ML20064J915
| ML20064J915 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 01/17/1983 |
| From: | Dennis Bley, Richardson D CONSOLIDATED EDISON CO. OF NEW YORK, INC., PLG, INC. (FORMERLY PICKARD, LOWE & GARRICK, INC.), POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | |
| References | |
| ISSUANCES-SP, NUDOCS 8301180242 | |
| Download: ML20064J915 (39) | |
Text
. - _
i JW.S4D opartrMONDou UNITED STATES OF AMERICA 00LKETED MUCLEAR REGULATORY COMMISSION gyc ATOMIC SAFETY AND LICENSING BOARD Before Administrative Judges:
'83 JAtl 17 P3 :08 James P.
Gleason, Chairman Frederick J.
Shon Dr. Oscar H.
Paris
)
In the Matter of
)
)
CONSOLIDATED EDISON COMPANY OF
)
Docket Nos.
NEW YORK, INC.
)
50-247 SP (Indian Point, Unit No. 2)
)
50-286 SP
)
POWER AUTHORITY OF THE STATE OF
)
Jan. 17, 1983 NEW YORK
)
(Indian Point, Unit No. 3)
)
)
LICENSEES' TESTIMONY OF DENNIS C.
BLEY AND DENNIS C.
RICHARDSON ON CONTENTIONS 2.1(a) AND 2.1(d)
ATTORNEYS FILING THIS DOCUMENT:
Brent L.
Brandenburg Charles Morgan, Jr.
CONSOLIDATED EDISON COMPANY Paul F.
Colarulli OF NEW YORK, INC.
Joseph J.
Levin, Jr.
4 Irving Place MORGAN ASSOCIATES, CHARTERED New York, New York 10003 1899 L Street, N.W.
(212) 460-4600 Washington, D.C.
20036 (202) 466-7000 03 8301180242 830117 PDR ADOCK 05000247 T
s r
TABLE OF CONTENTS A.
Introduction...................................
1 B.
Capability of the Indian Point Containments.... 4 C.
Core Melt Frequency Is A Poor Indicator of Risk........................................
7 D.
Comparative State of Knowledge of the Large l
Pressurized Water Reactor (PWR) Containment l
Versus FVCS and SCS............................
9 E.
Potential Benefits and Detriments of FVSC and SCS..................................
10 F.
Source Term Sensitivity Analysis..............
20 Figure 1....................................
22 Figure 2....................................
23 Figure 3....................................
24 Figure 4....................................
25 G.
Conclusions...................................
26 i
f i
l l
TESTIMONY ON FILTERED VENTED CONTAINMENT SYSTEM AND SEPARATE CONTAINMENT SYSTEM A.
Introduction My name is Dennis C.
Bley.
I am a consultant at Pickard, Lowe ana Garrick, Inc. in reliability, risk, and decision analysis for electrical generating plants.
I snus a principa] investigator on the Indian Point Probabilistic Safety Study.
A statement of my professional qualifications is attached.
My name is Dennis C.
Richardson.
I am the Risk Assess-ment Technology Manager in the Nuclear Safety Department of the Nuclear Technology Division of Westinghouse Electric Corporation.
I was a principal investigator on the Indian Point Probabilistic Safety Study.
A statement of my profes-sional qualifications is attached.
We have read the testimony of Messrs. Gordon R.
Thompson and Steven C.
Sholly addressing Commission-Question 2 and Board Contention 2.l(a) (filtered vented containment system (FVCS)) and 2.l(d) (separate containment system (SCS)) and have substantive causes for rejecting their line l
of reasoning that concludes that " implementation of a fil-I tered vented containment system or a compartment venting system is necessary at Indian Point Units 2 and 3."
UCS/
NYPIRG Testimony of Gordon R.
Thompson and Steven C.
Sholly I
on Commission Question Two, Contentions 2.1(a) and 2.l(d) at l
l 19 (Dec. 28, 1982).
I I
l l
l
[
We will show that Messrs. Thompson and Sholly are incorrect in claiming that "to achieve significant risk reductions, proposed solutions must address accident conse-quences."
Id. at 19.
Either preventive measures or mitiga-tive features can reduce risk.
Thompson /Sholly also fail to show that the risks from the Indian Point plants are unac-ceptable and that either preventive or mitigative features are warranted.
By citing documents which were published prior to the issuance of the Indian Point Probabilistic Safety Study (IPPSS), inappropriately using the containment failure modes identified in the Reactor Safety Study (WASH 1400), and applying generic concepts to plant-specific cases without a plant-specific assessment, Messrs. Thompson and Sholly overstate the feasibility, overstate the risk reduc-tion capability, understate the costs, and deemphasize the t
potential problems with such devices.
A plant-specific, site-specific safety study, the l
IPPSS, has been performed to ascertain the risk from opera-tion of Indian Point Units 2 and 3.
This study was issued in March, 1982, and has received extensive peer review.
It provides essential informationpfor evaluating Board Conten-tions 2.l(a) and 2.l(d).
As will be testified later in this proceeding under Commission Question 1, the risks of opera-tion of these units are very low.
The societal and ind iv id-ual risks fall well within the safety goals regarding off-site consequences just adopted provisionally by the Commis-l
_3-sion.
The results of the IPPSS demonstrate that the dominant scenarios identified in the Reactor Safety Study, such as early containment overpressurization, and uncritically assumed by UCS/NYPIRG to be important for Indian Point, are not applicable to the Indian Point plants.
Addressing key aspects of the Thompson /Sholly testi-many, we agree that the risk to the public from the opera-tion of Indian Point Units 2 and 3 is dominated by core melt accidents.
This is so because it is only such accidents which provide even a theoretical mechanism for releasing a large fraction of the radioactive inventory from the core.
Their statement that "[dl espite the considerable ef forts taken.
core melt accidents dominate risk," id. at 5, ignores this simple fact, Under the assumption that the risks must be reduced, Thompson /Sholly dismiss the effectiveness o efforts to lower risks by decreasing accident sequence frequencies.
They argue for this point by citations to inapplicable i
references, i,. e,., assuming generic studies are completely applicable to the Indian Point plants.
At the r.ame time, they suggest that the FVCS and SCS features would delay core melt by two days.
They are wrong on both counts.
The l
sequences for which the mitigative features would delay core melts do not apply to the Indian Point plants.
On the other
(
hand, effective measures do exist for reducing ccre melt l
frequencies.
In particular, we have pointed out in our l
l 1
testimony regarding the Director's Confirmatory Order that the IPPSS identified risk contributors amenable to improve-ment by plant modifications.
Those cost-ef fective changes are being implemented voluntarily by the licensees and yield substantial improvement in risk.
No such cost-effective gains have been demonstrated for the FCVS and SCS.
Lastly, the Thompson /Shally testimony makes assumptions about the quantity of radioactive material which would be released to the environment under the postulated accident sequences which it considers.
Their conclusions as to the achievable benefits from the EVCS and SCS directly depend upon such source terms.
No analysis of such devices can be valid without a critical assessment of source terms, which intrinsically affect results of any evaluation of the poten-tial benefits.
B.
Cacab ility of the Indian Point Containments A better understanding of the capabilities of the pre-sent Indian Point containments is necessary in order to evaluate the UCS/NYPIRG claims.
While Question 1 will address containment capability in detail, several facts are s ignif icant to this testimony.
These 2.6 million cubic foot structures are strong enough to withstand the largest earth-1.
Licensees' Testimony of Dennis C.
Bley and Dennis C. Richardson On Commission Question 2 and Board Question 2.2.1 (Jan. 12, 1983).
o quakes that can be reasonably postulated for this area.
Another measure of the strength of these containments is the very high internal pressures that would have to be achieved before containment integrity would be lost.
Although it has always been recognized that the actual fail-ure pressure was considerably in excess of the design pres-sure, no specific detailed analyses had been performed for the Indian Po.nt containments until the IPPSS analysis.
This analysis calculated the pressure limit to be about 141 pounds per square inch absolute (psia), some 2.3 tLmes higher than the design pressure.
The structural analysis
~
methods used for analyzing the Indian Point containments accurately predict structural behavior when applied within the buildings' elastic limits.
The structural criterion used for defining the Indian Point failure pressure is the onset of yielding (i.e.,
the point at which materials begin to deform beyond the elastic limit) at the most limiting locations.
The IPPSS containment capability analysis is well sup-ported by other independent analyses.
NUREG-0850 reports an initial calculation of the Indian Point containment failure l
l pressure of 133 psia and anticipates that refined analyses would raise this value to 148 psia.
Additionally, the NRC Staff in its testimony cites a containment failure pressure 1
of 141 psia.
Direct Testimony of James F. Meyer Concerning Contention 2.l(a and d) at 3 n.*
(Jan. 12, 1983).
1 I
Such high containment capability has the following effects on the risk.
First, the IPPSS has determined that pressure surges which would lead to an early containment f ailure due to potential steam spikes or hydrogen explosions occurring singularly or concurrently will almost always be below 140 psia at Indian Point.
Early containment overpres-surization is therefore not a major contributor to the risk of either early or latent health effects.
Second, assuming a core melt and a total loss of con-tainment heat removal capability, it would take a fairly long time for pressure conditions to build up in the con-tainment to reach the failure point.
The IPPSS demonstrates that delayed overpressure failures of this type do not con-cribute to early fatalities.
It was conservatively _ assumed in the IPPSS that containment overpressure failure would occur about 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> af ter accident initiation.
This repre-sents a significant conservatism because it is more likely I
greater than about one day after the initiating event before l
l the 140 psia level could be reached.
I Much can be done in this time, including operator action to terminate the pressure rise altogether; evacua-l tions could be implemented, if required; short-lived j
isotopes would decay away.
If no more than one containment l
fan out of the five installed fans were restored, or one out of the six containment spray modes were made functional, or other ad-hoc recovery measures taken, the containment could maintain its integrity.
The already low failure probability of the Indian Point containments is conservatively estimated in the IPPSS in that these recovery actions following core degradation were not quantified.
If minimum containment heat removal capability either existed throughout the accident sequence or was restored during the recovery period, failure pressures should not be reached and public health consequences from the accident will be negligible.
C.
Core Melt Frecuency Is a Poor Indicator of Risk In the Reactor Safety Study, detailed containment cap-ability studies of the type in the IPPSS were not done and containments were assumed to f ail at a pressure signifi-cantly below that calculated in the IPPSS.
The plant analyzed in the Reactor Safety Study was not representative of the Indian Point units and adequate consideration was not g iven to the detailed phenomena which would be associated with core melt.
Therefore, core melts were closely coupled in this earlier study with containment failure and, thus, radiological risk.
Our understanding of risk levels has changed signifi-cantly because of the advances made in the analyses of the Indian Point containment capability and our consideration of the phenomenology which would be associated with core melt at the specific Indian Point plants.
We now realize that risks from these plants are even lower than previously believed.
Contrary to the UCS/NYPIRG testimony, core melt frequency -- once thought a good indicator of radiological risk -- is, in fact, a poor indicator at Indian Point.
Instead, risks from the Indian Point plants are determined by a few dominant scenarios.
For example, more than 90 percent of the early fatality risk at Indian Point is due to one sequence, the interfacing systems LOCA.
This sequence would also bypass containment and would, therefore, also bypass UCS/NYPIRG's propc sed FVCS and SCS.
Yet, this sequence contributes less than one-half of one percent to the overall core melt frequency.
Many of the studies relied upon by UCS/NYPIRG are based on Reactor Safety Study-type containment failure assumptions found in the IPPSS to be incorrect for the Indian Point units.
Reliance on such studies is, therefore, seriously flawed.
Correspondingly, the reduction factors claimed by UCS/NYPIRG for their FVCS and SCS are based on outdated in fo rma tion.
Whereas UCS/NYPIRG refer to a reduction in the early fatality risk to just 8 percent of its present value l
(page 16, A.24, Thompson /Sholly Testimony), FVCS and SCS are ineffective for bypass scenarios and would not reduce the early f atality risk at Indian Point.
l l
l l
l d>
-9_
D.
Comparative State of Knowledge of the Large Pressurized Water Reactor (PWR) Containment Versus FVCS and SCS The state of knowledge of the design and performance of large PWR containments far exceeds that of FVCS and SCS.
Part of the extensive knowledge of large PWR containments stems from the hundreds of reactor years of day-to-day oper-ating experience and the experience gained from periodic pressure and leakage tests.
The IPPSS analysis and other recent dry PWR containment evaluations have added to this knowledge.
The UCS/NYPIRG mitigation concepts involve modification of the Indian Point containment boundary.
To the best of our knowledge, there are no filtered vented containment sys-tems or separate containment systems in commercial nuclear power plants in the United States.
Although UCS/NYPIRG has referred to Barseback, which is a boiling water reactor, the applicability of this concept to the Indian Point large dry containments is unknown without a plant specific analysis.
Similarly, the relevancy of a proposed filtered venting sys-tem for the Clinch River Breeder Reactor has not been estab-lished by UCS/NYPIRG (page 16, A.23, Thompson /Sholly Testi-mony).
It is difficult to understand how a proposed design for a liquid metal cooled reactor, for which only a few details of the filtered vont are available, would form a persuasive argument for altering the Indian Point plants.
As to a separate containment structure, UCS/NYPIRG did not
e -
identify any specific plant where this concept is even being considered.
Neither of these devices has been constructed and no operating experience exists.
We cannot know the true value of their specific proposals when no details are provided.
It appears that UCS/NYPIRG has not carefully evaluated the worth of these systems either, for they quote NUREG/CR-0165, NUREG/CR-0138, and SAND 80-0887 for reductions in early fatalities due to filtered vents.
Given our current state of knowledge, they do not apply to the Indian Point plants.
As stated previously, for early containment failures, FVCS would not reduce early fatality risk at Indian Point.
E.
Potential Benefits and Detriments of FVCS and SCS UCS/NYPIRG does not provide plant-specific probabili-ties and overstates the reduction in health effects associ-ated with these proposed systems.
Although they would have no effect on early fatalities due to early containment failures, FVCS and SCS might reduce latent health effects and property damage.
However, the Indian Point plants already~ meet the early and latent fatalities safety goals adopted by the Commission.
The FVCS and SCS do not prevent accidents; nor do they reduce the frequency or types of accident initiators or of core melt.
The principal health effect where the EVCS and SCS might have a role to play is in reducing the latent fatality
R.
risk.
Certain severely demanding design criteria must be met or these features may fail to realize even this limited role or indeed might even contribute to increasing the con-sequences of an accident.
The FVCS and SCS must be capable of withstanding all of the present initiating events which dominate the latent fatality risk in such a manner as to preclude failure modes that could cause an early release of radioactive material.
For example, the present pressure boundary of the Indian Point containments will not fail from any credible seismic event.
Similarly, the pressure bound-ary of the FVCS and SCS must not fail from any credible seismic event, otherwise a potentially late overpressuriza-tion scenario could become an early release scenario.
Such a failure would cause not only latent fatalities and property damage, but also add early fatalities.
Similar logic must be applied to fire and vind initiated scenarios which, with seismic scenarios, now influence the latent fatality risk at Indian Point.
Assuring that this criterion is met requires careful and detailed analysis which would have a large impact on the potential costs of these sys-l tems.
The best approach to this is to perform a probabilis-tic risk assessment on a specific design as was done in the IPPSS.
In order to properly evaluate in a quantative manner any proposed hardware change to a plant, including the FVCS and SCS, probabilistic risk assessment techniques should be utilized.
This was in fact clearly intended in the Commis-sion's Orders establishing this proceeding.
" Risks from nuclear power reactors are defined by the probabilities and consequences associated with potential accidents.
Despite these uncertainties (associated with quantitative risk assessment calculations), risk assessment methods offer the best means available for objective and quantitative comparison of the kind needed here."
Memorandum and Order at 8 (Jan.
8, 1981).
One part of this technique is the use of the contain-ment event tree which provides a thorough and structured study of the effect of the system or system changes on the progression of the core damage events and containment 1
integrity.
For each proposed feature, modifications should be made to the split fractions where the feature would affect the outcome of the containment event tree node (either positively or negativcly).
These split fractions should then be used to establish a revised containment matr ix for each mitigating feature.
These modifications must then be propagated through the risk calculations.
Comparing the family of risk curves thus calculated with the base risk identifies the risk reduction which the addition of the feature will afford.
This methodology is the manner in which the effect of a mitigation system on risk should be quantified.
This proce-dure should include the following:
o Indian Point plant specific accident initiation frequencies, The inclusion of a specific study of the o
effect of operation, maloperation or failure of the mitigation system on the probability of reaching different degraded core situations.
The effect of the mitigation system o
operation or misoperation on the ccn-tainment failure probability.
The effect of the mitigatica system o
operation or misoperation o.1 the timing of and type of radiological release.
o The inclusion of the above items in a detailed assessment of the probability of and consequences of core meltdown accidents at Indian Point.
Without performing this detailed plant specific assess-ment for internal and external events, the risk reduction of a mitigation system for the Indian Point units cannot be quantified.
In fact, without working with a plant specific
^
risk study (i.,e.,
the IPPSS), the risk reduction cannot even be estimated.
Before performing this detailed and costly assessment of a mitigation system, three questions must be considered; i
o First, is the analysis necessary, con-sidering the low risk already present at the Indian Point plants and the poten-tial for further reductions in estimates of risk due to source term research?
Second, does the technology exist?
o o
Third, are the costs appropriate?
If these systems were backfitted to the Indian Point plants, a detailed specification of design criteria and performance requirements would be necessary.
These speci-fications would have to account for a wide variety of condi-tions under which the systems may be implemented or called upon for use, including flow rates and steam / air / hydrogen compositions which cover a uide range of postulated acci-dents.
A preliminary checklist of design requirements as set forth below was taken from NUREG/CR-1410, referenced by UCS/NYPIRG.
However, the items on this list, which indicate the complexity of designing such systems, were not discussed in UCS/UYPIRG's testimony.
Checklist From NUREG/CR-1410 1.
Functional Requirements a.
Containment pressure reduction b.
Containment temperature reduction, if necessary c.
Mitigation of radioactive release 2.
Operational Requirements a.
Decontamination factors for impor-tant isotopes l
b.
Quality assurance criteria (espe-cially for sand filters)
Maximum filter loading capacity and c.
fission product re-entrainment characteristics d.
Maximum and minimum flow rates and pressure drops e.
Heat removal and condensate drain-age requirements f.
Capability to withstand operating environment g.
Instrumentation requirements 3.
Resistance to Hazards a.
Resistance to earthquakes, torna-does, and missiles b.
Resistance to fire and hydrogen explosions within filter system c.
Resistance to steam explosion from within containment 4.
Reliability a.
Valve actuation reliability b.
Reliability of mechanical compon-ents (air coolers, hydrogen recom-biners, and heat exchangers) c.
Likelihood of spurious operation d.
Likelihood and impact of human error e.
Filter failure or bypass modes and likelihood of occurrence f.
Emergency power requirements 9
Redundancy 5.
Control a.
Actuation logic b.
Flow rate control c.
" Z e ro-r eleas e" options 6.
Sabotage Protection a.
Passive operation versus operator control b.
Protection of piping, valves, and filters from unwanted access r-
- M
7.
Inspection Considerations a.
Ease of access b.
Frequency of inspection c.
Inspection objectives:
(1)
Evidence of struccural damage or degradation (2)
Water infiltration, weathering t
(3)
Contamination with foreign matter d.
Impact on plant operating procedures 8.
Testing Considerations a.
Frequency of testing b.
Testing objectives:
(1)
Efficiency degradation versus time (2)
Flow resistance versus time (3)
Component availability i
c.
Testing methods 9.
Maintenance Considerations a.
Ease of access b.
Periodic replacement of filter materials (especially charcoal) c.
Grooming of filters (especially sand bed) 10.
Post-Accident Safety and Repair-Restor-ation Considerations i
a.
Shielding criteria b.
Access to plant after accident
+
c.
Difficulty of restoring reactor to service Certain failure modes of a FVCS or a SCS could even j
aggravate accident consequences.
As mentioned in i
UC3/NYPIRG's testimony, the probability of basemat j
penetration could be increased.
Several other failure modes
)
are listed below.
The first two of these failure modes are l
listed in NUREG/CR-1410, a document referenced by i
UCS/NYPIRG.
l 1.
Subatmospheric Pressures in Containment Building Should an accident occur at one of the Indian i
Point plants, a mixture of steam, air, and entrained water droplets would be created in the present contain-ment.
This mixture, because of its higher pressure, would either pass through the FVCS or expand into the SCS upon opening of an isolation valve.
In either case, the original containment air would be pushed out of the containment leaving a mostly steam atmosphere.
Should the containment, now mostly filled with steam, go through a depressurization because of initiation of the sprays or fans, or just natural condensation cool-ing, the containment would likely become subatmospheric (less than 15 psia).
In general, containments have far greater superatmospheric capabilities than subatmos-pheric capabilities.
Such an event could then lead to early containment failure and early fatalities.
2.
Hydrogen Explosions UCS/NYPIRG expresses a concern about the effects of hydrogen explosions on containment integrity.
The l
analysis in IPPSS shows that this is not a significant concern for the Indian Point containments.
Yet the very designs UCS/NYPIRG suggests promote explosive hydrogen mixtures within the FVCS or SCS.
It is well established that hydrogen-air mixtures are more explosive than hydrogen-air-steam mixtures.
Both the FVCS and the SCS could remove the steam from the containment into the vent system, thereby possibly promoting explosive mixtures within the systen.
3.
Differential Motion, Should a seismic event occur, differential motion between the present containment and a FVCS or SCS, which would be located some distance away due to site space availability, might cause f ailure of any con-necting structure that links the containment to the proposed modification.
Such a failure could lead to early fatalities.
l 4.
Isolation Difficulties Valves that isolate the FVCS and SCS from the containment would be subject to a number of failure modes.
These valves may fail to open, negating the use of the FVCS and SCS.
They might be opened prematurely or stick open, thereby placing an unacceptable load on 1
I
the filterd or suppression pool.
Some of these fail-ures could be initiated by equipment problems or oper-ator errors.
In summary, the immature state of FVCS and SCS tech-nology suggests that there may be numerous unidentified failure modes.
Although some tentative findings as to the potential impact of a filtered source term are presented in the IPPSS, these results reflect a partial preliminary effort without any detailed evaluation of the factors set forth above at pages 14 to 17.
UCS/NYPIRG's testimony fails to offer an Indian Point-specific design for a filtered vent.
Accordingly, the IPPSS preliminary effort offers little insight into the value of the hypothetical filtered vent proposed by UCS/NYPIRG.
No comprehensive risk analysis has been made of any of these modifications joined to and interacting with the pre-sent Indian Point containments.
Consequently, their risk reduction worth has not adequately been established.
Fur-ther, no regulatory guidance has been issued with regard to the design, licensing, operation, or testing of the FVCS and SCS.
With an adequate research and development program, an actual design, and an associated probabilistic risk assess-ment, the frequency of these potential failure modes of the FVCS and SCS can be minimized.
However, because of the ity of these techno1cgies, such a process would be immate long and costly.
F.
Source Term Sensitivity Analysis As an outgrowth of the Three Mile Island accident, there is keen interest and a tremendous amount of ongoing work in private industry, government regulatory agencies, and various national laboratories regarding source terms, that is, the amount and mix of radionuclides that would be released under various postulated accident scenarios.
Releases of radionuclides smaller than those assumed in the IPPSS would result in a significant reduction in the risk reported in the IPPSS, as well as a significant diminution of the value of all mitigating features.
Testimony justify-ing the use of smaller sources will be presented elsewhere 4
in this hearing under Question 1 for the scenarios that dominate the Indian Point risk.
For the purposes of evalu-ating the potential for risk t' eduction from a FVCS or a SCS, it is useful to present a source term sensitivity analysis.
Figures 1 and 2 show the sensitivity of the whole body man-rem dose to source term reduction for Indian Point 2 and 3, respectively.
Three curves can be compared:
the IPPSS man-rem risk curve, a curve where the source term assumed in i
i the IPPSS is reduced by a factor of 10, and a curve where the reduction factor is 20.
Only the iodine and particu-lates are reduced in these latter curves; noble gas releases are unaffected.
6
~
.____ o Figures 1 and 2 show significant reduction in the latent effects with source term reductions of 10 and 20.
For example, a 20-fold reduction in the source term results in almost 10-fold reduction in the maximum man-rems.
Figures 3 and 4 show similar reductions for the latent f atality risk.
Even if the IPPSS source term is used, both Indian Point units are well within the Commission safety goal limits for latent f atality risk.
Smaller source terms will increase this margin further.
The Commission has recognized the great importance of updating its understanding of source term technology and has large research program well underway in the area, as does a
the industry in its IDCOR program.
Interim results are expected in the near future.
The benefits of mitigating devices such as FVCS and SCS are measured in terms of man-rems averted.
These benefits decrease with smaller source te rms.
Because of the lower risk associated with antici-pated smaller source terms, the Commission has deferred major policy decisions so as to base them on this newer inf o rma tion.
This same reasoning should also be applied to any proposal for a FVCS and SCS, the precise worth of which can only be evaluated after source terms appropriate for the purpose have been determined.
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4 10 10 10 10 10 10 10 10 CAtJCER FATALITIES (OillER TilAN TilYROID CANCER)
EFFECT OF REDUCED SOURCE TERM '0H LATENT CANCER FATALITIES - INDIAN POINT 3 Curve (1):
Totline an.1 particulates raluccxl by a factor of 10.
I Curve (2):
Rxline anl mrticulates raluccx] by a factor of 20.
G.
Conclusions The installation of a filtered vented containment sys-tem or a separate containment spstem at Indian Point 2 and 3 is not justified for the following reasons:
The risk from operation of these plants, o
as determined in the IPPSS, is very low.
o such systems are only potentially useful in reducing latent fatality risks, yet both Indian Point 2 and Indian Point 3 meet the Commission safety goal for this health effect.
o The preesure boundary of the present Indian Point containment should not be changed and possibly compromised.
o FVCS and SCS would not reduce the fre-quency or types of accident initiators.
o FVCS and SCS would not reduce the fre-quency of core melts.
o FVCS and SCS would not reduce the early fatality risk.
o FVCS and SCS technology is immature and unproven.
o The possibility of a negative impact of these features has not been considered by UCS/NYPIRG.
o The worth of any mitigating feature is highly dependent on the source term.
Consistent with previous Commission decisions, no decision to install a device should be made until the research program is concluded.
o The IPPSS identified risk contributors amenable to improvement by plant modifi-cations which are being implemented by the licensees and yield substantial reductions in risk.
Dennis C. Ricnardson
- Eisk Assessment Technoicgy Managar Fenn Stata University, 5.S. 4roscace Engireeri r.g 1963 M.S. Contrai Engi neering 1965 San Diego State University, M.S. Mathematics 1970 University of Pittsburgn, MBA 1980 Mr. Richardson has r,any years of professional and manager.ent experierce f n the nucl ear fiele.
Ha joirad the Pressurf red Water Reactor Division of Westinghcuse in 1972 'tiere he =anaced the Reactor Protection Analysis Group for perfoming nuciaar plant sarety analysis and, most recently, has =ar. aged the Risk Assessment Technology Organir.ation.
Prior to this, Mr. Richardson was wi-h Gulf General At::cic wnere he weried en design of centrcl and safety systems for the gas-cooled nuclear plant.
At Westinghousa, he has parcicipated in and directed a nt:sber of ri sk assessmer.: and safety analysis studies fer a wide variety of appitcaticns.
He was a. principal investigator in both tne Zion Sta-tion and Indian Fofnt Station Reactor Safety Studies.- He directed tne FRA studies for the Westingnousa Owrars Group that addressed the Post-TMI NUREG mguirecents on eoergency proceduras and operator display ncei recents.
Mr. Richardson was tecnnical and prograc canager for the Bri ti s (NHC) Refererce Water P.eactor Safety Stuoy.
He has aisc ied the develeccent of econocic and financial rf sk asuss=ent techniqJes for the use f n new reacter rJCdel design Concepts.
Mr. Richardson is a mecber cf tne IEEE and ANS and has served on the wrking g oups for two standards corrittees.
He is reviewing the sec-tions for the PRA ranual directed by NRC to be finished in 1981.
He f s author or co-author of more than 15 reports and papers dealing with ri sk assass=ent and varicus aspects of nuclear plant design
=,,e,
NAME
+
DENNIS C. BLEY E3UCATION
?h.D., Nuclear Reactor Engineering, Massacnusetts Institu a of Technol o gy, 1979.
Courses in nuclear engineering and c mouter science, Cornell University, 1972-1974 U.S. Navy Nuclear Power School,1963.
University of Cincinnati, 3.5.E.E.,1967.
Courses in Mathematics and Physics, Centre College of Kentucky, 1961-1963.
PROFESSIONAL EXPERIENCE General Sumnary A censul tant a t Pickard, Lowe & Garrick, Inc.,1979-p. esent.
Technical analysis of power plant availaoility and risk.
Cost-cenefit analysis of pcwer plant system changes.
Preparation of technical reports, expert testimony, anc procosals.
Supervision of the technical quality of PLG reports and direction of scme PLG projects.
Instrue:cr at availaoflity, ri sk, and decision analysis courses offered by PLG.
Oyster Creek Procabilistic Risk Assessment (OPSA).
Assisted in the ccmaletion and review of this c:mplete risk assessment of an ocerating SWR performed for Jersey Central Power & Lignt.
Work Order Scheduling System (WOSS).
Assisted in developing the San Onofre 2 and 3 plant model for a c:mputar based work order pricritizing, scheduling, and record keeping system for Soutnern Cali forni a Edision Comcany.
Staam Turoine Of agnostics Cost-Senefit Analysis.
Developed and applied a procedure for evaluating diagnostic alternatives for EPRI.
Reliability Analysis of Diaolo Canyon uxiliary Feedwater System for Pacific Gas & Electric.
Midland Pl an:
a Auxiliary eedwa ar System Reliacility Analysis for Consumers Pcwer.
Technical Review of the " Office of Emergency Services Recommended Emergency Pl anning Zone Considerations..." for Southern California Edison.
Prioriti:ation of NRC Action Pl an for NSAC.
Development of a methodology and participation in an AIF workshop to acply 1: for E?RI/NSAC.
Zion and Indian Point Probabilistic Safety Studies.
Methods development, systems analysis, and plan modeling.
Other ? ras--LaSall e, 3rewns Ferry, Midland, Pilgrim 1, and Ocones.
On USS Enterprise, Reactor Training Assistant, 5 months,1971.
Responsible for technical training of approximately 400 nuclear trained officers and =en prior to annual safeguards examination.
Propul sion
?l an: Station Officer, 9 mon:ns,1970-1971.
Responsible for maintenance and oceration of one propulsion plant (two reactors, eight staam generators, and asscciated ecuipment) during pcwer range tasting of new reactors and during deoloyment.
A: proxima:aly 50 enlistad cerscnnel were assigned to the plant.
Shift Propulsion Plant Wa:cn Officer,15 months, 1969-1370.
Supervised a crew cf about 20 navy enlisted opera: rs and many shicyarc wor'<ers on S-hcur snif rota:icn concu::ing maintenance
SLEY - 2 anc testing in one procuision plant during refueling-overnaul.
Shipboard cualifications:
Proculsion Outy Officer, responsible for all proculsion ecuipment during aosence of Reactor Officer and Engineer Officer.
Engineering Officer of the Waten, coerational waren in Central Control, resconsible for all procuision anc engineering equipment and watch standers.
Procuision Plan: Waten Officer, operational waren in one procui sion plant, directed and responsible for all operations in the plant.
At Cincinnati 3 ell, ?lant staff assistant, 4 months,1967.
Worked in central office and transmission group supplying technical assistance to the line organization.
Cooperative trainee, 3 years,1964-1967, work-study program with alternate three month periods at the University of Cincinnati.
Chronological Semnary 1979-Presen:
Consul tant, Pickard, Lowe and Garrick, Inc.
1974-1979 Massachusetts Institute of Technolocy.
Research assistant for Department of Energy LWR Assessment Project.
Teaching assistant in engineering of nuclear reactors.
Summer 1976 Northeast Utilities.
Engineer:
economy studies, plant startup, analysis of physics tests.
. sci-: e4 U.S. Naval Reserve, active duty.
Instructor of naval science, Corneli University, i
1971-1974; Reactor Department of USS Enter; rise, deployment and re fuel in g-overhaul, 1969-1971; l
Nuclear Power training crogram and Officer Candidate Scncol, 1967-1969.
l 1964-1967
~ Cincinnati Bell.
I
?l ant staff assistant and work-study program tr'ainee.
m l
MEMBERSHIPS, LI^ENSES, AND HONORS l
The Society for Risk Assessment.
Institute of I':ectrical and El ectronics Engineers.
American Nuclear Society.
American Association for the Advancemen: of Science.
The New York Acacemy of Sciences.
U.S. Naval Reserve, Cc==ancer.
Registered Nuclear Engineer, State of California.
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SLEY - 3 Sigma Xi (national science honors socie y),1975.
Sherman R. Knaop Fellcwship (Northeast Utilities), 1975-1975.
Sl oan Research Traineesnip, 1974-1975.
Eta Kappa Nu (national electrical engineering nonors society),1967.
REPORTS AND PUBLICATIONS "Seabrock Probabilistic Safety Assessment," Public Service Comoany of New Hampshire, to be published in 1983.
Pickard, Lowe and Garrick, Inc., " Midland Probacilistic Risk Assessment,"
Consumers Power Company, to be puolished in 1982.
Oconee ?rebabilistic Risk Assessment," a joint effort of the Nuclear Safety Analysis Center, Duke Pcwer, and otner participating utilities, to be publisned in 1982.
Tennessee Valley Aethority and Pickard, Lowe and Garrick, Inc., "Brcwns Ferry Probaoflistic Risk Assessment," to be published in 1982.
Acostolakis, G., M. Kazarians, and D. C. Bley, "A Methodology for Assessing the Risk from Cable Fires," accepted for puolication in Nuclear Sa fety, 1982.
Kaol an, S., H. F. Perl a, and D. C. Bl ey, "A Methodology for Sei smic Safety Analysis of Nuclear ?cwer Plants," procosed presentation a: :ne International Meeting on Thermal Nuclear Reactor Safety, Chicage, Illinois, August 29-Septemoer 2,1982.
Sl ay, D. C., S. Kapl an, and 3. J. Garrick, " Assembling and Decomposing PRA Results:
A Matrix FormaTism," proposed presentation at the In arnational Meeting on Thermal Nuclear Reactor Safety, Chicago, Illinois, August ?c-Septemoer 2,1982.
Garrick, 3. J., S. Kapl an, Tnd D. C. Blay, "Recent Acvances in Precanilistic Risk Assessa nt," prepared for One MIL Nuclear Pcwer Reactor Safety Course, Car.> ricge, Massachusetts, July 19, 1982.
Fleming, K. N., S. Kaplan, and 3. J. Garrick, "Seaorock Procacilistic Safety Assessment Management ?lan,"PLG-0239, June 1982.
Garrick, 3. J., " Lessons Learned From First Generation Nuclear Plant Probabilistic Risk Assessments," to be presented at the Workshcp on Lew-Precability/Migh-Consequence Risk Analysis, Arlington, Virginia, June 15-17,1982.
SLEY a
Garrick, 3. J., S. Xaolan, D. C. Iden, E. 3. Cl ev land, H. F. ?erl a, D. C. Si ey, D. W. Stillwell, H. V. Schneider, and G. Acostolakis, " Power Pl ant Avail ability Engineering:
Methods of Analysis, Program Pl anning, anc Applications,* E?RI NP-2158, PLG-0165, May 1982.
Sley, D. C., and R. J. Mulvihill, " Comments on Evaluation of Availacility Imorovemen: Options for Moss Landing Units 5 and 7," PLG-0225, Maren 1982.
Stillwell, D. W., G. Apos:cl akis, D. C. Bl ey, P. H. Raabe, R. J. Mulvihill, S. Xaolan, and 3. J. Garrick, "EEI Availacility Handbook," PLG-0218, January 1982.
Sl ey, D. C., L. G. H. Sarmanian, and O. W. Stillwell, " Reliability Analysis of Safety Injection System Modification, San Onofre Nuclear Generating Station - Unit 1,".PLG-0205, Oc coer 1981.
" Zion Probabilistic Safety Study," Commonwealth Edison Ccapany, Septamcer 1981.
Buttamer, D. R., " Analysis of Postulated Accidents During Low Power Testing at the San Onofre Nuclear Generating Station--Unit 2," PLG-0199, September 1981.
El ey, D. C., D. W. Stillwell, and R. R. Fray, " Reliability Analysis of Diablo Canyon Auxiliary Feecwater System," cresented at the Tenth Biennial Tocical Conference on Reactor Operating Experience, Cleveland, Ohio, Aucus: 17-13, 1981.
Garrick, 3. J., and O. C. Bl ey, " Lessons Learned from Current PRAs,"
presented to the ACRS Subcommittee on Reliaoility and Probacilistic Risk Assassnent, Los Angeles, California, July 23, 1931.
Xaplan, S., G. Apostolakis, 3. J. Garrick, D. C. Bley, and K. Woodard, "Methocolocy for ?robabilistic Risk Assessment of Nuclear ?cwer Plants,"
draf t version of a cook in preparation, PLG-0209, June 1981.
Perl a, H. F., "?roj ect Plan:
Probabilistic Risk Assessnent, Midland l
Nucl ear Power Pl ant," PLG-0150, May 1981.
Si ey, D. C., C. L. Cate, D. W. Stillwell, and 3. J. Garrick, " Midland Plant Auxiliary Feecwater System Reliability Analysis Synopsis,"
PLG-0156, March 1931.
Pickard, Lowe and Garrick, Inc., "A Methodolocy Oc Cuantify Uncertainty of Ces of Electri:ity for Alternate Designs of (Comoustion) Turcine Comoined Cycl e ?l ants," ?LG-0152, Marcn 1931.
3 LEY - 5
- Garrick, 3.. J., S. Ahmed, and D. C. Bley, "A Methodology for Evaluating tne Costs and Senefi s of Power Plant Diagnostic Technicues," sucaittec t
for presentation a-the Ninth Turbcmacninery Symposium, Houston, Texas, Decemoer 9-11, 1980.
Pickard, Lowe and Garrick, Inc., " Seminar:
Probacilistic Risk Assessment of Nuclear ?cwer ?lants," PLG-015a, Novemoer 1980.
Pickard, L we and Garrick, Inc., " Project Plan:
Prcoabilistic Risk Assess =ent, Browns Ferry Nuclear ?lan: Uni: 1," PLG-0149, Oc::cer 1980.
Garrick, 3. J., S. Kaplan, D. C. Iden, E. 3. Cl eveland, H. F. Perla, D. C. 31ey, and 0.~ W. Stillwell, "Pcwer Plant Availability Engineering, Me nods of Analysis - Program Planning - Applications," 2 Yol s.,
PLG-0148. Oc::ber 1980.
Bl ey, D. C., C. L. Cate, D. W. Stillwell, and 3. J. Garrick, " Midland Plant Auxiliary Feecwater System Reliability Analysis," PLG-0147, Oc coer 1980.
Bl ey, D. C., D. H. Wheel er, C. L. Cate, D. W. Stillwell, and
- 3. J. Garrick, " Reliability Analysis of Diablo Canyon Auxiliary Feedwater Sy stem," PLG-01a0, September 1980.
Garrick, 3. J., et ai, " Project Plan:
Oconee ?robacilistic Risk As ses sment," ?LG-0158, August 1980.
Garrick, 3. J., D. M. Wheel er, E. 3. Cl evel and, D. C. Bl ey,
L. H. Reichers, and C. 3. Morrisen, " Operating Ex:erience of Large
'J.S. 5:eam Turoine-Generators; Volume 1 - Data, Volume 2 - Utility Di rec; ry," PLG-0154, June 1980.
i Garrick, 3. J., S. Kaplan, and O. C. "Bl ey, " Seminar:
Pcwer Pl ant I
Probabilistic Risk Assessment and Reliacility," PLG-0127, May 1980.
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Garrick, 3. J., and S. Xaolan, "0yster Creek Probabilistic Safety Anal-ysis (CPSA)," cresented at the ANS-ENS Topical Meeting on Thermal Reac Or Safety, Knoxville, Tennessee, Acril 3-11, 1980.
Garrick, 3. J., S. Xaolan, G. E. Apostolakis, D. C. Bley, and T. E. Pctter, " Seminar:
Pr babilistic Risk Assessment as Applied to Nuclear Pcwer ?lants," SLG-0124, March 1980.
Garrick, 3. J., S. Ahmed, and D. C. 31 ey, "A Methodology for Evaluating the Costs and Benefits of ?cwer Plant Diagnostic Technicues," PLG-0113, J anuary 1980.
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SLEY - 5 Kaolan, S., 3. J. Garrick, and O'. C. Bl ey, "No as on Risk, ?robacility, and Decision," ?LG-0113, Novemoer 1979.
Sl ey, D. C., C. L. Cate, D. C. Iden, 3. J. Garrick, and J. M. Hudson,
" Seismic Safety Margins Research Program (?hase I), Pr0 ject VII - Systems Analy si s," ?LG-0110, Sectemoer 1979.
Cate, C. L., and 3. J. Garrick, "W-501 Combustion Turoine Starting Reliacility Analysis," ?LG-0103, June 1979.
Pickard, Lcwe and Garrick, Inc., "?lant Availaoility Program Specifica-tion, San Onofre Nuclear Generating Station," Maren 1979.
Pickard, Lowe and Garrick, Inc., " Work Order Scheduling System, Design Soeci fication," March 1979.
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION ATOMIC SAFETY AND LICENSING BOARD DOLMETED Before Administrative Judges:
U3!RC James P.
Gleason, Chairman Frederick J.
Shon 83 JAtl17 P3:08
'Dr. Oscar H.
Paris Lii;u:
3; u,.ci;,
G C N.iiN3 4 SEifvirF BRANCH
~
)
In the Matter of
)
)
CONSOLIDATED EDISON COMPANY OF
)
Docket Nos.
NEW YORK, INC.
)
50-247 SP 50-286 SP (Indian Point, Unit No. 2)
)
)
POWER AUTHORITY OF THE STATE OF
)
Jan. 17, 1983 NEW YORK
)
(Indian Point, Unit No. 3)
)
)
CERTIFICATE OF SERVICE I hereby certify that on the 17th. day of January, 1983, I caused a copy of Licensees' Testimony.of Dennis C. Bley and Dennis C.
Richardson on Contentions 2.l(a) and 2.l(d) to be hand delivered to those parties marked with an asterisk, and served by first class mail, postage prepaid on all others.
4 1
4 4
r-a
,r
.>n
-n
s e
- James P.
Gleason, Chairman Charles M.
Pratt, Esq.
Administrative Judge Stephen L.
Baum, Esq.
Atomic Safety and Licensing Board Power Authority of the 513 Gilmoure Drive State of New York Silver Spring, Maryland 20901 10 Columbus Circle New York, New York 10019
- Mr.
Frederick J.
Shon Administrative Judge
- Janice Moore, Esq.
Atomic Safety and Licensing Board Counsel for NRC Staff U.S. Nuclear Regulatory Of fice of the Executive Commission Legal Director Washington, D.C.
20555 U.S. Nuclear Regulatory Commission Washington, D.C.
20555
- Dr.
Oscar H.
Paris Administrative Judge Brent L.
Brandenburg, Esq.
Atomic Safety and Licensing Board Assistant General Counsel U.S.
Nuclear Regulatory Consolidated Edison Company Commission of New York, Inc.
Washington, D.C.
20555 4 Irving Place New York, New York 10003
- Docketing and Service Branch Office of the Secretary
- Ellyn R.
Weiss, Esq.
U.S.
Nuclear Regulatory Commission Hilliam S.
Jordan, III, Esq.
Washington, D. C.
20555 Harmon and Weiss 1725 I Street, N.W.,
Suite 506 Joan Holt, Project Director Washington, D.C.
20006 Indian Point Project New York Public Interest Research Charles A.
Scheiner, Co-Chairperson Group Westchester People's Action 9 Murray Street Coalition, Inc.
Box 488 White Plains, New York 10602
- Jeffrey M.
Blum, Esq.
New York University Law School Alan Latman, Esq.
423 Vanderbilt Hall 44 Sunset Drive 40 Washington Square South Croton-On-Hudson, New York 10520 New York, New York 10012 Ezra I.
Bialik, Esq.
Charles J.
Maikish, Esq.
Steve Leipzig, Esq.
Litigation Division Environmental Protection Bureau The Port Authority of New York New York State Attorney and New Jersey General's Office One World Trade Center Two World Trade Center New York, New. York 10048 New York, New York 10047
(
Alfred B.
Del Bello l
Westchester County Executive Uestchester County 148 Martine Avenue White Plains, New York 10601 Andrew S.
Roffe, Esq.
New York State Assembly l
Albany, New York 12248 l
l l
l L
... 4 Marc L.
Parris, Esq.
Atomic Safety and Licensing Eric Thorsen, Esq.
Board Panel County Attorney U.S.
Nuclear Regulatory Commission County of Rockland Washington, D.C.
20555 11 New Hempstead Road New City, New York 10956 Atomic Safety and Licensing Appeal Board Panel Phyllis Rodriguez, Spokesperson U.S.
Nuclear Regulatory Commission Parents Concerned About Indian Washington, D.C.
20555 point P.O.
Box 125 Honorable Richard L.
Brodsky-Croton-on-Hudson, New York 10520 Member of the County Legislature Westchester County Renee Schwartz, Esq.
County Office Building Paul Chessin, Esq.
White Plains, New York 10601 Laurens R.
Schwartz, Esq.
Margaret Oppel, Esq.
Zipporah S.
Fleisher Botein, Hays, Sklar and Hertzberg West Branch Conservation 200 Park Avenue Association New York, New York 10166 443 Duena Vista Road New City, New York 10956 Honorable Ruth W.
Messinger Member of the Council of the Mayor George V.
Begany City of New York Village of Buchanan District #4 236 Tate Avenue City Hall Buchanan, New York 10511 New York, New York 10007 Judith Kessler, Coordinator Greater New York Council Rockland Citizens for Safe Energy on Energy 300 New Hemstead Road c/o Dean R.
Corren, Director New City, New York 10956 New York University 26 Stuyvesant Street David H.
Pikus, Esq.
New York, New York 10003 Richard F.
Czaja, Esq.
I Shea & Gould Joan Miles 330 Madison Avenue Indian Point Coordinator New York, New York 10017 New York City Audubon Society 71 West 23rd Street, Suite 1828 Amanda Potterfield, Esq.
New York, New York 10010 Johnson & George l
528 Iowa Avenue Richard M.
Hartzman, Esq.
Iowa City, Iowc 52240 Lorna Salzman Mid-Atlantic Representative "Ruthanne G.
Miller, Esq.
Friends of the Earth, Inc.
Atomic Safety and 208 West 13th Street Licensing Board Panel New York, New York 10011 U.S.
Nuclear Regulatory Commission Stanley B.
Klimberg, Esq.
Washington, D.C.
20555 General Counsel New York State Energy Office 2 Rockefeller State Plaza Albany, New York 12223
e.* s Mr. Donald Davidoff Director, Radiological Emergency Preparedness Group Empire State Plaza Tower Building, Rm. 1750 Albany, New York 12237 Craig Kaplan, Esq.
National Emergency Civil Liberties Committee 175 Fifth Avenue, Suite 712 New Yo~rk, New York 10010 Michael D.
Diederich, Jr., Esq.
4 Fi tge rald, Lynch & Diederich 24 Central Drive Stony Point, New York 10980
- Steven C.
Sholly Union of Concerned Scientists 1346 Connecticut Avenue, N.U.
Suite 1101 Washington, D. C.
20036 Spence U.
Perry i
office of General Counsel Federal Emergency Management Agency 500 C Street, S.W.
Washington, D.C.
20472 Stewart M.
Glass Regional Counsel Room 1349 Federal Emergency Management Agency 26 Federal Plaza New York, New York 10278 Melvin Goldberg Staff Attorney i
l New York Public Interest Research Group 9 Murray Street New York, New York 10007 Jonathan L.
Levine, Esq.
P.
O.
Box 280 New City, New York 10958 b
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?aul F. 'e61a ruTIl
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