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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION DOCKETED UstaC BEFORE THE ATOMIC SAFETY AND LICENSING BOARD
'84 !?]-2 N3 :27 In the Matter of
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DUKE POWER COMPANY, Et al.
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Docket Nos. 50-413
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50-414~
(Catawba Nuclear Station,
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Units 1 and 2)
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PROPOSED FIEDIEGS OF FACT AUD C0hCLU3IOUS OF LAW IK THE FORM OF A PARTIAL INITIAL DECISION (Oh COETENTIONS Pa44/
CE3G18, REACTOR EMBRITTLEMEET AND DES 17, ADVERSS WSt.THER Intervenors Palmetto Alliance (PA) and Carolina Environmental Study Group (CESG) have addressed four contentions in the safety related phase of the Operating License proceeding for the Catawba nuclear station, units 1 and 2.
The Atomic Safety and Licensing Board (Board) herewith issues a partial Initial Dec'sion in regard to Contentions PA44/CESG18i concerned with the possible inadequacy of present means to predict end of life reactor embrittlement, in which we have designated CESG the lead intervenor, and joint PA/CSSG Contention DES 17 concerned with the possible inadequate consideration by NRC staff of extreme weather conditions in evaluating environmental consequences.
A separate Partial Initial Decision is issuing simultaneously deciding PA Contention 6, concerned with Quality Assurance deficiencies alleged, and PA Contention 17 concerned with possible underestimation of consequences of fuel pool accidents.
SUMFARY On examining the testimony, both oral and prefiled, the exhibits and supporting documents, and the proposed findings of fact submitted by the parties, this Board concluded that reaching a conclusion on the narrow question of whether the end of life nil ductility reference temperature. forecast variously by Applicant and Staff gave reasonable B403070037 B40229 PDR ADOCK 05000413 g
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assurance that Catawba reactor embrittlement with operation would not be excessive would not satisfactorily discharge its obligations.
The Board sees as its primary obligation in this matter adherence to 10 CFR 50 Appendix A Criteria 14, 30 and 31.
These criteria jointly require that in all ways possible the probability of rapid fracture or gross 'upture of a reactor be kept extremely low.
r The Board is not persuaded that adequate reactor ductility will be ensured by a measure which is directed to determining the temperature of the onset of complete brittleness.
The relation between the highest temperature where reactor material is completely brittle and the lowest temperature at which it is completely ductile is neither simple nor constant.
The Board finds that the RTNDT technology by which Staff would meet Appendix A requirements is defective and does not provide assurance that these requirements are met.
The Board agrees with Intervenor that there is no demonstration that capsule contained coupons are an adequate surrogate for the reactor vessel itself in terms of response to neutron fluence, the mechanical stresses of pressuriztion and depressurization, and thermal stress.
The Board is convinced that if a reactor fails, it will fail at the weakest region.
The record shows that cracks are a common enough threat to be dealt with calculationally.
No basis has been established that the crack sizes assumed may not be exceeded.
Charpy V notch tests on capsule contained specimens can not be expected to show the effects of cracks or crack growth.
The Board views Intervenor's proposal of real time monitoring of reactor welds for abnormal deformation a means of meeting Appendix A.
. 4 at the Catawba her conditions In the matter of extreme wrat Board found that no re bits os claimed by Intervenors theThe testimony of all p i
was ir is controversy existed.
of inversions and quiet a The incidence of population, Charlotte,
.ctnsistent.
center For fully one third of A large unusually frequent.
the plant.
is literally downwind from t carry over Charlotte.
l f's winds from the Catawba p anto what extent do Staf weather conditions?-
tha timeThe significant question is,t these extreme ences ovaluations in ths DE3 reflec ff's treatment of the consequ estimate The Board finds that the Sta effect on their essentially noThis is the result not only weather has of extreme or injuries.
nd taking of early fatalities nces, and accident types, asigning remar of averaging weather sequeand relocation, but of as which probabilities credit for lity of severe acidents evacuation The at risk.
low values to the'probabiconsequences to arrive a multiplier to definition of risk.
environment to it uses as Board does not accept this of the public and the ored The fact of exposure a fact that cannot be 16n nces is The Board is severe, possible consequecost benefit weighing un der NEPA.
peration arriving at a f avorabic balance to plant o in not able to find that a struck.
has been I
-h-
~'
A.-
Palmstto Contention 44/CESG Contention 18--
Embrittlement of the Reactor Vessel Intervenors' Contention 44/16 reads as follows:
The license should not 1ssue because reactor degradation in the form of a much more rapid increase in reference temperature than had been anticipated has occurred at a number of PWE's including Applicant's Oconee Unit 1.
Until and unless the NRC and the industry can avoid reactor embrittlement, Catawba should not be permitted to operate.
1.
The underlying concern of Contention 44/18 is with catastrophic failure of the reactor coolant boundary, specifically with reactor breach contributed to by reactor materials embrittlement in excess-of initial design objectives.
2.
This concern is expressed in 10 CFR Part 50, Appendix A.
Criterion 1h--Reactor co61 ant pressure boundary.
The The reactor coolant pressure boundary shall be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture.
Criterion 30--Quality of reactor coolant pressure beundary.
Components which are part of the reactor coolant pressure boundary shall be designed, fabricated erected and tested to the highest quality standards practical.
Means shall be provided for detecting, and, to the extent practical, identifying the location of the source of reactor coolant leakage.
Criterion 31--Fracture prevention cf reactor coolant pressure 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 (1) the boundary behaves in a nonbrittle manner and (2) the probability of rapidly propagating fracture is minimized.
The design shall reflect consideration of service temperatures and other conditions and the uncertainties in determining (1) material properties, (F) the effects of irradiation on material properties, ( ', ) residual, steady state and tr'ansient stresses, and ' i 1:e of flaws.
_ Criterion 32--Inspeco,n c.j r tor coolant pressure boundary.
Components which are part of the reactor coolant pressure boundary shall be designed to permit (1) periodic inspection and testing of important areas 4
and features to assess their structural and lenktight integrity, and (2) an appropriate material surveillance program for the reactor pressure vessel.
In Criterion 31, concerned with fracture prevention, the Offgets of irradiation are one of four explicitly mentioned and material ccncerns, which &lso include flaw size, stresses, prop;rties.
3 Applicant's representations, both in testimony prefiled, cnd in proposed findings of fact, (AFF), focuse on the changes in the RT f reactor vessel materials in response to irradiation UDT (AFF $31-533).
4 Radiation effects are stated to be extensively studied cnd well understood for reactor vessel materials.
5 Thors is no testimony by Applicant that the effects of radiction have been extensively studied and are well understood fpr material in a reactor which has gone through the anticipated number of thermal cycles and pressurization /depressurization cycles.
The brittleness of reactor vessel materials is characterized 6.
RTNDT, which is taken by the nil ductility reference temperature, as the greater of the drop weight nil-ductility temperature or tho temperature 60 F less than the 50 ft-lb and 35 mils lateral expansion temperature as determined from Charpy specimens (Mager end Meyer testimony, p. 4).
7 Irradiation causes embrittlement.
This is indicated by P. 4).
a chift upward of RT!iDT (M e
ie determined by Charpy V-notch tests on samples 6.
The RTgg7 of base metal, weld mets 1, and heat affected zone of base metal cimilarly exposed to neutron irradiation while contained in capsules in the reactor in the re6 on of the beltline where 1
The tests are performed at different temp's.
n;utron flux is highest.
n The findings are graphically represented (Riley testimony, Att. A-1 and A-2).
From this data curves are drawn, presumably representing the average relationship.
The posit' ion of this curve at the 35 mil lateral notch expansion is observed (.035 in.)
as is the crossing of the 50 ft-lb energy absorption level.
The temperature increase in comparison to unirradiated material is takan as the increase in RT The revised value of RT NDT.
NDT' is this increase plus the initial unirradiated value which may have been determined by the drop weight test rather than the Charpy. test.
9.
Intervenor challenges the use of the word " accurate" to describe RT values.
Sometimes Charpy data forms a locus NDT to which an " average" curve can readily be fitted as in Attachment A-1 tc Riley testimony where an RT f -16 F is obtained.
NDT Not infrequently the Charpy data are widely scattered as in Attachment A-2 with the rssult that an RT value can not be NDT derived.
In some instances it is not possible to obtain a curve on the Charpy data of unirradiated base metal to use as a reference point for irradiated sample data (Int. Exh. _-
Figure C-1, p. C-6 from Babcock and Wilcox Ihport 1699).
10.
The Board concludes, with intervenor, that Charpy determined RT values are not accurate and, indeed, may not NDT even be determinate in the sense that they either cannot be quantized, or, if quantized, would be characterized by so large a variance as'to make them useless.
I 11.
The testimony of both Applicant and Staff is found deficient and misleading in that neither testified to the high l
variance in sets of Charpy data; neither testified that some Charpy data sets are indeterminate by reason of high variance.
l l
In tha context of a Board finds Applicantsefoty matter of greatest importance and requirements of AppendiStaff have failed to the uncertainties do x A, meet the not assure Criterion 14 in that t rapidly propaEsting fail extremely he test method an and gross low probability of reference to Criterion 30 rupture; that with ure the highest quality materials have the This Board questions Sstandards prhetical.
not been te ted to 12.
s implement Criteria 14 taff judgment in and pressure Vessel Cod, 30, and 31 by ASME Boiler seeking to (10 at part of the CFR 50 App G II F)
NDT.
It is the This part of the ASMEe,Section II RT temperature nil ductility temperatu code defines full ductility; the iregion the ' shear fractu re.
In the nil ductility re (Riley Att. A-1 full ductilty it ismpset en whereas at at ess than 10 of the order ft-lb R; ports 1699 and 1436), A-2, Int. Exh. Fig. C-1 a of 100 ft-lb or more nd Fi. 7-3 from B&W failure it must be To protect a S
reactor well into the ductile against brittle from completely brittle to full condition.
about a 200 F temperature i y ductile The transition normally occurs Board is not able to find that critncrease (cf. foregoing ref over mat.
.).
This The Board m tter recognized eria 14, 30, and 31 have b these of reactor It finds fault with thcriteria a een safety.
implcmentation cho e
G whsrein the lowestsen by the Staff as in A e means of "edjucted reference temser/ ice temperature, RTppendix G II.E NDT + 100 F, and orittrion.
perature" both depend on the a nil ductility 13.
The estimates btaining these values of end-of-life RT HDT, and the
, reflect the deficiencies methods of noted foregoing.
Copper, nickel, and phosphorus are reported to increase the rate of embrittlement as a function of neutron flux. 'As a result of apparently unexpectedly high rates of embrittlement of coupons exposed to irradiation in a number of reactors, there has been much activity in the area of EOL prediction.
In addition to the Re6. Guide 1.99 method, testimony mentions four new approaches.
The answers show appreciable differences as would be expected considering the variance of the primary Charpy data.
14.,
Taking note of the relation of embrittlement rate to composition, Westinghouse (E), the vendor of the Catawba nuclear steam supply system, and the employer of Applicant's witnesses, developed trend curves (M&M, Attachment C).
These were used to estimate the EOL RT f Catawba reactor material.
(M&M, 6, 7, NDT 10-12.)
15 Using the projected EOL fluence value, W, in 1978, predicted EOL RT 58 and 94 for units 1 and a respectively.
NDT U
(M&M, p. 10).
The value for unit 1 has been corrected to 94,
a change of 36 (M&M, p. 10).
The resulting EOL RTNDT'8
"#8 86 and 109 for units 1 and 2, respectively.
16.
A more recent W trend analysis has resulted in an equation (M&M, p. 12).
[The equation improperly states that RT s e wal a funcMon of Cu and F.
R is change in M NDT NDT that is projected.]
Using this expression the respective RT
cycles would require an alternating stress (S )
f approximately 140,000 psi. "
a (AFF, pp. 544,5)
This is in conflict with Mr. Mager's testimony in which he uses an ultimate tensile stress of 80,000 psi in an example (Tr. 10,906; AFF p. 546).
And it does not accord with the properties of SA-508 which require a minimum ultimate strength of 80,000 psi (ASME, III, Table 1-1,1, p. 393 (1971)).
It is obvious that a material with an ultimate tensile strength of about 80,000 psi would fail if a single loading of 140,000 psi were attempted, not to mention 200 cycles.
Further, examination of the Design Fati ue Curve, Fig. 1-9-1, shows two loci, one G
for ultimate tensile strengths less than or equal to 80,000 psi, the other for ultimate tensile strengths from 115-130,000 psi.
Clearly the values of S, which at If cycles reach 580,000 psi, a
do not translate one-for-one to tensile stress.
Nor has Intervenor so testified.
Intervenor has stated that the 200 cycle value is 20% of thqinitial value.
From Fig. 1-9-1 the 200 cycle value for 80,000 UTS steel is about 153,00 psi for S,.
Extrapolated
to one cycle,.the value of S, is about 1,500,000 psi.
This is ten times the 200 cycle value.
Intervencr is conservative in claiming a reduction to 20% instead of 10%.
- 40. The Board does nt find the specific testimony in regard to the embrittlement of the vessel by service cycles of pressure stress and thermal stress of either Applicant or Intervenor persuasive.
Staff did not address the matter.
The Board is of the view that chances take place in the actual material in the reactor with repeated service cycles.
The Board notes that Applicant's witness, Mr. Majer, documents a strong background in non-destructive exadnadan but does not document a backEround in reactor design or tensile testing. Intervenor 's witness has a background of fatigue testing of polymer fibers, but not in reactor design or metals testing.
Considering the very substantial resources demonstrated by Applicant and Staff in this proceeding, tue Board notes with interest that neither offered an expert witness on the subject of the change in metal properties with stress fatigue.
In the absence of competent expert testimony on the matter of the change in reactor vessel embrittlement and decrease in ultimate strength as a result of 200 cycles of pressurization / heating, depressuriztion/ cooling, this Board cannot reasonably find that the capsule CVN test program accurately indicates the state of the pressure vessel itself as use accumulates.
- 41. Intervenor criticizes the ASMS Section XI reactor program for its inadequacy, the inspection of welds being only token.
(PA/CESG p. 4.)
Applicant apparently agrees.
It has stipulated to a 100 percent inspection of all
welds during the first 10 year inspection cycle (Tr. 11,148-9, McGarry 12/13/83)(AFF p. 547).
Intervenor refers to the need for knowing the condition of both
welds joining the ferritic components of the reactor vessel, and also the condition of the welded austenitic cladding.
The 4
Board places as a condition for the issuance of the Catawba Operating License the commitment by the Applicant to 100 percent inspection of all reactor vessel
welds, including cladding.
The record chows a difference in the thermal expansivity of the ferritic base metal and the stainless steel cladding.
The
SER calls attention to the possibility of cracking in the cladding and the deviation by vendor
Westinghouse in regard to high-heat-input welding practices fron the recommendation of Reg.
Guide 1 43 (
SER pp. 5-7&8).
Cracking of the stainless steel cladding permits
stress corrosion cracking in service (
SER p. 5-8).
42.
Intervenor notes the discrepancy between precise
PTS curves, no confidence limits or measures of variance are associated with the curves (PA/CSSG Att. B-1 and B-2), and the values on which they are based.
demonstrably inprecise ETNDT The Board views as a deficiency in Staff testimony, in which frequent reference is made to
PTS curves Elliot A.5, A.17, A.25, A.26) and the related
PTS screening criteria, that no basis is given for the establishment of the curves and the criteria.
Applicant regards the
PTS treatment of
SECY-62-465 as conservative (M&M p. 17).
Intervenor observes that the requirements of the
PTS curve are not necessarily attainable under faulted conditions (PA/CESG p.7).
Neither Staff nor Applicant rebutted this assertion though both would, if it were rebuttable, be in an excellent position to do so.
The Board finds the
PTS curves, with the large claimed conservatism for current EOL
RT!DT predictions, to be a useful adjunct to safe operation for a
somewhat embrittled reactor under normal operating conditions.
The Board is not able to find the degree of conservatism is as much as Applicant claims.
There is, flaws aside, the unanswered question of-the effects of thermal and mechanical stresses on reactor brittleness, the inappropriateness of using a nil ductility criterion rather than a full ductility criterion, and the uncertainties in the determination of both initial and projected EOL
RTNDT'8 43 Intervenor proposes a real time means of monitoring reactor degradation.
It would apply directly to the structural
welds in the ferritic~resctor shell.
It would utilize the placement of strain
gauges at intervals along each
weld, inputting the strain ~
gauge signals to "an appropriately programmed computer"
- (PA/CESG p. 8).
It is apparent to the Board that such a program would establish each strain
gauge deflection in the initial condition of the reactor.
The computer would select any deviations thet did not correspond to the calculated deviation for fault free material.. Such exceptions would be noted and the region of the reactor abnornal deflection scrutinized for flaws and repaired.
With/ hse the concern becomes one of' excessive flaw growth.
The computer would select-out signals which.showed progressive and continuing growth, leading to examination and repair.
If embrittlenent is accompsnied by changes in elastic modulus, these changes could be tracked.
44 The Bosrd finds Intervenor's suggestion for the real tine monitoring a credible means for meeting Criteria 14, 30, and
- 31 of Appendix A.
The Board finds signHicant uncertainties in in the deterr.ination of
RT and the formulae for the determinstion
NDT of SOL
RT 'DT.
The Eoard finds that a nil ductility criterion h
plus an arbitrary added temperature compensation is unsatisfactory
'os s' ductility criterion.
The Board sees no impe6iment to
' establishing a full ductility criterion in place of the present nil' ductility criterion.
The application of a full ductility criterion would reduce the level of uncertainty contributed by
RT The Board is not persuaded that capsule samples, subjected
NDT.
only to neutron fluence, give a correct indication of the onset
-tsmperature of full ductility of plate and forged base metal, the heat affected' zonec, and the
weldments of the reactor vessel itself as a result of neutron fluence, mechanical (pressurization and depressurization) stress cycles, and thermal (heating and cooling) cycles.
The Board applauds the spirit of the Applicent's offer to perform 1007 inspection of reactor vessel
welds efter ton years of operation.
'dowever the Board is not convinced that such an inspection program is adequate to detect flaw developments whien could occur within this period of time and possibly terminate with reactor breach by vessel fracture.
The Board finds that 100h monitoring of ferritic material reactor vessel
welds in raal time by a strain
gauge system with computerized detection end warning of anomalous responses to be the best presently svoilable means of meeting the requirements of Appendix A which of os11 for "an extremely. low probability of abnormal leakaEe s rapidly propagating failure, and of Eross rupture."
f
B.
Palmetto /CFS3 Contention DES-17 Adverse Meteorology 1.
Contention DES-17 of Intervenors Palmetto Alliance and 02SG ss admitted by this Board states:
The DES is concerned with environmenta] impacts.
Presumably, these are best represented as the entire range from triviel to serious, in conjunction with estimstes of likelihood.
The DES averages meteorological conditions in its consideration of accidents, 5.9 4 5 Because atmospheric inversions and quiet eir are a very common feature in thic region, secident consequences should be calculated for the extreme condition of inversion and very slow air movement.
In the matter of assessing serious accidents, the environmental assumptions are complex and assin do not appear to consider extreme weather, p. 5-37 The D2S, which differs fran the
CP FES in considerinc severe accidents, is at fault in not considering extreme, but frequently encountered, weather conditions.
2.
In its Memorandum and Order of December 1, 1982, at p. 21, the Board admitted Contention DES-17 and paraphrased it as
.. contend [ing] that the
DSS does not properly evaluate impacts of design basis and severe accidents because it does not isolate and analyse those impsets assuming severe weather."
3 There is concurrence among the parties that there is extreme weather in the reE on of the Catawba plant.
Applicant's
~
i witness does not expect a change to occur in the next forty years I
with respect to stable-air inversions and low wind speed conditions (Cesper, p. 2).
Intervenor's witness, who is particularly well L
qualified, having at one time been in charge of the Charlotte airport weather station, testified that there is a high incidence l
of calms and low wind speeds.
The frequency of atmospheric inversions, which inhibit plume rise and mixing, are among the highest in the United States (Purvis pp. 2, 3).
Staff defines very poor weather conditions as being characterized by inversion conditions and low wind speeds (Head et al A.4.).
Only one of
the 91 sets of weather data taken from a year of hourly observations could be characterized as nonadverse (Read et al, A.16. p. 14).
"A substantial fraction of the 91 meteorological sequences used by the CRAC code are modcrately adverse at high population locat$ons, and several are very adverse."
(Id.)
(Emphasis supplied.)
4 The prevailing wind direction is from the Catawba plant toward Charlotte.
The incidence of this wind direction is about twice a random incidence (Purvis, p. 2, PA/CESG Exh.
KUREG/CR-2239, Fig. A.4-1 Site Wind Rose Data).
5.. Weather conditions are an important parameter in the severity of consequences of severe accidents (Read et al, A.17.)
6.
In the sampling of just one year of data several very unfavorable consequences of a severe accident were encountered.
One was "e6regiously unfavorable as regards the nearest large popda'ti on" (Read et al, A.18., p. 13).
Intervence's witness testifies similarly to the'effect still air and inversions have on consequences of a release of radioactive Bases and particulates (Purvis, p. 3).
7 Of the 91 sites considered in the Sandia Laboratories study, Technical Guidance for Siting Criteria Development,
NUREG/CR-2239, the Catawba site has the fourth highest incidence of a prevailing wind direction (Table A.4-1).
Of these four sites it has by far the largest population in the downwind direction from the plant (Id. Table D.1-4 for population densities 0-5 mi, 0-10 mi, 0-20 mi, and 0-30 mi;
Affidavit of Jesse L. Riley, Nov. 22, 1983, p. 3).
- 8. The Board finds there is no dispute among the parties in relation to weather at the Catawba site.
There is a high incidence of quiet air and of inversions; the direction of the wind is from the plant toward the nearest large popnlation center about one third l
L
'of-the time; For a given release this combination of circumstances
oc4a.
.will tsnd to increased cons 9
" Average valuesThe Final and Drafequences.
reactor year," Tablof environmental rit E per atements present sks e 5 13.
resulting from value is 0.0011; for For accidents early fatalitiesevacuation to 10 mievacua 10 and 25 rrithe the value is two plus reactors 0.00002/ per reactor year fatalities Catawba, the totalFor 80 reactor y.reloc tion be at a
re pectively.
values s for the s
ear A similar are 0.09 and 0.002 10 mean early fatalitie study made by Sandi zero for SST2 and
SST s of 100xP a Laboratories find y for 3 accidents per an SSTl accident and fatalitie s
s for the lifetime reactor year.
of for accident probability of both units is 0 resulting The
, P, of 10 08 using the 11.. Considering s y
value conditions (DES, FES T pecific fatelities for EPZable 5.12)
leads toaccident consequence s for individual end relocation evacuation, 470 fat l calculation a
at 10-25 mi.
a ities for
EPZ ev of 19,000 12 The
- fatalities in theMinimal medic l associated probabilit acua tion a
tr eatment would y7is 10~0 fotelity level worst case from 19 000 increase the is number
{ne projected benefitsasociated with "sup to 24,000 of The 19,000 (DES, F3S, pp. Fof " heroic" medical tportive" med ovidsd.
treatment.
reatment is 13.
The average
-3, F-4, Fig. F-1) not and mean e Soing were 9-43),
obtained by using aarly fatalities in t' hisk, in this s 9 and 10 monssquences by thecontext, is definedrisk concept (
DSS 5 9 4 5(6),
probability.
as the product of
'14.
The Staff states th O probabilities used at there are pbsd in Table 5.10 (DESin regar'd substantial uncertainti to the 5 9 4 5).
accidents es i
These assumed as contribute to k:
uncertainties.
15 There are also uncertainties in the estimation of
'conscquences which may be as large as those in the probabilities (DES, 5.9 4 5(G), p.-5-38).
16.
The magnitude of the uncertainties in the view of Staff is that "the uncertainty bounds could be well over a factor 10, but not as large as a factor 100."
(DES 5 9 4 5(7), p. 5-46) 17 Adding to the uncertainty is the-exclusion from consideration of " Sequences initiated by natural phenomena such cs tornadoes, floods, or seismic events and those that could be initisted by deliberate acts of sabotaSe.
" (5.9 4 5(2) p. 5-36).
18.
This exclusion reflects the present " state of the art
.oftprobabilistic riste assessment."
(Id.)
19..-The consequence model "contains provisions for incorporating
.tha consequence reduction benefits of evacuation, relocation, and other protective actions."
(Id. p. 5-37) 20.
Staff does not know whether evacuation effectiveness would be better or worse than that assumed.
(Id. p.5-38)
- 21..The results.for Catawba " include the benefits of [early evacuation within and early relocation of people from outside the l
pluma exposure pathway EPZj."
(Id. p. 5-38) 22.
Alternative assumptions have not entered into the severe l
cccidont consequence calculations.
23 The Board finds that by excluding the consequences of L tcrn:does,' floods, seismic events, and acts of sabotage, and by Ossuming a high level of effectiveness of evacuation which does not weigh in evacuation problems; inadequate communication and
, warning, weather impeded travel, a demonstratedly functional emer5ency plan suitable for the largest release considered, the Staff has
undcrstated the ran6e of consequences and thereby the mean or m:dien consequences to an extent that the Board has no confidence
~in end does not accept the consequence probability, i.e. " risk"
' figures of Table 5 13 24 Given the bias toward understating consequences, 123, th9 Board rejects the elements of Fig. 5 3, 5 4, 5 5, and 5 6 r010ted to consequence severity for a given set of release and matcorology conditions.
25 The second factor in " risk", probability, contains a numbcr of elements.
One of these is the probability of a given type of eccident.
Four accident scenarios are assumed.
These accidents, end the associated probabilities, are sbown in DES and FES Table 5 10.
(DES p. 5-59) 26.
It is apparent to the Board that the selection of cecidsnts consid.msd will be a factor in the estimation of conscquences and, hence, risk.
The exclusion of more severe accidsnts and the inclusion of less severe accidents will clearly offcet the outcome. (head et al, A-19)
The very low probabilities projected for severe accidents 27.
there haVing (Table.5.10, p. 5-79) are not based on experience, bacn approximately 500 years of reactor operating experience at tho time the DES was drafted (DES 5 9.4 3, p. 5-29).
To verify thac9 probability astimates would take, if they are correct in m;gnitude, of the order of five million years based on the dvcnt V probability of once in 500,000 years and the statistical attidoline of at least a population sampling of 20 items to Ostimate cn average value and standard deviation.
28.
The extraordinarily low probabilities for severe impacts (DES Table 5.11, p. 5-80) of 10-8 result when low accident
probabilities are combined with low " worst case" meteorological probability.
29.
In the Summary of environmental impacts and probabilities it appears possible to expose 44,000 persons to greater than 200
rem, more or less the threshold for early fatalities (D3S 5 9 4 2(3),
- p. 5-28); 270,000 persons to over 25 rem, the threshold for acute dadiation sickness (id.); deliver a dose of 42 million person-rem cause 8 in a 50 mi radius and 160 million person-rem total,
[ sic, 16,000 ?],300 latent cancers in a 50 mi radius and a total of 6,100/; and result in an offsite cost of $6.7 billion dollars for mitigating actions.
30.
This obviously unacceptable set of consequences is assimilated by obtaining a risk value which is one tenth of a billion as large because the estimated probability is 10~0*
31.
This Roard has a different concept of risk than the one acted on in the DES.
The Board views it as the magnitude of t.he hazard to which an individuel, a collection of individuals, or a defined object is exposed.
Even where it is known that a risk is well established by experience, as death by vehicular accident, and though the chances for a person in this country to die by such accident are about one in 5000 per year, in a randomly selected Eroup of 5000 persons it will not be known before the event when that death will occur.
So it is with a nuclear accident.
Whatever the probability may be over a long period of time, when the accident will occur is not known.
It may or may not occur in the 40 year operating life of the plant.
32.
The Board finds the hazard, the maximum hazard to which people and environment are exposed by operation of the Catawba plant is that referred to in t29 foregoing with the worst case consequences referred to in t12 of 19,000 to 24,000 early fatalities.
33 The Staff places at one chance in 10,000 a set of zero impacts (
DSS Table 5 11).
The impacts are seen as still essentially nil at one chance in 100,000 (id.).
34 Given the figures of 533 it could be argyed that with so unlikely an average hazard there is no need for emergency
. planning as required by 10 0FR 50 47.
35.
The DES, without specifying.which, finds that the costs attributable to the
uranium fuel cycle and to plant accidents "are either negligible or range from small to moderate."
(
DSS p. 6-3)
The Board does not accept the-view of the DES that the costs related to'possible accident consequences can be disposed of so simply and so neatly.
In view of the inventory of radionuclides, DES Table 5.8, and the possible path; for release, DIS Appendiz E, which appear to be strongly supported as fact, and the weather, distance and demography considerations, which ue find to be fact, it is pcasible. for an sceident at the Catawba plant t6 have extremely severe consequences.
The exposure of literally hundreds of thnusands of people to this hazard can not be ignored.
It must be teken into balance.
'36.
This Board rejects the " risk" concept adv9nced by the Staff in the
DFS.
The probebilities assigned to major uccidents ere clearly speculative.
The experience does not exist on which
~
to base then.
Not one of the mejor accidents experienced in nucles.r: industry to date was forecast.
The probs3111ty assignable to each of these eccidents prior to the evmt was zero.
This was
~
true of the_ loosened graphite plate at Fermi which obstructed coolant flow; of the "nonflannable" polyurethane foam at Browns Ferry
-which burned and ignited the insulation of electric cables; and
.3 of the multiple failure sequences et Three Mile Island unit 2.
There is good cause to find it likely that there remain severe
' accident sequences, as yet unexperienced and unforeseen, for which the attributed probability is presently zero.
37 The Feard finds that the cost benefit weighing mendated by the
NEPA does not result favorably to the Catawba project.
A reasonable presentation has been quentified for a nunber of benefits for the Catawba plant.
However DES Table 6.1 esserts that Sec.~5 9 4
leads to the assessment of a "small" cost.
In fact, the cited section provides a blank bill for a cost which clearly the Staff cannot with specificity foretell but which also has an unquantifiable term pertaining to exposure to hazard.
A hacerd to which hundreds of tho.tcands of peerle are exposed can in our jud-went be considered very large and is not outweighed by the range of large to small benefite.
o.
C05CLUSIO::S OF LAW Embrittlement data obtained on reactor material specimens exppsed ta n;utron fluence do not provide adequate assurance that the Suetility of the reactor vessel is such
- to meet the requirenents
- f 10 CFR 50 Appendix A, Critaria 14, 30, and 31.
An operating licCnsa is denied.
Th3 Draft and Final Environmental Statements fail to reflect the increased exposure to hasard of the public due to extreme mitcor: logical conditions which can cause a severe reactor nccid:nt to result in massive fatalities and early injuries.,
Tha Stcff finding of mininal risk is an outcome of the Staff's d:finition of risk and a calculational procedure which assigns with2ut a sufficient basis extraordinarily lez prooabilities to scrious accidents.
The actual exposure of the public to hazard if tha plant were to operate is a cost which outweighs the sovcral benefits of plant operation.
An opertaing license is d$nied.
Februsry 29, 1954 Respectfully submitted, I' W l
1N
,h s L//
R;bert Guild, counsel for fesse' L. Hiley'/ spokesperson w-for Carolina Environmental Pc1metto Alliance Study Group
Ce UNITED STATES OF AMERICA
.M
EP, -2 E S NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD LFF;LE v? 3E' t:
GCCKEimG & SEN BRANCH In the Matter of DUKE POWER COMPANY, et al.
Docket No. 50-413 50-414 (Catawba Nu: lear Station Units 1 and 2)
AFFIRMATION OF SERVICE I h9reby' affirm that copies of " PROPOSED FINDINGS OF FACT AND CONCLUSIONS OF LAW IN T.dE FORM OF A PARTIAL INITIAL DECISION (ON CONTENTIONS PA44/CESG18, REACTOR EMBRITTLEMENT AND DESL&,
ADVERSE WEATHER)", in the above captioned proceeding, and a cover letter, were served on the following by deposit in the United States mail this 29th day of February, 1984, except for the copy to Judge Kelley which was hand delivered and the copios to Judges Foster and Purdom which were sent by Federal Express on the same day:
Janes L. Kelley, Chairman Richard P. Wilson, Esq.
Atomic Safety and Licensing Board Assistant Attorney General Panel State of
South Carolina U. 5. Nuclear Regulatory Commission P. O. Box 11549
Washington, D. C. 20555 Columbia,
South Carolina 29211 Dr. Paul W. Purdom Robert Guild, Esq.
235 Columbia Drive Attorney-at-Law Decatur,
Georgia 30030 P. O. Box 12097 Charleston,
South Carolina 29412 i
Dr. Richard F. Foster Palmetto Alliance
.P. D. Box 4263 2135 1/2 Devine Street Sunriver,
Oregon 97702 Columbia,
South Carolina 29205 i
Chairman Ron Shearin, Esq.,
Atomic Safety and Licensing Duke Power Company Board Panel P. O. Box 33189 U. 5. Nuclear Regulatory Commission Charlotte, N. C.
28242
Washington, D.C. 20555 Spence Perry, Esq.
Chairman Associate General Counsel Atomic Safety and Licensing Federal Emergency M'gmt Agency Appeal Board U. 5. Nuclear Regulatory Commission Room 840
Washington, D.C. 20555 500 C Street, S.W.
l
Washington, D.C.
20472 I
..-,_..,m
.. _.. ~.
e.
b*
George E. Johnson, Esq.
Karen E. Long Dffice of the Executive Legal Assistant Attorney General Director N. C. Department of Justice U. 5. Nuclear Regulatory Commission P. D. Box 629
Washington, D.C. 20555 Raleigh,
North Carolina 27602 Scott stucky Don R. Willard Docketing and Service Section Mecklenburg County U. S. Nuclear Regulatory Department of Environmental Commission Health
Washington, D.C. 20555 1200 Blythe Boulevard Charlotte,
North Carolina 28203 Michael J. McGarry, III, Eng.
Bishop, Liberman, et al O
(O
) //,,j 1200 Seventeenth Street, N.W.
/
Washington, D.C.
20036
/ Jesse L. Riley
-
