ML20002E100
| ML20002E100 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 01/20/1981 |
| From: | Levy S Atomic Safety and Licensing Board Panel |
| To: | |
| References | |
| NUDOCS 8101260405 | |
| Download: ML20002E100 (35) | |
Text
i
.o
-$\\sk f
'b f
0)
.j
'?
.i
?-
t)al.P C L.f id
' : Secret:ry M LIC.1/20/81 \\ A,
14c2 y h
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION 3EFORE THE ATOMIC SAFETY AND T.,ICENSING BOARD In the Matter of
)
)
METROPOLITAN EDISON COMPANY
)
Docket No. 50-289
)
(Restart)
(Three Mile Island Nuclear
)
Station, Unit No. 1)
)
e-6.
m.e 4
LICENSEE'S TESTIMONY OF IA ]
SALOMON LEW g
IN RESPONSE TO UCS CONTENTION NO. 13 (ALTEPNATIVE ACCIDENT SEOUENCES/ CLASS 9 ACCIDENTS) sioaseo4057
-i-CUTLINE The purposes and objectives of this testimony are to respond to UCS Contention 13, wnich asserts that the design of TMI-I does not provide protection against so-called " Class 9" accidents, that the process of selecting events to be consid-ered is faulty, and that there is therefore no reasonable assurance that TMI-1 can be operated without endangering the health and safety of the public.
The testimony demonstrates, in a semi-quantitative way, that the design, safety evaluation and licensing process employed for TMI-1 provides capability to cope with events not analyzed and with events beyond design basis accidents.
Accident sequences having a reasonable nexus p
to the TMI-2 accident are identified, and the prevention and mitigation of the consequences of such accidents are discussed qualitatively.
A quantitative, though approxim' ate, evaluation is presented of design modifications, improvements in operator training, and other actions taken in response to the TMI-2 i
accident, to illustrate the substantial increase in the margin of safety achieved for sequences having a nexu. to the TMI-2 accident.
. INDEX INTRODUCTICN................
................................ 1 RESPONSE TO rJCS CONTENTION NO.
13............................
2 i
. INTRODUCTION This testimony, by Salomon Levy, is addressed to the following contention:
UCS CONTENTION NO. 13 "The design of TMI does not provide protec-tion against so-called " Class-9" accidents.
There is no basis for concluding that such acci-dents are not credible.
Indeed, the staff has conceded that the accident at Unit 2 falls with-in that classification.
Of the realm of possi-ble accidents, the staff's method of determining which fall within the design basis accidents and those for which no protection is required is faulty in that the design basis accidents for TMI do not bound the credible accidents which can occur.
Therefore, there is not reasonable assurance that TMI-l can be operated without en-dangering the health and safety of the public and resumption of operation should not be permitted."
. PESPONSE TO UCS CONTENTION NO. 13 BY WITNESS LEVY:
The design, safety evaluation and licensing process of TMI-I requires the evaluation of a very large number of speci-fic events.
The events considered cover a broad range of fre-quency of occurrence, i.e.,
from normal operation or high fre-quency conditions to design basis accidents or very low fre-quency conditions.
Events to be considered are emot selected through application of a specific method or a numeric probabil-ity goal which subdivides events into two categories -- credi-ble and incredible; instead, the selection process is based upon years of evolving experience, and the careful composite engineering judgment of the NRC, ACRS, and the nuclear indus-try.
The events examined for TMI-1 are discussed in Sections 14 and 3.2.? of the TMI-1 Final Safety Analysis Report (FSAR)(1) and in Section 8 of the TMI-l Restart Peport (2).
In addition to identifying the specific events to be con-sidered, the design, safety evaluation and licensing process In recent years, a listing of events to be considered has been generated and is included in Regulatory Guide 1.70, Stand-ard Format and Content of Safety Analysis Reports for Nuclear Power Plants (Revision 2), issued in September 1975.
In some instances, specific plant designs or site features -(for example, aircraft impact) rescire consideration of additional design ba-sis accidents.
Accidents in this listing agree closely with the accidents recommended for evaluation in the American Nuclear So-ciety Pressurized Water.Peactor Criteria (ANSI No. 18.2).
. j also prescribes different levels of fuel integrity to be main-tained or releases of radioactivity not to be exceeded over-the entire spectrum of analyzed sequences.
The range of occurren-ces examined for TMI-1 can be subdivided into four major categories:
normal conditions, which occur at a very high to high level of frequency and for which a conservative margin is provided towards avoiding fuel failure or release of radioactivity.
anticipated operational events, which occur at a high to moderate frequency (i.e., once to several times per year) and which are required to not produce fuel failure or release of radioactivity.
abnormal operational events with a moderate to low probability of occurrence (in the range of once in ten to one hundred years),
which are required to not perforate more than one percent of the fuel rods and to release only the gaseous radioactivity present in the gap between fuel cladding and fuel pellet of such failed rods.
design basis accidents with a very low
. probability of occurrence (in the range of.
once per oae thousand to ten thousand years), for which fuel cladding peak temp-erature or fuel peak specific energy con-tent must be shown to fall below prescribed values to again limit the number and type of fuel failures and the releases of radio-activity that might ensue.
All the above postulated events were evaluated for TMI-l using conservative assumptions.
" Conservative" means that the performance of the TMI-l plant and safety systems are dell'aerately underestimated and that the fuel conditions and radioactive consequences are deliberately overestimated.
One example of such conservatism is the assumption of the worst or most limiting single failure in any safety-related system or function utilized in the required analyses.
Another example is the assumption that whenever Departure from Nucleate Boiling (DNB) occurs, the fuel cladding fails.
In fact, DNB corre-sponds to the point at which the fuel rod surface begins to be locally blanketed with steam; the cladding temperature at that location rises due to the steam blanketing but does not exceed the 3300 -3500 F temperature threshold at whien cladding perforation may occur. (3)
This will be discussed further later.
One final illustration of the " conservative" approach taken in licensing evaluations is the finding that licensing calculation methods overpredict by about 500* to 1000* F the
~
temperatures measured at the Loss-of-Fluid Test (LOFT) facility during simulated loss-of-coolant accidents.
The impact of such conservative assumptions is that the predicted conse'quences are overestimated substantially or that the probability of the predicted consequences actually occurring is high by at least one order of magnitude per year.
An effort was made herein to approximately quantify the design, safety evaluation and licensing process emplcved at TMI-1.
Figure 1 and Table 1 show the four major groups of occurrences examined for TMI-1 and the fuel failures estimated for such events in the TMI-3 FSAR.
The first group, normal conditions, includes the various modes of possible plant operation, e.a.,
refueling, shutdown, partial power and full power.
During normal operation, fuel failure is conservatively estimated to occur when the margin to DNB goes below 1.3.
(The factor 1.3 is used instead of 1.0 to provide a very high level of confidence that a very large percentage of the fuel rods are in no jeopardy of experiencing DNB.)
At 300 percent power at 4
TMI-1, the DNB margin is expected to be above 2.0 for the most probable conditions.
At reduced power and during shutdown and refueling, the margin would be well above 2.0.
For the worst nuclear, thermal and mechanical conditions during normal TMI-l operations, the DNB margin is expected to exceed 1.55, even at the-maximum assumed overpower limit of 114 percent (2).
The most probable margin to fuel failure and radioactive release in this very high to high frequency range of events is thus
, i estimated to be 1.5 (or 2.0 divided by 1.3) and.to never fall below 1.2 (or 1.55 divided by 3.3).
These margins are indicated by the cross hatching in Figure 1 in the very high to high frequency range of occurrence.
The second group of events, which range from high to moderate frequency, includes many anticipated transients, as listed in Table 1.
The transients are produced by system and control disturbances or component malfunctions.
Because the DNB margin for all such events is required to be above 1.3, no fuel failure or radioactive release is expected.
Table 1 and Figure 1 illustrate the results of the evaluations of such anticipated transients.
The third group of events, which occur from moderate to low frequency, result from a failure of components or operator error with and without an anticipated transient, or from a small break.
This category of events includes sequences having reasonable nexus to the TMI-2 accident since the TMI-2 accident resulted from the anticipated transient of loss of main feedwater combined with failure of a relief valve to reclose.
The events considered in this group are listed in Table 1, with their impact on fuel integrity.
The expected release of fission products is cross hatched in Figure 1.
In this category of events, the number of fuel failures is expected (and required) to be small.
The fourth group of events occur at low to very low frequency.
It includes loss-of-coolant accidents (LOCAs)
. beyond small breaks and extending over the entire spectrum up to and including rupture of the largest pipe diameter in the various plant systems; control rod assembly drop or ejection and fuel handling accidents are also considered.
For all such events, fuel failures are required to be limited and activity released must be kept well below the values specified in Regulatory Guide 1.4 For example, in the case of loss-of-coolant accidents, this is achieved by requiring Emergency Core Cooling System (ECCS). performance in accordance with 10 CFR 50.46.
Even so, under Regulatory Guide 1.4, issued in June 1974, twenty-five percent of the equilibrium iodine and one hundred percent of the equilibrium radioactive noble gas inventory developed from maximum full power operation is required to be assumed to be immediately available for release from the reactor containment for the purpose of calculating offsite dose consequences.
The estimated fuel failures and radioactivity releases for this final group of events are indicated in Figure 1; as discussed above, a conservative demonstration that fuel failures (thus, source terms) are considerably below the values of Begulatory Guide 1.4 is required.
As illustrated in Figure 1, the design, safety evaluation and licensing process employed at TMI-1 requires the considera-tion of a very large number of specific events and the assur-ance that fuel failures and radioactive releases are kept with-in specified limits which are more and more stringent as the frequency of occurrence c ' the event increases.
It was not
l l l intended (nor is it possible) to consider all events that could conceivably occur in a nuclear power plant.
The intent was to consider enough event types over a large and defined spectrum-of frequency to envelope the consequences of events not considered over the range of frequency shown in Figure 1.
Thus, the process does not stop at this point; rather, the process ensures additional capability to cope with events not analyzed and with events which might occur at lower frecuencies than postulated.
The source of this additional capability is the defense-in-depth approach employed in all light water power reactors.
The defense-in-depth concept relies upon multiple barriers to prevent radioactive fission product releases.
The defense-in-depth approach is supported by diverse but interrelated programs for redundancy in protection and safety systems, I
industrial standards and regulatory guides, quality assurance programs, surveillance and preventive maintenance requirements, j
and operator training.
The first barrier of defense-in-depth is the ceramic form of fuel and the fuel cladding employed.
The ceramic pellets of uranium dioxide retain about 98% of the radioactivity generated by the nuclear fission process.
Only approximately 2 percent of the radioactivity in gaseous form is present in the gap between fuel cladding and pellet.
It is released upon failure f
of fuel cladding.
Even when the fuel cladding is heated to a temperature several hundred degrees above the normal operating
. temperature, only the radioactivity in the gap space plus a small portion of the volatile fission products located at the surface of the ceramic pellets are available for release.
Even under the most extreme case of fuel melt, still only 15 to 25 percent of the total radioactivity is released from the molten portion of the fuel.
Any radioactivity released from the fuel is confired by the reactor coolant system piping and vessel, as long as they remain intact.
This second barrier is designed and constructed 3
to the highest quality standards.
Finally, even if the reactor i
coolant system boundary is breached, the containment building serves as a third barrier.
The containment houses the reactor coolant system and is designed to essentially confine any ra-dioactivity that should escape from the reactor coolant sys-tem.
The multibarrier concept of defense-in-depth ensures reserve capability.
For example, focusing upon events of moderate to low frequency (i.e., sequences having reasonable nexus to the TMI-2 accident), there are still one to two barriers at the termination of the event to minimize the transport of fission products outside the plant boundary, as i
indicated at the top of Figure 1.
In the range of moderate to Both siting and onsite and offsite emergency planning pro-vide additional barriers and further defense-in-depth.
These additional barriers are particularly important in consideration of beyond design basis events.
o low frequency of occurrence, only a few fuel cladding failures are permitted.
Thus, for those events involving a small break or leading to a small break of the reactor coolant system boundary:
the integrity of the containment is preserved and the containment is still available to prevent significant radioac-tive release outside the plant.
the pressure in the containment for the worst case analyzed will be below its design pressure.
(Even if the containment were to reach design pressure, it has a reserve capability and will therefore withstand static pressures at least twice the design pressure before it might fail.)
the fission product releases would be quite minimal compared to those prescribed by Regulatory Guide 1.4 and for which the containment has been designed.
Recently, the possibility of even more capability in terms of fission products released within the containment has been highlighted in the August 14, 1980 letter from W.
R.
- Stratton, A. P. Malinauskas, and D. O. Campbell to Chairman J.
Ahearne of the NRC (4).
The letter points out that the radioactive iodine
. released to the environment during the TMI-2 accident was about 100,000 to 1,000,000 times less than the corresponding release of Xenon-333, and recommends that "the frequently quoted fission products escape assumptions (from TID-14844 in 1962 to the more recent Regulatory Guides 1.3 and 1.4, and the Reactor Safety Study, WASH-1400) should be reexamined.
The present assumptions grossly overstate iodine release from a reactor
)
site in many types of loss-of-coolant accidents....."
This topic was addressed further at a Novemb'er 18, 1980 Commission briefing on Iodine Release from Accidents and Estimates of Consequences of Nuclear Accidents (5).
At that briefing, C.
Starr, M.
Levenson, and I. Wall of the Electric Power Research Institute (EPRI) discussed realistic estimates of the consequences of nuclear accidents.
Their comments, which have been amplified in a recent publication by M.
Levenson and F.
Pahn (6), reveal that the radiation releases from past controlled experiments and reactor accidents are one to two orders of magnitude less than the escape assumptions employed in nuclear safety and risk evaluations.
This difference is attributed to inappropriate recognition of "a number of physical processes which are al~ays operative and can be counted on to limit the consequences of a reactor accident."
In particular, the researchers emphasize that the chemical reaction and solubility of volatile fission products, particularly iodine, in water are not accounted for properly, and that aerosol coalescence and agglomeration, steam
1 condensation, and deposition on surfaces will act "to reduce the magnitude of the fission product release and change the character of the release in that iodine and particulates are greatly reduced relative to the noble gases."
The preceding discussion has particularly emphasized the extra capability provided by the containment for events having reasonable nexus to the TMI-2 accident.
However, as shown in Figure 1 and Table 1, similar additional capability can be demonstrated for other events of moderate to low frequency of occurrence which do not lead to a small break of the primary cooling system.
For example, in cases such as uncontrolled control rod group withdrawal, the primary coolant system barrier would remain intact and provide equivalent or extra capability to cope with any fission release during an uncontrolled control rod group withdrawal.
The multibarrier concept also ensures reserve capability for events with a low to very low frequency of occurrence, as illustrated in Figure 1.
The capability to cope does not drop to zero for accidents beyond the design basis events included in Figure 1 and Table 1, because the design process requires that the integrity of one barrier still be available for protection after the occurrence of a design basis accident.
Studies such as WASH-J400 (7) and the ' German risk assessment (8) have examined accident sequences beyond design basis events; they indicate that light water nuclear power plants can cope with events beyond design basis accidents.
In other
. I words, even though some accidents are not required to be analyzed or considered, it is not true that no protection is available or required for them.
The present process, which l
identifies specific events to be considered and imposes stringent limits and conditions on their consequences, ensures additional capability for other accidents.
In the case of TMI-2, an abnormal transient in the moder-ate to low frequency range degenerated to the point where it resulted in more fuel damage than one would have expected from Figure 1 and Table 1.
One of the issues which the NRC Staff and the industry have addressed since the TMI-2 accident is how to ensure that, in the future, a TMI-2 type event will have a much lower frequency of occurrence and/or produce fuel failures i
limited within the range shown in Figure 1 and prescribed by the present process.
4 The present process would put foremost emphasis on limit-ing the number of fuel failures to a very small number.
This is the correct approach because, at the probability level of sequences having reasonable nexus to the TMI-2 accident, it is clearly preferable to prevent the occurrence of a large number of fuel failures rather than attempt to cope with the degraded conditions of the TMI-2 accident.
An alternate process would be to consider events beyond de-sign basis accidents and to impose less stringent limits on their consequences.
This has not been done to date because I
the methods of performing such evaluations have not been fully developed or approved for use in the licensing process.
, Several actions have been taken since the TMI-2 accident to prevent its reoccurrence.
This testimony is not intended to comprehensively address all such actions.
However, in the section which follows, an effort is made to quantify the increased protection resulting from such actions, rather than speaking in qualitative terms only.
Many of the numbers that follow are judgmental.
However, even if one disagrees with the absolute values proposed, there is no doubt that substantial strides have been made to prevent the reoccurrence of the TMI-2 seguence and the occurrence of other sequences having a nexus to TMI-2 accident.
In the original TMI designs, a loss of main feedwater usually led to the actuation of a relief valve (PORV); the probability of such a relief valve failing to reclose, and thus cause a LOCA, is estimated to be about once every fifty times for this event sequence. (7)
Changes made to the main and emergency feedwater systems and differences between the design of the TMI-1 polishing demineralizers system and that of TMI-2 have reduced the probability of the transient being initiated.
Also, the pressure set point of the relief valve and of the reactor protection system will be changed and anticipatory reactor trips will be installed so that the relief valve will not be expected to open for a loss of main feedwater transient.
All these changes are judged to reduce the frequency of the initiating event which results in the opening of the PORV by a factor of at least 2 to 3.
. In the revised design, an inadvertent opening of the relief valve and its failure to reclose or a small independent reactor coolant system break would be required to reproduce the TMI-2 accident sequence.
It is estimated that a relief valve will inadvertently open at a frequency of once per 1000 per year. (7)
Combining this figure with the once in fifty times for the relief valve to stick open, the probability of a small break being produced by a relief valve following a loss of main feedwater transient is now estimated to be about once per 50,000 times instead of once per 50 times.
However, the probability of a small break occurring independently of relief valve actuation is estimated at about once per 1000 reactor years (7), so that the net effect of resetting the relief and reactor protection pressures to avoid a relief valve opening has, in fact, reduced the probability of the TMI-2 accident and its analogs by a factor of about 20 (small break probability divided by probability of relief valve to reclose).
The planned installation of a saturation meter and the associated procedures and instrumentation implemented to help the operator recognize inadequate core cooling (9) will improve the emergency cooling injection availability by a factor of 2 to 3.
The increased operator training required since the TMI-2 accident and planned staffing augmentation (including additional licensed personnel and a Shift Technical Advisor) will provide for another improvement factor of 3 to 5 (equivalent to that available for small breaks produced by
. other means than relief valves).
Thus, T conclude that all tha changes described above provide for an overall improvement factor of about two to three orders of magnitude for a TMI-2 type accident inititated by loss of main feedwater.
The probability of a stuck open relief valve for other transients involving the opening of the relief valve is the product of the probability of the relief valve being demanded
)
to open times the probability of the valve failing open on l
demand.
The raising of the relief valve setpoint from 2255 to 2450 psig, the lowering of the high pressure reactor trip setpoint from 2390 to 2300 psig, and the installation of an anticipatory reactor trip on turbine trip and/or loss of all main feedwater at TMT-1 has considerably reduced the number of demands on the relief valve.
A preliminary estimate indicates that the probability of the relief valve opening on an overpressure transient, or due to operator action, or due to instrumentation control faults is about once per 50 reactor years of operation.
When this figure is multiplied by the probability of once per 50 times of the relief valve sticking open on demand, the probability of a stuck open relief valve from all causes is calculated to be about once per 2500 reactor years of operation; this compares favorably to the probability of once per 1000 reactor years of operation for occurrence of a i
loss-of-coolant accident of that size from a leak in the reactor primary system (7).
-The preceding probabilistic estimates are not intended to
. address all actions taken since the TMI-2 accident or all event sequences having a nexus to the TMI-2 accident.
Substantial testimony is being presented in this proceeding by the witnesses of Licensee and the NRC Staff on the many actions taken in response to the TMI-2 accident.
The estimates presented here were generated to provide a quantitative illustration of the increase in the margin of safety achieved in many typical sequences having a nexus to the TMI-2 accident.
In summary, the present design, safety evaluation and licensing process employed at TMI-1 requires the consideration of a very large number of specific events and the assurance that fuel failures and radioactive releases are kept within specified limits which are more and more stringent as the frequency of occurrence of the event increases.
All the events are evaluated using " conservative" assumptions.
The process also relies upon a defense-in-depth approach which provides additional reserve capability to cope with events not analyzed and with events which might occur at lower frequencies than postulated.
This defense-in-depth approach, together with considerably overestimated releases of fission products, accounts for the "rc:gligible effect on the physical nealth of individuals" attributable to the TMI-2 accident, reported by the President's Commission on the Accident at Three Mile Island (10) and the NRC Special Inquiry Group (11).- The sequences of events having a reasonable nexus to the TMI-2 accident have a moderate to low probability of occurrence.
The correct
. approach at that probability level is to prevent ti'e occurrence of a degraded core rather than to cope with such a degraded core.
A substantial number of actions have been taken since the TMI-2 accident.
Some of the actions taken provide for an overall improvement factor of about two to three orders of magnitude for a TMI-2 type accident initiated by loss of main feedwater.
Similarly, substantial improvement has been made for other overpressure transients involving relief valves; overall, substantial actions bave been taken to ensure that sequences of events having a reasonable nexus to the TMI-2 accident are terminated long before the core reaches a degraded condition.
I L
, REFERENCES (1)
TMI-I Final Safety Analysis Report.
(7)
TMI-1 Restart Report.
(3)
Performance of Unirradiated and Irradiated PWR Fuel Rods Tested Under Power-Cooling-Mismatch Conditions,
{
~
by P. E. MacDonald et al., Fif th Water Reactor Safety Research Information Meeting, November 1977.
(4)
Letter by W.
R.
Stratton, A.
P. Malinauskas, D.
O.
Campbell to J.
Ahearne, dated August 14, 1980.
(5)
Nuclear Regulatory Commission Meeting on Iodine Release from Accidents and Estimates of Consequences of Nuclear Accidents, November 18, 1980, Washington, D.C.
(6)
The Need for Pealistic Estimates of the Consequences of Nuclear Accidents, by M. Levenson and F.
Rahn EPRI, December 1980.
(7)
Reactor Safety Study, WASH-1400 (NUREG-75/014),
October 1975.
(8)
Deutsche Risikostudie, Verleg TUV Rheinland, 1980.
(9)
Licensee's Testimony of R.
W.
Keaten, M. J. Ross and R. C. Jones, Jr. in Response to UCS Contention No. 7, ANGRY Contention No. V (B) and Sholly Contention No.
6(b), (Detection of Inadequate Core Cooling),
(9/15/80).
(10)
Report of the President's Commission on the Accident
~~
at Three Mile Island, October 1979.
(11)
Three Mile Island - A Report to the Commissioners and the Public, Vol.
I, Nuclear Regulatory Commission Special Inquiry Group, January 1980.
. go -
TABLE I.
EVENTS CONSIDERED AT TMI-l I.
NORMAL OPERATION (VERY HIGH TO HIGH FREQUENCY OF OCCURRENCE)
Shutdown, Refueling, Startup, Margin to fuel failure Partial, Full Power Conditions and always in excess of 1.2 Normal Load Following and most probably 1.5 II.
ANTICIPATED TRANSIENTS (HIGH TO MODERATE FREQUENCY OF OCCURRENCE)
Event Considered Impact on Fuel Decrease in feedwater temperature No fuel failure Increase in feedwater flow No fuel failure Steam regulatory malfunction No fuel failure Losa cf external electric load and/or No fuel failure turbine load without runback i
Closure of steam isolation valves No fuel failure Loss of condenser vacuum No fuel failure Loss of main feedwater flow No fuel failure Chemical addition malfunction No fuel failure Inadvertent operation of ECCS No fuel failure III.
ABNORMAL TRANSIENTS (MODERATE TO LOW FREQUENCY OF OCCURRENCE)
Event Considered Impact on Fuel Rod withdrawal accident at rated No fuel failure power operation Loss of external electric load and/
No fuel failure or turbine load without runback Loss of non-emergency AC power to the No fuel failure i
station auxiliaries Loss of coolant flow No fuel failure Startup accident No fuel fa:. lure i
l l
. IV.
ACCIDENTS (LOW TO VERY LOW FREQUENCY OF OCCURRENCE)
Feedwater piping break (entire spectrum)
No fuel failure Reactor coolant pump shaft break or 1% fuel failures
)
seizure and associated gap i
activity release Stuck-Out, Stuck-In, or Dropped-In No fuel failure Control Rod Fuel handling accident Failure of 56 fuel rods and associated gap activity release
(
Steam line break (entire spectrum) 1% fuel failures and associated gap activity release Control rod ejection 17.5% fuel failure and associated gap activity release Primary system piping break 100% fuel failure and (entire spectrum) associated gap activity release assumed
I E /
b
.MO_f....-l(lIlll...
. ~ - ~. - ~-
=
f g"-b
=
J
}
l w uJ, L,M M>- >-
=QO x cC Cg l
==
Ud et _J A O "J kQ UC tu >==
(
O >==
= cC.J i
O m
"f 23 JA
-8
>=
l M
{
M M
1It--------
LLJ rl----'
8
/
l M
Z j
=
M of O A
CC O
- -a l
Z O
.u.J 6 >=
Ld 3
- O
.Z y
cc uJ uJ MJ M
M _J >-
l Z
cC CD C Z
Z
'C O CD*C g )
uJ LaJ Cg >=
y JAT O **
uJ
==
3 *C U I
O J
6-->>=
H W
<C _ a
- C O
O M
Z O
CD Cf W F.
J D t.u aC OO Jg O
2O LaJ -
1 g
=
CD E uJ O
z >=
\\
- C
]----
- xel 0M cg
==
cg M
r-----
8
'd M
/
M U
F-ce z7 C
u La M C O
- = 0 a- '
LM u.
2 6 >-
~ h.J O
z M b-cc Ul O
cC uJ LM i Z CC
==
."D J b-
- C CE O
H CD O
% UJ i
E eC
>" O tu <:C M )
14 6--
O 3
x LuJLy OE c=y tu cg -
= <C U
)
uJ tu M
F-O F- > *=
cg g,;
- C _J
- C >-
w cg O
CD A
CL
>= J D uJ UO O_JL ct
{
=
3 >-
\\
Z l
-M
-...__________,l_..
.lt______
w ce i._
t cc M <C E D
'Z' %
d cc Z
uJ.E WE cc O
- cc Z Q -
OO "a"-
L cd cL.=== W F-cc eC U
l'
- C aC =>_ uJ uJ cm O E J >-
I cc O zOmO LM b-uJ== U <C CC cO-La O JL CC O O ~
O
-J*"*
= cC Z *C U
$.r -
JaC>-
U
<C J
M>
J E
co J JuJ J D O CC Z LJ cc w >= _f a (
DM-O O
O O
O O
N v
O CD
[
w > p.
\\
Q CO O
47 N
6
-MM O~
O-7 7
.i
,7 A11AI13V 031VI30SSV 30 A11 AI13V dV9 30 3SV3138 380 LIV 3 SNICOVl3 3SV3738 ONV il3W GNV 380 LIV 3 1303 01 NISBVW 1303 1N3383d 9NICOV10 1303 1N3383d 22 -
J
SALOMON LEVY l
Susiness Address:
S. Levy Incorporated 1901 South Bascom Avenue, Suite 725 Campbell, California 95008 Education:
3.
S.,
Mechanical Engineering, University of California at Berkeley, f
1949.
M.S.,
Mechanical Engineering, University of California at Berkeley, 1051.
Ph.D., Mechanical Engineering, University of California at Berkeley, 1953.
Experience President, S. Levy Incorporated, September 1977 to present.
Independent engineering consultant to several power utilities, national laboratories, Electric Power Research Institute, Nuclear Regulatory Commission Resear=h Division and several power ecuipment manufacturers; consultant to the staff of the President's Commission on the Accident at Three Mile Island and member of the Industry Advisory Board to Three Mile Island-2 recovery operations.
I General Manager, Boiling Water Reactor Operations, General Electric Company, April 1975 to September 1977.
In this position, was responsible for all the engineering and manufacturing aspects of General Electric's nuclear power business.
General Manager, Boiling Water Reactor Systems Department, General Electric Co.,
i 1973 to 1975.
Responsible for the design and development Of nuclear systems including fuel and for the manufacturing of control and instrumentation systems.
General Manager, Nuclear Fuel Department, General Electric Co., 1971 to 1973.
Responsible for the design, development and manufacture of nuclear fuels for light water reactor systems and for reprocessing of irradiated fuel.
D lD 1 0
^
) 8 omMeMl b
l l
Manager and Design Engineer, Atomic Power Equipment Department, General Electric Co., 1968 to 1971.
Responsible for the design engineering of all nuclear systems being offered by General Electric and for all project management functions associated with domestic nuclear j
systems.
l Manager, Systems Engineering, Atomic Power Equipment Department, General Electric Co.,
1966 to 1963.
Responsible for conceiving and definnng nuclear power plant systems for all requisition and proposal plants and all near-term improvements in nuclear power plant systems.
Manager, Heat Transfer and Reactor Projects, Atomic Power Equipment Department, General Electric Co., 1959 to 1966.
RespoJsible for heat transfer and fluid flow development in boiling water reactors.
Advance Nuclear Specialist, Atomic Power Equipment Department, General Electric Co.,
1956 to 1959.
Involved with advanced reactor concepts such as fast oxide breeders.
Supervisor, Atomic Power Equipment Department, General Electric Co.,
1954 to l'956.
Responsible for the design, safeguard analysis and development of small atomic power plants and test reactors.
Engineer Analyst, Knolls Atomic Power Laboratory, 1953 to 1954.
Worked with steam boilers and superheaters.
Research Engineer, University of California at Berkeley, 1950 to 1953.
Conducted i
boundary layer studies of heat trans:er in high speed flight. D**lD *]D T A
e o J\\\\ o J\\\\ 1 A o
i i
~
Honors and Professional Affiliations:
Member, National Academy of Engineering.
Fellow, American Society of Mechanical Engineers; Past Chairman, ASME Heat Transfer Division.
ASME Heat Transfer Memorial Award (1966).
ASME/AIChE Heat Transfer Conference Award (1963).
Member, Editorial Board, Nuclear Science and Encineerine.
Past Vice-Chairman, Management Committes.
CBI Nuclear, Memphis, Tenn.
Past Member, Editorial Board, International Journal of Heat and Mass Transfer, Heat Transfe:
in Russia, Heat Transfer in Jacan.
Past Member, AEC Task Force on Emergency Core Cooling.
U.S. Delegate, 1964 International Conference on Peaceful Uses of the Atom.
Publications:
Representative publications include:
" Heat Transfer to Constant-Property Laminar Boundary-Layer Flows with Power-Function Free-Stream Velocity and Wall-Temperature Free-S* ream Velocity and Wall-Temperature Variation," Journal of Aeronautical Sciences.
19, No.
5, 341-348 (May 1952).
" Skin Friction and Heat Transfer for Laminar Boundary-Iayer Flow with Variable Properties and Variable Free-Stream Velocity," Journal of Applied Mechanics, 20, No.
3, 415-421 (September 195 3) (with R.A. Seban).
" Local Heat-Transfer Coefficients on Surface of an Elliptical Cylinder, Axis Ratio 1:3, in a High-Speed Air Stream",
Transactions of the ASME, 75, No.
7, 1291-1302 (October 1953) (with oEhers).
"The Effect of Single-Roughness Elements on the Heat Transfer from a 1:3 Elliptical Cylinder," ASME Paper No. 53--A-86, ASME Annual Meeting, November 29 - December 4, 1953 (with others).
" Heat Transfer to Constant-Property Laminar Boundary-Layer Wedge Flows with Stepwise and Arbitrary Wall-Temperature variation,"
Transactions of the ASME, 76, No.
2, 279-286 (February 1954) (with S. Scesa).
" Integral Methods in Natural-Convecrion Flow," Journal of 79 plied Mechanics, 21, No.
4, 515-522 (Decemoer 1955).
" Heat-Conduction Methods in Forced-Conversion Flow," Transactions of the ASME, 78, No. 3, 1627-1636 (November 1956).
" Core and Facilities," Nucleonics, 15, No.
3, 44-47 (March 1957) (with othe?s).
" Generalized Correlation of Boiling Heat Transfer," Journal of Heat Transfer, 91, Series C., No. 1, 37-42 (February 1959T.
" Heat Transfer to Water in Thin Rectangular Channels," Journal of Heat Transfer, 31, Series C.,
No.
2, 129-143 (May 1959) Twith others).
" Hydraulic Instability in a Natural Circulation Loop with Net Steam Generation at 1000 Psia," GEAP-3215, July 15, 1959; and ASME Paper No. 60-HT-27 (with E.
S.
Beckjord).
" Steam Slip--Theoretical Prediction from Momentum Model," Journal of Heat Transfer, 82, Series C, No.
2, 113-124 (May 1960).
2
" Eccentric Rod Burnout at 1000 lbf/in with Net Steam Generation," Int. J.
Heat Mass Transfer, 5, 595-614 (1962) (with otners).
"Importance of High Power Density Boiling Water Reactor Development to Widespread j
Econcmic Nuclear Power," American Power i
Conference, 1962 (with D.
H.
Imhoff).
i D**
3
- b.
ow a
x
" Reliability of Burnout Calculations in Nuclear Reactors," Nuclear News, ANS, February 1963 (with A.P.
Bray).
" Thermal and Hydraulic Performance of Soiling Water Reactors," Papsr No. 89,
" 1962 Nuclear Congress, June 4-7,
- 1962, New York.
" Prediction of Two-Phase Pressure Drop and Density Distribution from Mixing Length Theory," Journal of Heat Transfer, SS, Series C.,
No.
2, 137-152 (May 1963).
" Investigations of Surnout in an Internally Heated Annulus Cooled by Water at 600 to 1450 Psia," ASMZ Paper No. 63-WA-149 (with others).
" Thermal and Hydraulic Performance of Boiling Water Reactors," Nuclear Engineering, 60, Part XI, No. 51, 110-118 (1964).
" Film Boiling of Steam-Water Mixtures in Annular Flow at 800, 1100 and 1400 Psi,"
Journal of Heat Transfer, 36, Series C, No.
1, 81-88 (February 19647 (with others).
" Critical Heat Flux Considerations in the Thermal and Hydraulic Design of Water-cooled Nuclear Reactors," Third International Heat Transfer Conference, Geneva, Switzerland, May 1964 (with others).
" Prediction of Two-Phase critical Flow Rate,"
Journal of Heat Transfer, 37, Series C.,
4 No. 1, 53-58 (February 1965T.
" Critical Heat Flux in Forced Convection Flow," University of California Lecture Series on Boiling and Two-Phase Flow, at Berkeley, April 1965.
" Experience with 3NR Fuel Rods Operating Above Critical Heat Flux," Nucleanics, 23, No.
4, 62-65 and 38 (April 1963) (wirh oEhers)..
oA O'
3"T' i
n D
=h
" Theoretical Predictions of Fully Developed Adiabatic Two-phase Flow,
" University cf Exeter, England, Symposium on Two-phase riow, June 1965.
" Prediction of Two-Phase Annular Flow With Liquid Entrainment," Int. J. Heat Mass
- Transfer, 9, 171-188 (1966).
" Plutonium Utilization in Boiling Water Reactor Power Plants," Commercial Plutonium Fuels Meeting, Washington, D.C.,
March 1-2, 1966 (with others).
" Turbulent Flow in an Annulus," Journal of Heat Transfer, 89, Series C, No.
1, 2!i-31 (February 1967).
"A Systems Approach to Containment Design in Nuclear Power Plants," IAZA Symposium on the Containment and Siting of Nuclear Power Plants, SM-89/51, vienna, Austria, J1ril 3-7, 1967.
"Ferced Convection Subccoled Boiling--Pre-diction of Vapor Volumetric Fraction,"
Int. J.
Heat Mass Transfer, 10, 951-965 (1967).
" Reactor and Fuel: An Integral System,:
Electrical World, 27-28 (April 20, 1970)
(with R.
3.
Ricnards).
" Nuclear Safety: Responding to the Critics (Round Table Panel)," Power Engineerine, Part I, 26-35 (May 1970), Part II, 25-49, (June 1970) (with others).
" Effluent Control for soiling Water Reactors,"
Symposium on Environmental Aspects of Nuclear Power Stations, International Atom _c i
Energy Agency, New York, August 11, 1970 (with others).
" Thermodynamic Developments in Boiling Water Reactors," IVI Nuclear Congress of Rome (Italy)
Developments of Thermodynamics in the Nuclear Field and Their Contribution to Other Uses, i
March 25-26, 1971 (with others).
i h
mm o
f u
_6_
U
l l
l "Large Soiling Water Reactors--Operations Confirm Design," American Power Conference, Chicago, Illinois, April 21-23, 1971 Owith others).
" Utility Involvement: How Much?," Nuclear Industry, Vol. 19, No.
2, February 1972.
" Fuel Reprocessing--A General Electric View," 6th Annual Conference of Japan Atomic Industrial Forum, Tokyo, Japan, March 7-9, 1973 (with others).
" Standardization and Safety Research and Development," AIF Workshop on Reactor Licensing and Safety, New Orleans, Louisiana, January 25-28, 1976.
"A Study of Simulation and Safety Margins in Light Water Reactors," SLI-7904, October 1979, prepared for the President's Commission on the Accident at Three Mile Island Owith J.E. Hench).
Natural Convection Chapter, Licuid Metals _
- Handbook, U.S.A.E.C.,
1955.
Fluid Flow Chapter, Vol.
2, The Technoloev of Nuclear Reactor Safety, edited by J. J. Thompson and J. G. Beckerley, MIT Press.
Wecort of Advisory Task Force on Power Reactor Emercency Cooling, U.S.A.E.C.,
1967.
Proceedines of the 1964 Heat Transfer and Fluid Mechanics Institute, edited by Warren H.
Giedt and S. Levy, Stanford Press. -
.. g-
.v,s
.hl
',b
\\
January 20, 1981.
,y\\.
.. ] > c k'C Q].&.....e *p :ri',Q
\\
N,7vd 'cp W
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
)
METROPOLITAN EDISON COMPANY
)
Docket No. 50-289
)
(Restart)
(Three Mile Island Nuclear
)
Station, Unit No. 1)
)
CERTIFICATE OF SERVICE I hereby certify that copies of " Licensee's Testimony Of Salomon Levy In Response To UCS Contention No. 13 (Alternative Accident Sequences / Class 9 Accidents)" were served upon those persons on the attached Service List by deposit in the United States mail, postage prepaid, this 20th day of January, 1981.
'4M
$ 2d2/i l
Deliss4L.A. Sicqray J' l
l Dated:
January 20, 1981 l
l
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSICN BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
)
METROPOLITAN EDISON COMPANY
)
Docket No. 50-289
)
(Restart)
(Three Mile Island Nuclear
)
Station, Unit No. 1)
)
SERVICE LIST Ivan W.
Smith, Esquire John A. Levin, Esquire Chairman Assistant Counsel Atomic Safety and Licensing Pennsylvania Public Utility Comm'n Board Panel Post Office Box 3265 U.S. Nuclear Regulatory Commission Harrisburg, Pennsylvania 17120 Washington, D.C.
20555 Karin W. Carter, Esquire Dr. Walter H. Jordan Assistant Attorney General Atomic Safety and Licensing 505 Executive House Board Panel Post Office Box 2357 881 West Outer Drive Harrisburg, Pennsylvania 17120 Oak Ridge. Tennesses.
37830 John E. Minnich Dr. Linda W. Little Chairman, Dauphin County Board Atomic Safety and Licensing of Commissioners Board Panel Dauphin County Courthouse 5000 Hermitage Drive Front and Market Streets Raleigh, North Carolina 27612 Harrisburg, Pennsylvania 17101 James R. Tourtellotte, Esquire (4)
Walter W.
Cohen, Esquire Office of the Executive Legal Director Consumer Advocate i
U. S. Nuclear Regulatory Commission Office of Consumer Advocate l
Washington, D.C.
20555 14th Floor, Strawberry Square Harrisburg, Pennsylvania 17127 Docketing and Service Section (3)
Office of the Secretary i
U. S. Nuclear Regulatory Commission Washington, D.C.
20555 l
l l
l l
l
.1
W
. Jordan D. Cunningham, Esquire William S. Jordan, III, Esquire Fox, Farr & Cunningham Harmon & Weiss 2320 North Second Street 1725 Eye Street, N.W., Suite 506 Harrisburg, Pennsylvania 17110 Washington, D.C.
20006 Ms. Louise Bradford Robert Q. Pollard TMI ALERT 609 Montpelier Street 315 Peffer Street Baltimore, Maryland 21218 Harrisburg, Pennsylvania 17102 Chauncey Kepford Ellyn R. Weiss, Esquire Judith H. Johnsrud Harmon & Weiss Environmental Coalition on Nuclear 1725 Eye Street, N.W.,
Suite 506 Power Washington, D.C.
20006 433 Orlando Avenue State College, Pennsylvania 16801 Steven C.
Sholly 304 South Market Street Marvin I. Lewis Mechanicsburg, Pennsylvania 17055 6504 Bradford Terrace Philadelphia, Pennsylvania 19149 Gail Bradford ANGRY Marjorie M. Aamodt 32 South Beaver Street R. D.
5 York, Pennsylvania 17401 Coatesville, Pennsylvania 19320 Attorney General of New Jersey Attention: Thomas J. Germine, Esq.
Deputy Attorney General Division of Law - Roon 316 1100 Raymond Boulevard Newark, New Jersey 07102