IR 05000260/1989039
| ML19325D459 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 10/10/1989 |
| From: | Cheng T, Hou S, Terao D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19325D457 | List: |
| References | |
| 50-260-89-39, NUDOCS 8910240178 | |
| Download: ML19325D459 (13) | |
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NUCLEAR RESULATCRY COMMIS560N t
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l UNITED STATES NUCLEAR REGULATORY C0m!$$!0N
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OFFICE OF NUCLEAR REACTOR REGULATION
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TVA PROJECTS D1Yl$10N
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Report Not 50-260/89-39
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Docket No:
50-260
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Licensee:
Tennessee falley Authority I
6N 38A Lookout Place
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1101 Market Street
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Chattanooga, Tennessee 37402-2801
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Facility:
Crowns Ferry Nuclear Plant Unit 2
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l Inspection At:
General Electric Company San Jose, California
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.o Inspection Dates:
Augustif-16,1989 a'
1nspectors:
_f)] As _{.I
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ICbh9 T. Cheng ~
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Consultants:
T. Tsai, G. Breidenbach Approved by:
'CM E loc
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D. Tereo, cnter Date
Engineering Branch i
TVA Projects Division Office of Nuclear Reactor Regulation
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8910240178 891013 gDR ADDCK 05000260 u
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i SPECIAL INSPEC1!ON l
i RELATING 70 BROWNS FERRY UNIT 2 SE!$NIC DESIGN PROGRAM
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1 BACKGROUNO i
i On April 24-25, 1969, during an inspection at General Electric of fices the NRC l
staff identified a modeling error in the nuclear steam supsly system seismic analysis which was recently perfomed as part of tie licensee {NSS$)
i s restart
activities for Browns Ferry Unit 2.
The error involved assuming the existence i
of seismic lateral restraints on the control rod drive (CRD) housing in the seismic analysis model when, in fact, none was installed at Browns Ferry. The i
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issue was identified as an unresolved item (URI 89-31-01) in the NRC Inspection Report IR 50-260/89-31 cated July 17, 1989.
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InaletterfromJ. Jones (GeneralElectric) tor.J. Smith (TVA)datedMay19, 1989, GE discussed its conclusion that the unrestrained CRD housing arrangement
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is in compliance with the Browns Ferry FSAR licensing basis. General Electric stated it was continuing its evaluation under 10 CFR Part 21 which showed that the CRD housings might exceed the licensing basis allowable stress limits when i
current seismic evaluation techniques are selectively updated as they are presently being done under the Browns Ferry restart activities. General
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Ilectric's Part 21 evaluation also would address the generic implications of the modeling error on othsr boiling water reactors (BWRs).
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On June 28, 1989, the licensee discussed with the staff its proposed corrective i
action to install a set of swirmic restraints to tie the bottom flanges of the CRD housings to the reinforceo concrete pedestal of the reactor pressure vessel
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l as documented in IR 50 260/89-31.
i 2 SCOPE l
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On August 14-16, 1989, tM NRC staff perforred an inspecticn of the final GE l
Part 21 evaluation and ruiewed the rvsclutica to the open itees identified in
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IR 60-260/89-31. A list of persons contacted is prwfdui in Appandices 1 and
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j 2 to this inspection report.
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3 INSPECTION FINDINGS l
3.1 Par (QJvaluation of the CRD Housing Modeling Error t
The staf' inspected the safety iroact of the modeling error on Browns Ferry
and the pneric implications of tte error on other SWRs.
In assessing the l
l impact on Browns Ferry, the staff reviewed the results of the analysis for the CRD housing to detemine the stresses and deflections under a safe-
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shutdown earthquake (SSE) when the CRD housing is unrestrained.
The staff also reviewed early GE operability tests to detemine the impact of the CRD housing deflections on the ability of the CRD tr.echanisms to perform their safety-related function.
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i 3.1.1 Imoact of Browns Ferr.y Modeline Error
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The original analysis of Browns Ferry 2 reactor pressure vessel and attachments was performed in the late 1960's and the results are reported in Specification
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No. 21A1111. Revision 9. ' Reactor Pressure Vessel". General Electric Company l
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dated August 26, 1970. The original seismic analysis discussed in the FSAR l
was based on the following design criteria.
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1.
Relatively simplified model of the reactor pressure vessel and internals i
as compa md with the new model currently used by GE.
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2.
North-south component of May 1940 El Centro warthquake ground motion f
time history.
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3.
Damping ratios as specified in the FSAR.
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No consideration of flexibility of structural elenents in vertical direction.
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Hydrodynamic water mass considered in the RPV model.
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1965 ASME Code stress criteria.
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7.
Operating Basis Earthquake (OBE) load combination classified as a normal / upset condition (Service Level B) with an allowable stress intensity limit of 1.0 Sm.
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Safe Shutdown Earthquake (SSE) load combination classified as an
emwrgency conditien (Service Level C) with an allowable stress
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intensity limit of 1.5 Sm.
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In the recent seismic reanalysis pertorned by GE for the Browns Ferry 2 reactor i
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pressure vessel, an updated and more refined analysis rodel was useo together
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with the El Centro earthquake ground motion time history as input. This retnelysis incorporated the original licensing basis criteria such as FSAR specified damping, stress allowebits and load combinations.
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The updated RPV reenalysis stroneously inr.luded lateral restraints 6t the CR0
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housings resultind in insignificant deflections and stresses for the housings.
Af ter it was discovered that the CRD housings were in fect not restrainted, the
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RPV model was revised to reflect the as-built condition and this model was
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reanalyzed. The results of this analysis are presented in "Sumary Report on
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Small Tas!. G066.204, Generation of Seismic Response Results for Reactor Building Drywell and Internals of Browns Ferry Nuclear Plant Using El Centro Time History Input (Revised RPV Model)", Bechtel North American Power Corporation dated
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May 10,1989 - (Bechtel Sumary Report).
It became apparent that the stresses at
the CRD housing connections to the reactor pressure vessel were higher than FSAR allowables for OBE and SSE leadings.
This then became the cause for the 10 CFR 21
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evaluation.
In order to demonstrate that the CRD housings are to be functional, comparisons were made to structural evaluation criteria which are currently in effect.
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The following updated criteria were considered:
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Regulatory Guide 1.60 response spectra
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Regulatory Guide 1.61 damping values
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1986 ASNE Code stress criteria l
OBE load combination classified as Service level B with an allowable stress intensity limit of 1.5 Sm.
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5.
$$E load combination classified as Service Level D with an l
allowable stress intensity limit of 3.6 Sm.
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6.
3-D earthquake and combining spatial responses by SRSS technique, i
The Part 21 evaluation conducted by GE, in order to demonstrate that no substantial safety hazard existed during a seismic event at Browns Ferry 2 due to the lack of lateral restraint of the CR0 housings, consisted of both
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recent seismic analysis of the CRD housings and functional testing which was
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conducted in 1965 and 1971 using production control rod drives and housings,
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The analysis evaluating the seismic adequacy of the CRD housings is presented in a report entitled, Browns Ferr Seismic Adequacy of CRD Housings",y 2 Potential Reportable Condition -by dated July 7, 1989. The analysis of the CRD housings is based on the loads from the Bechtel Susmary Report. The results of the analysis show that the CRD housings, both for OBE ar.d SSE, are overstressed when compared to the FSAR allowables(1965ASMECode).
The actual values are as follows:
Calculate,1 Allowable Stress Stress j
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OBE 19,734 pri 16,925 psi (SM)
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SSF 34, 060 psi 25,400 psi (1.5 Sm)
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hhen the OBE and SSC stresses based on the updated criteta are ccmpared to the
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1986 ASME Code allowables, the results arv as folicws:
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P Calculated Allowable Stress _
Stress OBE 24,986 psi 25.013 psi (1.5 Sm)
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SSE 44,564 psi 60,030 psi (3.6 Sm)
The differences in the calculated stresses, although based on the same leads r
obtained from the Bechtel Summary Report are due to the fact that the FSAR
requires only the use of the larger response from the two horizontal analyses combined with the vartical for calculating stresses, while current criteria
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requires the combination of the two horizontal and the vertical components by the SRSS method.
In the present case the vertical component, which was very
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small compared to the two horizontal components, was added directly. There is also a small difference in the Sm values for the CRD housing material (SA-312
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TP 304) between the 1965 and the 1986 ASME Code. The 1965 Code had a value of 16, 925 psi while the 1986 Code has a value of 16,675 psi at the operating temperature.
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l GE justified their analysis approach and the use of the 1986 ASME Code for the CR0 housing evaluation based on the following considerations:
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The new detailed model of the reactor pressure vessel and internals is based on current methodology.
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The new load analysis follows current practice by considering the
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i effects of Regulatory Guide 1.60 response spectra and Regulatory i
Guide 1.61 damping values.
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3.
Current methodology considers the SSE load combination to be a j
faulted event.
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The original FSAR model and the new detailed model are dynamically
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similar because the fundamental frequencies of the major components
in both models are close to each other.
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The difference between the average peak response from the one-percent
damped Regulatory Guide 1.60 Response Spectrum and the average peak
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response from the one-percent damped El Centro time history response
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spe:trum at the CRD housing frequency range of interest is within 101.
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6.
The damping values used in the original Browns Ferry 2 analysis are
based on the FSA9 and are generally much less than the damping values
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7.
Using the new refined RPV and internals model with Regulatory Guide 1.60
Response Spectra and Regulatory Guide 1.61 Damping Values would result in lower loads than using the new model with tie cSAR design spectra
i and damping values. Therefore it is conservative to use the loads from the Bechtel Sumary Report for the stress evaluation.
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i In addition to the analysis, GE also presented the results of functional
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testing which was perforced to demonttrate operability of the CRDs under
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simulatec earthquee unditions.
The testina was done in 1965 and 1971
using CR0 models of similt r design &nd functionally identical to the CRDs
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used in Browns Ferry 2.
The testt were conducted under a simulated reactor
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environment of pressure and tempercture.
The test results are documented
in the following reports
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1.
Browns Ferry CRD New Loads Assessment. E. Y. Gibo, General Electric Company, June 1989.
2.
APED-4815. Control Rod Drive Perfomance Under Simulated Earthquake Conditions, J. E. Benecki, Gsneral Electric Company, February 18, 1965 (1965TestReport).
3.
Document No. 383HA617, Rev. O, Evaluation of CRD Scram Characteristics
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Under Simulated Earthquake Conditions, General Electric Company.
December 8, 1971.
The tests which most closely simulate the Browns Ferry 2 CRD housing seismic response in the unrestrained condition are reported in the 1965 Test Report.
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During these tests, a 1-inch (2-inch " peak-to-peak") displacement was intro-duced at the drive flange at a forced excitation rate of about 3.8 cycles per i
second. The calculated horizontal acceleration corresponding to the flange
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displacement and natural f requency is 1.5g. The test data indicates that the l
mean scram time from scram circuit voltage interruption to 90% of the full i
144-inch stroke increased to 2.19 seconds from 2,09 seconds, still within scram time requirements which are given as 3.5 seconds in the Technical Specifications.
i No changes were observed in normal continuous drive motion or in jog operation.
No damage or abaormal wear was observed during the disassembly inspecticn following these tests.
f For comparison of the test conditions to Browns Ferry 2 seismic displacements,
it should be noted that the OBE displacenents at browns Ferry 2 are about
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0.65 inches and the SSE displacoments are about 1.3 inches. Therefore, the comparison with test data indicates that the displacements expected at the CRD
housings during the OBE event would not produce a measurable effect on scram perfomance. However, the test data does not conclusively demonstrate operability under $$E conditions.
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GE stated that scram performance will not deteriorate significantly during the SSE event for the following reasons:
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1.
The CRD will continue to insert the control rod while the CRD
housing oscillates through the zone of zero deflection during the SSh event.
2.
Nomal flanpe spacing between drives is 3/8 inch, and the length of the CRD housings vary from the center out toward the periphery.
This close spacing of the CRD housings and the different natural frequencies would probably impose a considerable limitation on the j
horizontal flange displacements.
The probability of a single CRD
i housing drive flan 3d reaching the full SSE horizontal displacement of 1.3 inches is low.
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3.
Since the tests demonstrated insignificant )erfomance degradation
I ar.d r.o internal damage or abnormal wear witi deflections of 1",
it is expected that, within limits CRD 1r.sertability would not be significantly impaired with deflections in excess of 1".
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The CRD housings with the largest seismic displacement are at the j
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periphery of the reactor where the control rods have the least
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reactivity worth. Since only 50% of the rods are required for
reactor shutdown, nomal operability of the outer rods is not absolutely essential.
The GE conclusions based on their Part 21 evaluation were that the CRD housings at Browns Ferry 2 are not overstressed as a result of an OBE or SSE when current structural evaluation criteria are consistently applied.
This is demonstrated by the analysis which used a response spectrum which is very close to the Regulatory Guide 1.60 response spectrum at the CRD housing frequency and used lower damping values than those currently
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Therefore the loads obtained in the
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analysis are higher than thosa which would be obtained using current
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allowable damping values. GE believes that the use of the 1986 ASNE
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Code stress criteria and the reclassification of the SSE load combination as Service Level D is justified for this evaluation based on the analysis approach which was used.
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In addition to structural consideration, GE concluded that functional tests I
have demonstrated that for an unrestrained CRD housing subject to cyclic l
deflection of 1", there is an insignificant effect on CRD scram perfomance.
The maximum calculated deflection of an unrestrained Browns Ferry 2 CRD housing subject to an OBE is 0.65".
Consequently, it is concluded from the i
above test results that an OBE will have an insignificant effect on the scram perfomance of Browns Ferry 2.
L During an SSE, the maximum oeflection is expected to double to a value of about 1.30". which is beyond the bounds of the test parameters. The CRD during an SSE would be expected to insert for the reasons given above.
t GE concluded that under the Browns Ferry 2 loading conditions, the CRDs will i
maintain their functional capability to insert the control rods during an
OBE or SSE seismic event in a manner which would not endanger the health and safety of the public.
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As a result of the Browns Ferry 2 Part 21 evaluation, GE reviewed other BWR
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plants to detemine if any updated structural evaluation was performed on the reactor pressure vessel and internals erroneously incorporating CRD housing lateral restraints in the analytical model. GE investigated all 24 Mark I BWRs and found that CRD housing seismic restraints were provided to 9 BWRs.
GE stated that of the remaining 15 BWRs.14 plants had seismic analyses that
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l did not assume nor show the need for CRD hout:ing seismic restraints. The only BWR still in question was Vermont Yankee which had been reenalyzed using seismic
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restraints at the CRD housings when in fact these were not installed. An updated recirculation piping evaluation for the Vermont fankee plant erroneously assumed the existence of lateral restraint springs between the CRD housings but not
between the CRD housings and reactor pedestal.
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The purpose of the Yerinont Yankee re-evaluation was to obtain refined seismic i
l response spectra at various points on the reactor pressure vessel for a recircu-lation piping analysis.
Forces and moments were not generated for the reactor
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l pressure vessel or associated internals.
A review by GE of the resulting analysis indicated that t.he incorrect modeling of the CRD housing lateral springs would have no significant effect on the spet.tra generated.
I GE concluded from the review of other BWR plants described above that no
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l substantial safety hazard has been identified in other BWR plants.
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i 3.1.2 Conclusions
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The staff concludes that the manner in which General Electric evaluated the i
i safety impact of the Browns Ferry unrestrained control rod drive housings was l
adequate. The staff review of General Electric's evaluation finds that the operability of the control rod drive mechanisms under a safe-shutdown earthquake would not have been impaired based on control rod drive housing deflections
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obtained from operability tests performed by General Electric in the 1960's and
I 1970's.
l Furthemore, the control rod drive housing stresses at operating conditions were found to satisfy allowable stress limits as permitted in later editions of i
the ASME Boiler and Pressure Vessel Code Section III (1986).
Thus, the modeling error did not cause and would not have caused the reactor to be placed i
in an unsafe condition.
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The staff also concludes that the manner in which General Electric evaluated the generic implication of the control rod crive modeling error was adequate.
The list of all boiling water reactors compiled by General Electric shows l
which plants were supplied centrol rod drive housing restraints (and which
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were not) as well as which plants were modeled with restraints (and which were
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not). This inforw6 tion provides sufficient evidence for the staff to conclude that the modeling error found at Browns Ferry was not couaitted at other BWR plants except for Vermont Yankee as discussed above.
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r Therefore, the staff concludes that the actions taken by General Electric i
in response to the modeling error found at Browns Ferry 6re appropriate l
and comprehensive to resolve the staff's concerns. The unresolved item (
(URI89-31-01) is closed.
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3.2 Resolution of Open items
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As documented in JR 50-260/09-31, three items fdentified remained open:
(1) j'astifications of using Methnds Non. 2 and 3 for the evaluation of NSSS
components, (2) redocumentation of Appendix J to the FSAR which documented the
original design basis and criteria for the BFN-2 primary system, and (3) incorrect
dynamic model of the control rod drive (CRD) housing (including supports) used
for generating seismic design loads to qualify the primary system. The following i
subsections document the staff's review results of these open items.
i 3.2.1 Methods Used for Impact Evaluation and Evaluation Results In order to correct the errors made in the evaluation of NSS$ components due to the use of the incorrect dynamic model (model with the CRD housing restraints).
TVAContractorBechtelPowerCorporation(BPC)usedthecorrectprimarysystem model (model without the CRD housing restraints), coupled with reactor building model, and regenerated a set of hSSS component seismic member forces.
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set of seismic loads was, following the proper procedures, transmitted to GE
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for the reevaluation of these components.
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When the primary system components (fuel elements, stabilizer, vessel stabilizar bracketandadjacentshell,reactorpressurevessel(RPV)supportskirtand ring girder, shroud support, top guide, core support, incore housing, control rod drive (CRD), orificed fuel support, control rod guide (CRD) tube. CRD
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housing, loads obtained from sechtel Power Corporation (BPC) a compon and CRD housing support) was originally evaluated by GE with the seismic considered qualified if one of the following methods was satisfied:
Method (1) The new seismic load or stress is less than the original
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seismic design load, stress or allowables specified in the FSAR.
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Method (2) The new seismic load or stress it less than the design
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load or stress of the similar component in other licensed j
plants of the same vintage as Browns Ferry for which GE i
was the NSSS supplier. "his approach is referred as a
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" sister plant approach".
l Method (3) The new seismic load or stress is less than the design
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capacity established generically by GE for the same i
component.
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I As documented in IR 50-260/89-31, the staff concluded that Method No. 1 is acceptable and Methods Nos. 2 and ' are also acceptable provided that
additional information should be provided to justify the adequacy of these j
methods.
During this inspection, the staff and its consultant reviewed the new seismic loads generated by BPC suonary (DRF B11-00547)(,B22890811101), GE's calculations and evaluation
and conclude the following:
(1)
Instead of providing additional information to justify its adequacy.
TVA/GE withdrew the use of Method No. 2 for the evaluation of NSSS components. The methods actually used for qualifying each of the i
13 components are tabulated below:
Component Evaluation Method Nos.
i Stabilizer
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Yessel Stabilizer Bracket l
& Adjacent Shell
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t RPV Support Skirt & Ring Girder
l Shroud Support
i Top Guide
Core Support
Fuel Elements 1. 3 Incore Housing
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I Comoonent Evaluation Method Nos.
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Control Rod Drive
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Control Roc Guide Tube 1, 3 l
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CRD Housing
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t (2) The use of Method No. I for qualifying NSSS Components is acceptable to l
thestaff(IR 50 260/89-31)
l (3) The use of Method No. 3 is acceptable provided that GE should identify the source documents for those components to which this method was
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applied.
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(4) The review of GE's calculations and e'faluation sumary found that the
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analysis method and calculation results for the impact evaluation are 1) finalize the document DRF
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acceptable provided that GE should (list presented previously B11-00457 to be consistent with the (Reference 4.1) and (ii) for those components qualified by the Method
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Nos. I and 3 as tabulated in Conclusion No. I above, identify which
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portion of the evaluation was based on Nethod No. 3.
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This item remains open.
j 3.2.2 Redocumentation of FSAR Appendix J By letter dated June 23, 1989 (Reference 5.2) TVA redocketed the original FSAR Appendices J. K and L (the original design basis of BFN NSSS Components)
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which were deleted from the FSAR in 1984.
During this inspection, GE also i
provided a copy of FSAR Appendix J for review.
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The review of this document found that the horizontal dynamic model shown in Appendix J is different from the latest dynamic model used for the seismic load
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generation in three aspects:
(1) the original rodel did not include the lateral
CRD housing restraints which were represented by the spring constants K and K
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in the latest model, (2) the original model did not include the refuel 1 Ag bel 18w, i
and (3) the distribution of the hydrodynamic mass were lumped differently in the two models. The use of the latest model (with the lateral CRD housing restraints
andrefuelingbellow)forgeneratingseismicloadsisacceptablebecauseTVAhas
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committed to install the CRD housing restraints before restart of BFN-2 and the latest model will represent the as constructed condition when the plant is back
to power. The staff and its consultant also reviewed the seismic loads generated
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by BPC based on the latest model and the comparison of the new loads with the
loads documented in Appendix J, and found that, in most cases, the new loads
were enveloped by the old loads except at two locations, the vessel shell and shroud. TVA/GE were asked to verify their design adequacy.
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In summary, this item is considered closed provided that GE should verify the l
design adequacy-of the RPV shall and shroud.
l 3.2.3 Concern and Desien of CRD Housino Restraints To resolve this open item, instead of qualifying the CR0 housing (without restraints) by analysis, TVA committed to install a set of restraints at the lower end of the CR0 housing before restart of BFN-2. During this inspection,
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TVA/GE presented their conceptual design of the restraints. As a result of
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the review, the staff found that the conceptual design of the restraints appears
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i reasonable. According to TVA, the final design of these restraints will be i
completed in mid October 1989 and the installation will be conducted in November
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1989.
The staff will review the final design when it becomes available.
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3.2.4 Conclusion
As a result of the staff's review, two items (justification of the methods used
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for the evaluation of NS$$ components and redocutentation of Appendix J to the
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FSAR) were still open and TVA was requested to provide additional information i
for closing these open items. As far as the CRD housing restraint installation, i
the staff will review TVA's final design during the next seismic design i
program inspection.
4.0 REFERENCES
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4.1 Letter from M. Ray (TVA) to hRC, " Browns Ferry Nuclear Plant Units 1, 2 and 3 - Original Final Safety Analysis Report Appendtx J K and L l
Submittals," dated June 23, 1989.
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APPEN0!X A
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INSPECTIONATGENERALELECTRIC(SANJOSE)
l BROWNS FERRY UNIT 2 SEISMIC DESIGN PROGRAM AUGUST 14, 1989 l
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(ENTRANCEMEETING)
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Wayne A. Massie TVA Browns Ferry Licensing l
Allen R. Smith GE Licensing
George B. Strambach GE Licensing
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Gerry Breidenbach BNL - NRC j
Shou-nien Hou NRC/ME8
Noel Shirley GE Licensing
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Pat Marriott GE I
Rick Cutsinger TVA
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Tom M. C. Tsai NRCConsultant(NLTEngineering)
i N. J. 81glieri GE - Engineering Services
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G. A. Deaver GE - Engineering Services
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T. M. Cheng NRC i
D. Terao NRC/TVAPD I
J. Wallach GE - Plant Analysis
!
E. Y. Gibo GE - Engineering Services
!
!
!
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!
!
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APPENDIX 8 NRC/TVA/GE J
BFN RPY EVALUATION EXIT MEETING AUGUST 16, 1989 Wayne A. P.assie
'TVA BFN Site Licensing David Tereo NRC Thones Cheng NRC r
!
Tom N. C. Tsai NRC Consultant
[
I Rick Cutsinger TVA l
Ned Biglieri SE - Engineering Services
!
Pat Me.rriott GE
'
'
Jerry Deaver GE - Engineering
Jot.n Wallach GE - Engineering Services i
t Allen Smith GE - Licensing l
.
i Bob Mitchell GE - Licensing
!
i Jim Ioakes GE - Engineering
!
D. K Henrie GE - Engineering Services
!
A. K. Kaul GE - Engineering Services
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!
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6-1 H