ML20153B222
| ML20153B222 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 03/11/1988 |
| From: | NRC OFFICE OF SPECIAL PROJECTS |
| To: | |
| Shared Package | |
| ML20127A683 | List:
|
| References | |
| NUDOCS 8803210502 | |
| Download: ML20153B222 (9) | |
Text
'o UNITED STATES NUCLEAR REGULATORY COMMISSION n
WASHINGTON. D. C. 20555 y
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SAFETY EVALUATION BY THE OFFICE OF SPECIAL PROJECTS FOR EPPLOYEE CONCERNS SE0V0YAH NUCLEAR POWER PLANT UNITS 1 AND 2 ELEMENT REPORT EN 215.1(B) "CIVIL / STRUCTURAL DESIGN-SE!SMIC CRITERIA" i.
SUBJECT Category:
Engineering Subcategory:
Civil / Structural Design Element:
Seismic Criteria Concern:
00-85-005-009 Sequoyah Nuclear Plant is sited on an earthquake fault that runs from around 1
Chattanooga to north of Knoxville.
If there were an earthquake, power plant structures could fail.
II.
SUK9RY OF ISSUE 1.
Sequoyah plant is on an earthquake fault that runs from Chattanooga to Knoxville.
2.
Plant structures could fail in an earthquake.
III. EVALUATION As discussed in the Subject Element Report and the SER issued in March 1979 (Reference 1), the nearest regional fault to both the Sequoyah and Watts Bar l
Nuclear Power Plant is the Xingston Fault, which lies about one mile to the northwest of the plant at its closest approach.
This fault is about 150 miles long, strikes northeast and dips at least 30* to the southeast.
Projection along the dip of the fault would place it at least 2000 feet beneath both plant sites.
The Kingston Fault is one of numerous low angle thrust faults that characterize the Southern Valley and Ridge Tectonic Province. These faults range in length f em several tens of miles to more than 100 miles.
They were for ned during the Applachian Orogeny in the late Paleozoic Era (more that 250 million year ago).
There is no evidence that these faults have been active since that time, however, outcrops that expose cross-cutting relationships between the faults and overlying younger strata are rare. The following are the bases presented 8803210502 080311 DR ADOCK 050 0
, in the Clinch River SER (Reference 2) to support the staff's conclusion that these faults are not capable in the meaning of Appendix A,10 CFR Part 100:
1.
Extensive filed research has been conducted in the region with the intent of finding evidence for recent displacement along these faults to explain current seismicity, and none has been found.
2.
Triassic dikes mapped in Virginia penetrate Valley and Ridge Province structures without being offset.
3.
In Alabama where Coastal Plain deposits overlie the southern part of the Valley and Ridge Province structure there is no evidence of offset.
4 Where subsidiary faults of the major thrust faults have been mapped in relation to overlying ancient terrace deposits, those terraces have not been offset (i.e., Phipps Bend and Watts Bar site fault investigation; TVA,1975;TVA,1974).
5.
Radicmetric age dating of gouge taken from the Copper Creek Fault, which is similar to the Kingston Fault and strikes parallel to it several miles to the east, indicates an age of at least 280 million years before present.
Seismological studies of instrumentally recorded earthquakes in eastern Tennessee and some of their aftershock sequences indicate that their hypocenters occur predominantly in the Precambrian basement well below the Paleozoic thrust faults.
Fault plane solutions of these events suggest that the earthquake source mechanisms are inconsistent with the structural trends and the sense of predominant displacement on these low angle thrust faults that are characteristic of Valley and Ridge. On the other hand, trends and senses of motion of these earthouake are consistent with structures imaged in geophysical data taken within the Precambriam basement.
IV. CONCLUSION For the reasons stated above, the staff reaffirms its conclusions made in previous licensing activities regarding sites in eastern Tennessee, specifically Sequoyah and Watts Bar, that the regional low angle thrust faults, including the Kingston Fault, do not represent a ground displacement or seismic hazard to nuclear power plants in that region and concurs with the conclusion drawn in the subject element report.
V.
REFERENCES 1.
NUREG-0011. "Safety Evaluation Report Related to Operation of Sequoyah Nuclear Plant, Units 1 and 2" dated March 1979.
2.
NUREG-0908, "Safety Evaluation Report Related to Construction of Clinch River Breeder Reactor Plant" dated March 1983.
SEQUOYAH NUCLEAR POWER PLANT, UNITS 1 & 2 SAFETY EVALUATION REPORT FOR EMPLOYEE CONCERN 5 ELEMENT REPORT 215.2-SQN "CIVIL /STRUCTUR AL DESIGN, CUT REBAR CONTROL" I.
Subject l
i Category: Engineering Subcategory: Civ"il/ Structural Element: Civil / Structural Design, Cut Rebar Control concerns: IN-85-297-005, IN-85-868-004 The bases for Element Report 215.2-SON, Rev. 1, dated January 27, 1987, are Sequoyah Employee Concerns IN-85-297-005 and IN-85-868-004 which questioned the structural integrity of the containment and the crane walls inside the reactor building because of over 2000 known releases for core i
drills due to penetrations of ducts, conduits and pipes.
These concerrs were evaluated by TVA as potentially nuclear safety-related and potentially applicable to Sequoyah (generic).
i II.
Summary of Issues The stated concerns as defined by TVA are: (a) cutting of rebar in the reactor containment and the crane walls inside the reactor building could have weakened the structure; (b) there are over 2000 known releases for core drills; and (c) procedural control / assessment of cut rebar to ensure structural integrity of concrete is in question.
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III. Evaluation Investigations by TVA personnel and consultants found that all three issues were valid and that there was a lack of procedural controls and adequate assessment of the cumulative effect of cut rebars.
The NRC staff reviewed TVA's investigations and concurred with their findings.
To re-solve the three issues, TVA developed a corrective action plan (CAP) that included three pre-restart and three post-restart action items.
The pre-restart CAP was tc (1) revise existing plant procedures to ensure co-ordination between plant operations and TVA's Division of Nuclear En-gineering, and develop a new procedure for documenting and controlling future rebar cuts, (2) develop a baseline map of cut rebars that ocurred in the reactor building during the construction phase, and assess the structural integrity of the shield wall and crane wall for the cumulative effect of both cut rebars and hanger loads (see Element Report 215.6-SON for employee concern on hanger loads), and (3) assess the structural in-tegrity of the most critically affected concrete elements in the auxiliary building - slabs at Elev. 714', 734' and 749', U-line wall, and other critical shield walls - for the cumulative effect of both cut rebars and hanger loads.
The percentage of cut rebar was assumed to be the same as the worst percentage of cut rebar developed from the data for the cor-responding concrete elements of the auxillary building at the Watts Bar Nuclear plant (WBN) because the concrete design is similar for the two plants and because such plant specific data for the SON auxiliary building were not available.
The assessments were done on the basis of the ultimate strength method as specified in design criterla SON-DC-V-1.3.3.1, and the combination of dead, live, and FSAR OBE or SSE loads was considered. -
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4 TVA has corpleted the implementation of the pre-restart CAP.
To assess the adequacy of the scope and implementation of the pre-restart CAP, the NRC staff performed a valkdown of the plant and audited a representative sample of the results of TVA's implementation.
In addition, TVA vas re-quested to compare the percentage of cut rebars between the SON and WBN reactor buildings based on the available data from both plants.
The com-parison showed tha,t the percentage of cut rebars in the reactor building was similar between the two plants, and the NRC staff accepted TVA's assurption for pre-restart CAP item (3) regarding the similarity in per-centage of cut rebars between the SON and WBN auxiliary buildings.
TVA was also requested to verify that the structural assessments, which con-sidered the seismic loads from the FSAR OBE and SSE, provided sufficient safety margins with respect to the seismic loads from the site-specific (84-percentile) SSE by evaluating the two most critically stressed lo-cations of the slab in the auxiliary building at Elev. 714'.
The eva-luation results obtained by TVA demonstrated that the floor does possess sufficient margin to withstand the 84-percentile SSE.
Based on the above evaluations, the NRC staff found the scope and implementation of the pre-restart CAP to be acceptable.
i For the post-restart CAP, TVA will (1) develop a plant-specific baseline of cut rebars for all Category I concrete structures at SON to facilitate the long term assessment of the cumulative effect of cut rebar and hanger loads, and also review the WBN cut rebar data and evaluations in detail because they were already complete, (2) revise Section 3.8 of the Sequoyah FSAR to clarify the use of the ultimate strength method as specified in design criteria SON-DC-V-l.3.3.1 for the evaluation of the reactor building and auxiliary building because the ACI working stress, _.
i method was the original FSAR criteria for the design of conctete for these l
two buildings, and (3) evaluate and document future cut rebar requests based on procedures developed f rom the pre-restart CAP.
The NRC staff found the scope of the post-restart CAP to be sufficient.
IV.
Conclusions 1
The NRC staff reviewed TVA's investigation of the employee concerns and the CAP developed by TVA to resolve such concerns, and found they were
. adequate.
TVA's implementation of the pre-restart CAP was also f ound acceptable.
The NRC staff therefore believes TVA's resolution for the concerns as described in Element Report 215.2-SQN, Rev. 1, is acceptable.
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SEQUOYAH NUCLEAR POWER PLANT, UNITS 1 & 2 SAFETY EVALUATION REPORT FOR EMPLOYEE CONCERNS I
ELENENT REPORT 215.6-SON "CIVIL / STRUCTURAL DESIGN, HANGER LCADS ON STRUCTURES" I.
Subject Category: Engineering Subcategory: Civ11/ Structural Element: Civil / Structural Design, Hanger Loads on Structures.
Concerns: IN-85-220-003, IN-86-173-001 The bases for Element Report 215.6-SON, Rev. 1, dated January 27, 1987, are Sequoyah Employee Concerns IN-85-220-003 and IN-86-173-001 which questioned the structural integrity of the supporting valls/ floors in the Unit 2 reactor building annulus areas in particular, and in other concrete structures in general, due to the weight of an excessive number of hanger attachments.
These concerns were evaluated by TVA as potentially nuclear safety-related and potentially applicable to Sequoyah (generic).
II.
Summry of Issues The stated concerns as defined by TVA are: (a) structural integrity of concrete valls/ slabs in the annulus area of the Unit 2 reactor building 1s questionable due to excessive number of hangers; and (b) design cal-culations have not evaluated individual and cumulative effects of hangers on concrete valls/ slabs.,
III. Evaluation i
Investigations by TVA personnel found both issues to be valid and identi-fled one additional deficiency, i.e., lack of control and documentation for hanger loads.
The NRC staff concurred with the findings from TVA's investigations.
To resolve the employee conerns and the related deft-ciency, TVA developed a corrective action plan (CAP) which consisted of both pre-and post-restart corrective actions.
The pre-restart CAP was to (1) perform live load evaluation of all Category I structure concrete slabs, (2) perform evaluation for two worst case shield walls, the reactor building shield wall,.and the auxiliary building U-line wall, and (3) re-vise DNE and plant procedures to control approval for all future hanger attachments, and develop a program plan for the long term evaluation of remaining Category I concrete walls not covered by the pre-restart eva-luation.
For the concrete elements in the reactor and auxiliary building, the cumulative effects of both hanger loads and cut rebar were considered simultaneously in the evaluations, as was discussed also in Element Report 215.2-SON, Rev. 1.
The implementation of the pre-restart CAP is complete, and the assessment of concrete structural elements was based on the ulti-mate strength method specified in design criteria SON-DC-V-1.3.3.1, con-i sidering the combination of dead, live and FSAR OBE or SSE seismic loads.
The NRC staff's evaluation included a walkdown of the plant and an audit of representative samples of TVA's implementation results.
The scope and implementation of the pre-restart CAP items were found acceptable.
The assumption regarding the similarity in percentage of cut rebar between the SON and VBN auxiliary buildings and the assessment of the concrete structural elements were found accepta'Ce as discussed in the staff safety evaluation for Element Report 215.2-SON, Revision 1.
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Regarding the post-restart CAP, TVA has committed to (11 revise Section 3.8 of sequoyah FSAR to clarify the use of the ultimate strength method from design criteria SQN-DC-V-1.3.3.1 for the structural integrity assess-ment of the reactor building and auxiliary building, and (2) perform the long term evaluation of Category I concrete walls not included in the pre-restart assessments, based on the program plan developed in pre-restart CAP item (3).
The NRC staff found the scope of the the post-restart CAP to be sufficient.
IV.
Conclusions The NRC staff reviewed TVA's investigation of the employee concerns and the CAP developed by TVA to resolve such concerns, and found they were adequate.
TVA's implementation of the pre-restart CAP was also acceptable.
The NRC staff therefore, believes TVA's resolution for the concerns as described in Element Report 215.6-SON, Rev. 1, is acceptable.
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SEQUOYAH NUCLEAR F0WER PLANT, UNITS 1 & 2 SAFETY EVALUATION REPORT FOR EMPLOYEE CONCERN ELEMENT REPORT 215.9 (B). STRUCTURAL STEEL CONNECTION DESIGN"
- 1. Subject Category:
Engineering (20000)
Subcategory:
Civil / Structural Design (21500)
Element:
Structural Steel Connection Design (21509)
The basis for Element Report 215.9 (B) Revision 1, dated Jannary 13, 1987 is employee concern IN-65-297-003 which states:
"Structural steel connections (I-beam to embed plates) are both welded and bolted. One method is for vibration and the other is for dead loads.
Both type connections are being used on the same-I-beam and these are not supposed to be mixed'.
Construction Dept, concern.
CI c'eclined to provide futher information."
This concern was evaluated by the licensee as potentially nuclear safety-related and potentially applicable to Secuoyah (generic). A similar concern was investigated under Sequoyah Element Report 222.5(B) entitled "Pipe Support Weld Design - Bol.ts Replaced by Weld."
II. Summary of Issues One issue was defined by the licensee:
Bolted and welds are used in the same connection to transfer loads frcm structural steel members to concrete walls.
These are not supposed to be mixed.
III. Evaluation The employee's concern about mixed connections was evaluated as related to welding and bolting at the same connection.
There is nothing unusual about a welded connection at one end of a beam and a bolted connection at the other end.
The FSAR and design criteria for the Sequoyah Nucit.ar Plant comit the licensee to design structural steel in accordance with the American Institute of Steel Construction (AISC) code.
Section 1.15.10 of the AISC code contains the design criteria for the use of bolts in combination with welds:
"A-307 bolts, or high strength bolts used in bearing-type connections j
shall not be considered as sharing the stress in combination with weld.
Welds is used shall be provided to carry the entire stress in the connection."
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2 The intent of this rule is that the relatively rigid weld will carry the shear load when connections have a ccmbination of bolts and welds. According to this rule, replacing one bolt in a four bolt connection with a weld means that
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the weld must accomodate the shear stress formerly taken by all four bolts.
However, the licensee also creates the potential for using a weld of insufficient size with Note 2 on Drawing 47A050-2 Revision 5 which states:
"Mechanical Hangar Drawings - General Notes - Note 2 - Where a bolt anchored plate overlaps an existing embedded plate, bolt anchors may be replaced by a minimum of 2 in, of 5/16 in, weld for each bolt eliminated due to some portion of the bolt hole being obstructed by the embedded plate.
Engineering design approval is required.
This note is not applicable to 1-1/4 in, wedge bolts or to any size of grouted anchors."
1 Significant Cendition Report SON CEB 8601 noted that this note had existed since 1978, but design calculations to justify this statement did not exist.
Similar notes are found in the licensee's drawing series 47A051, 47A052, 47A054, 47A055 and 47A056 which cover general notes for all seismic category 1 support structures for piping, electrical conduits and trays, HVAC ducts, and instrument tubing.
? licensee review found several structural steel supports with these mixed connections, e.g., supports for a large duct at elevation 710 that circles the reactor cavity wall.
The licensee submitted recent licensee calculations which concluded that all systems will be able to perform their intended functions and no failures would occur as a result of the drawing note.
The licensee's evaluation team observed that the loads were shared by the welds and bolts contrary to AISC rules and there wete several cases where tia yield stress of the weld was exceeded based on allowable design stresses, in addition, the shear strength of the base metal was not considered in determining the load-carrying capacity of the weld. While this is a programatic errcr. the NRC staff noted that it has a regligilbe effect for these welded base plates.
The licensee's evaluation team substantiated the employee concern that welds and bolts are used on the same connection.
The team also found that the calculations do not demonstrate licensee conformance to FSAR comitments.
For corrective action, Sequoyah randomly selected 60 baseplates with mixed connections that represent the structures throughout the plant.
These baseplates were analyzed by considering all of the shear forces applied to the baseplate as acting on the weld or welds.
Policy memorandum PM-86-17 was issued to provide instructions for designing these mixed connections and to prevent a reoccurrence of this type of problem.
The plan was to strengthen deficient welds, but none were found based on actual loads.
Since no welds were deficient. TVA claimed a 95* ccnfidence level inthe integrity of these types of connections at Sequoyah.
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IV. Conclusions The NRC staff believes that the licnesee's investigation of the concern was adequate, and their resolution of the concern as described in TVA Employee Concerns Special Program Report Number 215.9 (B) Revision 1 dated January 13, 1987, entitled "Civil / Structural Design - Structural Steel Connection Design" is acceptable for Sequoyah. The licensee admitted that the expansion anchors are designed to carry shear loads with welds contrary to AISC code requirements.
Sequoyah performed a random sampling program of 60 baseplates and performed a stress analysis based on the as-measured dimensions of the connections.
No connections required weld strengthening end the sample gave a 95% probability that less than 5% of all of these connections at Sequoyah may need strengthening. The licensee issues a policy memorandum to provide instructions for designing these mixed connections and to prevent a reoccurrence of this type of problem.
Technical
Contact:
Paul Cortland (301) 492-8734 l
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E WASHING TO N, 0. C. 20555 SAFETY EVALUATION BY THE OFFICE OF SPECIAL PROJECTS EMPLOYEE CONCERNS PROGRAM TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2 DOCKET N05.
50-327 AND S0-328 1.
INTRODUCTION The issues addressed in this Safety Evaluation Report (SER) for the Sequoyah Nuclear Plant (SON) are in the civil / structural and pipe support design areas. This report provides an evaluation of 2 individual concerns categorized in the following 2 element and/or subcategory reports:
ELEMENT / SUBCATEGORY DESCRIPTION 21510/25000 Feedwater Heater Monorail Design 22110/22100 Use of Snubbers if detemined to be valid these issues must be resolved for the Sequoyah Plant.
II.
EVALUATION The NRC consultant, Parareter, Inc., has reviewed the 2 employee and/or subcategory reports and prepared the attached Technical Evaluation Reports (TER).
The staff hcs reviewed the TERs and concurs in their bases and findings.
There were no allegations identified during the review pertinent to those reports.
Those elements that were initially submitted as non-restart justification issues were reviewed as part of a sub-category report.
The review included the evaluation of the employee concerns as well as addressing the SON restart issue.
Where corrective action has been warranted, the staff's acceptance is based upon satisfactory fulfillment of all comitments as described in the TVA corrective action plan.
III. CONCLUSION Based on the staff review of the attached TERs relating to the employee concerns program for SON, the staff concludes that TVA has adequately i
addressed the employee concerns and that their conclusions and corrective actions are acceptable.
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2 Certain corrective actions have been implemented for SON Unit 2 only.
It is the responsibility of TVA to assure that acceptable implementation of such corrective action will be performed for Unit 1.
Any additional program changes should be submitted for staff review and shculd not be implemented prior to review and approval by the staff.
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SEQUOYAH NUCLEAR POWER PLANTS, UNITS 1 AND 2 TECHNICAL EVALUATION REPORT FOR DIPI4YEE CONCERN.ELDENT -
REPORT 21510(B), "{'EEIMATER HEATER MONORAIL HANGER DESIGN" I.
SUBJECT Category:
Engineering (20000) t i
Subcategory:
Civil / Structural Design (25000) l Element:
Feedvater Heater Monorail' Hanger i
Design (21510)
The basis for Element Report 21510(B), Rev.0, 12/2/86 is employee concern LDA-86-001, which questions the structural integrity of the feedvater heater monorail hangers.
l II.
SLMtARY OF ISSUE The structural integrity of hangers for the feedvater heater monorails located in the turbine building is questionable.
j III.
EVALUATION i
i TVA subcategory report 25000, Rev.2, 10/26/87, and TVA element report 21510(B), Rev.0, 12/2/86, identify the issue as not safety related because of the monorail function and location within the turbine building. The TVA reports also identify the issue as l
not valid. The chronology of events affecting this issue is given as follows:
-The concern was expressed orally on or before August 5, 1985.-
-On August 6, 1985, the concerned employees met with the TVA design engineer who explained the design approach and details of the renorail hangers.
In a statement documenting the meeting, it i
is recorded that the employees expressed satisfaction and gave l
their assent to closing the issue.
-A TVA structural engineer made an independent review of the feedvater heater trnorail design on August 13, 1985 and affirmed its adequacy.
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-A 3rd party review vns mde on August 19, 1985 by Impe11 Corporation, j
vhich confirmed the design as adequate.
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-The scope of these reviews and the conclusion reached are stated within subcategory report 25000 as:
"The design calculations and drawings were reviewed for assumptions, icgic, analysis, code interpretations, m mber selections, connections, and clarity of presentations. The evaluation team found the design documents well organi::ed, cceplete, and meeting the AISC reqairements."
-TVA performed a load test of the system on August 25, 1985, using i
a Mad 40% heavier than the operating load to be carried. The i
test was successful, i
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IV.
CONCLUSION.
TVA eva hation, action, and resolution of the expressed concern is adequate and acceptable for Sequoyah Units #1 and #2 restart.
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SEQUOYAH NUCLEAR POWER PLANTS, UNIT 2 TECHNICAL EVALUATION REPORT FOR EMPIDYEE CONCERN ELE 4ENT REPORT 22110(B), "USE OF SNUBBER" I.
SUBJECT Category:
Engineering (20000)
Sabcategory:
Pipe support Design (22100)
Element:
Use of Snubber (22110)
The basis for Element Report 22110(B), Rev.1, 12/30/86, is employee concern SON-86-001-02 which states that the Upper Head Injection System vertical riser requires a rigid support where a snubber was used.
II.
SUS'ARY OF IS$UE A rigid type support is specified in the piping analysis for a specific location on the vertical riser of the Upper Head Injection (UHI) system, but the detail drawings and as-built condition show use of a snubber at this location. UHI has a plant safety-related function.
III.
EVALUATICN TVA element report 22110(B),Rev.1, 12/30/86 recognized the employee concern as valid.
In a subsequent letter J.A. Mcdonald (TVA) to B.J.
Youngblood (NRC), 2/17/87, responding to an NRC re uest for additional informtion, the root cause of this disparity totveen the pipe support analysis and the as-built condition was given as a lack of attention to detail, specifically, that an engineering judgement was made regarding support orientation and design without proper docunentaticn and ccenunication to interfacing groups.
The letter also identified a 10M engineering review of all snubbers in the plant against the piping analyses, and confirmed this instance to be a single, isolated case. The report indicates that TVA re-analysis of the UHI pipe restraint at this location utilizing a snubber demonstrated the use of the snubber to be an adequate design, able to sustain required seismic and thermal stress levels. 'Ihe TVA evaluation identified this as an acceptable resolution in the report, but also described TVA's decision and cc=mitment to replace the snubber type support with a rigid type support prior to re-start. TVA recognizes the necessity to fulfill applicable requirements of design centrol and configuration control of 10 CFR 50, Appendix B, Criterion III and ANSI N 45.2.11 in TVA work performed to SON Pipe Support Design Manual (pSEti), Volume III.
The depth and extent of the evaluation team review of this issue is adequate, including identification of root cause of the probicm. Corrective actions regarding both the engineering design activities and replacement of the pipe support are adequate.
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_2 The replacement action has been tracked under Corrective Action Tracking Document (CATD) 22110 SON 01, and is reported as completed and verified for Sequoyah Unit #2 only, on 8/27/87.
IV.
CONCWSION TVA evaluation and resolution of this employee concern is adequate, acceptable and appropriate for Sequoyah Unit No. 2 restart.
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4' SE000YAH NUCLEAR POWER PLANT, UNITS 1 & 2 SAFETV EVALUATION REPORT FOR EMPLOYEE CONCERNS ELEMENT REFORT 218.1(8), REVISION 1 "PIPE STRESS CALCUL,STIONS THERMAL ANALYSIS OF PIPING SUBJECTED TO TEMPERATURE LESS THAN 120' I.
Subject Category:
Engineering (20000)
Subcategory:
Pipe Stress Calculations (21800)
Element:
Themal Analysis of Piping Subjected to Temperature less than 120"F (21801)
Concerns:
SON-86-002-03, SON-86-001-03, IN-85-038-001, IN-85-039-001, IN-85-039-002 The tases for Element Report 21801, Revision 1 dated December 19, 1986 are Employee Concern Nos. SON-86-002-003, SON-86-001-03, IN-85-038-001, TN-85-039-001 and IN-85-039-002, which question the therinal analyses of piping performed by TVA.
II. Summary of issues The Employee Concerns Task Group (ECTG) report identified the following six issues from the employee concerns:
a.
Current operating mode drawings were not used for t.11 subsequent analyses.
b.
Site group stress inalysts were not allowed to evaluate the significance of the current operating mode definitions in the analysis of reccrd, c.
The environmental temperature in the annulus area r:ay reach 150*F 4
but site group stress analysts were not allowed to evaluate the effect of the environmental torperature on piping in that area.
d.
The operational mode procedure does not require evaluation of previously perforred thermal analyses when thermal conditions
- change, e.
Not all stress-analyzed piping included a code-recuired evaluation of themal expansion.
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Excess'ive levels of pipe support loads and pipe stress due to thermal expansion have been observed for some piping where the system operating temperatures were between 40*F and 120*F and a thermal expansion evaluation was not perfomed.
III. Evaluation A technical review of Employee Concerns Element Report ?l8.1(B), Revision 1 was performed by NCT Engineering, Inc. under NRC Contract to. 05-86-156.
The results of this review are summarized in the attached NCT technical evaluation report dated November 30, 1987 on Employee Concerns Element Report 218.1(B),
Revision 1.
Element Report P18.1(B), Revision 1 found that only issue d contained a valid concern fcr rigorously analyzed pipino systems.
The report further stated that based on the results of a sampling program at Watts Barr, the thermal operating modes used for Sequoyah were adequate and no corrective actions vere required. The report referred to Elemant Report 218.4(D) for evaluation of issues e and f tar alternate analysis piping. The NCT review of Element Report 218.1(B), Devision 1 found that the ECT6 evaluations of issues a, b, c and d was acceptable. Based on a finding by the NRC's Integrated Design Inspection (101) the NCT report concluded that issues e and f should remain open until the IDI finding is resolved.
The staff concurs with the conclusions presented in the NCT technical evaluation report.
The NCT technical evaluation report identified that an additional item has been raised by the ECTG based on a revised version of an employee concern.
This new issue has not been transmitted to the NRC and has not been reviewed.
The NCT evaluation also identified that TVA cemitted to issue new operating mode drawings for all Unit 2 piping systems and recomended that this effort be completed in a timely ranner.
TVA's implementation of this commitment to evaluate operating mode drawings should be reviewed as a post restart item for both Units 1 and 2.
IV. Conclusions Based on the review of Employee Concerns Element Report 210.1(B), Revision 1 the staf f concludes that Employee Concerns SON-86-002-03, SON-86-001-03, IN-85-038-001, IN-85-039-001 and IN-85-039-002 have been, in general adequately addressed for rigorous piping analyses for Secucyah restart.
Final resolution cf these concerns is contingent on the resolution of the NRC's Integrated Design Inspection finding on the ERCW themal analysis.
Alternately analyzed piping is addressed in the evaluation of Element Report 218.4(B).
In addition, the new issue raised by ECTG should be reviewed by the NRC staff prior to the Sequoyah restart to detemine whether the new issue has any inpact on the conclusions of this evaluation. TVA's implementation of the commitment to issue new operating drawings shculd be reviewed by the staff as a post restart item.
V.
Addendum See r. ext page.
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l V. Addendum (continued)
The safety evalujtion report for this element report specified the review of the revised employee concerns report as a restart item.
This element report was revised based on the identification of an additional technical concern by the employee concerns task group.
The additional issue involved TVA's failure i
to consider secondary stress range for alternate aralysis at Watts Bar.
The resolution of this item for Sequoyah as described in Element Report 218.1, Revision 2, is to address the issue in the Phase !! alternate analysis progran.
The Phase !! alternate analysis program will be performed af ter the Sequoyah Unit 2 restart.
The resolution of secondary stresses due to stress range considerations in the post restart effort is consistent with the staff safety evaluation on alternately analyzed piping and is acceptable.
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i SEQUOYAH NUCLEAR POWER PLANT,. UNIT 2 TECHtlICAL EVALUATIOtt REPORT FOR EMPLOYEE C0tlCERNS ELEMEt!T REPORT 218,1(B), REVISI0t! 1 "PIPE STRESS CALCULATIONS
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Thermal Analysis of Pipin*g Subjected to Temperature Less than 120 F" t
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SUBJECT:
This report summarizes the NRC audit of TVA investigation of SQN piping operating ~ mode identification, control and evaluation concerns.
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l By: Robert E.
Serb-Consultant NCT Engineering, Inc.
I Date: tiovember 30, 1987 l
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NCT ENGINEERING, INC.
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SEQUOYAH NUCLEAR POWER 'LANT, UNITS 1 & 2 TECHNICAL EVALUATION REPORT FOR EMPLOYEE CONCERNS ELEMENT REPORT 218.1(B), REVISION 1 "PIPE STRESS CALCULATIONS Thermal Analysis of Piping Subjected to Temperature Less than 120 F" I.
Subject Category:
Engineering (20000)
Subcategory:
Pipe Stress Calculations (21800)
Element:
Thermal Analysis of Piping Subjected to Temperature Less than 120 F (21801)
Concerns:
SQN-86-002-03, SQN-86-001-03, IN-85-038-001, IN-85-039-002 The bases for Element Report 21801, Revision i dated December 19, 1986 are Employee Concern Nos. SQN-86-002-03, SQN-86-001-03, IN-85-038-001, IN-85-039-001 and IN-85-039-002.
The concerns regard the identification, control, and pipe stress analysis of pipe system operating and design modes (op modes) for TVA nuclear sites and are potentially applicable to the Sequoyah Nuclear Power Plant (SQN).
II. Summary of Issue The ECTG report has translated the concerns into the following six issues:
i Current operating mcde drawings were not used for all a.
subsequent analyses,
- b. Site group stress analysts were not allowed to evaluate the significance of the current operating mode definitions in the analysis of record.
The environmental temperature in the annulus area may c.
reach 150 F but site group stress analysts were not allowed to evaluate the effect of the environmental temperature on piping in that area.
- d. The operational mode procedure does not require evaluation of previously performed thermal analyses when thermal conditions change, Not all stress-analyzed piping included a code-required e.
evaluation of thermal expansion.
- f. Excessive levels of pipe support loads and pipe stress due to thermal expansion have been observed for some piping where the system operating-temperatures were between 40 F and 120 F and no thermal expansion evaluation was performed.
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j III. Evaluation i
Issue "a" In order to assure that correct op mode data have been used at SQN, the ECTG revicwed the incorporation of op mode data in five calculations and for one of these verified that the proper data was evaluated in the piping analysis.
The ECTG review identified no instances in which erroneous data was used.
61though the ECTG identified no op mode data errors, the concern remains that some piping analyses performed prior to initiation of op mode drawing development may not have used correct I
operating temperatures.
This issue as well as procedural aspects of op mode data development, control, and application are addressed in the discussion of Issuo "d" below.
Issue "b"
To investigate the concern that analysts were not allowed to evaluate "currect" op modes, the ECTG separately interviewed four TVA pipe stress analysts.
All four analysts said they were not aware of any instance in which anyone was instructed not to use current op mode drawings.
Issue "c" The ECTG identified TVA Drawing 47E235-47, Revision 3 as the basin of containment annulus temperatures and that drawing was reviewed during the NRC audit of this concern.
The drawing defines the SQN containment annulus maximum normal and abnormal environmental temperatures as 120 F.
That drawing also defines a maximum faulted condition environmental temperature of 134 F.
SGN FSAR Table 3.9.2-2 specifies Safety Class B,C & D component i
support load combinations.
That table does not specify a thermal load for the faulted load condition.
FSAR Table 3.9.2-3 I
specifies stress limits for Safety Class B,C & D component load conditions and states that expansion stresses need not be evalunted for the faulted condition.
The FSAR is consistent with the requirements of the ASME Section III code which does not l
require evaluation of piping thermal expansion for plant faulted
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conditions.
Therefore, the ECTG conclusion that annulus area I i
g piping environmental condition thermal analyses should be based on the normal / abnormal maximum 120 F temperature is considered aceptable.
Evaluation of pipe thermal expansion due to the normal / abnormal 120 F annulus and other temperature conditions is addressed in the discussion of Issue "e"
of this report.
Issue "d"
The ECTG reviewed the Operating and Design Modes section of the SQN Rigorous Analysis Handbook, Section No. SQN-RAH-207 to verify that TVA procedures require piping design reevaluation when new op mode data is issued.
That procedure section requires design and operating modes to be issued in accordance with Mcchanical Design Guide DG-H5.1.1.
It also requires op mode information to be issued on drawings which must be referenced on all new analysis and reanalysis problem isometrics.
Therefore, consideration of current op modes has been proceduralized for analyses performed subsequent to issuance of the handbook.
In 1984, NCR SQNCEB8205 addressed three areas of nonconformance regarding issuance, revision control, and incorporation of op mode data in pipe analyses for analyses performed prior to the implementation of SQN-RAH-207.
The NCR identified the requirement to include op mode handling methodology in the Rigorous Analysis Handbook.
The adequacy of op mode data used in analyses prior to the use of op mode drawings was also addressed in the NCR.
The data was determined to be adequate based on the the following as excerpted from the NCR:
-WBU [Watta Bar), a Westinghouse sister plant to SQN, is a section III plant.
Availability of operating modes and design transients for WBN has resulted in a better understanding and better analysis of SQN systems than would normally be expected for that vintage plant.
-Operating temperature and pressures were furnished by Westinghouse and systems engineers.
As indicated the controlling equipment for these data was firmed up very early, and significant changes have been evaluated.
-In an effort to further ensure operating mode adequacy for SQN, a commitment was made to review any discrepancies determined by the WBN sampling program for their effect on SQN operating modes.
A sample of WBN problems was.docided in part because the more detailed consideration of operating modes at WBN would have turned up nonconforming items in some cases where the SQN procedures would not.
-The WBN sampling program was performed and a WBN report.
NCR WBNCEB8215 R5 (revised final), was completed March 7, 1984.
Only one problem was located for WBN (CEB 831025 003) and was immediately evaluated for SQN.
The problem was determined to be qualified for SQN (PHP 831202 016).
No other dinerepancie2 were noted.
Therefore, the adequacy of the consideration of SQN operating modes has been assured.
Based on the results of the HBN sampling program evaluations which demonstrated acceptable piping stresses, the ECTG concluded that thermal operating modes had been adequately defined for SQN.
During the NRC audit of employee concerns TVA commited to issue op mode drawings for all Unit 2 rigorous analysis prebloms by December 1988.
This was documented in the undated TVA Memorandum from H.J.Kagay to Rick Daniels; SEQUOYAH NUCLEAR PLANT
- UNIT 2 - GENERATION OF OP MODE DRANINGS FOR RIGOROUS PIPING ANALYSIS PROBLEMS, Per that memo, the potential effect of op mode drawing data on piping problems will be evaluated and the problems reanalyzed as required.
Issuo "e" Consideration of thermal expansion for SQN Alternate Analysis scope piping is addressed by TVA Element Report 218.4(B),
Revision 2.
Apparently, thermal expansion was not always considered and, therefore, this concern is valid for alternate analysis scope piping.
TVA has developed an alternate analysis review program to address deficiencies in alternately analyzed piping and this program has been the subject of other NRC review.
The discussion which follows is therefore limited to TVA rigorous analysis scope piping.
i TVA Design Criteria No. SQN-DC-V-13.3, Detailed Analysis of Category 1 Piping Systems, dated March 10, 1975 requires piping systems analyses to consider design and operating conditions.
The TVA memorandum f rom R.O.Barnett to CEB Files dated March 20, 1987, SEQUOYAH NUCLEAR PLANT (SQN) - DESIGN INPUT MEMORANDUM FOR DETAILED ANALYSIS OF CATEGORY l PIPING SYSTEMS, SQN-DC-V-13.3 specifies that thermal ranges for limiting operating modes are considered.
In addition, SQN Rigorous Analysis Handbook, Section SQN-RAH-207 requires evaluation of all op modes.
Therefore, TVA piping analysis procedures do require evaluation of thermal expansion for rigorously analy=ed piping.
The NRC Integrated Design Inspection (IDI) identified a deficiency in the implementation of these procedures.
An NRC j
letter to Mr.
S.A. White dated October 9, 1987, "ITEMS IDENTIFIED BY THE INTEGRATED DESIGN INSPECTION REQUIRING RESOLUTION PRIOR TO l
RESTART OF SEQUOYAH UNIT 2" identifies a "Draft Deficiency" regarding the failure to include a 35 F cold thermal modo in the op modes defined for an ERCH piping problem.
Evaluation of TVA operating mode procedure implementation should remain open pending resolution of that deficiency.
Issue "f"
Consideration of operating temperatures between 40 F and 120 F for SON Alternate Analysis scope piping is addressed by TVA Element Report 218.4(B), Revision 2.
Concern regarding consideration of all operating condition thermal expansion effects for rigorously analyzed piping is addressed at Issue " e" above.
1 IV. Conclusions Baned on review of the ECTG report and TVA criteria for
'rigoroun analysis piping system qualification, the ECTG evaluation of Issues "a",
"b",
and "c" is considered acceptable.
i TVA has commited to issue operating mode drawings for all Unit 2 piping systems.
This should assure evaluation of these piping i
3 systems for current operating modes.
It is recommended that evaluation of the drawing data and performance of resulting i
reenalyses, if any, be completed in a timely manner.
The ECTG evaluation, and the op mode drawing development and evaluation are sufficient to consider Issue "d"
evaluation acceptable.
The IDI dificiency regarding failure to consider the correct operating temperature'affects conclusions regarding Issues "e"
and "f".
The evaluation of these issues will be considered acceptable upon resolution of the IDI finding, During the NRC audit of these concerns, an additional issue 3
was identified by the ECTG based on receipt of a revised version of an employee concern.
Therefore, additional review of this concern will be necessary when the revised ECTG evrauation is completed.
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