ML20205T125
| ML20205T125 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 11/03/1988 |
| From: | Richardson S NRC OFFICE OF SPECIAL PROJECTS |
| To: | White S TENNESSEE VALLEY AUTHORITY |
| References | |
| TAC-60081, TAC-60409, TAC-R00354, TAC-R00370, TAC-R00371, TAC-R354, TAC-R370, TAC-R371, NUDOCS 8811140142 | |
| Download: ML20205T125 (72) | |
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November 3, 1988 g
Docket No"..*S0-327/328 Distribution:
DocketJHe4 J. Donohew G. Georgiev NRC/LPDR OGC E. Marinos Mr. S. A. White Projects Reading E. Jordan R. Pierson Senior Vice President, Nuclear Power J. Partlow M. Simms 1@ M9 Tennessee Valley Authority S. Richardson ACRS (10) g g6 6N 38A Lookout Place S. Black K. Jenison 1101 Market Street B. D. Liaw SON Rdg. File Chattanooga, Tennessee 37402 2801 F. McCoy D. Crutchfield L. Watson P. Cortland i
Dear Mr. White:
SUBJECT:
REVISION 1 0F THE PRELIMINARY SAFETY EVALUATION ON THE TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PERFORMANCE PLAN - SEQUOYAH UNIT 1 i
(TAC 60081, 60409 R00370, R00371, R00354)
By letter dated September 21, 1988, the Nuclear Regulatory Comission (NRC) issued its preliminary Safety Evaluation (SE) on the Tennessee Valley Authority's j
(TVA) response to the NRC's 10 CFR 50.54(f) letter of September 17, 1985, relating to site. specific issues at the Sequoyah site for Unit 1.
This SE
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represents the staff's evaluation of the TVA Sequoyah Nuclear Performance Plan through Revision 3, with supporting documents, for Unit 1.
Volume 2 of l
NUREG-1232 was issued by the NRC on May 18, 1988. and addressed the Sequoyab Nuclear Performance Plan for Unit 2.
l Enclosed is Revision 1 of this SE on the SNPP and the restart of Unit I from i
its current extended outage. This revised SE adds the staff's evaluation of the restart electrical design calculations in Section 2.3.3.
This revised SE also modifies the sections on the (1) Design Baseline and Verification Program, (2) Superheat Transient (Main Steam Temperature Issue), and (3) Containment Isolation Leakage Testing Program.
Changes to the SE are marked with a lir.e on i
the right hand of the affected page.
The revised SE contains the staff's evaluation of all the restart issues for Unit 1 in the SNPP.
l The staff plans to issue this evaluation, in final form, as a suoplement to NUREG-1232. Volume 2.
1 Sincerely.
Original Signed by Steven D. Richardson, Director TVA Projects Division Office of Special Projects
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Mr. S. A. White
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Sequoyah Nuclear Plant CC' General Counsel Peaional Administrator, Reginn if Tennessee Valley Authority U.S. Nuclear Pegulatory Commission 400 West Sur.mit Hill Drive 101 Marietta Street, N.W.
E11 B33 Atlanta, Georaia 30323 Knoxville, Tennessee 37902 Pesident inspector /Sequnvah NP Pr. R. L, Gridlay c/o U.S. Nuclear Reoulatory Commission Tennessee Valley Authority 2600 Igou Ferry Road SN 157B Lookout Place Soddy Daisy, Tennessee 37379 Chattanocoa Tennessee 37402-2801 Pr. John T. LaPoint Mr. Michael H. Mobley, Director Tennessee Valley Authority Division of Radiolohical Health Sequoyah Nuclear Plant T.E.R.R. A. Building, 6th Floor P.O. Box 2000 150 9th Avenue North Soddy Daisy, Tennessee 37379 Nashville, Tennessee 37219-5404 Mr. M. Ray Dr. Henry Myers Science Advisor Tennessee Valley Authority Committee on Interior Seouoyah Nuclear Plant and Insular Affairs U.S. House of Representatives P.O. Box 2000 Soddy Daisy. Tennessee 37379 Washington, D.C.
20515 Mr. D. L. Williams Tennessee Valley Authority Tennessee Valley Authority Rockville Office 400 West Suenit Hill Drive 11921 Rockville Pike W10 PP5 Suite 402 vnoxville, Tenne'see 3790?
Dockville, varyland 20052 County Judge Parilton County Courthouse Chattanoona, Tennessee 3740?
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4 Pevised Preliminary Sa'ety Evaluation Report On Tennessee Valley Authority Sequoych Nuclear Performance Plan Sequovah Nuclear Plant, Unit 1 Suonlanent to NUPrG-1232, Volume ?
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ABSTRACT The Safety Evaluation Report ($ER) Sequoyah Nuclear Performance Plan, NUREG-1232 Volume 2, was based on the information submitted by the Tennessee Valley Authority (TVA) in its Sequoyah Nuclear Performance Plan (!NPP),
through Revision 2, and on supporting documents.
It was issued on May 18, 1988, by the U.S. Nuclear Reculatory Commission staff for the restart of Sequoyah Unit 2.
The SNPP addresses the plant-specific concerns requiring resolution before startup of either of the Sequoyah units.
In particular, the SER addressed required actions for Unit 2 restart.
in most cases, the programmatic aspects for Unit 1 are identical to those for Unit 2.
TVA provided a description of the differences in programs between Unit 1 and Unit 2 in Revision 3 of the SNPP. This was submitted by TVA in its i
letter dated May 9, 1988. Where the Unit 1 program is different, the staff's evaluation is provided in this SSER which is a supplement to the staff's SER in NUREG-123?. Volume 2.
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On
'e basis of its review, the staff concludes that Sequoyah-specific issues have been resolved to the extent that would support the restart of Sequoyah lini t 1.
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TABLE OF CONTENTS Pace ABSTRACT.............................................................
1 INTRODUCTION......................................................
1-1 2 ADEQUACY OF DESIGN................................................
2-1 2.1 Plant Modification and Design Control........................ 2 2.2 Design Baseline and Verification Program.....................
2-3 2.P.1 Description and Scope.................................
2-3 2.2.2 Unit 1 and Unit 2 Program Differences............,....
2-3 2.2.3 TVA Independent Oversight Review.......................
2-4 2.2.4 Conclusions...........................................
2-6 2.3 De s i g n Cal cul a ti ons Prcg ram..................................
2-6 2.3.1 Nuclea r and Mechanical Calculations...................
2-7 2.3.2 Civil Calculations....................................
2-7 2.3.2.1 Description and Scope.........................
2-7 l
2.3.2.2 Ur,it 1 and Unit 2 Program Differences..........
2-7 2.3.2.3 Evaluation...................................
2-9 7
2.3.2.4 Conclusions..................................
2-10 2.3.3 El ec tri ca l Ca l cul a ti ons...............................
2-10
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t 2.3.3.1 Introduction.................................
2-10 7
2.3.3.2 Evaluation...................................
2-11 i
2.3.3.2.1 Auxiliary Power Systems................
?-11 2.3.3.2.2 Control Power Systems..................
?-14 I
l 2.3.3.2.3 Instrumentation and Control Systems Instrumentation Accuracy Calculations........................
2-16 2.3.3.2.4 Raceway Systems.......................
2-17 7.3.3.2.5 Short-Circuit Study - Medium Voltage i
Systems.............................
2-17 2.3.3.2.6 Technical Specification Surveillance I
Requirements........................
2-17 2.3.3.3 General Conclusions on the Sequoyah r
Electrical Calculations Program.........
2-18 e
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2.3.4 Branch Technical Position PS9-1.......................
2-18 t
2.4 Alternatively Analyzed Piping and Supports...................
2-19 i
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TABLE OF CONTENTS (Continued)
Pace 2-20 2.5 Cable Tray Supports..........................................
2.5.1 Interin Acceptance Criteria...........................
2-20 2.5.1.1 Evaluation...................................
2-20 2.5.1.?
Implementation of Interim Criteria...........
2-21 2.5,1.3 Anchoring in Concrete........................
2-22 2.5.1.4 Base Plate Analysis..........................
2-22 2.5.1.5 Concrete.....................................
2-22
?.5.1.6 Co n f i rma t o ry i t ems...........................
2-22 2.5.1.7 C o r, '. u s i o n...................................
2-22 2.5.2 Diesel Generator Buildino Supports Analysis...........
2-22 2.5.3 Cable Tray Support Base Plate Installations...........
2-23 2.6 Concrete Quality.............................................
2-23 27 Miscellaneous Civil Engineering Issues.......................
2-23 2.8 Heat Code Traceability.......................................
2-24 3 SPECIAL PROGRAMS..................................................
3-1 3.1 Fire Protection..............................................
3-1 3.1.1 Program Evaluation.....................................
3-1 3.1.2 S ta f fi ng of the Fi re B riga de...........................
3-2 3.1.3 Fi re Purp De s i gn De fici ency............................
3-3 3.1.4 Fire Protection Calculations Revision 9................
3-4 3.1.5 Inspection.............................................
3-4 3.1.6 Conclusion.............................................
3-4 3.2 Environrental Qualification of Electric Equipment Important I
to Safety....................................................
3-4 3.2.1 Compliance Vith 10 CFR 50.49..........................
3-4 3.2.2 Superheat Transient O'ain Stean Temperature Issue)....
3-5 3.3 Piece Part Ouelification (Procurement).......................
3-6 3.4 Sensing Line Issues...,.....................................
3-7 3.5 Weldino......................................................
3-7 3.0 Containrent isolation........................................
3-11 3.6.1 Containrent Isolation System Desion...................
3-11 3.6.7 Centainment Isolation Leakage Testing Program.........
3-11 2
3.6.3 Containment Leakage Testing............................
3-12 i
11 1
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s TAELE OF CONTENTS (Continued)
Pace 3-13 3.7 Containment Coatings.........................................
3-13 3.8 Moderate Energy Line Breaks..................................
3.9 ECCS Vater loss Outside Crane Wall / Air Return Fan 3-13 Operability.................................................
3-14 3.10 Platform Thermal Growth.....................................
3-14 3.11 Pipe Wall Thinning Assessment................................
3-15 3.12 Cable Installation...........................................
3.12.1 Program Evaluation....................................
3-15 3.12.2 Silicone Fubber Insulated Cable Er.vironmental 3-15 Qualification.......................................
3-17 3.13 Fuse Peplacement............................................
4-1 4 RESTART READINESS.................................................
4-1 4.1 Operational Peadiness........................................
4-1 4.1.1 Introduction..........................................
4-2 d.1.2 Evaluation............................................
4-4 4.1.3 Conclusions...........................................
4-4 4.2 Management...................................................
4-4 4.3 Cuality Assurance............................................
4-4 4.3.1 Quality Assurance Program..............................
4.3.? NPC Order EA 85 49.....................................
4-5 4-6 4.3.3 Changes to the CAQ Prog ram.............................
4-6 4.4 Op era ti ng Experience Improverrent.............................
4.5 Post-Modification Testing....................................
4-6 4-6 4.6 Surveillance Instruction Review..............................
4.6.1 Introduction..........................................
4-6 4-7 4.6.2 Evaluation............................................
4-8 4.6.3 Conclusions...........................................
4A 4.7 Op e ra bi l i ty "Look P a c k"......................................
4-8 4.8 Maintenance..................................................
4-9 4.9 Pestart Test Program.........................................
1-9 4.9.1 Introduction...........................................
4-10 4.9.2 Evaluation.............................................
4-11 4.9.3 Conclusions............................................
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TABLE OF CONTENTS (Continued) 4-12 4.10 Training....................................................
4-12 4.11 Security.....................................................
4-12 4.12 Emergency Preparedness......................................
4-13 4.13 Radiological Controls........................................
4-13 4.14 Restart Activities List......................................
4.14.1 Introduction......................................
4-13 4-14 4.14.2 Evaluation...........................................
4.14.3 Conclusions..........................................
4-15 5-1 5 EMPLOYEE CCNCERNS..................................................
6-1 6
/LLEGAT!0NS........................................................
APPENDICES A.1 A LIST OF CONTRIBUTORS iv
s 1
INTRODUCTION On September 17, 1985, the Nuclear Regulatory Commission (NRC) Executive Director for Operations issued a letter to the Chairman of the Aoard of Directors of the Tennessee Valley Authority (TVA) pursuant to Title 10 of the Code of Federal Reculationt Part 50.54(f) (10 CFR 50.54(f)). This letter reouested information on the actions TVA was taking to resolve NRC's concerns about TVA's nuclear program. These concerns were divided into four categories:
(1) corporate activities, (2) the Sequoyah Nuclear Plant (SQN1, (2) the Browns Ferry Nuclear Plant and (4) the Watts Bar Nuclear Plant.
TVA's Corporate Nuclear Perfornance Plan (CNPP), which was prepared in response to the NRC letter, was originally submitted to the NRC on November 1, 1985.
The revised plan was submitted on Parch 10, 1986, subsequent revisions were submitted to the NRC on July 17. July 31 and December 4, 1986 March ?6 and December 10, 1087.
The NRC staff Safety Evaluation on the revised CNPP, through Revision 4, was issued on July 28, 1987, as NUREG-1?32, Volume 1.
In addition to its corporate plan, TVA prepared separate plans to address site-specific problems at its Sequoyah and Browns Ferry nuclear plant sites.
A separate plan has yet not been submitted for the Watts Bar nuclear plant site.
Volume 2 of NUREG-1232 documents the staff's evaluation of the corrective 4
actions implemented by TVA to resolve problems at Sequoyah.
In rany cases, lono-term corrective actiers, extending beycnd startup, are required to fully resolve these issues.
The Seoucyah Nuclear Per'ormance Plan (SNPP) was submitted on November 1, 1985.
Devisions 1 to 3 to the plan were provided to the NRC by TVA on April 1 and July 2, 1987, and May 9, 1988, respectively.
Separate staff evaluations will be issued for Browns Ferry and Watts Bar at a later date.
j This Sucplemental Safety Evaluation Report (SSER) is a supplement to volume 2 of NUDEG-1737.
Volume 2 is the staff's evaluation of t5e restart of Sequovah Unit 2.
It was issued by the staff's letter dated May 13, 1980.
This SSER is the staff evaluation on the restart of Sequoyah Unit 1.
It is the staff's evaluation of the dif ferences in the Unit 1 SNPP programs from the prograns evaluated and aporeved for Unit 2.
These dif'erences were documented in TVA's letters dated March 31 and May 9, 19A8 and in the reeting with the staf' en April 14, 1988.
The reetinq sunrary was issued on May 4, 198P.
For Sequoyah Unit ?, TVA established a Sequoyah Task Force on March 19, 1986, to review irplementation o' the corrective actions applicable to Sequoyah, to initiate specific actions to address Sequoyah problems, to monitor and ensure 1VA SER Vol. ?, Supp 1 1-1 Preliminary Report, Pevision 1 4
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that a list of all known work items has been compiled, and to review the process and identification of those items required to be completed before 1
restart of Sequoyah Units 1 and 2, which were shut down by TVA in August 1985, i
This task force examined the distribution of Sequoyah-related issues that had i
been identified by the corporate level team of industry advisors, to confirm i
that root causes of problems were suitably addressed.
Sequoyah site-specific issues deal primarily with operations, maintenance, design control, end manage-f ment system implementation. The SNPP describes the programs and activities l
planned by TVA to improve performance in e6ch of these areas.
P To complete its assignment, the Sequoyah Task Force developed a list of Sequoyah plant activities (except for those of a routine nature) to be completed before restart.
The Sequoyah Activities List (SAL) was based on issues identified by NRC inspections TVA quality assurance (0A) audits, American Nuclear Insurers l
(ANI) audits, Institute of Nuclear Power Operations (INPO) inspection reports, Sequoyah corrective action reports (CAR) and discrepancy reports (DR). TVA p
Nuclear Safety Review Staff (NSRS) and Nuclear Safety Peview Poard (NSPR) l reports, erployee concerns, Sequoyah reactor trip reports and licensee event reports (LERs), and technical issues identified by TVA's Division cf Nuclear J
Enoineering (DNE).
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i The Sequoyah Task Force established criteria (Section IV.2.0 of the SNPP) to I
datermire which items were reautred to be resolved for restart.
The staff l
reviewed and accepted these criteria by letter dated June 9, 1987.
The task l
force reviewed the process the line organization used to identify, evaluate.
j disposition, and close out iters and reviewed the adequacy of planned actions l
to be taken before Sequoyah Unit 2 restart.
As new issues arose and work activities were developed, they were reviewed by Sequoyah management to l
j determire their importance to restart.
The Site Director had to approve all l
new items added to the restart list; however, only the Manager of the Office of Nuclear Power (ONP) (present title is Senior Vice President / Nuclear Power) l could delete items that had been desionated for restart.
I For Unit 1, the identification and tracking of restart items are being accom-plished by TVA's permanent trackirio system and reporting of open items (TROI) i J
computer system rather than by the SAL used for Unit 2.
The Unit I restart I
j list was developed by an item-by-iten review of completed and open Unit ? and comon restart activities and of open Unit 1 issues.
The criteria used to guide the line organizations in raising potential restart issues and making 4
j recorrendations to management have been the same restart criteria used for Unit 2.
The Site Director has desicnated eitha-the Restart Director or Assistant l
to the Site Director to evaluate proposed new activities and ascertain that 1
these activities meet the restart criteria.
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TVA described a number of special programs to ensure integrated corrective actions dealing with problems created by deficiencies in the past conduct of activities.
Secticn !!! of the original SNPP identified special prcgrams that needed to be resolved before restart of Sequoyah Unit 2.
These include i
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- p. grams to:
f complete the docurentation and resolve electrical equiprent environrental cualification questions initially raised at the tire Sequoyah was shut down TVA SER Vol. 2 Supp 1 1-?
Prelininary Report, Fevision 1
_7, j
i verify the adequacy, with regard to safe plant restart, of past i
selected safety-related design modifications keeping in mind the weaknesses in past design control programs j
reexamine cable tray support analysis for weaknesses in the analytical basis complete system analyses where proper design documentation did i
not exist in the past I
verify the adequacy of piping and supports that were t'ot rigorously analyzed and where alternate analysis methodolooy has j
heen poorly applied in the past resolve any differences in the effects of increased temperatures during main steam line breaks engendered by revised vendor analysis resolve identified areas of noncompliance with 10 CFR 50.
I Appendix R fire protection requirements i
assess the adecuacy of the welding program at Sequoyah, an issue i
raised through the employee concern program l
examine issues with regard to instrumentation sensing lines Since the original issuance of the SNPP, TVA added other special programs to Section !!! of the plan.
These include programs to:
determine if a problem exists with regard to pipe wall thinning, similar to that which led to a pipe rupture at the Surry Power Station, Unit 2, in Decer.ber 1986 establish a Restart Test Program review replacement components and parts and resolve those that do not meet the same quality recuirements as the installed equipeent l
f assess the adecuacy of cable ampacity design calculations
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resolve cable pullino concerns such as sidewall pressure, bend i
i radius, jaming, and vertical drop crrrect a misapplication of actuator fuses i
resolve an apparent nonconformance with 10 CFR 50, Appendir A, involving contairrent penetrations There are other programs as well to consider miscellaneous civil enoineering j
issues, mcderate eneroy line break flooding, containment coatings, emeroency i
core coolino system (FCCS) water loss outside the crane wall, platform thermal l
growth, and heat code traceability. Many of these programs are applicable to i
i TVA SER Vol. 2, Supp 1 1-3 Preliminary Report, pevision 1
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Units 1 and 2 although actual implementation for Unit 1 may not have been completed until after Unit 2 restart.
The programs mentioned above were evaluated for Sequoyah Unit 2 in Sections 2 i
through 4 of the Safety Evaluation Report (SER) on the SNPP through Revision
?.
This SER was issued as NUREG-1232, Volume 2, on May 18, 1988.
This SSER addresses the differences in the SNPP programs between Unit 1 and Unit 2.
These differences were described in Revision 3 to the SNPP which was submitted by TVA to the staff by letter dated Fay 9, 1988.
This SSER follows the same fornAt as NUREG-1232. Volume 2.
Where TVA has stated there are no differences, this SSER will refer to the appropriate section in the NUREG.
Another ra.ior orablem area included the concerns expressed by TVA employees regarding the quality of TVA's nuclear activities.
The programs relating to emplnyee concerns are br'efly described in Sectinn 5 of this evaluation.
The NDC plans for addressing altegations is discussed in Section 6 of this avaluation.
TVA SER Vol. ?. Supp 1 1-4 Preliminary Peport, Revision 1
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2.0 ADE00ACY OF DESIGN One of the root causes of the problems at Sequoyah was the failure to consistently document changes to the plant's design basis and to maintain the plant's configuration in accordance with that basis. TVA's efforts to strengthen its design control programs and to assess the effects of past weakresses on the plant are discussed below.
TVA's efforts in its design control programs were evaluated for Sequoyah Unit ?, in preparation for its restart, in the staff's SER, NUREG-1232, Volume 2 on the TVA Seqiioyah Nuclear Performance Plan (SNPP).
This section of this SSER is on the dif.
ences in the SNPP programs for Unit I concerning the adequacy of design. Wheie there are no differences in the Sf!PP program from that evaluated in NVPEG-1232 Volure 2. for Unit 2, the reader will be referred to the appropriate section of the NUREG.
This safety evaluatinn is a supplement (SSER) to the NUREG. Pecause this SSER has essentially the same table of contents as NUPEG-1232. Volume 2, the appropriate section is the same section as in this SSER.
By letters dated March 31 and May 9, 1988 and a meeting on April 14, 1988, TVA identified the differences in the SNPP prograns for Unit I from those approved by the staff for Unit 2.
The meeting summary for the April 14, 1988 meeting was issued on May 4, 1988.
These programs, for the adequac of design, are the design baseline and verification progran (DBVP)y(Section 2.2) and following:
design calculations (Section 2.3.2.).
TVA assumed Unit I was civil engineering (Mode 5) for many electrical design calculations.
These were in cold shutdown revised by TVA before the restart of Unit. and are cinussed in Section 2.3.3.
As with Unit 2, there are Phase 11 programs for DBVP, 1esign calculations program, alternatively analyzed piping and supports, and cable tray supports for Unit 1.
As discussed in NUREG-123?, Volume II, the Fhase I programs are cocpleted before the restart of the unit and the Phase !! programs will be completed after restart.
There have been meetings with TVA on June 72 and July Pl 1988 en Phase !! procrams.
The neeting sunearies were issued on July 1 and August 4, lo88.
The Safety Systen Quality Evaluation (SSOE) Inspection was conducted on the Unit 1 contairrent spray systen (CSS) by the staff in part to audit the adeCu3cy of the design progNes for Unit 1.
The SSQE was to provide additional assurance that the major programs had been properly implemented by TVA on Unit 1 and that the maior design and construction problems had been identified and resolved befort Unit I restart.
The S$QE is discussed in Inspection Report
(!R) 50-327/P,8 29 dated October 20, 1988.
All cemitments rade by TVA for these procrans for Units 1 and 2 are identified in the aceropriate sections below.
These comitrents may also be stated in NUREG-1232, Voluee 2.
TVA SER Vol. ?, Supp 1 2-1 Preliminary Report, Revision 1
2.1 Plant Podification and Design Contn' TVA did not identify any differences in the Unit 1 procran for plant modification and design control frot that of Unit ?.
The staff's evaluation of this program for Unit ? is ad Jressed in Section 2.1.2 of NUPEG-123?, Volume 2.
TVA's improved design change control program will be inclemented in two phases for current and future plant modifications.
The first phase was to be implemented before restart of Unit 2 and included a change control board and a transitional desian control system. This process requires that design changes that are to be implemented be contained in complete packaces specific to the appropriate unit.
This will facilitate the reviews required to ensure that each change has been quality engineered, that it can be installed and tested, and that documentation and safety analyses are complete and based on actual plant configuration.
A task engineer was assigned to coordinate these efforts.
The second phase in the development of the improved design control progran is to establish a permanent design control systen based on the plant modification package concept.
A procedure will be developed to ense e a comprehensive and focused evaluation of modifications and croper implementation and follow through.
Enhanced asnects of this program include the use of the actual plant configuration for design, updated desicn criteria, accurate reflection of the modification in licensing documents, and an integrated, pro.iect-oriented approach to handle changes tc the plant, as opposed to the fragmented work-plan approach used in the past.
In its December 11, 1986 letter, TVA committed to consolidation of the "as-constructed" and "as-designed" information on DBVP primary drawinos before the end of the second refueliro outage (Cycle 4) after restart of Unit 2.
The staff finds this cormitmert acceptable because (1) the fird refuelino is presently planned for saveral nonths af ter restart and (2'l in the interim, the actual confieuration will be depicted on marked-up drawings available for engineering and operational purposes.
By lettar dated December 15, 1087 TVA stated that Division of Nuclear Engineering procedures, which were needed to establish the crocess for preparing Sequoyah implementing procedures, have been implemented.
Site level procedures and training were completed by March 31, 1988.
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TVA has not cornitted to irplement a single drawin's systen for drawings other l
than DBVP drawings which are used by opera
- ions to operate the plant (primary drawings such as PAIDs).
Other drawings will apparently be produced only as needed to succort modifications.
The staff believes that a rore comprehensive approach, which includes schedulino details and identification of all other drawings to h maintained as configured, is needed.
In a letter dated April 1, in87(a), TVA stated that the detdils regardino compretvnsive scheduling of drawings to be maintained as-confioured is still bein4 Mevalnsed.
The staff considers this iten to be a rost-restart issue for both Units 1 and 7.
On tha basis of the f 4ndinos as documented in fl'DEG 193", VoIUme ?, the staff cercludes that TVA has ta6en the appropriate steps tn correct desion control problems at Sequoyah for the resort nf t! nit 1.
The staff will review the TVA SER Vol. ?, Supp 1 2e Preliminary Peport, cevision 1
transitinnal design control system during its review of the Phase !! portion of the DBVP.
The staff's evaluation of the DRVP is given in Section 2.2 below.
?.2 Design Baseline and Verification Program 2.2.1 Description and Scope TVA's special design baseline and verification program (DRVP) to assess the effect of past weaknesses in design and configuration control and to identify any corrective actions that may be required is addressed in SNPP Section !!!.?.
The DBVP is discussed in Section 2.7 of NUREG-1232, Volume ?.
The intent of this program is to provide additionti confidence that the *lant meets its oriainal licensing basis.
TVA's DRVP was developed to assess the effect of past weaknesses in desig, and configuration control at Sequoyah and to identify required corrective actions, i
The Unit 1 DBVP is described in Section III, Part 2 of the SNPP. The Unit 1 I
progi 3m includes (1) verifying and establishing the plant functional configur-ation, (2) reconstructing the design basis, (3) reviewing and evaluatins nodifications since the operating license was issued aoainst the design basis, and (4) performing the modifications developed from this review.
The Unit 1 PBVP was initially described in TVA's F' arch 31. 1988 submittal on the Unit I restart plan.
The Unit 1 DRVP is being implemented in two phases.
The pre-restart phase addresses the Unit 1 portion of the systems required to miticate accidents addressed in Chapter 15 of the Final Safety Analysis Report (FSARI and systems required tn ornvide safe shutdown.
The post-restart phase continues engineerino activities within the pre-restart phase that TVA consid-ered not essential to safe restart but are necessary to address identified desion control problems. TVA used the staff approved restart criteria to decide what was essential for safe restart.
The post-restart phase will also include other safety-related systens.
TVA defined the scope of the post-restart (Phase Ill portion of the DBVP in a v9y 12. 1987 letter.
The steff is enntinuing its review of the Phase !!
progran and this review by the staff is not essential to determine the acceptability of TVA's prograns to support restart of Sequoyah Units 1 and 2.
An evaluation of the Phase II program will be issued by the staff at a later datc.
The scope of the Unit 1 L'e-restart phase of the DBVP as described in Pevision 3 of the SNPP is ider.tical to the scope o' the Unit 2 program.
The staff review and acceptance of the Unit ? pre-restart phase of the ORVP is documented in Section ?.2.2 of NUDEG-1232. Vol. 2.
Based on the staff's previous review and acceptance of the Unit 2 pre-restart scope, the staff cencludes that the scope and systen selection for Unit 1 are the same and are thus acceptable.
2.2.2 Unit I and Unit 2 Pengram Differences l
Revisinr 3 of the SNPP identified two program differences between the Unit I and Unit 2 DBVP:
(li the Unit 1 procran takes credit for reviews that hati been perforrad under 'Fe Unit 2 procran and I?) resoonsibility for the review of testire has bean transferred to the restart test prograr.
The first itea is TVA SER Vol. 2, Supp 1
?-3 Prelininary Report, Revision 1
s l
I acceptable to the staff provided TVA identifies and evaluates all areas of the Unit 2 program reviews where Unit 1 differences exist.
In conjunction with the Unit 2 DBVP, functional test reouirements were identified by the DBVP and provided to the restart test program.
For Unit 1 the functional test requirements will be evaluated by the restart test program i
and the results provided to the DBVP for acceptance.
IVA's EA performed an l
t assessment of the management controls established for the conduct of Unit 1 l
j restart testing.
Based on its review. E' supported the DDW plan to review the l
results of the restart test prooram for ecceptability in order to satisfy the j
functional test requireTents.
The staff considers TVA's plan to transfer i
l responsibility for the review of testing to the restart test program l
acceptable.
Tre staff evaluation of the restart test prograu is further discussed in Section 4.9 of this supplement.
I Ouring a reeting with TVA on July 21,1988 (meeting sumary dated August 4 i
1988), the staff id?ntified two areas of the Unit 1 program description for l
l systm evaluations and corrective actions that we'e different from the Unit 2 i
pro; n n description.
TVA stated that these are.s were identical for both units ano committed to provide a revision to the Unit 1 program description to clarify these items.
In a subsequent conference call on September 16, 1988 TVA I
identified that one area, the use of the DNE review board, had been deleted l
from the Unit 1 program.
TVA provided justification for deleting the DNE j
review board in a letter to the NDC dated September 30, 1988.
TVA's letter l
I i
l states that the primary comments prnvided by the DNE review boaN were incorporated in the Unit I review process and procedures.
In additic1, TVA
[
i i
stated that the Unit 2 system evaluation reports which had been reviewed by the I
l DNE review board were used in the Unit 1 program.
The deletion of the ONE review board from the Unit 1 prooran was a procedural change which did not I
affect the scope of the Unit 1 DeVP. Based on TVA's incorporation of the I
I previous DNE review board corrents in the Unit 1 OBVP implementation and the l
results of the engineering assurance (EA) independent review of tiie OBVD l
f implementation, the staff finds that the deletion of the ONE review board from i
the Unit 1 DBVP does not result in an overall programatic change that would significantly alter the validity of the staff's conclusions on the adecuacy of f
the DSVP.
Thus, the staff considers this crocedural change to the 08VP for j
Unit I acceptable.
[
l TVA also' stated in its September 30, 1980 letter that the DRVP corrective i
j act m 1 reviews for the potential impact on equipment environmental c,ualifica-j i
tien (EO) program requirerents, which were associated with the DRVP in the y
1 l
$NPP, were actually performed as part of the plant design control process.
i This process was independent of CPvP.
By a telephone conference call on i
October 70, 198R ard letter dated October 26, 1988, TVA explained that these i
E0 reviews for both Units 1 and ? were conducted in the same manner.
The staff l
concludes that this environmental qualification review process is acceptable.
l 0.2.3 TVA Incerendent (versight As an integral part of the ORYP, TVA's Engineering Assurance IEA) aroup of the Divisine of Nuclear Engineerine eerforced an independent oversight review 6f the CBVP.
An in-depth description of the indepentent oversinht reybw process snd its results for Unit 1 is contained in TVA #erort EA.0R.003, "Engineering l
1
/ssurance Oversicht Recort, $nN Unit 1 DBVP," which w n submitted to NDC by j
TVA SER Vol. ?, Supp 1 2-4 Preliminary Peport, Devision 1
(
I
TVA also transmitted a supplemental TVA Report EA-0R-0035. "Supplemental Engineering Assurance Oversight Peview Report, SOM Unit 1 DBVP." by a letter dated September 19, IcP8.
The objectives of this independent review are the sare for Unit 1 as the objectives were for Unit 2.
The supplemental EA report addressed the open itens in the original EA oversight review report on the SBVP for Unit 1.
The original EA report is EA-0P-003.
The restart iters identified in the orioinal report have been completed in terms of DBVP Phase I.
One issue was determined to be a DBVP Phase II issue.
This issue and the verification in 08VP Phase II will be completed af ter the restart of (! nit 1.
Phase !!, as discussed above, is the post-restart portion of DBVP.
In an effort to gain further confidence in TVA's DBVP and in particular the independent oversicht activities of the EA organization relating to this 0BVP,
{,
the NDC staff has reviewed a sarple of five supporting back-up data packages.
l These documents address the results of the Engineering Assurance oversignt i
re<iew efforts including an independent evaluation and verification o' actions to torrect and close-out outstanding issues.
This review was in addition to the NRC inspections of TVA's DSVP (Reference Inspection Reports 50-527 I
328/S6-38. -45 and -55, 50-327, 328/87-14 and -31) which also assessed the effectiveness of the EA oversight effort.
The five data packages (Action itens C28 and C40 (Observation C6), E25 (Observation EE). E65 (Observation E3). E91 (Observation E14) and 0-12 l
(Observation 02)' consisted of results of analysis reviews and 'ierifications by the CNE and the EA organi:ations which support the action items / observations in the EA Reports EA-0R-001 and EA-00-0015.
As a result of resiewing each of the five packages, the staff found that the EA organization was actively and effectively involved in evaluating and reviewing the 04E DRVP efforts and in assessing and verifying the findinos, corrective actions and close-cut of these packages.
Further, the staff concludes that the restart open issues previously reported in EA-00-001 and -0015 have been adequately resolved and closed out.
Of particular interest has been the resolution of Action item 012 (Observation 21 This action iter resulted from a design deficiency identified in CAOP E6-03-6 2 ard rertains to:
A. desion criteria not being maintained; B. design I'
calculations not being maintaired; and C. plant configuration (as-built) desian doeurents beinq dif#erent from FSAD comnitrents.
In regard to Part A, Engineering Assurance has verified that all Seouoyah Unit 2 desien criteria are complete with the issuance of the restart design i
baste decurent (RDRD).
All post "estart design criteria develoorent is co~rM tted to be co?pleted by June 1,198c.
This latter 1:*.ue is a post restart I
ogen ite-i.
Ir regard to part P. Engineering Assurance ha: verified that ONE hat adequately reviewed all safety-related cal:ulations t ensure they are
]
technically adecuate and up-to-date and that a cross-refe< snee information l
systen has been established to maintain accountability of the status of l
j l
calculations acainst pertinent docuN nts, drawings and other calculatioas.
All 1
issues in Part B are, therefore, closed.
I i
the deficiency and corrective action associated with Part C was transferred to j
a separate ccrrective action rernrt SQ-CAR-R6 04-021.
The corrective action requires At the FSAP be updated ard verified to the current desion and j
TV/ JED Vol. ?, Suce 1
?-5 Prelininary Report, Pevision !
I
t l
~
as-built conditions. The schadule for completion of this action in is April p
1989. This issue remains open.
t The restart open issues previously reported in EA-0R-001 and -0015 iave been adequately resolved and cic ed cut.
The staff will verify that all comitted post-restart design criteria are corpleted by June 1989. The staff will also verify the completion of corree-l tive action report SQ-CAR-86-04-071 by April 1989.
In the orncess of complet-ing these actions, the resolution of design deficierties will involve consideration of unreviewed safety cuestions pursuant to 10 CFR 50.59 as approoriate.
Rased on the above and the staff's previous.eview of the program and its implementation en Unit 2, the staff concludes that the EA progran for Unit 1 is acceptable.
?.2.4 Conclusions TVA initiated the DBVP and EA independent oversight review as part of its i
effort to correct past design control deficiencies identified by emplevee concerns and design control reviews.
This program was extensively r(
$wed and inspected by the staff prior to Sequoyah Unit ? restart.
The staff concludes r
that the Unit 2 program with the modifications discussed above is sufficient to i
correct design control problems at Sequoyah Uni + 1.
The staff further concludes that the defined corrective actions, when completed, provide i
assurance that the plant will conforn to its licensing basis.
Based on the inspections en Unit 2, as discussed in Section 2.2 of NUREG-1
- Volume ? and the SSQE inspection on the Unit 1 CSS, the staff concludes that
[
Phase 1 of the DBVP has been sufficiently implemented for Ur.t I to restart.
The staff will review the transitional d gn control system durinq its review of the Phase Il portion of the 0;1VP.
S.3 Design Calculations Program j
Based en past findings N TVA and the NRC reviews, it was determined that there was inadequate donwntation or the calculations supporting the design
[
basis for TVA's nuclear plants.
Calculations were determined to be missing.
incomplete, or outdated. TVA's engineering disciplines (nuclear, rechanical, j
civil, and electrical) have cach developed programs to resolve these problems.
These efforts include (1) identifyino essential calculations; (?) verifying the existence of, or regenerating, essential calculations; (3) ensuring the tech-
[
nical adequacy of these calculations: and (4) ensuring tha calculations are current.
Essential cciculations are those which oualify existing plant systems or features whose failure could (1) result in a loss of integrity of the reactor coolant system, (2) result in the loss of ability to place the plant in a safe 1
shutdown condition, or (3) result in a release of radioactivity off site in excess of a significant traction of the 10 CFR 1 LOO quidelines.
The calculations review efforts for the engineering disciplines are discussed t
in detail in Section 2.3 of NUDEG-123?, Volume 2.
The following sections I
I TVA SER Vcl. 2. Supp 1 2-6 cre;iminary Report, Revision 1 f
h
0 O
discuss and evaluate the identified differences in the TVA design calculation procram for Unit 1.
2.3.1 Nuclear and Mechanical Calculations The nuclear and rechanical calculation review program for Unit ? is described l
in Section 111.4.2 and 111.4.3 of tne SNPP. TVA did not identify any differences in the programs fer Units 1 and 2.
Based on Section 2.3.1 of NUREG-1232, Volune 2, the staf f corcludes that the nuclear and mechanical engi-neering calculation review effort has been adeouately defined and implemented to identify the necessary essential calculations for the operation of Sequoyah; that the technical adequacy of the calculations has been adeouately demonstrated; and that necessary corrective actions are being scheduled in accordance with the staff approved restart criteria.
Therefore, the staff finds the TVA actions for resolution of nuclear and mechanical calculations concerns acceptable for the restart of Unit 1.
4 2.3.2 Civil Calculations 2.3.2.1 Description and Scope l
The civil calculation review program for Sequoyah Unit 2 is described in l
Section !!!.a.4 of Revision 1 to the SNPP.
The scope of the civil calculation review plan was describrd in greater detail in TVA submittals to the NRC on July 21,1987 (Gridley) and '.uqust ?l,1987 (Gridley). The staff safety evaluation of the SMPP through Revision 2 is contained in NUREG-1?32. Volume P.
2.3.2.?
Unit I and Unit. 2 Progran Differences l
TVA submitted revision ? to the SNPP to the NRC on May 9, 1988.
Part ? of Fevision 3 orevided a description of Sequoyah Unit I start-up programs that are different from the Unit 2 programs.
The dequoyah civil engineering program was identified as a program area where differences exist between Unit 1 and Unit 2.
The difference identified was that TVA would submit a final report on Inspec-tion and Enforcement (IF)Bulletin 79-14 for Unit 1.
IE Bulletin 79-14 had been conpleted by TVA on Sequoyah Unit 2; however, the bulletin was, still considered open by TVA en Unit 1.
TVA originally identified concerns with IE Bulletin 79-14 in Section 111.15.1 of the SNPP under the heading of miscellaneous civil engineering issues.
The staff's evaluation of this tupic was contained in Section 2.7 of NUREG-1232 Volume 2.
TVA had also covePeti the topic of piping and supports in the civil calculation program in Section 111.4.4 of the SNPP.
The staff evaluation of the civil calculation program was contained in Section 2.3.2 of NUREG-1232.
Vo'.ume 2.
For Secuoyah Unit 1. TVA combined the discussions of IE Bulletir.
79-14 and the pipe support cal:ulation effort under the heading of civil engineering program in Pevis1 n 3 to the SNPP.
l The Unit I civil ennineering program was iritially described in TVA's March 31, 19CR submittal on the Unit ! restart plan.
The TVA submittal ic'entified thit the Unit 1 civil engineerinq program was essentially the same as the Unit P proorm with the exception that a final recor' would be submitted on IE Rulletin 79-14 for Unit 1.
The Unit 1 IE Bulletin 70-14 implementation had been addressed by an employen concern re.) ort (EN 21202).
EN 21202 ic'ertified TVA SER Vol. 2, Suco 1 2-7 Preliminary Report, Kevision 1
c>.
that discrepancies existed with previous TVA pipe support inspections On Unit 1 and that TVA initiated a pipe support enhancement program as a corrective I
action.
J In a meetino with the staff on April 14, 1988 (Peeting Sumrary dated May 4, 1998), TVA presented additional details on the program scope of the IE Pulletin 79-14 and pipe support calculation efforts.
In addition TVA stated it wculd use the same criteria for determining required restart modifications for Unit 1 that was used for Unit 2.
TVA submitted the results of this IE Bulletin 79-14 f
evaluation in a letter dated August 4, 1988 (Gridley),
P IE Bulletin 79-14 requires that licensees verify that the seismic analysis of l
piping applies to the actual configuration of the plant.
As a result of concerns raised by the original NRC inspectiors of TVA's IE Bulletin 79-14 program at Seoueyah Unit 1, TVA initiated a sampling inspection program.
As part of this program, 20 piping isometrics inside the containment were inspected usino Special Paintenance Instruction MI-6.17.
In December 1985 and February 1986 TVA's quality assurance staff identified weaknesses in the MI-6.17 wal(downs and, as a result, two Corrective Action Peports (CARS) were issued.
In response to the CARS, TVA performed additional inspections of the 80 piping isometrics to Special Paintenance Instruction SMI-1-317-24 (SMI-24).
This program was reviewed by the employee concerns program (EN 21702).
The employee concerns report found the Seouoyah Unit 1 program had been substantially improved to correct past deficiencies and concluded that no l
further corrective action was required.
In its August 4, 19ER submittal. TVA identified a more comprehensive program fer the evaluation of rigorously analyzed piping at Seawoyah Unit 1.
This program ircluded the evaluation nf all open items that had been identified from previous programs against the piping analysis and support desions, and the upgrading of the suppnet calculations to the rew design criteria 50M-0C-V-?4.2.
]
The scope included IF? pioing analyses and approximately 2900 pioing supports.
TVA develnped Specisi t'aintenance Instruction SPI-0-317-69 (SMI-69) to control the collection of ecditional as-built data for the Unit I rigorously analyzed j
piping.
In its submittal, TVA identifiec that approximately one-third of the t
pipe supports and all but six piping isometrics were inspected to SMI-64 j
Sul-69 contained requirements for as-built dimensioning of pipinq that had not been obtained by some of TVA's previous walkdowns.
4 TVA used the criteria in SQN-DC-V-?4.? 'a evaluate suoports and CE8-CI-21.89 to identify the required restart modifications. TVA identified that 373 modifications were required to meet the criteria in SON-DC-V-24.2 and 179 restart redifications were required to meet the criteria in CEB-CI-21.E9.
TVA's submittal also identified that the closure of IE Bull? tin 79-14 for Unit 1 rigorously analy:ed piping in common plant areas was based on TVA's oricinal inspections supplemented by the adcitional SMI-24 inspections.
TVA also identified that the ccmmon area supports had been previously evaluated by the Unit 2 calculation program which included a functional verification inspection per CEB-CI ?1.83, Ir eddition to the rigorously analyzed pipina, TVA ider tified that alternately analy:ed piping within the scope of IE Bulletin 79-14 had been addressed by the Seouoyah alternate analysis program. This program had been previously described in Section !!!.5.0 of the SNpP.
TVA SER Vol. ?, Supp 1 2-8 Preliminary Peport, Revisinn 1
o 2.3.2.3 Evaluation TVA's civil calculation program for Sequoyah Unit 2 :s described in Section 111.4.4 of the SNPP involved the identification of essential calculations, verification of retrievability_ regeneration of r.issing essential calculatinns and verification nf the technical adequacy of existing calculations. The Unit 2 civil calculation progran was extensively audited by NRC calculation program inspections and the NRC integrated design inspections. The staff's evaluation of the Unit 2 progran es contained in NUREG-1232, Volume 2.
During the review of civil ergineering calculatioas, TVA determined that a large number r" rigorously analyzed pipe support calculations were not retrievable.
The Sequoyah Unit 1 prooram combines the regeneration of the pipe support calculations with the resolution of IE Bulletin 79-14 The Unit 1 pro-gran scope for rigorously analyzed pipe supports as described by TVA is mnre corprehensive than the Unit 2 proaram since additional detailed walkdowns were perforced for Unit 1.
Based on the staff's previous review of the Unit 2 crogram, the sare program for Unit I with the addition of a final report on IE Bulletin 79-14 is acceptable.
TVA identified that it did rot use the upgraded (SMI-69) IE-Bulletin 79-14 walkdowns for common areas where the supports had been previously evaluated by the Unit 2 program or for piping covered by the alternate analysis program.
The supports in Unit 1 connon areas had been functionally verified during the Unit 2 pipe suppret calculation effort.
in addition, the NRC's integrated design inspection of the essential raw cooling water system had performed as built inspections of the connon area for Unit 2.
TVA's alternate analysis program procedures for piping inspections had been previously reviewed and the staff's evaluation is contained in Seation 2.4 of NUCEG-1232, Volure 2.
Based on the staff's previous acceptPnce of the Unit 2 piDinq and support evaluations and the Unit 2 alternate analysis program, the sare programs applied to Unit 1 are also acceptable.
TVA identi'ied that it used the sare criteria for Unit I to evaluate rigorously analyzed pipe supports that oad been used for Unit 2.
The staff's evaluation ef these criteria is contained in Section 2.3.2 of NUDEG-123?, Volure 2.
The staff evaluation of SON-0C-V-24.2 determined that the criteria were acceptable for restart, and that the staf' rculd be performing additional evaluations of the standard component supports as a post-restart effert.
The staff evaluatien of CEB-CI-21.89 approved the criteria with certain restrictices in a letter to TVA dated February 23, 1988.
In addition, tb; staff identified several concerns with TVA's implementation of the pipe support criteria for Unit 2 in Inspection Report 50-327, ?28/88-12.
TVA's resolution of these inspection items is also applicable to Unit 1.
All supports must satisfy the restart criteria es accepted by the staff before the restart of Sequoyah Unit 1; the present schedule for compliance with the long-term criteria is the end of cycle 4 for Unit 1 (August 4, 1988 submittal).
TVA's use of the same criteria, as i
accepted by the staff, for Unit I that was used for Unit 2 is acceptable.
TVA's irplerentatien of IE Bulletin 79-14 for Unit I has been reviewed by several t.PC instections.
The '1C's original inspecticns identified concerns with TVA's IE Bulletin 79-14 pregram on Unit 1.
TVA's subsequent corrective actions were reviewed in Inspection Rcrort 50-327, 323/85-49.
The inspection report identified several pipe sucport discrepancies and TVA was cited with a TVA SER Vol. 2, Supp 1 2-9 Prelininary Peport, pevision 1 i
violation.
Inspection Report 50-327, 32P/86-16 identified additional discrep-ancies in a follow-up inspection.
Inspection Report 50-327, 328/86-55 closed the violation from inspection Report 85-49.
The inspection report also con-tained a review of the work being performed under Special Maintenance Instruc-tion SMI-1-317-74. The inspection report did not identify any violations or dev i a tior.s.
The staff also perforced a Safety System Quality Evaluation inspection of the Unit 1 Containnent Spray System.
This inspection included a sample review of pipe support calculations and pipe support as-built configura-tion.
In addition to these inspections the NRC conducted a special as built inspection of the essential raw cooling water system. Inspection Report 50-327, 328/P.7-52 which covered the Unit I and Unit 2 common plant piping.
Based on inspections performed by the staff of IE Bulletin 79-14, review of TVA's implerentation of the bulletin by the employee concerns program and TVA's odditional insoections usino SMI-69, the staff concludes that TVA's program to address IE Putletin 79-14 for Unit 1 is adequate to verify the as-built piping configuratien.
The staff plans to document the final close out of the bulletin t
for Sequoyah Unit 1 af ter the restart of l' nit 1.
2.3.?.4 Cot.a.l u s i o ns TVA initiated a civil calcelation procram to assess the adequacy of existing civi'. calculations and regenerate missing calculations.
This program was extensively reviewed and inspected by the staff prior to Sequoyah Unit ?
l; restart.
The staff cercludes that the sare pronram with the additional as-built verification perforred for Sequoyah Unit 1 is acceptable for restart.
l 3
2.3.3 Electrical Calculattens
?.3.3.1 Introduction The electrical calculation review program is described in Section III.4.1 of i
the SNPF.
The TVA electrical calculation review procram is divided into two
[
phases.
Phate ! for each unit is to be completed before plant restart of that t
unit ard covers the essential minimum set electrical calculations needed for restart.
Phsse II covers the remaining electrical calculations and will be completed after plant restart. The staff notes that TVA has committed to expand and formali:e its calculation control program over the long-term to cover all r
calculations, rot just those identified as the essential minirum set.
The staf f relies on this commitmert as the most effective treans tn assure that i
TVA's electrical calculations required to assure safety are maintained in the l
acceptable condition that the present program has established.
The staff evaluated the restart electrical desion calculations for tinit ? in
(
Section 2.3.3 of NUDEG-123?, Volume 2.
The staff concluded that there was l
1 reasonable reassurance that electrical systers are adecuate for the safe restart and operation of Unit 2.
The staff's conclusion on the general
[
adequacy of the electrical calculation program for Unit 2 did not extend to J
Lnit I restart because of the folleving reasons:
l (1) A number of calculations do rot assume two unit operation and require upgradico to suoport Unit 1 operation.
j TVA SER Vol. 2 Supp 1 2-10 Preliminary Pecort, Revision 1 l
c> -. - - - ---- - - - - - - -
O e
o (2) A number of deficiencies identified as required for restart have been completed for Unit 2 but not for Unit 1.
TVA provided information on the restart electrical design calculations for Unit 1 in its submittals dated August 4 and 11, and Septenber 15, 1988.
There are a nurber of deficiencies designated to be corrected after restart and there are a number of long-term programs TVA has connitted to undertake after restart.
These are listed in the various documents cited in Section 2.3.3.1 of NURE3-1232, Volume 2.
Expeditious ccepletion of these long term cornitments was assumed in the staff's evaluation of the adequacy of the TVA electrical calculations program for l' nit 2 and Unit 1.
l 2.3.3.2 Evaluation The staff's evaluation of the calculations for Unit 1 operations will follow the format of the evaluation for Unit ? included in Section 2.3.3.2 of NUREG-1?32. Volume 2.
2.3.3.2.1 Auxiliary Power Systers (APS)
(1) APS Load Systems and Voltage Calculations The staff's evaluation of the Auxiliary Power Systen (APS) load analysis for l' nit 2 is decurented in Section 2.3.3.2.1 and 2.3,4 of NUPEo-1232 Voluee 2.
This inad analysis was performed on the RADIAL computer code which is docurented in Section 2.3.3.2.1 of NUDEG-1232, Volure ?.
On August 4, 1988 TVA submitted an Electrical Calculations Plan to address calculations ree,uired for the Unit I restart, and on September 15,19RR TVA submitted the ref uits of the APS analysis for the two unit operation.
- However, this new analysis was perforred on a new computer code developed by Sargent and Lundy Engineers.
This ccie is called Electrical Load Ponitoring System for Alternating Current (ELMSAC).
TVA's decision to utilize ELMSAC for the Unit 1 restart was based upcn TVA's comitrent to the following:
o Standardized calculation methods.
Use industry proven ouality assurance (QA) sof tware to:
o Produce technically superior calculations Readily assessable I
Easily raintained The ELMSAC load flow computer code cerforms APS loadire, voltage and fault current analysis.
The calculation for Unit I restart is EER SQN-PS-T106 002, Revision 0, (ELPSAd. T*e input data to ELPSAC is frem a OA load and cable computer data base prooram the "TVA Electrical Auxiliary System", (TELAS).
TVA in their submittal letter of September 15. 198P,provided a confirratory analysis to demonstrate that (11 there is no appreciable difference from previously calculated voltages and (2) EtySAC results are within 3 percert of reasured salues established by the guidelines in Brarch Technical Position TVA SER Vol. ?, Supp 1
?-11 Preliminary Report, Pcvisica 1
I l
1 l
pBS-1. This confirmatory analysis used the Test I configuration listed in the j
table on Page 2-48 of NUDEG-1232. Volume 2.
This test configuration is based l
4 on TVA data of July 12, 1980.
l Load Analysis I
L l
The worst case APS loading was analyzed during full load operation. The elec-t trical power for full load operation would be supplied fron the unit station l
service transformers and the worst case fault currents would occur when the APS is supplied from these transformers because of the fault current available from
~
both the main generator and the offsite grid.
The analysis has demonstrated that the steady state maxir.um loads are within the APS equiprent ratings.
l 4
L Fault Current Analysis
[
1 r
The analysis bas demonstrated that the traasient available symetrical fault l
current is less than the Class 1E 6.9 KV Shut-Down Peard breaker interrupting l
rating and the asymetrical fault current h less than the breaker momentary
(
(close and latch) rating, when the emergency diesel generators (EDG) are not paralleled with the grid for testing.
When t3e EDGs are being tested. I hour j
each nonth, the available symetrical fault current exceeds the Shut-Ocwn Peard jl
}
breaker interrupting rating by 2%. However, sicce the 2% over the rating only t
l occurs during EDG testing, there is minimal effett in the capability of this
!I j
system to perform its safety function during emergencies.
t l
The analysis has demonstrated that the available symmetrical fault current for j
the Unit Board supply breaker is greater than the interrupting rating but ll within the one tire capebility test and for the feeder breakers the symetrical i,
l feult current is greater than the one time test capability. However, the I(
4 l{
)
available asymetrical fault current is less than all unit board nomentary ratings.
J The above results regarding available fault currents and Unit Board breaker r
j interrupting ratings and capability is similar to thnse results evaluated for j
Unit 2 restart.
The staff's evaluation is docu' rented in Section 2.3.3.2.5 of NUREG-1232, Volure 2.
TVA made a coccitment, in a letter dated August 10, 1987, to submit a j
corrective action plan to resolve the problem of unit board available fault i
i current before June 30, 1989.
This corrective action plan will assure that all l
circuit breakers will always operate within their service capability as defined by appropriate standards and verified by test or manufacturer's guarantee <-
l s
j Voltage Analysis
]
The TVA's APS voltace analysis was perforced for two unit operation, during l
normal and accident conditions, with electrical off-site power supplied from f
either the generator or the 161 KV switchyard.
During a unit trip, except for d
l a generator fault, the APS continues to receive power from the generator for ar additional 30 seconds and then power is transferred from the Unit Station Service Transforcers (USST) to the 161 KV Cornen Station Service Transformers i
{
However TVA's analysis conservatively considered tire zero as the tire i
(C5ST).
of transfer of the Unit Boards to CSST and includes ell initial starting i
(
accident loads een the CSST mit6;e taps are set at '
5'- (159 KV1. Roth tire
)
}
TVA SER Vol. 2. Supp 1 2-12 Prelininary Report, Revision !
i l
i zero and 5 seconds were analy:ed with both units tripping, one unit having an accident and the other at full load rejection.
1 The analysis indicated that the unit with a safety injection and containrent isolation phase B accident condition will have the following Shut-Down Roard voltages:
Shut-Down Board 1A-A 18 8 2A A 28-B Bus Voltage tire zero 6130 6275 6206 6192 time 5 sec 6648 6694 6623 665i The ELPSAC calculations of the APS for all plant operatino crnditiyns during two unit cperation, indicate that adequate voltage is available tr all electrical equipment, during transient and steady state conditiers, with the minimum source voltages.
Conclusion The staff cercludes that the ELP. SAC calculations for the APS are acceptable for two unit cperation during all plant operating conditions.
(2) Class IE MCC Control Circuit and Cable length Calculations The staff's evaluation of the 480V APS voltage for Unit 2 is included in Sect:en 2.3.3.2.1(3) of NUREG-1232, Volume 2.
The staff concluded that the Class 1E rotor control center (PCC) contactor and associated control devices will function properly during low voltage conditions associated with either the offsite grid or tne eeergency diesel generators (EDM.
The evaluation in Section 2.3.3.2.1(3) of NUREG-1232 Volume 2 is limited to Unit 2 operation sirre the emergency diesel generator load analysis, on which the analysis of PCC control performance during sequencing depends, assures Unit I was in cold shutdown.
The staff, therefore, concluded that the load analysis and the voltage from it will require recalculation for two-unit operation.
The APOV ac ADS vohage is related to its associated 6.9 KV ac APS.
The staff's evaluation of the 6.9 KV ac APS 'or Unit 1 and Unit 2 operation is included in Section 2.3.3.2.1(1) above.
The worst case transient voltage associated with the offsite source occurs af ter a unit trip concurrent with an accident condition.
This low voltage is less than the Class 1E APS degraded grid protective relays setpoint; however, the APS voltage recovers to above the protective relays setpoint after 5 seconds and blocks the 10 seennd time delay from causing systen separation.
The degraJed grid relays associated with both Unit I and Unit 2 provide protection to both the 6.9 KV and 480 V APS free transient or steady state low voltage condition when the pcwer is from o#fsite.
The APS voltace analysis, when supplied power from the onsite EDG, was i
evaluated by the staff in Section 2.3.3.2.lf4) of NUPEG-1232, Volume 2.
The basis for the staff's evaluation was TVA analysis of the worst case transient TCG voltages.
This worst case transient voltace was determined by TVA fron testina conducted on all four E N's in July 1987.
TVA used these test results l'
which were adjusted for the raxirum load conditien to deterrine the transient voltage response during the EDG loading sequence. Using inputs from this TVA SER Vol. 2, Supp 1 2-13 Preliminary Feoort, Pevision 1
6 I
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transient voltage response, TVA submitted a revised analysis on March 10, 1988 l
for the 480V APS voltage. The analysis evaluated the Unit 2 worst case actual FCC contactor circuit control wire length and parallel load combination concurrent with the lowest voltage, resulting from 6.6 KV notor starting, during EDG loading sequence. TVA also evaluated the worst case control circuit maximum current caused by low voltage for control circuit fuse protection i
coordination.
The staff finds the worst case used by TVA for Unit 2 analysis 1
also applies to Unit 1.
Therefore, the conclusion that the Class IE MCC j
contactor and associated control devices can perform their safety function during transient voltage conditions is applicable for both Unit 1 and Unit 2.
(3) Emergency Diesel Generater (EDG) Lead Analysis The calculations which repret nt the performance of the EDG for all plant conditions for Unit 2 operation were subn'itted to the NRC on February 79 and j
March 3 and 10, ins 8.
The staff review of these calculations was documented in I
Section 2.2.3.2.1(41 of NUREG-122?, Volume 2.
The calculations were based on f
EDG 28 B which TVA had determined was the most heavily load EDG, l
l On August 31, 198A, TVA submitted the results of the revised calculations for
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I two urit operation. The revised calculation SQN-E3-002, Revision 10, includes all the ECGS in all operating roces for both Units.
1 The EDG 28-P leading was not increased with both Unit I and Unit 2 in operation tecause no additional loads were applied with the restart of Unit 1.
Further i
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review of the EUG 2B-B Inading has demonstrated an overall reduction in load demand for this EDG.
The additional review has deroostrated that some loats i
were conservatively assured at a higher load derand than their actual require.
2 cent and, by letter dated September 28, 1988, TVA submitted a listing of the l
differences in inading.
The staff was able to reconcile the loading differences of EDG 2b 8 between i
Revision 7 and 10 of calculation 50N-E3-002 and found them acceptable.
I The staff's evaluation of EOG 28-8 transient and steady state loading for tro unit operatien indicates that for all 'lant operating conditions the loading is j
telow the EDG's rating.
The EDG 1A-A,18-B, 2A-A loading requirer ents are less than EDG 2B-B and are, therefore, also acceptable since all the EDG's have tne same capacity rating, i
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l-The staff concludes that the present EDG load analysis is acceptable to be used
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as the basis for determining the EDG Inads for all plant operating conditions.
Further, there is sufficient margin between the worst case trtnsient and steady l
state loading of the EOGs and the manu'acture's ratings to permit TVA to crerate t' nit 1 or both units at full power.
j 1
2.3.3.2.2 Control power Systems l
(11 1?5-Volt DC Vital Instru-ent Fower System Voltace Calculations I
The staff's evaluatier of the !?S volt de vital pcwer system voltace is ircluded in Section *.2.3.2.2 of NUREG-1232. Yelure 7 On the basis of its i
evaluation of the 12' volt de voltage calculatinns, the '.taff concludes that i
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TVA SER Vol. ?, Supp 1 2-la preliminary Report Fevision 1 l
l approximately 30 percent of the safety-related employee concerns pertain to various aspects of the TVA welding program.
By letter dated January 17, 1986. TVA formally submitted its program plan to address erployee concerns related to welding for staff review.
TVA formulated its program to evaluate the welding program at each TVA nuclear power plant in two separate work phases. The Phase I effort consisted of a review of the written TVA welding program (desian doc. rents, policies, and precedures) to ensure that the welding program correctly reflects TVA's licensing comittents The Phase 11 effort consisted of actual re-and regulatory requirenents.
inspection of selected welds and the inspection results were used to evaluate the implementation of the written welding program, in both phases of the program plan, TVA was to identify and catecorize any deficiencies in the existing program, correct the problens, and implement chant,es to prevent recurrence.
The staff has evaluated the TVA welding program for Sequoyah in Section 3.5 of NUR2G-1232. Volure 2.
TVA did not identify any differences in the Unit 1 prngran from that for Unit 2.
TVA has ccmitted to standardize among all nuclear plant sites the means of naintaintnc velder qualifications.
This will be accomplished by having the QC inspector or the welder foreman initial the welder's rod issue slip indicating that the welder has raintained qualification by the use of the process.
Section III.3 of TVA's revised SNPP provides an action plan that will impreve the design centrol progran for Sequoyah when iriplemented. This plan includes the reconciliation of "as constructed" and "as designed" drawings to achieve 5 single set cf plant drawings. This plan should address the irregularities identified above to ensure that the weld' and weldino requirements stated on the "as designed" drawings match the installed hardware.
On the basis of its evaluation in NUREG-1232. Volume 2, the staff concluded tFe following for Units 1 erd 2:
(1) During constructien of both Sequoyah units, TVA's implementatien of the OA/0C program in the area of welding, while gererally effective, was ineffective in certain instances.
For example, a significant number of deficient welds were found that required engineering calculations to demonstrate their suitability for service.
These calculations should have been performed during construction.
In addition, discrepancies between the design drawings and the actual hardware installed were identified.
Notwithstanding these findings, the fact that no welds required repair to reet desion code requirerents indicates an overall effective implementa-tien o' the QA/CC procram in the area of welding.
(2) The effectiveness of TVA's process for CC inspector training and qualification /certificaticn to visually inspect welds during plant con-struction and af ter operation is questionable.
The welainq deficiencies discussed above should have been detected and corrective actions should have been taker.
TVA SER Vol. 2. Supp 1 3-8 Pre 14ninary Report, Devision 1
(3)
In spite rf the deficiencies found in the implerrentation of the QA/QC progran fe welding activities, inclucing some that were of a programatic nature, the staff finds that these deficitreies have not significantly affected the suitability for service of plant hardware.
(4) With the exception of QC inspectors' training and qualification /certif-ication, the staff finds that other essential elements (i.e., welding procedures, welder qualificatien and training, weld cesign and configu-ration, and filler retal control) of a sound welding program were functioning and the resultant hardware is suitable for service.
Therefore, the staff concludes that TVA's welding re-evaluation program has teen carried cut adequately and that TVA has demonstrated that the hardware as cor.structed is suitable for service, that is, the design loao limits fer welded connections have been ret. The staff further concludes that restart of both Sequoyah units will not endanger the public health and safety.
For an overall improvecent of the welding program at Sequoyah, the staff endorses the following TVA proposed changes in its internal control documents conta1ned in the SNPP:
(1) Certtining the requirerents of General Construction Document G-29 and Process Specification iU3F2 into a single docurrent.
(2)
Replacle'g the general construction specification for each unit with sfeCific sfeulfiCaticns.
DJ Mainteir;irg indirect quality control of fit up inspection by monitoring processes as provided in 10 CFR 50, Appendix 5 (1) by having the welder and his forerran doCurent that fit bp is suitable for the QC inspector to verify weld size during final inspection and (2) by having the QC inspector selectively inspect a sampit of fit ups to verify this cocurentatior.
(4) Consolidate inspector training and certification into one program under tre control of a certified Level !!! h0E examiner.
(5) Provide training er orientation to engineers, designers, technical super.
visors, are engineering ranagers of the content and use of the internal control docureats.
(6)
Stancardize the process of maintaining welder's certification by having the (C inspector or welder foreman initial the rod issue slip indicating that tre specific welcer has used the process.
In a letter dated January 30, 1967 TVA committed to an ausrented and acceler-ated inservice inspection as recomencec by NRC staff.
The inspection program will include the eierents listed telo.<.
(1) A ICD-percent examinaticn of the ASME Class 1 and 2 piping fielo welds will be completed in the first 109 ear in-service interval.
Those welds that rer.ain to be exa.1ned will be scneculed for examinaticn in the next two consecutive refueling cutages folicwing the sutnittal of the revised plan anc the restart ef any unit.
T M SER Vol. 2. Supp 1 3-9 Freliminary Report, Eevision 1
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(2) A 100-percent exanination of the ASME Class 1 and ? pipe support field welds will be completed in the first 10-year in-service interval.
Those welds that remain to be examined will be scheduled for examination in the next two consecutive refueling outages following the submittal of the i
2 i;
revised plan and the restart of any unit.
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(3) Major component support welds riade in the field on the reactor vessel, stean generator, pressurizer, and reactor coolant pumps that have been d
identified to be examined in the first 10-year program will be completed.
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1 Those welds that remain to be exar.ined will be scheduled for exanination in the next two censecutive refueling cutages following the submittal of l
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- he revised program and the restart of any unit.
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(4) Where possible, the percentage of welds examined during the program will Fe maintained as required by the code in the Tables IW8 2412-1 and
!WC-2412-1 (Inspection Program B).
Note that the required percentages may l
not be ret for all categories of specific systems, or item numbers, i
l because certain systers contain a large number of socket welds that are field welds and the majority of pipe support welds are also field welds.
j Yhere conflicts arise with the percentage requirements, the revised j
augrented/ accelerated program will identify specific requirements for relief.
Credit for progran examination will be taken for all examinations perforced and l
i ro additional Class 1 and 2 field welds will have to be re-examined in the
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l remaining tire of the first 10-year interval, with the exception of the Code required acditional examinations resultino from unacceptable indicaticns in the l
initial or recuired successive examinations.
Future 10-year interval examina-l l
tions will follow their original schedule and will not be required to meet the accelerated prngram.
Because the first refueling cutane is scheduled to occur aproxinately 4 to 6 l
hs af ter restart of Unit 2, the short duration of the (wrating time may rn j
not provide the needed tire for the increased planning and scheduling, staffing j
and craf t support required to perfem the increased inspections of items 1, 2, t
j and 2 above.
In this case, the implementation of any accelerated program weuld be deferred to the second and third outages following restart of Unit 2.
Scheduling parts of the actual inservice inspection for Unit 2 for tre second
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and third refueling outage after restart rather than the first and sevind j
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refueling cutage after restart is acceptable to staff.
a Further, the staff reconrends that TVA consider the following:
(1) using industry-cenerated standards where possible, particularly using
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Arerican Heldino Society (AWS) standards for certifying the AWS scope weld inspectors; j
(2) amendina relevant FSM sectiens to reflect changes in commitments and to femalire the intent as stated above; and I
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(3) trainine personnel in the application of the standarce adopted.
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TVA SER Vol. 2. Supp 1 3-10 Prelirinary Report.]evision 1
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3.f Containrent Isolation l
3.6.1 Containment Isolation Systan Cesign General Desiqn Criteria (GrC) 54 through 57 of Appendix A to 10 CFR 50 cor.eain NRC design requirerents for isolation of piping systems penetrating l
4 containrunt.
In particular, GDC 54 contains general provisions for leak detection, redundancy, and reliability.
GCC 55 requires each line that is part of the reactor coolant pressure boundary (RCPB) and that penetrates the j
containment to have isolation valves as listed below, unles; it can be 7
demonstrated that the provisions for a specific class of lines are acceptable on sore other defined basis.
L The staff identified apparent discrepancies in system compliance with contain-rrent isolation requirements during an inspection conducted at Sequoyah on i
March 3-14. 1986.
Specifically, inspection Reoort 50-327/328 86-20 documents five containment penetrations of the chemical
' volume control systen (CVCS) that did not appear to meet 10 CFR 50. AprendL ' GDC for contairment I
isolation.
TVA submitted, by letters dated January 23 and February 3, 1987, requests for l
exemption to the reouirerents of 10 CFR 50 GDC 55 and 56 for the penetrations l
l in questiun.
Supplemental information to these requests was submitted by TVA l
en April 8, 1987.
1 l
in its evaluation in Section 3.5.1 of NUREG-1732. Voluce 2, the staff discusses i
each penetration not reeting the explicit GPC requirerents as identified by TVA j
in Table 2.? of its submittal of January 2, 1987.
Tte discussion includes the l
everption granted by the stef f t' have the Sequoyah certaineent isolation i
4 design in confornance with GDC 54 through 57 of Appendix A to 10 CFR 50.
l The staff normally requires that all power-operated containment isolation valves have ocsition indication in the rain control room. TVA recently l
confirred that with the exception of 22 valves, all other power operate:1 valves i
l have position indication in the main control room.
Position indication for the 22 exceptions are provided in either the auxiliary bJilding or the hot sample i
Installation of position indication for the 22 containrent isolation valves in the rain control room is planned for the cycle 4 refueline outage, l
rnom.
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1 Cn the basis of its evaluation the staff concludes that, with the approved exerstions, the Sequoyah containment isolation design is in accordance with i
Appendix A to 10 CFR 50 and, therefore, it is acceptable.
3.6.2 Containment Isolation Leakage Testing Program l
1 As discussed above, Inspection Deport 50-3D/37E P6-20 contained open iters i
l regardinc the containment isolation design for certain containment penetra-i tions.
By letter dated September ?A. 1986, and January 2, 1987, TVA proposed l
to partly resolve these open iters by redesicnating certain valves as con-I tairrent isolation valves.
The acceptability of these proposals is addressed i
above.
TVA also has evaluated the redesicnated contairment isolation valves in i
regarc to the requirements of Apperdix J to 10 CFR 50 concerning local leakace i
rate testing.
The staff's review of this issue is dncur.ented in Section 3.6.'
c' NU EG-1232. Volu~e ?.
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1S[!['2'5*I
By letters dated July l' and August 8. 1988. TVA requested an exerption to J for leak rat testing the ched valves of the.ontainment Spray r
Appendiy System (CSS) and Residu 1 Heat Rermal Spray System (RFRSS) for Sequoyah c
l' nits 1 and 2, which ar containment isolation valves.
Definition H in Acrendix J provides the criteria for a containment isolation valve to be type C tested. The CSS and RF SS reet criterion 3 where the system is required to operate interrittently Jnder postaccident conditions.
The Sequoyah desian for i
the CSS and PHRSS religd on a check valve inside containment and a manual valve
'utside containment to satisfy the requirements of GCC with seal water system i
- 56. This design is suc' that TVA has stated that it is impractical to test the inboard check valves; therefore, TVA has requested an exemption from the TVA proposed Appendix J 1eak rate testing requirererts for these check valves.
j to rely on the remote rarual valve and seal water system and the closed CSS and 2
FFRSS outside of containment as the basis for not Appendix J leak rate testing the check valves. This exemption was issued by the staff in its letter dated September 22, 19FF.
i Based on the evaluation in NU:EG-1232, Volume 2 the staff finds that with the above exe*ption, the proposed local leakar,e rate testing program for penetra-tiens is in accordance with the requirements of Appendix J to 10 CFR 60, and is, therefore, acceptable for Units 1 and 2.
3.6.3 Contain~ent Leakage Testing The staff requested that TVA perform a visual inspection of the Sequoyah Unit 1 containrent before restart of the unit. The purpose of the visual inspection r
is to demonstrate that the containment was not accidertally daraged during the extended cutage since the last integrated leak rate test o' the containment in I
Cecember 1985 i
TVA has reported that since the plant shutdown approxirately 3 years aco, there has been ro additional loading on the containment for that pe"iod.
Although there has been no centainment loading during the shutdcwn period, there have i
been rajor rodifications performed inside of containrent which increase the likelihu d of accidental damage to the containment.
Actual experience with i
other utilities has deronstrated that containment liners have been accidentally l
deraged during shutdown intervals ruch shorter than 3 years.
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TVA has conducted an audit of the work orders performed during the shutdown l
i interval to demonstrate that proper controls were in effect to prevent damage to the containment.
However, such audits would only reveal accidental damare to the containrent if it was reported. Unreported damage to the centainrent would not be identified by such audit.
A visual inspection cf the containtrent should identify any accidental damage.
l both reported and unreported, that ray have occurred during the 3 year shutdowr interval.
By the letter dated August 19, 1988, the licensee committed to perferr a visual inspecticn of the containcent under Surveillance Instruction i
(SI) 054, prior to restart.
If the inspectien demonstrates that tfe l
contain ent was damaged, the contairrent would need repairing and these repairs i
would need testing for leakage before restart, l
l r
- c. Supp 1 3-12 Preliminary Report. Revision 1 f
3.7 Containrent Coatinos TVA identified deficiencies found during a review of maintenance records relating to its programs for coatinos inside containment. These deficiencies are listed in Section 3.7 of NUREG-1232, Volure 2.
TVA did not identify any differences in the Unit 1 program from that for Unit 2.
Fellowing a loss-of-coolant accident (LOCA) or rain steam line break (MSLR),
water from the containment sump is used for makeup to the core and for cor.tainrent spray. The surp has a 6-inch trash curb around the base with 1/4-inch wire resh screens that slope upward and outward from the sump to prevent debris from entering.
Failure of coatings during a LOCA or MSLB could lead to bloc 6 ace of sump screens, thus an inadequate recirculation flow to the core or blockage of spray systemt.
TVA's corrective actions were evaluated in !ection 3.7 of NUREC-1232, Volume 2.
The staff cencluded that a sufficient area of the sump screen would remain unblocked following an MSLB or a LOCA to allew the containrent spray and RHR pumps to operate safely.
Therefore, the containment coatings issue is considered resolved for both Sequoyah Units 1 and 2.
3.8 Moderate Eneroy Line Breaks In Section III.15.2 of the SNPP, TVA identified the actions it would take before restart of Secunyah Units 1 and 2 to correct the moderate-energy line break (MELB) flooding issue. The staf f's evaluation is documented in Section 3.R of NUREG-1237, Volume 2.
TVA did not identify any differences in the Unit 1 progran fror that for Unit 2.
Pased on its evaluation in NU;EG-1?32, Volume 2, the staff accepted the licensee's procedures and assumptions for evaluating MELB f1 coding.
The staff further accepted the licensee's commitrent to complete the actions listed belew before restart of Unit 1.
(1) ensure adecuote sealing between the turbine building, control building, ard the auxiliary buildinc; (2) provide administrative control for possible flooding in the annulus; (3) verify trat the electrical eovierent and electrical boards on the 734-foot and 749-foot level are above MELB fined levels; and (4) update the previous review of unirplerented ECNs to determine if subse-quent ECNs irpact the floeding evaluation.
The staff cercludes that completion of these actions will be sufficient for restart of Unit 1.
However as a post-restart action, the staf f reco-rends that TVA be able to demonstrate quick respense to MELBs in safety-related areas.
3.9 ECCS h ter loss Outside Crane Wall / Air Return Fan Operability Ey letter dated July 8, 19E7, and as supplemented August 4, 1987, the licensee identified a condition involving the collection of water from the centainment and residual teat reroval sprays follewing a design-basis accident (PBA1 TVA SER Yol. 2 Supp 1 3-13 Prelir.inary Report. Pevision 1
O Spray water collecting on the operating deck floor could drain directly into areas outside the crane wall through the opening for the containment air return fan A-A.
The Concerns were that this drainage Could result in undesirably low water levels above the sump and in flooding of the air return fan A.A.
The staff's evaluation of TVA's actions, including modifications, to resolve this issue are in Section 3.9 of NUDEG-1232, Volume 2.
All efforts associated with the curb and drain modifications have been I
completed on Unit 2; those redifications for Unit 1 will be completed be'ere l
restart.
Based on its evaluation in h0 REG-123?. Volume 2, the staff concluded that the i
re-design of the contairrent drainage system will ensure that spray water will f
l not damage the air return fans or bypass the sump; therefore, the cisign is acceptable for Units 1 and 2, 3.10 Platform Trerral Growth In its preliminary evaluatinn dated March 25, 1988, the staff approved TVA's l
plan for the resolution of the structural thermal grewth issue as described in
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Section 15.5 of the $NPp. The staff has Completed a review of the details of the ifcensee's resolution of the issues that Mclude enhanced calculations.
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generic implications, and other eff, cts of th. et,rrective action.
The statf's
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evaluation is in Section 3.10 of NUREG-1232, Wlure ?.
TVA did not identify l
any differences in the Unit 1 program from that for Unit 2.
t TVA contracted Bechtel North American Power Corporation to review the correctise action plan; Bechtel recomended several additional items. TVA f
previded supplerental information on this issue in its letter of February 29, l
1988.
The reconnendations consisted of additional calculations for design justification and redificatter of some structures and their supports.
Examples I
to be reviewed in the future by the staff include structures within the rain l
stean line valve vault rooms as well as snubbers within the reactor buildine.
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TVA has determined usino tha staff approved restart criteria that these roc'ifications may be corpleted af ter Sequoyah Units 1 and 2 restart.
J On the basis of the discussion in NUREG-123?. Volure ?, as well as its trevious j
review of W P fection 15.5, the staff cencludes that the issue of the struc-l tural thereal growth has been adeauately addressed by TVA for Units 1 and 2.
l 3.11 Pire Vall Thinning Assessment On December 9.19Pf. Unit 2 at tFe Surry Porer Station experienced a catastro-phic failure of a rein feedwater pipe, caused by wall thinning due to erosion / corrosion in a carben steel pipe elbow.
Erosion / corrosion is a form of ficw assisted corrosion.
Although ripe failures due tn erosion / corrosion have occurred in piping containing a water-steam (two-phase) mixture and pipe failures due to erosion have occurred in viater systers containirg solids, there i
have been few, if any, previously reported failures in large dianeter carbon steel piping systers containing high-purity water (single-phase); thus, the l
licensee and tFe industry in general did not have an established program for the systeratic examinatinn of the thickness of the walls of the feedwater ard condensate pipine, I
TVA SER Vol. 2. Supp 1 3-14 Prelininary Report. Fevision 1 7
The staff's evaluation of TVA's response to the Surry 2 incident is based on the SNPP and treetings with the liwnsee on June 29, Septenbar 14 and 30, at t October 29, I?87.
Infornation was also obtained from the 1,:ensee's response to NRC Bulletii. No. 87-01. "Thinning of Pipe Walls in Nuclear Power Plants,"
which is being e/aluated separately. TVA's response of September 18, in87, included its tests and inspections of piping.
The licensee selected areas susceptible to erosion / corrosion based on base metal composition, flow velocity, pressure dif fere, dials, unusual flow path or geometry, and operating temperature, inspection was by visual and ultrasonic testing (UT) nethods.
The five susceptible systems are listed below.
coa.densate (single phase) feedwater (single-phase) extraction steem (two-ptase) heater drains and vent lines (tro-phase) turbine dra.n and vent lines (tt.a-phase)
TVA did rot identify any differences in this program fer Unit 1 from that for Unit 2.
The staff's evaluation is in Section ?.11 of NUREG-1232. Volume 2.
The NRC staff concluded that TVA's inspection and surveillance procran is acceptable.
The staff also concluded that renitoring TVA's implerrentation of the surveillance progran is not necessary at this time.
TVA plans to tronitre *.us-ceptible areas and trend the results.
3.17 Cable Insta.lation 3.12.1 Program Evaluation A number of erployee concerns were received relating to construction practices at Watts Ear, particularly with respect to cable installation.
The evaluation of these conce m was extended to the Sequoyah plants Units 1 and 2.
The staff's evaluation of TVA's cable installation practices at Sequoyah is provided in Section 3.12 of NUREG-1?32. Volure 2 for Units 1 and 2.
The staff has concluded that the cable installation practices were acceptable but there was a cuestien en the silicere rubber insulated cable installed in containment.
For Unit 2. the AIW cable was removed and the (VA test data on the Anaconda and Rockbestos cable, a partial qualification of the silicone rubber insulated cable for a period of 10 years, provided sufficient margin for the startup of Unit 2.
TVA would qualify these cables for the expected life of Unit 1 and Unit 2 before the return of Unit 2 to power from the next refueling outage.
TVA's test program to extend the qualified life of the Anaconda and Rockbestos cable is evaluated in Section 3.12.? below.
3.12.2 Silicone Ruhter Insulated Cabie Environrental Cualificatien By letter dated. Noverber 24,1QS7, TVA submitted the results of tests conducted by the Wyle Laboratories on silicore rubber insulated cables (cables) installed inside centainment at Seouoyab.
By letter dated December 28, 1987 TVA decurented its basis for concludito that the cables instelled iri containrent at Sequoyah are environmentally cualified to perform their intended function for a 10 year period following the original cable installation.
The TVA SER Vol. 2. Suce 1 3-15 Preliminary Report. Revision 1
staff reviewed the TVA data and concluded that the Wyle Laboratory environmental cualification tests of the Anaconda and Rocklestos cab'es and the replacement of A!W cables inside Unit 2 containment provided adequate a;surance that the functional integrity of the cables at Sequoyah Unit 2 is adequate to allow restart of that Unit.
By latter dated May 25,19U., the stsf f requested that TVA submit details of a cable test progran for extending the qualified life of the Rockbestos and Anaconda cabl4 to 40 years.
The staff accepted TVA's schedule for completing this testihg before the Unit 2 return to power from the Cycle 3 refueling outage.
In that letter the staff outlined the basic requirerents for an acceptable cable test program and by letter dated July 6, 1988, TVA submitted the details of their cable test program.
The staff, in its letter dated P.ay 25, 1988, requested that TVA submit a cable test program for testing silicone rubber insulated cables installed in containment at Sequoyah Unit 1 and supplied by all three manufacturers (Anaconda, A!W and Rockbestos) unless TVA decided to remove AIW cables from tFe t
Unit I containment. TVA has elected to remove all the AlW silicone rubber I
insulated cables from the Unit I containment and has proposed a test program for cables supplied by the rem 'ning two manufacturers.
This cable test l
program is the test program % extend the qualified life of the Rockbestos and Anaconda cables for 40 years which is discussed above, i
Tre cable test program requires removal of installed cables for testing, five j
from each manufacturer, selected frnn the worst-case conduit configurations i
located in containrent at the Watts Par Nuclear Plant.
TVA has identified
[
i criteria used to determine the worst case conduit cor. figuration.
These
(
l criteria are similar to the criteria identified in TVA's letter of July 31, t
1987, and include the length of cable pull, sidewall pressure, and 90' condulets. The test program elso includes thermal aging, radiatien aging Loss l
of Coolant Accident (LOCA) test (steam / chemical environment) as well as i
post-LOCA-high-pot test.
The only exception is that the post-LOCA-high-pot test will be perforced at twice the cables' rated voltage plus 1.000 volts instead of 240 Vde/ mil.
Aging and LOCA tests are sufficient to demonstrate the l
functional operability of the cables.
The post-LOCA high-pot test will be used l
l to demonstrate the margin available to account for test uncertainties. Fence.
)
the staff finds the proposed test pregram acceptable.
i The staff has reviewed TVA's proposed test program and has determined that the i
test progran meets the requirements outlined in the staff's letter of Fay 25, l
10Pd with the following clarifications:
1 1
Ill TVA has defined the scope of the test procram to include only the cables
[
j which are covered by 10 CFR 50.49 Category A and B.
The staff requires
' hat all 10 CFR 50." cables be included in the progran.
TVA has inferred i
tFe steff that all '
FR 50.49 cables are covered by Category A snd B.
i However, to clarify 9e ratter TVA will delete the reference to Categcry A
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I and R.
i
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)
(2)
Enclosure ?; "Sample Selection, Si:e and Demoval Process"; TVA shcule add l
a step between (4) and (5) to state that the cable sample will be selected j
from a conduit with no less tFan ? cables, unless justified.
TVA Fas j
l t
TVA SER Vol. 2, Supp 1 3-16 Freliminary Report, Revision 1
y,
o inforced the staff that their selection criteria already include this item and will add the criteria to the test program.
(3) Enclosure 3; "Resolution of Test Anomalies and Test Failures"; 3rd paragraph: TVA should add a requirement that, as soon as the deterruina-tion is made that a test anomaly is in fact an actual test failure, NGC will be prenotly notilied of such detennination.
TVA has agreed to add this requirenent to the test program.
Based on our evaluation, we conclMe that the proposed cable test progran is acceptatle provided TVA revises the program as discussed in items (1) through (3) above. TVA's removal of A1W cables from Sequoyah Unit I and the previous qualificatten tust of Anaconde anc Rockbestos cables at the Wyle Laboratories provides adequate assurance of the integrity of cables installed at Sequoyah Unit I for a period of 10 years. This is adequate for the restart of Unit 1.
Successful corpletion of the proposed test program will extend the environ-rentally qualified life of these cat,les to 40 years.
3.13 Fuse Peplacement TVA has exterier.ced problen's with fuses at Sequoyah.
This is discussed in Section 3,13 of NUREG-123?, Volume 2.
TVA did not identify any differences in thir. program for l' nit 1 from that for Unic 2.
Rased on the test results and errerience with the FLAS 5 cadmium solder fuses free. lots 4 and hiqher, the staff finds the rep?acerent fuses acceptable.
However, tecause the analysis perferred by TVA ce the service life of the solder junction is predicted to be F.0 ~onth on the average ard ?5 month r.inimum, TVA sFould either replace these fuses every PS months or extend the life of thw e fuses with further testing and analysis based on the arbient conditions ard failure rates of these fuses.
Prelininary c port, Revision 1 TVA SER '.'ol
?, supp 1 3-17 e
7, I
i 4 RESTART REf0! NESS There are a nunber of programs necessary for safe conduct of nuclear activities at Sequoyah discussed in the Seqenyah Nuclear Performance Plan (SNPP). These I
programs related to restart readiness are the following:
operational readiness, management, cuality assurance operating experience improvement, post-modification testing, surveillance instruction review, maintenance, restart test program, training, sec9rity, emergency preparedress and radiological controls. The programs, management centrols, iaitiatives and procedures f
related to these activities were evaluated for the restart of Segunvah Unit 2 l
in Chapttr 4 of NUREG-1232, Volume 2.
This NUREG was issued by the NRC letter to TVA dated fay 13, 1988. These activities will be evaluated here for the i
restart of Unit 1.
In its letters dated $8 arch 31 and Fay 9,19M TVA idantified Unit 1 $NPP progrars that were different from Uni * ? programs.
Those different Unit 1 programs that will be e
'ted in th s chapter are the following: operational readiress and Sequoyah,. svities list. Where the Unit 1 program is the same as the Unit 2 proeran, references will te made to NUREG-1242 Volume 2.
Any j
comitments made by TVA to NRC in resolving issues identified during the staff's l
evaluation of these progrars will be referenced below.
Ir.spections cf the efft a,tiveness of these prograns have been conducted by the staff and will continue to be conducted, i
a 4.1 neeratinnal Readiney 1
a.1.1 Introduction TVA has historicelly de-onstrated weakresses in perfomance of nuclear activ-ities as has been discussed in previous Systematic Assessrent of Licensee Performance (9tP) reports. On September 17, 1985, en the basis of continued poor performance as described in the fif th TVA SALP, the hRC issued a letter l
delineating their concerns rursuant to 10 CFR 50.54(f). to the staff's 10 CFR 50.54(f) letter posed certain cuestions to TVA regarding (1) equiprent cualification (0uestions 1 and 2)
(2) eperational readiness (Ovestion 3)
(3) cable tray support (0uestion 4) fa) design control IOuestion 5)
Items (l', (3), and Idi ere discussed in Sections 3.2, ?.5, and 2.1, reseec-tively, of this rernrt.
Operational readiress will ha discussed in this l
I section.
l TVA has urdertaken a significant effort to address and ccrrect operational readiress issues.
4 special SeQuoyah Task Force was established by the Panacer of Nuclear Power en Farch IQ. Inf6, to identify problens d initiate those l
l TVA SER Vol. 2. Suro 1 4-1 Prelininary Perort, Revision 1
o,
[
i I
actions necessary to resolve the problems before restart of either Seouoyah unit.
The Seouoyah Nuclear Performance Plan (SNPP), Pevision 1, provides the assessment and plans for resumina operation of the Sequoyah units and Section V i
I discusses those topics related specifically to operational readiness.
TVA has stated that the overall purpose of operational readiness is to provide f
the Site Director with verificatien that activities, programs, and comitments l
l required for restart are completed.
This is to be accomplished by desiqnating an Operational Readiress Manager who reports to the Senior Vice President.
J Nuclear Power and an Operational Readiness Manager who reports to the Site l
i.
Director.
The Operational Readiness Panager provides independent oversight of
(
4 the develoncent and frplementation of the operational readiness program and 1
assists the site in ensurirg the program adequacy while also providing iede.
l pendent assessments and evaluations to the Senior Vice President Nuclear 1
Power.
The Site Director will use the results of the operational readiness program and other status reviews to rake his recomendation for Unit I restart to the Senior Vice President Nuclear Power.
The Senior Vice President.
j Nuclear Power will not approve restart of Unit i until be is satisfied that all
[
j I
preparatiens for restart have been satisfactorily completed.
I The Orerational Readiness Panaoer assesses whether corrective action plans have been established to address the underlying causes of deficiencies or
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problem areas, evaluates the adequacy of corrective action, reviews the close-
{
out practices and orovides coments to improve the process and program centent, j
j The Operational Readiress Manager is responsible for working with the site i
and line creanizations to obtain verification of program irplementation, to 1
l obtain verification of organizatiorcl readiness through the evaluation of per.
forrance obiectises, and to develop the restart prertcuisite checklist.
The l
l checklist will te used to verify that hardware issues directly frpacting system I
eperability are closed be' ore applicable mode changes.
i 4.1.?
Evaluation Success of the operational readiness program is contineent upon the successful i
irplementation of the three program eierents:
the SNPP cerpletion of Volu e 2 programs, the establisbrent and assessment of perfomance obiectives, and the j
restart crerequisite verification (Restart Test Instruction 9 - Unit i v ster a
1 Test Stauence).
I 1rolementation of the first eierent will be to verify (11 that restart activ.
j ities as defined in the TVA Tracking Open Iters (TR0!) co m ter list have been I
l completed (2) that SNPP Volume 2 text statements of intentien have been i
i cenpleted, ard (3) that major proiects, having broad impact on other plant l
l l
activities, have been co*Pleted prior to restart.
Sore lono-tem program enhancerents will be oren at rest:rt and will be tracked through routire '3C l
nbservations of the TVA corporate commitment tracking system, i
i The purpose o' the perforrance ob,iectives evaluation is to ensure that site nrconizations function e Hectively and are prepered for plant restart ard
[
i cperati;n. Ceneric perferrance ebjectives and criteria have been established l
I and assigret to site orcanizations so that they ray address the areas of pro.
j i
cedures, staffirg, supervisory involverent, internally and exterrally identi-fied findings, housekeepiro, and readiness of support organi:ations during restart.
Additieral performance obiectives and criteria have been ceveloped tor tre functienal areas of ercani:ation and administratien, docu~ent centrol, I
i j
Tv2 SER Vol
?. Surp 1 4-2 Preliminary Peport Devisien 1 i
. a
i 1
l 0
i naintenance, training, licensing, enoineering, and configuration control, J
p i
Performance objectives in these functional areas also FAve been assigned to the l
appropriate site organi:ations, TVA's perforrance ob.iectives are based on the guidance provided by "Performance Objectives and Criteria for Operatire and Near Term Operatino License Plants "
[
INP0 85-001, Institute for Nuclear Power %erations, January 1985.
t This operational readiness evaluatic? will inclu 9 the following:
f establishiro appropriate objectives and criteria i
l evaluating readiness aoainst estabitthed criteria assessing impact of deficiencies identified l
develeping and implementinn additional corrective actions for identified deficiencies verifying that perforrance objectives have been met and readiness is
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j assured TVA has established plant instructions and tracking systems to ensure that l
hardware issues directly irpacting system operebility are closed before mode
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i chances. To ensure P. hat these hardware issues are complete, a restart pre.
r t
l quisite checklist has been developed.
This checklist was de, eloped by the Sf)N
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j operational readiness staff and serves to consolidate hnroware operability
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issues, including those listed below,
[
(
l raintenance or work request backlog outstanding clearances
]
redification status outstanding terporary alteration control forrs ITACFs) j outstanding preventive maintenance packages instrumentation availability
)
outstandino har 6 are-related PR0s and C M Ds j
J I
The restart prerequisite checklist will be provided to the Sequoyah Restart Test Manacer for inclusien in the plant r3 start test secuencing instruction.
This instruction will provide for PCCC review and plant manager approval of rn uits prior to leaving specified hold points, in addition t; incorporating the restart prerequisite recuireren's, this instruction will address tre com.
j pletion of required special testing auring the restart of Unit 1, i
[
)
A para 1*.el, incependent assessrent of operational readiness was perforced by
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the ONP Operational Reaoiness Manacer.
This review was conducted by senior l
personnel with plant experiet.re from both inside and outside TVA.
The team I
[
provided its findiras and recowendations to the Senior Vice President.
l' Nuclear Power in a letter dated August 23, 1988.
Further, the Senior Vice President Nuclear Pcwer has requested that the SON Nclear Safety review Boara j
i (NSFB) review the SNPP Volures 1 and 2 and the actual status of preparation for r
l restart of Secuoyab units from a safety perspective.
The NSAR has reviewed and
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l accepted the overall apprcach outlined in the SNPP, The Rnard also has reviewed
[
I l
the special programs and certain secondary hardware issues and the ensite safety review process, raintenance plermina and procedure developrent.
l
]
The staff has reviewed the Independent Readiress Review as part of +ke oneoing f
sta" evaluation of the implementation of the Operatienal readiness Revier l
j 1
t l
TVA SER Vol. 2, Supp 1 b3 Freliminary Report, Revision 1 1
b
-.--.I
s s
r Frocram.
In addition, the staff has conducted an operational readiness inspection at Unit 1.
1 4.1.3 Conclusions j
Initially, the sta. ' 'elieved that TVA needed to clarify the meaning of hardware issuts in tne paragraph dt.,,:ribing the restert prerequisite verifica, i
l tien eierent.
Provisions have been included to ensure that TVA asnsses j
bardware operability for the cumulative effect on system perfora nce. Overall t
I the staf' has concluded that the implementation portion of the operational readiness precram represents a realistic and systematic femat to ensure that 8
plant activities, programs, and commitments required for restart are completed.
The conclusion of the staff from its operational readiness inspection was that l' nit I readiness was acceptable.
On the basis of its review, staff finds that this program is acceptable.
As
[
j designed the program should provide the Site Director and Senior Vice President, Nuclear Power verification that activities, programs, and comitments required l
for restart for both l' nits 1 and 2 are completed, i
1 4? Manacrent TVA's SNDP states that in the past there has been a lack of clear assignrent of 4
l responsibility and authority to ranagers and their organizations.
To correct i
this weakness. TVA has reorgani:ed the Sequoyah site organization. TVA also has taken specific actions to clarify each manager's authority and area of l
1 responsibility and to establish accountability. TVA also has progrars under way to irorove the level of plant knowledge of plant ranagers and supervisors.
The staf' bas evaluated the efforts cade by TVA to improve the manacerent and ercanizatier, at Seouoyah in Section 4.?, Panagement, of NUPEG-123?, Volure 2.
l l
1.onq-term and short-tem actions are under vay to improve the plant procedures.
The short-tem effort corsists of tFe developrent or reviticn of those i
(
crocedures necessary to support plant restart. Work for l'rit 2 was completed l
j before tFe restart ef l' nit 2.
Changes that are not necessary prior to plant restart will te bandled as cart of the long-tem procedure upgrade prograr.
I i
The lone tem procedure upgrade procram is a corporate-wide effort that will extend beyond restart of a Sequoyah unit. As part of this program, the
(
'equoyah plert procedures sill be incorporated into an overall five tiered tackage of relicies, directises, standards, procedures anc instructions that l
A Site will govern the operatiens of TVA's entire Office of Nuclear Fower.
Procedures Group has been established on a perranent basis at Sequoyah to l
participate in this long-range program.
(n the basis of its evaluation, the NDC staf f concludes that TVA has acceptably 1
)
addressed the Sequoyah-specific ranagement concerns and weaknesses for the j
restart of l'rit 1.
l 4.3 Cuality Assurance l
i I
J 4.3.1 Ouality Assurance Procrar j
This section is en TVA's procran to rn olve conditions adverse to cuality in its nuclear activities and on its cuality assurance orocra9 These crocrams l
the sti ' ir Section 4.3 of NU;Eri.1'3?, Volve 2.
TVA did f
)
here evaluated by I
TVA SEP Vol. 2, $uro 1 4-4 Preliminary Report Pevisite 1
]
I
'not identify any differences in the Unit 1 program from that for Unit 2, It is important to rote that the staff's review and acceptance of the QA topical report reans only that TVA's comitrents reet the progrematic requirements of 10 CFR 50. Appendix P, as described in Section 17 o' the NRC Standard Review Plan (NUREG-080M, The staff will assess whether these cemitments are fully and effectively ret in its ongoino oversight and inspection of TVA'S technical and CA preorans, Recause of TVA's past problems in the OA area, the Feqion II Itt'f approved this revision (Revision 9) to the 04 topical report on January 30, loo 7, for a period nf 2 years. The staff's decision en extending the a-) proval of the topical report vill depend on how effectively TVA irrplements the program, Staff reviews and audits o' the TVA Conditien Adverse to Quality (CAP) process identified tectnical and acrinistrative programmatic wealnesses.
To address these weaknesses the licensee undertook a detailed and comprehensive progran to improve the TVA CACR (problem identification and resolutdon) process.
The staff
- 1usted the OA pregram. 0A topical and the CACR process as described in the iscensee's SNPP, i
The staff assessment of the QA program anc OA topical was that the Sequoyah The stiff prograts were acceptable and the Unit 2 irplerentation was adequate.
also conducted inspectiers ir this area as discussed in Inspection Feports l
50-327/32P-8B-15, and F8-19,
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Inspectiers f0-327/32c FF 15 and FP-19 found that CACR irplementation was l
These adequate and that identified weaknesses were addressed by the licensee.
findings were applicable to both Unit 1 and Unit 9 On the tesis of its reviews and the NRC iascections, the staff concludes that l
the CACR process is acceptable and that it is being adequately irplemented with respect to both Unit 1 and Unit 2.
The staff also finds that the Ouality Assurance program is acceptable fer the restart of Unit 1, t
j 2,3,2 Net Order EA 95 40 l
Cy letter dated June 14, 1985 NDC issued the Order EA 85 49 rodifying tha f
t licenses for Sequoyah.
The basis for the Order was the circurstances l
j surrounding the preparation of a nonconfomance report (flCR) related to the l
Secuoyah certainment pressure tranmitters.
As a result of the special review i
I conducted en March 27-29, 19T,5, hDC identified a breakdewn in the Panagement controls for evaluating and reporting potentially sierificant safety concerrs, I
TVA responded with letters cated July ?, July 26 August 13, September l' and TVA Noverher 15,199F; February 4 March 7 and July 2,1986; and March 2,19R7, concluded in its letter dated Varch 2, 1987 that, with the implerentation of the i
TVA revised (cnditions Adverse to Nality (CAQ) Program, it had met the recuire.
rents of the Order.
We habe reviewed these letters and the TVA revised CAQ prc;rar as descrited in the TVA Corporate and Sequovah Nuclear Performance he have also reviewed tre irplerentation of the prograr at Sequoyah in i
- Plans, 5f seral 'JC inscections, in the Safety Evaluation, enclosed in its letter dated P' arch 31, l
As discussed 19SS, the sta " concluded that 'VA had tecertebly addressed the Order for Seouoyah, j
The NRC sta
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l Tterefore, tre Creer nas cersidered satisfied fcr Seouoyah.
I TVA S!? Vci. O, Supp 1 4-5 Prelirirary Report, Revision !
l I
- I
o.
I I
stated that it would continue. however, to monitor the implementation of the CAO progran at Sequoyah as part of its normal inspection program for the units.
4.3.3 Changes to the CAO Program 4
TVA has recently revised its CAO Progran.
The previous program was evaluated
(
J by the staff and found acceptable prior to the restart of Unit ? and was the i
basis for the staff concluding that NRC Order EA 85-49 was closed for Sequoyah.
]
These evaluations are discussed in Sections d.3.1 and 4.3.2 above and in r
Section 4.3 of NUREG-123?. Volume ?.
l l
(
A meeting was held at NRC headqt arters in Pockville. Maryland with TVA on September 8. 1988. to discuss ths changes to the CAQ program.
The staff l
a concluded that the changes are an evolution of the program and do not affect I
3 j
the staff's conclusions in its safesy evaluation on Order EA 85 49 dated l
j Ma rch 31,1988.
The staff will centinue to ronitor the CAO progrim implementa-r tion in its normal inspection activity at Sequoyah to deternine the effective-l ness of the charees that TVA has made to the program.
The stenmary of the j
September 8. 1938. reeting will be issued by September 30. In88.
l 4.4 Ooerstine Experience Imoraverent f
i Item r ' of Enclosure 2 to the 10 CFR o0,54(f) letter requested a detailed t
descr10tien of the Sequoyah Operational Readiness Plan.
In response to this recuest. TVA described operating experience actions (in terms of enhancements made through reactor trip reduction 14mitation of spurious engineered safety i
features actuations, review of the Divis-Pesse event for lessces learned, and l'
review of nuclear operations experiencesi in the SNPP.
Each of these enhance-
]
rents were evaluated by the staff in Section 4.4 of N99EG-1232. Volume ?.
The staff concluded that the actions taken oy TVA were ac.:eptable for the restart j
of both Units 1 and 2.
Post v difiestien Testinn
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4.5 e
U Past NRC inspections have identified problems with respect to the adequacy of testino of systers and components following modification.
TVA instituted programs to address the deficiencies in its post-rodification testino.
These 3
programs ycre evaluated by the staff in Section 4.5 of NUREG-1?J?. Volure 2.
The staff concluded that the programs to address post-modification testine were I
acceptable for the restart of both Units 1 and 2.
TVA did rot identify any f
)
differences in the Unit 1 program from that for Unit ?.
f J
4.6 Surveillance instruction Peview j
- 4. fs.1 Introduction Staff reviews and audits of Sequoyah surveillance instructions (515) identi'ied
]
technical are acministrative weakressss in these instructions.
To remedy these J
weaknesses. TVA has t;ncertaken a comprehensive and disciplined program to
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j j
review and revise these irstructions.
The program has undergone several evolu-
)
tiens since it was iritiated in the su ner of 1986.
These changes hwe l
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)
l i
i TVA SEF Vol. ?, Supp 1 46 Prelirinary Report Revision 1 f
I
. t
c>
resulted in increasing the technical and administrative depth of reviews, the scope of reviews, the independent evaluations of the process and its products, the field verification of sis and their supporting instructions, and the technical content and specificity of $15.
4.6.2 Evaluation The staff assessrent of the descriptive material providinQ the basis for the TVA progran to review and revise certain Sequoyah Unit 2 S!s that implenant technical specification surveillance requirements before restart included the scope, methodology and orgsnization of TVA's surveillance review and revision program. The staff also conducted inspections in this area as discussed in Inspection Reports 50 327/'J8 87-36 and 87-50.
The basic obiective of the SI program is to ensure all technical specificatien requirements are addressed and that the $1s aed their supportino instructions covered by the program scnce are technically adequate to fulfill the surveil-lance recuirerents of the technical specificaticos, have an appropriate level of dependerce on the skill of the performer of the instruction, and comply with basic administrative recuirements that rtake perfomance of the S! reli ble.
This Unit 2 program was completed prior to Unit 2 restart.
Although tFa staff enncurs with TVA's ob.iectives, TVA should define the skill Itvel required to write, revise, and review the surveillance instructions and supporting precedures and TVA should describe, including starting and comple-tion dates, the long-term program which will be undertaken to ensure co%lete administrative consistency, achieve standard femat aad organization and nake ether improvements and enhancements as are determined to be needed.
Tre staff's evaluation of the Unit ? program is in SecMon 4.6 of NUREG 123?,
Volume 2.
The scope of TVA's Unit 1 phase of the SI review program includes those tectnt'sl specification S!s and supportino instructions that are recuired for startt,v, c:eration, and sa'e shutdown of Secuoyah Unit I to the poiet of the next refueling.
The program rethodology and the governinq ergani:stion, required traininq and qualification, and instruction verification were discussed in the l' nit 2 SER by the staff ard found to be acceptable.
These parameters are essentially the sare for Unit 1.
The program is currently under the control of the site director, ard it is implerented by the established plant oroanization under the day-to-day direction of the $! review pro.iect ranager.
Both the Unit 1 art' t' nit 2 phases of the $1 review progran call for a c'etailed checklist to be used during the technical review of an instructicn to identify technical deficiencies.
Pert I of this checklist focuses on the technical adecuacy of the instruction, 5:ith an operibility evaluation beino perforced only if the instruction is found to be technically inadecuate.
Part !? of the checklist focuses en the administrative acequacy of the instruction, but all itees within this section do not need to te fulfilled to ensure insti ction adequacy.
Part 11 of the checklist does not have to be cocpleted for this procran.
Certain iters in Pert II of sne checklist, such as SRO approval to TVA SEE Vol. 2. Supp 1 4-7 Preliminary Report, Revision 1
perform the test and verification or double verification signoffs, stem from etter docurents and are checked to ensuse necessary compliance.
TVA has adopted a progressive SI verification approach that obtains the best verification pemitted by plant conditions and the approval status of the instruction.
Du*ing the latter stages of instruction preparation, the responsible section perfoms normanipulative walkdowns to confim that the instruction is correct.
4.6.3 Conclusions On the basis of its review and the NRC inspectione, the staff concludes that the Surveillarce Instruction Review and Devision Irogram is producing adequate precedures to support Unit I ttartup. Powever, the staff belicves that the program for leng-tem control of surveillance instruction upgrades, includirr, resolution of the issues of temporary changes, qualification of reviewers, and schedule, needs to be provided to completely resolve this issue.
The staf' reviews of the Sequoyah Procedure Enhancerent Program indicated that this procram is net unit trecific and that the process being employed by the licensee is essentielly the same for Unit 1.
No additional inspection activities are recessary.
4.7 Orerability "terk Pack" As a result of violatiens recardino the adequacy and tireliness of corrective actions for repetitive equipment failures and out-of-tolerance conditions, the licensee incle' rented a trending and tracking program at Sequoyah. Because this program was geared toward identifying future deficiencies, the staff raised concerns regarding pctential operability questions resulting from past, undetected, repetitive failures.
TVA cenducted an coerability "look back" program that was designed io identify adve" e cen11tions associated with eculpment operability, to evaluate the scfe y siptificance of these conditions, to document the effectiveness of ccrrective actions, and to proCose further corrective actions where necessary.
This program was evaluated in Section 4.7 of NUDEG-1232 Volume 2.
The ttaff concluded that the sccre, guidelines, and implementation of the Secuoyah crerability lcok back review program satisfactorily accomplished its intended i
purpose fer both Units 1 and T.
4.8 Wintenance Previous h90 intrectiers at TVA nuclear units indicated progra m tic deficien-cies in the site rairtenance procrams.
In the ShrP, TVA discus',es specific i
reoblems identified by the NDC and TVA that have existed at Sequoyah.
These deficiencies it.clude failure to irpienent appropriate preventive maintenance programs, failure to provide adecuate planning of mainterance activities, and inaceouacies in the uaisirg procrars for the corcorate and site personnel ir.olved in rainterance activities.
l The 'DC staff evaluated the srece, orcani:ation, and rethodology of TVA's raintenance orocem ir Section 4.R of NUDFG-1032. Volume ?.
The steff TU SE0 Vol. 2, Sucp 1 4-S Preliminary recort, Fevision 1
l i
I i
concluded that the maintenance procram is acceptable.
TVA did not identify j
any differences in the Unit 1 program from that for Unit 2.
j i
in its evaluation in NUDEG 1232, Yoltre 2, the staff noted that ranacers do not d
adecuately address long tern program develope (nt and that improvements are l
needed in time Panagement, interface with support groups, and stabilization of
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the cerporate organization. It also stated that interviews have indicated that TVA has take0 the first steps in resolving these problers as evidenced by:
[
(1) TVA has conducted a tire study of managers at the plant and has identified l
i problem areas.
It is the staff's understanding that this study involved evaluaticns of maw gerent skills, work processes, climate and stress I
factors, facilities and tools and that a report with recommendations on improving the utili:ation of manacement talent has been provided to TVA.
l
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i (2)
The staff noted that the maintenance canaperent appears to be workirg with
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support groups to establish effective interfaces as evidenced by i
managerent planning reetings with CA and utilization of SROS in the work l
Planning process.
(3) The staff noted that the pernanent corporate organization is beginning to J
take shape with the hiring of several very capable managers.
The sta'f l
feels that the corperate orcanizations can have a sicnificant impact on the establish ent of an ef'ective procram, but believe that the stabili-j ration of the corporate staff is essential to rabing this a cositive irract and not a negative impact.
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The NRC will review the effect of these actions on the effectiveness of TVA's
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naintenpece program in a future irspection,
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4.9 Restart Tert Drocrim 1
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4.9.1 Introduction In response to errleyee concerns, TVA conducted a reassessment of its plants' i
operational safety.
A rafer re review of the Seauoyah Unit ? initial design, construction and operatine practices was conducted and a Restart Test Preeram (RTP) was instituted to ascertain the functHral integrity of the accident citigation and safe shutdown systems. The principal objective of the RTP is to instill confidence that certain pre operational tests conducted during initial plant licensing and surveillance inspectiens routinely cce' e.ted following i
j piant licensing and during the long plant shutdown are d tJ *,estt that can i
4 ensure the current functional integrity of safety sys vm m corponents.
This i
.av) been i
3 assurance is recuired because the functional integrity C
.ieopardized by plant rodifications, mainterance practices v the 74ke. This I
assurance is obtained by reviewinc post-modification and raintenance tests and any other tests, or programs that right have a potential irract on the validity i
of 'he subject tests, j
The stiff evaluated the PTP for l' nit ? in Section 4.9 of hUcEG 1?3? Velu e ?.
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TFe staff concluced that tre cTP will ersure the functional intecrity of safety systers at Unit ?.
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Preliminary r port. Feviston I I
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TVA identified minor differences in the Unit 1 RTP program from that evaluated for Unit 2.
These dif'erences are discussed in TVA's letters dated March 31 and May 9, 1988. These di'ferences were reviewen by the staff in NRC Inspection 50-327/88-?9 on the containment spray system (CSS).
4.9.2 Evaluation in the Inspection 50-327/88-;9, the staff reviewed the Unit 2 CSS RTP test I
matrix as it specifically applied to the Unit No.1 CSS as well as comparing The details of the general Unit 1 program against the Unit 2 complete progran.
the inspection objectives for this review and inspection finding are given in 1
Inspection Report 50-327/88-29 to be issued in September 1988. A summary of I
these details is provided below.
(1) Unit No. 1 CSS Restart Test Procram Review l
l The inspection effort included a review of the CSS to veri'y that the Restart Test Group (RTG) functional review process n 'aing adequately implemented, to verify that components / systems functir..s that are iden-tified as requiring testing are properly dispositioneo, to provide a sample assessment of the technical adequacy of several portions of previously completed preoperational tests that are being used to satisfy the functioral testing requirerents, and to verify that the functional analysis report (FAR) matrix package complied with the applicable documents including the FSAR and TS and contaired the necessary information.
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it was determined that, for the CSS, the requirements of the Unit I restart test program were either properly implemented or TVA agreed to correct the
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issue prior to Unit restar*
i (2) Concarison o# Unit 1 RTP to the Unit 2 Completed prnoram The ' ;rpose of this comparison was to determine the adequacy of the modified Unit 1 RTP as contrasted to t9e Unit 2 program that was acceDted by the staff and docunented in Section 4.n.sf NUPEG-1232, Volume ?.
TVA provided the details of the differences bttween :he Unit 1 RTP and the Unit 2 RTP in the enclosure to its May 9. 1988 lotter.
The RTP for Unit 1 is essentially the sare as that for Unit ? and the evaluation and conclusions discussed in NUREG-1232, Volume
?, are considered valid for both units.
However, the Unit 1 program scope was 1
l reduced from that applied to the Unit 2 based on lassons learned and as a result of modification to other Unit prograns that were process inputs to the RTP.
l These differences along with the team's comments are provided below:
h Once the design functions were established, the review of the impact o
of previous mcdifications was perforred by the RTP utilizing SIL-98 to generate the modificatien review report.
This was different fron the Unit 2 progran which utilired the DBVP output for the list of
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modifications which may affect the system.
I The tean identified a possible weakness with this approach.
l Specifically, the Unit ? program had also used red line drawing to depict the as constructed system at the time the preoperational tests were performed.
Cembining the CBVP output (i.e., mods since time c' licersing) with the red line drawing, the Unit 2 program could 1
TVA SER Vol. ?, Surp 1 a-10 Preliminary Report, Revision 1
--+.+----w,-
,~-----e-,-w-,---
r,--------,,--------n,enn m,_----r,-m-_-r-,-nn,,
- ~,
.,--n-m--r-
- -- - ~ ~.
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t evaluate the adequacy of pcst modification testing of all modifications subsequent to successful preoperational testing.
In corparison, the Unit 1 program which did not include the red line drawing process, creates a gap involving the adequacy of post-modification testing tetween the time the preoperational test was performed and thf. time of issuance of the operating licensing (OL).
This problem only affected those functions where the licensee was tsking credit for preoperational tests to validate adequate testing of the specified function. TVA has determined that 274 modifications fall into the post-preoperational testing and pre-OL category. Of these, 190 modifications were reviewed as part of the modifications review for Unit 1 and 16 wtre Unit 2 only leaving 68 modifications to be reviewed. Two of the 68 modifications were determined to have a potential impact on previously tested equipment and both of these modifications were determined to be adequately tested and had no impact on the function involved, The Unit 2 progran reouirement to review the results of the post-o maintenance test survey was not included in the Unit 1 program.
This decision was based on TVA lessons learned from the Unit 2 program which indicated that only approximately 6% of the maintenance requestt (MR) reviewed indicated either a lack of adequate test documentation or a lack of adequate testing.
Addftionally, the post-maintenance test survey was not conducted for Unit 1 as part of DBVP; therefore, the RTP could not use it as an inout to their process.
Additionally, the team was informed that the additional testing controls put in place at the station as a result of the Unit ? naintenance r,rogram upgrace should reduce tLe impact of possible inadequate post-maintenance testing on the validity 9 previous functional tests, The Unit 2 requiremer.t to review the impet of the niece ptets review n
was also deleted frcn the Unit 1 program.
TVA descrit'ed the reduction in l
the Unit 1 piece parts progt c in its letter dated February M, 196F.
This reductior. is evaluated by the staff in Section 3.3 of this ESER.
TVA l
stated that the RTP, therefore, did not identify a need to review the output of the piece part program for kract on functional test validation, i
Additionally, as stated above, the licensee feels that the improved maintenance program would ensure that any part replaced as a result of the piece parts review would be adenuately tested.
4.9.3 Conclusions As stated earlier, based on the above minor program implementation changes, the team concluded that the evaluation and canclusion for the Unit ? progran as stated in Section 4.9 of MUREG-1232, Volume 2, adequately bounds the Unit 1 program.
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TVA SER Vol. P. Supp 1 4-11 Prelininary Report, Fevision 1
4.10 Training Because of the programmatic concerns arising from licensed operator requalifi-cation deficiencies identified at Browns Ferry and deficiencies identified in i
operator and shift technical advisor (STA) knowledge of the safety parameter display system (SPDS), the staff determined that the Sequoyah training program would have to l'e reviewed for adequacy prior to startup.
l Section 11.2.3 of the SNPP documents TVA's review and evaluation of training and staffing.
In the SNPP, TVA committed to increase the reactor e-w 'mr certification program to 16 weeks and to increase the requalifica
.Jriod to 6 weeks. TVA also noted that training for assistant unit operators was l
increased from 1 week to 2 weeks in 1986 and will be 6 weeks in 1987 and thereafter.
The staff evaluated the training programs instituted by TVA in Section 4.10 of r
NUREG-1232, Volume ?.
TVA did not identify any differences in the Unit 1 program from that for Unit 2.
The staff concluded that these progr?ms were sufficiently acceptable to permit restart of Sequoyah Units 1 and 2.
- However, i
the staff will continue to monitor these programs to ensure proper implementation.
r 4.11 Security In the 10 CFR 50.54(#) letter (September 17,1985), the staff noted that there I
were several areas in which TVA had not been performing adequately.
These creas were identified from their low ratings within their respective SALP f
categories.
As a result of these concerns, TVA has init'ated several actions intended to upgrade performance.
In the most recent 5 ALP, the staff found an improving trend in the area of security, compared to the degradations previously noted.
However, to ensure that this improvement would continue, TVA undertool several actions. These actions, which are discussed in Item 4 of Appendix ? to the SNPP, are evaluated below.
I TVA identified in the SNPP those neasures it will take to enhance the knowledge l
of supervisors ard employees in their respcasibilities for conplying with security requirements.
TVA will trend all security degradations to identify areas for improvement and revise the training program for public safety to e
include exrerience from prior security incidents.
To ensure the planned improvements were being properly irplemented, the staff conducted physical security inspections at the Sequoyah plant as doeurented in Inspection Peport a
Nos. 50-327/328 86-30, and 50-327/328 86-47 The staff has reviewed the infornation provided in the SNPP and has performed several physical security inspections as part of its evaluation of the improve-l ments to the Sequoyah plant security.
Based on the results of its evaluation, the staf# concludes that the action taken by TVA to improve security addresses I
the sta'f's cencerns.
In addition, the staff finds that with the implementa-i tion of these actions. TVA will have an acceptable security program for restart
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1 of either Sequoyah unit.
4.l? Emergency Preparedness l
SNPP Appendix 2. Section 6, Pevision 1, documents TVA's actions taken in the f
Sequoyah emergency preparedness (EP) progran to resolve problems identified in j
J TVA SER Vol. 2 Supp 1 4-12 Preliminary Report, Revision 1 f
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i NRC SALP evaluations.
The corporate Emergency Preparedness Branch has been reorganized and additional staff identified to provide additional resources in the areas of emergency planning and procedures, state and local government interfaces, development and conduct of exercises and drills, and onsite and offsite facilities.
Additional staff has been identified at the sites for progran implementation.
TVA has completed installation of sirens and strobe lights in accordance with approved engineering change notices issued to meet the requirements of IE Bulletin 79-18, Audibility of Alarms in High-noise Areas. Tests to verify i
the system's effectiveness with the added sirens and strobe lights will be t
completed after restart of both units, when the equipment operating noise levels are normal.
Emergency preparedness for Sequoyah was evaluated by the staff in Sectien 4.12 of NUREG-1232, Volume 2.
This evaluation covers the improvements made by TVA to its Radiological Emergency Plan for Sequoyah.
The staff concludes that, with proper implementation, past EP problem areas should be satisfactorily resolved.
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4.13 Radiolooical Controls In Section 11.1.2.3 of the SNPP, TVA discusses its improvements to the radio-logical controls (DC) organization.
The staff evaluated these improvements in Section 4.13 of NUREG-1232. Volume 2.
TVA did not identify any differences in the Unit 1 program fror that for Unit 2.
The staff concluded that these l
measures will strengthen the RC program at Sequoyah.
The staff also concluded
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that the actions taken by the licensee, including correcticn of previous weaknesses in its program for maintaining expor,ures as-low-at-reasonably.
2 achievable (ALARA), tre suf ficient to support plant restart for botn Units 1 and c.
4.14 Postart Activities List l
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4.14.1 Introduction i
For Sequoyah Ur.it 2. TM established a Sequoyah Task Force on March 19, 1986, j
to review implementation of the corrective actions applicable to Sequoyah, to l
initiate specific actions to address Sequoy3h problems, to monitor ard ensure that a list of all known work itens hcs been compiled, and to review the l
process and identification of those iters required to be completed before restart of Sequoyah Units 1 and 2.
To ccmplete its assignment, the Sequoyah Task Force developed a list of Sequoyan plant activities (except for those of a routine nature) to be i
completed before restart.
The Sequoyah Activities List (SAli was based on j
issues identified by NRC inspections, TVA quality assurance (0A) audits, i
American Nuclear Insurers (AN!) audits, institute of Nuclear Power Operations t
(!NPO) inspection reports, Sequoyah corrective action reports (CAR) and j
discrepancy reports (DR). TVA Nuclear Safety Review Staff (NSPS1 and Nuclear r
i Safety Peview Board (NSRB) reports, employee concerns, Sequoyah reactor trie
- eports and licensee event reports (LERs), and technical issues identified by IVA's Divisien of Nuclear Ergineerir.g (DNE).
TVA SER Vol. 2 Supp 1 4-13 Preliminary Report, Revision 1 m.,n.
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I The Sequoyah Task Force had established criteria (Section IV.2.0 of the SNPPI to determine which items were required to be resolved for restart.
The staff has reviewed and accepted this criteria by letter dated June 9, 1987.
The Sequoyah Task Force reviewed the process the line organi:*ation used to i
identify, evaluate, disposition, and close out items and reviewed the adequacy l
As new issues arose of planned actions taken before Sequoyah Unit 2 restart.
t and work activities were developed, they were reviewed by Sequoyah management to determine their importance to restart. The Site Director had to approve all new items added to 1e restart list; however, only the Manager of the Office of Nuclear Power (ONP) [ presently ile Senior Vice President, Nuclear Power) could delete items that had been designated for restart.
4.14.2 Evaluation The identification, tracking and closure of restart items for Unit 1 is i
discussed in Section IV of Revision 3 of the SNPP. This was submitted by TVA in its letter dated May 9, 1988.
i For Unit 1, the identification and tracking of restart items is being accomplished by TVA's permanent tracking system and reporting of open items (TROI) computer program rather than by the SAL used for Unit 2.
This program i
lists the restart and non-restart items for Sequoyah in a database.
The status and responsible organi:ation for each item is accessible through computer terminals in computer printouts to plant personnel.
This capability was not available with the SAL. The staff has reviewed the data available from TROI
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l and finds it acceptable.
l TVA stated that the Unit I rest 3!t list was developea by an item-by-item review l
of comobted and open Unit 2 and comron restart activitias and of open Unit 1 issues.
Standard Practice SQA203 "l'so of TR0! for Unit 1 Restart Action List,"
wn isseed by TVA to specify the requirements for maintaining and controlling the Unic I restart list.
The criteria used by TVA to determine if issues must i
I t'e completed before restart is the same r? start e.riteria used for Unit 2.
Standard Practice SQA203 requires each SON Unit 1 ootential restart item to be evaluated against this criWia to determine whether associated corrective action is required to be corpleted before restart, TVA stated that the Site r
Director has desigrited eitter the Restart Director or Assistant to the Site Director to evaluate propnsed new activities and ascertain that these activities reet the restart criteria.
In describing its process to close out restart items, TVA stated that Standard Practice 50A203 specifies that existing site procedures will be used to ensure that Unit I restart items are dispositioned and closed in a verifiable manner.
Each site manager is responsible for: maintaining the status of his restart items through closure: adding new actions as necessary to resolve an open restart iter as the issue evolves; and ensuring that a specific discipline and manager within his organisation is assigned re,ponsibility for obtaining tirely closure of open restart items.
An item is considered closed for restart by TVA when all corrective actions that have been specified to be completed before restart are field completed, docurented, and verified as appropriate.
To coordinate the effort to designate new activities as restart items. TVA explained that the Site Director has identified a Unit 1 Restart Director who is responsible for coordinating the Unit I restart effort.
The Unit 1 Restart TVA SER Vol. 2, Supp 1 4-14 "reliminary Report, Pevision 1
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Director reports directly to the Site Director and has responsibility and authority to establish specific schedule priorities, to ensure tFat line manaaers are coordinating their activities to complete all restart actions, to establish site goals as appropriate to achieve a safe and timely restart, to call and conduct restart schedule status meetings, and to ensure performance of the individual aroups and integrated work activities. TVA stated that this position has been established in order to ensure that all restart requirements are properly completed in an integrated fashion and on a timely basis.
4.14.3 Conclusions Based on the above, the staff concludes that the use of TROI to identify, track status on and indicate closure of Unit I restart itens is acceptable.
TVA SER Vol. ?, Supp 1 4-15 Preliminary Peport, Pevision 1
5 EPPLOYEE CONCERNS During the spring of 1985, a number of TVA employees informed the NRC and selected members of Congress of safety concerns, primarily related to the Vatts Bar Nuclear Plant.
In addition TVA learned of many employee concerns through its own organization.
The concerns indicated that many TVA employees had lost confidence in TVA's nuclear nanagement and its ability to properly conduct nuclear activities.
In addition, some of these employees expressed fe6r of reprisal from TVA management if they raised their concerns directly.
Two l
programs relating to employee concerns have resulted; they are referred to as the new program and the special program.
These two programs are discussed in detail in the staff's Safety Evaluat;on Report on the Tennessee Valley l
Authority Revised Corporate Nuclear Performance Plan, NUREG-1232 Volume 1.
dated July 1987.
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The new employee concern program (ECP) was implemented at Sequoyah on February 1, 1986, as described in a TVA submittal of February 3, 1986.
The key element of the program is the ECP Site Representative at Sequoyah.
The ECP staff receive and investigate concerns from employees who feel that normal channels of resolution have failed. The procram is further described in other TVA subnittals includino the SNPP.
The staff issued its safety evaluation accepting the TVA new ECP on September 30, 1987.
In Fay 19F,5, TVA awarded the Quality Technology Company (0TC) a contract to develop and implement a program for conducting confidential interviews with TVA employees performing assignments for the Watts Bar buclear Plant. Concerns also were collected from TVA employees at ice Sequoysh and Browns Ferry plants.
This program, which e7phasized the ider.tification of employee concerns dealico with nuclear safety at all TVA facilities, identified more than 5000 employee concerns.
In February 1985, TVA initiated a program to evaluate and resolve these employee concerns. The employ 30 concern special program (ECSP) was developed to review the concerns received threagh the OTC or from the Nuclear Safety Review Staff (NSRS) for applic uility to Sequoyah.
This work was performed by the Watts Rar employee concern task group (ECTO).
The staff evaluation of the ECSP was issued to TVA by letter dated October 6,1987 The employee concerns were grouped into nine categories for evaluation and res-olution.
The categories are construction; engineerino; industrial material control; operations; quality assurance / quality control; welding; management and personnel; industrial safety; and intimidation, harassment, wrongdoing, or misconduct.
Because Sequoyah, Units 1 and 2, were scheduled to be the first TVA plants restarted, the concerns applicable to Seouoyah only, within each employee concern subcategory, were divided into individual element reports that addressed related concerns.
For Sequoyah, element reports were prepared covering six of the categories.
TVA has submitted over 300 element reports to address the resolution of employee concerns for Seouoyah.
These element reports have been divided into those needed to be resolved and evaluated before the restart of NUREG-1232. Vol. 2. Supp 1 5-1 Preliminary Report, Revision 1
The c?iteria the Sequoyah units and those that may be resolved after restart.
used was the staff-approved restart criteria.
The PRC staff has issued, by letter dated Parch 11, 1988, its "Preliminary Safety Evaluations on the Tennessee Valley Authority Employee Concern Element This preliminary safety evaluation Reports" for the restart of Unit 2.
addressed those e'ement reports that the staff considered had to be resolved before the restart of Unit 2.
The safety evaluation for the restart element reports for Unit 1 will be issued before the restart of Unit 1.
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Subcateoory and category reports will address the resolution of employee con-i cerns for the other TVA nucleer plants.
TVA will not submit any element report for the management and personnel and industrial safety categories because TVA The staff has con-has concluded these do not contain safety-related concerns.
cluded that employee enneerns in these two categories have been adequately
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addressed as discussed in letters to TVA (December 14, 1987, and August 24, i
1987, respectively).
Conc 3rns in the ninth category, relating to intimidation, harassment, wrongdoing, or misconduct, will be investigated and the results reported separately by the TVA Office of General Counsel or the TVA Inspector General.
The staff's review of TVA's handling of these concerns is discussed r
in an October 8, 1987 letter to TVA, i
On the basis of its review of the TVA employee concerns program, the NRC staff concluded in Volume 1 of NUREG-1232 that TVA now has a policy that promotes quality and safety and TVA has taken steps to ensure that this policy is under-t The actions l
stood by TVA employees and that the policy is strictly enforced.
a taker by TVA to improve employee confidence define an acceptable program for In combination with the other improvements in deal co with employee concerns.
the t.uclear program tnat TVA is implementing, these steps should improve the conficence of emplovees in TVA's management. The staff considers effact ve irplementation of the new employee concerns program necessary if TVA is to sig-nificantly change its prior performance record.
The staff will centinue to ronitor program implementation 'nd the effectivenest a
of Octions takin to deter intimidation and harassment.
1' The staff will not issue its evaluations on all of the element reports for Units 1 and 2 as Part 2 of NUDEG-1232. Volume 2.
The staff wills es stated above, issue its evaluations of the restart element reports for Unit 1 as a Safety Evaluation Report before the restart of Unit 1.
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huREG-1 U?, Vol. ?, Supp 1 5-2 Preliminary Report, Revision 1
6 ALLEGATIONS Many concerns about nuclear safety problems were nade to TVA and investigated under their employee concerns program; many concerns about nuclear safety and other issues were made directly to the NRC staff.
In a number of instances, the technical content of these allegations were provided to TVA for inclusion i
into the employee concerns program. The NRC staff used TVA's responses as i
well as independent reviews to evaluate the issues and corrective actions, The remaining allegations will be handled by the staff in accordance with l
established NPC policies for allegations. All potential nuclear safety L
significant Sequoyah-related allegations will be evaluated and resolved to the satisfaction of the NRC staff before the restart of Unit 1.
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hUREG-1232. Vol. 2, Supp 1 6-1 Preliminary Report, Pevision 1 i
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APPEt' DIX A LIST OF CONTRIBUTORS M. Branch Office of Special Projects P. Cortland Offi e of Special Projects J. Donohew Office of Special Projects J. Fair Office of Special Projects H. Garg Office of Special Projects E. Goodwin Office of Special Projects P. Hearn Office of Special Projects G. Hubbard Office of Special Projects X. Jenison Office of Special Projects E. Marinos Office of Special Projects F. Paulitz Office of Special Projects R. Piert:n Office of Special Projects T. Rotella Office of Special Projects D. Smith Office of Special Projects R. Wetteott Office of Special Projects NUREG-l?31. Vol. ?, Supp 1 A-1 Preliminary Report, Revision 1
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l adequate voltage is available for proper operation, during a loss of ac power, j
for all four vital de systems associated with Unit I and Unit 2.
There were no specific Employee Concerns regarding the DC system loads, battery and battery charger siring.
However, TVA's QA audit prior to mid 1987 indicated that proper documentation of calculations should be made to verify DC system loads, adequate sizing of batteries and battery chargers.
In addition I
1 the same concerns were mentioned in allegations.
TVA analyzed the DC power system to determine the adequacy of the 125 volt DC Vital Batteries and battery charger sizing for the conditions defined in the FSAR and TVA's Design Criteria SQN-DC-V-11.2. This analysis is included in calculation SON-CPS-004, Revision 4.
The staff's evaluation is given below.
i Vital Batteries The loading for Vital Battery I was determined by TVA to be the highest and this was used as the basis for sizing evaluation.
TVA used a Sargent & Lundy 1
validated computer code "Electrical Load Management System Direct Current" (ELMSOC) for battery sizing.
Although Vital Patteries II, III, and IV have lower load requirements the batteries are all Gould Type NCX 2100.
Future load l
could be put on these batteries or lighting loads shifted over if necessary from Vital Battery I.
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The staff has reviewed the battery load data and finds the load data comprehen-sive and representative of the DC loads associated with this type of facility.
s The assumptions used 11 the analysis appear to be reasonable. The cattery siring included an a9199 factor of 1.26 and d minimum electrolyte tenserature l
of 60'F.
This battery sizing indicated that there was a 1.1", remainir o rwrgin for the worst case StJtion Blackout (ne alternsting current off site or on l
Mte) condition.
TVA revised the calculation to include the following:
c Sinaliff the. timing of the switchcra" breaker irdicating lights.
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o The ligSting loads were applied over the total time period.
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The electrical operated 480 volt breakers were considered closed j
l for the accident condition.
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The inverter load was reduced from 17.5 KVA to 15 KVA which is the l
l limit imposed by the FSAR and Inverter Load Calculation SON-CPS-004.
1 This corrected the last unverified assurption.
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ELMSDC record numbers were revised for many loads due to the addition of multiple load records for switchgear and to simplify load sequence documentation.
Calculation !0N-CPS-004, Revision 6, was approved by TVA on September 26, 1988 i
and the results of the last revision changes the worst case, Station Blackout,
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l battery capacity margins from 1.1% to 7.7'.. This is in addition to margins l
ir.cluded for 60* F ambient temperature (1.11), battery aging (1.25) and the load I
l growth maroins built into the individual battery loads.
l TVA SER Vol. 2 Supp 1 2-15 PreliminaryReport, Revision 1l 1
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Therefore, the staff concludes that the analysis ICs demonstrated that the i
batte-ies are adequately sized.
Vital Battery Chargers The criteria in the FSAR and Design Criteria for the battery chargers is:
Condition 1, the charger ou u tt current must provide the continuous f
o load after a 2 nour station b ackout and recharge the batteries in l
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
Condition 2, the charger output current nust provide the continuous o
load for a
. 9.ftnt condition after 30 minutes and recharge the e
l battery in 12 ho ns.
i For Condition 1, the battn j charger requirement is 117 amperes and, for Condition 2, the recuirement is 85 amperes.
Since the battery chargers have been purchased with a rating of 150 anperes, the margin for Condition 1 is 25" l
and for Condition 2 is 43%.
Therefore, the 2taff concludet that the analysis has demonstrated that the battery chargers are adecua;e'}y si:ed.
t Conclusions Based on the acove, the staff concludes that the ELMSOC calculations for the direct curren+. systems are acceptable for all plant operating conditions.
Further, the 11alysis has demonstratczi that the batteries and battury chargers i
are adequatel. sized and h6ve suffic:ent margin between their r&ouirements and their reting.
Pased on the above and tre staff's evaluation in Section 2.3.3.2.2(1)in
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t!UREG-1232, Volume 2, on the 125-vnit PC soltage calculations, the staf f concludes th
- adequate voltage is available for proper operation of Units 1 1
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and 2 during o inss of AC power and no further corrective action by TVA 15 I
required.
j (2) 1EG-Volt AC Vital Instrament Plan System Voltage Calculstions
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i This is evaluated in Section 7.3.3.2.3(?) of NUREG-1232, Volume 2.
On the basis of its evaluation of the 120-volt ac calculations and TVA's corrective actions which were completed for the restart of Unit 2, the staff concludes j
that the safety-related 120-volt ac loads powered from the 120-volt ac vital j
instrument power boards will have adequate voltage for safe operatinn of l
I linits 1 and 2.
2.3.3.2.3 Instrumentation and Control Systems Instrumentation Accuracy Calculations a
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The staff's evaluation for the Sequoyah units is in Section 2.3.3.2.3 of NUREG-1232, Volume 2.
ibe staff concluded that the instrerent accuracy calculations for both units were satisfactory, j
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?-16 Preliminary Report Pevision 1
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2.2.3.2.4 Raceway Systems I
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The staff's evaluation of raceways systems for Units 1 and 2 is in Section 2.3.3.2.4 of NUREG-1232, Volume 2.
The staff concluded that TVA's justification for using its ampacity tables and the,iustification of these i
tables as arplied to control level cable trays, grouped conduits, and conduits with more than three cables and duct banks was acceptable.
Therefore, the staff concludes that this issue is resolved for the safe operation of Units 1
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and 2.
P.3.2.3.2.5 Short-Circuit Study - Medium Voltage System This is addressed in Section 2.3.3.2.1.2 above.
The staff concludes that TVA's corrective action plan will assure that all circuit breakers will always operate within their service capability as defined by standards and verified by a
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te,t or manufacturer's guarantees.
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2.3.3.2.6 Technical Specification Surveillance Requirements l
l During the Safety Systems Quality Evaluation (SSQE) inspection conducted by the l
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staff on the Containment Spray System, the Class 1E Electrical System Design was reviewed.
This review included surveillance testing of Class 1E Electric Systems, as required by thr Technical Specification (TS). The TS surveillance i
tests reviewed by the staff are listed in Inspection Report 50-327,328/88-29.
The latest tests reviewed were for the Emergency Diesel Generators (EDGs), EDGs 3
Batteries, Vital Station Batteries (VSBsi. Motor Control Center Protection Thermal Overload (TOL) heaters, and Containment Pene-tration electrical l
protection.
The si.aff's review of these Surveillance lest procedures indicates that TVA has been improvirg then with respect to the latest IEEE standards.
The staff's reviw c' Vital Eattery 11 cacacity test, which was conducted on August 1, J
H88, doand both the procedure, conduct ind results of the test acceptable.
s I
Tnis was also true of the Vital Rattery Ill service test, which was conducted l
' nis test was conducted using the latest loM prof sie from on August 2P, 1968.
i i
Vital Patttry I which has the highest load requirerents.
' vital Battery V ray be used to replace any of the other four Vital Batteries i
and has the same surveillance requirements for OPERA 3LE status.
The staff's i
review of Vital Battery V capacity test, which was conducted on May 10, 1988, indicates thet the battery is approaching the end of life.
The Technical i
Specification (TS) defines this battery as showing signs of degradation since
+
i its capacity is 83% of the manufacture's ratino and must be given an annual performance discharge test of battery capacity instead of the normal 60 months i
)
interval.
The TS requires that batteries have at least 82'; capacity to be l
considered operable, j
s l
TVA considers that Vital Battery V is OPERABLE and nay be used in place of the other batteries during mainterance and test activities for one year.
The staff j
agrees with this position because Vital Pattery V was purchased as a ?,250
[
l ampere-hour battery and the other four batteries were purchased as 2,100 l
ampere-hour batteries.
Further, the TS requires that 7 day and 90 day surveil-lance be conducted concerninq individual cell voltage and electrolyte specific i
l gravity.
These parameters rust also be within limits for the battery to be
[
considered operable, i
?-17 Preliminary Peport Revision 1
7 -.
Based on the above, the staff concludes, from the Surveillance Tests reviewed, that the tests conducted by TVA are adequate to determine operability of the electrical power systems in the TS.
2.3.3.3 General Conclusions on the Sequoyah Electrical Calculations Program J
The staff's conclusions on the electrical calculations program for Unit 2 are i
civen in Section 2.3.3.3 of NUREG-123P, Volume 2.
The staff concluded the I
followino for Unit 2:
TVA's analysis includes the essential auxiliary power systems required for safe plant operation.
The input data is sufficiently corrprenerisive and detailed for l
J consideration of all modes of plant operation.
The calculations assumed i
worst-case system and plant conditions.
The methodology used in these i
ant. lyses was appropriate for assessing problems in the systems.
TVA has i
i stated that it will correct the problems identified in the specific areas i
4 before restart.
(
[
i
]
TVA's proposed resolutions for each deficiency identified in the j
electrical calculations are acceptable.
TVA has provided a comitment to implement the proposed resolutions before restart.
l I
The content and format nf each system calculation is adequate for documentation purposes.
All doeurentation of the electrical calculations necessary tnr restart is in place and up tc da+e by comput e progran for easy manipulation (i.e.,
i data is retrievable for maintenance and update).
Thus, the staf f cencluded that there is reesonable assurance that the systems I
addressed will provide safn restart and op uation of Sequoyah Unit 2.
j Btsed on its review of the elect,r' cal calculations program for Unft 1 is
[
discussed above, tne staff concluuts that the above conclusicos for Unit 2 apoly also to Unit 1 and that there 45 reasonable assurance that the systems addressed will provide safe restart and operation of Sequoyah Units 1 and 2.
2.3.4 Branch Technical Position PSB-1 The staff's evaluation of Unit 2 against Branch Technical Position (RTP) PSR-1, is given in Section 2.3.4 of NUREG-1232, Volume 2.
The staff concluded that Unit 2 acceptably met the BTD.
The staff evaluated Unit 1 against the BTP PSB 1 as part of its evaluation of i
the Unit 1 electrical desien calculations.
This is discussed in Section 2.3.3 above.
The staff concluded in Section 2.3.3 that Unit 1 acceptably ret
[
BTP PSB-1.
IVA SER Vol. 2. Supp 1 2-18 Preliminary Report, Revision 1
- -~..- --..- - -- -.~-
2.4 Alternately Analyzed Piping and Supports SNPP Section III.5 describes a TVA program to verify the adequacy of piping and pice supports that had been installed and oualified by alternate analysis (AA) criteria.
TVA's AA criteria use design rules and guidelines to locate supports in lieu of rigorous pipino analysis.
This is discussed in Section 2.4 of NUPEG-1232, Volume 2.
TVA is conducting a two-phase procram to resolve the concerns on the Cateoory I (safety class) AA pipinq systems.' TVA provided a description of the Phase i program activities in Section III.E.?.1 of the SNPP.
The scope of the Phase I program includes those systems required to mitigate events addressed in FSAR Chapter 15 and safely shut down the plant.
This scope is consistent with the scope of phase I of the Design Baseline Verification Program.
The Phase !
review effort involved screening of AA pipino systems for specific deficiencies that had been identified in TVA's AA proaram as discussed earlier.
The Phase I score is evaluated in Section 2.4.2 of NUREG-1232. Volume 2.
The staff evaluation of restcrt procram implementation was based on an audit of the Unit 2 progran. On the basis of this audit, the staf f concluded that TVA had adequately defined and was adequately implementing a program to ensure that short-term safety concerns would be identified, evaluated, and resolved before Unit I restart.
TVA was unable to provide the basis for the deflection criteria that ensure that pipe supports are rigid.
In a letter dated January 28, 1987 TVA stated it will perform an evaluation during the long-term progran to.iustify the adequacy of the criteria.
This was acceptable to the staff.
IVA. in a letter dated August 18, 1986, defined a set of interim acceptance criteria for evaluating pipinq and pipe supcorts in the restart program.
TVA originally defined tre proposed interim criteria in terms of exceptions to FSAR corritments.
These exceptior:s and the staff's accepter.ce of them are listed in Section 2.4.2 of NU1EG-1?32 Volume 2.
In addition TVA proposed criteria for supoort evaluations taken from Section 3.0.4 ef the current NPC Standard Peview Plan ard 'ron Subsection NF of Section III of the American Society of Vetrar,ical Enqirears (ASPE) Code.
These criteria are not in accordance with the Saouoyah FSAR; r.onetheless, the use of these criteria en an interim basis is accettable to the staff.
Powever, the long-term program should use the criteria that meet the cotmitments in the FSAP.
TVA discussed the scope and activities of the Phase 11 effort in Section S.2.2 of the SNPP. This is evaluated in Section 2.4.2 of NUREG-1?32, Volume 2.
Phase !! will evaluate the remainino Categorv i AA safety class pipino systers reauired for restart for the areas of concern identified in the Phase i not procrar.
Phase 11 also will address instrurent lines and their supports.
The acceptance critetia for Phase !! will te TVA's established design criteria for piping and surports.
TVA presented the scope and the schedule for Phasn !! in a letter dated April 8, 10F7, in addition to the deficiencies evaluated ir the Phase I program, TVA also will address the areas of concern listed belcw in the Phase !! progran.
consideration of thermal flexibility analyses for piping systers with creratino temperatures between 120:F and ?00'F J
TVA SER Vol. ?, Surp 1 2-10 Preliminary Report, Revision 1 1
e.
consideration of the interface between AA piping and deadweight supported j
piping for pipe sizes less than or equal to 2 inches in nominal diameter consideration of the effects of long piping runs and large concentrated weights As discussed in Section 2.4.2 of NUREG-1232, Volume 2, the staff concludes that i
TVA has defined an adequate program for resolution of short-term safety j
concerns required for plant restart. On the basis of its audit of sample design packages and a field inspection of sarrple Unit 2 piping systems, the staff found that the program was adequately implemented. The staff concludes a
l that completion of the Phase I program for Units 1 and 2 will provide confidence that sufficient safety margins exist--in the design of AA piping /
)
support systems required to mitioate FSAR Chapter 15 events and safely shut i
down the plant--to allow Ur.it 1 to restart.
?.5 Cable Tray Supports i
TVA's neiginal desian criteria for cable trav supports were developed between 1977 and 1974 Although these design criteria included the effects of earthquakes, they did not consider the effects of desinn-basis accidents (DBA).
4 in 1975, TVA revised the original design criteria to include the DBA loads, but the original desinns were never reviewed to ensure that they ccmplied with the i
revised criteria.
This deficiency affected only the cable tray supports j
attaened to the steel containeent vessel (SCV); however, other deficiencies i
found in 1984 and loF6 dictated a thorough review of the adequacy of all the j
cable tray supports.
During that review, TVA discosered that the existing cable tray supports could not satisfy the basic comitments rade in the FSAP.
i At a neetino on July 17 and 18,1986, TVA proposed a set of interim acceptance
.riteria for cable tray supports that were less stringent than those in tne i
FSAP. As a cart of its recuest, TVA also committed to restore the original FSAP criteria for the affected cable tray upports in an orderly maneer after i
restart.
2.5.1 Interim Acceptance Criteria l
j 2.5.1.1 Evaluation i
1 1
The staff's evaluatine of the interim acceotance criteria propoted by TVA is 6
discussed in Section 2.5.1.1 of NUPEG-123?, Volume 2.
The a plicability of the l
q 1
staff's evaluation to Unit 1 is discussed below.
j l
l Ili Damping l
L 1
i TVA proposed to use 7 percent of critical damping for the cable tray 'or the safe-shutdown earthquake and design-basis accident (SSE/DBA) loadino, as j
compared with the 5 percent allowed in the FSAR.
For restart of Unit 1, the 7 percent darping proposed by TVA for DBA/SSE Inading is acceptable to the staff.
[
l t
j
(?) OBA/SSE Lead Combination l
t i
In the FSAR, TVA comitted in usino the absolute 4 enrbination of SSE and OPA I
I loading ef'ucts.
TVA now copeses to use the squai, root of the sum of tFe s
j squares (SpSSi cerbination for the interim acceptance criteria.
The staff r
i TVA SER Vol. 2, Supp 1 2 70 PreliminaryPeport, Revision 1l 4
i 3
o finds the SRSS method a reasonable load combination approach for Unit I restart and it is acceptable.
(3)
Elimination of 1/2 SSE Load Case in the FSAR. TVA ccomitted to considering the SSE and 1/2 SSE loads. TVA now I
proposes to use the SSE loading only for the interim acceptance criteria.
The proposed elimination of 1/2 SSE case is acceptable to the staff on an interim basis.
(4) Allowable Stresses In the FSAR. TVA makes a commitment that the cable tray support stresses be less than 0.9 times the yield strength for SSE/CBA loading.
TVA now proposes to chance this requirement to 1.7 times the American Institute of Steel Con-struction ( A!SC) allowables for SSE plus DBA loading, and 1.6 times the AISC allowables for the SSE alone.
The criteria proposed by TVA for cable tray support calculations are acceptable.
7.5.1.2 Implementation of Interin Criteria Tha staff's evaluation of the implementation by TVA of the interim criteria is aiven in Section 2.5.1.2 of NuoEG-12??, Volume 2.
The applicability of the staff's evaluation to Unit 1 is discussed below.
(1)
Cable Tray Supports Attached to Steel Cantainment Vessel The re-evaluation of supports attached to t'te steel containment vessel was required to resolve Nenconformance Report (NCR) SOFCER 8414 The PCR addressed the fact that the cable tray supports on the steel containment vessel were not desianed for DBA laadings. The staff corcluded that methods used in re-evaluatino the SCV cable trav supperis were adequate and that the interin 3cceptance criteria were appropriately implemented to qualify the supports for the restart of Unit 1.
(2) Cable Tray supports on the Peactor Building Shicld Wail Many cable tr4/s loc.ited in the annulus between the SCV and the shield wall are supported fro 1 the shield wall.
The staff concluded that TVA has demonstrated that each cable tray support attached to the shield wall had sufficient capacity to meet the interim criteria for the SSE load condition.
(3)
All Other Cable Tray Supports There are 2000 cable tray supports in Catecory I structures (excluding the steel containment building and the reactor building shield wall). "ost of these r 3 in the auxiliary buildino (1700) and the control buildina (850).
The staff tencluded that the proaran conducted by TVA for nualificatien of thete cable trav brackett and supports was adacuate and acceptable for Unit I restart.
?-21 kreliminary Peport Revision 1
~ - _
s 2.5.1.3 Anchoring in Concrete This discussion apolies to supports that are anchored in concrete bv means of base plates, anchor bolts, and embedded plates.
The staff's evaiua' tion is l
proviced in Section 2.5.1.3 of NUREG-1232, Volume 2.
(
The staff concluded that TVA should use, as a minimum, the original FSAR design f
criterion requiring 2.5 for wedae-type anchor bolts and 2.8 for self-drilling j
anchor bolts as safety factors for the interim period.
I As discussed in Section 2.5.1.3 of NUREG-1232. Volume 2, TVA comitted to the l
interim criteria proposed by the staff; therefore, this is acceptable.
For the long-tenn effort, TVA should determine the actual safety factors and evaluate them agairst the requirements of IE Bulletin 79-02.
2.5.1.4 Pase Plate Analysis The staff's evaluation is in Section ?.5.1.4 of NUREG-123?, Volume 2.
The staff concluded that the rodeling and analysis of base plates are acceptable.
?.5.1.5 Concrete The resolution of this issue is discussed in Section 2.6 of this report.
?.5.1.6 Confirmatory Items t
The staff identified ten confirnatory items in Section 2.5.1.6 of NUREG-1232, l
Volurre 2.
These items were identified during the audit of September 29 through
(
October 3.1986, and were required to be resolved by 1VA before restart of a-unit.
From revi.twing the inferration provided in TVA submittals dated January 14, and February 4, 1987, the staff concluded that TVA ' ad taken proper corrective action for the above ten confirmatory itens and that this is acceptable for i
Units 1 and 2 restart.
TVA conducted a test for the wedge bolt anche' in the area of the cracked concrete in accordance with 'iVA Construction SpecificatlMs i
i and found that no degradttion of the base plate anchor was observed.
Based on en enginet; ring,iudgment, this is con;idcred to be acceptable for restart.
However, an audit of the above items, including the cracked concrete, will be ccncocted following restart of the plant.
I 2.5.1.7 Conclusion I
The staff concluded that the interin acceptance criteria proposed by TVA for i
Sequoyah Units 1 and 2 restart, as modified in accordance with this report, are
[
t acceptable for the restart of both units.
i
?.5.2 Diesel Generator Buildino Supports Analysis r
All of the four diesel generators are required for the operation of either j
Unit.
The conclusion, as discussed in Section ?.5.2 of NUREG-1232, Volume ?,
{
applies to both units.
TVA has evaluated all cable tray support calculations in the diesel cenerator building and the additional diesel generator building
{
fer a failure to take the effect of Zero period acceleration (ZPA) into l
TVa SER Vol. 2. Supp 1 2-22 Preliminary Report, Revision 1 l I
account.
In those instances where the originally calculated acceleration was less than the ZPA, the ZPA was applied in the re-analysis.
Results of the re-analysis indicate that the existing cable tray supports are still able to serve their intended function during a seismic event. Therefore, on the basis of its inspection and its review of the inforration presented by TVA. the staff finds that no structural modifications are required.
2.5.3 Cable Tray Support Base Plate Installations Sixteen base plates (eight per unit) for the cable tray supports in the auxil-iary building were irrproperly installed in that every hole in the base plates was drilled per the engineering drawing with a diameter 3/8 inch larger than specified by TVA procedures.
Based on its evaluation in Section 2.5.3 of NUREG-1232 Volume 2, the staff concludes that TVA has completed all the necessary corrective actions regardinn the base plate installation deficiencies.
As a result, the redified connec-tions are judged to be able to serve their intended function as required by the design.
On this basis and its review of Section III.3 of the SNPP, the staff finds the issue nf oversize holes in the base plate has been acceptably resolved for both Units 1 and 2.
2.6 Concrete Quality The TVA avaluation of Eroinyee Concern IN-PS 095-002, related to the adequacy of the concrete cuality at the Watts Bar Nuclear Plant site, prompted the NRC staf' to renuast further evaluations of the in-place strength of the concre+e at the Sequovah site.
On the basis of its evaluation in Section 2.6 of NUREG-1232, Volume 2, the staff conclude', that all previous concerns related to adequacy of the structural criteria fc concrete strength and frequency of sa M 1ng and 1
centrols and standards for the beddlia cortar have been ecsolved for the restret of Unit 1.
2.7 Yiscellaneous Civil Er.gineering Issues TVA identified t, need to addrass t.be seismic oualification of components in meeting code and regulatney req;irements.
This effort ircludes the review of components. Diping, pipe supports, cable tray supports, corduit supports and heating / ventilating duct supports as well as structures, fection 15 of Part
!!! of the SNPP addresses miscellaneous civil engineering issues related to Seouoyah.
The staff evaluated TVA's special programs to resolve the miscellaneous civil engineering issues in Section ?.7 of NUREG-1232 Volume ?.
TVA stated that there were program dif'erences between Unit I and Unit 2 in this area; however, the differences were concerned with TVA's implementation of
!E Pulletin 79-14 This is discussed in Section ?.3.2. Civil Calculations, above of this SSER.
i As discussed in Section ?.7 of Nt!DEG-1?32 Volure 2, and on the basis of its l
?-23 Preliminary Report, Revision 1
i e
l re<iew of the TVA plans to execute these special prngrams, the NRC staff finds tFat with proper implementation of the plans the special issues should be fully resolved for Units 1 and 7, 2.8 Heat Code Traceability Section III.15.6 of the Secuoyah Nuclear Performance Plan (SNPP) describes a TVA commitrent to investicate materials control concerns involving FSAR comit-rents, desian requirerents, and traceability relative to pressure boundary Dioing components in the Seouoyah safety-related piping systems.
TFe issue of heat code traceability has been evaluated through the employee l
concern orcaren (element report MC 40703).
The staff's evaluation of this issue is discussed in Section 2.P of NUPE4 1232, Volume 2.
In addition, the issue of heat code traceability has been evaluated throuch the employee concern program in element report MC-40705-SQN, "Paterial Control "
Revision 4 and in Investigation Report ECP-87-SQ-510-09, "Heat Number Validation for " A Level Structural Steel Materials During the Construction Phase of SeQunyah Nuclear Plant."
The NDC concludes that TVA has properly characterized the problems with heat code traceability and has adequately addressed the employee concerns.
TVA SER Vol. ?, Supp 1 2-74 Preliminary Report, Pevision 1
le a
3 SPECIAL PROGRAMS The Sequoyah Restart Task Force identified a nun,ber of technical issues of particular interest that were to be addressed before the restart of either of the Sequoyah units.
These issues include major regulatory programs, such as environmental qualification of equipment and fire protection, as well as specific technical issues, such as adequacy of electrical cables. ~ % resolu-tion of these issues are discussed in the sections bele.:.
In sorne cases, there are related employee concerns; the individual evaluations of the element reports are discussed in Section 5.
In its letters dated March 31 and May 9,1988, TVA did not identify any differences in the Unit 1 SfiPP special programs from the Unit 2 programs tnat resolved the technical issues discussed in this Section.
The staff did conduct an additional inspection on Unit 1 on fire protection.
This inspection and the commitments made by TVA to fiRC to resolve these special programs will be discussed below.
3.1 Fire Protection 3.1.1 Program Evaluation Following a staff inspection of July 16-20, 1984, at Watts Bar on compliance with Appenoix R to 10 CFR 50, the staf f issued a Confirmatory Action Letter to TVA on August 10, 1984 This letter identified the actions to be taken by TVA to implement a con.plete review of the Appendix R program at Sequoyah.
In accordance with the Confirnatory Action Letter, TVA established roving fire-watches to provide continued surveillance of selected arer.s in the auxiliary N ilding, control builoing, and the turbine building.
These firewatches covered areas of the plant that contain cable / safe shutdown system interactions c%t did ret treet the requirements ef 10 CFR S0, Appendix R Sec*.fo, III.G.
In addition, these reving firewatches were required to cover their assigned areas at least once an hour 4.nd document their actions in accordance with TVA's Operations Section Letter Acministrative 73.
The staff evaluated the Appendix F program at Sequoyah in Section 3.1 of NUREG-1232, Volume 2.
This evaluation discusses the deviations requested by TVA from the requirements of Appendix R to 10 CFR Part 50 and the compliance of Sequoyah to Sections !!!.G, Ill.J and !!!.0 of Appendix R.,
On the basis of this evaluation, the staf f concluded that when the modifications and implementa-tion of the procedural corrective actions associated with TVA's deviation requests (as identified in the staff's SERs of May 29 and October 6,1986) and redifications and prJcedures (as toentified in Inspection Reports 50-327/88-37 and 50-328/86 37) are completed, TVA's Appendix R program will prcvide an acceptable level of fire protection, equal tc that required by 10 CFR 50, Appendix R, Sections !!!.G, Ill.J. !!!.L, and !!!.0 for Unit 1.
As a result of the recent inspection (July 25-29,1988), the staff found addi-tional interacticns that had to be cddressed.
This is oiscussed in Section 3.1.5 belcw.
-- OhALMRMASMmL 3-1 Freliminary Report, kevision 1
3.1.2 Staffing of the Fire Brigade By letter dated June 13, 1988, the licensee submitted a description of the reorganized fire brigade at Sequoyah Units 1 and 2.
This reorganization is part of an overall plan to upgrade the fire fighting capabilities at all TVA nuclear plants with professional fire brigade personnel.
The new fire brigade organization is in compliance with Sequoyah Technical Specification change 87-44 submitted to NRC by TVA letter dated Parch 1, 1988.
TVA is committed to meeting the requirements of Appendix A of Branch Technical Position (BTp) APCSB 9.5-1 at Sequoyah. The original NRC staff approval of the Sequoyah fire brigade is addressed in Section 9.5, Fire Protection Systems, V.
Administrative Controls, of Supplerent 1 of the Seouoyah SER NUREG-0011 which supports the licensing of Seouoyah.
The SER describes the Sequoyah fire brigade as consisting of at least five members equipped with breathing apparatus, portable communications equipment, portable lanterns and other fire fighting The SER also states that the fire brigade participates in periodic equipment.
drills and meets the requirerents of Appendix A to BTP ASB 9.5-1, NFPA recom-rendations and supplemental staff guidelines.
The original fire brigade was staffed by the Assistant Shif t Engineer as the fire brigade leader and four operations personnel. The reorganized fire brigade will be controlled by the assistant shift operations superv.sor (formerly assistant shift engineer).
The assistant shift operations supervisor will serve as the incident commander but the brigWe will be staffed by the brigade leader and four individuals from the onsite Fire Operations Unit.
The incident commander will respond to all plant fire emergencies and will provide the technical knowledge of safe shutdown systers to determine the effects of fire and fire suppressants on safety-related systems.
The incident commander will also remain in direct connunications with the shift operations supervisor / emergency coordination in order to provide any technical information that may be recuired for the plant Each duty shift cf operations staff to safely shut down an operating reactor.
the Fire Operations Unit is staffed by a fire captain (brigade leader) who has l
The fire j
professional fire service experience; and four fire operators.
operators Fave met the minimum standards for certification as firefighter il l
as defined by NFPA, have had 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> of class-ocm instructions on site 9
specific fire protection systems as well as on-the-jub-training, emergency l
health physics training, and emergency redical trainino, i
It is the licensee's position. and the staff agrees, that the reorganized five brigade meets or exceeds the existing fire brigade commitrent; v The licensee ha crovided a comparison of its fire brigade to requirements.
the requirerents of BTP CMEB 9.5-1 (which the licensee is rot committed to).
t The comparison showt that the reorganized fire Operations Unit meets the intent of the requirencnts of CMEB 9.5-1 in Section C.3 Fire Brigade.
\\
The staff concludes that the reorganized fire operations unit reets the staff guidance in Appendix A of BTP APCSB 9.5-1, in regard to the staffing, training l
i and eouipping of the plant fire brigade, and is acceptable.
t l
Preliminary Peport, Pevision 1
3.1.3 Fire Pump Design Deficiency TVt. identified a design deficiency on April 14, 1987 which could cause Class 1E electrical components to operate outside of their design limits during a postulated Loss of Coolant Accident (LOCA) event. TVA reported the design oeficiency to NRC in accordance with the requirements of 10 CFR 50.73 on August 18, 1987. This report was documented in a Licensee Event Report (LER)87-042.
The design basis of the plant does not provide for a LOCA condition concurrent with a fire.
This occurrence would be considered a low. probability event.
However, TVA identified the potential of the fire pumps starting automatically as the result of a LOCA condition.
The electrical analysis of the Class 1E electrical system by TVA also identified that the fire pump starting and running during a LOCA condition resulted in degraded voltage of the Class 1E electric auxiliary systems.
Further, if power is supplied from the standb.y emergency diesel generators (EOGs) these EDGs could also be overloaded.
There are two fire pumps associated with each Sequoyah unit. The pump motors are supplied electrical power from their respective redundant Class 1E electrical auxiliary systems. The pumps of each urit discharge into their respective headers. The headers from each unit are interconnected b.y a nomally open isolation valve.
The fire puro control logic was designed to provide an automatic start from the fire protection heat sensors within the containrrent.
The heat sensor actuation design setpoint temperature was specified as 220' F.
The containment temperature during a LOCA can exceed 240' F, causing the fire pumps to start.
The starting of the fire pumps during a LOCA condition would result in the degradation of he Class IE electrical auxiliary power system.
TVA took a shor'.- 9rn correction action to prevent problems during the Unit 2 restart by,nlacing the Unit 2 fire pumps control switches in the lockout petition.
This cp* cation would prevent the automatic start of the fire pumps, j
associated with Unit 2, should there be a Unit 2 LOCA event.
The Unit 1 fire pumps would be available, if needed, for both units.
TVA's ler19-term corrective action before Unit 1 start up involves rodification of all fire pump start logic.
This modification blocks the automatic start of l
Unit i fire purps 1A and 18 during a Unit 1 LOCA condition.
Similarly, the j
Unit 2_ fire pumps 2A and 2B would also be modified to prevent their automatic start during a Unit 2 LOCA condition.
l l
On the basis of our review, the staff concludes that the modification proposed by TVA, to correct the design deficiency, is acceptable.
The staff will review the electrical calculations for two-unit operations to verify that automatic i
starting of the fire pumps concurrent with other unit operating conditions does j
not cause degrading of the Class If auxiliary electrical systems.
This review l
is discussed in Section 0.3.3 above, i
f 4
[
TVA SER Yol. ?, Supp 1 3-3 Preliminary Report, Fevision 1
3.1.4 Fire Protection Calculations Revision 9 Ly letter dated June 10, 1988, TVA subnitted Revision 9 of the Sequoyah Appendix R shutdown logic calculations.
TVA stated that the Unit 2 plant configuration and associated Appendix R docurentation reflect this revision to the calculations except where interim compensatory measures exist.
TVA explainea that Unit 1 is in a verification process and any modifications identified during this process will be completed before Unit I restart except where interim compensatory measures exist.
TG stated thdt the Unit 1 verification process would be completed by July 11, 1966.
The staf f conducteo an inspection on July 25 through 29, 1988. The Revision 9 of these shutdo,vn logic calculations were reviewM by the inspection team during the inspection.
The staff's evaluation will be the Inspection Report 50-327,328/88-37 which will be issued in September 1988.
ihere are no unresolved Mcde 4 items.
3.1.5 Inspection An inspection cf the Unit 1 fire protection prograri was conducted on July 25 through 29, 19E6.
The cetails of the inspection and the conclusions of the staff were issued in Inspection Report 50-327,328/88-37 on October 26, 1988.
3.1.6 Conclusion The staff has evaluated the Sequoyah fire protection as discussed above.
The staff will br. issuing its evaluation of Revision 9 of the Sequoyah Appendix R shutdown logic and its Inspection Report 50-327,328/E8-37 in September 1983 before Unit 1 enters Mode 2.
The staff's evaluatien discussed atieve is sufficient to allow Unit I to restart from the current cutage, 3.2 Environmental Qualificat hn of Electric Equipment Orpertant to Safety 3.2.1 Cor liance with 10 CFR 50.49 A licensee must demor. strate that equipment that is usec tc perforn a necessary safety function is capable cf maintaining functional oreratility under all service conditions postulated to occur during its installed life for the tinc it is required to operate.
This requirement is appliceble to equipment 1scated 1
inside as well as cutt.ide containment, tiore detailed requirements and gaidance relating to the methods and procecures for demonstrating this electrical equipment capability are in 10 CFR 50.49, "Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants"; in f.UFEG-0588, "Interin Staf f Position on Environrental Qualit acation en Safety-Related Electrical Equiprent" (which supplements IEEE Standard 323 and sarious fiRC regulatory guices and irdustry standards); and "Guidelines for Evaluating Environmental (ualification of Class 1E Electrical Equiprent in Operating fenctors* (Civision of Operating Reactors (DOR) Guidelines).
The staff evaluation of the compliance of Sequoyah to requirements in 10 CFR EL.49, environmental qualification, is in Section 3.2 of fiUREG-1232. Volute 2.
for ident1 tying equiprent within the scope of 10 CFR 50.49(b)(1), (c)(2) y In this evaluatien, the staf f concluded that the rethodology teing used t' and TBA SER 801 20 Sus 9 1 3-4 Preiininary Report, Revision 1 0
~
(b)(3) is acceptable because it provides reasnnable assurance that equipment within the scope of 10 CFR 50.49 has been identified.
With regard to 10 CFR 50.49(b)(31 TVA evuluated existing system arrangements and identified equipment for the variables defined in Regulatory Guide (RG) 1.97, Pevision 2.
TVA has submitted a report outlining the results of the review and schedules for modifications.
Because the review is not complete, some of the ecuipment items jointly within the scope of NUREG-0737 and RG 1.97 have not been included in the 10 CFR 50.49 scope. When the RG 1.97 report and equipeent lists contained therein have been finalized and accepted by the staff, appropriate equiprent not already in the 10 CFR 50.49 scope will be added in accordance with the RG 1.97 implementation schedule.
TVA will complete environrental qualifi. tion of the applicable FSAR Class IE-designed instrumentation and the FSAR pu..-accident monitoring (PAM) instrumentation before Unit i restart.
For those instruments already added to the plant tecause of a commitment to meet post-TMI requirerents (NUREGs-0578 and -0737), TVA will complete its environmental cualification in accordance with its responses to those NUREGs or any extension granted with respect to those responses.
For instrumentation that is not considered operable or not installed but that will be complete by startup frem Unit 1, Cycle 4 refueling outage in accordance with the icolerentation schedule for RG 1.97 or post-TMI NUREGs, environmental qualification will be complete when the equiprent is installed and operable.
For that instrurentation that exists at the plants but that was not ircluded in the orioinal PAM irstrumentation set but that will be Category 1 or 2 PG 1.97 instrumentation, TVA will cceplete environmental cualification in accordance with tFe iTp;erentation schecule for RG 1.97 On tne basis of its 5. valuation in NURE3-1322, Volume 0, the staff has r? ached th? following conclusiens with regard to the qualification of electric equipment important to safety within the scope of 10 CFR 50.49:
(1)
The Sequoyah eliotrical equirrent environmental cualificaticn progren evolies with the require:ents of 10 CFP 50.49.
(2)
TVA's proposed re..olutions for ear.h of the environmental evalification deficiercies identified in the staff's SER and the FRC's TER ere acceptable.
The staff's findings regarding compliance with 10 CFR 50.49 rely on certain nodifications/ replacements that rust be corpleted for the affected eouipment to be qualified.
TVA will provide certification that all restart work is complete for Unit 1 tefore the entry of the unit into Mode 2.
3.2.2 Superheat Transient (Main Steam Temperature issue)
TVA desioned Sequoyah to withstard an unisolable break in a rain steam line either irside containTent or in the main steam valve vaults ("SVVs) located outside contain-ent.
As part of this design the electrical equipment used rdurinn this accident wculd be required to operate in the hich temper 3tures generated by such a line break.
After the plant was completed, the information on which the design was based was changed by Westinghouse.
This resulted in IVA SER Vol. 2, Supp 1 3-F Preliminary Peport, Revision 1
O g
ircreased accident peak temperatures in containment and the valve vaults. As a consequence, the design of the equipment located in these areas required re-evaluation. This issue is di3 cussed in Section III.6 of the SNPP and involves the Main Steam Line Break (PSLB) in the MSVVs and inside containment.
This issue is evaluated by the staff in Section 3.2.2 of NUREG-1232, Volume 2.
The staff concluded for Sequoyah Units 1 and 2 that this issue for the MSLP in the MSVVs was resolved.
The staff's conclusion that the containment temperature profile for the design basis MSLB inside containment is acceptable contingent on the verification that the analysis contained in the Westinghouse Reports WCAP-10986 and -10988 is The staff's review of these reports is being conducted on a generic accurate.
basis and the results of the generic review will be addressed separately. TVA has submitted information to the staff in its letters dated November 10, 1987 and February 10, June 1. August 31 and September 22, 1988.
3.3 Piece Part Qualification (Ficcurement)
TVA Nuclear Safety Feview Staff (NSRS) reports R-84-17-NPS and R-85-07-NPS identified deficiencies in TVA's practices for the procurerent of safety-related replacement items.
NRC hspection Report 50-327/328 86-61, dated November 14, 1986, cited related ceficiencies which were classified as a potential enforcement item (86-61-01) for failure to take corrective action.
Specifically, the TVA program could allow previously cualified ect.4pment to be degraded by purchasing replacement components and parts as commercial-grade, without documentation of its cualification and without adequate dedication of the items by TVA.
This is discussed in Section 12.0 cf Part 111 of the SNpP, Revision 1.
TVA has established the Secuoyah Replacement Items Project (RIP).
Throuch its RIP, TVA will establish a maintenance history of plant reple,cerent activities by reviewing maintenance requests, preventive maintenance activities, surveillance instructions, and work plans.
ONE will perform a documented engineering review and evaluation to establish the suitability of replacerent items for their intended application.
TVA responded to the staff's concern by letters dated April 1 and December 8, 1987 and provided a long-term program plan by letter dated February 10, 1988.
The staff TM staff evaluated the RIP in Section 3.3 of NUPEG-1232 Volume 2.
concluded that this process was sufficient to support plant restart of Unit 2.
The staff has reviewed TVA's supplerental program plan to the RIP which was submitted as an enclosure to TVA's letter to the NRC dated February 10, 1988.
This supplements the original RIP program plan which was submitted to the ?$C on April 1,1987 and addresses TVA's comitrent to provide a supplerental RIP procram plan in TVA's letter dated December 8, 1987.
TVA submitted the irclenentation schedule for the supplemental program plan in a letter to ';RC j
c'ated August 10, 1988.
TFe original procran plan provided for TVA's review and evaluation for f
i adecuacy of qualification, all installed replacerent items within the scope of l
10 CFP 50.a9 and seismically sensitive replacement items within the bouncary l
of SCN Unit 2 pre-restart phase of tFe Oesion Baseline Verification (OBVP).
i l
l l
TVA SER Yol. 2. Supp 1 3-6 Drelininary peport, Revision 1
o,
I All other Unit 2 installed safety-related replacement items were to be reviewed and evaluated post-restart.
The original program further provided for similar reviews and evaluations to be performed on Unit I with the same pre-restart and post-restart scheduling restrictions.
The pre-restart reviews and etaluations were performed for Unit 2 as required.
The supplement orogram changes the original program to allow for the substitution 01 a warehouse inventory review and evaluation of safety-related replacement items for adequacy of qualification instead of performing the reviews and evaluations on actual installed replacerv.nt items covered within the original scope of Unit ? post-restart items W Unit 1 pre-restart and pcst-restart items.
The plan also provides for review of deficiencies io?ntified during the Unit 2 pre-restart efforts and the warehouse inventory effor+.s relative to the need for corrective action on replacement items installed in the plant.
TVA provided justifications for the proposed changes in the supplemental program plan and the cover letter transmitting the plan.
The staff reviewed and evaluated the supplemental program plan and its schedule for the following:
(1) differences between the original program plan and the supplemental plan; (2) adeouacy of TVA's justifications for the program changes; and (3) adequacy, relative to restart of Unit 1 of TVA's actions toward the resolution of Unresolved Items (UR!) 50-327/87-40-01 from c
hRC Inspecticn Deport 50-327,329/87-40 dated Novenber 30, 1987.
Additionally, the supplemental plan was evaluated to determine if it provided an adequate level of confidence that l' nit 1 could be operated safely.
Based on the staff's reviews and evaluations, the staff finds that, with proper implementation of the plan, this special ist,ue (including actions toward resolution of the URI)
{
is satisfar.torily resolved for tFt restirt of Unit 1.
3.4 Se,n,sino line issues Issues were raised through the employee concerns progran concerning the instru-rent line slope, comprassion fittings and teflon tape.
These issues were evaluated by tFe staf f in Section 3.4 of NUoEG-1232, Volume 2.
TVA did not identify any differences in the Unit 1 progran from that for Unit 2, TVA has issued an electrical design stendard to 00 used for instrument line slope criteria in future modifications.
TVA has also issued an instrunentation engireering requirements specification that specifies the design standards and the required 0A inspections.
The staff has reviewed the new elet.trical design standard and believes that design standard together w1th the instrument specification will prevent the future recurrence of the problem.
Based on its evaluation in NUREr,-122?, Volute 2, the staff finds that these issues are adecuately resolved for the restart of Sequoyah Units 1 and 2.
As a lonq-ter-action, corporate auidance on the use of teflon tape and a single defin d Spe replacement plan will be issued.
l 3.5 Weldinq In Section !!I 8 of the SWp, TVA discusses the weldirq project program to evaluate the adequacy of tFe TVA welding program fer all of the TVA plants ard the suitability of welded structures and systems for service.
In addition.