ML20236B148
| ML20236B148 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 10/09/1987 |
| From: | James Keppler NRC OFFICE OF SPECIAL PROJECTS |
| To: | White S TENNESSEE VALLEY AUTHORITY |
| References | |
| NUDOCS 8710230369 | |
| Download: ML20236B148 (115) | |
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. UNITED !;YATES
~g 8-d NUCLEAR REGULATORY COMMISSION i'
WASHINGTON, D. C. 20555
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October 9, 1987.
H Docket Nos. 50-327/328 Tennessee Valley' Authority j
ATTN: Mr. S. A. White i
Manager of Nuclear Power 6N 38A Lookout Place 1001 Market Street.-
' Chattanooga, Tennessee 37402-2801
Dear Mr. White:
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SUBJECT:
ITEMS IDENTIFIED BY THE INTEGRATED DESIGN INSPECTION REQUIRING RESOLUTION PRIOR TO RESTAR1 0F SEQUOYAH UNIT 2 1
At the July 8,1987 meeting between TVA and NRC, the NRC presented its schedule for the conduct of the Integrated Design Inspection (IDI) of Sequoyah Unit 2.
One_of our schedular commitments was to formally identify, prior to issuance of the IDI report, the items requiring resolution prior to restart ~of Sequoyah Unit 2.
In fulfillment of our commitment' to identify fonnally the IDI, restart items by October 9, 1987, we have attached the specific Draft Deficiencies and Unresolved Items which in our. Judgment must be resolved prior to restart.
TVA must also evaluate the generic implications.of these items to Sequoyah Unit 2, prior to restart.
The basis for the staff's restart determinations is the NRC' approved restart criteria submitted by TVA as.part of the Nuclear Performance Plan. Although'the numbering of the items has been changed to' reflect how they will appear in the final report, each of the attached items has been fully discussed with the TVA staff and identified as'a restart. item during the-i inspection.
As a result of the IDI team's review of the calculations supporting the structural design of safety-related buildings and the design of equipment-i foundations, concerns were raised regarding the adequacy of the calculations.
The team found a number of examples where calculational assumptions were not representative of the as-constructed design, fundamental design considerations 1
had not been evaluated, and discrepancies existed between analyzed and installed' support configurations.
The team also questioned some' aspects of the reinforced concrete design, in particular rebar placement, as well as the development of the vertical response spectra for the steel containment vessel.
These concerns were described to TVA in an exit meeting held on September 11, 1987.
At the exit meeting, you stated that TVA had additional information relating to NRC's structural design concerns that were not available to the team during the IDI.
In order to maintain the IDI schedule, we could not extend the inspection to allow us to revisit the structural issues which had already been discussed at great length with your staff during the inspection.
8710230369 871009 PDR ADOCK 05000327 9Q PDR m
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Mr. S. A. White: October 9, 1987 requested that a substantial sample of structural calculations be reviewed.
This review was to be conducted by personnel external to TVA that had not been involved in-the original design of Sequoyah in order to provide a more objective review of the TVA structural'~ design practices.
At the exit meeting, you stated that TVA has additional information relating to NRC's structural design concerns that were not available to the team during the IDI.
In order te maintain the IDI schedule, we could not extend the inspection to allow us to revisit the structural issues which had.already been discussed at great length with your staff during the inspection. Two meetings were subsequently scheduled to discuss this new information, however, on both occasions these meetings were cancelled by TVA.
The NRC has received the TVA submittal dated October 2, 1987 which provided TVA's position on fourteen of the nineteen concerns related to the adequacy of structural calculations. We will, during the IDI follow-up inspection, fully evaluate this new information and reassess the scope of the structural calcula-tion review previously requested. The IDI follow-up inspection is currently scheduled for October 22 and 23 and the week of November 2, 1987. At this time we plan to issue the final IDI report on or ahead of our committed date of November 6, 1987.
Sincerely.
Original signed by:
James G. Keppler, Director Office of Special Projects Attachments:
As stated cc: see next page DISTRI80 TION w/ attachments:
Docket File JKeppler/JAxelrad IDI Team NRC PDR SEbneter RParkill Local PDR SRichardson JLeivo TVA Projects BDLiaw JHaller TVA R/F JZwolinski (5)
JHoughton Plant File GZech 0Mallon JPartlow SConnelly RMcFadden BGrimes BHayes RParkhill CHaughney.
CJamerson AVduBouchet GImbro OGC-Bethesda AUnsal RShewmaker FMiraglia Schen JFair EJordan BMasterson ACRS (10)
VFerrarini TVA-Bethesda
) HWang 4 (hM M (cu cum OFC
- NRR/RSIB 4
- NRR/RSIB 4
- NRR/DRIS 4
- 0SP:TVA/DD
- 0SP:TVA/D :0SP/DD :0SP/D D
____..:____............:.______....___:______________:... M ___.....g./ A........ g. ',W..__e. y f(_.
NAME
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- CHaughney
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- SRichardson
,tgEbneter :JAx -
- JKeppler
l Mr. S. A. White requested that a substantial sample of structural calculations be reviewed.
b This review was to be conducted by personnel external to TVA that had-ftot been involved in the original design of Sequoyah in order to pro de a more objective review of the TVA structural design practices.
At the exit meeting, you stated that TVA has additioyal information relating to NRC's structural design concerns that were not a flable to the team during the IDI.
In order to maintain the IDI schedule, could not extend the inspection to allow us to revisit the structural issy which had already been discussed at great length with your staff durinpthe inspection.
Two meetings were j
subsequently scheduled to discuss this new information, however, on both occasions these meetings were ca elled by TVA.
The NRC has received the
- submittal dated October 2, 1987 which provided TVA's position on four of the nineteen concerns related to the adequacy of structural calculatip s.
We will, during the IDI follow-up inspection, fully i
evaluate this new)t1 formation and reassess the scope of the structural calcula-tion review pre fously requested. The IDI follow-up inspection is currently scheduled for ctober 22 and 23 and the week of November 2, 1987. At this time we plan to ssue the final IDI report on or ahead of our committed date of Novembety,1987.
Sincerely, James G. Keppler, Director Office of Special Projects i
Attachments: As stated cc: see next page DISTRIBUTION w/ attachments:
Docket File JKeppler/JAxelrad IDI Team NRC PDR SEbneter RParkill Local PDR SRichardson JLeivo TVA Projects BDLiaw JHaller TVA R/F JZwolinski JHoughton Plant File GZech OMallon JPartlow SConnelly RMcFadden BGrimes 8 Hayes RParkhill i
CHaughney CJamerson AVduBouchet GImbro OGC-Bethesda AUnsal RShewmaker FMiraglia Schen JFair EJordan BMasterson ACRS(10)
VFerrarini i
TVA_Bethesda HWang OFC
- NRR/RSIB g 7
- NRR/RSIB g :NRR f IS
- 0SP:TVA/D
- 0SP/D
/:CHaughney
- JPc
- SEbneter
- JKeppler NAME
- EImbro:acm
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7 Mr. S.'A. White Tennessee Valley Authority Sequoyah Nuclear Plant
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General Counsel Regional Administrator, Region II i
Tennessee Valley Authority U.S. Nuclear Regulatory Commission' 400 West Summit Hill Drive 101 Marietta Street, N.W.
E11 B33 Atlanta, Georgia 30323 Knoxville, Tennessee 37902 Resident Inspector /Sequoyah NP Mr. R. L. Gridley c/o U.S. Nuclear. Regulatory Commission Tennessee Valley Authority
-2600 Igou Ferry Road SN 157B Lookout Place Soddy Daisy, Tennessee 37379 Chattanooga, Tennessee 37402-2801 Mr. Richard King 1
Mr. H. L. Abercrombie c/o U.S. GA0 Tennessee Valley Authority 1111 North Shore Drive Sequoyah Nuclear Plant Suite 225, Box 194 P.O. Box 2000 Knoxville Tennessee 37919 Soddy Daisy, Tennessee 37379 Tennessee Department of Mr. M. R. Harding Public. Health Tennessee Valley Authority ATTN:
Director, Bureau of Sequoyah Nuclear Plant Environmental Health Services P.O. Box 2000 Cordell Hull Building
' Soddy Daisy, Tennessee 37379 Nashville, Tennessee 37219 Mr. D. L. Williams Mr. Michael H. Mobley, Director Tennessee Valley Authority Division of Radiological Health 400 West Summit Hill Drive T.E.R.R.A. Building W10 B85 150 9th Avenue North Knoxville, Tennessee 37902 Nashville, Tennessee 37203 County Judge Hamilton County Courthouse Chattanooga, Tennessee 37402 4
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D2.2-1 (DEFICIENCY)
REVISION 4 October 7, 1987 1
DESIGN PRESSURE OF ERCW SYSTEM l
Background
The stated design pressure of the ERCW system is 160 psig except for the piping in the ERCW pumping station which is 180 psig.
The Code in effect for ERCW piping is ANSI B31.1 - 1967 which requires that the " design pressure not be less than the maximum sustained operating pressure within the piping and shall include allowance for pressure surges". TVA's associated design pressure s
calculation identifies many ERCW components with a design pressure of 150 psig l
or less, and attempts to justify lowering the system design pressure to 150 psig by using administrative measures to maintain a pressure of less than 124 psig in the pump discharge header. TVA's design pressure calculation utilizes frictional flow losses in the system to establish the maximum sustained operating pressure.
Description The design pressure calculation for the ERCW system (Ref. 1) has the following inconsistencies /non-conservatism not identified previously in the other IDI team findings:
1.
System realignments to accommodate equipment outages for maintenance, etc., was not considered.
Therefore, the frictional losses will be reduced and the pressure will increase.
2.
The maximum river elevation (726.8 feet) that occurs during the design basis flood was not used; instead the normal maximum river elevation of 683.0 feet was used - this difference increases the maximum sustained operating pressure by 19 psi. Also, no justification is provided describing how the equipment / piping will be protected from overpressurization when the river water level increases above 683.0 feet.
3.
The current design criteria and mechanical flow diagrams establish the associated portion of the ERCW as having a design pressure of 160 psig, not 150 psig as recommended in the subject calculation. Thus various components within the ERCW system that are currently rated at 150 psig co not meet the system design pressure of 160 psig.
4.
The governing code, ANSI B31.1 - 1967, does not permit lowering the desi ~
pressure (from 160 to 150 psig) by relying on pressure reducing devicer without providing relief valves.
Basis The code of record for the piping in the ERCW system is ANSI B31.1, which states that the design pressure not be less than the maximum sustained operating pressure. Contrary to the requirements of ANSI B31.1, a conservat h e DRAFT
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'" maximum' sustained operating pressure" has not been established and various components in the system are being pressurized to pressures more than their
. design pressure.
Impact on Design If the design pressure is' maintained above 150 psig, certain components would have to requalified to the new design pressure (e.g., Electrical Board Room Cooler,.CCP Oil Cooler, SI Pump 011 Cooler Station and Auxiliary Control Air.
Compressor Coolers, etc.) and'the ERCW system piping would have to be qualified
- to the higher design pressure by hydrostatic' test as specified by the code ~ of:
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1 record.
Extent.
f During a previous inspection,' concerns were raised about the design pressure of the Component Cooling Water system.
Review of design pressure adequacy for
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other TVA designed systems appears to be warranted.
References 1.
"SNP-ERCW Design Pressure Calculations" SON-KE-D053/ HCG-GEB-090386, Rev. 1, 3/25/87, 2.
ANSI B31.1, " Power Piping", 1967 1101.2.2, 102.2.4, 102.2.5a & b.
3.
Essential Raw Cooling Water System Design Critoria, SON-DC-V-7.4, Rev. 2.
7/11/86, 4.
Mechanical Flow Diagram - Essential Raw Water Cooling System sheets 1-6.
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02.2-2
-(DEFICIENCY)
' REVISION 3 October 6, 1987 PROCEDURE NOT AVAILABLE TO. LIMIT ERCW PUMP DISCHARGE PRESSURE.
Background
The design pressure for the ERCW system,. downstream of.the ERCW' pumping station H
header valves, is 160 psig. TVA has performed calculations to confirm compatibility between system design pressure and component design pressure.
ANSI B31.1.0 - 1967 requires'the system design pressure not be less than the
" maximum sustained operating-pressure" and "where pressure reducing valves are:
-used, one or.more relief devices or' safety valves.shall be provided on.the low pressure side" or the. system should be designed for the full initial pressure.
-Description
'The~TVA design pressure calculation identified several components with design pressures of 150 psig or lower. To protect those components with a' design pressure of 150 psig from overpressurization, the calculation. (Reference 2) reconsnanded administrative 1y limiting ERCW pump operation so that the pump discharge pressure does not exceed.124 psig.
In March, 1987, no approved l
procedure or methodology was identified to accomplish this recommendation.
d However, a memorandum dated July 31, 1987 from DNE to operations requested that j
the. operating procedures be revised to restrict ERCW' pump discharge pressures j
above 124 psig.
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a Basis j
FSAR Section 17.1A.5 states that, " activities affecting quality are prescribeo l
by documented instructions in the form of... procedures.". Contrary to this e procedure was not implemented to control the ERCW pump discharge pressure.
Impact'on Design The lack of an approved procedure to control ERCW pump discharge pressure c u result in overpressurization of the following safety-related components:
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Electric Board Room Cooler Air Conditioner.
CCP Oil Cooler IB-B.
g Turbine Driven AFW Pump.
SI' Pump 011' Cooler.
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Extent This lac' of approved operating procedures issued in a timely manner to addre s,
k design deficiencies' identified by the Division of. Nuclear Engineering of neea to be addressed generically by TVA.
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DRAFT.
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References
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FSAR Section 17.1A.5, " Instructions, Procedures and Drawings".
'l 2.
FSAR 9.2.2, Figure 9.2.2-1, R24 (47W845-1).." Mechanical Flow Diagram -
J Essential Raw Cooling Water System".
3.
TVA Calculation SON-KE-0053/ HCG-GEB-090386, "ERCW - Design Pressure Calculations" Rev.-1, 3/25/87.
1 4.
Memorandum to L. M. Nobles from J. B. Hosmer, " Recommendation'to Revise Operating Procedures", July 31, 1987,'B44 870731 014.
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D2.2-3 (DEFICIENCY)
REVISION 4 October 7, 1987 OVERPRESSURIZATION OF AUXILIARY AIR COMPRESSOR COOLERS
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Background
The design'pressureJof the ERCW system downstream of the ERCW pumping station
. header isolation valves is 160 psig. TVA has reviewed component design _
pressures and performed a calculation to confinn compatibi_lity between the system' design pressure and component design pressures._ The-ANSI-831.1.0~- 1967 and ASME'B&PV code Section III (ND) - 1971. requires the system to be designed such that the internal design' pressure. shall not be less than the " maximum '
d sustained pressure" and "where ' pressure reducing valves are used,'one or more H
relief devices'or safety valves shall beiprovided on the low pressure side'or the piping and equipment on the low pressure side shall meet the~ requirements for_ full' initial pressure".
Description In reviewing the TVA calculation for established ERCW design pressure-(Ref. 5),
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the IDI team identified an overpressure condition for' auxiliary air compressors cylinder jacket cooler and aftercooler. The design pressures for these components were 67 psig and 75 psig, respectively. LTVA has not identified any approved resolution to this overpressurization of the auxiliary air. compressor coolers. A memorandum from DNE to operations (Ref. 7) requested that throttling be employed as a means of overpressure protection. However, throttling is not in compliance with the governing code (ANSI B31.1) or ASME 1
Section III as an acceptable means of overpressure protection.
Basis The code of record for the piping to the auxiliary l air compressor'is ANSI B31.1, which states that the design pressure not be'1ess than the maximum sustained operating pressure. Contrary to the requirements of ANSI B31.1 the design pressure of the auxiliary air compressor is less than the maximum sustained pressure of the ERCW system.
Impact or Design Resolution may include addition of relief devices or re-rating the auxiliary air compressor to the system design pressure (including hydrotest). The overpressure condition could result in a system rupture that could prevent the ERCW system.from performing its safety-related function.
Extent As a result of-not providing-adequate overpressure protection, other similar methods, not in accordance with the code of record,'may be used to provide overpressure protection in other safety-related systems which could result in adversely affecting safety-related equipment or system functions ~.
DRAFT References.
1.
FSAR-9.2.2., Figure 9.2-2-4 R-24 (47W845-4), " Mechanical Flow Diagram -
Essential Raw Cooling Water".
2.
FSAR 3.2, Table 3.2.2-2, " Summary of Codes and Standards for Components of.
the Sequoyah Nuclear Plant for Procurement After April 2, 1973".
3.
ASME B&PV Code Section III, 1971 Edition.
l 4.
ANSI B31.1.0, " Power Piping", 1967.
l 5.
TVA Calculation SON-KE-0053/ HCG-GEB-090386, "ERCW Design Pressure
.j Calculations", Rev. 1,3/25/87.
6.
FSAR 9.2.2.8, " Design Codes".
7.
TVA memorandum, " Recommendation to Revise Operating Procedures" from J. B. Hosmer to L. M. Nobles, July 31, 1987, 844 870731 014.
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D2.2-4 (DEFICIENCY)
_ REVISION 4 October.7, 1987 j
i OVERPRESSURIZATION OF STATION AIR COMPRESSOR COOLERS i
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.Gackground The design pressure of the ERCW system downstream of the ERCW pumping station j
header isolation. valves is 160 psig. TVA has reviewed component design pressures and performed a calculation to confim compatibility between the-
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system design pressure and component design pressure. The ANSI:B31.1.0..1967 q
code requires that 'the -internal design pressure _ shall not _be less than the
" maximum sustained operating pressure" and "where pressure reducing valvesLare used, one or more relief devices or safety valves shall be provided on the low pressure side" or. piping or equipment on the low pressure side.shall meet the i
requirements for full initial pressure.
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Description.
I In reviewing the TVA calculation for establishing ERCW design pressure (Ref. 5) the IDI team identified that the station air compressor' cylinder jacket cooler 1
and intercooler had design pressures substantially lower than the system design
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pressure. The design pressures for these components were 67 psig and 50 psig, t
respectively. The ERCW system design pressure is 160 psig.
Basis Section 9.2.2.8 of the FSAR, " Design Codes for ERCW System". defines the code of record for the piping to the station air compressor as ANSI B31.1. ANSI t'31.1 states that the design pressure not be less than the maximum sustained operat-
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ing pressure.
Contrary to the requirements of 831.1 the design pressure of the station air compressor is less than the maximum sustained pressure of:the ERCW system.
Impact on Desian-Resolution may include addition of relief devices or rerating the station air compressor to the system design pressure (including hydrotest). The over-pressure condition could result in a system rupture that cculd. prevent the ERCW system from perfoming its intended safety function.
Extent As.a result of not.providing _ adequate overpressure protection, other similar non-code approved methods for providing overpressure. protection may be utilizec in safety-related systems and those interfaces with non safety-related systems which could result in adversely affecting safety-related equipment or system
' functions.
DRAFT.
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i References 1.
FSAR_9.2.2.8, " Design Codes".
2.
FSAR 9.2.2, Figure 9.2.2-5, R-15 (47W845-5), " Mechanical Flow Diagram -
l Essential Raw. Cooling. Water System".
3.
FSAR 3.2, Table 3.2.2-2,." Summary of Codes and Standards for Components of-the Sequoyah Nuclear Plant for Procurement After April 2, 1973".-
4 ANSL B31.1.0, " Power Piping",1967.
5.
TVA Calculation SON-KE-0053/ HCG-GDB-090386, "ERCW - Design Pressure I
Calculations", Rev. 1,~3/25/87.
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H 02.2-5 (DEFICIENCY)
REVISION'4 10ctober 7, 1987 EVALUATION OF FAILURE OF ERCW NON-SEISMIC PIPING
Background
.A portion'of ERCW system is non-seismical y designed and supplies cooling water l
to the station air compressors which are located in the turbine building.
Failure of.this non safety-related piping during a seiscic. event must be.
analyzed to evalaute>the consequences:of this failure on the operability of.the safety-related portion of the ERCW system.
Currently, failure of these lines
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is detected via high flow alarms in the main control room which,.in turn, require operator action'to isol. ate the. break by closing a single motor operated butterfly valve'in each train.
Description TVA has not evaluated the consequences of failure of the non-seismically designed piping on the operability of the safety-related portion of the ERCW system.- The analysis needs to demonstrate that-ERCW safety-related equipment i
receives the necessary cooling during the. time' period it takes the operator to isolate the non-seismic portion (Refer to Deficiency 05.2-10) assumingLa double-ended guillotine rupture.in the non-seismically-designed piping.
Further, during discussion of this deficiency,;TVA' attempted to. justify. system operability by performing a calculation which only postulated a. critical crack rather than a double-ended pipe break!.It is the opinion.of the. team, that this;is a misapplication of the guidance.provided in the-Standard Review Plan.
Chapter 3.6.1, that addresses postulated failures in fluid systems outside containment. The team is concerned that there may be other safety systems with non-seismic piping in seismic Category II buildings in which the. determination of system functionality was not based on postulation ofLa double-end pipe break.
Basis FSAR Section 9.2.2.1 states that, " Provisions are made to ensure a continuous flow of cooling water to those systems and components necessary for plant safety either during normal operation or under accident' conditions". Also, SQN-DC-V-7.4, Rev. 2, 13.7.1.2 states, "Non-safety related ERCW features shall be designed such-that their failure would not jeopardize. safety-related features." Contrary to these commitments, TVA has not demonstrated via calculation, that the ERCW system would remain functional.following a seismic event that results'in a break of the ERCW piping supplying the station air compressors.
In' addition,SQN-DC-V-7.4,13.7.1.8 states, "... the ERCW system design shall incorporate features.to detect'and isolate a' double-ended guillotine break in a nonqualified portionLof the. system. piping." Contrary to the requirement to postulate a double-ended guillotine break,'TVA only postulated a' crack in'the non-seismically design ERCW piping located in a Category II structure.
DRAFT.
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ERCW may not be able~to supply the necessary cooling water to the required safety-related equipment during the time period it takes the operator to detect and isolate this break. Automatic isolation may be' required to isolate the safety and nonsafety-related portions of the ERCW system following a break in the nonsafety-related piping.
Extent i
This finding raises concerns regarding how seismic /non-seismic system inter-faces were designed. Evaluation of system boundaries for other safety-related systems having an interface with non-seismically designed piping is warranted.
RefErgces 1.
ERCW Flow Diagram #47W845-5, Rev. 1, 2/28/87.
2.
FSAR Section 9.2.2.1, "ERCW - Design Bases".
3.
Design Cr.iteria SON-DC-V-7-4, "ERCW SYS (67)", Rev. 2, 7/11/86.
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02.2-7 (DEFICIENCY)
REVISION 4 October.8, 1987 DETERMINATION OF ERCW PUMP HOUSE AMBIENT-I TEMPERATURE'FOR ENVIRONMENTAL QUALIFICATION (MILD)
1 Background
- J Reference 1 is the original calculation which established the maximum and minimum temperature limits for the ERCW pump' house at Elevation 705'.
It was reviewed and concerns were identified by the IDI team regarding the technical basis for the maximum and minimum temperature limits.
In-response to these concerns, TVA provided (Reference 3) a more rigorous analysis based on steady state conditions. Additionally, in response to the IDI team's concerns, TVA's Division of Nuclear Engineering issued a memorandum to the' Division of Nuclear Operations which requested that the ERCW Pump House ventilation fan be turned off in winter months when the temperature in the ERCW. pump house is below 65 F to avoid a potential freezing condition in the building.
Description In reviewing Reference 1, the team observed that the upper temperature limit was not based on technically justified heat gains and losses, but rather-on unjustified assumptions.
In addition, Reference 1 contained unconservative 1
assumptions in that the lower temperature limit did not assume the-worst case loss of non-1E electrical equipment (i.e., the room heater not operating and the ventilation' fan in operation). Also heat losses through uninsulated concrete walls were incorrectly naglected.
Reference 3 dated 8/26/87 superseded Reference 1 dated 9/5/86 for the portion of the mild environmental calculations associated with the ERCW pump house.
However, Reference 3 did not adequately (establish the lower temperature limit associated with the abnormal condition i.e., LOCA) as it contained unconservative assumptions.
Specifically, heat gains utilized rated cable ampacity. The rated cable ampacities are the maximum heating values for control and power cables and are not realistic values-for establishing a lower temperature limit. The use of rated cable ampacities is conservative for establishing upper temperature limits, but it is not conservative for establishing lower temperature limits.
Furthermore, heat gains incorrectly included values from non-1E lighting and from sump pumps, which are not anticipated to be operating. Therefore, this calculation does not adequately demonstrate that freezing temperatures will not occur.in the associated ERCW rocms (Refer to Deficiency 05.4-1).
Also, as a result of IDI team findings in both References 1 and 3, both these calculations are examples of inadequate design verification.
Basis 10 CFR 50, Appendix A, Criteria 4 " Environmental and Missile Design Basis",
states equipment shall be compatible with environmental conditions. Also, F W DRAFT _ _ _ _
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3.11, Environmental Design of Electrical.and Mechanical Equipment states that safety-related mechanical and electrical equipment is capable of functioning properly in the worst-case local environments. Contrary to these comitments j
TVA did not ensure that the purchased equipment is compatible with the local j
environmental temperatures.
3 In addition, ANSI N45.2.11 - 1974, Section 6 requries that design verification o
be performed. Contrary to this an inadequate checking was performed for the l
l associated calculations.
j Extent These types of nonconservative assumptions may be used in other environmental calculations. Therefore, the technical basis of other environmental calcula-tions for mild environment should be reviewed.
j References 1.
Calculation, " Summary of Mild Environmental Conditions for Sequoyah and Watts Bar Nuclear Plant", Rev. 6, 9/5/86, pages 45 and 46, 845 860905 235.
2.
Drawing #47E235-34 " Environmental Data, Environment - Hild, EL 705',
i Rev. 2, 1/8/86.
3.
Calculation, "ERCW Pump Station HVAC Heat Transmission and Heating",
Rev. O, 8/26/87, 844 870826 012.
4.
Memorandum, from J. B. Hosmer to L. M. Nobles, 8/26/87, B25 870826 019.
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1 D2.3-3 (DEFICIENCY)
REVISION 3 October 7, 1987 ERCW SCREENWASH PUMP NOT INCLUDED IN TVA ASME XI PUMP INSERVICE TEST PLAN Backoround The TVA Pump Inservice Testing Program is in accordance with ASME Section XI, 1977 Edition through Summer 1978, Addenda, as identified in FSAR Section 6.8.
Appendix 6.8A, Table A in the FSAR lists each pump that TVA identifies as required for testing in accordance with ASME XI, IWP-1100.
Description The ERCW Screenwash Pumps have not been included in the TVA ASME XI Pump Inservice Test Plan and are not considered in its scope by TVA's Inservice Inspection (ISI) Section. However, the screenwash pumps are safety-related.
TVA Class C (per FSAR 3.2) and perform a safety-related function during normal plant operation. This function is to ensure cleanliness of the traveling water screen, which ensures the required suction pressure to the ERCW pump. Further.
the traveling water screen (TVA Class C) is electrically interlocked with the screenwash pump, such that it can not operate unless the screenwash pump discharge header pressure is at the pressure switch setting. Therefore, the screenwash pumps should be included in the ASME XI Pump Inservice Test Plan.
Basis The ASME Section XI Inservice Pump Test Plan is not in compliance with FSAR Section 3.2.2.3 since, the screenwash pumps are not included in that program.
Impact on Design The noncompliance with ASME XI Inservice Pump Test requirements (IWP) for screenwash pumps could lead to loss of function of traveling water screens are thus loss of ERCW pumps. Adequate instrumentation (i.e., flow, pressure, vibration, etc.) must be provided so that testing to meet ASME XI requirements can be demonstrated.
Extent This appears to be an isolated misclassification in the ERCW system.
However.
Other systems should be reviewed.
References 1.
FSAR 6.8, Appendix 6.8A, Table A.
2.
FSAR Table 3.2.1-2, " Summary of Criteria - Mechanical Systems Components
- ani! Table 3.2.2-2, "Sumary of Codes and Standards for Components of tre Sequoyah Nuclear Plant for Procurement after April 2,1973".
DRAFT
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FSAR Section 3.2.2.3,L" Class.C".'
3 '.. ASME Section XI 1977 EditionLthrough. Suniner 1978 Addenda.;
4 S.
Reg.~ Guide.1.26,Rev.1.(for. guidance)..
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DRAFT
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i D2.5-2 (DEFICIENCY).
REVISION 2 October 7. 1987 KEROTEST' pACKLESS Y-PATTERN: VALVES USED FOR THROTTLING
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Background===
During the field walkdown by the.IDI' inspection team. it was noted-that KER0 TEST valves were.being utilized for throttling flow through various ERCW.
heat exchangers. When asked if Kerotest valves were suitable for throttling '
- application,7TVA. checked with the valve manufacturer whoistated that Kerotest Packless Y-Pattern Globe valves are not recommended for. throttling but did
.y provide guidance that permits throttling in certain regions of percent open' versus flowrate. At the inspection team's request, TVA. reviewed the ERCW.
system for all valves that were throttled per. SI-682,.to determine if any such-Kerotest valves were usedlfor throttling. As a' result of this review, TVA discovered that 4 valvet were the Kerotest valve type to which throttling restrictions wereLapplicable. However, two of the. valves were' completely open
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and below maximum permitted flowrate. The other two valves were in the.
acceptable region of percent open versus flowrate.
Description TVA~ needs to restrict operation of the Kerotest Packless Y-Pattern Globe Valves, via procedure, to exclude' operation'in the throttling range defined by
'l the valve manufacturer. TVA should also verify that the valves do not exhibit disc noise or vibration as recommended by the valve manufacturer.
Basis 10 C.F.R. 50 Appendix B Criterion III Design Control states, " measures shall be established for the selections and review-for suitability of application of materials parts, equipment, and processes that are essential to the safety-related functions of the systems, structures'and components." Also, FSAR'Section 17.lA, " Quality Assurance During Design'and Construction" states that, quality assurance provides adequate confidence that a component will perform satisfactorily in service. Contrary to these requirements. TVA was'not aware of the valve manufacturer's recommendations restricting the use of Kerotest packless Y-pattern valves.
Impect on Design Damage to Kerotest valves and downstream piping could result from valve cavitation.
Flow restriction to safety-related components could occur in the event of a failure of valve internals. Valve cavitation could also cause excessive thinning of downstream piping.
DRAFT _
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Kerotest valves used elsewhere in the plant may be operating in the manufac-q turers unacceptable range. Use of these' valves, plant wide, needs to be I
assessed.
j References j
1.
Surveillance Instruction SI-682, ERCW Flow Balance Valve Position Verification, Rev. 18, 6/19/87.-
2.
Kerotest Graph, "2--inch Y-Globe Value, Dis Pos. % Open vs. GPM, Throttle Test at Pitt", 8/23/77.
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i 02.5-3' (DEFICIENCY)
REVISION 4 October 7, 1987 ENVIRONMENTAL QUALIFICATION (MILD) ERCW PUMP HOUSE COMPONENTS Background-
'In order to ensure safety-related equipment functionality, the ambient.
conditions to which.the equipment will.be subjected must be consistent with the-
. vendor's environmental qualification. Environmental qualification data were reviewed for:the ERCW' pumps,. strainers' screen wash pumps and traveling ' screens.:
This data was obtained from the original purchase specifications.(augmented by a
contractor data):and compared to the drawings entitled, " Environmental Data, Environment - Mild."
.ly Description j
H The temperature ranges stipulated.on the " Environmental Data, Environment -
1 Mild" drawings are 10*F to 120*F. (Normal) and 10*F to 130* (Abnormal) _ for the 720. foot elevation-(El 720') of the.ERCW ' pumping station.
For EL 705', the environmental temperatures are 65 to 104 F-(Normal) and 40 to 110*F (abnormal).
Contrary to the above, the specifications for the subject equipment have the following temperature, range specified for environmental qualification:
ERCW pumps
-20FJ to 108'F El 720',
)
ERCW strainers 60F to 104 F EL 705' Screen Wash Pumps
-20F to 104*F El 720' Traveling Screens 10*F to'120 F EL 720' Basis 10 C.F.R. 50 Appendix A - Criterion 4 " Environmental and Missile Design Basis" states that " structures, systems and components important to safety shall be designed to accommodate the effects.of and to be compatible with the environmental conditions associated with normal operation, maintenance, testino and postulated accidents".-. Additionally, FSAR Table 3.11.2-2 entitled,
" Environmental Design Criteria for ESFfSystem Equipment" states that ERCW components are " unaffected by the types of environmental conditions' calculated to be present". Contrary to these comitments,' no-justification.was-provided by TVA which reconciles the differences between the vendor data and the design drawings for mild enviornmental'-qualification of..various components.
Impact on Design Since the ambient environmental conditions exceed the vendors' environmental qualifications, the equipment may not be functional during al1 operating conditions.
DRAFT
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Extent It appears that TVA has not reviewed in detail, equipment environmental qualification, for equipment located in a mild environment. Of the four i
specifications reviewed, none were in compliance with the environmental data.
Therefore, TVA'should address the generic impact of this issue.
References i
1.
Environmental Data - Environment Hild EL-720, #47E235-38, Rev. 2,1/2/86.
2.
Environmental Data - Environment Mild EL-705', #47E235-34, Rev. 2,1/2/86, 3.
Traveling Water Screens, Specification No. 2635,9/15/75.
4.
ERCW Screen Wash Pumps,-Specification No. 2653, 6/28/76.
5.
ERCW Strainers, Specification No. 2654, 1/13/76.
6.
ERCW Pumps. Specification No. 2261, 8/21/74.
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.02.5-4 (DEFICIENCY)
REVISION 3 0ctober 6, 1987 INADEQUATE SUBSTANTIATION OF PROCEDURE FOR ERCW SCREENWASH PUMP MANUAL.' OPERATION
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Background-
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The'ERCW screenwash pumps are designed to be operated' automatically when traveling water screen differential. pressure is high The duration of the j
screenwash cycle.is controlled by an automatic timer. This logic provides. '
assurance for cleanliness of traveling water screens and thus assurance of operability of.ERCW pumps through provision of_ adequate suction pressure. The
' automatic operation.of the screenwash pumps was disabled pending replacement of i
~the traveling screen differential. pressure instrumentation.with a bubbler type.-
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Description e
The modification of traveling water screen differentia 1' pressure instruments--
tion to the_" bubbler type" has not been incorporated. A temporary change, TACF 1-82-258-67 removed the existing wiring and logic for-the differential pressure
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and timer, with a recommendation for procedurally controlled. manual operation of the screenwash pumps. TVA does not have an. approved procedure for manual operation of the screenwash pumps since the.temporacy chinge was made on.
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October 7, 1982.
Basis FSAR Section 17.1A.5, " Instructions, Procedures and Drawings" requires that j
activities affecting quality are prescribed by documented instructions in the i
form of drawings, specifications and procedures.
Contrary to this requirement.
TVA had no approved procedure.in place for. manual operation of the screenwash pumps as required by an implemented temporary change notice'5 years ago.
Impact on Design Without screenwash, the traveling screens could become blocked. This could affect ERCW pump operation by decreasing the available suction head to the pumps. Also, if the difference between the water level across.the screens exceeds several feet, the screens may collapse.
Extent As a result of operating procedures not being issued in a timely manner,
" temporary" changes utilized in lieu of design changes may' disable system-safety functions without providing compensatory action.
DRAFT O
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I References 1.
10 CFR 50 Appendix B ' Criterion V, " Instructions Procedures and i
. Drawings".
2.
FSAR 9.2.2.5.5, ERCW " Level Instrumentation".
3.
ECN L-5758, " Replace Diaphragm Type Level Differential System with a Bubbler Type for the Traveling. Screen Work Control",12/8/82.
4.
TACF 1-82-258-67, " Removal of Traveling Screen Differential Pressure and i
Timer Instrumentation",10/7/82.
5.
501-67.1, " System Operating Instruction - ERCW", Rev. 29, S/15/87.
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< REVISION 4
.0ctober 6, 1987 ERCW COLD THERMAL MODE'
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Background
Section 3.1.1, Normal ~ Functions, of.the~ERCW' design criteria (Reference 1) states in part, that "ERCW will be supplied to the various heat exchangers with the inlet water temperature between 35 degrees F and 83 degrees F."
Description The TVA' operating modes table (Reference '2) does not specify the 35 degree F cold thermal mode for the-ERCW pipe from header 2B to containment spray heat exchanger 2B. Therefore, the TVA piping analysis (Reference 3) does not analyze the associated piping and supports.for the cold themal mode or thermal range. However, the analysis did consider the containment. spray heat exchanger nozzle themal displacement due to the cold thermals mode and themal range.
Basis Section 3.1.1 of the FSAR specifies that'the range of temperatures for ERCW supply to the various heat exchangers will be between 35 degrees F and 83
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degrees F.
Criterion 4 of 10 CFR 50, Appendix A requires that systems l
important to safety shall be designed to accommodate the effects of and -to be compatible with the environmental conditions. associated with normal operation.
The basis for this deficiency is TVA's lack of a cold thermal mode and stress range calculation for the ERCW pipe (a safety system) from header 28. to the containment spray heat exchanger, as required by the ERCW design criteria.
Impact on Design TVA should review the ERCW operating modes tables to confim inclusion.of the-35 degree cold thermal mode. TVA should review the ERCW piping analyses to confirm that.the cold themal mode is considered. TVA should revise piping analysis N2-67-2A to include cold themal mode and thermal range effects for R
the ERCW pipe from header 2B to containment spray. heat exchanger 28.
l Extent The extent of this deficiency is not known at this time.
References 1.
TVA Design Criteria No. SQN-DC-Y-7.4, Essential Raw Cooling Water System (67),Rev.2,7/11/86(RIMSNo.B05860721505).
2.
TVA Drawing No. 47B466-67-22, Insulation and Operation Mode Analysis Data.
ERCy Supply Header 28, Rev. 1, 11/22/85.
3.
TVAPipingAnalysis.N2-67-2A,Rev.7,1/17/87(RIMSNo.B25870123814).-
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- 1 D3.2-7 (DEFICIENCY)
REVISION.2 September'30, 1987
'j ERCW PIPING SPOOL PIECES
Background
A spool piece has to be inserted into the ERCW line from header 2B to the component cooling water surge tank to supply emergency ERCW water to the tank.
during maximum flood mode. There are also several other spool pieces in the ERCW system which are to be available for rapid installation under certain
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flood conditions.
Description
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On September 17, 1986, TVA. prepared Drawing Deviation 86DD952 to document discrepancies-between the 3-inch diameter 1-foot long-. spool pieces fabricated for the ERCW piping running from heacer 2B.to the CCS. surge tank' for Units' 1 and 2 and the as-built dimensions of the installed piping. The 3-inch diameter blind flanged lines for Unit I had horizontal.and vertical misalignments of i
2-inchesLand 2-7/16-inches respectively, and a flange to flange distance of i
11-1/8-inches.
The corresponding Unit 2 piping had horizontal and vertical misalignments of approximately 3/4-inch and 1/8-inch, and a flange to flange distance of 11-3/4-inches. TVA has initiated Work Request No.~.B129163 to fabricate temporary (non-CSSC) spool pieces until pennanent spool' pieces are
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fabricated.
It appears that some of the spool pieces were fabricated to the i
nominal dimensions specified on the piping physical drawings rather than to the as-built dimensions of the installed piping.
Their use may induce substantial loading into adjacent-supports when installed since the piping needs-to be cold sprung to get the flanges to mate.
' Basis Criterion III of 10 CFR 50 Appendix B. " Design Control," states in part that-measures shall be established for the control of design interfaces and for verifying or checking the adequacy of design.
Contrary to this commitment, TVA did not verify that the spool pieces. installed in the ERCW system were fabricated to the as-built dimensions of the installed piping.
Impact on Design System accident-mode function could be impaired due to mismatch between spool pieces and piping.
Extent The extent of this condition is not known at this time.
References N/A DRAFT 3
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D3.2-8 (DEFICIENCY)
REVISION 3 October 6, 1987 VALVE OPERATOR FUNDAMENTAL FREQUENCIES
Background
FSAR Section 3.9.2.5.2 indicates that safety-related valves at SQN were procured to reflect a minimum extended structure fundamental frequency of 25 i
Hz. Valve extended structures with frequencies less than 25 Hz were modeled as flexible cantilevers in the piping analysis.
i Description
)
The Acton test report (Reference 1) which qualifies 3-inch valve 2-LCV-70-63 documents valve operators with resonance frequencies of 13 and 16 Hz. However, j
the valve extended structure was not modeled as a flexible structure in the j
computer analysis performed for alternate analysis calculation N2-70-39A.
I Basis CEB did not implement the commitment detailed in FSAR Section 3.9.2.5.2 to j
model flexible valve operators.
j Impact on Design Consideration of valve operator flexibility may increase pipe stresses and j
support reactions, as well as valve operator accelerations.
Extent The extent of this deficiency in both rigorously analyzed and alternately analyzed safety class piping subsystems is not known at this time.
References 1.
Acton Test Report No. 13253-2, " Seismic Vibration Testing of One 2" Control Valve, Model 38-20721, S/N N-137-2 for Masoneilan, Purchase Order No. 78288", 5/12/77.
DRAFT D3.3-1 (DEFICIENCY)
REVISION 4 October 6, 1987
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'ERCW SYSTEM PIPE SUPPORT CALCULATIONS N2-67-2A
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l Background-
)
i The piping stress ' analyst. prepares stress _ isometrics for. each safety-related piping subsystem which detail the types, orientations and locations of the.
l associated pipe supports. The pipe support calculation for each support. design 3
uses the output loads from the piping stress analysis to confirm the adequacy q
of the support design, a
Description Pipe supports IERCWH-71 and 1ERCWH-134..are both supported Dy.a 7'-9" span of W10X25 structural steel.
However, the pipe support calculation for 1ERCWH-71 which evaluates the~W10X25 beam does not include the support loads due to pipe support IERCWH-134 (References 1, 2).
Pipe support 1ERCWH-133 was modeled in the piping analysis as an axial restraint. The pipe support detail sheet shows a lateral clearance of 1/16" between the pipe and a vertical piece of 4"X4" tubing, which does not appear to be adequate to accomodate the 0.17" lateral motion specified in the pipe support calculation (Reference 3).
Basis
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FSAR Table 3.9.2-3 specifies B31.1 as.the piping design code of record for TVA safety Class B, C and D piping and supports.
Section.1G.~1 of ANSI B31.1 notes, in part, that the design of elements for supporting or restraining-piping systems, or components thereof, shall be based on all the concurrently acting loads transmitted into the supporting elements.
Impact on Design Pipe supports 1ERCWH-71 and -134 are not adequately qualified. --Pipe support IERCWH-133 may require modification to provide adequate lateral clearance.
Extent
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The extent of these conditions is not known at this time.
References-1.
TVA Pipe Support Calculation IERCWH-71, Rev. 902, 1/30/86..
2.
TVA Pipe Support Calculation IERCWH-134, Rev. 902, 2/25/86.
3.
TVA Pipe Support Calculation 1ERCWH-133, Rev. 903, 2/24/86.
DRAFT i
D3.3-2 (DEFICIENCY)
REVISION 4.
October 6, 1987 PIPE SUPPORT DISCREPANCIES g
Backgrobnd
.l Ten pipe support analyses for piping problems N2-67-10R and N2-67-11R were reviewed for compliance to CEB design criteria and FSAR commitments.
Description Three pipe support analyses associated with piping problems N2-67-10R and N2-67-11R were reviewed and found to have errors relating to assumptions.and dimensional data. One analysis was found to have used an unconservative-assumption during the development of the structural model. Two other supports were modeled with incorrect dimensions.
Pipe support analysis 47A450-25-344, Revision 2, dated 7/28/81, used a i
dimension-of 17 inches between support points at nodes 1 and 4 that attach to a
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surface mounted baseplate. Two separate dimensions.on the support drawing
'l indicated that this dimension was 11 inches,.while the difference between the elevation listings on the tube steel indicated that the dimension was 17 inches.
It appears that the analyst used the larger number based on the assumption that use of the 17 inch dimension was conservative.
In reality, the
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11 inch dimension between supports on the braced cantilever structure will i
result in higher bolt loads and structural loads. CEB concurred with this conclusion.
Pipe support analyses 47A450-25-348 and 349 Revision 2, dated 7/28/81, used incorrect dimensions for the vertical location of each support on a common 6" X 6" X 3/16" structural' tube. Pipe support 47A450-25-349 was modeled as connecting to the connon beam at a location of 26 inches above the baseplate, while the correct dimension-was determined to be 29 3/4. inches.
Similarly, pipe support 47A450-25-348 was modeled as connecting to the connon beam at a location of 44 inches above the baseplate, while the correct dimension was determined to be 46-1/2 inches. Since this common beam is essentially a cantilever beam, the location of the loads at a' greater distance above the baseplate will result in higher anchor bolt loads and structural loads. CEB concurred with the larger dimensions.
Basis Criterion III of 10 CFR 50, Appendix B, " Design Control," states in part that measures shall be established for the control of design interfaces and verify-t ing or checking the adequacy of design.- Criterion VI of 10 CFR 50, Appendix 8,
" Document Control." states in'part that adequate control shall be maintained for safety-related drawings. This requirement is also reiterated in the Sequoyah,FSAR paragraphs 17.2.5 and 17.2.6 (Amendment #20).
Contrary to these connitments, a piping support detail drawing specified inconsistent dimensions.
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and two pipe support calculations used incorrect dimensions to qualify the respective pipe supports.
' Impact on Design The unconservative assumption used in pipe support calculation 47A450-25-344 and the incorrect dimensions used in_ pipe support calculations 47A450-25-348 and 349 will result in higher anchor bolt loads and structural loads. These effects must be evaluated in order to determine their safety significance. The 1
team notes that CEB had scheduled all three pipe supports for regeneration to CEB design criteria SQN-DC-V-24.2 prior to Unit 2 restart.
Extent Since errors were detected in three out of ten support analyses evaluated, this deficiency suggests generic implications.
References 1.
Pipe Support Calculations 47A450-25-340, 348 and 349, 2.
FSAR Section 3.9.2, Safety Class B, C'and D Fluid Components.
3.
CEB Design Criteria 'w'B-DC-40-31.9, Rev. O, 8/29/75. -
4.
CEB Design criteria SQN-DC-V-24.1.
5.
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.c D3.3-3 (DEFICIENCY)
REVISION 3 October 6, 1987 INCORRECT PIPE SUPPORT ALLOWABLE STRESSES
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Background
Pipe supports for safety-related systems are analyzed using loads from the TPIPE piping analysis computer code.
Support loads are developed for various.
loading conditions and load combinations and are associated with specific systemserviceconditions(i.e., Normal, Upset, Faulted). TPIPE lists the pipe support loads as normalized loads, which are then used by the pipe support designer to analyze the pipe supports.
l Description Pipe support nonnalized loads derived from the TPIPE computer analysis are based upon the largest value arrived at when each specified load or load combination is divided by the appropriate factor associated with the allowable stress for each service condition.
This normalized load is then used in the pipe support analysis and the resulting stress is then compared to the Normal i
condition allowable stresses.
For example, in a Faulted condition the primary plus secondary support load would be divided by a factor of 1.6 (Sequoyah Pipe Support Design Criteria SQN-DC-V-24.1).
This resultant nonnalized load would then be applied in the pipe support analysis and the weak-axis bending stress for any support members.
It would be compared with 0.75 F F = material yieldstress).
Thefaultedfactorof1.6,therefore, allo %s(aXallowable stress for the Faulted condition of 0.75 x 1.6 = 1.2 F Both the' pipe support A-1)gn criteria (reiterated by the design manual in nole.
desi (a)ofTable7.17.4 and the FSAR require a Faulted allowable stress limit of 0.9 F.
y Basis Seedoyah FSAR Section 3.8.4.5.2 and Table 3.8.4-2 limit the allowable stress in steal structures for Faulted load cases (SSE) to 0.9 F This connitment is alsareiteratedintheSequoyahPipeSupportDesignMaXu.al, Volume 3, Section 7.17. Table A-1.
Contrary to this, pipe support allowable stresses can exceed 0.9 F for some types of stresses in linear supports (i.e., weak-axis bending)..
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Impact on Design l
Since the pipe support allowable stresses are accounted for in the TPIPE computer analysis by means of reduction factors, this item impacts the design of all safety-related pipe supports and must be evaluated.
Extent This item has generic implications and applies to all safety-related pipe supports, designed for rigorously analyzed piping.
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References 1.
Sequoyah FSAR Section 3.8.4.5.2
" Structural Steel" and Table 3.8.4-2, 1
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" Auxiliary Control: Building Structural Steel Loads, Loading Conditions,
.j and Allowable Stresses".
1 2.
Sequoyah Pipe Support Design Manual, Volume 3. Section 7.17. Table A-1, j
Rev. O, 4/22/83.
3.
US NRC Standard Review Plan, Section 3.8.3
" Concrete and Steel Internal Structures of Steel or Concrete Containments".
4.
Design Criteria SQN-DC-V-24.1, " Location and Design of Piping Supports and Supplemental Steel in Category I Structures", Rev. O, 6/23/86.
5.
Design Criteria WB-DC-40-31.9, " Location and Design of Piping Supports and Supplemental Steel in Category I Structures", Rev. 6,2/10/86.
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s D3.3-4 (DEFICIENCY)
REVISION 3 October'6, 1987 PULLOUT LOADINGS FOR BASEPLATE AND ANCHOR BOLTS
' Background TVA uses prequalified " typical" pipe supports where possible as part of their alternate analysis process.
These typicals have been qualified to loadings J
specified as part of the alternate analysis process, i
Description Upon reviewing the calculations (47A053-101, 102, 114) performed to verify f
these " typical" pipe support designs, it was observed that when calculating the bolt loads and baseplate thicknesses, only the moment loading was used with no consideration given to the pullout loading.
Since the pullout loading will increase both the baseplate stresses and bolt pullout loads, it should have been considered.
J Basis j
FSAR Section 17.1A.3 " Design Control" (Amendment #20) provides for the proper control and use of loadings. Contrary to these requirements, the applied i
pullout loads for the anchor bolts and baseplates were not considered.
Impact on Design The omission of a pertinent loading condition could result in undersized q
baseplates and anchor bolts.
j Extent This error may only be limited to the 47A053 series of alternately analyzed supports. TVA should verify the extent of the error.
References 1.
Pipe Support Calculations: 47A053-101, 47A053-102, 47A053-114, Rev. O, 1/7/80.
DRAFT
.y D3.3-5 (DEFICIENCY)
REVISION 3 October.6, 1987 l
INCORRECT NCR CORRECTIVE ACTION Background.
Unistrut clamps have connonly)been used at Sequoyah Nuclear Plant on small piping;(under6inchdiameter for multi-directional' loading. Unistrut c1 amp-test data were originally based-on one direction of. loading.- Designers have used the allowable load based upon the one. directional test in the appropriate direction regardless of the number of loading directions. On June 30, 1982 an NCR was written to address this issue.
Description-During a. review of design change documentation, the IDI. team determined that nonconfomance report SQNSWP8213, dated 6/30/82, n;as written-to address the use:
of Unistrut clamps. qualified for one directional loading but used in situations-with multiple loading directions acting simultaneously. The NCR concluded that an interaction equation should be used when these clamps were subjected to multi-directional loading. The TVA corrective action included an investigation and
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review of the clamp loadings consisting of testing of the-clamps for multi-directional loading. The testing accounted for separately applied clamp loadings in each.of two orthogonal.in-plane clamp loading directions, as well
-l as loading parallel to the pipe axis.
In order to address the interaction effects of the three loading directions, the resultant test individual maximum allowable loads were compared with the previously established maximum load applied for each loading direction. These resultant ratios were then evaluated by one of two methods as outlined in NCR evaluation calculation WBN SWP 8237, i
dated 1/25/83, with Revision 1 dated 7/6/84. A first attempt was made to interact the ratios using a straight line fomulation that algebraically adds the ratios and accounts for contributions.from two bending stresses and axial stress on the Unistrut clamp.
If the sum of these three ratios is less than or equel to 1.0, then the particular clamp (based on pipe diameter) is considered qualified to the multi-directional. test loads.
Compliance with this interaction equation would indicate that previous use of the particular clamp size was acceptable with no further evaluation required.. This straight line interaction equation demonstrates the clamp body capacity for the interaction of tension and bending on the clamp.
For those pipe clamp sizes, however, where the straight line interaction equation failed to demonstrate the clamp's i
qualification for prior usage, CEB used an ellipical interaction equation.
This type of interaction has historically been associated with the combination of tension and shear stresses (i.e., bolting), and is not readily applicable or justifiable for the interaction of tension and bending. CEB stated that the basis for using the elliptical interaction equation when the clamp failed the straight line. interaction equation was that the test data indicated bolt failure as the predominant test load failure. The IDI: team agreed that the'use of the elliptical interaction equation was suitable for bolt shear and tension interaction, but questioned CEB's method of combining tension and shear. The
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DRAFT _ _ _ - _ _ _ _ _ _
R sum of the squares of the ratios associated with each loading direction was added algebraically and compared with unity. This method does not account for i
the fact that two of the loading) directions provide-tension in the bolts and their contribution (i.e., ratios should be sumed before squaring, and then added to the square of the ratio associated with the bolt shear loading direction.
In addition, this interaction only qualifies the clamp assembly for bolt tension and shear and does not address the clamp body. Currently, the
. calculations show that the straight.line interaction equation for the evaluation of clamp bending and tension-does not provide adequate qualification for pipe. clamp sizes 1-1/2", 2", 2-1/2", 3" and 4" diameters. The clamp tension and bending interaction'needs to be addressed independently of bolt tension and shear interaction for these pipe clamp diameters in order to
'l qualify previous Unistrut pipe clamp usage.
In addition, the use of the straight line interaction equation may not be suitable to properly evaluate the contribution due to friction loading on the clamp.
This is discussed in further detail in related deficiency D3.5-3.
Basis FSAR section 17.2 refers to TVA Topical Report - TVA-TR75-1A, Section 17.2.
Section 17.2.16, " Adverse Conditions and Corrective Actions" of that report 1
stipulates that procedures shall provide for the identification and correction of adverse conditions.
Contrary to this, the corrective action for NCR SQNSWP8213, dated 6/30/82, did not properly consider the interaction of tension-and shear on the clamp bolts and tension and bending on the body clamp ~.
Impact on Design The incorrect application of interaction equations for tension and shear on the bolts and tension and bending on the clamp body may be significant for the evaluation of multi-directional loading on installed Unistrut clamps.
Extent Since this corrective action addresses the generic issue of Unistrut pipe clamp allowable loads for multi-directional loading, the failure to properly address interaction has generic implications.
References 1.
NCR SQNSWP8213, 6/30/82.
2.
ENDES Calculation, NCR Evaluation WBNSWP8237, 1/25/83 with Rev. 1, 7/6/84 3.
Salmon and Johnson, " Steel Structures:
Design and Behavior", 2nd Edition.
1980.
4.
ENDES Calculation, "Unistrut Pipe Clamp Load Ratings", 7/27/82.
DRAFT - _ _ _ -
d D3.4-1 (DEFICIENCY)
REVISION 4 October 6, 1987 MOTOR OPERATED VALVE DESIGN PRESSURE
Background
The system operating pressure for the four motor operated valves that regulate the flow of emergency makeup water from ERCW headers 2A and 2B to turbine-driven auxiliary feedwater pump 2A is'150 psig (Reference 1). The motor operated valves are TVA safety class C valves.
Description The valve vendor drawing (Reference 2) specifies a design pressure of 50 psig.
the vendor seismic qualification calculation (Reference 3) also specifies a i
design pressure of 50 psig, and uses 50-psig to compute the valve stem thrust and operational loads. This'is inconsistent with the specified system operating pressure of 150 psig.
Basis
)
FSAR Table 3.2.1-2 requires that TVA class C valves installed in the ERCW system in the auxiliary building be qualified to the requirements of ASME Section III.
FSAR Table-3.2.1-2 also requires that TVA class C ERCW valves installed in the auxiliary building be seismically qualified by test. Contrary to this comitment the valve seismic qualification is not adequate since they were qualified to the wrong operating pressure.
Impact on Design The valves must be requalified to the correct system operating pressure.
No hardware impact is anticipated.
Extent 1
\\
l The MOVs for turbine driven auxiliary feedwater pump 1A might have been I
similarly procured and qualified, j
q References l
1.
TVA Drawing 47B466-3-11, Insulation and Operation Mode Analysis Data,
{
Revision 0, 2.
Walworth Drawing A-7516-M-145D, TVA Revision 901, 8/3/86.
l 3.
Walworth Calculation, Seismic Acceleration Calculations,-12/3/73.
l DRAFT _ _ _ _ _ _ _
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i D3.4-2
'(DEFICIENCY)-
REVISION.5 October 6, 1987 SEISMIC QUALIFICATION OF-TURBINE-DRIVEN AUXILIARY FEEDWATER PUMP 2A
Background
TVA flow diagram 47K427-57 (Reference 1), specifies a design pressure of 1650 psig at the discharge nozzle of, turbine-driven auxiliary feedwater pump 2A ~. The pump is TVA safety class C.
The equipment nozzle loads that TVA supplies to' equipment vendors (Reference 2).
to use in the seismic qualification of: equipment requires consideration of.
nozzle axial load due to internally induced pressure in addition to the' forces and moments the piping imposes on the nozzle.
Description-The vendor's seismic qualification calculation for the turbine driven-auxiliary-1 feedwater pump (Reference 3) does not consider the' axial thrust at the pump 1
, i discharge nozzle due to the 1650 psig design pressure.
Basis i
FSAR Table 3.9.2-3 requires that TVA safety class C pumps be evaluated for the Normal, Upset and Faulted loading conditions in accordance with the ASME Section III, Class III, Draft Code for Pumps and Valves.
Sheet 7 of FSAR Table 3.2.1-2 also stipulates that the turbine driven pumps be evaluated in accordance with the Draft ASME' Code for' Pumps and Valves for Nuclear Power, 1
Class III, dated 1968, and March 1970 Addenda.
Subsection ND-3000 of the ASME Code requires that stresses resulting from pressure or other applied loads shall be included with stresses resulting from earthquake effects in the desiv 4
of pumps.
Impact on Design The anchor bolts which secure the pump to the. concrete slab must be reevaluated considering the axial thrust loads previously neglected.
Extent This may affect both turbine and motor driven pumps at Sequoyah.
j References 1.
TVA Drawing 47K427-57,.Rev. 9, 1/28/87.
2.
CEB' Report 82-1, Evaluation of Piping Analysis Loads-on Nozzles for Flocr Mounted Equipment, Rev. 1. July, 1984.
4 3.
Mcdonald Report No. ME-161, Seismic-Stress Analysis of Auxiliary Feedwate-Pumps, October, 1974.
i DRAFT
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03.4-3 (DEFICIENCY)
REVISION 5 October 6, 1987 CCW HEAT EXCHANGER CALCULATION
Background
During the'IDI team walkdown the component cooling water (CCW) heat exchanger 1
was observed to be supported by three supports. However, when the equipment
{
vendor perfonned the seismic and nozzle load calculations for the heat
]
exchanger shell only two supports were considered.
Description The CCW heat exchangers were furnished by Industrial Process Engineers (IPE) as part of TVA Contract 71C33-92691. These heat exchangers were' built to the requirements of Specification 1339 including Appendix E (seismic requirements).
-The calculations of record were performed using.only two supports. The J
manufacturer's drawings indicate only two supports.t TVA has been unable-to.
provide documentation such as drawings, calculations or correspondence with the manufacturer which substantiates the design of the additional support.
Basis FSAR Section 17, " Quality Assurance" references Topical Report'TVA-TR75-1.
Contrary to the requirements of TR-75-1, Section 17.1A.6, " Document Control".
TVA failed to control or keep records of design changes on safety-related i
equipment.
Impact on Design The addition of another support would. generally be considered a conservative modification which would tend to reduce the loasing of other supports.
However, the location and design of this support should take into consideration j
p(ossible unfavorable conditions relating to the equipment being supportedi.e., str "i
additional induced thermal stresses, etc.). Also, any modifications to the manufacturer's~ safety-related equipment should be-approved by the manufacturer.
Extent All CCW heat exchangers it Units 1 and 2 at Sequoyah are affecte6 References 1.
TVA Bid Package 171C 33-92691, 4/30/71.
I 2.
TVA Specification 1339,'" Component Cooling Water Heat Exchangers for l
Sequoyah Nuclear Plant, Units 1~and 2.
3.
Appendix E. " Guide for Qualification of Seismic Class I and II Mechanical and Electrical. Equipment", 2/11/71.
3 DRAFT 34 -
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- 4. -
IPE Nozzle Design Calcul'ations for-PO #71P 33-92691. 11/19/71.
f 5.
IPE Seismic Design Calculations.for PO #71P 33-92691, 1/1/72.
.)
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D3.4-4 (DEFICIENCY).
REVISION 4 October 6, 1987J CCW AND CS HEAT EXCHANGER N0ZZLE LOADING'
Background
Each of the component cooling water and containment spray heat exchangers has-several nozzles which the equipment vendor' analyzed to certain loading values:
The equipment and their supports have been-qualified to resist these loadings.
The pipe stress analyst must compare the actual nozzle loading to these allowable loadings to assure that the equipment is not overloaded.,
Description The CCW and CS heattexchangers were furnished.by Industrial Process Engineers (IPE) as part of TVA Contracts 71033-92691'and 71C33-92645.
IPE performed a-nozzle load analysis using the following loading criteria:
2' 0.1 x A x Sy Where A = Pipe area - in F
=
A 0.01 x Z x Sy S = Material: yield stress at ambi M
=
2 Y
ent temperature - lbs/in M9 0.01 x Z x Sy
=
3 Z = Pipe section modulus - in F = Axial' load - lbs,
= Pipe moment load - in-lbs M = Pipe torsion load - in-lbs The TVA criteria for nozzle allowable loading given in report number CEB 82-1 q
are listed below:
0.1 x Z x S Where F
= Forces in x, y, z F
=
0.0707xZlS x,y,z directions - lbs M*y'Y'Z
=
Mb,,z 0.01 x Z x S Y
Mby, z = Bending moment in y and t
y z directions - in-lbs-S =Materialyieldstressgt'ambi Y
ent temperature-lbs/in These allowable loadings differ from the loads used in the IPE calculations primarily because IPE did not consider the shear loadings in the two horizontal planes.
IPE only considered axial nozzle loadings.
Therefore, the CCW and CS heat exchanger nozzle loadings may have been compared to allowables'which were higher than those for which they were qualified.
Basis FSAR section 17. " Quality Assurance" references Topical-Report TVA-TR-75-1.
Contrary to the requirements of.TR-75-1 Section.17.1A.3.2, " Interface Control".'
which requires that TVA ENDES is responsible for controlling external interfaces with a vendor / contractor and assuring that proper submittals are received from the contractor. TVA failed to provide proper interface control' DRAFT
.o 1
to assure that the vendor loads were not exceeded when using internal design k
standards.
Impact on Design The additional nozzle loadings allowed by CEB 82-1 must be evaluated. The effect of these additional loadings could impact the design margins for the heat exchanger equipment supports and the civil foundations which TVA designed, in part, to the vendor's nozzle loadings.
Extent q
All equipment with vendor nozzle design loadings less than the CEB 82-1 loadings.
References 1.
TVA Bid Package #71C 33-92691, 4/30/71.
2.
TVA Bid Package #71C 33-92645, 12/28/70.
3.
IPE Nozzle Design Calculations for P0 #71P33-92691,11/9/71.
4.
IPE Nozzle Design Calculations for P0 #71P33-92645, 8/24/71.
5.
" Evaluation of Piping Analysis Loads on Nozzles for Floor Mounted 1
Equipment" Report No. CEB 82-1, Rev. 1, 7/6/84.
I DRAFT
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D3.4-5 (DEFICIENCY)
REVISION 3 October 6, 1987, VENDOR - SUPPLIED FLEXIBLE HOSE
.I
Background
i
.TVA utilizes'6-inch diameter flexible metal hose, FLEX 0NICS series 401M or-I equal, for TVA Class _C application in the ERCW system.
Since this material is I
used in.a safety-related system which is designed to resist seismic conditions, I
it is necessary that the material and-its configuration be seismically
. qualified.
' ~
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l Description i
l l
Section 5 of Appendix F to the procurement document (Reference 1) requires that "the documentation for each equipment type shall demonstrate that the equipment meets the specified seismic design requirements in both perfomance and structural integrity". However, Item No. 5'of Form SQN-QAP-III-2.1-A appended s
to the procurement document specifically exempts the vendor from providing the seismic valysis/ test information which fomed the basis for the vendor's technic, data sheet.
In addition, the nomal seismic requirements specified in Appendix F for.
seismic Category I components are preempted by the less stringent seismic requirement which is specified for the vendor under schedule XIV of the procurement document.
Basis Criterion IV of 10 CFR 50 Appendix B. " Procurement Document Control," requires in part that measures be established to assure that applicable design bases be included in procurement documents for equipment and material. Contrary to th'-
coninitment, TVA's procurement document for the flex hose did:not require the vendor to verify that the flex hose met the seismic design criteria specified for seismic Category I structures in Appendix F, or the Faulted load combination specified in FSAR Table 3.9.2-2.
Impact on Design
)
1 Documentation should be provided to demonstrate the flex hoses are seismicali, j
i qualified, j
Extent The extent of this condition is not known at this time.
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References q
1.
-TVA Appendix F', " Design Criteria for Qualification 'of Seismic Class l' and
.)
Setsmic Class ~II Mechanical and Electrical Equipment", Revision 1, 1
2/10/72.
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DRAFT.
l D3.4-6 (DEFICIENCY)
REVISION 3 October 6, 1987 ERCW UPPER CONTAINMENT VENT COOLER FREQUENCY CALCULATION l
Background
For equipment requiring seismic qualification, the equipment vendor was I
required to determine the equipment natural frequency in order to establish whether the equipment is rigid or flexible as required by the TVA' purchase-document.
j Description During the IDI team review of the seismic qualification for.the upper containment vent cooler, an error was discovered in the calculation-of the natural frequency of the. cooler structure.
This calculation was documented in
{
a report prepared for RG Products Inc. entitled, " Seismic Qualification of Air Cooling Unit Model HRV-192," dated October 23. 1972.
In the report, an error was ma'de in the calculation of the stiffness ~ of the motor foundation and.
casing, which resulted in incorrect natural frequencies. The original 1
calculation shows the natural frequency of 42.2 Hz for the first mode'and 84.5 Hz for the second mode. The corrected calculation indicates a natural frequency of 23.7 Hz and 80.5 Hz for the first'and second modes respectively.
This error raises doubt concerning the rigidity of the equipment:(equipment is f
considered flexible when the natural frequency is under 25 Hz) as stipulated in i
Appendix F of the purchase specification.
4 Basis Criterion III of 10 CFR 50, Appendix B, " Design Control," requires in part that.
measures be established for the identification and control'of design interfaces and for coordination among participating design organizations to include.the establishment of procedures for the review, approval, release, distribution, and revision.of documents involving design interfaces.' Contrary to.these commitments, TVA did not ensure the design review of the seismic qualification report for the upper containment vent cooler, which contained a calculation i
error which when corrected, resulted in a natural frequency for the equipment below 25 Hz.
Impact on Design Since the calculation error does not result in a natural frequency substantially-below the limit of 25 Hz, the design impact is minimal.
Extent Although the calculation error appears to be an isolated occurrence, the team has identified TVA's failure to ensure adequate design verification for equipment seismic qualification reports as a generic deficiency.
DRAFT _
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1 References
)
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1.
Seismic Qualification Report'of the Air Cooling Unit Model HRV-192, l
10/23/72.
{
2.
Purchase Specification TVA 72C35-92775, 2/18/72.
q 3.
TVA Drawing 010-142 for' Contract 72-92775, Rev. B, April, 1972.
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i D3.4-7 (DEFICIENCY)-
REVISION 3 October'6, 1987-CHILLER UNIT SEISMIC QUALIFICATION
Background
v Equipment must be seismically qualified to perform its safety. function. A I
i chiller unit designated as TVA Class ~C (safety-class) and its control box were
.to be seismically qualified.-
Description, l
l The seismic analysis of the chiller unit (Reference.1) yields a maximum j
I acceleration of 0.629 at the. support location of the control box. The seismic test of the control box.(Reference 2) was' conducted at 0.5g.and 0.6g levels.
1 These g levels are greater than the design horizontal:zero period acceleration j
(IPA)-acceleration of 0.339:for the auxiliary building at' elevation 734 feet.
The control box would therefore be; adequately qualified if mounted directly on.
the Lfloor. However,'since the control box is mounted on the chiller unit, the-provisions of Section 6.1.3 of the technical' specification attached'to the -
procurement document govern.
Section 6.1.3 requires that no device location on' 1
the support structure be permitted to have an acceleration greater than three-fourths of the floor design ~ response. spectrum, or three-forths of the actual device test acceleration.
-l
'I Basis l
The btsis for this deficiency is TVA's failure to implement a technical provision in the procurement document.. FSAR Section 17.1A.1.6, Item 10'
( Amendment #14), notes, in part, that TVA reviews and/or approves manufac-turer's qualifications -and test reports to ensure conformance with procurement l
requirements.
Impact on Design The chiller unit control box may not function as required. during 'a ' seismic v
event.
Extent s
The extent of this condition is not known at this time.
However, it may have generic implications regarding the seismic qualification of all panel or rack
. mounted equipment.
i m
References 1.
National Loss Control Corporation Report " Seismic' Response Analysis of an IPCX 230-0Q Chiller Unit for Dunham - Bush Incorporated",f 7/30/76.
2.
Aa N Nav Laboratories Test Report.No. ETL 5-6099,'12/29/75._
DRAFT -
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U3.5-1 (UNRESOLVEDITEM)
REVISION 5 October 6, 1987 PIPING CODE OF RECORD
Background
The piping code of record for design as stated in the FSAR is ANSI B31.1 - 1967 edition.
Since ANSI B31.1 - 1967 did not define combinations for the Normal, Upset and Faulted conditions, TVA used the stress allowable equations from ASME Section III, Subsection NC-3000, Winter 1972 Addendum for these plant conditions.
Description In addition to the stress equations, TVA used the stress allowable limits specified in the ASME Code.
This usage is documented in CEB's Rigorous Piping Analysis Handbook.
Potential Basis CEB's use of the ASME Code stress allowable limits is not consistent with FSAR Table 3.9.2-3, which commits to the use of ANSI B31.1 - 1967 stress allowable limits.
Impact on Design The stress allowable limits documented in ANSI B31.1 - 1967 and stipulated in FSAR Table 3.9.2-3 are generally more conservative for stainless steel materials than the ASME Section III allowable stresses.
Piping designs using stainless steel materials may, therefore, not meet the design limits specified in the Sequoyah FSAR.
Extent Stainless steel piping systems analyzed using ASME allowable stresses would be primarily impacted.
References
~
1.
ANSI B31.1 - 1967 Edition.
2.
Section 3.9 of Sequoyah FSAR, " Mechanical Systems and Components".
3.
ASME III, Subsection NC-3000 Winter 1972 Addendum.
4.
CEB Design Criteria 13.3, Rev. 3, 8/13/84 and 13.3.1, Rev. 2, 12/27/83.
l DRAFT i
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f 03.5-2 (DEFICIENCY)
REVISION 4 October 6, 1987 1
USE OF SELECTED B31.1 CODE RULES
.i
Background
The FSAR specifies ANSI B31.1 1967 as the piping l design. code of record for i
Sequoyah Nuclear Plant.
However, the FSAR also allows the use of ASME i
Section III, Subsection NC-3000, Winter 1972 Addendum for the Normal. Upset and Faulted stress equations. This combination has been used to qualify an overstressed piping member by using selected portions of both codes.
t Description The sumary of analysis report N2-67-2A for the ERCW system contains a l
l calculation-foran'overstressedcondition(pg.83).
This calculation uses the interpretation that for ANSI B31.1 - 1967-the seismic portion;of the' additive stresses did not require the use of the stress intensification factor (1) to increase-the stresses at a tee connection. The IDI team believes that the' B31.1 code requires the use'of the i factor for.both weight and seismic stress-calculations The team notes that the use.of a portion of a code just because it helps an analysis to document a tolerable condition is an unacceptable.
practice. This concern is directly related to unresolved item U3.5-1, which deals with the code of record for piping.
Basis Pcragraph 3.9.2.5.2 of the Sequoyah FSAR states in part that ANSI B31.11-1967 is the piping analysis code of record, but'that the operating condition stress
-l equations in ASME Section III, Subsection NC-3000, Winter 1972 Addendum may be used. However, TVA's use of the difference in~ the two codes to help qualify an analysis indicates that TVA believes that there is a significant difference in the B31.1 and NC 3000 requirements, which is contrary ~ to tha FSAR assertion that the B31.1 and ASME codes in question are equivalent.
i Impact on Design Use of this interpretation regarding the application of the stress intensifi-cation factor to qualify piping systems may result in'a design that b overstressed and does not meet.FSAR comitments.
- Enent, This interpretation may have been used on other systems. However, it is not part of any design guidelines and may only.have been used by. selected pipe stress analysis personnel.
DRAFT
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' References 1.-
Sumary of Analysis Report N2-67-29,'1/17/87.
2..
ANSI B31.1
.1967.
1Property "ANSI code" (as page type) with input value "ANSI B31.1</br></br>.1967.</br></br>1" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process. 3.
Section 3.9 of FSAR, " Mech'anical Systems and Components".
4.
ASME Section III NC-3000 Winter Addendum, 1972.
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1 D3.5 (DEFICIENCY)
REVISION 3
~0ctober 6, 1987-UNISTRUT CLAMP LOAD TESTING
Background
TVA uses Unistrut clamps for the restraint of small bore Seismic Category I l
piping. These clamps are used as part of_the restraints contained in the TVA.
3
" typical"' package. These clamps can be.used in the design of restraints for l
one, two or three force. directions. TVA chose to develop load ratings via the testing option of ASME Section III.- Subsection NF,1974 Edition, Winter 1976 1
i Addendum, rather than to the test load criteria specified in ANSI.B31.1 - 1967, the design code of record for piping at Sequoyah Nuclear Plant.
Description TVA's use of the load rating rules of ASME Section III, Subsection NF instead of the 831.1-1967 code of record is under current licensing review by the NRC's OSP. TVA calculation SMG-3060 established design loads (rated loads) based on the Unistrut load ratings detennined by test. The calculation contained the following errors and omissions:
j 1.
The calculation did not apply the 10% reduction in the load rating as required by NF when only one sample is tested.
2.
The clamp relies upon friction to resist loading along the pipe axis.
However, the A-307 bolting material cannot be used for friction connections according to AISC and NF.
3.
The load ratings do not consider temperature effects.
j 4.
The 3-inch clamp failed at approximately one-half the load rating of the i'
2-1/2-inch and 3-1/2-inch clamps without any justification or explanation for the failure. This anomaly should have been explained since'it may ropresent a manufacturing defect which could affect all clamp sizes, j
i 5.
The load rating test report did not discuss the method of test I
installation.
This consideration is important since these clamps can be j
installed in. the field either tight (to restrain axial movement) or loose j
(to allow axial movement). The resulting test load rating for each j
configuration could therefore differ.
.i 6.
In Revision 2 of the test report, TVA uses a square root of the sum of the squares formula to determine the interaction of loadings.
This formula is l
inappropriate for this application (see Deficiency.D3.3-5).
Furthermore, J
even a linear interaction equation may not be appropriate.
For example, I
the combination of tensior and axial. loading will result in a reduction of
]
the, normal force between tne Unistrut channel and the piping, thus reducing the resistance to axial loading.
DRAFT j l
I,c.'
I Basis FSAR Section 17.2 refers to the TVA Topical Report on Quality Assurance. TVA Topical Report.- TVA-TR75-1A in Section 17.2.11 " Test Control" requires in
.part that modifications, repairs and replacements shall be tested in accordance with the original design and testing requirements or acceptable alternatives.
The original testing requirements of Section 17.1.A.11 of that report required that the test program implemented during design, verify the adequacy of equipment design and fabrication.
Contrary to these requirements the tests
~
performed to establish load ratings for the Unistrut clamps did not follow established procedures and used material not suited for the service intended.
Impact on Design The deficiency can affect all designs using Unistrut clamps, especially designs with axial loading where only friction is relied upon to-resist the applied loading.
Extent All designs using Unistrut clamps.
References 1.
Calculation Number WBP-84-0801-037 R3, " Strut Pipe Strap Load Ratings" 8/23/84.
2.
Calculation Number 841-87-0707-001 R2, "Unistrut P-2558 Test Evaluation" 7/7/87.
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H D3.6 (DEFICIENCY).
REVISION 3 October'6, 1987 DESIGN REVIEW FOR ERCW EQUIPMENT ~
l
Background
1 The' ERCW cooling units supplied for the upper and lower containment ventilation j
-and the control rod drive systems are required to be seismically qualified by4 1]
their respective purchase specifications.
Description i
Seismic qualification reports were reviewed for the upper containment vent cooler, lower containment vent cooler and-the control rod drive vent cooler.
I The TVA purchase documents.for, each of these items were identified as 72C35-92775 for the upper containment, vent cooler and.72C33-92730 for both the lower containment vent cooler and' the control rod drive vent' cooler.
Both of 1
these procurement documents specify Appendix.A of the TVA Quality Assurance
. Item 2 of Appendix A states that' design computations shall be-Program.
independently. reviewed and certified to assure-compliance with all' requirements. The IDI review team could find no evidence of.a design review
~
for the equipment qualification' reports' identified above.
In a related finding (Deficiency D3.4-6), an error in the calculation of the natural' frequencies-of the equipment was discovered. The team notes that in the seismic qualification I
report that equipment had also.not received a design review.
Basis TVA Quality Assurance Program, Appendix A, attached to the equipment. procure-ment documents and Section'17.1A.3.9 of TVA Topical Report -' TVA-TR75-1A, referred to in Section 17.1 of the'FSAR require a design review by the vendor or by TVA for the equipment qualification reports identified.
Contrary'to thi' j
a design review was not performed for the Upper Containment Vent Cooler, the Lower Containment Vent Cooler and the Control Rod Dnive Vent Cooler.
P Impact on Design The IDI team reviewed the seismic qualification reports for four cooling units inside containment. Three of these units did not-have reports that were design reviewed, one of which contained incorrect natural frequency calculations (DeficiencyD3.4-6).
Based.on the above, this item as potential safety h
significance.
Extent This item has potential generic implications for vendor equipment seismic qualification reports which qualify installed safety-related equipment at Sequoyah, Nuclear. Plant.-
DRAFT :
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'j References 1.
Upper containment vent cooler seismic qualification report, 10/23/72.
~
2.
Lower containment cooler seismic qualification reports, 2/11/71',9/1/72 and 12/5/75.
1
'3.
.TVA procurement document 72C35-92775, 5/13/72.
/
4.
TVA procurement document 72C33-92730, 6/8/82.
5.
TVA procurement document 68C60-91934.
6.
TVA procurement document 73x-822718, 2/5/78.
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D4.2-1 (DEFICIENCY)
REVISION 6 October 7, 1987
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STABILITY OF ERCW ACCESS CELLS B_arkground The ERCW access cells are seismic Category I structures. The ERCW piping and conduit are routed through the access cells in passing from the ERCW pumping station to the other safety-related buildings.
Failure of the access cells would totally disable the ERCW system.
Boundary and behavior conditions assumed in the analysis and design process should be reflected in the actual as-built structure and its capability to behave as predicted. The
~
appropriateness of this part of the design should be verified as part of the design control process for seismic Category I structures.
Description The-seismic analysis report for ERCW access cells indicates that the six i
cylindrical sheet pile cells and interconnecting cells were assumed to act as a single unit. The assembly was analyzed as a "J-shaped" unit. Additionally,
~
the design postulates shrinkage of the interior cell concrete and gaps between interior concrete and the exterior sheet piles. The calculations also J
postulate vertical movement between the adjacent cells.
Beams have been i
designed and provided to tie the cells together in the horizontal direction but i
not vertically.
Foam was embedded above and below the beam which precludes j
vertical load transfer without significant beam deflections. The inability to i
transfer the vertical shear would make the original assumptions of unitized I
behavior invalid. Also, since the assembly of cells is a "J-shaped" structure and therefore not symmetrical, torsional loads should be considered in the l
analysis and design. Therefore, the assumptions used in performing the seismic analysis of the ERCW access cells may not be valid since they do not represent the as-built design.
The original seismic analysis (Ref. 1) evaluated the case where six ERCW access cells act independently and concluded that the factor of safety for overturning was less than one, i.e., during a seismic event the access cells would potentially overturn. There are other conservatism that have not been considered in the seismic analysis, e.g., geometrical interlocking of the access cells and -friction between individual cells, that when considered, may show that the structure is stable during a seismic event. However, the team could not conclude that the access cells would withstand a seismic event basec i
on the TVA analysis.
j Basis i
FSAR Section 3.7.2.2, in describing the methodology of the seismic analyses conducted for Category I structures, states that the structures were modeled and thatethe inertial properties of the models were characterized by the mass.
eccentricity and mass moment of inertia of each mass point so as to accurately j
model the structure.
DRAFT
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.Section 3.8.4.4.'7.of the FSAR, in describing the design and analysis procedures-
'of the ERCW access cells, states, "the structures were investigated to ensure-continuity of design. The structures were also investigated for stability against overturning,.floati,ng and sliding."
j GDC-2-states that safety-related structures must withstand natural events', such as ' earthquakes, floods, tornados, etc.
Contrary to the above commitments and regulations, TVA has not adequately demonstrated that the design of the ERCW access cells will withstand an earthquake.
l In this case the basic assumptions made in computing the moment of inertia of a section were not. consistent with ERCW access cell design so that the model does 4
I not accurately represent the structure.
)
l Impact'on Design 1
l The computations state that cells acting as individual cells may be. unstable.
J The ERCW pipes are supported on these cells and the failure or excess movement of these cells may cause the failure of the ERCW pipes and the electrical e
conduits embedded in the cells, s
j Extent i
l The extent of this specific problem is limited to the ERCW access cells.
]
l However -this raises a more fundamental issue concerning adequacy of design verification which should be examined by TVA.-
j j
References 1.
SQN Calculation, " Concrete Cells Providing Access to ERCW Pumping Station", Rev. O,1/18/78, CEB 800605 024.
i 2.
SQN Dynamic Earthquake Analysis of ERCW Pumping Station and Adjacent L
Cells, Report No. CEB-74-14, Rev. 2, 1/3/79, CEB 800603 005.
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.D4.2-2 (DEFICIENCY)
REVISION 5 October 6,1987
' CALCULATIONS WERE NOT CHECKED OR VERIFIE,0
Background
Design calculations for seismic Category I structures should be checked or.
Verified to ensure ~that the-input, methodology and computation are correct and
'l the results are reasonable.
Description The seismic analysis calculations (Ref.1, 2, 3, 8) contained pages that are neither signed by the originator nor-by the checker. A statement known as the
" technical position" was attached to the calculations and cited the corresponding seismic reports.(Ref.~4, 5, 6, 9) which summarized the.results of-the analyses that were signed by the preparer and checker. The statement 1
attested to the fact th.at the signer of the-reports did-indeed check the l
results of the computer analyses to verify the_ technical adequacy. However, the statement provides an insufficient basis for meeting the quality assurance procedure TVA had in-place.
1 Basis TVA. quality assurance procedure (Ref. 7) required that every design calculation i
be signed by the preparer and checker.
Each sheet of the design calculation should contain the name of the project and the general subj-Lect title followec 1
by any necessary subtitles. Contrary to the requirements of Ref. 7, the referenced calculations (Ref. 1,-2 '3, 8) contained unsigned and untitled pages. The referenced reports (Ref. 4, 5. 6, 9) are the by-products of tre original analyses. The original seismic analyses could he in error since *
.i have not been properly checked.. This is a violation of TVA quality assura~ ~
requirements and 10 CFR 50, Appendix B, Criterion III, " Design Control."
Impact on Design i
This finding may lead to the invalidation of the adequacy of the seismic analyses and any design based upon the results'of those analyses.. The se analysis of the steel containment vessel (Ref. 2) was based on the previc',s revision (Rev. 0 dated 11/6/86) which was not checked.
Extent All calculations containing the technical position statement should be re, to ensure that the basis of the position is founded upon properly reviewe.:
calculations rather than the seismic report.
DRAFT _
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. References.
1.
-SeismicLAnalysis.of Shield Buildings, Rev. 3, 4/11/87, 841 870416 003.
2.
Seismic Analysis of Steel Containment Vessel, Rev. 1, 3/9/87, B41 870310 007.
3.
Auxiliary Control Building. Seismic. Analysis, Rev. 3, 3/2/87, 841.
y 870303 006.
.j
.4.
SQN Dynamic. Earthquake Analysis and Static Wind-Tornado Analysis of the Shield Building, CEB 80-22-C, Rev. 1, 8/26/85, B41.850826 0G8.
5.
SQN Dynamic Earthquake Analysis of the' Steel Containment Vessel,-
CEB-75-03-CR1, 6/22/87, B41 870622 001..
6.
SQN Dynamic Earthquake Analysis of the Auxiliary-Control. Building, Rev. 1, 1/30/74, CEB 800603 002.
i 7.
Quality Assurance Procedure, SQN-QAP-III-1.3, " Preparation, Review and 1
Records of Design Calculations", Rev. 1,8/27/73.
8.
Reactor Building Seismic Analysis of ti:e Interior Concrete Structure, Rev.'7, 2/17/87, B41 870217 009.
9.
SQN Dynamic Earthquake Analysis of the Interior Concrete Structure and -
Response Spectra for Attached Equipment, CEB-80-23-C, Rev. 1,' 3/15/74,'CEB l
800603 004.
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. REVISION 3 0ctober 6,<1987' VERTICAL RESPONSE SPECTRA 0F THE STEEL CONTAINMENT VESSEL.
q
Background
Response spectra are generated'at elevations ~where these curves are needed for the seismic qualification of other systems, components, or structures.
For a
_i given. seismic Category I_ structure and input, the spectralgenerated by various 1
computer codes should be'nearly-the same at a given elevation-unless the
.l seismic model or the input to the analysis ~-(earthquake) has been changed.
)
Description The seismic. analysis of the steel containment vessel, a Category:I structure,
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(Ref.1).was performed in 1972 and the seismic report (Ref. 2) was subsequently H
issued on:1/24/75.
In the winter of 1985, the vertical response spectra were regenerated and the calculation was considered to be Rev. 1 of the-seismic analysis of the steel containment vessel. Revision 1 of the seismic report was-issued on 6/22/87. The regenerated. vertical! spectra curves show a significant
]
increase in acceleration within the 20 to 30 Hz range at every elevation.- At the higher containment elevations, vertical accelerations increased seven tc
- )
ten-fold. The reasons for the increase are unknown. This. leaves:in question the technical adequacy of the original methodology.used by-TVA to perform i
seismic' analyses.
It.also leaves in question the adequacy'of' equipment seismic i
qualification.which have apparently been based 'on the' earlier floor' response spectra.
q Basis 1
i TVA has not met the requirements of 10 CFR 50,' Appendix B, Criterion III, Design Control, which requires that calculations'be reviewed and checked or f
verified. Where computer codes are used they should be verified and documented.
In addition, TVA has not adequately demonstrated that Category I systems or components can adequately resist natural' phenomena, such as earthquakes, as required by 10 CFR 50, Appendix A, General. Design Criterion 2.
\\
Impact-on Design I
All seismic Category I systems, components and structures could be affected.due
'l to the fact that the seismic response spectra 'fonn the basis for seismic design or qualification.
Extent This,may have generic implications, particularly if the cause of the discrepancy in the vertical response spectra is attributed to the methodology used-in perfonning the analysis.
L DRAFT :i o______.____________.____
I l
References l
1.
Seismic Analysis of the Steel Containment Vessel, Rev. O, 11/6/86,841 i
861106 004; Rev. 1, 3/9/87, 841 870310 007.
2.
SNP Dynamic Earthquake Analysis of the Steel Containment Vessel, CEB-70-03-C, Rev. 1, 6/22/87, 841 870622 001.
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DRAFT 55 -
l D4.2-4 (DEFICIENCY)
REVISION 4 October 6, 1987 l
AUXILIARY BUILDING SEISMIC MODEL DOES NOT REPRESENT THE AS-BUILT.
Background
The seismic model used in the analysis of seismic Category I structures should be representative of the actual structure.
When structures are modified, the.
seismic model should be reassessed to assure that the model still represents the structure.
Description The seismic analysis for the auxiliary and control buildings, supported by a common base slab (Ref. 1), was performed in 1970.
The original model did not
' consider the concrete columns as part of the model. There were also certain modifica,tions as shown in the concrete drawings (Ref. 2) which together with the omitted concrete columns, could affect the results of the original analysis. No justification was available to confinn that the seismic analysis of the auxiliary and control buildings is adequate.
Basis FSAR Section 3.7.2.2 in describing the methodology of the seismic analysis of Category I structures states that the mathematical model used for the analysis accurately models the structure. The purpose of the analysis is to determine the response of the structure during a seismic event.
Since the model does not represent the actual structure, the analysis may not meet the TVA commitment.
In addition, TVA has not adequately demonstrated that systems and components in the auxiliary and control buildings will withstand a seismic event as required by 10 CFR 50, Appendix A, General Design Criterion 2.
Impact on Design The outcome of this deficiency could invalidate the current seismic analysis results for the auxiliary control building.
Extent All seismic Categ'ory I structures should be reassessed against the seismic model to ensure that the structural design changes do not affect the seismic analysis results.
References 1.
Auxiliary-ControlBuildingSeismicA6alysis,Rev.3,3/2/87, B41 780303 006.
2.
TVA, Drawings 41N700-1, Rev. 6, 2/4/93; 41N700-2, Rev. 6, 6/18/84; 41N700-3,Rev.7,8/24/83;41N7008,Rev.5,4/28/78;41N700-5,Rev.5, DRAFT _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ -
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4/28/78; 41N700-6, Rev. 5, 4/28/78; 41N700-7, Rev. 4, 4/28/78; 41N700-8, l
Rev. 5, 6/1/83.
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.I D4.3-1
'(DEFICIENCY)
REVISION 5 October 6, 1987 AUXILIARY BUILDING BASE SLAB DESIGN
Background
The base slab of the auxiliary building, a seismic Category I structure, is anchored to.the rock to minimize the bending stresses in the slab due to hydrostatic uplift pressure. ' Anchor rods made from different lengths of #11 reinforcing bars were used for this anchorage. The lengths of the anchor rods embedded in the rock and the'bar spacing are determined by the magnitude of.the s
net uplift force that acts on the base slab.
Description In order to determine the net uplift force acting o. the slab, TVA deducted the n
full unifonn soil pressure, equivalent to the whole building weight, from the j
hydrostatic uplift. force (Reference 1). However, only the weight of'the base j
and fill slab should have been deducted to determine the net uplift force, j
Since the base slab is relatively thin, the dead-loads transmitted by the walls
)
react primarily on the portions of the slab that are direct.ly.beneath the walls.
l I
Basis Section 3.8.4.3 of the FSAR and Table 3.8.4-1 provide information.on the concrete structural loads, load combinations and allowable stresses for i
concrete associated with the auxiliary building. The dead loads are defined as 1
the contributing stresses. The uncorrected net uplift pressure was used in'the
{
design of the anchor rods and the base slab reinforcement since the i
superstructure dead loads are not unifonnly distributed to the base slab. Thu 1s a violation of 10 CFR 50, Appendix B, Criterion III." Design Control."
Impact on Design The length and spacing of the anchor rods and the. amount of reinforcing.in the base slab are determined from the net uplift pressure acting on the base slab.
Therefore, the use of a lower uplift pressure may yield an unconservative-design.. This, in turn, might cause the base slab to develop cracks as. a resu;t of increased slab flexure if the postulated external flood level reached plant grade of~705 feet.
Extent Anchor rods are used in various seismic Category I structures at Sequoyah Nuclear Plant, including the control'and reactor buildings.
Similar errors n e.
exist in the design of the base mats of these buildings.
DRAFT
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5 References 1.
TVA Calculat1on',' Auxiliary Building, E1. 6'9.00'and Below, Rev. 2. 1/1/87,--
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6 B25 870101 300.
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DRAFT c
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LU4.3 (UNRESOLVEDITEM)-
REVISION 4 October 6, 1987 J
DEVELOPMENT LENGTH OF BASE SLAB ANCHOR RODL l
Background
>j i
The base slab-of the auxiliary building,-a seismic Category I structure, is anchored to the foundation rock by grouted #11 reinforcing. bars-to minimize the bending stresses.in the base slab due to hydrostatic uplift pressure. TVA used straight reinforcing bars with a 90 hook and an-extension to anchor the slab to the rock.
1 m
Description
.TVA' drawing 41N300-1 (Reference 1) shows that the #11 anchor rods are anchored in the' base slab with a 90' hook with'a straight extension. The same drawinp y
shows that the distance from the bottom of the base slab to the top of the hook is 21".
Such a detail cannot develop the full strength of the #11 reinforcing-bar since bar length extended beyond the bend is not fully-effective. The^ full strength of these bars was apparently used to detennine the spacing and length 1
l of the anchor rods.
l Potential Basis 1
f Sequoyah FSAR Section 3.8.4.3.2 states that ACI.318-63 (Reference 2) code was
]
used for the design of the auxiliary building. ACI.318-63 Code, Section 919, l
Anchorage of Web Reinforcements, states that-a standard hook can be considered develop'only-50% of the allowable stress in the bar. ACI-318-63 Code, Section 918, Anchorage Requirements - General, Subsection (h)' states in part that a standard hook in tension may be considered as developing 10,000 psi for working stress design. The' remaining stress in the bar has to be developed by; the bond between the concrete slab and the anchor rods along part-of their 21-inches vertical embedment length.
Part of the 21-inches embedment length shown on drawing 41N300-1 is not sufficient to develop the' remaining stress carried by the #11 reinforcing bar since the extension.beyond the standard hoia cannot be utilized to carry load.
Impact on Design The spacing and 'the length of the anchor rods were apparently determined by using the full strength of the #11 reinforcing bars. Therefore, the use.of a higher allowable than the bar could develop would yield an unconservative design of the base slab to resist the uplift' forces generated by the 1
hydrostatic pressure.-
Extent Anchor rods are used in various seismic Category I structures at'Sequoyah Nuclear Plant including the control and reactor buildings.
Similar designs n -
exist in the base mats of these buildings.
L DRAFT L
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- References i
l'.
TVA Drawing 41N300-1, Auxiliary Building Units 1 and 2 Concrete Anchor Rods, Rev. 2,12/16/70..
j 2.
ACI 318-63, Builo:ng Code Requirements for Reinforced Concrete, June 1963.
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DRAFT
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U4.3-3 (UNRESOLVEDITEM).
' REVISION 3-October 6, 1987 NEGATIVE MOMENT REINFORCEMENT IN BASE SLABS'AND WALLS
Background
The auxiliary building base slab, part'of a seismic; Category I structure, is placed on rock'and anchored with' rock anchors. The slab is reinforced on the top face to resist the moments induced by hydrostatic uplift. pressures.. The.
j design-of the base slab assumes that the slab'is continuously supported at each j
rock anchor.
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In addition, certain walls in the' auxiliary building below elevation 669 are.
1 placed against the rock, with reinforcing only on the inside face.' These walls are designed to. resist the horizontal hydrostatic pressures. The design of these walls assumes that the walls are fixed.at the bottom and top.
Description-TVA has not provided any reinforcement to compensate for the' negative moments developed, both in the base slab and the walls placed against the rock.
Since the designs assumed fixity at the rock anchors for the base-slab'and at the 1
tops and bottoms of the walls TVA should have provided' negative moment reinforcement (Ref. 1).
Potential Basis Sequoyah FSAR Section 3.8.4.4.1 states that the auxiliary building concrete structure is designed in accordance with the~ACI 318-63 building code-(Ref..I g
ACI 318-63 code states in Section 904, Frame Analysis - General.-that all members of frames or continuous construction shall be designed at all sectier-
]
to take the maximum effects of loads. These loads create negative moments which must be resisted.
Impact on Desian Since there is no negative reinforcement provided at the fixed or continuous
. supports, the refined concrete in the base slab and walls poured against rock could crack and. increase the positive moments. Cracking might also affect t N bond strength be. tween the concrete and the #11 anchor rods since any' cracking i
of the base slab would most likely occur at the locations where the #11'anc N r 1
rods are embedded in the base slab.
Both of these effects might lead to an unconservative design.
Extent This item is generic to most walls below grade, as well as the control builo" base slab.
It might involve other areas at Sequoyah Nuclear Plant.
DRAFT 1,
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References 1.
TVA Calculation, Elevation 669.0 and Below, Auxiliary Building Rev. 2,
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1/1/87, B25 870101'300.
j 2.-
ACI 318-63. Building. Code Requirements for Reinforced Concrete, June 1963.
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l 04.'3-5
'(DEFICIENCY)
REVISION 3
.0ctober 6, 1987-SHEAR CALCULATIONS FOR SLABS AND WALLS
.l Background.
g 1
The auxiliary building, a seismic Category I structure, is a reinforced I
concrete structure which is primarily made up of flat slabs and shear walls.,
j These reinforced concrete elements are designed for bending, axial and shear j
forces. The shear forces in slabs can be. critical at the edges where they are supported by walls.- In a similar fashion, shear forces for walls could be high where they are restrained by.the base mat and slabs at higher elevations.
It is important to evaluate shear forces, since these can result in brittle failure of the structural member that occurs suddenly without prior. warning.
Description building roofslab'(Reference'1) and the TVA calculations for the auxiliary (Reference 2) do not show' any evaluations for walls'on column lines Al and A15 i
shear forces at the possible critical sections.
Basis Sequoyah FSAR Section 3.8.4.4.1 states that the auxiliary building concrete structure is designed in accordance with the ACI. 318-63 code -(Reference 3).
j ACI 381-63 building code sections 1201, Shear Stress, and 1207, Shear Stress in Slabs and Footings. These require that shear stresses in a reinforced concrete-member be evaluated. The team could find no evidence that shear forces had i
been evaluated.
Impact on Design The design of the slabs and walls might be unconservative if the allowable shear stress values are exceeded at the critical sections.
Extent This item might be generic, since the two referenced calculations show that no evaluations were, performed for shear stresses.
References 1.
TVA Calculation, Roof Slab Elevation 778.0,' Auxiliary Building, Rev. 2, 7/28/87, B25 870728 450, 2.
TVA Calculation, Al and A15 Line Walls, Auxiliary Building, Rev. 2, 2/4/87, B25'870204 302.
3.
ACI 318-63, Building Code Requirements for Reinforced Concrete, June 1963.
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DRAFT
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u D'4.3-6 (DEFICIENCY)
REVISION 2 October 6, 1987 MINIMUM REINFORCEMENT FOR WALLS
-Background The auxiliary building, a seismic. Category.I structure,Lis a: reinforced concrete structure, mainly composed of slabs and walls. Walls are the major vertical load carrying. structural elements.. The interior and exterior walls 'in i
the auxiliary building are designed to resist. horizontal loads such as; seismic, tornado, soil and hydrostatic pressures.. The reinforcement required-in these i
walls was determined from analysis performed for these horizontal loads.
]
Description TVA provided a minimum horizontal reinforcing. steel area of 0.0020 bd where b, is the width of the wall and d is the distance from the extreme compression
{
fiber to the centroid of tension reinforcement '(Ref. I and 2). The minimum area of reinforcing steel provided should have been 0.0025 times the' gross' area -
)
of the wall, as required t'y ACI 318-63 building code Section 2202 (f)._
Therefore, TVA provided less than 80 percent of the reinforcing' steel required by the ACI Code.
j Basis Sequoyah FSAR Section 3.8.4.4 states that auxiliary building walls contain minimum reinforcing steel percentages in the horizontal and vertical directions as specified by the TVA temperature and shrinkage standards and the ACI Code.
318-63, Section 2202 (f). The ACI code requires a--higher percentage of minimum 4
horizontal reinforcement to be provided in the walls. TVA used a lower percentage of reinforcing steel without technical justification or prior NRC/AEC approval.
Impact on Design The ACI 318-63 building code requirements have apparently not been followed.
TVA must assess the impact of the use of lower reinforcing steel percentages. ca the structural adequacy of the auxiliary building.
)
Extent
}
It appears that this design basis was used in all Category I buildings. This should be addressed generically by'TVA.
References 1.
TVA Calculation, Al and A15 Line Walls Auxiliary Building, Rev. 2, 2/4/P',
B25870204302.
2.
TVA Calculation, A5 and All Line Walls Auxiliary -Building, Rev. -3, 4/6/8',
l B25 870408 374.
DRAFT i
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.D4.3 (DEFICIENCY)
REVISION 3 October 6, 1987 i
VERTICAL SEISMIC LOAD ON AUXILIARY BUILDING ROOF TRUSS
Background
j The auxiliary building roof at elevation 791.75', part of a seismic Category I structure, is supported on structural framing made up of girders spanning l
between trusses. The major load carrying elements are the trusses, which span l
80 feet. The trusses are required to be designed for dead, live, and seismic loads.
Description In the determination of the vertical seismic loads, TVA assumed that the trusses were rigid in the vertical direction. Therefore, the ' vertical acceleration used in the design of these trusses was'the rigid vertical acceleration at this elevation in the walls. A frequency analysis should have been performed to determine if the trusses were rigid.
During the IDI team inspection, a calculation was perfomed by TVA that indicated the live load on the roof had to be reduced from the original design of 50 psf to 30 psf. This i
40 percent reduction in live load was necessary to compensate for the 4
additional loads calculated when TVA reanalyzed the roof trusses taking their flexibility into account.
Basis i
Sequoyah FSAR Table 3.8.4-2 shows that seismic accelerations should be i
considered in the structural steel design. Section 3.7.2.3.8 of the FSAR states that a vertical lumped mass dynamic analysis was performed for all seismic Category I structures to detemine the vertical loads. Constant vertical load factors were not used unless the dynamic analysis indicated that the structure behaved as a rigid body in the vertical direction. TVA used an
.l unconservative vertical seismic acceleration in the design of the auxiliary buildingrooftrusses(Ref.1).
Impact on Design If the roof trus.ses are not rigid, the vertical seismic loads will increase.
Therefore, the allowable stresses in the truss members could be exceeded.
Extent A similar approach for vertical seismic loads could have been used on other structural steel systems which could result ir Stresses that exceed allowable stress levels. This must be generically reviewed by TVA.
i DRAFT L
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-i References
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TVA-Calculation. Auxiliary Building. Structural Steel Roof: Framing Rev 2,
7/7/81 SWP 810707 041.
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DRAFT
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D4.3-8
-(OEFICIENCY)
REVISION 3 October 6,'1987 f
OVERTURNING OF TANKS LOCATED ON THE AUXILIARY BUILDING ROOF
Background
There are' four tanks that are supported at elevation 791.75' on the roof of the aux 'ary building, a seismic Category I structure. These are, the two raw
- e water tanks, the demineralized water tank, and the cask washdown sers demineralized water tank. These tanks each weigh approximately 90 kips to 135 i
kips. These tanks are nonsafety-related.
Description The TVA calculations reviewed (Ref.1) show that only the vertical seismic loads from the tanks were considered.in the design of the auxiliary building roof. The loads generated to resist overturning moments of the tanks due to horizontal earthquake were not included in determining'the size of the structural steel members supporting the tank.
Basis Secuoyah FSAR Table 3.8.4-2 shows that seismic' loads'should be considered in the structural steel design. Contrary to this requirement, TVA failed to-include the horizontal seismic loads resulting from the seismic response of the tanks in the design of the auxiliary building roof steel framing.
Impact on Design The seismic overturning moments from the tanks will increase the total load acting on the girders and the trusses.
Since such loads were not considered, the allowable stresses for these structural steel elements could be exceeded.
Extent Omission of the seismic overturning moments could have occurred for other tant supports.
References 1.
TVA Calculation, Auxiliary Building Structural Steel Roof Framing, Rev.
7/7/81, SWP 810707 041, l
DRAFT i
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l 04.4-1 (DEFICIENCY)-
REVISION 3
. October 6, 1987.
DESIGN OF ERCW PUMPHOUSE STRUCTURE ROOF
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'TO RESIST TORNADO MISSILES
]
1
4 Background
{
l Tornado missile protection 'is to be provided for. the ERCW pumphouse structure,
.1 a seismic Category I structure. The protecting structure should be designed to j
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resist the missiles shown on Table 3.5.5.4 of the FSAR. The tornado missile 4
protection is provided by a ' series of-W21 X 49 beams with the webs. inclined at an angle of 45 and spaced at 9".
]
Description The tornado missile protection calculations for the ERCW pump house structure
)
are not. complete since only the bending mode of the beams.has been considered l
.by an energy absorption. calculation. Calculations not included are the shear.
l resistance, web crippling and-inelastic flange buckling. The critical missile 7
impact angle and location that produces the maximum loads'and resistance to.
1 smaller missiles that could. penetrate the structure should also be evaluated.
In summary, the TVA analysis only demonstrates that for one specific case, that of the utility pole. striking the W21 X 49 beams mid-span, that the beams'can absorb the missile's kinetic energy without structural failure. This analysis, however, considered only the energy absorption properties of the roof' structure-without investigating whether the beams themselves.would break free of their anchorages, and become missiles that could-strike.the ERCW pumps. During the'
- 1' inspection TVA analyzed the effect of~the utility pole missile striking the roof structure near the supporting walls rather than mid-span. The: analysis showed that the beam end connections would fail in shear.' However, the roof beams would drop only several inches to a concrete ledge, Basis H
The FSAR states that the roof structure is designed to withstand the-missiles y
of spectrum B of Table 3.5.5.4; however, the calculations (Reference 1).do not, for example, include consideration of various approach angles and zones of impact of the itsted missiles or potential penetrations of the smaller missiles. TVA has not adequately demonstrated that adequate tornado missile protection'has been provided for the' safety-related ERCW pumps as required by GDC-2.
~
Impact on Design Failure of the roof structure could generate secondary missiles and these could affect the operation of the~ERCW pumps.
DRAFT 1
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Extent
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This item is an isolated case and'is limited to the ERCW pumchouse. roof.
However, it is indicative of a more fundamental problem, inadequate design:
- i verification.
3 References.
j 1.
SON Calculation, "ERCW Pumping Station Structural. Steel and Miscellaneous Steel", Rev. 2, 3/19/87, B25 870319 302.
I 2.
Drawing 38N342, Rev. 4 -11/29/76.
3.
Drawing 38N343. Rev. 3,~11/29/76.
4.
Drawing 38N344, Rev. 5, 11/29/76.
-l 5.
Drawing 38N345, Rev. 2, 11/29/76.
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'6.
Drawing 38N346, Rev. 2, 11/29/76..
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DRAFT
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- sy D4.'4-2
'(DEFICIENCY)
' REVISION 4 October 6, 1987-ANALYSIS OF PILE SUPPORTS FOR THE_ BURIED ERCW PIPELINE 4
Background 1 Design. input'information such as ' assumed material properties' which are used in the. analysis and. design process for seismic Category.I structures, should-
. reflect the conditions which exist in the_ actual structure.
Description The rock fill; dike provides; lateral l support-and stiffness for the piles supporting the ERCW pipes and. electrical' conduits. The stiffness:of the piles-is assumed to be.small.in comparison-to the.rockifill, dike.- The calculation indicates that the dike will be unstable'under certain loading. conditions when the parameter D (theLinternall friction ang"le'of the rock) is taken as.
- 35. degrees.
If the dike fails, the piles supporting the ERCW piping and the electrical conduits may also fail. Two laboratory tests were performed on the rock fill material in 1979 to determine-the actual properties of the rock fill..
The' test results indicate that the material has an interna 1' friction angle.of-45 degrees. By utilizing the new 9 value, TVA-may._be.'able to demonstrate that the rock fill dike'will be stable during a seismic. event. The current calcu-lation does not consider the vertical earthquake component which must be considered acting with the horizontal seismic loading'.
Basis FSAR'Section 2.5.6.2.3 whic.h discusses the design criteria and' analysis of the dike slopes at the ERCW pumping station, states that the seismic Category:I-slopes of the dika leading to the ERCW pumping station are designed.such that they remain stable for the most critical design condition. The TVA calculation, which assumes an angle of internal-friction ~of 35 degrees' for the rock fill dike, shows that the rock fill dikes are' unstable and may fail during-Even with the new internal a seismic event possibly disabling the ERCW system.
friction angle of 45 degrees from laboratory tests, TVA still needs to demon-strate that the dike is. stable during all loading conditions.
Impact on Design.
Failure of-the ERCW system could prevent the safe shutdown of the reactor.
Extent The: concern is limited to the section of ERCW piping supported on piles..
However, it raises a broader, generic concern regarding adequacy of design verification which must be addressed by TVA.
DRAFT -
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.i References
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1.
SON Calculation " Seismic ~ Analysis of ERCW Access. Dike and Upstream Dikes", Rev. 2,,6/19/'/9. CEB 801124-020.
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04.6-1
-(DEFICIENCY)
REVISION 4
-October 6, 1987
]
DISCREPAN,CIES BETWEEN DESIGN CALCULATIONS AND CONSTRUCTION DRAWINGS
Background
A review of various safety-related mechanical component foundations was made'to detennine whether these foerdaiions were technically adequate. The review included concrete pads, as well as embedded plates and anchors. Both the design calculations and the drawings were reviewed to determine whether the construction drawings reflected the actual design as shown in the calculations.
Description Th,e team found the following discrepancies between the design calculations and the construction drawings:
1.
Component cooling heat exchanger support calculations (Ref. 1) require an embedded plate thickness of 3/4 inch. Contrary to this, TVA drawing 48N1269 (Ref. 2) shows this plate to be only 1/2 inch.
2.
Component cooling water surge tank support calculations show that 9 anchors are required (Ref. 3), with a spacing of 3 inches between the anchors.
However, TVA drawing 48N1271 (Ref. 4) shows at most 8 anchors, with spacing varyinn between 2 and 4 inches.
3.
Containment spray heat exchanger support calculations (Ref. 5) require the thickness of the embedded plate to.be 3/4 inch with four 5/8 inch diameter anchor studs that are 6-3/8 inches long. TVA drawing 48N1267 (Ref. 6) shows the embedded plate to be 1/2 inch. Also the anchors shown are 1/2 inch diameter and 5-3/16 inches long.
Basis The embedded plate thicknesses and the concrete anchor diameters as shown on the construction drawings are less than what is required by TVA calculations.
This indicates a. lack of design control, since TVA failed to confirm the adequacy of the design documents through a controlled checking and review process. This is counter to the requirements of 10 CFR 50, Appendix B, Criterion III, Design Control.
Impact on Design The embedded plates and anchors could be underdesigned, which might result in the failure of various mechanical equipment foundations.
DRAFT - _ _ _ _ _ _ _ _ ___ -
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1 Extent LThe structural' adequacy of mechanical equipment foundations should be evaluated generically.
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References 1.
TVA Calculation, Embedded Parts E1. _714.0',' Rev. O, 8/4/86, B25 860804 301.
2.
TVA Drawing 48N1269 Miscellaneous Steel Anchor Plates and Handrailing.
E1. 714.0 - Sheet 2, Rev. 11, 5/19/77.
3.
TVA Calculation, Embedded Parts E1. 734.0',,Rev. O, 8/4/86, B25.860804 302.
4.
TVA Drawing 48N1271,' Miscellaneous Steel Anchor Plates and Handrailing E1.
' l 734.0 - Sheet.2, Rev.'8, 8/2/77.
5.
TVA Calculation, Auxiliary Building Mise.LSteel Tank and Equipment Support,.Rev
- 1, 7/23/87, B25 870723.460..
6.
TVA drawing 48N1267.-Miscellaneous Steel Anchor Plates and Handrailing E1. 690.0 - Sheet 2, Rev. 7, 3/28/77.
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DRAFT
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I D4.6-2
'(DEFICIENCY)
Revision'3 October 6, 1987 ANCHOR BOLT DESIGN FOR TANK
Background
The component cooling water surge tank, a safety-related component, is supported.on a concrete floor:in the auxiliary building at elevation 734.0.
-The. tank is anchored to the concrete slab to resist the horizontal and vertical seismic loads.
Description The review of TVA and vendor calculations for.the tank (Refs. 2 and 1) shows that the anchor bolts were sized without considering the shear forces at the bottom of the tank. Both calculations show that the anchor bolts were sized considering only the forces required to resist the overturning moments.at the bottom of the tank.
Basis
'l Sequoyah FSAR S.ection 3.8.4.2 commits to the use of the AISC Code for the design of steel structures. AISC specification Section 1.6.3 and 1.22 state that both shear and tension should be considered in the design of anchor bolts.
Impact on Design Since the shear forces were not considered in calculating the size of-the anchor bolts, they could be undersized, leading to a failure of the tank support under the maximum design basis loading conditions.
Extent l
Failure to evaluate shear in anchor bolts for mechanical equipment could be a generic problem.
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References-l 1.' -
Modern. Welding Company'Inc. Calculation, Component Cooling Water Surge Tank Design, 11'/18/71.
J 2.
TVA Calculation Squad Check No. 72-2165, Design of Component Cooling Water Surge Tank, 12/15/71.
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AISC, Manual of Steel Construction, Seventh Edition.
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i 04.6-3 (DEFICIENCY)
REVISION 3 October 6, 1987' SEISMIC ANALYSIS OF STEEL TANKS l
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Background===
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Seismic Category I steel tanks and their supports. are designed to resist the
'l seismic loads during an earthquake. Appendix C to.the TVA purchase-specifications-(Ref. 7) required that tank vendors perform an analysis of the tanks using a lumped mass model to determine the natural frequency.of.the
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tanks. Tanks;that have a natural frequency above 33 Hz are considered:to be rigid. Tanks with natural. frequencies below 33 Hz will behave as. flexible structures during a seismic event. For these tanks, additional loads will'be
. imposed on the Lsupports, as a result of the increased,acce erat on due to'the l
i tank. flexibility.
Description The review of the seismic calculations (Ref.1) for the demineralized water.
tank shows that TID 7024 (Ref. 2).was used.to determine the dynamic pressures
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on the tank. 'As stated in TID'7024, Chapter 6 the analytical method used.
4 therein is valid only if the tank filled with water acts as a rigid body.- TVA
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calculations do not show that the frequency of the demineralized water tank was checked to make sure that it was above 33 Hz.
In. specification 72C 53-92725 (Ref. 7) Appendix C, TVA specifies the guidelines to determine the frequency of the tanks. Vendor calculation.s that determine the frequency of the l
demineralized water tanks could not be located by TVA.
1 Also, the review of TVA calculations for the component cooling water surge tank i
(Ref 4) showed that TID-25021 (Ref. 5), a procedure similar to TID-7024, was used to calculate the seismic effects without considering the frequency of the tank, even though the vendor calculations ^(Ref. 3) for this tank show that the frequency of the tank was calculated. The vendor calculation, however, failed.
to account for the shear stiffness of the tank. This is unconservative since-the omission will tend to raise the calculated natural frequency of the tank.
This calculational error was not discovered by TVA.
Basis The procedures used by TVA assume that the tanks are rigid, i.e...having a natural frequency of 33.Hz or. higher. TVA failed'to determine the frequency of the tanks before using analytical methods which are only valid for a-tank fra-quency of 33 Hz or higher. The.TVA purchase specification requirements require L
a seismic analysis of the tanks. This indicates both a' lack of design contro!
l and inadequate design verification. This is contrary to the requirements of l
10 CFR 50, Appendix B, Criterion III, Design Control and the TVA Quality L
Assurance Program approved by the NRC.
DRAFT- - _
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Imoact on Design
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1 If the natural frequencies of tanks are less than 33 Hz, then the seismic loads
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would increase possibly making the current' design of the tanks and their sup-l ports unconservative. This might lead to the failure of the tanks during a i
seismic event.
Extent This item may be generic, since TVA's Design Criteria for tanks (Ref. 6)
Section 2.3 states that the natural frequency of a. tank or reservoir, when considered to be' full of fluid, shall not be less than 33 Hz. Therefore, the calculations performed for tanks should show that the tanks have natural frequencies in the rigid range.
References.
l 1.
TVA Calculation, Squad Check 72-2174 Demineralized Water Tanks 47W310-11.
-12,1/27/72.
j 2.
TID 7024, Nuclear Reactors and Earthquakes, Chapter 6, August, 1963.
j 3.
Modern Welding Company Inc. Calculation, Design Component Cooling Water j
Surge Tank Job No. 7491,11/18/71.
j 4.
TVA Calculation Squad Check No. 72-2165, Design of Component Cooling Water l
Surge Tank, 12/15/71.
l 5.
TID-25021, Summary of Current Seismic Design Practice for Nuclear Reacter l
Facilities, 9/1967.
6.
TVA Design Criteria, SON-DC-V-13.6, Seismically Qualifying Tanks and Reservoirs and Their Supports, Re'v. I', 9/23/86.
7.
TVA Purchase Specification 72C-53-92725-2, Fabricated Tanks, 9/17/71.
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. 1 D4.7-1 (DEFICIENCY)
REVISION 4 October 6, 1987 i
RAWL EXPANSION ANCHORS Background, During.a recent inspection, NRC inspectors discovered that the Rawl expansion anchors used to restrain certain pipe support surface mounted base plates at -
Sequoyah Nuclear Plant have an ultimate. capacity approximately 10 percent lower than the Phillips Red Head anchors.
It was also discovered that TVA assumed that Rawl expansion anchors had the.same allowable tension and shear as the Phillips Red Heads. The factor of safety comitted to the NRC is based on the higher Phillips Red Head ultimate capacity.
Description Pipe support H10-727, a safety-related support, specifies that Rawl expansion anchors be installed, however, the design and factors of safety for the anchors were unconservatively based on the higher ultimate capacity of the Phillips Red Head anchors.
In addition, the Rawl anchors do not meet the NRC IE Bulletin
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79-02 criteria for the long term or restart, which call'for factors of safety of 5 and 2.8 respectively, because of their lower ultimate capacity.
i Currently, the pipe support designers are not evaluating the reduced capacity of Rawl expansion. anchors even though a CAQR has been written on this same subject. The designers are also increasing the allowable. capacity of the I
anchors to take advantage of the concrete strength gain due to the aging of the concrete even though the qualification tests for the anchors were performed with 4500 psi concrete which is higher than the concrete. strength used at-Sequoyah Nuclear Plant.
In response to this concern, TVA personnel responded that Rawl anchors had not been used even through they were specified on the support drawing.
If Rawl anchors were not installed as specified, this leaves in question the adequacy of quality assurance and quality control during original construction.
Basis The design is not in conformance with IE Bulletin 79-02 and with comitments te j
the NRC for restart criteria. The comitments to the NRC were contained in a document dated August 31,1987(Ref.2).
Impact on Design Pipe supports could fail during a seismic event as a result of the lower factors of safety and lower ultimate capacity of the Rawl anchors.
Extent This structural concern applies to all supports.using, or specified to use, Rawl anchors.
If in fact Rawl anchors were not used as specifically called (cr on the drawings, TVA needs to assess the adequacy of quality assurance and S
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quality control during construction which apparently did not detect the improper parts substitution.
j References l
1.
SON Calculation, " Pipe Support H10-727",'Rev. O, 5/30/87, B25 870531 311.
2.
Letter from Gridley to NRC, Sequoyah Unit 2, Support Modifications for 1
Restart Criteria for Rigorously Analyzed Pipe, 8/31/87.
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.D5.2-3 (DEFICIENCY)
REVISION'6
. October 6, 1987 4
. INEFFECTIVE ERCW ALARMS
' Back'oround Approximately 37 alarms are provided for the ERCW system. These alarms are intended to alert the operator to failures or misoperation of the ERCW system.
For, an alarmed condition to be valid.. operator action must be required to respond to the condition; an alarm is-considered invalid if the alarm.is'-
actuated but'no operator action'is required.
Fourteer of the ERCW' alarms have been designated priority alarms _ requiring imediate attention by the operator.
Description During a July 28, 1987 walkdown of the ERCW panel'OM-27A in the main' control room, the team observed that the following alarms had been actuated but had not cleared:
4 WINDOW LECEND1 8
ERCW 480V MCC IB-B, 28-B UNDERVOLTAGE*
15 1-FS-67-61 ERCW SUPPLY HDR A LOW *'
i 22 1-FS-67-62 ERCW SUPPLY HDR B LOW
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1-PDS-67-491 E/F ERCW HDR 1A-A STR DIFF PRESS HI 9
1-PDS-67-490 E/F ERCW HDR 1B-B STR DIFF PRESS HI 3
0-PS-67-17 ERCW HDR PRESS LOW 10 0-PS-67-18 ERCW HDR PRESS LOW i
5 0-SIS-67-301 AUX ERCW TRAY SCRN A-A, MALFUNCTION 19 0-SIS-67-472 ERCW TRAY.SCRN A-A, D-A MALFUNCTION 26 0-SIS-67-976 ERCW TRAY SCRN B-B -C-B MALFUNCTION 13 12 PDS-67-10B ERCW HDR B STR DIFF PRESS HI*
27 2-FS-67-62 ERCW SUPPLY HDR B~ LOW
- Priority alam (color coded red, per Reference 1)
Discussion with -TVA indicated that most of these alarms did not appear to represent valid alarm conditions, even though the ERCW system appeared to be operating in'a normal configuration and performing as would be expected at power. Another finding -(Deficiency D5.2-10) noted that false alams had ah o :
been observed for the two high ERCW flow alarms on panel _0-M-278.
Consequently, more than one' fourth of the ERCW alarms appeared to be invalic.
The-team believes that this represents an' excessive number of nuisance or' invalid-alarms for a safety system; accordingly, no credit should be taken Nr.
operator. action initiated by an alam until an acceptable level of effectiveness is achieved for the alarm system.
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Basis FSAR Section 3.1.2.3-(Ref. 4) states that all protection systems.are designed in accordance with IEEE-279 - 1971.
IEEE-279 (Ref. 3) requires that the protection system shall be designed to provide the operator with accurate, complete, and timely information pertinent to its own status and to generating station safety. The design shall minimize the development of conditions which would cause meters, annunciators, recorders, alarms, etc., to give anomalous indications confusing to the operator.
TVA_ design criteria SQN-DC-V-7.4, Table 2 requires specific ERCW alarms to 1
initiate operator action..The design and performance of the alarm system does not assure that true alarm conditions will initiate the operator action required by the design criteria.
Impact on Design No credit can be taken for operator action in response to ERCW alarms since there is no assurance that the alarm system will be effective during plant upset conditions.
Extent The alarms for other safety-related systems on the main control board should also be assessed by TVA for this deficiency.
1 The team understands that TVA is committed to correcting alarm system deficiencies as a result of findings from their control room design review program, and expects this IDI finding to be addressed in the TVA alarm system evaluation program.
References I
1.
Mechanical Instrument Tabulation 478601-55, Rev. G, 11/26/8A (shows annunciator front view).
2.
ERCW Design Criteria SQN-DC-V-7.4, Rev. 2, 7/11/86, Section 3.5.2, 3.9.
d 3.
IEEE-2791971, " Criteria for Protection Systems for Nuclear Power Generating Stations," Section 4.20.
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FSAR 3.1.2.3, " Protection and Reactivity Control Systems".
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05.2-4 (DEFICIENCY)
REVISION 4:
October 8, 1987 INADEQUATE ELECTRICAL' ISOLATION OF NON-1E TRAVELING-SCREEN SPEED SWITCH
Background
For each traveling screen.-a speed switch is provided for use in an alarm.
circuit. The circuit actuates.an alarm when the speed is abnormal, coincident
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with the traveling screen drive being energized.
Power for both the traveling screen drive control and the alarm circuit is supplied from a shared control i
power transformer. The function of the control circuit is. safety-related and
.1 the function of the alarm circuit is nonsafety-related. The alarm circuit hardware, including the speed switch, is non-Class.1E. Qualified devices must be provided to isolate the non-class IE circuits from the Class 1E circuits.
l Description
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The team reviewed the schematic di.agram (Ref. 1) for'the traveling screen drives and determined from TVA that fuses had been added to the circuit per ECM.
L 5637 (Ref. 2) with the intent of providing the required isolation.- NCR SCN
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NEB 8126 (Ref. 3) had identified the isolation deficiency.
The team determined that no analysis had been performed by TVA that would assure that the additional isolating fuse would clear a fault before the 1
control circuit fuse'would blow. An analysis performed by TVA during the inspection (Ref 4) demonstrated that there'was not sufficient assurance that the isolating fuse could perform.its function since the control. power fuse an:
isolating fuse current-time characteristics were not properly coordinated.
team concluded that a short circuit in the non-1E speed switch could directh lead to failure of the class 1E control circuit.
l Basis FSAR Section 3.1.2.3' states that all protection systems are designed in accordance with IEEE-279-1971.
IEEE-279 requires (Reference 5) that no credible failure at the output of. an isolation device shall prevent the associated protection system channel from meeting the minimum performance requirements specified in the design bases.
Improper fuse coordination leac-to a failure of a class 1E. control circuit as a result of a failure in the non-1E speed switch circuit is a violation'of IEEE-279-1971.
It is also a violation of TVA's design criteria (Ref. 6).
Impact on Design Failure of the non-1E speed switches during a seismic or other design basis event could lead to failure of all traveling screen drives, and loss of ERU capability for both units.
DRAFT,
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Extent All Class 1E circuits using fuses as isolation devices should be reviewed by TVA to assure that sufficient margin is provided in the coordination of 4
time-current characteristics of the isolating fuse and other Class IE fuses or i
protective devices serving the circuit.
References 1.
Schematic Diagram 35W 726-1, Rev. H (Identifies ECN), 12/27/86. Schematic Diagram 35W 726-1, Rev. 1 (As-configured),3/3/87.
2.
ECN L5637, Rev. O, 5/25/82 and WP.9818 (Reclassify traveling screen motors to Class 1E; change wiring so that a speed detector switch failure will not affect operation of IE components) 3.
NCR SQN NEB 8126, 5/8/81, " Electrical Components Required for the Operation of the ERCW Traveling Screens are not Class 1E".
4.
TVA informal calculation, 9/3/87 (Fuse coordination analysis.
Includes characteristic curves for Bussman. type FRN and. Shawmut type TRM fuses).
5.
IEEE Std. 279-1971, " Criteria for Protection Systems for Nuclear Power Generating Stations", paragraph 4.7.2.
6.
TVA Design Input Memo, DIM-SQN-DC-V-12.2-9, 6/25/87,
Subject:
Separation of Electrical Equipment and Wiring Design Criteria, SQN-DC-V-12.2",
i Section 4.4.7, " Electrical Isolation".
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D5.2-5 (DEFICIENCY)
- REVISION 5 October 8, 1987l INADEQUATE SEPARATION OF REDUNDANT MAIN CONTR01. BOARD WIRING 1
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Background
4 General _ Design: Criterion.22 (Ref.- 1) and IEEE-279 (Ref. 2) require that the independence of redundant safety-related systems be assured in part by. physical l
separation of redundant' instrumentation and control circuits.
In portions of.
l the main control board, it is often necessary to locate control switches of 4
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redundant safety-related divisions on the same' panel.so that human factors l
regarding operability lare-not compromised. Where minimum separation distances are not possible, barriers between redundant circuits can-be an acceptable.
L method of assuring independence.
The FSAR-(Refs. 3, 4) allows the use of braided metallic sheath as a barrier q
where 6 inch physical separation cannot be-provided within the main control l
board, provided that maximum air space between cables of different trains has been maintained. and that the braided metal sheaths of different trains never touch and cannot migrate with time to touch. These. requirements are also i
stated in TVA design criteria (Refs. 6 and 7).
Description During the July. 28, 1987. team walkdown and subsequent detailed inspections, the IDI team inspected the ERCW panel 0M-27A and observed that'in several places braided metal sheaths of redundant divisions were either touching,~could migrate with time to touch, or could touch if ' disturbed by maintenance personnel.
I Basis The installation of main control board internal wiring.in several. cases fails to conform to the FSAR commitments assuring independence of redundant safety-related circuits (Refs 3 and 4).and to TVA criteria (Refs. 6 and 7).
Contrary to the FSAR consnitments, control cable' of redundant divisions, enclosed by braided metallic sheaths, were observed to be.in contact or were' not adequately secured so that they could migrate with time to be;in contact.
Impact on Design 1
The installation is unacceptably vulnerable to common mode failures-of redundant' circuits since FSAR separation criteria were not followed in the installation..The independence of redundant engineered ~ safety features is l
compromised beyond the design basis.
- Extent, This nonconformance might occur anywhere in_the main control board where redundant safety-related circuit wiring appears.. Therefore, TVA'must inspect DRAFT 84 -
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and correct as required all other panel sections of the main control board for both units. Conformance to all separation criteria applicable within the main control board must be assessed.
References 1.
10 CFR 50 Appendix A General Design Criterion 22. " Protection System j
Independence",
2.
IEEE 279-1971, " Criteria for Protection Systems for Nuclear Power GeneratingStations",Paragragh4.6.
3.
FSAR 7.7.1.10. " Control Board.
4.
FSAR 07.31, Q7.43 (NRC questions /TVA responses regarding limitations-on use of Belden braid).
5.-
Westinghouse Specification 678879, Rev. 0, " Main Control Room Vertical Panels"..
6.
TVA drawing 45W1640, Rev. 4, " Wiring Diagrams / Control Boards Critical Wiring / Braid Installation", Notes 1, 2, and.3.
7.
TVA. Design Input Memo DIM-SQN-DC-V-12.2-9, June 25, 1987,
Subject:
" Separation of Electrical Equipment and Wiring Design Criteria, SQN-DC-V-12.2", Section 4.4.6.1, " Enclosures Containing Wiring for More Than One Division of Redundant Class IE and Non-Divisional Circuits".
8.
FSAR 3.1.2.3, " Protection and Reactivity Control Systems".
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1 US.2-6 (UNRESOLVEDITEM):
REVISION'4-October 5, 1987
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ADE0VACY OF ELECTRICAL SEPARATION OF ISOLATION'OEVICE INPUTS AND OUTPUTS
Background
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Electrical' independence bet' ween IE and non-1E circuits is necessary to prevent l
- a. failure.in a non-1E circuit from propagating into a Class:1E circuit.-
Isolation devices.are employed as a means of assuring this independence.
Description U
TVA issued <ECN 2480 to add surge suppression' networks to all solid state Crydom J
relays used for the Status' Monitoring System (SMS) in accordance with.TVA's cosmiitment to R.G.1;47. These relays and associated suppression networks were.
' installed in enclosures of the control-' equipment from which'the status signal l
had been derived (such as.switchgear:and motor control centers) and. provide-_
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electrical isolation between the Class 1E side of the equipment providing the e
signal and the SMS which is. classified as non-1E.. TVA drawings 6858075 and.
6858066 show that non-1E wiring connecting to the non-1E side of the electrical isolator has been bundled together with Class 1E signal wiring and wiring for j
the 125 volt DC switchgear control power.
Since the Class 1E and non Class 1E 1
wiring is bundled together, the Class 1E control power may be degraded due'to 1
the propagation of a credible ' fault in the non-Class 1E circuit to the Class IE 1
control power. circuit.
This would prevent the. Class 1E switchgear-from 1
performing its safety function.
Bundling the Class IE and non-Class'1E wiring together in this manner could defeat the purpose of the isolation device.
Potential Basis FSAR Section 3.1.2.3 (Ref. 10) states that all protection systems are designed in accordance with IEEE-279-1971.
IEEE-279. requires (Ref.7 11) that no credible
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failure at the output of an-isolation device shall prevent the assoicated j
protection system channel-from meeting the minimum performance requirements specified in the design bases.
Electrical separation criteria 4.4.7 of SQN-DC-V-12-2 (Ref. 9) states' that no credible failure of non-Class 1E circuit shall prevent the Class. IE circuit -
1 from performing its design' basis function.
TVA has not demonstrated that the wiring of these isolation devices meets these; criteria.
j Impact on' Design 1
Effect of the credible 1 fault in the.non-1E circuit.on the Class IE control' l
power circuit may degrade the control power to the.switchgear such that'the switchgear may not be able to perform its intended safety function.
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DRAFT o.
Extent According to TVA, this wiring practice is used extensively in the plant.
Therefore, TVA should justify this design practice on a generic basis.
References 1
1.
TVA Drawing No. 33-470350-664, Connection Diagram Panel-8, ERCW Pump F-B, Rev. 2.
2.
TVA Drawing No. 6858066, Low Voltage Metal Enclosed Switchgear/480V Shutdown Board 2B1-B, Rev. 7.
3.
TVA Drawing No. 6858D75, Low Voltage Metal Enclosed Switchgear/480V Shutdown Board 282-B, Rev. 8.
4.
TVA Drawing No. 33-47035-0474, Connection. Diagram Panel-8, ERCW Pump F-A, Rev. 1.
5.
TVA Drawing No. 6858046, Low Voltage Metal Enclosed Switchgear/ Shutdown Board 2Al-A, Rev. 9.
6.
TVA Drawing No. 33-47035-D565, Connection Diagram Panel-14, RHR Pump 2A-A.
Rev. 5.
7.
TVA Drawing No. 45N765-15, Wiring Diagram 6.9KY Aux. Power, Schematic Diagram SH.15. Rev. 7.
8.
Work Plan 12501.
9.
TVA Design Input Memo, DIM-SQN-DC-V-12.2-9, 6/25/87
Subject:
Separation of Electrical Equipment and Wiring Design Criteria, SQN-DC-V-12.2.
10.
FSAR 3.1.2.3, " Protection and Reactivity Control Systems".
11.
IEEE Std. 279-1979, " Criteria for Protection Systems for Nuclear Power Generating Stations", Paragraph 4.7.2.
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U S '. 2-7 (UNRESOLVEDITEM)
REVISION 3
_ October 5,:1987 USE OF COMMON PENETRATIONS FOR R50VNDANT INSTRUMENT LINES i
Background
For assuring independence of redundant safety-related instrument lines, q
redundant lines should not share the same penetration.
Such sharing of t
penetrations can be a. violation of single failure criteria, _ and the -
corresponding provisions in IEEE-279 for channel. independence.
Description Per Paragraph 2,JSection 4.4 of SQN-DC-V-10.5, '" Design Criteria for: Separation I
of Instrument Sensing Lines and Instrument Air Lines," redundant instrument
' lines may be routed through a common penetration when separate penetrations -
I cannot be used..This lack of spatial separation could cause a loss of-
)
redundant instrument lines as a result of an accident, internally generated i
missile or other hazard.
~
Potential-Basis FSAR Section 7.1.2.2 comits to IEEE 279-1971.
IEEE-279 Section 4.6, Channel Independence, requires that channels that provide signals for the same protective function shall.be independent and physically separated to accomplish decoupling of.the effects of unsafe environmental factors, electrical
.]
transients, and' physical accident consequences._ Sharing of penetrations'by.
redundant instrument lines could violate this criterion.
Impact on Design The safety function provided by the redundant instrument lines would not be available if. the comon penetration is destroyed during an accident or other event.
l Extent The practice of using a common penetration may have been~ extended to allow use-of a comon instrument rack or other common' structures for redundant instru-ments. TVA should identify the extent of this practice, and justify any 1
specific cases where penetrations or other common structures are shared by redundant safety-related instrument lines.
~
1 References 1.
Design Criteria No. SQN-DC-V-10.5, Rev. 1, 8/24/84, " Design Criteria for Separation of Instrument Sensing Lines and_ Instrument Air Lines."
2.
IEEE Std. 279-1971, " Criteria for Protection Systems for Nuclear Power 1
Generating Stations".
3.
FSAR Section 7.1.2.2, " Independence of Redundant Safety-Related' Systems"...
-DRAFT
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05.2-10 (DEFICIENCY)
' REVISION 6 October 5, 1987 ADEQUACY OF ERCW INSTRUMENTATION PROVIDED FOR DETECTION'0F BREAK I
IN NON-SEISMIC ERCW PIPING l
1-
' Background In the event of a pipe break in the non-seismic portion of the ERCW piping in l
the turbine. building, operator action is. required to isolate the-break and J
prevent degradation of ERCW capability.- In the plant design,-operator action 1
is initiated by a high flow alarm and status light in' the control room that monitors.each ERCW' header train that would supply water to the break in the 4
non-seismic portion of the piping.. The maximum credible break would'be l
equivalent to a double-ended break'of the two four' inch lines in the non-seismic portion of the piping. _ For this event, the operator.is required to 1
isolate the break from each header by closing the isolation valves 0-FCV-67-205 l
and 0-FCV-67-208.
l Description The team reviewed the ERCW design criteria, flow diagrams. control diagrams,-
connection diagrams, instrument tabulation and ECN L5414. The team also j
l discussed these alarm channels with engineering and operations. personnel, and 1
reviewed the system operating instruction (S0I).
Since the annunciator
]
hardware is not seismically qualified, no credit can be taken for the high-flow alarms in assuring timely operator action during a seismic event. Moreover, even if the alarms were seismically qualified, a. seismic event would likely-result in numerous alarms competing for the operators' attention. Likewise the-status lights, which.were stated by TVA to_be seismically qualified, would not assure timely operator action during a multiple alarm seismic event and its resulting plant transient.
The team also observed that:
(1). TVA was unable to retrieve.a calculation justifying ERCW system function-ality associated with the minimum required operator response time for a double ended break in the non-seismic ERCW piping..This is further l
discussed in Deficiency D2.2-5.
(2) The high flow alarm had not been identified as a high priority alarm.by-color code (Ref. 3).
(3) ' System operating instructions- (Ref.:5) require that the high flow alarm.be verified, but no specific procedure to verify the alarm is provided.
The 1
flow value -is not-displayed in the control; room.
Therefore, specific ar.d timely verification of the break under all system conditions may not be -
assured.
1
. DRAFT _
=__
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-(4) The high flow. alarm occurred during a supplemental.walkdown of the ERCW panel by.the team; the team observed-that the' corresponding white 4
indicating light did not light during the.high flow alarm, and was told that there had been prior. invalid occurrences of this alarm. The team also noted that TVA has initiated action to calculate a new setpoint for-1 these alarms-(Ref. 6). The. team had' planned to review these calculations when issued by. TVA; however, the calculations were not issued during the inspection.-
l Basis.
FSAR section 9.2.2.1 states that. provisions are made to ensure a continuous i
flow of cooling water to those systems and. components necessary for plant j
.i safety either during) normal operation or under accident conditions.SQN-DC-V
- j
- the nonsafety-related piping in the. turbine building which will require the j
main control room operator to intitiate' pipe break isolation. features.
The.
j design is not in accordance.with these requirements, since the annunciator is i
not seismically qualified, and the existing alarm does not appear reliable.
Moreover, there is no assurance that even a seismically qualified' alarm or indication would assure timely operator response required to assure continued j
operability of the ERCW during a multipe-alarm seismic event since, TVA has not i
provided an analysis,that establishes.the required operator response time.
Impact on Design The operator may fail to recognize and respond to a break'in non-seismic ERCW piping in a timely manner during a seismic event.
This would result =in' failure of both ERCW trains to supply sufficient' cooling water to accident or safe shutdown loads, and consequently could lead to failure of both trains of engineered safety features to perform their required safety functions.-
Extent Since TVA is taking credit in the ERCW analysis for timely operator action being initiated by this alarm, all safety-related systems where credit is tance for alarms initiating operator action may be deficient.in that'(1) the j
annunciator is not seismically qualified, (2) the alarm may'not assure' timely-1 operator response during a multiple alarm event and.(3) analyses may not exist-justifying operator action vs. automatic action or other design features.
References 1
1.
TVA Drawing 47W845-5, Rev. M.-3/22/86, " Mechanical Flow Diagram - ERCW",
2.
TVA' Drawing 47W610-67-5, Rev. M. 1/12/86,r" Mechanical Control Diagram -
l ERCW".
3.
Mechanical Instrument Tabulation 47B601-55, Rev. E,10/23/81 (shows i
annunciatorfront. view).
l 4.
ECW L5414', 4/21/81 (Replacement of flow elements with elbow taps).
5.
S0I-55-0M8278-XA-55-278-0, Rev. 4, Locations 15 and. 22 (Defines procedure 1
for response to alarm).
1 DRAFT.-_____
1 i
6.
CAQR'No. SQF 870132, 7/14/87 (Recommends corrective action to determine new setpoints-for FS-67-208, 204).
7.
FSAR 9.2.2.1, "(ERCW) Design Bases".
8.
ERCW Design Criteria, SQN-DC-V-7.4, Rev. 2, 7/11/86, Section.3.5.2.1.
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ORAFT 91 -
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i DS.3-1 (DEFICIENCY)
REVISION 5-October 5, 1987 INADEQUATE SHUTOOWN CAPABILITY OUTSIDE CONTROL ROOM:
TRAVELING SCREEN /SCREENWASH CIRCUITS
1 Background
]
The design basis for the plant includes an extensive fire in the control room-that would require control room evacuation (Ref. 1).
For such an event, credit 1
is taken for alternate controls outside the control room that would not be damaged by the postulated fire.
Operation of the traveling screen _and screenwash pumps _may be required to assure operability of the ERCW system for plant shutdown. 'Therefore,.the
{
design of the control circuits'must assure that the traveling screen and
{
screenwashspumps are capable of operation from outside the control room
- j following a fire in the control room.
l Description The team reviewed the schematics for the traveling screen drive and screenwash i
pumps (Ref. 2). While the control devices for these components are located outside the control room area, indicating lights in the control room are directly connected to the control circuits which must remain operable after th-fire. A control room fire could therefore render all of the ERCW traveling screens and screenwash pumps inoperable until. damage was identified and repan i
were made. The failure mechanism could be permanent short circuit of the j
controls and blown control power fuses, which would render the screens inoperable.
Basis The FSAR Section 7.4.1.3 (Ref. 3) requires that controls be provided outside the control room for the ERCW system such that safe shutdown can be achieved i
during a design basis fire. TVA design criteria-(Ref. 4) requires that screening of reservoir water be provided during emergency conditions.
The design of the screenwash drive and screenwash pump circuits is not in accordance with these requirements since the. circuits could be rendered inoperable after a design basis fire requiring control room evacuscion.
Impact on Design Failure of screen drives or'screenwash pumps could result in loss of suctier,
the ERCW pumps during conditions when debris is present. As a result, the l e.
system could become degraded and loss.of both trains of safe shutdown system could oc, cur.
4 4
DRAFT- -
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Extent The deficiency appears to be limited to the ERCW traveling screen and screenwash pump circuits.
However, since the root cause.of the deficiency may be a. failure to recognize vulnerability of local control circuits to fire indticed failures in portions of the circuits that provide status indication in -
the control room, all local control circuits for which credit is taken for safe shutdown outside the control room should be examined for this deficiency.
References
]
1.
FSAR 7.4.1.1, " Control Room Availabil.ity".
2.
TVA schematic Diagram 35W726-1-( As constructed) Rev. F.
3.
FSAR 7.4.1.3, " Systems Available for. Hot Shutdown".
4.
SQN-DC-Y-1.4.6, Rev. 3, 2/12/86, " Traveling Water Screens for ERCW Pumping Station", Section 3.3.1, "Importance to Safety".
j I
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D5.4-1 (DEFICIENCY)
REYlSION 5 October 5, 1987 I
ADEQUACY OF FREEZE PROTECTION FOR-INSTRUMENT LINES IN THE ERCW PUMPING STATION 1
J
Background
Worst case environmental conditions for areas containing safety-related:.
equipment are established in the FSAR (Ref.1).
Safety-related equipment not designed or qualified for these established environmental' extremes must be -
served by fully redundant environmental control facilities such that no loss of-engineered safety features. (ESF) equipment will occur from a single. failure.of
{
equipment provided for controlling the local' environment for ESF equipment (Ref.2).
In the ERCW pumping station, means must be provided for assuring that vital instrument lines will not freeze and defeat the function off the -
instrumentation.
.)
description The team observed that no heat tracing was provided for instrument lines for the ERCW-strainer differential pressure instruments and other instrument lines within the ERCW pumping station.
TVA indicated that.an earlier calculation -
(Ref. 3) had taken credit for electric space heaters in assuring that temperatures in the ERCW pumping station would be maintained above freezing; however, the space heaters are not Class 1E, therefore no credit can be taken for their operation.
H TVA stated that heat losses from sources inside the ERCW pumping station would maintain the environment above freezing for all equipment located below El. 720-ft., and'that' heat. tracing was provided at and above El. 720 ft. The team requested TVA to provide an analysis ~ demonstrating that'the instrument lines would not freeze without the use of space heaters or other environmental control measures.
Problems with these TVA analyses are discussed in Deficiency D2.2-7.
plant since-1979,.and observed that MRs A-236310 and(A-236311 datedIn ad December 26,1983 (Ref. 5) requested work to thaw transmitter sensing lines for transmitters 0-PT-67-461and0-PT-67-445(ERCWpumpQ-Aand'M-Bdischarge.
pressure transmitters). The corrective action was to replace a circuit card that had. failed on the heat trace controller.
There was.no record'of whether any untraced lines below El. 720-ft. (for example, at elevation 704-ft. or t
688-ft.)hadfrozen.
Discussions with TVA_ indicate that freezing appears to have been detected by responding to a heat trace system trouble alarm rather:
than by, direct inspection. The December 27, 1983 shift log sheet for surveillance instruction SI-606 (Ref. 6) indicates that the ambient temperature at El. 704-ft, near a transformer varied betweer. :: approximately 40-42*F'on that DRAFT 7.,.
4 I
1 day, while the outside temperature (auxiliary-building elevation' 763-ft.)
varied from approximately 0-1*F two days before.
Presumably, areas further 1
from the transformer could be significantly colder (particularly at lower elevations) since there do not appear'to be any other substantial Waste heat sources in the areas of interest. While all pertinent facts could'not be determined from the records available, the team concludes'that recent operating history suggests the existing design may' extreme cold weather.not preclude freezing of un lines in the.ERCW pumping station during Basis 10 C.F.R. 50 Appendix A. Criterion 4 requires that structures, systems and ~
components important to safety shall be. designed to' accommodate the effects of and to be compatible with the environmental conditions' associated with normal operation, maintenance,: testing and postulated accidents.
FSAR. Section 3.11.4' requires that redundant and qualified environmental control; facilities be provided such that no loss of ESF equipment will occur from a single failure of environmental control equipment (Refs. 1 and 2).
TVA Design Criteria SQN-DC-V-7.4 section 3.7.1.6 (Ref. 4) requires that all instrument lines exposed to freezing conditions be protected to maintain them at a non-freezing temperature.
The design doestnot assure that these requirements are fulfilled, since no Class 1E environmentally qualified heat sources are provided to prevent freezing conditions, and no analysis had been performed which adequately justifes the absence of qualified environmental control features.
Impact on Design Freezing of the instrument lines used for strainer differential pressure measurement could defeat automatic operation of all four strainers and defeat the differential pressure alarms which alert the operator to a clogged strainer.
As a result, both trains of ERCW for both units could-be rendered-ineffective; this could defeat both trains of ESF or safe shutdown equipment for both units.
Extent This deficiency could extend to other plant areas where credit is taken for local unqualified environmental controls or where acceptable analysis has not-been performed to assure operability of instrumentation during extreme environmental conditions. TVA should review safety-related instrumentaticr. o its environmental protection where it is located in such areas.
References 1.
FSAR 3.11 " Environmental Design of Mechanical and Electrical Equipment 2.
FSAR 3.11.4, " Loss of Ventilation".
3.
TVA Calculation 845860905235, Rev. 6. "Suninary of Mild Environment '
Conditions for Sequoyah and Watts Bar Nuclear Plant".
-DRAFT'
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4.
ERCW Design Criteria,~SQN-DC-V-7.4, Rev. 2, 7/11/86, Section 3.7.1.6.
5.
TVA Maintenance Requests A-236310 A-236311. 12/26/83.
1 6.
SI-606 Shift Log Sheet, page 2 of 2,12/27/83 and 12/25/83 (B0P Temperature Monitoring Program).
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05.4-2
'(DEFICIENCY)
REVISION 3 October 5, 1987 SEISMIC QUALIFICATION OF WESTINGHOUSE SWITCHES j
Background
Control switches'and the panels on which the switches are mounted must be 1
seismically q Switch function under specif. ualified. - They can be' qualified separately.ied seismic accelerationj accelerations at switch mounting ~ locations are. typically computed by analysis
'or determined by test.
If the switch is' tested to acceleration levels:
sufficiently higher..than the predicted panel acceleration' levels, the switch is considered seismically qualified for the intended service.
Descriptio'-
n Westinghouse Report WCAP-8540 (Ref.1)'sumarizes the seismic tests for Westinghouse type OT-2 and W-2 switches.. Wyle test; report' No - 42545-1-(Ref. 2) qualifies Westinghouse vertical auxiliary panel 0M-27A, on which a number of '
i the Westinghouse switches are mounted.
However TVA could not retrieve a j
calculation to confim that the maximum panel accelerations at the switch locations satisfied the commitments detailed in FSAR Section 3.10.2.
This FSAP section_ requires that the device panel mounting locations be' limited to maxim a accelerations less than three-fourths of the value of the Safe. Shutdown Earthquake peak response acceleration as determined from the appropriate response spectrum or.three-fourths of the actual. device test acceleration.
J This FSAR requirement was not evident'in the procurement documentation, nor.45 compliance with the criteria clearly demonstrated by test or analysis.
\\
Basis This deficiency is a failure to comply with FSAR Section 3.10.2 commitment (Ref. 3); it appears to result from a failure to comply with the FSAR Sectiv -
17.1A.1.6 requirement (Ref. 4) that TVA's Division of Engineering Design re,'.
specifications-and manufacturer's test results for NSSS furnished equipment.
Impact on Desion
~
Since qualification of the-main-control room panel mounted switches has. not been adequately demonstrated, there is-not sufficient assurance that virtual'-
any engineered safety features or safe shutdown circuits will function durin.
and following a seismic event.
Extent This deficiency applies to all control room panels.
DRAFT _ _ _ _ _
References 1.
WCAP 8540,." Seismic Qualification of the Full Size. Main Control Boards'-
Sequoyah and Watts Bar Nuclear Power Plants," May 1975.
2.
Wyle Test Report No. 42545-1, Rev. O, " Seismic-Simulation Test Program on One Vertical-Auxiliary Panel:0M-27A." 8/1/73.
3.
FSAR 3.10.2, " Seismic Analyses, Testing. Procedures, and Restraint Measures".
4.
FSAR 17.1A.I.6, " Quality Assurance During Design and Construction -
Engineering Division", Amendment 49.
5.
TVA Receivino Report No. SNP 73-5515, 6/11/73, and attachments (for Panels i
OM27A,OM27BJ.
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DRAFT
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1 U5'.4-3 (UNRESOLVEDITEM)
REVISION 4 October 5, 1987 ig ADE00ACY OF SEISMIC QUALIFICATION FOR' FIELD' LOCATED RELAYS, TIMERS AND TERMINAL BLOCKS
-Background
- When modifications areLmade'to seismically qualified electrical. or instrument ;
1 pa'nels. the modifications must not' result in a. change to the assembly configuration such that the new configuration departs significantly. from the seismic qualification bases of the original' configuration.. Such departure.
could result from
~)
1 1.
Changing the distribution'of mass ~on the panel..
2.
Locating new components in positions where the panel. accelerations either exceed or are unacceptably close to the. accelerations to which the component is qualified.
3.
Locating new components in, positions where panel accelerations are unknown.
Because of the potential for invalidating the qualification of the. devices or panels, engineering controls are required on modifications to seismically qualified electrical and instrument panels.
Description TVA issued ECN 2480 to add surge suppression networks to all solid state Crydam relays used for the Status Monitoring $ System (SMS) in.accordance with TVA's commitment to RG 1.47.. These relays and associated suppression networks were installed in enclosures of the control equipment from which the status. signal had been derived (such as switchgear and motor. control centers)'and provide electrical isolation between the Class 1E side of the' equipment providing the j
signal and the SMS which is classified as non-Class 1E. The connection diagrams show installation of various compon'ents (relays, timers, terminal!
blocks, etc.) incorporated by several other ECN modifications. The installation instructions for these items include the notations:
" field to' locate" and/or." field to order and locate."' Thus these instructions. appear to-direct the field personnel to;use their own. judgement when. installing.
1 additional, equipment 'in Class 1E panels, without proper analysis of.the effects
.of such installations upon seismic qualification of the. components or: che engineering regarding ' location (s)y evidence of field personnel infonnin panels. The team did not find an of mounting after installation'.
Field interfaces appear to be limited to situations where a problem is encountered in the post-installation testing. These tests are limited to electrical and functiohal characteristics.
Since many ECNs have~been completed this way,and if each ECN added some " field to locate" equipment in these panels, these panels may have significant added mass and the components-may have been. located i
DRAFT '
l; ;._.;
l u,
at points where the panel acceler'ations exces.J or are unacceptably ~is apparently-close to the accelerations for which the component is qualified.
Effects of th 1
uncontealled configuration on seismic qualification of the panels and H
components may not have been analyzed and therefore it is possible that qualification might have been voided or at least cannot be demonstrated.
)
Potential Basis J
Class. IE; panels are required to be qualified in ~accordance with IEEE-344 and if i
any parameter to which these panels were ' qualified changes significantly,'an j
analysis to verify continued qualification must be done. There is presently no evidenca Bat TVA has considered the~ significance.of these changes.~or performed this analpis; However, this finding was identified late in the inspection, and TVA required time to retrieve additional documentation not reviewed by the i
' team.
Impact on Design
~
Since effects of the uncontrolled location of components and of the added mass on the se:xaic qualification of the. panels and components apparently has not been analyzed, the existing configuration of switchgear, motor control centers y
and other panels may not be seismically qualified.
Extent This deficiency applies.to many similar ECNs. Therefore, TVA should evaluate 4
i the genern. implications associated with field located devices on seismically qualified electrical components such as switchgear, motor control centers and other electrical or instrument panels.
i References 1.
TVA Drawing No. 33-470350-664, Connection Diagram Panel-8, ERCW Pump F-8, I
Rev. 2.
2.
TVA Drawing No. 6858D66, Low Voltage Metal Enclosed Switchgear/480V Shutdown Board 281-B, Rev. 7.
3.
TVA Drawing No. 6858075, Low Voltage Metal Enclosed Switchgear/480V Shutdown Board 2B2-B -Rev. 8.
4.
TVA Drawing.No. 33-47035-D474 - Connection Diagram Panel-8, ERCW Pump F-,
Rev. 1.
5.
TVA Drawing No. 6858046, Low Voltage Metal Enclosed Switchgear/ Shutdown Board 2Al-A, Rev. 9.
6.
TVA Drawing No. 45N765 Wiring Disgram 6.9KV Aux. Power, Schematic Diagram Sheet 15,'Rev. 7.
7.
TVA Work Plan No. 12501.
8.
TVA Design Input Memo, DIM-SQN-DC-V-12.2-9, 6/25/87,
Subject:
' Separation of Electrical Equipment and-Wiring Design Criteria, SQN-DC-Y-12.2".
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05.6-1 (DEFICIENCY)
REVISION 4; October' 5' 1987 INADEQUATE SPECIFICATION OF BACKGROUND RADIATION FOR ERCW EFFLUENT LIQUID RADIATION MONITORS
===.
Background===
-Redundant radiation monitors for each unit continuously monitor the-ERCW effluent to assure that releases are below permissible levels during normal-operation and to detect leakage from the component cooling or containment. spray ;
j heat exchangers during an accident. JFor the monitors to function properly, J
their measured values must not be undu_ly affected"by background radiation.
l
' Description The team reviewed TVA Specification 1491 (Ref.1)' and determined that background values for ERCW effluent liquid monitors 0-RE-90-133, -134, _-140, 1
-141 were specified as 1.0 mR/hr per Table 21.1 of.that specification. TVA has
'{
indicated that post-LOCA dose rates for similar locations in.the auxiliary building were calculated in TI-RPS-21 (Refs.: 2, 3)'and exceed 1.0' R/hr. The manufacturer's-analysis E-199-331 (Ref. 4) indicates that the background responseforthemonitorgs11,800cpmfor1mR/hrCo-60.
The monitor scale is d
specified as 10 cpm to 10 cpm. Therefore,.the monitors could'be driven beyond their full scale range-by accident background, and thus would_not indicate additional radiation due to a radiological effluent'. release during an accident.:
The team also noted that one of the.monitorsL hao actuated onLhigh_ background on i
December 7, 1982 and that TVA had then investigated moving the monitor away-H from the interfering source' (Reference 5).
However.TTVA'apparently failed to j
recognize the_ post-accident safety significance of the. monitors at that' time, since the corrective action was apparently' limited to preventing actuation during normal. plant operation, and avoiding a potential LCO violation if the j
monitors were inoperable due to background actuation.
Basis FSAR paragraph 11.4.2.1.2 requires that these monitors: serve as an accident monitor to detect leakage from either the component cooling heat exchangers cr containment spray heat exchangers during an accident and that in the event of a high radiation signal, the operator will initiate manual isolation of the leaking heat exchanger. The present' design' and installation of, these monitors does not assure these requirements will be met during design basis accident.
background radiation conditions.
Impact on Design 7
No credit can be taken for these monitors'as a means of' leak detection Lor effluent monitoring-during an' accident, or for assuring isolation.of a-radiological release.
i __ ____RFfVE___ q L - __ __u _2____ -u. M1Ri-
i l ] a Extent l" Incorrect background levels may have been specified for other post-accident radiation monitors. References 1. TVA Specification 1491 (for TVA. Contract 72C-61-92759 dated 1/27/72), " Radiation Monitor Equipment for Sequoyah. Nuclear Plant Units 1 and 2 and Watts Bar Nuclear Plant Units 1 and 2. 1 2. QIR NEB 87270 (no date. indicated - developed by TVA during the inspection). 3. TVA Engineering Design-Calculation TI-RPS-21, Rev. 0, 8/15/80. 4. General Atomic Company Document E-199-331, Rev. 2, 6/76,'" Radiation Monitoring Equipment TVA Sensitivity Analysis". 5. RIMS No. 845 860512893 TVA letter A27830325 (Comittment to evaluate moving the. ERCW radioactive liquid _ effluent monitor). 6. FSAR 11.4.2.1.2, "ERCW Effluent Liquid Monitors RM-90-133, 134, 140, 141. ~ j I i f i I i 1 o 102.- .j DRAFT __ _ - _ _ _ a
f .D6.2-2 (LEFICIENCY) REVISION 6 .0ctober 8,1987 1 INSUFFICIENT DEMONSTRATION OF ADEQUATE CLASS 1E MOTOR STARTING AND RUNNING VOLTAGES 1 I
Background
Acceptable performance of Class 1E motors can not be assured unless adequate j starting'and running voltages are supplied to motor terminals under the worst-case conditions of loading and power supply, j 1 Description j l The SQN calculation OE2-EEJCAL001. T.ev. 8 "AC APS Voltage and Loading Analysis",~ determined the L.'arting and running voltage levels at both 6.9kV and 480V motor terminals for coititions.in which offsite power istavailable. :The i L study did not consider the cundition when AC power supply is from the onsite 1 diesel generators. Section 1.2 of both,the medium voltage and the low voltage sections.of the { study indicates that the design basis condition considers Unit 2 with a LOCA j and Phase B containment isolation and Unit 1 with a Lfull-load rejection. However, Section 5.A of the low voltage portion.of the study states that Unit 1. q is considered in cold shutdown. Basis 1 ~ The voltage and loading analysis does not demonstrate that the auxiliary i electric power system. tad the diesel ' enerators have sufficient capacity and g capability to supply adequate starting and running voltages at'the Class IE-motor terminals under the'potentially worst case conditions. Thus, full compliance with FSAR.Section 8.1.2 cnd General Design Criterion 17 cited in Section 8.1.5 has not been demonstrated. Impact on Design 1 Excessive voltage drops could prevent safety-related motors from starting during an accident, leading to failure of safe-shutdown systems. Reanalyses of 1 the auxiliary electric power system considering the worst case loading condition anc power supply from the diesel generators instead of offsite power are required. Extent This item involves the SQN auxiliary ele;tric power system and the' capacity atd capabilities of its power supplies. 103'- . DRAFT E ______-z- __z_.__________
i References ~1' TVA Calculation No. OEZ-EEBCAL001, Rev. 8, 6/8/87, "AC APS Voltage and Loading Analysis". 1 2. SQN FSAR, Section 8.1.2, " Plant Electrical Power System" and 8.1.5, " Design Criteria and Standards". 3. ~ General Design Criteria 17,10 C.F.R. 50, Appendix A. 4. IEEE Std. 308 - 1971, " Criteria for Class IE Power Systems for Nuclear ' Power Generating Stations". 4 4 i 9. u i l i i 1 3 l 1 l l 104.- ' DRAFT 4
,9 ' ' .e?, } l .] D6.2-4 (DEFICIENCY)
- REVISION 7 October 8, 1987-ABSENCE OF NEUTRAL GROUNDING AND GROUND FAULT DETECTION ON 480V' AUXILIARY-POWER-SYSTEMS
. Background-An electric. power system which-has no. intentional connection between an active. conductor and ground is: called " ungrounded." Ungrounded systems are highly-susceptible to voltage transients severe enough to damage the insulation of-conductors and equipment. Intermittent: phase-to-ground faults are the primary-cause of these transients. Therefore, the generally accepted industry practice .(Refs. 1,,2, 3) is to install system grounding. In those unusual cases where-ungrounded power systems are applied, each distribution substation should' be equipped with ground detector circuits that sense and alarm phase-to-ground-faults. Otherwise, such faults can persist indefinitely without any indication because the fault current typically amounts to less than.one ampere in i ungrounded low-voltage systems. s Description y All of the 480V ac auxiliary power systems at SQN are ungrounded and have no ground fault detectors. l Basis This condition is a deficiency because the lack of ground detection allows aL violation of the " single-failure" criterion in NRC' General Design Criterion l' (Ref.4)',IEEEStd 279-1971(Ref.5)andNRCRegulatoryGuide1.53(Ref.6), and is contrary to FSAR commitments (Ref. 7). This is explained in detail under " Impact oa Design". In addition, FSAR Section 8.1.5.3, p.- 8.1-6, commits TVA to adherence to NEMA Standard WCS/ICEA Standard S-61-402 (Ref. 3), TLble 3.2.of which declares the*. the cable in ungrounded power systems must be rated for 173% of nominal syster. voltage, unless the system is equipped with ground fault protection-assuring-the interruption of ground faults within one hour of.occu rence. This. requires a minimum cable insulation voltage rating of 830V, whereas much if not all. of the '480V cable at SQN is of 600V rating. FSAR Section B.3.1.1 (page 8.3-25) declares that the 480V auxiliary power (distribution substations).are equipped trith ground fault detectors.. Neither the "as-constructed" drawings nor the_ team's discussions with TVA/EEB personne. bear out this assertion. Impact on Design As noted previously, intermittent ground faults on low-voltage systems tend N cause severe transient overvoltages. In addition..if the' fault is permanent' z_ RAFT 1 _105 - D
~~ p.v. .v ~. 1 rather than intermittent, the solid connection of one phase to ground imposes ~a continuous 73% overvoltage.on the phase-to-ground insulation of the two 1 unfaulted phases'. The overvoltage accelerates insulation deterioration throughout the system supplied by the affected load center-and. eventually precipitates a second ground fault since the cable,-in this case, is.not sized for the 73% overvoltage. Combined with theLinitial fault, this is equivalent-to a phase-to-phase short circuit.through ground. Because the initial L fault ' ' current is too small _to trigger a phase-to-phase fault at the original ] location, the second-fault is more likely occur on a different circuit thereby, disabling two circuits on the same load center. I 1 Furthermore, there is-a possibility:of a single-failure criterion violation. Consider the following scenario: (a) A ground fault occurs somewhere in the system supplied by one of the Train "A" Class 1E 480V distribution substations. Being undetectable..this fault is not counted as a failure" in the IEEE 279 analysis. (b) An incident requiring safety system actuation occurs.- (c) An-unspecified random failure disables Train "B". ~This is the first failure that counts under IEEE 279. (d)TheTrain"A"powersystemsustainsasecondground ) fault initiated by overvoltages caused by the first fault, disabling 'one or more critical circuits.- -Therefore, redundant components may be disabled b a i single detectable failure. j l The 480V distribution substations should be physically inspected 3 to determine if ground detector circuits are actually present, as the FSAR ~ states. If'so, the applicable "as-constructed"-drawings should be corrected. If not, ground detection circuits should promptly be installed at' all 480V distribution substations connected to alarm locally and in the main control room. If 1 properly designed (Ref. 2), these circuits will also provide effective high-resistance system grounding to suppress transient overvoltages. Extent i This situation prevails.in all 480V power distribution systems in the plant. grounded and has ground fault relaying.) ystem, which is low-resistance-(It does not exist in the 6.9kV primary s References 1. IEEE Std. M1-1976, IEEE Recommended Practice for Electric Power Distribution in industrial Plants. 2. TEEE Std. 142-1982,_IEEE Recommended Practice for Groundino of Industrial and Commercia_1 Power Systems. 3. D. Beeman, Ed., Industrial Power Systems Handbook, McGraw-Hill, 1968. 4. NRC General Design Criterion 17, 10 CFR 50, Appendix-A. 5. IEEE Std. 279 - 1971, " Criteria for Protection Systems for Nuclear Power Generating Stations". 6. NRC Reg. Guide 1.53 - 1971, " Application of the Single-Failure Criter_ ion to' Nuclear Power Plant Protection Systems". 7. SQN FSAR Section 8.1.5, "Destgn Criteria and Standards". i . DRAFT' - 106 -
g-8. NEMA Standard WC5-1974/ICEA Standard S-61-402 (1974), " Thermoplastic-Insulated Wire and Cable for Transmission and Distribution of Electrical Energy." e w e h MOW. M /7hQ
4.: 7 II t i ] b D6.3-1 (DEFICIENCY)' REVISION'7 October 8,;1987.- INADEQUATE VOLTAGE DROP CALCULATIONS FOR- .125Vdc AND 120Vac CONTROL CIRCUITS Background. Acceptable operation of'125Vdc and 120Vac relays, solenoids and other control equipment'can not be assured if applied terminal voltages are outside the ranges identifien by-their respective ~ suppliers. Description 4 TVA has' made-several studies and calculations to demonstrate.that at least the. minimum level of operating voltages, as-identified by suppliers, will'be applied to 125Vdc and-120Vac C1' ass IE devices and equipment in the event of a 1 design basis incident. The team found the following shortcomings in the j voltage drop' calculations:.(1) all the calculations have unverified assump-H tions,'one of which. the minimum operating voltage of one type of relay in l SQN-VD-VDC-1, may be critical to the validity of the results, (2). corrective actions identified by Significant Condition' Reports LSCRSQNEEB8065 and 1 SCRSQNEEB8532 regarding calculations SQN-VD-VDC-1.and SQN-VD-VAC-2 are yet to j be completed; (3) calculation'SQN-VD-VDC-1) considers a de board voltage 1 evel 1 of 120Vdc when analyzing the 6.9kV and 480V switchgear control devices. -Based on FSAR commitments recognized by SCR SQNEEB8065, Rev. 1 p the' battery terminal. voltage should be considered as 105Vdc. 'Also, calculation'SQN-CPS-001 should consider a voltage level of 105Vdc rather than 120Vdc for the same reason as (3)above. Basis-FSAR Section 8.3.2.1.1 requires.the 125Vdc vital power system to have sufficient capacity to permit safe reactor shutdown and supply all' safety-related de loads during a 2-hour " loss-of-all-ac-power" condition. This implies that the. system must supply adequate _ operating voltages to the loads throughout the 2-hour period. The de voltage calculations failed to consider. the worst-case battery terminal voltages'(see " Impact on Design" below) and contain unverified assumptions as to device. operation at degraded voltage, ard therefore, fail to demonstrate compliance with the FSAR commitment. 1 Impact on Desian. 1 Thus the voltages delivered to dc. loads,.while demonstrated to be' adequate at 120Vc source voltage, may not be adequate' iri the worst case.- Insufficient-voltage could result in misoperation of critical: safe shutdown: equipment durina an accident..The de voltage drop calculations'should be ' revised to reflect tre worst-case battery voltages and the resolution of unverified assumptions.: il t - ~ may prove necessary to correct low-voltage conditions by; increasing conductor sizes, re-routing cables,.etc.
- DRAFT.
- 108.- y,
J 1 ) I i Extent This deficiency involves the Class 1E 125Vdc and 120Vac vital instrument and control' power circuits. References 1. SQN FSAR, Sections 8.1.4, " Design Bases"..and 8.3.2.2.2, " Analysis of.' 3' Vital 125V DC Control Power Supply Systems"; Tables 8.3.2-2 through -5. 2. TVA Calculation No. SQN-VD-VDC-1, "125V De Vital Instrument Power System Design Verification - Preliminary", Rev. 2, 11/24/86. 3. TVA Calculation No. SQN-VD-VAC-2, "125V Dc Vital Instrument Power System Design Verification - Further Analysis", Rev. 2,5/16/87. 4. TVA SCR No. SQNEEB8065, Rev. 1. 5. TVA SCR No. SQNEEB8532, Rev.'O. 6. TVA Calculation No. SQN-CPS-001, "120V Ac and De Solenoid Valve Voltage Studyl', Rev.1, 6/1/87. i l M FT . 109 - 1_ x______
l , w.. p L i D6.6-1 (DEFICIENCY)
- REVISION 5 October 8, 1987 l
UNSUBSTANTIATED MOTOR-0PERATED VALVE PERFORMANCE AT DEGRADED VOLTAGE
1 Background
Acceptable performance of. valve motor operators used with safety-related valves can not be assured unless motor characteristics are specified to match the capabilities of their power supplies. Not all of the MOV motors were purchasea under such specifications. ,.] Description -) i Calculation OE2-EEBCAL001, Rev. 8, which analyzes the capabilities of the I auxiliary. power system, assumes motor operated valves will open with 80%' rated. voltage applied to motor terminals. Also, the. study assumes motor' operated' j valves without brakes will close at 75% of rated terminal voltage, and valves i with brakes at 80%. The calculations show that the motor operated valves required to operate for a LOCA with Phase B containment isolation will operate, based on these assumptions. The team reviewed procurement documentation ~ indicating that while some ERCW motor operated valves were specified to operate with a minimum terminal voltage of 75%, the purchase specification of other safety-related ERCW MOVs had no-degraded-voltage requirement. Basis I The degraded-voltage performance of some of the ERCW safety related motor operated valves was not defined in procurement specifications. Thus the assumption in the calculation of operation at lower voltage may not be valid. The. ability of the power system to supply adequate' voltage at the motor operated valve terminals, and thus full compliance with the general system specifications in FSAR Section 8.1.2 and the design criteria in Section 8.1.c has not been demonstrated. Furthermore, the failure of motor specifications :- reflect an important performance requirement is a deficiency in design contro:, and violates FSAR Chapter 17 comitments to engineering quality assurance in the procurement process (Ref. 4). Impact on Design ~ Inadequate terminal voltage and motor torque could lead to misoperation of safety-related valves during an accident. It will be necessary to establist and document the minimum operating voltages for the ERCW system MOVs'and compare them with the worst-case voltages from the power system study. This may necessitate larger cable sizes, cable re-routing, etc. Extent i This condition could affect some of the safety-related motor operated valves r both units. DRAFT - 110 -
~ V' References c 1.- TVA Calculation No. OE2-EEBCAL001, "AC APS Voltage'and Loading Analysis"; -Rev. 8, 6/8/87. 2. TVA Procurement Contracts including but not limited to: 87Z26-829681,8/14/81, 85P73-837118,4/18/85, 84P46-834280,2/8/84, 77K34-822049,4/26/77, 3. SQN FSAR, Section 8.1.2, " Plant Electrical Power System" and 8.1.5, " Design Criteria and Standards". j 4. SQN FSAR' Amendment 49, paragraphs 17.1A.1.6 (7) and 17.1A.1.6 (10), " Division of Engineering Design". 1 'j \\ i i 4 o 111 - DRAFT i _ _ _ - _ _ _}}