ML20054J587

Responses to First Set of Interrogatories Re Questions 1 & 2.Certificate of Svc Encl.Related Correspondence

ML20054J587
Person / Time
Site:	Indian Point
Issue date:	06/25/1982
From:	Brandenburg B, Colarulli P CONSOLIDATED EDISON CO. OF NEW YORK, INC., MORGAN ASSOCIATES, POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:	NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD)
References
NUDOCS 8206290301
Download: ML20054J587 (65)

Text

. t 3 ELATED COINM*G UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION ,

ATOMIC SAFETY AND LICENSING BOARD Before Administrative Judges:

Louis J. Carter, Chairman Frederick J. Shon Dr. Oscar H. Paris

)

In the Matter of )

)

CONSOLIDATED EDISON COMPANY OF NEW YORK, INC. ) Docke t Nos .

(Indian Point, Un i t No . 2 ) ) 50-247 SP

) 50-286 SP POWER AUTHORITY OF THE STATE OF NEW YORK )

(Indian Poir,t, Unit No. 3) ) June 25, 1982

)

LICENSEES' RESPONSES TO NRC STAFF FIRST SET OF INTERROGATORIES CONCERNING QUESTIONS 1 AND 2 Preface Pursuant to 10 C.F.R. S 2.740b, 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 Staff's first set of l

interrogatories concerning Commission Questions 1 and 2.

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Brent L. Brandenburg , Charles Morgan, Jr.

Paul F. Colarulli Joseph J. Levin, Jr.

! CONSOLIDATED EDISON COMPANY MORGAN ASSOCIATES, CHARTERED i

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 l

8206290301 820625 C' (

], J '

l PDR ADOCK 05000247 l 9 PDR 1 --.

e 1 TABLE OF CONTENTS I. PRELIMINARY MATTERS................................. 1 II. RESPONSES TO INTERROGATORIES........................ 2 J

PRELIMINARY MATTERS In several instances the Nuclear Regulatory Commission Staff (Staff) has served interrogatories which are, in e f fect, requests for production of documents. While 10 C.F.R. S 2.740b 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 future time, in accordance with 10 C.F.R. Part 2 and the convenience of the licensees and the Staff.

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2-RESPONSES TO INTERROGATORIES Staff Interrogatory No. 1:

Please identify and describe all significant discrepancies between 1) the current, as-built-and-operated plant design and configuration for power generation and 2) the corresponding assumptions in the Indian Point Probabi-listic Safety Study (IPPSS) treatment of your unit. The scope of your response should include all parts of the IPPSS j leading up to the determination of the damage state vector for both internal and external events. By "significant" we -

mean to include discrepancies which would affect the IPPSS description of the qualitative failure modes or damage state likelihoods. You may exclude, as not "significant,"

modeling approximations that lead to small biases in the damage state likelihoods.

Response to Interrogatory No. 1:

Listed below are the assumptions used in the IPPSS which are different from the current, as-built-and-operated plant design and configuration for power generation. These items have been committed to but are not yet implemented.

1) System modification, procedural change, or verification testing to ensure that sufficient back-pressure will be maintained in the service water system to prevent service water pump overload for cases when only one service water pump is operating with the sys-tem in accident configuration.

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2) Rearrangement of diesel generator fuel oil transfer pump power supplies such that the primary transfer pump for each diesel is powered from one of that diesel's electrical buses. (Indian Point Unit 2 only)

3) Replacement of manual isolation valves with motor-operated isolation valves in certain of the fan cooler service water discharge lines. (Indian Point Unit 2 only)

4) Implementation of masonry wall upgrading modi-fications for station batteries in response to IE Bulletin 80-11.

5) Implementation of plant modifications for mitigation of ATWS.

Staff Interrogatory No. 2:

Please identify and describe all significant dis-crepancies between 1) the current technical specification allowable outage times, surveillance intervals or prevailing practices for maintenance or surveillance and 2) the cor-responding assumptions in the IPPSS which affect the damage state likelihoods.

Response to Interrogatory No. 2:

The versions of the plant technical specifications used in the preparation of the IPPSS are as follows:

o Indian Point Unit 2 through Amendment 68, March 27, 1981

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o Indian Point Unit 3 through Amendment 27, April 27, 1979 IP In addition, surveillance intervals and prevailing practices for maintenance and surveillance which were in effect the first quarter of 1980 were generally utilized in 1

I the IPPSS. An extensive ef fort to identify dif ferences between the current versions of these items and those utilized in the IPPSS has not been undertaken.

Listed below are the assumptions used in the IPPSS which are different from current surveillance practices.

These items have been committed to but are not yet implemented.

1) Implementation of a refueling interval sur-

' veillance test to' verify disc integrity for RHR valves MOV-730 and MOV-731.

2) Implementation of a refueling interval sur-veillance test to verify disc to stem integrity for spray valve 8.69A (such verification is already accom-plished for redundant valve 869B in the course of per-forming normal refueling operations).

3) Implementation of changes to refueling interval surveillance test to clarify test method and data recording for accumulator discharge valve flow verification check.

4) Implementation of procedures to periodically verify that: (a) the service water pumps designated

for essential header service are in fact aligned to the required essential services (i.e., field valve align-ment check), and (b) the position of the control room

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service water system mode selector switch is such that the correct set of service water pumps is selected for automatic essential header service. (Indian Point Unit 3 only)

Sta f f In terroga tory No . 3:

Please identify and describe each instance in which the operators, operational or maintenance personnel are credited in the IPPSS with actions during upset events or accidents which are not described 1) in current emergency procedures, or 2) in current personnel training.

Response to Interrogatory No. 3:

The evaluation of operating and maintenance personnel actions during specific event sequences employed in the IPPSS was based upon procedures and information generally available at the time of the analysis. The Indian Point Units 2 and 3 emergency operating procedures in effect in February 1980 were used as a basis for modeling operator actions. In addition to this specific procedural guidance, general operations and maintenance training programs, experience from similar events at other operating plants, and guidance from draft Westinghouse Owners Group emergency procedure guidelines were used in the evaluation of specific event scenarios.

Any more detailed identification in response to this interrogatory would be objectionable in that it is unduly burdensome and oppressive. Compiling such a list of

" instances" would require a complete review of all fault trees, event trees, and other analyses, as well as a concurrent examination of all current procedures and documented and undocumented content of current personnel training programs. Such a detailed compilation does not exist and its preparation would constitute a new, extensive 1

analysis. l Staff Interrogatory No. 4:

Please discuss the qualitative and quantitative effect on the damage state 1,ikelihood assessments of correcting the discrepancies noted in your response to 1) and 2) above and in crediting operational personnel with doing no more than they are instructed to do.

Response to Interrogatory No. 4:

The licensees object to this interrogatory as being unduly burdensome and oppressive. To respond to this inter-rogatory would require extensive new analyses which could not, in any event, be' performed in the time available.

Staff Interrogatory No. 5:

The risk to early fatalities presented in the IPPSS drops significantly when going from the point estimate of risk to the level two estimate. Please explain in more detail the use of U-factors to arrive at the level two

estimates as summarized in Sections 5, 6 and 8 of the IPPSS. Please also explain the role of accident pro-gressions and phenomenology in determining the U-factors.

Response to Interrogatory No. 5:

The so-called " point" estimate of risk is based upon mean values of release category frequencies and a " nominal" value of the S matrix. This nominal value is considered to be an overestimate by virtue of conservatisms in the source term, the isotopes released, and in the atmospheric modeling.

In the Level Two calculation of risk, uncertainty is included in the form of discrete probability distributions (DPDs) for both release frequencies and the S matrix. In the case of the S matrix, the source term and the site model uncertainties are expressed by using multiplicative factors, UP, on the source strength. To explain exactly what this means,letOfstand for the quantity of isotope i released in release category p . Also, let Q[ represent the nominal value of this quantity, i.e., the value used in computing the nominal value of the site matrix S. Then, to express uncertainty in the quantities of isotopes released, write Q[ = US Of whereQ[is,ofcourse,afixedconstantandUsisaDPD 0 Note that (probability histogram). Thus, 0i is also a DPD.

P U g does not have a subscript i. The same multiplier is used for all isotopes in a given release category.

The discretization chosen allowed the U f actors to have i

the values 0.1, 0.5, 1.0, and 2.0 (i.e., one-tenth, one-half, one, and two times the nominal source strength value). For example, see Figure 5.6-1 in the IPPSS.

This DPD expresses and quantifies the uncertainty about the quantity of isotopes actually released.

The atmospheric or " consequence" model has a great many parameters, all of which have some uncertainty about them.

There is a great deal of experience in doing sensitivity studies for these parameters, but a systematic propagation of parameter uncertainties through the model was not undertaken. Therefore, the aggregate of the uncertainty from all the parameters together with the uncertainties in the model itself is expressed in the form of an " effective source strength," or equivalently in terms of a multiplier on the source strength.

Thus, another multiplier, Uj'* which is dependent not only upon release category p , but also upon damage index x, is used. For example, see page 6. 3-7 in the IPPSS.

Now, writeQ['X to denote the effective quantity of isotope i released in category p for purposes of computing damage index x. Then, incorporating both sources of uncer-tainty, write

_g_

f p, x = U,0, x U g d[

This equation is the one actually used in the calcula-tions except for one additional modification as follows. In

! performing the product of DPDs, UP'X = Ug'X Uj the possible values for U P,x that arise see:

0.01, 0.05, 0.1, 0.2, 0.25, 0.5, 1.0, 2.0, and 4.0.

These values were rounded off to a higher value in the actual calculations. That is, 0.01 and 0.05 became 0.1 and 0.2, and 0. 25 became 0.5. Thus, the probability that should have been assigned to an ef fective source strength of 0.01 was actually assigned (i.e., added to) the probability of a source strength of 0.1. In this way, by making a conserva-

tive approximation, the number of discrete values of U were reduced.

Staff Interrogatory No . 6 :

Your limiting the values of the U-factors to the range 0.1 - 2.0 to represent uncertainties in the health consequences originating from site modeling area, appears to be equivalent to the assumption that the uncertainties in health consequences conditional upon release of a given source term are bounded by f actors of 0.1 and 2.0. Please

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provide basis for this underlying assumption and also provide basis for the probability values in Tables 8.5.8-2 and 8.5.8-3.

Response to Interrogatory No. 6:

The U values used for characterizing dif ferent levels ,

of uncertainty in the dose calculations are based upon experience with atmospheric dispersion and exposure models. S matrix results were computed for each of four U values. After quantitative consideration of the net effect of uncertainties in many aspects of the dose calculations, such as mitigation assumptions, rainfall washout, and release durations, it was determined that there is a small chance that doses are underestimated by a factor greater than 2.0. At the other end of the range, a value of 0.1 was chosen as a reasonable lower value based upon the determin-ation that atmospheric dispersion for many cases would be overestimated by at least that amount by the Gaussian plume model. Because probability values assigned would be very low, i.e., the four U values were suf ficient to characterize the uncertainty, it was not necessary to use a wider range of U values.

The basis for probability values in Tables 8.5.8-2 and 8.5.8-3 are provided on page 6.3-7 of the IPPSS.

Sta f f Interrogatory No. 7 :

In the event of an accident progression resulting in a blowout of core materials from the reactor cavity into the

containment proper, and a settling out on the containment floor:

a) Please describe the range of particle sizes expected from such an event and the basis for the range chosen; b) Please describe the settling characteristics I and final disposition of these core materials within the containment with fan coolers on and with fan coolers off and with the sprays on and with the sprays off; -

b) Please provide the magnitude and composition of particulates that you believe would fail the con-tainment sprays in the recirculation mode; d) Please identify all documents referenced or relied upon in responding to this interrogat' ory.

Response to Interrogatory No. 7:

a) Accident sequences in which the reactor vessel fails while the primary system is under substantial pressure would first release degraded core material into the reactor cavity, and the subsequent blowdown of the primary syst6m would likely eject this material onto the containment floor. If the material is released from the reactor pressure vessel in a solid state, the dynamic process of transporting the material from the reactor cavity to the containment floor would provide little effect upon such debris. For that

material which is molten, the analyses in the IPPSS show that the three major mechanisms of debris removal (hydraulic j ump, wave formation, and entrainment) would all result in accumulation of the debris on the sloped wall of the instrument tunnel. Given this reconstitu-tion of the molten debris into a liquid film, the major mechanism of debris removal would be through a liquid continuous wave transmitted through the instrument tun-nel onto the containment floor. Once the debris flows onto the floor, steam explosions could occur with the limited amounts of debris and water because of the 15 cm of water retained on the floor by the curb around the instrument t,unnel and the manway. The particle size resulting from such interactions could be expected to be within the range of 100 to 500 microns. These interactions with limited amounts of molten debris and water would be the principal means of producing any fine fragments as a result of this ejection.

b) Particulate sizes given in the above paragraph would have terminal velocities ranging from 4 to 9 m/sec. Volumetric discharge rates during the primary system depressurization result in velocities within the lower region of the containment of s 1 m/sec or less.

These are well below the velocities necessary to levi-tate the particle diameters for that limited amount of debris which would experience significant particulation

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due to steam explosions or rapid quenching. Thus, this debris would not be carried into either the upper regions of the containment or the annulus where the fan coolers are located.

Containment sprays in the recirculation mode draw suction from the recirculation sump which is located on the containment floor 90' away from the instrument tun-nel where core debris could exit during the postulated accident scenarios. In addition, this sump is behind a protective wall which allows limited access to the recirculation sumps from other regions on the contain-ment floor. As a result of this wall and the extensive structures supporting the steam generator and reactor coolant pump which also lie between the instrument tunnel and the recirculation sump, the amount of debris which would be transported into the sump would be mini-mal. In fact, the most likely path of debris transport to the sump would be through floor drains which would only transmit the smallest fragments to the sump.

Also, the recirculation sump is baffled to avoid sweep-ing sizable amounts of debris into the recirculation lines.

c) As discussed in the Response to Interrogatory No. 7b), mechanistic evaluations of particulate sizes and their transport within the containment show that the disposition of such materials following a dispersal

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event would generally be on the containment floor.

Major debris configurations would be large particles or slabs which were frozen when the material came into contact with the 15 cm of water retained on the con-tainment floor. As a result, no credible way was found for these particles to be transported into the fan coolers or for significant amounts of debris to either clog the recirculation sump or be swept into the recir-culation lines.

d)

References:

Indian Point Probabilistic Safety Study; L. D. Buxton and W. B. Benedick, " Steam Explo-sion Ef ficiency Studies," NUREG/CR-0947, SAND 79-1399, November 1979. ,

Sta f f Interrogatory No. 8 :

Please explain the derivation of the 0.02% probability for basemat penetration considering that:

a) the reactor cavity will eventually be dry for such damage stages as TE and SE; b) a large fraction of the core material will not be blown out of the reactor cavity (in particular, the 50% remaining in the vessel after initial blowdown);

c) basemat penetration times for non-coolable debris beds are in the range of one (1) day.

Response to Interrogatory No. 8:

a) On page 2.6-24 2 of the IPPSS, the probability for basemat failure assigned for the TE and SE states i

was conservatively estimated as 0.9999 and not 0.0002. No other split fraction for this class was a ssigned (i.e., via a conditional probability). This value could be conservatively assigned because the prior question "R" of late overpressure with 0.9999 probability ended in a late overpressure, which is a much worse release category.

b) and c) For the TE and SE type of plant states, the most likely accident progression is dispersal of the core debris which is generated initially. The dis-persal will drive the core debris out of the reactor cavity and onto the floor area in the containment.

This will be folJowed by a later degradation and dis-position of the bulk of the remaining core into the i

lower reactor cavity region.

The majority of the core debris which is dispersed from the lower reactor cavity will quench initially, and then begin to boil water from the containment floor. While this boiling is occurring, the water on the floor is being depleted due to the boiling-induced e

evaporation and is being replenished due to condensa-tion on structural surfaces in the containment.

Because the evaporation rate exceeds the condensation rate, the containment floor eventually dries out. At this point in the transient, the dispersed core debris will begin to attack concrete and transfer heat

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directly into the containment atmosphere. Calcula tions indicate that due to the large surface area available, the majority of the core debris will transfer heat directly into the atmosphere.

The core debris remaining in the lower reactor cavity will evaporate the wa ter in, that area , and then probably will begin to attack concrete.

The containment pressure and temperature transient will result from the integrated ef fect of these phe-nomena.

In the risk study performed, the containment was conservatively assumed to f ail at about 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> into the accident. For the best estimate accident sequence the lower reactor cavity did not dry out, and thus con-crete attack did not begin until after more than approximately 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> into the accident. Given that only a limited amount of core concrete reaction would occur, it was judged that only a rapid pressure rise (spike) or gradual overpressure would f ail the contain-ment.

The conclusion is, therefore, that for the TE and SE events concrete attack will occur, but will not cause basemat penetration prior to the end of an approximately 12-hour time interval. This is conserva-tive in light of code results which show much slower i

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penetration rates as possibly alluded to in Interroga- -

tory No. 8c).

1 Staff Interrogatory No. 9: l Please state where in Section 4 of the IPPSS a dry-cavity overpressurization analysis is located. l

,1 Please provide a containment temperature his-  !

a) tory plot consistent with your dry-cavity overpres- ,

suriza tion analysis. -

Response to Interroga tory No. 9 :

Accident Class V will eventually have a dry cavity.

This class is discussed in Sections 4.2.6 " Class V Sequence" and 4.3.5 " Class V Sequence." The Class V sequence is essentially the TE plant state in which the accident is typified by a transient event with loss of all offsite and, onsite power. Likewise, Appendix 4.4.9 Class V cases will result in a dry reactor cavity eventually.

a) See Appendix 4.4.9.

Staff Interrogatory No. 10:

Have you assessed the effect on the containment sprays and fan coolers as complete systems for the following

,; environmental conditions:

(1) pressure from gradual overpressurization (including steam conditions)

(2) pressure from significant hydrogen burns (which do not fail the containment)

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(3) temperature from gradual overpressurization events (with both flooded and dry cavity)

(4) temperature from significant hydrogen burns (5) aerosols released from primary system and as a result of core / water / concrete interactions.

Response to Interrogatory No. 10:

(1) No. Because gradual overpressurization f ailure of the containment is not predicted by the analyses when the sprays or fan coolers are operable, there is no logic in assessing such an effect.

(2) Ye s . During an energetic hydrogen burn, the containment pressure can be expected to rise by about 60 psig in about 20 to 60 seconds and then decay to the pre-burn value in a short time. The fan coolers and containment spray systems at Indian Point were designed to withstand differential pressure which could occur during a LOCA pressure transient from 0 to 47 psig in about 10 to 20 seconds. The fan cooler cooling coils.were designed to withstand an external pressure of 70 psig. Because of conservatism in the design analysis, the fan coolers are capable of withstanding a significantly more severe pressure transient than the design basis transient, and are capable of withstanding at least the containment failure pressure of 126 psig. In the case of the spray system, it is designed to withstand LOCA pressures and, basically being a system of

piping and pumps, would be expected to have a very high 44

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4 tolerance for any expected pressure surges in the con-tainment. Hence, the fan coolers and spray systems are expected to survive the pressure transient during a realistic hydrogen burn scenario.

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(3) No. Because gradual overpressurization f ailure (and associated temperature increases) of the containment are not predicted by the analyses when the sprays or fans are operable, there is no logic in assessing such an effect.

(4) Ye s . The duration of a thermal transient from a sig nificant hydrogen burn would be short. Components of fan cooler and spray systems are relatively large, rugged in construction, and hence should have substantial thermal inertia. During equipment survivability tests conducted at Fenwal, I:.c. and Acurex, Corp., the peak surface temperatures measured for equipment with relatively low thermal inertia (such as pressure transmitter and limit switch) exposed to hydrogen burns were generally under 300*F. As the components of the fan cooler and spray systems have substantially greater thermal inertia than the small equipment tested, the peak temperature of the former

- s O during hydrogen burns can be expected to be below 300 F.

All components of the fan cooler and spray syst;ms were designed to withstand, without impairing operability, a post-accident containment temperature of 271 F for 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. Because of conservatism in the design analysis, the fan cooler and spray systems can be expected to operate at l

temperatures substantially above 271 F for a few minutes. ,

The peak component temperature anticipated during hydrogen burns (about 300 F) is only slightly higher than the design basis (271 F), and its duration is subclantially shorter i

than that assumed for the design basis transient (3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />). Hence, a short duration thermal transient from a hydrogen burn should not have any adverse effects upon containment heat removal capability or fan operation.

(5) Yes. Aerosols would not significantly degrade the See also spray and fan cooler functional capabilities.

Responses to Interrogatory Nos. lla) and 12a).

Staff Interrogatory No. 11:

Have you assessed,the following specific effects on the ,

containment sprays and fan coolers as complete systems under the environmental conditions listed in Interrogatory 11

[ sic]:

a) aerosols on filter efficiencies, crud buildup on cooling coils, fan cooler efficiency; b) local hydrogen burns within the f an coolers on fan cooler performance; c) hydrogen burns and high ambient temperatures including on electrical subsystems within containment, any switch gear.

Response to Interrogatory No. 11:

a) An assessment of the ef fect of aerosols upon filter efficiences, crud buildup on cooling coils, and fan cooler

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efficiency has been performed. The effect is expected to be minimal, considering the facts that:

1) The HEPA filter efficiency is at least 99.97%

during accident operation, and the accumulation of aerosols on the filters could only increase their removal efficiency. The filters are not expected to fail by aerosol loadings which could exist inside the f an coolers under degraded core conditions.

2) Crud would not build up on cooling coils because of condensed water which constantly washes out the aerosols trapped by the cooling coils. The heat removal capacity of the cooling coils will not be affected by release of aerosols into the containment.

b) See Responses to Interrogatory Nos. 10(2) and 10(4). Further analyses would be unduly burdensome and oppressive in the time available, c) Several experimental and analytical studies have been made by the industry to assess the effects of hydrogen burns upon electrical equipment. These studies have shown that electrical equipment can perform its intended function during and after exposure to hydrogen burns. Based upon the data available to date, electrical equipment in Indian Point 1

is expected to survive the environmental conditions created

! by hydrogen burns.

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Staff Interrogatory No. 12:

If the assessments discussed in Interrogatories 11 and 12 (sic] have been performed:

a) describe in detail the results of such assessments, b) provide any such written assessments.

Response to Interrogatory No. 12:

a) In reference to Interrogatory Nos. 10(5) and lla),

detailed analyses on operability of the fan coolers under degraded core accident conditions have been performed. The conclusion is that the fan coolers would not fail due to the presence of aerosols. The detailed analysis is given below.

Aerosols could be generated during core degradation in-vessel or from core / water / concrete interaction in the ,

reactor cavity. Gap and melt releases are largely respon-sible for release of fission products and core materials from the primary system while the vaporization release during core melt-water interaction and subsequent concrete attack (if it occurs) is responsible for ex-vessel aerosol generation and release into the containment.

The aerosols generated in-vessel, which consist of fis-sion products and other core materials, would be effectively removed in the pathways, i.e., core and internals struc-tures, and piping, to the containment by natural deposition, either laminar or turbulent, and gravitational settling.

Because the steam content in the reactor vessel is high, fission product removal by steam condensation is also sig-

nificant. The retention of particulates in the primary system depends upon the thermohydraulic conditions involving a core melt. For accident sequences with low steam flow rates in the primary system, such as transient or small break events, a small fraction of the total aerosol mass from the core would be released from the primary system.

Given the long and winding paths, and low steaming rates for 1

most of the accident sequences following core melting, it is l expected that a very small fraction of the aerosol inventory in-vessel will escape into the containment.

Of the aerosols which do escape into the containment, aerosol plateout, agglomeration, and gravitational settling will remove the aerosols from the atmosphere quickly and prevent them from reaching fan coolers and containment spray headers near the dome. In addition, if the containment sprays are operating, considerable airborne aerosols would be washed out and the moisture separators of the fan cooler units would remove significant quantities of particulates prior to reaching the HEPA filters in those units.

The quantity of aerosols generated during core degrada-tion is primarily from core melt. As stated in the IPPSS, in the case in which water is available to form a coolable debris bed following vessel failure, very little aerosol would be released from the core debris because the core melt is quickly quenched. If the debris bed is non-coolable, the basic plant geometry ensures that the core debris will be

water-covered when fan coolers are operating. This water cover will remove most of the aerosols prior to their release into the containment. Therefore, whether a coolable or non-coolable debris bed is formed, aerosols generated from core-concrete interaction are not an issue.

The f an cooler units are located on the intermediate floor (elevation 68') between the containment wall and the primary compartment shield walls. The fan coolers take suc-tion from the immediate surroundings and discharge to the .

distribution duct work. To reach the fan cooler suction, airborne aerosols must take a tortuous path from the primary system to the lower containment outside the shield wall. It is expected that a large fraction of the aerosols will settle out in the transport path and will not be able to reach the fan cooler suction. The above arguments which support our view of an insignificant effect of aerosols upon the f an coolers are summarized as follows:

1) Aerosol generation would be limited to those generated in-vessel,

2) Aerosols would be mostly retained in the transport pathways, such as primary systems piping and structures, water in the reactor cavity and the containment sump, e tc .

3) The fan cooler location, associated ducting, and moisture separators will preclude aerosols from reaching the HEPA filters in a significant amount.

It can also be shown that the HEPA filters can absorb significant quantities of aerosols without significantly affecting the fan cooler performance. In Section 3. 2. 2. 6 of the NUREG-0850 report, a maximum of 3080 lbs. of aerosols was extrapolated from the highest available experimental results. It should be noted that this estimate is much higher than that given in the NUREG-0772 report. As dis-cussed above, a very small fraction of aerosols would reach the HEPA filters. For this analysis, it is conservatively assumed that 25% of the 3080 lbs. of aerosols would reach the HEPA filters. This amount of aerosols would cause only 3.1" and 2.2" w.g . pressure drops if three and five f an coolers were assumed to be operating, respectively.

As far as the effect of aerosols upon spray operation is concerned, the conclusion reached ir that there would be no significant adverse effects, based upon the fact that aerosols by definition would not be present inside the spray train unless the system is void of water and, therefore, not in operation. Thus, the only concern would be any impact aerosols might have upon the exposed exterior surfaces of the spray system. The impact would be minimal, considering the facts that:

1) When the sprays are in operation, the sprays will wash off exterior surfaces of the system and reduce the aerosol concentration significantly. The spray nozzles would not be affected by airborne

aerosols because they would be washed out before they reach the nozzles.

2) The spray system is designed for the harsh LOCA environment and the mechanical portions of the system consist primarily of steel piping and components which would not be adversely impacted by the aerosols.

3) The electrical systems, because of their LOCA design, should be able to withstand the aerosol-generated radioactive environment.

b) See Preliminary Matters.

Staff Interrogatory No. 13:

Please list all containment penetrations (include location, size, accessibility) for both Units 2 and 3 which are not being used and which could be used for additional piping and/or electrical penetrations.

Response to Interrogatory No. 13:

Included herein is a listing of spare containment penetrations for Indian Point "Jnits 2 and 3. For clarity, the response is provided in two parts: Part I, Indian Point Unit 2, and Part II, Indian Point Unit 3.

This data is based upon the latest available drawing information or from physical survey data as identified.

When some portion of a penetration has been used or is reserved for a pending modification, the remaining spare

pipes or cables have been identified by size and quantity.

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PART I: Indian Point Unit 2 Spare Penetrations: l a) Piping ILCATION(5) SIZE ACCESSIBILITY III MM 38', 56'-6" 18" OD Iccessible tbte 2 56',56'-6" 18" OD Accessible ?bte 2 b) Electrical PEtETRATION IOCATION(5) ACCESSIBILITY III CCM4E!7T H-32 Col . 12-13 El . 54 '-0" Accessible 1bte 3 - Spare H-46 Col. 12-13 El . 51'-0" Accessible !bte 3 - Spare H-62 Col . 13 El . 48 '-0" Meessible Ibte 3 - Spare H-20 Col. 12-13 El. 57' Accessible tbte 4 Partial 7 Triax H-21 Col . 12-13 El . 57 ' Accessible tbte 4 Partial 2 tb.16 SW H-22 Col. 13 El. 57' Accessible tbte 4 Partial 3 tb.16 SW H-23 Cbl . 13 El . 57 ' Accessible tbte 4 Partial 7 Triax H-24 Col . 13 El . 57 ' Accessible tbte 4 Partial 3 tb.16 SW H-25 Col . 13-14 El . 57 ' Accessible tbte 4 Partial 4 Triax H-26 Col. 13-14 El. 57' Accessible !bte 4 Partial 3 tb.16 SW H-27 Col . 13-14 El . 57 ' Accessible tbte 4 Partial 6 Triax H-28 Col. 13-14 El. 57' Accessible 1bte 4 Partial 6 Triax H-29 Cbl . 13-14 El . 57 ' Accessible tbte 4 Partial 1 tb.16 SW H-30 Col . 13-14 El . 57 ' Accessible tbte 4 Partial 5 Triax H-31 Col . 13-14 El . 57 ' Accessible tbte 4 Partial 3 tb.12 H-35 Col. 13 El. 54' Accessible ?bte 4 Partial 4 Triax H-37 (bl . 13 El . 54 ' Accessible Ibte 4 Partial 2 tb.12

H-38 Col . 13-14 El . 54 ' Accessible tbte 4 Partial 2 tb.12 H-40 Col. 13-14 El. 54' Accessible tbte 4 Partial 3 Triax H-41 Col . 13-14 El . 54 ' Accessible tbte 4 Partial 1 tb.16 STP H-42 Col. 13-14 El. 54' Accessible tbte 4 Partial 1 tb.12 H-43 Col . 13-14 El . 54 ' Iccessible tbte 4 Partial 11 tb.12 H-44 Col . 13-14 El . 54 ' Accessible tbte 4 Partial 2 tbte 12 H-50 Col . 13 El . 51' Accessible tbte 4 Partial 8 tb.12 c) Notes

1) Accessible means physical access to both sides of the penetration without consideration to radiation levels.

2) Wese penetrations are spare mechanical penetrations.

3) Electrical penetrations are 10" nan. diameter able to accept any standard low voltage electrical penetration of the several types in use at Indian Ibint thit 2. ,

4) Electrical penetrations are 10" nan. dianeter. Partial spare pene-trations indicate the nunber and size of cable available under the connents column. The following abbreviations are used:

16 STQ - #16 SW shielded twisted Quad Instrument Cable 16 STP - #16 AW shielded twisted Pair Instrument Cable TRIAX - Triaxial Cable 12 - #12 AW Ibwer Cable 8 - #8 AW Power Cable 4 - #4 AW Ibwer Cable 4/0 - #4/0 AW Power Cable 2 - #2 AW Ibwer Cable 35011CM - 350 FCI Power Cable

5) Iocation is given as degrees fran South centerline measured counterclockwise for piping penetrations and as proximity to containnent annular steel columns for electrical penetra-tions. Elevations are in feet above river level. O'-0".

l

PART II: Indian Point Unit 3 Spare Penetrations:

a) Piping IDCATION I7) SIZE ACCESSIBILITY I1I CCN!E!E 53', 56'-6" 18" OD Accessible ?bte 2 53', 59'-6" 18" OD Accessible 56', 56'-6" "hw" 18" OD Accessible ?bte 2 & 3 59*, 64'-0" 18" OD Accessible 62 , 56'-6" "ZZ" 18" OD Accessible tbte 3 62', 59 '-6" 18" OD Accessible

• 62', 64'-0" 18" OD Accessible 18'-3" E OF SE, 39'-6" "0 0 " 22" OD Accessible tbte 6 11 b) Electrical PEhTTRATION LOCATION (7) ACCESSIBILITY (1) ,

CCMMEtE H-19 Col . 12-13 El . 57 ' Accessible tbte 4 Partial 10 tb.16 STQ H-20 Col. 12-13 El. 57' Accessible !bte 4 Partial 7 Triax H-21 Col . 12-13 El . 57 ' Mcessible tbte 4 Partial 1 tb.16 SW H-22 Col. 13 El. 57' Accessible !bte 4 Partial 7 Triax H-23 Cbl. 13 El. 57' Accessible tbte 4 Partial 3 tb.16 SM H-24 Col. 13 El.,57' Accessible tbte 4 Partial 8 tb.16 SW H-25 (bl . 13-14 El . 57 ' Accessible tbte 4 Partial 6 tb.16 SM H-26 Col. 13-14 El. 57' Accessible tbte 4 Partial 3 Triax H-27 Cbl . 13-14 El . 57 ' Accessible tbte 4 Partial 116.16 STP l

H-28 Cbl. 13-14 El. 57' Accessible tbte 4 Partial 2 tb.16 S'IQ H-31 Col . 13-14 El . 57 ' Accessible tbte 4 Partial 18 tb.12

l l H-33 Col . 12-13 El . 54 ' Accessible tbte 4 Partial 22 tb.16 STP H-35 Col . 13 El . 54 ' Accessible tbte 4 Partial 17 Bb.16 STP

!!-36 Col . 13 El . 54 ' pccessible tbte 4 Partial 148 tb.12 H-38 Col. 13-14 El. 54' Accessible tbte 5 H-39 Col. 13-14 El. 54' pccessible tbte 5 H-42 Col . 13-14 El . 54 ' Accessible tbte 4 Partial 3 tb.16 STP H-43 Col . 13-14 El . 54 ' Accessible tbte 4 Partial 18 tb.12 H-44 Col. 13-14 El. 54' Accessible tbte 4 Partial 18 tb.12 H-46 Col . 12-13 El . 51' Accessible tbte 4 Partial 17 tb.12 H-47 Col. 12-13 El. 51' Accessible tbte 4 Partial 3 tb.12 H-48 Col . 13 El . 51' pccessible tbte 4 Partial 8 tb.12 H-49 Col . 13 El . 51' Accessible tbte 4 Partial 10 tb.12 H-51 Col . 13-14 El . 51' pccessible tbte 4 Partial 71 tb.12

!!-52 Col.13-14 El . 51' Accessible tbte 4 Partial 119 tb.12 H-54 Col. 13-14 El. 51' Accessible tbte 4 Partial 2 Triax H-55 Col.13-14 El . 51' Accessible tbte 4 Partial 11 tb.12 H-56 (bl . 13-14 El . 51' Accessible tbte 4 Partial 10 lb.12 H-58 Col. 12-13 El. 48' Accessible tbte 4 Partial 60 tb. 8 H-59 Col. 12-13 El. 48' Accessible tbte 4 Partial 80 tb. 8 H-60 Col . 12-13 El . 48 ' Accessible tbte 5 H-61 Col . 13 El . '48 ' pccessible tbte 5 H-62 Col. 13 El. 48' Accessible tbte 5 H-63 Col. 13 El. 48' Accessible tbte 5 H-64 Col. 13-14 El. 48' Accessible tbte 5 H-65 Col . 13-14 El . 48 ' Accessible tbte 5 H-66 Col. 13-14 El. 48' Accessible tbte 5

H-67 Col . 13-14 El . 48 ' Accessible tbte 4 Partial 2 Triax H-68 Col . 13-14 El . 48 ' Accessible tbte 5 H-70 Col . 13-14 El . 48 ' Accessible tbte 4 Partial 25 lb. 2 H-71 Cbl. 12 El . 48 ' Accessible tbte 4 Partial 15 lb. 2 H-72 Col . 12 El . 48 ' Accessible tbte 4 Partial 15 tb. 2 H-73 Col. 12 El. 48' Accessible !bte 4 Partial 17 tb. 2 H-74 Col . 12 El . 48 ' Accessible tbte 4 Partial 16 lb. 8 c) tbtes

1) Accessibility means physical access to both sides of the penetration without consideration to radiation levels.

2) Tb be used for future "PLOCAP".

3) Partial spare penetration consisting of one 3" and four 2" spare lines each. These penetrations "WW" and "ZZ" will becane partial spare penetrations only after the scheduled addition of the

• "RVLIS" modification.

4) Electrical penetrations are 10" ncm. diameter. Partial spare penetrations indicate the ntaber and size of cable available under the couments column. The followirg abbreviations are used:

16 S'IQ - #16 AWG 91ielded twisted Quad Instrtment Cable 16 STP - #16 AWG Shielded twisted Pair Instrtment Cable TRIAX - Triaxial Cable 12 - #12 AWG Ibwer Cable 8 - #8 A143 Power Cable 2 - #2 AWG Ibwer Cable

5) Electrical penetrations are 10 non. dianeter able to accept any standard low voltage electrical penetration of the several types in use at Indian Point Unit 3.

6) Spare contairinent step penetration.

7) location is given as degrees fran South centerline mea-sured counter clockwise for piping penetrations and as proxi:rdty to contairment annular steel columns for electrical penetrations.

Elevations are in feet above river level. O'-0".

Staff Interrogatory No. 14:

It is stated in Section 0.11.3 of the IPPSS that the uncertainties in the S-matrix (the site matrix) that may originate from the uncertainties in the source terms and those in the site modeling and site variables (such as dispersion, deposition, evacuation model and speed, shielding factors, dose models, etc.) were simulated by repeating calculations for three assumed source term magnitudes (resulting from U-f actors of 2, 0. 5, and 0.1) and then assigning probabilities to the 4 sets (including the one with U-factor 1.0 for source terms) of results. Please provide basis for the assumption that uncertainties origin-ating from site modeling and site variables can be scoped by uncertainties in source terms alone. Please also provide basis for the assigned numerical values of probabilities to the individual bins in histograms in -Figure 5.6-1.

Response to Interrogatory No. 14:

The uncertainty in the S matrix (expressed as the four elements included in the matrix and the probabilities assigned to each element) does not reflect uncertainty in source terms as they 'are presented in Section 6. The variation in the source term by the factor U was only used as a convenient way of simulating uncertainties in health ef fects resulting from the net ef fect of uncertainties in a combination of factors which affect dose. Some other variable could have been chosen. For example, dose (which

is directly proportional to source term) could have been ratioed by the U factors with the same effect. The S matrix elements were used independently in Section 5 to express un-certainty in health effects associated with uncertainties in source term.

The assigned values represent an estimate of the like-lihood that the source term values calculated with the Reactor Safety Study (RSS) methodology or some fraction of the RSS value might be expected to be released to the environment. The basic RSS methodology for calculating source term release to the environment was considered to be the following:

a) It is assumed that fission products released from the core are transported through the reactor coolant system to containment without attenuation.

b) Fission product depletion from the containment a tmosphere is calculated by applying the CORRAL Code.

c) It is assumed that there is no source term attenuation along the leakpath from the containment to the environment.

J For each of the histograms, a particular accident sequence was considered in developing the histogram for that release category. The a.cident sequence selected was the most probable sequence for that release category. The appropriate sequence is indicated in the discussion in Sec-tions 5.6.2.1 through 5.6.2.5. Probability distribution

histograms were only developed for those release categories which were significant contributors to risk. The bins selected for the histograms were a) the source value calcu-lated with the RSS methodology, b) one-half of the source value calculated with the RSS methodology, and c) one-tenth of the source value calculated with the RSS methodology.

A discussion of the basis for the probability values assigned to the bins follows:

1) Release Category 2 (see Section 5.6.2.1)

The V sequence is the dominant sequence considered for this case. For this sequence some retention of fission products during transport through the reactor coolant system and RHR system to the. break is expected. Add itionally , the auxiliary building is likely to experience local rather than gross failure as a result of the V sequence blowdown so that some holdup and deposition of fission products in the auxil-i iary building is likely. Overall reduction of the calcu-lated source term by at least a factor of 1/2 seemed to be j the most likely situation. A probability of .6 was assigned to this case. Values near the calculated source term might occur if gross failure of the auxiliary building were to occur, so a value of 0.25 was assigned. Reduction by as much as a factor of 10 is less likely and would only occur if the auxiliary building were to be relatively leak tight at the time of fission product release. Hence, a value of 0.15 wa s assig ned .

2) Release Ca tegory Z-1 ( see Section 5. 6. 2. 2. )

The dominant sequence for this category is containment failure prior to core melt as a result of an external event such as a seismic event. The basis for the assignment to the bins is discussed in Section 5.6.2.2. The degree of retention in the reactor coolant system is influenced by whether a hot leg or cold leg break occurs as a result of the seismic event, with cold leg breaks both more likely to occur and more likely to produce greater attenuation.

Hence, the f actor of 1/2 bin was assigned slightly higher probability than the factor of 1 bin. Reduction by as much as a factor of 10 could occur if a water lute existed in the flow path between the core and reactor coolant system break. Such a condition is of lower probability, and hence the 1/10 factor was assigned a probability of 0.1.

3) Release Ca tegory 2RW ( see Section 5.6. 2. 3)

The dominant accident sequence considered in estimating a source probability distribution for release category 2RW was a core melt sequence initiated by a transient with loss of all AC power and with loss of secondary heat sink. Loss of power continues for the long-term, and eventually a delayed containment failure occurs from pressure buildup in containment. Containment failure occurs 10 to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> af ter accident initiation. Containment sprays are not func-tional for this sequence. Such a sequence generally cor-responds to the TMLB' sequence of the RSS.

For such a sequence it is likely that the melt release component will flow through the hot leg to the pressurizer out the safety valve and to the quench tank for the period between core melt release and reactor vessel melt through.

Relatively complete retention, of radioiodine and particu-late material so transported, by the water in the pres-surizer and/or quench tank is expected. However, a signifi-cant fraction of the radionuclide inventory can still be in the vessel at vessel failure and vent out the melt hole. In addition, greater particulate deposition than that predicted by CORRAL is generally regarded as likely in the period of at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> between reactor vessel failure and contain-ment failure. Hence,'the largest probability was assigned to the 1/2 factor (0.55) based upon the expected retention of some of the melt release in liquid water. A probability of 0.15 was assigned to reduction by a f actor of 10, and it is now recognized that perhaps this probability value is too small. The value of 0.30 that a factor of 1 could apply to the source term would assume no reactor system deposition, now regarded as unlikply. Hence, the value assigned to this bin is probably too large.

4) Release Category 8B (see Section 5.6.2.4)

The dominant accident sequence considered in developing a source term probability distribution for release category 8B was a core melt sequence initiated by a small pipe break with failure of either ECC injection or ECC recirculation;

containment sprays are assumed to be functional. A signifi-cant fraction of the fission products released from the core debris are likely to be transported to the containment eventually, but most will be washed out by the sprays. Some retention in the reactor coolant system not accounted for in the calculation is also likely. CORRAL representation of the containment processes is regarded as reasonably good, although greater reduction of radioiodine than accounted for in CORRAL is likely. A value of 0.5 was selected for the factor of 1 bin, and 0.4 for the factor of 1/2 bin, to account for the situation in which reactor system retention is effective. For some small breaks or transient sequences when water lutes exish along the transport path, reduction by f actors as great as 10 can occur; a factor of 0.1 was assigned to this bin.

Staf f Interrogatory No . 15:

Please describe your assumptions regarding the effects of earthquakes on the S-matrix in Step 5 of Sections 0.17.1 and 0.17.6 of the IPPSS. Please provide the basis for these assumptions. ,

Response to Interrogatory No. 15:

In the IPPSS, the same site matrix S was used for seismic events as for internal events. That is, the analysis did not include any modifying ef fect of an earthquake upon the consequences of a radioactive release.

1 Staff Interrogatory No. 16:

Please provide (a) the basis for the assumption of 1 hr. as the base-value of evacuation delay time for all populations within 10 mi. and (b) basis for numerical values of the modifiers to adjust the delay time for adverse weather and schools-in-session scenarios, and the assigned probabilities for these modifiers (pages 6.2-9 and 6.2-10 of the IPPSS).

Response to Interrogatory No. 16:

Evacuee delay assumptions are discussed in detail on pages 6.2-9 and 6.2-10 of the IPPSS. Delay was treated probabilistically. The delay treatment for almost all of the EPZ population is, summarized in the following table: ,

Percent of Population Total Delay (hours) 90% of all 7% of all 3% of all scenarios scenarios scenarios 5 0.5 1.5 2.5 15 0.7 1.7 2.7 50 1.0 2.0 3.0 15 1.5 2.5 3.5 15 2.0 3.0 4.0 As explained on page 6.2-9, a small component of the population was assumed to be systematically delayed for much longer times for weekday-school-in-session scenarios because such delays were anticipated in the emergency plan at the time the analysis was performed.

The delay time distribution selected reflects an adjustment of delay time data derived from observed evacua-tions associated with transportation accidents. These data s uggest that the entire evacuee population would be delayed 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in 30 percent of all scenarios, 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> in 40 percent 4

of all scenarios, and 5 huurs in 30 percent of all scenarios (SAND 78-009 2, " A Model of Public Evacuation for Atmospheric Radiological Releases," Aldrich, D., e t al . , June 19 78 ) .

Delay times were shortened and probabilities for long delay times were reduced to reflect dif ferences between a severe accident at a nuclear power plant and the transportation accidents examined:

1) Early aJert of officials (in accordance with emergency action level guidance) soon after an event which initiates a nuclear plant accident would provide some time for partial mobilization before the need for emergency protective action develops. This margin could be as long as 1Y2 hours for release category 2, but would be small for release category Z-10

2) Pre-planning and drill for nuclear plant accidents would reduce the time required to diagnose and to decide to recommend protective action, par-ticularly in a severe accident. (Transportation accidents can happen in a wide variety of locations and

, can present a broad spectrum of hazards. These facts undoubtedly contribute to the time required for evalua-

tion and decisionmaking. Evaluation time we.s found to be a substantial component of the total delay for transportation accidents.)

3) Communication and alert systems for use in a nuclear plant accident would reduce the time required to communicate emergency response instructions to the public.

Staf f Interrogatory No. 17:

In order to justifiably take credit for shielding protection from the plume exposures in the regions beyond the 10-mi emergency planning zone (EPZ) that could be provided by sheltering (i.e., remaining indoors) it may be necessary to demonstrate with a degree of confidence that sheltering would indeed be the emergency response mode beyond the 10-mi EPZ of the Indian Point site. This would require that in a real radiological emergency situation at the Indian Point site, sheltering of the people beyond the 10-mi region be properly directed, i.e., people should be notified of the emergency and be advised to remain indoors until further notification. However, after start of the atmospheric release from the reactor, there would be delays before the people would actually be in shelters. Delays would be due to time to be taken by authorities to decide to notify the people to shelter, time to be taken by authorities to actually notify the people to shelter and time taken by people to move indoors. During the delay

periods the shielding factors only for the situation of normal activities of people as assumed elsewhere in your study and in WASH-1400 may be appropriate. Further, for not diminishing the benefit from the improved shielding factors (given the sheltering mode) it may also be necessary to advise the people to open windows and enhance ventilation to expel contaminated air trapped inside the buildings for exchange with the outside fresh air after the radioactive plume has lef t the area. Unless d 2.s latter action were taken, the dose from prolonged inhalation of the contamin-ated air trapped in the buildings would likely result in higher doses from plume inhalation exposure pathway (see WASH-1400, Appendix VI page 11-8 and Figure VI 11-5) .

a) Please explain the basis for your assumption in Section 6.2 at page 6.2-10 of the IPPSS of shielding protection f actors of 0.5 for cloud dose and 0.08 for ground dose for 90% of the population within a 10-50 i mile zone.

b) Please provide values of shielding protection factors and peoples' [ sic] transit times used in calcu-lations of ground dose during relocation af ter remain-ing in the sheltered mode for the assumed 24-hour period, c) Please provide the basis for and the value of the reduction factor, due to ventilation, for inhala-i

tion dose which is mentioned in Section 3.1.1.3.3.2 (sic] of the IPPSS.

Response to Interrogatory No. 17:

a) The sheltering shielding factors are based upon the assumption that persons with access to base-ments (90 percent of the population) would take shelter in them. For assessment purposes, it was assumed that the remaining 10 percent of the people continued their normal activities. Although emergency plans do not include explicit plans for protective action for plume explosure beyond the plume EPZ, the emergency plan does provide a resource base to take some actions beyond the EPZ as needed. The action assumed requires no special-ized shelter facilities. Evaluation and communication would be necessary, but the time required for the plume to travel through the EPZ would provide time for these activities. It was judged that the action would be relatively easily implemented, and probably would be implemented should an unlikely severe accident occur.

The shielding factors used for the sheltered popu-lation are derived from SAND 77-1725, reference 6-34 in the IPPSS. The reduction f actor assumed for shielding from plume exposure, 0.5, is the mean of 0.6 for the basement of a wood house and 0.4 for the basement of a masonry house (SAND 77-1725, Table 1). A representa-tive f actor for shielding from ground exposure is about

0.05 for basements of wood or mansonry houses (SAND 77-1725, Table 4). The reduction f actor assumed for shielding from ground exposure in the IPPSS, 0.08, is slightly higher to account for exposure during reloca-tion.

b) The shielding f actor selected for ground exposure for sheltered people incorporates relocation exposure equivalent to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> with a shielding factor of 1, or 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> with a shielding factor of 0.5 (auto-mobile shielding).

c) The inhalation dose reduction factor used in the IPPSS for sheltered people, 2.0, is based upon work which supersedes'the RSS: 1) NUREG/CR-1151, "Infiltra-tion of Particulate Matter into Buildings," Cohen, A.F.

and Cohen, B.L., November 1979) , and 2) SAND 77-1555, "Public Protection Strategies in the Event of a Nuclear Reactor Accident: Multicompartment Ventilation Model for Shelters," Ald rich, D.C. , and Ericson , D.M.,

January 1978. These documents indicate that a reduc-tion f actor of 2,is reasonable and that it would not be necessary or particularly helpful to attempt to " air out" the structure af ter the plume has lef t the area .

Staff Interrogatory No. 18:

In the calculation of early fatalities discussed in Section 6.3, page 6.3-3 of the IPPSS have any assumptions

l l

l l

been employed concerning the availability of medical trea tment?

Response to Interrogatory No. 18:

Yes. It was assumed that supportive treatment would be available to persons receiving life-threatening doses. See also Response to Interrogatory No.19.

Staf f Interrogatory No . 19:

If the answer to Interrogatory 19 [ sic] is yes, please describe the assumptions used, and provide the basis for those assumptions.

Response to Interrogatory No. 19:

Supportive treatment was assumed to be available for persons receiving life-threatening doses. This assumption is the same as that used in the RSS and is described in WASH-1400, Appendix VI, Section 9.2.1. The following basis for the assumption is similar to that used for the RSS and is a summary of information in the reference cited.

Supporting treatment for persons receiving doses to the bone marrow in the range of 350 to 600 rem would require specialized facilities: primarily adequate medical person-nel and laboratory support. Specialized treatment would not be required immediately, but could begin up to about three weeks after the accident. Because of the time available, it is reasonable to assume that national medical resources would be available to aid exposed people. A partial inven-tory of facilities which could provide such support, includ-

__ . . . - - . _ _ . . - . - - _ _ _ .~. - .

/

ing only those with approved progratis for" residencies in l

internal medicine, indicated that 2 500 to 5000 people could -

receive such treatment in U.S. hospitals. Medical trea tment for persons receiving less than 35C rem would not require - ,

special facilities.  :

t The limited availability of srecial facilites is a con-straint only in rare cases. In ana lyses used to produce the i S1 matrix in the IPPSS, only a few scenarios resulted in i

! estimates of more than 5000 personn in the dose range of 350 i

to 600 rem. Even in the most sevel:e scenario, limitation of

} supportive treatment to 5000 persons would result in less i

. than a 10 percent increase in the number of f atalities.

Explicit consideration of this resource constraint would not ,

change the risk curves significantly.

! Contention 2.2(b)

Staff Interrogatory No. 1:

Identify and sammarize any overcooling events that.have occurred, where the cooldown rates were determined to have exceeded the limits specified in plant. procedures or tech-

< a ,

nical specifications. Provide the results of any post event evaluations. ',

?  ?

Response to Interrogatory No. 1:

4 i Indian Point Unit 2 Following reactor coolant pump seal f ailure on July 2, 1977, the technical specification cooldown rate of 100*F/hr-f' Y

f J,

i l

was marginally exceeded during the ensued controlled shut-

$ down. A maximum cooldown rate of 105'F/hr was recorded between 4:00 A.ft. and 5:20 A.M., while the RCS pressure decreased from 2150 psig to 700 psig over the same time period. Fracture mechanics evaluation demonstrated accept-able results. There have been no other such events.

Indian Point Unit 3 There have been no overcooling events in which the cooldown rates were determined to have exceeded the limits specified in plant procedures or technical specifications.

Staff In terroga tory No . 2:

Provide a discussion of the limits, procedures and operator training for' terminating operation of the main or auxiliary feedwater pumps and/or safety injection pumps that will preclude a pressurized thermal shock event assuming an 1

accident or other conditions resulting in a cooldown rate l

. exceeding the allowable limits is in progress.

Response to Interrogatory No. 2:

Indian Point Units 2 and 3 Procedures for tprminating operation of the main or auxiliary feedwater and/or safety injection pumps have been incorporated into the Indian Point Units 2 and 3 Emergency

Operating Procedures, including loss of RCS coolant and loss of secondary coolant events. For example, the latter procedure requires that SI pump (s) be secured if RCS temp-erature is below 350' P nd pressure is above 700 psig and

is stable, or increasing. These procedures are based upon the Westinghouse generic emergency response guidelines.

Plant operators receive both classroom and plant-specific simulator instruction as part of their normal training program. Additional generic analysis and related activities, such as procedure review, are presently being performed by the Westinghouse Owners Group. Results of j these programs will be used to evaluate current Indian Point Units 2 and 3 procedures. A lecture program for plant operators on pressurized thermal shock has been initiated for Indian Point Unit 2. Additional training is planned for the near future.

Staff Interrogatory No. 3: ,

Provide a discussion on the operator guidelines, operating procedures and operator / shift technical advisor training program on pressurized thermal shock. Include a description of any testing given following training.

Re sponse to Interrogatory No . 3:

Indian Point Unit 2 See Response to , Interrogatory No. 2. Questions relat-ing to pressurized thermal shock will be included in the testing given following training.

Indian Point Unit 3 As part of the operator retraining program, all licensed operators review significant operating. events which occur throughout the industry. This " lessons-learned"

I approach informs operators of of f-normal incidents so that they will be able to be in a better position to recognize a similar event should it occur at their facility. As such, all licensed operators were issued a letter on September 4, 1981, on the Steam Generator overfill incident with detailed ,

information concerning the Rancho Seco incident which preci-pitated the concern of a pressurized thermal shock event.

Additionally, during the last quarter of the licensed operator retraining program which was conducted in early 1982, the Ginna Nuclear Facility steam generator tube rup-ture incident was discussed, one section of which specifi-cally addressed the thermal transient on the reactor coolant system.

Finally, all operators are aware of two system designs which are directly concerned with the pressurized shock event. The first is the incorporation of thermal sleeves into the design of the reactor vessel piping to prevent cracking of the vessel nozzles from thermal shock. The second is the installation of a reactor coolant overpres-surization system whiph is designed to limit an overpres-surization event during cold operating conditions.

Staff Interrogatory No. 4:

Identify and summarize the results of any post installation reactor vessel (internal) wall inspections.

1

Response to interrogatory No. 4:

Indian Point Unit 2 No reactor vessel (internal) wall inspection has been performed since initial fuel loading. The wall is scheduled for examination during the 1984 refueling outage as part of the 10-year inservice inspection program.

Indian Point Unit 3 The upper internals of the reactor were examined in 1978, 1979, and 1982. However, the vessel internal wall has not been examined. It is scheduled to be examined during the 10-year in-service inspection outage in mid-1986. This inspection will examine the longitudinal and circumferential shell welds using the' volumetric method. Further details can be found in the 10-year In-service Inspection Plan for Indian Point Unit 3.

Staff Interrogatory No. 5:

Identify the instrumentation available to the operators to assist them in recognizing a potential pressurized thermal shock event.

Response to Interrogatory No. 5:

Indian Point Unit 2 RCS wide range pressure and RCS wide range cold leg temperature instrumentation provide the principal informa-tion for recognizing a potential pressurized thermal shock event. In addition, the in-core thermocouples and the RCS i

hot leg temperature instrumentation provide useful informa-tion to the operators.

Indian Point Unit 3 l

The following pressure instrumentation can be used to assist the operators in recognizing a potential pressurized thermal shock:

Reactor Coolant System Temperature (Hot and Cold Leg s)

Pressurizer Temperature (vapor and Liquid Spaces)

Reactor Coolant System Pressure Pressurizer Pressure Incore thermo couples (Outlet temperature from fuel rods)

Staf f Interrogatory No. 6:

Identify and summarize the results of any accident analysis, transient analysis or probabilistic risk assess-ment study performed that could be applied to the pressurized thermal shock issn1 such as control system failures, steam line break accidents, small break loss of coolant accidents or pther conditions resulting in a cool-down which exceed (sic] the allowable rate limits and/or results in inside vessel wall fluid temperatures less than 300*F, with a potential for systems pressurization.

Response to Interrogatory No. 6:

Indian Point Units 2 and 3 u

WCAP-10019 1 and a subsequent report 2 transmitted to the Commission by the Westinghouse Owners Group provide accident i

analysis results and a generic assessment (including proba-bilistic risk assessment) of the reactor vessel integrity problems for all Westinghouse plants. The analyses have ,

included both operational transients and design basis events which result in a cooldown exceeding the allowable rate limits, and/or which result in inside vessel wall fluid temperatures below 300*F, including potential for system repressuriza tion . These analyses provide a conservative assessment of the yea'rs of operation prior to exceeding ves-j sol integrity acceptance criteria.

Sta f f Interrogatory No. 7:

Provide a discussion of the actions'taken thus far to lessen the probability and/or severity of pressurized thermal shock events.

i 1. WCAP-10019, " Summary Report on Reactor Vessel Integrity for Westinghouse Operating Plants." Transmitted by Le tter OG-66, O. D. Kingsley to H. Denton (NRC) ( De c . 3 0.

1981).

2. " Summary of Evaluations Related to Reactor Vessel Integrity" Transmitted by Letter OG-70, O.D. Kingsley to H.

Denton (NRC) (May 28, 1982).

l

Response to Interrogatory No. 7:

Indian Point Unit 2 A low leakage loading pattern is being incorporated (previously burned fuel assemblies will be selectively loaded at peripheral locations of the reactor core). This design will be incorporated in cycle 6 with anticipated fuel loading in the latter part of 198 2. This'is expected to result in a significant neutron fluence reduction, thus reducing the ef fects of radiation upon the reactor vessel materials.

The ef fects of radiation upon reactor vessel materials is monitored via the periodic removal of reactor vessel mater.ial specimens. These specimens are tested to evaluate the effects of radiation upon the vessel wall. Heatup and cooldown rate limitations are established based upon specimen analyses to assure that the effects of cyclic thermal stresses upon the reactor vessel are within acceptable limits.

A plant operator training / retraining program which incorporates the use of a plant-specific simulator provides futher assurance that plant operators are well-versed in effectively handling transient and accident situations, thus reducing the probability of subsequent pressurized thermal shock events. Furthermore, as discussed in the Response to Interrogatory No. 2, an additional specific training program

i on pressurized thermal shock has been initiated for the plant operators.

Indian Point Unit 3 During certain refueling outages which are stated in the technical specifications, vessel material specimen capsules which have been subjected to the same neutron flux as the vessel wall are removed from the reactor vessel.

These specimens are then tested to determine the effects from radiation upon their physical properties, i.e.,

fracture toughness. From this analysis, heat-up and cool-down curves are generated to lessen the effects of cyclic thermal stresses, thus reducing the probability and/or severity of a pressurized thermal shock event.

Additionally , the present fuel cycle incorporates a modified low leakage loading pattern which will result in a reduction of a fast neutron flux in the periphery of the core, with the subsequent ef fect of decreasing embrittlement of the reactor vessel walls.

Staff Interrogator? No. 8:

Provide a discussion and schedule for implementation of any actions planned to resolve the pressurized thermal shock concern, such as fuel management programs aimed at fluence reduction, increased ECC injection water temperature, and additional instrumentation.

Response to Interrogatory No. 8:

Indian Point Unit 2 At present there is considerable ongoing review analysis by the Commission, EPRI and others, including addi-tional generic analysis and related activities, such as procedure review, by the Westinghouse Owners Group. Results i

of these programs will be used to evaluate the desirability and/or need to effect other remedial actions, such as increased ECC injection water temperature. See Response to Interrogatory No. 7 as regards fluence reduction.

Indian Point Unit 3 The Power Authority is a participant in the Westing-house Owners Group wh'ich is involved in performing an analysis to evaluate the generic aspects of a pressurized thermal shock event. Additionally, the Power Authority's Fuels Group has been directly involved in addressing this Concern.

As was mentioned in the Response to Interrogatory No.

7, the present fuel cycle is designed to reduce the fast neutron flux which wi}1 lessen the effects of the radiation embrittlement concern over a period of time. Additionally, increasing the temperature of the ECC fluid has been identi-fied as part of the Westinghouse Owners Group generic study and will be considered in the plant-specific evaluation to determine if there is a need for this action to reduce the ef fects of a pressurized thermal shock condition.

Finally, a Reactor Vessel Level Monitoring System will be installed in the future which will assist the operators in evaluating core conditions.

4

As to Answers:

CONSOLIDATED EDISON COMPANY OF NEW, YORK, INC.

By

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Richard P. Remshaw Project Manager - Indian Point Hearings POWER AUTHORITY OF THE STATE OF NEW YORK By aerschel Spectef Project Manager - Indian Point Hearings As to Objections: ,

CONSOLIDATED EDISON COMPANY OF NEW YORK, INC.

By -

/A _

Brent L. Brandfnburg (/ /5'%

Assistant General Counsel ,

POWER AUTHORITY OF THE STATE OF NEW YORK MOR A O TES, C A 1

l By -

Paul F. Colarulli Attorney for Power Authority of the State of New York

VERIFICATION STATE OF NDi 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 interro-gatories were prepared under my direction and supervision  ;

and are true and corre,ct to the best of my knowledge, information and belief.

s L -

RKHARD P. REMSHAW Sworn to before me this 24th day of June, 1982

, , .," '  : 7,,

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n' Notary Public

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VERIFICATION STATE OF NBi YORK )

SS.:

COUNTY OF NSi 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 Power Authority; and that the foregoing answers to interrogatories were prepared under my direction and supervision and are true and correct to the best of my a knowledge, information and belief.

.fh HERSCHEL SPECTWR Sworn to before me this 24th day of June, 1982 N lN 4 l

Notary Public ,

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Respectfully submitted, Brent L.

4 Brahdenourgf,FL A l arles Morgan, Jr.

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Pa ul F. Colarulli CONSOLIDATED EDISON COMPANY Joseph J. Levin, Jr .

OF NEW YORK, INC.

Licensee of Indian Point MORGAN ASSOCIATES, CHARTERED Unit 2 1899 L Street, N.W.

4 Irving Place Wa shing ton, 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 F. Czaja David H. Pikus SHEA & GOULD 330 Madison Avenue New York, New York 10017 (212) 370-8000 Dated: June 25, 1982

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION ATOMIC SAFETY AND LICENSING BOARD Before Administrative Judges:

Louis J. Carter, Chairman Frederick J. Shon Dr. Oscar H. Paris

)

In the Matter of: )

)

CONSOLIDATED EDISON COMPANY OF )

NEW YORK, INC. ) Docke t No s . 50-247 SP (Indian Point, Unit No. 2) ) 50-286 SP

)

POWER AUTHORITY OF THE STATE OF )

NEW YORK )

(Indian Point, Unit No. 3) )

)

CERTIFICATE OF SERVICE I hereby certify that on the 25th day of June, 1982, I caused a copy of th2 Licensees' Responses to NRC Staff First

• Set of Interrogatories Concerning Questions 1 and 2 to be hand delivered to Henry J. McGurren, Esq.

Counsel for NRC Staf f Office of the Executive Legal Director U.S. Nuclear Regulatory Commission Washington, D.C. 20555 and served by first-class mail, postage prepaid on the following:

Louis J. Carter, Esq., Chairma n 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 U.S. Nuclear Regulatory Commission Of fice of the Executive Washington, D.C. 20555 Legal Director U.S. Nuclear Regulatory Commission Dr. Oscar H. Paris Wa shing ton , D.C. 20555 Administrative Judge Atomic Safety and Licensing Board Brent L. Brandenourg, Esq.

U.S. Nuclear Regulatory Commission Assistant General Counsel Washington, D.C. 20555 Consolidated Edison Company of New York, Inc.

Docketing and Service Branch 4 Irving Place Office of the Secretary New York, New York 10003 U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Ellyn R. We iss , Esq.

William S. Jordan, III, Esq.

Joan Holt, Proj ect Director Harmon and Weiss Indian Point Project -

1725 I Street, N.W., Saite 506 New York Public Interest Re search Washington, D.C. 20006 Group 9 Murray Street Charles A. Scheiner, Co-Chairperson New York, New York 10007 Westchester People's Action Coalition, Inc.

John Gilroy P.O. Box 4 88 Westchester Coordinator White Plains, New York 10602 Indian Point Project

• New York Public Interest Re sea rch Ala n La tma n , Esq.

Group 44 Sunset Drive 240 Central Avenue Croton-On-Hudson, New York 10520 White Plains, New York 10606 Ezra I. Bialik, Esq.

Jeffrey M. Blum, Esq. , Steve Leipzig , Esq.

New York University Law School Environmental Protection Bureau 4 23 Va nderbilt Hall New York State Attorney 40 Washington Square South General's Of fice New Yo rk , Ne w Yo rk 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 and New Jersey Westchester County One World Trade Center 148 Martine Avenue New Yo rk , Ne w Yo rk 100 48 White Plains, New York 10601 Andrew S . Roffe, Esq.

New York State Assembly Alba ny , New York 12248

Marc L. Parris, Esq. Stanley B. Klimberg , Esq.

Eric Thorsen, Esq. General Counsel County Attorney New York State Energy Of fice 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. Wa shing ton , D.C. 20555 Laurens R. Schwartz, Esq.

Margaret Oppel, Esq. Honorable Richard L. Brodsky Bo te in , Ha ys , 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 Honorable Ruth W. Me ssinger Member of the Council of the Zipporah S. Fleisher City of New York West Branch Conservation District #4 Associa tion City Hall 443 Buena Vista Road New York, New York 10007 New City, New York 10956 Greater New York Council Mayor 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, Coord ina tor '

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 Lorna Salzman Mid-Atlantic Representa tive 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

4-Ruthanne G. Miller, Esq.

Atomic Sa fety and Licensing Board Panel U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Mr. Donald Davidof f Director, Radiological Dmergency Preparedness Group Empire State Plaza Tower Building, RM 1750 Albany, New York 12237

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Pa ul F. Cbla r ulli N-I O

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