ML20058D579
| ML20058D579 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 07/20/1982 |
| From: | Remshaw R, Specter H CONSOLIDATED EDISON CO. OF NEW YORK, INC., POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD) |
| References | |
| ISSUANCES-SP, NUDOCS 8207270216 | |
| Download: ML20058D579 (31) | |
Text
\\
1 C0' F.I " 3 26 S'
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION ATOMIC SAFETY AND LICENSING BO RD Before Administrative Judges:
t".
t.
i Louis J.
Carter, Chairman r
Frederick J.
Shon
... a.:.,..,. / L * " " ' ~
Dr. Oscar H.
Paris
)
In the Matter of
)
)
cot!SOLIDATED EDISON COMPANY OF NEW YORK, INC. )
Docket Nos.
(Indian Point, Unit No. 2)
)
50-247 SP
)
50-286 SP POWER AUTHORITY OF THE STATE OF NEW YORK
)
(Indian Point, Unit No. 3)
)
July 20, 1982
)
LICENSEES' RESPONSES TO NRC STAFF SECOND SECOND OF INTERROGATORIES CONCERNING THE COMMISSION'S QUESTIONS 1,2 AND 5 Preface Pursuant to 10 C.F.R.
S 2.740b (1982), the Consolidated Edison Company of New York, Inc. and the Power Authority of the State of New York, licensees, hereby submit these responses to the Nuclear Regulatory Commission Staf f's second set of interrogatories concerning Commission Questions 1, 2 and 5.
Brent L.
Brandenburg Charles Morgan, Jr.
Paul F.
Colarulli Joseph J.
Levin, Jr.
CONSOLIDATED EDISON COMPANY MORGAN ASSOCIATES, CHARTERED OF NEW YORK, INC.
1899 L Street, N.W.
4 Irving Place Washington, D.C.
20036 New York, New York 10003 (202) 466-7000 (212) 460-4600
/'
b 8207270216 820720 PDR ADOCK 05000247 G
PDR a
O TABLE OF CONTENTS I.
P R E L I M I N A RY MATT E RS.................................. 1 II.
RESPONSES TO INTERROGATORIES.........................
2 i
1 1
I l
i l
l i
l i
I l
i PRELIMINARY MATTERS In several instances the nuclear Regulatory Commission Staff (Staff) has served interrogatories which are, in effect, requests for production of documents.
While 10 C.F.R.
S 740b requires responses to interrogatories within 14 days, 10 C.F.R.
S 2.741 allows 30 days for response to document requests.
Accordingly, the licensees do not intend to produce documents or make objections, except as otherwise specified, at this time.
The licensees are prepared, while reserving any claims of privilege or other objections to such production, to produce documents at a f uture time, in accordance with 10 C.F.R. Part 2 and the convenience of the licensees and the Staff.
s w e RESPONSES TO INTERROGATORIES Staf f Interrogatory No. 1:
Please provide the basis for the application of the definition of damage effective ground acceleration used in the IPPSS to structures such as buried pipe which depend on functional operation as opposed to ductile strength capacity.
Response to Interrogatory No. 1:
Effective peak ground acceleration (EPA) was chosen as a descriptor for the fragility curves developed by Struc-tural Mechanics Associates (SMA) because it was considered to characterize the majority of seismically-induced failure modes in the most appropriate manner.
Some modes of failure, such as those resulting from impact, could be characterized instead by parameters such as relative displacements.
However, such an approach using different parameters would have been difficult to implement in the overall risk assessment because it is difficult to rank the various failure modes as to order of occurrence, and because the seismic hazard curves must be developed in a manner l
l consistent with the fragility curve.
Such an extensive additional effort would not have a significant effect upon results.
For structures and many items of equipment, EPA is considered the most valid descriptor, even for items in which functionality is involved.
For modes of failure involving very little inelastic energy dissipation, very little or no credit is taken for ductility reduction in response.
Such modes of failure may involve elastic buckling, for instance, or a localized failure in a structure in which very little system ductility exists.
The seismic capacity of buried pipe is not considered to depend u~pon functional operation, but rather to be characterized by ductile strength capacity.
For buried pipe, the mode of failure, while ductile, is based upon the maximum strain in the soil (together with anchor point motions, if significant).
The soil strain, however, is dependent upon soil velocity rather than acceleration.
Normally, for large magnitude earthquakes, maximum ground velocities occur in the frequency range below 2 Hz.
It is expected that lower magnitude earthquakes considered for the Indian Point site would have frequency contents primarily in the 2 to 10 Hz range, so that the velocity would be i
i overpredicted using the median broad frequency spectra used for Indian Point keyed to an instrumental peak acceleration (IPA).
The buried piping for Indian Point has a very high i
expected seismic capacity.
The capacity for the buried pipe was estimated by using results for another plant which has softer site conditions and, consequently, higher expected soil strains.
Therefore, the results used in the IPPSS are i
conservative.
t For mechanical equipment with failure modes related to functionality, EPA is also considered to be a reasonable descriptor for many items.
For instance, shaf t binding in valve operators and pumps and fans does not usually occur 4
until elastic limits have been exceeded, and this normally requires several response cycles near resonance to develop malfunction rather than a single acceleration pulse.
Therefore, these mechanical components, which possess some ductility which may be relied upon before failure occurs, may be considered to behave in the same manner as structures for which EPA is considered a valid descriptor, even though the equipment mode of failure is functional.
This principle applies to such equipment which is mounted low in the structure where the equipment is subjected to essentially e
ground input motion, as well as equipment which is mounted high in the structure if the structure remains essentially l
elastic.
In addition, many of these components have I
significantly higher fragility levels than the structures so that for equipment mounted high in the structure, the structure fragility will control.
Again, the structure capacity is considered to be appropriately characterized by EPA.
Some electrical and a few mechanical equipment items are expected to be acceleratior. sensitive.
For instance, such items as electrical relays may trip or otherwise malfunction as the result of a single acceleration pulse.
l
No yielding or other structural damage may occur, so that in such cases IPA may be expected to be a more valid descriptor than EPA.
Most of these modes of failure are recoverable either automatically or by a manual reset, and many have relatively high acceleration capacities compared to other items required for plant safety.
However, most acceleration sensitive devices, such as relays, are mounted in flexible cabinets or racks which are structural elements and which require several cycles of strong motion to develop peak response.
Again, damage ef fective ground motion is considered appropriate.
References:
None.
Staf f Interrogatory No. 2:
Have uncertainties due to design and construction errors been considered in the seismic analysis in Section 7.9.3 of the IPPSS?
Response to Interrogatory No. 2:
Design and construction errors were a?t directly addressed in the IPPSS.
However, the possibility of design and construction errors was, in many cases, considered in establishing the uncertainty associated with a given failure mode.
For example, in the case of primary coolant piping, the possibility of a large throughwall flaw was considered as a lower bound on capacity, and limit moment capacities of other piping were biased below the test data median to account for possible flaws.
It should be recognized that design and construction errors do not always necepsarily result in a decrease in capacity.
Also, the inspection and quality assurance (QA) requirements for nuclear power plants are expected to produce fewer design and construction errors than in typical civil and mechanical construction projects.
i
References:
None.
Staff Interrogatory No. 3:
If the answer to Interrogatory 2 is no, please provide the basis for this lack of consideration.
Response to Interrogatory No. 3:
See Response to Staff Interrogatory No. 2.
Staf f Interrogatory No. 4:
Providc the basis for using SDOF (single-degree-of-freedom) models for MDOF (multi-degree-of-freedom) structures for determining the contribution of inelastic behavior.
Response to Interrogatory No. 4:
Only very limited information on nonlinear seismic response analyses of nuclear power plant structures near collapse is available in the literature, and no such information exists for the specific Indian Point structures under consideration.
Limited available studies tend to indicate that ductility reduction factors for multi-degree-of-freedom (MDOF) systems are normally somewhat less than single-degree-of-freedom (SDOF) systems for the same maximum
s
.'. r a
/
' /
+
- w element ductility, because not all elements of,an MDOF
/
- ff system normally experience the d6ctility.
The ductilityf reduction factor is related to the ove h11 system ductility j
' 'l rather than an element, ductility, and it is recognizgd that #-
s; system ductilities arg typi_cally less than element i
e j
7
/
ductilities unless virtually 100%.of the structure response _' JJ
?
7 occurs in a single mode.
'In the GPSS, systelyductilities
~
j,i appropriate for a.given mode of f ailure for MNF systems i
/
were estimated.
For localtzedi failures or ela's' tic. buckling '
modes of failure, low system ductilities were used.-
For'
{
g f ailure modes in which significant portions of,the/ structure '
are expected to be above yieldi higher system ductilities
+
were estimated.
When multimodce response is expected to be significant at response levels near failure, increased
/
uncertainties were assumed compared to structures with
/
nearly SDOF response.
Ductilities assumed in the IPPSS are i
3. <
'i typically somewhat less than those which are implicit in n
[
standard building codes.
References:
None.
j-e s
Staff Interrogatory No. 5:
~
~
J Provide the basis and the procedbre used for assigning F
1 numerical values to the randomness and uncertai,nty
.I components for fragility curves of critical s'tr'uctures.
~
Provide the basis for assigning any given va iabil ty to one or the other of these components;in Sdctio. n 7.,9.3 of the IPPSS.
f l
-]
/
.)
a
/
4 A
y f_
= - - - -
-I
7 77
< g-kJ Response to Interrogatory No. 5:
Randomness includes variabilities which cannot be substantially reduced based upon the current state-of-the-a art of seismic analysis and material behavior knowledge, i
Included in the values assigned to randomness are y-variabilities in material strength characteristics, earthquake characteristics, and estimates of modal analysis by square root of the sum of the squares (SRSS) with f
absolute sum responses.
These variabilities were based for the most part upon plant-specific data or data from other respresentative nuclear power plants.
Uncertainty includes estimated variabilities from lack of knowledge of a given parameter.
Uncertainty could normally be expected to be reduced, in some cases substantially, by further analysis or testing.
Typical sources of uncertainty include variabilities from modeling
,,, 0 assumptions and use of non-site-specific or generic data.
As an example, the factor of safety for the strength le i
contribution to the overall fragility evaluation of a l
containment wall shear failure considered both the steel and concrete material properties as well as modeling uncertainties.
Plant-specific results of concrete cy;. Mar tests or reinforcing steel tests were not available for the IPPSS.
Based upon data for other nuclear power plants, a 1.
See reference cited at the end of this response.
lognormal standard deviation ( S) for concrete fh, including aging, was determined to be approximately 0.13.
Similarly,
e for reinforcing steel strength was estimated to be 0.09.
Using the second moment approximation from statistics, these values were combined to give a 6 for material properties alone of 0.05.
However, the random variability in the ultimate strength of a concrete shear wall results from other sources in addition to the strengths of the steel and concrete.
Among these sources are the number of cracks and crack patterns, variations in bond splitting and local crushing around individual rebars, in-place versus cylinder concrete strengths, strain rate effects, and cyclic response characteristics.
A lognormal standard deviation of 0.09 was estimated for these ef fects, which was combined with the 0.05 material strength value to give a 6 for randomness for the strength factor of 0.10.
l Variabilities included in the uncertainty lognormal i
standard deviation include contributions from the modeling l
errors expected in the use of the shear wall strength equation, load distributions in the shell at response levels near failure, effective concrete area, interaction between the liner and reinforcing steel, flexure in the shell, and use of non-site-specific material properties for the strength determinations.
The lognormal standard deviation associated with uncertainty was estimated to be 0.20.
l l
Reference:
D.A. Wesley, eti al., " Conditional Probabilities of Seismic-Induced Failures for Structures and Components far Indian Point Generating Station Units 2 and 3," prepared for Pickard, Lowe and Garrick, revised October 1981 (Appendix 7.9.3 of the IPPSS).
Staf f Interrogatory No. 6:
Provide the basis for the assumption contained in Section 7.9.3 of the IPPSS that the collapse of a non-loadbearing masonry wall would essentially be a vertical collapse.
a) please identify which masonry walls in Indian Point Unit 2 [ sic] have been strengthened; b) provide the detailed basis for the development of masonry wall fragili-ties for both strengthened and non-strengthened cases.
Response to Interrogatory No. 6:
A review of observed failures of masonry walls indicates that walls which are restrained at both the top and bottom as in Indian Point do not normally fail as rigid body rotating cantilevers about the base.
Throughwall cracks are expected to form near the top and bottom courses, as well as approximately at midheight, and sometimes at other locations.
These segments of the wall are then expected to rotate as rigid segments until the P-A ef fects result in vertical collapse.
Therefore, most damage is expected within an area limited by approximately one-half the wall height.
The damage mode assumed for the IPPSS was that piping and equipment attached to the wall or
immediately adjacent to the wall would be damaged by wall f ailure, but that equipment separated one-half a wall dimension (typically, of the order of magnitude of an aisle) would have a high probability of survival.
a)
The masonry walls at Indian Point were analyzed in accordance with I&E Bulletin 80-11 dated May 8, 1980, separate from the IPPSS program.
Based upon the masonry wall failure mode, the only walls which were of concern to 1
the IPPSS for Unit 2 were battery rooms 21 and 22.
Design of modifications to the masonry walls of these two rooms has been completed, and their installation is committed for the near future.
For Unit 3, modifications have already been made for the masonry walls enclosing battery charger rooms 31 and 32.
b)
At low levels of seismic response, the non-strengthened masonry walls vibrate essentially elastically in the out-of-plane direction between their supports.
Behavior of the walls in the elastic range can be predicted adequately by closed-form solutions for the appropriate boundary conditions.
Out-of-plane seismic loads were determined using available in-structure response spectra from the original dynamic analysis.
The location of initial cracking and the response level at which this cracking occurs were based upon an estimated masonry modulus of rupture.
Towards the center of the masonry wall panels, cracking is expected to occur primarily in the horizontal direction near midheight of the walls.
Following cracking near midheight, the unstrengthened walls are separated into two sections:
one spanning from near the base to the midheight crack, and the other from the crack near the top support.
Under out-of-plane seismic response, these sections are expected to translate laterally as rigid bodies pivoting about their edges at the top and bottom supports and the crack near midheight.
Behavior of a wall in this condition can be idealized by a SDOF model.
Parameters for this model consist of an equivalent force acting at the intermediate crack and the horizontal translation of the rigid body system at this location.
Resistance to out-of-plane response consists of the restoring force associated with the weight of the wall itself.
The resistance function is a maximum in the initial, undeformed state and decreases linearly under horizontal displacement to zero when P-a effects lead to instability.
Seismic response of the rigid body wall behavior can be predicted using the reserve energy method.
Using this method, an equivalent elastic system having the same energy capacity as the actual nonlinear system was created.
- Thus, the equivalent elastic system has a maximum resistance equal to that of the ictual system and occurs when the ultimate lateral deflection corresponding to P-A instability is
reached.
Using the in-structure response spectra at 10%
estimated median damping generated by the original dynamic analysis, a median strength factor of 2.4 was calculated, i
No ductility is associated with this mode of failure, thus an inelastic energy absorption factor of unity was used.
Other factors and variabilities were computed for the unstrengthened masonry walls as was discussed in Section 7.9.3.
In order to develop representative results for a typical reinforced system, an assumed design based upon discussions with the designer and upon modifications to plants of similar age and seismic design bases was con-sidered.
In this system, vertical members are bolted to existing concrete floors at the top and bottom of the wall and interconnected by a series of horizontal members.
The structural steel grid is supplied on one side of the wall and through-bolted to bearing plates on the opposite side.
The strengthened masonry wall responds rigidly between the steel members and the capacity of the system is controlled by that of the steel framing.
The median factors of safety due to strength and inelastic energy absorption were estimated using previous experience with such a modified system.
The logarithmic standard deviation associated with the uncertainty of the strength factor was increased to 0.25 to account for lack of knowledge of the strength of the system actually to be employed,
The response factor for the masonry walls can be separated into two parts:
one associated with response of the structure, and one associated with the response of the wall (subsystem).
The subsystem response factor is similar to that developed for equipment and consists of variables accounting for spectral shape, damping, modeling, modal combination, and combination of earthquake components.
The subsystem median factors of safety and variabilities account for response characteristics of the wall alone, and were derived similar to the structure median f actors of safety and variabilities described in the report.
The structure response f actor consists of variables accounting for spectral shape, damping, modeling, and soil-structure interaction.
Derivation of the median factors of safety and variabilities is described in the report.
References:
None.
Staff Interrogatory No. 7:
What are the bases for assigning subjective probability for the seismogenic zones mentioned in Section 7.9.1 of the IPPSS?
Resoonse to Interrogatorv No. 7:
The basis for subjective zone probabilities is expert opinion on causes of earthquakes in the eastern United States.
Results of the TERA study, as well as geological expertise, were used in this assessment.
References:
None.
Staff Interrogatory No. 8:
Please provide the basis for the use of truncated exponential distribution used to represent frequency of earthquake occurrence.
Response to Interrogatory No. 8:
The mean number of earthquakes, N(m), of magnitude m or greater on a given source, is expressed using the Gutenberg-Richter relationship:
log N(m) = a - bm (1) or N(m)
= exp ( a - 6 m)
(2) in which a and b = empirical constants l
(1n 10)a a =
e and (ln 10)b a =
e The form of the Gutenberg-Richter relationship leads to an exponential distribution on N(m) [ Equation (2)].
A l
futher modification in Equation (2) becomes necessary when lower and upper bounds on earthquake size are used.
The lower bound, m is the minimum size of engineering a,
significance ( for example, on intensity IV).
The upper bound, mu, represents the size of a maximum credible earthquake that a given source is capable of generating.
A truncated exponential distribution N(m) becomes necessary in order to eliminate the possibility of earthquakes of size less than m or greater than m -
g u
The truncated exponential distribution is as follows:
(~1 1 - exp [-E(m - m Il 1 - exp [-8(m - m )]
o N(m)
A
=
u in which A = N(m ) = exp ( a - B m ).
This type of a o
u distribution has been used in a number of previous studies.1 Virtually all studies of seismicity in the eastern United States use the exponential distribution to describe the relative frequency of dif ferent earthquake sizes.
This distribution must be truncated; otherwise an infinite rate of energy release is implied, which is physically unrealistic.
References:
D.H. Weichert and W.G.
Milne, "On Canadian Methodologies of Probabilistic Seismic Risk Estimation,"
Bulletin of the Seismological Society of America, Vol. 69, pp. 1549-66 (1979);
L. Knopoff and Y. Kagan, " Analysis of the Theory of Extremes as Applied to Earthquake Problems,"
Journal of Geophysical Research, Vol. 82, No. 36, pp. 5647-57 (1977); A.
Der Kiureghian and A. Ang, "A Fault-Rupture Model for Seismic Risk Analysis," Bulletin of the Seismological Society of America, Vol. 67, No. 4, pp. 1163-94 (1977); C.A.
Cornell and H.A.
Merz, " Seismic Risk Analysis of Boston," Journal of the Structural Division, ACSE, 101, ST10, Proc. Paper 11617 (1975).
1.
See references cited at the end of this response.
. Staff Interrogatory No. 9:
In the IPPSS it is concluded that the annual frequency of a pipeline gas fire which threatens the plant is about 5 x 10-7 This value is based upon a large break near the plant.
Provide the bases for assuming that 5 x 10-7 per year is higher than the probability of a small break leading to ingestion of a flammable mixture of natural gas into the plant ventilation intakes.
Resoonse to Interrogatory No. 9:
A calculation was carried out to determine the annual frequency of ingestion of a natural gas flammable mixture into the plant ventilation intakes following a small break in the pipelines in the vicinity of the plant.
The results support the conclusion that the annual frequency of 5 x 10-7 per year based upon a large pipeline break is the upper bound for this scenario.
Flow rates through a small break (1 square inch) in the pipeline were calculated as a function of time, using a blowdown model.
Due to the high pressure and large dimensions of the pipeline, a small break does not cause a major blowdown.
Accordingly, the average flow rate, through the break, used for the dispersion calculations is close to the gas maximum flow rate.
The small area of the break determines a similar average flow rate from either the 26-inch or 30-inch diameter pipelines.
The calculated average 3
flow rates used were 14.75 and 14.6 M /sec., respectively.
l l
A dispersion model as presented by D. Bruce Turner was used.1 Meteorological data for the plant site was given in PASNY FSAR, Vol. lA (NYU Technical Report TR-73-1).
Using the meteorological data, the concentration at the Unit 3 PAB air intake was determined for leaks from various sections of the pipes and was compared with the lower flammability limit.
The annual frequency of wind speeds for each Pasquill stability category blowing in the worst direction from each pipe section was incorporated.
Once the dispersion calculations showed that a leak in a certain pipe section will reach the PAB's ventilation intakes in a flammable concentration, that section of the pipe was considered as a contributor for this scenario.
The annual frequency of a flammable natural gas mixture (taken as 5 percent volumetric concentration) reaching the plant is given by:
P gg =
l_
Ng
- f f
Di f
E f
,P s
t d jk L
i 3
k wsij k i
P g=
Annual frequency of a small pipeline break near the plant, causing an ingestion of a flammable natural gas mixture into plant ventilation intakes.
Total length of transmission pipeline in the L
=
United States (in meters).
Number of transmission line f ailures per year N
=
g in the United States (450).
1.
See reference cited at the end cf this response.
Fraction of failures which are small (12 out f
=
s of 13 = 0.93).
Fraction of failures due to construction-f
=
t related failures and corrosion (0.30).
Length of pipeline sections (meters).
r,
=
Annual frequency of wind speed in a certain f
=
ws direction and Pasquill stability category.
Probability of a flammable natural gas mixture P
=
d reaching the plant ventilation intakes.
Wind speed categories.
k
=
Pasquill stability categories, j
=
Pipeline sections.
i
=
Considering the two pipelines near the plant, the annual frequency of a small pipeline break causing a flammable natural gas mixture at Unit 3 PAB is:
gg =
2
- 450 0.93
- 0.3 3.12 = 1.7 x 10-6 P
280,000 x 1609 where 3.12 is the calculated equivalent length of the pipeline which, if a small leak develops, could result in a flammable mixture at the PAB, over the range of wind conditions:
E D-E E
f Pd j
k wsijk ijkj 1
Because Unit 2 is farther away from the pipelines, P gf calculated for this unit is even lower.
The analysis is considered conservative because topography ef fects, the dif ference in elevation between the ground and the air intake vent, and the associated possible vertical dispersions have been ignored.
On this basis, the large diameter break consequence is considered representative of the gas line hazard.
Reference:
D.
Bruce Turner, Workshop of Atmospheric Dispersion Estimates, U.S. Department of Health, Education, and Welfare, 1968, PHSP No. 999-AP-26.
Staf f Interrogatorv No. 10:
It is stated in Section 7.9.2 of the IPPSS that log-normal distribution is a good mathematical representation to describe observed sustained acceleration.
Can another type of distribution be used, and if not, why?
Resoonse to Interrogatorv No. 10:
A lognormal distribution has been used extensively in seismic risk studies to characterize the uncertainty in the attenuation of ground motion parameters such as acceleration and velocity.1 The logarithm of the ratio of observed to computed peak ground acceleration values was plotted in l
these studies on a normal probability paper and it was shown that, with reasonable accuracy, the variable has a normal d is tribu tion.
This supports the assumption of a lognormal distribution for the observed acceleration values.
A lognormal distribution has certain desirable proper-ties.
For example, the random variable can only assume 1.
See references cited at the end of this response.
___ _ __ ___ nonnegative values, the distribution can be conveniently represented by two parameters--median value and logarithmic standard deviation, and the dispersion in observed values around the median increases as the computed median value increases.
Although other distributions can be assumed for the attenuation of ground motion, they may not have some of these desirable properties.
For example, a normal distribu-tion can assume both negative and positive values and a beta distribution involves four parameters, some of which are d if ficult to compute.
References:
N.C. Donovan, "A Statistical Evaluation of Strong Motion Data Including the February 9, 1971 San j
Fernando Earthquake," Proc. Fifth World Conference on l
Earthquake Engineering, Rome (1973);
L.
Estava, " Seismic Risk and Seismic Design Decision," in Seismic Design for Nuclear Power Plants, MIT Press, Cambridge, Massachusetts, l
and London, England (1971).
Staf f Interrogatory No. 11:
In cases where design analysis results on plant-specific qualification reports were not available for deriving individual equipment fragility levels, these fragility levels were based on generic considerations, a)
Provide the justification for grouping equipment into these major categories for establishing fragility levels, b)
Would grouping by additional subclasses within these categories appreciably change the fragility results.
Provide the bases for your response.
c) Are there individual pieces of equipment that are outliers to their particular category.
Provide the basis for your response.
Response to Interrogatory No. 11:
a)
Grouping of equipment into generic categories is reasonable provided that one of the following criteria is met:
1.
Equipment fails in a structural mode and the seismic design criteria are similar.
2.
Equipment is similar in function and construction to equipment for which fragilities have been established by analysis or test.
3.
Equipment is similar in function and construction to equipment which has experienced major seismic events without failure or with an observed range of failure level.
Piping, vessels, heat exchangers, ducting, cable trays and electrical conduits which are designed to criteria such as ASME and AISC codes have code-defined safety factors and, as a result, can be grouped generically by design criteria.
b)
Further grouping of the generic categories by subclass would not be expected to make a major dif ference in core melt frequency.
Grouping by subclass would narrow the uncertainty bounds and may raise or lower the median fragility values a small degree.
One would not expect to define a very weak generic class of components which is safety-related and has undergone seismic qualification, because a minimum threshold of capacity has been established by the qualification.
c)
To the extent knowledge is available, there are no individual pieces of equipment which are outliers to their particular category.
In cases in which the uncertainty bounds on fragility descriptions have been large, the data base has consisted of fragilities of unqualified components and tends to conservatively bias the fragility description when applied to qualified components.
For instance, the fragility description for electrical and control equipment was derived from fragility data on non-qualified compo nents.
If unqualified components are installed in a safety system, then there is a low probability that the actual fragility level may be in the lower uncertainty bound, and the fragility descriptions define this case.
If the equipment has been qualified or if identical equipment l
has been qualified, the capacity would be even higher than indicated in the IPPSS.
References:
None.
As to Answers:
CONSOL AED EDISO o
NEW YO sK, I
By
/
yichard P.
(
Project Manager - Indian Point Hearings PG4ER AUTHORITY OF THE STATE OF NEW YORK By
' Herschel SpectqPt l
Project Manager - Indian Point Hearings t
VERIFICATION STATE OF NB1 YORK
)
- SS.:
COUNTY OF NEW YORK
)
I, RICHARD P. REMSHAW, being duly sworn, depose and say:
That I am the Manager, Indian Point Hearings for Consolidated Edison Company of New York, Inc., licensee of the Indian Point Nuclear Generating Station, Unit No. 2; that I am authorized to make this verification on behalf of said corporation; and that the foregoing answers to interrogatories were prepared under my direction and supervision and are true and correct to the best of my knowledge, information and belief.
I gICHARD P.
REMSHAW \\
Sworn to before me this
- th day of July,1982
.- /
l l' f 7'
~
w..
.c Notary Public DAVID H. PIKUS Notary Public. State of New York No. 31-4730506 Qualified in New York Ccun*y Commiss.on Exoires March 23,1003 i
VERIFICATION STATE OF N5i YORK
)
- SS.:
COUNTY OF NBi YORK
)
I, HERSCHEL SPECTER, being duly sworn, depose and say:
That I am the Manager, Indian Point 3 Hearings, Technical Support for Power Authority of the State of New York, licensee of the Indian Point 3 Nuclear Power Plant; that I am authorized to make this verification on behalf of said corporation; and that the foregoing answers to interrogatories were prepared under my direction and supervision and are true and correct to the best of my knowledge, information and belief.
- =
HERSCHEL SPECTER Sworn to before me this
- th day of July, 1982 a, 7 w
Notbry Public DAVID H. PtKUS Notary Public State of New York No. 31473S506 Qualified in New York Ceumy Commission boires March 20,1983
tf ly submitt f
/k, d
Brent L.
Brandenourg
/
M Trles Morgan, Jr.
/
Paul F.
Colarulli CONSOLIDATEDEDISONCpMPANY Joseph J.
Levin, Jr.
OF NEW YORK, INC.
f Licensee of India Point MORGAN ASSOCIATES, CHARTERED Unit 2 1899 L Street 4 Irving Place Washington, D. C.
20036 New York, New York 10003 (202) 466-7000 (212) 460-4600 Thomas R.
Frey General Counsel Charles M.
Pratt Assistant General Counsel POWER AUTHORITY OF THE STATE OF NEW YORK Licensee of Indian Point Unit 3 10 Columbus Circle New York, New York 10019 (212) 397-6200 Bernard D.
Fischman Michael Curley Richard Czaja David H.
Pikus SHEA & GOULD 330 Madison Avenue New York, New York 10017 (212) 370-8000 2et, 198 2 Da ted :
July
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION ATOMIC SAFETY AND LICENSING BOARD Before Administrative Judges:
I Louis J. Carter, Chairman Frederick J.
Shon Dr. Oscar H.
Paris
)
In the Matter of:
)
)
CONSOLIDATED EDISON COMPANY OF
)
NEW YORK, INC.
)
Docket Nos. 50-247 SP (Indian Point, Unit No. 2)
)
50-286 SP
)
POWER AUTHORITY OF THE STATE OF
)
)
(Indian Point, Unit No. 3)
)
)
CERTIFICATE OF SERVICE I hereby certify that on the 21st day of July, 1982, I caused a copy of the Licensees' Responses to NRC Staff Second set of Interrogatories Concerning the Commission's Questions 1, 2 and 5 to be served by first-class mail, postage prepaid on the following:
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, i 1
Louis J.
Carter, Esq., Chairman Charles M. Pratt, Esq.
Administrative Judge Thomas R.
Frey, Esq.
Atomic Safety and Licensing Board Power Authority of the 7300 City Line Avenue State of New York Philadelphia, Pennsylvania 19151 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 i
U.S. Nuclear Regulatory Commission Of fice of the Executive Washington, D.C.
20555 Legal Director U.S. Nuclear Regulatory Commission 3
l Dr. Oscar H. Paris Washington, D.C.
20555 Administrative Judge Atomic Safety and Licensing Board Brent L.
Brandenburg, Esq.
U.S. Nuclear Regulatory Commission Assistant General Counsel Washington, D.C.
20555 Consolidated Edison Company of New York, Inc.
4 Docketing and Service Branch 4 Irving Place j
Office of the Secretary New York, New York 10003 U.S.
Nuclear Regulatory Commission l
Washington, D.C.
20555 Ellyn R. Weiss, Esq.
William S. Jordan, III, Esq.
i Joan Holt, Project Director Harmon and Weiss i
Indian Point Project 1725 I Street, N.W.,
Suite 506 l
New York Public Interest Research Washington, D.C.
20006 Group 9 Murray Street Charles A.
Scheiner, Co-Chairperson l
New York, New York 10007 Westchester People's Action Coalition, Inc.
John Gilroy P.O. Box 488 Westchester coordinator White Plains, New York 10602 Indian Point Project New York Public Interest Research Alan Latman, Esq.
Group 44 Sunset Drive t
240 Central Avenue Croton-On-Hudson, New York 10520 White Plains, New York 10606 i
Ezra I.
Bialik, Esq.
Jeffrey M. Blum, Esq.
Steve Leipzig, Esq.
I New York University Law School Environmental Protection Bureau
(
423 Vanderbilt Hall New York State Attorney 40 Washington Square South General's Office New York, New York 10012 Two World Trade Center New York, New York 10047 Charles J.
Maikish, Esq.
Litigation Division Alfred B.
Del Bello The Port Authority of New York Westchester County Executive I
and New Jersey Westchester County One World Trade Center 148 Martine Avenue New York, New York 10048' White Plains, New York 10601 Andrew S.
Roffe, Esq.
New York State Assembly Albany, New York 12248 i
l
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Marc L.
Parris, Esq.
Stanley B. Klimberg, Esq.
Eric Thorsen, Esq.
General Counsel County Attorney New York State Energy Office County of Rockland 2 Rockefeller State Plaza 11 New Hempstead Road Albany, New York 12223 New City, New York 10956 Atomic Safety and Licensing Pat Posner, Spokesperson Board Panel Parents Concerned About Indian U.S. Nuclear Regulatory Commission Point Washington, D.C.
20555 P.O. Box 125 Croton-on-Hudson, New York 10520 Atomic Safety and Licensing Appeal Board Panel Renee Schwartz, Esq.
U.S. Nuclear Regulatory Commission Paul Chessin, Esq.
Washington, D.C.
20555 Laurens R.
Schwartz, Esq.
!!argaret Oppel, Esq.
Honorable Richard L.
Brodsky Botein, Hays, Sklar and Hertzberg Member of the County Legislature 200 Park Avenue Westchester County New York, New York 10166 County Office Building White Plains, New York 10601 lionorable Ruth U.
Messinger tiember of the Council of the Zipporah S.
Fleisher City of New York West Branch Conservation District 44 Association i
City Hall 443 Buena Vista Road New York, New York 10007 New City, New York 10956 Greater New York Council
!!ayor George V.
Begany on Energy Village of Buchanan c/o Dean R.
Corren, Director 236 Tate Avenue New York University Buchanan, New York 10511 26 Stuyvesant Street New York, New York 10003 Judith Kessler, Coordinator Rockland Citizens for Safe Energy Geoffrey Cobb Ryan 300 New Hemstead Road Conservation Committee Chairman New City, New York 10956 Director, New York City Audubon Society David H.
Pikus, Esq.
71 West 23rd Street, Suite 1828 Richard F.
Czaja, Esq.
New York, New York 10010 330 Madison Avenue New York, New York 10017 i
Lorna Salzman
!!id-Atlantic Representative Amanda Potterfield, Esq.
Friends of the Earth, Inc.
P.O.
Box 384 208 West 13th Street Village Station New York, New York 10011 New York, New York 10014 l
l l
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_ - =.
l.
Ruthanne G.
Miller, Esq.
Atomic Safety and Licensing Board Panel U.S.
Nuclear Regulatory Commission Washington, D.C.
20555 Mr. Donald Davidoff Director, Radiological Emergency Preparedness Group Empire State Plaza Tower Building, RM 1750 1
1 Albany, New York 12237
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