ML20077C818
| ML20077C818 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 07/20/1983 |
| From: | DEKALB AREA ALLIANCE FOR RESPONSIBLE ENERGY, LEAGUE OF WOMEN VOTERS OF ROCKFORD, IL, SINNISSIPPI ALLIANCE FOR THE ENVIRONMENT (SAFE) |
| To: | |
| References | |
| ISSUANCES-OL, NUDOCS 8307260258 | |
| Download: ML20077C818 (47) | |
Text
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UNITED STATES OF AMERICA - NUCLEARREGULATORYCOMMISSION[-[M'~ ' 2:.:$, ,
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s BEFORE THE ATOMIC SAFETY AND LICENSING BOARD [. Q c
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In the Matter of ) 7,,i av f,P'Se g
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COMMONWEALTH EDISON COMPANY ) Docket Nos. ISO M
) S 0-4_5 5.,OL '
(Byron Nuclear Power Station, ) Units 1 ti 2) -
)
f REVISED LEAGUE OF UOMEN VOTERS OF ROCKFORD, ILLINDIS AND DAARE/ SAFE FINDINGS OF FACT AND OPINION ON LEAGUE CONTENTIONS No. 22 and 0AARE/ SAFE CONTENTION.9 (c). July 20, 1983 l g ~ l 0307260258 830720 PDR ADOCK 05000454 0 PDR ,
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OPINION (League) Contention 22'and Rockford League of Women Voters'- Steam Generator Tube Integrity DAARE/ SAFE Contention 9(c) Prior to issuance of an operating license, the NRC must find reasonable assurance exists "that the activities authorized by the operating license can be conducted without endangering the health and safety of the public" and that such activities be conducted in compliance with t'he NRC's regulations. 100.F.R. & 50 57 (a)(3) section 50 57 (a)(3) is implemented with respect to steam generator tubes by General Design satisfying 10 C.F.R. Part 50, Appendix A Criteria 14-16, 30-32, which in pertinent part states The Criterion 14 -- Reactor coolant pressure boundary. reactor coolant pressure boundary shall be designe low probability of abornmal leakage, of rapidly pro- . pagating failure, and of gross rupture. The reactor Criterion collant 15 -- Reactor collant system design. system a protection systems shall be designed with sufficient margin to assure that the design conditions of there during any condition of normal operation, including
. anticipated operational occurrences.
Reactor containment 16 - Containment design. Criterion and associated systems shall be provided to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment and to assure that the containment design conditions important to safety are not ex'ceeded for (emphasis as long as postulated added)- accident conditions require. Criterion 30 -- Quality of reactor coolant pressure boundary. Components which are part of the. reactor coolant pressure boundary shall be designed, fabricated, erected, and tested to the highest quality standards
-practical . . .
s - Criterion-31 -- Fracture prevention of reactor coclant cressure boundary. The reactor coolant pressure boundary shall be designed with sufficient margin to assure that when stressed under operating, maintenance, testing and postulated' accident conditions the boundary behaves in i a nonbrittle manner and the probability of rapidly pro-pagating fracture is minimized. The design shall reflect consideration of service temperatures'and other conditions of the boundary material under operating, maintenance,
-testing, and postulated accident conditions and the uncertainties in determining material properties,.the effects of irradiation on material properties, residual, 1
steady state and transient stresses, and size of flaws. L
, Criterion 32 -- Inspection of reactor coolant pressure boundary. Components which are part of the reactor coolant a pressure boundary shall be designed to permit (1) periodic inspection and testing of important areas and features to assess their structural and leaktight integrity, and (2) an appropriate material surveillance program for the reactor pressure vessel.
- s Criterion 14 is extremely relevant because the steam
, generator tubes fall under the ambit of the " reactor. coolant pressure boundary." In order to establish steam generator
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tube integrity under-this criterion and under Criteria 30 and 31, Applicant must demonstrate that design measures (such as the technical fix for flow induced vibration) and fabri-cation methods (such as the method of fabricating tubes) and operationel procedures (such as ,the program for adherence to water chemistry guidelines) have been developed and tested, and that detection systems are in place (such as the loose parts monitoring system) so that under either normal operating l conditions or under accident conditions there is an extremely low probability of abnormal leakage, of rapidly propagating failure, of gross rupture, and of tubes-becoming so brittle that their degraded condition would eventuate in the above. Criterion 32, which requires-that reactor coolant pressure
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-boundary components be designed to permit periodic inspection and testing of critical areas, is, in essence, a means of satisfying Criteria 14, 30, and 31. Implementing guidelines to the above criteria have been issued as NRC Regulatory Guide 1.83, " Inservice Inspection of Pressurized Water Reactor Steam Generator Tubes." Criteria 15 is germane because the i balance of plant contai,ned in the~ secondary system comprise
< the associated auxiliary, control and protection systems. In order to establish steam generator tube integrity under this criterion, Applicant must demonstrate that this balance of plant (including condensers, feedwater systems, associated monitors and instrumentation, and' secondary water chemistry control program) have been designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operation. e
- Criterion 16 is relevant because, for. example, it is f essential for both the power operated relief valves on the i
t reactor and the safety valves on the steam generator to function properly in order to ensure that in the event of
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tube ruptures, the reactor operator's emergency procedure options
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l will not be' precluded by the above to the extent that inability to fully exercise these procedures in a timely manner results L - in a violation of this criterion. Improper operation of these valves could result in a release of radiation to the environment in the event of an accident, violating the criter-ion's provision for an essentially leaktight barrier. There- , I fore applicant must demonstrate that emergency procedure,s l will ensure an essentially leaktight barrier against the
r - _4 I l uncontrolled release of radioactivity to the environment. Additionally, Westinghouse steam generator tube integrity. is designated as unresolved safety issue A-3 Pursuant to Appeal Board decisions, individual NRC safety evaluation
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reports must describe those unresolved safety issues relevant I and potentially~significant to the facility under. review and
^ggg7 a systems analysis of why operation can proceed in advance
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(' of an overall solution. It held that the NRC Staff should cphTI A- make clear in the SER its perception of the nature and extent of the relationship between each significant USI and
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,- the extended operation of the reactor under scrutiny. The
.. s furnished information should determine whether (1) the problem 7jt: has already been resolved for the reactor under study; a
- restriction on the level or nature of operations adequate i
to eliminate the problem has been imposed (emphasis added): ac
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l . fj; ; or the safety issue does not arise until the later years of
- c w Gulf States Utilities Co. (River Bend Station, feh.Z operation."
%%HW ALAB-444, 6 -NRC 760 (1977) .
, M .f ti Units 1 and 2) s L fgg , It is against the foregoing guidelines that the Board
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-;gg.ct will weigh'the evidence on the issures raised by Contentions
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pl 22 and 9(c). [ 7] Rockford League of W[ omen Voters' (League) Contention 22, as litigated, provides: .eain < An extremely serious problem occurring at other plants $ 5fdsn such as Consuniers' Palisades plant and C.E. 's Zion plant, 3T70+ ^ ^ and likely to occur at C.E.'s Byron plant, is presented J^$; by degradation of steam generator tube intebrity due to < ~ corrosion induced wastage, cracking, reduction in tube diameter, and vibration induced fatigue cracks. This affects, and may destroy, the capability of the degraded - tubes to maintain their integrity, both during normal
operation and under accident conditions, such as a LOCA or a main steam line break. The Commission Staff has correctly regarded this problem _as a safety problem of a serious na-ture, as evidenced both by NUREG-0410 and the Black Fox test-inony cited above. As a result of this serious and unresolved problem the. findings required by 10 C.F.R.50.57(a) (3) (i) and 50.57(a) (6) cannot be . made. (Finding 1) DAARE/ SAFE Contention 9(c) as litigated, provides: Steam-generator tube integrity. In PWRs steam generator tube integrity is subject to diminution of corrosion, cracking, denting and fatigue cracks. This constitutes a hazard both during normal operation and under accident conditions. Prim-ary loop stress coirosion cracks will, of course, lead to radioactivity leaks into the secondary loop and thereby out of the containment. A possible solution to this problem could involve re (ign of the steam generator, but at FSAR, Section 10.3.5.3.the Applicant notes its intent to deal with this as a maintenance problem which.may not be an adequate response given the instances noted in Contention 1 above. (Finding 1)
-{o. Tog.addNesh the c~onb1ti$ns;,'-Applihnt'
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o.f--eiev'. emyftiesses. ./...Mi. ,. . Jo. .h. it. C'." Bl.omgren of..Comino~nwsItti"td3is'on " Company, addressed various measures that will be employed at the Byron Station to minimize steam generator tube degradation. Mr. Mahendra R. Patel, of Wecinghouse Electric Corp. , addressed the
" leak-before-break" principle and steam generator tube plugging criteria. Other witnesses from Wetinghouse included Mr. Daniel Malinowski, who addressed the inspection measures used to detect steam generator tube degradation; Dr. Michael J. Wootten, Who addressed the. water chemistry measures used tio minimize tube degra-l dation on the secondary side of the steam generators at Byron; Dr. Lawrence Conway, who addressed the design changes in D4 and D5 steam generators at Byron; and Mr. Thomas Timmons, who address-ed the flow-induced vibration phenomenon. Mr. Lawrence D. Butter-field, of Commonwealth Edison, addressed App '.icant's modifications f .,
with respect to the flow-induced vibration phenomenon. Applicant also presented the testimony.ofJf.r.-Kenneth ., s . ..n b J. Green, ~Mcchanical - . - - - : _:- ' ., :- lo...x
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, m Project Engineer ~ for Sargent and Lundy Engineers , who addressed the issue of whether the proposed modifications to the Byron steam I generators described by Messrs. Timmons and Butterfield might in-crease the liklihood of a waterhammer event in the Feedwater By-pass Systems of the Byron steam generators,'and Mr. Rodolfo Paillac f man, a Senier Quality Assurance Non-Destructive Examination Specs , l-,~ ialist with Ebasco Services, Inc., who addressed the pre-service .
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inspection of the steam generator tubes at Byron. Mr. Wilson D. Flet'cher, Manager of Steam Generator Materials ._. and Chemistry of. Westinghouse, provided an overview of the steam
,j. _ generator tube integrity issue and Mr. Michael Hitchler, Manager
!. of Probabilistic Risk Assessment with the Nuclear Safety Departmen9 i s. b f of Westinghouse, quantitatively assessed the probability.of steam
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generator tube ruptures under various conditions. The NRC Staff presented the testimony of Dr. Jai Raj N. Raj an,
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a mechanical engineer, who addressed the flow-induced vibration s ., C #? - phenomenon and the probability of tube rupture under various con-MW Q ditions; Mr. Ledyard B. Marsh, .who addressed- various steam gener-y ator tube degradiation issues; Mr. Louis Frank, a senior materials 17:%: en engineer, who addressed seco'ndary side water chemistry measures and inservice inspections; and Mr. Conrad McCracken, who addressed gy . steam generator design and secondary side water chemistry measures L to reduce corrosion. The League presented the testimony of Mr. Dale G. Bridenbaugh, G9_ H5 O a nuclear engineer and President of MH3 Technical Associates'in California, who addressed various aspects of . steam generator tube integrity issue & concerning the Byron Station. (Finding 2) I
O l The steam generator tubing, which is part of the reactor cool ._ ! l ant. pressure boundary, is an important barrier against the re-r lease of radioacitvity to the environment from the primary system. l Accordingly, design criteria for tube wall sizing have been es-tablished to assure structural integrity of the tubing under nor- ' l mal operating and the postulated design-basis accident condition loadings. (Finding,3)~ The normal wall thi~ckness of a' steam generator tube is .043 inches, which is thought to be sufficient to fit in with the above. (Finding 4) However, the operating experience of PWRs has shown that over a period of time under the influence of the operating loads and
- environment in the steam generator, steam generator tubes can be-come degraded and leak. Implicit in the language of General De-sign Criteria 14-16, and 31, which speak to abnormal leakage, rap-idly propagating failure, gross rupture, leak-tight barrier, un-controlled release, and exceeded marrgins, is the recognition that some degradition and leaY4ge will inevitably occur over the lifetime of a plant.
Given sufficient time and the forces that could be acting on a
, . tube, one mode of tube degradation could be more serious than an-other mode. The type and amount of degradation that occurs is spec-ific to the operating history of the plant. (Finding 5)
Tube degradation problems at W'estinghouse steam generators have included the following: wastage and thinning, corrosion, denting, pitting, intergranular attack, stress corrosion cracking we ar caused by flow-induced vibration, and wear and or impact damage as a result of foreign ' objects or loose parts. (Finding 6)
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Early immoderate use of phosphate water chemistry resulted in-tube wall thinning, a_ localized reduction in the tube thickness resulting from corrosion by phosphates in high concentrations. It , is caused by low sodium to phosphate ratio solutions, or a concen - trated' solution of sodium phosphate which was the corrosive agent which caused the thinning of the tubing. Tube wall thinning has been observed within sludge piles at the top of the tubesheet and at tube. support plate elevations, where lower flow velocities allow-ed concentration of phosphates to saturation levels. (Findings 7,8) The change to all volatile treatment (AVT) water chemistry con-
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trols stemmed the thinning of the steam generator tubing caused by acidic phosphate species. (Finding 9) While the change to AVT mitigated thinning caused by acidic phosphate species, in the immediate years following the conversion to AVT, another form of corrosion called " denting" was observed with AVT use. (Finding 10) Denting is a process whereby corrosive impurities are concen-trated in the space between the tube and tube support. The result-ant corrosion converts the base support metal to metal oxide. . . l. 1 . (Finding 11) . As tubes become more. highly stressed, they are more susceptible to stress corrosion cracking. (Finding 13) Byron Unit 1, D-4 Model steam generator will have carbon sted'
- support plates. Unit 2, D-5 will have ferretric stainless stuel support plates. (Finding 14)
Denting will still occur at Byron Unit 1. It will take from 3-10 years for initiation, depending on how well the water chemis-try guidelines and condenser inservice inspection program are ad-hered to. (Finding 15)
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- Pitting is a form of tube degradation that involves small discrete roughly circular regions of tube penetration typically less than 100 mils in diameter. Pits may occur separately or in bands wherein each pit sits independently of others within the . , '.
band. (Finding 16) One plant (Indian Point) experienced pitting of the Inconel
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tubing which is believed to be due to an acidic chloride condition involving copper and chloride ions.This?. incident wasecaused'by- ': excessive condenser inleakage , oxygen inleakage and the combin-ation acting on a copper alloy condenser and feed train over a protracted period of time. (Finding 17) Stopping condenser inleakage will stop the progression of virtually all corrosion mechanisms. The NRC is considering making it a requirement that condensers be inspected routinely as steam generators are. (Finding 18) Tube wall cracking generally occurs in a local region mad the
- crack may extend through the entire wall thickness. Depending up-on the orientation of tube wall stresses, cracks may initiate from the outside diameter or the inside diameter of the tube. All tube stress-corrosion cracks detected in Westinghouse-designed steam ~ generators have been intergranular in nature. Intergranul-ar attack is a form of tube degradation usually characterized by general grain boundary dissolution. It occurs in conj unction with stress-corrosion cracking on the outside diameter of the tube; it usually occurs within crevices between ths tube and the tube-sheet. (Finding 19) rube. wear is a form of tube degradation that results from a I l
mechanical abrasion.of the tube surface. Such wear progressively
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_10_ reduces the thickness of the tube area affected. Wear results from'the impact of adjacent structures or loose objects on the tubing; it has been observed at antivibration bar intersections , the baffle plates in preheat sections and locations in contact with' foreign obj ects. (Finding 20) Steam Generator Design
' Both Byron Unit 1 and Byron Unit 2 have preheat steam gener-ators which are functionally identical. Byron- Unit 1 is a model
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D-4 and Byron Unit 2' is a model D-5. (Finding 21) To reduce chemical concentration, areas such as at the tube sheet and between the tube and tube support plate, recirculation rates were increased, the ports in the blowdown pipe were modi-fied, and the tubes within the tube sheet hole were expanded to eliminate the crevices.at the tube sheet. (Finding 22) To minimize the tube stresses, the widest space in between
~ ' tube support plates-which is functionally acceptable was selected, the holes in the flow distribution baffle plates and th the top tube support plate were modified. (Finding 23)
! -In addition, the design of the Model D-5 in-Unit'2 has been enhanced over the D-4 in U. tit 1 by (1) utilizing ' stainless steel, l a more corrosion-resistant material, as the material for the tube support plates and baffles, (2) changing the shape of the holes in the-tube support plates from circular to a quatrefoil shape to improve flow, (3) expanding the tubes within the tube sheet by
- means of a hydraulic device in lieu of mechanical rollers to re-duce stresses, (4) thermally treating the Inconel 600 tubes to enhance resistance to corrosion, and (5) changing the holes in tha
flow distribution baffles from slotted to a circular shape to improve flow. (Finding 24) Flow-induced Vibration Since 1981, a design problem with flow induced vibration and subsequent wear of tubes in the preheater section of Model D steam generators has been identified at lead operating facilities I with Model D steam generators. (Finding 25) These include McGuire Nuclear Station, Unit 1 0Model D2) and the following three foreign facilities; Rpnghals, Unit 3 (Model D3), Almarez Unit 1 (Model D3)', and Krsko (Model D4). (Finding 26) Certain tubes in the preheater region of the steam generator i vibrate against the baffle plates as a result of the turbulence created by the feedwater flow entering from the main feedwater nozzle. (Finding 27) . The tube excitation mechanism appears to be a combination of i . a threshold type of fluid elastt c instability and turbulent buff-ering. (Finding 28) The oscillating drag forces are produced in large measure by :..~.. the presence of the impingement plate and the unsteadiness of the flow field in the inlet plenum. (Finding 29) Modei D4 and D5 steam generators employ counter flow type pre-heater design. The split flow design is employed for Model D2 and D3 steam generators. (Finding 30) Tube wear in Model D steam generators resulting from flow-induc-ed vibration was initially identified in Sweden at Ringhals Unit 3, a plant with Model D-3 steam generators in October 1981. The.de-
-gradation was in the form of a through-wall hole in a single tube
i l at a baffle plate. Eddy current testing performed also indicate @l tube wall wear in the outer 3 rows of the preheater tubes. l (Finding 31) ;
-Ringhals had operated at 100% power, 100% main feed for only ,
. I 2000 hours when the tube leak occured. (Finding 32)
Possible tube wear.was also detected at another D-3 plant,
- -m Almarez 1 which had operated only 1500 hours at 100% power in October 1981. (Finding 33)
Eddy current testing was also performed at MGuire 1, a domest; plant with Model D-2 steam generators and at Krsko, a non-domest-plant with Model D-4 steam generators. No indications were found The latter two plants had not operated above 50% power, MGuire o) eratfed for only 1,000 hours and Krsko for 300 hours at 40% r power. _ (Finding 34) S . Based upon the early eddy current testing indications, West-aMR) inghouse started a program to investigate, understand and define w. vibration and tube wear in Model D steam generators using and re<
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v]yh viewing tube wear data from operating plants and from model tests
.wiCs to conceive, develop, test, evaluate any modifications necessary
, [. hj to allow operation of Model D steam generators at full power. (Finding 35)
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p.u.x Accelerometers were placed in the McGuire, Almarez, Ringhal,
'J'" and Krsko plants between. December, 1981, and March,1982, to try to correlate tube vibration data to tube wear data to predict po<
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,; $ngQ;N Y j zw iEi tential tube wear over some operating period. (Finding 36) d; The Accelerometers at Krsko were placed in the 4 tubes thought to be the ones that would be most susceptible to tube vibration,,
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and would exhibit the highest levels of vibration. (Finding 37) In May of 1982, one tube was removed from the outer tube bundle at Krsko. It had 6% tube wear'in one location and 1% in another. This was also one of the tubes in which an accelerometer was placed . (Finding 38) Krsko shut down again in late October, 1982. Two previously in-strumented tubes were_ removed and one tube expanded. The two re-moved tubes had wear indications of 2-4%. Krsko f[t@bl again in November. Accelerometers were installed in the expanded tube. Vi-bration data was gathered at'this point for various feedwater flow configurations. (Findig 39) Atthetime, October,1982,i ese feedwater flow configurations were used to take accelerometer data and measure changes in vibra-tion levels with chang ~es in rn4En. feedwater into main nozzle, Krsko had only between 24 and 48 hours at 100% main nozzle flow. With only one original instrumented tube and a newly expanded tube, Krsko did not have a very sizeable data base for conclusions drawn from it. (Finding 40) Correlations of tube wear with vibration is a very complex func-tion. 90% or 100% main feedflow through the main nozzle contributes much more significantly co wear function than operation at lower power levels. "It is a strong function of the amount of flow", ace ...l..q cording to an applicant witness. . (Finding 41) As part of the generic Model D4/D5 program, a 16 full scale model was used to replicate in the laboratory tne tube vibration response observed in operating steam generators. A single tube vi-bration model was used to characterize tube response under various excitation and support conditions. (Finding 42)
, The model tests were all run at cold conditions requiring the a correction factor be used to obtain hot condition data. (100 -
cold condition) Adjustments in the model to give hot condition support conditions introduce an uncertainty of 10-20% in the - cow relation. (Finding '43) The 2/3 scale model tests did not allow measurement of a crit g ,a ical threshol,d water velocity at which vibration begins. All sc-models that were instrumented for vibration measurement purposes yF were instrumented with strain gauges e'xcept; the 16 .sodel%Ibe ai dUraci.of'the strain gauges is about plus or minus 20%. (Finding Fletcher and Rajan state that the tubes are being moved by a
.s eqpplex mixture of both turbulence and fluid-elastic vibration.
Fluid-elastic instability is present. (Finding 45) The object of ~ the preheat region is to extract more heat from the primary fluid. For that reason a certain amount of turbulene is necessary. However, a side effect of this increased turbulencL
- .~ is vibration of -some of the tubes in that region against the supd
' sr? $ [ port plates. This has resulted in less than acceptable tube wear jh 3
~c. r-on rsome of the tubes. (Finding 46) j[$..-ljic The problem of flow induced vibration has been at the support
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plate location. The support plate location is about a one -.c hwe..;g p@n$bh4j inch thick piece of steel where the tube goes through a hole tha9 GwT has a noninal 20 mil clearance. When the tube vibrates, the tube g n;eson within that support location becomes worn. If there is a one incB % e - support -location and thickness , a one inch spot on the tub'e is
-h -worn. (Finding 47)
It's not possible to measure drag forces on tubes in operating D2-D3 plants . (Finding 48)
41/ Premised upopdts investigation, Westinghouse has recommended
.that Applicant make the following modifications to the Byron plant to reduce the potential for significant tube vibrations in the Byron steam generators: (1) the expansion at baffle-plate loca-ko- tions in the prehe.ater region of approximately 100" tubes per steam generator and(2) the bypassing of approximately 10% of the
{: flow from the main feedwater nozzle to the auxiliary feedwater noz-zle.uThe' bypassing of -101cof_the main feed flow to the auxiliary nozzle of the steam generator will ~reduca the main feed flow at the inlet to the preheater to approximately 90%. (Finding 49) i The tube expansion process is scheduled to begin in mid-July prior to start up of the Byron Station. Changes to the control
- circuitry of the feedwater. preheater bypass valve and - the. instal-lation of a feedwater bypass line flowmeter will -o'ccur at' the same time as the tube expansion program. (Finding $0)'
( The tube expansion involves the insertion of tools into the -
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tube from the primary side of the steam generator tube sheet. The , E tubes are then used to locate the baffle plate intersection and . _x __ to expand the tube at the appropriate location. (Findingc51)n The 100 tubes identified as candidates for expansion at Byron I are regarded as a bounding number. It is not actgally known which tubes will be selected. ,(Finding 52) The precise location and number of tubes to be' expanded is not yet known. (Finding 53) - Timmons admits that a small number of expanded tub'es may' ave to be plugged over the lifetime of the plant. Timmons admits that [ some : of ~ the expanded steam generator tubes may wear to the po' int n -
-that they have to be plugged. However, projections of how many are only preliminary. (Finding 54)
Timmons saya some of his colleagues at Westinghouse may have specific tubes picked out which may have to be plugged due to vi-bration wear. (Finding 55) Later Timmons changed his testimony to remark that he now be-lieves no tubes will.have to be plugged at Byron over its life.
- He said a small fraction would have to be plugged. He says that he has received other information that indicates expanded tubes and bypassing feedwater flow is such that vibration would remain below a threshold level that would lead to tube plugging.
(Finding 56) It would take a maximum of five weeks to expand the one-hundred f tubes in each steam generator at the Byron plant and install the l flow metering devices . Five weeks it the maximum period necessary to complete the installation of the modification for flow induced vibration. (Finding 57) /
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The cormittment to install the tube vibration modification l dxpniision of abooth100 tubes /1,07. feedwater flow bypass to auxil-
,, liary nozzle) -before plant startup is a firm committment on behalf p of efG ' company (Commonwealth Edison)' which the company's attorney
- - ,:(Mr.' Joseph G' allo) reaffirms. (Firiding 56) j' -
7-The Atomic [$rfety and . Licensing Boards Appeal Board has recog-e
. nized that a committment made during adjudication is as enforcible
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.as a tech spec or li' censing cond3tionS.t (Finding ~59)
Judge Smith has accept'ed Commonyealth Edison's committment to
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the instillation of the sube vibhtion modification (expansion of
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Labout 100-tubes /10% feedwater flow bypass to auxilliary nozzle) as a disposition of that particular issue under the understanding
- that the committment has a force and effect as a license condi-tion. (Finding _60)
Timmons didn't know how many more' hours it would.take for the tube with 6% degradacion removed from Krsko in May to reach 40% degradation given the same power level ranges for the period- of
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January-May of 1982. (Finding 61) . Westinghouse intends to install accelerometers at the first plant at which the proposed modification is to be implemented in order to confirm that the vibrations seen in that plant are representative of those.seen in the test models and at Krsko. (Finding 62)
. Timmons did not know wheder a tube wear rate had been _ calculate '
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.. ed for the expanded tube at the Krsko plant. (Finding 63)
There is a possibility Byron may be that first instrumented plant. According to Timmons, Comanche Peak Unit 1 will be the first operational plant instrumented not Byron, but he is wrong b on the Comanche Peak fuel load dhte. Staff believes Byron and Com-anche Peak- will load Fuel in December of 1983. (Findings 64,65) Timmons states that instead of removing the expanded tube at Krsko in June 1983, it might be better to leave it as is and see whateffectexpansionof150othertubesaroundithasonthevi-
'bration of that tube. This still is not known. (Finding 66)
Special inspections were done over the past 6-9 months at McGuire on the tubes in the pre-heater area. (Finding 68) Staff indicated an enhanced inspection program like that at
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McGuire will be required for tubes expanded at the support plates (baffle plates), but neither applicant nor staff witness knew the 1 specifics about such a proposed program for Byron, including in-spection intervals and techniques, and evaluation. (Finding 69) The fact that n6 structural modification.would be required under a 90/10 flow split at Byron had some input into Westinghouse's choice of it for a final modification for Byron, (Finding 70) For an 80/20 split, you have to increase the_ resistance in the main feedline by installing either a different valve or an orifice l or something like that. Purchase and installation of an 18-inch orifice costs hundreds of thousands of dollars-. You also have to change the valves and decrease the flow resistance in the feed bypas s - line , depending on what system you install, valves used- , and line flow resistance..A 70/30 split would require new piping, installation of a flow restricter or flow resistance in the main us fee dline . The costs of either modification would range from some-
- L thing in the hundreds of thousands of dollars-to possibly a million i ,
/
- dollars or more. (Finding 71) .
The test simulation of various tube support heights and support b; plate hole did not occur in a field setup, but was a computer an-alytical model utilizing a stick or line to represent the tube and 7p a circle gap element to represent the plate hole. (Finding 72) With the exception of sponsoring research efforts at various lab-oratories, in this case at Argonne National Lab, no NRC sponsor or agency other than Westinghouse has reviewed or conducted testing on flow-induced tube vibration. Furthermore, Argonne, while familiar __ with data collection techniques and plant _ sites has not conducted
~
_la_ tests such as Westinghouse's, except for single tube models which are more routine. Argonne has merely reviewed the data accumulated by Westinghouse from operating _ plants and tests with respect to this recent tube vibration issue. (Finding 73) Staff has not received a final . report on the proposed modifi-cations from Argonne at this time, and will not receive the evalu-ation until within a.. week or two of the Westinghouse , report. (Finding 74) Loose Parts Monitoring System _ , Tube wear due to foreign objects can cause a tube to rupture.
. Two such recent incidents have occurred at Prairie Island Unit 1 in 1979 and at Ginna in 1982. (Finding 75)
The Prairie Island rupture was due to a coil spring which re-mained in the steam generator following an earliear outage for steam generator maintenantes Oa. (Finding 76)
;?he Ginna rupture was also due to the presence of a foreign ob-ject which impacted against and severed a previously plugged tube.
The severed tube subsequently wore against antmplugged tube pro-viding a long wear scar which ultimately led to tube rupture. (Finding 77) : The potential for damaging tubes as a result of foreign objects and loose parts being present in steam generators can be reduced by appropriate surveillance. Examples of available surveillance methe ; ods include visual inspecti'ons with the aid of fiber optics and/or
- radio camera devices, and loose parts (accoustic) monitoring of the steam generator during operation. (Finding 78)
Visual inspections on the secondary side of the steam generator i of. plugged tubes are not normally performed. (Finding 79) , I
1 Applicant intends further to conduct periodic visual inspec-l tions of the secondary side of .the steam generators during refuel-
-ing and maintenance outages . (Finding 80)
' Plugged tubes cannot be tested for loose parts potential be-cause no eddy current test exists for plugged tubes. (Finding 81)
Page IV. 7-1 of the S.A.I. Report, Section 7.1.1, indicates "PWR licensees shall ,be required to develop criteria and prodedures for stabilization of degraded tubes that may be subjected to pro-gressive degradiation mechanisms having the potential to cause sev-erance of the tube and consequently to damage adjacent tubes." (Finding - 82) Byron is being required to implement Regulatory Guide 1.133 which provides in part for a loose parts monitoring system. (Finding 83) This system includes two sensors on the secondary side of each steam generator. These sensors listen for noise generated by loose parts. (Finding 84) Mr. Blomgren is not " completely familiar with all the techniques that are involved with detecting loose parts and evaluating the size . and position 6f those parts." (Finding 85) The sensitivity of the loose parts monitoring system depends on a number of factors and involves a pretty complicated estimate to determine the sensitivity of a given sensor. (Finding 86) The ultimate scnsitivity of the loose parts monitoring system can not and will not be determined until some time during the early stages of the start-up of the plant. (Finding 87)
~
.To_say specifically that the Byron system will be able to do this or that is really going to be based only on the results of the start-up tests. It is possible to replace tubes with solid steel tubes. A detailed safety evaluation would be necessary to determine if there would be any detrimental effects from this.
~ ~ ' (Finding 88) - No evaluation has been conducted in the last several months of replacing the tubes chosen for expansion ,with solid steel tubee instead. (Finding 89) It is Bridenbaugh's understanding that the LPMS is not an acti% system, is there for periodic monitoring, .and is not required in technical specifications, i.e., is not a limiting condition for operation. He believes it should be. (Finding 90)
~
'The LPMS is expected to reduce the potential for the types of
-problems which have been experienced to date. However, some degrG
~
of degradation is.likely to occur at Byron during its lifetime. . _. Given the potential for degradation, surveillance requirements ars it-
- , .; ; ~ 'l essential to ensure adequate tube integrity is maintained against Mi rupture and excessive leakage during the full range of normal ~
operating and postulated accident conditions. (Finding 91) The maximum permissible leak rats during normal operation at Byron has been established through leak rate and burst pressure tests at the Standard Technical Specification limit of 0.35 gpm
- - : c ;L
,~ This corresponds to a maximum allowable per steam generator.
crack length of 0.43 inch. Yet the Ginna rGpture was a single tube rupture at 700 gpm. (Finding 92)
The critical crack length corresponding to the maximum accident condition pressure during a postulated Main Feedwater Line Break (MFLB) or Main Steam Line Break (MSLB) was conservatively deter-mined to be 0.51 inch using the results of the burst pressure tests. (Finding 93) Section VI of the ASME Boiler and Pressure Vessel Code (Code) provides guidelines for establishing the " limiting safe conditions" of tute degradation beyond which defective tubes must be repaired or removed from service. The limiting safe condition takes into account (1) the minimum tube wall thickness needed in order to sustain the imposed normal operating and postulated design basis accident condition tube loads ; (2) maximum (technical specification) permissible leak rate during normal operation to preclude a tube rupture during a postulated main steam line break accident and (3) allowance for continued degradation between inspections and
' eddy current measurement uncertainties. (Finding 94)
Applicant follows Section XI of the Code in lieu of the plant-specific program described in NRC Regulatory Guide 1.121. The amount of tube wall' degradation above which the tube shall not continue in service, the tube plugging criterion, established by Section XI of the Code, paragraph IWB-3521.1, is a depth of outside wall penetration not to exceed 40% of the wall thickness for tubing from SB 163 material.when the mean tube radius to wall thickness ratio is less than 8.70. (Finding 95) , i i l 1 l
- l l
. . l Pre-Service and In-service Inspections Commonwealth Edison has performed a-baseline examination of the tubes in the Unit 1 steam generators only. (Finding 96)
The purpose of performing the pre-service inspections is to l establish a baseline against which subsequent in-service l inspections can be compared. (Finding 97) A pre-service inspection was conducted of 100% of the tubes in Unit 1 using multi-frequency eddy current examinaticn. The eddy current probe is designed to measure any's'evere departure l from a nominal condition. (Finding 98) The data was reported in accordance with Article IV-6000 of ASME Section XI, which requires the reporting of tube wall pene- l trations in excess of 20% of the tube wall thickness and tube wall dents. (Finding 99) y Based upon the inspection, a baseline was established for Unit 1. The inspection revealed that two tubes were partially
. blocked. The cause of blockage could not be determined.
(Finding 100) Based on a Westinghouse recommendation, the two tubes were plugged. (Finding 101) - Some processing-induced denting was discovered. Most of the denting was located four or five inches away from the tubesheet. That is, in the preheater beltline. (Finding 102) Paillaman did not agree with the sentence as written by Paul which would highlight defects found in Row 1 tubes which is a suspected area for service-induced dents and cracking at the apex of U-shaped tubes as a secondary effect of denting. (Finding 103)
. There were from 2 to 3 dents on selected tubes for a total of' 600 dents in steam generator No. 1, 500 for steam generator No. 2, 450 for steam generator No. 3, and 550 for steam generator No. 4
-(Finding 104) l Steam generator No. 4 was found to have a large number (124) of tubes with indi~ cations of permeability; a -localized area of
~--
magnetic material or impurities in the Inconel which require a
, r.
different technique for an examination to verify int-grity of +- the tubes. (Finding 105) . - The condition of the Byron steam generators will be period-ically monitored through an eddy current inspection program. Inspections will be performed according to the provisions of the Byron Technical. Specifications and according to NRC Regulatory Guide 1.83. (Finding 106) The results of these periodic inspections will be compared to
, the 100% pre-service baseline examination. This comparison will 4p -- -
provide a periodic evaluation of the steam generator tubing s condition to allow time for the initiation of appropriate measures
- (
Li for steam generator maintenance prior to any occurrence of
*7 primary.to secondary leakage. (Finding 107) t- ,
NRC Regulatory Guide 1.83 requires that in-service inspections m be performed every 12 to 24 months. In cases where the degradation processes have been have been highly active, the Staff has re-quired that the-inspectio.ns be performed at more frequent inter-
'l vals, consistent with the rate at which degradation is occurring.
(Finding 108) 4 e a -
-e g em -r-w. - , . - ,-~,,,-w- . ~ . . -:,, -- < - , , - - .-- -w-,m---,- w--- a -- - - - - , ,-.m --
c g-- -- ,p..o -
Technical Specifications require sampling (at minimum) '37. of 'all the tubes in the plant in the first in-service inspection. The scope of the inspection may be expanded to cover 1007.-of the tubes in circumstances where either (1) greater than 10% of the tubes inspected have eddy -current indications greater than 20%, or (II). greater than 1% of the tubes inspected exceed the plugging criterion. (Finding 109) The total number of tubes to be inspected during an in-service
~
inspection will range between 3% and 100% of the total number of steam generator tubes. The, material and mechan'ical design differences between the D4 and D5 models, and the lack of empirical data on the efficacy of the modification for flov induced vibration suggest the need for a post modification in-service inspection. program based on Reg. Guide 1.121 and 1.183 rather than strictly based on the ASME code. (Finding 110) At Byron subsequent inspections will be conducted on all ..
. unplugged tubes previously identified to have eddy current indi-cations greater than 20% as.well as the 3% total sample provided by the Technical Specifications. (Finding 111)
Eddy current testing is the primary inspection technique and it is usually performed using an' instrument that impresses four different test frequencias on the coils simultaneously. The frequencies are selected on the basis of providing definitive information on tube degradation, support plates and external deposits. Because the responses for external discontinuities vary according to the test frequency, it is possible by linear combinations of the responses at different frequencies to reduce
' unwanted signals from a composite response. (Finding 112)
l 1..- 1 j l Bridenbaugh doesn't believe the industry has had enough j experience with multi-frequency eddy current testing to deter- l l cine whether it.will prove to be adequate. Because of the j 1 possibility of human error which exists in reading and inter-preting the signals from a tape, the test methods should use both differential and. absolute test frequencies and methods. ; (Finding 113) - 1. Multi-frequency eddy current testing has been used at some plants on a limited basis for about four years. I,t has not been utilized on an industry-wide basis , and it is not
, presently an NRC requirement. (Finding 114) ,
,~
- The sensitivity of' eddy current testing to tube wall degrada-tion varies depending upon the location, type, size, shape and extent of the degradation. Eddy current testing will detect the various types. of tube degradation at the following sensi-i'
' ~ tivity levels: tube wall thinning at 20% depth of the tube wall; f pitting at 20% tube wall depth; tube wall cracking at 40% tube 1 . wall depth; intergranular attack at 40% tube wall depth; tube wear at 20% of tube wall depth. Denting, a form of tube deforma-1- t tion, can be detected by eddy current testing. Eddy current ( development now in progress in the industry is expected to i further improve detection limits and characterization of possible , i tube degradation. However, the only absolute means of confirming the form of tube degradation indicated by eddy current tests is by direct observation, including the removal of the tube and by
- metallographic examination. (Finding 115) i a
----, --.--.-..v _%, ,
.wr_,..._m. , p ,- .
n , If a significant rate of tube degradation is determined as a
~ result of an in-service inspection, measures are available to reduce the probability that tube leakage will occur before the next scheduled inspection. These measures include (water chemie alterations such as temperature changes, water lancing, flushind and chemical cleaning to try to loosen and remove degradation- ^ 7.:k.?,"" causing impurities or deposits ; -foreshortening and reducing the interval between inspections or lowering the tube plugging limid
[ (emphasis added) (Finding 116) , Water Chemistry Operating plant experience has shown the need ~ for rigorous control of the water chemistry environment of the entire steam cycle, including the condensate and feedwater systems.
./
7 (Finding 117) Because of the extensive laboratory work that has-been per-
.; formed in evaluating the mechanisms of stress corrosion cracking
~ u-
' J.iIr. -
thinning and denting ~ of the steam generator tubes, it is clear that impurities admitted to the steam cycle which ultimately may
.w
],]Q.; reside in the steam generator, must be limited to a range that
^
.would lead to no more thAn 20-100 ppb on the secondary side of' (Finding 118)
_g,j;9; the steam generator. Impurities such as air, which contains oxygen, condenser cooling water such as fresh water sources that contain excess
,il
-4;hy .M alkalinity,- makeup water impurities and the like, must be exclus
^;2"'
,"c- w It has also been demonstrated that copper and. nickel bearing 1 alloys in the feed train 'can participate in corrosion reaction when transported to the steam generators. (Finding 119)
4 . Accordingly, the Applicant is implementing an AVT water chemistry program on the secondary side of the reactor systems at the Byron Station. (Finding 120) AVT chemistry control is based on a philosophy of minimum contaminant ingress through the practice of good initial design and material selection of condensers, feedwater heaters, makeup water systems and other components. (Finding 121) AVT control is maintained by appropriate inspection and main-tenance practices and operator actions during plant operation. Adherence to AVT guidelines enhances the long term' integrity of the rteam cycle, by minimizing the corrosion of condenser and feedwater heater materials , the steam generator and the turbine. This in turn minimizes the formation of corrosion products which are delivered to the steam generator. (Finding 122) AVT involves the addition of volatile chemicals as control agents. With proper monitoring these agents should not concen-trate in the steam generator but be removed via the steam to the~ remainder of the secondary system. Generally two chemicals are added, a volatile amine (usually ammonium gydroxide) for pH control of the feedwater and an oxygen scavenger (hydrazine). Hydrazine scavenges oxygen producing byproducts such as nitrogen, which is innocuous provided pH is controlled within a specified range. As the hydrazine moves through the feedwater system and is subjected to higher temperatures , any unreacted hydrazine can decompose to form volatile compounds such as ammonia, nitrogen and hydrogen. The addition of ammonium hydroxide for pH control must be continuously adjusted to compensate for the volatile compounds produced from.the excess hydrazine thermal decomposition. (Finding 123)
Applicant's AVT water chemistry program is based on Westing-house and EPRI guidelines. (Finding 124) In order to reinforce the need for vigorous chemistry control, EPRI has issued AVT guidelines as -a model'to be reviewed by the industry. (Finding 125)
; The' Westinghouse guidelines, introduced in 1977 and modified subsequently from time to time, recommend that (1) the guideline chemistry conditions should be achieved' prior to unit loading and i
maintained during power changes; (2) any source of contamination should be identified, the source corrected and no operation allowed with locatable contaminant ingres's ; (3) dissolved oxygen at the condensate pump discharge should be less than 10 ppb to minimize the inventory of corrosion product transported to the steam generator; (4) continuous monitoring of the chemistry of the steam generator blowdown should be performed; measured values should be compared to theoretical values in order to identifg whether or [ .not excess alkalinity or acidity is present; (5) copper bearing
~
alloys should be eliminated from the secondary system to permit greater flexibility and optimization in chemistry control; (6) ! main condenser integrit'y should be upgraded to minimize the in-gress of impurities in the condensate in order to improve the
. reliability of=the. steam generators and turbine; and (7) if a full-flow condensate polishing system is installed, it must be carefully controlled and properly operated in order to optimize s the quality of the treated condensate. (Finding 126)
, The so-called EPRI Guidelines were developed under the aegis .
of the Steam Generator Owner's Group (SGOG) . These guidelines i 1-
incorporate more restrictive water chemistry controls than the Westinghouse guidelines and include a staged corrective action plan. In addition to the more restrictive water chemistry con-trols, the SGOG Guidelines include recommendations for data management, surveillance, and analytical methods. The SGOG Guidelines include a recommendation that specific management
-g e.,c responsibilities regarding secondary water chemistry control
~2 be assigned from the plant chemist to senior corporate manage-
- - , - ment. (Finding 127)
- s. In addition to the Westinghouse Guidelines noted above, the Byron Station Chemistry Monitoring Progran incorporates the following elements from the SGOG Guidelines: (1) more restric tive SGOG water chemical controls coupled with a corrective action plan to require prompt station response to a chemistry excursion before unit shut-down is required; (2) a staged f.iz. corrective action plan based upon the level and duration of con
.m
' d '.;.c . taminant ingress, requiring specific corrective actions, includ
,;%g
_F5d.T ing staged reductions in power; (3) a data management and 9CIDIYS surveillance program providing for prompt identification of ffy21 yd.5n6j negative trends or inconsistencies in chemical control data; 9 4) (4) an analytical program to supplement and verify the continuo ys,se:gg on-line chemistry monitoring system data. Although not speciff ...,.y cally included in the Byron Chemistry Monitoring Program, the statement of management responsibilities recommended in the EPR y.;;;,; fN1?k? Guidelines is being addressed in a Commonwealth Edison corporat
.,; FWR Secondary Water Chemistry Control Program. (Finding 128)
Conway used illustration of flow slots on tube support plate to demonstrate how density caused hourglassing of the tube flow slots- in the upper tube support plates thus pinching the support plates and in turn pinching the tubes. Thus, compressive stresses inherent in .the pinched tubes caused tensile stresses at the U-bend. However, on the D-5 the widest spacing between tube support plates dhich is functionally acceptable was selected and the flow holes in the flow distribution baffle plates and in
- the top tube support plate were modified from a rectangular to 1
a circular design, thus reducing-the potential for U-bend cracking. (Finding 129) l The hydraulic tube expansion technique for expanding tubes at the tubesheet was used only for Unit 2 Model D-5. The tube expansion technique used for unit one was mechanical dilation j techniques with greater inherent residual stresses. (Finding 130) A form of tube thinning has also been observed at lower tube
, support plate elevations around the periphery of the bundle at
-two all-AVT plants. At the present time, no corrosion mechanism l has been identified for this phenomenon. (Finding 131)
Plants that have only operated on AVT have experi'enced some denting. (Finding 132) ! Denting is a localized radial reduction in the diameter of steam generator tubes, resulting from corrosion of the carbon steel tube support plates in the tube-tube support plate annulus, as in the D-4 Model. (Finding 133)
I Another source of denting identified through field tests is J
. the condenser in-leakage of contaminants from the tertiary water system such as copper, oxygen, and chloride ions. (Finding 134)
The 1977 Westinghouse AVT. Guidelines advocated rigorous con-trol of the condensate and feedwater chemistries during both shutdowns and power operation to minimize secondary system corrosion and transport of the corrosion products into the steam generators. (FLading 435) Concentrations of chloride in the levels of thousands of parts per million can build up in the tube support crevice even though the bulk steam generator solution is only at the part per billion level. (Finding 136) Chloride levels on the order of thousands of parts per million (ppm) have been seen in the corrosion deposits in the tube support plate crevice. (Finding 137) No method has been developed for determining the concentration of corrodants contained in the tube support plates annulus where
- denting is known to occur. A compiete sample of the chemical sol-ution contained in the ' tube support plate crevice has never been taken. Neither is there any empirical evidence on the concentra-tion of corrodants at the tube support plate coeurces or the min-imum concentration of corrodants necessary at the crevice for cor-rosion to occur. Thus, in operation, it would not be possib1'e to conclusively determine what concentration value in.the bulk water solution would indicate excessive corrosion, nor the corrosion product volume and rate which would lead to eventual tube leaks between inservice inspections. (Finding 138)
L . I 4
-32
~ f' ! The guidelines themselves don't indicate 4f ormat for reporting
- violations;.that is-left to the utility. Violations or failures to adhere to the water chemistry guidelines to be followed at ,. Byron will be reported only on an annual basis to the Nuclear Division Vice President. There is no requirement anywhere to re-port violations of the water chemistry' guidelines to the NRC offie tv of Inspection and Enforcement. (Finding 139)
The EPRI war chemistry guidelines do not specify the contents i i2 of reports prepa:Od by the utility on instances of failures to l
- l. l
- . meet the guidelines. -(Finding 140)
M.ssgr. Blomgren did not know whether there was any requirement, l that reports on water chemistry violations contain any_ document- l l ation at all (emphasis added) indi.cating whether appropriate ac-l tions were taken to correct the failure. (Finding 141) Bridenbaugh states that an independent review should be con-
~
l ducted to ensure that the principles that are generally included-r .
~
intheEPRIwaterchemistryguidelinesareinfactbeingimplemenf ed to the greatest degree possible at the Byron plant. Thisshould
~
)
, gs; be done prior to plant startup. (Finding 142) lighc The ' Resident Inspector for the NRC's Office of Inspection and Enforcement is incapable of reviewing all the plant's operational.
~ 57,c'i procedures for technical content. '(Finding 143) mz i- The function (s) of the Resident Inspector is limited to a curc 4 sory review of the p,lants operational procedures to assure that ursi
, . 9' these procedures are the latest updated version of all the proced<
- n.
ures being followed. (Finding 144) i The prevention of tube degradation in Unit 1 is almost totally dependent on the use and control of water chemistry, as o'pposed F
'n
, ~4- aw- --
, , , - - , --w.-,,--ne -4 ----n-r
to Unit 2 where several material selection and design changes will contribute to: reducing tube degradation. (Finding 145) d{Satushaschangedon26outof44steamgeneratorprocedures ... the Applicant is developing and identified in November, 1982. Edison has stated it will complete the remainder of these 44 pro-cedures before Byron fuel load, including secondary chemistry program descriptions and chemistry procedures Following completion of procedures, all procedures will have to undergo extensive test-ing and systems analysis to ensure they. do not cause a deviation from the secondary water chemistry guidelines. (Finding 146) When asked why he recommends that Byron operating chemistry - procedures should be reviewed by an independent body, Bridenbaugh says that NRC's ITE branch and resident inspector can notconduct an independent review. Their review is merely to' assure that prop-er procedures are in place and is limited to the assurance that the. procedure is identified (Finding 147)' This review wasn T t done when EPRI and SGSG drafted guidelines.
' ~
P :.' _ .. . : - . ' .a a . Bridenbaugh says NO! Bridenbaugh says they are Just guidelines. He's talking about a review of plant specific chemistry monitor-ing-equipment and procedures to en'ure s the guidelines are indeed being implemented to greatest deg'r ee' possible. An operating license shouldn't be issued ,_.,~until this is done. (Finding 148) g- - e
, . , . _ _ , . . . - - . . -~ - ' -
- Mr. Bridenbaugh testified that there is an increased probability that accidents will be initiated by tube failures during normal operation and an increased likelihood that accident sequences not I ,now considered in the safety analysis may occur as a result of the steam generator tubes degradation after some period of operation.
The accident' sequence could involve single or multiple tube fail-ures occuring in conjuction with other accident sequences which can impose transient or abnormal loading conditions on the tubes, resulting in common mode or systems interaction type failures that have not previously been analyzed in the licensing
- review of PWRs. (Finding 149)
Mr. Fletcher assimilated the coned'24sions provided by the App-licant's expert witnesses testifying with respect to their speci-fic disciplines and reached an overall assessment as to steam generator tube . integrity at the Byron Station. Based upon the de-sign, water chemistry, detection and remedial measures undertaken by Applicant, Mr. Fletcher concluded that steam generator tube de-redation at the Byron Station should not. be a safety concern and ' that tubs rupture should not occur, even under conditions of Main f Steam Line Break ~(MSLB) or Loss of Coolant Accident;s (LOCA's). However,.Mr. Fletcher's conclusion is based on the limited'oper-ational experience of steam, generators with AVT water chemistry, highly limited testing of new systems for detection such as loose , parts monitoring system, and a technical fix for the flow induced vibration problem which has not been fully tested for its adequacy in an operational steam generator. (Finding 150).
A Byron specific Pt.A which considers tube ruptures in conjunctit 4 with a-LOCA has not been done. (Finding 151) The NRC Staff's multiple steam generator tube rupture analysis s.m_ . has not been published and is still under staff review. NUREG-0937
- is only.one of the inpnts into this anal.ysisl (Finding 152)
- n. .~ The consequences of a large MSLB inside containment could be m.
adversely affected by the addition of reacter coolant invent ory j and stored energy to the containment. Calculations have shown tha9 provided there are no additional events such as operator error or f. safety valve failures, containment integrity should be effected, I the core should remain covered and cooled due to the' addition of emergency core cooling, and there should be ample water supply fos m .. , _ 1ong term cooling. However, if operator error and/or safety valve v p .W . . failure changes the accident sequence, the MSLB could be adversel) n affected and subsequent tube ruptures are possible. (Finding 153)- 11 3 41
;: y
~ a; > Calculations have been performed to evaluate the systems per-74[ wgm ij.5}Lpf formance, ofsite consequences and required. operator actions assum 3
%p ing' a stea n generator tube rupture concurrent with an MSLB outsid!
g,qw:- conteinment. These studies evaluate the effects of a main steam :.
- q ~g
' line break combined with one or five ruptured steam generator tub:
and a small break LOCA only. Calculations have not been performed V .: :: a ,. $q2J5 for multiple tube ruptures occuring with a uedium or large LOCA and improper depressurization of the' steam generator combined wig improper safety relief valve and PORV valve ~ functioning. (Finding i 154)
The results-of these analyses indicate that primary coolant shrinkage, caused by overcooling, and the simultaneuxs loss of primary coolant, aan be compensated by~the high pressure emergency core cooling system. The core remains covered, and the primary cool-and remains cool, except in the vessel'upperhead. The calculations and results are described in NUEEG-0937. (Finding 155) As part of the tedhnical resolution of the steam generator tube integrity unresolved safety issue, the Staff assessed the consequ- , ences of single and multiple tuba breaks -(in a single steam gener-ator) concurrent with a large main steam line break or large cold
- leg break LOCA. (Finding 156)
One of the main pugoses of this effort is to develop a statis-tically based inservice inspection program that affords a high degree of assurance that if a large MSLB or LOCA occurred concury
. rent with ruptured steam generator tubes in the effected steam generator, the offsite. ' dose and. fuel temperatures would be accept-able. (remain within the ~ range : defined in Appendix K to 10 CFR)
(Finding 157) l The consequences of a large, cold leg LOC:A could be adversely 4 affected by the flow of steam generator fluid into the primary loop 4hrough the broken steam generator tubes. Several computer studies performed over the years indicated the following: 1) rupture of a few tubes during a LOCA would have'very little effect, (2) if_a large num6ar of tubes ruptnred, the ~ additional fluid flow into the reactor vessel during a_large.br~eak would actually aid in cool-ing the core, and (3) an optimum number of tubes over a limited range (about 12',or about 1200 gpm total) could have a detrimental effect. (F.inding 158)
A series - of LOCA experiments with varying degrees of simulated tube failures performed in the semi-scale facility at the Idaho-National Engineering Laboratory several years ago confirmed the general behavior involved in small LOCA accidents. These experi- - ments did not show the same degree of degraded core cooling as the computer analysis did for the worst cases. In fact, the experiments did not indicate that-any core damage would occur. (Finding 159 The MSLB, M7LB and-cold and large cold leg break LOCA accidents ; are extremely low probability events on the order.gf 10 -5 t o 10
-6
.per reactor. year..The frequency'of multiple tube ruptures combined
~
as a' consequence of steam /feedline break events is 3x 10-5 per reactor year. (Finding 160)
.The steam generator tube rupture event is an infrequent event
__ ' . . , _ 4.n t w._ac on the order of 10-2 or 10-3 per reactor yeareIn the opinion of the Staff, the likelihood of a tube rupture concurrent with a LOCA, MFLB or MSLB is low. (Finding 161)
~
The Applicant performed an analysis whereby it predicted that tube rupture events in combination with accidents are predicted to result in severe core damage at frequencies of 10-7 per reactor year for the Byron. Station. By this calculation, a single tube
. rupture would occur about once every 33 years at Byron. (Finding 162) -
The calculation that a single tube rupture would occur about once every 33 years at Byron is bounded by the following confidence limits and uncertainty in the range factor: The range factor with an 80% confidence limit is 2.5. A total range factor of 5 is ap-plicable to the tube rupture probability when the condfidence or uncertainty level is 907.. Mr. Hitchler did not calcula~te a range
factor for tube rupture probability for confidence levels above 907.. - (Finding 163) The uncertainties on the results for one, two, and three tube rupturing is estimated to be factors of 5,25, and 125 respective: (Finding 164) Goldberg asks Bridenbaugh if the Byron FSAR analysis postulat< p failure of one tube and only 1 gpm leakage, and whether be belie-1 - p .. m. that? Bridenbaugh says no-he believes it could be as high as 760 l-L gpm since that was what was seen from the single-tube failure at Ginna. With a 760 gpm break, only 2 tubes would have to urpture to reach the. 1200 psi at which equilibrium between primary.and secondary side break flows should be reached. (Finding 165) Bridenbaugh explains how multiple tube failures could occur, r with a Ginna precursor as initiator. A plugged tube that is dama
,~
or comes loose may wear on more than one tube at the same time. could wear on all of the surrounding tubes and cause a reductios
..~
lT 'YM of wall thickness in a rapid fashion. (Finding 166) ! x y.xf Twelve potential requirements have undergone cost-benefit as-Ey@ ment by the Staff and its consultant. The consultant cost-benef. m. puM5 assessment is contained.in a final draft report, the S.A.I. 8eps b f' ; entitled Value Impact Analysis of Recommendations Conerning Stc
- Generator Tube Degradiation and Rupture Events" '(mar & d for ides fication as Intervenors' Exhibit 9)cThbse items are under consic
.i dhe ation as potential' Staff requirements. The Staff has continued evaluate these recommendations since.the issuance of the consui report, and has modified, altered and changed the majority of tb recommendations. A number of other items are also being conside v%
o . as Staff actions including about 15 procedural recommendations. (Finding 167) Gallo begins to question NRC witnesses on their judgement of whether or not they are going to recommend that the potential re-quirements in S.A.I. report will be proposed.as requirements. Each witness answers only with respect to his divisions's recom-mendations,8ecause th5re currently is not a staff poisiton on each of these requirements they j must still await CORGR and ACRS review
~
before being proposed as requirements. (Finding 168) Marsh states that the S.A.I. recommendations and NUREG-0844 will probably be appendices to a long memorandum on the requirer-at ! ments which must be implemented, and this memo will go to Licensees and Applicants. The requirements would be in the form of a 10 CFR 50.54 letter that allows agencies to require changes in the Standard j Review Plan. (Finding 169) Pages 4733-4750 of the Transcript of this proceeding enumerates ( the potential requirements in the S.A.I. report that the Division of i Systems Interaction and Engineering, the Division of Engineering, and the Division uof Safety Technology,in the NRC all agree be recommend-ed for implementation. (Finding 170) Marsh states that mayyof the S.A.I. requirements have been imple-mented at Byron and that there is no consideration to exclude any plant from back fitting requirements of NRC via S.A.I. review. But it is really up to the" Committee for Review of Generic Require-ments to decide on backfitting. (Finding 171) Marsh testifies that the S.A.I. Report is still under Staff review and is not yet complete. Anticipatesformulation of overall
recommendations 1 to be forwarded- to the coc=ittee for review of genericrequirementstowardsNend
~
i of May. Several steps have yet to be completed before S.A.I. recommendations become requirements, ! including ACRS and commission review and public comment period. I (Finding 172) I' As a result of the.TMI accident, TMI action plan 1.C.1 requires the . industry to upgrade emergency operating guidelines and procedures - t to cover multiple failure events which fall outside the required i design envelope assumptions for safety analyses. (Finding 173) The Staff and vendors are analysing a variety of such events , including coincident steam generator tube ruptures and LOCAs and - coincident steam generator tube rupture and steam line breaks. The results of a recent Staff analysis are discussed in NUREG-0937. (Finding 174) 2 1Due S. A.I. Report did not perform its own multiple tube failure , analysis but did review NRC's MTFA. (Finding 175) I - After TMI, and the concern. for improved operating procedures, the Westingh'o use Owners Group undertook a generic development of
- guidlines to cover all emergency operating procedures. These generic
. guidelines have been submitted to the NRC for approval. (Finding 176) . The Westinghouse generic emergency response guidelines have been used as a basis for the development of the Byron operating proced-l ures. Emergency operating procedures being developed should help the Byron operators to respond to the various compound accidents
- discussed at the hearing once they occur, provided that Byron spec-4 - ,
. ific procedures are adequately tested by computer traulation and the
- operators are thoroughly trained to allow an expedited use of these procedures in the event of an accident. (Finding 177)
If Byron is the first plant to experiment with the Westinghouse
,. - 41 ' -
NRC Staff Review Contrary to Applicant's opinion, tne efficacy of'the Westinghouse proposed modifications has not been the subject of an extensive review and verification process by NRC Staff. (Finding 179). Frank states that Applicant has to submit to Staff their recommendations.f'or operation with the modifications. Nothing has formally been' presented t'o Staff by Applicant or
~~' '
Uestinghousa. (Finding 180). Dr. Rajan is not a competent witnese for Staff in establishing
~
the empirical basis for Staff's interim conclusion on page 5 of his revised profiled testimony that, " Based on Staff's pre-liminary review of the proposed modifications, the objective h+. of minimizing tube. degradation associated with flow-induced' vibrations will be accomplished by these modifications." His
~
L..,
$:I$' familiarity with the data Westinghouse received from operating Nix
.:.' ;, plants, scale model testing, and Krsko tube expansion, is not
;q3 af;gi sufficient to objectively re.ach that conclusion. It is not an e-c
- dmpirical conclusion reached'by review of the actual data.
Instead he formed it on basis of View-graphs presented to Staff by Vestinghouse as data summations. The follouing are 4 examples of Mr. Rajans lack of competency for evaluating the 7 "~ ' proposed modifications by Westinghouse: Mr. Rajan was not aware that a tube had been expanded at KRSKO Plant. He was also not aware of any empi rical data which would enable Staff to reach their preliminary review of the proposed tube modification /fix for flow-induced vibration. (Finding 181).
'd r s A
/
^"
Except for'his personal notes tak'en from meetin.p uith Westinghouse, Rajan has received no formal submissi n or reoort.from Westinghouse on the proposed tube moditicdtion/fix for flow- induced vibration. (Minding 182). /' The Staff review of the proposed modifications will be completed prior to plant operation, and a safety evaluation - report issued followin,g its implementation at Byron. (Finding '183). !. .. \ % Unresolved Safety Issue A-Z ,+
' Q. .
- c. _ , s@
USI A-3 is an unresolved safety issue which will not.be, v fully resolved by the NRC for several more months. (Finding,184).' ,. A primary objective nf the steam generator USI program is to x
, ensure that tubes are plugged before they corrode to a .
significant degree that they have potential for rupturing in subsequent operation. The objective of the program is'to ' . t
- ' control degradation to enable the maintenance of tube"dall _. ,
~ .-
, integrity under both operating and accident conditions., -
[ w (Finding 185 ). _ 1
- r l
Staff resolution of the USI A-3 issue is not complete. It - is.almost exactly at the same point as in 1981 when-they said.'.
- ~
l . l it woul'd be resolved in early 1982 (Finoing 18E). c Staff has estimated previously the time of resolution of ,, NUREG - 0886 estimated early'P353. task A-3 and been in error. l Staff now says it will be done by mid-1983. The Ginna event'N
- - . " s :a ,
has delayed resolution by' introducing need to consider i. , _
?.
multiple tube failures. The Ginna t u t e r u p t u'r e e v e n t.ja o'd e'd ' - ,
- to previous corrosion problems the need to closely considar' ~
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- . . , , . . ,-. , ..n. , , _ . , - , . . . . . - ,--.,,-n..e_ -
, - , . , , . , , , - - -n, . - - , , , , - , , , , . , - - - . , , , ,,- - - -
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issues inutilving: damage by forei'gn,matelial,, deterioration
,, _ .a, c / .'
of prev'iously -~ plugged tub es', a76iiiYteractive failure of adjacent tubes. This, in turn, raised the need to consider h mult'iple tObe failures,and concf; gent accident sequences. (Finding 187). '
.M
[,, The recommendations.in NUREG-065'1 are also included in - my , , Staff's generic program 1as is NUREG-0844',
- s/ 1' a draft document,
^
not yet f'inal. (Fino'ing 188). ,
, n
~
,The decision as,-hoIbsther NRC requirements de.veloped as s- ,
otrt-~cf - the resol.ution of USI,A-3 mu:t be retrofitted into er operating nucleat. power plants-icia. decision that must be made by the' Committee for;the' Review of Generic Requirements.
.There is no consideration at.tnis., time-to exclude any plant from hackfitting in order td meet the NRC's requirements to e ', ,
resolve USI A-3. However, Staff's recommendations for appro-priate actions for resolution of UST A-3 have not been '
/. .-
forwarded to the' Committee for Reviowgof Generic Requirements ~ y e tt. This committ[sois the CinaIauthorityonwhetherornot to backfit. "(Finding 189).' , 'J W5 9 y a e , e e
^
A E wy j o i na w%$ g . 4 W -
>y -
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h
- 44 -
r.. i conclusion Eleven issues and sub-issues were reviewed in this proceeding with regard to e i the Byron plants corrosion-related degradation, steam generator design, pre-l service and inservice inspection, primary to secondary leak rate limits, water
;l i
chemistry, tube plugging, loose parts and flow-induced vibration, tube rupture i events and accident considerations, and the over-all resolution of NRC Unres-olved Safety Issue A-3, steam generator tube integrity. The Board finds that corrosion-related degradation will still cont-inue at Byron. Although Unit 2 is a D-5'with additional design margin over previous models, no Westinghouse steam generator model has been totally exempt f from corrosion-related degradation, and relatively little operational experience exists with either the D-4, or D-5 models to preclude it's occurrance at By-ron. Further, the Board finds that, in many respects the D-4 model of Unit 1 is not enhanced'against degradation. (F ndings, 16, 17, 23-26, 27, 28, 33-35) l l I v The Board finds that the steam generator tubes at the Byron plant ' t may be more suseptible Ebedegradationinthelongrun,duetofabrication defects on selected tubes at or about the level of baffle-plate in the pre-heater section in row one tubes, a suspected area for service-induced dents L p; and cracking at the.arex cf U-shaped tubes. (Findings, 134-140) AIARA is always a consideration for a proposed modification. The [': l Board finds that Applicant's coannittment to implement the proposed Westing-4 house modifications at Byron prior to operation be made into a license cond-
~
ition due to the substantial AIARA Considerations if that modification is not installed prior to an operating license. Futher, as there exists the possib- ! ility that Byron, and not Comanche Peak may be the first operational plant utilizing the Westinghouse proposed D-4, D-5 sodifications, the Board also finds it a license condition that
o .
* . a 45 -
+
- s .- - .
finds.it.a license conditi.on that Applicant, instrument the steam, generators of both units,with, accelerometers to monitor tube v.ibration levels on an onagoing
. basis,n and demonstrate t'he development of a special inspection program to en- ,
sure safe, operation of the plant. . , (Findings ,- 57,58,69,70-86) The Board finds that a review of the By,ron plant, specific chemistry monitoring equipment and procedures should be conducted by an independent body
. . . . ;.n ,
prior to plant oper'ation to ensure that'the EPRI and Westinghcuse guidelines are being implemented to the grea' test degree possible by Applicant. (Findings,
'r .:
173-180) -
~
The Board finds that as the loose parts monitoring system is not an active system, is there for periodic monitoring, and is not require;as a limit-ing condition for operatilon, be tested at Byron prior to operation to ensure it's operational effectivenesc, accurate calibration, and incorporate the gencr-ic surveillance requirements currently undergoing Staff development and review. ~
'M (Findings, 106-113) *
, ;, The Board finds that Applicant has failed to comply with 10 C.F.R.
< h Q '*
" . g;ngi' 50.57 (a) (3) as implemented with respect to steam generators by satisfying m ;, . _.
NM ~.t:'Wu General Design Criteria 30 and 31, wherein Applicant must demonstrate that de- ^$.3$5 sign measures (such as the technical fix for flow-induced vibration), fabric- %$$l$ ation methods and operational proceedures (such as the program for adherence
, yq to water chemistry guidelines) have been developed and tested; and that detect-D ion systems are in place (such as the loose part monitoring system) so that under either normal or under accident operating conditions there is an extre-
, . sl . ,
y Wg mely low probability of abnomal leakage, of rapidly propagating failure, of ( : gross rupture of tubes becoming so brittle that their degraded condition would eventuate in the above. Additionally, the Board finds that NhtC Staff has failed to provide i adequate explanation why the operation of the Byron plant can proceed in ad-
4 <^, 46 - ,
, o of an over-all solution of unresolved safety issue A-3, insofar as Staff has failed to conduct anadequateindependentinvestigationof$astinghouse'spro
,. posed modification for flow-induced vibration relating to steam generator tube integrity, so as to justify their eenelusion that the basis of their prelimin-ary review of the modifications is sufficient to ensure the public health and safety under plant operation. (Finding 223) 8 0* e O O 9 9 8 i h
~a, , ~ -+ -- , ., -
s . O - UNITED' STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )
)
COMMONWEALTH EDISON COMPANY ) Docket Nos. 50-454
) 50-455 (Byron Station, Units 1 and 2) )
CERTIFICATE OF SERVICE I hereby certify thaE copies of Revised Rockford League of Women Voters and DAARE/ SAFE Findings of Fact and Opinion on League Con-tention No. 22 and DAARE SAFE Contention 9 (c) in the above-caption-ed proceeding have been served on the following by deposit in the United Sates ma'il, first class, or as indicated by an asterisk, by express mail, oras indicated by double asterisks, by messenger this 21st day of July, 1983: Region III
*Ivan W. Smith, Chairman U.S. Nuclear Regulatory Commission Administrative Judge Office of Inspection & Enforc_ement Atomic Safety and Licensing Board 799 Roosevelt Road U. S. Nuclear Regulatory Commissio) Glen Ellyn, IL 60137 Washington, DC 20555 ..
Dr. Bruce von Zellen
*Dr. A. Dixon Callihan c/o DAARE'~'
Administrative-Judge P.O.!B6x 261 T.~.1 , Union Carbide Corporation DeKalb, IL 60015 P.O. Box Y Oak Ridge, TN 37830 Ms. Diane Chavez
- 326 N. Avon Street
*Dr. Richard F. Cole Rockfo'rd, IL :6fl03 Administrative Judge ~.
Atomic Safety and Licensing Board Jane Whicher,REs,q.'~ t U.S. Nuclear Regulatory Commission 109 N. Dearborn Street Washington, DC 20555 Chicago, IL 60602
** Michael Miller, Esq. Atomic Safety and Licensing Appeal Isham, Linc61n 6 Beale Board Panel Three.First National Plaza U.S.. Nuclear Regulatory Commission . Chicago, IL 60602 Washington, DC 20555
- Joseph Gallo, Esq. Docketing & Service Section Isham, Lincoln & Beale Office of the Secretary Suite 840 U.S . Nuclear Regulatory Commission 1120 Connecticut Avenue, NW Washington, DC 20555 Washington, DC 20036
-Atomic Safety and Licensing Board Panel _j j U.S. Nuclear Regulatory Commission ,, /h,9,[
Washington, DC 20555 BettyL nmyn, Chair League of Wo: nkotersofRockford Intervention ommittee O}}