ML20236C047
| ML20236C047 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 10/21/1987 |
| From: | Girdley R TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| NUDOCS 8710260471 | |
| Download: ML20236C047 (17) | |
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TENNESSEE VALLEY AUTHORITY-CH ATTANOOGA. TENNESSEE 374ot -
@g gL 5N.157B Lookout Place" I
00T 211987
. U.S. Nuclear Regulatory' Commission' ATTN: Document Control Desk
' Washington,.D.C.
20555
- Geritlemen t -
4 In the Matter of,
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Docket?Nos.' 5'0-3271
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- Tennessee Valley Authority
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.50-328-l SEQUOYAH NUCLEAR PLANT-(SQN) -: RESPONSE TO NRC REPORT NOS.'50-327/87-27.AND L
50-328/87-27 On June 5,1987, an NRC' inspection team from the' Office of Nuclear l Reactor?
l Regulation (NRR)'in.Bethesda,.MarylandP concluded a one-week inspection of the.
TVA Division of Nuclear Engineering-(DNE)' design l calculation. efforts.1'The-team. inspected the DNE,offorts to~ identify,eretrieve,~and review for'technica11 adequacy the essential calculations;within'each engineering branch,:as; i i described in a January 20, 1987 letter to NRC and as revi'ewediduring:an'NRC:
inspection during February 1987. The team reviewed the' individual branchi i
programs as well as specific calculations for technical adequacy. The resialts l
are identified as 13 observations'and arel documented in NRC Report Nos.
l 50-327/87-27 and 50-328/87-27 dated August _24, 1987'
[ consists of a restatement of the NRC observations, with the respective TVA response immediately following:the observation. Five observation categories were established., relating to.the DNE programmatic effort and the four DNE engineering branches.
.In' addition, following the.TVA response to observation:EEB-111is a discussion addressing NRC's general' comment concerning a lack of coordination among engineering disciplines regarding responsibility for determination of instrument setpoints.
This discussion describes TVA's_ corrective action-plan in response.to1the team's concerns.
DNE has performed a review to assess'the, adequacy of l
existing system performance requirements and supporting calculations'and/or L
tests. This review identified additional calculations.that are required, which are currently being generated inLaccordance:with restart criteria.
- Also, DNE will issue a division procedure to delineate the. Interdisciplinary; interfaces regardir.g the specification of ' system performance. criteria.
Additional' details of this effort are presented in' enclosure 1.
L l
h Q
An Equal Opportunity Employer -
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' :i U.S. Nuclear Regulatory Commission 00ff 21.1987 1
l l ' consists of new commitments for SQN identified in enclosure 1.
If you have any questions concerning this issue, please telephone Both 'L.
Hall-
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at-(615) 870-7459.
l To the best of my knowledge, I declare the statements contained herein are i
complete and true.
1 Very truly yours, TENNESSEE VALL 'Y ' AUTHORITY l
R.
"idley, Di ector Nuclear Licen ing and Regulatory Affairs Enclosures cc (Enclosures):
Mr. G. G. Zech, Assis', ant Director for. Inspection Prcgrams office of Special Frojects l
U.S. Nuclear Regulatory Commission
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101 Marietta Street.NW, Suite 2900 1
Atlanta, Georgia 30323 J
Mr. J. A. Zwolinski, Assistant Director for Projects Division of TVA Projects office of Special Projects L
U.S. Nuclear Regulatory Commission l
4350 East-West Highway EW 322 Bethesda, Maryland 20814 Sequoyah Resident Inspector Sequoyah Nuclear Plant 2600 Igou Ferry Road Soddy Daisy,. Tennessee 37379 1
I L
I ENCLOSURE 1 I
Following are TVA responses to observations documented in NRC Report Nos.
50-327/87-27 and 50-328/87-27.
The observations have been repeated as written in the inspection report, and the respective TVA response immediately follows each NRC observation.
Some of the responses reference corrective action j
identified as part of documentation of conditions adverse to quality (CAQs).
j Condition Adverse to Quality Reports (CAQRs) or other CAQ documentation will
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be dispositioned in accordance with TVA's CAQ process.
GENERAL OBSERVATIONS GEN Unverified Assumptions "TVA's policy on unverified assumptions is set forth in Nuclear Engineering' Procedure NEP-3.1, which requires all unverified assumptions in calculations to be identified, tracked, and eventually verified. Pursuant to NEP-3.1, the i
cover sheet of each calculation has a chec'. toff block to indicate whether the calculation contains unverified assumptions, and responsibility for following up on verification is assigned to the lead engineer on each project.
In addition to the written DNE policy, TVA management has agreed that all of the unverified assumptions contained in " essential restart" calculations for SQNP must be verified before the plant can be restarted.
The team is concerned that - notwithstanding the policy and TVA's management committ ent to verification - none of the technical branches has any procedure - ~
in place for ensuring that the unverified assumptions will be tracked and verified, and the corrected results will,be applied to calculations that rely on calculations containing unverified assumptions as a source of input.
In discussions with the team, DNE acknowledged this problem, but no solution was immediately available. The absence of a control mechanism for assuring the verification of unverified assumptions in design calculations is a significant u
issue that should be addressed by TVA in a timely manner."
TVA RESPONSE The DNE program for tracking unverified assumptions in engineering calculations was the lead engineers' official calculation logs as required by Nuclear Engineering Procedure (NEP)-3.1.
For Sequoyah Wuclear Plant (SQN) essential calculations, the DNE log is implemented by the Calculation Cross Reference Information System (CCRIS).
The verification of unverified j
assumptions for SQN essential calculations was initiated on June 5, 1987, by,
memorandum from H. E. Pennell to D. W. Wilson, SQU Project Engineer. This memorandum requested all disciplines to review the essential calculations and identify and resolve all unverified assumptions, revise the calculation to document this effort, and modify the CCRIS listing to reflect current status.
The disciplines modified their reapective calculation review programs as described in a TVA letter to NRC dated July 31, 1987.
These revisions included the commitment to review essential calculations for SQN and verify or justify all unverified assumptions contained in the SQN essential calculations.
This effort was to be accomplished in some areas in conjunction with the technical adequacy review and is to be completed prerestart. TVA considers that these activities will complete the prerestart corrective i
actions.
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' From a programmatic control standpoint. TVA acknowledges the NRC team's concerns regarding timely closeout of unverified assumptions. These concerns will be considered in future revisions to NEP-3.1.
MECRANICAL ENGINEERING BRANCH (MEB) OBSERVATIONS MEB Loss of Station AC Power Calculation "Sequoyah is committed per the FSAR to achieve and maintain safe shutdown following a loss of station AC power for a period of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The project does not appear to have a est of calculations and analysis that systematically show that adequate ambient temperature is maintained for essential equipment during this postulated event."
TVA RESPONSE TVA has been unable to verify the commitment as stated in this observation.
Instead it appears that the NRC team is basing this observation upon a commitment in Section 8.1.4 of the Final Safety Analysis Report (FSAR) that the " vital batteries shall have adequate capacity for a period of two hours.
without chargers, to provide the necessary DC power to maintain both reactors at hot shutdown, assuming the loss of all AC power sources." This commitment is stated again in Section 8.3.2.1.1.
TVA has demonstrated compliance to this commitment.
It should be noted that this issue is beyond the design basis of the plant and that the issue of total loss of plant ac power (station blackout) is currently the subject of unresolved safety issue rulemaking. TVA's position is that this issue be addressed in response to and consistent with the NRC rulemaking and industry initiatives. However, it is prudent to demonstrate that the actual critical components required to ensure core cooling be operable during the design duration of this event.
TVA has taken steps to demonstrate this design facet as described below.
The normal non-1E auxiliary building ventilation system is designed to remove the normal building heat load.
During a loss of station ac power event, a loss of normal building ventilation would occur; however, it would also be accompanied by a reduction in the building heat load that it is designed to remove.
Given the reduction in the heat load during this postulated sequence and the massive heat sink capabilities of the building for absorbing residual heat, resulting temperatures throughout the building are believed not to exceed the maximum values given on the environmental drawings. This conclusion is based upon existing thermal transient analyses of loss of heating, ventilating, and air conditioning (HVAC) with normal heat producing equipment continuing in operation and upon actual test date. A specific calculation for cach area is, therefore, not needed.
DNE has performed calculations in selected areas of the auxiliary building that confirm this position.
For example TI-874 determined that elevation 714.0 general floor area temperature will not exceed 104 degrees Fahrenheit during a loss of offsite power. The documented environmental abnormal maximum
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. value for this space is 110 degrees Fahrenheit. Although this calculation is 1
not for a total loss of ac, this scenario does result in the loss of normal l
ventilation. The major contributor to the heat load for this case is the shutdown board room water chillor equipment, which would also be eliminated in the blackout condition.
This analysis is representative of the effect of loss
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of all ac power on other plant areas.
Further, special tests were performed in the 6.9-kV and 480-V shutdown board rooms to determine the temperature in these areas during a complete loss of HVAC.
The results are documented in TI-ECS-95 and indicate that steady state j
temperatures are approximately 88' degrees Fahrenheit and 90 degrees f
Fahrenheit, respectively.
These values are both within the abnormal maximum temperature value of 104 degrees Fahrenheit.
The boards in these areas would i
not be energized during the station blackout scenario, and temperatures would be less.
ELECTRICAL ENGINEERING BRANCH (EEB) OBSERVATIONS I
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EEB Turbine AFW Time Delay Relay Sotpoint "An MEB calculation for the turbine driven AFW pressure switch setpoint j
(B44 870323 001 Rev. 5) established process safety limits of 50 and 110 psig with a 25 second maximum time delay for the electrical interlock controls.
QIR MEB 86021 communicated the 25 second time delay requirement to EEB; however, the present design provides for a 60 second time delay, which does j
not satisfy the 25 second time delay limitation.
In this instance, a hardware j
modification appears necessary. TVA initiated a CAQR during the inspection to l
correct the time delay relay setpoint.
This observation remains open pending description of the associated corrective action."
TVA RESPONSE TVA concurs that this observation identifies an oversight in EEB's calculation
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that time delay devices were not a part of EEB's minimum set of essential calculations.
EEB's minimum set of essential calculations has been
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revised to include time delay devices, as stated in the TVA letter to NRC dated July 31, 1987, regarding the DNE design calculation efforts for SQN.
3 All essential time delay devices have been identified, for which EEB is j
performing demonstrated accuracy calculations.
They are documented in a Chandley/ Wilson to Pennell memorandum for the HEB/ Nuclear Engineering Branch (NEB) and a Raughley to Pennell memorandum for EEB.
This effort is addressing j
both the 25-second time delay requirement in EEB-6 and the 15-to 25-second
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and 0.5-second timers in EEB-10.
j EEB HVAC Temperature and Flow Process Safety Limits "HEB 480 volt board room air handling unit temperature switch setpoint calculation B44 860819 004 Rev. O did provide both setpoint and accuracy values, but did not establish process safety limits for a large number of safety-related HVAC temperature and flow measurements.
Some switch safety limits were established at 50 percent of the instrument's tabulated setpoint; however, the adequacy of thi:n selectic.n was not justified for any of the flow instrument loops."
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. l TVA RESPONSE l
The HVAC Instrument Accuracy Calculation B44 860819 004, revision 0, has been superseded by calculation B44 871015 006, which addresses the issues i
identified in EEB-7 and EEB-8.
EEB Setpoint Accuracies for HVAC Temperature and Flow Instrumentation "HEB HVAC calculation B44 860819004 Rev. O, which addressed a number of HVAC
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temperature and flow instrumentation loops, contained predicted accuracies for
.i a number of instruments that did not conform with either the 40 degree minimum or the 104 degree maximum process safety limits.
This calculation did not provide any indication that these nonconformances were unacceptable or that additional resolution was required by EEB.
This calculation also stated that setpoint calculations were not required for flow switch setpoints used to initiate operation of the backup HVAC train even though these instruments perform a safety-related function."
TVA RESPONSE Please refer to the TVA response to EEB-7.
l EEB Containment Electrical Penetration Protection "SCR-SQN-EEB-8676 identified a concern that higher trip settings have been ~
l used to protect the circuits of the penetration assemblies Nos. 52 and 53 against continuous overcurrents. The conductor size used for these electrical penetrations was 12 AWG, and the maximum allowed current through these conductors, without damaging the penetration is 16 amps in accordance with IEEE-317-1983. A trip setting of 20 amps will allow the 16 amps limit to be exceeded without the short being detected in the 16 amp to 20 amp range.
In addition, the penetration manufacturer recommended the current to be limited to 6 amperes. The team feels that the allowed current, in' excess of 16 amps, may result in reduction in the life and/or leakseal capacity of the l
penetration assembly."
TVA RESPONSE l
TVA disagrees with the team's finding in EEB-9, based upon manufacturer's test results as documented in DNE calculation SQN-EPS-002, revision 3.
Section 6.2.1 (page AA) of the calculation documents a continuous current rating for i
penetrations 52 and 53 of 33.3 amps, significantly in excess of the 20-amp l
circuit breaker rating.
Only for the sake of consistency with Table Al of IEEE-317 (1983) was the continuous current rating of these penetrations specified on page AB of the calculation as 16 amps.
This figure is arbitrarily conservative and does not represent a physical limitation on the penetration conductors.
It must be noted, however, that even the 16-amp-rating is within the acceptance criteria of SQN-EPS-002, which considers as valid Exception No. 1, "Next Higher Overcurrent Protective Device Rating,"
from the National Electrical Code 1981 Article 240-3.
TVA therefore considers l
this item adequately addressed, with no corrective action required.
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f' f EEB Pump Start Time Delay Relay Setpoint Calculations "Using the Calculation Cross Reference Information System (CCRIS) database output, the team determined that no calculations had been prepared to support the setpoint or accuracy of 15 to 25 second and 0.5 second time delay relays used in pump start circuits for the ERCW, CCS, and AFW systems.
During the inspection, EEB stated that they are now preparing setpoint calculations for i
some of the safety-related time delay relays. There is no indication that all by the team as an indication of a coordination problem between EEB and MED safety-related time delay relays will be addressed.
This item is also viewed
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with regard to instrument setpoint calculations."
TVA RESPONSE Please refer to the TVA response to EEB-6.
i EEB Component Cooling System Setpoint Coordination "CCS flow switch setpoint calculation B44 MEB 870602 001 included records of telephone discussions between MEB'and EEB regarding flow alarm accuracy
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values, but additional information to technically justify the selection of these values was not documented. This item is also viewed by the team as an indication of coordination problem between EEB and MEB with regard to instrument setpoint calculations."
TVA RESPONSE The flow alarms addressed in this observation are not essential devices needed for safe shutdown of SQN. MEB has classified these alarms as desirable.
EEB has committed to perform demonstrated accuracy calculations on desirablo instrumentation postrestart, in accordance with Procedure Method (PM) 86-02 (EEB)-Electrical Calculations, Item 9a, under " Instrumentation and Controls (I&C) System Calculations."
General Overview for Instrumentation Setpoint Coordination During the subject inspection by the NRC of the SQN calculation validation program, five new observations were noted that related to system performance criteria (i.e., safety limits, setpoints, etc.).
This was in addition to a previous observation identified during the February 1987 inspection regarding a related matter.
These observations have given rise to the concern that some system-related essential calculations may not have been identified by TVA.
Another fecet of this' concern was the issue of consistent coordination among the responsible discipline branches who deal directly with system performance.
The particular concern stated by the NRC team was that documentation for TVA setpoint analyses and their acceptability for each system did not appear to be currently available nor identified as future calculation production work.
As an example, the Auxiliary Feedwater System has a time delay before the water is delivered to the steam generators following an accident. This time delay is composed of two parts:
the control circuit delay and the mechanical /
hydraulic delay once the pump has power.
The control circuit delay may be-
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. determined by analysis as well as demonstrated by preoperational testing.
The acceptability of the total (control + hydraulic) time delay would depend on the performance requirements of the system (safety limits). These requirements are set by the system design or by interface documents.with the Nuclear Steam Supply System (NSSS) vendor and should be documented in the design criteria.
Immediately after the conclusion of the NRC inspection in June, DNE, in response to this concern, conducted a review of the SQN system design criteria j
H documents to determine if (1) system performance requirements are adequately specified, and if (2) calculations and/or tests exist to demonstrate system performance.
As documented in C. A. Chandley and D. W. Wilson's memorandum to W. E. Pennell dated July 29, 1987, additional calculations have been identified as required to complete the definition of system performance criteria. This work is currently underway and scheduled for completion j
consistent with the restart' criteria as documented in Volume 2 of the Nuclear
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Performance Plan.
This work includes 10 calculations to be completed prerestart.
In response to the issue of inadequate coordination among the electrical, mechanical,'and nuclear discipline branches, EEB PM 86-14 has been in effect since August 1986, which defined the interfaces for validation of safety-related setpoints among the discipline branches.
These interface guidelines are currently being proceduralized on a divisional level.
This division procedure, to be issued by November 20, 1987, specifically defines
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the interdiscipline responsibilities regarding the specification of system performance criteria.
CIVIL ENGINEERING BRANCH (CEB) OBSERVATIONS CEB Regenerated CEB Pipe Support Calculations
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"The team reviewed the pipe support calculations which CEB' regenerated for pipe supports 1-H10-555, H10-635, H10-680 and H10-1219.
These pipe supports are located in the component cooling water (CCS) system.
The team notes the following:
(1) CEB's calculation for pipe support H10-635, dated April 28, 1987 (RIMS No.
B25 870429 306) demonstrated that the pipe support failed when CEB considered friction forces. CEB's restart pipe support design criteria requires consideration of friction forces.
However, CEB did not note this deficiency on the calculation cover sheet, or on the CAQR, and Bechtel did not address this deficiency in the supplemental calculation which Bechtel prepared to address the CAQR.
(2) CEB's calculation for pipe support H10-1219, dated April 21, 1987 (RIMS No. B25 870422 302) did not include a thermal check of the pipe support.
CEB's restart pipe support design criteria allows consideration of thermal loads post-restart. However, CEB did not note this unverified assumption on the pipe support calculation cover sheet, or on CEB's pipe support calculation log for post-restart resolution."
_7_
TVA RESPONSE These calculations were part of the calculation effort in support of the SQN Design Baseline and Verification Program.
Presently, they are being reviewed as part of the calculation regeneration effort for pipe supports on rigorously analyzed category I piping for SQN unit 2, as described in a TVA letter to NRC dated August 21, 1987. These calculations will be performed to design criteria SQN-DC-V-24.2, which has superseded the original criteria for design of the subject supports. This is an ongoing effort, and the final regenerated calculations for supports 1-H10-1219 and 1-H10-635 have not yet been issued.
It should be noted that the NRC team's findlngs were based upon preliminary copies of unissued Bechtel calculations.
The revised calculations are scheduled to be transmitted from Bechtel to TVA by November 11, 1987.
l CEB Engineering Assurance Acceptance of CEB's Corrective Action Program for Rigorously Analyzed Pipe Supports "EA Observation C-1 of Engineering Assurance audit report 87-09, dated February 10, 1987 (RIMS No. BOS 870210 001) noted that CEB could not retrieve many of the pipe support calculations required to demonstrate pipe support adequacy in accordance with design criteria SQN-DV-V-24.1.
EA also noted that CEB was generating calculations for modifications to pipe supports which lacked original calculations.
EA accepted CEB's responses to EA Observation C-1, subject to additional EA review.
However, the team considers EA's acceptance of CEB's program to identify and regenerate missing pipe supports-
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to be premature, based on the team's review of CEB's pipe support program during the period June 1-5, 1987. The team's review indicated that CEB has l
not yet documented a corrective action program to address the generic l
implications of CEB's design verification of 201 of the 791 pipe support calculations which the DBVP project identified as missing and which CEB regenerated."
i TVA RESPONSE i
l Engineering Assurance (EA) requested additional information from CEB by memorandum from A. P. Capozzi to R. O. Barnett dated June 12, 1987, concerning the expansion of the pipe support calculation regeneration effort.
EA will review any significant changes to the program and will perform verification of adequate implementation before closure of EA Observation C-1.
A general overview of TVA's program plan for the calculation regeneration effort was submitted to NRC on Au8ust 21, 1987.
The SQN unit 2 restart screening criteria for these calculations were transmitted to NRC as an enclosure to a letter dated August 31, 1987.
The regeneration effort for pipe support calculations on rigorous analysis piping includes the following steps:
1.
Issuing design critoria that meet the FSAR commitments.
2.
Reviewing all existing rigorous pipe support calculations.
3.
Regenerating all missing or inadequate rigorous pipe support calculations.
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. CEB Technical Adequacy of Miscellaneous Structural Steel "To determine the technical adequacy of miscellaneous structural steel, CEB reviewed 54 featuros from approximately 400 drawings.
In most cases there are more than 1 feature per drawing. Therefore, the number of features reviewed might be a small percentage of the total number of miscellaneous etructural steel features at Sequoyah.
The NRC team questions the validity of CEB's conclusion that miscellaneous structural steel is technically adequate without increasing their sample size."
TVA RESPONSE TVA's FSAR does not specifically define the requirements for the design of miscellaneous steel, but the steel was-originally designed to the requirements of design criteria SQN-DC-V-1.3.2, which was issued in 1970.
This design criteria addressed Operational Basis Earthquake (OBE) and Safe Shutdown Earthquake.(SSE) design conditions and specified a maximum allowable stress of 0.9Fy (where Fy equals allowable yield stress).
In 1980, design criteria SQN-DC-V-1.3.3.1 was issued that defined the requirements for new building additions. In 1983 this criteria was revised to allow its usage for evaluation and modification of existing structures.
This design criteria is compatible with the NRC Standard Review Plan (SRP) loads, loading combinations and allowable stresses. For a comparison of the two criteria, refer to
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Table A (attachment 1).
Since SQN-DC-V-1.3.3.1 is compatible with the industry' standard, it has been i
utilized in recent years for evaluations'and modifications of steel j
structures.
It was also utilized for the technical review of the calculations for the 54 miscellaneous steel features or portions of features discussed in
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the observation.
The review was conducted utilizing detailed Technical Review j
Plan SQN-CEB-87-02.
The review determined that 49 of the features met the l
dusign requirements of SQN-DC-V-1.3.3.1.
It also identified five CAQRs, three of which dealt with inconsistencies in the design loads utilized versus vendor i
loads. These CAQRs are being evaluated to verify that the structures meet operability requirements. After restart, miscellaneous steel calculations will be reviewed and revised as necessary to fully document the design basis.
For each of the CAQRs, corrective action has been developed and issued; and generic evaluations for the other nuclear plants have been initiated.
After restart, loadings will be reconfirmed and walkdowns completed to establish as-built configurations for features associated with the CAQRs; and the features will be reevaluated to ensure that they meet design criteria requirements.
These postrestart activities will be scheduled by December 31, 1987.
In addition to the above TVA has completed, or will complete, the following activities:
(1) TVA is conducting, as part of the evaluation of the CAQRs concerning vendor loads, a technical review of the additional miscellaneous steel calculations for equipment supports.
This review will ensure that the appropriate vendor loads have been utilized.
The review is scheduled fot-completion by November 30, 1987.
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p (2) TVA has regenerated 38 calculations :for miscellaneous steel st'ructures.
The calculations were found to meet design. criteria requirements for 36 features.
Two features did not meet. criteria requirements, and CAQRs have been issued. ' Corrective action, action required to prevent recurrence,.
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and restart evaluations have been. developed and issued for these CAQRs.
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CEB Conduit'and HVAC Duct Support Calculations.
I "CEB's review of recently. regenerated conduit-(5) and HVAC duct (4). support calculations showed numerous discrepancies between the calculations and the associated.designLeriteria. The team's review of the CEB's findings'on these
- 9 calculations showed that.the a'nalysis performed was incomplete and inadequate. specifically clamps and welds were not evaluated.
The findings also demonstrate to the team the contract personnel used to regenerate these
.j calculations lack knowledge about the applicable CEB design ~ criteria and need S
specific training regarding TVA standard practices."
TVA RESPONSE During the. regeneration of calculations that had been identified as noneetrievable'..TVA conducted independent technical reviews of: selected calculations. The conduit and HVAC support calculations were reviewed under.
q Detailed Technical Review Plan SQN-CEB-87-03.
During the initial review, five j
conduit and four HVAC support calculations were reviewed, and deficiencies in -
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all nine calculations were identified. The review was. expanded, and eight i
.s.
additional calculations were reviewed wit,h deficiencies noted in one j
calculation.
j In order to prevent recurrence of the deficiencies that were identified.by the j
technical review of the calculations, the following documents have been issued:
1.
A " Lessons Learned" memorandum has been issued to designers (TVA and contractor) defining the deficiencies and emphasizing need to improve quality.
1 2.
A branch instruction has been issued that clarifies the duties and responsibilities of each individual (TVA and contractor) involved in the development of design calculations.
In addition, because of. identified issues in the design of conduit, conduiL supports, HVAC ducts, and HVAC duct supports, TVA is -verifying the technical '
adequacy of these features through sampling programs.
CAQRs SQT870626 and SQT870843 have been generated to resolve these' issues.
The following corrective action is being implemented for the resolution of the CAQRs.
Conduit and Conduit Supports 1.
The design criteria have been revised to' correct design deficiencies described in the CAQR. The design basis is defined in attachment 2.
2.
Calculations for 60 representative " worst-case" supports and associated conduit runs have been developed.
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In addition to the worst-case supports, several of the conduit typical supports, representing approximately 90 percent of the support ropilation, were evaluated analytically.
4.
Some of the concerns in the CAQR were addressed through the usage of the earthquake experience data.
-l These activities have been completed. The conduit and conduit supports have j
been determined to meet operability requirements and thus are acceptable for
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restart.
(See attachment 3.)
HVAC Duct and Supports i
1.
The design criteria have been revised to correct design deficiencies i
described in the CAQR. The design basis is defined in attachment 2.
j 2.
Five repr-intative worst-case portions of ducts and supports have been identified and are being evaluated.
3.
Interfaces between rigidly and flexibly supported ducts have been identified and are being evaluated.
These activitlis will be completed by November 9, 1987.
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CEB CEB tirrective Action program Description "This observation was created to track a previously identified NRC request.
NRC letter from J. M. Taylor to S. A. White dated March 5, 1987 stated:
'We await receipt of written information requested during the inspection describing the current calculation review effort scope.
In particular we will be concerned with the description of the scope and depth of past reviews (beyond the standard quality or design verification calculation check) that TVA is relying upon to justify not examining civil engineering calculations in the current review program. '
Also NRC inspection report 50-327/87-06, 50-328/87-06 forwarded to TVA on April 8, 1987 stated in Section 4.3 for CEB the following request:
The team noted that the specific CEB program description needs to be formally submitted to the NRC in order that NRC can assess the CEB calculation effort scope of review. One particular concern is the scope and depth of past. reviews (beyond the standard quality or design verification calculation check) that TVA is relying upon to justify not examining civil engineering calculations in the current review program.
NRC reviewed two calculations in an area which TVA was not planning to review based on the CEB teview of previous verification programs.
In one of these calculations, the NRC found an unjustified assumption for concretc compressive strength of $300 psi vs. 4000 psi as stated in the FSAR and that allowable strcas in concrete and steel had been exceeded by 19% and 16%, respectively, also without justification (OBS CEB-1). This observation has given NRC cause to question TVA's methodology for
A:
... (determining which areas in'CEB require further review. LTVA was requested i-to provide justification of their. rationale for excludins certain. areas of CEB design from being reviewed on a. sampling basis'for technical adequacy.
In.this regard-TVA was also requested to provide descriptions.
of. the scope and depth of: previously conducted internal.and external-reviews that they are using as a basis for not performing current detailed
' I technical reviews of calculations similar to what-has been'done in the 1
other three technical branches."
1 TVA RESPONSE 1
Material has been submitted to URC that addresses the information requested as
'f follows:
1.
Overall description of CEB Calculation Program.
2.
Description of the scope and depth of past review.
I 3 '. Basis and results of CEB's Technical Adequacy Review process.
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d 4.
Pipe Support Calculation Regeneration Program for Rigorous Analysis Systems.
J The information is contained in the following transmittals:
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1.
R. L.'Gridley to Keppler dated July 31, 1987
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1 1
2.
R. L. Gridley to NRC dated July 31, 1987 3.
R. L. Gridley to NRC dated August 21,.1987
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As a result of the NRC. Integrated Design Inspection 'IDI), several calculation j
areas'have either received or will receive additional reviews.
These areas e
include concrete structure shear checks in Category I buildings, equipment supports, tank anchorage, and flexibility of building steel floors and roofs.
These additional reviews, along with the previous work that has been accomplished, ensure the adequacy of the CEB CaZeulation Program.
TVA will j
address the' issues identified during the IDI in a separate transmittal to NRC.
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- ATTACHMFMi.1 I
Table'A' SQN-DC-V,-1.3.3.1 SQN-DC-V-l'.3.2 Allowable-Allowable 1
Loads:
Stresses Loads Stresses 1
. OBE AISC-OBE.
~AISC
')
OBE + Thermal
-1.5 x AISC d
1 SSE_
0.9 Fy SSE'
.1~.6 x AISv Tension ~& Bending'= 0.96-Fy'(Compact Sections)
- Tension &; Bending =,1.06Fy(NoncompactSections){
Shear = 1.6 x 0.4:=.0.64 Fy Weak Axis Bending = 1.2 'Fy SSE'+ Pipe Break-
.1.7-x AISC Pressure-Loadsh' Tension & Bending ~m 1.7 x-0.6 Fy,
= 1.02 Fy (Compact Sections) ;.:
. Tension' & Bending = 1. 7 x 0.66
=<1.12 Fy (Noncompact Sections),
Shear = ' 1. 7 x 0.4 = 0.68 ' Fy
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Weak Axis. Bending =.1.28'Fy a
d
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NOTE:.Not all loading. combinations that are addressed'in SQN-DC-V-1.3.3.1-arc shown. For the specific loads, loading combinations, and allowable stresses, please refer to the design criteria'
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ATTACHMENT 2
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CONDUIT SUPPORTS
{
l Section 3.10.2, PC 3.10-7 1
The following is the FSAR rection on conduits and conduit supports:
" Conduit Banks and Supports 1
1.
Restraint Measures
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The Category 1 electrical exposed conduit supports,'and electrical j
conduit box supports have been designed to provide vertical and horizontal support for.the spacing recommended in TVA Construction Specification."
l TVA designers presantly use Design Criteria SQN DC-V-13.10 for design of
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conduit supports, HVAC DUCT AND SUPPORTS _
Design of HVAC duct and-supports is not addecssed in the FSAR.
1 TVA designers presently use Design Criteria SQN DC-V-13.8 for design of HVAC
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supports.
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'ATTACRMENT 3 Conduit Supports were reviewed under -CAQR SQT870626 for 60 representative worst-case supports with the following results.
'No. of Worst-Case-No. of Supports No.lof Spanaf l
Supports Meeting the.
of Conduit Meeting d
Evaluated Requirements of Requirements of Design Criteria Design. Criteria.
SON DC-V-13.10 SON-DC-V-13.10
]
60 58**
58*
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Conduit Operability Determination 1
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2 The allowable conduit stress limit for these two cases is 5,217 lb/in ;
2 however, yield is 13,000.lb/in.
The actual stress values were as follows:
a 1
1.
Support Tag No. AB25 i
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Max conduit stress 9,800 lb/in 2.
Support Tag No. AB45 s
2 Max conduit stress 8,800 lb/in Conduit Support Operabilty Determination 1
1 All 60 of the structural steel supports met design criteria
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requirements. However, two supports had clamps that' exceeded allowable values for the interaction equation. These interaction values were 1.3 i
and 1.06.
Minimum factors of safety of 2.0 were used to develop the allowable clamp loads. Therefore,'there would be no loss of function of-the support since'only a maximum of 65 percent of the capacity of.the clamp is used in any one direction.
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ENCLOSUREI2I i
o 5;,
' LIST OF NEW COMMITMENTS
'FOR:SEQUOYAH NUCLEAR PLANT.(SQN) UNIT 2 1;.
DNE will' issue a~ procedure by November 20[1987',. that-defines the-
- q interdiccipline' responsibilities'regarding the specification of system performance criteria.
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TVA wi11' schedule ~the postrestart activities. associated with~
miscellaneous, steel calculations by' December.31,.1987.
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3.
' By November 30,11987, TVA will complete'.the technicalireviews-of-the' additional miscellaneous steel calculations for' equipment-supports to
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ensure.that the' appropriate vendor loads have been,utilli:ed, n
4.
By November 9, 1987, TVA will' complete'the' evaluations of._five.-
representative worst-case portions of'HVAC ducts and. supports and the:
interfaces between-rigidly and; flexibly' supported ducts.
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