ML20028C935
| ML20028C935 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 01/12/1983 |
| From: | Meyer J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20028C925 | List: |
| References | |
| ISSUANCES-SP, NUDOCS 8301140324 | |
| Download: ML20028C935 (28) | |
Text
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L'NITED STATES OF AMERICA NUCLEAR REGULATORY CONilSSION BOARD _'
' BEFORE THE ATOMIC SAFETY AND LICE l
In the Matter of Docket Mos. 54 247-SP 50-286-SP CONSOLIDATED EDISON COMPANYOFNEWYORK(In 8
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POWER AUTHORITY OF THE STATEOFNEWYORK(I f-DIRECT TESTIMONY OF JAM s for the record Dr.
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d Please state your name and business ad res Q.1 U.S. Nuclear Regulatory Comission, Meyer.
My name is James F. Meyer.
A.1 Washington D.C.
F RC and describe your Please describe your position with the N Q.2 responsibilities in that position.
the reactor system and I am a Senior Task Manager responsible for(corem A.2 containment system portions of severe for risk assessment.
fessional qualifications?
Q.3 Have you prepared a statement of your pro f ssion Yes, I have prepared a Statement of my pro e A.3 attached to this testimony.
What is the purpose of your testimony?
Q.4 8301140324 830112 05000247
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.. -.. A.4 -The purpose of my testimony is to address the contention on mitigationfeatures,namelycontention2.1(aand d),whichstates:
Contention 2.1(a)
A filtered vented containment system for each unit must be installed.
Contention 2.1(d)_
A separate containment structure must be provided into which excess pressure from accidents and transients can be relieved without necessitating releases'to the environment, thereby reducing the risk of containment failure by overpressurization.
Please define the term " mitigation feature".
Q.5 A " mitigation feature" is defined here as an engineered system A.5 designed to mitigate the consequences of severe accidents, tha accidents that are beyond the design basis of nuclear reactor This is accomplished by reducing or eliminating containment buildings.
It is, one or several of the containment building failure modes.
however, important to note here that the existing containment bu adequately mitigate the consequences of a wide range of post r
accidents that are more severe than those considered in the ori A new mitigation feature, combined with an design of the building.
existing containment building design, will mitigate the consequences of an even wider range of severe accidents.
Of the various containment building failure modes that you have Q.6 studied, which is the most important from the standpoint of risk t the public?
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.3-d the NRC Study, the According to both the utilities study anomes from slow
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A.6 greates't single risk contribution c nt building by steam and5; overpressurization of the containmee the containment building
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non-condensibles to the point w er b
structurally fails.*
essurization protection Is there a need then to consider overprcontainment systems?
Q.7 i
beyond that provided by the exist ngdian Point, overpressurization If the goal is to reduce risk at In A.7 protection should be considered.
ssurization failure at How can the risk resulting from overpreThat is, how c this Q.8 Indian Point be reduced?
failure mode occuring.?
either the probability of the A.8 There are two general approaches:
pressurization failures can be accident sequences that lead to over ces to the public can be reduced (prevention) or the consequen pressurization capabili reduced by extending further the over The contention under tion).
the containment buildings (mitiga proach, namely to reduce risk
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consideration here takes the latter apidents that would otherw by mitigating the consequences of acc ssurization.
fail the containment building by overpre sive yielding of structurall ntainment building would take p ace t
- The NRC analysis determined that ex enW ssure.
members (reinforcing bars) for the co at 126 psig.
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- Please describe mitigation features that will prevent or Q.9 considerably delay containment building failure by overpressurization.
(a) accommodate the overpressurization A.9 Any design fix which can:
energies (resulting from decay heat, primary system stored energy, hydrogen production and burning and other sources) by providing ultimate heat sink of sufficient capacity; and (b) prevent or significantly reduce the release of hazardous radionuclides to the These environment should be a candidate for consideration.
candidates include:
FilteredVentedContainmentSystems(FVCSs)(Contention (1) 2.1.a),
Venting to separate Containment Structure (Contention (2) 2.1.d)
Passive Containment heat removal systems, and (3)
(4) Independent Auxillary Containment Spray Systems.
Q.10 Please describe in general tems, the infomation needed in ord Then continue by to make a decision regarding mitigation features.
providing the available information on the FVCS and the venting l
a separate containment structure.
A.10 There are basically five elements in the decision making process--whether or not to require a particular mitigation feature.
F They are:
engineering feasibility, that is, can a practical system be (1) engineered and built which meets the functional requirements imposed.
5-risk reduction, resulting from this mitigat on i
feature.
d recovery costs (2) cost _, of feature including installation, an h
costs (3) and down time costs of units 2 and 3 and how are factored into a value-impact analysis.
ithout this the risk to the public from Units 2 and 3, w f.
i (4) mitigation feature.
hould some measures I'
are Indian Point Units 2 and 3 safe enough or s i,
be taken to reduce risk?
lt) measures trade-offsbetweenprevention,(ofcore-me lt in similar risk (5) 3 and mitigation measures, when they both resu tions of reduction values, taking into account ques tion and completeness and uncertainties in both preven mitigation options.
(1) and (2) in some Of these five elements, I will discuss elements d containment detail at this time with respect to filtered vente I will t structure.
systems, and venting to a separate containmen t s for these two i
also present some order-of-magnitude cost est mitigationfeatures(element 3).
Staff has not perfomed h
of cost of these systems have been made, t e prerequisite for an detailed design studies, studies which would b
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accurate detemination of costs.
ing realistic design l
design studies, but only for the purpose of deve opd l
requiments and criteria such as those foun d in the broader context The fourth and fifth elements will be considereission Q of risk at Indian Point--the subject of Comm
6-Thus, the bases to be considered at a later time in this hearing.
this hearing.
for elements 4 and 5 cannot be presented until later in (2).
I will at this time, however, discuss elements (1) and Q.11 Please describe a filtered vented containment system.
device A.11 A filtered vented containment system (FVCS) is a hich incorporated into the design of the containment building w t osphere.
allows for controlled venting of the containment a m tion of the Controlled filtered venting is a process in which a por the environment containment atmosphere is deliberately released to gy' in a controlled manner through a system of filters and ener Such a pressure relief system would be actuated to ld otherwise lead absorbers.
reduce containment pressure, a pressure which cou t lled and to containment failure and thereby allow the uncon ro This system h e unfiltered release of radionuclides into the atmosp er.
ere accident, would only be used to mitigate the effects of a sev ident.
i.e., an accident which is beyond the design basis acc different Depending on the characteristics of specific designs, lized for these radioactive isotope attenuation factors could be rea For all designs considered by NRC the filtering systems.
iodine are better attenuation factors for particulates and molecular than 98%.
t d would The pressure level at which point the FVCS would be actua e damage states depend on the specific set of dominant-risk-contributor (The " dominant-risk-contributor damage for a given nuclear power plant.
i.f states" are those several damage states which, when using qua risk analysis, are considered the major contributors to the risk from These matters will be discussed in some a given nuclear power plant.
Basically, the pressure levels detail under Comission Question #1).
l cannot be too high so as to vent too late for the venting system to accomodate the pressure surge, nor too low so as to vent unec One pressure presently being considered for FVCS vent activa that approximately halfway between the design basis pressure o containment building and the structural failure pressure (e.g., for
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Indian Point, 100 psia).
Q.12 For what overpressurization events will the FVCS be effectiv A.12 Overpressurization events can be divided into three classes-overpressurization, resulting for example from a hydrogen burn; moderate-rate overpressurization, resulting for example from primary system blowdown and molten core quenching; and t overpressurizations, resulting for example, from core-concrete interactions and/or long-term decay-heat (less than 1%) conversion FVCS cannot accommodate the rapid overpressu-of water to steam.
rization events and therefore would be ineffective in preventing FVCS can be containment failure by such rapid pressure events.
designed to accomodate the moderate and gradual overpressu j
and can therefore prevent containment building failures from su overpressurizations.
Q.13 What are examples of the negative characteristics (a of FVCSs?
8-it can cause the A.13 The FVCS can fail or, even if it works correctly,
'I ction..
failure of.other safety features by adverse' system intera Furthermore,inadvertant thereby introducing other accident paths.
t have been operation can release radionuclides when they may no
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t ined low-2 otherwise' released thereby exacerbating normally con a
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The evaluation of the attendent risks requires consequence accidents.
tions and the main h
the identification of the main FVCS-systems interac Three FVCS-systems interactions have been i FVCS failures.
e asimportant:1/
injection Premature venting could negate the containment spray t i ment (1) system (CSIS) function, which is actuated by con a n If CSIS actuates after FVCS has rem j
overpressurization.
ould result.
most of the noncondensible gases, a strong' vacuum c FVCS A rapid depressurization of the containment building t spray (2) could cause the sump water to flash and the containme i
system may recirculation system and low pressure recirculat on 4
itive fail because of pump cavitation due to insufficient ne suction head.
The FVCS could affect the ability of the Emergency C i
l ith (3)
Injection (ECI) systems to keep water in the core evl If a LOCA has occurred with ECI the systems working.
ill lower activated, a venting of the containment building w f the LOCA. The the pressure acting on the ECI water at the point o f
increase and will lead escape of ECI water into the containment will to a greater boiloff of water in the core.
I 1
80 0167/,
" Report of the Zion / Indian Point Study," SAND i
1/
NUREG/CR-1410,Vol. 1, 1980.
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9-Two types of system failures should also be considered:
(1) Vent Failure to open on demand; Filter failure (filter bypass or loss of filter efficiency).
(2)
Q.14 Have these attendant risks been considered in your assessmen A.14 Yes, a number of'FVCS design options have been eliminated due to Also, attendant risks and uncertainties high attendant risks.
associated with attendant risks are included in assessmen realistic risk reduction.
Q.15 Have costs of these systems been estimated?
Il ranged between 12 and 32 million A.15 Yes, Costs in one study dollars, depending on the seismic design requirements and filtering capability.
0.16 What is the status of the filtered vented containment system (F studies to date?
A.16 Various types of FVCSs have been installed or are being installed For in fast breeder reactor facilities both here and abroad.2/
example, the Zero-Power Plutonium Reactor (IPPR) test facility, _
the Fast Flux Test Facility (FFTF), and the Gennan SNR-300 Lawroski, et al., Final Safety Analysis Report on the ZeroArgonne Plutonium Reactor (ZPPR) Facility, ANL-7471 (Argonne, IL:
2/
National Laboratory, June 1972),
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E all have FVCSs, or will be installing them.
Breeder Reactor (CRBR) prototype LMFBR r
Also, the present design of the Clinch River includes a FVCS.
k (LWRs) have received a
Vent-filter systems for light water reactorsd Swedish studies on
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attention since 1975, when Norwegian anof the surrounding soil an underground siting considered the useO Subsequently, a study rock as a filtering medium.
(UCLA)presenteda f
University of California at Los Angeles M comprised of a graded conceptual design of a vent-filter systemEPA and charcoal fil sand and gravel bed with downstream H burners to minimize the Their design included the use of hydrogen oling fans to prevent likelihood of hydrogen explosions and air coMore recently, overheating of the charcoal filter.
lt accidents was controlled vent-filtered system for core-meround nuclear plants d
considered in a conceptual study of un erg (CEC).6_
the California Energy Commission Off Gas Filter System of Bohn, S. Jordan, and W. Schikarski,"TheCleaning Conference
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the SNR-300," Proc.13th USAEC AirFrancisco, a Reactor Safet 3T.
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Standpoint,
% vember' 1 Rock Sitino of Nuclear Power Plants from TT6aT Report (sweden: centrala oraftTsTningen,as a Mea 4/
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Gossett, et al., Post-Accident Filtrat onContai December 1975.
t Evaluation of the Feasibility, Economic Im 5/
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Rwer PTaiits, Final C. Finlayson, et al., riidercround NucleT5T-14)-1Trerospace Corp Technical ReportMof t 6/
and Effectiveness
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1978).
Buried 'Benn Contained
E. Ward, et al.,Conce tual Design and Es6timated Cost of
& L Engineers.
7/
ower PTants_, c uc ear January 1978).
\\ l filtering structure being an i
. completely passive, with the princ pa shed rock and underground pressure relief volume filled with cru FVCS on Presently, Sweden is initiating-installation of a id ring the installation gravel.
two of their BWR systems and France is cons e j
of FVCS, on their 900 Mwe PWRs.
licensing and Q.17 Does the NRC have a program to address the saf value impact of FVCS7 t Authorization d
A.17 The U.S. Congress, in the Fiscal Year 1978 Bu ge epare a plan Act, directed the Nuclear Regulatory Comission to pr for nuclear for the development of new or improved safety systems In April 1978 the NRC submitted such a plan to be conducted power plants.
Congress, outlining seven key areas of research to f $14.9 million.
over three years at a total estimated cost o the NRC, a Of the various research projects being conducted by FVCS conceptual designs program for the development and analysis of NRC and the Advisory h
was accorded particularly high priority by t e CommitteeonReactorSafeguards(ACRS).
ific plants This program combines risk reduction analysis for spe M rformed at Sandia with actual conceptual design analyses being Results of this NRC/RES sponsored activity This t.aboratories.
specifically for Indian Point are reported infeatures for activity is presently considering mitigationA f
types of reactor systems.
ific application to program, NRR initiated a study of FVCSs for spec licensing actions including Indian Point.
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lusions reached by the Q.18 Based on these studies, what are the conc ftheFYCS(elem
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Staff regarding engineering feasibility otion events that could.
V A.18 Of the three types of overpressuriza as defined above it is the i
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potentially lead to containment fai ure ot be designed to Staff's judgement that a practical FVCS cannid overpressuriz I
This accomodate the first type, namely rap ts reported in judgement is based on RES-sponsored assessmenexchange l
NUREG-CR-1410 and presented at techno ogy f
d te the second type, A practical FVCS car, be designed to accomo ation;II,8_/ how i
namely the moderate-rate overpressur za gnitude of these pressure studies (NUREG-08S0) indicated that the madi rises are not likely to fail the In for these pressurizations.
i there is no need to provide mitigat on d te the third type, A practical FVCS can be designed to accomo ation, which is namely, the more gradual overpressur za Indian Point units. II,8/
i dominant overpressurization event of both f accidents initiated If the FVCS is to mitigate the consequences ohurricanes, to by external events (i.e. seismic events, iderably more stringent floods) the design requirements will be consHowever, FVCS l
and the costs will increase according y.
ideration.
remain feasible systems for further cons d Vented Containment y of meeting, J.F. Meyer Technology-Exchange Meeting No.4, "Filte 8/
21,1980.
and A. Marchese, NRC, July
,2 0.19 Based on these studies and your risk analysis for Indian Po h
isk what are the conclusions reached by the Staff regarding t e r reduction ratio (element 2)?
A.19 The risk reduction ratio can be large (approximately a factor
" as the of five*) for both units when considering " latent fatalities It is not as large when considering "early deaths" i
" risk measure".
2 and a as a risk measure (approximately a factor of 2 for Unit The risk reduction ratios quoted here are factor of 4 for Unit 3).
derived from the specific risk' analyses performed by the Staff Units 2 and 3 which will be discussed in response to Comission These risk-reduction ratios can only be realized if Question 1.
l events the mitigation features are qualified to withstand the externa that have been shown to be the major risk contributors.
Q.20 Based on the mitigation studies, what are the conclusio thecostofthesesystems(element 3).
As discussed earlier A.20 Costs can only be judged approximately.
These estimates have ranged from 12 to 32 million dollars.$500,000 estimates did not include reactor down time (which can ru d
per day) or costs to qualify the FVCS to withstan to $1,000,000 (which, as I already stated, can be the large external events.
considerable).
Q.21 Are you familiar with contention 2.1, part d?
A risk reduction ratio of 5 can be stated equivalently as a risk reduction percentage of 80%.
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A.21 Yes, contention 2.1, part d states:
d as "the following additional specific measure should be requir conditions of operation:
hich
'A separate containment structure must be provided into w L
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d.
excess pressure from accidents and transients can be
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without necessitating releases to the environment, t ere yii U
reducing the risk of containment failure by
. l; overpressurization."
Q.22 Please describe a " separate containment struc d
A.22 This concept simply provides for more containment vol i
by means attendant heat sink and source-term attenuation capabil ty to of a piping hookup from the existing containment building This concept has the potential to another (large) building.
b atly prevent or delay containment building failure and there y i ation reduce consequences to the public from severe overpressur i
t failure.
accidents that would have otherwise resulted in c i
is One such concept, referred to as a Containment Ventin shown in Figure 2.1d-1.
i Q.23 Has this concept been studied by the NRC?Office of A.23 Yes Sandia National Laboratory under contract to N i
Research conducted "A Value-Impact Assessment of A which considered l
Containment Concepts" (NUREG/CR-0165, 6/78)
A
" compartment venting" together with other mitigation fe d t d by more recent study, specific to Indian Point, also con uc e l
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_________________ ____________ _ _ ____ _ ___ _ - - Sandia National Laboratory, considered venting to a second bu tm as an option to a conventional filtered, vented containment sys e
(" Report of the Zion / Indian Point Study", Vol.1, NUREG/
i Q.24 What were the results of these studies?
f the A.24 Basically, they concluded that, if overpressurization failure o i
t containment building was a major contributor to risk, this concep The original study has significant potential safety benefit.
(NUREG/CR-0165) also concluded that costs would be high since aditional containment-like structure would have to b Q.25 1s such a separate containment structure part of any existing nuclear power plants?
A.25 Yes, the Canadians have incorporated this type of mitigation into some of their heavy water nuclear power plants. (E.W. Fee G.E. Shaw, " Vacuum Containment Systems for iWiti-Unit Nuclear Stations," 7th Int. Congress on the Confinement of Radioactivi the Utilization of Nuclear Energy _, Societe Franciase de la
_Radioprotection, Versailles, France, May 1974.).
Q.26 Are you familiar with the testimony filed in the proceedin Union of Concerned Scientists and the New York Pub Research Group (UCS/NYPIRG) concerning Contention 2.1?
A.26 Yes, I am.
Q.27 Please describe the system proposed by UCS/NYPIRG this contention.
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itigation feature would consist i
A.27 As I understand the testimony th s m t structure for each unit of a separate and dedicated containmen protection against i
which has the capability of provid ng overpressurization accidents.
g feasible for each unit f
l Q.28 Is such a separate containment vent vo ume
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t?
at Indian Point, and if so at what cosilable at the site, it is A.28 Assuming that sufficient space is ava however, the cost will t s feasible to build and maintain such sys em,
This cost dollars apiece.
i be very high, approximately $100 mill onactual costs of those i
estimate was determined by consider ng ctures which would a
components of existing containment stru constitute the vent volume system.
which achieves the same Q.29 Have you considered a vent volume system end at less cost?
tainment would be a candidate A.29 Yes, Unit 1 is shut down, and its conThe NRC is stu f
for further consideration. ding as a " separate containment using the Unit I containment buil structure" or " vent volume."
study to date (element 1)?
Q.30 What are the results of the feasibilityNRC study, A.30 Before presenting the results of thed a determination to note that NRC does not go beyon
--.y..
. in its ' consideration of the engineering aspects of this or any That is, NRC does not do detailed engineering mitigation' design.
This, quite properly, is the primary design studies.
NRC does perform conceptual responsibility of the utilities.
design studies for the purpose of developing realistic design A statement of feasibility at this requirements, and criteria.
First, some infomation time must have two important qualifiers.
required to make the final feasibility detennination is lacking, and, second, the feasibility is a strong function of the effect of external events (hurricanes, earthquakes, floods, tornados) on the With these qualifiers in mind, the present Unit I containment.
position on feasibility is:
unless the infomation still outstanding demonstrates new major problems, the use of the Unit I containment as a vent volume appears feasible to significantly reduce or delay the effects of severe overoressurization accidents caused by internal event initiators (including fires)
The use of the Unit I containment as a vent volume do appear feasible for accomodating severe overpressurization This accidents caused by seismic external event initiators.
l 1s because the nominal seismic resistance of IP-1 is l
Sufficient substantially less than that of IP-2 or IP-3.
information to assess other external initiators is lacking.
l l
l Q.31 Please list the pros and cons in drawing the above interim conclusions.
I D
m A.31 The features of the Unit I containment building'that make it an i
attractive candidate are:
The Unit I containment building is a steel sphere surrounded _
1)
It is of the the by a thick-walled concrete building.
i l
large-dry type and was originally designed to withstand overpressure (approx' 25 psig) resulting from large LOCA i
The Unit I containment building reactor accidents.
free-volume is considerable--1.75 x 10 ft3-(compared to 2.6 x
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6 l
10 ft eachforunits2and3). The passive energy absorption 6 3 l
(heat sink) capability of the containment building (exclusive a
of any active heat removal capability) was calculated to be greater than 650 x 10 Btu (equivalent to 6 hrs. of core decay 6
f heat at 1% in IP-2 or 3).
.f The Unit 1 containment building also has additional potential j
2) l active cooling capability (that is, a pot ; tial ultimate heat f
i sinkcapability)viacontainmentsprays(bothinternaland l
At least external to the steel sphere containment boundary).
the external sprays can be supplied with emergency cooling I
water sources (such as fire trucks).
i There appear to be sufficient containment penetrations 3)
I t
available (for Units 2 and 3 as well as Unit 1) to accomodat the necessary vent piping from Units 2 and 3 to Unit 1).
r f
The additional volume and surface area provided by IP-1 could f
4) l increase plateout and condensation effects and thereby reduce l
ambient radioactivity available for release to the l
The negative-pressure annulus between the steel environment.
and concrete containment structures may also enhance such reductions.
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ilding that make it The features of the Unit 1 containment bu unattractive as a vent volume are:d to withstand only l
Indian Point Unit I was originally designe Its normal seismic 1) a 0.1 (horizontal) seismic event.
below that of Units 2 i
capability appears to be considerably i
does not In addition, currently available Informat on i
ifically designed to and 3.
indicate that Indian Point Unit I was spec criteria that we i
meet other external event design criter a--Th l
l imposed on Units'2 & 3.
i st core-melt accidents f
worthy candidate for mitigation aga n initiated by external events, effect of,eight 1
h f
Indian Point Unit I was shutdown in 1974; t eand sys 2) it years of disuse on containment integr y f
ld l
is not known.
t known.
Costs of rehabilitating these systems is no functional (all Unless active containment cooling systems arei m 3) requiring a.c. power), the Unit I conta n bility. Unless f
an unlimited (i.e. ultimate) heat sink capad quate ab these cooling systems operate or unless a eit 1 at provided to re-isolate Indian Point Un tely fail and th i
times during venting, it too may ult ma (viatheannulus h
release fission products to the atmosp ere building).
between the steel sphere and the concrete tial The design of any vent lines would require su ld involve 4) hardening against external events that cou significant costs.
. The potential conbustion of any hydrogen transported to India 5)
Point Unit i under continued venting could require the.
addition of other mitigation devices and further consideration of pressure-reducing c6pabilities of the Indian Point Unit 1 4
containment.
Q.32 What are the Staff conclusions regarding cost of venting t Unit I containment building (element 3)?
A.32 Because of a lack of some very basic information (as discuss However, it can be stated above)thecostestimatecannotbemade.
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that if no further seismic upgrading is needed, then the costs On the other should be considerably below the costs of a FVCS.
hand, if extensive _ upgrading is needed to accomodate external events, then the costs would be equal to or possibly greater than those of a FVCS.
Q.33 What are the Staff conclusions regarding the risk reduction potential resulting from venting to the Unit I containment or venting to the two separate structures as posed by Intervenors.
(element 2)?
A.33 The conclusions are essentially the same for either of the two options, and furthermore are approximately the same as the namely:
1arge potential reactions when considering latent effects; moderate potential reduction when considering early deaths; I
l risk reduction is negligible unless the vent volume is ~
l qualified for the key external events.
l
s 21 -
of a FyCS or a separate vent Q.34 Do you believe that the addition enors should be requ Y
volume as described by the Interv f Indian Point Units 2 and 37 y
condition of continued operation ofiltered vented contain or i
g enors should be made A.34 No, I do not believe that avent volum a separate i
a condition of continued operat on hieve the level of risk ie These systems are costly ways /to acThere are better and le d
Our reduction that they may provi e.
s of risk reduction.
f ways to achieve the same end in term 1 and 5 will discuss testimony in reponse to.0uestionsmitigation and prevention d
risk at Indian Point as well as tandpoint of risk reduction an which may be warranted from the s the level of risk.
ding the use of Unit I containmen Q.35 What is your recommendation regar as a vent volume?
the cost benefit ratio for A.35 I am unable to conclude that t is so high as to obviously venting to the Unit I containmen nsideration of this matter.
preclude at this time further co t will be considered in ou Therefore, this particular concep(discussed fu subsequent testimony.
on Contention 2.1 (a) an Q.36 Does this conclude your testimomny A.36 Yes.
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,4 PROFESSIDHAL QUALIFICATIONS JAMES F. HEYER REACTOR SYSTEMS BRANCH DIVISION OF SYSTEMS INTEGRATION My name is James F. Heyer. I an employed by the U.S'. Nuclear Regulatory C' n-o mission. Washington, D. C.
20555. I as a Senior Task Manager and, as such, am responsible for analyzing and evaluating technical input for licensing actions in the general areas of severe (core degradation, melt or disruption) accidents in nuclear power plants.
i I attended Valparaiso Univeristy. Valparaiso. Indiana from 1958 to 1963. where I received Bachelor degrees in Electrical Engineering and Physics. Upon com-pletion of my undergraduate studies in 1963. I enrolled in the Nuclear Engi-nearing Department at the Pennsylvania State University. In 1965. I received my M.S. degree and in 1968 sqy PhD. both in the subject area of nuclear engi-ncerihg. Fo116 wing gradua'te studie's." I worked for Argonne National Laboratory
~
Physics Division at ANL on development, planning,)
for about six years. During the first five years. I worked in the Appli nd reporting of experiments on Zero Power (Plutonium Reactors. Myexperience uns in reactor analysis, fast reactor experiments, and general engineering (design and development). In addition tp.comple. ting th.e above tasks, sqy responsibilities included being a, reactor supervisor on two plutonium test reactors.
From October 1973 to November 1974 I was on loan from ANL to the Atomic Energy Co:rnission (now NRC) working in the liquid Metal Fast Breeder Reactor Branch.
For about 4 years (including the one year I was on loan from ANL)..iqy duties included analysis, assessment, and evaluation of safety / licensing issues associated with the Clinch River Breeder Reactor with specific responsibility in the areas of fuels, reactor physics, accident analysis, and core disruption analysis. Accomplishments during this period included publishing reports.
c:tablishing. licensing criteria and preparation for licensing hearings. From about August 1977 to August 1978. I had similar responsibilities for the Fast Flux Test Facility (FFTF) culminating in contributions to the Safety Analysis Report for FFTF.
The RRC, especially during the period from August 1978 to August 1979. parti-cipated 1n the Carter. Administration's Non-Proliferation Alternative Systems l
Assessment Program and the International Nuclear Fuel Cycle Evaluati'on Program.
I had lead responsibility, representing the Offica of Nuclear Reactor Regula-tion, in conducting an independent comparative evaluation of the safety.
environmental, safeguards, and licensing issues for the advanced reactors under consideration. Our preliminary assessment has been sent to the Department of Energy and-a Report to Congress was prepared.
Te:ause of my experience in the area.of severe (core melt) accidents for ahanced reactors. I was given similar. parallel responsibilities, during
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2-f the summer of 1979, in the area of PWR and BWR severe accidents and severe accident mitigation features. These responsibilities included analysis and evaluation of Filtered Vented Containment Systems, Hydrogen Control Systems.
and Core Retention Devices. Particular licensing applications included the Zion / Indian Point Task Force consideration of mitigating features for these Power plants and the consideration of degraded or melted cores in safety re-vi.ews for other nuclear facilities.
Expanding on the above activities, I am presently a Senior Task Manager responsible for guidance and coordination of specific NRR. licensing activities related to degraded core / molten core accidents.
While working in all of the above areas, I have been responsible for managing i
a large ($1 million) technical assistance program at various laboratories and universities. Also, I have made numerous presentations before the l
Advisory Committee on Reactor Safeguards.
- During this time period in the evenings'I have taught reactor physics tourses at the University of Maryland.. As a " Visiting Lecturer" I taught three 3-credit graduate level courses indhe Fal,1 of 1976 and 1977 and the Spring of 1978.
i My honors include Sigma X1, a number of Si:holarships and Fellowship awards, and a "Righ Quality Certificate" presented in July 1978. I am the author or co-author of several papers involving reactor physics and reactor safety.
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Publications Open Litersture
- 1. J. F. Meyer and A. M. Jacobs," Point Souree Green's I Functions for Neutral Particle Transport", Eucl.
t-Sci. & Eng., h_O, 239-2145_(1970) i; q
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Open Literature (Abstracts) l
- 1. J. F. Meyer and A. M. Jacobs," Point Source Green's i
Functions for Neutral Particle Transport", Trans.
Am Nucl. Soc. 12(1),161 (June 1969)
- 2. R. A. Lewis, K 7. Dance, E. F. Groh, F. H. Martins F. Meyer and T. W. Johnson, "The Argonne J.
Variable-Temperature Rodded-Zone Facility"), Trans.
' ~
Am. Nucl. Soc., 12(2), 696 (November 1969
- 3. J. F. Meyer, T. E Johnson and J. E. Sustman,"A l
Sodium-Vapor Monitor", Trans. Am. Nucl. Soc.1],(2),
794 (November 1970)
- 4. J. F. Meyer, E. M. Bohn and W. R. Robinson," Bhc Control Rod Worths and Reaction Rates, Sodium-Void Worths and 238-U Doppler Effects Nest BgC Control Rods in a Typical LMFBR Core", Tratis. Am. Nucl. Soc,'
(June 1972)
- 5. Ji (1), 501E. M. Bohn, J. F. Meyer and R. B. Pond," Measu i
ments of the 238-U Doppler Effect in Boron-Poiso: ed.
Zones in a LMFBR-Type Critical Assembly", Trans.
Am. Nucl. Soc. M(1) 502 (June 1972)
Reports
- 1. R. A. Lewis, K. D. Dance, J.F. Meyer and E. M.. Bohn
" Variable Temperature Rodded Zone Project Final Safety Analysis Report" ANL 7638
- 2. R. A. Lewis, K. D. Dance, J. F. Meyer and E. M. Bob
" Variable Temperature Rodded Zone Project Design
~
Summary Report" ANL-7639
- 3. R. A. Lewis, K. D. Dance, J. F. Meyer and T. W.
Johnson, "The Variable Temperature Rodded Zone (VTRZ) Program", Reactor Physics Division Annual Re-port, July 1,1968 to June 30,1969, ANL-7610,p.13 D. Dance, J. F. Meyer and E. F. Grc
- 4. R. A. Lewis, K.
"The Variable Temperature Rodded Zone (VTRZ) Projee Applied Physics Division Annual Report, July 1,196c to June 30,1970, ANL-7710, p.189
- 5. J. F. Meyer, T. W. Johnson and J. E. Sustman, "A Sodium-Vapor Monitor", Applied Physics Division /=
Report, July 1 1969 to June 30,1970, AHL-7710,p3C
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Publications Reports (contir.ued) b (continued)
- 6. K. D. Dance, J. F. Meyer, E. F. Groh and D. M.
Smith, "The Variable Temperature Rodded Zone
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Project", Applied Physics Division Annual Report,?
July 1,1970 to June 30,1971, ANL-7910, p.185 i
- 7. J. F. Meyer and E.M. Bohn," Description of Simulat B c Control Rods and Measurements of Control Rod h
W5rths in ZPR-6 Assembly 7", Applied Physics Division Annual Report, July 1 1971 to' June 30,1" ANL-8010
- 8. J. F. Meyer, E. M. Bohn and R. B. Fond," 238-U Doppler Effect Near the Simulated Control Rod Assemblies in ZPR-6 Assembly 7", Applied Physics Division Annual Report. July 1 1971 to June 30,1 AFL-8010
- 9. E. M. Bohn and J. F. Meyer," Sodium-Void Worth Nes the Simulated Control Rod Assemblies in ZPR-6 Assembly 7", Applied Physics Division Annual Repo; July 1 1971 to June 30,1972, ANL 8010
- 10. R. B. Pond, E. M. Bohn and J. F. Meyer,"Small Sample Reactivity Worths Near the Simulated Contr Rod Assemblies in ZPR-6 Assembly 7", Applied Physics Division Annual Report, July 1,1971 to June 30,1972, ANL-8010 (to be published)
- 11. E. M. Bohn, J. F. Meyer and R. B. Pond," Measureo ments in the Distributed Poison Zones in ZPR-6 Assembly 7", Applied Physics Division Annual Repe July 1,1971 to June 30,1972, ANL-8010
- 12. A.,B. Long, J. F. Meyer, R. A. Moore, and C. D.
Swanson," Correlation of Core Temgerature and Reactivity Drifts in the FTR-EMC, Applied Physid Division Annual Report, ANL-8010 C. D. Swsnson and
- 13. J. F. Meyer, J. W. Daughtry,8-U DoPP er Measure-R. A. Moore," Central Axis 23 l
ments in the FTR-EMC", Applied Physics Division Annual Report, ANL-8010 114. J. F. Meyer, E. F. Groh and W. McDowell,"Modifick tions to the ZPR-9 Doppler Equipment for Multip1q Position Capability", Applied Physics Division Annual Report', July 1,1971 to June 30, 1972, ANL 8010
- 15. J. F. Meyer, P. H. Kier, R. A. Moore and J. W.
Daughtry," Structural Material Doppler Effects in,
the FTR-EMC: Measurement and Calculation",
i Applied Physics Division Annual Report, July 1, to June 30, 1972
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JAMES FREDERICK MEYER_
(Up to' 1973) '
n Reports (continued)
Publications
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- 16. C. E. Cohn and J. F. Meyer, " Doppler.Re (continued) g Applied Physics Division Annual Report, July 1, l
1971 to June 30, 1972
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- 17. C. D. Swanson, A. B. Long and-J. F. Meyer Reactivity Measurements Using an Autorod",
Applied Physics Division Annual Report, July 1,'
1971 to June 30, 1972 e
UPDATING OF PUBLICATIONS Open Literature (Abstracts)
- 6. J. F. Meyer,"R. A. Moore, J. W. Daughtry, Structural Material Doppler P. B. Kier, Effects in the FFTF Engineering Mockup,"
16(1)
Trans. Am. Nucl. Soc. htry, C. D; Swanson, W. Dau 7..J. F. Meyer, J"b3OU Doppler Mappin in the FFTF Engineering Mockup' Critical g Tran. AM.
.R.
A. Moore, Nucl. Soc.16(1)
- 8. C. E. Cohn,' J. F. Meyer, " Doppler Reactivity Traverses by Inverse Kinetics," Trans. Am.
Nucl. Soc.16,(1)
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Publications Since 1973 I
R. Bari..W. T. Pratt, and J. F. ' Meyer, " Severe Accident Trends in Light Water
. a Reactors." International meetin on Thermal Muclear Reactor Safety. Chicago ' ~
111., August 29-Sept. 2,1982 to be'pubitshed) 7
'J. L. Carter and J. F. Meyer. "The Vapor Pressure Equation of State In Core Disassembly calculation." Transactions of the American Nuclear Society Vol. 2
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d J. F. Meyer, L. hois, J. Carter and T. P. Speis,'"An Analysis an Accident Evaluation of the Clinch River Breeder'Riactor Core Disruptive Energetics," NUREG-0122, March 1977.
e n d T. G. Th'eofanous, i
R. E. Alcouffe, L. Lois, J. Meyer, T. P. Speis a d t for LMFBRs,"
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Trans. Am. Nucl. Soc., & 22,, 402 (1975).
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. P. Speis, J. F. Meyer, R. P. Denise and T. G. Theofano i
t of t Destructive Accidents and Associated Energetics - An Ass State of Our Understanding, National Laboratory, May 1976.
Energ'etics held at Argonne i
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' J. F. Meyer.
T. P. Speis, R. P. Denise, R. W. Starostecki, L. Lois, d Their Role in and7. G Theofanous, "LMFBR Accident Energetics a l
f ty Licensing,"d Physics (CONF-761001), Oct. 6-8,1976, I
and Relate I
" Risk Reduction Associat; A Regulatory Perspecti
' T. P. Speis, J. F. Meyer, and T. E. Fenstermacher, with Severe Accident Mitigation Features -ing l
20-24,1931, pages 447 460 1
ment Sept.
idents and Their Impact On Licen l
king,"
From Specific Applications To Long Te J. F. Meyer and M. Silberberg, " Severe Acc In the United States: paper presented to the European Atom i
11 13, 1981 Sweden, May I
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