ML20153B096
| ML20153B096 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 03/11/1988 |
| From: | NRC OFFICE OF SPECIAL PROJECTS |
| To: | |
| Shared Package | |
| ML20127A683 | List:
|
| References | |
| NUDOCS 8803210378 | |
| Download: ML20153B096 (9) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION y
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,8 SAFETY EVALUATTON BY THE OFFICE OF SPECIAL PROJECTS FOR EMPLOYEE CONCERNS SE0VOYAH NUCLEAR POWER PLANT UNITS 1 AND 2 ELEMENT REPCRT EN 215.1(B) "CIVIL / STRUCTURAL DESIGN-SEISMIC 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 Chattanooga to north of Knoxville.
If there were an earthquake, power plant structures could fail.
II.
SUWARY 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 Nuclear Power Plant is the Kingston Fault, which lies about one mile to the northwest of the plant at its closest approach.
This fault is about 150 miles long, strikes ncrtheast 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 from several tens of miles to more than 100 miles.
They were formed during the Applachian Orogeny in the late Paleozoic Era (more that 250 million year ago).
There is r,o 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 hg 2gDO 28 p
PDR in the Clinch River SER (Reference 2) to supoort 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.
Radiometric 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 mechtnisms 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 regiun and concurs with the conclusion drawn in the subject element report.
V.
REFERENCES 1.
NUREG-0011, "Safety Evaluation Report Related to Operation of Sequoyah i
Nuclear Plant, Units 1 and 2" dated March 1979.
2.
NUREG-0968, "Safety Evaluation Report Related to Construction of Clinch i
River Breeder Reactor Plant" dated March 1983.
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SEQUOYAH NUCLEAR POWER PLANT, UNITS 1 & 2 SAFETY EVALUATION REPORT FOR EMPLOYEE CONCERNS ELEMENT REPORT 215.2-SQN "CIVIL / STRUCTURAL DESIGN, CUT REBAR' CONTROL" i
I.
Subject l
Category: Engineering i
Subcategory: Civil / Structural Element: Civil / Structural Design, Cut Rebar Control Concerns: IN-85-297-005, IN-85-868-004 i
The bases for Element Report 215.2-SQN, 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 drills due to penetrations of ducts, conduits and pipes.
These concerns were evaluated by TVA as potentially nuclear safety-related and potentially applicable to Sequoyah (generic).
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 /essessment 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 proced9ral 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 to (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 do-"menting 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 t.ie auxiliary building at the Watts Bar Nuclear Plant (VBN) because the concrete (esign 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 criteria SON-DC-V-1.3.3.1, and *he combination of dead, live, and FSAR OBE or SSE loads was considered.
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TVA has completed 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 walkdown of the plant and audited a representative sample of the results of TVA' implementation.
In addition, TVA was re-quested to compare the percentage of cut rebars between the SQN and WBN reactor buildings based on the available data from both plants. 'The.com-I 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 assumption for pre-restart CAP item (3) regarding the similarity in per-centage of cut rebars between the SQN 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 suf ficient safety margins *zith respect to the seismic' loads from the site-specific (84-percent 11e) 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-percent 11e SSE.
Based on the above t
svaluations, the NRC staff four.d the scope and implementation of the pre-i restart CAP to be acceptable.
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 cc:.plete, (2) revise Section 3.8 of the j
Sequoyah FSAR to clarify the use of the ultimate strength method as specified in design criteria SQN-DC-V-1.3.3.1 for the evaluation of the reactor building and auxiliary building because,the ACI working stress ~
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method was the ori;inal FSAR criteria for the design of concrete for these two buildings, and (3) evaluate and document future cut rabar requests based on procedures developed from the pre-restart CAP.
The NRC staff found the scope of the post-restart CAP to be sufficient.
IV.
Conclusions The NRC staf f reviewed TVA's investigation of the employee concerns and the CAF develop:d by' TVA to resolve such concercs, and f ound they were
. adequate.
TVA's implementation of the pre-restart CAP was also found 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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9 SEQUOYAH NUCLEAR POWER PLANT, UNITS 1 4.2 SAFETY EVALUATION REPORT FOR EMPLOYEE CONCERNS ELEMENT REPORT 215.6-SON "CIVIL / STRUCTURAL DESIGN, HANGER LOADS 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 Emplcyee Concerns IN-85-220-003 and IN-86-173-001 which r
questioned the structural integrity of the supporting walls / 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 con; erns were evaluated by TVA as potentially nuclear safety-related and potentially applicable to Sequoyah (generic).
II.
Summary of Issues The stated concerns as defined by TVA are: (a) structusal integrity of concrete valls/slacs in the annulus area of the Unit 2 reactor building is questionable due to excessive number of hangers; and (b) design cal-culations have not evaluated individual and cumulative effects of hangers' on concrete walls / slabs.
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III. Evaluation Investigattons 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 cor.:urred with the findings f rom 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 corcective 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 val), 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 overed 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-SQN, 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-sidering the conbination 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 eieraents were found acceptable as discussed in the staf f saf ety evaluation for Element Report 215.2-SON, Revision 1.
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Regarding-the post-restart CAP, TVA has committed to (1) 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 ths 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 ltem (3).
The NRC staff found the scope of the the' post-restart CAP to be sufficient.
IV.
Conclusions i
The NRC staff revleved TVA's investigation of the employee concerns and the CAP developed by TVA to resolve such concerns, and found they were i
adequate.
TVA's implementation of the pre-restart cap was also acceptable, i
The NRC staff therefore believes TVA's rtsolution for the concerns as described in Element Report 215.6-SQN, Rev. 1, is acceptable.
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