44). Based on the staff's safety evaluation report this
observation is closed.
i
(Clossd) Observation 6.3, Instrument Sense Line
Inspection Report 86-38 identified a concern that the walkdown inspections of
the auxiliary feedwater system turbine instrumentation and control was not
consistent for Sequoyah Units 1 and 2.
Inspection Report 86-55 indicated that
,
this observation was related to Observation 5.1 which involved the scope of the
.
walkdowns for the electrical and I&C areas. Observation 5.1 was closed in
Inspection Report 87-14.
Inspection Report 86-55 also identified a specific
concern with the scope of the instrument line walkdowns. TVA's response
(Reference 7) to Inspection Report 86-38 Observation 5.1 identified that 200
instruments had been inspected by the walkdowns. Based on the results of these
walkdowns Inspection Report 87-14 recomended a more complete walkdown of
safety-related HVAC instrument connections.
Inspection Report 87-31 documented
that TVA was performing additional walkdowns and that sketches were being made
of the installation of HVAC sensors that performed a protective or control
interlock function. The inspection report stated that TVA needed to confirm
the adequacy of the as-built installation shown on the sketches and provide a
schedule for issuing applicable design drawings.
Inspection Report 87-64 noted
that TVA was performing a technical adequacy review of the sketches and that
TVA had stated the sketches would be converted.into formal drawings when the
review was completed. The inspection report left the observation open pending
!
TVA's submittal of a schedule for completing the HVAC instrumentation drawings.
TVA's response (Reference 25) stated that they would issue the Unit 2 Phase 1
HVAC instrument line drawings by the Unit 2 Cycle 4 refueling outage. This
schedule is acceptable to the staff and, therefore, the observation is closed.
l
INSPECTION REPORT NO. 86-45
'
(Closed) Observation 8.1, Anchor Point Movement Loads
Inspection Report 86-45 identified the concern that anchor point movement loads
associated with a double ended guillotine break in the reactor coolant loop
were not included in Sequoyah's pipe support design criteria SQN-DC-V-24.1.
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The inspection report noted that these loads had been included in the Watts
Bar Design Criteria WB-DC-40.31.9 that had been previously used for the
design of Sequoyah supports.
The inspection report also noted that a NRC
letter (Reference 47) was sent to TVA requesting additional information on this
issue. The NRC letter requested TVA provide documentation of the basis for
concluding that the anchor point movements associated with the DBA are
si
sufficiently small to produce secondary, self-limiting type stresses.
Inspection Report 86-55 closed the observation based on the licensing review of
the issue. TVA's response to the NRC request for additional information
(Reference 48) stated that Sequoyah's FSAR had no direct reference to a
requirement to evaluate reactor coolant loop branch lines for the pipe break
anchor point motions. TVA's response also stated the basis for the conclusion
that these effects were small at Sequoyah were the results of the Watts Bar
analysis. The Watts Bar results cited by TVA in their response were movements
up to .5 inches for the broken loop and .25 inch for the unbroken loop.
Based'on subsequent discussions with the NRC staff TVA developed a new set of
design criteria, SON-DC-V-24.2, for the evaluation of pipe supports at
Sequoyah. This criteria was developed to obtain one set of design criteria
that was applicable to supports at Sequoyah and that met FSAR criteria. The
review of this criteria is contained in Section 2.3.2 of the staff's safety
evaluation of the Sequoyah Nuclear Performance Plan (Reference 44).
SQN-0C-V-24.2 does not require an evaluation of reactor coolant loop branch
connections for pipe break anchor motions. The staff agrees with TVA's
position that if the movements are sufficiently small they can be considered
The movement of
secondary and self-relieving for the pipe break evaluation.
.25 inch for the unbroken loop reported by TVA is considered sufficiently small
that this does not pose a safety concern.
In addition, the staff review of the
Sequoyah FSAR did not identify any commitment to analyze the reactor coolant
pipe branch lines for pipe break anchor point motions.
Based on the preceding
evaluation this observation is closed.
(Closed) Observation 8.2, Conformance to GDC for Containment Isolation
Inspection Report 86-45 identified the concern that Sequoyah was not in
compliance with the general design criteria for containment isolation.
. Inspection Report 86-55 closed the observation based on a licensing review that
was being performed. The licensing review of Sequoyah's containment isolation
design is addressed in Section 3.6.1 of the staff's safety evaluation of the
Sequoyah Nuclear Performance Plan (Reference 44). Based on the staff's safety
evaluation this observation is closed.
(Closed) Observation 8.3, Cable Tray Systems
Inspection Report 86-45 identified a number of concerns with cable tray support
systems.
Inspection Report 86-55 closed the observation based on a licensing
review that was being performed. The licensing review of cable tray supports
B-6
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is addressed in Section 2.5 of the staff's safety evaluation of the Sequoyah
Nuclear Performance Plan (Reference 44). The staff safety evaluation included
detailed reviews of TVA's cable tray support analyses.
Based on the staff's
safety evaluation this cbservation is closed.
(Closed) Observation 8.4, Piping and HVAC Systems
Inspection Report 86-45 identified a concern with unrestrained large
motor-operated valves on small lines and with the interaction of non-seismic
piping with safety systems.
Inspection Report 86-55 closed the observation
based on a licensing review that was being performed.
Both of these issues
were addressed by TVA's alternately analyzed piping program.
The review of
TVA's alternately analyzed piping program is addressed in Section 2.4 of the
staff's safety evaluation of the Sequoyah Nuclear Performance Plan
(Reference 44). Based on the staff's safety evaluation this observation is
closed.
INSPECTION REPORT 86-55
(Closed) Observation 3.8, Solencio Valve Mounting Seismic Qualification
Inspection Report 86-55 identified five items related to ECN L6487, Revision 1.
These items involved deficiencies in the documentation and seismic
qualification of a solenoid valve which was supported by control air tubing.
TVA's response (Reference 8) identified the corrective actions that would be
taken to address the deficiencies.
Inspection Report 87-14 found th3t TVA's
corrective actions were sufficient to document the adequacy of the solenoid
valve and closed the observation.
However, the inspection report noted a
possible discrepancy between TVA's Sequoyah Alternate Analysis Review Program
Description SQN-AA-001 and TVA's Sequoyah Nuclear Performance Plan on the
method of handling large concentrated weights for restart. The inspec': ion
report also noted that the alternate analysis program was the subject of a
separate staff review. This issue is similar to deficiency 03.2-2.
The
evaluation of TVA's alternately analyzed piping program is contained in
Section 2.4 of the staff's safety evaluation report on TVA's Sequoyah Nuclear
Perfomance Plan (Reference 44).
For alternately analyzed piping systems, the
torsional effects of large motor-operated and pneumatic-operated valves were
evaluated as prerestart items whereas TVA's evaluation of large concentrated
weights as a post-restart effort was considered acceptable.
Based on the
staff's safety evaluation report this observation is closed.
(Closed) Observation 6.15, Periodic Functional Test of Agastat Tirer Relays
in Pump Motor Start Circuits
Inspection Report 86-55 identified a concern that the .5 second time delay
relays, Agastat model 7012-PBL had not been periodically tested or calibrated.
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TVA's response (Reference 8) stated that the .5 secono reset timer would be
calibrated prior to Unit 2 restart and would be included in Sequoyah Standard
Practice SQE-8 for periodic calibration.
Inspection Report 87-31 contained a
review of TVA's corrective actions and identified concerns with the test
equipment, and method cf testing.
Inspection Report 87-64 contained a
discussion of TVA's method of calibrating the reset timers which involved
disconnecting the wiring leads. The inspection report stated that TVA should
verify the entire circuit on either an integrated or overlapping basis. TVA's
response (Reference 25) stated that the 22 relays identified by the observation
would be functionally tested in-circuit prior to Unit 2 Mode 2.
Based on this
comitment this observation is closed.
INSPECTION PEPORT N0. 87-14
(Closed) Observation 3.14, Evaluation of Masonry Block Walls
Inspection Report 87-14 identified that the DBVP project did not appear to be
evaluating unreinforced concrete masonry block walls in proximity to
safety-related piping and equipment in a consistent manner.
TVA's response
(Reference 15) referenced a TVA Final Report to the NRC for IE Bulletin 80-11
!
as evidence that adequate evaluations of block walls had been performed and
,
stated that the regeneration of these calculations would be performed as a
'
post-restart effort in accordance with the essential calculation verification
program. The resolution of issue of masonry block wall evaluations is
discussed in detail in Inspection Report 88-13, Deficiency 04.3-9.
Based on
!
the resolution of the issue in Inspection Report 88-13 this observation is
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closed.
(Closed) Observation 6.16, HVAC Flow Switch Calibration Data Records and
l
System 30 Surveillance Inst,ruction Procedures
Inspection Report 87-14 identified inconsistencies with the calibration records
for HVAC flow switches 2-FS-30-200 and 207.
In addition, the inspection report
identified a concern that no system level surveillance instruction existed to
test the various control logic interlocks developed by these sensors. TVA's
response (Reference 15) provided a discussion of TVA's surveillance
requirements for the switches and the tests performed.
inspection Report 87-31
found the additional information provided by TVA sufficient to close the issue
of the calibration data inconsistencies for the two identified switches but
still expressed a concern that a surveillance instruction procedure was needed
for the HVAC system.
Inspection Report 87-64 also identified that TVA had not
prepared or performed an appropriate surveillance instruction procedure. TYA's
response (Reference 25) comitted to test switches 2-FS-30-200 and 207 prior to
Mode 4 and to evaluate other control loops identified by the Restart Test
Program Function Matrix to determine which control loops should be added to the
periodic test program. TVA comitted to complete this evaluation by June 30,
1988.
Based on TVA's actions and commitments this observaticn is closed.
B-8
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(Closed) Observation 6.17, Diesel Generator Building Ventilation Fans Control
Logic and Surveillance Instruction Procedure
Inspection Report 87-14 identified drawing. errors in the logic diagrams and
control circuits that were not tested. The inspection report stated that
additional field inspection data was required to resolve the installation
configuration with the design. TVA's response (Reference 15) statad that CAQR
SQP 870171 was written to resolve the items prior to restart. TVA stated that
design drawing changes were to be accomplished by ECN-L6898 and under field
change request FCR 5351. TVA has also issued a CAQR SQT 871016 to resolve the
concern regarding periodic testing of these switches.
Inspection Report 87-31
identified a concern that TVA did not intend to prepare a surveillance
instruction (SI) to test the HVAC controls and interlocks and that the CAQR
corrective action had been changed to post-restart.
Inspection Report 87-64
acknowledged the TVA response to Inspection Report 87-14. However, the
inspection report restated the concern with the SI and testing of these
components. TVA's response (Reference 25) stated that the control loops would
be added to'the periodic test program post-restart as stated in the response to
Observation 6.16.
In addition, TVA stated that the functions of 2-FS-30-448,
450, 452 and 454 were evaluated under the restart test program. Based on this
program evaluation, TVA has determined that these switches were tested between
March and August, 1987. TVA's position was these functions did not require
testing again prior to restart. Based on TVA's corrective actions and
comitment this observation is closed.
INSPECTION REPORT N0. 87-31
(Closed) Observation 3.17, Solenoid Valve Mounting Support
Inspection Report 87-31 identified that ECN 5457 had resulted in the
replacement of solenoid valves in several piping systems that had variances
from standard typical drawing 47A054-33. The inspection report identified that
CEB was unable to retrieve seismic qualification calculations for these
variances. TVA's response (Reference 22) stated that no calculations could be
found for the variances.
Inspection Report 87-64 noted that TVA was generating
a calculation package to qualify the support variances. The inspection report
stated that this calculation would be completed prior to Unit 2 restart and
held the observation open pending the confirmation by TVA that the calculation
was completed. TVA's response (Reference 25) stated the calculation was
complete and issued under RIMS No. B25 871110 803.
Based on TVA's completto.n
of the seismic qualification documentation this observation is closed.
(Closed) Observation 4.8, Radiation Monitnring System
Inspection Report 87-31 identified an inconsistency between SCR SQNNEB8615 and
(QlR) NEB 86241.
The inspection report identified that punchlist item 4426
which had been written in response to SCR SQhNEB8615 had been reclassified from
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pre-restart to post-restart. Punchlist item 4426 identified corrective action
to provide safety-grade (seismically qualified) auxiliary control air to
System 90 radiation monitor supply valves in order to meet the requirements of
TVA's response (Reference 22) stated that the NRC's
original safety evaluation report (NUREG-0011) had identified that the airborne
particulate monito-ing system had not been specifically designed ta remain
functional when subjccted to an SSE and that Sequoyah's degree of compliance to
Regulatory Guide 1.45 constituted an acceptable basis for sotisfying the
requirements of General Design Criterion 30. Based on TVA's response,
Inspection Report 87-64 agreed with the system design and the post-restart
classification of the punchlist item. The inspection report left the
observation open pending(TVA's CCTS comitment to complete the corrective
action. TVA's response Refarence 25) documented that the QIR has been
revised.
Based on this response, this observation is closed.
(Closed) Observation 6.21, Change in Corrective Action for PAM Isolation
Inspection Report 87-3'. identified a concern that TVA was changing the
previously agreed upon corrective action for SCR SQNNEB8722. The original
corrective action called for separation of one post-accident monitoring channel
from non-safety-related wiring. TVA's response (Reference 22) stated that an
EEB disposition of SCR SONMEB8722 had determined that the separation was
consistent with the original design basis and acceptable. TVA also determined
that no qualified to non-qualified isolation problems are evident. TVA's
proposed corrective actions were to clarify the electrical separation
requirements in DIM-SQN-DC-V-19.9-1 and Section 7.5 of the FSAR.
In addition,
TVA comitted to upgrade the post-accident monitoring loops to meet Regulatory
Guide 1.97, Revision 2 requirerents in accordance with their previous
comitments by the end of the Cycle 4 outage for Unit 2.
Inspection Report
87-64 found TVA's interim and final implementation plans acceptable and held
the issue open pending TVA's submittal of the plan as a formal comitment.
TVA's response (Reference 25) documented that the FSAR would be revised in the
1989 update. Based on TVA's commitment to update the FSAR this observation is
closed.
(Closed) Observation 6.22, Auxiliary Control Air System Design Criteria
j
Inspection Report 87-31 identified a concern with the separation of the
auxiliary control air (ACA) headers from interactions with high and moderate
i
energy lines. TVA's response (Reference 22) described the previous TVA
evaluations for interactions and committed to perform an additional evaluation
to address the loss of ACA due to a small break LOCA.
Inspection Report 87-64
reviewed TVA's additional evaluation and requested that TVA verify the time
,
'
required for cperator action based on higher heat loads in the 480v shutcown
board rooms and that TVA review the operating procedures used by the control
i
room operator for the ventilation system process-auto control switches.
TVA's
'
response (Reference 25) stated these issues were addressed by a review of the
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control building HVAC design.
In this review, TVA has considered the A01-10 to
determine the operator actions and the higher heat load and has determined that
all equipment in the control building required for safe shutdowa or' Unit 2
remain functional to perform their safety function.
Based on TVA's revised
evaluation this observation is closed.
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TABLE B.1
DESIGN BASELINE ISSUES
. . _ .
Deficiency (D)
Inspection Report Number
Observation (Obs)
Unresolved Item (U)
86-27
86-38
86-45
86-55
87-14
87-31
87-64
02.1-1
0
C
02.3-1
0
C
D3.1-1
0
0
C
03.2-2
0
0
T
D3.2-3
0
0
C
D3.2-4
0
C
03.3-1
0
0
0
D3.3-2
0
C
03.3-3
0
C
03.3-4
0
T
D3.3-5
0
C
'
04.3-1
0
0
C
D4.3-3
0
0
0
0
04.3-4
0
C
!.
D4.3-5
0
C
04.3-6
0
C
U4.3-7
0
C
05.3-1
0
0
C
U5.3-2
0
T
U5.3-3
0
0
C
US.3-4
0
T
V5.3-5
0
T
D6.1-1
0
0
C
06.1-2
0
0
C
D6.1-3
0
0
C
!
B-12
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_
_ _ _ _ _ _
_
.
.
.
. -
. . .
. .
.
-
. - - . -
.
_
.
>
-
.
,
s
.
_
Deficiency (D)
Inspection Report Number
Observation (Obs)
UnresolvedItem(U)
86-27
86-38
86-45
86-55
87-14
87-31
87-64
~
06.2-1
0
C
06.3-1
0
0
T
V6.3-2
0
C
Obs 1.1
0
0
C
Obs 1.2
0
0
C
Obs 1.3
0
0
0
C
Obs 1.4
0
C
Obs 2.1
0
C
Obs 2.2
0
C
Obs 3.1
0
C
Obs 3.2
0
T
Obs 3.3
0
C
Obs 4.1
0
C
Obs 4.2
0
0
C
Obs 4.3
0
C
-
Obs 5.1
0
0
C
,
Obs 5.2
0
C
Obs 5.3
0
C
Obs 5.4
0
0
C
Obs 6.1
0
C
Obs 6.2
0
0
C
Obs 6.3
0
0
0
0
0
Obs 6.4
0
C
Obs 7.1
0
C
Obs 2.3
0
0
C
Obs 3.4
0
0
0
C
Obs 4.4
0
0
C
Obs 4.5
0
C
,
B-13
.
_ _ _ _ _ _ _ _ , - - - _ - - _ _
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.
_ .
.
_
_
.
_ . . .
__ -- --- - -
.
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,
(
.
.
M
Inspection Report Number
Deficiency (D)
'
Observation (Obs)
--
UnresolvedItem(U)
86-27
86-38
86-45
86-55
87-14
87-31
87-64
.
- Obs 4.6
0
C
.Obs 5.5
0
0
C
'
Obs 5.6
0
C
Obs 6.5
0
0
C
Obs 6.6
0
C
Obs 6.7
0
0
C
Obs 6.8
0
C
Obs 7.2
0
0
C
Obs 7.3
0
0
C
Obs 8.1
0
T
Obs 8.2
0
T
Obs 8.3
0
T
Obs 8.4
0
T
Obs 2.4
0
C
Obs 2.5
0
C
Obs 2.6
0
0
C
Obs 2.7
0
C
Obs 3.5
0
C
Obs 3.6
0
C
Obs 3.7
0
C
~
Obs 3.8
0
T
,
Obs 3.9
0
C
Obs 5.7
0
0
C
Obs 5.8
0
C
Obs 6.9
0
C
Obs 6.10
0
C
Obs 6.11
0
C
Obs 6.12
0
0
0
C
B-14
.
.
- . . .
.
. .
- .
-
.
- .
.
. . . .
..
-
.
'
.., '
,
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-
-Deficiency (D)
Inspection Report Number
Observation-(Obs)
Unresolved Item'(U)
86-27
86-38
86-45- 86-05
87-14
87-31
87-64
O
C
Obs 6.13.
0
0
0
C
Obs 6.14
0
0
0
0
-Obs 6.15
D
0
C
-Obs 7.4
0
C
Obs 2.8
0
C
Obs 3.10
0
C
Obs 3.11
0
C
Obs 3.12
0
0
C
Obs 3.13
0
Obs 3.14
0
C
Obs 3.15
0
C
Obs 4.7
0
C
Obs' 5.9
0
C
Obs 5.10
0
C
Obs 5.11
0
0
0
Obs 6.16
0
0
0
Obs 6.17
0
0
C
Obs 6.18
0
C
Obs 6.19
0
0
C
Obs 6.20
0
C
Obs 3.16
0
0
Obs 3.17
0
0
Obs 4.8
0
0
Obs 6.21
0
0
Obs 6.22
0
C
Obs 7.5
.
0 - Item is discussed as open issue in the Inspection Report
C - Item is closed in the Inspection Rt. port
T - Item is closed in the Inspection Report based on licensing review
B-15
_ _ _ - _ _ - _ - - _ _ _ _ _ _ _ _
"
.
.
.
APPENDIX C
LICENSEE. ACTION FOR PREVIOUS IDI FINDINGS
(Closed) Deficiency D3.3-3, Incorrect Pipe Support Allowable Stresses
Deficiency D3.3-3 identified a concern with the method used to evaluate the
faulted condition allowable stresses for pipe supports. TVA had used the TPIPE
piping analyses computer code which normalized loads to compare with allowable
stresses. This procedure could result in pipe support stresses exceeding the
allowable stresses for strucural steel specified in the Sequoyah FSAR. Review
of this deficiency was trans erred to the Office of Special Projects in
'
Inspection Report 87-74. Tb staff had previously identified a concern with
the criteria used by TVA to i taluate pipe supports.
This concern resulted in
TVA's issuance of pipe suppo t design criteria document SQN-DC-V-24.2 for the
evaluation of all rigorously analyzed pipe support calculations. During the
review of regenerated pipe support calculations, the staff confirmed that TVA
was implementing the criteria that limits pipe support allowable stresses to .9
yield. Based on the results of the review, this deficiency is closed.
(Closed) Deficiency 03.4-3, CCW Heat Exchanger Calculation
During a NRC field walkdown of the component cooling watee (CCW)heatexchanger
l
a discrepancy between the "as-built condition and vendor qualification
documents was identified. TVA's evaluation of the CCW heat exchanger found the
"as-built" condition acceptable, however, a detailed review of the
qualification of other Category I heat exchangers identified that the
I
containment spray heat exchanger required support modifications. The results
l
of this review are discussed further in Inspection Report 50-327,328/88-13.
This inspection report also identified the following three additional technical
I
concerns requiring resolution.
1.
TVA used damping va'ues of 2% for OBE loads and 3% for SSE loads to
evaluate equipment ioads. TVA has argued that these damping values are
consistent with ti~ criteria specified in IEEE 344-1975 which is used for
l
qualification of asectrical equipment.
IEEE 344-1975 specifies the same
l
damping values that are contained in Regulatory Guide 1.61, 2% for OBE and
3% for SSE. However, Table 3.7.1-3 of Sequoyah's FSAR specifies damping
values of 1% for OBE and up to 2% for SSE under the heading "Other Welded
)
l
Steel Structures." TVA's proposed resolution of this item is to revise
the FSAR to specify the higher damping values for equipment. Although
TVA's proposed danping values are consistent with current regulatory
criteria, the staff was concerned that the use of these values represented
a relaxation of the original liconsing basis for Sequoyah.
In order to
address the staff's concern with the conservation of the proposed damping
values. TVA provided an industry survey of measured damping values,
"Structural Damping Values as a Function of Dynamic Response Stress and
,
l
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..
,
.,
.
Deformation Limits" by J. D. Stevenson. This paper provides a survey of
in situ tests of nuclear power plant equipment. Table 7 of the paper
presents an average measured damping value of 7.7% for mechanical
components at a stress level corresponding to Sequoyah's SSE allowab?
l'mits.
Based on this test data, this issue is considered resolved for
restart. However, the staff still considers the resolution of this item
for conformance to TVA's FSAR commitments an open post-restart issue
(Unresolved Item URI 88-12-08).
2.
The issue of SRSS vs. absolute sum is discussed in Section 3.5 of this
inspection report. This issue is closed.
3.
For the analysis of piping attached to the heat exchangers, TVA decoupled
the heat exchanger analysis from the piping system analysis if the
calculated heat exchanger displacement at the piping nozzle attachment
point was less than 1/16 inch. This is consistent with the criteria for
allowable pipe support deflections contained in TVA's pipe support design
criteria, SQN-DC-V-24.2, and is acceptable. This issue is closed.
(Closed for restart) Deficiency 04.2-1, ERCW Pumping Station Access Cells
This deficiency was transferred to the Office of Special Projects in Inspection
Report 87-74. The ERCW pumping station access cells (access cells) consists of
six sheet pile cells and interconnecting cells which are filled with tremie
concrete. The ERCW piping and essential Class 1E conduits are also supported
by these cells.
The original seismic analysis of the access cells was based on the assumption
!
that the six cells and the interconnecting cells will act as a single
"J-shaped" unit. Contrary to this assumption, the design calculations predict
that shrinkage will occur in the interior concrete fill. This will cause a gap
,
I
l
between interior concrete and the exterior steel sheet piling. TVA design
criteria SQN-DC-V-104.5 states that "the sheet pile sections serve only as
l
o
forms for the tremie concrete; therefore, quality assurance is .ot required for
n
these sheet pile sections." The calculations also predict that tMre will be
l
vertical movement between adjacent cells.
Beams have been desir,ned to tie the
cells together in the horizontal direction but not vertically.
In fact
'
l
compressible material has been placed above and below these beams to preclude
l
load transfer in the vertical direction. TVA internal memorandum from J. H.
Coulson, Principal Civil Engineer to the Civil Engineering and Design Branch
files dated October 13, 1977, states, "cells A through F and the ERCW pumping
'
station are individual rigid bodies capable of moving vertically with respect
to each other."
The inability to transfer vertical shear between the cells makes the original
assumption of a single "J-shaped" unit invalid.
Furthermore, even if the
assumption was valid, torsional loads should have been considered in the
analysis and design since the "J-shaped" unit is not symmetrical. The
l
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.
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calculations also state the following:
"RJH & ROG analyzed the cells as both
individual cells and as a unit. The former case showed the cells were unstable
and the latter case showed a stable unit acting as a rigid body." The
calculations also show that cells are unable to transfer vertical shear
therefore making the original assumption of a single "J-shaped" unit invalid.
The calculation also state that the cells are not stable if they act as
individual cells. Also cores taken in November 1977 in several cells indicated
that the concrete at the bottom of these cells was soft, crumbly or contained
. gravel poc ets and cav t es.
TVA reanalyzed the ERCW access cells using a
k
ii
non-linear seismic time history response analysis. The revised seismic
analysis for the ERCW cell was based on a two-dimensional nonlinear tine
history analysis method in which the foundation was represented by discrete
springs and dampers with no tension capability in the vertical direction
(Reference 38). The staff's evaluation is discussed in the following:
,
1.
Soil-Structure Interaction Model - In the initial analysis model, the
upper bound modulus of tremie concrete was considered and the cell was
represented by a 5-mass stick model with a rigid base. Hydrodynamic
interaction between the submerged portion of the cell and the surrounding
water during horizontal vibrations was taken into account by including
tributary water masses in the lumped inass structural model. The struc-
tural damping was taken to be 5% for the concrete cell. The rock founda-
tion was represented by springs and dampers without tension capability for
,
the vertical springs and d:impers. The springs and dampers were derived
from the CLASSI computer code and then distributed to the cell-foundation
interface. Because the cell was partially submerged, the buoyant weight
of the cell was used as the effective dead weight in the nonlinear seismic
analysis. Since the initial analysis was not consistent with the as-bui!t
condition of the ce'), the staff requested that the cells need to be
reanalyzed. TVA revised both the structural model and the foundation
impedances based on the lower-bound concrete modulus. The 4' of very soft
concrete or gravel at the base of the cell was represented by discrete
springs and dampers that were combined with the rock foundation springs
and dampers.
The staff reviewed the nonlinear soil-structure interaction models used in
both the initial analysis (upper bound modulus of tremie concrete) and
subsequent analysis (lower bound modulus of tremie concrete), and found
,
I
them acceptable.
1
2.
Computer Code - The computer code UPLIFT was applied for the nonlinear
time history seismic analysis of the cell. The staff reviewed the
verification manua'. of the computer code and found it acceptable for the
uplift seismic analysis of the cell.
3.
Analysis - For the initial analysis, the four sets of SSE artificial
ground motion tim histories as described in the FSAR were used as input.
Each set of ground motion time histories contained one horizontal
component and one vertical component. To assess the significance of the
C-3
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.
" .
.
.
vertical ground motion to the uplifting of the cell base from the
foundation, four additional analyses were made with horizontal component
of ground motion as the only input. The results showeo that the maximum
base uplifting and the maximum horizontal displacement at the ERCW pipe
elevation were consistently about 76% and 0.18", respectively, for all
different ground n:otion input. TVA also evaluated the possible chipping
of the base concrete at the toe of the cell, and found that the maximum
possible chipping of the concrete would not exceed 1 foot. Additional
analyses considering the base dimension reduced by the 1 foot of concrete
chipping showed that it had a negligible effect on both the stability and
displacement response of the cell.
For the analyses using the lower-bound concrete modulus, the maximum base
uplift and maximum lateral displacement at the ERCW pipe elevation were
about 83% and 0.89", respectively. The maximum toe pressure was about 800
psi. The staff had concern on the magnitude of the seismic response toe
pressure with respect to the potentially low strength of the soft concrete
,
at the base of the cell.
TVA perfonned additional analysis which assume that the soil surrounding the
cell and sheet pile interlock confine the soft concrete and gravels. The
results of this analysis indicate a factor of safety of 1.05 against failure.
In a phone conversation, TVA comitted to [trform additional evaluations of
the ERCW access cells concrete foundation after the restart of the plant.
Special emphasis will be placed on determining the content and size of the
gravel pockets.
An evaluation program will be submitted to the staff for review and approval
(Unresolved Item URI 88-12-09). Once the as-built conditions have been
determined, TVA will reevaluate the stability and deflections of the access
cells. The intent is to confim that the ERCW piping will not be overstressed.
Based on tiie above, TVA has provided reasonable assurance that the ERCW access
cells can withstand the postulated SSE event. This deficiency is closed for
restart.
(ClosedRestart)
Deficiency 04.2-3, Vertical Response Spectra of the Steel
containment Vessel
During the IDI review an issue with the adequacy of the vertical amplified
response spectrum for the steel containment was identified. The issue involved
the adequacy of the time step used to generate the original vertical response
spectra which were used to evaluate piping and equipment.
As part of the
resolution of this issue, TVA identified two additional structures where the
newly generated vertical response spectra exceeded the original response
spectra. These structures are the reactor building interior concrete structure
and the auxiliary control building.
Inspection Report 88-13 discusses the
C-4
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,
.
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.
evaluation of the newly generated response spectra and the effect of these
spectra on piping and equipment attached to the steel containment. TVA
evaluated the effects of the newly generated spectra on a sample of piping and
equipment attached to the interior concrete structure and the auxiliary
The results of these evaluations are contained in a TVA letter to
building (. Reference 53). TVA's sample of equipment anchorages included seven of
the NRC
the 60 items that were previously evaluated for 101 Deficiency 4.6-1.
These
seven were located at the highest elevations of the auxiliary building to
evaluate the effects largest spectra changes. The results of these evaluations
showed the new spectra had no effect on the anchorage qualification.
In
e
addition, TVA evaluated two heat exchangers that had the highest anchorage
stresses; the containment spray and the component cooling water heat
exchangers.
For both cases the new spectra had lower accelerations than the
input that was used in the qualification of these components. To evaluate the
effects of spectra changes on piping systems, four piping systems were selected
for evaluation. The results of these evaluations showed small increase in
loads and stresses for the piping systems which did not affect the current
qualification of the piping, equipment and supports. Based on the results of
these sample evaluations, TVA's evaluation of the spectra changes for the
interior concrete structure and the auxiliary building is considerd acceptable
for restart.
TVA should provide additional evaluations of the remaining piping
and equipment as a post-restart effort (Unresolved Item URI 88-12-10).
During the inspection the staff requested that TVA address the staff concern
over the impact of the time step on response spectra that had been developed
for piping attached to the reactor coolant loop. The response spectra for the
reactor coolant loop had been generated by EDS Nuclear Inc. in January 1974.
TVA determined that EDS had used a .01 second time step with an EDS in-house
computer program to develop these response spectra. To assess the potential
impact of the time step on the reactor coolant loop spectra, TVA selected a
sample of four lines attached to the coolant loop. The results of the
assessment of these four lines are contained in a TVA letter to the NRC dated
March 2, 1988 (Reference 56). This assessment addressed the nozzle attachment
points and the first seismic restraint adjacent to eacn nozzle.
The results
of this evaluation showed that the nozzle stresses and supports met the
allowable limits.
Based on the results of these sample evaluations, TVA's
evaluation of the potential effects of the time step issue on the reactor
coolant loop attached piping is considered acceptable for restart. TVA
should provide additional evaluation of the remaining attached piping systems
as a post-restart effort (Unresolved Item URI 88-12-10).
(Closed) Deficiency 04.3-7, Vertical Seismic Load on Auxiliary Building Roof
Truss
Inspection Report 88-13 discussed the resolution of the concern with the
seismic vertical amplification of the auxiliary building roof due to its
flexibility. As part of the resolution of this issue, a concern with TVA's use
uf 2/3 of the horizontal spectra to represent the vertical spectra was
C-5
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,
..
identified. Three Category I structures were identified and reviewed by TVA
for this issue.
For one building, the diesel generator building, TVA generated
a set of new floor response spectra using the 84th percentile site specific
response spectra as input ground motion (Reference 54). The new vertical floor
response spectra exceeded the original vertical floor response spectra.
evaluated the effects of this new spectra on piping systems located in the
diesel generator building. This evaluation is contained in TVA calculation,
"Evaluation of CAQRS879242, N2-870242-Misc, Revision 1, dated February 25,
1988 (RIMS B25 880226 800).
The staff review of stress problem N2-82-3A
identified that TVA used interim criteria from CEB-CI-21.89 to qualify hanger
17A586-01-001. The criteria used by TVA for this evaluation involved a
modified fatigue evaluation for the secondary load case. This criteria is
contained in Section 3.2.8 of CEB-CI-21.89. This criteria was not accepted
by the staff for general use unless e case-by-case review and approval was
obtained (Reference 51). The staff requested that TVA identify all other
cases where similar criteria had been used without staff review and approval.
TVA stated that this evaluation had only been used on four typical cases on the
diesel generator exhaust lines. TVA used the criteria to evaluate a problem
identified with insufficient pipe clearance at a thermal travel stop.
The high
stress occurred ut the welded attachment and according to TVA the loading was
one directional and would not result in a full stress reversal during heatup
and cooldown. Therefore, TVA stated the application of the modified fatigue
evaluation would be conservative. The staff agreed with TVA's statement that
a thermal analysis that assumed a full stress reversal would be conservative
for this case.
However, the staff requested that TVA visually inspect these
four diesel generator exhaust support attachents prior to restart for signs
of distress.
The staff informed TVA that if the results of the inspection
did not show damage the post-restart modification of these supports would be
acceptable. TVA should document the results of this inspection (Unrevolved
item URI 88-12-11).
The staff's review of the evaluations of the remaining piping systems in the
diesel generator building for the new vertical response spectra did not
identify any additional probicms. Therefore, the piping evaluation performed
by TVA for the vertical spectra issue was considered acceptable.
This
unresolved item is closed.
(Closed) Unresolved Item U3.5-1, Piping Code of Record
Unresolved Item U3.5-1 identified that TVA used stress allowable limits for
piping specified in the ASME Code instead of the stress allowable limits in the
ANSI B31.1 - 1967 Code that was specified in Table 3.9.2-3 of Sequoyah's FSAR.
The B31.1 - 1967 stress limits are generally more conservative than the ASME
Code stress limits for stainless steel materials.
Review of this unresolved
item was transferred to the Office of Special Projects in Inspection Report
87-/4. TVA responded to this issue in a letter to the NRC (Reference 49).
TVA's response provided an evaluation of the differences between the B31.1 -
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dCodeandASMESectionIII1971throughtheWinter1972 Addenda. TVA's
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$1uationconcludedthatthebasicdefinitionofcriteriainbothcodeswere
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Jentical and the material requirements were similar. TVA's discussion
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ttributes the difference between the codes to a change in the definition of
WJU. i. f:/
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pyieldstressthatwaspickedupbythelater(1973)editionofANSIB31.1.
RW.C:-
[
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E TVA proposes to revise the FSAR to incorporate allowable stresses from
Since the FSAR criteria used to
-ifdpQav.
/
ASME III, 1971 through Winter 1972 Addenda.
define the stress allowable equations are based on the ASME Code criteria,
L)A'd
'
'
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-1
.
the use of the allowable stress limits defined by the ASME Code provides a
consistent design basis for piping stresses. Threfore, TVA's proposed FSAR
iM.
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."
revision is acceptable to the staff. This also resolves the open item URI
4.3.2 in Inspection Report 87-44 relating to the correct code allowable
d$.;j(' 'd
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stresses used in the evalt,ation of small diameter piping for the material
control issue. URI 4.3.2 had identified the same issue with code allowable
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stresses.
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1967 Code and ASME Section III 1971 through the Winter 1972 Addenda. TVA's
evaluation concluded that the basic definition of criteria in both codes were
identical and the material requirements were similar. TVA's discussion
-
attributes the difference between the codes to a change in the definition of
'
yield stress that was picked up by the later (1973) edition of ANSI B31.1.
TVA proposes to revise the FSAR to incorporate allowable stresses from
ASME III,1971 through Winter 1972 Addenda. Since the FSAR criteria used to
define the stress allowable equations are based on the ASME Code criteria,
the use of the allowable stress limits defined by the ASME Code provides a
consistent design basis for piping stresses. Threfore, TVA's proposed FSAR
revision is acceptable to the staff. This also resolves the open item URI
4.3.2 in Inspection Report 87-44 relating to the correct code allowable
stresses used in the evaluation of small diameter piping for the material
!
control issue. URI 4.3.2 had identified the same issue with code allowable
I
stresses,
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I
APPENDIX 0
POST-RESTART UNRESOLVED ITEMS
Unresolved Item URI 88-12-01, Thermal Monitoring of Supports
TVA's use of strain page measurements to qualify supports 2-H63-2, 2-H63-3,
2-H63-4 and 2H63-5 to the long-term pipe support design criteria SQN-DC-V-24.2
is not acceptable unless TVA performs an evaluation of the entire piping
analysis problem to obtain the correct distribution of loads for the
qualification of all supports in the analysis. TVA should provide a response
which specifies the method used to qualify these four supports'to long term
criteria.
Unresolved Item URI 88-12-02, Allowable loads for Standard Component Supports
TVA has specified allowable loads for standard component supports in
SQN-0C-y-24.2, Figure I-2 that allows the use of load rating provisions of the
ASME Code to establish allowable limits. The staff has accepted the use of
these allowable limits for restart (Reference 44) but the staff still has an
open issue with TVA's demonstration that these allowable loads meet the
Sequoyah FSAR allowable limits.
Unresolved Item URI 88-12-03, DBA ZPA Effects
The staff review of Employee Concerns Element Report 221.2(B) identified that
TVA had not followed the recommendations in civil engineering report CEB 80-58
l
for evaluating ZPA effects on the piping for the containment DBA analysis. TVA
l
addressed this concern for restart by evaluating a sample of five piping
systems attached to the containment to demonstrate that restart design criteria
were not exceeded. TVA should complete the evaluation of the remaining systems
attached to the containment to demonstrate these systems meet the FSAR
allowable limits.
Unresolved Item URI 88-12-04, Containment DBA Spectra
'
TVA requested the use of ASME Code Case N-411 damping for the analysis of
piping. The staff accepted the use of this code case for the containment DBA
analysis of piping p(rovided the response spectra had been derived on a
conservative basis Reference 34). The staff's review of the containment DBA
spectra raised concerns with the uncertainties involved in the analysis used to
generate the spectra. TVA's analysis method was considered acceptable for
restart. TVA was requested to confirm the adequacy of the double differentia-
tion technique and the adequacy of cutting off the analysis at .9 seconds as
l
post-restart items.
0-1
.-
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.
Unresolved Item URI 88-12-05, ERCW Pumphouse
During the review of the ERCW pumping station geophysical logs it was
discovered that one of the cores indicated that approximately 7 feet of void
exited between the two northern most intake liners.
TVA provided an evaluation
which was considered acceptable for restart. However TVA was requested to
provide additional evaluations as a post-restart effort.
,
Unresolved item URI 88-12-06, Feedwater Waterhammer
During the design calculation review it was identified that TVA had performed
an analysis of a waterhamer due to the feedwater check valve closure event but
had not formally issued the report. TVA subsequently performed an additional
analysis that was considered acceptable to the staff for restart. The staff
has not resolved the issue as to the appropriate long-term criteria for this
analysis.
Unresolved Item URI 88-12-07, HVAC Duct Support Calculations
CAQR SQT870843 was written by TVA to address concerns that had been identified
during the review of regenerated HVAC support calculations. TVA's resolution
of the CAQR was to evaluate the five worst case duct systems by computer
analysis.
The staff found the results of TVA's sample calculations acceptable
for restart. TVA used interim criteria in the qualification of these duct
supports. The staff considered the sample adequate for restart, however, the
staff requests that TVA evaluate an additional sample of HVAC duct supports as
a post-restart effort.
Unresolved Item URI 88-12-08, Component Damping Values
During the IDI review of the component cooling water heat exchanger calculation
it was identified that TVA was using damping values for component qualification
from Regulatory Guide 1.61 instead of the damping values specified in FSAR
Table 3.7.1-3 for weldeo structures. The staff considered TVA's use of current
l
li m nsing criteria acceptable for restart.
However, the staff still considers
the issue of appropriate damping values for mechanical compononts an open
post-restart issue.
!
Unresolved Item URI 88-12-09, ERCW Pumping Station Access Cells
During the IDI review a concern was identified with TVA's assumptions used in
the evaluation of the ERCW access crils. TVA performed additional evaluations
to demonstrate the adequacy of the access cells for the SSE event. The staff
considered the results of these evaluations acceptable for restart. The staff
requested that TVA submit an evaluation program to evaluate the stability and
D-2
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deflections .of the access cells using as-built conditions as a post-restart
item.
Unresolved Item URI 88-12-10, Seismic Analysis of the Steel Containment Vessel
TVA's review of steel containment vessel vertical response spectra for the time
step issue evaluated the effects on a sample of piping systems attached to the
affected structures including the reactor coolant loop. These samples were
considered adequate for restart. The staff considers that TVA should complete
the evaluation of the remaining piping systems as a post-restart item.
Unresolved Item URI 88-12-11, Diesel Generator Exhaust Piping
The staff review of TVA's evaluation of piping in the diesel generator building
for the effects of the time step on the seismic response spectra identified
that TVA had used interim criteria that had not been reviewed and approved by
the staff.
7 '. . staff considered the use of this criteria acceptable providing
TVA visually inspect the affected support attachments on the diesel generator
exhaust lines to determine if any damage had occurred. TVA should provide the
results of this inspection to the NRC.
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APPEllDIX E
E.1 ENTRANCE MEETING - FEBRUARY 15, 1988
NAME
ORGANIZATION
TITLE
J. R. Fair
NRC/OSP
Team Leader
T. M. Cheng
NRC/OSP
Team Member
A. I. Unsal
NRC/Censultant
Civil / Structural
Bill Neely
TVA/CEB
Senice Civil Engineer
Jim Rochelle
TVA/CEB
Senior Mechanical Engineer
George Sanders
G/C
Project Engineer
Salah Azzazy
TVA/CEB
Senior Project Engineer
Wayne Massie
TVA/SQN Licensing
Licensing Engineer
Peter Gulko
Bechtel
Technical Specialist
Hubert Nugent
Bechtel
Engineering Supervisor
A. V. duBouchet
NRC/ Consultant
Mechanical Components
Owen Mallon
NRC/ Consultant
Civil / Structural
Tony Capozzi
TVA/DNE
EA Manager
Carlo Brillante
TVA/CEB
Senior Mechanical Engineer
David Bogaty
TVA/EA
Civil /EA
L. Raghavan
Terry C. Price
TVA/SQEP
Design Basis Program Manager
Frank E. Denny
TVA/DNE/EA
Senior Engineering Specialist
Roy E. Hoekstra
TVA/SQEP
Principle Civil Engineer
J. C. Key
TVA/SQEP
Assistant Project Engineer
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L. A. Budlong
Assistant Project Engineer
Jack B. Thomison
TVA/SQEP
Principal Civil Engineer
Alan Perkins
TVA/SQEP
Technical Supervisor
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John Lockaby
Lead Engineer
Orhan Gurbul
Bechtel
Engineering Specialist
C. N. Johnson
TVA/CEB
Lead Civil Engineer
J. A. Graziano
TVA/CEB
Senior Civil Engineer
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Don L. Williams
TVA/DNLRA/ELB
Manager Engineering Licensing
Tom N. C. Tsai
NRC/ Consultant
Civil / Structural
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Kenneth L. Mogg
TVA/EMG/CEB
Lead Engineer EMG
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Karl S. Seidle
TVA/DNE/CEB
Asst. Chief, Civil
Engineering
Roy T. Holliday
TVA/DNLRA/ELB
Nuclear Engineer
Fred L. Moreadith
TVA/DNE
Engineering Manager
John K. McCall
TVA/CEB
Chief, Civil Engineering
Robert E. Serb
NRC/ Consultant
Mechanical Components
G. Harstead
NRC/ Consultant
Civil / Structural
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E.2 EXIT MEETING - FEBRUARY 19, 1988
NAME
ORGANIZATION
TITLE
i
Roy T. Holliday
TVA/DNLRA/ELB
Nuclear Engineer
Daivd H. Level
TVA/DNLRA/ELB
Nuclear Engineer
Bill Roberts
TVA/0NE/CEB
Civil Engineer
Carlo Brillante
TVA/DNE/CEB
Senior Mechanical Engineer
Ken Mogg
TVA/DNE/ CEB
LeadEngineer(EMG)
Wayne A. Massie
TVA/SQN Site Licensing
Licensing
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M. J. Ray
TVA/DNLRA
Deputy Director
Don L. Williams
TVA/DNLRA
Manager, Eng. Licensing
James M. Warren
TVA/0NE
Mechanical Engineer
F. L. Moreadith
TVA/0NE
Engineering Manager
L. G. Hersh
BWPC
Chief C/S Engineering
W. S. Tseng
BWPC/SF
Civil / Structural
J. A. Kirkebo
TVA/DNE
ONE Ofrector
D. W. Wilson
TVA/0NE
Chief, NTB
S. E. Gibson
TVA/DNE
Assist to Chief, MEB
L. A. Budlong
Asst Project Engineer
J. R. Fair
NRC/OSP
Team Leader
Robert A. Hermann
NRC/OSP
Chief. Engineering Branch
Rcbert E. Serb
NRC/ Consultant
Mechanical Components
A. V. duBouchet
NRC/ Consultant
Mechanical Components
A. I. Unsal
NRC/ Consultant
Civil / Structural
Gunnar Harstead
NRC/ Consultant
Civil / Structural
Thomas M. Cheng
NRC/OSP
Team Vember
Tom N. C. Tsai
NRC/ Consultant
Civil / Structural
Owen Mallon
NRC/Contul tant
Civil / Structural
Karl S. Seidle
Asst. Ch., Civil Engineering
Ruben 0. Hernandez
Asst. Ch., Civil Engineering
Bill Neely
Senior Civil Engineer
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Joseph A. Graziano
Senior, Civil Engineer
JoEnMcCall
Chief, Civil Engineering
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C. N. Johnson
Lead Civil Engineer
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S.'E. Azzazy
Senior Mechanical Engineer
Peter Gulko
Bechtel
Technical Specialist
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Roy Hoekstra
Principle Civfl Engineer
t.arry A. Katcham
' Principle Engineer
J. C. Key
Assistant Project Engineer
T. E. Bostrom
Bechtel
Project Engineer Manager
H. S. Nugent
Bechtel
Engineering Supervisor
K. S. Jadeja
Dep. Ch., Civil Engineer
Frank E. Denny
TVA-EA
Sr. Engineering Specialist
David Bogaty
TVA-EA
Civil Engineer
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A. P. Capozzi
Manager of EA
F. P. Carr
TVA/MEB
Engineer!ag Specialist
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E.3 REFERENCES
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1.
Inspection Report 50-327/86-27 and 50-328/86-27, forwarded by J. Taylor
letter dated April 22, 1986.
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2.
Inspection Report 50-327/86-38 and 50-328/86-38, forwarded by J. Taylor
letter dated September 15, 1986.
3.
Inspection Report 50-327/86-45 and 50-328/86-45, forwarded by J. Taylor
letter dated October 31, 1986.
4.
TVA response to Inspection Report 50-327/87-06 and 50-328/87-06 (Domer to
NRC) dated July 2, 1987.
5.
TVA Response to Inspection Report 86-27 (Gridley to Grace), dated July 28,
1986.
6.
TVA rt; vised rssponse to Inspection Report 86-27 (Domer to Grace), dated
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December 31, 1986.
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7.
TVA response to Inspection Reports 86 38 and 86-45 (Domer to Taylor).
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dated February 3, 1987.
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8.
TVA response to Inspection Report 86-55 and other 1hspection Items
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remaining open (Gridley to Ebneter), dated April 22, 1987.
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9.
Inspection Report 50-327, 328/86-55, forwarded by J. Taylor letter dated
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February 3, 1987.
10.
Inspection Report 50-327,328/87-06, forwarded by S. Ebneter letter dated
April 8,1987,
11. TVA Additional Information in Response to Inspection Report 86-27, (Domer
toToylor),datedJanuary 30, 1987.
12. Engineering Assurance Oversight Review Report, "Sequoyah Nuclear Plant
Unit 2 Design Baseline and Verification Program, " EA-0R-001, issued
April 29, 1987.
!
13. Sequoyah Nuclear Plant - Design Baseline and Verification Program Unit 2
Phase 1 Report, dated May 29, 1987.
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14.
Inspection Report 50-327, 328/87-14, forwarded by S. Ebneter letter dated
June 4, 1987.
15. TVA response to Inspection Report 50-327, 328/87-14 (Gridley to NRC),
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dated July 16, 1987,
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16. TVA revised response (Observation 5.7) to Inspection Report 50-327,
328/87-14 (Gridley to NRC), dated September 1,1987.
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17. TVA letter relating to control and processing nf changes to the punch list
(Gridley to NRC), dated August 20, 1987.
18.
Inspection Report 50-327, 328/87-27, forwarded by S. Ebneter letter dated
August 24, 1987,
,
19. TVA letter addressing SQN-DNE Design Calculation Efforts (Gridley to NRC),
dated July 31, 1987.
20. TVA response to Inspection Report 87-27 (Gridley to NRC), dated
October 21, 1987.
21. TVA letter addressing revised commitment date for interface guidelines
(Gridley to NRC), dated November 20, 1987.
22. TVA letter is response to findings identified during the final NRC
inspection of the DBVP (Gridley to NRC), dated October 27, 1987,
23.
Inspection Report 50-327, 328/87-64, forwarded by S. Richardson letter
dated February 23, 1988.
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24. TVAresponse(CalculationProgramIssues)toInspectionReport50-327,
328/87-64 (Gridley to NRC), dated January 19, 1988,
25. TVA response (DBVP Issues) to Inspection Report 50-327, 328/87-64 (Gridley
to NRC), dated January 20, 1988.
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26.
Inspection Report 50-327, 328/87-48, forwarded by S. Ebneter letter dated
'
November 6, 1987.
>
27. TVA response to Inspection Report 50-327, 328/87-48 (White to NRC), dated
December 29, 1987.
28.
Inspection Report 50-327, 328/87-74, forwarded by S. Richardson letter
dated February 22, 1988.
29. TVA response to Inspection Report 50-327, 328/87-74 (Ray to NRC), dated
April 21, 1988.
30.
Inspection Report 50-327, 328/88-13, forwarded by S. Ebneter letter dated
May 26, 1988.
31. TVA letter providing additional information on the Design Calculation
Review (Gridley to Keppler), dated July 31, 1987.
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32. NRC Trip Report addressing the review of employee concerns (Liaw to
Zwolinski),datedNovember 10, 1987.
33. TVA letter providing additional response (Observation MEB-3) to inspection
Report 50-327,328/87-27 (Gridley to NRC), dated February 18, 1988.
34. NRC Safety Fvaluation on ASME Code Case N-411 damping forwarded by~ letter
dated Februery 8, 1988.
35. TVA letter providing additional information on conduit and HVAC duct
support calculations (Gridley to NRC), dated March 2, 1988.
36. TVA letter providing additional information on the Essential Raw Cooling
Water (ERCW) Pumping Station concrete (Gridley to NRC), dated March 3,
1988.
37. TVA letter providing additional information on the Essential Raw Cooling
Water (ERCW) Pcmping Station concrete (Gridley to NRC), dated March 2,
1988.
38. TVA letter providing a supplemental response to IDI item 04.2-1 (Ray to
NRC),datedMarch2,1988.
39. TVA letter providing sdditional inforamtion on piping support design
criteria (Gridley to NRC), dated March 2, 1988.
40. TVA letter, "Effect of Zero Period Acceleration (ZPA) on piping during the
design basis accident (DBA)" (Gridley to NRC), dated March 2, 1988.
TVA letter providing adcitional information on swing (angle allowables,
41.
design basis accident spectra and U-bolt allowables Gridley to NRC),
dated March 2, 1988.
42. TVA letter providing additional information on the effect of square root
of the sum of the squaro (SRSS) versus absolute sum (ABS) (Gridley to
NRC), dated March 2, 1988.
43. TVA letter "Sequoyah Nuclear Plant (SQN) - NRC Comitments" (Gridley to
NRC), dated February 29, 1988.
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44. NUREG-1232, Volume 2 "Safety Evaluation Report on Tennessec Valley
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Authority:
Sequoyah Nuclear Performance Plan," dated May 1988.
45. TVA letter addressing the regeneration of pipe support calculations
(Gridley to NRC), deted August 21, 1987.
A.
NRC Safety Evaluation on Regulatory Guide 1.47 forwarded by letter dated
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1987.
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47.
NRC letter requesting additional information on anchor point movement
loads (YoungbloodtoWhite),datedSeptember 29, 1986.
movements (Gridley to NRC) quest for additional information on ancho
TVA response to the NRC re
48.
dated October 31, 1986,
,
49. TVA letter providing a revised response for 101 item U3.5-1 (Ray to NRC),
dated March 2, 1988.
50. TVA letter on the regeneration of Sequoyah Unit 2 pipe support
calculations (GridleytoNRC),datedAugust 21, 1987.
51. NRC Meeting Sumary on pipe support criteria dated September 4,1987.
52. TVA letter providing supplemental information on the Sequoyah Unit 2 pipe
support restart criteria (Gridley to NRC), dated November 17, 1987.
53. TVA letter providing additional information on 101 item 04.2-3 (Ray to
NRC), dated March 2, 1988.
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54. TVA letter providing additional information on IDI Item 04.3-7 (Ray to
NRC), dated March 2, 1988.
55. NRC letter on damping values for analysis of conduits (Zech to White),
dated February 18, 1988.
56. TVA letter providing additional information on the effect of the tine
step concern on the RCL spectra (Gridley to NRC), dated March 2, 1988,
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